ML20153D342

Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at Purdue University

ML20153D342
Person / Time
Site:	Purdue University
Issue date:	04/30/1988
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-1283, NUDOCS 8805090116
Download: ML20153D342 (69)

Text

- _. - - . -- - ..

NUREG-1283

Safety Evaluation Report relatec to the renewal of the operating license for the research reactor  ;

at Purdue University Docket No. 50-182 i

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation l April 1988 p g= %q

• ? '

l pEjWA"!29lp

\

., ( , T i

i l

NOTICE i I Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W. f Washington, DC 20655  !

i

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, l Washington, DC 20013 7082

3. The National Technical Information Service, Springfield, VA 22161 l

Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive. i Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection .

and Enforcement bulletins, circulars, information notices, inspection and investigation notices; l

Licensee Event Reports; vendor reports and correspondence: Commission papers; and applicant and  ;

licensee documents and correspondence.

l The following documents in the NUREG series are available for purchase from the GPO Sales l Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and  :

NRC booklets and brochures. Also available are Regulatory Guides NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Comminion issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic  !

Energy Comt.iission, forerunner agency to the Nuclear Regulatory Commission.

Documents available trom public and special technical libraries include all open literature items, such as books, joumal and periodical articles, and transactions. Federal Register notices, federal and '

state legislation, and congressionM reports can usually be obtained from these libraries.

I Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are avellable for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission. Washington. DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018.

NUREG-1283

- -_ . - - - _ - _ - _ . _ - _ _ _ _ _ . _ - - -__ -_ -_ . - - 3 Safety Evaluation Report related to the renewal of the operating license for the research reactor ~

at Purdue University Docket No. 50-182 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation April 1988

~$

f.

f. f

'6wi ,

l I

ABSTRACT This Safety Evaluation Report for the application filed by Purdue University for a renewal of Operating License R-87 to continue to operate a research reac-tor has been prepared by the Office of Nuclear Reactor Regulation of the U.S.

Nuclear Regulatory Commission. The facility is owned by Purdue University and is located on the campus in West Lafayette, Indiana. On the basis of its tech-nical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment.

l l

I i

)

NUREG-1283 iii I

I

l i I i

1 1

1 TABLE OF CONTENTS

. i i

Pa21 i i

ABSTRACT ............................................................... iii 1

i 1- INTRODUCTION ..................................................... 1-1 j l 1.1 Summary and Conclusions of Principal Safety j Considerations ..... ........................................ 1-2  !'

i 1.2 Reactor Description ......................................... 1-3

1. 3 Reactor location ............................................ 1-3 i i

1.4 Shared Facilities and Equipment and Special

! Location Features ........................................... 1-3

- 1.5 Comparison With Similar Facilities .......................... 1-3 1.6 Nuclear Waste Policy Act of 1982 ............................ 1-4

' 2-1 2 SITE CHARACTERISTICS ................... .........................

i ,

i 2.1 Geography ................................................... 2-1  :

) 2.2 Demography .................................................. 2-1 i i 2.3 Nearby Industrial, Transportation, and Military .

1 Facilities .................................................. 2-1 i 2-1 l 1 2. 4 Meteorology .................................................

! 2.5 Geology ..................................................... 2-4 1 2.6 Hydrology ................................................... 2-4 i 2.7 Seismology .................................................. 2-5 i

~

2. 8 Conclusion .................................................. 2-5 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS .................... 3-1 j 3.1 Reactor Building ............................................. 3-1 l 3.2 Wind and Water Damage ....................................... 3-1 l

3.3 Seismically Induced Reactor Damage .......................... 3-1

3.4 Mechanical Systems and Components ........................... 3-3 i 3.5 Conclusion .................................................. 3-3 i i

4 REACTOR .......................................................... 4-1 i 4.1 Reactor Core ................................................ 4-1

)

4

) 4.1.1 Fuel Assemblies ..................................... 4-1  !

4.1.2 Control Rods ........................................ 4-3 1 4.1.3 Neutron Source ...................................... 4-3 i l 4.2 Reactor Pool and Biological Shield .......................... 4-5 -

]

3 4.3 Grid Plates and Core Support Structure ...................... 4-5 4.4 Reactor Instrumentation ..................................... 4-5

{ 4-5

4.5 Dynamic Design Evaluation ...................................

J i

! NUREG-1283 y l.

1

-l TABLE OF CONTENTS (Continued) l W x 4.5.1 Excess Reactivity and Shutdown Margin ............. 4-6 4.5.2 Conclusion ........................................ 4-6 ,

4.6 Functional Design of Reactivity Control Systems ........... 4-7 i

4.6.1 Control Rod-Orive Assemblies ...................... 4-7 4.6.2 Control Rod Circuitry and Interlocks .............. 4-7 ,

4.6.3

~

Conclusion ........................................ 4-8 4.7 Operational Procedures .................................... 4-8 4.8 Conclusion ................................................ 4-8 5 REACTOR COOLANT AND ASSOCIATED SYSTEMS .......................... 5-1 ,

5.1 Primary Cooling System .................................... 5-1

5. 2 Process Water System ...................................... 5-1 Primary Coolant Makeup Water System .......................

5.3 5-1 '

5.4 Primary Coalant Chiller S 5-1 5.5 Conclusion ..............ystem ............................

.................................. 5-3 6 ENGINEERED SAFETY FEATURES ...................................... 6-1  !

6.1 Ventilation System ........................................ 6-1 f 6.2 Drain System .............................................. 6-1 l 6.3 Conclusion ................................................ 6-1  !

l 7 CONTROL AND INSTRUMENTATION SYSTEMS .............. .............. 7-1 I 7.1 Reactor Control System .................................... 7-1 [

t 7.1.1 Control Rod Drives ................................ 7-1 7.1.2 Servo Control System .............................. 7-1 i 7.1.3 Neutron Source Drive .............................. 7-1 7.1.4 Fission Chamber Drive ............................. 7-1 i 7.1.5 Annunciator and Alarm Systems ..................... 7-3 l

7.2 Reactor Instrumentation ................................... 7-3 I 7.2.1 Channel No. 1 - Startup Channel ................... 7-3  !

7.2.2 Channel No. 2 - Log N and Period Channel .......... 7-3 l 7.2.3 Channel No. 3 - Li near Power . . . . . . . . . . . . . . . . . . . . . . 7-3 t 7.2.4 Channel No. 4 - Safety Channel .................... 7-3 7.2.5 Temperature and Water Monitor Channels ............ 7-4 l 7.2.6 Radiation Monitoring Instruments .................. 7-4 i

7.3 Scram System and Interlocks ............................... 7-4 ,

7.4 Conclusion ................................................ 7-4 I i

i I

NUREG-1283 vi  !

I l

i

. - - - . - - -. - -~ - . -- - _ - .__ -- - . . ,

l TABLE OF CONTENTS (Continued) l l Pag 8 ELECTRIC POWER ..................................... ........... 8-1 8.1 Electrical Power System ................................... 8-1 8.2 Emergency Power ........................................... 8-1 8.3 Conclusion ................................................ 8-1 9 AUXILIARY SYSTEMS ............................................... 9-1 9.1 Ventilation System ........................................ 9-1 9.2 F i rc Protec ti o n System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.3 Fuel Storage System........................................ 9-1 9.4 Heating and Air Oonditioning System ....................... 9-1 9.5 Crane System .......... ................................... 9-1 9.6 Conclusion ................................................ 9-2 10 EXPERIMENTAL PROGRAMS ..... ..................... ............... 10-1 10.1 Experimental Facilities ................................... 10-1 10.1.1 Reflector Tubes ................................ .. 10-1 10.1.2 Orop Tubes ........................................ 10-1 10.2 Experiment Review .. ........ ............................ 10-1 10.3 Conclusion ................................................ 10-1 11 RADI0 ACTIVE WASTE MANAGEMENT .................................... 11-1 11.1 ALARA Commitment .......................................... 11-1 11.2 Waste Genaration and Handling Procedures .................. 11-1 11.2.1 Solid Waste ....................................... 11-1 11.2.2 Liquid Waste ...................................... 11-1 11.2.3 Airborne Waste .................................... 11-2 11 3 Conclusion ........... .................................... 11-2 12 RADIATION PROTECTION PROGRAM .................................... 12-1 12.1 ALARA Commitment . ........................................ 12-2 12.2 Health Physics Program .................................... 12-2 12.2.1 Procedures ........................................ 12-2 12.2.2 Instrumentation ................................... 12-2 12.2.3 Training .......................................... 12-2 12.3 Radiation Sources ......................................... 12-2 12.3.1 Reactor ...................... .................... 12-2 12.3.2 Extraneous Sources ................................ 12-3 NUREG-1283 vii

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

i l

TABLE OF CONTENTS (Continued)

.P., age i 12.4 Routine Monitoring ........................................ 12-3 12.4.1 Fixed-Position Monitors ........................... 12-3 12.4.2 Wipe Tests ........................................ 12-3 12.5 Occupational Radiation Exposures .......................... 12-3 l 12.5.1 Personnel Monitoring Program ...................... 12-3 12.5.2 Personnel Exposures ............................... 12-3 12.6 Effluent Monitoring ....................................... 12-4 12.6.1 Airborne Effluents ................................ 12-4 12.6.2 Liquid Effluents .................................. 12-4 12.7 Environmental Monitoring .................................. 12-4 12.8 Potential Dose Assessments ................................ 12-4 12.9 Conclusion ................................................ 12-5 13 CONDUCT OF OPERATIONS ........................................... 13-1 13.1 Overall Organization ...................................... 13-1 13.2 Training .................................................. 13-1 13.3 Operational Review and Audits ............................. 13-1 13.4 Emergency Planning ........................................ 13-1 13.5 Physical Security Plan .... ......... .................. 13-1 13.6 Conclusion .......................................... ..... . . . 13-2 14 ACCIDENT ANALYSES ............................................... 14-1 14.1 Fuel-Element-Handling Accident ............................ 14-1 14.1.1 Scenario ................................. ........ 14-1 14.1.2 Technical Assessment .............................. 14-1

.14.2 Maximum Reactivity Insertion .............................. 14-2 14.2.1 Scenario .......................................... 14-2

, 14.2.2 Technical Assessment .............................. 14-4 14.3 Flooding of an Irradiation Facility and Failure of a Movable Experiment ............. .......................... 14-4 14.3.1 Technical Assessment ................. ............ 14-5 14.4 Loss-of-Coolant Accident .................................. '4-5 14.4.1 Scenario .................. ....................... 14-5 14.4 2 Technical Assessment .............................. 14-5 i

NUREG-1283 viii 4

, , - - , . - ~ , , - - , , , , - - - - - , - , - , _ . , ,-r ,

, , - ,, .+--vr n-=- -,,-, y

TABLE OF CONTENTS (Continued)

P. age.

14.5 Maximum Hypothetical Accident ............................. 14-5 14.5.1 Scenario .......................................... 14-6 14.5.2 Technical Assessment .............................. 14-6 14.6 Conclusion ................................................ 14-7 15 TECHNICAL SPECIFICATIONS ........................................ 15-1 16 FINANCIAL QUALIFICATIONS ........................................ 16-1 17 OTHER LICENSE CONSIDERATIONS .................................... 17-1 18 CONCLUSIONS ..................................................... 18-1 19 BIBLIOGRAPHY .................................................... 19-1 FIGURES i

2.1 Purdue University Main Campus and the Lafayette-West Lafayette Vicinity .............................................. 2-2

2. 2 State of Indiana ................................................ 2-3 3.1 Floor Plan of Nuclear Engineering Laboratories, Including -

Reactor Room .................................................... 3-2 I 4.1 Facility Layout ................................................. 4-4 4.2 Core Configuration .............................................. 4-4 5.1 Reactor Water Process System .................................... 5-2  !

6.1 Reactor Room Ventilation and Cooling System ..................... 6-2 7.1 Reactor Control System .......................................... 7-2 12.1 Organizational Structure for PUR Operations ................... 12-1 ,

1 TABLES 4.1 PUR Principal Design Parameters ................................. 4-2  !

1 12.1 History of Personnel Radiation Exposure at PUR Facility ......... 12-4 14.1 Results of Power Transient Analysis With Ramp Insertion of Control Rod (Case A) ......................................... 14-3 14.2 Results of Power Transient Analysis With No Control Rods (Case B) ........................................................ 14-3 14.3 Comparison of Important Fuel Data for PUR and SPERT-1 ........... 14-4 14.4 Dose Rates in the Reactor Room From a Failed Fuel Experiment .... 14-7 14.5 Cose Rates at 100 m From a Failed Fuel Experiment ............. . 14-8 NUREG-1283 ix  :

- . , . _ _ _ . - . . . _ - - _ - ~ - , _ , - - _ . . _ . . . , _ -.,-. _ _ . _ ,.m .

1 \

l 1 INTRODUCTION Purdue University (Purdue/ licensee) submitted a timely application to the U.S.

Nuclear Regulatory Commission (NRC/ staff) for renewal of the Class 104 Operating License (R-87) for its open pool-type research and training reactor. The application, with supporting documentation, was transmitted by letter dated June 30, 1986, as supplemented, requesting renewal of the license for a period of 20 years. The licensee is permitted to operate the reactor within the con-ditions authorized in the existing license, as amended, in accordance with Title 10 of the Code of Federal Regulations, Paragraph 2.109 (10 CFR 2.109),

until NRC action on the renewal request is co npleted.

The renewal application references information regarding the Sriginal design of the reactor facility and contains information about modifications to the facility made since the initial licensing.

The application also includes a revised Final Safety Analysis Report (FSAR),

information for an environmental impset assessment, financial information, an Operator Requalification Program, and revised Technical Specifications.

Supplemental information included revisions to the Purdue University Physical Security Plan, which is withheld from public disclosure in accordance with 10 CFR 2.790.

The staff's technical review with respect to issuing a renewal operating license ,

to Purdue was based on visits to the facility and on the information contained '

in the renewal application and supporting documents, plus responses to requests for additional information. This material is available for review at the Com-mission's Public Document Room at 1717 H Street, N.W., Washington, D.C. 20555.

This safety Evaluation Report (SER) was prepared by R. E. Carter and A. Adams, 'r.,

Project Managers, Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission. Major contributors to the technical review include the Project Managers and C. H. Cooper and W. R. Carpenter of the Idaho National Engineering Laboratory under contract to NRC.

The purpose of this SER is to summarize the results of the safety review of the Purdue University reactor (PUR) and to delineate the scope of the technical i details considered in evaluating the radiological safety aspects of continued f operation. This SER will serve as the basis for renewal of the license for operation of the PUR facility at power levels up to and including 1000 W thermal (Wt). The facility was reviewed against Federal regulations (10 CFR 20, 30, 50, 51, 55, 70, and 73), applicable regulatory guides (principally Division 2, Research and Test Reactors), and appropriate accepted industry stendards ,

[American National Standards Institute /American Nuclear Society (ANSI /ANS) 15 t series]. Because there are no specific accident-related regulations for research 7 reactors, the staff has at times compared calculated hypothetical radiation #

dose values with related standards in 10 CFR 20, "Standards for Protection Against Radiation," for employees and the public.

NUREG-1283 1-1

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

The PUR was initially licensed for operation at 1.0 kWt in August 1962 as an open pool-type reactor, with fuel of the Materials Testing Reactor (MTR) type.

Only minor modifications have been made to the reactor since the initial licensing.

1.1 Summary and Conclusior: of Principal Safety Considerations The staff's evaluation considered the information submitted by the 1!censee, past operating history recorded in annual reports submitted to the Comission l by the licensee, reports by the Commission's Region III, and onsite observations, i In addition, as part of its licensing review, the staff obtained laboratory '

studies and analyses of credible accidents postulated for plate-type reactors.

The principal safety matters reviewed for the PUR and the conclusions reached follow:

(1) The design, testing, and performance of the reactor structure and systems and components important to safety during normal operation are inherently safe, and safe operation can reasonably be expected to continue.

(2) The expected consequences of a broad spectrum of postulated credible accidents have been considered, emphasizing those that could lead to a loss of integrity of fuel-element cladding. The staff performed conserva-tive analyses of the most serious credible accidents and determined that the calculated potential radiation doses outside the reactor room would not exceed 10 CFR 20 guidelines for persons in unrestricted areas.

(3) The licensee's management organization, conduct of training and research activities, and security measures are adequate to ensure safe operation of the facility and protection of its special nuclear material.

(4) The systems provided for the control of radiological effluents can be operated to ensure that releases of radioactive wastes from the facility are within the limits of the Commission's regulations and are as low as reasonably achievable (ALARA).

(5) The licensee's Technical Specifications, which provide limits controlling operation of the facility, are such that there is a high degree of assur-ance that the facility will be operated safely and reliably.

(6) The financial data provided by the licensee are such that the staff has determined that the licensee has sufficient revenues to cover operating ,

cests and eventually to decommission the reactor facility. j i'

(7) The licensee's program for providing for the physical protection of the facility and its special nuclear material complies with the requirements 1 of 10 CFR 73.  ;

(8) The licensee's, procedures for training reactor operators and the plan for operator requalification are acceptable. These procedures give reasonable assurance that 'he reactor facility will be operated with competence.

(9) The licensee's Emergency Plan provides reasonable assurance that the l licensee is prepared to assess and respond to emergency events.

NUREG-1283 1-2 2

. -g - -- ,,n-v -, - -r--,

1.2 Reactor Description The PUR is a heterogeneous, swimming pool-type nonpower reactor. The core is cooled by natural convection of light water, moderated by water, and reflected by water and graphite. The reactor core is located near the bottom of a water-filled tank surrounded and sipported by a concrete shielding structure. The reactor core rests on supports on the bottom of the tank, and the control mechanisms and d?tectors are suspended from a support plate at the top of the tank.

The reactor core is composed of approximately 16 fuel elements positioned in holes in an aluminum grid plate. The grid plate contains a rectangular matrix of holes to allow the cht.nging of fuel element locations and the insertion of graphite reflector elements to displace reflector water. Each fuel element consists of several thin metal plates assembled into a unit about 7 cm by 7 cm with an active fuel length of *0.60 m. Fuel elements of this general configuration were first designed for and used in the Materials Testing Reactor (MTR) and thus are referred to as MTR-type fuel elements. Three of the fuel elements were fabricated with the four middle plates missing, providing space for the positioning and movement of the reactor control rods.

Reactivity of the reactor core is changed by the operator moving the control rods that are suspended from fail-safe electromagnets. The ionization chambers used for sensing neutron and gamma-ray fluxes are suspended near the core. The control console, from which the operator can observe the reactor pool and the top structuras, is located adjacent to the reactor. The control console con-sists of typical read-out and control instrumentation.

1. 3 Reacter Location The PUR is housed in a small room designed and dedicated for that purpose in the Duncan Annex of the electrical engineering building on the east side of the campus of Purdue University in the city of West Lafayette. The nearest larger city is Lafayette, which is located about 2 km from the site.

1.4 Shared Facilities and Equipment and Special location Features The electrical engineering building is in close proximity to other buildings on ,

the campus and obtains utility services such as water, electricity, and sanitary sewage from the .nain campus systems. There are no special features associated with the facility location.

1. 5 Comparison With Similar Facilities The fuel used in the PUR is based on the MTR design and is very similar to the fuel used in approximately 50 other research reactors operating in the United States and at least 25 reactors operating in foreign countries. The control and instrumentation systems, although different in detail, are based on the same operating principles as those used for these otner 75 research or test reactors.

NUREG-1283 1-3

1. 6 Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test reactor, that the applicant shall have entered into an agreement with the Department of Energy (D0E) for the disposal of high-level radioactive waste and spent nuclear fuel. DOE (R. L. Morgan) has informed the NRC (H. Denton) by letter dated May 3,1983, that it has determined that universities and government agencies operating nonpower reactors have entered into contracts with DOE that provide that DOE retain title to the fuel and is obligated to take the spent fuel and/or high-level waste for storage or reprocessing.

Because Purdue University has entered into such a contract with DOE, the applicable requirements of the Waste Policy Act of 1982 have been satisfied.

NUREG-1283 1-4

2 SITE CHARACTERISTICS 2.1 Geography The PUR is located near the eastern edge of the Purdue University campur, in West Lafayette, Tippecanoe County, Indiana. There are few large centers of population in Indiana, and the nearest large city is Indianapolis, approximately 100 km to the southeast of the site.

The general terrain of and around the campus and the reactor site itself are located in a relatively flat area.

The location of the PUR within the campus is shown in Figure 2.1. The nearest off-campus residential area is approximately 50 m from the reactor building.

Figure 2.2 shows the location of Lafayette with respect to other major cities in Indiana, and to Chicago, Illinois.

2.2 Demography The daytime population on the campus near the reactor site is normally less than 40,000 people, including students and university staff. Because the campus is adjacent to a residential area, the permanent population within 2 km during werking hours is normally about 30,000 people. Most of the population of the Lafayette area resides within 8 km of the reactor site.

2.3 Nearby Industrial, Transportation, and Military Facilities There is no large industry, heavily traveled transportation route, or military installation in or near Purdue, nor is there a heavily traveled airport within several kilometers. ,

Because there are no industrial, military, or major transportation facilities l in the near vicinity of the reactor site that could directly or indirectly cause accidental damage to the reactor, the staff concludes that such accidents need not be hypothecized and evaluated.

2.4 Meteorology The general climate of Tippecanoe County is continental with hot summers and cold winters. The seasons are strongly marked, and the weather is frequently changeable. The average annual temperature is about 10 C. The mean temperature in January, the coldest month, is about -4 C, and in July, the warmest month, it is 23 C. Prevailing winds are from the west or southwest during the winter and from the south during the summer. The average annual precipithtion is about 0.9 m; July is the wettest month, and February is the driest. This region of the United States is subjected to tornado activity, primarily during the late spring and early summer. In Tippecanoe County as a whole, 25 tornados have .

been recorded during the past 35 years, with no significant damage occurring I on the Purdue campus.

NUREG-1283 2-1

'%g,

i 'a .'

i I -

~ . .,

\_ 4.s. N 4 l . )

I

• • • \ j \,1 1 l,

a J\ j "ws ', h,8r 1

~

, ~ ,,, ,

,;huk4**LNgy\ ,y i- em d r F f.,._.p'N j ',t

% gLJ

. =

- --i 3

G $, Q,,,m'fxv ~

L "'" "

e

.  ; _Q._

..,.I ,

g ,4,g,4

, i ,.:

' J=J

- / ' -

ij. -

g A f- '{

N / { f.;l,;,(..

?F c N 4nu *N M,

II;.

3

'( CM

,t

,p.g _. . . \

f ii-.- r .

p,jj 4

o i '

~ ,qp s- gr I{ [isic ,gyjO' ,

Duncan Annex l

7 h $l J

(PUR)

~

I:-]' Sf -i._..,,,i:femmy I gr f"y

~

~ 87 a~ .... .

& ) l ' .... ,. ; * *g l l

i' i r-l Figure 2.1 Purdue University main campus and the Lafayette-West Lafayette vicinity NUREG-1283 2-2

i l

MICHIG AN CITY CHICA

. es u

@ @ wGO HAMMONO U

69 30 FT. WAYNE 65 KOKOMO 9

LAFAYETTE e MUNICE ANDERSON y

g INDI ANAPO Lis g l RICHMOND n

(8 TERRE HAUTE l

e BLOOMINayon I

l 65 I

NEW ALBANY

/

g EVANSVILLE Figure 2 2 State of Indiana l

NUREG-1283 2-3 y n. , . m.--..-.m ___,--.y-- - 7,, - - , _ _ , m- . _ _ - _ , , . . _ , . , , m- m- ,e--,. ,,w.. -,,.,y y,m n _ . -, w - m--- yym?

2.5 Geology l

The county lies within the Tipton Till Plain of Indiana and is a section t. the Till Plains subprovince of the U.S. Central Lowlands physiographic province.

Most of the soils in this area are derived from the glacially deposited material.

Extensive upland areas are covered with a thin mantle of loose deposits. A few areas are covered with soils of alluvial, colluvial, or organic origin. Glacial drift covers the bedrock to a depth ranging from a few feet to more than 300 feet. The underlying '.,edrock, consisting of flint, shale, sandstone, and limestone of the Mississippian period, is exposed as rock terraces in the Wabash Valley and on the upland in the western part of the county. Purdue University is located above an extensive glacial deposit of sand and gravel.

The land surface of Tippecanoe County is flat to rolling, except where the major streams have cut deeply into the surface. The entire county lies within the drainage basin of the Wabash River and its tributaries. The land slopes generally southwestward with the streams flowing westw rd. Two main tributaries, the Tippecanoe River and Wild Cat Creek, enter the Wabash River upstream from the campus. Minor tributaries include Little Pine Creek, Indian Creek, Burnetts Creek, Hott's Creek, Sugar Creek, Buck Creek, Wea Creek, and Flint Creek.

2.6 Hydrology Most of Tippecanoe County is covered by glacial drift. The drift ranges in thickness from a thin veneer to about 435 feet and was deposited on a bedrock surface that was eroded by a preglacial drainage system. Much of the surface drift consists of glacial till. Water-laid cross-bedded sand and gravel are associated with the till. The subsurface glacial deposits also include much till with interbedded sand and gravel. Locally, clay deposits are as much as 106 feet thick. Within the drif t, five sheetlike water-bearing units are differentiated in parts of the county. Ground water within these units occurs under artesian and w:ter-table conditions. Leca11y these may occur within the same unit.

This area was repeatedly glaciated during the Pleistocene epoch. Before glacial times, a giant drainageway, now known as the Teays River, flowed from the Appalachian Mountains across Ohio and passed northwestward through the present site of Lafayette-West Lafayette. Illinoian ice dammed the preglacial Teays River channel and ponded the relatively small glacial Lake Lafayette. An outlet channel, developed to drain this preglacial lake, was subsequently perpetuated as the present Wabash River drainage line southwestward from the Lafayette-West Lafayette area.

The elevation of the Purdee University campus is approximately 706 feet, and the level of the Wabash River is approximately 510 feet. With this difference of over 100 feet, the flow of both surface water and ground water is in a generally easterly and southerly direction toward the Wabash River, which flows around two sides of the campus.

Any leakage of contaminated water frori. the PUR represents no potential hazard to either the West Lafayette or Purdue 'loiversity water supply, since these flows are away from the well fields of bath. The Wabash River represents a natural barrier between the reactor and t.'e Lafayette well fields, so no potential hazard exists there.

NUREG-1283 2-4

2.7 Seismology Information on seism'c activity in the Central United States (NUREG/CR-1577, j "An Approach to Seisr.iic Zonation for Siting Nuclear Electric Power Generating i Facilities in the Eastern United States," May 30, 1981) shows that the PUR is l located in that portion of Indiana that lies in a zone for low seismic activity, within which might result only minor damage to structures caused by distant earthquakes.

The three most significant seismic-source zones that are closest to West Lafayette are (1) the New Madrid area of southeastern Missouri (2) the Wabash Valley Fault system of southwestern Indiana and southeastern )

Illinois (3) the Anna, Ohio aren Reasonable estimates of the maximum magnitude events that could occur in those areas give values of 7.4, 6.6, and 6.3 (body wave motion) for the seismic zones, respectively. Based on the distance from these zones (400, 200, and 200 km, respectively) and attenustion curves, estimates for peak horizontal acceleration at West Lafayette for maximum magnitude events that are likely to occur at these three seismic zones show that these events would cause insignificant damage to Purdue buildings.

The staff concludes that the history of earthquake activity with no damaging i historic earthquakes near West Lafayette supports the conclusion that the risk l of seismically induced hazards to the PUR is not significant. i 1

2. 8 Conclusion The staff has reviewed and evaluated the PUR site for both natural and manmade hazards and concludes that there are no significant risks associateJ s',th the '

site that make it unacceptable for the continued operation of the reactor.

P NUREG-1283 2-5

3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS The licensee's Safety Analysis Report provides information on the design, construction, and functions of the reactor building, reactor systems, and auxiliary systems.

3.1 Reactor Building l The Duncan Annex of the electrical engineering building is constructed of brick, I concrete block, and reinforced concrete and was originally designed as a large high-voltage laboratory. It was subsequently subdivided into offices, class-rooms, and laboratories. The reactor is located in the southwest corner on the ground floor in a high bay area of the building. Figure 3.1 shows the floor plan of the nuclear engineering laboratories, including the reactor room.

The outside air supply and exhaust both pass through high-efficiency particulate air filters. The reactor room is maintained at negative air pressure (minimum 0.05 inch of water). All doors to the reactor room have foam rubber seals.

Steam heat is used to heat the room, and a room air conditioner circulates and cools the reactor room air.

The only floor drain to the sewers is sealed except for a vent opening. This vent is raised about 2 feet above the floor and has a filtered inverted opening.

Condensate from the air conditioner is released to this drain through an opening 12.0 feet above the floor.

3.2 Wind and Water Damage The Purdue University campus area experiences few extreme wind condition = such as tornados or inland hurricanes. Furthermore, the reactor building is constructed from concrete blocks and the reactor pool is formed of steel-reinforced poured concrete. The reactor site is well above the flood plain; therefore, wind or water damage to the PUR facility is very unlikely.

3.3 Seismically Induced Reactor Damage The information on past seismic activity and the likelihood of future earth-quakes in the area of the Purdue University campus indicates that the PUR is in a region where there is low probability of severe seismic activity. If an earth-quake should cause catastrophic damage to the reactor building and/or the reactor pool, water might be released. However, Section 14 of this SER shows that loss of coolant in the PUR would not lead to core damage, and mechanical damage to fuel cladding would release only a small fraction of the very low inventory of fission products. j HUREG-1283 3-1

B 71 -

1 I

i 7 t ii.@!.S.52d. IfIlBII::f;.:.% -

, :v - I i -.

. . . :.. :B:70 A::

. Reactor Room'i i 3,  ; , S70 8 ' 1 I

1

l l

8 72 h

-M::" !Q:i.

-' '.: 4 -

. . . ... . . . ( D. i _.

6.q} ..p a73

_hl

. . . ::g I

p:: [N _e

!:Etti: i:x . M: 1] *

- :. i;f:: . . . . . . , .

W::  ::e:.8. :.7..::

.::: o <, -

W. . .

.-.M. .:d. :e:.

.:.i.:. . .

37.6. A .  ::.N.

aj?i?.!: . i:x. . y:!.  ;:: ........

!s.:g:6:i

~: : .

E!N%.ii .. s.

[':: . . .

.g w.+. . . . .. ....

5 : . . . ::.w: :k..::.. ::. ... .

.p: . g-- .: .:.h.

~ ' :: ..:

i:s. .s. .d. .:i  :! ;i;j;:::

.:. .::.3 7.,.9. ::

.:g. .-

..:., . g .. 3

...f.. . ..... .

NObi  : s:a

>.)

.....::q 1 I i l!: .. :'rs: i:jM  :

0 10' 20'

....g: [ -Q ., :r:-

p . ,

. q: . I::

!:B:e$i. . . . . .

.  ::8. 54::.

-m . -j::.

i l

Figure 3.1 Floor plan of nuclear engineering laboratories, including reactor r)om NUREG-1283 3-2 1

3.4 Mechanical Systems and Components The mechanical systems of importance to safety are the neutron-absorbing con-trol rods suspended from the reactor superstructure. The motors, gear boxes, electromagnets, switches, and wiring are above the pool-water level and readily accessible for testing and maintenance. The staff has addressed the effects of aging on the continued performance of these components in Section 17 of this SER.

3.5 Conclusion On the basis of the above considerations, the staff concludes that the PUR facility was designed and built to adequately withstand all credible and likely wind, water, and seismic damage associated with the site. These considerations indicate that natural events would lead to small reactor-related consequences to the environment. Furthermore, the design and performance of Accordingly, the safety systems have been verified by more than 25 years of operation.

the staff concludes that the reactor systems and components are adequate to provide reasonable assurance that continued operation will not cause significant radiological risk to the health and safety of the public.

NUREG-1283 3-3 N

...m._ - . _ -. - -

4 REACTOR The PUR was built by Lockheed Nuclear Products and. initially attained critical-ity in August 1962. This reactor uses MTR-type 93% enriched uranium-235 (U-235) l aluminum-clad fuel plates that are assembled into fuel assemblies and placed into a gisphite-reflected region to form the reactor core. The reactor core is

! immersed in an open tank of light water that serves as the neutron moderator, coolant, and shield. The reactor operates at a maximum power level of 1 kW.

The reactor power is regulated by inserting or withdrawing neutron-absorbing control rods.

The reactor is used as a neutron source for activation analysis studies, aca-demic research, and the limited production of radioactive isotopes. It also is used as a training facility for the nuclear engineering educational program.

The PUR is operated for an average of about 13 kWh/yr. The principal design parameters for the current core configuration are listed in Table 4.1.

The PUR facility layout in the Duncan Annex of the electrical engineering building is shown in Figure 4.1.

4.1 Reactor Core The core of the reactor is 30.48 cm square and 60.95 cm high. It consists of ,

13 fuel assemblies and 3 control rod assemblies. Each fuel assembly consists of up to 10 aluminum enriched-uranium alloy plates. Et.ch control rod assembly consists of up to six plates and two aluminum guard plates with space for con- i trol rods. Adjustments to ensure that maximum excess reactivity is not exceeded are effected by substituting dummy fuel plates for uranium plates. The reactor 1

is cooled and moderated by a pool of light water. The 4 x 4 array of fuel assemblies is reflected on all sides with gaaphite-reflector elements and on the top and bottom with water. The 20 reflector elements are composed of l i

graphite waterproofed with epoxy resin and are contained in standard fuel j assembly cans. One row of six graphite-reflector elements is designed to hold samples for isotope production (see Figura 4.2).

4.1.1 Fuel Assemblies The MTR-type fuci plates are 93% enriched U-235 metal alloyed with 1100 alumi-num alloy and clad with 0.05-cm 1100 aluminum alloy, with a total thickness of 0.15 cm. These flat MTR-type fuel plates are then inserted in aluminum can-isters. Up to 10 fuel plates (7 x 64 x 0.15 cm overall dimensions) are con-tained in each of 13 standard fuel assemblies. Up to six plates plus two guard plates of 6061 aluminum alloy are contained in each of three control assemblies.

The number of fuel plates in the fuel assemblies can be adjusted to provide for a maximum excess reactivity of 0.6% Ak/k.

l l

l NUREG-1283 4-1

Table 4.1 PUR principal design parameters Parameter Value Maximum power level 1 kW Geometry of core 0.3 x 0.3 x 0.6 m Moderator-coolant Light water Maximum excess reactivity 0.6% ak/k Prompt neutron lifetime 77.2 x 10 8 s Fuel assemblies Number (total) 16 Standard-type 13 Control assembly-type 3 Number of plates per standard assembly 10 Humber of plates per control assembly 6 Plate dimensions 7.0 x 64 x 0.15 cm Active fuel length 59.4 cm U-235 per plate 16.5 g Water gap 0.53 cm Cladding 0.051-cm aluminum l Enrichment 93%

1 Reflector  !

Material on sides Graphite Number of graphite assemblies 20 Control rods and drives Number of regulating rods 1 Number of shim safety rods 2 Total number of control rods 3 Measured worth of control rods Regulating rod 0.26% ak/k Shin. safety rod no.1 5.0% ak/k Shim safety rod no. P 2.4% ak/k Rod speed out Regulating rod 45.0 cm/ min i Shim safety rods 11.2 cm/ min Scram time for complete insertion 1s l Material i Regulating rod Hollow stainless steel Shim-safety rods Solid borated stainless steel Size Regulating rod 1.3 x 5.7 x 64.8 cm Shim-safety rods 1.3 x 5.7 x 64.8 cm Maximum rate of reactivity change Regulating rod 0.006% ak/k/s Shim-safety rod no. 1 0.031% ak/k/s Shim safety rod no. 2 0.013% ak/k/s a

HUREG-1283 4-2

l Table 4.1 (Continued)

Parameter Value Average rate of reactivity change

' Regulating rod 0.0031% ok/k/s p Shim-safety rod no. 1 0.015% Ak/k/s Shim-safety rod no. 2 0.007% Ak/k/s '

i l

l Reactivity effects Temperature coefficient Calculated -2.1 x 10 2% Ak/k per C Measured -3.4~x 10 2% ak/k per C Void coefficient (measured) -2.6 x 10 2% Ak/% void Process water Resistivity >330,000 ohm-cm pH 5.5 1 1 Flow rate 1.89 L/s 4.1.2 Control Rods Three control rods are used to control and regulate the power levels in the PUR: one regulating rod and two shim-safety rods. Each of the three rods operates within a hollow gJide tube. The neutron absorber in the regulating rod is stainless steel, and the neutron absorber in the shim-safety rods is borated stainless steel. I'.ach control rod is 64.7 cm long and has a vertical travel of ~61 cm. The cross-sectional dimensions are 1.3 x 5.7 cm for all the rods. Tha maximum rate of withdrawal for the control rods corresponds to 0.031% ak/k/s and 0.013% ak/k/s for the two shim-safety rods and 0.006% ak/k/s for the regulating rod.

The neutron-absorbing sections of the shim-safety rods are supported by electro-magnets that release the rods in a scram. The scram time for complete insertion 3 of these shim-safety rods is 1 second. The regulating rod is mechanically con-nected to its drive and does not scram.

4.1.3 Neutron Source The PUR utilizes a 5-Ci plutonium-beryllium neutron startup source. The source is located in a special reflector element source holder adjacent to and just outside the graphite reflector (see Figure 4.2). The source can be withdrawn from its in-core position manually by means of an attached steel cable that is connected to the top of the source holder cap. An indicator light coupled to .

the startup meter at the control console shows whether the source is in or out of the core.

i NUREG-1203 4-3

-..- -.3 , ,,, ,,w_r ,-.--s_ ,- -%,_, yv.-- - ,-- . ,.-.,,.r- ,_.,f._,- , , ,, y,-- -r -

$talrway Office Storage Reactor b [ I"'

tab Countar(

Roon Office Hen Chemical tab Z

+ -6

]

LAJ LAJ

• ry1 ry] ryl ._,1 m _ Stores Electrical Pile Room Hater Loop Liquid tietal Shop "

Office Loop I"ht a y 1

3 3 3k, , r-F Figure 4.1 Facility layout Neutron Detectors

~~ '/ N s.

'O '/. 's:// ',' ',f,f,i

//,

7/,' , ',/ '

/ GRAP 11 TJ,'hE i,[

,'e,Y, _.j,/[

, 1,ECTOR , /,/ ,

' h.

Z', g / '/1 '

d

,f' FUEL ,

,3,//

• f, '

{y! Shim Shirn Source 9', Safet: Safet> j/ 's 1.27 cm Drop Tube

' ///:

'/' 2 I 7 .

/// @

/'N Y

'/, fl'h

.b .

'e ' Iteq Fissioi //.

,. j .

Rod Cha~mbei

./ /

' 7.62 cm Drop Tube j' / _ _

1n *

/.

. / / ,

, i s. ,/ /!/

/12.7 cm Drop Tube Figure 4.2 Core configuration NUREG-1283 4-4

4.2 Reactor Pool and Biological Shield The reactor core is located within two coaxial tanks that form the reactor pool.

The outer tank rests on a concrete pad 4.6 m below floor level. The reactor pool is built below floor level except for the 1-m wall that serves as a bio-logical shield for the operators and experimenters. The pool is contained in a cylindrical tank 5.3 m deep and 2.4 m in diameter. The core is located to one side to provide additional space for experiments. Opposite from the reactor core, two fuel storage racks are mounted cn the tank floor. These fuel storage racks are fabricated of aluminum and contain a boral sheet in their centers as a neutron absorption material.

j The supports for the drive mechanisms for the control rods, the fission chamber -

j and the source, and the neutron detectors are fastened to the support plate at the top of the tank. A traversing mechanism was mounted on the top of the reactor pool wall after the reactor was built. A lightweight, portable alumi-num bridge can be placed across the pool for mairitenance and fuel-handling i operations.

Shielding over the core is provided by 4 m of water, which reduces the radiati e level at the top of the pool to less than 1 mrem /h when the core is operati.g at 1 kW. The concrete pad, reactor tank, and distance reduce the maximum radia-tion level at the control console area to less than 0.1 mrem /h at 1 kW.

4.3 Grid Plates and Core Support Structure t A 7 x 11 position grid plate supports the 16 fuel assemblies and 20 reflector '

and isotope production elements. The approximate active core dimensions are 30.5 x 30.5 x 61 cm. The core structure is centered approxicately 76 cm from the center of the reactor tank and 9 cm from the bottom of tha tank.  ;

4.4 Reactor Instrumentation The nuclear operation of the PUR is monitored by four neutron sensitive channels '

(two of which are always on range) that indicate thermal power level over the entire operating range of the reactor. These channels initiate scram signals if preset neutron flux levels are reached. The bulk reactor coolant temperature is measured manually with a thermometer placed in the cool water. The instru-mentation and control systems are discussed in detail in Section 7.

4.5 Oynamic Design Evaluation The PUR is operated by manipulating control rods in response to changes in the ,

neutron flux (power) measured by the instrument channels. There are interlocks I to prevent inadvertent reactivity additions and a scram system to initiate a rapid shutdown (reactor scram) if a preset power limit has been reached. Addi-tionally, the measured temperature coefficient is negative over the operating temperature range. In the unlikely event of inadvertent high power operation  !

leading to high temperatures, this negative temperature coefficient of reactivity <

will tend to limit the reactor power.

t HUREG-1283 4-5 i

f

4.5.1 Excess Reactivity and Shutdown Margin Excess reactivity is defined as that value of reactivity that would occur if all control rods were completely removed from the reactor core. Reactivity is measured for a given core loading starting from a just-critical cold, clean core. A designated core loading may include irradiation facilities, such as the isotope production elements, or other facilities of such nature that they become a portion of the core when installed.

Excess reactivity must be built into the reactor core in order to compensate for a number of reactivity losses. Also, a sufficient reactivity must be avail-able to allow for an adequate reactor period for the PUR. This reactivity value has been determined to be not more than 0.6% ak/k and is the maximum allowed under any operating condition by the Technical Specifications.

The Technical Specifications require that the control rods provide a shutdown margin greater than 1.0% with the highest-worth control rod fully withdrawn and with the highest-worth experiment (0.4% ak/k for each secured experiment or 0.3% Ak/k for each movable or unsecured experiment) in its most reactive state under any conditions of operation. This is to provide assurance that the reactor can be shut down safely even if one control rod did not insert.

The current core configuration has an excess reactivity of 0.48% ak/k. The individual centrol rod worths are shown in Table 4.1. The total rod worth is l

7.66% ak/k. The shutdown margin for the current core configuration with the highest-worth rod fully withdrawn is 2.18% Ak/k.

Therefore, the current core configuration meets both the shutdown and the excess reactivity requirements. With all rods fully inserted, the core is subtritical by 7.18% ak/k.  !

4.5.2 Conclusion The Technical Specifications require that all three control rods be operable and the reactor can be brought to a subtritical condition even if the highest-worth control rod is totally removed from the core. These requirements ensure an adequate shutdown margin and provide sufficient redundancy in the unlikely event of a control assembly malfunction. Limiting the total excess reactivity of the core plus installed experiments to less than 0.6% ek/k allows for ade-quate reactor control under normal circumstances and prevents a prompt power excursion under any postulated abnormal circumstances.

On the basis of the above considerations, the staff concludes that excess reactivity will be limited sufficiently and adequate redundant shutdown

, capability is provided to ensure safe, controlled operation of the PUR. In addition, the 1.0% Ak/k shutdown margin with the highest-worth rod fully with-drawn and the highest-worth experiment in its most reactive position provides reasonable assurance that the ru ctor can be shut down adequately under all postulated operating conditions.

NUREG-1283 4-6 F

p.--~~--- ~ - -

-- . - -, ,, - - - - - , - - ,,.,c--. -

,, ,4 .

4.6 Functional Design of Reactivity Control Systems The power level of the PUR is controlled by three control rods: two shim-safety rods and one regulating rod. The two shim-safety rods can be scrammed, and solid borated stainless steel is used as the neutron absorber. The rods are connected to their drives by electromagnets. If electrical power to these electromagnets is interrupted for any reason, the armatures of the two shim-safety rods are released and they fall by gravity into the core. For the single regulating rod, hollow stainless steel is used as the neutron absorber.

This rod is mechanically attached to its drive and does not fall into the core during scram. The core locations of the three control rods are shown in Figure 4.2.

Each rod-drive system is energized from the control console through its own independent circuits. A manual scram at the control console is possible for the two shim-safety control rods, or they can be scrammed automatically by the safety circuits. Although the regulating rod cannot be scrammed, certain manual or automatic trips can cause the regulating rod to be driven downward into the core.

4.6.1 Control Rod-Drive Assemblies The tubular control rod-drive assemblies are mounted in a vertical position over the core. Motion is imparted to a control rod by a positive upward force on the extension rod, which is coupled to an electromagnet (except for the regulating rod), and to a screw mechanism, which is rotated through a fixed nut by the drive motor at the upper end of the tube. All rod drives are supplied with instant reversing induction motors. Downward motion results from the positive down-drive of the screw mechanism by the drive motor.

Limit switches are provided on the drive mechanism for the rod-drive units.

The following switches are provided: jam, up, 2/3 up, and down. In addition, the safety rods are provided with bottom and magnet-engage switches. One coarse and one fine position indicator located on the console indicate the position of each rod. These indicators are located on the console and provide readings to 0.01 cm.

4.6.2 Control Rod Circuitry and Interlocks Three identical control channels are used for the two shim-safety rods and the i regulating rod. A pushbutton switch selects an individual rod to be controlled. l All contro! rods can be inserted simultaneously into the core by a gang-lower j switch when shutdown is desired. This switch cannot cause the control rods '.o  ;

be gang raised under any circumstance. Control-console indicators for each rod include upper limit and lower limit. Additionally, engage (magnet coupled) and shim range (rod not bottomed) indications are provided for the shim-safety rods.

Rod positions are indicated on a coarse vertical scale and on a selectable digital readout device having a resolution of 0.01 cm. The true value of the rod position is known to 10.03 cm, which is equivalent to a maximum r tactivity uncertainty of less than 0.01% ok/k.

l NUREG-1283 4-7

Two types of automatic action are incorporated into the reactor safety system to correct abnormal or unplanned conditions: trip and rod insert. In a trip, the shim-safety rods are dropped by removing the current from the magnets. A rod insert (tet back) will cause all three rods to drive downward into the core.

Both actions are of the latching type, and manual reset of the safety system is required to return to the normal conditions. A complete description of the scram system with setpoints is contained in Section 7.

4.6.3 Conclusion The PUR is equipped with safety and control systems, control rods, rod drives, scram-logic circuitry, and interlocks that have performed reliably and satisfac-torily in the PUR for 25 years.

The control systems allow for an orderly approach to criticality and for safe shutdown of the reactor during normal and abnormal conditions. There is suf-ficient redundancy of control rods to ensure safe reactor shutdown, even if the most reactive rod fails to insert on receiving a scram signal. Interlocks pre-vent inadvertent rod withdrawal and, thus, inadvertent positive reactivity changes. A manual scram button allows the operator to initiate a scram indepen-dently for any conditions deemed to require a prompt shutdown.

On the basis of the above discussion, the staff concludes that the reactivity control systems of the PUR are designed adequately and will function to provide a reasonable assurance of safety.

4.7 Operational Procedures The PUR operates under Technical Specifications that direct the operation, audit, and surveillance of the reactor and provide procedural reviews for all safety-related activities. Written procedures have beer, established for safety-related and operational activities that include reactor startup, operation, and shutdown;

'v.r M,ame; and calibration of equipment and instrumentation. In addition, the retne is operated by trained NRC-licensed personnel in accordance with the xwnentioned procedures and Technical Specifications.

4.8 Conclusion The staff review of the PUR facility has included studying its specific design and installation, control and safety systems, and operational limitations, as identified in the Technical Specifications. The staff concludes that the PUR was designed and built according to good industrial practices. The staff further concludes that there is sufficient shutdown margin to ensure that the PUR can be adequatel conditions. y shut down under all anticipated normal and abnormal operating The design features of the PUR are similar to those of many pool-type research reactors operating in many countries of the world. On the basis of its review of the PUR and its experience with similar facilities, the staff concludes l that this reactor is capable of safe operation, as limited by its Technical Specifications.

NUREG-1283 4-8

. .. . - ~

5 REACTOR COOLANT AND ASSOCIATED SYSTEMS 5.1 Primary Cooling System The energy produced in the core is dissipated to the pool water as heat by the natural convection of the approximate 22,700 L of demin'iralized water in the reactor pool. The pool water is maintained at the ambient temperature of the environment by heat conduction to the ground and air and by some evaporation of water from the pool surface. To raise the temperature of the pool water 100 ,

l the reactor would have to operate continuously for more than 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> at full power (1 kW), assuming no heat loss. Thus, pool heatup poses no constraint on the anticipated operating schedule of the PUR (~13 kWh/yr).

5.2 Process Water System The process water system is assembled in one unit and contains a pump, filter, demineralizer, valves, flow meters, and a heat exchanger (see Figure 5.1). The heat-removal capacity of the heat exchanger is 10.5 kW. It was designed to maintain the reactor pool temperature at 75 F during continuous operation at 10 kW. The demineralizer contains a removable cartridge that is monitored continuously for radioactivity buildup. This system limits, by the use of filters and ion-exchange resin, the aluminum corrosion rate, corrosion product buildup, and neutron activation of impurities in the coolant.

5.3 Primary Coolant Makeup Water System Makeup water for the pool is taken batchwise from the Purdue University water line and is passed through the demineralizer enroute to the pool. A vacuum breaker excludes any possibility of siphoning pool water into the supply line.

The pool makeup water system, in addition to the demineralizer, also includes a normally closed manual shutoff and throttle valve and a check valve. '

5.4 Primary Coolant Chiller System Although the chiller is not needed for present operations, it remains available if required. Calculations indicate that the temperature rise rate while operat-ing the PUR at a power level of 1 kW would be 4.65 x 10 2 C/h, based on a mass of water equal to 1.85 x 104 kg. This takes no credit for heat loss to the surrounding sand and gravel or loss by evaporation. Experimentally, no temper-ature increase has been observed using the pool thermometer following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation at 1 kW. The chiller is designed with three loops to prevent the spread of radioactive contamination from the primary loop to the heat dump. '

The pool water passes through the primary loop, and a freon refrigerant is in i the secondary loop. Campus water in the third loop is used to remove the heat and is discharged into the campus sewer system. Radioactive contamination can-not pass through the three-loop system, unless at least two pipe failures were to occur simultaneously with significant abnormal radioactivity in the primary coolant, and the chiller system was in operation.

I NUREG-1283 5-1

I Float

'I" overfIev to

- Sever Drain r

~

' Conductivitya i h Lab.

Cell Water v 2 Check Valve

_ 8 $

Compact j y Water 1 Conditioner 2k[ Solenoid Valve Refrigeration - To Uni: _ Sever Drain

- llb., - >C Conductivity

^ ~~

Conductivity Cell B -

A I,

{ ~ E~ Float l g , -g Y' ' "

' Roto-

~"

Ntery E3 g .

g

i. 3 .N M a Lvn T'T x__.Pyep

\i d 4

, D,. , laac::r h'

N 7 :c' V 4 4  %

strainer (

t N

N Figuie 5.1 Reactor water process system HUREG-1283 5-2

5. 5 Conclusion The staff. concludes that the reactor coolant system at the PUR facility is of proper size, design, and condition and is maintained properly to ensure ade- '

quate cooling of the reactor at the power level specified in the PUR operating license. Also, the process water system can limit both corrosion and radio-activity problems associated with coolant contamination.

i e

f 9

1 I

i NUREG-1283 5-3 4

6 ENGINEERED SAFETY FEATURES Engineered safety features are those features or systems that mitigate the potential consequences of accidents. The only systems that could be considered as engineered safety features associated with the PUR facility are the ventila-tion system and the drain system. These systems are designed to limit the uncontrolled release of airborne and liquid radioactive materials under normal operating conditions as well as accident conditions.

6.1 Ventilation System The outside air supply and room exhaust are passed through high efficiency particulate air (HEPA) filters (see Figure 6.1). The reactor room is maintained at negative air pressure (minimum 0.13 cm of water). All doors to the reactor room have foam rubber seals. Steam heat is used to heat the room, and a room air conditioner circulates and cools the reactor room air. Under emergency conditions, the exhaust system and the air conditioner will be shut off and the sealed roos will prevent the rapid spread of contamination.

6.2 Drain System The only floor drain to the sewers is sealed except for a vent opening. This vent is raised about 1.2 m above the floor and has a filtered inverted opening.

Condensate from the air conditioner is released to this drain through an opening 3.66 m above the floot. During an emergency, the valve on the drain from the condensate holdup tank is shut off with the same switch that shuts off the-exhaust system. The condensate is held until it is tested for radiological contamination before it is released to the sewer.

6.3 Conclusion i

On the basis of the evaluation of potential accidents at the PUR facility that are discussed in Section 14 of this SER, the staff concludes that no significant I amounts of airborne radioactivity or liquid waste would be released into or out of the reactor room. Therefore, the available ventilation system is considered acceptable.

NUREG-1283 6-1

, s

t i

f

/ p uce A l s tha t 13 F alia m s t N f .

Blower on Reef

/ .-

pona Air in Caul Air out Air (buJ it limer

/ll

~~

1

)

llea t e r Outside Air In

- . ? e-

/r r_,ae ..eie

. J' 84.ilJ e.g toisk Ed pie.un Alr in f' '

Solenold Valve c.ait r ell e,I F rsws /

tcasole switcle ~ ' - ' - 1.-l-I -

One.deessat e g, g Segela Valve In S ewe r D e a le HLPA Filter NOTE: All floor drelas to the sewer are sealed. One

• Inverted went,1.2 e above the floor to prevent siphoning 15 equipped with a filter.

_x_ ..,.e Figure 6.1 Reactor room ventilation and cooling system l NUREG-1283 6-2

L 7 CONTROL AND INSTRUMENTA1ICS SYSTEMS The major components of the PUR control and instrumantation systems, including rod controls, annunciators, pen recorders, and meters, are located in the con-trol console. The cantrol console is designed to provide maximum visibility of Rhe instruments and accessibility to the controls and indicators. All indicators and controls necessary for startup and shutdown operations are located in one group in front of the operator.

l 7.1 Reacto- Control Sys g The reactor control system at the PUR facility, consisting of both nuclear ,

and process instrumentation, provides reactor control during normal operations and ensures safe shutdown in the event of abnormal operation (see Figure 1.1).

Interlocks are provided between the instrumentation system and the scram system to provide positive centrol of the reactor and essentially eliminate the c'ances of accident initiation.

i 2

7.1.1 Control Rod Prives '

Three identical an'i independent control channels are used for the hirt-safety  ;

and regulating rod-drive systems. A pushbuttnn switch selects an individual rod 'a be control ed. All control rods can be inserted simultaneously into the core by a gang lower switch when shutdown is desired. This switch cannot cause the control rods to be gang raised under any circumsta ce. A complete descrip-tion of the DUR control rod-drive systet.) is given in Section 4.

i 7.1.2 Servo Con'rol aystem j A servo control system provides automatic control once the reactor has reached the desired power level. The servo control system senses c'eviatic.95 from an adjustable setpoint on the channel no. 3 linear power recorder and adjusts the c6tition of the regulating rod to maini.ain the reactor at a con' tant power level. ,

Servo prmit circuitry actuates the console alarm buzzer if the reactor power deviates by more tnan 5% from the setpoint, indicating a malfunction of the )

system. A deviation meter is located on the console.

7.1.3 Neutrcn Source Drive i A motorized neutrop sourt:e drive is provided to raise the source through a travel of approximately L8 m to the "full out" position. The two-position ,

system i. operated by raise-lower switches at the console with limit switches  !

to indicate the source upper " ,'t or lower-limit resitions.

7.1.4 Fissio, Chamber Dr'  !

Controls for s s motor O w- c c:tannel no. 1 fission detector are alto located on the con'.' ( i ' se position indicator and a select-ab'.e fir:e position indiuto '

" ' , c'em is selected and coupled to the  !

drive s- tch in the saae e- . a , col rod drives. Indic& tor lights l NUREG-1263 7-1 1

l t

r s n s ee.

M R ea V .e A

. Se an.

m 3 4

aS a

. el z m.

. t a .

u Ma T

2 8 ,, O-  !)

. 3 s

-- . m 1 s

. - e

. , . .a - .

- - s e .a

. ._ w E

.s .

k a, m  ;

e L . .

6 T T s

T tS T

a F

e a

T s

1 y

T T y T 1 ,

Hat Q. .! . ii L1 iI i1 + I ll

,l

. ~ ___ .

8

• e. . f . .

- . g t

. '.=

.. n a

s m

e ks i.c u ** .,

. , R a

. > _ . I e

s. . c

- t. p '

• - t

, . . M t. aA

. . 8 tP t , - p s

- ,T 7 r 8 T FC .. 8( .s 7 t r lo .

s r - f. y

=

A! 7 g . ;i I l1 lj ll

~

i]

!L 1 lI ,I T

l s

o r

J t n

o

. c

- . r

~

o

. t

~ c a

4  ! e

- - R 1

II - 7 e

r

-. . u g

- 4 o

i F

s u

r I1 I1

!i l

u r w A

P a

m A

y

a * .

r. t s

o L

=

aa i t

a

! IL Ii e I n _

a r ti L

T F h w Ii W

E L

I m - S i

S a

~ w t

-,. E N

A i .

L R

~  :

c. - a

. .o

. .c L

E

- o t c 3

a.

N N . n L D N .- N E A

" a , N 3 C

N

( c N

II - s N A

E c I F

S 2c- = O. =00 Ne d

t ' .

_ j 1,

note the upper- and lower-limit positions; however, the detector may be placed at any position within its range.

7.1.5 Annunciator and Alarm Systems ,

When a system trip occurs, or when other abnormal system conditions' are sensed, an alarm (buzzer) sounds and an illuminated indicator is lighted on the control console indicating the source of the trouble. An annunciator acknowledge button may be used to reset the buzzer. ..

7.2- Reactor Instrumentation The function of the reactor instrumentation is to provide adequate information l for the operator and to generate signals to control the reactor or initiate  ;

trips. The nuclear instrumentation consists of a fission chamber, a compensated ion chamber, and two uncompensated ion chambers. All neutron detectors are arranged near the reactor core to ensure high sentitivity and to facilitate repair, maintenance, and repositioning. The detectors are in watertight aluminum  ;

tubes. The fission chamber is provided with a motor-driven positioning mechanism l and position-indication system; the other detectors are manually adjustable. l 7.2.1 Channel No. 1 - Startup Channel  !

The startup channel is used to monitor the neutron flux. It consists of a movable fission chamber, preamplifier, pulse nmplifier, scaler for accurate counting, log count rate and period amplifier, and count rate recorder, and j shares a pened recorder with chaanel no. 2. The ran from 1 to 105 counts /s (about 1.5 10 5 to1.5x100eofthisequipmentis W) with periods from i -30 to +3 s.

7.2.2 Channel No. 2 - Log N and Period Channel The log N channel indicates the reactor power level over the range from 0.0001 i to 300% power level (10 3 W to 3 kW). The detector for this channel is a com-pensated ionization chamber followed by a log N amplifier plus period instrumen-tation with outputs to the log N recorder and to the period recorder shared with channel no. 1.

7.2.3 Channel No. 3 - Linear Power l The linear level channel is ca able of measuring neutron flux in a reactor operating range from about 10- W (shutdown) to >100 kW. The sensing element is a BF3 ionization chamber coupled to a micro-microammeter.

i 7.2.4 Channel No. 4 - Safety Channel The safety channel utilizes a BFy ionization chamber and feeds directly into i the safety amplifiers. The sensitive range of this instrument is from a few '

percent to at least 150% of licensed power, linearly. Its output is indicated on the instrument chassis (instru,nent panel). The purpose of this channel is solely to provide reactor trip. ,

l NUREG-1283 7-3

7.2.5 Temperature and Water Monitor Channels Water temperature in the PUR pool is measured by a thermometer suspended by a string in the pool. Water level is measured by a scale immersed in the pool.

Water conductivity is displayed on the console by two meters registering the output of two conductivity cells that measure the pool water before and after it passes through the demineralizer.

7.2.6 Radiation Monitoring Instruments The radiation monitoring system consists of three fixed position remote area monitors (RAMS) and one continuous air monitor (CAM). All of the RAM alarms initiate a reactor scram. The CAM detects airborra particulates and alarms. A continuous sample is drawn from the reactor room through the CAM filter, which is checked semimonthly for gross beta gamma activity.

I 7.3 Scram System and Interlocks l Three types of action are incorporated into the control system to correct for abnormal reactor conditions:

fast scram - initiated by short reactor period on channel no. 2 or high <

flux on channel no. 4 - interrupts the current to the control rod magnets electronically )

1 slow scram - initiated by short reactor period on channel no.1 or no. 2 '

or by high flux on channel no. 2 or no. 3 - causes the rods to drop by removing power to the magnet power supply by means of a relay rod insert (setback) - initiated by short reactor period on channel no. 1 or no. 2 or by high flux on channel no. 3 or no. 4 - causes all three rods to be driven downward into the core All three actions are of the latching type ar.d require manual reset before return to the normal operating condition.

In addition to scrams initiated by high flux and reactor period, reactor scram  ;

or rod insert at PUR is initiated by any of the following signals:

- manual pushbutton compensated ion chamber (CIC) power supply failures safety amplifier trouble ,

area high radiation 7.4 Conclusion The control and instrumentation systems at the PUR are designed to provide i reliability and flexibility. There is adequate redundency and diversity in the ,

nuclear flux (power) monitoring circuits. In particular, nuclear power measure-  ;

ments are overlapped in the ranges of the startup, log-N, linear power, and i percent power level (safety) channels. In view of the simple nature of the [

open pool, the temperature and water monitor instrumentation is considered ade-quate. On the basis of the above information, the staff concludes that tne j 4

NUREG-1283 7-4

control and instrumentation systems at the PUR facility comply with the require- ,

ments and performance objectives of the Technical Specifications and applicable regulations and are acceptable to ensure safe operation and shutdown of the reactor.  ;

I 4

e b

i i

l l

T b

I W

i I

e I

1 i F l

j HUREG-1283 7-5 e - - - ,, . , - - - - - ,,

8 ELECTRIC POWER 8.1 Electrical Power System The electrical power for building' lighting and reactor instrumentation is single phase, 60 Hz, 120/240 V, which is furnished through a transformer and several control panels located throughout the building.

8.2 Emergency Power The reactor will scram in the case of an electrical power interruption because the control rods are supported by electromagnets. Because the decay heat generated in the core following a scram is not enough to cause fuel damage (see Section 14), emergency power is not required to maintain the reactor in a safe shutdown condition. Power for the facility intrusion detectors is supplied by a 12-V battery that is checked monthly and replaced biannually. If an electri-cal outage should occur, this battery would supply the necessary power for these instruments for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Battery powered emergency lighting is also available to facilitate personnel movement during a power outage. If a power outage should occur, no radiation monitors would be operating except for the portable, hand-held battery-powered type.

8.3 Conclusion The staff concludes that the design of the electrical power system, coupled with the fact that the reactor will scram in the event of a power failure, is acceptable for continued operation of the PUR. ,

l l

i i

i NUREG-1283 8-1 l

-=

l

9 AUXILIARY SYSTEMS The auxiliary systems considered are the ventilation system, the fire protec-tion system, the fuel storage system, the heating and air conditioning system, and the crane system.

9.1 Ventilation System The ventilation system is considered to be an engineered safety feature and is discussed in Section 6 of this report.

9.2 Fire Protection System The PUR and the building where the reactor is locat9d art intrinsically fire-proof, and in the event of a fire no special precautions are required. The fire protection system for the reactor facility is typical of those at most university low power research reactors. There are two portable fire extin-guishers in the reactor room located at either end of the room. If a fire should occur, the reactor would be shut down and the supervisor, or alternate, would be notified. Normal fire procedures for the building are in place and are expected to preclude accumulation of flammable materials. A fire station is located on campus 01/2 r.;ile from the reactor) and is available on short notice to assist in case of a fire.

9.3 Fut1 Storage System The only fuel at the PUR facility is the existing coro. The fuel is handled outside the core configuration only for experimental or inspection purposes.

During soise experiments, fuel elements are stored in one of the two fuel racks lo'cated inside the vessel on the bottom. Poison (boral steel plates) and geometry are used in these racks to ensure subcriticality. An annual fuel inspection involves removal of only one element outside the vessel at a time.

9.4 Heating and A,ir Conditioning System The PUR facility heating and air conditioning system is integral to the ventila-tion system discussed in Section 6 of this report. The air-operated dampers in this system are supplied pressurized air from the university system. The actu-ators are designed to close the dampers if a loss of Sir should occur.

~

9.5 Crane System The PUR facility has a 2-ton crane that runs along one axis over the center of  !

the reactor tank and is normally positioned at the end of the track. This  :

crane is used only occasionally under the supervision of the reactor supervisor for installing special experiments, and is maintained in good working order.

NUREG-1283 0-1 1

1

i 9.6 Conclusion >

1 l The staff concludes that the auxiliary systems at the PUR facility are designed, l operated, and maintained adequately and are capable of performing their intended l

functions of helping to ensure the safe operation of the facility.

9 b

l i

l l

l l

i I

[

l i I

l l

NUREG-1283 9-2 l

l

10 EXPERIMENTAL PROGRAMS The PUR is used in support of educational programs in physical, biological, and pharmaceutical sciences. The reactor also is a source of ionizing radiation and neutrons used for various research programs.

10.1 Experimental Facilities 10.1.1 Reflector TLbes Six positions in the graphite-reflector grid along one side of the core are utilized for special isotope production elements (see Figure 4.2). These ele- ,

ments are identical to the graphite-reflector elements except for central ac-cess holes to accommodate samples up to 3.8 cm in diameter and up to 61 cm long. Currently, the sample capsules loaded into these locations are 7.6 cm long with an outside diameter of 2.5 cm and are fabricated of aluminum. Ex-periments located in these tubes are considered secured in that they are not moved during reactor operation.

10.1.2 Drop Tubes Three drop tubes are currently utilized at the PUR. Their positions are shown

. in Figure 4.2. All of these tubes extend from the level of the activo core to a height above the pool water level that is sufficient to prevent inadvertent flooding. Reactivity addition, should accidental flooding occur, is discussed in Section 14. The inside diameters of the tubes are 1.3, 7.6, and 12.7 cm.

The 7.6-cm tube is fabricated of polyvinyl chloride; the other two are fabri-cated of stainless steel. Experiments contained in these tubes are properly secured to a suitable tether and then lowered into the tube. Because they may be moved during reactor operation, they are classed as nonsecured experiments.

10.2 Experiment Review i

Specific procedures to allow placement, operation, and removal of all experi-ments proposed for the PUR are reviewed and approved "oy the Committee on Reactor Operation to establish if they fall within an envelope of previously accepted reactivity addition and radiological consequences.

4 1 10.3 Conclusion i The staff concludes that the design of the experimental facilities, combined with the review and administrative procedures applied to all research activi-ties, is adequate to ensure that experiments are not likely to fail, are un-likely to release significant radioactivity to the environment, and are un-likely te damage the reactor systems or the fuel. Operating experience pro-vides further assurance that the experimental program at the PUR facility will i be conducted safely in the future. Therefore, the staff concludes that rea-i senable provisions have been made so the experimental programs and facilities  !

do not pose a significant risk to the reactor core or risk of radiation expo-sure to the public. )

NUREG-1283 10-1

i I

11 RADI0 ACTIVE WASTE MANAGEMENT The PUR produces essentially no rr.dioactive waste during normal operation because of the low power level and limited operating schedule.

l 11.1 ALARA Commitment i l  :

The university's commitment to the ALARA (as low as is reasonably achievable) i principle was established by the Radiological Control Committee in 1951 and was ,

recently (November 19, 1986) restated by the President of Purdue University. [

Purdue University is committed to a policy of making every reasonable effort to  !

keep radiation exposures as far below the specified regulatory limits as is l reasonably achievable. Thus, the underlying philosophy of the radiological g control operations of the university will be to maintain radiation exposures as low as is reasonably achievable, which is in keeping with the recommendations ,

of the National Council on Radiation Protection and Measurements, the National l Academy of Sciences-National Research Council, and other independent scientific organizations. The principle of ALARA is also codified as part of the NRC reg-  ;

ulations in 10 CFR 20.1(c), which state that licensees should "make every rea-sonable effort to maintain radiation exposures, and releases of radioactive ma-terials in effluents to unrestricted areas, as far below the limits specified ,

in the part as practicable."

j 11.2 Waste Generation ar.d Handling Procedures  ;

i 11.2.1 Solid Waste

{ The disposal of high-level radioactive waste in the form of spent fuel is not  !

anticipated for the term of the license. Low-level radicactive solid waste i generated at the facility consists of potentially contaminated paper and gloves and solid samples produced for experiments and is usually less than 1 ft8 /yr. ,

This waste is collected in specially marked containers and is disposed of under the university's By-Product License 13-02812-04. The water process system uses [

demineralizer resins to remove impurities from the primary coolant and collect any rauioactive ions in the water. These resins are continuously monitored and  ;

are periodically replaced. To date, no radioactive materials have been detected before shipment.

11.2.2 Liquid Waste v I

No radioactive liquid wastes except those that might be produced from stedent '

l or faculty experiments are generated as a result of normal reactor operations.

The pool water is analyzed periodically for radioactivity. Any detected short-  :

lived activity is allowed to decay and would not constitute disposable liquid

radioactive waste. Any wastes generated by research activities are disposed of

under the university's By-Product License.

t I 4 NUREG-1283 11-1 ,

r a F

^

11.2.3 Airborne Waste The airborne waste that could be present at the PUR facility would be composed of argon-41, tritium, nitrogen-16, and activated dust particles.

Argon-41 is produced by thermal neutron activatt;n of argon-40 in the air dis-solved in the pool water. No detectable traces os argon-41 from air dissolved-in the water have been observed or are expected at the PUR facility because of its low operating power levels.

The most likely source of tritium (H-3) is the pool water. From monthly water samples, the level of tritiuc has been found to be very low (much less than the most restrictive maximum permissible concentration).

The principal potential source of nitrogen-16 is from the fast neutron inter-action with oxygen in the pool water. The nitrogen must then diffuse to the surface of the pool before it is released to the atmosphere. In normal opera-tion, currents that might be established in the reactor pool would be very small and with the short half-life (7.14 seconds), most of the nitrogen-16 would decay before it reached the surface. No nitrogen-16 has been detected in the reactor room.

The reactor room air is sampled continuously for particulates by a CAM. Filters are changed and analyzed semimonthly for gross alpha and beta activity using a windowless flow proportional counter.

, 11.3 Conclusion The staff has reviewed the operational history of the PUR and concludes that no significant wastes are generated as a result of its normal operation. How-ever, should any significant waste ever be generated, acceptable provisions for the radiuactive waste management activities at the facility have been adopted and are expected to be continued to ensure compliance with 10 CFR 20 and the ALARA principle, i

i f

i i

NUREG-1283 11-2

)

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

12 RADIATION PROTECT.3N PROGRAM

! Purdue University has a structured radiatiun safety program. Policies for the program are determined by the Radiological Control Committee established by the President of the university (See Figure 12.1). The program is administered by the Radiological Control Of ficer and staff. The staff is equipped with radiation

> detection instrumentation to determine, control, and document occupational radiation exposures at the reactor facility and all laboratories using radio-isotopes at the university under By-Product License 13-02812-04. Routine surveys of the reactor room include analysis of the reactor pool water and reactor room air.

PRESIDENT PuRouE UNIVERSITY ll________________

DEAN RADIOLOGICAL CONTROL SCHOOLS OF ENGINEERlHG j COMMITTEE

- i ,

g I 8

l  :

l y  :

IIEAD 8 RADIOLOGICAL I

SciiOOL OF NUCLEAR ENGINEERING CONTROL 0FFICER j I  !

COMMITTEE ON i >

> I 8

REACTOR OPERATIONS 3 8

. i f i REACTOR l

> 4 ___________

SUPERVISOR i

i si i

REACTOR OPERATIONS Primarily AJainistration !

~ ~ ~ - - Pri mrily Saluty Figure 12.1 Organizational structure for PUR operations NUREG-1283 12-1

I 12.1 ALARA Commitment ,

The university is committed to the ALARA principle, and the Office of Radio-  !

logical and Chemical Controi makes every effort to keep doses o.1 low as is reasonably achievable (ALARA). All unanticipated or unusual exposures are  ;

investigated by the Radiological Control Committee and the operations staff to l

develop methods to prevent recurrences.  ;

12.2 H 1 alth Physics Program i At present, the university has a full-time health physics staff consisting of a a Radiation Safety Officer, Assistant Radiation Safety Officer, two health phy-sicists, environmental waste technician, and appropriate secretarial support. l The health physics staff performs all routine surveys and is available for consultation on all matters concerning radiation safety. -

a There are r.o documented procedures to ensure consultation on matters concerning radiation safety as opposed to ensuring availability of the health physics staff.

However, specific reactor operating procedures require that radiological control personnel be present when specified operations are performed. Because the Radiological Control Officer is an official member of the Committee on Reactor  !

Operations, this ensures the relationship is more than casual (see Figure 12.1).

12.2.1 Procedures Written procedures have been prepared that address routine health physics

monitori'g at the PUR facility. These procedures identify the interactions

J between the operations and health physics personnel and the administrative I

limits to control expuure. Copies of these procedures are available to the i operations and research staffs and administrative personnel.  ;

12.2.2 Instrumentation l l The university has a variety of detecting and measuring instruments for monitor-ing potentially hazardous ionizing radiation. Instrument calibration procedures and techniques are available to ensure that any credibl.e type of radiation and any significant intensities will be detected promptly and measured correctly.

12.2.3 Training All reactor-related personnel are required to attend a radiatiin safety training session before they begin work at the reactor. Additional training to illustrate i the ALARA principle and methods to minimize exposures is provided to those per-sonnel working directly with radioactivt materials. Retraining for reactor operators in radiation safety is also provided periodically.

l 12.3 Radiation Sources f 12.3.1 Reactor l

l The reactor core is the primary source of radiation directly related to reactor

! operations. Radiation exposure rates from the reacter core are reduced to

! acceptable levels by the water in the pool and concrete shielding.

}

! NUREG-1283 12-2 I

1 ___ _ _ _ _ _ - - - - ,_ _ - _ - _ . _ _ _ - _ . - _ , -_ -_

12.3.2 Extraneous Sources Sources of radiation associated with reactor use include radioactive isotopes produced for research, activated components of experiments, and activated samples. Personnel exposure from these sources is strictly controlled by fully developed procedures that follow normal health physics principles.

12.4 Routine Monitorina 12.4.1 Fixed-Position Monitors Three fixed position remote radiation area monitors (RAMS) with adjustable alarm setpoints and one continuous air monitor (CAM) are located in the reactor room.

The CAM air filters are changed and analyzed semimonthly.

The RAMS are set to alarm at 7.5 mR/h. This setpoint was calculated using the 10 CFR 20 limit of 3 R per quarter maximum whole-body dose and assuming 10 weeks per quarter and 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per work week. Instruments are calibrated semiannually by radiological control personnel. The RAMS are calibrated for exposure rate, and the CAM is calibrated for detection efficiency.

12.4.2 Wipe Tests Wipe tests of exposed surfaces of the reactor room are made monthly. Water samples are taken and counted monthly. All samples and material removed from the reactor are checked for levels of activity, and wipe tests are made for loose contamination.

12.5 Occupational Radiation Exposures 12.5.1 Personnel Monitoring Program Film badges and thermoluminescent dosimeter (TLD) finger rings are assigned to all approved reactor personnel. In addition, self-reading pocket dosimeters and dose rate instruments are used to administratively keep occupational expo-sures below the regulatory limits in 10 CFR 20. Students and visitors are provided with self-reading pocket dosimeters.

12.5.2 Personnel Exposures Exposures to reactor personnel are monitored with film badges and TLD finger rings, which are read quarterly. If doses are >100 mrem, a record is made of how and why the dose was received. Exposures are generally much less than 100 mrem, except during the annual fuel plate inspection, when a finger ring dose of >100 mrem may be experienced by the plate inspector. Because the reac-tor personnel and fast breeder blanket facility (FBdF) personnel are the same and use one personnel dosimeter system, it is difficult to determine how much of the "facility" dose is a result of the reactor operations. However, this is not considered necessary because the actual dosages are so low. A summary of the latest 5 years of whole-body exposure to reactor personnel is provided in Table 12.1.

NUREG-1283 12-3

'able 12.1 History of personnel radiation exposure at PUR facility Number of Individuals Whole-body -

exposure (rem) 1981 1982 1983 1984 1985 3

<0.1 8 9 6 8 7

>0.1 0 0 0 0 0 12.6 Effluent Monitoring 12.6.1 Airborne Effluents Potential celeases of airborne effluents are monitored with a CAM in the reac-tor room.

The CAM has never detected an unexplainable level of activity from either the accumulation of particulates or immersion exposure (i.e., immersion gas).

Analysis of the CAM filter generally indicates a concentration less than 3 x 10 16 pCi/cc for particulates, which is well below the most rostrictive maximum permissible concentration of 6 x 10 13 pCi/cc for an unknown alpha emitter.

Airborne waste is discussed in Section 11.2.3 of this report. .

12.6.2 Liquid Effluents The reactor generates no detectable radioactive liquid waste during normal operation. To prevent any release of potentially contaminated water to the i sewer system, sampler arn collected and analyzed by standard techniques.

Liquid waste is discussed in Section 11.2.2 of this report.

12.7 Environmental Monitorino l Under the environmental moni'oring program, the reactor pool water is sampled monthly, reactor room CAM air semples are analyzed semimonthly, one TLD has been placed in the reactor room, and one TLD has been placed in a classroom j above the reactor 'om. Reactor pool water samples are analyzed for gross gamma, gross alphe. and gross beta activity and for H-3. Reactor room air samples are analyzed for gross alpha and gross beta activity. Results indi-cate nothing beyond natural background has been detected in the reactor room air and reactor pool water samples. Typical exposures have ranged from those that are minimally detectable to 30 mrem.

12.8 Potential Dose Assessments Natural background radiation levels in the West Lafayette area result in an average exposure of about 100 mrem /yr. The maximum potential dose outside the NUREG-1283 12-4 1

4

reactor room is less than 1 mrem /yr based on film badge data foi the classroom above the reactor room; therefore, there is no significant contribution to the background radiation in unrestricted areas.

12.9 Conclusion The staff concludes that the radiation protection program receives appropriate support from the Purdue University administration. The staff further concludes that (1) the program staff is adequate and is equipped properly, (2) the reac-tor radiation safety-related staff has adequate autnority and lines of communi-cation, (3) the procedures are integrated correctly into the research plans, and (4) surveys verify that operations and procedures achieve ALARA principles.

Additionally, tne staff concludes that the PUR radiation protection program is <

acceptable because there have been no instances of reactor-related exposures of personnel above applicable guideline values and no significant releases of radioactivity to the environment have been identified. There is reasonable assurance that the personnel and procedures will continue to protect the health and safety of the public during routine or off normal reactor operations.

L NUREG-1283 12-5

)

a 1

l l 13 CONDUCT OF OPERATIONS i

13.1 Overall Organization Responsibility for the safe operation of the reactor facility is vested within the chain of command shown in Figt.re 12.1. The Head of the School of Nuclear i j Engineering is delegated responsibility, on behalf of the licensee, for overall  !

i facility operation, i 13.2 Traininj i I

Most of the training of reactor operators is done by in-house personnel. The j licensee's Operator Requalification Program was revised in February 1988 in (

conjunction with this license renewal application, and the staff concludes that ,

it meets the applicable regulations [10 CFR 50.54 (1-1) and 10 CFR 55] and is [

j consistent with the guidance of ANS 15.4. ,

13.3 Operation 11 Review and Audits  !

The Committee on Reactor Operations (CORO) provides independent review and audit l of facility activities. The Technical Specifications outline the qualifications and provide that alternate members may be appointed by the Chairman. The CORO >

? must review and approve plans for modifications to the reactor, new experiments,  ;

and proposed changes to the license or procedures. The CORO also is responsible  !

) for conducting audits of reactor facility operations and management and for

! reporting the results thereof to the university administration.

13.4 Emergency Planning '

l 10 CFR 50.54(q) and (r) require that a licensee authorized to possess and/or l l operate a research reactor shall follow and maintsin in effect an emergency ,

plan that meets the requirements of Appendix E to 10 CFR 50. An applicable i

Emergency Plan was submitted by the licensee on May 10, 1984, and approved by l the NRC on November 7, 1984.

l 13.5 Physical Security Plan f

The PUR facility has established and maintains a program to protect the reactor I and its fuel and to ensure its security. The NRC staff has reviewed the revised  !

Physical Security Plan and concludes that the plan meets the requirements of [

10 CFR 73.67 for special nuclear material of moderate strategic significance.  !

The PUR facility's inventory of special nuclear material for reactor operation i falls within that category.

i Both the revised Physical Security Plan, Revision 3, dated May 15, 1987, and the {

staff's evaluation are withheld from public disclosure under 10 CFR 2.790(d)(1). (

The amendment renewing facility Operating License R-87 incorporates this Physical  :

Security Plan as a condition of the license, j!

l i

NUREG-1283 13-1 i

f

i 13.6 Conclusion On the basis of the above, the staff concludes that the licensee has sufficient experience, management structure, and procedures to provide reasonable assurance  :

that the reactor will be managed in a way that will cause no significant risk - ,

, to the health and safety of the public.

i 4- ,

1 a

h 4

I i  !

4 l

1 1

i

} '

I i

1 i

4 l

I i

l NUREG-1283 13-2 1

I i

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

t 14 ACCIDENT ANALYSES The consequences of potential accidents in the PUR are limited by the low power level (1 kW) at which the reactor is operated. The low power levels and low use of the PUR result in a very small accumulation of fission products during normal operations and a correspondingly low level of decay heat and radioactiv-ity stored in the fuel elements. Therefore, some accidents normally postulated for nonpower reactors, such as loss of tank water and handling of irradiated fuel, do not constitute a major hazard in the PUR. To pose a significant haz-ard, an accident must generate and release a significant amount of fission  :

products.

The licensee and the staff both evaluated the potential consequences resulting from (1) a fuel-element-handling accident, (2) maximum reactivity insertions, (3) reactivity insertions from experiments, (4) a loss-of-coolant accident, and (5) failure of a fueled experiment.

14.1 Fuel-Element-Handling Accident Fuel-element maneuvers are always conducted under water in the reactor pool.

The fuel elements are removed from the core and moved into the storage space, 3 one at a time, using a hand-held fuel-handling tool. Normally, fuel is not removed from the pool except for an annual inspection of a fuel element. A fuel element weighs about 3.18 kg (7.0 lb) in air and only about 2.0 kg (4.4 lb) in water, 7

t

, 14.1.1 Scenario '

Three potential fuel-handling cases are considered. Case A assumes the element '

l is dropped on top of the core; case B assumes the element drops back into the core position as it is being raised; and case C assumes the element falls flat ,

on top of the core, ,

i 14.1.2 Technical Assessment For case A, if a fuel element should fall from the handling tool c'uring its transfer under water, it is not heavy enough to cause any significant damage.

The most severe damage likely to occur would be some denting of the end fit-tings because the fuel element, since it is an elongated object, would tend to fall in water in a rather upright position. For case B, no damage to the ele-ment would be expected if the element fell back into its original position be- ,

cause of the small distance (60.1 cm) it would fall. Also, no reactivity con-cerns are present because the core would have had sufficient shutdown margin  !

present before the fuel element was removed. For case C, if the element were to fall flat on top of the core, less foice would be placed on the element than if it fell on the end; therefore, t.o amage would be expected.

t NUREG-1283 14-1 i

14.2 Maximum Reactivity Insertion This hypothetical accident begins with the step insertion of the maximum li-censed excess reactivity of 0.6% adk into the critical reactor operating at the maximum licensed power of 1.0 kW.

14.2.1 Scenario Two general cases are considered: Case A assumes that the ssfety contrc1 cir-cuitry is operating and is activated by the reactor power exceeding the 120%

(1.2-kW) power setpcint. Scram redundancy is provided under these conaitions, because the period would be about 1 second and the short period trip would also initiate a reactor scram. During the scram, it is assumed that the most reac-tive shim-safety rod is stuck and the second shim-safety rod drops into the reactor providing a shutdoon margin of 1.8% ak/k. It is assumed that all nega-tive reactivity provided by the shim-safety rod is added as a linear ramp in 1 second, the time specified in the Technical Specifications for the rods to be fully inserted.

Case B considers the worst failure of the safety system where both rods fail to scram and the reactor is controlled only by the negative temperature feedback of the moderator. In this case, the calculated value of the moderator coeffi-cient is used because it is the lesser of the two values (calculated vs. mea-sured) reported in Table 4.1 and thus ensures conservatism.

The results for case A show reactor power would rise to about 2.8 kW within 55 ms following the step insertion of 0.6% ak/k, primarily as a result of the promptjump. The 120% (1.2-kW) trip, however, would initiate scram within about 3 ms of the step insertion.

Between this 3-ms scram initiation and the 55-m peak power, the scram reactivity is insufficient to control the power rise; however, once the prompt jump to 2.8 kW is completed, the scram reat-tivity becomes controlling and the PUR is quickly shut down. The calculated results for case A are sur.marized in Table 14.1. As shown, the power is above its initial value for less than 0.5 second and the resultant energy release is about 0.5 kW-s. This results in a negligible temperature rise in the PUR fuel plates.

The results for case B, where no control rods are inserted and the negative moderator are presented temperature coefficient is considered as the only shutdown mechanism, in Table 14.2. In this case, the analysis shows that the reactor power rises over a period of about 3 minutes to a level of about 380 kW. At this time the 0.6% ak/k step reactivity insertion causing the power rise is completely quenched by the negative moderator temperature coefficient. At this time or before, it is reasonable to assume that another scram or operator-initiated scram occurs and the power level is quickly reduced below I kW.

These results are consistent vith a number of the excursion experiments per-formed at the BORAX and SPERT facilities. Some of the results of the SPERT-1 experiment using the 00-12/25 core are applicable to !.he analysis of the PUR because the fuel geometry and composition are very similar. Table 14.3 com-pares the characteristics of the two reactors. A serier of self-limiting power excursien tests was carried out in SPERT-1 using five core loadings. The i'1put variab b referred to in these erperiments was the reactor period, induced by a NUREG-1233 14-2 s

l l'

Table 14.1 Results of power transient analysis with ramp insertion of control rod (case A)

Parameter Value Initial conditions Power level (kW) 1 Temperature ( C) 20 Input values Reactivity (step) added (% Ak/k) 0.6 )

Scram time (s) 1.0 Scram reactivity ramp (% Ak/k/s) -2.40 Calculated results Maximem power level (kW) 2.85 Elapsed time to maximum power (ms) 55 Elapsed time while P>l kW (ms) 275 Total energy released while P>l kW (kW-s) 0.5 Table 14.2 Results of power transient analysis with no control rods (case B)

Parameter Value Initial conditions Power levol (kW) 1 Pool temperature (*C) 20 Reactor temperature (*C) 20.6 Moderator temperature coefficient (% ak/k/*C) -2.1 x 10 2 Input values Reactivity (step) added (% ak/k) 0.6 Calculated results Maximum power level (kW) 380 Elapsed time to maximum power (min) ~3 Total energy released in 3 min (MW-s) ~54 stepwise reactivity insertion. The results of the calculations of the PUR po$tulated accident are consistent with the observed results of the SPERT-1 expariments using long periods of about 1 to 3 seconds where absolutely no fuei damage was observed. In fact, the SPERT-1 experiments showed that the fuel could withstand trant,ients with periods as short as 14 ms with no copartent damage to the fuel.

NUREG-1283 14-3

I l Table 14.3 Comparison of important fuel .

i data for PUR and SPERT-1 l i

Item PUR SPERT-1 l Geometry Plate Plate  ;

2 Length (cm) 61 61 j Width (cm) 7.0 7. 6 i Thickness (cm) 0.15 0.15 i j Water gap (cm) 0.53 0.45 J

Fuel Material U-Al U-Al ,

Enrichment (%) 93 93 j j Thickness (mm) 0.51 0.51 ,

Cladding Material Al Al l

Thickness (mm) 0.51 0.51  ;

i t i  !

l Because the results in Table 14.2 are consistent with long period SPERT-1 tests i j where no fuel failure was observed and because the available excess reactivity

• 2 is much less than was added to the SPERT-1 during the short period tests (as i low as 14 ms with no fuel failure), it is concluded that even during the very ,

unlikely event of a safety system failure during the maximum credible reac- '

tivity accident, the fuel would not melt and no fission products would be  ;

i released.  ;

! 14.2.2 Technical Assessment The staff believes that the reactivity insertion accidents considered in the i Purdue SAR are representative of the most severe transients that can credibly i occur at the PUR. The staff has reviewed the licensee's accident assumptions and calculations and finds them conservative, reasonable, and acceptable. The staff, therefore, concludes that it is unlikely that a credible nuclear excur-

, sion in the PUR would lead to fuel melting or cladding failure and, conse-

quently, such a transient would not pose a significant hazard to the public, l

j 14.3 Flooding of an Irradiation Facility and Failure of a Movable Experiment l A sudden replacement of a voided (i.e., air-filled) space next to the core by I water, such as that resulting from the flooding of an experiment tube, would cause a stepwise reactivity insertion, its magnitude depending on the void

! volume being replaced and its position relative to the core. Experiments have l shown that flooding of the 12.7-cm irradiation tube located outside graphite

reflector element F6 with water adds 0.3% ak/k to the core.

i l Another identified mechanism for suddenly adding reactivity to the critical

core at the PUR is the failure of a movable experiment. Similarly, the maximum j stepwise reactivity addition is limited by the Technical Specifications to

0.3% ak/k.

l NUREG-1283 14-4 1

i

14.3.1 Technical Assessment  ;

It is shown in Section 14.2 that a sudden reactivity insertion of 0.6% Ak/k into a critical core of the PUR can be tolerated with a sufficient safety mar- i gin. Therefore. 0.3% ak/k would be enveloped by that analysis, 14.4 Loss-of-Coolant Accident The reactor pool is designed to prevent unintentional drainage. The pool is constructed of a stainless steel liner and Theset tankinrests a second steel tank on a concrete padwith the about interstitial region filled with sand.

4.6 m below the floor of the reactor room, which is in the basement of the building. The pool has no drains or coolant pipes that could open or break.

Therefore, a sudden loss of coolant is considered to be extremely unlikely.

Furthermore, if the pool drained instantaneously, while the reactor was oper-ating, the loss of water (moderator) would shut down the reactor. i 14.4.1 Scenario i

Any reasonably conceivable leakage of water from the reactor pool is expected to be rather slow. In such a case, the radiation area monitor mounted directly  !

above the core would detect any additional radiation coming from the core as a result of a decreasing pool water level. Because the pool water level is i t

checked during daily routine operations, any significant leakage would be de-tected before reactor startup for the day.

4 14.4.2 Technical Assessment Although extremely unlikely, if the core were to become immediately uncovered I following a 1-kW power run for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, heit transfer would occur by natural convection of ambient air. The decay power of the PUR immediately after shut-down from full power (1 kW) is about 65 W. The decay power rapidly decreases 7 and is about 35 W after 1 minute of decay and 0.87 W in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For this case, the amount of heat removed is proportional to the cladding temperature, i No significant temperature increase of the fuel would be expected because heat l transfer would occur first to the aluminum fuel assembly cans and then by con-vection to the air. Even assuming adiabatic conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (that is, '

no heat transferred from the fuel), the fuel temperature rise would only be 9.5C', which is not a significant temperature rise for the PUR fuel. ,

14.5 Maximum Hypothetical Accident In this scction, the failure of an experiment during which fissile material has  !

been irradiated in the reactor is analyzed to assess the hazard associated with this accident. For this analysis, it is assumed that a capsule containing t irradiated fissile material breaks and a portion of the fission product inven-tory becomes airborne. The consequences of the release are analyzed for both j the reactor staff and the general public. Because the potential impact of this  ;

postulated accident is greater than that for any other accident analyzed, the failure of a fueled experiment is designated as the maximum hypothetical .

accident at the PUR. [

i I

NUREG-1283 14-5 r

t

1 14.5.1 Scenario In this analysis the consequences of a failed experiment generating 1 W of fission power were studied. The capsule containing the experiment is assumed .

to break as it is removed from the reactor. The fission products expected to become airborne are the noble gases and elemental iodine. Other fission prod-  ;

ucts and actinides are not volatile at the fueled experiment temperature (which is essentially room temperature). All of the noble gases and 25% of the radio-  :

' iodine are assumed to be released, which is consistent with Regulatory  ;

Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Conse-  ;

quences of a Fuel Handling Accident," March 1972. No credit for the absorp- i

' tion of iodine in water is taken because of the designation of this event as the maximum hypothetical accident, and because it is postulated to occur in l air.

1

' A conservative assumption that the irradiation time was infinite was made in thic analysis. Therefore, the fission inventories used in the analysis for

' some long-lived radionuclides (e.g., krypton-85 or even iodine-131) are overly conservative. Furthermore, it was assumed that the fission products are in-stantaneousl air volume. y released and uniformly distributed in the 424-m3 reactor room 14.5.2 Technical Assessment The calculated saturation activity for each respective radioiosotope and its concentration in the reactor room after experiment failure is shown in Table 14,4 for a 1-W experiment. This experimental specimen power level cor-responds to the amount of fuel that could be allowed by the relevant Technical Specifications.

whole body, skin,Also and shown thyroid.in this table are the calculated dose rates for the Under these conditions any one of the RAMS i

would controlcause room. anFrom automatic reactor shutdown and audible and visual alarms in the past experience, the reactor building can be evacuated I within 1.5 minutes. Therefore, it is assumed that the exposure time to the reactor staff is 1.5 minutes, which results in radiation doses of: whole body - 17 n: rem, skin - 11 mrem, and committed thyroid - 830 mrem. These doses are <10% of the dose limits as stated in Regulatory Guide 2.2, "Development of Technical Specifications for Experiments in Research Reactors," November 1973,

, for doubly encapsulated experiments.

This radiation exposure is less than the limitt ;<10% of the equivalent annual dose stated in 10 CFR 20) established in the Technical Specifications, Sec-tion 3.5.f, for a singla encsosulated experiment. This experiment corresponds

! to the irradiation of 1.1 gm of U-235 in the midplane of the isotope irradia-1 tion tube located in position F6.

l

' For the radiation calculations outside the reactor building, it was assumed that all fission products released in tue reactor building would leak out within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Because the reactor room does not have any windows and only

, a few doors and emergency procedures call for turning off the air exhaust

~

1 system, the leak rate assumption is considered to be reasonable.

i

l l NUREG-1283 14-6 1

- - _, _ . - - _ _ ._, _ . _ _ , ~ __ _ . . - - . _ . - _

T ale 14.4 Dose rates in the reactor room from a failed fuel experleent (power = 1 W)

Oose rate Released Total  % E gamma E-beta DCF8 Gamma ** Betat Thyroid Isotop* (Cl) (C1/cm3 ) (MeV) (MeV) (rem /C1) (ares /h) (area /h) (rea/h) 1-131 2.66 E-02 1.57 E-11 3.71 E 01 1.97 E-01 1.00 E+06 3.49 E+00 2.55 E+00 1.96 E+01 1-132 3.75 E-02 2.21 E 11 2.40 E+00 4.48 E-01 6.60 E+03 5.00 E+01 8.16 E+00 1,82 E 01 1-133 5.31 E 02 3.10 E 11 4.77 E 01 4.23 E 01 1.80 E+05 1.41 E+01 1.09 E+01 7.04 E+00 1 134 5.94 E-02 3.30 E-11 1.94 E+00 4.55 E-01 1.10 E+03 6.41 E+01 1.31 E+01 4.82 E+00 1-135 4.69 E-02 2.77 E 11 1.78 Ed0 3.08 E-01 4.40 E+04 4.54 E+01 7.02 C+00 1.52 E+00 Kr-33a 5.90 E 03 1.39 E-11 2.50 E 03 1.03 E 02 -- -- 3.41 E-02 1.18 E 01 --

Kr-3de 1.27 E 02 3.00 E-11 1.51 E 01 2.23 E 01 -- 4.27 E+00 5.50 E+00 --

Kr-35 2.53 E 03 5.97 E-12 2.11 E 03 2.23 E 01 -- 1.19 E-02 1.10 E+00 -

Kr-37 2.00 E 02 4.72 E-11 1.37 E+00 1.05 E+00 6.09 E+01 4.08 E+01 --

Kr-38 3.12 E 02 7.36 E 11 1.74 E+00 3.41 E-01 -- --

1.21 E+02 2.07 E+01 --

Kr-39 3.96 E-02 9.34 E-11 1.60 E+00 1.33 E+00 1.41 E+02 1.02 E+02 --

Xe-131m 2.53 E-04 5.97 E-13 2.00 E-02 1.40 E-01 -- 1.13 E-02 6.38 E-02 --

Xe-133e 1.35 E-03 3.18 E-12 3.26 E 01 1.55 E-01 *- -- 9.79 F-01 4.07 E 01 --

l Xe-133 5.48 E-02 1.29 E-10 3.00 E-02 1.46 E 01 -- 3.88 E+00 1.55 E+01 --

Xe-135m 1.77 E-02 4.17 E-11 4.22 E-01 9.74 E 02 -- 1.66 E+01 3.35 E+00 --

Xe-135 5.23 E-02 1.23 E-10 2.46 E 01 3.22 E-01 -- 2.86 E+01 3.27 E+01 --

Ke-137 5.31 E 02 1.25 E-10 1.50 E-01 1.37 E+00 1.77 E+01 1. 41 E+02 --

Xe-138 5.57 E-02 1. 31 E-10 1.10 E+00 8.00 E 01 1.36 E+02 8.66 E+01 --

7.11 E+02 4.32 E+02 3.3 E+01

'DCF = dose conversion factor.

• • Whole body.

15 kin.

Other conservative assumptions in the analysis include

- no radioactive decay (no decrease in source strength)

- no building wake effects and horizontal plume meandering

- release at ground level versus out the 15.2-m exhaust stack

- average wind speed of 1 m/s versus actual average of 3.4 m/s The calculated dose rates at 100 m (distance assumed as a reasonable distance at which evacuation and control could be accomplished within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) for an experiment power of 1 W are as shown in Table 14.5. If it is assumed that an individual is located at this point for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the fission product release from a postulated experiment failure, then that individual's resulting radiation dose to the whole body would be 0.51 mrem and 20 mrem committed dose equivalent to the thyroid. These doses are less than 10% of the dose limits in Regulatory Guide 2.2 and well within the applicable limits in 10 CFR 20.

14.6 Conclusion The staff has reviewed the credible potential accidents for the PUR. On the basis of this review, none of the postulated accidents are expected to release significant fission products to the environment or excessive doses to the reac-tor personnel or to the public. Therefore, the staff concludes that there is reasonable assurance that any accident resulting from continueJ operation of the PUR would not pose a significant risk to the health and safety of the public.

NUREG-1283 14-7

Table 14.5 Dose rates at 100 m from a failed fuel experiment (power = 1 W)

Oose rate Gamma

• Beta ** Thyroid Isotope (mrem /h) (mrem /h) (rem /h)

I-131 1.95 E-03 9.06 E-04 I-132 6.98 E-03 1.78 E-02 2.91 E-03 6.49 E-05 I-133 5.01 E-03 3.88 E-03 2.51 E-03 I-134 2.28 E-02 4.67 E-03 I-135 1.71 E-05 1.65 E-02 2.50 E-03 5.41 E-04 Kr-83m 1.21 E-05 4.20 E-05 --

Kr-85m 1.52 E-03 1.96 E-03 --

Kr-85 4.23 E-06 3.90 E-04 --

Kr-87 2.17 E-02 1.45 E-02 --

Kr-88 4.30 E-02 7.36 E-03 --

Kr-89 5.02 E-02 3.64 E-02 --

Xe-131m 4.01 E-06 2.45 E-05 --

Xe-133m 3.48 E-04 1.45 E-04 --

Xe-133 1.30 E-03 5.54 E-03 --

Xe-135m 5.91 E-03 1.19 E-03 --

Xe-135 1.02 E-02 1.17 E-02 --

Xe-137 6.31 E-03 5.03 E-02 --

, Xe-138 4.85 E-02 3.08 E-02 --

Total 2.53 E-01 1.73 E-01 1.01 E-02

• Whole body.

• • Skin.

2 NUREG-1283 14-8

15 TECHNICAL SrECIFICATIONS The licensee's Technical Specifications, evaluated in this licensing action, define certain features, characteristics, and conditions governing the contin-ued operation of the PUR facility. These Technical Specifications will be ex-plicitly included in the renewal license as Appendix A. Formats and contents acceptable to the NRC have been used in the development of these Technical Specifications, and the staff has reviewed them using the ANSI /ANS 15.1-1982 standard as a guide.

On the basis of its review, the staff concludes that normal reactor operation within the limits of the Technical Specifications will not result in offsite radiation exposures in excess of 10 CFR 20 limits. Furthermore, the limiting conditions for operation, surveillance requirements, and engineered safety features will limit the likelihood of malfunctions and mitigate the conse-quences to the public of offnormal or accident events.

l NUREG-1283 15-1 I

._- - - - - _ - - - - -- - a

i 16 FINANCIAL QUALIFICATIONS The PUR is owned and operated by a State educational institu+ ion in support of its role in education and research. On the basis of financial information supplied by the licensee in its submittal of June 30, 1986, as supplemented, the staff concludes that funds will be made available, as necessary, to support continued operations and eventually to shut down the facility and maintain it in a condition that would constitute no risk to the public. The staff reviewed the licensee's financial status and found it acceptable in accordance with the requirements of 10 CFR 50.33(f).

NUREG-1283 16-1

1 17 OTHER LICENSE CONSIDERATIONS Prior Reactor Utilization In previous sections of this SER, the staff concluded that normal operation of

• he reactor causes insignificant risk of radiation exposure to the puolic, and that only an offnormal or accident event could cause some measurable exposure.

Even a maximum hypothetical accident would not lead to a dose to the most ex-posed individual greater than the applicable guideline values of 10 CFR 20.

The staff concluded in its SER for the original Operating License that the reactor was initially designed and constructed to operate safely. During its current review, the staff considered whether prior operation would cause sig-nificant degradation in the capability of components and systen.s to continue to perform their safety functions. Because fuel cladding is the component most responsible for preventing release of fission products to the environment, pos-sible mechanisms that could lead to detrimental changes in inteority were con-sidered. Prominent among the considerations were the following: (1) radia-tion degradation of cladding integrity, (2) high fuel temperature or tempera-ture cycling leading to changes in the mechanical properties of the cladding, (3) corrosion or erosion of the cladding leading to thinning or other weaken-ing, (4) mechanical damage resulting from handling or experimental use, and (5) degradation of safety components or systems.

The staff's conclusions regarding these parameters, in the order in which they were identified above, follow.

(1) Nearly identical fuel has been laboratory tested elsewhere and has been

! exposed under similar irradiation conditions to much higher total radia-tion doses in operating reactors, such as at the Oak Ridge Research Reac-tor and the Omega West Reactor (Los Alamos National Laboratory). No sig-nificant degradation of cladding has resulted in any of these reactors, c

(2) The power density, coolant flow rates, and maximum temperature reached in the PUR fuel are far below similar parameters in some other nonpower reac-tors using similar fuel. No damage has occurred during normal operations in any of these reactors.

(3) The coolant flow rate at PUR is much lower than that used at several higher powered research reactors using MTR-type fuel. No erosion problems have been observed. At the PUR facility, corrosion is kept to a reason-able minimum by careful control of the conductivity and pH of the primary coolant water.

(4) The fuel is handled as infrequently as possible, consistent with required surveillance. Any indications of possible dame.ge or degradation are investigated immediately, and damaged fuel weuld be removed from service in accordance with Technical Specifications. All experiments placed near the core are isolated from the fuel cladding by a water gap and at least one barrier or encapsulation.

NUREG-1283 17-1

~

(5) The PUR personnel perform regular preventive and corrective maintenance and replace components, as necessary. Nevertheless, there have been some malfunctions of equipment. However, the staff review indicates that most of these malfunctions have been random one-of-a-kind incidents, typical of even good quality electromechanical instrumentation. There is no indi-cation of significant degradation of the instrumentation, and the staff further has determined that the preventive maintenance program would lead to adequate identification and replacement before significant degradation occurred.

Therefore, the staff concludes that there has been no apparent significant degradation of safety equipment, and because there is strong evidence that any future degradation will lead to prompt remedial action by Purdue, there is reasonable assurance that there will be no significant increase of in the component likelihood of occurrence of a reactor accident as a result malfunction.

NUREG-1283 17-2

e y 18 CONCLUSIONS On the basis of its evaluation of the applicatian as set forth in the previous sections, the staff has determined that (1) The application filed by Purdue University for renewal of Operating License R-87 for its research reactor, dated June 30, 1986, as supple-mented, complies with the requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I.

(2) The facility will operate in conformity with the application as supple-mented, the provisions of the Act, and the rules and regulatuns of the Commission.

(3) There is reasonable assurance (a) that the activities authorited by the-operating license can be conducted without endangering the health and safety of the public and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in 10 CFR Chapter I.

(4) The licensee is technically and financially qualified to engage ir, the '

activities authorized by the license in accordance with the regulations of the Commission set forth in 10 CFR Chapter I.

(4) The renewal of this license will not be inimical to the common defense and security or to the health and safety of the public.

NUREG-1283 18-1 l

m i

i 19 BIBLIOGRAPHY 1 American Nuclear Society (ANS), 5.1, "Decay Heat Power in Light Water Reactors,"

1978.

-- ,15.1, "Standard for the Development of Technical Specifications for Re-search Reactors," 1982.

-- , 15.4, "Selection and Training of Personnel for Research Reactors," 1977.

-- , 15.11, "Radiological Control at Research Reactor Facilities," 1977.

-- , N401-1974/15.6, "Review of Experiments for Research Reactors," 1974.

Code of Federal Regulations, Title 10, "Energy" (including General Design Criteria), U.S. General Services Administration, Office of Federal Register, National Archives and Records Service, U.S. Government Printing Office, Washington, D.C., January 1987.

Dietrich, J. R., "Experimental Determinations of the Self-Regulation and Safety Operating Water-Moderated Reactors," A/ Conf.8/P/481, Argonne National Labo-ratory, Argonne, IL, June 30, 1955.

Forbes, S. G., et al, "Instability in the SPERT I Reactor, Preliminary Report,"

100-16309, Idaho Operations office, Atomic Energy Commission, Idaho Falls, ID, October 1956.

Miller, R. N., et al, "Report of the SPERT I Destructive Test Prcgram on an Aluminum, Plate-Type, Water-Moderated Reactor," 100-16883, Idaho Operations Office, Atomic Energy Commissinn, Idaho Falls, 10, June 1964.

Nyer, W. E., et al "Experimental Investigations of Reactor Transients,"

100-16285, Idaho Operations Office, Atomic Energy 'ommission, Idaho Falls ID, April 1956.

NUREG-1283 19-1

g ,,- . i % e im.cos.,o., e-. .o% , ..4.......,-... .,,

=5'# slCUOGRAPHIC DATA SHEET NUREG-1283 it I thiY.VCt,0%$ 0% ?.6 .Iyt.s.

J ,1?Li .%o Sv.fiike 3 it t. it.%.

Safety Evaluation Report related to the renewal of the operating license for the research reactor at Purdue University a o." a n'oa ' c*u u o W%,. ....

j

• *v t aca

• Apri1 1988

. o.ri.e.o.i.iseio W% , se i ..

April 1988

, n.,os..%c, o.s..os.i .o% %.. .%o .v %o .oo., a u,= <, c , . .. m c , . . wo. . . , . . .

Standardization c.nd Non-Power Reactor Project Directorate Office of Nuclear Reactor Regult. tion ' ^ oa ca .% '

• wn a U. S. Nuclear Regulatory ColTeission Washington, D. C. 20555

% x .so.,%oo.c.s4u.o .... o...t,%c.uc,. u,. e,c , i i . e . . . o. . .v. ,

Safety Evaluation Report

. . ..o0 co. . o ny . .

13 S 6 wi%f.. 40f.3 Pertains to Docket No. 50-182

,3 .it..ciax ,

This Safety Evaluation Reg >rt for the application filed by Purdue University for a renewal of Operating License R-87 to continue to operate a research reactor has been prepared by the Office of Nuclear Reator Regulation of the U. S. Nuclear Regulatory Cccmission. The facility is owned by Purdue University and is located on the campus in West Lafayette, Indiana. On the basis of its technical review, the staff concludes that the reactor facility can continue to be operated by the University without endangering the health and safety of the public or the envircnment.

i l

.. _ .,%,.,. ..- .... W .m...,<... .. ............

Non-Power Reactor Purdue University Unlimited License Renewal . . ir e s. ,,, e t e s.. .c a t ,o*

u..,,,

. on =,,. ...$ o.<

• e n e e o vi..s Unclassified

,r.. ,.-u Unclassified

,,....o,..w

,....c.

.v.s.cc4.sw nr u tst:6 car a ct me s. r::.reri e n.$

UNITED STATES

- NUCLEAR REGULATORY COMMISSION ,o's"rl'o*We'iY)I,o usmac WASHINGTON, D.C. 20555 ein wr =. o o OFFCAL BU$iNESS PENALTY FOR PRIVATE USE. $330 120555078877 1 1AN US NRC-OARr.ADM OIV 0F PUR SVCS POLICY F. PUB 4GT BR-PDR NUREG W-537 DC 70555 WASHINGTON Q

v 1

4 4

t

_