ML090260292

Draft Safety Evaluation Report Related to the Renewal of Facility Operating License No. TR 5 for the National Bureau of Standards Test Reactor, National Institute of Standards and Technology

ML090260292
Person / Time
Site:	National Bureau of Standards Reactor
Issue date:	01/26/2009
From:	William Kennedy Research and Test Reactors Licensing Branch
To:
Kennedy W, NRR/ADRA/DPR/PRT, 415-2784
References
Download: ML090260292 (20)

Text

Draft Safety Evaluation Report Related to the Renewal of Facility Operating License No. TR-5 for the National Bureau of Standards Test Reactor, National Institute of Standards and Technology January 2009 Office of Nuclear Reactor Regulation

ABSTRACT This safety evaluation report summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. The staff conducted this review in response to a timely application filed by the National Institute of Standards and Technology (the licensee or NIST) for a 20-year renewal of Facility Operating License No. TR-5 to continue to operate the National Bureau of Standards Test Reactor (NBSR or the facility). The facility is located at the NIST campus in Gaithersburg, Maryland. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensees annual reports to the NRC) as well as inspection reports prepared by NRC personnel and first-hand observations. On the basis of this review, the staff concludes that NIST can continue to operate the NBSR, in accordance with the renewed license, without posing a significant risk to the health and safety of the public, facility personnel, or the environment.

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CONTENTS Page TABLE OF CONTENTS.............................................................................................................. ii LIST OF ABBREVIATIONS ........................................................................................................vi 1 INTRODUCTION ............................................................................................................. 1-1 1.1 Overview .................................................................................................................. 1-1 1.2 Summary and Conclusions on Principal Safety Considerations................................ 1-2 1.3 General Facility Description ..................................................................................... 1-3 1.4 Shared Facilities and Equipment .............................................................................. 1-3 1.5 Comparison with Similar Facilities ............................................................................ 1-3 1.6 Summary of Operation ............................................................................................. 1-4 1.7 Compliance with the Nuclear Waste Policy Act of 1982............................................ 1-4 1.8 Facility Modifications and History ............................................................................. 1-4 2 SITE CHARACTERISTICS .............................................................................................. 2-1 2.1 Geography and Demography ................................................................................... 2-1 2.1.1 Geography ....................................................................................................... 2-1 2.1.2 Demography ..................................................................................................... 2-1 2.2 Nearby Industrial, Transportation, and Military Facilities........................................... 2-1 2.3 Meteorology ............................................................................................................. 2-2 2.4 Hydrology ................................................................................................................. 2-3 2.5 Geology, Seismology, and Geotechnical Engineering .............................................. 2-4 2.6 Conclusions ............................................................................................................. 2-5 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS ..................................... 3-1 3.1 Design Criteria ......................................................................................................... 3-1 3.2 Meteorological Damage ........................................................................................... 3-1 3.3 Water Damage ......................................................................................................... 3-2 3.4 Seismic Damage ...................................................................................................... 3-2 3.5 Systems and Components ....................................................................................... 3-2 3.6 Conclusions ............................................................................................................. 3-3 4 REACTOR DESCRIPTION .............................................................................................. 4-1 4.1 Summary Description ............................................................................................... 4-1 4.2 Reactor Core............................................................................................................ 4-1 4.2.1 Reactor Fuel ..................................................................................................... 4-1 4.2.2 Control Rods .................................................................................................... 4-2 4.2.3 Neutron Moderator and Reflector ..................................................................... 4-4 4.2.4 Neutron Startup Source .................................................................................... 4-5 4.2.5 Core Support Structure..................................................................................... 4-6 4.3 Reactor Tank or Pool ............................................................................................... 4-6 4.4 Biological Shield ....................................................................................................... 4-8 4.5 Nuclear Design ........................................................................................................ 4-9 4.5.1 Normal Operating Conditions ........................................................................... 4-9 4.5.2 Reactor Core Physics Parameters.................................................................... 4-9 4.5.3 Operating Limits ............................................................................................. 4-11 4.6 Thermal-Hydraulic Design ...................................................................................... 4-11 4.7 Conclusions ........................................................................................................... 4-13 5 REACTOR COOLANT SYSTEMS ................................................................................... 5-1 5.1 Summary Description ............................................................................................... 5-1 5.2 Primary Coolant System........................................................................................... 5-1 ii

CONTENTS (cont.)

Page 5.3 Secondary Coolant System ...................................................................................... 5-2 5.4 Primary Coolant Cleanup System ............................................................................ 5-3 5.5 Primary Coolant Makeup Water System ................................................................... 5-4 5.6 D2O Experimental Cooling System ........................................................................... 5-4 5.7 Conclusions ............................................................................................................. 5-4 6 ENGINEERED SAFETY FEATURES .............................................................................. 6-1 6.1 Summary Description ............................................................................................... 6-1 6.2 Detailed Descriptions ............................................................................................... 6-1 6.2.1 Confinement ..................................................................................................... 6-1 6.2.2 Ventilation ........................................................................................................ 6-1 6.2.3 Emergency Core Cooling System ..................................................................... 6-2 6.3 Conclusions ............................................................................................................. 6-2 7 INSTRUMENTATION AND CONTROL SYSTEMS .......................................................... 7-1 7.1 Summary Description ............................................................................................... 7-1 7.2 Design of Instrumentation and Control Systems ....................................................... 7-1 7.3 Reactor Control System ........................................................................................... 7-1 7.4 Reactor Protection System ....................................................................................... 7-2 7.5 Engineered Safety Feature Actuation Systems ........................................................ 7-3 7.6 Control and Console Display Instruments ................................................................ 7-3 7.7 Radiation Monitoring Systems .................................................................................. 7-3 7.8 Conclusions ............................................................................................................. 7-4 8 ELECTRICAL POWER SYSTEMS .................................................................................. 8-1 8.1 Normal Electrical Power Systems ............................................................................ 8-1 8.2 Emergency Electrical Power Systems ...................................................................... 8-1 8.3 Conclusions ............................................................................................................. 8-2 9 AUXILIARY SYSTEMS .................................................................................................... 9-1 9.1 Heating, Ventilation, and Air Conditioning Systems .................................................. 9-1 9.2 Handling and Storage of Reactor Fuel ..................................................................... 9-1 9.3 Fire Protection Systems and Programs .................................................................... 9-3 9.4 Communication Systems.......................................................................................... 9-3 9.5 Possession and Use of Byproduct, Source, and Special Nuclear Material ............... 9-3 9.6 Cover Gas Control in Closed Primary Systems ........................................................ 9-3 9.7 Other Auxiliary Systems ........................................................................................... 9-4 9.7.1 Carbon Dioxide System .................................................................................... 9-4 9.7.2 Instrument Air System ...................................................................................... 9-4 9.7.3 Plant Chilled Water System .............................................................................. 9-4 9.7.4 Fuel Storage Pool Cooling System ................................................................... 9-4 9.7.5 Thermal Shield Cooling System........................................................................ 9-5 9.7.6 Thermal Column Tank Cooling System ............................................................ 9-5 9.7.7 Experimental Demineralized Water Cooling System ......................................... 9-5 9.8 Conclusions ............................................................................................................. 9-5 10 EXPERIMENTAL FACILITIES AND PROGRAMS ..................................................... 10-1 10.1 Summary Description ............................................................................................. 10-1 10.2 Experimental Facilities ........................................................................................... 10-1 10.2.1 Radial Beam Tubes ........................................................................................ 10-1 10.2.2 Through Tubes ............................................................................................... 10-1 10.2.3 Cold Neutron Source ...................................................................................... 10-1 10.2.4 Thermal Column ............................................................................................. 10-2 iii

CONTENTS (cont.)

Page 10.2.5 Pneumatic Tube System ................................................................................ 10-2 10.2.6 Vertical Thimbles ............................................................................................ 10-2 10.3 Experiment Review ................................................................................................ 10-2 10.4 Conclusions ........................................................................................................... 10-3 11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT .................... 11-1 11.1 Radiation Protection ............................................................................................... 11-1 11.1.1 Radiation Sources .......................................................................................... 11-1 11.1.2 Radiation Protection Program......................................................................... 11-2 11.1.3 ALARA Program ............................................................................................. 11-3 11.1.4 Radiation Monitoring and Surveying ............................................................... 11-3 11.1.5 Radiation Exposure Control and Dosimetry .................................................... 11-4 11.1.6 Contamination Control .................................................................................... 11-5 11.1.7 Environmental Monitoring ............................................................................... 11-5 11.2 Radioactive Waste Management............................................................................ 11-6 11.2.1 Radioactive Waste Management Program ..................................................... 11-6 11.2.2 Radioactive Waste Controls ........................................................................... 11-6 11.2.3 Release of Radioactive Waste ........................................................................ 11-7 11.3 Conclusions ........................................................................................................... 11-7 12 CONDUCT OF OPERATIONS ................................................................................... 12-1 12.1 Organization........................................................................................................... 12-1 12.2 Review and Audit Activities .................................................................................... 12-1 12.3 Procedures............................................................................................................. 12-2 12.4 Required Actions .................................................................................................... 12-3 12.5 Reports .................................................................................................................. 12-4 12.6 Records ................................................................................................................. 12-4 12.7 Emergency Planning .............................................................................................. 12-4 12.8 Security Planning ................................................................................................... 12-5 12.9 Quality Assurance .................................................................................................. 12-5 12.10 Operator Training and Requalification Program ..................................................... 12-5 12.11 Conclusions ........................................................................................................... 12-5 13 ACCIDENT ANALYSES ............................................................................................. 13-1 13.1 Maximum Hypothetical Accident ............................................................................ 13-1 13.2 Insertion of Excess Reactivity Accidents ................................................................ 13-3 13.2.1 Startup Accident ............................................................................................. 13-4 13.2.2 Rapid Removal of Experiments ...................................................................... 13-4 13.3 Loss of Primary Coolant ......................................................................................... 13-4 13.4 Loss of Primary Coolant Flow................................................................................. 13-5 13.4.1 Loss of Offsite Power ..................................................................................... 13-6 13.4.2 Seizure of One Primary Coolant Pump ........................................................... 13-6 13.4.3 Throttling of Coolant Flow to the Inner or Outer Plenums ............................... 13-6 13.4.4 Loss of Both Shutdown Coolant Pumps.......................................................... 13-7 13.5 Misloading of Fuel .................................................................................................. 13-7 13.6 Conclusions ........................................................................................................... 13-7 14 TECHNICAL SPECIFICATIONS ................................................................................ 14-1 15 FINANCIAL QUALIFICATIONS.................................................................................. 15-1 15.1 Financial Ability to Operate the Facility................................................................... 15-1 15.2 Financial Ability to Decommission the Facility ........................................................ 15-1 15.3 Foreign Ownership, Control, or Domination............................................................ 15-2 iv

CONTENTS (cont.)

Page 15.4 Nuclear Indemnity .................................................................................................. 15-2 15.5 Conclusions ........................................................................................................... 15-3 16 OTHER LICENSE CONSIDERATIONS ..................................................................... 16-1 16.1 Prior Use of Reactor Components .......................................................................... 16-1 16.1.1 Reactor Vessel and Components ................................................................... 16-1 16.1.2 Fuel Element and Control Rods ...................................................................... 16-2 16.2 Conclusions ........................................................................................................... 16-3 17 CONCLUSIONS......................................................................................................... 17-1 18 REFERENCES .......................................................................................................... 18-1 v

LIST OF ABBREVIATIONS Abbreviation Definition Page AC Alternating Current ........................................................................................ 8-1 ADAMS Agencywide Documents Access and Management System ........................... 1-1 AEA Atomic Energy Act of 1954, as amended ....................................................... 1-3 AEC Atomic Energy Commission ........................................................................... 1-4 ALARA As Low As Reasonably Achievable................................................................ 1-2 ALI Annual Limit on Intake ................................................................................... 1-2 ANSI/ANS American National Standards Institute/American Nuclear Society ................. 1-1 ARM Area Radiation Monitor .................................................................................. 7-4 BOC Beginning of Cycle ......................................................................................... 4-9 CDE Committed Dose Equivalent ........................................................................ 13-3 CFR Code of Federal Regulations ......................................................................... 1-1 CHF Critical Heat Flux ......................................................................................... 4-12 CHFR Critical Heat Flux Ratio ................................................................................ 13-4 CP Construction Permit ....................................................................................... 1-4 CP-5 Chicago Pile-Five .......................................................................................... 1-3 DAC Derived Air Concentration .............................................................................. 1-2 DC Direct Current ................................................................................................ 8-1 DNB Departure from Nucleate Boiling .................................................................. 4-11 DOE Department of Energy.................................................................................... 1-4 ECC Extended Continental Crust ........................................................................... 2-4 ECS Emergency Cooling System........................................................................... 6-2 EOC End of Cycle .................................................................................................. 4-3 EP Emergency Plan ............................................................................................ 1-1 EPZ Emergency Planning Zone ............................................................................ 2-1 ESF Engineered Safety Feature ............................................................................ 6-1 FEMA Federal Emergency Management Agency ..................................................... 2-3 HEPA High-Efficiency Particulate ............................................................................. 6-2 HVAC Heating, Ventilation, and Air-Conditioning...................................................... 9-1 I&C Instrumentation and Controls ......................................................................... 7-1 I-270 Interstate 270 ................................................................................................ 2-1 IRM Iaptean Rifted Margin .................................................................................... 2-4 LCO Limiting Condition for Operation .................................................................... 4-4 LSSS Limiting Safety System Setting .................................................................... 4-12 MCHFR Minimum Critical Heat Flux Ratio ................................................................. 13-4 MHA Maximum Hypothetical Accident .................................................................... 1-2 MTR Materials Test Reactor................................................................................... 1-3 MWD Megawatt-Days.............................................................................................. 4-3 vi

LIST OF ABBREVIATIONS Abbreviation Definition Page NBSR National Bureau of Standards test reactor ..................................................... 1-1 NCNR NIST Center for Neutron Research ................................................................ 1-1 NIST National Institute of Standards and Technology ............................................. 1-1 NOAA National Oceanic and Atmospheric Administration ........................................ 2-2 NRC U.S. Nuclear Regulatory Commission............................................................ 1-1 NVLAP National Voluntary Laboratory Accreditation Program.................................. 11-4 OFI Onset of Flow Instability............................................................................... 4-11 PSP Physical Security Plan ................................................................................... 1-1 RCS Reactor Control System................................................................................. 7-1 RMS Radiation Monitoring System ......................................................................... 7-1 RPS Reactor Protection System ............................................................................ 7-1 SAC Safety Audit Committee ............................................................................... 12-2 SAR Safety Analysis Report .................................................................................. 1-1 SEC Safety Evaluation Committee....................................................................... 10-2 SER Safety Evaluation Report ............................................................................... 1-1 SRO Senior Reactor Operator................................................................................ 9-2 SSC Structures, Systems, and Components .......................................................... 1-2 SU Startup Core Configuration ............................................................................ 4-1 TEDE Total Effective Dose Equivalent ................................................................... 13-3 TS Technical Specification .................................................................................. 1-1 UPS Uninterruptible Power Supply ........................................................................ 8-1 USGS U.S. Geological Survey.................................................................................. 2-4 vii

1 INTRODUCTION 1.1 Overview By letter (and supporting documentation) dated April 9, 2004, as supplemented on October 2, 2006; May 30 and August 14, 2007; and September 16 and October 21, 2008; the National Institute of Standards and Technology (NIST or the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC or the Commission) a timely application for a 20-year renewal of the Class 104c Facility Operating License No. TR-5 (NRC Docket No. 50-184). The renewed license would authorize continued operation of the National Bureau of Standards test reactor (NBSR) at the NIST Center for Neutron Research (NCNR) located on the NIST campus in Gaithersburg, Maryland. In accordance with Title 10, Section 2.109, Effect of Timely Renewal Application, of the Code of Federal Regulations (10 CFR 2.109), the current license will not be deemed to have expired until the Commission takes final action on the licensees application.

The staff conducted its review based on information contained in the renewal application, as supplemented. The renewal application includes the safety analysis report (SAR), proposed technical specifications (TSs), an environmental report, the operator requalification plan, the emergency plan (EP), the physical security plan (PSP), financial qualifications, and responses to staff requests for additional information. The staff also based its review on annual reports of facility operation submitted by the licensee and inspection reports prepared by NRC staff. The review staff conducted site visits to observe facility conditions.

The licensees application and other materials reviewed by the staff may be examined, and/or copied for a fee, at the NRCs Public Document Room, located at One White Flint North, 11555 Rockville Pike (first floor), Rockville, Maryland. The NRC maintains an Agencywide Documents Access and Management System (ADAMS), which provides text and image files of the NRCs public documents. Documents related to this license renewal dated on or after November 24, 1999, may be accessed through the NRCs Public Electronic Reading Room on the Internet at http://www.nrc.gov. If you do not have access to ADAMS or have problems accessing the documents located in ADAMS, or if you want access to documents dated before November 24, 1999, contact the NRC Public Document Room Reference staff at 1-800-397-4209 or 301-415-4737, or by email to pdr@nrc.gov.

This safety evaluation report (SER) summarizes the findings of the staffs safety review of the licensees application. This SER and the environmental impact statement (NUREG-1873, dated October 2007, ADAMS Accession No. ML072970861) will serve as the basis for issuance of a renewed license authorizing operation of the NBSR at power levels up to 20 megawatts thermal (MW(t)). In conducting its safety review, the staff evaluated the facility against the requirements of 10 CFR Parts 19, 20, 30, 50, 51, 55, 70, 73; and 100; applicable NRC regulatory guides; and relevant accepted industry standards, such as the American National Standards Institute/American Nuclear Society (ANSI/ANS) 15 series. The staff also referred to the guidance contained in NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, issued 1996.

Mr. William B. Kennedy from the NRCs Office of Nuclear Reactor Regulation, Division of Policy and Rulemaking, Research and Test Reactors Branch A prepared this SER. Other contributors to the safety review include Mr. Marvin M. Mendonca, Mr. William C. Schuster, Mr. Michael B.

Norris, Mr. Paul V. Doyle, Mr. Ronald B. Uleck, and Ms. JoAnn Simpson of the NRC staff. Mr.

William Watkins, Mr. James Willison, and Mr. James Wallace of Washington Safety 1-1

Management Solutions, LLC, provided a technical evaluation of the licensees SAR and TSs under contract to the NRC.

1.2 Summary and Conclusions on Principal Safety Considerations On the basis of its safety evaluation, the staff reached the following findings:

• The design, testing, and performance of the NBSR structures, systems, and components (SSC) important to safety during normal operation are acceptable. Safe operation of the facility can reasonably be expected to continue.

• The licensees management organization is acceptable to maintain and safely operate the reactor. The licensees management organization, training and research activities, and security measures continue to be acceptable to ensure safe operation of the facility and the protection of its special nuclear material.

• The licensee and NRC staff have conservatively considered the expected consequences of postulated accidents, including a bounding maximum hypothetical accident (MHA), using conservative initiating and mitigating assumptions. The calculated radiation doses resulting from the MHA are less than the doses specified in 10 CFR Part 20, Standards for Protection Against Radiation, for facility personnel and members of the general public and satisfy the regulatory dose requirements of 10 CFR Part 100, Reactor Site Criteria.

• Exposures from and releases of radioactive effluents and wastes from the facility are not expected to result in doses or concentrations in excess of the limits specified by Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage, to 10 CFR Part 20 and are consistent with as-low-as-reasonably-achievable (ALARA) principles.

• The renewed Facility Operating License and TSs, which state limits controlling operation of the facility, provide reasonable assurance that the licensee will operate the facility in accordance with the assumptions and analyses in the SAR. No significant degradation of SSCs has occurred, and the TSs will continue to provide reasonable assurance that no significant degradation of SSCs will occur.

• The financial data submitted with the application demonstrate that the licensee has acceptable access to sufficient funds to cover operating costs and to eventually decommission the reactor facility.

• The licensees procedures for training its reactor operators and the operator requalification plan give reasonable assurance that the licensee will continue to have qualified personnel who can safely operate the reactor.

• The licensees EP provides acceptable assurance that the licensee will continue to be prepared to assess and respond to emergency events.

• Continued operation of the NBSR poses no significant radiological risk to the health and safety of the public, facility personnel, or the environment.

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On the basis of these findings, the NRC staff concludes that NIST can continue to operate the NBSR in accordance with the Atomic Energy Act of 1954, as amended (AEA or the Act), NRC regulations, and Facility Operating License TR-5 without endangering the health and safety of the public.

1.3 General Facility Description The NBSR is a heavy-water-(D2O) moderated-and-cooled, enriched-fuel, tank-type reactor designed to operate at 20 MW(t) power. It is a custom-designed variation of the Argonne Chicago Pile-Five (CP-5) class reactor. It differs from the CP-5 in its power rating, core configuration, and cold neutron source, but retains the proven technology. The three most notable modifications to this basic design are: a 18-centimeter (cm) (7-inch (7-in)) gap between the upper and lower fuel regions in each fuel element to reduce the fast neutron background in the neutron beams; a double plenum at the bottom of the vessel to provide optimized cooling to the core; and the method for remote handling of fuel elements during refueling. This type of reactor was chosen because of its well-thermalized neutron spectrum, high neutron flux, flexibility for research, and inherent safety.

The NCNR has a wide range of research capabilities. The liquid hydrogen cold source provides cold neutrons (neutrons slowed to speeds of 1,000 meters per second (m/s) (2,200 miles per hour (mi/h)) or less) directly to experiments in the Confinement Building, and through a network of seven neutron guides to experiments located in the Cold Neutron Guide Hall. Beam tubes provide thermal neutrons for experiments located within the area immediately adjacent to the reactor. A pneumatic sample transfer system provides researchers with the ability to automatically inject samples into the core region of the reactor, while vertical thimbles provide locations for manual loading. The reactor utilizes U3O8 aluminum dispersion fuel clad in aluminum. Heavy water provides neutron moderation and core cooling. The closed primary coolant system circulates heavy water in an aluminum and stainless steel system. The heavy water pumped through heat exchangers transfers its heat to light water (H2O) before re-entering the core and returning to the pumps. The light water secondary cooling system transfers its heat to the atmosphere by means of evaporation from a cooling tower located outside of the Confinement Building.

1.4 Shared Facilities and Equipment The Confinement Building, constructed of reinforced concrete and situated partially below grade, adjoins a laboratory complex dedicated primarily to nuclear science-related research and other reactor support functions. The Confinement Building has an independent ventilation control system, capable of either an isolation mode of operation, or a dilution mode of operation when air is exhausted to the atmosphere through an elevated stack located adjacent to the building. Local utilities provide municipal water and sewerage, natural gas, and electricity to the NCNR.

1.5 Comparison with Similar Facilities The NBSR design is a variation of the Argonne CP-5 class reactor. The NBSR utilizes materials testing reactor (MTR) -type plate fuel, which is also in use at the Massachusetts Institute of Technology reactor, a 5 MW(t) tank-type research reactor. The Brookhaven National Laboratory and the Savannah River Site reactors also used similar fuel. The NBSR does not have any unique features that would preclude applying general knowledge and experience gained in the operation of these other reactors to operation of the NBSR.

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1.6 Summary of Operation The licensee has operated the NBSR in accordance with Facility Operating License No. TR-5 and established procedures to provide research services to government and commercial scientists. Research activities primarily take advantage of the cold neutron source which provides a unique resource for materials science and neutron property investigations. The reactor typically operates at 20 MW(t) for 24-hours-a-day, 7-days-a-week, with a routine shutdown every five-and-one-half weeks. During the routine shutdown, which lasts approximately 10 days, the licensee refuels the reactor and performs maintenance. According to annual reports submitted by the licensee, the NBSR operated for a total of 254 full-power days in fiscal year (FY) 2006 and 268 full-power days in FY 2007. These values represent expected annual facility operation during the period of the renewed license.

1.7 Compliance with the Nuclear Waste Policy Act of 1982 Section 302(b)(1)(B) of the Nuclear Waste Policy Act of 1982 provides that the NRC may require, as a precondition to issuing or renewing an operating license for a research or test reactor, that the applicant shall have entered into an agreement with the U.S. Department of Energy (DOE) for the disposal of high-level radioactive waste and spent nuclear fuel. DOE (represented by R.L. Morgan) has informed the NRC (represented by H. Denton) by letter dated May 3, 1983, that it has determined that universities and other government agencies operating non-power reactors have entered into contracts with DOE that provide that DOE retains title to the fuel and is obligated to take the spent fuel and/or high-level waste for storage or reprocessing. The NBSR is a federally-owned and operated non-power reactor that is part of the NIST. The NBSR and DOE have a contractual arrangement whereby DOE retains title to the fuel and is obligate to take the spent fuel and/or high-level waste for storage and reprocessing. All of the spent NBSR fuel has been returned to DOE pursuant to this arrangement. Accordingly, NIST is in conformance with the Nuclear Waste Policy Act of 1982.

1.8 Facility Modifications and History The Atomic Energy Commission (AEC) received the application for the NBSR construction permit (CP) and operating license on February 1, 1961. Construction began in 1963 when the AEC issued the CP. The CP was converted to Facility Operating License TR-5. The reactor achieved initial criticality on December 7, 1967, and began full-power operation at 10 MW(t) on February 9, 1969. On December 2, 1980, the NBSR requested an increase in the maximum licensed power from 10 MW(t) to 20 MW(t) and a 20-year renewal of the license. The NRC issued the license renewal, including authorization to operate at the increased power level, on May 16, 1984. Since the license renewal, the licensee has made no significant changes to the facility or facility operations.

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2 SITE CHARACTERISTICS 2.1 Geography and Demography 2.1.1 Geography The NBSR is located at latitude 39º 7 34 north and longitude 77º 13 6 west. The corresponding Universal Transverse Mercator coordinates are Zone Number 18, Northing 4333105 m, and Easting 308252 m. The NCNR facility is located on the southern portion of the 575-acre NIST campus in Gaithersburg, Montgomery County, Maryland, approximately 32 kilometers (km) (20 miles (mi)) northwest of Washington, DC. There are no prominent natural features in the immediate vicinity of the reactor, and the most prominent man-made feature is Interstate 270 (I-270) adjacent to the eastern boundary of the NIST campus. NIST is a non-regulatory federal agency of the U.S. Commerce Department within the Technology Administration.

The NIST Campus is located on the Maryland Piedmont Plateau approximately 48 km (30 mi) southeast of the Blue Ridge Mountains. The Blue Ridge Range rises to about 1,900 feet above mean sea level compared to the NIST elevation of 128 meters (m) (420 feet (ft)) above mean sea level. The general area within an 8-km (5-mi) radius about the NBSR has a gently rolling topography without any geographic features that could significantly affect the diffusion and dispersion of airborne effluents. There are few buildings over three stories high within the immediate area surrounding the NBSR.

2.1.2 Demography Montgomery County is the most populous county in the State of Maryland. Between 1990 and 2000, the county saw a population growth of 15.4% with much of the growth occurring in the southern half of the county. The population of the county is expected to increase by an additional 22% by 2025.

The NIST campus is surrounded by the city of Gaithersburg, Maryland. The cities of Gaithersburg, Washington Grove, Rockville, and the unincorporated areas of Germantown, Montgomery Village, Darnestown, and North Potomac all lie within 8-km (5-mi) of the NBSR.

According to the 2000 Census, more than 223,000 people live within this 8-km (5-mi) circle.

The facility Emergency Planning Zone (EPZ), a 400 m (1300 ft) circle centered on the NBSR ventilation stack, lies entirely within the NIST campus. NIST Security controls access to the campus. Access is limited to employees, contractors, and individuals who have business on the site. The area within the EPZ has laboratories and office buildings, but no residential buildings.

There is no part-time, transient, or seasonal occupation of any of the campus buildings. Thus the effect of County population growth on potential public exposure to radiation from accidents at the NBSR is not expected to be significant. The closest permanent residences are more than 400 m (1300 ft) directly to the east and directly to the west of the NBSR.

2.2 Nearby Industrial, Transportation, and Military Facilities The I-270 Technology Corridor is a major research and development center. Some manufacturing occurs in the area but there are no significant manufacturing plants near the NBSR. There are no chemical plants or refineries in the immediate area surrounding the NBSR.

Mining and quarrying operations in the vicinity are limited to those associated with constructing 2-1

new office buildings. A natural gas pipeline lies 3 km (2 mi) to the south of the reactor, and a liquid petroleum/gas pipeline is located 1.6 km (1 mi) to the north.

I-270 forms the northeast boundary of the NIST campus and is a major commuter and truck route serving the area. Three arterial and collector roads abut the NIST campus that serve the Gaithersburg area surrounding the NIST campus, providing truck routes serving the local economy. A rail line parallels the northeast boundary approximately 2 km (1 25 mi) from the NIST campus carrying goods through the region. The nearest waterway is the Potomac River.

Its nearest point is 10 km (6.4 mi) from the reactor.

The closest commercial airport to the reactor is Dulles International, 29 km (18 mi) from the reactor. Located 7.2 km (4 5 mi) from the reactor is the Montgomery County Airpark, a general aviation airport. Based on a review of airport maps, the runways and approach traffic for either airport are not in line with the NBSR facility.

Andrews Air Force Base, the nearest military base, is approximately 52 km (32 mi) away. Due to the distance between the two, operations at Andrews Air Force Base facility are not expected to impact safe operation of the NBSR.

Based on the historical lack of serious transportation accidents in the vicinity and the distances involved to the NBSR, the staff concludes that neither local industry, transportation, nor government facilities pose a significant risk to the continued safe operation of the NBSR.

2.3 Meteorology The National Oceanic and Atmospheric Administration (NOAA) maintains historical meteorological data sufficient to characterize the NBSR site. Wind data is available from the National Weather Service stations at Dulles International Airport (29 km (18 mi) SW) and at Reagan National Airport (40 km (25 mi) SSE). NIST has meteorological instrumentation located onsite. Although this instrumentation has not been in place long enough to provide useful site-characterization data when compared to the National Weather Service locations, it could be used for atmospheric dispersion calculations in the event of a release of radioactive material.

According to the NBSR SAR, monthly average temperatures range from a low of -4.6 degrees Celsius (°C) (23.8 degrees Fahrenheit (°F)) in January to a high of 29 °C (85 °F) in July. The annual mean humidity is 80% in the morning and 55% in the afternoon. Precipitation is evenly distributed throughout the year with an annual average of 81 cm (32 in). The maximum recorded precipitation in a 24-hour period was 20 cm (7.9 in) in June 1972. Snowfall occurs typically from November to March and averages 48 cm (19 in) annually. Snowfalls of several inches are typical and remain on the ground for several days before melting. Over the past 11 years, Montgomery County has experienced an annual frequency of 4.3 winter weather events, of which 2.3 per year were mixed rain and snow or ice and snow. The maximum-recorded 2-day snowfall total was 65.3 cm (25.7 in) in 1996. The 100-year return period snowpack including 2-day probable maximum precipitation was calculated for Montgomery County as 27.2 pounds per square foot (psf). The staff independently reviewed the licensees snowpack calculations and found them to be reasonable.

The average wind speeds at Dulles International Airport and Reagan National Airport are 3.3 m/s (7.4 mph) and 4.2 m/s (9.4 mph), respectively. The most common wind direction recorded is generally from the NW, with more frequent winds from the South during the summer months. Since 1950, only Hurricane Fran (1996) has affected the NIST site, though it has been 2-2

periodically affected by remnants of tropical storms. Tornados occur relatively infrequently in Maryland, with an annual frequency in Montgomery County of less than 0.22 events per year.

The tornadoes in Montgomery County were rated as F0 (40-72 mph) and F1 (73-112 mph) and represent the lowest categories on the Fujita Tornado Scale. The licensee calculated the 100-year return wind speed as between 35.7 m/s (79.8 mph) and 45.8 m/s (102.5 mph) for Montgomery County.

The frequency of hail events in Montgomery County has been 2.1 events per year. The frequency of lightning events in Montgomery County causing property damage is 1.8 events per year. Neither hail nor lighting has had any significant impact on the operation of NBSR.

Based on the metrological information supplied by the licensee and the staffs independent review of referenced data sources, the staff concludes that meteorology in the vicinity of the NBSR does not pose any significant risk to continued safe operation of the reactor.

2.4 Hydrology The topography in the vicinity of the NBSR is undulating and the relief is moderate. The confinement building and cooling towers are at an elevation of approximately 128 m (420 ft).

The elevation of the surface-water body at the nearest point to the confinement building is approximately 116 m (380 ft). The nearest naturally occurring surface water body to the site is a tributary of Muddy Branch, approximately 30 m (1,000 ft) to the west-northwest of the NBSR.

Surface water drainage at the site flows southwest to the Muddy Branch stream and its tributaries. Muddy Branch discharges to the Potomac River. The site is more than 16 river-kilometers (10 river-miles) from the Potomac River.

The nearest mapped flood zone to the reactor, the Muddy Branch floodplain, is located approximately 610 m (2,000 ft) south of the site. The highest 100-year flood elevation for this floodplain is 115 m (376 ft) at the confluence of the Muddy Branch tributary, approximately 580 m (1,900 ft) southeast of the site. A topographical rise separates the site from the 100-year flood zone for the Muddy Branch tributary.

The Federal Emergency Management Agencys (FEMAs) floodplain mapping and the topography of the site show that the site lies outside the 100-year and 500-year flood zones of the nearest surface-water bodies. There is no documented history of flooding occurring at the site either before or after the NBSR was constructed. The 100-year flood event for Muddy Branch and its tributaries is considered the controlling event for determining appropriate measures for flood protection. All the existing safety-related structures for the site are protected against it.

As the confinement building and support structures are outside the 100-year and 500-year flood zones of these water bodies, there are no additional flood-design considerations. There are sufficient surface-water drainage systems at the site to convey away the precipitation from local, intense events.

There are no probable maximum surge and seiche flooding considerations for this site as no large bodies of water are near the site where significant storm surges and seiches can form. As the site is not adjacent to a coastal area, tsunami flooding is not considered credible. There are no existing or proposed dams on Muddy Branch Creek or its tributaries upstream of the site, thus there are no seismically-induced potential dam failure considerations for the site.

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There is minimal potential for discharges of contaminated groundwater to surface waters from accidental releases at the site. The licensee conducts routine environmental sampling of grass, soil, and water in nearby streams and ponds. The samples are analyzed for possible neutron activation nuclides and fission product nuclides. Water samples are also assayed for tritium.

The results are reported to the NRC in the NBSR Annual Report. The staff reviewed the licensees annual reports from 1999 to 2007 and found that the licensee detected no contamination.

Based on the above information, the staff concludes that the local hydrology does not pose a significant risk to the continued safe operation of the NBSR, nor does it provide a credible pathway for contamination of the local water supply.

2.5 Geology, Seismology, and Geotechnical Engineering The NBSR is located in the southwestern portion of the city of Gaithersburg, Maryland. The reactor site lies within the Piedmont physiographic province. The eastern Piedmont is characterized by gently sloping upland areas, and broad, relatively shallow valleys. The Fall Line, which is the physiographic and tectonic boundary between the Coastal Plain and Piedmont provinces, is approximately 26.5 km (16.5 mi) to the southeast of the NBSR site. The eastern margin of the Blue Ridge Province, the Catoctin Mountains, is approximately 32 km (20 mi) west of the site.

No large-scale geologic structures have been mapped near the NBSR site. The deformations that have been mapped occurred more than 400 million years ago. The sites safety is not affected by these deformations. The only seismic source zones of concern are the Extended Continental Crust (ECC) and the Iapetan Rifted Margin (IRM). The NIST site lies in the western portion of Zone ECC, approximately 32 km (20 mi) from the eastern margin of Zone IRM.

The U. S. Geological Survey (USGS) updated its seismic hazard maps for the United States based on new seismological, geophysical, and geological information. The USGS employed a probabilistic methodology that uses a combination of gridded, spatially-smoothed seismicity, large background zones, and specific fault sources to calculate hazard curves for a grid of sites throughout the country. The USGS probabilistic analysis results show relatively low ground-motion risk for a broad area surrounding the NBSR site.

The NBSR site is located in an area that has experienced only minor earthquake activity. No earthquakes with magnitudes greater than those considered for earlier licensing actions have occurred within the ECC and the IRM since publication of NUREG-1007, Safety Evaluation Report Related to the License renewal and Power Increase for the National Bureau of Standards Reactor, in 1983. The licensees seismic analysis concluded that the maximum potential earthquake for the area would generate a maximum peak horizontal ground-acceleration at the site of 0.07 to 0.1 times the acceleration due to gravity. A 2008 National Seismic Hazard Map produced by the USGS shows only a 2% probability that in 50 years peak lateral ground acceleration will exceed 0.06 times the acceleration due to gravity.

No surface faulting has been documented for any earthquakes occurring in the ECC or the IRM.

The only faults mapped within 8 km (5 mi) of the site experienced deformation in the Paleozoic era. The potential for surface faulting at the site is negligible. The characteristics of the underlying soils and low level of seismicity, indicate that the potential for liquefaction is practically nonexistent.

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The NBSR buildings and facilities are supported on shallow and deep foundations. The shallow foundations are believed to rest on competent residual soils and/or transition materials.

Similarly, deep foundations are believed to be bearing into transition materials and/or bedrock.

This conclusion is supported by the successful performance of the buildings and facilities foundations since their original construction, and the lack of any reported signs of distress or movement in the foundations. The foundations are expected to continue to perform satisfactorily as long as the geologic, hydrogeologic, and superimposed loads continue to remain consistent with those adopted in designing the facilities.

Based on the above information, the staff concludes that the geology of the NBSR site is suitable for supporting the reactor building, structure, and systems, and that potentially damaging seismic events are unlikely to occur during the period of the renewed license.

2.6 Conclusions The staff concludes that the reactor site has experienced no significant geographical, meteorological, or geological change, and therefore the site remains suitable for continued operation of the NBSR. Infrequency of the occurrence of tornadoes and earthquakes continue to make the site suitable for operation of the reactor. Hazards related to industrial, transportation, and military facilities will not pose a significant risk to the continued safe operation of the NBSR. The demographics of the area surrounding the reactor have not changed and are not expected to change in any way that discernibly increases the risk to public health and safety from continued operation of the NBSR.

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3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Design Criteria The design criteria for SSCs are that the SSCs related to safe operation and shutdown of the reactor must be able to perform their intended functions as described in the NBSR SAR. The principal safety-related SSCs are the fuel, core support structure, reactor protection system, reactor coolant system, and the confinement building. The staff evaluated the following specific design criteria for the above-mentioned SSCs during normal operation and credible accident scenarios:

• The fuel must prevent the release of fission products.

• The core support structure must maintain its orientation, geometry, and structural integrity.

• The reactor safety system must be able to shut down the reactor.

• The reactor coolant system must be able to remove heat from the reactor core and keep fuel elements below temperatures which could result in cladding damage.

• The confinement building must isolate under upset conditions and prevent the uncontrolled release of radioactive materials to the environment.

The SSCs mentioned above were designed and constructed in accordance with the construction permit issued by the AEC in 1963. Design of the fuel, control elements, and the core support structure are discussed in Section 4.2 of this SER. Design of the reactor protection system is discussed in Section 7.2 of this SER. Design of the reactor coolant system is discussed in Section 5.2 of this SER. Design of the confinement building is discussed in Section 6.3 of this SER. TS 4.1, requires routine verification of the confinement function of the confinement building and functionality of the instrumentation that provides the confinement closure signal.

These SSCs have since been maintained and/or changed using license amendments or licensee review processes, including 10 CFR 50.59, Changes, Tests, and Experiments, maintenance, and special procedures, as appropriate, in accordance with the Commissions rules and regulations and Facility Operating License No. TR-5, as amended. The NRC staff previously evaluated all amendments to the facility license, and the NRC inspection program verified that the licensee conducted the proper reviews. In its safety evaluation supporting issuance of Amendment No. 5 dated May 16, 1984 (NUREG-1007), the NRC staff evaluated the licensees request for license renewal and an increase in the maximum reactor power level from 10 MW(t) to 20 MW(t). The NRC staff concluded that the SSC in place were adequate to support safe operation during the period of the license renewal at the increased power level.

Experience accumulated over the past 24 years of reactor operation supports the NRC staffs conclusion. Chapter 16 of this SER discusses age-related issues. Based on the above, the staff concludes that the design and construction of safety-related SSCs provide reasonable assurance that SSCs will continue to meet the design criteria.

3.2 Meteorological Damage Section 2.3 of this SER presents the meteorology of the NBSR site. While severe storms or tornados are possible at the NBSR site, the reactor and associated safety systems are housed 3-1

in a reinforced concrete structure which provides considerable protection. The reactor and associated shielding are supported by a central concrete column. Shielding walls are nominally 3- to 5-feet thick and are an integral part of the building foundation. The confinement structure is designed for a 100-mph wind load which is within the uncertainty for the 100-year return period wind load. The roof of the confinement structure is designed to withstand a snow load of 25 psf. The licensee calculated a 100-year return period ground snowpack of 27.2 psf, corresponding to a 22.8 psf confinement building roof snow load. The staff reviewed the licensees calculation and found it to be reasonable. The staff concludes that the design of the confinement building is sufficient to withstand likely meteorological conditions during the period of the renewed license. As discussed in Section 8.2 of this SER, there is reasonable assurance that failure of offsite power due to meteorological phenomena will not cause damage to the reactor or preclude safe shutdown.

3.3 Water Damage Section 2.4 of this SER presents the hydrology of the NBSR vicinity. There are no bodies of water in the immediate vicinity that could flood the NBSR site. Further, the licensee indicated that surface drainage measures and the watertight confinement building would prevent heavy rainfall from impacting the reliable operation of safety-related equipment. Therefore, the staff concludes that water damage poses no significant risk to safe operation or shutdown of the reactor.

3.4 Seismic Damage Section 2.5 of this SER presents the seismicity of the NBSR vicinity. The NBSR is located in a zone of low seismic activity. The building and reactor systems have been analyzed and shown to be able to withstand the stresses generated by peak lateral ground acceleration of 0.1 times the acceleration due to gravity. The probability of an earthquake resulting in peak lateral ground acceleration larger than 0.06 times the acceleration due to gravity is less than 2% in 50 years.

The NBSR reactor core and core components, including control rods, are designed to fail-safe.

On receipt of a scram signal or a loss of power, all four shim safety arm clutches are de-energized and the corresponding shim safety arms are inserted into the core. The shutdown-margin criterion for safe shutdown (TS Definition 1.3.30, Shutdown Margin, and TS 3.1.2, Reactivity Limitations) ensures that the reactor can be shut down even if the most reactive of the four shim safety arms fails to insert. The reactor vessel was designed in accordance with Building Officials and Codes Administrators Codes for seismic. The combined stress levels resulting from seismic loading plus all other design loads are well within the allowable limits for the various vessel sections. The NBSR accident analyses, discussed in Chapter 13 of this SER, include a seismic event. As shown by the licensees analysis, the postulated seismic event will not pose significant risk to the health and safety of the public.

3.5 Systems and Components The systems and components most important to safety are the reactor protection system and the fuel cladding. The reactor protection system consists of the control devices and safety-related instrumentation and controls. Section 4.2.2 of this SER discusses the design requirements of the control devices. Chapter 7 of this SER discusses the design requirements of the safety-related instrumentation and controls. Section 16.1.2.2 of this SER considers aging issues associated with control devices. These discussions show that the reactor safety system design bases and related TSs provide reasonable assurance that the reactor protection system will function as designed to ensure safe operation and safe shutdown of the reactor.

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Section 4.2.1 of this SER discusses the fuel cladding design requirements, Chapter 13 examines accident scenarios, and Section 16.1.2.1 addresses aging issues associated with the fuel. These discussions show that the fuel cladding design basis and related TSs are adequate to ensure fuel cladding integrity under all credible circumstances.

3.6 Conclusions On the basis of the above considerations, the staff concludes that the design and construction of the NBSR is adequate to withstand and/or ensure safe shutdown as a result of all credible and likely wind, water, and seismic events associated with the site. The design and performance of safety-related systems and components has been verified through safe operation during the period of the current license and routine NRC inspections. The staff also concludes that surveillance activities required by the TSs discussed in the above-referenced sections of this SER provide reasonable assurance that the safety-related functions of the facility SSCs will be operable. Accordingly, the staff concludes that the reactor systems and components are adequate to provide reasonable assurance that continued operation will not cause significant radiological risk to the health and safety of the public, licensee personnel, and the environment.

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4 REACTOR DESCRIPTION 4.1 Summary Description The NBSR is a D2O moderated and cooled, tank-type reactor designed to operate at a power level of 20 MW(t). The core is located in the lower section of an aluminum tank. The core contains thirty-seven fuel element locations and four semi-permanent irradiation thimble tubes.

Seven of the fuel element locations are specially adapted for thimble tubes, leaving only 30 positions available for fuel element assemblies. The NBSR fuel elements are plate-type elements consisting of U3O8 mixed with aluminum powder contained in aluminum clad plates.

Each fuel element contains an upper and lower fuel section separated by a gap, resulting in a split core design. This split-core design, with uranium fuel placed above and below the mid-plane of the reactor core, results in the thermal neutron flux reaching a peak in the center of the gap.

4.2 Reactor Core The reactor core is located in the lower section of an aluminum tank, 2 m (7 ft) in diameter by 5 m (16 ft) in height. The fuel elements are held in place by upper and lower grid plates. The grid plates contain 37 fuel element positions placed in 7 rows to form a hexagonal pattern with the rows oriented east to west in the core. The fuel is arranged in three rings within this hexagonal pattern, with the inner two rings having 6 fuel elements each and the outer ring having the remaining 18 fuel elements. The seven experimental thimble positions form a circular pattern about the center location. The overall length of the fuel element assembly is approximately 175 cm (68.8 in). Each fuel element fits into fixed openings in the grid plates.

Four semaphore-type shim safety arms and one automatic regulating rod provide control of the reactor. The four shim arms provide primary control of the reactor. They are used to attain criticality on start up, make major changes in the power level of the reactor, and compensate for reactivity changes that occur as a result of xenon, temperature, and fuel burn-up. The four shim arms are mounted on hanger brackets just under the upper grid plate. The regulating rod provides fine control of the reactor. The regulating rod is located in a 8.9-cm (3.5-in) vertical thimble.

4.2.1 Reactor Fuel The NBSR reactor utilizes only the MTR-plate-type fuel element consisting of enriched U3O8 mixed with aluminum powder contained in aluminum-clad plates. Extensive testing of fuel plates to determine the limits on fission density as a function of fuel loading has been performed.1 NBSR fuel is moderately loaded at 18% volume fraction with a maximum possible fission density of 2.6x1027 fissions per cubic meter (fissions/m3). With a burn-up of 73% in the 8-cycle fuel elements, the typical average fission density is 1.9x1027 fissions/m3. This is within the 2.0x1027 fissions/m3 limit specified in TS 3.1.4, Fuel Burn-up, and within acceptable levels for the prevention of unacceptable swelling reported in the literature.2 1 Hofman, G.L., J. Rest and J.L. Snelgrove, Irradiation Behavior of Uranium Dioxide - Aluminum Dispersion Fuel, Argonne National Laboratory, October 1996.

2 Snelgrove, J.L. and Hofman, G.L., Dispersion Fuels, Materials Science and Technology: A Comprehensive Treatment, Volume 10A: Nuclear Materials, Part I, R.W. Cahn, ed., New York, 1994.

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High-temperature testing of irradiated MTR fuel plates has been performed in order to determine limits for fission product release. Blistering of the U3O8 plates occurs between 450 C (840 F) and 550 C (1020 F)2. In order to preclude any blistering of the fuel cladding, the NBSR maximum allowable fuel clad temperature is 450 C (840 F), as specified by TS 2.1, Safety Limit.

Fabrication of NBSR fuel elements is in accordance with standard industry techniques for the manufacture of MTR plate-type fuel elements and the NIST specification for aluminum clad fuel elements. According to the licensee, new fuel element assemblies are subjected to stringent quality assurance prior to insertion into the core. The manufacturer inspects the fuel assemblies in accordance with U.S. Department of Energy requirements prior to shipment to the licensee.

The corrosion history of aluminum MTR-type fuel elements has been studied extensively. Fuel plates of the same basic configuration and the same material as those used in the NBSR fuel elements have been operated at higher coolant flow rates, higher temperatures, and much higher heat fluxes than are achieved in the NBSR. All of these factors generally increase the corrosion rate of the fuel, yet corrosion of fuel elements during lifetimes comparable to those in the NBSR has not been a problem from the standpoint of structural integrity. According to the licensee, no NBSR fuel element has exhibited significant signs of corrosion or symptoms of corrosion damage. Based on the licensees fuel utilization program and limited by TS 3.1.4, the lifetime of an NBSR fuel element is typically one year.

The outer shell of the NBSR fuel element represents the only major variation from the classic MTR plate type fuel element. Since this outer shell controls the establishment of the proper hydraulic regime for heat transfer purposes, confirmation of the structural and hydraulic design objectives was accomplished on a hydraulic stand, using a fuel element assembly fitted with dummy plates. For these tests performed by the licensee, flow rates over a wide range of velocities were employed to measure flow conditions in each channel and across typical channels as well as the total pressure drop, drag forces, bypass flow around the lower nozzle, and the vibration characteristics of the spring-loaded element lock. The tests confirmed the predicted performance of the NBSR fuel element design.

The staff reviewed the NBSR safety analysis and determined the NBSR fuel design is adequately supported by the MTR fuel development program. The staff also concludes that the NBSR limits on fuel temperature (TS 2.1) and fuel burn-up (TS 3.1.4) are supported by research and testing performed on similar U3O8 dispersion fuels. The historical evidence on the fuel integrity and corrosion resistance performance of the NBSR fuel, as described by the licensee in the SAR, offers additional support for the adequacy of the NBSR fuel design. Therefore, the staff concludes that continued operation as limited by the technical specifications offers reasonable assurance the fabricated fuel can meet the design objective of maintaining fuel integrity and thereby function safely in the reactor without undue risk to the health and safety of the public or the environment.

4.2.2 Control Rods The three NBSR reactivity control systems consist of four shim safety arms, a single regulating rod, and a moderator dump system. The shim safety arms provide the primary means of reactivity control and shutdown capability. Each shim arm is a 2.54-cm-thick by 12.7-cm-wide by 132-cm-long (1-in by 5-in by 52-in) poison tube with a hollow interior filled with helium. The poison material is a 0.102-cm-thick (0.040-in) cadmium plate clad with 1100 series aluminum.

The total reactivity worth of the four shim safety arms is approximately 26.4% at the end of 4-2

an operating cycle (end of cycle (EOC)). Reactor shutdown capability is maintained from the most reactive state with the most reactive shim safety arm stuck in the fully withdrawn position, as required by TS 3.1.2. This requirement satisfies the stuck rod criterion found in the guidance in NUREG-1537 and ANSI/ANS-15.1.

Redundancy is provided by the four separate shim safety arms as well as the independent moderator dump system required for reactor operation by TS 3.3.3, Moderator Dump System.

The NBSR shim safety arm control system is designed to ensure safe reactor control and shutdown under all operating conditions. This is achieved by using a design that relies on a passive feature (gravity) to achieve the safety function. All four shim safety arms are coupled to their drive motors by electromagnetic clutches. Thus, the only action required to effect a safe and rapid shutdown is to de-energize the electromagnetic clutches. The system is fail-safe in that:

• No power source is required to initiate a shutdown.

• Loss of electrical power automatically results in a shutdown.

• No mechanical action, such as the release of a latch, is required in order to insert a shim arm.

• Insertion of any three of the four shim safety arms will result in a reactor shutdown under the most reactive core conditions.

According to the licensee, the shim safety blades are subjected to fabrication quality control inspections and radiography to ensure all design requirements are met. Mechanical shock absorbers are included in the system design to ensure the shim safety arms can withstand all anticipated stresses during operation. Mechanical stops prevent over-travel of the shim safety arms in the event of a mechanical failure in the blade drive system.

The lifetime of the shim arms is affected by poison burn-up, corrosion, and radiation damage.

According to the licensee, under normal operating conditions, the shim safety arms have a lifetime of approximately 21,000 Megawatt-days (MWD). The use of similar shim arms in the CP-5 reactor for a period of 8 years demonstrated that poison burn-up, and not corrosion damage, limits the useful life of the shim arms. Over thirty years of NBSR operation have shown this to also be true for the NBSR shim arms. TS 4.1.2, Reactivity Limitations, requires annual determination of the reactivity worth of each shim arm. According to the licensee, radiation damage to structural materials of the shim safety arms is not significant during reactor operation because the shim arms are in the top reflector above the core where the fast neutron flux is relatively low. Shim arm sets have been replaced at the NBSR reactor three times, with no apparent radiation damage in the shim arms removed. Replacement is determined by acceptable reactivity margins necessary to maintain the shutdown capability required by TS 3.1.2.

The shim safety arm design provides digital position indication on the reactor control panel by a potentiometer coupled to the shim safety arm drive shaft. Each shim arm drive is driven by an electric motor, through a high ratio gear case and finally through an electromagnetic clutch. The shim arms may be controlled individually or as a bank. The four independent shim arm drives and control systems prevent a common mode malfunction.

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The shim arm drives are constant speed mechanical devices with a drive speed of 0.0445 degrees per second (°/s), consistent with the safety analysis for the insertion of excess reactivity accident and TS 3.2.1, Shim Arms. Scram is aided by a spring that opposes drive motion during arm withdrawal. The shim safety arms are considered operable for a scram if they drop the top 5 degrees of travel within 240 milliseconds, as required by TS 3.2.1. Withdrawal and insertion speeds or scram time should not vary except as a result of mechanical wear. The withdrawal and insertion speeds of each shim arm are determined at least semiannually, and scram times of each shim arm drive are measured at least semiannually, as required by TS 4.2.1, Shim Arms. These surveillance requirements, chosen to provide a significant margin over the expected failure or wear rates of these devices, meet or exceed the guidance for control rods in ANSI/ANS-15.1.

The regulating rod provides for fine control of core reactivity. The regulating rod consists of a solid aluminum cylinder, 6.35 cm (2.5 in) in diameter by 74 cm (29 in) long and is located in a vertical thimble 8.9 cm (5.5 in) in diameter. The regulating rod operates in a shroud which has the same configuration as the experimental thimbles. A fixed orifice in the nozzle of the shroud provides cooling water flow for the regulating rod. The regulating rod is designed with a reactivity worth of approximately 0.58% , and because the regulating rod is aluminum, poison burn-up is insignificant. Corrosion damage is similar to other core components and minimal.

The regulating rod is driven by a standard commercial design vertical drive mechanism mounted in the top plug. The regulating rod automatically drives into the core to the full in position upon receipt of a scram signal.

The moderator dump system provides a redundant shutdown capability. Moderator dump is initiated by manually actuating an air-operated diaphragm valve which allows the moderator level to drop to 2.54 cm (1 in) above the top of the core. Although the reactivity worth of the moderator dump is a function of the position of the shim arms, the moderator dump will make the reactor subcritical for any core configuration allowed by the TSs. TS 3.3.3 requires the moderator dump system to be operable when the reactor is operated and TS 4.3.3, Moderator Dump System, requires annual cycling of the moderator dump valve.

TSs 3.1.2, 3.2.1, 3.3.3, 4.1.2, 4.2.1, and 4.3.3 contain appropriate limiting conditions for operation (LCOs) and surveillance requirements necessary to ensure proper operation of the reactor control and safety systems, including the shim safety arms, regulating rod, and moderator dump system.

The staff reviewed the NBSR safety analysis and determined the control rod section adequately describes the reactivity control and shutdown systems of the NBSR. The descriptions include design, fabrication, acceptance testing, and reactivity worths of the systems. The analyses presented in the NBSR SAR (including accident analyses in Chapter 13) demonstrate sufficient reactivity worth for control of excess reactivity allowed by TS 3.1.2, adequate shutdown margin (TS 3.1.2), and acceptable control rod dynamic characteristics (TS 3.2.1) for both normal and accident conditions. Based on these considerations, the staff concludes that the reactivity control systems and related TSs provide reasonable assurance that the reactivity control systems will allow safe and reliable operation and shutdown of the NBSR.

4.2.3 Neutron Moderator and Reflector The tank-type design of the NBSR uses D2O as a moderator, reflector, and coolant. The core is immersed in D2O to thermalize fast neutrons to sustain the nuclear chain reaction, to remove heat created by the reaction, and to serve as the first stage of shielding. No other material is 4-4

used within or in the area immediately surrounding the core region to moderate the fast neutrons created by the fission process.

The side reflector is 50.8 cm (20 in) thick and the top reflector thickness is normally maintained at 300 cm (118 in). The top reflector level can be manually lowered to approximately 2.54 cm (1 in) above the core to effect an emergency shutdown of the reactor, as discussed in the previous section.

D2O chemistry is maintained by the primary coolant purification system to limit corrosion of the fuel elements and other materials in the reactor vessel and the primary coolant system. The purification system removes suspended particles and maintains the pH and conductivity by continuous flow of the D2O through filters and ion exchangers. TS 3.3.1, Primary and Secondary, requires all materials in contact with the primary coolant to be compatible with the D2O environment, thereby limiting degradation of the primary system chemistry. Evaporative losses and the products of radiolytic degradation are reduced by the helium sweep system which maintains a blanket of helium on the D2O in the reactor coolant systems. TS 3.3.1 limits D2 concentration to a maximum of 4% in the helium sweep system to ensure a substantial margin below the explosive value.

Thermal expansion of the D2O is accommodated by the D2O storage tank, which allows the primary coolant to expand and contract with variations in coolant temperature. The capacity of the tank is sufficient to hold the entire coolant inventory. D2O makeup is added directly from 55-gallon drums on an as needed basis.

Activation of the oxygen in the D2O produces a potential 16N radiation hazard. Passive shielding of the primary coolant system and access control at the NBSR limit worker exposures to radiation from 16N without the need for a dedicated active 16N control system.

The staff reviewed the NBSR safety analysis and determined the moderator and reflector section adequately describes the moderator and reflector design and safety considerations.

The description includes material compatibility with respect to chemical, thermal, and radiation environmental performance. The staff concludes that the NBSR moderator and reflector design adequately accounts for radiological degradation and the physical and chemical environment for the system. Continued operation as limited by the technical specifications offers reasonable assurance that the D2O moderator and reflector can continue to perform as designed and will not pose a significant risk to the NBSR or the health and safety of licensee personnel during the period of the renewed license.

4.2.4 Neutron Startup Source The normal power history of the NBSR produces a sufficiently strong photoneutron source for reactor startup, thereby obviating the routine use of an external neutron source. After extended shutdown, an encapsulated 1.9 Ci AmBe neutron source (half-life = 458 years), with a neutron yield of 2.2 x 106 neutrons per second per Curie, can be inserted into one of the vertical experimental thimbles to provide sufficient source neutrons for startup. The source is cooled directly by the D2O flowing through the experimental thimble. Once the reactor is critical and prior to raising the power level to 20 MW(t), the startup source is removed from the reactor and placed in its shielded storage container.

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To prevent the reactor operator from inserting positive reactivity into the core without having a visible indication of the power level in the reactor, shim blade withdrawal is blocked if the source range nuclear instruments indicate less than two counts per second.

The staff reviewed the documentation provided on the neutron startup source and concludes that the design is adequate to provide sufficient startup neutrons and source range indication for reactor startup.

4.2.5 Core Support Structure The core support structure is designed to ensure that all fuel elements, reactivity control devices, and in-core experimental facilities are properly secured against all anticipated loads including both the buoyant force of the coolant and the hydraulic forces associated with the primary coolant flow. The principal support features are the upper and lower grid plates. The upper grid plate is attached to four mounting brackets welded to the vessel wall. The lower grid plate is supported by the outer plenum flange plate which is welded to the outer plenum. The outer plenum is welded to the vessel bottom, so the vessel supports the load of the lower grid plate.

The grid structure in the upper and lower grid plates maintains the position of the internal core components. The fixed pattern in the grid plates aid in maintaining accurate positioning of the fuel elements, the reactivity control devices, and the experimental thimbles. This grid is designed to lock the fuel and other core components in place during reactor operation and to prevent movement of the core components by hydraulic forces. TS 3.1.3, Core Configuration, requires that core grid positions are filled with full length fuel elements or thimbles.

The selection of 6061-T6 aluminum alloy for the upper and lower grid plates makes them compatible with the materials of the vessel and primary piping. Aluminum is chemically compatible with the D2O coolant and exhibits excellent resistance to corrosion and erosion. It has low induced radioactivity and is resistant to radiation damage.

The staff reviewed the NBSR safety analysis and determined the core support structure section adequately describes the design for providing structural support for the core, accurate positioning of the fuel elements, and acceptable guides for other essential core components (shim safety arms, experimental thimbles, etc.). The core support structure is conducive to sufficient coolant flow as well as being compatible with the coolant and radiation environment.

The staff concludes that the core support structure is adequate for continued safe operation of the NBSR.

4.3 Reactor Tank or Pool The aluminum-alloy reactor vessel is 2.1 m (7 ft) in diameter and 4.9 m (16 ft) in height, and is designed in accordance with the ASME Boiler and Pressure Vessel Code for Unfired Pressure Vessels. The vessel is a vertical cylinder with an elliptical bottom and a flange top. The upper girth of the reactor vessel is made of 1.27-cm- (0.50-in) thick 6061-T6 aluminum alloy and extends down approximately 294 cm (116 in) below the surface of the reactor vessel flange.

The lower girth and the bottom are made of 2.22-cm- (0.875-in) thick 5052 aluminum alloy. The lower girth extends down from the upper girth approximately 422 cm (166 in) below the reactor vessel flange. The beam ports, through tubes, cold source, and rabbit tubes all attach to the reactor vessel in the lower girth. The reactor vessel is supported by the vessel flange which is bolted to the top of the thermal shield shim ring by twenty-four 2.54-cm (1-in) bolts.

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The design temperature for the reactor vessel is 121 °C (250 °F), and the design pressure is 50 psig. The normal reactor outlet temperature is 45.5 °C (114 °F) and the normal operating pressure is atmospheric. The maximum hydrostatic pressure, which occurs at the bottom of the reactor vessel, is 7.2 psig. The design of the vessel includes consideration of loads from constraining forces, along with those from steady state and transient thermal conditions. The vessels low heating rates and the excellent thermal conductivity of the aluminum combine to yield negligible stresses from internal temperature gradients. The very small temperature differentials between the coolant and the vessel components generate insignificant thermal transient loads. The NBSR vessel is fabricated entirely of aluminum alloys. Therefore, stresses resulting from differential expansion between dissimilar materials are negligible. The reactor vessel and its associated piping move freely under the influence of thermal expansion. Sliding pad-type pipe supports minimize reaction loads on the vessel, and in conjunction with all other loadings do not cause any stress levels above the maximum allowable working stress for various reactor sections.

No impact loads are transmitted to the vessel. Pressure surges that might be generated in the vessel by reactor power transients are small and would not cause the vessel to exceed the 50-psig design pressure. The licensees analysis of seismic forces from horizontal accelerations of 0.1 g resulted in combined stress levels from this loading plus all other design loads that were well within the allowable limits for the various sections of the vessel.

A 3 millimeter (mm) (0.125 in) corrosion allowance on all of the vessels pressure containment surfaces was incorporated into the design. Conservative estimates by the licensee predict a minimal corrosion rate. This was corroborated by a visual inspection of the vessels internal components that revealed little corrosion. The inspection was performed by the licensee in 1994.

The NBSR vessel is fabricated from aluminum alloys 5052 and 6061. Analysis of heavily irradiated (4.2 x 1023 thermal, 2.0 x 1022 fast neutrons per square centimeter (n/cm2) samples of the 6061-T6 alloy found in the literature indicates that the ductility retains approximately 70% of the original value, and the Charpy energy has dropped by over a factor of 6.3 The most heavily irradiated portions of the NBSR vessel, the tips of the beam tubes, will have accumulated less than 2 x 1023 n/cm2 thermal neutron fluence by the year 2024. Stress analysis performed by the licensee incorporating the reduction in Charpy energy and ductility indicate insignificant reductions in design safety margins.

The reactor vessel is penetrated by four D2O inlet and outlet pipes. The outer plenum is welded in the center of the vessel bottom while the inner plenum is located within and concentric to the outer plenum. The two outlet pipes are welded to the bottom on either side of the outer plenum pipe. The lower grid plate is bolted to both the inner and the outer plenums forming a watertight seal. The D2O holdup pan surrounds the core to a height just above the lower fuel section of the core and is attached to the lower grid plate, thereby trapping an inventory of cooling water during a loss of coolant accident.

3 Weeks, J.R., C.J. Czajkowski and K. Farrel, Effects of High Thermal Neutron Fluences on Type 6061 th Aluminum, Effects of Radiation on Materials: 16 International Symposium, AS Kumar, et al, eds., ASTM Publication Code Number 04-011750-35, Philadelphia, 1993.

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The reactor vessel design includes a 51-cm- (20-in) thick side reflector and a top reflector normally maintained at 300 cm (118 in). This D2O reflector surrounding the core serves as the first stage of shielding, followed by the thermal and biological shields.

Design features of the reactor vessel are specified in TS 5.2, Reactor Coolant System.

Material compatibility of all reactor vessel components with the D2O environment is ensured by TS 3.3.1.

The staff reviewed the NBSR safety analysis section on the reactor tank and determined the licensee has adequately demonstrated the design as capable of withstanding all anticipated mechanical and hydraulic forces to prevent loss of integrity which could lead to a loss of coolant or other malfunction. The licensee has also demonstrated by analysis and inspection that the design is capable of withstanding the corrosion and radiation environment in the tank for the period of the renewed license. The licensee also described reactor penetrations and provisions for maintaining core coolant coverage in the case of a loss of coolant accident (discussed in Chapter 13 of this SER). Based on the above considerations, the staff concludes that the reactor vessel design is adequate for continued safe operation of the NBSR.

4.4 Biological Shield The NBSR is enclosed in a shielding system consisting of the thermal and biological shields.

The combined effects of this system are a gamma-ray dose rate of approximately 0.02 millisievert per hour (mSv/hr) (2 millirem per hour (mrem/hr)) at the outer face of the biological shield and a negligible dose rate from neutrons. The gamma dose rate at the top of the center plug is about 0.001 mSv/hr (0.1 mrem/hr) and less than 0.005 mSv/hr (0.5 mrem/hr) in the immediate vicinity of a pick-up tool holding an element during fuel movement. All of these dose rates are well within the requirements of 10 CFR 20 and the guidelines of the facility ALARA program.

The thermal shield consists of 5 cm (2 in) of lead and 20 cm (8 in) of steel and except for the top of the vessel and experimental ports, nearly surrounds the Reactor Vessel. The thermal shield is cooled by H2O to remove the energy deposited by captured radiation. At full power, about 350 kilowatts (kW) is deposited in the thermal shield, preventing the concrete in the biological shield from excessive heating.

The design of the shield minimizes the effects of voids, necessary due to some structural features such as a pipe or shutter well, by adding enough lead to compensate for the gamma-stopping power of the concrete that was removed. Experimental beam holes designed to extract intense radiation beams from the reactor require extensive individual shielding. Each beam line was specifically reviewed by the licensee at the design stage, checked upon installation, and verified to have acceptable radiation dose rates during operation.

The staff reviewed the NBSR biological shield design and determined the biological shield section adequately describes the design and offers reasonable assurance that the shield design will limit exposures so as not to exceed the limits of 10 CFR 20 and the guidelines of the NBSR ALARA program.

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4.5 Nuclear Design 4.5.1 Normal Operating Conditions The normal operating cycle for the NBSR is a 38-day cycle beginning with less than 15%

excess reactivity (TS 3.1.2) and ending with no excess reactivity and all control elements fully withdrawn. A typical startup core excess reactivity of 6.57% is provided by example calculation in Section 4.5.1.3.2 of the NBSR SAR. The shim safety arms and the regulating rod are cooperatively utilized during the cycle to manage the reactivity changes resulting from temperature changes, fission product poisons, and burn-up.

A typical shutdown margin of 9.4% is provided by example calculation in Section 4.5.1.3.2 of the NBSR SAR with the highest worth shim safety arm fully withdrawn for the beginning of cycle (BOC) startup core. A redundant minimum shutdown margin of 7.4% as demonstrated by example calculation in the SAR is provided by the moderator dump system. These shutdown margins exceed the requirements of TS 3.1.2. Individual shim safety arm and regulating rod reactivity worths are determined annually or following any significant change in core or shim arm configuration, as required by TS 4.1.2.

The NBSR utilizes a combination of 7-cycle and 8-cycle fuel elements in a standardized fuel management configuration that results in 38 days of full power operation consuming approximately 970 grams 235U. The average burn-up in the 7 & 8-cycle fuel elements is 66%

and 73%, respectively. Due to the enrichment of the fuel, the total Pu inventory at EOC has a negligible effect on reactivity. At the beginning of each cycle, four fuel elements are replaced with fresh fuel elements, and the remainder rearranged according to the standardized fuel management scheme. Maximum fuel burn-up is limited to 2.0 x 1027 fissions/m3 by TS 3.1.4.

The experimental facilities in the NBSR represent large voids in the core reflector which may insert positive reactivity if flooded by a crack in the beam tubes or failure of a D2O-cooled experiment. The largest single reactivity insertion from such a failure would be flooding of the cold neutron source, resulting in a reactivity addition of 0.49% . This is below the technical specification reactivity limit of 0.5% for a single experiment, as specified in TS 3.8.1, Reactivity Limits. An analysis of a ramp reactivity insertion of 0.5% in 0.5 sec in Chapter 13 of the NBSR SAR demonstrated no loss of fuel integrity from the accident.

Technical specifications are provided to restrict the reactor core parameters to acceptable fuel loading, configuration, and burn-up (TSs 3.1.2, 3.1.3, 3.1.4); ensure adequate reactivity control (TS 3.1.2, 3.2.1); and limit the reactivity associated with experimental facilities (TS 3.8.1).

The staff evaluated the normal operating conditions for the NBSR and concludes that the TSs provide reasonable assurance that continued operation of the NBSR will not pose an undue risk to the health and safety of the public or the environment.

4.5.2 Reactor Core Physics Parameters The reactor physics parameters for the NBSR are determined primarily by calculations using the MCNP computer code. MCNP is a standard industry nuclear physics tool developed at Los Alamos National Laboratory. The code has been benchmarked against critical experiments and power reactor applications. The licensee benchmarked its MCNP results against measured values for the regulating rod and shim bank reactivity worths.

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Delayed neutron groups are presented in the 1980 Addendum to the original NBSR SAR and were taken from the literature.4,5 At a steady reactor power, the fraction of all the neutrons that are delayed, eff = 0.007574.

D2O moderated and reflected research reactors have prompt neutron lifetimes reported to be on the order of 700 microseconds (µs) (GTRR: 770 µs; expanded SPERT-II core: 750 µs; HFBR:

672 µs)6,7,8. The licensee calculated values of the prompt neutron lifetime using two-group diffusion theory and a homogenized core model of 500 µs to 800 µs, depending on the core volume and the fuel loading. Prompt neutron lifetimes calculated at Brookhaven National Laboratory (Hanson et al, 2004) using MCNP simulations of a pulsed neutron source in the subcritical NBSR produced neutron lifetimes of 774 +/- 35 µs (startup core configuration (SU))

and 819 +/- 48 µs (EOC).9 Although the fuel elements in the reactor core are widely spaced, the NBSR is an under-moderated reactor. The bulk temperature coefficient for the D2O moderator for the SU and EOC cores is -0.017%/ºF and -0.014%/ºF, respectively. The magnitude of the void coefficient depends on the location of the void. The values calculated for the moderator between the fuel elements are -0.043 %/liter (SU) and -0.030 %/liter (EOC). Therefore, a decrease in the D2O density anywhere in the reflector, the moderator, or the coolant inside the fuel elements results in a negative reactivity insertion, thereby contributing to the inherent safety of the core design.

Axial and radial flux distributions are calculated by the licensee using MCNP for SU and EOC cores. The maximum thermal flux, 3.5x1014 n/cm2/s, occurs very close to the core mid-plane in the unfueled region between the upper and lower cores. The calculated values are in good agreement with the measured peak thermal neutron flux in the central thimble of 3.5x1014 n/cm2/s at 20 MW(t). Calculated distribution patterns agree favorably with expected results based on fuel arrangement, the location of the control elements, and the location of the unfueled region in the core central plane.

For use in thermal-hydraulic analyses, the peak-to-average heat generation factors were calculated by the licensee using the MCNP results to determine the energy deposited in all 1080 coolant channels for the SU and EOC cores. Fuel element, axial, and lateral peaking factors are determined and then corrected for uneven burn-up to determine the relative power factors for the limiting thermal-hydraulic analyses. These factors were determined for both SU and EOC cores.

The staff has evaluated the reactor core physics parameters for the NBSR and concludes that the values and calculation methods are appropriate and consistent with methods for similar 4 Tuttle, R.J., Delayed Neutron Data for Reactor Physics Analysis, Nuclear Science and Engineering, 56, p. 37, Jan. 1975.

5 Johns, M.W. and B.W. Sargent, Canadian Journal of Physics, 32, p. 136, 1954.

6 Bretscher, M.M., Perturbation-Independent Methods for Calculating Research Reactor Kinetic Parameters, ANL/RERTR/TM-30, Argonne National Laboratory, Argonne, IL, 1997.

7 Grund, J.E., Self-Limiting Excursion Tests of a Highly Enriched Plate Type D O Moderated Reactor:

2 Part I, Initial Test Series, IDO-16891, Phillips Petroleum Company, 1963.

8 Hendrie, J.M., Final Safety Analysis Report on the Brookhaven High Flux Beam Reactor, BNL-7661, Brookhaven National Laboratory, Upton, NY, 1964.

9 Hanson, A.L., H. Ludewig and D. Diamond, Calculation of the Prompt Neutron Lifetime in the NBSR, Nuclear Science and Engineering, Vol. 153, 2006.

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reactors. Temperature and void coefficients are negative everywhere in the core and provide inherent safety characteristics during normal operation and transients. The staff concludes that flux distribution methods are adequate, and the selection of peaking factors acceptable for use in normal operating limit and accident analysis calculations.

4.5.3 Operating Limits TS 3.1.3 limits the core configuration to the 30 fuel element loading as analyzed in Chapter 13 of the NBSR SAR. The excess reactivity is limited to a maximum of 15 % , and the shutdown margin is required to be greater than $1.00 for any core condition, with all movable experiments in their most reactive condition. The reactor is also required to remain subcritical with the highest worth shim arm fully withdrawn. These restrictions on excess reactivity and shutdown margin are specified in TS 3.1.2. The staff reviewed the calculations provided in the NBSR SAR and determined that they demonstrate the adequacy of the shutdown margin for all core configurations.

Technical Specifications 3.2.1 and 3.3.3 require all four shim safety arms and the moderator dump system to be operable for reactor operation. TS 3.2.1 also limits the maximum reactivity insertion rate to 5x10-4 /sec. These limits conservatively ensure that adequate shutdown margin is maintained since only three of the four shim safety arms are required for shutdown of the most reactive core. In addition, the moderator dump system, which would provide adequate shutdown margin with all shim safety arms withdrawn, is required to be operable during reactor operation to ensure redundant shutdown margin capability. The maximum insertion rate limit ensures no fuel damage from reactivity insertion excursions. This insertion rate is analyzed in Chapter 13 of the NBSR SAR for a rod withdrawal accident using the maximum beginning-of-life rod worths with the rods operating at the design speed of their constant speed mechanisms.

The analysis shows that the most severe accident, a startup withdrawal accident from source level, is bounded by the maximum reactivity insertion accident, and will not result in core damage.

The reactivity associated with experiments in the NBSR is limited by TS 3.8.1 to a maximum of 0.5 % for a single experiment, and all experiments are limited to a total of 2.6 % .

Analysis in Chapter 13 of the NBSR SAR shows that if the most reactive single experiment were removed in 0.5 seconds, this ramp insertion into the NBSR operating at 20 MWt would not result in any fuel failure.

The staff has reviewed the NBSR safety analysis and determined the nuclear design section adequately describes the nuclear design characteristics necessary to ensure safe and reliable operation under normal operating conditions. Reactor core physics parameters are determined by acceptable analytical methods, and the technical specifications require operating limits that will ensure fuel integrity. The staff concludes that the nuclear design, as limited by the technical specifications, is adequate for continued safe operation of the NBSR.

4.6 Thermal-Hydraulic Design The thermal-hydraulic design basis for the NBSR is that there shall be no fuel damage resulting in the release of fission products during normal operation or from any credible accident. As previously discussed, the fuel cladding may begin to blister at 450 ºC (840 °F), and during the blistering process, cracks will develop that can release gaseous fission products. The criteria for ensuring this limiting fuel clad temperature is not exceeded are that no Departure from Nucleate Boiling (DNB) or Onset of Flow Instability (OFI) conditions in the coolant.

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The licensee selected nominal settings of flow and inlet temperature by the requirement that there be no nucleate boiling at the hottest spot on the fuel cladding. This was conservatively estimated by the requirement that the fuel clad temperature at the hot spot remain below the saturation temperature for D2O at the appropriate pressure. The licensee calculated the clad temperature with a heat transfer coefficient determined using the Dittus-Boelter correlation.

Having determined nominal flows, the limits of safe operation were determined using the most limiting criteria of DNB and OFI. For forced circulation, the licensee chose the Mirshak correlation to test for DNB, based upon the close similarity of the conditions for which the correlation was determined to those in the NBSR core. The Costa correlation was used to test for OFI.

For natural convection, the licensee utilized the Sudo-Kaminaga and Oh/Chapman correlations to check for DNB and OFI, respectively. These calculations were performed for both the inner and outer plenums, at both the hot spot and the exit of the hottest fuel channel of the upper fuel section. Critical heat flux (CHF) and OFI ratios calculated by the licensee show ample thermal safety margins for steady state operating conditions. The RELAP5 code was used by the licensee to analyze abnormal transients during 500 kW power operation with natural convection.

These analyses show peak clad temperatures less than the blistering temperature, thereby demonstrating the acceptability of the limiting safety system settings (LSSS) required by TS 2.2, Limiting Safety System Settings, for operation with natural convection. Although the licensee performed calculations for operation at 500 kW, TS 2.2 requires that the reactor power be limited to 10 kW during operation with natural circulation. This requirement provides a large safety margin for operation with natural circulation.

The licensee demonstrated that shutdown cooling will provide ample cooling for all shutdown conditions, including loss of offsite power followed by failure of both shutdown pumps. This scenario, as analyzed in Chapter 13 of the NBSR SAR, results in no damage to the fuel, thereby showing that natural convection cooling is adequate to provide cooling of the fuel, even immediately following a scram due to loss of all primary pumps.

The reactor safety limit (TS 2.1) states that the fuel cladding temperature shall not exceed 450 °C (840 °F) for any operating conditions of power and flow. Maintaining the cladding temperature below this limit will prevent blistering and thereby ensure fuel integrity. Ensuring the safety limit is not violated is provided for in the LSSS (TS 2.2). The LSSS are based on the onset of nucleate boiling temperature as determined with the correlation of Bergles and Rohsenow. Conservative calculations performed by the licensee in Chapter 13 of the NBSR SAR have shown that the LSSS limiting combinations on reactor power (130%), coolant outlet temperature (147 ºF), and coolant flow will ensure that any reactor transient caused by equipment malfunction or operator error will be terminated well before the safety limit is reached, including allowances for uncertainties in process instrumentation.

The staff has reviewed the NBSR safety analysis and determined the thermal-hydraulic design section adequately demonstrates the thermal-hydraulic characteristics necessary to provide the limits on cooling conditions that ensure fuel integrity will not be lost under any reactor conditions, including accident conditions. Thermal-hydraulic parameters are determined by acceptable analytical methods, and safety limits and LSSS are specified in the technical specifications that will ensure compliance with the design criteria of no DNB or OFI. The staff concludes that the thermal-hydraulic design, as limited by the technical specifications, is adequate for continued safe operation of the NBSR.

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4.7 Conclusions The staff concludes that the licensee has adequately described the bases and functions of the reactor design to demonstrate that the NBSR can be safely operated and shut down from any operating condition or accident assumed in the safety analysis. The systems provide adequate control of reactivity, containment of coolant, and barriers to the release of radioactive material as well as sufficient radiation shielding for the protection of facility personnel. Nuclear and thermal-hydraulic design and operating limits as established by the technical specifications adequately provide for the protection of fuel integrity. Therefore, the staff concludes that continued operation of the NBSR within the limits of the technical specifications and facility license will not result in undue risk to the health and safety of facility personnel, the public, or the environment.

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5 REACTOR COOLANT SYSTEMS 5.1 Summary Description The reactor coolant systems at the NBSR facility include the following systems: primary coolant system, secondary coolant system, primary coolant purification system, primary coolant makeup, nitrogen-16 control, and D2O experimental cooling system. The primary purposes of the reactor coolant systems are to remove the fission and decay heat generated in the core; to dissipate the decay heat to the environment; and to serve as one of the barriers to prevent fission product release to the environment. The primary coolant is D2O, and the secondary coolant is H2O.

5.2 Primary Coolant System The primary coolant system is designed to transfer 20 MW(t) of heat from the core to the secondary coolant system with nominal operating values of 9,000 gallons per minute (gpm) flow, 38.8 °C (100 °F) reactor inlet temperature, and 45.6 °C (114 °F) reactor outlet temperature. The reactor coolant systems at the NBSR facility are designed to remove sufficient heat to support continuous full-power operation at a power level of 20 MW(t) and remove the decay heat generated after shutdown from extended full-power operation.

The primary coolant system consists of pumps, heat exchangers, piping, and valves, and is located entirely within the reactor confinement building. While this system is not pressurized, it is closed to the atmosphere. Therefore, it serves as one of the three barriers to fission product release, the other two being the fuel cladding and the reactor confinement building.

The primary coolant system normally operates under conditions of forced flow in which primary coolant enters the bottom of the reactor vessel through the inner and outer plenums. The inner plenum feeds primary coolant to the center 6 fuel assemblies, while the outer plenum feeds the remaining twenty-four fuel assemblies. The coolant flows up through the fuel, removing the heat generated by fission, before exiting from the bottom of the vessel through two outlet pipes.

Then, the primary coolant flows through the D2O main circulating pumps to plate-type main heat exchangers, where the heat from fission in the core is transferred to the secondary coolant. The primary coolant passes through a strainer before returning back to the reactor vessel. A shutdown cooling system is provided to remove decay heat, although analyses in Chapter 13 of the NBSR SAR demonstrate that natural circulation cooling alone is adequate to protect the integrity of the fuel.

The process room contains the piping, strainers, D2O main circulating pumps, D2O shutdown pumps, main heat exchangers, control valves, and instrumentation associated with the primary coolant system. A curb captures any primary coolant that may leak from the system and collects it in a sump. A sump pump is provided to return spilled coolant to the overhead storage tank as part of the emergency core cooling system, and is required for reactor operation by TS 3.3.2, Emergency Core Cooling. Vent and drain lines are equipped with manual valves and flanges with quick-disconnect fittings to provide two barriers between the primary coolant and the atmosphere.

Potential leakage from the primary to secondary systems is monitored by detectors located in the secondary system. If a detector alarms, the secondary water is sampled for 3H. In addition, a leak into the secondary system can be detected by a change in the level of the D2O storage tank and by periodic sampling of the secondary water for 3H. These methods are sensitive 5-1

enough to detect a leak of above 135 liters (l) (36 gallons (gal)) in one day or 190 l (50 gal) in one week.

Either a secondary cooling water activity monitor or a D2O storage tank level monitor is required for reactor operation by TS 3.7.1, Monitoring Systems, and Effluent Limits. Primary coolant 3H activity is restricted to 5 Curies per liter (Ci/l) or less by TS 3.7.1 to ensure effluent 3H concentrations are below regulatory limits. Deuterium gas (D2) in the helium cover gas system is monitored and recombined in a catalyst bed to prevent concentrations from exceeding the 4%

limit (TS 3.3.1). This limit on D2 gas ensures a substantial margin below the lowest potentially explosive concentration.

Primary system instrumentation provides information on reactor inlet flow to each plenum, reactor outlet flow, reactor T, reactor vessel level, reactor overflow, and primary-to-secondary P in the main heat exchangers to the control room. This arrangement, including alarm panels to alert the operator to changing conditions, gives the reactor operator a single convenient location from which to monitor and operate the reactor. Low level scram and low reactor outlet flow channels are required for reactor operation, as specified in TS 3.2.2, Reactor Safety System Channels.

The staff has reviewed the primary coolant system design and technical specifications described in the NBSR SAR. Technical specifications ensure that necessary coolant system equipment and instrumentation are provided for reactor operation. Primary coolant system integrity is assured by technical specifications ensuring monitoring for system leakage, D2 gas accumulation, and pressure relief system functionality. Surveillance requirements for equipment operability and sampling frequency are provided in TSs 4.2.2, Reactor Safety System Channels, 4.3.1, Primary and Secondary, and 4.7.1, Monitoring System.

The staff concludes that the primary system design and technical specifications provide reasonable assurance of necessary primary coolant system operability for reactor operations as analyzed in the NBSR SAR while posing no undue risk to the health and safety of the public, licensee personnel, or the environment.

5.3 Secondary Coolant System The secondary coolant system is designed to transfer heat from the following heat exchangers associated with the reactor coolant and auxiliary support systems: main heat exchangers (HE-1A, 1B, 1C), D2O purification heat exchanger, (HE-2), thermal shield heat exchanger (HE-6),

thermal column heat exchanger, experimental demineralizer heat exchanger (HE-7), and the helium compressor secondary cooling heat exchanger. The heat load in the secondary coolant from these heat exchangers is transferred to the atmosphere via a 22 MW(t) hybrid wet/dry, plume abatement cooling tower. The heat-removal capacity of the system exceeds the total heat generated in all of these individual systems.

The secondary coolant system circulates H2O through two 10 MW(t) main heat exchangers to remove heat from the primary coolant that is generated by fission in the core. There are six parallel main secondary coolant pumps, arranged in two sets of three parallel pumps. When the reactor is shutdown, a single smaller pump provides secondary flow. Secondary system flow is measured by an installed flow element. The operator maintains a constant reactor inlet temperature by regulating the amount of secondary coolant bypassing the cooling tower.

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Upon leaving the main heat exchangers, part of the secondary coolant passes through a radiation detector and a test-coupon station. The radiation detectors monitor the secondary water for the presence of 16N, an indicator of a primary-to-secondary leak (TS 3.7.1). The test coupons monitor for any long-term effects that the secondary coolant might be having on the secondary piping.

A chemical addition system regulates corrosion and biological growth in the secondary system.

Water is continuously blown down to the sewer system to remove concentrated solids and to maintain a low concentration of dissolved solids.

Two secondary auxiliary booster pumps supply water from the discharge header of the main secondary coolant pumps to the D2O purification heat exchanger, the thermal shield heat exchanger, and the thermal column heat exchanger. Normally, one of these pumps is operating while the other remains in standby.

The helium compressor secondary cooling pumps supply water from the suction header of the main secondary coolant pumps to the helium compressor secondary cooling heat exchanger.

This removes the heat generated in the cold source refrigerator. Normally, one of these pumps is operating while the other remains in standby.

TS 3.7.1 and 4.7.1 are provided to ensure adequate monitoring for radioactivity in the secondary cooling system and to maintain 3H releases below regulatory limits. These TSs provide reasonable assurance that any primary-to-secondary coolant leaks will be readily detected and not result in any significant risk to public health and safety or the environment.

5.4 Primary Coolant Cleanup System The primary coolant purification system is designed to maintain the chemistry and purity of the primary coolant by removing both soluble and insoluble corrosion products and other foreign materials. The chemistry must be properly controlled to ensure that the components in contact with the primary coolant are not degraded over the life of the plant. The purification system maintains the primary coolant pH between 5.0 and 6.0 and the conductivity less than 1 micromho (µmho).

Mechanical filtration of the primary coolant removes particles 5 microns and larger. Purity of the primary coolant is essential to minimize the contaminants that might be exposed to the neutron flux, thereby decreasing personnel radiation exposure and limiting production of radioactive waste.

The primary coolant cleanup system design is consistent with recommended operational parameters for aluminum-water systems found in the literature and with systems for similar licensed non-power reactors.10 The staff concludes that continued operation within the parameters described in the NBSR SAR provides reasonable assurance that corrosion of the system components and contaminant levels will be acceptable.

10 DOE Fundamentals Handbook, Chemistry, Vol. 1 & 2, DOE-HDBK-1015, U.S. Department of Energy, Washington, D.C., January 1993.

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5.5 Primary Coolant Makeup Water System The D2O storage tank has sufficient capacity to hold the entire coolant inventory of the primary coolant system and its associated D2O systems. Accordingly, the NBSR reactor does not have a dedicated system for adding make-up water to the system. D2O is added directly from 210 l-(55-gal) drums on an as needed basis. Since the make-up water is added through existing return lines, the operation will not result in loss of primary coolant and cannot contaminate any potable water supply.

5.6 D2O Experimental Cooling System The D2O experimental cooling system distributes D2O from the primary coolant purification system to cool the cold neutron source; pneumatic sample tubes RT-1, RT-2, and RT-4; and other experimental facilities. The system utilizes cooled primary coolant from the discharge of the D2O purification heat exchanger and returns the heated coolant to the D2O storage tank.

In an emergency, D2O is available from the D2O emergency cooling tank to cool the system components. An emergency back-up supply of cooling water is also available from the domestic water system.

5.7 Conclusions The design of the NBSR cooling systems, as described in the NBSR SAR, are adequate for the removal of heat generated during continuous full power reactor operation and for the removal of decay heat after shutdown from extended full-power operation. The systems contain sufficient features to protect personnel from excessive radiation hazards, minimize corrosion of system components and fuel, prevent or detect losses of coolant, and provide one of the barriers to prevent fission product release to the environment. The staff concludes that the coolant systems of the NBSR are sufficient for continued safe reactor operation within the related limits of the facility license and TSs.

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6 ENGINEERED SAFETY FEATURES 6.1 Summary Description Engineered safety features (ESFs) are designed to prevent or mitigate accidents by controlling the release of radioactive materials to the environment. The ESFs at the NBSR include the emergency cooling system, the confinement building, and the ventilation systems. ESFs can be actuated automatically by the protection instrumentation that monitors various parameters during reactor operation, or manually by the reactor operator. The ESFs provide protection against (1) overheating of the core should forced flow of primary coolant be unavailable and (2) uncontrolled release of radioactive material to the surrounding environment.

6.2 Detailed Descriptions 6.2.1 Confinement The confinement building is a three-level structure with a volume of approximately 16,000 cubic meters (m3) (600,000 cubic feet (ft3)), and was designed and constructed to ensure minimum air leakage. The ventilation systems allow the building atmosphere to be maintained at a slight under-pressure during both normal and emergency conditions, assuring that any leakage is into the confinement building. All ventilation ductwork that penetrates the reactor building has automatically sealing closure valves or dampers. Personnel access to the confinement building is through entrances equipped with a sliding steel door with inflatable gaskets. During normal operations, these sliding doors are fully open. In an emergency, the sliding steel doors automatically close and the gaskets automatically inflate to seal the entrances to the building.

Because the building will be sealed during emergency conditions, a vacuum relief valve is incorporated to prevent any detrimental pressure differential from developing across the building walls or roof.

TS 3.4.1, Operations that Require Confinement, specifies the operating conditions that require confinement, and TS 3.4.2, Equipment to Achieve Confinement, prescribes the equipment operability requirements to establish confinement. Surveillance tests required by TS 4.4, Confinement System, periodically check the operability of the trip features of the confinement closure system and require an annual integrated leakage rate test to ensure the leakage rate remains within acceptable limits. These TSs ensure that the confinement system is available in the event of a significant radiological release and that the release rate will be within that analyzed in an MHA. The MHA analyzed in Chapter 13 of the NBSR SAR demonstrates for a worst case accident scenario that the confinement system adequately controls and mitigates the release of radiological material from the confinement system such that radiological doses to members of the public are within the regulatory requirements of 10 CFR Part 20 and 10 CFR Part 100.

6.2.2 Ventilation The intake air-supply for the normal ventilation system includes both fresh air from outside and recirculated air within the building. All effluent air exhausted from the confinement building is monitored for radioactivity. Each of the effluent pathways have filter banks which are monitored for particulate activity, air samples are withdrawn for counting of gaseous activity, and a monitor measures activity in the stack at the point of release to the environment. In the event high radiation levels are detected, the normal ventilation system will be shutdown, all building closure devices will be sealed, and the emergency ventilation system will be activated.

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Under emergency conditions, the air interior to the confinement building can be recirculated after being filtered through a system consisting of both a high-efficiency particulate (HEPA) and a charcoal filtering system. In the event the inside air warrants cooling and humidity control, the normal air conditioning system could be used to condition the inside air as appropriate during an emergency. In addition, the emergency exhaust system is designed to draw air at such a rate from the building that a pressure differential can be established across the building structure to ensure that any leakage is into the building regardless of likely outside pressure variations due to wind or barometer changes. The emergency exhaust system consists of two redundant subsystems each of which could draw air from the normal exhaust system ductwork. Air exhausted from the building in each subsystem passes through a filtering system consisting of both HEPA and charcoal filters before being released through the stack.

TS 3.5, Ventilation System, specifies ventilation system requirements necessary for reactor operation. Surveillance requirements in TS 4.5, Ventilation System, provide for periodic tests of valves, controllers, instruments, particulate and charcoal filters to ensure operability of the ventilation systems. The requirements of these TSs ensure the emergency functions of the ventilation system will be available and functional in order to provide for a controlled, mitigated release to the environment in the event of a radiological accident.

6.2.3 Emergency Core Cooling System The emergency cooling system (ECS) provides cooling for the reactor core should primary coolant be lost through leakage from the primary coolant system. An inner reserve tank located within the reactor vessel contains approximately 3028 l (800 gal) and would provide a minimum of 28 minutes of coolant flow to the core with no operator action. The D2O emergency cooling tank located above the reactor vessel can provide an additional 11,360 l (3,000 gal) of D2O to the core top or to each of the reactor inlet plena by manipulating control valves from the emergency cooling tank. In either case, emergency coolant flows through the reactor vessel and primary coolant system to the pipe rupture location where it drains out onto the process room floor. The concrete curb and floor drains in the process room direct the D2O to the emergency sump where it is either pumped back to the D2O emergency cooling tank or to the D2O storage tank located in the confinement building. There is sufficient D2O in the inner reserve tank and emergency cooling tank to provide 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of cooling on a once-through basis. The ECS also has the capability to add domestic H2O to the D2O emergency cooling tank through a spool-piece and double manual isolation valves.

TS 3.3.2 requires the ECS to be operable for reactor operation and requires a source of makeup water for the system. Surveillance tests of the valves and pump required by TS 4.3.2, Emergency Core Cooling System, check the proper operation of the ECS components.

Analysis in the NBSR SAR shows that the ECS provides adequate cooling during a loss of coolant accident. Therefore, the availability and operability of the ECS, as provided by these TSs, and the NBSR loss of coolant accident analysis provide reasonable assurance the ECS can continue to adequately protect against melting of the core and the associated release of fission products.

6.3 Conclusions The staff concludes that the design of the engineered safety features, as described in the NBSR SAR, provide adequate protection to prevent the loss of fuel integrity from a loss of coolant 6-2

accident and mitigate the release of radioactive material to the surrounding environment to acceptable levels, as demonstrated by the MHA analysis in Chapter 13 of the NBSR SAR.

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7 INSTRUMENTATION AND CONTROL SYSTEMS 7.1 Summary Description The NBSR instrumentation and control (I&C) system consists of five major subsystems: the Reactor Control System (RCS); the Reactor Protection System (RPS); the ESF; the Main Control Panel; and the Radiation Monitoring System (RMS). These subsystems consist of instrumentation, controls, and annunciators most of which have control or indication devices at the Main Control Panel to allow for remote operation and monitoring by the reactor operator.

7.2 Design of Instrumentation and Control Systems The NBSR I&C System design criteria consist of the following elements:

• Redundancy and diversity.

• Automatic initiation to mitigate the consequences of abnormal conditions.

• Fail-safe.

• Single failure will not prevent a safe shutdown of the reactor.

• Instruments required for the reactor to be placed in a shutdown condition are supplied by emergency power.

• Instruments important to reactor safety are redundant or diverse and their signal cables are routed in separate cable trays and cable chases to prevent common mode failures.

Specific design elements of the separate I&C subsystems are discussed in the following sections.

7.3 Reactor Control System The RCS provides for the withdrawal and insertion of the four shim safety arms and the regulating rod. The system consists of four individual shim safety arm withdraw/insert circuits, one regulating rod withdraw/insert circuit, their associated interlocks, and the automatic control circuit. The operating controls and positioning information are provided on the Main Control Panel. The system is enabled by the Startup Prohibit and the Withdraw Prohibit circuits. In the presence of a prohibit signal from either circuit, withdrawing any reactivity control device is prevented, regardless of whether the reactor is being operated in Manual or Automatic mode.

The Rundown function consists of selected plant parameters, any of which indicating an abnormal condition results in the four shim safety arms and the regulating rod being driven into the reactor core to automatically reduce reactor power. A reactor rundown, once initiated, continues until the condition clears or until the reactivity control devices are fully inserted.

Operability of the shim arms is required by TS 3.2.1, and periodic surveillance tests are required by in TS 4.2.1.

Eight separate channels of Nuclear Instrumentation monitor the reactor power level and period continuously from shutdown to full power. All eight channels are displayed and recorded on the Main Control Panel for use by the reactor operator. The detectors are located in instrument 7-1

wells located in the biological shield; hence, the detectors measure only the leakage flux from the core.

Source Range Channels NC-1 and NC-2 monitor the reactor power in the source range.

Intermediate Range Channels NC-3 and NC-4 provide reactor period trip signals to the RPS.

Linear Power Channel NC-5 provides a reactor power level signal to the automatic flux controller. Power Range Channels NC-6, NC-7 and NC-8 provide reactor power trip signals to the RPS. These units are designed to be failsafe such that a loss of power will cause a reactor scram. A minimum of one decade of overlap is designed into the transition between Source Range and Intermediate Range Nuclear Instrumentation and between Intermediate Range and Power Range Nuclear Instrumentation. Normally, the photoneutron source is sufficiently strong that the two Source Range Channels are not necessary. As a result, these channels are normally de-energized and their detectors removed from the instrument wells.

7.4 Reactor Protection System The RPS includes the Scram, Major Scram, and Moderator Dump methods for rapid shutdown of the reactor. A trip of any of the scram relays will cause a loss of power to all of the shim safety arm magnets, thereby decoupling the arms from their respective drive motors and dropping the arms into the core by gravity and compressed spring force. The scram functions are controlled by logic circuits containing relay contacts from the various Nuclear and Process Instrumentation circuits necessary for safe operation of the reactor. If any of the contacts in the scram logic strings are open, indicating that the plant parameters associated with these contacts are not in their normal range, then power is removed from the scram relays resulting in a reactor scram. Instrumentation and scram channel requirements during reactor operation are specified in TS 3.2.2, and the surveillance requirements for tests and calibrations to check operability of the system is prescribed in TS 4.2.2. These requirements provide protective action for nuclear and process variables to ensure the LSSS values are not exceeded.

The Normal Air Monitor Channel, Irradiated Air Monitor Channel, and the Stack Monitor Channel control the relays in the Major Scram circuit. Upon the detection of an excessive activity level by any of the three channels, the Major Scram relays open contacts in the scram logic string and initiate confinement building isolation. The relays also shut the doors at the entrances to the confinement building by tripping the Door Scram Relays , shift the ventilation lineup to recirculation mode by tripping the Fan Scram Relays , and close the Neutron Guide Isolation Valves. Two of the three gaseous effluent monitors (Normal Air, Irradiated Air, and Stack Air) are required to be operable during reactor operation by TS 3.7.1.

The Moderator Dump function is also part of the RPS. Negative reactivity can be inserted by removing the moderator from the area immediately above the reactor core. The Moderator Dump switch on the Main Control Panel provides a backup shutdown option to the reactor operator. Actuating the dump switch opens contacts in the scram logic string initiating a reactor scram. Additional contacts in this switch open the Moderator Dump Valve DWV-9, dumping primary coolant above the core into to the D2O Storage Tank. These contacts also open the circuit breakers supplying power to the Main Coolant Pumps. The moderator dump system is required to be operable during reactor operation by TS 3.3.3.

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7.5 Engineered Safety Feature Actuation Systems ESFs are designed to mitigate accidents and control the release of radioactive materials to the environment should an accident occur. The NBSR ESFs include the Emergency Cooling System, confinement building, and the Ventilation Systems. ESFs may be automatically actuated by the protection instrumentation that monitors various parameters during the reactors operation, or manually by the reactor operator. For example, in addition to shutting down the reactor, a major scram signal will actuate the Confinement Building Isolation System to prevent the release of radioactivity to the environment. This signal may be generated manually by the reactor operator, or automatically in response to high radiation levels detected by radiation monitors in the ventilation system.

7.6 Control and Console Display Instruments The NBSR Main Control Panel in the Control Room provides all of the information and controls needed by the operator to safely operate the reactor from a centralized location. The center, angled, portion of the console provides all of the instrumentation and controls associated with the reactor. Nuclear instrumentation provides overlapping indication of reactor power level from startup to full power as well as indication of reactor period. Controls and indications are provided on this section of the console to control the shim safety arms, the automatic regulating rod, the primary and secondary cooling pumps, and the operation of the cooling tower. Three annunciator panels, located at the top of the vertical backboard of the console, alert the reactor operator to off-normal conditions in these systems as well as indicate the source of scrams and rundowns. An additional, three-window annunciator panel located just below the center annunciator panel alerts the operator to scram, rundown, and withdraw prohibit conditions.

The console to the left of the reactor controls provides instrumentation and controls associated with the auxiliary systems, experimental facilities, and radiation monitoring equipment. Two annunciator panels on this portion of the console alert the reactor operator to off-normal conditions in these systems. Each alarm is labeled with the underlying cause (e.g., Reactor D2O Level High). In addition, the number of the annunciator panel and the individual alarm window (e.g., AN3-14) corresponds to the associated Annunciator Procedure number, thereby assisting the operator in quickly locating the appropriate alarm response procedure to take corrective action for each alarm condition.

The console to the right of the reactor includes the Nuclear Instrumentation Cabinet and additional instrumentation and controls associated with nuclear instrumentation, experimental facilities, and radiation monitoring equipment.

A second function of the display system is to provide essential information at the Emergency Control Station located outside of the Confinement Building in the basement level, B-2, of the NBSR office building, located adjacent to the Confinement Building. Information is available at this location for use during emergencies that result in the Confinement Building becoming inaccessible.

7.7 Radiation Monitoring Systems The RMS consists of both area and effluent monitors. The area monitors provide indication of radiation levels throughout the containment building. These have been positioned at locations where either experimental work is performed or where work involving radioactive material is likely to be undertaken. As a result, the reactor operator can observe radiation levels and warn 7-3

both experimenters and operations personnel of any unanticipated changes or hazards. The effluent monitors provide indication of the radioactivity of the air and water that leaves the building. All effluent paths are monitored. The monitored paths are exhaust air (gaseous and particulate), secondary coolant (beta-gamma) and sewer discharge (beta-gamma).

All area and effluent radiation monitors alarm in the control room. The effluent monitors can initiate a confinement building isolation automatically in response to high radiation levels detected by the radiation monitors in the ventilation system. The criteria for facility area radiation monitoring and secondary coolant activity monitor requirements are specified in TS 3.7.1. Periodic testing and calibration requirements for the area radiation monitors (ARMs) and secondary cooling water activity monitor are provided for in TS 4.7.1. Exhaust air monitor requirements are specified as part of the reactor safety system channel TSs 3.2.2 and TS 4.2.2, and operability tests of the confinement enclosure system prescribed in TS 4.4 includes annual testing of the exhaust radiation monitors with a radiation source.

7.8 Conclusions The staff has reviewed the design of the instrumentation and control systems, as described in the NBSR SAR, and concludes the systems and TSs are adequate to support normal reactor operation and to achieve safe reactor shutdown upon detection of abnormal conditions. The RCS and the nuclear and process instrumentation are sufficient to provide for safe control of reactor power and monitoring of reactor safety parameters. The RPS is adequate for maintaining operation within the LSSS and the ESF actuation systems are sufficient to respond to abnormal conditions for mitigation of the consequences of postulated accidents. The licensee has shown that all nuclear and process parameters important to safe and effective operation are adequately displayed at the Main Control Panel, and sufficient radiation monitoring is provided to detect abnormal radiation levels and prevent excessive radiation exposure to personnel or release to the environment.

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8 ELECTRICAL POWER SYSTEMS 8.1 Normal Electrical Power Systems The NBSR electrical distribution system is designed to supply the electrical power necessary to operate the NBSR during both normal and shutdown conditions by providing adequate sources of power to all of the equipment and instrumentation necessary for reactor operation. Electrical power is supplied to the NBSR by three independent, underground, 13.8 kilovolt (kV) primary feeders. Loss of any single feeder will not interrupt operation of the NBSR. Only one input feeder is required to support a normal shutdown of the reactor. Power from the primary feeders is not required to achieve and maintain safe shutdown conditions.

The redundancy and the protective scheme of the electrical distribution system prevent any single failure from causing loss of off-site power. The input switchgear is arranged in a linear configuration and is divided into three sections. Each section can be separately powered from its input transformer. The normal alignment has the three sections cross-connected through electrically operated tie-breakers.

Each of the electrical loads associated with the normal and emergency operation of the reactor are powered from one of the two parallel sets of motor control centers. The reactor does not require electrical power to shutdown and a loss of electrical power will cause an automatic shutdown of the reactor.

8.2 Emergency Electrical Power Systems The NSBR Emergency Electrical Power Distribution System is designed to provide emergency power should a complete loss of off-site power occur.

Each of the two 150 kW diesel-powered alternating current (AC) generators is capable of supplying emergency power to all necessary emergency equipment. Each diesel engine set is connected to a single bearing AC generator rated at 150 kW, 226 amps with a 0.8 power factor, 480 volts, 3 phase, and 60 cycles. Each control panel is equipped with automatic voltage regulator, voltage regulator switch, manual field rheostat control, generator voltmeter, generator ammeter, frequency meter, and an AUTO/STANDBY selector switch. TS 3.6, Emergency Power System, requires that at least one diesel generator be operable during reactor operation.

TS 4.6, Emergency Power System, requires that each diesel generator be tested for automatic starting and operation at least monthly and under a simulated complete loss of off-site power at least annually. Should one of the diesel generators become inoperative, the second diesel generator is start tested at least weekly.

The 125-volt direct current (DC) station battery is capable of independently supplying the emergency loads for a minimum of four hours. The station battery consists of sixty, 2-volt, lead acid type batteries with a capacity of 880 ampere-hours. This capacity allows the supply of the DC bus loads, which total approximately 100 amperes, for eight hours. If AC power is lost to the input of the on-line uninterruptible power supply (UPS), the trickle charge of the station battery ceases and the battery automatically supplies the loads on the DC distribution panel directly and the critical power panel loads indirectly through the inverter of the UPS. When AC power is restored, either from the diesel generator or from another source, the UPS rectifier automatically resumes charging the battery and the UPS automatically resumes supplying power to the critical power panel and the DC distribution panel. In order to ensure a safe shutdown condition and to 8-1

give an adequate response in emergency situations, the following essential loads are powered by the DC bus:

• DC powered shutdown cooling pumps,

• DC powered emergency ventilation system fans and controls,

• Valve control power,

• Emergency lighting,

• UPS Inverter,

• Reactor rod control,

• Reactor process instrumentation,

• Nuclear instrumentation.

TS 3.6 requires that the station battery be operable during reactor operation. TS 4.6 requires that the voltage and specific gravity of each cell of the station battery is tested annually and a discharge test of the entire battery be performed once every five years.

8.3 Conclusions The NRC staff reviewed the design of the electrical power system, as described in the NBSR SAR, and concludes the systems and TSs are adequate to support normal reactor operation and to achieve and maintain safe reactor shutdown under all abnormal operating conditions.

The electrical power system is sufficient to provide power to all equipment loads required for reactor operation and instrumentation needed for safe control of reactor power and monitoring of reactor safety parameters.

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9 AUXILIARY SYSTEMS 9.1 Heating, Ventilation, and Air Conditioning Systems Heating, Ventilation, and Air Condition (HVAC) Systems are described in Section 9.1 of the NBSR SAR. The intake air-supply for the normal ventilation system includes both fresh air from outside and recirculated air within the building, while for the emergency ventilation system the air interior to the Confinement Building is recirculated after being appropriately filtered. All effluent air which is exhausted from the Confinement Building is monitored for radioactivity. The monitoring of radioactive effluents is discussed in Chapter 7 and Section 11.2 of this SER.

Each of the effluent pathways have filter banks which are monitored for particulate activity, air samples are withdrawn for counting of gaseous activity, and a monitor measures activity in the stack at the point of release to the environment. In the event high radiation levels are detected, the normal ventilation system will be shutdown, all building closure devices will be sealed, and the emergency ventilation system will be activated.

Three separate exhaust systems operate during normal operation of the reactor. The Normal Exhaust System takes air from those areas supplied by conditioned air and combines with the exhaust air from fume hoods. The Reactor Basement Exhaust System draws air from the process equipment area. Finally, the Irradiated Air Exhaust System takes air from potentially contaminated areas. Air from each of these systems passes through similarly designed HEPA type filtering systems. The air is then released through the stack after being appropriately diluted and monitored for an acceptable level of radioactivity.

The NBSR contains several other HVAC Systems that are routinely in operation to provide a comfortable environment for personnel and equipment. These include the Pump Room HVAC System, the Cold Laboratory HVAC Systems, the Warm Laboratory HVAC Systems, and other miscellaneous office and laboratory ventilation systems. Each of these HVAC systems brings in fresh air from outside the building and has associated exhaust fans. The individual systems vary the amount of outside air based on environmental conditions and time of year. The failure of any of these systems will not affect the emergency safety features of the NBSR. TS 3.5 requires that the emergency recirculation system and emergency exhaust systems are both operable prior to reactor operations.

Based on the above discussion, the staff concludes that the HVAC systems are adequate to maintain conditions conducive to reliable reactor operation, including instrumentation and equipment temperature control and operator comfort. Additionally, the staff concludes that the ventilation system design and controls are adequate to control the release of radioactive materials during normal reactor operation and abnormal facility conditions.

9.2 Handling and Storage of Reactor Fuel Handling and storage of reactor fuel is described in Section 9.2 of the NBSR SAR. The fuel handling system for the NBSR allows used fuel assemblies to be removed from the core and moved to the fuel storage pool. It also allows fuel assemblies to be moved to any fuel location in the reactor, and is used for the addition of new fuel assemblies to the reactor.

Movement of fuel assemblies occurs entirely beneath the top shielding plug. Each fuel assembly location is located beneath a pickup head. The pickup heads can extend down and engage the top of the corresponding fuel assembly. Locking slots in the pickup heads engage pins in the top of the fuel assembly and ensure that the fuel assembly will not be dropped. The 9-1

pickup head is then raised, removing the fuel assembly completely from the core. The fuel assembly can then be engaged by one of ten fuel transfer arms. Once a fuel assembly has been latched by one of the transfer arms, the pickup head can be released. The fuel assembly can then be moved over any fuel location in the core by transferring the attached fuel from one transfer arm to another. Once over the desired fuel location, the appropriate pickup head is engaged, the transfer arm is moved out of the way, and the pickup head can lower the assembly into the core. TS 3.9.2, Fuel Handling, requires that fuel element latching is verified prior to further fuel movement. TS 3.9.2 also requires that a cooling time in hours equal to the operating power level in megawatts elapse prior to movement of fuel from the reactor. TS 6.1.3, Staffing, requires the presence of a senior reactor operator (SRO) in the facility whenever fuel is being moved within the reactor vessel.

One of the locations that the fuel transfer arms can reach is the fuel transfer chute. This chute extends upward to the top of the reactor shielding and is the pathway for new fuel assemblies to enter the core. The shielding plug over the transfer chute is removed and new fuel assemblies are lowered until they can be engaged by the fuel transfer arms. A cylinder is also lowered to minimize the surface area of D2O moderator that is available for evaporation during new fuel additions.

The fuel transfer chute also extends down to the lower levels of the reactor building where the spent fuel pool is located. To lower a fuel assembly, a hydraulic receiver is raised from the bottom of the transfer chute to a position where it can accept a spent fuel assembly. It is then lowered to the level of the spent fuel pool and rotated to a horizontal orientation for entry into the spent fuel pool. The spent fuel assembly will heat up once removed from the moderator and stabilize at a temperature that depends upon the operating history of the reactor and the amount of time since the reactor was shut down. A time limit of one hour for every megawatt of final power level prior to shutdown is required by TS 3.9.2 to limit the temperature that a fuel assembly may reach during transfer to the spent fuel pool. As the spent fuel pool contains H2O, the assembly is allowed to dry during the passage through the fuel transfer chute. As a further precaution, the chute is isolated from the reactor prior to its entry into the H2O of the spent fuel pool, as this will generate steam. Auxiliary cooling is available should a fuel assembly hang up in the transfer chute.

Fuel in the spent fuel pool is stored in specified locations along the edge of the pool. Each full-length fuel assembly is hung by its pickup head. Boral is located along the back wall and extends out from the wall between each fuel element. Latching mechanisms preclude placing any fuel assembly closer than the 3-inch minimum separation required for criticality safety.

For shipping, fuel assemblies are cropped to remove non-fuel portions. The trimmed fuel lengths are temporarily stored in special racks that contain Boral and are spaced for a critically safe configuration. TS 3.9.1, Fuel Storage, requires that fuel elements are stored in conditions that will not exceed a k-effective of 0.90 under optimum conditions of moderation and reflection.

TS 3.9.1 also specifies that water chemistry, level, and temperature in the spent fuel storage pool shall be maintained to ensure the integrity of the fuel elements.

The staff concludes that the fuel storage facility design, fuel handling procedures, and TSs requirements provide adequate measures to preclude inadvertent criticality and unauthorized fuel movement, and to minimize the risk of mechanical or chemical damage to the fuel during movement and storage.

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9.3 Fire Protection Systems and Programs The NBSR confinement building is constructed of steel and concrete. Most of the interior structures are made of fire-resistant materials, which limit the amount of combustible materials in the facility. Inventories of transient flammable materials are minimized. In the event of an upset condition caused by a fire, the shim arms would drop into the core by gravity with spring assist shutting down the reactor.

The building contains automatic fire detection systems supplemented by manual pull boxes throughout the facility which were observed during a site visit. Hose stations and fire extinguishers are also located throughout the facility. According to the NBSR SAR, all fire detection and alarm systems tie into the NIST fire system. This system is annunciated in the control room and includes visual and audible alarms throughout the reactor facility. The alarms are also sent to the NIST fire department which is onsite and operational 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day, seven days a week. The NIST fire department meets the National Fire Prevention Association requirements for fire fighting training and also has cooperative agreements with local fire departments. The NIST fire department also performs periodic inspections of fire extinguishers and hydrants at the reactor facility.

Based on these observations and the above discussion, the staff concludes that adequate measures are in place to prevent and mitigate fire at the NBSR, and that fire damage does not pose a significant threat to the safe operation or shutdown of the reactor.

9.4 Communication Systems The NBSR contains multiple systems for communication between the control room and the rest of the facility. These include the NIST telephone system, the page phone system, and sound-powered phones. The page phone system is connected to a speaker system that allows announcements to the entire building and can be accessed through the NIST telephone system.

A separate microphone/speaker system is part of the refueling system that connects the refueling pool, the control room, and the reactor top. Also available are handheld radios for portable communication. NIST also has an emergency alarm system tied into speakers throughout the building.

Based on the above, the staff concludes that adequate communication systems are in place at the NBSR to convey information between reactor operators, and facility personnel during both normal operations and abnormal conditions.

9.5 Possession and Use of Byproduct, Source, and Special Nuclear Material Aside from the fuel assemblies, byproduct, source, and special nuclear material at the NBSR is licensed under NRC License SNM-362 and is not discussed further here. An AmBe neutron startup source of approximately 2 Curies is kept in a shielded container in the reactor source storage room.

9.6 Cover Gas Control in Closed Primary Systems The Cover Gas Control and Processing System is described in Section 9.5 of the NBSR SAR.

The NBSR has helium cover gas system for tanks and vessels that contain D2O. These tanks are the reactor vessel, the D2O storage tank, the emergency cooling tank, and the purge tank.

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The cover gas system includes chillers, gas pumps, a helium recombiner, and associated instrumentation. The system maintains appropriate pressure in the cover gas and contains charcoal cartridges that collect radioactive gases rather than allow their release to the process room. The helium cover gas system also contains a cold trap for the removal of water vapor.

Makeup helium is supplied by six bulk helium tanks. A separate back supply of eight cylinders is available.

Based on this information, the staff concludes that the Cover Gas Control and Processing System is adequate to prevent the intrusion of ambient air into primary coolant systems and minimize radiation doses to facility personnel from the generation of Ar-41.

9.7 Other Auxiliary Systems 9.7.1 Carbon Dioxide System The NBSR has a carbon dioxide gas system for purging air from the cavity between the reactor vessel and the thermal shield. The pneumatic transfer tube is also purged and filled from the carbon dioxide system. Gas is supplied from a bulk storage tank located outside the Confinement Building. Elimination of normal air from areas containing significant neutron flux minimizes the production of Ar-41 from reactor operation.

9.7.2 Instrument Air System The NBSR has an instrument air system that supplies air at 100 psig to loads within the facility.

The system supplies air to pneumatic valve operators, ventilation control valves, air ejectors, and other loads within the Confinement Building, guide hall, pump house, and laboratory spaces. The instrument air system is required for proper operation of the confinement isolation doors, HVAC damper controls, and other systems throughout the facility.

Instrument air is supplied from the main NIST compressed air facility. Air receiver tanks are used to ensure that required air is available when needed even if the main supply is interrupted.

The actuators for the dampers and valves that are required for confinement isolation each have individual air receivers. Should the main supply of instrument air be interrupted, two air compressors are available in the facility that are powered from the diesel-backed emergency power buses. The pressure in the instrument air system is noted on a gauge in the control room and an annunciator is actuated if the air pressure drops below 85 psi. The two standby compressors will automatically start at 90 psi and 80 psi, respectively.

9.7.3 Plant Chilled Water System The plant chilled water system supplies water in support of experimental facilities and also to climate control equipment. The main NIST chilled water facility supplies chilled water to the NBSR.

9.7.4 Fuel Storage Pool Cooling System The Fuel Storage Pool Cooling System is designed to remove heat from fuel assemblies stored in the pool and also to remove particulate matter from the water to maintain water clarity. The pool is nominally filled with 114,000 l (30,000 gal) of demineralized water and makeup water is supplied from the water treatment system. The water is maintained at approximately 10 °C (50 °F).

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The system has a maximum heat transfer capacity of 131.9 kW to the Chilled Water System.

Water is circulated using two centrifugal pumps controlled from the control room. A mixed-bed ion exchanger maintains the purity and clarity of the pool water. A low water level in the pool will cause an alarm in the control room.

9.7.5 Thermal Shield Cooling System The Thermal Shield Cooling System is designed to remove heat from the Thermal Shield and also to maintain water clarity. Two centrifugal pumps are available to circulate the cooling water to a heat exchanger rated at 762 kW of heat removal. The system includes a storage tank that is used to maintain head for the pumps. Two mixed-bed ion exchangers maintain the purity and clarity of the cooling water. The system has extensive instrumentation including monitoring of system temperature, flow, and storage tank level. Shim arm rundown is caused by low flow in the system.

9.7.6 Thermal Column Tank Cooling System The Thermal Column Tank Cooling System removes heat from the bismuth shield and aids in the thermalization of neutrons in the thermal column. While the system uses 908 l (240 gal) of D2O, it is completely independent of the primary D2O system. Flow is provided by two pumps which circulated the D2O through a heat exchanger and a filter system. Low flow indication will result in a rundown of the reactor shim arms and cause an alarm. A helium bottle maintains a cover gas on the system.

9.7.7 Experimental Demineralized Water Cooling System The Experimental Demineralized Water Cooling System supplies cooling water at each experimental facility. It also provides water pressure for the refueling cannon. The system contains a heat exchanger and ion exchanger and contains a storage tank of 9,085 l (2,400 gal).

9.8 Conclusions Based on the above discussions, the staff concludes that the Auxiliary Systems at the NBSR support safe operation of the facility, and aid in the safe shutdown of the reactor. Further, the TSs ensure that fuel elements are appropriately handled and that there is no significant risk to the health and safety of the public from the storage and movement of fuel.

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10 EXPERIMENTAL FACILITIES AND PROGRAMS 10.1 Summary Description The NBSR serves as a source of radiation for use in the research programs of the NCNR. It provides a unique radiation source that allows studies that are currently not possible at other reactor facilities in the United States. The experiment types range from fundamental research on particle properties to applied questions of technology development. Experiments are performed by both university and commercial researchers. The TSs provide limitations for the effect on reactivity of all experiments and means for technical and safety review of experiments.

10.2 Experimental Facilities Experimental facilities include:

• Nine radial beam tubes.

• Two through tubes that pass completely through the reactor.

• A cold neutron source that supplies seven neutron guides in to an adjacent building.

• A 54-inch by 52-inch thermal column.

• Four pneumatic transfer thimbles.

• Eighteen vertical thimble locations (both core and reflector locations).

10.2.1 Radial Beam Tubes Section 10.2.1 of the NBSR SAR describes the design and construction of the nine radial beam tubes at the NBSR. The beam tubes range in size from 5 inches to 6 inches and have a thermal neutron flux at the core end of the beam holes of approximately 2.0 x 1014 n/cm2-sec during 20 MW(t) power operations. An aluminum plate seals the region between the reactor and the thermal shield. Each beam port has an 18-inch thick shutter consisting mostly of lead that provides shielding during reactor shutdown for removal and insertion of plugs and collimators outside of the shutter.

10.2.2 Through Tubes Section 10.2.2 of the NBSR SAR describes the design and construction of the two 4-inch tubes which pass completely through the reactor just below the core and off-center. One end of each tube has a shutter and the other end contains a plug.

10.2.3 Cold Neutron Source Section 10.2.3 of the NBSR SAR describes the design and construction of the cold neutron source which consists of a sealed reservoir of liquid hydrogen next to the core that lowers the energy of emerging neutrons. These lower-energy or cold neutrons are then transported through guides to instruments in an adjacent building.

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The heat absorbed by the hydrogen and its aluminum containment vessel is removed by the boiling of the liquid hydrogen. The hydrogen is then re-liquefied by a closed cycle helium gas refrigeration unit. The entire hydrogen system is designed to ASME code for pressure vessels with a maximum design stress of 6,000 psi, which is many times the maximum design working pressure.

All hydrogen components are surrounded by helium gas at higher than atmospheric pressure to preclude the introduction of oxygen with the hydrogen.

10.2.4 Thermal Column Section 10.2.4 of the NBSR SAR describes the design and construction of the thermal column which allows for irradiation of large specimens. Radiation from the core first passes through a bismuth gamma ray shield which is cooled by D2O. The remaining neutrons then travel through a 37-inch thick collection of graphite blocks that has a total cross-section of 54 inches by 52 inches. Some of the graphite blocks are removable. Access to the graphite for irradiation is available through a vertical hole above the graphite or through the face of the graphite. A Boral curtain is present that acts as a thermal neutron beam shutter. When not being used, the thermal column is covered by a shield door.

10.2.5 Pneumatic Tube System Section 10.2.5 of the NBSR SAR describes the design and construction of the pneumatic transfer system which is available to insert samples to four locations within the reactor vessel.

The sample containers are commonly known as rabbits, and have approximately 1-inch inner diameter. Samples can be sent into the reactor for irradiations lasting from a few seconds to several days. The pneumatic tubes send and receive rabbits from radiological hoods in a radiological laboratory in the reactor basement.

10.2.6 Vertical Thimbles Section 10.2.6 of the NBSR SAR describes the design and construction of Vertical Thimbles.

There are a number of irradiation positions available in both the core and in the reflector for experimental use. There are seven 3.5-inch thimble locations that are available, each of which has approximately 8 gpm of coolant flow. One of the seven locations is used for the regulating control rod. At the 3.5-inch thimble locations, the fast flux is depressed and the thermal flux is enhanced. There are four 2.5-inch thimble locations in the core, each cooled by approximately 10 gpm of coolant flow. The 2.5-inch thimble locations are in closer proximity to the fuel than the 3.5-inch thimble locations and have a higher fast flux. An additional seven thimble locations for experiments up to 3.5-inch are available in the reflector. While additional cooling can be provided for the incore thimble locations, no supplementary cooling can be provided to the reflector locations.

10.3 Experiment Review Proposals for the introduction of any new experimental instruments involving the NBSR are required to be submitted in writing to the NCNR Safety Evaluation Committee (SEC) in accordance with TS 6.2.3, SEC Review Function. Among other information, the proposals are required to include instructions on how to shut down the instrument and place it in a safe configuration. These instructions must be posted by the instrument during use and also provided to the reactor operations staff.

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Membership of the SEC is specified by TS 6.2, Review and Audit, and must include at least 2 members of the NCNR and one member from Health Physics. All members of the SEC are senior technical personnel.

Use of experimental facilities is reviewed and approved through the NCNR Safety Representative and the Hazard Review Committee. The Hazard Review Committee is composed of members from the scientific, technical and reactor operations staff, and includes at least one representative from the Health Physics Group, one representative from the SEC, and from Reactor Operations.

Use of experimental facilities requires a written proposal. The experiment proposals must include descriptions of the experiment in detail and the types of samples to be studied, and a description of the potential risks. The Safety Representative may authorize the use of materials that are within the approved safety envelope. The Hazard Review Committee is responsible for review of all materials that are to be introduced that are outside of the existing safety envelope.

No hazardous material may be introduced into the Confinement Building without the review and approval of the Director of the NCNR.

TS 3.8, Experiments, also includes specific criteria for evaluation and approval of experiments at the NBSR. These include a limit on the change in reactivity of 0.5% for any experiment and a limit on the sum of the absolute values of all experiments of 2.6% . These reactivity values are within the reactivity insertion rate limit assuming at least 0.5 seconds for removal of experiments. As the time to remove experiments from the pneumatic tube is less than 0.5 seconds, the total reactivity for all pneumatic tubes is 0.2 % . Additional requirements of TS 3.8 are that experiment malfunctions must not cause malfunctions in other experiments and reactor transients must not cause experiments to fail so as to contribute to accidents.

Containment measures are specified for explosive or metastable materials as well as for experiments containing materials that are corrosive to reactor components or reactive with reactor coolants.

The specific review criteria, along with the experiment review process, provide confidence that any experiment performed at the NBSR, can be performed safely and any unanticipated radioactive releases will be within 10 CFR Part 20, Appendix B limits.

10.4 Conclusions The staff concludes that the review process for experiments and use of experimental facilities provides high confidence that appropriate precautions are taken to minimize the risk to personnel from unintended radiation exposure. Further, the staff concludes that the review process provides reasonable assurance that the use of experiments or experimental facilities will not damage the fuel and thus not pose a significant risk to the health and safety of the public or licensee personnel, or the environment.

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11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT 11.1 Radiation Protection 11.1.1 Radiation Sources The primary source of radiation at the NBSR is the reactor itself. The reactor core is surrounded by deuterium moderator, and steel and concrete shielding to reduce direct radiation doses to surrounding areas.

During normal operations, the reactor generates neutrons for a number of research purposes.

A thermal column is available for large cross-sectional exposures, and when used, is typically controlled as a High Radiation Area. Beam ports for cold neutron experiments allow low-energy neutron to pass from the reactor through shielded neutron guides and pass through a 25-foot thick concrete wall into the Guide Hall. Each of the seven neutron guides has a keyed shutter along with status indication. The reactor also has a pneumatic transfer system for in-core experiments which can create radioactive materials.

As described in Chapter 11 of the SAR, the airborne radioactive materials generated during reactor operation of principal concern are 41Ar and 3H. Argon is a natural component of the atmosphere and becomes activated to 41Ar upon neutron bombardment. Minimization of 41Ar production is accomplished by surrounding the core structure with a helium blanket and conducting activities such as maintenance in ways that minimize air intrusion into volumes subjected to neutron flux. 3H is generated by the neutron bombardment of deuterium (2H) in the reactor moderator and builds up during reactor operation. The primary moderator is periodically changed out to maintain the concentration of 3H in the D2O to a level of 1 Ci/liter. TS 3.7.1 limits the concentration of 3H to 5 Ci/liter. TS 4.7.1 requires annual surveillance of the primary activity when the concentration is below 4 Ci/liter, and quarterly sampling otherwise. The action to keep primary coolant activity at 1 Ci/liter or less reduces airborne 3H levels from D2O releases and is an ALARA measure.

According to the Chapter 11 of the SAR, airborne concentrations of radioactive materials are significantly less than 1 DAC and calculated dose rates from 41Ar for typical occupancy times result in an annual dose rate to facility personnel of 0.02 mSv per year (mSv/yr) (2 mrem/yr). 3H doses to facility personnel average 0.4 mSv/yr (40 mrem/yr) or less for most personnel.

Reactor operators involved in activities that involve exposure to D2O such as refueling experience higher annual doses but normally not in excess of 1 mSv/yr (100 mrem/yr) and well within 10 CFR Part 20 limit of 50 mSv/yr (5,000 mrem/yr).

Liquid radiation sources at the NBSR consist primarily of activation products of the coolant and reactor components, principally 16N and 3H. 16N has a 7-second half-life and is only a radiation hazard during reactor operations or immediately after reactor shutdown. The primary coolant piping that presents an external radiation hazard is identified and shielded. 3H is a low-energy beta emitter and is not an external radiation hazard. Additional radionuclides are created in the primary coolant, but during operation, are less of a concern than the 16N. Dose rates in the vicinity of primary coolant piping during full power operation can be up to 60 rem/hr, though the dose rates drop to a few mrem/hr in shielded areas of the Process Room.

Radionuclides in liquids are also present in the Thermal Column D2O Tank Coolant and the Thermal Shield Cooling System due to neutron activation of the coolant and structural materials though at much lower activity levels than in the primary coolant. Small quantities of 11-1

radionuclides in liquids may also be present in the reactor secondary coolant and the Fuel Storage Pool from leakage or contamination from other systems. The secondary coolant system is monitored as required by TS 4.7.1 to detect any leakage through heat exchangers from the primary system.

Solid radioactive materials are generated by NBSR operations. Chief among these are the spent fuel assemblies. After irradiation in the core, the spent assemblies are moved to the spent fuel storage pool. Spent fuel movement and storage is discussed in Section 9 of this SER. Reactor shims are also activated after time in the reactor. These are also stored in the spent fuel storage pool and shipped offsite with non-fuel element metal pieces. The shim bodies may be stored in shielded dry storage wall cavities. Other solid radioactive sources include reactor resins and filters, shielding plugs, neutron beam shields, experiment components from high flux location and activated samples. Solid radioactive waste is disposed of in accordance with appropriate NRC regulations and is transferred to organizations authorized to receive the material.

Instrument check sources and calibrations sources are also used at the reactor facility, but are licensed under a separate Byproducts Materials License (SNM-362), and not by the reactor license.

TS 3.8.2, Materials, controls introduction of materials into the reactor as part of experiments and requires evaluation to ensure that appropriate safeguards are used for control of irradiated materials.

The staff concludes that the description and characterization of the radiation sources at the NBSR is reasonable for a test reactor of this type and size and that this information provides sufficient information to evaluate the radiation protection program and controls described in the remainder of this section of the SER.

11.1.2 Radiation Protection Program Chapter 11.1.2 of the NBSR SAR summarizes the radiation protection program required by 10 CFR 20.1101. This program includes the stated policy to employ the ALARA concept in all operations at NBSR, and that all radiation exposures are well within regulations and guidelines.

The responsibility for administering the radiation protection program at the NBSR belongs to the reactor health physics section as required by TS 6.3, Radiation Safety. This group is part of the NIST Health Physics Group and is in a separate reporting chain from the Director of the NCNR which includes reactor operations. A reactor senior health physicist oversees the activities of the reactor health physics section and is responsible for the implementing the radiation protection program at the NBSR. Normal staffing for the section is 2 to 4 Health Physicists and 2 to 4 Radiation Protection Technicians.

The responsibilities of the reactor health physics section include calibration of survey instruments, effluent monitoring, radiation and contamination surveys, training, sample analysis, and personnel monitoring. TS 6.2.1, Composition and Qualifications, requires a health physics representative serve on the SEC to review new experiment equipment installations.

Procedures for implementing the radiation protection program are required by TS 6.4, Procedures, and are reviewed by both operations and health physics personnel prior to 11-2

adoption. Plans and procedures not related to the NBSR are maintained under the NIST materials license (SNM-362).

Radiation safety training is given to all individuals granted unescorted access to the NBSR facility. This training covers information required by 10 CFR Part 19 including basic radiation information, signage and alarm responses, and proper use of dosimetry. Individuals who have duties involving radioactive material or reactor experiments receive additional training relating to their job duties and including proper use of survey instruments and specific regulations and procedures that applies to their work. The NBSR SAR indicates that refresher training is required every 24 months.

Records relating to personnel dosimetry or exposure investigations, as well as effluent records are retained for the life of the facility as required by TS 6.8, Records. The SAR indicates that other radiation protection related documents are retained for a minimum of 10 years.

The staff concludes that the structure and strategy of the Radiation Protection Program for the NBSR is consistent with the guidance of ANSI/ANS-15.11, Radiation Protection at Research Reactor Facilities and that the program as implemented is adequate to provide reasonable assurance that personnel are protected from radiation hazards.

11.1.3 ALARA Program As described in Chapter 11.1.3 of the SAR, NIST has established an ALARA program for the NBSR. The program provides specific emphasis on proposed experiments, planned activities that could result in significant exposures, and ambient radiation levels within NBSR. The program also includes ongoing supervisory reviews of worker and public doses to detect trends that may need corrective action.

The ALARA program includes extensive use of engineering controls to minimize facility dose rates. Shielding is utilized to reduce radiation levels to work areas. Activities above preset cumulative dose levels require a formal operating plan, pre-job meetings, identification of methods to reduce exposure and specific Health Physics staff oversight. Consistent with the ALARA program, the ventilation systems go into a recirculation mode of operation during upset or abnormal operating conditions with the operation of a standby charcoal filter.

These methods are typical for ALARA programs and the staff concludes that they provide reasonable assurance that personnel exposure to radiation will be minimized.

11.1.4 Radiation Monitoring and Surveying The Health Physics Group maintains numerous fixed and portable radiation detection instruments throughout the NCNR facility. Backup instrumentation is available at the Radiation Physics Building on the NIST site.

Section 7.7 of this SER discusses the ARMs used to alert staff and operators to changing radiation conditions. Ten fixed gamma area radiation monitors are located throughout the confinement building. These monitors have local readouts, warning lights, and alarms to alert workers in the vicinity to increased radiation levels. These ARMs also have readouts and alarms in the control room. The monitors are nominally set for 0.05 mSv/hr (5 mrem/hr) and are adjusted as needed for non-routine activities. One of the monitors is located in the spent fuel storage pool area and also serves as a criticality monitor, as does the monitor in the new fuel 11-3

storage area. TS 3.7.1 requires at least two ARMS be operable on floors C-100 and C-200 for the reactor to be operated.

Other fixed radiation monitors are used for detection of personnel contamination and include hand and foot monitors and portal monitors. These contamination monitors are located at the entrance to the reactor building and other locations, as needed.

Radioactive effluents released through the plant stack are monitored by a Geiger-Mueller detector as required by TS 3.7.1. The detector readings are calibrated by comparison with volumetric samples taken from the stack flow. 3H effluent is monitored by the building 3H monitoring system. The readings are compared with monthly samples. Additional monitoring is performed on an as-needed basis to support non-routine activities.

Portable instrumentation is available to survey areas in the NBSR facility for all types of radiation and radioactive contamination that may be present from facility operations in accordance with the requirements of 10 CFR 20.1501. The different instruments available can measure alpha, beta, gamma, and neutron radiation. These detectors cover a range from 0.2x10-3 mSv/hr (0.02 mrem/hr) to 10 Sievert/hr (1,000 rem/hr). Surveys using portable instrumentation are performed throughout the facility at least weekly. Survey frequency is adjusted during non-routine activities to enhance detection of changing conditions.

An installed gas-flow ion chamber system monitors the ventilation system in the reactor building for 3H. Additional air monitoring equipment is available for use as needed in the facility. This includes both particulate and iodine monitoring as well as 3H monitoring. Detection levels are a small fraction of applicable DAC levels.

The staff concludes that the installed and available radiation detection equipment and the program to use the equipment, satisfy the requirements of 10 CFR 20.1501(a)(b) and provide reasonable assurance that doses to personnel will be kept below the limits specified in 10 CFR 20.1201.

11.1.5 Radiation Exposure Control and Dosimetry The NBSR is located on the NIST Gaithersburg site which is a Controlled Area as defined by 10 CFR 20.1003. The site is fenced and access is limited. The NCNR building, which houses the NBSR is controlled as a Restricted Area. Access to the building requires training and a coded ID card for entry. Use of radioactive material within the facility is limited to designated areas.

According to the Chapter 11 of the SAR, all personnel accessing the NBSR are monitored for external radiation. The facility staff is issued record dosimeters, while temporary personnel may be issued pocket ion chambers as an alternative. Typically less than 20 personnel meet the criteria for which monitoring is required. Individuals that could exceed 5 mSv (500 mrem) annually are issued a pocket ion chamber in addition to their regular dosimeter. This practice allows for dose trending between quarterly processing of their regular dosimeter. Additional surveillance is performed for declared pregnant workers. Personal dosimeters worn by personnel at the NBSR are provided by a National Voluntary Laboratory Accreditation Program (NVLAP)-certified supplier as required by 10 CFR 20.1501(c). Extremity dosimeters are available if needed. This monitoring program meets the requirements of 10 CFR 20.1502 and is consistent with the guidance of ANSI/ANS-15.11.

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Internal monitoring is not normally required at the NBSR. 3H monitoring for selected personnel is performed as an ALARA measure. A whole body counting facility is available if needed for assessing potential internal uptakes. Respiratory protection devices are not used at the NBSR for limiting radiological uptakes. However, respirators are used for non-radiological purposes, such as pipe welding and cutting.

The principal concern for exposure control is the neutron beams both in the vicinity of the reactor and in the guide hall. Of particular concern is the area around beam experiments.

These experiments are reviewed prior to operation to minimize the radiation from activated materials, and to minimize the radiation levels surrounding the equipment. The vicinity of many of the experiments are posted as Radiation Areas. ALARA considerations are also included in experiment design.

For routinely accessed areas at the NBSR, the design requirement for long-term experiment shielding is less than 0.05 mSv/hr (5 mrem/hr) at one foot, with an ALARA goal of 10% of this value. Radiation levels of up to 1 mSv/hr (100 mrem/hr) are shielded as practicable, but allowed provided that impact is minimal. Additional controls are used for High Radiation Areas as required by 10 CFR 20.1601, including additional postings, physical barriers, and warning devices.

Based on the above, the staff concludes that the exposure control and dosimetry program at the NBSR is adequate to monitor and control exposures to personnel below 10 CFR Part 20, Subpart C limits.

11.1.6 Contamination Control Contamination control at the NBSR is accomplished through engineered controls and an aggressive program to eliminate any contamination below detectable levels. Hoods and glove boxes are used to contain potential contamination. Area surveys for contamination are preformed at least weekly. Any contamination found is cleaned to non-detectable levels, with the exception of contamination in non-traffic areas of higher external radiation that would be of minor benefit and inconsistent with the ALARA program.

11.1.7 Environmental Monitoring Environmental monitoring is performed by the licensee to verify that doses to the public are less than 10 CFR 20.1301 requirements. Chapter 11.1.7 of the NBSR SAR concisely describes an environmental monitoring program that includes radiation surveys, soil and vegetation sampling, and surface water sampling in the vicinity of the NBSR. Environmental samples are evaluated for activation products and fission products. Water samples are evaluated for 3H. The NBSR SAR indicates that samples are taken quarterly at a minimum of four locations for each sample type. Environmental samples are further supplemented by radiation readings from thermoluminecent dosimeters or other radiation detection devices appropriately placed at the NIST site boundary. The methodology for the collection and analysis of environmental samples is appropriate for determining compliance with 10 CFR 20.1301 requirements.

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11.2 Radioactive Waste Management 11.2.1 Radioactive Waste Management Program The NBSR radioactive waste management program is described in Chapter 11 of the NBSR SAR. The program is assigned to the Health Physics Group with a health physicist assigned primary oversight of the program. As described in the NBSR SAR, the program includes waste stream characterization.

The SAR describes a policy of radioactive waste minimization consistent with ALARA objectives. Waste minimization practices are used throughout the facility to minimize disposal costs. These practices include the use of materials with low neutron activation potential and through design of experiments to be reused and that minimize both activation and potential for contamination. Efforts to minimize radioactive waste also include disassembling and segregation so that only activated or contaminated materials enter the waste stream.

NBSR also has available 33 shielded concrete cavities for long-term storage of radioactive materials that have potential future use. These cavities are also used to allow radioactive decay of materials prior to disposal.

Sections 11.2.2 and 11.2.3 of this SER address characterization, monitoring, and release of wastes. TS 6.8 requires that records of gaseous and liquid effluents released to the environment be retained for the life of the facility and that records of solid radioactive waste shipped offsite be retained for a minimum of five years.

11.2.2 Radioactive Waste Controls As described in the SAR, radioactive waste at the NBSR is first controlled through waste minimization efforts to limit initial generation of radioactive material. These efforts also are designed to preclude the production of mixed waste. Identification of radioactive waste is accomplished either by process knowledge or by direct survey. For activation materials that are not easily surveyed, calculations are performed based on the materials present and the neutron irradiation characteristics.

Waste materials are collected at the point of generation in marked waste containers. The waste materials are then transported to another location of the NIST grounds for final disposition.

Surveys of potential waste material are performed in a low-background environment using the appropriate detectors for the type of radiation emitted. A commercial HEPA-filtered waste compactor is used to minimize waste volume.

Liquid radioactive waste is sampled and analyzed to confirm that waste released to the sanitary sewer meets 10 CFR 20.2003 requirements for solubility and concentration.

Gaseous radioactive effluent is monitored through a Geiger-Mueller detector on the plant stack flow path. In the event of upset or abnormal conditions, the ventilation system can be set to recirculation and a standby charcoal filter placed into operation. TS 3.7.1 requires that effluent monitoring be operable for operation of the reactor. TS 3.7.2, Effluents, limits effluent releases such that the total exposure to any person at the site boundary shall not exceed 1 mSv (100 mrem) per calendar year, excluding external dose from the facility. These controls provide 11-6

reasonable assurance that offsite doses to the public will be below the limits specified by 10 CFR Part 20.

11.2.3 Release of Radioactive Waste Releases of liquid and gaseous waste are described in Section 11.2.2 of this SER. All solid radioactive waste is disposed of by transfer to licensed disposal sites or processing facilities. All waste is packaged and transported as required by appropriate NRC regulations and applicable state licenses of the recipient.

Based on the above information, the staff concludes that the effluent radiation monitoring program at NIST is adequate to quantify and characterize the gaseous and liquid effluents released from the facility and keep effluent concentrations below the limits of 10 CFR Part 20, Appendix B. Further, the program is sufficient to provide reasonable assurance that doses to members of the public from effluents are well below the limits of 10 CFR 20.1301.

11.3 Conclusions The staff has reviewed the Radiation Protection and Radioactive Waste Management Programs, as described in the NBSR SAR, and concludes that the programs, systems and TSs are adequate to ensure that normal facility operation will not cause an undue risk to facility personnel, the public, or the environment from radiation or radioactive materials. The Radiation Protection Program has the organization, equipment, and personnel sufficient to maintain personnel doses below the limits specified in 10 CFR Part 20 and ALARA. The Radioactive Waste Management Program provides reasonable assurance that the licensee will properly handle and disposition radioactive wastes that are generated by the operation of the NBSR.

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12 CONDUCT OF OPERATIONS The conduct of operations involves the administrative aspects of facility operation, the facility emergency plan, and facility security. The administrative aspects of facility operations are the facility organization, training, operational review and audits, procedures, required actions, and records and reports.

12.1 Organization Responsibility for the safe operation of the NBSR is vested within the chain of command shown in TS Figure 6.1. TS 6.1.1, Structure, and TS 6.1.2, Responsibility, provide details of the management requirements of the NBSR. The Chief Reactor Operations and Engineering (Chief Nuclear Engineer) is delegated responsibility for overall facility operation. The Chief Nuclear Engineer is responsible to the Director, NCNR, for safe operation and maintenance of the reactor and its associated equipment. Individuals at the various management levels, in addition to responsibility for the policies and operation of the reactor facility, are responsible for safeguarding the public and facility personnel from undue radiation exposure and for adhering to all requirements of the operating license and TSs. The Chief Nuclear Engineer delegates the succession to this responsibility during his absence.

TS 6.1.3 contains the minimum staffing requirements for the NBSR. SAR section 12.1.3 states that the minimum crew complement during a shift shall be two persons, including at least one licensed senior operator, and that during normal operations, the crew complement for a shift shall be three persons. TS 6.1.3 states that the minimum staffing when the reactor is not secured shall be: (a) A Reactor Operator in the Control Room; (b) A Reactor Supervisor present within the reactor exclusion area; and (c) An SRO present in the facility whenever a reactor startup is performed, fuel is being moved within the reactor vessel, experiments are being placed in the reactor vessel, or a recovery from an unplanned or unscheduled shutdown or a significant reduction in power. This is consistent with ANSI/ANS-15.1 guidance and 10 CFR 50.54(m)(1).

12.2 Review and Audit Activities The SEC is described in Section 12.2.1 of the NBSR SAR. The SEC provides the NCNR with a method for the independent review of the safety aspects of reactor operations. The SEC assists the Director, NCNR in evaluating reactor operational activities, improving the quality of reactor operational programs, and recommending corrective actions for problem areas.

The SEC is composed of at least four senior technical personnel who collectively provide a broad spectrum of expertise in reactor technology, e.g. nuclear, electrical and mechanical engineering and radiation protection. The SEC members are appointed by the Director, NCNR.

At least two members are from the NCNR staff and one member is from the Health Physics Group. The SEC quorum is three members. The NCNR director may appoint alternates to serve during the absence of regular members. The SEC meets semiannually during reactor operations and as circumstances warrant. Written records of the proceedings, including any recommendations or concurrences, are maintained. The SEC reports directly to the Director, NCNR. The composition of the SEC is specified in TS 6.2.1.

TS 6.2.3 specifies that the SEC reviews the following:

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• Proposed changes to the NBSR facility equipment or procedures when such changes have safety significance, or involve an amendment to the facility license, a change in the Technical Specifications incorporated in the facility license or a change pursuant to the applicable criteria 10 CFR 50.59.

• Proposed tests or experiments significantly different from any previously reviewed or which involve a change pursuant to the applicable criteria of 10 CFR 50.59.

• The circumstances of all reportable occurrences and violations of Technical Specifications and the measures taken to prevent a recurrence.

• The SEC Charter on a biennial basis and recommends changes to the NCNR director.

The Safety Audit Committee (SAC) is composed of three senior technical personnel who collectively provide a broad spectrum of expertise in reactor technology. The Committee members are appointed by the Director, NCNR. Members of the SAC are not regular employees of NIST. At least two members are required to approve any report or recommendation of the Committee. The SAC meets annually and as required. The Committee audits the NBSR facility operations and the performance of the SEC. The SAC reports in writing to the Director, NCNR.

The staff reviewed the review and audit function against the guidance of ANSI/ANS-15.1 and found the SEC role as described in the SAR and the TS to be consistent with the standard.

12.3 Procedures The licensee has developed a comprehensive set of written operating procedures for all aspects of facility operation as required by TS 6.3. These procedures address the following:

• Startup, operation, and shutdown of the reactor.

• Fuel loading, unloading, and movement within the reactor.

• Surveillance checks, calibrations, and inspections required by the TSs or those that may have an effect on reactor safety.

• Personnel radiation protection, consistent with applicable regulations or guidelines and that include commitment and programs to maintain exposures and releases ALARA.

• The conduct of irradiations and experiments that could affect reactor safety or core reactivity.

• Use, receipt, and transfer of byproduct material, if appropriate.

• Maintenance of major components of systems that could have an effect on reactor safety.

Substantive changes to procedures require documented review by the SEC and approval by the Chief Nuclear Engineer. Minor modifications to procedures that do not change their original intent may be made by the Chief Nuclear Engineer or his deputy.

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12.4 Required Actions Certain events require specific licensee actions in accordance with the SAR and TS 6.6, Required Actions.

TS 6.6.1, Actions to be Taken in the Event the Safety Limit is Exceeded, requires that if a safety limit is exceeded, the reactor is shut down and operations are not resumed without authorization by the NRC pursuant to 10 CFR 50.36 (c)(1). A report to the NRC is then required in accordance with TS 6.7.2, Special Reports. This report shall contain a complete analysis of the circumstances leading to and resulting from the situation with recommendations to prevent recurrence. Further, The SEC shall review the circumstances of all reportable occurrences and violations of TS and the measures taken to prevent a recurrence and provide recommendations for action.

TS 6.6.1 requires that if a Limiting Safety System Setting is exceeded or a Limiting Condition of Operation is violated, reactor conditions are either returned to normal or the reactor is shutdown. If it is necessary to shut down the reactor to correct the occurrence, operation is not resumed unless authorized by the Chief, Reactor Operations, or his Deputy. The occurrence is reported to the Chief, Reactor Operations, or his Deputy, and to the NRC if required by TS 6.7.2. Further, the occurrence is then reviewed by the SEC at their next scheduled meeting.

TS 6.7.2 also requires that all Reportable Occurrences are promptly reported to the Chief, Reactor Operations, or his Deputy, and to the NRC, and subsequently reviewed by the SEC.

Reportable Occurrences include the following:

• Operation with actual safety system settings less conservative than limiting safety system settings specified in Technical Specifications.

• Operation in violation of LCO, unless prompt remedial action is taken.

• An uncontrolled or unanticipated significant reactivity change.

• An uncontrolled or unanticipated significant release of radioactivity from the site.

• An engineered safety system component malfunction or other component or system malfunction which could or threatens to render the affected system incapable of performing its intended safety function.

• Major degradation of one of the several boundaries which are designed to contain the radioactive materials resulting from the fission process.

• An observed inadequacy in the implementation of major administrative or major procedural controls, such that the inadequacy causes or could have caused the existence or development of an unsafe condition with regard to reactor operation.

The staff concludes that the required actions specified in the TS are consistent with the guidance of ANSI/ANS-15.1 and 10 CFR 50.36, Technical Specifications and provide reasonable assurance that the facility will respond to unanticipated occurrences in a manner that emphasizes reactor safety and protection of public health and safety.

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12.5 Reports TS 6.7, Reports, specifies reports that the facility is required to make to the NRC. These include an annual operating report and special reports. Annual operating reports include an operational summary, including the number of unscheduled shutdowns and reasons, information regarding safety-significant maintenance activities, effluent releases, environmental surveys, safety evaluations pursuant to 10 CFR 50.59, and any significant exposures to personnel or visitors.

The staff evaluated these reporting requirements and found they were consistent with ANSI/ANS-15.1 and provide reasonable assurance that the facility will report appropriate information regarding routine operation, non-routine occurrences, and changes to the facility and personnel to the NRC in a timely manner.

12.6 Records Section 12.6 of the NBSR SAR states that in addition to those records required by applicable regulations, the licensee commits to retaining the following records for period of at least 1 year:

• Records of all safety or safety-related equipment maintenance activities.

• Violations of Technical Specifications.

• Reportable occurrences.

• Those technical and safety considerations supporting the recommendations of the Safety Evaluation Committee, including action taken responsive to such recommendations.

• Records and logs of reactor operations.

• Records of principal maintenance activities.

• Records of surveillance activities performed in accordance with Section 4 of Technical Specifications.

TS 6.8 specifies records that the NBSR is required to maintain in a manner that facilitates convenient review. The licensee specified the types of records that the facility will retain and the period of retention to ensure that important records will be retained for an appropriate time. The staff evaluated the requirements of TS 6.8 and found that they are consistent with ANSI/ANS-15.1 and applicable regulations and give reasonable assurance that the facility will maintain appropriate records to facilitate NRC inspection and provide adequate history of the facility.

12.7 Emergency Planning By letter dated September 24, 2008 (ADAMS Accession No. ML082760410), the licensee submitted a revised emergency plan (EP) as part of the license renewal application. The staff reviewed the EP against NUREG-0849, Standard Review Plan for the Review and Evaluation of Emergency Plans for Research and Test Reactors, issued October 1983; Regulatory Guide 2.6, Emergency Planning for Research and Test Reactors, Revision 1, issued March 1983; ANSI/ANS-15.16, Emergency Planning for Research Reactors, issued 1982; and NRC 12-4

Information Notice 97-34, Deficiencies in Licensee Submittals Regarding Terminology for Radiological Emergency Action Levels in Accordance with the New Part 20, issued June 1997.

The staff concluded that the NIST EP is in accordance with the guidance and regulations. A memorandum dated November 9, 2009, from the NRC Office of Nuclear Security and Incident Response to the NRC Office of Nuclear Reactor Regulation (ADAMS Accession No. ML090080818) documented this review. The licensee has demonstrated the ability to make changes to the EP in accordance with 10 CFR 50.54(q). Accordingly, the staff concludes that the NIST EP provides reasonable assurance that the licensee can respond appropriately to a variety of emergency situations and that the NIST EP will be adequately maintained during the period of the renewed license.

12.8 Security Planning While Facility Operating License No. TR-5 authorizes possession of an overall quantity of Special Nuclear Material of strategic significance, the license stipulates that less than 5.0 kilograms of this amount be unirradiated. According to the licensee and based on the reported operations of the facility, the remaining 40.0 kilograms of uranium-235 meets the exemption from the requirements in 10 CFR 73.60(a)-(e). Therefore, the licensee must maintain security measures that satisfy the requirements of 10 CFR 73.67 and 10 CFR 73.60(f), as appropriate.

The licensee meets these requirements and the NRC inspects the licensees measures for physical security and protection of special nuclear material on a routine basis. A recent inspection and site visit verified that the licensees security measures satisfy applicable regulations and are acceptable.

12.9 Quality Assurance The NBSR SAR states that the licensee maintains an established quality assurance program that is consistent with the guidance found in ANSI/ANS-15.8, Quality Assurance Program Requirements for Research Reactors, issued 1995. The quality assurance program is maintained as part of the administrative rules and procedures at the NBSR.

12.10 Operator Training and Requalification Program This area of the licensees application is currently under review by the NRC Office of Nuclear Reactor Regulation. The licensee submitted a revised operator requalification plan based on the guidance document ANSI/ANS-15.4 (ADAMS Accession No. ML083510754).

12.11 Conclusions Based on the above discussions, the staff concludes that NIST has the appropriate organization, experience levels, and adequate controls through the TS to provide reasonable assurance that the NBSR is managed and operated in a manner which will not cause significant radiological risk to facility personnel, the public, or the environment.

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13 ACCIDENT ANALYSES To help establish safety limits, LSSSs, and LCOs for the NBSR, the licensee analyzed potential reactor transients and other hypothetical accidents for the effects of such events on the reactor fuel and the health and safety of facility personnel, the public, and the environment. None of the credible accidents postulated would lead to the failure of the fuel cladding or the uncontrolled release of fission products. However, the licensee postulated an enveloping event involving the complete blockage of flow to one element, leading to complete melting of the fuel plates. This event would lead to the maximum potential radiation hazard to facility personnel and members of the public. The licensee makes no assumptions as to the cause of the failure, but evaluated only the potential consequences of this event and not the likelihood or mechanisms of its occurrence. This worst-case accident scenario has been designated as the maximum hypothetical accident. The licensee has evaluated other possible accident sequences and none pose a significant risk of cladding failure or release of fission products.

Those accidents considered for evaluation and analysis are as follows:

• Maximum Hypothetical Accident (fuel channel blockage and fuel plate melting).

• Insertion of Excess Reactivity (including Experiment Malfunction).

• Loss of Primary Coolant.

• Loss of Primary Coolant Flow (including Loss of Normal Power).

• Misloading of Fuel.

For all of the accident scenarios, the reactor is assumed to be operating with all critical parameters at the most limiting extreme value of their normal range. This ensures that the analysis for each accident scenario uses the worst case initial conditions that might be anticipated, within the normal limits of operation. These conditions are shown in the following table.

Parameter Limit Value Reactor Power 102% of Nominal Rating 20.4 MW(t)

Reactor D2O Level Low 150 in. (3.81 m)

Core Inlet Temperature High 110 ºF (43.3 ºC)

Main Primary Coolant Flow Low 8700 gpm (549 l/s) 13.1 Maximum Hypothetical Accident The licensee has postulated the MHA as a complete blockage of flow to one fuel element by unspecified means. The flow blockage is assumed to result in complete melting of the fuel plates releasing all of the fission products to the primary coolant. The reactor is assumed to be shut down by one of the following mechanisms:

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• Fuel plates melt, dropping out of the core region, leading to loss of reactivity and shutdown.

• Manual reactor scram as a result of the fission product monitor alarm.

• Automatic major scram (TS 3.2.2, Table 3.2.2) actuated by stack monitor alarm.

Once the major scram is actuated by the high stack activity, normal ventilation is secured, confinement is isolated, and emergency ventilation is automatically established. This condition is assumed for the duration of the accident. Since the MHA does not involve a release of primary coolant, only volatile fission products are postulated for release. The inventory of noble gas and iodine fission products in the most heavily irradiated element is determined by the licensee using the ORIGEN2 computer code. Confinement operability for this scenario is required by TS 3.4, Confinement System, and emergency ventilation operability is required by TS 3.5.

All of the noble gas fission products are assumed to be released into the primary coolant and quickly collect in the helium space at the top of the reactor vessel. For the temperatures and coolant chemistry expected in the NIST MHA, most of the iodine is assumed be in the form of CsI, and would remain in solution in the primary coolant water. Accounting for radiolysis of CsI to Idoine (I2) gas, it is assumed that approximately 3% of the total iodine release is present as I2.

Based on a review of the literature on iodine releases from fuel failures, the staff concluded that this was a reasonable assumption for the conditions that would be expected for the NBSR MHA. ,

11 12 The gaseous fission products in the helium space at the top of the reactor vessel are released to the Confinement Building along with helium at a rate based on historical leak rate measurements characteristic of the primary system under emergency ventilation conditions (no normal building exhaust). Exhaust rates to the stack from these spaces are then determined by the emergency ventilation system. WSMS concludes that the removal rates assumed in the analysis by the licensee is reasonable relative to the maximum allowable leak rate in the annual integrated leak rate test required by TS 4.4.

The release of the fission product gases provides the source term for estimating doses to the general public at the 400-meter exclusion radius established in TS 5.1, Site Description. The doses to the public were calculated by the licensee following standard techniques. The codes HOTSPOT (first day) and CAP88 (>1 day) were used for doses from the passage of radioactive clouds. The direct doses were calculated following the methods used in NBSR-9, allowing for the depletion of the fission products in the building as the gases are released. The scattering from the air above the Confinement Building was calculated by the licensee with SKYDOSE. As a direct result of the aqueous iodine chemistry, the mitigating effect of the filters, and the closed primary system, the iodine dose to the public is entirely negligible. All dose components are below the regulatory requirements of 10 CFR 20 and the test reactor criteria in 10 CFR 100.

The predominant portion of the dose to the public were from the noble gas release; these results were verified by independent dispersion calculations performed by the staff.

11 Weber, C.F., E.C. Beahm and T.S. Kress, Models of Iodine Behavior in Reactor Containment, ORNL/TM-12202, Oak Ridge National Laboratory, October 1992.

12 McLaughlin, T.P., et al, A Review of Criticality Accidents, LA-13638, Los Alamos National Laboratory, May 2000.

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Postulated Doses to the Public (Unrestricted Area)

Location Type of Dose Dose mSv (mrem)

Exclusion Boundary (400 m) Whole Body (TEDE) 0.07 (7.0)

Thyroid (CDE) 0.001 (0.1)

The dose to the NIST staff was estimated by the licensee based on the fission product gas concentrations in rooms C-100 (the experimental floor) and C-200 (the operations level, where the control room is located) and the amount of time spent in the area. By procedure, the operators would evacuate the building of all non-essential personnel immediately upon seeing the high readings on the stack monitor and fission product monitor. They would then proceed to place the reactor in a safe condition, and leave themselves. For purposes of dose estimation, it is assumed that this takes 10 minutes. The results for the thyroid dose to the worker were verified as reasonable by independent calculations by the staff.

Postulated Doses to the Workers (Restricted Area)

Location Type of Dose 10 Minute Dose Sv (rem)

Experimental Floor (C100) Whole Body (TEDE) 0.0107 (1.07)

Thyroid (CDE) 0.0001 (0.01)

Operations Level (C200) Whole Body (TEDE) 0.0406 (4.06)

Thyroid (CDE) 0.0002 (0.02)

These calculated doses are based on conservative assumptions and show that the reactor can be put into a safe condition and all personnel evacuated within the dose limits allowed for an emergency. On the basis of these calculations, the projected doses from the MHA are acceptable for both the general public and the NBSR staff. Independent calculations performed by the staff using the same assumptions corroborate these conclusions.

13.2 Insertion of Excess Reactivity Accidents Accidents involving the insertion of excess reactivity were considered by the licensee to determine credible accident scenarios. A step reactivity insertion involving mishandling of a fuel element was determined to be incredible as a result of administrative controls, engineered safety features, passive safety features, and technical specifications. TS 3.1.3 requires all core grid positions to be filled, and TS 3.9.2.1, Fuel Handling within the Reactor Vessel requires all fuel be latched, thereby precluding any movement of fuel elements during reactor operation.

Two scenarios for a ramp insertion of excess reactivity were analyzed: startup accident and rapid removal of experiments. The consequences of these accidents bound any other credible scenario.

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13.2.1 Startup Accident The licensee evaluated a postulated startup accident assuming the reactor operator, in violation of training and procedures, continuously withdraws the shim safety arms from the reactor initially critical at a power level of 10-4 MW(t). The reactivity insertion rate is conservatively assumed to be 5x10-4 per second (a rate in excess of the maximum measured differential shim arm worth and limited to this maximum by TS 3.2.1). The transient is terminated by a high power level trip assumed to occur at 130% of full power (the TS 2.2 LSSS). An insertion time of 241 ms is assumed to insert the shim safety arms 5º, consistent with the 240 ms that is the limit specified in TS 3.2.1. Conservatively, no temperature or other reactivity feedback mechanism is credited. This scenario was analyzed for both the startup core and the EOC core. Using these assumptions, RELAP5 was used to study the transient behavior, with the results indicating a Minimum Critical Heat Flux Ratio (MCHFR) greater than 1.7 for the EOC core and 1.55 for the SU core. Both MCHFRs provide ample margin to ensure that no fuel damage will result.

13.2.2 Rapid Removal of Experiments The maximum absolute reactivity of any single experiment in the NBSR is limited to 0.5% by TS 3.8.1. Thus, the maximum credible excess reactivity insertion that could be caused by removal of a single experiment would be 0.5%. The licensee evaluated a postulated accident scenario in which an experiment containing the maximum allowed reactivity (0.5% ) is removed in 0.5 s (a 1.0% /s ramp). This postulated accident has been analyzed with the following assumptions:

• Initial power = 20.4 MW(t).

• Reactor power scram occurs at the LSSS of 26 MW(t) (130%) (TS 2.2).

• Negative feedback from increasing fuel and coolant temperatures is neglected.

• Shim safety arm insertion of 5º in 241 ms (consistent with TS 3.2.1 max = 240 ms).

• Prompt neutron lifetime of 650 µs.

The scenario was analyzed for both BOC and EOC cores, with the limiting case occurring at BOC. The lowest value of the critical heat flux ratio (CHFR) is 1.74, providing an adequate margin against fuel damage.

The staff concludes the inputs and assumptions for the startup accident and rapid removal of experiments analyses were reasonable and consistent with the TSs.

13.3 Loss of Primary Coolant The licensee evaluated a postulated loss of primary coolant scenario in which a major pipe break in the process room allows all of the primary coolant to drain from the reactor vessel into the process room located under the reactor while the reactor is operating at 20 MW(t).

Approximately 3,000 gal of primary coolant is trapped in the process room by a dam built for this purpose. The reactor scrams immediately on a loss-of-flow signal (TS 3.2.2). Primary coolant is passively added to the core from the Inner Reserve Tank which drains to a distribution pan that directs the coolant to the individual elements for over 20 minutes. Operation of a single valve adds the 3,000-gallon capacity of the D2O Emergency Cooling Tank by which the core is 13-4

fully protected for about 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. During this time, a system can be started by which lost primary water can be pumped from the dammed area in the process room up to the D2O Emergency Cooling Tank, providing virtually unlimited cooling time. Alternatively, through the addition of a single spool piece, H2O can be piped into the system to provide cooling. TS 3.3.2 requires the emergency core cooling system be available during reactor operation. With the cooling provided by this system, the temperature of the clad will remain well below the blistering temperature. Thus, no fission products will be released during this accident.

However, 3H contained in the primary water would be released as a result of the loss of coolant.

The Confinement System would mitigate the release to the environment, and is required during reactor operation by TS 3.4. The following conservative assumptions were made by the licensee in estimating the release:

• The 3H concentration in the primary coolant is at the maximum level permitted by TS 3.7.1 (5 Ci/l).

• After the break, emergency ventilation is immediately established. TS 3.5 ensures emergency ventilation is available during reactor operation.

• The process room is not isolated from the emergency ventilation system (ACV-10 is left open).

• The Emergency Ventilation System pulls the maximum design flow of 15 cubic feet per minute from this area.

• Equilibrium between the spilled D2O at an assumed temperature of 108 °F (42 °C) and the air in the process room is established immediately.

With these assumptions, the rate of 3H released to the stack is determined by the licensee to be 1.8x10-3 Ci/s. Using this release rate and a variety of weather conditions and EPA codes, the effluent concentration at or beyond the 400-meter boundary established by TS 5.1 is less than 1000 nCi/m3. This value was determined for extremely stable conditions and low wind speeds which could not persist over any significant length of time. Any release would be terminated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as remedial measures (pumping water into tanks, closing ACV-10, covering spilled water with plastic) would be taken immediately. Therefore, the maximum total dose to an individual member of the public at the 400-meter boundary would be 0.002 mSv (0.2 mrem),

well below 10 CFR Part 20 and 10 CFR Part 100 limits. The staff corroborated these results by independent calculations of the source term and dispersion analysis.

The NBSR SAR states that worker exposures would be determined by access to the process room, which is always strictly controlled. The 3H levels would result in a concentration approaching 1.25x104 DAC. The staff performed this calculation and independently verified the result. In the NBSR SAR, the licensee stated that if prolonged access were required, special provisions would be implemented to control exposure to acceptable levels.

13.4 Loss of Primary Coolant Flow The licensee evaluated five different scenarios for loss of primary coolant flow:

• Loss of Offsite Power.

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• Seizure of One Primary Coolant Pump.

• Throttling of Coolant Flow to the Outer Plenum.

• Throttling of Coolant Flow to the Inner Plenum.

• Loss of Both Shutdown Coolant Pumps.

Each of the scenarios was analyzed with the RELAP5 thermal-hydraulics analysis code with none of the scenarios resulting in fuel damage. The minimum value of the CHFR during the transients analyzed was found to be 2.19 for the case of loss of offsite power.

13.4.1 Loss of Offsite Power This accident scenario analyzed by the licensee assumes all three primary pumps trip upon loss of offsite power. The three primary coolant pumps will coast down, and primary coolant flow will drop to a value where one or more of the primary coolant flow monitors will generate a delayed scram signal. Flow monitoring scram signal channels are required by TS 3.2.2. The coolant flow coastdown in the RELAP5 model is validated by comparison with measured data. It is assumed that the scram occurs 400 milliseconds (ms) after flow has reached the trip value.

This allows for instrumentation sensing and scram actuation delays (max scram initiation time =

240 msec, TS 3.2.1). The transient analysis shows that the earliest flow trip is from the outer plenum flow at 0.896 seconds. After the 400 ms delay, a reactor scram is initiated terminating the transient. The minimum CHFR at the hot spot is 2.19 and the maximum fuel centerline temperature is 136.6 °C, which is below the TS 2.1 safety limit value of 450 °C, providing an adequate margin against fuel damage. The TSs cited in this section apply to all of the loss of flow accident scenarios below.

13.4.2 Seizure of One Primary Coolant Pump In this scenario, it is assumed that through some failure, such as a faulty bearing, the rotor of one pump suddenly becomes locked, imposing a rapid flow decrease. The RELAP5 code models the pump seizure as a 1/3 flow reduction over a time span of 3.5 s to assure the flow reduction transient is over and the discharge valve on the seized pump is closed (3 s closure time). At 3.5 s, it is assumed that a scram is initiated by flow sensing instrumentation. In response to the reduced flow, the minimum CHFR (at the hot spot in the outer plenum) decreases from an initial value of 2.67 to a new minimum value of 2.23, still providing an adequate margin against fuel damage.

13.4.3 Throttling of Coolant Flow to the Inner or Outer Plenums Because of the two-plenum configuration of the NBSR primary system, the possibility of inadvertent blockage of flow to either the inner or outer plenums exists. The licensee evaluated both scenarios. For the outer plenum scenario, the flow control valve DWV-1 is assumed to close, reducing the flow through the outer plenum. A scram is initiated when the flow reaches the LSSS low flow trip point of 4700 gpm (TS 2.2), and the scram is completed after 400 ms.

The complete closure of the flow control valve, after its 30-second stroke time, isolates the lower plenum of the reactor and cuts off the supply of forced coolant flow. The licensees RELAP5 calculation shows that closed loop recirculation flow paths are established between the inlet and outlet plenums and this recirculation flow removes heat from the core region by natural convection. When the buoyancy head is insufficient to induce closed loop recirculation, boiling 13-6

by itself is sufficient to remove decay heat from the core region provided that the power is below the flooding condition. The licensee calculated the flooding-limited power for the NBSR coolant channels to be 3.58 kW. The hottest coolant channel in the outer core region is estimated to be 1.5 times the power of an average channel. With the hottest channel at the flooding-limited power, the corresponding core power is 1.2 MW(t). The core power would fall below 1 MW(t) in less than 30 seconds after a reactor scram. Forced circulation (until full closure of DWV-1) and closed loop recirculation flow provide adequate cooling time for the reactor power to decrease below 1.2 MW(t), and therefore the throttling of a flow control valve would not lead to fuel damage.

For the inner plenum scenario, the flow control valve DWV-2 is conservatively assumed to close in 15 s, decreasing the flow through the inner plenum, with a scram completed 400 ms after the flow reaches the low flow trip point of 1200 gpm. In the most limiting BOC equilibrium core, the minimum CHFR at the hot spot (in the inner plenum) decreases from an initial value of 4.01 to a new low of 2.80. The transient is over once the shim arms begin to move in at 9.99 s. The discussions of heat removal by natural convection and boiling are applicable to the inner plenum flow transient, as well. The conclusion again is that no fuel overheating occurs due to the throttling of coolant flow to the inner plenum.

13.4.4 Loss of Both Shutdown Coolant Pumps TS 3.6 requires a diesel generator and a battery power system to be available when the reactor is operating. However, in this scenario, the loss of offsite power analyzed above is followed by a complete failure of all backup power sources. A reactor scram occurs on low primary flow with a 400 ms delay. Both the shutdown primary coolant pumps and all of the secondary coolant pumps also coast down due to the failure of all backup power sources. A RELAP5 simulation of this process is followed until the fuel reaches a relatively stable temperature, where it is being cooled by natural convective flow of water up through the fuel elements and down around the outside of the core. The coolant would gradually warm up; however, it is expected to take several hours for the bulk water temperature to reach the boiling point, allowing time for shutdown cooling to be restored.

13.5 Misloading of Fuel The fuel misloading accident scenario analyzed by the licensee assumes that an error occurs in the rotation of fuel elements and an irradiated fuel element is left in the M-4 location, which is normally filled by a fresh fuel element. Instead, it is assumed that the fresh fuel element was placed in the remaining empty position in the core. Power distributions were determined by the licensee for a set of 26 different postulated cases in which the positions of two elements were exchanged. The results indicate that the radial power distributions for the misloading of fuel accident have peak values of 1.68 and 1.51 in the inner and outer core regions, respectively.

The minimum CHFRs for the worst cases of the misloading of fuel accident can be inferred from the peak wall heat fluxes and the critical heat fluxes calculated previously. The minimum CHFRs are 2.50 and 2.01 for the misloaded fuel (worst case) in the inner and outer core, respectively. The minimum CHFRs for the misloaded fuel provide adequate safety margins, and therefore no fuel damage is anticipated for the misloading of fuel accident.

13.6 Conclusions The staff has reviewed the accident analyses presented in the NBSR SAR and concludes the licensee has considered a sufficient range of accident categories and analyzed limiting 13-7

scenarios for each category to bound all credible accidents for the NBSR. In addition, a hypothetical accident scenario (MHA) was analyzed that bounds all credible accident scenarios, and is shown to result in radiological consequences below the regulatory limits. Therefore, it is concluded that continued operation within the limitation of the TSs and facility license provides reasonable assurance that any credible accidents would not result in violation of facility safety limits or loss of fuel integrity.

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14 TECHNICAL SPECIFICATIONS The staff has evaluated the TSs as part of its review of the application for renewal of Facility Operating License No. TR-5. The TSs define certain features, characteristics, and conditions governing the operation of the NBSR. The TSs are explicitly included in the renewed license as Appendix A. The staff reviewed the format and content of the TSs for consistency with the guidance found in ANSI/ANS-15.1 and NUREG-1537. Other chapters of this SER discuss the evaluations of individual TSs. The staff specifically evaluated the content of the TSs to determine if the TSs meet the requirements in 10 CFR 50.36. The staff concluded that the NBSR TSs do meet the requirements of the regulations. The staff based this conclusion on the following findings:

• To satisfy the requirements of 10 CFR 50.36(a), the licensee provided proposed TSs with the application for license renewal. As required by the regulation, the proposed TSs included appropriate summary bases for the TSs. Those summary bases are not part of the TSs

• The NBSR is a facility of the type described in 10 CFR 50.21(c), and therefore, as required by 10 CFR 50.36(b), the facility license will include the TSs. To satisfy the requirements of 10 CFR 50.36(b), the licensee provided TSs derived from analyses in the NBSR SAR.

• To satisfy the requirements of 10 CFR 50.36(d)(1), the licensee provided TSs specifying a safety limit (SL) on the fuel temperature and limiting safety system settings for the reactor protection system to preclude reaching the SL.

• The TSs contain limiting conditions for operation on each item that meets one or more of the criteria specified in 10 CFR 50.36(d)(2)(ii).

• The TSs contain surveillance requirements that satisfy the requirements of 10 CFR 50.36(d)(3).

• The TSs contain design features that satisfy the requirements of 10 CFR 50.36(d)(4).

• The TSs contain administrative controls that satisfy the requirements of 10 CFR 50.36(d)(5).

The licensees administrative controls contain requirements for initial notification, written reports, and records that meet the requirements specified in 10 CFR 50.36(d)(1,2,7,8).

The staff finds the TSs to be acceptable and concludes that normal operation of the NBSR within the limits of the TSs will not result in radiation exposures in excess of the limits specified in 10 CFR Part 20 for members of the general public or occupational exposures. The staff also finds that the TSs provide reasonable assurance that the facility will be operated as analyzed in the NBSR SAR, and adherence to the TSs will limit the likelihood of malfunctions and the potential accident scenarios discussed in Chapter 13 of this SER.

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15 FINANCIAL QUALIFICATIONS 15.1 Financial Ability to Operate the Facility As stated in 10 CFR 50.33(f), Except for an electric utility applicant for a license to operate a utilization facility of the type described in 10 CFR 50.21(b) or 10 CFR 50.22, [an application shall state] information sufficient to demonstrate to the Commission the financial qualification of the applicant to carry out, in accordance with regulations of this chapter, the activities for which the permit or license is sought.

NIST does not qualify as an electric utility, as defined in 10 CFR 50.2. Further, pursuant to 10 CFR 50.33(f)(2), the application to renew or extend the term of any operating license for a nonpower reactor shall include financial information that is required in an application for an initial license. Therefore, the staff has determined that NIST must meet the financial qualifications requirements pursuant to 10 CFR 50.33(f), and is therefore subject to a full financial qualifications review by the NRC. NIST must provide information to demonstrate that it possesses or has reasonable assurance of obtaining the necessary funds to cover estimated operating costs for the period of the license. It must submit estimates for the total annual operating costs for each of the first 5 years of facility operations from the expected license renewal date, and indicate the source(s) of funds to cover those costs.

According to the application, the NBSR is operated within the NCNR. The projected operating costs for the NBSR are estimated to be $10,626,000 million per year for the fiscal years FY2009 thru FY2013. Funds to cover operating costs will come from the U.S. Department of Commerce appropriated budget. The applicant expects that this funding source will continue for the above-referenced fiscal years. The staff reviewed the applicants estimated operating costs and projected source of funds, and found them to be reasonable.

Based on its review, the staff finds that NIST has demonstrated reasonable assurance of obtaining necessary funds to cover the estimated facility operations costs for the period of the license. Accordingly, the NRC staff has determined that NIST has met the financial qualifications requirements pursuant to 10 CFR 50.33(f), and is financially qualified to hold the renewed license for the NBSR.

15.2 Financial Ability to Decommission the Facility The NRC has determined that the requirements to provide reasonable assurance of decommissioning funding are necessary to ensure the adequate protection of public health and safety. The regulation at 10 CFR 50.33(k) requires that an application for an operating license for a utilization facility contain information to demonstrate how reasonable assurance will be provided that funds will be available to decommission the facility. The regulation at 10 CFR 50.75(d) requires that each nonpower reactor applicant for or holder of an operating license shall submit a decommissioning report which contains a cost estimate for decommissioning the facility, an indication of the funding method(s) to be used to provide funding assurance for decommissioning, and a description of the means of adjusting the cost estimate and associated funding level periodically over the life of the facility. The acceptable methods for providing financial assurance for decommissioning are specified in 10 CFR 50.75(e)(1).

The application referenced a decommissioning cost estimate developed by Duratek, Inc., where the estimated cost was $56.1 million in 2003 dollars. The applicant updated the cost estimate to be $87,223,357 million in FY2009 dollars. The cost estimate shows costs summarized by labor, 15-1

radwaste shipment and disposal, energy, other (i.e., spent fuel shipment), and a contingency factor of 25 percent. The applicant stated that it will update its decommissioning cost estimate using an algorithm prepared by Duratek, Inc. based on NUREG-1307 where changes in labor, energy, and transportation costs will be estimated from producer price indices (available from the U.S. Bureau of Labor Statistics), adjusted for local conditions as described in the Duratek, Inc. cost estimate. Based on the staffs review of the information submitted by NIST regarding decommissioning the NBSR, the staff finds that the decommissioning cost estimate submitted by NIST is reasonable.

The applicant has elected to use a statement of intent (SOI) to provide financial assurance, as allowed by 10 CFR 50.75(e)(1)(iv) for a Federal, State, or local government licensee. The SOI must contain or reference a cost estimate for decommissioning and indicate that funds for decommissioning will be obtained when necessary.

The applicant provided a SOI, dated October 10, 2008, that states the signator will request as necessary from the U.S. Congress through the Department of Commerce and the Office of Management and Budget, external and NIST direct cost funds for decommissioning activities The decommissioning cost estimate is approximately $87,223,357 million for the DECON option (full scale decommissioning, decontamination, and demolition, restoring the site for unrestricted use as stated by NIST).

To support the SOI and the applicants qualifications to use an SOI, the application stated that NIST is a Federal agency, and included documentation which corroborates this statement. The application also provided information supporting NISTs representation that the decommissioning funding obligations of NIST are backed by the full faith and credit of the United States government. NIST also provided documentation verifying that Patrick Gallagher, Deputy Director of NIST (the signator of the SOI), is authorized to execute contracts on behalf of NIST.

The staff reviewed the applicants information on decommissioning funding assurance as described above and finds that the applicant is a Federal government licensee under 10 CFR 50.75(e)(1)(iv), the statement of intent is acceptable, the decommissioning cost estimate for the DECON option is reasonable, and NISTs means of adjusting the cost estimate and associated funding level periodically over the life of the facility is reasonable.

15.3 Foreign Ownership, Control, or Domination Section 104d of the AEA, prohibits the NRC from issuing a license under Section 104 of the AEA to any corporation or other entity if the Commission knows or has reason to believe it is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government.

The NRC regulation at 10 CFR 50.38, Ineligibility of Certain Applicants, contains language to implement this prohibition. According to the application, NIST is a Federal Agency within the U.S. Department of Commerce, and is not owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government. The NRC staff does not know or have reason to believe otherwise.

15.4 Nuclear Indemnity The staff notes that the applicant currently has an indemnity agreement with the Commission, and said agreement does not have a termination date. Therefore, NIST will continue to be a party to the present indemnity agreement following issuance of the renewed license. Under 15-2

10 CFR 140.71, NIST, as a Federal government licensee, is not required to provide nuclear liability insurance. The Commission will indemnify NIST for any claims arising out of a nuclear incident under the Price-Anderson Act (Section 170 of the Atomic Energy Act, as amended) and in accordance with the provisions under its indemnity agreement pursuant to 10 CFR 140.95, up to $500 million. Also, NIST is not required to purchase property insurance under 10 CFR 50.54(w).

15.5 Conclusions The staff reviewed the financial status of the licensee and concludes that there is reasonable assurance that the necessary funds will be available to support the continued safe operation of the NBSR and, when necessary, to shut down the facility and carry out decommissioning activities. In addition, the staff concludes there are no problematic foreign ownership or control issues or insurance issues that would preclude the issuance of a renewed license.

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16 OTHER LICENSE CONSIDERATIONS Previous sections of this SER concluded that normal operation of the reactor causes insignificant risk of radiation exposure to the public and that only an accident event could cause some exposure. Chapter 13 of the SER concluded that the maximum hypothetical accident is shown to result in radiological consequences that are below regulatory limits.

In this section, the staff reviews the impact of prior operation of the facility on the risk of radiation exposure to the public. The two parameters involved are the likelihood of an accident and the consequences if an accident occurred.

Because the staff concluded that the reactor was initially designed and constructed to meet nuclear safety requirements, the staff must also consider whether operation will cause significant degradation in safety features. Furthermore, because melting of a fuel assembly is the MHA, the staff has considered mechanisms which could increase the likelihood of fuel damage resulting in release of radionuclides. Possible mechanisms are:

• Radiation degradation of fuel element cladding strength.

• High internal pressure caused by high temperature leading to exceeding the elastic limits of the cladding.

• Corrosion or erosion of the cladding leading to thinning or other weakening.

• Mechanical damage as a result of handling or experimental use.

• Degradation of safety components or systems.

16.1 Prior Use of Reactor Components 16.1.1 Reactor Vessel and Components 16.1.1.1 Reactor Vessel Limited opportunities are available to inspect the NBSR vessel. In 2000, when the refueling head was removed, the licensee performed a visual inspection using binoculars. No flaws or unusual discolorations were observed. In January 2002, the licensee conducted an inspection using remote imaging equipment while the reactor vessel was fully defueled but still filled with D2O. This inspection included a detailed inspection of the vessel interior, the vessel exterior, and the interior of the thermal shield. The inspection revealed no flaws that indicated a decrease in the integrity of the welds or defects that would affect the vessel and component design function. Visual inspections are repeated if the refueling head is removed from the vessel. Remote imaging equipment inspections are repeated when the core is offloaded.

16.1.1.2 Reactor Vessel Exterior Inspection During maintenance activities, NIST personnel were able to insert a camera through a one-inch gap between the thermal shield and the reactor vessel to examine the reactor vessel wall and bottom and the thermal shield. The examination did not uncover any material condition that could lead to a failure.

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16.1.1.3 Reactor Vessel Embrittlement The NBSR vessel is fabricated from aluminum alloys 5052 and 6061. Oak Ridge National Laboratory studied heavily irradiated samples of the 6061-T6 alloy obtained from a control rod drive follower tube used in the High Flux Beam Reactor at Brookhaven National Laboratory.

The most heavily irradiated portions of the NBSR vessel are the tips of the beam tubes. These components will conservatively have accumulated less thermal neutron fluence than the irradiated samples by 2024. At that time, the ductility, while reduced, will retain approximately 70% of the original value. The Charpy energy will have dropped by over a factor of 6. The component materials will remain ductile, with reduced impact strength and toughness, but with reduced resistance to crack propagation under tensile stress, and reduced resistance to sudden pressure applications and impacts.

These components are never under significant tensile stress, therefore, tensile stress is not an issue. The surrounding D2O does exert minimal compressive force but this stress will not create cracks that can propagate quickly. As the vessel is entirely closed there can be no impacts against these components while the core is loaded. The staff concludes that the predicted 2024 embrittlement state of the reactor vessel and components creates no hazard to continued operation of the NBSR.

16.1.2 Fuel Element and Control Rods 16.1.2.1 Fuel Element Aging The NBSR core is completely replaced in 8 refueling outages over a nominal period of 1-year.

Therefore the staff concludes that fuel aging creates no hazard to continued operation of the NBSR.

16.1.2.2 Control Rod Aging The NBSR has three reactivity control systems consisting of four shim safety arms, a single regulating rod, and a moderator dump system. The shim safety arms are the primary means of reactivity control and shutdown capability. TS 4.1.2 requires annual determination of control rod worth. TS 4.2.1 requires semiannual verification of shim arm motion.

As discussed in Section 4.2.2 of this SER, the lifetime of the shim arms is affected by poison burn-up, corrosion, and radiation damage. The fact that corrosion does not limit the shim arm lifetime is demonstrated by the fact that similar arms remained in the CP-5 reactor for approximately 8 years until poison burn-up required their replacement. Over thirty years of operation of the NBSR reactor have shown this to also be true for the NBSR shim arms. The radiation damage to the shim safety arms is not significant during reactor operation since the shim arms are in the top reflector above the core where the fast neutron flux is relatively low.

Shim arm sets have been replaced at the NBSR reactor three times, with no radiation damage apparent in the shims removed. The staff also considered aging of the regulating rod. Since the regulating rod poison is aluminum, burn-up is insignificant. Corrosion damage is similar to other core components and minimal.

The staff concludes that normal replacement of the shim arms based on burn-up limits the effect of aging on the NBSR control elements. As a result, the staff concludes that control rod aging should not be a significant concern for the continued safe operation of the NBSR.

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16.2 Conclusions The staff has reviewed the prior use of reactor components as well as the aging of safety components, as described in the NBSR SAR, and concludes that there has been no significant degradation of reactor components to date. Further, the surveillance requirements in the TSs provide reasonable assurance that the reactor components will continue to be adequately monitored for degradation of systems and components.

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17 CONCLUSIONS On the basis of its evaluation of the application as discussed in the previous chapters of this SER, the staff concludes the following:

• The application for license renewal dated April 9, 2004, as supplemented on October 2, 2006; May 30 and August 14, 2007; and September 16 and October 21, 2008; complies with the standards and requirements of the Atomic Energy Act of 1954, as amended, and the Commissions rules and regulations set forth in Chapter I of Title 10 of the Code of Federal Regulations.

• The facility will operate in conformity with the application, as well as the provisions of the Act and the rules and regulations of the Commission.

• There is reasonable assurance that (1) the activities authorized by the renewed license can be conducted at the designated location without endangering the health and safety of the public and (2) such activities will be conducted in compliance with the rules and regulations of the Commission.

• As discussed in Chapters 4, 12, and 15 of this SER, the licensee is technically and financially qualified to engage in the activities authorized by the renewed license in accordance with the rules and regulations of the Commission.

• The issuance of the renewed license will not be inimical to the common defense and security or to the health and safety of the public.

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18 REFERENCES ANSI/ANS-15.1, The Development of Technical Specifications for Research Reactors, 2007 ANSI/ANS-15.4, Selection and Training of Personnel for Research Reactors, 1988 ANSI/ANS-15.8, Quality Assurance Program Requirements for Research Reactors, 1995 ANSI/ANS-15.11, Radiation Protection at Research Reactor Facilities, 1993 Atomic Energy Act of 1954, as amended Bretscher, M.M., Perturbation-Independent Methods for Calculating Research Reactor Kinetic Parameters, ANL/RERTR/TM-30, Argonne National Laboratory, Argonne, IL, 1997.

Code of Federal Regulations, Title 10, Chapter I, revised January 1, 2008, U.S. Government Printing Office DOE Fundamentals Handbook, Chemistry, Vol. 1 & 2, DOE-HDBK-1015, U.S. Department of Energy, Washington, D.C., January 1993.

Grund, J.E., Self-Limiting Excursion Tests of a Highly Enriched Plate Type D2O Moderated Reactor: Part I, Initial Test Series, IDO-16891, Phillips Petroleum Company, 1963.

Hofman, G.L., J. Rest and J.L. Snelgrove, Irradiation Behavior of Uranium Dioxide - Aluminum Dispersion Fuel, Argonne National Laboratory, October 1996.

Hanson, A.L., H. Ludewig and D. Diamond, Calculation of the Prompt Neutron Lifetime in the NBSR, Nuclear Science and Engineering, Vol. 153, 2006.

Hendrie, J.M., Final Safety Analysis Report on the Brookhaven High Flux Beam Reactor, BNL-7661, Brookhaven National Laboratory, Upton, NY, 1964.

Johns, M.W. and B.W. Sargent, Canadian Journal of Physics, 32, p. 136, 1954.

McLaughlin, T.P., et al, A Review of Criticality Accidents, LA-13638, Los Alamos National Laboratory, May 2000.

NUREG-1007, Safety Evaluation Report Related to the License Renewal and Power Increase for the National Bureau of Standards Reactor, September 1983.

NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, February 1996 Nuclear Waste Policy Act of 1982, as amended Snelgrove, J.L. and Hofman, G.L., Dispersion Fuels, Materials Science and Technology: A Comprehensive Treatment, Volume 10A: Nuclear Materials, Part I, R.W. Cahn, ed., New York, 1994.

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Tuttle, R.J., Delayed Neutron Data for Reactor Physics Analysis, Nuclear Science and Engineering, 56, p. 37, Jan. 1975.

Weber, C.F., E.C. Beahm and T.S. Kress, Models of Iodine Behavior in Reactor Containment, ORNL/TM-12202, Oak Ridge National Laboratory, October 1992.

Weeks, J.R., C.J. Czajkowski and K. Farrel, Effects of High Thermal Neutron Fluences on Type 6061 Aluminum, Effects of Radiation on Materials: 16th International Symposium, AS Kumar, et al, eds., ASTM Publication Code Number 04-011750-35, Philadelphia, 1993.

http://www.ncdc.noaa.gov http://www.earthquake.usgs.gov/research/hazmaps/products_data/2002/2002October/CEUS/C EUSpga2500v3.pdf http://www.marylandregionalaviation.aero/content/mdpublicuseairports/gai.html http://www.flightaware.com (for airport runway orientation) 18-2