ML19211D299
| ML19211D299 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/01/1979 |
| From: | Buck P NUCLEAR ENERGY SERVICES, INC. |
| To: | |
| Shared Package | |
| ML19211D294 | List: |
| References | |
| NUDOCS 8001170464 | |
| Download: ML19211D299 (47) | |
Text
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47 PAGE 1
nF NUCLEAR DESIGN ANALYSIS REPORT FOR THE CALVERT CLIFFS UNIT #1 NUCLEAR PLANT HIGH DENSITY SPENT FUEL STORAGE RACKS PREPARED UNDER NES PROJECT 5134 FOR THE BALTIMORE GAS & ELECTRIC COMPANY NUCLEAR ENERGY SERVICES, INC.
DANBURY, CONNECTICUT 06810 n(0 :=7 anum
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17 62 342 8001170 46 i-FORM # NES 204 9/78
i i
NUCLEAR ENEP.GY SERVICES, INC.
DOCUMENT NO.
El A05f;7 PAGE 2
OF 41 TABLE OF CONTENTS PAGE 1.
SUMMARY
6 2.
INTRODUCTION 8
3.
DESCRIPTION OF SPENT FUEL STORAGE RACKS 9
4.
CRITICALITY DESIGN CRITERION AND CALCULATIONAL' ASSUMPTIONS 12 4.1 Criticality Design Criterion 12 4.2 Calculational Assumptions 12 5.
CRITICALITY CONFIGURATIONS 14 5.1 Normal Configurations 14 5.1.1 Reference Configuration 14 5.1.2 Eccentric Configuration 14 5.1.3 Fuel Assembly Tolerance 15 5.1.4 Fuel Design Variation 15 5.1.5 Fuel Rack Variation 15 5.1.6 Cell Wall Thickness Variation 15 5.1.7 Poison Concentration Variation 16 5.1.8 Effect of Discrete B C Particle Size 16 5.1.9
" Worst Case" NormaIConfiguration 16 5.2 Abnormal Configurations 17 5.2.1 Single Storage Cell Displacement 17 5.2.2 Fuel Handling Incident 17 5.2.3 Pool Temperature Variation 17 5.2.4 Fuel Drop Incident 17 5.2.5 Heavy Object Drop 18 5.2.6 Seismic Incident 18 5.2.7
" Worst Case" Abnormal Configuration 19 6.
CRITICALITY CALCULATIONAL METHODS
'1 6.1 Method of Analysis 21 6.2 Reference Configuration 21 6.3 Uncertainties and Benchmark Calculations 22 6.4 Code Description 24 6.4.1 KENO IV 24 6.4.2 HAMMER 24-7 } }4}
6.4.3 EXTERMINATOR 2
FORM e NES 205 5/79
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DOCUMENT NO.
81A0567 PAGE 3
41 op TABLE OF CONTENTS (CONT'D)
PAGE 7.
RESULTS OF CRITICALITY CALCULATIONS 28 7.1 Reference Configuration 28 7.2 K,ff Values for Normal Configurations 28 7.2.1 Eccentric Configuration 28 7.2.2 Fuel Design Variation 28 7.2.3 Fuel Rack Pitch Variation 29 7.2.4 Fuel Rack Cell Wall Thickness Variation 29 7.2.5 Poison Content Variation 29 7.2.6
" Worst Case" Normal Configuration 30 7.3 K,ff Values for Abnormal Configurations 30 7.3.1 Fuel Handling incident 30 7.3.2 Spent Fuel Pool Temperature Variation 31 7.3.3
" Worst Case" Abnormal Configuration 31 8.
REFERENCES 40 1762 344 FORM = NES 205 5/79
i NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81A0567 PAGE 4
OF UI LIST OF FIGURES PAGE 3.1 Fuel Storage Rack Arrangement 10 3.2 PWR 10 x 10 Poisoned Fuel Storage Rack 11 5.1 Displaced Fuel Configuration 20 6.1 Reference Configuration 25 6.2 Bias between 16 Group KENO IV and Experiments 26 6.3 Bias between 123 Group KENO IV and Experiments 27 7.1 Ak vs. Enrichment 34 eff 7.2 Ak vs. Pitch 35 eff 7.3 Ak vs. Wall Thickness 36 eff 7.4 Ak vs. Poison Content 37 eff 7.5 Akeff vs. Temperature 38 7.6 Ak vs. H O Density 39 eff 2
1762 345 8 C R '.9 m NES 205 5/79
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DOCUMENT NO.
81A0567 PAGE 5
41 op LIST OF TABLES 7.1 Fuel Parameters 32 7.2 Parameters and Results of Exterminator Calculations 33 1762 346 FORM e NES 205 $/79
l NUCLEAR ENERGY SERVICES, INC DOCUMENT NO.
81A0567 PAGE 6
op 41
- 1.
SUMMARY
A detailed nuclear analysis has been performed for. the NES designed fuel storage racks for the Calvert Cliffs Unit No.1 Nuclear Plant. This analysis demonstrates that, for all anticipated normal and abnormal configurations of fuel assemblies within the fuel storage racks, the k f the system is less than the criticality criterion of 0.95 df for 4.10 w/o,14x14 Combustion Engineering fuel assemblies. Certain conservative assumptions about the fuel assemblies and racks have been used in the calculations.
Both normal and abnormal configurations were considered in the analysis.
The reference configuration consists of a square array, infinite in lateral extent, of storage cells spaced 10.09375 inches on centers. Each storage location contains one centrally located 14x14 Combustion Engineering fuel assembly. Poison sheets containing boron in the form of B C are located in the walls of the storage cells to provide criticality q
control. This reference configuration provides a base of comparison relative to which effects of normal and abnormal variations have been measured. Normal configurations include:
eccentrically positioned fuel, fuel enrichment variation, dimensional and material variations permitted by fabrication tolerances, and variation in the density of the boron in the poison slabs. Effects due to finite B C particle sizes were also g
considered.
Abnormal configurations include: pitch variation due to seismic events, spent fuel pool temperature variations and fuel handling accidents such as misplaced fuel assemblies.
I The principal method of calculation used to determine the k f the Calvert Cliffs eff spent fuel storage racks was transport theory using the Monte Carlo code KENO IV.
Two cross section sets were used in KENO IV: a 16 group Hansen-Roach set and a set using 123 energy groups.
Cross section input for the 123 energy group set was generated from the XSDRN library using the AMPX module NITAWL.
1762 347 FORM # NES 205 5/79
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DOCUMENT NO.
81A0567 PAGE 7
OF hl Parametric studies to determine the effects on k,ff of changes in fuel rack dimensions, temperature, and fuel assembly enrichment were performed with diffusion theory. Fuel, water and structural cross sections were determined using the HAMMER code, while blackness theory was used to determine boron cross sections. k,ff v-tues were calculated using EXTERMINATOR, a multigroup, two-dimensional diffusion theory code.
The k value calculated by KENO for the reference configuration is 0.9001.
eff Variations in k,ff due to normal configuration changes and calculational uricertainty were determined to be 0.0085.
The A k,f f due to the " worst case" abnormal configuration is 0.0000. Combining these two Ak,,f values with the k,ff for the reference configuration of 0.9001 results in a final k value equal t 0.9086. This eff value meets the criticality design criterion and is substantially below 1.0. Therefore, it has been concluded that the high density storage racks for the Calvert Cliffs Unit No.1 Nuclear Plant when loaded with the specified fuel are safe from a criticality standpoint.
1762 348 F O R *.' = N E S 205 5/79
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DOCUMENT NO.
81A0567 NUCLEAR ENERGY SERVICES. INC.
PAGE R _ 0F 41
- 2. INTRODUCTION The NES design for the Calvert Cliffs Unit No.1 Nuclear Plant high density spent fuel storage racks achieves high storage density through the placement of poison sheets in the walls of the storage cells. Details of the rack materials and structure are given in Section 3.
A detailed nuclear analysis has been performed to demonstrate that, for all anticipat-ed normal and abnormal conngurations of fuel assemblies within the fuel storage racks, the k,ff of the system is substantially below 1.0.
Certain conservative assumptions about the fuel assemblies and racks have been used in the calculations.
These are described in Section 4 along with the criticality design criterion for the fuel assemblies and racks.
The reference configuration which forms the basis of the criticality calculations represents the storage racks in nominal dimensions at 68 F with all fuel assemblies centrally located within their~ storage cells. Variations from this reference configura-tion were studied, and included effects of wall thickness and pitch variations, fuel enrichment and poison content variations, water temperature variation and eccentric fuel positioning. Fuel handling accidents were studied and their effects determined.
The~ configurations studied are described in detail in Section 5.
A description of the calculational methods, benchmarking results, and computer codes is given in Section 6.
The results of the criticality analysis are presented in Section 7.
- NES 205 5/79
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81A0567 PAGE 9
OF 41
- 3. DESCRIPTION OF SPENT FUEL STORAGE RACKS Three sizes of fuel storage racks, with 7x10, 8x10, and 10x10 storage cell arrays, will be used in the Calvert Cliffs Unit No.1 spent fuel storage pool (see Figure 3.1). The total number of fuel storage locations within the pool will be 830.
The inner wall of each storage cell is made up of a 0.060 inch t. hick sheet of 304L stainless steel, formed into a square with an inner dimension of 8-9/16 inches. On the outside of each of the four sides of this inner wall, a poison sheet 6-1/2 inches wide is sandwiched between the inner wall and an external 0.060 inch-thick stainless steel sheet (see Figure 3.2). The poison sheet is 0.090 inch thick and contains a minimum of 2
IO 0.024 gm/cm of B The external sheet extends over two fuel storage cells so that storage cells are grouped into 2x2 modules from which the storage racks are built up (Figure 3.2). The average center-tc-center pitch between all fuel storage boxes is maintained by the 1 03125 inches.
0 external sheets and welded spacers at 10.09375 1762 350 FORM eNES 205 5/79
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NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO.
81 A0567 PAGE 12 op 41
- 4. CRITICALITY DESIGN CRITERION AND CALCULATION ASSUMPTIONS 4.1 CRITICALITY DESIGN CRITERION A satisfactory value of k,ff for a spent fuel pool involves considerations of safety, licensability and storge capacity requirements. These factors demand k,ff substam tially below 1.0 for safety and licensability but high enough to achieve the required storage capacity.
The published position of the NRC on fuel storage criticality, stated in a communique to all reactor licensees
- is a f ollows:
"The neutron multiplication f actor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions".
Furthermore, NRC, in evaluating the design, will " check the degree of subcriticality provided, along with the analysis and the assumptions".
On the basis of this information, the following criticality design criterion has been established for the Calvert Cliffs Unit No.1 Nuclear Plant high density fuel storage racks: "The multiplication constant (keff) shall be less than 0.95 for all normal and abnormal configurations as determined by Monte Carlo calculation".
4.2 CALCULATIONAL ASSUMPTIONS The following conservative assumptions have been used in the criticality calculations performed to verify the adequacy of the rack design:
1.
The fuel is fresh and of a specified enrichment greater than or equal to that of any fuel available (4.1 wid 762 353
- See Section 8 for References.
FORM = NES 205 5/79
NUCLEAR ENERGY SFRVICES, INC.
DOCUMENT NO.
81A0567 PAGE I3 MI OF 2.
The reference configuration contains an infinite square array of storage loca-tions spaced 10-3/32 inches on centers. This is conservative because the array is not infinite, but finite.
3.
The absorption of the f uel assembly spacers is ignored.
4.
Any burnable poisons in the f uel assemblies are ignored.
5.
The vertical buckling is ignored, i.e., the f uel assemblies are considered to be infinitely long.
6.
Any soluble poison in the pool water is ignored.
1762 354 FOAV eNES 205 5/79
81 A0567 DOCUMENT NO.
NUCLEAR ENERGY SERVICES. INO.
PAGE 16 OF El
- 5. CRITICALITY CONFIGURATIONS In order to verify the design adequacy of the Calvert Cliffs Unit No.1 Nuclear Plant high density storage rack, it is necessary to establish the multiplication constants for the various arrangements or configurations of fuel assemblies and storage cells that are possible within the racks. These arrangements or configurations can be classified as either normal or abnormal configurations.
Normal configurations result from variation in the placement of fuel within the storage cell, variation in fuel assembly dimensions and/or fuel loading because of the manufacturing process, and the' variation
'in fuel storage rack dimensions permitted in f abrication. Abnormal configurations are typically the result of accidents or malfunctions such as the scismic event, a malfunction of the fuel pool cooling system (excessive changes in pool water temperature), a dropped fuel assembly, etc. The following sections present the normal and abnormal configurations which have been considered in this analysis.
5.1 NORMAL CONFIGURATIONS 5.1.1 Reference Configuration The reference configuration consists of an infinite array of storage cells having nominal dimensions (see Section 3) each containing a fresh 14x14 Combustion Engineering fuel assembly centrally located within the storage cell.
The water temperature within the rack is 68 F.
5.1.2 Eccentric Configuration It is possible for a fuel assembly not to be positioned centrally within a storage cell because of the clearance allowed between the -assembly and the cell wall.
This clearance is nominally 0.221 inches on each side of the fuel assembly.
Calculations have been performed to determine the effects of eccentrically located f uel. In these calculations it was assumed that four fuel assemblies were diagonally displaced within their storage cells as far as possible towards each other (see Figure 5.1).
1762 355 SORY e NES 205 SD9
81A0567 DOCUMENT NO.
NUCLEAR ENERGY SERVICES, ING.
PAGE 15 OF h1 5.1.3 Fuel Assembly Tolerance The important fuel assembly parameter determining k is the ratio of the amount of eff 235 235 U
to that of water. The amount of U per assembly is controlled to within a few tenths of a percent by weighing pellet stacks as the fuel is built and by using a known enrichment. The fuel assembly parameters which determine the volume of water in an assembly are the clad O.D. and the fuel rod pitch. These parameters are closely controlled to typically within 1 4%.
The effects of these fuel assembly 0
tolerances on k have been determined to be negligible on the basis of simple k df oo cell calculations. Consequently, fuel assembly tolerances were not considered further in this analysis.
5.1.4 Fuel Design Variation Calculations were performed to determine the sensitivity of k to variations of fuel gff enrichment from the base enrichment of 4.10 w/o. The criticality configuration used for these calculations was that of the reference configuration with the exception of fuel enrichment.
5.1.5 Fuel Rack Variation Calculations were performed to determine the sensitivity of k to changes in pitch, df the center-to-center spacing between storage cells. The pitch was varied from 9.75 to 10.25 inches.
The criticality configuration was similar to that of the r~eference configuration except for the obvious change in center-to-center spacing.
5.1.6 Cell Wall Thickness Variation The base case wall thickness was 0.060 inch for each of the stainless steel sheets forming the cell walls.
This thickness was varied from 0.070 to 0.050 inch to determine the effect on kdf*
- NES 205 5/79
NUCLEAR ENERGY SERVICES. INC.
DOCUMENT NO.
81 A0567 PAGE 16 OF 41 5.1.7 Poison Concentration Variation Experiments have shown that the poison material used in Calvert Cliffs experiences a loss of B C during radiation exposure equivalent to 40 years residence in the spent fuel 4
10 pool. The loss of B C reduced the B concentration by an average value of 15% with 4
the maximum redaction in any single sample being 19.2%. The pre-exposed material 10 2
for Calvert Cliffs has a minimum concentration of 0.024 gm B /cm.
For the 10 2
reference configuration this value is reduced to 0.020 gm B /cm to reflect the 15%
average loss. This concentration is varied by 310% to determine the sensitivity of k,ff to variations in this parameter.
5.1.8 Effect of Discrete B,,C Particle Size Calculations were performed with KENO in which the B C particles in the poison 4
sheets were represented as spheres of fixed diameter, regularly spaced throughout the sheet. The diameters studied were: 0.020 inch, 0.010 inch and zero (homogeneous).
For the 0.020 inch case, an increase of about 2% ak/k was found compared with the homogeneous case. However, for the 0.010 inch case, the increase in k was zero within the uncertainty of the KENO calculation. Since 50% of the particles are smaller than 0.005 inch and 90% are smaller than 0.010 inch the ak/k due to finite particle size is taken as zero.
5.1.9 " Worst Case" Normal Configuration The " worst case" configuration considers the effect of eccentric fuel assembly positioning, the minimum average pitch (center-to-center spacing) permitted by f abrication, the minimum wall thickness and the minimum poison concentration.
1762 357 FORM e NES 205 5/79
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81 A0567 PAGE 17 OF 41 5.2 ABNORMAL CONFIGURATIONS 5.2.1 Single Storage Cell Displacement Displacement of a single storage cell within the array is precluded by the welded construction and the presence of structure between cells. Therefore, the effect of such a displacement is taken to be zero.
5.2.2 Fuel Handling incident Accidental placement of fuel between the fuel racks or the racks and pool wall will be prevented by structural material. It is, however, conceivable that an assembly could be laid across the top of a fuel rack. In this case, the distance between the tops of the stored fuel and the bottom of the misplaced fuel will be greater than 25 inches, a distance which according to calculations effectively "decouples" the two groups of fuel. No increase in k,ff will result from this incident.
5.2.3 Pool Temperature Variation Calculations were performed to determine the sensitivity of k,ff for the reference configuration to variations in the spent fuel pool temperature. The pool temperature was varied from 39 F, where water density is maximum, to 250 F, the approximate boiling point of water near the bottom of the fuel rack.
5.2.4 Fuel Drop Incident The maximum height through which a fuel assembly can be dropped onto the fuel storage racks is limited. The dropped fuel assembly will mo.t likely impact the tops of the fuel storage rack cells. Because of the fuel rack design, damage will be limited to the upper 6 to S inches of the storage cells. Since the active fuel region is about 18 inches below this area, no significant change in fuel / cell geometry will occur.
However, it is possible for a dropped fuel assembly to enter a cell cleanly and impact directly on the fuel stored in the cell. The effect of this type of fuel drop incident was evaluated from a criticality viewpoint by assuming that the stored assembly would be compressed axially.
1762 358 FORM
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NUCLEAR ENERGY SERVICES. ING.
DOCUMENT NO.
81A0567 PAGE 18 OF 41 A calculation based on an axial compression of 2 feet yielded a 0.06 decrease in k of og the fuel cell. It has been concluded, therefore, that this incident would reduce k,ff and need not be considered further in this analysis.
5.2.5 Heavy Object Drop in the unlikely event that a heavy object is dropped on the storage rack with sufficient will decrease.
impact to cause structural deformation, it has been concluded that keff The basis for this conclusion is that the principal effect of dropping a heavy object will be to squeeze water from the rack. Both in the case of compacted fuel and voided pool water, depletion of water leads to a decrease in k,ff.
It would not be possible for a dropped heavy object to eject the poison material from the rack; the crushing effect of the heavy object could only act to compress the fuel and poison together.
5.2.6 Seismic Incident Seismic analyses have determined that during an SSE the pitch between two adjacent fuel assemblies could narrow locally by as much as 0.005 inches, due to oscillations about nodal points determined by structural members locating the cells within the racks. However, at the same time, the local pitch at other locations is greater by the same amount. Thus, the net effect, although the pitch may vary locally, is that the average pitch is unaffected. In the event that the entire rack is displaced by a seismic event, the average pitch will also be unaffected.
It is concluded, therefore, that if the fuel assemblies deflect independently in random directions or move together in a single direction, the average pitch between assemblies and, consequently, the k,ff are unaffected.
1762 359 FORM e NES 205 5/79
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81A0567 19 NI PAGE OF 5.2.7 " Worst Case" Abnormal Configuration The " worst case" abnormal configuration considers the effect of the most adverse abnormal condition in combination with the " worst case" normal configuration. The results of the " worst case" abnormal configuration are presented in Section 7.2.3.
1762 360 FORM e NES 205 5/79
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NUCLEAR ENERGY GERVICES, INC.
DOCUMENT NO.
81 A0567 PAGE 2I OF 4I
- 6. CRITICALITY CALCULATIONAL METHODS 6.1 METHOD OF ANALYSIS For the reference configuration discussed in Section 5.1.1, the k,ff was determined from a three-dimensional Monte Carlo calculation using KENO IV with two sets of cross sections: the 16 group group Hansen-Roach cross section set and the 123 group cross section set. Check calculations of the reference configuration as well as the parametric studies were performed with two-dimensional diffusion theory using HAMMER and EXTERMINATOR.
In both the Monte Carlo and diffusion theory methods, an infinite array of fuel assemblies loaded in spent fuel storage locations was represented by use of appropriate boandary conditions. An infinite array is used for two reasons: (1) an infinite array has a conservatively higher value of k and (2) the df problem can be suitably represented by a repeating portion of the array. Figure 6.1 shows a representation of one quarter of a storage location with reflecting boundaries on all sides. This duplicates an infinite array of storage locations.
6.2 REFERENCE CONFIGURATION In the reference configuration KENO IV calculations, each fuel pin and associated cladding and water was represented as a rectangular parallelapiped with height equal to the active fuel length and the width equal to one fuel rod pitch (0.580 inch).
Cladding and fuel were represented by concentric cylinders within the box with atom densities determined from the fuel parameters shown in Table 7.1.
Water at 68 F filled the region outside the cladding. Guide tubes were represented in a similar f ashion but with water inside the clad instead of fuel. The stainless steel sheets making up the box walls were represented as boxes with thickness of 0.060 inch and a width equal to the fuel storage cell edge. Poison sheets were represented by boxes 0.090 inch thick and 6-1/2 inches wide containing a homogenous mixture of B C.
q Water regions were boxes of appropriate sizes needed to fill the water gaps.
D * ]D D '9' } g P
_ o o m e4 J1 R o
1763.002
=
FOsW e NES 205 5/79
I NUCLEAR ENERGY SERVICES,INC.
DOCUMENT NO.
81A0567 PAGE 22 OF 41 Although the reference configuration KENO IV problem did not take into account the finite size of the B C particles, several supplementary problems were performed in 4
order to determine the effect on k,ff of particle sizes from zero to 0.020 inches in diameter. In these problems, the poison sheets were represented as layers of cubes with centrally located spheres of B C.
The diameters of the spheres were 0.000, q
0.010, and 0.020 inch in three separate problems, and the center to center spacing between spheres was chosen so as to maintain the areal density of boron constant.
6.3 UNCERTAINTIES AND BENCHMARK CALCULATIONS The errors in Monte Carlo criticality calculations can be divided into two classes.
1.
Uncertainty due to the statistical nature of the Monte Carlo methods.
2.
Errors due to bias in the calculational technique.
The first class of errors can be reduced by simply increasing the number of neutrons tracked. For rack criticality calculations, the number of neutrons tracked is selected to reduce this error to less than 1%.
The second class of error is accounted for by benchmarking the calculational method against experimental results. In the benchmarking process, the calculational method is used to determine the criticality value for a critical experiment configuration. The difference between the calculated criticality value and the experimental value is identified as the calculational bias. Once determined, this bias can be applied to other calculational results obtained for similar configurations to irnprove the degree of calculational accuracy. If the calculated criticality value found during benchmarking is less than the experimental value, then the bias is added to other calculational results to ensure a conservative criticality value consistent.with experimental results.
Conversely, if the calculational criticality value is greater than the experimental value, it is appropriate to subtract the bias from the other calculational results to improve the accuracy of the criticality determination.
W6$bN[hI D #
1 D**D FORM 8 NES 209 S/79
SIA0567 DOCUMENT NO.
NUCLEAR ENERGY SERVICES, INC.
PAGE 23 OF 41 NES has performed benchmark calculations with KENO IV using both sets of cross sections (16 group and 123 group) for several appropriate critical experiment per-formed by Babcock and Wilcox (Ref. 3), Battelle Northwest Laboratories (Ref. h), and Allis Chalmers (Ref. 5). Benchmark calculations performed by B&W and others using 16 group KENO IV show that the calculated criticality values are consistently 1.5 to 2% greater than the experimental value. This calculational method, however, has a p (the scattering cross section per absorber atom) which can be adjusted to factor o provide closer agreement with experimental results. NES has developed a method to select an appropriate o factor. Ak lattice calculation for the fuel pin used in the p
gg experiment is performed using the NULIF (Ref. 6) code which has been specifically developed to obtain highly accurate k values.
The lattice calculation is then og performed with 16 group KENO IV and the o factor adjusted until the k, values are p
g in agreement.
This o value is then used to perform the critical experiment p
benchmark calculations. Using the method, NES has determined the 16 group KENO IV bias as a function of the Dancoff factor associated with each critical experiment.
Figure 6.2 is a plot of the bias for a range of Dancoff factors associated with widely spaced pins (low Darcoff) to fuel pin spacing essentially equal to Calvert Cliffs. As can be seen, the alas is essentially zero within experimental and calculational uncertainty over the full range of Dancoff factors studied.
The bias for 123 group KENO IV was also determined as a function of the Dancoff f actor associated with the same experimen's, nd the results are shown in Figure 6.3.
Although the bias is zero for widr'y spaced pins, it increases to 2% at spacings essentially equal to Calvert Cliff.
Additional verification of the +2% bias was obtained by performing a k, lattice calculation using 123 group KENO for the Calvert g
Cliff pin spacing and comparing it with the k value determined by the lattice code, NULIF. The results of the comparison show that the 123 group KENO calculational method over estimates the k value by 2.7% (1.480 vs 1.441).
gg 1763 004 FORM eNES 205 5/79
DOCUMENT NO.
81 A0567 NUCLEAR ENERGY SERVICES. ING.
PAGE 76 OF 41 6.4 CODE DESCRIPTION 6.4.1 KENO IV KENO IV is a 3-D multigroup Monte Carlo code used to determine keff (see Ref. 7).
6.4.2 H AMMER HAMMER (see Ref. 8) is a multigroup integral transport theory code which is used to calculate lattice cell cross sections for diffusion theory codes. This code has been extensively benchmarked against D O and light water moderated lattices with good 2
results.
6.4.3 EXTERMINATOR EXTERMINATOR (see Ref. 9) is a 2-D multigroup diffusion theory code used with input from HAMMER to calculate k,ff values.
1763 005 h
FORM = NES 205 5/79
DOCUMENT NO.
81Aos67 NUCLEAR ENERGY SERVICES, INC.
25
- I PAGE 0F SS304L HO 2
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Poison Compound Illustration of Reference Configuration Fig. 6.1 1763'006 FORM e NES 205 9/78
E DOCUMENT NO.
81 A0567 NUCLEAR ENERGY SERVICES, ING.
PAGE 26 op 41 0.02 0.01 AK (calc-exp) 0)
(.)
0.00 G
-0.01
~
-0.02 O
0.1 0.2 DANCOFF FACTOR FIGURE 6.2 BIAS BETWEEN 16-GROUP KENO-IV AND EXPERIMENT l'763 007
DOCUMENT NO.
81A0567 NUCLEAR ENERGY SERVICES, INC.
27 OF 41 PAGE 4
ii 0.04 -
0.03 -
G e
0.02 AK (CALC-EXP) 0.01 0
0.00 6
-0.01 O
0.10 0.20 DANCOFF FACTOR FIGURE 6.3 BIAS BETWEEN 123-GROUP KENO-IV AND EXPERIMENTS 1763 008 n sa u e pg
. 1/yg
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
R1 An5A7 PAGE 28 OF 41
- 7. RESULTS OF CRITICALITY CALCULATIONS The k for the reference configuration was determined by means of KENO IV eff considering both 16 group cross sections und 123 group-cross sections. Parametric studies of enrichment, temperature, dimensional tolerances of the racks, and abnormal dislocations within racks due to seismic events, fuel handling incidents, fuel drop and heavy object drop were performed with either HAMMER / EXTERMINATOR diffusion theory or 123 group KENO.
7.1 REFERENCE CONFIGURATION The k,ff determined by KENO l', using the 16 group cross section set was 0.8859 with an uncertainty of 3 0055 at the 95% confidence level. The kdf determined by means 0
of KENO IV using 123 groups was 0.9201 which when combined with the bias of -0.0200 determined from benchmarking (see Section 6.3) gives a k f 0.9001. The higher of df these two KENO values,0.9001, is chosen for conservatism.
7.2 ' K VALUES FOR NORMAL CONFIGURATION df 7.2.1 Eccentric Configuration The a k f r the eccentric configuration described in Section 5.1.2 and shown in df Figure 5.1 (in which fuel assemblies are diagonally displaced towards each other) was deterrnined to be Akdf = -0.0075.
7.2.2 Fuel Design Variation The enrichment of the fuel was changed f rom 4.10 w/o to 3.90 w/o. The results are shown in Figure 7.1 and Table 7.2.
1763 009 FORM ANES 205 5/79
i DOCUMENT NO.
81A0567 NUCLEAR ENERGY SERVICES, INC.
PAGE 29 OF 41 7.2.3 Fuel Rack Pitch Variation A detailed study of the effects of variation in the rack pitch was performed with 123 group KENO. The pitch was varied from 9.75 to 10.25 inches and the results are shown in Figure 7.2.
The mechanical design of the fuel rack is such that the average pitch between boxes is maintained by structural members at 10.09375 3 03125 inches. The 0
change in k ;; for a decrease in average pitch of 0.03125 inch is +0.0062. (See Figure g
7.2.)
7.2.4 Fuel Rack Cell Wall Thickness Variation The value of the wall thickness used in the reference configuration calculation is nominally 0.060 inch. A variation of 3 0.010 inch was investigated and the results are shown in Figure 7.3. The material used for the wall will have a thickness toleranc of 2 005 inches and the Ak for this variation, as determined from Figure 7.3, is +0.0008.
0 7.2.5 Poison Content Variation The poison content was varied by 10% above and below the reference value of 2
10 0.020 gm/cm of B The results are shown in Figure 7.4 and Table 7.2.
The 10 maximum reduction in B concentration experienced by any single test sample (see 10 10 Section 5.1.7) was 19.2% which results in a B concentration of 0.01939 gm. B /cm 10 2
(when applied to the Calvert Cliffs pre-exposure value of 0.024 gm. B /cm ) as 10 2
compared with the reference configuration value of 0.020 gm B /cm. The ok value 10 10 2
for this additional B reduction of 0.00061 gm B /cm is +0.0015.
1763 010 FORM e NES 205 $/79
i NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81A0567 PAGE 30 op 41 7.2.6 " Worst Case" Normal Configuration Results for normal configuration can be summarized as follows:
K,ff or A K,ff 1.
Reference Configuration 0.9001 2.
Minimum Cell Pitch 0.0062 3.
Minimum Poison Concentration 0.0015 4.
Eccentric Positioning 0.0000 5.
Cell Wall Thickness 0.0008 6.
Enrichment Variation 0.0000 (maximum used) 7.
B C Particle Size Effect 0.0000 (negligible) q S.
Statistical Uncertainty in KENO 0.0055 The effects of the above normal variations are combined statistically as follows:
2 2
2 2
6 k,ff = (0.0062 + 0.0008 + 0.0055 + 0.0015 )B = 0.0085 The' result for the " worst case" normal configuration is thus 0.9001 1 0085.
0 7.3 K
VALUES FOR ABNORMAi. CONFIGURATIONS eff 7.3.1 Fuel Handling Incident Since it will not be possible to place fuel adjacent to a rack, and since the Ak caused by a fuel assembly lying horizontally on top of the rack is negligible, no allowance on k
is made for this abnormal configuration.
eff 1763 011 FORM e NES 205 5n9
I NUCLEAR ENERGY SERVICES.
DOCUMENT NO.
81A0567 PAGE 31 OF 41 7.3.2 Spent Fuel Pool Temperature Variation The kef; of the rack was studied for temperatures rc 1ging from 39 F to 250 F.
Results are given in W 7.5 and Table 7.2 and show that the rack has a negative temperature cc:!ficient with the highest k,ff occurring at the nominal 68 F tempera-ture.
7.3.3 " Worst Case" Abnormal Configuratiog The " worst case" abnormal configuration combines the change in k,ff due to the occurence of the most adverse abnormal condition with the k,ff value associated with the " worst case" normal configuration.
However, since none of the abnormal conditions gives a positive Ak, the " worst case" abnormal condition is simply equal to the " worst case" normal condition.
- II 1.
Worst Case Normal Configuration (per Section 7.2.6) 0.9086 2.
Most Adverse Abnormal Configuration 0.0000 3.
Final k for " worst abnormal" 0.9086 configuibion.
1763 012 FORM e NES 205 5/79
NUCLEAR ENERGY SERVICES, INC.
DOCUMENT NO.
81A0567 PAGE 37 OF 41 TABLE 7.1 FUEL PARAMETERS 14 x 14 Combustion Fuel Type Engineering Fuel Fuel Enrichment 4.10 w/o Mass of Uranium per Assembly 395. 2,.:g CladI.D.
0.388 in Clad O.D.
0.440 in Clad Thickness 0.026 in Clad Material Zircaloy-4 Pitch Between Rods 0.580 in Active Fuel Length 136.7 in Array Dimensions 14 x 14 1763 013 FORM e NES 205 5/79
i name TABLE 7.2 7
3 C
e PAR AMETERS AND R'ISULTS OF EXTERMINATOR CALCULATIONS P
Tx Average Cell HO Poison k
2 egg Enrichment Pitch Temp.gF Density Content or m
(w/o)
(inches)
(gm/cc) gm B/cm Ak 2
eff Ni Reference Configuration 4.10 10-3/32 68 0.998 0.020 0.9001
@5 Maximum Water Density 4.10 10-3/32 39 1.000 0.020 0.0000 h
150 F, Temo. Case 4.10 10-3/32 150 0.980 0.020
-0.0050 212 F, Te i.e 4.10 10-3/32 212 0.958 0.020
-0.0145 250 F, Tem. Case 4.10 10-3/32 250 0.942 0.020
-0.0165 e
Low Enrichment, 4.00 w/o 4.00 10-3/32 68 0.998 0.020
-0.0050 Low Enrichment, 3.90 w/o 3.90 10-3/32 68 0.998 0.020
-0.0101 Low Poison Content, D
- 10%
4.10 10-3/32 68 0.998 0.018
+0.0050 8
C Eccentric Fuel 4.10 10-3/32 68 0.998 0.020
-0.0075 h
2 N
Low Wall Thickness, 3
0.050 in.
4.10 10-3/32 68 0.998 0.020
-0.0016 El
,o a High Wall Thickness, 0.070 in.
4.10 10-3/32 68 0.998 0.020
+0.0016 3
A O
8 O
E
I NUCLEAR ENERGY SERVICES,INC.
DOCUMENT NO.
81 A0567 PAGE 34 op 41
/
0.000
-0.005 h K,ff
-0.010 I
t 3.9 4.0 4.1 Fuel Enrichment, w/o FIGURE 7.1 AK,ff vs. Enrichment 1763 015 FO AV = NES 205 S/79
NUCLEAR ENERGY SERVICES, INC.
'N O
- 81A0567 PAGE 35 _op 41 0.06 O
0.04 0.02 G
A K,ff 0.00 O
N
-0.02
-0.04 9.75 10.00 10.25 Pitch, Inches FIGURE 7.2 AKeff vs. Pitch FOAM eNES 205 5/79
NUCLEAR ENERGY SERVICES, INC.
PAGE 36 op 41 0.02 0.01 0.00 A K,ff
-0.01
-0.02
.05
.06
.07 Thickness, Inches FIGURE 7.3 AK vs. Stainless Steel Wall Thickness eff 1763 017 F O R M
- N ES 205 5/79
DOCUMENT NO.
81 A0567 NUCLEAR ENERGY SERVICES, INC.
37 NI PAGE OF 0.015 0.010 0.005 0.000
.018
.019
.020
.021
.022 10 B
Poison Density, gm/cm 1763 018 FIGURE 7.4 AKeff vs. Poison Content FORM
- NES 205 S/79
A E
"8 z
C P
Tx 0.00n
-O m
Zmna 4
en mx5o
-0.005 g
E O
AK egg 0.015 a
oCEm2
~
3 z
-0.020 i
i i
e e
i i
e i
i O
.O O
10 40 60 80 106 12b o
Temperature, OC M
o o
m FIGURE 7.5 AK
- n. Temperature eff
0 ooc 3mZH 2O-
= zCP Tnmz
< nmx5om(n. 29 r
=
y> O "'
w*
O" s
0 I'
0 1
99 0
I 8
y 9
t 0
isne 3
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r A
e t
6 a
9 W
6 I
0 7
l E
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G IF 59 0
s t
I 9
0 0
5 0
5 0
0 1
1 0
0 0
0 0
0 0
f 0
f e
K A
N&u h
a g* 5*
DOCUMENT NO.
81 A0567 NUCLEAR ENERGY SERVICES, INC.
PAGE 40 op 41
- 8. REFERENCES 8.1 Grimes, B.K.,"USNitC Letter to All Reactor Licensees," April 14, 1978.
8.2 "The Carborundum Company Test Program for Baron Carbide Material," CBO-N-78-299, October 1978.
8.3
- Bromley, W.D., Olszewski, 3.S., " Safety Calculations and Benchmarking of Babcock & Wilcox Designed Close Spaced Fuel Storage Racks," Nuclear Technology, Volume 41, Mid December 1978.
8.4
- Bierman, S.R.,
D irst, B.M.,
NUREG/ER-0073-RC, " Critical Separation 235 Bewtween Clusters of 4.29 w/o U Enriched UO Rods in Water with Fixed 2
Neutton Poisons," May 1978.
8.5 Weader, R.J., " Criticality Analysis of the Atcor Vandenburgh Cask," Nuclear Energy Services, Inc.; NES 81 A0260, May 1978.
8.6
- Wittkopf, W.A., "NULIF-Neutron Spectrum Generator, Few-Group Constant Generator, and Fuel Depletion Code," BAW-426, August 1976.
8.7 Petrie, L.M., Cross, N.F., " KENO IV - An improved Monte Carlo Criticality Program", ORNL - 4938, November 1975.
8.8 Sutich, 3.E., Honeck, H.C., "The HAMMER System" DP - 1064, January 1967.
8.9.
Fowler, T.B., et al, "EXTERMIN ATOR - 2", ORNL - 4078, April 1967.
1763 021 FORM 8 NES 205 5/79
I OCWEU NO.
NW NUCLEAR ENERGY SERVICES, INC.
REVISION LOG
"^o" E '
DATE PA$E DESCRIPTION APPROVAL gO g,
1 5/9/79 See CRA No. 821 2
12/10/79 See CRA No.1141 Since this CRA contains extensive changes to report 81 A0567 which affected every page, the use of a triangle / revision number adjacent to the revised area (per 80A9003, item 8.1.3) was not used.
1/63 U22
,ono es m n.