ML20098G318

Forwards Addl Info Re Application for Renewal of License R-79

ML20098G318
Person / Time
Site:	University of Missouri-Rolla
Issue date:	10/01/1984
From:	Bolon A MISSOURI, UNIV. OF, ROLLA, MO
To:	Thomas C Office of Nuclear Reactor Regulation
Shared Package
ML20098G319	List: ML20098G318 ML20098G320 ML20098G321
References
NUDOCS 8410040355
Download: ML20098G318 (15)

Text

e. .

l Nuclear Reactor Facility  ;

Nuclear Reactor II Rolla, Missouri 65401 Telephone: (314) 341-4236 UNIVERSITY OF MISSOURI-ROLLA Docket No. 50-123 October ', 1984 Mr. Cecil 0. Thomas, Chief Standardization and Special Projects Branch Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Thomas:

Attached is the additional information regarding the application for renewal of the operating license for the University of Missouri-Rolla Reactor which you requested in your letter dated August 27, 1984.

We hope that the enclosed information will be satisfactory to you and i your staff.

Sincerely, Albert E. Bolon Director AE8:gf Enclosures cc: Don Warner Robert E. Carter Appeared before /(//foA/ 5, #4 D CM a Notary Public for Phelps County, State of Missouri on first day of October 1984.

AA W E $ it $ 8 )

Signature of Notary Public it"TN S. f/ ATLOCK T OTMY FU . L!C, C TA TZ G; ^~

< CC: AIL ; u :i 8410040355 841001 h0 DR ADOCK 05000123 \

PDR \

an equal opportunity institution

D---*** Electronic Technician's Office Reactor Engineer's Office Upper Level Reactor Director's Office N

0--y ..

\Sj UP --

h counting N Room _

_. y

~

Control i Room r N h 4'A Outer _

"*i'"

49.5ft Office 13ft Storage Pool Q*"~'"

W 1f 8ay Area lc

, 9ft +

H

= 33.5ft  :

Main Floor Figure 9. Facility layout.

I Rest Room Storage- g

/

c: UP D.  ;: /

[

t __/

Y Work Bench Lavatory  ;

O j

^

l Sump Pump Demineralizer o O Compressor Intermediate Level C 33 ft =

H Retention Tanks -: UP 17ft P

Sump Rails i l O l Thermal l o l Column l _

I I

.\. Beam Port l l Basement Level l- Figure 10. Facility layout.

A*

Drive Assembly-Fission Counter

.I I Bridge Assembly WJ l 5 A

V' Y. - . iN

/ *. y*

e

' * ; a- Tower Assembly

._..._h. .

i

{

% 4. h-r q> .g.g

.i ,' . e w

• qg. J

-:.. . ., e t f:543 O  ?!.lia&

Compensated Ion Chamber Guide Assembly

%y

• i

.. .e Uncompensated Ion il ,f ',p,'yl.3,.; ' f~k . Thermal Column

r Chamber Guide -

..,[,. g 4.: e.1,,4 5p

. Door Assembly Assembly Grid Plate '.

.s.. s .

r 4:.g ..'* ,*:'y

p. I s..L 7'3 'Fue1

, e'*Thermal Column I ' '

i i n y -

. . . i

[

[

• r P4 Assembly  ?.'3..W..n[.%.h';lf){f,is

, ' , 2 Rack A*?y.n';p. ,:.,

r,g. ,.kg:,

L e;, "

+. *;.'9 .

~- .  ; ~, .,

. , ,; .

• O. .. . . :t : ?,.. ..,': .go: r a;

.4. . .~. , c' k2 -[

I-

$,s.P,e#ed:

. . ; . y.;,

[*

Figure 16. Cross section of the UMRR.

Mounting Assembly Control Rod

-.G p .

7 Drive Assembly, Safety / Shim Rod

. m -

, e-y Drive Assembly, Reg. Rod

., =  : = _

y 0 I [

Z :  :

i

.

I' : ,

e Q.

7 pp Q: u At:

i

^

M' l h.

T g '

[ Magnet Extension Assembly-Safety / Shim Rod 1;

/

/

/ g Drive Extension Assembly-

/

/ Reg. Rod

- -l ;

/ e Core Access Assembly-Model 59A

~ - ~

iff /

$ f A)[r' f

U

1. blli L

. . =  ; g!. -

V/a Core Assembly

, g  ; , , -

6 Beam Assembly

-9[:j 3IPI'_.  := '

r

[

n y b [

L-

,*f.E's

.  :*?d 4*.+

W.eygatf $, a :2 : '$ '

• e!?.'?.:;*.9.ry ,y,

, d ** #,=.' .

e y

r!?5.'$*.I'h:a'i SECTION A-A' Figure 17. Cross section of the UMRR.

e 5 6.12" e c 34' =

+ + 1.5" 4'3 +- -

A 8

's$9 4Y h h i I

, _ _ _ _ _ _ _ - _ _ , ---- --r------- l w :- - - - = - _ - - _ _ ~ . _ -

_r_______

43.6

-_r------- i_

" , _ _ _ - - - - - - g______[______,

l_ - - - - - - - -

g

_ -4_._  !

75.O" _ P - - _ _ __ _ __,_

32.6 W

• +A N 2. ,

M i y y +B

, jf * -as.5' ~

• !, '28.v .

I g, q

, p i [RI I I I I I DU I 1 , , __

l a 23.75" n

i l

+ 4 _

I l

g l' 478 I

+ i 1r l El l I I l l ':< 1 l + -- -

I ', i L____ __ _ __J U SECTION A-A , o7g.

< 48.0" % '

SECTION B-B Figure 20. Thermal column.

1 1

l

BEAM HOLE t.lNER SHUTTER PLUG ASSEMBLY ROTATING SHUTTER ASSEMBLY

78'

LEAD SHIELD PLATES (5 Totol)

' TOTAL THICKNESS = 2.25" W

, N / WATER l k \===-:==========.::====== ,-

,.g l

j -

M,\ See References N Dimensions of Shutter =

i For :i

= 38.5" = = - 23.25" c 64.25"Plue Assembly l

NEUTRON BEAM PATH APPROX.-1/8"TO 1/4' GAP i

I i

! Figure 21. Beam tube with shutter assembly in place,

. . . .-. _ _ . . _ _ _ . _ _ - . ~ . . . _ . . . . . . _ . . . . _ . _ _ . . . _ . _ _ _ . .. . . . _ _ . . _ _ - . _

• {

4 i

., g. .

7. .

p i

!. . RESPONSE T0 i-

. - OPERATING LICENSE RENEWAL

-FORMAL REVIEW QUESTIONS-t, i

f.

• University of Missouri-Rolla Reactor ,

i (UMRR) 1: '-

. Facility License No. R-79  ;

i ,

i Docket :.o. 50-123 t

i i

l l

j.

i-i L

September 28, 1984 G

e ,

• • "v-Pe r ,. ., , _ -. . e e m -ns- n- .-w-We+

L- ..

o j a

Introduction l

l: Attached is the additional infomation requested regarding the application

, for renewal of the operating license for the University of Missouri-Rolla

. Reactor, License No. R-79.

The information is presented in a question-and-answer fomat with references

. made to .the appropriate page in either the revised Technical Specifications -(TS) or the revised Safety Analysis Report (SAR).

I LO O

2 s.

1. .Show that K eff for the fuel storage pit will not exceed 0.9 if all

!) positions contain fuel.

Experiments have been conducted to determine k for fuel eff properly stored in the UMRR fuel storage area. The maximum value for one storage rack filled with standard MTR-type fuel elements was determined to be less than 0.6. No neutronic coupling was observed when additional fuel elements were loaded into the other parallel rack, which is 15 inches (38 cm) between centerlines from the filled rack.

[The fuel storage pit is mentioned in section 5.4.1 of the TS.]

2. Section 3.3(1) of your proposed Technical Specifications states that the reactor shall not be operated unless there is at least 16 ft of water above the core. How is this assured?

A U Immediately prior to starting up the reactor the Reactor Operator visually inspects the pool'and the core as a Standard Operating Procedure (S0P). In addition, the R0 on Duty can observe the level of water from the control room during reactor operation.

There are 5 feet of water above the intake and discharge pipes, which are themselves 16 feet above the top of the core. The water level is checked daily and recorded on the daily checklist in accordance with S0Ps.

At least one of two control room annunciations would be actuated if the pool level began to drop significantly. One ,

would be the Radiation Area Monitor on the bridge (which is a control channel), the other would be the High Sump Water Level on the facility sump.

. Thus, it appears very unlikely that the reactor would be v operated (intention 311y, or otherwise) with less than 16 feet L

w .

3 L

g of water above the core.

O

-[The'16 feet of water requirement is given in section 3.3(1) r of the TS.]

3. -What is the reactivity worth of a fuel element in (a) the most reactive position within the core and (b) at the periphery of the core?

(a) Since the current and proposed Technical Specifications both disallow a vacant grid position surrounded by fuel it has not been possible to determine the reactivity worth of a fuel element within the reactor core.

[The lattice position requirement is given in section 3.1(2) of the TS.]

(b) Experience at the UMRR has shown that the reactivity worth of a fuel element at the periphery of the core is less than 1.5%

X- delta k/k. .

[See section 9.6 of the SAR.]  :

4. The Final SAR should include an analysis of a hypothetical accident involving the step insertion of the normal maximum excess reactivity (1.5% delta k/k). There also should be an analysis of a step insertion for a core with 3.5% delta k/k excess reactivity. This latter analysis should be compared with the worth of a fuel element added to the core periphery.

No accident was identified which could possibly induce a stepwise insertion of the normal maximum excess reactivity (1.5%

. delta k/k) or the excess reactivity of 3.5% delta k/k. However, a hypothetical accident scenario was postulated in that a fuel element would be accidentally placed next to a barely critical 3

(V

4

,_ core. The worth of the fuel element was assumed to be 1.5%

b,s delta k'/k.

[An analysis of this hypothetical accident is presented in Section 9.6 of the SAR.]

' 5. Discuss the startup accident in detail referring to the original Hazard Summary Report (HSR) as a source.

[A startup accident is discussed in detail in Section 9.5 of the SAR.]

6. Provide an accident analysis for the irradiation of an experiment containing special nuclear material. Specify quantities and irradiation times. Consider the effects of loss of fission products to the reactor room and the environment.

p [ Provisions for fueled experiments are addressed in section

's J 3.7.2(3) of the TS, and analyzed in section 9.7 of the SAR.]

7. . Provide up-to-date (as-built) drawings of the reactor facility and the experimental facilities.

Up-to-date (as-built) drawings of the reactor facility are included as an enclosure to this letter.

8. Describe the administrative organization of the radiation protection program.

-[The administrative organization of the UMR radiation protection program is described in th'e first paragraph of Chapter 7 of the SAR.]

b G

5 X 9. Describe the responsibilities of the Radiation Safety Office for the U reactor facility operations.

[The responsibilities of the Radiation Safety Office for the UMRR operations are described in Sec. 7.2.2 of the SAR. The relationship between the Radiation Safety Office'and the Reactor Facility are shown on Figure 6.1 (in section 6.1) of the TS, and a similar diagram Figure 25 of the SAR.]

10. Identify the radiation safety related tasks that are performed routinely by the operations staff.

The only radiation safety related tasks that are routinely performed by the operations staff are the following:

i calibration of the radiation area monitors, calibration of the portable health physics instruments, collection of 11guld samples,

(]

V monitor samples and other radioactive objects as they are removed from the pool.

[The radiation safety related tasks routinely performed by the reactor staff are identified in the second paragraph of section 7.2.2 of the SAR.]

11. Describe thc radiation protection training for the non-Health Physics reactor staff.

[The radiation protection training for the reactor staff is described in section 6.1.4 of the TS and section 7.2.4 of the, SAR.]

12. Summarize your general radiation safety procedures. Identify the minimum frequency of. surveys, action points (levels), and appropriate responses.

t

[The general radiation procedures are described in section 7.2.2

6 7.

of the SAR. The frequency of surveys, action levels and appropriate O responses are all described in that same section.]

13. Descrioe your program to ensure that personnel radiation exposure and releases of radioactive material are maintained at a level that is "as low as reasonably achievable" (ALARA). Identify steps taken to implement the ALARA principle.

The campus administration's commitment to the ALARA Principle is very definite.

[The ALARA Principle is cited in the objectives of TS 3.4 and 3.6.1, which relate to the confinement and the radiation monitoring systems, respectively. Section 7.1 of the SAR further describes the program. A copy of the official policy as established by the Vice Chancellor, Administrative Services, is shown in Appendix B of the SAR.]

p d 14. Describe the liquid radwaste management program. Specify the locations and sizes of hold-up/ storage tanks; summarize the sampling procedures and analytical techniques.

[The Technical Specification requirements concerning releases of radioactive liquid effluents are given in section 3.6.2(2). The liquid radwaste management program is described in section 6.2 of the SAR.]

15. To the maximum extent possible, you are requested to incorporate the responses to the above questions in the revised Final Safety Analysis Report and revised proposed Technical Specifications that you have developed in recent months. This letter solicits their formal submittal.

However, please identify the answers to each question in a clear and explicit way.

The Technical Specifications and Safety Analysis Report include

{}

the appropriate information about the current design and operation of the facility, including responses to the formal questions from the

- .. .. .. . .. . . _ . . - . - . . - . . . _ . - - . _ . _ . - - . . . . - . = -- _ - - - - - - - . . . - . . . . - - .

A L.

p- . 7 0.

'L f,

.. NRC~ staff. The-two documents are hereby-presented as part of the

!b .

formal submittal for the operating' license renewal.

~.

^

4-I l t' l t i

j. '

t i J

I 1 ,

i 3

i +

h I

P s

1. i i.

~

i ,

}. .

(- s 1

I. i t

I i

b I k r l- I i .t

t. ,,

i.

t i

!. I c

1

, [

r

. i 5

l9 i i

b 1

'-.--