ML25119A286
| ML25119A286 | |
| Person / Time | |
|---|---|
| Site: | 07103034 |
| Issue date: | 01/21/2025 |
| From: | Shaw D Orano TN Americas |
| To: | Boyle R Office of Nuclear Material Safety and Safeguards, US Dept of Transportation, Pipeline & Hazardous Materials Safety Admin, US Dept of Transportation, Radioactive Materials Branch |
| References | |
| EPID L-2024-DOT-0002, E-64197 | |
| Download: ML25119A286 (1) | |
Text
Orano TN 7160 Riverwood Drive Suite 200 Columbia, MD 21046 USA Tel: 410-910-6900 Fax: 434-260-8480 January 21, 2025 E-64197 U.S. Department of Transportation Attn: Mr. Richard W. Boyle, Chief Pipeline & Hazardous Materials Safety Administration Radioactive Materials Branch 1200 New Jersey Avenue, S.E.
East Building, PHH-20 Washington, DC 20590
Subject:
Response to Request for Additional Information for the Revalidation of French Competent Authority Certificate F/313/B(U)F (Mbs) for the Model No. TN-BGC 1 (Docket 71-3034, EPID: L-2024-DOT-0002)
Reference:
[1] Letter from Don Shaw, TN Americas to Richard Boyle U.S.
DOT, Application for Revalidation of Competent Authority Certification, USA/0492/B(U)F for Revalidation of French Competent Authority Certificate F/313/B(U)F (Mbs) for the Model No. TN-BGC 1, dated July 23, 2024
[2] Letter from Pierre Saverot, U.S. NRC to Richard Boyle, U.S.
DOT, Request for Additional Information for the Revalidation of the French Certificate of Competent Authority No. F/313/B(U)F (Mbs) for the Model No. TN-BGC 1, dated November 22, 2024
Dear Mr. Boyle:
TN Americas LLC, on behalf of Orano NPS (NPS), requested a U.S.
Department of Transportation (DOT) revalidation of the French Approval Certificate F/313/B(U)F [1]. In response to the DOT request for recommendation, the U.S. Nuclear Regulatory Commission (NRC) requested additional information to complete their review of the application [2]. provides the responses from NPS to the additional information requested.
E-64197 U.S. DOT Page 2 of 2 Should you have any questions or require additional information to support review of this application, please contact Doug Yates by telephone at 434-832-3101, or by e-mail at douglas.yates@orano.group.
Sincerely, Don Shaw Licensing Manager cc:
Cheryl Laws, Project Manager, Orano TN (cover letter)
Renaud Le Blevennec, Project Manager, Orano TN (cover letter)
Enclosure:
- 1) RAIs and Responses SHAW Donis Digitally signed by SHAW Donis Date: 2025.01.21 13:19:34 -05'00'
RAIs and Responses to E-64197 Page 1 of 20 Structural Evaluation:
RAI 1
Provide documentation in the safety analysis report (SAR) for the TN-BGC 1 Type B(U) fissile transport package (e.g., CEA-DES-DDSD-DTEL-SGPE-DSEM-07600) describing how the aging mechanisms, on reusable structural components subjected to repetitive loading, have been addressed in accordance with section 613A of the 2018 edition of IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material.
Revision C of the SAR was reviewed for documentation of the consideration of the combined aging effects on the structural components of the TN-BGC 1 package (i.e., body, cover, and cage) and only the following were found: 1) a fatigue analysis for the tie-down stresses on the cage structure under normal overland transport conditions in attachment 7 of chapter 11, and 2) a qualitative assessment of the aging effects on some package components in section 4 of chapter 5. The expected service life of the package was not found in the SAR and therefore it does not appear to address package structural aging effects in accordance with the guidance of paragraph 613A.1 of the 2018 edition of IAEA SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material.
The combined structural aging effects on the package that need to be considered over the packages expected service life are those encountered during normal service, which includes assembly cycles, lifting/handling cycles, pressurization cycles, thermal cycles and vibration cycles. If certain categories of stress cycles are not applicable or negligible for certain package components, provide an explanation to justify their exclusion. Provide a quantitative assessment of the combined structural aging effects on the TNBGC 1 components in the SAR. With any fatigue evaluations performed in the scope of this assessment, provide copies of any referenced design fatigue diagrams with an explanation of the associated terminologies and units used.
The staff notes that, depending on the outcome of the package component structural aging management evaluations, the inspection and maintenance requirements for the package may also require updating.
This information is requested to determine compliance with the requirements in paragraph 613A of the IAEA SSR-6, 2018 Edition.
Response to RAI 1:
The TN-BGC 1 was designed to have a service life of 40 years. The possibility of extending it to 50 years is currently under review.
The thermal aging of stainless steel, wood, and neutron-absorbing resin is analyzed in [1]. The CEA further specifies that metallic materials such as aluminum bronze, bronze, and aluminum are not very sensitive to thermal aging at the temperatures reached in NCT.
It is specified in [1] that the wood of the TN-BGC 1 cover is dry and that it is not subject to variations in hygrometry because it is confined in the casing of the covers for which leaktightness is guaranteed, and because the temperatures reached in NCT are not likely to degrade its thermal or mechanical properties.
RAIs and Responses to E-64197 Page 2 of 20 It is also demonstrated in [1] that there is no risk of corrosion to the stainless-steel in contact with the resin. In addition, the risk of galvanic corrosion between the different metallic materials in contact in the package design is eliminated because the components are in a dry environment. More generally, the visual inspections performed during more than 30 years of maintenance operations, in order to detect any signs of external aging, have never revealed any signs of corrosion.
The maximum pressure likely to be reached in NCT in the cavity of a type B(U)F package model is 1.2 bar, which has no impact on the stressed components of the package model and is the reason why the impact of the pressurization cycles is not studied.
The CEA has not specifically studied the aging phenomena of the closure system parts due to the opening/closing cycles of the packaging because their frequency is very low between two maintenance campaigns (maximum 15 cycles over a period of 3 years). During these campaigns, the bayonet and clamping rings are visually inspected and checked using go/no-go gauges. If this check does not provide satisfactory results, the rings are replaced to avoid the risk of wear.
The number of lifting cycles during the handling and loading of the package on its conveyance between two maintenance campaigns is also very low and the stresses further reduced by the multiplicity of lifting devices (slings in horizontal position, forklift and lifting beam in vertical position). In addition, the maximum stress determined in the cage uprights on the basis of very conservative assumptions with regard to actual handling conditions (accelerations applied in the 2 g calculations) is only 114 MPa, which remains well below the acceptable criterion of Rp0.2/1.5 = 167 MPa. For information purposes, a fatigue analysis of the cage components stressed during the handling phases, carried out according to the methodology of Eurocodes 9 - EN 1993-1-3: 2007, shows that the minimum permissible number of cycles is 39,718, which is satisfactory with regard to the handling conditions described above.
The CEA, therefore, considers in [1] that only the components of the package model participating in its tie-down (screws for fastening the cage to the body of the packaging, plates and cage) are subjected to cyclic stresses that require a study of their fatigue strength. This study was carried out based on a so-called 3-band method, justified by the publication of Stamper E., 2017: Calculating Fatigue in a Random Vibration Environment, and assuming a distribution of accelerations recommended in SSG-26 according to a normal law (see figure below).
RAIs and Responses to E-64197 Page 3 of 20 The calculation applies the accelerations of [2] for the 3 band, 2/3 of the accelerations of [2] for the 2 band and 1/3 of the accelerations of [2] for the 1 band. The frequency of occurrence p for each band is p1 = 68.6% for the 1 band, p2 = 27.1% for the 2 band and p3 = 4.3% for the 3 band.
For each frequency of occurrence p, the service life N corresponding to a stress level is then obtained according to the criteria of Eurocodes 9 - EN 1993-1-3: 2007. Then, using Miner's rule (or linear damage rule) expressing the damage D according to the various stresses, the service life Ntotal is determined using the formula below:
=
1
= 1 1
+ 2 2
+ 3 3
This fatigue study shows a significant service life of these components (from 141,692 occurrences of acceleration, representative of extreme cases of braking or turning at a high speed, up to nearly two million occurrences depending on the tie-down configuration).
The CEA also reiterates that the lessons learned during operation and maintenance have not highlighted any damage to these components caused by the stresses encountered during the handling and tie-down phases.
References:
- 1. Safety Analysis Report CEA/DES/DDSD/DTEL/SGPE/DSEM 07600 Revision C, January 17, 2024.
- 2. Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, IAEA Safety Standards - No. SSG-26, 2018 Edition.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 4 of 20
RAI 2
Provide documentation in the SAR for the TN-BGC 1 Type B(U) fissile transport package describing how the package components employed for tie-down during normal conditions of transport have been designed employing the accelerations recommended during all modes of transport in the U.S., per table IV.2 of IAEA SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2018 edition, the guidance for section 638 of the 2018 edition of IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material.
Chapter 11 of revision C of the SAR was reviewed for documentation of the tie-down design of the structural components of the TN-BGC 1 package during both road and air transport. It appears that the normal road and air transport tie-down design conditions in chapter 11 attachments 7 and 8, respectively, consider the acceleration values presented in SSG-26, table VI.1 rather than those of table IV.2, listed for Type B(U) fissile packages transported in the U.S.
This information is requested to determine compliance with the requirements in paragraph 638 of the IAEA SSR-6, 2018 Edition.
Response to RAI 2:
The mechanical strength of the packaging in all authorized tie-down configurations for transport by air and land was indeed not studied while specifically taking into account the accelerations from Table IV.2 of Annex IV of [2], recommended for the transport of type B(U)F packages in the United States, which are:
10 g longitudinal, 5 g lateral, 2 g vertical.
Given the stresses determined by the French Alternative Energies and Atomic Energy commission (CEA) in the uprights of the aluminum cage, taking into account the accelerations recommended in Table IV.1 of Annex IV of [2] for transport by land and air, there is every reason to believe that the mechanical strength of the uprights could not be demonstrated while integrating the accelerations recommended for the transport of type B(U)F packages in the United States.
However, the package model is sized to withstand the drop tests representative of ACT, which generate accelerations on the package model of more than 200 g, much higher than those recommended for the transport of type B(U)F packages in the United States, as recalled above.
Thus, a failure of the aluminum cage uprights, which would be the first to yield, would have no impact on the safety functions of the package design. This complies with § 71.45 b/ 3) of Title 10 of the Code of Federal Regulations on the strength of tie-down systems.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 5 of 20 Materials Evaluation:
RAI 1
Provide an evaluation of degradation mechanisms and ageing processes for the TNBGC 1 package.
Per IAEA SSR-6, 2018 Edition, paragraph 613A: The design of the package shall take into account ageing mechanisms.
Per IAEA SSG-26, 2018 Edition, paragraph 613A.1: Packaging components and package contents are subjected to degradation mechanisms and ageing processes that depend on the component and the contents themselves and their operational conditions. Thus, the design of a package should take into account ageing mechanisms commensurate with intended use of the package and its operational conditions, as described in paras 613A.2-613A.6.The designer of a package should evaluate the potential degradation phenomena over time, such as corrosion, abrasion, fatigue, crack propagation, changes of material compositions or mechanical properties due to thermal loadings or radiation, generation of decomposition gases and the impact of these phenomena on performance of safety functions.
Except for the discussion of thermal aging in chapter 5 of the safety evaluation report, the staff was not able to locate an evaluation of degradation mechanisms and ageing processes as required by IAEA SSG-26, 613A.1.
This information is needed to determine compliance with the requirements in paragraph 613A of the IAEA SSR-6, 2018 Edition.
Response to RAI 1:
The CEA responded to questions concerning damage due to corrosion, thermal aging, and fatigue as part of the response to Structural RAI 1.
In addition, the CEA specifies that the TN-BGC 1 packaging, due to the non-irradiated contents it transports, is not subject to aging mechanisms under radiation, especially since this packaging is not intended for the interim storage of these materials. Additionally, the gas releases likely to occur in NCT consist of hydrogen and oxygen in very small quantities, which has no impact on the aging of the materials.
Regarding damage such as abrasion or the appearance and propagation of cracks in metal materials, the maintenance program implemented in response to Materials Evaluation RAI 2 makes it possible to prevent and, if needed, repair this damage, which keeps the packaging compliant with the associated safety requirements.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 6 of 20
RAI 2
Provide an inspection and maintenance program to address the effects of aging mechanisms on the package.
This program should include a description of any national or international codes, standards, and/or other methods, programs, or procedures that are implemented to ensure that package maintenance activities (including visual inspections, screening and evaluation of visual indications, and corrective actions such as component repairs and replacements) are adequate to manage the effects of aging in metallic package components that would see long-term use, such that the package components are capable of performing their requisite safety functions throughout the period of use.
The staff requests that this description address the following criteria:
- a. Inspection methods (e.g., bare metal visual exams and/or other types of nondestructive exams such as liquid penetrant exams or ultrasonic exams) for detection, characterization, and sizing of localized aging effects such as cracks, pits, and crevice corrosion.
- a.
Inspection equipment and personnel qualification requirements (e.g., lighting and visual acuity requirements for performing visual exams) to ensure reliable inspections that can adequately detect and characterize indications of localized aging effects prior to component failure or loss of safety function.
- b.
Acceptance criteria for aging effects such as early stage fatigue cracks and localized corrosion of stainless steel components, such as chloride induced stress corrosion cracking (SCC), pitting, and crevice corrosion. Examples of visual indications that may indicate potential localized corrosion of stainless-steel components include the accumulation of atmospheric deposits such as salts, buildup of corrosion products, rust-colored stains or deposits, and surface discontinuities or flaws associated with pitting, crevice corrosion, and/or SCC.
- c.
Describe any surface cleaning requirements that are implemented to ensure that bare metal visual inspections of component surfaces are capable of detecting surface flaws, and for ensuring adequate removal of atmospheric deposits such as salts or other chemical compounds that may contribute to localized corrosion of stainless-steel components.
- d.
Describe any flaw evaluation methods (such as flaw sizing and flaw analysis methods) and associated flaw acceptance criteria that may be used to determine whether components containing flaws are acceptable for continued service.
Per IAEA SSG-26, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2018 Edition, paragraph 613A.3: For packagings intended for repeated use, the effects of ageing mechanisms on the package should be evaluated during the design phase in the demonstration of compliance with the Transport Regulations. Based on this evaluation, an inspection and maintenance program should be developed. The program should be structured so that the assumptions (e.g., thickness of containment wall, leaktightness, neutron absorber effectiveness) used in the demonstration of compliance of the package are confirmed to be valid through the lifetime of the packaging.
RAIs and Responses to E-64197 Page 7 of 20 This information is needed to determine compliance with requirements in paragraph 613A of the IAEA SSR-6, 2018 Edition.
Response to RAI 2:
The inspection and maintenance program dealing with the effects of aging mechanisms on the metal components of the packaging is described below.
Dimensional inspection of the threads of the containment system closures The go/no-go dimensional checks of the threads are carried out according to the methods described in the standard NF ISO 1502: ISO general-purpose metric screw threads. Gauges and gauging using the corresponding calibrated plug or ring gauges.
Thread diameter Associated ring or plug gauges Ring gauges Plug gauges M12 M12 x 1.75 - 6g M12 x 1.75 - 6H M14 M14 x 2 - 6g M14 x 2 - 6H M20 M20 x 2.5 - 6g M20 x 2.5 - 6H M72 M72 x 2 - 6g M72 x 2 - 6H M230 M230 x 4 - 6g M230 x 4 - 6H The no-go ring and plug gauges must not enter by more than two turns. The go ring and plug gauges must enter freely.
Visual inspections The closures of the containment system, the body, cover, and cage of the TN-BGC 1 are subject to a visual inspection in accordance with standard "NF EN 13018 /2016: Non-destructive testing - Visual testing - General principles with appropriate lighting that does not dazzle the operator. The visual acuity of the inspection personnel is checked annually and meets the requirements of the standard NF EN ISO 9712 / 2012 - Non-destructive testing - Qualification and certification of NDT personnel.
Beforehand, the inspected surfaces are cleaned with appropriate products, excluding halogenated products in particular (chlorine, etc.).
Inspectors search for the following defects:
on the containment closures:
presence of foreign material, scratches, nicks, dents, impact, seizure, deformations, scratches on gasket seating surfaces.
on the cage, body, and cover: scratches, nicks, deformations, indentations, traces of impact, on the body/cage connection screws: scratches, notches, deformations, on the welds of the cage: cracks, crazing.
RAIs and Responses to E-64197 Page 8 of 20 In addition, the packaging cavity is inspected for traces of corrosion.
In the event of a defect, the component affected is repaired or replaced, depending on the extent of the defect. If replacement is not possible, the packaging is taken out of use.
Deformations of the cage tubes are repaired, which can range from weld metal buildup (minimum thickness 2 mm) to changing the complete tube. The defects on the welds are repaired by weld metal buildup followed by penetrant testing (*). Any repair on the upper oblique bars used for handling the packaging is followed by a load test at twice the unladen mass for at least 15 minutes, carried out by an approved inspection body, with a criterion of no permanent set of the bars. This is followed by a penetrant test (*) of the relevant welds, which must be compliant.
The indentations or traces of impact on the body and cover must have a maximum depth of 5 mm and pass a penetrant test (*). If a depression greater than 5 mm is observed on the part of the body containing resin, the packaging must be transferred to the neutronic test bench (see [1]
section 3.3 of Chapter 8 - maintenance instructions) to be declared compliant. If a depression greater than 5 mm is observed on the cover or on the part of the body containing wood, the packaging is taken out of use until an appropriate corrective action is defined.
(*) See description of the penetrant tests below Overload test of handling points (upper oblique bars of the cage)
The handling points are subjected to a static overload test by an approved inspection body for a minimum of 15 minutes at twice the unladen mass of the packaging; the acceptance criterion is the absence of permanent set of the cage bars.
Penetrant testing These tests are:
systematically carried out on the accessible welds in the upper and lower parts of the cage; these welds were stressed in tie-down situations and during handling by slinging or forklift, performed on the body and cover following observation of an impact or indentation.
Prior to these inspections:
the relevant parts are cleaned, rinsed, dried, and more generally cleared of any possible residues (corrosion, dirt, grease, surface defects, etc.) which could block the surface openings or interfere with the examination, a visual inspection as described above is carried out, in order to identify the defects present before penetrant testing, recording the presence of any significant defects.
These tests are carried out according to the requirements of the following standards:
NF EN ISO 3452-1 / 2021 Non-destructive testing - Penetrant testing - Part 1: General principles, NF EN ISO 3452-2 / 2021 - Non-destructive testing - Penetrant testing - Part 2: Testing of penetrant materials,
RAIs and Responses to E-64197 Page 9 of 20 NF EN ISO 17635 / 2017 - Non-destructive testing of welds - General rules for metallic materials, NF EN ISO 3059 / 2013 - Non-destructive testing - Penetrant testing and magnetic particle testing - Viewing conditions Inspectors shall be certified according to the standard NF EN ISO 9712 / 2012 - Non-destructive testing - Qualification and certification of NDT personnel in the penetrant testing method, level 2 (COFREND, CIFM sector), with a currently valid card. They perform the test and interpret the results.
The acceptance criteria for defects identified following a penetrant test shall comply with the requirements of Annex I1.A2 of CODAP 2005 for construction category A. They are summarised in the following table.
Table I1.A2.2 - Unacceptable indications for each construction category Construction categories A
B1 B2 Linear indications (Note 1)
Rounded or non-linear indications, the largest dimension of which is greater than 4 mm.
Three or more aligned indications, spaced less than 3 mm apart edge to edge.
Clusters of 5 or more indications, over a 100 cm² rectangular surface area that is chosen to be the least favourable in terms of the indications, but whose largest single dimension is no greater than 200 mm.
Clusters of 8 or more indications, over a 100 cm² rectangular surface area that is chosen to be the least favourable in terms of the indications, but whose largest single dimension is no greater than 200 mm.
Clusters of 10 or more indications, over a 100 cm² rectangular surface area that is chosen to be the least favourable in terms of the indications, but whose largest single dimension is no greater than 200 mm.
Note 1: An indication is said to be linear when its largest dimension is more than three times greater than its smallest dimension. Other indications are said to be rounded or non-linear.
Note: Only indications whose largest dimension is 2 mm or greater are taken into consideration.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 10 of 20
RAI 3
Provide the sources and/or references for the new thermal properties provided in table 3 of chapter 4 of the SAR for steel 39 CD 4, balsa, and poplar. Additionally, the staff requests reference 3 of chapter 12 of the SAR for the modified values of thermal conductivity and specific heat for resin. This information is needed for the staff to verify the thermal performance of the package during normal conditions of transport and hypothetical accident conditions.
This information is needed to determine compliance with requirements in paragraph 658 of the IAEA SSR-6, 2018 Edition.
Response to RAI 3:
The description of the packaging in Safety Analysis Report (SAR) [1] Chapter 4 has been supplemented to specify the thermal characteristics of 39 CD 4 steel, balsa and poplar wood at room temperature, which were not previously specified. They appeared only in the calculation notes attached to Chapter 12 - Thermal analysis [1]. These characteristics are taken from the literature available at ambient temperature.
For steel at temperature, they are derived from the RCC-M code (Design and Construction Rules for Mechanical Components of PWR Nuclear Islands).
For wood, the thermal conductivity is corrected for temperature according to the principles laid down in the Wood Handbook, Chapter 3: Physical properties and moisture relations of wood, which specifies that the thermal conductivity increases by 2% to 3% when the temperature of the wood increases by 10 °C. Conductivity is thus increased by 70% at 270 °C.
However, the CEA reiterates that these characteristics have not been modified in the calculation notes, where they have been in use for more than 20 years.
Regarding the resin, the thermal characteristics applied in the calculation notes in Chapter 12 have not been modified. The thermal characteristics specified in Chapter 4 have, however, been modified to bring them into line with the values used in the calculation notes, where, once again, they have remained unchanged for more than 20 years. For clarity, Chapter 12 now specifies the reference of reports from the tests performed in 1988, which made it possible to define these characteristics. These reports were attached to the safety analysis report up until 2003. It is not considered appropriate to transmit them again. However, they remain available to the American authority, in French, if necessary.
References:
- 1. Safety Analysis Report CEA/DES/DDSD/DTEL/SGPE/DSEM 07600 Revision C, January 17, 2024.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 11 of 20
RAI 4
Provide an evaluation of the compatibility of the new UO2F2 chemical form of contents no. 11 and the packaging components. This information is needed to ensure there is no significant corrosion or adverse reactions that could significantly reduce the effectiveness of the packaging.
This information is needed to determine compliance with requirements in paragraph 614 of the IAEA SSR-6, 2018 Edition.
Response to RAI 4:
UO2F2 is stable up to 300 °C. Beyond this, hydrofluoric acid is generated, which is a highly corrosive gas. The maximum temperature reached in the cavity of the package model is 136 °C under accident conditions of transport. At this temperature, there is no risk of degradation of the UO2F2 likely to damage the components of the package model.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 12 of 20 Thermal Evaluation:
RAI 1
Clarify that the chapter 16 section 4.3.2.2 thermolysis analysis considered relevant (or bounding) thermolysis rates and inputs.
Section 4.3.2.2 of chapter 16 relied on input from thermolysis rate test data of a polyethylene waste matrix. Test data were in units of moles/(year x drum) based on an average waste matrix mass of 51.4 kg, from which the application converted thermolysis rates to units of moles/(year x kg) and moles/(second x kg).
The analysis did not clearly derive the basis and conversion of thermolysis rates from original test data units [moles/(year x drum)] to the applications thermolysis rate units [moles/(second x kg)] at 343°K and accident condition transport temperature of 409°K (i.e., doubling of reaction rate for every 10°C rise in temperature) and, therefore, a review of the analysis could not be performed.
This information is needed to determine compliance with IAEA SSR-6 paragraph 618.
Response to RAI 1:
The data from the note Gas generation from transuranic waste degradation: data summary and interpretation, which is used in Chapter 16 of [1], specifies that the release of thermolysis gas Qgas is 1.9 mol/year for 51.4 kg of polyethylene at a temperature of 343 °K. [1] and authorizes 1 kg of polyethylene, i.e. Qgas = 3.7x10- 2 mol/year equivalent to 1.172x10-9 mol/s.
Safety Analysis Report (SAR) [1] conservatively postulates that thermolysis occurs for 1 day at a temperature of 409 K and then for 6 days at a temperature of 348 K.
Correcting Qgas for these temperatures according to Murphy's law, which is expressed as follows:
Q(T2) =
)
.Q(T 10 1
)
.T T
T T
(7400.
2 1
1 2
where:
Qgas = 3.552x10-6 mol/s at 409 K, Qgas = 2.393x10-9 mol/s at 348 K.
Thus, n = 0.307 moles of gas formed during the first day and then 1.242x10-3 moles during the following 6 days, i.e. a total of 0.308 moles of gas during the seven days of ACT.
This does not call into question the analysis carried out in Chapter 16 of [1], which concludes that the risk of explosion is excluded.
SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 13 of 20 Containment Evaluation:
RAI 1
Clarify the details of inerting the packaging boxes and internal fittings when pyrophoric content (e.g., uranium metal powders) is transported in the TN-BGC 1 package.
Chapter 3 section 2.5.1 and chapter 16 section 5 indicated that transport of pyrophoric content requires inerting the contents primary packaging boxes, internal fittings, and the packaging cavity. The operating procedures for closing the package (section 3.3 of chapter 7) provided instruction for inerting the package cavity; however, a similar instruction is not provided in section 3.1 and section 3.2 when loading the contents into the primary content box and internal fitting.
In addition, regarding chapter 7 section 3.3, clarify whether the package cavity inerting pressure (absolute pressure) after the inerting procedure is 2,104 Pa (i.e., approximately 2 kPa (absolute)).
This information is needed to determine compliance with IAEA SSR-6 paragraph 614.
Response to RAI 1:
Inerting of the primary packaging boxes and internal fittings is not described in Safety Analysis Report (SAR) [1] because it is the responsibility of the shippers, who must perform it according to their own procedures. In all likelihood, the inerting of the primary packaging boxes can only be done by packaging the material in an inert atmosphere, such as a glove box.
Inerting of the internal fittings is most often carried out in the same way as for the packaging, namely several successive cycles of depressurization followed by injection of neutral gas, allowing the container to be completely inerted. This method requires the use of primary packaging boxes capable of withstanding the very low pressure created in the cavity of the internal fittings before the injection of neutral gas.
There is also a method consisting of injecting argon into the bottom of the cavity using a pipe placed at the bottom of the internal fittings. As argon is heavier than air, it expels the air from the cavity of the internal fittings.
The internal fittings can also be loaded and closed directly under an inert atmosphere, as in the case of the packaging boxes.
The leaktightness of the internal fittings must, in all cases, be checked beforehand and be under 1.5 x 10-5 Pa.m3.s-1.
There is no depressurization of the primary packaging boxes or internal fittings after inerting.
The pressure in the packaging cavity after the inerting procedure is 2x104 Pa absolute.
References:
- 1. Safety Analysis Report CEA/DES/DDSD/DTEL/SGPE/DSEM 07600 Revision C, January 17, 2024.
RAIs and Responses to E-64197 Page 14 of 20 SAR Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 15 of 20
RAI 2
Clarify the timing and extent of certain packaging leak tests, including the fabrication leakage test and the test with a 5.5.10 -5 Pa m3/sec (SLR) acceptance criterion.
Section 3.1 of chapter 8 indicated that the closure system is leak tested during minor maintenance to an acceptance criterion of 1E-6 Pa m3/sec SLR and section 3.2 indicated that the containment system (including the containment boundarys stainless steel internal shell, bottom of the body, main plug, and quick-connect coupling cap as well as the O-ring gaskets associated with the plug and quick-connect coupling orifice cap) is leak tested during major maintenance (i.e., therefore, presumably the initial test at fabrication) to an acceptance criterion of 1. 10-7 Pa.m3/sec SHeLR (standard helium conditions). However, it appears there was no clear indication that a fabrication leakage rate test of the entire containment boundary with an acceptance criterion of 10-7 Pa.m3/sec (SHeLR) would be performed.
In addition, section 4.4 of chapter 13 indicated that the package transporting Content 11 and Content 26 is leak tested prior to transport to an acceptance criterion of 5.10-4 Pa m3/sec (section 4.2.1 indicated this test would take place at SLR conditions). However, although section 3.1.5 of chapter 8 indicated that Type B(U) packaging is leak tested to an acceptance criterion of 5.10-5 Pa.m3/sec (presumably SLR), there was no discussion as to the timing and extent of the leak test (i.e., the containment boundary with closure).
This information is needed to determine compliance with IAEA SSR-6 paragraph 809.
Response to RAI 2:
A helium leak test of the containment system was carried out at the end of manufacturing on each packaging. The leak rate was deemed satisfactory if it remained below 10-7 Pa.m3.s-1 SHeLR.
This test, described in Safety Analysis Report (SAR) Chapter 8, Section 3.2.3, is repeated every 6 years or every 40 transports during major maintenance operations.
During minor maintenance operations carried out every 3 years or every 15 transports, a leak test of the closure system is conducted, as described in Section 3.1.5, Chapter 8 of [1]. For packagings used under B(U)F approval, the leakage rate is considered satisfactory if it does not exceed 5x10- 5 Pa.m3.s-1 SLR.
As prescribed in Section 3 of Annex 0 of the certificate of approval and reiterated in Section 4.4, Chapter 13 of [1], a leak test of the closure system is performed before each transport and the resulting leakage rate must be less than 5 x 10-4 Pa.m3.s-1 SLR. It is this leakage rate, guaranteed under all conditions of transport, that is used in Chapter 13 of [1] to assess the release of the package model and ensure that it meets the criteria imposed by regulations under both normal and accident conditions of transport.
Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 16 of 20 Criticality Evaluation:
RAI 1
Provide additional justification that the modeled configurations of concentric spheres in the chapter 15 of the SAR (ref. CER/DES/DDSD/DTEL/SGPE/DSEM07615 Rev. B) provides the most reactive configurations in section 4.5.3 for content 11, as well as in section 4.13.2 for content 52 for determining the maximum permissible masses of uranium at the indicated enrichments.
In Technical Note 00048550.01A, Criticality Safety Analysis in Air Transport Configuration of the TN-BGC-1 Package Model Loaded with Metallic Uranium in Air Transport, the applicant used a series of concentric spheres to model the fissile material under air transport conditions.
The central spheres contain the fissile material divided between them, with one or more spheres containing a moderator (i.e., either variable water or a water and carbon mixture to simulate polymer) and the other sphere containing no moderator. Up to three additional concentric spheres in varying configurations of reflectors (i.e., water, steel or carbon) were evaluated by the applicant, leading to a series of calculations with up to six total concentric spheres.
Contrary to the six concentric spheres evaluated in the Technical Note, chapter 15, section 4.5.3 and section 4.13.2 cites that only five concentric spheres were evaluated. Clarification is needed to ensure that the modeled configurations are bounding and conservative for the packaging limits specified in table 11 and table 25 of the SAR.
This information is needed to determine compliance with IAEA SSR-6, Rev. 1, paragraph 676.
Response to RAI 1:
According to International Atomic Energy Agency (IAEA) Regulations for the Safe Transport of Radioactive Material [1], packages transported by air must be subcritical under the conditions resulting from the tests (mechanical and thermal) required for this type of transport and assuming reflection by at least 20 cm of water, but without penetration of water (§ 683.a). In the event that the accident conditions are not known (absence of mechanical and thermal test results), the evaluation of the previous configuration must be carried out by considering the most conservative values for the unknown parameters (§ 676). The interpretation of these regulatory requirements by the CEA was presented by IRSN at the Patram 2001 conference in Chicago, and was the subject of article SEC/T/02.153 Comments on the requirements for packages containing fissile materials. This article highlights the need to study at least three different configurations, in the absence of test results. These configurations are:
dry fissile material alone, surrounded by 20 cm of water, dry fissile material surrounded by the reflective materials of the packaging and by 20 cm of
- water, fissile material mixed with the moderating materials of the packaging and surrounded by 20 cm of water; neutron-absorbing materials may be taken into account.
The article also specifies that these configurations are not exhaustive and depend on the packaging and its contents, but that they can be used as a guide to meet the conditions of [1]
with regard to air transport.
RAIs and Responses to E-64197 Page 17 of 20 Accordingly, the CEA has chosen a fourth configuration that bounds those mentioned above, which consists in studying dry fissile material and/or fissile material mixed with the moderating materials of the contents and packaging, with neutron reflection provided by the possible reflective material of the contents surrounded by a layer of steel, corresponding to the steel of the packaging, and by 20 cm of water.
Subcriticality of an isolated package is understood to mean an effective multiplication coefficient (keff) no greater than 0.95 all uncertainties included (uncertainties related to the calculation methods used and their qualification). Given the conservative nature of the geometrical configurations considered, this acceptability criterion may be adapted in the case of air transport.
Technical Note 00048550.01A makes it possible to determine the maximum permissible masses in air transport for content Nos. 11 and 52.
Cases No. 1A to 1I are dedicated to content No. 11, modelled in the form of metallic uranium with a U-235 isotopic content equal to 100%. The following are specifically studied:
the absence of polymeric materials (case No. 1A),
the presence of 500 g of polymeric materials (cases No. 1B to 11E),
the presence of 1000 g of polymeric materials (cases No. 1F to 1I).
This polymeric material corresponds to the materials authorized for the packaging of the fissile material. Conservatively, it is modelled in the calculations as water.
In accordance with the justifications provided above, the computational model is composed of several concentric spheres comprising, from inside to outside:
a sphere of varying radius comprising a dry U fraction, a sphere of varying radius comprising a U fraction moderated by 500 g of water (cases No. 1B to 1E only) or 1000 g of water (cases No. 1F to 1I only) (*),
a concentric sphere of varying radius consisting of the remaining fraction of U moderated by 2 kg of water and 2.5 kg of carbon (corresponding to the wood of the packaging),
a concentric sphere consisting of 340 kg of steel (representing the packaging steel),
a concentric sphere 20 cm thick made of water ensuring the neutronic isolation of the package.
(*) this sphere is also studied positioned at the center, surrounded by the sphere containing a dry U fraction.
Four concentric spheres are therefore modelled in the absence of polymeric materials (case 1A) and five concentric spheres are modelled in the presence of polymeric materials (cases No. 1B and 1I).
RAIs and Responses to E-64197 Page 18 of 20 For content No. 52, which is the subject of cases 1K to 1S, the same methodology is retained, conservatively considering that the uranium is in metallic form with a U-235 isotopic content equal to 100%. However, to account for the presence of any amount of carbon in the cavity of the TN90 internal fittings, a reflective concentric sphere consisting of carbon is added. Its mass is calculated to fill the free volume of TN90 internal fittings.
On this basis, the computational model is composed of several concentric spheres comprising, from inside to outside:
a sphere of varying radius comprising a dry U fraction, a sphere of varying radius comprising a U fraction moderated by 500 g of water (cases No. 1L to 1O only) or 1000 g of water (cases No. 1P to 1S only) (*),
a concentric sphere of varying radius consisting of the remaining fraction of U moderated by 2 kg of water and 2.5 kg of carbon, a concentric sphere consisting of carbon (whose mass is calculated so as to fill the free volume of TN90 type internal fittings),
a concentric sphere consisting of 340 kg of steel (representing the packaging steel),
a concentric sphere 20 cm thick made of water ensuring the neutronic isolation of the package.
(*) this sphere is also studied positioned at the center, surrounded by the sphere containing a dry U fraction.
Five concentric spheres are therefore modeled in the absence of polymeric materials (case 1K) and six concentric spheres are modelled in the presence of polymeric materials (cases No. 1L and 1S).
Note: Cases No. 1A and 1J were studied, respectively for contents No. 11 and 52, in order to demonstrate that the steel model representing the packaging steel was conservative.
Reference:
- 1. International Atomic Energy Agency (IAEA), Regulations for the Safe Transport of Radioactive Material, Safety Standards Series No. SSR-6, 2018 Edition.
Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 19 of 20
RAI 2
Provide justification that the acceptance criterion used for determining the allowable keff limits are appropriate for TRIGA fuel and Uranium Metal. In Technical Note 00064690.01C, Criticality Study of the TNBGC 1 Package Model Loaded with TRIGA Fuel or Uranium Metal, section 4.2, Acceptance Criteria and Qualification, the applicant states that the qualification status of the V2 CRYSTAL criticality package has a lack of experience implementing fissile media in the form of UZrHx, and takes a keff penalty of 2000 pcm for these materials (i.e., TRIGA fuel).
Likewise, the applicant states that there is a similar lack of experience in evaluating moderately enriched uranium reflected by steel, CH2, or boron, and highly enriched uranium reflected by steel or aluminum, and imposes the same 2000 pcm penalty on keff.
Based on the calculated values for these types of materials (e.g., F3 and F4 TRIGA fuels and uranium metals) the maximum keff s encroach on the penalized limits proposed by the applicant.
Based on staff review of the Technical Note, it is not apparent how this 2000 pcm keff penalty was determined. Additional justification in necessary to ensure that this penalty is sufficient to account for the V2 CRYSTAL code uncertainties with these types of fissile materials.
This information is needed to determine compliance with IAEA SSR-6, Rev. 1, paragraph 676.
Response to RAI 2:
The technical note NT 00064690.01C studies two specific reference fissile media:
- UZrHx, uranium metal.
For each of these media, there is a lack of qualification of the CRISTAL V2 criticality package.
Thus, a fixed and conservative bias of 2,000 pcm is applied for each of the two media. This value of 2,000 pcm corresponds to the maximum calculation - experiment deviation observed in the qualification file of the CRISTAL V2 criticality package, which is based on several thousand experiments, about 80% of which are from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) manual.
These biases are not cumulative.
Impact:
No change as a result of this RAI.
RAIs and Responses to E-64197 Page 20 of 20
RAI 3
Provide justification that the acceptance criterion used for determining the allowable keff limits are appropriate for plutonium in non-powder metallic form.
Similar to the above RAI, in Technical Note 00046711.01, Criticality Study Package Model Loaded with 3.5kg of 239Pu Under Metallic Form in ACT, the applicant states that the qualification status of the V1.2 CRYSTAL criticality package has a lack of experience implementing fissile media in the form of non-powder metallic form and takes a keff penalty of 2000 pcm for these materials.
Based on staff review of the Technical Note, it is not apparent how this 2000 pcm keff penalty was determined. Additional justification in necessary to ensure that this penalty is sufficient to account for the code uncertainties with plutonium in non-powder metallic form.
This information is needed to determine compliance with IAEA SSR-6, Rev. 1, paragraph 676.
Response to RAI 3:
The fixed and conservative bias of 2,000 pcm corresponds to the maximum calculation -
experiment deviation observed in the qualification file of the CRISTAL V1.2 criticality package.
SAR Impact:
No change as a result of this RAI.