ML22277A777
| ML22277A777 | |
| Person / Time | |
|---|---|
| Site: | 07103097 |
| Issue date: | 08/03/2022 |
| From: | Boyle R, Shaw D TN Americas LLC |
| To: | Division of Fuel Management |
| Garcia-Santos N | |
| Shared Package | |
| ML22277A716 | List:
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| References | |
| A33010, L-2022-DOT-0008 | |
| Download: ML22277A777 (13) | |
Text
Unrestricted Orano Orano NPS SAFETY ANALYSIS CHAPTER 2.5 REPORT CRITICALITY-SAFETY ANALYSIS
Prepared by 0 orano FCC4 / FCC3 Checked by
Identification : DOS-19-021165-014-NPV Version : 1.0 Page 1 / 13
TableofContents Statusofrevision 2 Summary 3
- 1. Purpose 4
- 2. Inputdata 4
- 3. Calculationmethod 5
- 4. Analysisofpackagingloadedwithassemblytypes15x15, 17x17,16x16and18x18 6
- 5. Analysisofpackagingloadedwithrodsinrodboxesorchannels 8
- 6. Analysisofpackagingloadedwithassemblytypes14x148footand14x1410 foot 9
- 7. Analysisoftheincreaseinthecrosssectionofthecavity 10
- 8. Analysisofthepossiblepresenceofglycerine 11
- 9. Analysisoftheimpact ofthepresenceofsmoothwalleddummiesandEPRassemblymodel 11 10.Analysisoftheimpactofdamageonthedoorfortheinternal fittingsinFCC3andFCC4packagings,aftera1mdropontoabar atthemaximumtemperaturereachedduringNCT 11 11.Analysisoftheimpact ofachromethickness(inexcessthicknessornot)ontothe rodscladdingzirconiumalloy 11 12.Conclusion 12 13.References 12 14.Listofcriticalitysafetyappendices 13
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Status of revision
English French Date Purpose and record of changes Prepared by / Checked by version version
Old reference : DOS-13-00081778-500
First issue 00 00 04/2012 Revision of Appendix 11 of Safety Analysis Report TFXDC 2158 revision H and Appendix 12 of Safety Analysis Report TFXDC 2159 revision G.
Appendix 2.5-4 added 01 01 09/2012 Justification of the increase in cavity cross section, from x mm² to x mm².
02 02 12/2015 Inclusion of the potential presence of a maximum of 5 g of glycerine on the fuel assemblies.
Deletion of reference <2>,
FF DC 00561 rev C: Transport packages for fresh UO2 fuels. Determination of uncertainties to be applied in criticality studies performed with CRISTAL.
Appendix 2.2-5 added I Inclusion of contents XL and EPR used to increase the cavity cross section to x mm².
03 03 12/2016 Appendix 2.2-6 added Determination of uncertainties to be applied to criticality studies performed with CRISTAL.
Shift of the paragraph in Chapter 2.1-13, dealing with the impact on -criticality safety of the smooth-walled dummy to this chapter.
Addition of the impact of damage on the door for the internal fittings in FCC3 and FCC4 packagings, after a 1m drop onto a bar at the maximum temperature reached during NCT.
Integration of 17x17 GAIA 14-foot assemblies 1.0 04 10/2020 Addition of a potential thickness of chrome, excess thickness or not, onto the rods I cladding zirconium alloy I New reference: DOS-19-021165-014-NPV
Deletion of the case of limited number of rods without wedging.
See first Addition of reference <3>, which justify that the model of package arrays 1.0 1.0 page used in demonstration of Appendices 2.5-1 and 2.5-5 is penalizing compared to package arrays.
Formal modifications.
BRAND NAMES
AFA 3G, M5, Q12, MONOBLOC, TRAPPER, AGORA, HTP, HMP and ROBUST FUELGUARD are brands or registered brands of Fr amatome or its subsidiaries, in the USA or in other countries.
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Summary
The purpose of this Chapter is to describe the criticality-safety analysis of the FCC3 and FCC4 model packages in conditions of transport according to the regulations (reference <1>) applicable to industrial type containers loaded with fissile materials.
This analysis includes:
The criticality-safety analysis of the FCC3 and FCC4 packagings loaded with assembly types 15x15, 17x17 12-foot, 17x17XL, XLR and GAIA, 17x17 EPR type, 16x16 and 18x18,
The criticality-safety analysis of the FCC3 pa ckaging version 2, loaded with assembly types 14x14 foot and 14x14 foot.
The criticality-safety analysi s of the FCC3 and FCC4 packagings loaded with rods in boxes.
The analyses were conducted on the package in isolat ion and on an array of packages with the same criterion in both configurations: K eff 0.95, all uncertainties included.
Criticality-safety is ensured for the FCC3 and FCC4 packagings loaded with assembly types 15x15, 17x17 12-foot, 17x17 XL, XLR and GAIA, 17x17 EPR, with a CSI = 0.625.
Criticality-safety is ensured for the FCC3 and FCC4 packagings loaded with assembly types 16x16 and 18x18, with a CSI = 8.33.
Criticality-safety is ensured for the FCC3 and F CC4 packagings loaded with type 14x14 assemblies, with a CSI = 0.
Criticality-safety is ensured for the FCC3 and F CC4 packagings loaded with non-assembled fuel rods in boxes, with a CSI = 0.
Finally, gadolinium-bearing rods with a minimum content of 2 % on a matrix with a maximum uranium-235 enrichment of % can be transported irrespective of the number of rods and type of packaging used. These rods have a K infinite value less than 1.0 irrespective of the moderation used.
Respect for the criticality-safety criteria of the FCC4 Version 1 and FCC3 Version 1 packagings, loaded with non-assembled rods or fuel assembles, is not brought into question for cavity cross sections of up to mm x mm.
The presence of glycerine (max. 5g) does not make any difference to the conclusions of the criticality-safety analyses.
The presence of a smooth-walled dummy or EPR assembly model does not make any difference to the conclusions of the criticality-safety analyses.
The presence of chrome localized on the rods cladding (in excess thickness or not onto the zirconium alloy) does not make any difference to the c onclusion of the criticality-safety analysis.
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- 1. Purpose
The purpose of this Chapter is to describe the criticality-safety analysis of the FCC3 and FCC4 model packages in conditions of transport according to the regulations (reference <1>) applicable to industrial type containers loaded with fissile materials.
This analysis includes:
The criticality-safety analysis of the FCC3 and FCC4 packagings loaded with assembly types 15x15, 17x17 12-foot, 17x17XL, XLR and GAIA, 17x17 EPR type, 16x16 and 18x18,
The criticality-safety analysis of the FCC3 pa ckaging version 2, loaded with assembly types 14x14 foot and 14x14 foot.
The criticality-safety analysi s of the FCC3 and FCC4 packagings loaded with rods in boxes.
The analyses were conducted on the package in isolat ion and on an array of packages with the same criterion in both configurations: K eff 0.95, all uncertainties included.
- 2. Input data
2.1. Note on regulatory context
These containers are categorised as "Type II Fi ssile Industrial Packages" in accordance with IAEA Regulations (reference <1>) and must by des ign ensure sub-criticality for a container in isolation or a number of containers in Routine Conditions of Transport (RCT), in Normal Conditions of Transport (NCT) and in Accidental Conditions of Transport (ACT), all in accordance with IAEA R egulations (reference <1>).
Compliance with Criticality-safety is ensured for conditions of maximum neutron multiplication.
The analyses were conducted on the package in isolation and on an array of packages with the same criterion in both configurations: K eff 0,95, all uncertainties included (See Appendix 2.5-6).
2.2. Description of packagings and contents
A description of the packagings is given in Chapter 1.4 of the FCC3 and FCC4 Safety Analysis Reports.
A description of the contents is given in Chapt er 1.3 of the FCC3 and FCC4 Safety Analysis Reports.
The FCC packaging model involves confining the f uel inside a volume of known cross-section that is as small as possible and mechanically guaranteeing that this geomet ry is preserved on completion of regulatory tests.
The isolation system is descri bed in Chapter 1.4 of the Safety Analysis Reports for the FCC3 and FCC4 packagings.
The packages have no specific characteristics desi gned to prevent the infiltration of water; the presence of a given quantity of water is therefore assumed within the empty spaces of the package.
The damaged condition of the package is then defi ned as that resulting from the cumulative effect of the tests representative of normal c onditions of transport and ac cidental conditions of transport (See Chapters 2.1 and 2.2 of the FCC3 and FCC4 Safety Analysis Reports). This package condition covers the regulatory conditi ons defined by paragraphs 682 b) and c) for the package in isolation, and by paragraphs 684 and 685 for multiple packages.
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The undamaged condition of the package is then defi ned as that resulting either from routine conditions of transport or from drop tests in normal conditions of transport with regard to the slight damage possible to the shells of the packag ing (slight localised indentation; see Chapter 2.4).
The fuel rods remain intact and the cross-section of the packaging remains unchanged at the end of the regulatory tests. Details of the ca lculation cases considered are given below.
- 3. Calculation method
Appendix 2.5-6defines the uncertainties to be applied when using the CRISTAL form for the transportation of fresh UO2 fuels.
The number (N) of packages, as defined by Articles 684 and 685 of reference <1>, which can be transported from a criticality-safety standpoint is used to determine the Criticality-Safety Index (CSI) by the formula:
CSI = 50 / (min (NNCT ; NACT))
3.1. Isolated package
The package in isolation, damaged or in rout ine condition (undamaged), must be sub-critical with a total reflection by a 20 cm layer of water around the enclosure.
3.1.1. Package in isolation undamaged
The configuration studied is: package undamaged (cylindrical model) filled with water and surrounded by a reflector of 20 cm of water.
3.1.2. Isolated, damaged package
The geometrical configuration studied is: package damaged (rectangular model) surrounded by a reflector of 20 cm of water.
The space between the cavity and the shell is:
either full of water (the package is also filled with water),
or filled with vacuum (differential draining of the package is considered).
3.2. Multiple packages
Multiple packages modelled in the form of an array must be sub-critical, with the package array surrounded on all sides by a 20 cm layer of water.
3.2.1. Array of undamaged packages
The configuration studied is an infinite array of undamaged packages (cylindrical model).
Total reflection conditions are applied to all surfaces of the package.
The space between the cavity and the shell is:
either full of water: the package is also filled with water,
or filled with vacuum: differ ential draining of the package.
3.2.2. Array of damaged packages
An array of 2N damaged packages, in c ontact, is modelled under ACT (Paragraph 685 of Reference <1>).
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A reflector of 20 cm of water at the periphery of the array is taken into account.
In this configuration two variants are analysed:
The interior of the shell is empty, the fuel is moderated: differential draining.
the interior of the shell is full of water, the fuel is moderated.
3.3. Package modelling
The undamaged package is modelled by a cylindrical shell of mm in diameter. The damaged package is modelled by a rectangular shell of x mm with an unchanged neutron cavity.
Using a conservative approach, the bounding dim ensions of the rods are applied in the calculations, i.e. the maximum radius of the UO 2 pellets and the minimum outside radius of the zirconium alloy claddings.
The sensitivity study in reference document <2> defines these parameters as being conservative for the crit icality-safety studies.
3.4. Modelling of resin
In accordance with the results of the tests reported in Appendix 2.2-4, the model adopted for the resin in accidental conditions of transport is as follows:
% boron loss through the whole thickness,
% hydrogen loss over mm at the external surface.
- 4. Analysis of packaging loaded with assembly types 15x15, 17x17, 16x16 and 18x18
The criticality-safety study described in Appendix 2.5-1 serves to verify compliance with criticality-safety criteria of the FCC3 packaging version 1 (Version 100 and 200), the FCC4 packaging version 1.a (Version 400 and 600) or 1.b (Version 600) and the FCC4 packaging version 2 (Version 500) for the transport of fresh UO2 fuel assembly types:
15x15 with % 235U enrichment transported in the FCC3 packaging version 1,
17x17 12-foot, with % 235U enrichment transported in the FCC3 packaging version 1,
17x17 XL with % 235U enrichment transported in FCC4 packaging version 1.a and FCC4 packaging version 1.b, This study is applicable to 17x17 XLR assemblies (where only the altitude of the grids changes from the 17x17 XL assembly ) and GAIA assemblies. In fact, as shown in Chapter 1.3, the small differences between the 17x17 XL and 17x17 GAIA assemblies (mainly due to the external diameters of the guide tubes and instrumentation tubes) have no significant impact on the criticality-safety of the package model. The increase in these diameters has a tendency to reduce the moderating ratio, and therefore reactivity. Studies involving 17x17 XL assemblies therefore cover GAIA assemblies. Moreover, to assure conservatism, the guide tubes are replaced by water during the study. Thus, throughout the cr iticality study covered in this chapter, the justifications for the 17x17 XL assemblies are equally valid for 17x17 XLR & GAIA assemblies.
17x17 EPR with % 235U enrichment transported in F CC4 packaging version 1.b,
16x16 with % 235U enrichment transported in the FCC4 packaging version 2,
18x18 with % 235U enrichment transported in the FCC4 packaging version 2,
having the following prin cipal characteristics:
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Density 10.96 g/cm3 (100 % of the theoretical density),
Complete assemblies: any missing UO2 rods mu st be replaced by gadolinium-bearing rods, rods containing depleted uranium or a metallic material, or solid metallic bars (materials such as graphite and beryllium are strictly excluded) to avoid in creasing assembly moderation levels; the presence of neutron-absorbing poisons in the metal bars is permitted; glycerine may be present up to a content of 5 g max within the assemblies;
The uranium may originate from reprocessing as the 234U and the 236U present in enriched reprocessed uranium (ERU) are poisons which reduce the reactivity of the assemblies,
the pellets may contain capturing compounds (chromium oxide for example); their positive influence in terms of margins is disregarded.
The rods cladding material may include a chrome thickness (in excess thickness or not onto the zirconium alloy), neutron absorbing material which leads to reduced the reactivity,
If the cladding has an excess thickness of chrome, the moderating ratio into the assembly is reduced, which leads to a reduction of the reactivity.
The packages may only contain one or two assemblies of the same type and having a maximum enrichment of:
% 235U for arrays of 15x15, 17x17 12-foot, 17x17 XL, XLR, GAIA and 17x17 EPR assemblies,
% 235U for arrays of 16x16 and 18x18 assemblies.
The calculations are performed taking into account an expansion of the array over a third of its length, this being a bounding value resulting from the conclusi ons of the regulatory tests presented in Chapter 2.1. Differential slippage of the rods is analysed and has no significant impact on the reactivity of the rods. The maximum number of packages considered in the case of an array of packages is given in the following table for the different assembly types:
Cavity Number N of Assembly Number of cross-packages in the array
% U235 rods per Packaging section CSI assembly (mm x NCT ACT mm) 15 x 15 Infinite 80 0.625 5 204 17 x 17 FCC3 V1 foot 264 Infinite 80 5
x XLR & GA 264 FCC4 V1.a FCC4 V1.b 5
EPR 265 FCC4 V1.b Infinite 80 0.625
% U235 16 x 16 Infinite 6 8.33
% U235 236 FCC4 V2 x 18 x 18 6 8.33
% U235 300 Infinite
The analysis also includes a sensitivity study on neutron cavity misalignment and ejection of pellets (phenomena not observed following the tests representative of NCT and ACT). This sensitivity study considers grouping of the UO2 mass ejected in the form of a sphere, and remoderation of the assemblies by replacement of the burst rods by water.
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The above table takes into account the possibility of pellet ejection with remoderation of the assemblies. From a criticality standpoint, the number of rods per assembly that can eject the totality of their pellets is:
10 for assembly types 16x16 and 18x18 transported in FCC4 packaging version 2,
10 for assembly types 15x15 and 17x17 12-foot, transported in FCC3 packaging version 1,
7 for type 17x17 XL, XLR and GAIA assemblies transported in FCC4 packaging version 1.a and version 1.b,
7 for type 17x17 EPR assemblies trans ported in FCC4 packaging version 1.b.
- 5. Analysis of packaging loaded with rods in rod boxes or channels
Under certain circumstances it is necessary to transport variable quantities of rods. These rods are transported in boxes, as described in Chapter 1.3 of this Report, which replace the assemblies inside the neutron cavity.
The packagings used to transport the rods are the sa me as those designed to transport assemblies:
FCC3 packaging version 1 for the transport of r od types 14x14 "8-foot", 14x14 "10-foot", 15x15 and 17x17 12-foot,
FCC4 packaging version 1 for the transport of rod types 16x16,17x17 XL, XLR & GAIA, 17x17 EPR and 18x18, and also 14x14 "8-foot", 14x14 "10-foot", 15x15 and 17x17 12-foot.
In the study presented in Appendix 2.5-2, complian ce of the FCC3 packaging version 1 and the FCC4 packaging version 1 with critical ity-safety criteria is verified for the transport of fresh UO 2 fuel rods having the following prin cipal characteristics:
maximum enrichment: % U235,
Density 10.96 g/cm3 (100 % of the theoretical density),
nominal diameter mm or mm,
any length, therefore slippage possible,
any number of rods.
Any special rods (gadolinium-bearing rods, depleted uranium rods, metal bars, zirconium alloy bars) are covered by the UO2 rods.
The pellets may contain capturing compounds (chrom ium oxide for example); their positive influence in terms of margins is disregarded.
The metal or zirconium alloy bar s may contain a neutron poison.
The rods cladding material may include a chrome thickness (in excess thickness or not onto the zirconium alloy), neutron absorbing material which leads to reduced the reactivity.
The maximum fissile cross-section, per cavity, for the rods must be mm x mm.
The model is a 2D modelling scheme. The rod box is not modelled. The radial spacer is modelled either by water or air whichever is the more cons ervative case. The calculations are performed under optimum moderation conditions. Differential slippage of the rods is analysed and has no significant impact on the reactivity of the rods.
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The following results are presented in Appendix 2.5-2:
Keff max = 0.782 for the package in isolation,
Keff max = 0.830 for an infinite array of packages.
These results thus indicate a large margin compared with the adopted criteria (0.95 for isolated package and package array).
The number N for transportation of the rods is infi nite in NCT and ACT; we therefore have CSI = 0.
Finally, gadolinium-bearing rods with a minimum content of 2 % on a matrix with a maximum uranium-235 enrichment of % can be transported irrespective of the number of rods and type of packaging used. These rods have a K infinite value less than 1.0 irrespective of the moderation used.
- 6. Analysis of packaging loaded with assem bly types 14x14 foot and 14x14 foot
Fresh UO2 fuel assemblies are transported in FCC3 pac kaging version 2 for type 14x14 "8-foot" and 14x14 "10-foot" assemblies.
This study, described in Appendix 2.5-3, serves to ensure compliance of the FCC3 packaging version 2 with criticality-safety criter ia for the transport of fresh UO2 fuel assembly types:
14x14 "8-foot" with % 235U enrichment,
14x14 "10-foot" with % 235U enrichment,
having the following prin cipal characteristics:
Density 10.96 g/cm3 (100 % of the theoretical density),
Complete assemblies: any missing UO2 rods must be replaced by gadolinium-bearing rods, rods containing depleted uranium or a metallic material, or solid metallic bars (materials such as graphite and beryllium are strictly excluded) to avoid in creasing assembly moderation levels; the presence of neutron-absorbing poisons in the metal bars is permitted; glycerine may be present up to a content of 5 g max within the assemblies;
the uranium may originate from reprocessing as the 234U and the 236U present in enriched reprocessed uranium (ERU) are poi sons which reduce the reactivity of the assemblies and which are not taken into account in the criticality-safety assessment.
The rods cladding material may include a chrome thickness (in excess thickness or not onto the zirconium alloy), neutron absorbing material which leads to reduced the reactivity,
If the cladding has an excess thickness of chrome, the moderating ratio into the assembly is reduced, which leads to a reduction of the reactivity.
The model is a 2D modelling scheme. The cross-section of the neutron cavity is x (mm x mm) with 2 continuous pad recesses of maximum cross-section x (mm x mm). The assembly cross-section is considered in 3 configurations:
non-expanded ( mm x mm),
expanded to the cross-section of the neutron cavity ( mm x mm),
expanded to the cross-section of the neutron cavity and in the pad recesses.
The number of rods able to eject pellets is 20 rods.
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Differential slippage of the rods has no significant impact on the reactivity of the rods. Longitudinal wedging of the assemblies is not necess ary from the criticality standpoint.
The transportation of assemblies equipped with control cl usters is covered by this study as the control clusters contribute a neutron-absorbing material wi thin the assembly while reducing the moderation of the assemblies.
The number N for the transportation of 14x14 8-foot and 14x14 10-foot assemblies is infinite in Normal Conditions of Transport (NCT) and in Accidental Conditions of Transport (ACT); giving us a CSI of 0.
- 7. Analysis of the increase in the cross section of the cavity
Appendix 2.5-5 studies the impact of an increase in the cross section of the cavity when transporting fuel assemblies. The calculations are carried out with the FCC4 Version 1 packaging.The results of this study are also applicable to the FCC3 Version 1 packaging insofar as the FCC4 Version 1 is identical to the FCC3 Version 1 apart from the lon ger cavity. This appendix only covers FCC3 Version 1 and FCC4 Version 1 packagings.
7.1. Loading of the fuel assemblies
This impact study is carried out for a 17x17 XL or 17x17 EPR assembly, with a % 235U enrichment level, loaded into an FCC4 Version 1 packaging. The study is carried out for a array of packages (N=80) under accidental conditions (conditions leading to the maximum reactivity, as detailed in Appendix 2.5-1). The document in reference <3> shows that this modelling of the multiple packages in a array leads to obtain a K eff max statistically equivalent.
The cavity is modelled with dimensions of x mm. The other dimensional parameters of the packaging (thickness of plates and resin, lengths, shells, etc.) are unchanged. The assemblies are expanded to a thir d of their height, up to the cavity wall. The other modelling assumptions are identical to those of Appendix 2.5-1.
Thus, under regulatory transport conditions, the in crease in the reactivity of an array of packages, taking into account the increase in cavity section to x mm (or an increase of 2.3%) and assemblies covering a third of its height, up to the wall of the cavity, is between 700 and 800 pcm, which does not bring in to question the respect for the usually applied criticality-safety criter ia. The maximum reactivity for fuel assemblies loaded into a FCC3 Version 1 and FCC4 Version 1 packaging is 0.950, all uncertainties included, for a criterion of 0.95, therefore the result of th is impact study can be extended to assemblies transported in FCC3 Version 1 or FCC4 Vers ion 1 packagings. As explained in Paragraph 4, these results are also applicable to the transport of 17x17 XLR and GAIA assemblies.
7.2. Loading of non-assembled fuel rods
The impact study in Appendix 2.5-4 is comp leted for non-assembled fuel rods of mm in diameter (most conservative case dealt with in Appendix 2.5-2, loaded into an FCC Version 1 packaging). This study is carried out for an array of packages under accidental conditions (conditions creating the maximum re activity in Appendix 2.5-2).
The cavity is modelled with dimensions of x mm. The other dimensional parameters of the packaging (thickness of plates and resin, lengths, shells, etc.) are unchanged. The fuel rods fill the available space minus the thickness of the spacer - or, x mm (assuming a cavity cross section of x mm). The other modelling assumptions are identical to those of Appendix 2.5-2.
Thus, under regulatory transport conditions, the increase in the reactivity of the infinite grid of packages, taking into account the increased cavity cross section (to x mm) and with fuel rods at optimal moderation within a section of x mm (an increase of 3.6 %), is in the order of 600 pcm, but does not bring into question any of the usually applied criticality-safety criteria. The maximum reactivity for non-assembled rods loaded into a FCC3 Version 1
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and FCC4 Version 1 packaging is 0.830, all unce rtainties included, for a criterion of 0.95, therefore the result of this impact study can be extended to non-assembled rods transported in FCC3 Version 1 or FCC4 Version 1 packagings.
- 8. Analysis of the possible presence of glycerine
Assuming 5g of glycerine divided among 264 rods and over the length of the 900 MWe ( mm) rod, the thickness of the glycerine around the rods is equal to mm, which is negligible. This conclusion is also valid for PWR 15x15 and PWR 14x 14 assemblies, in fact, assuming the glycerine divided among 179 x 14x14 PWR fuel rods (which cove rs the case of 15x15 PWR rods) this thickness is mm, which is in the same order of magnitude as the 17x17 PWR assemblies. As an indication, in terms of the 17x17 PWR assemblies, even assuming a conservative length of mm, corresponding to a 17x17 assembly grid, the depth remains negligible ( mm).
In addition, glycerine contains less hydrogen than wate r (the chemical composition of glycerine being C3H8O3, with a hydrogen content of 6.59 x 10 22 atoms per cm3, which is lower than that of hydrogen in water (6.69 x 10 22 atoms per cm3).
Finally, glycerine is soluble in water.
As a consequence, assuming the three points covered above, the presence of 5g of glycerine will have no impact on the conclusions of the critical ity-safety study undertaken within this chapter.
- 9. Analysis of the impact of the presence of smooth-walled dummies and EPR assembly model As the smooth-walled dummy and the EPR model are free of fissile materials and are made up of an inert material (stainless steel and/or tungsten carbi de) there is a decrease in the reactivity of the package with regard to the calculated conf igurations containing fuel assemblies.
The presence of a smooth-walled dummy or model has no impact on the aforementioned demonstrations of criticality-safety.
- 10. Analysis of the impact of damage on the doo r for the internal fittings in FCC3 and FCC4 packagings, after a 1m drop onto a bar at the maximum temperature reached during NCT
During the drop on to a bar, the bar perforates the outer shell to impact on the door of the packaging, without piercing it. The assembly is therefore subjected to deformation during the drop onto a bar, caused by the deformation of the door.
The slight additional deformation of the door wi ll have the consequence of creating a localised reduction in the cross-section of the cavity, which wi ll have no impact on the criticality-safety studies (moreover, the conservative conditi ons, in terms of the criticality-safety modelling, correspond to a maximum cavity size with regard to the possi ble phenomenon of expansion of the array of rods and not to an indentation causing a reduction in the fiss ile cross-section and therefore a decrease in the reactivity of the package).
- 11. Analysis of the impact of a chrome thickness (in excess thickness or not) onto the rods cladding zirconium alloy
Chrome is a neutron absorbing, its presence onto the rods cladding has no impact on the aforementioned demonstrations of criticality-safety.
The same applies in the case of excess thickn ess of chrome around the cladding, which reduced the moderation ratio of the assembly.
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- 12. Conclusion
Criticality-safety is ensured for the FCC3 and FCC4 packagings loaded with assembly types 15x15, 17x17 12-foot, 17x17 XL, XLR and GAIA, 17x17 EPR, with a CSI = 0.625.
Criticality-safety is ensured for the FCC3 and F CC4 packagings loaded with assembly types 16x16 and 18x18, with a CSI = 8.33.
Criticality-safety is ensured for the FCC3 and FCC4 packagings loaded with type 14x14 assemblies, with a CSI = 0.
Criticality-safety is ensured for the FCC3 and FCC4 packagings loaded with non-assembled fuel rods in boxes, with a CSI = 0.
Finally, gadolinium-bearing rods with a minimum content of 2 % on a matrix with a maximum U235 enrichment of % can be transported irrespective of the number of rods and type of packaging used.
These rods have a K infinite value less than 1.0 irrespective of the moderation used.
Respect for the criticality-safety criteria of the FCC4 Version 1 and FCC3 Version 1 packagings, loaded with non-assembled rods or fuel assembles, is not brought into question for cavity cross sections of up to mm x mm.
The presence of glycerine (max. 5 g) does not make any difference to the conclusions of the criticality-safety analyses.
The presence of a smooth-walled dummy or EPR assembly model does not make any difference to the conclusions of the criticality-safety analyses.
The presence of chrome localized on rods cladding (in excess thickness or not onto the zirconium alloy), does not make any difference to the conclusions of the criticality-safety analyses.
- 13. References
<1> Regulations for the Safe Transport of Radioactive Materials - at the revision indicated in Chapter 1.2.
<2> Transnucléaire document 5532-B-2 Ver. 1 - "Envelope geometry for PWR and BWR fuel rods for criticality-safety studies".
<3> Framatome document FS1-0039573EN Rev. 1.0 - << Safety Criticality analysis FCC4 Fuel shipping casks Study of an array of damaged packages under Accident Conditions of Transport >>
Non-proprietary version Formulaire : PM04-4-MO-6 rév. 03 Unrestricted Orano Orano NPS Identification : DOS-19-021165-014-NPV Vers. 1.0 Page 13 sur 13
- 14. List of criticality safety appendices
Appendix Title Number of pages Framatome document - FFDC 00817 rev. 4.0 2.5-1 Criticality-safety analysis of FCC3 and FCC4 containers - Assembly types 2+37 15x15 and 17x17 - Assembly types 17x17 XL, 17x17 EPR,16x16 and 18x18 Framatome document - FFDC 01046 rev. 3.0 2.5-2 "Criticality-safety analysis of FCC3 and FCC4 containers - Transport of rods in 2+33 boxes" AREVA NP document - FFDC 01106 rev. A 2.5-3 Criticality-safety analysis of FCC3 containers version 2 - Assembly types 2+27 14x14 8-foot and 14x14 10-foot" Memo AREVA NP - FF DC 04191 Revision B 2.5-4 Criticality-safety analysis of FCC4 containers - Assembly types 17x17 XL - 2+7 justification for increase in cavity cross section AREVA NP document - FS1 0012857 rev. 1 2.5-5 Criticality-safety analysis of FCC4 containers - Justification of the increase in 2+5 cavity cross-section AREVA NP document - FFDC 00561 rev. 4 2.5-6 Criticality-safety analysis - Transport packages for fresh UO2 fuels. 2+14 Determination of uncertainties to be applied to criticality studies performed with CRISTAL.
TOTAL 135
Non-proprietary version