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| number = ML20236U547 | | number = ML20236U547 | ||
| issue date = 07/24/1998 | | issue date = 07/24/1998 | ||
| title = Summary of 980721 Meeting W/Nrc & Licensee Re Nuclear Criticality Safety Validation Rept, | | title = Summary of 980721 Meeting W/Nrc & Licensee Re Nuclear Criticality Safety Validation Rept, ,submitted to Support Licensing of Avlis Technology & Responses to Staff Question.W/Attendee List | ||
| author name = Persinko D | | author name = Persinko D | ||
| author affiliation = NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) | | author affiliation = NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) | ||
| Line 16: | Line 16: | ||
=Text= | =Text= | ||
{{#Wiki_filter: | {{#Wiki_filter:i | ||
.Tuly 24, 1998 i | |||
i l | i l | ||
TO; | l TO; Robert C. Pierson, Chief Special Projects Branch, NMSS FROM: | ||
Drew Persinko, AVLIS Project Manager l | |||
Special Projects Branch, NMSS | Special Projects Branch, NMSS | ||
==SUBJECT:== | ==SUBJECT:== | ||
| Line 31: | Line 31: | ||
==References:== | ==References:== | ||
: 1) AVLIS Criticality Code Validation Report, USEC to NRC, April 22,1998 | : 1) AVLIS Criticality Code Validation Report, USEC to NRC, April 22,1998 | ||
: 2) Staff Request for AdditionalInformation, NRC to USEC, June 10,1998 On July 21,1998, NMSS/SPB staff met with staff from the U.S. Enrichment Corporation | : 2) Staff Request for AdditionalInformation, NRC to USEC, June 10,1998 l | ||
(USEC) and their contractor, B&W, to discuss the nuclear criticality safety validation report (Reference 1), submitted to support the licensing of the Atomic Vapor Laser Isotope Separation (AVLIS) technology, and responses to staff questions (Reference 2) on the report. USEC is requesting NRC to license the AVLIS facility up to a uranium enrichment of 10 percent. The nuclear criticality safety validation report was submitted to support this enrichment level. The | On July 21,1998, NMSS/SPB staff met with staff from the U.S. Enrichment Corporation (USEC) and their contractor, B&W, to discuss the nuclear criticality safety validation report (Reference 1), submitted to support the licensing of the Atomic Vapor Laser Isotope Separation (AVLIS) technology, and responses to staff questions (Reference 2) on the report. USEC is requesting NRC to license the AVLIS facility up to a uranium enrichment of 10 percent. The nuclear criticality safety validation report was submitted to support this enrichment level. The j | ||
staff will consider the information presented at the meeting and review the information provided j | |||
at the meeting and determine whether any additional information is necessary for the staff to write a Safety Evaluation Report. | |||
Discussion points addressing the staff's questions on the nuclear criticality safety validation report are provided as Attachment 1. Slides used in the meeting are provided as Attachment 2. | Discussion points addressing the staff's questions on the nuclear criticality safety validation report are provided as Attachment 1. Slides used in the meeting are provided as Attachment 2. | ||
A list of meeting attendees is provided as Attachment 3. | A list of meeting attendees is provided as Attachment 3. | ||
USEC intends to submit an AVLIS license application in February 1999; however, non-site-specific information is being submitted for staff review before the application is submitted. The criticality safety validation report and this moeting were part of the ongoing pre-license application activities. | USEC intends to submit an AVLIS license application in February 1999; however, non-site-specific information is being submitted for staff review before the application is submitted. The criticality safety validation report and this moeting were part of the ongoing pre-license application activities. | ||
Attachments: As stated cc: R. Woolley, USEC | Attachments: As stated cc: R. Woolley, USEC | ||
/ | |||
DISTRIBUTION: | |||
Docket 70-3089 | l Docket 70-3089 Central File PUBLIC NMSS R/F FCSS R/F SPB R/F OFC | ||
_SPB bmSih,[ | |||
S_P_B SM SPB bPersinko:ij | |||
/ M in NAME L | |||
ey 73f[98 1 M/98 | |||
/ /98 DATE 1'Nf98 8 | |||
C = COVER E =' COVER & ENCLOSURE N = NO COPY mtsumcrt.721 OFFICIAL RECORD COPY l | |||
9 | |||
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( | ( | ||
t | t ATTACHMENT 1 DISCUSSION POINTS ADDRESSING STAFF QUESTIONS a | ||
ATTACHMENT 1 | i i | ||
l | |||
Discussion Material on the NRC's " Request for Additional Information Letter Dated April 22,1998 AVLIS Enrichment Plant f | Discussion Material on the NRC's " Request for Additional Information Letter Dated April 22,1998 AVLIS Enrichment Plant f | ||
for July 21,1998 Meeting with the NRC | for July 21,1998 Meeting with the NRC | ||
) | |||
I i | I i | ||
I l | I l | ||
l 1-l | l 1 - | ||
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I 1 | I 1 | ||
I 1 | I 1 | ||
* The following clarifications and additional information are provided to assist the NRC f | * The following clarifications and additional information are provided to assist the NRC f | ||
in their review of the AVLIS SCALE 4.3 Criticality Code Validation Report. | in their review of the AVLIS SCALE 4.3 Criticality Code Validation Report. | ||
: 1. Comment: | : 1. Comment: | ||
Reference the recommendation of ANSI /ANS-81-1983: " Bias shall be established by correlating the results ofcriticality experiments with results obtained for these same systems by the method being evaluated" and the recommendation of ANSI /ANS-8.17-1984: "The | l l | ||
criticality experiments used as benchmarks . should have physical compositions, configurations, and nuclear characteristics (including reflectors) similar to those of the systems being calculated." Based on these guidelines, discuss how subcritical margins were determined for extending the range of applicability for 7% and 7.5% fuel rod array experiments to moist oxides powders and nitrate solutions in the intermediate energy spectrum. ANSI /ANS8.1-1983 allows for extending the area of applicability of a calculational method beyond the range of experimental conditions over which the bias is established. The standard states that "where the extension is large, the method should be supplemented by other calculational methods to provide a better estimate of the bias in the extended area (s)" and that "the margin of suberiticality shall include allowances . for uncertainties due to any extensions of the area of applicability." Two actions are thus required to follow this ANSI recommendation: 1) use of an alternative code and/or data, and | Reference the recommendation of ANSI /ANS-81-1983: " Bias shall be established by correlating the results ofcriticality experiments with results obtained for these same systems by the method being evaluated" and the recommendation of ANSI /ANS-8.17-1984: "The criticality experiments used as benchmarks. should have physical compositions, configurations, and nuclear characteristics (including reflectors) similar to those of the systems being calculated." Based on these guidelines, discuss how subcritical margins were determined for extending the range of applicability for 7% and 7.5% fuel rod array experiments to moist oxides powders and nitrate solutions in the intermediate energy spectrum. ANSI /ANS8.1-1983 allows for extending the area of applicability of a calculational method beyond the range of experimental conditions over which the bias is established. The standard states that "where the extension is large, the method should be supplemented by other calculational methods to provide a better estimate of the bias in the extended area (s)" and that "the margin of suberiticality shall include allowances. for uncertainties due to any extensions of the area of applicability." Two actions are thus required to follow this ANSI recommendation: 1) use of an alternative code and/or data, and | ||
: 2) allowing margin due to the extension. In order to apply the alternative code, adequate independence must be established. Discuss why this ANSI recommendation was not followed in the validation report. | : 2) allowing margin due to the extension. In order to apply the alternative code, adequate independence must be established. Discuss why this ANSI recommendation was not followed in the validation report. | ||
===Response=== | ===Response=== | ||
I In response to item 1, we have followed the ANSI /ANS8.1 and ANSI /ANS8.17 | I In response to item 1, we have followed the ANSI /ANS8.1 and ANSI /ANS8.17 1 | ||
recommendations regarding establishing bias and the range of applicability for the bias. | |||
AVLIS plant normal conditions, accident scenarios and materials, along with consideration of i | AVLIS plant normal conditions, accident scenarios and materials, along with consideration of i | ||
our usual bounding calculational approach were examined to determine the appropriate experimental benchmark set. Table 1 of the validation report indicates that the selection of benchmark experiments is suflicient to validate the code system for criticality calculations | our usual bounding calculational approach were examined to determine the appropriate experimental benchmark set. Table 1 of the validation report indicates that the selection of benchmark experiments is suflicient to validate the code system for criticality calculations | ||
supporting Nuclear Criticality Safety Analyses (NCSA's) of AVLIS plant designs. | } | ||
supporting Nuclear Criticality Safety Analyses (NCSA's) of AVLIS plant designs. | |||
I To more clearly illustrate this point, please refer to the attached Tables A and B, which categorize the benchmark experimental data set according to material types, degree of moderation, enrichment, and energy spectra. The tables show the number of experiments that fall into each range within a category. Table A identifies materials (other than the fissile 1 | |||
material) expected to be encountered in the AVLIS facility that will contribute to the nuclear I | |||
characteristics of a system being calculated. The totals demonstrate that the experimental data set provides a significant number of data points for each material. Likewise, Table B provides l | |||
information regarding the important characteristics of the fissile materisl. Again, the totals, as well as individual data points, demonstrate coverage of the compositions, configurations, and nuclear characteristics expected in plant systems. | information regarding the important characteristics of the fissile materisl. Again, the totals, as well as individual data points, demonstrate coverage of the compositions, configurations, and nuclear characteristics expected in plant systems. | ||
For example, considering the comment regarding the extendon of range of applicability from 7% and 7.5% fuel rod array experiments to moist oxide powders and nitrate Page 2 of 24 l | For example, considering the comment regarding the extendon of range of applicability from 7% and 7.5% fuel rod array experiments to moist oxide powders and nitrate Page 2 of 24 l | ||
solutions in the intermediate energy spectrum; the complete set of benchmark experiments, notjust the fuel rod array expedments, is used to include the above systems in the area of applicability. Please note in Table B the significant number of solution experiments, experiments in the 5 to 10 wt. % U " range, as well as the number of experiments with | solutions in the intermediate energy spectrum; the complete set of benchmark experiments, notjust the fuel rod array expedments, is used to include the above systems in the area of applicability. Please note in Table B the significant number of solution experiments, experiments in the 5 to 10 wt. % U " range, as well as the number of experiments with H/U n less than 100. Additionally, please note the abundance of experiments with energy 2 | ||
spectra in the range of 0.625 eV to 0.1 MeV. Solution systems are not expected to have intermediate energy spectra. | |||
As noted in Table B, there are 11 cases that have Energy Corresponding to Average Lethargy Causing Fission (ECALCF) that fall into the intermediate range. The fission fraction in the three energy regions for each of these 11 experiments is shown in Table C. | As noted in Table B, there are 11 cases that have Energy Corresponding to Average Lethargy Causing Fission (ECALCF) that fall into the intermediate range. The fission fraction in the three energy regions for each of these 11 experiments is shown in Table C. | ||
These experiments were not specifically designed to produce an intermediate spectrum, but they do have a significant portion of fissions occurring in the intermediate region. These experiments were designed as either fast or thermal systems. The fast systems have a soller spectrum than a pure metal system because of the presence of moderators (carbon or polyethylene). The thermal systems have a harder spectrum than an optimally moderated system due to being under-moderated. By the virtue of design of the experiments, the intermediate energy region is tested. | These experiments were not specifically designed to produce an intermediate spectrum, but they do have a significant portion of fissions occurring in the intermediate region. These experiments were designed as either fast or thermal systems. The fast systems have a soller spectrum than a pure metal system because of the presence of moderators (carbon or polyethylene). The thermal systems have a harder spectrum than an optimally moderated system due to being under-moderated. By the virtue of design of the experiments, the intermediate energy region is tested. | ||
| Line 80: | Line 95: | ||
Finally, as an example, criticality safety limits for the separator pod are based on a bounding conservative model. The model consists of a hemisphere of an optimum heterogeneous mixture of uranium and borated water with graphite reflection on all sides. | Finally, as an example, criticality safety limits for the separator pod are based on a bounding conservative model. The model consists of a hemisphere of an optimum heterogeneous mixture of uranium and borated water with graphite reflection on all sides. | ||
Other materials (such as the iron in the uranium alloy and copper crucible) and more explicit geometries were conservatively not included in the limit calculation. Had they been included, they would only reduce the reactivity of the calculated system. | Other materials (such as the iron in the uranium alloy and copper crucible) and more explicit geometries were conservatively not included in the limit calculation. Had they been included, they would only reduce the reactivity of the calculated system. | ||
Page 3 of 24 | Page 3 of 24 iE________________-.____ | ||
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0 0.0027 0.0275 0.0844 0.0144 0.3 0.68 1.2 H/U Figure A: | |||
Correlation Coeiricient for 238 Group Fission Fraction from Damp Oxide Critical Spheres Compared to the 11 Intermt.diate Spectrum Systems | |||
: 2. Comment: | : 2. Comment: | ||
Reference NUREG/CR-6361, " Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages" for guidance on developing subsets of experiments that are applicable to cenain parameters. Table 2 of your report should be revised to include the ranges of the parameters for which each subset of experiments is applicable. From these subsets, a linear regression fit would allow determination of the correlation coefficient of key parameters, and consequently, the upper safety limit (USL) for those operations for which the subsets apply. | Reference NUREG/CR-6361, " Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages" for guidance on developing subsets of experiments that are applicable to cenain parameters. Table 2 of your report should be revised to include the ranges of the parameters for which each subset of experiments is applicable. From these subsets, a linear regression fit would allow determination of the correlation coefficient of key parameters, and consequently, the upper safety limit (USL) for those operations for which the subsets apply. | ||
| Line 109: | Line 131: | ||
l l | l l | ||
Page 4 of 24 | Page 4 of 24 | ||
: 3. Comment: | |||
* What is the basis for the additional margin of suberiticality chosen to be 0.02? An appropriate margin has traditionally been accepted as being 0.05 for closely matching experiments with many data points. For those operations where there is little or no data, from critical experiments, a larger margin would normally be applied, Respouse: | * What is the basis for the additional margin of suberiticality chosen to be 0.02? An appropriate margin has traditionally been accepted as being 0.05 for closely matching experiments with many data points. For those operations where there is little or no data, from critical experiments, a larger margin would normally be applied, Respouse: | ||
Regarding the appropriateness of the 0.02 margin of suberiticality addressed in item 3, the previous discussions indicate that the critical experiments data set sutliciently covers the range of applicability for AVLIS criticality calculations supporting operations and off normal conditions as they are currently defined. Since there is no simple correlation between k,g aiv, and variations in physical parameters, the safety of AVLIS operations where reactivity is calculated is based on an understanding of the safety margins provided by controlled parameters. For each controlled parameter, a determination is made using sensitivity analysis to determine the relationship between k,guir, and variations in the controlled parameter. This relationship is used to establish adequate safety margins. This approach shifts the focus from an arbitrary k,g ais, value as an indication of the available safety margin to an understanding of the sensitivity of the ken,aive to changes in controlled parameters. | Regarding the appropriateness of the 0.02 margin of suberiticality addressed in item 3, the previous discussions indicate that the critical experiments data set sutliciently covers the range of applicability for AVLIS criticality calculations supporting operations and off normal conditions as they are currently defined. Since there is no simple correlation between k,g aiv, and variations in physical parameters, the safety of AVLIS operations where reactivity is calculated is based on an understanding of the safety margins provided by controlled parameters. For each controlled parameter, a determination is made using sensitivity analysis to determine the relationship between k,guir, and variations in the controlled parameter. This relationship is used to establish adequate safety margins. This approach shifts the focus from an arbitrary k,g ais, value as an indication of the available safety margin to an understanding of the sensitivity of the ken,aive to changes in controlled parameters. | ||
For each controlled parameter, the values of the parameter that correspond to the Failure and Safety Limits are determined. The Failure Limit is defined as the lowest point at which the system may be critical based on the t. certainties in the benchmark experiments, the cross section sets, and the computer code system. Its calculated k,gea:v, value plus uncertainties therefore is the Lower Tolerance Limit value as seen in the validation report (0.9753 for the 238-group cross section set). The Safety Limit is set below the Failure Limit value as an added margin of safety. The Safety Limit e ther does not exceed a k,a.uis, of | For each controlled parameter, the values of the parameter that correspond to the Failure and Safety Limits are determined. The Failure Limit is defined as the lowest point at which the system may be critical based on the t. certainties in the benchmark experiments, the cross section sets, and the computer code system. Its calculated k,gea:v, value plus uncertainties therefore is the Lower Tolerance Limit value as seen in the validation report (0.9753 for the 238-group cross section set). The Safety Limit is set below the Failure Limit value as an added margin of safety. The Safety Limit e ther does not exceed a k,a.uis, of l | ||
We believe that a recommended Safety Limit of(LTL-0.02) provides a sufficient margin of safety because systems with enrichments lower than 10 wt. % U* are generally less sensitive to changes in reactivity parameters. As an example Figure B shows the results of k,nouiv, versus Fraction of Critical Mass for 10 wt. % enriched uranium. The system approaches criticality slowly"and nonlinearly. The nonlinearity of k,g aiv, with mass is significant (kegeais, ~ FRAC | LTL-0.02 (0.9553 for the 238-group cross section set) or 85% of the Failure Limit value when measured in terms of a controlled parameter. Either of these limits may be used independently to defme the Safety Limit. AVLIS shall be operated such that no single failure in a controlled parameter will exceed the Safety Limit value for that parameter. | ||
of k,g aiv,. There is very little added safety, and there are tremendous processing and | We believe that a recommended Safety Limit of(LTL-0.02) provides a sufficient margin of safety because systems with enrichments lower than 10 wt. % U* are generally less sensitive to changes in reactivity parameters. As an example Figure B shows the results of k,nouiv, versus Fraction of Critical Mass for 10 wt. % enriched uranium. The system approaches criticality slowly"and nonlinearly. The nonlinearity of k,g aiv, with mass is significant (kegeais, ~ FRAC | ||
). The fraction of critical mass at a 0.02 margin is approximately 84%, and at a 0.05 margin is approximately 65%. Failure to account for l | |||
nonlinearity of the relationship between k,geais, and fraction critical mass is one of the inherent shortcomings of any criticality safety regulation that prescribes a rigid limiting value of k,g aiv,. There is very little added safety, and there are tremendous processing and i | |||
economic drawbacks in going from a margin of 0.02 to 0.05 for these nonlinear systems. This Safety Limit is consistent with other licensees in low enriched operations. | |||
As supporting information for the selection of 0.02 margin of subcriticality, a statistical analysis comparable to that found in NUREG/CR-6361 was performed. In NUREG/CR-6361, a statistical method for establishing an Upper Safety Limit (USL)is Page 5 of 24 | As supporting information for the selection of 0.02 margin of subcriticality, a statistical analysis comparable to that found in NUREG/CR-6361 was performed. In NUREG/CR-6361, a statistical method for establishing an Upper Safety Limit (USL)is Page 5 of 24 | ||
described. This USL is analogous to the Recommended Safety Limit described in the AVLIS | described. This USL is analogous to the Recommended Safety Limit described in the AVLIS | ||
, Validation Report (the AVLIS validation report method takes into account the number of benchmark experiments in the validation). A statistical analysis code, USLS7M7S provided by ORNL, facilitates the analysis described in V TREG/CR-6361. This code determines a litiear regression fit between the calculated ken x., and the parameter ofinterest (e g., H/X, fission energy, enrichment, etc.). The code wih establish an USL using a user-supplied administrative margin and one using a closed interval approach. The closed interval method establishes the margin of subcriticality as the minimum ditTerence between the confidence interval and the tolerance limit. This definition is unique to NUREG/CR-6361 and does not appear in any consensus criticality safety standard. | |||
NUREG/CR-6361 presents a methodology to determine the adequacy of a selected arbitrary margin of reactivity. This methodology represents the minimum margin of subcriticality as the minimum difference of the confidence limit and the tolerance limit. | NUREG/CR-6361 presents a methodology to determine the adequacy of a selected arbitrary margin of reactivity. This methodology represents the minimum margin of subcriticality as the minimum difference of the confidence limit and the tolerance limit. | ||
Using this methodology, the USL was calculated for each of the three cross section libraries versus enrichment, H/X, and Energy Corresponding to Average Lethargy Causing Fission (ECALCF) using 0.95/0.99 confidence and an administrative margin of 0.02. The calculated minimum margin of subcriticality for the 238-group cross section library is below 0.014 Ak, for the 27-group cross section library is between 0.015 Ak and 0.017 Ak, and for the 16 group is between 0.017 Ak and 0.022 Ak (Figure C). The one set that does not have a calculated minimum margin less than 0.02 Ak, is the 16-group cross section library when trended against ECALCF. The difference is due to the variance in the data for this library. | Using this methodology, the USL was calculated for each of the three cross section libraries versus enrichment, H/X, and Energy Corresponding to Average Lethargy Causing Fission (ECALCF) using 0.95/0.99 confidence and an administrative margin of 0.02. The calculated minimum margin of subcriticality for the 238-group cross section library is below 0.014 Ak, for the 27-group cross section library is between 0.015 Ak and 0.017 Ak, and for the 16 group is between 0.017 Ak and 0.022 Ak (Figure C). The one set that does not have a calculated minimum margin less than 0.02 Ak, is the 16-group cross section library when trended against ECALCF. The difference is due to the variance in the data for this library. | ||
Page 6 of 24 | Page 6 of 24 | ||
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1.1 1.2 1.3 1.4 1.6 1.6 1.7 1.8 1.9 Fraction of Critical Mass 1 | |||
Figure B: | |||
k-effective vs. Fraction of Critical Mass for AVLIS 10 wt.% Enriched i | |||
Uranium and Water I | |||
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The Recommended Safety Limit in the AVLIS validation report was developed using a somewhat different statistical approach. The calculational results were tested to determine if trends existed. Linear least square fits were determined for selected key parameters (IVX, ECALCF, enrichment, etc.). The correlation coetlicient ( | R h" | ||
In comparing the USL and the Recommended Safety Limit, the Recommended Safety Limit is lower than the USL in all cases (see Figure D). The Recommended Safety Limit is the Lower Tolerance Limit (LTL) minus the margin of suberiticality. Since the Recommended Safety Limit is less than the USL, the AVLIS method of validation is more conservative than the method specified in NUREG/CR-6361. The USL and the l | 0 k | ||
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238 27 16 i | |||
Cross Section Library Figure C: | |||
Calculated Minimum Margin of Suberiticality Based on the Method in NUREG/CR-6361. | |||
The Recommended Safety Limit in the AVLIS validation report was developed using a somewhat different statistical approach. The calculational results were tested to determine if trends existed. Linear least square fits were determined for selected key parameters (IVX, 2 | |||
ECALCF, enrichment, etc.). The correlation coetlicient (R ) was determined for these fits. If it was below 0.5, no trend was assumed to exist. For the parameters ofinterest, no significant trends were evident. The Validation Report shows the data to be normally distributed. The Lower Tolerance Limit (LTL) was developed for a 0.95/0.99 percent confidence level. The pooled variance used in establishing the LTL was the sum of the variance about the mean, the variance of the calculations and the variance of the experimental uncertainties. The Recommended Safety Limit was then established as the LTL minus the margin of suberiticality (0.02 Ak). | |||
In comparing the USL and the Recommended Safety Limit, the Recommended Safety Limit is lower than the USL in all cases (see Figure D). The Recommended Safety Limit is the Lower Tolerance Limit (LTL) minus the margin of suberiticality. Since the Recommended Safety Limit is less than the USL, the AVLIS method of validation is more conservative than the method specified in NUREG/CR-6361. The USL and the l | |||
Recommended Safety Limit for each of the three parameters (enrichment, IVX, and l | |||
ECALCF) and the three cross section libraries are shown in Figures E-M. | |||
Page 8 of 24 | |||
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238 27 16 Cross Section Library Figure D: | |||
Minimum Difference Between the Recommended Safety Margin and the Calculated USL. | |||
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l l | l l | ||
Table C Fission Fraction in the Three Energy Regions l | |||
Fraction of Fissions l | Fraction of Fissions l | ||
Case | Case Thermal Intermediate Fast | ||
(<0.625 eV) | (<0.625 eV) | ||
l | (>0.1 NieV) l Ieu-comp-therm-001c1 0.504 0.375 0.121 i | ||
l Leu-comp-therm-026c2 | l Leu-comp-therm-024c1 0.674 0.244 0.082 i | ||
l Leu-comp-therm-026c2 0.800 0.132 0.068 Leu-comp-therm-026c4 0.670 0.221 0.109 y-dr-81_t2e4 0.220 0.171 0.609 y-dr-81_t2e5 0.212 0.186 0.601 y-dr-8l_t2e21 0.257 0.227 0.517 y-dr-81_t2e25 0.183 0.177 0.640 y-dr-8 l_t4e97 0.091 0.228 0.681 y-dr-81_t4e98 0.094 0.260 0.647 y-dr-81_t6e8 0.111 0.204 0.685 l | |||
Page 14 of 24 | Page 14 of 24 | ||
I i | I i | ||
a Table D Fission Fractions for a Water Reflected Damp Oxide Systems | a Table D Fission Fractions for a Water Reflected Damp Oxide Systems UO Mass Weight ECALCF Thernial Intermediate Fast 2 | ||
IIydregen (II/U) 5300 | (kg) | ||
(2.73x10'') | Fraction of (eV) | ||
4400 | (<0.625 eV) | ||
(>0.1 MeV) | |||
3300 | IIydregen (II/U) 5300 0 | ||
7200 0.180 0.334 0.486 (0) 5400 1.02x10'' | |||
7300 0.173 0.342 0.485 (2.73x10'') | |||
l 4 | |||
4400 1.03x10 4900 0.183 0.357 0.459 (2.75x10-2), | |||
4 1 | |||
3300 3.14x10 2900 0.189 0.393 0.417 (8.44x10-2) 2700 5.34x10" 1900 0.190 0.424 0.386 (1.44x10-2) 1800 1.12x10'' | |||
740 0.209 0.464 0.327 (0.30) 1000 2.49x10'' | |||
147 0.277 0.473 0.250 (0.68) 540 4.20x10'' | |||
32 0.380 0.428 0.191 (1.2) 1 I | |||
1 l | 1 l | ||
l Page 15 of 24 | l Page 15 of 24 | ||
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l l | l l | ||
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ATTACHMENT 3 ATTENDEES i | |||
i l | i l | ||
i | |||
Y. | Y. | ||
,.rj e | |||
Meetina Attendees Name Organization Drew Persinko NRC Robert Pierson NRC Dan Martin - | |||
NRC Kim Hardin NRC | |||
Robert Woolley | ' Christopher Tripp NRC Charles Nilson NRC Harry Felsher NRC James Slider USEC Marc Klasky USEC Larry Wetzel USEC/BWXT Brian Kidd USEC/BWXT Francis Alcorn USEC/BWXT via conference phone: | ||
Robert Woolley USEC Ron Koopman USEC Christa Boman USEC Mark Michaelson USEC l | |||
- - - _ -}} | |||
Latest revision as of 20:15, 2 December 2024
| ML20236U547 | |
| Person / Time | |
|---|---|
| Site: | 07003089 |
| Issue date: | 07/24/1998 |
| From: | Persinko D NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9807300268 | |
| Download: ML20236U547 (60) | |
Text
i
.Tuly 24, 1998 i
i l
l TO; Robert C. Pierson, Chief Special Projects Branch, NMSS FROM:
Drew Persinko, AVLIS Project Manager l
Special Projects Branch, NMSS
SUBJECT:
MEETING
SUMMARY
FOR MEETING WITH U.S. ENRICHMENT CORPORATION TO DISCUSS CRITICALITY SAFETY VALIDATION REPORT
References:
- 1) AVLIS Criticality Code Validation Report, USEC to NRC, April 22,1998
- 2) Staff Request for AdditionalInformation, NRC to USEC, June 10,1998 l
On July 21,1998, NMSS/SPB staff met with staff from the U.S. Enrichment Corporation (USEC) and their contractor, B&W, to discuss the nuclear criticality safety validation report (Reference 1), submitted to support the licensing of the Atomic Vapor Laser Isotope Separation (AVLIS) technology, and responses to staff questions (Reference 2) on the report. USEC is requesting NRC to license the AVLIS facility up to a uranium enrichment of 10 percent. The nuclear criticality safety validation report was submitted to support this enrichment level. The j
staff will consider the information presented at the meeting and review the information provided j
at the meeting and determine whether any additional information is necessary for the staff to write a Safety Evaluation Report.
Discussion points addressing the staff's questions on the nuclear criticality safety validation report are provided as Attachment 1. Slides used in the meeting are provided as Attachment 2.
A list of meeting attendees is provided as Attachment 3.
USEC intends to submit an AVLIS license application in February 1999; however, non-site-specific information is being submitted for staff review before the application is submitted. The criticality safety validation report and this moeting were part of the ongoing pre-license application activities.
Attachments: As stated cc: R. Woolley, USEC
/
DISTRIBUTION:
l Docket 70-3089 Central File PUBLIC NMSS R/F FCSS R/F SPB R/F OFC
_SPB bmSih,[
S_P_B SM SPB bPersinko:ij
/ M in NAME L
ey 73f[98 1 M/98
/ /98 DATE 1'Nf98 8
C = COVER E =' COVER & ENCLOSURE N = NO COPY mtsumcrt.721 OFFICIAL RECORD COPY l
9
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Table C Fission Fraction in the Three Energy Regions l
Fraction of Fissions l
Case Thermal Intermediate Fast
(<0.625 eV)
(>0.1 NieV) l Ieu-comp-therm-001c1 0.504 0.375 0.121 i
l Leu-comp-therm-024c1 0.674 0.244 0.082 i
l Leu-comp-therm-026c2 0.800 0.132 0.068 Leu-comp-therm-026c4 0.670 0.221 0.109 y-dr-81_t2e4 0.220 0.171 0.609 y-dr-81_t2e5 0.212 0.186 0.601 y-dr-8l_t2e21 0.257 0.227 0.517 y-dr-81_t2e25 0.183 0.177 0.640 y-dr-8 l_t4e97 0.091 0.228 0.681 y-dr-81_t4e98 0.094 0.260 0.647 y-dr-81_t6e8 0.111 0.204 0.685 l
Page 14 of 24
I i
a Table D Fission Fractions for a Water Reflected Damp Oxide Systems UO Mass Weight ECALCF Thernial Intermediate Fast 2
(kg)
Fraction of (eV)
(<0.625 eV)
(>0.1 MeV)
IIydregen (II/U) 5300 0
7200 0.180 0.334 0.486 (0) 5400 1.02x10
7300 0.173 0.342 0.485 (2.73x10)
l 4
4400 1.03x10 4900 0.183 0.357 0.459 (2.75x10-2),
4 1
3300 3.14x10 2900 0.189 0.393 0.417 (8.44x10-2) 2700 5.34x10" 1900 0.190 0.424 0.386 (1.44x10-2) 1800 1.12x10
740 0.209 0.464 0.327 (0.30) 1000 2.49x10
147 0.277 0.473 0.250 (0.68) 540 4.20x10
32 0.380 0.428 0.191 (1.2) 1 I
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ATTACHMENT 3 ATTENDEES i
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i
Y.
,.rj e
Meetina Attendees Name Organization Drew Persinko NRC Robert Pierson NRC Dan Martin -
NRC Kim Hardin NRC
' Christopher Tripp NRC Charles Nilson NRC Harry Felsher NRC James Slider USEC Marc Klasky USEC Larry Wetzel USEC/BWXT Brian Kidd USEC/BWXT Francis Alcorn USEC/BWXT via conference phone:
Robert Woolley USEC Ron Koopman USEC Christa Boman USEC Mark Michaelson USEC l
- - - _ -