Text

Enclosure 4: Nedo 32740, Revision 5, General Electric Nuclear Test Reactor Safety Report (Public)
	ML23086C028
Person / Time
Site:	Vallecitos Nuclear Center
Issue date:	03/31/2023
From:	Heckman D, Smyly J; GE-Hitachi Nuclear Energy Americas
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML23086C023	List: ML23086C024; ML23086C025; ML23086C026; ML23086C028;
References
	M230046 NEDO 32740, Rev 5
	Download: ML23086C028 (1)
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NEDO 32740 Revision 5 March 2023 GENERAL ELECTRIC NUCLEAR TEST REACTOR SAFETY ANALYSIS REPORT (PUBLIC)

Center and GE-Hitachi Dav .Id Digit*llysign,d by Jeff Smyly ON: c:n-.J11ff Smyly gn*Jeff Smyly c:-US United Digitally signed by Jeff Smyly StatOI i-us United StatH o*GEH ou*V1llecil01J Nuclear C.nt r Regulatory Compllanee David Heckman **i-ff,oy.sm)'lyOge com Ro aon: I am pp1ovin11 lhis document Prepared:

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Smyly, Regulatory Compliance

D. Heckman, VNC Licensing Lead Digitally signed by Thomas McConnell Reason: Approved Date: 2023-03-24 07:24-07:00 Approved: _ _ _ _ _ _ _ _ _ _ _ __

T. McConnell, Manager Nuclear Test Reactor

HITACHI 5 NOTICE This material was prepared by General Electric-Hitachi (GEH) for the United States Nuclear Regulatory Commission (NRC) to be used by the NRC to evaluate the relicensing of the GE Nuclear Test Reactor (NTR) located in Pleasanton, California (Facility License R-33, Docket No.

50-73). GEH assumes no responsibility for liability or damage which may result from any other use of the information disclosed in this material.

The information contained in this material is believed to be an accurate and true representation of the facts known, obtained, or provided to GEH at the time this material was prepared. GEH makes no warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this material, other than for the relicensing of the GE NTR in Pleasanton, California or that the use of any information disclosed in this material may not infringe privately owned rights including patent rights.

ii

HITACHI 5 ABSTRACT The GE NTR is described, and a summary of the facility safety evaluation is presented. The description includes the GE NTR history; the Vallecitos Nuclear Center Site and area characteristics; a detailed facility description; descriptions of Irradiation Facilities, instrumentation and control systems; and facility administration, including the Quality Assurance programs and shielding around the facility. The safety evaluation contains a summary of the analyses performed and the consequences of normal and off-normal conditions, and postulated reactor accident conditions.

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Table of Contents 1 THE FACILITY ................................................................................................................................................ 1-1

1.1 INTRODUCTION

.......................................................................................................................................... 1-1 1.2

SUMMARY

AND CONCLUSIONS OF PRINCIPAL SAFETY CONSIDERATIONS .............................. 1-1 1.3 GENERAL DESCRIPTION OF THE FACILITY ......................................................................................... 1-3 1.4 SHARED FACILITIES AND EQUIPMENT................................................................................................. 1-5 1.5 COMPARISON WITH SIMILAR FACILITIES ........................................................................................... 1-8 1.6

SUMMARY

OF OPERATIONS .................................................................................................................... 1-9 1.7 COMPLIANCE WITH THE NUCLEAR WASTE POLICY ACT.............................................................. 1-10 1.8 FACILITY MODIFICATIONS AND HISTORY ........................................................................................ 1-10 2 SITE CHARACTERISTICS ............................................................................................................................. 2-1 2.1. GEOGRAPHY AND DEMOGRAPHY ......................................................................................................... 2-1 2.2. NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES .................................... 2-9 2.3. METEOROLOGY .......................................................................................................................................... 2-9 2.4. HYDROLOGY ............................................................................................................................................. 2-10 2.5. GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING ................................................. 2-10 2.6. CONCLUSION............................................................................................................................................. 2-11 3 DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS ................................................................... 3-1 3.1. DESIGN CRITERIA ...................................................................................................................................... 3-1 3.2. METEOROLOGICAL DAMAGE ................................................................................................................. 3-2 3.3. WATER DAMAGE........................................................................................................................................ 3-2 3.4. SEISMIC DAMAGE ...................................................................................................................................... 3-2 3.5. SYSTEMS AND COMPONENTS ................................................................................................................. 3-4 4 REACTOR DESCRIPTION .............................................................................................................................. 4-1 4.1

SUMMARY

DESCRIPTION ......................................................................................................................... 4-1 4.2 REACTOR CORE .......................................................................................................................................... 4-1 4.3 BIOLOGICAL SHIELD ............................................................................................................................... 4-14 4.4 NUCLEAR DESIGN .................................................................................................................................... 4-17 4.5 THERMAL-HYDRAULIC DESIGN ........................................................................................................... 4-31 5 REACTOR COOLANT SYSTEMS .................................................................................................................. 5-1 5.1.

SUMMARY

DESCRIPTION ......................................................................................................................... 5-1 5.2. PRIMARY COOLANT SYSTEM ................................................................................................................. 5-1 5.3. SECONDARY COOLANT SYSTEM ........................................................................................................... 5-6 5.4. PRIMARY COOLANT CLEANUP SYSTEM .............................................................................................. 5-9 5.5. PRIMARY COOLANT MAKEUP WATER SYSTEM............................................................................... 5-11 5.6. NITROGEN-16 CONTROL SYSTEM ........................................................................................................ 5-12 5.7. AUXILIARY SYSTEMS USING PRIMARY COOLANT ......................................................................... 5-12 6 DESIGN BASES AND ENGINEERED SAFETY FEATURES ...................................................................... 6-1 iv

7 INSTRUMENTATION AND CONTROL ........................................................................................................ 7-1 7.1.

SUMMARY

DESCRIPTION ......................................................................................................................... 7-1 7.2. REACTOR CONTROL ROOM ..................................................................................................................... 7-5 7.3. SCRAM SYSTEM.......................................................................................................................................... 7-5 7.4. SAFETY-RELATED ITEMS ....................................................................................................................... 7-10 7.5. REACTOR REACTIVITY CONTROL SYSTEMS .................................................................................... 7-13 7.6. CONTROL CONSOLE ................................................................................................................................ 7-15 7.7. RADIATION MONITORING SYSTEMS................................................................................................... 7-15 7.8. NEUTRON SOURCE .................................................................................................................................. 7-15 8 ELECTRICAL POWER SYSTEMS ................................................................................................................. 8-1 8.1. NORMAL ELECTRICAL POWER SYSTEMS ............................................................................................ 8-1 8.2. EMERGENCY ELECTRICAL POWER SYSTEMS .................................................................................... 8-4 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................................................................................... 9-2 9.4 COMMUNICATION SYSTEMS................................................................................................................... 9-3 9.5 POSSESSION AND USE OF BYPRODUCT, SOURCE, AND SPECIAL NUCLEAR MATERIAL ......... 9-3 9.6 COMPRESSED AIR ...................................................................................................................................... 9-3 9.7 RADIOGRAPHY VACUUM SYSTEM ........................................................................................................ 9-4 10 EXPERIMENTAL FACILIITIES AND UTILIZATION ............................................................................... 10-1 10.1.

SUMMARY

DESCRIPTION ....................................................................................................................... 10-1 10.2. SECURED / MOVABLE EXPERIMENTS ................................................................................................. 10-2 10.3. EXPERIMENTAL FACILITIES.................................................................................................................. 10-2 10.4. EXPERIMENT REVIEW............................................................................................................................. 10-7 11 RADIATION PROTECTION PROGRAM / WASTE MANAGEMENT ...................................................... 11-1 11.1 RADIATION PROTECTION ...................................................................................................................... 11-1 11.2 RADIOACTIVE WASTE MANAGEMENT ............................................................................................. 11-12 12 CONDUCT OF OPERATIONS ...................................................................................................................... 12-1 12.1 ORGANIZATION ........................................................................................................................................ 12-1 12.2 REVIEW AND AUDIT ACTIVITIES ......................................................................................................... 12-7 12.3 PROCEDURES ............................................................................................................................................ 12-9 12.4 REQUIRED ACTIONS .............................................................................................................................. 12-12 12.5 REPORTS ................................................................................................................................................... 12-13 12.6 RECORDS .................................................................................................................................................. 12-14 12.7 EMERGENCY PLANNING ...................................................................................................................... 12-14 12.8 SECURITY PLANNING ........................................................................................................................... 12-15 12.9 QUALITY ASSURANCE .......................................................................................................................... 12-15 v

12.10 OPERATOR TRAINING AND REQUALIFICATION ............................................................................ 12-21 12.11 ENVIRONMENTAL REPORTS ............................................................................................................... 12-22 12.12 REFERENCES ........................................................................................................................................... 12-23 13 ACCIDENT ANALYISIS ............................................................................................................................... 13-1 13.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS .......................................................................... 13-1 13.2 EXPERIMENT DESIGN BASIS ACCIDENT ............................................................................................ 13-3 13.3 TRANSIENT MODEL ................................................................................................................................. 13-7 13.4 ANTICIPATED OPERATIONAL OCCURRENCES ................................................................................. 13-9 13.5 POSTULATED ACCIDENTS ................................................................................................................... 13-14 13.6 EXPERIMENT SAFETY ANALYSIS ...................................................................................................... 13-32 13.7 REACTOR SAFETY LIMITS ................................................................................................................... 13-40 14 TECHNICAL SPECIFICATIONS .................................................................................................................. 14-1 15 FINANCIAL QUALIFICATIONS.................................................................................................................. 15-1 15.1 FINANCIAL ABILITY TO CONSTRUCT A NON-POWER REACTOR ................................................. 15-1 15.2 FINANCIAL ABILITY TO OPERATE A NON-POWER REACTOR....................................................... 15-1 15.3 FINANCIAL ABILITY TO DECOMMISSION THE FACILITY .............................................................. 15-1 16 OPERATING EXPERIENCE ......................................................................................................................... 16-1 16.1 REACTOR FUEL ......................................................................................................................................... 16-1 16.2 SAFETY RODS............................................................................................................................................ 16-2 16.3 CONTROL RODS ........................................................................................................................................ 16-2 16.4 AREA RADIATION MONITORS (ARM) .................................................................................................. 16-2

16.5 CONCLUSION

............................................................................................................................................. 16-3 Figures FIGURE 1-1 REACTOR CELL, SOUTH CELL, AND CONTROL ROOM........................................................... 1-6 FIGURE 1-2 NUCLEAR TEST REACTOR FACILITY.......................................................................................... 1-7 FIGURE 1-3 DOE SPENT NUCLEAR FUEL DISPOSAL AGREEMENT (PAGE 1) ......................................... 1-14 FIGURE 1-4 DOE SPENT NUCLEAR FUEL DISPOSAL AGREEMENT (PAGE 2) ......................................... 1-15 FIGURE 2-1 SAN FRANCISCO BAY AREA MAP ............................................................................................... 2-1 FIGURE 2-2 VALLECITOS NUCLEAR CENTER AND SURROUNDING AREA ............................................. 2-2 FIGURE 2-3 TOPOGRAPHY CONTOUR OF VALLECITOS NUCLEAR CENTER ........................................... 2-3 FIGURE 2-4 VALLECITOS NUCLEAR CENTER SITE STRUCTURES ............................................................. 2-5 FIGURE 2-5 WILLIAMSON ACT PROPERTIES AND TRI-VALLEY CONSERVANCY EASEMENTS ......... 2-7 FIGURE 2-6 POPULATION DENSITY MAP ......................................................................................................... 2-8 FIGURE 3-1 BUILDING 105 FLOOR PLAN .......................................................................................................... 3-6 FIGURE 3-2 PART FLOOR PLAN AND ELEVATION ......................................................................................... 3-7 FIGURE 3-3 LINE DIAGRAM OF VENTILATION SYSTEM ............................................................................ 3-10 FIGURE 4-1 VERTICAL SECTION THROUGH THE NTR .................................................................................. 4-2 vi

FIGURE 4-2 FUEL CONTAINER ASSEMBLY ..................................................................................................... 4-3 FIGURE 4-3 FUEL DISK ......................................................................................................................................... 4-5 FIGURE 4-4 THERMAL NEUTRON FLUX AND CADMIUM RATIO TRAVERSE ........................................ 4-28 FIGURE 4-5 THERMAL NEUTRON FLUX TRAVERSES IN THREE MANUAL POISON SHEET SLOTS .. 4-29 FIGURE 5-1 PRIMARY PIPING AND INSTRUMENT DIAGRAM ..................................................................... 5-4 FIGURE 5-2 PRIMARY ISOMETRIC DIAGRAM ................................................................................................. 5-5 FIGURE 5-3 SECONDARY COOLING SYSTEM.................................................................................................. 5-8 FIGURE 7-1 NTR SCRAM SYSTEM SCHEMATIC DIAGRAM .......................................................................... 7-3 FIGURE 7-2 SIMPLIFIED BLOCK DIAGRAM OF ROD DRIVES .................................................................... 7-14 FIGURE 8-1 PARALLEL OFFSITE 60 KVA POWER ........................................................................................... 8-3 FIGURE 10-1 NTR NEUTRON RADIOGRAPHIC FACILITIES (SIDE VIEW)................................................. 10-3 FIGURE 10-2 MODULAR STONE MONUMENT NEUTRON RADIOGRAPHY FACILITY .......................... 10-4 FIGURE 12-1 NTR ORGANIZATION .................................................................................................................. 12-2 FIGURE 13-1 NTR TRANSIENT ........................................................................................................................ 13-10 FIGURE 13-2 INITIATION OF 5 KW HEATER FROM REDUCED COOLANT TEMPERATURE (65°F) ... 13-12 FIGURE 13-3 K-STEPS WITH 150-KW SCRAM ............................................................................................ 13-16 FIGURE 13-4 $1.3 STEP FROM 100-KW WITH SCRAM ................................................................................. 13-17 FIGURE 13-5 100-KW FINITE RAMP INSERTION WITH HIGH-FLUX SCRAM ......................................... 13-19 FIGURE 13-6 $4 RAMP IN 0.6 SECOND FROM 100 KW WITH SCRAM ...................................................... 13-20 FIGURE 13-7 $2 RAMP IN 0.3 SEC FROM 100 KW WITH SCRAM ............................................................... 13-21 FIGURE 13-8 $0.76 STEP FROM 100 KW - NO SCRAM ................................................................................. 13-22 FIGURE 13-9 K STEPS FROM 100 KW - NO SCRAM .................................................................................. 13-23 FIGURE 13-10 REACTOR POWER AND HOT SPOT FUEL TEMPERATURE VERSUS TIME, $0.76 STEP FROM SOURCE LEVEL, 55°F COOLANT INLET TEMPERATURE - NO SCRAM ............ 13-24 FIGURE 13-11 POSSIBLE REACTOR STATES FOLLOWING THE POSTULATED SEISMIC EVENT...... 13-24 FIGURE 13-12 DECAY HEAT RATE ................................................................................................................. 13-29 FIGURE 13-13 NODE STRUCTURE ADAPTED FOR TRACG ANALYSIS ................................................... 13-30 FIGURE 13-14 FUEL TEMPERATURE FOLLOWING LOSS OF COOLANT ACCIDENT ........................... 13-31 FIGURE 13-15 GRAPHITE HEAT-UP FOLLOWING LOSS OF COOLANT ACCIDENT ............................. 13-31 FIGURE 13-16 MULTI-CHANNEL CORE MODEL OF NTR (CORLOOP) ..................................................... 13-44 FIGURE 13-17 SCHEMATIC DIAGRAM OF THE NTR CIRCULATION LOOP MODEL (CORLOOP) ...... 13-45 FIGURE 13-18 REACTOR POWER VERSUS DNBR = DEPART FROM NUCLEATE BOILING RATIO.... 13-46 FIGURE 13-19 REACTOR POWER VERSUS CORE VOID FRACTION ........................................................ 13-47 FIGURE 13-20 REACTOR POWER VERSUS RELATIVE FLOW RATE ........................................................ 13-48 FIGURE 13-21 REACTOR POWER VERSUS CORE INLET TEMPERATURE .............................................. 13-49 FIGURE 13-22 LSSS AND SAFETY LIMIT FOR REACTOR POWER IN TERMS OF RELATIVE CORE FLOW RATE............................................................................................................................................ 13-51 vii

Tables TABLE 4-1 NUCLEAR PARAMETERS ............................................................................................................... 4-21 TABLE 4-2 NUCLEAR PARAMETERS (CONTINUED) .................................................................................... 4-22 TABLE 4-3 CORE MODEL CALCULATED POINT KINETICS PARAMETERS ............................................. 4-31 TABLE 4-4 TYPICAL NTR CORE THERMAL AND HYDRAULIC CHARACTERISTICS ............................ 4-33 TABLE 5-1 HEAT EXCHANGER SPECIFICATIONS .......................................................................................... 5-7 TABLE 7-1 SCRAM SYSTEMS .............................................................................................................................. 7-4 TABLE 7-2 SAFETY-RELATED ITEMS.............................................................................................................. 7-11 TABLE 11-1 STANDARD, CHECK, AND STARTUP SOURCES AT THE NTR .............................................. 11-3 TABLE 11-2 FISSILE AND FISSIONABLE MATERIAL AT THE NTR ........................................................... 11-3 TABLE 11-3 RADIATION MONITORING EQUIPMENT AT THE NTR .......................................................... 11-8 TABLE 11-4 STACK RELEASE ACTION LEVELS .......................................................................................... 11-16 TABLE 13-1 NTR EXPERIMENT DBA ISOTOPIC RELEASE TO REACTOR CELL...................................... 13-6 TABLE 13-2 NTR EXPERIMENT DESIGN BASIS ACCIDENT DOSES .......................................................... 13-7 TABLE 13-3 UNCERTAINTIES IN THE PRESENT METHODS FOR MEASURING IMPORTANT PROCESS VARIABLES ......................................................................................................................................................... 13-52 viii

1 THE FACILITY

1.1 INTRODUCTION

GE designed and constructed the Nuclear Test Reactor (NTR) as part of the experimental facilities at its Vallecitos Nuclear Center (VNC) Site in Alameda County, California. The reactor was designed as an experimental physics tool to advance the companys nuclear energy programs.

GE-Hitachi Nuclear Energy Americas LLC (GEH) currently operates the NTR facility for: (1) neutron radiography (neutrography) of radioactive and nonradioactive objects, (2) small sample irradiation and activation, (3) sensitive reactivity measurements, (4) training, and (5) calibrations and other testing utilizing a neutron flux.

The NTR is a heterogeneous, highly enriched-uranium, graphite-moderated and reflected, light-water-cooled, thermal reactor, licensed to operate at power levels not in excess of 100 kW (thermal).

. It has a confinement building to restrict the release of radioactivity to the environment, diversity and redundancy of instruments and controls and extremely low operating heat flux and temperatures.

The U. S. Atomic Energy Commission issued NTR a construction permit on October 24, 1957, and initial operating license on October 31, 1957. Renewals to License R-33 were issued on April 20, 2001 (Amendment No. 21, ML003775776) and December 28, 1984 (Amendment No.

18). Additional licensing history is included in Section 1.8.3.

Over 60 years of operation in the performance of a variety of experiments and testing for GEH and its customers has demonstrated the safety and effectiveness inherent in the reactors design and the companys operating methods.

1.2

SUMMARY

AND CONCLUSIONS OF PRINCIPAL SAFETY CONSIDERATIONS The operation and use of the NTR has no negative consequence to the health and safety of the public. The reactor facility is designed to contain radioactivity and monitor radioactive releases.

The facility is operated in accordance with approved procedures which limit radiation exposures 1-1

and off-normal operation of the reactor. In addition, built-in design features and automatic shutdown features prevent temperatures from exceeding heat flux limits.

The VNC site is not adjacent to a large population center and the weather is not prone to damaging extremes. The NTR core consists of an aluminum can filled with water in a graphite pack. The fuel is a stable aluminum-uranium alloy operated at low heat flux and thermal temperatures. The reactor is in a confinement building, which maintains the reactor cell at a negative pressure for inward air flow. Air for the confinement building is exhausted through a stack and is monitored for radioactive releases.

The NTR has a negative void coefficient of reactivity and a negative temperature coefficient of reactivity above 124°F, which is approximately the steady-state operating temperature.

Additionally, because of low stored heat content, the NTR fuel will not melt when the fuel coolant water is lost. These features greatly contribute to the protection of occupational workers, members of the general public, and the environment in the unlikely event of an accident.

The NTR has a scram system which automatically inserts enough negative reactivity to shut down the reactor and maintain it shut down. The system is activated by both manual and automatic switches when predetermined parameters approach preestablished limits.

The facility response to certain postulated credible events and less probable accidents which have potential safety significance has been evaluated. These events, further discussed in Chapter 13, include:

1. Loss of Normal Electrical Power

2. Loss of Secondary Coolant

3. Loss of Facility Air Supply

4. Inadvertent start of primary pump (Inadvertent Core Inlet Temperature Change)

5. Fuel Handling Errors

6. Uncontrolled reactivity increases Idealized Step Reactivity Insertions - with Scram Idealized Finite Ramp Reactivity Insertions - with Scram Idealized Finite Ramp Reactivity Insertions - with Scram Reactivity Insertions - without Scram Reactivity Insertions - without Scram

7. Reactor Loss of Flow Accident

8. Rod Withdrawal Accidents 1-2

9. Reactor Loss of Flow Accident

10. EXPERIMENT DESIGN BASIS ACCIDENT The three acceptance criteria for anticipated operational occurrences are the following:

1. Release of radioactive material to the environs does not exceed the limits of 10 CFR 20.1101(d) or 10 CR 20.1302 (Chapter 11).

2. Radiation exposure of any individual does not exceed the limits of 10 CFR 20.1201 for occupational workers or of 20.1301/1302 for members of the public (Chapters 11 and 13).

3. An established safety limit is not exceeded (Chapter 13).

The acceptance criteria for postulated accidents are as follows:

1. Release of radioactive material does not exceed the limits of 10 CFR 20.1201 for occupational workers or 10 CFR 20.1301/1302 for members of the public.

2. An established safety limit is not exceeded.

Because of the many safety features provided and the strong administrative control applied to operation of the facility, the possibility of an accident involving high radiation exposure or the dispersion of substantial quantities of radioactivity is considered extremely remote. However, the protection of the health and safety of the public is ensured further by housing the reactor in a thick-walled concrete cell that provides radiation shielding and permits controlled release of airborne contamination. Based on the descriptive and analytical information provided in this report and the proven performance of the facility over an extended operating period, it is concluded that the design and operating methods of the NTR facility provide the reasonable assurance required by the regulations that the health and safety of the public will not be endangered by continued operation of the facility.

1.3 GENERAL DESCRIPTION OF THE FACILITY The NTR is located at the Vallecitos Nuclear Center (VNC), which is largely undeveloped grasslands within the Livermore Upland physiographic area.

VNC is situated on the north side of Vallecitos Valley in Southern Alameda County within five miles of Livermore and Pleasanton and approximately 35 air miles east-southeast of San Francisco and 20 air miles north of San Jose. Vallecitos Valley is approximately two miles long 1-3

and 1 mile wide. The valley is at an elevation of 400 to 500 feet above sea level and is surrounded by barren mountains and rolling hills. There is very little commercial and residential development in the valley. The VNC Site slopes upward from about 400 feet at its relatively flat southern end to a 1,200-foot ridge on the north. The southern end of the property slopes slightly to the southwest where it drains through ditches to Vallecitos Creek which then discharges to Arroyo de la Laguna near the north end of Sunol Valley - two or three miles southwest of the property.

The NTR is a heterogeneous, enriched-uranium, graphite-moderated and -reflected, light-water-cooled, thermal reactor, licensed to operate at power levels not in excess of 100 kW (thermal).

The fuel consists of highly enriched uranium-aluminum alloy disks, clad with aluminum. The core is cooled either by natural or forced circulation of deionized light-water circulated in a primary system constructed primarily of aluminum. The reactor operates at very low temperature and low heat flux. Reactivity is controlled by up to six manually positioned cadmium sheets (although only 1/16 of a sheet is installed as of January 2023), four boron-carbide-filled safety rods (spring-actuated for reactor scram), and three electric- motor-driven boron-carbide-filled control rods. Conventional instrumentation is provided to indicate, record, and control important variables, and shut down the reactor automatically if assigned operating limits are exceeded. The reactors irradiation facilities include a central sample tube, penetrations through and into the reflector, the reflector faces, and the beams from any of these facilities. When used as a neutron 12 source, the reactor can provide unperturbed neutron fluxes (at 100 kW) of about 2 x 10 thermal 2 12 2 n/cm -sec and an epicadmium flux of about 1 x 10 n/cm -sec. When used as a detector,

-6 reactivity effects can be measured with a precision of 10 k/k without the use of a pile oscillator.

The reactor is located within a thick-walled concrete cell which, along with the control room, north room, setup room and the south cell, comprises the NTR facility. An overall view of the facility, except the north room and set-up room, is shown in Figure 1-1. Principal equipment in the concrete reactor cell includes the reactor, the reactor control mechanisms, the coolant system, and a fuel loading tank which provides radiation shielding and the primary water system reservoir. The control room contains the control console and provides space for experiment equipment, preparation, and an operator work area. The south cell is a concrete-shielded room which provides access to the thermal column, the horizontal facility and the horizontal facility south beam. The north room provides space for performing experiments utilizing the horizontal 1-4

facility north beam and the Cable Held Retractable Irradiation System (CHRIS). The set-up room is used for storage and setup of experiments involving irradiation or testing. There is a wall penetration into the south cell for long trays to utilize the horizontal facility south beam.

Release of radioactive materials is strictly regulated. Radioactive gas and particulates released from the reactor cell are monitored continuously and the reactor is shut down if required to reduce emissions below release limits. Solid and liquid radioactive wastes are collected by trained individuals in accordance with approved procedures and disposed of in accordance with applicable regulations.

The entirety of the NTR Facility is included within and constitutes a 10 CFR 20 Restricted Area as shown in Figure 1-2. Radiation protection of individuals is controlled by a variety of means.

A radiation protection program has established postings to notify workers of radiological hazards. Routine and special surveys assure that radioactive materials are controlled and that there is no unplanned exposure or movement. Also, there are ion chambers and filter sample stations strategically located in the facility to warn of unusual increases or releases of radioactive materials. An ALARA (as low as reasonably achievable) program also requires review of facility changes and new experiments to design for reduced radiation exposure. In addition, routine audits and reviews are conducted by personnel independent of reactor management and reactor operations personnel who are trained in radiation protection and work to approved written procedures.

1.4 SHARED FACILITIES AND EQUIPMENT The NTR Facility shares many facilities and equipment in Building 105 with other laboratory facilities. These include potable water supply, fire protection, emergency supplies and support, HVAC System, AC electrical distribution, compressed air system and the occupied spaces of Building 105.

Whereas small amounts of byproduct material may be handled in some of the laboratories in Building 105 under California Radioactive Material License 0017-01, there is no other federally licensed equipment or facilities in Building 105 such as hot cells, critical or subcritical assemblies, neutron sources, or irradiation facilities.

1-5

Figure 1-1 Reactor Cell, South Cell, and Control Room 1-6

Figure 1-2 Nuclear Test Reactor Facility 1-7

The NTR shared building spaces are adequately separated by walls to delineate the NTR facility from the other offices and laboratories. Other means of separation have been installed to adequately isolate the shared facilities and equipment. For instance, the potable water supply to the NTR contains an approved reduced pressure backflow preventer. Although there are shared load centers, reactor safety equipment is connected to electrical circuits which are not shared with other facilities and equipment outside NTR.

Other shared facilities and equipment have been established at NTR in order to increase the convenience and the strength of resources available to a small facility. These include the fire protection system (building sprinkler system, fire hoses, and portable fire extinguishers),

building emergency response teams, an emergency supply cabinet, HVAC system, and a compressed air system, none of which support reactor safety systems.

1.5 COMPARISON WITH SIMILAR FACILITIES The design of the NTR resulted from the evolution of a series of reactors designed by scientists at the GE Knolls Atomic Power Laboratory (KAPL) in Schenectady, New York. The earlier reactors were known as thermal test reactors (TTR). Three models were built and operated successfully. The GE TTR operated from 1954 to the mid-eighties at KAPL. The TTR No. 2 operated from 1955 until 1972 at the Battelle Memorial Institute Pacific Northwest Laboratory.

The third TTR, the Savannah River National Laboratory Standard Pile operated from 1953 to 1979.

The logical evolution which led to the design of the NTR produced a versatile and safe reactor.

Features which contribute to the safety of the reactor, and which were incorporated into the design and construction of NTR include:

1. Negative void coefficient of reactivity.

2. Small positive coolant temperature coefficient of reactivity which becomes negative at a water temperature slightly above the operating temperature.

3. A control system extremely sensitive to changes in reactivity so that minute changes are detectable.

4. Safety and control functions that are separate, except for an interlock which requires all safety rods to be fully withdrawn prior to withdrawing any control rod. This 1-8

ensures that negative reactivity is available if needed for scram before a control rod can be moved.

5. Manually positioned cadmium sheets that can be used to limit reactivity controllable from the console and to provide enough negative reactivity to preclude any possible danger or criticality during fuel loading.

6. An instrumentation system which includes fail-safe and redundant features as well as proven reliable components.

7. A system constructed from materials having properties compatible with their intended service.

Safety measures which have been incorporated into the operation of the facility include:

1. Very low heat flux, even at the maximum operating power.

2. Temperatures and pressures only a little above ambient.

3. Low operating power, resulting in a low fission-product inventory.

4. Rigid control by operations management of all experiments performed in the reactor facility.

5. Performance of all activities that can affect nuclear safety under the direction of an NRC-licensed reactor operator or NRC-licensed senior reactor operator, as required.

1.6

SUMMARY

OF OPERATIONS The NTR was originally built as an experimental tool for diverse applications. In the first 5 years of operation, it was used for pile-oscillator measurements of nuclear cross sections of materials, calibrations of foils and nuclear sensors, neutron activation analysis, studies of radiation damage in semiconductors, nuclear fuel enrichment measurements, and cryo-nuclear investigations.

Over the years the reactor has been used for a variety of purposes from neutron absorption measurements of material at a reactor power level of 10 watts to 24 hour/day irradiation of filter tape. More recently, the NTR has been used for sensitivity reactivity measurements, training, and calibrations utilizing a neutron flux. Currently the NTR is used for neutron radiography of radioactive and nonradioactive objects, and small sample irradiations.

1-9

The reactor can operate at extremely low power levels not in excess of 100 kW (thermal) and has operated in recent years at a nominal 800 annual EFPH.

1.7 COMPLIANCE WITH THE NUCLEAR WASTE POLICY ACT The NTR has entered into Contract DE-CR01-83NE4446 (Figure 1-3) with the Department of Energy (DOE) whereby the DOE will accept the NTR spent nuclear fuel. This satisfies the requirements of the Nuclear Waste Policy Act of 1982.

1.8 FACILITY MODIFICATIONS AND HISTORY 1.8.1 VNC Site History GE entered the nuclear power industry in the early 1950s and GE-Hitachi Nuclear Energy Americas LLC (GEH) continues to be an industry leader in Boiling Water Reactor technology and design. Earlier nuclear industry experience was gained while operating the Hanford reactors for the Department of Energy in the 1940s and 1950s and in the U.S. Navy Nuclear Power Plant Program.

The Vallecitos Nuclear Center, originally called the Vallecitos Atomic Laboratory, was established in 1956. Experience at the VNC site with NRC licensed activities include:

Radioactive Materials Laboratory (License SNM-960), June 1956 - Present.

Vallecitos Boiling Water Reactor (License DPR-1), August 1957 - December 1963.

Critical Experiment Facility (License CX-4), November 1957 - mid 1966.

General Electric Test Reactor (License TR-1), December 1958 - October 1977.

Empire State Atomic Development Associates Vallecitos Experimental Superheat Reactor (License DR-10), January 1964 - February 1967.

In addition, some activities have been and continue to be performed under State of California source and byproduct material licenses.

1.8.2 NTR History The NTR was constructed under construction permit CPRR-19, issued October 24, 1957, as requested by General Electrics application, dated June 5, 1957. Operation of the reactor up to powers of 30 kW was authorized by facility license R-33, issued on October 31, 1957. Initial 1-10

loading of the reactor began on November 7, 1957, and criticality was first achieved on November 15, 1957.

The NTR reactor was operated at powers up to 30 kW for more than 5000 hours0.0579 days 1.389 hours 0.00827 weeks 0.0019 months until, on July 22, 1969, the license was amended and revised in its entirety to authorize operation of the reactor at power levels of up to a maximum of 100 kW steady-state power (later amended to power level not in excess of 100 kW). Since then, the reactor has operated under license R-33, as amended, at power levels up to 100 kW for more than 45,000 hours0 days 0 hours 0 weeks 0 months while performing a wide variety of experiments.

In 1976, the reactor core can developed a leak in a weld area, necessitating replacement. The reactor fuel was removed and inspected, and a major portion of the reactor was dismantled.

The core can was replaced, as well as some of the graphite in the central area. Some modification of the irradiation facilities occurred at this time. The reactor was reassembled, utilizing the original fuel, and routine operation resumed.

Prior to 1985, many original instruments were replaced, including the picoammeters (linear wide range neutron monitors), log N (log wide range neutron monitor), area radiation monitors, and the stack effluent gas and particulate monitors.

1.8.3 NTR Licensing History Descriptive information for the NTR facility was originally contained in GEAP-1005, Safeguards Report, Nuclear Test Reactor. This document was part of the application, dated June 5, 1957, for a construction permit and facility license; pursuant to this application, as amended, construction permit CPRR-19 and facility license R-33 (Docket 50-73) was issued. In 1958, General Electric amended the license application to incorporate changes in procedures and equipment. At that time, the information in document GEAP-3068, Summary Safeguards for the Nuclear Test Reactor (October 7, 1958), was substituted for the related information in the original application.

A new Summary Safeguards Report for the NTR (APED-4444) was submitted February 1, 1965, which updated and provided new information about design, operation, and safety analysis. It accompanied a license application amendment requesting separate technical specifications and provided a documentation mechanism for simplifying amendatory actions.

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A revision of APED-4444 (APED-4444A) was submitted November 21, 1968 and amended March 31, 1969, and May 28, 1969, as part of a license application amendment requesting an increase in authorized maximum steady-state power level from 30 to 100 kW.

The SAR, APED-444A was again revised and reissued as NEDO-12727 in April 1977. The purpose of NEDO-12727 was to update the description of the facility and its organization and to summarize additional safety evaluation of the facility. NEDO-12727 was submitted as supporting material for the renewal of the operating license.

The SAR was again issued in August 1997 as NEDO-32740 and addressed minor licensing changes including an increased the 200-curie byproduct material limit to 2000 curies that was approved August 19, 1992, and amendment No. 19, dated June 2, 1989, that allowed one exception to Tech Spec 3.5.3.5 which requires separation of explosive and radioactive materials.

A revision (NEDO 32740, revisions 1) to the 1997 SAR was issued as replacement pages in June 2000.

The facility license was renewed (Amendment No. 21) on April 20, 2001.

The facility license was amended on October 22, 2007 (Amendment 23), to reflect a change in ownership of the facility from General Electric Company (GE) to GE-Hitachi Nuclear Energy Americas LLC (GEH).

A license amendment request was submitted on February 2, 2015, for the unconditional release (pursuant to 10 CFR 50.83) of approximately 610 acres on the north end of the VNC property.

This amendment included revision 2 to the Technical Specifications (NEDO 32765, revision 2) and change pages to the SAR (NEDO 32740, revision 2). This amendment was approved and issued as amendment 25 on June 29, 2022.

A license amendment request was submitted on July 1, 2019, and subsequently issued by the NRC as amendment 24 on February 12, 2021, to align the site emergency plan with the ANSI 15.16, 2018, Emergency Planning for Research Reactors.

This document is a revision to the current SAR, NEDO 32740, dated August 1997, and updated June 2000, and is intended to accomplish the following:

1. Update the description and organization of the facility.

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2. Support an application for the NTR license renewal to permit continued operation of the reactor to power levels not in excess of 100 kW (thermal).

The safe and efficient operation of the NTR is evidenced by the 62 annual reports submitted to the NRC on operating experience pertinent to safety as of 2022.

1.8.4 Recent Modifications In 2003 a change was made pursuant to 10 CFR 50.59(c) to route power to allow the control rods to be bottomed In 2015 the shop on the south side of the building across from the Control Room was converted into NTR office space. The NTR Setup Room was expanded to enclose the loading dock placing the wall access penetration to the south N-ray position in the Setup Room.

In March 2020, the entire radiation monitoring system was replaced with a digital system to increase reliability and eliminate installed check sources.

In 2022, the primary coolant thermopile was designated as abandoned in place as it was determined it could not be reliably placed in service. The thermopile failed in 1974, was replaced in 1999, but has never been returned to service. Pursuant to 50.59(c), the compensatory measure of using the core outlet (TC-2) and core inlet (TC-5) thermocouples to determine the core differential temperature has been made the permanent method for performing the heat balance determination.

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Figure 1-3 DOE Spent Nuclear Fuel Disposal Agreement (page 1) 1-14

Figure 1-4 DOE Spent Nuclear Fuel Disposal Agreement (page 2) 1-15

2 SITE CHARACTERISTICS 2.1. GEOGRAPHY AND DEMOGRAPHY 2.1.1. Site Location and Description Specification and Location The NTR is situated on the Vallecitos Nuclear Center (VNC) Site near Pleasanton, California (Figure 2-1). The VNC Site is owned by GE and is used for nuclear research and development.

The VNC is located on the north side of the Vallecitos Valley, which is approximately two miles long and one mile wide; its major axis is east-northeast and west-southwest. The valley is at an elevation of 400 to 500 ft (120 to 150 km) above sea level and is surrounded by barren mountains and rolling hills.

The VNC is located east of the San Francisco Bay in Alameda County, California; along SR 84 approximately 2.5 mi (4 km) east of where SR 84 intersects highway 680.

The reactor is located at: Lat/Long: 37.608741, -121.842296, UTM 10 N 602183 4163036.

Figure 2-1 San Francisco Bay Area Map 2-1

Boundary and Zone Area Maps The Site (Figure 2-2) is bounded on the west, north, and east by hilly terrain; in some places, the hills are about 400 ft (120 m) above the general Site elevation. Vallecitos Road (SR 84) is at the southern boundary of the Site, from which an expanse of gently rolling grassland extends north for about 1.2 mi (2 km) at which point mountains form a northern barrier; completing the geographical encirclement of the Site. The site and its boundary are according to Figure 2-3.

Figure 2-2 Vallecitos Nuclear Center and Surrounding Area A portion of the VNC Site to the northeast is rolling terrain that is gently sloping toward the southwest. The southern part of the Site, adjacent to the SR-84, is relatively flat and accommodates the VNC Site Developed Area; an approximate 135-acre (0.5-sq. km) parcel in the southwest quadrant of the VNC Site between the 400 and 600-ft (120 and 180-m) topographic contours. This area contains principal facility structures, including the NTR, laboratories, and administrative facilities.

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Figure 2-3 Topography Contour of Vallecitos Nuclear Center 2-3

Site Structures The NTR is located Building 105. The operations boundary for NTR is the boundary defined by the portions of Building 105 occupied by the NTR Facility and is also the 10 CFR 20 Restricted Area for the NTR Facility and the emergency preparedness zone (EPZ).

Building 105 and other Facilities located near the NTR are shown in Figure 2-4. The main laboratories are in buildings in the Site Developed Area, approximately 1/3-mi (500-m) north of the SR 84.

Building 102 contains the Radioactive Materials Laboratory, used for post-irradiation studies and research and development activities, and administrative offices Building 103 houses analytical chemistry laboratories and offices.

Building 104 contains Site storage and warehouse facilities.

Building 105 contains offices and laboratories and houses the NTR and another shielded cell that was formerly used as a critical experiment facility.

Building 106 contains machine, sheet metal and facilities maintenance shops.

Building 107 is the Hazardous Material Storage Building.

Building 400 and 401 are currently leased to ManTech.

There are three other reactor facilities within the VNC Site Developed Area: DPR-1, Vallecitos Boiling Water Reactor (VBWR); DR-10, Empire State Atomic Development Agency Vallecitos Experimental Superheat Reactor (EVESR); and TR-1, GE Test Reactor (GETR). All three of these facilities are currently in SAFSTOR (SAFe STORage) under possession-only licenses. Fuel from these facilities has been removed from the site.

Temporary storage of solid radioactive waste is accommodated at the hillside storage facility.

Facilities are available on the Site for handling, sorting, and processing liquid and solid radioactive wastes generated at all VNC nuclear facilities. A liquid waste evaporator facility is located near the hillside storage facility. There are no radioactive liquid effluents discharged from the facility. A nonradioactive liquid waste chemical treatment plant and sewage treatment plant are in the southwest corner of the Site.

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Figure 2-4 Vallecitos Nuclear Center Site Structures 2.1.2. Population Distribution Figure 2-2 provides visual testimony of population near the VNC. Inside an approximate half-mile (1 km) radius and immediately west of the VNC are a horse farm and about a dozen houses.

These and approximately 20 more residences are within a mile of the VNC and have direct access to SR 84 by way of Little Valley Road. Site property outside of the Site Developed Area is largely leased for livestock grazing. Neighboring land beyond the site boundary is devoted to agriculture and cattle raising. The population growth rate within a mile of the VNC has been negligible over the past few decades.

Analysis provided in Chapter 13 demonstrates that the effects of potential accidents involving the NTR would not place even this population at risk in that dose rates would not exceed 10 CFR 20.1301 limits at the Site Boundary.

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Population Within 2.5 Miles (4 km):

The City of San Franciscos San Antonio Reservoir lies within this radius. The rate of population growth within this area has increased by only 10% since 2010 but is expected to grow by about 30 percent in the more populated areas around the VNC by 2040, largely due to migration of residents from the San Francisco Bay area. In unincorporated East Alameda County, including Sunol, less growth is expected than in the entirety of Alameda County due to land use requirements set forth in the Alameda East County Area Plan.

A major part of the land around the VNC is rugged terrain which does not attract industrial or residential buildings. Substantial parcels of privately-owned land have been placed into the Alameda County and Preserve Program under the California Land Conservation Act of 1965, commonly referred to as the Williamson Act. Figure 2-5 (public domain) provides details on land currently under the Williamson Act and the Tri-valley Conservancy and indicates areas of ongoing road improvements.

The Williamson Act allows local governments to enter into contracts with private landowners to restrict the use of the land in return for lower property tax assessments than would be the case if they were assessed for potential market rate development. The Alameda East County Area Plan, adopted in 1994 and updated in 2002, places restrictions on the cancellation of Williamson Act contracts. According to plan projections, housing units in unincorporated lands were anticipated to increase from a total of 300 in 1990 to 470 at plan buildout, and jobs were anticipated to remain at a total of 100 in both 1990 and at plan buildout (Alameda County Planning Department 2002). The VNC Site Developed Area is designated as industrial, which allows for a maximum building density of 60 acres (0.25 sq. km) of the approximately 150-acre (0.5 sq. km) property (Alameda County Planning Department 2002). This allows for warehouses, storage, and low intensity office use. In the 310-acre (1.25 sq. km) Little Valley Specific Plan Area just north of SR 84 (Reference 4) at Little Valley Road, land is designated as planned development with one dwelling unit per each full 4.5 acres (0.02 sq. km) and a minimum parcel size of 2 acres (8,000 sq. m) (Alameda County Community Development Agency 1997).

Another parcel to the northeast of the VNC is under the Tri-Valley Conservancy. This land is under permanent restrictions to be maintain as agricultural property.

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Figure 2-5 Williamson Act Properties and Tri-Valley Conservancy Easements San Francisco Public Utilities Commission (SFPUC) as part of its Water System Improvement Program is preparing a Habitat Conservation Plan for the Alameda Creek watershed, which includes the Sheep Camp Creek facility which is southwest of the VNC. The Sheep Camp Creek facility is bordered by SR 84 on the south, Little Valley Road and the Little Valley community on the east, I-680 and Koopman Road on the west, and open space on the north. In addition to providing habitat for special-status species, the facility allows for cattle grazing to reduce fuel loads and fire risk.

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Population Within 4 Miles (6.5 km):

Minor land development has occurred within the 4-mile radius. One location is 2.5 to 3 miles (4 to 5 km) to the west and northwest, associated with the expansion of the town of Pleasanton and the unincorporated areas of Happy Valley and Sunol. The other location is approximately 3 miles (5 km) to the east associated with the expansion of the town of Livermore.

The Hetch-Hetchy underground aqueduct is within this area and runs approximately in the east west direction.

Population Centers Within 5 Miles (8 km):

The 5-mile (8 km) radius contains two towns. Pleasanton (population 83,000) is approximately 4 miles to the northwest. A 1200-foot (0.4 km) high mountain range is the outskirts of Livermore at 5 miles to the northeast. Livermore, the largest population center (population 90,300), is largely a bedroom community with some light industry and agricultural activities.

The Livermore Division of the Veterans Administration Palo Alto Health Care System (population 400) is also located approximately 4 miles to the east.

Population density in the bay area is depicted in Figure 2-6 (map is per public domain).

Figure 2-6 Population Density Map 2-8

2.2. NEARBY INDUSTRIAL, TRANSPORTATION, AND MILITARY FACILITIES Both Livermore and Pleasanton have some light industry; however, neither are centers of specific intensive industry.

Interstate 680 runs within the 2.5-mile (4 km) area to the west of the VNC in approximately the north south direction. California Route 84 (Vallecitos Road) lies directly south of the VNC and accesses the facility entrance and runs SW to NE across the sites southern border. The VNC property is entirely on the north side of Vallecitos Road which is a two to four-lane paved highway currently being improved and widened in a California Department of Transportation project in cooperation with the Alameda County Transportation Commission (Reference 4). This project is expected to be completed over the next few years and will add a 4-way intersection and turn signal at the entrance to the VNC and widen the balance of Vallecitos Road to four lanes, which is expected to greatly reduce traffic congestion and improve commuter safety to and from the site.

A Central-Pacific and a Western-Pacific rail line also runs within the 2.5-mile area west of the VNC.

There are no military bases within 5 miles (8 km) of the VNC.

2.3. METEOROLOGY The Bay Area has a Mediterranean climate characterized by wet winters and dry summers. The VNC is situated in the Livermore Valley, a sheltered inland valley within the Diablo Range near the eastern border of the San Francisco Bay Area Air Basin.

During the summer, the Livermore Valley is characterized by clear skies and relatively warm weather with maximum temperatures ranging from the high 80s to low 90s (°F). Cold water upwelling along the coast and hot inland temperatures during the summer can cause a strong onshore pressure gradient, which translates into a strong afternoon wind. A high-pressure cell centered over the northeastern Pacific Ocean results in stable meteorological conditions and a steady northwesterly wind flow that keep storms from affecting the California coast.

During the winter, the Pacific high-pressure cell weakens, resulting in increased precipitation and the occurrence of storms with a mean winter precipitation of about 14 inches. Winter temperatures are mild and usually range from the high 30s to low 60s (°F).

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Violent storms are infrequent in this area with their primary consequence being the possible interruption of electrical service to the VNC from the Pacific Gas and Electric (PG&E) system (described in the Electrical Power section). Upon a loss of normal electrical service, a reactor shutdown is automatic when power is interrupted to the safety rod magnets. Safety rods are injected by springs so that the reactor is shut down and maintained subcritical independent of electrical power.

There is only a remote likelihood of major flooding at the VNC, but a higher possibility of substantial sheet flow caused by heavy rainfall and resultant runoff from the surrounding hills.

All roadways and facilities are constructed with drainage to preclude damage caused by such an occurrence. Surface water drains away from the Site facilities to several natural ravines and man-made channels which empty into Vallecitos Creek.

The VNC meteorology was studied in 1976 by Nuclear Services of Campbell, California. The summary of the results of their investigations is published in a separate document (Reference 1).

Further studies were performed by the U.S. Geological Survey. Results of these studies are published in a separate document (Reference 1).

2.4. HYDROLOGY The hydrology of the VNC was studied in 1976 by Nuclear Services of Campbell, California. A summary of the results of their investigations is published in a separate document (Reference 1).

Further studies were performed by the U.S. Geological Survey. Results of these studies are published in a separate document (Reference 1).

2.5. GEOLOGY, SEISMOLOGY, AND GEOTECHNICAL ENGINEERING The Vallecitos Valley and Vallecitos Hills along SR 84 lie within an inland valley of the Coast Range Geomorphic Province of Central California, a series of northwest-trending mountain ranges and intermountain valleys bordered on the east by the Great Valley and on the west by the Pacific Ocean. The area is filled with Quaternary deposits and includes stream channel deposits, floodplain deposits, and young alluvial fan deposits. Above the valley floor are older alluvial fan deposits that include stream terrace deposits in some narrow canyons and on the margins of the Vallecitos Valley. The alluvial terraces at the mid-level elevations of the rolling foothills north and south of Vallecitos Creek are older sedimentary deposits known as the Livermore Gravels.

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These deposits exist on the faces of higher elevations as the result of uplift of the Diablo Mountain Range and subsequent erosion.

Comprehensive geological and seismological studies have been conducted at and near the VNC in the years 1977-1980. The results of these studies (Reference 2) and 3) have been reported and submitted to the Nuclear Regulatory Commission in relation to the General Electric Test Reactor, also located at the VNC. The NTR design and the excess reactivity limit under which it operates makes off-normal condition evaluations (see Chapter 13) insensitive to geological conditions and seismological parameter values.

2.6. CONCLUSION While both weather conditions and access to population centers make the area around the VNC attractive, geological and sociopolitical factors, including encumberments on land by the Williamson Act and the Tri-valley Conservancy, will ensure the land remains extensively agricultural and that population density will remain sparse into the foreseeable future.

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3 DESIGN OF STRUCTURES, SYSTEMS AND COMPONENTS 3.1. DESIGN CRITERIA The NTR is designed to operate at steady power levels up to 100 kW (thermal). The rate of power change is controlled by limiting the positive reactivity insertion. This is accomplished by limiting the control rod drive speed so that an operator may safely change power levels. A fast-period-scram will prevent the reactor power increase from being uncontrollably fast.

Facility shielding has been provided and beam shutters have been interlocked to maintain personnel radiation exposure less than 10 CFR 20.1201 limits.

The NTR has an extended life core. While irradiated fuel is not stored at the NTR, the original fuel was installed in 1957. Core modeling performed in 2015 projects the core will continue to operate without modification until approximately 2025. Business needs will dictate the need and modifications necessary to extend life of the facility beyond 2025. Potential excess reactivity has been maintained within limits by periodically removing Manual Poison Sheet (MPS) cadmium to compensate for positive excess reactivity loss from fuel burnup. As of January 2023, all MPS have been removed except for 1/16th sheet that remains in position #2.

The limiting accident at NTR is a step-positive reactivity insertion of $0.76 while the reactor is operating at full power. Assuming the reactor fails to scram from a fast-period-trip or an overpower trip, fuel damage does not occur. Also, assuming the primary coolant system failed at the time of the positive reactivity insertion, fuel damage would still not occur even if the reactor fails to scram on a low flow condition.

Building 105 contains an automatic fire sprinkler system to suppress fire in the operating areas.

The reactor cell and the north room do not contain fire sprinklers. Fire extinguishers are available and on-site responders are trained to use them.

Combustible materials in these areas are administratively controlled. The reactor graphite has been analyzed and found not to contain enough stored energy for combustion.

The NTR does handle devices for neutron radiography. These devices are prohibited in some areas and are strictly controlled in other areas to prevent damage to the reactor 3-1

and the control rods and to prevent the spread of radioactive materials outside posted radiation areas.

3.2. METEOROLOGICAL DAMAGE The meteorological conditions at the site are generally very mild. The reactor and the control room are contained in structures which are extremely unlikely to fail.

External cooling water is not required to maintain fuel temperature below the fuel and the fuel cladding melting points, so the secondary cooling water system is not required since there is no credible reactor event that could cause fuel melt or a significant release of fission products from the fuel. In addition, the reactor confinement building would effectively limit small radioactive releases.

There has been no damage to the reactor systems caused by the wind over the history of the facility.

3.3. WATER DAMAGE Major flooding at the NTR is extremely unlikely. The historical data show that flooding does not occur in this area. The NTR is situated on an area which slopes so that the whole site can accommodate significant runoff. Drainage ditches have been added when the site was built to prevent runoff from entering buildings. Modifications made to Lake Lee ensure the immediate release of all water from the Lake will not affect site buildings.

Any flood water intrusion into the reactor cell or the control room would not cause a hazard. The Safety Rods and Control Rods and the reactor core can are located two feet above the floor and any shorting or grounding of electrical systems would immediately scram the reactor.

3.4. SEISMIC DAMAGE California is a seismically active state. The San Francisco Bay Area contains many active faults, some near the Vallecitos Nuclear Center.

Active and potentially active faults in the area are the Calaveras, Verona, Pleasanton, Las Positas, Greenville-Marsh Creek, Mount Diablo Thrust, Hayward, Concord-Green Valley, and San Andreas faults. These major faults are considered capable of causing fault rupture or substantial ground shaking. The Verona Fault (considered part of the Pleasanton Fault) runs approximately 3-2

north to south just to the east of the VNC. The maximum moment magnitude (the largest earthquake that a given fault is mathematically capable of generating) for the Verona Fault is 6.6 (Reference 4).

Bounding seismic analysis performed in 2019 (Reference 2) confirms the 1980 analysis (Reference 3) that there is little risk of collapse of structural buildings at the VNC site.

Nevertheless, even if catastrophic failure of the NTR facilities is assumed, there are no potential consequences resulting in fuel melt or gross dispersal of radioactive material. Compaction of the fuel, while essentially impossible mechanistically, would not cause the reactor to go critical since water loss, increased self-shielding in the fuel and the geometry change due to flattening of the cylindrical core are all negative reactivity effects. Also, deformation of the core causing fuel to contact the core-can structure would improve heat-transfer and result in lower LOCA temperatures.

If a large seismic event occurs, it may be hypothesized that certain structures used to support the control and safety rod mechanisms as well as experiments might fail or move in such a manner as to withdraw the control rods and experiments from the core region and prevent operation of the safety rods. The cadmium poison sheets are manually positioned entirely within the graphite reflector, have no drive mechanisms, and will not move relative to the core during a seismic event. And, as previously noted, only 1/16th of a manual poison sheet remains. In order to preclude any mechanism which could lead to fuel melting, operation of the NTR is conducted in such a manner as to limit the reactivity available from control rods and experiments to less than or equal to $0.76. The analysis of a step reactivity addition without scram is analyzed in Ch 13 and will not result in fuel melting or release of fission products.

Given the inconsequence of any catastrophic failure of structures and systems at the NTR, major structures and systems have nevertheless been structurally analyzed and determined capable of surviving a 0.8g vibratory ground motion. These structures and systems include the following:

1. The reactor cell and roof.

2. The lead shield wall on the north side of the reactor graphite pack.

3. The Safety Rod and Control Rod support structure.

4. The reactor cell bridge crane.

5. The fuel loading tank.

6. The concrete shielding slab on top of the reactor graphite pack.

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3.5. SYSTEMS AND COMPONENTS The NTR is a low temperature, low heat reactor which is very forgiving of off normal and accident conditions. The NTR is operated in such a manner as to limit the potential excess reactivity to less than that required to cause fuel damage, assuming a failure to scram. This is accomplished by manually positioned cadmium poison sheets. The Manual Poison Sheets (MPS) are positioned entirely within the graphite reflector, have no drive mechanisms, and are mechanically restrained so they will not move relative to the core during a seismic event. The MPS are verified to be latched each time they are installed, and the potential excess reactivity is verified to be below limits at each reactor startup.

The worst loss-of-flow accident (instantaneous seizure of the rotor in the single recirculation pump in the system) does not result in fuel damage or release of fission products even with a failure to scram and the reactor operating at 100 kW.

The total loss-of-coolant inventory in the core as the result of a rupture in the primary system with the reactor operating at 100 kW combined with a failure to scram does not result in fuel damage or release of fission products.

The reactor fuel was fabricated in accordance with GE Specification AP-RG-56-8-1.1, dated October 23, 1956. This specification includes material type, dimensional tolerances, a helium leak test, and a corrosion test.

The reactor is installed in a room (the ) located in the Building 105. Figure 3-1 shows an approximate floor plan of Building 105, and Figure 3-2 shows plan and elevation views of the area that contains the NTR facility,

. These illustrations are not necessarily current with respect to the arrangement of the office and laboratory areas. Other details that are not pertinent to safety considerations for the NTR facility may not be as shown.

The reactor cell is a rectangular-shaped room with approximate internal dimensions of 22 ft (6.7 m) wide by 23 ft (7 m) long by 24 ft (7.3 m) high.

. Equipment of appreciable size located within the cell includes the reactor, reactor cooling system, fuel loading tank, holdup tank, and storage shelves. Approximate 3-4

gross volume of the cell is and, with the above-mentioned equipment installed, the net air volume is approximately 10,500 ft3 (300 m3).

Normal access to the cell is through the large doorway in the wall. During reactor operation, the doorway is normally closed The refueling of the reactor and the maintenance of equipment in the cell are performed only with the reactor shut down (i.e., sufficient manual poison sheets and safety rods inserted to satisfy shut-down margin requirements). Normally, the radiation and contamination levels are quite low; therefore, these activities can be performed with the cell door open. If expected radiation levels, results of radiation monitoring, or some other nonroutine nature of these activities makes closing the door desirable, either maintenance or refueling may be done with the door closed.

Although the reactor cell is not designed to provide gas-tight containment, in conjunction with the stack ventilation system, it helps maintain confinement in the event of a fueled experiment failure. In this scenario, the reactor cell would aid in containing any radioactive release while it is exhausted out the stack at between 1000 and 3000 cfm through the ventilation systems 99%

efficient filters.

The reactor is not operated above 0.1 kW unless the ventilation system is operating. A differential pressure alarm in the control room informs the operator prior to control room pressure dropping below that of the reactor cell and reactor power is reduced to 0.1 kW.

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Figure 3-1 Building 105 Floor Plan 3-6

Figure 3-2 Part Floor Plan and Elevation 3-7

3.5.1. Confinement For discussion, the penetrations into the reactor cell have been divided into two types - reactor and experiment services. Reactor service penetrations normally are used for passing water, electric power, and air into and out of the cell; their use generally does not change from day to day. Experiment service penetrations normally are used to move materials and equipment for experiment programs into the reactor cell. The list of penetrations below is the present arrangement. Future changes in the number, location, or type designation of any of the penetrations will be made in a manner to retain the effectiveness of the cell to control radiation, contamination and security.

, the existing reactor service penetrations include:

One secondary cooling water supply line; One deionized water supply line; One line to the Site retention basins. This drain has been disconnected and plugged.

Twelve electrical conduits for wiring between the cell and control room; Eight small-diameter ( ) electrical power and lighting conduits, including spares; One compressed air supply line; Two ventilation exhaust ducts through the cell roof; Four pipes to the adjoining laboratory; and Two holes approximately Existing experiment service penetrations are listed below. Use of these penetrations is controlled carefully to ensure that the effectiveness of the cell to contain radiation and radioactive materials is not significantly reduced.

South thermal column into .

Horizontal facility tube into .

A hole through the wall at approximately core centerline height

(

).

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Stepped hole ( ) through the wall approximately above the floor.

Hole for future thermal column through the wall (

).

Two holes ( ) in the wall, accessing the room.

The hole contains radiation area monitor cables and the hole is used for the CHRIS experiment facility.

3.5.2. Ventilation The NTR exhaust system (Figure 3-3) includes a 3,000-ft3/min fan located on the cell roof. The fan draws air from the cell, south cell (sample sink) and the room

. The air goes through a prefilter and a bank of absolute filters and is then discharged through a stack of adequate height to disperse the exhaust upward.

An air monitoring system provides continuous indication of the concentration of radioactive material in the ventilation effluent and energizes an alarm at the if the concentration reaches a set point which has been selected to ensure that the airborne release does not exceed established limits. Stack Release Action Levels have been established according to Chapter 11 (Table 11-4). Separate detection channels and alarms are used for particulate material and for nonfilterable radioactive gases. A continuous sample is drawn from the discharge of the NTR ventilation stack and passes through the particulate detector, a charcoal cartridge, the nonfilterable radioactive gas detector, flow control valve, and a central blower (Hoffman). It is then released through the building 105 NTR Furnace Exhaust. Particulate materials are collected on a high-efficiency filter paper and their emissions measured with a shielded Geiger-Müller detector. Nonfilterable radioactive gases are detected by an internal gas flow ionization chamber with a relatively high sensitivity for beta emitters. Current from the chamber is measured by a pico-ammeter. Each channel is recorded on a multipoint recorder. The charcoal cartridge and particulate filter are changed periodically (normally weekly) and counted by the VNC Counting Lab for I-131 and gross - and , respectively.

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Figure 3-3 Line Diagram of Ventilation System 3-10

4 REACTOR DESCRIPTION 4.1

SUMMARY

DESCRIPTION The NTR is a light-water-cooled, high-enriched-uranium, graphite-moderated and -reflected, thermal reactor with a nominal power rating of 100 kW.

4.2 REACTOR CORE Figure 4-2 is an assembly drawing of the present reactor fuel container. This container was put into service in 1976 after the previous container, which had been in service for approximately 18 years, developed a leak in a weld area. The annular ends of the container, 0.5-inch aluminum plates, are welded to the inner and outer cylindrical skins, which are rolled aluminum sheets 0.25 and 0.0625 inch thick, respectively. The outer cylinder is made from two pieces welded together opposite the loading chute. Attached to the inside surface of each end plate is an aluminum circular raceway which supports and guides the core reel assembly. The reel assembly, in turn, supports the fuel assemblies. The core reel assembly is described in more detail in Section 4.2.5.

Openings are provided in the north end plate for the 1.5-inch primary coolant inlet and outlet lines. The inlet pipe is connected to a flow-distributor tube located inside the container below the core. A row of 25 0.25-inch holes is drilled into the lower side of the 1.375-inch flow-distributor tube, with the holes near the core midplane closer together than those at the ends to distribute water flow to correspond to power distribution along the core. The center-to-center distance between the 10 holes nearest the midplane is 0.4375 inch; the next three holes, toward each end, are on 0.5-inch centers; the next two holes on 0.75-inch centers; and the last two holes on 1-inch 4-1

centers. An identical baffle tube located above the core (with holes on the top side of the tube) is connected to the outlet line.

A 3.25-inch opening in each end plate accommodates the drive mechanism for the reel assembly.

These openings are in the area just below the junction between the loading chute and the fuel container. This drive mechanism is discussed in Subsection 4.2.5.

Figure 4-1 Vertical Section Through the NTR 4-2

Figure 4-2 Fuel Container Assembly 4-3

Attached to the outer wall of the can, inclined upward at an angle of about 30 degrees with the horizontal, is a rectangular aluminum loading chute about 30 inches long, 20 inches wide, and 3 inches high. The chute is connected to the fuel loading tank. Slotted adapters fastened inside the chute provide a guide for the chute plug. The slots in the adapters line up with radial slots in the two circular raceways to guide the fuel loading tool to the core reel during refueling operations.

When not in use, the loading chute is filled with an aluminum-clad graphite plug, and the chute opening is at least partially covered with an aluminum gate located in the storage tank.

Eight 0.75-inch aluminum tubes supported from brackets attached to the end plates run horizontally along the outside surface of the fuel container to the north face of the reactor. These tubes are guides for the control, safety, and neutron source rods. Six slotted graphite ways attached to the north end plate, oriented parallel to these guide tubes, serve as guides for the manually positioned poison sheets. The only other attachment to the container is a bracket fastened to the south end plate to help support the 5-inch horizontal facility.

4.2.1 Reactor Fuel There are 16 fuel assemblies in the NTR core. Each fuel assembly consists of 40 fuel disks and spacers skewered on a shaft to form a shish kabob-type assembly. Lateral motion of the disks and spacers on the shaft is prevented by lock nuts placed on both ends of the shaft. All available spaces in the core support reel are filled by the 16 assemblies. There are 640 fuel disks in the core. These 640 fuel disks contained approximately 3.3 kg U-235 initially.

. All the fuel disks in the core are from the original fuel load fabricated in 1957. The fuel was fabricated in accordance with GE Specification AP-RG-56-8-1.1, which included corrosion and helium leak testing.

When the fuel container was replaced in 1976, the fuel was removed, inspected, and leak checked. No cleaning, replacement, or repair was necessary.

Each space between the disks contains a 0.180-inch-thick aluminum spacer, and an additional 0.031-inch-thick aluminum washer is alternately in every other space. This arrangement produces an assembly with an active length of approximately 15.25 inches with the face-to-face distance between disks alternating between 0.24 and 0.27 inch.

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Each fuel disk (Figure 4-3) is composed of a fuel bearing, flat, doughnut-shaped sandwich and an inner and outer edge ring. The three pieces were brazed together to clad the uranium-aluminum alloy meat. The sandwich consists of the uranium-aluminum alloy meat, which contains, on the average, depletion (for NTR operation as of 9/30/2019 at approximately 198.5 MWD) as a 23.5 wt. % alloy, plus 0.027 inch of 1100-series aluminum cladding on each face. This type of fuel was the industry standard for many years. The finished, flat, doughnut-shaped sandwich is 0.142-inch-thick and has a 2.68-inch OD, with a 0.58-inch-diameter center hole. The inner edge ring (0.516-in. inside diameter (ID), 0.033-inch-thick, and 0.20-inch-wide) fits into the center hole of the disk and is brazed to the faces of the sandwich cladding. A 2.75-inch OD outer edge ring, with the same width and thickness as the inner ring, fits around the circumference of the disk and is also brazed to the faces of the sandwich.

Figure 4-3 Fuel Disk A 0.75-inch length of each end of the 0.5-inch aluminum support shaft is machined to provide a tip suitable for supporting and positioning the fuel assembly accurately in the core reel.

Tolerances on the shafts, reel, and fuel container were set so that the maximum radial and circumferential movements of the shaft, and hence a fuel assembly, are less than 0.125 and 0.016 inch, respectively. The support tip extends past the ends of the core reel about 0.375 inch into the raceway; it is this section of the fuel assembly that is engaged by a tool during fuel handling.

4.2.2 Control and Safety Rods Movable neutron absorbers located about the periphery of the fuel container include:

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(1) Two boron-carbide-filled motor-driven coarse control rods; (2) One boron-carbide-filled motor-driven fine control rod; (3) Four boron-carbide-filled safety rods, motor-driven carriage with an electromagnet that attaches to the poison section, and scram force by cocked springs.

(4) Reactivity worth of control and safety rods are:

Control Rods Safety Rods Coarse Rod #1 $0.59 Safety Rod #1 $0.82 Coarse Rod #2 $1.01 Safety Rod #2 $1.10 Fine Rod $0.60 Safety Rod #3 $1.07 Safety Rod #4 $0.87 Orientation of these rods about the core is shown in Figure 4-1, and all run-in guides extend from the south end of the fuel container through the reflector and shield to the north face of the reactor.

The guides place the center of the poisons on a 9.5-inch radius or about 0.6 inch from the outside edge of the active core. The control and safety rods have horizontally mounted drive mechanisms that are supported from the north face of the reactor on a 5-foot-high aluminum support plate located about 4-1/2 feet in front of the north face. Both types of poisons were designed to perform a specific function.

The four safety rods were designed for rapid insertion to scram the reactor. The control rods (two coarse and one fine) were designed for the precise position control and indication required for analytical work during which the reactor is used as a detector.

Relative positions of the four boron-carbide-filled safety rods are shown in Figure 4-1. The poison section of each rod is 20 inches long and consists of a solid core of 1/2-inch-diameter boron carbide cylinders contained in a stainless-steel tube. A plug in the north end of the stainless-steel tube connects to an extension rod which has a rod stop armature assembly pinned to the other end. Two constant force spiral springs are attached to the extension rod so that withdrawal of the safety rod cocks the springs to store energy. Housings for the springs are secured to a support bracket attached to the north shield face. Also attached to this support bracket is a long piece of steel angle which passes through and is attached to the aluminum support plate to protect and support most of the hardware for the drive mechanism.

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The safety rod is held to the rod drive by an electromagnet that engages the armature attached to the extension rod. The electromagnet is attached to a drive nut that moves horizontally on a lead screw. Rotation of the lead screw is accomplished with an electric motor through a belt and pulley drive. The electric motor is a 1/12 hp, capacitor start and run, 57-rpm output, gear motor connected for instantaneous reversing and is provided with automatic reset overload protection.

Power to the motors is from the 115-Vac supply. Remote manual control by the operator is by pushbutton switches at the console. A circuit is provided to run the carriage automatically to the fully inserted position following a reactor scram, provided ac power to the console is maintained.

The sequence of operations to withdraw a safety rod is as follows; Run carriage in to engage armature; Energize the electromagnet; and Run carriage full out to withdraw safety rod and cock the constant force spiral springs.

Upon initiation of a scram signal, the following sequence takes place:

Scram signal deenergizes the electromagnets; Constant force spiral springs cause the armature to separate from the magnet and rapidly insert the safety rods; and Automatic signal runs all rod carriages to the fully inserted positions.

Deceleration of each scrammed safety rod is accomplished by an air dashpot-type shock absorber. The rod-stop armature begins to compress the air in the shock absorber housing about 4 inches from the full-in position. An orifice is provided to control the release rate of the compressed air and the deceleration rate of the rod.

Four microswitches are associated with each safety rod and drive mechanism. Listed below are the actions initiated by each switch:

(1) Drive-Out Limit Switch Interrupts motor circuit at outer limit of rod stroke.

Energizes yellow light at console.

Interlocked so that all safety rods must be withdrawn sequentially before any control rods may be moved, except when rod test panel is utilized.

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(2) Drive-In Limit Switch Interrupts motor circuit at fully inserted limit of stroke.

Energizes green light at console.

Interlocked to prevent energizing the electromagnets unless all control rods and neutron source are fully inserted.

(3) Safety-Rod-In Position Switch Energizes green light at console.

(4) Separation Switch Operates in series with carriage-out limit switch to energize yellow light, indicating that rods and carriage are out. This interrupts voltage in the safety system and deenergizes the safety rod electromagnets.

The scram mechanisms for the NTR are essentially the same as those that operated satisfactorily on the TTR reactors for many years. The present mechanisms have operated without showing appreciable wear. As required by the administrative procedures, scram time for the rods is measured periodically, and, if it is found that a rod does not meet the required insertion time, the rod will be considered inoperable until repairs are made.

As a result of the lack of symmetry in the arrangement of the nuclear poisons around the core and the possibility of strong shadowing effects, the reactivity worth of the individual safety rods vary. In the normal core, the reactivity worth of the most effective safety rod is about $1. The conservatively assumed worth of all four safety rods is $2. However, the total safety rod worth value, shown in Section 4.4.3, is larger than $2. If it is assumed the entire worth of the rod (for conservatism, use $1.5) is realized in the first 18 inches of rod movement at the withdrawal speed of 1 inch/second, the average reactivity addition rate by withdrawal of the most effective safety rod is only $0.083/sec, which is a reasonable addition rate for manual control. Since the actual full stroke of the safety rod is approximately 28 inches and the rods are interlocked so that each rod must be fully withdrawn before the next one can be started out, the reactivity in two safety rods cannot be added to the reactor in less than approximately 1 minute by normal withdrawal.

Figure 4-1 shows the location of the two coarse control rods with respect to the core and other neutron poisons. The poison section of each rod is 16 inches long and consists of a solid core of 1/2-inch-diameter by 2-inch boron carbide cylinders contained in a stainless-steel tube. A plug 4-8

in the north end of the tube is connected to an extension rod which is attached to a yoke that is positioned by the drive mechanism.

This yoke is fastened to a lead screw that runs through a sprocket and nut assembly connected through a chain drive to a gear motor. The gear motor is identical to the one described above for the safety rods. Power to these motors is supplied by the 115-Vac supply. Pushbutton switches at the console permit manual control. As with the safety rods, scram provides an automatic signal which takes control away from the operator and runs the rods to their fully inserted position, provided ac power to the console is maintained. Position indicators are provided to indicate the position of each rod, to the nearest one hundredth inch, over the full stroke for rod movements in either direction.

Two limit switches on each control rod drive mechanism perform the following functions:

(1) Rod-In Limit Switch Energizes green light at the console.

Interlocked to prevent energizing the electromagnets unless all control rods (in addition to the safety rods and neutron source) are fully inserted.

Interrupts the motor circuit at the full in position.

(2) Rod-Out Limit Switch Energizes yellow light at the console.

Interrupts motor circuit at the outer limit of the stroke.

Location of the fine control rod with respect to the other poisons is shown in Figure 4-1. The poison section is 18 inches long and consists of a solid core of 0.365-inch-diameter by 2-inch boron carbide cylinders contained in a stainless-steel tube that extends through the north face of the reactor to the drive mechanism. An aluminum rod is used to fill the remainder of the tube between the boron carbide cylinders and the drive mechanism.

The stainless-steel tube containing the poison connects to a nut block that travels on a lead screw. The lead screw is rotated through a right-angle gear box by a gear motor. The motor is 1/2-hp, 1725-rpm, (173-rpm output) with an electrically operated brake. Power to the motor is supplied by the 115-Vac supply. Pushbutton switches on the console are used for remote manual control.

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The fine control rod is automatically driven to the fully inserted position following a scram, provided ac power to the console is maintained.

Two limit switches actuated by the traveling nut block perform the identical functions discussed above for the control rods. Position indicators are provided to indicate fine rod position to the nearest one hundredth inch over the entire stroke for rod movement in either direction.

The control rods were calibrated using the rising period technique, with the rods in an essentially unshadowed condition. The results indicate the total worth of all the rods is approximately $2.3. The speed of withdrawal of each coarse control rod drive is 0.140 inch/second. The speed of withdrawal of the fine control rod drive is 0.145 inch/second. If it were assumed that all three rods were withdrawn simultaneously, the average reactivity addition rate would be approximately $0.02/sec, which is an amount that is easily manageable.

There is a third type of control designed to change reactivity, but the change does not take place during reactor operation. These are manually positioned poison sheets and are used to limit the reactivity available to the operator or to increase the shutdown margin. The manual poison sheets (MPS) are inserted or removed manually, during shutdown, in the graphite around the fuel container (Figure 4-1), through access holes provided in the north shield.

The sheets consist of 0.032-inch-thick by 19-inch-long cadmium laminated between two 0.08-inch-thick by 3-inch wide by 40 1/2 -inch-long 6061-T6 aluminum plates. The width of the cadmium in each sheet is as follows:

(1) Full sheet: 2.75 inches (2) 3/4 sheet: 2.06 inches (3) 1/2 sheet: 1.38 inches (4) 3/8 sheet: 1.03 inches (5) 1/4 sheet: 0.69 inches (6) 1/8 sheet: 0.34 inches (7) 1/16 sheet: 0.17 inches All manual poison sheets used are equipped with a spring-loaded latch handle that latches to a special latch plate on the north face of the aluminum box that contains the graphite reflector assembly. This latching assembly provides positive restraint of the manual poison sheets with 4-10

respect to the reactor assembly. The reactor shall be operated using only those manual poison sheet positions that have the latches installed - currently positions 1, 2 and 5.

The manual poison sheets do not have a drive mechanism, or any automatic functions associated with them. A sheet can only be inserted or withdrawn by entering the reactor cell to remove a shield plug in the shield face. The sheet is then positioned by engaging the sheet handle with a special latching tool and physically unlatching or latching it prior to removal or full insertion.

The removed sheets are stored in a rack in the reactor cell and are accounted for before reactor startup.

Reactivity worth of individual manual poison sheets in the core with all safety rods inserted were obtained by utilizing a pulsed neutron source. The worth of a full sheet was historically determined to be approximately $1, and a half sheet was worth about $0.5. In a typical core configuration (1/16 MPS in slot #2), the worth of the MPS is approximately $0.17 (subject to change based on experiment worth, etc.). Excluding the transient temperature worth (reactivity addition from the primary coolant temperature change) and the experiment transient worth, the typical excess reactivity available from the control rods at the console is $0.3 (subject to change based on experiment worth, etc.).

For NTR operation in July 2020, the worth from changing MPS size from the single installed 1/8 sheet in slot #2 to a 1/16 sheet in slot #2 was determined via measurement to be approximately $0.13. This MPS size change reactivity worth was determined with control rods in the critical position, safety rods fully withdrawn, the neutron radiography source log inserted, and a 1.0 °F reduction in primary coolant temperature (from 77 °F to 76 °F) between the pre-change and post-change conditions. Reactor modeling predicted this MPS change to result in a

$0.17 reactivity change.

4.2.3 Neutron Moderator and Reflector The graphite reflector-moderator is a 5-foot cube of reactor-grade Acheson Graphite Ordinary Temperature (AGOT) graphite which not only serves as a neutron reflector and moderator, but also physically supports the fuel container. The fuel container is centered in the reflector with the core cylindrical axis horizontal. Many small pieces (primarily 4 inches by 4 inches by varying lengths) were machined carefully and stacked together to form the 5-foot cube. The reflector is contained and supported by the aluminum box and base discussed in Section 4.2.5.

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Among the numerous items penetrating the reflector are 1) the fuel loading chute through the west face, 2) the control rod, safety rod and neutron source guide tubes, 3) the manually positioned poison sheet slots, 4) the cable held retractable irradiation system and 5) the core reel drive shaft through the north face. The horizontal facility tube and the experiment tube traverse the reflector from the north to the south face and continues through the thermal column.) The vertical facility extends from the top to the bottom of the reflector. Chapter 10 contains a discussion of the vertical and horizontal facilities and the thermal column. Several modifications were made to the main graphite pack during the major outage of 1976 that enable the removal of two special sections of the reflector: the set of blocks situated between the fuel container and the north face; and the group of blocks that fill the 11.5-inch- diameter hole formed by the inner skin of the fuel container. Removal of these sections makes it possible to inspect the fuel container without disturbing the rest of the reflector.

4.2.4 Neutron Startup Source A reactor start-up neutron source is installed on an electric motor drive mechanism, in a configuration like that of the control rod drives. The source drive has the same controls and indications as a control rod drive, with the exception that continuous position indication is not provided. The same interlocks as those on the control rods are provided (the safety rod magnets cannot be reenergized until the source is full in), except that it is not necessary to pull any safety rods to withdraw the source. Following a scram, the source automatically runs to the fully inserted position. The source travels in a guide tube identical to that used for the control rods, and the limit switches are adjusted so that the source moves about 30 inches from the full-in to full-out positions. A 0.2-Ci radium-beryllium source emitting about 106 n/sec is used for a startup source. It is an R-Monel encapsulation approximately 1/2-inch in diameter and 3-1/2 inches long, attached to an aluminum extension rod that connects to the source drive mechanism. The source-detector arrangement provides at least the minimum neutron flux signal required for the nuclear instrumentation for startup and also gives good indication of subcritical multiplication.

4.2.5 Core Support Structure The fuel container rests on the sections of the 5-foot graphite cube pack beneath it. The graphite was machined to close tolerance, then fitted around the container and loading chute to provide 4-12

maximum support. An aluminum box constructed of 0.375-inch plate and 2-inch angle aluminum contains the 5-foot graphite cube on all faces except the bottom and the south face.

The south face is joined to the 4-foot graphite cube thermal column. A 0.031-inch cadmium liner is provided for the north and east sides of the box. The box containing the graphite cube rests on a base consisting of a 0.625-inch-thick aluminum plate fastened to a framework of 5-inch aluminum I-beams. The I-beam base is clamped to steel support plates anchored to the reactor cell floor. At the time the reactor was installed, the base was shimmed level and grouted.

A 0.75-inch length of each end of the 0.5-inch aluminum support shaft is machined to provide a tip suitable for supporting and positioning the fuel assembly accurately in the core reel.

Tolerances on the shafts, reel, and fuel container were set so that the maximum radial and circumferential movements of the shaft, and hence a fuel assembly, are less than 0.125 and 0.016 inch, respectively. The support tip extends past the ends of the core reel about 0.375 inch into the raceway; it is this section of the fuel assembly that is engaged by a tool during fuel handling.

Located within the fuel container can is the core reel assembly, which consists of a pair of spur gears tied together with eight separator bars. Radial slots in these spur gears receive the machined tips of the fuel assembly shafts to support and position each fuel assembly. Stainless steel rollers attached to the outer face of each gear guide and support the reel in the radial raceways attached to the fuel container end plates (Figure 4-2). A reel drive mechanism is provided to rotate the entire reel assembly to any desired position with respect to the loading chute.

The two large spur gears are almost identical. These gears, made of 0.34-inch-thick aluminum, have a 16.3-inch outside diameter (OD) and a 12.9-inch inside diameter (ID). Eight stainless steel rollers are bolted to the outer face, and eight triangular-shaped separator bars are bolted to the inner face (through a 0.75-inch-thick spacer ring) of each gear. The roller and raceway on the north end are V-shaped to prevent lateral motion of the entire reel. Sixteen equally spaced slots, 0.189-inch-wide, are cut into each gear to receive the machined tips of the fuel assembly shafts. These radial slots terminate at an inner radius that places the center of fully inserted fuel elements on a 7.48-inch radius.

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The reel drive assembly consists of two pinion gears (3-inch OD by 0.34-inch-thick) keyed to a single 0.625-inch shaft. The shaft seal is a double O-ring seal with a tattle-tale petcock. Outside the reactor shield the shaft extends through a right-angle gear box, to the top of the fuel loading tank, to a hand-operated drive wheel with a dial indicator. The dial indicates the orientation of the reel and the position of any fuel assembly. The reel may be rotated to any desired position for core work; once the work is complete, the reel is no longer moved. Movement of the reel assembly is permitted only when the reactor is shut down. Since the reel can be rotated only from within the reactor cell and is locked in position, unauthorized or unintentional movement during reactor operation is not considered credible.

4.3 BIOLOGICAL SHIELD Reactor shielding is such that personnel radiation exposures throughout the building can easily be maintained within established limits with the reactor at full power. A list of typical levels in and around the facility, with the reactor at 100 kW, is given at the end of this section. During initial operations under new conditions, radiation dose rates and personnel exposures are closely monitored. If required, modifications are made to the shielding or the procedures to ensure continued compliance to established limits and consistence with As Low As Reasonably Achievable (ALARA) practices.

Reactor Shield At present, radiation shielding for the reactor includes the graphite and the cadmium-lined aluminum frame which were discussed previously, local shielding The arrangement of most of these materials can be seen in Figure 1-1, Figure 10-1, and Figure 10-2.

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Reactor Cell Section 3.5.1 contains a detailed description Radiation levels listed in the last part of this section demonstrate the effectiveness of the cell as a radiation shield. Whenever new operating or maintenance conditions are encountered, radiation surveys are made to determine that existing shielding is adequate and consistent with ALARA practices. Either temporary or permanent improvements in the shield are made if the results of the survey indicate they are necessary.

South Cell The main source of radiation in the cell is that which comes directly from the reactor and that which is induced in experiments and experimental equipment. As shown in Figure 1-1, 4-15

Shielding material has also been added above the cell entry ceiling and adjacent to the penetration through the east wall of the south cell.

In 1988 two-inch thick 7% boron-polyurethane sheets were added to the south cell ceiling and an 8-inch-thick high-density concrete shield wall 38-inches wide was built in the south cell doorway to reduce the dose rate in the control room.

Radiation coming from the reactor is reduced by the presence of a 4-foot-thick graphite thermal column. In front of that is approximately a 3 1/2 -foot-thick x 7-foot-wide density concrete block wall that stairsteps from 4-feet to 8-feet high and a 4-inch lead brick wall. A thick shutter consisting primarily of lead and borated polyethylene is operated from the control room and shields radiation from the horizontal cavity. An interlock alarm system is provided which:

- Prevents opening the door if higher than normal levels of radiation are present.

- Initiates automatic closure of the shutter to reduce radiation levels.

Initiates audible and visual alarms to warn personnel of higher-than-normal radiation levels.

In addition, a photo-cell alarm is provided at the south cell door access point which will sound whenever the light beam is broken when it is armed.

Room Modular Stone Monument (MSM)

The MSM, which is discussed in Chapter 10 (Figure 10-2) provides shielding in the north room from:

(1) The north neutron radiography beam; and (2) Radioactive objects during the neutron radiographic process.

The MSM is made of high-density concrete modular blocks (so design changes may be made easily in the future) and houses a borated lead polyethylene beam catcher. Additional lead shield closures may be utilized, as required, to further reduce the radiation from two of the penetrations, as shown in Figure 10-2.

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Radiation Levels A list of typical radiation levels in the areas of the NTR facility, while the reactor is operating at 100 kW, is given below. Unless shielding changes are made, the listed radiation levels are all proportional to reactor power. The values listed include contributions from fast, intermediate, and slow neutrons and gamma rays.

Shutters* Shutters*

Location Open Closed mRem/h mRem/h At reactor console 3.0 1 Hallway south of control room 1 1 Building 105 equipment room ( ) 1 1 cell roof directly above reactor (top shield slabs in place) 65 65 cell sample sink 57 2.5 Setup Room 1 1 Room (center of room) 5.5 1

cell horizontal cavity shield shutters 4.4 NUCLEAR DESIGN 4.4.1 Normal Operating Conditions The reactor consists of a core in the form of an annular cylinder that contains 16 fuel assemblies as discussed in Sections 4.1 and 4.2. The core is centered in a 5-foot cube of AGOT-grade graphite. Arrayed around the outside of the fuel container are four safety rods, three control rods and up to six manual poison sheets. All fuel assembles, control rods and safety rods are in fixed positions that are not changed.

Normal operation of the NTR is at powers no greater than 100 kW, with maximum temperature and pressure in the core at 150°F and 20 psia, respectively.

As a result of these very conservative operating conditions, none of the nuclear characteristics (except the water moderator temperature coefficient of reactivity) varies significantly with normal temperatures.

The reactor configuration is controlled to ensure that the potential excess reactivity is less than or equal to $0.76. The NTR burns ~$0.03 positive excess reactivity per year. The planned core configurations during the life of the reactor are to remove enough cadmium from the remaining 4-17

manual poison sheet to maintain normal operation and still ensure that the potential excess reactivity is less than or equal to $0.76.

Low power generation of the NTR makes reactivity changes from fuel burnup and fission product poisoning small. Since initial criticality, the reactor has accumulated approximately 198.5 MWD of operation. Based on this history (for NTR operation as of 9/30/2019), the total reactivity losses are estimated to be $2.1 from the fuel burnup, $1.8 from aggregate fission product poisoning (other than samarium-149), and $0.62 from samarium-149 poisoning (at equilibrium). Plutonium buildup in the NTR is negligibly small.

Selected reactivity worths of reactor components are listed in Section 4.4.2.

The following are administrative and physical constraints that prevent inadvertent addition of positive reactivity.

During reactor operation the reactor cell door is locked so that core changes are not possible. The manual poison sheets are physically latched and cannot move during operation. During operation, then, the only positive reactivity additions possible are from moveable experiments, coolant flow changes and movement of control and safety rods. Control and safety rods are manipulated by licensed operators in accordance with written procedures. These reactivity additions are limited physically (water coefficient of reactivity) and by design (control and safety rod drive speeds and experiment reactivity worth).

Entry into the reactor cell, when the reactor is critical, is authorized by special procedure (Engineering Release) describing the operation to be performed. This procedure must be approved by the Manager, NTR, and reviewed by the Manager, RC, prior to entry.

When the reactor is operating, there is a cell door/shutter interlock to prevent inadvertent entry into the cell with the cell shutter open. An electric photocell light mechanism causes an audible alarm to actuate when an entry into the south cell is made.

Entry into the cell is permitted when the reactor is critical if the power is stable, the entry does not distract the operator and no more than the minimum number to safely perform the task is permitted.

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During shutdown, positive reactivity changes are possible by safety and control rod movement, manual poison sheet movement and horizontal facility changes.

Safety and control rod movement defines the reactor as not secured. When the reactor is not secured the minimum staffing is composed of the following: A licensed operator in the control room. A second person at the site familiar with NTR emergency procedures and capable of carrying out facility written procedures. A licensed SRO shall be present at the facility (Bldg. 105) or be capable of reaching Bldg. 105 NTR within a reasonable time (1/2 hour / 30-mile radius) under normal conditions.

A licensed SRO shall be present at the NTR facility during manual poison sheet changes.

Each individual irradiation, in the horizontal facility, shall be reviewed to determine that the irradiation satisfies the approved irradiation criteria. This review is documented by the signature of an SRO.

4.4.2 Reactor Core Physics Parameters Several important features of the NTR that affect the nuclear characteristics result from an effort to enhance the performance of the reactor as a sensitive detector of reactivity changes. Among these features are the low critical mass, the fuel-to-sample geometry, and sensitive control system. The reactor is constructed so that samples placed in the horizontal facility are in a neutron flux that is higher than the flux in the fuel lattice. The sensitivity of the reactor as a detector is proportional to the ratio of the thermal flux at the sample to that in the fuel lattice.

Several nuclear parameters are listed in Table 4-1. The reactivity parameters in Table 4-1 are based on an average effective delayed neutron fraction () of 0.00704.

Figure 4-4 and Figure 4-5 illustrate the thermal neutron flux profiles in the horizontal facility and in three of the manual poison sheet slots. The profiles in the manual poison sheet slots are the profiles of the three upper slots on the east side of the core and are expected to correspond very closely to the axial neutron flux and thermal power distribution in the adjacent section of the core.

Discussions of temperature coefficients of reactivity usually separate the total coefficient into a nuclear cross section effect and an effect caused by density and volume changes in the system.

These two major effects are subdivided further according to the location of material that is affected (i.e., fuel, moderator, or coolant) and the speed with which the effect occurs. For an 4-19

NTR-type reactor, such a complete breakdown is not necessary. By far, the dominant effect for accident analysis is that of density changes, including displacement of cooling water by expansion of fuel within the fuel annulus. Although the results of earlier studies indicate that a positive effect may result from heating the reflector graphite, this temperature change would be too slow (on the order of minutes) to affect a nuclear excursion significantly. The effect from a temperature change in the fuel annulus is observed in fractions of seconds. The over-all temperature coefficient of the fuel annulus was measured and found to be positive up to 124°F.

As temperature is increased above 124°F (the turnover point), the coefficient becomes negative.

The coefficient was measured between 65 and 156°F, and, for analyzing accidents, it is assumed the data can be extrapolated to boiling. The measured coefficient is given by equation:

dp/dT = -5.7 x 10-3 (T-124) ¢/°F (Equation 4-1) where T is the primary coolant temperature in (°F) and is the reactivity of the system (¢). This coefficient is not affected by fuel burnup and is not expected to vary significantly with core life.

An experiment was performed to check the sign of the void coefficient of reactivity. In this experiment, the reactivity effect of moving pieces of aluminum from the core was positive; therefore, the void coefficient was negative, as required. The magnitude of the void coefficient was not measured directly but was determined from the results of the temperature coefficient experiment. In this determination, the source of reactivity change in the temperature coefficient is presumed the result of density changes only and is interpreted as an effect from void buildup.

Extrapolation of the temperature coefficient data yields a void coefficient of -5.7 ¢/% void above the temperature coefficient turning point of 124°F.

Changes in reactivity caused by inserting materials during experiments are largest for experiments in the horizontal facility. Several measured reactivity effects in the horizontal facility and the vertical facility are given in the table at the end of this section. As indicated by the fact that the thermal column increases the flux at the south face of the reflector, experiments at the face of the 5-foot graphite cube, which contain large quantities of reflector materials, could have a small reactivity effect. However, during experiments performed to date, such an effect has never been observed.

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Table 4-1 NUCLEAR PARAMETERS PARAMETER VALUE Fuel Loading

- Critical mass (cold, 0.28 inch between disks)

- Actual initial loading

- Actual loading after 198.5 MWD of operation Reactivity Worth of Movable Nuclear Poisons

- All three control rods 0.015 k/k ($2.2)

- All four safety rods (conservatively assumed) 0.014 k/k ($2.0)

Net Reactivity (console excess with typicala operational core)

- All four safety rods and three control rods withdrawn +0.002 k/k (+$0.3)

- All four safety rods withdrawn, and all three control rods -0.014 k/k (-$2.0) inserted

- All four safety rods inserted, and all three control rods -0.012 k/k (-$1.7) withdrawn

- All four safety rods inserted, and all three control rods -0.028 k/k (-$4.0) inserted Reactivity (Console Excess)

- All four safety rods and all three control rods inserted and -0.023 k/k (-$3.3)

MPS withdrawn (typicala operational core)

Reactivity Addition from Primary Coolant Temperature change +0.00048 k/k (+$0.07)

(from 75 to 124°F) a 1/16 MPS in slot #2, control rods in the critical position, neutron radiography source log in horizontal cavity, graphite in vertical and other experiment facilities (or similar arrangement); excludes temperature and experiment transient worth.

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Table 4-2 NUCLEAR PARAMETERS (continued)

PARAMETER VALUE Miscellaneous Reactivity Effects

- Removing graphite from central sample tube (3-in. -0.009 k/k (-$1.3) cavity)b

- Filling central sample tube with water (3-in. cavity) -0.02 k/k (-$3)

- Removing all graphite from vertical facility -0.008 k/k (-$1.1)

- Removing the fuel loading chute plug -0.006 k/k (-$0.89)

- Equilibrium xenon at 100 kW -2.3 x 10-3 k/k (-$0.3)

- Yearly fuel burnup (typical use) -2.3 x 10-4 k/k (-$0.03)

Coefficients of Reactivity

- Temperature coefficient in o Water coolant (measured) -5.7 x 10-3 (T-124) ¢/°F o Inner graphite (calculated)c +0.018 ¢/°F o Outer graphite (calculated)c +0.55 ¢/°F

- Average void coefficient -5.7 ¢/% void

- Doppler coefficient Negligible Mean Lifetime of Prompt Neutrons 2 x 10-4 sec Neutron Fluxes at 100 kW

- Average thermal flux in fuel 7 x 1011 n/cm2-sec

- Peak thermal flux in central sample tube 2.5 x 1012 n/cm2-sec

- Peak thermal flux in CHRIS 8.0 x 1011 n/cm2-sec

- Thermal flux at face of thermal column 7 x 108 n/cm2-sec

- Thermal flux at face of 5-ft graphite cube 5 x 1010 n/cm2-sec Miscellaneous Parameters After 198.5 MWD of Operation

- Reactivity lost due to fuel burnup 0.015 k/k ($2.1)

- Reactivity lost due to aggregate fission product poisoning 0.012 k/k ($1.8)

(other than samarium-149)

- Reactivity lost due to samarium-149 (at equilibrium) 0.0044 k/k ($0.62) b This miscellaneous reactivity effect includes removal of the neutron radiography source log and removable inner graphite sleeve from the horizontal cavity.

c The inner graphite region considered includes the removable inner graphite sleeve and graphite components within the horizontal cavity. The outer graphite region includes all other graphite reactor elements.

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4.4.3 Operating Limits The NTR operates with a single fixed core configuration. Reactor fuel is not reconfigured in any way. Outside the core, experiments and manual poison sheets may be altered. However, the potential excess reactivity, at the NTR, is always to limited to $0.76. Since a $0.76 step reactivity insertion will not cause fuel damage; even with a failure to scram, operation of the reactor will not pose a threat to the health and safety of the public.

Even if an instrument malfunction drives the most reactive control rod out in a continuous ramp mode in its most reactive region, the reactor period and neutron flux monitors would scram the reactor. If the reactor did not scram, and reactivity is introduced in either step or relatively long ramp (with the potential excess reactivity being $0.76 or less), a total reactivity addition of the control rods, experiments, and temperature effect will not result in fuel damage.

The shutdown margin for NTR is $2.0. This is calculated with: the strongest safety rod stuck in the full out position, all control rods full out and a xenon free core. The total safety rod worth of

$3.86 minus the maximum potential excess reactivity of $0.76 is $3.1. This value minus the strongest rod of $1.1 gives the shutdown margin of $2.0.

The Technical Specifications for NTR state that the minimum shutdown margin with the maximum worth safety rod stuck out shall be $1.0. Operation in accordance with this specification ensures that the reactor can be brought and maintained subcritical without further operator action under any permissible operating condition even with the most reactive safety rod stuck in its most reactive position.

Safety Limit Safety Limits are limits on important process variables which are found to be necessary to reasonably protect the integrity of the NTR fuel. The only accidents which could possibly cause fuel damage and a release of fission products from the NTR fuel are those resulting from large reactivity insertions. With the $0.76 potential excess reactivity limit, a large reactivity insertion is not possible. Therefore, there is no mechanistic way of damaging the fuel and Safety Limits should not be required.

The Code of Federal Regulations, however, requires a reactor to have Safety Limits. Therefore, a Safety Limit was chosen to restrict the ratio of the actual heat flux to the Departure from Nucleate Boiling (DNB) surface heat flux in the hottest fuel element coolant passage below 1.5 to preclude 4-23

any subsequent fuel damage due to a rise in surface temperature. Thermal-hydraulic analyses show that the DNB heat flux for the NTR is not significantly affected by the core flow rate or the core inlet temperature. Reactor power is the only significant process variable that needs to be considered.

The safety limit for the reactor operating under steady-state or quasi steady-state conditions is 190 kW. A DNB ratio equal to 1.5 was selected as a conservatively safe operating condition for steady-state and quasi-steady-state. The reactor thermal power level when DNBR=1.5 is 190 kW.

Another Safety Limit under reactor transient conditions is not required. Conservative transient analyses show that the potential excess reactivity limit of $0.76, fuel damage does not occur even if all scrams fail to insert the safety rods. Although the power level may safely attain 4000 kW during this transient event, the Safety Limit of 190 kW was conservatively selected to apply to the transient condition.

Limiting Safety System Setting The linear neutron power monitor channel set point shall not exceed the measured value of 125 kW.

Transient analyses were performed assuming greater than $0.76 maximum potential reactivity and an overpower scram set point at 150 kW. None of the anticipated abnormal occurrences or postulated accidents resulted in fuel damage using these values. The Limiting Safety System Setting (LSSS) of 125 kW (125% of nominal operating power is the currently preferred value for research reactors) is extremely conservative for the NTR.

Each linear neutron power monitor channel set point is set to trip at 120%. Full power of 100 kW is verified to indicate 100% on the linear channel. Therefore 120% trip point is within the 125-kW requirement. The trip points are verified on the Daily Surveillance Check sheet prior to each days operation.

Limiting Condition for Operation The reactor configuration shall be controlled to ensure that the potential excess reactivity shall be

$0.76. If it is determined that the potential excess reactivity is >$0.76, the reactor shall be shut down immediately. Corrective action shall be taken as required to ensure the potential excess reactivity is $0.76. This ensures that there would not be any mechanism for addition of reactivity greater than $0.76. Detailed analyses have been made of reactivity insertions in 4-24

Chapter 13. The analyses show that a reactivity step addition of $0.76 will not cause significant fuel degradation.

The reactor shall be subcritical whenever the rods are in their subcritical rod position; that is, whenever the four safety rods are withdrawn from the core and the three control rods are fully inserted. This ensures that criticality will not be achieved during safety rod withdrawal.

Adherence to the $0.76 limit also ensures that the reactor will not go critical during safety rod withdrawal.

The potential excess reactivity is verified to be $0.76 power to every startup. The calculation is recorded on the Startup-Shutdown Report. The calculation considers changes in reactivity with regard to temperature, manual poison sheets, horizontal facility, and any other possible reactivity changes.

The minimum shutdown margin with maximum worth safety rod stuck out shall be $2.0. This ensures that the reactor can be brought and maintained subcritical without further operator action under any permissible operating condition even with the most reactive safety rod stuck in its most reactive position.

Each manual poison sheet used to satisfy the $0.76 limit shall be restrained in its respective graphite reflector slot in a manner which will prevent movement by more than 1/2 -inch relative to the reactor core.

This ensures that the manual poison sheets will not be removed from the reactor core during the maximum postulated seismic event.

Any time a manual poison sheet is changed, it is verified to be properly latched in the new position.

The temperature coefficient of reactivity of the reactor primary coolant shall be negative above a primary coolant temperature measured value of 124°F.

This ensures there is no significant positive reactivity feedback from coolant temperature change during reactor power transients.

The over-all temperature coefficient of the fuel annulus was measured in earlier studies and found to be positive up to 124°F. As temperature is increased above 124°F (the turnover point), the coefficient becomes negative. The coefficient was measured between 65 and 156°F, and, for 4-25

analyzing accidents, it is assumed the data can be extrapolated to boiling. This coefficient is not affected by fuel burnup and is not expected to vary significantly with core life.

4.4.4 NTR Core Computation Model To analyze the NTR core, a detailed computational model was developed and benchmarked against measured data. This section provides information on the computational core model developed for the neutronics parameter and reactivity analyses.

The core model reactivity and neutronics parameter calculations were performed using MCNP6, versions 1 and 2 (Reference 36). MCNP6 is a Monte Carlo program for solving the linear neutron transport equation for a fixed source or an eigenvalue problem. The code implements the Monte Carlo process for neutron, photon, or electron transport or coupled transport involving all these particles and can compute the eigenvalue for neutron-multiplying systems.

MCNP6 uses pointwise (i.e., continuous) cross section data, and all reactions in a given cross section evaluation (e.g., ENDF/B-VII) are considered. In this evaluation, thermal neutron scattering with hydrogen and with graphite were described using an S(,) thermal scattering kernel for light water and graphite, respectively. The GEH/GNF version of the ENDF/B-VII continuous energy cross section library and the S(,) thermal scattering kernel inputs for light water and graphite were used for the reactivity and neutronics parameter calculations.

The models represent the entire reactor core, control rods, safety rods, MPS control elements, the graphite reflector and moderator, the irradiation facilities, the primary coolant regions, and other component geometry described in Section 4.2. All structural and fuel materials are included in the as-modeled fuel geometry.

The fuel depletion calculations were also performed using MCNP6 package (Reference 36),

which tracks a large number of fission products and the buildup of plutonium. The depletion model utilizes a unique fuel material for each set of two fuel disks along the active fuel assembly length to ensure that an appropriate axial fuel burnup distribution is achieved. This guarantees the axial variation in fuel burnup of each fuel disk region due to the variation in flux distribution is appropriately representative over the NTR core life. Unlike the model used for reactivity and parameter calculations, the depletion model assumes azimuthal symmetry, and models the axial variation of fuel isotopics along the active fuel length for one fuel assembly in the core.

Reflective boundary conditions were imposed to model the fuel depletion for one half of the 4-26

active fuel length for this representative fuel assembly. The BURN function in MCNP6 was used to deplete the fuel from beginning-of-life (BOL) to the current core exposure. The fuel depletion burnup scheme was modeled based on the cumulative burnup values from NTR records which closely corresponded to past NTR operation statepoints for MPS configuration changes and critical control rod calibration measurements.

The depletion modeling analysis fuel isotopic results were then used as the fuel material inputs for the core models at the different NTR operating statepoints corresponding to MPS configuration changes and critical control rod calibration measurements. The core models were used to simulate NTR operation and predict the reactivity worths for the various NTR components and core configuration changes. The core model calculated results were then compared with NTR measurements for critical reactor configurations and measured reactivity worths for MPS changes to benchmark the model and cross section library.

The differences between the core model-predicted control rod worths and net reactivity gains from MPS changes and the NTR measured data are within the overall model uncertainty for the past operation statepoints and for the current exposure statepoint. This validates that the core model and depletion analysis fuel isotopic results can be used to simulate NTR operation and calculate neutronics parameters and reactivity worth predictions for reactor components and core configuration changes.

Based on the core model calculations described in this section, there is no change to the safety case for the NTR core configuration for operation as a function of exposure.

Calculated Core Parameters Neutronics parameters and reactivity values are calculated for BOL (fresh fuel) and current exposure (as of 9/30/2019 at approximately 198.5 MWD). These results support the information provided in Table 4-1 and Section 4.2.2.

Core model calculations for the BOL and current exposure cores confirm the neutron flux parameter results and conclusions in Section 4.2.2. The core models are based on critical core configurations with safety rods withdrawn, control rods in critical rod positions, and with the neutron radiography source log inserted. The BOL core model includes a full MPS in slot #1 and a half MPS in slot #5 based on probable MPS configurations for initial NTR operation. The 4-27

current exposure core model includes a 1/16 MPS in slot #2 consistent with NTR operating records for the current exposure core configuration.

Figure 4-4 Thermal Neutron Flux and Cadmium Ratio Traverse 4-28

Figure 4-5 Thermal Neutron Flux Traverses in Three Manual Poison Sheet Slots 4-29

Reactivity Coefficients and Point Kinetics Parameters The graphite temperature coefficients of reactivity were calculated for the inner graphite and outer graphite regions.

The core model is used to calculate the reactivity effect from changes in the volume average graphite temperature for the inner graphite volume and outer graphite volume for the BOL and current exposure fuel compositions, while keeping other core parameters unchanged. The ranges on graphite temperature are based on the available ENDF/B-VII cross section library temperatures and S(,) thermal neutron scattering kernel inputs for graphite to cover expected transients for NTR operation. The core eigenvalue results were used to calculate the inner and outer graphite temperature coefficients of reactivity for the core model (Equation 4-2):

Equation 4-2 where TGraphite is the volume average graphite temperature (°F), is the reactivity of the system

(¢), and k is the core eigenvalue calculated for the graphite temperature case. The average effective delayed neutron fraction () from Section 4.2.2 is used to convert the results to units of reactivity.

The core model calculated results for inner graphite and outer graphite temperature coefficients of reactivity for the current exposure fuel composition are reported in Table 4-1.

The point kinetics parameters evaluated are the effective delayed neutron fraction, eff, and the prompt neutron generation time, l. Both the delayed neutron fraction and neutron generation time were calculated using the KOPTS function in MCNP6. The MCNP6 results for neutron generation time and core eigenvalues were multiplied to calculate the mean lifetime of prompt neutrons for the BOL and current exposure cores.

Table 4-3 provides the point kinetics parameters calculated for the BOL and current exposure cores. The BOL core model includes a full MPS in slot #1 and a half MPS in slot #5, and the current exposure core model includes a 1/16 MPS in slot #2. The control rods are positioned to achieve a critical condition for these calculations. The core model calculated results for mean lifetime of prompt neutrons support the existing value in Table 4-1.

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Table 4-3 CORE MODEL CALCULATED POINT KINETICS PARAMETERS 4.5 THERMALHYDRAULIC DESIGN Maximum authorized power for the NTR is 100 kW. High-power trips are routinely set at powers no higher than 125 kW and a core outlet high-temperature scram is set to ensure that the core outlet temperature is less than 222°F. For powers above 0.1 kW, forced circulation of deionized water is used to transfer the heat from the core to a heat exchanger, as described in Chapter 5.

When forced circulation is required, the reactor shall scram if flow is less than 15 gpm. At powers less than 0.1 kW, operation is permitted without forced circulation (i.e., the primary recirculation pump need not be operating, and the low-flow scram is bypassed). The 0.1-kW limitation for natural circulation operation is extremely conservative (established in the past) but will continue to be used even though more recent analysis for the loss-of-flow accident described in Chapter 13 shows the core can be adequately cooled by natural circulation at much higher powers. Under both operating conditions, natural or forced circulation, the performance of the core is good with regard to the avoidance of natural thermal limits. These thermal limits include melting of the fuel and cladding, and burnout of the fuel cladding.

The maximum authorized operating power, 100 kW thermal with a rated recirculation flow of 20 gpm, has been used for the reactor to establish values for the thermal and hydraulic characteristics of the reactor core. A summary of these characteristics given in Table 4-2 shows that the thermal loading on the core is quite modest. The core inlet coolant temperature is typically 90°F; the core average exit temperature is 120°F, and, in the hottest channel, the exit temperature is only 150°F.

The saturation temperature of the coolant corresponding to the average reactor pressure, 20 psia, is 228°F. Thus, the state of the coolant is far removed from boiling at the design operating condition.

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The cladding surface temperatures were established based on known coolant temperatures and the heat flux distribution in the core. The flow through the core is laminar, and the surface film heat-transfer coefficients were calculated from a known laminar correlation. Fuel-plate temperatures increase with power up to a certain point; however, when the surface temperature is elevated to a value that will support local boiling of the coolant, the heat-transfer mechanism undergoes a marked change. There is substantial increase in the heat-transfer coefficient, and, consequently, the plate surface temperature is practically held at a maximum value, corresponding to the value needed to establish local boiling. The Jens-Lottes correlation (Reference 5) was used to predict the local value of wall superheat necessary to establish local boiling. This phenomenon is important because metal temperatures are limited to values well below melting, which is particularly evident during certain accidental transients discussed in Chapter 13.

The core flow distribution out of the inlet header (described in Section 4.2) is such that adequate cooling of all portions of the core is achieved. The pipe is orificed to give higher-than-average flow rates in the horizontally central region of increased power generation, and lower-than-average flow to the end regions.

The peaking factors used in this evaluation were maximum expected values that result from operation of the reactor with neutron flux peaked on one side of the core. The circumferential power distribution used resulted in a circumferential power peaking factor of 1.25. The longitudinal shape is symmetrical, with a total axial peaking of 1.15. The total overall power peaking in the core is 1.58, which includes a local peaking factor of 1.1.

Of considerable importance is the ability of the recirculation system to maintain a mode of natural circulation flow when the primary pump is not operating, and core power is up. In the absence of pump head, the driving pressure difference around the recirculation loop is the net elevation head of the coolant. This is directly proportional to the density differences between the water in the core and riser section and the water leaving the heat exchanger. Again, this density difference is a function of core power. The length of piping over which this density difference exists is slightly more than 5 feet. System response to loss of recirculation pumping is discussed in Chapter 13.

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Table 4-4 TYPICAL NTR CORE THERMAL AND HYDRAULIC CHARACTERISTICS Maximum thermal power level (scram) 125 kW Maximum thermal power level 100 kW Average fuel disk surface heat flux 6600 Btu/h-ft2 Maximum fuel disk surface heat flux 10600 Btu/h-ft2 Total fuel to coolant heat-transfer area 52.7 ft2 Total core power peaking factor 1.58 Core average pressure level 20 psia Coolant flow characteristic

- Total core flow area 0.39 ft2

- Channel flow area 0.70 in2

- Channel hydraulic diameter 0.51 in

- Total recirculation flow rate 20 gpm (9800 lb/h)

- Inlet velocity, average channel 0.14 ft/sec

- Inlet velocity, hottest channel 0.07 ft/sec

- Mass flow rate, average channel 122 lb/hr

- Mass flow rate, hottest channel 64 lb/hr

- Coolant inlet temperature 90°F

- Coolant exit temperature, average channel 120°F

- Coolant exit temperature, hottest channel 150°F

- Coolant saturation temperature 228°F Fuel disk cladding temperature

- Average channel 170°F

- Hottest channel 195°F Maximum temperature difference, fuel-to-cladding surface 1°F Fuel plate steam-blanketing is a condition that may occur even in a pressurized water system and can be of considerable concern. This condition is caused by going from local surface boiling into film boiling upon reaching very high surface heat fluxes. This could be of concern because the steam film degrades the heat-transfer, and the fuel plate temperature increases greatly as a result.

However, during steady-state operation, this is of no real concern in the NTR for these reasons:

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Heat fluxes at maximum power in the reactor are quite small because of the low power rating, and the burnout heat fluxes, or the heat flux necessary to cause steam-blanketing, are very high for the coolant conditions existing in the reactor, as evidenced by experimental data. For instance, in the hottest channel in the core, the data indicates a burnout heat flux of 227,000 Btu/h-ft2 for the hydraulic conditions at which the channel is operating. The actual maximum heat flux in this channel, for 100-kW operation, is 10,300 Btu/h-ft2. Thus, the burnout ratio, or the ratio of burnout heat flux to maximum operating heat flux, is 22. This is a considerable margin and represents a highly safe condition.

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5 REACTOR COOLANT SYSTEMS 5.1.

SUMMARY

DESCRIPTION The reactor primary coolant system is an unpressurized light-water system which provides forced circulation coolant to the reactor core at powers above 0.1 kW; below 0.1 kW, no coolant circulation is required for typical operating periods, although it may be utilized if desired. The primary water system removes the heat from the core and transfers it through the two-pass, U-tube heat exchanger. Primary water flows through the shell side of the heat exchanger and is about 30 psig lower than the tube side. The secondary cooling water, to the heat exchanger, comes from the building 105 potable water supply, which is fed from the site raw water main supplied from the sites 500,000-gal storage tank. Upon leaving the heat exchanger, the water goes to the facility drain, which discharges to the site retention basins. The heat exchanger is designed to eliminate thermal stresses induced by temperature differentials.

Figure 5-1 shows the NTR coolant systems. Typical conditions for reactor power of 100 kW are:

35 gpm secondary water, 20 gpm primary water, 90°F core inlet temperature, and 124°F core outlet.

5.2. PRIMARY COOLANT SYSTEM 5.2.1. System Description The primary system is an unpressurized light-water system which provides the coolant to the reactor core. The cooling system contains a volume of about 28.5 gallons, of which 19.5 gallons are contained in the main flow path piping and 9 gallons are contained in the core tank.

The major portion of the primary system is constructed of 1-1/2-inch Schedule 40 aluminum pipe.

The internal parts of in-line equipment such as the pump and heat exchanger are stainless steel:

The primary coolant flow path is from the primary pump (Rated at 1 hp, 25 gpm at 55 ft. head) through a check valve and flow control valve, to the bottom of the reactor core tank. Water is distributed by a baffle tube and flows up around the fuel assemblies to the top of the core tank.

The water then flows out of the graphite pack, through a flow orifice, heat exchanger (U-type, 36 inches long, 3.4 x 105 Btu/hr.) an air trap and back to the primary pump. Refer to the primary piping and instrument diagram (P&ID) Figure 5-1 and the primary isometric Figure 5-2.

5-1

Primary water may be drained to the 500-gal holdup tank in the northeast corner of the cell where it can be retained or transferred to the other tanks for transfer from the facility. The holdup tank also receives the discharge from the primary system atmospheric vent line, which is connected to the inlet of the heat exchanger. This line is a design feature and is the highest point in the system.

It provides a continuous vent to atmosphere for air and other gasses and prevents over-pressurizing the primary system. An overflow line from the fuel storage tank to the holdup tank connects into the primary system atmospheric vent line. A sump pump is in a sump in the northwest corner of the cell. Any water collected in this sump is automatically pumped into the 500-gal holdup tank.

The air trap is a 5-foot length of 4-inch aluminum pipe that originally contained Cal-Rod-type heaters rated at 5 kW. The heaters have been removed from the system, but the tank remains and serves as a system air trap.

5.2.2. System Operation Maintaining water in the core tank ensures that there will be no reactivity insertions due to the removal of voids or the sudden addition of water into the core tank during reactor operation.

Therefore, the reactor is not operated above 0.1 kW unless the core tank is filled with water. If, during operation of the reactor, it is determined or suspected that the core tank is not filled with water, the reactor will be shut down immediately and corrective action will be taken as required.

Forced coolant flow is not required for reactor operation at or below 0.1 kW. Above 0.1 kW the reactor, required light-water forced coolant flow is ensured by maintaining the Primary Coolant Low Flow scram setpoint at no less than 15 gpm.

Primary system leakage is maintained below 10 gallons/day as an operational practice that ensures there are 10 days between the holdup tank high- and low-level alarms. This practice minimizes low coolant inventory-related shutdowns by providing ample time to manually add water to the system.

The maximum Primary Coolant High core outlet temperature scram set point is 222°F. This provides assurance that the reactor fuel temperature will not attain a temperature which will cause damage to the fuel.

A high core temperature alarm at 200°F gives warning to the operator of higher-than-normal primary coolant outlet temperature.

5-2

The temperature coefficient at temperature T (°Fahrenheit) is given by Equation 4-2:

-5.7 x 10-3 (T-124) ¢/°F The fixed NTR core configuration ensures a temperature coefficient turnover from positive to negative above the operational coolant temperature of 124°F and yields a negative void coefficient above that temperature. This ensures there is no significant positive reactivity feedback from coolant temperature change during reactor power transients.

The coefficient is not affected by reactor configuration and fuel burnup and is therefore not expected to vary significantly with core life; however, it could be affected by fuel, core, or moderator design changes. Such changes should include an evaluation to determine if the temperature coefficient needs to be reverified.

Primary water samples can be taken at the sample station located in the northwest corner of the south cell.

Typical cooling system conditions at reactor full power are as follows:

Flow Rate, gpm 20 Core Inlet Temperature (°F) 90 Core Outlet Temperature (°F) 124 Conductivity (µS/cm) <1.0 pH 5.5 to 7.5 5-3

Figure 5-1 Primary Piping and Instrument Diagram 5-4

Figure 5-2 Primary Isometric Diagram 5-5

5.2.3. System Disruptions In the event of a primary pump failure or seizure, or a significant leak or break in the primary cooling system, the reactor shall be shut down immediately. If the event causes a reactor scram, personnel should verify that all safety systems functioned as intended. If the reactor fails to scram, the console operator shall manually scram the reactor.

During a complete instantaneous loss off primary coolant flow without a reactor scram, fuel damage does not occur. Natural convection cooling is sufficient and therefore, forced coolant flow is only conservatively required above 0.1 kW.

A complete loss of coolant in the core tank with a simultaneous failure to scram the reactor at full power would result in a reactor shutdown because of moderator voiding. Peak fuel temperature would reach a maximum of 626°F about 30 minutes after coolant loss (Figure 13-14). No damage to the fuel will result, so the consequences of this accident are minimal.

5.3. SECONDARY COOLANT SYSTEM Secondary cooling system (Figure 5-3), for the NTR, uses water from the building 105 potable water supply, which is fed from the site raw water main supplied from the sites 500,000-gal storage tank. The 1.5-inch supply line to the NTR facility supplies the one heat exchanger. It passes through a filter, in the building 105 equipment space, across the roof of the building, and then enters through the ceiling of the control room. In the control room it passes through a shut-off valve, a check valve, and a flow indicator, and then enters the reactor cell. Inside the cell, the line goes directly to the tube side of the heat exchanger and then through a manual valve to the Site retention basins.

The heat exchanger can transfer 3.4 x 105 Btu/h. The tube bundle consists of 0.25-inch-ID, 36-inch-long stainless-steel tubes, and provides a heat-transfer surface of approximately 36 ft2. The heat exchanger is located on the east wall of the reactor cell about 6 feet above the reactor core.

Pressure at the inlet of the heat exchanger is normally about 70 psig. The design specifications pertinent to maintaining system integrity are given in Table 5-1.

5-6

Table 5-1 HEAT EXCHANGER SPECIFICATIONS Shell Side Tube Side Fluid (coolant) Primary Secondary Fluid Flow Rate (gpm) 20 35 Fluid Velocity (ft/sec) 1.48 3.43 Temperature in (F°) 124 60 Temperature out (F°) 90 80 Pressure Drop (psig) 1.6 4.0 Design Pressure (psig) 300 150 Test Pressure (psig) 500 300 Design Temperature (F°) 400 400 Material (stainless steel) 316 316 Primary water flows through the shell side of the heat exchanger. The probability and consequences of a leak between the two systems in the common heat exchanger have been evaluated. The primary side of the heat exchanger is <40 psig, and the secondary is, as mentioned, (~70 psig), therefore a heat exchanger leak would result in a secondary-to-primary leak. The evaluation showed that the probability of leaking contaminated water from the primary to secondary system is extremely low.

Secondary coolant flows by gravity through the tube side of the primary heat exchanger. Loss of secondary coolant flow would result in a primary coolant temperature increase to the alarm set point. A reactor scram could occur if remedial action is not taken. Loss of secondary coolant flow will cause the reactor to scram from high primary coolant temperature if operated at a reactor power level which would generate an appreciable amount of heat.

If there is a break in any of the coolant piping, in the reactor cell, the coolant will flow to the reactor cell sump pump where it is automatically pumped to the wastewater hold up tank. High water level in the reactor sump will activate an alarm in the control room and in the security building. This water can be held for evaporation inside the cell or transferred to the waste evaporator for processing. It is not added to other systems.

5-7

Figure 5-3 Secondary Cooling System 5-8

The NTR control room temperature recorder records the heat exchanger inlet and outlet temperatures of the secondary coolant water, along with other reactor temperatures. Any deviations, from normal, of secondary temperatures, as well as the other temperatures, would be noted by the operator.

A flow meter indicating secondary coolant flow is located on the wall in the reactor control room.

Any deviations, from normal flow, would be noted by the operator. The heat exchanger is regularly inspected per the NTR preventive maintenance procedures.

Although both the secondary coolant and the primary system makeup water come from the same potable water supply, they are separate systems. Each has their own check valve, manual valve, and solenoid valve (operated by keylock at the reactor console) so that one coolant system cannot flow to the other system. Addition of water to the primary system can only be accomplished from inside of the reactor cell; otherwise, the manual makeup valve remains closed at all other times.

Secondary cooling system water is not a potential radioactive liquid effluent because secondary heat exchanger pressure is higher than primary coolant pressure. The secondary coolant system is a single pass system as opposed to a closed loop system. Because the coolant is used only once and is initially drinking water quality, there are no radiation monitors or detectors incorporated into the secondary system. In the event the secondary should somehow become contaminated, the contaminated water would drain to the Site retention basins. Basin water is analyzed for radioactive material content and would not be released to the environment if found to be contaminated.

The secondary system is used only to remove heat from the primary system. There are no emergency core, experiment, or biological or thermal shield cooling systems at the NTR.

There are no limitations applied to the secondary coolant system.

5.4. PRIMARY COOLANT CLEANUP SYSTEM The entire primary coolant cleanup system is located inside the reactor cell, on the east wall.

The purity of the primary coolant system is maintained by two Barnstead Model BD-2 Pressure Bantam Demineralizers installed in parallel (Figure 5-1 and Figure 5-2). The cleanup system normally operates with both demineralizers online. If both units are used, the system will service 32 gph. The demineralizers contain replaceable cartridges. A conductivity monitor is located upstream of the demineralizers. To further ensure the purity of the system, a 5-micron cartridge-type filter is installed in the discharge line of the demineralizer system. The entire primary 5-9

coolant cleanup system is located inside the reactor cell, on the east wall. Water enters the cleanup system from the discharge side of the primary pump just before the reactor inlet and reenters the primary system on the suction side of the primary pump.

The primary system pH is ensured by monitoring the water conductivity. The conductivity of the primary coolant is checked prior to the first startup of the day in accordance with NTR Standard Operating Procedures. Both conductivity and pH are checked annually by Analytical Chemistry in accordance with NTR Preventive Maintenance Procedures. Contact radiation readings are taken on the demineralizers periodically to ensure conditions remain ALARA.

Normal radiation readings on the demineralizers are up to 2 R/hr. Resins should be scheduled for replacement when their radiation level reaches a consistent 3 R/hr. level. High radiation and contamination levels may be expected during the performance of resin replacement work. The resin cartridges are changed and staged in the reactor sump until they are transferred for disposal.

The filter is also changed as required.

An inadvertent release of excess radioactivity in the primary coolant system may cause a reactor cell area radiation monitor alarm. The area monitor detector is located on the reactor cell near the primary flow transmitter and is set to alarm at 1x 106 mrem/hr. The Senior Reactor Operator will determine the cause of an alarm and initiate corrective actions.

The piping for the primary cleanup system is 1/4-inch stainless steel tubing. The flow is preset by design to ~16 gph. If there is a break in any of the piping the loss of water would be noted by a low-level alarm on the fuel tank. The coolant will flow to the reactor cell sump pump where it is automatically pumped to the wastewater hold up tank. This water can be held for evaporation inside the cell or transferred to the waste evaporator for processing. It is never used to add to other systems.

The specific conductivity of the primary coolant water shall be maintained less than 5 µS/cm except for time periods not exceeding one month when the specific conductivity may exceed 5

µS/cm but shall remain less than 20 µS/cm. If the specific conductivity exceeds 5 µS/cm, steps shall be taken to assure the specific conductivity is reduced to less than 5 µS/cm.

The minimum corrosion rate for aluminum in water (<50°C) occurs at a pH of 6.5. Maintaining water purity below 5 µS/cm will maintain the pH between 5.5 and 7.5. These values are acceptable for NTR operation. High specific conductivity can be tolerated for shorter durations 5-10

during unusual circumstances. Operation in accordance with these limitations ensures aluminum corrosion is within acceptable levels and that activation of impurities in the primary water remain below hazardous levels.

Technical Specification surveillance requirements for the reactor coolant system are included in the NTR Technical Specifications, Table 4-1 and Table 4-2.

5.5. PRIMARY COOLANT MAKEUP WATER SYSTEM Water, for the site, is normally supplied to the Site from the Hetch-Hetchy Aqueduct. A 14-inch (36 cm), 3-mile (4600 m) long pipe has been installed from the aqueduct to the Site. The installed on-site pumps have a capacity of 1,000,000 gpd and the pipeline capacity is over 3,000,000 gpd.

The Calaveras Reservoir, located about 8 miles (13 km) south of the VNC, provides backup for Hetch-Hetchy.

Primary coolant makeup water (Figure 5-3), for the NTR, comes from the building 105 potable supply, which is fed from the site raw water main supplied from the sites 500,000 gal. storage tank. A potable water line, connected to the supply, feeds the building 105 deionizer unit located in the building 105 equipment space. The deionizer provides all the building 105 deionized water needs as well as NTRs primary coolant makeup water. The makeup line enters the reactor cell from a line located above the control room ceiling (Figure 5-1), through a penetration in the south wall of the reactor cell, and connects into the primary system, between the primary pump and the inlet side of the reactor, through a solenoid valve energized by the reactor console key lock switch and a manual valve in the reactor cell. Makeup to the primary system can only be done from inside of the reactor cell.

Through the reactor fuel loading chute, the makeup system also supplies the 1800-gal. fuel loading tank, which serves as a reservoir for the primary system. The fuel loading tank is discussed in Section 5.7. High and low water level, in the fuel loading tank, is indicated by level switches. The switches actuate annunciators and alarms in the control room. They also actuate alarms in the security building that is occupied around the clock. High and low tank level alarms are always investigated when they are received.

The primary system and the fuel loading tank are within the reactor cell. Regardless of whether the fuel loading tank overflows or there is a break in the primary line, the coolant will flow to the reactor cell sump pump where it is automatically pumped to the wastewater holdup tank. This 5-11

protects against leakage of contaminated coolant to the potable water supply. This water can be held for evaporation inside of the cell or transferred to the waste evaporator for processing.

5.6. NITROGEN16 CONTROL SYSTEM There is no nitrogen-16 control system at the NTR.

5.7. AUXILIARY SYSTEMS USING PRIMARY COOLANT The fuel storage tank is connected to the reactor core tank by a 3-inch by 20-inch by 30-inch-long (7.6 cm x 50.8 cm x 76 cm) chute inclined on a 30° angle. When not being used, the loading chute is filled with an aluminum clad graphite plug and the aluminum access gate in the tank is closed. The fuel storage tank is located on the west side of the reactor graphite pack and provides biological shielding for fuel which is removed from the core. The tank is 4 feet x 5 feet x 12 feet high (1.3 x 1.5 x 3.66 meters) and is constructed from 1/4-inch (0.635 cm) aluminum. There are two 4-inch (10.16 cm) diameter tubes and one 2-inch (5.1 cm) diameter aluminum tube mounted on the east side of the tank. These tubes contain neutron detection chambers for the reactor nuclear instruments. Access to the tank is from the mezzanine.

The tank water level is monitored by high- and low-level float-actuated switches. An overflow drains water to the holdup tank.

The fuel loading tank water low level set point is <3-ft below the overflow. This alarm gives assurance that there is adequate water in the primary system for operation of the reactor.

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6 DESIGN BASES AND ENGINEERED SAFETY FEATURES Accident Analysis in Chapter 13 does not credit Engineered Safety Features (ESFs) in the most limiting postulated accident scenario and supports the conclusion that ESFs are not needed for the NTR design.

Prior revisions of the NTR SAR, including the most recent June 2000 update, based bounding analysis on assumptions developed from an extremely conservative plutonium capsule fueled experiment accident scenario that is not performed at the NTR. Updated analysis uses assumptions based on a still conservative experiment involving the rupture of an irradiated U-235 capsule that might periodically be performed at the NTR.

As a result, ESFs have been removed from this chapter according to NUREG-1537. The NTRs robust confinement, and ventilation systems are described in detail in Section 3.5 and ventilation stack action levels are included in Chapter 11.

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7 INSTRUMENTATION AND CONTROL 7.1.

SUMMARY

DESCRIPTION The reactor is equipped with sufficient instrumentation to control operation of the facility, measure operating parameters, warn of abnormal conditions, and scram the reactor automatically if an abnormal condition occurs (Figure 7-1 and Table 7-1). All reactor scram functions cause a loss of energizing currents to electromagnets, which, when deenergized, permit rapid insertion of the spring-loaded safety rods. The energizing currents can be disrupted by contacts in the power switches, scram relays, or by a manual scram switch. The power switches are controlled by logic units which monitor the trip circuits, on a two-out-of-three coincidence basis, for high reactor power from 3 pico-ammeters and for loss of high voltage for the three Compensated Ion Chambers (CIC) in the pico-ammeter channels. Another logic unit monitors singly (noncoincidence) the fast reactor period trip, and high log N power trip. All other scrams, except the console manual scram, operate through the scram relays and are initiated by the following signals:

Log N amplifier mode switch position Log N CIC loss of positive high voltage Primary coolant high core outlet temperature Primary coolant low flow Loss of AC power Reactor cell manual scram.

Safety is also provided by having each scram (except loss of AC power) cause the control rod drives to run to their fully inserted positions. Also, a provided rod withdrawal permissive interlock blocks control rod and safety rod withdrawal if a pico-ammeter (in a two out of three-coincidence logic) is not indicating above a preset minimum level. The rod withdrawal permissive circuit ensures that instrumentation is seeing the neutron source for reactor startups.

Additional interlocks associated with the rod drive system include the following:

1. For initial startup, or following a scram, magnets cannot be energized unless all safety and control rods and the neutron source are at their inner limits.

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2. Safety rods must be drawn one at a time to their outer limits before more than one control rod can be withdrawn.

3. The rod test panel consists of a key-lock arm switch and a seven-position selector switch.

The seven positions on the selector switch are OFF, Safety Rod #2, Safety Rod #3, Safety Rod #4, Coarse Rod #1, Coarse Rod #2, and Fine Rod. This panel bypasses the sequential withdrawal interlocks and permits the selected rod to be withdrawn out of sequence. All other rods, however, must be fully inserted.

4. The safety rod timer panel consists of an electronic timer, a key-lock arm switch and a five-position selector switch. Selector positions on the latter switch are OFF, Safety Rod #l, Safety Rod #2, Safety Rod #3, and Safety Rod #4. The timer measures the time lapse between a trip signal from the nuclear instruments and the safety rod-in limit switch closure.

To keep the system as simple as possible, bypasses are not provided in most of the scram circuits. It is felt that simplicity and ease of operation are more important than continuity of operation. If important components become defective, the condition will be evaluated, and the reactor shut down until repair or replacement is completed. However, some bypasses are necessary; for example, an automatic bypass has been provided for low primary coolant flow while at powers less than 0.1 kW.

The fail-safe philosophy has been incorporated into the design as much as is practical. In most instances, circuits are completed by energized relays or actuated microswitches to give protection against loss of voltage, poor contacts, or broken wires. Manually operated switches are installed for the control rod drive circuits and wherever practical that spring-return to open circuit (more safe position).

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Figure 7-1 NTR Scram System Schematic Diagram 7-3

Table 7-1 SCRAM SYSTEMS Item System Condition Trip Point Function No.

No higher than Scram (2-out-of-3 or High reactor power 125 kW 1-out-of-2)

1. Linear Power No less than 90%

Loss of positive high voltage to Scram (2-out-of-3 or of operating ion chambers (if used) 1-out-of-2) voltage No less than +5 Fast reactor period Scram sec Amplifier Mode switch not in

2. Log N operate N/A Scram No less than 90%

Loss of positive high voltage to of operating Scram ion chambers (if used) voltage Primary Coolant

3. Temperature High core outlet temperature 222 °F Scram (Fenwall)

No less than 15 Primary Coolant

4. Flow Low Flow gpm when reactor Scram power > 0.1 kW

5. Manual Console button depressed N/A Scram Reactor console key in off

6. Electrical Power position (loss of AC power to N/A Scram console) 7-4

7.2. REACTOR CONTROL ROOM The location of the reactor control room is shown in Chapter 1, Figure 1-1.

Building 105 hallway on the , and other laboratories to the . The doors between the control room and the two cell areas are the only personnel entries to these areas. A flashing warning light at the cell doorway is actuated if the door is opened; however, the radiation level in the cell must be above a preset minimum for this warning device to actuate. Since the cell door will be open while performing some experiments, it also has an audible alarm which may be actuated by breaking the beam of light from an electric eye across the doorway. This alarm system alerts the reactor operator to traffic to or from the cell as required.

A communication system provides local communications between the control room, room, and the setup room. A loudspeaker page system is also available, which can be heard in all NTR areas. Communication between the control room and cell is by a microphone and speaker system or by face-to-face communications.

7.3. SCRAM SYSTEM The scram system (Figure 7-1 and Table 7-1) consists of manual, process, and nuclear scrams.

7.3.1. Manual Scram The manual scram system consists of a manual button/switch located on the reactor console.

When depressed, it directly opens the circuit supplying power to the safety rod magnets, providing a method for the reactor operator to manually shut down the reactor if an unsafe or abnormal condition should occur, and the automatic reactor protection action, as appropriate, does not function. A manual button/switch is also located in the reactor cell.

7.3.2. Process Scrams The process scram chain consists of relay contacts and switches connected in series between the

+24 VDC bus and the coils of scram relays (R-19 and R-20 in Figure 7-1). Normally open contacts of these relays are in the circuit supplying power to the safety rod electromagnets and in the circuits for the rod drive motors. Two additional normally open contacts of these relays are used in the process scram chain parallel to the rod-in limit switches; this parallel circuit requires all motor-driven rods to be fully inserted before the scram relays can be energized. Any off-7-5

standard condition of any component supplying action to either the switches or to the relay contacts in the scram chain will disrupt power to scram relays and cause them to deenergize.

Process scrams are as follows:

1. The log N channel has an amplifier position mode switch which is used for checkout and testing of the instrument. Not having the amplifier mode switch in the operate position will prevent the safety chain from being made up, or, if moved from the operate position during operation, will scram the reactor.

2. Loss of positive high voltage below a predetermined value to the log N Compensated Ion Chamber (CIC) will scram the reactor. Loss of positive high voltage provides assurance that the ion chamber is capable of detecting neutrons.

3. A thermally actuated switch in the core outlet line senses the primary water core outlet temperature. A high outlet temperature will cause the switch to deenergize the scram relays and scram the reactor.

4. The primary flow is measured with a differential pressure transducer sensing the pressure drop across an orifice in the primary water coolant loop. An electric signal from the transducer is indicated at the control console. The reactor will scram when reactor power is greater than 0.1 kW and the primary coolant flow drops below a predetermined value.

5. Loss of AC power to the console will cause the scram relays and the magnet power supply to deenergize and scram the reactor.

6. A manual scram button is positioned in the reactor cell to scram the reactor from this area, if required. Actuation of this button also deenergizes the scram relays and will scram the reactor.

When the scram relays are deenergized, the following actions take place:

The power being supplied to the safety rod electromagnets is interrupted to allow the spring-loaded safety rods to be inserted.

All rod and motor circuits are closed to cause the rod drives to drive in if normal AC power is still available.

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The process scram chain is blocked open until the scram condition is corrected and all rods are fully inserted.

In addition to the above actions, the scramming condition will cause an annunciator to actuate, give an audible alarm, and illuminate a pushbutton lamp on the console which indicates the source of the scram. The audible alarm will continue until an ACKNOWLEDGE push-button switch is actuated; the pushbutton lamp remains illuminated until the scram condition is corrected and the pushbutton lamp is depressed. Some conditions cause indicator lamps to illuminate, but do not cause audible alarm; these conditions are not scrams.

7.3.3. Nuclear Scrams The nuclear scrams consist of four power range channels. A block diagram of the system is shown in Figure 7-1. The power range instrumentation is used to monitor neutron flux (reactor power) and to protect the reactor against excessive power levels or rates of power rise. This instrumentation is required to be operable and connected to the safety system during each startup and the subsequent operating period. The system consists of four independent neutron detection channels; three are monitored by pico-ammeters and the fourth by a log N and period amplifier.

The three pico-ammeters have trip circuits which operate into a two-out-of-three (or one-out-of-two if one channel is inoperative) coincidence logic circuit capable of causing reactor scram.

The log N and period amplifier can cause a reactor scram on fast period. The pico-ammeters have 20 ranges covering 10 decades of power from 10-9 and 100. Two ranges are available for each decade of 0 to 40 and 0 to 125 percent of that decade. The Log N channel normally covers the power range from 15 milliwatts to 150 kilowatts.

A gamma-compensated ion chamber is used as a detector in each power range channel. The detectors are positioned in thimbles in the fuel storage tank or at one of the faces of the reflector.

The exact location selected for a chamber is determined by the intended use of the reactor, sensitivity of the system, and the desired meter reading. The desirability of seeing the neutron source for startup, the minimization of shadowing effects, and the provision of physical protection for the chamber are the primary factors considered when positioning the chambers.

The CIC output currents can be interpreted in terms of reactor thermal power through calibrations based on measurement of thermal power as determined by a heat-balance 7-7

measurement which utilizes the coolant flow through the core and the differential temperature across the core.

High voltage for the three pico-ammeters CICs is internal. When in a two-out-of-three situation, loss of positive high voltage on any two pico-ammeters will cause a reactor scram. When in a one-out-of-two situation, loss of positive high voltage on any one pico-ammeter will cause a reactor scram.

High voltage for the log N period amplifier CIC is supplied by a power supply in the log N period amplifier. Loss of positive high voltage from this power supply will initiate a scram through the process scram relays.

Three multirange pico-ammeters are normally used (although operation is permitted with two) to monitor the signals from three (or two) of the CICs. Sensitivity and range of the systems are such that the flux (with the reactor shut down) from the reactor neutron source will bring all channels well on scale, and maximum reactor power does not exceed the range of the instrument.

Each pico-ammeter amplifier output signal, in addition to driving the pico-ammeter meter, and remote meter, is connected to an internally mounted trip circuit and externally through a selector switch to a linear power recorder. Each trip circuit is set to trip when the meter reads 125 kW or less. When the instrument reading is less than the trip point, the trip circuit supplies 12 VDC to a coincident logic circuit, wired to cause reactor scram if 2 of 3 (or 1 of 2) inputs are tripped. Each pico-ammeter has a downscale alarm. When two indicate an alarm, the control and safety rod motors cannot be energized to withdraw. Reactor operation may continue with one pico-ammeter out of service, provided the trip circuit is set up so that a trip signal from either of the remaining pico-ammeters will cause reactor scram. If one pico-ammeter is out of service, the interlock at the low end of the scale for that pico-ammeter which prevents rod withdrawal may be bypassed. These automatic actions ensure that the pico-ammeters have the proper start-up sensitivity, and that the high-power scram trip point is always within a decade of operating power during operations which increase power.

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The log N and period amplifier receives its signal from the fourth CIC and displays the reactor period and reactor power on front panel meters. This system may be set up to cover the power range from source or reactor critical level, depending on CIC position, to 150% of power. Relay outputs from the period amplifier trip circuit and the log N amplifier trip circuits are connected through the noncoincident logic circuit to initiate a scram for reactor periods of less than 5 seconds. The log N power signal is recorded on a strip chart recorder. The mode (multiposition calibration) switch for the log N amplifier is interlocked to scram the reactor when the switch is not in the OPERATE position.

The diode logic element system consists of two units. One unit performs coincidence logic functions and the other performs noncoincidence logic functions on signals from the nuclear instrumentation system.

A coincidence logic unit contains five independently functioning component boards which can accommodate a total of 16 signals. Four of the component boards are identical and provide circuits for performing two-out-of-three coincidence logic. The fifth circuit component board (not used) is designed to perform selective two-out-of-four coincidence.

The 12 VDC trip output from each of the three pico-ammeters passes through contacts in the meter relays monitoring positive high voltage to the CIC in that channel. The trip outputs from the three channels are converted parallel to the 2-out-of-3 coincidence logic unit. A trip on any two pico-ammeters or loss of high voltage to any two CICs or a loss of voltage on one channel plus a high-power trip in another will cause trip outputs from the coincidence logic unit to be sent to the noncoincidence logic unit which deenergizes the power switches.

The noncoincidence logic unit contains two independent noncoincidence logic component boards, each of which accommodates nine input signals and provides one output signal.

Depending on the input signal levels, each noncoincidence logic component board provides one of two possible outputs to a power switch: either 16 VDC, or less than 1 VDC. For the output level to be 16 VDC, all inputs must be normal. If any one or more inputs drop to zero, the output signal drops to less than 1 VDC.

Input signals to each nine-channel noncoincidence logic board consist of the following:

Two 24-V signals that pass-through scram relay R19; Two 24-V signals that pass-through scram relay R20; 7-9

One 12-V signal from log N high power trip; One 12-V signal from log N fast period trip; and Three 12-V trip signals from one coincidence logic unit.

Output signals to each noncoincidence logic board go to a power switch which controls the electromagnet excitation current. If either power switch trips, current to all magnets will be interrupted.

Power for the safety rod electromagnets is supplied from a direct current power supply with the capacity of supplying all four electromagnets. Power to each electromagnet is routed through individual power-adjust modules so that minor variations in the electromagnets can be compensated.

7.4. SAFETYRELATED ITEMS Safety-related items consist of instrumentation and systems to assist in the operation of the facility, measure operating parameters, or warn of abnormal conditions. Setpoints for safety-related items are provided in Chapter 14, Technical Specifications, Table 3-2.

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Table 7-2 SAFETY-RELATED ITEMS Item Component Condition Function No.

1 Reactor Cell Low Visible and audible alarm; audible alarm may be Pressure differential bypassed after recognition pressure 2 Fuel Loading Low level Visible and audible alarm; audible alarm may be Tank Water Level bypassed after recognition 3 Primary Coolant High core Visible and audible alarm; audible alarm may be Temperature outlet bypassed after recognition temperature 4 Primary Coolant Core delta Provide information for the heat balance Temperatures temperature determination 5 Stack High Level Visible and audible alarm; audible alarm may be Radioactivity (Beta-gamma bypassed after recognition particulate and noble gas) 6 Linear Power Low power Safety or control rods cannot be withdrawn (1-indication out-of-3 or 1-out-of-2) 7 Control Rod or Rods not in Safety rod magnets cannot be reenergized; may Safety Rod be bypassed to allow withdrawal of one control rod or one safety rod drive for purposes of inspection, maintenance, and testing 8 Safety Rod Rods not out Control rods cannot be withdrawn; safety rods must be withdrawn in sequence; may be bypassed to allow withdrawal of one control rod or one safety rod or one safety rod drive for purposes of inspection, maintenance, and testing

1. A differential pressure switch measures the pressure difference between the reactor cell and control room. This switch actuates a visual and audible alarm if cell negative pressure drops below a preset level (not less than 0.5 inches of water). The reactor power must not be increased above 0.1 kW unless the cell negative pressure is as noted above.

If the cell negative pressure drops below the preset level and the reactor power is above 0.1 kW, then the reactor power shall be lowered to 0.1 kW immediately and corrective action taken, as required. This ensures that the direction of air flow is from the control room into the reactor cell and that potentially contaminated reactor cell air due to reactor operation is released through the ventilation system.

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2. Liquid level switches are provided on the fuel loading tank and actuate an alarm circuit when the tank is either low or too high. If the level is above the low-level alarm, it can be assured that the core tank is filled with water. The high-level alarm assures that adequate indication is given to the operator during the filling of the fuel loading tank that it will not overflow or denotes possible secondary system to primary system leakage.

3. A thermocouple in the primary water core outlet line senses the primary water temperature and reads out on a panel meter in the control room. A high-temperature warning alarm is actuated when the set point is reached to indicate a high primary water temperature.

4. Core inlet and outlet thermocouples provide for primary coolant core temperature differential. This information is utilized in combination with the primary coolant flow rate to provide information for a heat balance determination.

5. An NTR stack radioactivity air monitoring system is utilized (see Chapter 11). Separate detection channels and alarms are used for particulate material and nonfilterable radioactive gases to assure that the releases are acceptable.

6. A low power level rod block and alarm is provided on the linear power system. This rod block and alarm assures that the operator has a linear power channel operating and indicating neutron flux levels during rod withdrawal.

7. Interlocks are provided on the control rods to prevent outward movement unless the safety rods are all in a full-out position. This condition assures that the reactor will be started up by withdrawing the four safety rods prior to withdrawing the control rods. A bypass is provided for testing purposes.

8. Interlocks are also provided on the safety rods. Each safety rod must be withdrawn in sequence to assure the normal method of reactivity control. A bypass is provided for testing purposes which will allow any one safety rod or safety rod drive to be withdrawn.

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Other non-safety related items providing indication and alarm functions include:

1. A flow meter mounted in the control room to indicate the secondary coolant flow to the tube side of the heat exchanger.

2. A recorder to monitor thermocouples placed at several locations throughout the primary and secondary systems and the graphite pack.

3. A constant air monitor located in the control room to monitor the air activity in the reactor cell. The monitor is checked prior to each initial entry into the reactor cell.

4. A differential pressure switch senses the pressure difference between the core inlet and outlet lines. When core differential pressure, which is an indication of flow, falls below a preset value, an alarm circuit is actuated and indicates a low differential pressure in the core.

7.5. REACTOR REACTIVITY CONTROL SYSTEMS Three types of movable neutron poisons are included in NTR design to control core reactivity:

safety rods, control rods, and manual poison sheets. All these poisons are located about the periphery of the fuel container, and all run-in guides that extend from the south end of the fuel container through the reflector and shield to the north face of the reactor. The guides place the center of the poisons on a 9.5-inches radius or about 0.6-inch from the outside edge of the active core. The control and safety rods have horizontally mounted drive mechanisms that are supported from the north face of the reactor on a 5-foot-high aluminum support plate located about 4-1/2 feet in front of the north face.

The control rods (two coarse and one fine) were designed for the precise position control and indication required for analytical work during which the reactor is used as a detector.

The four safety rods were designed for rapid insertion to scram the reactor.

Figure 7-2 shows the control circuits for the safety rod and control rod drives. Refer to Section 4.4.2 for more details.

The manually positioned poison sheets are used to limit the reactivity available to the operator or to increase the shutdown margin. The manual poison sheets are designed to allow their manual movement or removal through access holes provided in the north shield. All but a partial sheet in position 2 have been removed as of January 2023.

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Figure 7-2 Simplified Block Diagram of Rod Drives 7-14

7.6. Control Console The reactor control console is a vertical metal structure approximately 6 feet high by 13 feet wide, designed to accommodate racks of standard 19-inch instrument chassis. Attached to the front of the vertical panel is a small sloping bench board that contains the controls and indicating devices for the rod drives and most of the lights and switches for the alarm system. Also attached to the front of the panel is a horizontal work surface for the convenience of the operator. The vertical panels contain visual readout, power supplies, and recording devices for nuclear process and stack effluent release parameters.

The reactor control console also provides a key lock switch by which incoming electrical power to the reactor control systems can be isolated.

Remote manual control of the control rod drives, including a manual scram button (Section 7.3.1), is by pushbutton switches at the control console along with indication lights for drive-in and drive out limit switches (Section 4.2.2).

7.7. RADIATION MONITORING SYSTEMS Radiation levels (gamma) are monitored by a five-station remote area monitor system and are provided for personnel safety (ALARA). Areas monitored are the (two stations). Radiation levels are indicated on the control console. Each channel is equipped with an alarm which will actuate visual and audible alarms in the room and the affected area. In addition, the cell monitor is interlocked with the cell shutter and door controls to prevent inadvertent exposure to the radiation beam from the reactor.

A functioning area monitor is required in the room during operations and in the subject experimental facility while experiments are in progress. A function area monitor may be an installed monitor or other gamma-sensitive monitor, but must provide a readout visible in the room and a local audible alarm 7.8. NEUTRON SOURCE A reactor start-up neutron source is installed on an electric motor drive mechanism, in a configuration like that of the control rod drives. The source drive has the same controls and indications as a control rod drive, with the exception that continuous position indication is not provided. The same interlocks as those on the control rods are provided (the safety rod magnets cannot be energized until the source is full in), except that it is not necessary to pull any safety 7-15

rods to withdraw the source. Following a process scram (de-energize scram relays), the source automatically runs to the fully inserted position. The source travels in a guide tube identical to that used for the control rods, and the limit switches are adjusted so that the source moves about 30 inches from the in-to-out positions. A 0.2-Ci radium-beryllium source emitting about 106 n/sec is used for a startup source. It is an R-Monel encapsulation approximately 1/2-inch in diameter and 3 1/2-inches long, attached to an aluminum extension rod that connects to the source drive mechanism. The source provides at least the minimum neutron flux signal required for the nuclear instrumentation for startup and gives good indication of subcritical multiplication.

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8 ELECTRICAL POWER SYSTEMS 8.1. NORMAL ELECTRICAL POWER SYSTEMS Electric service is supplied by PG&E to the Site substation where it is distributed to the Site facilities. The PG&E transmission line, passing just south of the Site, is fed from two directions in an electrical loop to ensure a most reliable and continuous parallel 60-kV supply to the substation.

This substation feeds electrical power to the NTR.

Power supplied to the console is used for reactor instrumentation and control rod and safety rod drive motors.

Upon loss of electrical power to the facility, the four safety rods will scram, and the three control rods will fail as is. All non-operations and non-nuclear safety personnel are evacuated from the control room, north room, and south cell. Procedures are in place to ensure nothing is done to increase the reactivity of the reactor. Radiation readings are to be taken to verify the reactor is shutdown. The rod insert bus breaker remains closed so that the control rods will insert automatically when normal power is restored.

Unless rods are fully inserted, the restoration of normal AC power will result in automatic inward control rod movement. Therefore, if normal AC power is lost and all rod drives are not fully inserted thereby securing the reactor:

a licensed operator will remain in the control room, and a second trained individual will be on site, and 8-1

a licensed SRO will be present or readily available on call; that is, capable of arriving at the NTR within a reasonable time (1/2 hour / 30-mile radius) under normal conditions.).

If the power outage occurs while the reactor is shutdown, a barrier is placed across the reactor cell doorway if it is open. If the power outage occurs during reactor shutdown and the reactor cell door is open and the reactor has been operating less than one-hour before, a plastic sheet is taped over the cell doorway to mitigate any uncontrolled release of radioactive material.

8.1.1 Safety Rod Magnet Power Supply This power supply is a regulated, constant voltage/constant current DC Hewlett Packard Model 4633A. The ranges are 0-150 VDC and 0-3 amps. The unit is operated at 60 VDC constant voltage resulting in a current of 0.75 amp to the safety rod magnets. This current will decrease approximately 50 mA as the magnet coils heat up.

The power supply is regulated to less than 18 mV change for 105 to 125 VAC input variation and less than 36 mV variation for a 0 to 10-amp load change. The power supply has an external jack located on the front panel.

8.1.2 Safety System Power Supplies There are two power supplies for Safety Systems.

8.1.2.1. The power supply that provides power to the scram system relays, the console annunciator lights and relays, and the rod drive motor controls, 8.1.2.2. The power supply that provides power to the power switches, the logic elements, and 8-2

Figure 8-1 Parallel Offsite 60 KVA Power 8-3

8.1.3 Log N Power Supply This power supply is a General Electric INMAC 194X606G1. It receives 120-VAC input and supplies positive and negative 24-VDC, 10-amp output for the Log N amplifier. It is temperature compensated to operate at a constant output between 5 and 50°C. The power supply is regulated to +/-5% for a 10% change in line voltage and to +/-5% for a 0-5 amp change in load.

8.1.4 Picoammeter Power Supply This power supply is identical to the Log N power supply described above. It receives 120-VAC input and supplies positive and negative 24-VDC, 10-amp output for the three pico-ammeters.

8.2. EMERGENCY ELECTRICAL POWER SYSTEMS The NTR has no emergency power system.

Semiportable emergency lighting units are installed at several locations in the facility. A battery maintains its charge from the normal 115 VAC circuits and energizes the lights upon loss of AC power. These lights provide for safe personnel egress and are not credited in Chapter 13 for accident response.

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9 AUXILIARY SYSTEMS 9.1 HEATING, VENTILATION, AND AIR CONDITIONING SYSTEMS The control room and office spaces at the NTR share a central HVAC unit with other non-NTR workspaces in Building 105. The Setup Room is also provided HVAC by way of an independent central HVAC unit. There are no safety features or controls associated with these HVAC systems.

Potentially contaminated air and other gases are collected in the reactor cell ventilation system.

Ducts draw air from the reactor cell, the ceiling of the south cell, and from the forward position (for neutron radiography of radioactive objects) in the Modular Stone Monument in the north room. Air is then passed through absolute filters before being discharged from the NTR stack.

This ventilation system does not have heating or air conditioning capabilities.

The reactor cell ventilation system operates whenever the reactor is operated above 100 watts to maintain negative pressure in the reactor cell or during activities that could release airborne radioactivity into the reactor cell. For more details refer to Chapter 11.

9.2 HANDLING AND STORAGE OF REACTOR FUEL The NTR core fuel assemblies were installed in 1957 and have operated satisfactorily since then.

There are no unirradiated fuel assemblies or discs stored on site. Refer to Section 4.2.1 for more information about NTR fuel.

Facilities are available to perform any fuel handling operations that might become necessary in conjunction with operation of the reactor. A handling tool is provided for remote underwater transfer of the assemblies between the core and the fuel loading tank. If it is necessary to remove more than one fuel assembly, special arrangements may be made to use a shielded transfer cask and storage facilities elsewhere on the Site. Proper authorization would be obtained before such transfers were made, and procedures would be developed to ensure safe handling with adequate consideration for radiation protection and criticality control.

The fuel loading tank, approximately 12 feet (3.66 m) high by 5 feet (1.5 m) long by 4 feet (1.2 m) wide, is located adjacent to the west face of the graphite reflector. An expansion joint connects the west end of the reactor fuel loading chute to the east side of the tank. An aluminum gate for the loading chute is attached to the inside of the east wall of the tank and is normally in a 9-1

partially closed position. A loading platform which essentially extends the loading chute into the tank may be attached to the inside of the east wall of the tank, as required. Affixed to the opposite tank wall is a storage rack for the fuel loading chutes aluminum-covered graphite plug.

A pulley for the plug cable-lift is attached to the storage rack. Two 3.5-in.-diameter and one 2-in.-diameter aluminum thimbles installed in the tank are used to hold detectors for reactor nuclear instrumentation or samples for irradiation in a low-flux region. Level switches indicate high and low water level in the tank by energizing annunciator lights at the console. Access to the loading tank is from the reactor cell mezzanine. With a normal level in the tank, there is about 5 feet of water for shielding above the loading platform.

9.3 FIRE PROTECTION SYSTEMS AND PROGRAMS The design of the building containing the reactor and the reactor itself makes maximum use of noncombustible structural material. A 500,000-gallon water storage tank on a hillside above the Site provides a gravity flow water supply to the Site, including the fire protection system. A fire alarm system initiates an alarm in building 105 when flow is initiated in the fire header.

Designated personnel are trained to extinguish incipient fires.

Equipment, buildings, procedures, etc., are in accordance with company-wide standards, state and local regulations, and the recommendations of insurance agencies.

Six-inch fire mains, which are legs of a loop surrounding Buildings 102 and 105, are located on the east and west sides of Building 105. These mains supply outdoor fire hydrants located at the northeast, southeast, and southwest corners of Building 105, and a sprinkler system located within the building. Fire hoses and nozzles are maintained in the building 105 hallway and the southeast corner of the building. In addition to the fire protection water system, conventional portable fire extinguishers are located throughout the NTR facility and Building 105. The balance of fire protection is according to the VNC sitewide fire protection plan.

The sites radiological emergency plan includes information about offsite organizations that provide emergency support.

Flammable liquids and combustible materials are limited and controlled in all areas. These are regularly checked by the building safety inspections and by insurance carrier audits.

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The reactor shall be shut down immediately if a fire occurs in the control room, the south cell or the adjacent hallway or laboratory rooms. In the event of a fire in another part of Building 105, reactor operation may continue if the fire is small and contained.

Due to the control of flammable materials, administrative controls, and facility design, a fire in the NTR would not result in more than a minimal release of radioactive material.

9.4 COMMUNICATION SYSTEMS The NTR is a small, simple, compact facility. Most work is performed in the south cell which is entered from the control room. Coordination with experimenters is minimal and is accomplished with an intercom between the control room to the north room. Standard telephones in these areas may also be used.

Numerous means of communication available for normal operating, maintenance and emergency situations include two-way radios (which are limited to areas where radio frequency sensitive pyrotechnic material is not present), a local public address system activated from the

, and a High-level Conference Circuit (HICON), that provides an open line to the security building.

9.5 POSSESSION AND USE OF BYPRODUCT, SOURCE, AND SPECIAL NUCLEAR MATERIAL The NTR facility license, R-33, applies to Byproduct, Source, and Special Nuclear Material needed for operation of the reactor and its experimental programs. All Byproduct, Source, and Special Nuclear Material used in other laboratory areas of Building 105, and other locations at the VNC, is possessed and used in accordance with other licenses issued by the NRC or the State of California. Material received for irradiation at the NTR is according to license R-33. The R-33 license also authorizes the receipt, possession, and use of activated solids and byproduct materials used for instrument calibrations, startup sources and as may be produced from reactor operations.

9.6 COMPRESSED AIR Compressed air for the facility is supplied from the building 105 service air compressor located in the second-floor mechanical equipment room. The compressor will deliver 50 scfm of air and is capable of a discharge pressure of 100 psig. A relief valve at the air compressor maintains system pressure at less than 120 psig. A low-pressure switch provides an audible and visual 9-3

alarm. Compressed air is supplied to the air piston operator for the south cell door and to an air-operated shutter used for radiation shielding for the south radiation beam. Conveniently located outlets are provided to supply service air. A loss of compressed air would cause the south cell door and the south shutter to fail in the as-is position. This failure would have no safety significance.

9.7 RADIOGRAPHY VACUUM SYSTEM Two vacuum pumps provide vacuum to the neutron radiography areas. One pump provides vacuum to the dark room, south cell, and control room. The second pump provides vacuum to the north room.

The neutron radiography vacuum film cassettes are connected to these vacuum sources so that the radiography film maintains intimate contact with the conversion screen in the cassette for a quality radiography image.

The vacuum system is not used for any reactor system and has no safety significance.

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10 EXPERIMENTAL FACILIITIES AND UTILIZATION 10.1.

SUMMARY

DESCRIPTION The NTR is used primarily for nondestructive testing of materials and for irradiating various types of materials. The experiment facilities at the NTR are defined as locations for experiments on or against the external surfaces of the main graphite pack and thermal column including the horizontal facility, vertical facility, fuel storage tank, Cable Held Retractable Irradiation System (CHRIS) facility, and the fuel loading chute. These facilities, according to types, are as follows.

10.1.1 Incore Facilities The Horizontal facility is used for three different types of experiments depending upon its configuration. It can be configured with a source log and pinhole unit and used for neutron radiography (nondestructive testing) or filled with graphite and used for reactivity testing of materials (nondestructive testing), or if completely empty it is used for an irradiation facility.

10.1.2 Inreflector Facility The vertical facility extends vertically from the top of the graphite pack, through the graphite reflector, tangential to the east side of the fuel container.

Another experiment location is the fuel loading chute which extends diagonally through the graphite pack from the fuel loading tank to the reactor core.

10.1.3 Inreflector and Automatic Transfer Facility The CHRIS is a dry tube that allows access for a cable-held carrier to an experiment position during reactor operation. The experiment position is a horizontal tube in the graphite pack in line with and parallel to the horizontal facility approximately 6 inches above the top of the reactor core can. The CHRIS is discussed further in section 10.3.4.

10.1.4 Thermal Columns There are three experimental areas that utilize the radiation coming from the external surfaces of the main graphite pack and thermal column. Two of them, the Top and East Face, are in the reactor cell on the top and to the east side of the main graphite pack. The third, the Thermal Column, is located on the south side of the reflector.

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Entry into the reactor cell is necessary for access to the Face facilities; therefore, objects to be irradiated are generally positioned before reactor startup or are provided with remote positioning devices.

Access to the Thermal Column is from the south cell. Sections of the biological shield may be removed to provide access to the graphite reflector or to permit use of the radiation beams.

10.2. SECURED / MOVABLE EXPERIMENTS Occasionally the need arises to conduct an Irradiation with either the vertical or horizontal facility occupied by an object which absorbs neutrons strongly, such as a boron-shielded counter or an ionization chamber. If the reactivity worth of the object Is more than $0.80, special precautions are taken to prevent removal of the object during operation:

(a) If the object is in the vertical facility, any leads which are brought out are secured to the reactor frame to prevent any lifting force from being applied to the test object.

(a) Motion of the object in the horizontal facility is prevented by blocking devices consisting of sets of graphite logs and/or metal tubes, which rest against either end of the object. At the south of the facility, these blocking devices are firmly secured to a heavy steel channel which Is anchored to the concrete walls of the cell. At the north end, the blocking devices are secured by metallic linkages to the reactor frame. In addition, microswitches are provided at either end of the facility to provide a signal If motion should occur in either direction.

10.3. EXPERIMENTAL FACILITIES 10.3.1 The Horizontal Facility The horizontal facility (Figure 10-1) is a 5-inch-diameter hole traversing the horizontal axis of the reactor. It contains a removable sleeve which makes it a 3-inch-diameter hole for current use. A source log which is 30.5 inches long and 2.96 inches in diameter is centered with the core can in the horizontal facility. The source log is an aluminum pipe containing pieces of graphite, lead and plastic to make the neutron beam more uniform. The thermal column end is stepped to 3.125, 5.0-, and 8.0-inches diameter. A pinhole is installed in the 3.125-inch-diameter area for proper focus of the neutron beam for neutron radiography.

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Figure 10-1 NTR Neutron Radiographic Facilities (Side View)

The horizontal facility is accessible from either the south cell or reactor cell. The south cell is provided with an air-operated radiation shield shutter that is also used for neutron radiography.

An electrically operated shutter is used at the penetration to the Modular Stone Monument (MSM), in the north wall of the reactor cell. Both shutters have their own timers and can be controlled from either the reactor control console or at the entrances to the south cell or the north room.

The MSM (Figure 10-2) is a dual neutron radiography facility, located in the north room, designed to allow neutron radiography on unirradiated or irradiated objects. The design involves six concrete blocks that make up the shield and structural unit. A 12-inch-ID stainless steel pipe, capped off at the bottom, penetrates the ground beneath the MSM for 20 feet. This penetration allows neutron radiography of long objects to be performed by lowering them into the pipe.

Mechanisms for changing neutron radiography imaging foils without returning irradiated objects to their casks have been incorporated into the design.

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Figure 10-2 Modular Stone Monument Neutron Radiography Facility Irradiated objects normally arrive at the NTR in large casks which are placed on the MSM, using the overhead crane. The objects can be then lowered down into the MSM in front of an imaging foil and the neutron radiography is then performed.

Unirradiated objects are moved into a facility on the north end of the MSM, usually by a trolley arrangement. The imaging system is placed behind the object and neutron radiography, or irradiation is performed.

Irradiations may be performed anywhere in the horizontal facility or in the radiation beams streaming from the tube ends. The pinhole or the pinhole and source log may be removed to install the sample. Material or equipment to be irradiated in the horizontal facility may be fastened to an extension rod and positioned manually from the south cell. Specially machined graphite logs suitable for use as sample holders for specific irradiations are available.

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Electrical leads, cooling lines, and associated equipment for instrumented devices can be brought out either end of the horizontal facility for connection to equipment in the reactor cell, south cell, or through available penetrations to or above the control room, set up room or north room.

During irradiation, reactor instrumentation, as well as any instrumentation associated with the irradiation device, is observed carefully. In the event of an unexpected change in neutron multiplication, critical rod position, radiation levels, or reactor irradiation device behavior, the operator will take whatever immediate action is required to ensure the safety of personnel, the reactor, and the irradiation device. The operator will then notify a Licensed Senior Reactor Operator, as required, who will evaluate the situation and initiate whatever further action is necessary.

Unloading of the horizontal facility is usually done with the reactor shutdown; however, if the reactivity effect of the sample and the radiation level permits, the sample may be removed with the reactor operating. If the radiation level from the sample is such that conventional tweezers or pliers that are normally used to handle samples are inadequate to properly control exposure, temporary shielding and remote handling tools will be utilized. To minimize radiation exposure to the operators, irradiated samples are normally allowed to decay to the extent practical before handling them.

10.3.2 The Vertical Facility The vertical facility is defined by a 4-inch2 by 5-foot-long aluminum can, which extends vertically through the graphite reflector tangentially to the east side of the fuel container. A piece of reflector graphite normally fills the facility when it is not in use. The facility is accessible only from the top of the reactor inside the reactor cell. Irradiations may be performed at any position within the aluminum can or in the beam streaming from the top.

Devices to be irradiated in the vertical facility are usually attached to a wire or extension rod supported from the top shield of the reactor. Since entry into the reactor cell during critical operation is normally forbidden, samples must be positioned manually before startup or provided with a means for remote positioning. Space not occupied by the device may be filled with graphite blocks. Leads from the device can be brought out of the top of the facility for connection to equipment in the reactor cell or through cell penetrations to or above the room, room, or room.

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The precautions and procedures during irradiation are the same as those discussed for the horizontal facility. The only significant difference is that the shut-down radiation level of the reactor core may contribute appreciably to the radiation exposure received during the unloading.

However, radiation monitoring is always required for sample unloading, which ensures that the operator is aware of radiation from all sources. In addition, the cell area radiation monitor (with readout in the control room) would indicate unexpected high radiation levels.

10.3.3 Fuel Loading Chute Facility The aluminum-covered graphite plug for the fuel loading chute may be removed to provide access to the inside of the fuel container for irradiation. Use of this facility is necessary for performing experiments, such as some of those required to determine the nuclear characteristics of the reactor. No experimental objects are permitted inside of the core tank when reactor power is greater than 100 watts. Experimental objects in the fuel loading chute will be secured (secured experiment) to prevent their entry into the core region during normal operating conditions.

10.3.4 CHRIS Facility The CHRIS is a dry tube that allows access for a cable-held carrier to an experiment position during reactor operation. Entrance to the irradiation system is from the NTR North Room mezzanine. The tube runs through the NTR reactor cell wall, across the reactor cell, and connects to the experiment position.

The peak thermal neutron flux in the experiment position (approximately at the core centerline) is approximately 8 x 1011 n/cm2-sec.

Samples to be irradiated are placed in a carrier. The carrier is a capped aluminum tube that slides through the irradiation system. One cap is removable for sample loading and unloading.

A flexible cable is attached to the carrier to insert it into and retract it from the irradiation system.

At least 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months of decay is allowed after irradiation before handling the irradiated portion of the cable.

10.3.5 Top Face and East Face Radiation escaping through any one of the faces of the 5-ft3 graphite reflector may be utilized for experimentation. The aluminum box (partially cadmium lined) that contains the reflector is provided with 4-foot by 4-foot removable sections on the top and east faces. The removable 10-6

section for the east face contains a 20-inch by 18-inch hinged section, which can be opened to eliminate the cadmium from this area. Limited space between the reflector and the top shield slab can be used without removing the plug in the shield.

Entry into the reactor cell is necessary for access to the Face facilities. Therefore, objects to be irradiated are generally positioned before reactor startup or are provided with remote positioning devices. Irradiation devices utilizing the Face facilities have a negligible effect on core reactivity.

10.3.6 Thermal Column The thermal column is a 4-ft cube of high purity graphite located against the south side of the reactors 5-ft cube graphite reflector. The thermal column is traversed by the horizontal facility.

A 20-inch-wide by 20-inch-high by 4-foot-long centrally located section, made of 4-inch by 4-inch by 4-ft graphite logs was designed to be removed partially or entirely to accommodate experiments inside the thermal column or to provide external radiation beams. The south face of the thermal column is accessible from the south cell. Radiation shielding on this face consists of a 0.375-inch boral plate, .

Sections of the biological shield may be removed to provide access to the graphite face of the thermal column or to permit use of radiation beams. An air-piston operated shutter is installed at the face of the lead brick wall to provide shielding from the horizontal cavity.

Irradiations using the thermal column normally have a negligible effect on core reactivity and may be loaded or unloaded with the reactor operating or shutdown. Radiation exposure to the operator is usually of more concern during such activities than the effects of the irradiation device on the reactor.

10.4. EXPERIMENT REVIEW Safety-oriented limits and restrictions applicable to experiment facilities and experiment programs follow. The limits and restrictions presented are derived from the reactor and experiment safety analyses, approximately 60 years of experience in conducting experiments at the NTR, and sound engineering practice. Most of these limits are contained in the Technical Specifications. Adherence to the limits and restrictions below is mandatory and provides assurance that:

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There is no anticipated mode of experiment operation that will endanger the health or safety of the general public or plant personnel.

No experiment will be performed that involves a technical specification change or that requires prior NRC approval according to 10 CFR 50.59.

A proposed experiment type will be evaluated in detail and its execution controlled to reduce any radiation exposure to the public and plant personnel to the lowest practicable level.

10.4.1 General Experiment Requirements A written description and analysis of the possible hazards involved for each type of experiment shall be evaluated and approved by the Area Manager or a designated alternate before the experiment may be conducted. Records of such evaluation and approval shall be maintained.

No irradiation shall be performed which could credibly interfere with the scram action of the safety rods at any time during reactor operation.

Experimental capsules to be utilized in the experimental facilities shall be designed or tested to ensure that the pressure transients, if any, produced by any possible chemical reaction of their contents and leakage of corrosive or flammable materials will not damage the reactor.

No experimental objects shall be inside the core tank when the reactor is operating at a power greater than 0.1 kW.

Experimental objects located in the fuel loading chute shall be secured (secured experiment) to prevent their entry into the core region.

10.4.2 Reactivity Limits 10.4.2.1 Requirements pertaining to the reactivity worth of experiments are as follows:

a. The sum of the potential reactivity-worths of all experiments which coexist plus the reactivity available from control rods and coolant temperature shall not exceed $0.76.

b. No experimental object shall be moved during reactor operation unless its potential reactivity worth is known to be less than $0.50 and the operation is 10-8

performed with the knowledge of the licensed operator at the console. All power operated, remotely controlled mechanisms for moving an object into the reactor core shall be energized from the reactor console; however, movement of the object may be initiated from another location. All manually operated mechanisms for moving an object into the reactor graphite pack shall be done with the knowledge and consent of the reactor operator at the controls of the reactor.

c. The potential reactivity worth of any component which could be ejected from the reactor by a chemical reaction shall be less than $0.50.

10.4.2.2 The potential reactivity worth of experiments shall be assessed before irradiation. If the assessment warrants, the reactivity worth of the experiment shall be measured and determined acceptable before reactor full-power operation.

10.4.3 Explosive and Flammable Material Lists Limits for explosives and flammable materials are developed in Chapter 13 and listed in Section 3.8.3 of the Technical Specifications.

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11 RADIATION PROTECTION PROGRAM / WASTE MANAGEMENT 11.1 RADIATION PROTECTION 11.1.1 Radiation Sources Airborne Radioactive Sources The most significant source of airborne radionuclides during the normal operation of the NTR is the thermal neutron activation of the naturally occurring argon gas in the air and dissolved in the cooling water in and around the reactor core. A small contribution to the total airborne radionuclides can occur from the gaseous fission products emitted from the trace quantities of uranium which may have contaminated the aluminum skin of the reactor fuel elements during fabrication. The radioactive noble gas isotope Ar-41 is the predominant radionuclide emitted from the NTR reactor cell through the exhaust ventilation system.

It can be demonstrated using fission product generation and decay codes that the activity inventory of fission produced halogens and fission produced noble gasses are approximately equal at shutdown (zero decay time). Further it can be shown by the same method that the ratio of the total fission produced noble gasses is approximately 46 times the I-131 activity at shutdown.

Therefore, if the measured I-131 release rate is known, and the measured total noble gas release rate is known, the quantity of the total noble gas release due to fission product noble gasses can be estimated. By this process, the fraction of the total noble gas release due to fission produced noble gas turns out to be a very small fraction of the total noble gas release, and therefore it is confirmed that most of the measured noble gas release during normal operation is due to Ar-41.

For example:

During 2018, which is typical of recent years of operation, the total measured release of I-131 from the NTR stack was calculated to be 6.2 µCi.

The total measured release of noble gas during the same period was about 190 curies.

A radioisotope inventory of the NTR core calculated with the Radioisotope Buildup and Decay, 1RIBD, computer code showed that the ratio of fission produced total noble gas activity to the I-131 activity at shutdown was approximately 46:1.

1 RIBD, Radio Isotope Buildup and Decay Code and Library, RSIC Computer Code Collection, CCC-137 11-1

Applying the 46:1 ration to 2018 data, the total activity of fission produced noble gas released during 2018 was approximately 285.2 µCi (46 times 6.2 µCi), assuming equal fractions of I-131 and noble gasses were released from the core to the exhaust system.

The difference of 190 curies minus 0.0002852 curies is the activity of Ar-41 released during the year.

The other general groups of isotopes which are measured in the stack effluent consist of gross beta particulate and gross alpha particulate. Both releases are equivalently low like the I-131 releases.

For example:

The total particulate activities released from the NTR stack during 2018 were measured to be:

Gross Beta Activity = 0.560 Ci Gross Alpha Activity = 0.021 Ci The committed effective dose equivalent due to exposure to these releases is demonstrated to be very low.

Liquid Radioactive Sources The only liquid radioactive source at the NTR is the primary coolant. Primary contributors to the radioactivity of the primary coolant are N-16, produced during reactor operation, and activated sodium (Na-24) from the aluminum primary coolant piping, also produced during reactor operation. The primary coolant is sampled periodically to monitor fuel leakage into the primary system. The primary coolant system is vented to a holdup tank prior to startup. The amount vented is small enough that the water in the holdup tank evaporates and the tank does not fill.

Dose rate measurements of the holdup tank indicate that no long-lived radioactive material accumulates in the tank. The only liquid radioactive waste generated is a result of sampling -

approximately one liter for each sample. This waste is placed in tanks with other laboratory generated liquid radioactive waste and subsequently disposed of in accordance with approved site practices and procedures.

Solid Radioactive Sources Solid Radioactive Sources at the NTR (Table 11-1 and Table 11-2) consist of: the reactor itself during operation; the reactor fuel; a Radium-Beryllium neutron source used for pre-startup instrument checks; an ion exchange demineralizer and filter system for the primary coolant system; spent power reactor fuel rod sections which are received from the hot cell facilities, 11-2

neutrographed as part of a NDE process, and returned to the hot cell facilities; activated experiments and neutrographed parts and fixtures which are exposed to the neutron beam or irradiated by placement in the horizontal cavity; byproduct material in experiments; instrument check sources; and solid radioactive waste.

Table 11-1 STANDARD, CHECK, AND STARTUP SOURCES AT THE NTR Isotopic Radiation Type / Sealed/

Source Activity Location Form Content Energy (MeV) Unsealed 1 Cl-36 - / 0.709 8 Ci CAM Cs7A2 Solid Sealed

- / 0.514 Stack Monitor, 2 Cs-137 8 Ci Solid Sealed

/ 0.662 Setup Room Stack Monitor, 3 Cl-36 - / 0.709 0.02 Ci Solid Sealed Setup Room Radium-4 Neutron* 0.2 Ci Graphite Pack Solid Sealed Beryllium 0.767 5 Sr-90 - / 0.546 Control Room Solid Sealed mCi

Neutrons from this source are characterized as 1.0E+06 n/sec.

Table 11-2 FISSILE AND FISSIONABLE MATERIAL AT THE NTR Original Enrichment Current Enrichment 93.17% (wt.% U-235) 91.54% (wt.% U-235) grams U grams U-235 grams U grams U-235 grams Pu NTR Fuel In-core The quantity of solid waste generated by NTR activities is very small. A more detailed description is provided in Section 11.2.3.

Solid radioactive sources in the form of experiments vary depending of the nature of the material and the experiment to be performed. Other radioactive materials present in the NTR facility are limited by the R-33 license.

Typical radiation levels for occupied or accessible areas of the facilities under license R-33 are discussed in Section 11.1.5. The radiation shielding of the facility is also described in Section 4.3.

Based on the last five years of dosimetry for personnel at the facility, the estimated maximum annual dose to a single worker is 862 mRem, and the average dose to all workers is 443 mRem.

No single worker at NTR exceeded a total of 2.5 Rem over this period. This is well within the 10 11-3

CFR 20.1201 requirements for occupational exposure. Exposure to individual members of the public as a result of NTR operations is essentially zero due to the existence of a restricted area, an access-controlled area, and the distance to the site boundary.

11.1.2 Radiation Protection Program Organization and Minimum Qualifications The radiation protection program for the site, including the NTR, is the responsibility of the Regulatory Compliance (RC) organization. The Manager, Regulatory Compliance (RC), has overall responsibility for the RP program. The Radiation Safety Officer (RSO) is responsible as the site radiation safety function leader for the ongoing implementation of the program and reports directly to the Manager, RC. The Manager, RC, may alternatively serve as the RSO if applicable radiation protection experience requirements are met.

The Manager, RC, shall hold a bachelors degree or equivalent and have two years of management experience in assignments involving regulatory activities or a high school diploma and five years supervisory or technical experience in a nuclear, manufacturing, or other technical field.

The RSO shall hold a bachelors degree in an engineering or scientific subject or equivalent (based on a combination of education and experience) and have more than two years of experience in applied radiation protection. Alternatively, a professional certification in health physics (CHP) may be credited for two years of experience in radiation protection.

Technical personnel within the RC component shall have at least an associate degree or equivalent technical experience in the nuclear industry, and enough professional experience to provide authoritative and competent discharge of assigned responsibilities. Radiation monitoring technicians are trained and qualified in accordance with a comprehensive VNC certification program.

Additional support is provided by GEH subject matter experts in the regulations and standards, quality, and operations organizations.

The staffing level for the RC organization is dependent on the level of activity at the site. Staffing for the RC activities for the NTR includes those necessary to perform health physics monitoring and nuclear safety oversight.

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RP Program Implementation The RP program for the NTR is a subset of the broader VNC site-wide RP program and is implemented by procedures that ensure compliance with all applicable federal and California regulations. Procedures implement the use of Radiation Work Permits to ensure safe, authorized work in restricted areas. RWPs provide information and instruction to the worker and prescribe necessary precautions and protective equipment when performing tasks in those areas.

The following is a list of program areas covered by implementing procedures:

- Startup, operation, and shutdown of the reactor.

- Defueling, refueling, and fuel transfer operations, when required.

- Preventive or corrective maintenance which could have an effect on the safety of the reactor, including the replacement of components.

- Surveillance checks, tests, calibrations, and inspections required by the Technical Specifications.

- NTR-specific radiation protection program implementing procedures for personnel safety consistent with applicable regulations or guidelines. Management commitment and programs to maintain exposures and releases as low as reasonably achievable are a component of the site-wide radiation protection program.

- Administrative controls for operation and maintenance and the conduct of experiments that could affect reactor safety or core reactivity.

- NTR-specific implementing procedures for the site-wide emergency and security plans.

- NTR-specific radiation protection program implementing procedures for the use, receipt, and on-site transfer of by-product material for such activities performed under the R-33 license.

The use of Change Authorizations and Engineering Releases for NTR activities provide documentation of changes and work and includes the determination whether the proposed change requires prior NRC approval pursuant to 10 CFR Part 50.59.

All procedures at VNC are administratively controlled to ensure procedures and changes that impact the radiation protection program are reviewed for adequacy, approved by authorized personnel, and distributed to the applicable staff.

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RP Training Radiation safety training is defined by procedure, managed by the RSO, and implemented by the responsible managers. Procedures describe radiation safety courses as well as how they are developed and maintained, delineate responsibility for ensure training is performed, and ensure training is documented.

Radiation Monitoring Technicians (RMTs) are trained and certified in accordance with a comprehensive Health Physics program that covers all site operations, including those at the NTR.

NTR personnel receive radiological safety training per the Reactor Operator Initial and Requalification programs. In addition, they receive annual radiation safety refresher training for site radiation workers and personnel assigned to site response teams.

RP Program Oversight The Vallecitos Technological Safety Council (VTSC) is the review committee associated with the activities of the site as a whole and the NTR (see Ch 12).

The radiation protection program is reviewed each year pursuant to 10 CFR 20.1101(c), in a report to the site manager. The VTSC reviews the report annually and determines the effectiveness of the program. The VTSC also receives and reviews incident investigation reports and countable event reports and uses all the information to implement program improvement and to ensure root causes are determined and effective corrective action is taken.

RP Program Records Radiation program safety records are generated per procedures. Procedures direct record retention requirements and reviews of those records. Records are required to be stored in a safe location and be easily retrievable. Radiation program safety records are used to develop trends, inform management, develop ALARA actions, and reporting to regulatory agencies. Radiation safety records include, employee exposure records, visitor and contractor exposure records, training records, medical records, radiation monitoring records, RWPs, VTSC meeting minutes, and the site decommissioning file.

11.1.3 ALARA Program The radiation program at NTR includes a commitment to maintain radiation exposure as low as reasonably achievable, ALARA. The VNC site manager, who has overall responsibility for the NTR license, and the Manager, RC, who has responsibility for radiation safety, are fully committed to the ALARA principle and oversee its effective implementation across the site. The 11-6

Manager, NTR, is fully committed to the ALARA principle and ensures its proper procedural implementation specific to the NTR.

During the first quarter of each calendar year, the Manager, NTR, reviews the NTR ALARA program and prior year accomplishments and establish goals for the current year. All licensed NTR operators review each radiation exposure report. The Manager, NTR, also reviews radiation exposure records for all NTR personnel. Any unusual exposure is discussed, and a probable cause determined. Individual workers, both NTR, and non-NTR personnel, are adequately trained for the job and periodically reminded of ALARA principles. ALARA is considered when facility changes are made, via the Change Authorization (CA) process, when new or changed experiments are reviewed, and when major maintenance is planned.

11.1.4 Radiation Monitoring and Surveying Radiation monitoring and surveying include the following:

Monitoring and Surveying routines performed by an RMT Special monitoring and surveying by RMT Fixed air sampling system Stack Monitoring system Continuous air monitor (CAM)

Area Radiation monitors (ARMs)

Smear surveys documented by NTR operations personnel Personal dosimeters Sampling and counting of Industrial Wastewater prior to release.

Radiation monitoring equipment at the NTR includes the equipment in Table 11-3.

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Table 11-3 RADIATION MONITORING EQUIPMENT AT THE NTR Description Location Function Control Room Dose rate (radiation fields);

Portable survey North Room , , n instruments Setup Room Hallway outside Control Room Contamination; , ,

Continuous Air Monitor Control Room Reactor Cell Air monitor (CAM)

Hand and Foot Counter Hallway near building exit Contamination detector Readout: Reactor panel, control room Detectors:

Area Radiation Monitor North Room (adj to CHRIS) Radiation field detection (ARM)* South Cell alarm Reactor Cell Control Room North Room (MSM)

Control Room South Cell Airborne contamination Fixed Air Filters Reactor Cell (2) detection North Room (3)

Stack particulate and noble Stack Monitor System Setup Room gas monitoring alarm Personal Dosimetry NTR Personnel Exposure monitoring

The ARM system is installed to ensure protection of personnel by ensuring exposures are maintained ALARA.

Records of monitoring and surveys conducted by RMTs are maintained, reviewed, and archived, and undergo independent review by RC personnel. Dosimetry and fixed air filter records are also maintained, reviewed and archived by RC personnel. Stack monitoring records and CAM records are maintained and archived by NTR personnel and reviewed by RC personnel.

Procedures are maintained to ensure the proper calibration of radiation protection instruments.

Calibration of radiation protection instruments used at the NTR is required upon initial acquisition, after major maintenance, and at least annually. Radiation monitoring instruments are calibrated on-site by VNC Instrument Maintenance or by approved offsite vendors. Calibration sources used for calibration are traceable to NIST standards. Radiation monitoring instruments are controlled, and timely calibration is assured by RC personnel.

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Monitoring and Surveying routines by a VNC RMT for the NTR are directed by procedures.

The frequency of routine surveys is based on potential for changes to affect the radiological situation.

11.1.5 Radiation Exposure Control and Dosimetry Radiation exposure control is achieved at the NTR by shielding, the ventilation system, security, entry control devices, an active ALARA program, the radiation protection program, environmental monitoring, equipment and materials, and through the VNC dosimetry program.

Shielding and typical radiation levels for occupied or accessible areas of the NTR facility are discussed in Section 4.3. Reactor cell ventilation is discussed in Section 3.5 and in Chapter 9.

The ALARA program, the radiation protection program, and the environmental monitoring program, are described in Sections 11.1.3, 11.1.2 and 11.1.7 respectively, of this Chapter.

Equipment and materials used in radiation exposure control consist primarily of protective clothing and respiratory protection equipment. Procedures provide for type and application of protective clothing in high radiation and/or radioactive materials areas.

The respiratory program ensures:

proper respiratory equipment is available and assigned, work environment respiratory hazards are evaluated, air sampling and analysis is adequate to identify hazards, individual personnel exposures are evaluated, respiratory protective equipment is properly selected and assigned in accordance with 10 CFR Part 20, minimum qualification requirements are set forth that include initial and periodic training, refitting, and medical clearance requirements.

Radiation exposure to NTR personnel varies somewhat from year to year, however, the expected average annual exposure, based on historical records, is ~440 mRem per person. Non-NTR personnel, working in the same building, again based on historical exposure records, are expected to receive < 100 mRem per year per person from radiation associated with the NTR. Site personnel providing service for the NTR, are expected to have a total annual exposure, from all 11-9

site sources of < 1.0 Rem, with an estimated < 30% of that exposure attributable to NTR sources of exposure.

Because it is unlikely that any individual would receive, in one year, an intake in excess of 10 percent of on ALI(s) of Table 1, Columns 1 and 2 of Appendix B to 10 CFR 20, committed effective dose equivalent (CEDE) is not typically added to external dose for determination of total effective dose equivalent (TEDE) at VNC. The philosophy at VNC and the NTR is to use procedures and other controls to limit the intake of radioactive materials during normal work in controlled areas. Air activity is controlled using ventilation systems and contamination control.

Intakes of radioactive materials is limited using respiratory protection devices, control of access and limitation on the time of exposure when other controls are impractical. Whole body counts are done routinely, based on the Manager, NTRs input, to confirm the lack of intake.

Exposure limits for occupational workers are pursuant to 10 CFR 20.1201 requirements and more restrictive administrative limits are established by procedures. Exposure to the Embryo/Fetus of declared pregnant workers is in accordance with 10 CFR 20.1208, 20.1502 and 10% of 20.1201 requirements as informed by Regulatory Guide 8.13, Instruction Concerning Prenatal Radiation Exposure, revisions 2, December 1987.

Procedures describe the exposure and dosimetry requirements for NTR Operations personnel.

Exposure control limits are given, over the full range of operations, including normal operations, emergency conditions and planned special exposures. Administrative dose action levels are established in accordance with the site RP Program and require management approval to exceed.

Personal dosimetry used at the NTR includes beta-gamma dosimeters, neutron albedo dosimeters, and electronic dosimeters (EDs). These are processed and read routinely or as necessary to estimate exposure between routine badge processing. Special use dosimeters such as TLD finger rings are issued for extremity exposure and high dose rate exposure according to the site RP program. Dosimetry records are kept indefinitely by RC for beta-gamma badge, neutron badge, and TLD finger ring exposures.

11.1.6 Contamination Control Contamination control at the NTR is accomplished through some of the elements of the Radiation Protection Program discussed in the previous sections. These elements include the ALARA program, routine surveys and monitoring, the dosimetry program including evaluation and testing for internal depositions, the use of anti-contamination clothing, training programs for staff and 11-10

visitors, and survey records. In addition, access control, area posting, and the use of RWPs are integral parts of contamination control at the NTR.

11.1.7 Environmental Monitoring The primary purpose of the environmental surveillance program is to obtain information essential to assessing and controlling the exposure of the neighboring population to industrial chemicals, radiation and/or radioactive materials. Secondary objectives include identifying the sources of specific contaminants that might be released, predicting trends in pollutant levels, and improving public relations by showing that the operations at VNC are not adversely affecting the health and safety of the public and surrounding areas. This program is responsive to several site licenses, permits, and procedures.

At VNC the overall environmental program is separated into two distinct categories: (1) effluent, monitoring and (2) environmental surveillance. Effluent monitoring for the NTR is limited to the ventilation stack. Sampling and analysis of water in the site retention basins is performed prior to release as a function of environmental surveillance and to coincide with required state industrial effluent sampling. This program provides measurements of the amount of radioactivity that is released to the environment in gaseous effluents and validates that licensed byproduct material does not make its way to the environment through the discharging of industrial wastewater.

Environmental surveillance covers all measurements and observations made of the environment on and adjacent to the site. This includes environmental air samplers and TLD stations; the sampling of water, vegetation, soil and stream bottom sediment; and the resulting radiological and chemical analysis. This program provides assurance that there are no deleterious impacts on the environment from operations conducted at the site.

A complete description of the current VNC environmental program is contained in the VNC Environmental Monitoring Manual. A method for determining action levels is established in implementing procedures and ensures specific actions are undertaken to determine the possible sources of the activity before regulatory limits are reached.

The specialist assigned to environmental protection is responsible to assure the requirements of the environmental protection program are met within the time frames established. This includes:

sample collection (method and frequency) and analysis (technique and sensitivity),

preparing required summary reports, and 11-11

assuring the proper installation, operation, and maintenance of environmental monitoring equipment.

The specialist assigned to environmental protection has been granted the necessary authority by management to meet these responsibilities.

Regulatory Compliance (RC) is responsible for reviewing the environmental protection program for adequacy and for recommending changes as necessary. Further, RC prescribes equipment in support of the environmental protection program and shall review periodically the activities of the specialist assigned to environmental protection.

11.2 RADIOACTIVE WASTE MANAGEMENT 11.2.1 Radioactive Waste Management Program Radioactive waste management at VNC addresses a broad range of activities across the site that are inclusive of the NTR. Radioactive Waste at the VNC is any material in which radioactivity is distributed or the surfaces of which are contaminated with radioactive material to levels that prevent release for unrestricted use, or which is potentially contaminated and cannot be shown to be less than these levels and which has no further functional or monetary value to the user or owner.

The radioactive waste management program at VNC is implemented by station procedures and sets forth management policy for the handling of low-level radioactive waste (radwaste) materials generated at VNC, including the NTR, establishes a radwaste handling program designed to meet the objectives of this policy, and defines the responsibilities for carrying out the program.

Responsibilities for program implementation for the NTR are appropriately divided among the Area Manager, MLO; RC; the Manager NTR; and those individuals performing radwaste activities. Procedures describe the criteria, method and responsibilities to be used at the NTR for collection, interim storage, identification, characterization, and transfer of low-level radioactive waste to the site inspection/packaging area.

VNC activities involved in the processing, packaging, transfer, receiving, interim storage, and shipment of radwaste comply with applicable federal and state regulations.

Low-level radioactive wastes undergo volume reduction as appropriate and are then packaged and ultimately shipped to licensed commercial radwaste disposal sites.

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Records and checklists associated with radwaste activities are maintained by the Area Manager, MLO. Such records consist of on-site transfers of radwaste to the MLO, shipment to off-site disposal facilities, packaging checklists, and personnel training records. These records are reviewed as part of the overall periodic radwaste program review.

11.2.2 Radioactive Liquid Waste Management The only liquid radioactive waste generated is as a result of the annual sampling, approximately one liter. This waste is placed in tanks with other laboratory generated liquid radioactive waste and subsequently disposed of in accordance with approved site practices and procedures. No liquid radioactive waste is released directly to the unrestricted environment.

Contaminated wastewaters created from NTR operation are processed in the site waste evaporator.

Evaporator bottoms are then processed and shipped as radioactive solid wastes. Industrial wastewater from the NTR single pass, non-contact, secondary cooling water heat exchanger is tested for radiological constituents as well as other potentially polluting constituents in accordance with a National Pollutant Discharge Elimination System (NPDES) permit prior to release to the environment.

11.2.3 Radioactive Solid Waste Management Low level dry active wastes, consisting of contaminated paper and plastic, filters, and resins, are infrequently produced and are also processed and shipped as a fractional component of the VNC solid waste stream. The quantity of solid waste generated by NTR activities is very small; estimated to be one to three cubic feet annually with the radioactive content measured in millicuries. Radwaste reduction techniques, include planning, decontamination, use of reusable vs. disposable materials, unpackaging (debulking) of supplies and equipment prior to transfer into a radioactive materials area (RMA), and dedication of appropriate tools and equipment to RMAs for reuse as needed. The site also uses commercial facilities, where cost effective, to perform radwaste volume reduction and recycling of contaminated materials, e.g., metal melt technology.

Solid radioactive waste is characterized, handled, packaged, surveyed and shipped in accordance with all applicable DOT and NRC regulations. Shipments of solid radwaste are intermittent at VNC and annual Curie makeup of those shipments in recent years have been dominated by unique shipments of byproducts of sealed source manufacturing and materials from the SAFESTOR reactor facilities.

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11.2.4 Radioactive Gaseous Waste Management Release of routine gaseous effluents is dominated by Ar-41, which is generated by neutron activation of Ar-40 in air. Airborne radioactive waste exiting through the NTR stack is monitored as radioactive effluent and is well within the Technical Specification and 10 CFR 20 requirements. Monitoring and alarms associated with the NTR stack have been discussed in this Chapters 7 and Section 3.5 of this SAR.

The reactor cell and stack ventilation system were originally required to mitigate an analyzed fueled experiment failure of a type that has not been performed at the NTR for more than 30 years. Nevertheless, the reactor cell will contain any radioactive release while it is exhausted through the ventilation system and out the stack.

11.2.5 Stack Release Action Levels The Stack Release Action Levels are defined as the release rates for each radionuclide group (noble gas, I-131, beta particulate, or alpha particulate) at which action should be taken to reduce the release rate. Ongoing operation below these Stack Release Action Levels ensures doses to members of the public due to airborne release are at or below the 10 CFR 20.1101(d) limit of 10 mrem per year. The method for establishing the Stack Release Action Levels is described below.

The VNC is a multi-licensed site that performs processes not associated with the NTR. These processes take place in other buildings apart from Building 105 that also emit radionuclides from stacks. This is relevant for two reasons.

First, contributions from all stacks to site boundary exposure must be accounted for in establishing the stack action levels for each stack. This is conservatively and simply done by assuming each stack produces half of the dose at the site boundary and applying an other stack reduction factor of 2.

Second, while all site stack action levels discussed in this section are tracked by the sites effluent program, only gases and particulates are included in Technical Specification 3.7.4. This is because the real-time NTR effluent monitors are only capable of detecting gaseous and particulate

(, ) releases, making these the only Stack Release Action Levels that are actionable by the NTR operators. Alpha particulate and halogen (I-131) releases are evaluated via samples collected and counted on a weekly basis and are monitored according to the sites effluent control program.

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Basis for Stack Release Action Levels The Stack Release Action Levels (weekly total or specific concentrations) for noble gas releases from the NTR stack ensure that the activity released will not exceed an annual average concentration of Ar-41 at the site boundary of 10% of the annual effluent concentration limit (ECL) in 10 CFR 20, Appendix B, Table 2 Column 1. Ar-41 has been shown to be the predominant noble gas in the stack effluent (Climent, 1969). Fission produced noble gases are a minor fraction unless fuel material is exposed to the effluent air. Ar-41 is produced by the neutron irradiation of the air passing through the reactor.

The Stack Release Action Levels for all other isotope groups ensure releases from each stack (including NTR) at the VNC will not yield an annual average concentration at the site boundary of 20% of the 10 CFR 20 effluent concentration limit for the restrictive, credible isotopes of each of the isotope groups: I-131, unidentified beta radionuclide, and Np-237.

Once reduced by a factor of 2 to account for other stacks, the Stack Release Action Levels ensure that any single stack release will not exceed an annual average concentration that is effectively 5%

of the ECL for Ar-41 and 10% of the ECL for other credible isotopes.

Computing Stack Release Action Levels To derive the Stack Release Action Level, a bounding release concentration rate (µCi/sec) must be determined that will not exceed the effective ECL (5% for gas and 10% for other). To do this, the effective site boundary ECL is divided by the atmospheric dispersion (/Q) factor (sec/ml) to determine the allowable release rate (µCi/sec) at the stack (point of release). The allowable release rate is then adjusted to a weekly release rate (µCi/week), which has been determined to be an actionable time period in which operator actions can be taken to ensure the stack release doesnt challenge the effective ECL. The allowable release rate is then adjusted for the average site building ventilation flow rate (1800 ft3/min) to determine the specific concentration Stack Release Action Level. For noble gas (Ar-41) and beta/gamma particulates, the specific concentration Stack Release Action Levels are applied in Technical Specification 3.7.4 as the Alarm Setpoints for the effluent monitors and are then further adjusted to provide weekly release Stack Release Action Levels based on a 30-hr operational week accounting for the fact that these concentrations are monitored in real-time.

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Table 11-4 STACK RELEASE ACTION LEVELS Nominal Noble Gas Halogen Alpha Beta Stack Flow Rate, Ci/wk mCi/wk µCi/wk µCi/wk cfm µCi/cc µCi/cc µCi/cc µCi/cc 9 3.48E+02 1.74E+01 1.74+E03 105, NTR 1.80E+03

9.5E-05 6.8E-07 3.4E-11 1.9E-08

Access to the reactor cell when the reactor is not operating is procedurally controlled by observation of CAM (see Table 11-3) activity readings.

Assumptions Stack flow rates fluctuate. For example, the NTR flow depends on the position of the cell door.

The flow in all filtered systems varies as the dust loading on filters increases and as containment systems are changed. The 1,800-cfm average stack flow rate for Building 105 is used for limiting concentrations and calculating measured releases.

The applicable effluent concentration limit values from Appendix B, Table 2, Column 1 of 10 CFR 20 are given below:

10 CFR 20 Effluent Limiting Release Category Concentration Limit, Isotope

µCi/ml Noble Gas* Ar-41 1.00E-08 Halogen I-131 2.00E-10 Alpha Particulate Np-237 1.00E-14**

Beta-Gamma Particulate *** 1.00E-12 The dilution-dispersion (/Q) factor and reduction factor to account for releases from other stacks on site are given below:

Other stack reduction Stack Location /Q, sec/ml factor Building 105, NTR 3.48E-11 2

The NTR noble gas inventory available to the boundary has been found to be primarily Ar-41, which is an activation product of air. Fission products would be of concern in the event of fuel failure, an abnormal condition.

- There are several isotopes with more restrictive limits, but they can be shown to be insignificant fractions of the typical mix of alpha emitters found at VNC.

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- - Unidentified isotopes, where several natural, transuranic, and other rare elements are known to be absent. These are mainly alpha emitters which would be accounted for in the alpha analysis.

The annual average dilution-dispersion factor for the NTR, and the other stacks at VNC, was calculated from valid hourly records of measured meteorological conditions for a two-year period in 1976 and 1977. The sector average /Q factors were conservatively computer calculated for each of 16 sectors (22.5 degrees each) using:

Scaled distances from a site layout map to determine the distances from the reactor to the center of the sector at the site boundary.

A building cross-section of 281 square meters, for wake effects.

A ground level release elevation.

No credit taken for plume depletion.

The single maximum calculated annual average /Q value of 3.48E-11 sec/ml was selected from the 16 sector average values. This value, which happens to occur in the east-southeast sector at 622 meters from the stack, is used to determine the NTR stack release limits.

The ECL release rate, i.e., the continuous release rate which would produce an annual average boundary concentration equivalent to the ECL, would be calculated by division of the ECL by the

/Q value. The Action Level rates are calculated as 5% of the ECL release rates for noble gas (10% of the ECL release rate, and a reduction factor of 2 for releases from other stacks). The Action Level Release Rates are calculated as 10% of the ECL release rates for other isotope groups (20% of the ECL release rate and a reduction factor of 2 for other stacks).

Isotope Group Action Level Release Rates µCi/sec Noble Gas 1.44E+01 Halogen 5.75E-01 Alpha 2.87E-05 Beta 2.87E-03 These conservative release rate limits are converted and presented as action levels based on cumulative weekly releases in Table 11-4 and the Technical Specification weekly release rate limits of Table 3-3 in Chapter 14, Technical Specifications.

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A normal maximum operating time for the NTR typically would not exceed 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months in a week.

Therefore, this partial operating time is used to calculate the operating stack effluent concentration limits.

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12 CONDUCT OF OPERATIONS 12.1 ORGANIZATION 12.1.1 Structure The establishment of functional levels and assignment of responsibilities is the prerogative of the organization authorized to operate the reactor facility.

The NTR facility organization and interrelationships are shown in Figure 12-1. This figure shows the relationship between the operating organization and the primary supporting organizations. The organization may be modified from time to time to reflect changes in programs and objectives.

The NTR facility is organized so that decisions are communicated via the proper levels and with adequate technical advice. Functions performed by one level may be performed by personnel at a higher level, provided they meet the minimum qualifications (i.e., Reactor Operators license, etc.). The Manager, VNC has overall responsibility for the reactor license. Reporting directly to The Manager, VNC, is the Manager, NTR, who is the Area Manager. The Manager, NTR is responsible for the safe and efficient operation, maintenance, and repair of the facility.

Operation of the reactor may be performed under the direction of a reactor Supervisor.

Contributing in a major way to the operating organization, but not reporting to the Area Manager, is the Regulatory Compliance organization (RC). Within this organization are specialists in nuclear safety, health physics, licensing, safeguards, security, and criticality. This organization also contains or is supported by health physics monitoring personnel and quality control personnel.

Also available to the Area Manager are many other highly specialized technical individuals on and off the VNC in the GEH organization.

12.1.2 Responsibility The responsibilities of selected NTR facility positions are as follows:

Manager, VNC Operations The Manager, VNC, has the overall responsibility for the NTR facilitys license.

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Manager, Nuclear Test Reactor (NTR)

The Manager, NTR, reports to The Manager, VNC. The Manager, NTR is the Area Manager and has the overall responsibility for the safe, reliable, and efficient operation of the NTR.

Figure 12-1 NTR Organization The Manager, NTR is responsible for maintaining a competent staff and an effective organization structure. All changes to the facility or facility procedures and all new tests and experiments require the approval of the Manager, NTR, or designated alternate. The Manager, NTR is responsible for an adequate safety review and may utilize resources of other GEH personnel (or outside consultants) not on the NTR staff. When the positions exist, the Manager, NTR, directs the activities of the reactor supervisor, and the NTR engineer(s). The Manager, NTR is also responsible for the development, maintenance, and implementation of written 12-2

operating and maintenance procedures, the coordination of operation maintenance, repairs, modifications, the training and requalification of operating personnel, and the safety of assigned and visitor personnel.

Reactor Supervisor The Reactor Supervisor (shift supervisors, Operations Supervisor, etc.) reports to the Manager, NTR and supervises the NTR operation in accordance with written operating procedures, administers planned work, and handles emergencies at the facility. The Reactor Supervisor is responsible for the safety of personnel working at the facility, ensures the security of fissionable materials, regulates entry into radiation and restricted areas, and trains new personnel. In the absence of a separate Reactor Supervisor, the Manager, NTR, has these responsibilities.

NTR Engineer (engineering and maintenance personnel)

The NTR Engineer (specialist, etc.) may report to the Manager, NTR and usually provides detailed direction and guidance in the installation and operation of experiment facilities and programs and overall plant maintenance, repairs, and modifications within the framework of operating procedures. In the absence of a separate Engineer, the Manager, NTR, has these responsibilities.

NTR Operators The NTR operators consist of trainees, licensed Reactor Operators, and licensed Senior Reactor Operators, who operate the reactor and experiment facilities in accordance with the operating procedures under the supervision of the Manager, NTR, or the Operations Supervisor. Licensed Reactor Operators may direct the activities of trainees, and licensed Senior Reactor Operators may direct the activities of licensed Reactor Operators and trainees, in accordance with the operating procedures.

12.1.3 Minimum Staffing Reactor Not Secured A licensed Reactor Operator or Senior Reactor Operator shall be in the control room with access to the nuclear console.

A second person present at the site who is familiar with the VNC Radiological Emergency Plan and Emergency Procedures relevant to the NTR and is capable of carrying out facility written procedures.

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A licensed SRO shall be present or readily available on call; that is, capable of arriving at the NTR within a reasonable time (1/2 hour / 30-mile radius) under normal conditions.).

Reactor Operating A Licensed Reactor Operator or Senior Reactor Operator shall be present at the controls at all times during the operation of the facility.

During Evolutions that Affect Reactivity Apparatus and mechanisms other than controls, the operation of which may affect the reactivity or power level of the reactor, shall be manipulated only with the knowledge and consent of a Licensed Reactor Operator or Senior Reactor Operator present at the controls.

A Senior Reactor Operator shall be present at the NTR facility during:

the first daily reactor startup and approach to power.

Recovery from an unscheduled shutdown.

Reactor fuel, safety rod, or control rod relocations within the reactor core region.

Manual poison sheet changes.

any experiment or facility changes with a reactivity worth greater than one dollar.

12.1.4 Selection and Training of Personnel Operations personnel shall have that combination of academic training, experience, health, and skills commensurate with their level of responsibility which provides reasonable assurance that decisions and actions during normal and abnormal conditions will be such that the plant is operated safely and efficiently in accordance with NRC license requirements and rules and regulations. Minimum qualifications shall include the following:

Manager, NTR At the time of appointment to the active position, the Manager, NTR, (Area Manager) shall have at least 6 years of nuclear experience. Additionally, the Manager, NTR shall have a baccalaureate or higher degree in an engineering or scientific field. Equivalent education or experience may be substituted for a degree. The degree may fulfill up to 4 of the 6 years of 12-4

nuclear experience required on a one-for-one basis. The Manager, NTR shall have or immediately pursue steps for obtaining an NTR Senior Reactor Operator license.

Reactor Supervisor At the time of appointment to the active position, the Reactor Supervisor, when utilized, shall have at least 3 years of nuclear-related experience. A maximum of 2 years equivalent full-time academic training may be substituted for 2 of the 3 years of nuclear-related experience required.

The individual should have a high-school diploma or have successfully completed a General Education Development (GED) test. The individual shall have an NTR Senior Reactor Operator license or shall immediately pursue steps for obtaining an NTR Senior Reactor Operator license.

NTR Engineer At the time of appointment to the active position, the NTR Engineer shall have at least 1 year of nuclear-related experience. Additionally, the NTR Engineer shall have a baccalaureate or higher degree. Equivalent education and experience may be substituted for a degree.

NTR Operators Senior Reactor Operator and Reactor Operator candidates are required to obtain licenses issued by the U.S. Nuclear Regulatory Commission in accordance with the provisions of 10 CFR 55.

Minimum qualifications as determined by the Manager, NTR are:

a. Senior Reactor Operator An NTR Senior Operators license.

A sufficient level of experience in NTR reactor operations, experiment setup and operation, and a high level of leadership.

b. Reactor Operator A high school diploma or equivalent, with a high degree of mechanical dexterity.

NTR Operators license.

Sufficient training or experience in related nuclear fields.

c. Trainee A high school diploma or equivalent.

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Initial Operator Training

a. Initial training for the Manager, NTR, and the reactor operators shall be sufficient for the individuals to obtain a Reactor Operator or Senior Reactor Operator license issued by the NRC. Topics shall include the following:

Fundamentals of reactor theory and operation, Facility design and operating characteristics, Instrumentation and control, Procedures and Technical Specifications, Radioactive material handling and exposure control, Code of Federal Regulations, Emergency response, Security.

Operator Requalification Program Licensed operators participate in a comprehensive Operator Requalification Program.

The program is designed to maintain the competence of the NTR operating personnel to handle abnormal events and to comply with the requirements and intent of 10 CFR 55.59. The NTR Requalification Plan approved by the NRC, is described in and is administratively controlled by the NTR procedures.

12.1.5 Radiation Safety Radiation protection is discussed in Chapter 11. Radiation protection functions at NTR are performed by NTR operators and radiation protection staff reporting to the Manager, Regulatory Compliance (RC). RC involvement ensures the radiation safety function is performed independently from the operations organization. All individuals performing a radiation safety function have stop work authority and the authority to raise a concern.

The Manager, RC, raises concerns to The Manager, VNC, the Manager, NTR, or by using the corrective action program.

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12.2 REVIEW AND AUDIT ACTIVITIES An effective independent review and audit process at NTR assures the following:

Operations comply with the facility license, the Code of Federal Regulations, and established procedures; The operating organization discharges its responsibilities consistent with good safety practices; and The records accurately and adequately reflect actual operation.

12.2.1 Composition and Qualifications Review and audit of the NTR is routinely conducted by members of the RC organization which provides a diverse resource of expertise in engineering and health physics. VNCs optional Vallecitos Technological Safety Council (VTSC) may be convened to provide an independent reviewing body. The VTSC is maintained via charter and is composed of personnel from a broad range of activities and technical disciplines. Membership is by appointment from The Manager, VNC. The VTSC is responsible to The Manager, VNC, and is independent of both the Regulatory Compliance and NTR operations organizations. Both, the RC and the VTSC are supported by offsite GEH organizations.

12.2.2 Charter The optional VTSC is a review body, independent of all operating organizations, that provides expert advice and counsel, but it is not responsible for conducting routine inspections. VTSC functions are conducted under a written charter. The VTSC reports its deliberations and recommendations to The Manager, VNC. Responses to VTSC safety- or compliance-related recommendations should be in writing, addressed to the Chairman, VTSC. The VTSC maintains records of its safety- or compliance-related recommendations and follow-on actions.

The VTSC meets quarterly unless there is no business to conduct in which case it meets annually at a minimum. The council may meet as frequently as necessary at the discretion of The Manager, VNC. A quorum is 50% or more of its members, but VTSC recommendations are based on a majority vote of members present. Minutes of each meeting are reviewed, approved, and disseminated to stakeholders.

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12.2.3 Review and Audit Functions The RC organization assists operating groups in developing methods of implementing the regulations, licenses, permits and solutions to safety issues and performs independent reviews and audits. Reviews are performed as documents are submitted. Audits are performed quarterly.

Comments and recommendations are made to the Area Manager. Disputes are resolved by The Manager, VNC.

The Regulatory Compliance organization is responsible for reviewing the following:

Determinations that proposed changes in equipment, systems, tests, EXPERIMENTs, or procedures are allowed without prior NRC approval as determined by 50.59 evaluation.

Determinations that new EXPERIMENTs or classes of EXPERIMENTs that could affect reactivity or result in the release of radioactivity do not require prior NRC approval as determined by 50.59 evaluation.

Determinations that proposed changes to the Fire Protection program as described in the Safety Analysis Report that do not require prior NRC approval, would not adversely affect the ability to achieve and maintain safe REACTOR SHUTDOWN of the NTR in the event of a fire as determined by 50.59 evaluation.

All new procedures and major revisions of existing procedures having safety significance that are required by the specifications in Section 6.4.

Proposed changes to the Technical Specifications or the FACILITY operating LICENSE.

Violations of Technical Specifications, and FACILITY LICENSE requirements.

Unusual or abnormal occurrences which are reportable to the NRC under provisions of the Federal Regulations or the Specifications in Section 6.7.2.

Periodic audit of facility operation, maintenance, and administration, to include:

o The conformance of facility operation to federal regulations, Technical, Specifications, and facility license requirements.

o The results of all actions taken to correct deficiencies or increase effectiveness in facility equipment, structures, systems, or methods of operation that affect nuclear safety.

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o The facility emergency procedures, security plan, requalification programs, and their implementing procedures.

The optional VTSC evaluates the overall effectiveness and relevance of safety studies and RC review activities.

The VTSC has the authority to review:

reportable incident investigations evaluations pursuant to 10 CFR 50.59 that conclude prior NRC approval is not required to make a change; operating standards, experiments, and receipt, possession, separation, use, processing, and transfer of radioactive material; and processes, operations and procedures which involve toxic, flammable, etc., materials.

Additionally, the VTSC has the authority to:

Consider and provide advice, as requested by The Manager, VNC, on problems of nuclear safety, criticality control, and industrial safety.

Review problems of nuclear safety, criticality control, and industrial safety.

Review the overall effectiveness and relevance of safety studies and RC review activities as they collectively influence the safety conditions.

Review and make recommendations on special topics as requested by the nuclear or industrial safety function or operations management.

Review any other matter which it conceives to be of safety importance.

12.3 PROCEDURES 12.3.1 Procedure Process The facility license, Technical Specifications, and Code of Federal Regulations establish the bounds within which the reactor must be operated. The VNC maintains a series of procedures for the NTR that are exclusive from those used for the broader operation of the site.

Site-specific procedures are reviewed by the applicable Area Manager and accepted by the Manager, Regularity Compliance or designated alternate.

Procedures exclusive to the implementation of administrative and operational requirements of the NTR Licensing basis and their revisions are approved by the Manager, NTR or his designated alternate.

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NTR and site-specific procedures address:

Startup, operation, and shutdown of the reactor Defueling, refueling, and fuel transfer operations when required Preventive or corrective maintenance which could have an effect on the safety of the reactor, including the replacement of components Surveillance checks, tests, calibrations, and inspections required by the Technical Specifications NTR-specific radiation protection program implementing procedures for personnel safety consistent with applicable regulations or guidelines. Management commitment and programs to maintain exposures and releases as low as reasonably achievable shall be a component of the site-wide radiation protection program Administrative controls for operation and maintenance and the conduct of EXPERIMENTs that could affect reactor safety or core reactivity NTR-specific implementing procedures for the site-wide emergency and security plans NTR-specific radiation protection program implementing procedures for the use, receipt, and on-site transfer of by-product material for such activities performed under the R-33 license 12.3.2 Procedure Change Process Because the NTR is subject to both site and corporate procedures, the procedure process that governs activities at the NTR bears some resemblance to that of a power reactor FACILITY. An effective procedure process includes approval and oversight of functional managers who provide expertise and oversight for processes governing programs such as those for corrective actions, procedure administration, radiation protection, emergency planning, and security. Because such programs and their implementing procedures are not limited in scope to the NTR, the Level 2 managers approval would be inappropriate for many changes. Therefore, approval of the Level 2 manager is not required for those procedures that fall under the authority of another function 12-10

manager. The Level 2 manager, however, is responsible to authorize any and all NTR-specific procedures that implement elements of these site-wide programs.

Major revisions to the NTR-specific procedures undergo independent review by the RC function.

Administrative changes, made in accordance with the sites procedure change process do not require independent review.

Temporary changes to an NTR operating procedure that are to be effective for 6 months or less are made according to the VNC / GEH temporary instruction change process. If a temporary change is to remain in effect longer than six months, a procedure revision shall be performed. A procedure revision is performed for a temporary change that is to remain in effect longer than six months.

An SRO may authorize temporary deviations from NTR-specific procedures during emergencies to prevent injury to personnel or damage to the facility. An SRO shall document the required emergency action in the logbook and notify the Manager, NTR by the end of the next working day.

Up-to-date copies of the NTR procedures are maintained and available to all personnel at the facility.

12.3.3 Engineering Release (ER)

An Engineering Release (ER) is issued, as required, to request work, distribute information, document actions, and otherwise ensure the safe and efficient operation of the NTR. An ER cannot contradict an effective procedure (A temporary instruction may be necessary.).

The ERs are written by NTR personnel and reviewed and approved in accordance with the procedure covering ERs. Independent review in accordance with Section 12.2 is required for ERs affecting those activities listed in Section 12.3.1.

12.3.4 Change Authorization (CA)

A Change Authorization (CA) is required for changes to the facility and changes to this document. The CA provides the documented description and a record of safety evaluations required by 10 CFR 50.59, and the review and approval of the change. A CA is required for changes, activities, or projects that are judged to involve significant safety considerations or a 12-11

potential Technical Specification violation or that may require prior NRC approval. A Change Authorization is also required for new types of experiments or changes to types of experiments.

Change Authorizations involving experiments (experiment type approval, as discussed in Chapter 10) require the following as a minimum:

a) All new types of experiments which could be postulated to affect reactivity or to result in unusual radiation exposure to personnel or an unusual release of radioactive materials, shall be reviewed for compliance with the facility license and the Technical Specifications.

b) Changes to approved experiments shall receive appropriate review and approval.

c) Approved experiments are implemented in accordance with written procedures.

Change Authorizations are administratively controlled by a procedure. The Change Authorization is reviewed independently by Regulatory Compliance to determine that the following criteria are satisfied:

The proposed change can be made without prior NRC approval (10 CFR 50.59).

The change does not violate any license requirement or federal regulations.

Special interim conditions which may exist during the period while the change is being made are analyzed to ensure that hazardous or unauthorized conditions do not exist during the modification or transition period.

More specific criteria and other review responsibilities are delineated in the Change Authorization procedure.

RC provides an independent review and concurrence of all Change Authorizations. RC may request evidence that the specified criteria are satisfied.

The Manager, NTR, or his designated alternate has the responsibility of approving Change Authorizations.

12.4 REQUIRED ACTIONS In the event of an abnormal occurrence, action shall be taken to assure the safety of the plant and personnel and to take appropriate corrective measures. If required, the reactor shall be shut down. If the reactor is shut down because of an abnormal occurrence, the reactor operation shall not be resumed until the cause is determined and required corrective action is completed.

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Should the true value of the reactor thermal power exceed 190 kW (as safety limit violation), the reactor shall be shut down and secured immediately and notification made to the Manager, NTR; Manager, Regulatory Compliance; and the Manager, VNC. The NRC is notified according to the Technical Specifications Section 6.6.1. The reactor shall remain shut down and secured until reactor operation is authorized by The Manager, VNC, and by the NRC.

If operation, maintenance, testing, or inspection reveals an unusual or unexpected result or a situation which is potentially reportable, the individual noting the occurrence shall notify the Manager, NTR immediately. If the reactor is in operation, the condition or situation shall be immediately returned to normal, or the reactor shut down. The Manager, NTR shall notify RC personnel. If the event is determined to be reportable, the NRC shall be notified according to the Technical Specifications Section 6.6.2.

12.5 REPORTS Reports shall be submitted to the NRC as required by the applicable portions of Title 10 CFR, Parts 20, 40, 70, 71 and 73.

12.5.1 Special Reports Special Reports of unplanned events at the NTR as well as planned major facility and administrative changes are addressed in Chapter 14 (Technical Specifications).

12.5.2 Annual Operating Reports Contents of the annual operating report are submitted to the NRC each year and include the following:

a. A narrative summary of reactor operating experience including the hours the reactor was critical and total energy produced.

b. The unscheduled shutdowns including, where applicable, corrective action taken to preclude recurrence.

c. Tabulation of major preventive and corrective maintenance operations having safety significance.

d. A summary of safety evaluations leading to the conclusions that they are allowed without prior authorization by the NRC according to Regulatory Guide 2.8.

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e. A summary of the nature and amounts of radioactive effluents released or discharged to environs beyond the effective control of the owner-operator as determined at or before the point of such release or discharge.

f. Summarized results of environmental surveys performed outside the facility.

g. A summary of exposures received by facility personnel and visitors where such exposures are greater than 25% of that allowed or recommended.

12.6 RECORDS NTR records are maintained, and retention periods specified according to Chapter 14 (Technical Specifications). Sitewide at the VNC, records are maintained in accordance with the applicable Federal Regulations such as 10 CFR 20.2103, 10 CFR 30.51, 10 CFR 40.61, 10 CFR 50.71, 10 CFR 55, 10 CFR 70.51, 10 CFR 73.70 and 10 CFR 74.19.

Records, in the form of logs, data sheets, recorder charts, and computer disks are maintained in file cabinets, binders, archive boxes, or electronic format.

12.7 EMERGENCY PLANNING The Radiological Emergency Plan is informed by ANSI/ANS-15.16, Emergency Planning for Research Reactors, and provides the framework of the sites response to emergency situations: to identify, communicate, respond to, and minimize the consequences of an emergency.

The objective of the Emergency Plan is to provide a basis for action, to identify personnel and material resources, and to designate areas of responsibility for coping with any emergency at the VNC that could impact public health and safety. This plan identifies both on-site and off-site support organizations that are required to be contacted for specialized assistance depending upon the nature of the emergency. It is the basis for detailed implementing procedures, which provide staff with the flexibility to cope with a wide range of emergency situations without requiring frequent revisions to the plan.

Provisions for reviewing, modifying, and approving the emergency implementation procedures are defined in the plan to assure that adequate measures to protect the staff and general public are available.

The emergency plan is reviewed biennially and revised as necessary.

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12.8 SECURITY PLANNING Since its inception, VNC has operated under a controlled-access security plan. The perimeter of the Site is posted No Trespassing. The VNC Site Developed Area includes an access road from SR 84 with a vehicle access gate and personnel turnstile at the access point to the permanent controlled access area. This controlled access area is the portion the VNC Site Developed Area that is within the vehicle access gate.

Several security measures are in effect at NTR to prevent unauthorized access to the facility and theft of materials.

Written procedures are available for incidents such

. The plan requires periodic testing of systems, recordkeeping, and reports to the NRC. Authorized individuals may refer to the current Security Plan for more details.

12.9 QUALITY ASSURANCE Design and construction of new and modification of existing structures, systems and components (SSCs) that are important to safety are subject to a comprehensive quality assurance program.

The objective of the program is to maintain an assurance of quality of the scram systems (Table 7-1) and safety-related items (Table 7-2) of the NTR.

12.9.1 Organization and Responsibilities This section describes the organizational structure and functional responsibilities for the quality program (Figure 12-1).

NTR Operations NTR Operations is responsible for operation of reactor and experiment systems in accordance with established procedures. NTR operations is also responsible for those items in Section 12.9.2 as required.

Engineering The NTR engineer (or the individual performing this function) is responsible for:

a) The performance of engineering on the NTR scram systems and safety-related items, b) The generation of designs and design changes, c) The preparation of specifications, work instructions, and procedures, 12-15

d) Participation in design reviews, e) Specifying which items in the scram systems and which safety-related items require quality assurance and the level of quality assurance required, f) Ensuring adequate proof of component and systems operability, g) Evaluating system and structural performance and effecting solutions, as appropriate, where operation is found to be inadequate, and h) Performing related engineering functions as required.

Regulatory Compliance The roles of Regulatory Compliance and the optional VTSC are discussed in Sections 12.1 through 12.3 of this Chapter.

Purchasing Purchasing performs activities related to procurement of materials and services required from outside vendors in accordance with procedures. Parts that are purchased for use in safety systems or safety-related items, from a vendor or original manufacturer having a 10 CFR 50, Appendix B Quality Assurance program, will have directions included in the Change Authorization for notifying the seller in the event the part fails or is found to be deficient according to purchasing documents.

Transportation and Materials Distribution The VNC Shipping and Receiving organization is responsible for receiving, shipping, and on-site movement of materials. The routine handling of materials is in accordance with procedures.

Instrument and Electrical Maintenance The Instrument and Electrical Maintenance components perform installation, calibration, repair, and maintenance services on electrical and instrumentation systems for the NTR. Records of work performed and calibration standards traceability, as required, are maintained.

Fabrication Fabrication may be performed on- or off- site as requested by the responsible engineer.

Mechanical Maintenance Facilities Maintenance components perform installation, repair, and maintenance services on mechanical systems not performed by the NTR staff. Records of work performed and calibration standards traceability, as required, are maintained.

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12.9.2 Instructions, Procedures, and Specifications Organizations responsible for work and/or performing work within the scope of this program are responsible for establishment and maintenance of documented systems and procedures for the performance of that work, unless provided for by NTR Operations or determined by NTR operations to be not required. Any changes of these documents are approved by the same function that authorized their issuance and use, unless otherwise specified within the document, or by governing Standard Operating Procedures.

Planning and/or implementing documents shall:

Provide, when warranted, space for sign-off by the person who performs the work to show that he has followed the prescribed instructions.

Call out essential controls and hold points, as required, which provide an independent assessment that the work was performed as prescribed and that the results meet specifications.

Include, as necessary, special instructions for handling and transportation.

12.9.3 Design Control Design Standards The responsible engineer identifies in the design drawings and specifications, required codes and standards and practices that provide the basis for design methods, material evaluation and process controls.

Design Verification Design verification is required for new systems or significant changes to existing systems for NTR safety-related items. This is accomplished by independent reviews (normally, RC review is adequate), alternate calculations, or the execution of a test program. The verification is performed by individuals other than those who performed the original design. The normal method for documentation is the Change Authorization, which is discussed in Section 12.3.4.

Engineering Change Changes to engineering definition documents are implemented and recorded by means of the Engineering Change Notice (ECN). Field changes during installation, as determined by the responsible engineer, may be implemented by redlining the drawing or specification, provided 12-17

the change is documented on an ECN and the change is evaluated by the same functions that approved the original prior to the operation of the component or system.

12.9.4 Procurement Control Procurement Flow Materials are ordered in accordance with the requirements of the engineering definition document, if applicable. Purchasing from outside vendors is performed by Purchasing in accordance with Purchasing procedures. Requests for Quotation (RFQs) and material to be purchased from outside vendors are documented on a Material Request form (MR). RC reviews MRs prior to submittal to Purchasing as required both for procurement and RFQs. Receiving inspection instructions, if required, are included on the requests. Receiving inspections are performed by the NTR personnel according to procedure to verify by objective evidence such features as proper configuration, identification, and cleanliness, and to determine any shipping damage, fraud, or counterfeit.

Vendor Selection and Surveillance Purchasing is responsible for soliciting quotes, negotiation of contracts, and procurement.

Vendor evaluation from a technical standpoint is performed by the responsible engineer. Vendor quality capability evaluation, if required, is performed by RC. The quality of purchased materials is verified by supplier-furnished evidence, source inspection, receiving inspection, or a combination of these, as appropriate.

12.9.5 Document Control Organizations performing work within the scope of this program generate documents such as procedures, drawings, specifications, and work instructions. Procedures are established describing the document control system. The document control system assures the proper review, approval, distribution, and control of documents and their revisions.

12.9.6 Material Control Procedures are established, as required, to control the identification, handling, storage, shipping, cleaning and preservation of safety-related items. The system provides measures to ensure the use of correct materials, to maintain traceability of components, and to clearly identify discrepant materials.

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Storage areas are provided, if necessary, to shelter material from natural elements, and to protect material in special environments. Materials held in storage are properly identified, adequately protected to preclude damage, and segregated to prevent the use of incorrect or defective parts.

12.9.7 Process Controls When required by engineering specifications or planning documents, production processes are accomplished under controlled conditions in accordance with applicable codes, standards, specifications, or other engineering criteria using appropriately qualified personnel and procedures.

Process Qualification Qualification of a production process is achieved by performing the process under controlled conditions on samples and then analyzing the output to determine acceptability. When the process can be duplicated on a repetitive basis by holding essential variables constant, and meet the requirements, the process is considered qualified. Qualifications are performed to written instructions based upon engineering specifications and include essential variables.

Personnel Qualification All personnel performing work activities have capabilities commensurate with their assigned functions, a thorough understanding of the operation they perform, the necessary training or experience, and adequate information concerning application of pertinent quality provisions to their respective functions. Supervisors responsible for directing work activities are responsible for assuring that personnel under their direction meet these qualification requirements.

12.9.8 Inspection Inspection Planning Inspections are performed to documented and approved plans for each work operation where it is necessary to measure quality. Inspection plans, as required, are incorporated into the detailed work instructions of the performing components.

Inspection Requirements Inspections are performed, as required, to written instructions and the inspection results are documented. NTR staff inspects raw materials, fabricated parts, assembly, and installation to the specifications provided. For purchased material, the receiving member of NTR staff identifies and matches quantities received with the purchase order and performs the receipt inspection.

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Hold Points - Approvals Hold points are stages in the planned activity beyond which work cannot proceed until the preceding work has been evaluated and approved. Hold points are determined by specific job requirements. Hold points and approval requirements for each organization are specified, as required, in the appropriate work instruction or procedure.

12.9.9 Test Control The responsible engineer identifies the need for development testing and/or for establishing test criteria for items not proven in design standard, mathematical analyses, or in state-of-the-art practices. Tests are aimed toward evaluation of performance capability under various conditions required by the design. Tests are conducted in accordance with written procedures; the test results are documented and evaluated to assure that the test requirements have been satisfied.

12.9.10 Control of Measuring and Test Equipment Each component which performs work is responsible for the inventory, identification, and calibration of gages and instruments used for measuring quality parameters as required or as specified by the requesting engineer. Inspection gages and instruments are calibrated, as required, with traceability to certified standards. If no certified national standards exist, the basis for calibration is documented.

12.9.11 Nonconformances Nonconforming Material Procedures Procedures will be provided, as required for the control of materials or parts as specified by the responsible engineer, which do not conform to requirements, in order to ensure their proper disposition. Nonconforming conditions shall be evaluated for further reporting to appropriate regulatory agencies.

Disposition of Scrap Materials Disposition of nonconforming materials shall be accomplished after a review by responsible personnel or groups and will consist of acceptance, repair, rework, or rejection. Technical justification for acceptability of a nonconforming item dispositioned repair or use as-is shall be documented in a non-conformance report and is subject to design control measures commensurate with those applied to the original design.

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12.9.12 Corrective Action Documentation of agreed-upon corrective action for conditions adverse to quality are governed by established procedures. In the case of a condition adverse to quality or a significant condition adverse to quality, the cause of the condition shall be investigated, and corrective action taken according to the GEH corrective action program.

12.9.13 Experimental Equipment This program provides, as applicable, controls over the fabrication and installation of experimental equipment to the extent that these relate to reactor safety.

12.9.14 Records Records are retained in accordance with the requirements of Section 12.6.

12.9.15 Audits RC conducts audits in accordance with established procedures to verify compliance with the various elements of this Quality Assurance program. Audits are conducted on a scheduled or random unscheduled basis, or both.

Procedures include a program for the performance of radiation safety audits. Audits/reviews to verify conformance of an item or activity to requirements, shall be planned, documented, and performed. Audits/reviews should include radiation safety, procurement, construction, modification, maintenance, and experiment fabrication, and should ensure effective implementation of the program is assessed, deficiencies identified, and corrective actions have been taken. Audits are scheduled to cover audited programs and activities over a two-year period. Reviews of several NTR Technical Specifications are performed four times per year.

The reviews are conducted by RC staff who are independent of the NTR organization and who have experience or training commensurate with the scope, complexity, or special nature of the activities to be reviewed. The written reports resulting from the reviews are sent to the Area Manager and/or Manager, RC, as appropriate. Follow-up action is performed by the reviewer and tracked to completion, to assure corrective action is accomplished.

12.10 OPERATOR TRAINING AND REQUALIFICATION All licensed operators participate in a comprehensive Operator Requalification Program. The program is designed to maintain the competence of the NTR operating personnel to handle 12-21

abnormal events and to comply with the requirements and intent of 10 CFR 55.59. Refer to the Requalification Program for the General Electric Nuclear Test Reactor for details.

12.11 ENVIRONMENTAL REPORTS Operation of the NTR has had minimal effect on the environment. Refer to Vallecitos Nuclear Center Environmental Report 2020, NEDO-12623.

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12.12 REFERENCES

1) Environmental Information Report for the General Electric Test Reactor, General Electric Company, Vallecitos Nuclear Center, Pleasanton, California, July 1976 (NEDO-12623).

a. Water-Quality Monitoring Network for Vallecitos Valley, Alameda County, California, U.S. Geological Survey, Water Resources Investigations 80-59.

2) Tier 1 Seismic Evaluation of the Vallecitos Boiling Water Reactor Shutdown Facility, prepared for GE-Hitachi Nuclear Energy, Sunol, CA, by Structural Integrity, July 2019.

3) Final Report, Seismic Risk Analysis for General Electric Plutonium Facility, Pleasanton, California, Part II, prepared by Tera Corporation, June 1980.

4) State of California, Department of Transportation and the Alameda County Transportation Commission, SR 84 Expressway Widening and SR 84/I-680 Interchange Improvements Project Final Environmental Impact Report/Environmental Assessment with Finding of No Significant Impact, April 2018.

5) W. H. McAdams, Heat Transmission, 3rd ed., McGraw-Hill, New York, 1954, pp. 362 and 393.

6) L. M. Petrie, and N. F. Cross, KENO-IV, An Improved Monte Carlo Criticality Program, November 1975 (ORNL-4938).

7) R. D. Carter, et al., Criticality Handbook, Volumes 1-3, Atlantic Richfield Hanford, September 25, 1975 (ARH-600).

8) H. K. Clark, Critical and Safe Masses and Dimensions of Lattices of U and UO2 Rods in Water, Savannah River Laboratory, February 1966 (DP-1014).

9) a.) G. E. Hansen and W. H. Roach, Six and Sixteen Group Cross Sections for Fast and Intermediate Critical Assemblies, November 1961 (LAMS-2543).

b.) L. D. Connolly, Los Alamos Group-Average Cross-Sections, July 1963 (LAMS-2941).

10) C.E. Newlon and A. J. Mallett, Hydrogen Moderation-A Primary Nuclear Safety Control for Handling and Transporting Low-Enrichment UF 6, Appendix 1, 12-23

Normalization of the Hansen-Roach Cross Section, by J. R. Knight, K-1663, ORNL, May 31, 1966.

11) H. F. Henry, J. R. Knight and C. E. Newlon, General Application of a Theory of Neutron Interaction, K-1309, ORNL, 1956.

12) D. R. Oden, et al., Critique of the Solid Angle Method, Battelle Pacific Northwest Laboratories, February 1978 (NUREG/CR-0005).

13) J. S. Tang, Investigation of the Solid Angle Method Applied to Reflected Cubic Arrays, October 1976 (ORNL/CSD/TM-13).

14) J. T. Thomas, Ed., Nuclear Safety Guide, Rev. 2, June 1978 (TID-7016).

15) J. K. Thompson, et al., Snake: Solid Angle Calculational System, Battelle Pacific Northwest Laboratories, February 1978 (NUREG/CR-0004).

16) H. F. Henry, J. R. Knight, and C. E. Newlon, Self-Consistent Criteria for Evaluation of Neutron Interaction, K-1317, ORNL, December 21, 1956.

17) E. P. Blizard, et al., Neutron Physics Division, Annual Progress Report for Period Ended 8/1/1964, Vol. 1, December 1964 (ORNL-3714).

18) J. W. Wachter, Ed., P-12 Plant Nuclear Safety Handbook (20th Edition), Oak Ridge National Laboratory, July 19, 1963 (Y-1272, TID-4500).

19) H. C. Paxton, Critical Dimensions of Systems Containing U-235, Pu-239, and U-233, June 1964 (TID-7028).

20) L. A. Bromley, Chem. Eng. Prog. 46, 221, 1950.

21) S. C. Skirvin, Users Manual for the THTD Computer Program, General Electric Co.,

San Jose, California, June 23, 1966.

22) Development of Technical Specifications for Experiments in Research Reactors, USNRC Regulatory Guide 2.2, November 1973

23) Recommendations of the International Commission on Radiological Protection, ICRP9.

24) J. G. Collier, Convective Boiling and Condensation, McGraw-Hill, London, 1972, pg.

253.

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25) R. V. Macbeth, Burnout Analysis, Part III, The Low Velocity Burnout Regime, Dorset, England, 1963 (AEEW-R-222).

26) P. T. Pon, et al., A Literature Survey of Critical Heat Flux Correlations for Low Pressure and Low Flow Conditions, General Electric Co., San Jose, California, May 1980

27) A. I. Yang, et al., CORLOOP Multi-Channel Core and Loop Model for the Nuclear Test Reactor, August 1980 (NEDE-24861).

28) R. Yahalom, GETR Multi-Channel Core Model for Simulating Internal Natural Circulation, General Electric Co., San Jose, California, June 1977 (NEDO-12663).

29) Regulatory Guide 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Revision 1, July 1977.

30) GEH Licensing Topical Report NEDE-32176P, Revision 4, TRACG Model Description, January 2008.

31) GEH Licensing Topical Report NEDE-32177P, Revision 3, TRACG Qualification, August 2007.

32) DOE-STD-1212-2019, DOE Technical Standard: Explosives Safety, November 2019.

33) NAVSEA OP 3565/NAVAIR 16-1-529, Electromagnetic Radiation Hazards -

Hazards to Ordinance, Vol. 2, Rev. 16, June 2007.

34) ANSI IEEE C95.1-2005, IEEE Standard for Safety Levels with Respect to Human Exposure to Radio Frequency Electromagnetic Fields, 3 kHz to 300 GHz, April 2006.

35) Regulatory Guide 2.2, Development of Technical Specifications for Experiments in Research Reactors, Revision 0, November 1973

36) C.J. Werner, Ed., MCNP Users Manual Code Version 6.2, Manual Rev. 0, October 2017 (LA-UR-17-29981).

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13 ACCIDENT ANALYISIS 13.1 ACCIDENTINITIATING EVENTS AND SCENARIOS This chapter contains an evaluation of the facility response to certain events that can be reasonably postulated to occur at the NTR, and which appear to have safety significance. The results of the analyses show that design features, equipment, and procedures are in place to ensure that the health and safety of the public and plant personnel are not jeopardized by the occurrence of any of the postulated events. The events analyzed include anticipated operational occurrences and potential accidents.

The maximum hypothetical accident (MHA) is an enveloping event that is postulated to involve a failure in a fueled experiment. Such event would lead to the maximum potential radiation hazard to the personnel. Also considered as an experiment design basis accident, the event is discussed in Section 13.3.

Reactor transients were analyzed by simulating reactor dynamics with a digital computer. The model used is discussed in Section 13.4.

Events categorized as anticipated operational occurrences are discussed in Section 13.4.

Anticipated operational occurrences are the results of single equipment failures, or malfunction, or single operator errors that can reasonably be expected during any planned mode of facility operation. The anticipated operational occurrences analyzed in this chapter are:

Loss of electric power Loss of secondary cooling Loss of facility air supply Inadvertent core inlet temperature change Fuel handling errors.

Unacceptable consequences for anticipated operational occurrences are:

Release of radioactive materials to the environs that result in exceeding the limits of 10 CFR 20.1301/20.1302 for members of the public; Radiation exposure of any person in excess of 10 CFR 20.1201 limits for occupational workers or 10 CFR 20.1301 limits for members of the public; and Violation of an established safety limit.

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Events categorized as accidents are discussed in Section 0. Accidents are defined as postulated events not expected to occur during the course of plant operation that appear to have the potential to affect one or more of the radioactive material burners. The postulated accidents analyzed in this chapter are:

Uncontrolled reactivity increases Loss of primary coolant flow (pump shaft seizure)

Rod withdrawal Loss of primary coolant.

Unacceptable consequences for postulated accidents are:

Radioactive material release to an extent that exceeds the guideline values of 10 CFR 20.1201 for occupational workers or 10 CFR 20.1301 for members of the public.

Violation of a safety limit.

Section 13.6 is an evaluation of experiment safety and shows that procedures, limits, and safety equipment are in place to ensure the proposed experiment program can be carried out without undue risk to the health and safety of the public and plant personnel.

There is a close relationship between the safety analyses for anticipated operational occurrences and accidents and the safety limits and limiting safety system settings. Development of proposed safety limits and limiting safety system settings are discussed in Section 13.7.

The results of the analyses show that there are no credible events that could cause fuel melt or a significant release of fission products from the fuel. Even if catastrophic non-mechanistic failure of the NTR facilities is assumed, there are no potential consequences more severe than those associated with the accidents analyzed in this section. Compaction of the fuel, while essentially impossible mechanistically, would not cause the reactor to go critical since water loss, increased self-shielding in the fuel, and the geometry change due to flattening of the cylindrical core are all negative reactivity effects. Loss of water shuts down the reactor and no fuel melting occurs, as discussed in Section 13.5.6. Also, deformation of the core, which causes the fuel to contact the core can structure, would improve heat-transfer and result in lower Loss-of-coolant Accident (LOCA) temperatures. The only accidents which could possibly cause fuel damage and release of fission products from the NTR fuel are those resulting from large reactivity insertions.

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Reactor configuration and the reactivity worth of experiments are controlled to ensure that destructive reactivity transients are not credible. Nevertheless, an assessment of the consequences of an assumed fission product release is presented in Section 13.2 to demonstrate the capability of the facility, even though such a release is not possible under the $0.76 reactivity limit.

13.2 EXPERIMENT DESIGN BASIS ACCIDENT 13.2.1 Introduction The experiment design basis accident is designated as the maximum hypothetical accident (MHA) for the NTR, resulting in the maximum potential radiation hazard to personnel and the public.

The material quantity limits, operating limits, and required safety equipment for irradiation experiments at the NTR has been developed based on the radiological criteria given in Regulatory Guide 2.2 (Reference 35). This analysis specifically addresses the limits for a singly clad U-235 powder fueled capsules and shows the capability of the facility and site to accommodate a radioactive material release with no credit taken for filtration of the release by the NTR stack filter system.

13.2.2 Accident Description Regulatory Guide 2.2, Part c.2.a (Material Content of Experiments) describes the release event as:

....a single mode nonviolent failure of the encapsulation boundary that releases all radioactive material into the immediate environment of the experiment or to the reactor building as appropriate....

and in addition, it states that:

The analysis should establish the most probable trajectory of the material, if any, into restricted and unrestricted areas. Credit for natural consequence-limiting features such as solubility, absorption, and dilution and for installed features such as filters may be taken provided each such feature is specifically identified and conservatively justified by specific test or physical data or well-established physical mechanisms.

13-3

Therefore, the design basis accident for an experiment in the NTR is described as follows:

1. Experiment material is 50 mg of U-235 powder in a singly encapsulated container.

a. Dose consequences for doubly cladded or pellet forms of U-235 equal to or less than 50 mg are bounded by the results of this analysis.

2. The most probable trajectory of the released material is from the experiment location to the reactor cell area. Since the event is a single-mode nonviolent failure, the established conditions would presumedly include the ventilation system being in operation; however, ventilation (filtration) is conservatively not credited in the analysis.

3. The release fractions of U-235 fuel and fission products to the environment are assigned as follows:

Release from capsule to reactor cell: Powder / Pellet (%)

U-235 100 Noble Gas 100 Iodine 100 All Remaining Fission Products 100 Release from reactor cell to the environment:

U-235 100 Noble Gas 100 Iodine 100 All Remaining Fission Products 100

4. Dose Limits, total effective dose equivalent (TEDE):

2-hour Fence-Post Dose to Member of Public 0.1 rem, TEDE Operator Dose, during 5-Minute Evacuation 0.5 rem, TEDE

5. The unrestricted area exposure will result from the diluted-dispersed cloud of isotopes released from the NTR stack, which reaches the nearest site boundary under type F meteorological conditions at 1 m/sec over a 2-hour period.

6. The restricted area (specifically, the reactor cell) exposure will result from the submersion in and inhalation of the isotopes released to the reactor cell for a period of 5 minutes during evacuation. The bases for this postulated exposure are as follows:

13-4

a. It is assumed that a complete release of the experiment capsule contents to the restricted area will occur uniformly over the two hours following the experiment failure.

b. The fission products from this release will cause high-activity alarms on the stack monitors.

c. The NTR operator will respond to the stack alarms and announce an area evacuation over the building public address system.

d. Evacuation to an upwind location will remove personnel from the stack concentration of released isotopes. On-site exposures can be controlled by use of the alarm system and evacuation procedures.

e. Assuming the operator is in the reactor cell where the accident occurs for the duration of the evacuation ensures the operator dose for this event bounds the operator doses that would be received at all other locations inside or adjacent to the NTR for a 5-minute evacuation.

13.2.3 Calculation Method ORIGEN2 Version 2.1 is used to calculate the fission products created during operation of the U-235 capsule, and the RADTRAD computer code (version 3.10) is used to calculate the TEDE dose resulting from exposure to the released radioactive materials. Decay and progeny have been accounted for in the RADTRAD dose consequence model. The inputs required for the evaluation of doses resulting from exposure to the released isotopes from a U-235 experiment in the NTR are:

1) Capsule operating power = 60 watts (for 50 mg U-235)

2) Capsule operating time = 1 day

3) Decay time after shutdown = 0 minutes (not credited)

4) ORIGEN2 cross-section library = BWRUE.lib (based on ENDF/B-V)

5) Meteorology type is Pasquill Type F with a wind speed of 1 m/sec (for the boundary dose)

6) Effective release height is at ground level 13-5

7) The breathing rate of the exposed subject is 350 cc/sec

8) The distance from the nearest boundary point to the NTR stack is 510 meters

9) Release to environment = 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months

10) Environmental release flow rate = 1000 cfm

11) Reactor cell airspace volume = 10,500 ft3

12) Dose conversion factors from Federal Guidance Reports No. 11 and 12.

13) Inventory of nuclides released in the reactor cell:

Table 13-1 NTR EXPERIMENT DBA ISOTOPIC RELEASE TO REACTOR CELL Nuclide Curies Nuclide Curies Nuclide Curies Nuclide Curies Kr85 1.78E05 Zr97 1.84E+00 Te131m 7.71E02 La140 3.11E02 Kr85m 6.20E01 Nb95 3.35E04 Te132 4.11E01 La141 2.84E+00 Kr87 1.29E+00 Mo99 6.71E01 I131 1.03E01 La142 2.86E+00 Kr88 1.82E+00 Tc99m 3.98E01 I132 3.69E01 Ce141 4.76E02 Rb86 1.50E06 Ru103 2.76E02 I133 1.82E+00 Ce143 1.16E+00 Sr89 3.23E02 Ru105 5.00E01 I134 3.83E+00 Ce144 6.63E03 Sr90 1.87E04 Ru106 3.93E04 I135 2.90E+00 Pr143 3.16E02 Sr91 2.43E+00 Rh105 1.48E01 Xe133 1.25E01 Nd147 6.86E02 Sr92 3.00E+00 Sb127 1.21E02 Xe135 7.67E01 Np239 1.74E12 Y90 4.99E05 Sb129 3.35E01 Cs134 2.05E08 Pu238 2.43E14 Y91 1.78E02 Te127 5.44E03 Cs136 1.84E04 Pu239 4.04E20 Y92 2.92E+00 Te127m 5.37E06 Cs137 1.94E04 Pu240 1.62E22 Y93 2.60E+00 Te129 2.98E01 Ba139 3.14E+00 Pu241 3.39E31 Zr95 3.43E02 Te129m 8.08E04 Ba140 1.65E01 U235 1.08E07 The source term for this evaluation is calculated by modeling the irradiation of a capsule containing 50 mg of U-235 at a capsule operating power level of 60 W for a period of 1 day simulated in an ORIGEN2 model. The model results include activities and masses for several hundred isotopes (>700 nuclides), however, many of these nuclides are not important contributors to dose consequences or are not present in sufficient quantity to impact the dose consequences of the NTR experiment DBA.

Thus, this analysis considers the 60 dose important isotopes used for offsite dose consequence evaluations for commercial nuclear power plants that use U-235 enriched fuel. This set of isotopes in commonly referred to as the Alternative Source Term (AST) nuclide set. There are 55 of the AST isotopes present in the ORIGEN2 NTR experiment DBA results. Along with 13-6

these 55 dose important isotopes, the remaining unburned U-235 is added to the release source term. The resulting source term comprised of 56 isotopes is shown in Table 13-1.

While these isotopes are known to be the main dose contributors for releases involving irradiated U-235, there may be a contribution from the remaining isotopes not being modeled. To account for this contribution a safety factor of 20% is added to the final dose consequences.

13.2.4 Results The results generated by the RADTRAD model shown in Table 13-2 confirm the material quantity and operating limits established for the NTR experiment DBA results in dose consequences at the unrestricted area boundary and the restricted area that are below the dose limits. Dose consequences for doubly cladded or pellet forms of U-235 equal to or less than 50 mg are bounded by the results of this analysis.

Table 13-2 NTR EXPERIMENT DESIGN BASIS ACCIDENT DOSES Dose Consequences, Dose Limits, Exposure Location and Duration TEDE TEDE Unrestricted Area Boundary with 66.2 mrem 100 mrem 2-hour Exposure (0.662 mSv) (1 mSv)

Restricted Area (NTR Reactor Cell) with 470.0 mrem 500 mrem 5-Minute Exposure (4.70 mSv) (5 mSv) 13.3 TRANSIENT MODEL The reactor dynamics were simulated with a computer model. TRACG computer code is used to simulate the thermal-hydraulics response of the reactor with kinetics feedback. The TRACG code (Reference 30) is the GEH proprietary version of the Transient Reactor Analysis Code (TRAC), a system code widely used for boiling water reactor safety analyses. The code capabilities include 3D and point kinetics models, a multi-dimensional, two-fluid model for the reactor thermal hydraulics, and an implicit integration scheme for numerical calculations. The code was reviewed and approved as part of the application methodologies for Anticipated Operational Occurrence (AOO), stability, Anticipated Transient Without Scram (ATWS), and Loss of Coolant Accident (LOCA) analyses. Its applicability range extend from full operating pressure of boiling water reactors to atmospheric pressures (Reference 33).

An important aspect of the analysis is the heat-transfer characteristic by which steam is formed during excursions (the steam voids provide the strongest negative reactivity feedback in addition 13-7

to scram). The general characteristics of the heat-transfer mechanisms have been described here and basic relations used in the heat-transfer analysis are given in (Reference 33).

Figure 13-1, illustrates the qualitative hypothetical high-power excursion without scram. For very fast transients which are not possible with the existing $0.76 reactivity limit, some of the sequences shown may not be the same; however, most mechanisms which appear are illustrated.

This hypothetical excursion develops according to the following sequence of events.

(1) Initially, all channels are in laminar flow with heat-transfer coefficients near 165 Btu/h-ft2-°F. Figure 13-1 shows three initial fuel temperatures for (a) average channel representing average power, flow, and temperature conditions, (b) a channel with less than average power (90%) but much less than average coolant flow (50%), and (c) a channel with highest power (130%) and highest flow (153% of the average). The highest power channel (c) usually produced the highest fuel temperature for transients in which the high-power peak was the dominant factor. As the transient progresses and power increases, fuel and water temperatures rise until the beginning of nucleate boiling.

(2) When the fuel surface temperature becomes high enough, nucleate boiling begins on the fuel surface. When this occurs at time, t1, heat-transfer conditions improve greatly, holding the surface temperature essentially constant, and increasing the rate of rise of channel water temperature. In TRACG, the values of fuel surface temperatures during nucleate boiling conditions are calculated with Chen correlation.

(3) In the transient shown in Figure 13-1, average channel exit water temperature reached saturation at time t3. The fuel temperature remained nearly constant throughout nucleate boiling. The formation of steam produced a large negative reactivity feedback, which turned the power excursion.

(4) In this example, the hot spot was steam blanketed. This phenomenon is triggered if the power rises high enough to produce a surface heat flux greater than the 450,000 Btu/h-ft2. When this occurs, the surface heat-transfer coefficient drops to about 10 Btu/h-ft2-

°F. The fuel temperature will rise sharply since this condition almost insulates the fuel.

As shown in Figure 13-1, this temperature will level off when the power is turned. The temperature will approach a new steady- state value which corresponds to the final 13-8

power level. This power level is dependent upon the type of accident and the extent of steam formation. The reactivity events typified by this general type of behavior are presented in Sections 13.5.1 through 13.5.4.

13.4 ANTICIPATED OPERATIONAL OCCURRENCES 13.4.1 Loss of Normal Electrical Power 13-9

Figure 13-1 NTR Transient 13-10

13.4.2 Loss of Facility Air Supply The only items affected by the loss of facility air are the air piston operator for the south cell door, and the radiation shield shutter for the horizontal facility in the south cell. The design specifications for the south cell door require that it be movable manually by one person. Loss of air to the beam shutter would have no effect on its position (i.e., it would remain in the position it was in at the time of air supply failure). Thus, there is no safety concern due to loss of facility air supply.

13.4.3 Loss of Secondary Coolant Secondary coolant flows by gravity through the tube side of the primary heat exchanger, as described in Section 15.3. Loss of secondary coolant or loss-of-coolant flow will cause primary reactor coolant temperature to gradually increase. This will cause power to increase until core temperature exceeds 124 °F at which time the negative coolant temperature coefficient will turn power over, and power will decrease. If reactor power level increases to a level that produces an appreciable heating rate, the reactor will scram from high primary coolant temperature. If reactor power is not high enough to produce a high-reactor power scram or result in a heating rate that could trigger a high primary coolant temperature scram, the loss of secondary coolant will soon be evident to the operator by:

The slightest changes in temperature, which cause an observable reactivity effect.

The temperature monitor system readout at the console.

The secondary flow control in the control room.

13.4.4 Inadvertent Core Inlet Temperature Change If the primary pump were inadvertently started, the effect would be to decrease the reactor inlet temperature. This and any other event that causes a decrease in inlet temperature while the average reactor coolant temperature is below 124 °F will cause reactor power to drop. There is no safety or radiological concern related to an inadvertent pump start.

Other events can be postulated that would increase reactor coolant temperature. An increase in inlet temperature will produce a rising power transient - a hot water transient. This transient is comparable to a cold-water accident for reactors that operate with a negative temperature coefficient of reactivity. There is currently no source of energy to produce an increase in primary coolant temperature. However, the system design includes a 5-kW heater that was 13-11

removed from the system years ago that could be reinstalled if needed. The amount of positive reactivity which could be added from this heater is less than $0.10 (from room temperature to turnover temperature); therefore, the resultant transient would, and could, be controlled by manipulation of the control rods.

The worst possible case would be a coolant heat up to 124°F from reduced temperature conditions. A case is performed with initiation of a 5-kW heater without scram. The temperature and power characteristics are shown in Figure 13-2. The power continues to rise until power increases sufficiently to raise the core coolant temperature above 124°F and reduce the net reactivity. Power and temperature stabilize at a higher power and temperature.

For such a slow transient, a high-power scram would clearly stop the excursion without fuel damage. For even higher heat additions, if the scram failed, bulk boiling would occur soon enough to prevent the power from reaching a level high enough to produce steam-blanketing. It has been shown that a step insertion of $0.76 of reactivity would not cause fuel damage, even if the reactor failed to scram. Therefore, it can be concluded that a transient caused by the small amount of reactivity from the temperature would also be safely limited.

220 100 200 95 180 90 Inlet Temperature (°F)

Power (kW) 160 85 140 80 120 75 Power Inlet Temperature 100 70 80 65 0 500 1000 1500 2000 2500 3000 Time (sec)

Figure 13-2 Initiation of 5 kW Heater from Reduced Coolant Temperature (65°F) 13-12

13.4.5 Fuel Handling Errors Fuel handling equipment and procedures are discussed in Section 9.2. The likelihood of such accidents is insignificant due to the design of the reactor, the design of the reactor fuel (discussed in 4.2.1), the historical infrequency of fuel handling (last done in 1976 to support core container replacement), and the unlikely sequence of events that would be necessary in order for the NTR to be refueled with LEU fuel. The 16 fuel assemblies on hand completely fill the core reel assembly and fill the fuel container to the extent that the only remaining space of appreciable size is in the fuel loading chute. In other words, a fuel assembly, once inside the fuel container, must either be in the provided positions in the reel or in the fuel loading chute. The physical arrangement of the fuel container is such that an element located in the loading chute results in a worse core geometry (one having a lower neutron multiplication factor) than the cylinder formed by having all assemblies in the core support reel. Dropping a fuel assembly could only cause an accident if the control rods were withdrawn during loading so that the reactor was almost critical before adding fuel. Such an act is contrary to operating procedures and requires errors by the console operator and fuel loaders. The only other means of getting fuel close to the core is by inserting it into either the horizontal or vertical facilities. Use of these facilities is discussed in Chapter 10.

In addition to the inherent safety feature provided by having all existing NTR fuel assemblies in their most reactive configuration in the core, the following additional safety features ensure safety during all phases of fuel handling:

Reactor design, fuel handling equipment, and administrative controls are such that not more than two elements can be handled at one time.

All fuel movement must be performed in accordance with written procedures.

The cell high-gamma-level alarm system will be in operation.

By using all the manually positioned poison sheets, the core can be made $6.1 subcritical (Table 4-1). Removal of the graphite plug from the fuel loading chute provides additional negative reactivity of approximately $1.25.

Any movement of source and special nuclear material within the NTR facility must have the approval of the licensed operator on duty.

Any storage arrangement used will be analyzed to ensure a subcritical configuration.

13-13

As a result, fuel damage during fuel handling would be unlikely, with no associated dose consequences.

13.4.6 External Events The only credible external event for the NTR is a seismic event, which is discussed as the initiating event for the bounding reactivity insertion accident in Section 13.5.1. Because fuel damage does not occur in this scenario, there is no release of radioactive products and no associated dose consequences.

13.4.7 Mishandling/Malfunction of Equipment Mishandling of equipment is precluded by the technical specifications and operating procedures.

Because bounding analysis in Section 13.4 addresses all events that could result from improper operation, fuel damage does not occur in this scenario, there is no release of radioactive products, and there are no associated dose consequences.

13.5 POSTULATED ACCIDENTS The transient model used to simulate the reactor dynamics is presented in Section 13.3, except that analysis in 13.4.1 and 13.4.2 do not use the TRACG code.

13.5.1 Idealized Step Reactivity Insertions - with Scram Transients resulting from step reactivity insertions up to $1.4 were studied; a range of different initial reactor powers and flows were used. The results for steps with high power scram occurring at 150 kW are shown in Figure 13-3. Only a very slight fuel temperature increase was observed for steps up to $1.0. In all cases, peak temperature rose sharply for reactivities above this value. Fuel melting and clad melting temperatures are within a few degrees and used interchangeably throughout Chapter 13 (e.g.; Figure 13-3,Figure 13-4, Figure 13-10, and Figure 13-14) due to the inherent imprecision of legacy data. The onset of any melting for the U-Alx alloy fuel meat is calculated to be a eutectic 1186°F and all legacy data is conservative relative to the eutectic temperature. This temperature limit is consistent with the time scales involved in the reactivity insertion accident transient, which are substantially shorter than the hours corresponding to the onset of blistering, for which the temperatures are 50-100C lower.

Therefore, the indicated Chapter 13 melt temperature is an appropriate metric to evaluate fuel melt during a reactivity insertion accident.

13-14

The transient due to a step reactivity insertion of $1.3 while the reactor is at 100 kW and at rated flow is shown in Figure 13-4 and the sequence of events for this transient is as follows.

Time Peak Fuel Temperature (sec) Event (°F) 0.0 $1.3 step insertion 195 0.0044 Scram circuit tripped 195 0.1460 Nucleate boiling began at hot spot 241 0.1558 Steam-blanketing occurred at hot spot 258 0.2046 Safety rods reached active core 504 0.218 (Power peak 1.04 x 105 kW) 652 1.0 Power dropping, temperature rising slowly 841 The transient is too fast for any channel bulk boiling to help the scram reduce power. The relatively high tail on the power curve is the result of delayed neutron groups which are controlling the rate of change of power. Even after an excursion has reached the steam-blanketed condition and the heat-transfer coefficient has dropped to 10 Btu/h-ft2-°F, a power level of 100 kW can be maintained without melting at the hot spot. The peak temperature characteristic is very sharp. A peak temperature of only 400°F resulted from a $1.2 step, compared to approximately 840°F for the $1.3 step. The results for lower initial power and flow show that fuel temperatures are lower for these other cases.

13-15

Figure 13-3 k-Steps with 150-kW Scram 13-16

Figure 13-4 $1.3 Step from 100-kW with Scram 13-17

13.5.2 Idealized Finite Ramp Reactivity Insertions - with Scram Large reactivity insertions over short periods of time were studied for finite reactivity ramps.

The study used a simplified point kinetics model with a lumped neutron precursor group with reactivity feedback. The feedback effects included, in addition to the scram, moderator temperature and voiding. Reactivity insertions of $2 and $4, with durations from 0.2 to 0.6 second, were analyzed. The results for initial powers of 100 kW with the overpower scram occurring at 150 kW are given in Figure 13-5. For the $4 insertion, fuel melting is not expected if the duration of the insertion is greater than 0.5 second. For the $2 case, the minimum acceptable insertion time was 0.24 second. Figure 13-6 and Figure 13-7 show near-limiting cases. In both cases, steam-blanketing and nucleate boiling occurred almost simultaneously so that fuel-surface, heat-transfer conditions were poor throughout the transients, and no bulk boiling was observed.

In each case, power dropped below the level at which the hot spot is cooled even with steam-blanketed conditions before peak fuel temperature reached melting. For transients starting from lower power levels, the temperatures will be slightly less than those shown in Figure 13-5 because of the lower initial temperature. The sharp characteristic, however, places the limiting reactivity insertion time at nearly the same value. The consequence of inserting these large amounts of reactivity too fast, or if the scram failed, would be partial core destruction. The primary shut-down mechanisms would be associated with the expansion and dispersion of the fuel.

13-18

Figure 13-5 100-kW Finite Ramp Insertion with High-Flux Scram 13-19

Figure 13-6 $4 Ramp in 0.6 Second from 100 kW with Scram 13-20

Figure 13-7 $2 Ramp in 0.3 Sec from 100 kW with Scram 13.5.3 Reactivity Insertions - without Scram There is no case by which insertion of excess reactivity can result in fuel damage to the NTR reactor when potential reactivity available from control rods and experiments is maintained at or below $0.76. To bound all possible insertion accidents as well as a worst-case external event accident, an extremely improbable event is hypothesized in which a beyond design basis seismic event results in the loss of all ability to insert negative reactivity into the core by presuming 13-21

control and safety rod drive and experiment mechanisms all fail in concert in such a way that control rods and experiments are withdrawn from the core region while safety rods are rendered inoperable. The cadmium poison sheets are manually positioned entirely within the graphite reflector, have no drive mechanisms, and are mechanically restrained so they will not move relative to the core during a seismic event. If the reactivity addition caused by control rod and experiment movement is sufficiently large, a power excursion not terminated by a scram could occur and result in fuel melting. The NTR will be operated in such a manner as to limit the potential excess reactivity to less than that required to cause fuel damage, assuming failure to scram.

From full power, the transient would be stopped by bulk boiling, even if all scrams fail, before fuel damage occurs for sizable step reactivity insertions. The results of a $0.76 step reactivity insertion are shown in Figure 13-8. Power peaked at 4 x 103 kW and PCT at 373 °F. The PCT and peak power characteristics versus magnitude of the reactivity step are shown in Figure 13-9.

10000 Power (kW) 1000 100 0 2 4 6 8 Time (Sec)

Figure 13-8 $0.76 Step from 100 kW - No Scram 13-22

10000 400 350 300 Peak Power (kW) 250 PCT (°F) 1000 200 150 Peak Power 100 PCT 50 100 0 0.20 0.30 0.40 0.50 0.60 0.70 0.80 K Step ($)

Figure 13-9 k Steps from 100 kW - No Scram To determine the effects of positive reactivity additions from less than full power and temperature, additional transients were run with an initial power level of 1 x 10-7 kW. Inlet water temperatures ranged from 55 to 90°F and the initial positive reactivity step of $0.76.

Results show that the positive reactivity feedback from the temperature coefficient, while not as important for the full-power cases because the feedback is very small, it is more important for the zero-power cases because coolant temperatures are lower.

Reactor power and peak fuel temperature versus time for a $0.76 step insertion from 1 x 10-7 kW and 55°F inlet water temperature is given in Figure 13-10. As can be seen from the results, limiting the step reactivity to $0.76 or less ensures that there are no mechanisms available which will cause fuel damage.

13-23

12000 1400 Clad Melting Temperature 1200 10000 1000 8000 Power (kW) 800 PCT (°F) 6000 600 Power 4000 400 PCT 2000 200 0 0 0 10 20 30 40 50 Time (sec)

Figure 13-10 Reactor Power and Hot Spot Fuel Temperature Versus Time,

$0.76 Step from Source Level, 55°F Coolant Inlet Temperature - No Scram Figure 13-11 Possible Reactor States Following the Postulated Seismic Event 13-24

The discussion above described the sequence of events for the first 2 minutes of the hypothetical event. A discussion of several possible sequences of events from the 2-min point in time out to the final state for the reactor follows.

The diagram presented in Figure 13-11 shows the various possible states for the NTR following the initial reactivity transient. If no operator intervention is taken, the final state of the reactor will always be State A (reactor shutdown caused by loss-of-coolant). The extremely conservative loss-of-coolant analysis presented in Section 13.5.6 demonstrates the loss-of-coolant for the NTR has no significant consequences.

The performance of the reactor in the near term after the postulated seismic event depends on the extent of damage to the remainder of the reactor system. The most significant items are:

a) the primary system piping, b) the primary pump, c) the secondary water supply system, and d) the electrical supply to the reactor system.

It is unlikely the primary system of the reactor would still be intact after a seismic event severe enough to result in the reactivity addition by the massive structural failure postulated here. If the primary system failed at the same time as the reactivity addition, the reactivity transient would not be significantly altered. The loss-of-coolant from the reactor results in power reduction by voiding the reactor core (Figure 13-11, State A).

We may also assume loss of electric power because: a) it is improbable that electric power to the site (including the NTR) would survive the event postulated here; and b) even in such an improbable circumstance, site emergency procedures call for the termination of all utility services to any buildings or facilities believed to have suffered damage.

As the loss of electrical supply automatically deactivates the primary system pump and automatically closes off flow to the secondary system, the structural fate of the secondary system becomes a moot question, and we need consider only the possibility of the primary system surviving the event. If the primary system does not fail, or leaks at a very slow rate, the system will arrive at State B. For this state, the reactor will operate in a natural circulation mode at low power and low flow. Since there is no secondary cooling, reactor power must be dissipated by the heat loss from the uninsulated reactor primary piping and by evaporation or boiloff of the 13-25

primary coolant. There are approximately 1000 of the 1800 gallons of water in the fuel storage tank which could drain into the reactor core can through the fuel loading chute to make up for the boiloff. The loss-of-coolant by boiling will be a less severe event than the loss-of-coolant event described in Subsection 13.5.6 for two reasons.

1. The reactor power is lower and, if a primary system leak was not developed, the loss-of-coolant is not complete. In fact, the slow loss-of-coolant will result in a slow decrease in power and only a partial loss-of-coolant will occur.

2. As the maximum fuel temperature for a loss-of-coolant occurring at a power level of 100 kW is less than the fuel melt temperature, there will be no fission product release from this accident. Since there is no release of radioactive products, there are no dose consequences associated with this accident scenario.

13.5.4 Rod Withdrawal Accidents The safety system and rod withdrawal procedures are designed to provide adequate control of the reactor at all times. Even if interlocks fail and the operator deviates from normal procedures so that the rate of power increase is not controlled by normal manual control rod movements, the reactor period and neutron flux level monitors would scram the reactor. If the reactor did not scram, the analysis in Section 13.5.3 is applicable. It is shown in the transient analysis the reactivity can be introduced in either a step or relatively long ramp without affecting the outcome. This analysis indicates that the transient which results from the total reactivity addition of the control rods, experiments, and temperature effect without scram (and the potential excess reactivity is $0.76) does not melt fuel. Therefore, the transient which would be caused by the withdrawal of all the rods can be accommodated.

13.5.5 Reactor Loss of Flow Accident To analyze the effects of a sudden loss of primary coolant recirculation pumping, it was assumed that the worst loss-of-flow accident (instantaneous seizure of the rotor in the single recirculation pump in the system) occurs. There is a reasonable probability of this scenario occurring because the coolant system has only a single pump. For such an accident, it is estimated that the pump flow will coast down to a natural circulation value within 0.1 second. The accident is assumed to occur while the reactor is operating at 100 kW. Although the transient would be terminated by the low-flow scram, in this analysis, it will be assumed that this scram does not function. After 13-26

the flow has decreased to the natural circulation rate, the coolant temperature and the natural circulated flow rate will increase. This trend will continue until either; a) bulk boiling at the hot spot produces enough voids to stop the power rise by reactivity feedbacks, or b) the average coolant temperature goes high enough to allow the negative temperature coefficient to halt the power rise. The initial core average coolant temperature is 106.2°F, and the initial excess reactivity is assumed to be zero. As the coolant temperature increases, the excess reactivity also increases to a turnaround temperature of 124°F, at which point the temperature coefficient becomes negative. Meanwhile, reactor power is on the rise, but will begin to slow down as the coefficient goes negative. The final steady-state operating point will correspond to a power and flow combination which gives the same reactivity contribution from temperature as for initial steady-state operation. Using the coolant temperature reactivity coefficient (Equation 4-2), this final coolant temperature level is 151°F. Thus, there is no bulk boiling in the average channel.

The heat flux is far below the heat flux necessary to initiate film blanketing. Moreover, the fuel plate surface temperature has been limited to a value well below the melting point, as a result of local surface boiling.

Maximum fuel temperature during the transient is 193°F at 54 seconds and then continues to decrease. Because fuel damage does not occur in this scenario, there is no release of radioactive products and no associated dose consequences.

13.5.6 Reactor LossofCoolant Accident The reactor loss-of-coolant accident involves the total loss-of-coolant inventory in the core as the result of a rupture in the primary system, combined with a failure to scram. The likelihood of this occurring is very low because it would require a catastrophic pipe sheer or core can failure that would prevent adding water to the core tank while completely draining the core tank. The accident is postulated to occur as follows:

Primary system ruptures at some point below the core entrance so that gross removal of core coolant supply occurs.

As the water in the core is removed, the fuel is uncovered; the uncovering of the fuel acts to shut down power generation to a decay heat level.

The rupture is taken as being large enough to cause a very rapid coolant loss so that all water is lost, and the core power is down to the decay heat level very shortly after the accident. It is 13-27

assumed there is no post incident cooling system in the reactor and, as a result, the only cooling of the fuel plates occurs by any natural convection air currents that may be set up and by radiation heat-transfer from the core to the graphite. For simplicity and conservatism, convective heat removal by natural air currents is neglected. It is further assumed that no heat escapes from the graphite stack to the outside environment.

The initial power level of the reactor is taken to be 100 kW, and the subsequent decay heating rates are given in Figure 13-12 as a fraction of the initial power. A power-peaking factor of 1.30 was assumed, which includes both the normal axial peaking and severe azimuthal skewing. The calculation was performed using the TRACG computer program. The nodal structure for the problem is adapted from earlier calculations using a version of the Transient Heat Transfer program (Reference 20) and is shown in Figure 13-13. Axial heat-transfer was neglected.

The peak fuel temperature and the volume-averaged graphite temperature are shown in Figure 13-14 and Figure 13-15 as a function of time after coolant loss. The fuel temperature reaches a maximum of about 626°F about 30 minutes after coolant loss and then begins to decline. The rise of the graphite temperature is almost imperceptible - only 15°F in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

The analysis was repeated using a higher peaking factor. The maximum fuel temperature for a loss-of-coolant accident with a 1.58 peaking factor is 800°F at about 20 minutes. Reactor power at the time of the peak fuel temperature is 1.5 kW. It has been shown that this power could be tolerated indefinitely without increasing graphite temperatures to over 150°F, assuming a natural convection heat-transfer coefficient of 0.6 Btu/h-ft2-°F on the exposed surface of the reactor.

Therefore, a second fuel temperature peak greater than 150°F is not possible. As a result, fuel damage will not occur and there is no release of radioactive products or associated dose consequences.

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Figure 13-12 Decay Heat Rate 13-29

Figure 13-13 Node Structure Adapted for TRACG Analysis 13-30

Figure 13-14 Fuel Temperature Following Loss of Coolant Accident Figure 13-15 Graphite Heat-up Following Loss of Coolant Accident 13-31

13.6 EXPERIMENT SAFETY ANALYSIS 13.6.1 Introduction Descriptions of the NTR experiment safety programs and associated facilities, equipment, and procedures are discussed in Chapter 10. As stated, before any experiment may be conducted, there must be a review and approval of a written description and safety analysis.

The purpose and requirements for experiment safety analyses are described in Section 13.6.2.

Considerations that will be addressed are identified and discussed.

The potential mechanical and radiological consequences of postulated accidents involving explosive material at the NTR facility are described in Subsection 13.6.4.3.

The limitations that will apply to all types of experiments are discussed. Adherence to these limitations is mandatory and will provide assurance of safe performance of experiment programs within imposed regulatory restrictions.

13.6.2 Safety Analysis The purpose of the safety analysis for experiments is to ensure that consideration is given to any feature of the design or conduct of an experiment, including intended functions and possible malfunctions, which could create, directly or indirectly, a radiological exposure hazard. When applicable, the analysis will consider:

Any interaction of an experiment with the reactor system that has potential for causing fission product release from the fuel.

Any interaction that could adversely affect an engineered safety feature or control system feature designed to protect the public from fission product release.

Inherent features of an experiment that could create beams, radiation fields, or unconfined radioactive materials.

Potentially adverse interaction with concurrent experimental and operational activities.

The safety evaluation for each experiment utilized in experimental facilities will consider:

The physical conditions of the design and conduct of the experiment.

The content of the material.

The administrative controls employed to evaluate, authorize, and carry out the experiment.

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A description of specific items that will be addressed, when applicable, follows.

Reactivity Effects The principal concern with a net positive reactivity effect is whether it is caused by the insertion of an experiment with a positive effect, or by the removal of an experiment having a negative reactivity effect. Every experiment or type of experiment, as appropriate, will be evaluated for:

Its potential reactivity worth.

The rate of change of reactivity of movable experiments.

Thermal and Hydraulic Effects An experiment will be evaluated to ensure its thermal limits are not exceeded and for its actual and potential thermal effects on reactor components and coolant. This evaluation will be made for the reactor at the extremes of its operating margin, as defined by limiting safety system settings.

The experiment design will be evaluated to ensure it will not adversely affect flux shape or reactor coolant flow considerations that were used to define or are implicit in the reactor safety limits.

Mechanical Stress Effects Materials of construction and fabrication and assembly techniques utilized in experiments will be evaluated, as appropriate, to provide assurance that no stress failure will occur from manipulation and conduct of the experiment or as a result of unintended but credible changes of, or within, the experiment. Every experiment or type of experiment, as appropriate, will be evaluated with respect to storage and possible uncontrolled release of any mechanical energy.

Material Content of Experiments Certain kinds of materials which may be used in experiments possess properties which could have significant safety implications. Limitations on the amounts of such materials can limit the consequences of experiment failures. The material content of an experiment will be analyzed and limited, as required, utilizing the following classifications as a guide:

Radioactive material Trace elements and impurities High cross section materials Highly reactive chemicals (explosives) 13-33

Corrosive chemicals Radiation sensitive materials Flammable material Toxic material Cryogenic liquids Unknown materials Administrative Controls Administrative controls are in place to ensure a written description and safety analysis are generated for each experiment type. Each experiment type must be reviewed by a technically competent independent review unit and approved by the Area Manager or his designated alternate. An experiment type includes repetitive experiments that involve common safety considerations and a similar reactor setup. Acceptance criteria for an experiment include compliance with regulatory requirements (including 10 CFR 20 and Technical Specifications),

GE procedures, and good safety practices. The independent review unit and its modulus operandi is discussed in Chapter 12.

Administrative controls applicable to all experiments are listed below:

1. An NTR Licensed Senior Reactor Operator must provide written approval of every experiment and must ensure that each person executing the experiment is knowledgeable in those procedures required to ensure safe conduct of the experiment.

2. Detailed written procedures must be provided for the use, or operation of, each experiment facility and each experiment type.

3. The Licensed Reactor Operator at the console must be notified just prior to moving any experiment (or series of experiments as specified by procedures) within the NTR facility.

4. Every experiment removed from the reactor must be subject to a radiation and contamination monitoring procedure, as applicable, that anticipates levels greater than those predicted.

13.6.3 Consequences of Accidental Explosions The facilities, equipment, and procedures used for experiment programs that involve explosive material are described in Chapter 10. To provide safe limits for the amounts of explosives 13-34

permitted in the NTR handling and radiography areas, separate Design Basis Accidents (DBAs) were defined for the south cell, the north room, and the set-up room. In general, these DBAs assumed a highly improbable accidental detonation of all explosive devices in the particular area and the consequences are evaluated in terms of both radiological and mechanical effects.

Radiological Consequences The radiological consequences of an accidental detonation of an explosive device are essentially nonexistent. Induced activities in explosive materials, structural materials containing the explosive, or structures used in neutron radiography are extremely small considering thermal neutron fluxes of 2 x 106 n/cm2-sec and normal exposure times of 103 seconds. However, if sufficient other sources of radioactive materials are present in the immediate area and become dispersed or airborne during the accidental detonation, the radiological consequences could be serious. Operations at the NTR include neutron radiography of uranium fuel pins and capsules containing significant amounts of fission products. Evaluation of the DBAs indicates that while it is virtually impossible to involve these materials in the accident, it is prudent to exclude these large sources of radioactive material from any area in which explosive devices are being handled.

Small amounts of radioactive materials (e.g., uranium contained in fission chambers or irradiated samples used in various experimental programs) may be safely stored in the south cell or the north room during the neutron radiography of explosives. By limiting these quantities to 10 curies of radioactive materials and to 50 grams of uranium, the health and safety of the general public will in no way be compromised. Storage locations are at least 5 feet from any explosive handling position and are normally either in concrete block caves or small lead casks. While accidental detonation of explosive devices might cause minor damage to the storage structures, the probability of releasing even a small percentage of the radioactive material from their contents is negligible. Assuming a 1% release and stable atmospheric conditions (inversion),

maximum site boundary doses are less than 20 mRem to the thyroid and 1 mRem to the whole body under this most pessimistic combination of circumstances. No radioactive materials other than those produced by neutron radiography are permitted in the set-up room if explosive devices are present.

Mechanical Consequences The primary safety criterion is that complete simultaneous detonation of all explosive devices in 13-35

a particular area will not increase the probability or consequences of accidents previously analyzed or create the possibility of a different type of accident not previously analyzed. While minor structural damage and possible injury to personnel will occur in the immediate area, damage to the reactor core, graphite pack, or control system is not expected, and injury to personnel is minimized. Damage to the reactor is prevented by limiting the amount of explosive material allowed in the particular areas (south cell, north room, and set up room) and by design and construction of an additional shield structure (south cell). Potential injury to personnel is minimized by strict adherence to safe explosive handling procedures. The mechanical safety analyses showed that neutron radiography of explosives can be accomplished safely in the reactor facility by limiting both the total quantity of explosive materials in pounds of equivalent TNT and the distance of the explosive material from sensitive components and structures.

Reactivity Effects There are no reactivity effects directly associated with neutron radiography of explosive or other materials. Objects undergoing inspection are located at relatively large distances from the reactor and have no effect on core reactivity. Even the large shutter in the south cell may be moved during reactor operation without affecting core reactivity. Some minor reactivity effects are associated with the neutron radiography beam preparation devices. Under normal circumstances, shock waves from accidental detonation of explosives will be attenuated sufficiently to make movement of the beam preparation device highly improbable. It is also noted that the reactivity added during removal or expulsion of the beam preparation device from the core region is included in the total amount that would be available, as discussed in Section 13.5.3.

Therefore, the consequences would be less severe than those analyzed, which assumed $0.76 step insertion both with and without scram.

13.6.4 Experiment Limitations Safety oriented limits and restrictions applicable to experiment facilities and experiment programs follow. The limits and restrictions presented are derived from the reactor and experiment safety analyses, approximately 60 years of experience in conducting experiments at the NTR, and sound engineering practice. Most of these limits are contained in the Technical Specifications. Adherence to the limits and restrictions below is mandatory and provides assurance that:

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1. There is no anticipated mode of experiment operation that will endanger the health or safety of the general public or plant personnel.

2. No experiment will be performed that involves a technical specification change or that has not undergone a 10 CFR 50.59 review.

3. A proposed experiment type will be evaluated in detail and its execution controlled so as to reduce any radiation exposure to the public and plant personnel to the lowest practicable level.

General Experiment Requirements

1. A written description and analysis of the possible hazards involved for each type of experiment shall be evaluated and approved by the Area Manager or his designated alternate before the experiment may be conducted. Records of such evaluation and approval shall be maintained.

2. No irradiation shall be performed which could credibly interfere with the scram action of the safety rods at any time during reactor operation.

3. Experimental capsules to be utilized in the experimental facilities shall be designed or tested to ensure that the pressure transients, if any, produced by any possible chemical reaction of their contents and leakage of corrosive or flammable materials will not damage the reactor.

4. No experimental objects shall be inside the core tank when the reactor is operating at a power greater than 0.1 kW.

5. Experimental objects located in the fuel loading chute shall be secured (secured experiment) to prevent their entry into the core region.

Reactivity Limits

1. Requirements pertaining to the reactivity worth of experiments are as follows:

a) The sum of the potential reactivity worths of all experiments which coexist plus the reactivity available from control rods and coolant temperature shall not exceed $0.76.

b) No experimental object shall be moved during reactor operation unless its potential reactivity worth is known to be less than $0.5 and the operation is performed with the knowledge of the licensed operator at the console. All 13-37

power operated, remotely controlled mechanisms for moving an object into the reactor core shall be energized from the reactor console; however, movement of the object may be initiated from another location. All manually operated mechanisms for moving an object into the reactor graphite pack shall be done with the knowledge and consent of the reactor operator at the controls of the reactor.

c) The potential reactivity worth of any component which could be ejected from the reactor by a chemical reaction shall be less than $0.50.

2. The potential reactivity worth of experiments shall be assessed before irradiation. If the assessment warrants, the reactivity worth of the experiment shall be measured and determined acceptable before reactor full-power operation.

Explosive and Flammable Material Lists

1. a) The maximum amounts of explosives (detonating and deflagrating, DOT Hazard Class/Divisions 1.1, 1.2, 1.3 and 1.4) permitted in the NTR facilities are as follows:

i. South cell: W = (D/2)2 with W 9 pounds and D 3 feet; ii. North room (without MSM): W = D2 with W 16 pounds and D 1 foot; iii. Set-up room: W = 25 pounds b) The maximum amounts of explosives allowed in the North room MSM (inclusive in the limit of 1.a.ii above) are as follows:

1. for DOT Hazard Class Divisions 1.1, 1.2, and 1.3 (detonating): W = 2 pounds

2. for DOT Hazard Class Division 1.4 (deflagrating): W = 4 pounds Where W = Total weight of explosives in pounds of equivalent TNT D = Distance in feet from the south cell blast shield or the north room wall.

2. A maximum of 10 curies of radioactive material and up to 50 grams of uranium may be in storage in a neutron radiography area where explosive devices are present (i.e.,

in the south cell or north room). The storage locations must be at least 5 feet from any explosive device. Radioactive materials other than those produced by neutron radiography of the explosive devices and imaging systems are not permitted in the 13-38

set-up room if explosive devices are present.

3. With the exception of communication equipment utilizing low-energy electromagnetic waves in radiofrequencies, such as mobile phones and two-way radios used by reactor personnel, unshielded high-frequency generating equipment shall not be operated within 50 feet of any explosive device.

4. No explosive device shall be placed in a radiation field greater than 1 x 104 roentgens or consisting of greater than 3 x 1012 n/cm2 thermal neutrons.

5. The maximum possible chemical energy release from the combustion of flammable substances contained in any experimental facility shall not exceed 1000 kW-sec. The total possible energy release from chemical combination or decomposition of substances contained in any experimental capsule shall be limited to 5-kW sec, if the rate of the reaction in the capsule could exceed 1 watt. Experimental facilities containing flammable materials shall be vented external to the reactor graphite pack.

TNT Equivalence The equivalence of an explosive material to TNT on a gram basis is determined by ratioing various parameters of the explosive to those of TNT. These parameters include brisance, ballistic mortar, trauzel test, and detonating velocity, and are described in Properties of Explosives of Military Interest, AMCP 706-177. This report contains pertinent data on many types of explosives and is used as a primary reference document. The equivalent grams of TNT for an explosive being handled or radiographed is determined by the following:

Parameter of explosive Gram equivalent TNT = grams of explosive x Parameter of TNT where the ratio of parameter is chosen to be the highest value of the brisance, ballistic mortar, trauzel test, or detonating velocity ratios.

If data are not available on the explosive, or the composition is proprietary, a factor of 2 is used for the parameter ratio, which is conservative and higher than any value found in AMCP 706-177.

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13.7 REACTOR SAFETY LIMITS 13.7.1 Introduction Safety limits for operation of the NTR are developed in this section. The safety limits presented also provide the basis for determining and specifying the Limiting Safety System Settings (LSSS) for important process variables.

Safety limits are developed for the reactor power, the only important measurable process variable with safety significance for reactor operation. Other process variables, namely core coolant inlet, temperature and reactor primary flow rate, have no significant effect upon the safety criterion over the entire range of core flow conditions, including natural circulation.

In this section, the safety criterion of Departure from Nucleate Boiling is discussed. The critical heat flux relationship and the thermal-hydraulic computer model used in the NTR safety limit analysis are described. The resultant safety limit curves are presented. Instrument uncertainties are applied to the safety limit curves to provide the LSSS for steady-state reactor operation.

13.7.2 Criterion for Development of Safety Limits Departure from Nucleate Boiling (DNB) has been selected as the most relevant criterion for development of safety limits for operation of the NTR. DNB is that stage of the boiling phenomenon when sufficient liquid is unable to reach the heating surface due to the rate at which vapor is leaving the surface. This restriction of the liquid flow causes an abrupt surface temperature increase above the saturation temperature in a heat-flux controlled situation.

The safety limits for the reactor power are chosen to restrict the actual heat flux in the hottest fuel element coolant passage below the DNB surface heat flux to preclude any subsequent fuel cladding damage due to a rise in surface temperature. The Departure from Nucleate Boiling Ratio, DNBR, is the ratio between the surface heat flux at DNB and at operating conditions; thus DNBR =

It was necessary to use two different correlations to evaluate the DNB for the NTR. The steady-state DNB condition is found to occur with saturated bulk boiling in a substantial portion of the core and is accompanied by a significant void fraction. The postulated reactivity transients presented in Section 13.5 reach a DNB heat flux with the core coolant significantly subcooled.

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13.7.3 Analysis for Development of Safety Limits Steady-State Critical Heat Flux Relationship The steady-state safety limit analysis required a DNB heat flux correlation which is applicable to low-velocity, low-pressure saturated boiling with a significant void fraction. As cited in Macbeth, (Reference 25), developed an empirical correlation of experimental data which presents the critical heat flux (Reference 25) as a linear function of the mass quality at the hottest surface location.

This correlation, which accounts for steam quality, is superior to other correlations which ignore the effect of void fractions and consider only other physical properties of the coolant. The Macbeth correlation states that the critical heat flux is proportional to the mass velocity in the low mass velocity region. The NTR core operates in the low mass velocity region for all operating conditions. The optimized correlation is H G 10 DNB 1 X 135 where DNB = departure from nucleate boiling critical heat flux (Btu/h-ft2)

Hfg = latent heat of vaporization (Btu/lb.)

G = mass velocity (lb./h-ft2)

Xmax = maximum quality =

The critical heat flux is calculated for the hottest location of the hottest channel.

Transient Critical Heat Flux Relationship The DNB correlation used to evaluate the reactor safety under transient conditions must be applicable to subcooled boiling. Macbeth developed the following empirical correlation for the DNB critical heat flux under subcooled, low-pressure, low-flow conditions:

. / / / /

DNB 0.247 H 0.00213 H 10 Hi where Hfg = latent heat of vaporization (Btu/lb.)

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= density of the vapor (lb/ft3)

= density of the liquid (lb/ft3)

D = hydraulic diameter (inches)

G = mass velocity (lb./h-ft2)

L = overall length (inches)

Hi = enthalpy difference between saturation temperature and channel temperature (Btu/lb.).

For a typical hypothesized transient, as presented in Section 0, the hottest surface location within the NTR core will attain DNB above a heat flux of 600,000 Btu/h-ft2.

The postulated accidents were originally analyzed using a DNB heat flux of 450,000 Btu/h-ft2).

Thermal Hydraulic Computer Model The computer model CORLOOP (Reference 27) was developed for the NTR natural circulation analysis. It is also used for forced convection analysis. CORLOOP includes a multi-channel core model and a circulation loop which includes the core, a heat exchanger, and a pump. The core model is illustrated in Figure 13-16; the circulation loop is illustrated in Figure 13-17.

Situations with the secondary coolant flow to the heat exchanger on or off and the primary coolant pump on or off were analyzed using the program. When the primary pump is off, the core is cooled by natural circulation. The core model is adapted to the NTR from the GETR multi-channel core model, CORFLO (Reference 28).

CORLOOP represents the parallel flow channels between the vertical fuel discs in the NTR core as four channels with six nodes per channel. This provides an adequate grid for determining the values for the measurable process variables. The CORLOOP program has been checked against hand calculations for steady state conditions and has been verified against actual operating conditions for the NTR.

In the current updated safety analysis report, the analyses of select accidents and postulated transients are reevaluated using TRACG code. TRACG is a two-phase, non-equilibrium code capable of simulating the thermal-hydraulics response of the reactor with kinetics feedback. The TRACG code (Reference 30) is the GEH proprietary version of the Transient Reactor Analysis Code (TRAC), a system code widely used for boiling water reactor safety analyses. The code 13-42

capabilities include 3D and point kinetics models, a multi-dimensional, two-fluid model for the reactor thermal hydraulics, and an implicit integration scheme for numerical calculations. The code was reviewed and approved as part of the application methodologies for Anticipated Operational Occurrence (AOO), stability, Anticipated Transient Without Scram (ATWS), and Loss of Coolant Accident (LOCA) analyses. Its applicability range extend from full operating pressure of boiling water reactors to atmospheric pressures (Reference 31) 13.7.4 Safety Limits The safety limit for the NTR was determined for two different kinds of events. The first analysis considers the steady-state high power operation of the reactor for various boundary conditions.

The second type of analysis considers the behavior of the reactor during various postulated transient events. This second analysis involves the indirect application of the safety limit concept. For the transient analysis, a scram trip point is assumed for important process variables, mainly reactor power, and a value is chosen for the DNB heat flux. After the transient analysis is performed, the integrity of the reactor fuel is evaluated, and the validity of the safety limit is determined.

The steady-state safety limits for the NTR were determined using the CORLOOP computer program. The analysis shows that the critical heat flux for the NTR is a strong function of the reactor power. Figure 13-18 shows, as one would expect, that the departure from nucleate boiling ratio approaches unity as the reactor power increases. Figure 13-19 shows the trend of increasing void fraction with increasing reactor power. The analysis shows, however, that the critical heat flux for the NTR is not significantly affected by the core flow rate, or the core inlet temperature (bottom plenum temperature), as shown in Figure 13-20 and Figure 13-21. The reactor power, therefore, is the only important measurable process variable to be limited. The safety limit for the reactor power assures that the actual heat flux never approaches the DNB heat flux.

The curve in Figure 13-19 is generated by running various sensitivity cases with different power levels by changing flow rates and inlet temperatures. The smallest power level that predicted subcooled boiling was 193 kW. The curve is then extrapolated (indicated by broken line) to zero power with zero voids; however, there is no voiding at power levels below 190 kW during normal operation.

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Figure 13-16 Multi-Channel Core Model of NTR (CORLOOP) 13-44

Figure 13-17 Schematic Diagram of the NTR Circulation Loop Model (CORLOOP) 13-45

Figure 13-18 Reactor Power Versus DNBR = Depart from Nucleate Boiling Ratio 13-46

Figure 13-19 Reactor Power Versus Core Void Fraction 13-47

Figure 13-20 Reactor Power Versus Relative Flow Rate 13-48

Figure 13-21 Reactor Power Versus Core Inlet Temperature In the analysis, the actual heat flux has been determined from the CORLOOP computer model.

The DNB heat flux has been determined from the Macbeth correlation for pool boiling conditions. At a reactor power of 430 kW, the DNBR reaches unity. At a reactor power of 190 kW, the DNBR = 1.5. As shown in Figure 13-20, a 30% decrease in DNBR corresponds to more than a 100% increase in reactor power. The analytical uncertainties present in the results 13-49

represent the RMS error of the empirical correlation, the physical differences in the flow conditions between the NTR core and the experimental apparatus used in two phase flow research, and the assumptions incorporated in the four-channel computer model of the core. A safety limit which corresponds to a minimum allowable value of DNBR = 1.5 provides a conservative and satisfactory margin to more than compensate for any analytical uncertainties.

The steady-state safety limit for reactor power is 190 kW, as shown in Figure 13-20 and Figure 13-22.

The curves presented in Figure 13-20 and Figure 13-22 do not extend below a relative flow rate of ~0.12. This flow rate is the value which would exist if the reactor is operated at 190 kW with the pump turned off. Steady-state operation below this flow rate at a power level of 190 kW or greater is not possible. Likewise, the steady-state operation of the reactor with inlet temperatures of less than 100°F, or greater than 150°F, is not possible at these power levels with reasonable secondary coolant inlet temperatures. Values of reactor power, flow rate, and core inlet temperature which fall outside these bounds do not represent steady-state conditions and should be evaluated on the basis of the transient safety limits and analyses.

The base transient analysis presented in Section 0, which required a reactor scram, were originally all performed assuming a scram occurred at 150 kW, a scram delay time of 0.200 seconds, and a DNB heat flux of 450,000 Btu/h-ft2. None of the anticipated abnormal occurrences or postulated accidents resulted in fuel damage using these values. The transients were reevaluated using the more realistic thermal-hydraulics tool predicting higher critical heat flux, confirming the adequacy of the thermal limits set for operation.

13.7.5 Instrument Uncertainties The instrument uncertainties are presented in Table 13-3 for each of the measured variables under consideration. These uncertainties, determined when the process variables were at their normal values and assumed unchanged over all acceptable LSSS, are both the systematic and random types. In general, systematic uncertainties include biases in calibrations, standards, signal transmitters, and recorders. Random errors include drift of instrument settings, signal-to-noise ratio of instrument electrical output, instrument instability, and operator-to-operator variation in interpretation within least count.

The uncertainty values for the three measurable process variables used in the heat balance for 13-50

reactor thermal power determination were determined by extracting the square root of the sum of the squares of the individual uncertainties in the contributing measurements.

Figure 13-22 LSSS and Safety Limit for Reactor Power in Terms of Relative Core Flow Rate 13-51

Table 13-3 UNCERTAINTIES IN THE PRESENT METHODS FOR MEASURING IMPORTANT PROCESS VARIABLES Reactor Power (Flux Monitoring)

Compensated Ion Chamber negligible High Voltage Power Supply negligible Pico-ammeter Setting Accuracy +/-0.25%

Pico-ammeter Calibration Accuracy +/-4%

Pico-ammeter Long-Term Drift +/-0.25%

Net Reactor Power Uncertainty +/-4.0%

Reactor Coolant Core Inlet and Outlet +/-3%

Differential Temperature Reactor Primary Flow +/-1.0%

Overall Instrument Uncertainty for Reactor Power = +/-8.0%

The LSSS have been chosen to ensure that reactor scram is initiated in time to prevent exceeding the safety limit for reactor thermal power during normal operation and anticipated abnormal occurrences, or violation of safety criteria during postulated accidents. The safety margin (the difference between the safety limit and the LSSS) includes systematic and random types of instrument uncertainties, and, for transient events, also includes the effect of safety system delay times. The LSSS appears as Curve D in Figure 13-22. The limiting safety system setting for the reactor power is 125 kW over the entire range of core flow conditions, including natural circulation. The value of 125 kW is a conservative setpoint well below the trip point of 150 kW used in the transient analysis of the postulated accidents and is used rather than the 190-kW steady-state safety limit because it is more restrictive.

Any quasi-steady event comprising changes in any process variables may be analyzed using the safety limit curves regardless of the rationale used for postulating the event. The most severe anticipated off-normal, quasi-steady event is one in which the reactor power is at its least favorable value of 150 kW. For this highly unlikely operating condition, the DNBR = 1.8.

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As a result of certain postulated accidents, the reactor power may exceed the specified safety limit without causing damage to the reactor fuel. The amount by which the safety limit may be exceeded is a time-dependent variable, with each case evaluated individually. Application of the limiting safety system settings for reactor power ensures that no damage to the fuel will occur for any transient resulting from the postulated accidents.

Curve A in Figure 13-22 shows the safety limit based upon the steady-state thermal hydraulics analysis. Curve B shows the safety limit curve adjusted to account for instrument uncertainties.

Curve C in Figure 13-22 shows the scram trip point used in the transient analysis. Curve D, the LSSS curve, represents Curve C adjusted to account for instrument uncertainties.

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14 TECHNICAL SPECIFICATIONS Technical Specifications have been developed for the NTR which follow the format of the 2013 revision to American National Standards Institute/American Nuclear Society (ANSI/ANS) 15.1.

Operation of the reactor within the limits of the Technical Specifications will not result in offsite radiation exposure in excess of 10 CFR 20.1201, 20.1301, and 20.1101(d) limits. Operation within the Technical Specifications also limits the likelihood and consequences of malfunctions and assures the health and safety of the on-site personnel and the public, and protection of the environment.

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15 FINANCIAL QUALIFICATIONS 15.1 FINANCIAL ABILITY TO CONSTRUCT A NONPOWER REACTOR Not Applicable 15.2 FINANCIAL ABILITY TO OPERATE A NONPOWER REACTOR The actual costs and estimates of costs to operate the NTR for the first five years of the renewal period are considered by GE Hitachi Nuclear Energy (GEH) to be proprietary. However, based on approximately 60 years of experience in operating the NTR at VNC, costs of NTR operation are well known and understood. GEH has significant assets and is capable of assuming total operating costs for the NTR for the duration of the license renewal period.

Current and anticipated sources of funding for the NTR, would come from the sales of services to various customers in the areas of neutron radiography, irradiation of test and research materials, and reactivity testing. The sales of these services cover all the NTR operating costs, excluding certain landlord type costs, which are required for operations of the site as a whole, and the expense of which is not directly attributed to the NTR. These landlord costs include cleaning, landscape maintenance, utilities, facilities maintenance, and security.

15.3 FINANCIAL ABILITY TO DECOMMISSION THE FACILITY GEH provides appropriate financial assurance instruments to demonstrate that sufficient funds will be available when needed for required decommissioning activities. In 2018, GEH began using a payment surety bond for a prescribed amount based on a site-specific NTR facility decommissioning cost estimate. The surety bond amount is reviewed annually, and a revised bond is submitted to NRC as necessary.

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16 OPERATING EXPERIENCE All reactor components of the NTR have been used exclusively in the NTR since initial criticality on November 15, 1957, with even the uranium-clad, uranium-aluminum alloy reactor fuel being manufactured on site exclusively for use in the NTR. The reactor was originally designed as an experimental tool to: (1) advance the Companys progress in the nuclear energy program, (2) provide a source of neutrons for sample irradiation or exponential experiments, and (3) provide a sensitive device for measuring reactivity. Since initial criticality, the NTR has performed innumerable experiments, sample irradiations (including rocks from the moon) reactivity measurements, sensor calibrations and uranium enrichment analyses. The NTR is now used predominantly for performing neutron radiography. The first neutron radiograph in the NTR was performed On August 30, 1966.

In order to accommodate all possible types of experiments conceived by GE, the original bounding for the facility was established using an extremely conservative fueled experiment involving the rupture of an irradiated plutonium capsule. Design of the facility followed this experiment by incorporating robust Engineered Safety Features that include confinement and ventilation systems that more closely resemble those of a large power reactor rather than those typical of a Class 2 research reactor.

The NTR is an easy to operate and easy to maintain facility. It is a low temperature, low pressure, low heat reactor so components are not unduly stressed. The primary system is constructed of aluminum and stainless-steel components and the primary coolant system is maintained at a high purity, so corrosion is not an accelerated concern. The reactor is also very accessible so that control rod and safety rod drives may be inspected and maintained regularly.

These inspections and tests have demonstrated that the NTR can be operated safely and that components with degraded performance may be detected and replaced.

16.1 REACTOR FUEL The fuel used in the NTR has an excellent performance history and does not release large amounts of fission gas without melting. Corrosion of the aluminum in water is minimized when the pH of the water is 6.5.

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The reactor primary coolant pH is maintained between 5.5 and 7.5 by maintaining the water purity below 5 µS/cm. High specific conductivity can be tolerated for short time durations during unusual circumstances.

The effect of aging of fuel was evaluated in 2020 and both the condition of fuel and the adequacy of monitoring fuel aging were determined to be adequate to support ongoing operation. Fuel cladding was concluded to be half of its original thickness after more than 60 years of operation and, while at least one fuel disc has a cladding breach, monitoring of Sr-91 and Sr-92 indicates the amount of fuel in the coolant has not been increasing at a significant rate over the last decade.

Chronic fuel degradation is monitored by primary coolant water conductivity and periodic sampling for Sr-91 and Sr-92. Significant degradation of the fuel cladding would be indicated in increased dose rates and escaped fission gasses by local area radiation monitors and the stack effluent monitor.

16.2 SAFETY RODS The safety rods have an exceptional performance history. The rods are accessible for inspection and testing. The tests include scram times, low-current magnet separation, rod withdrawal time, limit switches and interlocks, residual magnetism of the electromagnets, sliding friction, and spring force. Any component of the Safety Rods may be replaced if required.

16.3 CONTROL RODS The Control Rods have also had an exceptional performance history. The rods are accessible for inspection and testing. The tests include limit switches and interlocks, automatic insertion, rate of withdrawal, and position indication. Any component of the Control Rods may be replaced if required.

16.4 AREA RADIATION MONITORS (ARM)

The ARMs are accessible for periodic inspection, calibration, and testing. In March 2020, the entire radiation monitoring system was replaced with a digital system to increase reliability and eliminate installed check sources.

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16.5 CONCLUSION

The NTR is a simple, compact, accessible reactor. This is evidenced by the replacement of the primary; core can which occurred in 1976. At that time, the control rods, safety rods, startup sources and reactor fuel were removed. Most of the graphite blocks were relocated and the aluminum core-can and fuel reel assembly were replaced with new units and the reactor was reassembled. The extensive surveillance, testing and calibration program has resulted in a facility with an outstanding record of safe operation. Degradation of performance is evident when it occurs, and components replaced easily; assuring the NTR will continue to be a safe facility.

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v t e Letter Sequence Request
EPID:L-2020-RNW-0038, GE Hitachi Nuclear Energy, Requalification Program for the General Electric Nuclear Test Reactor - December 2021 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Questions Answers Results Approval, Approval Other: ML21053A071, ML21197A137, ML21197A139, ML22020A389, ML22132A272, ML22144A183, ML22144A196, ML22221A195, ML22221A199, ML22294A190, ML22339A140, ML23026A325, ML23033A672, ML23086C025, ML23101A185
MONTHYEAR2022-08-10 [Table View]

ML23086C028

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1.1 INTRODUCTION

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16.5 CONCLUSION

1.1 INTRODUCTION

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16.5 CONCLUSION

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