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{{#Wiki_filter:* | {{#Wiki_filter:* USGS science for a changing world Department of the Interior US Geological Survey Box 25046 MS-974 Denver CO, 80225 July 10, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Subj: Response to RAI dated June 19, 2014, regarding R-113 license amendment request (TAC No.ME9424)Gentlemen: | ||
The attached pages are submitted in response to your Request for Additional Information dated | The attached pages are submitted in response to your Request for Additional Information dated June 19, 2014. Please contact me if you need additional information. | ||
Sincerely, Tim | Sincerely, Tim DeBey USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct.Executed on 07/10/2014 Copy to: Vito Nuccio, Reactor Administrator, MS 911 USGS Reactor Operations Committee Responses to RAI Questions Response to Question 1: It is proposed that License Condition 2.B(2) be changed from the current restriction of enriched uranium at 93.00 percent to any level of enrichment. | ||
BASIS: The basis for the requested change is two-fold. | BASIS: The basis for the requested change is two-fold. | ||
First, the current prescribed enrichment of 93. | First, the current prescribed enrichment of 93.00 percent is assigning more precision to the enrichment value than can be determined, or is specified by the supplier, for any of the SNM possessed in our existing neutron detectors. | ||
Second, the specification of 93.00 percent enrichment is overly restrictive for the potential use at the facility. | Second, the specification of 93.00 percent enrichment is overly restrictive for the potential use at the facility. | ||
Fission | Fission chamber neutron detectors may use uranium enriched at various levels, as a method for changing the detector's sensitivity to fast neutrons. | ||
Most reactor instrumentation detectors use uranium enriched to 90% | Most reactor instrumentation detectors use uranium enriched to 90% or above, but other research may benefit from the use of detectors at lower uranium enrichments in order to enhance fast neutron response.JUSTIFICATION: | ||
JUSTIFICATION: | Research requiring fission chamber neutron detectors can benefit from those detectors having enriched uranium at a varying enrichment level. The prior specification of 93.00 percent is unnecessarily and unreasonably restrictive. | ||
Research requiring fission chamber neutron detectors can benefit from those detectors having enriched uranium at a varying enrichment level. The prior specification of 93.00 percent | Response to Question 2: Revised Response to Question 7 of RAI dated January 29, 2014: The amendment request proposes a very minor increase to the mass limit of special nuclear material currently authorized under license R-113 in TRIGA fuel. All TRIGA fuel elements received at the GSTR will be possessed under the existing SNM limit for TRIGA fuel at the facility. | ||
Response to Question 2:Revised Response to Question 7 of RAI dated January 29, 2014:The amendment request proposes a very minor increase to the mass limit of special nuclear | The SNM possessed at the GSTR will continue to be of low strategic significance, and therefore, within our current license and security plan. The SNM contained in TRIGA fuel elements possessed at the facility may be utilized but not separated. | ||
The SNM possessed at the | A new item of 2 grams of special nuclear material of any enrichment is proposed to allow for reactor experiments, detector calibration, and reference sources related to reactor operation. | ||
A new item of 2 grams of special nuclear material of any enrichment is proposed to allow | The effect of the increase of 2 grams of SNM at any enrichment is minimal. The proposed wording for SNM authorized on the license is: B. Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," in connection with operation of the facility to receive, possess and use (but not separate): | ||
The | (1) up to 9 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of TRIGA reactor fuel;(2) up to 15 grams of contained uranium-235 of any enrichment in the form of neutron detectors; and (3) up to 2 grams of special nuclear material of any enrichment in reactor-based experiments, calibration of radiation detectors, and reference sources for reactor based programs.(4) such special nuclear material as may be produced by the operation of the facility.1 Response to Question 3: The type of experiment that may be performed using up to 2 grams of special nuclear material could involve irradiation of that material or use of that material as a reference source. The worst case scenario would involve neutron irradiation of 2 grams of SNM to the point where a GSTR technical specification limit on iodine inventory (1.5 Ci of 1-131 through 1-135) or strontium inventory (5 mCi of Sr-90) is reached. These limits are given in T.S. 1.9.Any proposed experiment involving irradiation of SNM would have a safety analysis performed, as part of the GSTR experiment review process, to ensure that T.S. 1.9 would not be violated. | ||
The proposed wording for | For example, 2 grams of pure U-235 could be irradiated for approximately 9.5 hours at a neutron flux of 4e12 without exceeding T.S. 1.9 limits. An existing GSTR Experiment Authorization allows the irradiation of up to 1.5 g of natural U0 2 or 10.5 mg of HEU for up to 5 hours at a neutron flux of 4e12, so that approved experiment is well within the limits of T.S. 1.9.The table below summarizes the maximum allowable irradiation times for pure (100%) U-233, U-235, and Pu-239 isotopes, using masses of 2 grams each and a neutron flux (fast and thermal) of 4e12 for each energy range.Neutron 1-131 Total number flux Total thru I- of fissions Expt (each, Irrad number of 135 Sr-90 produced in fissions SNM Mass fast and time fissions in produced produced "hot" fuel rod as % of isotope (grams) thermal) (hrs) experiment (Ci) (Ci) (22 kW) MHA U-233 2 4.OOE+12 4.15 1.75E+17 1.49 1.18E-07 1.03E+19 1.70 U-235 2 4.OOE+12 9.5 4.12E+17 1.49 6.18E-08 2.35E+19 1.75 Pu-239 2 4.OOE+12 2.33 1.33E+17 1.49 2.68E-08 5.76E+18 2.31 As can be seen from the table, the maximum irradiations that can be performed with the SNM isotopes all produce small fractions of the total fissions (and therefore, fission product inventories) for the Maximum Hypothetical Accident (MHA) analyzed for the GSTR. The scenario that is closes to the MHA is the irradiation of 2 grams of Pu-239, which produces 2.31% as many fissions as the MHA.Using a ratio of the MHA analysis results, failure and dispersal of the Pu-239 irradiation experiment in air would conservatively give the following estimated doses to staff members who were present in the reactor room at the time of failure.2 Likewise, failure and dispersal of the Pu-239 irradiation experiment in air would conservatively give the following estimated doses to persons outside of the reactor facility.Distance CDEThyroid (no TEDE (no Location (m) water) (nrem) water).(mrem) | ||
in connection with operation of the facility to receive, possess and use (but not separate): | Building 15 south 11 1 1 door Emergency 32 0.3 0.1 assembly area Building 21 east entrance (West 49 1 0.1 of Building 15)Average of eastern 100 0.9 0.06 intersections Building 16 west 175 0.4 0.02 entrance 200 0.3 0.02-250 0.2 0.01 Nearest Unrestricted 475 0.06 <0.01 Access Location Residence 640 0.03 <0.01 School 720 0.03 <0.01 This analysis is based on the MHA analysis that has many conservative factors used, such as no decay time, the release occurs in air, no filtration of the air, and no containment in the reactor bay. It can be seen that failure of this experiment does not present a significant threat to the reactor staff or nearby members of the public. The use of more realistic factors would give significantly lower dose estimates. | ||
(1) up to 9 kilograms of contained uranium-235 enriched to less than | Response to Question 4: The USGS technical specification on iodine and strontium activities produced in fueled experiments will be observed for the irradiation of the SNM discussed in this request. As with all GSTR experiments, a safety analysis will be performed to ensure that each experiment authorization for irradiation of SNM will not violate any requirements of the license, T.S., or procedures. | ||
(4) such special nuclear material as may be produced by the operation of the facility. | As discussed in the response to Question 3 (above), the irradiation times for 2 gram samples of pure SNM would be restricted to the values shown in the table below, in order to meet the limits of T.S. 1.9. It is not expected that 2 grams would be irradiated in any one experiment, so the irradiation time values below are conservative for actual experiments that would be performed. | ||
1 Response to Question 3:The type of experiment that may be performed using up to 2 grams of special nuclear material | An experiment authorization for irradiation of 0.1 gram of SNM could allow a significantly longer irradiation time and still meet the T.S. 1.9 requirements. | ||
These limits are given in T.S. 1.9.Any proposed experiment involving irradiation of SNM would have a safety analysis performed, as | 3 Neutron flux Irrad time SNM Mass (each, fast (hrs) to reach isotope (grams) and thermal) T.S. 1.9 limits U-233 2 4.00E+12 4.15 U-235 2 4.OOE+12 9.5 Pu-239 2 4.OOE+12 2.33 A new technical specification is not needed, because the experiment review process at the GSTR will ensure that T.S. 1.9 requirements are met for all SNM irradiations. | ||
For example, | Response to Question 5: Our proposed license condition 2.C.l.c has the following sentence: | ||
(hrs) experiment (Ci) (Ci) (22 kW) | "(Note: following irradiation, if >99%of the radioactivity in the material has been produced in the GSTR, the byproduct material will then be considered to be entirely GSTR-produced.)" Our proposed license condition 2.C.1.e has a similar note.The purpose of these notes is to eliminate time-consuming and unproductive inventory accounting work that has no safety significance. | ||
The basis for these notes is licensed material received from other licensees, under proposed license conditions 2.C.1, needs to be inventoried on a periodic basis to ensure that the limits of these conditions are met. The licensed material received from other licensees will decay over time, while most (if not all)of the items will be irradiated at the GSTR. The result is that the originally-received isotope activity will decrease while GSTR irradiation will produce newly-activated isotopes in that item. If no limit is provided for the transferred isotope inventory requirement, then vanishingly small values will need to be calculated, tracked, and recorded with no safety significance. | |||
Justification of the proposed notes may be best provided by example. Assume that licensed material (a pneumatic sample terminus) containing 0.1 mCi of Cr-51 is received from another TRIGA facility, for use at the GSTR. The Cr-51 isotope has a half-life of about 27.8 days. The terminus is made of 6061 aluminum alloy, with 0.3 weight% chromium composition. | |||
Distance CDEThyroid (no TEDE ( | The terminus is installed in the GSTR where 500 grams of the terminus sees a routine neutron flux of 1.36e12 neutrons/cm 2-s. After 26 hours of irradiation, there will have been >10 mCi of Cr-51 produced in the terminus and the Cr-51 received from the other licensee will have decayed to <1% of the total Cr-51. In addition, there will have been many mCi of AI-28, Mg-27, and Na-24 also produced in the terminus. | ||
Building 15 south 11 1 | The GSTR-produced isotopes will continue to grow, while the isotope received from the other licensee diminishes. | ||
Response to Question 4:The USGS technical specification on iodine and strontium activities produced in fueled experiments | It is not reasonable at this point to continue to perform separate accounting of the small amount of Cr-51 remaining from the other licensee. | ||
As with all GSTR experiments, | The proposed license condition will allow 100% of the radioactivity in the terminus to be accounted for as if it was produced in the GSTR.4 Conversely, if an item is received from another licensee and it is not irradiated at the GSTR, then it would continue to be inventoried as transferred material and it would be subject to the limitations of proposed license conditions 2.C.1.c and/or 2.C.1.e.Response to Question 6: It is correct that we have not included a license condition to account for SNM produced during operation of the GSTR, so we appreciate your suggested addition of a new condition to include that material. | ||
As discussed in the response | We propose adding a new condition 2.B (4):... (4) such special nuclear material as may be produced by the operation of the facility.5}} | ||
the irradiation times for 2 gram samples of pure SNM would be restricted to | |||
An experiment authorization for irradiation of 0.1 gram | |||
3 Neutron flux Irrad | |||
T.S. 1.9 | |||
Response to Question 5:Our proposed license condition 2.C.l.c has the following sentence: | |||
"(Note: following irradiation, if >99%of the radioactivity in the material has been produced in the GSTR, the byproduct material will then | |||
Our proposed license condition 2.C.1.e has a similar note.The purpose of these notes is to eliminate time-consuming and unproductive inventory accounting | |||
The basis for these notes is licensed material received from other licensees, under proposed | |||
Justification of the proposed notes may be best provided by example. | |||
Assume that licensed material ( | |||
The terminus is installed in the GSTR | |||
The GSTR-produced isotopes | |||
It is not reasonable | |||
The proposed license condition will allow 100% of the radioactivity in the terminus to | |||
5}} | |||
Revision as of 13:21, 9 July 2018
| ML14205A300 | |
| Person / Time | |
|---|---|
| Site: | U.S. Geological Survey |
| Issue date: | 07/10/2014 |
| From: | DeBey T M US Dept of Interior, Geological Survey (USGS) |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards |
| References | |
| TAC ME9424 | |
| Download: ML14205A300 (6) | |
Text
- USGS science for a changing world Department of the Interior US Geological Survey Box 25046 MS-974 Denver CO, 80225 July 10, 2014 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Subj: Response to RAI dated June 19, 2014, regarding R-113 license amendment request (TAC No.ME9424)Gentlemen:
The attached pages are submitted in response to your Request for Additional Information dated June 19, 2014. Please contact me if you need additional information.
Sincerely, Tim DeBey USGS Reactor Supervisor I declare under penalty of perjury that the foregoing is true and correct.Executed on 07/10/2014 Copy to: Vito Nuccio, Reactor Administrator, MS 911 USGS Reactor Operations Committee Responses to RAI Questions Response to Question 1: It is proposed that License Condition 2.B(2) be changed from the current restriction of enriched uranium at 93.00 percent to any level of enrichment.
BASIS: The basis for the requested change is two-fold.
First, the current prescribed enrichment of 93.00 percent is assigning more precision to the enrichment value than can be determined, or is specified by the supplier, for any of the SNM possessed in our existing neutron detectors.
Second, the specification of 93.00 percent enrichment is overly restrictive for the potential use at the facility.
Fission chamber neutron detectors may use uranium enriched at various levels, as a method for changing the detector's sensitivity to fast neutrons.
Most reactor instrumentation detectors use uranium enriched to 90% or above, but other research may benefit from the use of detectors at lower uranium enrichments in order to enhance fast neutron response.JUSTIFICATION:
Research requiring fission chamber neutron detectors can benefit from those detectors having enriched uranium at a varying enrichment level. The prior specification of 93.00 percent is unnecessarily and unreasonably restrictive.
Response to Question 2: Revised Response to Question 7 of RAI dated January 29, 2014: The amendment request proposes a very minor increase to the mass limit of special nuclear material currently authorized under license R-113 in TRIGA fuel. All TRIGA fuel elements received at the GSTR will be possessed under the existing SNM limit for TRIGA fuel at the facility.
The SNM possessed at the GSTR will continue to be of low strategic significance, and therefore, within our current license and security plan. The SNM contained in TRIGA fuel elements possessed at the facility may be utilized but not separated.
A new item of 2 grams of special nuclear material of any enrichment is proposed to allow for reactor experiments, detector calibration, and reference sources related to reactor operation.
The effect of the increase of 2 grams of SNM at any enrichment is minimal. The proposed wording for SNM authorized on the license is: B. Pursuant to the Act and 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material," in connection with operation of the facility to receive, possess and use (but not separate):
(1) up to 9 kilograms of contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of TRIGA reactor fuel;(2) up to 15 grams of contained uranium-235 of any enrichment in the form of neutron detectors; and (3) up to 2 grams of special nuclear material of any enrichment in reactor-based experiments, calibration of radiation detectors, and reference sources for reactor based programs.(4) such special nuclear material as may be produced by the operation of the facility.1 Response to Question 3: The type of experiment that may be performed using up to 2 grams of special nuclear material could involve irradiation of that material or use of that material as a reference source. The worst case scenario would involve neutron irradiation of 2 grams of SNM to the point where a GSTR technical specification limit on iodine inventory (1.5 Ci of 1-131 through 1-135) or strontium inventory (5 mCi of Sr-90) is reached. These limits are given in T.S. 1.9.Any proposed experiment involving irradiation of SNM would have a safety analysis performed, as part of the GSTR experiment review process, to ensure that T.S. 1.9 would not be violated.
For example, 2 grams of pure U-235 could be irradiated for approximately 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a neutron flux of 4e12 without exceeding T.S. 1.9 limits. An existing GSTR Experiment Authorization allows the irradiation of up to 1.5 g of natural U0 2 or 10.5 mg of HEU for up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a neutron flux of 4e12, so that approved experiment is well within the limits of T.S. 1.9.The table below summarizes the maximum allowable irradiation times for pure (100%) U-233, U-235, and Pu-239 isotopes, using masses of 2 grams each and a neutron flux (fast and thermal) of 4e12 for each energy range.Neutron 1-131 Total number flux Total thru I- of fissions Expt (each, Irrad number of 135 Sr-90 produced in fissions SNM Mass fast and time fissions in produced produced "hot" fuel rod as % of isotope (grams) thermal) (hrs) experiment (Ci) (Ci) (22 kW) MHA U-233 2 4.OOE+12 4.15 1.75E+17 1.49 1.18E-07 1.03E+19 1.70 U-235 2 4.OOE+12 9.5 4.12E+17 1.49 6.18E-08 2.35E+19 1.75 Pu-239 2 4.OOE+12 2.33 1.33E+17 1.49 2.68E-08 5.76E+18 2.31 As can be seen from the table, the maximum irradiations that can be performed with the SNM isotopes all produce small fractions of the total fissions (and therefore, fission product inventories) for the Maximum Hypothetical Accident (MHA) analyzed for the GSTR. The scenario that is closes to the MHA is the irradiation of 2 grams of Pu-239, which produces 2.31% as many fissions as the MHA.Using a ratio of the MHA analysis results, failure and dispersal of the Pu-239 irradiation experiment in air would conservatively give the following estimated doses to staff members who were present in the reactor room at the time of failure.2 Likewise, failure and dispersal of the Pu-239 irradiation experiment in air would conservatively give the following estimated doses to persons outside of the reactor facility.Distance CDEThyroid (no TEDE (no Location (m) water) (nrem) water).(mrem)
Building 15 south 11 1 1 door Emergency 32 0.3 0.1 assembly area Building 21 east entrance (West 49 1 0.1 of Building 15)Average of eastern 100 0.9 0.06 intersections Building 16 west 175 0.4 0.02 entrance 200 0.3 0.02-250 0.2 0.01 Nearest Unrestricted 475 0.06 <0.01 Access Location Residence 640 0.03 <0.01 School 720 0.03 <0.01 This analysis is based on the MHA analysis that has many conservative factors used, such as no decay time, the release occurs in air, no filtration of the air, and no containment in the reactor bay. It can be seen that failure of this experiment does not present a significant threat to the reactor staff or nearby members of the public. The use of more realistic factors would give significantly lower dose estimates.
Response to Question 4: The USGS technical specification on iodine and strontium activities produced in fueled experiments will be observed for the irradiation of the SNM discussed in this request. As with all GSTR experiments, a safety analysis will be performed to ensure that each experiment authorization for irradiation of SNM will not violate any requirements of the license, T.S., or procedures.
As discussed in the response to Question 3 (above), the irradiation times for 2 gram samples of pure SNM would be restricted to the values shown in the table below, in order to meet the limits of T.S. 1.9. It is not expected that 2 grams would be irradiated in any one experiment, so the irradiation time values below are conservative for actual experiments that would be performed.
An experiment authorization for irradiation of 0.1 gram of SNM could allow a significantly longer irradiation time and still meet the T.S. 1.9 requirements.
3 Neutron flux Irrad time SNM Mass (each, fast (hrs) to reach isotope (grams) and thermal) T.S. 1.9 limits U-233 2 4.00E+12 4.15 U-235 2 4.OOE+12 9.5 Pu-239 2 4.OOE+12 2.33 A new technical specification is not needed, because the experiment review process at the GSTR will ensure that T.S. 1.9 requirements are met for all SNM irradiations.
Response to Question 5: Our proposed license condition 2.C.l.c has the following sentence:
"(Note: following irradiation, if >99%of the radioactivity in the material has been produced in the GSTR, the byproduct material will then be considered to be entirely GSTR-produced.)" Our proposed license condition 2.C.1.e has a similar note.The purpose of these notes is to eliminate time-consuming and unproductive inventory accounting work that has no safety significance.
The basis for these notes is licensed material received from other licensees, under proposed license conditions 2.C.1, needs to be inventoried on a periodic basis to ensure that the limits of these conditions are met. The licensed material received from other licensees will decay over time, while most (if not all)of the items will be irradiated at the GSTR. The result is that the originally-received isotope activity will decrease while GSTR irradiation will produce newly-activated isotopes in that item. If no limit is provided for the transferred isotope inventory requirement, then vanishingly small values will need to be calculated, tracked, and recorded with no safety significance.
Justification of the proposed notes may be best provided by example. Assume that licensed material (a pneumatic sample terminus) containing 0.1 mCi of Cr-51 is received from another TRIGA facility, for use at the GSTR. The Cr-51 isotope has a half-life of about 27.8 days. The terminus is made of 6061 aluminum alloy, with 0.3 weight% chromium composition.
The terminus is installed in the GSTR where 500 grams of the terminus sees a routine neutron flux of 1.36e12 neutrons/cm 2-s. After 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> of irradiation, there will have been >10 mCi of Cr-51 produced in the terminus and the Cr-51 received from the other licensee will have decayed to <1% of the total Cr-51. In addition, there will have been many mCi of AI-28, Mg-27, and Na-24 also produced in the terminus.
The GSTR-produced isotopes will continue to grow, while the isotope received from the other licensee diminishes.
It is not reasonable at this point to continue to perform separate accounting of the small amount of Cr-51 remaining from the other licensee.
The proposed license condition will allow 100% of the radioactivity in the terminus to be accounted for as if it was produced in the GSTR.4 Conversely, if an item is received from another licensee and it is not irradiated at the GSTR, then it would continue to be inventoried as transferred material and it would be subject to the limitations of proposed license conditions 2.C.1.c and/or 2.C.1.e.Response to Question 6: It is correct that we have not included a license condition to account for SNM produced during operation of the GSTR, so we appreciate your suggested addition of a new condition to include that material.
We propose adding a new condition 2.B (4):... (4) such special nuclear material as may be produced by the operation of the facility.5