ML20198P061
| ML20198P061 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 10/28/1997 |
| From: | Ten Eyck E NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | John Miller UNITED STATES ENRICHMENT CORP. (USEC) |
| References | |
| 70-7002-97-203, NUDOCS 9711060260 | |
| Download: ML20198P061 (6) | |
Text
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~ \ UNITED STATES
.[ I}C NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20565-0001
- October 28, 1997 Mr. J.ll. Miller Vice President Production United States Enrichment Corporation Two Democracy Center 6903 Rockledge Drive Bethesda, MD 20817
SUBJECT:
RESPONSE TO NOTICE OF VIOLATION NO. 70-7002/97-203-03
Dear Mr. Miller:
This refers to your July 28,1997, response to a Notice of Violation (NOV) transmitted to you by our letter dated June 27,1997, with Inspection Report 70-7002/97-203. In your response, you acknowledged two cited violations and denied cited Violation 70-7002/97-203-03 and -
provided a basis for your denial in Enclosure 2.
After careful consideration of the information provided by you, we have determined, for the reasons provided in Attachment 1, that Violation 70-7002/97-203-03 occurred as stated, and that your response to the Notice of Violation is inadequate. The response is inadequate because it does not provide any additional technical basis that the bCALE code is validated in the intermediate range, or demonstrate that the code is well-behaved between 5% and 92.5% enrichment. We believe it would be beneficial to hold a management meeting between yourself and NRC Headquarters staff to reach a common understanding regarding our expectations in this matter. You will be contacted shortly to set up a mutually agreeable date.
Pursuant to the provisions of 10 CFR 76.60, USEC is hereby required to submit a written //
statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with copies to the Regional Administrator, Region 111, h a and Chief, Fuel Cycle Operations Branch, Division of Fuel Cycle Safety and Safeguards, NMSS, and a copy to the NRC Resident inspector at the facility that is the subject of this Notice, within -
b 30 days of the date of this letter. This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further siolations, and (4) the date when full compliance will be achieved. Your response may reference or include previously docketed correspondence if the correspondence adequately addresses the required response. If an adequate reply is not received within ab time specified in this letter, an order or Demand for Information may be issued as to why the certificate should not be modified, suspended, or revoked, or why such other actions as may be 9711060260 971023 PDR ADOCK 07007002 C PDR
, . proper should not be taken. Where good cause i; shown, consideration will be given to extending the response time Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it she sild not include any personal privacy, proprietar/, or safeguards inf rmation so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response. then please provide a bracketed copy of your response that identifies the information that should be prot :cted and a redacted copy of your response that deletes such information. If you request withholding of such material, you me specificall; Mentify the portions of your response that you seek to have withheld, and provide in detail tue basis for your claim of withholding (e.g., explain why the disclesure ofinformation will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Sincerely, SU (E '- (d$
Elizabe i Q. Ten k, Director O Division of Fuel Cycle Safety and Safeguards, NMSS Docket No. 70-7002 i
Enclosure:
Analysis of Response
(
October 28, 1997 l
1 proper should not be taken. Where good cause is shown, consideration will be given to extending the response time. Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it con be placed in the PDR without redaction, if personal privacy or proprietary information is necessary to provide an acceptable response, then plerise provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such info,mation. if you request withholding of such material, you inusi specifically identify the portions of your response thct yeu seek to have withheld, and provide in detail the basis for your claim of withholding (e.g., explain why the disclosure ofinformation will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an ,
acceptable response, please provide the level of protection described in 10 CFR 73.21.
Sincerely, Elizabeth Q. Ten Eyck, Director Division cf Fuel Cycle Safety and Safeguards, NMSS Docket No. 70-7002 ,
Enclosure:
Analysis of Response QlSIILBMIIDN*
RPierson, SPB YFaraz, SPB FCOB r/f D. Hartland. Region 111 P. Hiland, Region ill Docket File 70-7002 PDR/LPDR Case File FCSS r/f NMSS r/f DOCUMENT NAME: G:\20upvio.wpd OFC FCOB b FCOB E FCOB 6 FCSS NAME DMorey Dhdl WSch [ PTing EQ Te Eyck DATE il /1/97 l' 97 io IVi97 10/f-[/97 C = COVER E = COVER & ENCLOSURE N = NO COPY J
~
ANALYSIS OF PORTSMOUTil RESPONSE TO NOTICE OF VIOLATIONS RESTATEMENT OF VIOldTION TSR Section 3.11.1 requires, in part, that "A Criticality Safety Prop am shall be established, implemented, and maintained as described in the Safety Analysis Report...".
SAR Section 5 ?.3.2 requires, in part, that "When NCS is based on computer code calculations of K,y, controls and limits are established to ensurc ' hat the maximum K,y complies with the applicable code validation for that type of system being evaluated."
Contrary to the above, as of May 30,1997, three Nuclear Criticality Safety calculations, NCS- CALC-97-009, NCS-CALC 97-010, and NCS-CALC-97-012, were observed to be based upon an enrichment of 20%, without justification that the results are bounded by the validation
, ceport.
SUMMARY
OF PORTS RESPONSE The regulatee denied the violation, claiming that the SCALE Code is adequately validated for the above referenced calculations. The response quotes SAR Section 5.2.3.2, which states, in part:
" Computer codes are validated using experimental data from benchmark experiments which, ideally, have geometries and materhl 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 suberiticality. The PORTS validation involves the modeling of benchmark critical experiments over a range of applicability. .. 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...".
The regulatee states that the SCALE Code is validated in accordance with the above passage from the SAR. The regulatee further states that the benchmark critical experiments used in the validation cover a large range of moderator and reflector conditions, and thus a wide range of neutron energy. They state that the validation test cases demonstrate " based on the low enriched and Ligh enriched results, that the 23'U microscopic cross sections are valid over a wide energy range." The graphs of K,g as a function of Average Energy Group (AEG) clearly demonstrate that the test cases "have thoroughly tested the SCALE cross section library over a wide range of neutron energy spectrums, for both low and high enriched cases." The response further states that "In order for the relative accuracy / bias of a particular SCALE calculation to be affected by 2
the enrichment of the fissile material, the2 "U or "U miscroscopic cross sections in the energy groups where the most neutrons exist would have to be different. Such a problem would have to manili:st itselfin the results of either the high or low enriched validation test cases.. ".
ENCLOSURE
o
. 1
, 1 l
4 NRC ANALYSIS '
SAR Section 5.2.3.2 states that benchmark experiments should have " geometries and material compositions similar to the systems being evaluated." The benchmark critical experiments used in the SCALE validation cover the ranges from 1.4% to 5,0% and 92.5% to 97.7% in2 "U enrichment. Ilowever, the uranium in the three calculations was enriched to 20%, which is substantially different from the material composition of the benchmark experiments. Moreover, statistical analysis has not been used to determine the calculational bias due to extending the area of applicability to the intermediate enrichment range. Therefore the code has not been validated in accordance with SAR Section 5.2.3.2, for the intermediate enrichment cases.
An examination of the SCALE validation report does show that the text cases cover a wide range of neutron energies. However, Figures C-2 and C-3 in the regulatee's response to Violation 97 203-03 show that the AEG range is almost the same for both the low and the high enriched bencluaark cases. Demonstration that a particular colculation yields an AEG within the range indicated in these figures is rat adequate demonstration that the enrichment falls within the validated range of applicability, since there is no apparent correlation between AEG and enrichment. Thus, the observation that there is no trend in the bias as a function ot neutron
~
energy does not demonstrate that there is no trend in the bias as a function of enrichment.
The calculational bias is, in general, a function of a number of neutronic and physical variables of the system. This is the reason that benchmark experiments should have physical characteristics similar to those systems being evaluated. The AEG is but one parameter averaged over the entire run and is not the only parameter that can exhibit a trend in the bias.
The response states that either the2 "U or 23: U microscopic cross sections would have to be in error for there to be a trend in the bias as a function of enrichment. The purpose of having a validation of computer codes used for nuclear criticality analysis is to confirm that the codes function as intended over a certain range of applicability. The SCALE code is a complex package of software modules and cross section libraries with the potential for errors to occur in either the executable coae or the nuclear data. This response does not consider the possibility that the computer code itself could be in error, rather than the neutron cross-section data. Such an error in the executable code could have unpredictable results and may only manifest itself where the macroscopic cross section involves significant quantities of both 2"U and 23 U.
Validation of benchmark cases over a well-defined range of parameters is required because of the complexity of the code and the potential for problems of an unpredictable nature to occur. There is an unsubstantiated assumption that the bias is approximately linear between the low enriched and high enriched ranges, without any demonstration of the behavior in the intermediate range.
The response also did not address the two calculations NCS-CALC-97-009 and NCS-CALC-96-012, whith did not use the AEG argument tojustify the validity of using 20 wt%
- enrichment in the computer model.
s; -
NRC CONCI USION The NRC concludes that the violation occurred as stated in the June 27,1997, Notice of Violation.
1
' p uay p-k, UNITED STATES s
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055!Ko01
- f October 28, 1997 Mr. J.II. Miller Vice President -Production Unitec States Enrichment Corporation Two Democracy Center 6903 Rockledge Drive Bethesda, MD 20817
SUBJECT:
RESPONSE TO NOTICE OF VIOLATION NO. 70-7002/97-203-03
Dear Mr. Miller:
This refers to your July 28,1997, response to a Notice of Violation (NOV) transmitted to you by our letter dated June 27,1997, with inspection Report 70-7002/97-203. In your response, you acknowledged two cited violations and denied cited Violation 70-7002/97-203 03 and provided a basis for your denial in Enclosure 2, After careful consideration of the information provided by you, we have determined, for the reasons provided in Attachment 1, that Violation 70-7002/97-203-03 occurred as stated, and that your response to the Notice of Violation is inadequate. The response is inadequate because it does not provide any additional technical basis that the SCALE code is validated in the intermediate range, or demonstrate that the code is well-behaved between 5% and 92.5% enrichment. We believe it would be beneficial to hold a management meeting between yourself and NRC Headquarters staff to reach a common understanding regarding our expectations in this matter. You will be contacted shortly to set up a mutually agreeable date.
Pursuant to the provisions of 10 CFR 76.60, USEC is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Centrol Desk, Washington, D.C. 20555 with copies to the Regional Administrator, Region Ill, and Chief Fuel Cycle Operations Branch, Division of Fuel Cycle Safety and Safeguards, NMSS, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, within 30 days of the date of this letter. This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, (2) the corrective s+.eps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your respense may reference or include previously docketed correspondence if the correspondence adequately addresses the required response. If an adequate reply is not received within the time specified in this letter, an order or Demand for Information may be issued as to why the certificate should not be modified, suspended, or revoked, or why such other actions as may be YkD0 kl'
proper thould not be taken. Where good cause is shown, consideration will be given to extending the response time. Because your response will be placed in the NRC Public Document l
Room (PDR), to the extent possible, it should not include any petsonal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or ~ aprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information, if you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld, and provide in detail the basis for your claim of withholding (e.g., explain why the disclosure ofinformation will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described la 10 CFR 73.21.
Sincerely, L- -
g Elizal i Q. Ten k, Director Division of Fuel Cycie Safety and Safeguards, NMSS Docket No. 70-7002
Enclosure:
Analysis of Response
- October 28, 1997 f
. . 2-l proper should net be taken. Where good cause is shown, consideration will be given to extending the response time. Because your response will be placed in the NRC Public Document j Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a
. bracketed copy of ycur response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you mug specifically identify the portions of your response that you seek to have withheld, and provide in detail the basis for your claim of withholding (e.g., explain why the disclosure ofinformation will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholdi'ng confidential commercial or financial information). If safeguards infbrmation is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.
Sincerely, Elizabeth Q. Ten Eyck, Director Division of Fuel Cycle Safety and Safeguards, NMSS Docket No. 70-7002
Enclosure:
Analysis of Response DISTRIBUTION RPierson, SPB YFsrat, SPB FCOB r/f D. Hartland, Region ill P. Hiland, Region 111 Docket File 70-7002 PDR/LPDR Case File FCSS r/f NMSS r/f DOCUMENT NAME: G:\2oupvio.wpd OFC FCOB b FCOB d FCOB [ FCSS NAME DMorey ([MI WSch [ PTing EQ Te Eyck DATE il l'l/97 l- 97 t o 14/97 10/f-(/97 C = COVER E = COVER & ENCLOSURE N = NC COPY
f ANALYSIS OF PORTSMOUTil RESPONSE TO NOTICE OF YlOLATIONS Rl!STATEhll!NT OF VIOLATION TSR Section 3.11.1 requires, in part, that "A Criticality Safety Program shall be established, implemented, and maintained as described in the Safety Analysis Report...".
S AR Section 5.2.3.2 requires, in part, that "When NCS is based on computer code calculations of K,n, controls and limits are established to ensure that the maximum K,n complies with the applicable code validation for that type of system being evaluated."
Contrary to the above, as of May 30,1997, three Nuclear Criticality Safety ' calculations, NCS. CALC 97 009 NCS CALC 97 010, and NCS CALC 97-012, were observed to be based upon an enrichment of 20R without justiliention that the results are bounded by the validation reg ort.
SUMMARY
OF PORTS RIISPONSE The regulatee denied the viola lon, claiming that the SCALE Code is adequately validated for the above-referenced calculations. The response quotes SAR Section 5.23.1, which states,in part:
(
" Computer codes are va'! dated using experimental data from benchmark experiments which, ideally, have geometries and raaterial compositions similar to the systems being modeled."
"Vallde' ion of the computei code determines its calculational bias or uncertainty as well -
as the effeulve margin of suberiticality. The PORTS validation involves the modeling of benchmark critical experiments over a range of applicability.... Statistical analysis is employed to estimate the calculational bias, which includes the uncenainty in the bias and uncertainties due to extensions of the area of applicability...".
The regulatee states that the SCALE Code is validated in accordance with the above passage from the SAR The regulatee further states that the benchmark critical experiments used in the validation cover a large range of moderator and reflector conditions, and thus a wide range of neutron energy. They state that the validation test cases demonstrate " based on the low enriched and high enriched results, that the 23: U microscopic cross sections are valid over a wide energy range." The graphs of K,aas a function of Average Energy Group (AEG) clearly demonstrate that the test cases "have thoroughly tested the SCALE cross section library over a wide range of neutron energy spectrums, for both low and high enriched cases." The response further states that "In order for the relative accuracy / bias of a particular SCALE calculation to be affected by the enrichment of the fissile material, the 23'U or 2)!U miscroscopic cross sections in the energy groups where the most neutrons exist would have to be different. Such n problem would have to manifest itselfin the results of either the high or low enriched validation test cases...".
ENCLOSURE
+
2 NRC ANALYSIS SAR Section 5.2.3.2 states that benchmark experiments should have " geometries and material compositions siullar to the systems being evaluated." The benchmark critical experiments used in the SCALE validation cover the ranges from 1.4% to 5.0% and 92.5% to 97.7% in2 "U enrichment. Ilowever, the uranium in the three calculations was enriched to 20%, which is substantially difTerent from the.naterial composition of the benchmark experiments. Moreover, statistical analysis has not been used to determ'.ne the calculational bias due to extending the area of applicability to the intermediate enrichr mt range. Therefore the code has not been validated in accordance with SAR Section 5.2.3.2, foi the intermediate enrichment cases.
An examination of the SCALE validation report does show that the test cases cover a wide range of neutron energies. Ilowever,l'igures C 2 and C 3 in the regulatee's response to Violation 97 203 03 shaw that the AEG range is almost the same for both the low and the high enriched benchmark cases, Demonstration ' it a particular calculation yields an AEG within the range indicated in these figures is not adequate demonstration that the enrichment falls within the validated range of applicability, since there is no apparent correlation between AEG and enrichment. Thus, the observation that there is no trend in the bias as a function of neutron energy does not demonstrate that there is no trend in the bias as a function of enrichment.
The calculational bias is, in general, a function of a number of neutronic and physical variables of the system. This is the reason that benchmark experiments should have physical characteristics similar to those systems being evaluated. The AEG is but one parameter averaged over the entire run and is not the only parameter mat can exhibit a trend in the bias.
The response states that cither the2 "U or 2nU microscopic cross sections would have to be in error for there to be a trend in the bias as a function of enrichment. The purpose of having a validadon of computer codes used for nuclear criticality analysis is to confirm that the codes function as intended over a certain range of applicability. The SCALE code is a complex package of software modules and cross section libraries with the potential for errors to occur in either the executable code or the nuclear data. This response does not consider the possibility that the computer code itself could be in error, rather than the neutron cross section data. Such an error in the executable code could have unpredictable results and may only manifest itself where the macroscopic cross section involves significant quantities of both "iU and 2"U.
Validation of benchmark cases over a well-defined range of parameters is required because of the complexity of the code and the potential for problems of an unpredictable nature to occur. There is an unsubstantiated assumption that the bias is approximately linear between the low enriched and high enriched ranges, without any demonstration of the behavior in the intermediate range.
- The response also did not address the two calculations NCS CALC-97 009 and NCS CALC 96-012, which did not use the AEG argument to justify the validity of using 20 wt%
enrichment in the computer model.
e
t ,
3 NRC CONCLUSION 171e RC concludes that the violation occurred as stated in the June 27,1997, Notice of 4
4 e'
-A
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