ML20212J775

Pilot Program:Nrc Severe Reactor Accident Incident Response Training Manual.Severe Reactor Accident Overview

ML20212J775
Person / Time
Issue date:	02/28/1987
From:	Giitter J, Hively L, Martin J, Mckenna T, Chris Miller, Sharpe R, Watkins R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
To:
References
NUREG-1210, NUREG-1210-V02, NUREG-1210-V2, NUDOCS 8703090131
Download: ML20212J775 (132)

Text

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. l NUREG-1210 Vol. 2 Pilot Program: NRC Severe Reactor Accident Incident Response

, Training Manual Severe Reactor Accident Overview 4

U.S. Nuclear Regulatory i Commission Office of Inspection and Enforcement

! T. J. McKenna, J. A. Martin, Jr., C. W. Miller, L. M. Hively, l R. W. Sharpe, J. G. Glitter, R. M. Watkins p* "%,,

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J NOTICE l i

Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

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NUREG-1210 Vol. 2 Pilot Program: NRC Severe Reactor Accident Incident Response Training Manual Ssvere Reactor Accident Overview Minuscript Completed: October 1986 Date Published: February 1987 T. J. McKenna, J. A. Martin, Jr., C. W. Miller *, L. M. Hively*,

R. W. Sharpe*, J. G. Giitter, R. M. Watkins*

' Oak Ridge National Laboratory

> Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Wcchington, DC 20555

\,....../

FOREWORD J

Over the past few years the Office of Inspection and Enforcement (IE), Division

-of Emergency Preparedness and Engineering Response, has undertaken a program to upgrade the NRC capabilities to respond to severe reactor accidents. As part of this effort, basic training sessions have been presented by IE staff to all response personnel (Headquarters and regions). Through the process of providing

'this training a standard student text has evolved.

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel.

This set of manuals is not licensing guidance. Rather, it is designed to pre-sent to NRC personnel the best understanding of response planning for a serious reactor accident.

These draft manuals are intended to change over time as NRC staff continues to gain experience. Suggestions are requested and should be sent to the Incident Response Branch.

M c -

, Edward . Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement 4

4

PREFACE i

Severe Reactor Accident Overview is the second in a series of l volumes that collectively summarize the U.S. Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident as se s sment. Other volumes in the series are:

4 Volume 1 -- Overview and S=--arv of Ma ior Point s

• Volume 3 -- Resnonse of Licensee and State and Local Of ficials

• Volume 4 -- Public Protective Actions--Predetermined Criteria 2

and Initial Actions e Volume 5 -- U.S. Nuclear Reaulatory Commission Resnonne Each volume serves, respectively, as the text for a course of instruction in a series of courses. The voinnes have been organized into these training modules to accommodate the scheduling and duty needs of participating NRC staff. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are 4

called out in the text.

l 111

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CONIENTS lAAR 1

PREFAG. ........ ........................ 111 LIST OF FIGURES... ... ........................ vil j l

LIST OF TABLES . ........................ . . . . lx ,

i LIST OF ACRONYMS AND INITIALISMS FOR VOLUMES 1-5 . . . . . . . . . . . . zi

1. SEVERE ACCIDENT OVERVIEW . . . . . . . . . . . . . . . . . . . . . . 1 1.1 OBJECTIVES. ............... . . . . . . . . . . . 1 1.2 TEE PROBLEM . . ........................ 2 1.2.1 Radioactive Material Required for Health Effects . . . . 3  ;

1.2.2 Plant Damage Required for Protective Action '

Guides to Be Exceeded. . . . . . . . . . . . . . . . . . 8 1.2.3 Plant Conditions Required for a Major Release. . . . . . 21 1.3 1HE CONTROL ROOM ENVIRONMENT DURING AN ENERGENCf. . . . . . . . 28 1.4 PREDICTING OFF-SITE ODNSEQUENGS. . . . . . . . . . . . . . . . 29 1.4.1 Protective Action Initiation Based on Plant Conditions . ...................... 30 -

1.4.2 Protective Action Initiation Based on Dose Proj ections. ...................... 35 l

1.4.2.1 Predicting the Release from the Plant . . . . . 35 '

1.4.2.2 Transport and Dose. . . . . . . . . . . . . . . 38 1.4.2.3 Estimates of Overall Uncertainties in Severe Accident Off-Site Dose Projections . . . 42 1.4.3 A Bad Example: Early Protective Action Decisions i During the Three Mile Island--Unit 2 Accident . . . . . 45 1.4.4 Role of Off-Site Dose Projection in Severe Accidents . . 48 1.5 ROLE OF FIELD MONITORING. . . . . . . . . . . . . . . . . . . . 50:

1.6 MAJOR POINTS. . ...... . . . . . . . . . . . . . . . . . . 51

, APPENDIX A. SLIDES RELATING TO VOLUME 2 0F THE SEVERE REACTOR ACCIDENT INCIDENT RESPONSE TRAINING MANUAL , . . . . . . . 55 ,

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1.1 Barriers to release of radioactive material 14

,for a pressurized water reactor . .. ..... ......

1.2 -Possible core damage states as core temperature

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.; z J 1?.'3 - ' Injected flow required to remove decay heat from m

the reactor vessel of a 3300-MW(t) plant by boillag S~ '

water and boiling water reactor systems available to provide the flow . ................... 18 1.4 Simplified diagram of typical boiling water reactor vessel water injection sources. . . .. . ... .. . ... 20

"' Major sequences in a severe light water reactor 1.5

• accident scenario . .................... 24 i ; /,

Event tree for severe accident consequences . .. . ... . 25

.1. 6 1.7 Examples of emergency action levels for a General Emergency . .................... 32 1.8 One-hour surf ace doses predicted by (a) Gaussian plume model, (b) puf f-traj ectory model, (c) complex 1

numerical model, and (d) doses actually observed. . ... . 40 1.9 Relationship between an actual plume and model projections . ......... .. . .. . . . .. . . . . 43 1.10 Hourly wind vector at Three Mile Island on M a rc h 2 8, 197 9. . . . . . . . . . . . . . . . . . . . . . . 46 i

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LIFT OF TABLES Table Eggt 1.1 Sammary of volatility of radioactive material in a large [3300-MW(t)] light water reactor core. . .. . . . . . . 5 1.2 Typical inventories of noble gases and lodine in reactor systems . . .. ... . ... . . . . . . . . . . . 6 1.3 Environmental Protection Agency (EPA) and Food and Drug Administration (FDA) Protective Action Guides

.~.. . . . . . . . . . . . . . . 9 (PAGs) . . . . . ......

1.4 Atmospheric release (C1) necessary under poor meteorological conditions to resnit in protective action guide levels at 1 mile. . . .. . . . . . . . . . . . . 10

. . . . . 12 1.5 Typ ic a l re l e a s e s . . . . . . . . . . . . . . . . . .

1.6 Ezanples of instrumentation and information available for determination of fuel (core) status. . . . . . . . . . . . 22 i

1.7 Typical containment design parameters. . . . . . . . . . . . . 27 1.8 Typical calculated containment radiation estimates. . . . .. ... . . . . . . . . . . . . . . . . . . 34 1.9 Estimated range of uncertainty between early projected dose and actual off-site dose for a severe accident (core melt). . . . .. . . . . . . . . . . . . 44 ix

LIST OF ACRONYMS AND INITIALISMS FOR VOLUMES 1-5 I

ALARA As low as reasonably achievable j AMS Aerial Weasurements System (s)

ARAC Atmospheric Release Advise y Capability ASC Administrative Support Coordinator AST Administrative Support Team BT Base Team (NRC Regional Office)

BWR Boiling Water Reactor CDPA Civil Defense Preparedness Agency CFA Cognizant Federal Agency CFR Code of Federal Regulations CL Congressional Liaison CRD Control rod drive CRDHS Control rod drive hydraulic system CSF Critical Safety Function DBA Design Basis Accident DOC Department of Commerce, U.S.

DOD Department of Defense, U.S.

DOE Department of Energy, U.S.

DOI Department of Interior. U.S.

DUT Department of Transportation, U.S.

DSO Director of Site Operations EAL Emergency Action Level ECCS Emergency Core Coollag System EDO Executive Director of Operations ENS Emergency Notification System EO Emergency Officer EOF Emergency Operations Facility EOP Emargency Operating Procedure EPA Environmental Protection Agency, U.S.

EPRI Electrical Power Research Institute EPZ Emergency Planning Zone ERC Emergency Response Coordinator ERM Emergency Response Manager ERO Emergency Response Organization EKI Emergency Response Team (FEMA organization)

ESF Engineered Safety Feature EST Emergency Support Team (FEMA organization)

ET NRC Executive Team ETA Estimated time of arrival FBI Federal Bureau of Investigation FDA Food and Drug Administration, U.S.

FEMA Federal Emergency Management Agency FRC Federal Response Center FRERP Federal Radiological Emergency Response Plan FRMAC Federal Radiological Monitoring and Assessment Center FRMAP Federal Radiological Monitoring and Assessment Plan FTS Federal Telephone System GLC Government Liaison Coordinator GLM Government Liaison Manager xi

GLO Government Liaison Officer-GLT Government Liaison Team HHS Bealth and Human Services, U.S. Department of HOO NRC Headquarters Operations Officer HPCI High pressure coolant injection HPCS High pressure core spray HPN Health Physics Network BQ NRC Headquarters HUD Bousing and Urban Development, U.S. Department of IE NRC Office of Inspection and Enforcement ICRP International Commission on Radiological Protection IDAS Interactive Dose Assessment System IRB Incidence Response Branch IRC Incident Response Center IRDAN' Interactive Rapid Dose Assessment System JIC Joint Information Center LC Liaison Coordinator LN0 Liaison Officer LOCA Loss of Coolant Accident LPCI Low-pressure coolant inj ection LPCS Low pressure core spray LT Liaison Team LWR Light Water Reactor NCS National Communication System NNSS NRC Office of Nuclear Material Safety and Safeguards NOAA National Oceanic and Atmospheric Administration NRC Nuclear Regulatory Commission, U.S.

NRR NRC Office of Nuclear Reactor Regulation NWS National Weather Service OC Operations Center OSC Operations Support Center (site)

PA Public Affairs PAC - Public Affairs Coordinator PAG Protective Action Gaides PAN Public Affairs Manager PAR Protective Action Recommendation PASS Post-accident Sampling Systems PAT Public Affairs Team PNC Protective Measures Coordinator PMN Protective Measures Manager PNr Protective Measures Team P-T Pre s sure-Tempe ra ture PWR Pressurized Water Reactor RA Regional Administrator RAT Radiological Assistance Team RBE Relative biological ef fectiveness RCIC Reactor core isolation cooling RCr Response Coordination Team RDO Regional Duty Officer R.G. Regulatory Guide RHR Residual hea t removal RI Resident Inspector RIRC Regional Incident Response Center xii

.~ -_

RN Resource Manager R0 Regional Office RSC Reactor Saf ety Coordinator RSM Reactor Safety Manager j RST Reactor Safety Team SC Safeguards / Security Coordinator SFO Senior FEMA Official SGC Saf eguards/ Security Coordinator SGT Safegasrds Team SI International System (of measurement)

SLC Standby liquid control SM Safeguards / Security Manager SO Status Officer ST Site Team SE Site Team Leader TLD Thermolumine scent dosimeter TNI-2 Three Nile Island-Unit 2 TSC Technical Support Center USDA U.S. Department of Agricni ture WHO World Health Organization xiii

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ACKNOWLEDGMENTS The authors wish to express their appreciation for the valuable assistance provided by the following people: Suzan R. Morris, Ursula F. Strong, and Malinda M. Hutchinson, for word processing and coordination; and Larry H. Wyrick and the staff of the ORNL Graphic Arts Department for preparing illustrations and view graphs.

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1. SEVERE ACCIDENT OVERVIEW Slides 1 and 2 1.1 OBJECTIVES Following completion of this section, the student should be able to describe e the dose levels required to produce oarly health effects; e basic radiation protection criteria, the U.S. Environmental Protection Agency (EPA) and U.S. Food and Drug Administration (FDA) protection action guide (PAG) levels, their relationship to health effects, and how they are used; e the enount of radioactive material that must be released to produce early health effects and to exceed the Environmental Protection Agency protective action guides at a distance of about 1 mile from the plant; 6 the location in the plant of the radioactive material required to produce early health effects, the damage required for its release, the means for detection of this damage, and the level of certainty of the detection; e conditions in the control room during the initial phase of an accident, including fundenental tasks the control room staff must accomplish;

• how a knowledge of plant conditions (e.g., core temperatures) is essential in projecting consequences of an accident and why in plant radiation monitors cannot be relied upon entirely to estimate plant releases; 1

2 e the uncertainties associated with dose assessment,-

including those between (a) what is predicted and what may actually be measured and (b) different model estimates; 4 why it may be impossible to project dose adequately during the early phase of a severe accident; e the role of dose projection during a severe nuclear power plant accident (i.e., General Emergency) and daring lesser events; and e why early protective actions for the population at greatest risk (those nearest the plant) should always be recommended for all directions near the plant and not just downwind.

1.2 THE PROBLEM Slide 3 Virtually all commercial nuclear power reactors in.the l United States are either pressurized water reactors (PWRs) or boiling water reactors (BWRs). These tycres of re2ctors are t i called light water reactors (LWRs) because the reac.or core is covered with water to allow the nuclear reaction to take place and to keep the core cool (see Appendix D of Vol. I for a general description of light water reactors). This manual discusses only accidents at light water reactors.

Another type of reactor, which uses helium gas to cool the core, is called a high-temperature gas-cooled reactor (HTGR).

The only high-temperature gas-cooled reactor in the United States is located at Fort St. Vain, Colorado. From a response standpoint, high-temperature gas-cooled reactors have a major advantage over -light water reactors. For a high-temperature gas-cooled reactor, it would take several hours af ter the loss of the systems designed to keep the reactor core cool before a maior release would be possible (vs minutes for some light water

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3 4

reactors).

1.2.1 Radioactive Katerial Reanired for Health Effects Slide 4 The basic sources of safety problems at an operating nuclear power plant are the very large amount of volatile radioactive materials that, if released, could cause off-site

-health effects and the energy in the core that, if not controlled, could release these fission products. Even if the reactor has shut down, substantial energy is stored in the reactor systems and is being generated by the decay of fission products (decay heat). During the first hour after shutdown, decay heat is on the order of 3% to 5% of full power. For a 3300-MW(t) [3300 megawatts (thermal)] [3,300,000,000 watts (thermal)]' reactor, initial decay heat is about 200 MW(t),

which, if not controlled, can be a substantial driving force for release of the radioactive materials in the core into the

, environment.

Slide 5 When discussing quantitles of radioactive material, curies are used; the symbol is C1. A curie is a direct measure of the enount of radioactivity. The number of curies required for various health effects can vary considerably, depending on the type of radiation and how it enters the body (i.e., the pathway). Although curies are helpful in discussing amounts of radioactive material (e.g., curies in the core), they are not useful in discussing the health effects that may result from their release. The amount of radioactivity absorbed by a body is called dose. One unit of dose is the rom. The rem is a measure of potential biological damage induced by radiation exposure to the body, so the rem is used in discussing health

! effects.

4 Scientific notation and units are discussed in Appendizes B and C of Vol.1.

Slides 6 and 7 Table 1.1 shows the principal components of the 5 billion or so curies of radioactive materials in the core of a light water reactor 30 min af ter shutdown. Table 1.2 shows inventories of the most volatile radioactive materials (noble gases and radiciodine) in various plant systems. The table also shows the percentage of the various radioactive materials that could be released under worst-accident conditions. Note that

the vast majority of radioactive material is contained in the*

f core. All other reactor systems contain less than one-half of 1% of the activity in the core.

! Noble gases receive special attention because they are the most .likely material to be released during an accident.

Radiolodine receives special attention because it can be a major source of the dose to the public early in a severs accident.

L Radiciodine also can concentrate in the thyroid and in the food l chain (i.e., milk). As a result, small quantities of radiciodine can cause damage to the thyroid gland.

Slide 8 As a result of a release of radioactive material to the atmosphere, a person can receive a radiation dose via three major routes (referred to as pathways): (1) shine, (2) inhalation, and (3) ingestion. Individuals receive shine dose externally from direct exposure to a passing cloud (plume) or from contamination deposited on the ground, on themselves, or on other obj ects. Inhalation dose is received from breathing radioactive material in the plume. Shine and inhalation are generally considered parts of the exposure pathway. Specific protective actions have been established for the plume exposure pathway because of the need for prompt actions. Ingestion dose l

5 l

Table 1.1. Summary of volatility of radioactive material in a large [3300-NW(t)] light water reactor core Volatility Inventory (alllions of C1)

Very volatile--100% release from fuel possible Xenon (Ie) and krypton (Kr)--noble gases 400 Iodine (I) 750

' Cesian -( Cs) 15 Moderately volatile--l-20% release from fuel possible 2

Tellurium (Te) 200 Strontian (Sr) and barium (Br) 370 Not volatile--small percent release possible Other 3500

--3 . - . _ _ - . _ _ _

6 Table 1.2. Typical inventories of noble gases and iodine in reactor systems Inventory (C1)

Noble gases (Ze, Kr) Iodine (I)

Reactor core total 400,000,000 750,000,000 Reactor core gap" 30,000,000 14,000,000 Spent fuel storage pool 1,000,000 500,000 b Primary coolant' 10,000 600*

Pressurized water reactor--other systems

(- Waste gas storage tank 100,000 1 Boiling water reactor--other systems Steam line 10,000 d 25 d r

Waste gas treatment system 5,000 0.25 l

Shipping cask 10,000 1 "See Sect. 1.2.2 for a discussion of " gap. "

One-third of the core is 30 days old; the rest is 1 year old.

' Nominal value at normal iodine levels can be much higher or lower (factor of 10) depending on fuel leakage.

C1/hr (circulating).

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- - - - =, ___

7 is received from eating or drinking contaminated food or water.

Ingestion of milk receives special attention because iodine from a plume can contaminate grass eaten by dairy herds; this radioactive lodine, which can be greatly concentrated in the allk, can then concentrate in the drinker's thyroid gland.

f These pathways can result in doses being received by various organs. Doses to the whole body and to the thyroid are commonly used as reference levels for protective actions. Cloud shine, groand shine, and inhalation all make major contributions to whole body dose. Inhalation is the major contributor to early thyroid dose.

Slide 9 A release to the atmosphere of major fractions of the radioactive material contained in the reactor core could result in two types of health effects that must be considered in emergency response. The first type is the potential for early

! deaths and injuries (weeks or months). The second type is the longer-term (latent) health ef fects (e.g., cancer) that would not be directly observable following an accident. Early injuries generally would appear at doses above 50 to 100 ren to the whole body, and early deaths would be expected at much higher doses (e.g. , 200 to 600 res) . It has been estimated that, with only supportive medical treatment, about 50% of the people who receive a whole-body dose of 350 rem would die within i

60 days.

j Risk of cancer is generally presumed to be proportional to dose,

no matter how small. The computer models assume that a collective dose of about 5000 person-rem (e.g.,1 ren to 5000 people) will result in one member of the affected population getting cancer. Because the release is spread over a larger area and therefore over a larger population the farther it moves from the plant, most of the cancers projected by the computer models will result from very small exposures bevond 50 miles from the plant.

I _, ___ _ _ . . _ _ _ _ - . _ _ , _ . _ . _ _ . _ , _ _ _ _ _ -. _ ._ .._ __

8 Slide 10 i

The current dose levels in the Federal guidance that indicate when protective actions are warranted are referred to as protective action guides. These guides have been conservatively established at levels well below the -level's that would cause early health effects. At these levels, no health effects would be detectable, even for senritive populations such as pregnant women.

The Environmental Protection Agency has established, protective action guides for plume exposures, and the Food and i Drug Administration has established protective action guides for food and agricultural pathways (see Table 1.3). It is important to note that protective action guides are nrofected doses. At any time, previously incurred doses are not to be considered.

{ 1.2.2 Plant Da==ne Reanired for Protect Ive -

Action Guides to Be Exceeded i

j Slide 11 l It is instructive to compare the amount of radioactive material in light water reactor systems to the amount necessary I

for release to the atmosphere to induce doses equal to Environmental Protection Agency and Food and Drug Administration protective action guides; note that dose levels ten or more i times higher than the protective action guides are required for early injuries or fatalities. Table 1.4 shows the number of curies that would have to be released to the atmosphere to j result in doses equal to the protective action guides under meteorological conditions that would, for a given release, produce higher-than-average dose levels at a distance of about 1 mile. Under average meteorological conditions, about ten times i

more radioactive material would have to be released.

[ _ _ . _ . _ . . _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . - - _ _ _ _ _ . - _ _ _ _ _ . . _ - .

9 Table 1.3. Environmental Protection Agency (EPA) and Food and Drug Administration (FDA) protective action guides (PAGs)

FDA PAGsb go, Organ EPA PAGs" for food and agricultural plume exposure (rea) products (rem)

Whole body (bone) 1-5 0.5-5 Thyroid 5-25 1.5-15 Other body organs 1-5 0.5-5 "At a lower projected dose, protective actions should be considered; at a higher projected dose, protective actions wonid be warranted.

At a lower projected dose, use of grazing land should be restricted; at a higher projected dose, contaminated milk shon1d be impounded.

10 Table 1.4. Atmospheric release (C1) necessary under poor meteorological conditions" to result in protective action guide levels at 1 mile o

Curies releasedb Radioactive, material Pathway 15 ren--thyroid

• 5 ren--whole body Iodine (I-131) Milk ingestion 2 -

Inhalation 600 -

Noble gases (samma emitters; Ie, Kr) Cloud shine -

1,500,000

" Conditions that resnit in doses higher than those projected under average conditions.

Approximate minimum.

Child's thyroid.

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i 11 Slides 12 and 13 Table 1.5 shows a summary of annual releases of noble gases and lodine during normal light water reactor operation. It is clear that release rates would have to increase by as much as

. .100,000 times above normal rates for protective action guides to be exceeded by a 1-hr release.

Slide 14 Comparisons of Tables 1.2 and 1.4 will show that the release of even a very small fraction of the core radioactive material inventory to the atmosphere could result la doses exceeding the protective action guides near the site. However, only the core, spent-fuel storage pool, and the primary coolant contain the requisite amount of the inventory. Accidents not involving those systems (e.g., gas-decay tank rupture) should 1 not result in off-site doses in excess of the Environmental Protection Agency protective action guides.* Furthermore, only I the reactor core contains sufficient radioactive material And enerav (e.g., decay heat) to result in prompt atmospheric releases that could result in early deaths and injuries off

! site.

However, even if the core failed, its failure alone would

! not be sufficient to cause off-site health effects. Other engineered safety features (ESFs) would have to fall also.

i

• One caveat is important: the Food and Drug Administration has proposed a preventative protective action guide of 1.5 ren for the ullk pathway; at this level, dairy animals should be removed from likely or actually contaminated pasture. Catastrophic accidental releases of 131 I from the waste gas storage tank at a pressurized water reactor site or from the effluent treatment system at a boillas water reactor could result, especially during a period of precipitation, in pasture contamination leading to a projected dose of 1.5 ren or greater via the contaminated allk ingestion pathway.

-- _ _! .a +-J 12 Table 1.5. Typical releases Release Source Noble gases Radiolodine Boiling water reactor I

Annual total, C1 100,000 1 Release rate to 10 0.0001

equal annual total, C1/hr" i Pressurized water reactor Annual total, C1 200 0.02 Release rate to 0.02 0.000002 equal annual total, C1/hr" 1-br release rate to exceed Environmental Protection Agency whole-body protective action guide via cloud shine, C1/hr 1,500,000 -

1-hr release rate to exceed Food and Drug Administration thyroid dose via inhalation, C1/hr -

600 "Using 10,000 hr/ year.

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2 13 Plants were designed to control most releases from these systems. In partionlar, the containment system is designed to limit releases from the fuel to the atmosphere.

Slides 15 and 16 Four barriers are designed to prevent the release of radioactive materials to the environment from t'careactor core (see Fig. 1.1). The first barrier is the structure of the fuel itself in the form of aranium dioxide (002 ) fuel pellets; UO2 is a ceramic. The individual fuel pellets are stacked upon each other in zirconium or stainless steel tubing (called fuel pins l or cladding) and sealed, forming the second barrier to release.

The space between the fuel pins is called the gap. Gaseous radioactive materials produced in the fuel during the fission process are trapped in this gay region. These trapped gases are called the " gap" activity. They are of special importance

, because they can be released very quickly from the fuel once the cladding is damaged. This would result in a large increase in the levels of radioactivity in the primary coolant at the start of core damage. As seen in Table 1.2, the " gap" contains about 105 of the noble gases and 2% of the iodine in the total core.

The fuel rods (pins) are assembled into fuel elements that are all contained within the primary cooling system (vessel and piping), which forms the third barrier to fission product release. Finally, the primary coolant system is contained within the containment building, the fourth barrier.

The term " core damage" is used to indicate major (20% or more) cladding failure (second barrier). A major release is not

! possible unless a major failure of cladding allows radioactive material to be released from the core into the primary coolant.

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>c (~1 cm)

~200 FUEL _ RODS __

IN EACH SUNDLE

,f Fig. 1.1. Barriers to release of radioactive material for a pressurized water reactor.

. ~ _ _ .

15

~ Slide 17 Core damage (or core melt) should be distinguished from fuel melt. The fuel (D0 ) 2has the highest melting point [5100*F (2800*C)] of all materials in the core. However, releases of radioactive material from the core that are sufficient to result in major off-site releases are possible at much lower temperatures. At temperatures above 1500 F, fuel cladding failure and release of the " gap from the core are expected.

Once the cladding falls, radioactive material in the fuel can escape into the primary system. At temperatures above 15000F, the volatile radioactive materials (e.g., iodine and cesium) are being released from the fuel pellets at a considerable rate.

Therefore, severe core damage is expected at core temperatures above 15000F. In a pressurized water reactor, these temperatures could be reached in 15 min if the core were not covered with water. Normally, temperature in the reactor pressure vessel, which houses the core, averages about 600*F.

Figure 1.2 illustrates the various levels of core damage, release rates of volatile radioactive materials as temperature increases, and the time af ter uncovery of the core that these temperatures could be reached in a pressurized water reactor.

Reaching these temperatures would take longer in a pressurized water reactor because more of the core must be uncovered before major heat-up starts.

Slide 18 Many systems are designed to protect the fission product barriers. The effectiveness of these systems can be assessed in terms of how well they perform a few critical safety functions i

16 3

. 5100*F MELTING OF FUEL PELLETS (UO2 I (3088 K)

(RELEASE OF ALL VOLATILE FISSION PRODUCTS FROM FUEL 60 min - --

3600*F FORMATION OF "LIQUEFlED FUEL"- FUEL DISSOLVES IN n (2255 K) MELTED COMPONENTS EEE

=a5 g2e TWE 5 o (VERY RAPID RELEASE OF IODINE, CESIUM, AND NOBLE GASES FROM FUEL AFTER

" 2200*F STEAM -ZlRCALLOY REACTION RELEASES H2 UNC0VERY --

0F (1477 K)

FWR CORE 2000*F o (1366 K)

E a

$ (CLADDING BURST; RELEASE OF GAP INVENTORY FROM FUEL PIN d

g._ 1400*F 15 min - E--('

E__ 1200 F CLADDING COLLAPSE POSSIBLE

$ (920 K) o a

-- 600 F NORMAL OPERATING TEMPERATURE Fig. 1.2. Possible core damage states as core temperature increases.

- . - . . . - - - _ - - .. . ~ . . _ - .- -.

17 (CEFs). The critical safety functions will, if naistained, prevent damage to the core. Each of these functions is performed by a number of redundant engineered safety features.

Basically these functions are i 1. shut down the reactor (control reactivity),

2. maintain coolant level (keep core covered),

i

3. provide coolant makeup (keep water flowing through the core), and

4. remove heat to an outside (of containment) sink.

These are the primary functions of the control room staff, whether the situation is an emergency or not. Specific emergency operating procedures (EOPs) and training are provided for this purpose. Therefore, the status of these functions and,

consequently, the projected status of the fission product barriers should be known in the control room. The control room should have early warning before a major release.

Slide 19 i

As an example, two of the critical safety functions are to keep the core covered (provide water) and cool (remove heat). <

Once the reactor has been shut down, prevention of core damage l

! is a fundamentally simple task. Ba sically, the task is to keep

{

removing decay heat (energy) from the core hr bolllna 21111 If the decay heat is not removed, it will ultimately result in failure of the primary system and possibly la release of fission I products to the environment. Only a relatively small flow of j water may be required to do this. About 1 hr af ter shutdown, a 1

flow of only a few hundred gallons of water per minute is needed to keep the core cool. Figure 1.3 shows the amount of water

{

I required to remove decay heat by boiling water vs time af ter

. - - - - - _ - _ . - . _ _ , - , - - _ . , - - _ _ . , _ , . . . . . . , , , . ~ . . _ - _ _ _ . . - .---._.,--_m.. , _ . _ _ . . _

18

@,000 -- -

FEEDWATER FLOW RATE (2 PUMPS)

CONDENSATE BOOSTER FLOW RATE 000 (3 PUMPS) 20,000 LPCI FLOW RATE (3 PUMPS)

E a 10,000 - -

S LPCS (1 PUMP) HPCS (LOW PRESSURE) NOTE CHANGE

!= 1,000 - HPCS FLOW RATE OF SCALE o

J (1 PUMP AT OPERATING PRESSURE)

UOO ~ ~

o n O

800 - -

m 700 RCIC FLOW RATE (1 PUMP)

E 600 - -

500 - _,,,,__ _,, _ _ __ _ _ __ _ _ _ _

400 - ~

WATER NEEDED TO 300 - REMOVE DECAY HEAT CRDHS -

200 - -

100 - ------------2---- -

0 SLC I l l l 1 0 1 2 3 4 5 6 TIME (hr AFTER REACTOR SCRAM)

' CALCULATED ASSUMING CONSTANTREACTOR VESSEL PRESSURE AND 100% INITIAL POWER DECAY HEAT PER I

1979 ANS STD.WITH ACTINIDE DECAY). COOLANT INJECTED l

AT 305 K (90*F) AND REMOVED AS DRY SATURATED STEAM.

! LPCI - LOW PRESSURE COOLANT INJECTION

! RCIC - REACTOR CORE ISOLATION COOLING

! LPCS - LOW PRESSURE CORE SPRAY HPCS - HIGH-PRESSURE CORE SPRAY CRDHS - CONTROL ROD DRIVE HYDRAULIC SYSTEM SLC - STANDBY LIQUID CONTROL i

Fig. 1.3. Injected flow required to remove decay heat from the reactor vessel of a 3300-MW(t) plant by boiling water and boiling water reactor systems available to provide the flow.

1

19 shutdown and the boiling water reactor systems available to perform this job.

As illustrated in Fig.1.3, once the boiling water reactor fission process is essentially terminated by a reactor scram (or trip), decay heat can be removed initially by boillag water at a i rate of less than 1000 gal / min. Af ter about 2 hr, boiling water at a rate of 200 gal / min would remove all decay heat. In either case, this would prevent further temperature increases and thereby protect the core.

Figure 1.3 also shows pumping capacities of some safety systems in a boiling water reactor. At high reactor vessel pressure [e.g., 1000 pounds per square inch sage (psis)], only a few systems are normally available to pump water [s.g., the high pressure core spray (HPCS) system in the figure]. Once the vessel pressure is reduced to below a nominal 300 psig or so (depending on the particular reactor design), high-capacity pumps would be available [e.g., low-pressure core injection j (LPCI), feedwater, condensate, and control rod drive hydraulic

. system (CRDBS) pumps]. Notably, during a severe fire at the Browns Ferry boiling water reactor in 1975, the reactor was 1

]

cooled by pumping water through the control rod drive cooling system.

4 j Slide 20 Figure 1.4 shows a diagram of water injection systems in a typical boiling water reactor. As can be seen, there are many redundant ways to get sufficient flow to keep the core cool by j

boiling water and/or to replace coolant lost through a break or i leak. Some of these systems are designed to maintain the critical safety functions even under very severe accident conditions such as a total break in the largest piping in the primary reactor coolant system.

Coro damage would require failure of several engineered l safety features [e.g., the emergency core cooling system (ECCS)]

l that have been designed to maintain the critical safety i

w,,,m_.,w-r,-.-.,,,+,,,,-,n_~,,,,-,e

20 ORN L-DWG 834481 R

/ NORMAL SOURCE f ------------ ALTERNATE SOURCE

4. _

FEEDWATER UNE gliMl!l r-i ,

D U -

. e RECIRCUI.ATION yg CRD WECHANISM ASSEElES {

NORMAL - 60 gpm O

\\\\\\\\\\\\\ SCnAu - 17o epm CRO pnp HYDRAULIC (4 pyyps) 1 PUMP 40,000 gpm S.000 gpm RCIC 600 gpm D ,,,,,,,,,,,,,,,)l CORE SPRAY CONDENSATE (4 PUWS) g .

STORAGE To . . . . . . . . . . . . . . . ....

u.500 gpm

. i l / .,

/

.- f '

~

~.

~.

ES SugE SUPPRESS \0* '

CRD - CONTROL ROD DRIVE HPCI - HIGH PRESSURE COOLANT INJECTION RCIC - REACTOR CORE ISOLATION COOLING RHR - RESIDUAL HEAT REMOVAL Fig. 1.4. Simplified diagram of typical boiling water reactor vessel water injection sources. (See Appendiz D of Vol. 1 for a more detailed discussion of light water reactor systems.)

21 functions. Future core damage may be estimated based on the status of these systems. In addition, the extent of core damage may be surmised (grossly) from several ley plant sa fe ty parameters (e.g., core temperature and coolant level in the reactor vessel and highly elevated levels of radioactivity in the coolant). Although core damage could be readily determined (e.g. , high temperatures, high radiation levels), its extent could be very difficult to determine early in an accident seque nce . Indicators of the status of key plant safety systems and parameters in conjunction with radiation monitoring provide the best, clearest, and probably only indicators of core damage and movement of radioactive material through plant systems.

Clear indications of actual or likely core damage should be evident in the control room before a release to the atmosphere.

Slide 21 Some of the instrumentation and information in the control room that could be used to indicate core damage are listed in Table 1.6.

1.2.3 Plant Conditions Reaufred for a Maior Release Slide 22 In addition to core damage, a major release sufficient to result in early injuries and/or fatalities would require a direct pathway to the environment and a driving force (e.g.,

steam). In essence, the containment systems would have to fall to perform their intended functions. The radioactive material released from the core must move rapidly through the primary coolant system (third barrier) and containment (fourth barrier) without being significantly filtered or reduced by other methods such as containment sprays. Even if the containment sprays

22 Table 1.6. Examples of instrumentation and information available for determination of fuel (core) status 1

Instrumentation and information Type of reactor Immediately available in control room Core temperatures--thermocouple Pressurized water reactor (PWR) 3 Core-emit temperature levels Bolling water reactor (BWR)

Containment radiation level BWR, PWR Radiation levels from condenser / air ejector BWR, PWR Neutron fluxes in core BWR, PWR Available af ter several hours Concentration or radiation level in BWR, PWR circulating reactor coolant Analysis of primary coolant- ganna spectrum BWR, PWR

Containment hydrogen level (from samples) BWR, PWR i

1 1

l i

i I

i i

a. + u e _a& *_. _ - 4 ._-_-A-M4_w-444 . . 3: . ._ .a4_, m. -e a.

l f 23

+

Y 1

fail, over time natural - removal processes (e.g., condensation and scrubbing) will remove most particulate fission products from tho' atmosphere of an Antact containment. Therefore, if the containment' holds for several hours and the containment spray systems work or if the release travels through many systems, early injuries or fatalities would be highly salikely.

Slide 23 J

On the other hand, severe-accident analyses have identified i

low-probability accident sequences that could result in i sisaf ficant of f-site releases. As can be seen from Fig.1.5, which summarizes the major sequences, the containment could fall either before or af ter the release of radioactive material from the core.

i 8116'e 24 l

Figure 1.6 uses the concept of an event tree to display the j potential consequences for public health due to severe accidents. Moving from left to right in the figure, 'Yes/mo" answers to scenarios at the top result in a series of branches, possibly to off-site consequences. For example, if only the ,

radioactive material contained in the fuel pins (gaps) is released with late containment failure, the off-site l consequences would be small (branch 7 in Fig.1.6) . If all ,

! answers are yes, branch 1 Indicates extremely severs off-site conseque nce s. Figures 1.5 and 1.6 illustrate two fundamental f

! public health questions during an emergency response at a light j water reactors 1

• What is the status of the reactor core?

i f

e What is the status of the reactor containment?

f

42:

ORNL-DWG 85-17682

h CORE COOLING

CORE MELTS CONTAINMENT FAILS FAILS

~

+

e CONTAINMENT CONTAINMENT COOLING FAILS FAILS CORE MELTS ACCIDENT INITIATING RELEASE OF EVENT RADIOACTIVE MATERIALS CORE COOLING CORE MELTS CONTAINMENT ^'

COOLING FAILS ALLS 3

SPECIAL DESIGN CASES, CONTAINMENT BYPASSED, AND L CORE MELTS CORE COOLING SYSTEM FAILED BY INITIATING EVENT i

Fig. 1.5. Major sequences in a severe light water reactor accident scenario.

l i

25 CORE STATUS CONTAINMENT STATUS CONSEQUENCES

? '

CORE CORE MELT i EARLY EARLY LATE (24-hr)

UNCOVERY i TOTAL MAJOR CONTAIN-

, (GAP , CONTAIN- CONTAIN- l MENT 1.

RELEASE MENT MENT I FAILURE I FROM FUEL FAILURE LEAKAGE l f PINS)  !. 1 (BYPASS) 1 YES

1. EARLY HEALTH EFFECTS LIKELY YES YES

2. EARLY HEALTH EFFECTS POSSIB LE NO

3. EARLY HEAL,TH EFFECTS YES UNLIKELY NO YES O

ji 4. NO EARLY HEALTH EFFECTS

5. EARLY HEALTH EFFECTS VERY

, UNLIKELY NO NO

6. EARLY HEALTH E FFECTS VE RY UN LIKE LY NO

7. NO EARLY HEALTH EFFECTS NO NO 8. NO EARLY HEALTH EFFECTS NO

9. NO EARLY HEALTH EFFECTS s

Fig. 1.6. Event tree for severe accident consequences.

A

26 The answers to these two questions scope the level of threat to the public and the need for off-site emergency response. This translates to an on-site responsibility for cooling the core and containing any releases from the core.

The control room staff will be able to scope the answer to the first of these questions. As we have seen, there will be considerable indicators of actual core damage. However, depending on the severe accident sequences, the control room staff may not be able to confidently assess containment performance.

Slide 25 The U.S. Nuclear Regulatory Commission (NRC), as part of its severe accident source term work, formed the containment performance working group. This group's findings were summarized in Reassessment of the Technical Bases for Estimatina Source Terms (NUREG 0956). The group concluded that, for most risk-dominant accidents, containments would not fail until several hours after core damage and resulting reactor vessel (third fission product barrier) failure. This conclusion was reached partly because containments are not estimated to fail because of overpressure until pressures reach two to three times the design pressurce.

Slides 26 and 27 Containments could f all because of overpressurization; Table 1.7 shows the design pressures and estimated failure pressures for several containment types. Actual failure pressure would be unpredictable. Failure could also result from a burning of hydrogen generated during the core damage process.

If the valves that are designed to 'tutton up" the containment

27 Table 1.7. Typical containment design parameters Allowable Design Estimated leak rate peessure failure Type of plant (vol %/ day) (psis) pressure (psig)

Pressurized water reactor Large dry 0.1 47 134 Substuospheric 0.1 45 119 Ice condenser 0.25 12 30 Bo111ag water reactor Mark I 0.5 62 117 Mark II 0.5 55 140 Mart III 0.4 15 60 n

28 failed to operate correctly, this could also result in a failure (to isolate). Finally, the containment could be bypassed by l I

accidents involving leakage from the primary coolant system I directly outside. While static pressures can be monitored and trended during an event, the actual point of failure and the other failure modes would be unpredictable.

1.3 IllE CDNTROL 200M ENVIRONMENT DURING AN EMERGENCY Slide 28 Any NRC (or other outside) respondent must realize that the reactor control room could be a very busy place under accident conditions. First of all, the plant may be well beyond design with many alarms (annunciators) sounding at once. For example, during the Three Mile Island--Unit 2 (TMI-2) accident, many alarms sounded at one time, including all audible alarms on the front control panels. When many alarms sound at once, it is difficult for the reactor operators to separate important, possibly critical, information from ancillary information.

f Additionally, instruments could be responding erroneously off scale.

11e reactor operators will also be faced with an extremely heavy workload, especially during the early stages of any accident. Based on their understanding of the situation, the operators must take immediate actions to assess and correct the problem and bring the accident to a halt. They must continuously monitor the plant's critical safety functions and attempt to restore them to normal and maintain them there. At the same time, the reactor operators are expected to classify the event; activate their emergency organizations; direct damage control; take on-site emergency actions (e . g. , site evacuation);

and, possibly, make protective action recommendations to off-site officials. In addition, the licensee is expected to notify the NRC and keep them informed of the status of the event and the condition of the plant.

1 m

29 l

The NRC staff's contact with the control room staff at such a time must be held to the level necessary to understand plant conditions and must be handled very efficiently so as not to substantially impede plant recovery by the operators. After the plant Technical Support Center (TSC) is established, communications with the plant can be more extensive.

1.4 PREDICTING OFF-SITE CONSBQUEN E S Slide 29 The most important aspect of predicting off-site consequences is to remember that (1) in the absence of core damage, there is little, if any, excessive risk to the public, but (2) given core damage, there must have been maior hanan error or equipment failure. Under these conditions, there may be little assurance that further failures or a major release is not poss ?. bis because the plant parameters are well beyond their design limits. Some have estimated that as many as one in ten core melt accidents would result in a major release sufficient to cause death and severe injuries off site if effective protective actions were not taken early in the accident sequence.

Slides 30 and 31 In the event of core melt, early, precautionary protective action decisions should be based on in-plant observables (control room indicators) and precalculated dose projections rather than on early, real-time dose proj ections. This conclusion is based on the relatively high risk of a major release (e.g., .10%) given core damage, the relative ease (of using a few key indicators) for the plant staff to detect major core damage, the large uncertainties associated with projecting containment failure, the great difficulties in making accurate and timely dose projections in the face of the latt'er n

30 uncertainty, and the fact that off-site protective actions would be most effective if initiated before a major release occurred (e.g., precautionary evacuation).

1.4.1 Protective Action Initiation Based on Plant Conditions Slide 32 Once a plant experiences imminent or actual core damage, it would be very difficult or even impossible to predict performance of the containment and other systems that are designed to mitigate the consequences of core damage. Because actions. to protect the public must be initiated before or ARAR a major release to the sonosphere for them to be fully effective, the control room staff must rely on core-condition indicators for recommending off-site action.

Slide 33 A major release will be preceded by a systems failure (start of accident), failure of one or more engineered safety features, failure to meet one or more of the critical safety functions, failure of fission product barriers, and movement of radioactive material through plant systems. Considerable instrumentation exists to indicate the status of the critical safety functions, fission product barriers, and (grossly) the movement of radioactive material.

Slides 34 and 35 As discussed in Vols. 3 and 3, current regulations require nuclear power plants to establish four levels of emergencies (classes) for which various levels of response are preplanned.

Licensees have established and incorporated into their procedures emergency action levels (EALs) based on control room l i

instrumentation that would indicate the class of emergency. The A

31 most serious level of emergency is a General Emergency, which should be declared when plant conditions indicate that immediate protective actions should be taken off site. Severe core-damage events have a very real potential for severe off-site health effects and are to be classified as General Emergencies, which would warrant immediate protective action recommendations off site. While some events have been postulated that could lead immediately (quickly) to core damage, most severe accidents studied would be classified as general emergencies by the emergency action levels well before a major release.

Slides 36 and 37 Figure 1.7 shows an example of the emergency action levels established by a licensee to indicate actual or projected core damage (General Emergency). Case A shows the criteria observable in the control room for declaring a General Emergency before core damage based on loss of a critical safety function (ability to shut down the reactor) and indication (containment temperature and pressure) of conditions that indicate imminent core damage. Case B shown the criteria for declaring a General Emergency based on radiological indications of core damage. In this case, core damage would be indicated based on the containment done monitor reading or a hand-held instrument dose rate reading outside the containment.

Slides 38 and 39 In addition to the emergency action levels, licensees have been required to develop procedures for assessing the status of the core. These procedures indicate the relationships of various plant instruments (e.g., containment monitor, water level, or thermocouple readings) and degrees of core damage.

These relationships must be used with caution and may be l considered as only aross indicators of core damage conditions.

n

32 ORNL DWG 85-17688 CASE A EXAMPLE OF NONRADIOLOGICAL INDICATORS OF IMMINENT SEVERE CORE DAMAGE REACTOR PRESSURE GREATER THAN SAFETY VALVE SET POINT REACTOR REMAINS CRITICAL AFTER R APIDLY INCREASING TRIP SIGN AL r AND OR CONTAINMENT PRESSURE (Failure to control.

reactivity)

RAPIDLY INCR EASING

- GENERAL EMERGENCY

^

T MPE

- lNITIATING CONDITIONS

• OBSERVABLES PERSIST FOR 15 min.

TRANSIENT REQUIRING OPERATION OF SHUTDOWN SYSTEMSWITH FAILURE TO SCRAM CASEB EXAMPLE OF RADIOLOGICAL trJDICATORS OF SIGNIFICANT CORE DAMAGE' CONTAINMENT DOSE RATE OUTSIDE MONITOR READING OF CONTAINMENT

>2000 R/hr** HATCH >10 R/hr**

OR ir GENERAL EMERGENCY

' Presumed to be corroborated (i.e., not merely a faulty reading).

, *

• Site-specific parameters - should be precalculated.

Fig. 1.7. Examples of emergency action levels for a General Emergency.

___-____n_-____-__ _.

i 33 l Slide 40 As an example, Table 1.8 shows typical derived relationships between cantainment monitor levels and postulated core damage states for a number of light water reactors as reported to the NRC. These would be key indicators of the level of core damage for accident sequences that release into the containment. These plant-specific estimates of containment

~

radiation levels associated with various levels of core damage vary greatly. This variation may be due to (1) basic accident scenario (release) assumptions used for the estimates or (2) assumed monitor efficiency, shielding, and location in the containment. Nevertheless, the containment radiation monitor readings would increase by several orders of magnitude (1000 times) for progressively more severe core damage levels, s uch that the plant operator could recognize the difference between a minor problem and a maj or one. Co re damage sufficient to declare a General Emergency would involve the release of about 20% or more of the gap inventory. In the examples in Table 1.8, the containment monitors would read between 100 and 20,000 R/hr for this level of damage. For purpose s of comparison, the radiation reading for a major leak of primary water into the containment, without significant core is expected 1 to 10 R/hr or less.

~

34 Table 1.8. Typical calculated containment radiation estimates (R/hr)"

Assumptions Plant 1005 gap 1% core 105 core 100% core activity Type of reactor activity activity activity A 400 4,000 40,000 -

Pressurized water B 1,900 1,900 18,000 200,000 Boiling water l

C 165,000 - - -

Pressurized water D -

5,000 50,000 500,000 Boiling water E 80,000 150 70,000 70,000 Pressurized water l F 100,000 10,000 100,000 1,000,000 Boiling water i

"R = roentgen, the international unit of x-radiation or gamma radiation.

i I

35 1.4.2 Protective Action Initiation Based on Dose Protections Slide 41 In the past, considerable attention has been given to the use of dose projections as the only basis for initiating off-site protective actions. However, it should be recognized that, for some very severe accidents, real-time dose projections would be available too late for initiation of an early (most effective) protective response off site. Studies have shown that, for a very severe accident with a major release shortly af ter core damage, evacuation of the population near the plant (2 to 3 miles) must start before or soon af ter the release to prevent early severe health effects. (The effectiveness of protective action is discussed in Vol. 4 of this manual.) Early evacuation of all nearby areas would considerably reduce all radiation exposure risis.

Slides 42 and 43 Predicting dose requires several steps: (1) predicting the quantity and timing of the release from the plant (source tera),

(2) predicting the movement of the plume (transport), and (3) predicting the dose from the plume.

1.4.2.1 Predictina the Release from the Plant 4

? 4.2.1.1 Predictina the Source Tern Followinn a Severe (Core Melt) Accident 1

Slides 44 and 45 During a severe (core molt) accident, predicting the release (i.e., the source term) would be the first step in projecting doses. Therefore, any errors associated with predicting the source term would have a direct impact on the

36 accuracy of dose projections.

As part of the upgrades that followed the Three Mile Island--Unit 2 accident, on-line radiation monitors capable of measuring the noble gases released through plant vents were -

installed. On-line monitors for iodine and other particulates were not considered practical. Therefore, the presence of iodine and particulates in a release is determined through analysis of samples taken during the release. This could require several hours. Note that noble gases are not considered as great a threat to the public as radioactive iodine and other particulates. Although current systems can characterize most releases, they cannot provide fast estimates of those very unlikely releases that pose the greatest threat to the public.

Even more importantly, the plants are designed to accommodate routine releases of radioactivity and to minimize releases resulting from abnormal conditions and accidents. However, because an accident resulting in off-site early health effects (death and injuries) would have to be fast, direct, and unfiltered, such a release would most probably be via an unmonitored pathway to ths atmosphere. The most important example is a release due to a major containment failure or major containment penetration failure. As a result, effluent-monitoring systems located in routinely monitored release pathways (e.g., stacks) will not be able to assess the extent and the characteristics of such a severe release.

For accidents where the total release is through a monitored pathway (e.g., the stack), it may be possible to obtain a good characterization of the release. At a minimum, the magnitude in relative terms (e.g., this release has the possibility to exceed Environmental Protection Agency protective action guides) can be estimated---if the monitors stay on scale.

By their very nature, however, releases resulting in off-site dose high enough to cause early health effects most likely cannot be characterized by effluent monitors.

j

i l

1 37 1.4.2.1.2 Current Rea s se s sment of Potential Source Terms f

Slide 46 Since the Three Mile Island--Unit 2 accident, considerable research has been conducted by the nuclear industry and NRC and its contractors on the reassessment of potential source terms.

Reassessment of the Technical Bases for Estimatina Source Terms (NUREG 0956) describes this effort and states the following interesting conclusions.

• Containment performance (sirvival, failure, or bypass) is a major factor affecting sonace term.

e The analyses performed to dare suggest that generalizations are inappropriate since large factors of reduction in source teru were not found for all sequences (accidents) as reported in other studies.

The study further showed that, even if all plant conditions were known, the current computer models could predict the source term only to within a factor of 100. This highlights the difficulty of source term estimation because, during an actual accident, detailed plant conditions would not be known.

Slide 47 The resnit is that, early in the response to a severe accident that results in a major release, it will be very difficult to predict the " source term" with a reasonable degree of accuracy.

I

38 Slide 48 Some simple tools have been developed that could at least provide a bound of possible off-site doses, based on consideration of plant (core and containment) conditions.

Figure 1.6 is a very simple example. Other examples are the 15 precalculated off-site dose projections for severa accident conditions and for many meteorological conditions presented in Dose Calenlations for Severe LWR Accidents (NURBG-1062).*

Such precalculated dose projections must be used with considerable caution, just as with a real-time dose projection, principally because of the major uncertainties of what the source term would be during a severe accident sequence.

However, projections of source terms based on consideration of plant conditions have the advantage that all possible accidents and release pathways can be considered.

1. 4 .2 .2 Transnort and Dose Once an estimate has been made of the radioactive material being released from the plant, movement of the material through the atmosphere must be characterized.

Slide 49 Even if one could accurately predict the radioactive material that may be or is being released from a plant during a severe accident, significant uncertainties would still be N associated with dose projections to off-site areas. Dose models give a very simple picture of a very complex situation.

• See Vol.1, Appendix A, 'Sibliography for Volumes 1-5. "

4 t

39 Slide 50 In a 1981 study conducted at the Idaho National Engineering Laboratory, a nonradioactive tracer (SF ) 6was released and the resulting air concentrations were compared with predictions made by various models to evaluate their potential use in emergency response situations. Figure 1.8 shows the actual air concentration (plume) pattern observed for one of the tests and

~~

the plume pattern predicted by three of the models tested under this program: (a) a simple, straight-line Gaussian plume model of the type used by many emergency response organizations, (b) a Gaussian puff trajectory model like that used in the NRC emergency response center, and (c) a more sophisticated windfield and topographic model used in the Department of Energy's (DOE's) Atmospheric Release Advisory Capability (ARAC) program. Even the most complicated model could not reproduce what actually occurred.

Slides 51 and 52 r

This result points out two concerns. First, typically, only one local meteorological tower is in the site vicinity.

The initial transport of radioactive material from a site af ter it is released to the atmosphere will be dominated by local conditions (e.g., hills, valleys, lakes, and precipitation).

, This single source of weather and wind information cannot give a definitive indication of winds away from the plant.

Nuclear power plants are typically located in very complex areas (e.g., in river valleys or on the coast) where wind direction and flows can vary considerably within a short t

distance of the plant. As an example, a 180* dif ference in wind direction could result from sea breeze effects at a coastal site. This is the basis for taking protective actions in all directions near the plant (within 2 or 3 afics). The events that occurred early in the Three Nile Island incident illustrate s

40 ORNL-DWG 8517692 I l l l 1 I I l l l l [

MAXIMUM DOSE MAXIMUM DOSE 150.1 212.3

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Fig. 1.8. One-hour surf ace dose s predicted by (a) Gaussian plume model, (b) puff-trajectory model, (c) coa. plex numerical model, and (d) doses actually observed.

I 1

l 41 ,

l 2 the' problems inherent in taking protective actions only in the downwind direction. This will be discussed later.

Slide 53 Second, differences should be expected in the estimates 4

produced by various dose models even if the same input conditions (e.g., source terms and meteorology) are used to estimate off-site dose s. Various response organizations will also be performing analyses based on different assumptions. For example, the NRC may be concentrating on dose projections based on possible additional plant failures, while the state is making dose projections based on estimates of actual releases.

Therefore, a 10- to 100-fold (or greater) spread in calculated doses must be anticipated among the various response organizations (licensee, NRC, state and local officials, and U.S. Department of Energy).

What may be more important than relying on a dose model in estimating plume movement is a knowledge of local meteorological conditions and trends (e.g., the winds shif t every morning at about 9:00 a.m.).

The basic point here is that the responder needs to understand the problem, the technical aids, and the results.

Indiscriminant use of technical aids such as dose projection l

models without access to staf f who understand the unpredictability of local conditions can provide misleading input to protective action decision making.

As emphasized earlier, in the early emergency response

[

l phase, dose projections for protective action decision making should be secondary to assessment of plant conditions and general weather at the time. After a release, environmental monitoring information should form the basis for additional protective actions.

42 Slide 54 Environmental monitoring would be the best way to characterize a release af ter it occurs. However, one must be sensitive to the differences between actual plume behavior and that sinalated by models. Dose projection models project the averaae dose as the result of pinae meander over a 15- to 30-min period. Therefore, as can be expected, a monitoring team within the actual pinne (Fig.1.9, point B) may observe arenter dose s than proj ected or, if the team is out of the pinae (point A),

lower doses than proj ected. Even if the model projections are

" correct," actual field-monitoring results conid differ considerably from projections because the projections show averages. One would A21 expect the first preliminary . field monitoring results to agree with model proj ections, even under the best circumstances.

1.4.2.3 Estimates of Overall Uncertainties in Severe Accident Off-Site Dose Protections Slide 55 Based on conclusions drawn from the preceding discussions about the difficulties of estimating source term and plume movement, an overall estimate of the uncertainties associated with dose assessment for severe accidents has been made by the NRC technical staf f. These estimates, given in Table 1.9, are estimates of the ratio of what a model may proj ect for an accident sequence and what the actual average dose rate may be.

It is apparent that, overall, the best that should be expected in the early time frame is that projected dose estimates may be within a factor of 10 of the true dose value; more likely, they will be even less accurate. Table 1.9 also shows that the largest single component of uncertainty is expected to be the uncertainty in the estimate of the source term. Unanticipated catastrophic containment failure is an example of a case where

ACTUAL PLUME MEANDER WHICH IS AVERAGED OVER 15-30 min TO OBTAIN AVERAGE CONCENTRATIONS MODEL AVERAGE CONCENTRATION p N:.:M ::h.

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! MONITOR LOCATIONS l (BOTH ARE IN PLUME j ACCORDING TO MODEL) 1 l Fig. 1.9. Relationship between an actual plume and model 1 proj ections. I

l 44 Table 1.9. Estimated range of uncertainty between early projected dose and actual off-site dose for a severe accident (core melt)" Uncertainty factor b At best Most likely Near worst

i Source tern 1 (event and sequence) 5 100-1000 100,000 l 1

Dispersion Diffusion (concentration) 2 5 10 Transport (direction) 22' 45 0 1800 Transport (rate) 1 2 10 (Iow wind spee d) Dosimetry 3 4 5 l Overall (dose and direction) 10, 100-10,000, 100,000, 22* 45" 180 "These estimates are for an averaged dose at a location (e.g., over 15 to 30 min), not for a specific or single monitor reading. b Ratio of a likely maximum or minimum value to the expected median value. I t

45 source term could be underestimated by a factor of 100,000. For lesser accidents (non-core damage) where the total release is through a monitored pathway and consists mostly of noble gases, the source term uncertainty can be reduced. However, the transport and dose uncertainties would remain unchanged. It is clear that one should not expect close agreement when comparing various dose projections with each other or with early field monitoring data. Dose projections should be viewed only as rough estimates. 1.4 . 3 A Bad Examnie: Enriv Protective Action Decisions Durina the Three Mile Island--Unit 2 Accident Slide 56 To highlight some of the points discussed in this section, certain aspects 'of the assessments of the Three M11e Island-- Unit 2 accident merit discussion. Figure 1.10 presents the hourly wind vector as measured by the site meteorological system during the first day of the accident. It is evident that wind direction 31 1hg slig varied dramatically throughout the 12-hr pe riod. Between 7:30 and 8:00 a.m., the State of Pennsylvania issued warnings of imminent evacuation to the west of the site. (A Site Emergency was declared at 7:20 a.m., followed by a . General Emergency at 7:30 a.m.) At 8:10 a.m., this preparedness was reduced to a standby notice because dose rate measurements to the west were "only" about 30 millires/hr (i.e. , about 10,000 times higher than the dose rate resulting from normal effluent rele a se s) . This reduction-to-standby notice came coincident with core uncovering.

• Actually, these measurements were not available to the NRC until l three days later because the plant computer crashed early in the accident progression.

1

46 ORNL-DWG 8517691 MARCH 28,1979 HOURLY WIND VECTOR

• TMI-2 N

r -160g

                                                 @*                      1 4Q 1300 1200 09 @

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                                                                                  \

W Os

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03 w E opg0 79 5 mph 0400 10 mph 03 15 mph S

             ' Arrows indicate direction toward which the on site wind was blowing at the local time indicated. Circles represent varying wind speeds.

Fig. 1.10. Hourly wind vector at Three Mile Island on March 28, 1979. l

47 Moreover, even if an evacuation to the west of the site had been initiated around 8:00 a.m., local wind conditions would have shif ted the potentially affected area to the north by 9:00 a.m., and then to the east by 11:00 a.m. Thus, the wrong people would have been told to evacuate. As the NRC Specisi Inquiry Group noted later, based on in-plant observations as set forth in the emergency plans and as emphasized in NRC emergency planning guidance in place even at the time (R.G. 1.101), annidirectional evacuation of the total low population zone (2.5-mile-radius area surrounding the site) would have been warranted no later than 7:30 a.m. Slide 57 By 9:00 a.m., indications of severe core damage were indisputable. Some of the core thermocouples showed temperatures over 2000*F (800*F beyond that required for cladding failures--see Fig. 1.7), and the containment done monitor increased from 600 to 6000 R/hr between 8:20 and 9:00 a.m. However, as indicated, the decision not to take action was made based on field-monitoring results. The NRC Special Inquiry Group found that the state offices should have been advised at 9:00 a.m. that "the core has been badly damaged and has released a substantial amount of radioactivity. Iha niant is in a condition not nreviousiv analvmed for cooline system nerformance. " The Inquiry Group went on to state, "The difficult question in this situation is whether to advise precautionary evacuation of the nearby population or to advise only an alert for possible evacuation. The recommendation to evacuate is consistent with what we think would then be the case, a prudent doubt that the core-cooling passages were still sufficient for cooldown. la addition, the contal== ant buildina was now fillina with intenselv radioactive nas and vanors. leavina the nearby nublic nrotected

48 by oniv one remainina barrier, the containment, a barrier with a known leak rate that needed only internal pressure to drive the leakage . " Finally, the Inquiry Group stated, 'Present emergency plans are inadequate because they do not provide a clear requirement to evaluate the need for protective actions based on deterioration of plant conditions." This example illustrates the importance (for core melt accidents) of implementing protective actions in the nearby areas as soon as core damage is detected and without regard for wind direction or detection of actual major releases. These are two of the foundations of current NRC staff emergency planning guidance. Early precautionary evacuation of the immediate area (approximately 2-mile radius) should A21 be recommended in only

                            " downwind" directions because of the inability to determine where downwind will be when the protective actions are actually implemented or when a significant release occurs.                In addition, when core damage is detected, the early recommendation to evacuate should not be based on early real-time dose project!9as but on the status of the core.                Indeed, the predetermined, early, initial evacuation for a General Emergency is called l                            '9recautionary" because a major release may never actually I

occur, as was the case at Three Mile Island--Unit 2. On the other hand, no immediate, early evacuation would be warranted for sequences less serious than core-melt accidents. 1.4 .4 Role of Off-Site Dose Protection in Severe Accidents Slide 58 The role of dose assessment during a nuclear power plant accident will depend on the type and phase of the accident. Dose assessment during the initial phase of a severe accident (General Emergency) provides a basis to establish priorities for the use of limited resources in the implementation of off-site actions such as deployment of field-monitoring teams. In an

                                   -_. -     -                        -_.       .          _     - - - ~ _ . . . _.

49 actual uncontrolled release of radioactive material to the environment, it would be imperative to obtain off-site monitoring team data as quickly as possible. However, for a - core-melt sequence, early protective actions in nearby areas (2 to 3 alles) should not await such results. In particular, the evacuation of nearby areas for a core-melt accident meanence (i.e., General Emergency) should be initiated on the basis of plant conditions. Af ter implementation of protective actions near the plant (based on an assessment of plant conditions), dose projections should be used to determine whether these actions shon1d be extended. The model projections may indicate the maximum distance from the plant where further actions are required. However, because of the difficulty of. projecting plume movement, actions should not be limited to just the downwind areas. Actions should be taken in all directions or at least in all areas considered to be the possible limits of the plume under f various conditions (e.g., inside a valley). Bounding dose calculations may be very useful in connarina the consequences of various plant response options (e . g. , venting the containment vs allowing later containment failure). For Site Ares emergencies and for the intermediate phase of a General Emergency (after early protective actions have been implemented for the population near the plant), the roles of dose assessment differ significantly from those discussed earlier. The first role is to assess the areas that may warrant implementation of protective actions according to radiation protection obj ectives. The second role of dose assessment under these conditions is to provide feedback regarding the magnitude and composition of a release based on the analysis of off-site

50 samples and monitoring results. 1.5 ROLE OF FIELD NONITORING Slide 59 Actual field monitoring can be used to determine accurately the actual dose rates and projected doses off site as the result of an accident. As discussed in Vol. 4, the role of field monitoring during the early phases of a severe accident would be to identify areas that may require further protective actions following a release. Reliance should be on field monitoring as soon as possible. Slide 60 Because the actual location of the plume or resulting ground contamination may not be known for some time, early, limited field-monitoring results should be used with great care. Even if the monitoring team is in the path of the release, the plume could be overhead (as shown in Fig. 1.9) or could meander or loop around the team. It will be difficult to obtain readings that are considered representative of the release. Slide 61 The biggest problem with field monitoring, as shown in Fig. 1.8d, is that the actual distribution of off-site dose could be very complex. The dose rate could change over sery short distance (hundreds of feet). " Hot spots could be surrounded by areas of lower dose rates. Therefore, an aircraft or large numbers of monitor teams would be needed to fully l

   +

51 characterize a major release in a short time. The teams should have instruments designed to monitor all radioactive material (iodine, cesium, strontium, and tellurium) that may be released during an accident. If air samples are t ake n, this analysis could take several hours. Monitoring teams typically will be dispatched into the emergency planning zone (EPZ) within an hour of ter initiation of a severe release. This subject is elaborated upon in Vol. 4 of this manual. 1.6 MAJOR POINTS j Slide 62 The major points covered in this section are summarized as follows:

• Accidents or incidents less grave than significant core damage, or the imminent threat thereof, would not warrant l

predetermined protective actions off site. e Keeping a core cool af ter scram is fundamentally a simple ? chore; it l'avolves using decay heat to boil water at a rate of 1000 gal / min or less [1000-MW(e) plant]. 4 The release to the atmosphere of even a small fraction of the inventory of radionuclides in the core of any ' operating reector would result in significantly elevated risks of

adverse health effects off site.

1 3 4 Coro damage and a fast, direct release pathway are required I i

                       . l;                               'Si5
                       >.s e

- t 52 for induction of early fatality and injury off site.

i. ,

• There would be clear indicators in the control room of actual core damage.

l 4 Early off-site protective actions must be driven by knowledge of in plant conditions, especially conditions of-the core.. These initial actions are taken based on predetermined in plant emergency action levels. Basically, core damage should warrant that people near the plant evacuate immediately.

• It is not wise to await a major release to the atmosphere (i.e., a major containment failure) before making I

i i protective action recommendations to the public. 6 The control room staff will be IgII busy during a severe accident. I A (

• Early in a severe core-melt accident, it would be difficult, if not impossible, to make a' confident projection of off-site doses.
• Protective actions agfg the site (2 to 3 alles), if warranted at all, should be implemented in all directions--not just in the downwind direction.
• Dose projections and actual field measurements will differ considerably, even if the dose projection model is doing a good job.
• Results of various dose models may be considerably different, even if each model is using the same laputs.

53 4 For the initial stages of a severe core-melt accident (General Emergency), off-site dose projection has a secondary role, independent of initiating protective actions near the plant.

• Field-monitoring resnits would be the most accurate indicator of off-site radiological impacts and their extent, but early field-monitoring results shon1d be used with caution.

l l

Appendix A SLIDES RELATING TO VOLUME 2 0F THE SEVERE REACIOR ACCIDENT INCIDENT RESPONSE TRAINING MANUAL 55 o

APPENDIX A SLIDES RELATING TO SEVERE REACTOR ACCIDENT INCIDENT RESPONSE TRAINING MANUAL: 3 SEVERE REACTOR ACCIDENT OVERVIEW VOL.2 NUREG-1210 ORNL/TM-9271/V2 Slide 1 ,

OBJECTIVES

• Describe the dose levels required to produce early health effects
• Describe basic radiation protection criteria, the EPA and FDA PAG levels, their relationship to health effects, and how they are used ,
• Describe the amount of radioactive material that must be released to produce early health effects and to exceed the EPA PAGs at a distance of about 1 mile from the plant g
• Describe the location in the plant of the radioactive material required for early health effects, the damage required for its release, the means for detection of this damage, and the level of certainty of the detection
• Describe conditions in the control room during the initial phase of an accident, including fundamental tasks the control room staff must accomplish Slide 2

\ - - _ _ _ - _ . - __. - - - - - - - - - - - - - -

OBJECTIVES (continued)

• Describe how a knowledge of plant conditions (e.g., core temperatures) is essential in projecting consequences of an accident and why in-plant monitors cannot be relied upon entirely to estimate plant releases
• Describe the uncertainties associated with dose assessment, including those between (A) what is predicted and what may be actually measured and (B) different model estimates g
• Describe why it may be impossible to adequately project dose during the early phase of a severe accident
• Describe the role of dose projection during a severe nuclear power plant accident (i.e., general emergency) and during lesser events
• Describe why early protective actions for the population at greatest risk (those near the plant) should always be recommended in all directions near the plant and not just downwind Slide 2 (Continued)

j LIGHT WATER REACTORS (LWRs) i l

• Pressurized water reactors (PWR) (70 +)
• Boiling water reactors (BWR) (20+)
• Major release from fue! possible in minutes following the loss of systems designed to keep the core cool 1

HIGH-TEMPERATURE GAS-COOLED REACTOR (HTGR) (only 1) j

• Hours before a major release would be possible following the loss j of systems needed to keep the core cool Slide 3

BASIC SAFETY PROBLEMS Reactor operation creates

• Radioactive fission products that must be contained a Decay heat that must be removed or else it could drive the fission products into the environment g Slide 4

7 CURIE Measure of radioactivity Not directly related to biological damage - dose REM Measure of potential biological damage Unit of dose Slide 5

SUMMARY

OF VOLATILITY OF RADIOACTIVE MATERIAL IN LARGE [3300-MW(T)] LWR CORE ' inventory Volatility (in millions of curies) Very volatile-100% release from fuel possible Xenon (Xe) and krypton (Kr)-noble gases 400 lodine (1) 750 g 15 l Cesium (Cs) Moderately volatile-1-20% release a from fuel possible Tellurium (Te) 200 I Strontium (Sr) and barium (Br) 370 Not volatile-small percent release possible Other 3500 Slide 6

l . I i 7 5 2

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CONTAMINATION Slide 8

HEALTH EFFECTS FROM RELEASES OF RADIOACTIVE MATERIAL Effects Dose Required I Early injuries 50-100 rem (whole body or bone) Early deaths 200-600 rem (whole body or bone) I Latent health effects 5000 person-rema l aCollective dose (not individual dose). Slide 9 L- -- - - - - - - - - - - - - - - - - - - - - -- - - - - - - - - - - - -

i i i EPA AND FDA PAGs'

                                                          .       FDA PAGs for EPA PAGs for       Food and Agricultural Organ       Plume Exposure (rem)      Products (rem)

Whole body (bone) 1-5 0.5-5 , j Thyroid 5-25 1.5-15 0 Other body organs 1-5 0.5-5 [ j Slide 10 i

ATMOSPHERIC RELEASE (curies) UNDER POOR METEOROLOGICAL CONDITIONS 1 TO RESULT IN PAGs AT 1 MILE , No. of Curies Released Radioactive Material Pathway 15 rem-Thyroid 5 rem-Whole Body 8; lodine (131 1) Milk ingestion 2 inhalation 600 Noble gases Cloud shine 1,500,000 (Xe, Kr) i Slide 11 n - - - - _- - . - - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - -

TYPICAL RELEASES l Curies Released i Sources Noble Gases Radioiodine ! BWR Annual total, Ci 100,000 1 i l Release rate to equal 10 0.0001 , l annual total,a Ci/hr i PWR i Annual total, Ci 200 0.02 l Release rate to equal 0.02 0.000002 ' annual total,a Ci/hr aUsing 10,000 hr/ year. Slide .12

f i 1-HOUR RELEASE REQUIRED TO EXCEED PAGs Release (Ci/hr) EPA whole-body PAG via cloud shine 1,500,000 a l FDA thyroid dose via inhalation 600 Slide 13

SYSTEM FAILURES REQUIRED (BUT NOT NECESSARILY SUFFICIENT) FOR VARIOUS CONSEQUENCES AT 1 MILE ] is the failure necessary to l System Cause early Exceed whole- Exceed l failure health effects? body PAGs? thyroid PAGs? l l Core fuel Yes Yes Yes Spent fuel No Occasionallya Yes Primary coolant No No Yes l Waste tank No No Yesb i I The fuel is relatively cool, so the driving force is small. b Milk only. l l Slide 14 i I

72

 . BARRIERS TO RELEASE OF RADIOACTIVE MATERIAL FOR A PWR

2. FUEL PELLET CLADDING FUEL PIN (~ 50,000 i CORE) g FISSION GAS PLENUM

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73 BARRIERS TO RELEASE OF RADIOACTIVE MATERIAL FOR A PWR (continued) STEAM GENERATOR TUBES, PART OF BARRIER 3

4. CONTAINMENT STRUCTURE

3. REACTOR COOLANT m BOUNDARY VESSEL O ft 9

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74 POSSIBLE CORE DAMAGE STATES AS CORE TEMPERATURE INCREASES

                 .. 5100*F      MELTING OF FUEL PELLETS (UO2 )
               ,        (3088 K)

(RELEASE OF ALL VOLATILE FISSION PRODUCTS FROM FUEL 60 min - --

                  . 3600 F      FORMATION OF " LIQUEFIED FUEL"- FUEL DISSOLVES IN n           (2255 K)    MELTED COMPONENTS EEE a5 gae TIME   5     o         VERY RAPID RELEASE OF IODINE, CESIUM, AND NOBLE GASES FROM FUEL 200*F     STEAM - ZlRCALLOY REACTION RELEASES H2 UNCO ERY           --

4 PWR CORE

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            =

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g. 1400 F i 15 min -

E. 1200 F CLADDING COLLAPSE POSSIBLE O (920 K) o a

                     - 600* F       NORMAL OPERATING TEMPERATURE Slide 17

CRITICAL SAFETY FUNCTIONS (EMERGENCY OR NOT) 1

• Control reactivity
• Maintain coolant level (water)
• Provide coolant makeup (water) y
• Remove heat i

1 i i l l Slide 18 1s l i

76 OHNL-OWG 85-17686R INJECTED REACTOR VESSEL FLOW REQUIRED TO REMOVE DECAY HEAT BY BOILING WATER [3300-MW(T)] AND BWR SYSTEMS AVAILABLE TO PROVIDE THE FLOW

                                     **       ~

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CONDENSATE BOOSTER FLOW RATE 000 - (3 PUMPS) 20.000 LPCI FLOW RATE (3 PUMPS) 10,000 - - LPCS (1 PUMP) HPCS (LOW PRESSURE) l 7 ~y NOTE CHANGE OF SCALE

                              $ 1,000 -                                   HPCS FLOW RATE                   _

8 (1 FUMP AT OPERATING PRESSURE)

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gmy o 700 G RCIC FLOW RATE (1 PUMP) g s00 _ _ 500 - , _ , , , , _ , _ _ _ _ _ _ _ _ _ _ _ _ _ 400 - WATER NEEDED TO

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• CALCULATED ASSUM1NG CONSTANTREACTOR VESSEL PRESSURE AND 100% INITIAL POWER DECAY HEAT PER 1979 ANS STD.WITH ACTINIDE DECAY). COOLANT INJECTED AT 305 K (90*F) AND REMOVED AS DRY SATURATED STEAM.

LPCI - LOW PRESSURE COOLANT INJECTION RCIC - REACTOR CORE ISOLATION COOL ING LPCS - LOW #RESSURE CORE SPRAY HPCS - HIGHPRESSURE CORE SPRAY CADHS - CONTROL ROD DRIVE HYDRAULIC SYSTEM SLC - STANDBY LIQUID CONTROL Slide 19 1

l 77 SIMPLIFIED DIAGRAM OF TYPICAL BWR REACTOR VESSEL WATER INJECTION SOURCES NORMAL SOURCE

                                                                                                             ------------ ALTERNATE S0WCE 4      .,

FEEDWATER UNE  !  !

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• Y WECHANISM AS$tueuES NORMAL - 60 gom O

                                                                                          \\\\\\\\\\ SCRAu - no gpm CR0                            RHR HYDRAUUC                     (4 PUMPS)

PUMP 40,000 gpm RCic l CORE SPRAY 600 gpm ' CONDENSATE "*""~~~~~~~,,) (4 PUwS) 12.500 gpm STORAGE tw ................ .... . i y^ ,, j __

 /                            ES SugE SUPPRESS \03                                                                     '

CRD - CONTROL ROD DRIVE HPCI - HIGH PRESSURE COOLANT INJECTION RCIC - REACTOR CORE ISOLATION COOLINO RHR - RESIDUAL HEAT REMOVAL Slide 20

EXAMPLES OF INSTRUMENTATION AND l INFORMATION AVAILABLE FOR DETERMINATION  ; i OF THE STATUS OF THE FUEL (CORE) l I instrumentation and Information Type of Reactor l Immediately available in control room Core temperatures-thermocouples PWR l Core-exit temperature levels BWR Containment radiation level BWR, PWR = Radiation levels from condenser / air ejector BWR, PWR , Neutron fluxes in core BWR, PWR ' Available after several hours Concentration or radiation ievel BWR, PWR in circulating reactor coolant i Analysis of primary coolant-gamma spectrum BWR, PWR ! Containment hydrogen level (from samples) BWR, PWR 1 i Slide 21 r

l l l PLANT CONDITIONS REQUIRED FOR EARLY HEALTH EFFECTS OFF SITE l

• Core damage and

• Fast-direct release to atmosphere Failure of safety systems (e.g., containment sprays)
• l Early major containment failure Early major release (less than 4-6 hr) l Slide 22 I

l .

MAJOR SEQUENCES IN A SEVERE LWR ACCIDENT SCENARIO CORE COOLING  ; CORE MELTS  : CONTAINMENT FAILS FAILS CONTAINMENT CONTAINMENT

CORE MELTS COOLING FAILS FAILS ACCIDENT RELEASE OF INITIATING RADIOACTIVE EVENT MATERIALS 8 CORE COOLING CORE MELTS CONTAIN ENT

                                                                                                     ^'

FALLS COOLING FAILS SPECIAL DESIGN CASES, CONTAINMENT BYPASSED, AND

CORE MELTS CORE COOLING SYSTEM FAILED BY INITIATING EVENT Slide 23

81 EVENT TREE FOR SEVERE ACCIDENT CONSEQUENCES CORE STATUS CONTAINMENT STATUS CONSEQUENCES r v 1 CORE EARLY EARLY LATE (24 hr) l

      ,UNCOVERY l CORE MELT             TOTAL       MAJOR                , CONTAIN- l (GAP                                                                      CONTAIN-    CONTAIN- l MENT                    l RE LE ASE                                          I MENT        MENT                I FAILURE      1 g FROM FUEL                                                                   i FAILURE     LEAKAGE l                          l l PINS)                                                   I                  II (BYPASS)                        I              i YES

1. EARLY HEALTH EFFECTS LIKELY YES YES

2. EARLY HEALTH EFFECTS POSSIB LE NO

3. EARLY HEALTH EFFECTS YES UNLIKE LY NO YES NO 4. NO EARLY JL HEALTH EFFECTS .

5. EARLY HEALTH EFFECTS VERY UN LIKE LY NO NO

6. EARLY HEALTH .

EF FECTS VERY  ! UN LtK E LY g E NO

7. NO E ARLY HEALTH EFFECTS  !

NO ( NO 8. NO E ARLY HE ALTH EFFECTS NO _

9. NO EARLY HEALTH LFFECTS S!ide 24

l s. e i CONTAINMENT LOADS WORKING

                                 GROUP CONCLUSION

• Early containment failure is unlikely during most core melt g accident sequences until several hours after core damage l

Slide 25

TYPICAL CONTAINMENT DESIGN PARAMETERS Allowable Design Leak Rate Pressure Estimated Failure Type of Plant (vol % / day) (psig) Pressure (psig) PWR, large dry 0.1 47 134 PWR, subatmospheric 0.1 45 119 PWR, ice condenser 0.25 12 50 e ! BWR Mark l 0.5 62 117 BWR Mark ll 0.5 55 140 BWR Mark Ill 0.4 15 60 I Slide 26

SOME IMPORTANT CONTRIBUTORS TO CONTAINMENT FAILURE

• Most important accidents Overpressurization several hours after core damage Hydrogen burn
• Some failures before core damage or unpredictable failures Isolation without scram Failure to isolate Containment bypass '

Interfacing system loss-of-coolant accident (LOCA) Slide 27

REACTOR CONTROL ROOM ENVIRONMENT DURING A SEVERE ACCIDENT o Plant beyond design o Many alarms may be sounding simultaneously ! o instruments may respond erroneously or go off scale o ' Reactor operators perform many tasks almost simultaneously (first hour) Identify the basic cause of the problem implement emergency operation procedures to control / correct problem Direct damage control / monitoring teams a Classify the event Possibly implement on-site emergency action (e.g., evaluation and accountability) Activate the site emergency organization Determine and recommend protective actions Notify the NRC and many other agencies ! Man the emergency notification system (ENS) and provide information to NRC Slide 28

MOST IMPORTANT ASPECTS OF PREDICTING OFF-SITE CONSEQUENCES l

• Little risk if core is not damaged a

Core damage is the result of major equipment or human failures

• No assurance that further failures are not possible

                                                                                  ?

• No assurance that release will be prevented
• Estimate 1 in 10 core damage accidents will result in major releases l

I Slide 29 l

t BASIS FOR INITIAL (EARLY) SEVERE ACCIDENT PROTECTIVE ACTION DECISIONS

• Plant conditions (general emergency action levels)~
• Precalculated dose projections
• Not real time dose projections Slide 30

l (.

                                                            ,         .w WHY PROTECTIVE ACTIONS ARE BASED l                                                    ON CORE CONDITIONS

• High risk of major release for core damage accidents
• Ease of detection of core damage
• Difficulty in projecting containment failure  :
• Effective protective actions are taken before containment failure Slide 31

e. . .. . .

l 1 THE PROBLEM l

• Protective actions for severe accidents (core damage) must be initiated before a major release to be effective
• Cannot miss core damage from control room
• Cannot confidently dismiss or predict major release (plant is well g beyond design)

Slide 32 e

n _ . _ _ _ . - _ l l ? EXAMPLE OF CONTROL ROOM INDICATION BEFORE A RELEASE N

                                                                                                                              @ RELEASE OF RADIOACTIVITY INTO ATMOSPHERE RADIOACTIVITY l

FUEL ( RELEASED FROM PRIMARY SYSTEM gCLADDING

                                       @ COOLANT                                            FAILS /

INJECTION g FAILS

• J b

                                      - ~ _;,;                                         -_

M a,,LOCA

                               ~~
                                                @ WATER                                                            i LEVEL                                                      "c DROPS                                                      ;).?

REACTOR CORE PRIMARY SYSTEM (FIRST AND SECOND (THIRD BARRIER) BARRIERS) CONTAINMENT (FOURTH BARRIER) { [XAMPLE LOSS OF COOLANT ACCIDENT (LOCA) CONTROL ROOM INDICATORS I A SYSTEM FAILURE -START OF ACCIDENT PRESSURE, TEMPERATURE f 8 ESF FAILURE FLOW, TEMPE RATURE C CSF FAILURE CORE TEMPERATURE D BARRIER FAILURE RADIATION, TEMPERATURE E MOVEMENT OF RADIOACTIVE MATERIAL RADIATION F RELEASE TO ATMOSPHERE (CONTAINMENT LEAKAGE) OFF-SITE DOSE RATE (e.g., AT GUARD SHACK) Slide 33

EMERGENCY ACTION LEVELS (EALs)

• Observables (e.g., control room instrument readings) l
• Indicate level of emergency (classification)

Activate on-site emergency organization (alert) Activate off-site organizations (site area emergency) Basis for off-site protective actions (general emergency) s ; l

• For a full range of accidents
• Most severe accidents should be classified hours (2 or more) before a major release Slide 34 i

EVENTS LEADING TO EVACUATION Core Damage ' u Potential for Early Health Effects Off Site

                                                                             ~

u Classify as General Emergency e v immediate Recommendation of Protective Actions u in General, Evacuation of Area Near Plant Slide 35

EXAMPLE OF NONRADIOLOGICAL INDICATORS OF IMMINENT SEVERE CORE DAMAGE REACTOR PRESSURE GREATER THAN SAFETY VALVE SET POINT REACTOR REMAINS _,

                       '^ ^                                    \         R APIDLY INCREASING ONTAMENT PRESSURE (F i e to co t of reactivity) 0 0                       R APIDLY INCREASING CONTAINMENT GENERAL EMERGENCY, ERATURE INITI ATING CONDITIONS
                                       *OBSERVABLES PERSIST FOR 15 min.

TRANSIENT REQUIRING OPERATION OF SHUTDOWN SYSTEMSWITH FAILURE TO SCRAM

Slide 36 F

z EXAMPLE OF RADIOLOGICAL INDICATORS OF SIGNIFICANT CORE DAMAGE

• CONTAINMENT DOSE RATE OUTSIDE .

MONITOR READING OF CONTAINMENT

                                         >2000 R/hr**                                                   HATCH >10 R/hr**

OR g U GENERAL EMERGENCY

• Presumed to be corroborated (i.e., not merely a faulty reading).

                                         *

• Site-specific parameters - should be precalculated. ,

l l Slide 37 l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - - _ . . _ _ _ - _ _ _.

i MAJOR INDICATORS FOR CORE DAMAGE ASSESSMENT

• Gross relationship of core damage and Radiation monitors (water, air)

Water level, flow rates ! Thermocouple readings I Pressure and pressure differences (head) Power levels (AC, DC, neutronics) Slide 38

96 EXAMPLE OF.lMPORTANCE OF TEMPERATURE AND PRESSURE MEASUREMENTS PERCENT OF FUEL RODS WITH RUPTURED CLADDING vs MAXIMUM CORE EXIT THERMOCOUPLE TEMPERATURE WHEN THE PRESSURE USE CURVE LABELED IS WITH TEMPERATURE

                                   < 100 psia                 1200 F
                                   < 1200 psia                1500 F
                                   < 1650 psia                1800*F 100  -

E g 1200 F Q 80 - d CLADDING RUPTURE o TEMPERATURE E R x 60 - ( 3 1500 F 8 @ 40 - d 0 I 1800 F O Z 20 0 Fu a. l I I I I o 1200 1400 1600 1800 2000 2200 2400 MAXIMUM CORE EXIT THERMOCOUPLE TEMPERATURE (*F) Slide 39

l TYPICAL CALCULATED CONTAINMENT RADIATION ESTIMATES (R/hr) Assumptions 100% Gap 1% Core 10% Core 100% Core Type of Plant Activity Activity Activity Activity Reactor A 400 4,000 40,000 s. PWR , B 1,900 1,900 18,000 200,000 BWR C 165,000 PWR D 5,000 50,000 500,000 BWR L E 80,000 150 70,000 70,000 PWR F 100,000 10,000 100,000 1,000,000 BWR Slide 40 -

                                            - mm - - mm

I 1 l l REAL-TIME SEVERE ACCIDENT DOSE-PROJECTING PROBLEMS FOR PROTECTIVE ACTION DECISION MAKING

• May be too late for effective protective actions off site
• For a major release, evacuation must start before or soon after g the release to be most effective
• Real-tirne dose projection would not be possible before a release
• Timing of a major release, if any, would be extremely difficult to predict in real time Slide 41 l

STEPS IN PREDICTING DOSE

1. Predicting quantity and of timing release (source term)

2. Predicting movement of the plume

3. Predicting dose from plume

4. Predicting dose vs distance and time 1

Slide 42

ORNL-DWG 86-16130 STEPS IN PREDICTING DOSE

                                                     ~

1 .

                                                                         .~....

SOURCE TERM c e-- PROJECTION  :

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xy j/ %PROJECTION p 7 Slide 43

i PREDICTING SEVERE-ACCIDENT SOURCE TERM

• Effluent radiation monitors.

  - Real time: noble gas only
  - lodine and other particulate samples:     E several hours

• Monitors are probably bypassed

  - Example: containment failure Slide 44

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• Containment response is most important
• Generalizations are inappropriate s
• Reductions in risk may not be found in all cases
• There are great uncertainties in estimating source terms l

Slide 46

SEVERE ACCIDENT SOURCE TERM ESTIMATION

• Early in a severe accident, it will be very a

                                                   ^

difficult to predict the " source term" l l l Slide 47 i

105 l EXAMPLE FROM NUREG-1062, DOSE CALCULATIONS FOR SEVERE LWR ACCIDENT SCENARIOS PWR #4 CASE 4 STABILITY CLA$S: D RAIN: NO WIND SPEED: 6 mph SHELTERING: NONE WHOLE BODY DOSE 24 hr 104 i 103 i i i a 7-day O CLOUD SHINE O 24 hour a GROUND SHINE 03 - 0 4-hour O WHAMD 102 _ , 2 10 _ _ h ~ 101 - - 0 10 1 - - 0 100 - 100 - - 10-1 I ' I I 10-1 0 10 10 1 102 103 100 10 1 102 j03 DISTANCE (miles) DISTANCE (miles) THYROID DOSE 245r 10 4 10 4 y y i y A 7-day O CLOU; SHINE 103 - 0 24-hour - 103 - a GROUND SHINE - 0 4 hour O INHALED

 $ 102     _                          _                                 j                             192                             _                         _

i o S g O 101 - - 0 10 1 100 - - 100 _ _ 1 10-1 1 10-1 1 1 0 2 10 0 jo3 1 10 10 10 103 101 102 DISTANCE (triles) DISTANCE (miles) Slide 48

s 106 ACTUAL VS MODEL ATMOSPHERIC PLUME t;7%.Mrw n. '

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MODEL Slide 49

107 ONE-HOUR SURFACE DOSES PREDICTED BY (A) GAUSSIAN PLUME MODEL, (B) PUFF-TRAJECTORY MODEL,(C) COMPLEX NUMERICAL MODEL, AND (D) DOSES ACTUALLY OBSERVED i i l 1 il i l l il l MAXIMUM DOSE MAXIMUM DOSE 150.1 212.3 i _ h,f i i l i i l i i I i (A) (B) i l l i i l l l l l MAXIMUM DOSE MAXIMUM DOSE 52.6 13.1 *

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h _ _ i. i I i i I i i I i i I i (C) (D) Slide 50

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Siide 52

DIFFERENCES IN DOSE MODEL RESULTS

• The results of various dose models differ because of

   ~
                                               - Different methods
                                               - Different assumptions
                                               - Different objectives                                                          -

s

• Expect a 10- to 100-fold spread
• Knowledge .of local meteorological trends and conditions is most important Slide 53

ORN L-DWG 85-17700R RELATIONSHIP BETWEEN ACTUAL PLUME AND MODEL PROJECTIONS ACTUAL PLUME MEANDER WHICH IS AVERAGED OVER 15-30 min TO OBTAIN AVERAGE CONCENTRATIONS MODEL AVERAGE CONCENTRATION p %:.7,p.y:.%b. y; I 'h .

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                                                                                                        '              ";i" MONITOR LOCATIONS (BOTH ARE IN PLUME ACCORDING TO MODEL)

Slide 54

l l ESTIMATED RANGE OF UNCERTAINTY BETWEEN PROJECTED AND ACTUAL OFF-SITE DOSE l FOR A SEVERE ACCIDENT (CORE MELT) Uncertainty Factora Element At Best Most Likely Near Worst Source term (event and sequence) 5 100-1,000 100,000 Dispersion y Diffusion (concentration) 2 5 10 Transport (direction) 22 45 180 Transport (rate) 1 2 10 (low wind speed) Dosimetry 3 4 5 Overall (dose and 10, 100-10,000, 100,000, direction) 22 45' 180 aThese estimates are for an averaged dose at a location (e.g., 15-30 min), not for a specific or single monitor reading. l Slide 55 l

113 HOURLY WIND VECTOR AT TMI-2 ON MARCH 28,1979 MARCH 28,1979 HOURLY WIND VECTOR' TMi2 N f" 1609  % f* 1 409 1300

                                                                    '1200 09 1100  )
                                                                                       '0200 S mph 0400 10 mph 9

03 15 mph S

                ' Arrows indicate direction toward which the on site wind was blowing at the focal time indicated. Circ'cs represent varying wind speeds.

Slide 56

TMI-2 PLANT CONDITIONS AT 9:00 A.M.

• Clear indications of core damage Containment monitor - 6000 R/hr Thermocouple - 2000 F (Cladding failure at 1200 F)
• Plant beyond analyzed conditions
• Large amount of radioactive material in containment (only 1 barrier remains)

Problems

• No protective action taken because projected doses off site did not
• exceed PAGs
• Action not recommended on plant (core) conditions l *300 wind shift in 3 hr Findings
• Initiate early protective actions off site in all directions
• Initiate early protective actions off site on detection of core damage Slide 57 1 _ _

1 l l THE ROLE OF DOSE ASSESSMENT DURING VARIOUS PHASES OF NUCLEAR POWER PLANT ACCIDENTS

• Initial phase of general emergency t

                       - Determine areas to be monitored first
                       - Compare consequences of various plant response options
                       - Determine maximum distance requiring further protective E-actions

• Site area emergency and intermediate phase of a general i emergency I - Assess additional areas for implementation of protective i actions

! - Provide feedback on source term based on monitoring results } Slide 58

FIELD-MONITORING ROLE IN PROTECTIVE ACTION DECISION MAKING AFTER A MAJOR RELEASE

• After major release Identify areas in which early. protective action decisions should be modified '

Can provide indication of actual off-site radiological impacts Slide 59

s t l FIELD-MONITORING PROBLEMS l ! - 1. Large variations-requires a large number of monitoring results to characterize a release

2. May not be in plume-'over teams-meanders or loops around team

3. Concentrations could change quickly i 4. Release could be very complex Ei I

Slide 60

   ~

J ONE-HOUR SURFACE DOSES PREDICTED BY DOSES ACTUALLY OBSERVED i  ; i i i i _ MAXIMUM DOSE _ 13.1 _ ... e - l hh::... .-

        ~
                                               ~

i I 3 , yg i

       ~
                         .. O s ~ f         _
                                                           =
       -                      /               -

i l i i I i Slide 61 l 1

I f l MAJOR POINTS i l

• Early, predetermined protective actions off site unwarranted for less than general emergencies
• Core can be kept cool by using decay heat to boil water (1000 gal / min)
• Release of only small fraction of core inventory could result in health effects s
• Core damage and fast, direct release required for severe health effects

 ,

• Can't miss core damage-can't predict release
• Early off-site protective actions based on core damage-don't wait for dose assessments-use predetermined EALs
• Core damage-evacuate near the site if possible
• Control room very busy during an emergency Slide 62

MAJOR POINTS (continued)

• During a major accident, accurate early dose projection may be impossible
• Early protection could be highly uncertain with the major problem being source term estimates
• Considerable differences will occur between field measurements and model predictions
• Considerable differences can exist between dose projection models even si if came inputs are used
• Dose assessment has a secondary role during early stages of a major accident-use dose assessment after initial early actions are taken based on core conditions
• Take protective actions in all directions near the plant, when warranted at all
• Field monitoring is most accurate indication of extent of off, site radiological impacts-but early results should be used with caution Slide 62 (Continued)

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    %'">'2,                          BIBUOIRAPHIC DATA CHEET                                                                                             N                   G-1210, Volume 2

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2. TITLE ANO SUST4TLt Pilot Program: NRC Severt: Reactor Accident j Incident Response Training Manual , , , , , , , , , , , , , , , , ,

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                                                                                                                                          ,                                                         i October                                              1986
    .2uf oais.

T. J. McKenn_, J. A, Martin, Jr., C. W. Miller, I- . oAn E, oaf .uuto L. M. Hively',,R. W. Sharpe, J.G. Guiitter, R. M. Wa ins February woNTa l vtAa 3 37

7. FimFOnuiNG ORGAN #2 Afiose manet ANo uAiLING Acon tSS (#sica w ee de Cees 8 Pfe0JECTIT A5EttwomE UNIT Nuustm

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t la SPONSOmiNG OmGANi2 Af SON N.wt AND M A4 LING AooRES$ ff=siese to Ceser lia TYPE OF REPORT Office of Inspection arid. Enforcement Division of Emergency Preparedness - . Pt a oo Cov t a t o <,-N, e .-, and Engineering Response, U.S. Nuclear Regulatory Commission e Washington, D.C. 20555 . [ 13 SUPPLtWENTAnv 40Tts 13 ASSTRACT (200 swores se sense

                                                                                              \          .

i This is one in a series of volumesjtjiat collectively provide for the U.S. Nuclear Regulatory Commission (NRC) emergen y response personnel the necessary background information for an adequate respon lto severe reactor accidents. The volumes in the series are: ' \ o Volume 1 -- Overview and Summ _ y of ajor Points o Volume 2 -- Severe Reactor Accident Overview o Volume 3 -- Response of Licerjtee and State and Local Officials o Volume 4 -- Public Protectivf Actions -Y Predetermined Criteria and Initial Actions o Volume 5 -- U.S. Nuclear Reclulatory Commission Response Each volume serves, respectiv ly, as the text (for a course of instruction-in a series of courses for NRC response rsonnel. These (n.aterials do not provide guidance or license requirements for NRC licensees or stategor local response organizations. Each volume is accompanied an appendix of slides that can be used to present this material. The slides re called out in the\ ext. 5.'

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