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MONTHYEARML0620600262006-10-0303 October 2006 Request for Additional Information Regarding License Renewal, Oregon State University Triga Reactor Project stage: RAI ML0632101822006-11-10010 November 2006 Oregon State University Triga Reactor (Ostr), Request for Extension 60 Days to Respond to the Request for Additional Information Regarding License Renewal Project stage: Request ML0633205002006-11-21021 November 2006 Oregon State University Response to Request for Additional Information Regarding License Renewal, Osu Triga Reactor, Dated October 3, 2006 Project stage: Response to RAI ML0713000102007-05-21021 May 2007 Oregon State University - Request for Additional Information Regarding License Renewal Request Project stage: RAI ML0714503072007-06-0707 June 2007 Determination of Acceptability and Sufficiency for Docketing and Opportunity for a Hearing Regarding the Application from the Oregon State University for Renewal of the Facility License for the Oregon State University Triga Reactor (Ostr) Project stage: Other ML0723305292007-06-25025 June 2007 Oregon State University Triga Reactor, Request for Extension of 45 Days to Respond to the RAI Re License Renewal, Dated May 21, 2007 Project stage: Request ML0721503622007-07-10010 July 2007 Oregon State University Response to RAI Regarding License Renewal, Oregon State University Triga Reactor Dated May 21, 2007 Project stage: Response to RAI ML0721503632007-07-27027 July 2007 Additional Oregon State University Response to RAI Regarding License Renewal, Oregon State University Triga Reactor Dated May 21, 2007 Project stage: Response to RAI ML0721900432007-07-31031 July 2007 Oregon State University Triga Reactor (Ostr), Decommissioning Funding Statement of Intent Project stage: Other ML0723405802007-08-0606 August 2007 Response to Request for Additional Information (RAI) Regarding License Renewal, Oregon State University Triga Reactor Dated May 21, 2007 Project stage: Response to RAI ML0807101392008-03-19019 March 2008 Oregon State University - Request for Additional Information Regarding License Renewal Request Project stage: RAI ML0822703832008-08-11011 August 2008 Oregon State University Submission of Relicensing Technical Specifications Revision J Project stage: Request ML0616501972008-09-0202 September 2008 Oregon State University Triga Reactor Environmental Assessment Regarding Amendment for Renewal of Facility Operating License No. R-106 Project stage: Other ML0825305092008-09-10010 September 2008 Technical Specifications for the Renewed Facility License No. R-106 for the Oregon State University Triga Reactor Project stage: Other ML0825200472008-09-10010 September 2008 Issuance of Renewed Facility License No. R-106 for the Oregon State University Triga Reactor Project stage: Approval ML0830203032008-10-30030 October 2008 Massachusetts Institute of Technology-Request for Additional Information Regarding Amendment Request Project stage: RAI ML0832204262008-11-0707 November 2008 Massachusetts Institute of Technology - Response to Request for Additonal Information Regarding Amendment Request Project stage: Request ML0831801352008-11-0707 November 2008 Massachusetts Institute of Technology - Response to Request for Additional Information Regarding Amendment Request Project stage: Response to RAI ML0907506442009-04-24024 April 2009 Massachusetts Institute of Technology - Issuance of Amendment No. 38 Number of Heat Exchangers Required by Technical Specifications Project stage: Approval 2007-07-31
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MONTHYEARIR 05000243/20232012023-08-24024 August 2023 Oregon State University U.S. Nuclear Regulatory Commission Safety Inspection Report No. 05000243/2023201 ML23143A3132023-07-18018 July 2023 Examination Result Letter No 50-243/OL-23-01, Oregon State University ML23047A5482023-03-0808 March 2023 Examination Confirmation Letter No. 50-243/OL-23-01, Oregon State University ML22199A3092022-07-26026 July 2022 Oregon State University - Issuance of Amendment No. 26 to Renewed Facility Operating License No. R-106 for the Oregon State Triga Reactor ML22102A2842022-04-28028 April 2022 Oregon State University - Review and Approval of the Revised Oregon State Triga Reactor Requalification Plan IR 05000243/20222022022-04-23023 April 2022 Oregon State University - Us Nuclear Regulatory Commission Routine Safety Inspection Report No 05000243-2022202 ML22062B3052022-03-10010 March 2022 Examination Confirmation Letter No. 50-243/OL-22-01, Oregon State University ML22019A2982022-01-19019 January 2022 Oregon State University ,Response to RAI Regarding License Amendment Request, Oregon State University Triga Reactor Dated June 17, 2020 ML21258A0732021-10-22022 October 2021 Oregon State University Request for Additional Information (RAI) Changes ML21299A2502021-10-22022 October 2021 Oregon State University Triga Reactor (Ostr), Docket No. 50-243, License No. R-106 ML21250A3432021-09-20020 September 2021 Oregon State University Closure of Confirmatory Action Letter No. NRR-04-001 on the Site-Specific Compensatory Measures Implementation Plan ML21236A3462021-08-24024 August 2021 Oregon State Univ., License Amendment Request to Revise Requalification Plan for Licensed Operators ML21225A0832021-07-29029 July 2021 Oregon State Univ., Modifications to the OSTR Physical Security Plan (PSP) to Close Out Confirmatory Action Letter No. NRR-04-001 IR 05000243/20212012021-06-0303 June 2021 Oregon State Routine Inspection Report 05000243/2021201 ML21141A2892021-05-21021 May 2021 Oregon State University Triga Reactor (OSTR) - Amendment to Letter Dated June 17, 2020, License Amendment Request to Remove Technical Specification Requirements Related to Instrumented Fuel. ML21119A2362021-04-22022 April 2021 Oregon State University Triga Reactor (Ostr), Follow Up Report of Event Notification 55184 ML21092A1372021-04-0909 April 2021 Oregon State University - Regulatory Audit Re License Amendment Request to Remove Technical Specification Requirements ML20318A3802020-12-18018 December 2020 Oregon State University - Temporary Exemption from the Requirements of 10 CFR Part 50, Appendix E, Section IV.F.2.b Related to Biennial Emergency Exercise (Covid 19) ML20350B7262020-12-10010 December 2020 Extension of Timeframe Required to Complete Biennial Emergency Exercise Per Oregon State Triga Reactor Emergency Response Plan, Facility Operating License No. R-106 ML20338A1302020-11-25025 November 2020 Oregon State University, Change in Level 1 Personnel Update ML20295A4242020-11-16016 November 2020 Notification of Mailing Address Change Regarding Submittal of Fingerprint Cards ML20318A0352020-11-0303 November 2020 Extension of Timeframe Required to Complete Biennial Emergency Exercise Per Oregon State Triga Reactor Emergency Response Plan, Facility Operating License No. R-106 ML20307A4132020-10-22022 October 2020 Oregon State University Triga Reactor (OSTR) ML20247J5482020-09-0303 September 2020 Examination Confirmation Letter No. 50-243/OL-21-001, Oregon State University IR 05000243/20202012020-04-20020 April 2020 Oregon State University - Nuclear Regulatory Commission Routine Inspection Report No. 05000243/2020201 ML19298A4412019-10-21021 October 2019 Oregon State University - Radiation Center and Triga Reactor Annual Report 07/01/2018 - 06/30/2019 ML19290D8132019-10-14014 October 2019 Oregon State University, 10 CFR 50.54(q) Report of Changes to the Ostr Emergency Response Plan Which Do Not Decrease the Effectiveness of the Plan ML19281C3412019-10-0202 October 2019 Oregon State University Triga Reactor (Ostr) 10 CFR 50.54(q) Report of Changes to the Ostr Emergency Response Plan Which Do Not Decrease the Effectiveness of the Plan ML19199A1632019-07-24024 July 2019 Examination Report No. 50-243/OL-19-01, Oregon State University ML19210B8882019-07-22022 July 2019 Oregon State University - Transmittal of Reply to Notice of Violation ML19011A3502019-06-18018 June 2019 Oregon State University Issuance of Amendment Number 25 to Renewed Operating License No. R-106 for the Oregon State Triga Reactor, Technical Specifications ML19098A1672019-04-0303 April 2019 Oregon State University Triga Reactor (Ostr) - Modification of Ostr Physical Security Plan, Docket No. 50-243, License No. R-106 ML19065A0512019-02-28028 February 2019 Oregon State University - Supplement to Letter Dated of January 24, 2019, License Amendment Request to Remove Requirement for Fuel Temperature Measuring and Safety Channels While Transient Operation Modes Are Precluded IR 05000243/20192012019-02-15015 February 2019 Oregon State University - U.S. Nuclear Regulatory Commission Inspection Report No. 50-243/2019-201 ML19052A0512019-01-24024 January 2019 Oregon State Univ. Tiga Reactor (Ostr) - Replacement of November 5, 2018, License Amendment Request to Remove Requirement for Fuel Temperature Measuring and Safety Channels While Transient Operation Modes Are Precluded Letter of ... ML19009A0912019-01-0404 January 2019 Oregon State University - Proposed New Basis for the Limiting Safety System Setting ML18347A2112018-12-0707 December 2018 Oregon State Univ., 10 CFR 50.54(q) Report of Changes to the Ostr Emergency Response Plan Which Do Not Decrease the Effectiveness of the Plan ML18334A1002018-11-26026 November 2018 Oregon State University - Withdrawal of License Amendment Request from October 30, 2018 and Amendment to License Amendment Request from November 5, 2018 ML18312A0612018-11-0505 November 2018 Oregon State University Triga Reactor (Ostr) - License Amendment Request to Remove Requirement for Fuel Temperature Measuring and Safety Channels While Transient Operation Modes Are Precluded ML18312A0602018-10-30030 October 2018 Oregon State University - License Amendment Request to Receive a 20/20 Instrumented Fuel Element ML18297A0512018-10-19019 October 2018 Oregon State University, Request for Removal of Mr. Todd Keller as a Reviewing Official ML18297A1202018-10-19019 October 2018 Oregon State University Triga Reactor (Ostr) - Annual Report for the Period July 1, 2017 Through June 30, 2018 ML18297A0522018-10-19019 October 2018 Oregon State University, Change in Level 1 Personnel Update ML18101A0352018-04-0505 April 2018 Oregon State University - Modification of Physical Security Plan IR 05000243/20172022018-02-20020 February 2018 Oregon State University - Nuclear Regulatory Commission Routine Inspection Report 50-243/2017202 ML17352A4022017-12-13013 December 2017 10 CFR 50.54(q) Report of Changes to the Oregon State University Triga Reactor Emergency Response Plan Which Do Not Decrease the Effectiveness of the Plan ML17353A2292017-12-13013 December 2017 Modification of Oregon State University Triga Reactor Physical Security Plan ML17305A0572017-10-30030 October 2017 Oregon State University - Submittal of Annual Report for the Period July 1, 2016 Through June 30, 2017 ML17193A1612017-07-12012 July 2017 Examination Report No. 50-243/OL-17-01, Oregon State University Nuclear Radiation Center ML17114A2122017-05-0505 May 2017 Oregon State University - Nuclear Regulatory Commission Security-Related Inspection Report No. 50-243/2017-202 2023-08-24
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Radiation Center Oregon State University, 100 Radiation Center, Corvallis, Oregon 97331-5903 T 541-737-2341 1 F 541-737-0480 1 http://ne.oregonstate.edu/facilities/radiation_center Oregon State UNIVERSITY V -t 0, l V P July 27, 2007 ,,a, c Mr. Alexander Adams Of h U. S. Nuclear Regulatory Commission Research and Test Reactors Branch A Office of Nuclear Reactor Regulation Mail Stop 012-G13 One White Flint North 11545 Rockville Pike Rockville, MD 20852-2738
Reference:
Oregon State University TRIGA Reactor (OSTR)Docket No. 50-243, License No. R-106 Request for Additional Information (RAI) Regarding License Renewal, Oregon State University TRIGA Reactor (TAC NO. MC5155) dated May 21, 2007
Subject:
Additional Oregon State University Respornse to RAI Regarding License Renewal, Oregon State University TRIGA Reactor dated May 21, 2007 Mr. Adams: In a letter dated May 21, 2007, the U.S. Nuclear Regulatory Commission (NRC)requested that Oregon State University (OSU) provide additional information with regards to the OSU license renewal application of October 5, 2004, as supplemented.
Enclosed is a further OSU response to the request. If you have any questions, please call me at the number above. I declare under penalty of perjury that the foregoing is true and correct.Executed on: *7/z- ?.Sincerely, Steve Reese Director Enclosure cc: Document Control, NRC Al Adams, NRC Craig Bassett, NRC John Cassady, OSU Rich Holdren, OSU Todd Palmer, OSU , Mike Hartman, OSU d020)
Oregon State University Further Responses to RAI Letter of May 21, 2007 Section 3.4. What is the relationship between the UBC 1964 Zone 3 seismic requirements and the maximum ground accelerations given in Table 2-4?Unfortunately, because the accelerations listed in Table 2-4 are peak ground accelerations and accelerations of the 1964 UBC method are building design level accelerations, one cannot do a direct comparison without further calculation.
The 1964 UBC does not have a method to calculate peak ground accelerations to compare to Table 2-4, but the building design level forces can be calculated using 2006 IBC criteria to provide a basis of comparison with the 1964 UBC method.The criteria calculated using the 1964 UBC results in a building design level acceleration.'
Design level acceleration is a scaled form of acceleration that includes methods to account for the building response characteristics.
This scaling is intended to give forces that are comparable with those observed in actual events and testing. The different Codes have used methods of different sophistication for this scaling. The 1964 method was very rudimentary, while the most recent 2006 IBC methods are more sophisticated and include factors to account for soil response characteristics and the hazard to the public of the building occupancy which were not in the 1964 method. A comparison of the design level forces follows.The 1964 UBC method is a single prescriptive formula that results in an acceleration that is intended for use in building element design. The seismic response characteristics of the structure and empirically observed behavior are built into the formulas.
The maps included in the 1964 UBC indicate the area in zone 2 (zones ranged from 0 to 3, with 3 being the zone of highest seismic concern), however zone 3 was conservatively used. The resulting building design level was calculated to be 0.2W. Using the 2006 IBC and corresponding site specific data now available, the design level acceleration was calculated to be 0.14W. The 2006 IBC value is lower and also significantly more representative.
This means that the facility still conforms to current seismic design level standards.
- 2. Section 4.2.1.9. The discussion of fuel swelling at high burnups refers to the agglomeration of fission gases at room temperatures above 1300'F.This swelling is time and temperature dependent.
Provide a discussion if there should be a steady state temperature limit to control this type of swelling.
Reference
4.1 shows
that the swelling as a function of time increases with increasing time and increasing temperatures.
The tendency for the curve to flatten out as temperatures decrease would suggest that swelling at our current steady-state operating temperature of approximately 350'C would be minimal, if present at all. The same conclusion could also be reached from the correlation between swelling and temperatures at end of life. However, the limited safety system setting for fuel temperature at 510'C is sufficiently below the 705'C temperature referenced that this type of swelling is precluded.
- 3. What is the maximum fuel element power for possible core loadings (#8)?What are the peaking factors? Tables 4-11 and 4-12 only contain average power per element.The power per fuel element was calculated using MCNP5 with a model for a core similar to core #8 at a power of 1.1 MW. The new (MCNP5) model represents the reference HEU core (i.e., the original core loaded in 1976) and contains 4 less elements than core #8. The MCNP5 model for the reference HEU core was found to have very good predictive capability over a wide range of reactor conditions and was deemed to be well-suited for performing neutronics calculations.
The results of this model can be seen in the following table: OU9o 0 eeei 4. TS 2.2. Discuss the derivation of the limited safety system setting (LSSS)value of 510"C. Discuss how LSSS protects fuel from exceeding the safety limit considering issues such as instrumented fuel element placement in the core versus the core hot spot, the thermocouple placement in the instrumented fuel element versus the fuel element hot spot, the accuracy of the measuring instrumentation and transient behavior of the reactor safety system.
The value of the LSSS is designed to protect thee& fuel from exceeding the maximum fuel temperature safety limit (SL) of 1,150°C forqfuel during non-pulsing reactor operation.
It is not applicable to pulsing operations.
The value of the LSSS at 510'C was conservatively chosen to be slightly lower than half the SL to account for uncertainties in measurement.
Based upon the analysis described in the answer to question three above, the highest power per fuel element location for the most reactive core was found to be in locations B-3 at a value of 11kW. The Instrumented Fuel Element (IFE) located in grid position B-4 has a calculated power per fuel element ofý kW. This is a difference of approximately 3.5%. The IFE is calibrated annually.
Experience has shown that 5% error or less in the true temperature is commonly observed.
Thus, the calculated difference in the power per elements is well within the error of the measurement.
Axial flux measurements were made for the In-Core Irradiation Tube (ICIT)core configuration.
The results of these measurements are shown in the figure below along with the locations of the three thermocouples within the IFE.The error associated with the flux measurements are slightly larger than the symbols used. The thermocouple at the lowest elevation (which reads highest during steady state operation) is exposed to flux levels no more than 6% less than peak axial flux levels. Per Figure 7.2 in the SAR, the thermocouples are located in the fuel meat halfway between the outside edge of the fuel and the inside edge of the ZrH rod on the interior of the fuel, 0.425 inch from fuel center.ICIT Core Measured Axial Flux Distribution 1.2E13 4A:0TC Locations 1 .2E+13 X 1.OE+13 8.0E+12 C14< 6.OE+12 Q 4.OE+12 2.0E+12 O.OE+00 0 10 20 30 40 50 Height (cm)Typical fuel temperatures observed at full power are approximately 350'C.The analysis in section 13.2.2.2.1 shows that an uncontrolled withdrawal of a control rod at an initial power level of 1 MW would result in a trip signal being initiated within 0.28 seconds resulting from a reactivity insertion of
, ..4$0.15. For an uncontrolled withdrawal of a control rod at an initial power level of 100 W, the trip signal would be initiated in 5.06 seconds resulting from a reactivity insertion of $1.06. Because fuel temperature lags behind power and the power is so low, each of these scenarios would result in high power trips before the fuel temperature trip is reached. This is confirmed by our experience of observed instrument behavior after a pulse. For the loss of coolant accidents described in section 13.2.3, the primary water temperature would trip the reactor or the low level alarm would annunciate and alert the operator long before enough water is lost to initiate a high fuel element temperature trip. Regardless, section 13.2.3.2.2.1 clearly shows that natural convective air cooling of the fuel will keep the maximum fuel temperature well below the SL even after an instantaneous complete loss of primary water at 1.5MW or below.The TC is located 0.3 inches from the edge of the central zirconium pin, thus giving it an axial position of 0.425 inches. The actual temperature distribution in the fuel pin will be quadratic (convex downward).
Assuming a water temp of 30'C, no temperature drop across the clad, temperature at the TC of 350'C, and using a parabolic curve fit, the temperature at the inner radius of the fuel is calculated to be: T(r) = 501.4-838r 2 T(0.425) = 488.3°C The last step needed is to show that given all the uncertainties, measurement errors and biases that the LSSS is sufficient to guarantee that the SL is never exceeded.
In other words calculate (nominal IFE temperature) x (max reactor power / nominal reactor power) x (max power per element in the B-ring / nominal IFE power) x (peak azimuthal flux / nominal azimuthal flux) x (temperature measurement uncertainty factor)(350'C) x (1.10) x (1.04) x (1.40) x (1.05) = 589°C For a nominal IFE temperature of 510'C (i.e., the LSSS), the calculated value is 858°C. Both of these values are significantly less than the 1150'C. This assures that the IFE can adequately protect the SL during steady-state operations.