ML20212J632
| ML20212J632 | |
| 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-V01, NUREG-1210-V1, NUDOCS 8703090078 | |
| Download: ML20212J632 (111) | |
Text
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._ l NUREG-1210 i Vol.1 1
l Pilot Program: NRC Severe Reactor Accident Incident Response Training Manual .
1 Overview and Summary of Major Points 1
U.S. Nuclear, Regulatory Commission Office of Inspection and Enforcement T. J. McKenna, J. A. Martin, Jr., C. W. Miller, L. M. Hively, ;
R. W. Sharpe, J. G. Giitter, R. M. Watkins
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- $ $ 87022e pyg I
NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
- 1. The NRC Public Document Room,1717 H Street, N.W.
Washington, DC 20555
- 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082
- 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Docu-ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bu!Ietins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers;and applicant and licensee documents and correspondence.
The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federal Regulations, and Nuclear Regulatory Commission issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commission.
Decuments available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.
Documents such as theses, dissertations, foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com-mission, Washington, DC 20555.
Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be s
purchased from the originating organization or, if they are American National Standards, from the American National Standards institute,1430 Broadway, New York, NY 10018.
NUREG-1210 Vol.1 1
Pilot Program: NRC Severe Reactor Accident Incident Response Training Manual i
Ov:rview and Summary of Major Points
_ _ _ . _ _ - _ _ _ _ . _ _ . . _ _ . _ _ _ _ _ _ . _ - . _ _ _ __. __ _ ____~~
M:nuscript Completed: July 1986 )
l Dite Published: February 1987 T. . . McKenna, J. A. Martin, Jr., C. W. Miller *, L. M. Hively*,
R. W. Sharpe*, J. G. Giitter, R. M. Watkins*
'o;k Ridge National Laboratory Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission
, W=hington, DC 20666 l ,+'* *'%
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FOREWORD 1
l 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.
h -
Edward . Jordan, Director ,
! Division f Emergency Preparedness i and Engineering Response
! Office of Inspection and Enforcement i
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l PREFACE Overview and S----rv of Maior Points is the first in a series of 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 assessment. Other volumes in the series are
, 4 Volume 2 -- Severe Reactor Accident Ove rview e Volume 3 -- Resnonne of Licensee and State and Local Officials e Volume 4 -- Pubile Protect ive Actions--Predetermined Criteria j
and Initial Actions 6 Volume 5 -- U.S. Nuclear Reaulatory Commission Reasonne Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do R21 provide guidance or license requirements for NRC licensees. The volumes have been organized into these training modules to accommodate the scheduling and duty needs of participating NRC staf f. .
Each volume is accompanied by an appendix of slides that can be used to present this material. The alldes are called out in the text.
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G)NTHE I
s 233R PREFA m . . . . . . . . . . . . . . ,. .'. . . . . . . . . . . . . . . . 111
(
' LIST OF FIGURES . ...,......... . . . . . . . . . . . . . . vil
~
LIST OF TABLES . .. .....?. . ....... . . . . . . . . . . . ix i
ch; I,IST OF AMWYMS AND INITIALISWFOR VOLUES 1-5 . . . . . . . . . . xi e i
,a p**
- 1. OVERVIEW .... . .. l . .'. . . . . . . . . . . . . . . . . . 1 V 1.1 GBJECTIVES . .. . ,4.................... 1
1.2 INTRODUCTION
'IU 1HE SEVERE REACTOR ACCIIENT INCIDENT RESPONSE 1 RAINING' MANUAL . . . . . . . . . . . . . 1 h,y, , 2. RESPONSE M0 GRAM GUIEELINES . . . . . . . . . '. . . . . . . . . . . 3
- ./ .. _ 2 .1 FIRST GUIDELINE: UNDERSTAND 11E PHASES OF AN :
L' , EERGENCY RESPONSE . ..........'..... . . . . . 3
/ 2.2 SECOND GUIDELINE: UNDERSTAND THE SEVERITY AND
, SG)PE OF AN EERGENCY/ . ........ . . . . . . . . . . . 4 l d 'c 2.3 HIRD GUIDELINE: UlOERSTAND RESPONSIBILITIES
! ', OF THE NRC, LIGNSEE, AND OFF-SITE OFFICIALS . . . . . . . . 6 :
, 2.4 FOURE GUIDELINE: . UNDERSTAND THAT EXTENSIVE i '-
- EERGENCY PREPLANNING HAS BEEN ACCDNFLISHED . . . . . . . . S I
2.5' FIF11 GUIDELINE: UleERSTAND HAT 1HE &lC IS PART
, . - 0F A VERY LA'tGE RESPONSE ORGANIZATION AND UNDERSTAND YOUR PART ,1N IT . . . . . . . . . . . . . . . . . 9 i
'e -
SII11 GUIDELINE: UPSERSTAND THAT NRC PUNCTIONS CAN
] . ' 2 .6
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. BEST BE MRPORED BY REGIONAL PERSWNEL Ar OR NEAR 9
',; THE SITE . ...... . ................. .
i , .2 .7 SEVENE GUIIELINE: Ul@ERSTAND THE IMPORTANT ASPECTS OF SEVERE ACCIDENTE AND THE DIFFIQJLTIES IN HEIR 10
~
ASSESSENT , ... . ..... . . . . . . . . . . . . . . .
2.8 EIGH1H GUIDELINE: UPE)ERS* SAND THAT NRC ASSESSENTS
/ , SHOULD ALWAYS BE CDNPARED' AND DISGSSED W13 OIEER !
. RESPONSE ORGANIZATIONS . . . . . . . . . . . . . . . . . . 11 C 3. MAIM POINTE . .... . ......... . . . . . . . . . . . 13
! j" APPENDIX A. BIBLIOGRAPHY ICR VG UES 1-5 . . . . . . . . . . . . . . 15
,AF$NDIX B. SCIENTIFIC NOTATION' . . . . . . . . . . . . . . . . . . 25 i ',
A N CIDIX C. E11 TIC UNITS . . .... . . . . . . . . . . . . . . . . 31
%/
f AFMNDIX D. L1Ultr WATER REAcr0R FUEEAR POWER PLANT fi- -
DESIGN PEATURES. ......... . . . . . . . . . . . 35 i
- 1,, . APPENDIX E. GLOSSARY OF TERNS PDR V0 LUES 1-5 . . . . . . . . . . . 59 (J % .
APPENDIX F. EIDES RFLATING *ID VG.UE 1 0F THE SEVERE REACTOR l ,.,.,
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ACCIDENT INCIDENT RESPONSE 1 RAINING MANUAL . 75
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LIST OF FIGURES Flaure fait B.1 Example of scientific notation: popuistion (p) inside 29 circle of radius x (miles) . . . . . . . . . . . . . . . . . . .
D.1 Pressurized water reactor nuclear power plant s coolant loop, simplified diagram . . . . . . . . . . . . . . . . 38 Typical pressurized water res.: tor layout . . . . . . . . . . . . 39
[ D2 Spent ^ fuel pool enclosure. . . . . . . . . . . . . . . . . . . . 42 D.3 D.4 Barriers to release of radioactive material for a 44 pr essurized water reactor. . . . . . . . . . . . . . . . . . . .
il D.5 Typical (a) assembly and (b) fuel pin (second fission product barrier) . . . . . . . . . . . . . . . . . . . . . . . . '45
', D.6 Typical pressurized water reactor primary system (third fission product barrier). . . . . . . . . . . . . . . . . 46
/
i D7 Relationship among critical saf ety functions, maintaining fission product barriers, and preventing a release . . . . . . . 48 Large high pressure containment. . . . . . . . . . . . . . . . . 50 D.8 Ice condenser de sign . . . . . . . . . . . . . . . . . . . . . . 51 D.9 D.10 Bolling water reactor nuclear steam supply system. . . . . . . . 53 D.11 Boiling water reactor, Mark I containment de sign . . . . . . . . 55 D.12 Boiling water reactor, Mark II containment de sign. . . . . . . .. 56 D.13 Boiling water reactor, Mark III containment de sign . . . . . . . 57 i
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LIST OF TABIES Table Eszt 1.1 bergency Class response. ..... . . . . . . . . . . . . . . 5 1 l
1.2 Emergency class descriptions . . . . . . . . . . . . . . . . . 7 l B.1 International System of Units (SI) prefixes . . . . . . . . . . 30 4
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LIST OF AMONYMS AND INITIALISM FOR VG.UES 1-5 ALARA As low as reasonably achievable AM Aerial Measurement s System (s)
ARAC Atmospheric Release Advisory Capability ASC Administrative Support Coordinator AST Administrative Support Team BT Base Team (NRC Regional Office)
BWR Boiling Water Reactor CDPA Civil Def ense Preparedne ss Agency CFA Cognizant Federal Agency CPR Code of Federal Regulations G, Congressional Liaison MD Control rod drive MIIIS Control rod drive hydraulic system GF Critical Saf ety Function DBA De sign Basis Accident DOC Department of Commerce, U. S.
D0D Department of Def ense, U. S.
DOE Department of Energy, U.S.
DOI Department of Interior, U.S.
DOT Department of Transportation, U.S.
DSO Director of Site Operations EAL Emergency Action Level ECCS Bsergency Core Cooling System EDO Executive Director of Operations ENS Bsergency Notification System EO Emersency Officer EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protectica Agency, U.S.
EPRI Electrical Power Research Institute EPZ Energency Planning Zone ERC Emergency Response Coordinator ERM Emergency Response knager ERO Faergency Response Organiz ation ERT Emergency Response Team (FEMA organization)
ESF Engineered Saf ety Feature EST Emergency Support Tesa (FEMA organization)
ET NRC Ezecutive Team ETA Estimated time of arrival FBI Federal Bureau of Investigation PDA Food and Drug Administration, U.S.
FEMA Federal borgency Management Agency PRC Federal Response Center FRERP Federal Radiological bergency Re.ponse Plan PRMAC Federal Radiological Monitoring and Assessment Center FRMAP Federal Radiological Monitoring and Assessment Plan FFS Federal Telephone System GLC Governsent Liaison Coordinator GLM Government Liaison Manager xi
GLO Government Liaison Officer GLT Government Liaison Team HHS Health and Human Services, U.S. Department of B00 NRC Headquarters Operations Officer HPCI High pressure coolant inj ec tion HPCS' High pressure core spray HPN Health Physics Network BQ NRC Headquarters HUD Housing 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 Regional NRC Incident Response Center IRDAM Interactive Rapid Dose Assessment Model JIC Joint Information Center LC Liaison Coordinator LNO 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. I 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 Guides 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 PNT Protective Measures Team P-T Pressure-Temperature PWR Pressurized Water Reactor RA Regional Administrator RAT Radiological Assistance Team RBE Relative biological effectiveness RCIC Reactor core isolation cooling RCT Response Coordination Team RDO Regional Duty Officer R.G. Regulatory Guide 4
RHR Residual heat removal RI Resident Inspector xii s
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RM Resource Manager RO NRC Regional Office RSC Reactor Saf ety Coordinator RSM Reactor Saf ety Manager RST Reactor Safety Team SC Saf eguards/ Security Coordina tor SFU Senior FEMA Official SGC Saf eguards/ Security Coordina tor SGT Safeguards Team SI International System (of measurement)
SLC Standby liquid control SM Saf eguards/ Security Manager SO Status Officer ST Site Team S1L Site Team Leader ILD Thermoinmine scent dosimeter TMI-2 Three Mile Island-Unit 2 ISC Technical Support Center USDA U.S. Department of Agriculture WHO World Health Organization I
xiii
AGNWLEDGlENTS The authors wish to express their appreciation for the valuable )
assistance provided by the following people: Sus an R. Morris, )
Ursula F. Strong, and Malinda N. Hutchinson, for word processing i
i and ocordination; and Larry H. Wyrick and the staff of the ORNL 1
Graphic Arts Department for preparing illustrations and view graphs.
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XV
- 1. OVERVIEW l
1.1 OBJECTIVES Slides 1 and 2 Following completion of this section, the student should be able to l e describe the basic guidelines upon which the U.S. Nuclear Regulatory Commission (NRC) response to a major nuclear power reactor accident is based.
1.2 INTRODUCTION
101HE SEVERE REACIOR ACCIDENT INCIDENT RESPONSE TRAINING MANUAL Slide 3 Individual members of NRC's emergency response teams should have an appreciation of the " big picture. " Each should be aware of answers to such basic questions as:
4 Why an I here?
e What is expected of me?
e Who are all these other people?
- Do I interact with them? How?
- How is this different from my normal job?
e What is important?
e How do I scope the level of emergency?
1
- 2. RESPONSE PROGRAN GUIDELINES In this section, answers to the preceding questions are presented in the form of a se t of guideline s. These concepts are basic to the U.S. Nuclear Regulatory Commission (NRC) incident response program and should be understood by all response personnel.
Slide 4 2 .1 FIRST GUIDELINE: UNDERSTAND THE PHASES OF AN EERGI!NCY
RESPONSE
The three phases of emergency response are
(
i e Early Phase--O to 6 hr, i
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- Intermediate Phase--2 hr to days, and l
e Late (Recovery) Phase--hours to months.
l Specific chains of command, communication flows, and l actions have been predetermined for the early emergency response phase. Experience indicates that these provisions are adequate to implement an appropriate response. Furthermore, they have suf ficient flexibility to be adj usted to accommodate change s that may be required as the event situation unfolds, when the time constraints are less critical.
This training manual concentrates on the early NRC response (first few hours) to serious reactor accidents. Once NRC's emergency response organization (ERO) becomes fully activated with NRC personnel on site, and as events have been more fully analyz ed, the NRC will adapt its response to the existing situation. Actions can be planned most effectively for only the 3
4 first few hours of the response. Beyond that, tne response organization must deal with the situation as it evolves.
Slide 5 2 .2 SECOND GUIDELINE: UNDERSTAND THE SEVERITY AND S00PE OF AN EERG!!NCY One of the basic questions is: "How bad is this accident?"
To help answer this question, a standard emergency classification scheme has been established. The four emergency classe s are
- 1. Unusual Event,
- 2. Alert,
- 3. Site Area Emergency, and
- 4. General bergency.
Rose emergency classe s cover the full range of reactor accidents free relatively common events that pose no threat to irradiated fuel (Unusual Event) to actual or imminent core damage (General hersency) . In 1984, 224 Unusual Events, 8 Alerts, and no Site Area or General Emergencies were reported to the NRC Operations Center.
True General Emergencies will be very rare. [Only one has occurred to date, the nree Mile Island-Unit 2 (TMI-2) accident.] Obviously, this level of emergency indicates conditions well beyond plant design and warrants early actions to protect the pubilc.
For each class, specific response actions have been prede t ermine d. Table 1.1 provides an overview of the basic initial actions taken in response to declaration of each emergency class.
5 Table 1.1. Emergency class response Class Plant action Local and state agency action Unusual Event Provide notification Be aware Alert Mobilize plant resources Stand by*
Man centers (help for control room)
Activate Technical Support Center (TSC)
Site Area Emergency Full mobilization Mobilize Nonessential site Man emergency centers personnel evacuate and dispatch Monitoring Team Activate TSC, Operations Inform public--activate Support Center, and warning system Emergency Operations Facility Take protective actions in accordance with Protective Action Guides (PAGs) or an ad hoc Dispatch Monitoring Team basis Provide dose assessments l General Emergency Full mobilizaticu Recommend predetermined Recommend predetermined protective actions to the protective actions (within public based on 15 min) after plant conditions declaring emergency Precautionary evacuation (2 to 5 alles)
"The NRC will typically staf f its response centers and implement its emergency response plan at the appropriate level for an Alert.
6
' Slide 6 Ihnergencies are declared based on predetermined plant instrument readings [ Emergency Action Levels (EALs)] . Table 1.2 provides an overview of the plant conditions associated with each class.
The basic for declaration of each class of emergency and the resulting actions are further discussed in Vols. 2 and 3.
Slide 7 2 .3 'IBIRD GUIDELINE: UNDERSTAND RESPONSIBILITIES OF HE NRC, LIGNSEE, AND OFF-SITE OFFICIALS He licensee is responsible for mitigating the consequences of an accident and recommending protective actions to off-site of ficials, who are then responsible for providing protective action recommendations to the public, who implement them, as nece ssary or de sirabl e.
The NRC shonid not have a major role in any predetermined actions very early in an emergency response (0 to 2 hr) .
Normally, it won 1d take about 1 hr af ter notification for the NRC's emergency response teams to assemble and initiate independent assessment of the situation. In addition, the NRC has only limited knowledge of the specific plant and off-site conditions and plans. This is further complicated when you ,
consider that early NRC contact with the plant may be limited to a few telephone line s and that only limited plant data are avail abl e.
Hus , the NRC may not be in a position to effectively assess licensee or off-site actions early in an event. He NRC's primary role is to monitor the actions of the licensee and off-site of ficials. In the event that NRC observes what it believes to be inannrontiate protective actions, NRC should first discuss the difference with the licensee to establish the basis for the dif ference, ne intent is, whenever possible, to arrive at a common understanding or asse ssment and to
I 7 I Tabl e 1.2. Energency class descriptions Cl as s" Core status Radiation Unusual Event No threat to irradiated No relea se above fuel technical specifications (or annual limits)
Alert Actual (or potential for) Release is maall fraction substantial degradation of saf ety of Environnental Protection Agency (EPA) Protective Action Guides (PAGs) beyond the site boundary Site Area Msjor failures ef functions needed Release is less than Emergency for public protection EPA PAGs beyond the site boundary General Actual or Laminent core Dose may exceed Emergency de grada tion EPA PAGs
" Classifications are based on plant instrument levels (i.e. , Energency Action Level) .
8 communicate that to the State. If NRC continues to have a differing assessment and is confident of its basis, NRC should provide its independent assessment to the State. NRC involvement is not to interfere with the licensee in making immediate protective action recommendations for a rapid-breaking general emergency. These immediate actions should have been predetermined in the licensee's and State's emergency action plans; thus, if these plans are implemented appropriately, the NRC should not have an active role in these early actions.
One other principal role of the NRC is to act as the lead federal technical expert and clearinghouse for technical assessments in events at licensed nuclear facilities. This is one of its roles under the Federal Radiological Emergency Response Plan. The functions and roles of the NRC and other organizations are discussed in more detail in Vols. 3 and 5.
Slide 8 2.4 FOURTH GUIDELINE: UNDERSTAND '111AT EXTENSIVE EMERGENCY PREPLANNING HAS DEEN ACCOMPLISHED Since the Three Mile Island accident, emergency planning and preparations have been improved extensively at nuclear power plants and in areas near the plants (i.e., within the 10- and 50-alle Emergenc: Planning Zones). This includes improvements in procedure, in the plant control room used to direct operators' response, to recognize the severity of an event, and to recommend protective actions off site. The ability to monitor the course of an accident and to determine appropriate public response has also been improved. Extensive facilities such as centers and communications have been pre-positioned. It is important that these licensee, State, and local preparations be recognized by the NRC response staf f.
NRC staff must be aware that, if they question State or off-site officials about nreolanned initial actions, recommend other actions, or otherwise fall to recognize the scope of the
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9 existing preparations, implementation of preplanned actions could be delayed.
Volane 3 discusse s the more important aspects of the licensee and off-site planning and preparations for severe s ocident s.
Slide 9 2.5 FIF111 GUIDELINE: UNDERSTAND THAT ME NRC IS PART OF A VERY LARGE RESPONSE ORGANIZATION AND UNDERSTAND YOUR PART IN IT The NRC is part of a very large structured emergency response organization and communications system. In the event of a severe accident, the licensee, state and local officials, other federal agencies, national laboratories, vendors, and others will all be responding. During Three Nile Island and a recent full activation exercise, in addition to the licensee personnel, more than 1000 people were involved. Therefore, it is important that you understand not only where you fit in your organization but also what other organizations are doing.
Volumes 3 and 5 will discuss the NRC and other response orga niz ations.
Slide 10 2.6 SIIlli GUIDELINE: UNDERSTAND THAT NRC FUNCTIONS CAN BEST BE PERPOR)ED BY REGIONAL IERS(NNEL AT OR NEAR THE SITE The NRC regional staff is more familiar with the plant and off-site conditions than the headquarters (HQ) staff. In addition, the NRC will be able to perform its roles better from the site, where f ace-to-face discussions can be held and adequate information will be available. Therefore, as discussed in Vol. 5, the NRC response is directed toward two goals: (1) keeping the lead with the regional response staff, to the extent possible (and in all cases keeping the NRC lead clearly
10 designated), and (2) e stablishing an NRC presence at the site as soon as po ssible. Only during the initial activation mode, when the regional staff is on the way to the site, will headquarters have the lead NRC role. Thereaf ter, headquarters staf f will support the regional staff.
Slide 11 2 .7 SEVENTH GUIDELINE: UNDERSTAND 'lllE IMPORTANT ASPECTS OF SEVDtB ACCIDENTS AND 111E DIFFIGLTIES IN THEIR ASSESSENT The basic role of the NRC response organization is to monitor to ensure that the appropriate public protective actions are being implemented. Since prompt action is crucial for severe (core damage) accidents, these accidents and their asse ssments must be understood. Almost everyone in the NRC emergency response organization would either conduct severe accident asse ssments in real time or use the results of these a sse s sment s. Therefore, all anergency response organization members (not just the technical team members) need some understanding of these topics. Consequently, these manuals will concentrate on severe (core damage) accidents.
As a first benchmark, only actual, or clearly Laminent, severe damage to irradiated f uel (core, spent f uel storage pool) would pose a clear and imminent threat to public health and safety. Lesser events should not lead to predetermined protective actions of f site.
As a second benchmark, everyone must recognize that great dif ficulties would be involved in asse ssing severe (core molt) accident conditions and potential of f-site consequence s. In the early phase, it is not likely that anyone (NRC, licensee) could definitely predict the course of reactor conditions and off-site conseque nce s. How ev er , actual or imminent core damage should be easily recognizable. Beyond that, it may be possible to perform bounding analyse s ba sed on some "what-if" assumptions; these results would have to be used with a great deal of caution.
11 Limited assessment tools and large nacertalaties la plant conditione during severe core melt accidents womid not allow
- 1 more to be done with confidence. I Normal NBC staff analyses of accidents for licensing purposes are based on a highly structured system of guidance, eriteria, rates, and technical positions, whlek has been built os conservative, of ten unrealistic assumptions. However, 13 emersemov reasonne, it is best to use the most reallatio '
ass-----tions nossible based on the laformation at h==d.
On the po sitive side, the more serious the accident, the less detail womid be required for protective action decision
- m aking. In partienlar, serious core damego should be classified immediately as a General Ibergency. Volmses 2 and 4 discuss
- those topics la more detail.
i j Slide 12 l
j 2.8 EIGirIH GUIDELINE: UM)ERSTAND 11 TAT NRC ASSE853EIN1B ERTI,D
- ALWAYS BE MARED AND DIS (IISSED Willi UIMIR RESPONSE ORGANIZATIONS he NFC should not make totally independent asse ssments; rather, assessments should be conducted in parallel with those of the licensee (and possibly those of off-site officials) . De NRC assessments shon1d be compared and discussed with the licensee and appropriate officials. Dese organisations have j better data and understandings of the plant and off-site
- conditions. His is true not only for technical assessments but I
also for information released to the public. Care must be taken that NRC information does not add to the confusion at the site or of f site. During drills, it has been foned that the 11eensee I
and off-site officials are of ten ahead of the NRC in their
, a s se s sme nt s. In addition, the licensee and of f-site of ficials j have peoplanned actions based on consideration of local factors
! that may act be known to the NRC.
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- 3. MAJOR POINT 5 Slide 13 he major points of this volume may be summarized as f oll ow s.
- U.S. Nuclear Regulatory Commission (NBC) staf f members are part of a very large emergency response organization.
- A standard event classification scheme has been established on which levels of response are based.
e no NRC is not likely to be part of the immediate (0 to 2 hr) implementation of predetermined actions in emergency re sponse .
- Early in an event, NRC analysis and asse ssment will concentrate on the need for and adequacy of protective actions.
e ne !!censee, local governments, and f ederal agencies have proplanned their early emergency responses.
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6 he NRC shon1d not interf ere with emergency response decisions or actions that basically conform to NRC guidelines.
e no NRC role is best performed at the accident site.
Responsibility shon1d be transf erred f rom NRC headquarters to the Site Team as soon as possible.
- NRC assessments ahon1d always be discussed with the licensee and appropriate off-site officials. especially where areas of disagreement occur.
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14 e Great amoertainties are associated with condsoting assessments of severe accidents (core melt accident segue ace s) .
- In amersency response, it is best to use the most realistic a ssumptions po ssible.
Appendix A BIBLIOGRAPHY FOR VG.UES 1-5 I
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. = _
Appendix A BIBLIOGRAPHY FOR VOLUMES 1-5 i Aldrich, D. C., and J. L. Sprung, Technical Guidance for Sitina Criteria
! Develonnent, NUREG/CR-2239, Sandia National Laboratories, l Albuquerque, N.M. , December 1982.
Bradshaw, R. W., and R. D. Ivany, " Core Damage Assessment Procedure Based Upon Post Accident Chemistry and Radiation Sample Analysis,"
l i
Trans. Am. Nucl. Soc. 45:267-68 (1983).
i i
Burke, Richard P., and Carolyn D. Helsing, In-Plant Considerations for Ontimal Of f site Resnonse to Reactor Accidents, NUREG/ CR-2925, SAND 82-2004, Sandia National Laboratories, Albuquerque, N.N.,
November 1982.
i
\
Camp, A. L., and J. C. Cummings, Limht-Water-Reactor Hydroman Manual, l NUREG/CR-2726, SAND 82-1137, Sandia National Laboratories,
! Albuquerque, N.M., June 1983.
Cook, D. H., S. R. Greene, R. M. Harrington, and S. A. Hodge, Loss of l
i l- DER Seanence at Browns Ferry Unit One--Accident Seanence Analysis, NUREG/CR-2973, ORNL/TM-8352, Oak Ridge National Laboratory, Oak Ridge, Tenn., May 1983.
Dynes, R. E., E. L. Quarante111, and G. A. Kreps, "A Perspective on Disaster Planning," TR-77, Civil Defense Preparedness Agency, Washington, D.C., December 1982. Available from Federal Emergency Management Agency.
I Examination of Offsite Radioloalcal Emeraency Protective Measures for l Nuclear Reactor Accident Involvina Core Melt, NUREG/CR-1131, Washington, D.C., October 1979.
l l
1 17 e
. . - - . - - - - - - _ _ , . . , , . _ _ . . _ --,---n- ..-,--n---mnn-nn-~~ ---,,,r--,-.- ---. - , - -
18 Federal Emergency Management Agency, 'Pederal Radiological Emergency Response Plan (FREEP); Publication for Public Review, Comment, and as the Basis for a Large-Scale Field Exercise," Fed. Reaist. 49:3578 (Jan. 27, 1984).
Hans, J. M., Jr., and T. C. Sell, Evacuation Risk--An Evaluation, EPA-520/6-74-002, U.S. Environmental Protection Agency, Office of Radiation Research, National Environmental Research Center, Las Vegas, Nev., June 1974.
Harrington, R. M., and L. J. Ott. The Ef fect of Small Canacity. Hinh-Pressure Infection Systems on TUUV Seanences at Browns Ferry Unit h NUREG/CR-3179, ORNL/TM-8635, Union Carbide Corp. Nuclear Division, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1983.
International Commission on Radiological Protection, ' Protection of the Public in the Event of Major Radiological Accidents: Principles for Planning, " ICRP/84/C4-5/2, Report for Committee 4, Adopted by the Main Commission, May 1984.
Krep4, G. A., " Individual and Societal Effects of Peacetime and Wartime Nuclear Disasters," in Proceedinas of the Svenosium on the Control of Exoosure of the Public to Ionizine Radiation in the Event of Accident or Attack, April 27-29, 1981, Reston, Va., N6tional Council on Radiation Protection and Measurements, Bethesda, Md., May 1982.
Larkins, J. T., and M. A. Cunningham, Nuclear Power Plant Severe Accident Research Plan, NUREG-0900, U.S. Nuclear Regulatory Commission, Office of Regulatory Research, January 1983.
Lewellen, W. S., R. I. Sykes, S. Parker, and F. C. Kornegay, Comnarlson of the 1981 INEL Disnersion Data with Results from a Number of Different Models, Aeronautical Research Associates of Princeton, Inc., Report No. 505, October 1983.
19 Lewin, J., Evaluation of Instr-- ntation for Detection of Inadaanate Core Coo 11mm in Bolllan Water Reactors, NUREG/CR-3652, ORNL/TM-9029 Oak Ridge National Laboratory, Martin Marietta Energy Systems, Inc.,
Oak kidge, Tenn., April 1984.
Margulies, T. S., and J. A. Martin, Jr. , Dona Calenlations for Severa LWR Accidents, NUREG-1062, U.S. Nuclear Regulatory Commission, Division of Risk Analysis and Operations, Office of Nuclear Regulatory Research, Washington, D.C., May 1984.
Mart in, J. A. , Jr. , ' Doses While Traveling Under Well Established Pl ume s , " He a l t h Phy s . 32:305-07 April 1977.
Martin, J. A., Jr., ' Objectives of Emergency Response and Potential Benefits of Evacuation and Shelter," in Proceedinas of the Mid-Year Tonical Svanosina of the Health Physics Society. January 6-10. 1985.
Ma rt in, J. A. , J r. , ' Perspective on the Role of Radiological Monitoring in an Emergency and Radiolodine Monitoring in an Emergency,-A Case for I-13 5, " Trans . Am. Nucl . Soc. 34:727-29, May 1980.
Martin, J. A. Jr., ' Potassium Iodide: Predistribution or Not? The Real Emergency Preparedness Issue " Health Ehza. 49(2):287-89, August 1985.
Perry, R. W., Citizen Evaluation in Resnonse to Nuclear and Nonnuelaar Threats, BEARC-400/81/013, Battelle Human Affairs Research Center, Seattle, Wash., September 1981.
Perry, R. W., M. R. Greene, and A. Nushcatel, American Minority Citimann in Disaster, NSF-PFR-80-19297, Battelle Human Affairs Research Center, Seattle, Wash.,1983.
20 Perry, R. W., M. K. Lindell, and N. R. Greene, Evacuation Flammina in Emermancy Mananament. D.C. Health and Co., Lexington, Nase.,1981.
Price, John M., Douglas W. Cooper, and Charles 8. Tee, Ennadiant Methods of Raneiratory Protection: III. Snhaleron Particle Testa and e--- rv of G==11tv Factors, NUREG/CR-3537, SAND 83-7450, Sandia National Laboratories, Albuquerque, N.N., July 1985.
Rogovia, Mitchell, Three Mlla Island. A Ranort to the r---Analomara and the Public, U.S. Nuclear Regulatory Commission. Washington, D.C.,1979.
Sh1eien, B., G. D. Schmidt, and R. P. Chiacchierini, Backaronad for Protective Action Rae--- ndations: Aceldental Radioactive contamination of Food and Animal Feeds, FDA 828196 U.S. Food and Dras Administration, Bureau of Radiological Health, Rockville, Md., August 1982.
Silberberg, M., J. A. Mitchell, R. O. Noyer, W. F. Pasedag, C. P. Ryder, C. A. Peabody, and M. W. Jankowski, Renassamment of the Tachaleal Ra se n for Es t a== t ina sourse Terma . NUREG 0956, U.S. Nuclear Regulatory Commission, Accident Source Tern Office, Office of Nuclear Regulatory Research, Washington, D.C., July 1985 (Draf t Report for Comment).
Turner, D. B. , Workbook of Atmannharlo Diana ralon Est imate s. TD-18, U.S.
Environmental Protection Agency, Office of Air Prograns, Research Triangle Park, N.C., January 1973.
U.S. Department of Health and Human Services and U.S. Food and Drug Administration, " Accidental Radioactive Contanination of Human Food and Aalsel Feeds: Recommendations for State and Local Agencies,"
Fed. Ramlat. 47:47073 (Oct. 22, 1982).
_. . - _ - -. . . _ -. - - .-. - ~ _ . _ ~ .- . . . - - _ _ _ _ - -
! i 81 U.S. Environmental Protection Aseney, Mammal of Prataative Astina Galdaa ,
4 and Pratmative Astiaan for halaar Insidenta. IPA-520/1-75-001, U.S.
4 Environmental Protection Ageney, Environmental Analysis Division, -
{ Office of Radiation Programs, Washington, D.C., September 1975, rev.-
{ June 1980.
i
,I U.S. Food and Drug Administration, 'tediosotive Contamination of Euman '
L I
and Animal Feeds and Potassium Iodide as a Blocking Agent in a Radiation Beersoney, Fad. Ramlat. 48:242, Part VII Dec. 15, 1978.
i U.S. Nuclear Regulatory Commission, Amamar Prasadaran for the IRC i j Inaldaat Raamamma Plant Flaal Ranart. NUREG-0845, U.S. Nuclear i l Regulatory Commission, Division of Emergency Preparedness and [
i Engineering Response. Offlee of Inspection and Enforcement, 1 1
! Washington, D.C., February 1983. [
{ U.S. Nuclear Regulatory Commission, " Assumptions Used for Evaluating the Potential Radiomative Consequences of a Loss of Coolant Aeoident for j Pressurised Water Resotors," Rav. 2. U.S. Nealear Raamlatarv i c---analen Ranulatory Gulda 1.d, June 1974.
l U.S. Nuclear Regulatory commission, " Assumptions Used for Evaluating the !
! Potential Radiological Consequences of a Loss of coolant Aeoident for
! Boiling Water Reactors," Rav. 2. U.S. Naalear Raamintory e - lanlam ,
Raamlatorv Gulda 1.3, June 1974.
t f
I j U.S. Nuclear Regulatory Commission, Clariflaatina of TMI Astlan Plan
! Raamiramentat Raamirementa for "- rammer Raamanas Camah111tv, i I
2 NUR50-0737, suppl.1, No.1. U.S. Nuclear Regulatory Commission, !
Washington, D.C., January 1983. !
i i
I s
2 i
L l
i
. _ _ . . . . ~ _ - . _ _ , ~ . _ _ , _ . , - . . _ . _ _ _ , _ . - _ _ _ . - - . _ _ _
22 U.S. Nuclear Ross1 story Commission, Criteria for Presaration ==d kvaluation of * 'ioloniaal "- raamer Roanonae Plana ==d Premaredmana in Emmeert of Naalear Power Plant a, WRM-0654, FRNA-RRP-1, rev.1 U.5. Noelesr Reestatory Commioeion aad Federal Emersesoy Nanageneat Asemey, Washtaston, D.C. , November 1980.
U.S. Noetear Regulatory Commission, Demoarashis Statistisa Portalaims to Naalear Power Reactor Eltaa, WREG-0384, Office of Nuclear Rosetor Regulation, Division of site Saf ety and Ravironmental Analysis, Weakington, D. C. , October 197 9.
U.S. Noelear Resstatory Commission, Final Raylronmental Innant Statemente u..a11 a ..a storane of seemt Limht Water Power Reactor Fuel. WRBU-0575/V1-V2, U.S. Noelear Regulatory Commission,
- Wa shington, D. C. , Assuet 1 M 9.
U.S. Noelear Regulatory Commission Faastional Criteria for "- raemov ,
Beamonse Faellities, WREG-0696, Office of Inspection and Enforcement, February 1981.
1 U.S. Nuclear Regulatory Commission NRC lacident Roanosse Plan, WREG-I 0728, rev.1 U.S. Nuclear Regulatory Commission, Office of Inspection and Ref orcement, Washington, D.C. , April 1983.
r U.S. Nuclear Regulatory Commission, Reactor Safety St udy-- Am Aa se n ame nt of Accident Blaks la U.S. C:-- relal Nuclear Power Planta, WASN-1400, NURNG-7 5/ 014, Oct obe r 1 M S .
U.S. Noelear Regulatory Commission and U.S. Ravironmental Protection
) Agency Task Force on Emergency Planning, Plaamina...Baala for the Develument of State and Local Government Radioloalsal_Imerasacy Roanonse Flama in Smaaort of Linkt Water Nuclear Power Plants, NURME-0369. EPA 520/1-75-016, U.S. Noelear Regulatory Commission and U.S. Ravironmental Protection Agency, Washington, D.C. , December 1978.
23 U.S. Nuclear Regulatory Commission, TMI-2 lannona 1aarmadr Tank Foras Statua Ranort ==d hart Tara Raa--- =dattoma, NUREG-0578, Of flee of Nueleer toastor tosolation, July 1979.
U.S. Nuclear Regulatory Commission. Tenhalaal Raman for Rata time Flanian Prodnat Rahavior Durina LWR Amaidenta, NUREG-0772, U.S. i Nuclear Regulatory Commission, Office of Nuclear Regulatory Research ,
l and Offlee of Noelent Reactor Regulation, Washington, D.C.,
June 1981.
i Wicher, R. P., C. F. Weber, R. A. Lorens, W. Davis, Jr., S. A. Bodge, and A. D. Mitchell, Station Blackout at Browna Ferry Unit Oma--Indina
(
and Noble Gaa Distrihmtioma, NUREG/CR-2182, ORNL/NUREG/TN-455/V2, Oak
- Ridge National Laboratory, Oak Ridge, Team., August 1982.
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Appendix B SCIENTIFIC NOTATION l
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Q '.g l SCIENTIFIC NOTATION
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' 611de le 8elestific notation is based on logarithms. A logarithm is the esponent that indicates the power to which a number is
^
ra'ised to provide a given anmber. For example, the loaariths of
~
100 to the ba se 10 is 3..
ne following shows a list of the simple powers or
. . - esportent s of 10 and their logarithms.
+
,/ Number Power Logarithm
,4 v
I 0.00001 10
-5 -5 1 0.0001 10-4 -4 0.001 10~3 -3 4' -
O.01 10
-2 -2 0 .1 10~1 -1 0
,, .- ~ 1.0 10 0 1
10.0 10 1 2
100.0 10 2 1,000.0 10 3 3 4
., 10,000.0 10 4
.l 100,000.0 10 5 5
/ 'Silde 15 ~
As an saa.nple of the use of scientific notations, assume that the following numbers of people (y) live within a radius equal to or less than the following nasber of miles (a) from a
- nuclest power plant 1
l l
l .
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l 28 No. of Scientific No. of people (y) notation miles (x) f 6
100 10 5 e
1 1,000 10 10 100,000 10' 20 1,000,000 0 10 50
]
s' I S11de 16 Figure B,.1 shows x graph of the same information as'in our l
table. This information can now be graphed using a logarithmic scale. Occasionally, scientific notation (using logarithms) may l be used to indicate numbers, especially those requiring several digit's to express.
l Slide 17 For ease of use, a set of prefixes and symbols are sometimes used in place of scientific notation. For example 5 mCl is equivalent to 5 x 10 -3 C1. Table B.1 shows the standard prefixes. Care must be taken in using prefixes during an emergency. For example, if one used Ci when one intended mci, the error would be of factor of 103 (1000).
l l
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29
' I I I I I 10 5 - g 4 -
10 "i
8 o
S -
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2 ,
102 _
10 1 -
l I I I l 30o 20 30 40 50 5 10 MILES Fig. B.1. Example of scientific notation: population (p) inside circle of radius x (miles).
I i
30 Tabl e B.1. International System of Units (SI) prefixe s Factor Prefix Symbol 10 18 esa E 10 15 peta P 12 10 tera T 9
10 giga G 0
10 mega N 3
10 kilo k 2
10 hecto h 1
10 deka da
-1 10 deci d 10" conti c 10
-3 milli a
~0 10 micro p 10
-E nano a 10-12 pg,,
p
-15 10 femto f
-18 10 atto a l
l__-__-- _ _ _
Appendix C MEIRIC UNITS l
I l
l l 31 i
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33 Appendix C METRIC UNITS Slide 18 To convert from to multiply by curie (Ci) becquerel (Bq) 3.7 00 000 E+10 degree (angle) radian (rad) 1.745 329 E-02 degree Fahrenheit (*F) degree Celsius ('C) C = (F-32) /1.8 gallon (U.S. liquid) cubic meter (m3 ) 3.7 85 412 E-03 liter (L) cubic meter (m3 ) 1.000 000 E-03 ,
mil e (U. S. statute) meter 1.609 347 E+03 mile per hour meter per seccad (m/s) 4.470 400 E-01 pound (Ib avoirdupois) kilogram (kg) 4.535 924 E-01 rad (absorbed dose) gray (Gy) 1.000 000 E-02 rem (dose equivalent) sievert (Sv) 1.000 000 E-02
1 l
Appendiz D LIGHT WATER REACIt)R MM MER PLANT IESIGN FEATURES 35
l i
l Appendix D LIGHT WATER REACIUR NUCLEAR POWER PLANT IESIGN FEATURES he fundamental distinction between a pressurized water reactor (PWR) and a boiling water reactor (BWR) is that the heated water coolant in a BWR is allowed to boil. PWR coolant I is pressurized so that no boiling occurs. A brief description of each reactor type follows.
l D.1 PRESSURIZED WATER REACTORS In the United States, PWRs are designed by three companies:
We stinghouse , Ccabustion Engineering, and Babcock y Wilcox.
Each de sign has its own peculiarities, but the three concept s are suf ficiently similar to be described generically.
l D.1.1 Primary System Slides 19 and 20 j
The PWR primary coolant loop (Fig. D.1) consists of the reactor vessel and core, reactor coolant pumps, steam generators, pressurizer, and interconnecting piping (see i Fig. D.2 for a typical plant layout) . Controlled fission ,
reactions in the core heat the coolant, which is then pumped to the steam generator, where the heat is transferred to a secondary loop through U-type tube s. The reactor coolant then returns to the reactor vessel, where it is reheated. An electrically heated pressurizer maintains a high pressure so that the water coolant does not boil. Most radioactivity is confined to this loop.
The PWR secondary loop (Fig. D.1) use s the dry steam frcat
(
the steam generator to drive the turbine generator for electrical power generation. no expanded steam is cooled in the condenser and becomes water, which is pumped back to the steam generator for rebolling. Unusable heat (removed f rce the 37
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40 condensed steam) is rejected via the tertiary loop to the
-environment (river, lake, ocean, or cooling tower). About two-thirds of the power generated in the core is rejected as thermal heat; about one-third is converted to useful electricity.
D.1.2 Secondary Swstems
.A'dditional PWR systems are connected to the reactor coolant system to perform specific functions: reactor shutdown, coolant purification, decay heat removal, and emergency core cooling.
The reactor shutdown system controls the chain reaction in the core by neutron-absorbing rods that can be withdrawn or inserted from above the reactor core. If there is an accident, these rods can be quickly inserted to stop the chain reaction.
However, the radioactive decay of fission products in the core continues'to produce heat even after the chain reaction is terminated. - The decay heat removal system removes this d:::7 heat during shutdown. In the event of a loss of coolant accident (LOCA), the emergency core cooling ' system (ECCS) rapidly injects water into the reactor coolant system via electrical pumps. Pressurized water tanks also automatically force water into the reactor vessel if the pressure becomes too low. The containment cooler system removes heat from the containment building during normal operation and during a LOCA via fan-driven air flow over water-cooled coil s.
D.I.3 Waste Gas Tank A gas treatment system provides for low-temperature holdup, filtering, and decay of radioactive gases which are continuously released from the steam condensers. One type of radioactive gas is due to nuclear activation--that is, gas which becomes radioactive by absorbing radiation from the fission reaction.
The most important of these gases are N, I'N, II N, I00 , I'O, and F. Also, fission products, such as radioactive forms of krypton, menon, and iodine, are slowly released in gaseous form from leaks in the fuel pins. These gases dissolve in the
I 41 primary coolant water but during the steam generation process are released as noncondensable gases (off gase s), which are then removed frcm the steam condenser air ej ector.
Noisture is removed from the stean-diluted off gas by the recombiner, cooler-condenser, and desiccant dryer. Shor t-lived krypton radioisotopes and their decay products are retained in the charcoal desiccant dryer material. In the waste gas tank, off-gas is refrigerated (to O'F) to eliminate remaining water and thoroughly filtered through charcoal. In this waste gas tank, the retention time is 96 hr for krypton and 42 days for zenon. Finally the off gas is released throagh a high-efficiency filter on the roof vent. The actual quantity of radioactive material contained in the waste gas tank is limited by the technical specifications. Even if the tank ruptured, off-site doses would not be in excess of Environmental Protection Agency (EPA) Protective Action Guide s (PAGs) .
D.1.4 Fuel Pool Slide 21
'Ihe fuel building is adj acent to the reactor building (Fig. D.2) and is used for receiving, storing, and servicing uranima fuel. The fuel pool provides for underwater, rack storage of new and spent fuel assemblies (Fig. D.3) . The assemblies are placed in vertical cells incorporating a neutron absorber with enough spacing between assemblies to prevent a
{ self-sustaining fission reaction. The pool is filled with borated water to further absorb neutrons and to provide radiation shielding f or plant personnel. The pool is constructed of reinforced coccrete with a stainless steel lining and has no gravity drains to prevent accidental water loss.
Water purification and decay heat cooling systems for the pool water are located in the fuel building. The fuel pool building has radiation monitors and filters to detect and limit any
- release as a result of damage to old fuel. How ev er, fission products that are the most important in terms of off-site health
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43 effects have decayed and are no longer in the old fuel. In addition, by the time spent fuel is stored, a minimum of three days af ter ' reactor shutdown, relatively little heat is generated by fission product decay; therefore, there is little energy to drive fission products out of the spent fuel into the a tmosphere.
D.1.5 Fission Product Barriers Slides 22 through 25 Enriched uranium pellets are forned into cylindrical l pellets of uranium dioxide. As shown in Fig. D.4, these pellets are stacked to form a long fuel qrlinder inside stainless steel or Zircaloy tube s, which are about 0.5 in. In diameter and 14 f t long. The tubes (cladding) are then sealed to protect the uranium fuel from attack by the coolant and to prevent escape of radioactive fission products. There are about 50,000 of these pins in a typical PWR core. These fuel rods are grouped into square arrays (Fig. D.5), which are termed " fuel bundles" or
" assemblies. " The reactor fuel is handled in the form of these f uel assemblies during transportation storage and in-core use.
Minimal risk to public health and saf ety demands that normal radioactivity releases from the reactor site be as low as reasonably achievable (ALARA) . For the radioactive fission products to escape to the environment, four barriers must be br ea ch ed. The first barrier is the ceramic matrix of the fuel pellet itself. The metal cladding surrounding the fuel pellets is the second barrier. The " gap inventory" ref ers to gaseous fission products, which collect in the gap between the cladding and the fuel and also in the plenum regions at the end of each f uel rod. The reactor coolant boundary (i.e. , the reactor vessel and primary coolant loop) and the containment building are the third and fourth barriers. The first three barriers are bounded by the primary coolant loop, which was sketched in Sect.
D.1.1 and shown in Fig. D.6; the containment building is de scribed in Sect. D.1.6. Core damage during a severe accident l
I i
l
44 STEAM GENERATOR TUBES, PART OF BARRIER 3
- 2. FUEL PELLET CLADDING
- 4. CONTAINMENT STRUCTURE FUEL PIN (~ 50,00015 CORE) pFISSION GAS
- 3. REACTOR PLENUM COOLANT w BOUNDARY g< CLADDING
- 1. FUEL VESSEL s' t "-
a <w <gg PELLET
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As FUEL-CLADDING 20 ft GAP (6 m) CORE ip; R
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~200 FUEL RODS IN EACH BUNDLE
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Fig. D.4. Barriers to release of radioactive material for a pressurized water reactor.
l l
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46 O STEAM OUTLET (TO TURBINE) g STEAM GENERATOR C FEEDWATER INLET (FROM CONDENSERI o \
PRESSURIZER M AIN COOLANT PUMP 68 f t !
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47 ref ers to suf ficient f ailure of the second barrier (i.e. , the fuel cladding) to release 20% of the gap activity. This level of damage was chosen because it would be beyond that expected if the safety gratoms designed to protect the core work as designed even under extreme stresses (accidents) .
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). The radioactive material released from the core must move rapidly through the primary system (third barrier) and containment (fourth barrier) without being significantly filtered or reduced by other methods such as contaissent spray s.
If the release from the fuel is held up or filtered, natural processes (e.g. , condensation and scrubbing) will renove most of the most important fission products.
Slide 26 Tnere are a number of critical saf ety functions (CSFs) that, if maintained, will prevent damage to the fission products barriers. Each of those functions is performed by a number of redundant engineered safety features (ESFs). Figure D.7 shows the relationship between the CSFs and fission product barriers.
Basically the CSFs are (1) shut down the reactor, (2) maintain the coolant flow and coolant level (keep oore covered and cool),
(3) remove heat from the primary system and containment, and (4) maintain radioactive control (e.g. , isolate containment) .
Monitoring and maintenance of CSFs are the primary function of the control room staf f during an energency. Specific amergency
- operating procedures (EOPs) and training are provided for this purpo se .
Ther ef or e, a major release must be preceded by a systems f ailure (start of accident), failure of one or more ESFs, j failure to meet one or more of the CSFs, failures of fission product barriers, and movement of the radioactive material through the plant rystems. Considerable instrumentstion exists l
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l 89 to indicate the status of Oke CSFs, fission product barriers, and movement of radioactive material.
D.1.6 Reactor Containment Buildina Slide 27 The final barrier to radioactivity releases is the containment building that houses the reactor coolant system.
The containment is designed to withstand the loads due to accidental depressurization of the reactor coolant system as well as to retain radioactivity released within the building.
Two major PWR design variations are currently in use.
The large, high pressure containment uses a combination of large-volume and high-pressure capability and is able to withstand the pressure increase resulting f rom depressurization of the reactor coolant sy stem; see Fig. D.8. It has been estimated that PWR containment would not f ail actil internal pressures were two or three times their design pressure. PWR containments are equipped with water spray Erstems to condense steam within the containment building atmosphere; these systems would also remove fission products from the containment a tmosphere. Chemical additives are mixed with the water spray, thus making it more effective in rasoving elemental iodine. A i
variation on this design has internal filters for cleaning recirculating air before it passes through the containment air cool er s. All containment gratoms have an automatic system to close none ssential vents, line s, and other penetrations through 4
- the building walls to minimize radioactive leakage during an
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a ccide nt. This is called containment isolation.
Slide 2 8 i
The ice-containment de sign (Fig. D.9) employs a i f undamentally dif ferent approach to control steam release during
- accidental depressurization of the reactor coolant erstem. The 4
released steam is directed to large ice beds, where it is cooled and condensed. Thus, the voinne and pressure rating of the
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52 building are reduced. A spray system in the upper contalment condense s steam that bypasse s the ice beds. The spray system and ice beds are expected to be effective in scrubbing fission products frce the containment atmosphere. A f an system circulates air from the upper to lower compartments and through the ice beds during an accident.
D.2 BOILDiG WATER REACIORS In the United States, BWRs are currently manuf actured only by General Electric Company. There are several design generations of this reactor type. However, it is sufficient to discuss the various reactor coolant designs as one, while outlining the three associated contalment designs.
D. 2 .1 Primary Systems Slide 29 he BWR reactor coolant system provides coolant water flow through the core, absorbing heat frce the fission reaction to produce steam in the primary loop which directly drives the i steam-turbine generator to make electricity. As illustrated in Fig. D.10, this is done in one coolant loop, rather than in two i loops as occurs in a PWR. Two external recirculation loops are I used to pump water into the reactor core. If an accident occurs, isolation valves in the steam line to the turbine would
- be closed to maximize radioactivity contalment.
D.2.2 Secondary System A nasber of systems that can af fect the course of an l accident are related to the operation of the reactor coolant sy st em. The BWR reactor shutdown system use s control rods, which are blades that fit between the fuel element s. They are inserted from the bottom of the core and can be rapidly forced
- into the core via pressurized air. The reactor core isolation i cooling system provides coolant to the reactor vessel via
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54 steam-turbine-driven pumps. He coolant level can be maintained during anticipated transient events and during some small pipe break accident s. He standby ECCS provides coolant to the reactor vessel if there is a LOCA. Spray systems cool the fuel from above, while the core is being reflooded from below, ne residual heat removal system removes decay heat from the reactor during shutdown and prevents the suppression pool (see Sect. D.2.3) from overheating during an accident. !
D.2 .3 Bo111am Water Reactor Containment Buildina 1 1
Slides 30 through 32 nere are three maj or BWR contalment de signs. Each uses a
[ 1arge water pool to provide pressure suppression during an I
a ccide nt. The Mark I de sign (Fig. D.11) has a separate toroidal pool (wetwell) that is connected to the main part of contalment (drywell) by large vent pipe s. This concept has a small volume and a high design pressure, no Mark II design (Fig. D.12) is called the over-ander design because the drywell is located directly above the wetwell. De Mark III design (Fig. D.13) is more like the PWR ice-condenser plant. He wetwell is in an I annular space at the contalment periphery. De vapor space of the wetwell forms the upper contalment compartment. He volume f is larger than the previous two BWR contalment designs with a substantially lower design pressure.
In addition to condensing steam during an accident, the BWR l
l suppression pools would remove fission products from the gas f flow passing f rom the drywell into the wetwell vapor space.
Spray systems are also included in these designs but would not necessarily be activated during an accident.
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Appendix E
' GIA6SARY OF TERM POR VM,UES 1-5 Alert: The " Alert" classification would apply if events are in progress or have occurred which involve an actual or a potential substantial degradation of the level of saf ety of the plant. Any relea se s are expected to be small fractions of the Environmental Protection Asener (EPA) Protective Action Guide (PAG) exposure levels.
Atmoseheric Dilution Factors: A constant used in estimating the dose to an individual from a radioactive plume.
Atmosnheric Dinnersion Model: A model used to predict the turbulent diffusion of the radioactive plume (release) caused by atmospheric eddy currents (e.g., a seulempirical Gaussian plume model) .
Atmoseheric Transnort Models: A model used to predict the movement of the radioactive plume (release) as a function of time.
Atto-: Prefix for 10-10 (0.000 000 000 000 000 001). ( Symbol "a ")
Becanerei (Ba): A term to denote one radioactive disintegration per second.
Catastronhic Accidental Release: Any releart tbit substantially exceeds the U.S. Environmental Protection Assuti Prr ctive Action Guides (i.e., by an order of magnitude or ac.e).
l-l Catastronhic Containment Failure: A failure of the containment that l
allows most ot the radioactive material in the containment to be
! released in a short time (e.g., <1 b).
l i Centi-: Prefix for 10-2 (0.01). ( Symbol "c ")
i; 61 s
62 Claddina: The fuel pellets in the reactor core are contained inside pins. The pin walls are the fuel cladding.
Connimant Federal Amency (CFA): The federal agency that has responsibility for technical assessment of radiological accidents.
In the case of reactor accidents, the CFA is the Nuclear Regulatory Commission.
Containment: The system whose functional responsibility is to prevent inadvertent release of fission products to the atmosphere and to l
! provide biological shielding in both normal and accident conditions.
Coolant: The fluid that is used to remove the heat from the core of the reactor. Typically it refers to the water that covers the core (primary coolant).
SRza: The central part of a nuclear reactor containing the uranium fuel, surrounding structure, and coolant water, where fission reactions occur to generate power.
Core Damame (Melt) Accident: An accident at the nuclear power reactor in which there is severe damage to the fuel that could result in release of significant amounts of radioactive material from the core.
Critical Safety Functions (CSFs): Functions that must be performed, both during normal operation and following accident initiation, to preclude fuel damage and/or maintain barriers to fission product release.
10 Curie: A term to denote 3.7 x 10 disintegrations per second = 3.7 x 10 10 Bq.
Deci-: Prefix for 10~I (0.1). ( Symbol "d ")
Deka-: Prefix for 10 (10). (Symbol "da ")
n
63 E211: The amount of radioactivity absorbed by the body.
Dose Calculations (Steadv State): Estimating doses to the individual using a measured or best estimate of the actual or potential source term and environmental dispersion. Simple and sophisticated dispersion models are available to provide reasonable estimates of time to exposure, exposure rates, and location of impact areas.
Dose Commitment Estimates: Estimating the off-site doses to selected individuals and the integrated dose to the entire population and certain subpopulations at risk. Using the dose-monitoring evaluatio. formation about the population (including location during the it,1 dent and eating and living habits), and information about the fate of the released radioactivity, an estimate of the population dose commitment can be made to be provided to affected individuals and public health workers for future use.
Dose Monitorina Evaluations: Evaluation of the measured impact of off-site radiation exposure rates from plume exposures and deposited radioactivity plus food contamination. These evaluations piece together the various types of actual measurements made at numerous locations to arrive at a consistent and understandable picture of the off-site impact of a release. This information is vital to modifying or taking further public protective actions, providing useful feedback to those doing dose calculations or dose projections and those groups who will calculate population dose commitments.
l Dose Protections (Dynamic): Estimating doses to the individual based on j the actual or potential degradation of reactor systems and I environmental dispersion as a function of time. An order-of-l magnitude estimate of an off-site impact can be made before an actual l release is initiated in a situation where the core and containment I are threatened.
Emeraency Action Levels (EALsl: Specific plant instrumentation readings or other appilcable observable indicators which, if exceeded, will n
64 initiate classification of an accident and other response actions.
For example, a reactor coolant leak rate of more than 50 gal / min would trigger _ an Alert, whereas _ a General Energency wonid be declared if core damage _ were imminent.
Emermency Core Coolina System (ECCS): An Engineering Saf ety Feature (ESF) designed to maintain core cooling following a Loss of Coolant Accident (LOCA).
Emeraency- Notification System (ENS): A ne twork of dedicated telephone lines which originate at many NRC licensed f acilities and terminate at the NRC regional offices and Headquarters (BQ) Operations Center (OC) . These lines provide a direct communication link between NRC and the licensee.
Emeraency Officer (EO): The NRC headquarters has an EO on call 24 hr a day, 7 days a week. This is usually a senior Inspection and Enforcement- (IE) management official who is charged [usually along with the Regional Administrator (RA)] with initial evaluation of a potential problem at a f acility and detenmining the appropriate NRC response actions, which are dependent on the severity of the event.
Emermency Onerations Facility (EOF) : A facility for management of the overall licensee emergenqy response (including coordination with federal, state, and local of ficials), coordination of radiological and environmental asse smaent s, and detennination of recommended pubilc protective action. It is typically located within 20 miles of the site.
i!
Emermency Onorations Procedures (EOPs): Procedures that direct the control room operators to bring the plant to a safe condition following an accident. These procedures are triggered by plant j rymptoms that indicate the status of the critical saf ety functions.
t Emermency Plannina Zone (EPZ): An area defined around a nuclear
. f acility for which specific preparations have been made to ensure n
65 that prompt and effective action can be taken to protect the public.
Enmineered Safety Features (ESFs): Regardless of the design' details of a particular reactor, ESFs perform a set of functions that include:
(1) reactor shutdown or " trip to stop the fission process; (2) emergency core cooling to keep the core cool by maintaining sufficient coolant flow and keeping the core covered, ' thereby minimizing the release of radioactive material from the core; (3) postaccident. radioactivity removal to remove from the containment atmosphere radioactivity released from the core; (4) postaccident heat removal to remove the core decay heat from the containment in order to prevent its failure; and (5) containment isolation to prevent any radioactivity not removed by the radioactivity removal system from being released to the environment.
Evacuation Zone (Sector): An area for which evacuation of its population has been planned to be conducted / directed as a unit.
These zones are typically bounded by local roads and other features so the people can understand if they are in the zone or not.
EXA;: Prefix for 10 (1 000 000 000 000 000 000). (Symbol E)
Federal Radiolonical Monitorina and Assessment Center (FEMAC): A center usually established at an airport near the scene of a radiological emergency from which the Department of Energy (DOE) off-site
- j. Technical Director conducts the Federal Radiological Monitoring and Assessment Plan (FRMAP) response. This center generally need not be
- located near the on-site or federal-state operations centers as long as its operations can be coordinated with them.
2 i
Federal Radioloalcal Monitorina and Assessment Plan (FEMAP): A plan to i
provide coordinated radiological monitoring and assessment assistance to the state and local governments in response to radiological emergencies. This plan, authorized by 44 CFR Part 351, is a revised version of the Interagency Radiological Assistance Plan.
m
[
l 66 Federal Roanonae Center (FRC): A center established by the Federal Emergency Management Agency (FEMA) at a location identified in conjunction with the State that serves as a focal point for Federal response team interactions with the State.
~1 Femto-: Prefix for 10 (0.000 000 000 000 001). (Symbol "f")
Finalon Product: Radioactive nuclei (material) formed during the fissioning process.
l Gan (Gan Release): The radioactive materisi released from the fuel during normal operation that is trapped in the fuel pin. If the pin fails (cladding fails), this material is a release of the gap into the reactor coolant.
I General Emermancv: A ' General Emergency" classification would indicate a severe accident involving imminent or actual core damage (melt) or loss of a plant to intruders. Full mobilization of emergency
) response organizations would be recommended along with appropriate protective actions. Radioactivity releases can be reasonably expected to exceed U.S. Environmental Protection Agency Protective Action Guide exposure levels beyond the immediate site area.
t 9.133-: Prefix for 109 (1 000 000 000). (Symbol 'ti")
i Grav (Gv): A term denoting absorption of 1 J of radiation energy in l 1 kg of mass (100 rads).
l Gro==d Contamination: Radioactive material deposited on the ground as a radioactive cloud passes.
Hecto-: Prefix for 102 (100). (Symbol "h")
Incident Resnonse Center (IRC): Each regional Base Team (BT) has an operations center called an IRC out of which are conducted vital l
emergency response functions. The IRC is always activated during l
l l
67 standby.
Inaastion Boundarw: See Emergency Planning Zone.
Immention Ernosure Pathgag: An area within about a 50-mile radius of the site for which plans and provisions have been made to protect the public from exposure from ingestion of-contaminated water or foods such as milk, fresh vegetables, or aquatic foodstuffs.
Isotone: Two or more species of atoms of a chemical element having the same atomic number and nearly identical behavior but having differing atomic mass (or mass number) and different physical properties.
Joule (J): A unit of energy, equal to the work done in moving 1 m against 1 newton of force.
Kilg-: Prefix for 103 (1 000). (Symbol "k")
Kiloaram (ks): The fundamental unit of mass in the International System of measurement (about 2.2046 lb).
Loss of Coolant Accident (LOCA): A breach in the reactor coolant pressure boundary that usually results in a decrease in the reactor coolant system pressure and, if unmitigated, may result in an uncovering of the reactor core. Following a large LOCA the reactor coolant would flash to steam as the reactor coolant pressure rapidly decreases. The reactor coolant pressure decreases at a much slower rate following a small LOCA. Because a large LOCA occurs on such a rapid time-scale, mitigation of a large LOCA would require automatic functioning of emergency core cooling system. Depressurization would occur over a much longer time period following a small LOCA.
333-: Prefix for 100 (1 000 000). (Symbol 'W")
Meter (m): The fundamental unit of length in the International System
, of measurement (about 3.2808 f t) .
l l
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Micro-: Profix for 10 (0.000 001). (Symbol 'v")
Milli-: Prefix for 10~ (0.001). (Symbol %")
l l
Nagg-: Prefix for 10 ' (0.000 000 001). (Symbol %")
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- l 2
Newton (N): A unit of force equal to 1 kg accelerated at 1 m/s ,
4 Noble Games: Chemically inert gases (e.g., xenon and krypton) which are present primarily in the reactor fuel pellets and fuel-cladding gap.
Because these gases are chemically inert, they cannot be filtered and l are the most likely radioisotopes to be released from the plant in the event of a core damage accident.
l I
Notification of an Unusual Event: This classification would apply if
. unusual events are in process or have occurred which indicate a
- potential degradation of the level of safety of the plant. No releases of radioactive materials requiring off-site response or
- monitoring are expected unless further degradation of safety systems
! occurs.
Onorations Center (OC): The NRC headquarters maintains an OC in Bethesda, Maryland. The OC is designed to provide a space out of which response functions are performed in reaction to reported events at licensed nuclear facilities. Technical personnel, licenses teams,
.and decision-makers perform various functions in support of the NRC response through the OC. In addition, normal mode functions are carried out by the Headquarters Operations Officer (HOO) during nonevent time periods. Functions performed in the OC are similar to' those performed in the regional Incident Response Centers.
Onerations Sannort Center (OSC): A " locker room-type " f acility located on site to house the various support teams that will assemble (e.g.,
in plant monitoring, health physics, and damage control). The OSC will have a reliable communication link to the Technical Support Center (TSC).
. _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - . - - - . . _ . _ - -.._.. - - _ - _ - - a
69 Part 100 Asan=ntions: Assumptions about the release of the fuel (100% noble gas and 25% iodine) and containment used in the licensing of a reactor. It is not appropriate for emergency response. [See Regulatory Guides (R.G.) 1.3 and 1.4.]
l Particulate / Aerosols: Dust / mist.
Peta-: Prefix for 10 (1 000 000 000 000 000). (Symbol '1P ")
4
-1 Pico-: Prefix for 10 .(0.000 000 000 001). (Symbol ')")
Pinme: Windblown cloud of radioactive material released from a nuclear power plant.
Plume Ennosure Emermancy Plannina Zons: An area within about a 10-mile radius defined by local conditions in which plans and provisions exist for prompt (e.g., <1 hr) initiation of protective action (e.g.,
shelter or evacuation) to protect from short-term exposure from the plume (shine, inhalation, and ground contamination).
Plame Excosure Pathway: The principal exposure sources from this pathway are: (a) whole body external exposure to gamma radiation from the plume and from deposited materials and (b) inhalation exposure from the passing radioactive plume. The duration of principal potential exposures could range in length from hours to days.
Projected Dose: An estimate of the radiation dose which affected population groups could potentially receive if protective actf ons are not taken.
Protective Action: An action taken to avoid or reduce a projected dose (sometimes referred to as a protective measure).
i
70 Protective Action Guide (PAG): The U.S. Environmental Protection Agency (EPA) and the Food and Drug Administration (FDA) recommended projection doses at which various protective actions are warranted.
These are doses that would be received if no protective actions were taken. They do not indicate any dose received prior to the time of proj ecting the dose.
Protective Measures Team (PNT): A group of NRC personnel located in the Headquarters Operations Center whose primary responsibility is to assess the adequacy of the protective measures being recommended by I the licensee and being implemented by off-site officials.
Puff Release: A very short ((1-hr duration) release of material.
1 Rad (radiation absorbed dose): Absorption of 0.01 J of radiation in 1 kg of mass (1/100 Gy) .
I Radionuclide: A radioactive form of a chemical element.
I I
Reactor Safety Team (RST): A group of NRC personnel located in the Headquarters Operations Center whose primary responsibility is to l
assess the condition of the core, containment, and engineered safety features in order to provide the Protective Measures Team with estimates of core and containment damage (including potential telease pathways) so that the Protective Measures Team can estimate potential of f-site consequences.
71 Ram (roenteen manivalent ===): The dose of any ionizing radiation which produces the same biological effect as 1 rad of I rays. The rem equivalent dose is the relative biological effectiveness (RBE) of the type of radiation times the absorbed dose of that radiation. Typical values of the RBE are shown in the following table:
Radiation type RBE y and X rays 1 fast electrons (p rays) 1 slow electrons ($ rays) 1.7 thermal neutrons 3 fast neutrons 10 protons 10 helium nuclei (a rays) 10 l
heavy ions 20 Resident Inanector (RI): Because the RI is resident on site, he or she would be the first NRC representative on site in the event of an incident.
Resnonse Coordination Team (RCT): The RCr is composed of individuals from the Incident Response Branch (IRB) trained in the NRC Incident Response Plan and supporting procedures so they can act as facilitators and logistical support functionaries in a response.
1 They train personnel and develop programs i~n support of incident response in addition to conducting drills and exercises, and they organize the NRC response structure and its components. An RCT liaison is assigned to each technical and liaison team and supports that team in any way possible.
Safsanards Team (SGT): The SGT's primary function is to evaluate each event for security and law enforcement significance. Every event has safeguards significance until proved otherwise. The principal focus of the team is to support the Federal Bureau of Investigation (FBI)
72 i
with critical information relating to the safeguards contingency plans at the facility. Safeguards information will be provided to the NRC Executive Team (ET) to be used in making recommendations to the state regarding protective actions for the public.
Sc ram:. Quick shutdown of a nuclear reactor by rapidly inserting rods of neutron-atsorbing material into the core.
Sector: A term used to indicate areas that may be affected by an off-site release. It is typically an area encompassed by a 22.58 arc. ,
It can be designated by standard compass designations (e.g., NNE),
letters, or numbers.
Shine: Radiation from an airborne cloud containing radioactive material or from contaminated ground.
Slevert (Sv): The equivalent dose associated with 100 rem.
Site Area Emeraencv: This classification would apply if events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to exceed U.S. Environmental Protection Agency Protective Action Guide caposure levels near the site l boundary.
I Site Team (ST): A group of NRC personnel who will travel to the site and assume the responsibilities of the Operations Center af ter the initial stages of the event. The ST will receive assistance from the
. Operations Center, however.
- Source Ters
- The radioactive material that is being released or that is projected to be released from a nuclear power plant. It is used in estimating off-site doses.
Technical Suncort Center (TSC): A protected and shielded area loca ted I near the control room that will accommodate the licensee technical
73 personnel who will be attempting to communicate with the operators in the control room in diagnosing and mitigating the event.
h: Prefix for 10 (1 000 000 000 000). (Symbol 'T")
- Wind Direction (downwind): Normally wind direction is given in terms of the direction from which the wind is blowing. One should always be certain as to the directions from which and to which the wind is blowing. The affected area or situation is downwind. Because of wind shif ts, plume meander and topological features, the downwind direction would vary in time and space.
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Appe ndix P ' i l
SLIBES RELATING 10 VSE8E 10F TEE l SEVRE REACIOR ACCIBENT INCIDENT
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- APPENDIX F i
SLIDES RELATING TO SEVERE REACTOR i ACCIDENT INCIDENT RESPONSE TRAINING MANUAL:
- OVERVIEW AND
SUMMARY
OF MAJOR POINTS y VOL.1 NUREG-1210 ORNL/TM-9271/V1 i
I Slide 1
1 i
I i
I 1
l
- OBJECTIVE l
l Describe basic NRC response guidelines =
I l
l f
I i
i
- Slide 2 I
1 l
BIG PICTURE
- Why am I here?
i
- What is expected of me?
- Who are all of these other people?
l
- Do I interact with them? How?
- How is this different from my normal job?
- What is important?
4
- How do I scope the level of an emergency?
l i
1 I Slide 3 t
\
! t
l l
t i
! UNDERSTAND RESPONSE PHASES l
'
- Early (0 to 6 hr)
Predetermined
!
- Intermediate / late (hours to months) j More ad hoc i
i i
- Slide 4 1
^
81 EMERGENCY CLASS RESPONSE Local and State Class Plant Action Agency Action Unusual Event Provide notification Be aware Alert Mobilize plant resources Stand by a Man centers (help for control room)
Activate Technical Support Center (TSC)
Site Area Full mobilization Mobilize Emergency Nonessential site Man emergency centers personnel evacuate and dispatch monitoring team Activate TSC, Operations Inform public - activate Support Center, and warning system Emergency Operations Facility Take protective actions in accordance with PAGs or on an ad hoc Dispatch Monitoring Team basis Provide dose assessments General Emergency Full mobilization Recommend predetermined Recommend predetermined protective actions to the protective actions (within public based on 15 min) after declaring plant conditions emergency Precautionary evacuation (2 to 5 miles) aThe NRC will typically staff its response centers and implement its emergency response plan at the appropriate level for an Alert.
Slide 5 1
i i
I j EMERGENCY CLASS DESCRIPTIONS i
l j Classa Core Status Radiation i Unusual Event No threat to irradiated No release above fuel technical specifications (or annual limits)
Alert Actual (or potential for) Release is small fraction j substantial degradation of EPA PAGs beyond the
! of safety site boundary a i
Site Area Major failures of Release is less than l
j Emergency functions needed for EPA PAGs beyond the
! public protection site boundary General Actual or imminent Dose may exceed Emergency core degradation EPA PAGs
- Classifications are based on plant instrument levels (i.e., EALs).
~
Slide 6 i
I
UNDERSTAND RESPONSIBILITIES
- Licensee Mitigate consequences Make protective action recommendations
- Off-site officials Assess and recommend protective actions to the public e
- Public Take actions
- NRC Monitor Help (if requested)
Serve as federal technical expert and clearinghouse
- Discussed in Vols. 3 and 5 Slide 7
UNDERSTAND EXTENT OF PREPLANNING
- Extensive preplanning Plans and procedures Organization Centers, facilities, and communications
= if preplanning efforts are not recognized, the preplanned response could be delayed by questions or conflicting recommendations
- Discussed in Vol. 3 Slide 8
j i
i i
UNDERSTAND EXTENT OF RESPONSE i
- Very large response involves more than 1000 people !
1 Licensee i
j State officials 4 l i Local officials '
l Federal officials (NRC, the Department of Energy, the Federal a l Emergency Management Agency, the Department of Health and
- t i
Human Services, the Environmental Protection Agency, etc.)
National laboratories i
. Vendors '
Discussed in Vols. 3 and 5 l Slide 9 i
i UNDERSTAND THAT NRC CAN BEST RESPOND FROM THE SITE ,
i l
l
- NRC personnel should be on site as soon as possible i
- NRC headquarters performs an interim leadership role until the l regional staff arrives at the site
- Headquarters supports the regional staff once it takes the lead
- Discussed in Vol. 5 i ,
i I
l i
l i
SMe 10
!l '
l
i l
l i
1 I
! UNDERSTAND SEVERE ACCIDENTS AND THEIR ASSESSMENT
)
i
- Understanding is crucial to protective action assessment !
- Only major fuel damage accidents should lead to a general
, emergency (predetermined bases) 1
- Severe accident is very difficult to assess -
- For a severe accident, however, less detail is needed for protective action decisions
- Discussed in Vols. 2 and 4 i :
k l
i Slide 11 ,
r
1 l
l l
l
\
! UNDERSTAND NEED TO DISCUSS ASSESSMENTS l
- Licensees and off-site officials are generally ahead of the NRC in j assessing the situation i
l
- Most early actions are preplanned i
- NRC is not familiar with local conditions g
- NRC conclusions should be discussed with licensees and off-site officials l
l
)
! Slide 12 i
i r ! it ! ! '
- e i
v t
c t
e t a
f o h d o pr t s
_ e n n n
- h o o o s i t
i t
i l
b t a t e c a
a n a
- t s e t r e e m n s .
e e n _
o n c l
p p e
n s e m o e c
e se i
- b r s d S
T a
t ioan l u
h t
o i
h dp
- N e e es h w I
O s mme s d e e 3 1 -
n e n r r e fe c P o s n p h i r
e t n n r n e R s c s ni t
s e a ea d i
e i l l p t
- O r n t f- m e nid S -
. ~
J o
rf ao s r iu A e g i t pr s e p r g o
M r a a c eo s s n e oR nC l i k e a e iN y
r f
i s t aes C b t s .
_. e s a on R s eo ut v t e a a
l c
l l yic N h qm t r ke d e
t f
o n s s oo
- e v
i e t n n nf n t
r e l
n e o do a
p d r
ome t i nr e
v p s
e l
hl uc oy
- s a st i n s r sac l
yo n i
d ie y C n Cd e l l Ci s R ta Rr rit ac r
a Ra
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- 7 - - _; . . . . .. . - . _ . - - . _ - - . -
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MAJOR POINTS (CONTINUED) i i
}
- NRC's role is best performed at the site l
- NRC assessments should be discussed with licensees and other organizations C ;
= Great uncertainties are associated with assessments of severe (core ;
melt) accidents
- Emergency response staff should always use the most realistic assumptions possible ;
t 4
I i I
Slide 13 (Continued)
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EXAMPLE OF USE OF SCIENTIFIC NOTATION No. of Scientific No. of People (Y) Notation Miles (X) 100 10 2 5 e 1,000 10 3 10 100,000 10 5 20 1,000,000 10 6 50 Slide 15 i
93 EXAMPLE OF SCIENTIFIC NOTATION:
POPULATION INSIDE CIRCLE OF RADIUS X (MILES) 10 6 I I I I I 10 5 _
g _
4 10 - -
"i 8
0 9
10 3 -
S -
8 5
0 8
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g _
10 1 - -
I I I I I 30o 5 '10 20 30 40 50 MILES Slide 16
94 S1 PREFIXES Factor Prefix Symbol 10 18 exa E 10 15 peta P 10 12 tera T 10 8 giga G 10 6 mega M 10 3 kilo k 10 2 hecto h 10 1 deka da 10 -1 deci d 10 -2 centi c 10 -3 milli m
- 10 -6 micro g 10 -8 nano n 10 - 12 pico p '
10 - 15 femto f 10 - 18 atto a Slide 17
METRIC CONVERSIONS To Convert From To Multiply By curie (Ci) becquerel (Bq) 3.700 000 E+10 degree (angle) radian (rad) 1.745 329 E-02 degree Farenheit ( F) degree Celsius ( C) C = (F-32)/1.8 3 3.785 412 E-03 gallon (U.S. liquid) cubic meter (m )
2 liter (L) cubic meter (m )
3 1.000 000 E-03
- mile (U.S. statute) meter (m) 1.609 347 E+03 mile per hour meter per second (m/s) 4.470 400 E-01 pound (Ib avoirdupois) kilogram (kg) 4.535 924 E-01 rad (absorbed dose) gray (Gy) 1.000 000 E-02
- rem (dose equivalent) sievert (Sv) 1.000 000 E-02 Slide 18
PWR NUCLEAR POWER PLANT COOLANT LOOP, SIMPLIFIED DIAGRAM N SECONDARY LOOP MOISTURE SEPARATOR g : , r ~
AND REHEATER PRESSURIZER l l TURBINE g GEN ERATOR l l g STEAM r ', ,
f!
GENERATOR
, s , % CONDENSER ki f/ x e e j ~
y ?
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p ,
r 1 i TERTIARY LOOP PRIM ARY LOOP 1 j
'" CONTAINMENT WALL Slide 19
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99 BARRIERS TO RELEASE OF RADIOACTIVE MATERIAL FOR A PWR
- 2. FUEL PELLET CLADDING FUEL PIN (~ 50,0001 CORE) pFISSION GAS )
PLENUM
,hgCLADDING
< gE inwil
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~200 FUEL RODS IN EACH BUNDLE i-c'd
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i 100 BARRIERS TO RELEASE OF RADIOACTIVE MATERIAL FOR A PWR (continued)
STEAM GENERATOR TUBES, !
PART OF BARRIER 3 l
- 4. CONTAINMENT STRUCTURE
- 3. REACTOR COOLANT m BOUNDARY l VESSEL h 9
0 ft -
fs
( *- (15 m) C; l .
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(6 m) CORE STEAM y....
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GENERATOR
- f,4;:
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Slide 23
101 TYPICAL (A) ASSEMBLY AND (B) FUEL PIN (SECOND FISSION PRODUCT BARRIER)
CAP :
GAP SPRING
- x. .
SPACER -
- h
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(A)
SPACER r-CAP (B) l Slide 24
~ _ - _ _ _ - _ _ -
102-TYPICAL PWR PRIMARY SYSTEM (THIRD FISSION PRODUCT BARRIER)
O STEAM OUTLET (TO TURBINE) ew STEAM GENERATOR C PRESSURIZER v7== %
v MAIN COOLANT PUMP s
68 f t 4h 'V.
$"m
<a 3
v M v v x v
l [.. [
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Slide 27 l
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105 ICE CONDENSER DESIGN SPRAYS
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Slide 28 m
BWR NUCLEAR STEAM SUPPLY SYSTEM
+
l TURBINE GENERATOR REACTOR EXTRACTION (,) [_jCONDENSER SEPARATORS LINES AND D RS
{
CORE E o u l
- L".1 r
JET } JET O 4 PUMP PUMP h
_ f I) _
z 1 DEMINERALIZER FEED PUMP I U I
~
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DEMINERALIZER Slide 29 w -- -
107 BWR MARK I CONTAINMENT DESIGN I
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~
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- SUPP SiON J L J L Slide 30
108 BWR MARK ll CONTAINMENT DESIGN
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109 BWR MARK lli CONTAINMENT DESIGN P
OUTER CONTAIN MENT e
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Slide 32
g.o-- u s. .,vetu.u uut.Yo.v Co ..o i
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2O BIBLIOGRAPHIC DATA SHEET Sgt ekSTRUCTeoN$ oe.T E ntvinst NUREG-1210, Volume 1
- 3. TeYtt amo sueisYtt 3ttavtstANK Pilot Program: NRC Severe Reactor Accident 1 Incident Response Training Manual f , ,,,,,,,,,,, _ ,,,,
" pT- l vt..
wT o...e Jup 1986 T. J. McKenna, J. A. Martin,Jr., C. W. Miller, L.M. Hivel) , f . o.Yt.t.o.T,uuto R. W. Sharpe, J. G. 'itter, R. M. Watkins f wo=T- vtaa l
f ebruary 1987 T. et ,oa..=o caG.=12.Tio . t . o .. t o.oo tu ,, ,,c , e oaCm .oa.u T=U t.
O fin on GRANY NUMOER
- 10. 5PO45onsNG omGami2AYaoN N.est amo ssa.Linc aoon g,asa,e, t, cese, f tie TvrtoretPoni Office of Inspection and Enfo ment f Division of Emergency Prepared ss "
Instructional and Engineering Response U.S. Nuclear Regulatory Comissio ,
Washington, D.C. 20555
- 13. AS$TR ACT (200 er.rve .r isess This is one in a series of volumes that ol ctively provide for the U.S. Nuclear R:gulatory Commission (NRC) emergency re se personnel the necessary background information for an adequate response to s "ere reactor accidents. The volumes in the series are:
o Volume 1 -- Overview and Summary of pajo iPoints o Volume 2 -- Severe Reactor Accident /Dvervhew o Volume 3 -- Response of Licensee afd State \and Local Officials o Volume 4 -- Public Protective Actjbns -- Pr4determir.ed Criteria and Initial Actions o Volume 5 -- U.S. Nuclear Regulat#y Comissi%n Response Each volume serves, respectively, s the text fo a course of instruction-in a series ~
of courses for NRC response perso!nel . These mat f als do not provide guidance or license requirements for NRC lic 1 sees or state o ' local response organizations.
Each volume is accompanied by a appendix of slide that can be used to present this materiil. The slides are alled out in the te .
t 14 ooCoutNT ANALv818 - e utvvvonos/DESCnarfons ! h il A asLA Lif v emergency response protective actions power reactor accidents training accident assessment ,
response , , , , , , , , , , , , , , , , ,
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