ML13190A052
| ML13190A052 | |
| Person / Time | |
|---|---|
| Site: | U.S. Geological Survey |
| Issue date: | 07/15/2013 |
| From: | Geoffrey Wertz Research and Test Reactors Licensing Branch |
| To: | Adrian B US Dept of Interior, Geological Survey (USGS) |
| Wertz G | |
| References | |
| TAC ME1593 | |
| Download: ML13190A052 (10) | |
Text
July 15, 2013 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 THE U.S. NUCLEAR REGULATORY COMMISSION 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 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, submitted by USGS letter dated May 17, 2013 (a redacted version is available in ADAMS Accession No. ML13162A662), to our request for clarification sent by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13059A027),
and has identified, in the attached table, additional clarification that is needed. Please provide your responses to the enclosed table 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,
/Alexander Adams, Jr. 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
Enclosure:
RAI 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
ML13190A052 *concurrence via e-mail NRR-088 OFFICE NRR/DPR/PRLB/PM*
NRR/DPR/PRLB/LAIT NRR/DPR/PRLB/LA NRR/DPR/PRLB/BC NRR/DPR/PRLB/PM*
NAME GWertz PBlechman GLappert AAdams GWertz (AAdams for)
DATE 7/10/13 7/10/13 7/10/13 7/15/13 7/15/13
Enclosure OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION RENEWAL OF THE FACILITY OPERATING LICENSE FOR THE UNITED STATES GEOLOGICAL SURVEY TRIGA REACTOR LICENSE NO. R-113; DOCKET NO. 50-274 The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the United States Geological Survey (USGS) staffs responses to our requests for additional information (RAIs) for the Geological Survey TRIGA Reactor (GSTR) provided by USGS letter dated May 17, 2013 (a redacted version is available in Agencywide Documents Access and Management System (ADAMS) Accession No. ML13162A662). The NRC staff has identified several RAI responses that need additional clarification as described in the attached table. Please provide your responses to the enclosed table within 45 days of the date of this letter. The page number in the brackets
[ ] below references the GSTR information provided by the USGS letter dated May 17, 2013.
RAI No.
Original RAI Information Needed 9
Describe the limiting core configuration (LCC).
Provide the peak fuel temperature for aluminum clad fuel element in the operating core (OC) (Figure 1) and LCC (Figure 2). [page 1]
Describe any restrictions on the locations of any aluminum clad fuel elements based on the peak fuel temperature in either the OC or LCC. [page 1]
The NRC staff has independently calculated the shutdown reactivity of the OC configuration using the information provided in the response to RAI No. 9 (excess reactivity $4.84; shim1, shim2, and the transient control rod worths, respectively, as -$2.16, -$2.25, -$2.06). The NRC staff calculations result in a shutdown reactivity of -$1.63. Explain the USGS calculated result of -$1.75. [page 4]
Provide numerical data for information from Figure 7. [page 7]
12 Describe the departure from nucleate boiling ratio (DNBR) analysis.
The NRC staff calculated a flow area per rod of 4.776 square centimeters (cm2) for the hot fuel element and 5.981 cm2 for the average fuel element based on the core plate figure provided in your RAI response dated November 24, 2010 (ADAMS Accession No. ML103340090). These values differ significantly from the USGS RAI response values of 5.855 cm2, and 8.382 cm2, respectively. The NRC staff calculation results are provided in Table 1. Provide details of your calculations. [page 8]
The NRC staff confirmatory analysis of the LCC at the limited safety setting system steady state power of 1.1 megawatts (MW) using the TRACE computer code indicates significant flow oscillations occur when the core inlet temperature exceeds 42 degrees Celsius (C). Discuss the results of the GSTR RELAP analysis including any observed flow instabilities, and its affect on fuel temperature over the range of the core inlet temperature up to 60 degrees C. [page 9]
NRC staff performed their confirmatory analysis of the LCC using a closed loop (CL) model which produced coolant flow velocities that were consistent with independent analysis and measurements from similar research reactors and power levels and, consistent with the NRC analytical solutions for single phase natural circulation flow (see Figure 1). The NRC staff has also investigated the use of other modeling techniques (fill & break and large tanks) and determined that the coolant flow velocities were too high as compared to the CL benchmarks. The NRC staffs concern is that models using coolant flow velocities higher than actual may result in a non-conservative estimate of the DNBR. NRC staff review found that the GSTR flow velocities were approximately 30 percent higher than NRC staff results and the results of models for other facilities (see Figure 1). Please justify the use of the GSTR model given the apparent variance shown in Figure 1. [page 9]
Provide the maximum fuel temperature of the aluminum clad fuel element for the pulsing analysis.
[page 9]
The temperatures in Table 3 do not match the temperatures provided in Figure 7. Explain the differences. [page 10]
The results of the Characterization of the response of GSTR to a reactivity pulse and an uncontrolled rod withdrawal transient event did not provide peaking factors or explain their use. Provide the peaking factors, and explain their use in the analyses (i.e., were they included in the model or applied to the results obtained from a model representing the average core or channel). [page 10]
For the uncontrolled rod withdraw analysis, explain if the withdrawn control rod continues until fully withdrawn, or if it inserts upon receiving the scram signal. [page 11]
GSTR provided a temperature of 1020 degrees Kelvin (K) and then converted it to 727 degrees C. The NRC staff converted 1020 degrees K to 747 degrees C. Explain the difference. [page 11]
15.3 Explain the methods used to determine the maximum hypothetical accident (MHA) doses The GSTR response to RAI No. 15.3 provided a volume-averaged fuel temperature of 423.2 degrees C.
The NRC staff confirmatory analysis calculated a significantly lower average fuel temperature. In order for NRC staff to fully understand the GSTR fuel temperature, provide an explanation of the fuel element nodal structure and a description of the nodal fuel temperature calculations for the volume-averaged and peak fuel temperatures. [page 13]
The NRC staff has completed confirmatory analysis of the MHA results using HOTSPOT 2.07.2, and was unable to reproduce the GSTR dose results for the atmospheric release scenario at the Emergency Assembly Area and the Building 21 East Entrance using the assumptions stated in the response to RAI No. 15.3 (see shaded area in Table 2 for NRC staff confirmatory calculations). Provide details of the total effective dose equivalent (TEDE) dose calculations at these locations. [page 22]
24.3 Basis for Shutdown Margin Definition The GSTR analysis provided in response to RAI No. 9 appears to establish the GSTR Core Excess Reactivity for Technical Specification (TS) 14.3.1.1.2 as $6.18. Using the GSTR control rod worths as provided in the GSTR response to RAI No. 9, the NRC staff confirms that the shutdown reactivity of the GSTR using the stuck control rod criteria would be -$1.20 which satisfies the existing TS shutdown margin specification of -$0.55 (TS 14.3.1.1.1). The proposed GSTR TS 14.3.1.1.1, Shutdown Margin, of
$0.30 appears to unnecessarily reduce the shutdown margin TS requirement. Advise if GSTR still seeks to establish TS 14.3.1.1.1, Shutdown Margin, of $0.30, and the basis for this reduction given the analyzed core excess reactivity of $6.18. [page 24]
Table 1 USGS Confirm.
Source diamter radius B-Ring Distance from Center (cm) 4.053 7.9772 cm ML103340090 6.2813 3.1406 in C-Ring Distance from Center (cm) 7.981 3.6513 cm ML103340090 2.8750 1.4375 in Inner Radius of Flow Channel (cm) 2.026 1.9050 cm SAR 10.2.1 1.5 0.75 in Outer Radius of Flow Channel (cm) 6.017 5.8142 cm (7.9772-3.6513)/2 + 3.6513 Fuel Element Cross Section (cm2) 10.949 11.0241 cm2 SAR Table 4.10 1.475 0.7375 in Wetted Perimeter of a Single Fuel Element (cm) 11.73 B-Ring Total Flow Area (cm2) 35.132 28.6567 cm2 area using outer radius-area of inner radius-6*area of FE Flow Area/Rod (cm2) 5.855 4.7761 cm2 calc Effective Hydraulic Diameter (cm) 1.977
- Elements 110 127.0000 lattice positions
4 1 TR and 3 CR Element Cross Section (cm2) 10.949 11.0241 cm2 using 1.475 in OD Control Rod Cross Section1 (cm2) 9.58 9.5799 cm2 using 1.375 in FF OD Transient Rod Cross Section (cm2) 7.917 7.9173 cm2 using 1.25 in OD Central Thimble Cross Section (cm2) 11.401 11.4009 cm2 using 1.5 in OD Core Cross Section (cm2) 2208.99 2140.6389 cm 2
plate diameter + 1 FE pitch (1.7658 in)
Total Flow Area (cm2) 824.209 747.6463 cm2 total-CT-TR-3CR-122FE Area/Rod2 (cm2) 8.382 5.9812 cm2 total FA/125 (122 FE + 3 FF)
Hydraulic Diameter (cm) 2.858 Figure 1 0.14 0.19 0.24 0.29 0.34 0.39 0.44 10 20 30 40 50 60 70 Bulk Water Temperature (°C)
Hot Channel Velocity (m/s)
Conf. CL ANL/WSU 1MW ANL/OSU 1MW ANL/TAMU 1MW USGS Conf. F&B Analytical Table 2 Location Distance (meter)
Radiation Dose to Member of Public NRC Confirmatory Calculation (GSTR SAR Table 13.6)
TEDE (mrem)
Building 15 south door 11 53 (53)
Emergency assembly area 32 10 (6.1)
Building 21 east entrance (West of Building 15) 49 7.8 (4.3)
Average of eastern intersections 100 2.9 (2.4)
Building 16 west entrance 175 1
(0.96) 200 0.8 (0.76) 250 0.52 (0.5)
Nearest Unrestricted Location 475 0.15 (0.14)
Residence 640 0.082 (0.081)
School 720 0.065 (0.065)