ML20205N440
| ML20205N440 | |
| Person / Time | |
|---|---|
| Site: | University of Illinois |
| Issue date: | 04/12/1999 |
| From: | Marsh L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20205N439 | List: |
| References | |
| R-115-A-010, R-115-A-10, NUDOCS 9904160277 | |
| Download: ML20205N440 (28) | |
Text
v (DO NGQ l
f.
UNITED STATES t
y*
E E
NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 4 r#
1
%...+ /
UNIVERSITY OF ILLINOIS AT URBANA-CHAMPAIGN DOCKET NO. 50-151 AMENDMENT TO FACILITY LICENSE Amendment No.10 License No. R-115 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that A.
The application for an amendment to Facility License No. R-115 filed by the University of Illinois at Urbana-Champaign (the licensee) on October 5,1998, as supplemented
)
on January 12, and February 3 and 11,1999, conforms to the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the regulations of the Commission as stated in Chapter I of Title 10 of the Code of Federal Regulations (10 CFR);
B.
The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance that (i) the activities authorized by this amendment can be conducted without endangering the health and safety of the public and (ii) such activities will be conducted in compliance with the regulations of the Commission; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E.
This amendment is issued in accordance with the regulations of the Commission as stated in 10 CFR Part 51, and all applicable requirements have been satisfied; and F.
Prior notice of this amendment was not required by 10 CFR 2.105, and publication of notice for this amendment is not required by 10 CFR 2.106.
9904160277 990412 PDR ADOCK O Q1 P
~
l 4
2 2.
Accordingly, the license is amended by changes to the following paragraphs, which are hereby amended to read as follows:
2.A. Pursuant to Section 104c of the Act and Title 10, Chapter I, CFR, Part 50,
" Licensing of Production and Utilization Facilities", to possess, but not to operate the reactor in accordance with the procedures and limitations described in the application and in this license; 2.B. Pursuant to the Act and 10 CFR Part 70, " Domestic Licensing of Special Nuclear Material," to receive and possess up to 9.6 kilograms of contained uranium-235, of which 7.0 kilograms of contained uranium-235 was for use in connection with the operation of the reactor and up to 2.6 kilograms of contained uranium-235 was for use in connection with operation of a subcritical assembly in the Bulk Shielding Facility of the reactor; up to 20 grams of the contained uranium-235 is of any enrichment in the form of fission chambers and the balance is contained uranium-235 enriched to less than 20 percent in the isotope uranium-235 in the form of reactor fuel; up to 1.005 grams of plutonium in the form of reactor fuel transferred from Facility Operating License No. R-117; and to possess, but not to separate, such special nuclear material as may have been produced by the operation of the facility; and 2.C. Pursuant to the Act and 10 CFR Part 30," Rules of General Applicability to Domestic Licensing of Byproduct Material," to receive and possess up to 0.2 curie of byproduct materialin the form of components of a subcritical assembly in the Bulk Shielding Facility of the reactor transferred from Facility Operating License No. R-117; to receive and possess, but not to separate up to 0.5 curie of byproduct materialin the form of reactor fuel of a subcritical assembly in the Bulk Shielding Facility of the reactor transferred from Facility Operating License No.
R-117; and to possess, but not separate except for byproduct material produced in experiments any amount of byproduct material as may have been produced by the operation of the facility.
3.A. Maximunjfower Level The li snsee shall not operate the reactor.
3.
Acco dingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment, and paragraph 3.B. of Facility License No. R-115 is hereby amended to read as follows:
3.B. Technical Soecifications The Technical Specificanuns contained in Appendix A, as revised through Amendment No.10, are hereby incorporated in the license. The licensee shall maintain the facility in accordance with the Technical Specifications.
)
s
v
'O 3
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b.
{ " &
NLedyard B. $iarsh, Chief Evehts Assessment, Generic Communications and Non-Power Reactors Branch Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation l
Enclosure:
Appendix A, Technical Specifications Changes Date of Issuance:
April 12, 1999 I
I
u
- s ENCLOSURE TO LICENSE AMENDMENT NO.10 FACILITY LICENSE NO. R-115 DOCKET NO. 50-151 Replace the following pages of Appendix A," Technical Specifications," with the enclosed
.pages. The revised pages are identified by amendment number and contain verticallines indicating the areas of change.
Remove Insert i
i ii ii 1
1 17 17 19 19 20 20 20b 20b 21 21 22 22 23 23 24 24 l
25 25 26a 26a 26b 26b 27 27 I
30 30 31 31 32 32 33 33 34 34 36 36 37 37 40 40 41 41 em.
v 7
.i.
University ofIllinois Technical Snecifications l
TABLE OF CONTENTS Pace 1.0 DEFINITIONS 1
2.0 SAFETY I 1MITS AND SAFETY SYSTEM SETTINGS 2.1 Safety 1.imit - Fuel Element Temnerature 4
2.2 I.imitine Safety System Settine 5
3.01.lMITING CONDITIONS FOR OPERATION 3.1 Reactivity 8
3.2 Llich Power Oneration 10 3.3 Pulse Operation 11 3.4 Reactor Instrumentation 12 3.5 Reactor Safety System 14 3.6 Release of Arcon-41 16 3.7 Ventilation System 17 3.8 I imitations on Exneriments 18 3.9 Suberitical Experiments and Fuel Storace l
lisine the Bulk Shieldine Facility 19 3.10 Primary Coolant Ouality 20b 4.0 SURVEll I ANCE REOUIREMENTS 4.1 Fuel 21 4.2 Control Rods 22 4.3 Reactor Safety System 23 4.4 Emereenev Snrav Cooline System 24 April 12,1999 Amendment No.10
?
-ii-4.5 Radiation Monitoring Eauinment 25 4.6 Maindenance 26 4.7 Suberitical Experiments and Fuel Storace l
Usinu the ilulk Shieldine Facilitv 26a 4.8 Primary Coolant Ouality 26b 5.0 DESIGN FEATIIRES 5.1 Reactor Fusel 27 5.2 Reactor Buildine 28 5.3 Fuel Storage 29 5.4 Emereenev Removal of Decav Ileat 30 6.0 ADMINISTRATlVE CONTROI.S 6.1 Organization 31 6.2 Review and Audit 33 6.3 Badiation Safety 35 6.4 Procedures 36 6.5 Experiments Review and Anoroval 37 6.6 Aetton to be taken in the Event a Safety I.imit is Exceeded 39 6.7 Action to be Taken in the Event of an Abnormal Occurrence 40 6.8 Renorting Reauirements 41 6.9 Plant Operating Records 43 April 12.1999 Amendment No.10
u
. 1.0 DEFINITIONS 1.1 Reactor Shutdown - The reactor is in a shut down condition when sufricient control rods are inserted so as to assure that it is suberitical by at least $1.00 of reactivity, and a senior reactor operator is in charge of any work in progress.
1.2 Reactor Secured - The reactor is secured when:
a.
It contains insufficient fissile material or moderator present in the reactor, adjacent experiments, or control rods, to attain criticality under optimum available conditions of moderatie and reDection, or
- b. All of the fbliowing conditions are met:
- 1) Sufficient control rods are inserted so as to assure that it is suberitical by at least $1.00 of l
reactivity:
- 2) Power to the control rod magnets and actuating solenoids is off, and the key removed; and l
- 3) No work is in progress involving fuel or in-core experiments or maintenance of the core structure, control rods, or control rod drive mechanisms.
1.3 Reactor Oneration - The reactor is in operation when it is not secured or shut down.
1.4 Standard Control Rod - A standard control rod is one having rack and pinion, electric motor drive, and scram capability.
i 1.5 Transient Control Rod - A transient control rod is one that is pneumatically operated and has scram I
capability.
t l.6 Onerable - A system or device is operable when it is capable of performing its intended functions in a normal manner.
1.7 Cold Critical - The reactor is in the cold critical condition when it is critical with the fuel and bulk water temperatures the same (~ 40'C).
1.8 Steadv-State Mode - The reactor is in the steady-state mode when the reactor mode selection switch is in the steady-state or automatic position. In this mode. reactor power is held constant or is changed on periods greater than three seconds.
1.9 Souare-Wave Mode - The reactor is in the square-wave mode when the reactor mode selection switch is in the square-wave position. In this mode, the reactor power is increased on periods less than one second, is held at constant power by automatic motion of the control rods, and is then reduced by shutting the reactor down.
I April 12,1999 Amendment No.10 l
1
v
. 3.7 Ventilation Svstem Ap,licability This specification applies to the operation of the reactor faciliiy ventilation system.
Obiective The objective is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radioactive materials resulting from reactor operation or during fuel movements.
Specification The reactor shall not be operated and fuel shall not be moved unless the facility ventilation l
system is in operation, except for periods of time not to exceed two days to permit repairs to the system. During such periods of repair:
- a. The reactor shall not be operated at power levels above 1 Mw;
- b. The reactor will not be operated in the pulse mode: and
- c. The reactor shall not be operated with experiments in place whose failure could result in the release of radioactive gases or aerosols, and
- d. Fuel shall not be moved.
13 asis it is shown in Section XIV of the SAR that operation of the ventilation system sufficiently reduces off-site doses to below 10 CFR Part 20 limits in the event of a TRIGA fuel element failure. The specifications governing operation of the reactor while the ventilation system is undergoing repair preclude the likelihood of fuel element failure during such times. It is shown in Section IV of the SAR that, if the reactor were to be operating at a power level of 1 Mw, fuel element failure will not occur, even if all the reactor tank water were to be lost.
April 12.1999 Amendment No.10
v
. 3.9 Suberitical Experiments and Fuel Storace Usine the Bulk Shieldine Facility l
Annlicability This specification applies to suberitical arrays and storage of fuel elements located external to l
the reactor in the bulk shielding facility.
Objective The objective is to assure that accidental criticality of the stored fuel or suberitical experiment will not occur, proper radiation monitoring is present and pool level is maintained for radiation protection.
Specifications
- a. The effective multiplication constant (kod of the suberitical facility shall not exceed 0.95 for assemblies of fuel elements using natural uranium fuel and shall not exceed 0.99 fbr assemblies of TRIGA fuel elements.
- b. For an assembly w here it is expected that kor could exceed 0.90, a step-wise procedure, in which kor is determined using the inverse multiplication method, shall be followed for the initial loading of the assembly.
- c. During the first loading of any TRIGA fuel suberitical assembly, a safety control rod worth at least 80 cents in the final assembly shall be provided in the assembly. The control rod shall be held in the withdrawn position by an electromagnet, and shall have scram capability provided by manual switches and by a high radiation signal from a monitor located near the assembly.
The maximum setpoint fbr the high radiation scram shall be 100 mr/hr.
- d. The initial use of the reactor as a source of neutrons for the suberitical assembly shall follow a step-wise procedure for steady-state power increases and power transients.
I c.
A portable radiation monitor shall be used during the initial assembly and startup of the experiment to determine dose rates in its vicinity.
f.
During periods when the Bulk Shielding Facility (BSF) or TRIGA pool is used for fuel storage a continuous air monitor shall be in operation in the reactor bay and an area radiation monitor shall be in operation above the pool. The continuous air monitor and/or area radiation monitor (s) may be out of service for up to ten days provided that no fuel handling takes place.
- g. I)uring periods when the Bulk Shielding Facility (BSF) or TRIGA pool is used for fuel storage the pool level will be maintained at a level at least six (6) feet above the top of the fuel elements.
April 12,1999 Amendment No.10
U C Basis The performance of subcritical experiments external to the reactor was evaluated and authorized for the original TRIGA Mark 11 reactor at The University ofIllinois by Amendment No. 6 to License No. R-69. It was concluded at that time, and subsequently shown by actual operation, that the above specifications provided adequate assurance of safe operation. Since it has been shown that the presence of the subcritical assemblies external to the reactor had negligible effect on reactor operation, it is concluded that such experiments can be performed with a similar degree of safety adjacent to the Illinois Advanced TRIGA. Experience has shown through historical usage of the fuel storage racks that a minimum level of six feet of water above the fuel provides adequate radiation shielding.
i April 12,1999 Amendment No.10 J
i
-20b.
1 i
3.11 Primarv Coolant Ouality 1
l l
Annlicability This specification applies to the quality of the primary coolant water m contact with the cladding l
i l
of the fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility.
l Obiective a) To limit the possibility for corrosion of the cladding on the fuel elements.
l b) To limit the concentration of dissolved materials which could be activated by neutron l
exposure.
)
l l
Specification The Advanced TRIGA and/or the suberith 1 assembly in the Bulk Shielding Facility shall not be operated if the conductivity of the primr y coolant water in the associated tank is higher than 4 pmho/cm.
l JJasis a) Corrosion may occur continuously in a water-metal system. In order to limit the rate of corrosion, and thereby extend the life and integr:ty of the fuel cladding, a water clean-up system is required. Experience with water quality control at many reactor facilities has shown that maintenance within the specified limit provides acceptable corrosion control.
b) Limiting the concentration of material dissolved in the water limits the radioactivity of neutron activation products. This tends to decrease the inventory of radionuclides in the entire coolant l
system, which will decrease personnel radiation exposure during both maintenance and operations. This trend is consistent with the ALARA principle.
l l
i 1
April 12.1999 Amendment No.10
c-u
'o 4.0 SURVEll l.ANCE REOUIREMENTS l
4.1 Eugj Annlicability This specification applies to the surveillance requirement for the fuel elements.
Obiective The objective is to assure the dimensions of the fuel elements remain within acceptable limits.
_S, neci6 cations a.
The standard fuel elements shall be measured for length and bend at intervals separated by not more than 1000 pulses of magnitude greater than 51.00 of reactivity or by an integrated reactivity of $3.000. I.ow hydride elements shall be measured annually not to exceed 14 l
months or at intervals separated by not more than 50 pulses, whichever is the lesser, if they se used for pulsed operation in the TRIGA core. New standard fuel elements shall be measured at intervals not to exceed 500 pulses until 1000 pulses have been exceeded.
- b. Standard thermocoupled fuel elements shall be checked at the same intervals as in a. above by the removal of the element from the core region and a visual check of the cladding.
c.
A fuel element indicating an elongation greater than 1/4 of an inch over its original length or a lateral bending greater than 1/16 of an inch over its original bending shall be considered to be damaged and shall not be used in the core for further operation.
- d. Fuel elements in the B-and C-hexagonals shall be measured for possible distortion in the event that there is indication that fuel temperatures greater than the limiting safety system setting on temperature may have been exceeded.
Basis The most sever stresses induced in the fuel elements result from pulse operation of the reactor, during which differential expansion between the fuel and the cladding occurs and the pressure of the gases within the elements increases sharply. It is shown in Section 111 of the SAR that the above limits on the allowable distortion of a fuel element correspond to strains that are considerably lower than the strain ext eted to cause rupture of a fuel element.
April 12,1999 Amendment No.10
I l 4.2 Control Rods Applicability l
This specification applies to the surveillance requirements for the control rods.
Obiective l
The objective is to assure the integrity of the control rods.
Sneeifications a.
b.
The control rods shall be visually inspected for deterioration biennially not to exceed 30 c.
months.
i 1
d.
l Bases l
The visual inspection of the fuel follower control rods specified has been shown to be adequate based on prior experien;e with a lack of fuel cladding deterioration over time.
1 I
April 12,1999 Amendment No.10
v
\\
o 4.3 Reactor Safety System Annlicability I
This specification applies to the surveillance requirements for the measuring channels of the reactor safety system.
Obiective The objective is to assure that the safety system will remain operable and will prevent the fuel temperature safety limit from being exceeded.
Snecifications A channel test of each of the reactor safety system channels shall be performed prior to each a.
day's operation or prior to each operation extending more than one day,
- b. A channel check of the fuel element temperature measuring channels shall be performed daily a henever the reactor is in operation at power les els greater than 50 kw or when pulse operation is planned.
c.
A channel check of the power level measuring channels shall be performed daily whenever the i
reactor is in operation.
l d.
e.
11 asis The daily tests and channel checks will assure that the safety channels are operable. The semi-annual calibrations and verifications will permit any long-term drift of the channels to be corrected.
April 12,1999 Amendment No.10
-24 Section 4.4 Emergency Spray Cooling System intentionally deleted.
j l
l l
April 12,1999 Amendment No.10
- 4.5 Radiation Monitoring EauinmenJ Applicability This specificatica applies to the radiation monitoring equipment required by Section 3.4 and 3.9 of these specifications.
Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings.
Specification The alarm set points fbr the radiation monitoring instrumentation shall be verified monthly not to exceed six weeks.
Ilasis Because of the redundancy of radiation monitoring instrumentation provided, monthly l
surveillance of the equipment will be adequate to assure that sufficient protection against radiation is available.
1 I
l
[
I April 12,1999 Amendment No.10
m
! 4
-26a-4.7 Suberitical Experiments and Fuel Storace Usine the Bulk Shieldine Facility l
l Annlicability l
This specification applies to the surveillance requirements associated with the suberitical assembly and storage of fuel elements in the bulk shielding facility.
Obiective To ensure safe operation of the suberitical assembly, to ensure that the radiation monitoring equipment is operating properly, and that the pool level is maintained for radiation protection.
Specification a) The reactivity worth of the control rod shall be determined annually (interval not to exceed 6fteen months). The surveillance may be deferred indennitely when the suberitical assembly is not being utilized, but shall be the first operation performed u hen the suberitical assembly is to be operated.
b) Control rod drop time shall be determined annu:.lly (interval not to exceed 6fteen months'; The drop time from fully withdraw n to 90 percent of full reactis ity insertion shall be less than one second. The surveillance may be deferred indefinitely w hen the suberitical assembly is not being utilized, but shall be performed prior to operation of the assembly.
c) The radiation monitor utilized for a high radiation signal scram shall be calibrated and veri 0ed operable annually (interval not to exceed fifteen months). The surveillance may be deferred indefinitely when the suberitical assembly is not being utilized, but shall be performed prior to operation of the assembly.
d) Approximately 10 % of the fuel elements in the Bulk Shielding Facility or TR!GA pool, shall be l
visually inspected annually for any indication of deterioration or distortion (interval not to exceed Ofteen months) such that all of the elements are inspected over a ten year period (interval not to j
exceed ten and one half yerrs). If any indication of deterioration or distortion is noted the element l
shall be removed to other storage and all elements shall be inspected within one week. Other storage j
shall be any other approved fuel storage area at the Nuclear Reactor Laboratory.
e) The manual and high radiation scrams shall be veri 0ed operable daily prior to operation of the suberitical assembly. This speciGeation is only applicable on days w hen the suberitical assembly is to tie operated f) The Bulk Shielding Facility pool level and TRIG A pool level shall be checked on a weekly (not to exceed ten days) basis uhen used for fuel storage.
i Basis The reactivity worth is measured to assure that control of the subcritical assembly can be maintained. The control rod drop time veri 0es the scram capability of the control rod. Calibration ar.d veri 0 cation of operability of the radiation mnaitor serifies the scram capability of the monitor. The visual inspection of the fuel elements speciGed had been shown to be adequate based on prior experience with a lack of fuel deterioration over time.
April 12,1999 Amendment No.10
-26b-4.8 Primary Coolant Ouality Anolicability This specification applies to the surveillance t 1 the quality of the primary coolant water in contact with the cladding of the fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility.
l Obiective The objective is to ensure that the quality of the primary coolant water in contact with the fuel cladding does not deteriorate over extended periods of titre even if the reactor is not operated.
Specifications The conductivity of the primary coolant water in contact with the cladding of Ge fuel in the Advanced TRIGA pool and in the Bulk Shielding Facility shall be measured at least once every l
two weeks (interval not to exceed 21 days) and shall not exceed 5 pmho/cm for more than 5 g
consecutive days. If the conductivity of the water exceeds 4 pmho/cm the sampling frequency i
shall be increased to daily until the conductivity drc ps below 4 pmho/cm. If the conductivity exceeds 5 pmho/cm for more than five consecutive days the fuel shall be removed from the tank to storage until such time that the conductivity has been restored to below 4 pmho/cm.
Basis Section 3.11 ensures that the water quality is acceptable during reactor operation. Section 4.8 ensures that the fuel cladding is not exposed to a significantly more corrosive environment for an extended period of time in the event that the reactor is not actually operated.
April 12,1999 Amendment No.10
r v
1
-2 7-l 5.0 DESIGN FEATURES 5.1 Reactor Fuel Annlicability j
l This specification applies to the fuel elements used in the reactor core for a critical mass.
l l
Obiective The objective is to assure that the fuel elements are of such a design and fabricated in such a manner as to pennit their use with a high degree of reliability with respect to their mechanical i
integrity, t
l l
Snecifications l
a.
Standard Fuel Element: The standard fuel element shall contain uranium-zirconium hydride, clad in 0.020 inch of 304 stainless steel. It shall contain a maximum of 9.0 weight percent uranium uhich has a maxi'r.um enrichment of 20 percent. There shall be 1.55 to 1.80 hydrogen atoms to 1.0 zirconium atom.
- b. l ow Hvdride Fuel Element: This fuel element shall contain uranium-zirconium hydride, clad in 0.030 inch of aluminum or 0.020 inch of 304 riainless steel. It shall contain a maximum of 9 weight percent uranium which has a maximum enrichment of 20 greent. There shall be 0.9 to 1.54 hydrogen atoms to 1.0 zirconium atorn.
c.
Ioadine: The elements shall be placed in a closely packed array except for experimental facilities or for single positions occupied by control rods and a neutron start-up source.
Basis These types of fuel elements have a long history of successful use in TRIGA reactors.
I I
April 12,1999 Amendment No.10 1
k l
-3 0-I l
f Section 5.4 Emergency Removal of Decay lieat intentionally deleted.
I I
\\
l l
i 4
l l
i April 12,1999 l
Amendment No.10 I
m z
1 6.0 ADMINISTRATIVE CONTROLS 1
l l
6.1 Oreanization l
l 6.1.1 Structure and Resnonsibility The reactor facility shall be an integral part of the Department of Nuclear Engineering of the a.
University ofIllinois. The reactor shall be related to the University structure as shown in l
Chart I.
- b. The reactor facility shall be under the supervision of the Reactor Administrator who shall have been qualified as a licensed senior reactor operator for the reactor. He shall be responsible for assuring that all operations are conducted in a safe manner and within the limits prescribed by I
the facility license and the provisions of the Nuclear Reactor Committee.
I c.
There shall be a Reactor llealth Physicist responsible for assuring the day to day and routine radiological safety activities at the Nuclear Reactor Laboratory. The University ofIllinois Radiation Safety Officer shall be responsible for monitoring, planning and promoting radiological safety at the Nuclear Reactor Laboratory. lie has the responsibility and authority to stop, secure or otherwise control as necessary any operation or activity that poses an unacceptable radiological hazard.
CII ART I Ilead of Department of Division of Environmental i
Nuclear Engineering IIcalth and Safety l
l l-l Radiation Safety Ollicer l
Nuclear Reactor i
Committee Reactor Adm..imstrator i
i s______
L l
I Reactor llealth Physicist l
l l
1 l
CliART 1: Administrative organization of the reactor facility. Dashed lines indicate reporting paths outside the operational chain of supervision, indicated by solid lines.
1 0
April 12,1999 Amendment No.10 l
l
l 32-l I
6.1.2 Staffine a.
The minimum staffing at the Nuclear Reactor Laboratory shall be:
1.
Reactor Administrator. This individual shall meet the requirements of ANSl/ANS-15.4-1988 "American National Standard for the Selection and Training of Personnel for Research Reactors" for a Level Two individual.
2.
Reactor 11ealth Physicist. This individual shall meet the requirements of ANSI /ANS-15.4-1988 "American National Standard for the Selection and Training of Personnel for Research Reactors" for a Level Three individual in addition to training in health physics.
b.
A list of reactor personnel by name and telephone number shall be readily available to the UlUC Division of Public Safety dispatcher. One of these individuals shall be reachable and able to respond to the facility within approximately one hour. The list shall include:
1.
Campus Radiation Safety Oflicer 2.
Reactor Administrator 3.
lie.id Department of Nuclear Engineering 4.
Reactor llealth Physicist 5.
Licensed operators Events requiring the presence at the facility of a Senior Reactor Operator:
c.
1.
Initial startup and approach to power.
2.
All fuel or control rod relocations.
l 3.
Relocation of any in-core experiment with a reactivity worth greater than one dollar.
4.
Recovery from unplanned or unscheduled shutdown or significant power reduction (In these I
instances, documented verbal concurrence from the Senior Reactor Operator is required).
6.1.3 Selection and Trainine of Personnel The Reactor Administrator is responsible for the training and requalification of the facility reactor operators and senior reactor operators. The selection. training, and requalification of operations personnel shall be consistent with all current regulations and guidelines.
1 April 12,1999 Amendment No.10 l
l
. l 6.2 Review and Audit 6.2.1 Charter and Rules a.
The Reactor Committee shall be composed of at least five voting members, one of whom shall be a llealth Physicist designated by the campus Radiation Safety Of6cer for the University, one whom shall be the Reactor Administrator, and one whom shall be the Reactor Health Physicist. The remaining members shall be appointed by the Head of the Department of Nuclear Engineering, so as to maintain a balanced knowledge of reactor safety and regulation.
b.
The Reactor Committee shall have a written statement de6ning such matters as the authority of the committee, the subjects within its purview, and other such administrative provisions as are required for the effective functioning of the Reactor Committee. Minutes of all meetings of the Reactor Committee shall be kept.
c.
A quorum of the Reactor Committee shall be a majority of not less than one half of the members and the reactor staff shall not constitute a voting majority.
- d. The Reactor Committee shall meet at least semiannually not to exceed nine months 6.2.2 Review Function The review function of the Committee shall include, but is not limited to the following:
Determinations that proposed changes in equipment. systems, tests, experiments, or procedures do not a.
involve an unreviewed safety question.
b.
All new procedures and major revisions thereto having safety significance, proposed changes in reactor facility equipment, or systems having safety significance.
All new experiments or classes of experiments for determination that an unreviewed safety question c.
does not exist.
j l
d.
Proposed changes in the technical specifications or beense.
Violations of technical speci6 cations or license, e.
f.
Operating abnormalities having safety signi6cance.
g.
Reportable occurrences as listed in 6.8.
j
- h. Audit reports.
A written report or minutes of the findings and recommendations of the Committee shall be submitted to the llead.
Department of Nuclear Engineering, and the Reactor Committee members in a timely manner after each meeting.
April 12,1999 Amendment No.10 1
. 6.2.3 Audit Pmetion The audit function of the Reactor Committee shall include selective (but comprehensive) examination of l
records, logs, and other documents. Discussions with cognizant personnel and observation of evolutions l j
should be used also as appropriate. In no case shall the individual immediately responsible for the area perform an audit in that area. The following items shall be audited:
l l
Facility operations for conformance to the technical specifications and license, at least once per a.
calendar year (interval between audits not to cxceed 15 months).
- b. The requalification program for the operating staff, at least every other calendar year (interval between audits not to exceed 30 months).
l The action taken to correct those deficiencies that may occur in the reactor facility equipment, l
c.
systems, structures, or methods of operations that affect reactor safety, at least once per calendar year (interval between audits not to exceed 15 months).
- d. The reactor facility emergency plan, and implementing procedures at least once every other i
calendar year (interval between audits not to exceed 30 months).
Deficiencies uncovered that affect reactor safety shall immediately be reported to the 11ead, Department of Nuclear Engineering. A written report of the findings of the audit shall be submitted to the Reactor Committee within three months after the audit is completed and then forwarded to the llead. Department of Nuclear Engineering.
l I
l l
i April 12,1999 Amendment No.10
t.
l '
6.4 Procedures Written procedures shall be prepared, reviewed and approved prior to initiating any of the activities listed in this section. The procedures shall be reviewed by the Reactor Committee and approved by the Reactor Administrator and such review and approval shall be documented in a timely manner. The procedures shall be adequate to assure the safety of the reactor, but should not preclude the use ofindependent judgment and action should the situation require such.
a.
Startup, operation, and shutdown of the reactor,
- b. Installation or removal of fuel elements, control rods, experiments, and experimental facilities.
c.
Actions to be taken to correct specific and foreseen potential malfunctions of systems or components, including responses to alanns, suspected primary coolant leaks, and abnormal reactivity changes.
- d. Emergency conditions involving potential or actual release of radioactivity, including provisions for evacuation, reentry-entry, recovery, and medical support.
e.
Maintenance procedures which could have an effect on reactor safety.
f.
Periodic surveillance of reactor instrumentation and safety systems, area monitors and continuous air monitors.
g Personnel radiation protection, consistent with applicable regulations or guidelines. The procedures shall include management commitment and programs to maintain exposures and releases As Low As is Reasonably Achievable (ALARA).
- h. Implementation of physical security plan.
Substantive changes to the above procedures shall be made only after documented review by the Reactor Committee and approval by the Reactor Adminetrator. Minor modifications to the original procedures which do not change their original intent may be made by the Reactor Administrator. Temporary l
deviations from the procedures may be made by the responsible Senior Reactor Operator or higher individual present, in order to deal with special or unusual circumstances or conditions. Such deviatiens l
shall be documented and reported to the Reactor Administrator.
April 12,1999 Amendment No.10
r u
l P
-3 7-l l
6.5 Experiments Review and Anoroval i
All new experiments or class of experiments utilizing the reactor shall be evaluated by the a.
experimenter and a staff member approved by the Reactor Committee. The evaluation shall be reviewed by a licensed senior reactor operator of the facility and the Reactor llealth Physicist to assure compliance with the provisions of the utilization license, the Technical Speci0 cations and 10CFR20.
l If, in theirjudgment, the experiment meets uith the above provisions and does not constitute a threat to the integrity of the reactor, they shall submit it to the Reactor Administrator for review. If the l
Reactor Administrator agrees e ith the evaluation by the senior reactor operator and the Reactor l
licalth Physidst he shall submit the experiment to the Reactor Committee for review as indicated in Section 6.2. The experiment shall be approved in writing by the Reactor Administrator prior to initiation. When pertinent, the evaluation shall include the following:
1 1.
The reactivity worth of the experiment.
i l
2.
The integrity of the experiment, including the effects of changes in temperature, pressure, or j
chemical composition.
3.
Any physical or chemical interaction that could occur with the reactor components.
4.
Any radiation hazard that may result from the activation of materials or from external beams.
l b.
The Reactor Committee review of an experiment shall be performed prior to the Grst experiment and shall be documented in writing and shall consider at a minimum the following:
1.
The purpose of the experiment.
l 2.
A procedure for the performance of the experiment.
3.
The evaluation approved by a licensed senior reactor operator.
4.
Determination that the experiment does not involve an unreviewed safety question.
c.
Substantive changes m previously approved experiments shall be made only after review by the l
Reactor Committee and approved in writing by the Reactor Administrator. Minor changes that do not signincantly alter the safety analysis of the experiment may be approved by the Reactor l
I Admiristrator.
l d.
For the irradiation of materials, the applicant shall submit a request to the Reactor 11ealth Physicist.
l This request shall contain at a minimum information on the target material including the amount, chemical form, and expected radiological hazard for the desired irradiation period. For routine l
irradiations (which do not contain nuclear fuel or known explosive materials and which do not constitute a significant threat to the integrity of the reactor or to the safety ofindividuals), the approval for the Reactor Comn.ittee may be made by the Reactor IIcalth Physicist.
l April 12.1999 Amendment No.10
r
. 6.7 Action to be taken in the Event of an Abnormal Occurrence In the event of an abnormal occurrence, as defined in Section 1.14 of the specifications, the following action shall be taken:
a.
The reactor shall be shutdown and the Reactor Administrator shall be notified and corrective l j
i action taken prior to resumption of operations. The Reactor Administrator shall authorize l
resumption of operations.
1 b.
A report shall be made which shall include an analysis of the cause of the occurrence, efficacy l
of corrective action and recommendations for measures to prevent or reduce the probability of reoccurrence.
c.
The occurrence shall be reviewed by the Reactor Committee at the next schedulea meeting.
d.
A report shall be submitted to the USNRC in accordance with Section 6.8 of these l
specifications.
l i
April 12,1999 Amendment No.10
o-3 I
1
,, 6.8 Reoortine Renuirements l
In addition to the requirements of applicable regulations, and in no way substituting therefore, reports shall be made to the USNRC as follows:
There shall be a report not later tt n the following working day by telephone and confirmed in writing a.
by facsimile or similar conveyance to the Regional Administrator, USNRC, Region 111 and the USNRC headquarters operations center to be followed by a written report that describes the circumstances of the es ent within 14 days to the Document Control Desk. USNRC 11eadquarters, and a copy to the l
Regional Administrator, USNRC, Region ill of any of the following:
1.
Release of radioactivity frem the site above allowed limits.
2.
Violation of safety limits.
1 l
l 3.
Any signiGeant variation from measured values from a corresponding predicted or previously measured value of safety-connected operating characteristics occurring during operation of the reactor.
4.
Incidents or conditions relating to operation of the facility which prevented cr could have prevented the performance of engineered safety features as described in these specifications.
5.
Any abnormal occurrences as defined in Section 1.14 of these specifications.
b.
A report within 30 day s (in writing to the Document Control Desk, USNRC lleadquarters) of:
1.
Any substantial variance from performance specifications contained in these specifications.
2.
SigniGeant changes in the transient or accident analysis as described in the Safety Analysis Report.
3.
Pennanent changes in the facility organization involving the Reactor llealth Physicist, Reactor l
Administrator or Department liead.
l l
A report within 60 days after criticality of the reactor (in writing to the Document Control Desk, c.
USNRC lleadquarters) upon receipt of a new facility license or an amendment to the license authorizing an increase in reactor power level or the installation of a new core, describing the measured values of the operatir.3 conditions or characteristics of the reactor under the new conditions, including:
)
1.
Total control rod reactivity worth.
2.
Reactivity worth of the single control rod of highest reactivity worth.
' 3.
Total and indisidual reactivity worths of any experiments inserted in the reactor.
l 4.
Minimum shutdown margin both at room and operating temperatures.
l i
i i
April 12,1999 Amendment No.10