ML20054J723

Amends to 800229 Application for Renewal of License R-71, Changing Pages to Include Recent Financial Data & Improve Clarity

ML20054J723
Person / Time
Site:	05000142
Issue date:	04/30/1982
From:	Oneill R CALIFORNIA, UNIV. OF, LOS ANGELES, CA
To:
Shared Package
ML20054J718	List: ML20054J717 ML20054J723 ML20054J726
References
NUDOCS 8206290492
Download: ML20054J723 (41)

Text

- - _ __ _________-___ _

&q l O ^xtnosents To Application for Renewal of the UCLA Research Reactor License R-71,1980 Section Pages to be Removed Pages to be Inserted (4-30-82)

Front Pages " General Contents" following Title Page (1 page) transmittal letter (1 page)

Table of Contents following Table of Contents

" General Contents" (14 pages) (1 page)

Appendix I I/1-1 (1 page) I/1-1,2 (2 pages)

I/2-1 (1 page) I/2-1,2 (2 pages)

Appendix II II/ii (1 page) II/ii (1 page)

II/2-la (1 page) II/2-la (1 page)

II/A-4 (1 page) II/A-7,8,9,10 (4 pages)

Appendi. III III/i.ii (2 pages) III/i,ii (2 pages) 111/1-1,1-3,1-5 (3 pages) I I I/1 -1,1 -3,1 -5 ,1 -6,1 -7 (5 pages)

III/6-2,3,4,5 (4 pages) III/6-2,3,4,5 (4 pages)

III/8-1 (1 page) III/8-1 through 8-13 (13 pages) 111/10-1 (1 page) III/10-1,2 (2 pages)

Attachments A & B (16 pages) Attachments A & B (2 pages)

Appendix IV ALL (including Attachments Reserved A.B.C.D New Appendix IV New Appendix V Appendix V ALL O

8206290492 820623 PDR ADOCK 05000142 P PDR

O APPLICATION FOR A CLASS 104 LICENSE FOR A RESEARCH REACTOR FACILITY Based on Code of Federal Regulations, Title 10, Part 50 to U.S. Nuclear Regulatory Comission O

R. R. O'Neill, Dean School of Engineering and Applied Science University of California Los-Angeles February 1980 AMENDED: April 1982 O

Title Page 4-30-82

. -m.. . . - _ . .. . . . . . . = . _ ._ _ _ _ _ _ . _ _ . _ _ __ ____ __ _ _ . . ~ . . _ . _ _ _ . . _ . _

i LO AeetICATI0n F0a A CtASS iO4 LICENSE 4

[ FOR A i

i RESEAERCH REACTOR FACILITY i TABtE OF CONTENTS i

4

---------- Application I

j Appendix I Financial Qualifications of the Applicant i

i Appendix II Environmental Impact Appraisal 1-

{.

Appendix III Argonaut Safety Analysis Report (ASAR) i j Appendix IV Emergency Response Plan j...

Appendix V Technical Specifications 1 Appendix VI Operator Requalification Program.

i i

4 1

t I

IO F

4-30-82 .

j u .. _ . . _ - - _ . . . . . . _ . , , _ . , _ . _ . . _ . _ . . = . _ , _ _ _ . . _ . . . _ , - . . . _ . , - . . . ~ - . . _ _ _ _ , . - _ _ _ . . . . _ , . _ _ _ _ _ , , - - . . _ . - . . , _

FINANCIAL QUALIFICATI0'15 0F THE APPLICANT 1.0 GENERAL DESCRIPTI0il The University of California is a land grant college that is financially supported by:

a) annual appropriations from the State of California; b) federal and state contracts and grants; c) student and other user fees; and d) private gifts and endowments.

There are nine (9) campuses of the University and several laboratories.

The "Systemwide Administration" of the University is located in Berkeley, California, adjacent to the University's Berkeley campus.

This application pertains to the Nuclear Research Reactor operated by the Nuclear Energy Laboratory (NEL) of the School of Engineering and

@ plied Science (SEAS) on the University's Eos Angeles campus (UCLA).

To the extent that the State Legislature provides funds annually to support the University of California, those funds are distributed to each of the campuses. The funds received by UCLA are further distributed to the various Colleges, Schools and Departments and support activities. Direct support of the NEL derives principally from the operating budget of the SEAS. The SEAS budget is part of the budgeted support for the Los Angeles campus of the University of O California. (See UCLA Financial Report, which appears here as Attachment "B"). In addition, the NEL is directly supported by the services of a resident health physicist who is funded out of the budget of the Office of Research and Occupational Safety, an administrative support unit of the campus. UCLA provides indirect support in administrative, custodial, maintenance and surveillance services. Although the actual dollar amount of this indirect support to the NEL cannot be ascertained, an approximation of this amount can be made by applying the rate that is negotiated periodically by the University and the federal government for the recovery of the indirect costs of supporting federal contracts and grants. The current indirect cost rate is used in the total cost analysis that is provided below.

In addition to the direct and indirect support provided by the University through the SEAS at UCLA, the NEL is supported by the recharge income it receives from technical work performed by the NEL staff on contracts and grants of other departments and, to a lesser extent, the fees that are charged for providing reactor services to both academic and non-academic users of the research reactor. The total amount of recharge income and user fee income that is received varies widely from year to year. The initial SEAS budget appropriation is based on an estimate of total expenditures and total income from whatever sources. During financial closing at the end of each fiscal year, the SEAS NEL appropriation is adjusted upwards or I/1-1 4-30-82

downwards to ensure that it equals the total of NEL expenditures less all sources of NEL income.

Funds of the School of Engineering and Applied Science support a broad range of academic programs in furtherance of the University's teaching and research mission. The UCLA Nuclear Energy Laboratory is one such program. Periodically, these programs are subjected to academic review by the faculty of the School. Based on these reviews, recommendations are made to the Dean for continuing financial support. Subject to the availability of funds from the State of California, continuing programmatic need, and continuing positive recomendations by the faculty, the NEL will be maintained at a relatively constant level of financial support adjusted, as needed, for normal increases in costs of operation.

O O

I/1-2 4-30-82

O 2.0 Es11na1E0 ANNUAL C0sT Or OPERATIONS-The estimated total annual cost of operating the UCLA Research Reactor is the cost of operating the NEL, adjusted to exclude costs associated with the non-reactor-related activities of the laboratory and to include other direct and indirect costs that do not appear in the budget or expenditure statements of the NEL. For the 1980/81 fiscal year these costs are given in the following table which is 6dapted from the cost accounting data prepared by the UCLA Finance Office. A more completc explanation of NEL operating costs can be obtained from UCLA's letters of January 25 and April 19, 1982 to the Commission.

UCLA Nuclear Energy Laboratory 1980-81 Financial Cost Statement Total Net NEL Non-Reactor Reacter Budget Costs Costs Salaries - Permanent Staff of $163,531 $49,805 $113,726 6 FTE Salaries - General Assistance 38,265 0 38,265 Employee Benefits 34,288 10,459 23.829 Supplies: $43,406; Equipment: $3,641; Travel: $712 47,759 0 47,759 TOTAL NEL Expense $283,843 $60,264 $223,579 Additional Expenses not reflected in above totals:

Health Physicist- Salary 28,000 0 28,000 Health Physicist - Employee Benefits 7,266 0 7,266 TOTAL EXPENSE $319,109 $60,264 $258,845 Indirect Costs 031% MTDC 97,795 18,682 79,113 TOTAL NEL COSTS $416.904 $78.946 TOTAL Reactor Operating Costs E-Total NEL Expense represents the amount that the NEL had to budget in -

fiscal year 1980/81 for all its operations. Budget support for the Health Physicist is provided by the Office of Research and Occupational Safety. The precise amount of the indirect costs ~of reactor operations are unascertainable; however, they are well approximated by the indirect cost rate that has been estabitsbed for the University as a percentage of modified total direct costs (MTDC),

! "I 4-30-82 m_

~~N (V that is, direct costs less eq11pment. Indirect costs are recovered for the campus as a whole and are not identified in the budgets of individual units such as the NEL. It should be noted that the University's accounting systeu does nor ordinarily distinguish, within

> the NEL accounts, reactor-related costs from non-reactor related costs.

As one consequence, all of N"L " Salaries - General Assistance" are reported as reactor-related expense. In fact, it is only the salaries of part-time student reactor :;perators (perhaps $2500 of expense) that is reactor-relatcd. The bal mce of the part-time salary expense in this category is related to non-reactor projects and activities of the NEL.

In addition to the SEAS appropriated support, the Nuclear Energy Laboratory derives funds by recharging other campus units for

' technical assistance provided to specific contracts and grants and by charging fees to both academic and non-academic users for reac+.or services. Support for the Health Physicist (who is budgeted out of the Office of Research and Occupational Safety) and for indirect costs (which are recovered far the campus as a whole) are not considered as sources of funds for NEL operations. The NEL does not regularly issue annual reports of a fiscal nature, however, the approximate distribution of fund sources for the past four (4) fiscal years is shown below.

NEL Sources of Funds O rIScAt nAa: auiy ist to aume 30tn 1977-78 1978-79 1979-80 1980-81 SEAS Appropriation $131,187 $127,636 $151,735 $189,724 Reactor User Fee Income 9,170 11,130 21,000 33,855 Non-Reactor Income 71,675 55,923 67,180 60,264 TOTAL SOURCES OF FUNDS $212,032 $194,689 $239,915 $283,843 k

V I/2-2 4-30-82

APPENDIX II ENVIRONMENpl_._IMPACTAPPRAISAL LIST OF FIGURES j PAGE Figure II/2-1 Film Badge Locations Nuclear Energy Laboratory - 1st Floor . . . . II/2-2 Figure II/2-2 Film Badge Locations Nuclear Energy Laboratory - 2nd Floor . . . . II/2-3 Figure II/2-3 Film Badge Locations Roof View Overlooking Stack . . . . . . . . . II/2-3 Figure II/A-1 Roof View Overlooking Stack . . . . . . . . . .

II/A-5 ,

i Figure II/A-2 Average Quarterly Dosimeter Readings. . . . . II/A-6  ;

Figure II/A-3 II/A-9 j TLD Locations . . . . . . . . . . . . . . . .

1 i

_ LIST OF TABLES 3

1

Table II/2-1 History of Annual Releases ......... 11/2-5 1 Table II/A-1 TLD Readings. . . . . . . . . . . . . . . . . II/A-8 LO i

i s

i i

4 i

]

i i

j O  :

i II/ii

4-30-82 i

i 1

O O O.  ;

1 i

i

IDENTIFIER BADGE NUPEER HIGT ST Nt0AL DOSE RADIATIO 4

! a

x 219 0 mrem -----

4 b

x 2048 0 mrem ----- l c [* Note - this badge includes x 1965 1200 mrem 8,y, x-ray

• x-ray radiation from the  !

d Tokamak Laboratory]

i x 230 0 mrem -----

r x' 1581 0 mrem ----- I f

! x 220 0 mrem -----

2 eM i .C x9 1914 0 mrem -----

1 -N h

j h x 218 110 mrem 8,y i x i 265 350 mrem 8 only ,

4  ;

l x 3 203 425 mrem B only 1 .

l x 820 0 mrem -----

[

1 x 302 0 mrem -----  !

4 1

LOCATION AND MEASUREMENT GUIDE TO FIGURES II/2-1, 11/2-2, and 11/2-3  :

?

m t

m ,

N

i

. 5 f.

4  !

f c

l D

b UPDATE OF ENVIRONMENTAL MEASUREMENTS Several developments have occurred subsequent to submittal of the Renewal Application dated February 1980. Firstly, as a result of Question 8 posed by the Nuclear Regulatory Comission on July 31, 1980, UCLA performed a theoretical analysis of plume dispersion based on a Gaussian plume model and showed that such analy!'s correlated with the previously described dispersion measurements of Rubin.

(Analytical results forwarded to the NRC on 9-5-80). Using this dispersion model, the Commission performed calculations of the attendant radiation levels on the roof of the Mathematical Sciences building assuming (conservatively) that the prevailing wind would be realized 100% of the time. These calculations resulted in an estimated dose of 1.4 mrem per year, and hence lead the Commission to respond negatively (on September 24,1980) to a petition to shutdown the UCLA Research Reactor (Director's Decision under 10 CFR 2.206, DD-80-30).

In addition to these calculations, UCLA initiated a new environmental measurement program utilizing Thermoluminescent Dosimetry (TLD),

beginning on August 20, 1980. As a result of what was learned in the 1976-79 monitoring program, dosimeter locations were chosen to minimize the effect of the natural rad *oactivity of concrete.

In general, all dosimeters were placed on non-concrete structures (wood or metal); however, two dosimeters were located in concrete O

V parking structures remote from the reactor to assess radiation levels attributable to concrete. All dosimeters are changed and read quarterly (every three months). Commencing with the second quarter of the study and thereafter, four dosimeters were transferred from raingutters to lead bricks with the bricks interposed between the TLD and the nearest proximate concrete.

The results of the six quarters of TLD observations are shown in Table II/A-1. The geometrical locations of the TLD's specified in that table are graphically illustrated in Figure II/A-3. Starting in the second quarter, lead bricks, 4 x 4 x 2 (inches) were used at locations A, B, D, and E. The bricks were placed on the top surface of the flat roof structure with the TLD fastened to the top of the brick, The brick orientation provided 2 inches of lead shielding between the TLD and the concrete structure. Dosimeters in locations C, G, H, I, J, K, L, and M were fastened to, respectively: the sheet metal of ventilation systems (C, J, M); telescope and planetarium domes (H, K); a wooden housing for meterological equipment (I); and cooling tower windscreens (G, L). TLD F was placed within the exhaust fan inlet plenum chamber and is analagous to TLD No.3 mounted on the stack top in the 1976-79 series.

This monitoring program was initially designed to use thirteen (13) dosimeters at locations A through M. The vendor pricing policy favored using sixteen (16) dosimeters, hence locations 0 and P were added for the specific purpose of assessing radiation from concrete. Location N II/A-7 4-30-82

/ T l }

LJ

- Se ao e e e n n n n sa e o + o o a m o + 2 m m e + *

e. e, = .e.

.e .e i

ee e e M N tt tt es e o e e o

• e .*

• e e ce e .re a. e ce to e.e .o e e e-sue

-m l a

e e < H H H H $

4e ' e o em o o

.s.

as e o se s e 4 -

ce en.

e, re -e se en l

i.. I.

e e i n

• x x e o o o e 2 + ~ s e 2 - +

or te o s

+4 - 4-e e me l

^ l Eg i a-

e. "

Cr

.e=.e m N M o o a o .

tt

c. s a e == in s a ye e s =? z - .--

w e en

.e 0

7~ q m f 4 % &

\'j ' W

.C.- ag u e e e e e a ea em ce o e s e o o ==

g g 4 eo

1. e.

- e. ee

& O e e

.O CE =a 4

& Q J

Y.

r o

o eu

. s l 2 b

.a >

e. r g e 5 v - 4 as 2 4 - V v >

4 a 2 a + g g . g 0.a l

- , < - 2.

. . - . .r

.x .

e. . .. a O o e a e # se

• I l

o o -Z e. . . em e.w c.

.se. -

d.,

e ,

o t .

5. ,

. . 2 3- -

.z -.- o~ - - .

. .. ., . v -

, . ..o,e E =, ,. ., .,.

- 3 3 2

- . es. .. e,e C O O O > . O O O O o O ee J 0 - . m

== .> > .Z >- e se v > > > > 2

• 2 2 . **

l

- , 3 - 8 e . c o .e o o z o, o o ooo o o o o .o . . . o, .-

l 80

. . r o. o.

sa es, , . . . .o .o . > . .

4 g e (

(

a .4 .

i r

, , o- --

C 2 O e 4; e u O w k o I - "3 . ., E Z o . N 3 l

n)

(

v'%/

/

l 1

I II/A-8

! 4-30-82 l

l l

l v

l ll l

O

_ l iIN X

/ -

S E

E L

P C

N L M E L O I

C J A C S H E C

L A g g LM RE N

E C T I 31 I T _kN L E

C S

A y# h O H M j B E E T

E H

i i

A \ RR A T /p UU M A TT N CC M UU O RR S R I P

T S T N O S O I

T A SG G T I

' C M OI UNN s N = A A d[

n NI K K I

O E C o OR O R

AR R I

i T D L E

VPA A t

A M a L E S P G c O I K T

N A E T Y O E

I R

R L S

A NOP

(

- N L

P I

L D

o VIN MT O TK M M A

L T

OA .

S NV C C V '5 OR A AE87 E TL F .

C}I RE TS SB RSE4 O O

3 A

R /

AO I I

] " e r

B u F  % g A

O O SL F p/% G i

F O

RM O SL R* 9 RM O8 O4 R '5 L . O7 FVE O4 L .

h L t FVE 9E h L t

8E O

. CI* , n l l

was chosen to replicate an earlier location where the average measured dose of 8.4 mrem per quarter was somewhat intermediate between values typical of concrete-mounted dosimeters and non-concrete mounted dosimeters. It was the only dosimeter mounted on a lead brick in the first quarter. The value for the first quarter was very low, but successive thefts of the lead bricks in the second and third quarters discouraged the continued use of that location. Therefore, dosimeter N was relocated on a wooden tower; however, it was somehow displaced during the quarter and the reading for the fourth quarter was compromised. Although this badge remained on the tower during the fifth quarter, a decision was made to mote the dosimeter to an entirely different location. For the sixth (and current) quarter, the dosimeter has been mounted on the windscreen surrounding the stack.

The location is symetrical relative to concrete walls and parapets, and relative to the TLD in the exhaust fan intake plenum. The objective has been to distinguish between an immersion dose and a background dose in otherwise similar locations.

TLDs 0 and P were placed in parking structures north of the reactor building for the first three quarters and then placed in parking structures generally west of the reactor for the next three quarters.

The location change was made to broaden the sample base.

The radiation levels seen by the TLDs in parking structures (12 readings) averaged 66 mrem per year whereas the TLD in the exhaust a fan intake plenum averaged 51 mrem per year. The conclusion that V concrete is a source of radiation is inescapable, but the quantitative contribution of this radiation source to arbitrarily placed TLDs is not readily estimated. The TLDs placed on lead bricks showed zero or slightly negative background values even tnough these locations were in the general downwind direction of the plume. The zero or negative background values are to be expected in that the lead bricks shield out the normal terrestrial component of the natural background radiation, and the reactor exhaust plume contributes no measurable increase in the background downwind from the stack. The average value of all other dosimeters (8 in number, 48 observations) in the roof top vicinity of the stack is 13.6 mrem per year.

The results of this second TLD program indicate that radiation from the plume is low, but that individual observations are probably sensitive to geometry, proximity of concrete, and shielding. A complete separation of the low level plume radiation from natural and man-enhanced (concrete) radiations does not appear to be feasible using TLDs.

l l

Ov II/A-10 4-30-82

APPENDIX !!!

f ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

U CONTENTS Chapter 1 Introduction and General Description. . . . . . . . III/1-1 1.1 C cription. . . . . . . . . . . . . . . . . . III/1-1 1.2 Opera tiors . . . . . . . . . . . . . . . . . . III/l-1 Chapter 2 Supporting Facilities . . . . . . . . . . . . . . . III/2-1 2.1 Subcritical Assembly . . . . . . . . . . . . . III/2-1 2.2 Neutron Generator. . . . . . . . . . . . . . . III/2-1 2.3 Activation Analysis Laboratory . . . . . . . . III/2-1 2.4 Heat Transfer Laboratory . . . . . . . . . . . III/2-1 2.5 To kama k La bo ra tory . . . . . . . . . . . . . . III/2-1 2.6 Auxiliary Laboratory Areas . . . . . . . . . . III/2-1 Chapter 3 General Plant Description . . . . . . . . . . . . . III/3-1 3.1 Si te Location. . . . . . . . . . . . . . . . . III/3-1 3.2 Si te Geol ogy . . . . . . . . . . . . . . . . . III/3-1 3.3 Site Hydrology . . . . . . . . . . . . . . . . III/3-1 3.4 Si te Sei smol ogy. . . . . . . . . . . . . . . . III/3-1 3.5 Site Meteorology . . . . . . . . . . . . . . . III/3-2 3.6 Population Distribution. . . . . . . . . . . . III/3-3 O Chapter 4 Plant Arrangement . . ........... . . . . III/4-1 Chapter 5 Nucl ea r Reactor . . . . . . . . . . . . . . . . . . III/5-1 5.1 Reactor Core . . . . . . . . . . . . . . . . . III/5-5 5.2 Reactor Cooling System . . . . . . . . . . . . III/5-8 5.3 Reactor Instrumentation and Control. . . . . . III/5-12 5.4 Reactor Fuel Handling. . . . . . . . . . . . . III/5-15 5.5 Reactor Waste Disposal Control . . . . . . . . III/5-15 Chapter 6 Compa r i s o n Ta bl e s . . . . . . . . . . . . . . . . . I I I / 6-1 6.1 Comparisons with Similar Facility Designs. . . III/6-1 6.2 Comparisons o' Final and Preliminary Design. . III/6-1 Chapter 7 Indentification of Agents and Contractors . . . . . III/7-1 Chapter 8 Credible Accidents for Argonaut Reactors .... . III/8-1 Chapter 9 Requirements for Further Terhnical Information. . . III/9-1 Chapter 10 Referenc es . . . . . . . . . . . . . . . . . . . . . III/10-1 Attachment A Analysis of Credible Accidents for Argonaut Reactors. . . . . . . . . . . . . . . . . . . . . III/A Attachment B - Fuel Temperatures in an Argonaut Reactor Core Following a Hypothetical Design Basis Accident (DBA) . . . . . . . . . .. .. .. . . . . . III/B III/i 4/30/82

APPENDIX III ARGONAUT SAFETY ANALYSIS REPORT (ASAR)

LIST OF FIGURES PAGE Figure III/1-1 Area Map - West Los Angeles . . . . . . . . 111/1-2 Figure III/4-1 Aerial View of Central Campus . . . . . . . . III/4-2 Figure 111/4-2 UCLA Campus Map . . . . . . . . . . . . . . 111/4-3 Figure 111/4-3 Nuclear Energy Laboratory - 1st Floor . . . . III/4-4 Figure 111/4-4 Nuclear Energy Laboratory - 2nd Floor . . . . III-4-5 Figure 111/4-5 Reactor Building - Elevation Sections . . . . III/4-6 Figure III/5-1 Reactor - Longitudinal Section. . . . . . . . III/5-2 Figure III/5-2 Reactor - Transverse Section Through Core Center . . . . . . . . . . . . . . . . . III/5-3 Figure 111/5-3 Reactor - Horizontal Section at Beam Tube Level. . . . . . . . . . . . . . . . . 111/5-4 Figure 111/5-4 Fuel Plate. . . . . . . . . . . . . . . . . . III/5-6 Figure III/5-5 Typical Fuel Cluster. . . . . . . . . . . . . III/5-7 Figure III/5-6 Fuel Boxes and Coolant Connections. . . . . . III/5-9 Figure III/5-7 Cooling Systems Piping Diagram. . . . . . . . III/5-11 LIST OF TABLE _S_

Table III/1-1 Chronology . . . . . . . . . . . . . . . . . III/1-4 Table III/1-2 Reactor Annual Use. . . . . . . . . . . . . . III/1-5 O Table III/1-3 Class Use of the UCLA Reactor . . . . . . . . III/1-6 Research Usage of the Reactor . . . . . . . . III/1-7 Table 111/1-4 Table 111/3-1 Habitable Area & Population Within R Miles of the Reactor Site . . . . . . . . . . . . . III/3-4 Table 111/5-1 General Cooling System Characteristics. . . . III/5-10 Table III/6-1(a) Comparison Table - General . . . . . . . . . . III/6-2 Table III/6-1(b) Primary Coolant. Nuclear Data . . . . . . . . III/6-3 Table III/6-1(c) Reactor Characteristics . . . . . . . . . . . III/6-4 Table III/6-2 Training Reactor Characteristics. . . . . . . III/6-5 Table III/7-1 Identification of Agents. . . . . . . . . . . III/7-2 Table 111/7-2 Organizational Relations. . . . . . . . . . . III/7-3 Table III/8-1 Inventory of One Fuel Plate Containing 0.57%

of the Core Inventory . . . . . . . . . . . . III/8-10 Table 111/8-2 Releases, Concentrations, and Dose in Reactor Room for Release of 2.7% of the Gaseous Fission Products in One Fuel Plate. . . . . . III/8-12 O

III/ii 4-30-82

ARGONAUT S_AFETY ANALYSIS REPORT (ASAR) c

b. 1.0 IJNTRODUCTION AND GENERAL DESCRIPTION 1.1 Description The Argonaut Safety Analysis Report has been prepared for submission to the U.S. Nuclear Regulatory Commission in support of reapplication for an Operating License. The application is made by the Regents of the University of California for the continued operation of the reactor, licensed as R-71, at the Los Angeles Campus.

The plant housing the reactor is located in the northwest wing of Boelter Hall at the University of California, Los Angeles. The 400-acre campus is located on a coastal plain and is approximately five miles east of the Pacific Ocean and fifteen miles west of the Los Angeles civic center. To the south of the campus is a business and shopping district, and to the north, west and east are residential areas. A map of the general area is shown in Figure III/1-1.

The reactor is located at the Nuclear Energy Laboratory in a 2 story, reinforced concrete structure with a floor area of approximately 75 x 90 ft. and a height of 27 feet. The construction of the reactor facility began in 1959, with the assistance of a $203,350 grant from the U.S. Atomic Energy Commission, through the efforts of the founding Director, Dr. Thomas E. Hicks. This grant was disbursed in construction and reactor equipment.

Subsequent construction, completed in 1968, has surrounded the reactor room on the north, east and south sides with additional laboritory space that provides a buffer zone between the reactor room ani adjacent but unrelated facilities. On the west side, first floor labo atory spaces and the second floor control room intervene between tha reactor room and the exterior building wall.

The third floor (roof) of the reactor building is bridged by new con-struction at the fif th, sixth, seventh, and eighth (roof) levels.

The void region between the third and fifth floors is a limited access region which contains a small structure housing air conditioning and water demineralization equipment.

The nuclear reactor is an Argonaut type; water-cooled and moderated, graphite reflected, 93% enriched uranium thermal reactor, that is current',y licensed for a maximum core thermal power of 100 kw. By special amendment, the reactor has operated in the past for brief periods of up to 500 kw. It appears that the reactor could safely operate up to 1,000 kw with modifications to the shielding, the cooling system, and special provisions for reducing argon-41 emission.

1.2 Operations Historically, the UCLA reactor reached criticality on October 21, 1960, o at 6:54 p.m. After a program of low-power testing at 10 watts, the

\J reactor went to its then licensed power of 10 kw in February of 1961.

The reactor was modified slightly, license amendments were approved, III/1-1 4/30/82

and in October of 1963, the reactor reached its present licensed full

/l thermal power output of 100 kw. The chronology of these and other U events is shown in Table III/1-1.

The reactor generates no electricity and is used primarily for activa-tion analysis, class instruction, student experiments and faculty, staff, and student research. To provide this flexibility, the reactor has three vertical irradiation holes (1.9" ID), a 78 cubic foot removable graphite thermal column with a one cubic foot irradiation volume, two 6" ID and four 4" ID horizontal beam ports, and a 3,000 g'allon water-filled irradiation volume. A pneumatic transfer system (" rabbit ) provides sample irradiation in the west vertical port with rapid transfer to a counting laboratory.

The variety of irradiation ports has provided great flexibility in the kinds of experiments that can be conducted with the reactor. The fast and thermal flux is maximized in the vertical ports, the thermal to fast flux ratio is maximized in the thermal column and a neutron and/or y beam may be extracted from the horizontal beam ports. Table III/1-2 gives a brief description of the annual reactor use fron 1973 to 1981 Variations from year to year are attributed to research de-mand, changes in technology, random maintenance requirements, class scheduling, and enrollments.

Class instruction includes the instruction of undergraduate and graduate students of the UCLA School of Engineering and Applied Sciences and other departments in basic nuclear engineering theory and applications.

Class instruction also includes general health physics and reactor operator training. Table III/1-3 lists the current class offerings which require use of the research reactor and the total annual student hours of reactor dependent instruction projected for the 1981/82 acade-mic year.

When not being used for class instruction, the reactor is made avail-able to assist both academic and non-academic users in activation analysis, delayed-neutron counting, fission track dating projects, and other experimental techniques. All such non-instructional uses of the reactor have been categorized as research. A number of the academic users of the facility are from other colleges and universities in the area. Recently, a non-academic user of the reactor has been employing activation analyses techniques in his ore-assaying business.

All research users of the facility are charged a fee for the reactor services provided. The fee is based on " port-hours" of reactor operation .

Although up to four (4) experimental ports may be used during one hour of actual reactor operation, such use is rare because of demand and incompatibility of desired irradiation conditions. The port-hours of use by each category of research user during the past ten calendar years is given in Table 111/1-4.

III/1-3 4/30/82

O Table III/1-2 REACTOR ANNUAL USE Year Number of Runs Megawatt-Hours Actual Operating Hours 1973 76 13.8 1974 76 14.8 1975 91 11.9 1976 82 13.1 184 1977 106 15.9 238 1978 132 20.3 271 1979 149 29.0 372 1980 131 28.9 381 1981 134 23.9 364 9

O III/1-5 4/30/82

Table III/1-3

( <

UCLA NUCLEAR ENERGY LABORATORY Table of Class Use of UCLA Reactor 1981 - 1982 Academic Year O a > E =

0 en E u Se *: " ha5 5

'5 f* 87" E;" E" 9" yt I

+ *E *E OER 85E E2R #E EE 5" E :

CLA55 -s 5e b5 e.

55

=ns T!s) entwnt- sytroR unIAHul Elserfs CDNT4]LS, E11CTMRICS topfwC.L*GNTIO.S (RIGlpmL h IEstGO Foo umt h (1963) *(1966) *(1957) 0FEATIE PDet opmL P0e.1%'rt, 28.1h/E U255 . .

LLINEAR, SPECIFIC PD O)

OPUtAT!E v21Amt,13-15 acAsaTT eets ve!Antz,15 To 3 es rets rp YtAR vm!Anz Soeur roi wAn PRINCIPAL ACTlvAT!@ AfdLY115, ilEACTOR TRAIN!E Af0 EDIATim Or MKlIAA IEACTIR OPDATOR TRAIN!E, LAE OF OPEAATDR TRA!NIE Af0 EDIAtl0M EwA. Af0 SCIENTISTS, flE.ACTnR Cuss D006TRATICB6, IEACT3t OFMQ1ARDrit,k$Tt10ff5 OPDjATCRS, ACTIYATIGe ApenLYSTS & LA50t ISOTDPE PMx1ETION,ledA EP $. RESEARCH OPD ATING 3 Fat siirt,1 SMirt 4 emi init 2 put shirt, I s<lri uwettY 2PART 2 ret siirT, I siirt ITJJF STJENT Of9Aftsts flPE OPGATORS O STATUS:

DATE FIRST G ITICAL M 1S 4/61 2411IM3 V59, M2 %1, 5'E7 E lb SOLID Full:

ELDeff *

• sdPE FtAT PLATE (MTR TYK) 3 Ful ' '

CXpM151 TION 13.4 itT.! u-a AtloY CJ Ful hY D CLAIElE qTtRIA.

74* x 2" x .0 c 113) ALmimm 23.5"x2.3"x.0F El (1A:EIE *

• TylCuK.S$ .25 IN.

FJ TYPE OF - -

r.Auo. ,uits .uw -Do gY o,cos ,o m - -

11 D{a0610E 5 5/8 x N /B x . R IN.

II SB-ggy E RI,c) x Di,ecio,c J) f 0F *

• no's 2 14 *

• x) Al%Nrl- f4 E19(.NT5 Ig EA04 & 6 BOES PtNT U $$- ARWs6 IN 4 PARAt1EL IKM A$$EMLY *

• U LIFETI4 IMI. FINITE

• PtWE & MANA., USIE HAf0 TOOLS, TRAf6FER IEM. LIE CAS4 IUCTE MifMR

O 111/6-2 vwe l

l

TABLE lil/6-1(b) PRIMARY COOLANT l

nuiD y tject1ATim A) DIRECTION

• uwoS w)

B ruw ruu -G ruD, .A. Rt.

i

, 'puttD Bv 23 m 40 e y 16 m

" '

• 4cWstc 2.1 ovste l 1.3 1sec AVDUGE CDRt i

• e(AT MATER-T&iehitR TtBE-S(11 S.S. *1 T.EE Ig4RY, $1M1.1 FASS D1551PATim SecDoRY-u Pti RATG) c oo HAT Octuen )IDLaLI-PASS sumoRv(nat

• *(93) PL/ MIN)

MANS DF PR19Av SYSTEM: CDNTIMDUS % W P@lFICATION BYPASS TW CARTRIDC.E FILTER TMJs T% DEN 14.AAL120t$ APO RETLMt To DM TAfE SHIELD T .! ItOEFOHNT, CQ>r-T!4DJS, G% FILTER =O 2 PARAU.fl. DENIM.RALIZG CARTRIDCL' NUCLEAR DATA F% uh4Dl%

AJ CILD

~"''""'"

74 ant C) " . 3,3 m u235

[$,c"ocess x, 3.6 m U235

• k u 2.3%as, y*a 10h 14 x 1012do8 sec "ew.

T rua 1.5 x INo8 sec

• 5.0 x 10NcN2 3ge I TYuu 1.5 x 10408 sce cV Prx N Fua 1.0 x 104o8 sec 3.0 x 10408 sec ReactivtTv

),

E"TtAS".

A) Ttw -0.48 x 10-2 1 avg (-0.74u F) -1.0x10~4 aduMC -05F

+ sg/adF

14. (cRAPw Tt)

B) WID dN WID -0.3 adu/1 y WiD . av votD C) 8%SS U ICitNTS

)

• 235 0.311 add U-235 0.011a v v m U D) cDRE

• octss 2.3% avn $3.%

iLahai *

• Polwe m eorem sutct 6.6 di M or To 5 ci fa-BE =01 cu M Se-Bt (2 Ci M ir w.crD)

O 111/6-3 vsyc

TABLE lil/6-1(c) REACTOR CHARACTERISTICS O F- . 6"gD I

'- 2" "r$' -

mm %, . .m B I'RDGlCAL 941 ELD CD M TE 00ppETE CDOETE Fial SCR M $) 500 FMat B) C)Mwant A) MmAL BAR A) pogleFm1PE 3) pwyn . e (W u & tFlat SqRAra LDest C sed CJ 5 l@

• 3 SEC NATER) Pl#13) (LISS FLUX (fodUt

• u'ya T(tP9

) SqRTHIG4 5AFETY OtadM 2)s D tot RE- D 134 Fun]T its E) OPEN E) Di0.D M LS SLT5 th LDe FT, LIVEL, A(p 10 Fl.D8 DLW VALVE OftN AdsNE 1 mTT F D M VALVE $0 TAM. INTULOLK1 WITH tt PLw. EY OrtN nmm wF =ltH TOR eueE suu.s tr

- Fut sow wlTH mTut am 50P G A) ptY nsefD OFF A) IF Opu I BUM LP

• Pw(s CLUTOES A) TtstN OFF B) LDt FLDd SCMM HIG4 BAY VENTILATION B)C)LDSS ID$$ OF OFOFF 8) PRIPt4RY C LDi CDRE LEWL m: MARY Ptn pop spe C)LCss tDssWorDitMTIE et OR vtNT FANImARY PmO IDss PRImnPY FtDe Mtl FLDsD) E )(Dss tDe CDfEor DJ GPM FIDtIO LDf SHIELD TArm mTD LIVEL teATUt LIVfL F LDf SHIELD TAfet PC)FLDt E PC tenTD DPDPS 10 llVTL N0P5 Agu F 6 RIT70N G) DILUTION tenTUt tral E HIG4 sFIDcNrY loots (STitt AsmE FLEL PLATES FAN 1R (sDHNJ5TJ U # lAT!GI F 9941 DOP = 00 VARI LIVIL 981704

(> x .cl/4 1- NM 1 6 O saw. *EDLY $4m1 ANALYZED H) DOP G SUTTON 1 LDGTC CDOITIGd 1861191T5 A) es1/T10g START SOLJtC1 CDLNT A) $Aff AS SutAM BAR B) 0F D A) PUti lit (T *sTrg 10 B) y LISS IHAM 4 8 PUtlCD Lf35 teLTD AgtM IE. PC TDv> Gt C) *m b SEC HIGO C) 10E DOP IN elm 0N CPS /SEC 1 W) C )(DG-es C1DStatEAr9tlFIUtOPDe01AP9Dt HIC (4 %g.TEL, E.!Tte w 2 W ICT IN OPUIATE #GE RMtlE5 1804181T OF LP Ut tvt Y DJ)ZCPS E LDGIC - CANOT WITH-OR FORE RADES $14LTAfGDUSLY F SEg PUtl3) UP IVE G IO Hi w!DE RAfE N+

STAREP(s OW9EL ADeLYS IN OPUIAfl0N.

1964181T IF CALINtATE Gt TRIP 9e ICT IN OF9 ATE PG)E FOR LP A) HIG4 PR Dif TD+- 13ff " ALE!RI ALAft1 B) 0F IGH MttPtWrY IT SET 194 & PDAftstE - HICH AfD 4) a AREA puelTuts = ALAfM LIGC S B HIG4 60NN) RADIATION *5 soRTH & tot - AmiRI Auwe L.uGn a RAotATI RDcTUR SOLITH HIG4 DAY

• RADI> 66t . 140 CHAf0413 a IJ 801/tSt ALW) = RABIT p ETIVE ST3 W E

• pAQATigM ALAfM SIRD6. LDss OF 9 tedsFLIX)R

• Dlf a * . HR ltEACTQR TOP pft HIGH WPIIN N1 RAaBIT IU M CL AMDe IC IN 10 FAIL uGC IF OWetL PGIAI ICT STEK

• 1.8 X 1TMthL IN AIR REGIVIE $1Gg4L FIOM AREA PUNITGt C)5Afff SOIM Gt IQ) NOP OtfP90t C) 4 075 mTUt IN PC PIT SM = A@lRI ALAfM IN CDMTML SEDt LIGG D) ALAfm LIVIL, STACK EFFupT ARAJSTAILE MIDHDIE 10 PDER OF W UtAfl0M Am.NM A) LDsITI S- A) LDS$ OF POO OR OFee CIRC 2JIT IN A)# 89timARY IvlTY LIG4T (PLY TivlTV

• 1 x - - B) E' * %U,dD 999 B) 14TIClUt th HIG4 FLLBt > 0. mTTs BIAQJAfloh $1RDe *n/N PC uno selw4,0NITORs LDssBu1TGF W4E.De LIGH Fum lairs>TmTTs C) MHIG4 CJ AffY lle11Bli Af0 OfflIT IESISTivlTY As.A(P9 LIG4T MLAETAALE ON !sETitMNT Ci Al'.

CDICITlose TRIP NY $1R04 " IED LIGH D) B M0RNT10lgt CDLJi'Ut 400#

DOGlZH) H45 IG EXTDt G LIG4. AUT3% TIC W Qlf 0FF S CPS E) FAST PEgl3) 1064181T LIG4T &

Ss! TOE 5 Fi FAST 500 II64tBIT LIGHT 196418 17

,&or 9elTOES BOT IN OPUIATE

'" 9 sCmM No"i'{J"vl AFTU",t A.T .OR ABINEim uGe '?"tDss 0 SECDO uGas -

FAlustES IN RfD 90M $TAf48Y BATTERY PDO PACK CJ ALOlBLE Am.NM & RED LIG4T IF DACK IDOR 11 OPDO. RED Ap0 GtEDs LIGR5 PUtlTDR OLITUt acres $

DOORS 10 FACILITY 111/6-4 m

O

TABLE lil/6-2 TRAINING REACTOR CHARACTERISTICS wt no Q in ma anmaae. umt 10 m 120 no rum uwt (AT 10 m) 1 x 10llvcM2stc 1.5 x 1012 g ,2 ,c exass atAcriv Tv (no4 secc unti) 0.58: e AT ISF 2.3% o AT m uw cxctss EAcTiv Tv tusT=un 1.S! e At nom nw 1.E e AT nom nw ofm cad cs:TscAL res 32G) m tF25

• EntcTavt mwi eturn urtTis 2.4 x N 2 x 10%

ts.trom utta w D cuiracitxT 4.12 en w D -0.164I o wD itstuits carricitar -0.48 x 1(T2 I EMF .481 x IfTgZ e LL25 russ carracitxi +0.311 en M euss +.3: e/: lFE sTAnT-tr ma 2 cwit Pu BE 6.6 mC: RA Bt gracims

• I cruwtit (IR aWcc) naufm 30mesww TE

• 2 truYtn wuta FucTim 0.QL8 .QE5 FEL PLATts FEL E DF OG, lb4L AMDY

• Fat tocirc 3,145.2 m lL235 3512mLF235 etAn TwicMss 0.0 D IN.

• wita owe.tt 0.137IN.

• Autuu to wita mTio (wL) 0.51

• etat anos: Tim 13.4 wrf (FAL

• CCXUNT

• N/)

Ftm IDcm 10 mm it m mr m IN 105 F IQff umurm ouT 11!PF 142F U ca<rnaL axrs CD, SW EIPc VM, GRAVITY FALL

• O ama 3 surTv 1 =auf w.

• V av.Enfim Tart 0.I*i ste (c4LauttD) 0.5sec(stAsse) meal Tim R) ste (rumlc0 10) stC axt eTH, sutTY 3 mos 13 e - 4.5% e 3 noos s 1.Q o = 4E axt eTH, sauTim 1 mm 0.c e = 0.C e I m o s II o tot =. 5.1% o TOTALS 5.2:o EAcTiv:TY No:T Ctd RATt, Pux. 0 5 : e/ste .fC: '/ste Shit 1D (C0KMTE) i s Es, cam n C rT. 0 IN. cast, nocTin

• s:Es, sHitLD Th Do {FT.8N. CAST, ORD MnY

• I s 25, DEMmL CDuM. Do 6 FT. 8 IN cast, MecTITt

• 5 PJ:I11 cast CONCETt SLDCEs

• Aax cmc 5 n. 10 N. n o cTi n nLocas

• rus 39" or acuTtD ramAr:N tus 3 n. 4 IN. meetin nGtxs

• OftR 8thTAL FAc LITits l DtML COM)h, H[RIMIAL 5 n. x 5 rT. x 4 n.11 IN. tac CDIN.x52N. x 43 N. Lac amovAna

( TEMmL CIUph, WRT: CAL P40VistG4 FOR NSTALLAT 0N

• l 2 tG nst Tm 5 n. x 5 n x 14 FT. 6 IN. HIGi

• EmRieTAL eoLts 2 - tonimo 6 IN. DiMTre

• 5-HaRam b 4 N. D W TER 4 - PCRIZmTO 4 N.D E nR 3 crico h IN. DIMTER 3

• UT CL 1-7/8 IN. DiMTER D@tRittNTAL M11s, DERM. GL. 15 seovAsi starHITE sTR MERs

• Fo:L E0il 11*POR2DNTO 1/8 IN. x I/2 N.

16 - erico .V8 :N.x1IN.

• O 111/6-5 i

i vne f*

8.0 CREDIBLE ACCIDENTS FOR ARGONAUT REACTORS

8.1 INTRODUCTION

The original accident analysis for the UCLA Argonaut-Type reactor, which was completed in 1960, postulated an accident involving local melting of the fuel core and assumed that a release of 10% of the volatile fission products into and away from the reactor building would result [1]. The authors of that work provided no basis for the postulate, but noting the inherent self-limiting characteristics of the reactor they did state that they regarded any core melting ae not plausible. The UCLA reactor technical staff has always considered accidental core-melting to be implausible. With the recent release of certain generic analyses by the NRC, the implausibility of a core-melting accident has been confinned for all Argonaut-type research reactors.

The postulatory basis for the core melt scenario discussed in the original UCLA accident analysis was probably related to the concept of a " Maximum Credible Accident" for power reactors. In power reactors, core-melting damage can be causally related to inadequate heat removal following a loss of coolant accident. In Argonaut-Type research reactors, decay heat power density is far less, and loss of coolant is a designed back-up shutdown or " scram" system - a safety feature rather than a hazard, f}

The NRC's generic analyses, which are discussed below, have served to identify the self-limiting characteristics of Argonaut reactors.

These generic studies, which in all cases were based upon very conservative assumptions and analyses, demonstrate that accidental core melting of a 100 kw Argonaut reactor (such as the UCLA research and teaching reactor) is a non-credible event. Thus, for an Argonaut reactor, there is no equivalent to the " Maximum Credible Accident" of power reactors and because that phrase carried the connotation of core melting it is not used further in this analysis.

The general conclusion of the generic studies is that credible accidents hypothesized for Argonaut reactors predict relatively minor radiological consequences. This conclusion is fully applicable to the specific case of the UCLA research reactor. Among the accidents examined in the generic studies, a fuel-handling accident was determined to be the worst credible accident and has therefore been adopted by UCLA as the design basis for emergency response planning, and is discussed in detail below.

8.2 GENERIC STUDIES Two generic accident analyses of Argonaut reactors have been released:

one by Battelle Pacific Northwest Laboratory [8]; and one by Los Alamos National Laboratory [9]. These two studies have been incorporated herein by reference as Attachments "A" and "B", respectively, to this Appendix. Drawing on these studies, the NRC has produced O its Safety Evaluation Review (SER) of the UCLA b -

III/8-1 4-30-82

O 8.3 GENERIC STUDIES EXTENDED FOR THE CASE OF THE UCLA REACTOR The Argonaut generic studies and other matters particular to the UCLA facility have been reviewed by the NRC. In its SER for the facility, the NRC noted that extremely conservative assumptions and analyses were used in the Battelle and Los Alamos studies. As a result unduly conservative estimates were made of the predicted consequences of either a fuel-handling or earthquake-core-crushing accident. Therefore, the SER treated the fission product release that might result from a fuel-handling accident as calculated in the Battelle study as equivalent to the fission product release that might result from a severly damaged fuel core caused by a building collapse during a major earthquake.

The conservative analyses of the Battelle and Los Alamos studies served to strongly support the NRC's general conclusion that no significant radiation hazards to individuals in either restricted or unrestricted areas would result from accidents at the UCLA reactor.

However, in examining the consequences of credible accidents at the UCLA facility for the purpose of planning emergency responses, it is necessary to extend the generic studies to take into account certain site-specific factors. In the discussion which follows, it will be useful to distinguish radiological accidents that might occur as a secondary result of some naturally occurring event, such as an O earthquake, and accidents that might occur during the ordinary course of reactor or facility operations. It may be assumed, with respect to accidents that occur as a secondary result of some natural event, that any additional hazard that might be hypothesized due to the existence of the reactor would be inconsequential relative to the general disaster caused by such an event.

8.3.1 CATASTROPHIC SEISMIC EVENT Due to the fact that the predicted consequences of a crushed reactor core are relatively insignificant [8,9], a detailed seismic analysis of the UCLA facility is not warranted. The following remarks are only intended to suggest certain of the factors that would be relevant in the prediction of the radiological consequences of a seismic event at the UCLA facility.

The known geological faults closest to the UCLA campus are the Newport-Inglewood fault to the east and the Santa Monica-Hollywood fault to the South [12,13]. Since the Newport-Inglewood fault is estimated to be capable of generating an earthquake of magnitude 7 to 7.5 (Richter) with a recurrence period of 1000 years, it is regarded as potentially more dangerous than the Santa Monica-Hollywood fault, which has an estimated potential of generating a magnitude 6 (Richter) earthquake with a recurrence period of 10,000 years.

Although the effects of a major seismic event on the reinforced concrete buildings surrounding the reactor are uncertain, it will be

/]

I11/8-3 4-30-82

assumed for this discussion that such an event is capable of pb collapsing one or more of those buildings. Furthermore, both the Los Alamos study and the Safety Evaluation Report assumed that a collapse of the reactor building could result in a collapse of the reactor biological shield and the reactor core. Two effects of the hypothesized reactor collapse resulting in a crushed core have been investigated.

In the Los Alamos investigation it was assumed that up to the immediate moment of the reactor's collapse the reactor had been operating continuously at full power (100 kw) for a sufficiently long period of time (months to years) to reach a near-equilibrium fission product inventory. Los Alamos examined whether reduced convective heat transfer in a collapsed configuration of the core could lead to core melting by decay heat accumulation. The Los Alamos study concluded that in such circumstances the core could not melt and fission product release by that mode was not possible. It should be noted that the reactor has never been operated under conditions that would result in attaining full power fission product equilibrium. The UCLA reactor cperates at an annual average power level of less than 5 kw; the long term historical average is approximately 3 kw.

Based on the Battelle study which considered a core crushing event, the SER assumes that seismically-induced core damage could sever fuel plates and release fission products to the environment. In the SER it is calculated that atmospheric dispersion of the radioiodines in p the crushed core situation could yield an estimated thyroid dose of d 30 rem to individuals at the boundary of the demolished reactor room.

When considering the credibility of any core crushing scenario, it should be recognized that the reactor is a dense concrete and graphite structure. The thick, short spans of reinforced concrete blocks above the reactor have enormous compressive strength relative to any conventional building structure. It is by no means certain that the reactor core would be crushed in the event of the collapse of the reactor building.

During periods of major core maintenance, the core may be exposed and more vulnerable to a major seismic event. Core maintenance at the UCLA facility occurs no more frequently than once in five years. In order to minimize radiation exposure of personnel, core maintenance is not begun until three weeks after the last shutdown. At that time the core is ex30 sed and the fuel unloaded in a single day, any required maint? nance is performed, and subsequently the fuel is reloaded and t H core covered in a single day. The fuel is not in the reactor while @intenance is in progress. The period for which the l

reactor is bott exposed and at least partially loaded is no more than l

16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> during any five-year period. Without speculating on the l

probability of an open, partially loaded core and the simultaneous occurrence of a seismic event severe enough to collapse the reactor building and crush the core, it can be pointed out that the.

radiological releases postulated for such a case would not be (O

l 111/8-4 4-30-82 l

O quantitatively different from those postulated for the closed core O cases.

8.3.2 CATASTROPHIC SEISMIC EVENT WITH FLOODING The Battelle study considered the possibility of a criticality-type accident in the event there occurred a catastrophic rearrangement of the core with subsequent flooding of the reactor. Battelle assumed that a shock sufficient to produce the precise structural rearrangements of the core needed for a criticality accident would also lead to a loss of the existing reactor water due to the severing of the water lines. In such a case the water would have to be replenished from some source to restore the moderator necessary for such an accident.

It is conceivable that subsequent flooding of the reactor room could occur as the result of earthquake-induced failure of the Stone Canyon Reservoir which is positioned in the hills to the north of the UCLA campus [14,15]. If the dam were to fail, a portion of the UCLA campus would be flooded. The flood resulting from instantaneous dam failure is hypothesized to be of a magnitude capable of destroying a substantial part of west Los Angeles [14].

In the absence of core crushing, flooding alone will not produce fission product releases. Various scenarios were considered in the Battelle study which assumed a critical reactor, structural

(' rearrangement of the core or stuck control blades, and loss of water with subsequent replenishment of the water-moderator by flooding.

Battelle found that structural rearrangements of the core into some more optimal geometry of reduced minimum critical mass and large excess reactivity was not credible and, it may be added here, appears to imply some interpenetration of graphite and fuel while ignoring the intervention of the cadmium control blades. Moreover, the reactor is considered to be near optimally moderated in the sense that additional moderation drives the reactor less critical.

It has long been known that wetting of the graphite results in a loss of excess reactivity, an effect which could alternatively be described as loss of reflector efficiency. Flooding beyond the optimum moderation level would be expected to lower the system reactivity.

Accepting the assumptions of the SER for the case of core crushing, subsequent flooding of the reactor could result in the dispersion of fission product releases in the flood water, which would be expected to disgerge to the Pacific Ocean southwest of the UCLA campus.

8.3.3 GRAPHITE FIRE The Battelle study considered a general building fire as an initiator of an accident, but discounted the credibility of the cause except for the case of a fire fueled by reactor materials or other combustibles which might be used in the reactor room. The initiation of a reactor accident by a fire external to the reactor III/8-5 4-30-82

O room is not credible in that the reactor building and the surrounding V complex are constructed of reinforced concrete.

The principal combustible material routinely present in the reactor room is graphite. Ordinarily the graphite is contained within the reactor. During major core maintenance, which occurs only rarely, the reactor graphite may be stacked outside of the reactor. There is also a graphite sigma pile of approximately 64 cubic feet in the northeast corner of the reactor room. Small amounts of other combustibles, such as wood, cloth and paper products, are often present in the reactor room. During major core maintenance organic solvents may be brought in for decontamination purposes. Borated paraffin blocks, which are not readily combustible, are used as additional shielding.

Battelle noted that the plausibility of a graphite fire within a reactor core or thermal column enclosed in concrete shielding is limited by the available oxygen supply. However, Battelle assumed an air flow rate through an Argonaut reactor of 250 cubic feet per minute. The air flow rate through the UCLA reactor is actually less than 100 cubic feet per hour, approximately 0.7% of the rate assumed by Battelle. Under the UCLA conditions, it is much more likely that any graphite fire that managed to get started would suffocate due to lack of air and the buildup of combustion products. Battelle also discussed the possibility of a graphite fire occurring when the core is exposed. However, when maintanance is performed on the graphite or other elements of the core, the fuel is not in the core. Any O'- scenario involving an open core, fully or partially fueled, unattended, with the graphite exposed and in contact with a substance capable of causing graphite ignition is not credible.

8.3.4 REACTIVITY INSERTION ACCIDENTS It may be assumed that the investigations of Battelle and Brookhaven were designed to set a conservatively safe limit upon the excess reactivity to be permitted in an Argonaut reactor. The Battelle investigation concluded that melting and consequent fission product release would not occur with the rapid addition of excess reactivity in the amount of 2.6%. Their choice of a prompt neutron life time of 1.4 x 10- seconds adds an element of conservatism to the calculation because this parameter is at least 1.9 x 10-4 seconds for i Argonaut reactors.

Brookhaven examined the ramp insertion of $3.00 of excess reactivity defined with S = 0.00714 (effective delayed neutron fraction). UCLA uses 8 = 0.0065 and the same excess reactivity would be termed $3.30.

By either definition, the excess reactivity is approximately 2.14%.

The Brookhaven effort was aimed at examining the safety of this amount i of reactivity. The study confirmed that the ramp insertion of 2.14%

l excess reactivity is safe, but no conclusions were drawn concerning the maximum safe upper limit.

l fj Neither Battelle nor Brookhaven addressed the question of how such i

III/8-6 4-30-82

a large reactivity insertion could occur. Brookhaven did suggest that G one or more large cadmium sleeves having a total negative reactivity on the order of $3.00 inserted in a vertical port might fall out of the reactor. A negative reactivity of this worth is conceivable. But the Brookhaven study does not suggest why such an object would be introduced into the reactor, nor how once introduced, it could be made to deviate from the normal gravitational forces, and fall "up" and out of the reactor. As postulated, the event is not credible.

8.4 FUEL-HANDLING ACCIDENT UCLA has adopted the fuel-handling accident proposed in the Battelle study as the most credible accident and has used this event for the purpose of planning emergency responses. However, in extending that generic study to the UCLA circumstances, three modifications are in order. First, although Battelle assumed 365 days of continuous operation at 100 kw, the UCLA reactor operates an average of less than 5% of that time. Continuous operation is typically no more than four hours at a time and only very rarely more than eight hours. To the knowledge of the current staff, the reactor has never operated continuously for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; the last occasion of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> continuous operation occurred in 1974. Second, Battelle assumed that the fuel handling accident occurred immediately upon shutdown at the end of the 365 days. However, a holding period of three weeks is observed at UCLA to reduce the potential of radiation exposures to the staff. Third, although the NRC review assumed that the fuel area g exposed in the core collapse event would be equivalent to one entire fuel bundle, ir. a fuel-handling accident the denuding of a single fuel plate is considered to be the worst possible credible consequence.

8.4.1 CORE INVENTORY The twenty years of UCLA reactor operations have generated a cumulative energy of 19.4 megawatt days, approximately half of the total energy assumed to be generated in one year in the generic study. Most of the historically generated energy was produced in the last seventeen years and the long term average energy generation is approximately 27.4 megawatt hours per year or an average power level of about 3 kw. The most intensive years of operation were in the middle 1960's, and the most intensive three-month interval identified in that era was the fourth calendar quarter of 1966. The energy generated in that quarter was 17.5 megawatt hours for an average power level of 8 kw.

The statistical history is relevant because long lived isotopes such as Krypton-85 accumulate slowly over a long period of time. Isotopes of intermediate life (Iodine-131 and Xenon-133) are present in quantities reflecting the prior several months of operation. The inventory of shorter lived isotopes depends upon the most recent operational history of the reactor. However, these generalizations must include consideration of precursor decay, particularly for any short lived g gaseous isotope that arises as a decay product of a longer lived III/8-7 4-30-82

I q reactor [10]. Subsequent to the release of the SER, Brookhaven V National Laboratory released a related analysis of Argonaut reactors [11].

Although differing substantially in scope and focus, these studies and reviews reached similar cnnclusions.

The Battelle study examined a broad spectrum of accident potentials including large reactivity insertions, core crushing, flooding, fire and fuel handling. The Battelle investigators concluded that the only credible accident that would result in significant radiological releases was a fuel-handling accident.

The Los Alamos study examined the properties of a crushed core and found that the altered configuration would not subject the core to melting by radioactive decay heat under the reduced convective cooling conditions within the crushed core. The Los Alamos investigator reported that even after long-term continuous operation of the Argonaut reactor at 100 kw, the maximum fuel temperature (following shutdown) in a core-crushing episode was calculated to be 358 C, well below the aluminum-uranium alloy melting point.

A transient analysis of the Argonaut reactor was conducted by the Brookhaven National Laboratory using computer modeling. The Brookhaven report concluded that a rapid ramp insertion of excess reactivity would not drive the peak core temperatures to the melting point, a conclusion in qualitative agreement with the Battelle finding.

A V All of the studies were concerned with accidents which might lead to radiological consequences. Specifically examined are the possibility that release of radioactive material due to core melting could be brought about by excess reactivity insertions (Battelle or Brookhaven) or by core crushing (Los Alamos). Each of the investigations determined that core melting was a non-credible event and that fission product release by this mode is not a credible consequence of an Argonaut reactor accident.

Battelle, Los Alamos, and the SER found that some fission product release could result in the case of a mechanically damaged and crushed core. Battelle noted that flooding of the core during or shortly after crushing would result in some release of fission products to the flood waters. It was generally assumed that the core crushing scenario could be produced by a major seismic event, although neither the probability of such an event nor the proposed mechanism of the crushing was examined by any of the investigators. Battelle discussed the possibility of a graphite fire and found that it would not create sufficient damage to melt any fuel or initiate a metal-water reaction [8 - Abstract].

Among the various accident potentials considered by these investigators, a fuel-handling accident was found to be the most credible accident that might result from ordinary facility operations, as distinguished from accidents which might be initiated by catastrophic natural events.

Accordingly, a fuel-handling accident has been adopted as the design O basis accident for reactor emergency planning.

O III/8-2 4-30-82

i precursor, w)

An extended holding time prior to core entry for fuel transfer is conventionally practiced because of the relatively modest shielding (6 inches of water) that remains after removal of the sixty inch concrete biological shield and a twelve inch lead and graphite plug.

A holding time of three weeks was observed in the 1974 core entry and led to acceptable personnel radiation dosages in the subsequent core entry and fuel-handling operations. The holding time can also be regarded as an accident control parameter and it is appropriate to demand a minimum holding (non-operating) period of three weeks prior to any fuel-handling operation (Technical Specification 3.6.3.4).

Therefore, the following operational schedule will be assumed.

a. Operation for two or more months at an average power level of 15 kw. (That average level is five times the historical long term level and approximately twice the highest intensity identified in any quarterly period.)

b. A final run of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 100 kw.

c. A holding period of three weeks prior to core entry for fuel-handling.

p V

Note that this operational schedule is considerably more intensive than is suggested by operating experience and any short period excesses can be limited by a restriction of operational intensity to less than 2.0 megawatt hours in any consecutive seven day interval (Technical Specification 3.8.3.C). This condition supplements and does not replace the existing limiting of 5% of the total potential of 8760 full power hours in any consecutive 365 day period.

8.4.2 LOSS OF CLADDING AND FISSION PRODUCT INVENTORY OF INTEREST The 264 rolled fuel plates used in the UCLA reactor consist of an aluminum cladding tightly bonded to the aluminum-uranium core " meat".

No examples of " peeling" could be identified by the University of Michigan where similar fiel had been handled frequently at much higher burnup than UCLA can ever expect to realize [21].

Aluminum is readily attacked by acids and alkalies. Such chemicals are not used in connection with reactor operations and are not stored or used for any other purpose in the reactor room. The presence in the reactor room of a tub or vat containing such chemicals, and of sufficient size to immerse a fuel element is not credible.

Aluminum metal is highly malleable and ductile, and hence deforms rather than shatters under impact. A loss of cladding accident would require abrasion, shearing, or tearing forces, and no specific event has been proposed to describe the mechanism of such damage. The III/8-8 4-30-82

p greatest conceivable area which could be exposed would result from the V complete removal of the cladding from the two exterior flat surfaces of a fuel element. The postulated exposed area is equivalent to that of a

, ingle fuel plate, and therefore the inventory of interest is that of a single plate. If we assume that the inventory of the most active element or plate is 50% greater than that of the average plate, then the fraction of the total inventory which will be present in the most active plate in the core is:

f = 1*6g p

= 5.7 x 10~' or 0.57%.

Using this fraction, the inventory of gaseous fission products of interest is shown in Table III/8-1.

Except for Krypton-85, the entries are those identified in [8], which remain in significant quantity af ter the 21 day holding period. The Krypton-85 is approximately 1/1.1 of the Battelle value and is close to the equilibrium value that would be~ approached in forty years of operation at an average power of 5 kw.

8.4.3 FISSION PRODUCT RELEASE In discussing the possible release of fission products in a fuel handling accident, Battelle assumed that cladding removal would release all gaseous fission products within one recoil length of the exposed surface. The fission fragment recoil length in aluminum is pJ approximately [19]

1.36 x 10'3 cm.

The prompt release of fission products from unclad fuel elements has been discussed theoretically by Olander [16]. Within the operating reactor with fission events in progress, the prompt release of a fission fragment can occur only if the fragment is formed within one recoil length of the surface of the fuel element. For specific fragments created at unifonnly distributed sites within one recoil length of the ' surface, Olander shows that only 1/4 of those fragments will be emitted in directions which will carry them to this surface.

The other fragments remain trapped in the fuel matrix. The prompt release terminates when the reactor is shut down.

For a fuel plate with a cladding thickness greater than one recoil length; the cladding can be expected to absorb and trap almost all of the fission fragments emitted from the fuel meat. The subsequent release of embedded fission products will be governed by diffusion rates in the solid matrix of fuel meat or cladding. Fission fragment diffusion in aluminum and aluminum-uranium alloys has been the subject of a number of investigations [17,18,20]. The rates are extremely slow at room temperature, and significant releases are observed only if the material is raised to a temperature of 400*C or higher.

v III/8-9 4-30-82

$t j ,

4 O

Table III/8-1 Inventory of One Fuel Plate -

Containing 0.57% of the Core Inventory, Curies Nuclides At Shutdown At 21 Days s .

Kr-85 0.09 0.09 .

Xe-133 5.98 0.62 .

I-1 31 2.83 0.49 .

I-132 5.81 0.07 -

I-133 20.20 -

g ""

+

g f a

!O '

s. n m - .

s , s e

i, i -

' ,s

~

1

,- m

' si

. ~

. ~

g e

~

s O '

S.r

~

111/8-10  ; .

J J' 4-30-82(

Y @

Inasmuch as the hypothetical fuel handling accident occurs long after b" ~

active fit.sioning has ceased, an assumed release of all of those gaseous fission products formed within one recoil length of the surface, yields a highly conservative overestimate of the expected release. The release fracticn is two recoil len ths (2.74 x 10~3cm) divided by the fuel matrix thickness (0.102 cm or 2.7%.

> The reactor room is not a sealed structure, hence the common practice of attempting confinement by shutting down the ventilation system is inappropriate. Table III/8-2 shows the release to the reactor room, the concentrations in the room and at the stack exit, and the personnel dose in the room under the assumption that the ventilation system continues to withdraw 9000 CFM from the reactor room and exhaust 14000 CFM at the stack exit. The entire' release and sweep-out is assumed to occur in one hour. The consequences are not sensitive to the rate of release, but do depend upon the amount of material released.

8.4.4 CONCENTRATION AND DOSE STANDARDS The thyroid uptake of radioiodines leads to a cumulative dose. The calculable dose for an exposure to concentration C for time T is proportional to the product C T and the same dose results if the co'ncentration is, coublad and the exposure time is cut in half.

The maximum permissible concentration (MPC) of iodine-131 for the general public is 1010 microcuries per milliliter [24]. Iodine-131 is s the. longest lived of the iodines considered here and the permissible (j concentration is for continuous exposure. 10 CFR 20-106a permits annual averaging and implicitly, an annual dose limit. An exposure of one hour per year to a concentration of 8760 times MPC will produce the same cumulative dose as continuous exposure to one MPC for one year.

Thus, the permissible concentration for an exposure of one hour occurring no more frequently than once per year, is 8760 x 10~10 =

0.876 x 10 6 microcuries per milliliter. The iodine-131 concentration in the stack effluent, resulting from a fuel handling accident (Table 111/8-2) is approximately 64% of the permissible one-hour, once per year release. Consequently, the exhaust stack plume cannot expose anyone to a thyroid dose greater than that which would result from continuous exposure for one year to the permissible concentration of 10 CFR 20, Appendix B,-Table II, column 1.

Tighter standards have been developed to define Emergency Action Levels (EAL's) for the purpose of emergency preparedness [22]. At the lowest EAL, the standard requires notification of the Commission if a release exceeds 10 times MPC when the concentration is averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Equivhlently, the Commission is to be notified if a release of one hour duration has a concentration exceeding 240 times MPC. Whether treated as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 times MPC or one hour at 240 times MPC, the thyroid

~

dose due to iodine-131 under this standard is the same and is less than 3% of the dose that-would accumulate in one year of continuous exposure to the maxirum permissible concentration of iodine-131.

()

f~^

111/8-11 m 4-30-82

4 O Table 111/8-2 Releases, Concentrations and Dose In Reactor Room for Release of

, 2.7% of the Gaseous Fission Products in One Fuel Plate Release Concentration, Ci/m3 Dose In Curies- Room, REM In Room At Stack Kr-85 0.0024 0.16 x 10 6 0.10 x 10 6 --

Xe-133 0.0170 1.10 x 10 6 0.71 x 10 6 0.2 x 10 3 1

Total Whole Body Dose Equivalent From Nobel Gases 0.2 x 10 3 1-131 0.0130 0.87 x 10 6 0.56 x 10 6 1.57 (thyroid)

I-132 0.0020 0.13 x 10- 0.08 x 10- 0.01 (thymid) .

Total Thyroid Dose Equivalent From Radioiodines 1.58 l

f t

s/

III/8-12

4-30-82 e - a ., - . - - - - e -- - - , - , ,r -- r 4 , e ev

8.4.5 SITE B0UNDARY AND AREA 0F RADIOLOGICAL CONCERN The emergency preparedness standard referred to above applies to concentrations at a site boundary and attempts to distinguish between on-site personnel and the general public beyond the boundary. The site boundary also encloses the area controlled by a licensee, and therefore distinguishes between on-site and off-site forces and resources available to a licensee for dealing with emergencies. For this last purpose, the site boundary is the campus boundary. However, the UCLA boundary is not fenced and no clear distinction can be made between the

. campus population (faculty, students, staff, and visitors) and the general public. flore significantly, the boundary is far from the reactor and concentrations at the boundary due to radiological releases from the worst possible accident at the reactor will be extremely small.

The foregoing can be illustrated by noting that the campus boundary nearest the reactor is about 830 meters east of the reactor exhaust stack. Employing the Gaussian plume model with a stack radius of 0.41 meters, an air exit velocity of 12.5 meters per second and a wind speed of 3.5 meters per second [23], the plume center line concentration 830 meters downwind will be reduced by a factor of 2000 from the value at the stack. When averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the concentration will be 13%

of MPC and 1/75 of the lowest Emergency Action Level of 10 times MPC.

The emergency preparedness standard may be used in reverse to define an area of radiological concern, beyond which the concentration will not p exceed 10 times MPC when the release is averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

O Using the previous parameters and the stack concentration from Table III/8-2, the calculated boundary is a circle, one meter in radius surrounding the stack. The ninth floor (roof) of the Mathematical Sciences Addition is an unrestricted area and the closest approach that the general public may easily make is about 8.2 meters from the stack.

The dilution factor at that distance is approximately 160, and therefore the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> averaged concentration at that point is about 1.3 times MPC or 1/6 of the criterion for invoking the lowest Emergency Action Level.

8.

3.6 CONCLUSION

S The radiological consequences of the worst credible accident will not expose any member of the general public to concentrations in excess of permissible limits. Further, the projected exposures are not at the level sufficient to require Notice of an Unusual Event under the applicable emergency preparedness standard [22].

O III/8-13 4-30-82

O 10. RErEaEnCES

1. R. D. MacLain,"UCLA Training Reactor Hazards Analysis, Final Report," Department of Engineering, University of California, Los Angeles, Report No. 60-18, March 1960.

2. AMF Atomics and General Nuclear Engineering Corporation,

" Start-Up Report AMF-GNE Educator Reactor, University of California, Los Angeles, January 1961.

3. American Society of Mechanical Engineers, Nuclear Reactor Plant Data, Research and Test Reactors, Volume 2. March 1959, McGraw-Hill Book Company, New York, New York.

4. C. E. Ashbaugh, " Nuclear Energy Laboratory - Report,"

School of Engineering and Applied Science, University of California, Los Angeles, 1977.

5. A. L. Babb, " University of Washington Training Reactor Final Hazards Sumary Report, College of Engineering, University of Washington, Seattle, Washington, May 1960.

6. Personal communication with Pat Miller, Reactor Supervisor, University of Washington, January 1980.

7. Personal communication with N. J. Diaz, Reactor Supervisor, p) y University of Florida, January 1980.

8. S. C. Hawley, et al., " Analysis of Credible Accidents for Argonaut Reactors," NUREG/CR-2079, April 1981.

9. G. E. Cort, " Fuel Temperatures in an Argonaut Reactor Core Following a Pypothetical Design Basis Accident (DBA),"

NUREG/CR-21h, June 1981.

10. " Safety Evaluation Report Related to Renewal of the Operating License for the Research Reactor at the University of California at Los Angeles, July 1981. Available through the U. S. Nuclear Regulatory Commission undet Docket No. 50-142, License R-71.

11. Brookhaven National Laboratory memorandum Neogy to Carew,

" Transient Analysis of Argonaut Reactors," November 28, 1981.

12. R. Greenfelder, " Maximum Credible Rock Acceleration From Earthquakes in California. Map Sheet 23," California Division of Mines and Geology, 1972. Revised, August 1974. j

13. Lindvall, Ricter & Associates, " Seismicity, Geology and Design:

Earthquake Report for Stone Canyon Dam," March 1976.

g 14. Ayyaswamy, P. et al., " Estimates of Risks Associated with Dam  !

Failure," UCLA-ENG-7423, March 1974.

()

III/10-1 4-30-82

15. City of Los Angeles, Department of Water and Power, " Stone p).

u. Canyon Dam Stability Evaluation," Report AX-213-34, March 1977,

16. D. R. Olander, " Fundamental Aspects of Nuclear Reactor Fuel Elements," Technical Information Center, TID-26711-PI, 1976,

p. 289-293.

17. D. S. Billington, " Radiation Damage in Reactor Materials,"

Peaceful Uses of Atomic Energy, Proceedings of the International Conference in Geneva, August 1955, Volume 7, p.425.

18. M. T. Simnad and L. R. 7amwalt, " Diffusion of Fission Products (and Transmutation Products) in Non-Fuel Materials," General Atomic Division of General Dynamics, GA-6542, December 1965, p.76_77.

19. C. E. Weber and H. H. Hirsch, " Dispersion Type Fuel Flements,"

Peaceful Uses of Atomic Energy, Proceedings of the International Conference in Geneva, August 1955, Volume 9, p. 197.

20. J. A. L. Robertson, _ Irradiation Effects in Nuclear Reactors, Gordon and Breach, New York, 1969, p.262-264.

21. Reed R. Burn, Personal Communication. University of Michigan, January 19, 1982. Confirmed no degradation at the high burnup values shown in "Research, Training, Test and Production Reactor Directory," American Nuclear Society,1980, p.1627.

22. " Standard for Emergency Planning for Research Reactors,"

ANS 15.16, Draf t 2, November 29, 1981. (Available from the American Nuclear Society, 555 North Kensington Avenue, LaGrange Park, Illinois, 60525).

23. Letter, Wegst (UCLA) to Tedesco (USNRC), September 5, 1980.

24. Title 10, Chapter 1, Code of Federal Regulations, Part 20, Appendix B, Table II, Column 1.

l l

! (

j 4-30-82 l

O APPENDIX III ARG0 NAUT SAFETY ANALYSIS REPORT (ASAR)

Attachment A Analysis of Credible Accidnets for Argonaut Reactors l

l NUREG/CR-2079-PNL-3691 Battelle Pacific Northwest Laboratory Richland, Washington April 1981

[ incorporated by this reference]

7 O

III/A 4-30-82

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

h O APPENDIX III ARGONAUT SAFETY ANALYSIS ~ REPORT (ASAR)

Attachment B i

Fuel Temperatures in an Argonaut Reactor Core Following a Hypothetical Design Basis Accident (DBA) 4 L

NUREG/CR-2198, by G. E. Cort Los Alamos National Laboratory l

Los Alamos, New Mexico June 1981 1 [ incorporated by this reference]

O 1

1 l

l l

i I

O I III/B 4-30-82 ,

O APPENDIX IV Emergency Rt,ganse Plan This section is reserved.

O O

4-30-82

__ _ _ .