NUREG/IA-0028

I ntematonal Agreement Report

Review of

LOFT Large Break Experiments

OECD LOFT Project

Prepared by

S. M. Modro

Austrian Research Center Seibersdorf

S. N. Aksan

Paul Scherrer Institute

V. T. Berta

Idaho National Engineering Laboratory

A. B. Wahba

Gesellschaft fuer Reaktorsicherheit

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

October 1989

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report was prepared under an international cooperative

agreement for the exchange of technical information. Neither

the United States Government nor any agency thereof, or any of

their employees, makes any warranty, expressed or implied, or

assumes any legal liability or responsibility for any third party's

use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use

by such third party would not infringe privately owned rights.

Available from

Superintendent of Documents

U.S. Government Printing Office

P.O. Box 37082

Washington, D.C. 20013-7082

and

National Technical Information Service

Springfield, VA 22161
NUREG/AA-0028

International

Agreement Report

Review of

LOFT Large Break Experiments

OECD LOFT Project

Prepared by

S. M. Modro

Austrian Research Center Selbersdorf

S. N. Aksan

Paul Scherrer Institute

V. T. Berta

Idaho National Engineering Laboratory

A. B. Wahba

Geselischaft fuer Reaktorsicherheit

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

October 1989

Prepared sa part of

The Agreement on Research Partdipatton and Technical Exchange

under the Inteenation Thermal-Hydrauik Code Assessment

and Applcatlon Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report documents work performed under the joint sponsorship of the

Austrian Research Center Seibersdorf, Austria, the Paul Scherrer Institute,

Switzerland, the Idaho National Engineering Laboratory, USA, and the

Gesellschaft fuer Reaktorsicherheit (GRS), Federal Republic of Germany.

The information in this report has been provided to the USNRC under the

terms of an information exchange agreement between the United States and

the Swiss Federal Office of Energy (BEW) (Agreement on Research Participation and Technical Exchange Between the United States Nuclear Regulatory

Commission and the Swiss Federal Office of Energy (BEW) for and on Behalf

of the Government of Switzerland in the USNRC Thermal Hydraulic Transients

Program and the ECCS-Reflood Program of the Swiss Federal Institute for

Reactor Research (EIR) Covering a Four-Year Period, May 2, 1984), and an

information exchange agreement between the United States and the Federal

Republic Republic of Germany (Technical Exchange and Cooperation Arrangement

Between the United States Nuclear Regulatory Commission and the Bundesminister

Fuer Forschung und Technologie of the Federal Republic of Gernany in the field

of reactor safety research and development, April 30, 1981). The BEW and GRS

have consented to the publication of this report as a USNRC document in order

that it may receive the widest possible circulation among the reactor safety

community. Neither the United States Government nor the BEW nor the GRS or

any agency thereof, or any of their employees, makes any warranty, expressed

or implied, or assumes any legal liability of responsibility for any third

party's use, or the results of such use, of any information, apparatus,

product or process disclosed in this report, or represents that its use by

such third party would not infringe privately owned rights.
ABSTRACT

Six non-nuclear and five nuclear large break loss-of-coolant experiments

were performed in the Loss-of-Fluid Test (LOFT) PWR facility at the Idaho

National Engineering Laboratory. These experiments provided a large amount

of data necessary for evaluation and refinement of reactor system-computer

codes and had major impact on the understanding of large break

loss-of-coolant accidents. An overview of these nuclear large break

experiments performed under NRC and OECD LOFT programs is given and the major

research results are presented.

*1

EXECUTIVE SUMMARY

One of the most prominent reactor safety research facilities in the world

was the Loss-of-Fluid Test (LOFT) facility. This unique facility, located at

the Idaho National Engineering Laboratory, was a 50 MW(th) pressurized water

reactor which was designed on the principal of volume scaling to simulate the

major components and system responses of a four-loop commercial PWR during a

hypothetical loss-of-coolant accident. Extensive research programs were

conducted at the LOFT facility under the sponsorship of the U.S. NRC and

later under the auspices of the Organization for Economic Cooperation and

Development (OECD) with funding from a consortium of ten countries.

Forty-four experiments were completed over a nine year period ending with

a severe fuel damage experiment in July 1985. These experiments were

conducted at typical initial and boundary conditions associated with loss of

coolant accidents and anticipated transients in commercial PWRs. The

research program included six nuclear large break LOCA experiments the

primary objective of which was to obtain data on LOCA phenomena and system

response for a range of initial and boundary conditions which could be used

for reactor system code development and assessment. The objectives, design,

and principal results of the nuclear large break experiments are described.

The important thermal-hydraulic phenomena measured in the large break

transients and their significance are discussed in the principal areas of

analysis that have been undertaken.

The sequence of large break LOCA experiments was conducted with

increasing transient severity wherein the initial and boundary conditions

increasingly approached licensing limits. The L2-2 Experiment was the first

nuclear experiment conducted in the LOFT facility. This experiment was

conducted with a maximum linear heat generation rate of 26.2 kW/m and with

continuous primary coolant pump operation. Subsequent experiments were

conducted with larger power densities and variations in primary coolant pump

operating boundary conditions extending to an immediate trip at break

initiation with a decoupling of the flywheels. The emergency core cooling

iii
system (ECCS) operating conditions incorporated various degrees of

degradation such as loss of ECC in the broken loop, as an example. The

principal finding from the large break experiments is that, for the degrees

of severity in initial and boundary conditions, the measured fuel cladding

temperatures remained well below the peak cladding licensing limit

temperatures.

The data obtained from the LOFT large break LOCA experiments provided new

insight into phenomena associated with large break LOCA. One of the most

important phenomena, observed for first time in the LOFT transients, is fuel

cladding cooling/quench during blowdown. This phenomenon is very important

to the degree of transient severity because it removes a large part of the

stored energy from the fuel early in the transient. Extensive research was

conducted to investigate the source and magnitude of the phenomenon. The

cooling/quench phenomena was determined to be caused by system hydraulics in

response to the operational characteristics of the primary coolant pumps

relative to the transition from subcooled to saturated choked flow at the

break. Two of the large break experiments, L2-5 and LP-LB-1, incorporated

pump characteristics, by design, which did not produce the cooling/quench

phenomena. The cooling/quench phenomena was allowed to occur by intent in

other experiments, in order to quantify the phenomena and provide proof of

the source. The significant finding was that the cooling/quench phenomena

would occur in all conditions except for a pump trip concurrent with break

initiation and decoupling from the flywheels. Similar limiting conditions

are expected to be required to suppress the phenomena in commercial PWRs.

Because of the significance of this early cooling/quench phenomenon and

because the systems codes at the state of development at that time were not

able to calculate this phenomenon accurately, several specific investigations

were conducted to determine quantitatively the effect of the LOFT cladding

thermocouples on the measured cooling phenomena. The concern was that these

thermocouples, by providing additional surface to the fuel cladding (fin

effect), could affect the heat transfer characteristics and also may measure

only very localized phenomena. Separate effect experiments in other

facilities and analysis of LOFT data showed conclusively that the blowdown

iv
cooling/quench in LOFT large break LOCA experiments is real. However, the

thermocouples do reduce the blowdown peak cladding temperature because of an

induced delay to DNB. Fin cooling subsequent to DNB was not found to

adversely affect measurement accuracy. In contrast, surface cladding

thermocouples are recognized to accelerate reflood quenching which occurs at

much slower rates and lower pressure compared to the observed quenching

during blowdown.

Examination of the LOFT large break LOCA experiments provided important

insight on emergency core cooling (ECC) performance during large break

transients. In general, the experiment results have shown that the ECCS

operation even in degraded conditions was effective in core quench and

transient recovery. The hot wall delay time was at most 2 s. Only a small

part of the ECC water is lost through downcomer bypass to the broken loop

cold leg indicating that the "downcomer bypass", which is one of the concerns

in licensing, is not of concern. Cooling phenomena during blowdown can

reduce the time to final quench by about 30% because the reflood quench is

strongly dependent on cladding temperature levels at the end of the refill

phase. Experiment LP-FP-1 which included upper plenum ECC injection showed

that ECCS mode as being highly effective and that relatively small amounts of

water can quench the core.

Predictions of the LOFT large break LOCA experiments were performed using

older generations of computer codes such as RELAP4/MOD6. Newer codes such as

RELAP5/MO02 and TRAC-PF1/MOD1 are in the process of being assessed using LOFT

data. Review of several recent calculations indicates that the hydraulic

conditions in the LOFT experiments are calculated relatively well. However,

cooling/quench phenomena associated with blowdown and reflood are not well

calculated. The calculations do not correspond to measured cladding

temperatures during blowdown even though the hydraulics appear to be

reasonably calculated. These results indicate the need for acquiring better

understanding of the early cooling/quench phenomena and consequent

improvement in post-CHF heat transfer modeling.

v
In summary, the LOFT experiments showed that the core thermal response in

a large break LOCA is much less severe than initially anticipated, and the

ECCS as designed is effective in plant recovery.

vi
ABBREVIATIONS AND ACRONYMS

BWR Boiling Water Reactor

CCFL Counter Current Flow Limitation

CFM Central Fuel Module

CHF Critical Heat Flux

COUPLE General purpose heat conduction computer code

DNB Departure from Nucleate Boiling

ECC Emergency Core Coolant

ECCS Emergency Core Coolant System

FEBA Solid type electric heater rod

FRAP Computer code for analysis of LWR fuel under transient condition

FRG Federal Republic of Germany

GRS Gesellschaft fuer Reaktorsicherheit mbH.

HPIS High Pressure Injection System

ILCL Intact Loop Cold Leg

INEL Idaho National Engineering Laboratory

KFK Kernforschungszentrum Karlsruhe

LANL Los Alamos National Laboratory

LB Large Break

LOCA Loss-of-Coolant Accident

LOCE Loss-of-Coolant Experiment

LOFT Loss-of-Fluid Test

LPIS Low Pressure Injection System

LTSF LOFT Test Support Facility

LWR Light Water Reactor

MLHGR Maximum Linear Heat Generation Rate

NEPTUN Reflood test facility - Paul Scherrer Institute, Switzerland

NRC Nuclear Regulatory Commission

OECD Organization for Economic Cooperation and Development

PBF Power Burst Facility

PCP Primary Coolant Pump

PPS Plant Protection System

PSI Paul Scherrer Institute (previously EIR (Switzerland))

PSS Pressure Suppression System

vii
ABBREVIATIONS AND ACRONYMS

(Continued)

PWR Pressurized Water Reactor ,- .

QOBV Quick-Opening Blowdown Valve

RCP Reactor Coolant Pump

REBEKA Cartridge type electric heater rod -.

RELAP Thermal-hydraulic computer system code for.LWR transient and LOCA

analysis - INEL

RCS Reactor Coolant System

RNB Return to Nucleate Boiling

SG Steam Generator

SPND Self Powered Neutron Detectors

TC Thermocouple

T/H Thermal-Hydraulic

TRAC Thermal-hydraulic computer system code for PWR transient-and LOCA

analysis, three-dimensional_.capability.-T..LANL

viii
CONTENTS

ABSTRACT ................................................................ i

EXECUTIVE SUMMARY ....................................................... iii

ABBREVIATIONS AND ACRONYMS .............................................. vii

1. INTRODUCTION ........................................................ 1

2. THE LOFT FACILITY ................................................... 8.

3. LOFT NUCLEAR LARGE BREAK EXPERIMENTS ................................ 15

3.1 Experiment L2-2 ................................................ 15

3.2 Experiment L2-3 ................................................ 19

3.3 Experiment L2-5 ................................................ 23

3.4 Experiment LP-02-6 ............................................. 28

3.5 Experiment LP-LB-1 ............................................. 32

3.6 Experiment LP-FP-1 ............................................. 37

3.7 Experiment LP-FP-IA ............................................ 41

4. SIGNIFICANT LB LOCA PHENOMENA ....................................... 44

4.1 "rhe Blowdown Bottom-up Core Quench ............................. 45

4.2 Effects of Primary Coolant Pump Operation ...................... 51

4.3 The Blowdown Top-down Core Quench .............................. 54

4.4 Blowdown and Reflood Heat Transfer ............................. 56

4.4.1 Surface Thermocouple Effects on Cladding Quench ......... 58

4.4.1.1 LTSF Experiments ............................... 59

4.4.1.2 PBF Experiments ................................ 64

4.4.1.3 Analysis of LOFT Data .......................... 65

4.4.2 Nuclear Fuel Versus Electrical Fuel Rod Simulators:

Simulation Limitations During Blowdown .................. 78

4.4.3 Reflooding and Boil-off: External Cladding

Thermocouple Effect, and Nuclear Fuel Rod and

Electrical Heater Rod Behavior .......................... 83

4.5 The Break Flow ................................................. 89

4.6 ECC Performance ............. ............................. 91

ix
5. SUMMARY AND CONCLUSIONS ............................................. 113

6. REFERENCES .......................................................... 115

APPENDIX A--CODE CALCULATIONS ........................................... A-1

A.1 The thermal-hydraulic code DRUFAN-02 .......................... A-2

A.2 RELAP5/MOD2 ................................................... A-9

A.3 TRAC-PFI/MOD1 ................................................. A-1O

FIGURES

1. Axonometric projection of LOFT system configuration for

Experiment L2-2 .................................................... 9

2. LOFT reactor vessel and core ....................................... 10

3. Predicted and measured fuel rod cladding temperature near

midplane of center fuel module for Experiment L2-2 ................. 18

4. Mass flow rates in cold legs of intact and broken loops during

Experiment L2-2 .................................................... 18

5. Temperature of fuel rod cladding during Experiments L2-2 and

L2-3 ............................................................... 20

6. Calculated and measured cladding temperature on fuel rod 5J4

at 0.53 m above bottom of the core for Experiment L2-3 ............. 22

7. Maximum temperature on fuel rod cladding during Experiments

L2-3 and L2-5 ...................................................... 24

8. Temperature of fuel rod cladding in upper half of central fuel

assembly during Experiments L2-3 and L2-5 .......................... 24

9. Temperature of fuel rod cladding in a peripheral fuel assembly

during Experiments L2-3 and L2-5 ................................... 26

10. Predicted and measured maximum temperature of fuel rod cladding

for experiment L2-5 ................................................ 26

11. Core exit and upper plenum fluid temperatures compared with

saturation temperature during Experiment L2-5 ...................... 27

12. Temperature of fuel rod cladding at different elevations during

the second core heatup in Experiment L2-5 .......................... 29

13. Fuel rod cladding temperatures at different elevations in the

center fuel module during Experiment LP-02-6 ....................... 29

14. Peak fuel rod cladding temperatures during Experiments L2-3,

L2-5, and LP-02-6 ................................................. 31

x
15. Measured and predicted peak fuel cladding temperatures for

Experiment LP-02-6 ................................................. 31

16. Measured and predicted differences between intact and broken cold

leg mass flows for Experiment L2-6 ................................. 33

17. Upper central fuel assembly cladding temperatures during

Experiment LP-LB-1 showing top down quench ......................... 35

18. Peak cladding temperature during Experiment LP-LB-1 ................ 35

19. Measured and calculated peak cladding temperature for Experiment

LP-LB-I ............................................................ 36

20. Fuel rod cladding temperatures measured at the peak power

elevation during Experiment LP-FP-1 ................................ 39

21. Fuel rod cladding temperatures measured at different elevations

in the center fuel module during Experiment LP-FP-1A ............... 42

22. Fuel rod cladding temperatures measured in a peripheral bundle

during Experiment LP-FP-1A ......................................... 42

23. Broken loop hot leg mass flow rates during Experiments L2-5 and

LP-02-6 ............................................................ 47

24. Broken loop cold leg mass flow rates during Experiments L2-5 and

LP-02-6 ............................................................ 47

25. Relative SPND output and cladding temperatures for Experiment

L2-3 ............................................................... 49

26. Upper plenum temperature during Experiment L2-3 .................... 49

27. Cladding thermocouple response during Experiment L2-3 at 38 cm

(15 in) elevation .................................................. 50

28. Initial cladding cooling versus axial elevation for center fuel

assembly during Experiment L2-3 .................................... 50

29. Coolant mass flow rates in the intact loop cold leg during

Experiments L2-5 and LP-02-6 ....................................... 53

30. Fuel cladding temperatures measured at rod 5G06 at different

elevations during Experiment LP-02-6 ............................... 53

31. Fuel cladding temperatures in the peripheral bundle 2 during

Experiment LP-02-6 ................................................. 55

32. Fuel cladding temperatures in the peripheral bundles 2, 4 and 6

during Experiment LP-02-6 .......................................... 55

33. Fuel cladding temperatures in the peripheral bundles 2, 4, and 6

during Experiment L2-5 ............................................. 57

xi
34. LOFT fuel rod external thermocouple ................................ 57

35. LTSF single rod quench test thermocouple response for initial

cladding ........................................................... 60

36. LTSF test nine-rod bundle configuration ............................ 60

37. Cross section of FEBA and REBEKA fuel rod simulators ......... ...... 62

38. REBEKA and FEBA heater rod 7 temperature responses for runs IA

(REBEKA) and IF (FEBA) ........................................ 63

39. Blowdown quench behavior of a REBEKA heater rod (from LTSF bundle

test) and a nuclear fuel rod (from PBF-TC tests) with external

cladding thermocouples ............. .............................. 63

40. Embedded and internal cladding thermocouple response in rod 04

(bare) and rod 02 (externally instrumented) during blowdown of

PBF test TC-4F ..................................................... 66

41. COUPLE/MOD5 calculation of fuel rod temperatures using FRAP-T5

boundary conditions originating from RELAP5 code .................. 68

42. Fuel centerline and cladding temperature at the 27-in elevation

on fuel rod 5D07 during experiment LP-02-6 ....................... 68

43. Fuel centerline temperatures at the 27-in. elevation during

Experiment LP-02-6 ....... . .............. ............... ......... 70

44. Measured and calculated with TRAC-PFI/MODI fuel centerline

temperature for experiment LP-02-6 at the 0.6858 m (27-in) core

elevation ............................... .......................... 71

45. Measured and calculated with TRAC-PF1/MODl cladding temperatures

for experiment LP-02-6 at the 0.6943 m (27-in) core elevation ...... 71

46. Fuel rod centerline temperature with and without cladding surface

thermocouple and cladding temperature measurement during

Experiment LP-LB-1 at the 27-in elevation ........................ 73

47. Fuel rod centerline temperature with and without cladding outer

surface thermocouple and cladding temperature measurement at the

43.8-in elevation ................................................ .73

48. Fuel centerline and cladding temperatures on rod 5104 at the

43.8 in elevation during Experiment LP-LB-I ....................... 74

49. Fuel centerline and cladding temperature on rod 5108 at the

43.8 in elevation during experiment LP-LB-1 ........................ 74

50. Fuel centerline and cladding temperature on rod 5114 at the

43.8 in elevation during experiment LP-LB-1 ........................ 75

xii
51. Comparison of cladding temperatures at the 27-in elevation in the

central fuel assembly during Experiment LP-FP-1 .................... 77

52. Cladding temperatures at the 27-in. elevation on rod 5D07 during

Experiments LP-02-6 and LP-LB-I .................................... 77

53. Calculated cladding temperatures for nuclear fuel rod,

cartridge-type and solid-type heater rods for LTSF single rod

blowdown quench test ............................................... 80

54. Comparison of measured Semiscale heater rod initial cooling rates

from LTSF single rod quench test with calculated nuclear fuel rod

initial cooling rates .............................................. 80

55. Calculated peak linear Semiscale heater rod power needed to

duplicate nuclear rod temperature response shown in Figure 53 ...... 82

56. Comparison of Semiscale core power used for Experiment S-06-3

with that required to simulate LOFT Experiment L2-3 ................ 84

57. Reactor liquid level during final reflood for Experiments L2-3

and L2-5 ........................................................... 84

58. Cladding temperatures during NEPTUN-I and NEPTUN-II experiments,

with and without external cladding surface thermocouples .. ...... 86

59. Cladding temperatures during NEPTUN-I and NEPTUN-II experiments,

with and without external cladding surface thermocouples .......... 86

60. Cladding temperatures of electric heater'rods and nuclear fuel

rods during quench experiments in Halden ......................... 88

61. Cladding temperatures measured at two different axial levels in

FEBA and REBEKA rod bundles ..................................... 88

62. Cladding temperatures measured in NEPTUN experiment with External

cladding surface (LOFT) thermocouple and with embedded

thermocouples .................................................... o90

63. Mass flow rates in the broken loop Cold leg during Experiments

L2-5, LP-02-6, LP-LB-1, and LP-FP-1 .................. ............. 92

64. LOFT ECCS simplified schematic diagram .............................. 94

65. LOFT accumulator and variable standpipe ............................ 95

66. Accumulator A liquid level during Experiments L2-5, LP-02-6, and

LP-LB-i ...................................................... 98

67. HPIS injection rates during Experiments L2-5, LP-02-6, and

LP-LB-i ............................................. . ............ 100

68. LPIS injection rates during Experiments L2-5, LP-02-6, and

LP-LB-i ..................... ......... ......................... 101

xiii
69. Temperatures measured at the lower end box during Experiments

L2-5, LP-02-6, and LP-LB-1 ......................................... 101

70. Cladding temperature in the central bundle during Experiments

L2-5, LP-02-6, and LP-LB-I ............................. ........... 103

71. Cladding temperature in the central bundle during Experiments

L2-5, LP-02-6, and LP-LB-1 ....................................... 103

72. Temperatures measured in the upper plenum during Experiments

L2-5, LP-02-6, and LP-LB-1 ......................................... 104

73. LP-FP-l upper head injection piping ................................ 106

74. LP-FP-1 CFM upper structure ..................................... .106

75. HPIS volumetric flow during Experiment LP-FP-1A ................... 107

76. Liquid temperatures at the lower end boxes during Experiment

LP-FP-1A .............. ............................. ............. 107

77. Cladding temperatures at the lower core region during Experiments

LP-FP-1 and LP-FP-1A ............................................... 109

78. Cladding temperatures in the upper core region during Experiments

LP-FP-1 and LP-FP-1A ............................................... 109

79. Cladding temperatures in the central core region during

Experiments LP-FP-l and LP-FP-1A ................................... 110

80. Cladding temperatures in the central fuel element during

Experiment LP-FP-1 .......................... .................. ... 110

81. Cladding temperatures in a peripheral bundle during Experiment

LP-FP-1 ............................................................ 111

Al. Measured and calculated with DRUFAN-2 cladding temperatures

during Experiment L2-5 ............................................. A-7

A2. Measured and calculated with DRUFAN-02 primary system pressure

for Experiment L2-5 ............................................... A-7

A3. Measured and calculated with DRUFAN-02 cold leg break flow for

Experiment L2-5 .................................................... A-8

A4. Measured and calculated with DRUFAN-02 hot leg break flow for

Experiment L2-5 ........ ......................................... A-8

AS. Measured and calculated with RELAP5/MOD2 primary system pressure

for Experiment LP-02-6 ....... ................................... A-11

A6. Measured and calculated with RELAP5/MOD2 peak cladding

temperatures for Experiment LP-02-6 ................................ A-11

xiv
A7. Measured and calculated with TRAC-PFI/MODI primary system

pressure for Experiment LP-02-6 ..................................... A-13

AB. Measured and calculated with TRAC-PFI/MODI cold leg break flow

for Experiment LP-02-6 ............................................. A-13

A9. Measured and calculated with TRAC-PFI/MODI peak cladding

temperatures for Experiment LP-02-6 ................................ A-14

A10. Measured and calculated with TRAC-PFI/MODl cladding temperatures in peripheral bundles for Experiment LP-02-6 ....................... A-14

TABLES

1. LOFT experimental approach to a large break LOCA at typical PWR

conditions and subsequent nuclear large break experiments .......... 3

2. LOFT - commercial PWR comparisons .................................. 12

3. LOFT PWR operating conditions prior to the large break

experiments ........................................................ 16

4. Large break experiments investigated ............................... 96

5. Accumulator configurations ......................................... 97

Al. Code application to LOFT LB experiments ............................ A-3

A2. Performance of current thermal-hydraulic system codes .............. A-4

xv

REVIEW OF LOFT LARGE BREAK EXPERIMENTS

1. INTRODUCTION

The Loss-of-Fluid Test (LOFT) facility in its form during the U.S.

Nuclear Regulatory Commission (NRC) sponsored experimental program and during

the OECD sponsored experimental program, had its beginnings in May 1967. At

that time the basic mission of LOFT was changed to be in compliance with the

developing emphasis in the nuclear industry to include engineered safeguards

in nuclear plant designs which would bring a nuclear plant to safe shutdown

condition following a loss-of-coolant accident (LOCA). Over the succeeding

two years the LOFT program objectives [1] and the facility design required to

meet those program objectives (2] were developed and finalized. The LOFT

program objectives were:

1. Provide data required to evaluate the adequacy of and improve the

analytical methods currently used to predict:

a. The LOCA response of large Pressurized Water Reactors (PWRs)

b. The performance of engineered safety features with particular

emphasis on Emergency Core Cooling Systems (ECCSs)

c. The quantitative margins of safety inherent in the performance

of the engineered safety features.

2. Identify and investigate any unexpected events or threshold(s) in

the response of either the plant or the engineered safety features

and develop analytical techniques that adequately describe and

account for the unexpected behavior(s).

3. Provide experience in the application of standards from the Division

of Reactor Development and Technology, and other standards and codes

generally applicable to large PWRs by their use and evaluation by

the LOFT Program. This objective was satisfied during design and

construction of the LOFT facility.

1
The new facility design was required, therefore, to model, as nearly as

possible, a "typical current generation large PWR primary coolant system,

reactor system, and ECCS."

The LOFT PWR design was based on the contention at that time that the

large break LOCA (double-ended offset shear of a primary coolant pipe) in a

cold leg of a PWR primary coolant system would provide the most severe test

of the ECCS. Consequently, the LOFT PWR design incorporated the intact

loop-broken loop concept wherein the intact loop was an operating loop with

active components and the broken loop was a "simulator loop" with inactive

components which simulated pressure differentials only [2].

Technology available to understand and calculate LOCA phenomena and

performance of the ECCS was extremely limited at the time LOFT became

operational. There were few other experimental facilities to aid in the

evaluation of ECCS performance and available computer codes were too

simplified to calculate complex LOCA phenomena that were theorized to occur.

The LOFT experimental program, therefore, was based on a step-wise approach

to an "integral" system large break LOCA with the core at typical commercial

plant power densities and with a fully operational ECCS at typical commercial

plant set points. The experimental approach to this LOCA is summarized in

Table 1. The integral system LOCA at typical PWR operating conditions was

achieved in Experiment L2-3. Subsequent large break LOCA experiments, also

listed in Table 1, were conducted to study LOCA phenomena and system response

for other sets of initial and boundary conditions, and to provide wide range

data for code development and application to commercial plant designs. The

six LOFT large break nuclear experiments, and the LOCA phenomena and ECCS

behavior in them, comprise the subject of this report. More detailed

information on the approach to the LBLOCA and the results of the NRC

sponsored experiments is contained in References 1 and 3. A summary of each

of the nuclear experiments is given in Section 3.

2
TABLE 1. LOFT EXPERIMENTAL APPROACH TO A LARGE BREAK LOCA AT TYPICAL PWR CONDITIONS AND SUBSEQUENT NUCLEAR

LARGE BREAK EXPERIMENTS

Experiments

Nonnucl ear LO Series

LO-2

LO-3

LO-3A

LO-3B

LO-3C

LO-4

LO-5

LO-8

LO-9

LO-10

Objectives (summarization)

Qualify components and

operational procedures for valves

simulating a pipe break.

Verify pressure suppression

system capability to withstand

structural loading.

Provide data to determine maximum

vent submergence in the pressure

suppression tank for complete RCS

blowdown

Vent submergence variations from 13.5 cm to

56.2 cm.

Coolant injection limited to 0.048 m3 through

each of two quick-opening blowdown valves (QOBVs).

Experiments included either one or both valve

openings. Injected mass is sufficient to test

dynamic loading capacity of the PSS.

Injected coolant temperature typical of PWR steady

state cold leg temperature.

Sionificant Plant Characteristics/Parameters

W Nonnuclear LI Series

L1-1 Verification of system and

component performance at less

than maximum severity blowdown

conditions.

Evaluate QOBV and PSS

performance.

Provide thermal-hydraulic data

for an isothermal LOCA with

maximum break area and no ECCS

operation.

Core simulator in place for pressure drop

simulations.

Break size reduced to one-half maximum size for

double-ended offset shear.

Primary coolant at typical PWR pressure and cold

leg temperature.

High intact loop flow resistance.

ECCS initiated after system thermal stabilization

to obtain data on hot wall effects.

L1-2

Core simulator in place.
TABLE 1. (Continued)

Experiments

Li-3

LI-3A

Li-4

Li-5

Objectives (summarization)

Provide thermal-hydraulic data

for an isothermal LOCA with ECCS

injection into the lower plenum.

Provide data on intact loop flow

resistance for comparison to

L1-2.

Provide thermal-hydraulic data

for an isothermal LOCA with ECCS

injection into the cold leg.

Provide data on PCS operation and

rapid depressurization with

borated water.

Provide data on ECC bypass and

mixing (in conjunction with

LI-3A).

Evaluate core mechanical response

to LOCA depressurization loads.

Evaluate system data with nuclear

core installed.

Provide isothermal base case LOCA

for comparison for nuclear LOCAs.

Provide operator training and

operating procedure verification

for nuclear test control with the

reactor shutdown.

Provide thermal-hydraulic data

for ECCS injection into the cold

leg.

Significant Plant Characteristics/Parameters

Low intact loop flow resistance.

Core simulator in place.

Low intact flow resistance.

Core simulator in place.

Borated water used for the first time.

Nuclear core installed, all control rods in and

coolant boration > 3000 ppm.

Low intact loop resistance. This condition

maintained for all subsequent nuclear large break

LOCA experiments.

Pressure and temperature conditions typical of PWR

hot stand by conditions.

Unpressurized fuel rods.
TABLE I. (Continued)

ExDeriments Objectives (summarization)

Nuclear L2 Series

L2-2

L2-3

Integral system large

°I break LOCA at typical

PWR conditions.

Provide data to analyze.

1. DNB and RNB

2. Post-CHF heat transfer

3. Fuel rod thermal

MLIIGR of 26.2 kW/m

4. Thermal-hydraulic for 67%

nominal hot-leg-to-cold

leg OT.

5. ECCS performance and

system refill, reflood,

and core quench.

Provide data on core-wide and

spatial variations of fuel rod

cladding thermal response.

Provide thermal-hydraulic data

for a large break LOCA at typical

PWR steady state operating

conditions to identify phenomena

and effects on fuel rod cladding

thermal response.

Provide data on ECCS performance,

system refill and reflood, and

core quench for these LOCA

conditions.

Provide data to determine

conservatisms in Appendix K

assumptions for LOCA from typical

PWR operating conditions and for

the case of early ANB

suppression.

Significant Plant Characteristics/Parameters

Configuration identical to L-5 except for nuclear

core power generation and system initial

conditions in the core-to-SG inlet region.

System configuration identical to L2-2 except for

reactor power increase to a MLHGR of 39.4 kW/m.

L2-5 System configuration identical to L2-2 except for

reactor power MLHGR of 40.1 kW/m.

Early PCP trip with flywheels disconnected for

rapid coastdown.
TABLE 1. (Continued)

Experiments

OECD-LOFT Program

LP-02-6

LP-LB-I

Objectives (summarization)

Provide T/Il data on LOCA with

design basis boundary conditions

on both unpressurized and

prepressurized fuel rods to

determine safety margins in ECC

licensing.

Maximize the fraction of the core

that has not rewet by the end of

blowdown.

Provide system and core

thermal-hydraulic data for a LOCA

under these conditions.

Provide data on reflood

initiation of high temperature

and significant downcomer head

conditions.

Obtain data on the release of

fission products form the fuel

cladding gap both into the vapor

environment in the core during

heatup and into water after the

subsequent reflood.

Obtain data on fission product

transport through and out of the

reactor coolant system in the

vapor environment.

Significant Plant Characteristics/Parameters

System configuration identical to L2-2 except for

CFM which had all but outer row fuel rods

prepressurized to 2.41 MPa.

Boundary conditions included loss-of-offsite power

coincident with LOCA initiation and United States

minimum ECC injection assumptions.

System configuration identical to L2-2.

MLHGR of -52 kW/m.

Boundary conditions included loss-of-offsite

power coincident with LOCA initiation and United

Kingdom (UK) minimum safeguard ECC injection.

PCP trip and flywheel disconnect within is of LOCA

initiation.

System configuration identical to L-2 except for:

(1) Special CFM containing 24 6-wt% enriched fuel

rods (all others standard enriched; 22 of the

6-wt% rods pressurized) and a flow shroud to

channel fission product release.

(2) Addition of an upper plenum ECC injection

nozzle.

(3) ECCS designed and scaled to FRG KWU 1300-MWe

reference plant ECCS.

(4) Addition of a special fission product

sampling and measurement system.

ECCS delayed until fuel cladding ballooning and

rupture occurred on 6% enriched fuel rods.

Fission product transport path was the broken loop

hot leg. Broken loop cold leg closed after

LP-FP-1
The large break loss-of-coolant accident (LBLOCA) data from the LOFT

experimental program represents the only integral system data from a facility

that has a nuclear core. As such, the LOFT data is instrumental in the

current USNRC effort to quantify calculational uncertainty for the

thermal-hydraulic system codes. The LOFT system response to the simulation

of a LBLOCA revealed phenomena which had not previously been observed.

Specifically, the first nuclear experiment, L2-2, showed significant fuel

cladding cooling during blowdown which included a complete quench.

Subsequent analyses, the results of which were experimentally confirmed,

showed that this phenomena was linked to system hydraulics inclusive of the

operating characteristics of the primary coolant pumps. Section 4 includes

an analysis of this phenomena.

Also included in Section 4 is an extensive review of separate effects

experiment results in concert with the evaluation of a spectrum of LOFT

LBLOCA experimental data which addresses the question of the validity of the

observed blowdown cooling phenomena. Questions arose concerning this data

because (1) current generation codes cannot predict the degree of cooling

observed during blowdown, (2) externally mounted cladding thermocouples are

known to introduce fin cooling effects, and (3) the LOFT nuclear rod

temperature data differed significantly from electric heat rod data. In

summary, analyses of the available data have led to the conclusion that the

observed blowdown cooling phenomena in the LOFT LBLOCA experiments actually

occurred. Further, the externally mounted thermocouples do have some bias

that results in lower measured temperatures relative to bare fuel rod

cladding temperature. An appropriate correction must be applied to both

blowdown and reflood cladding temperature data. Finally, the external

thermocouples did influence but did not mask the fuel cladding temperature

response to the system hydraulics. The appropriate correction factors to be

applied to the temperature data have not been resolved as yet; however, in

view of the conclusions reached, the LOFT LBLOCA experiments and the

phenomena within them are discussed in Section 3 using measured fuel cladding

temperature data. Determination of the correction factors will not affect

the interpretation of the phenomena observed.

7
2. THE LOFT FACILITY

The LOFT Experimental Facility was a 50 MW (th) pressurized water reactor

system designed to simulate the major components and system responses of a

commercial PWR during postulated loss-of-coolant accidents (LOCA) and

anticipated transients. The LOFT Experimental Facility shown in Figure 1 is

described in detail in Reference 2. The facility consisted of five major

systems: the Reactor System with the nuclear core (Figure 2); Primary

Coolant System; Blowdown Suppression System; Emergency Core Cooling System;

and a Secondary Coolant System. These systems were instrumented to measure

the behavior of system parameters during the experiments.

The Reactor System included a nuclear 1.68 m long core arranged in nine

fuel rod bundles (four triangular and five square). The LOFT fuel was

designed to have the same physical, chemical and metallurgical properties as

commercial fuel.

The Primary Coolant System consisted of an operating loop (with a steam

generator, two primary coolant pumps in parallel, pressurizer and connecting

piping) representing three intact loops of a four loop PWR, and a "broken

loop" which simulated the broken loop of a four loop PWR during LOCA

conditions. The broken loop consisted of hot and cold legs which connect the

Reactor System to the Pressure Suppression System, and are equipped with

steam generator and pump simulators and quick-opening blowdown valves. The

piping arrangement was variable to simulate hot or cold leg breaks.

The LOFT Blowdown Suppression System was designed to simulate the

containment back pressure in large PWRs during LOCA events. It consisted of

a large pressure suppression tank, downcomers and a header connected to the

primary system via the quick-opening blowdown valves.

The emergency core cooling system (ECCS) consists of the same three

systems currently in commercial PWRs -- the high pressure injection system

(HPIS), the accumulator, and the low pressure injection system (LPIS). The

systems are actuated similarly to their generic counterparts and inject

8
Intact loop ,A

BL.1

experimental

measurement

Broken loop

Figure 1. Axonometric projection of LOFT system configuration for

Experiment L2-2.
;OVO O0Q 90909;

900- -00000000000ee, 9000 ee ee~Ooo.9ee@ae'

E)G) /ee @eee Woeeeee oegeoeei

0 Broken loo

hot t"g

Soeee~e

)0E~eee@0e

kv ... ..... A

Core key

N

Idonlftcattof key

Guide tube

Rod

GUide tube

eeeeseoe Eeseee o~eeeea0 oe0eooeoee

00@ @eegeae300o Se oo00M -e000o4e09.0@O0o 0@deOeeeoo 09909@d09000

deo99oooeee~ede te9e~ e9e. e09G) 0%eoce~9eaeeoe aeeeeoee0eG~eOidBeeo

~eeee~eeeeeeeeeseee9eoeee eeeeeee ee ~ee A Asoano toP0

00 0 ~ceea ~ L e~9~ ~eeee ~ceee eeeeE ~ee ~'o ~' *ee003seeeoeoer0 3E98o30oo,

Intact loop

hot log

,I'0ee90- e@e~eeeoeeeee \oe94996566

Ot 00O

AA

Oe09000000000

oeoooeeooeeoee

e

Intact fooW

I cold lg

0 A 1 3?07LA, 

-Control rod

drive mechanism

Fuel module

Instrument

penetration

Reactor vessel

Downcomer

instrument

stalk (2ST) -

Core support barrel

- Fuel module upper

support structure

-Reactor vessel

filler assembly

-Downcomer

-Fuel bundle

* Lower core

support structure

"th / Broken loop

hot leg

270" INEL.L9IL15 000

Figure 2. LOFT reactor vessel and core.

10
scaled amounts of emergency core coolant (ECC) typical of the ECC delivery

behavior in commercial PWRs. The LOFT ECCS has the capability of injecting

ECC to any of several locations including the intact loop hot or cold legs,

and the reactor vessel downcomer, lower plenum, or upper plenum. An

identical backup ECCS is also available which functions separately from the

ECCS used in a LOCE.

The LOFT Secondary System was designed to remove the heat transferred

into the steam generator to the environment. This system however could not

be controlled for full simulation of secondary system response in large PWRs.

The component and system volumes of the Reactor and Primary Coolant

Systems were designed proportional to their respective volumes in a

commercial PWR with the ratio of the two system powers as the proportionality

factor [4]. The design objective for the LOFT facility was to produce the

significant thermal hydraulic phenomena with approximately the same

conditions and sequence of events that could occur during postulated

accidents in commercial PWR systems. The LOFT scale model of the generic PWR

that resulted is summarized in Table 2 which contains comparisons of

geometric and physical parameters between LOFT and commercial PWRs. The

physical parameters listed are nominal operating conditions in the

Westinghouse 4-loop ZION PWR and in the LOFT system prior to the LOCE

designated L2-3.

The LOFT reactor core is about one-half the length of typical reactor

cores in commercial plants. However, this is the only compromise made in the

nuclear fuel for the LOFT core. PWR 15x15 array fuel rod assemblies are used

in the geometry shown in Figure 2. The triangular corner assemblies are

partial square assemblies for simulation of a more circular core. The outer

four square fuel assemblies have reactor control rods in the guide tubes.

The center fuel assembly is the most heavily instrumented assembly with

instruments placed in the vacant guide tubes as well as on the fuel rods.

The LOFT fuel assemblies are complete with upper and lower end boxes and fuel

rod spacer grids at five elevations.

11
TABLE 2. LOFT - COMMERCIAL PWR COMPARISONS

Item LOFT TROJAN

Volume (m3 ) % Total % Total Volume (m3)

Reactor Vessel

Outlet Plenum 0.95 12.51 15.95 55.47

Core and Bypass 0.31 4.12 7.50 26.05

Lower Plenum 0.71 9.32 8.58 29.73

Downcomer and

Inlet Annulus 0.69 9.00 5.89 20.42

Subtotal 34.95 37.95

Intact Loopa

Hot Leg Pipe 0.35 4.60 1.94 6.71

Cold Leg Pipe 0.37 4.85 2.08 7.22

Pump Suction Pipe 0.33 4.38 3.09 10.70

Steam Generator 1.45 18.97 26.40 91.49

Pump 0.20 2.60 1.96 6.80

Subtotal 35.40 35.47

Broken Loop b

Cold Leg to Break 0.16 2.16 1.72 5.97

Vessel to Steam

generator 0.15 1.98 0.65 2.24

Steam Generator 0.52 6.88 8.80 30.50

Pump 0.05 0.72 0.65 2.27

Additional

Volume Part 0.19 2.46 N/A N/A

of Outlet Plenum

Additional Volume

Part of Inlet

Plenum 0.22 2.83 N/A N/A

Subtotal 17.03 11.17

Pressurizer 0.96 12.62 14.7 50.97

Total 7.63 100.00 100.00 346.60

12
TABLE 2. (Continued)

Item

Core (Loft L2-3, ZION nominal

conditions included)

Fuel rod number

Length (m)

Inlet flow area (T2)

Coolant volume (mn)

Maximum linear heat generation

rate (KW/m)

Coolant temperature rise (K)

Power (MW)

Peaking Factor

Power/coolant volume (MW/m 3)

Core volume/syste• volume

Mass flux (Kg/s-mr)

Core mass slow/system volume

(Kg/s-m )

LOFT ZION

1300

1.68

0.16

0.295

39.4

32.2

36.7

2.34

124.4

.038

1248.8

39372

3.68

4.96

20.227

39.4

32.2

3540.5

1.60

175.0

.057

3707.3

25.6 51.7

a.

b.

TROJAN values are for three loops combined

Includes pump suction piping

13
The LOFT facility is augmcnted with an extensive "experimental"

measurements system (2) in addition to the normal PWR instrument systems for

reactor operation and control; State measurements of the coolant in the

primary system provide the capability of following the redistribution of mass

and energy in the primary coolant system following the initiation of a

transient. Extensive thermal measurements in the nuclear core provide

detailed information on the thermal response of the fuel cladding. Nuclear

measurements in the core assist in determining the initial or steady state

energy distribution. The philosophy followed on measurement locations in the

nuclear core, as shown in Figure 2, was to instrument one-half of the core on

a circular symmetry basis with emphasis on the center fuel assembly.

14
3. LOFT NUCLEAR LARGE BREAK EXPERIMENTS

Six nuclear large break LOCA experiments were conducted in the LOFT

facility as described in Table 1. Three of the experiments, L2-2, L2-3 and

L2-5 were conducted as a part of the U.S. NRC LOFT Program. The other three

experiments, LP-02-6, LP-LB-1, LP-FP-1 were conducted as part of the

international OECD LOFT Program. The principal phenomena and events in these

large break LOCA transients are summarized in this section. The initial

conditions for these experiments are summarized in Table 3. Additionally to

these six experiments we will discuss the experiment LP-FP-1A. This was

aborted experiment LP-FP-1. Although it was only short blowdown, data of

this experiment were very valuable to understand the thermal-hydraulic

processes in the actual experiment LP-FP-1.

Along with the experiment description, a review of experiment predictions

and post-experiment calculations are presented. Appendix A summarizes some

recent code analyses of the LOFT large break LOCA experiments and gives a

brief description of the three codes used in these analysis: RELAP5, TRAC

and DRUFAN.

3.1 Experiment L2-2

The basic objective of this first nuclear large break experiment [5],

which was conducted in December 1978, was to provide integrated system data

on thermal-hydraulics and fuel behavior during a 200% cold leg break

(double-ended offset pipe shear) LOCA. The configuration of the facility for

this experiment is shown in Figure 1. The experiment was conducted from 50%

power (25 MW, 26.38 kW/m maximum linear heat generation rate), a

specification that resulted from the planned stepwise approach to the large

break LOCA at typical PWR operating conditions. Prior to experiment

initiation, the reactor was operated at steady state to build in decay heat

equivalent to approximately 90 percent of that for infinite operation at the

initial condition power level. This criteria was used for all LOFT nuclear

experiments.

15
TABLE 3. LOFT PWR OPERATING CONDITIONS PRIOR TO THE tARGE BREAK EXPERIMENTS

Parameter

Primary Coolant System

L2-2 1.2-3 1.2-5

Mass flow rate (kg/s)

Pressure (MPa)

Cold leg temperature (K)

Hot leg temperature (K)

Boron concentration (ppm)

Reactor Vessel

Power level (MW)

Maximum linear heat generation rate (kW/m)

Control rod position (above full-in position) (m)

194.2

15.64

557.7

580.4

838

24.88

26.37

1.37

0.353

0.607

619

15.62

1.089

199 + 6.3

15.06 _ 0.03

560.7 + 1.8

592.9 + 1.8

679 + 4

36.0 + 1.0

39.0 + 3.0

1.37 + 0.01O

192.4

14.94

556.6

589.7

668

+ 7.8

± 0.06

7 4.0

+ 1.6

+ 15

LP-02-6

248.7 + 2.6

15.09 + 0.08

555.9 + 1.1

589.0 + 1.0

512 + 15

46.0 + 1.2

48.8 + 3.6

1.381 ± 0.0002

0.39 + 0.02

0.607 + 0.02

615.6 ± 5.8

15.3 ± 0.11

1.04 - 0.04

305.8 +

14.90 ±

556 ±

587.8 +

513 +

2.6

0.08

1

1.2

15

486.7

14.77

563.2

577.6

612

+

+

+

+

+

2.5

0.07

1.1

0.8

15

LP-LB-I LP-FP-I

36.0 + 1.2

40.1 + 3.0

1.376 _ 0.010

49.3

51.7

1.455

+

4-

+

1.2

3.6

0.0002

37.0 +

51.2 +

1.38 '

1.2

3.6

0.002

Pressurizer

Steam volume (m3)

Water volume (m3

Water temperature (K)

Pressure (MPa)

Liquid level (m)

ay Broken Loop

Hot leg temperature (K)

Cold leg temperature (K)

Steam Generator Secondary Side

0.293

0.670

615.3

15.06

1.19

+ 0.008

± 0.008

3.0

± 0.03

+ 0.01

0.32 +

0.61 +

615.0 ±

1.14 +

(aT

0.02

0.02

0.3

0.3

0.37 t 6.02

0.56 + 0.02

615 ± 5.8

14.92 ± 0.11

1.04 + 0.04

0.27

0.66

616.2

14.73

1.23

+

+-

+

+

+

0.02

0.02

5.8

0.11

0.04

561.2 565.5 + 1.8 561.9 + 4.3

555 554.3 7 1.8 554.3 - 4.2

560 + 6

553 + 6

561 + 6

552 + 6

564.8 + 1.8

561.4 + 1.5

3.10 + 0.06

548.5 _ 1.5

6.41 ± 0.08

19.0 ± 0.4

Water level (mn)

Water temperature (K)

Pressure (MPa)

Mass flow rate (kgls)

3.14

553

6.35

12.67

3.11 + 0.025

482.1 + 3

6.18 + 0.08

19.5 + 0.4

(a)

547.1 ± 0.6

5.85 ± 0.06

19.1 0 0.4

3.28 + 0.6

538 ± 1.5

5.63 + 0.2

24.3 + 0.4

3.19 + 0.02

528 + 1.5

5.53 _ 0.02

25.4 + 0.4

(a) Measurement failed.
The experiment was initiated by opening the quick-opening blowdown

valves. The reactor was scrammed automatically on low system pressure in

1.7 s. The ECC flow was directed into the intact loop cold leg, beginning

with HPIS flow at 12 s after blowdown initiation and accumulator injection at

18 s at 4.2 MPa primary system pressure. The primary coolant pumps were

operating during the blowdown and were tripped at 200s. Detailed experiment

results are presented in Reference 6.

While the hydraulic behavior of the system was approximately as

predicted, the'thermal behavior of the core was surprising. The cladding

temperature rose initially as predicted but an unpredicted core wide cooling

started at about 5.5 s leading to a complete core quench. The maximum core

temperature of 789K was reached during the initial temperature rise.

Figure 3 shows the measured and predicted fuel cladding temperatures for the

center fuel module.

The early rewet was caused by resumption of positive core flow when the

broken loop cold leg break flow transitioned from subcooled to saturated

critical at about 3.4 s which resulted in a decrease of the mass flow rate.

At this time the cold leg coolant in the operating loop was still under

subcooled conditions with pumps operating. This condition resulted in an

increase in reactor vessel coolant inventory as more coolant flowed into the

vessel than flowed out (Figure 4). This additional fluid was carried, due to

the positive core flow, through the core causing a complete fuel rod rewet at

5.5 s.

The large difference between the predicted and measured core thermal

behavior was concluded to be caused by at least one and possibly all of the

following:

- inadequate modeling of initial fuel stored energy

- inadequate heat transfer models

- insufficiently accurate system hydraulic calculations

- presence of thermocouples on the surface of fuel cladding

17
2

E

,2

1000

900

800

700

600

500

400

Figure 3.

0 5 10 15 20

Time after rupture (s) L221-KM258-01

Predicted and measured fuel rod cladding temperature near

midplane of center fuel module for Experiment L2-2.

U,

C)

Co

4-

Co

0

CO

C',

Co

750

625

500

375

250

125

0

0 2 4 6 8

Time after rupture (s)

10

L221-KM258-02

Figure 4. •~:z •I, ratas in cold legs of intact and broken loops during

Experiment L2-2.

18
The problem of inadequate modeling of initial fuel stored energy was

eliminated as a possibility after inspection of the experiment data which

showed that the difference was primarily caused by heat removal mechanisms

rather than the heat supply mechanisms.

Detailed examination [7] of the predictions revealed that the code

calculated in the 3 to 10 s interval a stagnation at the core outlet, whereas

the experiment data indicated strong upward flow. Comparisons made of the

predicted and measured mass flow rates in the intact loop cold leg showed

reasonable agreement between prediction and calculations. The comparison of

break mass flow rate showed however strong overprediction of the cold leg

break flow by as much as 120 kg/s. Overprediction of the break flow was

responsible for miscalculation of core hydraulics.

3.2 Experiment L2-3

Experiment L2-3, the second nuclear large break LOCA experiment, was

performed in May 1979. The major objective [5] of this experiment was to

measure thermal-hydraulic phenomena and effects on fuel rod cladding thermal

response for a higher initial power level. The LOFT system configuration for

this experiment was identical to the configuration for experiment L2-2 with

the exception that the power was increased to 36.7 MW (39.4 kW/m maximum

linear heat generation rate). The core power density, the core coolant

temperature rise and the system pressure corresponded to typical operating

conditions in a pressurized water reactor power plant.

The experiment was initiated by opening the quick-opening blowdown

valves. Reactor scram was completed 1.7 seconds later, HPIS injection was

initiated at 14 s, accumulator injection at 17 s at 4.18 MPa system pressure,

and LPIS injection at 29 s. The core was reflooded at 55 s. During this

experiment, as in experiment L2-2, the primary coolant pumps operated

throughout the experiment and were tripped at 200 s. Detailed results of

this experiment are presented in Reference 8. The core thermal response in

this experiment was similar to the core thermal response in the

Experiment L2-2, (Figure 5). The differences in the temperature magnitudes

19
2

C)

C.

E

C,

1000

900

800

700

600

500

400

-- LOCE L2-2

LOCE L2-3 ....... 

..................

:,°-.,

.. °°.o°

I I

H

-10 0 10 20 30

Time after rupture (s)

40 50 60

L221-KM258-03

Figure 5. Temperature of fuel rod cladding during Experiments L2-2 and

L2-3.

20
and event occurrence times is consistent with the core power and initial core

fluid temperature rise. Measured peak cladding temperature was 789K in

Experiment L2-2 and 914K in Experiment L2-3. In both experiments the maximum

peak cladding temperature occurred during the blowdown phase. In both

experiments an early cladding rewet was observed which was caused by the same

phenomena. As in Experiment L2-2 the transition from subcooled to saturated

critical flow in the broken loop cold leg occurred about 2 s earlier than

the fluid saturation in the cold leg of the operating loop resulting in more

coolant being delivered to the downcomer than removed. This additional

coolant traversed the core from bottom to top and caused fuel cladding rewet

for a short period of time.

The experience from the sensitivity studies performed after Experiment

L2-2 was applied to the prediction of the Experiment L2-3, however the early

rewet again was not predicted (Figure 6). Analyses [9] have shown that the

difference between the predicted and measured fuel cladding temperatures is a

result of post-critical heat transfer modeling in the core. The analyses

indicated that the critical heat flux (CHF) correlation used in RELAP4/MOD6

is not suitable for LOFT cases where the rewet occurred at core coolant mass

fluxes of 100 to 600kg/s m2 . An alternate approach, the Biasi CHF

correlation which covered the mass flux ranges from 100 to 6000kg/s m2 , was

built into RELAP4/MOD6. The calculation with this correlation, with measured

initial conditions, keeping all other modeling the same as for the

predictions showed better agreement with experimental data. Figure 6 shows

the prediction and posttest calculation of cladding temperature compared with

the measured cladding temperature.

The calculated maximum peak cladding temperature was approximately 70K

higher than measured in the experiment. Further study indicated that lack of

a fuel cracking model (from power ramping) in the RELAP code results in a

larger fuel gap width and consequently in higher stored energy and higher

peak cladding temperature. Calculations with the FRAP-T5 code, which

contains a fuel relocation model, using boundary conditions from RELAP4/MOD6

calculations, showed excellent agreement of calculated with measured cladding

temperature.

21
1400

1200

1000

800

CL

E

,T

600

400

-10 0 10 20 30

Time (s)

40

L221-KM268-04

Figure 5. Calculated and measured cladding temperature on fuel rod 5J4

at 0.53 m above bottom of the core for Experiment L2-3.

22
3.3 Experiment L2-5

The third large break LOCA experiment was performed in June 1982. This

experiment, designated L2-5, simulated a 200% break in the cold leg piping of

a commercial PWR simultaneous with loss of offsite power. This experiment

was designed to provide experimental data to demonstrate that Appendix K

assumptions result in a conservative prediction of peak cladding temperature,

even without early rewet. The configuration of the facility was identical to

the configuration used in experiments L2-2 and L2-3. The core steady state

power level was 36 MW (40.1 kW/m maximum linear heating rate). The reactor

was scrammed on a low pressure signal at 0.34 s. Following the scram, the

operators tripped the primary coolant pumps at 0.94 s. The pumps were not

connected to the flywheels in this experiment. This was done in order to

provide an early rapid pump coastdown which would prevent the early core

rewet phenomena and result in higher fuel cladding temperatures.

Reference 10 provides detailed experimental results for this experiment.

The core thermal response was quite different, and more complex, than in

Experiments L2-2 and L2-3. Figure 7 shows typical cladding temperature in

the lower half of the central fuel bundle compared to cladding temperature

measured in this region during the L2-3 experiment. The cladding

temperatures in L2-5 increased quickly in response to degraded cooling as in

the previous large break experiments. At about 5 s the temperature rise rate

decreased, and about 10 s after reaching approximately 1050K the cladding

temperature began to decrease slowly. 20 s after experiment initiation the

temperature increased again and reached a maximum of 1077K at about 30 s.

From this time on a gradual cooling of the cladding occurred in response to

the injection of ECC water. The fuel rod cladding was completely quenched by

65 s.

In the upper part of the central fuel bundle and high power regions of

the peripheral bundle the measured thermal response was quite different from

that in the high power region. As shown in Figure 8, initially the cladding

temperatures increased similarly to the high power region. However, at about

15 s after experiment initiation a strong top-down quench was measured which

23
1200

1000

2

:3

CO

12

800

600

400

0 20 40 60 80 100

Time (s)

120

1221-KM256-05

Figure 7. Maximum temperature

L2-3 and L2-5.

on fuel rod cladding during Experiments

2)

C,

E

a,

1000

900

800

700

600

500

400

0 20 40 60 80 100

Time (s)

Temperature of fuel rod cladding in upper half of

assembly during Experiments L2-3 and L2-5.

24

120

L221KM4258-O8

Figure 8. central fuel
lasted for a few seconds followed by another temperature excursion.

Subsequently, during reflood these core regions were quenched in the same

manner as the high power region.

The fuel cladding thermal behavior in the lower power regions of the

peripheral fuel bundles was also unique. The cooling was sufficient in these

regions to maintain the cladding at saturation until about 20 s (Figure 9).

This part of the core behaved similarly in L2-3 and L2-5 and experienced only

relatively small temperature excursions before final quench occurred from

ECCS reflood.

In summary the core thermal behavior in Experiment L2-5 was similar to

the core thermal behavior in Experiment L2-3 with the exception of the early

bottom-up core wide rewet which occurred in Experiment L2-3. This difference

was caused by the mode of primary pump operation and demonstrates the

sensitivity of fuel cladding temperature on hydraulic phenomena attributable

to primary pump operation.

Predictions for Experiment L2-5 [11] were performed using the RELAPS/MODI

code. Figure 10 shows the predicted and measured maximum temperature of the

fuel rod cladding in the core. The calculated maximum peak cladding

temperature was 1082 K in comparison to the measured maximum peak cladding

temperature of 1077 K; however, the maximum was calculated to be much earlier

than measured. This is in part due to the failure of the code to calculate

the top down quench which was measured in the upper regions of the core.

Also the later heatup in the peripheral fuel bundles was not calculated.

In the post L2-5 recovery phase, a potential reactor operating procedure

was studied wherein the reactor vessel liquid level was to be maintained

below the reactor vessel nozzles and above the core. The reactor operators

were required to cycle the HPIS and LPIS flows to control the coolant level

based on observation of temperature measurements at several elevations in the

upper plenum. Temperatures observed by the operators and also the coolant

temperature measured at the core exit are shown in Figure 11. These

temperature measurements did not indicate that core heatup, or inadequate

25
650

600

CO

E

0,

550

500

450

400

0 20 40 60 80 100

Time (s)

120

L221-KM258-07

Figure 9.

1200

1000

Temperature of fuel rod cladding

during Experiments L2-3 and L2-5.

in a peripheral fuel assembly

Q

CO

Q)

E

C,

800

600

400

0 20 40 60 80 100

Time (s)

120

L221-KM258-08

Figure 10. Predicted and measured maximum temperature of fuel

for experiment L2-5.

26

rod cladding
a,

E.

4-

600

550

500

450

400

350

0 100 200 300 400

Time (s)

500

L221-KM25 8-09

Figure 11. Core exit and upper plenum fluid temperatures compared with

saturation temperature during Experiment L2-5.

27
core cooling, had occurred until about 370 s. Figure 12 shows that

inadequate core cooling occurred before 200 s and that the core was in

significant heatup prior to 370 s. These results clearly show that upper

plenum coolant temperature measurements are not adequate for liquid level

control and maintenance of adequate core cooling.

3.4 Experiment LP-02-6

Experiment LP-02-6, the fourth nuclear large break experiment was

conducted in October 1983. The major objective of this experiment was to

provide data to assess the capability of computer codes to predict PWR system

response during a design basis accident [12] . This experiment simulated a

double ended offset shear of a commercial PWR main cold leg coolant pipe.

The initial conditions for this experiment were representative of USNRC

licensing limits in commercial PWR and included loss of offsite power

coincident with LOCA initiation and minimum United States emergency core

coolant injection. The experiment was initiated from a power level of 46MW

(49kW/m maximum linear heat generation rate). Prepressurized fuel was used

in this experiment in the central fuel assembly. Results of this experiment

are discussed in Reference 13.

The experiment was initiated by opening the quick-opening blowdown valves

in the broken hot and cold legs. The reactor scrammed automatically when the

hot leg pressure reached 14.8MPa at 0.1 s. The primary coolant pumps were

tripped and allowed to coast down until 16.5 s, when they were disconnected

from the flywheels. The flow in the core reversed almost instantaneously

with experiment initiation, and the fuel rod cladding experienced DNB within

I s after experiment initiation. The cladding temperature increased until

about 5 s when the positive core flow was reestablished. Figure 13 shows

cladding temperatures at four different elevations in the core CFM for the

first 60 s of the transient. The data show that in the first 10 s the fuel

cladding was quenched in the lower 2/3 of the core while the upper part of

the core was cooled but not quenched. This is quite different core thermal

response from that observed during experiment L2-5 in the same time frame.

The difference in thermal response is the result of different pump operation

28
1200

1000

0.

E

a)

800

600

400

200

0 100 200 300 400

Time (s)

500

L221-KM258-10

Figure 12. Temperature of fuel cod cladding at different elevations during

the second core heatup in Experiment L2-5.

1200

1000

C2

E

800

600

400

0 10 20 30 40 50

Time (s)

60

L221-KMd258-13

Figure 13. Fuel rod cladding temperatures at different elevations in the

center fuel module during Experiment LP-02-6.

29
modes in these experiments. During the L2-5 experiment the primary pumps

coasted down almost immediately, while in experiment LP-02-6 the pump speed

was almost constant for the first few seconds of the transient. Because of

the short coastdown during L2-5 the early quench at the hottest elevation in

the core is not observed. Figure 14 compares the maximum cladding

temperatures measured during these two experiments and experiment L2-3. The

core thermal response in experiment L2-3 is similar to the response in

experiment LP-02-6 but with reduced peak cladding temperatures as a result of

continuous pump operation and lower initial power level. The difference in

pump operation is the reason that the bottom up quench was core wide in the

L2-3 experiment and extended only for 2/3 of the core in experiment LP-02-6.

A second DNB began at about 10 s (Figure 13) but at about 15 s a top down

quench was initiated which extended over the upper 1/3 of the core. CHF and

heatup occurred again in this region at about 20 s and at about 30 s the

entire core was above saturation temperature. ECCS reflood quenched the core

at 56 s. The maximum cladding temperature reached during the reflood phase

was 840K.

One of the major objectives of this experiment was to determine whether

fuel rod damage would occur during a design basis accident for unpressurized

and for prepressurized fuel rods. Therefore, a center fuel module (CFM)

containing prepressurized fuel (2.41 Mpa) was used. Post-experiment fluid

samples taken from the primary system indicated no fission products in the

coolant and fuel rod plenum pressure measurements did not indicate cladding

rupture. Analysis [13] also indicated no fuel rod failure and no appreciable

fuel rod ballooning.

Predictions [14] for this experiment were made using the TRAC-PD2/MODI

and FRAP-T6/MODI computer codes. The FRAP code was used to analyze in detail

the thermal fuel rod response using TRAC-calculated thermal-hydraulic

conditions. Figure 15 shows the TRAC-calculated and measured peak cladding

temperature. The initial cladding heatup is calculated very well but the

calculations indicate only relative slow cooling of the hot spot after 5 s

whereas the data show a rapid quench. The lack of a quench in the

calculation is responsible for higher peak temperatures after blowdown and

30
2

C)

E

93

1300

1200

1100

1000

900

800

700

600

500

400

0 10 20 30 40 50 60

Time after rupture (s)

70

L221.KM266.'4

cjure 14. Peak fuel rod cladding temperatures duri~r Experiments L2-3,

L2-5, and LP-02-6.

1200

1000

800

-- TE-5L06-026 ........ TRAC

............ .. °........... ............

2

(D

E

a,

600

400

0 10 20 30 40 50

Time after rupture (s)

60

L221-KM258 -15

si-..re 15. Measured and p:... fual cladding temperatures fc.-

Experiment LP-02-6.

31
also partially for delayed final quench. The time to core quench was also

affected by the low minimum film boiling temperature calculated by TRAC. The

comparison of measured and calculated reactor vessel cold leg flow balance

(Figure 16) also shows some deficiencies in the calculations. The initial

coolant depletion from the reactor vessel is calculated to be smaller than

measured and reestablishment of coolant addition to the reactor vessel occurs

later and is smaller in magnitude than measured. The core thermal behavior

is very sensitive to small differences in hydraulic phenomena as the vessel

mass balance data indicate. The accuracy in prediction of the hydraulic

conditions in the core must be increased for appropriate assessment of

low-flow film boiling models that are used for calculation of thermal core

behavior during the early quench time.

3.5 Experiment LP-LB-1

Experiment LP-LB-1, the fifth large break experiment in LOFT was designed

to reproduce conditions representative to United Kingdom licensing limits

[15]. This experiment simulated a double-ended offset shear of one inlet

pipe in a four loop PWR. The experiment was initiated from conditions

representative of a PWR operating near its licensing limits. The boundary

conditions for this experiment were set to simulate loss of offsite power

coincident with LOCA initiation and United Kingdom minimum safeguard

emergency core coolant injection. These assumptions resulted in utilization

of 70% of the accumulator volume and 50% of the pumped ECC injection of that

used in the LP-02*6 experiment which represented the U.S. licensing limits.

An early rapid primary coolant pump coastdown was included to attain maximum

cladding temperatures by suppression of the early rewet phenomena.

The experiment was initiated by opening the blowdown valves from a core

power level of 49.3 MW (51.7 kW/m maximum linear heat generation rate)[16].

The reactor was scrammed on a low pressure signal at 0.13 s and the primary

pumps were tripped slightly later and disconnected from the flywheels at

0.63 s. The fuel cladding went into DNB in less that 1 s in the high power

region. The early decoupling of the primary pumps from their flywheels

resulted in insufficient flow into the vessel from the intact cold leg to

produce a bottom up flow into the core and an early fuel cladding quench that

32
300

- Data

200 ........ TRAC

100

0 .. .. .... " 

........ V -----.....

0 -1000

U) -200

-300

-4°°0

.500 i I I I I

0 2 4 6 8 10

Time after rupture (s) L221-KM258-1O

Figure 16. Measured and predicted differences between intact and broken cold

leg mass flc;.,is for Experiment L2-6.

33
occurred in experiments L2-2, L2-3, and LP-02-6. The rapid cladding

temperature rise had stopped at about 13 s because of liquid fallback from

the upper plenum. This top-down liquid flow resulted in quench of the upper

part of central fuel assembly (Figure 17) and more extensive cooling in the

peripheral fuel bundles. The maximum cladding temperature during the

blowdown phase was reached shortly before the top-down cooling trend started

and reached 1261K (Figure 18). The top-down cooling lasted until about 25 s

when fuel cladding heatup started again. ECC injection from the accumulators

began at 17 s and from the LPIS at 32 s and resulted in a core quench at

about 34 s. The core quench started at both bottom and top of the core and

progressed toward the peak power elevation and was completed at 72 s. The

maximum cladding temperature recorded during the ECC injection phase was

1257K.

One of the concerns was whether cladding damage or deformation would

occur on the unpressurized fuel rods used in the LP-LB-1 experiment. The

very high temperatures reached in this experiment would cause cladding

structural weakening and possible cladding collapse onto the fuel pellets.

Analysis [16] and coolant samples indicated that cladding was not ruptured

but possibly deformed.

The TRAC-PD2/MOD1 code was used to predict this experiment. The version

of the code which was used contained an error in the gap conductance model

which was believed would affect significantly the calculated cladding

temperatures. Immediately after the experiment posttest calculations were

performed using a corrected version of the code and measured initial and

boundary conditions. Figure 19 shows the pre- and posttest calculation

results for peak cladding temperature compared with the measured peak

cladding temperature. The predictions show the initial heatup rate in

agreement with experiment data but after 4 s the calculated temperatures

deviate significantly from the data. These deviations are a direct result of

the code error. There are also significant differences between the

calculated and measured cooling in the reflood phase. In the experiment the

cooling rate increased as the reflood progressed. In the calculations the

cooling rate was reduced as the temperature difference between cladding and

fluid was reduced.

34
E

a,

1100

1000

900

800

700

600

500

400

"/ -,45 . in. K I. . . - 49 in.

54 in.

/,i,"•--L----62 in.

4"

TE-5G06-062

........ TE-5106-054

- TE-5H07-049 ............... TE-5G06-045

.- " ....... .7 

....

.......-

0 10 20 30

Time after rupture (s)

40 50

L221-KM268-17

Figure 17. Upper central fuel

Experiment LP-LB-I

assembly cladding temperatures during

showing top down quench.

CL

E

a,

1300

1200

1100

1000

900

800

700

600

500

400

0 20 40 60 80

Time after rupture (s) L221.•KM258.a

Figure 18. Peak cladding temperature during Experiment LP-LB-1.

35

r~.
CO

CL

E

a,

1300

1100

900

700

500

300

0 10 20 30

Time

40 50 60

after rupture (s)

70 80

PSOO-WHT-488-01

Figure 19. Measured and calculated

LP-LB-1. peak cladding temperature for Experiment

36
The initial heatup was calculated correctly in the post-experiment

calculations. The reduction of the heatup after 6 s was also calculated but

was slightly more extensive than in the experiment. The time of the peak

cladding temperature was calculated correctly though the calculated

temperature was approximately 60 K lower than the measured temperature

(1261 K). The posttest calculated thermal response during reflood was also

less than measured but was in better agreement with measured data than was

the pretest calculation.

In general, the code calculated the hydraulic conditions quite well with

the exception of the underpredicted depressurization rate during accumulator

injection. The code calculated the top-down quench during the blowdown, but

underpredicted the extent. The code also calculated properly the

simultaneous bottom-up and top-down quench during reflood. The strong,

hydraulically controlled azimuthal asymmetry measured in the thermal response

of the peripheral bundles was also partially calculated. The major

differences between the experiment and the calculations were in the

temperatures at the peak power location. The code did not calculate

correctly the initial cooling during blowdown, the peak cladding temperature,

and the cooling during reflood. These deficiencies in the calculations

indicate or reveal limitations of the post-critical heat flux models used in

the TRAC code.

3.6 Experiment LP-FP-1

Experiment LP-FP-1 was conducted in December 1984 and was

programmatically a fission product release and transport experiment [17].

However, the thermal-hydraulic boundary condition for this experiment was

based on a design basis large break LOCA in a FRG PWR. In LP-FP-1, the ECCS

was intentionally delayed until cladding rupture had occurred on fuel rods of

a CFM specifically designed for this purpose. The ECCS was designed to

simulate the FRG PWR ECCS with both hot and cold leg injection.

For this experiment the reactor core was equipped with a 15X15 center

fuel assembly with a thin zircaloy shroud which enclosed the inner 11xI1 fuel

rod array in which 24 of the fuel rods were enriched to 6-wt%. Twenty two of

37
these fuel rods were also prepressurized (2.41 MPa). The experiment was

designed to cause cladding ballooning and rupture to occur 60-90 s after

initiation. In order to accomplish this, experiment initiation began with a

reactor scram which was followed I s later by opening of the quick-opening

blowdown valves. This sequence removed sufficient stored heat to cause a

delay in reaching high temperatures. This delay was necessary to provide a

well-defined set of boundary conditions for fission product release and

transport. The primary coolant pumps were tripped and disconnected from

their flywheels I s after QOBV opening. This provided conditions similar to

experiment L2-5.

Figure 20 shows that the core thermal behavior was quite different from

the behavior observed in the previous large break experiments. The major

characteristic of the core temperature transient is that the expected early

cladding temperature rise was prevented by several quenches and the actual

core heatup started very late in the transient.

The first core heatup began at about 3 s and continued to about 6 s when

the first quench occurred (Figure 20). This was a bottom-up quench which

influenced only the lower half of the core. This quench was quite similar to

the early quenches observed in the experiments L2-2, L2-3 and LP-02-6. The

attempt to eliminate this early quench by tripping the pumps and

disconnecting the flywheels failed in this experiment. There are two reasons

for this. First, the primary coolant pumps were operated initially at higher

speeds than in experiments L2-5 and LP-LB-1 which resulted in higher initial

mass flow rate and fluid inertia. This resulted in more delivery of coolant

from the intact loop to the downcomer than in the other experiments. Second,

the reactor was scrammed before blowdown (intentionally) which removed some

of the initial stored heat.

At about 9 s a second quench started, this time a top-down quench. The

top-down quench was not uniform across the core as was the bottom-up quench.

It started near the intact loop hot leg in peripheral bundle No. 4 as a

result of liquid falling back from the hot leg intact loop. At about 10 s a

second heatup in the CFM started and was followed again by a quench at 12 s.

This quench was a top-down quench and propagated through the entire central

38
1250

1000

2

CO

a)

0.

E

a)

750

500

250

0 100 200 300

Time (s)

400

P590-WHT-488-02

Figure 20. Fuel rod cladding temperatures measured at the

elevation during Experiment LP-FP-1. peak power

39
bundle reaching the bottom at approximately 18 s. The cause of this quench

was premature injection of the ECC water initially stored in the ECCS

injection line routed to the upper plenum. Detailed analysis of this

injection and effects on core behavior are presented in Reference 18.

A third heatup of the core began at about 21 s. It was also not a

uniform core heatup. The top of the center fuel module remained at

saturation temperature. Also, in the peripheral assemblies most of the

thermocouples in the upper parts of the modules indicated saturation

temperature. The reason of this extended cooling of the upper parts of the

core was the unintentional injection of cold water from the upper plenum ECC

injection line. This cold water from the injection line resulted in complete

quench of the CFM between 23 and 26 s and also caused cooling to some parts

of the peripheral assemblies. Most of this unintentionally injected water

was evaporated before reaching the lower regions of the peripheral fuel

bundles.

The main cladding heatup in the central fuel assembly started at about

80 s and progressed from the bottom up. The heatup was not uniform radially;

fuel rods closer to the broken loop heated up later and quenched earlier.

This nonuniform thermal behavior of the center assembly was due to the

effects of the unintentional ECC injection.

There was a second unintentional injection at about 266 s which resulted

in additional cooling that effected the peripheral fuel assemblies more than

the center fuel assembly. Some of this water was able to penetrate to the

lower part of the core and cause cladding quenches.

At 344 s the experiment was terminated on a high temperature limit for

the peripheral bundles by ECC injection into the upper plenum and the intact

loop cold leg. The quench began at the top of the core followed by the core

bottom and then the high power region. The quench was not uniform through

the core. The CFM was quenched at 370 s. Fuel assemblies 2 and 6 did not

quench until 380 s. The maximum cladding temperature recorded in the CFM was

1210K and occurred at 347 s. This temperature is above the temperature

required for cladding ballooning and rupture. Posttest analysis revealed

that 8 fuel rods had ruptured cladding.

40
Predictions for this experiment were performed using TRAC-PD2/MOD1,

RELAP5/MOD2 and DRUFAN codes. IDetailed description of these calculations is

given in [19] for TRAC calculations and in [20] for RELAP and DRUFAN

predictions. Since an unexpected subcooled liquid injection occurred during

this experiment which significantly affected the reactor thermal response, a

full comparison of the measured data with the predictions is not meaningful.

Discussion of the predictions and of the experimental data is included in

Reference 20.

3.7ý Experiment LP-FP-1A

The first trial to conduct the experiment LP-FP-1 was carried out on

December 12, 1984. The Quick Opening Blowdown Valve (QOBV) in the cold leg

opened immediately but that inithe hot leg did not open. At about 10 s the

Plant Protection Systems (PPS)Iwas actuated with following ECCS realignment:

HPIS pump A, accumulator A andILPIS pump A were aligned to inject in the

lower plenum, while HPIS pump B, Accumulator B and LPIS pump B were aligned

to inject into the downcomer. HPIS flow started at 15 s and achieved full

capacity of 1.95 1/s at about 17 s. Accumulators began to inject at about

19 s. At about 30 s the LPIS pump started to inject and the core was

completely quenched at about 35 s. Most of the fluid thermocouples just

below the core have shown superheating starting at 23 s with a subsequent

quench at 31 s. This indicates that the refill phase was completed at 31 s.

Post-test analysis and comparison with earlier double ended break tests have

shown that the hot leg QOBV opened sufficiently at time zero to allow maximum

flow. Only the position indicator of the QOBV did not operate properly. All

quench phenomena observed during the blowdown phase in LP-FP-1 and discussed

in detail in [18] were also present in LP-FP-1A. The bottom-up quench

started at about 6 s in the lower core region and the top-down quench reached

the lower core region at about 16 s as shown in Figure 21. The complete

quench of this part of the core was reached at about 32 s.

Bottom-up quench occurred in the peripheral bundles simultaneously with I

the central bundle, while the subsequent top-down quench was heterogeneous

specially in the lower core region as shown in Figure 22. Bundle 4 did not

41
Ca

E

,T

850

750

650

550

450

350

0 10 20 30 40 50

Time after rupture (s) L221-KA258o22

Figure 21. Fuel rod cladding temperatures measured at different elevations

in the center fuel module du-ing Experiment LP-FP-1A.

750

650

2L

E)

I550

450

350

0 10 20 30 40

Time after rupture (s)

50

L221-KM258 -23

7icu,-e 22. Fuel rod cladH4-.c temperatures measured in a peripheral bundle

during EAperimen: LP-FP-IA.

42
show dryout until 22 s and bundle 6 did not quench during the top-down quench

at 16 s. Only bundle 2 was analogous to the central bundle in the quench

behavior.

Comparison of LP-FP-IA with LP-FP-1 and a discussion how this experiment

was used to explain some phenomena in LP-FP-1 is given in [18].

43
4. SIGNIFICANT LB LOCA PHENOMENA

Before March 1979 the small break LOCA was of little concern by

regulatory bodies and the entire research was concentrated on the large break

LOCA. The TMI-2 accident shifted the research efforts towards small break

LOCA, anticipated transients, and severe accidents. Since then, with the

exception of LOFT experiments L2-5 through LP-FP-1, little was done in the

area of large break LOCA research. Today, the nuclear industry appears to

have a conviction that the phenomena associated with the large break LOCA are

well understood and well-defined by models in codes such as TRAC. In this

section we will reassess our understanding of the large break LOCA to confirm

this position by reviewing the phenomena observed during the LOFT large break

experiments. These experiments up to now are the most important because of

the facility scale, the nuclear core of the facility and the experimental

results which provided some insight into what may happen in large PWRs during

a LOCA.

The first and most important phenomenon is the core-wide fuel cladding

cooling and quench during blowdown. This phenomenon, observed for the first

time in the LOFT experiments, changed the perspective of large break

accidents and also led to a reevaluation of critical and post-critical heat

transfer models used in the systems codes. We will discuss the reactor

coolant pump operation mode and coolant flow distribution during the early

blowdown phase which contribute to this phenomenon. We will also review some

experimental aspects related to this phenomenon, specifically the fin-cooling

problem of external cladding thermocouples and nuclear fuel rods versus

nonnuclear heating elements.

The blowdown cooling in LOFT LBLOCA experiments is not unique to that

facility. This phenomena is calculated to occur in 4-loop plants with

nominal pump operation and trip criteria. Only under a typical pump

operating conditions, such as impeller seizure in one pump, can the cooling

phenomena be suppressed. Also, because the cooling phenomena is strongly

dependent on the mass inflow/outflow hydraulic balance in the reactor vessel,

the cooling phenomena is likely never to occur in PWR configurations with two

inlet pipes (cold legs). The cooling phenomena may or may not occur in PWR

44
designs with three inlet pipes depending on pump trip criteria and coastdown

characteristics. As previous stated, the cooling phenomena is considered

highly likely in PWR configurations with four inlet pipes.

We will also discuss the effects of top-down quench phenomena, break flow

phenomena and finally ECC injection including the effect of accumulator

nitrogen injection. The reproducibility and dependence of all of the above

identified phenomena on initial and operating conditions imply that the

phenomena are real and are associated with this type of transient.

Consequently these phenomena can be expected to occur during large break

accidents in large power reactors and should be appropriately modeled in

systems codes to quantify the magnitudes in the several vendor plant

geometries.

4.1 The blowdown bottom-up core quench

The early fuel cladding cooling during blowdown was observed during

Experiment L2-2, the first LOFT nuclear experiment. This behavior was

different from the expected (and predicted) core thermal behavior. At that

time according to the understanding of reactor system behavior during a large

break LOCA, the cladding temperature should increase rapidly after LOCA

initiation due to equilibration of stored heat and then continue to increase

slowly from decay heat to a maximum during the reflood phase. A precusory

cooling due to droplet entrainment from the lower plenum after ECCS

initiation would reverse the core heatup and begin a slow cooling trend.

Finally the cladding temperature would be quenched to the saturation

temperature as a result of core reflood with ECC water. This classic large

break LOCA scenario was supported with experimental evidence from facilities

such as Semiscale and with code analysis.

In experiment L2-2 (and in other LOFT experiments with similar boundary

conditions) the cladding temperature increases as expected; however, the

temperature increase is stopped in a few seconds and is followed by a core

wide bottom-up cladding quench. The cladding temperature enters CHF

approximately 5 s later and reaches a second maximum during the reflood

phase; however, the highest peak cladding temperatures occur during the first

45
heatup. This unexpected core thermal behavior is in response to hydraulic

phenomena within the primary system during blowdown which cause an upward

coolant mass flow through the core. The hydraulic phenomena and effects on

core thermal behavior provide a difficult challenge to best estimate systems

codes.

After the opening of the blowdown valves in the broken loop the reactor

vessel voids rapidly. The flow direction in the core reverses in response to

the larger liquid flow out of the broken loop cold leg as shown in Figures 23

and 24 for experiments L2-5 and LP-02-6. Fuel rod cladding DNB occurs at

approximately 1 s in the high power region. The reverse flow through the

core lasts only a short time (i.e., in experiment L2-2 the flow becomes

positive after 2.5 s). During the initial blowdown the flow out of the

vessel through the cold leg break greatly exceeds the coolant flow into the

downcomer from the intact loop cold leg as it was shown in Figure 4.

However, the break flow transitions from subcooled critical flow to saturated

critical flow which reduces the magnitude of the break flow below the inflow

from the intact loop cold leg. More liquid is being delivered to the

downcomer than is being expelled out of the downcomer. The additional liquid

penetrates down to the lower plenum and up to the core resulting in a core

bottom-up quench of the cladding as measured by thermocouples on the cladding

exterior surface. The evidence for core flow reversal, and a coolant density

increase within the core is provided by momentum flux transducers at the core

exit, a densitometer in the hot leg of the broken loop and in-core

self-powered neutron detectors (SPND). The SPNDs contain cobalt emitters

which are sensitive to neutron and gamma radiation. In the reactor shutdown

state the SPNDs are sensitive to variations in local fluid density through

the gamma flux sensitivity and therefore provide good indication of the

additional liquid, or sometimes referred to as a density wave travelling

through the core [21], [22]. The positive mass flow and density increase

causes the bottom-up cladding quench which is relatively uniform radially.

The quench lasts for about 5 s at which time the continued mass depletion

causes CHF to occur. Also, the coolant flow from the intact loop cold leg

decreases below the cold leg break flow at about this time which contributes

to reactor vessel coolant depletion.

46
250

0,

04

CO

('O

(0

150

50

-50

-5 0 5 10 15

Time after rupture (s)

20 25 30

L221-KM258-25

Figure 23. Broken loop hot

LP-02-6. leg mass flow rates during Experiments L2-5 and

CD

0

CO

('

700

600

500

400

300

200

100

0

-100

-5 0 5 10 15

Time after rupture (s)

20 25 30

L221-KM258-24

Figure 24. Broken loop cold leg mass flow rates during Experiments L2-5 and

LP-02-6.

47
An estimate of the initial quench wave density is possible from the

response of the SPND's located at the axial peak power zone of the LOFT core,

as the SPND output signals changed rapidly coincident with the initiation of

the cladding temperature quench. The SPND response is shown in Figure 25 for

the L2-3 experiment and is compared to the cladding thermocouple response at

the same elevation. This change in the SPND output has been correlated to

coolant density change through reactor physics calculations. Aksan [23]

estimated the value of the quench front coolant density in the center fuel

assembly to be 615 ± 80 kg/m3. The calculated densities from the SPND

indicate that even at the hot spot region the quality was about I to 3%.

Therefore, the core inlet flow had to have been at saturation conditions.

Another indication of low-quality flow upwards through the core was

obtained from the upper plenum thermocouples, which measure coolant

temperatures directly above the core. Figure 26 shows the measured coolant

temperature and indicates that from approximately 3 to 6 s, the coolant in

the upper plenum nearest the core was superheated vapor. However, at

approximately 6 s, the upper plenum coolant temperature was rapidly reduced

to saturation temperature.

The cladding thermocouple data are also useful in establishing the

general behavior of the quench as it progressed from the bottom to top of the

reactor core. Figure 27 shows the measured cladding temperature at the 38 cm

(15 in.) axial location (measured from the bottom of the fuel rods). Notice

the well-defined time at which the coolant flow initiated the rapid cooling.

This behavior was consistent at all axial levels and is summarized in

Figure 28 showing the initial quench cooling time versus axial position for

each of the 20 axial cladding thermocouple locations. The velocity of the

coolant wave as measured from the initial, rapid cladding cooling time versus

axial position is estimated from Figure 28 to be approximately 1.00 to

1.50 m/s. (The upper plenum coolant thermocouple quench occurred just after

highest elevation fuel cladding thermocouples began to quench). The core

inlet mass flux was estimated to be approximately 515 to 1050 kg/m 3 s from

the coolant velocity and density estimates.

48
900 1 1 1 -4.5 o...........

800 -5.0-

. . p Ez 70 -5.5 EL

500 - a

Time after rupture (s) P5.0-WHT-488-04

Figure 25. Relative SPND output and cladding temperatures for Experiment

L2-3.

E

.CD

CL

CL

D)

800

750

700

650

600

550

500

-2 0 2 4 6 8 10

Time after rupture (s) P590-WHT-488-03

Figure 26. Upper plenum temperature during Experiment L2-3.

49
CL

E

1000

900

800

700

600

500

-2 0 2 4 a 8

Time after rupture (s)

10

P690-WHT-488-06

Figure 27.

6.9i

Cladding thermocouple response during Experiment L2-3 at 38 cm

(15 in) elevation.

6.7

Upper plenum thermocouple data

i Fuel - rod thermocouple data - i

6.5 II.-

I6.3

-I -!

6.1 h

-I

5.9

5.7

5.5

I

I Note: Vertical bars Indicate data

spread from three or more

thermocouples. i1

v.V 0 25 50 75 100 125 150

Elevation from bottom of active fuel (cm)

175 200

INEL 2 2061

Figure 28. Initial cladding cooling versus axial elevation for center fuel

assembly during Experiment L2-3.

50
In summary, the rapid cladding cooling was primarily a result of low

quality, high upward core flow at a time when the system pressure was still

relatively high (7 MPa). References 6, 9, 23, and 24 discuss the thermal

hydraulic conditions leading to the early quenches during experiments L2-2

and L2-3, in more detail.

4.2 Effects of primary coolant pump operation

The LOFT experiments have shown that the early bottom-up quench is a

result of the fine balance of coolant inflow and outflow from the downcomer

and flow reversal in the core. Review of the break flow (more discussion of

the break flow is included in the Section 4.4) in all of the large break

experiments shows that the break flow and cold leg broken loop flow are

dependent on primary system pressure and coolant temperature upstream of the

break. These parameters were nearly the same for all the experiments.

Therefore, the early bottom-up fuel cladding quench depends on hydraulic

parameters within the reactor vessel and intact loop.

Three operational modes of the reactor coolant pumps were used in the

LOFT large break experiments:

- continuous pump operation

- early pump trip with typical pump coastdown

- early pump trip with fast pump coastdown (decoupled flywheels)

The early quench did not occur only in the experiments in which the pumps

were tripped within 1 s of transient initiation and disengaged from the

flywheels (L2-5 and LP-LB-1). The experiment results show that the early

quench is a function of the pump operation mode, pump characteristics, and

initial flow conditions in the intact loop.

The first two large break experiments, L2-2 and L2-3, were conducted with

reactor coolant pumps running. The coolant mass flow rate in the cold leg

remained almost constant until 6 s during both experiments. Due to this and

the decreasing mass flow rate in the broken loop cold leg, the intact loop

cold leg mass flow rate exceeded the broken loop cold leg mass flow rate. At

51
5 s this difference was about 60 kg/s. The difference in the flow rates

resulted in an excess of 700 kg of water being delivered to the downcomer

between 4 an 6 s. The excess mass of water resulted in the propagation of a

density wave upward through the core during this time. In these experiments

the early quench was also the most complete, extending through the entire

core radially and axially.

In experiment LP-02-6 the pumps were tripped at the beginning of the

transient and allowed to coast down under the influences of the flywheels.

Figure 29 shows the mass flow rate in the intact loop cold leg. Despite pump

trip at about I s the mass flow rate in the intact loop cold leg remained

almost constant until 5 s (similar to experiments L2-2 and L2-3 where the

pumps were not tripped). The amount of the additional coolant entering the

downcomer in this early portion of blowdown was initially calculated for

LP-02-6 to be 135kg [13], about 5 times less than during the L2-3

Experiment. This mode of pump operation provided enough coolant and head to

initiate the bottom-up quench, but the quench front did not propagate through

the entire core. This is illustrated by the behavior of thermocouples

mounted on fuel rod 5G06 in the center fuel module at four different axial

positions shown in Figure 30. The thermocouple at the lowest position

TE-5G06-11 (11 inches above the bottom of the core) quenched at about 6 s,

while the thermocouple at 45 inch elevation was rewetted at 9 s. The upper

thermocouple at 62 inch showed only some cooling effects but not quench.

In experiment L2-5 the reactor coolant pumps were tripped also at about

1 s but the flywheels were disconnected from the pumps resulting in a very

fast pump coastdown. The effect of fast coastdown is illustrated in

Figure 29 which compares the mass flow rate in the intact loop cold leg for

experiments L2-5 and LP-02-6. The coolant mass flow rate decreases rapidly

at 2 s in experiment L2-5 compared to 5 s in experiment LP-02-6 with typical

pump coast down. This early mass flow rate decrease in the intact loop cold

leg happened about 1.5 s before saturation in the broken loop cold leg.

Consequently, only about 9 kg of additional coolant [13] could be delivered

to the downcomer which was insufficient to quench the core.

52
"O

0

02

400

350

300

250

200

"50

100

50

0

-50

-5 0 5 10 15 20 25 30

Time after rupture (s) L221-KM268-27

Figure 29. Coolant mas

Experiments

flow rates in the'intact

L2-5 and LP-02-6.

loop cold leg during

CL

E

I11o0

1000

900

800

700

600

500

400

-5 0 5 10 15 20 25 30

Time after rupture (s) L221-KM258-28

Figure 30. Fuel cladding temperatures measured at rod 5G06 at

elevations during Experiment LP-02-6.

53

different
Experiment LP-LB-1 was performed with reactor coolant pumps in the same

operation mode as in Experiment L2-5. This experiment also did not contain a

bottom up core quench.

Experiment LP-FP-1, in which the reactor coolant pumps were tripped early

and disconnected from the flywheels, contained an early quench. The quench

occurred in this case because of the higher initial, or steady state, mass

flow in the intact loop. The additional mass flow caused the magnitude of

the intact loop cold leg flow to be larger than the broken loop cold leg flow

at the time of transition to saturated critical flow at the break.

4.3 The blowdown top-down quench

The core cooling during blowdown contains two phenomena: the bottom-up

cooling, discussed in the previous sections, and the top-down cooling.

Again, LOFT experiments were the first to show the top-down cooling

phenomenon. Cladding temperatures measured during the LP-02-6 experiment, as

shown in Figure 30, indicate a second quench in the upper part of the core

which moved downwards and rewetted the cladding at the 45 inch elevation at

17.5 s. The thermocouple measuring the highest temperature at the 30 inch

elevation did not indicate the top-down quench.

This top-down quench affected fuel module 2 as shown in Figure 31. The

top-down quench was not uniform across the core as shown in Figure 32. The

quench reached the lower half of bundle 4 (Figure 32) earlier than

bundle 2. Bundle 6 was not affected in the lower half at all by this

quench. Bundle 4 was closer than the other fuel elements to the intact loop

hot leg which is a source of water which drains into the reactor vessel. As

shown, the top-down quench is multidimensional in contrast to the bottom-up

quench which can be treated as one-dimensional as it rewets the center fuel

module and the peripheral modules at the same time. Analogous top-down

quench phenomena were detected in other LOFT large break experiments.

Experiment L2-5 was performed with a rapid RCP coastdown to prevent the

bottom-up quench, however this operation did not prevent the top-down

quench. In Figure 8, which compares the L2-3 and L2-5 experiments, only the

54
C)

E

a)

1000

900

800

700

600

500

400

-5 0 5 10 15 20 25 30

Time after rupture (s) L221-KM268-28

Figure 31. Fuel cladding temperatures in the peripheral bundle 2 during

Experiment LP-02-6.

,2

1000

900

800

700

600

500

400

-5 0 5 10 15 20

Time after rupture (s)

25 30

L221-KM258-29

Figure 32. Fuel cladding temperatures in the peripheral bundles 2, 4, rnd

6 during Experiment LP-02-6.

55
top-down quench is indicated in the upper part of the CFM in experiment

L2-5. Figure 33 shows that only the top-down quench was present in the

peripheral fuel modules.

The top-down quench phenomenon is concluded to be very significant

because it decreases cladding temperatures which reduces the time needed to

quench the fuel cladding during core reflood with the ECCS.

4.4 Blowdown and Reflood Heat Transfer

The LOFT nuclear core had special experimental instrumentation which

included 186 thermocouples [2] that were laser welded to the external surface

of 76 fuel rods as shown in Figure 34. The fuel rod cladding external

thermocouples indicated that the reactor core was quenched early in the

blowdown transient as explained in the previous sections. Since the

phenomenon of the early quench is very important with regard to removal of a

substantial amount of stored energy from the fuel and with regard to the

ability of computer codes to predict large break LOCA peak cladding

temperature cooling and quench phenomena have been extensively studied.

However, the true nature of this early cooling and quench phenomenon remains

in question.

The postulation has been made that the LOFT external fuel thermocouples

indicate only local quenches of the thermocouple itself or of small cladding

area around it. Our position is that there is enough evidence that the LOFT

cladding surface thermocouples were indicating a true complete quench during

blowdown. However we recognize that the external thermocouples do not

measure the cladding surface temperature accurately because of the

fin-cooling effect. In the following sections we will discuss in detail the

problems associated with the external cladding thermocouples with regard to

the blowdown quench and the reflood quench. We will review separate effect

experiments conducted to study the thermocouple effects and we will discuss

the evidence from the LOFT experiments indicating complete fuel rod quenches

during blowdown.

56
C2

E

'U

1000

900

800

700

600

500 -•-

400 . 1 1 1 1 1 1

-5 0 5 10 15 20 25 30

Time after rupture (s) ,L221.KM25.30

Figure 33. Fuel cladding temperatures in the peripheral bundles 2, 4, and 6

during Experiment L2-5.

Z•CALOY CLADOIW LASER WELDS

... .... _ __ .... ...... .. _.

DUMMY SEWMEIrT MEASURING JUNCT1ON

Figure 34. LOFT fuel rod external thermocouple.

57
We will also address the differences in behavior of nuclear and

electrically heated rods since these differences influence the understanding

of the surface thermocouple effect. It can be concluded from the review of

the work on nuclear rod versus electrical heater rod behavior under rapid

flooding conditions that solid-internal heater rods of the Semiscale design

cannot simulate the rapid quenching of a nuclear rod, due to the relative

high thermal diffusivity of the electrical heaters. Because the U02

conductivity and the fuel-to-cladding gap limit the energy delivery rate to

the nuclear fuel rod cladding during a rapid cooling transient, the nuclear

rod cladding can be quenched by removing only the energy in the cladding.

For the solid-type heater rod, not only the cladding energy but also a

significant portion of the rod internal energy must be transferred before

cladding quench can occur. In other words, a nuclear fuel rod is conduction

limited and a solid-type electrical rod is convection limited. Consequently,

a large amount of energy must be removed from the solid-type electrical rod

before it can quench. The calculated nuclear rod cooling rates can be from 4

to 5 times greater than for Semiscale electrical rods, depending on the inlet

coolant flooding velocity.

4.4.1 Effect of Cladding Surface Thermocouples on Blowdown Heat Transfer

The effects of cladding external thermocouples on the early quench

phenomena were analytically and experimentally investigated at the INEL. Two

different sets of experiments were conducted in the LOFT Test Support

Facility (LTSF). These tests were performed in a single rod geometry and in

a nine rod bundle geometry and involved solid heater rods and rods with

simulated pellet-to-cladding gap. Another set of experiments with nuclear

fuel was conducted in the Power Burst Facility (PBF). This section includes

the results of these experiments and also analyses of LOFT data, specifically

the comparison of fuel centerline temperature measurements with cladding

surface temperature measurements.

58
4.4.1.1 LTSF Experiments

The LTSF quench experiments provide a simple geometrical configuration

with well-quantified inlet hydraulics and the capability to maintain system

pressure as high as 7 MPa. The detailed description of the test facility is

given in References 25 and 26. The first series of tests in LTSF were

performed using a single Semiscale rod which is a typical "solid" internally

heated fuel rod simulator used in many light water reactor research

projects. Four external thermocouples had been laser welded to the outer

surface of this heater rod similar to LOFT cladding thermocouples. The

heater rod also had four internal cladding thermocouples to measure the rod

temperature response. A second rod was built without the external

thermocouples. Identical experiments were conducted with both rods. The

results of one of the 20 quench tests, conducted in LTSF are shown in

Figure 35, for the boundary conditions given in the Figure. The data

presented were taken from cladding internal and external thermocouples

located at the heater rod hot spot. The time of coolant arrival at the

thermocouple location is indicated by the rapid change in the test section

gamma densitometer response. Thus, the quench times can be estimated with

respect to coolant arrival. During the high-pressure (7 MPa) tests the

heater rod with external thermocouples consistently quenched in about half

the time required by the heater rod without surface thermocouples. It can

also be seen from the data that the surface thermocouple is preferentially

cooled and quenches much sooner than the cladding, as indicated by the

internal thermocouple data. However, further analyses [23,28] indicated that

solid heater rods of Semiscale design cannot simulate the rapid quenching of

a nuclear rod, due to the relative high thermal diffusivity of the electric

heater rods. The differences between nuclear fuel and solid heater rods will

be discussed in more detail in Section 4.4.2. Details of these LTSF

experimental results can be found in References 25 and 27.

The second series of LTSF tests investigating quench behavior of a

different fuel rod simulator design was conducted with a nine-rod (3x3)

bundle. The tested rod was always in the center of the nine-rod

configuration (Figure 36). The eight surrounding rods were solid-type FEBA

heater rods (similar to Semiscale heater rods).

59
1200

1000

Ile

CL

E

a)

800

Rod with surface amd internal thermocouples

Surface thermocouple ------------ Internal thermocouple

-.-.- Densitometer

-'-,,trouBare rod with internal ... ~ thermocouple

- \

Pressure , 7 MPa

Flooding rate - 1.8 m/s

Cladding temperature - 1175 K

-inlet quality - 0

600

400

0 5 10

Time (s)

15 20

P590-WHT-468-07

Figure 35. LT$F single

cladding.

rod quench test thermocouple response for initial

External

thermocouple

(4 places)

FEBA rod

(8 places)

INEL 2 1172

Figure 36. LTSF test nine-rod bundle configuration

60
A REBEKA cartridge-type heater rod and a FEBA solid-type heater rod

(Figure 37) were each tested in the center position in the nine-rod bundle,

which provided a geometry and thermal-hydraulic environment typical of a

nuclear fuel rod cluster. The REBEKA heater rod has Zircaloy cladding and

aluminum oxide pellet construction with a pellet-cladding gap to simulate the

thermal characteristics of a nuclear fuel rod. This heater rod was tested

with and without cladding external thermocouples. The main objectives of

this experimental program were to evaluate the effect of cladding external

thermocouples on the early blowdown phase quench behavior of a cartridge-type

nuclear fuel rod simulator, to determine how accurately cladding external

thermocouples measure cladding temperature during a blowdown phase quench,

and to compare the high-pressure quench behavior of a cartridge-type heater

with that of a solid-type heater rod under thermal-hydraulic conditions that

occurred during the blowdown phase (0 to 10 s) of LOFT experiments. The

experimental program and the results of the tests are given in detail in

Reference 26.

This research program showed that the REBEKA rod satisfactorily simulates

the thermal response of a nuclear rod. It was shown also that the quench

behavior of FEBA rods is significantly different than that of REBEKA and

nuclear fuel rods. Due to the higher thermal diffusivity of solid-type heater

rod and lack of pellet-cladding gap, the rod undergoes a lengthy period of

precursory cooling before quenching; whereas a cartridge type heater rod and

nuclear fuel rod quench very rapidly from high temperatures when subjected to

rapid flooding conditions. The REBEKA rod quenched in less than 3 s from

about 900 K, whereas, the FEBA heater rods experienced an extended period

(10 s) of precursory cooling before quenching at about 700 K (Figure 38).

The results of the experimental program indicate that cladding external

thermocouples had a negligible effect on the cooldown rate and quench

behavior of a REBEKA cartridge-type heater rod under rapid (I to 2 m/s)

flooding conditions at high pressure (Figure 39). Rods with or without

external thermocouples undergo the same quenching under the same hydraulic

conditions. However the cladding external thermocouples are preferentially

cooled during the quenching process and do not accurately measure cladding

61
10.64

9.20

9.10 - .I

6 l.1•-- 

6.02

4.80

3oron nitrid Gas annulus

Inconel heater Heater (Inconel) rod sheath

Magnesium oxide' CZiraloy cladding

Annular pellets (AI 2 0 3 )

Dimensions are In millimeters

INEL. 2 1166

Dimensions are in millimeters

INEL 2 1132

REBEKA FEBA

Figure 37. Cross section of FEBA and REBEKA fuel rod simulators

62
1000

900

2

a)

L.

COI..a)

D.

E

Co I800

700

REBEKA ................ FEBA

Inlet quality - 0 percent

Inlet velocity - 1 m/s

Mass flux - 746 kg/s-mr

Pressure 7 MPa

600

500

0 5 10 15

Time (s) PSOO-WHT-488-O0

Figure 38. REBEKA and FEBA

(REBEKA) and 1F

heater rod 7 temperature responses for runs IA

(FEBA).

1000

900

REBEKA rod predicted

cladding temperature

REBEKA rod TE-RES-X2 ................ PBF nuclear rod

cladding temperature

PBF rod external

thermocouple

pressure = 7 MPa

CL

E

a)

800

700

600

500

0 5 10

Time (s)

15

P590-WHT-488-09

Figure 39. Blowdown quench beahvior of a REBEKA heater rod (from LTSF

bundle test) and a nuclear fuel rod (from PBF-TC tests) with

external cladding thermocouples.

63
temperature during reflood. Since the REBEKA rod has been shown to

satisfactorily simulate the thermal response of a nuclear rod, [28], the

REBEKA rod results are considered applicable to LOFT nuclear fuel rods.

4.4.1.2 PBF Experiments

Three series of light water reactor fuel behavior experiments

(Thermocouple Effects Experiment series TC-1, TC-3 and TC-4) were performed

in the Power Burst Facility (PBF) at the INEL to specifically evaluate the

influence of cladding surface thermocouples on the thermal behavior of

nuclear fuel rods under LOCA conditions. A total of twelve experiments were

conducted. Each experiment was performed with four LOFT-type fuel rods

contained in individual flow shrouds. Two of the rods were instrumented with

four LOFT cladding external thermocouples located near the high power region

of the fuel rods. All four rods were also instrumented with internal

thermocouples at the same axial level as the external thermocouples. Details

of the experiment design, conduct and results are presented in References 29,

30, and 31. The analysis of the PBF data contains larger uncertainties than

the analyses of other facility experiment data because the hydraulics were

not exactly the same among all four separately shrouded rods. However, the

following is considered to be qualitatively accurate.

Evaluation of the measured temperature difference across the cladding

indicated that the cladding surface thermocouples measured cladding surface

peak temperatures during blowdown that were only slightly lower (20 to 30 K)

than the actual cladding temperatures. However, comparison of externally

instrumented rods with bare rods showed that the surface thermocouples

influenced the cladding temperatures during the blowdown phase of the TC

tests in two respects:

1. The cladding surface thermocouples increase the surface heat

transfer area of the fuel rods and enhance the heat transfer during

the initial few seconds of blowdown. As a consequence, CHF is

delayed on the externally instrumented rods which results in a

reduction of stored energy in the fuel rods at the time-of-CHF.

64
2. The presence of external thermocouples influences the fuel rod

thermal response through a "fin cooling" effect during cladding

heat-up.

These two effects resulted in a reduction of blowdown peak cladding

temperature ranging from 101 to 115 K (Figure 40). Garner [30] estimated

that 35-58% of this reduction is related to delay in time-to-CHF with peak

cladding temperatures reduced 74 K for each second of delay, and 42 to 65% of

peak cladding temperature reduction is due to fin cooling.

The PBF-TC experiment results showed that fuel rods with and without

external surface thermocouples were quenched. The effect of surface

thermocouples on rod thermal response during this blowdown quench (Figure 40)

appears to be relatively small. If the quality of the flow is very low and

the rods are quenched extremely fast, the surface thermocouple effect is

negligible. In addition, the effect of surface thermocouples on nuclear fuel

rod thermal response during the blowdown quench decreases as the rod initial

power decreases, and at low power the effect disappears.

4.4.1.3 Analysis of LOFT Data

In December 1981, analysis of fuel rod perturbations was begun for

several geometries of internally located thermocouples. The purpose of this

work was to determine the thermal perturbations of placing thermocouples

inside the fuel rod pellets. Fuel centerline thermocouples were being

designed for placement in LOFT CFMs as a further study of the thermal

response during blowdown. This analysis [32] used the COUPLE/MOD5 heat

conduction code [33] in both steady state and transient modes. Transient

boundary conditions of (1) power generation in the fuel as a function of

time, (2) heat transfer coefficient at the fuel rod surface, (3) coolant

temperature, and (4) the fuel to cladding gap conductance were obtained from

FRAP-T5 prediction calculations of the then designated NRC LOFT Experiment

L2-6, which subsequently became OECD LOFT Experiment LP-02-6. The

COUPLE/MOD5 calculations of fuel and cladding temperatures are typically as

65
CL

E

'U

1300

1200

1100

1000

900

800

700

600

500

-5 0 5 10

Time (s

15 20 25 30

;) P690-WHT-48B-10

Figure 40. Embeded and internal cladding'thermocouple response in rod 04

(bare) and rod 02 (externally instrumented) during blowdown of

PBF test TC-4F.

66
shown in Figure 41 (taken from Reference 32). The results indicate that the

fuel centerline temperature is sensitive to the occurrence of a short

duration quench of the cladding.

FRAP-T5 calculations were based on the thermal-hydraulic behavior

calculated by the RELAP series of codes. The thermal-hydraulics of L2-6 were

calculated to be similar to the thermal-hydraulics in Experiments L2-2 and

L2-3 [7]. Post-experiment calculations with appropriate revisions to the

existing models resulted in good agreement with the measured quench phenomena

[7]. Quench phenomena also was calculated to occur in Experiment L2-6

because of the similar thermal-hydraulics. However, the formal prediction of

the approved OECD LOFT Experiment LP-02-6 was done with the TRAC-PD2/MODI

code [14]. The following conclusion is taken from Reference 14:

An early rewet was not calculated to occur during

Experiment LP-02-6. However TRAC-PD2/MODI calculated core

hydraulic conditions for Experiment LP-02-6 which were

similar to the hydraulic conditions which were responsible

for the early rewets in previous LOFT large break

experiments. Since the early rewets in the previous

experiments were also not calculated by TRAC there is a

definite possibility that an early rewet could occur in

Experiment LP-02-6.

The preceding information provides two principal conclusions which sets

the basis for the succeeding discussion. These conclusions are:

1. Fuel rod heat transfer is strongly dependent on fuel

cladding-to-coolant heat transfer and cladding temperature. The

dependency extends to and includes the fuel center.

2. Systems codes have difficulty calculating quench behavior during the

blowdown phase. However, specific use of correlations such as the

Biasi correlation can lead to calculated quench behavior very

similar to that indicated by cladding thermocouples.

The OECD LOFT Experiment LP-02-6 did include early quench phenomena [13]

similar to that observed in Experiments L2-2 and L2-3 [7] as discussed in

detail in Section 3.4. Figure 42 shows the response of the fuel centerline

67
CL

E

a)

1800

1700

1600

1500

1400

1300

1200

1100

1000

900

800

700

600

500

i

Fuel centerline

• °...........................

Cl pellet surface at TC location

Eme d e th r o o pe j n to

Cladding outside surface at TC location -

Cladding outside surface 180* from TC

0 10 20 30

Time (s)

40 50 60

P500-WHT-488-11

Figure 41. COUPLE/MOD5 calculation of fuel

boundary conditions originating

rod temperatures using FRAP-T5

from RELAP5 code.

2000

1500

TC-5D307-27

TE-51307-027 -........

CO

CL

E ................

1000 I500

n,

-50 0 50

Time after rupture (s)

100 150

P590-WHT-488-12

Figure 42. Fuel centerline and cladding temperature at the 27-in elevation

on fuel rod 5D07 during experiment LP-02-6.

68
temperature to the measured cladding temperature at the same elevation. The

fuel centerline temperature responds to the change in cladding temperature

shortly after the blowdown quench at approximately 8 s and similarly after

the reflood quench at approximately 53 s. Figure 43 shows the fuel

centerline temperature for two fuel rods which did not have externally

mounted cladding thermocouples compared with the fuel centerline temperature

shown in Figure 42. The three centerline temperatures are essentially

identical in behavior. These data conclusively show that the hydraulics

cause complete fuel cladding quench and not just thermocouple quench or

localized cladding quench.

Post-transient calculations of LP-02-6 with the RELAP5/MOD2/CYCLE36 code

[34] showed cladding quench behavior at all core elevations except at the 26

in elevation which showed significant cooling but no quench. Post-transient

calculations with the TRAC-PF1/MOD1 [35] showed only some cooling of the

cladding at all elevations. Comparison of calculated and measured fuel

centerline and cladding temperatures, shown in Figures 44 and 45, reveals

that the calculated cladding temperature must be in error since (1) the fuel

centerline temperature comparison is poor, and (2) the fuel centerline

temperature is strongly dependent on the cladding temperature (or

cladding-to-coolant heat transfer) as has been shown in Reference 32

and actual LP-02-6 data.

The data in LP-02-6 shows that cladding quench occurs at high

temperatures, well above the values that have been used in codes such as

TRAC-PF1/MOD1 [35]. Nelson [24] describes forced-convective heat transfer

mechanisms and explains how minimum wall superheats greater than the

homogeneous nucleation temperature result. The conclusion is that quenching

can occur at high temperatures in forced-convective water systems. An

extensive collaborative effort by Gottula, Condie, and Nelson of EG&G Idaho,

Sundaram, Yankee Atomic Electric Company, and Neti and Chen, Lehigh

University produced a large experimental data bank from forced-convective,

post-CHF heat transfer experiments [36]. To quote from the report,

"Quasi-steady state (slow moving quench front) experiments were conducted at

pressures of 0.4 to 7 MPA, mass fluxes of 12 to 70 kg/m2.5, inlet

69
2000

2 1500

CL, 10

500

0 10 20 30 40 50 60

Time after rupture (s) P59oWHT-488.14

Figure 43. Fuel centerline temperatures at the 27-in. elevation during

Experiment LP-02-6.

70
0

IS

a.

E

S

IS

S

S

U

S

I'.

o TC-5D09-27

A TC-5010-27

Cco"Vanw"I so

Rod 2

E Iev = 0.6385 tm

3WWdIq 5

-20 0 20 40 60

Time (S)

80 m0 120

Figure 44

1000-

I- •0

800

0 ago

CL

E 700

V

50

Measured and calculated with TRAC-PFI/MODI fuel centerline

temperature for experiment LP-02-6 at the 0.6858 m (27-in) core

elevation.

STE-5006-027

aTE-5D07-027

*TE-SH06-025

Rod I

Eiev in 0.6943 t"

BndI. 5

-20 0 20 40 60 80 ¶00 120

Time (s)

Figure 45. Measured and calculated with TRAC-PFI/MOD1 cladding

for experiment LP-02-6 at the 0.6943 m (27-in) core

71

temperatures

elevation.
qualities of -7 to 47%, and heat fluxes of 8 to 225 KW/m 2 ... ."The data

above 0.4 MPA extend the existing data base. The data was compared with

currently used wall heat transfer correlations, and the'results were

unsatisfactory. A regression analysis of the data showed thermal

nonequilibrium, quench front quality, and distance from the quench front to

be significant factors in the correlation of the data. These effects are not

included in current correlations and are thought to be the major reason for

the poor comparisons." The authors further conclude that there is a need for

further post-CHF model development that include such factors as quench front

conditions and elevation. These are needed for advanced codes to predict the

post-CHF heat transfer, thermal nonequilibrium and quenching phenomena. This

recent work combined with the experimental results of Experiment LP-02-6 and

the results of code calculations such as those in Figures 44 and 45 clearly

indicate that the current generation systems codes do not adequate predict

post-CHF heat transfer and quenching.

OECD LOFT Experiment LP-LB-1 also contains data showing the strong

dependency of the fuel centerline temperature on the cladding temperature and

heat transfer. In Experiment LP-LB-1 the early bottom-up quench phenomena

was suppressed. However, there was a weak partial top-down quench that

occurred in the 10-30 s time interval and extended over approximately the top

third of the core [16]. Figure 46 shows fuel centerline temperature at the

27-in. elevation for rods with and without cladding surface thermocouples.

The centerline temperature behavior indicates no quench in agreement with the

measured cladding temperature. Figure 47 shows similar temperature data at

the 43.8 in. elevation. A small early cooling occurs in this region as

indicated by the cladding temperature in Figure 47 compared with that in

Figure 46. The fuel centerline temperature is sufficiently sensitive to show

even this small cooling. The data in Figures 46 and 47 during the final

quench does show that the final quench occurs up to approximately 20s earlier

on fuel rods with thermocouples. These results are consistent with fuel rod

results in PBF [30]. Comparison of several other pairs of fuel rods in

Reference 16 all show the same trends. Figures 48, 49, and 50 show fuel

centerline and cladding temperature measurements at the 43.8 in. elevation.

These fuel rods experienced larger degrees of cooling than those in

72
2000

1500

12 1000

500

0

-50 0 50 100 150 200 250

Time after initiation (s)

300

P590-WHT-488-16

Figure 46. Fuel rod centerline temperature with and without cladding

surface thermocouple and cladding temperature measurement during

Experiment LP-LB-I at the 27-in elevation.

2000

1500

I I

----------------

i I i

TC-5J07-43.8 (centerline without TC)

TC-5L07-43.8 (centerline with TC)

TE-5L07-43.8 (cladding TC)

2

(,

CL

E

91

1000

500 F0

-50 0 50 100 150 200 250

Time after initiation (s)

300

P690-WH4T-488-16

Figure 47. Fuel rod centerline temperature with and without cladding outer

surface thermocouple and cladding temperature measurement at the

43.8-in elevation. 73
2000

1500

2

1.=

aE

IT

II I I I

TC-5104-43.8 (centerline) ................ TE-5104-43.8 (cladding)

------------------- ---- ------- - - ----

1000

500

0

-50 0 50 100 150

Time after initiation (s)

200 250 300

P690-WHT-488-17

Figure 48. Fuel centerline and cladding temperatures on rod 5104 at the

43.8 in elevation during Experiment LP-LB-1.

90nn(

2000

1500

1000

2

CL.

E

1,2

I i I I 

TC-5108-43.8 (centerline) TE-5108-43.8 -.----..------.. (cladding)

-------------------- --- ------------

500

U

-50 0 50 100 150 200

Time after initiation (s)

250 300

P590-WHT-488-18

Figure 49. Fuel centerline and cladding temperature on rod 5108 at the

43.8 in elevation during experiment LP-LB-1.

74
2000 III I I

TC-5114-43.8 (centerline) ................ TE-6114-43.8 (cladding)

CL.

E

IT

1500

1000

500

0

-50 0 50 100 150 200 250

Time after initiation (s)

300

P590-WHT-488-19

Figure 50. Fuel centerline and cladding temperature on rod 5114 at the

43.8 in elevation during experiment LP-LB-1.

75
Figures 46 and 47. The correlation between fuel centerline and cladding

temperature is consistent with previous comparisons and indicates that more

than just thermocouple cooling is occurring. The data in Figures 48, 49 and

50 also show that at specific times the cladding temperatures exceeds the

fuel centerline temperature. The reason is that the surrounding rods were

not cooled as much and as a result heat transfer occurs from those rods. The

fuel centerline data shows that the heat transfer occurs through the rod at

the time constant for the fuel thermal properties.

Experiment LP-FP-I [20] provided cladding temperature data for the case

where the cladding thermocouple is quenched but not the fuel cladding on the

same or adjacent fuel rods. Typical data is shown in Figure 51. Rods 5G05

and 5105 did not show thermocouple dryout until shortly after 200 s whereas

thermocouple dryout occurred before 100 s on rods 5G11 and 5111. The higher

rate of temperature increase on 5G05 and 5105 relative to the other rods

indicates that adjacent rods had significant cladding temperature at the time

of thermocouple dryout on rods 5G05 and 5105. The thermocouple on 5G05 was

quenched again at approximately 270 s. The more rapid rate of heatup

following subsequent dryout indicates that only the thermocouple was

affected. The phenomena in Figure 51 has not been observed in any of the

cladding thermocouples in Experiments L2-2, L2-3, and LP-02-6 during the

quench-dryout period early in the depressurization.

Cladding temperature data in the hot region of the core shown in

Figure 52 for LP-02-6 (early quench phenomena) and LP-LB-1 (no early quench),

can be used directly to assess the validity of the assumption that only the

thermocouple in LP-02-6 was quenched. If the assumption is true then, (1)

the rate of temperature increase following subsequent dryout should be

greater than that following DNB, and (2) the maximum temperature following

dryout should equal or exceed the maximum temperature reached following DNB.

Neither result is evidenced in the data in Figure 52. The conclusion is that

the assumption is incorrect and that significant cooling occurred on all fuel

cladding. The cooling was sufficiently large initiate and precipitate early

quench.

76
1250

1000

C2

E

a)

750

500

250

0 100 200 300

Time (s)

400

P590-WHT-488-20

Figure 51. Comparison of cladding temperatures at the 27-in elevation in

the central fuel assembly during Experiment LP-FP-1.

12

1200

1100

1000

900

800

700

600

500

TE-5D07-027 (LP-LB-1) .......... TE-5D07-027 (LP-02-6)

C,p

! . . " ° .. 

&IUU

-50 

0

-50 0 50

Time after initiation (s)

100 150

PBgO-WHT-488-21

Figure 52. Cladding temperatures at the 27-in. elevation on rod 5D07 during

Experiments LP-02-6 and LP-LB-I.

77
There is one final topic for discussion. As mentioned in Section 4.1 the

early quench phenomena involves high pressures (5-8 MPA) and rapidly moving

density waves through the core. The latter is calculated to be approximately

1.8 m/s. This moving density wave causes a quench propagation of

approximately 1.5 m/s in the core. Nelson [24] states that for those

conditions (1) no applicable data base exists, and (2) conduction-controlled

quenching cannot be the controlling factor. He states further that these

conditions comprise a second case where downstream quenching is influenced by

both the quenching front and convection (the quench front in this case is

termed a conduction-convective propagating quench front). It is apparent

that currently used heat transfer correlations cannot handle this case. The

research reported in Reference 36 provides data to extend the data base into

the required range and shows that current heat transfer correlations are in

poor agreement with the new data.

In summary, the early blowdown quench phenomena in LOFT is concluded to

be real and not an artifact of thermocouple-only or thermocouple induced

quenching. Further, current systems codes cannot calculate this phenomena

because adequate heat transfer correlations have not been developed.

4.4.2 Nuclear Fuel Versus Electrical Fuel Rod Simulators: Simulation

Limitations During Blowdown

The typicality of the blowdown quench behavior of a solid-type electrical

heater rod relative to that of a nuclear fuel rod has been questioned because

of the different thermal properties and lack of a simulated fuel-pellet

cladding gap. In this respect, LTSF experiments, which investigated the

blowdown quench behavior of a Semiscale solid-type heater rod with only

internal thermocouples, can be used as a basis for evaluation of code model

calculations. Details of such calculations using the RELAP4/MOD6 computer

code [37] are given in References 23 and 28. Having established the validity

of the heat transfer models to calculate the initial cooldown rate of a

quench, a series of RELAP4 calculations were performed to compare the initial

cooldown rates of the nuclear fuel with gap, REBEKA cartridge-type electrical

heater rod with gap, and Semiscale solid-type electrical heater rod for rapid

cooling transients. These calculations under typical LTSF single-rod

78
experiment conditions were performed by substituting the individual rods in

the RELAP4/MOD6 LTSF model. The calculated results for LTSF experiment 12

are given in Figure 53. These calculations show that the REBEKA heater rod

simulates the nuclear fuel rod cooling very well for the conditions

investigated.

The nuclear fuel rod with gap was calculated to cool approximately five

times faster than the Semiscale solid-type heater rod. These comparison

calculations [28] were carried out over a range of inlet flooding velocities

(2 to 6 m/s). The predicted initial cooling rates are summarized in

Figure 54. Also indicated in this figure are the initial cladding cooling

rates measured on nuclear fuel rods from the PBF Thermocouple Evaluation

Experiment Series [30]. The limited number of nuclear fuel rod data suggest

that the calculated cooldown rates may be 10 to 20% too high. The results of

those PBF experiments also indicate that the thermal decoupling of the

cladding and fuel was apparently significant, allowing the cladding to

rapidly quench during the blowdown phase. This thermal decoupling of fuel

and cladding demonstrates the importance of in-pile experiments or

out-of-pile experiments where the fuel-to-cladding gap is properly simulated.

Additional experiments were conducted in LTSF using a REBEKA

cartridge-type fuel rod simulator with gap and zircaloy cladding and thermal

diffusivity much closer to the nuclear rod diffusivity. The experimental

results without external thermocouples show very rapid cooling (150 to

20OK/s) and quench times (2 to 3 s) similar to the nuclear fuel rod data at 4

m/s inlet flooding rates (Figure 54). A comparison of the cladding

temperature response of the REBEKA rod with external thermocouples ,and a

nuclear fuel rod with external thermocouples, where the initial temperatures

of the rods prior to quenching were about the same (900 K) is shown in

Figure 39. Similar results exist for rods without external thermocouples as

mentioned above. The quench behavior of the REBEKA rod is similar to that of

a nuclear fuel rod, which is also consistent with the results of calculations

performed by RELAP4/MOD6 code (Figure 53).

79
:3

E)

L.

1100

1000

900

800

700

600

500

0 2 4 a 8 10 12

Time (s)

14

P590-WHT-488-22

Figure 53. Calculated cladding temperatures for

cartridge-type and solid-type heater

blowdown quench test.

nuclear fuel rod,

rods for LTSF single rod

-S

o4-

a)

0)

C

.5

0

U

0)

C

-c

a)

.5

-i

250

200

150

100

50

I

----------------

I I I I I

LTSF data (Semiscale rod)

Calculated nuclear rod response

for LTSF test conditions

Nuclear data from

PBF Test TC-3

0

0 1 2 3 4 5 6 7 8

Inlet flow velocity (m/s) P590-WHT-488-23

Figure 54. Comparison

rates from

fuel rod

of measured Semiscale heater rod

LTSF single rod quench test with

initial cooling rates.

80

initial cooling

calculated nuclear
Additionally, the bundle experiments performed in the LTSF produced

valuable evidence with respect to the behavior of cartridge-type REBEKA and

the solid type FEBA fuel rod simulators (Figure 38). The analysis indicates

that the solid-type heater rod temperatures are controlled by convective heat

transfer at the cladding surface. This is due to the high thermal

diffusivity of the rod allowing rod internal energy to be transferred rapidly

to the cladding. As the rod begins to cool down in film boiling, internal

rod energy is conducted to the cladding about as fast as the surface

convective heat transfer removes the energy; thus, the cladding cools slowly

since the rod energy must be transferred before the cladding temperature is

low enough to allow surface quenching. This is clearly seen in Figure 53,

where the film boiling cooldown lasts for 8 s. The nuclear rod, bycontrast

is internally conduction limited by the greater thermal resistance of the

U02 fuel and the fuel-cladding gap thermal resistance. The inability of

the nuclear rod to rapidly transfer internal energy to the cladding, together

with a much smaller zircaloy cladding heat capacity, significantly changes

the energy balance at the cladding surface, causing a more rapid cooldown

during the film boiling (see Figure 53). In comparison, at these high flow

rates the nuclear rod cooling is controlled more by the cladding stored

energy, while the cooling of the solid-type heater rod is controlled more by

total rod internal energy.

The detailed investigations performed on the effect of changing power

history for both nuclear fuel and solid-type heater rods [23] indicated that

an attempt to simulate nuclear fuel rod behavior with solid-type of high

thermal diffusivity heater rods would require unrealistic changes in

electrical heater rod input power, even including applying negative power.

Figure 55 indicates the required heater rod power necessary to simultaneously

duplicate the nuclear rod surface temperature from Figure 53 and

corresponding heat flux obtained from RELAP4/MOD6 calculations. A large

amount of negative power is needed to force the solid-type heater rod to

duplicate nuclear rod response. The unrealistic amount of negative power

needed to simulate nuclear fuel rod response indicates that observed

differences in electrical and nuclear fuel rod response result from inherent

limitations in any solid heater rod design.

81
4000

E

2000

0

.0

-0

0

-4000

-1 0 1 2 3 4

Time (s)

5

P590-WHT-488-24

Figure 55. Calculated peak linear Semiscale heater rod power needed to

duplicate nuclear rod temperature response shown in Figure 53.

82
Similar types of analysis were performed for the Semiscale counterpart

experiments to the LOFT experiments. The experimentally used power history

for the Semiscale large break LOCA experiment was obtained from predicted

LOFT nuclear fuel rod cladding temperatures. But the cladding temperature

response during the LOFT nuclear experiments was very different from the

predicted values used to specify the electrical power for the Semiscale

experiments. As stated previously, during the first 10 s of the LOFT L2-3

experiment, more than 60% of the stored energy in the core was transferred

from the fuel rods due to the high core inlet flow, causing a cladding rewet.

This early cooling was not predicted by the code calculations to occur in the

LOFT experiments and therefore was not included in the calculation to specify

cladding temperatures for the Semiscale solid-type heater rods. Thus, the

Semiscale electrical rods were overpowered compared to the LOFT measured

data. Tolman and Carboneau [38], shows that to duplicate the LOFT cladding

temperatures, negative rod powers are required for substantial periods of the

transient (Figure 56). This is a condition that can not be achieved and it

also indicates that the solid-type heater rod can not exactly simulate the

nuclear rod thermal response under the same hydraulic conditions.

4.4.3 Refloodinq and Boil-off: External Cladding Thermocouple Effect, and

Nuclear Fuel Rod and Electrical Heater Rod Behavior

Large break experiments in LOFT were intended to validate the performance

of the emergency core cooling systems for the design basis loss-of-coolant

accident. As discussed in previous sections the L2-2, L2-3 and LP-02-6

experiments showed that about 60% of the initial steady state stored energy

is transferred to the primary coolant prior to emergency core coolant

delivery to the core. The final core quenches are primarily due to

accumulator fluid delivery. The characteristics of the relatively rapid (10

to 15 cm/s) core quenching for the L2-3 and L2-5 experiments are compared in

Figure 57. The core reflood behavior was very similar during both

experiments even though significant difference is initial stored energy and

cladding temperatures existed. All of the other LOFT large break

loss-of-coolant experiments (L2-2, LP-LB-1, LP-02-6), showed similar type of

behavior during rapid core reflooding. A significant observation is that the

reflooding rates always exceeded the 2.5 cm/s licensing regulation

83
4

I.-

0

CL

0

0

0,

N

Co

E

0

2

0

-2

-4

-6

Power used for Semiscale

Test S-08-3

------------ Power required for LOFT

LOCE L2-3 simulation

j • .I

0 5 10 15

Time after rupture (s) P590-WHT-488-25

Figure 56. Comparison of Semiscale core power used for Exper-iment S-06-3

with that required to simulate LOFT Experiment L2-3.

70 I

60 H

L2-3

• 50H

0

.0

S

- 0

2" 20-

0-4 10

10 20 30 40 50 60 70

Time that liquid level detectors indicate ECC water entry

INEL 2294

Figure 57. Reactor liquid level during final reflood for Experiments L2-3

and L2-5.

84
limitation. Additionally the LOFT large break loss-of-coolant experiments

showed that the blowdown hydraulics and heat transfer (early blowdown quench)

is more important in removing the initial fuel rod stored energy than reflood

heat transfer.

The relatively rapid reflooding observed during the LOFT large break

loss-of-coolant experiments was questioned because of the fin effect of

external thermocouples. As a consequence, separate effect bundle reflooding

experiments using electrical heater rods instrumented with both external and

internal cladding thermocouples were conducted in the NEPTUN test facility

[39] in Switzerland at the Paul Scherrer Institute (formally EIR). The

NEPTUN-I series of experiments were performed with five central heater rods

instrumented with both external and internal cladding thermocouples and the

NEPTUN-II series of experiments were performed only with internal cladding

thermocouples. The results and comparison of experimental data from these

two experiment series [40, 41], indicated that electrical heater rods

instrumented with LOFT external thermocouples experience preferential cooling

during reflooding compared to heater rods with internal embedded cladding

thermocouples (Figure 58). The effect is reduced with higher reflooding

rates (e.g., 15 cm/s). During the precursory cooling until the quench, the

rods with external thermocouples show comparable temperature histories as the

rods with internal thermocouples but heater rods with external thermocouples

quench at higher temperatures and earlier than the other heater rods. An

overall comparison between repeat experiments NEPTUN-I (five central rods

equipped with external thermocouples) and NEPTUN-II (all thermocouples

embedded on the cladding of the heater rods) is shown in Figure 59.

Differences between similar experiments are small especially during

precursory cooling.

The ability of electric heater rods to duplicate nuclear fuel rod thermal

response during reflooding was also questioned because of the large

differences in electric and nuclear fuel rod thermal properties. In this

respect, the Halden Project Test Program Instrumented Fuel Assembly 511

(IFA 511) in Norway, in Holden Research Reactor was designed to

systematically evaluate the ability of electric heater rods (Semiscale

85
0

0)

I.-

4-

to

3.-

a,

0.

E

a, I1000

900

800

700

600

500

400

300

LOFT Rod. level 4 (external TC) LOFT Rod. level 5 (external ---.-.---------. TC) -

NEPTUN Rod. level 4 (embedded TC)

NEPTUN Rod. level 5 (embedded TC)-

200

100

0

0 50 100 150 200 250

Time P690-WHT-488-26 (s)

Figure 58. Cladding temperatures during NEPTUN-I and NEPTUN-II experiments, with and without external cladding surface thermocouples.

0

CD

E

a,

1000

900

800

700

600

500

400

300

200

100

0

0

Figure 59.

50 100 150 200 250

Time (s) P590-WHT-488-2

Cladding temperatures during NEPTUN-I and NEPTUN-II experiments, with and without external cladding surface thermocouples.

86
solid-type) to simulate the response of nuclear fuel rods during the heatup

and reflood phases of a large break loss-of-coolant accident. The experiment

rods were also instrumented with both external and internal cladding

thermocouples to determine if external thermocouples provide an accurate

measurement of cladding temperature. The experiments consisted of one series

with nuclear fuel rods [42] and two other series with electric heater rods,

semiscale solid-type rods (43, 44] and REBEKA heater rods with cladding gap

[45]. In these experiments the nuclear fuel rods were quenched substantially

earlier (four times faster) than solid-type electrical heater rods

(Figure 60). REBEKA heater rods closely simulated the actual nuclear fuel

rod behavior under reflooding conditions. Also, the electric rod, unlike the

nuclear fuel rod, is characterized by a well defined quench. Experimental

data also shows that the response of the external thermocouples was

significantly different than the comparative internal cladding thermocouples

during reflooding at about 7 cm/s flooding rates. The indicated temperature

of the external thermocouple was at least 50 K less than that indicated by

the internal thermocouples throughout reflood and the external thermocouples

indicated quench 20 s earlier. The different thermal behavior indicated by

the external thermocouples was primarily caused by fin cooling effects.

Additional experiments were performed in the FEBA test facility (at KFK

Karlsruhe, Federal Republic of Germany) within the SEFLEX test program in

order to quantify the influence of the design of different fuel rod

simulators on the cladding temperature transients under reflood conditions.

These experiments were done by employing a solid-type FEBA heater rod bundle

and a corresponding bundle of REBEKA fuel rod simulators with gap (for rod

cross-sections see Figure 37). The experimental data (46], indicates that

the reflooding behavior between the two bundles consisting of 5x5 FEBA and

5x5 REBEKA rods is significantly different (Figure 61). At an inlet flooding

velocity of 3.8 cm/s, the influence of the rod design on the peak cladding

temperature is around 100 K lower for REBEKA rods. The reasons for the lower

cladding temperatures and the faster quench front progression for the REBEKA

,,rod bundles are the lower heat capacity of the zircaloy cladding and the

pronounced decoupling of the cladding from the heat source due to the

cladding gap.

87
2

a,

L.

Cu1~

0

0.

E

0

I1200

1100

1000

900

800

700

600

500

400

300

0 50 100 150 200 250 300 350 400

Time (s) P50O-WHT-488-28

Figure 60. Cladding temperatures of electric

rods during quench experiments in

heater rods

Halden.

and nuclear fuel

U

ES

1000

800

600

400

200

0

loOC

800

0 I00 200 300 400

Time (s)

500

P&IO.-W1-4?.488A

0

E

9

600

400

200

0

Figure 61.

0 100 200 300

Tre (s)

Cladding temperatures measured at

FEBA and REBEKA rod bundles.

88

400 500

PIO-WHT-480-298

two different axial levels in
Core uncovery (boil-off) experiments were conducted in the NEPTUN

experiment facility which has been already mentioned above. The results of

NEPTUN boil-off experiments at 5 bar [47] showed that the external cladding

thermocouples do not cause a significant cooling influence in the rods to

which they are attached. The dry-out times of the internal and external

cladding thermocouples were within 10 seconds of each other at any axial

elevation for all rods in the bundle. The cladding external thermocouples

measure the cladding temperatures that would have been measured in their

absence within 0 to -20 K (Figure 62). The experimental data from IFA 511

experiments for the heat-up phase at low pressures showed that the response

of the external and internal thermocouples was nearly identical through

heat-up until temperatures exceeded 700 K. However, after about 700 K, the

cladding surface temperature measured by the external thermocouple was lower

than that measured by internal thermocouples and the difference increases

thereafter. The measured cladding peak temperature was 25 to 40 K less, for

both electrical heater and nuclear fuel rods. These results confirm the

findings of the NEPTUN boil-off experiments.

4.5 The Break Flow

The break flow is a principal parameter in the reactor safety research

because of its strong influence on primary system coolant inventory and

consequently core thermal behavior. The break flow influences almost every

feature of a LOCA sequence. The most important factor which was driving the

research associated with the break flow and development of special mass flow

rate measurement systems was the need to obtain accurate data on break flow

for assessment of the computer system codes against experimental data. The

LOFT facility was very well instrumented to provide good resolution on the

break flow for interpretation of the system behavior during a large break

LOCA.

There are generally four phases for the break flow: subcooled flow,

saturated water flow, two phase flow and steam flow. Subcooled break flow

ended in the hot leg (Figure 23) at about 0.2 s compared with 3.4 s in the

cold leg (Figure 24). Saturated water flow out of the cold leg ended at

about 5 s followed by two-phase flow which ended at about 20 s. The break

89
800

CL

E

CD -O

600

400

200

0

0 200 400 600 800

Time (s)

1000

PSOO-WHT-488-30

Figure 62. Cladding temperatures measured in NEPTUN experiment with

External cladding surface (LOFT) thermocouple and with embeded

thermocouples.

9O
flow consisted of steam for the remainder of the transient. Phase changes in

the hot leg occurred much faster because the initial temperature was higher

and no cooling was present in the steam generator simulator. Break flow of

saturated water out of the hot leg ended at 0.6 s. The following two-phase

flow decreased rapidly from 190kg/s at 0.6 s to 20 kg/s at 3 s. From the

minimum value of 20 kg/s (steam flow) it increased again to a maximum of

80 kg/s at 7 s and then decreased again to steam flow at about 20 s. The

mass flow rates were determined from measured densities and measured momentum

flux using a procedure given in [9]. The uncertainty of the mass flow rate

magnitudes shown in Figure 23 and 24 is approximately ± 20 kg/s, which is on

the order of the differences in mass flow rates between L2-5 and LP-02-6.

The same break size and geometry was used in all LOFT large break

experiments which resulted in similar break flow for all experiments as

illustrated in Figure 63. Figure 63 shows break mass flow rate measured in

the broken loop cold leg for experiments L2-5, LP-02-6, LP-LB-1 and LP-FP-1.

The similarity in the initial break flow for experiments L2-5, LP-02-6,

LP-LB-1 despite different initial power and/or pump operation mode indicates

that during the first few seconds of the blowdown the break flow depends only

on break geometry and initial coolant temperature and pressure. A slight

influence of pump operation mode is visible for the time period between 4 and

6 s when smaller mass flow rates are measured for experiments L2-5 and

LP-LB-1. A rapid coastdown of the pumps in these experiments caused less

coolant to flow into the downcomer and to the cold leg break. The initial

mass flow rate during the Experiment LP-FP-I varies from the massflow rate

measured in the other experiments because the reactor was scrammed before

break initiation in this experiment.

4.6 ECCS Performance

In 1967 evaluation of the ECCS performance became the main objective of

the LOFT program. Two equivalent but independent ECC systems were designed

for the LOFT facility to satisfy two objectives:

- Plant protection

- Simulation of ECCS variations in large pressurized water reactors

(LPWR).

91
600

Co

-~400

0

Va 200

0

0 2 4 6 8

Time

10 12

after rupture (s)

14 16 18 20

L221-KM268-38

Figure 63. Mass flow rates in the broken loop cold leg during Experiments

L2-5, LP-02-6, LP-LB-1, and LP-FP-1.

92
Each of the ECC systems included an accumulator, High Pressure Injection

System (HPIS) and Low Pressure Injection System (LPIS). Figure 64 shows a

simplified schematic diagram showing the connections of the different

components in system A and system B. This diagram also shows that the water

from system A can be injected in the intact loop cold leg or directly into

the lower plenum. In case of system B water can be injected in the intact

loop hot leg, in the upper plenum or in the downcomer. Most of the

experiments were carried out by utilizing system A, while system B was kept

as a backup system. If the ECCS is activated for plant protection, system A

will inject in the lower plenum and system B in the downcomer. Some details

of the accumulator and its instrumentation are shown in Figure 65. The

difference between the initial water level in the accumulator tank and the

position of the lower edge of the variable standpipe determines the amount of

water to be injected. When the water level decreases below the low end of

the variable standpipe, N2 penetrates into the injection line.

The ECCS was used in all LOFT large break experiments. However, in order

to determine the influence of the injection mode on the refill and reflood

processes, we will discuss here ECCS performance only for the following

experiments: L2-5, LP-02-6, LP-LB-I, LP-FP-1 and LP-FP-IA. These

experiments were selected because of differences in ECCS operation mode and

phenomenological results of these experiments. Tables 4 and 5 show the major

characteristics of these experiments with regard to ECCS.

ECCS injection in the first three experiments was in the same location

(intact loop cold leg) but different amounts of emergency coolant were

injected. During the LP-FP-I experiment a combined injection into the upper

plenum and intact loop cold leg was used. The LP-FP-1A test was aborted at

about 10 s into the transient by activation of the Plant Protection System

(PPS).

First, the experiments with intact loop cold leg (ILCL) injection will be

compared and discussed. Figure 66 shows the liquid level behavior in

accumulator A during experiments L2-5, LP-02-6 and LP-LB-1. Similar amounts

of water were injected into primary system during experiments L2-5 and

LP-02-6. However, despite higher initial water level, much less ECC was

93
Accumulator B 6-in. tine - - 4-in. line

h-pressure I t

Low-pressure

injection system

heat exchanger B

Hig

injection system,

pump B I 6 -in.6 re- - 4-in line

a

6 x4-in. reducerj W ----

1-1/2-in. li!ne' '1-1/2 Steamx 4-in.

generator-"

Primai

coolant p

I•4e._1-1/2-

ire I

emI

Test chamber

4-in, line XRO-66

ulator A 6-in. line

reducer

Hot

Downcomer

injection point

High-pressu

injection syst

pump A

injection system

heat exchanger A injection system

pump A

INEL-A" ;O,5B

Figure 64. LOFT ECCS simplified schematic diagram.
Control

mechanism

AP measurement

for liquid level

LIT-P120-44 (ACCLIT-P120-30 (ACO-

-Accumulator flow

Inlet

• Borated water

Inlet

Flow reversing

vane

-Variable

standpipe

.r Manhole

-Liquid temperature,

thermocouple

TE-P120-41 (ACC.A)

TE-P120°27 (ACC.B)

Accumulator

tank pressure

PT-P1 20-43 (ACC-A)

PT-P120-29 (ACC-B)

Borated water experimental

ECC Injection line

IN EL-MCL-4900

Figure 65. LOFT accumulator and variable standpipe.

95
TABLE 4. LARGE BREAK EXPERIMENTS INVESTIGATED

Reactor

Power

Excerlment MW

Pump

Coast

down

Injection Initiated Initiated Acc.

ECC Acc. HPIS LPIS Empty

Mode s s s s

Core

Quench

S

L2-5

LP-02-6

LP-LB-1

36.0 rapid ILCL

46.0 normal ILCL

NRC

49.3 rapid ILCL

UK

16.8

17.5

17.5

19.0

344.5

23.9

24.1

15.0

515.8

37.3

37.1

30.8

53.0 65

50.0 56

40.0 72

LP-FP-IA 37.0 rapid PPS

LP-FP-1 37.0 rapid U.P.

ILCL

a. Unintentional injection in the U.P.

30.0 A 63.0 33

B 72.0

-- 507.0 37 6a

in LP-FP-1 started at about 2 s. Highest injection rate

was 25 to 75 s. 80% of the core was completely quenched at 60 s with 200 kg of water.

96
TABLE 5. ACCUMULATOR CONFIGURATIONS

Liquid Gas Liquid Liquid

Inject Level Vol. Vol. Injict.

Experiment _t. Acc. m m m m

L2-5 A ILCL 2.1 0.84 1.52 1.96

LP-02-6 A ILCL 2.1 0.95 1.236 1.69

LB-LB-I A ILCL 2.362 0.66 0.724 1.18

LP-FP-IA A L.P. 2.15 1.189 2.93 3.37

B D.C. 2.10 1.133 2.81 3.38

LP-FP-I A ILCL 2.18 1.189 1.64 2.08

B U.P. 2.12 1.133 2.17 2.17

97
2.4

2.0

E

-5 1.6

1.2

0.8

Figure 66

LE-ECC-01A (LB-1) .......... LE-ECC-01A (L2-5)

LE-ECC-O1A (LP-02-6)

................ . ......................

, I , , , I , , , , I , , I , , , I , , , I , I , , , , I ,J

-10 0 10 20 30 40 50 60

Time after rupture (s)

S. Accumulator A liquid level during Experiments

LP-LB-1.

70 80

P472 MM-1087-O1

L2-5, LP-02-6, and

98
injected from the accumulator during experiment LP-LB-1. In all three

experiments the liquid level in the accumulator started to decrease about

18 s after experiment initiation. HPIS (Figure 67) was activated in L2-5 and

LP-02-6 at nearly the same time but different amounts were injected. In

LP-LB-1 no HPIS was activated. LPIS injection was activated in L2-5 and

LP-02-6 at the same time and the injection rates were very similar as shown

in Figure 68. In LP-LB-1 the LPIS was activated earlier, at 31 s but with

only about half the capacity of L2-5 or LP-02-6. The general shape of the

injection rate curve was analogous to previous experiments L2-2 and L2-3.

The differences shown in Figure 68 are mainly due to the earlier activation

and lower flow rates.

The ECCS affected the core thermal behavior differently in these three

experiments due to differences in operation. The effects of ECCS injection

on phenomena in the reactor vessel during these three experiments can be

easily understood if we keep in mind the following facts:

- Most of the water injected during the initial 20 s after ECCS

initiation originates from the accumulator (HPIS and LPIS amounts to

about 0.2%)

- The injection point in all three experiments was the same

- The injected ECC was partly lost through a bypass to the broken loop

cold leg

- LP-LB-1 and LP-02-6 are alike with regard to the initial power level

but different in pump behavior

- LP-LB-1 and L2-5 are alike in pump behavior but different in initial

power

- The least amount of ECC was injected in LP-LB-I.

The completion of the refill phase is indicated by quenching of fluid

thermocouples just below the core at the lower end box. Figure 69 shows the

behavior of such a thermocouple during the experiments. In L2-5 and LP-02-6

the thermocouple quenches at nearly the same time (31 s) but 2 s later in

LP-LB-I. This small delay is attributed to a slightly smaller injection rate

99
2.0

1.5

Coc 1.0

0

S0.5

0.0

-0.5

KL FT-P128-104 (LB-1)

- - ----- FT-P128-104 (L2-5)

FT-P128-104 (LP-02-6

~7

•i . .... ,./ • .• . .. .. ... .. j

FLJ

-110 0 0 20 30

Time after

40 50

rupture (s)

60 70 80

P472 MM-1087-02

Figure 67. HPIS injection rates during

LP-LB-1.

Experiments L2-5, LP-02-6, and

100
8.0

6.0 1-1-"IZU-Ubb (LF,

(0

c 4.0

0

02.0

0 .0 --------------------------... .... ...

-2.0

-10 0 10 20

Time

30 40 50 60 70 80

after rupture (s) P472 MM-1087-03

Figure 68. LPIS injection

LP-LB-1.

rates during Experiments L2-5, LP-02-6, and

800

700

600

CL0.

E 500

400

300

-10 0 10 20

Time

30 40 50

after rupture (s)

60 70 80

P472 MM-1087-04

Figure 69. Temperatures measured at the lower end box during Experiments L2-5, LP-02-6, and LP-LB-1.

101
in LP-LB-1 (Figure 4). Approximately 512 kg of water were delivered from the

accumulator between 18 s and 31 s in LP-LB-1 versus 625 kg delivered in the

same time in L2-5 and LP-02-6.

Larger differences among the experiments are seen by comparing the

behavior of cladding thermocouples in the hot region of the core as shown in

Figure 70. For experiment LP-LB-1 the quenching time of the 5H06-027

thermocouple is representative of the entire core (72 s in Table 3). This

thermocouple indicated the highest temperature measured in LP-LB-1. The hot

spot in L2-5 and LP-02-6 was located at a different position as shown in

Figure 71. The quenching time of thermocouple 5104-27 was representative of

the entire core in L2-5 and LP-02-6.

The quenching time of thermocouples in the upper plenum near the top of

the core is a good indication of the end of the reflood phase. Figure 72

shows the behavior of one such upper plenum thermocouple. The final quench

of this thermocouple, indicating the end of the reflood phase, occurred at

nearly the same time in L2-5 and LP-02-6 (57 and 58 s) but occurred about

10 s later (68 s) in LP-LB-1. This delay in quench in LP-LB-1 can be related

to the reduced amount of water injected in that experiment (Figures 66

through 68). The end of the reflood phase was coincident with the time of

complete core quench only for the LP-02-6 experiment. In L2-5 and LP-LB-1

the end of the reflood phase was about 8 s and 4 s earlier than the total

core quench, respectively. This indicates that in the case of higher

cladding temperatures the reflood liquid level passes the hot spot without

quenching the cladding. During experiment LP-02-6 the clad temperatures were

significantly reduced due to the early bottom-up blowdown quench which

allowed the core to quench simultaneously with the reflood front.

As mentioned earlier the first fission product release experiment in LOFT

was attempted on 12.12.84 and conducted successfully on 19.12.84. During the

first attempt (LP-FP-IA) the experiment was terminated during the blowdown

phase with PPS (Plant Protection System) initiation. The second attempt on

19.12.84 (LP-FP-1) was successful despite the occurrence of an unintentional

ECC injection in the upper plenum which substantially delayed fuel rod

rupture. The experiment was terminated as planned through combined ECC

102
1400

1200

2

21000

0

E 800

600

400

-10 0 10 20

Time

30 40 50 60 70 80

after rupture 9 (s) P472 MM-O187-06

Figure 70. Cladding temperature in the central bundle during Experiments

L2-5, LP-02-6, and LP-LB-1.

1400

1200

2

21000

E 800

600

400

-10 0 10 20 30 40 50

Time after rupture (s)

60 70 80

P472 MM-1087-06

Figure 71. Cladding temperature in the central bundle during. Experiments

L2-5, LP-02-6, and LP-LB-1.

103
600 ....

TE-

- TE1• •

500

400

-10 0 10 20 30 40 50

Time after rupture (s)

Figure 72. Temperatures measured in the upper pl

60 70 80

P472 MM-1087-07

enum during Experiments

L2-5, LP-02-6, and LP-LB-1.

104
injection in the upper plenum and intact loop cold leg. In summary, three

ECC injection modes occurred in the first LOFT fission product release

experiment:

- Combined lower plenum and downcomer injection (PPS in LP-FP-1A)

- Upper plenum injection (unintentional ECC pipe draining in LP-FP-1)

- Combined upper plenum and intact loop cold leg injection (ECC in

LP-FP-1)

The upper plenum ECC injection system in LOFT was specifically designed

for LP-FP-I using accumulator B and the connecting pipes to the Reactor

Vessel as shown in Figure 73. The total volume of the 30 m long piping was

0.458 m3 . The volumes of the different parts of the piping are given in

Figure 73. The configuration of the injection nozzles in the upper plenum is

shown in Figure 74. Eight nozzles arranged to inject towards the peripheral

bundles were located about 13 cm above the central bundle and 6 nozzles

arranged to inject in the central bundle were located 42 cm above core

outlet. During the PPS operation in LP-FP-1A the entire water volume in the

accumulator and some N2 were injected. In the week between 12.12 and

19.12.84 the pipes shown in Figure 73 were not degassed. The unintentional

injection during LP-FP-1 was caused by the expansion of the residual N2 in

the ECCS line.

LP-FP-1A results show that the HPIS initiation about 12 s after rupture

(Figure 75). Accumulator initiation occurred several seconds later, at

nearly the same time as in L2-5, LP-02-6 and LP-LB-1. The end of the refill

phase occurred at about 31 s approximately the same time as in the other

experiments, as shown in Figure 76. The accumulators inject directly in the

downcomer and the lower plenum in the PPS mode to exclude bypass losses to

the broken loop cold leg. In view of this and since the end of the refill

phase in LP-FP-1A occurred at the same time as in the earlier large break

LOCA experiments, the bypass losses during intact loop cold leg injection are

concluded to be negligible.

105
4

CV

6 in.-ECC-107.VB.

4

WF CV.P.120-1l

P-120

31

Containment

vessel

wall

in.-ECC-103-VA

V-6332-IV A-/

V-5112-IVV

CV.P-120-64-

67

in.-ECC-109-VS

Figure 73.

Central ECC

reflood nozzles (6)-'

LP-FP-1 upper head injection piping..

86.62

Steam sample line Sj

Peripheral ECC reflood nozzles

Figure 74. LP-FP-1 CFM upper structure.

106
2.5

2.0

1.5

_ 1.0

0.5

0.0

FT-P128-085 (LP-FP-1A) .............. FT-P128-104 (LP-FP-1A)

, ,--- I , - , ,- , ,- I ,

-10 0 10 20 30 40 50

Time after rupture (s) P472 MM-1087-10

Figure 75.

600

2 500

,o

E

1. 400

300 . . . .

HPIS volumetric flow during Experiment LP-FP-1A.

-10 0 10 20 30 40

Time after rupture (s)

50

P472 MM-1087-11

Figure 76. Liquid temperatures at the lower

LP-FP-IA.

107

end boxes during Experiment
Review of cladding temperatures throughout the core during LP-FP-1 and

LP-FP-1A provides some insights on ECCS effectiveness with regard to

different injection locations. Figures 77, 78 and 79 compare cladding

temperatures measured at three levels during both experiments. All three

figures indicate similar thermal core behavior at the three elevations for

the early part of the transient. Strong differences between the two

experiments begin after 22 s when the core showed more tendency to heatup in

LP-FP-1A than in the LP-FP-I. The unintentional injection, explained in

Section 3.6, which started early in the blowdown phase and was effective at

about 24 s, is the cause of this difference. This unintentional injection

penetrated the entire core causing complete quench at about 25 s (Figure 77)

and 24 s (Figure 79) in the lower part and upper part of the core,

respectively. In the hot region of the core (Figure 78) the unintentional

injection also caused a quench; however, the quench was not complete until

about 44 s. Figures 77-79 illustrate the effectiveness of upper plenum

injection in mitigation of a core temperature escalation. The upper plenum

injection seems to be even more effective than a full PPS action because only

200 kg (estimated) resulted in almost complete core quench within about

44 s. The short quench time in LP-FP-1 and also in case of LP-FP-1A was

possible because of the low maximum temperature (less than 700 K) which was a

result of the early quench phenomena described in Section 4.1.

The combined upper plenum and cold leg injection in LP-FP-1 started at

344 s when predetermined termination conditions were reached. The central

fuel bundle was quenched at about 369 s (Figure 80) and the whole core was

quenched at about 374 s (Figure 81).

In summary, the conclusions are:

1. The ECCS configuration in LOFT has significant effect on core quench

but not on lower plenum refill

2. Early quench phenomena reduced the time of complete core quench by

about 30%

108
800

700

600

CL

E500

400

300 '

-10

Figure 77.

0 10 20 30 40 50

Time after rupture (s) P472 MM-1087-12

Cladding temperatures at the lower core region during

Experiments LP-FP-1 and LP-FP-1A.

E

a,

900

800

700

600

500

400

300

-10

Figure 78.

0 10 20 30 40 50

Time after rupture (s) P472 MM-1087.13

Cladding temperatures in the upper core region during

Experiments LP-FP-1 and LP-FP-1A.

109
900

800 . I -U-U l~r-rr- U .

700

600 ........ ..... '

5. 600

400

.300 . . . ... . . .. .

-10 0 10 20 30 40 50

Time after rupture (s) P472 MM.1087-14

Figure 79. Cladding temperatures in the central core region during

Experiments LP-FP-1 and LP-FP-1A.

.o.°. .. .... ............. °'

1000

800

4-I

600 I

I-I

400 ___ TE-5K09-021 (LP-FP-1) "".-*-•v\ 

-........ TE-5E09-021 {LP-FP-1)

- TE-5E07-043 (LP-FP-1)

300 310 320 330 340 350 360 370 380 390 400

Time (s) P472 MM-1087-08

Figure 80. Cladding temperatures in the central fuel element during

Experiment. LP-FP-1.

110
1200

1000

800

E 600

t2

400

200

300 310 320 330 340 350 360 370 380 390 400

Time (s) P472 MM-1087-09

Figure 81. Cladding temperatures in a peripheral bundle during Experiment

LP-FP-1.

111
3. The shortest quench time was achieved during PPS operation which

involved a higher.net rate of accumulator injection

4. Upper plenum injection is highly effective (about 200 kg of

unintentionally injected.water quenches 80% of the core)

5. Upper plenum injection reduced the final quench time-in LP-FP-1-by

about 30%.

112
5. SUMMARY AND CONCLUSIONS

Forty-four experiments were completed in the LOFT facility over a nine

year period ending with a severe fuel damage experiment in July 1985. These

experiments were conducted at typical initial and boundary conditions

associated with loss of coolant accidents and anticipated transients in

commercial.PWRs. The research program included six nuclear large break LOCA

experiments, the primary objective of which was to obtain data on LOCA

phenomena and system response for a range of initial and boundary conditions

which could be used for reactor system code development and assessment. The

objectives, design, and principal results of the nuclear large break

experiments are described. The important thermal-hydraulic phenomena

measured in the large break transients and their significance are discussed

in the principal areas of analysis that have been undertaken. The principal

finding from the large break experiments is that, for the degrees of severity

in initial and boundary conditions, the measured fuel cladding temperatures

remained well below the peak cladding licensing limit temperatures.

The data obtained from the LOFT large break LOCA experiments provided new

insight into phenomena associated with the large break LOCA. One of the most

important phenomena, observed for first time in the LOFT transients, is fuel

cooling/cladding quench during blowdown. This phenomenon is very important

to the degree of transient severity because it removes a large part of the

stored energy from the fuel early in the transient. The cooling/quench

phenomena was determined to be caused by system hydraulics in response to the

operational characteristics of the primary coolant pumps relative to the

transition from subcooled to saturated choked flow at the break. The

significant finding was that the cooling/quench phenomena would occur in all

conditions except for a pump trip concurrent with break initiation and

decoupling from the flywheels. Similar limiting conditions are expected to

be required to suppress the phenomena in commercial PWRs.

Separate effect experiments in other facilities and analysis of LOFT data

showed conclusively that the blowdown cooling/quench in LOFT large break LOCA

experiments is real. However, the thermocouples do reduce the blowdown peak

cladding temperature because of an induced delay to DNB. Fin cooling

113
subsequent to DNB was not found to adversely affect measurement accuracy. In

contrast, surface cladding thermocouples are recognized to have noticeable

effects on reflood-quenching which occurs at much slower rates compared to

the observed quenching during blowdown.

Examination of the LOFT large break LOCA experiments provided important

insight on emergency core cooling (ECC) performance during large break

transients. In general, the experiment results have shown that the ECCS

operation even in degraded conditions was effective in core quench and

transient recovery. The hot wall delay time was at most 2 s. Only a small

part of the ECC water is lost through downcomer bypass to the broken loop

cold leg indicating that the"downcomer bypass", which is one of the concerns

in licensing, is not of concern. Cooling phenomena during blowdown can

reduce the time'to final.quench by about 30% because the reflood quench is

strongly dependent on cladding temperature levels at the end of the refill

phase. Experiment LP-FP-1 which included upper plenum ECC injection showed

that ECCS mode is highly effective and that relatively small-amounts of water

can quench the core.

114
6. REFERENCES

1. C. L. Nalezny, "Summary of Nuclear Regulatory Commission's LOFT Program

Experiments", NUREG/CR-3214, EGG-2248, July 1983.

2. D. L. Reeder, "LOFT System and Test Description (5.5-Ft Nuclear Core

LOCEs)", NUREG/CR-0169, EGG-2037, March 1983.

3. C. L. Nalezny, "Summary of the Nuclear Regulatory Commissions LOFT

Program Research Findings", NUREG/CR-3005 EGG-2231, April 1985.

4. L. J. Ybarrondo et al., "Examination of LOFT Scaling", Proceedings of

ASME Winter Annual Meeting, New York, New York, November 17-22, 1974.

5. P. A. Harris et al., "Power Ascension Test Series L2, LOFT Experiment

Operating Specification", Vol. 2, NE L2 Series, Rev. 2, July 1978.

6. M. McCormic-Barger, "Experiment Data Report for LOFT Power Ascension

Test L2-2", NUREG/CR-0492, TREE-1322, February 1979.

7. W. H. Grush, et al., "Results and Predictions of Scaled, Nuclear Large

Break Loss-of-Coolant Experiments", ASME Winter Meeting, November 16-21,

1980, Chicago, Illinois, (80-WA/HT-49).

8. P. G. Prassinos, B. M. Galusha, D. B. Engleman, "Experiment Data Report

for LOFT Power Ascension Experiment L2-3", NUREG/CR-0792, TREE-1326,

July 1979.

9. J. C. Lin, "Posttest Analysis of LOFT Loss-of-Coolant Experiment L2-3",

EGG-LOFT-5075, March 1980.

10. P. D. Bayless and J. M. Divine, "Experiment Data Report for LOFT Large

Break Loss-of-Coolant Experiment L2-5", NUREG/CR-2826, EGG-2210, August

1982.

11. P. N. Demmie, T. H. Chen, S. R. Behling, " Best Estimate Predictions for

LOFT Experiment L2-5", EGG-LOFT-5869, May 1982.

12. E. M. Feldman, ,"OECD LOFT Project Experiment Specification Document Cold

Leg Large Break Experiment LP-02-6", OECD LOFT-T-3402, July 1983.

13. J. P. Adams, K. G. Condie and D. L. Batt, "Quick-Look Report on OECD

LOFT Experiment LP-02-6", OECD LOFT-T-3404, October 1983.

14. K. G. Condie, P. N. Demmie, E. W. Coryell, "Best Estimate Prediction for

the OECD LOFT Project Large Cold Leg Break Experiment LP-02-6", OECD

LOFT-T-3403, September 1983.

15. J. C. Birchley, "OECD LOFT Project Experiment Specification Document for

Cold Leg Large Break Experiment LP-LB-1",OECD LOFT-T-3502, December

1983.

115
16. J. P. Adams, J. C. Birchley, "Quick-Look Report on OECD LOFT Experiment

LP-LB-1", OECD LOFT-T-3504, February 1984.

17. V. T. Berta, E. W. Coryell, "OECD LOFT Project Experiment specification

Document Fission Product Experiment LP-FP-1", OECD LOFT-T-3702, October

1984.

18. A. B. Wahba, "Thermal-Hydraulic EASR of OECD LOFT LP-FP-1", OECD

LOFT-T-3709, November 1986.

19. E. W. Coryell, H. G. Glaeser, "Best Estimate Prediction for OECD LOFT

Project Fission Product Experiment LP-FP-1", OECD LOFT-T-3703, October

1984.

20. J. P. Adams et al., "Quick-Look Report on OECD LOFT Experiment LP-FP-1",

OECD LOFT-T-3704, March 1985.

21. J. P. Adams, "Monitoring PWR Reactor Vessel Liquid Level with SPNDs

during LOCAs", Proceedings of the Topical Meeting on Advances in

Reactor Physics and Core Thermal Hydraulics, Kiamesha Lake, New York,

Sept. 22-24 1982, NUREG/CP-0034, Vol. 2, August 1982.

22. J. P. Adams and V. T. Berta, "Monitoring Reactor Vessel Liquid Level

with a Vertical String of SPNDs", ANS Winter Meeting, San Francisco,

California, 1983.

23. S. N. Aksan, "Evaluation of Analytical Capability to Predict Cladding

Quench", EG&G-LOFT-5555, August 1982.

24. R. A. Nelson, "Forced-Convective Post-CHF Heat Transfer and Quenching",

Journal of Heat Transfer, Vol. 104, February 1982.

25. R. C. Gottula, J. A. Good, "The Effects of Cladding Surface

Thermocouples on the Behavior of an Electrical Heater Rod", EG&G Idaho

Report LO-00-80-115, 1982.

26. R. C. Gottula, "Effects of Cladding Surface Thermocouples and Electrical

Heater Rod Design on Quench Behavior", NUREG/CR-2691, EG&G-2186,

February 1984.

27. R. C. Gottula, M. L. Carbonneau, E. L. Tolman, "An Assessment of LOFT

Fuel Rod Quench Behavior Based on Electric Rod Quench Tests,"

Proceedings of the International Symposium on Fuel Rod

Simulators-Development and Application, Gaithersburg, Tennessee, USA,

October 1980, (Conf-801091, May 1981).

28. S. N. Aksan, E. L. Tolman, R. A. Nelson, "Application of Analytical Capability to Predict Rapid Cladding Cooling and Quench During the

Blowdown Phase of a Large Break Loss-of-Coolant Accident", Proceedings

of the Second International Topical Meeting on Nuclear Reactor

Thermal-Hydraulics (Vol. 3), Santa Barbara, CA, January 11-14, 1983.

116
29. T. R. Yackle, "An Assessment of the Influence of Surface Thermocouples

on the Behavior of Nuclear Fuel Rods During a Large Break LOCA", The 8th

Water Reactor Safety Research Information Meeting, Gaithersburg,

Maryland, USA, NUREG/CP-0023, Vol. 4, October 1980.

30. R. W. Garner, P. E. MacDonald, "Power Burst Facility Thermocouple

Effects Test Results Report, Test Series TC-I, TC-2, and TC-3",

NUREG/CR-2665, EGG-2190, June 1987.

31. R. W. Garner, "PBF Thermocouple Effects Tests, Test Series TC-4, Quick

Look Report", EGG-TFBP-5465, August 1981.

32. D. S. Seegmiller, "LOFT F1 and F2 Bundle Thermocouple Perturbation

Analysis", EG&G Idaho Internal Technical Report L0-14-82-082, Released

Report RE-A-82-044, USNRC-P394, February 23, 1983.

33. H. G. Kraus, "COUPLE/MODS, A General Purpose Two-Dimensional Finite

Element/Finite Difference Heat Conduction Code", RE-A-82-035, EG&G

Idaho, December 1981, EG&G Idaho, Inc.

34. J. J. Pena, "Post Experiment Analysis of the OECD LOFT Experiment

LP-02-6, OECD LOFT-T-3408", May 1985.

35. T. D. Knight, "TRAC Analysis of The OECD LOFT Experiment LP-02-6", OECD

LOFT-T-3409, October 17, 1985.

36. R. C. Gottula, K. G. Condie, R. K. Sundaram, S. Neti, J. C. Chen, R. A.

Nelson, "Forced Convective, Nonequilibrium, Post-CHF Heat Transfer

Experiment Data and Correlation Comparison Report", NUREG/CR-3193.

37. S. R. Fisher et al., "RELAP4/MOD6: A Computer Program for Analysis of

Nuclear Reactors and Related Systems, Users Manual," CDAP-TR-003,

January 1978.

38. E. L. Tolman, M. L. Carboneau, "Comparison of Semiscale MOD-i and LOFT

Large Break Loss-of-Coolant Experiments: A Case for Nuclear

Experiments", ANS Winter Meeting, San Francisco, California 1981.

39. H. Grutter et al., "NEPTUN Bundle Reflooding Experiments: Test Facility

Description", EIR Report No. 386, Wurenlingen, March 1980.

40. S. N. Aksan, "NEPTUN Bundle Boil-off and Reflood and Experimental

Program Results", The 10th Water Reactor Safety Research Information

Meeting, Gaithersburg, Maryland, USA, NUREG/CP-0041, Vol. 1 (1982).

41. M. Richner, "Comparison of NEPTUN-I and NEPTUN-II Reflooding

Experiments: External Thermocouple Effect", (In German) EIR Internal

Report, (TM-32-87-06) Wurenlingen, May 1987.

42. C. Vitanza et al., "Blowdown/Reflood Tests With Nuclear Heated Rods

(IFA-511.2)" OECD Halden Reactor Project, HPR-248, May 1986.

117
43. T. Johnsen, C. Vitanza, "Blowdown/Reflood Tests With Semiscale

Heaters (IFA-511.3, Data Collection)", OECD Reactor Project, HWR-17,

May 1981.

44. T. Johnsen, C. Vitanza, "Results of LOCA Tests With Semiscale Heaters

(IFA-511.5)", OECD Halden Reactor Project HPR-313, May 1984.

45. C. Vitanza, T. Johnsen, "Results of Blowdown/Reflood Tests With

REBEKA Electric Simulators (IFA-511.4, July 2)", OECD Halden Reactor

Project, HWR-85, May 1983.

46. P. Ihle, K. Rust, "PWR Reflood Experiments Using Full Length Bundles

of Rods With Zircaloy Claddings and Alumina Pellets (Results of the

SEFLEX Program)", Proceedings of Third International Topical Meeting

on Reactor Thermal Hydraulics, Newport, Rhode Island, USA,

October 15-18, 1985 (Vol. 1).

47. S. N. Aksan et al., "Boil-off Experiments With the EIR-NEPTUN

Facility: Analysis and Code Assessment Overview Report", EIR Report

No. 623, September 1987.

118
APPENDIX A

CODE CALCULATIONS

A-i
APPENDIX A

CODE CALCULATIONS

The predictions for LOFT large break LOCA experiments (discussed in the

section "LOFT Nuclear Large Break Experiments") were performed using codes of

the earlier generation such as RELAP4/MOD6, RELAP5/MOD1 or TRAC-PD2.

Although, these codes are currently being used in specific applications they

do not represent the current knowledge in reactor safety. In this section we

present examples of the application of current advanced thermal-hydraulic

codes for analysis of LOFT large break LOCA experiments. We understand that

these thermal-hydraulic codes will continue to be refined and maintained in

the future. These codes are: DRUFAN, RELAP5/MOD2 and TRAC-PFI/MODl.

A few relatively recent calculations for LOFT large break LOCA

experiments were selected and are briefly discussed to illustrate performance

of these codes. Table Al shows which codes were applied to pre- and

post-experiment analyses of LOFT large break experiments, and Table A2

summarizes performance of the current codes in presented calculations.

Additional information on code performance in simulation of the large break

LOCA can be found in Reference Al which provides review of analyses of

Experiment LP-02-6 with RELAP5/MOD2, TRAC-PD2/MODI, DRUFAN-2 and

TRAC-PFl/MODl codes.

Al. The Thermal Hydraulic Code DRUFAN-02

The Code DRUFAN has been developed in the Gesellschaft fur

Reaktorsicherheit (GRS) mbH, Germany for the simulation of LWR reactors. The

code is applied to the analysis of large, medium sized and small breaks and

selected transients.

The physical system is described by "lumped parameter" control volumes

which are connected by flow paths. Also, valve, pump, accumulator, steam

generator and pressurizer models are available for simulation. The numerical

method used in DRUFAN is the lumped parameter approach. The ordinary

differential equation system of the thermo- and fluid-dynamic model is based

A-2
TABLE Al. CODE APPLICATION TO LOFT LB EXPERIMENTS

_L2-

P PA

X

L2-3

P PA

X

L2-5

P PA

X

X X

LP-02-6

P PA

LP-LB-1 LP-FP-l

P PA P PA

RELAP4/MOD6

RELAP5/MODI

TRAC-PD2

DRUFAN-02

RELAP5/MOD2

TRAC-PFI/MODI

X

x

X X X

X

X

X X X

X

X

X

X

X

X

X

X

X

P - Predictions

PA - Postexperiment Analyses

A-3
TABLE A2. PERFORMANCE OF CURRENT THERMAL-HYDRAULIC SYSTEM CODES

DRUFAN-02

1.2-5

PSC

Pressure: Good

ILCL Flow:

RCP:

Break Mass Flow

Subcooled cold leg

break flow

underpredicted.

HIL break flow Initial

2-phase flow

underpredicted. Effects

of intact loop draining

into upper plenum

(increase of break flow)

not calculated.

Good. Shows

osccilations during

accumulator injection.

Blowdown Quench

Top-down quench not

calculated.

Effects in peripheral

Bundles not calculated.

PCT Reflood Quench

Satisfactory

ECC

RELAP5/MOD2 Pressure: Good

ILCL Flow:

Initially

underpredicted

(5s).

RCP:

Calculated

coastdown speed

deviates from

measured.

3=, !

4•

Bottom-up quench

calculation only using

reflood option. Quench

not as complete as in

experiment. No rapid

return to DNB.

No heatup in upper part

of core. Top-down

quench too early; too

much water kept above

the core.

No bottom-up quench.

only some cooling in

lower core.

Top-down quench: Some

cooling and quenching

calculation in upper

core, not as complete

as in experiment.

Mostly missing in

peripheral bundles.

Good with good

timing.

At lower elevation

correct timing and

rate. delayed at PCT

elevation.

Initial injection rate

good. Later

overpredicted.

TRAC-PF1

MODI

LP-02-6

Pressure: Good

ILCL Flow:

Initially

underpredicted

(ss).

RCP: Initial

condition speed

too high.

Later coastdown

overpredicted.

Good Blowdown PCT quite

correct.

Reflood PCT

overpredicted

significantly.

Reflood quench

delayed and lower

temperature than

measured.

Accumulator empties

earlier than in

experiment calculation

injection. Condition

results in PCS pressure

drop and overprediction

of accumulator flow.
on the conservation laws for vapor mass, liquid mass, overall energy and

overall momentum. The liquid and vapor phases are treated as a homogeneous

mixture, or in the case of mixture level-tracking as a nonhomogenous mixture

[A2].

The velocity difference of the liquid and vapor phase may be determined

by a drift flux model. The differential equations are integrated by an

explicit-implicit integration method with automatic control of time step,

order of consistency and local discretization error [A3].

The entire range from subcooled liquid to superheated vapor including

nonequilibrium effects is simulated by assuming either the liquid or vapor

phase to be saturated.

The table for the determination of critical discharge rate at the break

is calculated by a one-dimensional nonequilibrium model which is based on the

same four conservation equations used for the "lumped parameter" control

volumes. This model takes into account the geometry of the discharge flow

path [A3].

For the simulation of structures, electrical heaters and fuel rods a heat

conductor model and point neutron kinetics model are used. The heat transfer

coefficients coupling the structure and thermal hydraulic model are

determined by a comprehensive heat transfer package. The heat transfer

package also contains a set of critical heat flux correlations.

The LOFT input model was developed for the L2-3 post experiment

calculation and was then used for prediction of L2-5 and LP-02-6. The

primary and secondary loops of LOFT were simulated by 79 control volumes, 98

junctions and 118 heat conductors. All parts of the LOFT facility except the

blowdown suppression tank were simulated. The LOFT core was simulated by two

parallel channels. The hot channel simulated the central region and included

two heat conductors with a radial power factor of 1.4 and 1.2. The second

channel simulated all fuel rods outside the central region and included two

heat conductors with the power factors 1.0 and 0.75. Cross flow was allowed

A-5
between the two parallel channels and was simulated by horizontal junctions.

The heat generation in the active core was determined by a point neutron

kinetic model. A value of 6000 W/m2K was used for the gap conductance.

The upper plenum was divided into three levels and represented by 4

control volumes. The central part was simulated by two parallel control

volumes to account for differences in geometry above the fuel bundles. The

bypass between the downcomer and the upper plenum was modelled allowing flow

of 5% of the total steady state loop flow.

Two parallel channels were used to represent the nonuniform transient

behavior of the fluid in the downcomer. Each channel was divided into four

axial cells connected by vertical junctions. Cross flow was permitted

between the parallel channels through four horizontal junctions. The fluid

between the core filler pieces was represented by one cell, which admits a

core bypass of 6% of the steady state flow.

The inner and outer heat structures were modelled. The heat losses on

the primary side were assumed to be equally distributed and totaled 150 kw.

All essential parts and structures of the LOFT steam generator were modelled.

The total heat loss through the outer structures was 50 kW. The heat

generation of the primary coolant pumps was also taken into account.

The pressurizer was simulated by one control volume and a second control

volume simulated the surge line. The broken loop nodalization included the

steam generator and pump simulators. The critical discharge rate was

determined by the 1-D (BIASI) discharge model.

DRUFAN-02 was used for pre and post-experiment calculations of the large

break LOFT experiments. Experiment L2-5 was chosen to be international

standard problems No. 13 (ISP 13). DRUFAN-02 was used for the blind

calculations [A4]. The agreement between measured and calculated peak clad

temperature was satisfactory as shown in Figure Al. Also good agreement was

formed by comparing other thermal hydraulic parameters such as system

pressure (Figure A2) and break flow (Figures A3 and A4). However,

A-6
LOFT L2-5: PREDICTIONS USING DRUFAN-02

1000

U

w

LU

LU

I0

z

a

a

U

800

600

400

200

TIME AFTER RUPTURE (S)

Figure Al. Measured and calculated with DRUFAN-02 peak cladding temperatures

for Experiment L2-5.

LOFT L2-5: PREDICTIONS USING DRUFAN-02

20

16

a 12

(tu

0.

0.

10 15

TIME AFTER RUPTURE (S)

Figure A2. Measured and calculated with DRUFAN-02

Experiment L2-5.

primary system pressure for

A-7
LOFT L2-5: PREDICTIONS USING DRUFAN-02

I I YIv%

0

0

0

-J

I'

0

0

80o600-

400-

200-

O G(46) CALCULATED

o FR-BL-OO1 MEASURED

-I I

0 -I 0 5 10 15 20 25 30

TIME AFTER RUPTURE (S)

Figure A3. Measured and calculated with DRUFAN-02 cold leg break flow for

Experiment L2-5.

LOFT L2-5 PREDICTIONS USING DRUFAN-02

0

0

30

TIME AFTER RUPTURE (S)

Figure A4. Measured and calculated with DRUFAN-02 hot leg break flow for

Experiment L2-5.

A-8
multidimensional phenomena due to quench processes could not be calculated.

Also rewetting time and subsequent CHF did not agree consistently with

measured values specially in the peripheral bundles.

A2. RELAPS/MOD2

The RELAP5/MOD2 code, developed at the Idaho National Engineering

Laboratory, is a best-estimate thermal-hydraulic system code for simulation

of a variety of hydraulic and thermal transients in both nuclear and

nonnuclear systems involving steam-water-noncondensible fluid mixtures [A5].

The RELAP5/MOD2 hydrodynamic model is a one dimensional, transient

two-fluid model for flow of a two-phase steam-water mixture that can contain

a noncondensible component in the steam phase and/or a nonvolatile component

in the liquid phase. The basic field equations for the two-fluid

nonequilibrium model consist of two phasic continuity equations, two phasic

momentum equations, and two phasic energy equations. The system model is

solved numerically using a semi-implicit finite difference technique. For

steady state and very slow transient there is a user option for a

nearly-implicit finite difference technique which allows violation of

material Courant limit.

The code includes many generic component models from which general

systems can be simulated. The component models include pumps, valves, pipes,

heat structures, reactor point kinetics, electric heaters, jet pumps,

turbines, separators, accumulators, and control system components. In

addition special process models are included for effects such as form loss,

flow at an abrupt area change, branching, choked flow, boron tracking, and a

noncondensible gas.

The first independent application of RELAP5/MOD2 for analyses of a LOFT

experiment was for post experiment calculations of Experiment LP-02-6 [A6].

Code version 36 was used. In these calculations the input model was used

based on a RELAP5/MOD1 model for predictions of this experiment. This model

included a split downcomer and split core channel for better simulation of

the 3-D effects strongly present in a large break LOCA. The parallel

A-9
channels were connected with cross flow junctions. The gap conductance model

was used only for the prepressurized fuel rods in the center fuel module.

Recorded setpoints and initial conditions were used in this analysis.

Figure A5 shows the calculated and measured primary system pressure. The

calculated pressure follows the measured pressure, relatively close although

it is lower than the measured during initial 10 s of the transient and higher

in the latter part of the depressurization. The code calculated a period of

positive core flow during the blowdown. However, the balance of flows in the

broken loop cold leg and the intact loop cold leg indicates a possible

smaller flow through the core than in the experiment which is attributed to a

less-than-measured calculation of the early intact loop mass flow rates. The

blowdown bottom-up quench was predicted by the code however not to the extent

as measured in the experiment. The cladding at the hot elevation was cooled

rather than quenched (Figure A6). The reason may be the lower mass flow rate

calculated during that time or that the film boiling correlation used is not

valid for low quality flows. The code also calculated the top-down quench

but earlier than measured. It was concluded that the CCFL conditions are not

adequately modeled by the code. The calculation places too much liquid on

the top of the core which acts to prevent cladding heatup in the upper part

of the core.

A3. TRAC-PFI/MODI

The TRAC-PF1/MOD1 code was developed at Los Alamos National Laboratory

under the sponsorship of the U.S. NRC. TRAC-PFI is the successor of the

TRAC-PD code.

TRAC-PFI/MOD1 is a best-estimate computer code for analysis of postulated

accidents in light water reactors. The code [A7] features a three

dimensional treatment of the reactor vessel and associated internals, two

phase nonequilibrium hydrodynamic models, flow regime dependent constitutive

relations, optional reflood tracking capability for both bottom-reflood and

falling-film quench fronts, and consistent treatment of entire accident

scenarios, including the generation of consistent steady state conditions.

A-10
:3

Ca

C20

15

10

5

0

0 10 20 30 40 50 60

Time after rupture (s) L221-KM258832

Figure A5. Measured and calculated

for Experiment LP-02-6.

with RELAP5/MOD2 primary system pressure

cc

E

0

1100

1000

900

800

700

600

500

400

0 10 20 30 40 50

Time after rupture (s)

60

L221-KM258-31

Figure A6. Measured and calculated with RELAP5/MOD2 peak cladding temperatures for Experiment LP-02-6.

A-II
The hydrodynamic model is a two-fluid, six equations model in both oneand three-dimensional components. The partial differential equations are

solved by finite differences. The stability-enhancing two-step (SETS)

numerical algorithm is used in the one dimensional hydrodynamics and permits

this portion of the fluid dynamics to violate the material Courant

condition. This technique permits large time steps and, hence, reduced

running time for slow transients. The three-dimensional vessel option uses

semi-implicit differencing. The finite-difference equations for hydrodynamic

phenomena form a system of coupled, nonlinear equations that are solved by a

Newton iteration procedure. The heat-transfer equations are treated using a

semi-implicit differencing technique. Reactor components in TRAC-PFI/MOD1

consist of accumulators, breaks, fills, cores, pipes, pressurizers, plenums,

steam generators, tees, turbines, valves and vessels with associated

internals.

The calculations of Experiment LP-02-6 presented here were performed at

the Los Alamos National Laboratory and are documented in Reference A8. The

input model used in these analyses was adopted from the TRAC-PD2 input model

used in pre-experiment calculations with code required modifications and

measured initial conditions. Code version 12.0 was used on CDC-7600

computer.

The code calculated the primary system pressure quite well as shown in

Figure A7. The cold leg break flow also agrees well with the data

(Figure A8). However, the break mass flow rate increase calculated after

40 s as result of nitrogen injection and system pressure increase was not

measured. The peak cladding temperature was calculated to contain fewer

cooling effects during the transient as shown in Figure A9. The code was not

able to calculate the early bottom-up quench. The final quench is calculated

to occur later and from lower temperatures than in the experiment. The

inability to calculate the rapid quenching during blowdown is attributed [A81

to limitations in heat transfer correlations in the code. Figure AIO shows

the measured and calculated cladding temperatures in bundle 4 in which the

top down quench was most effective. The comparison indicates that TRAC was

also not able to calculate this phenomenon.

Aý-12,
0

Figure A.

Time (s)

Measured and calculated with

for Experiment LP-02-6.

TRAC-PF1/MOD1 primary system pressure

Annl. I I

700-

6,

ox

0

a.

09

0,

500.

400-

300-

200.

- rRAC

0 FR-BlL-105

10.

O-44

-~u4 -

-20 0 20 40 60o 0

Time (s)

,DO 120

Figure A8. Measured and calculated with TRAC-PFI/MODI cold leg break flow for

Experiment LP-02-6. A-13
0

V

V

E

0

U

'Figure A9.

'V

0

E

0

U

4T 6m

Time (s)

Measured and calculated with TRAC-PF1/MOD1

temperatures for Experiment LP-02-6. peak cladding

40 60

Time (s)

Figure AIO. Measured and calculated with TRAC-PFI/MOD1 cladding temperatures

in peripheral bundles for Experiment LP-02-6.

A-14
In summary, the most advanced thermal-hydraulic systems codes perform in

an acceptable manner in a macroscopic sense. That is, parameters such as

system pressure and break flow are well-calculated. However, in a

microscopic (or localized) sense, the codes do not do as well although, in

general, the trends in the calculations indicate that the phenomena in

question, such as fuel cladding cooling/quench phenomena, are being sensed

however incorrectly in magnitude. These codes are projected to be able to

calculate the significant phenomena in a large break LOCA with improved

models for phenomena such as post-CHF heat transfer and forced convective

cooling.

A-15
APPENDIX A

REFERENCES

Al. S.Guntay,. "A Comparative Analysis on OECD-LOFT Experiment LP-02-6",

OECD LOFT-T-3410, November 1987.

A2. F. Steinhoff, "DRUFAN-02 Interim Program Description Part 1", GRS-A-685,

March 1982.

A3. M. J. Burwell et al., "DRUFAN-Ol/MOD2 Volume III", GRS-A-846, July 1983.

A4. J. D. Burt, S. A. Crowton, "International Standard Problem 13 (LOFT

Experiment L2-5)", EGG-NTAP-6276, April 1983.

AS. V. H. Ransom, et al.,"RELAP5/MOD2 Code Manual, Volume 1: Code

Structure, Systems Models, and Solution Methods", NUREG/CR-4312, August

1985.

A6. J. J. Pena, "Postexperiment Analysis of the OECD LOFT Experiment

LP-02-6", OECD LOFT-T-3408, May 1985.

A7. "TRAC-PFI/MODI: An Advanced Best-Estimate Computer Program for

Pressurized Water Reactor Thermal-Hydraulic Analysis", NUREG/CR-3858,

July 1986.

A8. T. D. Knight, "TRAC Analyses of LOFT LP-02-6", LA-UR-85-3723,

October 1985.

A- 16
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12.89) (Assigned by NRC. Add Vol., Supp., Rev., NRCM 1102, and Addendum Numbers, if any.) 3201,3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/IA-0028

2. TITLE AND SUBTITLE OECD LOFT-T-3900

3. DATE REPORT PUBLISHED

Review of LOFT Large Break Experiments MONTH YEAR

October 1989

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

S. M. Modro, S. N. Aksan, V. T. Berta, A. B. Wahba

7. PERIOD COVERED (Inclusive Daw•se

B. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC, provide Divisin, Office or Reion, U.S Nuclear Regulatory Commission, end mailing address;'it contractor, provid

name and mailing address.) Austrian Research Center Seibersdorf Idaho National Engineering Laboratory

A-2444 Seibersdorf, Austria Idaho Falls, ID 83415

Paul Scherrer Institute

CH-5303 Wurenlingen, Switzerland

Reactor Safety Corporation

Gesellschaft fuer Reaktorsicherheit

Porschungsgelaende, 8046 Garching

The Federal Republic of Germany

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If ARC, type "Same as above'" if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

Office of Nuclear Regulatory Research

U. S. Nuclear Regulatory Commission

Washington, D.C. 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Six non-nuclear and five nuclear large break loss-of-coolant experiments were

performed in the Loss-of-Fluid-Test (LOFT) PWR facility at the Idaho National

Laboratory. These-experiments prcvided a large amount of data necessary for

evaluation and refinement of reactor system computer codes and had major impact

on the understanding of large break loss-of-coolant accidents. An o erview of

these nuclear large break experiments performed under NRC and OECD LOFT programs

is given and the major research results are presented.

12. KEY WORDS/DESCR!PTORS (List words orphrases that w.7l assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Loss-of-Fluid Test (LOFT) Unl imited 1.SECURITY CLASSIFICATION

Organization for Economic Cooperation and Development (OECD) 1ThisER Pe)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89)




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

SPECIAL FOURTH-CLASS RATE

POSTAGE & FEES PAID

USNRC

PERMIT No. G-7