ML22277A776
| ML22277A776 | |
| 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:
|
| References | |
| CAC A33010, EPID L-2022-DOT-0008, E-61285 | |
| Download: ML22277A776 (9) | |
Text
AREVA TN AREVA UNRESTRICTED DISTRIBUTION NUCLEAR LOGISTICS OPERATIONS A SAFETY 2.4. ANALYSIS OF EXTERNAL DOSE RATES AREVA Form: PM04-4-MO-6 Rev. 00 ANALYSIS REPORT Prepared by Identification DOS-13-00081778-400-NPV FCC3-FCC4 Checked by Rev. 01 Page 1 / 9 TN International Table of Contents Revisions history 2
- 1. Purpose 4
- 2. Description of the package 4
- 3. Evaluation under routine conditions of transport 4
- 4. Evaluation under normal conditions of transport 4
- 5. Impact of the smoothwalled dummy on radiation protection 6
- 6. Conclusion 6
- 7. References 6 List of figures 7 Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 2 of 9 Revisions history Rev. Date Purpose and record of changes Prepared by / Checked by First issue 0 04/2012 Revision and amplification of Appendix 16 of the Safety Analysis Report TFXDC 2158 revision H (FCC4) and Appendix 17 of TFXDC 2159 revision G (FCC3)
Update of the maximum mass of the loaded FCC4 package to 5,550 kg See first Correction to the distance value of the fuel assembly - external shell for FCC3 1
page under NCT Addition of impact of the smooth-walled dummy Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 3 of 9
SUMMARY
The purpose of this Chapter is to describe the analysis of external dose equivalent rates for the FCC3 and FCC4 model packages in routine and normal conditions of transport in accordance with the regulations (reference <1>) applicable to industrial type containers loaded with fissile materials.
The dose equivalent rate criteria applicable to industrial type packages loaded with fissile materials are as follows:
Routine conditions of transport: 2 mSv/h in contact with the package.
Normal conditions of transport: less than 20% increase in the maximum radiation intensity in contact with the package.
The evaluation of radiation intensities in contact with the package gives the following results:
FCC3. FCC4.
Routine transport conditions mSv/ h < 2 mSv/h mSv/ h < 2 mSv/h (maximum)
I Variation in maximum radiation intensity under normal transport 9.0 % < 20% 9.7 % < 20%
conditions The regulatory criteria are satisfied.
Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 4 of 9
- 1. Purpose The purpose of this Chapter is to describe the analysis of external dose equivalent rates for the FCC3 and FCC4 model packages in routine and normal conditions of transport in accordance with the regulations (reference <1>) applicable to industrial type containers loaded with fissile materials.
The dose equivalent rate criteria applicable to industrial type packages loaded with fissile materials are as follows:
Routine conditions of transport: 2 mSv/h in contact with the package, Normal conditions of transport: less than 20% increase in the maximum radiation intensity in contact with the package.
- 2. Description of the package A description of the FCC3 and FCC4 packagings is given in Chapter 1.4 of this Report. A description of the contents of FCC3 and FCC4 package models is given in Chapter 1.3 of this Report.
- 3. Evaluation under routine conditions of transport 3.1. Radiation level in contact with the package Figure 2.4-2 gives the variation in dose equivalent rate versus the fuel assembly surface-shell surface distance.
The shortest distance between the fuel assembly and the surface of the shell is mm for the FCC4 packaging and mm for the FCC3 packaging (see figure 2.4-1).
Thus the dose rate at the surface of the package is:
Distance between fuel Dose equivalent assembly and outer shell rate (mm) (Sv/h)
FCC4.
FCC3.
These radiation level values are below the criterion of 2 mSv/h.
- 4. Evaluation under normal conditions of transport It is demonstrated that during tests simulating normal conditions of transport the increase in the maximum dose rate in contact with the package is less than 20%.
The regulatory tests, designed to prove the capacity of the package to resist NCT, are as follows:
water spray test, free-fall drop test, stacking test and penetration test.
The water spray test has no negative effect on the behaviour of the packaging in relation to radiation protection risks. On the contrary, water would have a slight effect of biological shielding by decreasing the dose rate.
Chapter 2.1 of the safety analysis reports for the FCC3 and FCC4 packagings shows that the stacking test does not result in any damage or deformation of the container. There is therefore no variation in the dose rate on the surface of the package following this test.
The penetration test which consists of dropping a 6 kg bar from a height of 1 metre can produce local deformation of the shell but without rupturing it (see Chapter 2.1 of this Report). This deformation remains less than that observed during a free drop of the package.
For this reason the conservative case corresponds to the free drop test. The following analysis is therefore based solely on this configuration.
Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 5 of 9 The principle of this demonstration is to assess the effects of a 1.2 metre drop (FCC3) or a 0.9 metre drop (FCC4) on the geometry of the packaging based on the observations made following the 9-metre drop of prototype 2 (FCC4) and to deduce therefrom the variation in maximum radiation intensity at the surface of the package. This approach is conservative for the FCC3 package model in that the deformation observed on the prototype is representative of the deformation of an FCC4 package model for which the potential drop energy is greater.
The types of damage which can have an impact on the maximum radiation intensity are deformation of the outer shell and deformation of the cradle which can reduce the distance between the fuel assembly and the external surface of the packaging.
The minimum distance between the fuel assembly and the outer shell is the distance between the corner of the door and the shell (see Figure 2.4-1). This distance is used below and it will be assumed that the deformation of the shell is directly related to the drop energy (i.e. the potential drop energy of the package). The mass of prototype 2 is 5262 kg. The crushing observed after the drop test are also corrected by determining the ratio of the maximum masses of the FCC3 package (4385 kg) and the FCC4 package (5550 kg) and the mass of prototype 2.
The deformation after the 1.2 metre or 0.9 metre drop is calculated by applying the following ratio of the deformation of the shell after the 9-metre drop.
h1 M package Reduction ratio h2 M proto Where:
h1: the distance between the middle of the packaging (lowest point) and the ground, for the 1.2 metre or 0.9 metre drop; where h1 is dependent on the packaging in question.
h2: the distance between the middle of the packaging (lowest point) and the ground, for the 9 m drop; where h2 is calculated from the real value of the drop height for Prototype 2 (or 9.14 m). h2 is determined from Prototype 2 = 9.14+ = m.
=--
The lowest point in the middle of the packaging is calculated taking into account the inclination of the packaging and the reduction ratios for the FCC3 and FCC4 packagings are as follows:
=--
FCC3 reduction ratio FCC4 reduction ratio According to Appendices 4 and 5 to the Prototype 2 Test Report attached as Appendix 2.1-10 to this Safety Analysis Report, the combination of deformation of the outer shell facing the frame and the frame-door assembly leads to a maximum reduction of the distance between the outer shell and the assembly of mm, which thus means a maximum reduction of this distance of mm after the 1.2-metre drop for the FCC3 packaging and mm after the 0.9 metre drop for the FCC4 packaging. Deformation shapes at the ends of the shell are ignored as the dose equivalent rates are maximised in the continuous section of the packaging, opposite the fuel assemblies being transported.
Figure 2.4-2 gives the variation in maximum dose rate depending on the distance between the fuel assembly and the outer shell. Maximum radiation intensities after the drop representative of normal conditions of transport are therefore evaluated:
Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 6 of 9 Distance between fuel Dose rate assembly and outer shell (Sv/h)
(mm)
FCC4, before drop FCC4, after drop FCC3, before drop FCC3, after drop This corresponds to a maximum increase in radiation intensity of 9.7 % for the FCC4 package model and 9.0 % for the FCC3 package model.
Following the drop representative of normal conditions of transport, the maximum increase in radiation intensity in contact with the surface of the packaging is below 20%.
- 5. Impact of the smooth-walled dummy on radiation protection The inclusion of the smooth-walled dummy has no impact on regulatory dose equivalent rates around the package, since it is made up of an inert material.
- 6. Conclusion The evaluation of radiation intensities in contact with the packaging gives the following results:
FCC3. FCC4.
Routine transport conditions (maximum) mSv/ h < 2 mSv/h mSv/ h < 2 mSv/h Variation in maximum radiation intensity under normal transport 9.0 % < 20% 9.7 % < 20%
conditions The regulatory criteria are satisfied.
- 7. References
<1> Regulations for the Safe Transport of Radioactive Materials - at the revision indicated in Chapter 1.2.
Non-proprietary version
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400-NPV Rev.: 01 Page 7 of 9 List of figures Figure Title No. of pages Location of the shortest distance between the surface of the assembly and the surface of the 2.4-1 1 package Variation in dose rate versus the distance between the assembly surface and the package 2.4-2 1 surface TOTAL 2 Non-proprietary version
AREVA RESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400 Rev.: 01 Page 8 of 9 Figure 2.4-1 Location of the shortest distance between the surface of the assembly and the surface of the package Shortest Section A-A UPPER SHELL distance Scale: 1/5 assgmbly See drawi ng No. 220 K 041 0 between assembly CO NFI GURATION FOR CO NFIGURATION FOR 14-FOOT ASSEMBLI ES OF THE surface and shell 14-FOOT ASSEMBLIES XL TYPE OR CHANNEL FOR NON-ASSEMBLED RODS surface before OF THE XLR TYPE See drawing No. 220 K 0400 RIGHT HAND DOO R sub-assembly drop See drawing No. 220 K 0400 LEFT HAND DOOR sub-assembly See drawi ng No. 220 K 0470 Shortest SUPPORT FRAME sub-assembly distance See d rawing No. 220 K 0445 between assembly surface and shell surface after CRAD LE sub-assembly drop See drawing No. 220 K 0430 INTERIOR EQUIPMENT assembly See drawi ng No. 220 K 041 5 LOWER SHEU asS8mbly S9e drawi ng No. 220 K 0405 This document and the information that it contains are the property of TN International and may not be communicated, reproduced or used without its prior written authorisation.
AREVA UNRESTRICTED DISTRIBUTION AREVA TN Form: PM04-4-MO-6 Rev. 00 NUCLEAR LOGISTICS OPERATIONS Identification: DOS-13-00081778-400 Rev.: 01 Page 9 of 9 Figure 2.4-2 VARIATION IN DOSE RATE VERSUS THE DISTANCE BETWEEN THE ASSEMBLY SURFACE AND THE PACKAGE SURFACE The values of dose equivalent rates are calculated by varying the distance between the surface of the assembly and the outer shell of the package. The variation of this distance simulates the denting of the shell following the drop of the package.
The calculations were performed using the qualified MICROSHIELD computer code version 5.05.
The calculations were made assuming a package loaded with ERU type fuel assemblies with a U235 enrichment of 4.95% and a proportion of U232 of 50 ppb (main parameter influencing the dose equivalent rate).
The equation given above the Figure corresponds to the trend curve established from the dose rate values calculated versus the assembly-shell distance.
Non-proprietary version