ML20209H237
| ML20209H237 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 07/14/1999 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20209G255 | List: |
| References | |
| 70-7002-97-203, NUDOCS 9907200173 | |
| Download: ML20209H237 (7) | |
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, DOCKET: 70-7002.
L : CERTIFICATE HOLDER: United States Enrichment Corporation -
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Portsmouth Gaseous Diffusion Plant '
I' Piketon, Ohio
SUBJECT:
. COMPLIANCE EVALUATION REPORT: Validation Report POEF- I
~ LMUS-13 Revision 1, " Validation of the CSAS25 Sequence in SCALE-4.3 and the 27-Group Cross Section Library for Use at the Portsmouth Gaseous Diffusion Plant."-
BACKGROUND Violation 70-7002/97-203-03 was issued on June 27,1997, for the failure of USEC to develop -
an adequate validation report that supported the use of the SCALE computer code at the L~
Portsmouth GDP (PORTS). The inspectors had observed that several PORTS calculations had -
been performed to establish the safety margin for systems based on an enrichment of 20 wt%
"U. The PORTS validation report at the time of inspection 97-203'did not contain any - .
bencnmarks in the range between 5 wt% to 93 wt%. The PORTS validation report concluded that the Average Fargy Group of neutrons causing fission was similar between the calculated results and the benchmarks below 5 wt% and above 93 wt% so that the range could be extended by interpolation. The NRC determined that this methodology did not adequately .
. address certificate commitments in thet ihe extension was not statistically supported.
As a result of this finding; USEC committed to identify additional benchmarks and revise the validation repor* t) include benchmarks related to the range of desired parameters. USEC
~ completed the validation report revision in' July 1998 and submitted the report to inspectors do ring inspection 70-7002/98-208 as evidence of completion of open item 70-7002/97-203-03.
Subsequent to the inspection, the staff performed ariindependent review of the revised validation report in order to resolve the open item. This Compliance Evaluation Report closes this issue by determining that the revised PORTS validation report meets certificate - l
- requirements and USEC commitments to resolve the violation. '
CERTIFICATION COMMITMENTS -
- Three categories of commitments are applicable to a PORTS validation report, the Portsmouth
. Safety' Analysis Report which is also the certificate application, the Technical Safety - j Requirements, and American National Standards Institute (ANSI) standard ANSI /ANS-8.1-1983. l yThe commitments are straightforward and nonprescriptive, allowing considerable flexibility in j how USEC prepares the validation report. !
Enclosure l l 9907200173 990714-PDR ADOCK 07007002 :
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Portsmouth Safety Analysis Report Section 5.2.3.2, under the heading " Computer Calculations,"
requires, in part:
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Critical experiments which have specific, well-defined parametric values and are adequately documented are termed benchmark experiments. Computer codes are validated using experimental data from benchmark experiments which, ideally, have geometries and material compositions similar to the systems being modeled.
Validation of the computer code determines its calculational bias or uncertainty as well as the effective margin of subcriticality. The PORTS validation involves the modeling of benchmark critical experiments over a range of applicability. Because the k,, value of a critical experiment is essentially 1.0, the bias of the code is taken to be the deviation of the calculated values of k,, from unity. Statistical analysis is employed to estimate the calculational bias, which includes the uncertainty in the bias and uncertainties due to extensions of the area of applicability, as well as the effective margin of subcriticality. Uncertainty in the bias is a measure of both the precision of the calculations and the accuracy of the experimental data.
The validation of the computer code specifically defines the maximum acceptable k , used to determine suberiticality.
1 The margin of suberiticality used at PORTS shall result in a k,, upper safety
. limit which will ensure that there is a 95 percent confidence that 99.9 of all future k,, values less than this value will be subcritical. The minimum margin of subcriticality of 0.02 in k,, shall be used to establish the acceptance criteria (i.e., upper safety limit) for criticality calculations. The upper safety limit for PORTS calculations is established as <.9605 in accordance with the validation report POEF-T-3636, Revision 1, " Validation of Nuclear Criticality Safety, Software and 27 Energy Group ENDF/B-IV Cross Sections" dated i January 1996. !
i When NCS is based on computer code calculations of k,,, controls and limits are established to ensure that the maximum k , complies with the computer code validation for that type of system being evaluated.
i Scoping and analysis calculations can be performed utilizing various unvalidated computer codes; however, computer calculations of k,, used as i the basis for NCS approvals are confirmed by, or performed using, I configuration-controlled codes and cross-section libraries for which j documented validations are performed with at least the same degree of conservatism as that presented in the validation report POEF-T-3636, Revision 1, and are in accordance with the American National Standards Institute (ANSI) standard ANSI /ANS-8.1-1983.
Technical Safety Requirement 3.11.3 states that "A minimum margin of suberiticality of 0.02 in I
k,,, shall be used to establish the acceptance criteria for criticality calculations."
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Technical Safety Requirement 3.11.4 states that "The k , for criticality calculations shall be 5 0.9605 which includes the bias, uncertainty, and the margin of suberiticality."
ANSI /ANS-8.1-1983 Section 4.3 under the heading " Validation of a Calculational Method" requires:
There are many calculational methods suitable for determining the effective multiplication factor (k ) of a system or for deriving suberiticallimits. The methods vary widely in basis and form, and each has its place in the broad
- spectrum of problems encountered in the nuclear criticality safety field.
However, the general procedure to be followed in establishing validity is common to all.
4.3.1 Bias shall be established by correlating the results of all criticality experiments with results obtained for these same systems by the method being validated, Commonly the correlation is expressed in terms of the values of k., calculated for the experimental systems, in which case the bias is the deviation of the calculated values of k , from unity. However, other parameters may be used. The bias serves to normalize a method over its area (s) of applicability so that it will predict critical conditions within the limits of the uncertainty in the bias. Generally neither the bias nor its uncertainty is constant; both should be expected to be functions of composition and other variables.
4.3.2 The area (s) of applicability of a calculational method may be extended beyond the range of experimental conditions over which the bias is i established by' making use of the trends in the bias. Where the extension is I large, the method should be supplemented by other calculatienal methods to provide a better estimate of the bias in the extended areas (s).
4.3.4 A margin in the correlating parameter, which margin may be a function l of composition and other variable, shall be prescribed that is sufficient to ensure subcriticality. This margin of subcriticality shallinclude allowances for the uncertainty in the bias and for uncertainties due to any extensions of the area (s) of applicability.
4.3.5 Nuclear properties such as cross sections should be consistent with l' experimental measurements of these properties.
4.3.6 A written report of the validation shall be prepared. This report shall: 1 (1) Describe the method with sufficient detail, clarity, and lack of )
ambiguity to ' allow independent duplication of results.
(2) State computer programs used, the options, recipes for choosing l mesh points where applicable, the cross section sets, and any numerical parameters necessary to describe the input.
(3) Identify experimental data and list parameters derived therefrom for use in the validation of the method.
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(4) State the area (s) of applicability.
(5) State the bias and the prescribed of subcriticality over the area (s) of applicability. State the basis for the margin."
The staff believes that for a particular safety limit to be considered adequately validated, the above requirements must be complied with and documented in a validation report combined with analyses of the specific situations to which the safety limits apply. All of the above )
requirements do not need to be addressed explicitly in the validation report but must be I addressed in the combination of validation report and analysis to establish confidence in the safety limit.
BENCHMARK SELECTION Fissile materialis found in a variety of forms at PORTS, thereby requiring a relatively broad area of applicability to be established. USEC enriches uranium using the gaseous diffusion process.
UF, is brought into the facility in cylinders and withdrawn into the process cascade in gaseous or liquid form. UF,is with6tawn from the cascade in liquid form and returned to cylinders. While in the process cascade, UF, encounters atmospheric moisture which causes 'JO2 F2 deposits in the cascade equipment.
UF, deposits are removed from cylinders by flushing with steam. USEC has facilities for removing UO2F2 deposits from equipment by spraying with nitric or citric acid. Uranium is recovered from waste by a solvent extraction process involving uranyl nitrate and tri-butyl phosphate. Calciners in the uranium recovery area produce U3 0, powder. Other significant uranium systems encountered are uranium-freon, uranium-alumina, and uranium contaminated lube oil.
The USEC validation report uses 301 benchmark cases divided into areas of LEU, IEU, and HEU which are further subdivided into 10 areas:
LEU - UO2F2 l LEU - UO2 Arrays i LEU - UO2 (NO3)
HEU/IEU - U-metal
- IEU - UF, The staff believes that the benchmarks selected explicitly model t,he major uranium systems in the facility with the exception of UF, and various waste configurations. USEC argues that UF. l and waste configurations are accurately represented by UF, and Uranium /U-metal benchmarks. ;
This argument is reasonable but not supported by anything in the validation report. Therefore, the adequacy of UF, and Uranium /U-metal benchmarks for UF. and waste systems must be l
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determined by review of specific analysis. This does not detract from the overall usefulness of the revised validation report.
STATISTICAL ANALYSIS Benchmarks were tmken down by USEC into the ten representative groups for statistical analysis. USEC performed standard tests for normality on the data within the groups using the Shapiro-Wilk and Lilliefors tests which are widely accepted tests. Benchmark data was analyzed by USEC using one of three methods based on the results of the normality tests:
weighted, single sided, lower tolerance limit
+ weighted confidence interval
- distribution free confidence interval based on the sign test The staff believes that the lower tolerance limit method incorporates relevant uncertainties in the bias as required by certificate commitments. The staff further believes that confidence interval methods may not adequately represent uncertainty in the bias, depending on the specific formulation, and therefore require additional discussion conceming how uncertainties are addressed. The staff believes that both the standard and the non-parametric confidence interval methods used in the validation report were otherwise a suitable methods based upon the wide acceptance of these methods in the literature and their previous wide and successful use in j validation.
The PORTS validation report analyzed benchmarks for trends based on two parameters; enrichment and average energy group causing fission. The staff believes that, due to the relative simplicity of and long experience with the systems concemed, these two parameters adequately to represent the physics of thermalized fissile systems involving uranium with q enrichments between 1.4 and 92.3 wt% and will allow a suitable comparison of analytical models with benchmarks.
The PORTS validation report displays the results of the statistical analysis in Table 3 of the report as either an Upper Safety Limit or a km.. The validation report does not explicitly state that the lower of either the TSR limit, USL, or km, must be used to support calculations which 1 rely solely on the validation report to comply with validation requirements. In accordance with l certificate commitments, a PORTS analysis is allowed to produce a result which exceeds the l USL or km, so long as it does not exceed the TSR limit, in this situation, the staff believes that additional justification is required to support the reduction in the established margin. The staff believes that this position is supported by the USEC comrnitment to comply with ANS-8.1.
l AREA OF APPLICABILITY The area of applicability that is encompassed by the USEC validation report is explicitly stated in Table 4 of the repon. UF is absent from the " Fissile Material" section of the table along with oil from the " Moderating Materials" section. Although these materials can be dealt with by ;
discussion in separate analyses, they are such broad issues that the staff believes they should reasonably be covered in the validation report. The staff believes that a validation report does not necessarily need to cover all situations in a facility in order to be valid for that facility. Other 5-
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I situations may be' addressed by an analysis that'uses the validation report as a base to extend the area of applicability or by selecting applicable benchmarks for the specific situation and
- i. performing separate statistical analysis.
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( - The staff believes that a validation report is not a procedure and does not establish
- requirements'L Therefore, a calculation which does not fall within the area of applicability established by the validation report must independently establish that all applicable certificate
- commitments have been met even though additional restrictions imposed by the validation
, report have been complied with by the analysisi An example is using UF, and Uranium /U-metal benchmarks for. UF and waste systems as discussed above.
- i. .f The USEC miidation report includes limitations on code options in section 6.0 " Conclusions."
The staff belidves that allowed code options should be tabulated in the area of applicability, statement wt ere there are issues regarding their use. Justification for the use of additional code options where an impact on the bias may result should be justified in individual analyses through '
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s sensitivity drialysis.
. The validation report does not include benchmarks between 10.07 and 29.8 wtE USEC argues that enrichment is equivalent to the concentration of 23sU, an isotope for which the physics is well understoed and that the benchmarks selected give a good feel for the behavior of the subject-systems across the stated range of enrichments. The staff believes that enrichment can be
, adequately viewed as 23s0 concentration and that adequate benchmarks are now available in the valirjation report to predict the behavior of related systems so long as additional margin is used ir, accordence with ANS-8.1 paragraph 4.3.4.
1' QQBOLUSION -
The staff believes that the validation report POEF-LMUS-13 Revision 1, is adequate to resolve the issue underlying Open item 70-7002/97-203-03 which was inadequate validation due to lack 1' of benchmark experiments for uranium systems with between 5 wt% and 20 wt% enriched 23sU.
/ Several weaknesses were noted in the validation report: i i
a : Allowed code options should be' tabulated in the area of applicability statement where l
- there are issues regarding their use. l l
= ' UF, should be included in the " Fissile Material" section of Table 4.
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- .: ? Oil should be included in the " Moderating Materials" section of Table 4.
' The report does not explicitly state that the lower of either the TSR limit, USL, or k. l
. must be used to support calculations which rely solely on the validation report to comply lwith validation requirements.
. Confidence interval methods may not adequately represent uncertainty in the bias.
. The Safety' Analysis Report refers to a different validation report than the one in use. !
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- item 4 on page 32 states that care should be taken with enrichments between 40 - 80 wt% 2"U. Actually, additional margin should be used between 10.07 and 29.8 wt% and 36 and 92.6 wt% or a justification for not using additional margin included in the analysis.
The staff believes that these issues may be easily corrected in a subsequent revision of the validation report and do not detract from the overall acceptability of the submitted report to resolve the original open item. The validation and determination of adequate margin for specific cases may be documented in the relevant analysis, as stated previously. Although there is no i benchmark data available at 20 wt% (where the original calculations were performed), and j additional margin may be warranted in this region, the staff believes there is no safety issue for l the particular cases leading to the violation. The licensee is authorized to operate up to 10 wt% j enrichment; modeling at 20 wt% instead of 10 wt% provides more than sufficient margin to i compensate for increased uncertainty in the bias in this range. The staff concludes that the !
revised PORTS validation report now contains adequate benchmarks to validate calculations between 5 wt% and 20 wt% 2"U, in principle. The staff concludes that the subcritical margin and statistical analysis for highly thermalized systems with compositions similar to the benchmark experiments is adequate between 5 and 10.07 wt%. The staff also concludes that l the revised PORTS validation report identifies benchmarks with geometries and material compositions similar to the systems being analyzed; employs a suitable approach to grouping the benchmarks for analysis; identifies a k, upper safety limit which will ensure that there is a 95 percent confidence that 99.9% of all future k, values less than this value will be suberitical; and !
does not view bias as constant. The documentation is therefore adequate for the purposes of resolving this violation. Open item 70-7002/97-203-03 is considered closed.
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, PRINCIPAL CONTRIBUTORS Dennis Morey j Chris Tripp :
I REFERENCES
- 1. ANSIIANS-8.1 1983,
- Nuclear Criticality Safety in Operations with Fissionable Materials ;
Outside Reactors," American Nuclear Society, La Grange Park, IL.
- 2. D. Lurie and R. H. Moore, " Applying Statistics," NUREG/-1475, U. S. Nuclear Regulatory Commission, Washington, DC (1994).
- 3. M. Hollander, and D. A. Wolfe, " Nonparametric Statistical Methods," John 'Niley & Sons, 1973.
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