ML12270A415
| ML12270A415 | |
| Person / Time | |
|---|---|
| Site: | U.S. Geological Survey |
| Issue date: | 10/02/2012 |
| From: | Geoffrey Wertz Research and Test Reactors Licensing Branch |
| To: | Adrian B US Dept of Interior, Geological Survey (USGS) |
| Wertz, Geoffrey 301-415-0893 | |
| References | |
| TAC ME1593 | |
| Download: ML12270A415 (6) | |
Text
October 2, 2012 Ms. Betty Adrian Reactor Administrator Department of the Interior U.S. Geological Survey PO Box 25046 MS 975 Denver Federal Center Denver, CO 80225-0046
SUBJECT:
UNITED STATES GEOLOGICAL SURVEY - ADDITIONAL CLARIFICATION REQUESTED RE: RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION DATED SEPTEMBER 29, 2010 (TAC NO. ME1593)
Dear Ms. Adrian:
The U.S. Nuclear Regulatory Commission (NRC) is continuing its review of your application for the renewal of Facility Operating License No. R-113 for the U.S. Geological Survey TRIGA Reactor (GSTR), dated January 5, 2009 (a redacted version of the safety analysis report is available on the NRCs public Web site at www.nrc.gov under Agencywide Documents Access and Management System (ADAMS) Accession No. ML092120136). As part of our review, the NRC staff submitted requests for additional information (RAIs) by letter dated September 29, 2010 (ADAMS Accession No. ML102510077).
The NRC staff has reviewed your responses to our RAIs and has identified several RAI responses needing additional clarification in the attached table. Please provide responses to the enclosed request for additional information within 45 days of the date of this letter. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.30(b), you must execute your response in a signed original document under oath or affirmation. Your response must be submitted in accordance with 10 CFR 50.4, Written Communications.
Information included in your response that is considered security, sensitive, or proprietary, that you seek to have withheld from the public, must be marked in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding.
If you have any questions about this review or if you need additional time to respond to this request; please contact me by telephone at 301-415-0893 or by electronic mail at geoffrey.wertz@nrc.gov.
Sincerely,
/AAdams for RA/
Geoffrey Wertz, Project Manager Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-274 cc: See next page
U.S. Geological Survey TRIGA Reactor Docket No. 50-274 cc:
Environmental Services Manager 480 S. Allison Pkwy.
Lakewood, CO 80226 State of Colorado Radiation Management Program HMWM-RM-B2 4300 Cherry Creek Drive South Denver, CO 80246 Mr. Timothy DeBey Reactor Director U.S. Geological Survey Box 25046 - Mail Stop 424 Denver Federal Center Denver, CO 80225 Test, Research, and Training Reactor Newsletter Universities of Florida 202 Nuclear Sciences Center Gainesville, FL 32611
The NRC staff has reviewed your responses to our RAIs and has identified several RAI responses needing additional clarification in the attached table. Please provide responses to the enclosed request for additional information within 45 days of the date of this letter. In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.30(b), you must execute your response in a signed original document under oath or affirmation. Your response must be submitted in accordance with 10 CFR 50.4, Written Communications.
Information included in your response that is considered security, sensitive, or proprietary, that you seek to have withheld from the public, must be marked in accordance with 10 CFR 2.390, Public inspections, exemptions, requests for withholding.
If you have any questions about this review or if you need additional time to respond to this request; please contact me by telephone at 301-415-0893 or by electronic mail at geoffrey.wertz@nrc.gov.
Sincerely,
/AAdams for RA/
Geoffrey Wertz, Project Manager Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Docket No. 50-274 cc: See next page DISTRIBUTION:
PUBLIC DPR/PRT r/f RidsNrrDpr RidsNrrDprPrta RidsNrrDprPrtb GWertz, NRR GLappert, NRR PTorres, NRR ADAMS Accession No.: ML12270A415 *concurrence via e-mail NRR-088 Office PRLB:PM PRLB:LA PRLB:ABC PRLB:PM*
Name GWertz GLappert AAdams GWertz Date 9/25/2012 10/1/2012 10/2/2012 10/2/2012
OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION RE: RENEWAL OF THE FACILITY OPERATING LICENSE FOR THE UNITED STATES GEOLOGICAL SURVEY TRIGA REACTOR LICENSE NO. R-113; DOCKET NO. 50-274 The NRC staff has reviewed your responses to our requests for additional information (RAIs) and has identified several RAI responses that were incomplete or needed additional clarification in the attached table. Please provide responses to the enclosed request for additional information within 45 days of the date of this letter.
RAI No.
Original RAI Information Needed 9
Describe the limiting core configuration.
Please provide the results of the U.S. Geological Survey TRIGA Reactor (GSTR) neutronic analyses that document the Limiting Core Configuration (LCC) and operating core (OC) including:
A core map showing the contents of the core lattice positions for the LCC and the OC.
The enrichment and cladding type for the fuel elements used at GSTR.
Diagrams and dimensions for fuel elements, control elements and other occupants of lattice positions.
The effective delayed neutron fraction (eff) used for the LCC and the OC.
The all-control-rods-out k-effective (keff) and the excess reactivity (excess) for the LCC and the OC.
The control rod worths for each of the 4 control rods including the keff values determined for the LCC and the OC.
The comparison of the excess and the control rod worths calculated and measured from the OC.
The shutdown reactivity of the core with the maximum worth rod removed (stuck rod condition) and the resulting maximum experiment worth that can be inserted into the core without violating shutdown margin requirements for the OC.
A power distribution graphic for the LCC and the OC showing power in kW per fuel element.
10 Provide coefficients and power distribution estimates.
Please provide the fuel temperature coefficient for the LCC and the OC as a function of fuel temperature over the temperature range expected for the GSTR operation.
12 Describe the DNBR analysis.
Please provide thermal-hydraulics data for the LCC consistent with the following:
Identify the unit cell used to define the RELAP model - graphically and with dimensions.
Identify and justify any entry/exit loss coefficients employed in the RELAP model.
Provide a diagram of the RELAP model used.
Document input assumptions used to analyze DNBR for the LCC such as fuel element power, peaking factors employed, inlet temperatures assumed, etc.
Document the RELAP model calculated results such as the core flow rate, peak fuel and cladding temperatures, the location of the minimum DNBR, and the value of the minimum DNBR using the Bernath correlation.
Characterize the response of GSTR to a reactivity pulse and an uncontrolled rod withdrawal transient event.
Please provide similar information as with DNBR results but also include the final power achieved in the event, the duration of the pulse or event, and the sequence of events (e.g., the initial power prior to the event, the scram circuits that are active, when the transient rod is released and inserted, when the other control rods are inserted, etc.).
14.2 Describe how LSSS and SCRAM setpoints protect the safety limit.
Please provide the analysis of the uncontrolled rod withdrawal for the LCC; this analysis should be consistent with evaluated control rod worths and should demonstrate the acceptability of scram setpoints, control rod drop times, and control rod withdrawal rates and speeds in the technical specifications.
15.3 Explain the methods used to determine the thyroid doses Please explain the use of 22 kW per fuel element, and why it was used rather than using the hot rod power determined from the LCC. Clarify the basis for the value used in the accident analysis.
The hot rod inventory was calculated using the fission yield factors for uranium-235, and the assumption of saturation conditions for the halogens and noble gases. However, NRC staff calculations using the fission yields in the Chart of the Nuclides, or those of the ENDF/B-VI in Summary Documentation Report (BNL-NCS-17541, ENDF-201, 1991) could not reproduce the licensees radiological inventory. The major noted differences were in the estimation of Kr-85, where the NRC staff estimate was higher, and Br-82, where the licensees estimate was higher. Please provide the fission yield data used, or explain how the source term inventory is calculated in sufficient detail to allow independent confirmation. Please explain whether GSTR is using 1 year of operation or saturated results.
The calculations of offsite dose were based on an elevated release with the ventilation working. This analysis does not include a scenario that could lead to ground release which is typically included in TRIGA MHA dose calculations. In addition, an elevated release can only be used if the release point is 21/2 times greater than the height of the adjacent solid structures, or higher (see RG 1.1.45); no statement is made concerning the applicability of this assumption to GSTR. GSTR is requested to provide the following:
o There is no explanation of the HVAC system in SAR Section 9.1.3 including differentiating between normal exhaust and emergency exhaust. SAR Figure 9.1 refers to a "filtered exhaust" that employs a HEPA filter. However, in the MHA analysis, the release is assumed to be instantaneous with no HEPA filtration, or decay of fission product gases that were released into the reactor bay. Please clarify the assumptions used in the accident analysis regarding the HVAC system (e.g., normal ventilation or emergency ventilation mode of operation).
o Please include in your revised response the public dose estimates assuming a ground release, or clarify why such estimates are not required.
o Please include in your revised response a justification for using the assumption of elevated release.
o Because the results of HotSpot calculations are input dependent, please provide the complete input scenario along with the source term used for all HotSpot calculations.
o Please provide doses estimates for adjacent or nearby offices, where non-involved workers could be present or clarify why such estimates are not required.
o If a decision is made to use all possible modes of HVAC operation, then evaluate corresponding occupational and public doses for all such modes and demonstrate that regulatory requirements of 10 CFR Part 20 are satisfied. Please state clearly all assumptions such as actuation speed, manual activities required, flow rates, damper conditions, fan conditions, etc.
o RAI 17.1 response provides a distance to the fence line of 968 feet, as opposed to 350 meters in RAI 15.3 response. Please provide dose calculations based on consistent distances, or explain the differences.
16.1 Explain excess reactivity insertion analysis.
The value of the fuel temperature coefficient cited in SAR Table 13.7 was a linear function. As can be seen in Figure 1 below, General Atomics (GA) and NRC staff confirmatory analysis shows that this function was not linear.
The GSTR linear function provides additional negative reactivity feedback at elevated temperatures that is not consistent with the GA or NRC staff confirmatory analysis. Please justify the use of the GSTR linear fuel temperature coefficient or provide a revised the fuel temperature coefficient.
8.5 8.5 8.5 8.5 8.5 8.5 8.5 12 12 12 12 12 12 12 G
G G
G G
G
-0.030
-0.028
-0.026
-0.024
-0.022
-0.020
-0.018
-0.016
-0.014
-0.012
-0.010
-0.008
-0.006 0
200 400 600 800 1000 Fuel Temperature (°C)
($/°C)
GA-7882 SS 8.5 8.5-20-0.
12 12-20-0.
G GSTR Figure 1: Comparison of GA, NRC staff confirmatory calculation, and GSTR fuel temperature coefficients.
16.2 Provide analysis for 8 and 8.5 wt
% fuel.
Please provide an explanation of why the 12 wt% fuel provides the limiting results for fuel used in GSTR.
17.1 Provide details of the dose calculations.
Please provide the following:
The parameters used in determining the scattered dose at the fence line location (about 259 meters from the center of reactor bay). This should include all data and calculations with and without the credit for the 1 ft concrete wall of the reactor bay.
In response to RAI 15.3, the fence line distance to the reactor bay is identified as 350 meters, as opposed to 968 feet in the response to RAI 17.1. Provide dose calculations based on a consistent value of the distance to a member of the public.
The analysis of offsite dose was limited to one location beyond the GSTR fence line, and did not consider locations within the owner controlled area between the fence line and the reactor bay (e.g.,
parking lot, office locations, etc) where individuals (members of the public) could be exposed. Please provide an analysis of the potential radiation exposure to individuals between the fence line and the reactor bay. Incorporate any assumptions, as described in the GSTR emergency plan, concerning evacuation of individuals from the owner controlled area.
The NRC staff notes what seems to be a typo in the definition of as 0/cm.
23.1 Criteria for significant change in core configuration.
The response to this RAI does not provide criteria for determining what constitutes a significant change in core configuration. Please provide and justify the criteria chosen.
24.3 Basis for the SDM Value The proposed SDM of -$0.30 is less than the guidance provided in NUREG-1537 (-$0.50). That guidance is predicated on the licensees ability to be capable of accurately measuring reactivity +/-$0.50. To justify the SDM of -$0.30, please explain or demonstrate USGSs ability to consistently discern a reactivity change of this magnitude.
24.9 No calculational reference for values provided.
The NRC staff observed that the input to COMPLY has resulted in a failed calculation for screening level 1, and that for screening level 2, the RAI response provided an input of 2.266e-6 Ci/sec. The NRC staff cannot reproduce this number using the technical specification values for allowed release concentration, volume, and flow rates. Furthermore, even if this numerical value is correct, the NRC staff notes that the calculated exposure from COMPLY is 0.5 mrem, and not 5 mrem, the value in the technical specifications. Please provide an RAI response that demonstrates that the technical specification limit on release concentrations is justified by the statements in the basis.
26.d Provide details of Ar-41 dose calculations in unrestricted area.
The use of an acute airborne release model (HotSpot) for chronic airborne release analysis does not appear applicable as HotSpot Version 2.07 was intended to estimate doses from short-term airborne releases of less than a few hours following nuclear accidents and significant releases, and was not designed to estimate doses from annual releases. It does not provide conservative best estimates of the annual dose to individuals. Further, the input parameters shown in the response include deposition velocities for large and small particle material, while Ar-41 (as a noble gas) is not associated with particles. If computational analysis is desired then CAP-88 PC (the U.S. EPA COMPLY software) that is commonly used for clean air act compliance calculations is appropriate. Alternatively, Regulatory Guide 1.111 provides another acceptable method. Please provide dose calculations using acceptable code.
26.e Provide the basis for SAR equation 11.31.