Machiels, Wells

Electric Power Research Institute, Inc.,

Palo Alto, California, United States of America

Abstract. “Burnup credit” refers to taking credit for the burnup of nuclear fuel in the performance of criticality safety analyses. Historically, criticality safety analyses for transport of spent nuclear fuel have assumed the fuel to be unirradiated (i.e. “fresh” fuel). In 1999, the U.S. Nuclear Regulatory Commission (NRC) Spent Fuel Project Office issued Interim Staff Guidance – 8 (ISG-8) with recommendations for the use of burnup credit in storage and transportation of pressurized water reactor (PWR) spent fuel. The use of burnup credit offers an opportunity to reduce the number of spent nuclear fuel shipments by ~30%. A simple analysis shows that the increased risk of a criticality event associated with properly using burnup credit is negligible. Comparing this negligible risk component with the reduction in common transport risks due to the reduced number of spent fuel shipments (higher capacity casks for transporting PWR spent fuel) leads to the conclusion that using “burnup credit” is preferable to using the “fresh fuel” assumption. A specific objective of the EPRI program is to support the Goals of the U.S. Industry. These goals are consistent with the original U.S. Department of Energy (DOE) goal defined in 1988: a burnup credit methodology that takes credit for the negative reactivity that is practical (all fissile actinides, most neutron absorbing actinides, and a subset of the fission products that account for the majority of the available credit from all fission products). The determination of the optimum number of fission products to consider in a practical burnup credit methodology validates the approach advocated by researchers from France to first focus on a handful of isotopes that include Sm-149; Rh-103; Nd-143; Gd-155; and Sm-152.

BURNUP CREDIT – U.S. INDUSTRY GOALS

The development of a practical burnup credit methodology must be focused on achievable goals that add value to the use of burnup credit. An NEI/EPRI meeting identified the following goals for a useful commercial burnup credit approach:

(1) Be practical, based on the U.S. Nuclear Regulatory Commission’s Interim Staff Guidance 8 (ISG-8).

(2) Support initial enrichments up to 5.0 wt% U-235 without a loading offset.

(3) Include burnable absorbers.

(4) Provide standard axial burnup profiles.

(5) Allow various cooling times.

(6) Include the addition of fission products.

(7) Allow for extension to high burnup (> 40GWd/MTU).

(8) Allow burnup measurements to be replaced by reactor records.

(9) Provide for BWR applications.

(10) Be based upon standard (consensus) parameters (11) Provide standard benchmark methodologies (12) Support standard isotopic depletion codes

The first goal is simply that a burnup credit methodology must be practical to be of value to the U.S. commercial nuclear industry. The burnup credit methodology should be simple enough that existing reactor records would allow the determination whether a specific assembly could be stored or transported in a particular cask. Extensive data collection efforts, such as performing cask criticality calculations for each discharged assembly using isotopic data obtained from core-follow calculations, would not be desirable. A go/no-go decision for each assembly should be possible based upon existing reactor records.

The second through fifth goals are to add flexibility to the existing U.S. NRC guidance (Interim Staff Guidance 8, Revision 1, or ISG-8, Rev. 1) [Nuclear Regulatory Commission, Spent Fuel Project Office, Interim Staff Guidance, ISG - 8, Revision 1, Issue: Burnup Credit

in the Criticality Safety Analyses of PWR Spent Fuel in Transport and Storage Casks,"

August 8, 1999]. These have been addressed by an NRC-Research-sponsored PIRT (Phenomena Importance Ranking Table) Panel. If the NRC implements the suggestions of the PIRT Panel, Goals 2 through 5 will have been accomplished.

The sixth goal, the addition of fission products, is key to the value of burnup credit, and is discussed in some detail later in this paper. One important question is just how many fission products are needed.

The seventh goal, higher burnups, is independent from the desire to add fission products, but would likely be effectively addressed by considering fission products, since both the sixth and seventh goals depend upon the extension of current available data.

The eighth goal relates to the NRC’s desire for a check of the burnup of each assembly, which is somewhat independent of reactor records.

The ninth goal is the foreseeable need for burnup credit for BWR plants. ISG-8, Rev. 1 does not currently permit the use of burnup credit for BWRs.

Goals 10, 11, and 12 are factors important to the practical implementation of burnup credit.

Standard parameters and methodologies, and standard benchmark validation, would streamline the use of burnup credit, and also simplify the NRC review process. These goals may be considered as “cookbooks” for burnup credit calculations for spent fuel casks. For example, some of the parameters required for the burnup credit process are obscure, such as the UO2 pellet temperature; these parameters are not easily available in existing utility or cask vendor documents. The collection of such parameters and determination of conservative, but reasonable, values would certainly simplify the burnup credit calculations and the NRC review.

Fission products — Determination of optimum number of fission products for a practical commercial burnup credit methodology

The sixth U.S. Industry Goal, the addition of fission products, is a most challenging technical issue because extensive confirmation of our knowledge of nuclear data is required. Nuclear data of two types are required for burnup credit calculations: (i) isotopic build-up and decay data, and (2) nuclear cross-section data. The first type of nuclear data facilitates the calculation of the isotopic concentrations of actinide and fission products in spent fuel. The second type of data allows the reactivity (k-effective) of spent fuel to be determined. This is shown graphically in Figure 1.

At the ICONE 9 meeting, Toubon et al. [Toubon et al. “Current Applications of Actinide-Only Burnup Credit within the COGEMA Group and R&D Programme to Take Fission Products into Account”, ICONE 9, Nice, France, April 2001] presented two examples of the potential benefits of making allowance for fission products in criticality studies. For this study, similar types of calculations were performed using on one hand the Yucca Mountain repository project list of 16 fission products that are reasonably stable over geological time scales, and on the other hand five (Sm-149; Rh-103; Nd-143; Gd-155; and Sm-152) of the six fission products selected by Toubon et al. The Cs-133 isotope was not included in this study because of potential volatility considerations.

FIG. 1. Burnup credit validation.

The Yucca Mountain approach is very complete and includes all isotopes (except for Cs-133) that make a useful contribution to criticality safety, but the large number of isotopes adds to the level of complexity. The effectiveness of fission product burnup credit was compared, using both the “Yucca Mountain” and “French” (minus Cs-133) approaches, to the actinide-only approach. The commercial VSC-24 storage cask burnup credit loading curve was selected as a means of providing a perspective of what an “acceptable” loading curve might look like.

Calculational method

The Yucca Mountain project had performed a series of studies of the burnup credit loading curves for PWR and BWR fuel assemblies. The PWR studies employed a generalized model of the Crystal River 3 B&W fuel assemblies, and included pre-closure calculations that are analogous to the calculations performed for storage and transport, i.e.the fuel assemblies remain intact. The axial burnup profiles were based upon reactor vendor core following calculations, and had 18 axial nodes. The SCALE 4.3 computer code system SAS2H module was used to obtain the isotopic contents of irradiated fuel for a range of initial enrichments up to 6.0 weight percent U-235 and for burnups up to 60 GWd/MTU. The isotopic data files for the Yucca Mountain loading curve study were extracted and used to prepare a series of input files for the MCNP4 computer code. The spent fuel assemblies were modeled in the VSC-24 cask to allow comparison to a previously existing burnup credit loading curve.

The process of generating a loading curve involves calculation of the required minimum burnup to produce a k-effective of 0.95 for various initial enrichments. The required minimum burnup, expressed as a function of initial enrichment, is the burnup credit loading curve.

Initial calculations were performed for 4.0 wt% U-235 initial, and are shown in Figure 2. The cask basket for these calculations was the actual VSC-24 Multi-Element Sealed Basket (MSB) loaded into the MSB Transfer Cask (MTC). The basket for the VSC-24 consists of 24 square ferritic steel tubes with a 0.200-inch wall thickness. There is no boron neutron absorber in this basket. The k-effective is plotted as a function of burnup in order to determine what burnup provides a k-effective of 0.95, including bias and uncertainty (a bias of 0.02 is assumed for irradiated fuel calculations). Such curves are used to obtain a single data point for the burnup credit loading curves. In Figure 2, the k-effectives are plotted for three different fission product sets: “Actinide-Only” (no fission products), “French Isotopics” (five fission products), and “Principal Isotopes” (Yucca Mountain, 16 fission products). Note that the curves for both the Principal Isotopes and French Isotopics are substantially different from the Actinide-Only curve. The Principal Isotopes and French Isotopics curves require burnups of 44.8 and 49.9 GWd/MTU respectively, while the Actinide-Only curve requires as much as 62 GWd/MTU. The required burnup difference between the Yucca Mountain and French fission product sets is 5 GWd/MTU, which is substantial, but still much less than the 16 GWd/MTU difference between the Yucca Mountain and Actinide-Only burnup requirements. Thus, Figure 2 shows that the French Isotopics list (minus Cs-133) is less effective than the Yucca Mountain isotopic, but much better than Actinide-Only. Figure 2 also shows that the required minimum burnups for all three isotopic selections are excessive, since a 4.0 wt% U-235 initial fuel assembly will typically achieve a burnup of approximately 40 GWd/MTU. The logical conclusion is that some additional means of criticality control is needed, and this is why all large, multi-assembly transport cask baskets contain a supplementary neutron absorber built into the basket structure.

Required Minimum Burnup, 4.0 wt%, No BPRAs

49.9 61.8

44.8

0.750 0.800 0.850 0.900 0.950 1.000 1.050 1.100 1.150 1.200 1.250

0 10 20 30 40 50 60 70

Burnup, GWd/MTU

k-effective

Yucca Mtn Isotopics French Isotopics Actinide-Only

FIG. 2. Comparison for VSC-24 basket.

To investigate the effect of fission product selections in a “poisoned” SNF basket, the square ferritic steel fuel tubes of the VSC-24 basket were changed to stainless steel tubes containing boron. This approach was studied in EPRI TR-102001, “Evaluation of the Transportability of the VSC Basket”. The particular boron/stainless alloy chosen for this new basket was SS-316 B6A, an alloy of stainless steel 316 with 1.6 weight percent of natural boron. Actual basket designs may use a variety of neutron absorber materials, including aluminum/boron carbide composites, boron/aluminum alloys, and stainless steel 304 alloys with up to 2.0 wt% boron.

The SS-316 B6A alloy was chosen partly because it was used in conceptual designs for PWR and BWR Waste Packages for Yucca Mountain, and partly because the effective boron areal density is 0.023 g/cm2 B-10. The NRC has allowed credit for only 75 percent of the actual boron content, and the 0.023 g/cm2 B-10 value would require a total boron content of 0.30 g/cm2 B-10. This boron content places the conceptual basket into the middle of the typical PWR neutron absorber content range for transport baskets.

The k-effective calculations were repeated for the Yucca Mountain Principal Isotopes, the French Isotopics, and the Actinide-Only fission product selections. The results of these calculations are shown in Figure 3. Inspection of this figure shows that, as before, the French Isotopics list is closer to the Yucca Mountain fission product set than to the Actinide-Only case. The effect of the addition of boron to the basket is manifested in the expected reduction in the amounts of burnup required for all three fission product sets, and also in a reduction in the spacing between the three curves. The required minimum burnup for 4.0 wt% U-235 initial PWR fuel is less than 30 GWd/MTU for all three fission product lists, which is a much more reasonable burnup requirement. The difference between the Yucca Mountain and French fission product lists is only 1.7 GWd/MTU, one third of the 5 GWd/MTU difference obtained for the non-borated basket calculations. The difference between the French fission product list and Actinide-Only is still a substantial 6.5 GWd/MTU, showing the benefit of the inclusion of fission products. In other words, we obtain most of the benefit of fission products with the simpler French fission product list.

Required Minimum Burnup, 4.0 wt%, No BPRAs

21.9 28.4

Waste Package Yucca Mtn Isotopics French Isotopics Actinide Only

FIG. 3. Comparison for “transportable, poisoned” VSC-24 basket.

These calculations demonstrated that burnup credit alone is not enough for criticality control in a closely packed 24-assembly PWR basket, and that a neutron absorber such as boron must be added to the basket structure to obtain a practical burnup credit loading curve. The calculations for 4.0 wt% U-235 initial showed that the French fission product list (without Cs-133) achieves much of the benefit of the more complex 16-fission product list.

A comparison of the complete loading curves for enrichments up to 5.0 wt% U-235 initial is desirable, so calculations like those shown in Figure 3 were repeated for enrichments of 2.0, 3.0, and 5.0 wt%. In addition, the ability of the “poisoned” VSC-24 basket to accommodate fresh, unirradiated fuel was determined. The results of these calculations are shown in Figure 4. For comparison purposes, the loading curves for the VSC-24 Technical Specifications are also shown in Figure 4, for Westinghouse 14x14 PWR assemblies and for Combustion Engineering 15x15 assemblies. Although the VSC-24 Tech Spec loading curve was not derived with modern burnup credit assumptions (credit was taken for Xenon and Samarium), it has successfully been used by several utilities without undue restriction on the assemblies which can be loaded. An unofficial estimate is that 75 percent or so of the fuel inventory at PWRs could be stored in the VSC-24 based upon the burnup credit loading curve.

Inspection of Figure 4 shows the comparison of loading curves obtained with different fission product isotopic selections. The French Isotopic set (minus Cs-133) is nearly as effective as the 16-fission product set of the Yucca Mountain Principal Isotopes. The Actinide-Only case is not as effective as the French or Yucca Mountain fission product selections. The VSC-24 Tech Spec loading curves are less effective, but it must be recalled that this basket does not contain a neutron absorber and is hence much less expensive to fabricate.

Comparison of Yucca Mtn versus French F.P.

Isotopics

0.030 g/cm2 B-10 in Basket (VSC-24 has no boron⁾

ACCEPTABLE

Fresh Fuel Enrichm ent, w t%

Required Minimum Burnup, GWd/MTU

FIG. 4. Comparison of fission product isotopic selections.

In summary, there are three basket criticality solutions based on hardware: no poison basket cells, poison basket cells, and poison with flux trap basket cells. Adding a neutron absorber like boron adds to cost, but adding a flux trap (a neutron absorber/water gap sandwich affair) adds to cost and reduces payload. By itself, burnup credit alone cannot accommodate the spent fuel inventory. With a neutron absorber built into the basket, burnup credit can accommodate most of the fuel inventory (Figure 5). Actinide-only burnup credit captures many of the fuel assemblies, but misses some under-burned higher enrichment fuels. The simpler French Fission Product isotopic approach captures almost all of the assemblies accommodated by the Yucca Mountain isotopic selection.