ML20235J076

Proposed Tech Spec Changes Revising Fission Density Limit & Amending Percentage of Voids Permitted in UA1(x) Type Fuel

ML20235J076
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	02/13/1989
From:	MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE
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ML20235J065	List: ML20235J062 ML20235J076
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NUDOCS 8902240154
Download: ML20235J076 (26)

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.o 4 -I - Appendix A Proposed Technical Specification Channes

1. Attached is-. a copy of the proposed technical -specification

'#3.11.2(e) with the change nottd -by margin marks. (Note:

Specification #3.11 and its basis occupy pages 3-40, 3-40a,.3-41, 3-42, 3-43, and-3-43a. Only mate'r ial on pages 3-40, 3-43, and 3-43a .is being altered. However, the other pages have also been (

retyped so as to eliminate the need for.page 3-40a. Also minor  !

typographical errors have been corrected and the license number j mentioned. in the last paragraph of the basis has been updated.

Specifically, SNM-81 was consolidated with several other licenses to form SNM-986.

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p 3;11 Limitina Core Operatina Conditions'

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Applicability ~

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'This specification applies to core conditions:during operation of.

the MITR.

Obiective-To assure that core conditions are maintaine'd within the ' bounds used to establish the. safety. limits.

i Specification

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1. The reactor shall not be made critical unless all fuel ele-ments and other core components are secured in position and the hold-down plate is latched in position.

2. :The reactor shall not be operated at power levels of greater 1 than 100 kW unless:

a. All positions in the core tank are filled with either a

' fuel element or another approved unit.' For changes in core configuration F and F shall be evaluated.

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b. Primary coolant flow is established,

c. At least five operable shim blades are within 2.0 inches of a banked (average shim blade height) position and any inoperable blade.is at the average height or above, ex-cept that greater imbalance may exist when one or more I I

shim blades are being inserted to make the reactor sub-critical.

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d. The reactor top shield is in position,

e. Peak fuel burnup does not exceed 1.3(10)88 fissions /em8 i

if the fuel is uranium-aluminum (UA1) alloy; or, f

.2.3(10)8 fissions /cm8 if the fuel is intermetallic UA1 x with 4 to 11% volds.  !

Amendment No. 12 3-40

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j 3. Pu-Be ' neutron sources shall not be used in the reactor above a reactor power of 500 watts.

Basis All fuel elements and core components must be secured in position ,

to prevent mechanical damage of the components and core reactivity changes from movement and to assure proper flow distribution and 4

cooling.

The safety limits in Specification 2.1 were derived on the basis of an approximately even distribution of H 0 flow among fuel elements.

To ensure the validity of the Safety Limits, measures must be taken to assure that flow is adequate and evenly distributed through the fuel l elements. Therefore, all positions in the core must be filled.

Further, the established Safety Limits includes the effect of i

banked shim blade height. Unbalanced shim blade configurations might {

lead to higher power per element conditions. Calculations have been l 1

made to determine the Increase in power peaking that might occur in  !

the extreme case of operation with one shim blade fully inserted. l These calculations involve a three-dimensional solution to the neutron  ;

i diffusion equations in order to account for the azimuthal non-  !

i symmetry. Two methods have been used. The first method involved the use of the two-dimensional code EXTERMINATOR as described in Chapter 3 of the SAR. The R-Z calculation was made to get values of the local buckling terms. at each mesh spacing in a plane near the bottom of the t shim blade full in position. This was done with all the blades at the 10.0 inch position. These values of the buckling were used for leak-age terms in an R-0 calculation through the plane of interest, near the bottom of the core. One R-0 calculation was made with no shim 3-41

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'[ blades at this plane and one calculation was made with one shim blade inserted at this plane. .The use of the same axial bucklings for both cases produces a very conservative estimation of the power peaking.

The results of these calculations indicate that the flux peaking. in the hot channel at the edge of the core opposite to the inserted shim i

blade will be increased by about 14%. "

Another method was also used to calculate the power peaking for the one shim inserted case. In this method a full three-dimensional solution was made . by using the computer code CITATION discussed in Section 15.9 of the SAR. In order to solve the problem within the memory capability of the MIT computer, it was necessary to go to a coarser mesh spacing in the R, Z and 0 directions. A calculation was I made with all blades at the 10.0 inch position and one was made with one shim blade fully inserted and the remaining rods still at the 10.0 inch position. This gives a conservative estimate of the peaking fac-tor increase since, in actual operation, the other five blades would be withdrawn further to compensate for the insertion of one blade, j The result of these esiculations indicate an increase in the peaking i factor at the hot channel, on the side of the core opposite to the inserted blade, to be approximately 4%.

In summary, these two methods predict that the maximum increase in the power peaking factor for a non-uniform shim blade setting even as far as one blade fully inserted will still be less than 14%. Such an increase in the peaking factor will not cause the het channel to approach the safety limit. It would take an increase of 33% in the peaking factor to approach the conservative safety limit equivalent to j 8 MW, assuming that the initial hot channel factors were at their max-

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• ( .imum values and that the reactor was operating at the 6 MW operating -

i limit. The specification of a maximum imbalance of *2.0 inches at the j blade position of maximum reactivity worth corresponds to about 10% of l

'the' total worth of'a single blade, and'therefore will.not appreciably affect the core power distribution.

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.During blade insertion for shutdown, the power will decrease att all points in a closely-coupled core such as the MITR-II. Although an  ;

imbalance in blade height will change the shape of the flux distribu-tion, the power will not increase at any point unless other blades are i

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raised to maintain criticality. i 1

The reactor top shield acts as a biological shield. To facili-i tate the performance of various experiments placed in the core, the i reactor may be operated at power levels below 100 kW with this shield removed. The total dose rate at 100 kW is conservatively estimated to be approximately 2.5 rem /hr. This dose rate is not in excess of those occasionally encountered during certain maintenance operations, and it has been demonstrated that administrative controls can provide ade-i

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quate controls under such conditions. Adequate controls will be instituted during such experiments to prevent excessive personnel exposure.

The peak fission density limits are based on information devel-oped and tested as part of the fuel designs for the Engineering Test Reactor (ETR) and the Advanced Test Reactor (ATR) located at the Idaho Testing Station. These fuel plate designs and tests cover the range l 1

of fuel loading expected for the MITR-II core. The information used to set these limits is provided in Section 3.3.5.2 of the SAR.

The use of the two one-curie Pu-Be neutron sources presently Amendment No. 12 3-43 l

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possessed under license SNM-986 (formerly SNM-81)' has been analyzed and determined'to be' safe ~up to a reactor power of 500 watts. lThis

' analysis was submitted.to the NRC by-letter dated March 28,1975.

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Amendment'No. 12 3-43a l

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-I Appendix B

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Technical Literature Supportina Revision of the  !

Fission Density Limit for UA1, Fuel. l i

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1. Attached is a copy of an article by J. M. Beeston, R.~R. Hobbins,. 1 G. W. Gibson, and W. C. trrancis entitled, " Development and-Irradiation Performance of Uranium'Aluminide Fuels in Test Reac-  !

tors", Nuclear Technology, Vol. 49, June 1980, pp 136-149.  :

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s I Appendix C  !

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SAR Revision No. 37

1. Attached is'SAR Revision No. 37.

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. Summary of Channes SAR Revision No. 37 Remove Panen Insert Pages Description of Change  !

i 3.3.5-2 SAR Changes fission density limit for UA1x t 3.3.5-2 fuel from 1.8 to 2.3 108L fissions /cm'. j Changes the percentage. of . voids l pe rmitted in UA1 x type fuel _f rom '4 to 7%' to ' '4 to 11%' and deletes a parenthetical note indicating that the fuel is 35 weight-percent U-235.

Indicates that reference material on j swelling and blistering (Ref. l' 3.3.5.2-1) is now only applicable to UA1 type fuel. I i

3.3.5-3 SAR Adds technical discussion of swelling 3.3.5-3 and blistering for UA1 x fuel based on new reference, #3.3.5.2-2.

3-ref-4 SAR Adds new reference #3.3.5.2-2.

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SAR 3.3.5 V Lo Fuel elements will be discharged from the. core as required either for reac-- I tivity changes or whenever a fuel burnup limit is approached. The fuel burnup

11mit is to be determined on the following basis

if the fuel elements are made with uranium-aluminum (UA1) alloy, the 1) peak fission density shall not exceed 1.3(10)85 fissions /cm'.

2) if the fuel elements are made with -intermetallic UA1 x with 4 to 11%

voids, the peak fission density shall not exceed 2.3(10)81 fissions /

cm8 .

The evaluation of the peak fission density in each fuel element will be based on the total time the element is operated in each position combined with the meas-

) ured power distribution factors for that position. The hot channel facters and uncertainties that apply to the peak heat flux as discussed in Section 3.3.4 and Section 11.3 will be applied to provide a conservative estimate of the peak fis-sion density in each element.

The above peak fission density limits are based on information developed and tested as part of the fuel designs for the Engineering Test Reactor (ETR) and the Advanced Test Reactor (ATR) located at the Idaho Testing Station. These ,

fuel plate designs and tests cover.the range of fuel loading expected for the MITR-II core. The information used to set the UA1 limit has been collected by Caudle Julian (Ref. 3.3.5.2-1). The information used to set the UA1x limit is given by J. M. Beeston et al (Ref. 3.3.5.2-2). Figures 3.3.5.2-1 and'2, taken from Ref. 3.3.5.2-1, give a summary of the basis for the fission density limit for UA1. (Note: Information given in these figures for. UA1 x is now superseded by data contained in Ref. 3.3.5.2-2). At the limit, the volumetric change due to core (fuel teat) swelling is AV/V = 8.5%. An 8.5% fuel platt j swelling will lead to a reduction of the coolant channel of 0.0013 inches on each side of the fuel plate. Such a reduction is within the manufacturing  !

tolerance and is covered in the power peaking and hot spot calculation by the conservative uncertainty factors discussed in Section 3.3.4.3.3.

l The possibility of feel plate failure by blistering for UA1 is presented in l

Figure 3.3.5.2-2. The hot spot in the MITR-II core is always predicted to be well below 200*C under limiting conditions of operation and also under loss of flow conditions with natural convective cooling. Hence, by the information in Fig. 3.3. 5.2-2 no f ailure by fuel plate blistering is expected to occur during the operation of the MITR- I core. )

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SR #0-88-13 DEC 20 1958 1

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. ' 3.3.5-3 j Relative to UA1x fuel performance, the change due to swelling is estimated by.Beeston (Ref. 3.3.5.2-2) to be 2.6%F/1088 where F is the fission density in fissions /cc. At the limit of 2.3 1088 fissions /ce, the swelling would be 6%.

This is well below the 8.5% that was shown above to be acceptable.

As regards the blistering of UA1 ,xdata from Ref. 3.3.5.2-2 show that the blister temperature for UA1x at a fission density of 2.3 1088 fissions /cc is i above 700 K even after allowance for three standard deviations in the measured data. MITR-II operating temperatures are well below this.

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SR /10-88-13 DEC 20 1988 l

OAR 3-ref-4

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3.3.5.2-1 Julian, Caudle, " Evaluation of a 6.2 Kilogram U-235 Core Loadirg f'or i the Missouri University Research Reactor," University of Miss ouri, July 28, 1970.

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3.3.5.2-2 Beeston, J.M., Hobbins, R.R., Gibson, G.W., and W.C. Francis, l

" Development and Irradiation Performance of Uranium Aluminide Fuels on Test. Reactors," Nuclear Technoloav, Vol. 49, June 196.: , pp 136-149. ,

3.3.5.4-1 Furtado, P.M., " Longitudinally Finned Test Section Experimental Data I' Analysis," MIT Nuclear Engineering Department (unpublished).

3.3.5.4-2 Kennedy, D.J., " Beam Port Optimization for the Proposed High-Flux MITR," Nuclear Eng Department, MIT, ScD Thesis, August 1969.

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i SR #0-88-13 DEC 20 1988

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