TMI-2: A Textbook in Severe Accident Management...Nuclear Safety Analysis Center (NSAC), 1980, “Analysis of Three-Mile Island – Unit 2 Accident,” NSAC Report No. NSAC-80-1 (NSAC-1

TMI-2: A Textbook in Severe Accident Management

R. E. HenryNISD Professional Development

Workshop

2007 ANS/ENS International Meeting

November 11, 2007

A Severe Accident Primer for Young Engineers

Our Goals for Today

• Learn about, or review, the accident at Three Mile Island Unit 2 (TMI-2).

• Learn about, or review, accident management in general.

• Discuss the accident management insights and lessons derived from the TMI-2 accident.

1. Akers, D. W. and McCardell, 1989, “Core Materials Inventory and Behavior,” Nuclear Technology, Vol. 87, pp. 214-223.

2. Anderson, J. L. and Sieniki, J. J., 1989, “Thermal Behavior of Molten Corium During the Three-Mile Island Unit 2 Core Relocation Event,” Nuclear Technology, Vol. 87, pp. 283-293.

3. Bandini, B. R. and Baratta, A. J., 1989, “Potential for Recriticality of the Relocated Core,” Nuclear Technology, Vol. 87, pp. 926-931.

4. Benedick, W. B. et al., 1984, “Combustion of Hydrogen: Air Mixtures in the VGES Cylindrical Tank,” JUREG/CR-3273, SAND83-1022.

5. Burges, P. S. et al., 1982, “Flammability of Mixed Gases,” U.S. Bureau of Mines Report of Investigations 8709.

6. Chu, T. Y. et al., 1997, “An Assessment of the Effects of Heat Flux Distribution and Penetration on the Creep Rupture of a Reactor Vessel Lower Head,” 12th Proc. of Nuclear Thermal Hydraulics, Nov. 16-20, Albuquerque, NM, ANS Thermal Hydraulics Division.

7. Cronenberg, A. W. and Tolman, E. L., 1989, “Thermal Interaction of Core Melt Debris with Three-Mile Island Unit 2 Vessel Components,” Nuclear Technology, Vol. 87, pp. 273-282.

8. Deitrich, J. R. et al., 1954, “Transient and Steady-State Characteristics of a Boiling Reactor, the BORAX Experiments, 1953,” Argonne National Laboratory Report ANL-5211.

9. Deitrich, J. R. et al., 1955, “Design and Operating Experience of a Prototype Boiling Water Power Reactor,” Paper Presented at the 1955 Geneva Conference on Peaceful Uses of Nuclear Energy.

10. Electric Power Research Institute (EPRI), 1992, “Severe Accident Management Guidance Technical Basis Report,” EPRI Proprietary Report TR-101869, Vols. 1 and 2.

11. Electric Power Research Institute (EPRI), 1994, “Experiments to Address Lower Plenum Response Under Severe Accident Conditions,” EPRI Proprietary Report TR-103389, Vols. 1, 2 and 3.

12. Electric Power Research Institute (EPRI), 2007, “MAAP4 User’s Manual,” EPRI Proprietary Report.

13. Epstein, M. and Fauske, H. K., 1989, “The Three-Mile Island Unit 2 Core Relocation – Heat Transfer and Mechanism,” Nuclear Technology, Vol. 87, pp. 1021-1035.

14. Hagen, S. and Hain, K., 1986, “Out-of-Pile Bundle Experiments on Severe Fuel Damage (CORA Program),” Kernforschungszentrum Karlsruhe, KfK 3677.

Reference Materials for Further Review

15. Henrie, J. O., 1989, “Timing of the Three-Mile Island Unit 2 Core Degradation as Determined by Forensic Engineering,” Nuclear Technology, Vol. 87, pp. 857-864.

16. Henry, R. E. et al., 1991, “Direct Containment Heating Experiments in a Zion-Like Geometry,” 26th National Heat Transfer Conference, AIChE Symposium Series 283, Vol.. 89, p. 86-98, Minneapolis, MN.

17. Henry, R. E. and Fauske, H. K., 1996, “A Different Approach to Fragmentation in Steam Explosions,” International Conference on Nuclear Engineering, Volume 1 - Part A, p p. 309-316.

18. Henry, R. E., 1997, “Is Fuel Fragmentation Needed to Understand the KROTOS Experiments?,” Presented at the 5th International Conference on Nuclear Engineering (ICONE-5), Nice, France, May 26-30.

19. Henry, R. E., 1998, “An Approach to Calculating the Peak Pressures in a Steam Explosion,” Presented at the 6th International Conference on Nuclear Engineering (ICONE-6), San Diego, CA, May 1-014.

20. Hobbins, R. R. et al., 1989, “Molten Material Behavior in the Three Mile Island Unit 2 Accident,” Nuclear Technology, Vol. 87, December, pp. 10051-12.

21. Hofmann, P. et al., 1988, “Dissolution of Solid UO2 by Molten Zircaloy and Its Modeling,” Int. Symp. on Severe Accidents in Nuclear Power Plants, Sorrento, Italy, Paper No. IAEA-SM-296/1.

22. Hohmann, H. et al., 1995, “FCI Experiments in the Aluminum Oxide/Water System,” Nuclear Engineering & Design, Volume 155, pp. 391-404.

23. Huhtiniemi, II. Et al., 1995, “FCI Experiments in the Corium/Water System,” Proceedings of the International Meeting on Nuclear Reactor Thermal-Hydraulics, NURETH-7, Saratoga Springs, New York, Sept. 10-15, NUREG/CR-CP-0142, pp. 1712-1727.

24. Huhtiniemi, I. Et al., 1996, “Test Results and Analysis of Recent KROTOS FCI Experiments,” ANS Proceedings of the 1996 National Heat Transfer Conference, HTC-Volume 9, pp. 27-42.

25. Korth, G. E., 1993, “Peak Accident Temperatures of the TMI-2 Lower Pressure Vessel Head,” Contribution to the TMI-2 Vessel Investigation Project Integration Report, NUREG/CR-6197.

26. Kumar, R. K. et al., 1984, “Intermediate Scale Combustion Studies of Hydrogen-Air-Steam Mixtures,” EPRI Report NP-2955.

27. Lanning, D. D. et al., 1988, “Data Report: Full-Length High-Temperature Experiment 4,” Pacific Northwest Laboratories ReportPNL-6368.



28. Magallon, D. and Hohmann, H., 1995, “High Pressure Corium Melt Quenching Tests in FARO,” Nuclear Engineering and Design, Volume 155, pp. 253-270.

29. Magallon, D. and Hohmann, H., 1995, “Experimental Investigation of 150-kg-scale Corium Melt Jet Quenching In Water,” Proceedings of the 7th Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-7, Saratoga Springs, New York, September 10-15, NUREG/CP-0142, pp. 1688-1711.

30. Magallon, D., Huhtiniemi, I. And Hohmann, H., 1996, “An Overview of the FARO and KROTOS Test Results,” Proceedings of the International Topical Meeting on Probabilistic Safety Assessment PSA’96, Park City, Utah, September 29-October 3, Volume III, pp. 1351-1358.

31. Marshall, B. W., 1986, “Hydrogen:Air:Steam Flammability Limits Combustion Characteristics in the FITS Vessel,” NUREG/CR-3468, SAND 84-0383.

32. Martinson, Z. R. et al., 1986, “Volume 1: PBF Severe Fuel Damage Test 1-1, Test Results Report,” NUREG/CR-4684, EGG-2463.

33. Maile, K. et al., 1990, “Load Carrying Behavior of the Primary System of PWRs for Loads Beyond the Design Limits, Part 2: Creep and Failure Behavior of the Piping Section Under Internal Pressure and High Temperature,” Nuclear Engineering and Design, 199, pp. 131-137.

34. Nuclear Safety Analysis Center (NSAC), 1980, “Supplement to Analysis of Three-Mile Island – Unit 2 Accident,” NSAC Report, NSAC-1 Supplement.

35. Nuclear Safety Analysis Center (NSAC), 1980, “Analysis of Three-Mile Island – Unit 2 Accident,” NSAC Report No. NSAC-80-1 (NSAC-1 Revised).

36. Olsen, C. S. et al., 1989, “Application of Severe Fuel Damage Experiments to Evaluating Three-Mile Island Unit 2 Core Materials Behavior,” Nuclear Technology, Vol. 87, pp. 884-896.

37. Pilch, M. et al., 1994, “The Probability of Containment Failure by Direct Containment Heating in Zion,” NUREG/CR-6075, SAND93-1535, Sandia National Laboratories, Albuquerque, NM.

38. Pilch M. et al., 1995, “The Probability of Containment Failure by Direct Containment Heating in Surry,” NUREG/CR-6109, SAND93-2078, Sandia National Laboratories, Albuquerque, NM.

39. Pilch, M. M. et al., 1998, “Resolution of the Direct Containment Heating Issue for Combustion Engineering Plants and Babcock & Wilcox Plant,” NUREG/CR-6475, SAND97-0667.


40. Porter, L. H. and Austin, W. E., 1989, “Disassembly and Refueling of the Three-Mile Island Unit 2 Reactor Vessel Lower Core Support Assembly,” Nuclear Technology, Vol. 87, pp. 595-608.

41. Ratzel, A. C., 1985, “Data Analyses for Nevada Test Site (NTS) Premixed Combustion Test,” NUREG/CR-4148, SAND85-0135.

42. Rausch, W. N. et al., 1986, “Data Report: Full-Length High-Temperature Experiment 1,” Pacific Northwest Laboratory Report PNL-5691.

43. Strain, R. B. et al., 1989, “Fuel Relocation Mechanisms Based on Microstructures of Debris,” Nuclear Technology, Vol. 87, pp. 187-190.

44. Thinnes, G. L. and Moore, R. L., 1989, “Comparison of Thermal and Mechanical Responses of the Three-Mile Island Unit 2 Reactor Vessel,” Nuclear Technology, Vol. 87, pp. 1036-1049.

45. Thompson, L. B. et al., 1984a, “Large Scale Hydrogen Combustion Experiments,” Presented at the ANS International Conference on Containment Design, Toronto, Canada.

46. Thompson, L. B. et al., 1984b, “EPRI Large Scale Hydrogen Combustion Experiments,” Joint ANS/NRC Conference on Design, Construction and Operation of Nuclear Power Plants, Portland, Oregon.

47. Thompson, R. T. et al., 1988a, “Large-Scale Hydrogen Combustion Experiments, Volume I: Methodology and Results,” EPRI Report NP-3878, Volume I.

48. Thompson, R. T. et al., 1988b, “Large-Scale Hydrogen Combustion Experiments,Volume II: Data Plots,” EPRI Report NP-3878, Volume II.

49. Tuunanen, J. et al., 1988, “Long Term Emergency Cooling Experiments with Aqueous Boric Acid Solution with the REWET II and VEERA Facilities,” NECSAFE ’88 Proceedings, Avignon, France.

50. Wierman, R. W., 1979, “Experimental Study of Hydrogen Jet Ignition and Jet Extinguishment,” HEDL-TME 78-80.

51. Wolf, J. R. et al., 1993, “TMI-2 Vessel Investigation Project Integration Report,” NUREG/CR-6197 TMI V(93)EG10 EGG-2734.

Key References

Nuclear Technology (A Journal of the American Nuclear Society), Volume 87, 1989. October, November and December issues devoted to technical papers describing the accident behavior as well as the clean up and defueling of the facility following the accident.

NSAC-80-1, (NSAC-1), 1980, “Analysis of Three Mile Island – Unit 2 Accident”

Acknowledgments

• Dana Powers for sending me the slides he has used in lectures on the TMI-2 accident.

• Hans Fauske for looking over all of the slides we are using today.

Presentation Outline

• Overview of the TMI-2 design• Overview of the accident progression• Summary of Accident Management Goals• Detailed review of the accident progression

combined with the technical basis for the severe accident phenomena

• Summary of the key Accident Management insights from the TMI-2 accident

What are the Key Accident Management Insights to be Discussed

• Timing of the coolant inventory loss and the core overheating

• Rapidity of the core damage once the fuel cladding oxidation begins

• Available fission barriers• Available debris barriers• Cooling of a compacted core• Ultimate cooling of the debris• Hydrogen Generation and Combustion

Overview of the TMI-2 Design

• What was the TMI-2 accident?• TMI-2 Reactor Coolant System (RCS)• TMI-2 Once-Through Steam Generators

(OTSGs)• TMI-2 containment building• TMI-2 reactor core design• Measurements recorded during the event

TMI-2: What Happened onMarch 28, 1979?

• Due to a lack of makeup/injection, the reactor core was starved of water, i.e. the core was eventually uncovered.

• As a result of decay power the core was overheated sufficiently that a significant oxidation reaction occurred between the Zircaloy fuel cladding and steam in the core.

• The chemical energy release caused the core to overheat faster and eventually melt or liquefy the individual constituents.

• Most of the molten core material eventually relocated outside of the original core boundaries.

• The core material was eventually quenched by water in the Reactor Pressure Vessel (RPV).

Most Important Insight

• “An operator must never be placed in a situation which an engineer has not previously analyzed.”

• Quote from “Design, Training, Operation –The Critical Links: An Operator’s Perspective” by Ed Frederick control room operator for TMI-2.

TMI-2 Design Features

• B&W Pressurized Water Reactor (PWR) with a thermal power of 2,720 MWt

• 2 Once Through Steam Generators (OTSG) with 2 Hot Legs connecting the RPV & SGs.

• 4 Reactor Coolant Pumps (MCPs) & 4 Cold Legs• Steam Generators are located at about the same

elevation as the Reactor Pressure Vessel (RPV).

Isometric of the TMI-2 RCS

Once-Through Steam Generators

TMI-2 Design Features• The pressurizer has a loop seal in the connection, i.e.

the lowest point of the connecting pipe is below the entry into the hot leg piping.

• Safety valves and a Pilot Operated Relief Valve (PORV) were located at the top of the pressurizer. The tailpipes from these valves discharged into the drain tank in the containment.

• To mitigate the high temperatures of the pressurizer, the PORV and SV had loop seals upstream of the valves.

• A block valve was located upstream of the PORV in case of excessive leakage.

• The hot leg piping from the RPV to the OTSG has the shape of a “candy cane” and the top of these pipes are the high points of the Reactor Coolant System (RCS).

TMI-2 Containment• The containment, or reactor building, is a steel

lined, cylindrical concrete structure with a volume of approximately 2 million cubic feet. This building is designed to contain a pressure of at least 55 psig; containments of this type are called large dry containments.

• Typically these containments are leak tight to a value of 0.1 % of the building volume per day at the design basis pressure. This is a hole about 2 mm in diameter.

• As a result of the leak tight design, this is an important fission product barrier.

Cross Section of the TMI-2 ReactorContainment Building (Facing South)

TMI-2 Containment• The containment includes safety systems such as

the containment sprays that are mounted in the top of the building and fan coolers that are located near the operating deck.

• Both of these are designed to depressurize & cool the building by removing steam from the containment atmosphere, i.e. each of these systems is designed to remove the decay power from the containment.

• The sprays initially inject a NaOH solution to buffer the boric acid in the RCS coolant.

TMI-2 Containment

• In addition to the RCS and the OTSGs,the containment houses the core flood tanks, the seal table for the in-core instruments and the pressurizer drain tank.

• The bottom of the containment building serves as the sump for recirculation of the borated water after the injection phase has emptied the Borated Water Storage Tank (BWST) and perhaps the core flood tanks.

TMI-2 Reactor Core• The reactor core was composed of 177 fuel

assemblies• Each fuel assembly had 208 fuel pins.• 16 control rod guide tubes and 1 in-core instrument

position per fuel assembly.• Each fuel pin was 0.430 inch in diameter and 12 feet

in length.• The fuel was enriched uranium dioxide pellets in

Zircaloy tubes/cladding. This is an important fission product barrier as is the RCS boundary.

• These fuel pins were fabricated in open lattice bundles with spacer grids holding the pins in place along with upper and lower end fittings.

Original Core Materials Inventory

(From Akers and McCardell)

Reactor Vessel Components

Lower Head Core Support Assembly

Cross Section of the Reactor Vessel and Internals

Fuel Assembly

TMI-2 Reactor Instrumentation• The reactor core had in-core instrumentation that

entered through the RPV lower head and extended through the entire core region.

• These instrument probes contained Self-Powered Neutron Detectors and a core exit thermocouple with a calibration chamber in the center.

• Each instrument thimble had a water annulus that was sealed at the seal table. This design enables each instrument tube to be retracted during refueling.

TMI-2 In-Core Instrument Design

Typical Instrumentation Nozzle - Vessel Weld for a PWR System

Plant Reactimeter• This is an instrument recording package that

was to be used only during startup. However, this captured much of the system data that has been used to interpret the accident progression.

• These records are in addition to the plant computer and included much of the standard measurements such as:

- the RCS pressure,- the pressurizer water level,- the OTSG pressures,- the OTSG water levels,

and many others.

Primary System Reactimeter Measurement

Plant Computer

• Core outlet temperatures were monitored on the plant computer but were cutoff at 700 F

• This caused considerable confusion in the hours after the accident occurred.

• Numerous other parameters associated with the operation of the plant were also recorded on the computer.

Discussion

Overview of the Accident Progression

• Loss of coolant inventory• Core heatup• Fuel cladding oxidation• Core compaction• Relocation of core materials• Final disposition of the core debris• Hydrogen burn

Loss of Coolant Inventory

• With the initial LOFW transient, the OTSGs dried out quickly and the RCS pressurized.

• Eventually the pressurizer PORV opened and the water in the loop seal likely bent the valve stem causing it to stick open.

• RCS coolant was lost through the PORV over about a 100 min. period.

• As a result of the coolant loss, vibrations in the MCPs caused the operators to shutdown these pumps and the water and steam separated.

Drain Tank Behavior From NSAC-1

Influence of Decay Power

• Decay power is due to the continued decay of fission products in the reactor fuel.

• As long as the undamaged core is covered with water, the fuel will be cooled.

• Once a fuel pin is uncovered locally, that part of the fuel pin will begin to heatup.

• The rate of temperature escalation due to decay power can be easily estimated.

Influence of Decay Power

• Decay Power can be represented by the empirical expression:

• Q = Qo t-0.283

• At t = 10 secs Q/Qo = 0.06• At t = 3600 secs Q/Qo = 0.015• At t=10,000 secs Q/Qo = 0.010

Core Heatup

• Consider the decay power to be 30 MW.• Assume 100,000 kg of core materials.• Assume an average specific heat for the

core materials of 600 J/kg/C.• This would cause a heatup rate of 0.5 C/sec.• In 30 minutes the average core temperature

would increase 900 C.

Fuel Cladding Oxidation• The Zr in the Zircaloy cladding will oxidize in a high

temperature steam environment: hydrogen and energy (heat) are released by this reaction

Zr + 2H2O → ZrO2 + 2H2 + ΔHR

The heat of reaction, ΔHR, is about 6.5 MJ/kg.• At about 1000 C, the rate of chemical energy release

approximately equals the decay power.• The oxidation rate increases with increasing

temperature, which leads to an escalating core heatup rate.

• Therefore, the core damage was generally caused by the cladding oxidation.

Example: Core Heatup Rate Escalation Due to Cladding Oxidation

• Important TestsOut-of-Reactor – CORAIn-Reactor SFD, FLHT, LOFT-FP2 andPHEBUS

CORA-12 and –13 Bundle Arrangement and Radial Nodalization

Temperature History at Lower Part of the Test Bundle for CORA-12

System Pressure and Cladding Temperature Histories

The marked events are:1. ILCL Break Initiated2. LPIS Break Initiated3. ILCL Break Closed4. ILCL Break Opened5. PORV Opened6. ILCL Closed7. PORV Closed8. LPIS Closed9. ECCS Initiated

Temperatures When MaterialsCan Liquefy and Relocate

Stainless steel 2500°F/1640 K.

Zircaloy ~ 3320°F/2100 K.

Uranium dioxide 5100°F/3100 K.

Lower melting point eutectics can be formed, such as

• stainless steel and boron carbide,

• stainless steel and Zircaloy,

• Zircaloy and uranium dioxide.

Temperature Indicators for Nuclear Reactor Core Materials

Core Compaction• As individual materials reach their melting

temperatures, or are liquefied as a eutectic, these liquids will drain downward.

• In general due to the core power shape and flow through the core, melting would begin near the top of the core and progress downward.

• With cooler temperatures lower in the core, these liquids can freeze resulting in coolant channel blockages and compaction of the core region.

Postaccident Damage State Within the TMI-2 Reactor Pressure Vessel(taken from Hobbins et al., 1989)

Cross Section of the Lower Crust, Showing Filling ofFlow Channels Between Fuel Rods and Melt Penetration

Into Cracks and Pellet-to-Pellet Interfaces Within Fuel Stacks

Relocation of Core MaterialsOutside of the Core Boundaries

• Once core compaction has occurred, the rate at which heat can be removed from the debris surface decreases even if it is submerged.

• If the energy generation rate exceeds the energy removal rate, the debris will develop a molten central region.

• In this state, internal circulation within the molten pool will cause the maximum heat flux and thermal attack to be at the top-side of the pool.

• Hence, if not cooled, the thermal attack will radially progress into the core outer structures.

• Once the molten debris penetrates the core baffle plates, there is a more open flow path to move downward into the lower plenum

Core Former Geometry

Elevation Positions of the TMI-2 Core Former Plates

Artist’s Rendition of Baffle Plate Melt-Through Noted fromVideo Inspection of TMI-2 Reactor Vessel After Defueling

Final Debris Disposition

• Over 60 tonnes of the core debris was molten at the time that the core former plates were penetrated.

• At this time the entire core had been submerged for at least 30 minutes.

• The molten materials flowed into the water baffle region.

• Some of the debris froze in the baffle region and about 20 tonnes flowed into the lower plenum.


Estimated Postaccident Core Materials Distribution

(From Akers and McCardell)

Final Debris Disposition• In the lower plenum, some of the debris was

particulated and some formed a continuous layer, sometimes called the reconstituted layer.

• The particulated debris quenched rapidly.• The continuous layer heated the a large

region of the RPV lower head with the hot spot estimated to have reached 1100 C.

• The RPV wall then cooled rapidly.

Hydrogen Burn

• Considerable Hydrogen had been generation as a result of the cladding oxidation.

• This was released into the containment building through the pressurizer PORV and drain tank.

• Eventually the hydrogen accumulated to a sufficient concentration that a global burn occurred in the containment.

Discussion

Summary of Accident Management Goals

• Control the power generation.• Manage (protect) the fission product

barriers.• Manage (protect) the debris barriers.• Cool the debris.

Accident Management

• What is it and what are the goals?• What are the principal actions to be

taken?• What are the basics?• What is the technical basis for each of

the recommended actions?

The Basics

1. There are only two types of conditions that can lead to a core damage event.

• Loss of coolant accident.• Adequate coolant inventory but inadequate heat

removal.2. The primary goal of accident management is to take

appropriate corrective actions to eliminate or counter the accident condition.

3. Depending on the accident state, the possible corrective actions must be evaluated to assure their implementation does not challenge the existing fission product barriers.

Key Events

• Accident initiated (secs. - tens of mins. - hrs.)

• Core is uncovered (tens of secs. - tens of mins.)

• Major core damage (1-2 hrs.)

• RPV failure (hrs. to tens of hrs.)

• Containment failure.MAJOR INSIGHT

• There is time to stop the accident progression andprotect one or more fission product barriers.

Barriers to PossibleFission Product Release

1. Fuel pin cladding.

2. Reactor pressure vessel/reactor coolant system.

3. Containment.

BWRs: Primary containment plus MSIVs.

PWRs: Containment building plus the steam generatortubes.

Barriers to Debris Transport

• Fuel pin cladding: If the cladding is not melted,the core geometry remains intact and the fuel is easily cooled.

• Reactor pressure vessel: If debris can be retained in the RPV, issues related to ex-vessel debris behavior are eliminated.

- Ex-vessel debris coolability.

- High pressure melt ejection.

• Containment: Debris must be cooled in the containment to prevent attack of concrete structures.

Severe Accidents Involve the Interactionof Several Complex Phenomenon

• Hydrogen generation/cladding oxidation.• Hydrogen combustion.• Movement/relocation of core materials.• In-vessel natural circulation.• Creep rupture of RCS pressure boundary.• Coolability of a damaged core.

- In-vessel.- Ex-vessel.

• Thermal attack of the RPV wall/lower head penetrations.• Fission product release, transport and deposition.• Fission product revaporization.• Challenges to containment integrity.

Major Concerns forAccident Management

1. RCS and containment isolation.

2. Cooling of the debris/debris retention (maintaining a “debris barrier”).

3. Core reactivity.

4. Heat removal capability.

5. Hydrogen/carbon monoxide combustion.

6. Consequential damage to any existing fission product barrier.

Cooling a Damaged Core

• Cooling by single phase water flow.

P = 1000 psia (7 MPa)

( )SP D fW Q / h h= −

DExample : Q 30 MW=

6fh h 1.2 x 10 J / kg− = ∼

SPW = 25 kg/sec ~ 420 gpm

Cooling a Damaged Core (Continued)

• Cooling by two-phase (boiling) flow.

P = 1000 psia (7 MPa)

( )TP D gW Q / h h= −

DExample : Q 30 MW=

6gh h 2.7 x 10 J / kg− = ∼

TPW = 11.1 kg/sec ~ 183 gpm

Reactivity Considerations

• 1$ of reactivity = the fraction of neutrons due to delayed reactions ~ 0.0065 (0.65% keff).

• Approximate worth of controls rods ~ 10$.• Available means to offset any possible increase in reactivity.

- Boron injection.- Negative void feedback

• Increased dissolved boron concentration.1 ppm boron concentration = -0.01% keff (-1.5¢/ppm).

• Increased void.~ -0.14% keff/% void (-22¢/% void).

Temperatures When MaterialsCan Liquefy and Relocate

Stainless steel 2500°F/1640 K.

Zircaloy ~ 3320°F/2100 K.

Uranium dioxide 5100°F/3100 K.

Lower melting point eutectics can be formed, such as

• stainless steel and boron carbide,

• stainless steel and Zircaloy,

• Zircaloy and uranium dioxide.

What Are the Symptomsof an Overheated Core?

• Low water level in the RPV.

• Core exit temperatures > Tsat.

• Measured superheat in the hot leg thermocouples.

What Are the Symptomsof Core Damage?

• Core exit temperatures approaching the steel melting temperature (low water level).

• Hydrogen in containment.

• High radiation in containment.

Possible Significant Uncertainties

• How much hydrogen has been created?

• Could the accident state change suddenly?

- Creep rupture of the RCS pressure boundary.

- RPV failure.

• What is the core reactivity?

• How coolable is the debris configuration?

How Could the UncertaintiesBe Accommodated?

• Hydrogen.

- Consider the uncertainties in the containment measurement.

- Consider the hydrogen that could be resident in the reactor coolant system.

• Anticipate possible changes in the accident state.

- Hydrogen could be released from the RCS.

- Debris released to the containment (RPV failure).

- Core material could return to a critical state.

• Consider that the debris may be only marginally coolable or notcoolable.

Hydrogen Generation

• Evaluations of current plants are based on the reaction of 75% of the active cladding.

• Plant specific and accident specific evaluations for the accident progression (IPEs) have shown that the hydrogen produced during the core degradation is less than that obtained by reacting 75% of the active cladding.

For Example:

- MAAP sample problems 25% - 55%.

- NRC (MELCOR & SCDAP/RELAP5) 20% - 60%.

• Accident management considerations must consider the hydrogen that is produced as the accident progresses as well as that which could be generated as a result of recovery actions.

Hydrogen Generation(Continued)

• Current accident management evaluations are based on a total hydrogen generation equal to the reaction of 75% of the active cladding.

- If most of the cladding is reacted during the accident progression, there will be little to react during therecovery.

- If little of the cladding is reacted during the accident progression due to “steam starvation”, there will be substantialamounts of high temperature metal to react during the recoveryphase. This was observed in the LOFT-FP2 and CORA tests.

Hydrogen Combustion

• Hydrogen combustion evaluations for accident management considerations include the hydrogen that could be produced as a result of:- uncovering and overheating the core,- recovery actions,- core-concrete attack, and- HPME.

• Also accident management assessments should consider the influence of steam on combustion of the hydrogen.- ~ 1 atm of steam is sufficient to inert the atmosphere.- Large scale tests at SNL have shown that no burning

occurs in a classically inerted system even under HPMEconditions.

Examples of Consequential Damage

• Steam generator tubes: possible overheating of dry tubes due to forced or natural convection.

• Containment liner attack by molten debris.

What Are the Consequencesof Boron Addition?

• Compacted debris may require boiling to remove the energy generated.

• If boiling occurs, boron may be concentrated in the boiling region.

• Concentration of the boron could jeopardize the cooling.

How Could the Water Injection Ratebe Used to Control Core Reactivity?

• With the strong negative void coefficient of reactivity, the injection rate can be used to control the power level.

• If the core could return to a critical condition, the power level would be determined by water injection rate.

• Example: consider W = 1000 gpm (61 kg/sec).Q = W (hfg).Q ~ 61 (2 x 106) ~ 122 MW.

• This power, if it were to occur, could be relieved by the RCS relief valves.

How Could the RPV beChallenged by Molten Debris?

• For designs with lower head penetrations, the molten debris could melt the penetrations and enter the calibration chamber.

• Molten debris could also challenge the support welds for the penetration(s).

• Debris could also possibly heat the vessel wall sufficient to result in creep (and possibly rupture) of the vessel lower head.

Vessel Failure

• Experiments have focused on the possible failure of lower headpenetrations if molten debris should drain into the lower plenum.- These experiments show that penetrations would not fail as a result

of the material drainage or near term thermal attack.- The observations of the experiments are consistent with the post-

accident examinations of the TMI-2 reactor vessel.

• SNL Lower Head Failure (LHF) experiments show that significant strainis necessary for failure to occur. However, the vessel wall strain may alsostrain the penetration support weld to failure well before the vessel isstrained sufficiently for failure.

• The modes for reactor vessel failure must also consider the potential forexternal cooling. Specifically the influence of material creep must includethe influence of a temperature gradient through the RPV wall.

How Could the DebrisProgression be Stopped?

• Once the core has been damaged, the debris may drain into the lower plenum and thermally attack the RPV wall.

• In-vessel cooling of the debris if the RCS water inventory can be restored and maintained.

• Cooling the RPV on the outside (external cooling) could retain the debris within the RPV.

• The combination of in-vessel and external cooling minimizes the uncertainties to be addressed.

Debris Cooling

• Numerous experiments have demonstrated that cooling rates for molten material poured into water can approach 10 MW/m2 to 30 MW/m2.

• Experience from the TMI-2 RPV examinations show that a reconstituted solid layer of limited coolability accumulated at the bottom of the debris bed and caused elevated RPV wall temperatures.

• Current focus on ex-vessel coolability experiments is on the gas evolution rate from core-concrete attack and the influence that this would have on debris cooling.

• The current technical basis suggests that accident management evaluations with respect to debris cooling needs to consider both ends of the spectrum, i.e., the possibility of significant steaming rates if the debris quenches rapidly as well as the implications if the debris is not coolable.

Conclusions

• The phenomena associated with a severe accident are complex.

• Measurements from the RCS and containment would provide only a general representation of the phenomena.

• However, the objectives for accident management are clear.

• Given the clear objectives, the actions to be taken to recover from a severe accident can be structured essentially independent of the phenomenological uncertainties.

Discussion

Detailed Review of the TMI-2 Accident Progression & the Accident

Management Technical Basis

Detailed Review of the AccidentProgression & the Technical Basisfor Important Physical Processes

• Coolant inventory loss.• Core heatup.• Fuel cladding oxidation.• Thermal attack of the instrument penetrations and control rod fingers.• Liquefication of the uranium dioxide fuel.• Downward relocation of the molten debris.• Reflooding of an overheated core.• Energy transfer from a submerged compacted core.• Relocation of molten debris out of the original core boundaries.• Rapid steam generation.• Debris configuration in the lower plenum.• Thermal attack of the RPV lower head.• Ultimate cooling of the core debris.• Release of hydrogen to the containment.• Hydrogen burn

Coolant Inventory Loss• After the PORV stuck open, the pressurizer filled

with water and the RCS continued to loose coolant through the valve to the drain tank.

• Since the pressurizer was full, the operator training at the time steered the operators in the direction of reducing the RCS coolant inventory, i.e. shutting off the HPI injection because there was too much water in the pressurizer/RCS.

• Also, as the water decreased, the ex-vessel source range monitor indicated an increasing count rate which was interpreted as a return to power.

Short-Term Source Range Trace for TMI-2 (3/28/79)

Coolant Inventory Loss

• Faced with a possible return to power, the operators asked for a chemical analysis of the boron concentration in the coolant; expecting a value approaching 1000 ppm.

• This took at least 30 minutes and the analysis showed 400 ppm, i.e. confirmation that the core could be returning to power.

• This caused a diversion of the operators attention away from the loss of coolant.

• Note that the continued loss of water also lead to a continued increase in the source range detector, i.e. a sustained indication of a return to power.


• Hind-sight is 20-20• Both RCS flow meters exhibited a decreasing flow rate

while the MCPs were delivering a constant volumetric flow.

• This decreasing pressure difference was due to a decreasing coolant density.

• The pressurization and subsequent depressurization of the drain tank indicated a loss of coolant.

• An increasing source range monitor is indicative of a decreasing coolant density.

• Interpreting these trends was not, and should, not be part of the operator’s training/duties in the control room.


• As the primary side void fraction increased, the MCPs began to vibrate with increasing amplitude.

• Eventually the vibration level alarmed and the MCPs were turned off.

• Without the MCPs the steam & water phases separated with much of the water inventory trapped in the OTSGs.

Water Level on the Secondary Side of the A Loop Steam Generator

Water Level on the Secondary Side of the B Loop Steam Generator

Comparison Between the Calculated Flow in the B Loop and the Data

Comparison Between the Calculated Flow in the A Loop and the Data

Short-Term Source Range Trace for TMI-2 (3/28/79)

Calculated and Measured Primary SystemPressure During the First Five Hours of the Accident


• “An operator must never be placed in a situation which an engineer has not previously analyzed.”

• Quote from “Design, Training, Operation –The Critical Links: An Operator’s Perspective” by Ed Frederick control room operator for TMI-2.

Additional Quote from Ed Frederick

• I am often asked the question,”If you were back in the control room faced with the same circumstance – and allowed to use only one of the innovations installed or created since the accident – which change or improvement would you choose and would that prevent the incident?”

• …, my response is that if I must choose only one, then the training on the symptom based emergency procedures is all that I would need.


• AM Lesson: In the midst of an accident, it is easy to be side-tracked by some of the accident signatures. These may even have confirming indications.

• In the face of a developing accident, stick with the basics, i.e. the symptoms. Is there any indication that the core is, or may be uncovered? For example superheated steam temperatures in the hot leg or high core outlet temperatures.

Core Heatup

• With the loss of coolant, shutdown of the MCPs and separation of the phases, the core was barely covered by water.

• Boiling in the core caused some level swell to extend the core coverage interval but this was not significant since the RCS was at an elevated pressure.

• With continued boil-off due to decay power, the core eventually uncovered.

Cross-Section of THTF Test Section(Taken from Anklam, et al., 1982)

Cross-Section of FLHT-4 Fuel Rod Bundle and Shroud:(a) Full Cross Section; (b) Detail of the FLHT-4 Shroud (dimensions in inches)

(Taken from Lanning, et al., 1983)

Schematic of the IIST Facility(Taken from Lee, et al., 1989)

Schematic of the IIST Core Heater Rods Layout

Comparison of the VFVOL Drift Flux ModelWith the THTF, FLHT and IIST Experiments

Core Heatup• Uncovering of the core begins at the top and

progresses downward with the decreasing boiled-up level.

• As individual portions of the fuel pins become uncovered, they begin to heatup due to the internal heat generation from decay power.

• The rate of heatup is determined by the core power distribution (chopped cosine) and the steam flow rate from the covered portion of the core.

• The average core heatup rate due to decay power was approximately 5 C/sec

Fuel Cladding Oxidation• As the boil-off of the water in the core continued,

the uncovered region continued to heatup with the highest cladding/fuel temperatures being at about the ¾ core height location.

• Increasing temperatures caused the Zircaloy oxidation rate to increase which was accompanied by an increased release rate of chemical energy.

• At about 1000 C, the oxidation energy release rate equaled the decay power. From this point on, the core was in a thermal-runaway state. During this interval the Zircaloy reaction was limited by the rate of steam generated in the covered part of the core which decreased as the water level decreased.


Comparison of the Calculated and Measured Pressurizer Water Level

Comparison of the Measured andCalculated A Loop Hot Leg Temperatures

Comparison the the Measured andCalculated B Loop Hot Leg Temperatures

System Pressure and Cladding Temperature Histories

The marked events are:1. ILCL Break Initiated2. LPIS Break Initiated3. ILCL Break Closed4. ILCL Break Opened5. PORV Opened6. ILCL Closed7. PORV Closed8. LPIS Closed9. ECCS Initiated

Temperature History at Lower Part of the Test Bundle for CORA-12

Calculated Hydrogen Generation History

Thermal Attack of the InstrumentPenetrations and Control Rod Fingers

• With cladding oxidation the heatup rate increased and was only limited by the steam generation rate (steam starvation) and the eventual change in the core geometry.

• Melting of the core materials began with the lowest melting constituent which was the stainless steel control rod fingers and Inconel stainless steel in-core instrument probes.

• At the time of melting, the control fingers also contained molten silver, indium and cadmium. Failure of the stainless steel cladding initiated downward relocation and also generated an uncertainty with respect to the location of the absorber material.

• Melting of the instrument probes opened numerous small but important paths from the core region to the containment atmosphere through the calibration channels in the center of the probes

Cross Section of the Reactor Vessel and Internals

Fuel Assembly

TMI-2 In-Core Instrument Design

Temperature Indicators for Nuclear Reactor Core Materials

Thermal Attack of the Instrument Penetrations and the Control Rod Fingers

• With this flow path opened, airborne radioactive fission products, along with hydrogen and perhaps steam could escape to the containment atmosphere.

• At a slightly higher temperature the Zircaloy cladding melted and with this the first two fission product barriers where breached.

• In the TMI-2 accident, the block valve had been closed before the fuel pins, control fingers and instrument probes reach temperatures sufficient to fail the structures, i.e. the RCS was completely closed.

• While the RCS was “buttoned-up” the containment radiation monitors showed rapid increases thereby illustrating that this path was opened.

Typical Instrumentation Nozzle - Vessel Weld for a PWR System

Test Element Configuration

Release of FissionProducts to the Containment

• From the accident chronology in NSAC-1, the PORV block valve was closed at 06:22:37 and the reactor coolant system pressure began to increase.

• This indicates that the RCS is essentially a closed system.

Three Mile Island – Unit 2 Reactor BuildingLocations of Radiation Monitors Along Section B-B

Three Mile Island – Unit 2 Area Personnel Monitors Response

Fission ProductsReleased to the Containment

• Again from the NSAC-1 chronology, several radiation monitors simultaneously began to increase at 06:39:00

• At the same time, the Letdown Cooler A rad monitor pegged off scale high.

• At 06:40:00 two boron analyses indicating a boron concentration of 400 ppm. Emergency boration was started.

Reason for the DecreasedReading for Boron Concentration

• In the TMI-2 RCS design, the coolant sampling line is in the letdown line which is in the pump suction piping for the A loop.

• After the phases (steam & water) separated, only the steam was transported from the core to the OTSGs, i.e. it was distilled and the boron was left behind in the core region.

• Conclusion: the RCS configuration caused the reduced reading: the core had a higher boron concentration.

Fission Products to the Containment

• Release of radioactive fission products of this magnitude into the containment gas space demonstrates that the first two fission product barriers were breeched.

• It is important to note that this occurred when the RCS was a closed system.

• AM lesson: the in-core instruments provide a release path from the RCS that opens at nearly the same time that the cladding begins to fail.

Downward Relocation of the Core Debris• As molten debris is formed, this drained

downward into cooler regions of the core where it froze. This change in the core geometry removed material from the top regions of the core and collected it in the coolant channels below. This began the core compaction process.

• This process of melting, downward relocation and freezing continued with the compacted region growing and moving lower in the core.

• The major impact of the debris movement is that the available surface area for cooling the debris and removing decay is dramatically decreased. We also note that debris relocation and compaction also tends to decrease the cladding oxidation rate.

Cross Section of the Lower Crust, Showing Filling ofFlow Channels Between Fuel Rods and Melt Penetration

Into Cracks and Pellet-to-Pellet Interfaces Within Fuel Stacks(From Hobbins et al.)

Neutron Radiographs of SFD Test BundlesDimensions are in Centimeters

Reflooding of an Overheated Core

• At 174 mins. into the accident, the operators successfully started the 2B MCP and this pushed an estimated 700 cubic feet of water into the core.

• At this time the core had undergone considerable cladding oxidation and relocation of core materials.

• Reflooding caused rapid steam generation, most likely rapid hydrogen generation and a rapid increase of the RCS pressure as well as a surge of water into the pressurizer.

Hypothesized Core Damage Configuration at 173 Minutes

Hypothesized Core Damage Configuration (175 to 180 Minutes)


Comparison of the Calculated and Measured Pressurizer Water Level

Calculated Hydrogen Generation History

Coolant Inventory in the RPV

• The 2B pump start at 174 mins. and added considerable water for an interval of 3-9 seconds.

• Even though the pump continued to run for many minutes, the B loop flow meter recorded no flow, i.e. there was insufficient water in the suction piping .

• At 200 mins. makeup pump 1C was started and the core was submerged shortly thereafter and essentially remained submerged during the remainder of the accident.

Energy Transfer from a Compacted Core

• Energy transfer from a compacted core is more limited for two reasons

• (1) the surface area is dramatically reduced.• (2) the conduction length in the debris is

longer, i.e. a length of about 0.1m means that the internal region is molten.

Azimuthal Locations of Main Flow Holes in the PlenumCylinder in Relation to Plenum Assembly Damage Pattern

Configuration for Considering In-Vessel SteamGeneration During a Reflood of An Overheated Core

Comparison of the Calculated Pressure Rise with that of TMI-2

Relocation of Molten DebrisOutside of the Core Boundaries

(While it is Submerged)• Due to internal circulation of the molten debris, the peak

heat flux from the molten core is at the top outer radius of the compacted volume.

• As a result, the molten region would tend to propagate (spread) radially. This continued until the debris reached the core baffle plates which also have 2 inch dia. holes for the water to exchange between the water baffle region and the reactor core.

• Once the debris breaks through one of these holes, there is essentially a open volume (albeit submerged) in the water baffle region for the molten materials to spread.

• The melt spread around the outside of the core and also drained downward through the core former plate holes.

• The continued flow of molten debris through the breach in the core baffle plate caused ablation (melting attack) of the steel plate and enlarged the size of the breach.

Artist’s Rendition of Baffle Plate Melt-Through Noted fromVideo Inspection of TMI-2 Reactor Vessel After Defueling

Ablation Calculation

( ) ( )

( )( )( )

s F F,m s s s s,m s s

1/ 2

s s sF s

F F F1/ 2

s s s

F F F

F F

F F,m

s s s,m s s

drhA T T A c T Tdt

k cT Tk c

Gk c1k c

c Uh f2

h T Tdr Bdt c T T

⎡ ⎤− = ρ − + γ⎣ ⎦

⎡ ⎤ρ+ ⎢ ⎥ρ⎣ ⎦=⎡ ⎤ρ

+ ⎢ ⎥ρ⎣ ⎦

ρ=

−= =ρ − + γ

Crust Formation

Flow of Core Debris Through a Failure Site

Mass Discharged as a Function of Time for Gravity Driven Flow


Baffle Plate Damage of the East Side of the TMI-2 CSA

Debris Configuration in the Lower Plenum

• Drainage of the molten debris into the lower plenum resulted in particulation of some of the debris and accumulation of the remainder on the RPV lower head.

• Debris particulation increased the debris coolability and also resulted in rapid steam generation.

• Accumulation of the debris on the lower head caused a heatup of the RPV lower head.

Debris Configuration in the Lower Plenum

The large scale FARO tests provide the most complete technical basis for this aspect of the accident progression.

FARO Test Arrangement(taken from Magallon and Hohmann, 1995b)

Summary of Experimental Conditions

Comparison of the MAAP Calculated Pressure Increaseand the Measured FARO Vessel Pressure for Test L-11

Potential for Steam ExplosionsDuring the TMI-2 Melt Relocation

• The accumulated technical basis shows that an elevated pressure (p > 1-2 MPa) will prevent a thermal explosion. These is 5 to 10% of the thermodynamic critical pressure.

• The large scale FARO tests are part of this technical basis but are supported by a large number of other smaller scale tests with real and simulant materials.

• At the time of the TMI-2 melt relocation, the RCS pressure was of the order of 10 MPa. Therefore, the system pressure alone was sufficient to suppress any potential for an explosion.

Experiments Demonstrating a High Pressure Cut-Off

Laboratory Materials Used

Expl. PressuresMeasured

(MPa/psia)

System PressureReq. to EliminateExpl. (MPa/psia)

ReducedPressure

RatioArgonne (Henry & Fauske, 1979) Freon-22 and Mineral Oil (Fig. G-11) 2.5/363 0.2/29 0.04Argonne* (Henry & McUmber, 1978) Freon-22 and Mineral Oil (Fig. G-10) 2.0/290 0.5/73 0.10Ispra (Hohmann, 1979) Water and Molten Sodium Chloride 6.0/870 1.0/145 0.05Ispra* (Hohmann, 1982) Water and Molten Sodium Chloride 4.0/580 3.1/450 0.15Sandia* (Nelson and Buxton, 1978) Water and Molten Corium 1.5/218 0.75/109 0.04Winfrith (Bird & Millington, 1979) Water and Molten Uranium Dioxide 3.0/435 0.9/131 0.05Sandia (Corradini, 1981) Water and Molten Iron Thermite 3.5/508 > 1.01/ > 147 > 0.05(Avedisian, 1982) η-Octane and Glycerin Not Measured > 0.101/14.7

< 0.687/100> 0.04< 0.27

(Cho, 1991) Water and η-Paraffins Not Measured 0.28/40 0.014(Frost and Sturdevant, 1986) Ethyl Ether and Glycerol Not Measured 0.3/44 0.085JAERI (Yamano, et al. , 1993) Molten Iron Thermite and Water 10.0/1500 > 0.1/15

< 1.6/230> 0.005< 0.08

* Externally triggered systems.

Cutoff Pressure for Water ~ 1 MPa/145 psia


Illustration of (a) TMI-2 Lower Plenum Structures and(b) Assumed Melt Penetration/Jet Impingement Geometry

How Could the RPV beChallenged by Molten Debris?

• For design with lower head penetrations, the molten debris could melt the penetration and enter the calibration chamber.

• Molten debris could also challenge the support welds for the penetration(s).

• Debris could also possibly heat the vessel wall sufficient to result in creep (and possibly rupture) of the vessel lower head.

Typical In-core Nozzle with Seal and Retaining Weld

Grid Map of TMI-2 Core Showing Locations of Nozzles Examined

Elevation View of Nozzle Segment M-9

Schematic of the Source of TMI-2Metallographic and Mechanical Property Samples

Thermal Contour Map of Peak Temperatures Constructed as Best Estimate Based on Results of the Metallographic Examinations of the Boat Samples

TMI-2 Lower Head Cross Section of Hard Debris, Row 7

Nozzle Damage Profile

Conclusions(from Korth, 1993)

From the examinations of the 15 TMI-2 lower head samples,the following conclusions were formulated:1. All elliptical hot spot approximately 1 x 0.8 m on the inside

surface of the lower pressure vessel head was heated to –800 to 1100°C for approximately 30 minutes due to the relocated fuel debris.

2. The remainder of the lower head remained below 727°C, but some areas may have been close to this temperature.

3. The temperature gradient through the thickness of the vessel wall was approximately 2 to 4°C/mm.

4. The thermal excursion of the lower head was “quenched”, i.e., cooled at a rate in the range of 10-100°C/minute.

LHF-3

Failure for LHF-3

LHF-4

Failure for LHF-4

Additional TechnicalBasis for Material Creep

• Maile et al performed a creep rupture on a full size hot leg for the KWU design.

• This hot leg was approximately 0.7m in diameter, a thickness of 47mm and pressurized to 16.3 MPa.

• External heating was used to raise the metal to 700 C with a measure temperature difference through the wall.

• The carbon steel properties are almost identical to the TMI-2 RPV steel.

Material Characterization for the TMI-2 Vessel(as reported by Wolf, et al.)

Measured Inner and Outer Pipe Surface TemperaturesDuring the Material Creep (taken from Maile et al.)

Current Dynamic Benchmark Results

Calculated Inner Vessel Wall Temperaturesas a Function of Initial Gap Resistance

In-Vessel Cooling of the Debris

• Metallographic studies show that the RPV wall cooled rapidly once it began.

• There are two mechanisms that could have resulted in this rapid: (1) coolant penetrating through cracks in the solidified debris layer or (2) a slight creep of the RPV wall that developed a small gap between the debris and the vessel wall.

• In either case it is important to understand the insight that RPV wall was cooled rapidly after it had reached elevated temperatures, i.e. in-vessel cooling is an effective mechanism.

Hydrogen Release to the Containment

Release of Hydrogen to the Containment

• Hydrogen captured inside the RCS was eventually to the containment through periodic openings of the block valve.

• This occurred over a 10 hour period.• From an AM viewpoint, once core damage

is suspected, it must be assumed that the hydrogen will eventually escape to, and accumulate in, the containment.

Hydrogen Burn

• Hydrogen released to the containment over a 10 hour period eventually reached a combustible condition and experienced a global burn.

• The potential for a global burn to occur is a function of the steam concentration as well as that of the hydrogen.

• Obviously an ignition source is needed.

Effect of H2 Ignition on Measured Steam Pressures

Reactor Building Temperatures During Hydrogen Ignition

Pressure Rises Resulting from Ignition ofHydrogen-Air in 12-ft-diameter Sphere at 18°C

Influence of Steam on Combustion

• Increase steam partial pressure in the containment atmosphere is a consequence of most accident conditions.

• Steam can slow the combustion process and even inert the Containment Atmosphere (prevent combustion).

Normalized peak pressure (Pmax/Po) for hydrogen air:diluent mixtures, comparing CO2 and steam (AICC = adiabatic isochoric complete combustion, Rh - relative humidity). From NUREG/CR-3273.

Hydrogen:air:steam flammability data/fans on.From NUREG/CR-3468.

Hydrogen:air:steam flammability data/fans off.From NUREG/CR-3468.


Material Characterization for the TMI-2 Vessel(as reported by Wolf, et al.)

External RPV Cooling

• The positive role of external RPV cooling under severe accident conditions has received a great deal of attention.

• Of particular interest are large, molten pools which are cooled at the vessel outer surface. The energy transfer rate from the molten pool to the RPV is limited by natural convection in the pool (high Rayleigh numbers).

• The spectrum of experiments at high Rayleigh numbers have demonstrated the applicability of the Mayinger correlations for natural convection of molten material.

• Several experiments, at different scales, have shown the substantial capabilities to remove heat downward from a reactor vessel.

Water Accumulation on the ContainmentFloor After RWST Injection

Possible Debris Accumulation and HeaterTransfer in the RPV Lower Plenum

Relevance of External RPVCooling to the TMI-2 Accident

• While the water lost from the RCS was generally accumulating in the containment, at the time of melt relocation to the lower plenum, there was insufficient water in the reactor cavity to contact the RPV.

• Therefore, external RPV cooling played no role in the stabilization (cooling) of the debris in the lower plenum.

Discussion

Accident Management Insights

Key Events

• Accident initiated (secs. - tens of mins. - hrs.)

• Core is uncovered (tens of secs. - tens of mins.)

• Major core damage (1-2 hrs.)

• RPV failure (hrs. to tens of hrs.)

• Containment failure.MAJOR INSIGHT

• There is time to stop the accident progression andprotect one or more fission product barriers.

Accident Management Issues Demonstrated bythe TMI-2 Accident (Particularly In-Vessel Cooling

and Temperature Distribution in the Vessel)

1. Substantial core damage can occur in a few tens of minutes after the core has been more than half uncovered.

2. Considerable hydrogen can evolve from a severe accident and result in combustion in the containment.

3. Melting of core materials and downward relocation can result in a relatively tightly packed geometry with limited coolability.

4. Core material relocation, including drainage into the lower plenum, can occur even though the material is completely submerged in water.

5. Drainage of material into the lower plenum can cause substantial overheating of the RPV wall.

6. The debris which overheats the RPV wall may experience a relatively rapid cooling after substantial wall overheating has occurred.

7. The upper plenum temperatures varied from 1800ºF (982ºC) immediately above the core to 800ºF (427ºC) at the dome plate. These temperatures are meaningful to address thermal/mechanical challenges to the RCS integrity.

Summary of the Key Accident Management Insights from theTMI-2 Accident

• In the face of a developing accident, stick with the basics, i.e. is there any indication that the core is, or may be uncovered?

• Is the water injection more than sufficient to remove the decay power?

• Is the RCS heat removal sufficient to extract the decay power?

• A submerged condition is not necessarily sufficient to ensure cooling of the debris, i.e. the debris can relocate in to the lower plenum. Do everything possible to protect the RPV lower head.


• The first materials to melt and move in the core are likely the control materials.

• The fission product barriers of the fuel pin cladding and the RCS pressure boundary are likely to be challenged at about the same time even when the RCS is a closed system.

• When examining the hydrogen concentration in the containment, remember that much of the gas could still reside inside of the RCS.


• The accident was terminated by adding water to the reactor core region.

• Water in the lower plenum protected the RPV wall.

• The containment is a very affective fission product barrier even in the presence of a global hydrogen burn.

AppendixAccident Chronology

Documents

TMI-2: A Textbook in Severe Accident Management...Nuclear Safety Analysis Center (NSAC), 1980, “Analysis of Three-Mile Island – Unit 2 Accident,” NSAC Report No. NSAC-80-1 (NSAC-1