ML20325A205

NEDO-32740P, Rev 3, General Electric Nuclear Test Reactor Safety Analysis Report. Part 1.(Redacted)

ML20325A205
Person / Time
Site:	Vallecitos Nuclear Center, 05570888, 05573531, 05573532
Issue date:	11/30/2020
From:	Heckman D GE Hitachi Nuclear Energy
To:	Office of Nuclear Reactor Regulation
Shared Package
ML20325A193	List: ML20325A195 ML20325A196 ML20325A205 ML20325A206
References
NEDO-32740P, Rev 3
Download: ML20325A205 (110)

Text

NEDO 32740, Rev 3 NOVEMBER 2020 GENERAL ELECTRIC NUCLEAR TEST REACTOR SAFETY ANALYSIS REPORT Prepared by the Technical Staff of the Vallecitos Nuclear Center and GE-Hitachi Approved: --i,-a

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T. McConnell. Manager Nuclear Test Reactor

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

This public version of the NTR Safety Analysis Report is identified as "Rev 3" to align with the non-public version of the same.

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 .............. ... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ..... 1-7 1.5 COMPARISON WITH SIMILAR FACILITIES ................ .. .......... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. .. .. ......... 1-8 1.6

SUMMARY

OFOPERATIONS .................................................................................................................... 1-9 1.7 COMPLIANCEWITHTHENUCLEAR WASTEPOLICY 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 7 INSTRUMENTATION AND CONTROL ......................... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. ............. 7-1 7.1.

SUMMARY

DESCRIPTION ......................................................................................................................... 7-1 7.2. REACTOR CONTROL ROOM ..................................................................................................................... 7-5 iv

HITACHI 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 EXPERIMENTAL FACILITIES ...................................... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. .. .. ....... .. .. .. .......... 10-2 10.3 EXPERIMENT REVIEW ......................................................................... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. .. .. .. 10-7 11 RADIATION PROTECTION PROGRAM I 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 ....... .. ... ... .. ... ... .. ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... .. ... ... 12-6 12.3 PROCEDURES ............................................................... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. .. .. ....... .. .. .. .......... 12-9 12.4 REQUIRED ACTIONS .............................................................................................................................. 12-11 12.5 REPORTS ................................................................................................................................................... 12-12 12.6 RECORDS ....................................................................... .. .. .. ....... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. .. .. ........ 12-13 12.7 EMERGENCY PLANNING ....... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ..... .. .... 12-13 12.8 SECURITY PLANNING ............................................................................................................................ 12-14 12.9 QUALITY ASSURANCE ................................................... .. .......... .. .. .. .......... .. ........... .. .. .. .......... .. .. .. ........ 12-14 12.10 OPERATOR TRAINING AND REQUALIFICATION ............................................................................. 12-20 12.11 ENVIRONMENTAL REPORTS .................................... .. .. .. .......... .. .. .. ....... .. .. .. .......... .. .. .. .......... .. .. .. ........ 12-21 12.12 REFERENCES .................... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. .. .. .... 12-21 13 ACCIDENT ANALYISIS .......................................................... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ... .. .. .. .. .. .. .. .. 13-1 13.1 ACCIDENT-INITIATING EVENTS AND SCENARIOS ............. .. ... ........ .. ... ........... .. ... ........... .. ... ........... 13-1 13.2 TRANSIENT MODEL ................................................................................................................................. 13-3 13.3 ANTICIPATED OPERATIONAL OCCURRENCES .... .. ... ........... .. ... ........ .. ... ........... .. ... ........... .. ... ........... 13-6 v

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 13.4 POSTULATED ACCIDENTS ..................................................................................................................... 13-9 13.5 EXPERIMENT SAFETY ANALYSIS ...................................................................................................... 13-26 13.6 EXPERIMENT DESIGN BASIS ACCIDENT .......................................................................................... 13-33 13.7 REACTOR SAFETY LIMITS .......................................... ... ............. ... ............. ... ............. ... ............. .......... 13-38 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 REACTORFUEL .... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ..... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. 16-1 16.2 SAFETY RODS ............................................................... .. .. .. .......... .. .. .. .......... .. .. .. .......... .. ........... .. .. .. .......... 16-2 16.3 CONTROL RODS ........................................................................................................................................ 16-2 16.4 RADIATION AREA MONITORS (RAM) .................................................................................................. 16-2

16.5 CONCLUSION

................................................................. ... ............. ... ............. ... ............. ............. ... ............ 16-3 Figures FIGURE 1-1 REACTOR CELL, SOUTH CELL, AND CONTROL ROOM ........................................................... 1-5 FIGURE 1-2 NUCLEAR TEST REACTOR FACILITY ........................................................... .. .. .. .......... .. .. .. ......... 1-6 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 FIGURE 4-2 FUEL CONTAINER ASSEMBLY .... .. ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ... ... .. ..... 4-3 FIGURE 4-3 FUEL DISK .......................................................................... .. .. .. .......... .. .. .. .......... .. ........... .. .. .. ............ 4-5 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) ................................................. 10-2 FIGURE 10-2 MODULAR STONE MONUMENT NEUTRON RADIOGRAPHY FACILITY .......................... 10-3 vi

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 FIGURE 12-1 NTR ORGANIZATION .................................................................................................................. 12-2 FIGURE 13-1 NTR TRANSIENT .......................................................................................................................... 13-5 FIGURE 13-2 INITIATION OF 5 KWHEATERFROMREDUCED COOLANT TEMPERATURE (65°F) ..... 13-8 FIGURE 13-3 ~-STEPS WITH 150-KW SCRAM .................. ... ............. ... .......... ... ............. ... ............. ... .......... 13-11 FIGURE 13-4 1.3$ STEP FROM 100-KW WITH SCRAM ............................................... ... ..... ... ..... ... ..... ..... ..... 13-12 FIGURE 13-5 100-KW FINITE RAMP INSERTION WITH HIGH-FLUX SCRAM ......................................... 13-14 FIGURE 13-6 4$ RAMP IN 0.6 SECOND FROM 100 KW WITH SCRAM ...................................................... 13-15 FIGURE 13-7 2$ RAMP IN 0.3 SEC FROM 100 KW WITH SCRAM ............................................................... 13-16 FIGURE 13-8 0.76$ STEP FROM 100 KW- NO SCRAM ................................................................................. 13-17 FIGURE 13-9 ~STEPS FROM 100 KW- NO SCRAM .................................................................................. 13-18 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 .......................... 13-19 FIGURE 13-11 POSSIBLE REACTOR STATES FOLLOWING THE POSTULATED SEISMIC EVENT.. .... 13-19 FIGURE 13-12DECAYHEATRATE ................................................................................................................. 13-23 FIGURE 13-13 NODE STRUCTURE ADAPTED FOR TRACG ANALYSIS ................................................... 13-24 FIGURE 13-14 FUEL TEMPERATURE FOLLOWING LOSS OF COOLANT ACCIDENT ........................... 13-25 FIGURE 13-15 GRAPHITEHEATUP FOLLOWING LOSS OF COOLANT ACCIDENT.. .. .. ... ... .. ... ... .. ... ...... 13-25 FIGURE 13-16 MULTI-CHANNEL CORE MODEL OF NTR (CORLOOP) ..................................................... 13-42 FIGURE 13-17 SCHEMATIC DIAGRAM OF THE NTR CIRCULATION LOOP MODEL (CORLOOP) ...... 13-43 FIGURE 13-18 REACTOR POWER VERSUS DNBR =DEPART FROM NUCLEATE BOILING RATIO .... 13-44 FIGURE 13-19 REACTOR POWER VERSUS CORE VOID FRACTION ........................................................ 13-45 FIGURE 13-20 REACTOR POWER VERSUS RELATIVE FLOW RATE ................ .. ... .. ... ... .. ... ... .. ... ... .. ... .. .... 13-46 FIGURE 13-21 REACTOR POWER VERSUS CORE INLET TEMPERATURE .............................................. 13-47 FIGURE 13-22 LSSS AND SAFETY LIMIT FOR REACTOR POWER IN TERMS OF RELATIVE CORE FLOW RATE .......................................................................................................................................................... 13-49 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 STACKRELEASEACTIONLEVELS .......................................................... .. ... .. ... ... .. ... ... .. ... ...... 11-14 TABLE 13-1 NTR EXPERIMENT DBA ISOTOPIC RELEASE TO REACTOR CELL.. .................................. 13-37 TABLE 13-2 NTR EXPERIMENT DESIGN BASIS ACCIDENT DOSES ........................................................ 13-38 TABLE 13-3 UNCERTAINTIES IN THE PRESENT METHODS FOR MEASURING IMPORTANT PROCESS VARIABLES .............................................................................................................................................. 13-50 vii

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 company's 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 April20, 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 reactor's design and the company's 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 is maintained 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 degrees Fahrenheit, 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 the following:

1. Loss of electrical power

2. Loss of secondary cooling

3. Loss of facility air supply

4. Inadvertent start of primary pump

5. Fuel handling error

6. Uncontrolled reactivity increases

7. Loss of primary coolant flow (pump shaft seizure)

8. Rod withdrawal

9. Loss of primary coolant

10. Experiment failure The three acceptance criteria for anticipated operational occurrences are the following:

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

1. Release of radioactive material to the environs does not exceed the limits of 10 CFR 20.1101(d) or 10 CR20.1302 (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 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 1-3

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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, 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 reactor's 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 source, the reactor can provide unperturbed 12 2 neutron fluxes (at 100 kW) of about 2 x 10 thermal n/cm -sec and an epicadmium flux of about 12 2 1 x 10 n/cm -sec. When used as a detector, reactivity effects can be measured with a precision 6

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 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.

1-4

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Figure 1-1 Reactor Cell, South Cell, and Control Room 1-5

Figure 1-2 Nuclear Test Reactor Facility 1-6

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 ofunusual 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.

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 1-7

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 ofNTR 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 ensures that negative reactivity is available if needed for scram before a control rod can be moved.

1-8

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

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

OFOPERATIONS 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.

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.

1-9

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 1950's 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 ofEnergy 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 Electric's application, dated June 5, 1957. Operation of the reactor up to powers of30 kW was authorized by facility license R-33, issued on October 31, 1957. Initial loading of the reactor began on November 7, 1957, and criticality was first achieved on November 15, 1957.

1-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 The NTR reactor was operated at powers up to 30 kW for more than 5000 hours0.0579 days <br />1.389 hours <br />0.00827 weeks <br />0.0019 months <br /> 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 <br />0 hours <br />0 weeks <br />0 months <br /> 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), remote area gamma 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.

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.

1-11

HITACHI The SAR, APED-444A was again revised and reissued as NED0-12727 in April 1977. The purpose ofNED0-12727 was to update the description of the facility and its organization and to summarize additional safety evaluation of the facility. NED0-12727 was submitted as supporting material for the renewal of the operating license.

The SAR was again issued in August 1997 as NED0-32740 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 was submitted on February 2, 2016, to address the unconditional release (pursuant to 10 CFR 50.83) of approximately 610 acres on the north end ofthe VNC property.

This amendment includes revision 2 to the Technical Specifications (NEDO 32765, revision 2) and change pages to the SAR (NEDO 32740, revision 2). This amendment is pending approval with the NRC.

A license amendment was submitted on July 1, 2019, to address a revision to the site emergency plan to align with the ANSI 15.16, 2018, Emergency Planning for Research Reactors. This amendment is currently pending with the NRC.

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.

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 60 annual reports submitted to the NRC on operating experience pertinent to safety as of 2020.

1-12

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 1.8.4 Recent Modifications In 2003 a change was made pursuant to 10 CFR 50. 59(c) to route power from the site's backup diesel generator to the rod drive bus so that the reactor operator can insert safety and control rods during an extended loss of commercial (normal) power. This is not credited as emergency power and was installed in order to allow the control rods to be bottomed and the reactor secured without having to wait until normal power is restored.

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.

1-13

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Department of Energy JUL 1 :t 1983 Washington, D.C. 20585 Jolr. R. W. Darmitze1

~anagers lrradfat1on Genera 1 Electric Company 175 Curtner Avenue San Jose. California 95125

Subject:

Contract Number OE-CROl-83~£44426 Enclosed he r ewith is one fully execute<2 *c opy of subject contract.

Please acknowlel.lge receipt of this contract by completing the "Acknow-ledgement .. below and returning it to:

u.s. Department of Energy Office of Procurement Operations Attn; Thomas s. Keefe. MA-453.1 koom Number lJ-027 1000 Independence Avenue. SW Washington, DC 20585 Sincerely,

.tJLJ-X-c--

/.t' lhomas S. Kufe Contra~ting Offieer Office of Procurement Operations Enclosure ACKNOWLEDGEMENT:

R. W. Darmitzel , Manager

• I r radia tion Processing Operat ion contractor's Authorized Representative Name and ~itle (Type/Print) signature Dat M * ~~~ /7$'3 Figure 1-3 DOE Spent Nuclear Fuel Disposal Agreement (page 1) 1-14

GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 APPENDIX A NUClEAR POWER REACTOR (S) OR Ont£R FACiliTIES COVERED Purchaser. General Electric Company

-~ntract Number/Date))£-C r!pqo JJf trQ' I _...;:'-;;;.:4:..:uZ;.:.I..&.r,.....l:.-----

~~~;tor/Ftc:11ity N&me Vallecitos Nuclear Center Location:

Street VaHecitos Road, P.O. Box 460 C1ty Pleasanton Count)'/State Alameda 1 Cal ;fornia

--~~~------------

Z1p Code 94566

~pac:1ty (MWE)

• Gross __N_o_ne_ _ _ __

Rea eta r T>'pe:

• See below.

BWR 0 pWR 0

.~ner . (I dent 1fy) --------------"!"""""----

Fit1U~y Description . Nuclear research and development facility including research reactors, hot cens. and various laboratories.

O.te of CQmlenc:ement of* 0p,er1.tion _1_9_5_6_ _ _ __

-(actua 1 or est 1uted)

IRC License 1: SNM-960, TR-1, TR-33 By Purchaser:

--~Slgn&ture ~

Manager, Irradiation Processing June 15, 1983 title 6'perat1on-- Date

• The SNF to be disposed of by the Vallecitos Nuclear Center under this contract has been irradiated in one or more of the folloWing reactors: General Electric Test Reactor. Nuclear Test Reactor, -Mont;cello~

• Quad Cities-1, Big Rock Point.

Vermont Yankee. Millstone-1, Humboldt Bay;3and Brunswick-1.

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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 2 SITE CHARACTERISTICS 2.1. GEOGRAPHY AND DEMOGRAPHY 2.1.1. Site Location and Description

2. 1.1.1 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 500ft (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 ofwhere SR 84 intersects highway 680.

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

24 Lafayette Fre Gate Berkeley Byron nal t10n v a Danville Tassajara Mountain San Francisco Alameda San Ramon iij House

• Banta CD _ Altamont Tracy v San Leandro Lyoth Dublin DalyCity §)

v Ulmar Carbona Hayward v Uvermore South San Pleasanton Francisco San Bruno Vallec*tos Nuclear Center Pacifica @@

CD Union City Sunol Mendenhall Spnngs San Mateo Fremont V Bair Island @

Don Edwards El Granada G§ San Francisco Redwood City Bay Nati onal Hall W*ldhle._

Moon Bay Palo Alto Milpitas Mountain Lobltos View CD Joseph

@ San Jose D. Grant San Gregorio Cupertino County Park La Honda Figure 2-1 San Francisco Bay Area Map 2-1

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 2.1.1.2 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 400ft (120m) 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.

2-2

GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 g

- - r

/

. ~

~ , '

lf j'll I

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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 2.1.1.3 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 Gryphon.

There are three other reactor facilities within the VNC Site Developed Area: DPR-1, Vallecitos Boiling Water Reactor (VBWR); DR-I 0, 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 comer of the Site.

2-4

GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 I

J V:!llecito:s RoOOi. SR :114 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 20.1301limits at the Site Boundary.

2-5

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Population Within 2.5 Miles (4 km):

The City of San Francisco's 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.

The 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 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 ofWilliamson Act contracts. According to plan projections, housing units in unincorporated lands were anticipated to increase from a total of300 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 31 0-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 full4.5 acres (0.02 sq. km) and a minimum parcel size of2 acres (8,000 sq. m) (Alameda County Community Development Agency 1997).

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

2-6

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 livl!m1om I:J Ro<ild! i n~rovemBnts

- l ltlill~ad trbcb

• WIIramsoo A.ct

- 1P11<M A,g.!lc:ultut~I!Jiod I Non,P!i~ l\bi(;;.ltur;, Ui!ldl

,tltlrJ-ReMwal I~HNOI!OO I.Md Urban *nd B141l ~ Iandi D W,Icr lri.Va,l ey Co~oancy

tos.eme~l 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.

2-7

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 in 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) .

.-.... . . ~

• • e - - **

• Canoord Riohmanct San Francisco .a Oaldand :. ___ **** * -**

• -*-* ..,.('

Daty Ci.ty \

San Francisco Bay Area @ Heyward ~

S1111 lee \. ~monl Red>liood C11y ,

,..-* Sun11~

100000 -~000

' San Jose 50.000

• UIO.OOO  :*. , Sal\'l,aCiar:a

..-!11.000

--==--==-* 501il11 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 toNE across the site's 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 (degrees Fahrenheit).

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 (degrees Fahrenheit).

2-9

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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. GEOLOG~ SEISMOLOG~ 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.

2-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 These deposits exist on the faces of higher elevations as the result ofuplift 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.

2-11

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 20.1201limits.

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 bumup. At present, all MPS have been removed except for 1/161h 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 failed 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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. WATERDAMAGE 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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. 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.

3-3

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 reactor cell) located in

-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, 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 of22 ft (6.7 m) wide by 23ft (7 m) long by 24ft (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

HITACHI ort NEDE 32740P, Rev 3 gross volume of the cell is and, with the above-mentioned equipment installed, the net air volume is approximately 10,500 ft 3 (300m 3 ).

Normal access to the cell is through the large doorway 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 minimum 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 system's 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.

3-5

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Figure 3-1 Building 105 Floor Plan 3-6

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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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:

• O n e - secondary cooling water supply line;

• O n e - deionized water supply line;

• O n e - drain 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;

• O n e - compressed air supply line;

• T w o - ventilation exhaust ducts through the cell roof;

• F o u r - pipes to the adjoining laboratory; and

• T w o - 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

• Horizontal facility tube

• hole through t h e - wall at approximately core centerline height 3-8

HITACHI ort NEDE 32740P, Rev 3

• Stepped hole through the. wall approximately - above the.floor.

• Hole for future thermal column through the. wall

• Two holes in t h e - wall, accessing t h e -

• T h e - hole contains radiation area monitor cables and t h e - 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 t h e - cell roof The fan draws air from the - cell, . . cell 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-Muller 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 picoammeter. 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 1-131 and gross ~-Yanda, respectively.

3-9

GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

,..... ~--

1crm aoo 106 FURNACE EXHAUST I

I KANNE CHAMBER/

PAR11CULATE MONITOR PIG NTA STACK I

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L Q SAMPLE STATION I

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OUT OF SERVICE Figure 3-3 Line Diagram of Ventilation System 3-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 of25 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 l-inch 4-1

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 a:

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"-::l II ;II Figure 4-2 Fuel Container Assembly 4-3

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 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.

4-4

HITACHI 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 ofthe 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 Rods Movable neutron absorbers located about the periphery of the fuel container include:

4-5

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 (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.

Orientation of these rods about the core is shown in Figure 4-1, 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-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.

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 4-6

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 spnngs.

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.

(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 4-7

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

• 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, flight 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 8.3 ¢/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 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 4-8

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 .

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 4-9

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 approximately 2.3$. The speed of withdrawal of each coarse control rod drive is 0.140 inch/second. The speed ofwithdrawal ofthe 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 2¢/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 of0.032-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 respect to the reactor assembly. The reactor shall be operated using only those positions which 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.

4-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 (3/8 MPS in slot #5), the worth of the MPS is approximately 0.7$ (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 February 2015, the worth from changing MPS size from the single installed 1/4 sheet in slot #2 to a 1/8 sheet in slot #2 was determined via measurement to be approximately 0.2$. 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 0.4 °F reduction in primary coolant temperature (from 70.4 °F to 70 °F) between the pre-change and post-change conditions.

4.2.3 Neutron Moderator and Reflector The graphite reflector-moderator is a 5-foot cube of reactor-grade (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.

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 ofblocks situated between the fuel container and the north face; and the group of blocks that fill the 11.5-inch- diameter hole 4-11

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 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 1-beams. The 1-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.

4-12

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

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 0-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.

4-13

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

4-14

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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, Shielding material has also been added above the . . cell entry ceiling and adjacent to the penetration through the east wall of the south cell.

4-15

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

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.

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.

4-16

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 r o o m - ) 1 1

- c e l l roof directly above reactor (top shield slabs in place) 65 65

- c e l l roof 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 degrees 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 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 (forNTR operation as of9/30/2019), the total 4-17

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 reactivity losses are estimated to be 2.1$ from the fuel bumup, 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.

4.4.1.1 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).

4.4.1.2 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.

4.4.1.3 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.

4.4.1.4 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.

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

4.4.1.6 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 4-18

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 within two hours.

4.4.1. 7 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.

Figures 4-4 and 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 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.

4-19

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 COF) and p 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.

4-20

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 0.016 L1k/k (2.3$)

- All four safety rods (conservatively assumed) 0.014 L1k/k (2.0$)

Net Reactivity (Console Excess with Typicala Operational Core)

All four safety rods and three control rods withdrawn +0.002 L1k/k (+0.3$)

All four safety rods withdrawn, and all three control rods -0.014 L1k/k (-2.0$)

inserted All four safety rods inserted, and all three control rods -0.012 L1k/k (-1.7$)

withdrawn All four safety rods inserted, and all three control rods -0.028 L1k/k (-4.0$)

inserted Reactivity (Console Excess)

- All four safety rods and all three control rods inserted and -0.023 L1k/k (-3.3$)

MPS withdrawn (typicala operational core)

Reactivity Addition from Primary Coolant Temperature change +0.00048 L1k/k (+0.07$)

(from 75 to 124°F) a 3/8 MPS in slot #5, 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.

4-21

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Table 4-2 NUCLEAR PARAMETERS (continued)

PARAMETER VALUE Miscellaneous Reactivity Effects

- Removing graphite from central sample tube (3-in. -0.009 L1k/k (-1.3$)

cavity)b

- Filling central sample tube with water (3-in. cavity) -0.02 L1k/k (-3$)

- Removing all graphite from vertical facility -0.008 L1k/k (-1.1$)

- Removing the fuel loading chute plug -0.006 L1k/k (-0.89$)

- Equilibrium xenon at 100 kW -2.3 X I0-3 L1k/k (-0.3$)

- Yearly fuel burnup (typical use) -2.3 X I0-4 L1k/k (-0.03$)

Coefficients ofReactivity

- Temperature coefficient in 0 Water coolant (measured) -5.7 X I0-3 (T-124) ¢/°F 0 Inner graphite (calculatedY +0.018 ¢/°F 0 Outer graphite (calculatedY +0.55 ¢/°F

- Average void coefficient -5.7 ¢/%void

- Doppler coefficient Negligible Mean Lifetime of Prompt Neutrons 2 x I0-4 sec Neutron Fluxes at 100 kW

- Average thermal flux in fuel 7 x IOll n/cm 2 -sec

- Peak thermal flux in central sample tube 2.5 x 1012 n/cm 2 -sec

- Peak thermal flux in CHRIS 8.0 x lOll n/cm 2 -sec

- Thermal flux at face of thermal column 7 x 108 n/cm 2 -sec

- Thermal flux at face of 5-ft graphite cube 5 x 1010 n/cm 2 -sec Miscellaneous Parameters After 198.5 MWD of Operation

- Reactivity lost due to fuel burnup 0.015 L1k/k (2.1$)

- Reactivity lost due to aggregate fission product poisoning 0.012 L1k/k (1.8$)

(other than samarium-149)

- Reactivity lost due to samarium-149 (at equilibrium) 0.0044 L1k/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.

4-22

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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$. This is calculated with: the strongest safety rod stuck in the full out position, all control rods full out and a xenon free core. The total safety rod worth of 3.86$ minus the maximum potential excess reactivity of0.76$ is 3.1$. This value minus the strongest rod of 1.1$ gives the shutdown margin of 2$.

The Technical Specifications for NTR state that, the minimum shutdown margin with the maximum worth safety rod stuck out shall be 1$. 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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- and quasi steady-state. The reactor thermal power level when the 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 of0.76$, 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 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 day's 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Chapter 13. The analyses show that a reactivity step addition of0.76$ will not cause significant fuel degradation.

The reactor shall be subcritical 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 1$. 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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(a,~) 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(a,~) 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

4.4.4.1 Calculated Core Parameters Neutronics parameters and reactivity values are calculated for BOL (fresh fuel) and current exposure (as of9/30/2019 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 halfMPS in slot #5 based on probable MPS configurations for initial NTR operation. The 4-27

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 current exposure core model includes a 1/8 MPS in slot #2 consistent with NTR operating records for the current exposure core configuration.

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GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 4.4.4.2 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(a,~) 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):

acraphiteTemp (¢)

0 F = dT dp Graphlte Equation 4-2 where Taraphite is the volume average graphite temperature (°F), p 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, jJeff, 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 BOLand current exposure cores. The BOL core model includes a full MPS in slot #1 and a halfMPS in slot #5, and the current exposure core model includes a 1/8 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.

4-30

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Table 4-3 CORE MODEL CALCULATED POINT KINETICS PARAMETERS Effective Mean Delayed Standard Standard Lifetime of Generation NTR Fuel Composition Neutron Deviation for Deviation for Prompt Time, l [sec]

Fraction, Peff l [sec] Neutrons Peff [sec]

BOL 0.00732 0.00013 1.99 x w- 4 1.02 x w- 6 1.99 x w-4 Current Exposure 0.00727 0.00013 2.29 x w- 4 1.09 x w- 6 2.2s x w- 4 4.5 THERMAL-HYDRAULIC 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 psi a, is 228°F. Thus, the state of the coolant is far removed from boiling at the design operating condition.

4-31

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 ofthe 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 over-all 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.

4-32

HITACHI Table 4-4 TYPICAL NTR CORE THERMAL AND HYDRAULIC CHARACTERISTICS Maximum thermal power level (scram) 125 kW Maximum thermal power level 100kW 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 ft 2 Total core power peaking factor 1.58 Core average pressure level 20 psia Coolant flow characteristic Total core flow area 0.39 ft 2 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 122lb/hr Mass flow rate, hottest channel 64lb/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 Hottest channel Maximum temperature difference, fuel-to-cladding surface 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:

4-33

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

4-34

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 site's 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, ofwhich 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 is utilized 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 The temperature coefficient at temperature T (Fahrenheit) is given by Equation 4-2:

-5.7 X 10- 3 (T-124) ¢/°F To ensure the temperature coefficient remains negative above 124°F, it must be verified whenever changes are made to the reactor that could affect the temperature coefficient. 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.

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 COF) 124 Conductivity (mhos) <1.0 pH 5.5 to 6.5 5-3

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 necessary 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 site's 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-offvalve, 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 of0.25-inch-ID, 36-inch long stainless-steel tubes, and provides a heat-transfer surface of approximately 36 ft 2 . 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 psi g. The design specifications pertinent to maintaining system integrity are given in Table 5-1.

5-6

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 0 ) 124 60 Temperature out (F 0 ) 90 80 Pressure Drop (psig) 1.6 4.0 Design Pressure (psig) 300 150 Test Pressure (psig) 500 300 Design Temperature (F 0 ) 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; furthermore, should such a leakage occur, the contaminated water would drain to the Site retention basins. The basin water is analyzed for radioactive material content before it is released.

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 rapid 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.

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 valves and solenoid valve (operated by key lock at the reactor console) so that one coolant system cannot flow to the other system. Addition to the primary system can only be accomplished from inside of the reactor cell; otherwise the 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 required for 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

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 controlled by controlling the water conductivity. The conductivity is operationally maintained at or below 2 11S/cm. The pH then, will be between 5.6 and 9.0, which is compatible with aluminum/stainless steel systems. 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, of high enough level, would cause the reactor cell remote area monitor to alarm. The area monitor detector is located on the reactor cell near the primary flow transmitter and is set to alarm at 106 mR/hr. The Senior Reactor Operator will determine the cause 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 10 11S/cm except for time periods not exceeding 7 consecutive days when the specific conductivity may exceed 10 11S/cm but shall remain less than 20 11S/cm. If the specific conductivity exceeds 10

!J.S/cm, steps shall be taken to assure the specific conductivity is reduced to less than 10 11S/cm.

5-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 The minimum corrosion rate for aluminum in water (<50°C) occurs at a pH of6.5. Maintaining water purity below 10 !J.S/cm will maintain the pH between 4.8 and 8.7. These values are acceptable for NTR operation. High specific conductivity can be tolerated for shorter durations 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, Tables 4-1 and 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 em), 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 Retch-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 site's 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 NTR' s 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.

5-11

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 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. NITROGEN-16 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 em x 50.8 em x 76 em) 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 em) aluminum. There are two 4-inch (10.16 em) diameter tubes and one 2-inch (5.1 em) 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.

5-12

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 NTR's robust confinement, and ventilation systems are described in detail in Section 3.5 and stack action levels are included in Chapter 11.

6-1

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 picoammeters and for loss of high voltage for the three Compensated Ion Chambers (CIC) in the picoammeter 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 picoammeter (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.

7-1

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

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 #1, 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).

7-2

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 Table 7-1 SCRAM SYSTEMS Item System Condition Trip Point* Function No.

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

1. Linear No less than 90%

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

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

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

3. Temperature High core outlet temperature :5 222 °F Scram (Fenwall)

No less than 15 Primary Coolant

4. Low Flow gpm when reactor Scram Flow 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)

• Trip points are the nominal measured values and need not consider the uncertainty in the channel 7-4

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 h2. REACTORCONTROLROOM The location of the reactor control room is shown in Chapter 1, Figure 1-1.

Building 105 hallway on the south, and other laboratories to the west. The shielded doors between the control room and the two cell areas are the only personnel entries to these areas. A flashing warning light at the south cell doorway is actuated if the door is opened; however, the radiation level in the south cell must be above a preset minimum for this warning device to actuate. Since the . . cell door will be open to perform 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, north 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 reactor 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. 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-standard 7-5

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 R-19 and R-20 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 (R-19 and R-20) 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.

7-6

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

• 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 picoammeters and the fourth by a log Nand period amplifier.

The three picoammeters 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 Nand period amplifier can cause a reactor scram on fast period. The picoammeters have 20 ranges covering 10 decades of power from I0-9 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 measurement which utilizes the coolant flow through the core and the differential temperature across the core.

7-7

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 High voltage for the three picoammeters CICs is internal. When in a two-out-of-three situation, loss of positive high voltage on any two picoammeters will cause a reactor scram. When in a one-out-of-two situation, loss of positive high voltage on any one picoammeter 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 picoammeters 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 picoammeter amplifier output signal, in addition to driving the picoammeter 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 picoammeter 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 picoammeter out of service, provided the trip circuit is set up so that a trip signal from either of the remaining picoammeters will cause reactor scram. If one picoammeter is out of service, the interlock at the low end of the scale for that picoammeter which prevents rod withdrawal may be bypassed.

These automatic actions ensure that the picoammeters have the proper start-up sensitivity and the high-power scram trip point is always within a decade of operating power during operations which increase power.

The log Nand 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 7-8

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 through the noncoincident logic circuit to initiate a scram for reactor periods of less than 5 seconds. At powers of less than 0.1 kW, a signal from the log N recorder actuates automatic bypass of the primary coolant low-flow scram. At powers greater than 0.1 kW, a signal from the log N amplifier actuates a relay which automatically switches the signal to a single function recorder from the start-up channel (source range monitor) if utilized to the thermocouple pile (millivolts) signal, which is used in the heat balance calculation. The log N power signal is also recorded on a strip chart recorder. The mode (multi position 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 picoammeters 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 picoammeters 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 R19B;

• Two 24-V signals that pass-through scram relay R20B; 7-9

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

• 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.

h4. SAFET~RELATEDITEMS 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.

7-10

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 water Visible and audible alarm; audible alarm may be Tank Water Level level bypassed after recognition 3 Primary Coolant High core Visible and audible alarm; audible alarm may be Temperature outlet bypassed after recognition (TC7) temperature 4 Primary Cooling Core delta Provide information for the heat balance Core Temperature temperature determination Differential 5 Stack Beta-gamma Visible and audible alarm; audible alarm may be Radioactivity particulate high bypassed after recognition level Noble gas high level 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 Rods not in Safety rod magnets cannot be reenergized; may 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.

7-11

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3

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. A thermocouple pile is provided which indicates the primary coolant core temperature differential. This 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.

7-12

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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.

7-13

GEH Nuclear Test Reactor S I I I~

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 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 and are provided for personnel safety (ALARA). Areas monitored are t h e - cell, . . c e l l , - room, and . .

room (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 control 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.

h& NEUTRONSOURCE 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 rods to withdraw the source. Following a process scram (de-energize R19 and R20), 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 7-15

HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 n/sec is used for a startup source. It is an R-Monel encapsulation approximately liz-inch in diameter and 3 liz-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.

7-16

HITACHI GEH Nuclear Test Reactor NEDE 32740P, Rev 3 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 remain 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 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 on site, and

• a licensed SRO will be present or readily available on call.

8-1

HITACHI GEH Nuclear Test Reactor NEDE 32740P, Rev 3 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 of0.75 amp to the safety rod magnets. This current will decrease approximately 50 rnA as the magnet coils heat up.

The power supply is regulated to less than 18 m V change for 105 to 125 V AC 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 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

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HITACHI GEH Nuclear Test Reactor S ort NEDE 32740P, Rev 3 8.1.3 Log N Power Supply This power supply is a General Electric INMAC 194X606G 1. It receives 120-V AC 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-V AC input and supplies positive and negative 24-VDC, 10-amp output for the three picoammeters.

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 V AC 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.

8-4