ML072210846

Texas A&M - Nuclear Science Center Reactor SAR Relicensing Report Redacted Version

ML072210846
Person / Time
Site:	05000128
Issue date:	02/27/2003
From:	Reece W Texas A&M Univ
To:	Document Control Desk, Office of Nuclear Reactor Regulation
WASSON, D NRR/DPR/PRTA 415-2862
References
2003-0024
Download: ML072210846 (157)

Text

Nuclear Science Center Reactor LICENSE: R-83 DOCKET: 50-128/

RELICENSING REPORT REDACTED VERSION SECURITY RELATED INFORMATION REMOVED Redacted text and figures have been blacked out

TEXAS E N G I N E E R I N G EXPERIMENT STATION TEXAS A&M UNIVERSITY 3575 TAMU COLLEGE STATION. TEXAS 77843-3575 NUCLEAR SCIENCE CENTER 9791845-7551 FAX 9791862-2667 February 27,2003 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Reference:

Application for Renewal of R-83 Docket Number 50-128 This memo and the enclosed Safety Analysis Report (SAR) serve as application for 20-year renewal of the Texas A&M Nuclear Science Center with the listed license and docket number. The combination of this letter and the enclosed SAR meet the requirements of 10CFR54.

The Nuclear Science Center Emergency Plan, Physical Security Plan and Operator Requalification Plan will have no associated changes. The enclosed SAR addresses the financial and environmental issues concerning the continued operation of the Nuclear Science Center.

Sincerely, Warren D. Reece, Director TAMU Nuclear Science Center xc: 2.1 Ifcentral File Marvin Mendonca, USNRC USNRC, Region IV RESEARCH AND DEVELOPMENT FOR MANKIND httpJ/nsc tamu edu

SAFETY ANALYSIS REPORT for the Nuclear Science Center Reactor Texas A&M University Texas Engineering Experiment Station February 2003

ABSTRACT T h ~ document s supports of the renewal of L~censeR-83 and supercedes all prev~oussubmittals in Docket 50-128.

T h ~ Safety s Analys~sReport (SAR) is a consohdated and updated safety analysis for the contmued operat~onof the Nuclear Sc~enceCenter Reactor (NSCR) uslng standard TRIGA and/or FLIP TRlGA fuel and contams prev~ously reviewed mater~alfrom the August 1967 and June 1979 SARs and thelr supplements The purpose of t h ~ SAR s 1s to provlde a description and safety analysis of structures, systems and components in terms of thelr ability to prov~deproper operat~onalperformance and functions for the twenty-year term of the l~cense renewal The continual upgrad~ngthat has been implemented smce the j n ~ t ~ operation al of the NSCR has improved reactor safety and prevented the need for restrictions on reactor operations due to age of structures or equipment.

CONTENTS ABSTRACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .I1 THE FACILITY ......................................................................... 1.1 Introduction ....................................................................................................1.1 Summary and Conclusions on Prlnclpal Safety Considerations .............................................. 1-1 General Description of the Facility ............................................................................1-1 Shared Facilities and Equipment ................................................................................................ 1-2 Comparison wlth Similar Facilities ..................................................................................................1-2 Summary of Operation ....................................................................................................................1-2 Compliance with the Nuclear Waste Policy Act of 1982............................................................... 1-2 Facility Modifications and Hlstory ......................................................................................... 1-3 SITE CHARACTERISTICS ............................................................................................ 2-1 Geography and Demography ................................................................................................... 2-1 Nearby Industrial, Transportation, and Mllitary Facilities ....................................................................2-5 Meteorology .................................................................................................................... 2-6 Hydrology ......................................................................................................................... 2-8 Geology, Seismology, and Geotechnical Engineering .................................................................2-8 DESIGN OF STRUCTURES. SYSTEMS. AND COMPONENTS ..................................................3-1 Design Criter~a........................................................................................................3-1 Meteorological Damage ...................................................................................................... 3-1 Water Damage ........................................................................................................................ 3-1 Seismic Damage ......................................................................................................................... 3-1 Systems and Components ............................................................................................................... 3-1 REACTOR DESCRIPTION ......................................................................................................... 4-1 Summary Description ..................................................................................................................4-1 Reactor Core ...............................................................................................................................4-1 Reactor Pool .................................................................................................................................... 4-9 Biological Shield ................................................................................................................................... 4-9 Nuclear Design ............................................................................... ...................................... .............. 4-10 Thermal-Hydraulic Design ............................................................................................................. 4-23 REACTOR COOLANT SYSTEMS ...................................................................................................... 5-1 Summary Description .............................................................................................................................. 5-1 Primary Coolant System ......................................................................................................................... 5-3 Secondary Coolant System .................................................................................................................. 5-4 Primary Coolant Cleanup System .......................................................................................................... 5-5 Primary Coolant Makeup Water System.................................................................................................. 5-7 Nitrogen-16 Control System .................................................................................................................... 5-7 Auxiliary Systems Using Primary Coolant ............................................................................................... 5-7 ENGINEERED SAFETY FEATURES ............................................................................................ 6-1 Summary Description ....................................................................................................................... 6-1 Detailed Descriptions ............................................................................................................................ 6-1 INSTRUMENTATION AND CONTROL SYSTEMS ............................................................................7-1 Summary Description .......................................................................................................................7-1 Design of Instrumentation and Control Systems............................................................................... 7-1 Reactor Control System .............................................................................................................................7-6 Reactor Protection System ........................................................................................................... 7-17 Engineered Safety Features Actuation System ................................................................................... 7-18

7.6 Control Console and Display Instruments .....................................................................7-38 7.7 Radiation Monltor~ngSystems ................................................................ 7-20 8 ELECTRICAL POWER SYSTEMS ......................................................... 8-1 81 Normal Electrical Power Systems ...................................................... 8-1 82 Emergency Electrical Power Systems........................................................... 8-1 AUXILIARY SYSTEMS ................................................................. 9-1 Heatlng. Ventilation. and Alr Condltionlng Systems.................................................. 9-1 Handling and Storage of Reactor Fuel .........................................................................................9-2 Flre Protections Systems and Programs ................................................................................9-3 Communicat~onSystems ........................................................................................................... 9-3 Possession and Use of Byproduct. Source. and Speclal Nuclear Materlal ............................................. 9-3 Cover Gas Control in Closed Primary Coolant Systems ...................................................................... 9-3 Other Auxiliary Systems ................................................................................................ 9-3 10 EXPERIMENTAL FACILITIES AND UTILIZATION .............................................................. 10-1 10.1 Summary Description ........................................................................................................................10-1 10.2 Experimental Fac~lities........................................................................................................... 10-1 10 3 Experiment Review .................................................................................................................. 10-7 11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT .................................11-1 11 1 Radlatlon Protection ............................................................................................................... 11-1 11.2 Rad~oact~ve Waste Management ........................................................................................11-5 11.1 Bibhography .............................................................................................................. 11-5 CONDUCT OF OPERATIONS .......................................................................................................... 12-1 Organization ...................................................................................................................... 12-1 Reactor Safety Board (RSB) Review and Audit Actlvlt~es............................................................12-3 Procedures ....................................................................................................................... 12-4 Required Actlons .................................................................................................... 12-5 Reports ...................................................................................................................................12-6 Records .............................................................................................................................................12-7 Emergency Planning .................................................................................................................. 12-8 Security Planning ........................................................................................................................... 12-9 Quality Assurance .................................................................................................................. 12-9 Operator Training and Requalification .............................................................................................. 12-9 Startup Plan .......................................................................................................................................... 12-9 Environmental Reports .................................................................................................................... 12-9 13 ACCIDENT ANALYSIS ....................................................................................................................... 13-1 13.1 Accident-Initiating Events and Scenarios ................................................................................................ 13-1 13.2 Accident Analysis and Determination of Consequences ......................................................................... 13-1 13.3 Summary and Conclusions .................................................................................................................. 13-13 14 TECHNICAL SPECIFICATIONS FOR FACILITY LICENSE NO . R-83 ............................................ 14-1 14.1 Definitions ............................................................................................................................................. 14-1 14 2 Safety Limlt and Limiting Safety System Setting ................................................................................. 14-5 14 3 Limiting Conditions for Operation .....................................................................................................14-7 14 4 Surveillance Requirements ......................................................................................................... 14-19 14.5 Design Features.............................................................................................................................. 14-24 14.6 Administrative Controls .................................................................................................................... 14-29 15 FINANCIAL QUALIFICATIONS .......................................................................................................15-1 15.1 Financial Ablllty to Construct a Non-Power Reactor .....................................................................15-1 15 2 Financial Ability to Operate a Non-Power Reactor ............................................................................. 15-1

15.3 F~nancialAbll~tyto Decomm~ss~on the F a c ~ l ~....................................................

ty 15-1 16 OTIIER LICENSE CONSIDERATIONS ................................................. 16- 1 16 1 Prior Use of Reactor Components .............................................................. 16-1 16 2 Med~calUse of Non-Power Reactors. ...................................................... 16-1

LIST OF FIGURES FIGURE1.1 NUCLEAR SCIENCE CENTER REACTOR AND LABORATORY BUILDINGS . . . . . . . . . . . . . . . . . . . . . . . 1 .1 FIGURE2-1: NSC AND EASTERWOOD AIRPORT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 FIGURE2.2 NSC SITE........................................................................ 2-2 a

FIGURE2-3: MAJOR ROADSAROUND NSC AND EASTERWOOD ..............................................................2-3 FIGURE2-4: NSCR BUILDING CROSSSECTION ............................................................................2-3 FIGURE2-5 UPPERRESEARCH LEVEL...............................................................................2.4 FIGURE2-6: LOWERRESEARCH LEVEL..............................................................................................2-4 FIGURE2-7: LABORATORY AND REACTOR BUILDING WITH PNEUMATIC INTERCONNECTION ..................................2-5 FIGURE2-8: AVERAGE WINDFREQUENCY DISTRIBUTION .............................................................................. 2-7 FIGURE2-9: EARTHQUAKE PROBABILITY MAP.....................................................................................2-9 FIGURE4-1: FOUR-ELEMENT BUNDLE ..........................................................................................................4-2 FIGURE4-2 THREE-ELEMENT BUNDLE WITH FUELFOLLOWED CONTROL-ROD ...................................................4-3 FIGURE4-3: FUELFOLLOWER INSTALLATION B ~ O .................................................................................

M 4-3 FIGURE4.4 . FUELFOLLOWER INSTALLATION TOP................................................................................... 4-3 FIGURE4.5 . CONTROL RODBUNDLE WITH GUIDE TUBE..................................................................................... 4-4 FIGURE4-6: DETAILED DRAWING OFFLIPFUELROD ......................................................................................4.5 FIGURE4-7: INTEGRATED FILAMENT THERMOCOUPLE FUELROD................................................................. 4-5 FIGURE4-8: CONTROL RODDETAILED .......................................................................................................... 4-6 FIGURE4-9: FUELBUNDLE WITHGUIDETUBE............................................................................................ 4-6 FIGURE4.10 . COREGROD PLATE................................................................................................... 4-7 FIGURE4-1 1: CORECONFIGURATION VIII-A ............................................................................................. 4-8 FIGURE4-12: FLIP PULSING CHARACTERISTICS ........................................................................................4.10 FIGURE4-13' INVERSE PERIOD AND INVERSE WIDTHA T HALFMAXIMUM POWERVERSUS PROMPTREACTIVITY INSERTION ........................................................................................................................................... 4-11 FIGURE4.14 . FULLWIDTHAT HALFPEAKPOWER VERSUS PERIOD .................................................................... 4-11 FIGURE4-15: PEAKPOWERVERSUS FULLWIDTHAT HALFPEAKPOWER ...................................................... 4-11 FIGURE4-16: PEAKPOWER vs .REACTOR PERIOD ........................................................................................ 4-12 FIGURE4-17: PEAKPULSEPOWERvs . PROMPT REACTIVITY SQUARED ............................................................4-12 FIGURE4-18: NSCR CORE111-A ................................................................................................................... 4-12 FIGURE4- 19: EXPERLMENT VERSUS CALCULATED FLUXUSINGEXTERMINATOR-2............................................ 4-13 FIGURE4.20 . EXPERIMENTAL VS . CALCULATED PULSE TEMPERATURES .................................................................. 4-14 FIGURE4-2 1:CORE111-A PULSEDATAIF POSITION E5 (SE) ................................................................................ 4-14 FIGURE4-22: TEMPERATURE COEFFICIENTS OF TRIGA FUELS .............................................................................. 4-15 FIGURE4-23: COMPARISON OFFLIP TO STD. TRIGA PULSEFORSIMILAR REA~IVIT INSERTIONS Y ................. 4-15 FIGURE4-24: PROMPT NEGATIVE TEMPERATURE COEFFICIENT TRIGA-LEU 20-20 FUEL................................... 4-16 FIGURE4-25: COMPARISON OFFLIP ANDLEU PULSES(BOL) ......................................................................... 4-16 FIGURE4-26: PULSING TEMPERATURES FOR LEU AND (BOL)...................................................................... 4-17 FIGURE4-27: COMPARISON OF^ AND LEU PULSES FOR$1.80 INSERTION (EOL) ............................................ 4-17 FIGURE4-28: PULSING TEMPERATURES FOR $1.80 INSERTION (EOL) .................................................................. 4-18 FIGURE4-29: PROMFTNEGATIVE TEMPERATURECOEFFICIENTTRIG/! LEU JTJEL4-ROD CLUSTER ........................ 4-19 FIGURE4-30: PEAKPULSING POWER...................................................................................................................... 4-20 FIGURE4-3 1: PULSING ENERGY ........................................................................................................................ 4-20 FIGURE4-32: FULLWIDTHAT HALFPEAK POWER ................................................................................................ 4-21 FIGURE4-33: PEAKTEMPERATURES ..................................................................................................................... 4-21 FIGURE4.34 . PEAKTO THERMOCOUPLE TEMPERATURE RISE RATIO ...................................................................... 4-22 FIGURE4-35: NOMINAL FUELRODSPACING IN THE NSCR CORE........................................................................... 4-23 FIGURE5-1: POOLWATERSYSTEMS .................................................................................................................... 5-1 FIGURE5-2: NSCR POOLWATERELEVATION ......................................................................................................5-2 FIGURE5-3: REACTOR POOLSECTIONS AND PENETRATIONS ............................................................................. 5-3 FIGURE5-4: REACTOR POOLCOOLING SYSTEM ..................................................................................................... 5-4 FIGURE5-5: WATERPURIFICATION AND DISPOSAL SYSTEM ................................................................................5-6 FIGURE5-6 SKIMMER SYSTEM ............................................................................................................ 5-7 FIGURE7-1' LOGPOWER CHANNEL .................................................................................................... 7-2

FIGURE 7-2: WIDE-RANGE LINEAR DRAWER . . .. . . .... ... .. .. ... . .... . . .. .... .. .. ..-... . ... . .. ... ... . . . ......... .....7-3 FIGURE 7-3 FUELTEMPERATURE CHANNEL . . . . . .. . ... . .. . . ........ . . . .... . .. . . . ... .. .. . . . . ... . . .. . ...7-5 FIGURE 7-4 Stllhl SAFETY ARhlATURE A N D DAMPENLVG DEVICE ASSEMBLY. . . .. . . . ... .. .. ... . . . .. . .... .... 7-7 FIGURE 7-5. CONTROL RODASSEMBLY SUPPORT MECHANISM .. . .. . . . . ... . .. . . . .. . . . .. . . . .... . ... .. .. . . ... .7-8 FIGURE 7-6. CONTROL ROD~NSTALLATION. . . . . . . .. . .. . . .......... ... ..... .... . .. . . . . .... .. .. . ... ....... .. . . . . 9 FIGURE 7-7: CONTROL RODDRNEMECHANISM FOR SHIM SAFETY CONTROL RODS. . ... .. ....... . . . ... . .. . . . . .7-11 FIGURE7-8: TRANSIENT RODDRIVE... ....... .. . . .. . . .. . .. ....... . ... . ........ .. . . .. . . . .. ...... . .. . . . . . . ........ ...7-13 FIGURE 8-1: NSCR ELECTRICAL D~STRIBUT~ON. .. . .. .. ... . . .. . . .... . . ... ....... .. . ..... . . . . .. .... . ... . . . .... ... .....8-2 FIGURE 10-1: POOLEXPERIMENTFAC~L~T~ES ... . .... ... . ...... ... .... . ... . .... . . .... .... ... .. ....... ... ... . . . .. ......... 10-1 FIGURE 10-2: STALLBEAMPORT~NSTALLAT~ONS WITH BISMUTH TROUGH.. ................................................. 10-2 FIGURE 10-3. STALLBEAMPORTINSTALLATIONS WITH GRAPHITE COUPLER BOX AND THERMAL COLUMN EXTENSION ..... ...... . . ......... . . ....... .. .... ....... . .. . . . ... . .... ........ .. . ... .. .... . ...... .... .. .... ....-.............. .. 10-2 FIGURE 10-4: PNEUMATIC SYSTEM .. .... .. ..... . . ....... .............. ..............-............ ............. ........... .....................10-5 FIGURE 10-5: IRRADIATION CELL... ................. . .......................... ......................... ... ................. . . ..................... 10-6 FIGURE 10-6: BP-~/~ADIOGRAPHY CAVE. .. ... . . . .......... . ......... . . . ...................-.-.. . ...... .. ..... . . ... ............. 10-7 FIGURE 12-1: ORGAN~ZAT~ON CHART FOR REACTOR ADMINISTRATION ........................ ................. ............ ........ 12-1 FIGURE 13-1: STEADY STATEFUELTEMPERATURE AS A FUNCTION OF POWER GENERATION ....... . ......................13-1 FIGURE 13-2: PULSETO PRODUCE %O0C PEAKTEMPERATURE .................... .. . .. ........-............. . ...... ..... . .......13-3 FIGURE 13-3. TRANSIENT RODINTEGRAL RODWORTH........................................ ............................ . ... ............ 13-4 FIGURE 13-4. TIMEDEPENDENT PULSEREACTIVITY INSERTION USEDTO OBTAINRAMPTABLE........ . ...... . ....... 13-5 FIGURE 13-5. TMEDEPENDENT REACTIVITY INSERTION USEDTO GENERATE SCRAM TABLE.................... ......... 13-5 FIGURE 13-6: TEMPERATURE COEFFICIENTS 01;TRIGA FUELS................... . . ....... . . ......... ....... ....................... 13-6 FIGURE 13-7: INTEGRAL TEMPERATURE COE~CIEN ....T.. ................. ....... ...................................................... 13-7 FIGURE 13-8: STRENGTH AND APPLIED STRESSAS A FUNCTION OF TEMPERATURE FOR 1.7 AND 1.6 H-ZRTRIGA FUEL............... . . .....,. ........................................................ ......... . ............................................................ 13-10 FIGURE13-9: DECAY HEATPOWERGENERATION FOLLOWING LOSS OF COOLANT FOR INFINITE R E A ~ R OPERAT~ON AND PERIODIC REACTOR OPERATION ........... ..................................... ........................................ 13-11 vii

LIST OF TABLES TABLE1.1:

SUMMARY

OF REACTOR DATA....................................................................................... 1-2 TABLE2- 1: POPULATIONDISTRTBUT~ON lN THE NSC VlCINlTY ..........................................................2-5 TABLE4-1 :

SUMMARY

OFREACTOR DATA..............................................................;................... 4-1 TABLE4-2: PRINCPALFUELELEMENT DESIGN PARAMETERS ...............................................................4-4 TABLE4-3: NSCR STANDARD TRIGA CORECHARACTERISTICS ..................................................................4-10 TABLE4-4 COMPARISON OF MEASURED AND CALCULATED VALUES OF K, ............................................... .4-13 TABLE4-5 OPERATING CHARACTERISTICS OF NSCR CORE111-A........................................................................4- 14 TABLE4-6. COREPARAMETERS OF GA TESTCASE................................................................................4-18 TABLE4-7: PEAKING FACTORS FROM GA TESTCASE...............................................................................4-19 TABLE4-8. ESTIMATED PEAKTHERMALFLUX AT 2 MW 4-RODCLUSTER TRIGA-LEU FUEL........................4-19 TABLE4-9 NSCR LEU COREPROPERTIES .............................................................................................. 4-22 TABLE7-1: MMIMUM REACTOR SAFETYCHANNELS .......................................................................................... 7-17 TABLE7-2:

SUMMARY

OFINFORMATION DISPLAYED AND RECORDED ON REACTOR CONSOLE .............................7-19 TABLE7-3:

SUMMARY

OF ALARMS DISPLAYED ON REACTOR CONSOLE.......................................................... 7-20 TABLE13-1:

SUMMARY

OF RADIATION EXPOSURES FOLLOWING CLADDING FAILURE OF THE HIGHEST POWER DENSITY FLIP FUELELEMENT ...................................................................................................13-2 TABLE13-2: BLOOST RESULTS FOR PULSEFROM 1 MW ........................................................................ 13-8 TABLE13-3 BLOOST RESULTS FOR PULSEFROM 300 W ................................................................................ 13-8 TABLE13 PULSEVALUES TO EXCEED TEMPERATURE ................................................................................. 13-8 TABLE13-5: FUELBUNDLE ROTATION STUDYFOR MAXI~IUMPOWER AND MAXIMUM TEMPERATURE ............. 13-12

1 THE FACILITY 1.1 Introduction The Texas Englneering Experiment Station owns and operates the nuclear reactor fachty located at Texas A&M University named the "Nuclear Sclence Center" The faclllty is a university-operated research reactor designed to prov~dea center for the university's students in vanous disciplines and for outside research and commercial users.

The fachty, located on the West end of Texas A&M Univers~tycampus In College Stat~on,Texas, houses a TRIGA (Teach~ng,Research, and Isotopes, General Atomic) reactor utilizing FLIP. LEU and/or Standard TRIGA fuel with a maximum operating power level of 1.0 MW.

Pr~nclpaland inherent safety features include passive shutdown (SCRAM) capablllty and negative temperature/power feedback. This Safety Analysis Report contains documentation and basic information as well as conslderatlons to support the conclusion that the Nuclear Science Center (NSC) can operate safely. This document supports the renewal of NRC Llcense R-83.

1.2 Summary and Conclusions on Principal Safety Considerations The primary safety features of a TRIGA-type reactor are from the use of a fuel with a strong negative prompt temperature coefficient of reactivity, which l~mltsthe excursions from reactivity insert~ons,thus preventing fuel damage from credible react~vityaccidents. Ejectlon of the transient rod from the core, when the core 1s operating at the power-level scram point will result in no fuel damage. Smce experiments are limlted to less reactiv~tyworth than the transient rod, experiment failure cannot cause a more severe accident The operating power level of 1 M W (Ilmited to 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> per week) results in decay heat potential in the fuel small enough that a loss of reactor coolant does not result In fuel damage or release of fission products The only case where significant exposure occurs requires the simultaneous failure of the fuel element cladding, catastrooh~cfailure of the ~ o o andl liner and a failure of the ventilation svstem with ~ersonnelremaining within the reactor kclllty for a perioi of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after release. This would result in i U Thus, no realistic hazard of consequence will result from the Design Basis Accident.

1.3 General Description of the Facility The Texas A&M Nuclear Science Center houses the TRIGA pool-type reactor in a dedicated buildrng off the main part of campus. F ~ g u r e1-1 is a photograph of the front of the Reactor Building and adjacent Laboratory Bu~lding.

Figure 1-1: Nuclear Science Center Reactor a n d Laboratory Buildings

The reactor IS a 1-MW pool-type nuclear reactor The current configuration uses 4-element bundles with TRIGA-FLIP fuel Llght water flows through the reactor by natural convection Graphlte serves as a reflector.

A suspension frame supported by a brldge that spans the pool supports a grid block which in turn supports the fuel, reflector, control rods, samples and any other ln-core material Four shlm-safety control rods, a translent control rod and regulat~ngrod control reactivity. Table 1-1 is a summary of reactor data Table 1-1: Summary of Reactor Data Responsible Organ~zation I Texas Engineering Experiment Statlon Location 1 College Station, TX I Purpose I Teaching and Research Fuel Type TRIGA-Standard, TRIGA FLIP, and TRIGA LEU Number of elements (nominal) 94 (Including Fuel-Followed Control Rods) control I Safety Elements ( 4 Shlm-Safety Control Rods Regulatmg Element ( 1 Servo-controlled Control Rod Transient Control I 1 Pneumatic Operated Control Rod 1.4 Shared Facilities and Equipment The Nuclear Sclence Center shares utilltles wlth an accelerator facility that is part of the Nuclear Engineering Department. This accelerator shares a bullding with Nuclear Sclence Center auxiliary shops and thus shares a wall and electrical distribution system. This building is external to the confinement building. The Nuclear Science Center shares no other facilities or equipment.

1.5 Comparison with Similar Facilities The reactor has operated with a full FLIP core or FLIPJStandard mixed core since 1973. Thls provides the greatest operational history with wh~chto compare.

The University of Wisconsin and Washington State University operate I-MW reactors using TRIGA FLIP and Standard fuel Unlike the Texas A&M reactor, General Electric built both of these reactor pools and so the pool sizes and configurations differ from between the three reactors Nevertheless, the reactor behavior and accident analysis between the reactors is simllar.

1.6 Summary of Operation The NSCR provides the following:

Laboratory exercises for undergrad and graduate students at TAMU, Neutron activation analysis facilities for numerous departments at TAMU, Neutron activation analysis facilities for educational institutions without a research reactor, A source of radioisotopes for various research and educational projects, Radioisotopes for the medical mdustry, Radioisotopes for several commercial organizations, and A neutron radiography facility for research and commercial use 1.7 Compliance with the Nuclear Waste Policy Act of 1982 In accordance with a letter from the U.S. Department of Energy (R. L. Morgan) to the U.S Nuclear Regulatory Commission (H. Denton) dated May 3, 1983, it has been determined that all universities operating non-power reactors have entered into a contract with the Department of Energy (DOE). The contract provides that DOE retain title to the fuel and DOE 1s obligated to take the spent fuel and/or high-level waste for storage or reprocessing.

Because Texas A&M Unlverslty has entered Into such a contract wlth DOE, the Texas A&M Nuclear Sc~ence Center has satisfied the applicable requirements of the Nuclear Waste Pollcy Act of 1982.

1.8 Facility Modifications and History The init~alplanning for the NSCR began in 1957. At that tlme, the University was embarking on a program of expandmg graduate education and research programs Texas A&M Adminlstrat~onrecognized that a research reactor that would be able to serve many departments and support a large var~etyof research actlvit~eswould s~gnificantlycontribute to thls development The appllcat~onfor a construction permlt and operating llcense was submitted in March 1958 along with the Hazard Summary Report. Supplement I to the Hazard Summary Report was submitted in 1959. The construction permit, Number CPRR-38, was issued in August 1959. This perrmt was converted to operatmg llcense R-83 which authorized operat~onof a MTR swimmmg-pool type reactor at 100 kW.

The reactor first went critical on December 18, 1961. Smce that tlme, the use of the fac~lityhas increased steadlly to 11scurrent posit~onof supporting an act~venuclear engmeering educational program and various other research endeavors. The facility serves many campus departments, other universities and colleges, several city and state agencies, and other industrial and research organlzatlons By January 1965, the use of the facility had ~ncreasedto necessitate the operation of a two-shlft bass three days per week and one shift operation for two days After July 1966, the reactor routinely operated two shlfts for five days per week In 1968, the reactor was converted to TRIGA fuel and the power level was increased to 1,000 kW. Only three years had elapsed from lnitlal reactor operations before a comprehensive upgradmg program was implemented In December 1965, proposals were submitted to the Natlonal Science Foundation (NSF) and the Atomic Energy Commission (AEC) for funds to support a long-range expansion program.

The expansion of the faclllty included four separate phases and are described briefly below:

Phase I: Pool Mod~ficatlonand Liner The large reactor pool was mod~fiedby installing a multipurpose irradiation cell. This faclllty allows exposure of large anlmals or other objects to the radiation from the reactor core. A permanent stainless steel liner was installed as part of the pool mod~ficatlonto ehminate problems of pool leakage, a source of previous significant operational problems.

Phase 11: Coolmg System To allow steady-state operation at power levels up to 1.0 m V , a cooling system was provrded for the reactor. The 10 MW reactor power improved a number of existing research programs and encouraged the initiation of new projects.

Phase 111: Conversion of the Reactor Core The reactor core was converted to employ standard TRIGA fuel elements, and on July 31, 1968, an amended facillty llcense allowed the NSCR to be operated at a maximum steady-state power level of 1,000 kW and pulsing up to $3.00 reactivity insertion The inherent safety of the TRIGA fuel allowed increased flexibility and utlllzatlon of the reactor Pulsing was possible due to the prompt negative temperature coefficient of reactivity and the integrity of TRIGA fuel at pulse peak temperatures Phase IV: Laboratory Building The original research space w~thinthe Nuclear Science Center confinement building was quite limited. A laboratory buildlng was constructed to adequately accommodate the increased research load and to allow for anticipated expansion of programs

From mtiatlon, the plan covered a perlod of 3 !h years to completion In mld-1969. The plan not only changed the lnltlal facllity physlcal plant but also established a new reactor program Operatmg experience wlth standard TRIGA fuel revealed a high fuel burnup rate resulting in fuel additions to maintam sufficient reactlvlty. Modifying the reactor grid plate In late 1970 extended Core hfe by approximately 1 %

years and provlded for the installatlon of fuel-follo\ved control rods Subsequent operation, however, eventually required the addition to the core of all standard TRIGA fuel on hand, which led to dramatically reduced fluxes avahble for experiments. The solutlon to thls problem was the lnitlatlon of a program to provide a core loadmg utilmng TRIGA FLIP (Fuel Llfetlme Improvement Program) fuel. In June 1973, the NSCR licensed to operate full standard, mixed, or full FLIP TRIGA cores. The mixed cores licensed to operate at a maximum steady-state power of 1,000 kW with maximum pulse reactivity msertions of $2.00. In July 1973, the first NSCR mixed TRIGA core, containing 35 FLIP and 63 standard elements, went into service (Core 111) In July 1975, the maximum permissible pulse reactivity lnsertlon increased to $2.70 On September 27, 1976, during a loading operation, four "lead" elements failed to pass through a "golno-go" gauge.

Steady state hydrogen migration followed by rapid hydrogen pressurization during reactor pulses caused the damage (GA-A16613, 1981) The reactor was not pulsed agaln untll a complete analysis was submitted to the NRC In 1981.

The maxlmum pulse allowed by the NSCR reduced to that amount wh~chwould not cause the reactor to exceed a temperature lmlt of 830°C (1525°F).

The Nuclear Science Center Reactor (NSCR) operated from 1962 untll 1967 w~thMTR-type curved aluminum plate elements. During thls tlme, the reactor operated extensively at a maximum power level at 100 kW. In 1968, the reactor began using TRIGA fuel at power level of 1,000 k ~ . ' .The ~ initla1 core loading was quite satisfactory, but fuel burnup and samarium bulldup soon affected experimental ~ a ~ a b l l lTo t ~restore

.~ excess reactivity, the NSC period~callyadded addltlonal fuel to the core and graph~tereflectors to all core faces This eventually led to a 126-element core with a resultant decrease In the flux of almost 40% and the ellmination of most of the irradiation facilities.

In August 1970, the installatlon of fuel followed control rods lead to a gain In excess reactivity and helped solve the problem of maintaining excess reactivity. This installatlon required modification of the grid plate to allow passage of the fueled portion of the control rod through the gr~dplate.4 Speciftcally this modification achieved an average

$1.10 increase per fueled follower, whlch extended the core life nearly two years The high fuel burnup rate of standard TRIGA cores continued to be an operational problem for the NSCR The NSCR has operated approximately 100 MW-days per year slnce 1969.

It was obvious that a solution was needed that could fit within the constraints of a university budget and limited federal support. Replacement of the core w~thnew fuel would have lead to considerable expense with a very short effective hfe of a standard core. Since the average core-burnup was only 10% and a reasonable amount of fuel would only provide small reactivity Increases, Cycling new fuel into the core was no more attractive. The solution to the problem was in a new fuel developed and marketed by General Atomic It is almost identical to the standard J.D. Randall, "Power Uprating Experience Following Conversion of a Pool Reactor From Plate-Type to TRIGA Fuel Elements," Nuclear Safety, Vol. 10, No. 6, December 1969.

• w.B. Wilson, et a]., The Installation and Operating Characteristics of the Texas A&M University Reactor, Technical Report 23, Nuclear Science Center, Texas A&M University, August 1969.

3 ~ . Schad~ . & J.D. Randall, "Operatlonal Reactivity Considerations of the Texas A&M TRIGA," TRIGA Seminar, Denver, Colorado, 1970.

4 ~ . EFeltz,

. P.M. Mason, J.D. Randall, Modificatlon of a BSR-MTR Type Grid Plate to Accept Fueled Followers,"

Conference on Reactor Operating Experience, American Nuclear Society, Denver, Colorado, August 8-11, 197 1.

5 ~ Foushee,

. J R. Shoptaugh, G B. West, W.L Whlttemore, "TRIGA FLIP-A Unique Long-Lived Verslon of the TRIGA Reactor," Trans. Amer. Nucl Soc., Vol. 14, No. 2, Miami Beach, October 1971.

TRIGA fuel except that the enrichment was 70% rather than 20%. The hydrogen to zlrconlum ratlo decreased from approximately 1.7 to 1.6, and 1.5-ive~ghtpercent natural erblum was added as a burnable polson The fuel designated as FLIP (Fuel Life Improvement Program) has a calculated lifetime of approximately 9 MW-years Thls contrasts wlth experience for a standard core, where r t was posslble to operate only six months (approximately one-seventh of a MW-year) w~thouta file1 add~tion.

Inasmuch as funds for a complete FLIP core were not ava~lable,~t\vas necessary to conslder operat~onwith a core comprised of a mixture of FLIP and standard TRIGA fuel A precedent for this had been established by General Atomic when they operated a standard core loaded wlth e~ghteencentrally located FLIP elements in a fuel test program.6 Calculations at Texas A&M led to the conclusron that satisfactory core arrangements were possible with a mlxed core.7 As funds became avarlable, the amount of FLIP fuel could increase until the core was completely FLIP fuel. This concept provides the additional satisfaction of producing substantially greater bumup in the standard fuel used in the mixed core.

Sufficient funds provided for a partlal loadlng of FLIP fuel in a 98-element core with a 35-element FLIP region.

Thls configuration ach~evedcrltlcality in July 1 9 7 3 . ~The burnup data indicated that the burnup rate was initially 0.5$ per MW-day and after samarium buildup the rate dropped to 0.2$ per MW-day. Thus, the lncorporatlon of FLIP fuel had increased the lifetime of the core by a factor of three.

The NSCR has operated w t h two mlxed core loadlngs contamng 35 FLIP and 5 9 FLIP, elements each since lnit~al approval in June 1973 Smce the late 1970, the core has operated with all FLIP fuel.

~ . and J.R. Shoptaugh, "Experimental Results From Tests of 18 TRIGA FLIP Fuel Elements in the Torrey 6 ~ . West Pines Mark F. Reactor," GA-9350, September 1969.

~ . M. Hardt & J.D Randall, "Feasibility Studles of a Mlxed Core Using Standard TRIGA and FLIP Fuel,"

7 ~ . Feltz, TRIGA Owners Conf. 11, College Statlon, 1972.

~ . T.A Godsey, M. Hardt & J.D Randall, "Performance Testlng of a Mixed Core Utilizing TRIGA-8 ~ .Feltz, Standard and TRIGA FLIP Fuel," Trans. of Amer. Nucl Soc., Vol 17-1, Myrtle Beach, August 1973.

2 SITE CHARACTERISTICS This chapter covers the geographical, geological, se~srnolog~cal, hydrological and meteorolopical characteristics of the NSCR site and its vicmty 2.1 Geography and Demography 2.1.1 Site Location and Description 2.1.1.1 Specification and Location The Texas A&M Nuclear Sclence Center (NSC) 1s on a rectangular SIX-acresite on the Texas A&M University campus 1,500 feet from the North-South runway of Easterwood Airport. Figure 2-1 shows the relationship between the Nuclear Science Center and the Easterwood Airport runways. The facillty 1s SIX miles south of the clty-center of Bryan (pop 65,660). three miles southwest of the maln campus of Texas A&M and two and one-half mdes west-southwest of the city of College Statlon (pop 67,890) In Brnzos County, Texas The facillty location is N30°35',

W96O23'.

Figure 2-1: NSC and Easterwood Airport Land owned and controlled by Texas A&M University and Easterwood Airport surrounds the slte. A chain-link steel fence that provides reasonable restriction of access to the slte defines the indemnity confines of the site. The main entrance into the site is through an electrically operated chain-link steel gate at the east end of the site. The entire area inside the perimeter fence of the NSC 1s a "Restricted Area". Located within the boundaries of the site are the reactor confinement building, reception room, laboratory building, mechanical equipment room, cooling system equipment, holding tanks and other storage and support buildings.

2.1.1.2 Boundary and Zone Area Maps The Nuclear Science Center Site (Figure 2-2) defines the operation boundary for the NSC. In addit~onto th e NSC.

the site contalns a linear accelerator and associated laboratory.

A map of the area surroundmg the NSC (Figure 2-3) shows the major roads in the area up to a distance of 8 km from the NSC The figure includes the area for the Texas A&M University campus as well as College Station and most of Bryan. Figure 2-4, Flgure 2-5, Figure 2-6 and Figure 2-7 show the floor plans for the Reactor Buildmg a nd Laboratory Building.

Figure 2-3: Major Roads around NSC and Eastenvood

2.1.2 Population Distribution Table 2-1 shows est~matedpopulation maximums. The dormitories for Texas A&M are greater than 4 km and less than 6 km from the NSC. Therefore. the estimated population for 6 km wlll increase by approximately 10,000 during the fall and sprlng semesters No residences exist or are likely within 1 km of the NSC as this property is owned by Texas A&M and Easterwood Airport A firefighter training school and a few other small facilities, within 1 km, employ dozens of people. As many as a few hundred firefighters attend training a few weeks each year. The area within 2 km of the NSC is mostly owned by Texas A&M and houses Easterwood Airport. The population in this area is low and quite stable.

The area wlthln 8 krn lncludes much of College Station, part of Bryan and surroundmg areas. The population of the 8 km area may continue to increase.

Table 2-1: Population Distribution in the NSC Vicinity 1 Dlstance from Facility (kilometers) 1 Estimated 2000 population I 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2.1 Locations and Routes No industrial facilities are near the NSC.

The nearest ratlroad runs through the maln campus and comes ro \v~thln3.5 km of the NSC.

The NSC 1s located approxmately 1.5 km West of FM 2818 and 2 km South of Hlghway 60 Highway 6 is approximately 7 km East of the NSC. There 1s no Interstate highmay in the area. Refer to Flgure 2-3 Easterwood Alrport is in the mmedlate vlcinlty of the NSC The nearest runway 1s 300 meters from the NSC at its closest point The NSC 1s approximately 700 meters from the prlvate and non-passenger commerclal termlnal and is over 1 km from he main termlnal. Refer to Flgure 2-1 There are no milltary facilltles In BryanlCollege Station area wlth the exception of the National Guard Facility 2.2.2 Air Traffic Easterwood Airport, the only commercial airport near the NSC, is immediately adjacent to the slte. Although one of the three runways is close to the NSC, none have trajectories that take commerclal traffic dlrectly over the reactor.

Although commercial, prlvate and mhtary trainlng flights use Easterwood Airport, arrival frequency 1s low and the local control tower rarely places inbound traffic 1n holdmg patterns 2.2.3 Analysis of Potential Accidents at Faciliti'es No ~ndustnal,transportation or mllitary facilities within the vlcinlty of the NSC pose sufficient risk to the reactor to render the site unusable for operation of the reactor facllity Although an alrport is nearby, the construction of the NSC (the reactor is below the surface of the ground and protected by thlck pool walls) and the trajectory of the runways make the magnitude of a casualty resultmg from an alrcraft colllslon and the probability of such an event low.

2.3 Meteorology 2.3.1 General and Local Climate The BryadCollege Station area is located approximately 160 kilometers (100 miles) inland from the Texas Gulf Coast Largely, the high-pressure areas that are predominant over the Gulf of Mexico determine the local weather.

As a result, warm southeasterly winds occur a large majority of the time on an annual basis (Figure 2-8). Average annual rainfall is between 7 6 and 89 cm (30-35 inches). Snow occurs only rarely and temperature reaches sub-freezing infrequently for brief periods during the winter. Northwest winds normally accompany the passage of the frontal systems Calms occur an average of 10% of the time, and wind speeds above 3 8 kph (21 knots) rarely occur.

-SPEED-KNOTS 11 11-21 >21 TOTALS CALM 10.2 KNOTS 75.0Oto KNOTS 14.3*lo KNOTS 0 5%

Figure 2-8: Average Wind Frequency Distribution Tornadoes are common in Texas. Data on tornado frequency between 1950 and 2002 indicates fifteen tornadoes in Brazos County Texas during that fifty-two year period Sixty percent of those occurred in May and over half occurred in the afternoon between 2:00 and 6 00 pm The season usually starts in March and reaches its peak in May.

l ~ a t i o n a Oceanic l and Atmospher~cAdministration, National Climatic Data Center, http:/lwf.ncdc.noaa gov/oalncdc.html

The reactor build~npdeslgn requires that it wthstand 207 P a (30 psi) over pressure wlth the exception of the domed roof. It w~llw~thstandonly 2344 Pa (50 psf or 0 34 psi) In the case of a tornado passlng nearby, the roof would probably act as a pressure rellef mechanism The b a s ~ csteel structure in the roof would probably remaln Intact unless a tornado made d~rectcontact on the bu~ldlng.The reactor build~ngdes~gnrequlres that 11 wthstand a straight 145 Kph (90 mph) wind The remforced concrete construction and round shape of the buildmg provlde a considerable strength to wthstand high winds.

In the event of a tornado wthin an 8 km (five mile) radlus of TAMU, the r a d ~ ooperator at the TAMU Communicatlons Center will notlfy the NSC or the first ava~lableperson on the NSC emergency notification roster.

The r a d ~ oroom recelves not~ficationof tornadoes from both the TAMU weather radar and the Brazos County, BryanICollege Statlon Disaster Emergency Plannmg Organlzatlon. The method of tornado detection is by TAMU radar, area spotters and the Nat~onalWeather Service (NWS) 2.4 Hydrology Drainage of the slte IS by way of White Creek to the Brazos Rlver three mdes to the southwest. The facility IS on high ground and the entire area drains well via a number of tr~butarlesof White Creek Based on hlstory, the slte, which IS approx~mately92 meters (304 feet) above sea level, IS not In flood area The highest recorded crest on the Brazos River at Bryan (December 1913) was 16 meters (54 feet) above flood stage or 75 meters (246 feet) above sea level.

The probabll~tyof contaminating dr~nklngwater supphes is low since the Brazos Rwer is not a source of water and there are no open reservoirs in the surroundmg area. The publlc water supply comes from deep wells several miles from the Nuclear Science Center.

Ground water should not present any problems. The NSC is on a formation known as the Easterwood Shale. The thickness of the formation is between 3 and 90 meters (10 and 300 feet). The bulldmgs in College Station and those on the campus have this shale as a foundat~on.The shallowest aquifer is the Bryan Sandstone, whlch underhes the Easterwood Shale. It is well below the depths requ~redfor building excavation 2.5 Geology, Seismology, and Geotechnical Engineering 2.5.1 Regional Geology Texas lies on the North American tectonic plate, several hundred miles from the nearest edge. In addition, there are no active faults in Texas.

2.5.2 Seismicity Texas lies in a region of minor seismic activity. Extreme west Texas, over 965 lulometers (600 miles) west of College Station, is the closest to the active belt along the west coast of Mexico and the United States. There are occasional minor shocks of very small magnitude in the state. There IS only record of one earthquake of any significance in Texas; this shock was at 30.6 N and 104.2 W on August 16, 1931, near El Paso In extreme west Texas and was a Class C (6 4 in magmtude) shock.

Reinforced concrete structures provide good protection against earthquakes. The heavlly reinforced NSC wall structure and reactor pool walls would wthstand any minor shock that may occur.

2.5.3 Maximum Earthquake Potential F~gure2-9 shows the relative probabhty for se~smicact~vityIn the Un~tedStates Central Texas has the lowest hazard Earthquake potential at the NSC is unhkely.

Figure 2-9: Earthquake Probability Map2 Probabilistic S e i s m ~ Hazard c Maps of Alaska by Robert L. Wesson, Arthur D. Frankel, Charles S.

Mueller, and Stephen C. Harmsen, U.S. Geological Survey Open-Fde Report 99-36,43 p.

3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Design Criteria The prlmary design criterla for the safe operatlon of a TRIGA reactor 1s that the faclllty 1s able to wtthstand any credlble accldent wlth negllglble hazard to the public, wlthout relying on actlve safety systems. TRIGA FLIP, Standard and LEU fuel with stamless steel claddlng meets thls crlterlon TRIGA Standard and FLIP fuel exhlbit a prompt negattve temperature coefficient responsible for reactor shutdown for all credlble temperature excursions Many references not specifically llsted here document the characterlstics of TRIGA fuel. Chapter 13 deals in detail with the most credlble loss of coolant accident. The fuel and cladding construction, rather than structure or control systems, meet the design criteria that the Texas A&M Nuclear Science Center Reactor (NSCR) can withstand credible accidents with negligible hazard to the public.

The building that houses the NSCR was completed in 1958. The criteria for the deslgn allowed control of the airflow into and out of the NSC through ground level suctions and an exhaust stack of a specific height Sect~on1.8 of this report describes the hlstory of modlficatlons to the NSC Among these are the additions of a larger 1 MW cooling system to accommodate higher power levels and llning the reactor pool with stainless steel to reduce the loss through leakage 3.2 Meteorological Damage The accident analysis In Chapter 13 does not use crlterla for the condltlon of the buildlng or equipment for accldent analysls. Therefore, there are no crlterla for safe operation based on possible meteorological damage. The NSC building has stood since 1959 without suffering meteorological damage. The budding meets all local codes for withstandlng meteorolog~caldamage 3.3 M7aterDamage The accident analysls In Chapter 13 does not use criteria for the condltlon of the buildlng or equipment for accident analysis and Sectlon 2.4 addresses hydrology. There are no criterla for safe operatlon based on possible water damage. The NSC bulldtng is equlpped w ~ t ha sump pumping system. This keeps the lower level of the building dry durlng heavy rams In any case, if the sump system fails and the lower level becomes flooded, water damage cannot affect the safe operation of the NSCR The buildlng meets all local codes for withstandlng floodmg.

3.4 Seismic Damage The accident analysis in Chapter 13 does not use criteria for the condltion of the buildlng or equipment for accident analysis and section 2.5 addresses seismic activity for the region. Therefore, there are no criteria for safe operation based on seismology. This SAR only assumes the fuel and claddmg are operable and intact for accident mitigation.

3.5 Systems and Components The accident analysis in Chapter 13 does not use criteria for the condltion of the systems, components or other equipment for accident analysis. Therefore, there are no criteria for safe operation based on systems or components.

Thls SAR only assumes the fuel and cladding are operable and intact for accident mitigation

4 REACTOR DESCRIPTION 4.1 Summary Description 4.1.1 Introduction The NSCR IS a pool-type TRIGA reactor wlth Standard TRIGA, TRIGA-LEU and/or TRIGA-FLIP fuel. The NSC has used 70%-enriched FLIP fuel srnce the 1970s but may use 20% emrched or a combmation of the three types.

Pool water cools the reactor vla natural convection and serves as a brological shreld and moderator As of 2003, the NSC uses graph~temoderators on two srdes of the core and primarily uses the other two sides for sample irradiation.

The reactor support is a 7x9 grid Each location in the grid supports a fuel bundle with 4 positions for either fuel elements, control rods or other non-fueled elements. The Reactor Bndge, mounted on rails along the top of the pool, supports the frame that in turn supports the reactor grid. When the reactor is critical, the frame rests on the floor of the pool However, when the reactor IS shutdown the bridge can support the reactor and move the reactor to any locatlon along the centerline of the pool In addltlon to supporting the reactor, the grid provrdes a gu~defor the fuel-followed Shim-Safety control rods When fully Inserted the fueled portion extends through guide holes in the gnd. A safety plate below the grid prevents the rods from fallrng out of the core should they become detached Gulde tubes attached to the fuel bundles guide the transient rod and regulating rod 4.1.2 Summary of Reactor Data Table 4-1: Summary of Reactor Data Responsible Organlzatron Texas Engineering Experiment Station Locatlon College Station, Texas A&M University Purpose Teaching, Research and Isotope Production Fuel Type TRIGA FLIP, Standard andfor TRIGA-LEU Control Safety Control Rods Four Fuel-Followed Shlm-safety SCRAM rods Regulating Rod Non-followed Control Non-SCRAM Rod Transient Control Rod Void Followed SCRAM Rod Experiment Facilities Beam Ports Five permanently installed beam ports in reactor pool Pneumat~cTubes Pneumatic receiver various and changing core locations and periphery locations Reactor Materials Fuel-Moderator U-ZrH 2 3 5 Enrichment

~ 70% (FLIP), 20% (Standard) or 20% (LEU 20-20) 4.2 Reactor Core General Atomlc has successfully operated Mark I11 standard fuel elements and FLIP elements in TRIGA cores at steady-state power levels of up to 1.5 MW. The arrangement of fuel in the NSCR is such that the minimum nominal spacing between the fuel rods provides adequate convection cooling of cores up to 2.0 MW. T h ~ spacings and the extra cooling holes at the comers of the bundle enhance core cooling. Cooling of the NSCR is also improved due to the increased depth of pool

4.2.1 Reactor Fuel The fuel bundles are in three or four-element bundles that allowed convers~onto TRIGA fuel w t h the exist~ngg r ~ d The four-element fuel element assembhes of the TRIGA core provide easy passage of c o o l ~ nwater

~ between the eleientsT

hydride homogeneously combined with partially enriched uranlum fuel, provides moderation Table 4-2 shows the physical characterist~csof the three fuel types Table 4-2: Principal Fuel E I

STANDARD Fuel-moderator material U-ZrH U-ZrH U-ZrH Uranium content 8.5 Wt-5% 20 Wt-%

U enrichment 20%

Average Z 3 5 content

~

123 grams 99 grams 35 grams Natural erblum (1.5 Natural erbium (0.59 Burnable poison None weight %) weight 8 )

Cvlmdrlcal

- a Cylindrical Cylmdrical Length of fuel meat 15 inches 15 inches 15 inches Dlameter of fuel 1.37 1 inches 1.37 1 inches 1.37l inches Type 303 SS Type 303 SS Type 304 SS Cladding thickness 0 020 lnches 0 020 lnches 0 020 inches

4.2.2 Control Rods Six motor-drwen control rods (four shim-safety rods, a regulating rod, and a translent rod) control the reactor and prov~deSCRAM and shutdown capability. The shim-safety and translent control rods provide scram capability.

They fall into the core whenever power is lost to a valve solenoid or electromagnets. The regulating rod maintains constant power during steady state operation and does not have SCRAM capabilities. Section 7.3 details these funct~ons.

The shim-safety control rods are fuel followed. Each consists of a fueled reglon and a poison region. The poison region 1s borated graphite wlth the same cladding as the fuel elements. The fueled region is 15" of active fuel-moderator identical to the other fuel elements. When fully inserted, the fueled portion of the control rod extends through the grid plate below the reactor core with the poison section in the core. F ~ g u r e4-8 shows the fuel-followed

control rod. The fuel-followed control rods do not have gu~detubes, as a guide tube \vould limit cool~ngto the fueled Sectlon movlng with the control rod should it bmd The blade attaches to the slde of the tube that houses the control rod extension When the rod drwe unlt is secured to the reactor support structure, there 1s a one-eighth mch clearance between the foot and This clearance permits small thermal expansion of the fuel without vertical restriction.

The transient control rod is void followed. It consists of a poison section and an evacuated section. Thle poison sectlon is borated graphlte with aluminum claddmg The follower portion is also aluminum clad. Figure 4-9 shows a bundle with a gulde tube that keeps the transient rod in place. The guide tube surrounds the rod and has holes for proper cooling.

The regulating rod is water followed (no physical follower sect~on).The poison section of the regulating rod is a BjC powder. The regulating rod uses a guide tube similar to the transient rod

4.2.3 Neutron Moderator and Reflector Reflectors, excluding experiments and experimental facll~t~es, are water or a combmatlon of graphite and water.

Graphlte elements are machmed to fit flush agalnst a machined spacer f i t Into the grid.

4.2.4 Neutron Startup Source The neutron startup source is located in the core to provlde good mult~pllcatlondata on the startup channel The source strength is such that the startup channel wdl change by greater than 2 cps upon the lnsertlon or removal of the source from the core at ~nltialstartup 4.2.5 Core Support Structure A bridge that spans the reactor pool supports the reactor core, the control rod drives, the nuclear lnstrumentatlon detectors, and the dlffuser system. Mounted on four wheels, the bridge travels on rails provlded at the sides of the pool; thus, the reactor can move from one operating position to another along the centerline of the pool The bridge is hand operated and its speed of travel 1s ltmlted due to the large gear ratios involved. The brldge receives electric power, control-circuit wring and compressed alr. A cable that hes In a covered trough, which is parallel to the south wall of the reactor pool, provides slack for the bridge movement Quick disconnect valves are mounted just below the grating on the upper research level at each end of the pool to facilitate water and air connections upon movement of the reactor from the stall positlon to the large pool section or irrad~at~on cell operating posit~on. ~ i e x l b l quick e

disconnect hoses for the d~ffusersystem and the transient rod air allow operatlon at any location In -

N the pool. Quick disconnects for the pneumatic A B C D E F sample transfer system allow operatlon of the system in a llmited area near the East end of the pool.

An adjustable frame on the west side of the bridge called the brldge yoke serves as the mountmg for the reactor suspension system A large crank wheel and jack mechanism mechanically ralses or lowers the yoke and allows approximately a SIX-inch vertlcal adjustment of core positlon A seven-inch I-beam on the yoke frame insures that the reactor frame can support the weight of the translent rod mechanism An aluminum suspension frame supports the reactor grid plate (Flgure 4-10). The suspension frame is a welded structure of three-eighths by two by two inch aluminum angle. The west side of the frame 1s open toward the large section of the pool. Thls angle construction allows unrestricted flow of the coollng water. An aluminum stabilizer frame, bolted to the bottom of the grid plate, prov~desfor vertical support. Stainless steel guides on the bottom of the stabilizer fit between tracks on the pool floor. This allows accurate repositioning of the reactor core, which is essential for numerous experiments The stabilizer also allows lowerlng the core untd it rests on the bottom of the pool. Thls Figure 4-10: Core Grid Plate prevents swaying that could introduce reactivity variations.

One-quarter inch stamless steel plns attaches the aluminum frame to the grid plate on a11 four corners T h ~ gris plate supports the TRlGA fuel elements TRIGA elements are considerably heavler than the aluminum-plate-t fuel elements the grid plate mitially supported The grid plate contams 54 holes arranged In a nine by SIX array to accommodate fuel bundle assembl~es,graph Ite, instruments, and experiment locations A reactor core loading could have several options for location in the gr.id plate. Figure 4-1 1 shows a typlcal core loadmg conta~n~ng 98 elements and graphlte reflectors. In this loading;, the

'A' row of the grld plate 1s ava~lablefor positioning experiments To accommodate a fuel followed control roc3, a one and three-quarter Inch d~ameterclearance hole through the grld plate allows passage of the fueled sectlon ()f the rod (F~gure4-10) Twelve clearance holes are compatible w t h the four-rod TRIGA assembly des~gn.Each hcAe is located at the southwest corner of the four-rod fuel assembly.

A safety plate assembly beneath the reactor g r ~ dplate stops a control rod follower two inches below its normal position should ~tbecome detached from its mountlng

4.3 Reactor Pool The concrete pool structure and the pool water provide sh~eldlngof the reactor The s h ~ e l dcapaclty 1s for a reactor operatmg at 5 MW, w h ~ c hIS well above the current 1 MW TRIGA maxlmum operatmg level The movable reactor br~dgeperm~tsoperatlon of the reactor at any positlon on the pool centerhe, which runs approximately east to west.

The pool has a stall section and a m a n pool section (See Figure 5-3). An aluminum gate can Isolate these sections to allow draming only one sectlon The pool 1s thlrty-three feet deep and eighteen feet wide In the main pool. The stall section is nine feet across and has an 180"-curved surface with a four and one-half foot radius The upper seventeen feet of the pool wall IS standard concrete. The lower portion of the pool wall 1s barites concrete and llght concrete. The Irradiation Cell 1s a shielded structure adjoming the main pool (Figure 5-3). The Irradlatlon Cell may support reactor experiments or serve as pool water storage. An irrad~atlonwindow, located in the shield wall, separates the reactor pool and lrrad~atloncell The reactor can operate any desired dlstance from the window for irradiation of experiments in the cell.

Stainless steel (Type 304) lmes the reactor pool for maxlmum water containment and water pur~ty.The pool walls are 10 gauge and the floor is one-quarter lnch thick. A ten-inch line provides a drainage system beneath the liner drainmg possible h e r leakage mto the sump of the valve plt. T h ~ leakage s ultimately goes to the 11qu1d-rad~oactive waste storage tanks.

Experimental penetrations consist of the thermal column, pneumatic tubes, beam ports and the irradiation cell wmdow (Figure 5-3) The removable ends of these penetrations have bolted flanges with mechanical seals for water t~ghtness.

The 250-gpm Transfer pump interconnects the two pool sections, the irradlatlon cell and the demineralizer room.

The Transfer pump and associated piping are in the valve pit of the cooling equipment room. The system can transfer water to and from pool sections for storage, to the waste sump for d~sposal,or to the demineralizer room for purification. A pump switch 1s located at the pump and in the reactor control room on the water system control panel for operatlon of the system A single three-inch crossover h e connects the demineralizer system and water transfer-storage system for flex~bilityof operation. Flgure 5-5 shows the pool-water transfer-storage system.

4.4 Biological Shield Concrete and water serve as a b~olog~cal shield to protect personnel and visttors from the intense rad~ationthat the reactor produces. Normally, 26 feet of water covers the NSCR core. The core normally operates in the stall area of the reactor pool. In the stall, five feet of high-density concrete provldes most of the shielding for personnel in the Lower Research Level near the stall area (approximately the same level as the reactor).

When the reactor is operating in the large part of the pool, approximately eight feet of water and three feet of high-density concrete provide shielding to personnel closest to the reactor in the Lower Research Level.

The Diffuser System pulls water from the main area of the pool and discharges it through a nozzle above the core.

This forces the ' 6 in ~ the coolant flowing dlrectly out of the core into the deeper part of the pool, thereby allowing most of the ' 6 to ~ decay before it has a chance to reach the surface. The result is lower radiation levels at the surface of the pool.

Radlation levels on the reactor bridge, which is directly above the reactor, are less than 10 mlUhr with the reactor operating at 1 MW. Radlation levels in the Lower Research Level, with the reactor operating at 1 MW in the center of the stall are less than 0.5 mR/hr. Levels are higher in the immedlate vlcinity of the beam ports when extracting a beam or operating the reactor adjacent to the graphite coupler box.

The reactor operator can monitor the reactor bridge at all times, thus llmit~ngaccess to that radlation area. If an individual enters an area in the Lower Research Level that could be a radiation area, an alarm will alert both the reactor operator and the ind~vidual. Other devices flood the beam ports and/or SCRAM the reactor to reduce the radiation levels if personnel enter the area around a neutron beam

4.5 Nuclear Design 4.5.1 Normal Operating Conditions Note- The Normal Operating Cond~tionssection 1s lncluded In the Reactor Core Physics Parameters section below.

4.5.2 Reactor Core Physics Parameters 4.5.2.1 Standard TRIGA Cores This class of TRIGA reactors has well known operating characterlstics and inherent safety chara~teristics.~

The first NSCR Standard TRIGA core loading reached crltrcality In August 1968. The following were the operating characterlstics from the initial operatlonal standard TRIGA core for the NSCR:

Table 4-3: NSCR Standard TRIGA Core Characteristics I Steadv-state Power Level I IMW Crtt~calMass 2,830 grams '-')u Core Mass 3,325 grams L - " ~

Max~mumExcess React~vity $3.77 Prompt Negatlve Coefficient of React~vity -1 2 x Aktk-OC Power Coefficient (1 MW) $3.60 I

Maxlmum Pulse Energv ($2.00 Total Control Rod Worth

- insertlon) 14.7 MW-sec 1 $11.23 1 I Maxlmum Pulse Reactivity Insert~on I $3.00 1 The pulsing of standard TRIGA cores at a pulsing llmit of $2.35 resulted in safe cond~tionssince operat~onalNSCR cores were regularly pulsed at $3.00 insertions The pulsing charactenstics for a NSCR operatlonal standard TRIGA core are rn Figure 4-12. The pulsing analysis for standard TRIGA fuel 1s different from that of FLIP fuel due to a rather constant negative temperature coefficlent (-1.2 x lo4 Akk-"C) for standard TRIGA as compared to a variable temperature coefficient for FLIP fuel. More than 50% of the temperature coefficlent for standard TRIGA cores comes from the "cell effect" or dependent disadvantage factor, and -20% each from Doppler broadening of the resonances and temperature dependent leakage from the core.

Extensive measurements were made of the various parameters relating to the pulsing operation of the General Atomic Figure 4-12: FLIP Pulsing Characteristics Prototype reactor. The most important of these are below for step insertions of reactivity up to 2.1% 6klk ($3 00) l G ~ - 3 8 8 6mev. A), TRIGA Mark I11 Reactor Hazards Analysis, Feb. 1965.

Dunng puls~ngoperation, the reactor IS super-prompt-cr~tica The asymptotic perlod exh~hitsan inverse relationship to the prompt reactlvlty lnsertlon Figure

/

4-13 shows the results of plott~ngthe reciprocal of the measured perlod versus /

the prompt reactivity lnsertlon The minlmum perlod for reactlvlty msertions of $3 00 ($2 0 0 prompt) IS approxlmately three mdl~seconds

/

Flgure 4-13 also shows a plot of the reciprocal of the measured w~dthat half maxlmum power versus prompt INVERSE PERIOD react~vityinsertion.

Flgure 4-13, Figure 4-14, F ~ g u r e4-15, Flgure 4-16 and Figure 4-17 show the interrelationship between maximum transient power, pulse widths and per~od.When considered together, these plots serve to descrlbe the general features of the TRIGA Mark 111 core performance in the pulsing mode. For a glven core configuration, the amount of reactlvlty Inserted determines the peak INVERSE WIDTH AT HALF power, integral power in the prompt MAXIMUM POWER burst, and width of the pulse The plots show that the peak power 1s controllable over a wide range slnce this parameter IS very nearly proportional to the square of 1 I I 6kfk minus $1.00 Pulse width and 0 100 2 00 3 00 4 00 lntegral power, on the other hand, are PROMPT REACTIVITY INSERTION, DOLLARS approx~matelyh e a r functions of Figure 4-13: Inverse Period and Inverse Width at Half Maximum reactlvlty insertions above prompt Power Versus Prompt Reactivity Insertion critical, so, thelr range IS more limited 0 ~ 4 6 0 I 2 1 4 ~ ~

PERIOD. M~LLSECOMOS Figure 4-15: Peak Power Versus Full Width at Half Peak Power Figure 4-14: Full Width at Half Peak Power Versus Period

2 -

LS -

'00  !! :;a:  !  ::;;  !, : A : ! : ::;:I Z)(1 Reactor Period. mllllseconds Figure 4-16: Peak Power vs. Reactor Period Figure 4-17: Peak Pulse Power ts. Prompt Reactivity Squared 4.5.2.2 Mixed Standard-FLIP and Full F L I P Cores The NSCR has previously operated a mixed core reactor composed of both TRIGA standard and FLIP elements (F~gure4-18)

Considerable information from General Atomic on the TRIGA and FLIP systems reinforced the decision to convert to a mlxed core from a standard core. Stud~esmade by the NSC for a variety of cores from all standard fuel to all FLIP fuel indicated that a core with a mixed loading would safely satisfy all operational requirements. FLIP fuel elements are located in a contiguous central region of the core. Future add~tions of FLTP fuel were made such that the FLIP region grew outward as the outside core dimensions remained essentially the same.

Texas A&M investigators performed an analysis using Exterminator-2, a two-dimensional multigroup diffusion theory code. The code employs the technique of variable dimensioning, therefore the llrmtation on problem size was actual machine storage space available. An optional total thermal flux printout andlor computer plot helped facilitated the core studies An additional program calculated the average power generation

in each fuel element for the cores under Investlgatlon General Atomlc provlded bas~c-proveninput data such as cross sections, number denslties and bucklmg Homogenized calculations, In which each fuel element and ~ t associated s moderator were a cell and each control rod and ~ t associated s moderator were a cell, were necessary because the spaclng between fuel elements 1s small and dlffusion theory IS not vahd for the heterogeneous problem Fuel cell dimensions for NSCR cores ivere 3 854 centlmeters by 4 050 centlmeters Comparisons between calculated and experimental CORE C E N T E R L I ~ ~ E results tested the accuracy of the Input data and insured meaningful results using Exterminator-2 Three cases In particular show the valid~tyof the code. The first test of the Exterminator-2 program was to repeat some of General 2

0

.@ - Atomlcs' calculations for a a FLIP core. Both programs gave 3

w essentially identical results The 2

second test was a comparison of 3 4 z

a W

- the calculated and measured values of a relative thermal flux x

4 - Q ~ F ~ U S I O ITHEORY J dlstrlbution in an operational w KXTERM1NATOR-111 Z

5

- 4- EXPERIMENTAL NSCR TRIGA core. The results, Figure 4- 19, show good

-I W

a .2 - agreement considering diffusion theory calculation gives only

- \ flux values averaged over a cell.

u.

I The third test was a comparison O I b k  :  : 5  :  :  :  : :  : , 2  :  : x DISTANCE INTO CORE, INCHES DIST&NCE VfTO REFLECTOA.1HCMES measured and values (Table 4-4) of kcrf for Figure 4-19: Experiment Versus Calculated Flux Using Exterminator-2 several TRIGA cores.

Table 4-4: Comparison of Measured and Calculated Values of KII kfr Core Calculated Measured NSCR Core IA (1) (all standard) 1.032 1.026 Puerto Rico (all FLIP) 1.068 1.059 The cores for the comparisons were both 5 x 5 array (100 cell) cores. The NSCR IA (1) core had 95 standard TRIGA fuel elements, 3 shim-safety control rods. 1 transient rod and 1 regulating rod. The Puerto Rico all-FLIP core had 94 FLIP fuel elements, 3 shim-safety rods, 1 transient rod, one regulating rod, and one alr void experiment hole. In both cases, the calculated values were sllghtly higher than the measured values, and for safety consideration, the results were conservative.

These results established confidence in the Exterminator-2 code and the GA cross-sections. The conclusion is that any safety calculations obtained from this program would yield realist~cand reliable results.

As ment~onedearher, Exterminator-2 computed the power generation In a glven core for all fuel cells. However, the computed peak adiabatic fuel temperature during a pulse was on a selective basis following examination of calculated cell power values Weighting the average power with the flux distribution, for the cells exhibiting the hlghest average power, provided the drstrlbut~onof power density In each The flux dlstrlbution was the product of the average flux from Exterminator-2 and the ~ntra-cellflux from General Atomlc. Thls method was in good

agreement wlth two dimensional cell calcula~ionsof h~gherorder, w h ~ account h separately for cell effects A knowledge of the temperature as a function of the volumetr~cheat content allowed calculating the maxlrnum fuel temperature for a pulse of given energy Flgure 4-20 shows a comparison of 108 - measured and calculated pulse

- EXTERMINATOR I1 temperatures for smge thermocouple locations In mlxed TRIGA cores. Pulse 0 PULSING. I.F. 7 5 2 3 - F T W temperatures were for a $2 04 pulse E

W 0 The relat~vevalues for the pulse 2

w 104- temperatures determined by the code and I-the experimental values agree very well.

The next comparrson was the experimental and calculated ratios for $2.00 ~ u l s e

+

q W

I000 0 a

thermocouple for each set of values I

0 I

90 t

180 t

270 I

3 60 obtamed For a 35 FLIP element m~xed core, the observed and calculated ratlos I F ROTATION. DEGREES agree to within 4%. The design core for the first NSCR loading was 35 FLIP and 63 standard TRIGA elements. This core went Figure 4-20: Experimental vs. calculated pulse temperatures Crltical on July 1, 1973, and pulsed on July 4, 1973. Figure 4-18 shows the core loading designated as 111-A. Table 4-5 shows the operating characteristics for core 111-A.

Table 4-5 Operating Characteristics of NSCR Core 111-A Steady-state Power Level Critical Mass Core Mass I Maximum Excess Reactivity I Power Coefficient (1 M@) 1 $2.50 I 1 15.7 MW-sec Maximum Pulse Energy ($2.00 insertion)

Total Control Rod Worth 1 $15.57 I Maximum Pulse Reactivity Insertion 1 $2.00 Reactor core studies using Exterminator-2 predicted a core excess of $6.20 for Core 111-A as compared to the measured value of $6 09. Pulsing characteristics observed for 111-A are in F~gure4-21.

The design considerations for pulsing FLIP fuel as compared to standard TRIGA fuel are different due to a temperature-dependent negative temperature coefficient for FLIP fuel (refer to F~gure4-23). For a TRIGA FLIP fuel element, the uranium loading is about three and one-half times that of a standard TRIGA element, which makes the neutron mean free path FLIP element much shorter. For this reason, heating the fuel-moderator does not greatly reduce the

resonance escape probabdlty for neutrons (The longer slowing domn length of Standard fuel increases the probability that a neutron wdl be absorbed \vdl slo\\lng down In a U-238 resonance) In the TRIGA FLIP fuel, the temperature-hardened spectrum decreases reactlvlty through

~ t Interaction s with a low energy resonance material. Thus erblum, with its double resonance at -0 5 eV, serves both as a burnable polson and enhances the prompt negative temperature coefficient In the TRIGA FLIP fuel. The neutron s ectrum shift pushes more of the thermal neutrons into the ' ~ resonance r as the fuel temperature increases In TRIGA fuel, the temperature coefficient 1s prompt because there 1s a lager portion of moderator In the fuel-moderator materlal (This effect is greater for Standard than for FLIP) The fuel and moderator temperatures rise simultaneously producing the temperature dependent spectrum shift.

In a TRIGA FLIP core, the results of cell structure on the temperature coefficient are small. Almost the entlre I 1 coefficient comes from temperature dependent changes in 0 I I I I I 0 100 ZOO 300 400 500 600 100 qfw~thlnthe core and -80% of this effect is independent of the cell structure. The calculated temperature coefficients are in F ~ g u r e4-22 for standard, mixed and FLIP cores. The Figure 4-22: Temperature Coeff~cientsof TRIGA temperature dependence of the temperature coefficient of a Fuels -

TRIGA FLIP core is advantaneous in that a minlmum 146 Mw-SBC - FLIP reactivity loss is incurred in reaching F) = 1.040 MW

-- STD. TRlGA normal operating temperatures, but any sizable increases in the average core temperature result in a sizably increased prompt negative temperature coefficient to act as a shutdown mechanism. The burnup calculations indicate that after 3000 MW days of 14.1 M w - S ~ C operation, the 2 3 5 concentration

~

a;eraged over the core is -67% and the I 6 ' ~ rconcentration is -33% of the beginning-of-life values. The end-of-life coefficient for a FLIP core is less temperature dependent than the beginning-of-life coefficient since the sizable loss of 16'~rand the resulting increased transparency of the -0.5 eV resonance region to thermal neutrons.

Figure 4-23 demonstrates the effect of the temperature coefficient for standard and FLIP cores upon the pulse shape. Note that the FLIP pulse peak power is considerably higher than that in standard fuel, thus higher fuel temperatures for the same reactivity insertions.

Figure 4-23: Comparison Of FLIP To Std. TRIGA Pulse For The time for full width at half maximum Similar Reactivity Insertions (FIVHM) for a F L P pulse is considerably less than that for the standard core pulse, yet the total pulse energy is approximately the same.

4.5.2.3 FLIP and LEU Cores Promp fc'e~abwTempratwe Cofflicicnt TRICA-LEU 20-20 Fuel Standard and FLIP TRIGA cores have well nith 0.590 v+t% Frburn known Inherent safety charactenstics2. The 18 pulslng of TRIGA-FLIP cores with $2 00 reactivity insertlons results In very safe conditions. Earlier operat~onalNSCR cores were regularly pulsed wlth $3 00 insertlons The pulslng characterist~csfor a NSCR operational TRIGA-FLIP core are in Flgure 4-12. The pulsing analysis of LEU cores indicates performance similar to that of FLIP cores.

In TRIGA FLIP and LEU fuel, the temperature hardened thermal neutron spectrum adds negatlve reactlvlty through its interaction wth erbium For TRIGA LEU fuel, the 2 3 5 loadlng

~

1s less than that of TRIGA FLIP fuel Thls results in an increased concentration of 2 3 8 ~ .

This allows the higher 2 3 8 ~loading to generate a similar prompt negative coefficient of reactivity with a lower erbium concentratlon 1

Another important design feature of TRIGA fuel is the ZrH moderator present In the fuel Figure 4-24: Prompt Negative Temperature Coefficient structure. The ZrH moderates neutrons in the TRIGA-LEU 20-20 Fuel fuel rod. As the fuel temperature Increases, the density of the material decreases, resulting in a negatlve feedback effect lmOl 26 W W W m Figure 4-24 shows the calculated temperature coefficients in for both FLIP and LEU cores. The temperature dependent character of the temperature coefficient of a TRIGA-FLIP or LEU core is em o 19 W 1Pu-s.x advantageous because reaching normal operatmg temperatures f provides an acceptable negatlve reactivity add~tlon Further increases in the average core temperature result in a sizably z scoo i!

FWN 17 w increased prompt negative temperature coefficient to act as a shutdown mechanism. The burnup calculations indlcate that after 400 o 1-1 3000 MWD of operat~onfor FLIP fuel, the 2 3 5 concentration

~ 2w o averaged over the core is -67% and the 1 6 7 ~ concentration r is -33%  : 1%

of the beginning-of-life values. For LEU fuel at 1300 MWD, the OD A 00 0I 02 03 04 1

05 2 3 S ~concentration is -47% and the 1 6 7 concentration

~r is -11% of rlnu(XCO&I the beginning-of-11fevalues. The end-of-11fe coefficient for both fuel types is less temperature dependent than the beginning-of-life Figure 4-25: Comparison of FLIP And LEU Pulses (BOL) coefficient because of the sizable loss of ' 6 7 ~and r the resulting increased transparency of the 0 5 eV resonance to thermal neutrons.

Flgure 4-25 demonstrates the effects of the temperature coefficlent for FLlP and LEU cores upon the pulse shape at beginning of life. Note that the LEU pulse peaks at a power higher than does the FLIP pulse; thus, higher fuel temperatures will result during the LEU pulse as Flgure 4-26 shows. This relationship between the two fuel types w l l change as the core hfe progresses By the end of core life, the FLIP pulse will generate hlgher temperatures than the LEU pulse. Figure 4-27 and Flgure 4-28 show this behavror.

GA 3886 (Rev A) TRIGA M a r k I11 Reactor Hazards Analysis. Feb. 1965 4-16

Durmg pulsing operatlon. the reactor 1s In a super- 033 C --.-- -.------. -Fo. $ 1 - lnse-wn U3L i

L-prompt-crit~calcond~t~on The asymptotic pertod is inversely related to the prompt reactlwty ~nsertlon.

Figure 4-13 sho\vs the results of plottlng the i ,..,

-- - - ----- .\

reciprocal of the measured perlod versus the prompt i reactivity insertion The m ~ n ~ m uperiod m obtalned for 4230 reactivity insertions of $3.00 ($2.00 prompt) is 3 1

msec Flgure 4-13 Also shows a plot of the reciprocal of the measured width at half maximum power versus prompt reactiv~tyInsertion See Figure 4-14, Flgure 4-15, Flgure 4-16 and the accompanying text for a f

3 0 0 0 ~

% mJo , --ZiXj

---HIP d~scusslonof pulse response. 1000 The NSCR may operate a core composed of TRIGA oo 00 50 to D LEU fuel elements (Table 4-2). The LEU fuel will T ~ m ~r r m s >

have essentially the same properties as the TRIGA FLIP fuel.

The investigation of the design of a core composed of thls type of fuel required the use of a computer code capable of producing all the necessary design parameters dunng both steady state operatlon and during accident scenarios Accuracy of computations was of the utmost importance. Performing benchmark calculations for FLIP cores and comparing results for the LEU core to those for the FLIP core assured Accuracy.

The code selected for the generation of few group neutron cross sectlons representative of various sub-regions in the core and 11s surroundmgs is WIMSd4/m, a one-1 a#$ 1 EO lnrcruoq EOI.

dimensional neutron transport code capable of 1209 o r solving for flux distributions using many fine energy groups. WIMSd4/m created spatially and lm 0 26 49 sw uc F7\'11\1. 11 muc 1Ll energetically averaged sets of cross sectlons In the 2 b M 4lW K-seven standard energy groups used in TRIGA , 19- m p 8000.

analyses for each core sub-region. Since neutron 5

cross sections are highly temperature dependent,

6300 -

Z WIMSd4Im calculated many cases covering the s d

entlre operational temperature range for both 4330 -

normal and accident conditions for each type of core sub-region. A sub-reglon can consist of a fuel pin and its surrounding water, a water hole, a 2000 .

control rod and its surroundmg water, a section of the graphite reflector or any other material present in the core or surrounding experimental irradlatlon OOoo -_

o1 -_--- 02 o3 mmc~ r ~ m b r )

o4 areas of interest.

Figure 4-27: Comparison of FLIP And LEU Pulses For

$1.80 Insertion (EOL)

The code DIF3D. using three dimensional neutron diffusion calculations, modeled the core, the reflector and the irradiation facilltles. DF3D used the generated temperature and burnup dependent library of homogenized cross sectlons.

This code is capable of computing the hrr for different fuel and control rod configurations, as well as the spatially dependent power and neutron fluxes for each energy group. The code allows the investigation of different control rod positions and fuel loadlng arrangements and generate the prompt temperature coefficients of reactivity necessary for the translent calculat~onson a full core basis.

To prove the accuracy of these codes for model~ngthe NSCR, lnvestlgators at Texas A&M Unlvers~tyrecreated a test case run by General Atomlcs on a two-megawatt Figure 4-28: Pulsing Temperaturcs for $1.80 TRIGA core3 This test case models a core contaming fuel Insertion (EOL) slmilar to the TRIGA-LEU fuel proposed for use at the NSC and thus served as a benchmark calculation. The fuel In both the NSCR and In the test case has essentially the same properties for each element The major difference 1s the add~tlonof a shroud around each four-rod cluster In the GA test case. In addition, the GA test case core has a modeled power level of 2 MW while the NSCR core wdl operate at 1 MW. This does not adversely affect the vahdity of the test case Table 4-6 shows a summary of the core parameters for thls test case Table 4-6: Core Parameters of GA Test Case Fl~elCli~cter: TRTGA-TX1J 21) wt% 7 J in IT-7rH Fuel rods per cluster:

Standard Cluster: 4 Control Cluster: 3 Nom~nalFuel Rod Dimension Fuel 0.D . 32.4 mm Clad 0.D : 33.5 mm Fuel Height: 508 mm Fuel Loading: 548 mm U (20% enriched)/rod 2.2 Kg U (20% enriched)/cluster 440 gm U-235/std cluster

-0 59 wt% Erblum as burnable absorber Number of fuel clusters in core: 26+1 Standard Clusters 21 Control Clusters: 51tl Reflector: Water Core size: 78f 2 h e r s U-235 contentlcore: 10.6 Kg Core Geometry: 4x6 arrangement Grid Plate: 6x9 positions (normal conversion)

Burnup Status of the core Equilibrium core Average core burnup: -20%

Thermal-Hydraulic data:

Average Power Density: 26 Kwfliter Coolant Flow Rate: 1000 gpm Core inlet temperature: 38°C General Atomics, "Generic Enrichment Reduction Calculations for Rod-Type Reactors" Research Reactor Core Conversion From t h e Use of Highly Enriched Uranium t o t h e Use of Low Enriched Uranium Fuels Guidebook. International Atomic Energy Agency, Vienna. 1980

Table 4-7 l~stsThe peaking factors generated and reported by GA for the test and the correspondme values from the NSC calculations using WMSD3/m and DIF3D Table 4-7: Peaking Factors From GA Test Case I

Tape of Peakmg I % General Atomics 0 MWD 900 MWD Core Radlal 1 57 1.61 1.62 Core Axial 1.36 1.30 1.31 1D Cell ( 2 3 ' ~ ) 1.48 1.41 1.29 1D Cell (310°C) 1 52 1.44 131 1D Cell (700°C) 1.61 1.49 1.34 The core radial and core axial values agree quite well, but there is a small discrepancy in the 1D cell values. The GA test case is at a burnup of approximately 20%, wh~chcorresponded to approximately 900 MWD. TAMU values would be in better agreement with the no burnup on the core.

Table 4-8 compares of the peak thermal flux values in the core and the water reflector. The sum of the values of groups 5 , 6 and 7 in the 7-group model provlded the thermal flux values. These flux values are in close agreement.

Table 4-8: Estimated Peak Thermal Flux a t 2 MW 4-Rod Cluster TRIGA-LEU Fuel General Atomics TAMU Core 1.5 x l0l3 1.5 l0l3 Reflector (water) 2 x 10'~ 1.9 x l0l3 The final parameter that GA generated in their report was the prompt negative temperature coefficient.

Figure 4-29 shows the Texas A&M values along with the GA values for this parameter.

Promfi Ficrative T c m ~ r r t u r eCocflicicnt These are the only parameters verified TRICA-LEU Fuel-l-RodCluster against General Atomics data for the 'fZXAS ABrRl hlODDt benchmark case. However, they show that the TAMU models are capable of producing accurate results as compared to the GA approved models DIF3D is also capable of producing output in much more detail than this document shows. With the aid of an external plotting program (TECPLOT),

investigators generated three dimensional flux maps that might prove useful in future operations at the NSCR 4--- I....... G* /

A steady state thermal hydraulic analysis, using the DIF3D power distribution and NCTRIGA, determined the maximum fuel and clad temperature during operation.

NCTRIGA is a one-dimensional thermal o 1mxa m 4 m m a 703 m m ~ a x ~

hydraulic code that calculates temperatures

'IEbIPRL\TURE(T) at several nodes along a single fuel rod channel and determrnes the natural Figure 4-29: Prompt negative temperature coefficient TRIGA LEU convection lnduced coolant flow rate. fuel 4-rod cluster NCTRIGA uses the power distrlbutlon

generated from the neutronic analysls and information on the geometry and material composit~onof the flow channel to produce temperatures in the fuel and to pred~ctthe coolant flow rate.

Investigators used the code PARET from Argonne Nat~onalLaboratory to perform calculations to determine peak fuel and clad temperatures dur~ngnormal pulsing and accident transients for the both the FLIP and LEU cores. This code 1s capable of performing 1-D radlal heat transfer calculations under non-steady state conditions at several axial nodes along a fuel rod. They used NCTRIGA to generate the initla1 temperatures and flow rates needed for lnput to PARET and DIM (with temperature dependent cross sections taken from WIMSd4/m) to generate the temperature coefficients of reactivity to input to PARET. Flgure 4-24 shows these for FLIP and LEU fuels at various values of burnup. T o acquire inltial power distribut~ondata for pulsing, investigators ran a 'DIM job' at 300 watts steady state power with the transient rod fully inserted. Compiling this data into a PARET deck produced a model for predicting performance of the peak fuel element and core during pulsing Comparisons of peak pulsing power data from experiments on the Texas A&M Nuclear Science Center Reactor and data from the model provide a benchmark for the method. Figure 4-30, Figure 4-32, Flgure 4-31 and Figure 4-33 show the relationship between predicted and measured pulse values and that the predicted values are reliable.

Peak Pulsing P m r 1

Figure 4-30: Peak Pulsing Power Pulrf ng Energy Rtrclidty tnrtrfcd 6)

Figure 4-31: Pulsing Energy The pulse energy mismatch at $1.80 (Figure 4-31) is most likely because the experimental value IS too small caused by the detector overloading due to the excesslve amount of energy generated In this pulse.

2 4 6 8 10 I2 14 16 18 PERIOD, MILLISECONDS Figure 4-32: Full Width a t Half Peak Power Ptsk Temperatures

-m- PARETcent. fine.

[-&-PARET fuel surface 0

1.00 1.10 120 1-30 1.40 1.50 1.M) 1.70 180 R c r c t l r i l ~lnsttttd (5)

Figure 4-33: Peak Temperatures This peak fuel temperature 1s difficult to exactly quantify due to the nature of the experimental data. The is from a thermocouple mid-way between the fuel surface and the centerlme of the (F~gure4-7). Smce the measured temperature falls between the predicted centerhe and fuel surface temperatures, the results are reasonable

Peak to Thermocouple Temperature Rise Ratio Cent. Lint Ttmp (drt C)

Figure 4-34: Peak to thermocouple Temperature rise ratio Investigators used a curve fit to develop a functional relationship for the ratio of the peak fuel temperature to peak thermocouple temperature as a funct~onof peak centerline temperature This related the thermocouple temperature to the temperature In the element Figure 4-34 shows a plot of this for the data.

A second order polynomial fit to this data results In:

Where:

R(T,3 IS the ratio of the peak centerline temperature to the Ta, is PARET calculated peak centerline temperature of the Smce LEU and FLIP fuel heat transfer properties are similar, this ratlo should hold for both fuel types 4.5.2.4 2. LEU Cores Designing a core for the NSCR from the methods described above, a LEU core for the NSCR has the properti Table 4-9.

Table 4-9: NSCR LEU Core Properties l~teadvState Power Level: 1 1MW I bumber of Fuel Elements: 1 90 I Critlcal Mass: 6 836 grams 2 3 5 ~

Core Mass. 8,230 grams ='u haximum Excess Reactivity: 1I $3.43 I Total Control Rod Worth: $16 91 SS1: $2.78 1 RR: 1 $1.39 I TR: $2.80 Minimum Shutdown Margin: $0 91

The core map 1s the same as In Flgure 4-1 1 All of the deslgn parameters meet the requirements of the Technlcal Speclficat~onsfor the NSCR. The procedure indicated In the Technical Speclficat~onsprovides the method to calculate the shutdown margln The deslgned core exceeds the requlred h m ~ of t $0.25.

4.5.3 Operating Limits The Technlcal Specifications (Sect~on14 of this SAR) speclfy the operating limlts for the NSCR 4.6 Thermal-Hydraulic Design The NSCR operates at 1 MW steady state wlth natural convectlon cooling The NSCR can operate anywhere along the pool centerline except In the gateway between the stall and large pool section. Pool water constantly surrounds the reactor core. This water flows freely from the bottom and sides of the core during the convection coollng process.

F~gure4-1 shows that the four-rod fuel element assembly provides easy passage of coollng water through the assembly. Water flows by natural convectlon through the 2" diameter hole in the gnd plate adapter. It passes through the large cruciform openlng and then over the entlre element untd it leaves the core through the numerous openlngs in the aluminum handle In additlon to the coolant passages through the grid plate adapters, the NSCR grid plate has additional coolant holes 112" in diameter located at the corner of each four-rod element.

General Atomics has successfully operated Mark 111 standard fuel elements and FLIP elements operated in TRIGA cores at steady state power levels of up to 1.5 Megawatts. The arrangement of fuel in the NSCR is such that the mmmum nominal spacing between the fuel rods provides adequate convection cooling of cores up to 2.0 MW.

Figure 4-35 shows the nominal spacing of the fuel rods In the NSCR core. This Increased spacing and the extra coollng holes at the corners of the bundle considerably enhance Core coollng. .

Figure 4-35: Nominal Fuel Rod Spacing in the NSCR Core

5 REACTOR COOLANT SYSTEhlS 5.1 Summary Description The various pool water systems accompl~shheat removal, pur~ficat~on, recirculation, transfer, storage, make-up water addlt~on,pool surface cleaning and 11quidwaste d~sposal. System components and piping handl~ngpool water are stainless steel, aluminum, and plastic to maintain maxlmum pool punty. Welded piping systems with mechanical seals insure mlnimum leakage.

A heat exchanger system with pool water on the primary slde and cool~ngtower water on the secondary s ~ d cools e the reactor pool The entlre cooling system consists of the pool, heat exchanger, coolmg tower, primary and secondary pumps and associated pipmg The maximum operatmg water pressure occurs in the heat exchanger tubes. The maximum pressure for other pool water systems corresponds to the reactor pool depth of thirty-three feet. The maximum heat exchanger tube pressure of approximately twenty-two psl 1s \ire11 below the des~gnpressure of 150 psi for all systems.

The Control Room is the remote operating statlon of the pumping components of the pool water systems. Figure 5-1 IS a schematic of the pool water systems

F~gure5-2 shows the elevations of the water systems.

Two two-inch drain Ilnes, one on the floor of each pool section, terminate in the dermneralizer room. These Iin~

are for drainage and recirculation. Two three-inch demineralizer recirculation and fill lines are located near the of the pool The pool surface sklmmer system has two one and one-half inch lines at the top of the pool for operation A ten-lnch h e beneath the liner at the center of the stall section routes Pool liner leakage to the re pit The irrndlation cell floor has a three-mch drain h e . Flgure 5-3 shows the piping penetrations In the reactor >I

5.2 Primary Coolant System Three ten-lnch water-coolmg h e s are located on the floor of the reactor pool. The Primary Cooling pump take:s suctlon on the single ten-inch h e located on the centerline of the main pool. The pump d~schargesthrough the heat exchanger and back to the pool through one of two ten-mch lines (these d~schargein the stall and main pool).

Diffusers are on the d~schargeof the two return Ilnes.

The pool cooling system (Flgure 5-4) has a design capacity of 2 MW with nominal pool operatmg temperature between 70°F and 80°F. Reactor pool water flows through the tube side of the heat exchanger for cooling and 1 returned to the reactor pool. This primary system 1s a closed loop with a deslgn flow rate of 1,000gpm

I Drains 1 Primary Coolant Loop Figure 5-4: Reactor Pool Cooling System The secondary cooling water flows from the basm of the coolmg tower through the shell side of the heat exchanger and back to the coollng tower. The cooling tower uses evaporative coolmg to remove heat from the secondary water to the atmosphere at the cooling tower. The secondary loop has a nominal flow rate of 1575 gpm The cooling tower will dellver 83°F water at 78°F wet bulb alr temperatures The primary loop components are stainless steel Thls helps to preserve pool water purlty during the cooling process Components for the primary coolmg loop are located in the cooling equipment room on the lower research level The tubes, tube sheet and header of the heat exchanger are stainless steel and the shell is carbon steel. Design operating rnlet pressures of the heat exchanger are 30 psi for the prlmary side and 22 psi for the secondary side.

The convection cooled TRIGA core does not present a problem of fuel melt down and resultant fission product release when there is a loss of coolant flow through the heat exchanger. Loss of the cooling system with the reactor in operation would result in a gradual pool temperature increase. Therefore, ample tlme is available before it would be necessary to terminate reactor operations due to a hlgh pool temperature. It follows that loss of electrical power to all coolant systems would not result in a hazardous condition 5.3 Secondary Coolant System

The Secondary Coohnp System conslsts of a pump, secondary s ~ d of e the heat exchanger coohng tower and assoc~atedpiplng "Auto-Off-Hand switches allow local operat~onand are located at the secondary pump and cooling tower. When the local sw~tchis in the "Auto" positlon "On-Off' switches in the control room operate the components.

The Auxlllary Alarm Panel in the Control Room prowdes alarms In for pr~maryand secondary pump power fa~lures and for secondary loss of flow. Local detectors monitor heat-exchanger lnlet and outlet temperatures A computer or electronic system displays temperatures for Control Room operators.

Chemlcal treatment of the secondary loop extends the life of the components and reduces scale deposlts In the heat exchanger. A system to control secondary chem~strysamples the water and activates chemical injection or initiates a blow down 5.4 Primary Coolant Cleanup System 5.4.1 Demineralizer/Recirculation System The purposes of the Demineral~zer/Recirculat~on System are:

1) Maintain pool water purity,

2) Provide a filtering mechan~smfor makeup water and

3) Provide a path for makeup water for filling the pool durlng both normal and emergency pool fills.

The Demineral~zer/Recirculationsystem uses regenerative mixed bed demneralizer unit in conjunction wlth micron filters, activated charcoal and gravel filters The Recirculation pump takes ~ t suctlon s In the Southwest corner of the pool and drscharges to the Northeast corner of the main pool area.

This system (F~gure5-5) 1s located In the demineralizer room on the lower research level It has a design flow rate of 75 gpm wlth an output conductivity of less than one m~croseirnenper cm3

,,-Filter

'-'-= f

--Activated Charcoal Mixed Bed Demineralizer I

'- Raw Water Make-up A

Filter: Cotton Wound I From Pool Water Purification System TO Pool I Waste Discharge to Creek Tanks 1-12,000 gal 2-12,500 gal Demiernalizer Room Sump Liquid Waste Collection Liquid Waste Disposal System Figure 5-5: Water Purification and Disposal System A remote "on-off' switch is located in the reactor control room for operation of the demineralizer recirculation pump. The local "Auto-Off-Hand" switch is located next to the pump in the Demlneralizer room If the local switch is In the "Auto" position, the remote switch In the Control Room controls the pump. "Off' prevents the remote switch from starting the pump and "'Hand" starts the pump regardless of the remote sw~tchposition.

Regular maintenance includes manual regeneration of the dernlneral~zeras required.

5.4.2 Skimmer System The purpose of the Sklmmer System IS to maintain the surface of the pool free of dust and debris. This system has lrttle effect on the actual purlty of the pool water The Sk~mmerpump takes suction from a surface suctlon filter and discharges In the Northwest corner of the pool The pump and filter are located in the chase level Filter Cotton Wound 1 r Strainer I llll 1111 1111 1111 Skimmer Return -

Skimmer Intake Drain to Demineralizer Room '

1 Liquid Waste Sump Figure 5-6: Skimmer system 5.5 Primary Coolant Makeup Water System The Deminerahzer system provides makeup water by processing raw water before it enters the pool. Raw water enters the Deminerahzer system downstream of the Reclrculatlon pump. The flow path to the pool 1s through the filter, charcoal bed, gravel bed and demineralizer. Figure 5-5 shows the system with the Raw Water Connection.

5.6 Nitrogen-16 Control System The NSCR core diffuser system draws water from the pool and discharges it through a nozzle above the core. The result~ngcirculation pattern reduces the dose rate at the pool surface from ' 6 and

~ 4 1 ~produced r in the coolant water as it passes through the core. The diffuser pump and associated piplng is located in the mechanical chase as in Flgure 4-10. Two outlets permit operation of the system when the reactor is in the large pool or stall section. A flexible, quick disconnect water hose connects the bridge piping to the d~ffuseroutlets. The On-Off switch for the system is in the Control Room on the water systems control panel 5.7 Auxiliary Systems Using Primary Coolant Primary coolant provides cooling and shieldmg. It has no auxlllary uses

6 ENGINEERED SAFETY FEATURES 6.1 Summary Description Sectlon 13 2.1 states that "no realist~chazard of consequence wdl result from the Design Basis Acc~dent" As a result. the NSCR does not requlre engineered safety features. T h ~ analysis s considered simultaneous fa~luresof the reactor pool integrity, fuel cladd~ngand the ventilation system operabll~ty.

6.2 Detailed Descriptions 6.2.1 Confinement The reactor confinement b u ~ l d ~ n1sga cylindrical steel remforced concrete structure, approxlmately seventy feet in diameter and seventy feet high. Approx~matelyfifty-five feet of the structure is above grade. An exhaust blower The upper research level (Figure 2-5) is the largest by volume of the three levels. The exterior walls of thls level and those of the central mechanical chase are reinforced concrete slabs between concrete-encased steel columns The next level down and approxlmately at grade level is the central m he reactor pool walls take up a major portion of the available space on this level. Signal and power cables, which connect the reactor to the control room, pass through trays attached to the ceiling of the chase.

The reception room is located outside the south side of the confinement structure A master control panel for operation of exhaust and air condit~oningsystems in the confinement structure is located on the north wall of the reception room.

A laboratory building, on the south end of the reception room and outside the confinement building, contalns pneumatic receivers (Figure 2-7). Each pneumatic system to the laboratory building lies within a large airtight tube and air within this tube flows through the existing exhaust system and Facil~tyAir Monitors before release from the stack. This design allows for monitoring and controlled release of rad~oactivegases associated with operation of the pneumatic system Four air handlmg units and an exhaust fan control pressure, temperature and hum~dltywithln the confinement building. The confinement building has three zones of negative pressure for effective rsolation of possible contaminated areas. The zone of least negative pressure includes the control room, locker areas and the bu~lding entry where contamination lest likely. Air recirculates in this zone but exhausts through the monitoring system and the stack. An Intermediate zone of negative pressure includes the upper and lower research levels where infrequent contamnation might occur. Air also reclrculates in t h ~ zone.s The third zone of maximum negative pressure lncludes areas where contamlnatlon of activation is most likely to occur, 1.e.. beam ports, thermal column, and

through tubes Also included In this zone of maxlmum negatlve pressure 1s Laboratory 1. An- does not recrrculate In t h ~ zone s and exhausts directly to the stack and monitormg system.

Air-handl~ng units prov~deand c~rculatefresh alr In the b u ~ l d ~ n gAll four units have controls on the control panel In the receptlon room and will shutdown s~multaneouslyw t h the central exhaust fan when a~rbornerad~oact~vlty reaches alarm levels on the exhaust particulate monltor, the Xe-125 monltor or the fission product rnonltor Dampers are located at the alr Inlet to all handling units, the fresh air bypass to the exhaust fan, and in the 84-feet The air handhng system is comprised of two sections. One sectlon handles fresh air, controls temperature and humidity, and rec~rculatesbu~ldlngalr. The second sectlon controls buildmg pressure and exhaust A control panel 1s located in the receptlon room for operat~onof the system The air-handllng units, exhaust-fan and assoc~ated dampers can be operated from t h ~ panel.

s Emergency air handllng operations are performed at this panel.

6.2.2 Containment The Central Exhaust Fan and the Central Exhaust Bypass Dampers mamtain the buildmg at a negatlve pressure T o prevent or rmnimize the uncontrolled release of rad~oactlvityto the environment surrounding the Nuclear Science Center. In addition, the Fac~lltyAir Monitors continuously monitor air discharged from the b u ~ l d ~ n If g a splll or a release increases radioactivity levels above a pre-selected set point, the F a c ~ l ~Air t y Mon~torsgenerate a slgnal to shutdown the air-handlers and close the Inlet dampers Thls isolates the building. Operators can control the alr handling system, including the Emergency Exhaust Filter-bank, and monitor the actlvity in the air from the control panel in the reception room 6.2.3 Emergency Core Cooling System Emergency coolmg for the NSCR is the 106,000 gallons of water contained in the pool and stall portlon of the reactor pool. The large heat capacity of t h ~ amount s of water can cool the reactor for several hours at 1 MW In the event of failure of the coohng system.

The two coolant return lmes and the coolant extraction line in the bottom of the pool have manual closures to isolate the pool in the event of cooling system component failure.

7 INSTRUR/IENTATIONAND CONTROL SYSTEMS 7.1 Summary Description The reactor operates in two modes:

Steady State Mode - steady power levels up to 1 MW Pulse Mode - a rapid Translent Rod withdrawal (the technical specifications set the limit) causes a large power excursion A Mode Selector Switch at the Reactor Console allows operators to select between Steady State and Pulse modes All reactor operations are at the Reactor Console, which provides for reactor and reactor systems controls and indication of reactor and reactor systems parameters 7.2 Design of Instrumentation and Control Systerns Five rad~ation-basedmstruments provlde indication of reactor power from intrinsic source range levels to full power. Two of these instruments, A Wide Range Linear Drawer (with multlple scales) and a Log Drawer (with multiple instruments), provide indication over the entire range. Two, the Safety Drawers, provide ind~cationonly above 10kW. One, the Pulse Drawer, provides indication above IOkW, and provides mdicatlon of peak power and energy during reactor pulsing. The Fuel Temperature Instrument provides jnd~cationof fuel temperature and records maximum temperature durlng pulses The Safety Drawers, Log Drawer and Fuel Temperature Instrument provide SCRAM capability to the Reactor Safety System.

The two Steady State methods of controlhng the reactor are Manual and Automatic In Automatic (I e. Servo), the Wide Range Linear Drawer provides the power level input to a servo controller. The servo controller generates a signal to drive the Regulating Control Rod (Reg Rod) as required to maintain a constant-preset power level.

Various experiment and manual scrams exist as interlocks that wdl automatically shut down the reactor.

7.2.1 Design Criteria The instrumentation and control system provides the following:

Information on the status of the reactor The means for insertion and withdrawal of control rods Automatic control of reactor power level The means for detecting over-power, fuel over-temperature or loss of detector voltage and automatically shutting down the reactor to terminate operation 7.2.2 Design-Basis Requirements The primary design basis for TRIGA reactor safety is the limit on fuel temperature. In the pulse mode, the reactor SCRAMS when temperature reaches the Limiting Safety System Set point for a fuel temperature. In Steady State mode, fuel temperature trip and the power level trips independently prevent exceeding a safe operating temperature.

7.2.3 System Description 7.2.3.1 Log Power Channel The log power channel consists of a fission chamber, preamp, amplifier and rate meter.

The log power channel performs the following:

Provides reactor power indication over a of range of ten decades of reactor power Provides the following:

Low Count Interlock o Prevents withdrawing control rods without at least two counts per second

Per~odScram o Provides a scram signal to the Safety Ampl~fierwhen reactor per~odIS three seconds or less and the Pcriod Scram is not bypassed The Log Drawer contains two overlapping Instruments. The low-range instrument converts pulses from the fisslon chamber to a logarithmic power indlcat~on The hlgh-range Instrument converts detector current to a logarithm~c power indlcat~on The ~nstrumentsoverlap to provide contmuous indication F ~ g u r e7-1 shows a simplified dlagram of the log power channel n WIDE RANGE MONITOR *

- 1 KW INTERLOCK COWCOUNT RATE INTERLOCK 3 SECOND PERIOD SCRAM DlGlTPL PERIOD COUNTER METER G

Figure 7-1: Log Power Channel 7.2.3.2 Pulse Channel The Pulse Channel consists of an uncompensated Ion chamber and the Pulse Drawer.

The Pulse Drawer provides the following indications in the associated mode Steady State Mode o Percent Power Pulse Mode o Percent Power o Peak Power o Energy (Mw-Sec) 7.2.3.3 Wide Range Linear Channel The linear power channel consists of a compensated Ion chamber and a Linear Wlde Range Drawer.

The Linear Wrde Range Drawer provides the following:

Power ind~cat~on over the entlre range of operating and shutdown levels

Input to the servo controller for automatic power control The detector is above a tapered graphlte reflector element that scatters the neutron flux from the core face This configuration provides excellent linearity and significantly reduces the contribution due to gamma rays so that the system is sensitive and accurate at low power levels even after extended operation at ~ M W . '

7.2.3.4 Servo Control System In automatic control, the servo controller compares the slgnal from the Linear Wlde Range Drawer to a preset signal. It prov~desa shim-rn or shim-out signal to the Regulating Rod Drive Controller to adjust power, as indicated by the Wlde Range Lmear Drawer, to the preset level. Regulating rod control automatically shifts back to manual if the actual level drifts excessively from the preset level The Regulat~ngRod Drlve Controller receives a s~gnalfrom both the servo controller and the manual control sw~tchon the Rod Control Module. Figure 7-2 shows the h e a r power channel 0 COMPUTER HIGH VOLTAGE SIGNAL VUDE RANGE COMP VOLTAGE LINEAR METER Ir ,

li 1f SEW0

& REGULAflNGROD CONTROL u

COMPENSATED JON CHAMBER Figure 7-2: Wide-Range Linear Drawer

7.2.3.5 Safety Pon e r Channels Each of the two Safety Power Channels consists of an uncompensated Ion chamber, a Safety Drawer and an external hlgh-voltage power supply The two channels are identical and isolated from each other. Each channel provldes an Independent scram input to the Safety Amplifier located between the Safety Drawers The Safety Power Channels can be the Limitlng Safety System wlth the LSSS at 125% Full Power in Steady State ~f the Fuel Temperature IS not available There are two scrams associated \blith the Safety Drawer.

• Hlgh power scram o Provides a scram when the reactor power reaches 125% of full power Loss of high voltage scram o Provides a scram signal when the detector voltage drops below 150V 7.2.3.6 Safety Amplifier The Safety Amplifier supplles current to the control rod magnets providing the mechanism for scramming the reactor. When the Safety Ampl~fierreceives a scram signal, it stops supplying current to the electromagnets that hold the control rods in positlon. Without the magnets, the control rods gravity-fall to the fully inserted positlon.

The Safety Amplifier also receives scrams that are not internal to the Safety Power Drawer. Following are the scrams provided to the Safety Amplifier-

• Hlgh Power (125% Full Power)

Low Voltage (150V)

Per~od(c3sec)

Fuel Temperature (975")

Manual (Console)

Bridge lock scram Various Experiment Scrams allow experimenters to Independently and locally scram the reactor. These are manual scram buttons located as follows:

o Beam Port Areas o Irradiation Cell o Reactor Bridge o Pool Slde Interlocked scrams that ensure the reactor is shutdown when:

o Beam Port 4 Cave Door IS open and the reactor is near the Thermal Column Graphite Coupler BOX o Cell Door is open and the reactor is within 8 feet of the cell window 7.2.3.7 Fuel Temnerature Channel The Fuel ~ e m ~ e r a t uChannel re consists of a thermocouple embeddeddnin a-a a Fuel Temperature Instrument. A second temperature indicator on the Reactor Console, with a thermocouple selector switch is available to read out the temperature of thermocouples in the fuel, the pool water and the irradiation cell.

Fuel Temperature is also available in the Reception Room normally from a lower reading thermocouple. Figure 7-4 shows a diagram of the Fuel Temperature Channel.

CELL T C 7

FOOLTC \ 4 DIGITA INDCATOR

/ I DlGITAL IhQlCATOR Figure 7-3: Fuel Temperature Channel The Fuel Temperature Channel is the Limiting Safety System with the LSSS at 975°F. If the Fuel Temperature Channel is not available, the Safety Channels can act as the Limiting Safety System.

The Fuel Temperature Instrument performs the following functions:

Scrams the reactor if the thermocouple temperature reaches 975°F Captures peak fuel temperature during pulse The Fuel Temperature Instrument normally operates in continuous indication mode. It captures and displays peak pulse temperature only when operating in the peak mode.

~ h e ( 1 is located

~ adjacent

) to the central bundle excluding the comer positions and observed temperatures are proportional to maximum fuel temperature experienced by the fuel. This IF can be in any of eight locations in the core. Three chromel-alumel thermocouples, embedded in the IF, are located at the vertlcal center and one inch above and below the vert~calcenter and 0.3 inches from the center (Figure 4-7).

7.2.3.8 Preset Timer The Preset Tmer scrams the Translent Rod 15 seconds or less after a pulse The deslgn of the Preset Tmer 1s such that the actual tlme is adjustable. but the maxlrnum sett~ngis 15 seconds.

The purpose of the Preset Tmer is to prevent the reactor from restarting following a pulse 7.2.4 System Performance Analysis The instrumentation and control systems have been i n routme operatlon for over 40 years Solid-state electron~cs have replaced nearly all measuring Instruments The result 1s improved rellabllity.

Limiting Safety System Sett~ng,L ~ m ~ t l nCondltlons g for operatlon, surveillance requirements and actlon statements concerning the control and instrumentation systems are In the Technical Specifications 7.2.5 Conclusion The Safety Power Channels and the Fuel Temperature prevent exceed~ngthe operating llmlts for fuel temperature and reactor power. The operatmg llmits for temperature and power mdependently protect from exceedmg the Limiting Safety System Setting Other scram cond~tions~nclud~ng loss of ac power, Loss of Safety detector voltage and Manual Scram ensure that the safety equipment wdl operate as planned.

7.3 Reactor Control System The Reactor Control System consists of Rod Control Modules, Control Rod Dnve Mechanisms and external inputs (i.e. s h ~ midshim out signal and ~nterlocks)to the Rod Control Modules for Steady State and Pulse modes. The Rod Control Systems for each Control Rod consist of a hold-down device, Control Rod barrel, electromechanical Control Rod Drive Mechanism (CRDM), Rod Control Module and associated control circuits.

The Rod Control System performs the following functions:

Provides method for controlled add~tionof reactivity Provides scram capabdity Holds Control Rods and Control Rod bundles in position Provide indication of Control Rod posit~on Provides Rod Withdrawal interlocks Rod Control Modules in the Reactor Console:

Provide signal for indwidual rod motion Provide Logic interlocks Provide rod and carriage position indication Each Control Rod Drwe Mechanism (CRDM) motor drives either a lead screw (Shim Safeties and Regulating Rod) or a chain-driven externally threaded cylinder (Transient Rod). The CRDM couples the rod extension to the carnage to move the control rod. The hold-down assembly assures control rod bundle remains in place.

The CRDM recelves a signal that controls carriage motlon from its associated Control Rod Module. This signal controls the motor, which is the source of rod motion. When the rod is coupled to the carriage, the motor controls rod motlon; when it is not coupled, the rod remains fully inserted regardless of carriage positlon The CRDM also has various swltches that provlde information to the Rod Control Module about the status of the rod (the following sections explain these In detail).

7.3.1.1 Shim-Safety Rod Control The rod control system for the Shim Safety control rods allows the operator to control these four rods individually or as a group. Each rod drive has a Rod Control Module, Control Rod Drive Mechanism (CRDM), control rod barrel

with offset and hold-down tube In add~tion,the Shim Safety Rod D r ~ v eSystems share control circuitry for mterlocks ;~ n group d motion and a Power Supply for rod motion.

F~gure7-4 shows the control rod magnet and armature and rod assembly-dampening device for the sh~m-sal les The piston act~onprovides dampening of the control rod towards the end of its fall into the core. Water relit slot in the barn:I allow the rod to drop freely untd the rod begins the last six mches of travel. At this pomt, the p ton ring forces the water out of the bottom of the control rod barrel. When the piston enters the piston receiver, le reduced clt:arance dampens the rod's fall Stainless steel and alummum components provide smooth mover :nt a reduced wt:ar of slidmg parts

The shim-safety control rods have e ~ g hopt~onal t control rod posit~ons The offset assembly funct~o ns s in~larl

~ Y to the 1>oh act~onof a rifle. Spacers, w h ~ c hseparate the two plston rods, are In slots that allow vertlcal mov'emen11but rest101ct lateral movement. The offset barrel can rotate In 45" Increments. The hold-down assembly proviides a mea ns to enclose the control rod extension and prevent acadental lifting of a fuel element The hol d-do\an tul exte nds downward to the reactor core and fits over the upper end the cross bar of the fuel bundle The control rods attach to a horizontal plate on the upper port~onof the reactor frame structure w ~ t hmaclm e d slots and clamps to hold the rod drives in posltlon (F~gure7-5). A support rmg holds the S h ~ mSafety Ccmtrol Rod asse mbly. This assembly perm~tsremoval of the assoc~atedcontrol drlves for maintenance without movimg th e asscmated control rod from the core. F~gure7-6 shows the installat~onof a shlm-safety control rod

Each shlm-safety rod mechanism connects to a Rod Control Module at the reactor console. Push buttons permt operation of each rod drive independently of the other control rods. A Gang Switch, located on the Reactor Console near the Modules, permits operatmg all Shim Safety Control Rod drives simultaneously. The Gang Switch does not necessarily override the ind~vidualrod drive buttons on the individual Modules. Rather, an IN signal -from either the individual or Gang switch- overrides an OUT signal.

Travel speed for the shlm-safety rod drives is 11.3 centimeters per minute.

To provide rod height indication, the Module receives a signal from a dlgital encoder that rotates with the lead screw through the stepplng motor drive shaft The Module uses thls signal to provide Carriage Height indication in unlts of percent withdrawn and to provide logic interlocks for carriage full out (100%) and carriage full in (0%). If the rod is coupled to the carriage (as indicated by the Engaged light), rod height and carriage helght are the same.

The Rod Control Moddes for the Shlm Safety Control Rods perform the following funmons:

Provides dlgltal ~nd~catlon of carrlage height Provldes Rod InlRod Out signal to CRDM Probldes mdlcatlon of the followng o Rod Engaged o Rod Down o Rod Jammed o Carnage Down o Carriage Up Resets the rod posltlon mdlcation to 1 0% when:

o The engaged switch changes from disengaged to engaged

-while-o Carriage 1s drivmg In

-and-o The Rod Down swltch is made Provldes the following logic interlocks o Prevents rod ~nsertlonfor jammed rod o Prevents rod lnsertlon ~f Carriage height is 0 0

-and-Rod 1s engaged o Prevents rod wlthdrawal lf rod height 1s 100.0 o Prevents rod withdrawal if the gang switch is in the Gang Down position Provides lndlvidual rod wlthdrawal and rnsertion capablllty Interlocks associated with the SItrnl Safety Control Rods are as follows.

Rod Jammed o Prevents driving carriage down when lead screw presses the Jam switch

• Rod Down o Prevent drlvlng carrlage down when:

Carriage height indlcat~onat 0.0%

-and-Rod engaged Rod Out Interlock o Prevents rod wrthdrawal lf rod height 1s 100.0%

Rod In Override o If Gang Switch is in the Gang Down position or the individual Rod Down button is pushed, the rod will drive in regardless of the positlon of the other switch Shim Safety Pulse Interlock o Prevents withdrawing Control Rods in the Pulse Mode Low Count Interlock o Prevents withdrawing Control Rods with c 4mW on the Log Power Channel

Flgure 7-7: Control Rod Drwe Mechanism for Shim Safety Control Rods The Control Rod Drive Mechanisms (CRDM's) (Figure 7-7) are electromechanical assemblies that provide the motive force to move control rods and hold them In position. The motor provides the movement, while an electromagnet (attached to the bottom of the lead screw) holds the rod. When energized, the electromagnet holds the iron armature against the Engaged switch Each Shim Safety CRDM has an Engaged, Jam and Rod Down switch. These three physical switches perf(Irm the following and have the following physlcal descriptions:

Engaged Switch o Functions Llghts Engaged light on Rod Control Module

Provides mput to rod posltlon reset circult to reset indlcat~onat 1 O% when the control rod extenslon presses the Engaged swltch.

= Provldes Input to Rod Down Interlock (prevents further insertion ~frod at 0 0% and rod engaged) o Descriptlon Push button on the bottom face of the electromagnet Rod Down Switch o Functions Lights Rod Down I~ghtwhen control rod is <5% withdrawn Provldes input to rod positlon reset circult o Descriptlon

= Magnetlc reed switch external to the CRDM and adjacent to the armature on top of the control rod extenslon assembly Rod Jammed o Functions L~ghtsRod Jammed hght Provldes input to rod jammed mterlock, which prevents rod msertion lf rod is jammed o Descriptlon A mlcro-switch pressed when the lead screw drives in without the carriage lowering 7.3.1.2 Transient Rod Control The rod control system for the Translent control rod (Translent Rod) allows the operator to control this rod individually in both the Steady State and Pulse Mode. The rod drive system has a Rod Control Module, Control Rod Drive Mechanism (CRDM), control rod barrel with hold-down tube. In addition, the Transient Rod Drive System shares control circuitry for lnterlocks with neutron detection instruments and other Rod Control Modules.

Figure 7-8 shows the pneumat~c-electromechanicalControl Rod Drive Mechanism for the Translent rod. The pneumatic portlon of the CRDM is a single acting pneumatic cylmder. A piston within the cylinder attaches to the transient rod by means of a connecting rod. The piston rod passes through an air seal at the lower end of the cylinder. Compressed air, admltted at the lower end of the cylmder, drives the piston upward. As the piston nses, the compressed air above the piston exhausts through vents at the upper end of the cylinder. During the final inch of travel, the shock absorber slows the rod to minimize mechanical shock when the piston reaches its upper limit stop.

An accumulator tank mounted on the reactor bridge stores compressed air for operating the pneumatic portion of the CRDM. A three-way solenoid valve controls the air. De-energizing the solenoid valve interrupts the air supply and relleves the pressure in the cylinder so that the piston drops to its lower limlt by gravity.

Following describes the operation of the Transient Rod for both modes of operation.

Steady State o High-pressure air actmg on a piston holds the transient rod against its rod dr~vecarriage (specifically the shock absorber)

Pulse o Hlgh-pressure air pushes the transient rod rapidly to the carriage position o Air vents when the Preset Timer times out

The transient rod dribe attaches to a support frame that bolts to the reactor bridge. The Transient control rod is in the center of the core The hold-down assembly provldes a means to enclose the control rod extension and prevent accidental llftlng of a fuel element The hold-down tube extends downward to the reactor core and fits over the upper end of a control rod-gulde tube.

The Translent rod mechanism connects to a Rod Control Module at the reactor console. This module 1s similar to those for the Shim-Safety control rods. Push buttons permlt operation of the rod drwe.

Travel speed for the Translent rod drive 1s 11 3 centimeters per mlnute To provide rod height indication, the Module recelves a slgnal from a chain driven dlgital encoder that rotates wlth the motor and worm assembly. The Module uses this slgnal to provide Carriage Height indication in units of percent withdrawn and to provide logic rnterlocks for carrlage full out (100%) and carriage full in (0%). If the rod 1s coupled to the carriage (as indicated by the Air Apphed hght), rod height and carrlage height are the same.

The Rod Control Module for the Translent Control Rod performs the followingfiinctions-Prov~desdlgital indlcatlon of carriage height Prov~desRod InRod Out signal to CRDM Provldes Indication of the following o Air Applied o Rod Down o Carriage Down o Carrlage Up o TR Fire Ready Resets the rod position ind~catlonto 1.0% when o The Carriage Down swltch makes

-wh~le-o Carrlage is drlving in

-and-o The Rod Down switch 1s made Provides the following loglc interlocks o Prevents rod lnsertlon if:

Carriage height 1s 0 0

-and-

= Carriage Down swltch 1s made o Prevents rod withdrawal ~frod height 1s 100 0 Provides signal for pulsing For steady-state reactor operations, the electromechanical portlon of the transient rod drive controls the transient rod position. The pneumatic cylmder must be in the fully inserted position in order to apply alr to the plston for steady-state operation of the rod. Once air is applied, the pneumatic cylinder movement controls the transient rod at a rate of approximately 11.3 centimeters per minute.

Interlocks assoc~atedwith the Transient Control Rods are as follows:

RodDown o Prevent driving carriage down when:

Carrlage height indication at 0 0%

-and-

= Rod Carriage is Down Rod Out Interlock o Prevents rod withdrawal if rod height is 100.0%

Air Apphed interlock o Allows applying air when:

Mode Selector swltch In Pulse

-and-

Power 1s less than 1kW

-or-Mode Selector Switch is In Steady State

-and-

• Carrlage is down

-0r-Alr 1s apphed (Havlng air apphed satisfies the electronic 1og1c interlock and allows air to continue to be applied when rod is withdrawn In steady state. If alr IS vented, a ~ cannot r be re-applied untd the carriage 1s down )

TR Withdrawal o Prevents w~thdrawingControl Rods in the Pulse Mode Low Count Interlock o Prevents withdrawmg Control Rods with < 2cps on the Log Power Channel The Control Rod Drlve Mechan~sms(CRDM) (Flgure 7-8) IS an electromechanical assembly that provides the motive force to move the Translent Rod and hold ~t In p o s ~ t ~ o The n motor provides the movement, while high-pressure air holds the rod The Transient Rod CRDM has a Carr~ageDown and Rod Down switch. These physical swltches perform the following and have the following physical descnpt~ons:

Carriage Down Swltch o Functions Provides input to rod pos~tlonreset circuit to reset ~nd~cation at 1 0% when The carrlage presses the Carriage Down swltch

-while-

• Rod is dr~vingIn

-and-Rod isdown.

Provides input to Rod Down Interlock (prevents further ~nsertion~frod at 0.0% and carriage 1s down) o Descript~on Mlcro switch activated by carrrage Rod Down Switch o Functions Lights Rod Down hght when control rod 1s 4%withdrawn Provides input to rod posltlon reset circuit o Description Micro swltch act~vatedby the plston rod 7.3.1.3 Regulating Rod Control The rod control system for the Regulating control rod (Reg Rod) allows the operators to control the rod manually or automatically via the servo controller. The Reg rod has a Rod Control Module, Control Rod Drive Mechanism (CRDM), control rod barrel and hold-down tube. In addition, the Reg Rod shares control circu~tryfor interlocks and a Power Supply for rod motion with other control Rods.

The regulating rod control assembly is similar to the Shim Safety control rods in Agure 7-7 except that the barrel contams a lower guide piece with no piston action slnce the control rod extension bolts to the lead screw and does not scram The control rods attach to a horizontal plate on the upper portion of the reactor frame structure with machmed slots and clamps to hold the rod drives in posit~on(s~milarto the Shim Safety Control Rods). A support ring holds the Control Rod assembly. This assembly permlts removal of the associated control drives for maintenance without moving the assoc~atedcontrol rod from the core; however, since the lead screw for the Reg Rod physically attaches to the connector rod, the CRDM must be disassembled

The Reg Rod CRDM connects to a Rod Control Module at the reactor console Push buttons permit operation of the rod drive.

Travel speed for the Reg Rod is 11.3 centlrneters per mlnute.

To provide rod helght ~ndlcatlon,the Module receives a signal from a digltal encoder that rotates w t h the lead screw through the stepping motor drlve shaft The Module uses this signal to provlde Carnage Height ind~catlonin units of percent withdrawn and to provlde logic interlocks for carnage full out (100%) and carnage full In (07k) Rod helght and carnage helght are the same The Rod Control Module for the Regulating Rod performs the follow~ngfinctions.

Provides digital rndlcation of carrlage helght Provides Rod In/Rod Out signal to CRDM Provides indlcatlon of the following o Carrlage <20%

o Carnage >80%

o Rod Jammed o Carriage Down o Carrlage Up Resets the rod position lndicat~onto 1.0% when:

o Carriage 1s drlvlng in

-and-o The Rod Down switch is made Prov~desthe following loglc mterlocks o Prevents rod insertion for jammed rod o Prevents rod insertion if:

Carrlage height is 0.0

-and-Rod Down Switch is made o Prevents rod withdrawal ~frod height is 100.0 Provldes signal for Shimming Required alarm o Alarms when Red Rod 1s <20% for >80% fully withdrawn Interlocks associated with the Regulating Rod are as follows:

Rod Jammed o Prevents driving carriage down when lead screw presses the Jam switch RodDown o Prevent drlv~ngcarriage down when:

Carnage height indication at 0.0%

-and-Rod Down switch is made Rod Out Interlock o Prevents rod withdrawal if rod height is 100.0%

Shim Safety Pulse Interlock o Prevents withdrawing Control Rods in the Pulse Mode Low Count Interlock o Prevents withdrawing Control Rods with < 4mW on the Log Power Channel The Control Rod Drive Mechanism (CRDM's) for the Regulating Rod operates exactly like the Shim Safety Control Rod Drwe Mechanisms except that there is no electromagnet.

The Reg Rod CRDM has a Jam and Rod Down swltch. These physical switches perform the followng and have the following physical descript~ons-Rod Down Switch o Functions

a Prov~deslnput to rod posltlon reset clrcuit o Descr~pt~on Micro sw~tchactivated by carr~age Rod Jammed o Funct~ons Llghts Rod Jammed llght Provldes input to rod jammed interlock, whlch prevents rod insertion ~frod 1s jammed o Descr~ption A m~cro-swltchpressed when the lead screw drives In wlthout the carnage lowermg 7.3.1.4 Mode Selector Switch The Mode Selector Sw~tchselects between Steady State and Pulse modes It ensures the appropriate interlocks for both pulse and steady state and prevents puls~ngIn Steady State Mode.

The Mode Selector Swltch has two positions that provlde the following functions.

Steady State o Prevents applylng Translent Rod air unless the following condltlons are met.

Transient Rod Carr~ageDown

-0r-Transient Rod Air applled (Having alr applied satisfies the electronic logic interlock and allows air to continue to be applied when rod is withdrawn in steady state. If air is vented, air cannot be re-applied unt~lthe carriage is down )

Pulse o Allows Translent Rod Air apphed o Activates Preset Timer Preset Timer scrams the Transient Rod less than 15 seconds after applying alr o Prevents control rod withdrawal o Disables Safety Drawer ampl~fiers o Bypasses Perlod Scram 7.4 Reactor Protection System Table 7-1 indicates the minimum reactor safety circuits and interlocks that are necessary for reactor operation.

Failure to comply with any of the safety criteria will result in an immediate reactor scram Table 7-1: Minimum Reactor Safety Channels Number Effective hlode Safety Channel Function Operable Steady-State Pulse Fuel Element Temperature SCRAM @ LSSS 1 X SCRAM @ 125% 2 X Safety Power SCRAM on loss of supply voltage to 2 X detector power supply Console SCRAM Button SCRAM 1 X X Transient rod SCRAM less than X Preset Timer 1 15seconds after pulse Prevent shim-safety withdrawal at less X Log Power 1 than 4 x lo5 W Prevent application of air unless fully X Transient Rod Air Apply 1 inserted Shim-safety & Regulating Prevent withdrawal in Pulse Mode 1 X Rod Pulse Interlock

7.5 Engineered Safety Features Actuation System There are no engmeered safety features actuat~onsystems 7.6 Control Console and Display Instruments The NSCR operates In two standard modes: Steady State and pulse. Steady-state mode 1s for operation at power levels up to 1000 kW (thermal). Pulsed mode is for the condlt~onresulting from the rapid withdrawal of the transient rod, which introduces a step lnsertlon of reactivity that results In peak powers of up to about 1,600,000 kW.

The reactor console displays all pertinent reactor-operating conditions and allows for reactor control. The console also d~splaysinformation about the cooling system, environmental monitoring and experimental facllltles. The control system consists of five power measuring channels utilizing three uncompensated ion chambers, one compensated Ion chamber and one fission counter. Descrlptlons for the specific controls for the reactor and associated water systems are In those sectlons At all times when the console is turned on, a licensed reactor operator or llcensed senior reactor operator will be in the control room. Reactor operators in training may operate the reactor in the presence of a llcensed reactor operator or llcensed senior reactor operator in the control room All fuel addltlons to the reactor core and crltlcal experiments require the presence of a member of management as designated by the director.

Table 7-2 lists indications and controls on the reactor console,

Table 7-3 hsts alarms displayed on the man reactor console.

Table 7-2: Sumrnarg of Information Displayed and Recorded on Reactor Console Reactor Safety S)stems Control Indication Record Log Power: Power Ind~cation I X X Log Power: Period Indlcat~on 1 X I I L~nearPower I I X l X l I Safety Amplifier X I Fuel Temperature I X I X Rod Drives X X I Mnnual SCRAM I X l X l I Other SCRAMS I X I Facihty & Reactor Condit~onalAlarms X Water Systems I Pool Water Coohng System X 1 X 1 Pool Recirculat~onSystem X X Pool Skimmer System X X Diffuser Svstem X X Transfer System Secondary Treatment System X

X I X X

I 1

[ Air Handling System Shutdown X X 1 Emergency Evacuation Horn X X Irradiation Cell Exhaust X Televis~onMonitors X Facilitv "Door Oven" Alarms X Experimental Facilities Pneumat~cSystem X X Sample Rotisserie Motor X X "C-2" Exueriment Personnel Control Alarm X X

Table 7-3: Summary of Alarms Displayed on Reactor Console Alarms Bridge Unlocked Fuel Temperature Scram Perlod Scram Safety Amphfier Scram Manual Scram Experiment Scram Manual Scram I Servo Fault I I Perlod Scram Bypass 1 Bridge Interlock Air Handler shutdown Bypass Re~ulatlneRod S h i m m l n ~Reaulred I Area Radlatlon alarm 1 Fachty Alr Monitoring Emergency shutdown a n Handling System Buildlng Pressure Svstem Fadure Cell Door Open Pool Level Alarm 7.7 Radiation Monitoring Systems Two systems for monitoring radlation in the facillty are Area Rad~atlonMonitors (ARM) and Facility Air Monitors (FAW 7.7.1 Area Radiation Monitors (ARM's)

Area Rad~ationMonitors are located throughout the faclhty to monitor levels in areas where radlation levels could exceed normal levels One ARM above the reactor provldes radlatlon levels In the reactor bay area. Other ARM's are by the beam ports accesses, demmerallzer room and radloactlve sample handing areas.

The ARM's provide audlble and visual for Alert and Alarm. Operators can adjust these alarm settings in the Control Room The indicators are In the Reception Room (Emergency Support Center), Control Room and locally for each ARM.

7.7.2 Facility Air Monitors SIXFacility Air Monltors (FAM's) detect airborne activity In both gaseous and particle form. They monitor air in the building and leaving the building. Following is a llst of the detectors by channel includmg their functions and sample points.

1) Channel 1 - Stack Particulate a Monitors for radioactive particles in the air entering the exhaust stack

b. Automatically shuts down the air handllng system

2) Channel 2 - Fission Product

a. Monitors for radioactive particulate above the reactor core

b. Automatically shuts down air handlmg system

3) Channel 3 - Stack Gas a Monitors for Ar-41 entermg the exhaust stack

4) Channel 4 - Buildmp Particulate a Monitors for rad~oactlvepart~clesin the confinement buildlng 5 ) Channel 5 -Xenon Monrtor (Note Shares a detector with Channel 3)

a. Monitors for Xe-125 enter~ngthe exhaust stack b Automatically shuts down alr handlrng system

6) Channel 6 - Bulldlng Gas

a. Monltors for Ar-4 1 In the confinement bu~ldlng Each FAM channel provldes indication in the FAM Equipment Room, the Control Room and in the Recept~onroom Each FAM channel also provides an audible Facil~tyAir Monitoring alarm in the control room and an alarm light in the Reception Room. The FAM channels that shut down the air handlers also provide an Emergency Shutdown Air Handling System alarm in the control room.

T.A. Godsey and J.D. Randall, "A Solution to the Varying Response of the Linear Power Monitor Induced by Xenon Poisoning," Presented at TRIGA Owners Conference 111, Albuquerque, NM, 1974

8 ELECTRICAL POWER SYSTERIS 8.1 Normal Electrical Power Systems No electrical power supplies are cr~tlcalfor malntainlng the faciltty in a safe shutdown condit~on Figure 8-1 shows the electrical dlstributlon system for the Nuclear Science Center. Texas A&M plant servlces supplies electrical service to the facility from the distribution system through power poles on the NSC site Emergency disconnects are in place at the transformer stations on the NSC site 480 VAC 3 Phase Electrical Power Power panels MCC'MA' and MCC'MB', in the mechanical equipment building, supply the majority of the loads in the reactor and laboratory buildmgs MCC'PA' IS located in the heat exchanger room in the lower research level and supplies power to the reactor cooling system equipment motors MCC'RA' 1s located on the chase level of the confinement buildmg and supphes the majority of the loads on the chase.

120f208 VAC Electrical Power LB panels LB'A' and LB'B', located In laboratory seven Laboratory building, supply loads for the laboratory area 8.2 Emergency Electrical Power Systems Rechargeable, battery-operated emergency floodlights are located throughout the building. In the event of a power falure, these lights, which are normally off, provide sufficient l~ghtingto permit evacuation of the reactor buildmg or the performance of emergency activities in the building

Figure 8-1: NSCR Electrical d~stributlon 9 AUXILIARY SYSTEMS 9.1 Heating, Ventilation, and Air Conditioning Systems 9.1.1 Heating and Air Conditioning Chillers and a furnace provlde chllled water and hot water for the ventilation system. As alr enters the buildlng, ~t flows through a heat exchanger wlth both hot and cold water from the furnace and chdler. Thls heatmg and coolmg system removes humidity from the air and malntalns the bulldlng at a comfortable temperature 9.1.2 Air Handling Units Four alr handlmg unlts and an exhaust fan control airflow, pressure, temperature and humidity wlthln the reactor buildlng. The fachty has three zones of negative pressure for effectwe lsolatlon of possible airborne radloactlve material. Following is a llst of the three zones and the areas they cover.

1) Least negative pressure zone a Control Room and locker areas where contamination is least likely

2) Intermediate zone of negatlve pressure a The main research areas where infrequent contamination might occur

3) Maxlmum negative pressure lncludes areas where radioactive contammation is likely a Beam ports b Thermal column

c. Through tubes The Air Handling units supply alr to the following areas of the building:

1) Air handler A a Upper research level

2) Air Handler B

a. Lower research level

3) A n Handler C

a. Control room

4) Air Handler D a Restrooms b Electronics shop The Central Exhaust Fan takes suction on all areas and discharges dlrectly to the stack or through an Emergency Filter Bank. The height of the exhaust stack above ground level is 85 feet.

A Bypass Damper, at the suctlon of the Central Exhaust Fan, controls pressure in the building Controls for all four units are in the control panel in the reception room. An interlock prevents running units A, B, C or D unless the Central Exhaust fan is running. This ensures the building will not be at a positive pressure.

The Central Exhaust Fan will shut down and the inlet and exhaust dampers close when:

The FAM's generate an Emergency Shutdown signal The Air Handling Shutdown button is pressed on the Reactor Console The H ~ g hTemperature Sensor above the Emergency F~lterBank exceeds its set point When the Exhaust Fan shuts down, the rest of the air handlers also shut down

9.1.3 Dampers and Filters Dampers are located at the alr Inlet to all alr-handlmg un~ts,the fresh alr bypass to the exhaust fan. and in the exhaust stack. In cases of emergency, a sw~tchIn the reactor control room can close these dampers and s~multaneouslysecure the alr handlers to sola ate the buildmg and stop alrflow (see 9 1.2). An Emergency Exhaust Air Fllter Bank is between the exhaust fan and building stack. The Emergency Fdter Bank conslsts of two part~culatefilter banks and one bank of activated carbon filters Controls for the filter bank are in the Recept~on Room 9.1.4 Emergency Operation The Emergency Control Panel in the Reception Room provides all the controls for operating the ventilation system for both emergency and normal operations 9.2 Handling and Storage of Reactor Fuel Techn~calspeclficat~onsrequlre stored fuel to be In a configuration with kffto be less than 0 8 for all conditions of moderation. The storage arrays for irradiated fuel permlt sufficient natural convection coolmg by water or alr such that the fuel elements or fueled dev~cetemperature will not exceed design values.

Fuel elements are stored and handled in two general areas at the NSC. These areas are the fuel storage room and the reactor pool handllng and storage areas Unirrad~atedfuel can b e temporarily stored in approved shipping containers used by the fuel manufacturers for shipment In the reactor.

9.2.1 Fuel Handling 9.2.2 Fuel Storage 9.2.2.1 Fuel Storage Room or Fuel Vault 9.2.2.2 Reactor Pool Storage Areas Special storage facilities prov~def o and fuel-followed control-rod storage.

9.2.3 Fuel Bundle Maintenance and hleasurements The Maintenance J I supports

~ the entire fuel bundle and prevents any lndlv~dualelement from falling when 11 1s unscrewed from the lower gulde of the bundle This j ~ gprov~desaccess to the lower end of the fuel elements T h ~ s allows operators to completely d~sassembleand reassemble a bundle In the j ~ g .

Once an element 1s unscrewed from the lower pulde, the single element tool holds the fuel element for visual Inspection.

The Fuel Measuring Device provides a method of a golno-go test for transverse bend and length measurement for element elongatlon. An element will not fit into the devlce ~fthe transverse bend exceeds 0.125 Inches over the length of the claddmg. The dev~ceholds one fuel element and allows operators to measure the length difference between a given fuel element and the standard. The difference between these two changes over the life of the element and provides the elongatlon information 9.3 Fire Protections Systems and Programs Smoke detectors that alarm off-site and numerous fire extinguishers throughout the faclllty provlde fire protection at the Nuclear Science Center Add~t~onally, the College Statlon F ~ r Department e provtdes the NSC with fire protection servlces and 1s on call twenty-four hours a day. Flre department personnel receive train~ngin radiological hazards and NSC site famil~ar~zatlon.

9.4 Communication Systems The NSC is equipped with several commercial telephone lines, all of which are available in the Control Room, Emergency Control Center (Reception Room) and several other locations within and outslde the Reactor Bulldlng.

The system allows public address from any telephone.

Two-way radlos provide additional communication between the NSC and the 24-hr staffed Communications Center at Texas A&M.

Anally, the Communications Center malntains an emergency recall roster listing home phone and pager numbers for key personnel 9.5 Possession and Use of Byproduct, Source, and Special Nuclear Material The NSC receives, possesses and uses, In amounts as required, any byproduct material without restriction to chemical or physlcal form that has a definite research, development or education purpose. It may also have any byproduct material generated by the licensed activities, but may not separate such fueled byproduct material.

All activities covered by the NSC license take place on the NSC site and adjacent NSC controlled facilities.

9.6 Cover Gas Control in Closed Primary Coolant Systems The NSCR does not have a closed primary loop; ordinary light water at atmospheric pressure in an open pool is the primary coolant. Therefore, no cover gas control is necessary.

9.7 Other Auxiliary Systems There are no other auxiliary systems requlred for safe reactor operation

10 EXPERIMENTAL FACILITIES AND UTILIZATION 10.1 Summary Description 10.2 Experimental Facilities 10.2.1 Beam Ports Five permanent beam ports of Type 304 stainless steel are cast into the pool wall at the lower research level. Bea port 5 is in the north wall of the main pool (F~gure10-1); the other four beam ports are in the stall end of the pool The thermal column modification accommodates a film irradiation system. The film irradiation system displaces beam ports 6,7 and 8 in Figure 10-2. Figure 10-2 shows the stall section of the pool with an unmodified thermal column Normally, Beam Ports 2 and 3 are flanged at the inner pool wall See the Film Irradiation section for specifics on the thermal column Figure 10-3 shows the arrangement wlth the Thermal Column extension and Graphite Coupler Box.

The beam ports are stamless steel in sections of SIX,ten, and nineteen Inches in dlameter dlvlded longltudlnally into three. tic0 and one-half. and one foot segments, respectively Thls deslgn prevents neutron streaming when concrete shield plugs are in place. The six and ten inch sectlons have one-quarter inch Boral lining wlth the exception of the SIX-lnchsection of beam port 4 Each of the above ports ends flush wth the external face of the pool wall and is sealed by a hrnged two-foot square, four-mch thick, carbon steel clad lead door. The doors are equlpped with an O-rlng seal and tightening lug to provlde a water barrler in the event of port flood~ng.A mlcro swltch actuates an enunciator light on the console when these doors open.

Beam port plugs are alummum cylmders filled wlth barltes concrete, each about one foot long wlth a handle recessed In the exposed end for ease of handllng Three of these plugs can fit Into the SIX-inchdiameter beam port-sectlon and two can fit Into the ten-inch dlameter section when the port IS not in use. A nineteen-inch dlameter, one-foot thlck sectlon 1s available to plug the final recessed section of the beam port.

Each beam port has a two-inch dlameter pipe connecting it to the central exhaust system, which maintains a constant negatlve pressure in the tube The vent connection to the tube 1s nearer the inner pool wall to ensure the removal of any gases before they can reach the external end of the tube These beam ports enable a varlety of experiments such as the extraction of a well-coll~matedbeam of neutrons and/or gamma rays from the reactor. Varieties of extenslons can attach to the beam ports as required by different experiments The extensions prevent interference with the movement of the reactor frame and grid plate when the graphlte coupler box is not in place. A short extension suspended between the tips of beam ports 1 and 4 and the graphite coupler box removes water from between the coupler box and the beam port. Beam ports 2 and 3 are radial ports with extenslons removed A "C-2" alarm device indmtes to the Control Room when personnel enter into the beam port areas A system to magnetically lock doors to the lower research level is available for activation during beam port usage Two separated segments of a single through tube, constructed of 303 stainless steel, penetrate the stall section of the pool. Their construction is essentially identical to that of the beam ports except that they have no boral liners or outer doors. Smce the tubes sit along a collinear axis, a straight six-inch dlameter connectmg tube can be bolted to the flanged pool ends of the tubes providing a continuous six-mch diameter passage completely through the pool Concrete plugs similar to those described above provide the necessary shielding in this tube to prevent streaming of radlatlon The through tube also vents to the central exhaust system.

The through tubes can facilitate transit experiments that pass through this tube or fixed experiments Each segment may be used as a separate beam port by fitting an extension tube between the reactor and the end of the through tube segment 10.2.2 Thermal Column The stainless steel and aluminum thermal column is located in the east end of the stall portion of the pool. It consists of a three and one-half-foot square section on the inside of the pool that enlarges to a four-foot square opening on the experimenter's side that penetrates the pool wall at core level. The walls of the thermal column are welded to the stainless steel pool h e r . An aluminum cover plate with gasket, bolted to the inside flange of the cavity, provides the water seal.

A graphite coupler box, adjacent to the thermal column couples the thermal neutron flux from the reactor. The reactor can operate wlth its east face adjacent to the coupler box for maximum thermal neutron density to the thermal column and the beam ports.

A vent h e from the thermal column cavity extends dlrectly to the central exhaust system, where the air goes through the Facility Air Monitoring system before leavlng the building. A movable thermal column door, constructed of lead and concrete shielding material, is on tracks embedded In the lower research level floor.

The current thermal column arrangement makes it a film madlatlon fachty. The film Irradiation system extends intk the thermal column penetration Into the pool so that film comes close to the graphite coupler box.

10.2.3 Pneumatic System The NSC pneumat~csystem consists of an electronic control, exper~mentreceivers, in-core receivers, a gas supply system and interconnect~ngtubing The control system allows the experimenter to control the length of irradiation and the control room operator to prov~deor prevent permit to the experimenter. The experiment receivers provlde a means for the experimenter to load and retrleve the samples The in-core receivers receive and support the sample in the core. The gas supply system prov~desthe pressure to 'shoot' the sample into the core and return the sample from the core The pneumatic tube Itself cons~stsof a core receiver, polyethylene tubmg, protective metal sheathing at the reactor b r ~ d g eand a receiver in any of several laboratories (Figure 10-4) The pool wall pneumatic penetrations in Figure 10-1 are unused because of inconven~encein maintaming the system within the reactor pool. At present, the pneumatic system lines enter the pool at the reactor bridge and pass over the top of the pool walls.

The pneumatic tubes are for the product~onof short-lived radioisotopes primardy to support neutron activation anal ysls.

10.2.4 Irradiation Cell The irradiation cell is located at the west end of the reactor pool. This cell is approximately eighteen feet wide by sixteen feet deep by ten feet high. The frame for the concrete roof is an eight by eight inch steel I-beam column connected with six by fifteen inch steel I-beam joists. An overlay of four by six inch t~mbersprovide decking for the concrete blocks which are two by two by four feet. The blocks are stacked four feet high with an opening of approximately five by five feet left directly over the cell window. A motor driven concrete shield covers the opening the opening (Figure 10-5). The concrete roof of the lower irradiation cell provides the floor for the upper irradiation cell.

A steel ladder that extends from the observation deck to the upper irradlation cell provides access to the upper irradiation cell. An extendable ladder or an electro-mechanical lift provides access to the lower irradlation cell.

With the exception of an opening to accommodate a sample platform to carry samples and people to the lower irradiatlon cell, the observation deck over the cell is steel plate. A small-hinged section of the deck plate provides access to the ladder that runs from the upper level to the top of the concrete shield.

Concrete steps lead up to the observation deck on the south side of the pool. This area provides an excellent vantage point for facility visitors. The irradiatlon cell w~ndowis in the two feet thick wall that separates the cell from the reactor pool. The window is two feet square on the pool side and flares out to four feet square on the cell side. A one-half inch aluminum plate bolts to the pool side of the cell window to provide a watertight barrier. The pool side flange is large enough to prevent the cell window from projecting inside the reactor frame. A boral plate can be hung over the window to shield samples in the irradiation cell from excessive neutron flux; a box can also be hung over the window to accommodate various sources for gamma irradiations.

The breaker that supphes electrical power for the motor driven shield serves as a manual interlock for personnel safety. Locking the breaker open prevents openmg or shutting the shield door. Mechanical stops on the rail prevent inadvertent movement of the reactor closer than eight feet atyay from the irradiat~oncell wndow. In addltlon, a bridge mterlock provides a scram in the event the lrradiat~oncell door is open and the reactor 1s wlth~neight feet of the cell T o handle removal of 4 1 ~duer to activat~onIn the cell, an exhaust duct extends to the bottom of the cell for continuous removal of alr from the cell. The duct discharges to the central budding exhaust ahead of the stack gas monitor. The Facil~tyAir Mon~torsmonitor the cell air before release to the environment. The controls for the cell alr exhaust are located on the reactor console. An experimenter SCRAM button and an Intercom are located lns~de the cell.

10.2.5 Neutron Radiography Cave The neutron radiography facility is a concrete block structure on the lower research level located adjacent to the pool sh~eldwall. It contams and shields a thermal neutron beam extracted from beam port 4 (Figure 10-6). The cave structure surrounding the beam port provides for remote posit~oningof samples wlth the beam port In operation A hydraulic shutter at the beam port exit can shield the neutron beam between exposures. A sample preparation room and a dark room are available for loading and unloading cassettes and film processing.

An alarm in the reactor control room will alert the operator upon personnel entry into the sample preparation room film loadlng access area or the cave and a "C-2" device is visible to the person entering. An entry device on the cave door will cause a SCRAM ~fthe cave door opens when the reactor is against the radiography reflector (graphite coupler box). The bridge rail stop restricts movlng the reactor any closer than eighteen inches from the reflector when in place for cave entry.

10.3 Experiment Review

The NSC Standard Operat~ngProcedures (SOP'S) glve guidelines for review and approval of any new experiment or class of experiments. In addition, the Technical Specifications provide specific revlew requirements for the Reactor Safety Board The Senlor Reactor Operator on duty can author~zethe conduct of routlne experiments.

All experiments are subject to the llmitat~onsIn the Technical Spec~ficat~ons.

11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT Activities at the Nuclear Science Center wlll comply w t h 10 CFR Part 20 "Standards for Protectron Against Radiat~on". The NSC wdl assess and control exposure of rnd~v~duals and release of radloactlvity to the envlronment to maintain compllance \wth all appl~cablesectlons of the regulations. The following sections outlme methods and mstrumentatlon that establish compllance.

11.1 Radiation Protection The intent of the radlation protection program at the reactor faclllty is to malntain radiation exposures as low as reasonably achievable (ALARA) given current technolog~es The design of the experimental facilltles, reactor pool and the reactor shield includes protective measures and devices that limit radlation exposure and radioactive material release. The Standard Operating Procedures govern general requirements such as dosimeter use and records, certlficatlon of training, survey frequency, leak testing of sources, and ALARA program 11.1.1 Radiation Sources 11.1.1.1 Airborne Radiation Sources Releases from abnormal reactor operations The fuel retains its fission products, with releases to the envlronment only durmg a breach of fuel cladding. This posslbillty is one of the acc~dentsconsidered in Chapter 13 of this report in the analysis of the deslgn bas's accldent.

Releases form normal reactor operations Production of radloactlve gases, primarily 4 ' ~ rresult

, from the irradiation of air and dissolved gases in the cooling system, open beam port tubes, the dry tube, pneumatic irrad~atlonsystems, and the irradlatlon cell.

Nuclear Science Center Technical Report Number 32Em0r! Book"arknot defined. documents that the NSC releases approximately 4 7 Ci of 4'Ar on an annual basis. Applying a dilution factor of 5 x the releases produce approximately 0 8% of the effluent concentration of Argon-41 as specified in 10 CFR Part 20, "Standards for Protection Against Rad~atlon". The result of 4.7 Ci assumed 100 MW-day operation of the NSCR The EPA COMPLY' program lndlcates that the maximally exposed receptor would receive a dose of 0 2 r n r e d y for continuous discharge.

There are several important findings concerning 41Arproduction at the NSC.

1) On a long-term basis, the pool accounts for more than 95% of the facility's production.

2) The measured peak 4 1 ~concentration r in the Irradiation Cell for ind~vidualfour-hour runs were:

a. 6 . 7 ~ 1 0 "pCllcc with the cell exhaust fan on and,

b. 1 . 2 ~ 1 0pCi/cc

~ with the exhaust fan off.

3) The maximum 4'Ar release from firing the pneumatics system was 208 pC1; this occurred after the reactor operated at one MW for six hours. Firing the pneumatic system with the reactor shutdown for a long time resulted in a 6.8 pCi 4 1 ~release.

r Firing 5 times always purged the pneumatics system of argon.

4) As expected, the dry tube did not contribute to 4 ' ~ release, r but the beam port measurements showed a level of 2.15~10"pCikc at 1MW in beam port 1 closest to the core.

Although the pool is the major production source over a long period, the other sources can rlval the pool release rate on occasion

The buildlng central exhaust pulls the 4 1 ~Inr the beam ports and in the irrad~ationcell directly to the exhaust stack.

The " ~ from r air In the pool water goes from the pool Into the building and then through the bulldlng ventrlatlon system to the central exhaust The central exhaust disposes of the gaseous waste ~ n t othe environment through the 85-feet hlgh b u ~ l d ~ nstack.

g The equations for developing the dllution factors below are from F.A. G~fford,Jr. 2,3. These calculat~onsare based on release at ground level and utlhze the bulld~ngdllution factor ( D B= cAzf ), where A 1s the cross sectional area of the buildlng normal to the wind and 11 IS wlnd speed in meterslsecond The estimated value of c is 0.5. The cross sectional area of the Nuclear Science Center IS 357 m2 The equation for the atmospheric dllution factor is:

Equation 11-1 where X is the concentration in grams or curies per cubic meter; Q is the orlginal source strength In grams or curles per second; ri is the mean wind speed In meters per second; y is the crosswind in meters from the plume axis; 11 is the source helght In meters; and 0,and (rz are the dispersion coefficlents in m2.

By combining the bullding dilutlon factor, DB,with the atmospheric dllution factor and in the downwind direction

( y = 0).the formula becomes:

Equation 11-2 The average wind speed as determined from U S . Weather Bureau data for this location IS 10 mph. The following calculation utillzes dispersion coefficients of a, and 0,for stable conditions and a wind speed of 1 m/sec (2 mph) to determine the dilution factor available under pessimstic conditions (Q = 1) at a distance of 100 meters from the point of release.

Equation 11-3 1

X = Equation 11-4 7 ~ +O.5-357 2

X= 11203 Equation 11-5 This calculation indicates that the minlmum dllution at 100 meters is 200 under the most adverse conditlons. From the wind rose diagram shown in F ~ g u r e2-8, these conditlons are indicated approximately 10% of the time; however,

most calm cond~tlonsoccur at nlght whde the majorlty of operat~onsoccur durlng the daylight hours Assuming an average wind veloclty of ten mph. the dilution factor becomes x = 11903 Equation 11-6 Again, this 1s a pessimistic approach stnce the d~lutlonis at only 100 meters (approximate boundary of the NSC site). The calculat~onat 1500 meters under stable conditions and with a wind speed of ten mph yields a dllution factor of 6,920 With a wind speed of only 2 mph, the dllutlon factor 1s still 1,570 The calculations presented in t h ~ ssectlon clearly show that the Nuclear Sclence Center can use a ddutlon factor of 200 for stack release without endangering the publ~chealth and safety.

Section 7.7.2, Facility Air Monitors, hsts and summarizes the functions of the instruments that monitor airborne radioactivity in the NSC facillty and in the exhaust leaving the facility.

Sect~on7.7.1, Area Radiation Monltors (ARM'S), describes the system for mon~toringradlatlon level in the workspaces at the NSC.

Section 10.2, Experimental Facllitles, describes the Pneumatics System, Irradiation Cell and Beam Ports.

11.1.1.2 Liquid Radioactive Sources Nitrogen-16 (I6N) is the only isotope in the pool from normal reactor operations capable of exposlng personnel to hazardous levels of radlatlon The reactor produces a significant amount of 1 6 when ~ coolant passes through the core at power levels greater than 400 kW. The dlffuser system controls theI6N exposure (section 5 6) and reduces the dose rate at the pool surface to 2 to 3 mrem /h durlng full power operation. If the d~ffusersystem falls durlng full power operation, the dose rate at the pool surface is less than 100 rnremlh.

Radiation levels from the llquld radioactwe waste are extremely low and do not present radlatlon exposure hazards.

Section 11.2.2 addresses dlsposal of lrquid waste.

11.1.1.3 Solid Radioactive Sources The major source of radlation and radloactlvity is the fisslon product in the reactor fuel. Typical four-element fuel bundles will generate fields of 100 to more than 1000 Rlh In air at 3 ft ~fremoved from the reactor pool. As long as the fuel is contained withln the pool filled wth water, this source of radlatlon presents no personnel hazard. Chapter 13 considers a loss of pool water. The pool design makes a complete loss of pool water highly unlikely.

Other possible sources of significant radlation exposure from solid radioactive material are the fission gammas from the operating TRIGA core, samples Irradiated for isotopic production, reactor components which have spent a long tlme In or near the core and the reactor startup source. The non-fuel sources are all small compared to fission product activity in the operating core. Activity produced during irradiations is significant; and the NSC estimates final activities of samples before the irradiation Equipment and procedures are In place to deal wlth the activity after the irradiation is completed 11.1.2 Radiation Protection Program All personnel entering the facility will have appropriate personnel monitoring devices. Personnel momtoring devices may include beta-gamma and neutron film badges and pocket ionization chambers.

Protective clothing including coveralls, boots, shoe covers and gloves are available for use at the NSC. Use of protective clothing will be as prescribed by the health physics staff.

At least one shower connected to the "hot" drain on the lower research level provides for decontamination of personnel The laundry also has "hot" drains for cleaning contammated clothing.

A rad~oactlvematerral handling area. located adjacent to the reactor on the upper research level, prov~desa space for processrng and packaging mdloactrve materials. Protectrve clothlng and equ~prnentare available for use In thls area.

10 C I 3 Part 20 governs access control and posting requrernents for thls area.

A standard radiochemistry laboratory on the lower research level IS available for research exper~mentsand health physics use.

Texas A&M University and the Texas Department of Health have agreed to an environmental monrtoring program Through this program, NSC staff collects sediment samples from the NSC Creek and mrlk samples from a downstream Dairy. The NSC analyzes these samples for radroactivity and radrorsotope rdentificatron. Data from these samples have remained unchanged since 1974. None of the results shows a significant Impact on the environment.

NSC staff use portable survey meters to survey operations in restricted areas and dunng potentially hazardous experimental activities to assure personnel safety and compliance wrth 10 CFR Part 2 0 limits and local ALARA limits The NSC maintains approprrate counting equipment to survey for surface contamination on equrpment removed from the building, to determine extent of contamination in the event of a radioactive splll, to conduct a routrne rad~ologicalsafety surveillance program, and to conduct analyses of hquid waste and other samples 11.1.3 ALARA Program The NSCR standard operating procedures rnclude an ALARA plan and procedures that reflect the management's commitment to ALARA princrples. A Reactor Safety Board member, Ex-officro member or designee conducts an annual ALARA review and the Board reviews the results 11.1.4 Radiation Monitoring and Surveying NSC standard operating procedures dictate the requirements for periodic radration and contamination surveys.

Sect~on7.7 of this report describes installed radiation detectors and facrlity air monitors. In general, NSC Health Physics conducts regular surveys for both radiation levels and contamination, and the procedures and forms set requirements for radiation monltors when removing samples from irrad~atrondevices.

11.1.5 Radiation Exposure Control and Dosimetry NSCR standard operatmg procedures specify requirements on radiation control and dosimetry. NSC Health Physics staff administers the dosimetry program. TLD, film-badge or equivalent dosimeters detect exposures for operating personnel and students using the NSC on a regular basis, while pocket ion chambers monitor exposures for tour groups and visitors All Radiation, High Radiat~onand Very H ~ g hRadration areas are subject to the requirements of 10CFR20.

Visitors and tour groups receive very low radiation doses; and tour groups d o not have access in any area with dose rate exceeding 2 mremlhr. No student has ever received a measurable exposure from reactor operation.

Occupational exposures of operatrons and maintenance personnel have been low, seldom exceeding 1 Rem TEDE in a year.

11.1.6 Contamination Control NSCR standard operating procedures specify requirements on Contamination Control. NSC Health Physics monrtor for rad~oactivecontamination as noted in section 11.1.4.

11.1.7 Environmental Monitoring Environmental TLD on the fence surrounding the NSC provide indication of environmental radiatron levels.

11.2 Radioactive Waste Management NSCR standard operating procedures speclfy requirements for dealrng \wth radloactive waste 11.2.1 Radioactive Waste Control Liquid waste from radioactlve laboratory floor drams, laboratory slnks, decontammatlon showers, demlnerallzer regeneration collects in a sump In the Deminerallzer Room Sump pumps transfer the waste mto one of three 12,000-gallon liquld waste hold up tanks.

11.2.2 Release of Radioactive Waste A normal operation associated with the Nuclear Science Center Reactor generates sohd radioactlve waste in the form of gloves, paper towels, used laboratory equipment, sample containers, aluminum, and used experimental hardware.

The NSC accumulates Low-level sohd waste in plastic-lmed waste containers located at strategic points throughout the facility. When filled, these plastlc contamers remain sealed In the radloactive waste storage building (Flgure 2-2. NSC S ~ t e ) Short-lived radioactive waste decays and ends up as non-rad~oactivewaste in a local landfill Long-llved waste stays at the NSC awaiting final disposal.

Activated equipment normally stays In the high-level waste storage area adjacent to the outs~dewall of the irradlation cell If equipment is not reusable, ~tbecomes either short-hved or long-lived waste.

Low-level liquid waste originates from four primary sources at the Nuclear Science Center. These sources are: floor drains, laundry, showers, and laboratories on the lower research level; the demineralizer room filter and ion bed; condensate from air handlmg units on mechanical chase; and the valve pit sump in coollng equipment room.

Liquid waste flows through common headers to a liquld waste sump located below the grade of the lower research level. A sump pump transfers waste to one of three storage tanks located above grade 200 feet northwest of the building. These tanks have a total storage capacity of 34,000 gallons. Each tank 1s equ~ppedwlth an ~ n l e valve.

t outlet valve, volume indicator and samplmg line. There is a locked valve on the master outflow line. Fresh water is ava~lableto the master outflow line for dllutlon.

The NSC d~schargesliqurd waste from the hold up tanks to the environment as approved by the Texas Natural Resource Conservation Commission (TNRCC) discharge permit.

Liquid waste from the reactor building drains to the hot waste sump in the deminerahzer room. Two 100-gpm-sump pumps lift the liquid waste for storage in collection tanks located on the northwest comer of the reactor site. The sump pump is below the base elevation of the reactor pool. Liquid waste from the pool liner and cooling equipment room drains to the valve pit sump. The valve-pit sump pump transfers the water to the deminerallzer sump. F ~ g u r e 5-5 shows the liquid waste disposal system.

11.1 Bibliography I U.S. Environmental Protection Agency, COMPLY Program Rev. 2, October 1989 F.A. Gifford, Jr., Nuclear Safety, December 1960 F.A G~fford,Jr., Nuclear Safety, July 1961

12 CONDUCT OF OPERATIONS All operations involving the reactor will be conducted in compliance with the regulations specified in 10 CFR Part 50 and 10 Cl3 Part 55. The reactor will be operated within the hmits of the license and technical specificat~ons.

12.1 Organization The Nuclear Science Center 1s operated by the Texas Engrneering Experrment Station (TEES) The Dlrector of the Nuclear Science Center 1s responsible to the Dlrector of the TEES for the administrat~onand the proper and safe operatlon of the faclllty. Figure 10-1 shows the adm~nlstratlonchart for the Nuclear Science Center.

The Reactor Safety Board advises the Director of the NSC on all matters or pollcy pertaining to safety.

The NSC Radiological Safety Officer provides "onsite" advice concerning personnel and radiological safety and provides technical assistance and revlew in the area of radlatlon protection 12.1.1 Structure Director Texas Engineering Experiment Station pkq Safety Board L --

1 I

Manager Technical Reactor Operations t r S a E $ " d ~ ! e r Services Senior I ~ e a c t ooperators r I Figure 12-1: Organization Chart for Reactor Administration A h e management organizational structure provides administration and operatlon of the reactor facihty.

The Deputy Director of the Texas Engineering Experiment Station (TEES) and the Director of the Nuclear Science Center (NSC) have line management respons~billtyfor adhering to the terms and conditions of the Nuclear Science Center Reactor (NSCR) license and technical specifications and for safeguardmg the public and facility personnel from undue rad~ationexposure. The facil~tyshall be under the direct control of the Director (NSC) or a llcensed senior reactor operator.

12.1.1.1 Rlanagement Letels Level 1 - Deputy D~rectorTEES (Licensee) Responsible for the NSCR faclllty l~cense Level 2. D~rector(NSC) Responslble for reactor facility operation and shall report to Level 1.

Level 3: Senior Reactor Operator on Duty: Responslble for the day-to-day operation of the NSCR or shlft operation and shall report to Level 2.

Level 4: Reactor Operating Staff. Licensed reactor operators and senlor reactor operators and tramees These indlvlduals shall report to Level 3.

12.1.1.2 Radiation Safety A qualified, health physicist has the responslbillty for implementation of the radiation protection program at the NSCR. The individual reports to Level 2 management.

12.1.1.3 Reactor Safety Board (RSB)

The RSB is responsible to the Licensee for providing an independent review and audit of the safety aspects of the NSCR.

12.1.2 Responsibility Responsibihty for the safe operat~onof the reactor facility shall be In accordance with the line organlzatlon established above.

12.1.3 Staffing 12.1.3.1 The minimum staffing when the reactor is not secured shall be as follows:

1) A l~censedreactor operator wlll be In the Control Room ( ~senior f operator hcensed, may also be the senior reactor operator below)

2) A designated senior reactor operator shall be readlly available at the facility or on call (i.e., capable of getting to the reactor facillty wlthin a reasonable time)

3) A second designated person present at the site able to carry out prescribed written instructions.

4) The Dlrector (NSC) or his designated alternate is readily available for emergencies or on call (i.e ,capable of getting to the reactor facillty within a reasonable time).

5 ) At least one member of the health physics support group will be readily available at the facility or on call (i.e., capable of getting to the reactor facility within a reasonable time).

12.13.2 A list of reactor facility personnel by name and telephone number shall be readily available for use in the control room. The list shall include:

1) Administrative personnel

2) Radiation safety personnel

3) Other operations personnel 12.13.3 The following designated individuals shall direct the events listed:

1) The Director (NSC) or his designated alternate shall direct any loadmg of fuel or control rods within the reactor core region.

2) The Director (NSC) or his designated alternate shall direct any loading of an in-core experiment wlth a reactivity worth greater than one dollar.

3) The senior reactor operator on duty shall direct the recovery from an unplanned or unscheduled shutdown other than a safety limit violation.

12.1.4 Selection and Training of Personnel A tralnlng program for reactor operatlons personnel exists to prepare personnel for the USNRC Operator or Senlor Operator examlnatlon Thls trainmg program normally contalns twenty hours of lecture, outside study, and requlres several reactor startups 12.1.4.1 The selection and training of operations personnel shall be in accordance with the following:

a) The Director (NSC) or his designated alternate IS responsible for the tra~ningand requalification of the faclllty reactor operators and senior reactor operators

2) Requalification Program a) Purpose.

I) To Insure that all operating personnel mamtain proficiency at a level equal to or greater than that required for Initial kensing b) Scope:

i) Scheduled lectures, wrltten examinations and evaluated console manipulations insure operator proficiency.

12.1.5 Radiation Safety Members of the health physics staff routinely perform radiation safety aspects of facility operatlons, including routine surveying for radiation and contamination and sampling water and alr. Chapter 1 I details the radiation safety program for this license.

12.2 Reactor Safety Board (RSB) Review and Audit Activities A Reactor Safety Board (RSB) acts as a review panel for new reactor experiments, procedural changes and facility modifications The RSB thus provldes an Independent audit of the operatlons of the Nuclear Science Center. Issues concerning nuclear safety are immediately brought to the attention of the RSB. The University Radiological Safety Office provides Health Physics assistance for the Nuclear Sclence Center. This organizational arrangement thus provides another independent review of reactor operations (F~gure10-1).

12.2.1 RSB Composition and Qualifications The Reactor Safety Board (RSB) shall consist of at least three voting members knowledgeable In fields that relate to nuclear safety. The RSB shall review, evaluate and make recommendations on safety standards associated wth the operational use of the facility. Members of NSC operations and health physics shall be ex-officio members on the RSB. The review and advisory functions of the RSB shall include NSCR operations, radiation protection and the facility license. The Chairman of the Reactor Safety Board under the direction of the Deputy Director of TEES shall appoint the board members.

12.2.2 RSB Charter and Rules The operations of the RSB shall be in accordance with a wntten charter, including provisions for:

1) Meeting frequency: not less than once per calendar year and as frequent as circumstances warrant consistent wlth effective monitoring of facility activities.

2) Voting rules

3) Quorums

4) Use of subcommlttees

5 ) Revlew, approval and dlssemlnation of m~nutes 12.2.3 RSB Review Function The review responsiblllt~esof the Reactor Safety Committee shall Include, but are not 11mltedto the follow~ng.

Review and approval of new experiments utilizmg the reactor facilltles; Renew and approval of all proposed changes to the faclllty, procedures, license and techn~cal specificat~ons, Determination of whether a proposed change, test or experiment would constitute an umeviewed safety question or a change In Technical Specification; Review of abnormal performance of plant equipment and operating anomalies havlng safety significance; Revlew of unusual or reportable occurrences and incidents that are reportable under 10CFR20 and IOCFR50; Review of audit reports; and Review of vlolatlons of technical speclficatlons, license, or procedures and orders hav~ngsafety significance RSB Audit Function The RSB or a subcommittee thereof shall audlt reactor operations and radiation protection programs at least quarterly, but at intervals not to exceed four months. Audits shall include but are not l~mitedto the following.

1) Facillty operations, includmg radiation protection, for conformance to the techn~calspecifications, applicable license conditions, and standard operating procedures at least once per calendar year (Interval between audlts not to exceed 15 months),

2) The retraming and requalification program for the operating staff at least once per calendar year (interval between audlts not to exceed 15 months);

3) The facility security plan and records at least once per calendar year (interval between audits not to exceed 15 months);

4) The reactor facility emergency plan and implementing procedures at least once per calendar year (interval between audits not to exceed 15 months).

The licensee or his designated alternate (exclud~nganyone whose normal job function is within the NSCR) shall conduct an audit of the reactor facility ALARA program at least once per calendar year (~ntervalbetween audits not to exceed 15 months). The results of the audit shall be transmitted by the licensee to the RSB at the next scheduled meeting.

12.3 Procedures The philosophy of nuclear safety at the Nuclear Science Center assumes that all operations utihzing the reactor will be carned out in such a manner as to protect the health and safety of the publlc. This ph~losophy1s augmented in practice by deta~led,wvntten procedures. All personnel using the facil~tiesof the Nuclear Science Center follow the procedures. The loadlng or unload~ngof any core 1s performed according to detailed written procedures. Startup and operation of the reactor is also performed according to detailed written procedures.

Wrltten operating procedures shall be prepared, reviewed and approved before initratrng any of the activities l~sted in this section The procedures shall be rewewed and approved by the Drrector (NSC), or his deslgnated alternate.

the Reactor Safety Board, and shall be documented In a tlmely manner. Procedures shall be adequate to assure the safe operatlon of the reactor but shall not preclude the use of independent judgment and actlon should the situation require such Operating procedures shall be in effect for the follow~ngItems:

Startup, operatlon, and shutdown of the reactor, Fuel and experiment loading, unload~ng,and movement within the reactor, Control rod removal or replacement, Routine maintenance of the control rod, drives and reactor safety and interlock systems or other routine maintenance that could have an effect on reactor safety; Testing and calibration of reactor instrumentation and controls, control rod drives, area radlatlon monitors, and faclhty air monitors; Crvil disturbances on or near the facrlity site; Implementation of required plans such as emergency or security plans, and Actions to be taken to correct specific and foreseen potential malfunct~onsof systems, lncludlng responses to alarms and abnormal reactivity changes The Director (NSC) and the Reactor Safety Board shall make substantwe changes to the above procedures effective only after documented review and approval. The Director (NSC) or his designated alternate may make only minor modificat~onsor temporary changes to the original procedures that do not change their onginal intent. All such temporary changes shall be documented and subsequently revlewed by the Reactor Safety Board 12.4 Required Actions 12.4.1 Action to be Taken in the Event a Safety Limit is Exceedcd In the event a safety lim~t1s exceeded:

1) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2 ) An rmmedlate report of the occurrence shall be made to the Chairman, Reactor Safety Board, and reports shall be made to the NRC in accordance with Section 6.6.2 of these specifications, and

3) A report shall be prepared whlch shall include an analysis of the cause and extent of possible resultant damage, efficacy of corrective action, and recommendations for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Board for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

12.4.2 Action to be Taken in the Event of a Reportable Occurrence In the event of a reportable occurrence, the following action shall be taken:

1) NSC staff shall return the reactor to normal operating or shut down condrtions. If it is necessary to shut down the reactor to correct the occurrence, operations shall not be resumed unless authorized by the Director (NSC) or his deslgnated alternate.

2) The Drrector (NSC) or his designated alternate shall be not~fiedand corrective actlon taken with respect to the operations involved

3) The Director (NSC) or his desrgnated alternate shall notrfy the Chairman of the Reactor Safety Board

4) A report shall be made to the Reactor Safety Board whlch shall lnclude an analysis of the cause of the occurrence, efficacy of correct~veactlon, and recommendat~onsfor measures to prevent or reduce the probability of recurrence, and 5 ) A report shall be made to the NRC in accordance with Section 6 6 2 of these specificat~ons 6 ) Occurrence shall be revlewed by the RSB at their next scheduled meeting 12.5 Reports 12.5.1 Annual Report An annual report coverlng the operation of the reactor facility during the previous calendar year shall be submitted to the NRC before March 3 1 of each year providing the following information-A brief narrat~vesummary of (I) operating experience (including exper~mentsperformed), (2) changes in facil~tydesign, performance character~stics,and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and Inspections; Tabulation of the energy output (In megawatt days) of the reactor, hours reactor was critical, and the cumulatrve total energy output since initial cntlcality; The number of emergency shutdowns and inadvertent scrams, including reasons thereof, Discuss~onof the major maintenance operat~onsperformed during the period, lncludlng the effect, if any, on the safety of the operat~onof the reactor and the reasons for any corrective maintenance required; A brief description, includlng a summary of the safety evaluations of changes in the fac~lityor in procedures and of tests and experiments carried out pursuant to Section 50.59 of 1 0 CFR Part 50; A summary of the nature and amount of radioactive effluents released or d~schargedto the environs beyond the effective control of the licensee as measured at or before the point of such release or discharge. If the estimated average release after dilutlon or d~ffusionis less than 25% of the concentration allowed or recommended, a statement to thls effect is sufficient.

I) Liquid Waste (summarized on a monthly basis)

(1) Rad~oactivitydmharged during the reporting period.

(a) Total rad~oactivityreleased (In Curies)

(b) The Effluent Concentration used and the isotopic composition if greater than 1 x pCuries/cc for fission and activation products.

(c) Total radloact~vity(in curies), released by nucllde, during the report~ngperiod based on representative isotopic analysis.

(d) Average concentration at point of release (in pcurieslcc) during the reporting period.

(2) Total volume (in gallons) of effluent water (includlng dilutent) during periods of release.

11) Gaseous Waste (summarized on a monthly bass)

(1) Radloactivity discharged during the reporting period (in Curies) for:

(a) Argon41 (b) Part~culatesw t h half-hves greater than e~ghtdays iil) Sohd Waste (1) The total amount of solid waste transferred ( ~ cnu b ~ cfeet)

(2) The total actlvlty involved (in Cunes).

(3) The dates of shipment and disposition ( ~ shlpped f off We).

g) A summary of radiation exposures received by facility personnel and v~sitors,including dates and time where such exposures are greater than 25% of that allowed or recommended.

h) A description and summary of any environmental surveys performed outside the fac~lity.

12.5.2 Special Reports In addltion to the requirements of applicable regulations, reports shall be made to the NRC Document Control Desk and special telephone reports of events should be made to the Operations Center as follows:

1) There shall be a report not later than the followmg workmg day by telephone and confirmed in writing by telegraph or slmtlar conveyance to be followed by a wrltten report that descr~besthe circumstances of the event wlthin 14 days of any of the following:

a) Violation of safety l ~ m i t (See s Required Actions).

b) Any accidental release of radioactlvlty above permiss~blelimits In unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure; c) Any reportable occurrences as defined in the Specificat~ons. The written report (and, to the extent possible, the prellm~narytelephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outllne the corrective measures taken or planned to prevent reoccurrence of the event;

2) A written report with~n30 days of:

a) Personnel changes in the facility organization involv~ngLevel 1 and Level 2.

b) Significant changes In the transient or accident analysis as described in the Safety Analys~sReport.

12.6 Records A dally reactor operations log is maintained by the reactor operator, and contains such information as core loading.

experiments in the reactor, time of insertion and removal of expenments, power levels, tlme of startup and shutdown, core excess reactivity, fuel changes, and reactor instrumentation records.

Records are maintained which ind~catethe revlew, approval and conditions necessary for the production of rad~o~sotopes or performance of irradiation experiments.

Records of facihty operations in the form of logs, data sheets or other suitable forms are retained for the period indicated in the following sections-12.6.1 Records to be retained for a Period of at Least Five Years or for the Life of the Component Involved

1) Normal reactor faclllty operation

Prmcipal mamtenance operations Reportable occurrences Surveillance actlvltles requlred by the Techn~calSpec~ficatlons Reactor faclllty radlatlon and contammatlon surveys where required by applicable regulations Experiments performed with the reactor Fuel inventories, receipts, and shipments Approved changes in operatmg procedures Records of meet~ngand audlt reports of the RSB Records to be retained for at Least One Training Cycle Retraining and Requallficatlon of certified operations personnel Records of the most recent complete cycle shall be mamtained for ~nd~viduals employed Records to be retained for the Lifetime of the Reactor Facility Gaseous and liquld radloactwe effluents released to the environs.

Off-site environmental monrtormg surveys required by the Technical Specifications Radiation exposure for all personnel monltored Drawings of the reactor facility.

12.7 Emergency Planning The Emergency Plan for the Texas A&M Nuclear Science Center was prepared to meet the requirements of ANSVANS 15.16-1978 as amplified by Nureg-0849 The NSC submitted thls plan to the NRC for review in August of 1994. The plan has subsequent revisions in September 1995 and December 1999. Thls version 1s the current version in use at the faclllty.

The Emergency Plan applles to Texas A&M University System Texas Engineering Experiment Station Nuclear Science Center (NSC) facility.

Texas A&M University has a campus wide radiological emergency plan which is intended to integrate rad~ological emergency planning at all campus facilities using radioactive materials or radiation producing devices. The NSC Emergency Plan is an integral part of the Texas A&M University emergency plan and specifies the objectives and implementing procedures to be followed for emergencies occurring at the NSCR.

The Emergency Plan indicates response capabilities for emergency conditions arising in connection with operation of the reactor. It includes Identification of various precursor conditions (loss of electrical power, fires, reactor pool leaks, etc.) and the consequences for various independent or simultaneous precursors. The plan includes the event classification system. Deta~ledemergency implementing procedures have been developed and are referenced in the plan The NSC Director has primary responsibihty for emergency planrung and response. The plan specifies delegation of responslblllty and authority in the absence of the Director. The Emergency Plan and implementing procedures are revlews annually to assure that any required changes are incorporated into the plan

12.8 Security Planning The Nuclear Sc~enceCenter Securlty Plan ind~catesthe measures provided to protect specla1 nuclear material,

~ncludmgdetails of the protectlte equlpment and police agencles As result, ~t is not for publ~caccess. The NSC D~rector1s responsible for admmstering the securlty program and assurrng that ~t is current The Physlcal Securlty Plan prov~desthe NSC with specific criterla and d~rectionto protect the NSC from acts of sabotage and theft, which m ~ g h endanger t the health and safety of the public or the integrity of the facllity The plan fulfills the appl~cablesecurity plannrng requlrements of 1OCT;RSO and 10CFR70 Spec~fically,the NSC maintains a non-power reactor license and implements physrcal protectlon based on 10CFR73.60, "Additional requirements for the phys~calprotectlon of specla1 nuclear mater~alat non-power reactors "

12.9 Quality Assurance Smce the NSC is not seekmg a construction permit, this SAR does not lnclude a description of a quallty assurance program for the design and construction of the structures, systems and components of the facility This sectlon describes the Qual~tyAssurance program that 1s in place to govern safe operation and modlficatlon of the facility.

This program meets the appl~cablerequirements of Regulatory Guide 2.5 and ANSVANS-15 8-1995.

The NSC Dlrector has responslbil~tyfor the quahty assurance actlvit~es,and thus has the authority to rdentify problems, to initlate corrective actlons and to insure that correctwe actions are complete. He exercises a Q A oversight by assuring that operatmg and mamtenance procedures ~ncludespecific requirements to assure that mod~ficat~on, maintenance and calibrat~onof safety-related systems malntaln the qual~tyand r e l ~ a b ~ lof ~ty equtpmcnt. Further, exper~mentreviews use written requlrements to assure that installat~onand operation of the experiment does not degrade the performance of safety equlpment. Planning and reviewing modification of safety-related equipment usmg formal written checklist-type procedures assure equipment continues to meet NSC specificat~ons.Most of the reactor equipment In use in the facll~tydoes not have a formal QA documentation. The provisions of section 4 of ANSVANS-15.9 cover this equipment. After-maintenance checks, alignment and cal~bratlonof the replacement equipment assure that equipment meets the or~ginalequipment specifications.

Procedures include schedules of equipment maintenance and callbration, and provide records that such functions are completed. Calibration procedures ~ncluderequlrements that critlcal equipment and instruments used in the callbrat~onsare themselves currently calibrated 12.10 Operator Training and Requalification The Standard Operating Procedures covers the detailed requirements for Requallfication of licensed operators. The NSC is comm~ttedto maintain~ngthe hlghest level of operator qualification The program consists of lectures followed by detaded and in-depth exams to verify level of knowledge, a number or reactor manipulations to ensure proficiency and operator evaluations to ensure an effective skdl level 12.11 Startup Plan This Safety Analys~sReport does not include a startup plan because the facillty has been in routme operation for many years.

12.12 Environmental Reports The Atomic Energy Commission concluded "that there will be no significant environmental impact associated with the l~censmgof research reactors or critical fac~l~ties designed to operate at power levels of 2 Mwt or lower and that no environmental impact statements are required to be written for the issuance of construction permits or operatmg licenses for such facilities." This is from a letter date January 23, 1974 from D. R. Miller.

The NSC expects no change in land and water use because of extending the NSC llcense for an additional 2 0 years.

Emissions of radioactive materials or other effluents will not change because of extending the llcense term.

13 ACCIDENT ANALYSIS 13.1 Accident-Initiating Events and Scenarios Thls sectlon 1s ~ncludedIn the speclfic subsections in 13 2 13.2 Accident Analysis and Determination of Consequences 13.2.1 Maximum Hypothetical Accident The Des~gnBasis Accident 1s the loss of fuel claddlng integr~tyfor one fuel element and the simultaneous pool-water loss resulting In fission product release.

In an effort to estimate the total release of fisslon product In the case of fuel-claddrng failure, Texas A&M conducted a series of experiments to predict the maximum fuel temperature The experimenters measured temperatures In a few core locatlons Experiments used thls data to estlmate the actual peak temperature Inn those locations To find the actual power denslty (PD) in these locatlons, the expenmenters used the program Exterminator-2. With this information, along w t h the peak temperatures, they were able to graph the relationship between PD and peak fuel temperature (Figure 13-1). The results show hlgher peak fuel temperature than predicted for the Puerto Rico Nuclear Center The Puerto R ~ c Nuclear o Center also obtained one measurement for a PD of 23 kW per fuel element The Texas A&M relatlonshlp shows a higher temperature for that PD Figure 13-1: Steady State Fuel Temperature as a Function of Power Generation 13-1

Finally, t h ~ analysis s assumes a maximum temperature of 535°C for a PD of 28 kW, the maximum allowed for the NSCR With this max~mumtemperature and PD, the surface or "m~nimum"temperature would be 150°C.

To estlmate the fission product release, this analysis uses the fission-product release fraction of 2 6x10.~.General Atomics experimentally determined this fract~onaveraged over the fuel volume Usmg the above assumptions, the saturated activities of the significant fission products at 1 Mw in a s~nglefuel element are Total lodme fission products - 6,432 Ci Total halogen fission products - 7,611 CI Total gaseous fission product - 10,760 Ci Applying the release fraction of 2.6 x 10.' to the total inventory In a single element operatmg at 1 Mw ylelds the followmg activities that would be released In a cladding falure.

Total gaseous activity - 280 mC1 Total todine activlty - 167 mCi Total halogen activity - 198 mCi If the release accident occurred with water in the pool, the halogens will remaln in the water. The resulting concentration would be 3 65x10-~ pc/cm3. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, t h ~ value s would decay to 8.34 x lo-' pclcm2. The Deminerahzer system would be removed these soluble fiss~onproducts and they would eventually go Into the liquid waste system Texas ABM calculated the results of the release of fission products from a single fuel element wlth and without water in the reactor pool, and with and without the ventilation system in operation. Table 13-1 shows the calculated exposure to population outside the building and exposure to operatmg personnel inside the facihty. The only case where significant exposure occurs requires the simultaneous failure of the fuel element clad, catastrophic failure of the pool and liner, and a failure of the ventllatlon system with personnel remaining within the reactor facility for a period of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after release. The maximum exposure is 49 R to the thyroid. Thus, no realistic hazard of consequence will result from the Design Basis Accident.

Table 13-1: Sumnlnry of Radiation Exposures Following Cladding Failure of tlle highest Power Density FLIP Fuel element A. Building Ventilation Operating:

1. Maximum Exposure to Population Outside Building WBGD* Thyroid Dose Pool Water Remaming 3.5~10" --

Pool Water Drained 1.4x10-' 3.7 mR

2. Exposure to Operatmg Personnel in One Hour After Release Pool Water Remalning 0.84 mR --

Pool Water Drained 1.75 mR 10.5 mR B. Building Ventilation Shut Down

1. Maximum Exposure to Populat~onOutside Budding (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) WBGD* Thvroid Dose Pool Water Remaining 3.6~10'~ --

Pool Water Dralned 2.1x10-~ 18 mR

2. Exposure to Operating Personnel in One Hour After Release Pool Water Remaining 1.75 mR --

Pool Water Drained 4.2 mR 49 R

• WBGD = Whole Body Gamma Dose

13.2.2 Insertion of Excess Reactivity 13.2.2.1 Accidental Pulsing F r o m Full P o u e r T h ~ analys~s s 1s for the accidental prompt reactlwty add~tlonfor the NSCR operating at power. Normally these pulses are from belo\+,the point of adding heat It is necessary to examme t h ~ ss~tuationin splte of the interlocks that will prevent this from happening General Atom~cperformed the calculations uslng the BLOOST 2 code assumlng ad~abaticprocesses. The details of the calculations are below Texas A&M supphed the exper~mentalparameters and core power d~stributions The study was to find the react~v~ty insertion from power that would produce a peak core temperature of 950" C. Figure 13-2 shows the results from both 1 MW and 300 watts as a function of the number of FLIP elements In the mixed core at the beginning-of-llfe. F ~ g u r e13-2 also shows the calculated end-of-hfe case for a full FLIP core. As Figure 13-2 shows, ~fthe reactor is pulsed from 1 MW, considerably more prompt reactivity is requlred to obtain 9.50" C Therefore, it is pulsing from low power that l ~ m ~the t s amount of insertion and not the "acc~dent" sltuatlon o Pulse From 1 MW Pulse From 300 w EOL (8.2 MW-Years) 0 EOL (8.2 MW-Years) 0 30 40 50 60 70 80 90 100 Number of FLIP Elements in 98 Clemenl Core Figure 13-2: Pulse to Produce 950°C Peak Temperature 13.2.2.2 The Pulsing Accident in Mixed a n d Flip Cores The scope of this study performed by General Atomic was:

1) Determine the size of the pulse producing 950°C peak temperature as a functlon of the number of FLIP elements in the core and the burnup by assuming ad~abaticprocesses, and

2) If the results of 1) do not allow pulses of an acceptable magn~tude,to refine the calculations by lncludmg the effects of heat transfer wh~chwould give a more real~sticassessment of the effects of the reactivity add~t~on The results of 1) indicated that 2) was not requ~red Figure 13-2 shows the slze of the pulse that would produce maximum temperatures of 950°C as a function of the number of FLIP elements In the core. The results are for pulses from 1 Mw and from 3 0 0 steady~ state power. The curves represent beglnnlng-of-hfe (BOL) conditions for the prompt neutron lifetime and the temperature coefficient General Atom~csused the end-of-hfe (EOL) values for these parameters to calculate the two polnts for the full FLIP core Following is a summary of the process lnvolved in acquiring this ~nformat~on:

The calculat~onswere made using BLOOST 2 The input parameters that were common to all problems were.

No of elements = 98 Delayed neutron fract~on= 0 007 Fuel specific heat = 720 0 + 1.48 T (w-secI0C-element)

Water speclfic heat = 860 (W-secI0C-element)

Fuel thermal resistance = 10000 ("CIMWlelement)

Coolant thermal resistance = 1175 ("CIMWIelement)

Initial average fuel temperature at 1 MW = 238" C Initial average coolant temperature at 1 MW = 45" C

10) Pulse insert~ontlme = l00rnsec

11) Scram delay t m e = 15 msec

12) Rod drop tlme = 0.985 sec For the pulses from 300 W, General Atomics assumed that the system had an initial temperature of 25°C and that there was no scram F~gure13-3 shows the measured transient rod integral worth The "ramp" table assumes this rod uniformly accelerates from 11s in~tlalposit~onto 605 units (I e., $3.25) The ln~tlalposition provides the deslred worth of the pulse. The $3 25 posltlon for the upper end of the Insertion sharpens the pulse as the integral worth curve flattens out drastically above that point Using the full out posit~onas the final point would tend to clip the pulse. Flgure 13-4 shows the ramp as a functlon of tlme for $2 00, $2.25 and $2.75 pulse. These curves are plots of the ramp table.

Figure 13-3: Transient Rod Integral Rod Worth

/"

-52 15Pulse

/./ ---- $2 25 Pulse

.? --- $ 2 00 Pulse 0 02 W .06 08 I Time. Seconds Figure 13-4: Time Dependent Pulse Reactivity Insertion Used to Obtain Ramp Table Construction of the scram table assumed the total excess available was $7.00, that $3.50 was held down by the 1 M w temperature, total worth of the rods available for the scram was $12.00, the shape of the rod worth is represented by the General Atomlc "standard" shape and the rods fall with uniform acceleration Figure 13-5 is a plot of the scram table.

0 .2 .4 .6 .8 1.0 Time, S e c o n k Figure 13-5: Time Dependent Reactivity Insertion Used to Generate Scram Table General Atomics included three different core configurations to determine the effect of adding FLIP fuel. All assumed the FLIP fuel was In the central region surrounded by standard fuel (if any). All the cores consisted of 98 fuel elements The values of the several parameters that are dependent on configurations were estimated by interpolation between data points already acquired.

Peaking factors for the power distribution were 35 FLIP elements: axial = 1.36; radial= 2.00; cell = 1.95

59 FLIP elements: axial= 1 36, radlal = I .91, cell = 1 95 98 FLlP elements: axlal = 1 36; radlal = 1 56, cell = 1 95 The following calculated data provlded the input to estimate the prompt neutron hfetimes-BOL - 18 FLIP153 standard I= 28 0 psec BOL-AlIFLIP 1= 17.5 psec EOL- All FLIP 1= 2 1.0 psec Lmear interpolation, based on the fract~onof FLIP elements In the core, Elves the followmg 35 FLIP (35 7% FLIP) BOL 1 = 26.5 psec 59 FLIP (60.2% FLIP) BOL 1 = 23.1 psec 98FLIP(100%FLIP) BOL1=17.5psec 98 FLlP (100% FLIP) EOL 1 = 21 0 psec Temperature, OC Figure 13-6: Temperature Coeff~cientsof TRIGA Fuels

Figure 13-6 shows the calculated prompt negative temperature coefficients for the 18-rod FLIP at BOL and the full FLIP at BOL and EOL The BOL coeffic~entfor the mixed cores between 18 rod (which corresponds to 25% FLIP) and full FLIP was estimated by mdang a h e a r ~nterpolat~on between the two curves with the ratio of FLIP to total as the proportionahty constant Flgure 13-7 provldes plots of the Integral temperature coefficlents. used In BLOOST 2, for the four core configurat~ons

-BOL, 35 n w (35 7%)

--- BOL, 59 FLIP (60 2%)

Figure 13-7: Integral Temperature Coefficient The integral water temperature coefficient 1s the standard FLIP coefficient in the inset to F~gure13-7.

Since the BLOOST 2 does not calculate the peak temperature properly when the pulse is from power, hand calculations provlded these values. The following steps describe the calculation procedure.

1) Determine the average power density in the "hottest" element at 1 M V .

2) From this power density, determine the steady state temperature at the periphery of the fuel at the axial centerhe.

3) Calculate the energy content at the point at 1 MW

4) Calculate the energy added at that point in the pulse Determine the adiabatic temperature that would result from that energy content. The peak edge temperature In the fuel before the pulse IS:

35 FLIP Max SS edge temp = 165°C Energy content = 0.139 MWIelement 59 FLIP Max SS edge temp = 170°C Energy content = 0.144 MU'Ielement 98 FLIP Max SS edge temp = 180°C Energy content = 0.154 MWIelement The following tables give the principle results of the BLOOST 2 calculations

T o find the values of the reactlvlty lnsertlon that would result in peak fuel temperatures of 950" C, an Interpolated value was found from the data in the tables above. These values are In the next table It was felt that these results were sufficient to allow effective operation so ~twas not necessary to do the second part of the program that would consider heat flow from the fuel after the pulse.

Table 13-4: Pulse Values t o Exceed Temperature REACTIVITY TO G I V E 950" C PEAK TEMPERATURE No. F L I P M w Burnup Pulse from 300w Pulse from 1Mw 35 BOL $2.45 $3.32 59 BOL 2.35 3.29 98 BOL 2.36 3.65 98 BOL 2.08 2.42 13.2.3 Loss of Coolant The strength of the fuel element clad is a functlon of its temperature. The stress imposed on the clad 1s a function of the fuel temperature as well as the hydrogen-to-zirconium ratio, the fuel burnup, and the free gas volume within the element. In the analysis of the stress imposed on the clad and strength of the clad uses the following assumptions:

1) The fuel and clad are at the same temperature.

2) The hydrogen-to-zircon~umratio is 1.7 for standard fuel and 1.6 for FLIP fuel

3) A space one-eighth Inch hlgh within the clad represents the free volume within the element

4) The reactor contains fuel that has experienced burnup equivalent to 7700 MW-days.

5) Maximum operating temperature of the fuel is 600" C.

The fuel element internal pressure P is given by:

where.

P, 1s the hydrogen pressure, P/, is the pressure exerted by volat~lefiss~onproducts, and Pa,, is the pressure exerted by trapped alr For hydrogen-to-zirconium ratios greater than about 1.58, the equil~briumhydrogen pressure can be approximated by:

where x is the ratlo of hydrogen atoms to zlrconlum atoms, and Tkis the fuel temperature (K)

The pressure exerted by the fiss~onproduct gases is given by.

n RT, P,, = f--E E V where.

f is the fission product release fraction; It

- is the number of moles of gas evolved per unit of energy produced (mol/MW-day),

E R is the gas constant (8.206 x 1u2L-atm/mol-K);

V is the free volume occupied by the gasses (L), and E is the total energy produced in the element (MW-day)

The fission product release fraction is given by:

-1 . 3 4 ~ 1 0 ~

f = l S x l ~ - +3.6x103

' exp where:

Tois the maximum fuel temperature in the element during normal operation (K).

n The fission product gas production rate, -, varies slightly with the power density. The value 1.19 x 10'~mollMW-E day is accurate to within a few percent over the range from a few kilowatts per element to well over 40 kW per element. The free volume occup~edby the gases is assumed to be a space one-eighth ~ n c h(0.3175 cm) high at the top of the fuel so that V = 0.3 l75n - r,2 where.

r, is the inside rad~usof the clad (1.745 cm).

For standard TRIGA fuel, the maximum burnup IS about 4 5 MW-days per element, but the TRIGA-FLIP fuel 1s capable of burnup to about 77 MW-days per element As the fission product gas pressure is proport~onalto the energy released, assume that the FLIP fuel in the reactor has experienced max~mumburnup Finally, the air trapped w t h m the fuel element clad \+A1 exert a pressure where it is assumed that the initial specific volume of the air is 22.4 Umol. Actually, the air forms oxides and nitrides with the zirconium, so that after relatively short operation the air is no longer present in the free volume Inside the fuel element clad. The results of the stress Imposed on the clad for standard and FLIP fuels are in Figure 13-8.

400 500 GOO 700 BCO 900 1000 1100 TEMPERATURE. OC Figure 13-8: Strength and Applied Stress as a Function of Temperature for 1.7 and 1.6 H-Zr TRIGA Fuel General Atomic developed A two dimensional transient-heat transport computer code for calculating the system temperatures after the loss of pool water. They were derived the heat removal parameters for the calculations. The assumption was that the reactor was shutdown fifteen minutes before the core was uncovered (the tlme between the actuation of the pool level alarm and the uncovering of the fuel for a catastroph~cfailure of a ten-inch stainless steel line underneath the pool).

If the reactor operates for seventy MW-hours or less per week, power generatlon per element values approximately 20% higher are sufficient Thus, 25 kW/element for standard and 28 kWIelement for FLIP fuel are adequate power denslt~es A comparison of decay heat generatlon versus tlme following loss of coolant for infinite reactor operntlons and 70 hlW-hours per week cycle operatlon are in hgure 13-7.

I\\ ,- Reactor Operation Infinile Reactor O p c r a t ~ o n70 MW-HI p e r w e e k (3300 MW-Days)

Figure 13-9: Decay Heat Power Generation Following Loss of Coolant for Infinite Reactor Operation and Periodic Reactor Operation 13.2.4 Loss of Coolant Flow The NSC Reactor uses natural convection flow; therefore, this casualty is not applicable.

13.2.5 Mishandling or Malfunction of Fuel This acc~dentis included in the Maximum Hypothetical Accident. The worst case would be mishandling or malfunction leading to cladding falure. The Maximum Hypothetical Accident addresses this with a catastrophic failure of the pool and a failure of the ventilation system 13.2.6 Experiment Malfunction The Reactor Safety Board must approve any new experiment involving the reactor. This committee reviews all experiments for safety and for compliance with the operating license and NRC regulations. The Senior Reactor Operator controls loading, unloadmg or moving Experiments affecting the reactivity of the core.

The reactivity effects of experimental facilities used with the present core present no significant problems The values reported for sirmlar experimental facilities at other TRJGA installations appear to be comparable and therefore no hazard exists. The reactivity worth of any non-secured experiment shall have reactivity wroth less than

$1.00. This specification provides assurance that the worth of a single unfastened experiment will be lim~tedto a value such that suddenly inserting the positive worth of the experiment will not exceeded the safety limit. Removal of experiments of $0.30 worth or more from the reactor at full power often requires a power decrease by the operator to prevent high power levels 13.2.7 Loss of Normal Electrical Power The NSCR will shutdown upon loss of electrical power, as electricity is required to maintain magnet current in the control rod drives to keep the control rods rased out of the core. Once electrical power is lost, magnet current is lost

and control rods drop ~ n t othe core. The reactor will have sufficient shutdown margin even w ~ t hthe most reactwe control rod stuck in 11sfully withdrawn posltlon No emergency power supply IS available.

13.2.8 External Events Floods. humcanes, earthquakes and tornados are the three areas of concern for naturally occurring external events In the case of floods and humcanes, a complete flood~ngof the Lower Research Level will have no effect on reactor safety or the safety of the publ~c.Nor 1s ~t likely that hurricane winds wdl affect the Integrity of the bulld~ng(a fallout shelter) or the pool As chapter 2 of thls report states, earthquakes are extremely unl~kelyIn t h ~ area.

s Finally, tornados occur near the NSC. However, damage to the reactor pool, w h ~ c hprotects the reactor, IS not credlble The NSC reaches the same conclusion for the NSC Reactor.

13.2.9 Mishandling or M a l f u n c t i o n of Equipment Potentla1 Hazards Considered

1. Fuel Bundle Rotation T o achieve symmetry with the central FLP region it is often necessary to load both FLIP and standard elements in a single bundle. Smce the fuel bundles can be physically rotated 180" and still fit the grid plate, the inadvertent rotatlon of a four-element fuel bundle containing one FLIP and three standard elements was considered In the event of such a rotatlon, standard fuel would surround the FLIP element by. The analysis showed no appreciable increase in average power generated either in the FLIP element, or in the standard elements. The maximum fuel temperature that would be obtained during a 30 Mw-sec pulse with the rotated bundle is 1450°F (788°C) which is well below the safety l~mit.Table 13-5 shows the results of the calculations of peak average power per cell during 1 M w steady-state operation and the maximum fuel temperature due to a 30 Mw-sec pulse with and without rotatlon of one bundle. There is no safety problem due to inadvertent rotation of a fuel bundle.

Table 13-5: Fuel Bundle Rotation Study for Maximum Power a n d Maximum Temperature I Limiting Criterion 0 -

I Core with unrotated 1 O n e bundle bundle rotation M a x i m u m power p e r e l e m e n t 18.18 kW 17.94 kW Maximum fuel temperature d u e to a 30MW-sec 1400°F 1450°F pulse (76 1O C ) (788OC) 2 Control Rod Run-Out

The magnrtude of the result of the withdrawal of all control rods at maxlmum speed was considered for the PRNC reactor The magnltude of the effect of thls accident is dependent primarrly on the speed of rod withdrawal and the value of the temperature coeffictent Srnce proposals for the Texas A&M reactor included core loadlngs that vaned from all, standard to all FLIP TRIGA fuel the temperature coefficient had to be exammed for these variatrons.

General Atomlc performed calculations for a varrety of cores wlth the results shown In Flgure 13-6 As can be seen, the standard fuel with no poison had a temperature coeffic~entthat was relatively constant wlth temperature The add~tlonof FLIP fuel resulted In temperature coefficient that Increased linearly with temperature Most of the core loadrngs antlcrpated for the NSCR wdl Ire between 18-element FLIP and all FLIP curves It is doubtful that an entlre core would ever achleve the 3000 MWIday burnup indmted for the lowest curve due to the planned loadlng sequence. Even for this llmltlng case, however, the magnltude of the temperature coefficient was large enough to allow safe operation of the reactor.

For the calculation of the PRNC control rod run-out accident, the withdrawal time used was 16.2 seconds The withdrawal tlme of the shim-safetles of the NSCR reactor is 347 mmutes. The NSCR will be set to scram at 1.25 Mw (or less) as opposed to 2.2 Mw for the PRNC reactor. Smce the temperature coefficient will be the same or larger and the control rod removal rate IS so much slower, the reactor power level will follow the rod insertion so that the excess reactivity wlll be near zero. Thus, when the trlp occurs the core temperature will nearly correspond to the case of a reactor operating at steady state. The maxlmum power generated in any cell will be approximately 25 kW, which 1s well below the maximum permitted 13.3 Summary and Conclusions None of the accidents here will result in consequences to the publlc health and safety.

14 TECHNICAL SPECIFICATIONS FOR FACILITY LICENSE NO. R-83 Included in this sect~onare the Technical Specificat~onsand the "Bases" for the Technical Specifications. These bases, which provide the techn~calsupport for the ~ndividualtechnical spec~ficat~ons, are included for mformational purposes only. They are not part of Technical Speclficat~onsand they do not constitute lim~tat~ons or requirements to uvh~chthe l~censeemust adhere.

14.1 Definitions 14.1.1 Abnormal Occurrence An abnormal occurrence is an unscheduled incldent or event that the Nuclear Regulatory Commission determines is significant from the standpoint of pubhc health or safety 14.1.2 ALARA The ALARA program (As Low As Reasonably Achievable) IS a program for ma~ntainingoccupational exposures to r a d ~ a t ~ oand n release of rad~oactiveeffluents to the environs as low as reasonably achievable.

14.1.3 Channel A channel is the combmation of sensors, llnes, amplifiers and output devices, which are for measuring the value of a parameter.

14.1.3.1 Channel Test A channel test is the introduction of a signal mto the channel for verification that it is operable 14.1.3.2 Channel Calibration A channel callbration is an adjustment of the channel such that its output corresponds, wlth acceptable accuracy, to known values of the parameter that the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and constitutes a channel test 14.1.3.3 Channel Check A channel check is a quahtative verification of acceptable performance by observation of channel behavior. T h ~ s verification, where poss~ble,shall include comparison of the channel with other ~ndependentchannels or systems measuring the same variable.

14.1.4 Confinement Confinement means a closure of the overall facil~tythat results in the control of the movement of air into it and out of the facillty through a defined path.

14.1.5 Core Lattice Position The core lattice position is that region in the core (approximately 3" x 3") over a grid-plug hole. A fuel bundle, an experiment or a reflector element, may occupy the position.

14.1.6 Experiment An operation, hardware, or target (excluding devices such as detectors, foils etc.) which is designed to investigate non-routine reactor characteristics, or whlch is intended for madlation within the pool, on or in a beam port or irrad~ationfacility, and which is not rig~dlysecured to a core or shield structure so as to be a part of their design.

14.1.7 Experimental Facilities Experimental facil~tiesshall mean beam ports, ~ncludingextension tubes wlth shields, thermal columns wlth shields, vertical tubes, through tubes, in-core lrradlatlon baskets, irradiation cell, pneumatic transfer systems and in-pool irrad~ationfacilities.

14.1.S Experiment Safety Systems Experiment safety systems are those systems, lncludlnp them associated Input circuits, which are deslgned to Initlate a scram for the prlmary purpose of protecting an experiment or to provlde informat~onthat requires operator Intervention 14.1.9 FLIP Core A FLIP core is an arrangement of TRIGA-FLIP fuel In a reactor grid plate.

14.1.10 Fuel Bundle A fuel bundle is a cluster of two, three or four elements and/or non-fueled elements secured in a square array by a top handle and a bottom grid plate adapter. Non-fueled elements shall be fabricated from stainless steel, aluminum or graphite materials.

14.1.11 Fuel Element A fuel element is a smgle TRIGA fuel rod of either standard, FLIP, or LEU type.

14.1.12 Instrumented Element An instrumented element 1s a special fuel element in which a sheathed chromal-alumel or equivalent thermocouple is embedded In the fuel near the horizontal center plane of the fuel element at a point near the center of the fuel body.

14.1.13 LEU Core A LEU core is an arrangement of TRIGA-LEU fuel in a reactor grid plate.

14.1.14 Limiting Safety System Setting The llmiting safety system settlng is the setting for automatic protective devices related to those variables having significant safety functions 14.1.15 Measuring Channel A measuring channel is the combination of sensor, lnterconnecting cables or lines, amplifiers, and output device that are connected for the purpose of measuring the value of a vanable.

14.1.16 Measured Value A measured value is the value of a parameter as it appears on the output of a channel.

14.1.17 Mixed Core A mixed core is an arrangement of standard TRIGA fuel elements with at least 35 TRIGA-FLIP andlor LEU fuel elements located in a central contiguous region of the core.

14.1.18 Movable Experiment A movable experiment is one for wh~chit is intended that the entire experiment may be moved in or near the core or into and out of the reactor while the reactor is operating 14.1.19 Operable Operable means a component or system is capable of performing its required function.

14.1.20 Operating Operating means a component or system is performing its required function

14.1.21 Steady State Operational Core A steady state operational core shall be a standard core, FLIP core, LEU core, or mlxed core for which the core parameters of shutdown margin, fuel temperature and power calibration ha\e been determmed 14.1.22 Pulse Operational Core A pulse operat~onalcore is a steady state operational core for which the maximum allowable pulse reactlvity insert~onhas been determmed 14.1.23 Pulse Mode Pulse mode operation shall mean any operation of the reactor wlth the mode selector swltch in the pulse posltion 14.1.24 Reactivity Worth of an Experiment The reactivity worth of an experiment 1s the maximum absolute value of the reactlvity change that would occur as a result of intended or antlclpated changes or credible malfunctions that alter the experiment position or configuration 14.1.25 Reactor Console Secured The reactor console is secured whenever all scrammable rods have been verified to be fully inserted and the console key has been removed from the console.

14.1.26 Reactor Operating The reactor 1s operating whenever it is not secured.

14.1.27 Reactor Safety Systems Reactor safety systems are those systems, including their associated Input channels, which are designed to initiate automatic reactor protection or to provide information for mtiation of manual protective actlon Manual protective actlon is considered part of the reactor safety system 14.1.28 Reactor Secured A reactor 1s secured when:

1) It contains insufficient fisslle material or moderator present in the reactor and adjacent experiments to attain criticality under optimum available conditions of moderation and reflection, or 2 ) The reactor console is secured no work is in progress involving core fuel, core reflector material, installed control rods, or control rod drives unless they are physically decoupled from the control rods.

14.1.29 Reactor Shutdown The reactor is shut down if it is subcritical by at least one dollar with the reactor at ambient temperature, xenon-free and wlth the reactivity worth of all experiments included.

14.1.30 Reportable Occurrence A reportable occurrence is any of the following that occurs durlng reactor operation:

1) Operat~onwlth actual safety system settlngs for required systems less conservative than the limiting safety-system settings specified In the Technical Specificatlons 2.2.

2) Operation In violation of llmiting conditions for operation established in the technical specificatlons.

3) A reactor safety system component malfunction that renders or could render the reactor safety system incapable of performing its Intended safety function unless the malfunction or condition IS discovered during maintenance tests or periods of reactor shutdowns. (Note: Where components or systems are

provided in addltion to those requlred by the technical speclficatlons, the fallure of the extra components or systems 1s not considered reportable provlded that the mintmum number of components or systems speclfied or requlred perform their Intended reactor safety function.)

An unanticipated or uncontrolled change In reactlvlty greater than one dollar Abnormal and significant degradation in reactor fuel or claddmg, or both, coolant boundary, or containment boundary (excludmg minor leaks) where applicable which could result in exceedmg prescribed radlatlon exposure limits of personnel or environment, or both An observed inadequacy In the implementation of admlnlstratlve or procedural controls such that the inadequacy causes or could have caused the existence or development of an unsafe condltlon wrth regard to reactor operations 14.1.31 Rod-Control A control rod 1s a device fabricated from neutron absorbmg material and/or fuel that is moved up or down to control the rate of a nuclear reaction It may be coupled to ~ t drlve s unlt allow~ng~tto perform a safety functlon (scram) when the coupl~ngIS disengaged.

14.1.32 Rod-Regulating The regulating rod is a low worth control rod used primarily to maintain an Intended power level that need not have scram capablhty and may have a fueled follower Its percent withdrawal may be varied manually or by the servo-controller.

14.1.33 Rod-Shim Safety A shim-safety rod is a control rod having an electric motor drive and scram capabillties. It may have a fueled follower section 14.1.34 Rod-Transient The translent rod 1s a control rod with scram capabillties that is capable of providrng rapid reactivity insertion to produce a pulse.

14.1.35 Safety Channel A safety channel is a measuring channel in the reactor safety system 14.1.36 Safety Limit Safety h i t s are llmits on Important process variables that are necessary to reasonably protect the integrity of those physical barriers that guard against the uncontrolled release of radioactivity. For the Texas A&M NSC TRIGA reactor the safety llmit is the maximum fuel element temperature that can be permitted with confidence that no damage to any fuel element cladding will result.

14.1.37 Scram Time Scram tlme is the time measured from the instant a simulated signal reaches the value of the LSSS to the Instant that the slowest scrammable control rod reaches its fully inserted position.

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

14.1.39 Shall, Should and May The word "shall" is used to denote a requirement, the word "should" to denote a recommendatlon, and the word "may" to denote permlsslon, neither a requirement nor a recommendatlon In order to conform to t h ~ standard, s the user shall conform to ~ t requirements s but not necessarily to its recommendatlons.

14.1.40 Shutdown Margin Shutdown margin 1s the mlnimum shutdown reactivity necessary to provide confidence that the reactor can be made subcr~tlcalby means of the control and safety systems, starting from any permissible operating condlt~on.It assumes that the most reactlve scrammable rod and any non-scrammable rods are fully withdrawn, and that the reactor will remaln subcrltical without any further operator action 14.1.41 Standard Core A standard core IS an arrangement of standard TRIGA fuel In the reactor grid plate.

14.1.42 Steady State Mode Steady state mode of operation shall mean operation of the reactor with the mode selector swltch in the steady state position.

14.1.43 True Value The true value IS the actual value of a parameter 14.1.44 Unscheduled Shutdown An unscheduled shutdown 1s any unplanned shutdown of the reactor caused by actuation of the reactor safety system, operator error, equipment malfunction, or a manual shutdown in response to conditions that could adversely affect safe operation. It does not include shutdowns that occur durlng testing or check out operations 14.2 Safety Limit and Limiting Safety System Setting 142.1 Safety Limit Fuel Element Temperature Appllcab~lity This speclficatlon applles to the temperature of the reactor fuel.

Objective The objective is to define the maximum fuel element temperature that can be permitted with confidence that no damage to the fuel element cladding will result.

a) The temperature in a stainless steel-clad TRIGA-FLIP or TRIGA-LEU fuel element shall not exceed 2100°F (1 150°C) under any condltions of operation.

b) The temperature in a stainless steel-clad Standard TRIGA fuel element shall not exceed 1830°F (1000°C) under any condltions of operation.

The important parameter for a TRIGA reactor is fuel element temperature. This parameter is well suited as a single specification because ~tcan be measured directly w ~ t ha thermocouple or inferred indirectly through reactor power.

A loss in the integrity of the fuel element cladding could arise from a buildup of excessive pressure within the fuel element if the fuel element temperature exceeds the temperature safety hmit The fuel element temperature and the ratlo of hydrogen to zirconium in the fuel-moderator material determine the magnitude of the pressure buildup. The

mechanism for the pressure bu~ldupis the dissociation of hydrogen from the zircon~umhydrlde moderator that has been blended w~thuranium to form the fuel mlxture encased w t h m the fuel element cladding The temperature safety llmit for the TRIGA-FLIP or LEU fuel element is based on data whlch lndlcates that the internal stresses wlthln the fuel element due to hydrogen pressure from the d~ssoclat~on of the zircon~umhydrlde

~ 1 1 not 1 result In compromlse of the stamless steel claddlng lf the fuel temperature is not allowed to exceed 2100°F (1 150°C) and the fuel element cladding 1s water cooled The temperature safety hmlt for the Standard TRIGA fuel element 1s based on data, including the large mass of experimental evidence obtained durlng high performance reactor tests on thls fuel, whlch indlcate that the stress In the claddlng due to hydrogen pressure from the drssoc~ationof zirconium hydride wdl not result in compromlse of the stainless steel claddlng if the fuel temperature is not allowed to exceed 1830°F (1000°C) and the fuel element claddlng is water cooled.

14.2.2 Limiting Safety System Setting Appl~cabilitv Thls specification applies to the scram settlngs that prevent the fuel element temperature from reachmg the safety hmit Obiective The objective is to prevent the fuel element temperature safety hmits from bemg reached a) For steady state operation

1) The hm~tingsafety system setting shall be 975°F (525OC) as measured in an instrumented fuel element. The instrumented element shall be located adjacent to the central bundle with the exception of the corner positions, or,

2) The limlting safety system setting shall be 125% of 1MW as measured on the two high power level safety channels. This safety channel LSSS 1s more conservative than the temperature LSSS.

b) For pulslng operation The l~mitingsafety system settlng shall be 975°F (525°C) as measured in an instrumented fuel element. The instrument element shall be located adjacent to the central bundle with the exception of the comer positions Pulsing is not allowed if this limiting safety system channel is not operable.

The limiting safety system setting (LSSS) is a temperature or reactor power setting that, if exceeded, will cause a reactor scram to be initiated preventing the safety lim~tfrom being exceeded.

The temperature safety hmit for TRIGA-FLIP or LEU fuel is 2100°F (1150°C), while the llrmt for Standard TRIGA fuel 1s 1830°F (1000°C) Due to various errors in measuring temperature in the core, it is necessary to arrive at a Limlting Safety System Settlng (LSSS) for the fuel element safety hmit that takes into account these measurement errors. In doing this analysis, the NSC picked the Standard fuel element temperature limit as the base case to derive the LSSS for all fuel types. One category of error between the true temperature value and the measured temperature value is due to the accuracy of the fuel element channel and any overshoot in reactor power result~ngfrom a reactor translent during steady state mode of operation Although a lesser contributor to error, a minimum safety margin of 10% was apphed, on an absolute temperature basis. Adlusting the Standard fuel temperature safety limit to degrees Kelvin, OK,and applying a 10% safety margin results in a safety l~mitreduction of 150°C. Applying this first margin of safety, the safety setting would be 850°C for Standard fuel and 1000°C for FLIP or LEU fuel. However,

to arrive at the final LSSS r t 1s necessary to allow for the major difference between the measured temperature value and the true temperature value (peak core temperature), whlch is a functlon of the location of the thermocouple within the core. For example, if the thermocouple element were located In the hottest pos~tionin the core, the d~fferencebetween the true and measured temperatures would be only a few degrees smce the thermocouple junction I S at the mid-plane of the element and close to the antlclpated hot spot. However, at the NSC this core posrtlon is not available due to the locatlon of the translent rod For the NSC the location of the instrumented elements 1s therefore restricted to the posrtions closest to the central element Calculations lndicate that, for thls case, the true temperature at the hottest location in the core will d~fferfrom the measured temperature by no more than 4 0 4 . When applylng this 40% worse case measurement scenario to Standard fuel and considering the previously mentioned sources of error between the true and measured values a final LSSS temperature of 975°F (525°C) is lmposed on operat~onwith all types of fuel. Viewed on an absolute temperature scale, OK, this represents a 37% safety margin in the FLWLEU fuel safety 11mlt and a 44% reduction in the Standard fuel safety limit.

In the pulse mode of operatlon, the temperature limltlng safety system setting will apply. However, the temperature channel will have no effect on l~mitingpeak powers generated because of its relat~velylong tlme constant (seconds) as compared wlth the w d t h of the pulse (mdliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting the "tail" off the energy translent in the event the pulse rod remalns stuck In the fully w~thdrawnposition.

t TRIGA-FLIP, LEU or Standard fuel during steady state operatlon will The reactor hlgh power level safety l i m ~ for be a measured power of 125%.(1.25 MW) on either of the two power safety channels. The hlgh power safety drawers are alrgned wlth the linear power monltor durrng annual callbration for nominal 1.0 MW operatlon These safety channels Independently measure reactor power and have been a part of Texas A&M University reactor operation safety systems for over 30 years D u r ~ n gthe years of IMW operation of the TAMU TRIGA reactor the LSSS temperature lrmit of 975°F (525°C) has never been reached although several scrams of the safety channels have been recorded indicating that the LSSS of 125% for the high power level safety channels IS more conservative than the temperature setting.

14.3 Limiting Conditions for Operation 14.3.1 Reactor Core Parameters 14.3.1.1 Steady State Operation Applicability Thls specification applies to the energy generated in the reactor during steady state operatlon.

Obiective The objective is to assure that the fuel temperature safety hmit will not be exceeded durlng steady state operation.

Specifications The reactor power level shall not exceed 1.3 megawatts (MW) under any condition of operatlon. The normal steady state operating power level of the reactor shall be 1.0-1.2 MW. However, for purposes of testlng and calibration, the reactor may be operated at higher power levels not to exceed 1.3 M W during the testmg period.

Thermal and hydraulic calculat~onslndlcate the TRIGA fuel may be safely operated up to power levels of at least 2.0 MW with natural convection coollng.

14.3.1.2 Pulse Mode Operation Applicability Thls specification apphes to the peak temperature generated in the fuel as the result of a pulse insertion of reactivity.

14-7

The objectwe is to assure that respectwe pulsing will not lnduce damage to the reactor fuel a) The reactlvlty to be inserted for pulse operatlon shall not exceed that amount \vh~chwill produce a peak fuel temperature of 1526°F (830°C). In the pulse mode the pulse rod shall be l~mltedby mechanical means or the rod extension physically shortened so that the react~vltyInsertion w ~ l not l inadvertently exceed the maximum value.

b) U n t ~ any l new fuel core has been cahbrated, maxlmum pulse shall be hmited to $2.00 After callbratlon, a new maximum pulse insertion value shall be adhered to.

TRIGA fuel 1s fabricated wlth a nominal hydrogen to zirconlum ratlo of 1 6 for FLIP fuel and 1.65 for standard.

Thls ylelds delta phase zirconlum hydr~dethat has high creep strength and undergoes no phase changes at temperatures over 1000°C. However, after extensive steady state operation at 1 MW the hydrogen will red~stribute due to mlgratlon from the central hlgh temperature regions of the fuel to the cooler outer regions When the fuel is pulsed, the instantaneous temperature distrlbut~onis such that the highest values occur at the surface of the element and the lowest values occur at the center. The higher temperatures In the outer regions occur in fuel with a hydrogen to zirconium ratlo that has now substant~allyIncreased above the nom~nalvalue. T h ~ produces s hydrogen gas pressures considerably in excess of the expected for ZrHl If the pulse insertion is such that the temperature of the fuel exceeds 874°C. then the pressure will be sufficient to cause expansion of microscopic holes in the fuel that grows with each pulse The pulsing limit of 830°C 1s obtained by examining the equihbrium hydrogen pressure of zirconium hydride as a function of temperature. The decrease in temperature from 874OC to 830°C reduces hydrogen pressure by a factor of two, which is an acceptable safety factor. Thls phenomenon does not alter the safety lrmlt since the total hydrogen In a fuel element does not change Thus, the pressure exerted on the clad will not be significantly affected by the distribut~onof hydrogen within the element.

In practice, the pulsing limit of 830°C will be translated to a reactivity Insertion l m i t for each specific core. The peaking factors from the thermocouple element to the hottest spot In the core must be calculated for each core configuration that is to be used. Temperature would then be measured for small pulse insertions For new uncalibrated cores, the pulse insertions shall be increased by small increments to a maximum of $2.00 to allow an extrapolation of peak temperatures, thereby establishing the maximum allowed pulse insertion for a given core. Following approval by the NRC staff of the callbration of the new core, the $2.00 restriction shall be removed.

14.3.1.3 Shutdown Margin Appl~cability These specifications apply to the reactivity condition of the reactor and the reactivity worths of control rods and experiments. They apply for all modes of operatlon.

Obiective The objectwe is to assure that the reactor can be shutdown at all times and to assure that the fuel temperature safety limit will not be exceeded.

Specifications The reactor shall not be operated unless the shutdown margm provided by control rods is greater than $0.25 with.

a) The highest worth non-secured experiment in its most reactive state,

b) The h~ghestworth control rod and the regulating rod (if not scrammable) fully withdrawn. and C) The reactor in the cold condit~onwithout xenon The value of the shutdown margln assures that the reactor can be shut down from any operat~ngc o n d ~ t ~ oeven n ~fthe h~ghestworth control rod should rema~nIn the fully wthdrawn pos~tion If the regulating rod is not scrammable, ~ t s worth 1s not used In determming the shutdown react~vlty.

14.3.1.4 Core Configuration Limitation Appl~cabil~ty This specification applies to mixed cores of FLIP, LEU and standard types of fuel and to full FLIP, LEU or standard cores.

The objective is to assure that the fuel temperature safety lim~twill not be exceeded due to power peaking effects in full FLIP, LEU, or Standard cores and in rn~xedfuel cores a) The TRIGA core assembly may be Standard, FLIP, LEU or a comb~nat~on thereof (m~xedcore) provided that any FLIP and/or LEU fuel core be comprised of at least thirty-five (35) FLIP and/or LEU fuel elements, located in a contlguous, central region b) The instrumented element, if present and servlng as the Limit~ngSafety System, shall be located adjacent to the central bundle w ~ t hthe exception of the comer positions a) In mixed cores, ~tis necessary to specify the minimum number of FLIP and/or LEU elements and arrange them in a contlguous, central reglon of the core to control flux peaking and power generation values in individual elements.

b) Reference. 14.2.2 Limiting Safety System Sett~ng 14.3.1.5 Maximum Excess Reactivity Avvlicabilitv This specification apphes to the maximum excess reactlvity, above cold critical, which may be loaded into the reactor core at any time.

Obiective The objective is to ensure that the core analyzed in the safety analysis report approximates the operational core within reasonable hmits.

Specifications The maximum reactlvity in excess of cold, xenon-free critical shall not exceed 5.5% Aklk ($7.85)

Although ma~nta~ning a mrnrmum shutdown margin at all times ensures that the reactor can be shut down. that specification does not address the total reactiv~tyavarlable u i t h ~ nthe core This specrfication, although over-constrainmg the reactor system. helps ensure that the Ircensee's operat~onalpower densit~es.fuel temperatures and temperature peaks are maintained withrn the evaluated safety lim~ts The specified excess reactivity makes up for negative reactivity due to power coefficrents, xenon poisoning, experrments and fuel deplet~on 14.3.2 Reactor Control and Safety Systems 14.3.2.1 Reactor Control Systems Applicabilrty This specification applies to the informatlon that must be avallable to the reactor operator durrng reactor operation The object~veis to requlre that sufficient informatlon is avallable to the operator to assure safe operation of the reactor Specrfications The reactor shall not be operated unless the measuring channels l~stedin the followrng table are operable I Measuring Channel I

I Mzn. No. Operable I Operating Mode S-S I Pulse 1

Hlgh Power Level I 2 X I Fuel Element Temperature I 1 I I X Llnear Power Level 1 1 X I Log Power Level I 1 I X I Integrated Pulse Power 1 I X Fuel temperature displayed at the control console gives continuous information on this parameter, which has a specified safety limit. The power level monitors assure that the reactor power level is adequately monitored for both steady state and pulsing modes of operation. Monitoring of the high power level channel is important since it is used to ensure the temperature safety limit 1s not reached, slnce the power level is related to the fuel temperature.

14.3.2.2 Reactor Safety Systems Applicabilrtv This specification applies to the reactor safety system circuits.

Obiectrve The objective is to specify the minimum number of reactor safety system channels that must be operable for safe operation

Specifications The reactor shall not be operated unless the safety clrcults described In the following table are operable

- - .- ~ - .

SCRAM @ LSSS (975°F) X Fuel Element Temperature SCRAM @ LSSS (125%) X Hlgh Power Level Hlgh Power Level i SCRAM on loss of supply voltage, or low I X I I

~ e i e c t oPower r Supply power supply I SCRAM at operator's discretion. l x l x Console Scram Button I I Transient rod scram tlme to be 15 seconds I 1 X Preset Tlmer or less after pulse. I Prevent wthdrawal of shim safeties at <4 I X 1 Log Power x l o 3 W (Low count ~nterlock).

Prevent appl~cationof a ~ inr steady state X Transient Rod position mode unless transient rod 1s fullv inserted. 1 I I Shim Safeties & Prevent withdrawal of s h m safeties and X Repulatmg Rod Position regulating rod while in pulse mode The fuel temperature and high power level scrams provide protection to assure that the reactor can be shutdown before the safety limit on fuel element temperature will be exceeded.

In the event of failure of the power supply for a high power level safety channel, operation of the reactor wlthout adequate instrumentation is prevented.

The manual console scram allows the operator to shut down the system ] f a n unsafe or abnormal condition occurs.

The preset timer lnsures that the reactor power level wdl reduce to a low level after pulsing The interlock to prevent startup of the reactor at power levels less than 4 x W which corresponds to approximately 2 cps assures that sufficient neutrons are available for proper startup The interlock to prevent application of air to the transient rod unless the cyllnder is fully inserted 1s to prevent pulsing of the reactor in steady state mode.

The interlock to prevent the withdrawal of the shim safeties or regulating rod in the pulse mode is to prevent the reactor from being pulsed while on a positive penod.

14.3.23 Scram Time Applicability This specification applles to the time requ~redfor the scrammable control rods to be fully inserted from the instant that the fuel temperature safety channel or the hlgh power level safety channel variable reaches their respective Limiting Safety System Setting.

Obiective The objective is to achieve prompt shutdown of the reactor to prevent fuel damage.

The scram time measured from the Instant a simulated signal reaches the value of the LSSS to the Instant that the slo\\est scrammable control rod reaches ~ t fully s Inserted posltlon shall not exceed 1 2 seconds Thls specification assures that the reactor wdl be promptly shutdo\vn when a scram signal is in~tlated Experience and analysis have lndlcated that for the range of transients ant~cipatedfor a TRIGA reactor, the specified scram time IS adequate to assure the safety of the reactor 14.3.3 Confinement 14.33.1 Operations that Require Confinement Avdicab~l~ty This speclficatlon applles to confinement requirements during operatlon of the reactor and the handling of radloactlve materials.

To malntaln normal or emergency alrflow Into and out of the reactor buildmg during operations that produce or could potentially produce alrborne radloactlvity.

Specification Confinement of the reactor bulldlng wdl be required during the following operations.

a) Reactor operating.

b) Handlmg of radioactive materials wlth the potential for airborne release Note: For perlods of maintenance to the central exhaust fan, entry doors to the reactor building will remaln closed except for momentary opening for personnel entry or exlt.

a) This basls applies during the conduct of those actlvlties defined as reactor operations. Argon41 is produced during operatlon of the reactor in exper~mentalfacilities and In the reactor pool; thus, air control within the buildlng and the exhaust system in necessary to maintain proper airborne radiation levels In the reactor building and release levels in the exhaust stack. Other radioactivity releases to the reactor building must be considered during reactor operation, such as fission product release from a leaking fuel element or a release from fixed experiments In or near the core b) The handling of radioactive materials can result in the accidental or controlled release of airborne radioactivity to the reactor buildmg environment or direct release to the buildlng exhaust system. In these cases, the control of alr into and out of the reactor building is necessary.

14.33.2 Equipment to Achieve Confinement Applicability This specification applies to the equlpment and controls needed to provide confinement of the reactor buildmg.

Objective The objective is to assure that a minimum of equlpment 1s In operation to achleve confinement as specified in 3.3.1 and that the control panel for this equ~pmentIS available for normal and emergency situations.

a) The minimum equipment requlred to be In operation to achleve confinement of the reactor building shall be the central exhaust fan b) Controls for estabhshlng the operation of the venrllation system durlng normal and emergency cond~tlons shall be located In the receptlon room Note: During perlods of mamtenance to the central exhaust fan, entry doors to the reactor bulldlng will remain closed except for momentary opening for personnel entry or exit a) Operation of the central exhaust fan will achleve confinement of the reactor buildmg during normal and emergency conditions when the controls for air lnput are set such that the central exhaust fan capaclty remains greater than the amount of air being delivered to the reactor buildmg. The exhaust fan has sufficient capacity to handle extra alr Intake to the building during momentary opening of doors.

b) The control panel for the ventilat~onsystem provldes for manual selection of air input to the reactor budding and the automatic or manual selection of air removal. The am-supply and exhaust systems work together to maintain a small negatlve pressure in the reactor budding. These controls are located In the receptlon room for access~blhtyd u n g emergency conditions.

14.3.4 Ventilation System Au~licabilltv T h ~ specification s applies to the operation of the facillty ventilation system.

The objectwe is to assure that the ventilation system is in operation to mitigate the consequences of the possible release of radloactlve materials resulting from reactor operation Specification The reactor shall not be operated unless the facillty ventilation system is operable, except for periods necessary to permit repair of the system. In the event of a substantial release of airborne radioactivity, the ventllation system will be secured automatically by signals from the appropriate facdity air monitor.

During normal operation of the ventilation system, the concentration of Argon-41 in unrestricted areas is below the effluent concentration (Section 11). In the event of a substantlal release of airborne radioactivity, the ventilation system will be secured automatically. Therefore, operation of the reactor with the ventllation system shutdown for short periods of time to make repairs insures the same degree of control of release of radioactive materials.

Moreover, facility air monitors withln the building independent of those in the ventllation system wdl give warning of high levels of radiation that might occur during operation with the ventilation system secured 14.3.5 Radiation Monitoring Systems and Effluents 14.3.5.1 Radiation Monitoring Avplrcabilitv This specification applles to the radiation monitoring information that must be available to the reactor operator during reactor operation.

The objectlve is to assure that sufficlent radlation monitoring lnformatlon 1s available to the operator to assure safe operation of the reactor.

The reactor shall not be operated unless the radlatlon monitoring channels llsted in the following table are operable.

- Radatzon Monztorrng Channels Functron Number Channel Backup Durzng Mamtenance Area Radiation Monitor (ARM) - Monitor radiation levels 1 Area Radiation Monltor (ARM) -

Reactor Brldge I withln the reactor bay I Materlal Handling Area Facility Air Monitor (FAM) - I Monitor radiation levels I 1 1 Faclllty Air Momtor (FAM) -

Fission Products withm the reactor bay ~ i i l d i particulates n~ .

Facility Air Monitor (FAM) - I Monltor radiation levels of I 1 1 Facll~tvAir Monitor (FAM) -

Stack Exhaust Gas the stack exhaust gases Buildlng Gas Facility Air Monitor (FAM) - Monitor radiation levels of 1 Fac~lttyAir Monitor (FAM) -

I Stack Exhaust Particulates I the stack exhaust particulates I I ~ b i l d i Part~culates n~

Note: For perlods of maintenance to the Rad~atlonMonitoring Channels. the Intent of thls s~eclficationwill be satisfied ;ithey are replaced with the Channel Backups 11stef1nthe precedmg table. 1f two'of the above Radiation Monitor Channels are not operating, the reactor shall be shutdown.

Bases The radlation monltors provide information to operating personnel of any impending or exlsting danger from radlation so that there wlll be sufficlent time to evacuate the facility and take the nccessary steps to prevent the spread of radioactw~tyto the surround~ngs.

14.3.5.2 Argon-41 Discharge Limit Applicab~lity Thls specification applies to the concentratlon of Argon-41 that may be discharged from the TRIGA reactor facility.

Objective To insure that the health and safety of the publlc IS not endangered by the discharge of Argon-41 from the TRIGA reactor facdlty.

Specification The concentration of Argon-41 in the effluent gas from the facility as diluted by atmospheric air in the lee of the facillty due to the turbulent wake effect shall not exceed 1.0 x lvs pCi/ml averaged over one year.

The maximum allowable concentratlon of Argon41 ( 4 1 ~ rin) air in unrestricted areas as specified in Append~xB, Table I1 of 10CFR20 is 1.0 x 10" pCi/ml Sectlon 11 of the SAR for the NSCR substantiates a 5.0 x 1 v 3 atmospheric ddution factor for a 2.0 mph wind speed. This dilut~onfactor represents the conditions at the slte building for a wind speed of 2.0 mph, which occurs less than 10% of the time on an annual basis.

14.3.5.3 Xenon and Iodine Rlonitoring Appl~cablllty This specificatlon applles to the radmtion monltorlng systems necessary to monltor and control the concentration of any effluent releases during the productlon of '"I from the radioactive decay of 12'xe The objectwe 1s to assure that sufficient radiation monitoring information 1s available to the operator to lnsure that the health and safety of the general public is not endangered during the productlon of "'1.

No experiment that involves actlve handling of I2'xe and/or '"I may be performed unless thelr respective radiat~on monitoring system is operable. No experiment may be performed, except decay of ' " ~ e , unless the I2'xe FAM channel is operable.

Radiation Monitoring Channels Funcrlon Number Facihty Air Monitor (FAM) - Monitors IL'xe radiation level 1 Stack Exhaust Gas 1Iz5xe) in stack exhaust gases 1L>

I Alr Monitor Monitors work area lL'I 1 radlation levels Note: For periods of maintenance to the 12'xe FAM channel, the intent of thls specificatlon wlll be satisfied if it is replaced by buildlng gas samples.

The required facility air monitors (14.3.5.1) are not calibrated for 12'1 The ' " ~ e FAM channel provides information to operators in the event there is a significant release of lZSxeduring the production of 12'1 The 12'1 air monitor measures "'I radiation levels in the immediate area where the iodlne handllng is being performed 14.3.6 Limitations on Experiments 14.3.6.1 Reactivity Limits Apvlicabillty This specificatlon applies to the reactivity limits on expenments installed in the reactor and its experimental facilities.

Obiective The objective is to assure control of the reactor during the handling of experiments adjacent to or in the reactor core.

Svecifications The reactor shall not be operated unless the following conditions governing experiments exist a) The reactivity worth of any single, non-secured experiments shall be less than one dollar.

b) The reactivity worth of any single experiment shall be less than two dollars

a) T h ~ specification s 1s intended to provide assurance that the worth of a s ~ n g l eunfastened experiment wlll be hm~tedto a value such that the safety hmlt w11I not be exceeded ~fthe posltlve worth of the experiment were suddenly inserted Thls does not restrlct the number of non-secured experlments adjacent to or in the reactor core.

b) The maxlmum worth of a slngle experiment 1s limlted so that its removal from the cold crltlcal reactor will not result In the reactor achieving a power level high enough to exceed the core temperature safety hmit Slnce experlments of such worth must be fastened in place, its removal from the reactor operating at full power would result in a relatively slow power lncrease such that the reactor protective systems would act to prevent high power levels from belng attained.

14.3.6.2 Material Limitations Applicability This specification applles to experlments mstalled in the reactor and its experimental facilities The objective is to prevent damage to the reactor or excessive release of radloactivity by limtlng materlals quantity and radioactive material inventory of the experiment.

a) Exploslve materials in quantltles greater than 5 pounds shall not be allowed wlthm the reactor building.

Irradiation of explosive materials shall be restricted as follows:

1) Exploslve materlals in quantltles greater than 25 m~lllgramsshall not be irradiated in the reactor pool.

Explosive materials In quantities less than 25 milligrams may be irradlated provided the pressure produced upon detonation of the explosive has been calculated and/or experimentally demonstrated to be less than the design pressure of the container.

2 ) Explosive materlals in quantities greater than 25 milligrams shall be restricted from the reactor pool, the upper research level, the demineralizer room, coollng equipment room and the interior of the pool containment structure.

3) Exploslve materials in quantities greater than 5 pounds shall not be irradlated in experimental facilities.

4) Cumulative exposures for explosive materials in quantities greater than 25 milligrams shall not exceed 1012nlcm2 for neutron or 25 Roentgen for gamma exposures.

b) Each fueled experiment shall be controlled such that the total inventory of lodlne isotopes 131 through 135 in the experiment is no greater than 10 CI.

a) This specification is intended to prevent damage to the reactor or reactor safety systems resulting from failure of an experiment involving explosive materials.

1) This specification is intended to prevent damage to the reactor core and safety related reactor components located within the reactor pool In the event of fallure of an experiment involving the irradiation of explosive materials. Limited quantities of less than 25 milligrams and proper containment of such experiment provide the required safety for in-pool irradlatlon.

2) This specification is Intended to prevent damage to vital equipment by restricting the quantity and location of explosive materials within the reactor building. Explosives in quantities exceeding 25

m~ll~grams are restr~ctedfrom areas contain~ngthe reactor bndge, reactor console, pool water coolant and pur~ficatlonsystems and reactor safety related equipment

3) The failure of an experiment involving the lrradiat~onof up to 5 pounds of explos~vemater~alIn an experimental facihty located external to the reactor pool structure w111not result In damage to the reactor or the reactor pool contamment structure

4) Thls specification 1s intended to prevent any increase In the sensitivity of explosive materials due to radiation damage durlng exposures b) The 10 CI limitat~onon Iodine 131 through 135 assures that in the event of failure of a fueled experiment leading to total release of the iodlne, the exposure dose at the exclusion area boundary wdl be less than that allowed by 10 CFR 20 for an unrestricted area.

14.3.6.3 Failures and Malfunctions Applicabihty This speclficatlon applies to experiments installed In the reactor and its experimental facd~ties.

Objective The objective is to prevent damage to the reactor or excess~verelease of rad~oacttvematerials in the event of an experiment failure.

a) Experiment materials, except fuel materials, which could off-gas, subhme, volatilize, or produce aerosols under (I) normal operating condlt~onsof the experiment or reactor, (2) cred~bleaccident cond~tionsin the reactor, or (3) possible accident conditions In the experiment shall be limited in activity such that if 100%

of the gaseous activity or radioactive aerosols produced escaped to the reactor buildrng or the atmosphere, the airborne concentration of radioactivity averaged over a year would not exceed the llmlt of Appendlx B of 10CFR20.

b) In calculations pursuant to a) above, the following assumptions shall be used.

1) If the effluent from an experimental facility exhausts through a holdup tank that closes automatically on hlgh radiation level, at least 10% of the gaseous activity or aerosols produced will escape.

2) If the effluent from an expenmental facllity exhausts through a filter installation designed for greater than 99% efficiency for 0 3 micron particles at least 10% of these vapors can escape

3) For materials whose boiling point is above 130°F and where vapors formed by boiling this material can escape only through an und~sturbedcolumn of water above the core, at least 10% of these vapors can escape.

c) If a capsule fails and releases material that could damage the reactor fuel or structure by corrosion or other means, removal and physical inspection shall be performed to determrne the consequences and need for corrective action. The results of the inspection and any corrective action taken shall be reviewed by the Director (NSC) or his designated alternate and determined to be satisfactory before operation of the reactor is resumed.

a) This specification is intended to reduce the likelihood that airborne activities in excess of the h i t s of Appendlx B of 10 CFR 20 will be released to the atmosphere outside the facillty boundary of the NSC.

b) These assumptions are used to evaluate the potentla1 alrborne radioactivity release due to an experiment failure.

C) Operat~onof the reactor with reactor fuel or structure damage 1s proh~b~ted to avoid release of fission products. Potentlal damage to reactor fuel or structure must be brought to the attent~onof the Director (NSC) or hts des~gnatedalternate for revlew to assure safe operation of the reactor.

14.3.6.4 Xenon Irradiation for Iodine Production Appllcab~l~ty This specification applies to the experlments that produce 1-125 from the activation of enriched l2%e and the decay of Iz5xe.

Obiective The objectlve is to prevent excessive release of radloactlvlty by limltlng the quantity and radioactwe materlal Inventory of the experiment.

a) Iz4xeactivation experlments shall be controlled such that the total single experiment activity produced is hmlted to no more than 2000 Ci of Iz5xe b) The total faclllty Iz5xeinventory of all experlments shall not exceed 3500 CI a) The 2000 CI limitation on Xenon-125 produced in any one experiment assures that in the event of a failure of an experiment leadlng to the accidental release of xenon, the exposure to the general public would be less than 0.05 rem per year (10 CFR 20 1301).

b) Xenon-125 production In excess of thls limit is not necessary.

14.3.7 As Low As Reasonably Achievable (ALARA) Radioactive Effluents Released Applicability This specification applles to the measures required to ensure that the radioactive effluents released from the facility are in accordance wlth ALARA criteria.

The objectlve is to limit the annual radiation exposure to the general publrc resulting from operation of the reactor to a level as low are reasonably achievable below the limits llsted in 10 CFR 20.1301.

1) In addition to the radiation monitoring specified in Section 14.5.4, an environmental radiation-momtoring program shall be conducted to measure the integrated radiation exposure in and around the environs of the facility on a quarterly basis

2) The annual radiation exposure (dose) to the public due to reactor operation shall not exceed the hmits defined In 10 CFR 20 1301. The facility perimeter shall be monitored to ensure this specification is being met.

3) The total annual discharge of Argon41 into the environment may not exceed 30 CI per year unless permitted by the RSB.

4) In the event of a significant fission product leak from a fuel rod or a slgnlficant alrborne radioactive release from a sample being irradiated, as detected by the continuous faclllty air monltor (FAM). the reactor shall be shut down untll the source of the leak 1s located and eliminated However, the reactor may contmue to be operated on a short-term basis, as needed, to assist In determmng the source of the leakage.

5 ) Before discharge, the fachty liquid effluents collected In the holdup tanks shall be analyzed for the nature and concentrat~onof radioactive effluents The total annual quantlty of radioactivity in liquld effluents (above background) shall not exceed 1 CI per year.

The simplest and most rellable method of ensuring that ALARA release llmits are accomplishing then objectwe of mlnimal facil~ty-causedradiation exposure to the general public is to actually measure the integrated radlatlon exposure in the environment on and off the slte 14.3.8 Primary Coolant Conditions Applicability This specification apphes to the quality of the prlmary coolant In contact wlth the fuel cladding The objectlves are (1) to mlnimize the possibility for corrosion of the claddlng on the fuel elements and (2) to mlnim~zeneutron actlvatlon of dissolved materials.

Specifications

1) Conductivity of the bulk pool water shall be no higher than 5 x mhoslcm (5 psiemenslcm) for a period not to exceed two weeks

2) The pH of the bulk pool water shall be in the range 5.5 and 8.0 (inclusive). Deviations of pH values outside t h ~ range s shall not exceed a period of two weeks A small rate of corrosion continuously occurs in a water-metal system. In order to limit this rate, and thereby extend the longevity and integrity of the fuel cladding, a water cleanup system is required. Experience with water quality control at many reactor facilities has shown that maintenance wthin the specified llmits provides acceptable control By limiting the concentrations of dissolved materials in the water, the radioactivity of neutron activation products is limited. Thrs is consistent with the ALARA principle, and tends to decrease the inventory of radlonuclides in the entlre coolant system, which wlll decrease personnel exposure during maintenance and operations.

14.4 Surveillance Requirements 14.4.1 General A~vlicability This specification, applies to the s u ~ e i l l a n c erequirements of any system related to reactor safety.

The objective is to verlfy the proper operation of any system related to reactor safety.

Any addrtions. modlficatlons, or maintenance to the vent~latlonsystem, the core and ~ t associated s support structure, the pool or ~ t penetrations.

s the pool coolant system, the rod drwe mechanism, or the reactor safety system shall be made and tested in accordance with the specifications to whlch the systems were orrplnally des~pnedand fabricated or to specifications approved by the Reactor Safety Board. A system shall not be considered operable until ~tIS successfully tested.

This specification relates to changes in reactor systems, which could dlrectly affect the safety of the reactor. As long as changes or replacements to these systems contlnue to meet the original deslgn specifications, then ~tcan be assumed that they meet the presently accepted operating cnterla 14.4.2 Reactor Core Parameters 14.4.2.1 Steady State Operation Apd~cabllltv This specification applies to the surveillance requirement of the power level monitoring channels The objectlve IS to verlfy that the maximum power level of the reactor meets the license requirements Svecification A channel callbration shall be made of the power level monitoring channels by the calorimetric method annually but at intervals not to exceed 15 months.

The power level channel calibration will assure that the reactor will be operated at the proper power level.

14.4.2.2 Pulse Mode Operation Appllcabillty This specification applies to the surveillance requirements for operation of the reactor in the pulse mode.

Objective The objective is to verlfy that operation of the reactor in the pulse mode is proper and safe and to determine if any significant changes in fuel characteristics have occurred.

Specification The reactor shall be pulsed semiannually to compare fuel temperature measurements and core pulse energy with those of previous pulses of the same reactivity value or the reactor shall not be declared operational for pulsing until such pulse measurements are performed.

The reactor is pulsed at sultable Intervals to make a comparison with previous similar pulses and to determine if changes In fuel or core characterlstlcs are taking place.

14.4.2.3 Shutdown hlargin Applrcab~l~ty T h ~ specificatlon s applies to the surve~llancerequirement of control rod callbrations and shutdown margln Oblective The objective 1s to verlfy that the requ~rementsfor shutdown margins are met for operational cores.

The reactivity worth of each control rod and the shutdown margin shall be determined annually but at intervals not to exceed 15 months.

The reactivlty worth of the control rods is measured to assure that the requlred shutdown margin is available and to provide an accurate means for determmng the reactivlty worth of experiments inserted in the core. Experience with TRIGA reactors gyves assurance that measurement of the reactivity worth on an annual basls is adequate to Insure no significant changes in the shutdown margin 14.4.2.4 Reactor Fuel Elements A~plicabil~tv This specificatlon applles to the surveillance requirements for the fuel elements.

Oblective The objective is to verlfy the continuing integrity of the fuel element cladding and to ensure that no fuel damage has occurred a) All fuel elements will be inspected vlsually for damage or deterioration and measured for length and bend within a 5-year perlod b) If any element is found to be damaged, the entlre core will be inspected.

c) The reactor shall not be operated knowingly w ~ t hdamaged fuel.

d) A fuel element shall be considered damaged and must be removed from the core if:

1) In measuring the transverse bend, the bend exceeds 0.125 inch over the length of the cladding.

2) In measuring the elongation, its length exceeds its original length by 0.125 inch, or

3) A clad defect exists as indicated by release of fission products.

The frequency of inspection and measurement schedule 1s based on over 30 years of operating experience and on the parameters most llkely to affect the fuel claddmg of a pulsing reactor operated at moderate pulsing levels and ut~lizingfuel elements whose characteristics are well known.

The llmit of transverse bend has been shown to result in no d~fficultyIn disassembling fuel bundles. Analysis of the removal of heat from touching fuel elements shows that there will be no hot spots resulting in damage to the fuel caused by this touching. Experience with TRIGA reactors has shown that fuel element bowing that could result in

rouchlng has occurred wlthout deleterious effects The elongat~onllmlt has been speclfled to assure that the claddlng maferlal wlll not be subjected to stresses that could cause a loss of lntegrlty in the fuel containment and to assure adequate coolant flow 14.4.3 Reactor Control and Safety Systems 14.4.3.1 Reactor Control Systems A~pl~cabll~tv These specificatlons apply to the surveillance requirements for reactor control systems The objective is to verify the condltlon and operabillty of system components affecting safe and proper control of the reactor Specifications The control rods shall be visually inspected for deterloratlon at intervals not to exceed 5 years.

The visual inspection of the control rods is made to evaluate corrosion and wear characteristics caused by operation of the reactor.

14.43.2 Reactor Safety Systems A~plicabillty These specificatlons apply to the surveillance requirements for measurements, tests and calibrations of the control and safety systems.

The objective is to verify the performance and operabillty of the systems and components that are directly related to reactor safety.

a) A channel test of each of the reactor safety system channels for the intended mode of operation shall be performed before each day's operation or before each operation extending more than one day, except for the pool level channel which shall be tested weekly.

b) Whenever a reactor scram caused by hlgh power level or high fuel element temperature occurs, an evaluation shall be conducted to determine whether the fuel element temperature safety limit was exceeded.

c) A callbration of the temperature measuring channels shall be performed semiannually but at intervals not to exceed 8 months.

d) A channel check of the fuel element temperature measunng channel for pulse mode operation and the high level power channels for steady state operation shall be made daily whenever the reactor is operated by recording a measured value of a meaningful temperature or high power level indication.

Channel tests will assure that the safety system channels are operable on a daily basis or prior to an extended run.

Operational experience with the TRIGA system gyves assurance that the thermocouple measurements of fuel

element temperatures and the hlgh power level channels have been sufficiently reliable to assure accurate ~ndicatlon' of these parameters 14.4.3.3 Scram Time Appllcabillty This specification apphes to the surve~llanceof control rod scram times.

The objective is to verify that all scrammable control rods meet the scram tlme requirement The scram time shall be measured annually but at intervals not to exceed 15 months Measurement of the scram time on an annual basls 1s a check not only of the scram system electronics, but also is an indication of the capability of the control rods to perform properly.

14.4.4 Equipment to Achieve Confinement: Ventilation System Auphcabdlty This specification apphes to the building confinement ventilation system Obiective The objective is to assure the proper operation of the ventilation system in controlling releases of radioactive material to the uncontrolled environment During periods of extended operation, or radioactive material handling, the ventilation system shall be verified operable weekly. This specification IS not required dunng periods of non-operation, e.g., holidays, extended maintenance outages.

Experience accumulated over several years of operation has demonstrated that the tests of the ventilation system on a weekly basis are sufficient to assure the proper operation of the system and control of the release of radioactive material.

14.4.5 Radiation Monitoring Systems and Effluents Applicability This specification applles to the surveillance requirements for the area radiation monitoring equipment and the contmuous facillty air monitoring (FAM) system Obiective The objective is to assure that the radiation monitoring equipment is operating and to verify the appropriate alarm settings

The area r a d ~ a t ~ omonltoring n system (ARM) and the f a c ~ l ~air t y monitoring system (FAM) shall be calibrated annually but at intervals not to exceed 15 months and shall be verlfied to be operable at weekly ~ntervals.

Experience has shown that weekly verlficatlon of area radiation and air monltoring system operations in conjunction w t h annual calibration 1s adequate to correct for any varlatlon in the system due to a change of operating characteristics over a long tlme span 14.4.6 Experiments Applicab~l~t~

This speclficatlon apphes to the surveillance requ~rementsfor experiments installed In the reactor and its experimental facilltles and for ~rrad~atlons performed In the irradiation facilities The objectwe is to prevent the conduct of experiments or irradlatlons that may damage the reactor or release excessive amounts of radioactive materials as a result of failure.

a) A new exper~mentshall not be installed in the reactor or ~ t experimental s fac111tlesuntd a hazard analysis has been performed and renewed for compllance with Section 14 6 of the Technical Specificat~ons.Minor mod~ficat~ons to a reviewed and approved experiment may be made at the discretion of the senior reactor operator responsible for the operation with concurrence from a person qualified in health physics. The senlor reactor operator and health physicist must review the hazards associated with the mod~ficat~ons and determine that the modifications do not create a significantly different, a new, or a greater safety risk than the original approved experiment b) The performance of an experiment classified as an approved experiment shall not be performed until a licensed senior operator and a person qualified In health physlcs has reviewed ~t for compllance.

c) The reactivity worth of an experiment shall be estimated or measured, as appropriate, before reactor operat~onwith s a d experiment.

It has been demonstrated over a number of years of experience that experiments and irradiations reviewed by the Reactor Staff and the Reactor Safety Board as appropriate can be conducted without endangering the safety of the reactor or exceeding the limits in the Technical Specifications.

14.5 Design Features 14.5.1 Reactor Fuel Avpl~cabilitv This specification apphes to the fuel elements used in the reactor core.

Obiective The objective 1s to assure that the fuel elements are of such a design and fabricated in such a manner as to permit their use with a high degree of reliablllty wlth respect to their physical and nuclear characteristics.

Specifications a) TRIGA-FLIP Fuel

1) The ind~vldualunlrrad~atedFLIP fuel elements shall have the followmg character~stlcs

2) Uran~umcontent- maxlmum of 9 Wt% enr~chedto nominal 70% Uranium-235.

3) Hydrogen-to-zircon~umatom ratlo (in the ZrHx): nominal 1 6 H atoms to 1.0 Zr atoms

4) Natural erb~umcontent (homogeneously distnbuted): nominal 1.5 Wt%.

5) Cladd~ng.304 stainless steel, nominal 0.020 inch thick.

6) Identification: Top pieces of FLIP elements will have characterist~cmark~ngsto allow visual identificatlon of FLIP elements employed In m ~ x e dcores.

b) Standard TRIGA fuel The individual unirradiated Standard TRIGA fuel elements shall have the following characteristics:

1) Uran~umcontent: maxlmum of 9.0 Wt% enrlched to a nominal 20% Uranlum-235.

2) Hydrogen-to-z~rcon~um atom ratio (in the ZrH,): nom~nal1.7 H atoms to 1.0 Zr atoms.

3) Claddmg 304 stainless steel, nominal 0.020 inch thlck.

c ) TRIGA-LEU Fuel

1) The individual unirradiated LEU fuel elements shall have the following characteristics:

2) Uranium content maximum of 2 0 Wt% enriched to nominal 20% Uranium-235

3) Hydrogen-to-zirconium atom ratio ( ~ the n ZrH,). nominal 1 6 H atoms to 1 0 Zr atoms.

4) Natural erbium content (homogeneously d~stributed):nominal 0 59 Wt%.

5) Cladding: 304 stainless steel, nominal 0 020 inch thick.

a) A maximum uranium content of 9 Wt% in a TRIGA-FLIP element is about 6 % greater than the design value of 8.5 Wt%. Such an increase in loadlng would result in an increase in power density of about 2%.

Similarly, a minimum erbium content of 1.1% in an element is about 30% less than the design value. This variation would result in an increase In power density of only about 6%. An increase in local power density of 6% reduces the safety margin by at most ten percent. The maximum hydrogen-to-zircomum ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than the value resulting from a hydrogen-to-zirconium ratio of 1 60. However, this increase in the clad stress during an accident would not exceed the rupture strength of the clad.

When standard and FLIP andlor LEU fuel elements are used in mixed cores, visual ldentlficatlon of types of elements is necessary to verify correct fuel loadings. The accidental rotatlon of fuel bundles containmg standard and FLIP andor LEU elements can be detected by visual inspection. Should this occur, however, studies of a single LEU element accidentally rotated into a standard fuel reglon indicate an insubstantlal increase in power generation in the LEU element.

b) A maxlmum uranlum content of 9 Wt% 7cn a standard TRIGA element IS about 6% greater than the deslgn value of 8 5 Wt%. Such an increase In load~ng\vould result in an increase in power density of less than 6%. An Increase in local power density of 6% reduces the safety margln by at most 10%. The maxlmum hydrogen-to-nrconium ratlo of 1 8 wdl produce a maxlmum pressure withln the clad durmg an accldent well below the rupture strength of the clad c) The Department of Energy (DOE) 1s facilitating the conversion of HEU reactors such as the Texas A&M University NSC TRlGA reactor. The DOE controls the timetable for conversron to LEU fuel 14.5.2 Reactor Core Appllcabillty This specification applies to the configuration of fuel and In core experlments The objectrve IS to assure that provisions are made to restrict the arrangement of fuel elements and experlments to provide assurance that excessive power densities w ~ l not l be produced The core shall be an arrangement of TRIGA uranium-zlrconlum hydride fuel-moderator bundles positioned in the reactor grid plate.

The reflector, excluding experiments and experimental facilities, shall be water or a combination of graphite and water or D,O.

Standard TRIGA cores have been in use for years and then characteristics are well documented. FLIP cores have been operated at General Atomics and the Puerto Rico Nuclear Center and thelr operational characteristics are available. General Atomlcs has also performed a series of experiments uslng standard and FLIP fuel in mixed cores In addition, studies performed at Texas A&M for a varlety of mixed core arrangements and operatlonal experience with mixed cores lndlcate that such loadings would safely satisfy all operatlonal requirements General Atomics and Texas A&M has done a serles of studles documenting the viabll~tyof using LEU fuel In TRIGA reactors.

The core wlll be assembled In the reactor grid plate that is located in a pool of llght water. Water In combination with graphlte reflectors can be used for neutron economy and the enhancement of experimental facdity radlatlon requirements Control Rods This specification apphes to the control rods used in the reactor core.

The objective is to assure that the control rods are of such a deslgn as to permit thelr use with a hlgh degree of reliability with respect to their physical and nuclear characteristics.

Specifications a) The shim-safety control rods shall have scram capability and contain borated graphlte, B4C powder or boron and ~ t compounds s in sohd form as a poison in aluminum or stamless steel cladding These rods may incorporate fueled followers that have the same characteristics as the fuel region in which they are used.

b) The regulating control rod need not have scram capablllty and shall be a sta~nlessrod or contaln the materials as specified for shim-safety control rods. Thts rod may incorporate a fueled follower.

c) The translent control rod shall have scram capabil~tyand contaln borated graphlte or boron and its compounds in solid form as a polson In an alum~numor stainless steel clad The translent rod shall have an adjustable upper l ~ mto~ allow t a variat~onof reactivity insertions. T h ~ rods may Incorporate an aluminum or alr follower.

Usmg neutron absorb~ngborated graphite, B4C powder or boron and ~ t compounds, s satisfies the poison requirements for the control rods. Since the regulat~ngrod normally is a low worth rod, using a sohd stainless steel rod could satisfy 11s function. These materials must b e contained in a suitable clad material, such as aluminum or stainless steel, to Insure mechanical stability during movement and to isolate the polson from the pool water environment. Control rods that are fuel followed provlde add~tionalreactivity to the core and Increase the worth of the control rod. The use of fueled followers In the FLIP region has the additional advantage of reduclng flux peak~ngin the water filled reglons vacated by the withdrawal of the control rods Scram capabilities are provided for rapid insertion of the control rods, w h ~ c his the primary safety feature of the reactor. The transient control rod is designed for a reactor pulse The nuclear behavior of the alr or aluminum follower that may be incorporated into the transient rod 1s slmllar to a void. A voided follower may be required in certain core loadings to reduce flux pealung values.

14.5.4 Radiation Monitoring System Applicabil~ty This specification describes the functions and essential components of the area radiation monitoring (ARM) equipment and the facil~tyair mon~toring(FAM) system for continuously monltoring airborne radioactivity.

The objective IS to descr~bethe radiation mon~tor~ng equipment available to the operator to assure safe operation of the reactor.

Speclficat~on The radiation monitoring equipment listed in the following table will have these charactenstics.

I Rodration Monitoring Channel 14.5.4.1 Detector Type 14.5.4.2 Function Area Radiat~onMonltor (ARM) Gamma sensitive instruments. II Monitor radiation fields in key locations.

Alarm and readout in the control room and readout in the reception room.

Beta-Gamma sensitive I Monitors concentration of airborne Particulates detector. radioactive particulate activity. Alarm and readout in the control room and I readout in the reception room Fac~lityAir Monitor (FAM) - Gamma sensitive 1 Monitors concentration of radioactive Gases detector. gases. Alarm and readout in the control room and readout in the reception room The rad~ationmonitoring system is intended to provlde information to operating personnel of any impending or existing danger from radiation so that there will be sufficient time to evacuate the facility and take the necessary steps to prevent the spread of radioactivity to the surround~ngs.

14.5.5 Fuel Storage Appllcabllitv This specification applies to the storage of reactor fuel at tlmes when l t IS not In the reactor core Obiective The objectwe IS to assure that fuel that is bemg stored n.111 not become crltical and will not reach an unsafe temperature.

Specifications a) All fuel elements shall be stored In a geometrical array for which the k-effect~veis less than 0 8 for all conditions of moderation b) Irradiated fuel elements and fueled devlces shall be stored in an array which wlll permlt sufficient natural convection cooling by water or air such that the fuel element or fueled device temperature will not exceed design values.

The limits Imposed by Specifications 14 5.5 a and 14 5 5.b are conservative and assure safe storage.

14.5.6 Reactor Building and Ventilation System Applicability This specification applies to the building that houses the reactor.

The objective is to assure that provisions are made to restrlct the amount of release of radloactlvlty Into the environment a) The reactor shall be housed in a fac~lltydesigned to restrict leakage. The minimum free volume in the facility shall be 180,000 cublc feet b) The reactor budding shall be equipped with a ventllation system deslgned to filter and exhaust air or other gases from the reactor buildmg and release them from a stack at a minimum of 85 feet from ground level.

c) Emergency shutdown controls for the ventilation system shall be located in the reception room and the system shall be designed to shut down in the event of a substantla1 release of fission products.

The facility is designed such that the ventllation system will normally maintain a negative pressure with respect to the atmosphere s o that there will be no significant uncontrolled leakage to the environment. The free air volume within the reactor buildmg is confined when there is an emergency shutdown of the ventilatlon system. Controls for startup, emergency filtering, and normal operation of the ventilatlon system are located in the reception room.

Proper handling of airborne radioactive materials (in emergency situations) can be conducted from the reception room minimizing exposure to operating personnel

14.5.7 Reactor Pool Water Systems Appllcabllity This spec~ficat~on apphes to the pool contaming the reactor and to the cooling of the core by the pool water.

Obiect~ve The object~veis to assure that coolant water shall be avadable to provide adequate cooling of the reactor core and adequate radiation shielding Specifications The reactor core shall be cooled by natural convective water flow The pool water lnlet and outlet pipe for the demineral~zer,d~ffuserand skimmer systems shall not extend more than 15 feet below the top of the reactor pool when fuel is In the core.

Pool water inlet and outlet pipes to the heat exchanger shall have emergency covers w~thinthe reactor pool for manual shut off In case of pool water loss due to external plpe system failure.

A pool level alarm shall indlcate loss of pr~marycoolant before or equal to the pool level dropping to 10%

below the normal operating level.

This specification is based on thermal and hydraulic calculat~onswhlch show that the TRIGA-FLIP core can operate continuously in a safe manner at power levels up to 2,700 kW with natural convection flow and sufficient bulk pool coollng. A comparison of operation of the TRIGA-FLIP and standard TRIGA Mark 111 has shown them to be safe for the above power level Thermal and hydraulic characteristics of mixed cores are essentially the same as that for TRIGA-FLIP and standard cores In the event of accidental siphoning of pool water through inlet and outlet pipes of the demineralizer, skimmer or d~ffusersystems, the pool water level will drop to no more than 15 feet from the top of the pool.

Inlet and outlet coolant hnes to the pool heat exchanger terminate at the bottom of the pool. In the event of pipe fallure, these llnes must be manually sealed from within the reactor pool. Covers for these lines will be stored in the reactor pool. The time required to uncover the reactor core due to failure of a single pool coolant pipe system is 17 minutes.

Coolant water loss of 10% or more requires corrective action. This alarm 1s observed in the reactor control room and in the reception room.

14.6 Administrative Controls 14.6.1 Organization The Nuclear Science Center is operated by the Texas Engineering Experiment Stat~on(TEES). The Director of the Nuclear Science Center is responsible to the Director of the TEES for the administration and the proper and safe operation of the facility. Figure 10-1 shows the administration chart for the Nuclear Sclence Center.

The Reactor Safety Board advlses the Director of the NSC on all matters or p o k y pertaining to safety.

The NSC Rad~ologicalSafety Officer provides "onsite" advice concerning personnel and rad~olog~cal safety and provides technical assistance and review in the area of radiation protection.

14.6.1.1 Stri~cture A 11nemanagement organlzatlonal structure provldes admlnistratlon and operatlon of the reactor faclllty.

The Deputy Dlrector of the Texas Engmeering Experiment Statlon (TEES) and the Dlrector of the Nuclear Sclence Center (NSC) have h e management responsiblllty for adhermg to the terms and condltlons of the Nuclear Science Center Reactor (NSCR) llcense and technical specifications and for safeguarding the publlc and facllity personnel from undue radiation exposure The faclllty shall be under the direct control of the Dlrector (NSC) or a licensed senlor reactor operator 14.6.1.1.1 Management Levels Level 1: Deputy Dlrector TEES (Licensee). Responsible for the NSCR facllity license.

Level 2: Director (NSC). Responsible for reactor facility operatlon and shall report to Level 1 Level 3: Senior Reactor Operator on Duty: Responsible for the day-to-day operation of the NSCR or shift operatlon and shall report to Level 2.

Level 4: Reactor Operatmg Staff. Llcensed reactor operators and senior reactor operators and trainees These individuals shall report to Level 3.

14.6.1.1.2 Radiation Safety A qualified, health physicist has the responsibility for implementation of the radiation protection program at the NSCR. The indlvldual reports to Level 2 management.

14.6.1.1.3 Reactor Safety Board (RSB)

The RSB is responsible to the Llcensee for providing an independent review and audit of the safety aspects of the NSCR.

14.6.1.2 Responsibility Responsibility for the safe operatlon of the reactor facillty shall be in accordance with the lme organization established above.

14.6.1.3.1 T h e minimum staffing when the reactor is not secured shall be as follows:

1) A licensed senior reactor operator and either a licensed reactor operator or trainee shall be present at the facility.

2) A llcensed reactor operator or senlor reactor operator will be in the Control Room

3) The Director (NSC) or his designated alternate is readily available for emergencies or on call (i.e.. capable of getting to the reactor facillty within a reasonable time).

4) At least one individual qualified in health physlcs will be readily available at the facility or on call (~.e.,

capable of getting to the reactor facility within a reasonable time) 14.6.1.3.2 A list of reactor facility personnel by name and telephone number shall be read~lyavailable for use in the control room The list shall include:

1) Administratwe personnel

2) Radiation safety personnel

3) Other operations personnel

14.6.1.3 3 The following designated individuals shall direct the events listed:

1) The Director (NSC) or hls designated alternate shall dlrect any loading of fuel or control rods wthin the reactor core regton 2 ) The Director (NSC) or hts destgnated alternate shall direct any loading of an m-core experiment ~vltha reactlvtty worth greater than one dollar

3) The sentor reactor operator on duty shall dlrect the recovery from an unplanned or unscheduled shutdown other than a safety hmit violatton.

14.6.1.4 Selection and Training of Personnel A training program for reactor operations personnel exlsts to prepare personnel for the USNRC Operator or Senior Operator examination This training program normally contams twenty hours of lecture, outside study, and requires several reactor startups 14.6.1.4.1 The selection and training of operations personnel shall be in accordance with the following.

1) Responslbility:

a) The D~rector(NSC) or h ~ designated s alternate is responsible for the training and requaltfication of the facillty reactor operators and senlor reactor operators.

2) Requal~ficationProgram a)

Purpose:

To insure that all operating personnel malntam proficiency at a level equal to or greater than that requlred for initlal licensing b) Scope.

Scheduled lectures, written examlnatlons and evaluated console man~pulattonsinsure operator profic~ency.

14.6.1.5 Radiation Safety Members of the health physics staff routinely perform radlatlon safety aspects of faclllty operations, Including routine radlatlon and contammation surveys, and alr and water sampling. Chapter 11 details the radlatlon safety program for this license.

14.6.2 Reactor Safety Board (RSB) Review and Audit Activities A Reactor Safety Board (RSB) acts as a revlew panel for new reactor experiments, procedural changes and facillty modifications. The RSB thus provides an independent audit of the operations of the Nuclear Sclence Center. Issues concerning nuclear safety are immediately brought to the attention of the RSB. The University Radiological Safety Office provides Health Physics assistance for the Nuclear Science Center. This organizational arrangement thus provides another independent revlew of reactor operations (F~gure10-1).

14.6.2.1 RSB Composition and Qualifications The Reactor Safety Board (RSB) shall consist of at least three voting members knowledgeable in fields that relate to nuclear safety. The RSB shall review, evaluate and make recommendatlons on safety standards assoc~atedwith the operational use of the facility. Members of NSC operations and health physics may be ex-officio members on the RSB. The review and advisory functions of the RSB shall include NSCR operations, radiation protection and the facility license. The Chairman of the Reactor Safety Board under the direction of the Deputy D~rectorof TEES shall appoint the board members.

14.6.2.2 RSB Charter and Rules The operations of the RSB shall be in accordance with a written charter, including provisions for.

1) Meeting frequency: not less than once per calendar year and as frequent as circumstances warrant consistent with effective monitoring of facility activities.

2) Votmg rules

3) Quorums

4) Use of subcommittees

5) Revlew, approval and dissemination of minutes 14.6.2.3 RSB Review Function The revlew responsibllltles of the Reactor Safety Committee shall include, but are not llmited to the following.

Revlew and approval of new experiments utlllzing the reactor facllities; Review and approval of all proposed changes to the facility, procedures, license and technical specifications, Determination of whether a proposed change, test or experiment would constitute an unrevlewed safety question or a change in Technical Specification; Review of abnormal performance of plant equipment and operating anomalies having safety significance, Review of unusual or reportable occurrences and incidents that are reportable under 10CFR20 and 10CFRSO, Review of audit reports; and Review of violations of technical specifications, Ilcense, or procedures and orders having safety significance 14.6.2.4 RSB Audit Function The RSB or a subcommittee thereof shall audit reactor operations and radiation protection programs at least quarterly, but at intervals not to exceed four months Audits shall include but are not limited to the following:

1) Facility operations, including radiat~onprotection, for conformance to the technical specifications, applicable license conditions, and standard operating procedures at least once per calendar year (interval between audits not to exceed 15 months);

2) The retraining and requalificat~onprogram for the operating staff at least once per calendar year (~nterval between audits not to exceed 15 months);

3) The facility security plan and records at least once per calendar year (interval between audlts not to exceed 15 months);

4) The reactor facility emergency plan and implementing procedures at least once per calendar year (interval between audlts not to exceed 15 months).

The licensee or his designated alternate (excluding anyone whose normal job function is within the NSCR) shall conduct an audlt of the reactor faciltty ALARA program at least once per calendar year (interval between audits not to exceed 15 months). The licensee shall transmit the results of the audit to the RSB at the next scheduled meetmg.

14.6.3 Procedures The philosophy of nuclear safety at the Nuclear Science Center assumes that all operations utiliztng the reactor will be carrled out in such a manner as to protect the health and safety of the publlc. This philosophy is augmented in practice by detailed, written procedures. All personnel using the facilities of the Nuclear Science Center follow the procedures. The loading or unloading of any core is performed according to detailed written procedures. Startup

and operatlon of the reactor 1s also performed accord~ngto detalled wrrtten procedures Wr~ttenoperating procedures shall be prepared, revre\ved and approved before lnltlatlng any of the actlvltles listed In this sectlon. The procedures shall be rewewed and approved by the Dlrector (NSC), or his designated alternate, the Reactor Safety Board, and shall be documented In a trmely manner Procedures shall be adequate to assure the safe operation of the reactor but shall not preclude the use of independent judgment and action should the situation requlre such Operatmg procedures shall be rn effect for the followrng items Startup, operatlon, and shutdown of the reactor, Fuel and experiment loadmg, unloadmg, and movement wrthln the reactor; Control rod removal or replacement, Routine maintenance of the control rod, drwes and reactor safety and rnterlock systems or other routine maintenance that could have an effect on reactor safety; Testmg and calrbratlon of reactor instrumentation and controls, control rod drives, area radiation monitors, and facrllty air monitors; C~vrldisturbances on or near the facrlity site; Implementat~onof required plans such as emergency or security plans; and Actions to be taken to correct specific and foreseen potential malfunctions of systems, lncludlng responses to alarms and abnormal reactivity changes The Director (NSC) and the Reactor Safety Board shall make substantrve changes to the above procedures effective only after documented review and approval. The Director (NSC) or his designated alternate may make only minor modifications or temporary changes to the orlglnal procedures that do not change thelr original intent All such temporary changes shall be documented and subsequently reviewed by the Reactor Safety Board 14.6.4 Experiment Review and Approval Approved experiments shall be carried out in accordance wlth establrshed and approved procedures

1) All new experiments or class of experrments shall be reviewed by the RSB and implementation approved in writing by the Director (NSC) or his designated alternate.

2) Substantive changes to previously approved experiments shall be made only after review by the RSB and implementation approved rn writing by the Director (NSC) or hls designated alternate. The Director (NSC) or his designated alternate may approve minor changes that d o not significantly alter the experiment.

14.6.5 Required Actions 14.6.5.1 Action to be Taken in the Event a Safety Limit is Exceeded In the event a safety hmit is exceeded:

1) The reactor shall be shut down and reactor operation shall not be resumed until authorized by the NRC.

2) An immediate report of the occurrence shall be made to the Chairman, Reactor Safety Board, and reports shall be made to the NRC in accordance with Section 14.6 6.2 of these specifications, and

3) A report shall be prepared which shall include an analysis of the cause and extent of possible resultant damage, efficacy of corrective action, and recommendatlons for measures to prevent or reduce the probability of recurrence. This report shall be submitted to the Reactor Safety Board for review and then submitted to the NRC when authorization is sought to resume operation of the reactor.

14.6.5.2 Action t o be Taken in the Event of a Reportable Occorrence In the event of a reportable occurrence, the follow~ngactlon shall be lahen:

1) NSC staff shall return the reactor to normal operatrng or shut down cond~trons If rt 1s necessary to shut down the reactor to correct the occurrence. operatrons shall not be resumed unless authorrzed by the D~rector(NSC) or his des~gnatedalternate.

2) The D~rector(NSC) or hrs designated alternate shall be notrfied and corrective actron taken with respect to the operations involved

3) The D~rector(NSC) or h ~ desrgnated s alternate shall notify the Chairman of the Reactor Safety Board.

4) A report shall be made to the Reactor Safety Board wh~chshall mclude an analysis of the cause of the occurrence, efficacy of correctrve action, and recommendations for measures to prevent or reduce the probabil~tyof recurrence, and

5) A report shall be made to the NRC In accordance wrth Section 6.6 2 of these specifications.

6) Occurrence shall be reviewed by the RSB at their next scheduled meetlng 14.6.6 Reporting Requirements 14.6.6.1 Annual Report An annual report covermg the operation of the reactor fac111tyduring the prevlous calendar year shall be submrtted to the NRC before March 31 of each year prov~drngthe following information A) A brief narrative summary of (1) operating experience (~nclud~ng experiments performed), (2) changes in facility des~gn,performance characterist~cs,and operating procedures related to reactor safety and occurring during the reporting period, and (3) results of surveillance tests and inspections; B) Tabulation of the energy output (in megawatt days) of the reactor, hours reactor was cr~tical,and the cumulative total energy output since initial crit~cality, C) The number of emergency shutdowns and inadvertent scrams, includmg reasons thereof; D) Discuss~onof the major maintenance operations performed during the period, including the effect, if any, on the safety of the operation of the reactor and the reasons for any corrective maintenance required; E) A brief description, includmg a summary of the safety evaluations of changes in the facil~tyor in procedures and of tests and experiments camed out pursuant to Section 50.59 of 10 CFR Part 50, F) A summary of the nature and amount of radioactive effluents released or d~schargedto the environs beyond the effect~vecontrol of the licensee as measured at or before the point of such release or d~scharge.If the estimated average release after d~lutionor diffus~onis less than 25% of the concentration allowed or recommended, a statement to this effect is sufficient.

1) Liquid Waste (summarized on a monthly basis) a) Radioactivity discharged during the report~ngpenod.

(1) Total radioact~v~ty released (in Curies)

(ii) The Effluent Concentration used and the rsotopic composition if greater than 1 x pClIcc for fission and activat~onproducts

(lil) Total radloactluty ( ~ cunes),

n released by nucllde, during the reportlng perlod based on representatlye isotopic analysrs (IV) Average concentration at polnt of release (in pCl/cc) durlng the reporting perlod b) Total volume (In gallons) of effluent water (mcluding dilution) during per~odsof release.

2) Airborne Waste (summarlzed on a monthly basis) a) Radloactivlty discharged during the reportlng per~od(in Cunes) for, (ii) Particulates wlth half-lives greater than eight days.

3) Solid Waste a) The total amount of sohd waste transferred (in cubic feet).

b) The total actlvity mvolved (in Curles) c) The dates of sh~pmentand dlsposit~on(if shipped off site).

G) A summary of r a d ~ a t ~ oexposures n recewed by fachty personnel and vis~tors,includ~ngdates and time where such exposures are greater than 25% of that allowed or recommended.

H) A description and summary of any environmental surveys performed outside the facil~ty.

14.6.6.2 Special Reports In add~tionto the requirements of appl~cableregulations, reports shall be made to the NRC Document Control Desk and special telephone reports of events should be made to the Operations Center as follows:

1) There shall be a report not later than the following working day by telephone and confirmed In writing by telegraph or similar conveyance to be followed by a written report that describes the circumstances of the event within 14 days of any of the following:

a) Violat~onof safety limits (See Required Actions).

b) Any accidental release of radioactivity above permissible limits in unrestricted areas whether or not the release resulted in property damage, personal injury, or exposure; c) Any reportable occurrences as defined in the Specifications. The written report (and, to the extent possible, the prelimnary telephone or telegraph report) shall describe, analyze, and evaluate safety implications, and outline the corrective measures taken or planned to prevent reoccurrence of the event;

2) A written report within 30 days of.

a) Personnel changes In the facility organization involving Level 1 and Level 2.

b) Significant changes In the transient or accident analysis as described in the Safety Analysts Report.

14.6.7 Records A daily reactor operations log is maintained by the reactor operator, and contains such information as core loading, experiments in the reactor, time of ~nsertionand removal of experiments, power levels, time of startup and shutdown, core excess reactivity, fuel changes, and reactor instrumentation records.

Records are mamtained whlch indicate the review, approval and conditions necessary for the production of radioisotopes or performance of irrad~at~on experments Records of facll~tyoperations in the form of logs. data sheets or other suitable forms are retamed for the per~od mdicated In the followng sections 14.6.7.1 Records to be retained for a Period of at Least Five Years or for the Life of the Component ln~olved Normal reactor facility operation Principal maintenance operattons Reportable occurrences Surveillance activitm required by the Technical Specifications Reactor facility radiat~onand contamination surveys where required by applicable regulations Experiments performed with the reactor Fuel inventories, receipts, and shipments Approved changes In operating procedures Records of meeting and a u d ~reports t of the RSB 14.6.7.2 Records to be retained for at Least One Training Cycle

1) Retraining and Requalification of certified operations personnel.

2) Records of the most recent complete cycle shall be maintained for ind~vidualsemployed 14.6.7.3 Records to be retained for the Lifetime of the Reactor Facility

1) Gaseous and liquid radioactive effluents released to the environs.

2) Off-site environmental monitoring surveys required by the Technical Specifications.

3) Rad~ationexposure for all personnel monitored.

4) Drawings of the reactor facility.

15 FINANCIAL QUALIFICATIONS 15.1 Financial Ability to Construct a Non-Power Reactor Texas A&M IS not seeking to construct.

15.2 Financial Ability to Operate a Non-Power Reactor The Nuclear Sc~enceCenter is part of the Texas A&M Engmeermg Experiment Station (TEES). TEES provides an annual budget to the NSC deslgned to mantain the facdity ava~lablefor both academics and research. Whde the NSC has its own independent budget, ~t does share personnel with other departments, prlmar~lythe Nuclear Engineering Department in the College of Engineering For example, the NSC Director is a professor of Nuclear Engineer~ngand so the NSC pays a portion of his salary whlle the Nuclear Engineering Department pays the remamder. In thls information, there 1s no effort to take into account the total cost of the shared personnel This information reflects the actual independent budget of the Nuclear Sclence Center. As a result, thls ~nformatlon1s subject to change from year to year depending on the needs and resources of the Nuclear Sclence Center and other departments.

The operating budget for the Nuclear Science Center is not fixed. Whlle TEES provides a fixed amount to the NSC budget, most of the operating budget comes from servlces provided to commerc~alusers, researchers and educators The information here is for the 2001-2002 fiscal year. These values have not changed significantly for several years The total operating budget for the 2002 fiscal year was $773,000 Of this, earned commercial Income covered to

$415,000, State Funding $250,000, earned income through academic services $50,000, earned income through Texas A&M Experment Engineering Statron's Research Enhancement fund $22,000; fees for servlces paid through Department of Energy's Reactor Sharlng program $24,000, and fees for services provided to other Texas A&M Departments $13,000 The largest part of our expenses is salar~esand varies with demand for services TEES administration has been supportwe of the reactor facillty and continuation of the Nuclear Science Center to support various curricula including Nuclear Engineering curr~culum.

Much of the capital equipment funding in recent years has come from the DOE program to update the mstrumentation and experimental equipment for non-power reactors.

15.3 Financial Ability to Decommission the Facility The University of W~sconsinestimated the cost of decommissioning a similar facillty in twenty years to be between

$3 million and $8 million. TEES is a state agency and will be able to obtain the funding when necessary.

16 OTHER LICENSE CONSIDERATIONS 16.1 Prior Use of Reactor Components The Texas A&M Nuclear Science Center uses FLIP fuel from a research reactor in Puerto Rlco. Chapter 4 of this SAR addresses the measured characterlstlcs of thrs fuel. Chapter 13 considers this fuel for accident analysls The NSC has had thls fuel In operation for over 20 years; the characteristics and operating parameters are well known 16.2 Medical Use of Non-Power Reactors Texas A&M Nuclear Sc~enceCenter does not engage in nor is it llcensed to engage In any actwlties for medlcal use of the facility.