ML19165A021
| ML19165A021 | |
| Person / Time | |
|---|---|
| Site: | University of Maryland |
| Issue date: | 06/06/2019 |
| From: | Andrea Johnson Univ of Maryland - College Park |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19165A039 | List: |
| References | |
| Download: ML19165A021 (9) | |
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4418 Stadium Dr UNIVERSITY OF College Park, Maryland 20742-2 l 15 MARYLAND 301-405-7756 TEL ajohns37@umd.edu GLENN L.MARTIN INSTITUTE OF TECHNOLOGY A. JAMES CLARK SCHOOL OF ENGINEERING Department of Materials Science & Engineering Radiation Facilities Amber Johnson, Director June 6, 2019 Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001 RE: Maryland University Training Reactor (License #R-70) responses to Request for Additional Information dated April 16, 2019 Enclosed please find the University of Maryland's responses to the 13 requests for additional information (ML19024A291) regarding the license amendment request dated the March 28, 2018 (ML18092A086).
I declare under penalty of perjury that the foregoing response is true and correct.
Sincerely, Amber S. Johnson
RAI #1: Provide a copy of the OSU Radiation Center report titled , "Analysis of the Neutronic Behavior of the Maryland University Training Reactor," dated July 2017 .
See attachment 1.
RAI #2: Provide a thermal-hydraulic analysis for the proposed core configuration or explain why the thermal-hydraulic analysis for the current core configuration , as referenced in the LAR, bounds the proposed core configuration. Alternatively, justify why additional information is not necessary.
A new thermal-hydraulic analysis is unnecessary. A comparison of Figure 1, the power distribution of the proposed core with Figure 2, the current core power distribution (from the neutronics analysis report) , shows that the average power per element at full power will be reduced from 2. 00 kW to 1.95 kW, and the power of the most reactive single element will be reduced from 3. 86 kW to 3.52 kW This indicates that the power peaking factor will be reduced and the current thermal hydraulic analysis is a conservative estimate.
8 7 6 ) . 3 7390 7191 7:378 Tf19 rr 7J55 7395 739J 7168 7169 1.m 7115 1.30 1.65 __OJ 2JS 256 2.16 2.70 SB 2.3:S 2.06 L.67 1..36 "f mo U56 7334 73S9 7m 7377 7353 1WI 7396 7167 1166 7336 1.48 Ul9 2.41 3-07 l09 3.20 3.19 l09 3.0i 2.45 t93 L55 7161 7026 739S 259 7368 1365 73'74 7375 304 7406 7342 7343 l.6S Ui 2.!}7 0..00 174 Ho U6 171 0.00 2!J7 2.20 Ui9
- £ 702S 1(!11 1399 7400 1361 1366 7316 7404 7405 nn 7344 l.78 2-29* 2..95 :l65 1.7 3.SS 161 3..58 2.,89 129 1.7:S 7408 7400 ms 7346 731U 7383 7311 nn 7290 7'.UO 7164 71.65 2.21 U9 Bl ;
HB 3.77 173 3.51 H6 2.73 2.21 DO 0 7LM 734S: 7347 73S1 7384 7310 1369 7>32 7.Ul 7163 7162 2Jl9 3.02 3.33 3..64 3.36 3.13 l.84 2.47 2.00 1.55 1349 7.362 7lo1 HS 132 C 7350 1364 t9S 1.12 Figure 1: Power distribution of the current core. The upper numbers of each pair are the element serial numbers, and the lower numbers are their power in kW at full power. The highest power element is highlighted in red.
8 J IS 5 4 3, 7390 7391 7378 737.9 7354 1155 7395 7393 7168 7169 7133 7335 1.12 1.40 1.71 1.99 2.19 2.28 : 227 2.16 l.98 l.73 L l 1.16 r 7389' 7392 7377 nso 7353 7356 1JfJ7 7396 1167 7166 7334 7336 L2.S l .6:3 2.08 2.64 2.65 2.73 l.71 2.~ :?.58 2:JJ(S 1.63 U2 il6l 1026 7398 2S9 1368 16S 73;4. 737$
- 304 7406 7342 734 1.48. 1.90 :uo ();00 3.2 ' 3.17 3.15 :US 0,00 2.52 I.st> i .46 E 7021 707J 7399 7400 7367 7366 7J7.3 7376 1404 1405 7341 7344 l.63 2J)9 :u1 1.29 H . 3.44 Ml 3.lS 3,16 2.S4 .02 LS5 7408 7409 7345 73% 7382 7383 7371 7372 7290 7330 7164 7Hi5 l.158 2.15 2.71 3.11 336 3.47 3.42 3.22 2.9l 2 .51 2 . 03 1.56 I) 7401 7160 7348 7347 7331 7370 7369 7332 7331. 7163 711S2 1..63 2.10 2c62 2J.)9 3.22 2.42 l.93 1.50 7360 1l51 73$2 7349 1401 73<52 7363 l.49 l.90 2.37 2.66 3.(ll 0.00 3.02 l ..82 l..lS C 7359 7358. 7351 7350 7403 7400 7J.S7 7Ml 7l64 1.27 l .63 2.00 l.21 2.37 2.60 2.33 l.61 us 6286 6284 5861 6281 '6287 6289 ,*73:33 73T7 6282 '6277 l .01 1.28 1.55 1.78 UI6 1.16 1.76 1.59 1.21 0.93 B
~ ~s.s $862 SS64 6219 ~290 : 7339 7340 62$$ 62Ci8 o:n 0.92: LU UtS 1.43 us i 1.}9 1.0S O..S6 0,6$
Figure 2: Power distribution of the proposed core. The upper numbers of each pair are the element serial numbers, and the lower numbers are their power in kW at full power. The highest power element is highlighted in red.
RAI #3: Confirm that TS 4.1 , Specification 4 , in the LAR should include the inspection of bundle C6. If not, explain the apparent discrepancy between the information provided in the LAR and current MUTR TSs .
This is a typo in the LAR. Inspection of bundle C6 is required. The current MUTR Tech Spec reads "A visual inspection of a representative group of fuel bundles from row C column 8, 7, 6, 5, 3 and row B column 4 shall be performed annually, at intervals not to exceed 15 months. If any are found to be damaged, an inspection of the entire MUTR
. core shall be performed. "
RAI #4: Provide an explanation of the substantial difference between the calculated and measure reactivity worths for Shim 1 and Shim 2. Additionally, state any ranges of acceptability between the simulated and measured reactivity worths and what action will be taken if the range is exceeded.
In the neutronics report individual simulated rod worths are simply the absolute value of "all rods in" minus the absolute value of the worth of the individual rod. Thus the control rod worths of the rods are Shim I: $3. 80 +/- $0.05, Shim II: $3.87 +/- $0. 06, Reg: $2. 82 +/-
$0.06 and the total rod worth of $10.48 +/- $0.10.
Rod worths are typically calculated using the asymptotic period method, but this can suffer from the effects of rod shadowing. We tried measuring the rod worths again by the rod drop method to reduce these effects. This resulted in worths of $3.22 for Shim 1,
$4.25 for Shim 2, and $2.48 for the Reg Rod, or a total worth of $9.95. This is a much better agreement with the simulated values (within 20% in all cases, and within 6% in total worth.
Rod worths can differ by approximately 20% from year to year so this is considered an acceptable deviation from the simulated values.
RAI #5: Provide a justification that the rod withdrawal analysis provided by the December 18, 2006, letter bounds the proposed excess reactivity LCO of $3.50 . If not, provide a revised rod withdrawal analysis that considers an excess reactivity of $3 .50 .
Alternatively, justify why additional information is not necessary.
The rod withdrawal analysis from the December 18, 2006 RA/ response does not provide a bounding condition for the proposed excess reactivity LCO of $3.50. A hypothetical rapid insertion of $3. 70 in excess reactivity is described in sections 13. 1.2 and 13.2.2 of the SAR. The hypothetical rapid insertion of $3. 70 at 0.01kWand 250kW corresponds to peak fuel temperatures of 692°C and 988°C respectively. This temperature is lower than the safety limit temperature of the fuel cladding, 1000°C, thus the fuel integrity would not be lost. Therefore, this hypothetical instantaneous insertion of $3. 70 would not result in the release of any fission products from the primary barrier.
A slow ramp insertion of excess reactivity might be expected to produce higher fuel surface temperatures in a fuel with smaller or slower negative temperature feedback.
However, this is not a concern for TR/GA fuel, which introduces a large prompt negative feedback as the fuel meat heats up, greatly reducing the power density at that location. A reactor with fuel designed for the harsh conditions produced by a pulse can withstand a continuous rod withdrawal. Further information is not necessary.
RAI #6: Provide an analysis that shows the relationship between the measured fuel temperature in the IFE D8 position and the maximum fuel temperature within the proposed core configuration . Alternatively, justify why additional information is not necessary.
The /FE is located in position 08 NE comer. The power of this element is 2. 10 kW in the proposed core. The hottest element in the proposed core is 06 NE with a power of 3.52
kW (Figure 3) . The power of these elements differs by Jess than a factor of 2. The basis for TS 2.2 establishes that the temperature of the JFE can differ from the hottest element by no more than a factor of 2 to allow for a 650G safety margin with an JFE trip setting of 175G. Since the power of 06 NE is Jess than twice the power of 08 NE, the basis for TS 2.2 is still being met and further analysis is unnecessary.
8 7 6 s 4 3 7390 7391 7378 7379 i354 1395: 7393 7168 7169 7333 7335 *
, 1.12 7JS9 l.40 7392 1.71 7377 l.99 7380 2.19 7353 22:1 7397 2.16 n~
- 7167 L98 1.73 71~
1.28 l.63 2.08 2..M 1.65 2.71 l .d:1 . 2.58 1.06 Figure 3: Power 7HSl 7QU 7J!JS 2$9 il~ n;.a. 1S 304 74()6 1 distribution of the 1.48 l.90 2.60 0.00 1 .24 J..15 :HS 0.00 2'.5l l.86 proposed core. The E 70"...8 7400 7367 7376 7404 7405 734]
10Z7 7399 7373 1.63 2.67 J.3 J..41 3.2 .54 2.02.
upper numbers of 2 .09 329 3]6 740& 7409 7345 7346 7371 7372 7290 7330 7i64 7165 each pair are the 1.68 2.15 2.71 J.U 3.42 3.22 2.91 2.51 2.03 1.56 element serial 0 74CYI 71-60 7348 i.347 7370 7369 7'332 7'331 7163 7162 numbers, and the 1.63 2.10 2.62 2.99 323 :3.0l 2.76 2..42 l.93 1.50 lower numbers are 1360 1'1S1 7352 7349 7401 260 7.3fl: nss 7362. 7363 the ir power in kW at 1.49 l.90 2.37 2.66 3.01 0 .00 3.02. 2 .71 1.82 138 C 7359 7358 7351 7350 7403 140.2 7387 7386 ' 73-61 73-64 full power The /FE 1.27 1.63 2.00 2.21 2.37 :uo 2.13 2 .30 l.61 us is serial number 6286 6284 5861 6281 6287 6289 73311
- 7337 6282 6277 7160.
1.01 12'8 1.55 1.78 L86 U6 1.76 1.59 1.21 0.93 B
m3 62 ' S862 S864 6279 6290 73 9 7340 CP..SS 626S
().73 0.92 l .l l. 6 L4 us ll9 LOS 0.86 0.6$
RAI #7: Clarify the units of measure used in Figure 2 of the LAR or explain why units of measure are not needed.
The units of Figure 2 are kW for the lower of each pair of numbers. The upper of the number pairs are element serial numbers.
RAI #8: LAR Figure 3, "Proposed core configuration power peaking factors ," provides information on power peaking factors . However, the ratio used to develop these peaking factors are not clear. Provide an explanation of the ratio used to develop the power peaking factors presented in LAR Figure 3. Alternatively, justify why additional information is not necessary Figure 3 is miscaptioned; it should be labeled "Proposed core power distribution". The units of Figure 3 are kW for the lower of each pair of numbers. The upper of the number pairs are element serial numbers.
RAI #9: Provide an illustration of the proposed core configuration showing the locations of fuel bundles that may be displaced for the in-core pneumatic experimental system, plutonium-beryllium source , neutron detectors, and graphite reflector elements.
Alternatively, justify why additional information is not necessary.
See below. Elements are displaced from grid position 86 for the plutonium-beryllium source and grid position C4 for the in-core pneumatic experimental system. Neutron detectors and graphite reflectors are located on the periphery of the core.
Thermal Column
@ Instrumented Rod
~9 ~9
- Rabbit
~ Reflector
~-~---~~
F f-9 8818888888888 @
E oo.o e oooo e ooo.
00!0000000000 a
'\; C..CD"§~CIJ w
~
B 00 00 00 00 00 00 00100 00 00 1--******-***-+******-----*--+----***--**- **-*******-**..**-*** *----
ooooo e oo 00 00 00 00 00 00 !,.-."'- 00 00 00 00 00
**-*-*- ****-**********--+----*+
oo 00 *-+--+-----1
+----l 0000 ~ 00 000 9 8 7 6 5 4 3 2 1
\\ Through Tube
_J RAI #10: Provide a shutdown margin analysis that includes relevant uncertainties, error limits, and worst-case conditions and takes into account the difference between the simulated and measured control rod worths for the current core configuration . Provide an explanation of why the proposed core configuration simulation provides acceptable predictions of control rod worths for determining that the shutdown margin will be maintained for the proposed core configuration . Alternatively, justify why additional information is not necessary.
Further information is not necessary, adding fuel to the core will increase the control rod worths. Assuming a worst case scenario where the worths of the control rods do not change from their current values (LAR table 2), while the excess reactivity increases to
the proposed limit of $3.50, the shutdown margin of the reactor, given by the total rod worth minus the most reactive rod minus the excess reactivity is:
($2.08+$2. 75+$2.20)-($2. 75)-($3.50)=$0. 78. Comfortably above the Tech Spec limit of
$0.50.
RAI #11: Provide an explanation for the apparent discrepancy between .the statements in the LAR above referring to a $4.00 insertion of excess reactivity and MUTR TS 3.6, Specification 2. In addition, for the proposed core configuration, provide the basis for selecting $4.00 of reactivity as a bounding analysis for a credible prompt insertion of reactivity, especially given that your proposed excess reactivity is $3.50.
$4.00 was chosen as a conservative estimate of maximum possible reactivity insertion.
The central fuel bundle has an estimated worth of $4. 70. (SAR 13.2.2.3) If the central bundle were to be dropped into the core while critical, the reactivity insertion above prompt critical would be approximately $4. 00. This was analyze_d as the maximum possible reactivity insertion although there is no conceivable scenario in which this could happen. This situation is analysed in section 13.2.2.3 of the SAR with an insertion of $3.70.
RAI #12: Clarify the request for the grammar change of the current LC 2.8.2.d.
The licensee request was to update current license condition 2.B.2.e. However, due to the requested deletion of 2.B.2.b, 2.B.2.e would become 2.B.2.d. The licensee now understands that the NRG prepares the updates to the license and providing the text is unnecessary.
RAI #13: Provide an explanation if UMD can perform an adequate visual inspection of the remaining fuel bundles (i.e., other than fuel bundles in rows Band C) in the core. If so, state which fuel bundles can be adequately inspected. Include additional information in TS 4.1, Specification 4, to explicitly state which and how many fuel bundles will be inspected on an annual frequency. Alternatively, justify why additional information is no_t necessary.
UMD proposes a change to MUTR TS 4.1, "Reactor Core Parameters," to clarify the fuel bundles that are inspected annually as follows:
- 4. 1 Reactor Core Parameters
- 4. A visual inspection of 2 fuel bundles from rows B and C shall be performed annually at intervals not to exceed 15 months. The bundles inspected shall change each year so that in a 5 year period the entire group will be inspected. If any are found to be damaged, an inspection of the entire MUTR core shall be performed.
In response to RA/ Nos. 4, 9, and 31 (Refs. 15, 55, and 96), we provided justification for not inspecting and measuring all fuel elements in the core, or performing length and bend measurements. The MUTR does not pulse, does not use a forced circulation coolant system, has relatively low fuel burn up given its operating history, uses stainless steel fuel elements, has a low risk of damage to instrumentation, and is only currently licensed for a power of 250 kW Therefore visually inspecting the fuel in grid plate locatiqns listed in TS 4. 1, Specification 4 provides an adequate representative profile of all other fuel elements in the core. However, if an annual inspection identifies damaged fuel, then *the entir~ core would be visually inspected for damage in accordance with TS
- 3. 1 Specification 4. *,n order to measure the fuel elements, the fuel bundles would need to be diS8SSf!mb/ed, Which has never been undertaken because it presents risk of fuel and instrumentation damage. Therefore, TS 4. 1, Specification 4, is a reasonable alternative*to the guidance in NUREG-1537, Appendix 14.1, Section 4.1.
Attachment 1 "Analysis of the Neutronic Behavior of the Maryland University Training Reactor"