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

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· 1-7

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· 1-9

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1-9

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• 1-10

· 1-10

· 1-11

· 1-11

· 1-11

· 1-11

· 1-12

1-12

1-12

· 1-12

· 1-13

· 1-13

• 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

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· 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

<to o z I--'~ s:: t"i ;3 I (]) U1

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BN~V'L-500

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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l 33t J 51i ~ i .ekt ~ Ei = SHiELD A SSE'V'SLY

0 2 SCALE:

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li ~= .(

t

............ j::::, .... -'-. "'_. ;:;I ..

~""~Es ..

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-I ~~

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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

•

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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

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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

c

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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:--~

~£-

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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

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" - 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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•

D

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

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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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'-.---------------------------.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

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C

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...

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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CONCEPTUAL

tHIS DRAWING COMPLIES WITH THE DEC 1968 CONCEPT GUIDEUNES

~2·ISSUE 5-1-69 t! ," ISSUE 3-28-69

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INSTRUMENT TREE DRIVE SEQUENCE

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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

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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

Rn Me Fleischman (ZPPR)

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

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Dlstr-2

No. of Copies

1

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4

2

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11

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BNWL-500 Volume 31

Distr-3

BNWL-SOO Volume 31

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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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