ML23027A211

GE-Hitachi Nuclear Energy Americas, LLC, NRC Ntr License Renewal Audit Questions Set 1 Rev 2

ML23027A211
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Site:	Vallecitos Nuclear Center
Issue date:	01/27/2023
From:	GE-Hitachi Nuclear Energy Americas
To:	Office of Nuclear Reactor Regulation
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ML23027A209	List: ML23027A210 ML23027A211
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Download: ML23027A211 (1)

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Enclosure 1 - GE NTR License Renewal Questions and Responses Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 003 FSAR section 4.4.2 state that the coolant temperature coefficient is calculated based on overall temperature changes and describes determination of the void coefficient by extrapolating the temperature coefficient data. FSAR figure 1319 shows that some void fraction is expected for steady state operation at all nonzero power levels.

a. Provide the experimental data and calculations of the coolant temperature and void coefficients of reactivity.

b. Based on the discussion in this same paragraph, the NTC staff interprets that the coolant temperature coefficient includes effects of both water density change and void buildup, ant that the void coefficient of 5.7¢/% void is only applicable over 124°F.

i. Is the NRC staff interpretation correct? If not, explain the effects included in the coolant temperature coefficient.

ii. Explain over what ranges the reactivity coefficients listed in Table 42 are applicable.

a. According to historical documents from 1959 to 1964, values for temperature coefficient were painstakingly derived by experimentation and measurement and required some modification to the plant that included installation of a 10 Kw heater and the addition and removal of aluminum disk spacers to adjust fuel rod length. While not provided in tabular form, the data was plotted as a figure in an internal 1964 document (NUSA114) produced by the GE Nuclear Safety Analysis Group (see attachment). This resulting relationship provides for the coolant temperature coefficient that is used even today.

b. i. Yes. The NTR coolant temperature coefficient of reactivity, which is described on page 420 of NEDE32740P, Rev. 3, is the result of density changes only, so are converted to a void coefficient (in that this is also a net density change), and thus yields a void coefficient of 5.7 /% void above the temperature coefficient turning point of 124 °F. It is only applicable above the turning point, where the sign of the coefficient will be negative.

ii. The Coefficients of Reactivity listed in Table 42 are applicable for the following ranges:

Temperature coefficient in: Applicable Temperature Range Water coolant Taken as applicable between 60°F and ~200°F. Supporting data taken were in the range ~65°F to ~150°F (internal memo Sager, March 15, 1965.

Inner graphite Calculated as average over: 74°F to 236°F Outer graphite Calculated as average over: 74°F to 236°F Average void coefficient 124°F to ~200°F Doppler coefficient All temperatures 1

Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 003 2

Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 003 The relationship for the coolant temperature coefficient of reactivity is converted to an effective void coefficient of reactivity for temperatures above the turnover point (124°F) where bulk coolant voiding may be possible. While the details of this conversion first performed in the 1960s are not available, the value of 5.7/%void may be demonstrated to be conservative by the following analysis.

The coolant temperature reactivity coefficient, which was previously determined by NTR reactivity measurements (prior to 1965), is given by the expression:

5.7 10 124 Using properties of water at 1 atm to develop a relationship for the change in voids (i.e., net density) with a change in temperature, this can be expressed as:

5.7 10 124 Voids are calculated relative to 124 °F density by using the relationship:

124

% 100 124 The table below provides data of density vs. temperature at 1 atm for water, along with the calculated effective void fraction (in percent), as well as the ratio of the incremental differences in temperature and void, or dT/dV.

T(°F) Density(g/cc) V(%) dT/dV (°F/%)

124 0.98754 0.00 125 0.98729 0.03 39.502 130 0.98597 0.16 37.407 135 0.9846 0.30 36.042 140 0.98319 0.44 35.019 145 0.98173 0.59 33.820 150 0.98023 0.74 32.918 155 0.97868 0.90 31.856 160 0.9771 1.06 31.251 165 0.97547 1.22 30.293 170 0.9738 1.39 29.925 180 0.9704 1.74 29.045 190 0.9668 2.10 27.432 200 0.963 2.48 25.988 Avg 32.35 3

Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 003 Using that relationship, the equivalent voids at 200 °F is ~2.5%.

To apply this relationship and have it represent a reasonable void coefficient across the entire temperature range, the average temperature in this range is chosen. So the above equations are applied using the temperature coefficient of reactivity at 155 °F, and the average of the variation in temperature with effective voids. This results in an evaluation of a void coefficient of:

5.7 10 124 5.7 10 155 124 32.35 5.7 ¢/%

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 004 In section 4.4.3, the FSAR states that manual poison sheets shall be restrained in a manner which will prevent movement by more than 1/2inch relative to the reactor core.

b. How much movement is expected once the poison sheets have been latched in place?

c. Is this movement in the radial or azimuthal direction?

d. Discuss the effect on the reactivity worth of the 1/2inch of movement of these poison sheets.

a. Movement of the manual poison sheets (MPS) is approximately 1mm once locked in place (Figure 2). This latching mechanism was installed by CA 117 in 11/1977. There are 6 available MPS slots, of which only 3 were updated with the locking mechanism. The other 3 are padlocked to prevent usage of those slots.

b. The MPS are configured radially around the core (Figure 1), and only move parallel to the axis of the horizontally configured core (Figure 2 & 3).

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 004 The MPS cannot move radially (increase / decrease distance) away from the core (Figure 1), nor azimuthal (tilting in any of the 3 planes) in relation to the core, nor even circularly (360degree type motion) around the axis of the core (Figure 3). Nor can the MPS itself rotate on its own axis (Figure 3). Figure 3 shows the shape of the slot that they can slide into which is a T shape that prevents rotations. The wide portion of the MPS is parallel with the axis of the core (Figure 1), and therefore only allows movement along that plane the same type of movement that the Control and Safety Rods engage in. The locking mechanism prevents movement (Figure 2), in (toward the core) or out.

c. The 1/2inch relative movement was entered into the previous (1997) SAR to bound conditions for the 0.76$

potential excess reactivity limit analysis. As stated in SAR 4.2.2 and as explained in b above, the MPS movement is limited to approximately 1 mm in a parallel plane with the core axis.

If we assume the MPS movement could move the analysis bounding limit of 1/2, and noting that the only direction of movement available to the MPS is the same type of movement as the Rods, it would have a similar affect for that 1/2inch of movement as does the Rod movement. The difference will depend on the size of that MPS. We have several variations of poison thickness, currently installed is a 1/16 thickness. Full thickness is 2.75 of cadmium and the thickness range is full, half, etc., down to 1/16 thickness. The analysis bounding limit of 1/2 may have been for a full thickness MPS, which will not be inserted at this time due to core age.

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 006 FSAR Section 4.4.4, states that differences between the core modelpredicted control rod worths and net reactivity gains from MPS changes and the NTR measured data are within the overall model uncertainty.

a. Does this apply to safety rod worths?

b. What is considered in calculation of the overall model uncertainty?

6.a. This applies to core model predictions of safety rod worth(s) since the core model and depletion analysis fuel isotopic results were validated to simulate NTR operation and calculate reactivity worth predictions for reactor components and core configuration changes. It is noted that NTR measurements of safety rod worths were not available from current operation to use for comparison to core modelpredicted safety rod worths. Therefore, the core model predictions were compared with the historical SAR Chapter 4 safety rod worth values, and were found to support the historical values.

6.b. The overall model uncertainty is a combination of the uncertainty associated with the MCNP6 code, depletion analysis fuel isotopics, asmodeled NTR configuration, and changes in control rod positions from the different past operation statepoints and core configurations modeled.

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 023 FSAR section 16.1 states that the fuel cladding thickness is 50% of its original value. Discuss and describe the limiting thickness where a fuel rod would need to be replaced.

In addition to corrosion, some fuel aging mechanisms considered for the NTR aluminumclad UAlx fuel include the following:

Fatigue Creep Ductility Fretting wear Fission product leaching The NTR fuel is of a nonstandard design that is described in section 4.2.1 of the submitted SAR. The fuelbearing meat is a uranium/aluminum (UAlx) composite hot rolled between two layers of aluminum into a physically bonded, cladmeat strip. The aluminum fuel cladding is not a structural member; therefore, aging mechanisms that affect its mechanical state such as fatigue, creep, and ductility have limited importance to the residual cladding thickness. Rather, the aluminum claddings primary purpose is to provide a physical barrier that prevents direct contact between the coolant and the UAlx fuel which in turn provides a barrier for unrestricted release of fuel and fission products to the coolant.

The cladding can serve this physical barrier purpose with only a minimal residual thickness, for example with as little as 0.001inch residual thickness, which is about 5% of the original cladding thickness. Aging mechanisms such as fretting wear and fission product leaching could be delayed to some extent by a thicker residual cladding thickness, but these aging mechanisms, if active, would still be possible whether the thickness is asfabricated or reduced by corrosion to 0.001 inch. Since 25% of the original cladding thickness is estimated to remain after 37 more years of operation, the cladding is deemed adequate over the next licensing period to serve its primary purpose as a physical barrier against unrestricted fuel and fission product release.

Cladding breach due to corrosionrelated clad metal thinning, should it occur, is expected to be a leak before break condition whereby serious breach would be readily detected and appropriate actions taken. Cladding breach, leading to escape of fission products and fuel to the coolant, is monitored through continuous monitoring of stack particulate and gaseous activities, and periodic sampling of the primary coolant for Strontium 91 and 92 activity.

Stack particulate and gaseous activities would provide the first indication of escaping fission products. Stack alarm action levels are very close to the normal operating activity levels. See table below. Stack activity that exceeds the alarm level limits would prompt an immediate reactor shutdown according to procedure SOP 8.3, Abnormal Operation, to evaluate the cause, which would include drawing a primary sample for analysis.

Gaseous Activity (µCi/cc) Particulate Activity (µCi/cc) 8

Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 023 Normal Operations 4.0E05 3.0E09 Alarm Level 9.5E05 2.1E08 Periodic strontium 91 and 92 activity monitoring of the primary coolant provides an indication of a less serious breach condition than that which would be detected by continuous monitoring of the stack alarms. Strontium 91 and 92 are measured in accordance with NTR planned maintenance procedure SOP 12.15, Primary Chemistry. Starting in 2021, samples are drawn and analyzed 3 times per year, while in past years, an annual sample was taken. There are no action levels on primary sample activity results; however, strontium 91 and 92 trends are monitored to look for unexpected/unfavorable trends over time. Actions taken based on unfavorable strontium 91 and 92 trends would be dependent on the nature and severity of the trend, but would include actions up to and including suspending reactor operation to evaluate the condition and to identify appropriate corrective actions.

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 026 TS 1.2.15 defines Potential Excess Reactivity as That excess reactivity which can be added by the remote manipulation of control rods plus the maximum credible reactivity addition from primary coolant temperature change plus the reactivity worth of all installed experiments.

a. Provide an example calculation to show how potential excess reactivity is calculated in order to meet TS 4.1.3.1.

b. Describe how changes in graphite temperature factor into the excess reactivity calculation and the change in graphite temperature that is expected during normal operation.

c. Could reactivity be added by other means of control rod movement exceed that added by remote manipulation?

a. First, record current coolant temperature at T/C7. Let's assume it's 80 degrees. Next, record yesterday's actual excess reactivity from rods. Let's assume it was 21.5 cents. Now, using today's T/C 7 temp, yesterday's T/C 7 temp, and the associated table of reactivity to temperature, determine change in reactivity from temperature. Subtract today's temperature reactivity from yesterday's temperature reactivity. Assuming yesterday was 78 degrees, this comes out to 6.0 5.5 = 0.5 cents. Under normal operating conditions, we do not change MPS sheets or any other reactor configurations, so our predicted excess in rods will just be 21.5 (yesterday's actual) + 0.5 (difference from temperature change) = 22 cents.

Now we add 6.5 cents for the experiment worth (this is covered in SOP 66, but in short, this is due to pinhole and sourcelog positioning), and add 5.5 cents (temperature reactivity for 80 degrees) = 34 cents. This is our predicted total excess reactivity.

Attached for visual reference are two reactivity calculations from days when unusual testing was taking place. In the first, dated July 17 2020, Experiment Worth is listed as 0 due to removal of the pinhole and source log. In the second, dated July 20 2020, MPS p is listed as +17 due to changing which MPS was inserted.

b. Changes in graphite temperature are not actively factored into excess reactivity calculations. However, it is passively included, in that primary coolant temperature is being maintained by the heat from the graphite.

With slight seasonal variability, graphite temperature is generally 7080 degrees at startup. Over the course of 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> of operation at full power, graphite temperature will stabilize around 120F at 6" and 140F at 18".

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 026

c. Rods cannot be manually repositioned to add reactivity. For safety rods, the shutdown position of the rod follower and magnet physically block the rod from being withdrawn.

For control rods, there's no electromagnet blocking rod withdrawal, but the motor that drives the rod follower will drive the rod back in the moment the "Rod In" limit switch is no longer made up. Even if this were not the case, the motor cannot be manual hand cranked.

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Vallecitos NRC NTR License Renewal Audit Questions Set 1 Rev 2 QUESTION 026 12