ML052510383

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U. S. Geological Survey Providing Clarification and Additional Supporting Information in Response to Request for Additional Information
ML052510383
Person / Time
Site: U.S. Geological Survey
Issue date: 08/25/2005
From: Timothy Debey
US Dept of Interior, Geological Survey (USGS)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC5120
Download: ML052510383 (10)
v • d • e


Text

USGS science fora changing worfd Department of the Interior US Geological Survey Box 25046 MS-974 Denver CO, 80225 August 25, 2005 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Gentlemen:

The U.S. Geological Survey is herein providing clarification and additional supporting information in response to your request for additional information (TAC No. MC5120) dated March 10, 2005.

This concerns the USGS amendment request to its research reactor facility license (No. R-113, Docket 50-274) to allow the use of aluminum-clad TRIGA fuel in the core.

Sincere]

7~4 Ly, Tim DeBey Reactor Supervisor ff22cfd2

&~

5b -97(4 I declare under penalty of perjury that the foregoing is true and correct.

kecuted on 8/25/05 t

6e.2 /

A ozD

CLARIFICATION AND ADDITIONAL SUPPORTING INFORMATION FOR TBE LICENSE AMENDMENT REQUEST FOR USE OF ALUMINUM-CLAD FUEL AT THE UNITED STATES GEOLOGICAL SURVEY; DOCKET NO. 50-274

1. In our response datedApril 8, 2005 we proposed to revise Technical Specification D.3:

to limit the measured fuel temperature to 750 0C for B-ring measurements and 667 0C for C-ring measurements. However, our analysis only supported actual peak fuel temperatures at those levels, not measured fuel temperatures. The measured fuel temperature limits need to be 15 degrees lower to allow for possible inaccuracies in the measurement instrumentation. As a result, the proposed change in Technical Specification D.3 needs to be for measured fuel temperatures of 7350C and 652 0C, respectivelyforthe B and C rings. The revised proposed change is given below. We propose to not specify a limit on the aluminum-clad fuel in the F and G rings for two reasons: (1) the limits being imposed upon the B and C-ring fuel temperatures will inherently protect the F and G ring fuel elements from exceeding 500 0C, and (2) no instrumented aluminum-clad fuel elements are available to provide the associated measurements. The necessary calculations to support the safety of the aluminum-clad fuel elements in the F and G rings have been provided in the documents supporting the license amendment request.

Curzent wording:

D.3. Fuel temperatures near the core midplane in either the B or C ring of elements shall be continuously recorded during the pulse mode of operation using a standard thermocouple fuel element.

The thermocouple element shall be of 12 wt* uranium loading if any 12 wt* loaded elements exist in the core.

The reactor shall not be operated in a manner which would cause the measured fuel temperature to exceed 8000C.

Proposed wording:

D.3.

Fuel temperatures near the core midplane in either the B or C ring of elements shall be continuously recorded during the pulse mode of operation using a standard thermocouple fuel element.

The thermocouple element shall be of 12 wt* uranium loading if any 12 wt* loaded elements exist in the core.

The reactor shall not be operated in a manner which would cause the measured fuel temperature to exceed 7350 C in a stainless steel clad element in the B ring or 652°C in a stainless steel clad element in the C ring.

2.

In order to help summarize and clarify calculated fuel temperature data under various operating conditions, we have prepared the following table of data. The table includes calculated peak fuel temperatures in the USGS reactor under the following conditions:

a.

normal full power (1MW) operation with the current core condition of 125 fuel elements and typical water temperature.

b.

overpower operation at 1.1 MW (maximum high power scram setting) with the current core condition of 125 fuel elements and typical water temperature.

1

c.

overpower operation at 1.1 MW with the current core condition of 125 fuel elements and water temperature at the maximum allowed (60C).

d.

overpower operation to create a peak temperature of 750C in a B-ring fuel element (approximately 2.1 MK) with maximum water temperature.

e.

normal full power operation (1 MW) with the minimum core loading (100 fuel elements) and typical water temperature.

f.

overpower operation at 1.1 MW with the minimum core loading and typical water temperature.

9.

overpower operation at 1.1 MW with the minimum core loading and maximum water temperature.

h.

a maximum allowed pulse ($3.00) with the minimum core loading and maximum water temperature.

i.

overpower operation to create a peak temperature of 750C in a B-ring fuel element (approximately 1.2 MW) with maximum water temperature.

The data for 125 elements in the core are based on empirical values measured during 1 MW operation and then scaled appropriately for the changed conditions. The scaling was performed by conservatively assuming that the heat transferred from the fuel elements is by free convection (i.e., ignoring radiation heat transfer) so that the rate of heat transferred (i.e., power transferred) to the tank water is directly proportional to the difference in temperature (AT between the fuel element's temperature and the cooling water's temperature. (Q = hA (47) where Q is the heat transfer rate, h is the heat transfer coefficient, A is the surface area, andATis the temperature difference.

The data for 100 elements in the core are based on element power calculations performed using the MCNP (monte cardo) computer code for a 100 element configuration of the GSTR core. These data were examined and reviewed by the NRC for our license amendment number 8, dated March 16, 1998. The data are appended to this letter as. The scaling of core conditions was performed in the same manner as it was for the 125-element core, as described in the preceding paragraph. Another data analysis that we performed to further aid in our understanding of the fuel temperatures resulted in the graph shown as Attachment 2. This graph provides a correlation between fuel element power production in kW and the peak delta T rC) between the fuel temperature and the pool water temperature. The graph was developed by fitting a line to empirical fuel temperature measurements vs the respective MCNP power calculations.

Since the MCNP output is kW produced in each fuel element, the graph can then be used to find the peak fuel element temperature by reading the delta T and adding the pool water temperature.

For example, an element producing 5 kW with a pool water temperature of 25C would have a peak fuel temperature of 130.9 + 25 =155.9 "C.

A core of 100 elements causes more power peaking in the center of the core and lower power production in the outer elements because of the loss of fuel and graphite that was present in the 25 fuel elements that were eliminated. The net result is that the G-ring element temperatures are lower in the 100-element core when compared to a 125-element core. However, the F-ring element temperatures are higher in the 100-element core when compared to a 125-element core.

2

The 100-element core also reaches the 750 0C limit of the B-ring temperatures at a much lower total power than in the 125-element core.

This makes the F and G-ring temperatures lower for the smaller core under these conditions than it is for the larger 125-element core.

We believe the table below provides a concise summary of al operational conditions that could cause high temperatures in the fuel elements.

the F and G ring elements stay well below 500'C.

In all cases, the fuel temperatures of Calculated peak fuel temp Description of Operation Core configuration and power level 125 elements, I MWSS 125 elements, 1.1 MWSS 125 elements, 1.1 MW SS 125 elements, B-rng at 750C (note: the power level required Is - 2.1 MW) 100 elements, I MW SS 100 elements, 1.1 MW SS 100 elements, 1.1 MW SS 100 elements, $3 pulse 100 elements, B-ring at 750C (note: the power level required is - 1.2 MW)

Coolant temperature 25C 25C 60C 60C 25C 25C 60C 60C 60C Calculated peak fuel ternp (VC)

B-ring Crnn F-ing G-dnna 348 309 206 176 380 l 337 l 224 l 191 415 373 259 226 750 667 447 383 604 662 702 500 558 612 652 442 224 244 284 296 159 173 213 252 750 695 299 224 Summary of other proposed technical specification changes:

Section D Reactor Core Cur2ent iordinrg:

6.

Each standard fuel element shall be checked for transverse bend and longitudinal elongation after the first 100 pulses of any magnitude and after every 500 pulses or every 60 months, whichever comes first.

The limit of transverse bend shall be 1/16-inch over the total length of the clad portion of the element (excluding end fittings).

The limit on longitudinal elongation shall be 1/10 inch.

The reactor shall not be operated in the pulse mode with elements installed which have been found to exceed these limits.

Any element which exhibits a clad break as indicated by a measurable release of fission products shall be located and removed from service before continuation of routine operation.

3

Proposed wording:

6. Each standard fuel element shall be checked for transverse bend and longitudinal elongation afler the first 100 pulses of any magnitude and after every 500 pulses or every 60 months, whichever comes first During the first 5 years of aluminum-clad fuel usage, annual fuel transverse bend and longitudinal elongation measurements will be made on 20' of the aluminum-clad fuel elements that have been in the core at any time during that year.

The measurement schedule will be controlled such that different fuel elements are measured each year for this initial 5-year period. After this initial 5 years of aluminum-clad fuel usage, if no generic problems have been detected, the inspection schedule will revert back to the standard fuel 60-month schedule.

The limit of transverse bend shall be 1/16-inch over the total length of the clad portion of the element (excluding end fittings).

The limit on longitudinal elongation shall be 1/10 inch for stainless steel clad elements and h-inch for aluminum clad elements.

The reactor shall not be operated in the pulse mode with elements installed which have been found to exceed these limits.

Any element which exhibits a clad break as indicated by a measurable release of fission products shall be located and removed from service before continuation of routine operation.

Fuel elements that have been removed from service do not need to be checked for transverse bend or longitudinal elongation.

Section D Reactor Core Current vording:

7. The power produced by each fuel element while operating at the rated full power shall be calculated if the reactor is to be operated at greater than 100 kW with less than 100 fuel elements in the core. Recalculations shall be performed:

a) at 6 + 1 month intervals, or b) whenever a core loading change occurs.

Power per element calculations are not required at any time that the core contains at least 100 fuel elements or if reactor power is limited to 100 kW.

If the calculations show that any fuel element would produce more than 22 kW, the reactor shall not be operated with that core configuration.

Proposed wording:

7.

Observance of the license and technical specification limits for the GSTR will limit the thermal power produced by any single fuel element to less than 22 kW if the reactor has at least 100 fuel elements in the core.

Therefore the reactor must have at least 100 fuel elements in the core if it is to be operated above 100 kW.

Operations with less than 100 fuel elements in the core will be restricted to a maximum thermal power of 100 *W.

Proposed new (additional) technical specifications:

4

Section C.

Reactor Pool and Bridge Proposed additional specification wording:

3.

The control console shall have an audible and visual water level alarm that will actuate when the reactor tank water level is between 12 and 24 inches below the top lip of the tank.

This water level alarm shall be functionally tested monthly, not to exceed 45 days between tests.

This item is not applicable if the reactor is completely defueled and the pool level is below the water treatment system intake.

Section C.

Reactor Pool and Bridge Proposed additional specification wording:

4.

The pool water shall be sampled for pH at quarterly intervals, not to exceed 4 months.

The pH level shall be within the range of 4.5 to 7.5 for continued operation.

This item is not applicable if the reactor is completely defueled and the pool level is below the water treatment system intake.

5

.Attach rnent I

~MCNP ANALYSIS of GSTR CORE (100 elements)

GSTR REACTOR ANALYSIS Eight 12 wlo in B.C rings I

2122/95 100 elements GRID DESCRIPTIO N ISerial No. 'Powe atr ro

a~w(W fu l o l W /o
  1. rbl 17832:

2.1 I O.

j1

1.

1 1 fuei rod

Ub2

~O i1.73!

2.1 17.473fI 0 fu lr, d 12w/

  1. b I

7884 2.1 0.01 i 21.4120 fu lrod I)

/o #b 7865.

0 1

  • 211.71 so fuel rod
  1. b5 3701:

1.5910.011, 18.0590 fuel rod 12 w/o #b6t

7867, 2.061 0.011 20.8060 fulrod 0c 531.

1.2 0011!

12.8397 fuel rod #cllI 9842 1.33- 0.01111.40 fuel rod 49ci 4

2

/

1240, 1.91 001 1 19.3920 fu el rod #3 4

7932 1.381 Qooit___

FFCR in #c4 1i0252:

1.581 0.0111 15.9738 fuel rod

  1. c5 1

_9532-1.38)-

0.01 1 13.9380 fuelro 12 wto #c6 I 7869' 1.981 0011 19.9960 fuel rod

  1. c;7
9533, 1

.1

.13.5340 123.91 0.0 19.19 00 fuel rod

  1. tcl 3007 13 0.01 14.03,90 FF R in#

I 7 50.96, 0.012 i 9.71 52 inFdlO__in__

5980 1

1.

39!

140 2 fuel rod #dll j

7200!

1.25! 0.0111 1

2-.6375 fuel rcd ftdl2

~

7927 1.1Ii7

0. 11 1I8 8 fuel rod
  1. d13 I

5007.

1.161 0.01 11.7276 fT!ro

  1. I1 3321 1.231 0.0111 12.4353 fuel rod dAdl5 53

.ll-O.

r.

il1

1. 2221..

fuel rod

  1. dl7 60 0

1.141 0.012 1.36 fuel rod #dl7 70905

1. 1 11.9298 fue ro #d 4 2 0 1.1 9) 0.0111 2 0 0 fuel rod
  1. d3 13 7929 1.331 0.0111 13.4483 fuel rod
  1. d4 73250 1.291 0.01l1 12.0309 FLueli ro d fd7 79261 1.18 0.011 11.9298 fuel rod
  1. d8 31341 1.23 0.011 12.4353 fuel rod
  1. do 7030; 1.25 0.0111 1~21.6375 fuel rod
  1. el ____

41 2-8 1

096) 0.013'___ 9.7248 fuel rod

  1. elO 0 30171 1.02J 0.012

__10.3224 fuel rod

  1. -ell 1 8 311

. 1 11 2

fuel rod

  1. el12 1

38601 1.071 R012 10.8284 fuel rod #el3 31161 1.031 0.012 10.4236 fuel rod

  1. el14 24451 1

0.012 10.1200 fuel rod #el 5 j5952 1.02 0.012 10.322A fuel rod

  1. el6 3022__

9___

__8261___

fuel rod, #el17 3697 0.99 0.012 10.0188 fuel rod #el 8 6587 1.02 0.012

'10.3224 fuel rod

  1. el191 57511 0.99 0.012 ____i.0188 fuel rod #e2 68.43 0.89 0.013 9.0157 fuel rod
  1. e20 5957 0.92 0.013 9.3196 fuel rod
  1. e21 51991 0.-79 0.013 8.0027 fuel rod #e22____

57041 0.82 0.013 8.3066 fuel rod 1*e23 57051 06.811 0.01 31 8.2053 ~

GSTR REACTOR ANALYSIS 2

Eight 12 wlo in B,C rings 100 elements GRID DESCRIPTION

!Serial No. Power factor I Error Max power (kW) fuel rod #e24 3361 0.93L 0.013 9.4209 fuel rod

  1. e3 6839 0.97 0.012 9.8164 fuel rod _ #e4_!__

57611

_ 0.96! 0.012 9.7152 fuel rod ffeS_

r F57c5 1T 0.012 101200 fuel rod #e6 i

57541 1.03 0.012 10.4236 fuel rod

  1. e7 1

68401 1.071 0.012!

10.8284 fuel rod

  1. e8 3857f 1.03j 0.0121 10.4236 fuel rod
  1. e9 5013 1.03' 0.012i 10.4236 fuel rod
  1. f1 5726 0.6 0.015!

6.0900 fuel rod Xfl0 i

5759 0.74 0.0141 7.5036 fuel rod

  1. f1l 574858 o

6 0.67!

0.0141 6.7938 fuel rod

  1. 112 5728j 0.75! 0.0141 7.6050 fuel rod
  1. . 3 5735 0.731 0.0141 7.4022 fuel rod
  1. f14 l

57440.741 0.01' 7.5036 fuel rod

  1. f15 j5737 0.751 0.0141 7.6050 fuel rod
  1. f16 5716 0.7i 0.0141 7.0980 fuel rod #f17 5730 0.73 0.014 7.4022 fuel rod uf18 5743 0.7 0.014$

_7.0980 fuel rod #f19 57401 0.691 0.0141 6.9966 fuel rod Wf2 l

57071 0.71 0.014' 7.0980 fuel rod

  1. f20 1

5706 0.72, 0.014' 7.3008 fuel rod

  1. f21 5731 0.63L 0.01_i 6.3945 fuel rod
  1. f122 I

5732 0.691 0.014!

6.9966 fuel rod

  1. f23 1 5729' 0.bro.015 6.9020 fuel rod
  1. f24 5753 0.681 0.015 6.9020 fuel rod Xf25 I

5745 0.681 0.014 6.8952 fuel rod

  1. f26 5725 0.62] 0.015' 6.2930 fuel rod
  1. f27 5747 0.651 0.0151 6.5975 fuel rod
  1. f28 1

57271 0.661 0.0151 6.6990 fuel rod #f29 1

5741 0.651 0.0151 6.5975 fu rod

  1. f3 5717 0.67! 0.015.

6.8005 fuel rod #30 1 5751 0.65 0.015j 6.5975 fuel rod

  1. t4 5719 0.69 0.0151 7.0035 fuel rod #f5 `

5734 0.71 0.0141 7.1994 fuel rod

  1. f.

57601 0.68i 0.014!

6.8952 fuel rod

  1. r7 5 7 3 9j 0.73! 0.0141 7.4022 fuel rod
  1. f8 57_

I 5708 0.74 0.0141 7.5036 fuel rod 5750 0.73 0.0 1i 7 402 fuel rod

__1 _

9472 0.491 0.017 4.9833 fuel rod 10 1

5720 0.5' 0.016 5.0800 water

  1. gl1 0.0000 0.0000 fuel rod
  1. 913 i

5736 0.51 0.016 5.1816 water

  1. g14 1

. 0.0000 water

  1. g15 I

0.0000 fuel rod #g16 5701 0.51 0.016 5.1816 water

  1. g17 0.0000 water
  1. g18 0.0000 fuel rod #q19 5682 0.51 0.016 5.1816 water
  1. g2 0.0000 water
  1. 1g20 0.0000
t.

2/22/9!

W

-J GSTR REACTOR ANALYSIS Eight 12 w/o in BC rings 3.

10Q elements GRID DESCRIPTION tSerial No.lPower -fctor LError Max power (

water_

  1. I1j 0.0000

_ e ~ ~

fuel rod

  1. 22 1

5B6761 0.5 0.075085 water_#23 0.0000 fuel rod

  1. g25 I

57581 0.49F~17 1

4.9833 wate.

__.26 _ ________

0.0000 water

  1. g 2

0.0000 fuel rod L

28 1

5678 0_4g5 0.017 _4.578 water

  1. g29

-. do water

  1. 3 i__3 0-0000 water
  1. g9730 0.0000 fuel rod

'3 I5 5686 0.44i 0.0171 4.4748 waterj #a32 i

0.0000 vvater_

-13 0-.000-0 water

  1. 9 j

0.0000 waer__g5____II___I0.0000 water

  1. 93 0 0000 LI 4.7.7.9, fuel rod t#31 56831 0.44i0.0171 4.4748 water

._000 water

  1. g6 r

I 0.0000 fuelrod ftg 5762j 0.481 0.01-7_

4.8816 water I

j 0.0000 water

  1. g5 I

0.0000 Attachment I 11 TIT.

I

40o GSTR Fuel Element AT vs Element Power 375 350 325

/ o 300

/,

/

/

275

//

250 jog 225 XO200 o

(U 175 i

y = 26.18x (+0.71 1 95% Cl);

Easo

=

r20.9839 ISOI 1251 100

//

75 -

50 I

0

0.

1 0

1 2

3 4

5 6

7 10 12 13 14 12 kW/element

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