ML20199K462

SAR for Purdue Univ PUR-1 Reactor

ML20199K462
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Site:	Purdue University
Issue date:	06/30/1986
From:	Clikeman F, Stansberry E PURDUE UNIV., WEST LAFAYETTE, IN
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l SAFETY ANALYSIS REPORT for the PURDUE UNIVERSITY l

PUR - I REACTOR LICENSE NUMBER R-87 DOCKET NUMBER 50-182 Prepared by:

F.M. Clikeman E.R. Stansberry West Laf ayette, IN 47907 00 ADO 000 82 l

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C-SCHOOL OF NUCLEAR ENGINEERING L

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West Lafayette, Indiana 47907

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CM TEN'I5 1.

IN'IRODUCTION AND G ENERAL DES CRIPTION OF FACILITY..............

1-1 1.1 Introduction.............................................

1-1 1.2 General Facility Description.............................

1-1 1.3 Modifica tions to the Reactor Facility....................

1-2 2.

S ITE CH AR A CTERISTICS..........................................

2-1 2.1 Location.................................................

2-1 2.2 Demography...............................................

2-1 2.3 Geology and Seismology...................................

2-2 2.4 Hydrology................................................

2-4 2.5 Meteorology..............................................

2-5 2.6 References...............................................

2-6 3.

REACr0R.......................................................

3-1 3.1 Introduction.............................................

3-1 3.2 Reactor Building and Reactor Room........................

3-2 3.3 R e a c t o r Co r e.............................................

3-3 3.4 Reactor Poo1.............................................

3-5 3.5 R e a c t iv i ty Pa r am e t e r a....................................

3 -6 3.6 Water Process System.....................................

3-8 3.7 Rea c tor In strumentation and Contro1......................

3-8 4.

ENVIRNMENTAL IMPACT AND RADI0 ACTIVE WASTE MANAG EMENT.........

4-1 4.1 Construction.............................................

4-1 4.2 Environmental Ef f ects of Facility Operation..............

4-1 4.3 Enviro nm ent a l Lapa c t s o f Ac c ide nt s.......................

4-2 4.4 Alterna tiv es to Operation of the Facility................

4-2 4.5 Long-tern Effects, Costs and Benefits, and Alternatives.......

4-3 5.

RADIATION PROTECTION PR0G RAN..................................

5-1 5.1 ALARA Comm i tm ent.........................................

5-1 5.2 He a l th Phy s i c s Pr o g r am...................................

5-1 5.3 Radiation Sources........................................

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5.4 Ro ut i ne Mo ni t or i n g.......................................

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5.5 Occupa tional Ra diation Expo sur e s.........................

5-3 5.6 Effluent Monitoring......................................

5-4 5.7 Environmental Monitoring.................................

5-4 5.8 Po t e n t i a l Do s e A s s e s sm en t s...............................

5-5 6.

00N DU CI 0F O PER ATION S.........................................

6-1 6.1 Organization.............................................

6-1 6.2 Reactor Procedures......................................

6-2 6.3 Audits...................................................

6-2 l

6.4 Staff Training Program...................................

6-3 6.5 Emergency Planning.......................................

6-4 6.6 Physical Security P1an...................................

6-4 6.7 Records and Reports......................................

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6.8 Reporting Requirements...................................

6-5 7.

AC CI DENT AN AL YS ES.............................................

7-1 7.1 Fuel Element Handling Aeoident...........................

7-1 7.2 Flooding o f an Irradiation Facility......................

7-2 7.3 Loss of Coolant Accident.................................

7-2 7.4 Fa il ure of a Movabl e Experiment..........................

7-4 7.5 Maximum Re a c t iv ity In se rti on.............................

7-5 7.6 Failure of a Fueled Experiment...........................

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l Ligi gf Tables Page 2.1 Population Distribution Around Purdue University 2-8 2.2 Monthly Mean Climactic Elements for Period 1953-1970 2-9 2.3 Mean and Extremes for Period 1965-1974 2-10 2.4 Tornado Frequency, Tippecanoe County, 2-11 Indiana 1950-1985 6.1 Records and Logs Retained for Five Years 6-8 6.2 Records and Logs Retained for Life of Facility 6-9 7.1 Delayed Neutron Fraction 7-15 7.2 Results of the Power Transient Analysis 7-16 with Ramp Insertion of Control Rod 7.3 Results of the Power Transient Analysis 7-17 with No Control Rods 7.4 Comparison of Important Fuel Data 7-18 7.5 Dose Rates in the Reactor Room from Failed 7-19 Fuel Experiment 7.6 Dose Rates at 100 Meters 7-20 i

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L111 sf Finures Page 2.1 Purdue University Main Campus and the 2-12 Lafayette-West Lafayette Vacinity 2.2 Laf ayette-West Laf ayette and Tippecanoe County 2-13 2.3 State of Indiana 2-14 2.4 Seismic Zone Map of the United States from 2-15 the Unified Building Code,1985 edition 3.1 Nuclear Engineering Laboratory 3-15 Showirs the Reactor Room, B-70A 6.1 Organization Structure 6-10 7.1 Fission Product Decay Heat Power as a 7-21 Function of the Time Af ter Shut Down iV

4 1.

INTRODUCTION AND GENERAL. DESCRIPTION OF FACILITT 1.1 Introduction This report is submitted in support of the application for renewal of the operating license (R-87) for the Purdue University Reactor (PUR-I) for a period of 20 years.

The reactor is located in the Nuclear Laboratories in the Duncan Annor of the Electrical Engineering Building on the eastern edge of the campus in West Lafayette, Indiana. The Duncan Annex is of brick and concrete block construction and was originally built as a high voltage laboratory.

In 1962 the reactor was built in half of the existing high voltage laboratory, which was a high bay area.

Offices, classrooms, and laboratories had been built in the remainder of the original building.

t 1.2 General Facility Descrintion The PUR-I is a 1 kW pool type reactor, utilizing KIR type enriched fuel plates, which are graphite reflected, and light water moderated and cooled.

It was designed and built by Lockheed Nuclear Products of Lockheed Aircraft Corp., Marietta, Georgia.

The reactor is controlled by three blade-type control rods located in the core region of the reactor. There are two shin-saf ety rods made of solid borated stainless steel, utilizing a magnetic clutch between the blades and the lead screw operated drive mechanisms, and a regulating rod which is a screw operated direct drive and made of hollow 1_1

stainless steel. Each control blade is protected by an aluminum guido plate on each side within the fuel assembly.

Fuel movement is only by a fuel handling tool, which is stored securely to the aluminam superstructure, when not in use.

Security of the fuel handling tool is under administrative control of the licensed senior operators.

1.3 Modifications to the Reactor Facility Modifications to the reactor facility and either approval dates or completion dates are as follows:

May 1964 Amendment 1 - Permit 10 kW operation.

December 1965 Installation of pool traversing mechanism completed.

July 1966 Amendment 2 - License renewal.

October 1968 Change 1 - Install stainless steel liner.

September 1969 Installation of air conditioner.

l January 1972 Change 2 - Change pH of pool water.

I February 1974 Change 3 - Regeneration of domineralizer.

November 1978 Amendment 3 - Technical specifications.

i August 1980 Amendment 4 - Physical security plan.

February 1981 Installation of catwalk around air conditioner.

March 1981 Amendment 5 - Physical security plan.

September 1982 Amendment 6 - Technical specifications - revision 1.

October 1982 Amendment 7 - Technical specifications - revision 2.

April 1983 Amendment 8 - Technical specifications - revision 3.

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SITE CHARACTERISTICS 2.1 Location The PUR-I reactor is located in the Duncan Annez of the Electrical Engineering Building on the campus of Purdue University in West Laf ayette, Tippecanoe County, in the State of Indiana, as shown in Figure s 2.1, 2.2, and 2.3.

The Lafayette-West Lafayette area is about 60 miles northwest of Indianapolis, the State Capitol, and about 140 miles south-southeast of Chicago, Illinois. Generally, the land surface is flat to rolling, except where the Wabash River and it's tributaries have eroded deep valleys.

2.2 Denoaranhv The campus popuistion of student and staff during the period of highest usage for the year 1985-1986 was approximately 40,500 (32,000 +

8,500).

According to the 1980 census summary for Tippecanoe County, the total population was 121,702. As the Table of Popuistion Distribution i

(Table 2.1) shows, the popuistion living within a 10 mile radius is 117,212, with the zone of greatest concentration being within a radius i

of 1.25 miles of the reactor site.

Although nearly the same number of i

people reside in the area between the 1.25 alle and the 2.5 mile radius i

the increased area reduces the population density.

Beyond the 10 mile radius the popuistion is mainly rural and drops of f rapidly.

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2.3 Geolony and Seismolony 1

2.3.1 Gaolgir The county lies within the Tipton Till Plain of Indiana and is a section of the Till Pisins subprovince of the U.S. Central J

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.1 Glacial drift covers the bedrock to a depth ranging from a few feet to more than 300 feet. The underlying bedrock 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.2 Purdue University is located above an extensive glacial deposit

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of sand and gravel.

i The land surf aces of Tippecanoe County is flat to rolling, except j

where the major streams have cut deeply into the surf ace.

The entire county lies within the drainage basin of the Wabash River and its tributaries.3 The land slopes generally southwestward with the streams flowing westward. Two main tributaries, the Tippecanoe River and the Wild Cat Creek enter the Wabash upstress from the campus.

Minor tributaries include Little Pine Creek, Indian Creek, Burnetts Creek, Nott's Creek, Sugar Creek, Buck Creek, Wes Creek, and Flint Creek.

2.3.2 Seismolony Figure 2.4 shows that the PUR-I is located in that l

portion of Indiana that lies in zone 1 for seismic activity.

Within this zone might be found minor damage to structures caused by disttat 2-2 s.

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earthquakes.

The three most significant seismic source zones which are closest to Want Lafayette are:

1.

The New Madrid area of southeastern Missouri; 2.

The Wabash Valley Fault system of southwestern Indians and southeastern Illinois; 3.

The Anr.a, Ohio area.

Reasonable estimates of the maximum magnitude events which 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 attenuation curves, estimates for peak horizontal acceleration at West Lafayette for maximum magnitude events which could occur at these three seismic zones are approximately S-15% G.A l

The way in which the reactor facility was constructed by modifying l

an existing building with no reinforcing bars tied into the original structure, the reactor pool can be considered a free standing unit in the event of any seismic activity. The reactor pool consists of steel i

cylinders containing compacted magnetite sand between the cylinders and the 1/4 inch carbon steel tank. The inaide of this tank was later lined l

with 1/16 inch stainless steel. With these barriers to contain the reactor pool water and considering the reactor pool as a free standing unit it is highly unlikely that any reactor water would be lost during 2-3

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any severe seismic activity.

2.4 Hydroloav Most of Tippecanoe County is covered by glacial drif t.

The drift ranges in thickness from a thin veneer to about 435 feet and was deposited upon a bedrock surf ace that was eroded by a preglacial drainage system. Much of the surf ace drif t consists of glacial till.

Water-laid crossbedded sand and gravel are associated with the till.

The subsurf ace glacial deposits also include much till with interbedded sand and gravel. Locally, clay deposits are as much as 106 feet thick.

Within the drift. five sheetlike water bearing units are differentiated in parts of the county. Ground water within these units occurs under artesian and water-table conditions.

Locally these may occur within the same mait.3 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.2 Illinoian ice dammed the pregiacial Tesys River channel and ponded the relative asall Glacial Lake Laf ayette.

An outlet channel, developed to drain this proglacial lake, was subsequently perpetuated as the present Wabash River drainage line southwestward from the Laf ayette-West Lafayette area.1 The elevation of the Purdue University campus is approximately 706 feet and the level of the Wabash River is approximately 510 feet.

With 2-4

this difference of over 100 feet the flow of both surf ace water and ground water is in a generally easterly and southernly direction toward the Wabash River, which flows around two sides of the campus.

Any leakage of contaminated water from the PUR-I represents no potential hazard to either the West Laf ayette or Purdue University water supply, since these flows are away from the well fields of both. The Wabash River represents a natural barrier between the reactor cud the Lafayette well fields, so no potential hazard exists there.

2.5 Meteoroloav 2.5.1 Climat.o The climate of the county is continental with hot summers and cold winters. The seasons are strongly marked, and the weather is frequently changeable.

Climatological data available from the Purdue University Agronomy Department are sr amarized in Table 2.2.

The table shows the conditions as measured at West Lafayette, where the l

0 latitude is 40*28', the longitude is 87 00', and the ground elevation is 706 feet.

The average annual temperature is about 50'F.

The mean temperature in January, the coldest month, is 23'F, and in July, the warmest month, 0

is 73.3 F, About nine days per year the temperature falls below zero, and about 137 days per year the temperature goes below freezing (32*F).

The average annual precipitation is 35.68 inches.

July is the wettest month with 4.74 inches, and February is the driest month with 1.41 inches of precipitation.

4 2-5

3 REACTOR 3.1 Introduction The 1 kW PUR-I reactor described herein was deaigned and constructed by Lockheed Nuclear Products of Lockheed Aircraf t Corp. of Marietta, Georgia.

Safety and other operational characteristics of this reactor system are similar to other reactors using the MTR sype fuel assembly. The power and flux level of the PUR-I are of adequate range and the experimental facilities are sufficiently flexible to encompass a wide variety of training and research experiments. The reactor is designed so that a minimum of restrictions are imposed on the experimenter, and the console can be readily operated by one person.

Safety is an overriding requirement in a training reactor.

Self-ILuiting features of the PUR-I core, coupled with carefully designed control instrumentation, assure the highest degree of safety. The safety record of this f acility, demonstrated over the past 24 years gives proof Chat the design, construction, and installation of the l

reactor system, coupled with the administrative control over operation, maintenance, and utilization, are more than adequate to provido protection for the public health and saf ety.

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3.2 Reactor Balldina and Reactor Room 3.2.1 Descrintion The Duncan Annex of the Electrical Engineering Building is of brick, concrete block and reinforced concrete construction which was originally designed as a large high voltage laboratory.

It was subsequently subdivided into offices, classrooms and laboratories. The reactor is located in the southwest corner on the groundi 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.

3.2.2 Ventilation The outside air supply and exhaust are both passed through HEP'A filters. The reactor room is maintained at negative air pressure (minimum 0.05 inches of water). All doors to the reactor room have foes rubber seals.

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

3.2.3 Drains The only floor drain to the sewers is sealed except for a vent opening. This vent is raised about two 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.

3.2.4 Emersency During emergency conditions the exhaust system is shut off and the sealed room will prevent the rapid spread of contmaination.

During an emergency the air conditioner and the valve on the drain from the condensate holdup tank are shut of f with the same switch that shuts off the exhaust system. The condensate is held natil it is tested by l

Radiological Control before it is released to the sewer.

If 3-2

contamination la found, it is disposed of as radioactive liquid waste.

3.3 Reactor Core Pertinent design parameters are listed in Table 4.1.

.ISEnE 11 Maximum power level

'I kW Geometry Ift. x ift. x 2ft.

Moderator-coolant Light water Nazimum excess reactivity 0.6% Ak/k Average thermal neutron lifetime 77.2 10-5,,,,

Fuel assemblies Number 16 Standard assemblies 13 Control assemblies 3

Number of plates per standard assembly 10 Number of plates per control assembly 6

Plate dimensions (inches) 2.76 x 25.12 x 0.060 Active fuel length (inches) 23 3/8 Enrichment 93%

U-235 per plate 16.5 gm Water gap 207 inch Cladding 0.020 aluminum

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. Reflector Material on sides Graphite Number of graphite assemblies 20 Control rods and drives Number of regnisting rods 1

Number of shin safety rods 2

Total number of control rods 3

Rod worths Regulating rod

.25% Ak/k Shim safety rod #1 5.0% Ak/k Shim safety rod #2 2.5% Ak/k Rod speed-out Regulating rod 19.7 in./ min.

Shim safety rods 4.4 in./ min.

Scram-time for complete insertion 1 sec.

Material Regulating rod hollow stainless steel Shim safety rods solid borated stainless steel Size Regulating rod (laches) 1/2 x 2 1/4 x 25 1/2 Shim safety rods (inches) 1/2 x 2 1/4 x 25 1/2 Maximum rate of reactivity change 3-4 l

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Regn1 sting rod 1.1 10-4 Ak/k/sec.

-3 Shim safety rods 9.9 10 Ak/k/sec.

Average rate of reactivity charge Regulating rod 5.8 10 Ak/k/sec.

Shim safety rods 5.4 10-3 Ak/k/sec.

Reactivity effects Temperature coefficient (calculated)

-2.1 10-4 Ak/k per *C.

-4 (measured)

-3.4 10 Ak/k per C Void coefficient (measured)

-3.0 10-6 Ak/k per cm Process water resistivity

>330,000 GBM-CM pH 5.5 + 1 Flow rate 30 GPM 3.4 Reactor Pool The reactor pool is built below floor level except for the three foot wall that serves as a biological shield for the operators and experimenters. The pool is contained in a cylindrical tank 17 feet, 4 inches deep and 8 feet in diameter. The core is located to one side to give additional experimental space.

The supports for the drive mechanisms for the control rods, the fission chamber 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 af ter the reactor was built.

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A light weight, portable slaminum bridge can be placed across the I

pool for maintenance and fuel handling operations.

An overhead crane has been installed from the ceiling beams to assist in moving heavy obj ects into and out of the pool. Its capacity is 2 tons.

3.5 Reactivity Parameters This section discusses some important reactor parameters for j

reactor operation and control.

3.5.1 Moderator Temeerature Coefficient Many of the parameters which deteomine the multiplication f actor depend on the reactor temperature.

As a result, a change in the moderator temperature leads to a change in the multiplication f actor, and hence alters the reactivity. This dependency is best expressed in terms of the moderstor temperature coef ficient of reactivity.

It is defined as the ratio of the change in reactivity to the change in the moderator temperature.

It is desirable that the moderator temperature coef ficient be negative since an increase in temperature will then lead to a decrease in the reactivity with a consequential reduction in the reactor power.

Usually, the value of the moderator temperature coef ficient is detemmined experimentally. The PUR-I moderator temperature coef ficient was calculated to be -2.1 10-4 Ak/k/ *C.

Experimentally it has been measured to be -3.4 10~4 Ak/k /*C.

3-6

3.5.1.1 Void Coefficient When water is removed from the core, changes occur in the moderation, leakage, and absorption of neutrons. These changes manifest themselves as reactivity changes. The void reactivity coef ficient is defined as the ratio of the change in reactivity to the voided volume.

For the purpose of reactor safety and stability, it is desired that the void reactivity coef ficient be negative. For the PUR-I the experimental void reactivity coef ficient has been determined to be

-3.0 10-6 Ak/k per cm.

3 3.5.1.2 Excess Reactivity The excess reactivity is defined as that value of reactivity which would occur if all control rods were completely removed from the reactor core.

It is measured for a given core loading starting from a clean, cold 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.

An 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 available to allow for an adequate reactor period.

Therefore, the maximum excess reactivity which is allowed for normal operation is 0.6% Ak/k.

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3.6 Water Process Syston The water process system includes a 30 GPM water pump, a water filter, a domineralizer, flow meter, a chiller, conductivity cells that measure the pool water before and af ter passing through the domineralizer and appropriate valves.

Although the chiller is not needed for present operations, it remains available whenever required.

Calculations indicate Okat the 2.

temperature increases af ter operating the PUR-I at a power level of 1 kW would be 4.65 10-2 oC/ hour. This is based on the mass of water as i

1.85 104 KG.

This takes no credit for heat loss to the surrounding sand and gravel or loss by evaporation.

Experimentally, no temperature increase has been observed with 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 contaninstion in an emergency. The pool water passes through the primary loop while a freon refrigerant is in the secondary loop. The third loop uses campus water to remove the heat and l

is discharged into the campus sewer system. The chance of contamination I

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passing through the three loop system is small.

l 3.7 Reactor Instrumentation and Control l

l The function of the reactor instrumentation is to provide adequate information 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 3-8

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feel 11 tate repair, maintenance, and repositioning. They are in water-tight aluminum tubes. The fission chamber is provided with a motor driven positioning mechanism and position indication system; the other detectors are manually adjustable.

3.7.1 Reactor Safety System Two types of action are incorporated into the reactor safety system to correct for abnormal or unwise conditions; trip, and rod insert.

In a trip the shim-safety rod or rods are dropped by the removal of current from the magnets. A rod insert (set back) will cause all three rods to drive downward into the core.

Both actions are of the latching type and manual reset is required to return to the normal condition.

See the Technical Specifications for the setpoints at l

which trips and rod inserts are initiated.

3.7.2 Channel #1 --Startan Channel The startup channel is used to monitor the neutron flux. The channel consists of a fission chamber, a preamplifier, a pulse amplifier, a scaler for accurate counting, a log count rate and period amplifier, a los count rate recorder, and shares a period recorder with Channel #2.

The range of this equipment is from 1 to 104 counts /second with periods from -30 to +3 seconds.

In addition l

to the outputs shown on the recorders, readout is also provided by a los count rate meter and a period meter shared with Channel #2 on the console and instrument panel. The complete reactor power range may be monitored by this instrument by appropriate repositioning of the detector by means of the fission chamber drive mechanism. The fission chamber may be raised into a cadmium shield by means of a drive mechanism similar to the control rod drive units.

The controls and 3-9

i Position indication for this drive are located on the console. Two set po int s, specified in the Technical Specification and based on the reactor period, provide for a reactor setback and trip in the event of a short reactor period.

3.7.3 Channel #2 --Los N and Period Channel The los N channel indicates the reactor power level over the range from 0.0001 to 300 percent power level. The detector for th'is channel is a compensated ionization chamber followed by a log N mmplifier plus period instrumentation with outputs to the los N recorder and to the period recorder shared with Channel #1.

Indication, in addition to the j

recorders, is provided by log N and period meters on both console and instrument rack, with the console period meter shared with Channel #1.

This channel is not 'on scale' at startup, but will be indicating before the range of the fission is exceeded.

j A reactor trip will be initiated if this channel indicates power levels in excess of 120% of the licensed power. Two set points, specified in the Technical Specifications and based on the period, provide for a reactor setback or trip in the event of a short reactor period.

In the event of the loss of high voltage to the compensated ion ch ambe r, a trip will be initiated.

3.7.4 Channel #3 --Linear Power The linear level channel is capable of measuring neutron finx in a reactor operating range from shutdown to >

100 kilowatt. The sensing element is a BF ionization chamber coupled 3

to a micro-microammeter. The range of the instrument is adjustable by e

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l means of a range switch located on the instrument (instrument panel) i' from 0 - 10.0 10-12 to 0 - 10.0 10-4 amperes.

Detector characteristics, however, limit its maximum output to 10-4 amperes. This channel will thus read from startup to full power by adjustment of the range switch.

The output is recorded on the linear power recorder on the instrument panel and indicated on meters on the consolo and on the instrument

panel, r

This channel has two set points that will initiate a reactor set back at either zero or 110 % range. These set points insure that the instrument is kept on range at all times during reactor operation.

There is also a 120 % range set point that will initiate a reactor trip.

3.7.5 Channel #4 -Safety Channel This channel utilizes a BF ion 3

chamber and f eeds directly into the safety amplifiers. The sensitive range of this instrasent is from a few percent to at least 150 percent of power, linearly.

Its output is indicated on the instrument chassis (instrument panel). The purpose of this channel is solely to provide a trip at the measured value as specified in the Technical Snecifications.

3.7.6 Control Console The reactor console is designed to provide maximum visibility of the 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.

Colors for the indicator lights on the console show the operator the status of the reactor at a glance.

All trip and warning indicators are red or yellow. Operating procedures, as well as interlocks, keep 3-11

e the operator from withdrawing the control rods when a warning indicator is showing.

3.7.6.1 Onorational. Controls 3.7.6.1.1 Control Rod Drives Three identical control channels are used for the shin-safety and regulating rod systems. A push-button switch selects an individual rod to be controlled. All control rods can be inserted simultanously into the core by a gang lower switch when shutdown is desired. This switch can not cause the control rods to be gang raised under any circumstace.

Control console indicators for each rod include the following:

Lower limit Engage i.e. magnet coupled (not applicable for regulating rod)

Shim range (not applicable for regulating rod)

Upper limit Rod positions are indicated on a coarse vertical scale and on a selectable digital readout device having a resolution of 0.01 ca.

3.7.6.1.2 Servo Control System A servo control system provides automatic control once the reactor has reached the desired power level.

The servo control system senses deviations from an adjustable set point on the Channel #3 linear power recorder and adjusts the position of the regulating rod to maintain the reactor at a constant power level.

Servo permit circuitry actuates the consolo alarm buzzer if the reactor power deviates by than 5 % from the set point, indicating a malfunction of the system. A deviation meter is located on the console.

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3.7.6.1.3 Neutron Source Drive A motorized neutron source drive is i

i provided to raise the source through a travel of approximately six feet to the ' full out' position. The system is operated by raise-lower switches at the console with limit switches to indicate the source

'apper limit' and ' lower limit' positions.

3.7.6.1.4 Fission Chamber Drive Controls for the motor drive system for the Channel #1 fission detector are also located on the console, with both a coarse position indicator and a selectable fine position indicator. The drive system is selected and coupled to the drive switch in the same manner as the control rod drives.

Indicator lights note the apper and lower limit positions.

3.7.6.2 Annunciator and Alarm Systems When a system trip occurs, or when other abnormal system conditions are sensed, an alarm (bazzer) will sound and an illuminated indicator will be lighted on the control console, indicatir.g the source of the trouble.

An annunciator acknowledge button may be used to reset the buzzer.

3.7.7 Radiation Monitors 3.7.7.1 Radiation Area Monitors (RAM) The radiation area monitoring (RAM) system consists of three scintillation type detectors for monitoring samma radiation. These detectors measure the radiation level above the pool, near the dominera112er cartridge, and at the control console. These monitors use a los scale to cover a wide range of radiation levels and have readouts both on the instrument and on the instrumentation rack.

Set points on each instrument will initiate a 3-13

reactor trip if the dose rates exceed predetermined levels. These levels are specified in the Technical Specifications.

3.7.7.2 continnans Air Monitor. (CAN)

A continuous air monitor (CAN),

which utilizes a GM tube as a detector, is in operation in the reactor 4

room to indicate long term levels of radiation and to monitor any radioactive particulates released to the air in the room.

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19.28. 3 873 L_

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I

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),8

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.. -(.. ~ ~ '

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w sm i

. J.E.L.*.

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I 1

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.,..... yU.;., :T.:k. /."..[

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. g. q...:. ::.M i

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Figure 3.1 Reactor Room - B-70A.

i 3-15 L

4.

ENVIRONMENTAL IMPACT AND RADIOACTIVE WASTE MANAGEMENT 4.1 Construction The reactor is located in an existing building at a developed site and there are no plans for additional construction.

4.2 Environmental Effects of Facility Onoration 4.2.1 Thermal Discharmes The PUR-I reactor, under normal modes of operation, discharges no heat to the environment.

Although a heat exchanger, operating on the University water supply and discharging into the sewer system is available, it is not operated since operation at the licensed power of 1 kW does not significantly raise the temperature of the pool water.

4.2.2 Radioactive Discharmes 4.2.2.1 Airborne Waste Argon-41 Argon-41 is produced by thermal neutron activation of Argon-40 in the air.

No detectable traces of Ar-41 from air dissolved in the water or in the isotope irradiation tubes has been observed.

Tritium No detectable tritium has been observed in the pool water.

Even at the design power levels of 10 kW the neutron flux is too low to produce detectable quantities.

Nitrogen-16 The main possible source of nitrogen-16 is from the f ast neutron interaction with orygon in the pool water. The nitrogen 4-1

must then diffuse to the surf ace of the pool before it is released to the atmosphere.

In normal operation, no strong currents are established in the reactor pool and with the short half-life (7.14 seconds), the nitrogen decays before reaching the surface.

No nitrogen-16 has been observed in the reactor room.

4.2.2.2 Lianid Waste Normal reactor operations produce no radioactive lignid waste except those that might be produced for experiments on campus. These wastes are disposed mader the University's By-Product license.

4.2.2.3 Solid Waste Solid waste generated at the f acility consists of potentially contmainated paper and gloves and solid semples produced for experiments. These wastes are disposed under the University's By-Product license.

4.3 Environmental Innects of Accidents Accidents ranging from the failure of experiments to fission product release are considered in the chapter on Accident Analysis.

Effects are considered negligible with respect to the environment.

l 4.4 Alternatives to Oneration of the Facility Nach of the educational and research activities using the PUR-I cannot be done using other suitable or economic means. The PUR-I is the i

j only operating reactor in the state of Indiana.

1 l

I e

~

4-2

4.5 Lona-tern Effects. Cos_ts and Benefits, and Alternativet pf, Facility Onoration Since the f acility is an existing one, the capital costs are low.

The operational costs are minimal while the benefits in the education process of nuclear engineering students, radiation health scientists and technicians are great, both to the individual people and to the national interests.

No reasonable alternatives exist to the wide versatility of research/ training reactors such as the PUR-I in contributing to education and scientific knowledge.

b t

l i

4-3 l

5.

RADIATION PROTECTION PROGRAN Purdue University has a structured radiation saf ety program.

Policies for the program are determined by the Radiological Control Committee established by the President of the University. The program is administered by the Radiological Control Officer and his staff. The staf f is equipped with radiation detection instramentation to determine, I

control and document occupational radiation exposures at the reactor facility and all laboratories using radioisotopes at the University under the By product License 13-02812-04 (Broadscope). A ' Radiological Control and Health Physics Handbook' has been published which contains the rules and radiation safety procedures for all laboratories using radioisotopes and/or producing ionizing radiation, including the reactor. Routine surveys are performed of the reactor room and include analysis of the reactor pool and reactor room air.

5.1 ALARA Commitment The Universeity is committed to the ALARA principle, the Office of Radiological and Chemical Control makes every effort to keep doses as low as reasonably achievable (ALARA). All unanticipated or sansaal exposures are investigated.

5.2 Health Physics Promram At present, the normal in11-time health physics staff consists of a Radiation Saf ety Officer, an Assistant Radiation Saf ety Officer, two Health Physicists, an Environmental Waste Technician and appropriate j

I 5-1

secretarial support. The Health Physics staf f performs all routine surveys and is available for consnitation in matters concerning radiation safety.

5.2.1 Proceduras Written procedures have been prepared that address routine health physics monitoring at the University's research reactor facility.

5.2.2 Instrumentation The University has a variety of detecting and measuring instruments for monitoring potentially hazardons ionizing radiation. The instrument calibration procedures and techniques ensures that any credible type of radiation and any significant intensities will be detected promptly and measured correctly.

5.2.3 Trainina All reactor-related personnel are required to attend a radiation safety training session before they begin work at the reactor.

5.3 Radiation Sources 5.3.1 Reactor Radiation from the reactor core is the primary source of j

radiation directly related to reactor operations.

Radiation exposure rates from the reactor core are reduced to acceptable levels by the water in the pool and concrete shielding.

5.3.2 Extransons Sources Sonrces of radiation associated with reactor use include radioactive isotopes produced for research, activated components of experiments and activated samples.

5-2

-.. ~..

i

)

5.4 Routine Monitorina 5.4.1 Fixed - Position Monitor The PUR-I has 3 fixed position l

radiation area monitors (RAN) with adj ustable alarm set points and 1 contianons air monitor (CAN) in the reactor room. The CAN sir filters are changed and analyzed semi-monthly.

5.4.2 Experimental Wipe tests of exposed surf aces 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 j

activity and wipe tests made for loose contamination.

5.5 Ocennational Radiation Exnosares 5.5.1 Personnel Monitorina Proarsa Film badges and '1LD finger rings are assigned to all approved reactor personnel.

In addition, self reading pocket dosimeters and dose rate instruments are used to administrative 1y keep occupational exposures below regulatory limits in 10 CFR 20.

Students and visitors are provided self reading pocket dosimeters.

1 5.5.2 Personnel Exmosures Approved reactor personnel are monitored I

with film badges and TLD finger rings.

Exposures are generally minimal except during the annual fuel plate inspection when a finger ring dose

> 100 mrom may be experienced by the plate inspector.

Because the reactor personnel and PBBF (fast breeder blanket facility) personnel are the same and use one personnel dosimeter system, it is not possible to determine how much of the ' facility' dose is a result of the reactor I

5-3 4

,--g--

-, ~.., -, -

---

,-,.-.._,,_._,,-.n_-

,,,.,e

,..--,.s,---------

l l

operations. A sammary of the last (5) years whole body exposure to reactor personnel is provided below.

j History of personnel radiktion exposure at Purdue University roastor facility i

Whole Body Number of Individuals i

exposure range (roms) 1981 1982 1983 1984 1985 2

minimal 1

3 4

7 7

(.1 7

6 2

1 0

.1

.25 0

0 0

0 0

i

.25

.75 0

0 0

0 0

.75 - 1.0 0

0 0

0 0

> 1.0 0

0 0

0 0

5.6 Effluent Konitorina 5.6.1. Airborne effluent No airborne effinents have been released from l

the PUR-I.

Potential releases are monitored with a CAN in the reactor l

room,

.i 5.7 Environmental Monitorine I

The environmental monitoring program consists of monthly senples of l

the reactor pool's water, semi-monthly analysis of reactor room air i

samples, one TLD placed in the reactor room, and one TLD in a classroom i

above the reactor room.

Reactor pool water semples are analyzed for gross samma, B3, gross alpha, and gross beta.

71eactor room air samples are analyzed for gross alpha and gross beta. Resnits indicate nothing r

beyond natural background has been detected on the reactor room air and 5-4 9

-n--me-e--.

. _ ~

reactor pool water samples. ILDs have generally shown levels ranging from minimal - 30 mree/ quarter.

5.8 Potential Dose Assessments Natural background radiation levels in the West Lafayette area resnit in an average exposure of about 100 arem/yr.

On the basis of normal reactor nee, the maximum potential non-reactor room dose wonid be less than 1 mram/yr, so there should be no significant contribution to the background radiation in unrestricted areas.

1 i

O

'I 5-5

l 6.

CONDUCT OF OPERATIONS 6.1 Ormanization The reactor facility is an integral part of the Schools of Engineering at Purdue University as shown in Figure 6.1. The reactor supervisor has direct responsibility for the operation of the PUR-I. He is responsible for assaring that all operations are conducted in a safe manner and within the limits prescribed by the facility license, Technical Specifications, and other applicable regulations.

In all matters pertaining to the administrative aspects of the operation of the reactor, the reactor supervisor reports directly to the l

Head of the School of Nuclear Engineering or his designated alternate.

Financial budgets for the operation of the reactor are handled through the School of Nuclear Engineering.

In all matters pertaining to radiation safety the reactor l

supervisor is responsible to the Radiological Control Committee, usually through the Radiological Control Officer. The Radiological Control Committee was established by the President of the University under an executive memorandam A-50. The duties of the committee, revised under executive memorandan B-14, include responsibility, from the standpoint f

of safety, for all University programs involving radioactivity or producing radiation.

The committee determines the policies and reviews all applications for use of radioactive materials and radiation on the Purdue campuses. The Radiological Control Program is administered by the 6-1

l i

Radiation Control Officer and his staf f. The Radiological Control Committee also has several saboommittees reporting directly to it, J

including the Committee On Reactor Operations (CORO).

Qualifications for the reactor supervisor, remotor staff, and Radiological Control Officers are established in Section 6.1.4 of the r

l Technical Specifications. The duties of the reactor staf f are also specified in Section 6.1 of the Technical Specifications.

6.2 Reactor Procedures i

Written procedures for all reactor operations are prepared by the reactor staff and reviewed by the Committee On Reactor Operations (CORO). Those proposed changes to procedures, equipment or systems that

)

change the original intent or use and/or are non-conservative or those

l I

that involve an unreviewed safety question as defined in Section 50.59, 1

10 CFR are reviewed and approved by 00R0 before being implemented.

6.3 43411g i

)

Review of the facility activities shall be performed by the 00RO.

I j

Audits of the f acil.8 ty activities shall be performed under the cognizance of the 00k e but in no case by the personnel responsible for the item audited. Individual audits may be performed by an individual who needs not be an identified CORO member. These audits shall examine the operating records, procedures, retraining of the f acility staf f, j

resnits of actions taken to correct deficiencies, facility emergency 6-2

plan, and the facility security plan at intervals designated in the Section 6.2.6 of the Technical Specifications.

6.4 Staff Trainian.Proaram The PUR-I reactor f acility has an NRC-approved operator requalification program that all licensed reactor operators and senior reactor operators must complete as a condition for renewal of their licenses. Persons who are preparing to take the NRC operator's licensing examination participate in essentially the same training program as well as receive extensive ' hands on' reactor operations training at the console. All licensed operators at the PUR-I participate in the program I

and must satisf actorily complete this program during each license o

renewal period. Each licensed operator or senior operator includes in his/her license renewal application a statement that he/she has satisfactorily completed the requirements of the requalification Program. The requalification program is divided into three major areas which are designed to provide assurance that all operators maintain competence in all aspects of the licensed activities. The three areas are as follows:

1 Lectures followed by examinations of various parts of the reactor operations Technical Specifications, emergency plans, and security plans. Special lectures or makeup studies are used to retrain those operators who demonstrate deficiencies in any part of the examinations.

o 6-3

2.

An annual written examination is used to verify the operator's overall knowledge level of reactor operations.

3.

An ananal evaluation of the operator's performance on the reactor console to actual and/or sinalated plant conditions.

At the present time there are only two licensed operators for the f acility, both are senior operators and both ananally teach courses in reactor operation and the facility. Therefore, in a letter dated March 23, 1982, the facility has received permission to suspend the retraining yxc3 ram for the staf f. It will again be reinstated at such a time as the reactor staff is increased.

6.5 Emermancy Plannina The PUR-I reactor emergency plan includes the guidelines, policy, and organization required to mitigate the consequences of an emergency.

Specific implementation procedures are provided for each type of emergency in the standard operating procedures for the PUR-I reactor.

l The revised emergency response plan dated May 10, 1984, has been approved by the NRC.

6.6 Physical Security Plan There is a physical security plan for the PUR-I reactor f acility which describes the physical protection system and the security l

organization which provides protection against radiological sabotage and l

detection of thef t of special nuclear material from the f acility and related laboratories. The physical security plan was submitted as 6-4

Amendment 4 and Amendment 5 of the reactor license, and is on file with the NRC. It is withheld from public discloenre pursuant to 10 CFR 2.790(d).

6.7 Records and Renorts 6.7.1 Onoratina Records The records listed in Table 6.1 are prepared and retained for at least five years. The-records and logs listed in Table 6.2 are retained for the life of the f acility.

6.8 Renortint Reanirements 6.8.1 Annus1 Onoratina Renorts A report covering the operations for the previous year is submitted to the directory of the appropriate NRC regional of fice by March 31st of each year. It includes:

a.

Changes in plan design and operation b.

Power generation c.

Unschednied shutdowns d.

Maintenance e.

Changes, tests, and experiments f.

Radioactive effinent releases.

6.8.2 Non-Routine Renorts In the event of a reportable occurrence, notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and/or telegraph to the Director of the Regional Regulatory Operations Office, followed by a written report within ten days. The written report on these 6-5

abnomaal occurrences, and to the extent possible the preliminary telephone and/or telegraph notification shall:

a.

De scribe, analyze, and evaluate safety implications, b.

Outline the measures taken to assure that the cause of the condition is determined, c.

Indicate that corrective action taken to prevent repetition of the occurrence and/or similar

't occurrences involving similar components or

systems, d.

Evaluate the saf ety implications of the incident in light of camulative experience obtained from the record of previous failures and malfunctions of similar systems and components.

6.8.3 Ennsual Events A written report shall be forwarded within 30 days to the director of the Regional Regulatory Operations Office in the even of:

1.

Discovery of any substantial errors in the transient or accident analysis or in the methods used for such an analysis, as described in the Hazards Summary Report on the bases for the Technical Specifications.

2.

Discovery of any substantial variance from 6-6

perfoemance specifications contained in the the Technical Specifications or in the Hazards Summary Report.

3 Discovery of any conditida involving a possible single f ailure which, for a system designed against assumed single failures, conid result in a loss of the capability of th6 system to perform its saf ety function.

9 6-7

Table 6.1 Resords and Lega Retained 121 f.iIt J.1A21 a.

Normal facility operation and maintenance.

b.

Reportable occurrences c.

Tests, checks, and measurements documenting compliance with surveillance requirements.

d.

Records of experiments performed.

e.

Recorder charts.

f.

Records of radioactive shipments.

g.

Changes to operating procedures.

h.

Facility radiation and contamination surveys.

l a

6-8

Table 6.2 Records n,4 h Retained fs,g M of, Facility a.

Gaseous and liquid waste released to the environs.

b.

Off site environmental monitoring surveys, c.

Radiation exposures for all PUR-I personnel.

d.

Fuel inventories and transf ers.

e.

Updated, corrected, and as-built facility drawings.

f.

Minutes of the CORO meetings.

g.

Records of transient or operational cycles for those components designed for a limited number of transients or cycles.

h.

Records of training and qualification for members of the facility staff.

i.

Records or reviews perfor: sed for changes made to procedures or equipasnt or reviews of tests and experiments pursuant to 10 CFR l

l 50.59.

l l

e 6-9 l

7.

ACCIDEPfr ANALYSES In this chapter, details of the analysis of various accident scenarios are presented. The results of some of these analyses validate the saf ety system settings established in the Technical Specifications for the PUR-I. The potential effects of the accidents on the health and safety of the staff and public are analyzed.

7.1 Fuel Element Handlinn Accident Fuel element maneuvers are always conducted in the reactor pool.

They are removed from the core and moved into the storage space, one at a time, using a hand-held fuel handling tool. Annually a fuel element is removed from the pool for inspection. A fuel element weighs about 3.18 kg (7.0 lb) in air and only about 2.0 kg (4.4 lb) in water.

therefore, even if a fuel element should f all from the handling tool during its transfer it is not heavy enough to cause any considerable damage. The most severe consequence likely to occur would be some denting of the end fittings since the fuel element, being an elongated obj ect, would tend to f all in water in a rather upright position.

The PUR-I Standard Operating Procedures define administrative steps which are intended to prevent a fuel handling mishap.

They are:

i 1.

All fuel handling is done in accordance with written procedures.

2.

Loading operations are done by qualified personnel under direct supervision of a Senior Operator.

e 7-1 ll --

3.

The fuel handling tool is kept locked with the key secured to prevent unauthorized movement of fuel.

It is concluded that no adverse consequences are to be expected to the health and safety of the public or the staff from this type of accident.

7.2 Floodina of an Irradiation Facility I

A sudden replacement of a voided, i.e. air filled, space next to the core by water would cause a stepwise reactivity insertion.

Its magnitude depends on the void volume being replaced and its position relative to the core. Experiments have shown that flooding of the 5' irradiation tube located outside of the graphite reflector element F6 with water adds 0.3 % Ak/k to the core.

It is shown in Section 7.5 that a sudden reactivity insertion of 0.6 % Ak/k into a critical core of the PUR-I can be tolerated with a sufficient safety margin. Therefore, it is concluded that flooding of any irradiation f acility would not endanger the reactor and would not pose any hazards to public health and safety.

7.3 Lpsm of. Coolant Accident The reactor pool is designed to prevent the possibility of an unintentional drainage.

It is constructed of steel and set in a second steel tank with the interstitial region filled with sand.

The tank rests on a concrete pad about 15 feet below the floor of the Reactor Room, which is in the basement of the building. The pool has no drains.

Therefore, a sudden loss of coolant is considered to be extremely 7-2

\\

l i

remoto.

If the pool drained instantaneously, while the reactor was operating, the loss of water (moderator) would shut the reactor down.

The most severe problem identified in this accident scenario is the removal of decay heat during and af ter loss of coolant. There is no danger of significant fuel overheating as long as the core stays immersed and heat can be removed by the water.

If the core were to become uncovered, heat transf er would occur by natural convection of ambient air. For this case, the maount of heat removed is proportional to the cladding temperature. Decay heat generation after reactor shutdown is shown in Figure 7.1.

According to this Figure, the decay power of the PUR-I benediately af ter the shutdown from full power (1 kW) is about 63 watts. The decay power rapidly decreases as indicated in Figure 7.1, being about 35 watts after 1 minute of cooling. At these power levels, no heating problem exists.

In any accident which is reasonably conceivable, the 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 due to a decreasing pool water level. The pool water level is checked during daily routine operations.

It is concluded, that a slow leak of pool water would be discovered early and specific actions could be taken to mitigate its consequences.

It is concluded that no adverse consequences are to be expected to the health and saf ety of the public or the staf f from this type of 9

7-3 i

i

l l

accident.

7.4 Failure of a Movable Exneriment A sudden (stepwise) introduction of a positive reactivity into the critical reactor will cause a transient power increase.

Its magnitude and course depend on the amount of the inserted reactivity. At the PUR-I, the maximum reactivity worth of a movable experiment is limited by Technical Specifications to 0.3% Ak/k In the following analysis an assumption is made, although it is highly unlikely under current operational practice, that an experiment with the maximum reactivity worth suddenly moves out of the core.

This wonid result in a positive stepwise reactivity change of 0.3 % Ak/k. A number of other conservative assumptions are:

1.

The reactor power is 1.0 kW.

2.

All control rods are in a position with the least differential reactivity worth.

3.

The most reactive control rod (SS1) cannot be scrammed (stuck rod criterion).

4.

The power excursion does not start to reverse until the second shin safety rod is fully inserted (600 as af ter scram).

5.

No thermal feedback ef fects are taken into account.

Using the prompt j ump approximation, the calculated power jumps to 1.58 kW before a scram is initiated. The stable reactor period, 7-4

corresponding to the reactivity insertion of 0.3% Ak/k, is 9 seconds.

j Two '120 % Full Power' trips based on separate detectors and electronic systems would be activated. Therefore, there is both redundancy and diversity available to terminate a mild power excursion such as described above. The power goes up to 1.69 kW before the least reactive shim safety rod (-2.2 % Ak/k) is fully inserted.

Again, this accident is similar to t'ho situatloc analyzed in section 7.5 where a maximum of 0.6% Ak/k is inserted into the reactor.

It is concluded that no adverse consequences are to be expected to the health and safety of the public or the staf f from this type of accident.

7.5 MAXIMUN REACTIVITY INSERTION 7.5.1 GENERAL This hypothetical accident begins with the sudden insertion of all of the licensed excess reactivity into the critical reactor operating at full power. It is assumed that the reactor operator takes no corrective action.

Two general cases are considered:

Case A assumes that the safety control circuitry is operating and is activated by the reactor power exceeding the 120 % power set point. Under these conditions, the period would be about 1 second and the short period trip would also initiate a reactor trip. At this point it is assumed that the most reactive shin

(

l saf ety rod is stuck and the second shim saf ety rod drops into the l

l reactor providing a shutdown margin of 1.9% Ak/k.

It is assumed that the negative reactivity provided by the shim saf ety is added as a linear ramp, ortending over 1000 milliseconds; the time specified in the 7-5

technical specifications for the rods to be fully inserted. This is a conservative assumption because (a) the rods are slowed near the end of the travel in order to prevent damage and the actual drop times over the most effective portion of their range is much less than 600 milliseconds and, (b) the most reactive region of the control is near the mid-range.

This case assumes no temperature feedback effects and no heat transport from the core.

Case B considers the situation when both rods are stuck and the reactor is controlled only by the negative temperature feedback. In this case the reactor pool is assumed to remain at a constant temperature over the time frame considered, providing sub-cooled water at 20 degrees

~

C.

7.5.2 THE CALCULATIONAL MODEL The power transient analysis for the PUR-I is based on the results obtained from a computer program designed to solve the point-kinetics equations coupled with a single-mode temperature feedback equation.

Since the magnitude of the reactivity inserted in each case is less than one dollar, the point-kinetics solutions should closely approximate the actual reactor behavior.

The equations used to describe the kinetic behavior of the reactor are:

6 jf=o(t)-B n + iAC (1) gg k

i=1 W

7-6

O

=

n -AC (2) gg and p(t) = P (t) + mT(t) + p,(t-t )

(3) g d

where average moderator temperature,1/2(T,, + Tont),( C),

T

=

g T,

g inlet moderator temperature, (20 C)

=

p(t) reactivity; may include moderator temperature feedback

=

and shutdown reactivity insertion, as indicated, temperature coefficient, a

=

t,

(

%A k),

p, reactivity inserted starting at time,

=

d and jf prompt neutron generation time, (77.2ps).

=

The temperature of the reactor is a function of the power, but also depends on the flow of the water through the reactor. A relationship for the PUE-I, based on material in Ref.1, and measurements, is given by T=T,+1.159*(f)2/3 (4) g and

(

d =.9 982 - 2.105 e-4(T-T,) - 3.95 e-6(T-T,) 2 (5) g where P is the reactor power in kilowatts and d is the density of the water as a function of temperature. The temperature coef ficient, a, is considered to be constant over the temperature range of interest.

9 7-7

Additional values of the input data used in the equations are summarized in Table 7.1.

7.

5.3 CONCLUSION

S Based on the results presented in Table 7.2, for Case A where the reactor safety systems terminate the chain reaction, it is concluded that the postulated transients do not create conditions which would endanger personnel or render equipment inoperative. The reactor power would rise to about 2.8 kW in 55 as, primarily due to the prompt jump followed by a 1 sec period. However, the 120 % trip would be reached within 3 as initiating a reactor trip. The assumptions in the calculational model are conservative since no credit for negative temperature feedback is taken and the ramp insertion of negative reactivity underestimates the actual insertion function.

It should also be remembered that the reactor is actually designed for steady-state operation at 10 kilowatts, although licensed at 1 kilowatt. The minimum shutdown margin under these conditions,1.9% Ak/k, is sufficient to maintain the reactor in the shutdown condition and the power is rapidly reduced before significant energy release occurs.

The results for Case B, where no control rods are inserted and the negative temperature coef ficient is considered as the shutdown mechanism, are presented in Table 7.3.

In this case the analysis shows that the reactor power would rapidly rise over a period of about three I

minutes to a power level of about 360 kilowatts. These results are consistent with a number of the excursion experiments performed at the BORAI AND SPERT Facilities.2,3 Some of the results of the SPERT-1 experiment using the DU-12/25 core are applicable to the analysis of the 7-8 1

PUR-I reactor since the fuel geometry and composition are very similar.

A detailed comparison of the reactor characteristics are given in Table 7.4. A series of self-limiting power excursion tests was carried out in SPERT-1 using five core loadings. The input variable referred to in these experiments was the reactor period, induced by a step wise reactivity insertion. The results of the calculations of the PUR-I experiment are consistent with the observed results of the SPERT-1 experiments using long periods on the order of 1-3 seconds. The SPERT-1 experiments showed that the fuel could withstand transients with periods as short as 14-asec with no apparent damage to the fuel. From the tests it was concluded that the mechanism responsible for self-limiting the power excursion consist of fuel and moderator thermal expansion and boiling (the latter being the dominant shut-down mechanism). Thus, fron the experiments with various stepwise reactivity insertions, it is concluded that the above postulated accident would be safely terminated by this self-limiting shutdown mechanism.

Such an accident can, therefore, be terminated even if the safety instrumentation were inoperable.

l It should be pointed out that the assumptions leading to this accidsat are very unlikely, and therefore, it is not believed that such an accident would ever happen. The analysis, however, has been useful in showing the inherent saf ety capacity of the PUR-I.

No effects on the health and safety of the public nor the reactor staff are to be expected from this type of accident.

l 7-9

7.6 Failure of a Fueled Exneriment In this section an analysis is perf omeed to assess the hazard associated with the f ailure of an experiment in which fissile material has been irradiated in the reactor.

In the scenario of Okis accident it is assumed that a capsule containing irradiated fissile material breaks and a portion of the fission product inventory becomes airborne.

The consequences of the release are analyzed for both the reactor staf f and general public.

Since the potential impact of this postulated accident is greater than in any other accident analyzed, the failure of a fueled experiment is designated as the maximum hypothetical accident of the PUR-I.

In this analysis the consequences of a f ailed experiment generating 1 W 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 products and actinides are not volatile at the temperatare (which is essentially at room temperature) at which the experiment would be performed. The amount of noble gases and radiciodine is assumed to be that specified in Ref. 1, i.e.

100 % of the noble gases and 25 % of the iodine inventory. If the experiment were to break in the reactor l

pool a credit for the absorption of iodine in water can be taken.

However, this is not considered in this analysis.

A conservative assumption that the irradiation time was infinite was made in the analysis. Therefore, the fission inventories used in 1

l l

1 7-10

the analysis for some long-lived radionuclides, e.g. Er-85 or even I-131, are overly conservative. Furthermore, it was assumed that the fission products are instantaneously released and uniformly distributed in the Reactor Room air. The free volume of the Reactor Room is 3

approximately 424 m.

The external dose rate (in area /hr) due to y - and S-radiation was calculated using the relationships given in Ref.1 11 D = 9.43 x 10 xIzE (6) y 3

where I = radionuclide concentration (C1/cm )

I average 7-energy per disintegration (MeV) and

=

7 E = 8.24 x 1011 xIzE (7) g p

where Yg = average $-energy per disintegration (MeV).

The dose rate b represents the skin dose.

The dose rate to the g

thyroid (in res/hr) due to the inhalation of radiciodines is given by k=DCFzBzI (8) dose-conversion f actor for the thyroid (res/C1) where DCF

=

3 breathing rate (cm /hr) and B

=

radiciodine concentration (C1/cm }

3 I

=

The standard breathing rate recommended is 1.25 x 106,,3/hr.

The calculated saturation activity for each respective radioisotope and its concentration in the Reactor Roon af ter experiment failure is shown in Table 7.5 for an experiment of 1 T.

Also shown in this table are the calculated dose rates f or the whole-body, skin, and the thyroid.

7-11

With a 7 -dose rate in the Reactor Room as high as 50 mres/hr any one of the radiation area monitors wonid cause an automatic reactor shutdown and audible and visual alaces in the control room.

From the past experience, it is known that the reactor building can be evacuated within 1.5 minutes. Therefore, it is assumed in the following analysis that the exposure time to the members of the reactor staf f is 1.5 minutes. The resulting radiation doses are: whole-body dose 22.3 arem, skin dose 13.4 arem, and the thyroid dose 1.43 ren.

This radiation exposure approaches the limits established in the Technical Specifications, Sec 3.5.f for a singlely excapsulatd experiment. This experiment corresponds to the irradiation of 1.1 ga of

~

U-235 in the mid plane of the isotope irradiation tube located in position F6.

For the radiation calculations outside of the reactor building it was assumed. that all fission products released in the reactor building would leak ott within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Since the reactor room does not have any windows and has only a few doors and emergency procedures call for turning of f the air omhaust system, the assumption about the leak rate

]

is considered to be reasonable. Another conservative assumption was 1

)

made in that no radioactive decay and hence no decrease in the source strength was taken into account while calculating the dose rates outside the reactor building. The radionuclide concentration at a distance of 100 m from the release point was calculated using the atmospheric dispersion f actor recommened in Ref.1;

)

I

~

7-12 l

1

-d'= _ I (9)

Q nuayag where X is the concentration of radioactive material (C1/m )

3

'Q = source rate (C1/ s) a = average windspeed (m/s) y = lateral plume spread (m) a 2 = vertical plume spread (a) a In the above expression for the atmospheric dispersion f actor no credit was taken for so-called building wake ef fects and horizontal plume meandering, both of which help in spreading the radioactive plume.

Also, no credit was taken for the f act that air from the basement ares is exhausted at a minimum height of 50 feet.

Using an average windspeed of 1 m/s and Pasquill type F atmospheric conditions the dispersion 3

factor at 100 m is calculated to be 1.78 x 10~4 s/m.

(Actually, the average windspeed is about 3.4 m/s. Therefore, the results of this analysis are conservative at least by a factor of 3).

Calculated dose rates at 100 m for an experiment power of 1 W are shown in Table 7.6.

If it is assumed that an individual is located at this point for 2 hrs following the fission product release from a l

l postulated experiment f ailure then his/her resulting radiation dose to l

the whole body would be.51 area and to the thyroid.02 rem.

These doses are only fractions (about 1%) of those which are referred to in 10 CFR 100 in conj unction with the determination of an exclusion area.

It is concluded that experiments using fissile material can be irradiated at the PUR-I within the power limits 7-13

Esferences 1.

Lewis, E.E., ' Nuclear Power Reactor Safety', John Wiley and Sons, Inc. 1977.

4 2.

Dietrich, J.R.,

' Experimental Determinations of the Self-Regulation and Saf ety Operating Water-Moderated Reactors',

Proceedinas p.f the h U.N. International Conference spa Peaceful Uses p.1 Atomic Enerzy, Ypo.112., Geneva, 88 (1955).

3.

Nyer, W.E.,

et al.,

' Transient Experiments with the SPERT-1

)

Reactor', NagJeonies 13., No 6, 44 (June 1956).

O 4

7-14

Table 7.1 Delayed Neutron Fraction Half-life Si =.00021 t1 = 56 sec

$2 =.00142 t2 = 23 S3 =.00128 t3=

6.2

$4 =.00257 t4=

2.3 S5 =.00075 t5=

0.61

$6 =.00027 t6= 0.23 The ' offective' delayed neutron function is given by $*

= yp where y = 1.15 for the PUR-I.

analyzed in this section. There is no nadne hazard to the general public nor to the reactor staff in the very hypothetical case of a f ailed experiment as postulated above.

I

.k 7-15

Table 7.2 Resnits.91.1A1 f.25.71 Transient M sh y.,13A h Insertion.91 Control _4pf Case A Initial Conditions Power level (kW) 1 Temperature ('C) 20 Plow Rate (kg/s) 0 J.g g.t Valnes Reactivity (step) added (Mk/k) 0.6 Insertion time (as) 1000 Reactivity (ramp) added (Mk/k)

-2.50 Calculated Resuits Nazimum power level (kW) 2.85 Elapsed time to maximum power (as) 55 Elapsed time while P)1 kW (ms)

.275 Total energy released while P)1 kW (kW-s)

.503 l

\\

l I

~

7-16

T.

Table 7.3 1,JgLt1 31 tig h Transisnt Analysis wltA 22 Control 194g Initial Q aditions Power Level (kW) 1 1

Pool Temperature (supoC) 20 20 Reactor Temperature ('C) 20.6 20.6 Temperature Coefficient (Ak/k) / *C (30 C - 50*C) calculated

-2.le E-4 measured

-3.4 E-4 h Values Reactivity (step) added (%Ak/k)

.6

.6 Calculated Rejgita Maximum Powel Level (kW) 380 170 Elapsed Time to Max (min)

>3 12 Total Energy Released in 3 min (Nw-s) 54 24 Pool temperature at I hour ('C) 32 28 W

7-17 j

Table 7.4 Comparison 31 Important f,ggi Dgt.g PUR-I SPBtT-1

)

Geometry Plate Plate Length [ca]

61 61 Width [ca]

7.0 7.6 Thickness [ca]

0.15 0.15 Water gap [ca]

0.53 0.45

f. Pal Material U-Al U-Al Enrichment [%]

93 100 Thickness [mm]

0.51 0.51 Claddina Material Al A1 Thickness [mm]

0.51 0.51 7-18

Table 7.5 Dose Rates in the fleactor Room from a Failed Fuel Experiment (Power = 1 Watt)

Isotope As E-gamma E-beta DCF DR-gama DR-beta DR-thyroid (Cl)

(C1/cm3)

(MeV)

(MeV)

(res/Cl)

(aram/hr)

(ares /hr)

(rum /hr)

I-131 2.66E-02 1.57E-11 3.71E-01 1.97E-01 1.00E+06 5.49E40 2.55E+00 1.96F41 I-132

3. 5E-02 2.21E-11 2.40E40 4.48E-01 6.60E+03 5.00E41 8.16E+00 1.82E-01 1-133 5.31E-E 3.13E-11 4.77E-01 4.23E-01 1.80E45 1.41E41 1.09E+01 7.04E+00 I-134 5.94E-02 3.50E-11 1.94E+00

4. 5 -01 f.10E+03 6.41E+01 1.31E41 4.82E-02 I-135 4.69E42 2.77E-11 1.78E+00 3.08E41 4.40E+04 4.64E+01 7.02E+00 1.52E+00 Kr-83s 5.90E-03 1.39E-11 2.60E-03 1.03E-02 3.41E-02 1.18E-01 Kr-85s 1.27E-02 3.00E-11 1.51E-01 2.23E-01 4.27E+00 5.50E+00 Kr-85 2.53E-03 5.97E-12 2.11E-03 2.2 I-01 1.19E42 1.10E+00 Kr-87 2.00E-02 4.72E-11 1.37E+00 1.05E+00 6.09E+01 4.08E+01 Kr-88 3.12E-02 7.36E-11

1. 74E+00 3.41E-01 1.21E+02 2.07E+01 Kr-89 3.96E-02 9.34E-11 1.60E+00 1.33E+00 1.41E+02 1.02E+02 Ie-131s 2.53E44 5.97E-13 2.00E-2 1.40E-01 1.13E-02 6.88E-02 Ie-133s 1.35E-03 3.18E-12 3.26E-01 1.55E-01 9.79E-01 4.07E-01 Ir133 5.48E-02 1.29E-10 3.00E-02 1.46E-01 3.66E+00

1. 5 +01 Ie-135m 1.77E-02 4.17E-11 4.22E-01 9.74E-02 1.66E+01 3.35E+00 Ie-135 5.23E-02 1.23E-10

2. 46E-01 3.22E-01 2.86E+01 3.27E+01 Ie-137 5.31E-02 1.25E-10 1.50E-01 1.37E+00 1.77E+01 1.41E+02 fe-138 5.5E-01:

1.J'E-10 1.10E+00 8.00E-01 1.36E+02 8.66E+01 7.11E+02 4.92E+02 2.84E+01 7-19

[

Table 7.5 Dose D.les at 100 meters Isotope DR -9amma DR -Deta DR-thyroid (mrem /hr )

(mrem /hr )

(rem /hr )

1-131

1. 9 5E -0 3 9. 0 6E -0 4 6. 9 8E -03 I-132

1. 78E -02

2. 91E-03 6. 4 9E -05 l-133 5. 01E -0 3 3'. 8 8E -0 3 E. 51E -0 3 1-134 2. 2 SE -0 2 4. 6 7E -0 3

1. 71E -05 l-135

1. 6 5E -0 2

2. 5M -03

5. 41E-04 Kr - 8 32

1. 21E -05

4. 2M -05 Kr-8 5m

1. 5 2E -0 3

1. 9 6E -0 3 Kr - 85 4. 2 3E -0 6

3. 9M-04 Kr - 8 7 2.17E -02

1. 4 5E -0 2 Kr - 8 8

4. 3M-02

7. 3 6E -0 3 Kr - 8 9 5. 0 2E -0 2 3. 6 4E -0 2 Xe-131m 4. 01E -06 2. 4 5E -05 Xe-133m 3, 4 8E -04

1. 4 5E -04 Xe-13 3

1. 3M -0 3 5. 5 4E -0 3 Xe-135m

5. 91E-03 1.19E -0 3 Xe-135 i. 02E-02 1.17E -0 2 Xe-137

6. 31E-03 5. 0 3E -0 2 Xe-13 8 4. S SE -02 3.08E-02 2. 53E -01

1. 75E -01

1. 01E -02 7-20 4

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10 TIME AFTER SHUT 00WN. sac Fisuon-product decay heat power as a funcuon of tune after shutdown (from ANS-5.1/N18.6L Figure 7.1 i

e l

7-21