ML20211L707

Response to Request for Supplemental Information - Tn Americas LLC, Renewal Application for the TN-68 Dry Storage Cask System, Certificate of Compliance No. 1027 (Docket No. 72-1027, CAC No. 001028, EPID: L-2020-RNW-0014)

ML20211L707
Person / Time
Site:	07201027
Issue date:	07/29/2020
From:	Narayanan P Orano TN Americas
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
CAC 001028, E-57065, EPID: L- 2020-RNW-0014
Download: ML20211L707 (6)

Text

July 29, 2020 E-57065 Orano TN 7160 Riverwood Drive U. S. Nuclear Regulatory Commission Suite 200 Columbia, MD 21046 Attn: Document Control Desk USA One White Flint North Tel: 410-910-6900 11555 Rockville Pike Fax: 434-260-8480 Rockville, MD 20852

Subject:

Response to Request for Supplemental Information - TN Americas LLC Renewal Application for the TN-68 Dry Storage Cask System, Certificate of Compliance No. 1027 (Docket No. 72-1027, CAC No. 001028, EPID: L-2020-RNW-0014)

Reference:

Letter from Christian Jacobs (NRC) to Prakash Narayanan (TN Americas LLC), Request for Supplemental Information - TN Americas LLC Renewal Application for the TN-68 Dry Storage Cask System, Certificate of Compliance No. 1027 (Docket No. 72-1027, CAC No. 001028, EPID: L-2020-RNW-0014), dated June 30, 2020 TN Americas LLC (TN) hereby submits our response to the Request for Supplemental Information (RSI) forwarded by the letter referenced above. Enclosure 1, herein, provides the response to the RSI. The RSI response did not require changes to the Application for Renewal of Certificate of Compliance (CoC) No. 1027 for the TN-68 Dry Storage Cask.

Should you have any questions regarding this submittal, please do not hesitate to contact Mr. Douglas Yates at 434-832-3101, or by email at Douglas.Yates@orano.group.

Prakash Narayanan Chief Technical Officer cc: Christopher Markley, NRC DFM Doug Yates, TN

Enclosure:

1. RSI and Response

RSI and Response Enclosure 1 to E-57065 Materials RSI:

RSI-M1:

Provide supplemental information that addresses the differences between the cladding alloys to demonstrate how the M5 cladding included in the Department of Energy (DOE) High Burnup Dry Storage Cask Research and Development Project can be used as a surrogate for the high burnup Zircaloy-2 clad fuel in the TN-68 Dry Storage Cask.

The supplemental information should address the differences between the high burnup Zircaloy-2 cladding and the M5 in the following areas: (1) the potentially higher hydrogen content in the cladding, (2) the higher fuel burnup, and (3) the lower threshold stress for reorientation. In addition, the supplemental information should address the maximum cladding temperatures for the TN-68 dry storage cask and the temperatures measured in the DOE High Burnup Dry Storage Cask Research and Development Project to show how it can be used as a surrogate.

Additional information is provided below.

As background, the CoC No. 1027 Renewal Application High Burnup Fuel Aging Management Program Section 4.5.1 states:

The parameters of the surrogate demonstration program are applicable to the Design bases HBU fuel, as:

• maximum allowed burnup of the design-bases HBU fuel (i.e., 60.0 GWd/MTU) is on the order of the nominal burnup of the fuel in the surrogate demonstration program (i.e., 58 GWd/MTU),

• the similar cladding texture between Zircaloy-2 and M5 of recrystallized annealed (RXA) and higher hoop stresses in the M5 PWR fuel cladding when compared to Zircaloy-2 BWR fuel cladding, M5 PWR fuel cladding can be considered an enveloping surrogate for Zircaloy-2 BWR fuel cladding, and

• the cladding temperature of the HBU fuel is limited to the values in ISG-11 and the cladding temperature in the surrogate demonstration program is as close to the ISG-11 limits as practicable.

The staff notes that, in addition to cladding texture and hoop stress, cladding hydrogen content as a result of fuel burnup, threshold stress, and temperature are important factors for hydride reorientation in spent fuel cladding. The staff is seeking information regarding the differences between the high burnup Zircaloy-2 cladding and the M5 in the following areas:

1. A comparison of Zircaloy-2 and M5 cladding alloys in NUREG-2214 Section 3.6.1 shows that above 45 gigawatt-days per metric ton of uranium (GWd/MTU), the hydrogen uptake in Zircaloy-2 cladding is greater than the hydrogen uptake for M5 cladding (NRC, 2019),

which leads to greater hydrogen content. In addition, the Zircaloy-2 cladding appears to have more variability in hydrogen uptake cladding when the fuel burnup exceeds 50 GWd/MTU (Geelhood and Beyer 2011) compared to M5 with a similar burnup (Hanson et al. 2016, Figure 3-11).

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RSI and Response Enclosure 1 to E-57065

2. The actual maximum burnup for the M5 cladding in the DOE High Burnup Dry Storage Cask Research and Development Project is 53.5 GWd/MTU (Hanson et al. 2016 FCRD-UFD-2016-000063, Figure 6-1), whereas the maximum allowed burnup for the Zircaloy-2 cladding in the TN-68 Dry Storage Cask is 60 GWd/MTU as stated in the CoC No. 1027 Amendment 1 Technical Specifications (ML073050262).

3. A review of data in NUREG-2214 Section 3.6.1 indicates that the threshold stress for Zircaloy-2 is reported to be as low as 70 megapascals (MPa) which is substantially lower than the value of 90 MPa reported for the other Zirconium alloys used in fuel cladding (NRC 2019).

4. The DOE High Burnup Dry Storage Cask Research and Development Project uses a TN-32 Dry Storage Cask. The maximum measured temperature during loading and cask placement for the DOE High Burnup Dry Storage Cask Research and Development Project was 237°C [459°F] (Hanson 2018). The maximum fuel cladding temperature for the TN-68 dry storage cask is 343°C [649°F] as stated in Table 3-3 of the TN-68 renewal application (ML20100F295).

This information is needed to determined compliance with 10 CFR 72.240(c)(3).

References Geelhood, K., and C. Beyer, Hydrogen Pickup Models for Zircaloy-2, Zircaloy-4, M5' And ZIRLO', Paper T2-011, 2011 Water Reactor Fuel Performance Meeting, Chengdu, China, Sept. 11-14, 2011 (ML12093A469).

Hanson, B.D., S.C. Marschman, M.C Billone, J. Scaglione, K.B. Sorenson, S.J. Saltzstein, High Burnup Spent Fuel Data Project Sister Rod Test Plan Overview, FCRD-UFD-2016-0000632016, PNNL-25374, April 29, 2016.

Hanson, B., High Burnup Spent Fuel Data Project & Thermal Modeling and Analysis, Presentation at the NWTRB Meeting, Albuquerque, NM October 24, 2018.

U.S. Nuclear Regulatory Commission, NUREG-2214, Managing Aging Processes in Storage (MAPS) Report, July 2019 (ML19214A111).

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RSI and Response Enclosure 1 to E-57065 Response to RSI-M1:

The High Burnup Spent Fuel Data Project and the testing of sister rods are focused on bridging the technical data gaps associated with the cladding for high burnup (HBU) pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies. This project aims to be representative of all HBU PWR and BWR fuel assemblies above 45 GWd/MTU. For this reason, the HBU spent nuclear fuel (SNF) in the Research Project Cask has been chosen to be representative of typical HBU fuel assemblies, with M5 cladding alloy expected to cover all recrystallized cladding such as Zy-2. Meanwhile, the sister rods testing has not only been defined to provide an initial picture of the fuel assemblies studied in the Research Cask project but also to study individual specific rods with peak conditions such as peak burnup rods and/or high peak cladding temperature. Nevertheless, according to the High Burnup Spent Fuel Data Project Sister Rod Test Plan Overview [1], the burnups in this High Burnup Spent Fuel Data Project are fully representative of fuel not only discharged to date, but of expected future discharges as well. Moreover, the peak cladding temperature, as well as the assembly-average burnup mentioned in the TN-68 CoC, are conservatively higher than those actually expected or measured in the HBU demonstration program.

Item 1:

Hydride reorientation embrittlement threshold: The Electric Power Research Institute (EPRI) report [2], based on studies such as [3] and [4], shows that the 70 MPa threshold mentioned by Kamimura [5] is not applicable for Zy-2 with liner when a low cooling rate is applied. Indeed in that case, hydrogen in solution has more time to diffuse into the pure Zr inner liner (due to the terminal solid solubility precipitation (TSSp) difference between Zr and Zy-2), leading to large hydride depleted zones and little radial hydride reorientation. This statement is consistent with the SPAR II report from IAEA [6]. Moreover, the burnups of some irradiated fuel claddings under investigation can exceed the 55.5 or even 60 GWd/MTU considered in the application. In the study from the CEA [3], Westinghouse Sweden supplied the LK3 cladding material, which was irradiated in the Leibstadt BWR for seven cycles up to 68 GWd/tU (local burn-up). Thus, Zy-2 cladding with liner would be immune to hydride reorientation embrittlement during dry storage and is expected to retain, or even increase ductility (due to irradiation defects annealing).

Most of BWR HBU fuel claddings are expected to have an inner liner, since according to [7], this technology was introduced in the 1980s and most of HBU fuel assemblies have been generated after this date. The ORNL report [8] does discuss the introduction of liners or barrier fuel into the GE design. The report also states that barrier fuel was offered commercially by GE beginning in 1983. By 1986, GE expected that 100% of the fuel produced at their Wilmington facility would be barrier fuel (See page 27 of 165 in [8]). The general licensee also confirmed that all HBU fuel assemblies have a liner in the fuel cladding.

Item 2:

High burnup range: The maximum allowed burnup for the Zy-2 cladding as reported in the Technical Specifications for Certificate of Compliance (CoC) No. 1027 is 60 GWd/MTU.

However, according to the DOE test plan overview [1], no HBU BWR fuel assembly from the United States would be expected to see an assembly-average burnup above the range of 55 GWd/MTU.

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RSI and Response Enclosure 1 to E-57065 Moreover, as stated in the sister rod testing plan, peak rod-average burnups are typically 5 to 9 percent higher than the assembly-average burnup (Geelhood and Beyer 2013). Where possible, a sister rod representing the peak burnup of its donor assembly was chosen for this testing program, so it is expected that individual rod-average burnups as high as 59 GWd/MTU will be tested. Thus, other than for lead test assemblies (LTA), the burnups in this High Burnup Spent Fuel Data Project are fully representative of fuel not only discharged to date but of expected future discharges as well.

Item 3:

Hoop stress threshold: End of Life rod internal pressures in BWR fuel cladding (Zy-2) are expected to be lower than for PWR (M5) leading to hoop stresses in the range of 40 MPa as reported in NUREG 2214 [9] (based on Raynauds work [10]). This hoop stress is lower than the 70 MPa threshold mentioned above where a loss of ductility is induced, especially when combined with peak clad temperature around 300 °C. Note that in the study from Kamimura [5],

a threshold stress of 40 MPa is mentioned but deals only with initiation of hydride reorientation on lined Zy-2 without inducing loss of ductility and it becomes not applicable for slow cooling rates associated with dry storage (as explained above).

Moreover, some margin can be added to the demonstration, since the hoop stress value reported at 40 MPa in [9] and [10] has been evaluated with a cladding temperature of 400 °C.

With a more realistic peak cladding temperature (PCT) combined with the lower temperature of the gas in the plenum, a lower hoop stress value could be derived using the ideal gas ratio between the pressure and the temperature.

Consequently, even for Zy-2 cladding without liner, it is expected that the fuel assemblies are not affected by hydride reorientation embrittlement. The HBU demonstration program is aimed to provide further arguments and demonstration of this statement.

Item 4:

PCT range: Regarding the PCT, the maximum measured thermocouple temperature on the HBU project was 237 °C, which corresponds to a PCT of about 240 °C as provided in Section 6.2.1.3 of [11]. The temperature limit on the radial neutron shield of the TN-32 cask limited the High Burnup Demonstration Cask Project from being able to further increase the heat load and cladding temperatures for the Research Cask. It must be noted as a fact, however, that for the High Burnup Research Project Cask, the calculated PCT for the license application was originally 348 °C as provided in Section 6.2.1.3 of [11]. So for the TN-68 license renewal application, the calculated PCT of 343 °C is comparable to the original design basis calculation for the HBU demonstration of 348 °C and the actual PCT (if measured) in the TN-68 would be expected to be on the order of 240 °C, similar to the HBU demonstration. Note also that the sister rods testing will simulate higher peak cladding temperatures reached by some fuel claddings, to force hydride reorientation and demonstrate the fuel cladding performance.

In conclusion, the BWR Zy-2 cladding is not expected to be sensitive to hydride reorientation embrittlement, either through the inner liner immunity or due to the low hoop stress on BWR. As a final demonstration of this statement, the DOE High Burnup Dry Storage Cask Research and Development Project (including the sister rods testing) can be used as a surrogate for the HBU Zy-2 clad fuel in the TN-68 Dry Storage Cask.

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RSI and Response Enclosure 1 to E-57065

References:

1. HighBurnup Spent Fuel Data Project Sister Rod Test Plan Overview- Hanson, B.D- FCRD-UFD-2016-000063, PNNL-25374, April 29, 2016

2. EPRI TR 3002016033, Effect of Hydride Reorientation in Spent Fuel CladdingStatus from Twenty Years of Research, Final Report, June 2020

3. Hydride Reorientation and its Impact on Ambient Temperature Mechanical Properties of High Burn-Up Irradiated and Unirradiated Recrystallized Zircaloy-2 Nuclear Fuel Cladding with an Inner Liner- Q. Auzoux- Journal of Nuclear Materials 494 (2017) 114e126

4. Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage- Masaki Aomi- Journal of ASTM International, Vol. 5, No. 9

5. K. Kamimura (2010 November), Integrity Criteria of Spent Fuel for Dry Storage in Japan at International Seminar on Spent Fuel Storage - ISSF 2010

6. IAEA-TECDOC-1680, Spent Fuel Performance Assessment and Research: Final Report of a Coordinated- Research Project (SPAR-II)

7. C.D. Williams, M.O. Marlowe, R.B Adamson, S.B., Wisner, R.A. Band, J.S. Armijo (1996),

Zircoloy-2 Lined Zirconium Barrier Fuel Cladding in Zirconium in the Nuclear Industry- 11th International Symposium

8. ORNL / TM-10902, Physical Characteristics of GE BWR Fuel Assemblies - R. S. Moore, June 1989

9. NUREG-2214 Managing Aging Processes in Storage (MAPS) Report, U.S. Nuclear Regulatory Commission, July 2019, ADAMS No. ML19214A111

10. Cladding Stress During Extended Storage of High Burnup Spent Nuclear Fuel- Patrick A.C.

Raynaud- Journal of Nuclear Materials 464 (2015) 304-312, ADAMS No. ML15180A411

11. EPRI TR 3002015076, High Burnup Dry Storage Research Project Cask Loading and Initial Results, Final Report, October 2019 Impact:

No change as a result of this RSI.

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