ML20034F567
| ML20034F567 | |
| Person / Time | |
|---|---|
| Site: | 07000152 |
| Issue date: | 02/16/1993 |
| From: | Schweitzer J PURDUE UNIV., WEST LAFAYETTE, IN |
| To: | Adensam E NRC |
| References | |
| NUDOCS 9303040007 | |
| Download: ML20034F567 (21) | |
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4 PURDUE U N IVE RSITY 5%
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't DEPARTMENT OF RADIOLOGICAL AND ENVIRONMENT AL M ANAGEMENT U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Ms. Elinor Adensam Re: Docket No.70-152 License No. SNM-142
Dear Ms. Adensam:
This letter is in response to your January 5,1993 request for further information regarding the renewal of our special material license. We will address the requests in the same sequence as your letter. We intend to modify the original application by replacing pages into the original document when they are approved by your office. This will allow us to reference one document and reduce any ambiguity that multiple submissions might introduce. Pkase find the entire document text enclosed with the requested changes in their corrected form. Also included are addnional attachments-as updated and supplemental information.
1.
References to the Radiological Control Officer on pages 2,5, and 9 have been changed to Radiation Safety Officer for consistency.
2.
The training described in Section 8.2 will be provided by the RSO or a staff health physicist (described in Section 7.5).
3.
The effectiveness of the training will be determined by testing the individuals on radiation safety principals and procedures covered in the training.
4.
Section 8.3 has been changed to indicate that Principal Investigators are required to attend the radiation safety worker training.
5.
Exemptiou from the requirement for criticality monitoring is requested only for the Duncan Annex. The Civil Engineering Building is no longer a storage area for special nuclear material. The justification for the exemption and a description can be found in Attachment 9-13.
6.
An evaluation, showing that a maximum dose to a member of the public offsite due to a release of radioactive materials would not exceed 1 rem effective dose equivalent or an intake of 2 milligrams of soluble uranium, can be found in Attachment 914.
If you should have any questions regarding the information submitted please contact me at 317-494-2350.
S'. erely,
.c A"
/
AQOORO I
9303040007 930216
' James P. Schweitzer, Ph.D.
DR ADDCK 0700 2
Radiation Safety Officer i
Civit ENGINEERING BUILDING. B173
- WEST LAFAYETTE. IN 479o7 y
N E O,
6.0 Purpose for Which Licensed Material Will Be Used 1
a.The proposed use of U-235 enriched solid helices and discs is to measure the effects of rare carth additions on the thermal properties of UO2 and to measure the effects of elevated temperatures on the mechanical l
properties of UO. Typical experimental procedures willinvolve heating solid discs or helices of various U-235 2
enrichment (3-20%) to a maximum temperature of 500 C for differential thermal analysis and 1600 C for thermal diffraction measurements. llelices will be heated in a vacuum furnace for measurement of mechanical properties. Other various samples of both enriched and natural uranium may be obtained for non-destructive testing, chemical and thermal analysis, and other analytical and developmental techniques.
b.The encapsulated Pu-Be neutron sources are employed in activation analysis studies, for instrument calibration, and for neutron studies in a sub-critical exponential pile. (Note: The natural uranium subcritical pile is separately licensed under Source Material License No.SUD-296)
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- c. Uranium-235 in the form of Spert fuel rods enriched not to exceed 4.8 weight percent, fuel rods enriched not to exceed 1.3 weight percent, natural UO2 pellets clad in aluminum, and californium in up to 10 doubly encapsulated sources will be used in the operation of a subcritical Fast Breeder Blanket Facility (FBBF). The FBBF is a small subcritical facility in which blankets of fast reactors can be mocked up in a realistic geometrical configuration. The FBBF has a central region composed of 4.8% - enriched fuel rods so as to provide the surrounding blanket mockup with neutron spectrum typical of that found in the core blanket interface of a large LMFBR. Na* ural uranium fuel rods are used in the blanket mock-up. The 13% enriched fuel rods were no!
used in the initian FBBF loading. The FBBF will be subcritical with K rr of less than 0.43. The FBBF will be e
of ], n by Cf spontaneous fission neutron sources distributed along its center line. Up to a maximu dri)
Cf possession is requested in order that the FBBF can be gfueled to or near its original levels cf Cf at i
2 some time in the future. This would not occur until the original Cf has decayed to levels not sufficient for j
experimental purposes. Neutron and gamma-ray transport will be studied in the-blanket mock-up. The experimental studies will include measurements of neutron spectra, foil reaction rates, fission distribution, and gamma-ray heating.
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i Procedures used in operation of the FBBF are reviewed for safety by the Radiological Control Committee and the Radiation Safety Officer and/or his staff. The Radiological Control Committee is responsible for:
1.
Reviewing plans, specifications, and procedures for operation of the FBBF and taking appropriate action.
2.
Resiewing any significant proposed changes in design or new operations for the FBBF. Committee approval or Radiation Safety Officer approval is required in advance of initiating such changes.
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Approving research projects utilizing the FBBF.
4.
Authorizing pers<mnel to work in the facility.
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Overseeing all radiation safety aspects of the operation of the FBBF. including routine surveys, leak tests, personnel monitoring, instrument calibration, emergency procedures, and inventory.
d.
U-233, Np-237, Pu-238, Th-230, Ac-227 and Cm-244 (Sections e-j of item 5) will be used as calibration sources.
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7.0 Indhiduals Responsible for Radiation Safety Program and Their Training and Experience 7.1 Senior Administration The governing body of the University is the Board of Trustees which selects the president of the University.
The president, Steven C. Becring, as the chief executive officer, is responsible to the board for the internal -
administration of the University. The executive vice-president and treasurer, Frederick R. Ford, serves as the chief business and financial officer for the University with responsibility for all business offices, the physical plant, residence halls, facilities development, investments, and trusts. He acts for the president with administrative authority for managing, developing, and planning in all of thes areas. The executive vice-president and treasurer serves as certifying official for NRC license documents. The vice-president for physical facilities, Wayne W. Kjonaas, assists and acts for the executive vice-president and treasurer with
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administrative respcmsibility for operation of the physical plant and facilities planning and construction. The director of Safety and Security, T. Grant Kepner acts for the vice-president for physical facilities in matters of safety. police and fire, parking, and emironmental management which includes radiation safety. Charts of the organization are in Attachment 7-1.
7.2 Radiological and Emironmental hianagement (REht) i The head of REh! is Stuart W. Kline. This department is responsible for general safety, industrial hygiene, chemical safety and management, and radiological safety. This comprehensive program allows interaction l
and coordination of all facets of emironmental and laboratory safety, j
7.3 Radiological Control Committee (RCC)
The RCC is empowered by Executive hiemorandum B-14 (hiay 15,1973) and subsequent charter (Attachment 7-2) to act as the body responsible for all University programs invohing radioacthity or a
radiation producing devices. The RCC meets at least quarterly and an attendance of at least 50% is considered a quon.m. Membership of the committee is diversified and representation from different areas of use (e.g. life sciences, nuclear engineering or veterinary medicine) is maintained to ensure appropriate review of research uses of radioactive material.
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All new investigators (authorized users) are apprtwed by the RCC at the quarterly meetings. All major changes in existing authorizations are also approved at these meetings. The Radiation Safety Officer (RSO) is authorized to give interim approval to investigators whose use is similar to uses that have been approved l
in the past by the full committee. This interim approval must also be reviewed by another RCC member prior to the issuance of the authorization for work with radioactive material. Ifin the judgement of the RSO j
or the reviewer from the RCC the use is substantially different from past uses or invohrs an interpretation
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l of guidelines or regulations then the full committee must approve the authorization.
j The Ilead of REM (management representative), Radiation Safety Officer (RSO), and a health physicist (secretary) serve as ex officio committee members. The current RCC is listed in Attachment 7-3. The membership of the committee may change due various reasons however membership will be limited to individuals having significant knowledge of and experience with radioactive materials. Resumes of the Chair and RSO are in Attachment 7-4 l
In addition to those duties described by Executive hiemorandum B-14 the RCC performs these duties:
j Review and renew authorized user permits at two year intervals.
Review annual audit findings and approve recommendations of radiation safety staff for appropriate action.
Review license amendment submissions to the NRC and resuhs of periodic inspections.
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Establish and approve minutes of all proceedings and actions taken by the committee.
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7.3 Radiation Safety Officer (RSO) t The current RSO is James F. Schweitzer, Ph.D. Dr. Schweitzer received his M.S. degree in licalth Physics l
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(1981) and Ph.D. in Environmental Toxicology (1985) from Purdue University. lie was employed at the l
Illinois Department of Nuclear Safety from 1986-87 where he worked in the emironmental monitoring and I
radon section. lie was employed as the Radiation Safety Officer at Purdue in 1987.
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The primary duties of the RSO include but are not limited to the following:
j Administration of the Radiation Safety Staff (RSS) with the overall responsibility of managing the radiation safety program.
Ensuring compliance of the radiation safety program with state and federal regulations and NRC license i
conditions.
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Providing training and recommendations to individuals that use radioactive materials.
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Act as the agent of the Radiological Control Committee to ensure that use of radioactive material is consistent with recommendations and requirements of the committee.
i Serve as representative of the University to regulatory agencies to act in licensing matters and providing i
corrective action when deficiencies are identified.
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7.5 Radiation Safety Staff i
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Radiation Safety Officer. The duties of the RSO are described above. The RSO should meet the requirements as described in the Draft Regulatory Guide (April,1982). *Oualifications for the Radiation l
2 Safety Officer in a Large-Scale, Non-Fuct Cycle Radionuclide Program" Heahh Physicists (3). The ficahh Physicists are required to have a B.S. degree in llealth Physics or a related area. If the degree is in a related area, experience in a medical or university health physics program is highly desirab!c.
i Environmental Technicians (2). The technicians are required to have a high school diploma but no other relevant experience. These indhiduals work under the supenision of the RSO or a heahh physicist.
i Radionuclide Purchasing Agent and Secretary. These two positions require no formal training in radiation safety ahhough database management experience and basic radiation terminology is highly desirable.
Student Assistants. Work-study students and undergraduate interns are hired to perform basic heahh physics tasks or other tasks under the supenision of various radiation safety staff. No previous experience is required.
7.6 SNM for which possession is requested in this application is under the direct control of indhiduals of the l
Nuclear Engineering Department who have been authorized by the Radiological Control Committee. Direct j
control over the design and operation of the FBBF and the storage of SNM, source materials and byproduct material will be under the direction of the Department of Nuclear Engineering.
f FBBF Project - FBBF Laboratory and Assistant Laboratory Directors will be appointed by the licad of the Department of Nuclear Engineering. Indhiduals appointed to these positions will have had formal training l
in their fic!d and or sufficient practical experience to be considered qualified to assume the duties associated with the appointment. The responsibilities of the Psoject Director include the definition of the experimental and theoretical research topics, which will be done in cooperation with all principal investigators, and in coordination with specially designated representatives of the supporting agencies Dept. of Energy and
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Electric Power Research Inst.). The Laboratory Director and his Assistant are responsible for devising and 4
performing the experiments on the FBBF, including the FBBF operation. The safety and security l
responsibilities of the Laboratory Director and his Assistant include:
I 1.
Ensuring that the FBBF facility is operated in compliance with license conditions.
i 2.
Preparing procedures for Radiation Control Committee approval.
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Assuring that personnel have been approwd by Radiological and Environmental Management.
4.
Maintaining Physical inventory and daily records.
In matters regarding radiation safety, the line of authority for reporting purposes is directly to the Radiation j
Safety Committee through the Radiation Safety Officer. The current project director is Frank M. Clikeman.
l A resume of his training and experience is in Attachment 7-5.
i 8.0 Training For Individuals Working In Or Frequenting A Restricted Area j
4 The extent of training for individuals in a restricted area is dependent on factors such as responsibility of the indhidual, duties in the area, and frequency in the area.
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Personnel are divided into two main categories: radiation workers and support staff.
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8.1 Training for Support Staff All support staff upon employment by the University are required to attend an orientation program which describes the employce's rights and responsibilities. The training covers right-to-know information (chemicals, biohazards, etc.) and radiation safety information. A videotape " Radiation Safety for Support Personnel"is shown and indhiduals are given an opportunity to ask questions. When any doubt exists as to a situation in a laboratory the indhiduals are advised to call REM or University police. Examples of
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indhiduals that attend this training are maintenance personnel (plumbers and electricians) and all custodial staff.
8.2 Training for Radiation Workers
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All radiation workers receive instruction in accordance with 10 CFR 19.12 prior to beginning work with
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licensed material. This instruction is provided by a qualified staff member which is the RSO or a health
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physicist. Each user regardless of prior experience is required to attend this orientation and training session.
At the conclusion of the training cach indhidual receives a training booklet and Radiation Safety Manual (Attachment 8-1) and is tested on radiation safety principles and procedures covered in the training. The training consists of the following:
4 Videotape series (Indiana University Produced) 1 1
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" Introduction to Radiation Safety"
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- Laboratory Techniques" i
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" Emergency Procedures" Purpose of Radiological and Environmental Management Principles of ALARA Special Notices (e.g. Results from latest NRC inspection)
Glossary of Radiation Terms Notes from the Videotape Series
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Instruction Regarding Prenatal Exposure Regulatory Guide 8.13 l
Summaries of 10 CFR Parts 19 and 20 1
Personnel Dosimetry and Exposure Limits I
liazards Associated with Commonly Used Isotopes t
i Decontamination and Accident Procedures
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Waste Management Procedures i
Facility Classification
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Marking and Labeling of Facilities and Equipment Laboratory Demonstration i
Examples of Labeling l
Shielding for Beta and Gamma Radiation 3
Demonstration of GM Sunry Techniques Wipe Survey Techniques Survey and Imtntory Records 8.3 Principal investigators (authorized users) are those to which an authorized use of radioactive material is granted. The principal investigator (PI) may have radiation workers such as graduate students, postdocs, technicians, or other Purdue staff or students that work under his authorization. The PI must be a l
permanent Purdue staff member with a college degree at the bachelor level or have equivalent training and experience in the physical or biological sciences or in engineering and attend the radiation safety training. In addition, the PI must have at least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of training and experience in the safe handling of radioactive materials, and in the characteristics of ionizing radiation, units of radiation dose and quantities, radiation i
detection instrumentation, and biological hazards of exposure to radiation appropriate to the type and forms.
of byproduct material to be used.
Prmeipal investigators not meeting this criteria may be asked to work under the authorization of an existing user until he has received the requisite training and experience. The RCC reviews each PI to evaluate the training and experience and the amounts and types of licensed material requested. Approval is given when
.l the RCC is satisfied that the PI can work safely with the given type and amounts oflicensed material.
8.4 Radiation workers (users) are authorized as full users when their training and experience is sufficient to work with the quantities and types of radioactive material that the principal investigator is licensed to use.
Some users may have littic or no experience working with licensed material. In that case, the individual is authorized as a restricted user who may only use smaller quantities until he receives additional training and experience. Once the PI has certified that the individual has received the training and experience the user may become fully authorized to use licensed material.in those procedures for which the laboratory is authorized to perform.
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t 8.5 Continuing Education i
The continuing education program is performed in a number of ways. Examples of the methods used to j
accomplish this are:
i REM Newsletter (distributed quartcaly)
This publication is designed to distribute timely information on radiation safety, chemical safety, and general safety in the laboratory.
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Radiation Safety Directives
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These publications are sent to authorized users when situations are identified that pose hazards to a number of users or potential violations of license conditions or regulations. Results of NRC inspections and corrective action have been communicated in this manner.
i Facility Audits
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Health physicists will communicate with laboratory personnel during routine audits and provide information on A1 ARA, surveys, shiciding, etc.
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Special Seminars or Insenice Training These programs may be provided by REM or requested by the principalinvestigator.
i 8.6 Authorization Procedures i
5 Af ter attendance at the orientation and safety training the principalinvestigator must submit the appropriate l
documentation for authorization. A Form A-1 describes the isotopes and amounts, proposed uses, procedures, associated hazards, and techniques to prevent contamination and keep exposures ALARA. The l
Form A-IS provides a description of the facility and the amount oflicensed material to be used in each area.
The Form SM-1 provides a description of the survey meter (if required). The Form A-4 is required for each l
authorized user that intends to work under the supenision of a particular PL An example of each form is in Attachment S-2.
i 8.7 Additional Training students in Nuclear Engineering associated with the FBUF facility receive the didactic instruction and l
iaboratory training in traditional nuclear engineering courses. Evaluation of the didactie instruction is by
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examination since the material is covered in existing formal courses in Nuclear Engineering, and the students receive college credit for their efforts. Also seminars on a regular basis reinforce training covered i
in coursework and on the job.
9.0 Facilitics and Equipment i
9.1 SNM will be used and/or stored in two locations on the Purdue University campus. The three buildings are the Duncan Annex and the Physics Building. Their location is marked on the map of the Purdue University campus. (Attachment 9-1) In addition, other locations may be used from time to time when approved by the Radiological Control Committee.
a Duncan Annet Storace Facility: SNM are stored in Room B-84 of the Duncan Annex which is attached to the Electrical Engineering Building. The storage room is approximately 18 ft. 4 in. by 30 ft. 8 in. by 23 ft.
high, as shown in Attachmcut 9-2 and 9-3. The floor level is about 9 ft. below the outside ground level. The
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building is ccmstructed of steel frame with concrete and brick and no window openings.
- b. Physics Buildine Fast Breeder Blanket Facility Laboratory: This laboratory will be operated as part of the f
School of Nuclear Engineering. The laboratory is located in rooms in the Physics Building on the Purdue Campus (See Attachment 9-5). The building is constructed of steel frame with concrete and brick. The floor
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plan of the laboratory is shown in Attachment 9-6. The steel fuel storage cabinets (Attachment 9-7 and 9-8) are fastened to the southwest wall of room U28 and are surrounded by a wire-mesh cage. The dimensions of 7
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I room B28 are approximately 51 ft. by 50 ft. by 20 ft, high. The floor level is about 8 ft. below the outside i
ground level and is about 30 inches above the floor level of the newer portion of the Physics Building.
The Fast Breeder Blanket Facility is located in room B28C as shown previously. The fueled portion of the facility is cylindrical with an outer diameter of 58 inches and a height of 48 inches. The facility is set on a concrete base approximately 24 inches below the floor level of room B28(See Attachment 9-9). The center i
of the facility is approximately 32 ft.11 in, from the center of the nearest steci storage cabinet. Room B28C l
is a concrete housing constructed within room B28 to enclose the FBBF. The walls and roof of room B28C have been designed so that radiation levels outside of room B28C will be acceptably low.
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9.2 Criticality Alarms f
i Two of the Remote Area Monitors are installed in the FBBF to serve as a radiation and criticality accident
'i alarm system to meet the requirements of 10 CFR 70.24 and ANSI N16,1969. Four criticality dosimeters are -
i used to monitor the area. These dosimeters are Reactor Experiments type 708 dosimeters (or equivalent) containing six activation foils for neutron dosimetry and a thermoluminescent dosimeter (TLD) for
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gamma-ray dosimetry in accordance with requirements stated in 10 CFR 70.24 j
i 9.3 Request for Exemption to 10 CFR 70.24.
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l We request an exemption to 10 CFR 70.24(a)(1) requiring an area monitoring system for the storage facilities for SNM located in Duncan Annex and Chil Engineering Building. The exemption is requested on I
the basis that the facilities are for storage only of the fuel and helices. Numerous survey meters are available in the area at all times.
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9.3 instrumentation Purdue University has extensive instrumentation for the determination of contamination, exposure rates, and radioactivity in solid and liquid samples. A list of instruments is in Attachment 9-10 and 9-11.This list is representative of the type and quantity of instrumentation and may be changed due to program requirements. Authorized users also possess survey equipment in their own laboratories if required to'do so by the RCC.
i Procedures for calibration of survey instruments and sources utilized in those calibrations are in
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10.0 Radiation Safety Program 10.1 Personnel Monitoring Devices Film badges are generally used as peisonnel monitors of radiation exposure. Any person working with SNM
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is required to wear personnel monitoring devices whenever entering a restricted radioisotope area under conditions where he is likely to receive, in an calendar quarter, a dose in excess of 10 percent of the MPD (1.25 rems / quarter). In addition, pocket dosimeters may be worn by personnel in addition to film badges in 1
cases where exposures would reasonably be anticipated. Pocket dosimeters may be utilized in lieu of film badges for individuals entering restricted areas on an infrequent or temporary basis. Criticality foils are re-quired when an indhidual enters the room housing the FBBF assembly.
The film badges are read monthly using a NVLAP accredited supplier. An annual review, not to exceed fourteen months, is made of personnel exposures. In cases where indhidual monthly doses exceed 100 millirem, the Radiation Control Office will notify the indhidual (or his/her supenisor) of the exposure as a means of alerting the indhidual to the occurrence of the dose.
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10.2 Bioassays i
No formal bioassay program has been established for individuals working in the FBBF facility. Bioassay can, however, be required of certain individuals who work with SNM depending upon the level of activity and circumstances of use. The Radiation Safety Officer or his staff may establish the need for such bioassay evaluation which could indude analysis of urine or other excreta, depending on the nuclib, its chemical and physical form, and the mode of intake. The Radiation Safety Officer is authorized to require the submission of other excreta (such as fecal samples, nose wipes, or breath samples) in addition to or in lieu of urine samples.
103 Surveys and Monitoring
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Surface contamination surveys are conducted at the FBBF facility every month, not to exceed six weeks, to determine levels of removable contamination. Smears are taken of 100 cm' areas. Acceptable surface contamination lesels are shown in Attachment 10-1. The amount of removable radioactive material per 100 cm' of surface area is determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels are reduced proportionally and the entire surface is wiped.
The average and maximum radiation levels associated with surface contamination resuhing from beta-gamma emitters should not exceed 0.2 mrad /hr at I cm and 1.0 mrad /hr at the surface, respectively, f
measured through not more than 7 milligrams per square centimeter of total absorber.
l When observed contamination levels reach the above levels, decontamination must take place. From an operational point of view, every effort is made to keep contamination levels below those set forth in the table above. Ahhough not an official action level,in practice, any time removable contamination levels above 10%
of the table value is reached, decontamination is accomplished.
No maximum time allowed before decontamination is actually started has been established. Contamination problems are assessed on an individual incident basis and after review of the situation a decontamination l
plan is decided upon. After the tabled levels are exceeded, the area is restricted except for decontamination activitics.
11A Personnel Surveys
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Protective clothing and/or devices will be used for all manipulations with unsealed sources where the possibility of contamination exists. In particular, suitable gloves will be worn whenever hand contamination is probable. Surgical glove techniques are used for putting on and removing gloves in order to avoid i
contaminating the inside surfaces. Thorough monitoring of hands, feet and clothing is mandatory whenever leaving a radioisotope laboratory where work with radioactive materials is in progress. Each individual radioisotope user is personally responsible to check himself for contamination every time he leaves the radioisotope area.
Laboratory clothing or protective garments (such as lab coats) used in radioisotope laboratories are monitored soutinely during the course of the work and when work with the radioactive materials is completed. Such garments will not be released for washing or cleaning until they have been monitored by the user and found to be free of contamination. Contaminated garments will not be worn elsewhere, (outside the radioisotope laboratory or in
- clean" areas.) Articles which show contamination will be left in 9
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- the work area, or other areas designated for this purpose. Such clothing will be marked by the user with his -
name, date, and the nature and degree of ccmtamination and held for storage until the activity has decayed to background level; or decontaminated; or disposed of as radioactive waste.
i in the case of skin ccmtamination, decontamination by soap and water washing will be initiated as soon as is practicable. In stubborn cases, the use of chelating agents, mild oxidants, solvents, and isotope dilution i
solutions will be used to reduce skin contamination levels to background levels.
Leak Tests i
- a. Each plutonium source will be tested for leakage at intervals not to exceed six (6) months. In the absence of a certificate from a transfer indicating that a test has been made within six (6) months prior to the
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transfer, the scaled source shall not be put into use until test /
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- b. The test will be capable of detecting the presence of 0.005 microcuries of alpha contamination on the test sample. The test sample will be taken from the source or from appropriate accessible surfaces of the device i
in which the scaled source is permanently or semipermanently mounted or stored. Records of leak test
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results will be kept in units of microcuries and maintained for inspection by the Commission, j
- c. If the test reveals the presence of 0.005 microcurie or more of removable alpha contamination, the.
I licensee will immediately withdraw the scaled source from use and shall cause it to be decontaminated and repaired by a person appropriately licensed to make such repairs or to be disposed of in accordance with the Commission regulations. Within five (5) days after determining that any source has leaked, the licensee will file a report with the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, describing the source, the test results, the extent of contamination, the apparent or suspected cause of source failure, and the j
e corrective action taken. A copy of the report shall be sent to the Director of the nearest NRC Inspection and j
Enfor:cment Office listed in Appendix D of Title 10, Code of Federal Regulations, Part 20.
- d. The periodic leak tests do not apply to scaled sources that are stored and not being used. The sources f
excepted from this test will be tested for leakage prior to any use or transfer to another person unless they have been leak tested within six (6) months prior to the date of use or transfer.
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- c. Leak tests of the Cf-252 scaled source will be performed only at times when it is removed from the FBBF assembly for purposes of source transfer, replacement, or similar acthities. When the Cf-252 source is leak l
tested, conditions b and c, above, will apply.
j 10.5 Procedures for Opening Radioactive Materials Packages
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All packages received at Purdue University will be monitored in accordance with 10 CFR 20.1906 for f
external contamination and dose rate. General procedures for opening packages are as follows:
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Prior to opening packages put on appropriate personnel protection such as gloves.
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Note any specific instructions by the supplier with regard to special precautions for opening the package.
3.
Smear the outside of the package to verify that ccmtamination limits are not in excess of that specified l
by 10 CFR 71.87(i). Notify RSO if these limits are exceeded.
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Use an ion chamber or GM survey meter to verify that external exposure rates are within those specified by 10 CFR 71.47. Notify RSO if these limits are exceeded, j
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Packages which have obvious damage to the outer container should have the mtegrity of the inner containers verified. If there is a question as to the integrity of any inner container it should be smeared to verify that the contents have not leaked out.
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Personnel opening packages should use the principles of ALARA (time, distance, and shielding) to maintain exposures as low possible.
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Any records such as package k>gs, calibration data, and inventory must be recorded if required.
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If material ordered is received in a satisfactory condition then the user should be notified for pickup.
I 10.6 Procedures for Maintaining Inventory and Accountability All radioactive material is accounted for in REM by using a computer database program called Radioisotope Inventory and Tracking System (RITS). As previously mentioned, each individual order is checked versus the possession limit for each particular PI inventory. Once the material is received the amount is added to the inventory of the PI possessing the material. The PI is also resp (msible for keeping l
records within the laboratory for receipt and disposal.
As waste material is picked up from each laboratory this amount is subtracted from the running total that is maintained by RITS. At any given time,it is then possible to retrieve the amount that a particular PI has in inventory (RITS also decays activity). Each PI verifies his inventory on an annual basis to correct for variations in estimates of material that has been assigned to waste or has decayed.
A printout of the entire inventory is generated on a quarterly basis so the RSO or his designate can verify that license quantities have not been exceeded. The amounts on the inventory are compared with the license I
amounts and plans can be made to accommodate future research needs without undue delay.
Yearly totals are also generated for material that has been disposed of via the sanitary sewer and through incineration. Since the MPC is never exceeded on a single day it is not necessary to generate a running total I
to ensure that the yearly MPC is not exceeded. However the database is in the process of being modified so that information can be retrieved in a report form that can be presented to the RCC or management representatives.
10.7 General Rules for the Safe Use of Radioactive Material Laboratories are strongly encouraged to utilize good laboratory practices and maintain the facility with good housekeeping in mind. General rules apply to all individuals working with radioactive material unless specific exemptions have been granted by the RCC or RSO or use of the general rules are incompatible with other safety provisions or procedures.
1.
All areas where radioactive materials are used or stored wi!! be posted in accordance with 10 CFR 19.11 and 10 CFR 20.1902.
2.
Eating, drinking, food preparation, food storage, and smoking is not permitted in laboratories where significant amounts of radioactive materials are used or stored.
3.
The use of food containers for handling or storing radioactive materials is not permitted. Any other containers used must be clearly marked as containing radioactive material.
4.
The pipetting of radioactive solutions by mouth is strictly prohibited.
r 5.
A trial run without radioactive material should be conducted for all new procedures prior to the use of i
radioactive material.
6.
Any work performed with volatile radioactive material (such as sodium iodide) or operations that have the potential for personnel exposure or contamination must be performed in an appropriate hood or glove box.
i 7.
Protective equipment such as gloves and lab coats must be used for all manipulations of unsealed j
sources. In addition, eye protection must be worn when working with materials that may be hazardous i
6 11
to the eyes. Eye protection is required when handling greater than 10 millicuries of high-energy beta emitters such as P-32.
i 8.
Protective equipment must not be worn outside the laboratory unless it has been monitored and found i
to be free of contamination.
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9.
Work surfaces that may be subject to contamination should be covered with absorbent paper that is i
changed on a regular basis. Any work with large volumes of material or a process with a high spill i
probability should be done in a spill tray.
10.
A radiation survey should be performed by the radionuclide user at appropriate intervals. The survey l
may be conducted with a survey instrument or wipes depending on the isotope used. Items found to be contaminated should be placed in a suitabic area, decontaminated, or disposed as radioactive waste.
1 11.
Contamination must not be allowed to remain in any area for an extended period of time. If j
contamination is found outside the immediate use area, REM should be notified immediately.
?
i 12.
Radioactive material use, survey, and inventory records must be maintained at all times by the principal i
investigator.
l 10.8 Emergency Procedures 1
Emergency procedures are designed primarily to protect personnel from radiation and other physical j
hazards in the workplace. Secondarily, these procedures protect facilities and confine any contamination to I
the immediate area. All personnel are instructed in emergency procedures during initial training and are l
given copies of emergency procedures which appear in the Purdue Radiation Safety Manual and are in i 0-2. Radiation workers are encouraged to refer to this manual on a regular basis. Numbers for i
emergency response are posted on all laboratories that utilize radioactive material. Specific procedures for the FBBF are in Attachment 10-3.
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i One member of the radiation safety staff is on call (wears a pager) at all times to respond to emergencies t
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invohing radioactive material. Other members of the staff are listed with the police department, so these indhiduals may be reached if further assistance as needed.The Purdue Fire Department is also trained in hazardous material response which includes radioacti e material. Mutual aid agreements exist for additional I
assistance from other local municipalities.
i i
i 10.9 REM Laboratory Suntys and Monitoring i
Surveys and monitoring are performed by REM staff on a regular basis according to the hazard j
classification of the particular laboratory. Frequencies for Class A labs are weekly, Class B labs are monthly, j
Class C labs are quarterly, and Class D labs are annually. A typical laboratory audit sheet is shown in 3 0-4. The audits of each lab check the following items:
j i
l 1.
Resuhs oflast suncy to identify that any problem areas have been corrected.
4
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2.
Amounts and isotopes used in the laboratory.
i 3.
Direct radiation survey throughout the laboratory to identify any contamination areas or areas where l
posting is required by 10 CFR 20.1902.
l 4.
location and verification of operation of laboratory survey instrument (if required).
1 5.
Proper waste disposal practices.
1 6.
Security of radioactive materials if lab is not attended.
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. Proper labeling of equipment and radmactwe storage areas.
8.
Utilization of proper protective equipment such as lab coat and ghwes.
f 9.
Observance of eating, drinking, and smoking prohibitions.
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- 10. Personnel dosimetry is worn if necessary.
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- 11. Wipe tests of selected areas to determine that contamination has not been inadvertently missed by i
laboratory personnel, i
- 12. Inteniews with laboratory personnel will be conducted on a periodic basis to verify that all individuals' i
have been authorized by REM and appropriately trained by the Pl. Training or information regarding
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an aspect of radiation safety is usually covered at this time.
- 13. Inspection of survey records maintained by the laboratory, f
i 10.10 ALARA Program Purdue University is committed to providing a working place relatively free of recognized hazards. To thi end an AIARA program is an important operating feature of radiation safety (See Attachment 10-5). Since any exposure to ionizing radiation is thought to incur some risk of cancer or genetic effects, the goal is to j
j keep exposures low while allowing research with radioactive materials to proceed without undue hardship.
1 The AIARA program uses the following methods to keep radiation exposures as low as reasonably achievable:
i 3
Training i
1 A
l All individuals are made aware of radiation effects and methods to keep exposures low. Laboratory j
l demonstrations during training illustrate the principles of time, distance, and shielding through the use of actual sources and survey instruments. In addition, proper survey techniques are demonstrated so users can
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minimize any contamination which may have the potential to be inhaled or ingested.
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l Personnel Dosimetry i
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Although few indhiduals meet the requirement for external personnel dosimetry this senice is provided to l
all users of energetic beta and gamma emitting radionuclides. This program allows Purdue to determine j
easily any problem areas that may be developing especially for users that lack extensive experience in j
j handling radionuclides.
Exposure Notification Exposure reports are usually reviewed by all members of the professional radiation safety staff. Subsequ;at I
to that review any exposure excceding 100 millirems in a given month is identified. Exposures to the l
extremities, skin, and those to the whole body are included in this process. Indhiduals receiving doses l
l exceeding this trigger limit are identified and receive a F-1 Form (Attachment 10-6). This form alerts the individual of the dose and the location (whole body, hand, etc.) recching the dose. The indhidual is i
required to sign and return the form to confirm his awareness of the dose. The indhidual must also indicate the reason for the dose and indicate actions that will be taken in the future to reduce those exposures.
i Exposure Investigations i
When exposures exceed greater than 25 percent of the limits specified in 10 CFR 20.1201 an investigation l
into the causes is initiated. Procedures are examined and recommendations are made by heahh physicists to
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assist in reducing exposure to the affected indhidual(s). In many cases investigations procecd when users l
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request assistance in keeping their exposures as low as possible even prior to reaching 25 percent of the applicable limits.
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Laboratory Audits l
l Routine laboratory audits are an ideal method to informally observe procedures and aethities in the
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laboratory. Independent measurements in the lab can also identify areas where shielding and placement of j
radioactive storage areas can reduce exposures to lab occupants even further.
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11.0 Waste Managernent 1
All waste management operations at Purdue University are carried out by trained technicians. Technicians are responsible for picking up radioactive waste from alllaboratories since researchers are prohibited from j
sink disposal or any other direct means of disposal unless specifically exempted from the requirements.
11.1 Containers All containers for waste are supplied by REM unless a PI has requested to use alternative equivalent containers. The containers that are delivered to investigators upon request include the following:
i Plastic carboys: 1 liter,4 liter,20 liter Fiberboard drums: 8 gallon,30 gallon
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Metal drums: 8 gallon Bags: 8 gallon plastic,30 gallon reinforced paper Prior to pickup by REM, all containers must be properly labeled to include isotope, amount, authorization number, investigator name, date, and any solvents or hazardous materials present.
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11.2 Segregation of Waste i
Waste is required to be separated into a number of different categories so that it may handled. At this time the categories for waste include.
i Solid short half-life (< 30 days) is placed into drums or bags as appropriate.
i Long half-life material is separated into combustible and non-combustible and placed into bags or drums.
I Liquid waste is separated into short and long half-life. Any liquid waste containing hazardous or non-i dispersible components is identified and handled separately.
j Vials are required to be returned to their original carton to remain upright during transport or placed in a drum which is double-bagged to prevent leakage.
l l
Biological waste is placed in plastic bags and kept frozen until pickup.
Sharps are required to be in a rigid or semi-rigid container so that handling the container would prevent cuts f
or punctures to the technicians.
14 l
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i 11.3 Pickup Procedures i
When a PI is ready for a waste pickup REM is called to schedule the pickup. Technicians verify the label to ensure that all information necessary is complete. The packages are all smeared, surveyed for exposure rate l
at the surface and at I meter, and checked to see that the enclosures are secure. Any labeling, marking, notices, shipping papers, and placarding requirements are observed pursuant to 49 CFR prior to transport j
on public highways.
l 11.4 Waste Processing After pickup waste is delivered to various handling facilities for processing and disposal. Description of those facilities are provided elsewhere in this application.
Short half-life materials are stored for a minimum of 10 half-lives prior to disposal. Solid material is s
surveyed with an end-window GM survey meter and any material with radiation levels essentially equal to background are disposed as normal trash. Prior to disposal all radioactive labels and markings are defaced or destroyed. A record of all radioactive material that has been disposed of by decay in storage (DIS) will be i
maintained. lj uid material will be sampled and analyzed for radioactivity by liquid scintillation counting.
q Any material less than 100 dpm/ml will be disposed in the sanitary sewer provided the material is readily dispersible and non-hazardous. If the activity is above this amount the material will be allowed additional decay time. If further sampling reveals that the acthity is still present it will be treated as long half-life material (see below).
Combustible solid long half-life material will be incinerated at Purdue facilities using an incinerator described in our broad scope license. Procedures for loading waste and ash handling are also described therein Short half-life material may also be incinerated prosided that this material be allowed to decay for at least 10 half-lives.
Non-combustible long half-life waste (metal and glass) is compacted using Tcledyne Industries compactor or a " Slugger" compactor (International Dynetics). The Teledyne compactor is equipped with an external air exhaust and HEPA fiher which is monitored on an annual basis for aethity and integrity.
No air sampling is done in the area since the mow volatile compound undergoing compaction is I-125.
?
Personnel bioassays, probably the most sensitive exposure indicator, have indicated a maximum dose equivalent of 13 millirem per year to the thyroid. This is less than 0.03% of the 50 rem occupational limit for the thyroid and falls well under our 100 mrem ALARA trigger level. Direct radiation surveys and wipe testing of the areas are performed on a regular basis.
long half-life liquid is sampled and analyzed according to the radionuclide(s) present. Pure beta emitters 2
are analyzed by liquid scintillation counting and gamma emitters are analyzed by Nal or Ge spectroscopy.
An aethity is calculated and a computer program compares the actiity with disposal limits and makes recommendations for disposal.
Liquid waste is then disposed of by the sanitary sewer as provided in 10 CFR 20.2003. Attempts are made to limit daily disposal to less than ten times the Appendix C value in the interest of ALARA. However,if this l
limit would increase exposure to radiation safety staff (such as subdhiding waste packages) the former limits would be observed. Records are maintained at all times of material that has been disposed of in this manner.
I Scintillation vials are processed in two ways depending on the type of scintillation fluid used. Vials that utilize a hazardous component such as xylene is packaged in DOT 17H 55 gallon drums and shipped via ADCO Ser ices for ultimate incineration at Quadrex. All applicable DOT regulations regarding the l
shipment of hazardous materials are observed.
Vials utilizing the biodegradable cocktail are collected and crushed using a Vyleater (S and G, Inc.) and the resuhing liquid is disposed of in the sanitary sewer system pursuant to 10 CFR 20.2003. Records are i
15
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i maintained for all material disposed of in this manner Glass from vials containing less than 0.005 microcurics per gram of 11-3 or C-14 will be disposed of as if it were not radioactive. Glass from crushing other isotopes will be rinsed prior to disposal. If a sample of the glass is < 10 dpm/ gram the material will be disposed of as normal trash. If the material has activity greater than 10 dpm/ gram then the material will be rinsed until that level is achieved. Operating procedures for the vial crusher are found in our broad scope license.
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11.5 Waste Shipments
(
All waste shipments will be transferred to a licensed waste broker or facility. Currently waste is transferred
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to ADCO Services, Inc. for ultimate disposal at Richland, Washington under permit number 5730. Scaled j
sources may also to transferred to other license for disposal or reuse such as J.L. Shepherd and Associates.
In all cases, we will follow Appendix F to 10 CFR 20.1001 to 20.2401 for land disposal of all wastes.
4 11.6 tong Term Storage of Radioactive Wastes Purdue University has constructed a facility for long-term storage of radioactive waste once access to f
currently operating sites is denied. The facility will be adequate to store the types and amounts of radioactive
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wastes that are currently being generated at the University. A description of the facility and procedures to i
maintain security and integrity of the waste according to NRC Informational Notice 90-09 are found in our broad scope license.
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APtil 1972 Vice President for Physical facilities 49 48000 Kay Bronten, Secretary 49-40766 Doris Mills l
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Grant tecrwr Bob Bela Don Mufford Wayne tjenees Jemy Couvi Tom Scet, Director Ofrector Director Director Director Assistent Director of Feellities Plaming Safety and Security Fiscal Services Sulldings and Crowns:
Utttttles Persomel Services
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February 1993 Wayne Kjonaas Vice President for Physical Facilities 49-48000 Key erenson, Secretary 49-40766 Doris Hills
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',Irector Director Director Acting oirector As:Istent ofrector of Fecilities Ptemin9 Sede.y ord Security Fiscal services Buildirgs eruf Grounos Utltitles Persomel Services
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. Attachment 9-13 Justification for exemption for criticality monitoring for the Duncan Annex will be a combination of storage geometry, procedural controls, and amount of material in a given location at any one time. These methods will ensure that a criticality accident can not take t
place in these storage facilities.
The Duncan Annex Storage Facility Room B-84 (Attachment 9-3), which will be used for the storage of 1.3 3ercent enriched uranium oxide fuel,is constructed of a steel frame with-i-
concrete and bric < walls. The cabinets (Attachment 9-7,9-8), designed to hold the rods vertically in a safe slab geometry are fastened to the walls of the room. The 1.34 m long fuel rods (active fuel length 1.22 m) are stored in six steel cabinets in two adjacent arrays of three cabinets each.The maximum slab thickness in the storage cabinets is 14 cm, which is ~
well below the subcritical limits of 20 cm specified in ANSI /ANS-8.1-1983.
This room also contains a natural uranium metal graphite subcritical' assembly with a k-eff of 0.642. This natural uranium is licensed under source material license SUD-296. Spare fuel (natural uranium slugs weighing 1.9 kg each) may be stored in cabinets in the same room with no modifying material present.
A second room, B-77A in Duncan Annex (Attachment 9-15), will also be used for the j
storage of enriched uranium. The uranium oxide fuel (4.8 percent enriched) is stored in 1
metal racks designed to hold the rods vertically and in a slab geometry. The spacing i
between rods is maintained by the use of top and bottom grid plates. The slab thickness is ma'ntained at or below the maximum safe thickness (8.7 cm) for this fuel as specified in ANJ /ANS-8.1-1983.
Fuci when transferred from one storage or use area to another, will be transferred in 8 inch
- I diameter cylinders (similar to DOT 6M 2R containers) with only a single cylinder being i
i moved at one time. Eight inches is below the subcritical limit for'4.8 percent enriched fuel i
as specified in ANSI /ANS-8.1-1983.
l These storage areas will be used for storage only of the type of fuel, amount of fuel, and l
geometry specified in this application. Deviations from these requirements must be j
3 approved in advance by the Radiation Safety Committee or through license amendments i
from the NRC.
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!i Although Purdue University possesses a significant amount of enriched uranium, a criticality 3
accident is not a credible occurrence. At this time the FBBF, a subcritical facility, operates at a k-eff of 0.45. The calculated k-eff does not exceed 0.75 if the facility becomes flooded with water (a situation that has never occurred in the history of the building). A criticality could j
only occur if an individual were to disassemble the converters, remove the boron carbide, and i
configure the uranium in a potentially critical mass. The area would also need to be flooded or j
a suitable moderator added to achieve cAicality.The security at Purdue University is such that this type of activity could not go undetected and would be stopped before this could occur.
l Therefore an evaluation for a criticality accident, in our opinion, is not justifiable.
l 1
A more likely occurrence, although still highly remote, would be a large fire which could l
involve some of the special nuclear material. A large fire in the facility or m a glove box could 4
release some material offsite, but the following calculations will show that this release would he minimal.
A worst case analysis would assume that all the special nuclear material is involved in a fire.
l The total inventory of 550 kg of contained U-235 and 10910 kg of U-238 would be involved in j
this incident. According to Table 13 in NUREG-1140, the release fraction for uranium is O.001. Since nearly all the material is uranium oxide pellets, the release fraction is probably closer to 0.0001. Using the equation found in NUREG-1140 Section 2.3.1.3, one can compute the quantities required to deliver a 1 rem effective dose equivalent:
1 1
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RFj(li j + IIGCi + HCSi) l The effective dose equivalent is calculated to be approximately 9-90 millirem for U-235 and 30-302 millirem for U-238 utilizing the release fraction of 0.001 to 0.0001 for the uranium j
j oxide fuel. The terms IIGCi and 11CSi, for ground contamination and cloudshine can be j
ignored since the inhalation pathway would dominate. However this worst case situation would be lessened due to the following factors as outlined in 10 CFR 70.2.2(i)(2):
l i
(i)
The radioactive material is physically separated so tht only a portion of the i
material would be involved. Some of the material is stored in cage area l
cabinets (Attachment 9-6) and other material is stored in Duncan Annex.
l 1
(ii)
The uranium oxide is clad in stainless steel or aluminum.
(iii)
In the case of fires or enlosions the uranium oxide pellets should be j
considered non-volatile solics with a release fraction of 0.0001 according to NUREG-1140. A high temperature fire would be unlikely due to lack of I
combustible materials in the area.
l 3
(iv)
The material is in the form of insoluble uranium oxide.
This would sigmficantiv lower the dose to offsite individuals.
i (v)
The facility is located withm a concrete structure with walls three feet thick which would contain any explosion that takes place. The HEPA filters installed for the glove boxes and Room B28C would most likely remain intact l
l for any accident that takes place.
i These factors could reduce the consequences of a severe accident by several orders of i
l magnitude. Any offsite dose would be very low or nonexistent. The intake of soluble uranium j
would also not be likely since material handled in any quantity is in the insoluble form of i
uranium oxide.
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