ML072340580
| ML072340580 | |
| Person / Time | |
|---|---|
| Site: | Oregon State University |
| Issue date: | 08/06/2007 |
| From: | Reese S Oregon State University |
| To: | Alexander Adams NRC/NRR/ADRA/DPR/PRTA |
| References | |
| TAC MC5155 | |
| Download: ML072340580 (7) | |
Text
Radiation Center Oregon State University, 100 Radiation Center, Corvallis, Oregon 97331-5903 T 541-737-2341 I F 541-737-0480 http://ne.oregonstate.edu/facilities/radiationcenter Oregon State UNIVERSITY August 6, 2007 Mr. Alexander Adams 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 Response 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 in regards to the OSU license renewal application of October 5, 2004, as supplemented. On July 27, 2007, we submitted a letter and enclosure that involved answers to the RAI request. We respectfully request that this letter and enclosure supersede that letter and enclosure (i.e.,
the letter sent by us to you dated July 27, 2007) in its' entirety. The need to update these answers is based upon discussions with you on July 30, 2007, and information obtained since thedate of the letter. 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:_____
Sincerely, Steve e es~se Director Enclosure cC:
Document Control, NRC Rich Holdren, OSU Al Adams, NRC Todd Palmer, OSU Craig Bassett, NRC Mike Hartman, OSU John Cassady, OSU
,4oc
Oregon State University Further Responses to RAI Letter of May 21, 2007
- 1.
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 I1BC 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 13000F.
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 most conservative maximum fuel temperature estimate of 566°C for a nominal IFE temperature of 350'C (see answer to question #4 below) is sufficiently below the 705'C temperature referenced such 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. Since 1976, five elements have been removed from the core periphery and the core configuration has changed. This will have an effect of increasing the power per element by approximately 250 W, but not more than 1 kW. The results of this model can be seen in the following table:
- 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 thejfuel from exceeding the maximum fuel temperature safety limit (SL) of 1,150'C formif fuel during non-pulsing reactor operation. It is not applicable to pulsing operations. The value of the LSSS at 51 0°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~l~kW. The Instrumented Fuel Element (IFE) located in grid position B-4 has a calculated power per fuel element of MkW.
This is a difference of approximately 4%. 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 highest elevation (which reads lowest during steady state operation and is therefore the most conservative value) is exposed to flux levels approximately 18% less than peak axial flux levels.
ICIT Core Measured Axial Flux Distribution TC Locations 1.2E+13 X
.OE+13 U) 8.OE+12
< 6.OE+12 Q
r 4.02+12 2.OE+12 IPSE S
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
$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. Thisis.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 0.175 inches from the edge of the central zirconium pin.
The zirconium pin has radius 0.125 inches, thus giving the TC a radial position of 0.300 inches. Given a uniform power density, the temperature distribution in a fuel rod would be a quadratic function. In actuality, the neutron flux and therefore power density drops off rapidly inside the fuel.
Therefore the actual temperature profile will be flatter between the TC and the zirconium pin than predicted by a quadratic temperature distribution. This will cause the actual maximum fuel temperature in the fuel pin to be lower than the maximum temperature predicted by a quadratic distribution. If it is assumed that pool temperature is 30'C, there is no temperature drop across the clad, and temperature at the TC is 350'C, then using a quadratic (conservative) curve fit, the temperature distribution within the fuel meat is.
calculated to be: T(r) = 411 - 677r 2. At the inner radius of the fuel, temperature is calculated to be: T(O. 125) = 400'C.
The last step needed is to show that given all the uncertainties, measurement errors and biases that the LSSS is sufficiently conservative 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 radial temperature / radial temperature at the TC) x (temperature measurement uncertainty factor) x (peak axial flux / axial flux at the TC)
(350'C) x (1.10) x (1.04) x (1.14) x (1.05) x (L.18) = 566'C For an indicated IFE temperature of 51O0 C (i.e., the LSSS), the calculated maximum fuel temperature is, 824°C. This value is significantly less than the SL of 1150'C. This assures that the LSSS of 51O0 C as measured at the IFE maintains fuel temperature below the SL during steady-state operations.
- 5.
Section 13.2.1.1. For the Maximum Hypothetical Accident Scenarios B and C, is exposure from building shine from the source term inside the reactor room considered? If not, discuss why this is not a significant
contributor to dose outside the reactor room. [This replaces question 33 in our letter submitted July 10, 2007.1 Microshield version 5.05 was used to determine the dose rates from each isotope in Table 13.1 with and without pool water. The dose rate at each distance for each isotope was used in the following equation to determine the total dose:
Dose = z Dose ratei x (Ie"'t' )
tst = stay time of personnel and 2ý, =f + X, where X1 is the decay constant for the ith isotope and X \\is the ventilation constant for the appropriate scenario.
The initial dose is at t = 0 and the other is for the exposure time for each scenario. The thyroid dose was calculated by multiplying the total by 0.03.
RX Bay Volume Source Scenario A (Microshield)
I1 I
Distance trom Wall Location (in) 10oo1 Water (mR)
Thyroid DDE (after 8.52 s)
Whole Body (after 8.52 s) 10 1.16E-07 3.87E-06 50 1.37E-08 4.58E-07 100 3.30E-09 1.1OE-07.
150 1.25E-09 4.16E-08 200 5.79E-10 1.93E-08 250 3.02E-10 1.01E-08 RX Bay Volume Source Scenario A (Microshield)
Distance from Wall Location (m)
W/O Pool Water (mR)
Thyroid DDE (after 8.52 s)
Whole Body (after 8.52 s) 10 2.1OE-07 7.OOE-06 50 2.48E-08 8.27E-07 100 6.01E-09 2.OOE-07 150 2.29E-09 7.64E-08 200 1.07E-09 3.57E-08 250 5.62E-10 1.87E-08
RX Bay Volume Source Scenario B (Microshield)
Distance from Wall Location (in)
Pool Water (miR)
Thyroid DDE (after 14.7 m)
Whole Body (after 14.7 m) 10 1.78E-04 5.93E-03 50 2.1OE-05 7.02E-04 100 5.05E-06 1.68E-04 150 1.91E-06 6.36E-05 200 8.83E-07 2.94E-05 250 4.60E-07 1.53E-05 RX Bay Volume Source Scenario B (Microshield)
Distance from Wall Location (m)
W/O Pool Water (mR)
Thyroid DDE (after 14.7 m)
Whole Body (after 14.7 m) 10 2.25E-04 7.51E-03 50 2.67E-05 8.89E-04 100 6.38E-06 2.13E-04 150 2.40E-06 7.99E-05 200 1.11E-06 3.68E-05 250 5.73E-07 1.91 E-05 RX Bay Volume Source Scenario C (Microshield)
Distance from Wall Location (m)
Pool Water (mR)
Thyroid DDE (after 63.7 h)
Whole Body (after 63.7 h) 10 6.07E-03 2.02E-01 50 7.21E-04 2.40E-02 100 1.68E-04 5.59E-03 150 6.15E'05 2.05E-03 200 2.78E-05 9.28E-04 250 1.42E-05 4.74E-04 RX Bay Volume Source Scenario C (Micros Distance from W/O Pool Water (mR)
Wall Thyroid DDE Whole Body Location (m)
(after 63.7 h)
(after 63.7 h) 10 1.26E-02 4.19E-01 50 1.49E-03 4.98E-02 100 3.44E-04 1.15E-02 150 1.25E-04 4.18E-03 200 5.53E-05 1.84E-03 250 2.77E-05 9.22E-04 shield)