3 3679 00060 6378
FAST FLUX TEST FACILITY
CONCEPTUAL SYSTEM DESIGN DESCRIPTION
FOR
THE REACTOR CORE SYSTEM
NO. 31
August 27, 1969
PACIFIC NORTHWEST LABORATORIES Rlchland, Washington 99352
Operated by Battelle Memorial Institute
for the
BNWL-500
Volume 31
U"Sc Atomic Energy Commission under Contract No. AT(45-1)-1830
BNWL-500 Volume 31
encompassing the total vertical array of assemblies. To
assure predictable response of active core components during
all phases of operation and to limit permanent deformation
to values which are within clearance limits, a radial restraint
system utilizing reflector row assemblies for load application
is included in the radial support system. The radial restraint
system is designed to offset the effects of irradiation and
thermal bowing and creep by packing the assemblies together
at contact rings prior to startup of the reactor. The restraint
load is applied by a compliant beam system which in turn is
actuated by an over-center lever attached to the shield assembly
cover plate. The plate is retained vertically by bolts with
oversized heads which, through clearance holes in the plate,
permit limited lateral shift. This lateral motion is included
to allow this plate to be located relative to the top of the
core elements through the radial restraint and instrument tree
lower support point. Out of vessel actuation for the lever is
supplied by a ball-nut jacking system driven by a removable
ganged drive motor system.l
Shutdown for maintenance of the reference core involves two
parallel phases, an out of vessel test handling machine and
the In-Vessel Handling Machine utilized for all normal outage
in-vessel component manipulation. 2 The following sequence is
required to initiate in-vessel handling once the reactor has
been verified to be subcritical and the primary sodium tem
perature has been reduced to the shutdown condition. The reduc
tion in temperature is required to provide clearance through
1. Refer to Drawings, Appendix F, SK-3-14433. 2. Refer to References, Appendix A, Item 10.
2-7
,------- 1'---! I : BATTELLE-NORTHWEST!
DESiGN DESCRIPTiON CHANGE I~OTICE li-12-69 i PAGL I RICHLAND, WASHINGTON
bl~. "C" __ pC °NNo~ E_~_T_U_~_L. _--.D_~_R_E_L.I_MI_N_A:_Y ____ ~ FIN A L.
1--,-: 2. DOCUMENT TITL.E AND DATE
)' A-0121F Conceptual System Design Design Description
for the Reactor Core System No. 31
-------~------------
o PROPOSED
BNWL-500 Volume 31
.1 August 27, 1969
i-4:--c'O-N T R A C:-::T:-------+-;S:-. -C:C-'-O-N -:-T RO:-A:-:CC-:::T'-'U-'-A-L. -'-A-'-'U-::CT ':-H O:-:R-:-CI=-T Y,-:---------------I-·------·
.1 I
AT(45-1)-1830 Configuration Control Board Directive No. A-0131A November 12, 1969
7
I~---------'-----------------------·---~------------6· TEXT CHANGE
Remove and destroy pages iii, 1-3, 1-6, 1-7,1-12 and 1-13.
Replace with the attached pages iii, 1-3, 1-6, 1-7, 1-12 and 1-13.
Place this sheet in the front of the subject document as a change control lo~.
Redirection of Conceptual Design.
Date
Recommended by ),{., ( I,J·. i;f'd . Evaluation Board! i
Date
Approved by ~£ ~ Configuration Control Boar
Date
BNWL-500
Volume 31
4/1/69
5/13/69
6/2/69
Approved by AEC Letter, J. J. Shivley, Project Administrator, FFTF Project Office, Richland Operations Office, to E. R. Astley, FFTF Project Manager; Subject, CSDD for the Reactor Core System (No. 31), dated August 27, 1969.
ii
1.0
1.1
1.1.1
1.1.1.1
1.1.1.2
1,Ll.3
1.L1.4
1.1.1.5
1.LL6
L1.1. 7
1.1.1.8
1.1.1.9
DDCN-l 1. L 1.10
1.1.2
1.1.2.1
1.1.2.2
1.103
1.1.3,1
1.2
DDCN-l
10 2.1
1.2.1.1.1
1.2.1.1.2
1.201.1.3
1.2.1.1.4
1.2.L1.5
1.2.1.L6
1.2.1.2
1.2.1.2.1
1.2.1.2.2
1.2.1.203
1.2.1.3
BNWL-500 Volume 31
CONTENTS
INTRODUCTION
FUNCTIONS AND DESIGN REQUIREMENTS .
FUNCTIONS .
Core Subsystem.
Fuel Assemblies
Reactor Nuclear Control Components.
Reflector Assemblies
Radial Shielding Assembly .
Open Test Facilities
Closed Loop Assemblies (In-Core Portion)
Core Restraint Devices.
Special Assemblies.
Core Support Structure.
Fuel Subassembly Length
Instrument Tree and Plug Subsystem.
In-Core Instrumentation
Instrument Tree and Plug
In-Vessel Fuel Storage Subsystem
In-Vessel Storage 0
DESIGN REQUIREMENTS
Basic Design Requirements .
Core Height
Neutron Reflector .
Core Component Accessibility
Interchangeability.
Neutron Shield.
Core Subdivision
Performance
Neutron Flux
Core Power.
Advanced Cases.
Control and General Safety.
.x
· 1-1
· 1-1
· 1-1
· 1-1
· 1-1
· 1-2
· 1-2
· 1-2
· 1-2
· 1-2
· 1-3
· 1-3
· 1-3
· 1-3
· 1-3
· 1-3
· 1-4
• 1-4
· 1-4
· 1-4
· 1-4
· 1-4
· 1-5
· 1-5
· 1-5
· 1-5
· 1-6
· 1-6
· 1-6
· 1-7
o 1-7 iii
1.2.1.3.1 Nuclear Control Provisions.
1.2.1.3.2 Fuel Melting Limitation
1.2.1.3.3 Mechanical Response
1.2.1.3.4 Stability.
1.2.1.3.5 Power Coefficient.
1.2.1.3.6 Void Coefficient
1.2.1.3.7 Reactivity Insertions.
1.2.1.3.8 Limits of Propagation (Driver and Open Test)
1.2.1.3.9 Limits of Propagation (Closed Loops)
1.2.1.3.10 Short-Term Transients.
1.2.1.3.11 Radial Neutron Shield.
1.2.1.3.12 Core Support Structure.
1.2.1.4 Thermal Hydraulics.
1.2.1.4.1
1.2.1.4.2
1.2.1.4.3
1.2.1.4.4
1.2.1.4.5
1.2.1.4.6
1.2.1.4.7
1.2.1.4.8
1.2.1.4.9
Coolant
Coolant Inlet •.
Coolant Temperature
Coolant Flow Direction - Operation.
Coolant Flow Direction - Shutdown
Coolant Pressure Loss .
Heat Removal Balance
Coolant Velocity
Emergency Core Cooling System .
1.2.1.4.10 Entrapped Gases
1.2.1.4.11 Peak Primary Temperature
1.2.1.4.12 Minimum Primary Temperature
1.2.1.4.13 Flow Distribution.
1.2.1.5 Mechanics - Materials
1.2.1.5.1 Materials.
1.2.1.5.2 Sodium Exposed Surfaces
1.2.1.5.3 Contacting Surfaces
1.2.1.5.4 Fuel
1.2.1.5.5 Independent Positioning
1.2.1.5.6 Codes and Standards
BNWL-500 Volume 31
· 1-7
· 1-7
· 1-7
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· 1-8
· 1-9
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1-9
· 1-10
• 1-10
· 1-10
· 1-11
· 1-11
· 1-11
· 1-11
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1-12
1-12
· 1-12
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• 1-13
· 1-13 1-14
· 1-14
1-14
1-14
· 1-14
1-15
· 1-15
· 1-15
· 1-15
· 1-15 iv
1.2.1.5.7
1.2.1.5.8
1.2.1.5.9
1.2.1.6
1.2.1.6.1
1.2.1.6.2
1.2.1.6.3
1.2.1.6.4
Allowable Material Properties.
Support Structure Materials
Support Structure Damage
Instrumentation
Driver Assembly Instrumentation
Failed Fuel Detection and Location
Clad Failure Provisions
Minimum Test - Instrumentation -Capability
Flux Monitors.
Minimum Test Instrumentation - Type
System Integration
Support Structure History.
Open Test Instrument Interface
Testing Capability
BNWL-500 Volume 31
· 1-16
· 1-16
· 1-16
· 1-17
· 1-17
1-17
1-17
· 1-17
· 1-18
1-18
· 1-18
· 1-18
· 1-19
· 1-19
1.2.1.6.5
1.2.1.6.6
1.2.1.6.7
1.2.1.6.8
1.2.1.6.9
1.2.1.7
1.2.1.7.1
1.2.1.7.2
1.2.1.7.3
1.2.1.7.4
1.2.1.7.5
1.2.1.7.6
1.2.1.8
1.2.1.8.1
1.2.1.9
1.2.1.9.1
1.2.1.9.2
1.2.1.9.3
1. 2.2
1.2.2.1
Closed Loop Capability - Primary Function. 1-19
Closed Loop Capability . 1-19
Closed Loop Location . · 1-19
Open Test Positions - Outlet Temperature . 1-20
Axial positioners.
Open Test Position - Mechanical Design
Quality Assurance.
General
Miscellaneous.
Verification of Component Operability.
Surveillance .
Availability .
Concept Related Design Requirements
Core Related Requirements.
1.2.2.1.1 Configuration.
1.2.2.1.1.1 Core Array
1.2.2.1.1.2 Core Cross Section
· 1-20
1-20
· 1-20
1-20
· 1-20
1-21
1-21
· 1-21
1-21
· 1-21
· 1-21
· 1-22
· 1-22
v
1.2.2.1.1.3 Test Position Access.
1.2.2.1.2 Mechanics.
1.2.2.1.2.1 Total Core Radial Positioning.
1.2.2.1.2.2 Axial Positioning and Holddown
1.2.2.1.2.3 Radial Restraint.
1.2.2.1.2.4 Assembly Clearance (Operation)
1.2.2.1.2.5 Assembly Clearance (Handling).
1.2.2.1.3 Operations
1.2.2.1.3.1 Component Handling Access.
1.2.2.1.3.2 Component Positioning.
1.2.2.2
1.2.2.2.1
Instrument Tree Plug Requirements.
Configuration.
1"2.2.2.1.1 Number of Assemblies.
1.2.2.2.1.2 Test position Interface
1.2.2.2.1.3 Fuel Assembly Clearance
1,2,2.2.1.4 IVHM Interface
1.2.2.2.1.5 Seals and Shielding
1.2.2.2.2 Mechanics.
1.2.2.2.2.1 Backup Holddown Clearance.
1.2.2.2.2.2 Holddown Sensing.
1.2,2.2.2.3 DBA Containment
Operation.
BNWL-500 Volume 31
· 1-22
· 1-22
· 1-22
• 1-22
· 1-23
• 1-23
· 1-23
· 1-23
· 1-23
· 1-24
• 1-24
· 1-24
· 1-24
· 1-24
· 1-24
· 1-25
• 1-25
· 1-25
· 1-25
· 1-25
• 1-26
· 1-26
1.2.2.2.3.1 Disengagement and Removal. . 1-26
1.2.2.2.3.2 Maintenance . 1-27
1.2.2.2.3.3 In-Vessel Handling Machine Compatibility. 1-27
1.2.2.2.3.4 Rotational Compatibility. . 1-27
1.2.2.2.4 Instrumentation . 1-27
1.2.2.2.4.1 Instrument Housing
1.2.2.2.4.2 Instrument Support
1.2.2.2.4.3 Instrument Replacement
1.2.2.2.4.4 Failed Element Detection and Location (FEDAL) Connection
· 1-27
· 1-28
• 1-28
· 1-28
vi
1.2.2.2.4.5 Control/Safety Rod Instrument Probe Clearance
1.2.2.2.4.6 Peripheral Instrumentation
BNWL-500 Volume 31
· 1-28
· 1-29
1.2.2.3 In-Vessel Storage Related Requirements · 1-29 1.2.2.3.1 Configuration
1.2.2.3.1.1 Cross Section (Interface with Stored Components) .
1.2.2.3.1.2 Azimuthal Location
1.2.2.3.1.3 Fuel Storage.
1.2.2.3.1.4 Cooling Requirement.
1.2.2.3.1.5 DBA Considerations
1.202,3.2 Mechanics
1.2.2.3.2.1 Positioning.
1.2.2.3.3 Operations
1.2.2.3.3.1 Number of Positions.
1.2.2.3.3.2 Position Access and Manipulation.
1.2.2.3.3.3 Maintenance.
1.2,2.3.3.4 In-Vessel Transfer
102.2.4 Testing Capability
1.2.2.4.1 Test position Interchangeability.
1.202.4.2 Total Number of In-Core Test Positions
1.2.204.3 Test Position Location
2.0 PHYSICAL DESCRIPTION OF THE SYSTEM
201 SUMMARY DESCRIPTION.
2.2 DETAILED DESCRIPTION.
2.2.1 Fuel Assembly
2.2.2 Nuclear Control Components
2.2.3 Reflector Assemblies.
2.2.4 Radial Shielding Assembly
2.2.5 Open Test Positions.
2.2.6 Closed Loop Assembly.
2.2.7 Core Restraint
· 1-29
· 1-29
· 1-29
· 1-30
· 1-30
· 1-30
· 1-30
· 1-30
· 1-30
· 1-31
· 1-31
· 1-31
· 1-31
· 1-31
· 1-32
· 1-32
· 1-32
· 2-1
· 2-1
· 2-14
· 2-14
· 2-22
· 2-27
· 2-29
· 2-31
· 2-32
· 2-32
vii
2.2.8
2.2.9
2,2.9.1
2.2.9.2
2.2.9.3
2.2.1Q
2.2.10.1
2.2.10.2
2.2.10.3
2.2.10.4
2.2.10.5
2.2.11
2.2.12
3.0
3.1
3.2
4.0
4.1
4.2
4.3
4.4
4.5
5.0
Special Assemblies
Core Support Structure
Inlet Plenum.
Ring Girder .
Core Barrel .
In-Core Instrumentation •
Fuel Assembly Instrumentation
Control/Safety Rod Instrumentation
Open Test Instrumentation
Closed Loop Instrumentation .
In-Core Flux Monitor Positions
Instrument Tree and Plug.
In-Vessel Storage
SAFETY CONSIDERATIONS
HAZARDS .
PRECAUTIONS 0
PRINCIPLES OF OPERATION •
STARTUP •
NORMAL OPERATION.
SHUTDOWN.
SPECIAL OR INFREQUENT OPERATIONS.
EMERGENCY
MAINTENANCE PRINCIPLES
APPENDIX A - References
APPENDIX B - Support Information Requirements.
APPENDIX C - Interfaces
APPENDIX D - FFTF Design Data Summary.
APPENDIX E - Alternate Core Designs
APPENDIX F - Drawings.
BNWL-500 Volume 31
• 2-42
• 2-42
• 2-43
• 2-44
• 2-45
• 2-46
• 2-46
• 2-47
· 2-47
· 2-47
· 2-48
• 2-48
• 2-52
• 3-1
• 3-1
• 3~3
• 4-1
• 4-1
• 4-2
· 4-3
• 4-3
• 4-4
• 5-1
A-I
B-1
C-l
D-l
E-l
F-l
viii
2.1
2.2
2.3
2.4
2.5
FIGURES
Reference Core Map
Arrangement of Test Facilities
Central Integrated Flux Spectrum
Normalized Radial Power Density and Flux Profiles at the Axial Mid-Plane
Normalized Axial Power Density and Flux Profiles at Radial Center
2.6 Fuel Assembly Size Versus Fuel Pin OD for 127, 169, and 217 Pins Per Assembly and Peak Fuel Assembly and Core 6T's of 250, 300, 350
2.7
2.8
and 400 of
Bases for Fuel Assembly Size Selection
Number of Fuel Assemblies Required Versus Core Height
BNWL-500 Volume 31
2-2
2-4
2-10
2-11
2-12
2-15
2-16
2-18
2-19 2.9
2.10
Peak Total Flux Versus Core Volume
Influence of Pin Bundle to Duct Clearance, Wall Thickness, and Duct-to-Duct Clearance Peak Total Flux and U/Fissile Pu Ratio
Duct
2.11
2.12
2.13
2.14
2.15
2.16
2.17
E.l
E.l
Core Volume Versus Fuel Pin OD for Various Core 6T's, Design Based on Use of 3 In-Core Safety Rods and Full Peripheral ShimRegulating Control
Fuel Pin Parameters (3 In-Core Rods)
Fuel Design Parameters (9 In-Core Rods)
Initial FTR Core Thermal Deflection Curve Above and Below Core Support (5th Row)
Initial FTR Core Assembly Swelling Deflection (5th Row) Support Pads Above and Below the Core
Radial Restraint Loading Sequence
Initial Core FTR Radial Restraint - (50 0 T Across Duct)
Alternate Core Map
Alternate Core Layout
on 2-21
2-23
2-24
2-25
2-36
2-38
2-39
2-41
E-2
E-3
ix
REACTOR CORE SYSTEM
INTRODUCTION
BNWL-500 Volume 31
This document defines and presents specific core design require
ments; describes the selected reference concept;l indicates
unresolved problems and the proposed plan of action to resolve
them; and identifies design constraints and freedom of choice
for preliminary reactor design constraints and freedom of choice
for preliminary reactor design by the Reactor Plant Designer.
Supporting documentation containing comprehensive design guides
and criteria have been developed and provide part of the technical
bases for the reactor core design.
These documents include:
FTR Reference Nuclear Parameters and Parametric Studies 2
3 Design Safety Criteria for Reactor Core
f t . . 4 Bases or Reac or Core Deslgn Requlrements
FTR Fuel and Core Parametric Studics 5
Separate Desig"n Descriptions have been completed for the First
Core Fuel Assembly Component, Reactor Nuclear Control
Components, Reactor Refueling System, including the In-Vessel
Handling l-..f.achine (IVHM) and Closed Loop System.
1. Refer to References, Appendix A, Item 19. 2. Refer to References, Appendix A, Item 2. 3 . Refer to References, Appe.ndix A, Item 3, Sec·cion 12. 4. Refer "to References, Appendix A, Item 5. 5. Refer to References, Appendix A, Item 1
~.
x
BNWL-500 Volume 31
Section 1.0 of this Design Description is "baseline" data;
the remainder is reference design information. Several
of the requirements of Section 1.0 are presented in the form
of performance ranges, e.g., 350 - 400 Mw. In such cases,
the reference core description in Section 2.0 should be
consulted for nominal design values.
The contents of this document support and expand the
requirements established in the Overall Conceptual Systems , 't' 1 Deslgn Descrlp lone
1. Refer to References, Appendix A, Item 16.
xi
BNWL-500 Volume 31
SECTION 1.0 FUNCTIONS AND DESIGN REQUIREMENTS
1.1 FUNCTIONS
The function of the Reactor Core System is to provide a testing
environment and space for testing adequate to fulfill the FFTF pro
gram objectives. l The Reactor Core System, as defined by this
CSDD, consists of all removable assemblies within the reactor
vessel with the exception of the in-vessel handling machine
(IVHM) , the vessel liner, and the closed loop and short-term
test facilities. The major components of the reactor core
assembly, and their basic functions are further categorized (to
facilitate division of functions between the major components
of the core). These subsysterns and their functions are:
1.1.1 Core Subsystem
Includes the components associated with the neutronic perfor
mance of the core.
1.1.1.1 Fuel Assemblies
Functions are described in the Conceptual Component Design
Description for the First Core Fuel Assembly.
1.1.1.2 Reactor Nuclear Control Components
Functions are described in the Conceptual Component Design
Description for the Reactor Nuclear Control Component.
1. Refer to References, Appendix A, Item 12.
1-1
BNWL-SOO Volume 31
1.1.1.3 Reflector Assemblies
To minimize neutron leakage and to maintain flux in the core
at the high energy levels desired for fast neutron irradiation
experiments.
1.1.1.4 Radial Shielding Assembly
To reduce the total fluence and thereby limit the associated
material degradation of the reactor vessel, thermal liner,
core support structures.
1.1.1.5 Open Test Facilities
To provide instrumented testing capability at reactor
coolant conditions.
1.1.1.6 Closed Loop Assemblies· (In-Core Portion)
Functions are described in the Conceptual System Design
Description for the Closed Loop System.
1.1.1.7 Core Restraint Devices
To restrain motion of the core such that response to steady
state an6 transient conditions is both predictable and safe.
1-2
1.1.1 0 8 Special Assemblies
BNWL-500 Volume 31
To accommodate specialized equipment that may later be
designed and installed, e.g., special neutron sources,
special driver fuel assemblies to substitute in place
of the open or closed loop test assemblies, etc.
1.1.1.9 Core Support Structure
To provide a support system which allows predictable
core component response to steady-state, transient, and
abnormal loading conditions. Included is the removable
portion of the inlet plenum, the core barrel, and ring
girder support structure.
DDCN-l 1.1.1.10 Fuel Subassembly Length
The length of the fuel assembly shall not exceed 12 ft.
1.1.2 Instrument Tree and Plug Subsystem
Includes those components making up the instrument tree
and the associated drive mechanisms.
1.1.201 In-Core Instrumentation
Functions are described in the Conceptual System Design
Description for the Reactor and Vessel Instrument
System.
1.1.2.2 Instrument Tree and Plug
To provide the support structure for the driver fuel and
control/safety rod instrumentation, to provide backup
holddown for the core, and to provide rotary and axial
motion to: (1) remove the tree from the core region,
and (2) in conjunction with the IVHM, perform handling
operations. 1-3
BNWL-500 Volume 31
1.1.3 In-Vessel Fuel Storage Subsystem
Includes the components pertaining to in-vessel storage of
driver fuel and other core cor::lponents.
1.1.3.1 In-Vessel Storage
To provide interim storage for new and spent fuel assemblies
and other core components.
1.2 DESIGN REQUIREMENTS
The reactor core design requirements have been classified as
"Basic," and "Concept Related." Concept related requirements
for the core components, in-vessel storage, and the instrument
tree have been segregated under the appropriate heading.
Additional descriptive comments and basis for each requirement
are contained in two supporting documents. l ,2
1.2.1 Basic Design Requirements
1.2.1.1 Core Arrangement
1.2.1.1.1 Core Heightl
The height of all driver assemblies shall be equal and the
reactor core active zone shall be greater than 32 inches.
1.2.1.1.2 1 Neutron Reflector
The driver core shall be reflected neutronically in the radial
and axial directions.
1. Refer to References, Appendix A, Item 5. 2. Refer to References, Appendix A, Item 3, Section 12.
1-4
BNl'JL- 5 fJ 0 Volume 31
1.2.1.1.3 Core C9mponent Accessibilityl
:Reactor GorE! components shall be designed for access ibili ty
to facilitate operation, maintenance, surveillance and in
place testing, and replaceme~t. These features shall be
consi stent with the irradiated core corr.ponent handling
requirements dnd plant availability goals.
1.2.1.1.4 Inter~hanqeabilityl
All in-core assemblies shall conform to a uniform core lattice
dimension and geometry.
1.2.J .1.5 Neutron Shield
Neutron shield in; will be provided within the reactor vessel
and will be designed to li~it the total dose of damaging
radiation on the vessel wall to assure end-of-life properties
that exceed minimum requirements. 3 (DSC 5.1)2
Neutron shielding design shall also assure end of life
properties exceeding minimum requirements4
for other critical
in-core components including the grid-plate, core barrel,
and vessel thermal liner.
1.2.1.1.6 Core Subdivision
The subdivision of the core will assure that the potential
reactivity insertio~ rate and power transient from meltdown
and S'ravity collapse of anyone driver assembly ~·,rill be
1. Refer to References, Appendix A, Item 5,. 2. Refer to References, Appendix A, Item 3, Section 12. 3 . Refer to Support Information Requirements, Appendix B, .
Item 24. 4. Refer to Support Information P.equirernents, J\ppendix B,
Item 15. 1-5
DDCN-l
BNWL-500 Volume 31
terminated by normal operation of the Plant Protective
System with no more than "moderate damage"l to other
driver fuel assemblies in the core resulting from the
over-power transient condition. (DSC 2.1)2
1.2.1.2
1.2.1.2.1
Performance
3 Neutron Flux
Flux (~) available for closed loop testing shall be the
highest attainable, consistent with testing allocations,
reactor safeguards, driver fuel and total core power
limitations. First generation cores shall provide a
total peak neutron flux, averaged over the reactor opera
tion cycle greater than 7 x 1015 n/cm2-sec to at least
one closed loop positiono The design goal for advanced 16 2 cores is a total peak flux of 1.3 x 10 n/cm -sec.
1 Core Power
The core power level shall maximize flux while not
exceeding the steady state design power level of the
DDCN-l heat transport system. First generation core power
shall be 400 MWt at t. = 600 of and delta-t = 300 °Fo ln The design shall consider operation with t. = 500 of ln but at reduced power.
DDCN-l 102.1.2.3 Advanced Cores
The design shall provide the capability to accommodate
reactor modifications necessary for advanced cores of
1. Refer to References, Appendix A, Item 3, Section 1. 2. Refer to References, Appendix A, Item 3, Section 12. 3. Refer to References, Appendix A, Item 5.
1-6
BNWL-SOO Volume 31
higher flux and power rating. These modifications shall be
performed with minimized expendability of existing core
components.
1.2.1.3 Control and General Safety
1.2.1.3.1 Nuclear Control provisions l
The core shall include provisions for accepting the control
and safety rods. Sufficient positions shall be included to
assure that reactivity margins are available to meet all
operating, shutdown, and accident requirements for initial
and advanced cores. 2
102.1.302 Fuel Melting Limitationl
The fuel temperature shall not reach incipient melting during
steady-state or transient overpower conditions resulting
from "operational incidents.,,3
The mechanical response of the core assembly to steady-state
and transient operating conditions shall result in predictable
safe reactivity changes_ Mechanical response analysis shall
include the effects of thermal expansion and creep, radiation
induced transient and steady-state creep, coolant pressure,
dead and live mechanical loadings. These effects may be
applied separately and together as a function of time to the
1, Refer to References f Appendix A, Item 5. 20 Refer to References, Appendix A, Item 8. 3. Refer to References, Appendix A, Item 3, Section 1.
1-7
BNWL-500 Volume 31
core and reactor structures as dynamic'response systems.
Mechanical response shall include 'consideration of both
mechanical and thermal inertial effects under time-dependent
loadings of random and periodic character.
1.2.1.3.4 Stability
The FTR will be designed to assure inherent stability with
respect to spatial power distribution and total power level
throughout the operating range. (DSC 1.1)1
1.2.1.3.5 Power Coefficient
The FTR will be designed to provide both a negative prompt
and a negative overall power coefficient of reactivity. The
magnitude will be such that:
A. Inherent stability of the reactor is assured for all
operational transient and steady-state conditions.
B. Power transients initiated by accidental reactivity
insertions from any "minor accident,,2 condition can be
terminated by normal safety system action without
exceeding "moderate damage,,2 to the fuel. "Operational
incidents,,2 will be terminated by normal safety system
action with "no damage" to the fuel.
C. The energy release of any "disruptive accident,,2 will
be within the coordinated design bases of the core and
containment systems. (DSC 1.2)1
1. Refer to References, Appendix A, Item 3, Section 12. 2. Refer to References, Appendix A, Item 3, Section 1.
1-8
BNWL-500 Volume 31
1.2.1.3.6 Void Coefficient
The sodium void positive reactivity worth throughout the core
will be limited by design such that:
A. Spatial voiding of any single channel as a result of
loss of coolant flow or meltdown of the fuel assembly
without protective action will not result in damage
severity greater than a "Major Accident. "1
B. Voiding by molten fuel-sodium interactions during a
severe nuclear power transient leading to core disas
sembly will not result in an energy release greater than
the design basis of the containment system.
C. The voiding of any region or regions of the core which is
possible with the thermal, hydraulic, and neutronic
characteristics of the reactor and physical constraints
concurrent with failure of the core protective system will
not result in an energy release greater than the design
basis of the containment systemo (DSC 1.3)2
1,20103c7 Reactivity Insertions
Potential reactivity insertions due to fault conditions will be
limited to design in either rate or magnitude to assure that
any damage attributed to that fault is no greater than the
limit as defined in the Criteria for Accident Severity Levels. l
(DSC 1.4)2
lo2el,3.8 Limits of Propagation (Driver and Open Test)
Driver fuel assemblies and open test facilities will be designed
to 11mit the potential for propagation of the effects of local
disturbances. It is a design objective that disturbances
10 Refer to References, Appendix A, Item 3, Section 1. 2. Refer to References, Appendix A, Item 3, Section 120
1-9
BNWL-500 Volume 31
initiated. by some local fault condition be effectively isolated
to the extent that the ability to shut the reactor down is
not impaired, and the damage cannot progress mechanically or
hydraulically beyond the initially affected assembly. (DSC 2.2)1
Closed loop tubes will be designed to limit the potential
for propagation of the effects of local disturbances to
adjacent assemblies. Failure of a closed loop test
assembly will be effectively isolated to the extent that
the ability to shut the reactor down is not. impaired.
~amage within a closed loop tube from a condition resulting
in mUltiple pin failures vdthin that tube will be such
that core damage external to the closed l.oop is limited
to the severity of an "operational incident.,,2 (DSC 2.3)1
1.2.1.3.10 Short-Term Transients
The core will be capable of sustaining anticipaten short-time
transient imbalances which may occur upon "operational
incidents,,2 such as loss of primary pumping power or scram,
with no core damage. (DSC 3.1)1
1.2.1.3.11 Radial Nuetron Shield (DSC 5.3)1
The internal shielding will be designed to assure that any
single "unlikely fault,,2 affecting position or function, e.g.,
1. Refer to References, Appendix A, Item 3, Section 12. 2. Refer to References, Appendix A, Item 3, Section 1.
1 ... 10
BNWL-SOO Volume 31
any single structural failure, will result in a predictable
effect (as rel~ted to core reactivity, coolant flow, materials
interactions) to limit the severity of that fault to a "minor
accident. "1 (DSC 5.3)2
1.2.1.3.12 Core Support Structure
Core reactivity changes under normal operational or abnormal
conditions resulting from thermally, hydraulically, and mechanj
cally induced deflections of the core support structure will be
limi ted to assure conformance to the design sa.fety criteria ::or
reactivity coefficients. (DSC 2.8)2
1.2.1.4
1.2.1.4.1
Thermal Hydraul_~~.
3 Coolant
The core coolant shall be sodium.
1.2.1.4.2 3 Coolant Inlet
The flow inlet regions to the core inlet plenums and fuel
assemblies will be designed to prevent passage into these
regions of any foreign matter, deposited in the primary system
by any single "u.nlikely fault"l which would be capable of
leading to accidents, through flow blockage, whose magnitude
exceeds that of a "minor accident."l (DSC 2.6)2
Control safety rod and open test units shall also incorporate
features to meet this requirement.
1. 2. 3.
Refer to References, Appendix A, Item 3, Section 1. Refer to References, Appendix A, Item 3, Section 12. Refer to References, Appendix A, Item 5.
1-11
1.2.1.4.3 1 Coolant Temperature
BNWiL-500 Volume 31
The core coolant design temperatures shall be as follows:
Design Temperature
(Reactor Capability)
Average Coolant Inlet Temperature
Average Core Coolant Outlet Temperature
Average Driver Fuel Coolant Temperature Rise (~T)
500-900 of
800-1200 of
300-400 of
1.2.1.4.4 Coolant Flow Direction - Operation1
Coolant flow through the core shall be upward.
1.2.1.4.5 Coolant Flow Direction .... Shutdown
Design Temperature (First Generation
Reactor Core)
500-600 of
800-900 of
300-350 of
(DSC 3.3)2
Coolant flow through the fuel assemblies during shutdown or
emergency cooling conditions will be in the same direction as
normal flow during power operation. (DSC 3.4)2
1.2.1.4.6 Coolant Pressure Loss1
The coolant pressure loss increment (which is primarily dependent
upon driver fuel design) when combined with the remainder of the
DDCN-l primary system shall be within the 500 ft11.ead:c.a.pabi1ity of existing
sodium pump designs for flows in the FTR range.
1. Refer to References, Appendix A, Item 5. 2. Refer to References, Appendix A, Item 3, Section 12.
1-12
1.2.1.4.7 Heat Removal Balance
BNWL-500 Volume 31
The thermal-hydraulic design of the core will be such that,
in conjunction wit.h the heat removal and control systems, a
stable equilibrium between heat generation and heat removal
does exist and can be maintained throughout the normal operat
ing range of the FTR. (DSC 3.2)1
1. 2 .. 1. 4.8 Coolant Veloci -t:.y'2
Coolant velocity in the core shall not cause excessive or
unpredictable damage to core components.
Mean velocity in any channel or major subchannel such as the
fuel bundle shall not exceed 30 feet per second. Coolant
DDCN-l velocities in other regions (such as channels in dynamic
seals) in excess of 30 ft per second shall be considered
on an individual basis.
1.2.1.4.9 Emergency Core Cooling System
The core design will be coordinated with emergency core
cooling system so that continuity of reactor core cooling is main
tained for the intact core even in the unlikely event of a
"major fault.,,3 (DSC 3.5)1
1.2.1.4.10 Entrapped Gases
The core design will be such that any entrapped gases will not
lead to a degree of vapor blanketing which causes damage
greater than that of a "minor accident.,,3 (DSC 3.6)1
1. Refer to References, Appendix A, Item 3, Section 12. 2. Refer to References, Appendix A, Item 5. 3. Refer to References, Appendix A, Item 3, Section 1. 1-13
1.2.1.4.11 Peak Primary Temperature
BN~rr.-· 5 0 0 Volume 31
The peak primary coolant temperature during transient
conditions corresponding to a major accident, will be
limited to a value so as not to exceed either: (a) the
sodium boiling temperature, or (b) that temperature at which
the corresponding fuel cladding temperature will allow
excessive cladding strains l to occur. (DSC 3.7)2
1.2.1.4.12 Minimum Primary Temperature
The minimum primary coolant temperature will be greater than the
temperature at which plugging of coolant flow paths can occur.
l. 2.1. 4 .13 1 ' '1.-. t' 3 F O".v Dlst.rLJU lon (DSC 3.8)2
Features which affect the distribution of flow, such as inlet
regions and orifices (if required), shall be designed to assure
that mismatch in outlet temperatures are within acceptable
" 't 4 _lrnl s.
1.2.1.5 Mechanics - Materials
1.2.1.5.1 Materials
Materials used in the reactor core components and radial neutron
shield, will be limited such that exposure of that material to
other core materia.ls., e. g., sodium, under normal and abnormal
conditions will not initiate fault conditions or increase the
severity of imposed fault conditions beyond the limits corres
pondin'J to the fault conditions as defined in the Criteria for
Accident Severity - FFTF Reactor systems. 3 (DSC 2.5 & 5.2)2
1. Fuel cladding temperature and allowable cladding strains are to be determined.
2. Refer to References, Appendix A, Item 3, Section 12. 3. Refer to References, Appendix A, Item 3, Section 1. 4. The degree of acceptable mismatch is to be determined ..
1-14
2.1.5.2 1
Sodium Exposed Surfaces
BNWL-500 Volume 31
Core component surfaces in contact with sodium shall be fabri
cated of austenitic stainless steel. Exceptions, e.g., hard
faced surfaces, may be required but such exceptions must be
individually approved.
1.2.1.5.3 Contacting surfaces l
Design provisions shall be included to protect contacting
surfaces in the core against damaging interactions. Raised pads
shall be utilized for contacting surfaces between in-core ducts.
1.2.1.5.4 1
Fuel
The initial driver fuel for the FTR shall be mixed plutonium and
uranium oxides.
1.2.1.5.5 Independent Positioning
Independent features for positioning of the core components
(driver fuel assemblies, test assemblies, control assemblies)
",nIL be provided such that any single "unlikely fault" 3 affecting
the positioning function; e.g., failure of hydraulic balance
features, any single structural failure, will result in a pre
dlctable react.ivity effect due to component movements to limit
the severity of that fault to a "minor accident."2
(DSC 2.4)3
1.2.1.5.6 Codes and Standardsl
Core design shall conform where applicable to appropriate AEC
standards and specifications. The design shall also fulfill the
1. Refer to References, Appendix A, Item 5. 2. Refer to References, Appendix A, Item 3, Section 1. 3. Refer to References, Appendix A, Item 3, Section 12.
1-15
BNWL-500 Volume 31
:Lntent of the ASME Pressure Vessel Code Section III and Code
Case 1331-4 for similar temperature regimes.
1.2.1.5.7 Allowable Material Properties 1
The core support structure will be designed to establish
allowable stresses which are based upon specified temperature-
dependent mechanical properties of structural materials. .2 (DSC 2.10)
The effect of other environmental conditions such as fluence and
sodium exposure shall be included in determining the allowable
value of material properties. 3
1.2.1.5.8 Support Structure Materials
Materials used in the core support structure will be chosen to
assure compatibility with the environmental conditions and to
maintain satisfactory end-of-life properties to fulfill their
intended funciton over the range of design conditions and life
time requirements. (DSC 2.11)2
1.2.1.5.9 Support Structure Damage
The core support structure will be designed to sustain no
damage under "minor accident ll4 conditions, and to limit core
movements under "major accident" ll conditions e.g., the Design
Basis Earthquake, to prevent reactivity additions by fuel
movement which will potentially lead to significant additional
fuel damage. (DSC 2.9}2
1. Refer to References, Appendix A, Item 5. 2. Refer to References, Appendix A, Item 3, Section 12. 3. Refer to References, Appendix A, Item 4. 4. Refer to References, Appendix A, Item 3, Section 1. 1-16
~.2.1.6 Instrumentation
1.2.1.6.1 Driver Assembly Instrumentationl
BNWL-500 Volume 31
Each driver fuel assembly shall be instrumented to determine
bulk outlet temperature and bulk "flow. Accuracy of each flow
meter shall be such as to confirm that adequate heat removal
capability is available to ascend from shutdown to normal
power operation without incident.
1.2.1.6.2 Failed Fuel Detection and Locationl
Instrumentation shall be provided for each driver fuel position
to detect fission product release and to locate the affected
assembly.
1.2.1.6.3 Clad Failure provisions l
Provisions shall be included in the core design to allow opera-
tion with clad failures within appropriate operational limitations.2
1.2.1.6.4 Minimum Test - Instrumentation - capabilityl
A minimum of bvo open test positions and all closed loops shall
be capable of accommodating special test instrumentation in
contact with test specimens. 3
1. Refer to References, Appendix A, Item 5. 2. The specific operational limits which can be tolerated are
to be determined. 3. Refer to References, Appendix A, Item 13.
1-17
1 6 t:; 1 . 1 1.2._ .. _ F.ux Monltors
BNWL-500 Volume 31
A minimum of three in-vessel flux monitor posi·tions shall be
provided on the periphery 0f the core for acco:rrur.odat.ing shut
down flux rr.onitors.
1.2.1.6.6
Open test positions and closed loops, at a minimum, shall be
instrll.111ented for inlet2
and outlet coolant temperatures,
coolant flow and ft'.el failure detection and location.
1.2.1.6.7 System Integration
The core design will be coordinated with the Plant Protection
System to incorporate the Protective Subsystems required to
monitor, control, initiate and carry out the protective
actions required. 4 (DSe 4.1)3
1.2.1.6.8 1
Support Structure H}stort--
.Heans ~7ill be provided to determine the thermal and mechanical
load conditions of the core support structure to assure con-
tinued operation within the design ranges. (DSe 4.2)3
---.--1. Refer to References, Appendix A, Item 5. 2. Bulk reactor inlet temperature for open test positions. 3. Refer to References, Appendix A, Item 3, Section 12. 4. Refer to References l Appendix A, Item 3, Section 31. 5. Refer to Reference~, Appendix A, Item 21.
1-18
Bm-vI,-500 Volume 31
As a minimum, the design shall include provisions for maintain
ing a history of pressure and temperature environments to
which the primary support structures are exposed.
1.2.1.6.9 Open Test Instrument Interface
Provisions for connection of open test instrumentation to the
out-of-reactor test cabling shall be provided at each reactor
cover penetration where an open test assembly may be potentially
located. 1
1.2.1.7 Testing Capability
1.2.1.7.1 Closed Loop Capability - primary Functio~2
The core shall provide the capability to accept closed loops
for testing of fuels and materials in dyna.mic sodium coolant.
Closed loop test modes shall range up to and include failure
short of planned test meltdown.
1. 2.1. 7.2 Closed LOC:>E Capabili ty2
The core shall provide the capability for the installation of
at least six closed loops.
1.2.1.7.3 Closed Loop Location2
A closed loop installation position shall be provided near
the peak flux region of the core.
1. The specific number and type of connections for each open test is TBD.
2. Refer to References, Appendix A, Item 5.
1-19
BNWL-500 Volume 31
1.2.1.7.4 Open Test Positions - Outlet Temperature 1
Outlet temperatures in open test positions shall be limited 2 to avoid excess thermal stresses 'in adjacent core components.
1.2.1.7.5 Axial Positioners l
The core shall be capable of accommodating experiments
utilizing axial positioners in at least one contact
instrumented, open test position.
1.2.1.7.6 Open Test Position - Mechanical Designl
The mechanical design of the open test position shall assure
safe and predictable displacement of the experiment during
thermal and hydraulic transients.
1.2.1.8 Quality Assurance
1.2.1.8.1 1 General
This system shall satisfy the criteria establishing quality
which will be defined and documented during design. These
criteria shall cover the following areas: 3
Design
Fabrication and Construction
Operation
Maintainability.
1.2.1.9 Miscellaneous
1. Refer to References, Appendix A, Item 5. 2. The degree of acceptable mismatch is to be determined. 3" Refer to Support Information Requirements, Appendix B,
Item 18. 1-20
BNWL-500 Volume 31
1.2.1.9.1 Verification of Component operabil~!Zl
Means shall be provided to monitor position indication of
control and safety rods, and actuation and operability of
core radial and axial restraint.
1.2.1.9.2 Surveillance
The design will accommodate an integrated program of sur
veillance and in-service testing, the function of which will
be to ensure continual safe operation of all structural mate-
rials and components subjected to high neutron fluences. 2 3 (DSC 6.1)
1.2.1.9.3 Availabilityl
The reactor core system shall be designed for reliability
and maintainability consistent with achieving an FFTF overall
plant availability goal of 75 percent. The actual increment
of the overall goal assigned to the reactor core is to be
determined. 4
1.2.2 Concept Related Design Requirements 5
1.2.2.1 Core Related Requirements
1.2.2.1.1 Configuration
1. Refer to References, Appendix A, Item 5. 2. Refer to Support Information Requirements, Appendix B,
Item 14. (Specifics regarding Surveillance are to be determined. )
3. Refer to References, Appendix A, Item 3, Section 12. 4. Refer to Support Information Requirements, Appendix B,
Item 10. 5. Refer to References, Appendix A, Item 5, for bases for all
concept related requirements. 1-21
1.2.2.1.1.1 Core Array
BNWL-500 Volume 31
The core components shall be arranged in a vertical array.
1.2.2.1.1.2 Core Cross Section
The core components shall be arranged on an equilateral triangu
lar pitch. Provision for test positions and in-core controls
shall be provided at various fixed radial positions. (See
1.2.1.1.3)
1.2.2.1.1.3 Test Position Access
Space for closed loop and open test position connectors and
piping and access for the instrument probes shall be provided
by placing these test positions-along three radial corridors.
1.2.2.1.2 Mechanics
1.2.2.1.2.1 Total Core-Radial-positioning
The core assembly out to and including the shield assemblies
shall be radially positioned by an external support structure
co.re barrel.
1.2.2.1.2.2 Axial Positioning -and Holddo'\ATn_
Primary holddown for all assemblies except closed loops out to
and including the second reflector row shall utilize a combina
tion of hydraulic balance and assembly weight; backup holddown
for each of these assemblies shall be afforded by the instru
ment tree holddown plate. 1-22
B~WL-500
Volume 31
1.2.2.1.2.3 Radial Restraint
In addition to the radial positioning structure noted in
1.2.2.1.2.2, the design shall include a radial restraint
system which, by a combination of initial positioning to
assure proper operational confisuration and a compliant
loading system, shall result in predictable reactivity changes
during all operating and transient conditioning and shall allow
handling, utilizing normal procedures. The design of the
restraint system shall also insure that in the event of a given
accident severity, the resultant damage shall not be increased
by further radial core motion.
1.2.2.1.2.4 Assen~ly Clearance (Operation)
Adequate clearance shall be provided between in-core assemblies
(with the exception of contacting surfaces) to accommodate a
cornbination of tolerance accumulation, radiation-induced
swelling, and creep and thermal deformations.
1.2.2.1.2.5 Assembly Clearance (Handling)
Clearance shall be provided to significantly reduce the
probability of galling or scratching at contacting surfaces
during assembly withdrawal or insertion, during operation.
1.2.2.1.3 Operations
1.2.2.1.3.1 Component Handling Access
All positions in the active core and reflector rows shall be
accessible utilizing the in-vessel handling machines. Core
1-23
BNWL-SOO Volume 31
component design in conjunction with Ivm·l design shall negate
the need for special grappling fixtures.
1.2.2.1.3.2 Component Positioning
Features shall be incorporated to assure placemen-t of components
in the designatedl
lattice position during a normal outage.
The versatility to intentionally rearrange the core map through
design of the more permanent structures, i.e., tubesheet, shall
be maintained. (See 1.2.1.1.4.)
1.2.2.2 Instrument Tree Plug Requirements
1.2.2.2.1 Configuration
1.2.2.2.1.1 Number of Assemblies
A total of three instrument tree assemblies (one for each
trisected core segment) shall be utilized to provide a support
structure for instrumentation to all driver fuel positions.
1.2.2.2.1.2 Test Position Interface
Test position support shall be separated from the tree. Test
position and instrument tree movements shall be independent
of each other.
1.2.2.2.1.3 Fuel Assembly Clearance
During refueling operations the instrument tree assembly shall
be positioned so that the instrument tree does not interfere
with IVHM access to core components.
1. Refer to Interfaces, Appendix C, Item 6. 1-24
1.2.2.2.1.4 IVHM Interface
BNWL-500 Volume 31
If required, the plug shall include an opening for mounting
the IVHM. The opening shall include provisions for meeting
requirements 1.2.2.2.2.3 and 1.2.2.2.1.5.
1.2.2.2.1.5 Seals and Shielding
The plug shall incorporate motions required for refueling
and maintenance. Seals, steps, and shielding shall be
incorporated at all penetrations to allow manned access to the
operating deck during reactor operation.
1.2.2.2.2 Mechanics
1.2.2.2.2.1 Backup Holddown Clearance
Sufficient clearance shall be provided for these core
components to accommodate anticipated thermal and radiation
induced expansion, as well as tolerance accumulation prior
to engagement of the backup.
1.2.2.2.2.2 Holddown Sensing
In addition to restricting axial motion of the core components,
in the event of hydraulic balance failure, designs of the instru
ment tree shall consider the optimum tradeoff of structural,
neutronic, and load sensing functions. The following items
should be considered in the tradeoff:
If practical, total holddovm of all components serviced
should be accomplished.
1-25
BNVJL-500 Volurl1e 31
React i vi ty in3e:ct-io::1s due to liftt-,d cO!np'::m'allts falling back
into t:'le core ::;hould be li(1lited to a maximum of 0.6$ (minor 1
accident level) .
Hydraulic failuce indication should be investigated for
operati.onal indication of potential pro~lems.
1.2.2.2.2.3 DBA Co~tainment
Upon installation in the reactor vessel cover, the plug shall
be cdpable of meeting structural requirerlents of the cover.2
The instrument tree shall provide necessary motion in order to
disengage the instrument probe from the fuel assembly outlets
and clear the area above the reactor core for fuel handling.
To assure safe conditions exist duri:1g probe removal, the
following features shall be inherent in operation of the tree:
Control/Safety Rod Check: During axial withdrawal of the
instrument tree, each control/safe-::'y rod position shall
be checked to assure that the i!!strune::1t probe elllU control
rod duci.: core disengaged. ~ll>xial movement at the time thi~;
check is perfcrmed shall not result in a reactivity increase
in excess of the minor accident levell
in the event that
all of the r0ds attached to the tree were lifted with the
tree.
1. Refer to References, Appendix A, Item 3, Section 1. 2. Refer to Re:feren:.!ei~, Appendix A i Item 7.
1-26
BNWL-500 Volume 31
Stripper: To preclude lifting of the fuel and control rod
assemblies with the tree, an axial load between the duct
and instrument probe shall be applied during axial motion
of the tree.
Final Disengagement Check: Prior to lateral motion of the
instrument tree, a final check shall be performed at all
lattice positions serviced by the tree to assure total
disengagement at all positions.
1.2.2.2.3.2 Maintenance
The complete instrument tree/plug assembly (including the IVHM,
if incorporated in this plug) shall be removable as a unit for
maintenance procedures.
1.2.2.2.3.3 In-Vessel Handling'Machine Compatibility
The plug shall include features for accepting and mounting
the in-vessel handling machine, if required. The plug shall
operate in conjunction with the IVHM assembly to position the
handling grapple during transfer operations p if required.
1.2.2.2.3.4 Rotational Compatibility
Plug rotation and location shall be compatible with both
instrument tree and IVHM operations.
1.2.2.2.4 Instrumentation
1.2.2.2.4.1 Instrument Housing
Instrumentation sufficient to meet fuel assembly instrument
requirements (See 1.2.1.6.1 and 1.2.1.6.2.) shall be housed
in the instrument probe section of the instrument tree. 1-27
Bm1L-500 Volume 31
1.2.2.2.4.2 Instrument Support
The instrument probe shall provide structural support for the
instrument assembly and FEDAL components housed within the
probe. The instrument tree shall provide support for the
instrument leads.
1.2.2.2.4.3 Instrument Replacement
1 The thermocouples and flo\~eter for each fuel assembly shall be
removable from the ins"trument tree without removing the tree
or plug from the reactor.
1.2.2.2.4.4 Failed Element Detection and Location (FEDAL)
Connection
An under-sodium joint shall be provided between the instrument
tree portion of the FEDAL lines and the FEDAL in-vessel
transition section.
1.2.2.2.4.5 Control/Safety Rod Instrument Probe Clearance
Adequate clearance and guidance shall be provided for the control/
safety rod instrument probe, hanger rod, and the control/safety
rod duct to prevent rod binding. Guidance provisions
shall be provided for the anticipated maximum nunilier of 2 control/safety rods.
1. Refer to Drawings, Appendix F, SK-3-l2896. 2. Refer to References, Appendix A, Item 8.
1-28
1.2.2.2.4.6 Instrument Connections
BNlvL-500 Volume 31
Instrument connections at the plug shall be designed for ease
of mating with the connecting out-of-reactor cabling.
1.2.2.2.4.7 Peripheral Instrumentation
In addition to servicing assemblies to meet the subdivision
requirements of 1.2.1.1.6, the instrument tree shall include
provisions for accommodating up to 15 additional instrumented
driver assemblies spaced about the periphery of the core.
1.2.2.3 In-Vessel Storage Related Req"ll:irements
1.2.2.3.1 Configuration
1.2.2.3.1.1 Cross Section (Interface with Stored Components)
The in-vessel storage containers shall be capable of accepting
during refueling and storing during reactor operation, components
whi.ch would be removed during a normal refueling outage. These
include control/safety rod poison subassemblies, driver fuel
assemblies, reflector assemblies, radial restra.int/reflector
assemblies and open test positions. Excluded items are those
components which are removed directly from the core, such as
test assemblies.
1.2.2.3.1.2 Azimuthal Location
Azimuthal location of in-vessel storage positions shall
consider the following:
IVHM Access
Neutronic Coupling 1-29
BNWL-500 Volume 31
Local thermal and neutronic effects on the vessel and
core barrel walls
Relation to Ex-Vessel Flux Monitor Positions.
1.2.2.3.1.3 Fuel Storage
In-vessel storage shall be at stationary locations near
the vessel periphery.
1.2.2.3.1.4 Cooling Requirement
Assemblies stored at the in-vessel storage positions shall be
cooled with bypass flow from the primary sodium system. Adequate
cooling shall be available for removing the maximum heat load
from anticipated advanced driver fuel assemblies.
1.2.2.3.1.5 DBA Considerations
Design of in-vessel storage positions shall prevent criticality
and be such that the effect of stored fuel on accident situations
will not lead to an energy release greater than the design basis.
1.2.2.3.2 Mechanics
1.2.2.3.2.1 Positioning
Independent features for positioning shall be provided for the
assemblage of in-vessel storage positions such that any single
'Unlikely Fault' affecting the location of individual positions
as an assembly, will result in a predictable or negative
reactivity effect to limit severity of that fault to a 'Minor
Accident. '
1.2.2.3.3 Operations 1-30
BNWL-SOO Volume 31
1.2.2.3.3.1 Number of Positions
Sufficient storage positions shall be provided to store and
shuffle the anticipated number of fuel assemblies plus other 1 2 core components required for a normal refueling cycle. '
1.2.2.3.3.2 Position Access and Manipulation
Storage position shall be located within the vessel at locations
which are accessible with the IVHM and which minimize the
shuffling required to transfer units from the in-vessel
handling machine to the out-of~vessel handling machine.
1.2.2.3.3.3 Maintenance
In-vessel storage components shall be removable from the core
for maintenance and replacement.
1.2.2.3.3.4 In-Vessel Transfer
Common transfer points shall be provided between the three
sectors of the core to allow shuffling between regions serviced 3 by the three IVHM's.
1.2.2.4 Testing Capability
1. Refer to References, Appendix A, Item 2, for preliminary fuel management procedures.
2. The minimum acceptable number of positions is to be determined.
3. Refer to Interfaces, Appendix C, Item 6. 1-31
BNWL-SOO Volume 31
1.2.2.4.1 Test Position Interchangeability
Capability to interchange closed loop and contact instrumented
open test positions shall be maximized. As a minimum, the
capability to replace any closed loop by a contact instrumented
open test position shall be provided.
1.2.2.4.2 Total Number of In-Core Test positions
A minimum of eight positions with contact instrumentation
capability shall be provided in the active core.
1.2.2.4.3 Test Position Location
A minimum of two positions with contact instrumentation
capability shall be located within 8 inches of the core
centerline. The remaining in-core test positions shall
be located along the radial arms.· (See 1.2.2.1.1.3 and
1.2.2.2.1.2.)
1-32
BNWL-500 Volume 31
SECTION 2.0 PHYSICAL DESCRIPTION OF THE SYSTEM
2.1 SUMMARY DESCRIPTION
The primary function of the FTRI is to provide the capability
of testing candidate materials for the fast breeder program in
environments approximating those-to be found in fast breeder
reactors. The reactor core is the most critical component in
fulfilling the stated function~' The core provides the neutron
flux, establishes the power-densityand.operating temperature,
and provides locations for experiment placement. 2
The FTR reference core3
(see Figure 2.1) operates at a power
level of 400 Mw, is cooled by sodium and provides a variety
of test locations. Reference design parameters are tabulated
in Appendix D Table I. The reference reactor core consists
of a 91 position hexagonal array of-driver fuel assemblies,
test assemblies, control assemblies, reflector assemblies and
associated structure arranged in a vertical array.
Test positions are located in a"Y" shaped-array with the
intersection of the legs located on the -central element of
the active core. The arms then extend to the center of the
flats of the overall hexagonal cross section. The test position
array was selected so that a 120 0 access area could be gained
to core components other than tests without disturbing the
test positions, particularly the test position extensions
which protrude above the top elevation of the other core
components.
1. Refer to Drawings, Appendix F, SK-3-l4545 and SK-3-l4544. 2. Refer to Support Information Requirements, Appendix B,
Item 3. 3. Refer to Alternate Core Designs, Appendix E.
2-1
IV I
IV
~ OPEN TEST ASSEMBLY WITH PROX. INSTR. - 1
~ OPEN TEST ASSEMBLIES - 2
~ CLOSED LOOPS - 6
~ SAFETY RODS - 3
C]]J REFLECTOR/RESTRAINT POSITIONS - 42
~ PERIPHERAL CONTROL RODS - 15
REFLECTORS - 66
t::J DRIVERS - 76
~ FLUX MONITOR POSITIONS - 2
~ IN-CORE SHIM/SCRAM RODS - 3
~ COMBINATION STIF AND FLUX ~ MONITOR POSITION - 1
FIGURE 2.1. Reference Core Map
<b:I o Z 1-':8 s:: t-' ;::J I CD V1
o WO I-'
BNWL-500 Volume 31
Access to positions other than the test items is required to
monitor instrumentation during operation and to manipulate
fuel from the core to in-vessel storage positions located in
the periphery of the vessel.
Another factor influencing-test"positioning is the routing of
closed loop coolant piping and open test failed element
detection and location (FEDAL) lines away from the reactor
cover nozzles. Ideal routing is a straight corridor extending
radially from the central test positions to the edge of the
reactor. To prevent interferences with the in-core control
rods and open test and closed loop nozzles, the most efficient
arrangement develops when the piping is located laterally one
half lattice position in one direction off of the radial leg
and the test position is off-set one-half position in the
opposite direction. with this cross section, the central core
position is utilized for a fuel assembly with the second row
containing three tests and three fuel positions. (See Figure 2.2)
In addition to the fuel and test location, shutdown and control
margins for the reactor control systeml dictate the use of
6 in-core control and safety positions plus 15 peripheral
controls. Locations for the in~core controls is dictated
by the nozzle spacing at the reactor cover. Each control
and test position occupies an 8-inch OD cross sectional
envelope above the core. The envelope for these items
plus the closed loop piping requirements restricts the location
of the in-core and peripheral rods to specific locations 2 preferably away from the center of the core.
1. Refer to References, Appendix A, Item 8. 2. Refer to Drawings, Appendix F, SK-3-l425l~
2-3
tv I
01:::>
TEST CORRIDOR
~-"'--- --""-- "--" ~',
PIPHG t /
t
= CONTROL/SAFETY CONTINGENCY POSITIONS
= TEST POSITIONS
CONTROL AND SAFETY POSITIONS
TESf l FIGURE 2.2. Arrangement of Test Facilities
BNWL-500 Volume 31
Provisions for 3 additional control or safety contingency
positions are also shown in Figure 2.2. To meet criteria
(refer to section 1.2.1.6) for fuel assembly instrumentation,
the outlet of each fuel assemblyl mates with an instrument
probe containing a vortex generator to concentrate fission
gas for recovery by the failed fuel detection system, an
array of thermocouples and an eddy current flowmeter. The
thermocouple and flowmeter assembly are individually removable
through the instrument tree plug. The individual probes are
mounted in the fuel array to an instrument support plate which
contains all of the instrument probes for a one-third section
of the core. The support is in turn attached to the instrument
shaft. This assemblage of probes, support plate and
instrument shaft form the instrument tree. The instrument
tree in conjunction with the instrument tree plug perform a
variety of functions in the FTR. Primary functions include:
Providing the structure for supporting the instrument
assemblies.
Providing backup holddown stops to prevent ejection of
fuel assemblies in case of hydraulic balance failure.
Positioning the core assemblies into the operating con
figuration from the fuel handling clearance configuration
prior to restraint mechanism actuation.
Providing axial and rotary motion to clear the are~ above
the active core for fuel handling operations.
Assuring disengagement of the fuel assembly from the
instrument probe by utilizing a holddown plate to strip
probes away from the fuel assembly joint.
In conjunction with the IVm~, providing the translatory
motion to locate the core components during refueling.
1. Refer to Drawings, Appendix F, SK-3-l458l.
2-5
BNWL-500 Volume 31
Providing the routing for fuel assembly instrument leads
and FEDAL system tubing.
Checking the top of the safety/control rod and fuel
positions during and after instrument probe removal
to assure separation.
The core support may be divided into axial and lateral com
ponents. Axially the assemblies making up the core (with the
exception of the closed loops) are supported at the lower 1
inlet seat of the hydraulic balance inlet receptacle. In
addition to providing the primary reference point, the inlet/
receptacle provides the hydraulic balance which in conjunction
with the dead weight of the assembly creates the primary hold
down for the components. The core support structure, as a
system, must be designed to limit flow and mechanically 2 induced vibrations to acceptable levels. In the event of
failure of the hydraulic balance, secondary holddown is
provided by the instrument tree holddown plate which prevents
the core assembly from being ejected from the core.
Shield assemblies 3 and the outer reflector assemblies 4 cooled
by by-pass sodium are supplied from a secondary lower pressure
plenum created by the support structure. Hydraulic balance is
not required for these positions due to the reduced inlfrt
pressure. Primary holddown is supplied by assembly weight
with secondary holddown furnished by a perforated plate
structure at the top of the assemblies. 5 Lateral support
is supplied by a core support barrel, a cylindrical container
l. Refer to Drawings, Appendix F, SK-3-14585. 2. Refer to Support Information Requirements, Appendix B,
Item II. 3. Refer to Drawings, Appendix F, SK-3-14570. 4. Refer to Drawings, Appendix F, SK-3-14499. 5. Refer to Drawings, Appendix F, SK-3-14433. 2-6
BNWL-500 Volume 31
differential thermal expansion between the instrument tree
holddown plate and the top of the core elements. l
Disengage the control drive extension at the upper and
lower joint leaving the extension rod resting on the
in-core portion of the control assembly.
Disengage the upper joint and retract the upper shaft
of the extension rod into the drive mechanism.
Remove the drive mechanism and other instrumentation, etc.,
protruding above the operating floor. 2
Disengage the intermediate shaft utilizing the special
disengagement tool and raise this section to a position
flush with the bottom of the reactor vessel cover (this
motion also deactivates the lower collet of the main
shaft) .
Raise the instrument tree checking to assure that the
assemblies are free from the probes by actuating the 3 holddown plate.
Disengage the radial restraint mechanism.
Back the instrument tree free from the fuel assemblies
by a combination of rotation of the instrument support
post and the instrument plug.
Secure the tree into the fuel handling position.
Proceed with the fuel handling sequence transferring
fuel between the common loading point, the core, and the
in-vessel storage positions.
Location of in-vessel storage is dictated by several restraints,
these include:
1. Refer to Drawings, Appendix F, SK-3-l4606 for sequence. 2. Refer to Drawings, Appendix F, SK-3-l4604. 3. Refer to Support Information Requirements, Appendix B,
Item 18. 2-8
BNWL-SOO Volume 31
It must be accessible with the IVHM.
It must be located with relation to the reactor vessel wall
and the core barrel to reduce additional fluence on the
components due to spectrum shift from coupling with the
t ' 1 ac lve core.
A common removal point must be accessible with the fuel
handling machine while the IVHM is still operable.
2 The in-vessel storage positions shown were selected to give
the maximum number of positions consistent with the above
restraints. Each storage position consists of a hole in the
support structure and a lateral support system. Flow is
supplied to the storage position from the secondary support
structure plenum.
Studies of the neutronics of the reference core have been
completed to optimize the design. l The neutron energy
spectrum for the reference core is shown in Figure 2.3.
Total normalized flux as a function of radial position is
shown in Figure 2.4 and 2.5 tor the active core.
Coolant flow through the core is upward from a high pressure
plenum into a removable inner inlet plenum. This l:emoval.Jle
plenum provides an initial restrictor to prevent foreigfr
objects frmil blocking the individual assembly inlets. A
second restrictor is provided by incorporating radial holes
around the individual inlets as well as a bottom opening. 3
1. Refer to References, Appendix A, Item 2. 2. Refer to Drawings, Appendix F, SK-3-14S44. 3. Refer to Drawings, Appelldix F, SK-3-14585.
2-9
1'0 I
I-' o
---"--------------------,
1.0 r-----___ ~
>L':) a::: w z w
>< ::>
.8
-.I lJ...4
lJ.. a z a ....... IU ~ a::: lJ.. .2
o .oot .01 . 1
ENERGY (Mev)
FIGURE 2.3. Central Integrated Flux Spectrum
N I
I--' I--'
>< :::J --l l.J....
Cl z: <:(
>-I-
(/')
z: w Cl
0:: w 3: 0 CL
Cl w N
--l <:( ::E 0:: 0 z:
1.4
1.2
1.0
.8
.6
.4
.2
o
ZONE I
o 10 20
FIGURE 2.4.
'"
30 40
,P07NSITY '\: .
" " "
ZONE II
50
RADIUS (em)
,
60
REFLECTOR
70 80
Normalized Radial Power Density and Flux Profiles at the Axial Mid-Plane
90 95
N I
I-' N
::I--Vl z: w a 0::: w 3: o CL
a w N ----l <::( L 0::: o z:
1.2.,---, -'.,----.. --'---------r--. ---,
Power Density 1.0
-- Flux
.6
.4
.2
COR E t REF L E C TOR
OL-____ ~ ______ -L ______ L-______ L_ __ ~_J ______ _L ______ J_ ______ ~ ____ ~~~
o 10 20 30 40 50 60 70 80
DISTANCE FROM CORE CENTER (Cm)
FIGURE 2.5.' Normalized Axial Power Density and Flux Profiles at Radial Center
90 95 <td o Z I-'~ ~ t"i S I (j) U1
o WO I-'
Bm'~L-500
Volume 31
Flow then continues into ~he hydraulic balance receptacle
and into the individual assembly. Hydraulic balance is
maintained by venting che cup below the end of the fuel
assembly through small vent lines located within each fuel
assembly.l
The major increment of pressure d:cop for: the primary coolant
system is contained in the fuel assembly. Present calculations
indicate a maximum pressure drop for the fuel assembly is 120 psi
which includes orificing2
and insJcrument probe contingencie::;.
Maximum allowable pressure drop for the reactor (inlet tu outlet)
is 145 psi.
Some of the coolant flow by-passes the instrument probe through
leakage around the probe/fuE:!l assembly receptacle. ThE:! major
portion of the flow, however; passes through the instrument.
probe containing a vortex generator which concentrates fission
gas for capture by a removable probe. Flow then exists
through peripheral and annular openings at the upper end of
the prube.
In addition to the primary flow path through the core ~ucts a
secondary low pressure plenum formed by the support structure
is fell from the removable inlet pler!unl by the third ref lector
row assemblies which extend into special receptacles in the
plenum and incorporate outlet passages in the support structure. 3
This secondal:y plenum has penetrations to provide by-pass
coolant flow to the shield assemblies, to the in"vessel storage
lucations and to the vessel thermal liner annulus.
1. Refer to Drawings, Appendix P, SK-3-14581. 2. Refer to Support Information Requirements, Appendix B,
Item 19. 3. Refer ~o Drawings, Appendix F, SK-3-l457u.
2-13
BNWL-500 Volume 31
By·-pass coolant around the asser.,blies is pruvided by l:elief
[low irum ~~he inlet hydraulic balance features.
2.2
2.2.1
DETAILED DESCRIP1'ION
1 FUE::1 Assembly
A complete description of a fuel assembly may be found in the 2
CSDD for the First Core Fuel Assembly Component. Items which
cO~1sti tute an interface with other core compone:!.1ts or other
systems are discussed here.
Primary areas of interface include:
Assembly Cx'oss Section and overall lattice dimensions
Flow iElet
Outlet area
Radial Restraint Pads
Fuel Handling Slots.
Fuel a~sembly size is the primary factor affecting core lattice
size and overall core dimensiuns. The fuel assembly size iF;
a function of fuel pin diameter, which in turn is based on
heat removal characteristics and pin spacing. 3 Figure 2.6
shows -the effect of fuel pln OD on duct cross section as a
function of core !:.T while Figure 2.7 relates these dimensions
to the nlinimum required closed loop test diameter of 2.5 inches.
Other factors af .Lecting the COl.'e la tti(;e, volume and hence
the peak flux ir.cludes the number of assemblies required for
1. Refer to Drawings, Appendix F, SK-3-l~58l. 2. Refer to References, Appendix A, Item 9. 3. Reier to Support Information Requixements, Appendix B,
::::tem 8.
~--------- ~~~-
I I
tv I
...... en
V')
V') l-ou c::: => uo c::r:
>-z V') o V')
-- c::r: V')
z......J Li.J Li.J ::::E: => ...... l.J... Cl
l.J... Li.J 0 0 __ V')
V') l-I- c::r: => ......J Ol.J...
. Z --
:).0
4.:)
4.0
3.S
3.0 0.200 0.210 0.220 0.230 0.240 0.250
FUEL PIN 00 (IN.)
FIGURE 2.7. Basis for Fuel Assembly Size Selection
MININJM REQ'O. LOOP 00 (IN. )
<to g.~ Ct"i S I CDUl
o wo ......
BNWL-500 \701ume 31
f 'd' 1 sa ety conSl eratlons. Number o~ assemblies to achieve the
reference power based on the reference design is shown in
Figure 2.8. The net result of the above considerations is the
peak flux which may be directly related to the core volume as
shown in Figure 2.9.
'1he fuel a.ssembly inlet consists of a taperGd r.osepiece to
aid in locating the assembly in its recep-::acle. Flow enters
the assembly through peripll.eral slots iTl the nosepiece. The 2
nosepiece is seated in ~he hydraulic balance receptacle.
Hydraulic balance is maintained by venting the lower end of
the receptacle through small by-pass tubes in the fuel nose
piece to the sodium pool. By maintaining the same leakage
flow into the lower section of the receptacle a~ is by-passed
into the pool, a floating or balance condition is achieved.
The assembly is then held in place oy the dead weight.
The outlet section of the fuej, assembly interfaces with the
instrument probe through a collar on the fuel assembly.
Although a significant percentage of the flow must by-pass
this joint,3 sufficient flow must be maintained to reliably
monitor temperature flow and fission gas release for each
assewbly. A loose fit at this joint is required to prevent
seizing of the assembly during instrument probe withdrawal.
rhe upper surface of -the assembly contacts the instrument
tree holddown plate during operation to provide the backup
holcldmm for the fuel assemblies.
1. Refer to References, Appendix A, Item 3, Section 12. 2. Refer to Drawings, Appendix F, SK-3-14585. 3. Refer to Support Information Requirements, Appendix B,
Item 7.
2-17
N I
I--' co
90
85 217 PINS/ASSEMBLY
Cl -0' w 0:::
V) 80 >-V) V)
c::t:
--.l W => l...L
l...L 0
0::: 75 w cc ::E => ::z:
70
65L---------~------~~----------1~------~--------~
..... 0 32 34 36 38 40
CORE HEIGHT (IN.)
FIGURE 2.8. Number of Fuel Assemblies Required Versus Core Height
<to o Z 1--'::8
~ ~ CD lTl
e we I--'
1. 0 r---.------~----------------.-------'-------......,
0.9
lD
I 0
>< 0.8 >-:z::
>< :::)
---l l.L
---l 0.7 <::( I-0 I-
:::.::: <::( w 0....
0.6
0.5
700 800 900 1000 11 00 1200 1300 1400 <:tl:1 o z
CORE VOLUME (LITERS) 1-'::8 ~ t; (1) Ul
0
FIGURE 2.9. Peak Total Flux Versus Core Volume wo I-'
BNWL-500 Volume 31
The fuel assembly outlet also mates with the instrument tree
holddown plate during instrument probe withdrawal. The
pIa te contacts the upper edge of ·the fuel assembly to assure
that the assembly and probe are decoupling as the probe is
lifted.
Present studies indicate tnat to assure a predictable response
of the core under steady state and transient conditions,l a
lateral restraining or restraint load must be applied to the
core elements at a position near the core. The restraint
mechanism is described in a later portior. of this section.
The application of the load is through loading pads located
just below the active core region.
This position was selected to reduce the thermal stresses on
the thicker loading pad. Additional pads utilized for applica
tion of secondary loads are also located just above the core
and at the top of the ducts. The pads are 0.050 inch in
thickness giving a 0.100 inch gap between assenililies. This
gap is required to provide clearance for tolerance buildup and
for thermal and irradiation induced distortion effects. The
effect of variations in the gap size on flux is shown in
Figure 2.10.
Attachment provisions for the fuel handling machines consists
of a set of slots located at the points of the hex can below
the instrument probe mating ring. The handling machine grapple
pilots into the fuel asserr;bly outlet, fingers then engage
the fuel assembly slots. To assure that fuel may not be placed
in a control or reflector position due to operator error, the
design must incorporate features to prevent such an error.
1. Refer to Support Information Requirements, Appendix B, Item 5. 2-20
.. . 41
QJ .,.... 41 In IG - VI U IG
COU QJ CI ,..... VI 0 0\41 IG - o VI
..u 1.1 1.1 a.. • IaI ,..... .a
II II .
41 CI CI . 0 .,.... 0
t.I ....J t.I
co C ....J ,..... z: CI
CI 0\ :::> z: 0 0 eo :::> - - eo a.. ::c -BASE CASE u ::c II II I- U
( (; .e .• PlOD .... I-
CI CI 1. 0918) a.. .... 0 0 = a..
t.I l.J..J W U l.J..J
....J ....J ...... U CI CI I- .... :z: :z: 1.01.- 1.0 I-' I-:::> :::> ct: I-eo c::o ....J ct: - - ....J
::c :I: l-I-U U ct:
l- I- ct: .... ..... ~
a.. a.. a.. ~ a..
t.I t.I t.I U U ....J t.I ..... ..... ..... ....J
l- I- '" '" l- I- ~ V'l c:t: c:t: .... .... ....J ....J 1.1.. 1.1.. - -l- I- :::> :::> ct: c:t:
>< >< 0
~ ~ ....
0.9 I-
0.9 c:t: ~ ~ c:::
1.02 1.04 1.06 1.08 1.10 1.12 1. 14 ..
0 RATIO: LATTICE PITCH/FUEL PIN BUNDLE OD ( = PlOD) .... <:tJ:l I- 02: ct: c::: 1-':8
FIGURE 2.10. Influence of Pin Bundle to-Duct Clearance, Duct Wall ~ t'"i E3 I
N Thickness, and Duct-to-Duct Clearance on Peak Total CD U1
I 0
N Flux and U/Fissile Pu Ratio wo I-' I-'
BNWL-500 Volume 31
The selected method is to vary the axial location of the slots
and then provide stops on the IVHM to restrict axial movement
below a given height.
2.2.2 Nuclear Control Components
A complete description of the control components may be found
in the CCDD for the Reactor Nuclear Control Components. As with
the fuel assembly, only items constituting an interface with other
core components or other systems are discussed here.
Primary interface areas differing from those already described
for the fuel assembly, such as the inlet region and radial
restraint pads, include:
Control Effects on core flux
Flow outlet region
Extension rod.
Control rod interfaces not directly related to the core are
detailed in the control rod component design description.
A combintation of in-core and peripheral control has been
selected for use in the FTR to concurrently provide the
needed control margins while maintaining the flux at as high
a level as possible. In order to maintain the minimum closed
loop test volume with current core ~T designs complete periph
eral control is not feasible as shown in Figure 2.11. By
utilizing a combination of in-core and peripheral controls
adequate control margins are available. A comparison of
in-core and peripheral control effects as a function of peak
neutron flux and U/Fissile Pu ratios which determines the
Doppler coefficientl is shown in Figures 2.12 and 2.13.
1. Refer to Support Information Requirements, Appendix B, Item 1.
2-22
I\J , I\J W
V')
0:: w f-I--i
-l
LJ.J ::E: :=J -l 0 :::-LJ.J 0:: 0 u
12CJG (0[:[ uESIGiiS IN ThIS REGION, i.e., l,JITH VOLUfv1ES GREAl ER THAN'\, 95C 1 l-JILL REQUIRE ~10RE THAN 3 IN-CORE RODS TO MEET THE FTR CONTROL SYSTEM WORTH REQUIREM T'
"~" ~~ 1100 ~ ~
1000
900
800
USE OF REENTRANT LOOP TUBE WITH A MINIMUM TEST SECTION lD. OF 2.5 IN.
700~~~~~ ________ ~ ______ ~ ________ ~ ______ ~
300°F (1 T
350°F fiT
400°F ilT
0.200 0.210 0.220 0.230 0.240 0.250 FUEL PIN 00 ( IN.)
FIGURE 2.11. Core Volume Versus Fuel Pin OD for Various Core ~T'S Design Based on Use of 3 In-Core Safety Rods and Full Peripheral Shim-Regulating Control
0:: W f-w ::;:: c::( >-< Cl -...... :r: u f-...... CL
z: ...... CL
--J W :;) l.J..
1. 6.7 KW/FT AVG NOM. LINEAR HEAT RATING.
2. 10 PSI/FT FUEL PIN BUNDLE FRICTION 6P. 1.1~ __________________________________________ -,
1.3
1.2
1.1
A" a --9
~EAKC)TOTAL FLUX,
Nv x 10- 16 (TYP)
U/FISSILE Po RATIO (TVP) )
o ',&
1.19· 28
3·15
PEAK FUEL ASSY AT, of (TYP) ASSUMING A 1.4
RADIAL PEAKING FACTOR
1.0~~ ____ ~ ________ ~ ______ ~ ________ ~ ______ ~
BNWL-500 Volume 31
.220 .230 .210 .250
FUEL PIN 00 (IN.)
FIGURE 2.12 FUEL PIN PARAMETERS (3 IN-CORE RODS) 2-24
er: w I-W ::t: 0:::: ....... Cl ........ ::c u I-....... 0-
z: ....... 0-
.....J W => LI..
1. 6.7 KW/FT AVG. NOM. LINEAR HEAT RATING. 2. 10 PSI/FT FUEL PIN BUNDLE FRICTION 6P. 1.4~ ________________________________________ ~
1.3
1.2
1 . 1
U/FISSILE Pu RATIO (TYP)
PEAK FUEL ASSY 6T, OF (TYP) ASSUMING A RADIAL PEAKING FACTOR OF 1.4
\
1.0 ~~----~~----~~------~~----~~----~ 00 .210 .220 .230 .240 . 50
FUEL PIN 00 (IN.)
FIGURE 2.13 FUEL DESIGN PARAMETERS (9 In-Core Rods)
BNWL-500 Volume 31
2-25
BNWL-500 Volume 31
Since axial access must be available to position the control
rods, the instrument probe control position interface must be
significantly different from the similar interface for driver
fuel positions. For control rods, the instrumentation consisting
of a flow indicator and outlet thermocouples are located
in a tube supplied by an annular probe extending about 2
inches into the control rod duct.
The extension rod for the control rod, a two piece member
consists of one section which is positioned by a guide tube
attached to the instrument tree and an upper section which
may be withdrawn into the reactor vessel cover after removal
of the control rod drive mechanism. When the lower section
is disconnected from the in-core section and the upper rod,
it rests in special guides attached to the instrument tree.l
A detent on the rod is engaged by axial movement of the
instrument tree during retraction of the tree from the fuel
assemblies. The rod is thus lifted clear of the in-core
portion of the rod during refueling and remains with the
instrument tree.
The poison section of the control rod is housed in a duct
similar in configuration to the driver fuel assembly.2
Interfacing areas with adjacent fuel assemblies are the same
as those described for the fuel positions. Poison section
removal, based on allowable irradiation effects, is accomplished
utilizing the IVHM.
1. Refer to Drawings, Appendix F, SK-3-14604. 2. Refer to Drawings, Appendix F, SK-3-14560.
2-26
2.2.3 Reflector Assemblies
BNWL-500 Volume 31
Three rows of assemblies containing reflector material are
located around the periphery of the active core region. In
addition to reflector assemblies, these rows also contain the
peripheral control rods in the inner row, flux monitor and
radial restraint loading beams in the middle row, and additional
restraint beams in the outer row. By utilizing a high nickel
content material in this region the neutron leakage is reduced.
A separate duct configuration is required for each of the three
rows due to functional requirements in addition to the basic
reflecting function. All three rows, however, are similar in
the core region. The in-core portion consists of a rod bundle
with a cross section similar to the fuel assembly. The bundle
is contained in a stainless steel duct of similar design to
other core components.
The inner reflector row positionsl
are externally the same as
the driver assemblies. Internally, the first row configuration
contains 0.5 OD pins, the in-core portion consisting of a
stainless clad nickel rod which extends 6 inches above and
below the core. The remainder of the rod, extending from
approximately one foot above the inlet to one foot below the
top of the duct is a stainless rod which provides shielding
for the vessel and core barrel. All reflector assemblies are
handled by the IVHM with replacement rates based on allowable
fluence on the individual positions.
In addition to containing the reflector material, the central
reflector row is also the primary loading beam for the core
1. Refer to Drawings, Appendix F, SK-3-l4636. 2-27
BNWL-500 Volume 31
radial restraint system. The in-core portion contains a
nickel rod array similar to that described for the inner row.
The height of the duct portion of the central reflector row
is shortened to provide clearance for the radial restraint
loading member. Radial restraint functions of this assembly
are described in Section 2.2.8. To achieve the proper loading
pattern for radial restraint, the two central positions of the
central reflector row do not contain a restraint mechanism
reflector. The restraint for these center positions is located
in the third reflector row (see Figure 2.1). The two central rows
center of flat positions contain first row reflector assemblies.l
The third row reflector positions also perform a dual role.
In addition, to provide the outer boundary for the radial
reflector region, the lower end of the assembly is utilized to
connect the high pressure inlet plenum with the secondary low
pressure plenum formed by the support structure.2
The inlet
end of this position passes through the support structure and
sockets into an inlet receptacle connected with the high
pressure plenum. The inlet receptacles are orificed to regulate
flow and pressure into the secondary plenum. In the secondary
plenum region a flow split occurs with the major portion of the
flow passing into the secondary plenum through peripheral
openings in the duct. The remaining flow passes through an
additional orifice into the rod array. The rod configuration
for the third row is similar to that of the first row assemblies.
For the first and second row positions, hydraulic balance
receptacles are utilized for primary holddown. Backup hold
down for the first reflector row is provided by the instrument
tree.
1. 2.
Refer to Drawings, Appendix F, SK-3-l4636. Refer to Drawings, Appendix F, SK-3-l4570.
2-28
BNWL-500 Volume 31
Backup holddown for the second reflector row is provided by
the restraint loading beam.
The reduced diameter of the third row inlet eliminates the
need for hydraulic holddown. In addition, the space available
for the penetration precludes incorporating this feature. For
these positions holddown is afforded by the assembly weight
with the backup holddown maintained by the restraint loading
plate.
Although the third row is accessible with the IVHM, the restraint
plate limits replacement of these positions to major maintenance
shutdowns when the restraint system is removed. Characteristics
for the reflector assemblies are shown in Table I.
2.2.4 Radial Shielding Assembly
Radial shielding for the FTR is required to reduce the flux
on the core barrel and reactor vessel wall to values consistent
with established material properties for critical structural
components. The radial shield also reduces the problem of
neutronic coupling between the in-vessel storage positions
and the reactor core.
To meet the above objectives a nominal radial thickness of
approximately 55 cm is required. The shield assemblies form
the transition between the hexagonal cross section of the
active core and the circular core barrel. The shield thickness
thus varies from about 46 cm to 61 cm. Partial lattice
positions at the core barrel are formed by welding the partial
ducts to the core barrel internal diameter. The remaining
shield assembliesl
are made up of ducts similar in configura
tion to the fuel assembly containing 0.500 OD stainless steel
1. Refer to Drawings, Appendix F, SK-3-l4499. 2-29
BNWL-SOO Volume 31
TABLE I. Reflector & Shield Reference Concept Characteristics
units Reflector Shield
1.
2.
3 .
4 .
S.
6.
7.
8 .
9.
Duct Outside Dimension Across Flats
Duct Wall Thickness
Lattice Dimension
Configuration
Number of pins
Pin OD
Pin Lattice
Overall Assembly Length
Drawing
10. Volume Fractions
Stainless
Sodium
Inconel
11. Coolant Flow/Assembly
12. Residence Time In-Reactor
13. Method of Handling
14. Primary Method of Holddown
IS. Backup Holddown
in.
in.
in.
in.
in.
ft.
%
4.61S
0.140
4.71S
4.61S
0.140
4.71S
Pins on triangular pitch
61
1/2
0.S38
14 to 12
61
1/2
0.S38
10 ft.
SK-3-14636 SK-3-14499 SK-3-14S68 SK-3-14600
29.9
26.2
43.9
TBD
TBD
IVHM
Hydraulic Balance
Instrument Plate
73.8
26.2
TBD
TBD
Special Maintenance
Dead Weight
Barrel Cover Plate
rods in a triangular array. The assembly receives by-pass
primary coolant from the support structure plenum. Flow is
orificed to these positions to maintain an outlet temperature
consistent with that from the active core. Primary holddown
is maintained by the dead weight of the assembly with an upper
perforated plate structure attached to the core barrel providing
2-30
BNWL-500 volume 31
the backup holddown. Shield assemblies are foreshortened to
provide clearance for the restraint and FEDAL systems.
Characteristics for the shielding assembly are shown in Table I.
2.2.5 Open Test Positions
Open test positions provide locations for utilizing primary
sodium for cooling test specimens occupying a full lattice
position. with the exception of one test position noted below,
all open tests will include provisions for contact instrumenta
tion on the test specimen. The in-core portion of the open
test l consist of a duct similar in configuration to the driver
fuel duct. The assembly contains provisions for mounting
either fuel pin bundles or other test specimens. Instrumentation
included in the open test position includes:
Thermocouples
Pressure sensors
Gas sampling lines Flux thimble FEDAL sampling system.
Instrumentation is routed from the core region through a
3 1/4 inch hanger tube attached to the duct section. The
connector design at the reactor cover2 allows an open loop
to be inserted in any of eight in-core test locations. The
ninth test position must include the capability for removal
of the instrumentation hanger rod during refueling. Removal
of this rod is required in order to remove and relocate the
instrument probe at the central driver fuel position. The
open test position will be capable of being examined and
replaced during reactor shutdown. By slipping the outer duct
1. 2.
Refer to Drawings, Appendix F, SK-3-l4586. Refer to Drawings, Appendix F, SK-3-l4461. 2-31
BNWL-500 Volume 31
off, the test specimen may be visually examined with the duct
replaced afterward for further irradiation.
2.2.6 Closed Loop Assembly
Primary features and requirements for the closed loop are
contained in the CSDD for the Closed Loop System. Primary
interface areas with the reactor core in addition to neutronic
performance include:
Lower seat connection
Interaction at contact pads.
The closed loopsl like the open test positions may be located
at any of the eight contact instrumented in-core test positions.
Since the test positions are interchangeable for open or closed
tests, the closed loop, which does not require cooling from the
inlet plenum, must provide a plug for sealing the inlet plenum
hole.
The interaction at the contact pads is particularly important
with the closed loop as well as with the open positions due to
the extension of these components to the reactor cover. The
double supported, e.g., cover and tubesheet, test positions
are less compliant than other core components, the radial
support must account for this anomoly. (Refer to Section 2.2.7)
2.2.7 Core Restraint
To assure a predictable neutronic response due to radial motion
of core components, a radial restraint systeln is require6 for
ble FTR. This sys·tem must assure that under all operating
1. Refer to Drawings, Appendix F, SlZ-3-l4515. 2-32
BNWL-SOO Volume 31
conditions, including transi8nts, that component motion will
result in an overall negative feedback effect.
Several factors can result in radial motion of core components.
These include:
Differential thermal expansion
Stainless steel swelling
Radiation Induced Creep
Thermal creep.
The first effect results in a dynamic response under transient
conditions as well as a steady state effect. The latter items
are effects which occur as a function of burnup, temperature
and fluence and therefore are dependent on residence time in
the reactor.
Empirical relations regarding stainless steel swelling and
radioactive induced creep are presently incomplete. The best
data available however, has resulted in a compliant loading
system utilizing the second and third reflector row for
applying a radial load to the inner assemblies. The primary
loading position is directly below the active core with secondary
loading points above the core and at the top of the assemblies.
The compliant restraint is one in which elastic deflections are
limited by core packing at reaction pads. A compliant loading
member utilizing the reflector positions is used to accommodate
radial expansion of the core.
The radial restraint system incorporates features in:
the fuel assembly duct
the radial reflectors
the instrument tree probes
the radial restraint mechanism.
2-33
Interfacing areas inGlude:
the tubesneet receptacles
the core barrel
the reactor cover
the instrument tree.
BNWL-50~
Volume 31
The factors involved in restraint system design, as noted above,
affects the system selected to varying degrees.
Thermal camber results from differential thermal expansion
across the duct wall due to the power gradient across the duct.
This effect can be easily analyzed but must be compensated
for during transient conditions to assure that reactivity
effects are negative. l Figure 2.14 shows the resultant deflec
tion and pad reactions for the lattice positions as a function
of the thermal gradient.
Irradiation induced swelling is known to be a function of the
flux and thermal gradient occurring across the duct. It may
also be dependent on the stress state of the duct. The phenomena
is not well definea at the present time,2 but initial analysis
of test data indicates the differential swelling effect (~L/L)
is significant at the FTR temperatures and burnup. The effect,
however, is tilue dependent and therefore manifests itself as a
function of time in reactor (as fuel burnup) and exposure
temperature. The resultant effect of swelling is duct distortion
which can result in handling difficulties. 'I'he present design
approach for first core oesign is to limit the burnup and outlet
telllperatures to a level which will result in differential
1. Refer to Support Information Requirements, Appendix B, I-tern 2.
2. Refer to Support Information Requiremen-ts, Appendix B, Item 13.
2-34
N I l;J -
Vl
+0.040
+0.030
+0.020
+0.010
0.000
1 -0.010
-0.020
-0.030
-0.040 . z: .....
!z: LoJ 2: LoJ U
:5 L CIt ..... C
SUPPORT NtJ4BER
, ...... '4I--__ ACT=IV=-E ___ • I r CMEI
I
2 3
20 40 60 80 INCHES 100 120
TE,.,. SUPPORT REACTIONS - L8S 2 3 4 5
100°F 17 -170 752 -972 372 75°F 12 -120 564 -727 279 50°F 8 -80 376 -485 186 25°F 4 -40 188 -242 93
SU8ASSEMBLY LENGTH
FIGURE 2.14. Initial FTR Core Thermal Deflection Curve Above and Below Core Support (5th Row)
140 160
<:td o Z 1--':8 C t"' ;::1 I CD Ul
o LJJ 0 I--'
BNWL-500 Volume 31
5tain2.ess swelling «1%) which can be accormnodated by ·the
present reactor core arrangement. l Several methods are feasible
for reduction of swelling distortion. The best method is to find
a metallurgical treatment which reduces the metal;s susceptibility
to swelling, other methods include shuffling of fuel from high
to low gradient (both thermal and neutron) zones and rotation
of assemblies. Representative deflections 6ue to stainless
swelling from preliminary swelling models2
are shown in Figure
2.15 along with resultant reaction loads.
Radiation induced creep, like stainless swelling, is a time
dependent phenomena. Present data, which is limited, indicates
that the tiILle regime in which the creep occurs is of much shorter
duration than for the swelling phenomena. The data indicates
that creep begins at reactor startup and continues to some
residual stress factor (0/0 ) as yet not defined up to a o
fluence of about 1 x 1020
nvt. Above this fluence the limited
data available indicates that further creep does not occur unless
a stress tnreshold of 15,000 psi is exceeded.
During the completion of the initl.al or primary creep period,
which extends from 6 :nours to more than a full reactor cycle,
depending on the relative distance from the axial core centsr
line, the restraint system must compensate for the reduction
in applied load due to the creep by excess additional load at
startup or by an adjustable restraint system.
1. Refer to Support Information Requirements, Appendix H, Item 6.
2. Refer to References, Appendix AT Items 17 and 18.
2-36
N I
W --<
0.020
0.010
0.000
-0.010
-0.820
z: 0 .... I-U L4.I ....J ~
~
2 3
1ST REACTOR CYCLE ~ ~2ND REACTOR CYCLE ~
/ //lST & 2ND REACTOR CYCLE \ .-~ 3RD .. 4TH REACTOR CYCLE--.I
II
NONROTATED
ROTATED 180· AFTER EACH CYCLE
__ l __ ~ __ ~ __ --.l 20 40 60 80 INCHES 100 120 140
CYCLE NO • SUPPORT REACTIONS - ROTATED SUPPORT REACT lOftS - NOMROTATED
2 3 4
1 2 3 4 5 1 2 3 4 5
-9.3 89 -70 -37 28 -9.3 89 -10 -37 28
-8.2 79 -63 -33 25 -28 265 -211 -111 85
-25 237 -188 -99 76 -53 510 -408 -214 165
-24 232 -185 -98 15 -83 795 -636 -335 256
FIGURE 2.15. Initial FTR Core Assembly Swelling Deflection (5th Row) Support Pads Above and Below the Core
160
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Volume 31
Operation of the radial restraint can best be described by
following a s<;quence from -thE: shutdown/refueling configuration
to an operating mode.
During sln:.tdmvn,the core: assembly clearance is increased
from a contacting configuration at the support pads to a
"petalled out" arrangemenL with a 0.1 inch clearance between
adjacent ducts. During preparation for startup the instrument
tree is positioned over the ducts and lowered axially. The
instrument probesl
are graded in length so that assemblies
are sequelltially piloted back into the contacting position
from the center of the core outward until all positions are
located and the instrument tree is in the full down position.
The radial restraint mechanism is then actuated. Actuation
involves lifting a rod exb2nding from a roller nut jack
located in the cover. The six jacks are actuated by extension
rods from penumatically driven motors located at the operating
floor.2
The rods extend to toggle mechanisms a~tached to the
core barrel. The toggle mechanism, an over center latch
system moves a slide plate imlard. 'I'he slide plate contact
pads located on the tapered beam section of the radial
restraint/reflector positions.3
contact pads on the slide
plate are ground so that the point elements are engaged
first resulting in an overall across points tightening. The
two positions located one lattice out from the points are
engaged next and so 011, resulting in a triangular wedging
tt 1 " 2 - 6 4 pa ern as s~own 1n F1gure .~.
'I'he thermal creep does not apl?ear to be a significant factor
for the first core due to the lower outlet temperature. Above
900 of, thermal creep becomes significant.
1. Refer to Drawings, Appendix F, SK-3-14604. 2. Refer to Drawings, Appendix F, SK-3-l4433. 3. Refer to Drawings, Appendix F, SK-3-l4568 and SK-3-l4600. 4. Refer to Support Information Requirements, Appendix B,
Item 17. 2-38
N I
w
'"
$ OPEN LOOP WITH PROX. INSTR. - 1
@ OPEN LOOP S - 2
@ CLOSED LOOPS - 6
~ CONT ROL.' SAFETY RODS
(ill) REFLECTOR/RESTRAINT . i ...
~ PERIPHERAL CONTROL
Q REFLECTORS .. 66
0 DRIVERS T3
e OPTIONAL TEST POSITIONS - 3
LOADING SEQUENCE IS DENOTED BY CD FOR EACH LOADING PATH.
FIGURE 2.16. Radial Restraint Loading Sequence
~
<:b:J o Z f-J~ ~t-l 8 I ro Ul
o WO f-J
BNVlL-500 Volume 31
Compensation for the stainless swelling ra1:..st be provided by
a restraint mechanism which ccJ.n load the core into a tigh·tly
packed lattice at the reaction pcJ.ds and maintain the compliancy
to allow radial grow"ch of the assembly cans. 1 ~'he combined
resultant of stainless swelling and therma:L deformation is
shown in Figure 2.17. Figure 2.17 does not account for creep
or stress relaxation effects.
As irradiation induced creep occurs, the radial restraint
system must either provide an excess load at startup which
at shutdown will still possess a minimum loading concH tion
or thE: capability must be available to sense the load and
provide a::1justrnent during opera·tion. The former method is
preferred due to its simplicity, however, with the uncertainty
in the creep effect at the present time, an adjustable system
is shown as an alternate2 to the preferred compensating load 3
sys"tem.
Both systems utilize the driver duct for load transmission.
Presen~ data indicates that by imposing the performance
limitations noted above, the resultant deformations and
stresses are within acceptable values. As additional results
cf radiation effects on material properties become available,
the single piece duct must bE: evaluated. to assure that deforma
tions are not excessive.
'l'he load is transferred from tLe contact point wiLh the slide
plate through the radial restraint/reflector assembly beam.
The beam is supported by the outer inlet receptacles at the
lower end. The applied load is reacted at the lower end and
1. Refer to Support Information kequirements, Appendix B, Item 12.
2. Refer to Drawings, Appendix F, SK-3-l4434. 3. Refer to Drawings, Appendix F, SK-3-l4433.
2-40
N I
01:::> I-'
5
2
1
FREE THERMAL CA'(BER
FIGURE 2.17.
-0.01
SUPPORT 3, 4 • 5 ACTInG
a_-- -.020
CCMBIIlED 'l'HERMAL • S'IllLLIIG DEFECTION @ ElfD OF FIRST CYCLE
Initial Core FTR Radial Restraint -(50 0 T across Duct)
<:trJ o z 1-':8 § ) (]) U1
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BN\~;L"500
Volume 31
in turn loads the fuel assemblies at contact pads located
below the active core, above -the Gore and at the top of the
duct. Tte beam loads the ducts into a tightly packed geometry
at these contact pad areas.
~o release the core for refueling the above sequence is
reveJ:-sed with the exception that the instrume.nt tree is with
drawn prim: to releasing t~'1e ra.dial restraint mec~'lanism.
By balc..ncing the moment of inertia for the d.uct with the
application point and magnitude of applied load, a relatively
stress free operating mode may be obtained.
2.2.8 Special Assemblies
It is recognized that assemblies other than those previously
described may De added to the core. Special assemblies
identifiec at this time include an oscillator position and
special asserr~lies with fractional fuel loading to replace
closed and open test positions. Primary emphasis on special
assemblies during the conceptual phase has been to identify
and survey those components which may be anticipated at this
t ' 1 lme.
2.2.9 Core Support Structure
The core support structure for the reference concept consists
of three ~i1ajor assemblies, 111 addition to the radial restraint
system. These asser,lblies are the:
Inlet Plenum
Ring Girder and Core Barrel Support
Core Barrel.
1. Refer to References, Appendix A, Item 6. 2-42
BNT.N'L-SOO Volume 31
All core support s-.:.ructure components are removable from the
reactor vessel utilizing special maintenance procedures. 1
The functions and description of each of trIese assemblies
are covered individually below:
2.2.9.1 Inlet Plenum
The inlet plenum2 i:; a removable section of the hig:-l pressure
zone of the reactor vessel. Removability for -::his flat plate
which is a primary structure is a function of the allowable
fluence which will result ir. an end of life mi:;'limum ductility
of 10 %. The plenum is rei:ai::1ed by an interrupted thread with
an anti-rotation ring placed in the thread interruptions. The
inlet plenum performs the following functions:
Forms a portion of the high pressure inlet for the
reactor ve.ssel.
Prevents foreign obj 13Ct:S from blocking the assembly inlets.
Supports a meltdown dispersion grid for containing and
dispersing a parti~ll fuel meltdown.
:!?rovides the inlet se.a·t for all in-core assei.Tlblies.
Incorporates hydraulic holddown features for in-core
assemblies.
The upper plate of the plenum contains hydraulic balance
receptacles in , ... hich all in-core assemblies are seated.
Hydraulic balance is achieved by maintaining a low pressure
area below the nosepiece of ea.eh assembly. The low pressure
is obtained by venting this ~hiDble to the rea~tor pool
through three 1/8 inch ·tubes located wi thin each core assembly.
1. Refer to Support Information Requiremerts, AppeLdix B, Items IS and 16.
2. Refer to Dravlings, Appendix F, SK-3-l4S8S. 2-43
BHWL-500 Volume 31
The outer row of penetrations provide Q seat for accepting
the outer reflector positions which provide flow to ".:he low
pressure pler.um formed by the lower section of the core barrel
and the ring girder.l
The cylindrical wall of i:he pler..um is perforated with linch
holes for inlet flow frlJI:1 the high pressure zone of the reactor.
The small diameter holes serve as a strainer to prevent foreign
objects from blocking the individual receptacle inlets. The
small holes also prevent vortices formed by inlet flow inter
actions from propagating into a fuel assembly ir.let.
The lower closure of the pler.um is contoured on the inteL'nal
face to dis?erse liquid fuel into a non-critical configuration
in the event of a mel tdo'v'Jn accident. Preliminary analysis
indicates that sufficient heat removal capdbility is provided
by dispersing the nel ted fuel through a combina·tion of contour
and hole pattern dispersion to contain a pD.rtia.1. meltdown
accident.
2.2.9.2 RinS Girder
'I'he ring girder is a circular hollmv vveldment resting on a
sheer ledge of the reactor vessel wall. The ring girder
performs the following functions:
1.
Locates and supports the core barrel.
Locates and sup~orts the in-vessel sto~age positions.
Provides a ·tertiary plenum for metering and distributing
f1mV' from th,;:, secondary plerum to the vessel thermal
liner in-vessel storage positions, fuel transfer pots.
anc1 the common fuel transfer/storage positions.
Supports and positions the vessel thermal liner.
Refer to Drav.:ings, AppE:ndix F, 8K-3-1454 -1. 2-44
BNWL-500 Volume 31
structurally the ring grider supports the core barrel, shield
assemblies, the radial restraint system and the in-vessel
storage positions. Flolv is introduced i.nto thi= cavity throug-h
holes from the seco~dary plenum formed by the base section of
the core barrel.
2.2.9.3 Core Barrel
The core barrel provides lateral as well as an axi~l support.
In addition to providing lateral support for the shield
assemblies, the shield cover plate on the barrel provides the
reactive load path for the radial restraint mechanism. This
plate is retained vertically but limited lateral motion is
provided. By allowing lateral adjustment in this plate in
conjunction with prepositioning the tops of the core assemblies
~"i th the instrument probes, relati ve aligr;m{~nt, of the instrur,1fmt
SUPi)Ort tree, the core ass(~ri1b] ies and the radi2.1 restrair.'t ma~i
be accomplished.
The lower closure plate also for~s tile s2condary plenum. The
plenum coolant is supplied by special third row refh~ct_orsl 'vV'hich orifices flow from the hisrh pre:;sur,~ plenum into tre
secondary plenum. The .3econdary plenum in turn supplies the
shield and outer refle8tor assemblies and the tertiary plenu~.
In summary, the core barrel performs the follm.rir.g functions:
Late1':'ally suppcrts the shield assEmb ties.
Provides the reactive load path for the radial restraint
mechanism.
Axially sup:r;:.cr-ts the shield and outer reflector assemblies.
Provi':!,es coolant flo~.v to tte shi.elc_ <lno. ()l)J: er reflecto]-
assemblies and to the tertiary plenum.
1. Refer to Drawings, Afpendix P, SK-3-14570. 2-45
BNWL-500 Volume 31
Provides a reactive load path for the x".:1di3.1. res trai.nt
~eam r2ceptacles in conjunction with the instrument
tree a_nc lower FE Dl':,IJ extensions ..
2.2.10 In-Cere Instrument:ad on
A complete descri.ption of the in-core instrumentation may be
found in the InstrumentaU_on & Contxol System CSIJD. 1 H:ems
covered here constitute funGtir.ll~a2- or mechanical interface
considerations. The instrume~tation for fuel assemblies and
control positions are located in the instr~ment probes of the
in3t::UIne~t tree. InstruIPEntation for ~-he open test positions
and closed loop~, 'x~..:ilize direct access and may include cont~ct
instrumentation on the test item. Other in-rea.ctor instrumenta-
tion includes in-co:r'€ flux moni tors, leve.l seD,Bors 3.:1.0 sodiulU
pod thermal sensors.
2.2.10.1 Fuel Assembl.y IDs-crumeni:ation2
Fuel J:.ssembly Instrumentation housed in tbe h:s'c.rument probes
incluc:1es:
Vortex Generator
Flow Straightener
Sample
.'Cddy Current Flovffitet,er
Thermocouples
Pt=_ll Tube,
The Vortex Generator, Flow Straightener and Sample tube are
attached directly to the inside dial".letet' of 1:he inst.rurnen't
probe. The Sample tube is connected to the FEDAL J.ea(L-out,
'!:ubE:.'s, a portion of th(~ i~'st,rllment tree. ':!'he latter three
2. Refer to References, Appendix A, I~ems 13, Refer to Drawings. Appnndi~ F. SK-3-12896.
14 and 1.5.
2-46
BNHL-500 Volume 31
CCI'1pOne::1ts noted a~ove are assembled dS a unit, and lc.ay be
extracted or inserted through the guide tube cf the instrument
treE; v;i thout removal of the tre'8 from th.e reaCT-or.
2.2.10.2 Control/Safety Red Instrumentation
Control/Safe ty rod ins tn:mlent~(l t:iOll lnc] '..lues :FJ ow indica.tion
and outlet t.e::Upt;:;!ra t·J.re. Inscru.l1lentcl tio"1 :i:or ccmtrols is
housed in a special package 'l))·.)ve the F'~l!~2. Instru:nent: probes.
The instruments are of~set from the reactor exteGsion rods.
FIOVJ is directe~ to the instrument package by the extension
rod guide tube. To assure that the poison section is not
lifted with the instrument tree, the instrument guide tube
length inserted in the duct must be reduced to a minimum value
so that the sweep may be performed to assure disengagement
prior to completing withdrawal of the tree from the fuel
assemblies. Mechanical interfaces at the control rod probes are
the same as those described for the fuel assemblies.
2.2.10.3 Open Test Instrumentation
In addition to the instrumentation described for the fuel
assemblies, flux thimbles and contact instruments may be
included in the open test assemblies. A maximum capability 1 instrument package is shown on the open test assembly.
Instrument lead out for these positions (including PEDAL) is
through the hanger rod to the reactor cover test connection
blocks.2
2.2.10.4 Closed Loop Instrumentation
Closed loop instrumentation is described in detail in the 3
CSDD for the Closed Loop System. Instrumentation interfaces
for the closed loop occur at the reactor cover.
1. 2. 3.
Refer to Drawings, Appendix P, SK-3-14S86. Refer to Drawings, Appendix P, SK-3-14461. Refer to References, Appendix A, Item 11.
2-47
2.2.10.5 In Core Flux Monitor positions
BNWL-SOO Volume 31
During reactor operation flux levels outside the FTR are
sufficient magnitude to allow monitoring to be performed with
out-of-reactor monitors. l During shutdown and startup flux
levels are too low for sufficient signal strength to be
generated outside the vessel. Therefore, during shutdown,
startup, and normal maintenance procedures in-core flux
monitors must be utilized.
To prevent the in-core flux monitor from being damaged during
reactor operation, the flux monitor will be withdrawn above
the core during operation.
Three positions with reactor cover access must be provided
during startup and shutdown. To prevent interference with the
instrument tree, locations on the radial test rows are utilized
as shown in Figure 2.1. Two of these are designated as flux
monitor locations, the third position would utilize the STIF
location for inserting a flux monitor during startup to provide
the required minimum number. During normal maintenance and
refueling, only two positions would be required.
2.2.11 Instrument Tree and Plug
In addition to containing the instrumentation for the driver
fuel and control/safety rods the instrument tree and plug
perform a variety of structural and fuel handling functions.
The operational sequence for instrument tree operation provides
the best description of the various functions the instrument
tree and plug perform. During refueling, the instrument tree
1. Refer to References, Appendix A, Item 15. 2-48
BNWL-500 Volume 31
is retracted into a position under the plug while the plug
rotates in conjunction with rotation of the in-vessel handling
machine to positions over the various assemblies being removed
and replaced. When the refueling sequence is complete, the IVHM
is stored in a retracted position under the plug. The plug and
instrument tree are then rotated simultaneously to position the
tree over the lattice positions which each individual probe
services. The instrument tree is then lowered with the longer
instrument probe located near the central core region engaging
the top of the fuel ducts and laterally moving these central
positions into a compacted configuration. l Further lowering
of the tree engages the next row of assemblies and locates them
in a tight geometry and so on until all rows are sequentially
assembled in a compact hexagonal array.
During this lowering operation, the intermediate control rod
extensions pilot onto the top of the poison section. Once
the tree is lowered into its final position, flow into the
tree FEDAL lines, and the FEDAL transition lines are sealed
by the bushing to each probe. Flow is then extracted from the
individual FEDAL line to each instrument probe rather than from
the pool.
The radial restraint mechanism is then actuated and the upper
extension for the control rod is lowered into position and the
collet latches at the two extension rod joints are engaged.
During operation, if a hydraulic balance feature fails at one
of the in-core assemblies, the assembly would be lifted by the
hydraulic force until the stop, provided by the tree holddown
1. Refer to Drawings, Appendix F, SK-3-l4604 and SK-3-l4606.
2-49
BNWL-500 Volume 31
plate, prevents further axial motion of the lifted assembly.
The plate is located with sufficient clearance relative to the
upper end of the assembly to allow for thermal expansion and
stainless swelling. Engaging the instrument holddown plate
causes a vertical displacement of the plates which actuates
an indicator attached to the upper end of the holddown plate
support column.
In the event of failure of a flow meter or thermocouple during
operation or shutdown, it will be possible to remove the
complete instrument module utilizing the pull tube attached
to the instrument connector. l Failure of the FEDAL system
components located in the instrument probe require removal of
the total plug assembly for maintenance or replacement.
Upon shutdown, the CRDM is removed and the upper extension rod
is detached simultaneously disengaging the lower collet on the
intermediate extension rod. The special tool utilized for
detaching withdraws the upper extension rod to a position flush
with the reactor head forming a gas and radiation seal at the
cover, penetration. The intermediate extension rod is now seated
on the poison section attachment column but is not attached.
The instrument tree then raises approximately eleven inches 2
which disengages all of the probes from their mating assemblies.
During this axial movement the control/safety rod continues to
rest on the in-core poison section. Once the probes are free,
the holddown plate rotates approximately 1/4 inch which places
the holddown plate clearance hole eccentric with the retaining
ring on the in-core poison section.
1. Refer to Drawings, Appendix F, SK-3-l2896. 2. Refer to Drawings, Appendix F, SK-3-l4606.
2-50
BNWL-SOO Volume 31
Lifting of the tree then continues which engages the ring on
the control/safety rod intermediate extension which lifts the
intermediate rod free of the poison section. Lifting of the
poison section in the event of sticking at the lower collar
is prevented by the eccentric configuration between holddown
plate and ring on the poison section.
During all phases of upward movement until all probes are
disengaged the instrument tree holddown plate remains stationary
to aid in disengaging probes from the assemblies.
Upward motion then continues until the support bushing/FEDAL
connection is disengaged and sodium flow through the instrument
tree lines ceases. The tree is then rotated into the storage
position utilizing simultaneous, programmed, rotation of the
plug and instrument tree.
In addition to the functional requirements noted above, the 1
instrument tree plug, which forms an integral portion of the
reactor head must also contain all anticipated pressure extremes
generated within the reactor, including the DBA loading. 2
Excessive thermal gradients combined with the precision location
requirements for the tree require a relatively stiff structure
fabricated from thin sections. 3 Axial support for the instrument
tree is from the instrument plug with lateral support provided
by both the lower seat and the instrument plug. An expansion
joint to provide for the differential thermal expansion must
therefore be provided at the lower support joint while maintain
ing the FEDAL seal function. 4
l. Refer to Drawings, Appendix F, SK-3-14605. 2. Refer to Support Information Requirements, Appendix B,
Item 4. 3. Refer to Support Information Requirements, Appendix B,
Item 2. 4. Refer to Support Information Requirements, Appendix B,
Item 9. 2-51
2.2.12 In-Vessel Storage
BNWL-500 Volume 31
The in-vessel storage positions are located at the periphery of
the vessel and provide storage space for decay of spent fuel,
temporary storage of green fuel and storage locations for other
core components removed during normal refueling. The latter
include radial restraint assemblies and control/safety rod
poison sections.
Subsequent studies will determine the sequence of refueling
providing the most efficient utilization of these storage
positions. storagel
positions are located in the ring girder
and may be reached by the IVHM. Three positions will accommodate
a finned fuel handling cask which is withdrawn and inserted
through a valve in the reactor vessel cover by the ex-vessel
handling machine (EVHM).
During refueling in-core assemblies are lowered in the finned
cask by the EVHM into the cask retaining structure. The IVHM
then attaches to the handling slot on the assembly and moves
the a$sembly to its preassigned core location. The IVHM is
then attached to a spent assembly which is raised and transferred
to one of the in-vessel storage positions or alternatively,
to the common position within the core barrel. The common
position may be reached by adjacent IVHM's allowing transfer
of an assembly around the vessel periphery for storage or
repositioning in another third of the core.
The in-vessel storage positions are cooled by by-pass flow
from the tertiary plenum which directly enters the inlet of
the assembly.
1. Refer to Drawings, Appendix F, SK-3-l4544. 2-52
BNWL-500 Volume 31
The in-vessel storage position configuration is a hollow
tube supported by the ring girder with a penetration into
the tertiary plenum. The EVHM position is larger in diameter
to accommodate a finned cask. The common storage position
is similar in configuration to the other storage locations.
In-vessel positions must be located in an array which precludes
criticality and minimizes coupling with the active core. The
positions must also be located with sufficient sodium space
from the vessel wall and core barrel to prevent irradiation
damage to these components during plant lifetime.
2-53
SECTION 3.0 SAFETY CONSIDERATIONS
BNWL-500 Volume 31
Safety considerations applicable to the design of the reactor
core components may be categorized into two general areas:
(a) Those considerations related to the prevention of potential
abnormal fault conditions, and (b) those considerations
related to the mitigation of the consequences of abnormal
fault conditions in the event of their occurrence. Emphasis
upon design characteristics and conservative margins to failure
for prevention of abnormalties is an obvious objective of
the core component design effort. Design features incorporated
to ensure mitigation of consequences to limit damage to
acceptable levels requires integration between the core design
effort, design characteristics of inherent and actuated
protective actions, and definition of the accident severity
levels based upon the estimated frequency rates of the various
occurrences (less damage is accepted for more probable
occurrences). Preliminary definitions of the severity levels
for various accidents have been established for the FFTF.l
3.1 aAZARDS
Hazards which require consideration in design of the reactor
core components are those which affect the core reactivity or
its cooling integrity. These conditions in turn affect the
capability of the fuel pin to effectively contain fission
gases and fissionable material. Specific hazards, but not to
be construed as all inclusive, are delineated as follows:
1. Refer to References, Appendix A, Item 3, Section 1.
3-1.
BNWL-500 Volume 31
A. Excessive radial movement of the core due to mechanical,
thermal, or hydraulic loads could have an adverse affect
upon the magnitude and sign of the overall power reactivity
coefficient.
B. A positive prompt power reactivity coefficient could
aggravate power instabilities and prevent adequate
mitigation of accidental power transients.
C. An excessive spatially positive coolant void coefficient
could lead to severe damage for accidents in which
spatial voiding is initiated.
D. An excessive local disturbance initiated within a single
core assembly could initiate additional failures by its
progression autocatalytically across adjacent core
channels.
E. Failure of assembly holddown during operation could
initiate excessive reactivity disturbances - especially
if the failed assembly moves suddenly into a more
compact core configuration.
F. Application of incompatible materials could initiate
excessive corrosion, mass transfer, or leaching of
materials leading to subsequent failures.
G. Foreign matter in the core inlet regions could restrict
flow through various portions of the core initiating
cladding failure and possibly fuel melting.
H. Insufficient thermal margins to failure thresholds
during operation could lead to premature failure of
the components for various core disturbances.
I. Momemtary stoppage or reversal of the core coolant
flow could initiate excessive local overheating and
cladding strains.
J. Passage of large volumes of entrapped gases through the
core could initiate excessive power disturbances by
its effect upon core reactivity. 3-2
BNWL-500 Volume 31
K. Excessive coolant temperatures could result in local
sodium vaporization or severe cladding strains.
L. Inadequate coolant temperatures could initiate coolant
freezing and subsequent loss of coolant integrity to
the core components.
M. Inadequate monitoring of core parameters and inadequate
protective action for off-normal conditions could initiate
component failures or excessive radiation damage and
subsequent failure.
N. Inadequate neutron shielding of core components could
lead to excessive radiation damage and subsequent failure.
o. Inadequate surveillance of critical components could
allow extended application of those components beyond
the designed allowable limits and thus lead to unantici
pated failures.
P. Insufficient removal of the core decay heat could lead
to subsequent fuel failures.
Q. Excessive vibration of core components could initiate
cyclic fatigue and subsequent component failures.
3.2 PRECAUTIONS
The design of the reactor core components will assure that
precautions are taken against possible hazards as delineated
in the foregoing section. Precautionary measures corresponding
to those hazards are as follows:
A. A core radial support structure will be designed to
limit core compaction with increasing power to assure
an overall negative power reactivity coefficient.
B. The core and fuel designs for the FTR will assure that
the prompt reactivity coefficient is negative throughout
the core lifetime and of sufficient magnitude to ensure
that severe accident conditions leading to core disruption
are within the design basis of the containment system. 3-3
BNWL-500 Volume 31
C. The design of the core configuration and components
will assure that the effects of spatial coolant voiding
are limited within the design bases of protective actions
and engineered safeguards.
D. To the extent possible, flow ducts will be designed to
prevent autocatalytic progression of damage to other
assemblies.
E. At least two independent methods of holddown for the
core assemblies will be provided by the design with
failure detection means provided.
F. Contacting surfaces within the core assembly will be
protected against damaging interactions. Fretting of
these surfaces will be considered in the component
designs. The design will allow adequate allowances
for erosion and corrosion for the coolant conditions.
Materials will be compatible with environmental conditions
and minimum end of life requirements.
G. The core inlet passages and assembly inlets will be
designed to limit the possibility of coolant flow blockage
by foreign matter or objects.
H. The core design will incorporate adequate thermal margins
by integration of the design overpower and overheating
factors with the designed protective actions.
I. The core will be designed with flow in an upward
direction for all operating and shutdown modes.
J. The core design will limit the passage of entrained
gases through the core to minimize coolant voiding effects.
K. The maximum coolant temperatures under transient conditions
will be limited by design to a level which does not result
in coolant vaporization or excessive cladding strains.
L. Minimum coolant inlet temperatures will be greater than
those at which sodium plugging occurs for the impurity
levels in the sodium. 3-4
BNWL-500 Volume 31
M. Instrumentation will be provided to detect off-normal
conditions in the core and to actuate protective action
where inherent protection is not provided by design
features.
N. Internal neutron shielding will be provideu to limit
damaging radiation to components to levels which will
allow cont.inueo. safe operation over their design lifetimes.
O. A program of surveillance and in-service testing will be
incorporated to assure predictable operation of the
components througrlOut their design lifetimes.
P. Adequate removal of decay heat will be assured under
all conditions. Fuel assenillly outlet coolant temperatures
as well as coolant flow will be measured.
Q. 'I'he fuel pin assemblies will be designed so that their
integrity will not be jeopardized by fatigue effects.
Vibration analyses and tests will De performed and the
results incorporated in the design.
3-5
SECTION 4.0 PIUI\CIPLES OF OPERATION
BNWL-500 Volume 31
This section outlines the step by step procedures which will
be required during the various phases of FTR operation. The
FTR operations may be divided into the following subdivisions:
Startup - the normal procedures employed in bringing
t.he reactor from shutdown to the normal operating mode.
Normal Operation - the primary operation mode for the
FTR, e.g., the irradiation of test specimens at prototype
flux levels LM?BR.
Shutdown - the normal procedures employed in bringing the
reactor from the normal operating mode to a shutdown
condition.
Special Operation - includes special test procedures
which may be required to verify special operating condi
tions or other changes in core loading.
Emergency - conditions not anticipated which can lead
to varying degrees of damage to core and reactor components.
The range of conditions would include a span of accidents
from pin cla6 rupture to massive fuel meltdown leading
to the Design Basis Accident.
4.1 STARTUP
Prior to reactor startup, perform and complete all prestartup
checks as detailed in t~e operational procedures. Check the
calibration, trip settings and functions of all instrument
systems, both safety and surveillance, and place these in their
startup operating mode. The number and type of instruments
which must be operative for startup to proceed will be specified
in other documents. Emergency power systems and emergency
cooling systems must be operating or in standby, depending
upon their normal mode, and the dynamic test loop.s must be
4-1
BNWL-SOO Volume 31
circulating coolant, and their respective instrumentation
calibrated, trip settings checked and functioning properly in
order for the startup to proceed. The various utility support
systems (gas, air, sodium, etc.) will be operating, as required,
during startup.
Verification of adequate flow and temperature indication on
every assembly channel is required before opera"tion at a
pmver level which mig-ht result in sodium vaporization or fuel
meltdown within a plugged channel. The initial irradiation
period for any fuel assembly must be performed at ~&rtial
reactor power to enable tile mixed oxide pellets within the
fuel assembly to form their characteristic annular configura
t1.on and to close the gap between the fuel cladding and the
pellet. Power increases during reactor startup are made on
an incremental basis. A complete survey of safety and
surveillance instrumentation is performed at each level of
increase to verify that all operating parameters are normal
and another incremental increase in power can be safely
attained. The rate of temperature increase will be limited
to values consistent with allowable thermal shock conditions.
Experiments, particularly initial operation of experiments,
may dictate special startup requirements.
4.2 NORMAL OPERATION
During normal operation, both safety and surveillance instru
ment systems must function prof'erly. Definition of required
instrumentation may be different than that required for
.startup. Emergen~y power systems and emergency cooling
systern must b8 operating or in standby, depending upon their
normal mode. The various utility and process support systems
(gas air, sodium, etc.) will be operating. Test facilities 4--2
BNWL-500 Volume 31
and loops will be operating as individually specified. The
plant availability goal is 75%. Planned. operating cycle is
nine weeks on line and three weeks shutdown for maintenance,
refuelLng, etc.
4.3 SHUTDOWN
There are two basic mddes of·shutting the reactor down, i.e"
planned shutdowns and those initiated in response to abnormal
operatLng condit.Lons. The latter may be divided into several
subcategoTLes, e g., manual shutdown, programmed shutdown,
power set back. and scram. Operatlng philosophy will be to
respond to lndications of abnormal operating conditions with
corrective actlon which is adequate to protect the plant. but
least perturbs the operation of the plant During plant
shutdown, the emergency power system and emergency cooling
systems wilL be operative. Certain components of the lnstru
mentation system, both safety dnd survelilance wlll remain
operatl-ng during shutdown periods. VarlOUS utility and
process systems ~gas, air, sodium, etc., must remal.n operating
during reactor shutdown periods.
4.4 SPECIAL OR INFREQUENT OPERATION
Test runs on the reactor ltself constitute special operatlons.
The nature of these tests will range from an extenslve test
program during plant checkout, initLal critlcality, and initial
ascent to power to periodIc measurements performed as part
of the Standard Operating Procedures. During testing periods,
the reactor may be operated at conditions other than those
specified for normal operations. Deviations from normal
operatlng parameters wlll be speclfied in detailed testing
procedLlres and will be examined for safety .lmplications 0
4-3
BNWL-SOO Volume 31
The insertion of experiments into the reactor will, on occasion,
necessitate special reactor operations. Such special operations
will most likely be required "during the shakedown periods for
new experiments. These special "operations may consist of
extended low power operations; restrictive heatup or cool-
down rates, etc. Such requirements will be examined for safety
implications and effect upon the overall program objectives of
the reactor.
Reactor refueling represents a condition which can be categorized
as a special operation even though the reactor is not critical.
Refueling is an operation requiring particular care. Emergency
power systems, emergency cooling systems, instrumentation
systems, and,gas systems must all be operative. The refueling
philosophy is to move only one core component at a time which
has an effect on core reactivity. The reactive condition of
the core will be known and predictable, within reasonable limits,
at all times.
4.5 EMERGENCY
A number of emergency conditions can be visualized ranging
in magnitude from an indication of high temperature, through
loss of site power, to massive fuel meltdown leading to the
Des.ign Basis Accident. It is essential that the instrumentation
system be as discriminating and "fail safe as possible so that
acknowledged abnormal conditions are controlled with a minimum
of damage to the reactor~ Signals which are received by the
instrument system will be assumed to be true and will be
acknowledged as such. The aCknowledgments required will be
specified as a result of detailed safety studies.
4-4
BNWL-500 Volume 31
If the response to an indication of abnormality can be
identified and corrected operations may continue or be resumed.
If failure has occurred, the resultant action will be determined
either by standard procedures covering such a failure or by
a series of ad hoc analyses and-decisions if the action' is
not covered by procedures.
4-5
SECTION 5.0 MAINTENANCE PRINCIPLES
BNWL-500 Volume 31
The basic principles of maintenance on the reactor core
internals is modular removal and replacement. The modules
are moved from the reactor to another area for repair or
disposition. It will be impossible to perform maintenance
on core components in position with the reactor operating.
Only minor repairs and/or adjustments will be performed on
core components in place while the reactor is in a shutdown
condition. The modular replacement philosophy is consistent
with the goal of maximizing reactor operating time.
Reactor core components can be divided into two categories:
1) those which are replaced systematically, e.g., driver fuel,
reflectors, restraint beams, control elements and 2) those
which are essentially fixed ~tems, e,g., shield assemblies,
core clamping mechanisms, etc. Failure of one of the former
components would lead to the early replacement of the failed
module. Failure of one of the latter components would
probably lead to an extensive reactor shutdown. Removal of
one of the fixed components which has failed would require
removal of a refueling port or the reactor head. Spares
for fixed components will be selectively available. Therefore,
an evaluation of each failure or problem would be assessed
to determine whether fabrication of a replacement unit,
removal and repair, or devising a technique for in-place
repair would be the most feasible. The failure rate for
fixed components is expected to be small. They will be
designed for long life in core, most are static components,
and, in most cases, they are non-load bearing.
The components which are systematically replaced are expected
to have a much higher failure rate than the fixed components.
5-1
BNWL-500 Volume 31
The driver fuel assembly, in vessel fuel handling equipment,
and associated instrumentation represent the most critical
subsystems of the reactor core and the ones most likely to
fail. If a fuel failure of a magnitude which activates
a number of sensing elements occurs, the failed element
will be located and removed from the core for inspection,
analysis, etc. If a failure occurs which activates one
or more alarms for a single assembly, isolation of the fault
location will be attempted and replacement of the sensing
elements will be performed during the next outage.
The need for maintenance on routinely replaceable core
components will be established by surveillance or inspection.
Since routine visual examination of in-place core components
will be virtually impossible, process surveillance instru
mentation which provides trend information and advance
warning of developing problems must be relied upon. Neutron
flux, flow, and temperature distributions are the speciflc
in-core conditions that will be monitored. Periodic functional
testing of critical core components, e.g., rod drop time
measurements and safety system circuitry checks, will be
performed to assure continued operation within specified
design limits. Surveillance requirements and techniques
for core components which are not easily replaced have not
been defined but are under study.
5-2
APPENDIX A
REFERENCES
APPENDIX A
REFERENCES
BNWL-SOO Volume 31
1. A. F. Lillie, FTR Fuel and Core Parameter Studies, BNWL-II08, Battelle-Northwest, Richland, Washington.
2. W. W. Little and L. L. Maas, Nuclear Parameters and Parametric Studies for the Fast Test Reactor (FTR) , BNWL-I067, Battelle-Northwest, Richland, washington.
3. Design Safety Criteria for the Fast Flux Test Facility, BNWL-823, Battelle-Northwest, Richland, Washington, June 17, 1968.
4. R. A. Moen, FFTF Materials Design Data, BNWL-891, Battelle-Northvlest, Richland, Washington, October 1968.
S. G. R. Waymire, Bases for Reactor Core Design Requirements, A-OIO S, Battelle-North\V'est, Richland, l.qashington.
6. P. K. Telford, Design Survey and Conceptual Definition of a Reactivity Oscillator for the FFTF, BattelleNorthwest, Richland, Washington.
7. D. P. Shively, Conceptual Component Design Description for the Reactor Vessel and Shield, Component No. 32, BNWL-SOO, Vol. 32, Battelle-NorthvV'est, Richland, Washington.
8. D. Marinos, Conceptual Component Design Description for the Reactor Nuclear Control Components, No. 33, A-0048-R3, Battelle-Northwest, Richland, Washington.
9. P. D. Cohn and E. G. Stevens, Conceptual Component Design Description for the First Core Fuel Assembly, Component No. 3S, BNWL-SOO Vol.3S,. Battelle-Northwest, Richland, lvashington, August 8, 1969.
10. E. Ruane, Conceptual System Design Description for the Reactor Refueling System, No. 41, A-004S-R3, BattelleNorthwes t , Richland, ItVashington.
11. M. K. Hahaffey, Conceptual Component Design Description for the Closed Loop S¥stem No. 61, BNWL-500, Vol. 61, Battelle-Northwest, Rl.chland, Washington, December 19E8.
A-I
BN1vL-500 Volume 31
12. Letter, Milton ShaT'; to D. C. Williams, "FFTF Prograrr. Direction," January 28, 1966.
13. W. Dalos, Conceptual System Design Description for the Reactor and Vessel Instrumentation System No. 92, A-0052-R2, Battelle-Northwest, Richland, Washington.
14. Prepared by Westinghouse. Conceptual System Design Description for the Fuel Failure Monitoring System No. 94, C-OOOI-R, Battelle-Northwest, Richland, Washington.
15. L. W. McClellan, Conceptual System Design Description for the Flux Monitoring and Control System No. 95, A-0056-R, Battelle-Northwes~, Richland, Washington.
16. C. C. Steele, FFTF Overall Conceptual SJLstems Design Description, BNWL-500, Vol. 1, Battelle-North\<lest, Richland, Washington, July 7, 1967.
17. ,J. J. Holmes, Fast Reactor Induc'3d Swelling in Austenitic Stainless Steel, Bln~L-SA-2126, Battelle-Northwest, Richland, 'Washington, Decernber 1968.
18. C. Cawthorne, E. J. Fulton, "Voids in Irradiated Stainless Steels", Nature, vol. 216, p. 577, 1967.
19. R. C. Walker, FFTF Reference Concept, Summary Description, Bln>\]L-9 55, Battelle-Northwest, Richland, Washington, January 1969.
20. E. R. Astley, FFTF Quarterly Progress Report: June, July, August, 1968, BNWL-917, Battelle-Northwest, Richland, Washington, December 1968.
21. J. P. Thomas, Conceptual System Design Description for the Plant Protection System, A-OIOI-R, Battelle-Northwest, Richland, Washington.
A-2
APPENDIX B
SUPPORT INFORMATION REQUIREMENTS
Item
1
2
tJj I
f-'
3
4
5
6
7
8
.A:PPENDIX B
SUPPORT INFORMATION REQUIREMENTS
Information Required
Determine the required minimum Doppler coefficient
Establish the maximum inlet and outlet temperature transients
Establish number and type of initially installed loops
Define the Design Basis Accident
Determine the effect of corrosion
Establish fuel performance limits and determine failure mechanisms under transient conditions
verify hydraulic design of FFTF fuel assemblies
Establish that critical heat flux is > twice the peak heat flux of the reactor. (Heated Pin)
Type of Effort
Study
Study
Study
Study
Experimental
Experimental Investigation
Experimental Investigation
Experimental Investigation
Information Source
BNh'
BNW
BNW
BNW
BNW
BNW
BNW
BNW
When Required
Early preliminary design
Early preliminary design
Mid-term preliminary design
Mid-term preliminary design
End of preliminary design
End of preliminary design
End of preliminary design
Mid-term detail design
Item Information Required
9 Determine effect of irradiation on shield materials
10 Component and System Reliability Analysis
11 Determine core and outlet plenum vibration characteristics
12 Investigate core restraint mechanisms -More Concise-
13 Determine radiation effects on metals, -More Breakdown~
14 Determine surveillance and in-serviCe testing and inspection requirements for core components and materials
15 Determine allowable fluence limit on components
16 Determine maintenance procedure for remova 1 of semi-permanent components (tubesheet, shield assemblies, etc. )
17 Determine optimum sequence for applying the radial
~ restraint load , tv
Type of Effort
Study and Experimental Investigation
Study
Study and Experimental
Experimental
Experimental
Study anCi Experimental
Study and Experimental Investigation
Study
Experimental
Information Source
BNW
RPD
BNW
BNW
BNW
BNW/ RPD
BNW
RPD
When Required
Mid-term detail design
Early preliminary design
End of preliminary design
Mid-term preliminary design
End of preliminary design
Mid-term detail design
Check by beginning of preliminary design, Experimental Confirmation by end of preliminary
End of Preliminary design
Hid-preliminary
Item Information Required
18 Determine procedures for core access in the event of instrument tree/core assembly galling or self welding
19 Determine orificing system for the core
to I w
Type of Effort
Study
Study and Experimental
Information Source
RPD
RPD/ BNW
When Required
Early Final Design
Complete by end of preliminary
APPENDIX C
INTERFACES
Interfacing System Item Number Title
o I
I-'
1
2
3
4
5
6
24
27
32
33
35
41
Radioactive Waste System
Reactor Containment System
Reactor Vessel & Shield Component
Reactor Nuclear Control Components
First Core Fuel Assembly Component
Reactor Refueling System
APPENDIX C
INTERFACES
Interfacing Area
Nature and quantity of both routine and emergency wastes created by the reactor core places size and type requirements on the Radioactive Waste Facility
Interface between ring girder, removable inlet plenum and reactor vessel and between test positions, instrument tree plug, radial restraint and reactor vessel head. Nature of the reactor core and in-vessel storage imposes size requirements on the reactor vessel.
Interface between core lattice configuration and control elements. Functional interface between core nuclear characteristics and control system characteristics.
Interface between core lattice and fuel assembly. Interface with radial restraint loading system, instrument probe and hydraulic balance receptacle. Nuclear and thermal hydraulic functional interface.
Interface between in-core assemblies and in-vessel handling machine, and between in-vessel storage and the IVill1 and EVHM. <: tJ1 o Z
1-':::8 s:: Ll S I ro Ln
o WO I-'
Interfacing System Item Number Title Interfacing Area
7
8
9
10
11
12
13
14
() I
IV
43
44
51
61
71
81
82
91
Irradiated Fuel Handling System
Central Maintenance System
Reactor Heat Transport System
Closed Loop System
Inert Gas Cell Examination Facility
Functional interface through size and configuration of core components.
Configuration of core components determines sizing of maintenance casks and radioactive maintenance areas.
a) Functional interface between core coolant requirements and heat dissipation capacity, including pressure, flow and LMTD functions.
b) Functional interface through primary sodium volumes and sodium flow paths in the core. (For ECCS design)
Interface between core lattice configuration, radial restraint, and closed loops.
Functional interface through size of core components and interim exam characteristics.
Sodium Receiving and Functional interface through sodium purity Processing System requirements.
Inert Gas Receiving and Processing System
Functional interface through gas purity requirenents.
Central Control and Functional int~rface through control and surData Handling System veil1ance requirements of core,
Interfacing System Item Number Title Interfacing Area
15
16
17
18
19
20
n I
W
92
93
94
95
96
99
Reactor and Vessel Instrumentation System
Plant Instrumentation System
Functional interface through control and surveillance requirements of core. Physical interface between fuel assemblies and in-core instrumentation sensors.
Fuel Failure Monitor- Interface in the instrument probes and in-vessel ing System FEDAL lines with the FEDAL vortex generator,
straightener and pickup tube. Interface at the reactor cover with the ex-vessel FEDAL system.
Flux Monitoring and Con~rol System
Radiation Monitoring Systerr.
Plant Protection Systerrl
Interface at the lattice positions continuing the flux monit:or. Functional interface through the reactor control systeFl.
Function~l interface through control requirements of the core.
Nature of reactor core and potential failures places functional requirements on Plant Protection Systerr: .
APPENDIX D
FFTF DESIGN DATA SUMMARY
APPENDIX D
FFTF DESIGN DATA SUMMARY
BNWL-500 Volume 31
SUMMARY OF REFERENCE CONCEPT CHARACTERISTICS AND DATA
A. General Plant Data
1. Core arrangement
2. Design life
3. Peak flux
4. Total power
5. Reactor coolant
6. Core volume
7. Core coolant flow (Bypass not included)
8. Pressure drop (design maxima) fuel assembly (includes instruments)
9. Reactor (nozzle to nozzle)
10. Reactor bulk inlet temp. initial core design maximum design maximum (Advanced cores)
11. Reactor bulk outlet temp. initial core design maximum (Advanced cores)
12. Core temperature rise avg. initial design maximum (Advanced cores)
13. Reactor cover gas
units
years 2
nlcm Isec
MWt
liters
Ib/hr
psi
psi
Values
vertical
20
~7 x 1015
400
sodium
1033 7 1.5 x 10
120
145
600
900
900
1200
300
400
Argon
D-l
B. Reactor Vessel
1. Diameter
2. Height
3. Wall thickness
4. Wall fluence, total
5. Material
C. Core Design
1. Number of core lattice positions
2. Number of driver fuel subassemblies
3. Number of closed loops
4. Number of contact instrumented in-core open test positions
5. Number of proximity instrumented open test positions
6. Number of in-core safety rods
7. Number of in-core control rods
8. Number of peripheral control rods
9. Equivalent core diameter
10. Active core height
11. Reflector material axial radial
12. Reflector thickness axial radial
13. Fuel pin heat transfer area
14. Pin bundle coolant velocity, max.
15. Direction of coolant flow
Units
feet
feet
in.
nvt
inches
inches
inches rows
sq. ft.
ft/sec
BNWL-500 Volume 31
Values
17
55 I 9"
2
10 21
304 88
91
76
6
2
1
3
3
15
46.8
36
stainless steel nickel
6 3
2980
30
upward D-2
D. Driver Fuel
1. Fuel composition
2. Cladding material
3. Linear heat genera~ tion rate, average
4. Overpower factor
5. Peak linear heat generation at overpower (equivalent hot channel)
6. Target burnup, average
7. Target peak
8. Cladding heat transfer coefficient
9. Max. expected temperature (Steady-state)
10. Fuel assembly length
11. Fuel geometry
12. Pin diameter
13. Spacer wire diameter
14. Number of pins per assembly
15. Subassembly crosssection outside dimension
Units
KW/ft
KW/ft
MWD/Tonne (metal)
MWD/Tonne (metal)
2 Btu/hr-OF-ft
ft
inches
inches
(across flats) inches
16. Lattice spacing inches
17. Duct wall thickness inches
BNWL-500 Volume 31
Values
18-25 vol% Pu0
2 75-80 vol% U02
316 SS 20% cold worked
7.3
1.20
18.0
45,000
80,000
37,500
4,050
14
Triangular pin cluster
0.230
0.056
217
4.615
4.715
0.140
D-3
E. Physics Data
1. Delayed neutron fraction
2. Neutron lifetime
3. Doppler
4. Power density (peak)
5. Power distribution (peak/average) radial axial total
units
seconds dk
T(dt)
MW/liter
F. Test Facilities
1. Closed loops -number -location
-power handling capability
-test flow rate
-test section outlet temp.
-test section length
-test section diameter
-pumping head-primary
-test section pressure drop, maximum
-material (in-core tube)
MW
l/min
inches
inches
lb/in. 2
lb/in.2
2. In-core open test positions
-number with contact instrumentation with proximity instrumentation(a)
BNWL-500 Volume 31
Values
0.003 -7 4.0 x 10
-0.004
0.75
1.40 1.24 1. 74
6 2 adjacent to core center
1 (~ mid-radius
3 @ core periphery
3-5
30-350
1,400 (bypass flow permitted)
36
2.5-3.0
250
90
316 88
2 1
(a) One proximity instrumented test position will have superior instrumentation accessibility. Driver fuel positions can be used as open test positions with standard driver instrumentation.
D-4
2. cont'd Units
-power
-coolant flow rate
-test assembly length
-test assembly cross section
-coolant
3. Short-term irradiation facility
-type
-number
-location
-minimum irradiation time at constant flux
-sample length
-sample cross section, maximum
H. Shielding
1. Within reactor vessel
-material
-configuration
-thickness
2. Cover Shield
-material
-thickness
-coolant
minutes
inches
inches
inches
feet
BNWL-500 Volume 31
Values
Same as driver fuel assembly
Sodium
trail cable
1
core periphery
1
TBD
2.0
stainless steel
hexagonal assemblies
21.2 (4.5 concentric rows)
Low alloy steel
4
Argon
D-5
APPENDIX E
ALTERNATE CORE DESIGNS
APPENDIX E
ALTERNATE CORE DESIGNS
ALTERNATE CORE DESIGNS
BNWL-500 Volume 31
Variations in performance characteristics for the core such
as fuel peaking factors, fuel melting temperature and neutronic
parameters can result in significant changes in the ability to
meet the requirements noted in Section I such as peak flux
and total power.
To provide for these variations two alternate core arrangements
are included. The first configuration is simply an extension
of the present reference core map as shown in Figure E.l.
By utilizing the first reflector row of the reference core map,
an additional 15 fuel assemblies may be added to the core. The
reference plug and instrument tree are positioned so that access
to these positions as well as all positions out to and including
what is the first shielding with the reference concept is
possible without any modification to the reactor cover and
plug arrangement. Replacement of the radial restraint loading
bar would be required. This arrangement provides the contingency
for adding additional fuel with a minor modification to the
overall design.
Another alternate which is discussed in detail in the Parametric
studyl utilizes an arrangement of test positions located on
true radial tri sections. This cross section, shown in
Figure E.2, provides an additional test position at the center
1. Refer to References, Appendix A, Item 1.
E-l
I:Ij , N
0 @J @
• ~ ..
• 0 e 0 e
FIGURE E-l. Alternate Core Map
OPEN LOOP WITH PROX. INSTR. - 1
OPEN TESTS - 2
CLOSED LOOPS - 6
SAFETY RODS - 3
REFLECTORiRESTRA INT POSITIONS - 42
PER I PHERA L CONT ROL RODS - 15
REFLECTORS - 88
DRIVERS - 91
IN CORE SHIM SCRAM RODS 3
FLUX MONITOR POSITIONS - 2
COMBINATION FLUX MONITORS STIF POSITION - 1
<to o Z 1-':8 s:: t:-' ;3 , (!) U1
o LV 0 I-'
M I
w
SR
• OT
CR
CL
PC
R
r
0
FIGURE E-2. Alternate Core Layout
Safety Rod 3
Flux Monitors 3
Open Test 4
Control Rod 6
Closed Loops 6
Peripheral Control 6
Radial Restraint/Reflectors42
Reflectors 63
Driver Fuel 84
<to o Z 1-'::8 s:: t""i ;:l I CD Ul
o W 0 I-'
BNWL-SOO Volume 31
of the core. The design shownl
utilizes 9 in-core safety/
control positions along with six peripheral controls.
With the arrangement shown, the central test position cannot
be an interchangeable closed and open configuration. Access
is limited by the safety rods which rules out a closed loop
at this position. By eliminating the central safety positions
the central loop can be made interchangeable, but would
probably require the use of additional peripheral controls
which would move the plug back and increase its diameter
(a smaller plug is one of the advantages of this desLgn .
Piping access to the test positions are along both sides which
reduces the number of available peripheral control positions.
Summarizing advantages and disadvantages for this alternate:
Advantages
Reduces plug diameter
Reduces instrument tree size
Reduces IVHM arm length
Reduces vessel diameter
All three trees become symmetrical
Provides one additional open test position.
Disadvantages
Utilizes more in-core control which can result ~n local
flux depressions
Flexibility inherent in peripheral control is unavailable.
1. Refer to Drawings, Appendix F, SK-3-14S89.
E-4
APPENDIX F
DRAWINGS
SK-3-12896
SK-3-14251
SK-3-14433
SK-3-14434
SK-3-14435
SK-3-14436
SK-3-14461
SK-3-14499
SK-3-14515
SK-3-14521
SK-3-14S44
SK-3-14545
SK-3-14560
SK-3-14567
SK-3-14568
SK-3-14570
SK-3-14581
SK-3-14585
SK-3-14586
SK-3-14588
SK-3-14589
SK-3-14600
SK-3-14604
SK-3-1460S
SK-3-14606
SK-3-14607
SK-3-14636
-....
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I A' SECTION Z1:..-
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SODIUM LEVEL
INSTRUMENT POST
hf)1V1O<JAL F£DAL TUBES
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PARTIAL £L~VATION
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tTYA'CAL 1<08£ _TRUAIEIff P
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DETAIL [
FULL SIZE
l' I I
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rjc SPIDER SUPPORT
SCALE ;>X' FULL
FLOWMETER EDDY cuRRENT
STRA/GHT£lIl£R TOP OF"FLOW
DETAIL .II
FULL SIZE
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CURVED GUIDE TUBES
CX/PL E LEADS PAL THERMoe PULL rUBE, s;;:,OUND CENTR~f.vOLUTlON PER APPROX ONE._ THREE FEET
GRAPPC.E GROovE
DETAIL !ll
T SURFACES CONTAC
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GUIDE TUBE
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HDLDDOWN FOOT
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'. r
LEGEND
~_-1 PARTIAL PLAN ABOVE REACTOR COVER
FIXED STORAGE 16 POSITIONS TYP l PLACES
PARTIAL BELOW REACTOR
PLAN
COVER
e e 0 e e ~ ~
e r~"" ~ rs~~i "J
CORE RESTRAiNT MECHAN 15M HP 6 PLACES
o I" !
LOOP - CLOSED
TEST -OPEN
TEST - OPEN WITH PROXIMITY INSTRUMENTATION
CONTROL ROD - IN CORE
CONTROl ROD - PERIPHERAL
CONTROL ROD - SHIM
DRIVERS - OPTIONAL PERIPHERAL
FLUX MON I TO R
SHORT TERM I RRADIATION FACILITY
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FINNED POT LOCATION TYP 3 PLACES
CLOSED lOOP TRENCH TVP l PLAC Es
FEDAL CROSSARM PIPING DISCONNECT CATCH TRAY
PLUG
CONCEPTUAL
THIS DRAWING COMPLIES WITH THE DEC 1968 CONCEPT GUIDELINES
th- ISSUE au·· ISSuE
I ST ISSUE
5-13-69 3-28-69 2-7-69
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LOOP CHASEWAY
SECTION C-C
SECTION 8-a
THIS DRAWING COMPLiES WITH THE DECEMBER 1968 CONCEPT GUIDE LINES
CONCEPTUAL
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fA· =-~ __ ---=1
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".120 DIA X D.187wALL
U IS"------I-------I--- 3"t-1 --+------
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SECTION B-B SECTION C-C
!l II 10
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CONCEPTUAL THIS DRAWING GO'*LIES WITH THE: CEC 1988 C.ONCEPT GUIDELINES
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INSTRUMENT PRCt3E SUPPORT PLATE
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~ B.J~ ~ r"c--c~ '~4 ~
-I ~~
~i ~-'- I z I' 271. ·1, a"-a ,. I __ 4_
10'- Z· I ..
~----------------------------------------------------------------------------------------------IO-Z·------------------------------------------------------------------------------~I
r--------------------------------------------------------------e"-ei"------------------------------------------------------------------------------~ D~
0.180
SECTION A-A SECTION 8-8
l il
Q
SCALE
0.500 0.0. (III TYP)
2 3 4 ~ 6 IN.
SECTION C-C
o 2 liN.
SCALE: eL:1ooL ..... ~....,"""=====~
0.140 WALL
-- :' ...... CNC. DRAWINGS
"UEL HANDLING SLOTS (6 PLACES TV?)
4.615
SECTION 0-0
DRAWINQ STAT'.JS
..
LOADING PAD (3 PLACES TVP)
NOTE
I. THIS DRAWING COMPLIES WITH THE DECEMBER IgS8 CONCEPT GUIDELINES.
ilz", ISSUE 2-Z0-69
CONCEPTUAL ". III ISSUE 2-3-69 -=--==:' u. S. ATOMIC ENERGY COMMISSION IItICHLAND OPERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY :. OP.:IltATll:D .Y .... TT.u.... M'Ih401'lIAl. 11'l'!llTTTUT'I
SHIELD ASSEMBLY
FAST FLUX TEST FACILITY ~;. ~-300 GEN ,._- 2200.01
Sl<0I0~
SK-3-14499
_ I
•
:I
c
A
S T SHEET LINER TO lLQL SPACER mcr Nan 4 u Ums
SECTION 8 - 8
WJX W N I T O rnlMBLl
REMWABLE SHR WEDGE LOCKS SlXTION TO FAC FOR MCUNTING INTERIM E X A M . RhlWABLE SHROUD
SECTION
I
SECTION C - C
I GRIDS
SECTION D-D S E C T I O N E - E
a i t o p OF PRESS WPER LIMIT OF TOP OF
TUBE SHROUD $8' REMOVABLE SHROUD s 6. M L ZONE --I
LOWER LIMIT OF SINTERED SST TANTALUM LINER REMOVABLE SHROUD 0- WUUTH)N ron MELT-DOWNT
1 - 1 f
PRESSURE TUBE SHROUD
•
,
o
Il T
','
12
INERT GAS
ELEC - LEAD TO TRACE HEATER
II 10 • PERSONNEL SHIELD
: \ t- . . ~--.---] ¢=' ~_~--- I i
------r--~- '= · r-i ------ .~--------..;:-~~--. __ t_.
I 1- --- -- -- -r
1=, ---==-=--=-~-~ c2;-r r--' ---~ ---~-=------..
-Ijf----- --------.---- ------------ ---~+__
~--. .~--------=, ~l \i ~~IGil I~ '" STLF REACTOR COVER PLUG ARRANGEMENT
I: 1O~7' I SEE REACTOR COVER PLLG --r ...... ----------
CORE MAP 1eAL..('" .I <I • It ";2 ~ ...,
REF' SK-l- i4!1t57
•
F'Ef'FCRATED CARRIAGE BODY
SEC110N A-A ruu. SCALE
.---------------------- ------ ------------------------------1 34:0'
---------------------------------------i
1 I AARANGEMENT ABOVE ! i'I -- ~~-~'1~~--' ~~~~-~~-t~~~---~--------:J t, ~PANSION
~- . - - __ -= ____ -~ - ___ ~ l+ __ ~~~~_~~::".p::'!ATOR . ------- ~--.--
~·-o ... ______ _______ _ __---.£.O .. --'·COI'O='-'E'----________ ~~~"_'O'_· ______ ____1 r I
,-1 __ --J __ ~~-E3--) -, --=-
! iBYPASS ~ cRESTR!CTION ORlnCE
~=;;:~-='-:--=:-~--===~~-"=-"'~~~~±fr' ! ~--:-*,*t;>--,
S H 0 R T T E R M R R A D A T o N F' A C L T Y
N-R E ACT 0 R TUB E A R RAN GEM E N T
..
..
NOT..L SoT I.~. IMP05ES A RESTRIC TION r:F 2 CLOSED LOOPS MAX . • M OPEN/CLOSED LOOP TES T BLOCK_
!THIS DRAWING COMPLlESII ITH THE DECEMBER 1968 NCEPT GUIDE LINES
CONCEPTUAL ~I IV ISSUE I·2e-69
--""""----- R.R. OERJ~£AU
'NONE
U. S. ATOMIC ENERGY COMMISSION "ICHLANO OFiERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY OPtIlATl:D .Y .... TTI[u... IoII!MOAtAL I .... STlTUTE.
S:r.I.F. IN-REACTOR TUBE
ARRANGEMENT FAST ~LUX TEST ~ACILITY
300 -GEN, 1'--- 2200.01
o
C
A
~~ ____ ~1~ ____ ~ ____ ~1~ ____ ~ ____ ~':' ______ ~~~!-~ __ ~~ ____ :-____ ~ ____ ~ ______ ~ ____ ~ ____ ~~----~----~------~----~----~.~----~----~------~----~-------------------1r--
o
c
A
COMMON FUEL TR~VSFER POSITION TYP FOR jJJ AT Uf7 - PERM! TS FUEL SHUFFLINf:3 BETWEEN J VH M·
'lIEl. TRANSFER F!N/v=~ por·---_. (FOR CHARGE-OISCHARr..E)
C'C1RE RA2/AL REST'flAINT----.-__ !J IJECHAJoSM- ryp FOR (~)AT 00'
FIXED IN~VESSEL -JE:" ------_.iL. STOQAGE POSiT/eNS(66) TOTAL
FlOATING SHIELD COvE,'? --PLATE - SELF CEVT£RING ON CORE I'YI-t£N ,'flAD/~L Il£STRAINT IS ACTUATED
FEDAL D/SCOII/,VEer -4NO -----.;1---__ RECEPTACLE ~OR :NSTRI./M5VT TR£E - 80TH RAD/AL RESTR·HVT AND F£DAL IN-VESSEL TR4/1SITION ARM ARE SUPPQRTE;J ON THE SHIELD COVER PLATEN TURN THE SHIELD COvER PLATE IS SUP.~ORTED ON THE CORE BARREL
PLAN AT tJ-8' .---... SEE SK·')·14545 (EL (-1)(6;
REFLECTOR POSITIONS ThAT ,4CC£PT ----~--------~~~~ BY-PASS FLOW FROM I!'.'LET PLE.~.:"U
RECEPTACLES FOR SHIELD a..E'-fENTS
FINNED POT STOR.4./£ R£C£PTACL£-----~~-------
JNSTfiHJMENT ~_
•
IN-VESSEL FUEL HAlVDUNG VACHINE
SIZE IN R~ACrr:IH HEA/)
PLAN AT i4 -A· /" SEE SX-3'14545 (EL I-II/O")
, _1#1f tt.~~' ~f
__ ---~~_----ACTlJ.ATlNG SHAFT TO RADIAL RESTRAINT
·--'-~~c-_';;""-+=-:I+----t-+-- ---1.----0---- -------I~ ~#ff!/f-!:-'(Jy~_T_- 0 PO" (J) AT 1<0'
\
CORE SAPREL SUPPOPT 8Q:(EO RING GIROCR, SUPPORTED ON SHEAR LEDGE OF REAC TOR VESSEL .... FIXED IN-VESSEL FUEL STORAGE AND POT TRANSFeR POSITIONS BUILT INTO rHIS GIRDeR
8'r"-PASS FLON ANNULUS
REACTOR vesSEL ".I4LL
THIS DRAWING COMPLIES WITH THE DECEMBER 1968 CONCEPT GUIDE LINES
CONCEPTUAL ----~ --~-
U. S. ATOMIC ENERGY COMMISSION "rlCHLAND OPERATIONS OFFICE
PLAN AT "C-C· PACIFIC NORTHWEST LA80RATORY 0P'VtA1"ED .v .ATTIIL.1....l: MEMORIAL IN5TfTUTI!
SEE SX·3·14545 (ELt-J 40:0j REACTOR VESSEL
It:: =tl ~±=l.=.HC.=O"A=w .. a=. =~~I-J+-,,:-.:::-+--~-l-L-~:.I-'--h-I ::L~.-o':.:_. ~-.,""-. ~ ---11-:' 1=~~~-=i~l_C~OM='P;c:;O~S/;:;;:;:TE~A='L_;;:;;AN-'S_t
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SOOIUN JNf..£7 u'SCH «lS PIPE, TYP FOR (e)
REACTOR a£"RFLOW Ttl SURGE TANI{ VIA REGEN£RATIVE fEAT £XOiANGER
n
REMOVA8U INLET Pi. ENUM WITH CO~MtLTMDOWN DISPERSI I GRIO
) HIGH PRESfURt" ;roNE-(
I
CONTRot ROD DRlWS
.REACTOB._ ELEVATIDf;/.
II
TYPICAL HANGER liEY BLOCK
SDDI/M 0117/.£7 ,.'sa; 4DS PlP£, Tm FOR (.V
...
me « U,fSSfW"1 lL ... .11:. ..
TOP CZ SHIELD , -"!!(.~ "~'.J. ~ ~
GUARO VESSEL, ZI'" all -,' *LL
CORE BARREL, g'.,' III _,,'WALL
INOIVIOI.1AL HYDRA/JLIC _ANC£O INLET MC£PTACLES
-L---:!---SHEAR LEor;E, RIH(; filRO£II SIJPPOIIT
RING r;IROtR, CORf IJARR£L SlJPPQRT
IOTrpM or M<ltW m;sa A .,g',.
•
'7 I. to. •
.------------------~--------~------~--------~--~~--~------------~------~~------~--------~--~----~--~--~----. - ,
~ _______________________________ ~Z4~·~R~E~·----------------------------------~!--------------------------------------------------------------1~4!OVERALL TIP UNGTH--------------------------i 1. __ -----------------------------------------4Z.'--------------------'~------------------------~~4-----CORE~ __ V--'OP OF ~RID PLATE __ .. < , Jt-r'-~-<--<-<-,-,-«------<--, --,--<-.-. -------~--I---l.'STROKE-------------------------------t-----2---l' 'A'~ --I I-
~ ~i+-
--
O©©,©t:) .~ ~:-~ ©©@©©©O©~ ; t-'-eP.~.O ©el2Jf:;J~ ~:_L-. f$p ©~©teJe;., 1-
I, \i.~ ~ !(i ~ © 0 ©~vljf)r'1 ~!-+---~ .. ,.
r:--~
~£-
_-t : -
I-:-..=. I-
-.-r-=p~ \~I . :~ ~ ~-----=--=----------+1 . O©f)©@©@"iti-+-----sr.:::< 1~-~ - ~~ ~~. - -+ @@© if ~
I. .I-oI.-----------------------------~-----------37{"EFLECTOR FOLLOWER---------------tl--------~· B4 C I I
__ 'Ao~ j: 1-------------------------------------------------------------------------------------------------------------14'-0" OVERALL ROD DUCT LENGTH ------------------------------------1-1
F
• ~~®("---~~I E T---------------- 12'-4rOVERALL TIP LENGTH-----
r-----36-CORE·------------i1 n- - I ~. r--- zf =1· I{-j
_ ~ ~: .!!!fffl!!"":!!':1111~r ~>J't~ !~""iiEJ;~" "" On m nO""" m' ~=m,,"'" m ,;gr: :;' "~"I ~ E;;:::; :: ~: ~:'"' ';;;3-f I +r=- 1~_L =-< ----- _____ ~~~+---\--__ -----'------------r.:--J!j~:,:.:llilil"d."'.~!~rD-IA__l.--I ~ . -'. + +-r~6' t-- I 3, tal ~~.--. -. --.-. -?-."'\ _____________ --1 ~ - -lIIWlllllililliPlllililiil.// I
~.. . 30· t C=-r-,H _ !~ ... ! .:::==~ :,.;.." ' ... ~ . ~ 'L... _____________ ...J. ____ ....Lt
D
: ~ ~§j .. t, f i ~L.-,-._ I ~}. ~-=-~ . . -1.L~. -- - .. >. Y :; =-=:=::=] ... ' LL ,=-==-=-fS:. ,r-" ~ , i F ~ ~~ --.. ~:;.,'
D~ i '- i* ft; ~-X';~1· ~=g; ++T,C~ , Z' " ; ,#?? ,-=~;;;;;;;);/ [ _J r-- .~~. ~~ ==~ - ~ 's :4: :A:==~ ~ ~ " , s " ":r£ =::£' § S " S , " '< " , , " " " " " .. , , , , " , , .. , .... , " " " 'i
C
.,
" - MATCH LiNE@ / ~4'-0" OVERALL ROD DUCT LENGTH ____________________________________________________ ----: _________________ L __ 'i_·~~I~.~~~~~~4_.~~_-_-~ -l:--!I POISON TIP I!. DUCT ASSEMBLY ~
, 0 I 2 3 4 SiN I ~ SCALE ~I ~-----....i 0.043" TYPICAL- ~ , . .,;".: "''''~ '" ~347"DIA W,TH O_043'CLAD
_t ~ ~ r ,,@@@.@@~~, t,'---~.-' 4 15' ~I e~l4!:.f)@@@t2J~, ~l .
"'--STEEL STAMP OR ELEC-;ROETCH Tep I @~@ @ ©. L !f~ END CAP IN ~ CHARACTERS ("..,MBE"S I !~~~l7:>©C':\@O'\©~i'" TO BE SUPPLIED By Ci..'STGNlER') I ~~'4J~'<:.tJ~ , TI!~~v("\~O,wC),;:.n', -
TYP PIECE MARK !!'i"@<:;VO<;.,;JO avo""': =O'W'.L!. I~r r;;;lJoOoOOOCY"1
,@@~~go()~ j/ " ~~ p-:7 ".115-
~". 3B?$ OUTSIDE
_ ' INSIDE
~ l.O'oDiNG PAD 4.715"
SECT~·'OA_A ~0 TSSFLATS
~'WALL _ ~
f--
c
r'-------------------------------------- 14'- O· CONTROL/ $AfETY ROD DUCT -----------------------------------...,01
r"r1---!-- 241--0-~-~~:!:=======--4-Z.~----------------l-f~=~-LZ_'-4_r_C_ONT::,~/SAFETY ROID TIP .~: I I t.....--::- Q ---==-! .. ---'\--------"r----~~ ~CORE LOADING PADS---~"" L
THE l.li=~e r~" ~"'g§1 '0.' f'fu ~ . ';~,~' __ _ -F1J J£._-- .. ~ ~ ~L--+-----+I------<>--/ --L! ;--ITt-~~~~~~---~=r! '~f ~~ t2i~i ~-~===::::JB-"
~ THIS DRAWING COMPL,ES WITH DECEMBER 1968 CONCEPT GUIDELINES
CONCEPTUAL ::4TH ISSUE &-lo-e9
_3 RO ISSUE 3-e7·69 _ iS2NO ISSUE 2i!1-69 ... 1ST esSUE 1·31-69
CON,.ROL / SAFETY ROD TIP & DUCT ASSEMBLY ~ I ~<-"-------I.I ':::==- r-=- u. s. ATOMiC ENERGY COMMISSION
:. PACI;;I~H~~~~~~~N~~';~TORY $CAL[ ~I ___ .;.0 ___ .;..' ___ .:..Z __ .......;3ft =D_ t.lARINOS I'''!IS CONTROL/SAFETY ROD
I I I I I I I = -J POISON TIP ASSEMBl..Y ~~~~~~~~------------~-~~-----+~ ....... _~~ _______ ~. -~ - ~-~ - - I 1_ ~ PAVENPORT Him ....... __ +=+-__________ --,,'~":.":!--:..=-:::1 - 1: ... 1 aac.at~ - ~OTEO I FAST FLUX TEST FACILITY __ =- _ ... _ .......... _ .. .-.__ ~ ItEVISIO"". ~_ J~"--- 300 GEN I' -. ;90101
1-.!EXT-u--.. o~:= .. ·=·NC=. o= .. .::.::~ '='"' --:..---'-=-·=· .. =~="-=·~:...O--.. --~ .. -""Q-;.=~~c=':·=~='''"=·~'---'M''''-~ -=: ~'Kio NE ,,," '-SK -3 -145 ~o r', ~!T' "
,.
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C
"
IO"DIAUETER INSTRUt.4ENT -- ------.,.
TREE SUPPORT POST (3 Pl..ACES TYP)
5~ DIAt.4ETER PLUG
l (3 PLACES TYP)
1->---------- 37.872 ----------"I!-""---------37.872----------i
SCALE: 10 2 • ~
e 8 10 12-H ;;;;;J
SHIELDING (8 Pl..ACES TYP)
211-2.50 (TYP)
-- :'
@ @) <@) @ o o ~ ® o
CONTROL/SAFETY RODS ------
OPEN LOOP WITH PROX. INSTR.----
OPEN LOOPS ---------- 2
CLOSED LOOPS --------- • REFLECTORS ---------- 78
DRIVERS ----------- 73
PERIPHERAL CONTROL RODS ----- 15
REFLECTOR / RESTRAINT POSITIONS -- 33
SHIELDING
If .If t" Ism {. ,.~, CONCEPTUAL '~I" ISSUE 12-13-68 -=.:==- U. S. ATOMIC ENERGY COMMISSION
:.
-G.R.WAYMIRE n
ItICHLAND OPERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY OP"&RATKO .... aAnYu..- fllltI!MORr .... L INSTTTVTE
REFERENCE CORE MAP
CONCEFT ~(a)
FAST FLUX TEST FACILITY ~- 300 GEN 1'"-- 1.500.01
I
•
•
D
c
•
A
J
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o
C
A
n II 10 , • - r • 'r r I
li-8" OVERALL LENGTH ---------------------1' f 'I I EI4.00--l--! ---- 25 *:: ~ [.g 113. [t+1 .~
PRIMARY LOADING BEAM RADIAL RESTRAINT
o
Ii-s' I
14.00
1~2"m"'1 ) 84.50 B"-' I
r---a.625 "I 33.25 -,. 51.25 I I
7.125 A9 I t
I 1-2.1251 I I ~ \ l
V ~ ~ .• :>' 7 f- 7 'l I ( ~ /
~ '---.J \
.-f- ---- - . - ~--
-~ ". '/:: l - Q ~ i_ --~ - f-- /.
A~ -::~
~ > \ ~ ~ J ~ I
\.... CORE LOADING PAD \....-FOR OVERALL BEA" ARRANGE"ENT SEE
(SK-3-14600. DETAIL F-F)
.,.
-------------------------------------------------.---------------------13-S"------------_------------------------------------------------------------------~
J~--------------S4.50---------------~----------------------------------------------------------------6655 .. 5500 ----------------------------------------------------------~
~----------V.50-------------~ 51. 25 --------------~----'l r - - So -~------I
TAPERED LOADI"G BEA"
J~_
-----l------_. ---------------r----- -- - ----- .. -----, , ~I-+-~~------+- ------~~~====~~------
o 2 3 4 5 6 IN. SCALE: """'== ........... = ..... 0=-,.
0.100
SECTION A-A
SECTION B-B
II to , •
SECTION C-C
'- -- - - - -,- ---- -.----- j
ACTUATOR LOADING PAD/
(SEE DETAIL E-E)
0.050 PAD
SECTIOi\J 0-0
TOP OF WEB SUPPORTS 0 ~
St ----------~
1---, -------',
L-.. ---I\,------Aj
DETAIL E-E
!!.2L!;; THIS DRAWING CO"PLIES WITH THE
DECE>ABER '968 CONCEPT GUIDELINES.
~2'" ISSUE 2-28-69
CONCEPTUAL "',", ISSUE 12-'4-&8
-=.:~_-:::::::- U. S. ATOMIC ENERGY COMMISSION RICHLAND OPERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY OPCltAn.a aT ."TTII:U.E MIlIoIIORIAL INSTITUTE
- G. R. WAYMIRE '~' .. PRIMARY LOADING BEAM
J.-+--I-----4--+--I'-=-+------------I---t-=w;--:. u.-:--JA":":' --::C-:-::K~SO::-:N--l''"'''.-:-i. RA D I A L RES T R A I N T 1----1--+---------11 :: FU .... &NOTED FAST FLUX TEST FACILITY
-- =- ~~ ;.;-~ -- 300 GEN 1--- 2200·01
.,.....,.=-:-.... =-D-=::::"'.::·:::"""=·-=D.:::.:.:: .. "'IHG=·'---__ 11----:-:::-=~=-::.;'=-;::u.-=-::_="":-c'W-:::''''HQ~OT='C:'''='=::';:::=''-----''''-I'- ~':'~ NON E
• 4
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., '7 '. r
I ;2:- 2.-1" ..)VERAi..i.. LENGil-l
.. 1·----30 ----.r . ...-----33-l--------t.I--------- I _I • 51 "
~ ~ f 8 OUTER REFLECTOR ASSEMBLY
SCALE: o
~1·-------------Z4------------~
r---rl------------------------------+-----------------li-zf------------------------------------------------------------------------~--~ ~--~,~----------30.-------------------~~---------~--------~====-36-------------------------------------------------------+~----__________ _
SrRUCTURE (SEE ~ 48---------.
............... ----'-,,---' D .. r3T2-l LOWER SHIELD I
----Rt3iE~~~-.. (=~-~~ --~~~--~-,
,.'
B~ D.J 0.050 COLLAR GRID PLATE
(SEE SK-3-14540) INLET RECEPTACLE
r-------------------------------------------. ,-. 12-Z2----------------------------------------------------------------------------------~
r--------------------------48--------------------------------~----------~------------------------------- I 'j REFLECTOR
->
(TYPJ
0.205
SECTION A-A
SECTION B-B
•
SECTION C-C
.'1 32 2 -------------1-
1
1 E...-; I UPPER SHIELD I ~ =----- -= ~----== - » ~. ~~
::::---:-~ -----. ~-------------. t----t------ --
F==- .5 C:::--c .- - :'"-<
r 0 I 2 3 4 5 6 IN. SCALE: .. Ioo! -----
0-140
SECTION 0- 0 SECTION E-E
.v E -== r-- -'--- -
:..-----
~~ -'.
FUEL HANDLING SLOT7
(6 PLACES TYP) E~ LOADING PAD (3 PLACES TYPJ
I. THIS CRANING COMPLIES WITH THE DECEMBER
Ieee CONCEPT GUIDEUNE S.
CONCEPTUAL ~i~:::~; ~~::~:: -:.-==:' -~ u.s. ATOMIC ENERGY COMMISSION
,.ICHL.AND OPERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY ~.A"'I:) .v .... ~u.. JoIII:'-IOR1Al. lNSTIT\1'T1[
SCALE: I!!!5;i=OZO==~iiiiiiiiiiiiiiiiii2~=,;3 IN.
IJJ -G..R. WAYMIRE /2" OUTER REFLECTOR
ASSEMBLY FAST FLUX TEST FCILITY ~- 300 GEN ,._- 2200.01
A
i ... I j
,
IE
D
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'. ------------------------------ . .,'_.J" a"'E.~ALl .•. :....~NvTH I t----'8 -------- .D'4 -------------- ~, i' --------
_ __ ---=--;&~i ~~tth-=======s~-t==-==-=-=-=~======--=:~-·~ .. t=r ===============-2 1
'-.---------------------------.21 44 11-0 _____________ 111-------------1 SCALE: I" la3
~1 13~ ------~ ~-------------------------------------42-----------------------~--------------~.------
DRIVER FUEL ASSEMBLY
i: ! ~ i 4M8.7S0---1 II. ---~"' r PARTICLE TRAP- DUCT INTERNAL Bj I ! I- zi 1 A ~ 0-140 WALL) / CONI'IGURATION SI~ILAR TO SECTION B-B . LOWER REFLECTOR i
~~~=~~I--_- ~~=::='~~~-~i~~I-~~~'~~:~~~~~~~~~~~~~f~-I-I-~allIIIIIIIII~_ f > /~~;; ~ I - ~ [rt= ...
~ ~~ \~~ ~~ I =--"~~ --' .. l (:::: ~'" -",,: n ~ j ;;'rr "G'''-n.~ ""m" ,_, ~~" CO",""," -.-/ n 1 r B.J j I INLET RECEPTACLE
l~ ~-------------------------------------------------_'4_'-_O·_------------------------------------·-------------------------·-----------i
.,.
r--------------------------------------------------------------------------------r-OT-A-L-F-U~:·~:-L-E-~-G-T-~------9-3------------------------------------------- I
r-----------,.-C-T-"V :31!E
81
c ·-O-R-E-------------+I-u-p-P-ER- :~,,=1 Bj ~. "'""" !-----C--'--P-L-E-N-U-M--C-O-NT-,-N-G-E-N-C-Y-''':'''""'"' ".G D i
• F <====-= Em;±P?
~-.-. -~i-'- ---
LEAKAGE FLOW vENTS HYDRAULIC BALANCE
DETAIL E- E
SECTION A-A
..
- ,y'-t-s
SECTION B-B
,~
~k ---- ~- - -~ ~-
2 3 4 S 8 'N.
SECTION C-C
-- ::
·==----c' I mc ","",.G ~O" D ..J .,J (8 PLACES TYP)
O.OSO
SECTION 0-0
THIS DRA .... ,NG COMPLIES WITH THE
DECE~BER li68 CONCEPT GUIDEL'NES. b!ll!l ISSUE ~-14-69 E'>4!!! ISSUE 3-Z1-69 ., r- ISSUE 3- 3- 69
A.1Z-.! ISSUEI!I 2-19-69
CONCEPTUAL I!! ISSUE 2-8-69
":::===" u. S. ATOMIC ENERGY COMMISSION "ICHLAND OPERATIONS OFFICE
PACIFIC NORTHWEST LABORATORY OPWltAno .., aATTIU.. Mlp,jIOI'IIAL ',",STTnJTI:
-G.R.WAYMIRE ~.. DRIVE R FUE L
ASSEMBLY -w.u JACKSON 1~>6
NOTED FAST ' • .He -- 300 GEN
r
o
C
, I
,.
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C
...
j /
/ ,,--y_ L __ _
'-- REACTO~ VESSEL
(SEE SK-J-12899)
" DIAMETER INLE"'!" HOLES
'UEL MELT SEcnON
DETAIL 0-0
SCALE:
" r
SEE ~NLARGED DETAIL OF HYD~AULIC BALANCE RECEPTACLE
SCALE:
--;y
I I
I I
.-----e------ -a-----
9 B
B 9 o 9
-- -e------e
-~----a a
---- e
o 2 l 'T. ~========~~~
I 'i
DETAIL W
"
.,--- -", ..._------J y __
\
61-2
DETAIL D-D
24~
i CORE --.1. (SEE SK-3-14545)
HYDRAULIC B . .o..L.A.NCE RECEPTACLE
-- ::
SHADED AREAS INDICATE REACTION PADS AT
RA::lIAL RESTRAINT POSITION
0 0
0 0
0 0 0
0
0
0 0
·0
SECTION C-C
NOTE:
THfS ORAWI~G ~Ot.APLIES NITH THE DECE~SE:R Ieee CONCEPT GUIDELINES.
~~ I'>SUE J-27-89 CONCEPTUAL '~'SSUE 2-18-89
--- U. 5. ATOMIC ENERGY COMMISSICN ... ----.. "'CHLAND OPERATIONS OFFICE"
PACIFIC NORTHWEST LABORATOR, OP.:ftAnc .T .... TTCu...& WIEMO.UAl.. !1'I,Trn.JTII:
INLET PLENUM
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1ICtc.s.s. e£-.:'INS F"OR n...t:L PIN PQ E.!.5 ~.a.o ..vrTH Nar( c..A.PlLLARY ~~-\
II
SECTION A-A
" !G •
SECTION B-B
• INS'T. LEADS 'NCLUDE:
/
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CONCEPTUAL
tHIS DRAWING COMPLIES WITH THE DEC 1968 CONCEPT GUIDEUNES
~2·ISSUE 5-1-69 t! ," ISSUE 3-28-69
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~I'" ISSUE 4- 24 - 69
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.::AU< NOTE D
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GUIDE LINES
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PACIFIC NORTHWEST LABORATORY
INSTRUMENT TREE DRIVE SEQUENCE
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DETAIL E-E
12
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II 10
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SECTION C -C
_... ::
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SECTION 0- D
THIS DRAWING COMPLIES .... iT ... THE DECEMBER IgS8 CONCEPT G:....·i!)!::L!NES.
CONCEPTUAL .~ I!! iSSU E ]-14-69 -----. __ c..",-...
:. 'U.S, ATOMIC ENERGY COMMISSION
"ICHLAND OPEq"TIC~5 OFFICE
PACIFIC NORTHWEST LABORATORY OP'I:JltAT1:0 .Y .ArTl:L!.1: ""Eo,ja~!"1. IN5TlruTII!
INNER REFLECTOR
ASSEMBLY I FAST FLUX TEST ·ACIL!TY
~.,. ~-~ 300 GEN 1'--- 2200.01
--J I
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No. of Copies
31
2
1
2
BNWL-500 Volume 31
DISTRIBUTION
u.S. Atomic Energy Commission Division of Reactor Dev & Tech Washington, D.C. 20545
M. Shaw, Director, RDT Asst Dir for Nuclear Safety Analysis & Evaluation Br, RDT:NS Environmental & Sanitary Engrg Br, RDT:NS Research & Development Br, RDT:NS Asst Dir for Plant Engrg, RDT Facilities Br, RDT:PE Components Br, RDT:PE Instrumentation & Control Br, RDT:PE Liquid Metal Systems Br, RDT:PE Asst Dir for Program Analysis, RDT Asst Dir for Project Mgmt, RDT Liquid Metals Projects Br, RDT:PM FFTF Project Manager, RDT:PM (3) Asst Dir for Reactor Engrg, RDT Control Mechanisms Br, RDT:RE Core Design Br, RDT:RE (2) Fuel Engineering Br, RDT:RE Fuel Handling Br, RDT:RE Reactor Vessels Br, RDT:RE Asst Dir for Reactor Tech, RDT Coolant Chemistry Br, RDT:RT Fuel Recycle Br, RDT:RT Fuels & Materials Br, RDT:RT Reactor Physics Br, RDT:RT Special Technology Br, RDT:RT Asst Dir for Engrg Standards, RDT EBR-II Project Manager, RDT:PM
AEC Chicago Patent Group
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To An Nemzek
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BNW ReE!esent~tive
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BechteL~.ration
BNWL-500 Volume 31
T. Yo Mullen, Project Administrator, FFTF (5) M. 0 0 Rothwell (Richland)
Combustion Engineering 1000 MWe Follow-On Study
W. Po Staker, Project Manager
Dlstr-2
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Oak Ridge Natlonal Laboratory
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United Nuclear Coreoration Research and Engineerlng Center
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Westlnghouse Electric Corp Atomic Power Division Advanced Reactor Systems
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BNWL-500 Volume 31
Distr-3
BNWL-SOO Volume 31
No. of Copies
101 Battelle-Northwest
S. O. Arneson (2) J. Wo Mitchell Jo M. Batch Co N. Orsborn A, La Bement, Sro J. A. Perry Ro A. Bennett R. E. Peterson La E. Besel 0 0 Wo Priebe Co La Boyd Jo J o Regimbal Do Co Boyd Ho C, F. Rl.pfel C. L. Brown W. Eo Roake We. L, Bunch D p, Schively J o R. Carrell J, .M. Seehuus Wo Lo Chase Fe Ho Shadel T. To Claudson p. F, Shaw J. C. Cochran C. Ro F. Smith D. L. Condotta R. J. Squl.res R. Ro Cone D. D. Stepnewski J. H. Cox En G. Stevens D. Rn Doman G. H. Strong G. Eo Driver C. Do Swanson R. Va Dulin Po Kc Telford J. F" Erben J, C, Tobin E. A. Evans R. C. Walker To Wo Evans G. R. Waymire (10 ) L. Me Finch J. H. Westsik En E. Garrett J. Fa Wett E. D. Grazzl.ni To W. Withers V. Wo Gustafson No Go Wittenbrock In Wn Hagan M. Ro Wood J. P. Hale FFTF Fl.les (10) J. Eo Hanson Technical Publications Ko Mo Harmon Technical Information (5) R. Ac Harvey Legal - 703 Bldgo R. Eo Heineman Legal - ROB 22l-A P. L. Hofmann B.c Wolfe H, E. Hylbak R< J. Jackson En Mo Johnston F" J.-. Kempf D. P. Koreis J. Kolb G. A. Last H. Do Lenkersdorfer C. E. Love Do Marinos Wo Be McDonald ;J, So McMahon
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