ML20079J069
| ML20079J069 | |
| Person / Time | |
|---|---|
| Issue date: | 03/19/2020 |
| From: | Office of Nuclear Material Safety and Safeguards |
| To: | |
| Kallan P | |
| References | |
| Download: ML20079J069 (3) | |
Text
Road Map on Source Terms (Recommendations III-1, III-2 and III-3)
DRAFT To establish a road map for the RIRP on source terms, a proposed list of actions is presented below, followed by additional background information.
- 1. Proposed List of Actions for the RIRP on the source terms:
- Compile a complete list of principal inputs parameters to the source term calculations
- For each parameter, establish the following o Typical values / approaches used in the design basis calculations (FSAR) o Typical actual (real assembly / site specific) values / approaches o Sensitivity of source terms to the parameter This has to be done separately for the difference source terms, e.g. decay heat; neutron source; gamma source; Co-60 source.
This should be based as far as possible on existing public studies, e.g. NUREG/CR reports from ORNL o (estimated) uncertainty of source term
- Review of any relevant benchmarking of the standard computer code used for developing source terms (SCALE SAS2H/TRITON/ORIGENS etc.)
- Evaluate the effect of the number of assemblies per cask, and the location/distribution of assemblies within the cask o For example, can uncertainties on source term calculation results be considered statistically, or is there a strong case that correlations are important?
- Evaluate conservatisms and uncertainties already considered in the analyses that use the source terms.
o For example, there is no point determining a source term value within 0.1% if the calculation that uses the value has an uncertainty of 3% and that is considered acceptable.
- Finally, all this information and the evaluations should be combined, and recommendations should be compiled into one or more tables that show the list of parameters, and for each parameter a recommendation, such as use design basis value without any added uncertainty; or use actual value but add uncertainty of xx%. Different Tables may be developed to address the different recommendations III-1 through III-3.
- 2. Background Information The discussions below provide additional background information, with a specific focus on the differences between the evaluations documented as part of the safety evaluations (FSAR), and those performed to support cask loading. Two types of source terms (thermal and radiological) are addressed separately.
- Decay heat This is used as the input to the thermal analyses, to calculate various temperatures of the systems.
In most cases, temperatures are compared to corresponding allowable temperature limits. This is used in a two-step fashion
o In the FSAR, decay heat values are essentially assumed (albeit informed by some source term calculations for assumed fuel), then the temperatures resulting from those decay heat values are calculated. They are then compared to the corresponding limits to show they are acceptable. Since the temperatures are acceptable, the initially assumed decay heat values are then defined as limits in the technical specifications. In other words, decay heat limits in the TS are defined as a substitute for specifying temperature limits in the TS. Note that in the context of the source term calculations in the FSAR, decay heat values may be presented or discussed, but they are not necessarily linked to the values assumed and used in the thermal calculations.
o When casks are loaded, it then needs to be demonstrated that the decay heat values of the fuel assemblies are below the TS limits, which then would assure that the temperatures will be below the temperature limits.
o The advantage of this approach is that the thermal evaluations (calculating temperatures based on decay heat) can be completely documented in the FSAR, and no further thermal calculations need to be performed when a cask is loaded, or to show that a loaded cask meets the applicable temperature limits.
o The disadvantage, in addition to the somewhat backwards two step process, is that FSAR does not necessarily need to specify the approach and the details for calculating the decay heat values. This presents some uncertainty to the user (the entity that loads the cask).
Depending on the design and analysis philosophy of this entity, it may feel that it needs to add additional margin or uncertainties, not intended in the initial calculations. At a minimum, it may add margin and uncertainty to make sure that under no circumstances it will be above the decay heat value limits specified in the TS, so it will never be in the situation of potentially violating a TS limit.
- Radiological source terms o These include fuel neutron source strength, fuel photon (gamma) source strength, and the source strength from Co-60 in any of the structural materials of the fuel, predominantly in the steel and maybe in Inconel.
o For these, the entire calculation from the fuel characterization to calculated dose rates is documents in the FSAR, including inputs and details of the calculational methodology.
o A complication in this area results from a different aspect, namely that the FSAR cannot show compliance with dose or dose rate limits by comparing calculated dose rates with such limits. However, this is a subject that will be addressed in a separate workshop, and hence is not further discussed here. Hence overall, the calculations in the FSAR encompass the following steps Select fuel characteristics (assembly type, burnup, initial enrichment, cooling time, uranium weight, cobalt content of structural materials). Some of these parameters are specified in the TS.
Select operating parameters (power densities, temperatures, soluble boron concentration for PWR, void fractions for BWR, cycle and outage lengths, flux fraction for structural materials). Conservative or bounding assumptions are typically made for these parameters, and they are not specified in the TS.
Perform the source term calculations, resulting in the neutron, gamma and Co-60 source strengths for that fuel and operating conditions.
Use the source strengths in or with the radiation transport calculations to calculate doses and dose rates at the desired locations.
- Two options are available here, both of which result in the same dose rates.
The first option is to use the strengths directly as inputs in the radiation transport calculations. This way, the calculation directly determines doses or dose rates for the given strengths. If dose rates are needed for several fuel characteristic sets, this means radiation transport calculations need to be repeated for each set, which can be rather computationally time consuming. The second option is therefore to perform radiation transport calculations for standard source strengths (e.g. 1 particle per second), and the subsequently multiply those results with the source strengths to determine dose rates. In this case, only the multiplication needs to be re-performed for subsequent fuel characteristics.
In the implementation, it is now necessary to show compliance with the regulatory limit, such as the site boundary annual dose limit, which is site specific. There are several option how that can be demonstrated
- In all cases, compliance with any fuel related TS limits need to be assured.
Such limits include, at a minimum, limits on burnup, enrichment and cooling time.
- The simplest approach would be to use the source strength from the conservative or bounding calculations documented in the FSAR for all fuel assemblies in all casks to be loaded. If the corresponding doses or dose rates still meet the regulatory requirements, no further calculations or refinements many be needed. In this case it is however important to clearly highlight that the resulting doses and dose rates are unrealistically high compared to realistically expected doses and dose rates, potentially by orders of magnitude, so that the values are not misinterpreted as expected values. If this approach does not result in satisfactory results, several levels of refinements are possible, with essentially all of them changing the inputs to the source term calculations that are used to generate the source strengths used in the dose or dose rate analyses. For that, the typical approach is for burnups, enrichments, cooling times, uranium mass, and core operation parameters to be adjusted and brought into alignment with the actual fuel to be loaded, to the extent necessary to meet the dose or dose rate limits.