ML20137L875
| ML20137L875 | |
| Person / Time | |
|---|---|
| Site: | Pennsylvania State University |
| Issue date: | 01/31/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1158, NUDOCS 8601280120 | |
| Download: ML20137L875 (79) | |
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l Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at Pennsylvania State University U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation January 1986 f***%,,
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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sourc 1.
The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box Washington, DC 20013 7082 ,
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publication it is not intended to be exhaustive. ,
Referenced documents available for inspection and copying for a fee from the NRC Public D ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspe and Enforcement bulletins, circulars, information notices, inspection and investigation no Licensee Event Reports; vendor reports and correspondence; Commission papers; and app licensee documents and correspondence.
The following Program: documents in the NUREG series are available for purchase from the GPO formal NRC staff and contractor reports, NRC-sponsored conference proceedings a NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in th Federal Regulations, and Nuclear Regulatory Commission issuances.
Documents available from the National Technical information Service i reports and technical reports prepared by other federal agencies and reports prepared by the A Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Documents available from public and special technical libraries include all open literatu such as books, journal and periodical articles, and transactions. Federal Register notices, state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non-NRC proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draft reports are available free, to the extent of supply, upon w to the Division mission, Washington,ofDCTechnical 20555. information and Document Control, U S Nuclear om- Regulatory C Copies of industry codes and standards used in a substantive manner in the NRC reg are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are ava there for reference use by the public. Codes and standards are usuallyy copyrighted be and m purchased from the originating organization or, if they are American National Standards f American National Standards institute,1430 Broadway, New York, NY 10018 .
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-_-__--_2- - _T2 : -- - T :- - -~ 2 Safety Evaluation Report related to the renewal of the operating license for the Research Reactor at Pennsylvania State University U.S. Nuclear Regulatory .
Commission Office of Nuclear Reactor Regulation January 1986 f* **%.
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ABSTRACT This Safety Evaluation Report for the application filed by the Pennsylvania State University for a renewal of Operating License R-2 to continue to operate the Pennsylvania State University Breazeale reactor (PSBR) has been prepared by the Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facility is owned and operated by the Pennsylvania State Uni-versity and is located on the campus in University Park, Pennsylvania. On the basis of its technical review, the staff concludes that the reactor facil-ity can continue to be operated by the university without endangering the health and safety of the public or the environment.
I PSU SER iii
i 1
TABLE OF CONTENTS l
P_ age iii t
ABSTRACT.............................. ...........................
i
'*****'''' 1-1 1 1 INTRODUCTION................ - -
t
' 1.1 Summary and Conclusions of Principal Safety. 1-1 Considerations........................................
1-2
- 1. 2 History.................................................
.' 1. 3 Rea c to r De s c ri p t i o n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
1-3 ,
1-3 1.4 Shared Facilities and Equipment.........................
13
- 1. 5 Comparison With Similar Facilities...................... 1-3 !
' 1. 6 Nuclear Waste Policy Act of 1982........................
4 21 2 SITE CHARACTERISTICS.............. ..........................
2-1 2.1 Geography............... . .............................
2-1
- 2. 2 Demography..............................................
2.3 Nearby Industrial, Transportation, and Military
~
2-1 Facilities............................................
l 2- 1 2.4 Meteorology............................................. 2-4 ;
- 2. 5 Hydrology............................................... 2-4
- 2. 6 Geology and Seismology..................................
j 2-6 ,
2.7 Conclusion.............................................. >
I DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS................ 3-1
- 3 S
3-1
- 3.1 Wind and Water Damage................................... 3-1 l 3. 2 Seismically Induced Reactor Damage......................
3.3 Mechanical Systems and Components....................... 3-1 3-1 3.4 Conclusion.............................................. ,
4-1 4 REACT 0R......................................................
4-1 4.1 Reactor Core............................................
Reflector Element: and Core Support Structure.... 4-1 4.1.1 4-1 4.1.2 Fuel Elements.................................... 4-4
~
4.1.3 Neutron Source and Holder........... ............ 4-4 4.1.4 Control Rods.....................................
4.2 Reactor Pool and Biological Shield...................... 4-4 4-7 4.3 Reactor Instrumentation................................. 4-7 4.4 Dynamic Design Evaluation...............................
4-7 4.4.1 Excess Reactivity and Shutdown Margin............ 4-8 4.4.2 Normal Operating Conditions...................... 4-8 4.4.3 Assessment.......................................
1 V
PSU SER ,
TABLE OF CONTENTS (Continued)
Page
- 4. 5 Functional Design of Reactivity Control System..... .. 4-9 4.5.1 Standard Rod Drive Assembly.... 4-9 4.5.2 Transient Rod Drive Assembly.... . ..... . .....
.... .......... 4-9 4.5.3 Scram-Logic Ci rcuitry and Interlocks. . . . . . . . . . . 4-10 4.5.4 Assessment...... .. ................ . ... ... 4-10
- 4. 6 Operational Procedures. . . . . . . . . . . .... ... 4-11 4.7 Conclusion........... .. ......... ............. ......
..... ... 4-11 5
REACTOR COOLANT AND ASSOCIATED SYSTEMS..... .. ..... ... ... 5-1 1
5.1 Cooling System..... ....... . .. ................ .... 5-1
- 5. 2 Primary Coolant Purification System. . . .. . ...... ... 5-1 5.3 Primary Coolant Makeup System. . . . ........... 5-1
- 5. 4 Conclusion............................................. ..... .... 5-1 6
ENGINEERED SAFETY FEATURES. .... .............. ............ 6-1 6.1 Ventilation System. .. ...... .. ....... .......... . 6-1
- 6. 2 Conclusion. ....... ............... .... .......... .. 6-1 7 CONTROL AND INSTRUMENTATION SYSTEMS....................... . 7-1 7.1 Reactor Control System..... ................... ....... 7-1 l 7. 2 Instrumentation System......... ............. ......... 7-4 7.2.1 Nuclear Instrumentation......... ............... 7-4 7.2.2 Nonnuclear Instrumentation............ ... ..... 7-5
- 7. 3 Conclusion....................... .................... 7-6 8
1 ELECTRIC POWER SYSTEM.................................. .... 8-1 8.1 Normal Power .............. ....... ................... 8-1 1
8.2 Emergency Power............ ................ ........ 8-1 8.3 Conclusion............................................. .. 8-1 9
AUXILIARY SYSTEMS.... ...................................... 9-1 '
9.1 Ventilation System............ ........................ 9-1 i 9.2 F i re P ro t e c t i o n Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1
, 9. 3 Communicolion System. ......................... ....... 9-1 9.4 Compressed Air System.................................. 9-1
! 9. 5 Air Conditioning System................................ 9-1
- 9. 6 Fuel Handling and Storage.. ........................... 9-1 9.7 Conclusion..... ....................................... 9-2 PSU SER vi
i TABLE OF CONTENTS (Continued)
Page ;
........................ .. 10-1 j 10 EXPERIMENTAL PROGRAM............
10.1 Experimental Facilities.... ........... .............. 10-1 10-1.
10.1.1 Beam Ports................... .................
10-1 10.1.2 Thermal Column....... .........................
10-3 10.1.3 Central Thimble...... ... ..................... 10-3 10.1.4 Vertical Tubes.............................. ..
10-3 10.1.5 Pneumatic Transfer Systems. . .................
1 10.2 Experimental Review................................... 10-4 10-4 10.3 Conclusions........................ ............ .....
............ ........ 11-1 11 RADI0 ACTIVE WASTE MANAGEMENT.. .......
11-1 11.1 ALARA Commitment... .... .............................
11.2 Waste Generation and Handling Procedures.............. 11-1 11.2.1 Solid Waste...................... ............ 11-1 11.2.2 Liquid Waste.................................. 11-1 11.2.3 Airborne Waste................................ 11-1 11.3 Conclusions................ .......................... 11-2 RADIATION PROTECTION PR0 GRAM...................... ......... 12-1
! 12 i
12.1 ALARA Commitment...................................... 12-1 l
12.2 Health Physics Program.............................. . 12-1 12.2.1 Health Physics Staffing....................... 12-1 12.2.2 Procedures.................................... 12-1 12.2.3 Instrumentation........... ................... 12-1 12.2.4 Training...................................... 12-2 12-2 12.3 Radiation Sources.....................................
12.3.1 Reactor....................................... 12-2 12-2 12.3.2 Extraneous Sources............................
12-2 12.4 Routine Monitoring....................................
12.4.1 Fixed Radiation Monitoring System............. 12-2 12.4.2 Experimental Support.......................... 12-3 12.5 Occupational Radiation Exposures.... ................. 12-3 12.5.1 Personnel Monitoring Program.................. 12-3 12-3 12.5.2 Personnel Exposures...........................
vii PSU SER
TABLE OF CONTENTS (Continued)
Page 12.6 Effluent Monitoring................................... 12-4 12.6.1 Airborne Effluent............. .. ............ 12-4 12.6.2 Liquid Effluent................ .............. 12-4 12.6.3 Environmental Monitoring...................... 12-4 12.6.4 Potential Dose Assessment..................... 12-4 12.7 Conclusions.... ......... ....... ......... .. ... ... 12-4 13 CONDUCT OF OPERATIONS.......... ........................... 13-1 13.1 Overall Organization....... ........ ................. 13-1 13.2 Training............. ........................ ....... 13-1 1
13.3 Operational Review and Audits..... ................... 13-1 13.4 Emergency Planning.................................... 13-1 13.5 Physical Securi ty Plan. . . . . . . . . . . . ... . ............. 13-1 13.6 Conclusion...... ..................................... 13-1 14 ACCIDENT ANALYSIS. . . .. ... ........................ ..... 14-1 14.1 Fuel-Handling Accident................................ 14-1 14.1.1 Scenario...................................... 14-2 14.1.2 Assessment........... ........................ 14-3 14.2 Rapid Insertion of Reactivity....................... . 14-3 14.2.1 Scenario.......... ........................... 14-4 14.2.2 Assessment.... .... .......................... 14-4 14.3 Loss-of-Coolant Accident.............................. 14-5 14.4 Misplaced Experiments................................. 14-5 14.5 Effects of Fuel Aging................................. 14-6 14.6 Conclusion............................................ 14-7 15 TECHNICAL SPECIFICATIONS.................................... 15-1 16 FINANCIAL QUALIFICATIONS.................................... 16-1 17 OTHER LICENSE CONSIDERATIONS................................ 17-1 17.1 Prior Reactor Utilization............................. 17-1 17.2 Conclusion............................................ 17-2 18 CONCLUSIONS................................................. 18-1 19 REFERENCES.................................................. 19-1 PSU SER vill
LIST OF FIGURES 2-2 2.1 University Park Campus. .................. ....... ... ...
2-3
- 2. 2 Pennsylvania State Breazeale Reactor............ . .......
2-5 2.3 Spring Creek Drainage Basin.. . .. .. ............... ...
4-2 4.1 Cutaway View of the PSBR Facility..... ............ ...... 4-5 l 4.2 Typical PSBR Fuel Element.... .............. .. . ......
4-6 4.3 PSBR Typical Core Configuration.. ... ....................
PSBR Coolant System. . . ..... ... .............. . . .... 5-2 l 5.1 7-2 7.1 PSBR Control Console Instrumentation.............. ..... ..
7-3 7.2 Schematic of the PSBR Control Console Instrumentation..
Locations of PSBR Facilities for Experimenters............ 10-2 10.1 13-2 13.1 Organization Chart....... .... .. . ..... ..............
LIST OF TABLES Principal Design .. ... . ............... .. .... 4-3 4.1 . ...
Minimum Reactor Safety System Channels...... ......... ... 7-5 7.1 7-5 7.2 Console Alarm Settings.. ... ............... .........
.. ......................... 10-1 10.1 Beam Port Measurements.....
Number of Individuals in Exposure Interval........... .... 12-3 12.1 Doses Resulting From Postulated Fuel-Handling Accident.. . 14-3 14.1 i
I, 1
l
, PSU SER ix.
i
1 INTRODUCTION The Pennsylvania State University (PSU/ licensee) submitted a timely applica-tion for a 20 year renewal of the Class 104c operating license R-2 (NRC l Docket No. 50-5) for its TRIGA research reactor facility to the U.S. Nuclear l
Regulatory Cop ission (NRC/ staff) by letter (with supporting documentation) dated March 1, 1985. The university currently is permitted to operate the PSBR (Pennsylvania State University Breazeale reactor) within the conditions autho-rized in past amendments in accordance with Title 10 of the Code of Federal R m -
lations (10 CFR), Paragraph 2.109, until NRC action on the renewaT request is completed.
The staff's review, with respect to issuing a renewal operating license to PSU, was based on the information contained in the renewal application and supporting supplements, plus responses to requests for additional information. The renewal application included Financial Statements, the Safety Analysis Report, an Environ-ment Report, and Technical Specifications. This material and the previously ap-proved Emergency Plan and Operator Requalification Plan are available for review at the Commission's Public Document Room at 1717 H Street N.W., Washington, D. C.
The approved Physical Security Plan is protected from public disclosure under 10 CFR 2.790(d)(1) and 10 CFR 9.5(a)(4).
The purpose of this Safety Evaluation Report (SER) is to summarize the results of the safety review of the PSBR and to delineate the scope of the technical details considered in evaluating the radiological safety aspects of continued operation.
This SER will serve as the basis for renewal of the license fur operation of the PSBR at thermal power levels up to and including 1 MW and pulsed operation with reactivity insertions up to 2.31% Ak/k. The facility was reviewed against the requirements of 10 CFR 20, 30, 50, 51, 55, 70, and 73; applicable regulatory guides; and appropriate accepted industry standards [American National Standards Institute /American Nuclear Society (ANSI /ANS 15 series)]. Because there are no specific accident-related regulations for research reactors, the staff has com-pared calculated dose values with related standards in 10 CFR 20, the standards for protection against radiation, both for employees and the public.
This SER was prepared by Robert E. Carter, Project Manager, Division of Pressurized-Water-Reactor Licensing-B, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission. Major contributors to the technical review include the Project Manager and John Dosa of NRC and C.A. Linder, C.L. Faust, and A.E. Sanchez-Pope of Los Alamos National Laboratory (LANL) under contract to
[ the NRC.
1.1 Summary and Conclusions of Principal Safety Considerations The staff's evaluation considered the information submitted by the licensee, past operating history recorded in annual reports suteited to the Commission by the licensee, inspection reports by Region I,.and c:,ite observations. In addition, as part of its licensing review of several TRIGA reactors, the staff l
obtained laboratory studies and analyses of several accidents postulated for the TRIGA reactor. The staff's conclusions, based on evaluation and resolution of the principal issues reviewed for the PSBR, are as follows:
PSU SER 1-1
(1) The design, testing, and performance of the reactor structure and the systems and components important to safety during normal operation were adequately planned, and safe operation can reasonably be expected to continue.
(2) The expected consequences of several postulated credible accidents have been considered, emphasizing those likely to cause a loss of integrity of fuel-element cladding. The staff performed conservative analyses of the most serious, hypothetically credible accidents and determined that the calculated potential radiation doses outside tne reactor site are not likely to exceed to guidelines of 10 CFR 20 for doses in unrestricted areas.
(3) 'The licensee's management organization, conduct of training and research a activities, and security measures are adequate to ensure safe operation of the-facility and protection of its special nuclear material.
(4) The systems provided for the control of radiological effluents can be operated to ensure that releases of radioactive wastes from the facility are wit'.in the limits of the Commission's regulations and are as low as is reasonably achievable (ALARA).
(5) The licensee's Technical Specifications, which provide limits controlling operation of the facility, are such that there is a high degree of assur-ance that the facility will be operated safely and reliably.
(6) The financial data provided by the licensee are such that the staff has determined that the licensee has reasonable access to sufficient revenues to cover operating costs and eventually to decommission the reactor facility.
(7) The liceniee's program for providing for the physical protection of the facility and its special nuclear material complies with the requirements of 10 CFR 73.
(8) The licensee's procedures for training its reactor operators and the plan for operator requalification are adequate; they give reasonable assurance that the PSBR will be operated competently.
(9) The licensee's approved Emergency Plan provides reasonable assurance that the licensee is prepared to assess and respond to potential emergency events.
1.2 History PSU began construction of a research reactor using plate-type (MTR) fuel in January 1954. In July 1955, the Atomic Energy Commission issued Operating License R-2 for the PSBR at power levels up to 100 kWt. Criticality was first reached on August 15, 1955. The authorized maximum operating power level was increased to 200 kWt in 1960. The current zirconium hydride fuel (TRIGA) was installed in 1965 to accommodate increased experimental demands. On December 30, 1965, the PSBR was licensed to operate at a maximum of 1 MWt, and in the pulse mode, with the TRIGA core and control systems.
PSU SER 1-2
1.3 R_eactor Description The PSBR is a heterogeneous, pool-type reactor. The core is cooled by natural convection of light water, moderated by zirconium hydride and light water, and reflected by graphite and light water. The core is suspended from a movable bridge that operates on rails mounted on top of the pool. The pool (withThe a total capacity of 71,000 gal) is 9.1 m long, 4.3 m wide and 7.3 m deep.
pool can be divided by a removable gate allowing independent drainage of either side of the pool.
The reactor core consists of stainless-steel-clad uranium zirconium hydride The uranium (U-ZrH fuel elements with a mixture of 8.5 and 12 wt % uranium. The cylindrical elements used is) enriched to less than 20% in the 23su isotope.
(with a diameter of 3.73 cm and an active length of 0.38 m) are arranged in a hexagonal lattice. Reactivity of the reactor core is changed by the operator by moving three fuel-followed control rods and an air-followed transient rod.
1.4 Shared Facilities and Equipment The reactor facility shares its utilities--electricity, water, heating steam, and nonradioactive sewage--with the laboratories and offices in the Breazeale Nuclear Reactor Building. The reactor bay has dedicated heating, air condition-ing, and air exhaust systems.
1.5 Comparison With Similar Facilities The reactor fuel rods are similar to those in most of the 52 TRIGA-type reactors in operation throughout the world, 25 of which are in the United States and 22 of these are licensed by the NRC. The instruments and controls are typical of TRIGA reactors and similar in principle to most of the nonpower reactors licensed by the NRC.
- 1. 6 Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test reactor, that the applicant shall have entered into an agreement with the Department of Energy (DOE) for the disposal of high-level radioactive wastes and spent nuclear fuel. DOE (R.L. Morgan) has informed the NRC (H. Denton) by letter dated May 3, 1983, that it has determined that univer-sities and other government agencies operating nonpower reactors have entered into contracts with DOE that provide that DOE retain title to the fuel and be obligated to take the spent fuel and/or high-level waste for storage or repro-cessing. Because PSU has entered into such a contract with DOE, the applicable requirements of the Waste Policy Act of 1982 have been satisfied for the PSBR.
PSU SER 1-3
2 SITE CHARACTERISTICS 2.1 Geography The PSBR is located on the eastern edge of the University Park Campus (see Figure 2.1) of PSU in central Pennsylvania in the County of Centre. The campus is bordered by commercial and residential areas of the Borough of State College to the east, west, and south. North of the campus are university-owned athletic fields and farms. The reactor site exclusion area is defined by the chain-link fence surrounding the PSBR facility as shown in Figure 2.2.
2.2 Demography The PSU campus has a student population of about 32,000 and State College Borough has a permanent population of about 36,000. Dormitories and residential areas are located approximately 300 m from the reactor building in all directions except the north. Classroom facilities are located within 60 m of the reactor site. The nearest population center (other than State College Borough) is 16 km to the north.
2.3 Nearby Industrial, Transportation, and Military Facilities There is no major highway o~ railway in the vicinity of the PSBR or the Borough of State College. The University Park Airport, which serves only small private or commercial aircraft, is more than 2.4 km from the reactor building.
There are no large industries or major military establishments in the State College area that will cause heavy use of local transportation systems.
Because there are no major transportation routes and no significant industrial or military facilities in the near vicinity of the reactor site that could cause accidental damage to the reactor, the staff concludes that such accidents need not be hypothesized or evaluated.
2.4 Meteorology The average annual rainfall recorded at the university weather station is 0.98 m with a maximum 24-hour rainfall of 0.12 m and a maximum monthly rainfall of 0.33 m. The average annual snowfall is 1.16 m with a maximum snow accumulation in one winter of 2.49 m.
Wind direction varies with the season, but on an annual basis the prevailing winds are generally westerly. The fastest wind speed recorded is 88 km per hour.
The State College area averages 40 thunderstorms per year with 25 occurring during the summer. These storms ara occasionally severe enough to cause prop-erty damage from hail, wind, lightning, and local flash flooding.
PSU SER 2-1
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PSU SER 2-3
Central Pennsylvania is an area of low tornado activity. During the past 30 years, six tornadoes have been recorded within a 40-km radius of the reactor building.
The closest tornado during this time was a Category II (117-179 km/hr wind speed) about 10 km southwest of the reactor building.
PSU is not in the normal hurricane path. Hazel, in 1954, was the most recent storm to be classified as a hurricane as.it. entered Pennsylvania. This storm was less severe in the University Park Campus area than normally occurring thunderstorms.
2.5 Hydrology-The PSBR site is in the' Spring Creek Drainage Basin, an area of about 450 km 2 (see Figure 2.3). The basin is underlined primarily by limestone.
The outlet for the drainage basin is at Milesburg Gap, about 16 km north of the PSBR site.
The PSBR site is not in an area of the basin with a past history of floods.
The hydrology of the area is typical of limestone ter.ains, with water entering -
the underground aquifer system by way of sinkholes, caves, fissures in rocks, and percolation through stream bottoms and soil covers. In the University Park Campus area, the water table is typically 68 m below ground level, well below the reactor building foundations.
The university water supply (including the PSBR facility) is' served by wells located approximately 3.2 km north of the reactor building. The State College Borough and many surrounding areas are served by well fields and a spring-fed mountain reservoir located several kilometers to the south of the PSBR site.
2.6 Geology and Seismology The PSBR is located in the northern Valley and Ridge Tectonic Province about-16 km east of its boundary with the Appalachian Plateaus Tectonic Prnvince.
The Valley and Ridge Tectonic Province is characterized by closely spaced, northeast southwest striking anticlines that are assymetrical to the west.
Faults are usually present on the overturned limbs of these anticlines. There is no evidence that these faults or other faults in the region have been active since the Mesozoic Era (more than 60 million years ago); therefore, they do not represent a hazard to the PSBR.
The physiography of the site region is characterized by broad valleys underlain by limestone and bounded by steep ridges composed predominantly of sandstone.
The site lies within one of these limestone valleys. Relief in the area ranges from less than 100 to more than 300 m.
Eighty-four seismic events have been reported in Pennsylvania from 1737 to 1974 (Stover et'al., 1981). Of these earthquakes, only two have been reported to have occurred within 50 km of the University Park Campus: . a maximum modified-Mercalli-intensity tively.
(MMI) III and VI on March 25, 1937, and July 15, 1938, respec-From October 1975 to September 1983, nine seismic events have been re-ported by the Northeastern U.S. Seismic Network (1984). Because of the low his-torical and instrumental seismicity of central Pennsylvania, the likelihood of ground motion frort an earthquake that would constitute a hazard to the facility during its operating life appears to be low.
PSU SER 2-4
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2.7 Conclusion On the basis of the above considerations for natural and man-made hazards, the staff concludes that there are no significant risks associated with the site that would make it unacceptable for the continued operation of the reactor.
PSU SER 2-6 l
3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS The licensee's Safety Analysis Report provides information on the design, con-struction, and functions of the reactor building, reactor systems, and auxiliary systems.
3.1 Wind and Water Damage The University Park campus area experiences few extreme wind conditions such as tornados or, inland hurricanes. Furthermore, the reactor building is constructed from concrete blocks and the reactor pool is formed of steel-reinforced poured concrete. The reactor site is well above the flood plain; therefore, wind or water damage to the PSBR facility is very unlikely.
3.2 Seismically Induced Reactor Damage The information on the past seismic activity and the likelihood of future earth-quakes in the area of University Park Campus indicates that the PSBR is in a region where there is low probability of severe seismic activity. In the event of an earthquake causing catastrophic damage to the reactor building anJ/or the reactor pool, water might be released. However, Section 14 of this SER shows that loss of coolant in the PSBR does not lead to core damage, and mechanical damage to fuel cladding would release only a small fraction of the fission product inventory.
3.3 Mechanical Systems and Components The mechanical systems of importance to safety are the neutron-absorbing control rods suspended from the superstructure. The motors, gear boxes, electromagnets, switches, and wiring are above the pool-water level and readily accessible for testing and maintenance. The staff has addressed the effects of aging on the continued performance of these components in Section 17 of this SER.
3.4 Conclusion Based on the above considerations, the staff concludes that the PSBR facility was designed and built to adequately withstand all credible and likely wind, water, and seismic damage associated with the site. These considerations indi-cate that natural events would lead to small reactor-related consequences to the environment. Furthermore, the design and performance of the safety systems have been verified by more than 20 years of operation. Accordingly, the staff con-cludes that the reactor systems and components are adequate to provide reasonable assurance that continued operation will not cause significant radiological risk to the health and safety of the public.
PSU SER 3-1
4 REACTOR The Pennsylvania State University Breazeale reactor (PSBR) is a 14M hexagonal grid core, Mark III TRIGA pool-type reactor incorporating solid uranium zirconium hydride fuel-moderatar elements enriched to less than 20% 2a30. The reactor core is submerged in a large, open pool of light water that acts as moderator, coolant, and radiation shield. The reactor power is controlled by inserting and withdraw-ing neutron-absorbing control rods. Pulse and square wave operation An of the re-overall view actor is initiated by the pneumatic ejection of a transient rod.
of the PSBR is given in Figure 4.1, and the principal design parameters are listed in Table 4.1. The PSBR is used principally for education, research, and training.
4.1 Reactor Core The current core configuration consists of a lattice of s95 cylindrical U-ZrH fuel elements, graphite elements, three fuel-followed control rods, and one air-followed transient control rod. These components are positioned by upper and lower aluminum grid plates. The active (or fueled) region of the reactor core forms a hexagonal array and contains 3.5 kg of 23sU. Water moderator / coolant occupies sl/2 the core volume.
4.1.1 Reflector Elements and Core Support Structure Radial neutron reflector graphite elements may occupy grid positions not filled by fuel-moderator elements or other core components. Top and bottom axial neutron reflection is provided by 8.74-cm-long graphite plugs incorporated into individual fuel elements.
The fuel elements are positioned laterally at the top and bottom by two grid plates, 1.59 cm and 7.38-cm thick, respectively. The lower grid plate supports the weight of the fuel elements.
The top grid plate is supported by spacer rods bolted to the top and bottom grid plates. The bottom grid plate is bolted to the vertical alumimum members of the suspension tower structure. In addition, a safety plate of 3.18-cm-thick aluminum is provided to preclude the possibility of control rods falling out of the core in the event of failure of their normal support systems. The safety plate is located s32 cm below the bottom grid plate.
4.1.2 Fuel Elements The reactor uses standard stainless-steel-clad cylindrical fuel elements in which the enriched uranium is mixed homogeneously with a ZrH moderator.
The fuel meat consists of a cylindrical rod of U-ZrH contaiNingeither s8.5 wt % or 12.0 wt % uranium enriched to (20% 23sV7 The nominal weight of 2asU in each 8.5 wt % and 12 wt % fuel element is 37 and 55 g, respectively.
The fuel section of each element is 38.1 cm long and 3.63 cm in diameter.
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Table 4.1 Principal design Parameter Description Reactor type TRIGA Mark III Maximum licensed power level 1 MW Maximum pulse 2.31% ak/k (3.305)
Fuel element design Fuel-moderator material U-ZrHy
- Uranium content 8.5 wt% and 12 wt%
<20% 2asy Uranium enrichment Shape Cylindrical Length of fuel 38.1 cm overall Diameter of fuel 3.63 cm outside diameter Cladding material 304 stainless steel Cladding thickness 0.051 cm Number of fuel elements Critical core s74 Operational core s95 Excess reactivity, maximum 4.9% ak/k (7.00$)
Number of control rods 4 Iransient (air-followed) 1 I
Shim (fuel-followed)
Safety (fuel-followed) l Regulating (fuel-followed) 1 Total reactivity worth of rods ** ~7.8% ak/k (11.16$)
Reactor cooling Natural convection of pool water p effective 0.007
- The nominal H/Zr ratio for the 8.5 wt% elements is 1.70. For the 12 wt% elements, the nominal H/Zr ratio is 1.65.
- For PSBR typical core configuration.
PSU SER 4-3
l serve as axial neutron reflectors. The fueled section and graphite end plugs are steelcontained fittings atinthe a 0.051 cm bottom.
top and thick stainless-steel-walled can welded to stainless-s3.2 kg. Each element is s72.1 cm long and weighs A schematic view of a TRIGA stainless steel-clad fuel element is shown in Figure 4.2.
Special instrumented fuel elements, otherwise similar to standard elements, contain three thermocouples embedded in the fuel region. The sensing tips of the thermocouples are located at the vertical centerline and $2.54 cm above and below it, respectively. The thermocouples monitor the fuel element temperatures and provide a scram signal on sensing a preselected value. Graphite elements may be used to fill peripheral grid positions not occupied by fuel-moderator ele-ments, control rods, or other core components. These elements are of the same general dimensions and construction as the standard fuel elements but contain only graphite and are clad with aluminum. A plan view of the PSBR core showing the positions of a typical core loading is shown in Figure 4.3.
4.1.3 Neutron Source and Holder An antimony beryllium photoneutron source is used for startup. The neutron source holder is made of aluminum, is cylindrical in shape, has a cavity to hold the source, and has outside dimensions similar to standard fuel elements so that the source holder can be installed in any fuel element location. A 0.235-mg 2s2Cf spontaneous fission neutron source in a platinum matrix, doubly encapsulated in Zircaloy-2, also is available. In addition, the licensee is considering PSBR.
the acquisition of an Am-Be neutron source for future use in the 4.1. 4 Control Rods Power levels in the PSBR are regulated by four control rods, as listed in Table 4.1. The neutron poison material in all the control rods is borated graphite.
4.2 Reactor Pool and Biological Shield The reactor pool is approximately 9.1 m long, 4.3 m wide, and 7.3 m deep.
The pool is constructed of steel reinforced cnncrete. The pool walls are 0.46 m thick below actor the level of the reactor bay floor and 0.31 m thick above the re-bay floor.
watertight pool liner. The inside of the pool wall is coated with epoxy to form a of the south side, The pool is surrounded by earth fill with the exception lhe Beam Hole Laboratory (BHL) is outside the south side of the pool at the elevation of the reactor core. Additional shielding is provided by 1.1 m of high-density concrete on the outside of the pool wall in this room.
The total pool volume is approximately 71,000 gal. The pool can be divided into two units by a removable gate that permits draining either part of the pool pool. while maintaining the reactor under water in the other part of the Seven beam ports penetrate the pool wall from the BHL to provide access to reactor radiation. The pool is equipped with two floor drains, one for each side of the pool when it is divided with the removable gate. Four other pool PSU SER 4-4 I
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wall penetrations exist: the two located approximately 3.1 m above the pool floor are the water recirculation loop inlet and outlet, and the two located approximately 5.2 m above the pool floor are for the heat exchanger. A cutaway showing the reactor pool and biological shield is shown in Figure 4.1.
4.3 Reactor Instrumentation The reactor instrumentation consists of a fission chamber, two compensated ion chambers, and a gamma ion chamber that provides indication from 10 3 W to >l MW.
Additionally, there are temperature sensors to measure fuel element and bulk coolant temperatures. A detailed description of the reactor instrumentation is provided in Section 7.
4.4 Dynamic Design Evaluation The operation of the PSBR is accomplished by manipulating control rods in response to observed changes in measured reactor parameters such as neutron flux (reactor power) and in temperature provided by the instrumentation channels. In addition, interlocks prevent inadvertent reactivity additions, and a scram system initiates rapid, automatic shutdown when safety settings are reached. Further stability and safety during normal operating and pulsing conditions are incorporated into TRIGA reactors by virtue of the large, prompt, negative temperature coefficient inherent in the U-ZrH fuel moderator material. The negative temperature coeffi-cient is primarily a result of the neutron spectral hardening properties of ZrH, at elevated temperatures, which increase the leakage of neutrons from the fuel-bearing material into the ambient water where they are absorbed preferentially.
The reactivity decrease is a prompt effect because the fuel and ZrH are mixed homogeneously; thus, the ZrH, temperature rises essentially simultaneously with reactor power. An additional contribution to the prompt, negative temperature coef ficient is the Doppler broadening of 2a8U neutron absorption resonances at high temperatures, which increases nonproductive neutron capture in these reso-nances (General Atomic (GA) 0471; GA-4314; Simnad, et al., 1976).
This inherent property of U-ZrH, fuel has been the basis for designing TRIGA reactors with pulsing capability. The large, prompt, negative temperature coef-ficient rapidly and automatically compensates for step insertions of excess reactivity. In the pulse mode, it terminates the resulting excursion without depending on electronic or mechanical safety systems or operator action. In the nonpulsing mode, the prompt, negative temperature coefficient serves as a backup safety feature, mitigating the effects of accidental reactivity inser-tions (GA-0471; Go-4314; Simnad, et al, ,1976).
4.4.1 Excess Reactivity and Shutdown Margin The Technical Specifications for the PSBR limit the maximum excess reactivity The Technical Spect-to 4.9% Ak/k (7.0$) in the cold xenon-free condition.
fications also require a minimum shutdown margin of 0.18% Ak/k (0.25$) with the highest worth control rod fully withdrawn, with all movable experiments in their most reactive state, and with the reactor in any operating condition.
The reactivity worth of any movable experiment in the reactor is limited by the ,
Technical Specifications to less than 1.4% Ak/k and that of any single experiment to less than 2.31% Ak/k. The control rod worths for a typical core configuration PSU SER 4-7
^ ________
are 1.92% ak/k for the transient rod, 1.53% ak/k for the shim rod, 2.85% ak/k for the safety rod, and 1.53% ak/k for the regulating rod, for a total rod worth of 7.81% ak/k. Under those conditions, the maximum excess reactivity that can be loaded into the core and still meet the minimum shutdown margin requirements is 4.78% ak/k.
Thus, in certain core configurations, it would not be possible to load the maxi-mum allowed excess reactivity without violating the minimum shutdown margin, which is the overriding requirement.
4.4.2 Normal Operating Conditions The Technical Specifications impose a fuel element safety setting limit of 700 C in an instrumented fuel element to provide assurance that the maximum fuel tem-perature will not reach the safety limit of 1150 C.
The safety limit for high-hydride (ZrH>l.6) stainless : steel-clad fuel elements is based on preventing excessive stress buildup in the cladding because of hydrogen pressure f rom disassociation of ZrH . Based on theoretical and experimental evi-dence (GA-4314; Simnad, et al., 1976), the above limit represents a conservative value to provide confidence that the fuel elements will maintain their integrity and that no cladding damage will occur. Further limitations are imposed on reactor power level and pulse reactivity insertion to provide assurance that the safety limit will not be exceeded. At the maximum licensed nonpulsing power level of 1 MW, the maximum allowed fuel temperature is 700 C. Scrams are pro-vided to shut the reactor down if the nonpulsing power level causes the measured fuel temperature in a B ring element (Figure 4.3) to exceed /00"C. During the maximum allowed 2.31% ak/k (3.305) pulse, fuel temperatures will not exceed 667 C.
On the basis of radial and axial power distributions, these requirements ensure that the above safety limits are not exceeded anywhere in the core.
4.4.3 Assessment The staff concludes that the inherent large, prompt, negative temperature coef-ficient of reactivity of U-IrH, fuel moderator provides a basis for safe opera-tion of the PSBR in the nonpulsing mode and is the essential characteristic supporting the capability of operation of the reactor in a pulse mode.
Furthermore, the Technical Specifications require that the core excess reactivity and experiment reactivity worths be limited so that the reactor always can be brought to a subcritical condition even if the highest worth control rod is fully withdrawn. The current core configuration meets all of these limitations.
The safety limits at the PSBR are based on theoretical and experimental investi-gitions and are consistent with those used at other similar reactors. Adherence to these limits provides confidence that fuel element integrity will be main-tained. Operating data at maximum licensed nonpulsing power and at maximum pulse reactivity insertion show that the maximum fuel element temperatures remain well below the prescribed safety limit. TRIGA reactors similar to the PSBR have demonstrated safe and reliable operation at nonpulsing power levels up to 1.5 MW and pulse reactivity insertions up to 3.5% ak/k (5.00$) (GA-4314; Simnad, et al., 19/6). On the basis of the above considerations, the staff concludes that, under normal operating conditions, there is reasonable assurance PSU SI.R 4-8
that the PSBR can be operated safely at 1 MW and with a pulse limit of 2.31% ak/k (3.30$) as prescribed by the Technical Specifications.
4.5 Functional Design of Reactivity Control System The power level in the PSBR is controlled by three standard control rods (one shim, one safety, and one regulating rod) and one transient rod. The three standard control rods and the transient rod contain borated graphite as the neutron poison. The locations of the four control rods are shown in Figure 4.3.
Rod movement is accomplished using rack-and pinion electromechanical drives for each standard control rod and a pneumatic electromechanical drive for the tran-sient rod. Each control rod drive system, through its own independent electrical cables and circuits, is energized from the control console; this independent circuitry tends to decrease the probability of multiple malfunctions of the drives resulting from a single cause. When a scram signal is received, all four control rods fall by gravity into the core, thereby shutting dcwn the reactor.
4.5.1 Standard Rod Drive Assembly The control rod drive assemblies are mounted on a bridge assembly over the pool.
Each assembly consists of a reversible single phase electric motor coupled to a rack-and pinion drive system. A draw tube, connected to the rack, supports an electromagnet that, in turn, engages an iron armature attached to the upper end of a long connecting rod. The neutron absorbing section of the control rod is attached to the lower end of the connecting rod. During normal operation, the electromagnet is energized and the motorized system inserts or withdraws the shim and safety control rods at a rate of s0.04 cm/s, corresponding to reactivity insertion rates of s0.05% ak/k/s (0.07$/s) and s0.09% ak/k/s (0.135/s), respec-tively. The regulating rod is moved at a maximum rate of s1.0 cm/s, correspond-ing to a reactivity insertion rate of s0.06% ak/k/s (0.08$/s). The transient rod, when used as a control rod during nonpulsing conditions, has a reactivity insertion rate of s0.09% ak/k/s (0.13$/s). If power to the electromagnets is interrupted for any reason, the connecting rods are released, and the control rods fall by gravity into the core, rapidly shutting the reactor down (scramming).
Limit switches mounted on the drive assemblies actuate circuits that indicate on the control console the up (fully withdrawn) and down (fully inserted) position of the magnet, the fully inserted position of the rod, and whether the magnet is in contact with the armature. In addition, helipots connected to the pinion gear generate position indications for the shim, safety, and regulating rods that are displayed on the control console.
4.5.2 Transient Rod Drive Assembly The transient rod drive is mounted on a frame bolted to the bridge and is operated by a pneumatic drive system consisting of a single-acting pneu-matic cylinder whose piston is attached to the transient rod by a connecting rod. For pulse operation, compressed air is admitted to the bottom of the cylinder through a solenoid valve, driving the piston upward in the cylinder and driving its connected transient rod out of the core. At the end of its stroke, the piston strikes the anvil of a shock absorber and decelerates at a controlled rate. Adjustments of the cylinder position, in relation to the piston head, control the piston's stroke length and hence the extent of tran-sient rod withdrawal from the core and the corresponding amount of reactivity PSU SER 4-9
l inserted for a pulse. The adjustment is performed electrically at the rod drive housing. When the solenoid valve is deenergized, the air is vented from the cylinder and the transient rod drops by gravity into the core.
Limit switches mounted on the drive assembly actuate circuits that indicate on the control console the up (fully withdrawn) and down (fully inserted) position of the cylinder and the down position of the rod. In addition, a helipot con-nected to the motor shaf t generates transient rod cylinder position indication and is displayed on the control console.
In the nonpulsing mode, an interlock prevents the application of air to the transient rod unless the transient cylinder is inserted fully. Additionally, with air pressure continuously applied, the transient rod cylinder can be moved so that the rod may be used as an ordinary control rod.
4.5.3 Scram-logic Circuitry and Interlocks The PSBR is equipped with a scram-logic safcty system that receives signals from core and other system instrumentation to initiate a scram by interrupting the electrical power to the control rod magnets and the transient rod solenoid air valve.
The reactor parameters that initiate these scrams are (1) high reactor power (2) high fuel temperature (3) voltage failure on the detectors (4) manual scram (5) preset timer on the transient rod (6) high radiation level 4 (7) building evacuation alarms (8) external manual scrams Several safety interlocks are incorporated into the control rod circuitry to prevent inadvertent reactivity insertions. During nonpulsing operation, inter-locks prevent the simultaneous withdrawal of two standard control rods, and another interlock prevents the application of air to the transient rod unless the transient rod cylinder is inserted fully. In addition, an interlock prevents any control rod withdrawal unless an adequate neutron source signal is detected.
In the pulse mode, only the transient rod can be withdrawn.
Additional details concerning the safety-logic circuitry and interlocks are provided in Section 7.
4.5.4 Assessment The PSBR reactor safety and control system design uses proven state-of-the art components and technology. The control rods, rod drives, and scram and inter-lock logic have performed reliably and satisfactorily in the PSBR, and similar equipment has shown satisfactory performance in many other TRIGA reactors over long periods of time.
The control system design all w for an orderly approach to criticality and for safe shutdown during both normal and emergency conditions. There is sufficient redundancy of control rods to ensure safe shutdown even if the most PSU SER 4-10
reactive rod f ails to insert when receiving a scram signal. The reactivity worths and speed of travel of the control rods are adequate to allow complete control of the reactor system during operation f rom a shutdown condition to full power. Interlocks are provided to preclude inadvertent rod movement that might lead to hazardous conditions. Independent scram sensors and circuits are incorporated to shut the reactor down automatically and mitigate the consequences of single malfunctions. A manual scram button allows the operator to initiate a scram whenever such action is needed.
In addition to the active electromechanical safety controls for normal and abnor-mal conditions, the large, prompt, negative temperature coefficient of reactivity inherent in the U-ZrH fuel-moderator (discussed in Section 4.5) provides a unique backup safety feature. The reactor shutdown mechanism of this fuel mix-ture limits the power level and terminates reactor transients that produce large increases in temperature. Because this inherent shutdown mechanism acts to limit the magnitude of a possible transient accident, it will mitigate the consequences of such accidents and can be considered to be a fail-safe safety feature.
In accordance with the above discussion, the staff concludes that the reactivity control systems of the PSBR are well designed and will function adequately to ensure safe operation and safe shutdown of the reactor under all credible conditions.
4.6 Operational Procedures PSU has implemented administrative controls that require review, audit, and written procedures for all reactor safety-related activities. The Penn State Reactor Safeguards Committee reviews all aspects of current reactor operation to ensure that the reactor facility is operated and used within the terms of the facility license considering the health and safety needs of the public as well as of the operating personnel. The Committee also reviews operating pro-cedures, experiments, and proposed changes to the f acility or its Technical Specifications.
Written procedures reviewed by the Penn State Reactor Safeguards Committee are established for safety-related activities, including reactor startup, operation, and shutdown; preventive or corrective maintenance; and periodic inspection, testing, and calibration of reactor equipment and instrumentation.
The reactor is operated by trained, NRC-licensed personnel in accordance with the above-mentioned procedures.
4.7 Conclusion The staff review of the PSBR has included studying its design and instal-lation and control and safety instrumentation. As noted earlier, these features are similar to those typical of TRIGA-type research reactors operating in many countries of the world, more than 20 of which are licensed by the NRC.
There are currently 11 TRIGA reactors operating at 1 MW or greater with no safety-related problems. On the basis of its review of the PSBR and emperience with these other facilities, the staff concludes that there is reasonable assurance that the PSBR is capable of safe operation as limited by its Technical Specifications.
PSU SER 4-11 l
5 REACTOR COOLANT AND ASSOCIATED SYSTEMS l
The PSBR core is cooled by the natural convection of deionized pool water. The i
primary reactor coolant is pumped to a heat exchanger, where the reactor heat is removed before it is returned to the pool. The quality of the reactor cool-ant is maintained by a purification system. A schematic of the reactor coolant system is given in Figure 5.1, and the coolant system instrumentation is de-scribed in Section 7.
5.1 Cooling System The primary cooling system suction and return lines both penetrate the pool walls 5.2 m above the pool floor. The primary cooling system pump takes water from the pool, forces it through the shell side of two heat exchangers in series, and sends it back to the pool. Secondary cooling system water is pumped from a spring-fed pond 200 m from the facility, through the tube sides of the two heat exchangers, and sent back to the pond through the storm sewer system. This sys-tem is maintained at a higher oressure than the primary cooling system to ensure that any heat exchanger leakage will be contained in the primary loop of the reac-tor facility. If this differential pressure falls below 1.5 psi, an alarm sounds at the control console.
5.2 Primary Coolant Purification System Approximately 40 gal / min of primary coolant are drawn through a skimmer located at the south end of the pool, a recirculating pump, a filter, and a mixed-bed deionizer and then are sent back to the pool. The conductivity of the pool water, as measured at the demineralizer, is maintained below 5 pmhos/cm. The purifica-tion system is shown schematically in figure 5.1.
5.3 Primary Coolant Makeup System Makeup to replace pool water lost to evaporation ordinarily is supplied by waste evaporator distillate or the university water supply. Flows from both these sources are supplied to the pool through the purification loop. If neces-sary, water can be added rapidly directly to the pool using fire hoses from the secondary cooling system or the university water supply.
5.4 Conclusion The staff concludes that the PSBR cooling system is of proper size, design, condition, and maintenance level to ensure adequate cooling of the reactor under routine operating conditions specified in the PSBR operating license. The cool-ing and water purification systems at the PSBR facility have the same design features as used in many comparable operating nonpower reactor facilities. There is no new or unproven technology involved in the system.
On the basis of the above considerations the staff concludes that the cooling and purification system at the PSPR is adequate for continued safe operation.
PSU SER 5-1
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6 ENGINEERED SAFETY FEATURES f
The only system designed to mitigate the consequences of a radiological accident
! at the PSBR facilities is the ventilation system.
6.1 Ventilation System i
One of two separate exhaust systems maintains the reactor bay at a negative pres-sure with respect to the atmosphere and the remainder of the building. Incoming
! l air to the room is supplied by leaks around penetrations such as doors. Normal ventilation of the reactor bay is accomplished by one of two roof fans that exhausts air to the atmosphere at roof level. This system is secured automati-l t cally when a building evacuation alarm is sounded. The second system, an emer-I gency exhaust system, is started automatically when the building evacuation alarm is sounded. This emergency system exhausts the reactor bay air through
,; roughing filters, absolute filters, and charcoal filters (all in series) before discharging to the atmosphere 0.9 m above the reactnr bay roof.
1 j A control / status panel for the emergency exhaust system is located in the recep-j tion area for the 60Co facility attached to the north side of the reactor build- ,
ing. This instrument panel is located so that the instruments on it could, in l
an emergency, be read from outside the building. Instruments on the panel con-i J
sist of four differential pressure gauges, three of which indicate pressure I drops across the prefilter, the absolute filter, and the charcoal filter, respec-tively. The fourth differential pressure gauge indicates the air flow in the
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j stack. Two pilot lights also are located on the panel: one indicates that the
) system is energized and the other indicates that there is airflow in the system.
The second light receives a signal from a flow switch in the duct. There is also a switch on this panel that permits manual activation of the emergency system.
l l 6.2 Conclusion 1-On the basis of the above and the considerations in Section 14 of this SER, the staff concludes that the reactor room ventilation system and equipment are
- adequate to control the release of airborne radioactive effluents in compliance J
with regulations and to limit releases of airborne radioactivity in the event j of abnormal or accident conditions.
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PSU SER 6-1 i
7 CONTROL AND INSTRUMENTATION SYSTEMS The PSBR facility control and instrumentation system is designed to provide the reactor with safe, reliable operation. The control console, located in the con-trol room, ditiplays the reactor parameters, including power level, fuel element temperature, pool water temperature, and control rod positions. The reactor con-trol console is shown in Figure 7.1. A schematic of the control and instrumenta-tion system is shown in Figure 7.2.
7.1 Reactor Control System The control system consists of those components that control the operation of the reactor control rods as well as associated equipment appropriate to the reactor operating mode selected. The reactor control rods and rod drive mech-anism are described in Section 4.
There are five modes of operation associated with the PSBR: manual, automatic, square wave, pulse high, and pulse low.
Manual Mode: Regardless of which mode is selected subsequently, the reactor always is started in the manual mode. The rods are controlled manually by the switches on the rod control panel.
Several interlocks limit the movement of the control rods in the up direction.
These are (1) scrams not reset (2) magnet not coupled to armature (3) neutron source level below minimum count rate (4) two UP switches depressed at the same time (5) mode switch in the pulse position (the mode switch is placed in " pulse" af ter criticality and a predetermined power level have been reached)
(6) mode switch in AUTOMATIC position Automatic Mode: Automatic power control can be obtained by switching from manual operation to automatic operation. The regulating rod then is controlled automatically in response to a power level and period signal. Reactor power level is compared with the demand level set by the operator and the signal dif ference is used to bring the reactor power to the demand level on a flxed preset period.
The purpose of this feature is to automatically maintain the preset power level during long-term power runs. The shim rod is moved automatically only if the regulating rod approaches the upper or lower 2bT, of its maximum travel length.
Square-Wave Mode: In square wave operation, the reactor is taken manually to critTcal (below 1 kW), leaving the transient rod at its Im,er limit. The tran-sient rod is ejected from the core to a preset height by means of the translent rod flRE button. When the desired power level reaches the demand level, the automatic control system maintains the power level in the same way as in the automatic mode.
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Figure 7.2 Schematic of the PSBR control console instrumentation 4
i I
i Pulse-High and Pulse-Low Modes: The pulse-low mode is used for pulses with l peaks up to 400 MW, and the pulse-high mode is used for pulses up to 2000 MW.
Reactor, control in each pulsing mode consists of establishing a constant power i below 1 kW. This is accomplished using the motor-driven control rods, leaving the transient rod at its lower limit. The transient rod then is ejected to a preset height using the FIRE button. The MODE SELECTOR switch automatically connects the gamma pulsing chamber to monitor and record peak power in the i
appropriate range.
J 7.2 Instrumentation System
)
The instrumentation system is composed of nuclear and process instrumentation that provides the operator with the information necessary for the operation of the facility.
1 7.2.1 Nuclear Instrumentation i
The nuclear instrumentation uses two independent systems to process the input signals from the various nuclear detectors.
' Four detectors provide wide-range (from 10 3 W to 2 x 106 W) indication from i source range up to >100% power, with appropriate overlaps among the channels.
1
' The startup channel consists of a fission counter that is withdrawn automatically when the reactor power level reaches 1 kW. The outputs (three) from the fission chamber go to the log count rate meter, an interlock bistable, and the second pen of a dual pen recorder, where the count rate is displayed on j a log scale.
A compensated ion chamber (CIC) provides the signal to the log power amplifier.
The first output signal from the log amplifier goes to-the second pen of the log count-rate recorder that indicates power from 0.3 W to 2 x 106 W. The second output goes to the 1-kW interlock bistable that causes the movement of j the fission bypass chamber in or out of the core and also to the control rod interlock function.
' The third primary output provides the input to the period channel, where four secondary outputs are provided. Two outputs go to a period
, meter (range of -30 to +3 s) and a decade / minute meter. The third output goes 1
to a bistable that provides a scram signal if the period decreases below +3 s. !
j The last output provides a period signal to the automatic controller (servo system). l A second CIC provides a signal to the linear power channel. A 15 position reactor power switch provides range selection from 0.1 W to 1 MW. The three outputs f rom the linear amplifier go to (1) the blue pen of the dual pen recorder, providing linear power indication from 0 to 100% on any range; (2) a bistable that provides a scram signal if the power level exceeds 110% of any range; and (3) the linear power level input to the automatic controller.
1 i
A gamma ionization chamber (GIC) that feeds into a linear amplifier and a per- )
i cent power unit indicates power from 0 to 110% of 1 MW rated power and provides '
a scram function at 110% power.
The reactor scram system provides a protective action (scram signal) that j interrupts the magnnt current to the transient rod air solenoid valve and to J
l i PSU SER 7-4
-,m- ,,- , _ _ . . _ . . - , , - - . - -
_ - -,- - - - - - - - - - - - - - -m- - - - - - - - - - - - - - - - .-v, ,-,.,w -
the holding magnets of all other control rods. The minimum required safety system channels and their set points are listed in Table 7.1.
~.
t J Table 7.1 Minimum reactor safety system channels Safety channel Function Set point Manual scram Scram Manual Fuel temperature Scram 700 C Linear power level Scram 110% of full scale Percent power level Scram 110% of full scale l Peak pulse power Scram 110% of full scale High voltage Scram Loss
! Magnet current Scram Loss i Minimum period Scram Available as desired (<3 s) j External safety switches Scram As required
! In the event of loss of electrical power, the deenergized magnets release all four control rods for insertion by gravity.
i 7.2.2 Nonnuclear Instrumentation The nonnuclear process instrumentation measures reactor fuel and bulk pool tem-peratures and displays them for operator information. The output of chrome 1/
l alumel thermocouples embedded in reactor fuel (described in Section 4) is pro-cessed and displayed continuously on a meter with a range of 0*C to 700*C on
- the control console and also provides for a high fuel temperature scram to j avoid reaching the temperature safety limit.
The reactor pool water temperature indication that is displayed continuously on a control console meter receives its signal from a' resistance bulb thermometer suspended in the pool from the reactor bridge. The bulk pool temperature is measured over a range of 0 C to 60 C. A second resistence bulb thermometer, also suspended in the pool, monitors the bulk pool temperature continually and provides for an alarm at a preset high pool water temperature. Table 7.2 shows the console alarm settings for the PSBR coolant systems.
Table 7.2 Console alarm settings Alarm Setting Pool water level >6.1 m above bottom of core AP between primary and secondary [1.5 psi coolant systems Primary coolant flow 0-400 gal / min Pool water temperature <43.3*C In addition to the instrumentation displays in the control room, there are several flow, temperature, and pressure gauges in the coolant treatment area for local readout only.
PSU SER 7-5
_. , _ _ _ _ ~ _ . . _ . _ _ _ _ _ _ _ ___
7.3 Conclusion The control and instrumentation systems at the PSBR facility, which are similar to those in other NRC-licensed nonpower reactors, are designed to provide reli-ability and flexibility. All power and instrumentation wiring is protected from physical damage by conduits and is well identified. There is redundancy and diversity in the important nuclear and temperature monitoring circuits. 'I n particular, nuclear power measurements are overlapped in the ranges of the log-N, linear power, and percent power level channels.
On the basis of the above considerations and the formal administrative cont.rols required by the Technical Specifications, the staff concludes that the control ;
and auxiliary instrumentation systems at the PSBR comply with the requirements '
and performance objectives of the Technical Specifications and are acceptable I and provide reasonable assurance of the continued safe operation of the reactor.
l I
1 PSU SER 7-6 l
. - ._ _ - . = _ - . _
8 ELECTRIC POWER SYSTEM The electrical power system at the PSBR facility is a standard electrical supply system designed and constructed to specifications similar to those at other research reactor facilities.
8.1 Normal Power i
i Electric power is supplied to the facility through a dedicated three phase transformer located inside the reactor site boundary fence. The power is supplied by the West Penn Power Company.
8.2 Emergencv Power Emergency electrical power is supplied by an uninterruptible power supply system maintained in the reactor bay. A battery bank in this system is maintained in a charged state by normal building power. The battery bank in turn supplies
~
power to the following facility devices so that each device continues to operate normally in the event of a power failure.
(1) control room annunciator panel~
(2) reactor bridge east and west radiation monitors (3) reactor bay east and west air monitors (except for pumps) i (4) 80Co bay area monitor (5) beam hole laboratory area monitor (6) several evacuation alarms (7) building. intrusion alarm system In the event of a power failure, emergency lighting is provided in 12 locations throughout the PSBR building by. individual battery packs.
8.3 Conclusion The staff concludes that the electrical power system at the PSBR, as a standard electrical supply system typical of research reactor facilities, is adequate for continued operation of the reactor. The staff also concludes that emer-gency power, in addition to-that currently available, is unnecessary.
PSU SER 8-1
9 AUXILIARY SYSTEMS 9.1 Ventilation System The ventilation system is considered an engineered safety feature and is described in Section 6.
9.2 Fire Protection System The reactor building is equipped with an internal (local alarm only) fire alarm system. Fire extinguishers of either the CO2 or compressed air and water type are located at strategic locations throughout the building. The interior of each of the two hot cells is equipped with automatic water sprinklers. When activated, the hot cell sprinklers send a signal to police service and sound an alarm bell on-the hot cell loading dock. Firefighting protection for all university buildings, including the reactor building, is provided by the State College Alpha Fire Company. A fire hydrant is located just outside of the reactor site boundary fence, approximately 84 m from'the
' building.
9.3 Communication System Three internal communication systems and a commercial telephone are available to the control room operator. The internal communication systems allow (1) two-way conversation with anyone on the reactor bridge and with the experi-menters using the pneumatic transfer systems at any of the sending / receiving stations, (2) two-way conversation between 24 offices or laboratories, and (3) the use of a page system that has speakers servicing all parts of the building.
9.4 Compressed Air System Compressed air is supplied by two air compressors. A 1 -horsepower compressor is dedicatted .to supply compressed air to the reactor transient rod drive. A 20-horsepower compressor supplies compressed air for general use throughout the building.
These compressors are located in an equipment room below the loading dock.
- 9. 5 Air Conditioning System The air in the reactor bay and control room is heated and cooled by a dedicated reactor bay air conditioner. This unit recirculates, heats, cools or dehumidi-fies reactor bay air, as required. No air is interchanged with any other part of the building or outside of the building by this unit.
9.6 Fuel Handling and Storage Irradiated and new fuel elements are stored in the pool in wall-mounted storage racks. Fuel elements are rearranged in the core or moved to or from storage PSU SER 9-1
racks manually using long-handled tools.
~
The Technical Specifications prescribe acceptable conditions of fuel storage.
9.7 Conclusion The staff concludes that the auxiliary systems at the PSBR facility are designed and maintained appropriately and are adequate for their intended purposes.
PSU SER 9-2
1 10 EXPERIMENTAL PROGRAM The PSBR is used to help educate students in nuclear science and engineering and to instruct reactor operators. The reactor also serves as a source of ionizing radiation for research and for radionuclide production, ,
10.1 Experimental Facilities Facilities available to experimenters include neutron beam ports, a 0 20 thermal column, a central thimble, dry vertical irradiation tubes, and two pneumatic transfer systems. The locations of some of these facilities are shown in ,.
Figure 10.1 and are described below.
10.1.1 Beam Ports Seven beam ports penetrate the south end of the concrete shield. The pool side l
- of the beam hole is sealed with a gasket and a blank flange. The other end of j the beam port terminates in the Beam Hole Laboratory where shielding for each of the ports is provided by lead-filled stainless-steel and concrete-filled aluminum plugs.
The beam hole inside diameter measurements are shown in Table 10.1.' The pool wall around beam port No. 4 is protected from gamma radiation by a 61 cm by 61 cm by 2.54 cm lead shield.
1 I Table 10.1 Beam port measurements Beam port Diameter >
f No. 2, 4, 6 17.8 cm 10.2 cm
- No. 1, 7 J No. 3, 5 7.6 cm
In addition to the Geiger-Mueller area monitor described in Section 12, the BHL is equipped with a television camera to scan the laboratory with a receiver in j the reactor control room and _with a light beam alarm with local and control room
- annunciation. The television camera and the light beam relay are used whenever
- a beam port is.open.
i 10.1.2 Thermal Column The thermal column is an aluminum tank measuring 0.86 m in diameter and 0.69 m
. long. Although the thermal column is movable, it normally is associated'with beam port No. 4. Built into one side of the thermal column is a 0.23 m-long, t
' O.25 m-diameter air-filled flux trap.
4 4
PSU SER 10-1
S.S. Lead Filled
- 4 j/ Rolling Nors i i s o* O i i p i ,i s
c '
S 0 sN '
Lead ef #g N\ g g f, s ,
,p/ Shield e #
@ s N NN -
u -
'e f p sj
\ \ '
N g DO s //
Boral/ 2 f g Shields- Tank ,
N --
/ j "f)E' Core d
%_ Central Thimble Additional Shielding e ca Beam Hole Experiment Tubes Laboratory 7-.-Q
, ,u Trolley
[ ,' Tower \. /
y Instrument Bridge 8
3
- Carriage l
/
W l h NORTH Figure 10.1 Locations of PSBR facilities for experimenters PSU SER 10-2 l
l_ _ - - - - -..-- --- - - - - _
- _ I
i t
i, Mechanical obstructions prohibit the thermal column from contacting the beam
! port; therefore, an air-filled extension may be bolted to the beam port flange to remove, water from the neutron beam path.
4 Reactor instrumentation and the reactor pool wall are protected from neutron and gamma radiation by boral shields placed on the thermal column. The thermal column is filled with D 02 moderator. However, because the tank is sealed, release of tritium is controlled.
10.1.3 Central Thimble i
I A central thimble allows samples to be irradiated within the core.in the region of maximum neutron flux. The thimble is an aluminum tube that extends from the j reactor bridge, through the bottom grid plate, 'and- to the safety plate. Because there are holes near the bottom of the thimble, it is filled with water, so streaming of gamma rays and neutrons into the reactor room does not occur.
10.1.4 Vertical Tubes l Air-filled aluminum tubes are available for dry irradiations. These tubes
[ extend from the reactor core level to just above the reactor bridge floor.
The tubes are weighted, and shielding plugs can be installed by a jib crane, J which is on the reactor bridge. j h 10.1.5 Pneumatic Transfer Systems j
i Two pneumatic transfer systems are available. Pneumatic transfer system I l
(PTSI) enables samples to be transferred from the reactor core to the labora-
- tory wing of the facility. Pneumatic transfer system II (PTSII) enables the
! transfer of samples between a hood in the reactor bay and the reactor core or j the thermal column.
I Pneumatic Transfer System I PTSI (laboratory to core) uses C02 as a carrier gas, thus minimizing the production of activated contaminants, such as 4!Ar, from the activation of air. The system also is designed to minimize the leakage of air into the system.
Components in this system that are likely to leak are enclosed in a gas-tight l
container that is vented through a system filter. Overpress'urizations are vented through an absolute filter and a charcoal filter.
PTSI use is controlled from the reactor console with a communication link to the experimenter in the laboratory. The experimenter may choose one of two core termini: a bare aluminum tube or an aluminum tube lined with cadmium.
Pneumatic Transfer System II PTSII (reactor bay hood to core or D2 0 tank) uses nitrogen as the working gas. Components in this system that could release activated gases are located
~
in hoods that ventilate through an absolute filter. .
PTSII use also is controlled from the reactor console with communication links to the experimenter. The experimenter may choose one of three termini: a PSU SER 10-3
1 I
stainless-steel tube in the reactor core, a cadmium-lined aluminum tube in the reactor. core, or an aluminum tube in the thermal column.
10.2 Experimental Review f
The Penn' State Reactor Safeguards Committee-(PSRSC) performs a safety evalu-ation of all in-core nuclear reactor experiments. The PSRSC is composed of_7 j to 10 members who are knowledgeable in subjects.related to reactor operations.
- At least one member of the PSRSC has health physics expertise. Written proce-l i
dures are in place for experiment evaluation and authorization. Experiment re-views are based on Federal regulations, license requirements, internal operating
-procedures, and safety considerations. ALARA considerations are included in experiment planning. Changes in experiments that do not alter the original j intent can be approved by the facility director.
i 10.3 Conclusions
, The staff concludes that the design of PSBR experimental facilities, combined with administrative and ' procedural controls and technical specification limits on experiments, ensures a safe experimental program. Therefore, the staff concludes that reasonable provisions have been made so that experimental programs do not pose a significant risk of radiation exposure to the PSBR staff, the students, or the public.
5 I
i l l 1 l i
1 i
i i
i 4
PSU SER .10-4 l
11 RADI0 ACTIVE WASTE MANAGEMENT ,
The major radioactive waste generated by reactor operation is activated gases, principally 41Ar and tritium. Small volumes of liquid and' solid radioactive j waste also are generated, primarily in connection with the experimental uses of the reactor, j i
11.1 ALARA Commitment ;
i l
The PSBR is operated with the philosophy of limiting the release of radioactive materials to the environment, consistent with the ALARA (as low as reasonably
~
, achievable) principle, which has recently been reaffirmed by the university l administration.
I 11.2 Waste Generation and Handling Procedures i 11.2.1 Solid Waste l The disposal of spent fuel is expected to occur infrequently, if at all, during i the term of this license renewal. . The largest volume of solid radioactive waste consists of slightly contaminated paper and plastic material. Most of the activ-
]
ity in the solid radioactive waste is found in activated samples, components,.
and equipment. Solidified evaporator sludge also is treated as solid radioactive- ,
waste (see Section 11.2.2). The solid waste is collected by the university's I health physics staff in specially marked barrels. The waste is held temporarily
! before being packaged and shipped to an approved disposal site in accordance r 1 with applicable regulations.
{ 11.2.2 Liquid Waste
? The largest volume of contaminated water is produced by the regeneration of the
! demineralizer. This liquid, plus waste liquid from some floor drains and pump seals, is collected in holdup tanks. This waste subsequently is evaporatedt The sludge is solidified and shipped as solid radioactive waste, and the dis-tillate is used as makeup water for.the reactor pool.
l
! 11.2.3 Airborne Waste
- l Airborne discharges from the reactor facility include small amounts of Tritium, 41Ar from the. neutron activation of air dissolved in the cooling water, and 16N from-activation of. oxygen in the cool.ing water.
i
- Average annual releases of 41Ar are estimated to be 600 mci. A dilution-factor l of 100 m3/s gives an average annual concentration of 2 x 10 10 pCi/ml at ground l
1evel outside the building, which is 0.5% of the 10 CFR 20 maximum permissible concentration (MPC) in unrestricted areas. Exposures of personnel in the bay are limited by a ventilation turnover rate half-time of 31 min.
~
Average annual releases of tritium are estimated to be 7 mci which results.in
! an average annual ground level concentration of 2 x 10 12 pCi/ml, which is 10 3%
i i PSU SER 11-1 1
..-p, _--e-- re--.,.w-, , ~, ,<- ,
of the MPC for unrestricted areas. The concentration of tritium in the reactor room air is reduced by dehumidification of the air.
Nitrogen-16 is produced in sufficient quantity to result in measurable exposure rates (~15 mR/h) in the reactor room at the pool wall, but because of the short half-life (7.1 s), the radioactivity released to the environment is not significant.
11.3 Conclusions The staf f concludes that- the waste management activities at the PSBR f acility have been conducted and are expected to be conducted in a manner consistent with 10 CFR 20 and the ALARA principles. Among other guidance, the staff review has followed the methods of ANSI /ANS 15.11, " Radiological Control at Research Reactor Facilities."
l Because 4 tar is the only significant airborne radionuclide released by the reac-tor to the environment during normal operations, the staff has reviewed the history, current practices, and future expectations of reactor operations with respect to this radionuclide. It is concluded that the doses in unrestricted areas as a result of actual releases of 41 Ar have never exceeded or even l
approached the limits specified in 10 CFR 20 when averaged over a year.
l 1
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l PSU SER 11-2
12 RADIATION PROTECTION PROGRAM PSU has a structured radiation safety program with a health physics staff to implement the program. The reactor facility has the equipment to detect, mea-sure, and control area and personnel radiation exposures. Use of radioactive material and radiation sources is controlled carefully, and releases of radio-active material to the environment are kept to a minimum.
12.1 ALARA Commitment PSU has adopted a policy that environmental releases of radioactive material and exposure of individuals to ionizing radiation be kept as far below regula-tory limits as is reasonably achievable. A training program has been developed, and equipment and procedures are designed to implement this policy. There are <
- personnel and environmental monitoring programs to ensure that radiation expo-sures are kept ALARA.
l
- 12.2 Health Physics Program t
12.2.1 Health Physics Staffing i.
! The PSU Health Physics Office technical staff consists of the university's j health physicist and three health physics assistants. The normal radiation
! safety staff at the reactor facility consists of a health physics assistant who j devotes about half of his time to reactor health physics duties. The health I physicist provides oversight to the radiation protection program at the reactor t facility and participates in experiment review and approval through his position
! on the Reactor Safeguards. Committee. Reactor operations staff and student assistants perform many health physics-type activities, with assistance and f consultation from the Health Physics Office.
12.2.2 Procedures I Rules and procedures for the use of radioactive materials at PSU were written l b, the University Isotopes Committee (UIC) in 1980 and were revised last in January 1985. These procedures specify training requirements, administrative policy, and health physics procedures for the control of radiation around the -
facility. Requests for authorization to use radioactive materials are reviewed
( by the health physicist and the VIC. Experimenters-must have UIC authorization before activated samples will be released to them. (When an activated sample j leaves the reactor pool, it comes under UIC jurisdiction.)
i 12.2.3 Instrumentation l PSU has acquired a variety-of detecting and measuring instruments for monitoring l potentially hazardous ionizing radiation. The instrument calibration procedures l
and techniques ensure that any credible type of radiation and any significant l
intensities will be detected promptly and measured correctly. ,
i PSU SER 12-1
I 12.2.4 Training All students, staf f, faculty, and independent users of the PSBR facility are l required to participate in a health physics orientation program. The orienta-tion lecture includes information on university and Federal regulations for radioactive material use as well as instruction . radiation protection principles and practices.
1 Reactor operator training and requalification include lectures on radiation control and safety, )
12.3 Radiation Sources 12.3.1 Reactor Sources of radiation directly related to reactor operations include the reactor core, the ion exchange columns, the cooling water cleanup systems, and radio-active gases. ,
The fission products are contained within the fuel's stainless-steel cladding.
Radiation exposures from the reactor core.are reduced to acceptable levels by water and concrete shielding. The ion exchange resins are changed routinely before high levels of radioactive materials have accumulated, and access to the demineralizer room is controlled, thus limiting personnel exposure.
Personnel exposure to the radiation from chemically inert U Ar is limited by dilution and prompt removal of this gas from the reactor area and its discharge 2
to the atmosphere, where it dif fuses further before reaching occupied areas.
I Nitrogen-16 production results in measurable radiation exposure rates around the pool; however, these are in areas where personnel occupancy is limited, so annual accumulated personnel exposures from this source are well within 10 CFR 20 limits.
12.3.2 Extraneous Sources
, Sources of radiation that may be considered as incidental to the normal reactor operation but associated with reactor use include radioactive isotopes produced for research, activated components of experiments, and activated samples or
! specimens.
Personnel exposure to radiation from intentionally produced radioactive material as well as from the required manipulation of activated experimental components is i
controlled by operating procedures that use the normal protective measures of I time, distance, and shielding.
12.4 Routine Monitoring 12.4.1 fixed Radiation Monitoring System There are six detectors in the PSBR fixed radiation monitoring system: (1) an ionization chamber on the east side of the bridge, (2) an ionization chamber
" on the west side of the bridge, (3) a Geiger-Mueller (GM) detector in the BHL, (4) a GM detector in the 60Co facility, (5) a particulate air monitor (GM) on PSU SER 12 _ _ _ _ _ _ _ _ - _ _ _ _ - _. - - .-. .
the east side of the reactor bay, and (6) a particulate air monitor (GM) on the west side of the reactor bay.
All of these are monitors read out in the control room. An alarm on any of these detectors results 'in a reactor scram and a building evacuation alarm.
Additional remote reading radiation detectors are located in the demineralizer room, hot cell areas, 60Co facility basement, lobby, lunchroom, and various l laboratories.
A GM detector monitors radiation levels in the PTSI containment box. An alarm on this monitor will open a valve to bypass the pressure relief manometer to relieve any pressure in the system, thus limiting leakage to the reactor room.
1 12.4.2 Experimental Support An experimenter must receive approval from the university's health physicist and
- the UIC to remove activated samples from the reactor' pool and use them in re-search. The health physicist reviews all proposed procedures for~the use of radioactive material for methods to minimize personnel exposures and limit the generation of radioactive waste.
1~
12.5 Occupational Radiation Exposures 12.5.1 Personnel Monitoring Program f Faculty, staff, students, and other facility users are issued either film badges or TLDs by the Health Physics Office to monitor their radiation exposure.
- Visitors are issued pocket dosimeters'.
I 12.5.2 Personnel Exposures The PSBR personnel annual exposure history for the last 5 years is given in Table 12.1. The results indicate that the management of potential radiation exposure at PSBR is acceptable and well within 10 CFR 20 guidelines.
1 Table 12.1 Number of individuals in exposure interval Number of individuals in each range Whole-body exposure range (rem) 1980 1981 1981 1983 1984 i
2 8 13 17 6 i No measurable exposure 10 8 5 17 Measurable exposure less than 0.1 15 l 1 0 0 0
- 0.1 to 0.25 1 0 0 0 t over 0.25 0 0 t
18 19 21 22 23 i
Number of individuals monitored i
l PSU SER 12-3 i
12.6 Effluent Monitoring 12.6.1 Airborne Effluent As discussed in Section 11, the major radioactive airborne effluent of the reac-tor facility is 41Ar. The average annual concentration of this radionuclide is a small fraction of the MPC for unrestricted areas and should result in insignif-icant radiation exposures to the general public.
12.6.2 Liquid Effluent Because of the manner in which liquid waste is handled, there is no radioactive liquid ef fluent resulting from operation of the reactor (see Section 11.2.2).
12.6.3 Environmental Monitoring Environmental monitoring consists of integrated (90-day) radiation readings from TLDs. Typical results of this program over the last 2 years indicate that the reactor contribution to radiation levels in the surrounding areas is negligible.
12.6.4 Potential Dose Assessment Natural background radiation levels in the central Pennsylvania area result in an exposure of about 115 mrems/yr to each individual residing there. At least an additional 7% (approximately 8 mrem /yr) will be received by those living in a brick or masonry structure. Any medical diagnosis X-ray examination will add to the natural background radiation, increasing the total cumulative annual exposure.
Conservative calculations based on the effluents of the facility and the results of the environmental monitoring program indicate that reactor operations do not ,
l contribute significantly to the annual exposures in unrestricted areas.
12.7 Conclusions The staff concludes that radiation protection' currently receives appropriate support from the university administration. The staff further concludes that (1) the program is staffed and equipped properly, (2) the reactor health physics staf f has adequate authority and lines of communication, (3) the procedures are integrated correctly into the research plans, and (4) surveys verify that operations and procedures follow ALARA principles.
Additionally, the staff concludes that the PSU radiation protection program is adequate based on the results of personnel monitoring and the environmental monitoring programs. Futhermore, there is reasonable assurance that personnel and procedures will continue to protect the health and safety of the public during continuing operations.
1 i
PSU SER 12-4
13 CONDUCT OF OPERATIONS 13.1 Overall Organization Responsibility for the safe operation of the reactor facility is vested within the chain of command shown in Figure 13.1. The PSBR Director is delegated responsibility, on behalf of the licensee, for overall f acility ' operation.
13.2 Training Most of the training of reactor operators is done by in-house personnel. The licensee's Operator Requalification Program has been reviewed, and the staff concludes that it meets the applicable regulations [10 CFR 50.54 (1-1) and Appendix A of 10 CFR 55] and is consistent with the guidance of ANS 15.4.
13.3 Operational Review and Audits The Penn State Reactor Safeguards Committee (PSRSC) provides independent review and audit-of facility activities. The Technical Specifications outline the quali-fications and provide that alternate members may be appointed by the Chairman.
The PSRSC must review and approve plans for modifications to the reactor, new experiments, and proposed changes to the license or procedures. The PSRSC also is responsible for conducting audits of reactor facility operations and manage-ment and for reporting the results thereof to the university administration.
I 13.4 Emergency Planning 10 CFR 50.54(q) and (r) require that a licensee authorized to possess and/or 4 operate a research reactor shall follow and maintain in effect an emergency plan that meets the requirements of Appendix E of 10 CFR 50. A revised Emergency Plan was submitted by the licensee and approved by the NRC on January 22, 1985.
13.5 Physical Security Plan j
The PSBR facility has established and maintains a program to protect the reactor and its fuel and to ensure its security. The NRC staff has reviewed the Physical Security Plan and concludes that the plan meets the requirements of 10 CFR 73.67 for special nuclear material of-low strategic significance. The PSBR facility's l
inventory of special nuclear material for reactor operation falls within that category.
Both the Physical Security Plan and the staff's evaluation are withheld from pub-lic disclosure under 10 CFR 2.730(d)(1). Amendment No. 21 to the facility Operat-ing License R-2 dated November 25, 1981, incorporated the Physical Security Plan
~
- as a condition of the license.
i 13.6 Conclusion ,
J On the basis of the above discussions, the. staff concludes that the licensee has sufficient experience, management structure, and procedures to provide reasonable PSU SER 13-1 v.. n - - , . . . . - . , - - - - , - . , , - - , , .
assurance that the PSBR will continue to be managed in a way that will cause no significant radiological risk to the health and safety of the public, j
'l 1
i Vice President Research and Graduate Studies l
l i
Dean, College of Engineering i l
I
' Nuclear Engineering Department Head University Health i
Physics Penn State Reactor i
Safeguards Committee I
I l l l
i L_____ oirector ____J Penn State Breazeale Reactor i
i Reactor Staff i
Figure 13.1 Organization chart PSU SER 13-2
14 ACCIDENT ANALYSIS The NRC staff has evaluated the applicant's documentation and analyses of poten-tial site-specific events. These analyses included the various types of possible accidents and the potential contcquences to the public.
The following potential accidents or effects were considered to be sufficiently credible for evaluation and analysis.
(1) fuel handling (2) rapid insertion of reactivity (nuclear excursion)
- (3) loss of coolant
- (4) misplaced experiments j (5) fuel aging l
i Of these potential events, only one, the fuel-handling accident with the loss j of cladding integrity of one irradiated fuel element in air in the reactor room, i
would have a potential effect on the environment outside the PSBR. Thus, the
, fuel-handling accident will be designated as the maximum hypothetical accident (MHA). An MHA is defined as a hypothetically conceived accident for which the risk to the public health and safety is greater than that from any other event that can be postulated mechanistically. The staff assumes that the accident occurs but does not attempt to describe or evaluate all of the mechanical details ,
of the accident or the probability of its occurrence. Only the consequences are i - considered.
The results of the analyses of accidents with less severe consequences than the MHA are included to demonstrate the extent of the staff investigation.
3 14.1 Fuel-Handling Accident i
This potential accident, designated as the MHA, includes various incidents to j at least one or more irradiated fuel elements in which the fuel cladding might
- be breached or ruptured. To remain conservative, the staff did not try to
! develop a detailed scenario but assumed that the cladding of one fuel element completely fails and that this occurs outside the reactor pool, instantly ;
i releasing the volatile fission products that have accumulated in the free 4 volume (gap) between the fuel and the cladding.
i Several series of experiments by the fuel vendor [ General Atomic (GA)] have
[
given data on the species and fractions of fission products released from U-ZrH*
1 under various conditions (GA-4314; GA-8597; Foushee and Peters, 1971; Baldwin, Foushee, and Greenwood, 1980; Simnad,'1976). The noble gases were the principal 2 species found to be released. When the fuel specimens were irradiated at tem-
{ peratures below 350 C, the fraction released could be summarized as a constant equal to 1.5 x 10 5, independent of the temperature. At temperatures greater than 350 C, the species released remained the same, but the fraction released increased significantly with increasing temperature.
1
- PSU SER 14-1 l
'i
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GA has proposed a theory describing the release mechanisms in two temperature regimes that appears to be valid, but the data do not agree in detail (Foushee and Peters, 1971). It seems reasonable to accept the interpretation of the low-temperature results, which imply that the fraction released for a standard TRIGA fuel element will be a constant, independent of operating history or details of operating temperatures, and will apply to fuel whose maximum tem-perature is not raised above 400 C. This means that the 1.5 x 10 5 release fraction reasonably could be applied to TRIGA reactors operating up to at least 800 kW. However, the PSBR has a licensed power level of 1 MW with a correspond-ing peak fuel temperature of ~467 C, so a higher release fraction must be assumed to be conservative.
The theory in the fuel temperature regime above 350 C is not as well established.
The proposed theory of release of the fission products incorporates a diffusion process that is a function of temperature and time. Therefore, in principle, details of the operating history and temperature distributions in fuel elements would be required to obtain actual values for release fractions at the higher temperatures. In situations where a fuel cladding failure was assumed, the staff used the GA results (GA-8597; Foushee and Peters, 1971; Simnad, 1976) to estimate fission product release fractions. The staff considers these results to be conservative in that they represent a theoretical maximum release greater than corresponding experimental observations. Thus, for the fuel-handling accident, the staff estimated a fission product release fraction of 7.5 x 10 5 of the inventory of both noble gases and halogens.
Because the noble gases do not condense or combine chemically, it is appropriate to assume that any released from the cladding will diffuse in the air until their radioactive decay. On the other hand, the iodines are chemically active and are not volatile below about ~180 C. Therefore, some of the radiciodines will be trapped by materials with which they come in contact, such as water and structures. In fact, evidence indicates that most of the lodines either will not become or not remain airborne under many accident scenarios that are appli-cable to nonpower reactors (NUREG-0771). However, to be certain that the fuel-cladding-failure scenario discussed below led to upper-limit dose estimates for all events, the staff assumed that 100% of the lodines in the gap become airborne and that all fission products had reached their saturated activity levels.
These are very conservative assumptions considering the historical operating environment at the PSBR. Normally, a significiant amount-of time elapses before removing fuel from the reactor; nevertheless, no activity decrease was assumed for radioactive decay during the time between reactor shutdown and removal of the fuel element from the pool into the air. These assumptions lead to computed doses principally resulting from iodines that may be at least a factor of 100 higher than in more realistic scenarios, for example, those in which the cladding fails while the fuel element is immersed in the pool water.
14.1.1 Scenario The staff's analysis assumed that a cladding failure occurs in a B ring fuel element immediately following an extended run at the authorized maximum power.
All the noble gases and halogens in the fuel cladding gap are released from the fuel element and instantly form a uniform distribution in the reactor room air.
- No plate-out was assumed. Scenarios incorporating realistic estimates of the above conservative assumptions could reduce the resulting computed doses significantly.
PSU SER 14-2
The staff calculated the potential whole-body gamma-ray (immersion) and thyroid doses by iodine inhalation to an individual in the reactor room and in an
- unrestrictqd area immediately outside the building. For the occupational doses, it was assumed that the ventilation system was shut down at the time of the accident and that all the fission products remained in the reactor room.
, Additionally, immersion in a finite cloud was assumed. For the outside doses, i
it was assumed that the normal ventilation system was operating at its rated capacity. Dose calculations for the outside receptor assumed immersion'in a semi-infinite cloud, which leads to computed whole-body doses that are much i higher than could occur in a realistic finite cloud (NUREG-0851).
14.1.2 Assessment The calculated doses for the above assumptions and locations are presented in Table-14.1. Because there is no credible way in which the postulated accident could occur without operating personnel being alerted immediately, orderly l
evacuation of the reactor bay would be accomplished within minutes. As a result of the underlying calculative and atmospheric assumptions, the calculated doses within and outside the facility, shown in Table 14.1, are higher than could occur realistically.
- Table 14.1 Doses resulting from postulated fuel-handling accident j Whole-body Thyroid committed
! Dose and location immersion dose dose l 10-min exposure in the reactor bay 1 0.115 arem 0.125 rem 30-min exposure immediately outside the building 2 0.381 mrem 0.01 mrem IFinite cloud model.
2 Semi-infinite cloud model.
14.2 Rapid Insertion of Reactivity As discussed in Section 4.4 of this report, theoretical calculations have pre-dicted and experimental measurements have confirmed that U-ZrH* fuel exhibits a strong,' prompt, negative temperature coefficient of reactivity This tempera-ture coefficient not only tends to terminate a pulse or nuclear excursion but also causes a loss of reactivity as the temperature of the fuel is raised during nonpulsed operation. These results have been verified at many operating TRIGA reactors. Although it may be possible theoretically to rapidly add sufficient excess reactivity to a U-ZrHx reactor to create an excursion that would not be terminated before fuel damage occurred, the limits imposed by the design and Technical Specifications of the PSBR make such an event incredible.
In some PSBR configurations, full withdrawal of the transient rod could result in a reactivity insertion greater than the authorized maximum pulse insertion.
In such case, administrative controls are applied to the adjustment of the PSU SER 14-3
transient rod stroke to ensure that the maximum allowed pulse reactivity is not exceeded. The Technical Specifications for the PSBR limit the maximum allowed pulse reactivity insertion and the maximum worth of the transient rod to 2.31% ak/k (3.30$) and 2.59% ak/k, respectively.
14.2.1 Scenario The staff knows of no credible method of rapidly inserting the total authorized excess reactivity into the core. Therefore, in this postulated event, it is assumed that the maximum excess reactivity available in a single credible event is inserted into the reactor instantaneously.
The staff has considered the scenario of pulsed operation from a power level I between 0 and 100 kW and simultaneously the failure of a movable experiment inserts 0.49% ak/k positive reactivity into the core. The analysis neglected reactivity loss as a result of the buildup of 135Xe, a conservative assumption.
l Analyses have found that the higher the temperature at which the rapid insertion is initiated, the lower the final temperature of the fuel immediately after the transient. Therefore, the staff assumed the worst case: initiation of a 2.59% ak/k transient with the core at ambient temperature and essentially zero initial power. This corresponds to the technical specification limit for the maximum positive reactivity worth that a pulse and movable experiment can insert.
The potentially significant consequences of the reactivity insertion accidents considered by the staff are melting of the fuel or cladding material and failure of the cladding as a result of high internal gas pressures and/or phase changes in the fuel matrix. The primary cause of cladding failure at elevated tempera-tures in stainless steel-clad elements would be excessive stress buildup in the cladding caused by hydrogen pressure from disassociation of the ZrH . Calcula-tions performed by GA and confirmed in many reactor pulses indicate *that water-immersed cladding integrity is maintained at peak fuel temperatures as high as 1175 C (GA-4314; GA-6874; Simnad et al., 1976).
1 14.2.2 Assessment GA has performed many experiments with reactivity insertions as high as 3.5% ak/k (5.00$) in an 85-element TRIGA core. GA measured, among other parameters, the temperature of the fuel in the hottest core position and examined fuel elements afterward (GA-6874; Simnad et al., 1976). There was no indication of undue stress in the chdding and no indications of either cladding or fuel melting. The mea- t sured maximum temperature for the 3.5% ak/k pulse was 750 C, and the estimated peak transient temperature at any localized point in the fuel was 1175"C. Because the radial temperature distribution in a fuel element immediately following a pulse is similar to the radial power distribution, the peak transient temperature immediately after the pulse is located at the periphery of the hottest fuel ele-ment. It will fall rapidly (within seconds) as the heat flows toward the clad-ding and toward the fuel center. It also was observed that for a 3.5% ak/k pulse, the maximum measured pressure rise within an instrumented fuel element was far below the predicted equilibrium value at the peak temperature (GA-9064; GA-6874; Simnad et al., 1976). On the basis of these considerations, the staff concludes
, that the excess reactivity in the PSBR available for rapid insertion is not suf-ficient to support a transient that would lead to cladding failure. or fission product release.
PSU SER 14-4 i
L I
14.3 Loss-of-Coolant Accident A potential accident that would result in increases in the fuel and cladding temperatures is the loss of coolant shortly after the reactor has been operating.
Because water is required for adequate neutron moderation, its removal would terminate any significant neutron chain reaction. However, the residual radio-activity from fission product-decay would continue to deposit heat energy in the fuel.
. The staff assumed that the reactor has been operating at the licensed power of I 1 MW long enough to achieve fission product equilibrium (a conservative assump-tion based on a technical specification limit as well as expected usage) and is shut down at the initiation of a gross cooling-water leak. It is further assumed that heat is removed by convective water cooling until the top of the core L becomes uncovered, after which heat is removed only by air convection.
Several investigations have analyzed such scenarios under various assumptions I (GA-6596; GA-9064; Texas A&M, 1979; Oregon State University, 1968). In the PSBR, ,
i the core will remain completely immersed in water as long as the water level is l 1.8 m above the tank bottom. That would require a decrease of 4.9 m in the pool i
l water level before the top of the core becomes uncovered. The largest credible r loss-of-coolant accident (LOCA) is one where the 15.24-cm pipe connected to the l
i bottom of the pool fails. The core would remain partially or totally covered for
- i at least 23 min. At full licensed power, the PSBR generates a total of ~23 kW in the hottest fuel element. GA has performed experiments on a similar type reactor and has shown that the maximum fuel temperature reached was 761*C.
I Additionally, analytical studies performed by GA have shown that a fuel element I generating ~23-kW power before the LOCA will attain a maximum fuel temperature j of ~870 C. During the LOCA, the maximum temperature reached would be well below j 900*C, which is a conservative upper limit recommended by GA for air-coole'd,
- stainless steel clad TRIGA fuel (GA-4314; Simnad, 1976).
i A pulse performed immediately before water loss would not contribute signifi- [
j cantly to the fission product decay heat, and the heat generated during the i pulse would be removed during the 23-min interval before the core becomes uncovered.
The Technical Specifications require that corrective action be taken or the
.l reactor be shut down if the water level falls below a level of ~4.9 m~above the top of the core. In addition, water level and radiation monitors would alert the operating staff to a low water condition. Even if the coolant loss were preceded by an extended reactor run at the maximum authorized power level of j 1 MW followed by a 2.3% Ak/k pulse, the resultant maximum fuel and/or cladding I temperatures would not cause fuel damage or fission product release.
q On the basis of the above,.the staff concurs with the licensee that there is l reasonable assurance that the reactor cannot suffer a loss of coolant that would i raise any fuel temperature above 900*C.
I 14.4 Misplaced Experiments This type of potential accident is one in which an experimental sample or device inadvertently is located in an experimental facility where the irradiation condi-j
- tions could exceed the design specifications. In that case, the sample might i
PSU SER 14-5 j f
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become overheated or develop pressures that could cause a failure of the experi- i ment container. As discussed in Sections 10 and 13, all new experiments at the PSBR facility are reviewed before insertion, and all experiments in the region of the core are separated from the fuel cladding by at least one barrier, such as the pneumatic transfer and irradiation tubes, the central thimble, or the core screen.
l The staf f concludes that the experimental facilities and the procedures for experiment review at the PSBR are adequate to provide reasonable assurance that failure nf experiments is not likely, and even if failure occurred, breaching of the reactor fuel cladding will not occur. Furthermore, if an experiment should fail and release radioactivity within an experimental facility, there is reasonable assurance that the amount of radioactivity released to the environ-ment would not be more than that of the proposed MHA.
i
, 14.5 Effects of Fuel Aging Fuel aging is considered normal and is expected to occur gradually. The possi-
- bility of internal reactions is discussed in this section.
4 There is some evidence that U-ZrH fuel tends to fragment with use, probably
' because of the stresses caused by*high temperature gradients and the high rate of heating during pulsing (GA-4314; GA-9064). Some of the possible consequences of fragmentation are (1) a decrease in thermal conductivity across cracks, lead-ing to higher central fuel temperatures during normal operationi and (2) an '
increase in the amount of fission products released into the cracks in the
, fuel.
j With regard to the first item above, hot cell examination of thermally stressed i
hydride fuel bodies has shown relatively widely spaced radial cracks that would cause minimal interference with radial heat flow (GA-4314; GA-9064). However, after pulsing, TRIGA reactors have exhibited an increase in both steady-state fuel temperatures and power reactivity coefficients. At power levels of 1 MW, j}. temperatures have increased by ~23 C and power reactivity coefficients have increased by ~20% [ Armed Forces Radiological Research Institute (AFRRI), 1960;
} GA-5400). GA has attributed these changes to an increased gap between the fuel j
material and the cladding (caused by rapid fuel expansion during pulse heating) that reduces the heat transfer coefficient. Experience has shown that the ob-served changes occur mostly during the first several pulses and have essentially saturated after 100 pulses. Therefore, the PSBR should not experience any further changes in the fuel-cladding gap caused by pulsing.
As noted in Section 14.1, two mechanisms for fission product release from TRIGA fuel have been identified by GA (GA-4314; GA-8597; Foushee and Peters, 1971). The first mechanism is fission fragment recoil into gaps within the fuel cladding.
j This effect predominates up to ~400 C and is independent of fuel temperature.
I
- Temperature distributions during pulsing would not be affected significantly
- by changes in conductivity because a pulse is completed before significant I heat redistribution occurs.
1 PSU SER 14-6
GA has postulated that, in a closed system such as exists in a TRIGA fuel ele-ment, fragmentation of the fuel material within the claddina will not cause an incrcase in the fission product release fraction (GA-8597). The reason for this is that the total free volume available for fission products remains constant within the confines of the cladding. Under these conditions, the formation of a new gap or widening of an existing gap must cause a corresponding narrowing of an existing gap at some other location. Such a narrowing allows more fission frag-l ments to traverse the gap and become embedded in the fuel or cladding material on the other side. In a closed system, the average gap size, and therefore the fission.prodJCt release rate, remains Constant, independent of the degree to ?
which fuel material is broken up.
Above ~400 C, the controlling mechanism for fission product release is diffusion, and the amount released depends on fuel temperature and fuel surface-to-volume ratio. However, release fractions used for the safety evaluation are based on a conser.vative calculation that assumed a degree of fuel fragmentation greater ;
i than expected in actual operation.
l i As the two likely effects of aging of the U-ZrH fuel moderator will not have
- asignificanteffectontheoperatingtemperatufeofthefuelorontheassumed 4
-release of gaseous fission products from the cladding, the staff concludes that ,
j there is reasonable assurance that fuel aging will not significantly increase .l l
the likelihood of fuel-cladding failure or the calculated consequences of an 1
accidental release in the event of the loss of cladding integrity.
l 14.6 Conclusion i
i
- The staff has reviewed the credible accide*,.s for the PSBR. On the basis of j this review, it has been determined that the postulated accident with the greate:t potential effect on the environment is the loss of cladding integrity of an irradiated fuel element in air in the reactor room. The analysis of this accident has indicated that even if several fuel rods failed simultaneously, l the expected dose equivalents in unrestricted areas still would be below the 10 CFR 20 guideline values. Therefore, the staff concludes that the design of l the facility and the Technical Specifications provide reasonable assurance that
! the PSBR can continue to be operated with no significant risk to the health and j safety of the public.
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PSU SER 14-7 5 - - - - . - - - - - - - , _ . . .- .. , - _ . _ , - _ - - , _ _ _ _ _ _ _ , _ _ _ _ __ , _ _
15 TECHNICAL SPECIFICATIONS The licensee's Technical Specifications evaluated in.this licensing action define certain features, characteristics, and conditions governing the continued operation of this facility. These Technical Specifications are explicitly included in the renewal license as Appendix'A. Formats and contents acceptable to the NRC have been used in the development of these Technical Specifications, l
l and the staff has reviewed them using ANS 15.1, "The Development of Technical !
I Specifications for Research Reactors" as a guide.
l On the basis of its review, the staff concludes that normal operation of the f PSBR within the limits of the Technical Specifications will not result in offsite l
radiation exposures in excess of the guidelines of 10 CFR 20. Furthermore, the limiting conditions for operation and surveillance requirements will limit the likelihood of malfunctions and mitigate the consequences to the public of off-normal or accident events.
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PSU SER 15-1
t 16 FINANCIAL QUALIFICATIONS i The Pennsylvania State University Breazeale reactor is owned and operated by a '
l state educational institution in support of its role in education and research. ~
On the basis of financial information supplied by the licensee in its March 1, ,
1985 submittal, the staff concludes that funds will be made available, as neces- i sary, to support continued operations and eventually to shut down the facility cnd maintain it in a condition that would constitute no risk to the public.
The licensee's financial status was reviewed and found to be acceptable in i accordance with the requirements of 10 CFR 50.33(f).
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I PSU SER 16-1 I '
. . _ . . _ _ _ . . . . , - - - . . _ . - . . . . . _ _ - . . _ _ _ , _ - . - . . . . . _ _ _ . . . _ _ _ _ . . _ _ _ , _ _ _ . . ~ . _ _ . _ _ _ , ._
17 OTHER LICENSE CONSIDERATIONS 17.1 Prior Reactor Utilization Previous sections of this SER concluded that normal operation of the reactor causes insignificant risk of radiation exposure to the public and that only an off-normal or accident event could cause some significant expusure. The maxi-mum hypothetical accident (MHA) was shown to result in potential radiation exposures within applicable guideline values of 10 CFR 20.
The staff has reviewed the impact of prior operation of the facility on the risk of radiation exposure to the public. Although the staff has concluded that the reactor was initially designed and constructed with both inherent safety and acceptable engineered safety features, the staff considered whether continued operation would cause significant degradation in these features.
Because loss of integrity of fuel cladding is possible, the staff considered l mechanisms that could increase the likelihood of failure. Possible mechanisms are (1) radiation degradation of cladding strength, (2) high internal pressure caused by high temperature leading to exceeding the elastic limits of the clad-ding, (3) corrosion or erosion of the cladding leading to thinning or other weakening, (4) mechanical damage as a result of handling or experimental use, and (5) degradation of safety components or systems. .
The staff's conclusions regarding these considerations, in the order in which l
they were identified above, are as follows:
I (1) The stainless-steel-clad TRIGA fuel in the core has beenSome in use sincefuel 1965 and has been subjected to a maximum of 36% burnup of zasu. TRIGA at more extensively used reactors has been in use for even higher burnup, with no observable degradation of cladding as a result of radiation.
(2) The possibility of approaching excessive pressures would occur if the entire fuel element including the cladding were to be heated to more than 930 C.
Although it is likely that some points in the fuel would exceed this temperature i
for a few seconds following a 3.35 pulse, only a simultaneous and instantaneous 4 total loss of coolant could cause the cladding temperature to exceed a few hun-dred degrees. Because the staff considers that there is no credible scenario involving all of these assumptions, the staff concludes that there is no realis-tic event that would cause the elastic limit of the cladding to be exceeded.
(3) Water flow through the core is obtained by natural thermal convection, so the staff concludes that erosion effects that might result from high flow veloc-ity will be negligible. High primary water purity is maintained With conductivity by continuous below passage through the filter and demineralizer system.
about 5 pmhos/cm.1, corrosion of the stainless-steel cladding is expected to be negligible.
(4) The fuel is handled as infrequently as possible, consistent with periodic surveillance. Any indications of possible damage or degradation are investigated i
PSU SER 17-1 l
promptly. The only experiments that are placed near the core are isolated from the fuel cladding by a water gap and at least one metal barrier, such as the pneumatic tubes or the central thimble. Therefore, the staff concludes that the loss of' integrity of cladding through damage does not constitute a signifi-cant risk to the public.
(5) The PSBR staf f performs regular preventive and corrective maintenance and replaces components as necessary. Nevertheless, there have been some malfunc-tions of equipment. However, the staff review indicates that most of these malfunctions have been random one-of-a-kind incidents. There is no indication of significant degradation of the instrumentation, and the staff concludes that there is strong evidence that any future degradation will lead to prompt reme-dial action by the PSBR staff. Therefore, there is reasonable assurance that there will be no significant increase in the likelihood of occurrence of a reactor accident as a result of component malfunction.
17.2 Conclusion On the basis of the above considerations, the staff concludes that there are no credible events resulting from reactor aging that could produce offsite radio-logical hazards greater than those already analyzed in Section 14.
PSU SER 17-2
18 CONCLUSIONS On the basis of its evaluation of the application as set forth above, the staff has determined that (1) The application for renewal of Operating License R-2 for its research reac-tar filed by the Pennsylvania State University', dated March 1, 1985, as sup-plemented, complies with the requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I.
(2) The facility will operate in conformity with the application as amended, the provisions of the Act, and the rules and regulations of the Commission.
(3) There is reasonable assurance (a) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public and (b) that such activities will be conducted in compliance with the regulations of the Commission set forth in-10 CFR Chapter I.
, (4) The licensee is technically and financially qualified to engage in the activities authorized by the license in accordance with the regulations
- of the Commission set forth in 10 CFR Chapter I.
(5) The renewal of this license will not be inimical to the common defense and security or to the health and safety of the public, i
i a
PSU SER 18-l'
19 REFERENCES American National Standards Institute /American Nuclear Society (ANSI /ANS),
15 series.
American Nuclear Society (ANS) 15.1, " Standard for the Development of Technical Specifications for Research Reactors," September 1982.
--_, ANS 15~.16, " Standard for Emergency Planning for Research Reactor," Draft 2, November 1981.
-- , ANSI /ANS 15.11, " Radiological Control at Research Reactor Facilities," 1977.
Armed Forces Radiobiology Research Institute (AFRRI), " Final Safeguards Report for the AFRRI TRIGA Reactor," Appendix A, Docket 50-170, November 1960.
Baldwin, N. L., F. C. Foushee, and J. S. Greenwood, " Fission Product Release From TRIGA-LEU Reactor Fuels," Proceedings: Seventh Biennial .U.S. TRIGA User's Conference, San Diego, CA, March 2-5, 1980.
Foushee, F. C. and R. H. Peters, " Summary of TRIGA Fuel Fission Product Release Experiments," Gulf Corporation report Gulf-EES-A10801, San Diego, CA, Sep-tember 1971.
General Atomic Company, GA-0471, " Technical Foundations of TRIGA," August 1958.
-- , GA-4314, Simnad, M. T., "The U-ZrHx Alloy: Its Properties and Use in TRIGA Fuel," E-117-833, February 1980.
-- , GA-5400, "Thermionic Research TRIGA Reactor Description and Analysis,"
transmitted by letter dated February 28, 1966 (Docket No. 50-227), Rev.- C, November 1, 1965.
-- , GA-6596, Shoptaugh J. F. , Jr. , " Simulated Loss-of-Coolant Accident for TRIGA Reactors," transmitted by letter dated September 22, 1970 (Docket No. 50-227). ,
-- , GA-8597, Fou;4 hee, F. C. , " Release of Rare Gas Fission Products from U-ZrH Fuel Material," March 1968.
-- , GA-9064, West, G. B. " Safety Analysis Report for the Torrey Pines TRIGA Mark III Reactor," transmitted by letter dated Janaury 29, 1970 (Docket No. 50-227),
January 5,1970.
Northeastern U.S. Seismic Network, " Seismicity of the United States," Bulletin No. 32, compiled and edited by J. E. Foley, C. Doll, F. Filipkowski, G. Lorsback, and J. Matigue, Weston Observatory, Boston College, January 1984.
Oregon State University, "SAR for the Oregon State University TRIGA Research Reactor," Docket 50-243, August 1968.
PSU SER 19-1
Simnad, M. T., F. C. Foushee, and G. B. West, " Fuel Elements for Pulsed TRIGA Research Reactors," Nuclear Technology 28, 31-56 (1976).
Stover, C. W. , B. G. Reagor, and S. T. Algermissen, " Seismicity Map of the State of Pennsylvania," U.S. Geological Survey Map MF-1280, U.S. Department of Interior,.1981.
Texas A & M, SAR for the Nuclear Science Center Reactor, Texas A & M University (Docket 50-128), June 1979.
PSU SER 19-2
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'C'"ds' BIBLIOGRAPHIC DATA SHEET su ,~st ucri s os 1-1 au se NUREG-1158 ,
,ma.s3suoiva 2a.on.u Safety Ev luation Report related to the renewal of the operating icense for the research reactor at Pennsylvania
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Office of Nuclear Rea or Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 4 3 3 os se..so 0.c.v z u cs ... .so ..u .p. s, ,, <, c , , , , ... o, i.on Division of PWR Licensing-B Office of Nuclear Reactor gglationRe\ Safety Evaluation Report U.S. Nuclear Regulatory CommisFon '. n a'a cou aio "~ ~~
Uashington, DC 20555 12 is ,LiWiNT.R V NC'e s Docket No.50-005 o . sta.cracc.-,.- m This Safety Evaluation Report for t e applica on filed by the Pennsylvania State University for a renewal of Operating Lic se R-2 o continue to operate the Penn State Breazeale Reactor (PSBR) has been prepared y th Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Comission. .he acility is owned and operated by the Pennsylvania State University and is located h the campus in University Park, Pennsylvania. Based on its technical review, e staff concludes that the reactor facility can continue to be operated by the U i rsity without endangering the health and safety of the public, or the environment
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