ML24192A184
| ML24192A184 | |
| Person / Time | |
|---|---|
| Site: | 07109383 |
| Issue date: | 06/28/2024 |
| From: | Bussmann D, Kattner F GNS Gesellschaft fur Nuklear-Service mbH |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards |
| References | |
| T1213-CO-00021, EPID L-2021-NEW-0002 | |
| Download: ML24192A184 (1) | |
Text
@ GNS
GNS Ge sellschaft fur Nu klear-Service mbH
- Postfa c h 10 12 53. 45012 Essen
Document Control Desk Ou r reference: T121 3-CO-00021 Director, Division of Fuel Management Co n tact perso n : Dom in ik B u ssmann Office of Nuclear Material Safety and Safeguards Phone : +49 (0 )201 / 109-1891 U.S. Nucl ear Regulatory Commission Washington, DC 20555-0001 Fax : +49 (0)20 1/ 109-1186 USA Email: do m inik.b uss mann @gn s.de
Date: 06/28/ 2024
Subject:
10 CFR 71 Application of the CASTOR geo69 Spent Nuclear Fuel Transportation Package (Docket No. 71-9383 ; EPID No. L-2021-NEW-0 002)
Response on Second Request for additional informat ion and submission of related revised documents
By letter dated February 22, 2024 NRC submitted a second reques t for additional information on the application for approval acc. to 10 CFR 71 of the CASTOR geo69 spent fuel package and the GNS Quality Assurance Program Description.
The response to this request and the accordingly revised parts of the SAR as well as the revised QAPD are e nclosed to this letter. A list of the revised sections ca n be found in section 0.2 of the SAR.
This response and the revised SAR contain ce rtain information that is proprietary, confidential and a trade secret to GNS. Therefore, a non-proprietary version of the response and the SAR together with an affidavit prepared pursuant to 10 CFR 2.390 providing the basis for withholding of the GNS proprietary inform ation from public disclosure is enclosed to this letter.
Should the NRC staff require additional information to support review of this response, please do not hesitate to contact Mr. Dominik Bussmann at +49 201 109 1891, or by email at dominik.bussmann@gns.de.
Sincerely, GNS Gesellschaft fur Nuklear-Service mbH D ig ital sign iert va n Digital signiert van Kattner, Fran z Bu ssman n,
@ i.V. Da tum : 2024.0 6. 28 @i.A. Do m in ik 09 :40 :40 +02'00' Datum : 20 24.06.28 GNS GNS 08 :28 :45+02'00'
- Affidavit pursuant to 10 CFR 2.390 : Response on Request for additional information (Proprietary Version) : Response on Request for additional information ( Non -Proprietary Vers ion )
En c losure 4: T1213-SR-00003, Rev. 4, Safety Analyses Report ( Proprietary Version) : T1 213-SR-00003, Rev. 4, Safety Analyses Report (Non-Proprietary Version) : GNS-QAPD-001, Rev. 2, Quality Assurance Program Description
GNS Gesellschaft fiir Nuklear - Service mbH Froh nh a us er Stra ~ 67 Comm erzban k AG, Essen USH dNr. DE 171892 160, Steue r-Nr. 111/5714/1234 DE-45127 Essen !BAN OE17 3604 0039 0124 3237 00, SIC COBADEFF Cha irman of the Supervisor, Board : Dr. Guido Knott Telephone +49 201 109-0 Bayer ische Landesban k Dusseldorf Ma n ag ing Directors:
Telefa x +49 201109-1100 !BAN DE56 7005 0000 0004 3662 43, SIC BYL AOEMM Da n iel Oehr {Ch a irman}
www.gns.de Deutsch e B a n k AG, Essen Df.-lng. Jen s Schro der HRS Essen 11213 !BAN DE28 3607 0050 0151 7804 00, BIC DEUTDEDEXXX
~
fl
] Enclosure 1 to T1213-C0-00021
PURSUANT TO 10 CFR 2.390
I, Dr.-lng. RAINER NORING, depose and say that I am Divisional Director of Cask Projects of GNS Gesellschaft fur Nuklear-Service mbH (a company duly organized under the German Law, having its seat at Frohnhauser Strasse 67, 45127 Essen, Germany), duly authorized to execute this affidavit.
I, Dr.-lng. SASCHA KLAPPERT, depose and say that I am the Divisional Director of Engineering of GNS Gesellschaft fur Nuklear-Service mbH (a company duly organized under the German Law, having its seat at Frohnhauser Strasse 67, 45127 Essen, Germany), duly authorized to execute this affidavit.
We, Dr.-lng RAINER NORI NG and Dr.-lng. SASCHA KLAPPERT, have reviewed or caused to have reviewed the information which is identified as confidential and referenced in the paragraph below. We are submitting this affidavit in conformance with the provisions of 10 CFR 2.390 of the Commission's regulations for withholding this information.
The information sought to be withheld from public disclosure is contained in Enclosure 2 and 4 of letter T1213-C0-00021 and are listed below:
Enclosure 2, Response on Request for additional information, Proprietary Version Enclosure 4, Safety Analyses Report (SAR), Type B(U)F Transport Package CASTOR geo69, GNS Report number 1014-SR-00001, Revision 4, Docket 71-9383, Proprietary Version
This documents has been appropriately designated as proprietary.
We have personal knowledge of the criteria and procedures utilized by GNS Gesellschaft fur Nuklear Service mbH in designating information as a proprietary trade secret, privileged or as confidential commercial or financial information.
Pursuant to the provisions of paragraph (b) (4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure, included in the above referenced document, should be withheld.
(1) The information sought to be withheld from public disclosure involves certain design details associated with the SAR analyses and SAR drawings of the CASTOR geo69 package design, which are owned and have been held in confidence by GNS Gesellschaft fur Nuklear-Service mbH.
(2) The information is of a type customarily held in confidence by GNS Gesellschaft fur N_uklear Service mbH and not customarily disclosed to the public. GNS Gesellschaft fur Nuklear-Service mbH has a rational basis for determining the types of information customarily held in confidence by it.
(3) The information is being transferred to the Commission in confidence under the provisions of 10 CFR 2.390 with the understanding that it is to be received in confidence by the Commission.
(4) The information, to the best of our knowledge and belief, is not available in public sources, and any disclosure to third parties has been made pursuant to regulatory provisions or confidentiality agreements which provide for maintenance of the information in confidence.
(5) Public disclosure of the information is likely to cause substantial harm to the competitive position of GNS Gesellschaft fur Nuklear-Service mbH because :
(a) A similar product is manufactured and sold by competitors of GNS Gesellschaft fur Nuklear Service mbH.
(b) Development of this information by GNS Gesellschaft fur Nuklear-Service mbH required expenditure of considerable resources. To the best of our knowledge and belief, a competitor would have to undergo similar expense in generating equivalent information.
(c ) In order to acquire such information, a competitor would also require considerable time and inconvenience related to the development of a design and analysis of a cask for the dry storage and transport of spent nuclear fuel.
(d) The information requ ired significant effort and expense to obtain the licensing approva ls necessary for app lication of the information. Avo idance of this expense would decrease a competitor' s cost in applying the information and marketing of the product to which the info rmation is appl icable.
(e) The information consists of design features, analyses methods and calculation results related to the design and analyses of a cask for the dry storage and transport of spent fuel, the application of wh ich provide a competitive economic advantage. The availability of such information to co mpet itors would enable them to modify their product to unfa irly get a better competitive position with GNS Gese llschaft fur Nuklear - Service mbH, take marketing or other actions to improve their product's position or impair the position of GNS Gesellschaft fur Nuklear-Serv ice mbH ' s product,
while avoiding the expense of developing similar data and analyses in support of their processes,
methods or apparatus.
(f) In pricing GNS Gesellschaft fur Nuklear-Service mbH ' s products and services, significant research, development, engineering, analytical, licensing, quality assurance and other costs and expenses must be included. The ability of GNS Gesellschaft fur Nuklear-Service mbH 's competitors to utilize such information without similar expend iture of resources may enable them to sell at prices reflecting significantly lower costs.
Date : Date :
.............. 03.-......
Dr.-lng. RAINE R N
- Divisional Directo
GNS Gese llsc h aft.. GNS G ese llschaft fur Nuk lear -S ervic e m b N u klear-Service mbH u~ 1/2kv" ll ~ 1
Subscr ibed and sworn to me before this..Ac.?.. ~~ day of... J.l.-i.~ 2024. Q)\\ )Jok,rid's
...................... W-................... n_,,f,k d~ Jo] 4
Notary Public/2; nterschrift des Notars)
Nummer GO-153 des Urkundenverzeichnisses fur 2024 Number GO-153 of Notary's List of Deeds for 2024
Die vorstehenden, heute vor mir ge-I herewith certify that the foregoing fertigten Namensunterschriften von signatures today is given by
Dr.-lng. Rainer Noring, geb. am Dr.-lng. Rainer Noring, born on 27.11.1968, von Person bekannt, 27.11.1968, of known identity,
und and
Dr.-lng. Sascha Klappert, geb. am Dr.-lng. Sascha Klappert, born on 21.08.1972, von Person bekannt, 21.08.1972, of known identity,
beide dienstansassig Frohnhauser both with their business address at Str. 67, 45127 Essen, Frohnhauser Str. 67, 45127 Essen.
beglaubige ich hiermit.
Die Unterschriften wurden heute im The signatures were given today in Hause Frohnhauser Str. 67, 45127 Frohnhauser Str. 67, 45127 Essen.
Essen, geleistet.
Obligatory Statement according to Pflichtvermerk nach deutschem German law :
Recht:
According to Sec. 3 Para. 1 S. 1 No. 7 Die Unterzeichner verneinten eine German Notarisation Law the Notary Vorbefassung im Sinne des § 3 Public asked the appearers whether Abs. 1 S. 1 Nr. 7 Beurkundungsge he or any member of his firm had setz, nachdem ihnen der Notar den acted in the matter which is the sub lnhalt dieser Vorschrift erlautert hat. ject of this instrument, except in a no tarial capacity. The appearers replied Essen, den 10. Juni 2024 in the negative.
Essen, June 10, 2024
Cfo t~ers ~ l
Notary Public APOSTILLE
(Convention de La Haye du 5 octobre 1961)
- 1. Country/ Land:
Federal Republic of Germany / Bundesrepublik Deutschland
This public document I Diese offentliche Urkunde
- 2. has been signed by/ ist unterschrieben von Notary Public Dr. Joachim Gores I Nota r Dr. Joachim Gores
- 3. acting in the capacity of/ in seiner Eigenschaft als Notary Public in Essen / Notar in Essen
- 4. bears the sea l I sie ist versehen mit dem Siegel of the Notary Public Dr. Joach im Gores in Essen / des Notars Dr. Joachim Gores in Essen
Ce rt if i e d /Best at i gt
5. at/ in E s s e n 6. the I am t 4. Juni 2024
- 7. by/ durch the President of the Regional Court Essen /
die Prasidentin des Landger ichts Essen
8. No./unterNr. 91 E 1 /Sa,l h )t~3/cl...OJ.l(
- 9. Seal / Stamp Siegel / Stempel: 10. Signature/ Unterschrift:
\\n des lnzng 16..................... ~........
Kretschme r
- -,,-,,1"';.:,.,.... ;-,f,. 1r1+
28/06/2024 to letterT1213-CO-00021 @ GNS
- 1. Response on 2. Request for additional information dated February 22 2024 Model No. CASTOR geo69 Docket-No.: 71-9383, EPID No. L-2021-NEW - 0006 Non -Proprietary Vers ion
General
Additional There is not a clear consistency in the pressure values provided in different Question SAR tables.
by Mail dated 1. For example, SAR table 8. 1-2 indicated that the internal canister absolute 03.22.2024 test pressure is greater. However, it appears SAR table 2. 7-30 listed the canister test pressure as much less than
- 2. In addition, SAR table 2. 7-30 indicated an internal cask pressures of either slightly less than or slightly greater than was applied for HAG structural calculations (e.g., impact, fire); however, these pressures are less than the potential internal cask pressure (much greater than ll reported in SAR section 4.3.1.
Answer :
First of all, it should be noted, that the pressure data determined in Chapter 4 are absolute values, whereas in the structural evaluations deal with ga uge pressures.
- 1. The test pressures for cask and canister required by the design and evaluated within the mechanical evaluations differ depending on whether the transport or storage SAR is being considered. This is because of the d iffering cask pressure ( during transport and dur ing storage, see RAI 1-1). Table 8. 1-3 has been adjusted to the requ iremen ts and boundary condit ions of the transport SAR. A Note has been added to Section 8. 1.4. 1 dealing with the change of the test pressures if the CASTOR geo69 is also used as a storage cask. The resulting higher test pressure assessment is part of the Storage SAR. During manufacturing pressure tests will be performed only once but with the covering test pressure.
2. The comment is justified, there are inconsistencies regarding mentioned pressures in SAR Rev. 3. These were eliminated in Rev. 4, among others, under consideration of the pressure calculations in Chapter 4. It should be noted, that all cask pressure calcu lations regard ing a failure of the canister containment, which would lead to a significant pressure rise in the cask
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are omitted in Rev. 4, since a failure of the canister containment is excluded as shown in Chapter 2.
RAI 1-1: Clarify the reason for differences in cask helium backfill pressures that are mentioned in the Part 71 SAR and Part 72 SAR operations and confirm that the package's analyses (e.g., pressure calculations, structural calculations) are bounding.
The calculations (e.g., Chapter 4 pressure calculations) within the Part 71 SAR were based on the pressure listed in operation procedure SAR table 7. 1-2, which indicated a cask helium backfill pressure of (absolute). This pressure is less than the Part 72 SAR operation procedure SAR table 9.3-1, which indicated a cask helium backfill pressure of (absolute). A discussion that reconciled the different cask helium backfill pressures for transportation and storage was not found and, therefore, it is uncertain if an intermediate operation procedure is required for transportation and it is uncertain if the package's analyses are bounding.
This information is needed to determine compliance with 10 CFR 71. 35 and 10 CFR 71.41.
Answer:
The differences in cask helium backfill pressures mentioned in the Part 71 SAR and Part 72 SAR operations chapters are correct. The reason is as follows:
During transport operations cask and canister are both provided with an underpressure with regard to the ambient atmosphere of (absolute). In the unlikely event of a minor leak, ambient atmosphere would leak into e.g. the cask, but initially no cask atmosphere would leak out. This pressure scenario is considered consistently for the safety evaluations in the Part 71 SAR.
Prior to start of the storage of the CASTOR geo69 the pressure inside the cask must be adjusted to (absolute). This is essential for the proper functionality of the pressure switch, that is installed in the cask lid instead of the blind flange. To be precise, when storing the cask after transportation, the cask interior must be ventilated via the SVK under the protection cap, the blind flange must be replaced with the pressure switch and the cask must then be pressurized with helium. For sure, LT shall be applied to the metal gaskets of pressure switch and protection cap, afterwards. Finally, the pressure switch can be connected to the monitoring system of the storage facility. This pressure scenario is considered consistently for the safety evaluations in the Part 72 SAR.
If the cask shall be transported after storage, this must be reversed.
Section 7.4.1 of the SAR Rev. 4 has been expanded to include a corresponding explanation.
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A. Structural Evaluation (St)
RAI 2A-1: Provide the package type for CASTOR geo69 based on the payload for which the package is designed to store.
Packages licensed in the United States (U.S.) have a package type assigned to the package design. The designer uses this package type categorization for component design that support the package safety functions. The categorization in addition, informs the designer of the amount of radioactive material to contain, control and shield against, in maintaining the required level of package safety.
This information is required to demonstrate compliance with 10 CFR 71.33 (a)(1)
Answer:
Section 1.0.2 of the SAR identifies the CASTOR geo69 as a Type B(U)-F package.
RAI 2A-2: The CASTOR geo69 transportation package SAR identifies the cask as Type B(U)-F using the requirements in 10 CFR 71.4. Explain how the cask type was translated to the design codes in Table 1.0-1 and 1.0-2.
The staff uses the guidance in NUREG/CR 6407 and NUREGICR 3854 to ensure that the spent fuel transportation cask is designed to the level of safety as achievable using appropriate ASME codes for its design, fabrication, and inspection. The selected codes provide criteria for fabrication processes which are related to materials control, forming, heat treatment, examination, and acceptance testing. Implementation of these criteria will ensure the structural integrity of shipping containers at levels consistent with the radioactive materials being transported.
This information is required to meet the requirements of 10 CFR 71.35(a), 10 CFR 71.37(a) and 10 CFR 71.107(b)
Answer:
The way in which SSCs (and also services) are classified as ITS (Class A, B or C) or not-ITS (D or " ") items is regulated at GNS as a so-called graded approach via the QM system (and orients to NUREG/CR 6407). This classification is typically made in the manufacturing parts list. However, the item classification is also relevant for the design of the transport and storage cask. Thus, within Rev. 4 of the SAR, GNS has revised the design parts lists in the Appendixes 1-6 to 1-10 of Chapter 1 to not only classify, whether an item is ITS or not, but also to grade ITS items from A (failure of the item could directly result in a safety case), B (failure or malfunction of the item could indirectly result in a safety case) to C (failure or malfunction of the item is unlikely to create a safety case) or D if not-ITS, but quality relevant. Cf. penultimate
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column of the respective design parts list which is headlined by Cl. standing for "classification level".
As part of the relevance for the safety report, a corresponding declaration was made in section 1.2.
RAI 2A-3: Use the guidance in NUREGICR 6007 for the computation of bolt stresses or provide a comparison showing the equivalence of stress computations using VD/ 2230: 2014-12 and VD/ 2230: 2015-11 with bolt stress computed using NUREGICR 6007.
The staff reviews the stress computation of highly stressed bolts using the guidance in NUREGICR 6007 hence the bolt stress shown in the SAR will be evaluated using the methodology of the guidance document.
This information is required to meet the requirements of 10 CFR 71. 51 (a)(1) and (2).
Answer:
In the following it is shown, that the stresses of bolts are computed in equivalence to NUREG/CR-6007.
How stresses in the bolts are calculated is depicted in [APPENDIX 2-1 to 1014-SR-0001] section 3.1.2 "Evaluation of the stress states of the lid bolting". The following is cited in this section: "The evaluation of the stress states of the lid bolting is done according to section 2.2 in accordance with [4]." (reference [4]
is NUREG/CR-6007).
In the following the procedure is depicted in detail.
The occurring forces and moments in the bolting are derived from the performed FE-calculations by integration of the node stresses over a correspondent cutting plane for a reference point on the bolting axis in the free loaded shank and thread of the bolting (i. e. in the clamp length of the bolting).
The torsional stress from assembly conditions is taken into account in the assessments.
In the FE-calculations the geometry is explicit modelled, while in NUREG/CR-6007 the formulas for calculating forces and moments are based on simplified geometries.
The loads and combination of loads to take into account according to NUREG/CR-6007 are all considered in the SAR (see section 4.0 in NUREG/CR-6007: Bolt Forces/Moments generated by Preload, Gasket Loads, Pressure Loads, Temperature Loads, Impacts loads, Puncture loads, Vibration Loads, Combination of Bolt Forces/Moments from Different Loads).
The admissible stress criteria for normal conditions and hypothetical accident conditions for bolts are taken covering from NUREG/CR-6007 and ASME
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Code Sec Ill Division 3 WB-3230 (see section 2.2.1 "Normal conditions (NC)"
and section 2.2.2 "Hypothetical Accident Conditions (HAC)" in APPENDIX 2-1 to 1014-SR-0001 ). Therefore, the acting forces and moments are assessed against the admissible stresses according to NUREG/CR-6007 and ASME Code Sec Ill Division 3 WB-3230.
For the bolts, the stress criteria of the normal conditions (Level A) acc. to section 2.2.1 are applied under HAC. According to Division 3, WB-3234 (Level D service limits) the requirement of leak tightness of the closure may be satisfied by using the rules of Division 3, WB-3232 (Level A service limits).
In conclusion it is shown, that the stresses of bolts are computed in equivalence to NUREG/CR-6007 and additional that the assessments are performed with admissible stress criteria according to NUREG/CR-6007 and ASME Code Sec Ill Division 3.
RAI 2A-4: Explain how the NCT and HAG conditions envelope the condition where a canister drop from a failed lifting gear is encountered.
The lifting gear is not considered as a component important to safety (see Storage SAR section 1.2.2.1.3) and hence not addressed in the SAR. Thus, the failure of the lifting gear poses an unanalyzed condition not addressed by other safety requirements. The staff is of the opinion that this should be addressed under the handling operations that prepare the transport cask for transportation.
This information is required to meet the requirements of 10 CFR 71.
Answer:
According to the following explanations safety evaluations for a canister drop event are not necessary.
The ancillary equipment lifting gear is not relevant for the safety evaluation of the transport cask. However, since it is used during the cask loading, it is assumed to be designed single failure prove in accordance to ANSI N14.6 and
.NUREG-0612. The same assumption is also made for the crane systems of the nuclear facility. Furthermore, the load attachment points (threaded holes) on the canister lid are proven to resist stripping (Appendix 2-3 of SAR Section 2.12). The SAR demands the usage of sc_rews with a tensile strength of at least 1000 MPa. However, the screws are part of the lifting equipment (e.g. lifting pintle) which is also assumed to be designed single failure prove.
This guarantees a fall-proof load chain and thus, canister drop during transshipment is not a credible accident.
Subsection 1.2.4.1.2 is revised in Rev. 4 of the SAR to express this fact.
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RAI 2A-5: Make the item numbering in Tab. A 2 consistent with that of SAR table 2.1-2.
Aligning the item number between the two tables allow an easy evaluation of the appropriate items of interest in the table.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
Answer:
The numbering in Tab. A 2 "Design temperatures [°C] " in APPENDIX 2-1 is adapted to SAR table 2.1-2.
RAI 2A-6: Use the methodology of NUREG-6007 to compute the preload stresses in the bolts or show that the stresses computed using the VD/ approach results in the same values of the stress parameters that are used in the acceptance criteria from NUREG-6007.
The staff reviews the stress computation of highly stressed bolts using the guidance in NUREGICR 6007 hence the bolt stress shown in the SAR will be evaluated using the methodology of the guidance document.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
Answer:
The comparability of German code VOi 2230-1 with US code NUREG-6007 concerning the computation of preloads is shown in the following either by statement or by calculation. The preloads determined by means of both codes are comparable or German code VDI 2230-1 leads to covering preloads.
The answer is integrated in the SAR in the new section 2.1.4.1. "Comparison of German and US standards".
RAI 2A-7: For the NCT and the HAG drop analysis, provide information on the physical properties of the foam, along each direction, used in the LS-DYNA material model for the impact limiter simulations. Provide a comparison of these physical properties of the foam derived from testing to that from a simulation using the same LS-DYNA material model.
The staff needs assurance that the material models used can replicate the actual response of the foam under the specified drop conditions of NCT and HAC.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
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Answer:
In the following the information requested is depicted and summarized by showing the relevant passages in the SAR.
II
In addition, the relevant statement of the before mentioned reference [4]
"Numerical and Experimental Investigations of Polyurethane Foam for Use as Cask Impact Limiter in Accidental Drop Scenarios" is depicted here: "Additional investigations proved that the behavior neither depends significantly on the angle between specimen foam growth and load axis nor on its edge size.
Hence, the results are valid for arbitrary oriented components in all usual dimensions."
In conclusion the used material model is able to replicate the material behavior of the PU-foam.
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Fig. 1: Uniaxial compression tests at +23°C and a strain rate of 0.005 s-1 -
simulation compared to the tests
[4] Numerical and Experimental Investigations of Polyurethane Foam for Use as Cask Impact Limiter in Accidental Drop Scenarios-12099 Eva M. Kasparek, Holger Volzke, Robert Scheidemann, Uwe Zencker BAM Federal Institute for Materials Research and Testing, 12200 Berlin V\\IIVl2012 Conference, February 26-March 1, 2012, Phoenix, Arizona, USA
RAI 2A-8: For the NCT and the HAG drop analysis, provide information on how the contact area between the cask and the impact limiter was developed for the transfer of loads to the AN SYS cask and canister design models.
The impact force transferred to the cask and canister design elements from the drop is a function of the contact area post-drop. The staff needs to know how this contact area was developed from the LS-DYNA results.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
Answer:
The accelerations calculated with LS-DYNA are transferred to the AN SYS cask and canister model.
The post-drop contact area and the here acting pressure distributions resulting from the accelerations are derived from the LS-DYNA models for the ANSYS cask model. For the various drop orientations the derivation of the post-drop
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contact areas a n d the pressure distributions is shown below in detail for the ANSYS cask model.
In case of flat drops uniform pressure distributions, as ca lculated in the LS DYNA mode ls, are applied in the ANSYS cask mode ls. (see Fig. 1 and Fig. 2 be low, same as Fig. A 16 a n d Fig. A 27 in SAR APPENDIX 2-1 ).
NSYS cask mode l for flat drop onto lid side
Fig. 2 : Uniform pressure d istribution in the ANSYS cask model for flat drop onto bottom side
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In case of edge drops the procedure is described in SAR APPENDIX 2-1 section 10.8.6 "A 2 - 9 m Edge Drop onto Lid Side". The pressure distribution is depicted in Fig. 3 and Fig. 4 below (same as Fig. A 17 and Fig. A 18 in SAR APPENDIX 2-1 ). The pressure distribution at the time point of maximum forces are slightly simplified and applied onto the ANSYS cask models (compare Fig. 3 and Fig. 4 with Fig. 5 and Fig. 6; Fig. 5 and Fig. 6 same as Fig. A 21 and Fig. A 22 in SAR APPENDIX 2-1).
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In case of side drops the LS-DYNA results show no uniform pressure distribution for the lid and bottom impact limiter. The pressures occur in axial direction in the areas where the impact limiters overlap. In radial direction the pressure has approximately a sinusoidal distribution with its maximum in the symmetry plane and acts in inversed drop direction. The sizes of the post-drop
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contact areas in circumferential direction are geometrically derived from the deformation of the impact limiters after the drop. The post-drop contact area between the unyielding foundation and the impact limiter is projected onto the cask model. The procedure is shown in Fig. 7 below exemplary. The procedure conservatively underestimates the post-drop contact area size in circumferential direction of the ANSYS cask model, which leads to a higher load introduction.
Fig. 7: Procedure of deriving post-drop contact area size for ANSYS cask model The answer is integrated in the SAR in APPENDIX 2-1 in the section 10.7.2, 10.8.5, 10.8.6, 10.8.7 and 10.8.8.
RAI 2A-9: Provide a numeric tabulation of the deformation in the finite elements (FEs) leading up to the contact surface between the impact limiter and the cask extending the tabulation a few FE's beyond the contact surface.
The staff wants to be sure that the drop does not contribute to an inelastic deformation of the cask body resulting in internal work done by the FE's that represent the contact boundary. This is difficult to see for the presented time stepped pictures of the foam deformation.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
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Answer:
The cask body is modelled as a rigid in the LS-DYNA FE-Analysis (see Appendix 2-4 to 1014-SR-00001 section 4.1.1 ). The energy is solely dissipated by the impact limiter.
The following points must be guaranteed for the contacts which are significantly involved in the load transfer to exclude a reduction of the impact force respectively energy:
- No or negligible small negative contact energies (slave+ master) occur during the impact.
- No or negligible small penetrations in the contact regions occur.
Exemplarily, for the load case 9.3 m side drop, Pas. 2 -29 °C the before mentioned checks are depicted in the following.
The contacts which are significantly involved in the load transfer are:
- for the lid impact limiter: cask body to spacer, spacer to aluminum housing and aluminum housing to PU foam (see Fig. 4)
- for the bottom impact limiter: cask body to spacer and spacer to PU foam (see Fig. 5)
In figure Fig. 1 and Fig. 2 the contact energies for the above-mentioned contacts are depicted. There are no negative energies.
Fig. 1: Contact Energies of the lid impact limiter
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Fig. 2: Contact Energies of the bottom impact limiter In Fig. 4 and Fig. 5 deformation plots at various time points are provided from which possible penetrations of the above-mentioned contacts can be seen (Fig. 3 shows the locations of the sections of the deformation plots).
Additionally, in Fig. 7 to Fig. 11 time-history plots of selected points in the contact areas show that the penetrations are negligible small with values of 0 µm - 34 µm (Fig. 6 shows the locations of the selected points).
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Fig. 4: Deformation plots at various time points, section view of lid impact limiter (1/2)
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Fig. 4: Deformation plots at various time points, section view of lid impact limiter (2/2)
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Fig. 5: Deformation plots at various time points, section view of bottom impact limiter (1/2)
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Fig. 5: Deformation plots at various time points, section view of bottom impact limiter (2/2)
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Fig. 6: Locations of selected points for time-history plots of the distance between the significantly involved contacts
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In conclusion the contacts work as intended and a reduction of the impact force respectively energy is excluded.
The answer is integrated in the SAR in APPENDIX 2-4 in the section 6.1.
RAI 2A-10: Provide the criteria used for eroding the elements of the impact limiter as the drop proceeds.
As the drop energy is dissipated through the impact limiter some portions of the foam will be eroded i.e., unable to participate in resisting the drop. The staff wants information as to what strain level was this erosion set at in the model and how does it compare with the rupture strain of the foam.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
Answer:
This ensures that the impact forces are assessed conservatively, resulting in higher loads on the cask and providing a robust evaluation of its structural integrity.
If erosion were included, it would lead to a softer impact due to a greater impact distance covered by the impact limiter. Portions of the foam exceeding a certain strain level would be considered non-contributory. This softer impact would result in lower calculated impact forces transmitted to the cask, potentially underestimating the loads experienced.
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By not incorporating erosion, we ensure the foam's full energy absorption capacity is considered, leading to higher calculated loads on the cask. This approach deliberately accounts for the covering scenario, where the foam fully engages without losing any material effectiveness, resulting in a harder impact.
This methodology results in a more stringent and conservative evaluation, ensuring the cask can withstand higher impact forces.
The remaining height of the impact limiters after the drop, as illustrated in the attached figures below (excerpts from figures referenced in SAR APPENDIX 2-4 in section 6.1 (see Fig. A 16, Fig. A 18, Fig. A 20 and Fig. A 22), clearly indicate that the cask maintains a sufficient distance from the unyielding surface, providing sufficient safety. This remaining height emphasizes that even under severe impact scenarios, there is ample assurance that no direct contact occurs.
In summary, the absence of erosion in the model ensures a conservative assessment that applies higher loads on the cask, ensuring its integrity and safety under HAC.
Fig. 1: 0.3 m side drop, Pos. 2 - -29 °C: Max. deformation of impact limiter-sectional view
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Fig. 2: 9.0 m flat drop on top - -29 °C: Max. deformation of impact limiter - sectional view
Fig. 3 9.0 m edge drop on top, Pos. 2 - -29 °C: Max. deformation of impact limiter:...
sectional view
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RAI 2A-11: Provide information on the conditions to which the aluminum cladding of the impact limiter is exposed to during contact with the rigid surface. Discuss the effect of any drag i.e., the effect of friction and the stress in the aluminum sheet and its potential to tearing. Information is requested on any increase in temperature of the aluminum sheet as a result of dragging and the state of both the aluminum and the foam when exposed to this elevated temperature.
The staff wants to know if the aluminum sheet remains intact and that the elevated temperature does not degrade the modeled performance of the foam.
This information is required to meet the requirements of 10 CFR 71. 71 and 10 CFR 71.73.
Answer:
The lid and the bottom impact limiters consist of an aluminum housing with a PU-foam filling.
The cells are separated by thin aluminum sheets (see Fig. 1 and Fig. 2 below).
The innovative design of the within the impact limiters ensures exceptional durability and reliability.
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In conclusion, any impact resulting from dragging is localized and does not affect the overall performance of the impact limiter.
Moreover, the outer aluminum housing of the impact limiters may experience higher temperatures due to dragging and friction with a rigid surface. This exposure benefits from the material properties of aluminum. The fracture strain of aluminum, which is 15% at room temperature, increases at higher temperatures. Consequently, this characteristic ensures that there is no adverse effect on the tearing potential of the aluminum housing when subjected to elevated temperatures due to dragging. This robust design and material behavior ensure the continued effectiveness and safety of the impact limiters.
The impact duration of all evaluated free drops is less than - (see figures in section 6.1.1 (Fig. A 34 - Fig. A 37) and 6.1.2 (Fig. A 38 - Fig. A 51) of Appendix 2-4 to 1014-SR-00001 (1014-TR-00029)). In this short time period only a small amount of the PU-foam filling will be heated by the heat transfer of the aluminum sheets (see detailed discussion below). Therefore, it is only a local effect and the overall performance of the impact limiter is not impaired.
The temperature-increase after the drop test due to sliding by friction within the total impact limiter (mean temperature hub) and within the friction zone of the impact limiter (local temperature of sliding outer aluminum housing) are approximated in the following.
The covering assumed boundary conditions are:
- The mass of the loaded cask with impact limiters amounts to me =
150000 kg.
- The assumed sliding distance after the drop test is about Sci= 0.5 m.
- The friction coefficient in the sliding region is set to µ = 0.5.
- The minimum mass of one impact limiter is mIL = 6190 kg.
- The friction zone is set to A= 4 m 2 (see Fig. 3)
- The mass of the outer aluminum housing in the friction zone is about mAL = 53 kg.
- The minimum specific heat is c = 900 J/(kg*K).
The heat impact by friction is calculated to:
01 = m
- g * µ
- Sci = 367875 J The mean temperature increase within the impact limiter is calculated to:
I:::.. T1L = Qtf(m1L
- c) = 0.07 K The local temperature increase of the friction zone of the outer aluminum housing is calculated to:
l:::..TFz = Qtf(mAL
- c) = 7.7 K The calculations show that the mean temperature increase in the impact limiter is negligible. The mean temperature increase of the PU-foam is covered by the mean temperature increase of the impact limiter.
The heat transfer from the friction zone through the outer aluminum housing is much higher than through the PU-foam with a much lower heat conductivity.
The mean temperature increase within the PU-foam is negligible.
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RAI 2A-12: Use US equivalent codes for German design codes or show the equivalence by computing the stresses using both methods.
The staff is familiar with US codes used in the design and fabrication of spent fuel cask. While the acceptance criteria in the application uses US standards several of the design methodologies follow VD/. KTA, DIN or Eurocodes.
Though these methodologies are acceptable in Europe the review staff is unfamiliar with the analysis steps used in these codes.
This information is required to ensure that subsequent compliance with the requirements of 10 CFR Part 71 assures the safety of the transportation cask
Answer:
The comparability of German standards with US standards is shown in the SAR Rev. 4 in the new section 2.1.4.1 either by statement or by calculation. If the comparability is shown by calculation, all analyses are fulfilled, regardless the use of German standards or US standards.
The comparability of VDI 2230-1 concerning the determined preloads is shown by the answer to RAI 2A-6.
The comparability of VDI 2230-2 concerning the determined minimum safety for trunnion bolt stresses is shown by calculation according to
,,Machine Design" (August 17, 1967). The determined safety is greater than 1, regardless the use of German code or US code.
The comparability of KTA 3201.2 concerning the determined required length of engagement is shown by calculation according to,,Machinery's Handbook" (Edition by E. Oberg et al, 2012 Industrial Press New York).
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All verifications can also be successfully carried out by using the US technical code.
- The comparability of KTA 3201.2 concerning the verification concept of the fuel basket is shown by statement with regard to ASME BPVC.II1.3-2017 and ASME BPVC.III.A-2017.
- The verification concept used by GNS is applicable for the mechanical design of fuel baskets.
- The standards DIN EN 1993-1-5 (Eurocode 3) and DIN EN 1999-1-1 (Eurocode 9) concerning the analytical stability analyses are replaced by
,,Roark's Formulas for Stress and Strain" in report TR-1014-00028.
B. Materials Evaluation
RAI 2B-1: Provide additional information on the cast iron package coatings that demonstrate coating durability and support the emissivity values credited in the thermal analysis.
The specifications and minimum requirements for inner and outer coatings are provided in SAR section 2.2. 7, which does not provide any details about materials qualification -data, manufacturer data sheets, or other bases that demonstrate the performance of the coating under the elevated heat and radiation exposures of the internal package environment. The applicant states the coating is important to safety (ITS) as the thermal analysis relies on emissivity values. However, this is not indicated in the bill of materials or the drawing package. The staff notes that the emissivity specification and how the emissivity value will be verified is important, as well as what standard the applicant is using to verify and validate these values.
This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5) and 10 CFR 71.43(d).
Answer:
The requirements on the coatings in SAR Section 2.2.7 are supplemented with the explicit designation of two coating systems, one for the inner and another one for the outer cask surface. The proof of suitability of these coatings regarding the requirements is provided in a qualification report
1014-TR-00083 Rev. 0 Material Qualification Coatings
as Appendix 2-11 to Section 2.12 of the SAR Rev. 4. It shall be treated as proprietary information to be withheld from public disclosure. Amongst others, it provides manufacturer data sheets, results from GNS testing and reference to literature.
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The thermal evaluation does not take any credit from the emissivity of the cask cavity coating (which serves solely for corrosion protection), but from the emissivity of the outer coating, which is proven by literature. Table 2-22 is adjusted accordingly.
RAI 2B-2: Justify that the fuel basket plates will have adequate fracture performance in a drop accident to maintain the assumed fuel configuration in the criticality analyses.
The SAR does not include toughness testing requirements to verify that brittle fracture will not affect the structural integrity of the basket in a drop accident.
The staff notes that the criticality analysis relies on the configuration of the neutron absorber plates and fuel assemblies being maintained.
Although nonferrous materials are generally excluded from fracture acceptance testing in consensus standards, the proprietary metal matrix composite is a noncode material that contains boron carbide ceramic particles that may diminish fracture performance relative to the conventional aluminum material that are considered in the ASME BPVC.
This information is needed to demonstrate compliance with 10 CFR 71.33(a)(5)(ii), 10 CFR 71.31(c) and 10 CFR 71.55(e).
Further explanation by NRG (E-Mail 24.01.2024):
GNS provided material qualification data for the fuel basket material, however fracture toughness data was not provided which was the original RA/. If the fuel baskets are relied upon for mechanical performance including under accident conditions, then GNS needs to establish that material has adequate strength, ductility and fracture toughness to keep geometry control of neutron absorber plates and maintain fuel orientation.
Answer:
Section 2.4 of the material qualification report 1014-TR-00011, that is attached to the SAR as Appendix 2-8 has been revised, demonstrating the ductile behavior of the basket plates even at low temperatures. The assessment is supplemented by a discussion of scanning electron microscopy observations of the fracture surfaces of dynamically cracked test specimen (Charpy V-notch test).
RAI 2B-3: Provide analysis to demonstrate that the package drying criteria is adequate to remove residual moisture such that it limits hydrogen generation and clarify operational procedures that would prevent accumulation of hydrogen during loading operations.
The SAR section 7.1.2 provides process steps but is unclear why this will provide adequate dryness and removal of moisture. Additionally, in table 7.1-2 "Operations for loading of contents," it is unclear if and how hydrogen levels
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are monitored to ensure that the flammable gas mixture is monitored and mitigated.
This information is needed to demonstrate compliance with 10 CFR 71.43(d).
Answer:
In Rev. 4 of the SAR, Subsection 7.1.2 has been adjusted as follows:
The vacuum drying process reduces the MPC pressure in stages to prevent the formation of ice. Step by step, the vacuum pressure is reduced to 1 kPa.
The boiling temperature at that pressure is 7 °C, which is far enough away from the freezing temperature of water, thus icing of the system can be excluded.
Subsequently, after a sufficient drying time to remove most of the liquid water from the cavity, the vacuum pressure is further reduced to 0.4 kPa. Even if the dew point of water at this pressure is about -5 °C, icing of the system is negligible due to the very low residual moisture in the cavity. After a sufficient drying period, the interior is disconnected from the vacuum pump and the maximum permissible residual moisture is verified. The proof of maximum permissible residual moisture is considered to be provided if the 0.4 kPa pressure criterion can be kept stable for at least 30 minutes. In accordance with NUREG-2215 (Section 8.5.15.2.3, examples of accepted methods for drying),
the amount of residual water is constrained to only a few grams as shown by
[Knoll/Gilbert]. Thus, fuel cladding degradation due to oxidizing gases (e.g., 02, CO2 and CO) arising as a result of the conversion of the residual water over the operating/storage time is within an acceptable level. To the same extent, the amount of hydrogen produced is also tolerable. After successful completion of drying, the cavities are backfilled with the inert gas helium (
) to a pressure of at to promote heat transfer and prevent cladding degradation.
RAI 28-4: Provide additional information to demonstrate that the fuel assemblies are not exposed to a non-inert environment during the dewatering, drying and helium backfilling process during loading operations.
The SAR section 7. 1.2 provides process steps, but it is not clear if this provides adequate protection of fuel assemblies without any adverse effects to the material (fuel cladding).
This information is needed to demonstrate compliance with 10 CFR 71.43(d) and 10 CFR 71.55(d)(2).
Answer:
The work steps dewatering, drying final inertization with helium and subsequent closure are carried out sequentially and without interruption. The canister is never left with an uncontrolled/ undefined interior atmosphere (e.g. open). Only the canister atmospheres flooded with water, water vapor and helium (and their relevant mixed states) need to be considered. Thus, the fuel assemblies are
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not exposed to any non-inert environment and adequate protection of fuel assemblies without any adverse effects to the fuel (cladding) material is permanently provided.
A respective paragraph has been added to subsection 7.2.1 in Rev. 4 of the SAR
RAI 2B-5: Provide material qualification for ductile cast iron and discuss how the procurement of the cast iron cask body is contr(!lled to ensure that the casting facility conducts required sampling and testing to qualify the material performance and ensures that the casting facility does not perform repairs of casting defects, per the ASME Boiler and Pressure Vessel (B&PV) Code criteria. Absent such controls, describe how repairs are performed, documented, and examined to ensure mechanical properties of the cask are uniform and meet the minimum requirements of the ASME B&PV Code.
SAR section 1.2.1.1 states that the cask body is made of ductile cast iron, composed of a hollow cylinder with a closed bottom end, cast in one piece.
ASME B&PV Code, Section Ill, Division 3, WB-2573 states the following:
"Casting shall not be repaired by plugging, welding, brazing, impregnation, or any other means."
The staff notes that these ASME B&PV Code requirements are in place to eliminate specific issues with each of the repair methods to potentially degrade the microstructure and mechanical properties of ductile cast iron. If controls are not in place to ensure that defe~ts are not repaired, the staff requests information on how defects will be repaired and examined to ensure that code required properties will be achieved in these localized areas.
The information is needed to determine compliance with the regulatory requirements in 10 CFR 71.31(c).
Answer:
A subsection regarding material testing has been added to Section 8.1 of Rev.
4 of the SAR. The procurement procedure of material including controls is explained in detail for the DCI cast body as an example. All specifications and acceptance criteria are provided to the manufacturer via fabrication and test plans, material test specifications, and fabrication specifications.
In addition, Chapter 2 is supplemented by another material qualification report (1014-TR-00048 Rev. 0) regarding DCI as appendix. It shall be treated as proprietary information to be withheld from public disclosure.
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C. Thermal Evaluation
RAI 3-1 Discuss the thermal impact of package temperatures and pressures if the surface enhancement factor was applied to the radiant energy component during the 30-minute engulfing fire.
SAR section 3.3.1.4 indicated that the package's radial fins were not explicitly modeled; rather, a surface enhancement factor was applied to the heat transfer coefficient at the model's corresponding unfinned surface. In addition, SAR figure 3.4-1 indicated that radiant heat transfer is the dominant heat flux component to the package. However, there was no discussion regarding the impact of increased thermal input to the fin's additional area from increased radiation heat transfer during the fully engulfing fire accident condition. Staff notes that the concept of "angle of incidence", as indicated in RA/ response 3-11 and mentioned in SAR section 3. 3. 1.4, generally would not be a factor for the assumed fully engulfing fire.
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
The SAR Rev. 4 has been advanced by a Subsection 3.4.3.3, providing a variation calculation considering an increased effective emissivity of the cask lateral surface during HAC fire. The subsection additionally describes why a surface magnification factor for radiative heat transfer during the fire phase is not appropriate.
RAI 3-2 Clarify the process for determining the need for a personnel barrier if thermal measurements during pre-shipment operations indicate excessive surface temperatures and confirm the personnel barrier would not adversely impact package thermal performance.
- a. While reviewing the responses to RA/s, staff noted that SAR table 7.1-3 (footnote 2) was revised; it stated that a personnel barrier is to be placed on the package if measurements during pre-shipment operations indicate surface temperatures (or doses) do not meet regulations (e.g., 10 CFR 71.43, 10 CFR 71.47). The procedure did not provide instruction to first investigate the reason for the excessive temperature (e.g., incorrect loading, source term methodo/ogy1) in order to confirm the package is in compliance with the certificate of compliance
- b. SAR section 3.3.2.3 indicated that thermal model results showed margin between the maximum package accessible surface and the 85 deg C exclusive use shipment requirement of 10 CFR 71.43(g). Likewise, SAR section 3.3.1.5.4 and section 3.3.2.1 listed conservative assumptions which should indicate the maximum package accessible surface temperature is conservative. However, the SAR did not offer potential explanations for a surface temperature measurement exceeding the 85 deg C exclusive use temperature limit.
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- c. There was no thermal analysis to demonstrate acceptable package thermal performance with a personnel barrier, recognizing that heat transfer from the package could be adversely impacted depending on the personnel barrier design.
This information is needed to determine compliance with 10 CFR 71.35(a),
71.43(g), 71.47.
Answer:
Thank you for pointing this out. Unfortunately an error has crept into footnote 2 of table 7.1-3. The use of the personnel barrier is only associated with increased dose values as described in detail in Subsection 7.0.2. There are no credible reasons for the surface temperature to exceed the 85 °C limit. If, contrary to expectations, the limit is exceeded, this is not permitted as it does not comply with the safety assessments, and suitable measures must be taken as it can be assumed that the loading is incorrect.
The thermal references have therefore been removed from footnote 2 that is only correlated with dose rate measurements and a footnote 3 has been added to demand suitable measures in the unlikely event of a thermal limit exceedance.
RAI 3-3: Provide the comparison inputs and calculation that demonstrate CASTOR geo69 pressures are bounded by the assumed 0. 15 fission gas release fraction of high-burnup fuel, relative to Jow-burnup fuel that has a 0.30 percent fission gas release fraction.
The response to RAJ 3-4 and SAR section 4. 2. 1 appeared to indicate that the high-burnup fuels listed in SAR table 1.2-12 result in greater fission gas amounts and canister pressures than low burnup fuels. However, there was no calculation for staff to review and make findings.
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
In the revised SAR section 4.2.1 (and 4.3.1 ), it is justified that high-burn up fuels result in greater fission gas amounts and canister pressures than low-burnup fuel for physical reasons which are compiled in Appendix C.5 of NUREG/CR-7203.
Neglecting these physical reasons, however, in Rev. 4 of the SAR also the 30 % fission gas release fraction for low-burnup fuel, has been addressed and assessed in the form of variation calculations in the containment (section 4.2.1 and 4.3.1) and thermal (section 3.5.4) evaluations and by re-evaluation of the structural analyses (section 2.6.11 and 2.7.8):
- A comparison of temperatures for varying fraction of fission gas release (15 % to unphysical 30 %) shows that all maximum permissible
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component temperatures are complied with. The maximum temperature rise occurs in the fuel rods and amounts to 6 K (NCT) and 17 K (HAC fire).
- In section 4.2.1 and 4.3.1, it is shown that the calculated (unphysical) pressures are higher than for high-burnup fuel in NCT ( *
) and HAG-fire ( 11.
- To demonstrate the robustness of the CASTOR geo69 design, GNS has re-evaluated all load cases in the structural analyses due to the new pressures in section.
RAI 3-4: Clarify in SAR chapter 7 Operations that the condition and cleanliness of the fins are maintained prior to shipment to ensure fin effectiveness during transport.
The SAR section 3. 3. 1.4 indicated that the effectiveness of the cask fins is an important factor to achieve thermal performance of the package. Fin effectiveness is dependent on maintaining fin geometry (i.e., no large areas of deformed fins) and cleanliness (i.e., no fouling factors, maintaining emissivity factors). However, the response to RA/ 3-2 did not provide explicit instruction in SAR table 7. 1-3 to repair the fins if the H3 inspection step finds unacceptable fin condition (e.g., deformations) and cleanliness (e.g., fouling factors).
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
Spatially limited changes of the conditions of the cask fins due to dirt, debris or small deformations would only lead to negligible changes in the thermal behavior and the effectiveness of the heat removal. (As a comparison during transport the relative large contact area of the cask surface with the supports of the transport frame are, which has no significant impact on the heat removal during NCT). However, prior to each transport (and also periodically during storage at an ISFSI) visual inspections of the cask surface verify the proper condition of the fins. If contrary to expectations soiling or local deformation of the fins is detected, at least the soiling shall be removed.
Therefore, an additional inspection step (H3.2) regarding fin cleanliness has been added to Table 7.1-3 in the new revision of the SAR also considering cleaning if applicable.
A rather massive, large scale deformation of the fin geometry, which would have an influence on the thermal behavior must have been the consequence of HAC or an handling accident. Such deformations are thus not relevant here.
RAI 3-5: Clarify in the SAR the basis for applying approximately 18.4 kW decay heats (rather than the 18.5 kW and approximate bounding decay heat described in SAR section 3.1.2 and section 3.3.2.1) on the thermal models.
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SAR sections 3.3.2.1, 3.4, 3.5, and 3.6 indicated that thermal analyses considered decay heats less than 18. 5 kW. However, the basis for demonstrating a loading maximum decay heat of 18. 5 kW, as noted in SAR section 3.1.2, was not clearly described or shown in the results.
This information is needed to determine compliance with 10 CFR 71.35(a).
NRG addendum (dated 01/25/2024) following GNS inquiry (dated 01/17/2024):
SAR section 3. 3. 2. 1 states the following: "The heat power of the cask loading applied to the FE model is by higher than defined Section 3. 1. 2 for the thermal requirements."
It would appear that indicates a decay heat was applied to the thermal model (for example,
If the above understanding is incorrect, then possibly the wording in the SAR can be adjusted. Note that the RAJ response should clarify the decay heats applied to the thermal models (e.g., NGT, fire HAG, impact HAG, short-term operations), especially as it relates to the SAR statement that up to 18. 5 kW can be loaded. For example, if thermal requirement 2 is applied (which is Jess than 18. 5 kl/ll) to a thermal model, then what is the demonstration for a finding that 18. 5 kW can be loaded?
Answer:
The total decay heat power of all thermal requirements is overestimated by
- for NGT (see Sec. 3.3.2.1; the respective bullet point has been precised).
An overestimation is considered for all calculation NGT (section 3.3), HAG (section 3.4) and short-term operations (section 3.6) such that the total decay heat of the most unfavorable thermal requirement is always greater than 18.5 kW. This is stated in various sections of the report.
Section 3.4.1.1 states "For the analysis under HAG, the same geometric 30 FE model is utilized as for NCT described in section 3.3.1." (see page 3.4-2 of SAR Rev. 4)
Section 3.5.2.1 states "The same numerical model is used as described in Section 3.3.1 for NGT." (see page 3.5-5 of SAR Rev. 4)
Section 3.5.3.1 states "The same numerical model is used as described in Section 3.4." (see page 3.5-9 of SAR Rev. 4)
Section 3.5.5.1 states "For scenario I of HAC impact, the same numerical model is used as described in 3.4 for HAC... "(seepage 3.5-20 of SAR Rev. 4)
Section 3.5.6.1 states "The same numerical model is used as described in Section 3.3.1 for NCT." (see page 3.5-30 of SAR Rev. 4)
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RAI 3-6: Clarify that the effective thermal conductivity applied to the 3-dimensional (30) model basket's homogenized fuel correctly considers the appropriate amount of thermal radiation heat transfer.
The response to RAJ 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. These thermal analyses, including those for normal conditions of transport a!1d hypothetical accident conditions, were based on SAR section 3.3. 1. 5.2, which stated that radiation heat transfer among fuel rods and inner surfaces of the fuel channel, and among the fuel channel and the inner surface of the basket sheets is accounted for in the detailed 2-dimensional (20) model and the simplified 20 model. However, SAR figure 3.3-7 indicated that the half-symmetric 30 model of the canister and cask includes the fuel channel, the helium between the fuel rods and fuel channels, and the helium between the fuel channels and outer basket sheets (i.e., as part of the homogenized fuel's effective thermal conductivity and explicitly modeled in the 30 ANSYS model). This may appear to indicate that radiation heat transfer within the basket is accounted for twice, which would not be an accurate or bounding assumption.
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
The simplified model of a FA contains the homogenized fuel rod zone, fuel channel, helium gap and basket sheets.
As described in section 3.3.1.5.2, in the simplified model of a FA, the fuel rods, the water rods and the gas atmosphere between the fuel rods are substituted by a homogenized zone. The homogenisation is performed for the region within the fuel channel. The helium gap between the fuel channel and the basket sheets is not part of the homogenized fuel rod zone.
Hence, the radiation in the helium gap of the FA is not counted twice and the radiation heat transfer is modelled correctly.
Additional Question by Mail dated March 22, 2024:
Although SAR table 3. 6-3 provided the thermal conductivity of water vapor, it was not clear that the thermal conductivity values represented a pressure of 0.01 bar (i.e., only density property mentions 0.01 bar). In addition, considering that SAR table 3.3-5 only listed effective thermal conductivity of the fuel assembly based on helium (i.e., no tabulated values based on water vapor),
there was no discussion that demonstrated the effective thermal properties (e.g., thermal conductivity) of the fuel basket assembly used during the steady=-
state vacuum drying thermal analysis were based on 0.01 bar water vapor properties.
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Answer:
The thermal conductivity of stationary gases is pressure independent for all pressures above 1 mbar (=0.001 bar) [1 ]. The source in German can be found below.
[1] Grundlagen derVakuumtechnik, Dr. Walter Umrath (translated by GNS)
It is confirmed that the fuel assemblies for vacuum drying and helium backfilling have been homogenized for 100% water vapor and 100% helium atmosphere respectively. The thermal properties of the homogenized fuel assemblies during vacuum drying have been added to the SAR Rev. 4
RAI 3-7: Demonstrate as part of the SAR chapter 3 thermal analyses that spent fuel baskets and other components that are important to safety (i.e., canister,
_ transfer cask, cask loading unit) for operations within the reactor building would not be affected by the high temperature listed in SAR section 3.6.4.3 and without the use of air conditioning within the reactor building. In addition, clarify if the following considerations would impact the conclusions in SAR section
- 3. 6. 6 and section 3. 6. 7 regarding there being no temporal restrictions and no need for additional calculations for transfer operations inside the reactor building and loading of the canister in the cask(s).
- a. The response to RA/ 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. Staff notes that Part 72 SAR section 12.1.2 stated that off-normal temperatures during cask loading unit (CLUJ handling operations in the reactor building are not credible and Part 72 SAR section 2.2.5.1 indicated there are no off-normal environmental temperatures within the reactor building because it is assumed that the building is air conditioned. However, active cooling of the CASTOR geo69 system's heat sink (i.e., internal reactor building temperature) via air conditioning is inconsistent with regulations requiring only passive cooling.
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- b. Part 72 SAR section 12.1.2 stated that it was assumed that off-normal temperatures during CLU handling operations in the reactor building are covered by normal temperature evaluations for the Part 72 application.
Although the thermal analysis of the off-normal storage condition at a 52 deg C ambient temperature discussed in Part 72 SAR section 4. 5.4 indicated that ITS components were below allowable temperatures, it was not demonstrated that the content and ITS component temperatures within the canister and transfer cask (which does not include fins) is bounded by the finned CASTOR geo69 storage system thermal analysis at the potential higher 52 deg C ambient temperature. Likewise, the 53-hour time limit to begin the dewatering process discussed in the Part 71 SAR section 3. 6. 3 would be reduced if ambient temperature was raised from 35 deg C to 52 deg C (per item a, above).
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
New calculations for off-normal conditions (52°C for both, ambient and SNF pool water temperatures) during vacuum drying and helium backfilling are performed. Gaps in the transfer cask (see RAI 3.8) have been considered thereby. Corresponding sections in the SAR have been updated in the new revision.
RAI 3-8 Clarify and demonstrate as part of your thermal analysis in SAR chapter 3 the appropriateness of the thermal model assuming no gaps between transfer cask components.
The response to RAJ 3-14 was to add thermal analyses to the Part 71 SAR for short-term operations. Although Part 72 SAR section 1.2.2.1.1 indicated the transfer cask is fabricated from a number of radial sections (e.g., outer shell, water jackets, lead), Part 72 SAR section 1. 2. 2. 1. 6 indicated an absence of air gaps in the transfer cask body. Likewise, a visual review of the ANSYS model indicated there were no contact resistances between the radial components.
However, there was no discussion that demonstrated assurance that fabrication would be possible without the presence of gaps between components (e.g., inner liner and lead shielding, which is poured or composed of multiple cast parts or lead sheets, per Part 72 SAR section 10. 1. 6. 3) and the corresponding thermal contact resistance that would result in increased component temperatures. For example, Part 71 SAR table 3. 6-5 appears to indicate that increased thermal resistance through the transfer cask (i.e.,
presence of gaps within transfer cask) may result in some ITS components (e.g., basket sheets, polyethylene shielding CLU components) being above allowable temperatures.
This information is needed to determine compliance with 10 CFR 71.35(a).
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Answer:
Gaps have been modelled in the transfer casks sandwich structure and considered for the recalculations in the short-term operation calculations. The new revision has been updated accordingly.
RAI 3-9 Demonstrate and clarify in the short-term section 3. 6 of the SAR that all ITS components are within allowable values and whether there are time limits during those operations, including when the canister is within the transfer cask on top of the CASTOR geo69 cask using the cask loading unit.
The revised SAR section 3. 6. 7 of the recent submittal indicated that there was no need to analyze the CASTOR geo69 system with the cask loading unit, such as the condition of the transfer cask containing the fuel rod filled canister being positioned on top of the CASTOR geo69 cask with the bottom lid closed (as described in SAR section 1.2.4.3.4) and in which the canister would be enclosed in the transfer cask (which has no cooling fins and no improved external convection heat transfer capability). However, there were no temperature comparisons of the ITS components with their allowable temperatures (e.g., polyethylene shielding material within the transfer lock, per Part 72 SAR section 1.2.2.1.2) and no discussion of potential time limits.
This information is needed to determine compliance with 10 CFR 71.35(a).
Answer:
For the operation phase of vacuum drying of the transfer cask interior, a temporal restriction amounting to - is required (see section 3.6.4 in the new revision).
The temperatures of all components during helium backfilling are within the allowable limits. No time restrictions are need for this operation. (See section 3.6.5)
Section 3.6.6. deals with the configuration 'Transfer of the canister via transfer cask inside the reactor building'. The section clearly explains why this constellation is covered by the assessments in Section 3.6.5 and that therefore no further calculations and assessments need to be carried out. There are thus no time restrictions for the transfer inside the reactor building.
In Rev. 4 of the SAR the explanations in section 3.6.6 have been extended.
Further extensions (e.g. comparisons with limit temperatures) of this section are not necessary from our point of view since the maximum temperatures presented in Section 3.6.5.4 are also valid here.
Section 3.6. 7 deals with the configuration 'canister inside transfer cask positioned on the transfer lock on top of the CASTOR geo69 cask'. This section has been extended in Rev. 4 of the SAR to provide further argumentation to prove, that this constellation is covered by the assessments in Section 3.6.5 and that therefore no further calculations and assessments
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need to be carried out. There are thus no time restrictions.
Further extensions (e.g. comparisons with limit temperatures) of this section are not necessary from our point of view since the max temperatures presented in Sect. 3.6.5.4 are also valid here.
D. Containment Evaluation
RAI 4-1 Revise Table 1. 0-2 "List of BPVC Alternatives for the CASTOR geo69 DSS Transport Cask" on page 1.0-9 and provide additional justification for the request for the approval of a significant departure from the ASME B&PV Code Section Ill, Division 3, WB-6120.
Table 1.0-2 "List of BPVC Alternatives for the CASTOR geo69 DSS Transport Cask" on page 1.0-9, provides a justification for an alternative to ASME B&PV Code Section Ill, Division 3, WB-6120, for hydrostatically, or pneumatically pressure testing of the transport system. The justification provided includes the following statement:
The requirements of ASME B&PV Code Section Ill, Division 3, WB-6120, applicable to transportation containments, include pressure testing in accordance with WB-6200 (hydrostatic test) or WB-6300 (pneumatic test) and then, in addition, helium leak testing in accordance with WB-6700. The alternative listed in table 1. 0-2 does not address the combined requirement of pressure testing and helium leak testing required for a transportation containment designed, constructed, and tested in accordance with the requirements of ASME B&PV Code Section Ill, Division 3, Subsection WB -
Class TC Transportation Containments. This exception, if granted by the NRG as requested, would provide relief from the ASME code requirement for any pressure testing of the geo69 transportation containments in the course of serial production of both the "cask" and "canister", based solely on the structural analysis of the geo69 transport "cask" and "canister" presented in section 2 of the SAR. The alternative proposed represents a significant
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departure from the ASME code requirements, one that NRG staff has not previously granted to any applicant for certification of a transportation package design; therefore, the alternative requested would need additional justification in order to be further considered by the staff.
This information is needed to determine compliance with 10 CFR 71.31(c) and 71.BS(a).
Answer:
The application to waive the pressure test and full scale LT on each single cask as part of serial production is withdrawn from Rev. 4 of the SAR. Both tests will be performed regularly.
Table 1.0-2 has been revised with respect to deviating from ASME B&PV Code Section Ill, Division 3, WB-6120. The corresponding line in Table 1.0-2 has been remove in the SAR Rev. 4.
In addition, the statement of omitting full scale LT on both independent containment boundaries was removed in Sections 1.0, 8.1, and subsections therein.
E. Quality Assurance Analysis
The following questions are on the GNS-Quality Assurance Program Description (QAPD) submitted August 30, 2023 (ML23257A096).
RAI 17-1: Revise chapter 17 to clarify and describe what applicable codes and standards are used in the storage, retention, and maintenance of both electronic and hardcopy quality assurance records.
The GNS QAPD chapter 17 states, in part, the requirements and responsibilities for records transmittal, retention (such as duration, location, fire protection, and assigned responsibilities), and maintenance shall be consistent with applicable codes, standards, and procurement documents. The staff needs to know what the applicable codes and standards are and if this information is within the procedures and instructions that control the storage, retention, and maintenance for both electronic and hardcopy quality assurance records. The NRG regulatory guide (RG) 7. 10, revision 3, "Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material,"
and RG 1.28, revision 4, "Quality Assurance Program Criteria (Design and Construction)," which endorsed Nuclear Quality Assurance (NQA)-1-2008, and NQA-1a-2009 Addenda provide regulatory information about the control of electronic records. Both RGs provide reference to an NRG, Regulatory Issue Summary 00-18, "Guidance on Managing Quality Assurance Records in Electronic Media," (ML003739359), which provides applicants and licensees
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with a way to satisfy the requirements for the maintenance of quality assurance records.
This information is needed to demonstrate compliance with 10 CFR 71.135.
Revise chapter 18. 2 to remove a description that allow the internal audit interval to be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities (i.e., the quality assurance program (QAP) elements) are being satisfactorily accomplished. The NRG RG 7. 10, revision 3, and RG 1.28, revision 4, that endorse the ASME NQA-1 took exception to the statement that the interval may be extended up to two (2) years.
The GNS QAPD chapter 18.2 states, in part, that all applicable QAP elements shall be audited at least once each year or at least once during the life of the activity, whichever is shorter. Further, section 18. 2 goes on to state that this interval may be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities are being satisfactorily accomplished in accordance with the applicable QAP elements. This language is consisted with the ASME NQA-1-2008 and NQA-1a-2009 Addenda, Requirement 18. However, in NRG RG 7.10, revision 3, "Establishing Quality Assurance Programs for Packaging Used in Transport of Radioactive Material,"
and RG 1.28, revision 4, "Quality Assurance Program Criteria (Design and Construction)," which endorsed NQA-1-2008, and NQA-1a-2009 Addenda, Requirement 18, provide a regulatory position that all applicable QAP elements shall be audited at least once each year or at least once during the life of the activity, whichever is shorter. The NRG RGs took exception to the statement that the interval may be extended up to two (2) years based on the results of an annual evaluation and objective evidence that the activities are being satisfactorily accomplished.
This information is needed to demonstrate compliance with 10 CFR 71.137.
Revise chapter 18.3.2 to provide a description on how the lead auditor qualifications meetthe NRG RG exception provided in RG 1.28, revision 5 as committed to and described in ASME NQA-1-2015. For example, the description can be as simple as the lead auditor qualifications and certifications are described in implementing procedures and account for the exceptions identified in NRG RGs for endorsed ASME NQA-1 standards.
The RG 1.28, revision 5 endorses ASME NQA-1 2015 with certain clarifications and exceptions. The exception to the lead auditor qualifications is as follows
"Prospective lead auditors, with comparable industry experience
[emphasis added], may satisfy the lead auditor qualification requirement of participating in a minimum of five quality assurance audits within a period of 3 years prior to the date of qualification by alternatively demonstrating the ability to properly implement the audit process, effectively organize and report results, and
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participate in at least one nuclear audit within the year preceding the date of qualification, subject to review and acceptance by the responsible QA organization [emphasis added]."
This information is needed to demonstrate compliance with 10 CFR 71. 105 and 10 CFR 71.137.
Answer:
REGULATORY GUIDE 1.28, REVISION 6 requests:
For the management of electronic records, appropriate controls on quality include the following:
- No deletion or modification of records is allowed unless authorized pursuant to the record retention rule.
- Redundancy (e.g., system backup, dual storage) is provided.
- Legibility is required of each record.
- Records media are properly maintained.
- Random inspections ensure no degradation of records.
- Records are acceptably converted into any new system before the old system is taken out of service
Implementation within GNS-QAPD-001 Rev. 2 Chapter 17.5:
"Storage, retention, and maintenance of both electronic and hardcopy quality assurance records shall be performed on the basis of:
- ISO 9001
- ISO 27001
- IAEA GSR-2, requirement 8
- BAM-GGR 011 and
- KTA 1401, chapter 11.
The requirements of RG 7.10 Rev. 3 and RG 1.28 Rev. 6 are consistent with the above-mentioned codes and standards."
To demonstrate compliance with 10 CFR 71.137 the following is implemented within GNS-QAPD-001 Rev. 2 Chapter 18.2:
"All applicable Quality Assurance Program elements shall be audited at least once each year or at least once during the life of the activity, whichever is shorter."
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To demonstrate compliance with 10 CFR 71.105 and 10 CFR 71.137 the following is implemented within GNS-QAPD-001 Rev. 2 Chapter 18.3.2:
"Prospective Lead Auditors, with adequate education and industry experience, may satisfy the Lead Auditor qualification requirement
- of participating in a minimum of five quality assurance audits within a period of three years prior to the date of qualification
- or alternatively
- demonstrating the ability to properly implement the audit process,
- effectively organize and report results, and
- participate in at least one nuclear audit within the year preceding the date of qualification subject to review and acceptance by the MD
/QD."
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