ML20091R600

Forwards Replies to Questions Re Relicensing of Michigan State Univ Reactor,License R-114

ML20091R600
Person / Time
Site:	05000294
Issue date:	06/11/1984
From:	Terry S MICHIGAN STATE UNIV., EAST LANSING, MI
To:	Thomas C Office of Nuclear Reactor Regulation
References
NUDOCS 8406150178
Download: ML20091R600 (49)

Text

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MICHIGAN STATE UNIVERSITY ,

VICE PRESIDENT FOR FINANCE AND OPERATIONS AND TREASURER EAST LANSING

• HIQIIGAN

• 48824 412 ADMINISTRATION BUILDLNG June 11, 1984 Mr. Cecil 0. Thomas, Chief .

Standardization and Special Projects Branch -

Division of Licensing U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

Docket 50-294, Formal Review Questions--Your Letter Dated May ll, 1984

Dear Mr. Thomas:

Pursuant to your request, I am enclosing 3 copies of the. replies to your questions regarding the relicensing of the Michigan State University Reactor, R-ll4. Additionally requested material (Revisions to the Safety Analysis and revised Technical Specifications) are being prepared and will be submitted shortly.

V ry truly ours.

R.)

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Stephen,H. Terry i Assistant Vice President for Fin e SHT/bbc/0398F cc: Bruce Wilkinson

, Enc.losures STATE OF MICHIGAN ss. .

COUNTY OF INGHAM Subscribed and sworn to before me, a Notary Public, this 11th day of June,1984. .

g- g y _

Notaff Public, Ingham Co., Mich. Q My p ission Expires March 6, 1985.

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Michigan State University Docket 50-294 Formal Review Questions, MSTR J

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Reference:

NRC Letcer Thomas -> Wilkinson 5/11/84) {

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1. What are the principal uses of the MSTR? What is the current use in f megawatt-hours per year? how frequently is the reactor pulsed? l I

The three principle uses for MSTR are neutron activation analysis, f isotope generation and engineering training. Between January 1, 1983 and I

December 31, 1983 the total operation of the reactor was 5166 kilowatt-hours. The reactor is pulsed semi-annually with at least 1.0% Ak/k.

2. Provide a plan view of the current core configuration showing the number {

e and location of the fuel elements (differentiate between 8.5 and 12 i weight-per cent elements), control rods, graphite reflector assemblies, experimental tubes, and the startup source.

Figure 1 shows the current core configuration. All positions are filled with fuel elements except those marked "G", representing graphite reflector elements, the three control rod positions, marked reg. shim, and pulse, and the source position marked source. Currently there are no -

experiment tubes in the core grid plate except the central thimble, marked C.T. j

3. What is the total 235U content in the core? What is tl.a current startup i source and what is its strength? j I

The total 235U content of the core is 2528 grams. The startup source is an Americium-Beryllium source of 1.88 curies. 1 h

4. What is the fuel to moderator ratio for the current core?

235 U in core 25289 l 1487069 brincore(Zr- 5

- 8706) 6.4759 x 10 Atoms U[U=25 x 6.02 x 1023 - 6.4759 x 1024) 6 x 6.02 x 1023 9.8374 x 1026]

9.8374 x 10 Atoms ' Zr (Zr = 9 1.67 x 1027 Atoms H in core Fuel (9.8374 x 10 26 x 1.7 = H] .

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core volume = (H) = 4.127.64 in ~

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=4.127.64-91[ A 4 14  !

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= 4.127.64 - 2251.34 = 1876.3 in = 307479 HO 2

23 27 Atoms H in core water [30747 x 6.02 x M x2=M 2.057 x 10 33, 1 L

i, Total core H = HFuel + H, '

27

= 1.67 x 10 + 2.057 x 10 = 3.73 x 10 Fuel to moderator ratio = t 24 27 235U in core t H in core = 6.4759 x 10 1 3.73 x 10 {

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= 1.736 x 10 f s

5. What are the excess reactivity and control rod worths in the current core? I' When k = 1 the reactivity inserted in $3.83 or 0.02681 Ak/k. Total reactivity' of the current core is $6.83 or 0.04781 Ak/k. Therefore excess reactivity is $6.83 - $3.83 = $3.00 or 0.021 Ak/k.

The individual control rod worths are:

Reg = $1.85 - 0.01295 Ak/k Shim = $3.03 - 0.02121 Ak/k Pulso = $1.95 = 0.01365 Ak/k

6. Provide schematic drawings of the primary cooling system and the water ,

purification system.

These f Figure 2 shows the cooling system and water purification shunt.

two systems are interdependent.

3

7. What isotopes are found normally in the reactor pool water and in what concentrations?

The pool water analysis of February 2, 1984 shows that gross S exclusive of tritium is 2.52 x 10-9 p ci/ce. The tritium analysis on that date  !

shows 1.1 x 10-4 p Ci/cc.  :

8. What are the normal evaporation losses from the pool? Describe the water f make system.

Normal evaporation loss equaled 51.6 gallons per month averaged between September 1983 and March 1984. Distilled water is brought into Room 184 in a 50 gallon carboy. The water is drained manually to the pool.

9. Is there more than one conductivity monitor in the coolant purification system? Where are the readouts for the conductivity.monitorsf How often l

are the conductivity monitors calibrated?

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b There are two conductivity cells, one before and one after the dominera- jf lizer on the filtration shunt of the primary water system. The monitor ,

I readouts are at the console. The cells are compared with each other as part of the daily check list. The cells are not calibrated unless [

y readings are suspect.

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10. Provide a description of the heating system at the MSTR.

There are two sources of heat in Room 184. Building air is brought in j through the console room and outside air is brought in through the unit  !

i ventilator on the east wall of Room 184. In the ventilator, the air moves through thermostatically controlled heating coils, past automatic dampers and inte the room.

11. Provide a schematic drawing of the ventilation system by-pass through the absolute filter. How are the monitoring system and the damper function monitored and controlled from the console? ,

Figure 3 is a schematic of the ventilation by-pass. Damper function is i j

monitored and controlled at the console. The operator has a button marked " vent" with dual lights to indicate damper position. {

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12. W at kind of fire protection systems are available at the reactor  ;

facility? W ich fire departmants respond to fire-related emergencies at the reactor facility? W o is responsible for maintaining fire extin-guishers at the facility?

l Fire extinguishers are on site. A fire hose is available in the hallway. l The City of East Lansing operates a fire station 3 blocks west of the i Engineering Building. The emergency plan contains a letter of agreement  ;

with the City of East Lansing Fire Department to provide fire protection l

support. The Department of Public Safety through the Office of the Campus Safety Engineer maintains the fire extinguishers. l,

13. Wat kinds of communication systems (intercom, internal telephone system, i commercial telephone services, and so on) are available at the reactor l facility?  !

- t Commercial telephone system is the only commsunication system at the  ;

reactor facility. .

4

14. Describe the compressed air system (s) at the reactor facility, including ,

details of installation and uses. .

Figure 4 shows a schematic of the compressed air system to Room 184. The ,

only use for compressed air in Room 184 is to drive the pulse rod.

15. How many spare fuel elements are there? How many have been irradiated?

Were are they stored? ,

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t There are 14 total spare stainless steel fuel elements in reactor pool storage racks. The 14 elements are stainless steel clad 8.5% by weight.

. In addition, 12 aluminue clad elements are stored in the reactor pool storage racks (these elements will not. be used in the core). All spare fuel has been irradiated.

16. What volume of spent ion exchange rasins are generated annually from the coolant maintenance operations? What kind of radioactivity levels have been observed in these estorials in the past?

'The io'n exchange resin was last reolaced in 1973. Our records indicated very low radiation levels at that time. Because we use distilled make-up water we do not foresee changing the exchange resin the near future.

There is approximately 4 cubic feet of resin in the ion exchange resin canister.

17. Describe the liquid radweste management program.

There is no significant radioactive waste produced except occasional liquid irradiated samples.

18. Describe the solid radweste management program.

All solid redweste from room 184 is accumulated and held at the reactor facility until time of shipment. At the time of shipment a DDT approved 55-gallon steel drue, lined with 4 mil poly liners, is brought to the reactor facility. The drum is packed, monitored, labeled and loaded on the commercial waste hauler's truck and sent to a burial site at Richland, Washington.

19. Summarise the quantities of liquid and solid radioactive waste resulting from reactor operations for the last 5 yr (total activity of each physical fore at times of release or shipment for each year).

There were 3 pickups of solid radweste from the reactor facility over the past 5 years. There were no liquid pickups. The solida consisted of rubber gloves, paper and used sample holders. The total activity of each pickup was < 0.05 pC1/ge of material.

20. Describe the facility electrical pwer systes and list all controls and instrumentation that are provided with emergency back-up power.

1 Normal electrical power is 110/220/440 V line power. The following systems are on an emergency back-up power system as wollt

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a) Intrusten alare  ;

b) Area stare , ,

c) particulate alare d) Argen-41 monitst 4 l

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There are 14 total spare stainless steel fuel elements in reactor pool  !

• storage racks. The 14 elements are stainless steel clad, 8.5% by weight, f In addition, 12 aluminum clad elements are stored in the reactor pool storage racks (these elements will not be used in the core). All spare fuel has been irradiated. , 4

16. What volume of spent ion exchange resins are generated annually from the g coolant maintenance operations? What kind of radioactivity levels have i been observed in these materials in the past? I The ion exchange resin was last replaced in 1973. Our records indicated (

i very low radiation levels at that time. Because we use distilled make-up i

water we do not foresee changing the exchange resin the near future.

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! There is approximately 4 cubic feet of resin in the ion exchange resin canister. }

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17. Describe the liquid radweste management program. e There is no significant radioactive waste produced except occasional g 1 liquid irradiated samples.

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18. Describe the solid radweste management program. t i

All solid radwaste from room 184 is accumulated and held at the reactor j facility until time of shipment'. At the time of shipment a DOT approved 55-gallon steel drue, lined with 4 all poly liners, is brought to the b reactor facility. The drum is packed, monitored, labeled and loaded on l

the commercial waste hauler's truck and sent to a burial site at ft 1 Richland, Washington.

19. Summarine the quantities of liquid and solid radioactive waste resulting 'f

' from reactor operations for the last 5 yr (total activity of each t physical form at times of release-or shipment for each year).

! There were 3 pickups of solid radweste from the reactor' facility over the [

l past 5 years. There were no liquid pickups. The solids consisted of rubber gloves, paper and used sample holders. The total activity of each j pickup was < 0.05 pC1/gm of material.

20. Describe the facility electrical power system and list all controls and [

f: instrumentation that are provided with emergency back-up power.  !

a Normal electrical power is 110/220/440 V line power. The following systems are on an emergency back-up power system as wollt ,

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[ a) Intrusion alarm  ;

b) Area alarm c) Particulate alarm l d) Argon-41 monitor ,

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21. Provide definitions for the following terms as used in your Technical

. Specifications.

'(1) Secured experiments (2) Nonsecured experiments (3) Movable experiments (4) Irradiations i

Secured E..periment

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A secured experiment is any experiment, experiment facility, or component of an experiment that is held in a stationary position relative to the reactor by mechanical means. The restraining forces must be substantially greater than those to which the experiment might be subjected by hydraulics, pneumatic, buoyant, or other forces which are normal to the operating environment of the experiment, or by forces which can arise as a result of credible malfunctions.

Nonsecured Experiment A non-secured experiment is one which does not comply with the

' requirements set forth for a secured experiment.

Movable experiment A movable experiment is one where it is intended that the entire experiment may be moved in or near the core or into and out of the core while the reactor is operating.

Irradiations Exposure of samples to neutrons and/or gamma radiation in either rotary specimen rack, central thimble or other experimental assembly.

22. Describe the administrative organization of the radiation protection program, including the authority and responsibility of each position identified.

The administration organization of the Office of Radiation, Chemical, and Biological safety is listed below as a flow chart of authority.

9 Director, and Safety Officer 1

Health Physicist 4

Radiation Monitor The authority and responsibility of each position is_ detailed as follows:

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a Radiation Safety Officer

. Evaluate exposure' records and determine required changes in procedures, equipment, and shielding to reduce hazards.

Serve as the University's radiological safety officec.

Determine that radiation usage is in compliance with the U.S. Nuclear

. Regulatory Commission and the Michigan Department of Health regula-tions. .

Approve all procedures and changes in University policies that may involve personnel radiation exposure.

Supervise all aspects of University radiation protection and measurement activities such as personnel monitoring, x-ray monitoring, maintenance j of exposure records, and sealed source leak testing as required by the

U.S. Nuclear Regulatory Commission licen'se and state regulations.

. Inspect areas where radiation is used to identify radiological health j and safety hazards.

Prepare and disseminate information on radiation safety practices and .

procedures to staff and students. l Develop and supervise an environmental radioactivity monitoring program i for the University.

j Supervise the radiation ' testing and monitoring laboratory.

Organize and train radiation monitoring teams which are responsible for radiation monitoring and control in the event of an emergency.

l Act as liaison with University architects and engineers for the purpose

! of including radiologically nafe construction in all new and remodeled l- buildings in which radiation usa'ge is planned.

j consult with equipment manufacturers ar.d suppliers regarding the design 8 and installation of radiation facilities.

Develop and supervise a radioactive was e collection and disposal 4 system.

j Develop control systems and records of receipt, transport, testing, and

! disposal of radioactive materials.

Develop and administer program for obtaining U.S. Nuclear Regulatory .

. Commission licenses and State of Michigan authorizations for the use

! of radiation facilities and supplies.

j Participate in radiological health and safety programs and conferences. ,

Prepare and submit budget needs for the radiological health and safety i, functions.

1 Authorize the purchase of equipment and supplies.

Health Physicist Assist in the development and coordination of a radiological health and safety program. .

Assist in the development of an expanded bionssay program, covering a comprehensive range of frequently used radionuclides on-campus.

Coordinate nuclear emergency planning'and procedures with' appropriate j University units. .

coordinate daily operational activities for control,' receipt, moni-  ;

toring, transport, testing, and disposal of radioactive materials.

I Develop radiation monitoring procedures for University facilities, 1 i including medical and research.

! Maintain liaison with computer personnel in the development and in-j plementation of ' computerized systems / services.

Act as a resource' authority regarding administrative unit policies and procedutos.

Train support staff in new or revised work methods and procedures.

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Interview and recommend employment of support staff.

Evaluate and review performance of support staff.

I Radiation Monitor Survey radiation personnel, facilities, and work areas to monitor i radiation levels and detect radioactive contamination.

Distribute, collect, and record data from personnel monitoring equipment

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such as various film badges and dosimeters.

Calibrate radiation measuring instruments and develop new methods of

{ calibration.

Advise University employoas using radioactive materials, equipment, and areas.

Collect samples and analyze and maintain records on air, water, and

! population radiation exposure.

+ Inspect incoming radioisotope shipments and notify delivering carrier and

U.S. Nuclear Regulatory Commission of contamination.

Perform bioassays to determine internal disposition of radioactive compounds and institute corrective measures when necessary.

! Conduct training and lecture programs for radiation workers.

Maintain up-to-date knowledge of U.S. Nuclear Regulatory Commission and other regulations.

Serve on the Radiation "On Call" lists for campus radiation emergencies.

Prepare and maintain appropriate computer and other records..

! 23. Describe any radiation protection training for the non-Health Physics

staff.

i A slide presentation using a narrative cassette in used by all police and fire fighting personnel to inform them of emergency procedures. A visit

to the reactor facility by the same personnel is used to familiarise them j with the area. Slides, videotape and lectures are available to them and i other non-emergency staff for general radiation training. The presenta-l tions are one hour in length'to allow for questions, answers and general j discussion after the planned presentation. On going training, specifi-

! cally for police and fire departments includes =an annual tour of the reactor lab.

1

24. Summarise your general radiation safety procedures. Identify the minimum-

frequency of surveys, action points (levels), and appropriate responses.

Radiation surveys are done as needed on an emergency basis.- Regular radiation surveys take place every other week. Action is required on any I area that registers two times background-action varies from complete

clean-up to explanation of higher reading (i.e.,usample storage area).

i -

! 25. Describe your program to ensure that personnel radiation exposure and releases of radioactive material are maintained at a 1evel.that is "as

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! low as reasonably achievable" (ALARA). Identify steps taken to implement l the ALARA principle.

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F Routine bi-weekly surveys are conducted of the reactor area, control room I and associated labs to ensure there is no spread of contamination. Area -

and particulate monitors as well as environmental monitors have been installed. Personnel film badges are changed and read biweekly. The ,

reactor is shielded to minimize exposure in the office, control room and i surrounding passages.

26. Describe any gaseous and air particulace effluent sampling equipment with l

respect to sampling location, sampling rates, and probe geometry.

Argon-41 monitoring System probe geometry: side window G-M-Tube location: Probe - in room 198 monitor readout - console, room 184 t stack flow: 700 CFM sample rate: several cfm Air particulate monitoring system i

probe geometry: end window - G-X-Tube location: probe - room 184 monitor readout - console, room 184 sample rate 2.5 cfm

27. For the fixed-position radiation and effluent monitors, describe the generic types of detector and their efficiencies and operable ranges, also specify the methods and frequency of instrument calibrations and

. routine operational checks.

Monitor Type of Ranges Efficiency Calibration Detector and Frequency Area Ion chamber 0.1 to 10,000 MR/HR Known source (Co60), annual ~

i Air- G-M Tube 20 to 200,000 counts / min NBS standard source, annual particulate l Argon-41 G-M Tube 1 to 999,000 counts /30 sec known source (Ar-41), annual 1 i

Criticality Ion chamber 1 to 15 mR/NR known source (CS-137)' annual l

Operational checks Type of Check Frequency Monitor (s) l Alarm check daily check list air and ares Source calibration daily check list ares

! H.V. operation check daily check list air i Background count check daily check list Ar-41, sir, ares Alarm check, source weekly check list criticality cal.

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28. For the radiation monitors that are alarmed, specify the alarm set points and indicate the required staff response to each alarm. ,

Alarm Set Point Response (for all alarms) .

Air particulate 10 K cap 1. Scram reactor Area 100 mR/ hour 2. shut off primary pump -

Criticality 15 mR/ hour 3. pull building evacuation alarm -

4. evacuate room, locking doora

5. seet members of the DPS, Fire Department and Reactor Safety i Committee at the northwest corner .

' of the Engineering Building for re-entry procedure.

' 29. Identify the generic type, number, and operable range of each of the l

portable Health Physics instruments routinely available at the reactor installation. Specify the methods and frequency of calibration.

Quantity Range Calibration & Frequency Type 4 5 to 500,000 NBS Source - semi annually

, G-M Tube

! pulse rate counts / min meter Ion chamber 2 0.1 to 199,000 MR/HR Known source (CS 137) and semi-annually

' O.1 to 100,000 MR/He

30. Describe your personnel monitoring program.

Film badges are the only in-house personnel monitoring system. They are changed bi-weekly and supplied by R.S. Landauer Co., Chicago, IL.

31. Provide a summary of the reactor facility's annual personnel exposures

[the number of persons receiving a total annual exposure within the designated exposure ranges, similar to the report described in 10 CFR 20.407(b)] for the last 5 yr of operation.

Number of Individuals in Each Range Whole-Body Exposure Ranae (Rees) 1979 1980 1981 1982 1983 Total i No measureable exposure 79 49 125 57 69 375 8 2 0 10 l

Less than 0.1 1 1

7 i 0.1 to 0.2 0 7 O O O O.2 to 0.3 0 1 0 0 0 1 i more than 0.3 0 0 0 0 0 0

32. Describe your environmental monitoring progrant summarize the results for past 10'yr and compare recent measurements with any performed before the initial reactor criticality. Provide your analysis of the environ-mental monitoring results.

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Environmental Radioactivity Data Environmental radioactivity data have been collected on a continuing l Included are air, water and popula- j basis for the past several years.

tion exposure measurements. The latter measurements are obtained by the j l

use of a new calcium sulfide thermoluminescent (TLD) dosimetry system i which enables precise determination of doses for comparison with i governmental control reference dars. The long term record enables an l evaluation of contributions of radioactive material releases by campus i nuclear facilities. The university is required to estimate the popula- -

tion dose resulting from nuclear facilities for which it is licensed. ,

Summary of past 10 years:

Quarterly Airborne Water (pci/cc) Quarterly Average Particulate gross gross Averagel Background Year (uci/ce) a 8 TLD (mR) TLD (mR) 1983 5.58E-15 1.36E-10 1.46E-8 35.28 23.52 1982 1.42E-15 2.08E-10 2.35E-8 42.51 21.12 1981 2.56E-15 2.70E-10 2. 50E-8 34.48 20.62 1980 3.61E-15 3.68E-10 4.34E-8 29.35 20.78 1979 2.37E-15 4.50E-10 3.45E-7 45.92 16.26 1978 1.91E-14 1.8 E-10 3.14E-7 28.56 14.28 1977 5.77E-13 1.15E-11 1.12E-8 27.14 19.97 l

3 1976 1.76E.13 4.08E-11 2.54E-8 9.85 12.11 1975 9.18E-15 6.08E-11 4.26E-8 13.60 16.55 1974 4.06E-14 3.44E-11 2.61E-8 11.48 11.10

, 1973 1.74E-14 4.3 E-11 3.4 E-7 7.60 6.48

! The date of initial reactor criticality is: March, 1969.

In 1976, the Office of Radiation. Chemical, and Biological Safety re-calibrated and repaired the TLD measuring equipment which explains the sudden rise in TLD readings.

l

33. Comment on the ability of the reactor components and systems to continue to operate safely and withstand prolonged use over the term of the requested license renewal. Include the potential effects of aging on I

fuel elements, instrumentation, and safety systems.

f Reactor es uponents have withstood approximately 20 years of exposure with little failure. The thermocouples in the fuel typically fail after several years and this necessitates the installation of backup thermo-l couples or the replacement of instrumented fuel assemblies. The distortion of fuel has been minimal at the present operating conditions both at MSU and at other TRIGA facilities. Reactor instrumentation is relatively old and experiences some component failures. The design of reactor instrumentation is such that components can be readily removed and replaced. Redundancy in safety systems assures that a system safety 1 Includes background. Nessured at reactor exhaust stack.

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is maintained even with component failure. Mechanical safety systems have experienced essentially no failure during the MSU experience and it is believed that continued operation will be possible.

34. What is the esasured temperature in a B-ring fuel element immediately following a 1.4% Ak/k (2.00$) pulsef What is the nazieue measured temperature in the element and how long after the pulse does this maxieue temperature occurf on January 26, 1984 we pulsed the reactor with a $1.98 reactivity insertion (1.39% Ak/k). The fuel temperature meter showed temperature of just over 250*C. The temperature was sustained for approximately 2 seconds.

35. What is S effective for the MSTR current core configurationt 8 effective is 0.007 i i 36. In the steady-state mode, why is the interlock that prevents the 1

application of air to the safety transient rod unless the " safety trar.sient rod cylinder" is full "in" not listed in the proposed Techni-i cal Specifications? Justify.

j j In the steady state mode, the transient rod cannot be withdrawn unless j the res & shie rods are fully "in." Thus the reactor is suberitical.

j Since the maxieue reactivity insertion by the transient rod occurs when the rod is withdrawn from the full "in" position, any reactivity l

j insertion from a partially withdrawn position will be less than the situation normally encountered (i.e. startup with transient rod fully "in.") Thus, no interlock is necessary.

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37. The proposed Technical Specifications state the control red drop time j for the slowest rod as 2 s. The old Technical Specifications give a rod j drop time of less than 1 e (except for the pulse rod). Justify this longer time interval.

i i The proposed 2 second time is a simplification of the original require-j ment of I second for Reg and Shis rods or 2 seconde for pulse rods. The j pulse rod has always been the slowest one and this was defined as the

limiting case. We will revert to the 1 and 2 second base as requested.

l There has been no problem with compliance under the current license.

i

38. What are the actual numbers of 8.5 and 12 weight-per fuel elements l

loaded into the core? Where are they positioned? What administrative 4

limits or requirements are placed on fuel loadings?

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! There are eight 12% by weight fuel elements in the core. Six'in the

, B-ring and one in the C-9 position and one in the C-11' position. The

l. other 62 elements in the core are 8.5% by weight.

. The administrative requirements of fuel loading are as follows:

i

! 1. All core loading must be done under the supervision of a licensed

! . Senior Reactor Operator.

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2. During core loading, the normal procedures of reactor operation must be observed, includingt ,

s. Daily and weekly check liste must be completed with no abnormal 1 circumstances.

b. Roos must be isolated (doore and windows closed). {

i

! 3. The pulse and shie rode must be in-an up (or cocked) condition

} during critical loading or movement of more than one element to t

] serve as a safety system.

i j 4. A licensed reactor operator must be at the console during core  ;

loading to determine critical core configuration when more than one j element is being moved.

(

f 5. No aluminue clad fuel will be loaded into the core without prior j safety analysis written and approved by the Reactor Safety Commit-

. tee and the NRC.

i 6. There are no special restrictions on the location of 12 wt.% fuel in the core.

[

! 39. What is the annual release rate of A'I r from the reactor facility to the

! environment, What is the noresi 4IAr concentration in the emperimental are.e, i

The annual relece rate varies from year to year depending on the amount >

of time the reactor is running. A typical release rate is 400 pC1/ year.

The " normal" A41 concentration in the reactor room when operating at t y full power for extended time is approximately 4 x 10-8 pC1/cc.  !

40. Specify assumptions and provide calculations of the expected radiation .

levels at the following locations. L

a. Beyond the limits of the reactor facilities (as a result of all ,

airborne releases of radioactive materiale from the reacter facil-

, ity) t b. In the reactor room (as a result of 'I r Aconcentrations when such .

I concentratione are a maximus)  !

i

c. Above the reactor pool (as a result of the manieue I'N roleseos .

I possible) i j d. In the reactor room (because of continuous espesure to I'N and 'IA r for some specified period)

(a) Typical A41 release rate = 400 pC1/yr = 20 pCi/ Full power Operation he

4 . ** ,00 m .0 ,.. , . 3,. , - > 4 > < ' *c u -  ;

! = 40% of Mpc to unrestricted area j If Mpc e 300 er/yr or h = 0.09 er/hr l

Then does rate to public = 0.4 a 0.09 = 'J.036 er/hr i

. 5/s4

F ,

-6 -

(b) Max A conc = 2 x 10 gg,,,,,,,,gg,gg) e 5000 er/2080 hre = 2.5 er/hr (c) N-16 in pool water shielded from operating area by 10 ft of stagnant water.

N-16 release rate in outgassing (assume sa=w en.tgas rate as A41)

Aton Ration A/0/N = 1/40/158 in dissolved air Fast $ = Thermal $

4 rate of prodn yl6 , g ,0,) 0.017 x 10" harne) = 1.1 x 10'3 rate of prodn A'I 1 0.62 barns rate of decay N16 T 4/2 of A 1.83 x 3600 "

rate of decay A'l " T r ui I'l3 1/2

/. Ameission rate of N16 e that of A41 if outgassing occurs without delay. (delay likely and will reduce N16 due to decay)

/. Max release rate NI6 = 1.6 x 10-8 pC1/cc air e

This is 53% of NPC to uncontrolled areas and equivalent to 0.048 er/hr (d) Sun of A'I + N16 = 0.036 + 0.048 = 0.084 er/hr

41. Provide justifications for not having a reactor pool water level alare.

Loss of pool water very unlikely due to facility corstruction and location. Small (few feet) loss of pool water would not generate safety hasard. Radiation monitors over pool give ample warning of radiation hasard.

42. What controls are placed on the spare aluminue elements currently located in the reactor pool storage racks surrounding the reactor core I

to prevent them from accidentally being loaded into the caret Administrative record keeping to assure that proper fuel loading is being performed. All fuel elements are numbered for identificatAon.

43. Provide a block diagram of the nuclear instrumenta control and set points. It should include the following information.

(1) Type of channel (2) Detector (3) Range of detector (not meter readout)

(4) Autenetic controle (5) Set points (minieve and easieue where appliesble) 17 5/84

Set Points Fission Fm Amplifier 8 2 cys (mia)

Counter Amelifiet Count Batt , ,

interlock Ing n Escorder re ced ion Iag a Feriod i 6 7 sec (min) hr Amplifier Circuit E Ed Amplifier W .g s

i*naar FWR I I harh"

+

8 r- I SCRAM $ To Rods Ion Power Inval , 8  % Power 6 10% of Scale Chamber Moaltor ,' [ _ (max)

-- - - - - i + - - 3 Resulatina

,Fo r - r _ _ _ _ _ _ _ l _ ' + _ _ d_%1er _

,F---- ,od

• t=1- - J Fuel Temperature & 450*C (max)

C 1

- o so c h [ T e rature F W (AA=inistrative)

Figure SA. Block Diagram of Reactor Insteunentation' for Steady-Sta:e Operation I

Set Points PULSE PEAx power g y I

10M CHAMgER AM PLI FI ER G UNEAR RECORDER

{ MIP l y yn gggg , gug I Y I LOG R ECOR DER l

FUEL TEMPERATURE

h. 450'c (max) 4 5 TEMPERATURE PROSE 50*C (Administration) 4 i

Figure 5B. Block Diagram of Reactor Instrumentation for Pulsing Operation E

t 10' .

4 T 1 iO' - 1 I

.L UNCOMPENSATED 10 7 - ION CHAMSER ,

PEAK POWER (PULSE POWER)-

SAME CHAMSER Id -

d - -- -- --

5

' UNCO' MPENSATED 38 . 10N CHAMSER PER CENT POWER i

(STEADY-STATE OPERATION) 3 D -

-COWENSATED 10N CHAMSER LINEAR RECORDER

, g , (STEADY-STATE OPERATION)

G #

!' h 8d '

COMPENSATED 10N CHAMSER J IO* CfSEC LOS n RECORDER

{s (BTEADY-STATE- OPER ATION)

,I -

W FISSION COUNTER g

COUNT-RATE CHANNEL

,. g6a ,

OURCE LEVEL I C /SEC 4

16

[. Figure ~6. Detector Ranges i

20 5/84

)

i l

Figure SA + B are block diagrams of the instrumentation during steady

! - state operation and pulsing respectively.

Figure 6 is a graph of the detector ranges for the MSTR.

l 44. Provide a bar chart drawing of the operating ranges of the in-core l nuclear detectors.

Figure 6 is a bar chart of the operating ranges of the drtectors.

45. Please address the following accidents; include all assumptions made, calculative' methods used, and accident scocarios.

(1) Loss-of-coolant accident (2) Fuel element failure in air (3) Rapid insertion of reactivity (step nuclear excursion)

(4) Mechanical rearrangement of fuel (consequences of dropping 600-lb i

lead cask into core).

(See attached Safety Analyses p. 34-48)

46. What are the flow rates of the coolant through the heat exchanger system and the domineralizer loopf The present flow rate of the primary water through the heat exchanger is 150 gallods per minute. Flow through the dominerlizer loop is 3 gallons per minute.

. 47. Describe the 16N diffuser system at the NSTR.

The design of the primary water return from the heat exchanger provides for 16N diffusion. The cold water return is directed across the top of the core. The primary cooling system is always on during normal operation.

48. What is the current use of the storage vault and storage pipes in the pit next to the reactor in room 1847 .

The storage pipes are not used at this time but would provide emergency storage of fuel elements. The vault is used to store a small amount of tritiated water. It provides extra radioactive sample storage space.

( 49. Describe any equipment and services shared by the reactor facility with i other parts of the University.

i Facilities shared with other parts of the building include:

1. Pressurized air - building

2. Electricity - line power - campus l 3. Steam - for heat - campus

4. heter - to sinks - campus -

There is no gas service to room 184.

l l

21 5/84 -

.r

F i I

50. Describe the primary power supply to the campus and to the reactor facility. g j i Power plant no. 65'is owned and operated by MSU. It provides power and j, steam to the University. It has a 40 MW, capacity and generates 850,000 lbs of steam / hour. The plant is on intertie with Consumer's [!

Power Co. with which it exchanges power during peak demand periods. (

l 4

51. Provide detailed geographic data for the site, including location with i respect to area features, nearby industry, transportation, military and I

manufacturing facilities.

[

The following is a list of major geographic points and their direction )

and distance from the reactor lab. Note that East and South of the g reactor lab is mostly farm land.  ;

E I-496: 1 1/2 miles West  !:

1-96: 2 1/2 miles South I-69: 1 1/2 miles North y

• i Red Cedar River: 1400' (North)  ;

East Lansing downtown ares:- 4200' (North) L Grand River junction: 4 miles (West)

Lansing downtown area: 4 miles (West) i[

j Major industrial (Oldsmobile): 4 miles (West) '

C&O railroad: 3500' (South)

Grand Trunk railroad: (leased to passenger) = 1400' (South)

Lansing Capitol Airport: 7 1/2 miles (North-West) '

Abrams airport - Air National Guard post: 16 1/2 miles (North-West)

State Capitol: 4 miles (West).

52. Discuss the ability of the Engineering Building to withstand the most severe credible tornado. What would be the effects of tornado damage on the reactor room and on any safety-related equipmentf.

The Engineering Building was constructed to withstand winds to 88 aph.

Data received from T.T. Fujeta at the University of Chicago indicates that although East Lansing is in the southern half of lower Michigan,

{

an area of high tornado frequency, East Lansing has been historically free of tornados. From data collected between 1930 and 1974 all tornados passed North and West or South and East of East Lansing. . East. t

' ' Lansing is on the Northern edge of a pocket of relative calm. The storms to the North and West have all been of the F orF3 2 intensity on the Fujeta scale (113 to 157 mph). Only one tornado of the F4 or F5 intensity was reported between 1930 and 1974. This was about 10 miles North of the reactor site. It is estimated that, in the case of

several () 88 mph) winds, the non-supporting curtain walls of the

! . building would collapse'but the floor / ceiling structure of the building would survive. Damage to the below ground reactor core would be limited.

! to debris falling into the pool. Since the building is not designed as a containment structure for the reactor.. loss of curtain walls would

,- not be serious. Monitoring for radioactivity would be done by battery-operated instruments since electric power would certainly be dis .

~

rupted.

f 22 S/84-i-

i

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

.. = . . - - - - ,

, SHOCK ABSORDER VENT HOLES '

EXTERNALLY THREADED CYLINOER PISTON SUPPORT

-. P'1 F"1

~ , , .

~

.l I M I I L_ 1 3 J_1 -

BEARINGS g  : WORM WORM GEAR -

MLL NW u

HOUSING 1 tir-PISTON ROD

VENT

" SUPPLY  !

AIR SUPPLY {

HOSE l SOLENOID VALVE

-h% LIMIT SWITCH 80TTOM LIMIT '

= 1 0 CONTROL ROD h '

a Fig. 7. Transient-rod drive mechanism 23 5/84  !

e

• i 1

i

53. Provide more detailed information on the geology of the site and surrounding areas. Are there any faults in the ares? What is the history of seismic activity in the area?

I The North Western corner of In' g ham county lies near the center of the Michiganian basin. This is an interacratonic basin developed on precambrian crust. The basin is filled with primarily paleozoic sedimentary rocks approximately 15,000 feet thick.

l In the Lansing area, Bedrock, which consists of shales and sandstone of i the Saginaw formation and Pennsylvanian age, are covered with a layer

, of glacial drift approximately 75 feet thick. Bedrock is essentially underformed and faults do not off set younger formations.

Thsre is one recordable event in recent history. August 9, 1947 in coldwater, Michigan on earthquake occurred. It could be felt in

. Lansing at intensity IV and is the largest event reported in Michigan.

4 This event occurred along a fault line in the Coldwater area. It is

, estimated that this fault line could produce earthquakes of magnitude

{ IV. Coldwater is approximately 50 miles from East Lansing.

f 54. Describe in detail thp transient rod pneumatic drive system and tlie

! system operation in the pulse, scram and manual modes. Address the absence of an electromagnet.

j Figure 7 details the rod drive mechanism for the~ pulse rod. ' Operation of the pulse rod is controlled from the console.

! The drive is mounted on a steel frame that bolts to the center ch'annel'-

cover plate. Two steel covers keep the' mechanism clean. 'From zero to a maximum of 15 inches of rod may be withdrawn from the core; however, administrative control is exercised to restrict the travel so as not to exceed the maximum permissible step insertion of reactivity.(2.00 or.

1.4% Ak/k).

The transient-rod drive is a single-acting peneumatic cylinder with its

piston attached to the transient rod through~a connecting rod assembly.'

The piston rod passes through an air seal at'the lower end of the cylinder. Compressed air at 75 pounds per square inch is supplied to the lower:end of the cylinder from an accumulator tank mounted beneath t the center channels when a three-way solenoid valve located in the l piping between the accumulator and cylinder is energised. The com :

pressed air drives the piston upward in the cylinder and causes the rapid withdrawal of the transient rod from the core. ;As the piston; rises,- the air trapped above it is pushed 'out through vents at the

-upper end of the cylinder. ~At the end of its travel the piston strikes the anvil of an oil-filled hydraulic shock absorber, which has a: spring-return, and which decelerates the piston at a controlled rate over its j, last inch of travel. When-the solenoid is de-energised, the valve cuts-l .off the compressed-air supply and relieves the pressure in the l cylinder, thus allowing the piston to drop by. gravity to its original' position with the' transient rod fully. inserted in the reactor core. . No electromagnet is present to hold the rod in position as in the-reg and l ' '

_. 24 l 5/84 , ,

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

-. .- . . =

! shim rods. Instead pressurized air holds the transient rod up. The

- operator reads solenoid position at the console, not electromagnet contact.

Provision is made by raising or lowering the cylinder, thereby controlling the distance the piston travels. This adjustment deter-mines how far the transient rod is withdrawn from the reactor core, and thus determines the amount of excess reactivity inserted. The cylinder has external threads running most of its length, which engage a series of ball bearings contained in a ball nut mounted in the drive housing.

As the ball nut is manually rotated by a crank-driven worm sear, the i cylinder, which is. prevented from rotating, moves up or down, depending i on the direction the crank is turned. An extension of the crank drives i a mechanical revolution counter that indicates the position of the cylinder and also indicates the distance the transient rod will be ejected from the reactor core. As soon as the cylinder has been positionad, the crank is removed from the mechanism and placed in storage to prevent unauthorized operation.

4 Attached to, and extending downward from the drive housing, is the rod guide support, which serves several purposes. A bar attached to the bottom of the cylinder projects through a slot in the rod guide and prevents the cylinder from rotating. Attached to the lower end of the piston rod is a flanged connector that is attached to the connecting-rod assembly that moves the transient. rod. The flanged connector stops the downward movement of the transient rod when the connector strikes the damper pad at the bottom of the rod guide support. A microswitch is mounted on the outside of the guide tube with its actuating lever 1 extending inward through a slot. When the transient-rod is fully inserted in the reactor core, the flanged connector engages-the actua-ting level of the.microswitch, and indicates on the instrument console that the rod is in the core.

Pulse and normal steady state operation is the same for pulse rod. Rod interlocks are provided in steady-state mode to limit upward travel of the pulse rod only when both reg s'nd shim rod are in the fully inserted position.

All',scrans apply to the pulse rod as do to the reg and shim in both pulve and steady state mode. However, in the pulse mode a 15 second-tim 4r will insert the pulse rod is well.

55.. Provide a brief description of the Radioisotope Committee and the Office of Radiation, Chemical and Biological Safety, including member-ship and general. responsibilities.

+

The Radioisotope Committee is an advisory committee to the University sanctioned to help'the-University meet' federal and state regulations as regards radioactive materials. This committee gives approval to qualified individuals and their laboratories to'une radioactive materials on campus. The committee also develops policy and guidelines for the university in the use, transport, receipt and shipping of radioactive materials. Current membership on the committee includes I

faculty membersffrom the departments of Biochemistry, Cyclotron, Food

+

'25~ 5/84 y _ - - , , , _ - . - - , . . --

- .m , - - ,- 4 - - . - - , _ . .

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

- lI ll l

I Sciences, Animal Science,and Biology as well as the campus Radiation  :

Safety Office, and Health Physicist. All ,of these people have.had i 2 extensive experience with the use of radioactive materials. Committee ,

membership is appointed by the MSU Provost.

The Office of Radiation, Chemical and Biological Safety (ORCBS) l membership was discussed in Question 22. This office is the enforce-

' ment' arm of the radioisotope committee. The office maintains the university broad license as well as the license from the State of  ;

' Michigan to operate i~onization" equipment and cyclotrons. The office l l

provides radiological expertise to the committee, as well as other data that will be helpful to the committee in granting approval to indivi-duals and laboratories for the use of radioactive materials. ORCBS also provides health physics support to the reactor lab in the form of t-f expertise and back-up emergency equipment.

56. Describe selection and training of reactag and senior reactor opera-tors.

Nuclear reactor operators are personnel who have basic technical l

> education (physical science or engineering) and an interest in nuclear l j reactor operations. Most of the operators are undergraduate students in [

l engineering who are employed as part time. student assistants.

-Senior operators are generally people with at least a BS degree in a physical science or engineering who appear to have the qualifications and aptitude to serve in a more advanced position. Senior operators are generally full time employees of the University (rather than i students).

Training of operators involves self study of suitable text material and~

one-on-one tutoring by a Licensed Operator or the Reactor Supervisor and the-ORCBS. Wherever possible, formal training by MSU courses on Reactor Theory or in courses conducted at the University of-Michigan is used to supplement the individualized instruction.

57. Please provide clear, reproducible MSTR figures for the following:

1. Reactor Facility Layout Current Core Loading Diagram (See Question 2) 2.

3. MSTR Fuel /Noderator' Element Showing Construction and Dinensions

4. Cutaway View of MSTR Installation. .

~

5. Schematic Drawing of. Primary Coolant and' Purification '3ystema (See Question 6)

6. MSTR Domineraliser Loop .

j 7. ~ Ventilation Bypass System (See Question 11)-

l 8. Transient-rod Pneumatic Drive l '9. Reactor Instrumentation Block-Diagram for Steady-State Operations __

10. Reactor Instrumentation Block Diagram for Pulsing Operations

-11. Block Diagram of Nuclear Instrumentation and Setpoints (See '

Question 43) 12._ Figure 6. ' Engineering Building Floor Plan (From 1967:SAR)

13. Figure 20.- Block Diagram ~of Reactor Instrumentation (from 1967 corrected SAR) 0 '

12 6 :

___.-.u a _ u . _ , - , _ - , . . - _ _ _---~ ,- . . .

l 1

14. Figure 25, Loading Diagram (from 1967 SAR)

15. Figure 1. East Lansing Census Tracts (from March 1984 SAR)

16. Figure 2. MSU campus map (from March 1984 SAR) j e

See:

1. Figure 8

2. Figure 1

3. Figure 9

4. Figure 10

5. Figure 2

6. Figure 2

7. Figure 3

8. Figure 7 1

9. Figure SA

10. Figure SB

11. Figure SA and B

12. Figure 11

13. Figure SA

14. Figure 2

15. Enclosed I copy. Figure 12.

16. Enclosed I copy. Figure 13.

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MSU Engin:Gring Building,(Southeast Wing)

Offices - Offices

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Basement Figure 11.

5/84 31 _ _ . _ _ _ _ _ _ _ _ . . _ . _ . _ _ _ _ _

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1 SAFETY ANALYSIS 45-1 LOSS OF COOLANT ACCIDENT The MSTR operates at a maximum of 250 KWg and the present core contains 70 elements. The average power density, thus, is 250/70 = 3.57 l

KW g per element. A portion of the fuel contains 12 wt% uranium and, thus, is operated at a 41% higher power density than the remaining 8.5%

l uranium fuel. Thus, the maximum power density would be 3.57 x 1.41 or 5.03 KWth per element. Under-conditions speculated to involve the complete loss of power operation, it is estimated that the fuel temperature would I reach approximately 225"C (see Fig. LCA-1). Under st$h conditions the i estimated stress imposed on the cladding by internal gas is approximately

1200 psi (see Fig. LCA-2). This stress is considerably less then the approximately 37,000 psi yield stress for the stainless steel fuel i

i cladding. It is, therefore, concluded that the loss of coolant would not I

result in the failure of the fuel cladding and that the fission product containment would not be lost.

]

4 j Radiation Levels from Loss of Coolant If it is assumed that the reactor water is suddenly lost, the radi-l ation emitted from the contained fission products would potentially pose a l - threat to personnel in the reactor room and in Room 284 directly over the

reactor. The operating floor of the MSTR is 20 feet above the reactor core and the estimated radiation levels from both direct radiation and scattered j from the concrete room ceiling are given below for various decay times ,

immediately following full power operation.

l

! Calculated Radiation Dose Rates For Loss of Reactor Pool Water Direct Radiation Scattered Radiation i Time r/hr r/hr

10 sec 2,500 0.65 1 day 300 0.075 1 week 130 0.035 1 month 35 0.01 i-

34

, 5/84 [ j

>e _

2000 -

1800 -

COOLING TIME (SEC 0

1400 -

3 10 1200 -

I a ,

8  : .

Q 1000 5

W 800 -

1' 600 -

6 -

400 -

200 -

• ' i - a i~ n' =

0 i 0 5 to .15 20 25 30 35 40 45 OPERATING POWER DENSITY-KW/ ELEMENT EL-1872-~

! Fig. LCA-1. Maximum fuel temperature versus power density %

after loss of coolant for various cooling j timas between reactor shutdown and coolant loss.

35 .

. 5/84:

[ __ ,

t 10' -

ULTIMATE STRENGTH

---~_.%s' s N N YlELD STRENGTH \

\

\

10 - \

\

\

\

N S \

m

= \

N m

\

N

\

SinESS IMPOSED ON CLAD 103 -

2 10 , , , _ , , , ,

400 600 800 1000 1200

. TEMPERATURE (*C)

EL-1873 Fig. LCA-2. Strength and applied stress as a function of temperature, U-ZrH uel, uel and clad at same temperature. 1.65 36 5/64

)

The~ table indicates that, except for the direct beam from the core radiation exposure in the reactor room would be high but tolerable even immediately after loss of coolant and that emergency operations could be ,

1 carried out with limited time of action.  !

Of greater concern is the radiation levels in the uncontrolled area, l Room 284, located directly over the reactor. The distance from the core to the occupied area of this room is 30 feet and the radiation levels i calculated above (for 18 feet) will be reduced by inverse square (18/30) = 0.36x and the shielding of the concrete floor (approximately 6 inches) = 0.2x.

On this basis, the direct radiation in Room 284 is estimated to be Direct Radiation Time r/hr 10 see 180

I day 22 l 1 week 9 I

1 month 2.5 These levels of radiation are sufficiently high to necessitate immediate evacuation of the classroom in case of total loss of pool water.

The integrated exposure to personnel in this room as a function of time delay in evacuation is given below.

i Time, Min. Radiation Dose, R 5 14.97 15 44.77 30 '89.07 5 45 132.9 ,

60 176.3 Evacuation of Room 284 following loss of pool water during or shortly after reactor operation (the came assumed above) could most certainly be done in a time much less than an hour since a reactor operator on duty could assure building evacuation in such a catastrophe. If pool water were to be lost during non-working. hours, a longer evacuation time might be experienced as the result of the time delay inherent in emergency _ response.

The radiation levels would be reduced in'this case however since-the short

' lived fission products would have decayed away. As an example, assume a 6 l

l 37-

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l 1

hour shutdown period followed by loss of pool water and a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> evacuation delay during which time Room 284 is occupied. The exposure to a person in such a situation would be 132 R. Such an exposure, although high, would not be life threatening, t

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SAFETY ANALYSIS 45-2 FUEL ELEMENT FAILURE IN AIR In this analysis, it is assumed that one of the 12 wt% Uranium fuel elements is ruptured in air after a long (1,000 day) exposure to full power (250 kwth) Operation of the reactor. The radioactive material released to the air consists of 100% of the rare gases and halogens which, during reactor generation, have migrated to the gap between the fuel meat and cladding. No credit is assumed for deposition of the isotopes on walls, in pool or in any other manner.

(a) Exposure to reactor room occupants. In this case, the rare gases and halogens are assumed released to the reactor room, and the ventilation in the room is assumed to be zero. A room occupant is assumed to remain in the room for 10 minutes while evacuation takes place.

The fission product gas concentration in the core operated at 750 kwth for 1,000 days is given below (from Lamarsh, p. 535--scaled to MSTR power output). (Note: core loading 70 elements, power density in 12 wt% fuel = 41% greater than core average.)

Curies, Single Curies, Single Radioisotope Curies, core Element, Avg. Element 12 wt%

Kr85m 2800 40 56.4 Kr85 98 1.s 1.97 Kr87 5,000 71.4 100.7 Xel338 400 5.71 8.06 Xe133 14,320 204.6 288.4 Xe135" 2,220 31.7 44.7 Xe135 3,900 55.7 78.6 I131 5,850 83.6 117.8-1132 8,270 118.1 166.6 Il33 14,300 204.3 288.0 Il34 15,200 217.1 306.2 I135 13,550 193.6 272.9 i

j 39 5/84

For fuel element ruptures in air, the gaseous fission product inventory in

, the gap is estimated to be 1.5x1.0-5 fraction of the total gaseous fission products (GA 4314). If all of this fraction is released into the reactor roca (volume 2x108 ) with no air change rate, tne concentration of radioisotopes in the room air is:

i Curies x 106 x 1.5 x 10-5 / 2 x 108 = pCi/s1 4

Assuming a person occupies the room for a total of 10 minutes following fuel rupture (while evacuation takes place) and that a respiration rate of 3.47 x 10-4 m 3/sec (Reg. Cu.ide 4) is experienced, it is possible to project the personnel exposures and isotope ingestion. The following equations were used: ,

"#* **

• E *"

i 1. Whole Body Exposure = MPC (Table 1, 10CFR20) x x

yr "I" 2000x60 yr

(Assumes that continuous exposure to 1 MPC is equivalent to 5 rems /yr whole body dose)

2. Isotope ingested = conc. of radioisotope x 3.47 x 102 x 600 see The results of such a calculation for a 12 wt7. element are given below.

Conc. MPC 10 Min.

Radioisotope pct /ml Table 1 , 10 Min.se Dose, Ingestion, pCi i

Ke85m 4.23x10-6 6x10-6 0.29 0.88 Kr85 1.48x10-7 lx10-5 0.006 0.03 Ke87 7.56x10-6 - 1x10-6 3.15 1.57 ye133e 6.04x10-7 1x10-5 0.02 -

0.13 Xel33 2.16x10-5 1x10-5 o,90 4.50 Xel35" 3.35x10-6 4x10-6(,,,) 0.35 0.70 X3135 5.90x10-6 4,10-6 0.61 1.23 1131 8.83x10-6 9x10-9(s) 409.0 1.83 1132 1.25x10-5 2x10-7(s) 26.0 2.60 1133 2.16x10-5 3xio-8(s) 300.2 4.50

-Il34 2.30x10-5 5x10-7(s) 19.2 4.78 1135 2.05x10-5 lx10-7(5) .85.3 4.27 845.0

(

40 .

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Thus, the maximus whole body exposure to a person in the reactor room for 10 minutes immediately following a fuel element rupture in air would be 845 mress--a value well within the requirements of 10CFR Part 20.

The ingestion of iodine isotopes identified above would result in the concentration of the isotopes in the thyroid with a resultant thyroid exposure. The extent of oxposure can be calculated by multiplying the iodine ingested by the corresponding internal dose effectivity factor.

10 Min Effectivity Ingestion Factor Thyroid Radioisotope pCi res/Ci Dose, reas ,

1131 1.83 1.486x106 2.72 I132 2.60 5.288x104 0.14 1133 4.50 3.951x105 1.78 I134 4.78 2.538x104 0.12 I135 4.27 1.231x106 5.25

• Total thyroid dose 10.01 rems This resultant thyroid dose is reasonable based on the conservative assumptions made.

(b) Exposure to the general public. In case of a fuel element rupture, some of the released radioisotopes would be swept from the reactor room by the ventilation system and discharged through the hood vent into the uncontrolled area. For the sake of the analysis of exposure to the general public, it is assumed that, at the time of the hypothetical fuel element rupture, the ventilation system would

- he switched to its "seersency" operation in which the exhaust air

' is diverted through an absolute filter before being vented.

Furthermore, the air intakes to the room would be isolated to maintain the reactor room under negative pressure and inhibit radioactivity release except through the exhaust vent. The air flow through the exhaust under these conditions is 150 cfm which represents an air change rate in the room of about 1 change per hour.

The concentration of radioisotopes released from the fuel would be l

reduced by the air flow through the room. A common engineering calculation for a well'aixed vessel assumes essentially complete l

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41 5/84'

flush of the vessel in 5 volume changes. However, for the purpose of the present calculations, only one room volume is assumed, and, thus, the concentration of isotopes in the room exhaust would be the same as was calculated above. However, the air is discharged at the roof level, approximately 40 feet above grade. Thus, a dilution due to wind will be encountered.

The Dilution Factor due to the building is defined by Lamarsh as:

DB = cAv .

where c = shape factor, experia'ntally e between 0.5 and 0.67 A = cross section area of the building 9 = average wind speed Sir.co the prevailing winds near the MST2 are form the west, the 3

, north-south area of the Engineering Building is used, thus l

A = 500 ft x 40 ft - 20,000 ft8 9 = 10.1 mph = 14.8 ft/sec (from National Weather Service Bureau) c = 0.5

. Thus, DAB = (0.5)(20,000)(14.8) = 1.48x105 ft f

With a stack flow rate of 150 (= 2.5 , ), the concentration reduction due to wind dilution will be 3,4 = 1.69 x 10-5, The concentration of nuclides.will,.thus, be reduced by this amount outside the building. The dose received by a member of the public j over an hourl of occupancy may be estimated in a manner similar to that'used for occupational exposure given above.

1. Whole Body Exposure = MPC(Table 2,10CFR20) x yr -x 8760 hrs 4

(Assvees that continuous exposure to IMPC is equivalent to 500 ar/yr whole body dose) i 2. Isotope ingested = cone. of radioisotope x 3.47x102 x 3600

..c. .

1 One hour exposure assumed because released activity will be dispereed after this time.

42

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The results of these calculations are given below for the rupture of one 12 vt% fuel element in air with 100% release of halogens and rare gases in the air gap of the fuel.

1 hr dose, I hr ingestion Radioisotope pCi/ml Table 2 er pCi Kr858 7.15x10-ll 1x10-7 4.08x10-5 0.89x10-4 Kr85 2.5x10-12 3x10-7 4.76x10-7 0.03x10-4 Kr87 1.28x10-10 2x10-8 3.6x10-4 1.6x10-4 Xe1338 1.02x10-Il 3x10-7 2x10-6 0.13x10-4 Xel33 3.65x10-10 3xio-7 6.9x10-5 4.56x10-4 Xe135m 5.66x10-Il 1x10-7(ass.) 3.2x10-5 0.71x10-4 Xe135 9,97x10-Il 1x10-7 5.7x10-5 1.25x10-4 4

1131 1.49x10-13 1x10-10(S) 8.5x10-5 1.86x10-7 I132 2.11x10-10 3xio-9(S) 4.02x10-3 2.64x10-4 II33 3.65x10-10 4x10-10(S) 5.2x10-2 4.56x10-4 1134 3.89x10-10 6x10-9(S) 3.7x10-3 4.86x10-4 I135 3.46x10-10 1xto-9(s) 1,98x10-2 4.32x10-4 Total Whole Body Exposure 0.080 nr Similarly, the thyroid exposure from the ingestion of iodines would he 1 hr Effectivity Ingestion Factor Thyroid

Radioisotope WCi Rem /Ci Dose Rees -

1131 1.86x10-7 1.486x106 2.76x10-7 1132 2.64x10-4 5.288x104 1.40x10-5 I133 4.56x10-0 3.951x105 2.0x10-4 I134 4.86x10-4 2.538x104 1.23x10-5 i I135 4.32x10-4 1.231x106 5.0x10-4 l

l Total Thyroid Dose 7.58x10-4 Rees These calculations indicate that the exposure,to the general public as the result of a fuel element failure in air after extended reactor operations I

would be negligible.

43 G471 _

Bibliography Lamarsch, John R., " Introduction to Nuclear F.ngineering." Addison Wesley, December, 1977.

USNRC Regulatory Guide 4. j Simnsd, M. T., "The U-ZrHx Alloyt Its Properties and Use in Triga Fuel,"

G. A. Techrologies Report E-117-833 (February, 1980).

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t SAFETY ANALYSIS 45-3 REACTIVITY INSERTION The present MSTR core consists of 70 elements, all of which are Eight of stainless steel clad U/ZrH with a nominal H to Zr ratio of 1.7.

the elements are 12 wt % Uranium (<20% enriched) while 62 are 8.5 we %

! Uranium (<20% enriched). The 12 wt % elements are located in the B and C i rings of the core and two of these (one each in the B & C rings) are

' instrumented with chronel-alueel thermocouples.

I Experiments conducted at MSU involving the rapid insertion of 1.39%

Ak/k into the core described above by means of a pneumatic control rod results in the production of a power pulse of approximately 258 MWth and an '

integrated power output of about 15 MWeec. The temperature measured in the l

instrumented fuel elements located in the B and C rings, 12 wt % fuel, is ,

i approximately 275'C. It is expected that this fuel concentration in these 4

j locations would experience the highest neutron flux and temperature.  ;

1 By comparison, experiment, conducted at General Atomic (GA6216) with a l j

l Stainless Steel Core containine 8.5 wt % (<20% U235) Stainless Steel Fuel

~

and subjected to pulses of up to 3.5% Ak/k reactivity insertion compared i

the temperatutes produced. The temperatures in both the B and D rings were seasured and a linear relationship between the fuel temperature and the quantity (Ak/8-1) was observed. (8 = effective delayed neutron fraction =

l l

0.7% for this fuel). Although experiments were not conducted at i

Ak = 1.37%, the temperatures produced may be estimated by linear extrapo-f lation of the data obtained between 2.1 and 3.5% down to the 1.37% value

! corresponding to the MSU tests. This yields an estimated peak fuel i temperature of 230*C (in 8.5 wt % fuel). This is consistent with the MsTR data. (Note that the MSTR fuel contains 12 wt 1 U as compared to 8.5% in the GA tests. The higher fuel content would be expected to result in higher power density and a correspondingly higher temperature. This increase might be expected to be approximately 12-8.5/8.5 = 41% to project a 12 wt % fuel temperature in the GA tests of (230-30)(1.41) + 30 = 322'C.)

The same GA report (6216) gives the results of theoretical calcula-tions of peak and integrated power as well as fuel temperature use due te .

the reactivity insertion. The predicted temperatures compare closely with the observed values for large reactivity insertions. Similar calculations 45 l 5/84 l

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{ .

i demonstrated that the peak 'emperature in the fuel occurs not at the

midplane between the centerline and cladding where the thermocouple is located. The peak temperature is approximately 21% higher than the

{ midplane tsoperature which would extrapolate the measured MSTR fuel temperature to (200-30)(1.21) + 30 = 332'C.

( The maximum excess reactivity of the MSTR is 2.25% Ak/k; and, thus, it is theoretically possible to insert this amount of reactivity into the core. Again, based on GA 6216, such an insertion as a step increase would result in a bulk fuel temperature of 450"C. Again, correcting for the peak to bulk calculation and 12 vs 8.5% fusi as above results in a peak tempera-

} ture of 716'C. This temperature exceeds to the phase transition temper-i ature of low hydrogen (H/Zrs1.5) fuel (GA7882) but the MSTR fuel with i

H/Zr:1.7 shows no such phase transition; and, thus, temperatures well above j 1000*C are permissible. Thus, the tapid insertion of the full 2.25% Ak/k I reactivity into the MSTR is not expected to result in a thermal degradation of the fuel.

i hi kh fuel temperatures might be expected to result in increased gas j pressuras within the fuel due to expansion of the gas between the fuel seat i

and the cladding as well as the release of fission product gases.

Similarly, the UZr/H1 ,7 will exert a partial pressure of hydrogen which is a function of fuel temperature.

Tests reported in GA6216 indicated that the maxieue pressure generated in the fuel element was only 7 peig with a reactivity insertion of $4.20 <

i (2.95% t.k/k). It is reasonable to assume, then, that the peak pressure in ,

the MSTR reactor with 2.25% insertion will be less than this (note that at

$4.20 insertion, the peak fuel temperature according to thi same report would be 610"C (bulk)-30 x 1.21 (peak / bulk) + 30 = 732"C which is consis-I tent with the anticipated MSTR peak of 716*C). A peak pressure of 7 psig I -

would result in a stress of about 1500 poi in the fuel cladding. Report i.

GA6216 givos the tensile strength of 30485 at 650*C to be 44,500 psi. Thus, j the pressure generated by a 2.25% Ak/k pulse will not even approach'the .

bursting pressure of the cladding.

on the basis of this analysis, it is concluded that the rapid inser-tion of the full excess reactivity into the MSTR would not result in damage to the fuel.

1 I

v 46 5/84

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SAFETY ANALYSIS 45-4 MECHANICAL REARRANGEMENT OF FUEL When it is necessary to remove irradiated fuel from the MSTR, the following procedure is typically followed. A 400 lb lead transfer cask (such as is used with the BMI-1 shipping cask) is lowered into the reactor l pool by means of a chain falls and A-frame superstructure. The cask is capable of holding three standard Triga fuel elements which are remotely loaded into the cask by a fuel handling tool. The cask is then removed from the pool and the lead shielding protects personnel from the radiation emitted by the fuel.

During the cask handling process it is possible that the transfer cask could be dropped into the pool. Such an accident might impact upon the

, reactor and the reflector. It is proposed to analyze the consequences of 2 ouch an incident.

Several accident scenarios are possible:

1. Deflection of the control rod extensions with the resultant control rod withdrawal. Since the cask will impact in a downward direction, the principle force exerted will tend to drive the control rods into the core (a " safe" configuration). Even if a horizontal force is exerted on the extension rods, the removal of sufficient control rod worth to enable the reactor to go critical is unlikely in light of the shutdown margin inherent in the MSTR.

2. Crushing of fuel. In case of impact of the cask'onto the fuel, deflection or crushing of the fuel is the likely result. The fuel j cladding of several elements would probably be damaged and the rare gas / halogen fission products in the gas gap between the fuel an the cladding would escape. Since the only situation involving the fuel transfer cask would involve a flooded pool, the fission product release described above (for fuel failure in air) would be reduced by the dissolution of halogens in the pool water. Aa indicated in

( the air rupture analysis, the largest radiation is due to the '

Iodine release. For the present analysis, more than one element might be damaged resulting in the releases of greater amounts of rare gaser ad Iodine. However, the presence of the pool water would rosace the amount of Iodine released to the room and, thus, compensate for the increased number of elements involved. It is, 47

.5/84

< f thus, estimated that the radiation exposure for this accident .

t postulation would approximate that of the one analyzed previously (Fuel Elesunt Failure in Air).

3. ImpactonReElectorIonChambers. The reflector of the MSTR will i serve as an impact shield for the reactor core. It will, therefore, reduce the consequences of a falling cask over that described in #2 above. Since the reflector contains little radioactivity and no fuel, damage to it will not result in a significant radioactivity release.

1

'it is, therefore, concluded that the worst case resulting from a cask dropping into the reactor pool would result in no greater consequences than the fuel element failure in air previously analyzed.

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