ML20235K703

Safety Evaluation Supporting Amend 11 to License R-37

ML20235K703
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	11/17/1975
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235K610	List: ML20235K607 ML20235K641 ML20235K649 ML20235K678 ML20235K681 ML20235K689 ML20235K700 ML20235K703
References
NUDOCS 8902270211
Download: ML20235K703 (8)

Text

.

'/ , , D ' , __

\'

- ^

UNITED STATES I - NUCLEAR REGULATORY COMMISSION

~

WASHINGTON. D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIOil SUPPORTING AMENDMENT NO. 11 TO LICENSE NO. R-37 MASSACHUSETTS INSTITtJTE OF TECHNOLOGY DOCKET NO. 50-20 Introduction The Nuclear Regulatory Commission requested, by letter dated August 29, 1975, that the Massachusetts Institute of Technology (MIT) review its requirements for special nuclear material (SNM) and determine.

the amount of SNM that constitutes the " lowest acceptable quantity" (LAQ) of SNM necessary to sustain current operation of the MIT Research Reactor (MITRR). The letter further requested that if the amount of SNM currently authorized by the MITRR Operating License is greater than the LAQ, that MIT should request that lts authorization for possession of SNM be

, reduced to its LAQ. By letter dated September 25, 1975, MIT requested that

's its possession limit for SNM be reduced from 45 kilograms to 29 kilograms of contained Uranium 235.

Evaluation We have reviewed the analysis provided in support of MIT's September 25, 1975, requested license amendment and have determined that the licensee has accurately calculated the lowest acceptable quantity of SNM required to sustain current operations. Furthermore, our evaluation of HIT's September 25, 1975, submittal supports the conclusion that reduction of the.

authorized amount of SNM to 29 kilograms of contained Uranium 235 will allow sufficient operational freedom with respect to: (1) the storage of new, unirradiated fuel elements, (2) the storage of irradiated fuel elements, and (3) the reactor core uranium inventory. The proposed reduction in the allowable quantity of special nuclear material authorized for receipt, possession and use at the MITRR is an administrative change only and does not affect the conduct of operations at the reactor. The reduction of the quantity of SNM is consistent with the Commission's objective of authorizing only the " lowest acceptabic quantity" of SNM to licensees, and should not affect the licensee's ability to sustain current operation l

for the immediate future. On the basis of the above, we find the proposed p

change acceptable, i 8902270211 ADOCK890214 050000gG l

PDR F DL. l p

'., b-. ,

4 e

,# N..

CONCLUSION We have concluded, based on the considerations discussed above, that:

(1) because the change does not involve a significant increase in the probab.lity or consequences of accidents previously considered and doer.

not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

I November 17, 1975 l Date:

tN

\_.Y

\

l l

l

/^\

e l

~

5 O

TRANSACTIONS OF THE AMERICAN NUCLEAR SOCIETY 10TH BIENNIAL CONFERENCE ON l REACTOR OPERATING EXPERIENCE

" NUCLEAR OPERATION IN THE COMING DECADE" Supplement Number i to Volume 38 August 16-19, 1981 TANSAO 38 (Suppl.1) 1-66 (1981)

Cleveland. Ohio ISSN: 0003-018X f

O l

'. , . L - ,

. Test, Research, and Training Reactors-l 17 m

b

6. Use of Element Rotetion/ Inversion to '

increase MITR il Fuel Depletion, J. Bernard, Jr., D. D. Lanning, L. Clark, Jr. (MIT)

This paper describes the use of element rotation and inversion to increase fuel depletion at the modified Massa-chusetts Institute of Technology (MIT) Research Reactor.

The onginal core, which was cooled and moderated by heavy water, used widely separated 1691 elements. The fully rnoderated core had a low power density and achieved fuel {

burnups of 45%. The redesigned reactor, MITR il, uses a45 g l elements and, while surrcunded by a heavy water reflector,is light water cooled and moderated. The core, a compact hexagon ~15 in, wide X 2 ft long, has a high metal to-water ratio. Its 27 rhombic shaped etern .its are arranged in rings designated by the letters A, B, and C.They contain 3,9 and 15 elements, respectively. Six peripherally located shim blades and one regulating rod provide reactiv;ty control.

Soluble poisons are not used.

A The redesign provided an improved flux for beam tube,

Vi trace analysis, and matenals research. Its drawback was the potential for reduced element burnups due to flux peaks.

Figure i shows measured A. and C-Ring axial copper wire scans. ,2 The kRing, not being significantly affected by absorbers has peaks due to water at both ends. The C Ring, whose upper plates are adjacent to the shim blades, peaks at its base and drops substantially over its upper portion (B-Ring profiles resemble either the A or C-Rir's, depending on the plate.) The normalized radial flux vatses from t.25 in the A Ring to 0 90 in the mid B Ring to 1.54 at the oute*

C. Ring it was recogniaed that while these peaks only affected part of each element it would be necessary to discharge the entire element once that part attained the allowed fission density of I 8 x 102' fissicm .8The solution was to design the MITR-il elements so that they could be rotated 180 des and/or inverted. A calculational procedure, based on the CITATION code.3 was established to monitor fuel depletion 8 lt consists of the following-

1. CITATION calculations to obtain a three< dimensional full < ore. fission power distnbution at the equihbnum shim bank height to obtain a relative power for each element.

2. Modification of the calculated flux peaks by using measured shapes, This corrects the diffusion theory CITATION calculation at interfaces where matenal properties change abruptly.

3. Calculation of the fuel depletion using dedicated codes.

The MITR Il recently completed its first C Ring fuel pi cycle Each element was rotated twice and inverted once I g Rotations were performed when the elements were at 25 to d 30% and 75 to 80% of their useful life. The inversion *di l

l C______.________..___

( v, n . i 18- Test, Research, and Training Reactors-l 22'a n..-

[ /* u- .m  %

)

l'Q 19.6 -

t C-atng 16p  ;

is.g _

\s  : a n-2gsygn .,

_ . ggt,gge j-a i,

a.at Rin0 f :-

(*--I" II D'8I'I'8 O ble Ran* I I"'3

• m178-!! Core f -

0 la.0 -

- 3 f*[

j  % 4 4 j 9

11.2 4

1 n.2 -

\4 l . 15 810tesg]

.4 -

i, 5 4 n . ..

3

.a -  ;

)

l k

=

8.a -

\ {! 4 W J

l h 4 mita.il (lownt s sing y' C-ping

~5'6 il s.s -

\

e.

l f 2.8 'L k l

j 2.g -

fi .-

),;  ! I f 1 i 1....,,,* 0 3. 3 0. 9 1.0 t ,3 g,;

, sto so!trad U 235 trenar

( 0.5 1.0 1.5 2.0 2.5 pjg, 7, gxggj s28U distnbution.

Isonnali2nd Therri f!ur Fig. l. Axial thermal flux shapet 0 40 x 445 = 178 m MITR.il. Thus, the MITR il design including a rotationhnversion capability allows a factor of done at 50 to 55%.Similar strategies cust for the other nngs. 2.3 fewer elements in the fuel cycle,less frequent spent fuel (O/ but the timms is complicated since fresh fuel is gradually introduced and the custing elements are removed for reuse shipments, and possibly less concern about the diversion of nuclear matenals.

Figure 2 shows data from C Ring element 4 Mil. Its overall burnup was 40.4%. Had it not been inverted. it is estimated I G. C. ALLEN. D D. LANNING, J. W. GOSNELL, and that its burnup would have been 30 6%. inversion allowed a L. CLARK. The Reactor Engineenns of the MITR !!

32% increase m element lifetime. Without rotation, the over. Construction and Startup." MITNE.186. Massachusetts all burnup would have been even lower.Maumum depletion Institute of Technology. Department of Nuclear Enge occurred at the mid plane, with both ends having almost the neenng (1976).

name density. The radial assU distnbution has essentially no 2.1. A BERN ARD "MITR il Fuel Management. Core De-otation is performed using standard tools. Inversion E I'"' " b k. p fr DiUusion requires a special tool to protect the fuel plates and takes o gram . esis. . amchusens 5 mm per element. Reactivity effects caused by these "" " ' 8E' 'E' 'O "C "E* " '"

g9 manipulations have been measured to be small (a few cent / element t 3 T. B FOWLER et al . " Nuclear Reactor Core Anahsis

!! should be noted that the 3HU burnup per fuel element Code CITATION," ORNL-TM 2496. Rev. 2. Oak Ridge has been extended from 0 45 x 169 = 76 g in MITR l to National Lab (July 1971).

A b <

l I

\

.~ l

. 6 O

TRANSAC ICJNS OF THE AMERICAN NUCLEAR SOCIETY llTH BIENNIAL CONFERENCE ON REACTOR OPERATING EXPERIENCE "PLA':T STARTUP AND OPERATIONS IN THE '80s" Supplement Number 1 to Volume 44 August 1-3, 1983 TANSAO 44 (Suppl.1) 1-106 (1983)

Scottsdale, Arizona ISSN: 0003-018X O

l O

1

.L v ~ e .

g $.

=14' ' Training, Research, and Test Reactors-l 3

and moderated by hght water is heavy water renected. Six

i peripherai!y ' located shim blades and one regulating rod n()

k/

provide reactivity control. This design. which was selected for its many experimental advantages results in severe l

-l Cux gradients. Figure I shows A and C-Ring axial dus I profiles measured using uranium foils. Radial gradients also exist with the Oux across some elements varying by as )

{

much as 1.4 Refuehngs occ.ir at roughly six *eek intervals. Fresh.. i; fuel is introduced te, the B Ring where peaking. but not l element power production. is lowest. Partially spent fuel is {

cycled through the A and C Rings. The equivalent of eight j elements are used per year. Power distnbutions must be

]

_ accurately characterized to demonstrate comphance with heat transfer hmits as well as to allow the use of element rotation inversion thereby avoidir.g the premature discharge of fuel. Given the frequent changes m core configuration necessitated by refuehngs, a rehable means of theoretically predicting threedimensional G D) power distnbutions was needed. Initial studies revealed two problems. First, the available codes, such as CITATION 2 were based on dtffusion theory and, hence, were inherently incapable of accurately predicting Oux shapes at interfaces where matenal propertie$

changed abruptly. This is discussed by Henry.3 Second, most codes did not permit suffident differentiation of matenal densities to accura;ely model the partially spent fuel.

I I I I I I 26 -

. g* _. m e-ft eer.

Sham peg

} 22 i Plad "

\ . l h(c.d

/ /%  %

?1 -

1. #% N .=

l _

is =

$; -Y \

A-Rin - ing -

- 16 g e-Rina

, tsni- e.o 4 t 3 -

3 . .- .

_: 8 u

d -;. .

m -,

p A in .? . .,

J M i o .

[4 8 d 5 ---- : - a m -

E j

1

5. Calculational Procedures Used to Extend i MITR-ll Fuel Sumup, J. Bernard. K. Kwok D. [q* ~

Lanning, L. Clark (MIT) ~

This paper desenbes calculational and expenmental 2 - e -.

methods developed to estend fuel burnup at the Massachu- . .

setts Institute of Technology's 5 MN th e research reactor. o 1 a MITR II. This report complements an earber one w hich ~ ~~

discussed element rotation and inversion a a neans of increasing fuel depletion.' - -'

The MIT reactor was redesigr,ed and renuilt bet w ee n -

j. -

i i i 1974 and 1976. The modified factht). MITR il. has a . ore ' I I l i I that is a compact hexagon 15 in. wide a : tt long *ith 05 "O "5 U M l0

/ positions for 27 rhombic shaped elements # 445 g 2 "l i in e r-a l u e:f de r-a l rio

\ rings designated by the letters A. B. and C Thes .ontain 3.

9. and 15 elements. respectively. The core. * *uch is cooled Fig I Aual thermal flux shapes i.

-- . _ . . _= .. _ - _ - . . _ _ _ _ _ _ _ _ _ _ _ - - . _ _ - _ - _ - _ _ _ _ _ _ __ - _ - ___

3 ,, -

7. - -*

Training, Research, and Test Reectors-l - 15.

l The first problem-was solved by correlating theoretical. 4 Modification of the calculated fission power peaking -

p)

D' and measuhd power shapes and then using that correlation to modify subsequent theoretically generated shapes. Mea-surements were made using plate scans, copper wire scans,-

by using measured shapes.

5. Calculation of an equilibrium senon distnbution.

or uranium foil scans. The correlations for.a given fuel 6. Repetition of steps 1-4 at the tenon equihbrium bank channel did not vary with core configuration but did change height. "

as a function of shim bank height. Figure : shows the .,

correlations obtained for the & and C Ring Gux proGles M d@w m MM shown earlier. The quantity ' plotted is measured power # d

divided by predicted power as a function of height. Unity Fuel elements are routinel> depleted to 40 to 447 for imphes a perfect predictions As expected, the correlations an average consumption of I?8 to 105 g element. This are poor at the ends of both elements and worse overall for calculational methodology would benent any facibty with the C Ring, since that is where there are abrupt changes in severe flux gradients. Furthermore, as noted in Ref. I. this matenal properties techmque, when coupled with the MITR il design and the

'The second problem was solved by decouphng the capabihty to rotate invert elements, has allowed a factor eigenvalue. power proGle ' and fuel depletion calculations. of 2.3 fewer elements in the fuel cycle than was the case Separate ; codes were wntten that computed the fuel deple- with MITR L tion by first modifying the CITATION 3 D power shapes in .

accordance with the expenmentally determined correlations I. J. BERNARD. D. LANNING. and L. CLARK "Use of and by then applying the mod Ged shapes to a detailed core Element Rotation inversion to increase MITR II Fuel model. In addition, CITATION was . modified at MIT to Depletion / Trans A m. .Vuct ' Soc., 30. Suppl. l. I-allow $00 matenal descriptions. ( Aug.1981 t The overall procedure

• for reactor physics and fuel depletion calculations on the MITR il is  : T. B FOWLER et al.. " Nuclear Reactor Core Anal > sis Code; CITATION." ORNL TM 2496, rev. 2. edition J.

~ 1. Construction of a xenon free explicit eore model Oak Ridge National lab. tfuly 1971).

with shim blades at the cold entical position.

3, A. F. HENRY, .Vuclear Reactor Analvsrs. pp.1:1 124

2. Homogenization of that model to be compatible with MIT Press. Cambndge, Mass. (1975 L CITATION. (Ref. 4 discusses cross section denationsa 4 J. BERNARD, "MITR Il Fuel Management. Core Deple-

3. CITATION calculations to obtain a FD. full-core.. tion, and Analysis: Codes Developed for the Diffusion Gssion power distnbution. The 6 R Z incore mesh is Theory Program CITATION," NE Thesis. MIT, Depart.

30 x 10 x 14. ment of Nuclear Engineenns (19h I

i i s ii s a g i r i i r i i

- 12 .=

p".- ,

20 -

) -

4 -

r

)

g -

is ,i g.. (

Im -

4-i - i -mg -

0 n.

0 i4 =

N .

' q!

p ji _

g' q,j 10

,, m . v .-

A- Aifia f

e'

~+

y

.').

J . -

,b,, - ractor a ."'.*. A ri 3F ""* we r ', 5 d.

~

l

-l 3 ,- k, . - w. . . . - 1 I f f I t i t iI t ' ' ' ' '

.i 3 i:

[

\ .

power 03rrelas. Ts  ;

Fig. : Asul power correlation ru tor l

l l

_- _