STATUS OF

LMFBR REHEAT IN WESTERN EUROPE--1972

Report of the United States of America

LMFBR Sodium Reheat Team Visit

to France, Germany

Netherlands, and

May 22-Jun

(FRG),

United Kingdom

e 6, 1972

Division of Reactor Development

and Technology metadc784183

States Atomic Energy Commission

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The cover shows a schematic of the heat-transfer systems for

the European sodium-cooled fast breeder reactor, SNR-300,

to be located near Kalkar, West Germany.
WASH- 1219

Liquid Metal Fast

Breeder Reactors

(UC- 79)

STATUS OF

LMFBR REHEAT IN WESTERN EUROPE- -1972

Report of the United States of America

LMFBR Sodium Reheat Team Visit

to France, Germany (FRG),

Netherlands, and United Kingdom

May 22-June 6, 1972

March 1973

Division of Reactor Development

and Technology

United States Atomic Energy Commission

For sale by the Superintendent of Documents, U.S. Government Printing Office

Washington, D.C., 20402

PREFACE

The decision by the AEC to have a representative group of LMFBR

steam-generator experts visit Europe to determine the status of LMFBR

sodium/steam cycles in Europe was based on the need for additional information needed to determine if sodium reheat could be successfully developed

for the LMFBR Demonstration Plant Nuclear Steam Supply System at the

Tennessee Valley Authority John Sevier Power Plant. Information obtained

by the members of the team was used by the management of the LMFBR

demonstration plant to help arrive at the final decision to locate the demonstration plant at a new site on the Clinch River near Oak Ridge, Tennessee.

The discussions with European counterparts and visits to their establishments left the team members with the impression that the Europeans

consider that a prerequisite to successful LMFBR demonstration and to

commercial reality is a comprehensive, coordinated, and disciplined development program with considerable emphasis on testing to the degree necessary for ensuring successful operation of plant components and systems,

involving full-scale testing in some cases, large-model testing in others,

and component- part testing in many cases . Most impressive were the largescale sodium-water reaction tests being conducted in France, Germany,

and the U.K. and the large steam-generator model testing in France and

Netherlands. Such programs provide encouragement to this country to vigorously pursue its program to provide safe, reliable and economic commercial LMFBRs, based on extensive development and testing of component

parts, components, and systems using in-depth engineering and strong

quality- as surance practices.

The exchange of information with our friends overseas and their

gracious hospitality were most gratifying to all the members of the team

and not soon to be forgotten. The work of Max Weber, of the Argonne

National Laboratory, in compiling and editing the manuscript has been most

appreciated.

John G. Yevick John A. Ford

Robert V. Laney John S. McDonald

Raymond E. Hoskins Robert S. McQueen

William F. Rolf Louis A. Sturiale

Edgar Rodwell

i

TABLE OF CONTENTS

NOMENCLATURE. ....................

Page

.. . xiv

. . . . . . . . . . . I. SUMMARY. ... . . ............... .

A.

B.

C.

D.

E.

F.

Status of Reheat in Western Europe.................

Steam Generator and Associated Systems Design . . . . . .

1. Phenix . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2. French CFR Steam Generators . . . . . . . . . . . . . . . .

3. PFR....... ...................................

4. SN R -300 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

Steam-generator Development . . . . . . . . . . . . . . . . . . .

Turbine Development for Steam Reheat. . . . . . . . . . . . .

French Approach to Developmental Testing . . . . . . . . . .

Sodium-Water Reaction Programs and Test Sites . . . . . .

G. Demonstration Plants . . . . . . . . . . . . . . . . . . . . . . . . .

H. Commercial Plants . . . . . . . . . . . . . . . . . . . . . . . . . .

II. INTRODUCTION. .............. . . . . . . . . . . . . . . . .

III. EUROPEAN STATUS AND POSITIONS ON SODIUM REHEAT. 

1

1

2

2

2

3

3

4

6

6

7

7

8

9

13

A. Introduction......... . . . . . .......

B. Status of Sodium Reheat . . . . . . . . .

1. Status Summary . . . . . . . . . . . .

2. Status Comparison . . . . . . . . . .

a. French...................

b. British . . . . . . . . . . . . . . . .

c. SNR Consortium. . . . . . . . . .

C. Reasons Expressed for the Plant-cycle

1. Introduction and Summary . . . . .

2. SNR Consortium. . . . . . . . . . . .

3. French.......................

4. United Kingdom . . . .. ..... ....

. . . . . . . . . . . . . . 13 . . . . . . . . . . . . . . 13

. . . . . . . . . . . . . . 13

. . . . . . . . . . . . . . 14

. . . . . . . . . . . . . . 14

. . . . . . . . . . . . . . 15

. . . . . . . . . . . . . . 16

Selection. . . . . . . . 16

. . . . . . . . . . . . . . 16

. . . . . . . . . . . . . . 17

... . ....... 17

0.0. .........18 . . .

IV. STEAM GENERATORS AND ASSOCIATED SYSTEMS DESIGN. 

A. Summary. . . . . ................... . . .. . . ...........

111

19

19
TABLE OF CONTENTS

Page

B. France.. . . . . .. ..................................... 20

1. Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

2. Phenix Steam Generators. . . . . . . . . . . . . . . . . . . . . 21

3. Steam Generators under Development for 1000-MWe

Plant. ... . . . . . .. .. . . . . . . . . . . . . . . . . . . . . . . 28

a. Babcock-Atlantique. . . . . . . . . . . . . . . . . . . . . . . 28

b. Stein Industrie........ .......................... 28

c. Comparison of Phenix-type, B-A, and Stein Industrie

Commercial Units . . . . . . . . . . . . . . . . . . . . . . . 31

d. Les Renardieres Steam Generator Test Facility . . . 31

e. Sodium-Water Reaction Test Facility at Cadarache . 32

C. United Kingdom: PFR Steam Generators. . . . . . . . . . . . . 32

1. Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32

2. PFR Startup Procedures . . . . . . . . . . . . . . . . . . . . . 39

D. SNR Steam Generators . . . . . . . . . . . . . . . . . . . . . . . . . 41

1. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41

2. Steam-generator Design. . . . . . . . . . . . . . . . . . . . . . 41

3. Steam-generator Tube Leak and Rupture-disk

Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . 48

4. Sodium-reheat Component and Systems Considerations . 50

5. M iscellaneous . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50

V. DEVELOPMENT PROGRAMS FOR STEAM-GENERATION

SYSTEMS AND RELATED COMPONENTS. . . . . . . . . . . . . . . 51

A. Summary and Comparison. . . . . . . . . . . . . . . . . . . . . . . 51

B. France.. . . . . .. ..................................... 51

1. Phenix Steam-generator System Development . . . . . . . 51

2. Steam Generator under Development for 1000-MW

Plant.... ................... ....... .. ... 53

a. Babcock-Atlantique. . . . . . . . . . . . . . . . . . . . . . . 54

b. Stein Industrie........ .......................... 54

3. Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54

C. United Kingdom. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56

D. Germany--INTERATOM . . . . . . . . . . . . . . . . . . . . . . . 58

1. Development of SNR Steam Generator. . . . . . . . . . . . . 58

2. Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58

iv
TABLE OF CONTENTS

Page

E. Netherlands......... . . . . . . . . . .... 0 ...... . . 59

1. Tests on SNR Steam-generator Thermal Model at

Hengelo. .... ... .. ................ . ..

2. Steam-generator Development Test at Apeldoorn

(TNO)..................................

3. Component and Weld Surface Cleaning... .... ...

VI. SODIUM-WATER REACTION PROGRAMS AND TEST SITES

A. Summary and Comparison... ............... ....

B. France....................................

59

62

63

65

65

67

Large-leak Test Facilities . . . . . . .

Large-leak Test Results. . . . . . . . .

Large-leak Mathematical Models .

Large-leak Future Plans . . . . . . . .

Large-leak Test Instrumentation . . .

Large-leak Test Facility Operation .

Relief Systems . . . . . . . . . . . . . . .

Design and Analysis for Sodium-Water

Small-leak Test Facilities . . . . . . .

Results of Small-leak Tests . . . . . .

Leak Detection.... ....... . . . . . . .

Reactions 

C. United Kingdom... . ...... 0. . .. . ..... .0.... 0...

Large-leak Test Facilities . . . . . . . . . . . . . . . . . .

Large-leak Test Results. . . . . . . . . . . . . . . . . . . .

a. Test 1...... ...............................

b. Test 2....... ...............................

c. Test 3........ ..............................

Large-leak Mathematical Models . . . . . . . . . . . . . .

Large-leak Future Plans . . . . . . . . . . . . . . . . . . .

Operation of Large-leak Test Facility. . . . . . . . . . .

Large-leak Test Instrumentation . . . . . . . . . . . . . .

Simulant Tests . . . . . . . . . . . . . . . . . . . . . . . . . .

Relief Systems . . . . . . . . . . . . . . . . . . . . . . . . . .

Design and Analysis for Sodium-Water Reactions . . .

Small-leak Test Facilities . . . . . . . . . . . . . . . . . .

Small-leak Test Results . . . . . . . . . . . . . . . . . . . .

Wastage Models . . . . . . . . . . . . . . . . . . . . . . . . .

Leak Detection . . . . . . . . . . . . . . . . . . . . . . . . . .

Damage Location and Assessment ...................

V

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

. . . 67

. . . 67

. . . 69

69

69

. . . 69

. . . 69

70

71

71

. . 72

. . . 74

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

74

74

76

76

77

77

77

78

78

79

79

79

79

80

80

80

81
TABLE OF CONTENTS

Page

D. SNR Consortium . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81

1. Large-leak Test Facilities . . . . . . . . . . . . . . . . . . 81

2. Large-leak Test Results. . . . . . . . . . . . . . . . . . . . 83

3. Large-leak Mathematical Models . . . . . . . . . . . . . . 86

4. Large-leak Future Plans . . . . . . . . . . . . . . . . . . . 87

5. Operation of Large-leak Test Facility. . . . . . . . . . . 87

6. Large-leak Test Instrumentation . . . . . . . . . . . . . . 87

7. Simulant Tests . . . . . . . . . . . . . . . . . . . . . . . . . . 88

8. Relief Systems 88

9. Design and Analysis for Sodium-Water Reactions . . 89

10. Small-leak Test Results . . . . . . . . . . . . . . . . . . . . 90

11. Leak Detection . . . . . . . . . . . . . . . . . . . . . . . . . . 91

VII. LMFBR DEMONSTRATION PLANTS . . . . . . . . . . . . . . . . . 92

A. Summary and Comparison. . . . . . . . . . . . . . . . . . . . . . 92

B. France--Phenix . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96

1. Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96

2. Status of Construction . . . . . . . . . . . . . . . . . . . . . 99

C. U.K.--PFR. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100

1. Plant Description . . . . . . . . . . . . . . . . . . . . . . . 100

2. Status of Construction . . . . . . . . . . . . . . . . . . . . . 103

3. Steam-generator System. . . . . . . . . . . . . . . . . . . . 104

D. SNR.. .......... .................................... 105

VIII. COMMERCIAL PLANTS . . . . . . . . . . . . . . . . . . . . . . . . . 110

A. Summary and Comparison. . . . . . . . . . . . . . . . . . . . . . 110

B. France.. . . . .. . . ................................... 111

C. United Kingdom. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117

D. Germany . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 118

IX. STEAM-GENERATOR TEST SITES . . . . . . . . . . . . . . . . . . 121

A. Summary and Comparison. . . . . . . . . . . . . . . . . . . . . . 121

B. Les Renardieres Steam-generator Test Facility, France . 121

1. Facility Description. . . . . . . . . . . . . . . . . . . . . . . 121

2. Status.. . . . . .. . ................................ 122

3. Future Facility Usage . . . . . . . . . . . . . . . . . . . . . 125

vi
TABLE OF CONTENTS

Page

C. U.K. Steam-generator Testing . . . . . . . . . . . . . . . . . . . . 125

D. Hengelo Sodium Component Test Facility, Netherlands . . . . 126

1. Facility Description . . . . . . . . . . . . . . . . . . . . . . . . 126

2. Status.......... .................................. 133

3. Sodium Leak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 133

4. Additional Problems . . . . . . . . . . . . . . . . . . . . . . . . 134

5. Steam-generator Testing . . . . . . . . . . . . . . . . . . . . . 135

6. Future Facility Usage . . . . . . . . . . . . . . . . . . . . . . . 135

7. Leak-detection Experience. . . . . . . . . . . . . . . . . . . . 136

E. 5-MWt Test Facility, Bensberg, Germany. . . . . . . . . . . . . 136

X. OTHER TEST SITES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138

A. TNO--Central Technical Institute Facilities at Apeldoorn,

Netherlands. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 138

1. Introduction........ .............................. 138

2. Corrosion Loops... . ... . .......................... 139

3. Chemistry Loop . . . . . . . . . . . . . . . . . . . . . . . . . . . 140

4. Thermal-shock Loop. . . . . . . . . . . . . . . . . . . . . . . . 141

5. Cold-trap Design . . . . . . . . . . . . . . . . . . . . . . . . . . 142

6. Pump Loop . . .. . . . . . . . . . . . . . . . . . . . . . . . . 143

7. Sodium Cleaning. . . . . . . . . . . . . . . . . . . . . . . . . . . 145

8. Miscellaneous . . . . . . . . . . . . . . . . . . . . . . . . . . . . 145

9. Laboratory Visits. . . . . . . . . . . . . . . . . . . . . . . . . . 145

10. IHX Tube Plugging and Welding. . . . . . . . . . . . . . . . . 145

11. Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 146

B. EDF Les Renardieres Test Facility. . . . . . . . . . . . . . . . . 146

C. CEA--Cadarache Nuclear Research Center. . . . . . . . . . . . 147

1. Introduction........ .............................. 147

2. Component Test Facilities . . . . . . . . . . . . . . . . . . . . 148

a. HR-1 Building . . . . . . . . . . . . . . . . . . . . . . . . . . 148

b. HR-4 Building . . . . . . . . . . . . . . . . . . . . . . . . . . 150

3. Corrosion Test Facilities--SEML . . . . . . . . . . . . . . . 151

4. Rapsodie. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 152

D. INTERATOM, Bensberg, Germany. . . . . . . . . . . . . . . . . . 153

1. Nitriding of Materials . . . . . . . . . . . . . . . . . . . . . . . 153

2. SNR Pump-test Facility. . . . . . . . . . . . . . . . . . . . . . 153

3. SNR Rotating-plug Test . . . . . . . . . . . . . . . . . . . . . . 155

vii
TABLE OF CONTENTS

Page

APPENDIXES

A. Itinerary of U.S. LMFBR Sodium Reheat Team . . . . . . . .

B. Agendas........ ...................................

1. CEA/EDF: Saclay, May 23. . . . . . . . . . . . . . . . . . .

2. United Kingdom Atomic Energy Authority: Dounreay

Experimental Reactor Establishment . . . . . . . . . . . .

3. Meeting USAEC- Team, INTERATOM, Bensberg,

Germany; Sodium Reheat Cycle Design; May 29-30,

1972...........................................

4. TNO/NERATOOM- -Hengelo, Netherlands, June 1. . . .

a. Mechanical . . . . . . . . . . . . . . . . . . . . . . . . . . .

b. Operational . . . . . . . . . . . . . . . . . . . . . . . . . . .

c. Electrical . . . . . . . . . . . . . . . . . . . . . . . . . . . .

d. Water Chemistry . . . . . . . . . . . . . . . . . . . . . . .

e. Measurements . . . . . . . . . . . . . . . . . . . . . . . . .

f. Steam Generators . . . . . . . . . . . . . . . . . . . . . . .

g. Detection of Water Leaks in Steam Generators... ..

5. CEA, Cadarache, France, June 6..

C. Participants............................. 0...........

1.

2.

3.

4.

5.

6.

7.

8.

9.

Saclay: May 23 .... . . . . . . . . . . ... . . . . . .

Clamart, France: May 24 . . . . . . . . . . . . . . .

Dounreay, Scotland: May 25 and 26 . . . . . . . . .

Bensberg, Germany: May 29 and 30. . . . . . . . .

Apeldoorn: May 31 . . . . . . . . . . . . . . . . . . . .

Hengelo, TNO/NERATOOM, June 1 . . . . . . . . .

The Hague, Netherlands: June 2...............

Marcoule, France (Phenix): June 5 . . . . . . . . .

Cadarache, France: June 6...................

D. Address List of Places Visited . . . . . . . . . . . . . . . . . .

REFERENCES..... ........... . . . . . .. . . . . . . . . . . . . .

156

157

157

157

158

159

159

159

159

160

160

160

160

. . . . . . . . . . 160

162

. . 162

. . 162

. . 163

163

164

... . 164

164

. . 164

.0 ... 165

166

168

viii
LIST OF FIGURES

No. Title Page

1. Flow Diagram for LMFBR Nonreheat System . . . . . . . . . . . . 9

2. Flow Diagram for LMFBR Sodium-reheat System . . . . . . . . . 9

3. Flow Diagram for LMFBR Steam-reheat System. . . . . . . . . . 10

4. Arrangements for Sodium-reheat Steam Generator . . . . . . . . 10

5. Phenix Steam Generator . . . . . . . . . . . . . . . . . . . . . . . . . . 21

6. Phenix Steam Generator- - Water Inlet Header . . . . . . . . . . . . 23

7. Temperature Profiles for Phenix Steam Generator; Normal

Power Level (563 MW); Cold Condition . . . . . . . . . . . . . . . . 25

8. Temperature Profiles for Phenix Steam Generator; 0.25 of

Normal Flow Pressure: 168 bars; Cold Condition . . . . . . . . . 26

9. Temperature Profiles for Phenix Steam Generator; 0.25 of

Normal Flow Pressure: 42 bars; Cold Condition. . . . . . . . . . 27

10. Babcock-Atlantique Steam Generator. . . . . . . . . . . . . . . . . . 29

11. Stein Industrie Steam Generator . . . . . . . . . . . . . . . . . . . . . 30

12. Schematic of Heat-removal System for PFR Steam Generator 33

13. Temperature/Heat Diagram for PFR Steam Generator . . . . . . 33

14. PFR Steam-generator Heat Exchangers . . . . . . . . . . . . . . . . 35

15. Internal Thermal Sleeve for PFR Superheater and Reheater . . 36

16. U.K. PFR Tube-to-Tube-sheet Butt/Fillet Joint . . . . . . . . . . 38

17. Prototype of SNR Steam Generator . . . . . . . . . . . . . . . . . . . 43

18. Plots of Sodium Flow for SNR Steam Generator. . . . . . . . . . . 45

19. NERATOOM Tube-to- Tube-sheet Connections Developed for

SNR Steam Generator . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46

20. Phenix Secondary-system Isolation Valve. . . . . . . . . . . . . . . 55

21. Assembly Arrangement of Phenix Secondary-system Isolation

V alve. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55

22. SWISH Loop Thermocouple Connections to Steam-generator

Tube.. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63

23. Sodium-Water Reaction Test Loop for Phenix Modules . . . . . . 68

24. BSB Type DV Rupture Disks Used at Les Renardieres Steamgenerator Test Site . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70

ix
LIST OF FIGURES

No. Title Page

25. Schematic of Small-leak Sodium-Water Reaction Facility at

Cadarache.. . . . . .. . . ................................... 71

26. French Small-leak Test Results for 24 Cr- 1 Mo-1 Nb . . . . . . 72

27. Location of Leak-detection System in Phenix Secondary

Circuit................................................. 73

28. Leak-detection System for Phenix Steam Generator . . . . . . . . 73

29. Super-NOAH Rig, Simplified Flow Sheet . . . . . . . . . . . . . . . . 75

30. Plot of Pressure vs Time for PFR Steam-generator SuperNOAH Sodium-Water Reaction Test 2 . . . . . . . . . . . . . . . . . 76

31. INTERATOM Large-leak Test Facility . . . . . . . . . . . . . . . . 82

32. Relief System for INTERATOM Large-leak Test Facility . . . . 83

33. Plot of Pressure vs Time for INTERATOM SNR SodiumWater Reaction Tests 4 and 5. . . . . . . . . . . . . . . . . . . . . . . 83

34. Plot of Pressure vs Time for INTERATOM SNR SodiumWater Reaction Test 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . 84

35. Location of Thermocouples in Reaction Zone for INTERATOM

SNR Sodium-Water Reaction Test 3. . . . . . . . . . . . . . . . . . . 85

36. Thermocouple Temperature Readings for Locations Shown in

Fig. 35 for INTERATOM SNR Sodium-Water Reaction Tests . . 85

37. Plot of Pressure vs Time- -Prediction vs Experiment--for

INTERATOM SNR Sodium-Water Reaction Test 3 . . . . . . . . . 86

38. Plot of Temperature vs Time--Assumed, Calculated, and

Measured--for INTERATOM SNR Sodium-Water Reaction

T ests. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86

39. Plot of Pressure vs Time-- Theoretical vs Experimental--for

INTERATOM SNR Sodium-Water Reaction Test 4 . . . . . . . . . 87

40. German Rupture Disk . . . . . . . . . . . . . . . . . . . ... . . . . . . . 88

41. Flow Diagram for Phenix System . . . . . . . . . . . . . . . . . . . . 93

42. Flow Diagram for PFR System. . . . . . . . . . . . . . . . . . . . . . 94

43. Flow Diagram for SNR System. . . . . . . . . . . . . . . . . . . . . . 95

44. Portion of Phenix Secondary Sodium System, Showing Some

Features to Minimize Effects of Sodium-Water Reaction. . . . . 98

45. Organization of the SNR Project . . . . . . . . . . . . . . . . . . . . . 105

x
LIST OF FIGURES

No. Title Page

46. Sodium-reheat and Steam-reheat Parameters for 780-MWe,

3000-rpm Turbine . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 112

47. EDF 1000-MWe LMFBR Concept: Heat-balance Diagram for

Sodium Reheat. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 114

48. EDF 1000-MWe LMFBR Concept: Heat-balance Diagram for

Steam Reheat . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 115

49. Cost Savings vs AT between Primary and Tertiary. . . . . . . . . 119

50. EDF Les Renardieres Steam-generator Test Facility--Flow

Diagram . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123

51. EDF Les Renardieres Test Site.. ................. . .. 124

52. First Steam-generator Module Tested and Removed from the

Les Renardieres Facility . . . . . . . . . . . . . . . . . . . . . . . . . 125

53. Model of Hengelo Steam-generator Test Facility . . . . . . . . . . 126

54. Furnaces, Air Preheaters, and Central Chimney of Hengelo

Steam-generator Test Facility . . . . . . . . . . . . . . . . . . . . . . 127

55. Mixing Tee Used at Hengelo Steam-generator Test Facility. . . 128

56. Sodium System Schematic Diagram, 50-MW Sodium Component

Test Facility, Hengelo . . . .,. . . . . . . . . . . . . . . . . . . . . . . 129

57. 50-MWt Sodium Component Test Facility, Hengelo. . . . . . . . . 130

58. Cutaway View of Steam Hall at 50-MWt Sodium Component

Test Facility, Hengelo . . . . . . . . . . . . . . . . . . . . . . . . . . . 131

59. Plan View of 50-MWt Sodium Component Test Facility,

Hengelo........... ..................................... 132

60. Section of Loop Showing Specimens for High-velocity Sodium

Tests at Pump-bearing Clearances, TNO-CTI, Apeldoorn. . . . 140

61. TNO-Apeldoorn Thermal-shock Loop . . . . . . . . . . . . . . . . . 141

62. Vessel-nozzle Thermal-shock Model, TNO-CTI, Apeldoorn. . . 142

63. Schematic of 5000-m3/hr Pump-bearing Thermal-shock-test

Container, TNO-CTI, Apeldoorn . . . . . . . . . . . . . . . . . . . . . 142

64. Visco-seals in Tandem . Arrangement. . . . . . . . 143 . . . . . . . . .

65. Visco-seal Principle.. . . . . .............................. 143

66. TNO-CTI 280-m3/hr Pump Loop at Apeldoorn. . . . . . . . . . . . 144

67. Cadarache Nuclear Research Center . . . . . . . . . . . . . . . . . . 149

68. Prototype Pump and Drive. . . . . . . . . . . . . . . . . . . . . . . . . 154

xi
LIST OF TABLES

No. Title Page

1. Parameters for LMFBR Demonstration Plants: U.K., France,

and Germ any . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

2. Comparison of West European Plant-cycle Status . . . . . . . . . 13

3. French Commercial LMFBR Steam-cycle Conditions . . . . . . . 15

4. Nominal Phenix Steam-generator Characteristics . . . . . . . . . 23

5. Comparison of French Commercial 1000-MWe LMFBR Steamgenerator Concepts.. . . . . .. . ............................. 31

6. Main Parameters for Heat-transport System . . . . . . . . . . . . 34

7. U.K. PFR Steam Generator: Material Composition . . . . . . . . 38

8. U.K. PFR Steam Generator: Tube and Tube-sheet

Dimensions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38

9. SNR Straight-tube Steam-generator Data . . . . . . . . . . . . . . . 42

10. Summary of Demonstration-plant Design Conditions. . . . . . . . 92

11. Summary of PFR Parameters . . . . . . . . . . . . . . . . . . . . . . 101

12. Stein Industrie 1000-MWe Concept; Phenix-type Steamgenerator Data; Sodium Reheat vs Steam Reheat . . . . . . . . . . 112

13. Stein Industrie 1000-MWe Concept Using Phenix-type Steam

Generators: Weight Comparison of Sodium Reheat vs Steam

Reheat............. ...................................... 112

14. Differential Costs of Stein Industrie 1000-MWe LMFBR

Concept: Sodium-reheat Plant minus Steam-reheat Plant. . . . 113

15. EDF 1000-MWe LMFBR Concept of Sodium Reheat vs Steam

Reheat: Steam-generator Data. . . . . . . . . . . . . . . . . . . . . . 113

16. EDF 1000-MWe LMFBR Concept; Differential Costs; Sodiumreheat Plant minus Steam-reheat Plant . . . . . . . . . . . . . . . . 114

17. System Parameters for EDF 1000-MWe LMFBR Optimization

Studies........... ...................................... 116

18. Fixed Design Values for EDF Optimization Studies . . . . . . . . 117

19. Capital Costs Associated with EDF 1000-MWe LMFBR

Optimization Studies. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117

20. Cost Change with SNR-1000 vs Reference Design I. . . . . . . . . 119

21. Legend for Figs. 57 and 58 . . . . . . . . . . . . . . . . . . . . . . . . 132

22. Design Characteristics of 5000-m3/hr SNR Prototype Primary

Sodium Pump on Test at Bensberg . . . . . . . . . . . . . . . . . . . 155

xii


NOMENCLATURE

xiii
APDA

B-A

BNDC

BSB

CEA

CEGB

CFR

CGVS

CTI

DNB

EDF

GAAA

INTERATOM

LMFBR

LWR

NERATOOM

PFR

SEML

SNR- 300

TNO

TNPG

UKAEA

NOMENCLATURE

Atomic Power Development Associates, U.S. Designer of Fermi Fast Breeder

Babcock- Atlantique, French manufacturer of steam generators

British Nuclear Development Company, one of two U.K. consortia for design and

construction of nuclear power plants

Black, Sivalls, and Bryson, U.S. manufacturer of sodium steam-generator rupture

disks

Commissariat a l'Energie Atomique (French AEC)

Central Electricity Generating Board (United Kingdom National Electric Utility)

Commercial Fast Reactor, U.K. designation for first large-scale (about

1300-MWe) LMFBR

Circuit d'essais de Generateur de Vapeur chauffes au Sodium

TNO Central Technical Institute

Departure from nucleate boiling

Electricity de France (National French Electric Utility)

Groupement Atomique Alsacienne Atlantique (private nuclear design and construction firm)

German industrial consortium for design and construction of reactors

Liquid-metal-cooled Fast Breeder Reactor

Light-water reactor

Netherlands industrial consortium for nuclear power design and fabrication

Plutonium Fast Reactor, U.K. LMFBR Demonstration Plant located at Dounreay,

Scotland

Section Etude Metalle Liquide, division for liquid-metal studies at Cadarache,

France

Schnelle Natriumgekihlte Brutreaktor, 300-MWe LMFBR Demonstration Plant to

be located in Germany

Nederlandsche Organizatie voor Toegepast- -Natuurwetens choppelijk Onder zoekNetherlands Organization for Applied Scientific Research

The Nuclear Power Group, one of two U.K. industrial consortia for designing and

constructing nuclear power plants

United Kingdom Atomic Energy Authority

xiv
1

I. SUMMARY

The U.S. Sodium Reheat Team trip was made to those Western

European countries designing and constructing LMFBR demonstration plants

(and presently designing LMFBR commercial units) for the purpose of determining the directions being taken by these countries to provide acceptable

quality steam to commercially available steam turbines. Based on this information, evaluations can be made by the U.S. nuclear industry to better

guide the U.S. in its program to develop a safe, reliable, and economical

commercial LMFBR. Visits were made to France, Germany, Netherlands,

and the United Kingdom. Though the agendas emphasized steam-generator

development, other areas such as sodium isolation valves, turbines, and

demonstration-plant progress were covered, as well as general test

capability.

A. Status of Reheat in Western Europe

The European nations visited (France, Germany, Netherlands, and

United Kingdom) initially chose sodium reheat for their LMFBR demonstration plants as the means for re superheating steam exhausted from the highpressure turbine (Phenix, SNR and PFR), the decisions being based mainly

on availability of standard turbine design and utility preference. However,

reevaluations have been made or are being made by all the countries. A

reevaluation by the SNR Consortium (Belgium, Germany, Luxembourg, and

Netherlands) convinced them that it would be more economic and more

reliable to use steam reheat in the SNR- 300 LMFBR demonstration plant.

Also, though their reference for the commercial SNR- 1000 is steam reheat,

a final evaluation will be made before final selection. The French

(CEA/EDF) are using sodium reheat in the 250-MWe Phenix LMFBR demonstration plant. The French EDF, in conjunction with the CEA and the

French nuclear industry, has made a decision to use steam reheat rather

than sodium reheat in their commercial 1200-MWe LMFBR. This decision

was based on a slight economic edge for steam reheat and improvement in

plant reliability and availability. France, Germany, and Netherlands considered nonreheat as too costly.

The U.K. CEGB/AEA/TNPG groups are using sodium reheat in the

250-MWe PFR-LMFBR demonstration plant. For commercial LMFBRs,

an evaluation is in process to determine whether to undertake the development of large-scale turbines (1000 MWe or larger), providing an opportunity at the same time to reassess the need for sodium reheat as contrasted

to steam reheat or nonreheat. At the time of the visit, the reference CFR

design used two standard 600-MWe turbines operating in parallel and utilizing sodium reheat.

All the groups visited indicated that, though sodium reheat represented added complexity and required some system and component developing, varying in degree, none expected that sodium reheat presents
2

insurmountable problems or requires major technology development beyond

that required for sodium- cooled evaporators and superheaters.

B. Steam Generator and Associated Systems Design

1. Phenix

The 250-MWe (563-MWt) Phenix plant has 15.6-MWt modular

steam-generator units. Each of the three secondary loops contain 12 oncethrough steam-generator units in a large rectangular casing. Each unit

consists of a counterflow evaporator, a counterflow superheater, and a

counterflow reheater. The evaporator units are made of 2 Cr- 1 Mo lowalloy steel, stabilized and unstabilized. Each unit shell contains seven

tubes. The casing (an engineered safety feature) serves as containment,

can be used for emergency cooling, and can be flooded with inert gas if

necessary. The design uses no tube sheet, has an external orificed water

header, and has no internal gas space.

Each group of 12 steam-generator units can be isolated on the

sodium side by quick-acting (6-sec) isolation valves, in the event of a

sodium-water reaction. The steam-generator package is designed for

rapid dumping of both water/steam and sodium. A manually controlled

valve, adjusted once for full load, distributes the sodium flow between the

superheater and reheater, which are in parallel on the sodium side. A

mixing tee joining the sodium outlets of the superheater and reheater and

the sodium inlet to the evaporator is designed to mitigate the temperature

differentials. The superheater and the reheater can be isolated on the

steam side, allowing for water and steam circulation through the remainder

of the system at startup.

2. French CFR Steam Generators

The four-loop, 1200-MWe LMFBR will contain only one (at most

two) steam-generator units per loop. Two designs are being developed, the

Babcock-Atlantique (B-A) helical-coil unit and the Stein Industrie straighttube unit with a Z-type tube bend below the upper head for expansion. Results of sodium-water reaction tests from Les Renardieres steam-generator

tests and early Grand Quevilly tests on a small B-A model indicated that the

reliability of the large units in the CFR should be good. Further, the economics is dominantly in favor of the large units.

The B-A helical unit (with evaporator and superheater in one

shell), either 290 or 580 MWt in size, has tubes of Incoloy 800, a Type 304

stainless steel shell, water and steam connections of 24 Cr- 1 Mo, and a gas

space in each unit. There are four tube sheets at the top steam exit, with

tubes penetrating the shell at the bottom.
3

The 580-MWt Stein- Industrie unit with separator evaporator

and superheater has tubes of unstabilized 24 Cr- 1 Mo. Shell and tube sheets

are Type 316 stainless steel. It is interesting to note that the weight, in

tonnes per MWe, is reduced from Phenix-type units of 1.05 to B-A unit of

0.56, a reduction of about 50%. The volume occupied is even more striking--

15.8 m3/MWe for Phenix type and 1.73 m3/MWe for the B-A unit.

For each concept, 45-MWt models are being built, to be tested in

Les Renardieres and then sodium-water-reaction tested in Cadarache.

3. PFR

One forced- recirculation steam-generator unit, consisting of a

U-tube evaporator, a U-tube superheater in parallel on the sodium side with

a U-tube reheater, a steam drum, and two recirculation pumps, is located

in each of the three PFR secondary sodium loops. Forced recirculation is

used to provide design and operational conservatism. Also, in the U.K.,

there is a large body of experience with forced recirculation. The superheaters and reheaters are counterflow; the evaporator uses parallel flow.

Each tube of the evaporator is orificed on the water side. A

controlled gas space over sodium is provided in each unit. The superheater

and reheater are' geometrically similar, except for dimensions. Interesting

is the heat-transfer surface requirement for the reheater--30% of total

while only transferring 14% of the plant heat. The steam volumetric flow

is four times that of the superheater. The evaporator tubes are 2 Cr- 1 Mo

stabilized with niobium. The evaporator tube sheets are of unstabilized

24 Cr- 1 Mo. The superheater and reheater tubes are Type 316 stainless

steel. Backface welds are used for the tube-to-tube- sheet joints. Six

6-in. lines from the reheater and six 8-in. lines from the superheater to

the evaporator inlet mixing chamber were used to eliminate the need for a

header system and to provide good flow distribution to the chamber. Fixed

orifices are installed in the lines to provide for proper flow division.

Two 6-in. sodium-bypass lines run from the discharge side of

each secondary loop pump (consisting of two 14-in. lines) to a mixing tee

at the sodium inlet of the reheater. Valves are located in each bypass line.

The bypass line is used during startup to provide for equalization of

temperatures.

4. SNR- 300

Each secondary sodium loop will have three once-through steamgenerator units (each rated at 82 MWt) consisting of an evaporator and a

superheater. For two of the loops, the evaporators and superheaters are

similar-- straight tubes from tube sheet to tube sheet, with bellows in the

shell to provide for tube expansion and no gas space in the units. Backface welds are used for the tube-to-tube- sheet joints. The evaporator tubes
4

are orificed, and each unit has a protective shroud around the tube bundle.

The third loop will have helical-coil evaporators and superheaters. It appears that the steam-generator material used in all cases is niobium (1%)--

stabilized 24 Cr-1 Mo with 0.5% nickel. In all cases, steam separators are

installed between the evaporator and the superheater, which allows for 95%

quality steam exit into the superheater. SNR- 300 will not have sodium reheaters. It will use steam reheaters, reheating steam from the IP turbine

section to the LP turbine sections. The steam generators are being designed

and built by NERATOOM, a Netherlands nuclear-industry consortium. A

50-MWt model of the straight-tube steam generator incorporating a sodium

reheater (initially planned for SNR-300) is being tested by the Netherlands

Research Institute, TNO, at Hengelo.

The helical- coil unit (again with no gas space) will have no tube

sheets; tubes will penetrate the lower and upper heads through thermal

sleeves and will be externally headered. Though the Germans/Dutch favor

modular units, the size of the helical- coil units for the commercial LMFBR

may range as high as 250 MWt.

The German sodium-water reaction tests indicate that the presence of a gas space in the steam-generator unit allows secondary pressure

pulses to resonate to high peaks. The Germans depend on rapid rupturedisk failure and flow-through of reaction products to keep the pressure peaks

under control. The absence of a gas space also eliminates level control.

The Germans chose the once-through steam generator as less costly, eliminating recirculating pumps, steam drums, pipes, valves, and controls. Sufficient experience has been obtained on waterside dryout and corrosion to

provide confidence in the once-through units. Further, most of their fossilfired units use once-through boiler units. The Dutch/German concept contrasts with the British, who have a gas space in the PFR units, and the

French, who plan gas spaces in their 1200-MWe LMFBR. In the opinion of

the British and French, the gas space mitigates the initial pressure wave.

C. Steam-generator Development

For the demonstration plants, both the French and the SNR

Consortium decided that it was necessary to test steam-generator components on a large enough scale to provide a meaningful proof test mechanically, thermally, and hydraulically of each design. Tests are being

conducted by the French in the Les Renardiere s facility and by the SNR

Consortium in the Hengelo facility. The British have been negotiating for

CFR steam-generator model testing at the Hengelo facility. Alternatively,

they have explored the construction of their own steam-generator test

facility. No decision was reached at the time of the visit. The French and

the SNR Consortium are both proceeding to qualify two steam-generator

manufacturers and their designs through the testing of 50-MWt steamgenerator models.
5

None of the countries considered the sodium reheater, as a component, a major developmental item. The U.K. strongly emphasized reheater

system complexity, which requires careful matching of the reheater to the

turbine conditions for startup and shutdown. This matching has resulted in

added system- control complexity. For example, the U.K. has added a cold

sodium bypass with valving to degrade sodium temperature to the reheater.

The French have not considered this matching a problem and apparently had

not yet, at the time of the visit, issued a set of detailed operating instructions for Phenix such as the U.K. has devised for PFR.

Though valves are used to establish proper flow to the superheater

and reheater, this is done on a one-time basis only; i.e., no continuous flow

control is provided in the Phenix and PFR systems. The Germans have

three modules per circuit in SNR and use trim valves (set only once) to

establish flow between the three modules. All three nations use fast-acting

(2 to 6 sec) sodium valves to isolate the steam-generator system in event

of a water-to- sodium leak. Extensive valving is also required on the steam

side in Phenix and PFR for reheater steam control. All three nations have

extensive sodium-valve development programs.

All countries consider the evaporator to be the major development

problem. The SNR Consortium plans to test only this component during the

next phase of testing, since they will have system information from their

initial systems testing in Hengelo. For the commercial LMFBRs, the

French and the British plan to integrate a test of both evaporator and superheater models. The British indicated that if sodium reheat is chosen for

the CFR, and a reheater test were desirable, it would be done separately.

There is a strong trend in all countries toward the development of large

steam-generator units, since they have confidence that larger units can be

built to a high enough integrity so that small-size modular designs are no

longer justified. On the other hand, none of the countries appear to have

reliable means of locating.and assessing tube damage in situ (such as might

be caused by tube leaks).

The French, Germans, and Dutch have extensive high-quality test

facilities for developing steam generators. Though there were no largemodel tests for the U.K. PFR, an in-depth steam-generator development

program for component parts was evident. Emphasis in all cases is placed

on evaporator development because of its complex features such as

thermal/hydraulic stability. Both the French and the Dutch have included

simulated turbine response in their integrated tests of steam generators

that incorporated sodium reheat.

Tests by the French at Les Renardieres on three full- scale Phenix

modules included steam- state heat-transfer performance, flow stability,

dynamic response to flow changes, simulated plant startup, shutdown, and

various transients.
6

The British, French, and Germans have conducted, and are continuing

to independently conduct, extensive sodium-water/steam reaction analysis

and testing with a major objective of developing advanced analytical methods

capable of describing the behavior of such reactions. Consensus was that,

for the near future, each new concept would require large-leak testing.

All the countries visited have developed mixing tees or mixing

chambers for mixing sodium streams of two unequal temperatures and do

not consider this a major developmental item. The British, French, and

Germans have developed and are testing advanced rupture-disk designs

capable of operating several years at normal conditions without loss of

capability to rupture at predetermined levels during an emergency. All the

nations are developing hydrogen detectors capable of quickly detecting

water/steam leaks. In situ calibration is being used in all cases. Fastacting sodium- isolation valves have been tested extensively by the French

and the Germans for valves of up to 600-mm (~24-in.) diameter--in a valve

loop at Les Renardieres by the French, and in the INTERATOM (Bensberg)

full- scale pump test loop by the Germans.

Industrial consortia are dominant in steam-generator design and

fabrication. Such consortia include INTERATOM (Germany), NERATOOM

(Netherlands), and Babcock-Atlantique and Stein Industrie (France). Largemodel testing and much of the sodium-water reaction testing are being

performed by government groups such as EDF and Les Renardieres (steamgenerator test), TNO at Hengelo (steam-generator test), CEA at Cadarache

(sodium-water reaction test), and UKAEA at Dounreay (sodium-water

reaction test).

D. Turbine Development for Steam Reheat

The Dutch, French, and Germans do not consider turbine development for steam reheat to be a problem. High-pressure turbines would be

similar to those used for fossil-fired plants, and the low-pressure turbines

are similar to those used in the light-water- reactor (LWR) industry. Steam

reheaters and moisture separators are similar to those used in the LWR

industry. Both the Dutch and French estimate a 1000-MWe steam-reheat

turbine to be lower in cost than a 1000-MWe sodium- reheat turbine. The

Germans expect no difficulty in building 3000- rpm steam- reheat units up

to 800 MWe and see no reason for not being able to extend this to 1000 MWe.

E. French Approach to Developmental Testing

The French divide their tests into four categories:

1. Tests on reduced-scale models whose results are extrapolated

to prototype conditions: for Phenix: 1/4- scale models of reactor jacket;

1/10-scale entire reactor unit for hydraulic tests; hydraulic tests of fullscale tubes for 1/4 segment of IHX.
7

2. Tests of prototypes in an environment simulating actual environments: full-scale tests of Phenix primary and secondary sodium pumps in

water loops.

3. Prototype tests in actual environment: Phenix reactor fuelhandling equipment; Phenix steam- generator modules; small equipment

such as control mechanisms; full- scale valve tests in sodium.

4. Safety te sts: Phenix steam- generator hydrogen- detection

system; sodium-water reaction tests.

F. Sodium-Water Reaction Programs and Test Sites

Major sodium-water- steam reaction programs have been and are

being conducted by the British, French, and Germans. Testing is centralized at one location and under the direction of one engineer. In the United

Kingdom, this testing is done by the UKAEA at Dounreay under A. Bray;

in France, by the CEA, at Cadarache under Lions; and in Germany by

INTERATOM at Bensberg under Dumm. Full-scale tube ruptures in

representative steam-generator models using many tubes is supplemented

by extensive mathematical modeling and by small-leak testing. Each new

concept is tested.

At Cadarache, the French plan to perform sodium-water reaction

tests in the two 45- MWt steam-generator models (B-A and Stein Industrie)

after they have been tested in the Les Renardiere s steam-generator test

facility.

The value of such testing is evidenced by following results obtained:

1. Effects of cover gas in the steam generator.

2. Optimum design and location of rupture disks.

3. Reaction-product dumping capability of the dump system.

4. Amount of reaction products left behind.

5. Capability for system recovery.

6. Pressure and temperature peaks--their location and duration.

7. Effects of steam-versus-water leaks.

8. Means of correlating mathematical models to tests.

G. Demonstration Plants

At the time of the visit, it was estimated that both PFR and Phenix

would be on line in 1973. SNR-300 was yet to start construction.
8

Important to note is the extensive developmental and proof-testing

conducted by each nation.

The French have conducted full- scale tests, at Cadarache, of the

Phenix reactor fuel-handling mechanisms; at Les Renardieres, full-scale

tests of Phenix steam-generator modules; full- scale water tests of the

Phenix pumps; full- scale water test of a 1/4 Phenix IHX segment; full- scale

sodium-valve tests conducted in a sodium loop at Les Renardieres; and

numerous other component and subcomponent tests in sodium, water, or gas.

The British have emphasized the developmental testing of component

parts. For example, steam-generator tube-to-tube- sheet joint development

went through 2 years of testing. Radial vibration of tubes against tube supports was tested, as well as fretting of tubes against tube supports from

vertical expansion. A 6000-gpm sodium pump was tested in sodium at

Risley at one-half the head and one-third the flow of the PFR pump. At the

time of the visit, the PFR rotating plug had been exhaustively tested at

Dounreay.

The Germans performed or were performing full- scale SNR- 300

pump and valve tests in a 600-mm (24-in.) sodium loop at INTERATOM,

Bensberg; full-scale gas and sodium tests of the SNR rotating plug; largescale sodium-water reaction testing at Bensberg; small- scale steamgenerator testing at Bensberg. At Hengelo, the Dutch were in the process

of testing a 50-MWt steam-generator model (compared to 82 MWt in

SNR- 300) and a 50-MWt SNR- 300 IHX.

H. Commercial Plants

Though all of the nations have announced intentions of building

commercial LMFBRs, there have been no. firm commitments. The latest

announcement by the British is that a commitment will be made in the latter part of the 1970's. The French may make a commitment in 1975.

Successful operation, in the near term, of both PFR and Phenix will

be a prerequisite for making commercial commitments. The commercial

Phenix-type design appears to be quite active and may well be the first

committed as indicated above. Reference designs have been made for a

1200-MWe LMFBR with four sodium loops. In July 1971, it was announced

that an association of three utilities was formed with the intention of building

such a plant. The utilities are EDF, ENPEE (1'Ente Nazionale Per l'Energia

Elettrica, Italy), and RWE (Rheinisch- Westfalisches Elekrizit.tswerk A. G.,

Germany).

Only the French have made a firm commitment of steam reheat for

their 1200-MWe LMFBR. The Germans are using steam reheat for reference, but have made no final decision, and the British, using sodium reheat

as reference, are in the process of reevaluating.
9

II. INTRODUCTION

Three thermodynamic systems can be considered for an LMFBR

plant: (1) nonreheat, (2) sodium reheat, and (3) steam reheat with live

steam or extraction steam. Moisture separation means can be used with

any of the schemes. A nonrehe-' system is shown in Fig. 1, a sodiumreheat system in Fig. 2, and a steam-reheat system in Fig. 3. Sodiumreheat arrangements can be made in several ways, as shown in Fig. 4.

The United Kingdom Plutonium Fast Reactor (PFR) and the French-Phenix

reactor use the parallel-series arrangement. Table 1 compares the main

thermodynamic parameters for the three European demonstration plants.

SUPERHEATER

INTERMEDIATE I

EACTOR HEAT

REIACONDENSER

EVAPORATOR

DEAERATOR

HIGH-PRESSURE LOW-PRESSURE

H THEATERS HEATERS

HP = High-pressure turbine L P= Low-pressure turbine

GEN. Generator

Fig. 1. Flow Diagram for LMFBR Nonreheat System

SUPER- HP IP LP LP LP GEN.

HEATER

REACTOR CONDENSER

IHX LOW-PRESSURE

I HEATERS

YEVAPO- RATOR DEAERATOR

HP = High-pressure turbine

IP = Intermediate-pressure turbine

HIGHPRESSURE LP= Low-pressure turbine

HEATERS GEN. - Generator

IHX - Intermediate heat exchanger

Fig. 2. Flow Diagram for LMFBR Sodium-reheat System
10

STEAM

REHEATER

SUPERHEATER HP IP LP LP LP GEN.

INTERMEDIATE

REACTOR HEAT

EXCHANGER CONDENSER

EVAPORATOR

DEAERATOR

I I HIGH-PRESSURE LOW-PRESSURE

I i HEATER S HEATERS

HP = High-pressure turbine

P - Intermediate-pressure turbine

LP - Low-pressure turbine

GEN. = Generator

Fig. 3. Flow Diagram for LMFBR Steam-reheat System

FROM REHEAT SODIUM

IHX TRIM VALVE

_

W = _= W

REHEAT SODIUM MIXING TEE

CONTROL VALVE

0 0

W- 0

TO IHX

PARALLEL-SERIES PARALLEL-SERIES SERIES

NOTE: STEAM/FEEDWATER LINES NOT SHOWN

Fig. 4. Arrangements for Sodium-reheat Steam Generator
11

TABLE 1. Parameters for LMFBR Demonstration Plants: U.K., France, and Germany

French

U.K. PFR Phenix SNR

Thermal output, MWt 600 563 720

Electrical output, MWe 270 250 300

Reactor outlet temp, OF 1044 1050 1000

Reactor inlet temp, *F 752 750 700

Sodium temp, steam generator out, OF 400 662 644

Sodium temp, steam generator in, OF 990 1022 977

Feedwater temp, OF 550 475 545

Steam temp, high-pressure turbine, *F 955 955 950

Steam pressure at main turbine, psig 2300 2400 2400

Type of reheat Sodium Sodium Steam

Reheat steam temp out, OF 960 955 428

Reheat pressure at turbine, psig 440 500 132

Steam generator type F.C.a O- Ta O- Ta

Modular No Yes Yes

No. of units or modules 9b 1 0 8 c 1 8 d

aF.C. = Forced recirculation; O- T = Once-through.

bThree evaporators, one per loop; three superheaters, one per loop; three sodium

reheaters, one per loop.

cEach of three secondary circuits contains a steam-generator unit consisting of

36 modules--12 evaporators, 12 superheaters, and 12 reheaters- -located in a boxlike

structure (see Item 9, Fig. 5).

dNine superheaters, three per loop; nine evaporators, three per loop.

The visit of the United States LMFBR Sodium Reheat Team from

May 22 through June 6, 1972, was made to determine the status of reheat

in LMFBR demonstration and commercial plants in France, Germany,

Netherlands, and the United Kingdom. In the process of determining this

status, discussions were also held on steam-generator design and development, associated components and systems design and development, sodiumwater reaction testing, steam-generator testing, and demonstration-plant

status.

Ten sites were visited, as indicated in Appendix A. These included

two 250-MWe LMFBR demonstration plants (the U.K.-PFR and the French

Phenix) and two 50-MWt steam-generator test sites: one at Les Renardieres,

France, and one at Hengelo, Netherlands.

Members of the team were as follows:

John G. Yevick (Team Leader)

Office of Program Analysis

Division of Reactor Development and Technology

U.S. Atomic Energy Commission

Washington, D.C.

Robert V. Laney, Associate Director

Engineering Research and Development

Argonne National Laboratory

Argonne, Illinois
12

Raymond E. Hoskins, Supervisor

Energy Research Section

Tennessee Valley Authority

Chattanooga, Tennessee

William Francis Rolf

Manager of Engineering

Commonwealth Edison

Chicago, Illinois

Edgar Rodwell, Project Manager

Steam Generator Development

Liquid Metal Engineering Center

Atomics International

Canoga Park, California

John A. Ford, Head*

Systems Section

Atomic Power Development Associates, Inc.

Detroit, Michigan

John S. McDonald, Manager

FBR Heat Transfer Systems Components

Atomics International

Canoga Park, California

Robert S. McQueen, Manager**

Heat Transfer Equipment Engineering

General Electric Company

Sunnyvale, California

Louis A. Sturiale, Manager

Liquid Metal Components, Engineering Division

Westinghouse Electric Company

Tampa, Florida

*Now with Division of Reactor Development and Technology, Atomic Energy Commission.

**Now Principal Project Engineer, Breeder Reactor Plant Engineering.
13

III. EUROPEAN STATUS AND POSITIONS ON SODIUM REHEAT

A. Introduction

Each of the Western European LMFBR development programs,

specifically the French, British, and the West- German/Dutch/Belgian

Luxembourg Consortium (hereinafter referred to as the SNR Consortium),

has some form of resuperheated steam. The following paragraphs cover

the status of the reheat programs, the results of economic studies, and the

overall reasoning that has led to the present reheat status of the above

European countries' LMFBR development programs.

B. Status of Sodium Reheat

1. Status Summary

Initially, all the Western European LMFBR programs included

the development of sodium reheat. The sodium-reheat plant cycle had been

selected early in each program as the basic cycle for the plant, primarily

because the standard turbine design apparently required some reheat.

Utilizing fossil-fired reheat was most adaptable to the use of sodium reheat,

and the user preferred this cycle. All the Western European programs have

carried out development work on at least the heat-transfer components;

some programs have carried the development effort through to the point of

inclusion of sodium reheat in the respective demonstration plants, as indicated in Table 2.

TABLE 2. Comparison of West European Plant-cycle Status

Demonstration Plant Commercial Plant

Scheduled Steam-generator Committed Steam-generator

West European Size, Service Model Test Service Model Test

Program MWe/Name Cycle Date Completed Size, MWe Cycle Date Completed

French 250/Phenix Sodium reheat 1973 a. Small heat-transfer model 1000 Steam reheat Not announced a. Plan to test a

test of -5000 hr completed (now being 45-MWt helicalbefore 1972. designed) tube model in

b. Three modules (E, S, and 1973.

RIa of Phenix steam gen- b. Plan to test a erator with -6000 hr to 45-MWt straightbe completed in 1972. tube model in

1974.

British 250/PFR Sodium reheat 1973 No model testsb 1300 Decision on cycle to be Not announced May test a full-length

made in early 1973.c (now being and tube-diameter, (Present reference-- designed) partial-tube-sheet

sodium reheat.) segment model, probably around 1974.

SNR Consortium 300/SNR-300 Steam reheatd 1978 Three-module 50-MWt steam- 2

000 Steam reheat is Not announced No test plans or (West Germany, generator model (E, S, and RI present reference. (making trade- schedule announced.

Netherlands, test to end in 1973; now on off studies)

Belgium, test.e

Luxembourg)

aE evaporator; S = superheater; R - reheater.

bAlthough the British did not do a scale-model test of complete PFR steam generators, heat-transfer stability tests were made along with fabrication tests. Also, the British,

like the West Germans and French, did do a model test of the response of the demonstration plant's steam-generator system to sodium-water reactions by simulating tube

failures. (See Section V for details of steam-generator development programs.) cPresent reference is sodium reheat.

dThe SNR Consortium was initially committed to a sodium-reheat plant cycle and initiated a steam-generator development effort that included sodium reheat. In 1971, the

program was reoriented to develop steam generators without sodium reheat. SNR-300 plant is being designed with a steam-reheat cycle. eFollowing completion of tests of the early-design SNR steam-generator model, a 50-MWt helical-tube evaporator model for SNR-300 will be tested, scheduled to be completed in 1974. IThe SNR Consortium's commercial-plant reference size is 2000 MWe.
14

At this time, for commercial LMFBRs, the French and probably the German LMFBR programs have elected to use steam reheat in

lieu of sodium reheat. The British are reevaluating their reference CFR

sodium reheat. The SNR Consortium will make a reevaluation prior to a

final decision.

Experience with sodium-reheat equipment and components at

this time is limited to the data obtained from model and module sodiumloop tests by the French in the Phenix plant's steam-generator developmenttest effort, and the Dutch have started to test an early design of the SNR-300

steam-generator system, which includes a sodium-reheater model, though

the SNR-300 does not have sodium reheat- -instead, it has steam reheat.

(For more details on tests on the steam-generator loop, see Section IV.)

Although European experience with sodium reheaters in test loops appears

successful, no plant to date has operated with a sodium-reheat cycle. The

first plant experience with a sodium-reheat system will be in 1973, when

both the U.K.-PFR and the French-Phenix demonstration plants are scheduled to operate.

2. Status Comparison

Table 2 indicates the status of the plans for the plant cycles for

the various West European demonstration and commercial plants. Also

indicated in the table are the status and plans for the related steamgenerator test programs. (For more details on steam-generator test programs, see Section V.)

a. French

The French have developed a sodium-reheat system for

their LMFBR demonstration plant- -Phenix, a 250-MWe (nominal) unit

scheduled for service in 1973. The steam-generator model consisting of

three full-sized modules (evaporator E, superheater S, and reheater R) of

the Phenix plant's steam-generator design will have completed more than

6000 hr of testing in a steam-generator test loop by the middle of 1972

(tests scheduled to end in July 1972). The Phenix plant contains three loops

with a 36-module steam generator in each loop: 12E, 12S, and 12R.

For the French commercial plant, the designers have determined that (1) the large number of modules used in Phenix is uneconomical and (2) the sodium-reheat plant cycle is not as economical as the

expanded- steam- reheat plant cycle, and (3) there is a reliability advantage

with its concomitant availability advantage in favor of steam reheat. Consequently, the steam-generator development plans for the commercial plant

call for testing two different once-through steam-generator designs, neither

of which includes sodium reheat.
15

The commercial plant's steam-generator model is a 45-MWt

Babcock-Atlantique design with helical tubes. This first model is to be

tested in 1973. The commercial plant's layout incorporating this type of

design will have one or two steam generators in each of the four plant loops.

Each steam generator will consist of one evaporator and one superheater

combined in a single shell.

The commercial plant's alternative steam-generator model

is a 45-MWt Stein Industrie design with straight tubes. It is to be tested in

1974. The commercial plant's layout using this alternative type of design

will have one steam generator in each of the four plant loops. Each steam

generator will consist of one evaporator and one superheater, each in a

separate shell. The preliminary data for the commercial plant's steam

cycle are shown in Table 3.

TABLE 3. French Commercial LMFBR Steam-cycle Conditions

Main steam *C-F 510-950

bars-psig 165-2378

Main steam flow kg/sec-106 lb/hr 1040-8.4

Intermediate-pressure-turbine *C-F 164-329

exhaust steam to steam reheater bars-psig 7-87

Reheated steam to low pressure * C-F 285-545

turbines bars-psig 6.6-81

Steam from high-pressure C- F 473-883

turbine to steam reheater bars-psig 85-1218

The turbine will be a 3000-rpm tandem-compound machine

with a Peerless-type steam dryer (similar to Westinghouse PWR). (For

additional details on plant cycle, see Section IX.)

b. British

The British demonstration plant (PFR), rated at 250 MWe

(nominal), is scheduled to start up in 1973 and will use a sodium-reheat

plant cycle. As noted in Table 2, the British, in lieu of testing models of

the PFR steam generators, have performed supporting tests to ensure design and fabrication requirements and have made extensive analyses including the development of comprehensive operating procedures.

Reference plans for the British Commercial Fast Reactor

(CFR) call for the utilization of two 650-MWe "standard" high-temperature

reheat turbine generators (T-G's) with reheat steam supplied by sodium

reheaters. Experience with the development of PFR operating procedures,

including those concerning startup, transients, and shutdown, have caused
16

concern that these problems may be multiplied with the two parallel turbinegenerator units in the CFR plant. The British are now in the process of reevaluating the plant cycle of the CFR. Plans are to reach a decision on the

CFR plant cycle in early 1973.

Large-scale model tests of the CFR plant's steam generators will be made if the CFR incorporates a new steam-generator concept.

No firm plans or schedule were presented, but general ideas were discussed,

and it is assumed that the schedule for such model tests would be about 1974.

(See Section V for more details on British steam-generator testing

considerations.)

c. SNR Consortium

As noted in Table 2, the initial design of the SNR-300

demonstration plant included a sodium-reheat cycle. The design of the

plant progressed to the point where a 45-MWt straight-tube, bellows-inshell steam-generator model, which included a sodium-reheat module, was

fabricated for testing at Hengelo. The model, including the sodium reheater,

was installed in the Hengelo Test Loop late in 1971. The decision to eliminate sodium reheat in SNR was also made in 1971. Redesign of either the

model or the loop to accommodate elimination of the reheater would have

taken too long and been too expensive.

After completion of the tests on this steam-generator

model, probably 1973, an SNR 50-MWt helical-tube evaporator will be installed and tested. Its testing is scheduled to be completed in 1974.

The commercial SNR plant reference design uses a steamreheat plant cycle similar to that now being designed for SNR-300. There

were indications that the SNR Consortium would review the commercial

SNR cycle decisions prior to final choice.

C. Reasons Expressed for the Plant-cycle Selection

1. Introduction and Summary

This section discusses the economic and technical reasoning

that led the various Western European nations to the present reference

cycles.

The overall observation is that all the Western European programs have reconsidered, or are now in the process of reappraising, the

initial selection of the sodium-reheat plant cycle for the future plants now

in the design stage. The general reason expressed for the initial selection

of sodium reheat for the Western European programs, and specifically for

the demonstration plants, was that the sodium-heat transfer system was
17

designed around a "standard" turbine-generator (T-G) steam cycle. This

"standard" T-G cycle was developed for fossil-fueled plants and included

reheat requirements that had been optimized for fossil-fueled plants.

Two of the demonstration plants, the PFR plant of the British and the

Phenix plant of the French, were developed to use 250-MWe (nominal)

"standard" T-G units. Also, initial studies indicated favorable economics

for a sodium-reheat cycle.

2. SNR Consortium

In 1971, the SNR Consortium decided to switch from a sodiumreheat cycle to a steam-reheat cycle for the SNR-300 demonstration plant.

Both NERATOOM and INTERATOM studies indicated that the economics of

the steam-reheat cycle were more attractive than those for sodium reheat.

NERATOOM, which made parametric economic studies for the

SNR-300 plant only, indicated that the studies showed about 1*% total-energyproduction cost advantage for the steam-reheat cycle over the sodiumreheat cycle. It was indicated that the complexity of the sodium-reheat

cycle was of concern and a small economic penalty could be taken to avoid

the problems specific to the sodium-reheat cycle.

INTERATOM made-parametric economic studies of various

plant cycles for both the SNR-300 and 1000-MWe plants. Results of the

INTERATOM study indicated a 0.25 and 0.3% total-energy-production cost

advantage for steam reheat over sodium reheat for the SNR-300 and

1000-MWe plants, respectively. It was indicated that had the economics

been in favor of sodium reheat by the same magnitude, sodium reheat would

have been recommended.

3. French

The French made the decision (in 1971) to change the reference

plant cycle from sodium reheat to steam reheat. The change, for the French,

applies to the commercial 1000-MWe plant cycle and not to the French demonstration plant Phenix, because the change was made after the Phenix plant

was well along in the construction stage.

The French economic studies for the 1000-MWe plant indicated

a one- to two-percent total-energy-production cost advantage for the steamreheat cycle over the sodium-reheat cycle. This economic incentive, plus

the French assessment of the potential availability advantage (roughly estimated in the range of 1/2 to 1%) in favor of steam reheat, were the reasons

indicated for deciding to use the steam-reheat cycle in lieu of sodium reheat for the 1000-MWe French commercial plant. Earlier preliminary

studies indicated nonreheat was economically unacceptable, and it was

discarded at an early stage in the French commercial-plant considerations.
18

4. United Kingdom

The British have not made a final decision on the type of commercial LMFBR steam cycle. The stated plan is to make the plant-cycle

decision for the 1300-MWe CFR in early 1973.

Reference design for the CFR is to use two "standard" 650-MWe

T-G's in parallel, and sodium reheat. The experience gained on PFR indicates that the startup and shutdown thermal-transient control systems and

procedures would make the parallel operation of the two CFR T-G's difficult.

Also, the British are studying the reliability and operational considerations

associated with the added components and systems in the sodium-reheat

system and apparently plan to weigh this against reliability considerations

of developing T-G concepts associated with steam-reheat or nonreheat

cycles. Reliability, operation, equipment vendors' capability, and energygeneration costs are factors considered in the studies.
19

IV. STEAM GENERATORS AND ASSOCIATED SYSTEMS DESIGN

A. Summary

European LMFBR R&D programs, in general, emphasize testing of

large sodium-component prototypes or large-scale models, such as pumps,

heat exchangers, and valves. The French have tested full-scale steam

generators for Phenix, and the SNR Consortium is testing large steamgenerator models for the SNR-300. The French are fabricating two 45-MWt

models of a different design than Phenix for their commercial LMFBR

plant. Full-scale sodium valve tests have been performed by all the nations for their LMFBR demonstration plants. Full-scale multitube steamgenerator sodium-water reactions have been and are being conducted by

France, Germany, and the U.K. The various component tests are indicative

of a general desire to verify component and system reliability by proof

testing before operation. In this respect, the SNR Consortium emphasized

that all major SNR components have been or will be tested at full scale or

as close to full scale as possible. This includes testing of heat exchangers,

pumps, and valves.

Full-scale proof testing of components has not been pursued in the

PFR program. There are indications that the British will test a large

steam-generator model if their first commercial LMFBR plant adopts a

steam-generator concept different from the PFR.

To date both the French and the SNR Consortium have tested the

steam generator as part of a system. In this approach, the component

performance can be individually determined in addition to the verification

of the component's interaction with the system and other components. The

French and the SNR Consortium have in operation 50-MWt facilities for

testing large steam generators- -namely, the CGVS at Les Renardieres and

the Hengelo Steam Generator Test Facility in the Netherlands.

The Phenix steam generators are once-through, modular, serpentinetype units. The SNR-300 steam generators are modular, straight-tube units,

which incorporate a bellows in the shell for differential expansion. The PFR

evaporator, superheater, and reheater are basically variations of a rather

conventional U-tube bundle. Phenix and PFR are both sodium-reheat plants;

the SNR-300 uses steam reheat. The design philosophy for Phenix, aimed

at achieving high availability, was dominated by consideration of a potential

sodium-water reaction. This is evidenced both by the extreme modularization of the steam generator (108 modules per three-loop plant) and by

features provided for removal of modules from the loop, such as rollers

installed on each module assembly. As stated by the French, sufficient

uncertainties and unknown consequences were associated with the sodiumwater reaction to have warranted conservatism in the Phenix design. The

French feel that the tests for Phenix have provided sufficient confidence in
20

steam-generator performance and reliability to enable a less conservative

and more economical approach for the large commercial plants. This will

be confirmed in Phenix operation.

In France, Babcock-Atlantique (B-A) and Stein Industrie are each

developing large, full-size steam generators for the 1200-MWe Phenixtype commercial-plant. A 45-MWt model of B-A's 600-MWt unit has been

designed and fabricated, and is scheduled for testing at Les Renardibres

in February 1973; testing of a Stein Industrie 45-MWt model is presently

scheduled for completion by the end of 1974. One of these steam generators

will be selected as the reference design for the 1200-MWe plant by early

1975. Results of the Les Renardieres tests will be an influential factor in

the choice.

None of the countries visited considered the sodium reheater, as a

component, to be a major developmental task. The U.K. steam generator,

which includes sodium reheat, is characterized by an extensive control

system and detailed operating procedures to ensure proper match of sodium reheater with turbine operation. (See Section IV.C.2 for details.)

In summary, the Europeans are well into the hardware and operational phase of steam-generator development for the near-term LMFBR

power plants. Emphasis is on large-scale model testing with emphasis on

the verification of performance and the development of sodium-water

reaction-analysis technology.

Although the steam generators being developed may not be prototypic

of commercial-plant units, the experience obtained from large-scale testing and from the operation of the demonstration-plant units will provide an

excellent springboard from which to design and/or develop commercial

LMFBR steam generators.

B. France

1. Introduction

Phenix is a sodium-cooled fast-breeder-reactor demonstration

plant scheduled for criticality in 1973. The first reactor in the French

LMFBR development program was the Rapsodie experimental reactor, which

has been operative at the Cadarache Nuclear Research Center since 1967.

Rapsodie has become, in the main, a fuel-development facility similar to

the U.S. EBR-II reactor. Rapsodie has no electrical power-generating

equipment (i.e., no steam generators); in 1970, the thermal power of the

facility was increased from 20 to 40 MW.

Phenix, with a plant rating of 250 MWe (563 MWt), represents

the transition to electrical power generaL:,n ad will provide the industrial
21

experience essential for development of the larger commercial power

plants. The Phenix plant uses a sodium-reheat cycle; the 1200-MWe plant

design will use a steam-reheat cycle.

2. Phenix Steam Generators

The steam generators developed for the Phenix are not prototypic of the steam generators being developed in France for the large commercial LMFBR plants. For the three-loop Phenix reactor there are

108 steam-generator modules: 36 evaporators, 36 superheaters, and 3 6 reheaters. Each loop contains 12 sets of once-through steam-generator

modules. (See Fig. 5.)

12

11

10

S67

STEAM

GENERATOR

1 sodium inlet

2 outlet collector

3 water separator

4 sodium outlet

5 to hydrogen

separator

6 rupture disk

7 water drain valve

8 water inlet

9 insulated casing

10 superheated

steam

11 to reheater

12 reheated steam

A reheater

B superheater

C economiser-evap.

Fig. 5. Phenix Steam Generator
22

Modular steam generators were preferred for the Phenix because of economic and reliability considerations. The use of modules was

intended to limit or decrease the consequences of sodium-water reactions.

A significant benefit derived from modularization of the steam

generators was that a full-scale modular system could be prooftested at

Les Renardieres. Prooftesting with full-scale modules would permit direct

verification of the thermal-hydraulic performance and vibration characteristics. Also, in the event of any sodium-water reaction during the tests,

the observed behavior of a full-scale module would be of potential benefit

to Phenix operation. A successful test of a Phenix module at Les Renardieres

for more than 6000 hr has substantiated the design choices.

With modules, in-service maintenance tasks would be much

easier to perform. For example, in the event of a tube failure, tasks such

as physically locating the leaking tube and/or removing the failed module

from the loop would appear to be less difficult with modules. Simplifying

maintenance activities, or in essence, decreasing the time required for

maintenance would provide a realistic and practical way to increase component and plant availability. Presumably, fabrication of modules (versus

full-size units) could be more stringently controlled, and this specific control should result in a higher-quality unit. The French stressed that high

integrity was essential for safe sodium systems. Nevertheless, it is interesting to note that the French commercial plant will not be modularized

(see Section IV.B.3).

For safety purposes, the modules for each loop and the water

and sodium headers are installed inside an insulated, self-supported, rectangular containment box. The modules themselves are not insulated. In

the event of loss of the water/steam system, the box can be opened and aircooled by natural convection. For firefighting purposes, argon could be

substituted for the air atmosphere. Each evaporator, superheater, and

reheater module is a counterflow heat exchanger, consisting of a seventube bundle. The units do not require tube sheets because the tubes are

bent and penetrate the shell, terminating in an outside water or steam

header. Local thermal shielding is provided at the welded tube-shell junction to guard against damaging thermal shocks. Orifices are installed in

the inlet to the water tubes (evaporator), at the header, to ensure stable

operation at low power levels. Nominal steam-generator characteristics

are presented in Table 4.

To permit tube plugging in the event of a leak, a small nipple

is provided at each tube end (see Fig. 6). Direct access to the orificed

tube end is achieved by removing the nipple. For other than in-place repair, the modules can be removed to a work area for repair and/or replacement. The individual modules are mounted on small rollers for ease

of removability.
23

TABLE 4. Nominal Phenix Steam-generator Characteristics

Total Steam

Evaporator Superheater Reheater Generator

Thermal power, MWt 356 121 86 563

Sodium flow rate, kg/sec 2210 1215 995 2210

Sodium inlet temperature, *C 475 550 550 550

Sodium outlet temperature, *C 350 475 475 350

Water or steam inlet temperature, *C 246 375 308 246

Steam outlet temperature, *C 375 512 512 512

Water or steam inlet pressure, bars 193 172.7 36.2 193

Steam outlet pressure, bars 174.9 168 34.9 168

Water flow rate, kg/sec 209 209 186 209

Total number of modules 36 36 36 36 x 3 = 108

Material Ferritica TP-321 M TP-321 M

Tube, OD x wall, mm 28 x 4 31.8 x 3.6 42.4 x 2

aFor test purpose, two sets of 12 modules are made of 2.25 Cr-1 Mo, and one set of 12 modules

is made of 2.25 Cr-1 Mo-1 Nb.

pU

Fig. 6. Phenix Steam Generator--Water Inlet Header

Rapid-closing (6-sec) sodium shutoff valves (ball-type) are included at the exit and entrance of the steam-generator unit in each loop to

isolate it during a sodium-water reaction. The heat exchangers are designed and instrumented to provide rapid dumping of both the water/steam

and sodium.

Ferritic steel (24 Cr-1 Mo) is used in evaporators to eliminate

the potential chloride stress corrosion in regions in which water is boiling.

In an attempt to gain maximum knowledge from operation of the Phenix

steam generators, both stabilized (with niobium) and unstabilized ferritic

steel is used. The superheaters and reheaters use austenitic TP-321 steel,

primarily because this steel shows good creep properties. Additionally,

experience has been gained through the frequent use of stainless steel in

the superheaters of the Electricite de France (EDF) 250-MWe fossil plants
24

with 565*C (1050 F) superheated steam. Welding problems have been encountered in welding stabilized Croloy, but welding feasibility has been

established. EDF confirmed that Stein Industrie intends to use unstabilized

Croloy for the 1200-MWe steam generator.

Steady-state temperature profiles are shown in Figs. 7-9. The

French were confident that the Phenix steam generators could safelyhandle

plant transients. The worst transient occurs when a feedwater pump trips

off. Since sodium can continue to flow, the evaporator dries out. The evaporator is then hit with 150 C water from the emergency feedwater supply,

developing the worst transient case. Thermal-shock tests have been conducted to ensure that there will be no failure for such a condition. This

transient was repeated for about 22-24 cycles before failure occurred.

The French expressed concern about the movement between the

tubes and shell caused by differential expansion and also about the effect of

plugged tubes where the tube penetrates the shell. The latter concern was

due to a potential for fatigue failure. Large forces are imposed on this

penetration if no relative motion between the tube and its support is assumed. Calculations indicate that the Phenix modules are safe for several

thousand cycles if this motion occurs. The problem was originally anticipated in the superheater and reheater, where both the tubes and the tube

supports were stainless steel. No such problem is now believed to exist,

however.

A butterfly-type control valve is used to proportion the sodium

flow between the superheater and reheater. This flow-control valve is set

during the initial operation and not varied with load. Phenix has a mixing

tee at the sodium outlet of the superheater and reheater. However, the

French stated that this tee is not essential because the temperature differentials expected between superheater and reheater sodium outlet (Figs. 7-

9) are not excessive. This temperature difference does not change in magnitude during a turbine trip. Large thermal transients induced by turbine

trips would not occur, since the steam flow is quickly reduced in sufficient

time to prevent damage. A steam-side bypass around the reheater is provided. The steam flow is turned off within 0.2 sec, and this provides sufficient time to prevent damage to the steam generator. Means are provided

on the steam side to isolate both the superheater and the reheater, allowing

water and steam circulation through the evaporator, moisture separator,

feedwater heaters, and reheaters.

EDF indicated that the steam-generator tests at Les Renardieres

model or prototype the Phenix-plant system conditions as closely as possible. This includes transients such as turbine trip. Although everytransient

could not be simulated at Les Renardieres (because of possible damage to

the sodium heaters), EDF calculations indicate that the Phenix steam generators could safely accommodate the turbine-trip transient. Rapid shutdown
PRESSURE: I CM= I BAR

BROKEN CURVES' TEMPERATURE OF INSIDE AND OUTSIDE WALLS

186.621,

- I

350.0-

- s

30

LENGTH, m

40

-47.9L

I I

I I

37. I

I I L

50 56.28 0

SODIUM TEMPERATURE

STEAM.

TEMPERATURE

175.17

PRESSURE

SUPERHEATER

I0

LENGTH, m

-1550.01

-5I2.I

1475.9

I I

I I

I I.

308.01-

1

SODIUM TEMPERATURE

STEAM

TEMPERATURE

37.20

- ---- PRESSURE

REHEATER

-J5 5 0 .0

A512.0

I 

~3404

I I I

20 26.21 0 10 20.49

LENGTH, m

Fig. 7. Temperature Profiles for Phenix Steam Generator; Normal Power Level (563 MW); Cold Condition

N

600

500E

400[

IQ

W

FPRESSURE

SODIUM TEMPERATURES . .--

. . . - - '75.75

.. .-- '"' "''" WATER TEMPERATURE

EVAPORATOR

3001

0 I0 20 , 
. PRESSUF -- - - - - - - - - -- ---------- U

Sc

PRESSURE, I CM = I BAR

BROKEN CURVES= TEMPERATURE OF INSIDE AND OUTSIDE WALLS

SODIUM I I TEMPERATURE ..

RE IEM

--------------- 168.50..--. PRESSURE 168.54 1 --- i---- - RES E 54.7 . 8 449.8-

)DIUM TEMPERATURE - - ' STEAM TEMPERATURE

.. - ......-.-." -- 3-7--- . 

.. :.--.-- . -7

WATER TEMPERATURE

EVAPORATOR

I0 20 30

LENGTH,m

40

-- I.

I I

I I

I I

I I

I I

I I

I I

I I

I I

L 50 56.28 0

SUPERHEATER

10

LENGTH, m

1519.81

j512.Og

I I

I I

'4 I

'443.1 L I I

I I

I I

I I

I I

I I

I I

I I

I I.

I I

I I

,

2550r I I

I I

I I

I I

I I

SODIUM

TEMPERATURE

STEAM

TEMPERATURE

9.50 I PRESSURE - -

REHEATER

-1519.8

.1506.2

18.42

SI II

20 26.21 0 10 20.49

LENGTH, m

Fig. 8. Temperature Profiles for Phenix Steam Generator; 0.25 of Normal Flow Pressure: 168 bars; Cold Condition

500

0 400

a.

300

200

16947r

- I

306.7-

- I

17

- 

0

I 

1 I I 

.

._--.'
PRESSURE: I CM= I BAR

BROKEN CURVES: TEMPERATURE OF

48.49- - - ------- - - - -_______ PRESSURE

-------------------

4.5

SODIUM TEMPERATURE

S. ..-

500

U

. 400

Ia

w 300

200

-I

1770

464.4

I456.0

I

S375.0

INSIDE AND OUTSIDE WALLS

SUPERATURE

'.-' STEAM TEMPERATURE

44.30

-- PRESSURE

42

SUPERHEATER

513.0 I 1512.0 

1 

442II

444.2

1328.0

I 

I

I II

I I

I 1

1 I

SODIUM

TEMPERATURES

-- STEAM

TEMPERATURE

9.30

PRESSURE

REHEATER

I I I

50 56.28 0 10

LENGTH, m

I I I 20 26.21 0 10 20.49

LENGTH, m

Fig. 9. Temperature Profiles for Phenix Steam Generator; 0.25 of Normal Flow Pressure: 42 bars; Cold Condition

WATER TEMPERATURE

EVAPORATOR

-4513.0

-1502.9

18.38

0 I0 20 30

LENGTH, m 40

_. _. w trs Tr "f1r T1 r

...

I i

l . . . l 1

- --

..--
28

transients have been conducted at Les Renardieres to ensure this capability. Turbine trip, followed by loss of water and steam flow, will be simulated during the last test phase at Les Renardieres.

Instrumentation used in the module tests included orifices in

the tube inlet, both to measure flow and to provide hydraulic stability. (The

orifice size was calculated for the smallest usable orifice initially and

enlarged as the tests progressed.)

3. Steam Generators under Development for 1000-MWe Plant

Two French companies--Babcock-Atlantique and Stein Industrie--

have joined Commissariat a L'Energie Atomique (CEA) and EDF in a steamgenerator program. Because of economics, the Phenix steam generators

will not be used in the 1000-MWe commercial LMFBR plants. Economic

studies conducted by the French indicate that for a large plant the number

of required modules would make the cost of components, piping, and instrumentation prohibitive. The decision to proceed with the large units for

the commercial plant was also based on the excellent experience obtained

in the steam-generator tests at Les Renardieres, the sodium-water reaction tests at Cadarache, and the sodium-water reaction analyses at Saclay

and Cadarache. Operational experience and sodium-water activities have

indicated that (1) sodium-water leaks could be detected at an early stage,

(2) the relief system is effective, and (3) the tubes and tube-to-tube-sheet

joints could be satisfactorily made for reliable service.

a. Babcock-Atlantique (Fig. 10)

The tube bundles for the B-A steam generator are helically

coiled. This type of generator is suited to a steam cycle without sodium

reheat. Sodium at 1020 F and water at 465F enter the unit and produce

superheated steam at 950 F (2450 psi). The tubes are made of Incoloy 800,

the water and steam connections of 24 Cr-1 Mo. The shell is of Type 304

stainless steel. The tubes--of which there are 174 if the steam generator

is 290 MWt or about 348 if 580 MWt--are 25-mm OD x 2.5-mm wall thickness. A small 1.5-MWt model has been tested at Grand Quevilly to determine thermal performance and stability. Also, nonsodium waterside tests

have been performed at Les Renardieres. A model (see Fig. 10) with 24tubes

forming three rows between two concentric shells will go on test in

Les Renardieres in February 1973.

b. Stein Industrie (Fig. 11)

The Stein Industrie steam generator is a counterflow unit

using straight tubes with a Z-type expansion bend in the gas space below

the upper head. Each module is 580 MWt; four units would be required for

a 1000-MWe plant. The steam-generator reference material for the tubes
29

Superheat

Steam

Level 18.8 m

-. ,Na

Separator

Na

Helical

Tube s - 0-1011

aN

Single N

Helical

Tube

Thermal Ruptureb Sleeve Disc o1 45 MWt Model

-,0 -w .For Lea Renardi'eres

fee Test Facility

Tubesheet Rupture

Disc

Tube-

- Na

Level 0

Feedwater See Detail "X"

One of 4 feedwater tubesheets

Feedwater toroidal header

Feedwater y

Na .Exit Pipe

Detail "X"

Note: - '2200 indicates Diameter =

2200 millimeters

Fig. 10. Babcock-Atlantique Steam Generator (two units in parallel; each 290 MWt)
Superheat Level - 22 m. Steam

Tubesheet

Separator Reverse t

bend

Na

Single

Tube

1440- 9- 1884

0

41

Sw

Rupture

Disc

Rupture

Disc Na.

Levei

See Detail "X"

Feedwater

Orifice

,4- -4

ii

-# 645

Superheater Evaporator

45 MWt - Model

For Les Renardi'eres

Test Facility

Detail "X"

Note: -#565 indicates Diameter =

565 millimeters

Fig. 11. Stein Industrie Steam Generator (580-MWt unit)

30

ibe
31

is unstabilized 2 Cr-1 Mo. Each evaporator has 1400 tubes of 18-mm OD x

3-mm wall thickness. Each superheater has 463 tubes of 21.3-mm OD x

2-mm wall thickness. Shell and tube sheets are Type 316 stainless steel.

A 45-MWt (see Fig. 11) model will be tested at

Les Renardieres in 1974. This model will have a 44-tube bundle. After

completion of these tests, the model will be shipped to Cadarache for use

in sodium-water reaction tests.

c. Comparison of Phenix-type, B-A, and Stein Industrie

Commercial Units

Table 5 compares on a per-MWe basis, the weight, volume,

and heat-transfer area for the commercial LMFBRs--Phenix type; B-A large

unit; and Stein Industrie large unit.

TABLE 5. Comparison of French Commercial 1000-MWe LMFBR Steam-generator Concepts

No. of Modules Total No. of Modules Weight, Volume, Heat-transfer

Concept per Loop for Four Loops Tonnes/MWe m3/MWe Area,a mZ/MWe

Phenix 36 144 1.05 15.8 11

B-A 1-2 4-8 0.56 1.73 8.65

SteinIndustrie 1 4 0.70 1.4 8.3

aOn OIS of tube.

d. Les Renardieres Steam Generator Test Facility

France has built and is operating many sodium facilities

in support of the LMFBR development. Facilities of interest to steamgenerator development and visited by the Sodium Reheat Team are located

at two major sites: Cadarache, operated by the CEA; and Les Renardieres,

operated by the EDF. Les Renardieres is the site of the 50-MWt steamgenerator test facility (CGVS). The French considered it necessary to

operate such a facility (1) to test industrial-scale steam-generator components under conditions simulating an LMFBR plant, and (2) to confirm

the endurance of the component over long time periods. Features of the

CGVS facility are:

(1) A gas-heated boiler, capable of heating sodium to

about 1200 F. The primary sodium test circuit is designed to simulate

temperature transients.

(2) A water/steam circuit, where the steam produced

by the steam generator is expanded, condensed, and returned to the steam

generator. This loop also includes water treatment for condensate and

makeup water.
32

The characteristics of these circuits are:

Sodium

Maximum heater power 50 MWt

Maximum sodium temperature 1200*F

Maximum sodium flow rate 1000 m3 /hr

Water/steam

Steam pressure (at outlet) 2450 psi

Maximum steam outlet temperature 1050*F

Feedwater temperature 480-535F

Flow 80 tons/hr

The test schedule for CGVS provides for the Phenix steamgenerator tests to be completed in 1972, the Babcock-Atlantique 45-MW

steam generator to be on site by November 1972, and tests of the

Stein Industrie 45-MWt model to be completed in 1974.

Steam-generator tests will concentrate on commercialoperation aspects and the resultant operating problems. The steamgenerator models will be tested under various plant transient conditions,

including turbine-trip and reactor-scram events.

e. Sodium-Water Reaction Test Facility at Cadarache

The CEA's sodium-water reaction test facility is located

at the Cadarache Technical Center. The sodium-water reaction studies

are focused primarily on large-leak tests. However, small leaks of water

in sodium have been investigated for the Phenix steam generator. The

large-leak investigations are concentrated on the experimental and theoretical analysis of effects of a large leak on the steam generator and the

associated sodium loop. Preliminary large sodium-water tests are carried

out in simplified models to develop theoretical codes for predicting the

time and pressure histories associated with a reaction. Further tests are

performed in a full-size evaporator model to confirm predictions for the

Phenix steam generator. Section VI.B contains further information on this.

C. United Kingdom: PFR Steam Generators

1. Description

The Prototype Fast Reactor (PFR) is a 250-MWe sodiumcooled fast reactor located at Dounreay, Scotland. The power output of

the PFR is based on the requirement that it should be the smallest reactor capable of providing, with confidence, the information necessary for the

installationand operation of commercial fast reactors in the United Kingdom.

The main features of the PFR, relative to the steam system, is the use of

single-wall-tubed heat exchangers and steam conditions of 960F and 2300 psi,

and a forced-recirculation system.
33

The PFR uses three heat-transfer loops. The steam generators

are shell-and-U-tube heat exchangers with provisions for isolating the effects

of a sodium-water reaction. Figures 12 and 13 present a schematic of the

2300 psiq, 1000 *F

- -674 *F

LPT IP HPTARN

SUPPLY

500psi , 600 F

_ _16

LPT=LOW-PRESSURE TURBINE 20~m 

I PT =INTERMEDIATE -PRESSURE TURBINE a -- REHEATER

HPT=HIGH-PRESSURE TURBINEUE

NEEDLE

-BYPASS II4 ~~~VALVEIIIIII

MIXING TEE 6" (2) _ 6(6

ORIFICE 1040F 14" (2)

TO IH X705 *F 14 (2)

24

Fig. 12. Schematic of Heat-removal System for PFR Steam Generator

TEMPERATURE

6W C

562C (1 044F)

550 C, 169 kg/cm2

(2 400 p.s.i.s.)

- 532C (990F)

T.S.V. CONDITIONS 514C' 516C (960FI

500C 513-C (955F) /163 kg/cm 2 (2 315 p s i.s.) (957F) N 32 5 kg/cm2

49 ,30 0O O

450C - 910 k 6as I 455-C (851F) N1 ;1

400C

1752'F) CO

370"C

(69 F) 1 250 kg/sec (9 91 - 106 lbs /hr ) 353C

350C (668'F) 173 kg/cm 2 (2 465 p s I a 

353C (668F) 174 kg/cm2 (2 475 p si a I 343'C (650F)

300CDRMH34 8 kg/cm 2 305C (581*F) 30C DRUM HEATING (495 p s i a)

288C (550F) CONTRA-FLOW

PARALLEL FLOW EVAPORATORS CONTRA-FLOW SUPERHEATERS REHEATERS

250C

0 10 20 30 40 50 60 70 80 90 100

% HEAT TRANSFERRED

Fig. 13. Temperature/Heat Diagram for PFR Steam Generator 1 (600 MW). (Numbers

do not necessarily coincide with Figs. 12 and 42 or Tables 5, 10, and 11.)
34

PFR heat-removal system and a temperature/heat diagram for the plant,

respectively. While the reheater and its associated reheat system were the

major topic for discussion during the visit, a brief description of the heat

exchangers is included here for information.

The choice of the PFR steam generators was influenced by what

was considered to be the necessity of demonstrating that single-wall sodiumto-water/steam heat exchangers could be designed and operated safely with

high availability. The Nuclear Power Group (TNPG) and the Central Electricity Generating Board (CEGB) feel that the single-walled-tube approach

is correct for commercial LMFBRs. To increase assurance of success, a

conservative steam-generator design using forced recirculation was selected for PFR, which, in comparison with once-through systems, has a

large body of experience available in the U.K.

Figure 14 shows the steam-generator evaporator, superheater,

and sodium reheater. These units, together with a conventional steam drum,

circulating pump, piping, and other appurtenances, form a forced-circulation

steam generator. The major design parameters are listed in Table 6.1*

Temperatures, transition steam conditions, dryout, etc., were all considered reasonably predictable by calculation and development within the

time available. The operational safety margins appeared much greater in

comparison with those for once-through systems, since they were affected

less by sudden plant-parameter changes and possible water contamination.

The prime design and development considerations were component simplicity and reliability of the sodium/steam-water interface,

essential ingredients to achieve high structural integrity, minimization

of water leaks, and high availability. High-quality seamless tubing subjected to stringent nondestructive testing was used in the heat exchangers.

To further ensure sodium-water interface integrity, no tube welds were

TABLE 6. Main Parameters for Heat-transport System (from Tables 10 and 11)

Thermal output, MWt 616

Electrical-generator output (gross), MWe 305

Sodium temperature at core inlet, C-F 425-797

Sodium temperature at core outlet, C-*F 585-1085

Sodium temperature at steam-generator inlet, C-F 560-1040

Sodium temperature at steam-generator outlet, C-F 374-705

Steam temperature at high- and intermediate-pressure turbine stop-valves, C-*F 537-1000

Feedwater temperature to drum, C- F 288-550

Sodium flow through core and breeder, kg/sec-106 lb/hr 2920-23.2

Sodium flow through three secondary circuits, kg/sec-106 lb/hr 2920-23.2

Steam generation rate, kg/sec-106 lb/hr 250-1.983

Steam flow to reheaters, kg/sec-106 lb/hr 171-1.361

Water-circulation rate in evaporators, kg/sec-106 lb/hr 1250-9.22

Steam pressure at main turbine stop-valve, bars-psig 159-2300

Steam pressure at intermediate-pressure cylinder inlet, bars-psig 30-435

*Superscript numbers refer to references listed at the end of this report.
Fig. 14. PFR Steam-generator Heat Exchangers

permitted under sodium. This latter criterion limited the size and geometry of the units to the use of maximum available single lengths of seamless

tubing.

The heat exchangers are all vertically mounted shell-and-tube

units using U-shaped tubes (see Fig. 14). The evaporator with 1-in.-ODtubes

uses parallel flow between the water in the tubes and sodium in the shell;

the superheater and reheater operate with counter flow. Another main difference between the components is the method of arranging the tubes--the ,v orator has a large single tube sheet and header; the steam units (reheater and superheater) have concentric annular tube sheets and headers.

35

52 -s S*GE

,.ELAEA.E2- . UE mUD SIDESOD'U DT T-EVLA2LV UUES

GAUGE HYDRGEN L EVIS GAUGES

.ELLOW. B ELLOWS

2-SODWM/WATER REACTION

S 5001 ENOZ005ZLES ON BIOS sOowMsLET 

DET--T0 SEEATE

TSEEWSREJEATE I LJ!I 1

.SDIUSLEL

tuef t

onAINNOZZLL

SUPERHEATER EVAPCTATN

REACTION NOZZLE

"|JCTM

gom - OGMOUT&T.

6*1 1.0.

SHELL

WATER 

RAIN 

NOZZLE
36

The annular tube sheets are separated by an internal thermal sleeve (expansion compensator)

- T400F

NOT TO SCALE

Fig. 15

Internal Thermal Sleeve for

PFR Superheater and Reheater

designed to cope with the 400 F AT (calculated) between the tube sheets (see Fig. 15). All the units

use trapped-gas spaces between the sodium-free

surfaces and the tube sheets. A double-valve arrangement injects or vents argon cover gas. This

system is operated only during startup to compensate for thermal expansion of the sodium.

The shell design pressure of the PFR is

500 psig. The British indicated that this design

pressure was selected by judgment, and they backfitted their analysis to this original selection. The

adequacy of the shell design to contain sodiumwater reactions has been confirmed by large waterleak tests in the Super-NOAH rig at Dounreay. The

design pressure for the steam generator.is approximately twice the design pressure of the test

vessel, which has already withstood 11 sodiumwater reactions (see Section VI.C).

One of the unique features of the heater exchangers is the pressure relief connection from the shell to the sodiumwater-reaction effluent system. The relief connection consists of a hinged

plate supporting a nickel membrane, which serves as the sodium seal; the

plate itself is held by a shear pin, designed to fail at approximately twice

the normal circuit operating pressure. The nickel membrane is supported

over most of its area, avoiding the creep problems to which a bursting disk

would be subjected at high sodium temperatures. Another advantage of the

relief-connection design is that the stressed component (the shear pin) can

be maintained (replaced, etc.) from outside the shells without disassembly

of the unit or sodium dumping. A safety-lock mechanism is used during replacement of the shear pin.

The main flange seal on the heat exchangers is a two-stage,

stainless steel asbestos seal-wound gasket. There is a leak-detection

groove between the two gaskets.

The sodium reheater is similar to the superheater, but with

different tube size, tube spacing, and shell dimensions. The reheater uses

stainless steel. The tube sheet is made of Type 321 austenitic material,

and the tubes of Type 316 austenitic material. The temperature differential between the two tube sheets is 400 F. An internal thermal sleeve (expansion compensator) connects the two tube sheets and is designed to lessen

the effect of the large AT (see Fig. 15). If the units were to be redesigned,

greater accessibility to the tube bundles would be incorporated.
37

Large transients in the reheater are associated with a turbine

trip. When the turbine trips, the reheater is taken out of service immediately, whereas the evaporator and superheater are not. This will cause

rapid transients in the reheater. The reheater steam-side transient is of

particular concern. The tube sheet may be relatively cold due to the flow

from the high-pressure turbine, while the sodium temperature is relatively

high. Thermal shock tests have not been performed on components. Reliance on the adequacy of the unit has been placed instead ontwo-dimensional,

finite-element stress analysis.

The reheater heat-transfer surface is 30% of the total steamgenerator surface and transfers 14% of the reactor heat. The overall heattransfer coefficient is 300 Btu/ftZ-OF, as contrasted to 1000 Btu/ftZoF for

the superheater. Twenty-five percent of the cost of the steam-generating

components is in the reheater. The steam reheater assumes a significant

penalty in pressure drop; each psi costs about $5,000. An important factor

in reheater design is its steam volumetric flow- -four times that of the

superheater with only one-third of the superheater sodium flow. This has

resulted in a 0.250-in, ligament between the tubes. This arrangement presents special design and development problems in the areas of tube support,

tube-sheet design, and inspection.

The selection of materials for the different heat exchangers

required the consideration of two potential problems. The first was carbon

and mass transfer on the sodium side; the second was the accelerated corrosion (by chloride concentration on the waterside) of materials commonly

used in sodium systems. The British selected 24 Cr-1 Mo stabilized with

niobium for the evaporator. Type 316 stainless steel tubes were selected

for the superheater and reheater (with tube plate and shell probably Type 321

stainless steel) since, at the time of material selection, mass-transfer data

and strength properties were not available above 475C for the low-alloy

steel. The U.K. (CEGB, TNPG, and UKAEA) is evaluating ultimate material selections for LMFBR steam generators. Particular emphasis is

placed on the use of existing technology and better operability for steam

temperatures to 495*C (925 F). There was no discussion at the meeting

concerning the British using Incoloy 800 in any of the PFR steam generators. The material composition for tubes and tube sheets is shown in

Table 7. The dimensions of tubes and tube sheets are shown in Table 8.

The tube-to-tube-sheet weld is shown in Fig. 16. The tubeto-tube-sheet joints (as well as the rest of the steam generator) had to

be designed to withstand the pressure and temperature changes associated

with an estimated 300 emergency shutdowns due to reactor faults, 100 emergency shutdowns due to turbine faults, and 1500 normal controlled plant

shutdowns and restarts.
38

TABLE 7. U.K. PFR Steam Generator: Material Composition

Chemical Composition, 

Evaporator Superheater and Reheater

Tube Tube Sheet Tube Tube Sheet

Carbon 0.06-0.10 0.12 max 0.04-0.08 0.04-0.08

Silicon 0.15-0.35 0.15-0.35 0.40-0.75 1.00 max

Manganese 0.40-0.60 0.40-0.60 1.00-2.00 2.00 max

Sulphur 0.020 max 0.040 max 0.030 max 0.030 max

Phosphorus 0.020 max 0.040 max 0.030 max 0.030 max

Nickel 0.30-0.80 - 13.00-14.00 12.00-14.00

Chromium 2.00-2.50 2.00-2.50 16.00-17.00 16.00-17.50

Molybdenum 0.90-1.10 0.90-1.10 2.00-2.60 2.00-2.60

Niobium 1.20 max - -

TABLE 8. U.K. PFR Steam Generator: Tube and Tube-sheet Dimensions

Evaporator Superheater Reheater

Tube OD, mm 25 15.8 20.7

in. 0.984 0.622 0.815

Tube-wall thickness, mm 2.3 2.0 1.3

in. 0.090 0.079 0.051

Tube-sheet diameter, mm 2230 1860 1720

in. 87.8 73.2 67.7

Tube-sheet thickness, mm 460 204 177

in. 18.1 8.0 7.0

Free from crevices, the

.4 Tube sheet weld has a smooth profile. Each

weld was automatically made with

a rotating TIG torch. Also, each

weld was inspected and radiographed. Thalium-170 sources were

used varying from 0.5-mm diam- Tube eter x 0.5 mm long to 1-mm diameter x 1 mm long. Fine-grained

Fig. 16. U.K. PFR Tube-to-Tube-sheet Butt/Fillet Joint X-ray films in plastic cassettes

produced a cone-shaped film with

an overlap. Exposure time varied from about 12 to 30 min, depending on

size and life of isotope. Each weld was helium-leak-tested. Production

welding records of current, voltage, speed, and dimensions were kept as

quality-control measures for assessing any changes, errors and other

variables. Each day's work was cleared for inspection--visual; radiography, leak testing. Less than 2% of the welds were rerun for weld faults,

and less than 7% for profile. Leak testing of about 6000 joints showed only
39

two defective joints, associated with tube-sheet inclusions. For a description of weld development, flow-distribution tests, and tube-fitting tests, see

Section V.C.

2. PFR Startup Procedures

Information was provided by TNPG regarding methods for both

"cold" and "hot" startup of PFR. Cold-startup procedures are applicable

after plant shutdown for 200 hr or more (e.g., shutdown for a turbine overhaul) with main and auxiliary systems shut down and drained. The hotstartup procedure is used when the plant has been shut down for 10 hr or

less and the turbine high-pressure casing temperature is 400C or above.

The cold-startup procedure will be described first, after which the modifications appropriate for hot startup will be discussed.

Initial conditions for cold startup include the following: primary

and secondary sodium systems charged with sodium temperature at 320-

350*C, one primary sodium pump at 30% flow with other two pumps on

standby, all secondary pumps available for operation, sodium bypass to reheaters closed, leak-detection systems in service, condenser steam space and

cooling-waterside drained, low-pressure feed train and deaerator drained,

feed pumps and high-pressure heaters isolated and drained, boiler drums

empty and blowdown valves shut, and turbine at rest with stop valves shut.

The initial step in the startup routine is to fill the low-pressure

feed system between the main condenser and the boiler feed-pump sections

with demineralized water and to clean up the system before delivering

water to the boilers. The auxiliary boiler pump is then started up on recirculation to the deaerator with the pump's main discharge valve shut.

The auxiliary boiler feed-pump valves are opened, and the drum circuits

are flushed and then filled to the specified drum level. The boiler circulating pumps on the three drum circuits are started, and the heat input

from these pumps raises drum-water temperature while the evaporators

are isolated from the drum circuits. The temperature of the feedwaterdeaerator section rises to about 180 C in 7 hr, and steam-drum temperature rises to about 350C in 22 hr. After the drum-water temperature

matches evaporator sodium temperature, the evaporator is pressurized

with steam from the top of the boiler drum and water is then admitted to

the evaporator. Circulation through the evaporator is established with

drum-evaporator circuit conditions of 1600 psi and 320C. Primary and

secondary sodium-pump flows are then established at the minimum flow

of 20%, after which the reactor is made critical. Reactor power is raised

to provide steam for pipe warming, gland sealing, and turbine runup.

While the foregoing activities are underway, the oil services for

turbine power, control systems, bearing cooling, and lubrication are brought

in, as are the generator cooling system, condenser circulating water, auxiliary steam-pipe warming system, and emergency boiler feed pump.
40

The superheaters are then brought into service through the

following sequence: (1) Set the primary-outlet temperature control at

320*C, and the dump depressurizer valve at 1100 psi (the depressurizer

setting ensures a superheat potential of approximately 30*C); (2) open the

superheater outlet valves; (3) open the inlet-valve bypasses; (4) open the

inlet valves. Pipe-warming steam is now supplied from the superheaters,

and so the superheater bypasses are shut. The final steps of the startup

operation are to (1) raise the vacuum in the condenser, (2) admit steam to

the turbine, and (3) bring the turbine to 3000 rpm for synchronization and

load acceptance. With full vacuum on the condenser, surplus "dump steam"

from the deaerator is rejected to the condenser rather than to waste.

Steam is admitted to the turbine by partially opening the throttle valves

and cracking the stop valve, allowing turbine speed to rise to 500 rpm.

Warming of the turbine is initially at this speed. Steam-temperature control and dump-valve settings are gradually increased, and the turbine speed

raised in stages to 1000, 1700, and 3000 rpm. After turbine valves are

fully opened, the generator is synchronized with the grid and loaded to

5-10%. The 5% load is required to prevent low-pressure blade flutter, and

the 10% upper limit is dictated by the capacity of the reheater bypasses.

The next stage of the startup operation is to commission the

reheaters and establish full steam conditions to the intermediate-pressure

turbine. Reheaters are brought into service by first opening the inletvalve bypasses and then the outlet and inlet valves themselves. The reheater bypasses are closed when the steam inlet temperature is within

83*C of the intermediate-pressure cylinder temperature. Steam is not

admitted to the reheater until it has a minimum 30C of superheat (for

protection against chloride stress corrosion). After full steam conditions

are achieved, the autopressure control is engaged and the load increased

toward full load in accordance with the planned startup profile.

The hot-startup sequence begins with conditions that include

primary and secondary sodium temperatures of 350*C, all primary and

secondary sodium pumps in service at 20% flow, the low-pressure feed

train and all high-pressure heaters in service, the cooling-waterside of

the condenser charged, all evaporators and drums in the circuits filled

with water at 2400 psi and 350 C, superheaters pressurized at drum pressure with outlet valves closed and bypasses open to dump, turbine bypasses

shut, reheater bypasses shut on the steam side, turbine glands sealed, and

generator gassed with cooling in service.

With these conditions, and all safety-circuit checks completed,

the reactor is made critical. Superheaters are brought into service, and

steam pressure and temperature controllers are engaged. Power is raised

by adjusting the set points. During the shutdown periods (up to 10 hr), the

steam needed for auxiliary service such as gland seal will have been generated by decay heat. After a power level of about 45 MWt is reached
41

(corresponding to a limit of 10% auxiliary boiler feed pumps), the main

boiler feed pumps are brought into service and power continues to be

raised. The superheaters are brought into service by opening the outlet

valves, and steam dump is transferred from the superheater bypasses to

the turbine bypass system by appropriate opening and closing of the isolation valves. The boiler feed pump is run up on a closed discharge valve

to full pressure and then brought into service by opening the discharge valve

while the auxiliary boiler-feed-pump discharge valve is gradually closed.

To bring the reheater into service and establish full steam conditions to

the intermediate-pressure turbine, the procedure previously described

for the cold startup is used. Turbine runup, synchronization, and loading

are accomplished in a manner similar to that followed in the cold-startup

operation.

D. SNR Steam Generators

1. Introduction

The SNR is a 300-MWe LMFBR demonstration plant under development by a Dutch-German- Belgian industrial consortium- -NERATOOM,

INTERATOM, and BelgoNucleaire, respectively.

The SNR uses a steam-reheat cycle. Each of the three secondary heat-transfer loops contains six modules: three evaporator modules

and three superheater modules. Steam reheaters are located between the

intermediate- and low-pressure sections of the turbine.

The NERATOOM (Dutch) share of the SNR program was to design and supply major components for SNR such as pumps, intermediate

heat exchangers, and steam generators. An early decision was made to

test a full-size tube segment model of the steam generator, a scaled-down

intermediate heat exchanger, and a full-size pump. The steam generator

and IHX were to be tested at the Hengelo test site, and the pump at

Bensberg, Germany.

2. Steam-generator Design

NERATOOM is developing two steam-generator concepts for

the SNR plant: a straight-tube unit and a helical-coiled-tube unit. Two

loops of the SNR will use the straight-tube units; the other loop will use

helical-coiled steam-generator modules.

The straight-tube steam-generator design that has been developed incorporates a bellows in the shell to accommodate differential

thermal expansion. Dimensions and other pertinent data on the steam generator are summarized in Table 9. Reference 2 describes the unit in

greater detail. The steam-generator prototype presently under test at
42

Hengelo is shown in Fig. 17. (Testing at Hengelo was delayed in April/

May 1972 by a sodium leak. Operation was resumed in the middle of June.)

TABLE 9. SNR Straight-tube Steam-generator Data

Feedwater inlet, *C- F 253-487

Main stream, *C-*F 500-932

bars-psig 167-2400

Evaporators

Number 9

Feedwater flow per module, kg/sec-10 6 lb/hr 69.4-0.56

Heat-transfer area per module, m2 -ft2 220.8-2377

Tube OD x wall, mm-in. 17.2 x 2-0.68 x 0.079

Overall height, m-ft 21-69

Superheaters

Number 9

Steam flow per module, kg/sec-106 lb/hr 84.5-0.67

Heat-transfer area per module, m2 -ftz 167.2-1800

Tube OD x wall, mm-in. 17.2 x 2.9-0.68 x 0.114

Overall height, m-ft 18-59

The design has a shroud around the tube bundle to enhance

sodium-flow distribution. The shroud also reduces pressure effects and

potential wastage damage on the shell resulting from a sodium-water reaction. Sodium enters and exits the tube bundle in expanded portions of the

shell at the upper and lower ends of the steam generator. The shell-side

sodium flows downward; the tube-side water/steam flow is upward. The

tubes are approximately 20 m long with tube support grids spaced at 1-m

intervals. There is no gas space in the steam-generator modules. The

tube sheets at both ends of each module are protected against thermal

shock from the main sodium stream by thermal shielding. Rupture disks

are located directly opposite the sodium inlet and outlet on each module.

The need for close coupling of rupture disks with potential leak sites was

based on sodium-water reaction investigations (both experimental and

analytical studies). The rupture-disk locations have also been influenced

by the structural limitations of the bellows in the shell. A backup ring

provides structural support for the bellows.

The modular approach is favored because it reduces the risk

of extensive damage due to tube leaks. INTERATOM feels there is insufficient evidence upon which to base a less conservative design. However,

INTERATOM would, for economic reasons, be inclined to use larger

modules if operating experience provides a sound basis for such a move.

Small leaks will be corrected by in-place repairs (although a satisfactory

procedure for accomplishing this has not been developed). For large leaks,

however, the faulted module will be replaced with a spare unit. Such a replacement would require 2-4 months; this estimated repair time would include the time necessary for replacing all rupture disks in the system and

for performing all necessary inspections.
STEAM

SODIUM 

BELLOWS IN SHELL

SODIUM ..

SHROUD

Flow stability of the evaporator

units is maintained with orifices in the

tube inlets. Earlier tests of KNK units

with steam pressures of 1200-1350 psi

indicated that flow stability could be

maintained in a low-pressure oncethrough arrangement with the use of

orifices. The higher steam pressure

of SNR, along with the orifices, should

provide a satisfactory stability margin.

The once-through arrangement

is preferred over the recirculation

steam-generator arrangement, primarily because of (1) the estimated

economic advantage and (2) the extensive satisfactory experience in Germany

with once-through units. Most conventional steam generators (fossil-fired)

in Germany are of the once-through type,

and experience has been satisfactory.

The question of corrosion and DNBdryout effects has been examined. To

investigate this question, experience of

conventional plants has been reviewed,

laboratory results have been studied,

and discussions have been held with

specialists in feedwater chemistry.

Where once-through feedwater chemistry is used, no problems have been

observed.

An oxygen level of about 50 ppb

WATERORSTEAM (up to perhaps 200 ppb maintained in

the feedwater system) helps stabilize

magnetite in the low-temperature parts

of the feedwater chain. This general

Fig. 17. Prototype of SNR Steam Generator area of concern developed as the result of erosion-corrosion effects that

had been observed in preheaters and at the evaporator inlet. Maintaining

the oxygen levels noted above, together with reduction of inlet velocities

to 2 m/sec or less, has resolved the-latter problem.

The SNR steam generator will be operated with 95% quality

steam at the inlet to the superheater under 100% load and 83% quality at

low loads. A separator will be used between the evaporator and superheater

sections (a single separator per loop). The reason for adopting this approach

43
44

is that calculations performed by the Dutch indicate that, with the original

20-30*C superheat at the evaporator exit, normal operating variations in

sodium flow would produce steam-temperature fluctuations as high as 100*C

at 25% load (for a flow perturbation of about 3%). Curves summarizing the

results of this study are presented in Fig. 18.

The material for the steam-generator module is 24 Cr-1 Mo

stabilized with 1% niobium and is considered by NERATOOM to be as

weldable as unstabilized 2 Cr-1 Mo. NERATOOM has developed techniques for welding tube-to-tube sheet joints based on internal butt welding,

which avoids crevices. The tube-to-tube-sheet joints (see Fig. 19A) are

made by inserting each tube approximately 2 mm into the tube sheet. The

position to the tube end and the internal-bore welding electrode must be

precise. The Dutch have produced a number of weld specimens in developing the adopted procedure; 300 of these tube-weld specimens have been

destructively examined to verify weld quality. Vacuum-melt forgings were

used for the tube sheet.

Brazing techniques have also been developed for tube-to-tubesheet joints (see Fig. 19B). In principle, the technique is cheaper than

welding, but if the first brazing operation is not successful, expensive repairs are required. To braze, the tubes are inserted roughly halfway into

the tube-sheet holes. The tubes use small chambers on their OD for containing the Microbraze. In contrast, for the welded joint, the tube is inserted into the tube-sheet hole only 2 mm. No filler is required for the

tube-to-tube-sheet weld. Ultrasonics were specified as the primary inspection method for the braze joint. NERATOOM stated that 0.5-mm2

defects are detectable; the acceptance criterion is based on accepting

defects of 1 mm2 . The brazing operation for the entire tube sheet requires 1 hr. The reheater for Hengelo was fabricated in this manner and

required no repairs. NERATOOM stated that although they were not

needed, repair techniques have been developed. Radiographic techniques

are used for inspecting the welded tube-to-tube-sheet joints. Based on

the advanced status of welding techniques and the advanced status of welding repairs as compared to brazing repair, the decision was made that

tube-to-tube-sheet joints in the SNR steam generator would be welded.

Sodium-water reaction tests in support of this design have

been performed at Bensberg using a full-scale-tube, 6-m-long segmental

model. The model has been subjected to series of single-tube failure tests

(see Section VI.D).

NERATOOM considers the straight-tube design extrapolable to

larger units of approximately 250 MWt; however, doubt was expressed as

to the extrapolation of the shell bellows to much larger sizes.

NERATOOM is presently developing an alternative steamgenerator design for SNR with helical tubes and no tube sheets. The
500

W 450

W

I 400

350

3001 1 I I IJ

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

100% %LOAD

TB

T8

TS

-T7

-T6

75/ %LOAD

Tg

500 T

T8

450 

T7

400 T6

350

T4

I I I

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

50% LOAD

TT

Ts

T6

T7

T4

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

30% LOAD

T9

500

450 -6

400 

350

T4 300

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

FLOW RATIO: SODIUM TO WATER/STEAM

25% LOAD

T

TT

450 6

TT7

400-

350

300

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

SOC

450

S400-

w

3500

3

500

450

4 400

a

ac~R

T4

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

300 T4

0.95 0.975 1.0 1.025 1.05

SODIUM FLOW

LEGEND

T2 - SUPERHEATER/REHEATER SODIUM

INLET

T4 - EVAPORATOR SODIUM OUTLET

T5 - EVAPORATOR INLET TEMPERATURE,

SODIUM SIDE

T6 - EVAPORATOR OUTLET TEMPERATURE,

STEAM SIDE

T7 - REHEATER SODIUM OUTLET

T8 - SUPERHEATER OUTLET TEMPERATURE,

SODIUM SIDE

T9 - SUPERHEATER OUTLET TEMPERATURE, STEAM SIDE

T10 - REHEATER STEAM OUTLET

Tl - REHEATER STEAM INLET

T12 - BOILER-FEED WATER

NOTE:

SODIUM FLOWS: 1.0 - NORMAL SODIUM

FLOW

FLOW RATIO: SODIUM TO WATER/STEAM

Fig. 18. Plots of Sodium Flow for SNR Steam Generator

U,

500

450

400

350

20% LOAD

TT

TT

- T6

TT

v

oc

D

a

ac

W

d

2

W

H

14% LOAD

TT g

T6

-T
TO X-RAY MACHINE Il

SOURCE

/ TUBESI

OLDER TUBE

A. WELDED JOINT

HEET

ULTRASONIC PROBE

I I

TUBE HEETNICROBRAZE

B. NICROBRAZE JOINT

Fig. 19. NERATOOM Tube-to-Tube-sheet

Connections Developed for SNR

FILMFILM HOL

Steam Generator Studies and test programs on the

helical-coil design have paralleled those

for the straight-tube design. Flow investigations are being made to determine vibration of the helical coils (which are more flexible than straight

tubes) and sodium-side pressure-drop data, for which limited information

is presently available.

The Hengelo model will subsequently be used as the sodiumwater reaction model for the Bensberg sodium-water reaction tests. Fabrication areas specific to this design that may cause some concern are

(1) the bending of the tubes and (2) the preparation and connection of the

tube end pieces needed to connect the helical coils to the head. NERATOOM,

however, does not consider the tube-to-head joints for the non-tube-sheet

design to be a problem. Although there is confidence in existing technology, the technique would require some proof testing on a prototypic single

tube-to-head connection.

Fretting tests between tubes and tube supports have been completed by TNO at Apeldoorn, Netherlands, and have led to finalizing clearances between tubes and tube supports. To avoid self-welding of 24 Cr-1 Mo

material, a 0.4-mm diametrical clearance has been determined as a good

compromise between self-welding, vibration, and fretting considerations.

NERATOOM considers the helical-tube design to be capable of

extrapolation to greater sizes that the straight-tube design, which, as

stated earlier, may be limited to about 250 MWt. In addition to the shell

46

individual tubes penetrate the shell of the

steam generator at the bottom and top

heads. One loop of the SNR will have three

evaporator and three superheater modules

of this design; the other two loops will use

steam-generator modules of the straighttube configuration. As with the straighttube unit, the helical-tube unit has no gas

space. Further, it has no bellows in the

shell. The tubes are of larger diameter.

Small sodium flows are introduced adjacent to the tube sheets to aid in maintaining

circulation and to facilitate leak detection

in this area. Injection fittings at the upper

end of the helical-tube unit will be used for

the introduction of either steam or hydrogen

for calibrating hydrogen-leak detectors. A

50-MWt evaporator prototype of this helical

design is being fabricated and will be tested

at the Hengelo facility in 1973-1974.
47

bellows, other limitations of the straight-tube design are: (1) the difficulty

in maintaining tube-sheet quality in larger sizes and (2) the limited capacity

of the vacuum-melt furnaces. Similarly, the heat treatment of the tube-totube-sheet joints is time-consuming; one must preheat the tube sheet to

make a ring of welds, then cool to inspect, etc.

NERATOOM is also considering a single-shell design for both

the superheater and evaporator. This is particularly suitable to the helicalcoil design, which does not require a long shell. One disadvantage of such

a design is the single-tube wall thickness and tube diameter for both the

evaporator and superheater. The helical-coil design of this type requires

tube-to-tube welds to obtain the required tube lengths. The present design

is based on 20-m (about 66 ft) sections of tube lengths, which are transportable without a serious problem (35-m sections have been considered).

Although 60-m-long (about 198 ft) tubes are available, they are difficult to

transport. NERATOOM indicated that the helical tubes were almost equal

in length, despite the different coil diameters. The maximum difference

between tube lengths is 2 ft.

The relative cost between the straight-tube and helical-coil

designs was not known for large sizes, but based on their experience with

the Hengelo models, NERATOOM considers the fabrication costs of the

helical-tube design to be more favorable than the straight-tube design in

large sizes. Individual tubes can be plugged in both designs, although the

techniques are different. The helical-tube design has tubes penetrating the

shell and terminating in headers on the outside. Plugging would be performed at the header. Ultrasonic transducers can be passed through the

tubes of either unit for inspection; however, the coiled-tube unit has better

access for initial insertion of an ultrasonic probe.

The helical-coil design uses a 5-mm-thick shroud surrounding

the tube bundle. Effects of pinhole-type tube leaks are considered to be

minimized at the shell by this shroud. A 30-mm sodium annulus between

the shroud and the shell also serves as a buffer against pinhole-leak effects.

The current Bensberg sodium-water reaction test model does not contain

a shroud because of interference with the required test instrumentation.

NERATOOM has begun a study on a double- and/or triplewalled-tube steam generator. The ultimate objective of this study is to

increase the integrity of a steam generator to a point where the secondary

sodium circuit could be eliminated. The steam generator would be inside

the primary containment, and the design would be such as to virtually prohibit a sodium-water reaction. Both helical- and straight-tube designs will

be evaluated. NERATOOM does not believe a single-wall tube will meet

the needs of secondary-circuit elimination. Long-term studies in the U.K.

are underway to determine if a single-wall-tube steam generator can be

designed in a manner to eliminate the secondary sodium system. The
48

multitube design would attempt to provide a leak-detection capability between the concentric tubes. Studies will initially concentrate on fabrication

techniques.

3. Steam-generator Tube Leak and Rupture-disk Considerations

INTERATOM does not provide a gas space over the sodium in

the SNR steam-generator modules. Cover gas is maintained in the pump

and in the expansion (surge) tank of the secondary loop. Experiments and

analytical studies have indicated that the presence of a cover-gas space

can result in very high pressures due to sodium hammer effects if sodiumwater reactions occur in the steam generator. The gas space contributes

to a delay in depressurization of the system through a delay in bursting of

a rupture disk and results in higher reaction-zone temperatures and transmission of stronger pressure pulses to adjoining modules, the secondary

system, and the IHX. Without a cover-gas space, rapid depressurization

through the relief system occurs in the event of a tube leak, and the design

can therefore be based on the guillotine failure of a single tube. Studies

did indicate that the depressurization-pressure pulsation problem might

be alleviated with proper design of the gas space. However, the control of

level and pressure in the gas space may present some difficulties.

INTERATOM considered the difficulties manageable within the current

state of control technology.

The initial pressure pulse that results from a guillotine tube

rupture dissipates within less than 10 msec and is of such little energy

that the steam-generator structure is not affected by the burst. The secondary pressure peak is of longer duration and must be treated as a static

load in the structural analysis. The bubble growth occurs at such a rate

that a second tube failure would have to occur within 30 msec to have any

significant effect on the pressure transient.

If a large leak occurs, INTERATOM proposes that the faulted

steam-generator module be replaced by a spare unit. Reaction products

would have to be removed from the loop. Removal procedures have not

yet been developed. In the event of small leaks, the unit will be repaired

in place. At present, there is no way of accurately evaluating the extent

of damage and required repairs. There is concern that small leaks may

plug themselves and become extremely difficult to detect after cooldown.

Such an experience occurred in the 5-MWt steam-generator test facility.

Following a tube leak, the water system would be blown down

while sodium remains in the system for a period of time. Sodium would

not be dumped immediately after detection of the leak, because of concern

over a possible secondary sodium-water reaction in the dump tank. The

cyclone separator has a capacity of about 6 m3 of sodium, which is greater

than the volume of one module. Test results, as well as other analytical
49

studies and laboratory results, indicate that no caustic corrosion should be

expected as a result of the large leak. The steam generator should empty

in about 40 msec. Only small amounts of reaction products are expected

to be transported to adjacent units. Maximum temperatures observed in

the reaction zone to date have been about 100 0 C. The rupture relief lines

are 250 mm in diameter, and lengths to the cyclone separator are about

20 m for the line at the upper end of the modules and 40 m for the line at

the lower end.

Calculations indicated that with cover-gas space in the steam

generator, large leaks cause pressures of 140 atm within the hydrogen

bubble. However, the sodium hammer effect in the presence of the cover

gas can result in a pressure of up to 500 atm. Tests have indicated the

bellows can withstand about 150 atm without failing. (For more detail,

see Section VI.D.)

Statistics on tube-failure rates (the sources of these statistics

were vague) have indicated that the most pessimistic failure rate that can

be reasonably assumed for small leaks in a steam generator is one tube

failure per year. INTERATOM assumes a gross failure rate of one to

two tubes for the life of the plant.

The sodium operating pressure is 8 atm, and the rupture-disk

burst pressure is 17 atm. The waterside blowdown period is estimated

at 30-40 sec following automatic shutdown in response to disk-burst indication. The rupture disks are 150 mm in diameter and 1.1 mm thick. The

estimated life of the rupture disk is approximately 6 yr, taking account of

corrosion and age-hardening effects. The disks are in the shape of a sector of a sphere with the convex side facing the sodium. Analytical studies

indicate that for disks of the present size and larger, the peak pressure is

about the same. The present rupture-disk diameter and collector-pipe

dimensions have been adopted as a tradeoff between peak pressures and

loadings on the relief piping. A number of tests have indicated that the

rupture disk fails consistently around the electron-beam seam weld that

joins the disk to its ring. Rupture disks of this type have now accumulated

17,000 hr under normal operating conditions in KNK. The disk material

is niobium-stabilized stainless steel. The rupture disks, after breaking

cleanly, pass down the relief pipe and are found in the cyclone separator.

In the initial arrangement, a single rupture disk was located in

each of the sodium headers--i.e., the sodium-inlet header to the superheaters and reheaters, the sodium-outlet header to the superheaters and

reheaters, the sodium-outlet header from the superheater and reheater

modules, and the evaporator-discharge header. Experimental studies, together with analytical studies, indicate that with the reference arrangement,

pressure oscillations occurred that would cause bellows failure. Bringing
50

the rupture disks closer to the tube-failure site (two rupture disks in series

at each end of a module unit) could eliminate the pressure oscillations and

thereby protect the bellows.

4. Sodium-reheat Component and Systems Considerations

INTERATOM considers that sodium reheat has a "slight" technical disadvantage, compared with plants using the other cycles. The

primary reasons contributing to this view are: (1) Startup of a sodiumreheat plant is more complicated than plants using other cycles, and (2) the

extra sodium heat-transfer area in the steam generator increases the risks

of tube leaks and resultant possible sodium-water reactions. They have

analyzed the reheater component for all design transients, including hotshock conditions. They consider the reheater to present a less difficult

design problem than the evaporator. The stresses resulting from the design transients considered are acceptable, and, although the dimensions

differ, the conceptual layout of the reheater is similar to that of the evaporator and superheater. In establishing design transients, they assumed

that the secondary-system pump coastdown to 5% flow required 50 sec.

5. Miscellaneous

INTERATOM indicated that, for commercial LMFBRs, they

would consider the use of four loops for reasons of safety and plant

arrangement. Such an approach would permit them to operate with three

loops while the fourth loop was down for repairs.

The occurrence of a fire at KNK in spring of 1971 was briefly

discussed. An electrical submerged-type tubular heater element in a

region of flow stagnation experienced a short that resulted in a meltthrough of the shell wall of the heater and loss of about 500-600 kg of

sodium. The result was a sizable fire.
51

V. DEVELOPMENT PROGRAMS FOR STEAM-GENERATION SYSTEMS

AND RELATED COMPONENTS

A. Summary and Comparison

All the countries visited were able to show an in-depth program of

engineering, development, and testing directed toward their demonstrationplant steam generators. In many areas, development work for their demonstration plants had been completed, and ongoing work is being directed

toward their commercial plants.

Each European program had focused its early development work

toward a single reference design. At present, the French are directing

their current work toward two new designs for the commercial LMFBR,

and the Dutch-German program has established an alternative design,

which will be used in one loop of the SNR. The British are also looking at

alternative steam-generator design concepts for the CFR.

The French and Dutch-German programs have taken the approach

of performing proof tests of large-scale models. The British did not perform a steam-generator model test of the PFR steam generator. For the

CFR, the British plan a steam-generator model test, if the steam-generator

concept is different from that of the PFR. This test may be in an existing

European facility or in a new facility to be built in the U.K.

The French program and the SNR Consortium program showed a

coordinated program of testing in support of many areas of steam-generation

system design. In particular, extensive work had been performed on rupture

disks and valves. The existence of an in-depth development program supporting the PFR design was made evident in the discussions. This section

reports on the information made available in support of the PFR, SNR, and

Phenix demonstration plants.

B. France

1. Phenix Steam-generator System Development

Development programs supporting the design of the Phenix

steam generators and the Phenix plant systems included a 5000-hr test at

the Grand Quevilly 5-MW test facility, and full-scale module testing at

Les Renardieres for over 6000 hr.

The CGVS has a capacity of 50 MWt. The Phenix steam generator (CGVS) first tested at this facility was made of unstabilized 2 Cr-1 Mo.

This unit, after completion of test, was removed from the building and was

located outside in the open atmosphere. A Phenix steam-generator module,

using stabilized 2* Cr-1 Mo plus niobium, was on test.
52

Steam-generator test instrumentation includes orifices used to

measure flow and also to provide hydraulic stability. Thermocouples are

placed at the inlet of each tube and on the outside of the shell. There are

no thermocouples in sodium.

The Les Renardieres facility flow is controlled by valves,

whereas in the Phenix plant, flow is controlled by the pump. Maximum

size of valves being used by the French is 300 mm. These valves will be

tested at Les Renardieres, where valves up to 500 mm can be tested.

EDF indicated that the Les Renardieres steam-generator facility

tests Phenix modules, simulating conditions as close as possible, including

transients such as turbine trip. However, not every transient would or

could be simulated because of possible damage to the sodium heaters. EDF

stated that Phenix could handle the turbine trip. Rapid- shutdown transients

have been conducted to ensure this capability. Turbine trip followed by

water and steam-flow stoppage has not been simulated in the Phenix test.

This was planned as the last test phase at Les Renardieres for the summer

of 1972.

Tests completed to date in the CGVS include (1) steady-state

heat-transfer performance, (2) flow stability, (3) dynamic-response tests

(e.g., response to ramp or step changes in flow), (4) simulated plant

startup, (5) simulated plant shutdown, and (6) simulation of various plant

normal, upset, and emergency operating transients. Preparations were

underway for the final emergency transients, such as loss of secondary

sodium pump. Tests in which plant operating conditions have been simulated

have essentially duplicated the planned Phenix conditions, except for the

turbine-trip case. In Phenix, turbine trip results in reheat-steam-flow

stoppage; in CGVS, reheat steam flow continues because of present limitations in test-loop flexibility.

The French are satisfied with test results to date. There is

evidence of good mechanical integrity of the Phenix steam generators.

Heat-transfer performance has been essentially as predicted. Oncethrough, steam-generator flow stability has been demonstrated (with evaporator tubes having inlet orifices) down to pressures of 90 bars. On the

basis of testing to date, EDF believes that plant operations with sodium

reheat can be carried out in a satisfactory manner.

The Phenix test was scheduled to end in July 1972; BabcockAtlantique was to have a 45-MWt unit on the site by November 1972 and

begin testing by February 1973.

The French consider that the main results to be derived from

superheater and reheater component testing are in the area of metallurgical

and mechanical effects. Heat-transfer performance and flow-distribution
53

effects can be predicted with acceptable accuracy, and testing of an evaporator together with analytical modeling can provide needed control

information.

The French position regarding required sizes of test units is

that an assembly of units totaling 45 MWt will provide the necessary confidence that full-size units will perform satisfactorily in a commercial

plant. Test units of this size will have essentially the same characteristics

as plant units in the areas of steady-state heat transfer, flow stability and

distribution, metallurgical effect, and (to a large extent, depending upon

design) mechanical behavior. With the test units having tubes of the same

dimensions as the commercial units, the dynamic performance of the models

would be similar to that of the plant units. The differences in shell time

constant due to differences in wall thickness should be a second-order

effect, which can be accounted for in analyzing test results for plantperformance predictions.

The French have a mixing-tee arrangement, which can be

either single- or double-stage. The design has a divided-honeycomb

arrangement parallel to the pipe flow. One of the flows to be mixed enters

perpendicularly to the mixing chamber; the other flows through the pipe

and the honeycomb. This type of mixing tee has been tested with 200 C

differential temperatures in the fluids and has performed satisfactorily.

The French are using BSB membrane-type rupture disks,

purchased in the United States. A major factor in the life of the rupture

disk is the fact that it must not operate near the yield point if a long life

is expected. The French situate the rupture disk approximately 1 m away

from the flow stream.

They have tested a number of these disks and found considerable

scatter in the test data; they plan to change the membrane in the plant when

they have performed further tests. They have two new types of rupture

disks under consideration: the BSB modified, which is a welded arrangement, and a device of their own, which is welded in a special support with

a cutter at the back. They feel that this membrane can be calibrated better.

In the 1000-hr tests on these disks, different types of tests are performed,

including thermal and pressure cycling. The disks are about 200 mm in

diameter and are made of Type 316 stainless steel. They believe that, of

the BSB disks, the RB 90 is the best. The test arrangement being used by

the French contains two disk assemblies and three chambers containing

gas and sodium.

2. Steam Generator under Development for 1000-MW Plant

Two French companies- - Babcock-Atlantique and Stein

Industrie--have joined CEA and EDF in a common program.
54

a. Babcock-Atlantique

The tube bundles are helical. This generator, 290 MWt (if

two units per loop) or 580 MWt (if one unit per loop), is most suited to a

steam cycle without sodium reheat. The main stage of this development

is the construction of a 45-MWt mockup for endurance testing in 1973 at

the Les Renardieres EDF steam-generator test facility. This model will

consist of a 24-tube bundle forming three rows between two concentric

shells, with a 1-m external diameter. It is fed by hot sodium at 550 C

(1022 F) and water at 2400C (554 F) and produces steam at 167 bars

(2455 psig) and 510 C (950 F). The tubes are made of Incoloy 800, the

water and steam connections of 2.25 Cr-1 Mo. The model, after test at

Les Renardieres, will be sent to Cadarache for a sodium-water reaction

test.

b. Stein Industrie

This steam generator is a counterflow module having a

bundle of straight tubes with a Z-type expansion bend in the gas space

below the tophead. The steam generator consists of separate evaporator

and superheater units. Such a module has 580-MWt power, or four units

for a 1000-MWe plant. It is proposed to fabricate the steam generator with

a stabilized ferritic steel (2.25 Cr-1 Mo-0.4 Nb-0.4 V).

A 45-MWt model will be tested at Les Renardieres in 1974.

This model will have a 44-tube bundle. After the Les Renardieres test, the

model will be shipped to Cadarache for sodium-water reaction tests.

3. Valve s

The French CEA and EDF are cooperating in the development

of LMFBR valves. The valves being developed include (a) shutoff valves,

(b) flow-regulation valves, and (c) trimming valves. The design requirements for a number of these valves are dictated by the steam-generator

system design. A specific example is the large secondary-system sodiumwater isolation valve. This unique valve is manufactured by Neyrpic. The

valve operates by the rotation of the main assembly through 900 and a

downward motion of 30 mm. The assembly includes a toggle arrangement,

which closes the valve. This valve is shown in Fig. 20, and the operating

mechanism is outlined in Fig. 21. The valve is located in the 494-mm-ID

secondary-system line, and the valve has an internal-port diameter of

380 mm.

Valve-testing facilities located at Les Renardieres include a

valve loop and a sliding friction testing machine. These facilities and

current valve-development work are described in Ref. 3.
55

Fig. 20

Phenix Secondary-system Isolation Valve

Fig. 21

Assembly Arrangement of Phenix

Secondary-system Isolation Valve

i
56

C. United Kingdom

The British steam-generator development problems requiring work

are outlined in Ref. 1. Though an outline of the British development programs that have been or are being conducted in support of the PFR Steam

Generator design was not presented, the following information was obtained

in response to specific questions.

Hydraulic-stability analysis and test were not required for the superheater or reheater, but the fabricator has performed hydraulic-stability

analyses and tests for the evaporator, using three tubes in parallel. These

tests were done without orifices. Instabilities were present only if dryout

was achieved in one tube, where dryout is defined as the point inside the

tube at which the tube is no longer wetted. Though confident that they may

not be needed, orifices have been installed in the evaporator tubes with a

pressure drop equal to the tube-side pressure drop without the orifices.

This approximately doubles the steam-side pressure drop.

The stability tests were electrically heated. Many dryout tests

have been performed. The margin of 1.6* is satisfactory, since it represents the condition that would exist with heavily fouled tubes and one completely cleaned tube.

Magnetite is an iron oxide layer (FezO3 ) formed inside the tube in

the absence of oxygen. A design goal was to avoid cracks in the magnetic

layer. The tests indicated a dryout of approximately 20C. This is not

considered sufficient to crack the magnetite layer. However, tests will be

performed with various water qualities to see what water-quality range

can be used for off-normal operating conditions. A problem is the unknown

corrosive mechanisms that may exist at the dryout point, even though the

AT is relatively small.

There is a 10/ continuous blowdown from the steam drum. This results in a 1001 concentration of solids in the recirculation system. The

system is dosed and has a pH of 9-9.5. If blowdown is required after a

sodium reaction, blowdown will be done from both the drum and the evaporator. With this arrangement, safety valves are needed on the evaporator

as well as on the steam drum. Such safety valves have been provided in

the PFR.

A significant amount of weld development has been performed in

parallel with the design and fabrication of the plant units. Initially, many

problems were encountered with minute cracks in the tailoff of the tubeto-tube-sheet welds. This problem was resolved by connecting the welding

torch to a computer, which programs the runoff so that the cracks are

minimized. Finally, a small dental-type drill is used in any suspect area

*Margin of 1.6 presumably means the margin of recirculation available.
57

for cleanup and final finish. Cracks observed were 10 mils long and 2 mils

deep. The cracks are probably due to chromium in the tube. The chromium

carbide balance must be correct. Also, the commercial Type 316 stainless

steel specification has too wide a range for chromium and ferrite. TNPG

establishes its own specification.

Research and development for the tube-to-tube-sheet welds was

done over a 2-year period. The tube holes were bored and welds made

(see Fig. 16). The degree of penetration into the tube sheet was very

important and had to be controlled to 2 mils. Before being welded, the

tube was lightly rolled into the tube sheet to ensure good contact and uniform current. The initial weld-development work was on single tubes, then

on seven-tube arrays using full-scale U-tubes, and finally on weld sections

containing 100 full-scale U-tubes. This last series essentially meant fabricating partial tube bundles.

A special "mouse" has been developed for measuring wall thickness

in the PFR steam generators. This is either an ultrasonic or eddy-current

mouse, which measures wall thickness to within 10 mils. This probe will

not go through the inside two rows of the units, because the bends are too

tight. Development work is being done by instrument developers in Germany

to permit the mouse to go through the inner two rows. The mouse is

greased, and the grease helps to maintain the contact during passage of the

mouse through the tubes.

The British have worked on nondestructive test (NDT) development

at Dounreay in parallel with the large sodium-water-reaction test work in

Super-NOAH.

The work done on flow visualization for the PFR steam generators

was with plastic models of 5/16 scale. A water-flow test of one of the

secondary loops will provide further information.

Fretting work has been done in 6000C (1112F) sodium containing

10 ppm oxygen. Separate high-amplitude sliding and tapping tests were

performed for the reheater and superheater; one simulated the expansion

and contraction due to thermal effects, and the other simulated vibration

effects. No damage of concern was found in these tests. The tests will be

reported after information is obtained from the secondary-loop water-flow

tests.

Proposals are being negotiated for a U.K. facility for heat-exchanger

testing. Also under consideration is the use of available European facilities

such as Hengelo. Economics of a test facility dictate a facility much

smaller than 200 MWt. A rig is being considered to handle full- size and

full-length tubes, but only a segment of the unit. The British consider it

important to pay careful attention to adequately instrumenting the test

program.
58

The British expressed a need for a steam-generator systems test.

For a nonreheat system, this would involve an evaporator and a superheater with associated piping and valves. For a sodium-reheat system,

the evaporator and superheater could be tested together and the reheater

tested separately, simulating other system inputs obtained from the

evaporator-superheater test. Such tests would be systems tests and

instrumentation and control tests, rather than proof tests of prototype

models.

The U.K. has abandoned the PFR primary-system hydrodynamic

water test. They intend to perform a water-flow test of one of the secondary

loops, which may have been done by the end of August 1972. The information

to be obtained from this water-flow test would include vibration, acceleration, level changes, gas entrainment, and pump performance. The British

feel they have assurance that gas entrainment will not occur in the primary

circuit as a result of model testing they have performed.

D. Germany- -INTERATOM

1. Development of SNR Steam Generator

The Dutch are taking the lead in the development of SNR steam

generators. The major steam-generator-related development at

INTERATOM is on sodium-water reactions (this work is reported in

Section VI.D).

INTERATOM has a 5-MWt steam-generator test facility at

Bensberg. Results were satisfactory, but larger-scale tests were necessary

to obtain extrapolable results. Stability tests were performed in the 5-MWt

model in which sodium flow was varied while the water flow was kept

constant.

INTERATOM has also performed steam-generator water-flow

tests at one-third scale. These tests were for water-distribution and

vibration effects. Particular emphasis on the tests was on the inlet region.

2. Valves

A special butterfly valve for use in sodium has been developed

for isolating the SNR steam generators. This valve has been tested in

600- and 350-mm sizes. The valves are of an unconventional offcenter

design with a sloped seat and nearly no friction at the seat. The valves

used for isolating the steam generators can be operated in 10 sec. Normal

flow velocity of the sodium is about 6 m/sec.

The freeze seal on the valve stem has been satisfactorily

tested. The freeze seal is backed up by a stuffing box. Leakage of the

350-mm valve has been measured at 1 m3/hr at a AP of 1 atm. The
59

isolation valves prevent sodium-water reaction products from entering the

intermediate heat exchanger. These valves are not intended to isolate the

system for maintenance purposes. Valve closure results in a sodium

hammer of only 4-5 bars.

A 350-mm isolation valve has been tested under thermal-shock

conditions, 4-5'C/sec over 100 C with no observable malfunction of the

valve. This valve has had a test life of about 5000 hr under various conditions. The valve has been operated under cavitation conditions at 15 m/sec,

and erosion and pits have been observed on the spindle and behind the disk.

The valve continued to function, and there was no damage to the bearing

around the valve stem. Cavitation was the result of low pressure during

pump test. Relocating the valve to a high-pressure region has avoided

further cavitation.

The 600-mm valve has been tested for 1500 hr. It has not been

subjected to thermal-shock tests.

The 250-mm sodium-flow-trim valve (nondrainable) for SNR is

a single-blade gate valve. Tests on a 200-mm trim vlave for KNK were

conducted in Karlsruhe. Based on these tests, further tests for SNR trim

valves were not needed.

The SNR primary valves are a Neyrpic French design. A

600-mm valve of this type was scheduled to be tested in July 1972. A

450-mm valve has been tested with a 5-liter/hr leak rate. This is a splitball rotating valve. In the open position, the ball is rotated to where there

is essentially unrestricted passage through the valve. To seat the valve,

it is rotated 900 and the shutoff disks are cammed into position on the valve

seats by vertical motion of the valve stem, which operates a series of

mechanical linkages attached to the disks. (See Figs. 20 and 21.)

E. Netherlands

1. Tests on SNR Steam-generator Thermal Model at Hengelo

At present, the straight-tube steam-generator design for the

SNR and the reference IHX design for SNR are being tested in the Hengelo

Steam Generator Test Facility. The steam generator consists of a straighttube evaporator, superheater, and reheater. The superheater and reheater

are in parallel, both in series with the evaporator. The reheater is no

longer included in the SNR reference steam generator. Approximately

3 weeks of steaming operation had been attained with the facility at the

time of the team visit, and a thermal power of up to 25 MWt had been

achieved.

The overall performance of the steam-generator model up to

the 25% load condition, reached for a short period of time, is close to
60

predictions. Some areas had lower heat transfer; others had higher. A

detailed analysis was yet to be made. The nucleate boiling region of the

evaporator heat transfer was better than predicted; the mist flow and

transition region of the evaporator yielded poorer heat transfer than

predicted.

The superheater and reheater performances were in good

agreement with prediction. The differences in heat transfer in the evaporator agreed with the qualitative results obtained from the Apeldoorn

single-tube SWISH experiments. No instabilities were noted in the units.

However, instability measurements were not made at very low load. No

orifices are included in the tube of the units. At 25 MWt, the superheaterinlet steam quality is 907.

There have been no reheater problems during startup. Proper

temperatures are maintained, since the steam temperature coming out of

the reheater and superheater is very close to the sodium temperature under

all loads. A steam bypass is used similar to the bypass arrangement used

for large fossil plants during startup and up to approximately 30% load.

The Hengelo steam-generator tests will (1) confirm the heat

transfer, (2) detect any possibility of instability on the waterside, (3) collect

detailed results of the effectiveness of the thermal baffles in protecting the

tube sheets (about 100 thermocouples were installed in the Hengelo model

for this purpose), and (4) monitor the shell bellows to confirm that the

displacements are as predicted.

Steam-generator tests that have been performed include fullscale water-flow distribution tests at Apeldoorn to obtain information on

radial-flow patterns and temperature distributions. These tests also provided detailed information to assist in the stress analysis of the units. The

Dutch considered an important factor in obtaining a good distribution is to

have the tolerances and s tr aightne s s e s of the tubes maintained to fine

tolerances. One area in which they have done a lot of work is in the

thermal baffles that protect the tube sheet and the shell. These are multilayered labyrinth-type baffles and have been tested to ensure they meet

required performance.

The Dutch believe they have no reliable data on sodium-side

pressure drop. They indicated that the testing program they planned for

the 50-MWt helical-coil evaporator was part of a development program

for the helical-coil unit, which was similar to that which had been performed

for the straight-tube unit.

The Dutch cannot test a large superheater, due to the sodium

flow required for the superheater test and the lack of a suitable steam

source onsite. Although the future test will be of an evaporator alone, some
61

of the NERATOOM staff feel that system tests are still important. People

at Hengelo stated that there is a trade-off in benefits between component

size and system information to be obtained.

The evaporator, in comparison with the other units, has major

problems associated with fabrication, heat transfer, and stability. The test

of as large a unit as possible is required to prove the evaporator capability

to the maximum, including the adequacy of fabrication. If tests of other

units were desirable, it would be sufficient to test the superheater, if it is

of the same design as the reheater.

Testing philosophy was discussed. The Dutch stated that they

would feel it necessary to perform an overall systems test first. With an

initial systems test and with confirmatory plant operation, future tests

need only be for large components rather than systems, unless there was

a major change in the system arrangement. The Dutch stated that 50-MWt

tests could be extrapolated to commercial-size units, and there are no

plans to test a larger unit than could be tested at Hengelo. If a new type

unit is to be designed, their approach would be to test it as a component at

Hengelo at the 50-MWt level and then extrapolate this unit to larger sizes.

Many minor problems have been experienced during the heatup

of the steam generator, and the Dutch were interested in the demands of a

reactor system with respect to startup of the steam generator. Problems

experienced have been with the evaporator when steam condenses in the

unit and some water drips to the lower sections. This can cause thermal

shocks and large differences in temperatures. Also, problems have

occurred in obtaining exactly the right evaporator temperature before

introduction of feedwater.

It was initially difficult to obtain the proper AT's on the tube

sheets, and it now takes approximately one-half day to preheat the steam

generator from a cold condition to a minimum power condition. It is easier

on a hot start to go down to 2000C and restart very quickly.

The Dutch were interested in knowing if it is worthwhile to

either artificially foul the waterside of the steam generator or conduct

very long-duration tests at steady state to determine fouling. Such tests

are being considered.

Cold thermal shocks are planned on the present units, about

110 C in 20 sec. Actual operation provides the most severe thermal shocks,

such as those ensuing after turbine trip and reactor scram. These shocks

have been unintentional and include a transient induced by steam safetyvalve operation, which occurred during the sodium fire. The maximum

shock that can be imposed at the steam generators intentionally is 5C/sec.
62

No vibration testing or measurements are being performed in

the steam-generator test. Vibration and flow- distribution measurements

were made in models tested in water at Apeldoorn. Only preheat-startup

and steady-state tests up to 25 MWt have been conducted on the present

units.

The hydrides and free-hydrogen concentration in the system

change with temperature. Thus, the Dutch are concerned at Hengelo that

the leak-detection system may have difficulty in detecting the hydrogen

due to a sodium-water reaction. The method proposed is to measure the

inlet and outlet conditions of the unit. Under consideration is the control

of hydrogen level in the system by flushing out the argon cover gas from

the expansion-tank cover space. Plans are to install a hydrogen detector

in the gas space in the expansion tank, to measure the hydrogen concentration in the system, flush out the argon and associated hydrogen, put clean

argon in, and allow the gas space to fill up with free hydrogen from the

system. There is some question as to the possibility of such a system

providing control of the loop hydrogen. (Note: The suggestion of having a

hydrogen sink in the secondary system was previously advanced by

Gene Hill of Varian. This was not discussed at the meeting.)

The hydrogen level at Hengelo is 0.5-1.0 part per million (ppm)

ranging up to 2.0 ppm. The Dutch intend to control the hydrogen level in

the sodium by flushing the gas space to a level of approximately 0.1 ppm.

The purging rate and other factors were yet to be determined, as well as

the sensitivity of the hydrogen meter.

2. Steam-generator Development Test at Apeldoorn (TNO)

The SWISH loop, originally used for supercritical water tests,

is now being used for subcritical heat-transfer studies simulating an SNR

10-m-long, 8-mm- inner-diameter, 2-mm-thick-wall steam-generator

tube. For reasons of strength the SWISH loop steam-generator tube is

made of Type 347 stainless steel, so that it is capable of 300 bars and

550 C on the steam side. Sodium surrounds the steam-generator tube,

which is filled with water. The TNO Metallurgical Institute specified the

Type 347 stainless steel for supercritical pressures. The loop is operated

at 6000C; it is a 0.5-MWt loop with 5-m3/hr capability and is used for heattransfer studies. The electrical heat source for the loop consists of

100 5-kWt heaters. The loop is divided for instrumentation into four

1-m-long sections and three 2-m-long sections. The subcritical tube is

operated from 40 to 190 bars. Flow, power, and subcooling can be changed.

Low flow conditions are simulated by a bypass arrangement. Temperatures

are measured at the inlet and outlet of the 10-m-long tube, as well as at

many points in the upper three 2-m-long sections. At these points, temperatures are measured on the inside and outside of the wall as well as of the

water. The thermocouple connections to the steam-generator tube are
63

shown in Fig. 22.

leg not insulated.

WATERr

The thermocouples are 20 mils in diameter with the hot

There are 100 nonreplaceable thermocouples, spaced at

5-cm intervals in each of the top

/STEAM-GENERATOR TUBE three 2-m sections of the tube.

-0.080" TEST ENVELOPE

- 0.070

0S020

SO'IUM

7

V

TEST ENVELOPE

ATR SODIUM 

Tc

ST EAM-GENERATOR TUBE

Tube fabrication alone (without design) cost about $10,000 and

took 9 months to complete. Stability

phenomena and other phenomena

associated with the Hengelo steam

generators are studied in the SWISH

loop. Void fractions can be measured

by high-speed photography and by

X ray at the exit of the tube. Waterside AP measurements are also

made. Studies to date have indicated

no instabilities on the single tube--

not even under low-load conditions.

Tests indicate that existing methods

of prediction result in DNB at too

high a quality. Also, the measured

heat flux with DNB was higher than

predicted.

Fig. 22. SWISH Loop Thermocouple Con- Low-load conditions have

nections to Steam-generator Tube been simulated by a bypass arrangement. For the helical-coil steamgenerator concept, the SWISH loop will be converted from the single straight

tube to two helical tubes, each of full length but of different coil diameters.

The Dutch stated that they had performed fretting tests in

sodium. The tests had established design clearances. The Dutch experienced fretting and metal transfer in these tests. For design, they had to

establish a clearance that was neither too small, so that fretting took

place, nor too large, so that vibration was possible. They established a

0.4-mm diametral clearance for the steam generators.

Water-test facilities are located at Apeldoorn. These facilities

had been used for waterflow tests of the SNR test-model steam generators

and IHX's. The flow-test model was 0.8 full scale. Only the evaporator

model was tested. The flow-test facility is currently being used to test a

mockup of the degasifier at the bottom of the SNR reactor core.

3. Component and Weld Surface Cleaning

Component cleaning is accomplished with an initial nitrogen

flow, followed by a slowly increasing moisture content until all the sodium

rBoss

TUBE

BRAZE TC

1

I
64

has reacted. The unit is operated and maintained in this oxidized condition.

An alternative method used is to heat the component and vacuum-distill the

sodium.

The experience with welding sodium- contaminated piping has

not been good, even after the sodium film was removed one to two feet from

the weld. The heat from welding melts the sodium, which flows into the

weld area. The resulting welds do not meet requirements. Specifically,

the root pass has been found to have defects. These defects were observed

using double-wall X rays and dye tests.

Discussion took place on the means of cleaning a system before

welding. At Hengelo it has been determined that clean surfaces are necessary to achieve welding standards. One method is to mechanically clean

the surface before welding. There is also concern about welding in remote

locations or in drain-tank nozzles when there is sodium in the system.

According to the Dutch, this area of technology has received limited

attention in the nations with LMFBR programs.
65

VI. SODIUM-WATER REACTION PROGRAMS AND TEST SITES

A. Summary and Comparison

The British, French, and Germans have had major sodium-water

reaction programs in progress for a number of years. These programs

are continuing and even expanding. Each of these countries has delegated

the direction of these programs to a single engineer: British, UKAEA:

A.Bray; French, CEA: N. Lions; and Germany, INTERATOM: K. Dumm.

The French at Cadarache, the Germans at INTERATOM in Bensberg, and

the English UKAEA at Dounreay have conducted large-scale sodium-water

reaction testing of full-scale tube ruptures in representative steamgenerator models using many tubes, and plan to continue such testing for

each new steam-generator concept.

Based on the French and German experience with small-diameter

modular units not containing a cover-gas space, relief is required at or

near the steam generator. The Germans found that including a cover-gas

space in a modular unit results in excessive oscillatory pressure peaks

having a period such that they are reflected as a static loading on the shell.

For this reason, none of the designs for SNR have a cover gas. The French,

on the other hand, are considering large units with a cover-gas space for

the commercial plant, and all the British designs, including the PFR, use

cover-gas spaces. Both the Germans and French find that the initial sharp

pressure peak is of such a short duration with respect to the shell natural

period that it is not significantly reflected in strain of the shell. This initial

peak was not present in the British tests. However, the method used to

initiate the reaction is such that the peak would not be expected. The British

are now modifying their injection system to more closely simulate an actual

tube rupture.

The large-leak tests indicate that the reaction with water more

probably goes to sodium hydroxide and hydrogen, rather than sodium oxide

and hydrogen. For the steam-sodium reaction, the German tests showed

that the pressure effects are considerably different than the corresponding

water reaction, indicating that the reaction mechanism is not the same for.

both cases. The German large-leak results indicate that for the same

failure, a steam leak is potentially more damaging than a water leak.

All countries consider the rupture of a single tube to be the maximum initial accident. However, the British design basis is a single tube

followed by six additional tubes failing within 20 sec. Although propagation

of a large leak to other tubes is considered an unlikely problem in modular

units, this potential problem is increased for larger units where hightemperature tube failures might occur.

The maximum temperatures measured during the large-leak tests

were 1300*C in the French tests and 1000-1100*C in the German tests. The
66

difference might result from the difference in size and geometry between

the two test sections. Also, a film of sodium is left on the tubes following

expulsion of the sodium by the reaction, and this film reacts with steam

from the leaking tube, causing high temperatures.

Turnaround times for the large-leak facilities vary considerably

between the three countries: 4-5 weeks for the British NOAH, 6-12 months

for the British Super-NOAH, 3-4 months for the Germans, and 1 month for

the French. The variation in turnaround time appears to be affected by the

size of the test section, the facilities available for cleanup, the procedures

used, and the priority of the program compared to other programs competing

for manpower.

Mathematical models that reasonably agree with experimental results have been developed in all three countries for predicting the effects

of large leaks. The German and French models include the effects of the

compressibility of the sodium; the British model does not. A sophisticated

treatment of the reaction mechanism using a diffusion-mass transfer model

is provided in the British model. All codes are one dimensional and provide

for variations in water leak rate as a function of reaction pressure. The

hydrogen-bubble temperature used in these models is generally 100 0 C.

Experience with simulant testing in Germany and Britain indicates

that these tests are worthwhile, but cannot duplicate a sodium-water reaction. Simulation could possibly be used to duplicate a predetermined

sodium-steam reaction and may be beneficial in the shakedown of a large

sodium-water reaction facility. Such tests can provide results for major

changes in the test-section geometry. However, quantitative simulation of

a reaction is not possible.

Existing steam-generator sodium-water- reaction relief-system

designs usually consist of a knockout tank and a cyclone-type separator and

have proved adequate for containing the effluent solid and liquid reaction

products. Rupture-disk technology appears well in hand, with special

designs developed where creep is not a problem, and replaceability has

been provided.

All three countries have conducted small water and steam-leak tests

at the same site as the large sodium-water reaction tests, and they are

building or have built larger small-leak facilities of a size similar to

APDA's Rig 10.4 Tests in France indicate that sodium velocity is an important parameter for wastage. The Germans have succeeded in confirming

the APDA wastage data for water leaks within 20%, and leaks with hightemperature steam produced greater wastage rates than equivalent leaks of

water. The French are concerned that corrosion at the site of a very small

leak could lead to a large leak problem.
67

The reference hydrogen leak-detection system is a hydrogendiffusion tube made of nickel; most countries use an ion vacuum pump on

the tube. The British currently use a thermal-conductivity cell, but they

are taking a second look at changing to an ion pump. Work on acoustic

detection is still in progress in all countries; however, its feasibility is

still inconclusive. Experience in the Hengelo steam-generator test facility

indicates that fluctuations in the background hydrogen level in the sodium

could be a problem for leak detection. The Dutch are attempting to reduce

this background level by using a gas-space diffusion device to remove

hydrogen. The Germans, Dutch, and French are working on advanced

hydrogen-diffusion tube-leak detector technology.

In addition to this report, Refs. 5 and 6 contain detailed information

on the sodium-water reaction programs and results for these three countries.

B. France

The French have been conducting sodium-water reaction tests for a

number of years and are planning a continuing and expanding program in

this area. Most of these tests are conducted by the CEA at Cadarache.

However, some special small-leak testing is being conducted at the

Les Renardieres and Chatou facilities of EDF, and digital-computer analysis is being done at Saclay.

1. Large-leak Test Facilities

Sodium-water reaction test facilities at Cadarache are capable

of testing full-scale steam-generator tubes in a model of the steam generator. More detailed descriptions of the test rig appear in Refs. 7-9, and a

schematic is shown in Fig. 23. For Phenix, the tests consisted of one fullscale module of seven tubes. Each test consisted of rupturing one tube out

of the seven. Water flow rate from the ruptured tube in the full-scale

sodium-water reaction test facility was measured indirectly by checking

the flow in a ruptured tube in a separate device and comparing this with a

formula used for the actual test facility. The single tube to be ruptured is

cut by a hydraulic shearing device, which deforms the tube during the

shearing. The actual cross-sectional area of the tube cut was determined

after the test. There was no indication as to how the effects of the reaction

on the cross-sectional area were accounted for.

2. Large-leak Test Results

The test results have been previously reported in Refs. 7-10,

and no new results were presented. The results are discussed briefly in

Section B.8 below.
EXPANSION TANK

--- S

T

PLUGGING -

INDICAJOR t

ICOLD

TRAP

PRO]

I

II T T -

1I n cr7 rrT 1I 

UIJLIRPTLR IC 

- INLET RUPTURE DISC

- OUTLET RUPTURE DISC

DRAIN TANK

, SPARK PLUG LEAK DETECTOR

PRESSURE SENSOR

IMMERSED THERMOCOUPLES

THERMOCOUPLES IN WELLS

URGE

'ANK

H20

CONDENSERLw

its. f 1 

TOTY PE PHEN IX MODULE 4f

CYCLONE

SEPARATOR

WALL THERMOCOUPLES

* FLOWMETERS

STRAIN GAUGES

Fig. 23. Sodium-Water Reaction Test Loop for Phenix Modules

00

1

_ 77 71111177 117T
69

3. Large-leak Mathematical Models

Sodium-water/steam analyses are accomplished by means of

digital computation at Saclay and by analog at Cadarache. The French

model is described in detail in Refs. 7 and 10.

4. Large-leak Future Plans

The Babcock-Atlantique (coiled-tube) steam-generator sodiumwater reaction tests have been initiated. A progression of five shots has

been made, using only one tube in a simplified model. The number of tubes

and extent of modeling will be increased as more information is gained on

the characteristics of the one-tube sodium-water reaction test. A new

large test facility adjacent to the existing facility is being constructed. The

framework was underway. This new facility will initially be used to test

full-scale tubes of Babcock-Atlantique (coiled-tube) and Stein Industrie

(straight-tube) steam generators. There is a planned sequence: (1) Use

the existing test facility to the fullest extent possible; (2) proceed to

larger models in the new facility; and (3) eventually test the 45-MWt

evaporator models shipped from Les Renardieres--the Babcock-Atlantique

unit and the Stein Industrie unit. These units would be shipped to Cadarache

already instrumented to a large degree after completion of steam-generator

tests at Les Renardieres.

5. Large-leak Test Instrumentation

To measure the sodium-water reaction wave front, the French

used 0.35- to 0.50-mm-OD thermocouples. These thermocouples have to

be replaced after two sodium-water reactions. A more detailed description

of the instrumentation is given in Ref. 7.

6. Large-leak Test Facility Operation

Approximately one month is required to reconstitute the test rig

in preparation for another test.

7. Relief Systems

Phenix uses BSB rupture disks, shown in Fig. 24, which have

been working satisfactorily. Three pairs of these 150-mm-dia disks are

in each of the three steam-generator units, a seventh being located at the

surge tank. Advanced rupture disks under development at Les Renardieres

and Saclay for the 1000-MWe Phenix-type reactor plant will be 200-250 mm

in diameter.

A tour of the sodium-water reaction test site at Cadarache

showed a one-stage separator at the bottom of the stack. The reaction
Argon Injection

(Intermediate iArgon Chamber Chamber

ps

The first large-leak experiments produced serious pressure

and velocity effects and severe dam-

- - - age because no surge tank was located

Spark Plugs for close to the sodium-inlet side of the Leak Detection

sodium-water reaction model. After

Fig. 24. BSB Type DV Rupture Disks Used at initial experiments, a surge or buffer

Les Renardieres Steam-generator tank was provided and the resultant

Test Site pressure at the IHX was reduced from

about 21 bars to about 9-10 bars. An

expansion tank contains the sodium pump, but this did not reduce the peak

pressures because the tank is located on the sodium-discharge side of the

steam generator. As a consequence of the test results, the Phenix design

was provided with a surge tank located on the sodium inlet to the steam

generator. The normal pressure in the Phenix surge tank is 8-10 bars.

The Phenix design is based on the rupture of only one tube. Lions indicated

that the Phenix IHX (designed for 15 bars) is the limiting component with

respect to failure from a reaction.

In addition to the buffer tank, protection against sodium-water

reaction pressures is provided by (a) a system of seven rupture disks per

steam generator (mounted two in tandem at each of seven points to protect

against inadvertent failure), discharging to a separator; (b) a sodium dump

tank, to receive secondary sodium from steam generators; (c) quick-closing,

nonautomatic sodium valves, to isolate the steam generator from the

70

N',

wI

products on the ground have amounted

to approximately 5 mg/m2 for each

shot. The test facility takes up pipe

expansion between the sodium-water

reaction unit and the separator by

means of a sliding-pipe connection

using a graphite seal. N. Lions (in

charge of the sodium-water reaction

program) indicated that a rupture

disk, which initially was in the Phenix

surge tank in the gas space, has been

relocated at the bottom, in sodium.

There was no confidence that the disk

would work in the gas space. Further,

excessive corrosion of the disk may

result in the cover gas. The Phenix

pump tank also has a pressure relief

valve.

8. Design and Analysis for

Sodium-Water Reactions

soDiu.
71

remainder of the secondary systems; and (d) a hydrogen-detection system

for each steam generator, fitted with a sodium-selector valve, which can

sample any of 12 modules or take a mixed sample from all simultaneously.

The 1000-MWe steam generators are designed to have a gas

space within the steam generator. The gas volume and pressures will be

varied using a sodium-level control.

9. Small-leak Test Facilities

Special tests are being conducted at the Les Renardieres facility

on caustic corrosion in small cracks. The test rig for this work is entitled

"Cacau." Small cracks are fabricated by tension fatigue using helium-massspectrometer and ultrasonic measurements to determine when a crack has

penetrated the wall and what the leak rate through the crack is. The thimble

representing a steam-generator tube is then put in a sodium-filled pot

(3 liters of sodium), and pressurized water (169 bars, 2350 psig) is provided

inside the tube. A helium leak rate equivalent to 10~ -4l/sec is achieved.

Caustic corrosion occurs in the crack, forming a brittle plug. The French

- Nitrogen are afraid that failure of this

plug could cause a large sodiumwater Reservoir Sodium water reaction. Materials to

Reheater - - be tested in this facility are

Croloy, followed by Type 321

stainless steel. The tubes are

immersed in sodium for - Flowmeter E. M. Pump m i SOdum f

---eEMPp1600 hr or more and then examined. Hydrogen-detection

capability is provided. Test Section Sample Tube

* Ii Wastage tests are

L. _

Injector 

]s Disk- Economizer

-" Filter

Nitrogen

Nitrogen Cold 

Trap --

- 2 Discharge

Reservoir

Fig. 25. Schematic of Small-leak Sodium-Water

Reaction Facility at Cadarache

being conducted at Chatou (in

addition to Cadarache) using

holes as small as 20 microns,

which are drilled with a laser

beam. The wastage tests at

Cadarache are conducted in

the facility shown in Fig. 25.

10. Results of Smallleak Tests

Most small-leak

tests are conducted at Cadarache.

The pinhole wastage tests have

been conducted for the past

several years. Several
72

specimens having pinhole effects were shown. Lions indicated that the

sodium flow velocity is important, as illustrated in Fig. 26 (from Ref. 8).

If the velocity is less than a certain value (about 3-5 ft/sec), damage

could be obtained in adjacent tubes. Small-leak (pinhole) tests have been

conducted at a rate of three or four per week. This indicates that the CEA

has accumulated a large amount of small-leak data.

_ " A 0.3 

S A 0.05-0.3 - _ -*

100

I

B

5mm

e28x 4

I

10

E

E

bm1

0.1

0,1

A

Leakage Flow g/sec

Fig. 26. French Small-leak Test Results for 2 Cr-1 Mo-1 Nb

11. Leak Detection

Leak detection in the Phenix modules is provided by a 0.3-mmthick hydrogen-diffusion membrane, made of nickel. An ion pump is coupled

with a sodium-sampling manifold system. The manifold system samples the

inlet and outlet of each module or a mixed sample from a group of 12 modules. The location of the leak-detection system in the Phenix secondary

circuits is shown in Fig. 27, and the system itself is shown in Fig. 28.

Mass spectrometers had previously been used, but have been dropped in

favor of the ion pump. The French have measured hydrogen-diffusion rates

through the tubes of a module equal to 10 mg/hr per module. This works

out to a diffusion rate of approximately 0.2 mg/hr per square meter. Initially, when the unit is first started up, the rate is three to four times as high.

Sodium Water Symbol Tube

Velocity,m/sec Temp.* C Temp.* C Pressure, bars

o A 2.5-3 350 320 170

* B 2.5-3_33

"+

100

A 0.8-1 r rI
73

RE RT ME

TO H2 vi

DETECTOR v W

--- 1 DISCHARGE TO

ATMOSPHERE

E E .. I

-- SP

- --REACTOR-- Z

- MAIN SECONDARY SODIUM CIRCUIT (1 OF 3) - SODIUM-WATER REACTION DISCHARGE PIPING - HYDROGEN DETECTION CIRCUIT

PURIFICATION CIRC~itT

ARGON CIRCUIT

KEY:

E INTERMEDIATE HEAT EXCHANGER R SODIUM REHEATER

P PUMP M MIXER

RE EXPANSION TANK WITH SECONDARY PUMP S SELECTOR

RT SURGE TANK SP SEPARATOR

CS DRAIN TANK VI ISOLATION VALVE

PF COLD TRAP VR CONTROL VALVE

IB PLUGGING INDICATOR ME RUPTURE DISKS

Fig. 27. Location of Leak-detection System in Phenix Secondary Circuit

--

Circuit identical to

that of the superheater -0

Arson Vacuum Pump

Selector Mass'

Spectrograph

tl aDetector 0--' 0

c J Mixer

C"" Circuit identical to

-- To main circuit that of the superheater

TostraeIan

v

aIW

E. M. Pump E. M. Flowmeter

To storage tank

Fig. 28. Leak-detection System for Phenix Steam Generator
74

Regarding other types of steam-generator leak detection, the

EDF engineers indicated that acoustic techniques for locating a leak in the

steam generator will be used as a confirmatory test. Apparently this concept is in the development stage and will be used only as a backup system.

Present plans are to insert the acoustic sensors in the sodium rather than

on the vessel. At present the Phenix steam generators use no acoustic

detection. Plans are underway to obtain results from acoustic tests in the

Phenix core for use in future sodium-water reaction tests. Steam-generator

background noise is quite high, a serious detriment to acoustic detection.

The French leak-detection work is reported in Refs. 3 and 11.

C. United Kingdom

The British have been engaged in sodium-water reaction testing on

a continuous basis for a number of years and are now pursuing a program

in support of the CFR. Alan Bray is the chief investigator in this work.

1. Large-leak Test Facilities

The purpose of the large sodium-water reaction testing in the

Super-NOAH test facility is to develop calculational techniques for predicting the effects of large sodium-water reactions. The third test in the

Super-NOAH facility was due to be run in the second half of 1972. A leak

through the packing of a water valve prevented running the test while the

team was at Dounreay. The Super-NOAH facility consists primarily of a

large vessel, 21 ft high and 3 ft in diameter. Also included are a waterinjection system and a relief system consisting of a large dump tank

followed by a cyclone separator. Figure 29 is a simplified flow sheet of

the Super-NOAH test rig. Tube bundles are installed in the reaction vessel

to simulate the section of the PFR evaporator to be modeled. The British

have purchased a spare reaction vessel for this facility.

The injection system for Super-NOAH consists of a tube containing a longitudinal defect. The tube diameter is larger than that of a

normal PFR evaporator tube so that the cross-sectional area for flow is

twice that of an evaporator tube, and a double-ended rupture of a single

tube is simulated. Before test initiation, the sodium fills the injection tube

up to the level of the sodium in the test vessel. Argon cover gas blankets

the section of the injection tube between the sodium and the rupture disk

that isolates the water reservoir from the injection tube. More detailed

descriptions of the NOAH and Super-NOAH facilities are given in Refs. 12-17.

2. Large-leak Test Results

The results of the first test in Super-NOAH were described in

the IAEA working-group meeting on sodium-water reactions at Argonne

National Laboratory in 1968.5 The pressure instrumentation did not function

during that test. The large-leak test results from Super-NOAH were

scheduled for publication in 1973.
\BURSTING

BUST. DISC.

ING

DISC.

1C

STORAGE

VESSEL

1I

I I

I I

II

II

I I

SUR'

V ES

II

II

Ii

FIjTER FITER

SODIUM PIPEWORK.

-- -- WATER PIPEWORK.

BURSTING r '- - - - - - - - - -- - -- - - - - - -1

DISC. 

-1 A ---- -- s-- 

BURSTING I

DISC. 1

III 

I

FLAME II

ARRESTER I I

I 1

FLAP\ I 

I-REACTION

VALVES -- D.P

TEST VESSEL CELL.

I II

GE III SE PRESSURIZER 

SEL II

I 1 1

II I

II

II

____ II L

--- D-- HAND OPERATED

VALVE.

RELIEF VALVE.

NON - RETURN

VALVE.

MAG. VALVE OPERATING

A PNEUMATIC VALVE.

- CONDENSATE POT.

Fig. 29. Super-NOAH Rig, Simplified Flow Sheet

U

C

STORAGE--

VESSEL

B

L_________1
76

a. Test 1

Test 1 was conducted in Super-NOAH at the end of 1968.

It consisted of a water leak rate of 10-15 lb/sec for 12 sec. The sodium

and water were at 300 C (572 F), and the rupture simulated a single-tube

rupture of a 3/4-in.-OD tube, which was for an earlier design of the PFR

evaporator. The water pressure was 2300 psi. A small leak occurred in

the injection system before the large reaction was to be initiated. For this

reason, initiation of the large leak was accelerated at the expense of having

the recorder for the reaction pressures jam; therefore no pressure information was obtained. Very little damage occurred to the tubes, and there

was no dimensional change to the shell. Based on calculations, the peak

pressure was on the order of 300 psi, while the shell is designed for 500 psi.

b. Test 2

Test 2 simulated a leak in the bottom end of the evaporator,

and the same temperatures and pressures were used as in Test 1. This test

used a duplicate of the PFR effluent system, whereas in Test 1 no centrifugal

separator had been employed. The tube size was the same as that for the

PFR evaporator. The waterflow started at 90 lb/sec and decayed down to

60 lb/sec. Approximately 200 lb of water were injected. The cyclone

separator extracted 220 lb of sodium, and virtually no sodium was discharged from the effluent system. No white cloud or flame appeared during

this test, whereas a spectacular cloud and ball of fire were evident during

Test 1 (in which the cyclone separator had not been installed). Reaction

products in the relief line reached velocities of 150. ft/sec at about 0.3 sec,

dropping to about 50 ft/sec at about 2 sec after the initial injection.

500

a Z 400

3:oo

J

Wa 300 CfW

2O 200

W 00

(nJ

a0 2

TIME, sec

3

Fig. 30. Plot of Pressure vs Time for PFR

Steam-generator Super-NOAH

Sodium-Water Reaction Test 2

tiating the reaction in which sodium

The pressure transient resulting from Test 2 is shown in

Fig. 30.

During this test, the shell

liner deformed and failed in the

reaction region. The maximum

pressure experienced by the shell

was 400 psi. No large initial pressure peaks were recorded, only a

very slow buildup to a peak pressure

followed by a decay. This result is

contrary to the results being predicted by the U.S. codes and shows

a much longer duration of the highpressure peak. The method of iniand cover gas are in the injection tube

could cause this behavior. Therefore the British are redesigning their

- 1 1 -
77

injection system so that water will be immediately adjacent to the defect

when the test is started. The reactions at the bottom of the evaporator are

located between the bottom two support grids. Considerably more reaction

products were retained in the test vessel during Test 1 than was experienced

in Test 2. Based on the pressure profile measured during Test 2, the British

are getting good agreement between their mathematical model and the test

results.

c. Test 3

Test 3 was about to be run and consisted of a simulation of

a rupture at the top end of the evaporator at the interface between the sodium

and the cover gas. The temperature for this test will also be 300 C.

Two tests on preweakened tubes were conducted in NOAH:

one test using a flattened tube and another test using two thinned tubes.

However, no effect on these tubes was seen. Additional discussions of test

results are given in Refs. 12-18.

3. Large-leak Mathematical Models

The British mathematical model is a one-dimensional model,

which includes accounting for the back pressure of the reaction and the

convective mass and heat transfer between the reactants. Prediction of the

water leak rate is based on experimental work conducted by Henry, published in an American journal. Water-leak-rate measurements have been

made using a turbine flowmeter and a drag-plate-type flowmeter. These

measurements were made during special tests of waterflow through a

defected tube. The model predicts well the results of the Super-NOAH tests.

4. Large-leak Future Plans

Alan Bray felt that the large integral proof-test approach to

sodium-water reactions was not the best approach, unless the customer

demanded it. He felt that simple geometry tests and tests directedparticularly to confirming the analytical model and developing design guidelines

would be more beneficial. The British are planning to perform some simple

geometry tests during shutdown of the Super-NOAH.

Additional tests to be run in Super-NOAH are as follows:

Test 4. This test simulates a defect at the bottom of the

evaporator and uses a sodium temperature of 450 C and water temperature of 350 C.

Test 5. This test simulates a rupture at the bottom of the

evaporator. The conditions will duplicate Test 2, with a sodium and
78

water temperature of 300 C. The duration of the test will be extended to

approximately 20 sec to test for the possibility of local high temperatures

in the unit during an extended reaction. Approximately 2000 lb of water

will be injected.

Test 6. This test will simulate the rupture of more than one

tube by using a 3-in.-dia water-injection pipe. The leak rate will be about

200-250 lb/sec.

Test 7. This test will simulate a rupture in a tube immediately

adjacent to the shell of the reaction vessel. All previous tests used a

rupture in the center of the tube bundle in the reaction vessel.

Additional tests up to Test 13 or 14 are planned on other

designs to be considered for the CFR. For these tests the British will

probably have to extend the facility. Until they have sufficient confidence

in predicting the pressures, they expect to require tests approaching full

scale. Tests including preweakened tubes in the reaction area are also

planned.

5. Operation of Large-leak Test Facility

The staff for the Super-NOAH facility is six engineers and five

or six technicians. Hardware cost is approximately . 10,000 ($26,000) per

test. The best estimated turnaround time for the facility is 5-6 months,

although this has not yet been achieved. The average turnaround time for

Super-NOAH for the first several tests has been 1 year. However, some

smaller testing has been done in the facility between each of the large

tests. Steam cleaning uses a one-to-one mixture of steam and nitrogen by

volume, followed by a water wash. It takes approximately 2 days to wash

the vessel and 1 week for cleaning after the solid products and sodium

have been removed. The removed sodium is cleaned by filtration.

The turnaround time for NOAH tests was 4-5 weeks. The

larger-scale Super-NOAH tests require 4-5 months for removal of the

bundle. With a leak at the bottom of the reaction vessel, approximately

1 tons of sodium are ejected to the dump tank. The remaining products

are very viscous, with a nominal melting point of 300*C. The vessel is

cooled down, the bottom head removed, and the solid reaction products

chipped out. The bottom is then reinstalled, and the sodium is melted out

and drained. The vessel and internals are then steam-cleaned and rinsed

with water. Alan Bray stated that a turnaround time of a few months for a

test program involving 500 lb of water was completely out of the question.

6. Large-leak Test Instrumentation

The type of strain gauges used on the reaction vessel are

Hawker-Sidley Dynamics Ltd. flame-sprayed strain gauges. Additional

instrumentation information is provided in Refs. 12-17.
79

7. Simulant Tests

A plan for gas-water model testing, including a model of the

entire secondary system, has been submitted for approval and funding.

The gas-water tests would be done in three stages from small to full size.

Estimates for conducting the full-size gas-water tests indicate that the cost

is approximately the same as conducting sodium-water reaction tests. The

tests will be conducted in the same geometry as the sodium-water reaction

tests.

8. Relief Systems

The relief system for PFR consists of a collection manifold,

which goes to the secondary dump tank in each loop. The relief line leaves

the top of the dump tank and goes to the cyclone separator, which has its

own retention tank.

The British conducted rupture-disk tests, using both gas and

water. However, they feel that the rupture properties of the disks are not

important. They are interested in the rupture-disk design developed by

INTERATOM and sold by Albert Zeifle, A.G., in Germany. The rupture

disks used in the PFR depend on a shear pin. The shear pin, which operates

on a hinged member, can be replaced. The rupture disk itself, therefore,

only serves as a seal and is not subject to high stresses. A safety-lock

mechanism is used during replacement of the shear pin.

9. Design and Analysis for Sodium-Water Reactions

The British are concerned about the impact forces on the upper

head of the steam generator from the impinging slug of sodium resulting

from a reaction. They prefer not to discuss a maximum-credible-leak concept, but rather deal in probabilities of occurrence. Bray felt that a singletube rupture with later propagation would be the maximum reaction that

could be expected. PFR uses sodium-side isolation valves, which are

closed in 1-2 sec and are actuated by rupture-disk failure. The valves

are butterfly-type units. The shell design pressure of the PFR is 500 psi.

This was selected by judgment, and the British have been backfitting their

analysis to this original' selection. Bray believed that a spare tube bundle

was desirable for the plant, but he recognized that economics and design

would be the prime factors in making such a selection.

10. Small-leak Test Facilities

The small-leak facility, similar in size and scope to the

APDA Rig 10, was scheduled for completion in the fall of 1972. The British

anticipate that they will be able to conduct two injections per test run in a

2" week period. In this facility, they intend to examine steam leaks and produce damaged tubes to send to those involved in developing inspectiondevices
80

for detecting wastage in a steam generator. The small-leak tests will be

conducted using about 0.1-in.-ID stainless steel capillary tubes, which will

be heated by passing current through the capillary. A 1/2-in, rupture disk

is in series with the capillary. During an injection, argon gas will first be

discharged; then the rupture disk will be broken, and the injection fluid will

be slowly changed over to liquid water. Following an injection, the sodium

flowing back into the capillary tube can be frozen and then remelted for

another injection.

11. Small-leak Test Results

The British have observed the shape of reaction products and

feel that earlier tests performed demonstrated that there is an optimum

jet size for maximum damage. A particularly important factor is that

steam leaks give greater wastage than water leaks for the same quantity

of water involved, as also indicated in tests at APDA. In the planned

wastage work, they intend to look particularly at reheater steam conditions

for wastage. At this time, the effect of reheat steam conditions on wastage

is not known. Previous small-leak test results are reported in Refs. 12,

13, 19, and 20.

12. Wastage Models

The British had tried to characterize small-leak shapes with

X rays, but this work was unsuccessful and has been terminated. A number

of erosion theories are now popular for explaining small-leak damage.

Harwell is studying jet shape and reaction-particles, and is using flow tests

to simulate sodium-reaction water jets and directing work to help establish

an erosion model.

13. Leak Detection

Hydrogen in sodium is detected in PFR by diffusion of hydrogen

through a nickel membrane, followed by measurement with a thermalconductivity cell. The British are now looking at an ion-pump-type detector again. The diffusion tube (13/16-in. OD) will also be used in the cover

gas. The British intend to calibrate the PFR hydrogen-detection systems

in situ. Water will be injected in the evaporators, and hydrogen in the

reheater and superheater. They have performed response-time and sensitivity tests of the hydrogen detectors in special tests in Super-NOAH by

injecting the water and hydrogen in both stagnant and flowing sodium. They

found differences in the results between injection of hydrogen as opposed to

water. This is due to the different types of bubble formation and their effect

on the reaction. Injections are required to calibrate the hydrogen detectors

in the plant. Large leaks are detected by pressure detectors in the gas

space. Additional studies on leak detection are reported in Ref. 21.
81

14. Damage Location and Assessment

The British are working on NDT development to determine the

amount of damage from wastage after a leak. This is being performed by

others at Dounreay in parallel with the small-leak tests to be conducted

soon. In addition to providing a surge volume, the argon blanket system

in the reheater, evaporator, and superheater assists in locating leaks

detected during plant operation. If a small leak is confirmed during operation, a controlled dump of the steam side will be followed by isolation of

the steam generator and pressurization of the steam side with argon, which

will bleed through the leak. Sodium level will then be changed and, using

acoustic detectors, the level of the leak will be located. Helium will then

be used on the shell side to try to locate the tube that has failed.

D. SNR Consortium

Most sodium-water reaction testing and analysis for the SNR Consortium are being performed by INTERATOM in Germany. Some leakdetection work and large-leak analysis work is being done in the Netherlands

in addition to the leak-detection experience being obtained in the Hengelo

facility. The INTERATOM experimental program on sodium-water reaction

and leak detection started in 1968. Computer codes have been devised to

predict the performance of large steam/water leaks, but many questions

remain, such as self-sealing of a leak, deformation of steam-generator

internals, maximum reaction temperature, performance of the bellows in

the straight-tube design, chemical- reaction mechanisms, rupture-disk

behavior, and other factors pertaining to what actually happens during the

design-basis accident, a guillotine failure of a single steam-generator tube.

1. Large-leak Test Facilities

The German sodium-water reaction facility (ASB plant) at

INTERATOM was constructed in 1968 to deal with the safety problems

associated with sodium-water reactions in steam generators, such as the

total rupture of one or more tubes and small water leaks (pinholes).

The tests led to layout criteria for the steam generator and the

development of leak-detection systems.

The test facility consists of five main parts:

a. A 200-mm-dia sodium loop (30 bars, 550 C) with storage

tank (11 m3 ), centrifugal pump (120 m3 /hr, 6 bars), combined heater and

cooler (150 kW), surge tank, and cold-trap system.

b. Pressurized-water system with a main vessel (0.5 m3

280 bars, 450 C) for large-leak tests and a small vessel for pinhole tests.

Water- and steam-injection rates up to 40 and 10 kg/sec, respectively, are

possible.
82

c. Test sections (steam-generator model with maximum

possible dimensions of 5800-mm length and 700-mm diameter) and several

small-leak test sections.

d. Reaction-products separation system with a centrifugal

separator and a wash column (10 m3 each).

e. Cleanup system of four cold traps.

Since 1968, the following tests have been performed:

a. Twelve large sodium-water reaction tests in a 0.260-m3

vessel to understand the reaction behavior in vessel geometries.

b. Five large sodium-water reaction tests in an about

1:3 scale model (1.7 m3 ) of the straight-tube steam generator of the

SNR-300.

c. More than 50 small-leak tests and several tests on leak

detection (development of acoustic methods; performance and calibration

of hydrogen-detection systems).

The sodium-water test rig is shown in Fig. 31 and currently

has installed in it an 81-tube bundle simulating the straight-tube SNR

steam-generator design. A comparison between the large-leak model and

the steam generator is as follows:

Length, m

Diameter, mm

No. of tubes

Steam Generator

23

600

211

Large-leak Model

5.6

325

81

63

69

7 4

5

KEY

1. 120-m3 /hr pump

2. Surge tank

3. Heater/cooler (150/100 kW)

4. Cold traps (4)

5. Storage tanks

6. 500-liter water-steam pressure vessel

7. 30-liter water-steam pressure vessel

8. Temperature-gradient section

9. Large-leak test section

10. Small-leak test section

11. Reaction-products tanks

Fig. 31. INTERATOM Large-leak Test Facility

The tube bundle in the model is pressurized up to 180 atm by

steam, water, or gas. The steam-generator shell bellows was also modeled

in this test rig. The initial design used pressure relief located in the main

loop, not at the steam generator. These early tests determined the necessity
83

to provide pressure relief at each end of each steam-generator unit. The

test rig was modified to provide this capability. The relief-line diameter

is one-fourth that of the diameter of the reaction vessel.

The first three shots were conducted with the initial pressurerelief system. Strong pressure peaks were observed in the sodium. The

cause of the pressure peaks was determined to be the dynamics of the hydrogen bubble. Relief is necessary before major acceleration of the sodium

slug has occurred. This consideration led to the present configuration

with the rupture disk at the sodium

inlet and outlet of each steamgenerator unit. The facility relief

system is shown in Fig. 32.

Additional information

on the test facility is given in

Refs. 22-25.

2. Large-leak Test Results

In their large-leak testing, the Germans started at lower

leak rates and built up to larger leak

if rates as follows:

Test 1: 5 kg/sec

Test 2: 9.5 kg/sec

Test 3: 12 kg/sec Fig. 32. Relief System for INTERATOM

Large-leak Test Facility Test 4: 18 kg/sec

The peak pressure observed for Test 4 was 51 atm. Results from Test 5,

which was a steam leak, showed that the pressure results are considerably

different, as indicated in Fig. 33.

TEST 5:

E STEAM, 4.5 kg/sec

0

Fig. 33 w 50 - ------- -- TEST 4:

WATER, Plot of Pressure vs Time for INTERATOM 18 kg/sec

SNR Sodium-Water Reaction Tests 4 and 5 Cr

0 

0 5-8 15-20

TIME, msec
84

The pressure results from a steam leak are similar to those

measured during injections of gas into water, and the pressure peak must

be applied to the shell as a static load.

All tests would indicate that the sodium reacts with water to

form NaOH plus -H 2 , including injections made with superheated steam.

The maximum temperature is 100 0 *C, rather than the 1300-1400 C reported by Lions in Ref. 7. This difference might be attributed to geometry

differences, although it might also be due to variations in heat transfer to

the tube, the presence of fog flow, or inertia of sodium in the system. (The

Germans had 81 tubes in their test unit; the French had only seven.) This

temperature is used in the mathematical models for the reaction. The

Germans expect that, for larger models, this temperature may become

even lower because of heat loss to the tubes and other structures.

Originally, pressure relief was off the main piping by two

450-liter vessels, each containing a rupture disk. After the first test,

the Germans changed to pressure relief directly from each module, since

very large pressure oscillations were experienced in the sodium, as described above. The frequency of these oscillations was such that the load

on the steam-generator shell was essentially static rather than dynamic,

and the bellows was not designed for such high pressure applied statically.

At first, the Germans thought the oscillations were due to not having a

bellows in the shell of the model. However, it was later shown to be a

result of hydrodynamic effects in

the sodium. If pressure relief

100 atm PRESSURE OSCILLATIONS occurs soon enough (before signifiE IN SODIUM cant acceleration of the sodium slug),

IrIno oscillations will result. The re-

! fsults of Test 3 illustrating these

oscillations are shown in Fig. 34.

Although the peak pressures

PRESSURE in the sodium reached 100 atm,

15-20 pressure in the real steam generator

TIME, msec could go to 150 atm from a tube rupture. (Note Section IV.D.3, which

Fig. 34. Plot of Pressure vs Time for INTERATOM indicates that with a gas space,

SNR Sodium-Water Reaction Test 3 pressures could reach 500 atm.)

These oscillations were damped out

by the time the pressure wave reached the IHX. Strain measurements

made on the shell reflected the shell response to these pressure oscillations

as being essentially a static load. Response of the strain gauges also showed

that the initial short-duration pressure peak did not affect the shell structurally and could be neglected from a structural standpoint.

Measurements (see Figs. 35 and 36) indicate that the hydrogen

bubble appears to expand isothermally rather than adiabatically. During
85

the German tests, the tubes surrounding the water-filled tube were filled

with nitrogen. Thermocouples installed away from the steam generator

wall would fail. No secondary tube failures have occurred during a reaction, despite very high apparent heat-transfer coefficients. Measurements

on a tube wall, as shown in Figs. 35 and 36, were used to determine the

heat-transfer coefficient.

T/C I

TEMP, 1100*C------

TUBE T/C020

WALL

H 2 BUBBLE

TEMP 

550 - ------

T/C3 -SODIUM/3503

TEMP TUBE IN 150 500

REACTION ZONE TIME, msec

Fig. 35 Fig. 36

Location of Thermocouples in Reaction Thermocouple Temperature Readings for

Zone for INTERATOM SNR Sodium- Locations Shown in Fig. 35 for INTERATOM

Water Reaction Test 3 SNR Sodium-Water Reaction Tests

An overall heat-transfer coefficient of 1.15 x 104 W/mZ- K was

obtained from many tests. The overall coefficient calculated for hot hydrogen would be only 1 x 103 W/mZ-oK. The Germans feel that the high apparent

coefficient is due to the reaction between flowing steam and a sodium film on

the surface of the tubes. They calculate a film thickness of 0.2-0.25 mm

after expulsion of the sodium; this agrees with earlier calculations on the

KNK steam generator and measurements made during sodium-boiling experiments at Ispra. Temperatures up to 1300 C were measured in the

relief pipes at a very late stage after passage of the hydrogen bubble.

This is apparently due to a reaction between the remaining sodium film

and the steam from the leaking tube. Failure of thermocouples away from

the wall (T/C 1) can be ascribed to these high coefficients.

The rupture disk fails within 5-7 msec, the hydrogen bubble

enters the relief line 150 msec after the tube break, and the unit is completely emptied in 500-700 msec. Measurements on the spark-plug probes

and thermocouples in the relief line indicate that the sodium slug actually

is initially well defined. However, there appears to be a fog-flow region

at the interface between the hydrogen bubble and the slug. This two-phase

flow results in higher flow-resistance factors and higher pressures

measured at later times than are predicted by the model, as illustrated

in Fig. 37.

Because of the higher resistance, the velocity of the effluent

in the relief lines is much lower and the measured values of loading on the
86

relief line were approximately onePREDICTION tenth of those predicted. This loadEXPERIMENT

ing was measured using strain gauges

on the supports for the cyclone

0 0 separator.

The maximum water leak rate 200 800

TIME, msec injected was during Test 4, when

18 kg/sec of water was injected for

Fig. 37. Plot of Pressure vs Time--Prediction 4 sec. The Germans feel that one

vs Experiment-for INTERATOM SNR must inject for more than 1 sec to

Sodium-Water Reaction Test 3 obtain the temperature effects that

may occur later in the reaction; however, for verifying a mathematical model and predicting pressure effects,

1-sec injection times appear adequate.

Previous large-leak test results are published in Refs. 22-25.

3. Large-leak Mathematical Models

The Germans feel that several series of sodium-water reaction

tests and several years of analysis using computer codes have confirmed

their assumptions and the analytical procedure used.

The Germans did blowdown tests to obtain the steam-water flow

rate using temperature and pressure measurements at the injection nozzle.

Bubble volume was calculated by measurements of sodium displacement.

With measurements of hydrogen pressure and knowing the water-injection

rate combined with bubble volume, the reaction temperature can be calculated if a particular reaction is assumed (Na + H2 O-NaOH + 2H2 ). Figure 38 compares the temperature calculated as indicated above, the

temperature actually measured in the reaction zone, and the temperature

assumed for the mathematical model.

ASSUMED

~890 - --

Fig. 38

Plot of Temperature vs Time--Assumed, Cal- taI CALCULATED

culated, and Measured--for INTERATOM SNR W

Sodium-Water Reaction Tests MEASURED

~100

TIME, msec

When the temperature in the bubble was calculated, the apparent

temperature curve derived indicated a hydrogen temperature below the sodium or steam temperature. Some temperature information was lost during
87

the test, but two points were obtained to corroborate the rest of the data.

The Germans expect that if they apply the calculated temperature to the

mathematical model, they may get better agreement between prediction

and test results. They think that the difference between the calculated and

measured temperatures may be a result of the formation of sodium hydride

initially during the reaction.

The Germans are interested in incorporating the reactionkinetics model developed by the British (based on mass transfer through

the hydrogen bubble) into their own code, since their own code does not

predict the results very well at early

times between 2 and 6 msec. The

140- 4 ------ THEORETICAL decrease in pressure immediately EXPERIMENTAL

\Efollowing the first pressure peak is

120- considerably overestimated by the

mathematical model, as shown in

E loo Fig. 39. This discrepancy may be

WI0due to (1) the assumption that the

S8O bubble is in the form of a wafer,

completely filling the cross section

60 of the model, or (2) the fact that not

all water reacts initially as predicted

40 in the British model.

20 The German mathematical

model is described in Refs. 22 and 25, 0

0 10 20 30 and the Dutch model in Ref. 26.

TIME, msec

4. Large-leak Future Plans

Fig. 39. Plot of Pressure vs Time--Theoretical vs

Experimental--for INTERATOM SNR

Sodium-Water Reaction Test 4 The next unit to be tested

in the large-leak facility is the

helical-coiled unit. Possible deformations of the bundle, which could reduce

the effective relief area are among the areas under study. The Germans feel

that only the steam generator needs to be modeled, and not the system, since

with the rupture disk at the component, they do not feel they will have severe

pressure oscillations in the system external to the steam generator.

5. Operation of Large-leak Test Facility

Normally it has taken the Germans 3 or 4 months to clean up

the system and make repairs between tests. Deducting the need for repairs,

the turnaround time could be reduced to 2 months.

6. Large-leak Test Instrumentation

The Germans measure hydrogen bubble pressure and other pressures at 16 locations. Temperatures are measured at 20-30 locations, and
88

the velocity of the sodium in the relief lines is measured using contact

probes and thermocouples. The thermocouples used are 0.5-mm-dia

Chromel-Alumel grounded-junction types. Their response time in liquid

lead is 17-20 msec. Sodium-ejection velocity was measured in the relief

line using spark plugs to give signals, backed up by thermocouple measurements. Hydrogen velocity was obtained from this measurement. The

steam-generator model shell used hermetically sealed SG-425 Microdot

strain gauges.

Additional information on large-leak test instrumentation is

provided in Refs. 22-25.

7. Simulant Tests

Before large-leak Test 4, seven reaction-simulation tests were

conducted in the steam-generator model filled with water. Nitrogen was

injected at 180 atm. No quantitative results were possible. Although qualitative results were obtained, quantitative simulation does not appear

possible. The tests did show the effect of changing the location of the

rupture disk. Oscillations of pressure similar to those experienced in the

reaction tests were obtained. These occurred at different frequencies than

those in the reaction tests. The computer codes predicted that these oscillations would occur in the gas-water testing. The Germans feel that simulant testing cannot replace the sodium-water reaction tests. These tests

cannot simulate complete rupture of one tube and result in a completely

different picture of pressure in the reaction zone. The acceleration of the

injected gas is very important. Simulant tests could possibly be used to

simulate a steam leak into water and are useful for rig shakedown.

8. Relief Systems

The rupture disks used by the Germans (shown in Fig. 40) are

capable of withstanding a vacuum on

DISK EJECTED IN

THIS DIRECTION

ELECTRON

SEAL 150mm

WELD

DISK-SPHERICAL

SHAPED

SODIUM

Fig. 40. German Rupture Disk (manufactured by A. Ziefle)

the sodium side. The estimated life

of the ruptured disk is about 6 years,

taking into account the corrosion and

age-hardening effects. The disks

(replaceable) are in the shape of a

sector of a sphere with the convex

side facing the sodium. The disk

collapses consistently by buckling at

the circumferential electron-beam

weld between the disk and its flange

ring, and the complete rupture disk is

expelled into the cyclone separator.

In all experiments, the rupture disk

was found in one piece in the separator

tank, wound up like a sausage.
89

Analytical studies, conducted to determine pressures as a

function of rupture-disk size, indicate that above the present size, the

peak pressure is fairly constant. The present rupture-disk diameter and

collector-pipe dimensions have been adopted as a tradeoff between peak

pressures and loadings on the relief piping. Rupture disks of this type

have now accumulated 17,000 hr under normal operating conditions in KNK.

The disk material is niobium-stabilized stainless steel. The sodium operating pressure is 8 atm, and the rupture-disk burst pressure is 17 atm.

The waterside blowdown period for the SNR is estimated at 30-40 sec

following automatic shutdown in response to disk-burst indication.

As noted in Section IV.D, the length of the 250-mm-dia rupturedischarge collection headers to the cyclone separator is about 20 m for the

line at the upper end of the modules and 40 m for the line at the lower

module. The diameter of the lines between the rupture disks and headers

is 150 mm.

9. Design and Analysis for Sodium-Water Reactions

INTERATOM does not provide a gas space over the sodium in

the SNR steam-generator modules. Cover gas is maintained in the pump

and in the cold-leg expansion (surge) tank of the secondary loop. However,

the exact location of the tank was not indicated. Experiments and analytical

studies have indicated possible adverse effects due to sodium hammer, which

can result from sodium-water reactions in the presence of a gas space in

the steam generator. The gas space contributes to a delay in depressurizing

the system through a delay in bursting of a rupture disk. This results in

possible higher reaction-zone temperatures and stronger pressure pulses

transmitted to adjoining modules and throughout the secondary system,

including the IHX.

Analyses conducted for the British by the Germans indicate that

for steam-generator concepts proposed for the British CFR, using a covergas space, pressures as high as 500-600 atm could occur at the tube plates

from sodium hammer for conditions in which the pressure in the hydrogen

bubble was only 150 atm. These peak pressures would occur as an

oscillation.

Without a cover-gas space, rapid depressurization occurs in

the event of a tube leak, and the Germans believe the design can therefore

be based on the guillotine failure of a single tube. Studies indicated that the

depressurization-pressure pulsation problem might be alleviated with

proper design of gas volumes. However, the control of level and pressure

in the gas space would present significant difficulties.

The initial peak pressure resulting from a guillotine tube rupture

in a system without a gas space dissipates within less than 10 msec and is of

such little energy that the steam generator is not affected structurally by the
90

burst. Also, the Germans predict no significant forces on adjacent tubes

during the very early times. The secondary pressure peak is of longer

duration and must be treated as a static load in the structural analysis.

The bubble growth occurs at such a rate that it would be necessary for a

second tube failure to occur within 30 msec for it to have any significant

effect on the pressure transient.

As noted in Section IV.D, if a large leak occurs, INTERATOM

proposes that the affected steam-generator module be replaced by a spare

unit. This will require 2-4 months. Resulting reaction products and loop

contamination would have to be removed. Procedures for such removal

have not yet been developed. For small leaks, the unit will be repaired in

place. However, the water header must be cut to gain access for plugging.

At present, there is no way of accurately evaluating the extent of damage

or the amount of repairs to be made, nor is there a way to verify that

repairs would be effective. There is concern that small leaks may plug

themselves and become extremely difficult to detect after cooldown. Such

an experience occurred in the 5-MWt steam-generator test facility.

British statistics on tube failure rates indicated that the most pessimistic

failure rate that can reasonably be assumed for small leaks in a steam

generator is one tube failure per year. INTERATOM assumes a failure

rate of one or two tubes per unit for the life of the plant.

Following an identified tube leak, the water system would be

blown down. The isolation valves isolating the steam generators from the

remainder of the secondary system would be closed, and the sodium would

be held in the system for a while, after which sodium would be dumped into

a contaminated sodium dump tank. Sodium would not be dumped immediately after detection of the leak because of the concern over a possible

secondary sodium-water reaction in the dump tank. The sodium cyclone

separator can hold about 6 cubic meters of sodium, which is greater than

the volume of one module. Test results, as well as other analytical studies

and laboratory results, indicate that no caustic corrosion should be expected

as a result of the large leak. The steam generator should empty in about

40 msec. Only small amounts of reaction products are expected to be

transported to adjacent units.

10. Small-leak Test Results

Though there was no discussion of small-leak test results,

target specimens were available that indicated a strong similarity between

the types of damage obtained in the German and U.S. tests. Dumm stated

that they had conducted several tests using water duplicating the APDA

wastage test conditions and obtained results within 20% of the APDA data.

For this reason, the Germans planned to explore ranges in parameters

beyond that covered in the APDA tests. Some previous test results appear

in Ref. 25.
91

11. Leak Detection

In the helical-coil steam generator, small sodium flows are

introduced adjacent to the tube sheets to aid in maintaining circulation and

facilitate leak detection in this area. Injection devices at the upper end of

the helical-tube unit will be used for the injection of either steam or hydrogen for calibrating hydrogen-leak detectors. The reference detector for

the SNR program is a nickel diffusion tube with an ion pump. The licensing

authorities want the SNR designers to have the signals from the hydrogen

detectors go directly to the reactor safety system.

Dutch tests at the Apeldoorn Laboratory indicate that a whole

spectrum of elements are coming through the 1/4-mm-thick Type 316

stainless steel diffusion tube operated at 300*C, and they are investigating

this phenomenon. The spectrum includes hydrocarbons. The Dutch also

indicate that certain elements, such as sodium peroxide, formed at high

oxygen contents can severely corrode the nickel membrane.

The TNO people at Hengelo have found significant variations

in hydrogen in the sodium using their in-sodium hydrogen detectors, as

described in detail in Section IX.
92

VII. LMFBR DEMONSTRATION PLANTS

A. Summary and Comparison

Table 10 summarizes important characteristics of Phenix, PFR, and

SNR; Figs. 41-43 are flow diagrams for these systems. Phenix and PFR

employ sodium-reheat systems that enable them to use standard turbinegenerator sets. The earlier SNR reference design involved sodium reheat,

but this approach has been replaced by a steam-reheat cycle in which steam

exiting from the turbine high-pressure cylinder is used in reheating steam

from the intermediate-pressure exhaust, which, when reheated, supplies

the low-pressure turbine units.

TABLE 10. Summary of Demonstration-plant Design Conditions

Phenix PFR SNR

Number of loops

Power, MW

Thermal

Gross electrical

Sodium temperatures, C-*F

Reactor inlet

Reactor outlet

Secondary hot leg

Secondary cold leg

Total sodium flow rates,

kg/sec-10 6 lb/hr

Primary

Secondary

Steam cycle

Steam temperature, C-*F

Throttle

IP cylinder inlet

Steam pressure, bars-psig

Throttle

IP cylinder inlet

Steam generator

Type

Design

Recirculation rate

Gas blanket in unit

Number per loop

Evaporator

Superheater

Reheater

Water/steam temperature, *C-*F

Feedwater inlet

Evaporator exit

Superheater exit

Reheater exit

Sodium temperature, C-*F

Superheater/reheater inlet

Superheater/reheater exit

Evaporator exit

Three Three

563

233

672

305

400-752

560-1040

550-1022

350-662

2760-21.9

2210-17.5

Sodium reheat

512-954

512-954

163-2345

33-470

O-T

(1)

No

12

12

12

246-475

375-707

512-954

512-954

550-1022

475-887

350-662

425-797

585-1085

560-1040

374-705

3230-25.5

2860-22.6

Sodium reheat

537-1000

537-1000

159-2300

30-435

F.R.

(2)

4.62:1

Yes

One

One

One

288-550

356-674

538-1000

538-1000

560-1040

523-873

374-705

Three

736

312

377-711

546-1015

520-968

335-635

3400-26.9

3140-24.9

Steam reheat

500-932

305-581

167-2400

40-565

O-T

(3)

No

Three

Three

253-487

360-680

500-932

520-968

335-635

NOTE- O-T = Once-through.

F.R. = Forced Recirculation.

(1) Module = S-shaped shell and seven S-shaped tubes.

(2) U-tube.

(3) Straight tube in two loops. Helical coil in one loop.
-{] SURGE TANK

955*F

lj 586 F _ _ _ _ _ _ _ _

*qHP

BYPASS A

I - - - I.

6 + DEAERATOR

FEEDWATER

TRAIN

- REHEATER

- SUPERHEATER

- EVAPORATOR

- HIGH-PRESSURE TURBINE

SODIUM

IP - INTERMEDIATE-PRESSURE TURBINE

LP - LOW-PRESSURE TURBINE

G - GENERATOR

C - CONDENSER

NOTE: REACTOR TEMPERATURE, 1050/750 F

STEAM AND FEEDWATER

Fig. 41. Flow Diagram for Phenix System NoJ

TO

IHX

530 psig

955 F 1

S.

TURBINE

BYPASS

H P IP LP G

wpw 245 0 psig

I L

f E.

448*F

R

S

E

I-P

IN - -lir Ms

r

I

6 71' F

2573 psig

MOISTURE

SEPARATOR

709'F

R.
H

1040"F

TO

IHX

- -jr

705 F

TURBINE BYPASS

DRUM

2600 psig

- 550F FEEDWATER

-4

- 1

FCP

C

H

2

6 LINES

6 LINES

STEAM FLOW TO HIGH-PRESSURE TURBINE 1.983 x 106 

lb/hr

STEAM FLOW TO REHEATER 1.361 x 106 

lb/hr

WATER CIRCULATION RATE 9.22 x 106 lb/hr -5x

1000 F 1000 F 435 psig

FED0 psig

IP IP G

O L.P.PUM580PF

O 3 480 psig

w 

C

C EW

FEED WATER

PUMPS AND

LEGEND: HEATERS

SODIUM

STEAM AND FEED WATER

- - STEAM AND FEEDWATER HEADER

VALVE

FCP FORCED-CIRCULATION PUMP

NOTES: 1. PFR IS LOAD FOLLOWING; STEAM PRESSURE

AND TEMPERATURE = CONSTANT

2. SODIUM CIRCUIT SHOWN IS ONE OF

THREE CIRCUITS

3. STEAM GENERATOR - FORCED CIRCULATION

4. REACTOR = 1085/797 F

Fig. 42. Flow Diagram for PFR System

4p

ro

I
520"C968* HDV STEAM REHEATER

#2 3750 t/h 165 ata

30 MWt L SUPERHEATERS 495C

105 FC3/LOOP OP923 F

4060 t/h 

193t/h

O-i- IP IP LP LP-

#3 LOOP IHX

REACTOR 3 I EVAPORATORS 66 t/h NKL = 312 MWe

711 F 3/LOOP36C

80 MWt = CORE IN EINII STATE

762 MWt = POSSIBLE POWER OF A FOLLOWUP COREt/h

(DESIGN OF HEAT TRANSFER SYSTEM)3 C

ata = ATMOSPHERES; 1 ata = 14.7 psi 8

t/h = tonnes/hr; 1 tornne 2200 lb

HP = HIGH-PRESSURE TURBINE

LP = LOW-PRESSURE TURBINE

IP = INTERMEDIATE-PRESSURE TURBINE 1259 t/h

180*C

3566"

Fig. 43. Flow Diagram for SNR System

U
96

INTERATOM indicated that the change in SNR reference design was

made primarily for economic reasons. U.K. (on the basis of engineering

to date) and the French (on the basis of system tests at Les Renardi'eres)

feel that the Phenix and PFR sodium-reheat systems will operate in a

satisfactory manner. An SNR-type steam-reheat system has not been

tested; however, the Germans feel that this system will operate satisfactorily. They are confident that SNR could operate satisfactorily with

sodium reheat, based on tests at the Hengelo facility. SNR will use a turbine consisting of standard cylinders that have never been employed in a

single turbine assembly before. SNR will have a steam reheater similar

to those used in LWR plants.

The French and Germans have had extensive, successful experience

with once-through steam generators and will use this concept in Phenix and

SNR. The British have had limited experience with once-through units, but

considerable experience with recirculation units. In the interest of conservatism, PFR will have forced-recirculation steam generators.

The British and French construction programs are moving well

toward completion. SNR construction was expected to begin in the fall of

1972. PFR sodium fill was scheduled for March 1973, and criticality was

expected by June 1973. The target date for Phenix sodium fill, as of the

team visit, was September 1972. Criticality was expected by March 1973,

and power operation by summer 1973.

B. France--Phenix

1. Description

Phenix, a 250-MWe sodium-cooled fast-neutron prototype reactor, located at the French Nuclear Production Center at Marcoule, about

35 km north of Avignon, was visited by the team on June 5.

The reactor is located on the bank of the Rhone, which supplies

cooling water. The fuel-handling building, reactor building, steam-generator

building, and turbine hall are located from north to south along the riverbank. The reactor building is of conventional design, constructed to withstand an internal pressure of 40 mbar. Table 10 summarizes important

features of Phenix; Fig. 41 is a heat-transfer and steam-system flow diagram. Structural and facilities design, reactor and fuel-element design,

and fuel handling are described in Nuclear Engineering International,

July 1971 (pp. 557-580).

Phenix is a pool-type reactor, the main tank containing, in

addition to core, blanket, and shield, three primary loops with three main

sodium pumps, six intermediate heat exchangers (IHX's), and 850 tons of
97

sodiumunder a controlled argon atmosphere. The primary tank (11.5-m diameter) is enclosed within two other vessels (a stainless steel leak tank and

a carbon steel containment tank), the spaces in between being charged with

nitrogen. The containment tank, essentially a cavity liner, is cooled on its

outer surface with approximately 3/4-in.-ID water pipes horizontally coiled

around and welded to the vessel.

The secondary coolant circuit consists of three independent

loops. Each loop, with its main pump located at the high point, feeds a

once-through steam-generator unit, with evaporator, superheater, and reheater sections. There are valves in each of the three loops to isolate the

steam generator and to dump the sodium, water, and steam in the event of

a sodium-water reaction. Each of the three loops consists of 12 identical

15.7-MWt units. Each of the 12 units consists of 12 evaporators, 12 superheaters, and 12 sodium reheaters.

In the event of a failure in one secondary loop, it is planned to

use IHX isolation valves to isolate the secondary side of the two IHX's in

the affected loop. In addition, sodium flow through the primary side of the

two IHX's would be interrupted by mechanically moving a sleeve to cover

the IHX sodium-intake area. The reactor would then be operated on two

loops while repairs are made. It is expected that a steam-generator module

can be removed and replaced in 10 days to 2 weeks. In the event of a leak

in the steam generator, only that portion of the module containing the leak,

i.e., evaporator, superheater, or reheater, would be replaced. Withdrawal

space is provided in the steam-generator building.

Water-reaction tests at Cadarache were made on a Phenix

prototype test unit consisting of one full-scale module of seven tubes. Each

test consisted of rupturing one tube out of the seven. The Phenix design is

based on the rupture of only one tube. These tests on the Phenix module at

Cadarache indicated that peak pressure at the IHX may reach 100-150 bars.

In Phenix there is a spherical expansion tank with an argon-gas

cover on the discharge side of the steam generator. This tank contains the

secondary sodium pump. Also, the secondary sodium system contains a

"buffer" or surge tank at the sodium inlet to the steam generator, with an

argon cover gas. The buffer tank was added after sodium-water reaction

tests indicated that it should be included to reduce the secondary-system

peak pressure resulting from a sodium-water reaction in the steam generator. Further tests verified the effectiveness of the buffer tank (see

Fig. 44).

The flow-dividing valves for Phenix are not power-operated

control valves, which would be required to change position during plant

operation. They are trim valves, set once to properly divide the sodium

flow between the reheater and the superheater.
98

SECONDARY

PUMP

EXPANSION 550 C BUFFER TANK

TANK C0

A

rFLOW-DIVIDING VALVE

REHEATER 

RUPTURE

DISKS

SUPERHEATER |

ATMOSPHERE

EVAPORATOR |

A

IHX IHX

MtXTNG TEE B SEPARATOR

g SODIUM-WATER

BLOWDOWN TANK

A. Sodium Isolation Valves (45-m 3 CAPACITY)

B. Sodium Dump Valves DUMP TANK

Fig. 44. Portion of Phenix Secondary Sodium System, Showing Some

Features to Minimize Effects of Sodium-Water Reaction

The steam generator is isolated by valves designed to close in

about 6 sec upon the sensing of a severe sodium-water reaction in the

steam generator. The closure of these valves limits the sodium-water

reaction residual impurities in the secondary sodium circuit and prevents

such impurities from entering the IHX in undue amounts.

The time required to drain a secondary sodium loop for the

Phenix plant is about 30 sec.

In addition to the buffer tank, protection against sodium-water

reaction pressures is provided by (1) a system of seven rupture disks per

steam generator (mounted two in tandem at each of seven points to protect

against inadvertent tube failure), discharging to a separator; (2) a sodium

dump tank to receive secondary sodium from steam generators; (3) quickclosing sodium valves to isolate the steam generator; (4) a hydrogendetection system for each steam generator, fitted with a sodium-selector

valve, which can sample any one of 12 modules or take a mixed sample

from all simultaneously. Phenix does not use the advanced rupture disk

being developed at Les Renardieres and Saclay. The French are using

Black, Sivalls, and Bryson (BSB) disks, which have been working satisfactorily. Six of these 150-mm-dia disks are in each steam generator,

and a seventh is at the surge tank.

The reference design for leak detection is the hydrogen-diffusion

detector tube in sodium. Vacuum ion pumps are used in conjunction with a

0.3-mm-thick nickel membrane. A mass-spectrometer system had previously been used. The vacuum ion pumps with the nickel membrane were

yet to be installed at Phenix. They were being tested elsewhere.
99

2. Status of Construction

The manpower employed on the project is supplied by the CEA,

the EDF, and an industrial group (The Groupement Atomique Alsacienne

Atlantique--GAAA). There were said to be about 700 workers onsite,

mostly on a 54-hr workweek.

All principal primary components, including pumps, IHX's, and

the rotating plug, were installed in the tank. Two of three pump motors

are in place. Main secondary sodium piping appeared complete, with one

circuit completely insulated, one partly insulated, and one bare, but with

heater and leak detection in place.

Two of the three 12-module steam generators were in place

and connected; 11 of 12 modules of the third steam generator were in place,

with connecting piping being welded. The 12th module was on the floor ready

to go in. Secondary pumps and buffer tanks were in place, connected, and

insulated. Main steam and reheat piping was complete to the turbine building, but not yet insulated. In the turbine building, the final fit-up of steam

piping and valves was in progress. The high-pressure turbine was in place;

medium- and low-pressure rotors were in place, with work in progress on

alignment, couplings, and bearings. The generator was in place.

Additional obs ervations showing the following construction

status:

a. Much electrical-cabling work remained to be done, including

some main-cable pulling and extensive hookup.

b. Control-room consoles were in place, with connecting up

in progress.

c. Sodium storage tanks, blowoff tanks, and cold traps appeared completed.

d. Fuel-handling system and equipment were in the final

stages of installation. Irradiated-fuel interim- examination cells were

structurally complete, but lacked windows, manipulators, and other

equipment.

e. Structural cells for component storage, cleaning, and repair appeared finished, except for viewing windows, through-wall manipulators, and other outfitting.

f. The pump and IHX removal coffin was on hand in the fuelhandling building. (The same coffin is used for both components.)

g. Neutron-measuring chamber-insertion devices, located

under the outer reactor containment, were in place.
100

h. Hydrogen-detector piping from each ,steam-generator

module appeared complete, but the detectors themselves and the ion pumps

were not installed. Detectors were still on test.

i. The three secondary sodium storage tanks contained sodium (150 tons per secondary loop).

The latest schedule showed that sodium fill was to take place

in September 1972, criticality by March 1973, and power operation by

summer 1973. The observed status of construction supported these dates.

Note: Phenix as of January 1973:

a. Sodium filling completed.

b. Cold- and hot-flow tests scheduled first half of 1973.

c. Criticality expected in June 1973.

C. U.K.--PFR

Until 1968, the responsibility for the design and construction of

PFR, together with the associated R&D, rested with the UKAEA, although

British industrial companies, many of whom are members of the British

Nuclear Forum, played a significant role. With the reorganization of the

nuclear industry, responsibility for the design and construction of all reactors passed to two nuclear consortia: TNPG and BNDC. TNPG was

subsequently awarded the contract to complete PFR, with BNDC member

companies contributing a significant fraction of the components.

1. Plant Description

Important features of PFR are summarized in Table 11, and a

heat-transfer and steam-system flow diagram is pictured in Fig. 42. The

PFR incorporates a single-tank design in which the entire primary circuit

is contained in a 12.2-m-dia, 15.2-m-deep cylindrical vessel (40 x 50 ft).

This tank, its leak jacket, and all its components are suspended in a concrete-vault from a roof structure. All connections to other parts of the

circuit pass through this structure, eliminating any penetrations of the

main tank. The hexagonal core and breeder assemblies are located in a

honeycomb pattern on the diagrid, which is suspended from the roof. Outside the shield and extending to the top of the tank is the reactor jacket,

which has in its upper part six pods containing the IHX's in three pairs.

The reactor jacket directs coolant flow and separates the hot sodium

emerging from the core from the cooler sodium in the outer part of the

tank. Between the pairs of IHX's, but outside the jacket, are the three

primary-sodium centrifugal pumps, which have outlet pipes leading to the

underside of the diagrid. Their drive motors are mounted above the roof.
101

TABLE 11. Summary of PFR Parameters

Max MWt, core

radial breeder

axial breeder

Gross MWe

Net MWe, max

design

Station net efficiency, 

Reactor inlet, *C-*F

Reactor outlet, *C-*F

Secondary-sodium IHX out, *C-F

IHX in, *C -*F

Primary flow, kg/sec-106 lb/hr

Secondary coolant flow, kg/sec-106 lb/hr

Turbine

Steam pressure, bars -psig

Steam temperature, *C -*F

IP inlet pressure, bars-psig

IP inlet temperature, *C-*F

Steam generator

No. of steam generators

MWt each

Feedwater, *C-F

Forced-recirculation rate

Superheater sodium-outlet, *C-*F

Pumps

Flow per pump, gpm

Delivery head, psi

Motor speed, rpm

Motor bhp

Output speed from fluid coupling to pump

shaft, rpm

Working temperature, *C-*F

Fuel transfer pot

Max heat dissipation, kWt

Bucket surface temperature, *C -*F

Primary circuit

Diameter of primary tank, ft

Weight of sodium in tank, tons

Cover gas

Operating pressure of cover gas, psig

Primary

18,500

117

960

2200

616

34

22

305

290

254

42.3

425-797

585-1085

560-1040

374-705

3230-25.5

2860-22.6

159-2300

537-1000

30-435

537-1000

3

224

288-550

4.62:1

468-873

Secondary

16,500

64

990

895

200-960 200-990

400-752 370-700

15

580-1076

40

1000

Argon

1
102

TABLE 11 (Contd.)

Secondary circuit

No. of circuits

Weight of sodium per circuit, tons

Cover gas

Fuel and breeder subassemblies and core

No. of fuel pins per subassembly

Outside diameter of fuel pin, in.

Fuel-cladding thickness, in.

No. of subassemblies- in core, inner zone

outer zone

No. of radial breeder subassemblies

No. of radial breeder reflector subassemblies

Length of fuel in pin, in.

Length of upper axial breeder, in.

Length of lower axial breeder, in.

Overall length of fuel pin, in.

Total length of subassembly, ft

Lattice pitch (triangular), in.

Weight of mixed-oxide fuel in core, tonnes

Weight of PuO2 in core, tonne

Max subassembly rating, MWt

Max center-pin can temperature, *C- F

Design burnup, % max

Residence time, days

Irradiation period, days

Fuel-transfer period, days

Decay period, days

Rods

No. of control-rod positions, tantalum

No. of shutoff-rod positions, boron

No. of safety-rod positions, boron

3

85

Argon

325

0.23

0.015

30

48

52

82

35

16

4

90

12.5

5.70

4.0

0.902

10

650-1200

71

238

47

2

30

5

5

1

The sodium coolant fills the main tank and is in constant circulation. The coolant flows upwards through the core and breeder and over

the inner lips of the pods into the IHX's, where it gives up the heat gained

in the core to a separate, secondary sodium circuit. From these IHX's, it

flows over the outer lips of the pods and into the pump intakes. It is then

driven through the pumps to repeat the cycle.

The secondary circuit consists of three separate loops, each

loop being connected to two IHX's. After passing through an isolating valve,

the hot sodium flow is shared between superheater and reheater sections of

the steam generators and then recombined to pass through the evaporator

section. From the evaporator, it is returned to the IHX by a secondary

mechanical pump through expansion tanks and a further isolating valve.
103

Each steam-generator unit consists of a separate superheater,

reheater, and evaporator. The design and operation of this unit are discussed in Section IV.C. The general arrangement is similar in all nineunits,

though there are dimensional differences. The hot sodium enters at the

bottom and flows upwards through a central duct. At the top of the duct,

the sodium flows down an inner annulus, which contains the return legs of

the U-tubes. The sodium then emerges through ports into an outlet header.

Annular inlet and outlet headers at the top of the unit deliver and collect

the steam, which then passes to the turbine. TNPG stated that a forcedrecirculation-type steam generator was adopted because this was considered

to be the most conservative approach. Most boilers in U.K. power plants

are of the recirculation type, and several years ago--when the design choice

was made--the British felt there were too many areas of uncertainty in

once-through steam generators.

The choice of a sodium-reheat cycle permits the use of a conventional 300-MWe turbine set at PFR. It is a 3000-rpm tandem machine

with one high-pressure, one intermediate-pressure, and two double-flow,

low-pressure cylinders. Steam conditions at the stop valve are 2300 psi

and 538*C (1000*F) with single reheat to 516C (960F).

Several features specific to the steam-generator system and

its operations are discussed in Section VII.C.3.

2. Status of Construction

PFR construction was nearing completion at the time of the

team visit. It was estimated that the system would be filled with sodium

by March 1973, and criticality was predicted by June 1973. Some civil works

were still in progress. The main reason that the PFR is behind schedule

is late delivery of components. As of May 25, five out of the six IHX's had

been installed and two out of the three primary pumps installed with the

third pump were being assembled. The evaporator tube bundles were installed. The superheater and reheater tube bundles were to be delivered

in June 1972. These units were completed, but were being kept at the

manufacturer's plant until the last possible moment to prevent the possibility of chloride stress corrosion of the stainless steel during storage at

the site.

Reheater and IHX tube bundles have been in fabrication 22 years,

from a firm-design point. Tube-to-tube-sheet weld development caused

considerable delay. TNPG representatives indicated that mounting the top

shield on a structure above the floor and checking'out assembly fit-ups and

circuitry in this position had been very advantageous. The previous plan

had called for checkout to occur after the shield was positioned in the roof.

Delay in roof delivery caused the British to take the approach indicated

above, and a number of fixes became necessary during checkout. It would

have been difficult and time-consuming to have accomplished the same

work with the shield in place in the roof.
104

The U.K. has abandoned the PFR primary-system hydraulic

water test because of the fear that they could not remove the water that

entered defects in the honeycomb insulation inside the mounting. However,

they were going ahead with a waterflow test of one of the secondary loops.*

The information to be obtained from this waterflow test will cover the

areas of vibration, liquid-level changes, gas entrainment, and pump performance. The British felt they have assurance that gas entrainment will

not occur in the primary circuit as a result of the model testing they have

performed. The steam system will be tested and commissioned using an

auxiliary boiler.

3. Steam-generator System

Several auxiliary features (valves, rupture disks, and leakdetection devices) of the steam generator deserve special attention.

Bypass lines with valves have been provided on both the sodium

and steam sides of the reheater. The primary function of these valves is to

facilitate the matching of main and reheat steam temperatures to turbine

rotor temperatures during startup and shutdown. The sodium bypass had

been added recently and consists of two 6-in. lines with a double-valve

arrangement in each line (see Fig. 12). The bypass line runs directly

from the cold-leg pump discharge to the reheater inlet line. The size was

selected as 6 in. because a 6-in. ball valve was available. These valves

are used either fully open or fully closed. The sodium lines from the reheater and superheater to the evaporator are orificed to ensure good flow

distribution between these units.

A steam bypass line around the reheater has a 50% flow capacity. This bypass ensures that exit steam from the high-pressure cylinder

has at least 50 F superheat before being admitted to the reheater. A valving

arrangement downstream of the bypass helps keep up the steam pressure in

the reheater (see Fig. 12). This valve consists of a main valve in the line

and a "pressure-sustaining" bypass valve. The latter is a proprietary

needle valve used for up to 15% flow and aids in keeping the pressure on

the steam side of the reheater higher than the pressure on the sodium side.

The differential pressure between the steam and sodium sides is monitored

and, when the AP drops to 20 psi, the steam side is backfilled with argon

(during shutdown operations).

TNPG and CEGB feel that satisfactory modes of PFR startup

and shutdown have been established. The startup (up to about 20% power)

would be manual and not by the use of a computer. Once the system startup

characteristics are better known, the system can be programmed

automatically.

*As of January 1973, this test was completed.
105

The steam-generator rupture disks incorporate shear pins,

which determine the burst pressure. The shear pin, which operates on a

hinged member, can be replaced. A safety-lock mechanism is used during

replacement of the shear pin. The design of the rupture-disk arrangement

is not proprietary. The bursting disks in the relief system have cross

wires, which rupture and initiate an automatic steam dump. In addition to

the main relief system, a smaller relief system is provided for intermediate

leaks. This system has a 1.5-in. line and relieves to the argon stack. This

relief line is opened by pressure transducers in the argon-gas space.

Safety valves permit blowdown of both drum and evaporator in the event of

a large tube leak.

Small tube leaks are detected by measuring the hydrogen concentration formed by the reaction between steam or water and sodium.

The hydrogen is allowed to diffuse through a heated nickel membrane (in

the form of a coiled tube) into an argon-gas purge flow; a katharometer

(a standard instrument that measures changes in thermal conductivity)

then measures the concentration of hydrogen in argon. The nickel membranes are included in all blanket-gas spaces in the steam-generator units

and in one position below the bulk sodium level at the evaporator exit. On

the basis of rig tests, these systems are expected in most cases to detect

leaks with sizes well below the level of concern. The detectors in the

cover gas have three katharometers (for a two-out-of-three indication),

and the under-sodium detector has four diffusion tubes, one of which is a

spare. The U.K. intends to calibrate the PFR hydrogen-detection system

in situ. Water will be injected in some locations, and hydrogen in others.

They have performed response-time and sensitivity tests of hydrogen detectors by injecting water and/or hydrogen.

D. SNR

The design, development, construction, and operation of SNR are

being undertaken as an international venture involving Germany, Belgium,

and Netherlands. Cooperation occurs at four levels- -government, national

laboratories, industry, and utilities. The general project organization is

pictured in Fig. 45. The industrial consortium consisting of INTERATOM,

PROJECT COMMITTEE

RMANY - BELGIUM - THE NETHERLANDS - LUXEMBOURG

GOVERNMENT NATIONAL LABS INDUSTRY UTILITIES

BASIS PROGRAM INDUSTRIAL PROGRAM

KARLSRUHE (G) INTERATOM (SIEMENSI (G) R W E (G)

MOL (8) BELGONUCLEAIRE (B) SYNATOM (B) PETTEN IN) NERATOOM (N) S E P (N)

BELGONUCLEAJRE B)

T N 0 N S N R-300 LUXATOM (L) DESIGN ORDER

REACTOR PHYSICS CONSTRUCTOR OPERATION

FUEL ELEMENT DEVELOPMENT

REACTOR SAFETY HEAVY NA COMPONENTS

ADVANCED CONCEPTS TEST FACILITIES

S N R- 1000

Fig. 45

Organization of the SNR Project

(Nuclear News, March 1972)
106

Belgonucleaire, and NERATOOM has the responsibility for design and construction of SNR. The research centers of Karlsruhe, Mol, TNO, and

Petten support the industry effort and contribute basic research. The reactor owner and operator SBK (Schnell Brueter Kernkraftgesellschaft),

comprised of three utilities--RWE of Germany, Synatom of Belgium, and

SEP of Netherlands--will place the order for SNR and will be the operator.

At the time of the visit, SNR construction was expected to begin in the fall

of 1972, and the start of power operation was scheduled for 1978.*

Table 10 lists important operating characteristics of SNR, and

Fig. 43 is a general flow diagram for the system. The SNR is a loop-type

reactor with three main primary and secondary circuits and a rotatingplug refueling system. All free-sodium surfaces in the reactor vessel, as

well as in the pumps and tanks of the primary cooling system, are covered

with a protective cushion of argon at a slight positive pressure. The secondary cooling system also has an argon cover gas, which is maintained at

a higher pressure than that in the primary cooling system because the

primary pump, being in the hot leg, would otherwise produce a higher pressure on the shell side of the IHX tubes and could cause leakage of radioactivity into the secondary system.

Heat produced in the reactor core is transferred from the doublewalled reactor vessel through three main primary coolant circuits to the

IHX's. The core inlet and outlet temperatures are 380 and 550*C, respectively. The IHX's are straight-tube units designed to permit the

identification of leaking bundles without dismantling and cleaning the entire

IHX. Because of cavitation considerations, the primary pumps are located

in the hot leg. The primary piping is arranged above a minimum admissible

sodium level of the reactor tank, preventing the possibility of siphoning and

subsequent loss of forced-convection core cooling after a major pipe rupture. For the same purpose, the primary pump and the IHX's are surrounded

by steel guard tanks. Shutoff valves in the inlet and outlet pipes allow a

separation from the reactor vessel whenever repair and maintenance of

primary components are necessary. These valves, however, have no safety

purpose.

The SNR will be equipped with six emergency heat exchangers inserted into the reactor vessel close to the radial blanket of the core so that

their heat-transfer surface is located lower than the outlet nozzles of the

primary circuits.

Three main secondary loops, with IHX inlet and outlet temperatures

of 340 and 525*C, transfer the heat to the steam generators. In contrast to

the primary circuits, the three main secondary pumps are located in the

cold leg. The steam generator consists of three evaporator modules and

*As of January 1973, the construction contract had been signed. Government contributions had yet to be

finalized.
107

three superheater modules per loop. The steam-generator design (developed by the Dutch) consists of a straight-tube unit that incorporates a

bellows in the shell to accommodate differential thermal expansion. (Dimensions and other pertinent data on the steam generator are presented in Section IV.D.) Two loops will use units of this design; the third loop will have

steam generators with helical-tube design. In this alternative steamgenerator design, the tube sheets have been eliminated. The individual

tubes penetrate the shell of the steam generator at the lower and upper

heads.

The SNR steam generators are of the once-through type. The oncethrough arrangement is preferred over the recirculation steam-generator

arrangement primarily because of economics. Most conventional steam

generators in Germany are of the once-through type, and experience has

been satisfactory. Corrosion and DNB-dryout effects have been examined

and are not considered a problem. The SNR steam generator will be operated with 95% steam quality at the superheater inlet at 100% load and

83% quality at low loads. A separator (with constant 5% drain flow) will be

used between the evaporator and superheater sections (a single separator

per loop). This approach was adopted to better control steam temperatures

between the evaporator and superheater sections.

The live steam leaves the superheaters at 500*C and 170 atm and is

conveyed to the turbine hall through three live-steam mains. After passing

the steam through a mixing section, the mains distribute it to the four turbine inlet valves. Steam-reheater units are located between the intermediateand low-pressure sections of the turbine. The steam is expanded in a

300-MWe single-shaft condensation turbine with a directly coupled, threephase ac generator, and then condensed in the condenser, which is cooled

by river water.

The live-steam conditions attainable in the plant permit the use of

conventional turbine parts. The turbine is made up of a single-flow, highpressure part and a double-flow, intermediate- and low-pressure parts.

Steam is resuperheated in a steam reheater located between the intermediateand low-pressure turbine sections. Steam is supplied from the exhaust of

the high-pressure turbine. Two steam reheaters are envisaged to be located

vertically alongside the turbine. This turbine is designed for stationary

high-pressure operation and has nozzle-group regulation.

The feedwater is preheated in seven stages up to about 246C by

means of steam bleeds from the turbine and, after further preheating to

about 253*C by the waste condensate from the separator between the evaporator and the superheater, is fed into the steam-generating unit.

In the event of turbine trip, the Germans follow a procedure that is

normal practice in fossil-fueled plants. The main steam supply bypasses
108

the turbine and is throttled to a lower pressure. The steam at reduced

pressure then enters the reheat steam line and returns to the reheater.

Thus, the reheater sees essentially a constant steam temperature at its

inlet, and steam pressure remains above sodium pressure following trip.

The steam flow exiting from the reheater then is dumped to the condenser.

(There is 100% steam-dump capacity in the system.) With turbine trip, the

reheater can experience a transient of 70*C in 10 sec. Thermal shields

placed between the active sodium stream and the lower tube plate provide

adequate protection against excessive stresses resulting from temperature

transients.

The steam generators of each circuit are housed in a separate

building, in an air environment. INTERATOM indicated that the German

licensing authorities are considering the need for use of an inert atmosphere

in place of air. Because of safety considerations, the SNR plant would be

shut down if one loop is out of service. Emergency cooling can also be

maintained by natural convection through one of the three main coolant circuits or through the backup emergency heat exchangers in the reactor vessel

through which the decay heat is transferred to air coolers. The use of the

emergency heat exchangers installed in the reactor vessel is avoided, except

in extreme emergency conditions, because of the adverse system temperatures that would result when they are in service.

A decay-heat removal system is provided for each steam-generating

unit. The system includes two water-storage tanks (300 tons total) and

four emergency feedwater pumps located in the turbine building. Steamgenerator flow is dumped to a decay-heat condenser and then returned to

the feedwater stream. To ensure the decay-heat removal, four emergency

feed pumps are provided. Three are directly connected, one to each steamgenerating unit, and the fourth is a reserve unit that can be connected as

necessary to each of the three steam generators. The pumps are electrically driven through hydraulic regulating couplings and can also be operated

from the emergency power supply. Pressures are maintained at 165 atm

for as long as possible. Each loop has a decay-heat condenser with 3% of

total capacity. For the first 45 hr, decay-heat removal requires 100 tons

of water.

Valves of the butterfly type are used in the secondary sodium system to isolate the steam generators in the event of a major sodium-water

reaction. Such isolation minimizes the introduction of sodium from the hot

leg into the steam-generator area, where entry of the sodium might promote

further reactions with steam or water after the initial expulsion of sodium.

The second function is to reduce the possibility of transporting sodiumwater reaction products into the secondary system.

Gate-type trim valves, identical to valves in service in the sodiumcooled KNK reactor, are located on the sodium supply lines to the individual
109

evaporator modules. These valves balance the sodium flow between the

three evaporators at full-flow conditions. Once set, the valves remained

fixed in position. Trim valves are not incorporated in the individual superheater, since temperature swings resulting from flow maldistributions are

acceptable.

INTERATOM does not provide a gas space over the sodium in the

SNR steam-generator modules. Cover gas is maintained in the cold-leg

pump and in a hot-leg expansion (surge) tank of the secondary loop. Experiments and analytical studies have indicated that the presence of a

cover-gas space in the steam-generator units could result in very high

pressures due to sodium hammer effects if sodium-water reactions occur

in the steam generator. The gas space contributes to a delay in depressurization of the system through a delay in bursting of a rupture disk and

results in higher reaction-zone temperatures and transmission of stronger

pressure pulses to adjoining modules, the secondary system, and the IHX.

Without a cover-gas space, rapid depressurization occurs in the event of

a tube leak, and the design can therefore be based on the guillotine failure

of a single tube.

Each evaporator and each superheater module has two sets of

double-rupture disks, one set at the upper and one set at the lower end of

the unit. Two cyclone separator units are associated with each secondary

loop. The cyclone separator can hold about 6 m3 of sodium, which is

greater than the volume of one module. The rupture relief lines are of

250-mm diameter, and lengths to the cyclone separator are about 20 m for

the line at the upper end of the modules and 40 m for the line at the lower

end. The rupture disks are 150 mm in diameter and 1.1 mm thick. The

estimated life of the rupture disk is approximately 6 years, taking account

of corrosion and age-hardening effects. The disks are in the shape of a

sector of a sphere with the convex side facing the sodium. Analytical

studies indicate that for disks of the present size and larger, the peak

pressure is about the same. The present rupture-disk diameter and

collector-pipe dimensions have been adopted as a tradeoff between peak

pressures and loadings on the relief piping.

The sodium operating pressure is 8 atm, and the rupture-disk burst

pressure is 17 atm. Following a tube leak, the water system would be blown

down, while sodium remains in the system for a period of time. Sodium

would not be dumped immediately after detection of the leak because of

concern over a possible secondary sodium-water reaction in the dump

tank. The waterside blowdown period is estimated at 30-40 sec following

automatic shutdown in response to disk-burst indication.
110

VIII. COMMERCIAL PLANTS

A. Summary and Comparison

France, U.K., and Germany have moved well into studies of commercial plants. France, Germany, and Italy have entered an agreement

under which two LMFBR plants will be constructed. The reference steam

cycle for these two plants involves steam reheat. Design effort is underway on the U.K. 1300-MWe Commercial Fast Reactor (CFR). The reference

design for this plant at present involves sodium reheat. The reference

1300-MWe CFR uses two standard 650-MWe turbine generators operating

in parallel. Concern has developed that this parallel arrangement may

cause operating difficulties. CEGB and TNPG are therefore currently

reevaluating the general area of future power-plant requirements to determine the size and operating conditions for which turbine-generator sets

should be standardized in the future.

Comparisons of sodium-reheat and steam-reheat arrangements for

commercial plants were studied by both CEA snd EDF. Preliminary studies

indicated that nonreheat was economically unacceptable, and this concept

was discarded at an early stage in their commercial-plant considerations.

In addition, the high moisture of the nonreheat turbine end point of 13- 14%

was a problem. Nonreheat turbines would have to operate at 1500 rpm,

costing 30-40% more than 3000-rpm turbines. Phenix-type steam generators were assumed in the CEA study; large steam generators designed by

Babcock-Atlantique were assumed in the EDF studies. Both studies indicated an economic advantage of steam reheat over sodium reheat amounting

to 2-4% of total plant cost. Both CEA and EDF stated that the "reliability"

advantages of steam reheat over sodium reheat, relative to having less

sodium- to-water/steam heat-transfer surface and providing easier plant

startup, were important in any evaluations made of steam cycles for commercial LMFBRs.

In Germany, INTERATOM studied a number of different steam

cycles for a 1000-MWe commercial plant. The Germans concluded that

steam reheat, which uses high-pressure cylinder exhaust to reheat exhaust

steam from the intermediate-pressure unit to the low-pressure units (a

steam separator is also employed in this arrangement, with two highpressure preheating stages), was most economical. This case offered an

advantage over sodium reheat of about 0.3% of total plant cost for an

assumed 1-mill/kWhr fuel-cycle costs for a 1000-MWe plant. INTERATOM

regards this cost differential to be significant. They also cited as a "slight

technical advantage" of steam reheat the simpler startup operation and

reduced sodium-to-water/steam heat-transfer surface area.

INTERATOM is now studying 1000- to 2000-MWe plants trying to

find where economies can be realized by additional development work. The

reference design is the steam-reheat cycle. When the next plant design is
111

undertaken, the economic parametric studies described above will be repeated to determine if they still hold true. The opinion was expressed that

the testing already completed at Hengelo provided grounds for believing

that sodium reheat could operate satisfactorily in a commercial plant.

CEGB did not wish to offer an opinion or preference as to sodium

reheat, steam reheat, or nonreheat. As a result of work done on PFR, they

feel they understand the sodium-reheat system and are confident of its

satisfactory performance in both PFR and commercial plants. Other systems have not been studied to the same level of detail as sodium reheat.

Startup operation of a sodium-reheat plant is more complicated than

desired, but with experience this operation could become routine.

B. France

The French are planning to start construction of a 1200-MWe prototype power plant in the mid-1970's. A cooperative agreement has been

established with Germany and Italy under which the first commercial-sized

plant would be constructed in Germany and a second in France.

Commercial-plant trade studies have been carried out by both CEA

and EDF. One CEA study was for the purpose of selecting a steamgenerator design for the 1200-MWe plant. (The steam-generator designs

considered are discussed in Section IV.B.) The study compared Phenixtype modules with large units designed by Stein Industrie and BabcockAtlantique. The economic results of the study favored the use of large

units over the modules. The decision to proceed with the large units for

the commercial plant was also based on the excellent experience obtained

in the steam-generator tests at Les Renardieres, the sodium-water tests

at Cadarache, and sodium-water analyses at Saclay and Cadarache, which

indicated that (1) sodium-water leaks could be detected at an early stage,

(2) the relief system was effective, and (3) the tubes and tube-to-tube-sheet

joints would be satisfactorily made for reliable service.

CEA conducted an additional study for the purpose of choosing a

steam cycle. Figure 46 compares sodium-reheat and steam-reheat parameters for a 780-MWe, 3000-rpm turbine used as reference in early economic

studies 

Stein Industrie conducted 1000-MWe studies for CEA, considering

the use of Phenix-type steam generators in four loops of about 550 MWt

each. There were 18 modules per Phenix-type enclosure per loop and

eight IHX's per plant. Steam-generator data for this plant are shown in

Table 12. Table 13 shows a weight comparison of the sodium-reheat and

the steam-reheat plants. Differential costs are shown in Table 14. The

economics showed a slight advantage for steam reheat. Though economics

was not a significant factor, this plus the reliability/availability advantages

favored steam reheat.
LO WPRESSURE

REHEATER TURBINE

HIGH- INTERMEDIATEPRESSURE PRESSURE TURBINE TURBINE CONDENSER

INTERMEDIATEPRESSURE

TURBINE LOW-PRESSURE

TURBINE

HIGHPRESSURE CONDENSER TURBINE

DRYER REHEATER

SODIUM REHEAT STEAM REHEAT

163 bars 510 C 2325 tonnes/hr Superheated steam 163 bars 510 C 2864 tonnes/hr

38 bars 510 C Reheated steam 4.42 bars 230 C

42.4 bars 318 C Steam to be reheated 4.82 bars 150.5 C

251 C Feedwater temperature 251.5 C

50 mbars (Cooling water 15*C) Condenser pressure 50 mbars (Cooling water 15 C)

1737 MWt Thermal power 1793 MWt

758.5 MWe Net electric output 756.5 MWe

43.67% Net efficiency 42.19%

Fig. 46. Sodium-reheat and Steam-reheat Parameters for 780-MWe, 3000-rpm Turbine

TABLE 12. Stein Industrie 1000-MWe Concept; Phenix-type

Steam-generator Data; Sodium Reheat vs Steam Reheat

Sodium Reheat Steam Reheat

Thermal power, MWt 2236 2304

Sodium inlet, *C 550 550

Sodium outlet, *C 350 350

Sodium flow, kg/sec 8841/3773 9040

Water inlet, *C 246 246

Steam outlet, *C 512 512

Steam outlet, bars abs 168 169.7

Waterflow, kg/sec 838 1020

Water steam (tp), no orifice, bars 19 28.2

Heat-transfer area, ID of tubes, m 2 Evaporator 3850 3850

Superheater 2100 2100

Reheater 2660 

TABLE 13. Stein Industrie 1000-MWe Concept

Using Phenix-type Steam Generators: Weight

Comparison of Sodium Reheat vs Steam Reheat

(in tonnes) (Steam-generator package only)

Sodium Reheat Steam Reheat

Modules 967 769

Headers 118 88

Support structure 1411 1050

Sodium piping 160 57

Water piping 249 283

Sodium-water relief system 224 185

3129 2432

112
113

TABLE 14. Differential Costs of Stein Industrie 1000-MWe LMFBR

Concept: Sodium-reheat Plant minus Steam-reheat Plant

106 Francs $106

Steam generator 30.5 6.5

Piping 14.33 3.0

Buildings 1.25 0.26

Differential cost -46.0 9.76

Turbine-generatora 8.5 1.8

Total differential cost 37.5 7.87

or

37.5 francs/kWe (capital cost)

Energy differential cost 18.3 francs/kWe gl ~$3.7/kWe

including steam reheaters.

As of July 1971, EDF had partially optimized sodium-reheat and

steam-reheat 1000-MWe LMFBR designs. The cost comparisons were

based on the vendor's detailed estimates for many items (Babcock-Atlantique

quotes on the steam generator, for example). The system parameters for

the sodium-reheat and steam-reheat cases compared are presented in

Table 15, and differential costs are summarized in Table 16. Heat-balance

diagrams for the two cases (power levels slightly different from conditions

given in Table 15) are presented in Figs. 47 and 48. Babcock-Atlantique

steam-generator designs are used in both cases. In, this design, the evaporater and superheater tubing is continuous and is in a single shell. The

superheater and evaporator on the sodium side are in parallel with the

sodium reheater for the sodium-reheat plant. In such a configuration, the

reheater is operating with a greater sodium-side A T than is optimum for a

plant-size configuration. However, this factor was not of major influence

in determining the results of the study.

TABLE 15. EDF 1000-MWe LMFBR Concept of Sodium

Reheat vs Steam Reheat: Steam-generator Data

Sodium Reheat Steam Reheat

Power (core), MWt 2250 2330

Power, MWe 962 962

Plant efficiency, % 42.75 41.25

Primary temperature, reactor

inlet/outlet, *C 400/560 400/560

Secondary temperature, steamgenerator outlet/inlet, *C 350/550 350/550

Feedwater temperature, *C 250 258

Feedwater pressure, atm 215 215

Steam temperature, *C 510 510

Steam pressure, atm 163 163

Reheat inlet temperature, *C 318 164

Reheat pressure (inlet), atm 42 6.9

Reheat outlet temperature, *C 510 285

Reheat pressure (outlet), atm 37.8 6.5

Steam flow, tonnes/hr 3023 3810
114

TABLE 16. EDF 1000-MWe LMFBR Concept; Differential Costs;

Sodium-reheat Plant minus Steam-reheat Plant

+106 Franc s

Building and site

Reactor

Fuel handling

Primary pumps

IHX

Auxiliaries

Secondary pumps

Sodium piping

Auxiliaries

Steam generators

Steam piping

Turbine

Steam reheat

Condensers

Feedwater heaters

Total

- 106 Francs

3.1

6.4

0.8

23.2

13.0

1.35

}

47.85

2.1

0.2

0.3

0.7

2.0

0.2

8.8

0.9

15.2

Col. 1 - Col. 2 = 32.65 x

for steam-reheat plant.

PRESSURE, TEMPERATURE,

bars *C

ENTHALPY, FLOW,

kJ/kg tonnes/hr r- -- 6.88 200

CONDENSER 3016.9 2291.5

0.050I

2381.2 1881.43I

- LP -

S2

1.811 143.8

2158.4 201.24

137.1 2423.3

I 10.85 58.4 .85 108.2

245 2423.3 451.3 24233

138. 423.3

S231.1 211820 .4

2400.55

HP - HIGH-PRESSURE TURBINE

IP - INTERMEDIATE-PRESSURE TURBINE

LP - LOW-PRESSURE TURBINES 13 DOUBLE FLOI

B - DEAERATOR

TPA - FEEDWATER PUMP TURBINE DRIVE

106 francs or about $6.9 million savings

37.6 510

3400 2001.5 REHEATER.

I 42 311.1BA

II 3004.9 2001.5 TL

I- ------ --- -- UNA

I I rAT

I I 163 510 3325.1 3141.5

IP - - HP

SUPERHEATER

S4I 1S5 IS6 EVAPORATOR

IIS3II L ---------

1.02$0.2 I - --- - ----

L. 13.5 364.5 22.05 433.4 421311.1 166.4 246.46 1.0 200.2 3110.5 111.55 3316.6 121.65 3004.61240 1013.4 3141.5

B TP

1.15 112.1 I 88.4 150.5 I 411.4 215.1

111.1 3141.5 111.4 111.4 W17.1 3141.5 921.1 3141.5

TIM.43111.5

114.4

13 .4 436.4

416.5 223.5

645.2 360.65 956.5 240

POWER OUTPUT - 1022 MWe

- 2331 MWt

W)

BCOCKLANT IQUE

IT

Fig. 47. EDF 1000-MWe LMFBR Concept: Heat-balance Diagram for Sodium Reheat
64.4 413.3

PRESSURE, TEMPERATURE, 516658.6

bars *C

ENTH ALPY, FLOW,

k J/kg tonnes /hr STEAM REHE ATER 8.08 184.2 ST EA M

2140.5 2541.8 fSEPA RA TOR 183 561

62.4 'C I2.83256871?33838.

CONDENSER

237 2055.7 03 322.217.61.2

LP HP a IP

883.2 1.6

131. 2541.3 

21

S1 S2 t5 215.2

028.6 141 1. 4S5

1POW.R2OUTPUT-.1 13 8.15. 2541..8 T.PCA S. 5.64

10..85 58..1 L8. 108..4

6.

5121B6.

POWER OUTPUT - 104

HP - HIGH-PRESSURE TURBINE

IP - INTERMEDIATE-PRESSURE TURBINE

LP - LOW-PRESSURE TURBINES 13 DOUBLE FLOW)

B - DEAERATOR

TPA - FEEDWATER PUMP TURBINE DRIVE

Fig. 48. EDF 1000-MWe LMFBR Concept: Heat-balance Diagram for Steam Reheat

BABCOCKATLANTI QUE

EVAPORATOR

B SUPERHEATER

?15.6 251.T

22.4 51.2

26 MWe

- 2410 MWt

U

. . v ... 
116

The basis for the system optimizations was to hold the reactor fuelcladding hot-spot temperature constant for the cases being compared. The

results indicated that, for the steam-reheat plant, the optimum reactor

sodium-outlet temperature was 540 C (1004 F) and the optimum steam

temperature was 470 C (888 F). (Evidence was not presented to indicate

that these temperatures are consistent with having a constant hot-spot'

temperature constraint.) For the sodium-reheat plant, the optimum temperatures appeared to be 533 C (991 F) for the reactor sodium outlet and

484 C (903 F) for the steam temperature. Thus, the smaller temperature

drives available in the sodium-reheat plant result in its required larger

heat-transfer surface area. The total cost differential is 2% in favor of

steam reheat. A preliminary evaluation indicates steam-reheat plant

availability should be 0.5-1% better than that predicted for sodium reheat.

After the team visit, EDF forwarded a later 100-MWe LMFBR plantoptimization comparison, which is being used as a guide to choosing the

parameters for the first commercial plant. The studies determined the

optimum value of parameters using the general criterion of energy-cost

minimization, which may be varied to minimize capital cost. Effects of

parameter changes on reliability and safety were not part of the study.

Optimization studies were done for the entire plant life. Fourteen independent variables were used. Table 17 indicates optimized values for these

TABLE 17. System Parameter's for EDF 1000-MWe LMFBR Optimization Studies

Sodium Reheat Steam Reheat

Power, MWt

Power, MWe

Net Plant Efficiency 

Sodium Conditions, *C-*F

Core inlet

Core outleta

Core ATa

Secondary Hota

Secondary Cold

Secondary ATa

Steam Conditions:

Turbine inlet pressure,a bar-psia

Turbine inlet temperature,a C-*F

Feedwater temperature,aoC-*F

Condenser pressure,a mb-in. Hg

Reheated steam temperature,a C-*F

Reheater steam pressure

AT between evaporator and reheater exita

Core

No. of pinsa

Pin diametera mm-in.

Pin pitch,a mm-in.

Core height,a cm-in.

alndicates the 14 independent variables used in the studies.

2556

966

37.8

399- 750

553-1027

154-277

540- 1004

310- 590

230-414

200-2900

483-901

183-361

92-2.7

245-493

N.A.

70,500

7.92-0.31

9.32-0.37

111-44

2570

964

37.5

390- 734

548-1018

158-284

536-997

309- 588

227-409

200-2900

- 470-878

190-374

80-2.4

256-493

N.A.

45-113

82,400

7.60-0.30

8.83-0.35

93-37
117

14 parameters (marked with the letter a) for both a sodium-reheat LMFBR

and a steam-reheat LMFBR. Table 18 indicates fixed input values to the

optimization studies. The reference 1000-MWe LMFBR (Phenix type,

excluding steam generators) used in the studies is based on GAAA

(Groupement Atomique Alsacienne Atlantique) concepts developed in conjunction with component and tur ine fabricators, and may not be fully

representative of the 1200-MWe LMFBR being planned. The turbine used

in the study consisted of five sections (HP, IP, and three LP sections).

TABLE 18. Fixed Design Values

for EDF Optimization Studies

Max cladding temperature, *C- F 670-1238

Load factor 0.69

Burnup, MWd/t 100,000

Fuel-fabrication cost, $/kg 300

Plutonium price, $/g 8

Total electric power, MWe 1000

Net electric power, MWe 965

Costs associated with the optimized designs are shown in Table 19.

TABLE 19. Capital Costs Associated with EDF 1000-MWe

LMFBR Optimization Studies

Sodium Reheat Steam Reheat

106 Francs $106 106 Francs $106

The total costs associated with each of these plants are:

Capital cost 1321 260 1276 251

First charge of fuel 209 42 199 40

Direct construction costs:

Civil works 105 21 100 20

Reactor maintenance and refueling 169 34 171 34

Sodium circuits, steam circuits, and

steam generators 265 53 236 47

Turbines and steam-plant equipment 109 22 113 23

Electric installations 96 19 93 19

744 149 713 143

C. United Kingdom

A reference design for the 1300-MWe Commercial Fast Reactor

(CFR) has been established, and the design effort is underway. At the time

of the visit, British plans called for the start of CFR construction in 1974,*

*Note: Issue of Atom 191, p. 161, Sept 1972, has a statement of Aug 8, 1972, by Davis, member of

U.K. Cabinet, that the. first full-size order is to be placed in the latter part of the 1970s.
118

with power operation scheduled to begin in 1979. The PFR concept is

suitable--with some modification--for use in the CFR.

The 1300-MWe CFR would require the use of two standard 650-MWe

turbine generators operating in parallel. Based on design work done to

date, CEGB has become concerned that such an arrangement may result

in operational complications. TNPG and CEGB are therefore currently

reevaluating the general area of future power-plant requirements to determine the size and operating conditions for which future turbine-generator

sets should be standardized and to identify turbine-generator development

requirements. In this study, operability and high availability are of greater

importance than small economic differences.

CEGB did not wish to offer an opinion or preference as to sodium

reheat versus steam reheat versus nonreheat. As a result of work done on

PFR, an understanding has been obtained of the sodium-reheat system and

there is confidence in its satisfactory performance for both PFR and commercial plants. Other systems have not been studied to the same level of

detail, and therefore there is little recognition of potential problems.

Though startup operation of a sodium-reheat plant is complicated, with

experience, operation with sodium reheat can become routine.

D. Germany

In 1971, INTERATOM made economic comparisons of various steam

cycles for SNR-300 and an SNR 1000-MWe system. 2 7 The different plant

arrangements and steam cycles considered were:

Case I: Sodium reheat, with three high-pressure feedwater preheating sections (this was taken to be the reference case).

Case II: Similar to CaseI, with a reduction to two high-pressure

preheating stages.

Case III: Steam reheat, using exit steam from the high-pressure

cylinder, with two high-pressure preheating stages.

Case IV: Steam reheat, using high-pressure cylinder exhaust to

reheat exhaust steam from the intermediate-pressure unit to the lowpressure units. A steam separator is also employed in this arrangement,

with two high-pressure preheating stages.

Case V: No reheat, with direct expansion of steam from a lower

live steam pressure, with two high-pressure preheating stages.

These cases were compared under conditions of constant reactor

exit temperatures and system parameters, which presumably were optimized in each case with respect to cost. In optimizing system. parameters,
119

pressure drops in the primary and secondary loops were held constant.

All cases were normalized to the same net electrical output from the plant.

With primary temperatures fixed and a specified steam temperature, other

temperatures through the system were optimized for minimum total cost

of IHX and steam generators. Fuel-cycle costs were treated parametrically.

Unavailability costs were not included. A discount rate of 14% and capacity

factor of 0.8 were used in the study. A 17-year basis for amortization was

used, with interest at 8% and taxes at 2.5%.

Results of the study are presented in Table 20 and Fig. 49. The

costs are given in terms of percent of total plant cost. The steam-plantline item includes the cost of the turbine generator. The turbine-generator

costs were essentially constant for the conditions studied for the different

cases. The safety-board requirement that the turbine hall be located well

away from the steam-generator buildings penalized the reheat cases because

of the longer pipe runs involved. Costs for the IHX and steam generator

were predicated on proposals or cost estimates provided by suppliers.

TABLE 20. Cost Change with SNR- 1000 vs Reference Design I

(in percent of plant cost at constant MWe)

Case: I 11(a) II(b) II(c) II(d) III IV(a) IV(b) IV(c) V

Plant-cost Fraction

IHX and steam generator 0 -0.71 - -1.72 -2.24 -2.54 -1.63 -1.39 -2.32 -2.61 -2.16

Reactor plant (without

IHX and steam generator) 0 +0.13 ? +0.50 +0.68 +0.60 +0.67 +0.84 +0.99 +1.25

Steam plant 0 -0.49 -0.32 -0.32 -0.24 -0.69 -1.12 -1.09 -0.85 -1.41

Total Cost Change (with various fuel-cycle costs)

0.2 pfenning/kWhr 0 -1.01 -1.56 -1.79 -1.72 -1.41 -1.48 -2.11 -1.93 -1.64

0.3 pfenning/kWhr 0 -0.97 -1.47 -1.66 -1.54 -1.26 -1.29 -1.89 -1.66 -1.30

0.4 pfenning/kWhr 0 -0.94 -1.39 -1.52 -1.35 -1.10 -1.11 -1.67 -1.38 -0.96

20 40 60 80

No TEMP/STEAM TEMPT, C

Fig. 49

Cost Savings vs AT between Primary and

Tertiary. Note: One pfenning ~ 3 mills

100

Case IV (steam reheat using high-pressure cylinder-exhaust steam

to reheat IP steam) offered an advantage over sodium reheat amounting to

U

IU

0

J

IL

U.

0

1' -1.0-

w

? -1.5-

U)

z

can -2.0

0

- CASE H

CASE S

SNR-300 0.5 PFENNIG/kWh

SNR-I000 0.3 PFENNIG/kWh -'

\ 
120

approximately 0.3% of total plant cost for an assumed 1-mill/kWhr fuelcycle costs for a 1000-MWe plant. INTERATOM regards this cost differential to be significant. If the results had been reversed, i.e., if the results

had indicated 0.3% in favor of sodium reheat, sodium reheat could well have

been retained as the reference design. Turbine generators could be provided for all cases considered in the study, either in the form of standard

units or by assembling together various parts of standard turbines. Nonreheat with lower steam pressure looked promising, but economically it

was found to be more sensitive to the effects of unfavorable fuel-cycle and

the rmal- efficiency conditions.

INTERA TOM is now studying 2000-MWe plants trying to find where

economies can be realized by additional development work. The reference

is steam reheat utilizing HP exhaust to heat IP exhaust steam. When the

next plant design is undertaken, the economic parametric studies described

above will be repeated to verify the reference choice. It was indicated that

scale-up to the 2000-MWe plant from the SNR-300 should not present any

major problem. The first component feasibility limit will probably be the

electric generator.

INTERATOM cited a "slight technical advantage" of steam reheat

because of the simpler startup operation and reduced sodium-to-water/

steam heat-transfer surface area. NERATOOM expressed the opinion that

the testing already completed at Hengelo provided grounds for believing

that sodium reheat could operate satisfactorily in a commercial plant if

this cycle was eventually adopted.
121

IX. STEAM-GENERATOR TEST SITES

A. Summary and Comparison

Large steam-generator test facilities in the 50-MWt range have

already been constructed and are in operation, at Les Renardieres in France

and Hengelo in Netherlands, the latter being part of the SNR Consortium effort. Both groups decided that testing of large steam-generator models was

necessary before final fabrication and installation of the demonstrationplant steam generators.

These decisions contrast with the earlier, British decision, which

was to fabricate and install demonstration-plant steam generators without

advance testing of a large model. The British consider that if the first

commercial LMFBR adopts a steam-generator design different from that

of PFR, a large-model test will be necessary. The building of a new

facility in the U.K. or utilization of Hengelo is being considered.

Although the Hengelo test facility is approximately 2 years behind

the Les Renardieres facility, the SNR Consortium schedule allows more

margin between start of steam-generator model testing and the start of

operation of the demonstration-plant steam generators (6 years compared

with 32 years).

Both existing facilities are gas-fired, with bypass loops to provide

transient simulations, and both contain most of the equipment within a

building. At Les Renardieres, the gas-fired heater is inside the building;

at Hengelo, outside. Construction costs and schedules were also similar

and are recorded below.

B. Les Renardieres Steam-generator Test Facility, France [Circuit

d'Essais a Grande Puissance de Generateurs de Vapeur Chauffes

au Sodium (CGVS)]

1. Facility Description

The facility was built by, and is operated by, the French national

electric utility, Electricite de France (EDF). The facility (50-MWt capacity)

is designed to test the performance of sodium-heated steam generators under

steady-state and transient conditions. The choice of 50 MWt was stated to be

a compromise between desiring to test steam generators as near LMFBR

plant sizes as possible, and minimizing the facility construction and operating costs. The facility cost $20 million, and the fuel (gas) costs are

approximately $4000/day. From initiation of design to operation at power

in early 1970 took 3 years. Approximately 6000 hr of steam-generator

testing had been achieved by the time of the U.S. reheat team visit.
122

According to the French, a steam-generator assembly (evaporator, superheater, and reheater) of 50-MWt capacity should provide confidence that full-size LMFBR steam generators will perform satisfactorily.

Test models of this size have essentially the same characteristics as fullsize units in the areas of steady-state heat transfer, flow distribution,

stability, metallurgical effects, and mechanical behavior. With the test

models having tubes of the same diameter and wall thickness, though somewhat shorter than LMFBR plant units, the dynamic performance should be

extrapolable to the larger unit. Differences in the shell time constant due

to differences in shell-wall thickness should be a second-order effect,

which can be accounted for when extrapolating test results to plant conditions.

On the sodium side of the facility, the sodium heater is gasfired, supplying up to 650 C sodium. There are two centrifugal, variablespeed sodium pumps, connected in parallel, and a 15-MWt, fan-driven

sodium cooler. The sodium loop contains several bypass circuits to permit

the rapid changes of flow and/or temperature necessary in the transient

tests. The sodium circuit is fitted with electrical-resistance heaters to

maintain 150 C temperature.

Although the Phenix steam generators are of subcritical pressure, the facility is designed to permit testing of possible future

supercritical-pressure units. The facility includes a mixing tee for mixing

sodium flows of different temperature. One of the flows enters along the

centerline of a mixing chamber and passes through approximately half the

tubes of a honeycomb of tubes. The other flow enters the chamber at right

angles and is baffled to flow through the other half of the tubes in the

honeycomb, transferring heat to or from the other flow in the process.

Physical mixing of the two flows takes place downstream of the honeycomb.

The mixing tee has been tested with 200 C difference between the two flows

and has performed satisfactorily.

The entire facility and test steam generator, with the exception

of the sodium-water reaction discharge stack, are enclosed within the

facility building which is 58 x 28 x 40 m high. The facility requires a staff

of 50, including five operating crews of four men each.

A flow schematic of the facility is shown in Fig. 50. A general

map of the EDF Les Renardieres research facility is shown in Fig. 51.

Further information on the facility is available in Refs. 28 and 29.

2. Status

Testing of the Phenix steam-generator modules was nearing

completion at the time of the visit. A complete, full-size Phenix module,

including evaporator, superheater, and reheater, built of unstabilized

24 Cr-l Mo steel, has been tested under steady-state conditions, without
PRESS REDUCERS A N 1-1 4 Ct. DESUPE GREATER P"5(0 P"5b DESUPE GREATER

----. h' T " 400 C T -"385 C & MIXER M ,3 e. A- - COOLING

50b

565 C P - 1b f- H~ESUPEHHEATER S3 T.1 110 C

HOT By PASS40C 500 C CONDENSER

> s - "; l : PRESS RED

NATURAL 0 P 806 P 80h

GAS - COOLER T !.30 C T 400 C -0

;:.1 's~l".:T -400 C

AIR AIR ", , . 167b, PRESS RED EUEH AE C

P4 " q 65 C EXTRACTION

S(OIIuM NY?7STEAM/WATER PUMPS

HI All IIC O'F S1 SEPARATOR

/TMWAE 

CONDENSED

SEPARATOR WAERIN

S1 S2 TETN

I 00C

- ~S30 MOMInER 410"C r , DEGASSIFIER SDU P -200t,

fMxE T- 160 ;-=1C

M IX E H0 

S2*CPRE HEATERS AE

CIRCUL ATION FEEDWA TE RCOLG P1MsPUMPS 

Fig. 50. EDF Les Renardieres Steam-generator Test Facility--Flow Diagram
EXPLANATION OF THE GENERAL MAP

I - Entrance

2 - Administrative Center

3 - Housing for staff on duty

4 - Medical and Welfare Center. The Restaurant

5 - 1.000 kV experimental-line station

6 - Power-lines building

7 - Ageing-test Imes

g - M.A.I. Hall and terminal station of computing center

9 - Erosion-Corrosion test shop

10 - Offices A

I I - COn circuit shop and Sodium

12 - CO circuit shop extension, and Sodium

13 - Civil-engineering shop

14 - Offices B

15 - Material-testing hail

17 - A.D.E - Experimental building

19 - A.D.E - Chmatron

19 - A.D.E. test shop

20 - Sodium-exchanger test circuits

31 - Extra-High Voltage laboratory

32 - Transformer heat-build-up station

33 - Tests under artificial pollution

34 - Revolving-machinery test station

35 -- Mechanical and climatic test stands

36 - Extra-high power test station (S.T.G.P.)

37 - S.T.P.G. reactances and transformers

38 -- 400.001 volt sub-station

39 - 225.000 volt sub-station

40 - 65.000 vott sub-station

41 - Compressor building

42 - 20.00/60.000 volt building

43 - Medium-power building

44 - Short-circuit alternator buiding

45 - High-power test station

- ----- General Map

Existg buildings

"- -... Foreseen buiddngs

C I ~- - ----UC --- -- ;o U 

® 

-s -- 1 

I~~~ -- - - NAAC - - - -- --

Fig. 51. EDF Les Renardieres Test Site. Note: Steam-generator test facility is in Bldg. 20. Other sodium facilities are in Bldgs. 11 and 12.

N
125

any major problems, and is now removed from the facility building. A

photograph of this module is shown in Fig. 52. The second module, under

test, is made of stabilized 24 Cr1 Mo steel.

Sodium temperature distribution is determined by thermocouples

mounted on the outside of the shell.

Thermocouples are mounted at the

inlet and outlet of each tube, which

also contains an orifice for flow

measurement and flow stability.

The orifice size was varied during

the test program, starting with the

- smallest orifice possible (highest

P ,pressure drop).

Fig. 52. First Steam-generator Module Tested and Tests completed to date

Removed from the Les Renardieres Facility include:

a. Steady-state heat-transfer performance.

b. Flow stability.

c. Dynamic response tests (e.g., response to ramp or step

changes in parameters).

d. Simulated plant startup and shutdown.

Transient testing has been reserved to the end and was to be

performed during the summer of 1972. Because of possible damage to the

sodium heater, the facility cannot duplicate all the possible Phenix transients. Turbine trip can be simulated, except that the steam would continue

to flow to the reheater.

Information on the performance of the steam generators under

test is included in Section V.B.

3. Future Facility Usage

A 45-MWt model of the Babcock-Atlantique helical-tube design

was scheduled for delivery to the Les Renardieres site in November 1972

and should be ready for testing in February 1973. The test of the Stein Industrie straight-tube, 45-MWt model should be completed by the end of 1974.

C. U.K. Steam-generator Testing

As noted before, if the British change the design of the steam generators for the first commercial LMFBR, they will test a 30- to 50-MWt

model in a test facility to be built in the U.K. or will test the unit at Hengelo.
126

As of the visit, an option had been negotiated or was in the process of being

negotiated to have the model tested at the Hengelo facility in Netherlands.

D. Hengelo Sodium Component Test Facility, Netherlands

1. Facility Description (see Figs. 53 and 54)

NERATOOM, the Dutch nuclear-industry group, is responsible

for the design, development and fabrication of the IHX's, steam generators,

and pumps for the SNR demonstration plant. To test a model IHX and a

model steam generator, a 50-MWt test facility has been constructed at

Hengelo, in East Holland, and was in the process of being transferred to

TNO (Netherlands Organization for Applied Scientific Research) as the

facility operator.

Although stated to be a nominally 50-MWt facility, the maximum

heat input to the test steam generator is 55-MWt, and to the IHX, 70-MWt,

because of the regenerative arrangement of the facility. Only a small

extrapolation in size will be necessary from test results to SNR conditions

for the steam generator (SNR unit = 82 MWt). Maximum values of main

parameters of the facility are 650*C (1202*F) sodium temperature, 1800-m3

hr (7900-gpm) sodium flow, 600 C (1112oF) steam temperature, and 215 bars

(3160 psi) steam pressure.

I II

1IFig. 53. Model of Hengelo Steam-generator Test Facility

a

C h 

Irv

s wa s art ^,

fit

i" r F

r

Ell
127

Thermal shocks can be produced, with a maximum rate of shock

of 10F/sec over a 20-sec period.

LMFBR plant transients, such as

turbine trip or reactor scram, may

be more severe. More severe transients than the above 10*F/sec have

been applied to the steam-generator

model as a result of facility incidents,

e.g., due to safety-valve operation

during a sodium leak and fire described below.

The heat is dumped by means

of an air-cooled condenser, waterside

pressure being kept above air pressure to preclude air leaking into the

water circuit. Full-flow demineralization is provided. A turbine can be

simulated by control of steamexpansion valves downstream of the

Fig. 54. Furnaces, Air Preheaters, and Central superheater and reheater under test.

Chimney of Hengelo Steam-generator

Test Facility The facility is capable of

simulating startup, shutdown, and

turbine trip using a turbine simulator. This includes simulation of reheater

conditions during startup and partial load. The turbine simulator consists

of nine cams, which actuate various control devices. The main controls of

the simulator are throttling valves, plus water attemperation of the steam.

Each cam represents a controlled parameter as the y axis and its variation

with main steam flow as the x axis. These parameters include reheaterinlet steam temperature, condenser pressure, deaerator pressure, and

reheater steam flow. A turbine trip can be simulated using this system.

The simulator is presently set up for use only above 20% load.

A trim valve on the sodium system is provided to obtain a

3:7 flow ratio between the reheater and superheater, respectively. The

valve is not changed following the initial setting, nor is it changed with

load. No consideration appears to have been given to special provisions

during startup to maintain a higher pressure in the steam side of the reheater than the sodium side.

The sodium circuits can be electrically trace-heated, and the

water/steam system preheated by means of a 5-MW auxiliary boiler.

The gas-fired heater consists of three units, with the capacity

of each part no more than twice the capacity of similar units from which

they were extrapolated. In each heater, 27 parallel sodium tubes coil in
128

close contact with each other to form the inner wall of the combustion

chamber near the top of the heater. Thus, no refractory is required, and

no continued sodium flow is necessary to remove stored heat in the furnace.

The convection system is a six-pass system, with concentric

coils of sodium tubes near the bottom of the heater. The Type 304 stainless

steel tubes have an ID tolerance of 0.02 mm to ensure correct flow distribution. The tubes are connected to 75-mm (6-in.)-ID toroidal headers, and

the three parallel heaters are connected, top and bottom, by valved piping

loops. The furnaces are of Swiss design (Bertrams) and Dutch fabrication

(N. V. Mashinfabriek, Breda). The furnace is kept preheated above 200*C

to prevent chloride corrosion from a nearby salt plant, even when no sodium

is in the unit. About 0.1 mm of corrosion was experienced in about 1 year

on the external surface of the furnace sodium-outlet pipe. This occurred

after installation and before startup, when the system was at ambient

conditions.

Most of the facility is enclosed, with separate sections for the

sodium plant, the steam/water plant, and the steam-generator model. The

sodium plant and steam/water plant are separated by two concrete block

walls, which enclose the stairs and elevator. The three-unit, gas-fired

sodium heater, the steam condenser, and the sodium/water reactionproducts vessel are outdoors.

The facility includes. a data-logging unit with 500 channels and

a sampling rate of 10,000 points per second. Data logging is performed in

a room separate from the control room, to protect proprietary data of the

steam-generator designer/fabricator. A data readout, but not logging,

capability has been added to the control room.

A patented mixing tee used at the facility was described (and

is shown in Fig. 55). No thermal shielding is included. The tee was tested

in water models, and no experiments have been performed on the tee in

sodium. The tee is designed for sodium temperature differences up to

4000C.

MIXING REGION A sodium flow schematic

of the facility is shown in Fig. 56.

Artist's sketches of the facility are

provided in Figs. 57 and 58. (The

numbered items in Figs. 57 and 58

LINE B HAS SODIUM UNDERGOING J STAGNANT SODIUM are identified in Table 21.) The

TRANSIENT; LINE A SODIUM MUST AT TEMPERATURE

NOT CHANGE IN TEMPERATURE OF LINE A layout of the major components is

RAPIDLY A

shown in Fig. 59. Further inforFig. 55. Mixing Tee Used at Hengelo Steam- mation on the facility is available

generator Test Facility in Ref s. 29-31.
AT - 170~220 0

C

MAX 6500C AT 230 kg/sec MAX AP = 0.8 atm

MAX 370 kg/sec AT 6350C MAX 370 kg/sec REHEATED STEAM

MIN 50 kg/sec.-

TCV 118B SUPERHEATED STEAM

WET STEAM STEAM MAX

I X = 095%- 600C AT 215 atm

MAX 6500

C AT 220 kg/sec

MAX 350 kg/sec AT 6300

C

TC MIN ALLOWABLE 370 C I-- TV

118AI

EVAP

SH

60ft RH

---- HX-

){ r I - FEEDWATER AIRMAX 3000C COOLER 235 atm 13.5 Mw H112 BUF BUF BUF 33 kg/sec

FILL AND STEAM TO DRAINL R E HEHAT ER DAN"HOLDUP" TCV 125 -- '-- - --- - $ - -R AE

t-rVESSEL FILL

H114 AND DRAIN FILL AND DRAIN FILL AND DRAIN

TCV

FILL 119

3-MODULE HEATER AND SODIUM PUMP FILL AND MAX ALLOWABLE AP = 2.0 atm

NOM 53 Mw DRAIN MAX 430 kg/sec DRAIN AT 170 ~300 0

C AT 350 kg/sec

MAX 58 Mw

AT = 170 -300 0

C

MAX AP = 15 atm

MAX 370 kg/sec

Fig. 56. Sodium System Schematic Diagram, 50-MW Sodium Component Test Facility, Hengelo

N
WJ

x V

2~j

T~~ I

td - ~

'"7

i

fv ry 

t

r

o , t.

Fig. 57. 50-MWt Sodium Component Test Facility, Hengelo. (See Table 21 for explanation of numbers.)

'I

W Sodium Component TeF'

DESG AND ONSTRUCTED BY

Z77A 7"

IL .x ~. I 7

:w
17

0 Y

Fig. 58. Cutaway View of Steam Hall at 50-MWt Sodium Component Test Facility,

Hengelo. (See Table 21 for explanation of numbers.)

131

f

2-1-7.

N,

1"

0

w i.
132

TABLE 21. Legend for Figs. 57 and 58

Sodium test hall

Steam hall

Sodium furnaces (three)

Main building

Control room

Steam-generator box

Evaporator

Superheater

Reheater

Sodium circulation pump

Cold trap

Dump tanks

Expansion vessel

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

Pipe bridge

Sodium furnace

Air preheater

Chimney

Auxiliary boiler

Deaerator steam system

Steam separator

Main entrance, steam hall

Main entrance, sodium test hall

Test instrumentation room

Power-supply room

Intermediate-heat exchanger frame

OFFICES WORKSHOPS

STE AM

HALL MEDICAL OFFICE

STEAM GENERATOR CELL SECURITY OFFICE

SODIUM

HALLZ SODIUM

COOLER

PIPING BRIDGES NATURAL GAS

CONTROL 

STATION 

CNESR

CONDENSERS

Q SODIUM- // HE ATE RS

iDOCK

CANAL

Fig. 59

Plan View of 50-MWt Sodium Component Test Facility, Hengelo

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.
133

2. Status

Facility design started toward the end of 1966 and was extensively revised in 1967. Construction started in June 1968 and was completed toward the end of 1971, although sodium was added to some components

in April 1971. Steam-generator testing started in February 1972. Facility

cost was $15.3 million. A staff of 50 is required, including 20 operators and

seven workshop technicians.

A building extension is now being incorporated to provide an

additional steam-generator test bay, so that one model can be installed

while another is under test, thereby reducing the turnaround time between

models to be tested. This improvement was undertaken because the SNR

demonstration plant will include two different steam-generator designs,

both to be proved out by a test model at Hengelo. Another change to be

made will be moving the cold traps into a separate cell to facilitate removal and replacement, which are proving to be more frequent operations

than originally anticipated.

The components now under test are the SNR IHX, and the

straight-tube steam generator that will be used for two of the three SNR

demonstration-plant loops. The steam generator installed at Hengelo includes a reheater unit in parallel with the superheater. The reheater has

already been deleted from the SNR design, but it was too late to change the

Hengelo model installation.

At the time of the visit, three weeks of steaming operation had

been attained, with thermal power up to 25 MWt. The maximum steam conditions attained were 170 bars (2500 psig) and 510 C (950F). In April 1972

a small leak in the sodium system at the IHX destroyed several cable trays,

causing a shutdown of 6-7 weeks. The facility was expected to be started

again at the end of June 1972.

3. Sodium Leak

Thermal-shock tests at Apeldoorn showed that the original IHX

head seal using a metal O-ring was not adequate during thermal transients.

For this reason, the IHX model installed in the test facility was disconnected

for modification. The modification consisted of a seal weld to replace the

O-ring. While the IHX was being modified, some of the sodium lines had to

be cut and capped. Before cutting and capping one of the lines, the Dutch

established a freeze seal about 12 m (40 ft) long to backup an isolation valve.

The freeze seal was established by removing the insulation on the line and

allowing the line to cool; no auxiliary cooling of the line was used. The

line remained frozen for several weeks.

The leak occurred when the cap was cut off for reconnection of

the line. It is estimated that 10-100 liters of sodium leaked out of the pipe.
134

Sodium carbonate was used to extinguish the fire, but restrictive access to

the leak made firefighting difficult. Additional platforms for access were

being installed. Also, special covers over cable trays are being installed.

It has been postulated that the leak through the freeze seal

occurred because a gas pocket formed in the freeze seal due to contraction

of the sodium as it solidified. Later, the isolation valve leaked slightly and

the liquid sodium came through the gas pocket of the freeze seal to the location of the leak. The program delay resulting from the leak was caused by

sodium spilling into cable trays underneath the leaking pipe, destroying

many facility cables. These cable trays were covered with sheet metal,

but the sodium entered at locations where vertical connections had been

made to the cable tray. The Dutch feel strongly now that electrical cabling

should be routed away from the main sodium system to avoid such problems

in the future.

4. Additional Problems

After filling, plugging of the sodium cold traps was experienced.

The first cold trap plugged soon after system purification was started, and

this unit was removed and examined. It was determined that very little of

the total capacity of the unit contained sodium oxide. The cooling design

of these cold traps was poor, causing the plugging problem. The second

cold trap also plugged in a similar fashion. A third cold trap, with modifications to the cold-trap internals and cooling system, has now been installed. The possibility of recycling cold traps is also being studied.

Difficulty has been experienced in obtaining exactly the right

evaporator temperature before introduction of feedwater and in obtaining

the proper AT's on the tube sheets. It now takes approximately one-half day

to preheat the steam generator from a cold condition to a minimum-power

condition. On a hot start, it has been found easier to go down to 200 C and

restart very quickly.

The Dutch have had poor experience with welding sodiumcontaminated piping, even after the sodium film was removed back 1-2 ft.

The heat from welding melts the sodium, which flows into the weld area.

The resulting welds do not meet their requirements. Specifically, the root

pass has been found to have defects. These defects were observed using

double-wall X rays and dye tests.

Additional problems include failure of some trace-heating elements and many small problems with valves and other equipment. No

problems were encountered with the loop-support system, and good flow

distribution was confirmed in the sodium heater.
135

5. Steam-generator Testing

The steam-generator tests are intended to (1) confirm the heat

transfer, (2) detect waterside instabilities, (3) confirm effectiveness of

the thermal baffles protecting tube sheets (approximately 100 thermocouples

installed for this purpose), and (4) confirm the predicted movement of the

shell bellows.

The overall performance of the steam-generator model up to the

25% load condition, reached for a short period of time, is close to predictions. The steam-generator test results are discussed in detail in

Section V.E.

The instrumentation for the reheater in the present installation

is not as great as for the superheater, since the reheater is of less interest,

being similar to the superheater. Besides size, the only difference between

the superheater and reheater is that the reheater tubes have larger diameters

and thinner walls. No problems have been experienced to date with the reheater. There is a cold-sodiumbypass to the superheater and reheater inlet. This was originally intended to be used for control of temperatures to

these units; however, it has not been used during current operation. Proper

temperatures are maintained, since the steam temperature coming out of

the reheater and superheater is close to the sodium temperature under all

loads.

No vibration testing or measurements are being made in the

steam-generator test. Vibration and flow-distribution measurements were

made in models tested in water at Apeldoorn.

6. Future Facility Usage

Following completion of testing of the present units, the helicaltube SNR evaporator will be tested. The Dutch are not interested in future

tests of an entire steam-generator system since they expect to obtain systems information in the present systems test, which was set up to study the

dynamic behavior of the system. Only an evaporator on as large a scale as

possible will be tested. In comparison with the other units, the evaporator

has major problems associated with fabrication, heat transfer, and stability.

The test of as large a unit as possible is required to prove the evaporator

capability to the maximum extent possible, including the adequacy of fabrication. If tests of other steam-generator units were desirable, it would be

sufficient to test the superheater, assuming the reheater to be of the same

basic design. Reheater conditions could be established in the superheater

test.

Consideration is being given to artificially fouling the waterside

of the steam generator or to conducting very long-duration tests at steady

state to determine fouling.
136

The tests of the helical-coil tube model should be completed in

early 1974. The Dutch were negotiating with the British and Italians concerning the possibility of their using the Hengelo facility for component

testing in 1974 and 1975. The test on the SNR IHX should be completed by

mid-1973, when capacity would be available for testing other IHX units.

7. Leak-detection Experience

Significant variations in hydrogen in the sodium have been

found using in-sodium hydrogen detectors. These detectors use nickelmembrane diffusion tubes with an ion pump, developed in Apeldoorn. The

system has a high background of hydrogen, which fluctuates frequently between 0.5 and 1 ppm. The fluctuations are in large part due to temperature

changes, with higher hydrogen levels occurring at higher temperatures.

These high background levels make it difficult to detect a leak; however,

since the Dutch have a hydrogen detector at both the inlet and outlet of each

module, they can detect a leak by the difference. However, the high background level and the fluctuation require that the readings be interpreted to

determine if a leak has occurred. Also, nonoperability of the cold trap

would result in the hydrogen level of the sodium increasing rapidly from a

very low value to approximately 1 ppm.

Cover-gas hydrogen detectors will be installed in the surge

tank to help detect hydrogen levels. Hydrogen will be scavenged from the

system above set points. Hydrogen gas has been injected to calibrate the

in-sodium detectors, which have responded well to the injections. Some

evidence of hydrogen diffusion through the steam-generator tubes from

waterside corrosion has been seen. A background level of 0.1 ppm is the

target for the facility. No accoustic-leak detectors are included.

E. 5-MWt Test Facility, Bensberg, Germany

In support of the overall German sodium-reactor program, including

the KNK reactors, the INTERATOM 5-MWt test facility at Bensberg has

tested many components, including steam-generator models, earlier and

smaller in capacity than the models now being tested at the 50-MWt facility

at Hengelo, Netherlands.

The sodium heater is gas-fired to provide up to 580 C sodium, and

the flow rate provided by the pump is 180 m3 /hr. The main components of

the sodium loop are fabricated from niobium-stabilized 24 Cr-i Mo steel.

The water-steam loop is designed for 250-bar, 550 C steam and

feedwater at 10-m3 /hr and 300 bars. Heat is dumped through an air-cooled

condenser.
137

The facility includes three test rigs: a 700-mm-dia, 10-m-high

vessel for tests of fuel-handling equipment, a test rig with an argon circuit

for testing sodium-vapor traps, and a fuel-element and control-rod test rig.

Operation of the facility started in 1965. Since then it has tested

steam generators, IHX's, pumps, valves, cold traps, carbon traps, plugging

meters, fuel-handling devices, control rods, and vapor traps. The facility

is still in use for testing such components, including a single-shell oncethrough steam generator.
138

X. OTHER TEST SITES

A. TNO- -Central Technical Institute Facilities at Apeldoorn, Netherlands

1. Introduction

The TNO (Nederlandsche Organisatie voor ToegepastNatuurwetenschappelijk Onderzoek, i.e., the Netherlands Organization for

Applied Scientific Research), is a government organization, which has two

sources of income--the government and the industry. The TNO Central

Technical Institute (C TI) is one of 40 institutes in TNO. As one of its many

functions, the TNO-CTI has the responsibility for sodium-technology basic

research in the Netherlands. NERATOOM supports the industrial portion

of sodium-technology development. The two are meshed. For example,

NERATOOM built the Hengelo Steam Generator Test Facility, which is to

be operated by the TNO-C TI.

Van Gasselt (10 years with C TI), head of the Sodium Technology

Section, reports to K. A. Warschauer, the head of the Division of Physical

Transport Phenomena. About 200 employees are employed by CTI at

Apeldoorn.

The Apeldoorn Facility has eight sodium loops; seven of the

eight are automatically operated 24 hr a day. Twelve people run all the

loops. Each loop can be dumped automatically at certain preset limits.

TNO-CTI designs its own loops, but uses two or three constructors for

fabrication and erection. CTI is responsible for loop safety and its own

safety criteria. The cost of the loops at Apeldoorn is: for each of the

two isothermal corrosion loops, 300,000 guilders; the anisothermal loop,

800,000 guilders; the chemistry loop (6 months to build), 500,000 guilders;

the pump testloop, 1,000,000 guilders; the SWISH loop, 1,500,000 guilders; the

thermal-shock loop, 800,000 guilders; and the Arend loop, 450,000 guilders.

(One guilder equals 31 cents.) Following are the loops:

a. Two small isothermal corrosion loops.

b. One anisothermal corrosion loop.

c. A chemistry loop.

d. A loop called SWISH, originally designed for supercritical

steam R&D, now used for subcritical work for the SNR

sodium-cooled steam generators (this loop is described

in Section V.E).

e. A thermal-shock loop.

f. A pump-test facility, which also can be used for thermal

shocks.

g. A loop called Arend, named after one of the people working

for TNO.
139

Apeldoorn also has sodium-analytical capabilities, capability

for postoperational cleaning components free of sodium, and a workshop.

Further, there is an extensive waterflow distribution and vibration test

loop to simulate shell-side conditions of IHX's steam generators, and reactor plenums.

Apeldoorn facilities, in existence for 6 years, have lately been

expanding. The CTI Division of Hydrodynamics and Aerodynamics and the

TNO-Metallurgical Institute of TNO, which was at Delft, are located, or

will be located, at Apeldoorn. Corrosion studies that normally would be

carried on by other divisions of the CTI are temporarily under the responsibility of Van Gasselt of the Sodium Technology Section.

2. Corrosion Loops

The oxygen content and the mass velocity can be varied in the

corrosion loops, and the linear velocity can be varied from 0.5 to 5 m/sec.

These loops have been used to test nicrobrazed tube-to-tube-sheet joints

for the SNR steam generator and to study the effects of corrosion and carburization of Type 304 stainless steel when oil from pump seals leaks into

sodium, as has occurred. Self-welding, cavitation, and fretting corrosion

studies are still being carried on at Delft.

The isothermal corrosion loops have been used to test 2* Cr-1 Mo

and 9 Cr-1 Mo specimens in Type 316 stainless steel corrosion loop.

Specimen-to-loop surface was 1/500. At 500*C, no carbon transfer was

found with 2* Cr-1 Mo, but at greater than 500*C, decarburization was

found. With 9 Cr-1 Mo, it was difficult to find decarburization. However,

oxidation did occur. The oxygen content was 20-25 ppm. To start with,

the actual carbon in the 2* Cr-1 Mo was 0.12%; after decarburization at

500*C, it was 0.02%. Test results are to be published in the TNO quarterly

reports. Findings above 500*C essentially do not agree with Thorley of the

UKAEA, who indicates decarburization begins at 450*C, and do not agree

with the U.S. Babcock & Wilcox tests, which indicate decarburization begins

only at 540*C . Original condition (heat treatment) of steel may be important,

i.e., how dispersed the carbides are in the original heat. Various heat treatments will be tested to try to get some correlation with Thorley and B&W

in their experiments.

In the process of testing hydrodynamic bearing materials in

high-velocity sodium, about 30 m/sec (100 ft/sec), with samples 50C below

isothermal loop temperature of 580*C, a sodium chromite deposition phenomenon was observed. Specimens are 50 mm wide, 1 mm thick, and about

200 mm long, and there is one specimen on each side of the sodium flow

passage (see Fig. 60). The spacing of the opening is 1 mm. Sodium chromite deposits have been measured as 0.1 mm thick. The deposition phenomenon occurred in a time period of 4-5 hr. As the deposition took place,
140

SPECIMEN sodium velocity decreased to

S0mm 16 m/sec in 4-5 hr after start of

SODIUM 1 mm 1 mm SODIUM deposition, the velocity remaining

PRESSURIZED constant for the rest of the test

- 

perature in this loop was 180-200*C.

Fig. 60. Section of Loop Showing Specimens for The initial oxygen content was

High-velocity Sodium Tests at Pump- 10 ppm, later increased to 20-

bearing Clearances, TNO-CTI, Apeldoorn 25 ppm. Chemical activity probably plays a role in this deposition

at high velocities, a type similar to that of carbon-transport activity of

ferrite. The deposition occurred even when the concentration of iron was

less than saturation. Cavitation is prevented in the tests by high pressure

on the downstream side. So far, the tests have been done with nickel specimens in the stainless steel loop, but other types of materials will be tested.

The anisothermal loop operates between 300 and 700*C at

1.75-2-m/sec sodium velocity. Specimens are generally 2 x 10 x 30 mm.

The high-temperature specimen can be stressed.

The Type 316 stainless steel 1-in. anisothermal corrosion loop

has the capability of testing specimens at 480, 600, 650, 690, and 700*C in

the heated-up portion and 700, 650, 600, and 500*C in the cooled-down portion. Fuel elements for INTERATOM are now being tested in these loops.

MSA (Mine Safety Appliance) EM pumps are used in the corrosion loops. They have not been satisfactory, since they cannot withstand

thermal shocks, multiple heatup and cooldown cycles, or high pressures.

They are capable of only 3 atm, whereas tests require several times this

number. French EM pumps are capable of 8-10 atm. The MSA pump is a

two-electrode pump; the French pump is a one-electrode pump.

3. Chemistry Loop

Sample-taking techniques have been developed for removing

samples from the 2-in. 700*C-capability sodium-chemistry loop to the

laboratory where analytical means are available for measuring oxygen,

hydrogen, and carbon concentrations. Hydrogen and oxygen in-line instrumentation is used. The oxygen meter they have used or have tried to use

is of the UNC design. The electrodes are sensitive to breakage because of

the AT on the electrodes. There is no development work on this meter at

the laboratory.

Plugging meters are used for on-line detection of oxygen.

Two hydrogen detectors using membranes of nickel, stainless steel, and

possibly other materials are used on-line to detect hydrogen. The Ultek

ion pumps are operating satisfactorily. There are indications that an
141

entire spectrum of elements is coming through the 1/4-mm-thick Type 316

stainless steel diffusion tube operated at 300*C, and the Dutch are in the

process of investigating this phenomenon. The spectrum includes hydrocarbons. The Dutch also indicate that certain elements such as sodium

peroxide formed at high oxygen contact can severely corrode the nickel

membrane. The sample harps used in the chemistry loop were shown and

are fairly sophisticated. An on-line distillation apparatus was being developed in the chemistry loop. Several of these have been sold by the

Dutch to other European countries. The automatic plugging indicator in the

chemistry loop is of the English Electric design. The chemistry loop has

two EM pumps.

4. Thermal-shock Loop

The 2-in.-dia thermal-shock facility (see Fig. 61) has been in

existence for 9 years; it is fairly extensive. Many subcomponents designed

for Hengelo and for SNR have

TH+ T

c been thermally shocked in this

fii ilit-i E. va l a of th1 c

BUFFER

r -- - - - - OR MIXING - - - - - SUBJECTED TO THERMAL TRANSIENTS

(REMAINDER OF LOOP

Q 10 IS ISOTHERMAL)

SPECIMEN 1 SPECIMEN 2

A B C D

X X PROGRAMMED

VALVES

EC "dH

COOLER HEATER

LcJL y . x LIjJJ. U'.LLLLS eLs 

are:

D PUMP

a. The O-ring seals

they had originally intended

to use in the intermediate heat

exchanger for SNR.

b. Full-size bearings

for the SNR 5000-m3

/hr sodium

pump.

C. Specimens of vessel nozzle attachments.

d. Multiple tube-totube-sheet connections for their

steam-generator facility.

The thermal-shock

IQQloop is capable of about

100C/sec from 550 to 380*C.

TRANSIENT OPERATION: Total capability is 300C/sec

for 10 sec. The AT's for this

VALVES A & C ARE PROGRAMMED FROM FULLY OPEN TO FULLY CLOSED AS VALVES B & D

ARE PROGRAMMED FROM FULLY CLOSED TO FULLY OPEN. SPECIMEN 1 THEN RECEIVES A facility are specified by COLD SHOCK FROM TH TO TC, AND SPECIMEN 2. A HOT SHOCK FROM TC TO TH. NERATOOM. The flow is

Fig. 61. TNO-Apeldoorn Thermal-shock Loop 5 m3/hr . A 2 -in. mixing tee

for the thermal-shock loop

was described, in which the mixing thermal sleeve comes in at the elbow.

Specimens to be shocked are generally contained in a 2-m-dia insulated

chamber, which can be filled with nitrogen in the event of a sodium leak.
142

When thermally shocked at 120*C in 3 sec, the Hengelo IHX

O-ring seal was damaged in

RING

REPRESENTING

NOZZLE

E

LZ

REPRESENTI N

VESSEL

-4 in.

-2 in,

T 1 in

SODIUM

Fig. 62

Vessel-nozzle Thermal-shock M

TNO-CTI, Apeldoorn. (Series of

zie models can be tested on this

about two cycles. As a result, the O-ring was

replaced with a seal weld.

G The model used for thermal shock of

vessel-nozzle specimens is shown in Fig. 62.

It consists of a 2-in.-dia, 2-ft-long center

tube, representing the shell. The nozzles

are represented by 0.25-in.-thick plates

formed into rings welded to the center tube.

-29 in.

Six of these 0. 2 5-in. segments were welded

to the tube at 4-in. intervals. Thermal shocking of vessel-nozzle welds has been successful.

The only effects observed are those on the air

side. Inclusions present in the weld will come

out, leaving a vacancy that apparently does not

affect the quality of the weld.

Seven full-scale bearings for the

SNR 5000-m3/hr pumps are being thermally

shocked, primarily to determine the effect on

the hard surfacing (see Fig. 63). Stellite 6

[odel, applied by welding and by spraying has been

noz- under test. A cermet hard facing is yet to be

rig.) tested. Van Gasselt indicated that Stellite

sticks, but cermets do not. Cermets grind,

but keep running. The team was shown a series of full-scale bearings for

the 5000-m3/hr SNR pump, which were or are being thermally shocked.

The first Stellite 6 bearing was tested

for 5000 hr, then thermally shocked, THERMOCOUPLES

a2nd it aeartd t he in ond sha e

though a rigid inspection is yet to be

made.

The thermal-shock facility

is usually used in conjunction with the

SWISH unit. Sodium from the thermalshock loop is used to heat the outside

of the SWISH loop. This loop also has

a hydrogen detector and has waterchemistry capability.

5. Cold-trap Design

TNO has had difficulty with

cold traps. Original designs used a

packing that extended across the cross

section of the cold trap and was very

ineffective due to crystallization for a

- 16 in.--

SODIUM -JOURNAL

BEARING

- -SODIUM

THERMOCOUPLES

NOTE: ACTUAL LOCATION OF SODIUM INLET

AND OUTLET NOZZLES NOT KNOWN.

Fig. 63

Schematic of 5000-m3 /hr Pumpbearing Thermal-shock-test Container, TNO-CTI, Apeldoorn
143

short distance at the entrance to the packing, blocking cold trapping above

this point. The Hengelo cold traps have been replaced twice. A third series

of tests was being conducted at Apeldoorn. A more effective cold trap could

be obtained if the packing was put into an annulus. The Arend loop was being

used for cold-trap testing. The cold trap was cooled with air. The cold trap

on test had an annular packing.

6. Pump Loop

The pump-loop capability at Apeldoorn was 280 m3/hr, 85.5-m

head, and design temperature of 600*C. Since commissioning in July 1968,

the pump had operated for about 10,000 hr. The 280-m3

/hr pump had two

interchangeable sodium-lubricated hydrostatic bearings with a mixed-flow

impeller and shaft sealing in the gas phase. The gas seal initially consisted

of a mechanical seal-type, then a visco-seal. The visco-seal (shown in

Figs. 64 and 65) was developed by Stork. Referring to Fig. 66, a specially

grooved bushing, flexibly mounted on the shaft, rotates within an outer stationary bushing. A high-viscosity lubricant is applied as shown. The special

groove pattern allows for a pumping action during shaft rotation, forcing the

lubrication into the annulus and providing a seal between the cover gas and

the upper atmosphere. Correct design of grooves also prevents lubricant

leakage. The lubricant also keeps the flexible rotating bushing centered on

the shaft.

Apparently the visco-seal can be operated under a vacuum.

A static seal is needed when the shaft is not rotating; this also allows for

maintenance of the visco-seal. The visco-seal has the advantages of long

life, nonleakage of lubricant, low cost, ease of assembly, and low sensitivity

STORK DOUBLE VISCO4SEAL

LUBRICATING OIL

SE-ALING ORL

-CLEAN ARGON

SEALIG OI

FAKK.

spiral er.....as

rotaing ris

directionof

ort .of g

0 

I- alss ebrio prsar

prasar. built Ito oil th

f reae 

ereer raaunr,

Fig. 64. Visco-seals in Tandem Arrangement

L

Fig. 65. Visco-seal Principle
144

PUMP

66 in.

FLOWMETER MIXING TEE

6 in. M 6 in.

3 in. BO E

THERMAES HOCK I 3 in.

S - SEMIVARIABLE FLOW

ORIFICE

3 in.

PROGRAMMED VALVE

U AMOTOR-OPERATED VALVE

Fig. 66. TNO-CTI 280-m3 /hr Pump Loop at Apeldoorn

to contaminants in the lubricant. Two seals in tandem provide for 100% seal

reserve, argon flushing between the seals, and helium-leak tests. A disadvantage of the visco-seal is its complex lubrication supply and pressurization system. It is possible that a lubrication-system pump can be driven by

the pump shaft, reducing some of the complexity.

The pump loop has been used for testing pump seals, bearings,

bearing materials, bearing clearances, vibration, cavitation, pump drives,

and moments of the axis.

The pump loop has thermal-shock capability (see Fig. 66).

Normally, the main sodium flow is through a semivariable-flow resistance.

A thermal-shock vessel, cooled by a blower, is located in a bypass line.

For thermal shocks, a sufficiently large temperature difference is created

between the sodium in the main circuit and that in the thermal-shock vessel.

This is done by heating or cooling the thermal-shock vessel at zero bypass

flow. Meanwhile, the temperature in the main circuit is controlled by running or stopping the pump. As soon as the desired temperatures are

reached, the pump is operated at the desired speed and the programmed

valve in the bypass circuit is operated so that the correct flow is initiated

through the thermal-shock vessel to provide the desired thermal shock on

the pump. A thermal shock of 200*C can be obtained in 3 sec.

Stork started manufacturing sodium pumps 9 years ago; the

first were 4700-rpm pumps with three bearings. The present ones are

9 6 0-rpm pumps. Stork has built the SNR 5000-m3/hr, 9 6 0-rpm pump,

being tested at Bensberg. The hydrodynamic tests of the hydrostatic bearing for the pump are done at Delft. Initially these tests are done in water.

Final tests of bearings are in sodium.
145

The Apeldoorn pump-test facility was being altered to be able

to test SNR core components at 200C per 4 sec at 280 m3/hr. This test

was scheduled to be done at the end of 1972.

7. Sodium Cleaning

Van Gasselt indicated that sodium cleaning is accomplished by

first vacuum-distilling in place, then removing the equipment, placing it in

water, drying it, and finally cleaning it with acetone. This is representative

of TNO sodium cleaning, satisfactory for laboratory use.

8. Miscellaneous

A NERATOOM 12-in. control valve will be tested at Apeldoorn

and then installed in Hengelo for further testing. The purpose of the valve

was not specified.

9. Laboratory Visits

A visit was made to the section of the laboratory used to fullscale water-test the flow distribution and vibration of the Hengelo IHX and

steam generators using plastic models. A plastic model of the SNR reactor

vessel was being tested, simulating a degassing section. This was not discussed. The loops were made of 12 -in. -dia Plexiglas.

Also shown was a water tunnel (not part of sodium technology).

The water velocity could reach 22 m/sec. The test section was 1 m 2 in

cross section; maximum flow volume was 22 m3/sec. The tunnel was- used

to study wind loadings on structures such as buildings. It could be used for

testing submarines and for cavitation studies. It may have been used to test

Hengelo and SNR component models.

A control room for the thermal-shock test facility was visited.

A cold stream of sodium was mixed with a hot stream of sodium from the

isothermal portion of the loop to alternately shock components (see Fig. 61).

10. IHX Tube Plugging and Welding

Tests have been conducted on SNR IHX explosive plugging and

welding. For welding, it was necessary to have very clean walls at the

tube-to-tube-sheet connection, and it was also necessary to have a good

design of the plug. Small amounts of oxygen were applied to the tube wall

to oxidize the sodium.' A rotating brush was used to remove the residue,

after which welding could be done.
146

11. Valves

An eccentric offcenter 6-in, butterfly-type valve based on a

Camflex valve has been redesigned for use at Apeldoorn. The valve company is Masoneilan, which has offices in Norwood, Mass. The original

bodies were cast. Apeldoorn purchased the valve internals and actuator

and provided its own forged body and freeze seal. The original valve, designed for use in the oil industry for fused salt flow, cost 5000 guilders.

With a TNO body and freeze seal, the cost was 10,000 guilders (about

$3,000). A chief reason for doing their own work on the valve body was the

need for providing quality assurance not available on the outside.

B. EDF Les Renardieres Test Facility

The EDF Research and Development Directorate has several R&D

Test Services:

1. Electrical Equipment

2. Machinery and Automated Production

3. Network Engineering

4. Data Processing and Applied Mathematics

5. General Technology, Applications of Electricity, and Hydraulics.

Services 1, 4, and 5 are located at Les Renardieres. The others

are located at Chatou, Clamart, and Saint Denis.

Personnel in the R&D directorate comprise 801 engineers and technical and administrative executives, and 1272 supervisors, technicians,

and administrators.

The General Technology Services has three departments:

1. Material Studies

2. Nuclear Technology

3. Applications of Electricity.

The Nuclear Technology Department has four divisions or groups:

1. Thermal Reactor Technology

2. Fast Reactor Technology

3. Sodium-heat Steam-generator Test-circuit Operation (CGVS)

4. Erosion-corrosion Tests.
147

The CGVS is described in Sections V.B and IX.B. The Fast Reactor

Technology Division is responsible for studies on sodium components.

Sodium testing is conducted in two sodium-test buildings. Tests include:

1. Corrosion loops

2. Materials creep in sodium

3. Valve tests

4. Steam-generator rupture disks

5. Insulation

6. Journals, bearings, and gears.

A chemistry laboratory provides for sodium monitoring and analysis.

Material testing is divided into several phases. Steam-generator

decarburization tests and carburization tests are done in the large Icare

loop. Fatigue testing of steam-generator tube-to-tube-sheet joints is conducted, as well as other technology tests such as those for tube supports.

The Icare loop is a small loop for the study of carburization and

decarburization of steam-generator materials. It consists of a mockup of

a Phenix secondary loop on a small scale. Parameters studied in this loop

are surface area, temperature and temperature gradients, sodium velocity,

surface-to-flow ratio, and time. The most important parameters were

temperature and temperature gradient, sodium velocity, and time. The

scaling down from the Phenix system was done using nondimensional parameters so that the test results would be valid for Phenix.

C. CEA--Cadarache Nuclear Research Center

1. Introduction

A large segment of the test facilities for the French LMFBR

program is located at the Cadarache Nuclear Research Center (CNRC).

Located at Cadarache are Rapsodie, a 40-MWt fast reactor;

Harmonie, a 2.5-kWt neutron-source reactor; Masurca, 1 5-kWt (max)

critical assembly; a Van de Graaff accelerator; a Plutonium Technology

workshop, for developing LMFBR fuels; the CABRI test reactor, used to

simulate LMFBR core excursions; and technological test buildings.

The team visit to Cadarache was made mainly to discuss and

observe the sodium-water reaction analysis and test facilities and to visit

the sodium-component support facilities. As such, it directed much of its

attention to:
148

a. Sodium-water reaction analysis and test.

b. Component test facilities.

c. Sodium systems corrosion analysis and test.

A short visit was also made to Rapsodie.

2. Component Test Facilities (technological tests)

The component test facilities are located in several buildings

(see Fig. 67): HR-1--1300 m2 , 30 m high (1400 ft2 , 100 ft high); HR-2--

850 m2 (10,000 ft2 ); HR-4--1200 m2 (13,000 ft2 ).

HR-1 includes sodium tests of main reactor components (pumps,

heat exchangers, fuel-handling machines, control-rod drives, etc.). The

building uses two 15-ton traveling-bridge cranes and a 60-ton semigantry

crane.

HR-2 contains test equipment for mechanical, hydraulic, and

thermal tests not involving sodium. Water loops are used to test for flow,

vibration, contraction, and mixing problems. These include small-scale

tests of Phenix reactor-vessel and fuel subassemblies.

HR-4 has a 10-ton traveling-bridge crane and houses sodium

test rigs to aid in the design and calibration of sodium detection and to

provide sodium heat-exchange and safety tests; friction, wear, and weld

tests; and acoustic detector tests. In addition, a small area called ZEFIR,

250 m2 (2700 ft2 ), is used to study sodium-air reactions and model studies

of reactor structures.

The team visited the HR-1 and HR-4 buildings.

a. HR-1 Building

The LMFBR component test facilities in this building

include:

(1) A Rapsodie core mockup, not in use, but available if

circumstances indicate the need, after an incident, to reactivate the mockup

to determine how to bring Rapsodie back on line.

(2) A series of test facilities consisting of seven pots

(pot-type containers), all with static sodium. Two of the pots are used for

fuel handling: one for the Phenix IVHM (Internal Vessel Handling Machine),

and one for the Phenix elevator mechanism. A total of 930 fuel-transfer

operations were made with the IVHM with sodium temperature between

250*C (482*F) and 360C (680*F). The elevator mechanism was operated
i 1 

pe

li.6.CdrceNula eerhCntr ra1lamni;Ae ,Msra

~~~A e 3, Ra so ie Are 4, Te h ol g Tes Buil ings A. *R 1 Bt .,^ " _ " . HR- . .. ya, ' ."t

.0
150

for 1000 cycles. A third pot is used to test control rods for Phenix and

Rapsodie. Two prototype control-rod mechanisms were cycled 1000-

2500 cycles. A fourth pot is used to study the convection of argon in the

cover gas. A fifth pot is used for testing of instrumentation. The use of

the other two pots was not indicated.

(3) A test rig, available for testing sodium-vapor traps.

The rig is a vertical unit having four layers of Raschig rings. Sodium

vapor enters at 450*C and leaves at 150*C.

(4) A sodium facility, for testing Phenix dummy subassemblies. Filters are located in each inlet of these subassemblies, prototyping the filters that will clean the sodium of Phenix as it goes through the

dummy core (now installed in Phenix).

(5) A test loop, available for testing the Phenix sodiumsystem insulation.

(6) A sodium rig for endurance and thermal-shock test of

Phenix-type subassemblies of all types. Each test is run for 1000 hr with

160*C AT. Two test sections are used. The subassemblies are removed

each 100 hr for inspection. A final design prototype is tested for 8000 hr.

The subassemblies are full scale and prototypal to a large extent, except

for dummy pins.

(7) A two-section rig, used for testing fuel subassembly

orifice cavitation. The orificing for Phenix is similar to that in EBR-II,

where the diagrid has a fixed orifice. The fuel-subassembly nozzle also

has orifices. The relative orientation of the fixed orifice in the diagrid

tube between the two diagrid plates and the nozzle holes determines the

orificing.

b. HR-4 Building

The HR-4 sodium-technology building contains 20 to 40 different sodium rigs. Most of these were not described. Pontier (CEA) described several of the rigs.

(1) A test loop is used to determine the self-welding

characteristics between fuel pins, providing for 1-mm relative movements.

(2) A second loop is used for testing small EM pumps.

(3) A third loop is used for wear measurements.

(4) A fourth and fifth rig are used for testing fuel-handling

bearing mechanisms.
151

(5) A sixth rig tests mechanical connections, which are

presumably the sliding fittings between Phenix piping from the pumps to

the high-pressure plenum of the reactor vessel. This mechanical connection (probably a sliding fit) connects the pump to the core inlet plenum,

allowing relative freedom of movement. The system consists of a sphere

in a cylinder and enables (a) pump structure strains to be reduced as much

as possible, (b) leaks to be reduced, and (c) the pump to be dismantled and

assembled, simply by inserting the sphere into the cylinder. A prototype

articulated sleeve was first tested in water. The piping runs in the primary

tank are short without any bends, using a sliding connection. This connection is tested for leakage, vibration, and other characteristics.

(6) Three rigs were shown that have been used for sodiumboiling studies. They were not in use, since it had been determined that

sodium boiling was no longer a problem.

(7) A large test area is devoted to ultrasonic testing with

a series of five rigs for ultrasonic flowmeters: (a) an ultrasonic detector

for onset of boiling, (b) an ultrasonic detector for under-sodium scanning,

(c) an ultrasonic detector for measuring vibration (in the order of 1 mm),

(d) a rig for ultrasonic level measurement, and (e) a rig for ultrasonic

temperature measurement.

Pontier indicated that 80-100 people were responsible for

operating both HR-1 and HR-4, including engineers, technicians, and operators, with five operators per shift around the clock for each of the

two buildings.

3. Corrosion Test Facilities--SEML

Section Etude Metalle Liquide (SEML) consists of 35 people at

Cadarache headed by Champeix. There are four groups:

a. One group in SEML does corrosion studies under Baque,

who was not present. There are eight loops of two types: one for flowthrough around samples, and the second where the sample is the loop itself.

b. SEML conducts chemistry studies of sodium, particularly

sodium analysis of carbon and oxygen. (A short presentation regarding the

sample type of loops was made in Detroit during winter 1971-1972.)

c. SEML has a metallurgical group, which conducts electronmicroscopy and microprobe work under Lafon.

d. SEML has a physical-chemistry group under Trouve, who

led the SEML discussion. This group studies impurities in sodium and is

just beginning to work on this problem. Ferritic versus stainless steel
152

loops, in which the loop itself is the sample, have been operating for about

18 months. The walls are analyzed for deposits of carbon. No analysis

was available as of June 6. A loop was scheduled to be cut up by the end of

June 1972. Temperatures and the ratios of surface areas have been modeled

within these loops to represent the Phenix secondary system.

A radioactive mass-transport loop was scheduled to be operating

within 1 year. The loop will be Type 316 stainless steel. No work had been

done to date by SEML on the effects of sodium-water reactions on materials.

SEML was beginning to study the sodium-oxygen-hydrogen system to determine the natural components of such a system in a representative LMFBR.

SEML will start studies on the sodium-carbon-oxygen system and hopes to

start studies on the sodium-nitrogen-oxygen system. The SEML people do

not conduct any studies or tests of rupture-disk or hydrogen-meter membrane material.

The team toured the corrosion loops. There are four carbontransfer loops, where the loop is the sample itself. The cost of each loop

is about 200,000 francs ($40,000).

4. Rapsodie

The team visited Rapsodie, which had just finished a 60-day run.

Mr. Gajac conducted the group in the reactor building, where refueling was

in progress using two refueling machines, one to bring in fuel and one to

remove fuel. In this manner, the refueling time was cut by approximately

half. When asked about tritium, Mr. Gajac indicated that this was not a

problem. Traces of tritium had been found in the oil of the pump bearing,

but not in the secondary system. When asked how the fuel-pin leaks were

located, Mr. Gajac indicated that a small group of approximately four suspect subassemblies could be isolated and removed from the core. The

details of how to detect the failed subassembly were not revealed. Also,

how the group of four suspects was located was not revealed.

The Rapsodie building was originally designed for 2.4 bars

(~35 psig). This pressure design is too high. A recalculation indicated that

the maximum amount of sodium (the basis for the original pressure design

is similar to EBR-II containment) that could escape from the reactor is on

the order of 200 kg.

Mr. Gajac indicated that the recycle time for one fuel element

is 70 min. An entire refueling is made in about 4 days from shutdown to

startup. Experimental s ubas s emblie s complicate the refueling.
153

D. INTERATOM, Bensberg, Germany

1. Nitriding of Materials

Nitriding of ferritic and austenitic steels in a nitrogen cover

gas was reported in a Monaco paper. 3 2 Although the Germans did not have

definite plans to use nitrogen cover gas in the steam generator, nitrogen

was used in a steam-generator test at Bensberg as an inert atmosphere in

an enclosure around the steam generator. Nitriding effects in the presence

of sodium were observed. Tests were made to confirm this observation

using the following materials:

a. 2* Cr-1 Mo unstabilized and stabilized with niobium and

with niobium and 0.5% nickel.

b. Type 304 stainless steel.

c. Inconel 600.

d. Type 347 stainless steel.

e. Type 321 stainless steel.

These tests were conducted at 500, 550, and 600C for times

up to 800 hr, and provided information on kinetic behavior of the nitriding

process. The tests confirmed that hydrogen apparently had no effect on

nitriding, but that sodium in the presence of nitrogen influenced nitriding.

The work on nitriding has been extended to determine the effects

of nitrogen in the SNR cover gas, in the range of 350-500 C. A literature

survey indicates many values for nitrogen impurity in argon, but gives no

indication that nitriding of components will occur. Grosser 32 will study

effects of nitrogen in the range of 500-20,000 ppm. He noted that the

UKAEA at Risley has such a program for PFR with temperatures ranging

to 600 C. SNR does use nitrogen in the primary-system cells. They do not

anticipate that they will have any difficulty with nitriding in the refueling

cell, where the temperature would be in the range of 300-400 F. To date,

one test with 4000 ppm of nitrogen at 500C has shown no evidence of nitriding.

2. SNR Pump-test Facility

The pump-test facility, made of Type 304 stainless steel, is

designed for 650C. The pump-test stand is shown in Fig. 68. The characteristics of this pump are shown in Table 22. The 600- and 350-mm steamgenerator isolation butterfly valves were tested in this loop.

The actual SNR pump will have a head of about 140 m. The

pump uses hydrostatic bearings and is driven by a synchronous motor with
154

ac converted to dc and then back to ac to control the pump speed. The pump

was tested in a water-test facility at Stork in Hengelo, Netherlands, in

October 1969 using demineralized deaerated water.

Tests in sodium at Bensberg started in the fall of 1970 and

were completed in July 1972. Except for minor difficulties, the experience with the pump drive has been good. Present plans are to use this

type of drive for the SNR pump.

/500 K , / PCv at .Jsm/eLt-AMOWTOR

MA/Mg

~-- FL//D COUPI/M6

POMP MOaTO //kW/

f viri TOR UE CaMvfrF"e

Lr

Fig. 68. Prototype Pump and Drive

v

J
155

TABLE 22. Design Characteristics of 5000-m3/hr SNR Prototype

Primary Sodium Pump on Test at Bensberg

(mechanical, free-surface)

Metric Units

Capacity

Design temperature

Working temperature

Specific weight of sodium

at working temperature

Total head

NPSH required

Impeller- -singlesuction diameter

Argon pressure

Shaft seal

Rpm

Rpm control, o

Type of speed control

Pump-casing design

pressure

Double shell of pump

casing

Argon pressure between

the two shaft seals'

Structural material

Engineering Units

22,000 gpm

1076 F

1040*F

5000 m3/hr

580*C560*C

820 kg/m3

85 m

10 m

880 mm

0.5-1.5 atm (abs)

51 lb/ft3

280 ft

33 ft

35 in.

7.4-22 psia

Mechanical or visco

960

3 to 100

Frequency changer

or fluid coupling

11 atm (abs) at 580*C.

Vacuum at 200 C.

1.8 atm (abs) at 500*C.

Vacuum at 200*C.

1.5 atm (abs) and vacuum at 80*C.

162 psia at 1076*C.

Vacuum at 392 F.

26 psia at 1076*F.

Vacuum at 392 F.

22 psia and vacuum at 176*F.

X 6 Cr Ni 18-11

(U.S. 304 SS)

Stork-Hengelo,

Netherlands

Manufacturer

3. SNR Rotating-plug Test

Adjacent to the pump-test facility is a full-scale mockup of the

SNR triple-rotating shield plug and refueling machine used to check out the

SNR fuel-handling system. The fuel elements will be transferred in sodiumfilled pots. Dry tests have been conducted for about 1 year. Sodium testing

was scheduled to begin during summer 1972. Sodium-vapor deposition will

be an important part of the test.
APPENDIX A

of U.S. LMFBR SodiumReheat Team

Monday, May 22

Tuesday, May 23

Wednesday, May 24

Thursday, May 25

Friday, May 26

Monday, May 29

Tuesday, May 30

Wednesday, May 31

Thursday, June 1

Friday, June 2

Monday, June 5

Tuesday, June 6

Location

U.S. Embassy, Paris

Saclay, France

Les Renar.dieres, France

Clamart, France

Dounreay, Scotland

Dounreay, Scotland

Bensberg, Germany

Bensberg, Germany

Apeldoorn, Netherlands

Hengelo, Netherlands

The Hague, Netherlands

Marcoule, France (Phenix)

Cadarache

Visit

Meeting of Team

CEA, EDF

EDF

EDF

TNPG

UKAEA

CEGB

TNPG

UKAEA

CEGB

INTERATOM

INTERATOM

TNO

TNO/NERA TOOM

NERATOOM

CEA/EDF

CEA

156

Itinerary
157

APPENDIX B

Agendas

1. CEA/EDF: Saclay, May 23

09:15

10:00

10:15

10:40

12:45

13:45

Afternoon

Welcome to visitors

Steam generators for Phenix and 1000/1200-MWe power

plant

Sodium or steam reheat in case of using a 1000-MWe

Phenix-type steam generator (CEA)

Coffee break

Sodium or steam reheat with a Babcock-Atlantique steam

generator (EDF)

General discussion

Departure to Les Renardieres

Arrival at Les Renardieres and lunch

Visit to the 50-MW steam-generator test facility at

Les Renardieres

2. United Kingdom Atomic Energy Authority: Dounreay Experimental

Reactor Establishment

Program for the visit of the USAEC

Fast Reactor Sodium Reheat Team on May 25-26, 1972

Thursday, May 25

11:40 Arrive Wick Airport

Coach to DERE

12:30 Arrive DERE

Welcome and introduction to the Establishment by

Director, DERE

In attendance: Mr. P. Mummery, Director, DERE

Mr. C. W. Blumfield, Deputy Director,

DERE

Messrs. W. Macrae and D. Taylor, TNPG

Mr. W. Kemmish, CEGB

Mr. J. Williamson, TOD

& If
158

13:00 Luncheon, Director's Dining Room

Director Mummery and Messrs. Blumfield, Macrae, Taylor,

Kemmish, R. H. Campbell, A. D. Evans, and Williamson

14:00 Coach to PFR Site

TNPG Conference Room No. 1

Discussions with Messrs. Campbell, Macrae, Taylor,

Kemmish, and Williamson

16:45 Coach to Pentland Hotel and Royal Hotel

18:00 Coach trip to Dunnet Head, returning to Royal Hotel for 19:30

19j 5 Dinner, Royal Hotel, Thurso

20':00 Messrs. Blumfield, Macrae, Taylor, Kemmish, and

Williamson

Friday, May 26

09:00 Coach, Pentland Hotel and Royal Hotel to DERE

09:25 Arrive DERE- - PFR Site

TNPG Conference Room No. 1

Continue discussions and visit the Prototype Fast Reactor--

Messrs. Macrae, Taylor, Kemmish, and Williamson

13:00 Luncheon, Director's Dining Room

Director Mummery and Messrs. Blumfield, Macrae, Taylor,

Kemmish, and Williamson

14:00/

14:00 Coach to Wick Airport 14:15

15:45 Depart Wick Airport

3. Meeting USAEC- Team, INTERATOM, Bensberg, Germany; Sodium

Reheat Cycle Design; May 29-30, 1972

Place of meeting: Atrium II

Beginning: 09:30

Monday, May 29

09:30 a. SNR design and development program

b. Large (1000-2000 MWe) LMFBR plant application and design

(1) Economic considerations

(2) Technical considerations
159

(3) Programmatic efforts

(4) Design

13:00 Lunch

14:30 Continuation of discussion

17:30 Dinner, Goethehaus

Tuesday, May 30

09:30 Continue discussion

13:00 Lunch

14:30 Visit APB plants

4. TNO/NERATOOM-- Hengelo, Netherlands, June 1

a. Mechanical

(1) Repairs of sodium systems and/or components

(a) Philosophy

(b) Methods and/or procedures

(2) Valves and valve choice

(a) Frozen seal versus bellows seal

(b) Control valves design and manufacturers

(3) Means of compensating thermal expansion

(a) Bellows, etc.

(b) Connection of systems of different temperature

b. Operational

(1) Sodium quality and quality control

(a) Definition of cleanliness in primary and secondary

circuits

(b) Measurement, sampling techniques

(c) Recyling problems

(d) Component design (cold traps, plugging meters)

(2) Heating of components where allowable temperature differences are limited

(a) Mode of internal heating

(b) Necessary means for safe operation

c. Electrical

(1) Trace heating problems and design

(a) Type of elements used, specific allowable heat loads,

lifetime experience
160

(b) Control systems

(c) Fixation to piping

(d) Temperature specifications of piping

(2) Cabling. Cabling protection against fire and sodium spills.

d. Water Chemistry. Requirements of feedwater for LMFBR

steam generators, degree of alkalinity.

e. Measurements

(1) Strain-measurement techniques

(a) High temperature out of sodium

(b) High temperature in sodium environment

(2) Flow- measuring techniques

(a) Calibration of (big) sodium flowmeters

(b) Measurement methods

f. Steam Generators

(1) Cycle design choice for LMFBR

(a) Outlet quality of evaporator, problems of water separation, recirculation, control concepts

(b) Starting up from different conditions

(2) Steam-generator testing

(a) Extent of test programme

(b) Instrumentation

g. Detection of Water Leaks in Steam Generators

(1) Methods

(2) Experience of detection of leaks; mode of leak versus

signal

(3) Influence of hydrogen diffusion through tube walls, especially in steam generators; production of hydrogen on

water side

5. CEA, Cadarache, France, June 6

08:45 Pickup at the Guest House

09:00- 09:15 Meet Mr. Estavoyer and staff in Bldg. 212 Conference

Room

09:15- 10:15 Discussions of the sodium-water reaction with Messrs.

Lions and Biscarel in Bldg. 212 Conference Room

10:15- 11:30 Inspection of the test a- a led by Messrs. Lions and

Falgeyrette s
11:35- 12:15

12:30- 14:15

14:30- 15:15

15:30- 16:30

16:30

Inspection of Test Halls HR-1 and HR-2 led by Mr. Pontier

Lunch at the Guest House

Discussion and inspection of the SEML led by Mr. Trouve

Inspection of Rapsodie led by Mr. Gajac

Departure

161
16Z

APPENDIX C

Participants

1. Saclay: May 23

Name Or ani zation

Emile Rodet

Gilbert Cornu

Maurice Barberger

Pierre Zaleski

Henri Procaccia

Etienne Cambillard

Marcel Robin

Lucien Duchatelle

Jean Birault

Louis Vautrey

Claude Moranville

Georgel Le Guelte

Pierre LeCocq

Louis Valibus

A. De Montardy

EDF/Equipment

EDF/Equipment

EDF/Equipment

EDF/Coordinator, Fast Reactor Activities

EDF/Des.ign and Research

CEA

CEA/Assistant Head Mechanical & Thermal

Studies Leader

CEA

CEA

CEA

CEA

CEA

EDF/Chief, Liquid Metals Division, Chatou

Research & Development Center

EDF/Design and Research

' EDF/De sign and Research

2. Clamart, France: May 24

Name

P. Zaleski

J. Faure

M. Cravero

P. LeCocq

A. Rafalovitch

R. Lheureux

G. Cornu

Organization

EDF

EDF

EDF

EDF

EDF

EDF

EDF
163

3. Dounreay, Scotland: May 25 and 26

Name Organization

R. H. Campbell

D. Taylor

W. Macrae

P. Mummery

C. W. Blumfield

J. A. Bray

Simpson

Tregoni

J. Williamson

W. Kemmish

A. D. Evans

G. Malcolm

4. Bensberg,

TNPB/Deputy Managing Director (5/25 only)

TNPG/Manager Engineering Design-- PFR

TNPG/Project Manager--Fast Breeders

DERE/Director

DERE/Deputy Director

DERE/Chief Sodium Water Reactor Analysis 

Tests

DERE

DERE

Technical Operations Directorate

CEGB/Head, Fast Reactor Engineering

DERE/Head, PFR Division

DERE

Germany: May 29 and 30

Organization

W. Ratzel

W. Jansing

K. Dumm

R. Fischer

H. Metzner

Trenz

A. Brandstetter

D. Grosser

G. Slopianka

Seitz

INTERATOM/Chief, LMFBR Heat Transfer

Systems and Auxiliary Branch

INTERATOM/Chief, LMFBR Large Fast

Facilitie s

INTERATOM/Chief, Sodium Water Reactor

Analysis and Test Branch

INTERA TOM/Chief, Design Optimization Branch

(all reactors)

INTERATOM/Chief, LMFBR Design Branch

INTERATOM

INTERATOM/Manager, LMFBR Engineering

Department

INTERATOM/Nitriding specialist

INTERATOM/Proje ct management- - visit

coordinator

Kraftwerk Union/Turbine specialist

Name
5. Apeldoorn: May 31

Name

M. L. G. van Gas sett

Ervin L. Chirer

Organization

TNO/CTI- Head Sodium Technology

TNO/Metallurgical Institute

6. Hengelo, TNO/NERATOOM, June 1

Name

N. A. Doets

D. E. L. Kruyt

A. R. Braun

A. J. van't Hoft

G. Huijben

Romeny

Smit

Organization

C TI/TNO

C TI/TNO

CTI/TNO- -Manager

NERATOOM N. V.

C TI/TNO

CTI/TNO (Electrical

CTI/TNO (Operation

Hengelo Test Facility

7. The Hague, Netherlands: June 2

Name

Wil J. C. de Cle r cq

W. W. Nij s

G. A. de Boer

Ruyte rman

Bakema

M. de Hes

P. W. Piet Hein Ludwig

Organization

NERATOOM

NERATOOM

NERATOOM

NERAT OOM

NERA TOOM

NERATOOM

NERATOOM

8. Marcoule, France (Phenix): June 5

Name

J. Befre

T. Le Marechal

Andy J. Polliart

Henry Noel

Organization

CEA/DCP

EDF/Phenix

CEA/DCP

CEA/DCP

164
165

9. Cadarache, France: June 6

Name Organization

Estavoyer

R. Pontier

G. Gajac

Trouv

Baqu6

N. Lions

Biscarel

CEA/Deputy Director

CEA/Test Facilities Chief

CEA/Chief of Rapsodie Operations

CEA/Chief of Physical Chemistry in SEML

CEA/Chief of Corrosion Studies in SEML

CEA/Chief of Sodium Water Reaction Tests

CEA/Sodium Water Reaction Tests

Note s:

1. R. Stella, USAEC Scientific Representative, U.S. Embassy, Brussels,

Belgium, accompanied the team to Bensberg, Apeldoorn, Hengelo, and

The Hague.

2. J. LaFleur, USAEC Scientific Representative, U.S. Embassy, Paris,

France, accompanied the team to Saclay, Les Renardieres, Clamart,

Marcoule, and Cadarache.

3. S. Rousso, USAEC Scientific Representative, U.S. Embassy, Paris,

France, accompanied the team to Saclay, Les Renardieres, and Clamart.
166

APPENDIX D

Address List of Places Visited

1. Centre d' Etude s

Nucleaires (CEN) de Saclay

Commissariat a l'Energie Atomique (CEA)

Boite Postale No. 2

9 1- Gif- sur- Yvette

(Seine- et- Oise), France

2. Electricity de France

Les Rendardieres

Route de Sens pres de Moret-sur-Loing

77- Ecuelle s

France Phone--422-48-20

3. Electricity de France

Direction des Etudes et Recherches

17, Avenue du Genral de Gaulle

92- Clamart

France Phone- - 644- 15- 15

4. Dounreay Experimental Reactor Establishment (DERE)

Dounreay, Thur so

Caithne s s, Scotland

United Kingdom Phone-- Thurso 2121 xt. 7242

5. INTERATOM

506 Bensberg/Ko1n

Friedrich Ebert-Strasse

Federal Republic of Germany Phone--0-22-04-51

6. TNO- - Head- - K. A. Warschauer

Division of Physical Transport Phenomena

Chemical Laboratory

372 Arnhemseweg

Apeldoorn Phone-- 05760-31700

7. TNO- - Hengelo Steam Generator Test Site

Head--A. R. Braun

50 MW Component Test Facility

Petroleumhavenstraat

Hengelo (0) Phone-- 05400- 10860 or 20163

8. NERATOOM

LAAN vanNieuw Oost Indic 129- 135

P.O. Box 539

The Hague

Netherlands Phone-- 07 0-8 14821
167

9. Centre de Marcoule

Commissariat a 1'Energie Atomique

Boite Postale No. 106

30, Bagnols- sur- Ceze, France

10. Commissariat a 1'Energie Atomique (CEA)

Centre d' Etudes Nucleaires de Cadarache

B. P. 1, St. Paul-lez-Durance

France
168

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