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Dry Cask 101: Making Sure They’ll Hold Up

Steven Everard
Structural Engineer

CASK_101finalEvaluating the structure of a spent fuel storage cask is a key part of our licensing process. In its application, the cask designer must provide an evaluation that shows the system will be strong and stable enough to resist loads that may be placed on it. NRC structural and materials engineers scrutinize this evaluation to make sure the design meets our regulatory requirements.

In an application, casks designers must provide evidence the cask system will:

  • Maintain confinement of the spent nuclear fuel
  • Maintain the fuel in a subcritical condition
  • Provide radiation shielding
  • Maintain the ability to retrieve or recover the fuel if necessary

In our structural review, we make sure the system can perform those functions even after experiencing a load, such as if the cask were dropped. We look at the structural design and analysis of the system under all credible loads for normal conditions—that is, planned operations and environmental conditions that can be expected to occur often during storage.

We also look at off-normal conditions, accidents and natural phenomenon events. “Off-normal” describes the maximum conditions that can be expected to from time-to-time, but not regularly. An example is the highest pushing or pulling force on a horizontal canister when it is being placed inside the storage overpack. Accident conditions and natural phenomenon include a dropped cask, earthquakes, tornadoes, flooding and any other credible accident or environmental condition that could affect the structural integrity of the system. These requirements are outlined in 10 CFR Part 72.

The structural review looks at whether the cask designer evaluated the proper loading conditions. It will also ensure the designer evaluated the system’s response to those loads accurately and completely. The reviewers must verify whether the resulting stresses in the material meet the acceptance criteria in the appropriate code.

As we explained in an earlier post, codes and standards are guidelines typically developed over many years of experience and through industry-wide and government agreement. Some of the more common codes an applicant may use come from the American Society of Mechanical Engineers, the American Society of Civil Engineers, the American Concrete Institute, the American Institute of Steel Construction and the American Welding Society.

Not all loads are likely to occur at one time, but some might occur together. So we look at several different combinations of loads that can be expected at the same time. These include dead loads (which come just from the weight of the material), live loads, (which come from the movement of the system or people and things near it), and environmental loads (including snow, ice, wind, temperature and seismic). For example, the cask could experience a dead load, live load, snow load and wind load together. But it is not reasonable to expect the cask to be in a snow storm, a tornado and an earthquake at the same time.

These cases are analyzed to determine the stresses placed on the material used to construct the cask system. This analysis may be completed by either hand calculations or by a computer model. Typically, we only look at the maximum stresses in the different materials—since lesser stresses would not be as challenging to the system.

The maximum stresses from the analysis are compared to the allowable stresses from the appropriate code to determine a margin of safety. These design margins are typically large. This is because designs must be robust enough to withstand the accident scenarios. To be conservative, we and the designers overestimate loads and underestimate material strength. Doing this adds conservatisms and enhances our assurance that the design is adequate.

12 responses to “Dry Cask 101: Making Sure They’ll Hold Up

  1. Anonymous April 15, 2016 at 10:56 am

    If it is structural stability is the main objective that the agency is trying to demonstrate and reassuring, staff ought to come out candid in accepting or denying if Mineral VA earth quake did not do anything, except shifting the position of the cask. Just a thought.

  2. Anonymous April 9, 2016 at 1:47 pm

    Has the Staff (NRC) reach a conclusion about the embrittlement of fuel cladding during the drying process of putting fuel into the dry cask?

    • Moderator April 13, 2016 at 1:44 pm

      Interim Staff Guidance (ISG)-11 and NURGEG-1536 provide guidance on this subject. During loading operations for all spent fuel, the maximum fuel cladding temperature should not exceed 400 degrees Celsius and the number of thermal cycles (repeated heatup/and cooldown) should be limited to 10. A thermal cycle is defined as a temperature change of over 65 degrees Celsius. If these parameters are met, the NRC staff believes that this will provide reasonable assurance that embrittlement will not prevent the spent fuel from being safely stored in the configuration analyzed in the application. The NRC currently is sponsoring confirmatory research on this phenomenon at Oak Ridge National Laboratory.

      Steve Everard

      • Anonymous April 13, 2016 at 3:19 pm

        When will the confirmatory research be completed and will it be available to the public or will it be treated as propriatery and does it address all cladding materials (i.e., Zirc-2, Zirc-4, ZIRLO, M5, MDA, etc.)?

    • Anonymous April 14, 2016 at 6:58 am

      The response given has been the Staff’s long standing position; however, as of 2010, articles published at conferences and additional testing (confirmatory) done by the staff overseas has shown that the position established in ISG-11 and NUREG-1536 may not be conservative and may need to be revisited. The fundamental issue is not one of safety, but one of compliance since fuel placed in a dry cask in one state, may not be the same state the fuel is in, if the cask were reopened in the future for removal purposes.

      This fundamental understanding of mechanisms may also impact operational burn-up limits besides the simple compliance issues of what occurs to the fuel if embrittled during the drying process.

  3. Anonymous April 8, 2016 at 4:43 pm

    From the NRC Moderator:
    Please use the Open Forum post for any further discussion about San Onofre steam generators or anything else unrelated to the topic of this blog post.

  4. drgenenelson April 8, 2016 at 1:50 am

    I hope that a future installment of “Dry Cask 101” will help readers to understand the robust construction of Independent Spent Fuel Storage Installation (ISFSI) such as the one at Diablo Canyon Power Plant (DCPP.) I hope that you will describe the robustness of the completed Holtec International HI-STORM storage casks bolted with massive bolts to an 8 foot thick pad of concrete located directly underneath the 500 kV power transmission lines leading from the facility, among other attributes of the DCPP ISFSI

  5. Gary Headrick April 7, 2016 at 5:39 pm

    Was the NRC being conservative when they approved the computer modeling done on the steam generators that failed at San Onofre? What if you are wrong again when it comes to safe storage containers? Do you have a realistic plan to implement if canisters begin leaking before they can be relocated, or do you excuse yourself by claiming it is too unlikely to take into consideration? These are serious questions that need to be answered. We can’t afford to make mistakes with so much at stake. Please don’t assume everything will go as planned. We know from several examples, especially from this industry, that they don’t.

    • drgenenelson April 8, 2016 at 11:13 am

      Creative fearmongering to attempt to create an equivalence between a system designed to move massive amounts of energy between a primary and secondary loop of a PWR – and the dry cask storage system. Dry casks need to passively dissipate a small amount of heat energy while maintaining structural integrity.

      Post-closure design reviews of the SONGS replacement steam generators indicate that the design changes implemented by the SONGS owners in conjunction with Mitsubishi Heavy Industries should have triggered the additional oversight of the License Amendment Review (LAR) process. For example, see;”NRC Oversight of Licensee’s Use of 10 CFR 50.59 Process To Replace SONGS’ Steam Generators (OIG Case NO. 13-006)” – the October 2, 2014 report from the NRC Office of the Inspector General.

      • CaptD April 8, 2016 at 3:03 pm

        DrG — You comment is off-base since the NRC Region IV AIT report, both the SCE and the MHI Root Cause reports had serious flaws since they listed the operating conditions of both Unit 2 and Unit 3 as being the same when intact they were quite different (which is why Unit 3 RSG’s suffered IPFEI and Unit 2 RSG’s did not). All 4 almost new RSGs were damaged because of their flawed unsafe design, yet the NRC never fined SCE, because to do so would have also implicated NRC Region IV for their 50.59 approval, when it is now obvious that they did not have a clue what they were approving. This is why the NRC IG’s report was just yet another coverup that has made #SanOnofreGate the multi-billion $ SCE-CPUC ripoff that it is. The above decisions are now being challenge in the NRC system and hopefully soon the nuclear industry will get a chance to learn what the correct Root Cause was, not the CYA Root Cause that SCE submitted to hide its engineering debacle.

        You want more proof, read the Independent Consultants Report (done by Beckman Associates for the Chairman of the NRC (which was submitted to the NRC about a week before the Region IV AIT report was made public). It was highly critical for the above name reports and for that reason (IMO) it was not released to the general public until AFTER the decommissioning of San Onofre was announced by SCE.

        It is important to note that SCE’s operators never had any idea that all four of the new RSGs were destroying themselves, until the Unit 3 RSG started leaking! This placed all of southern California at risk because RSG tube failure(s) can result in leakage of radioactive core coolant into the environment or much worse, even a meltdown of the reactor core itself, if there is a cascade of tube failures. Animation of RSG tube failures.

        The discovery of the extent of the RSG damage was made during the required inspections after the radioactive leakage. Comparatively, even though SCE operated the Unit 2 RSGs within their functional/testing limitations, they only lasted about twice as long as the Unit 3 RSG’s due to their design! The NRC said that the damage to all 4 of the RSGs was “unprecedented,” since the damage to the Unit 2 and Unit 3 RSG’s had more damage that all the rest of the US nuclear fleet combined! http://sanonofresafety.files.wordpress.com/2011/11/steamgeneratortubesplugged1.pdf

  6. Anonymous April 7, 2016 at 4:14 pm

    It’s actually the American Concrete Institute, not the American Concrete Association (about 5th paragraph down).

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