Table Of ContentNUREG/CR-6042
SAND93-0971
._erspectives on Reactor Safety
by
E. Haskin/UNM
L. Camp/SNL
and Nuclear Engineering Department
of New Mexico
landiaNational Laboratory
',ratedby
Corporation
for
Nuclear Regulatory Commission
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NUREG/CR--6042
SAND93-0971
!
I I II
Perspectives on Reactor Safety
ManuscriptCompleted: February1994
Date Published: March 1994
Preparedby
F.E. Haskin, University ofNew Mexico
A. L.Camp,Sandia National Laboratories
Chemicaland Nuclear Engineering Department
209Farris Engineering Center
University of New Mexico
Albuquerque, NM 87131
Under ContractTo: SandiaNationalLaboratories
Albuquerque, NM 87185
Prepared for
Division of Safety Issue Resolution
Office of Nuclear RegulatoryResearch
U.S. Nuclear Regulatory Commission
Washington, DC 20555--0001
NRC FIN L1914
MASTEfi
DI_'RtBUTION OF THiS DOCUMENT IS UNLIMITED
ReactorSafe_ Course1_R-800_) ReactorSafet/'Course
ABSTRACT
The U.S. Nuclear Regulatory Commission (NRC) maintains a technical training center at
Chattanooga, Tenne_-_ to provide appropriate training to both new and experienced NRC
employees. This document describes a one-week course in reactor safety concepts. The course
consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression
in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and
offsite consequences. The course text is accompanied by slides and videos during the actual
presentation of the course.
USNRCTechnicalTrainingCenter ,iil NUREG/CR.6042
ReactorSafet_Course_R-800) ReactorSafe_ Course
TABLE OF CONTEN'I_
FOREWORD ................................................. vii
ACKNOWLEDGEMENTS ........................................ ix
ENGLISH TO METRIC CONVERSION FACTORS ....................... x
1.0 HISTORICAL PERSPECTIVE ................................. 1.0-1
2.0 ACCIDENT SEQUENCES .................................... 2.0-1
3.0 ACCIDENT PROGRESSION IN THE REACTOR VESSEL ............ 3.0-1
4.0 ACCIDENT PROGRESSION IN THE CONTAINMENT ............... 4.0-1
5.0 ACCIDENT CONSEQUENCES AND EMERGENCY PLANS ........... 5.1-1
USNRCTechnicalTrainingCenter v NUREGICR-6042
ReactorSafet_Course(R-800) ReactorSafet_Course
FOREWORD
The USNRC maintains a technical training center (TTC) at Chattanooga, Tennessee. This
TTC is responsible for training and, in part, qualification programs for new employees and, at times,
for retraining. Inasmuch as the agency hires about 150 new technical staff per year (due to turnover
from retirement or other losses) there is a need to train these new employees for their NRC r_le.
The entering staff have varied backgrounds: fresh from college or university; from Naval Reactors
programs; from private industry. _, all cases there are some training needs. However, the NRC
must cope, in its training programs, with the nationwide deemphasis in nuclear power in the
universities. Thus, we see in the incoming interns educational background in other areas, such as
chemical or electrical engineering, or else degrees in mathematics or physics or chemistry. This shift
in emphasis has placed an added burden on the TTC. In particular, it is seen that the most
fundamental concepts in reactor safety, are not readily available to the college student as formal
courses. Further, many of the present employees have not had the benefit of formal training in the
bases for many of the regulations d.ealing with fundamental safety concepts. In this sense,
fundamental concepts include: the design basis loss of coolant accident; the core melt assumptions
which are embedded into the siting policy (Part 100); core melt progression and fission product
release; fission product inventories and biological effects; atmospheric diffusion and transport; offsite
effects; and, historical aspects of important rules such as station blackout.
This one-week course was developed to fill the gap in understanding of reactor safety
concepts. It started with an expression of need from the Director of AEOD to the Director of
Research, in the fall of 1990. The Research office engaged Sandia National Laboratories to develop
much of the work contained herein. Sandia in turn engaged Professor Eric Haskin of the University
of New Mexico who worked with Dr. Allen Camp at Sandia as the principal developers. Over the
last two years the course material has been developed, refined, discussed, and is now ready for trial
use. It consists of five modules: 1) historicld perspective; 2) accident sequence; 3) accident
progression in the reactor vessel; 4) containment characteristics and design bases; and 5) source terms
and offsite factors. Presentation slides have been developed, but are not included in this text,
although copies will be available for the course attendees. Several videos will be shown on topics
of the developing accident sequences, with scale model examples from the severe accident research
program at Sandia. A video on the Three Mile Island event will be shown. Hand calculations on
various accident phenomena (such as core heat up time) will be emphasized. Although most TTC
training courses culminate with a written examination, this Reactor Safety course does not have
exams.
USNRCTechnicalTrainingCenter vii NUREG/CR.6042
ReactorSafer]'course _R-800) ReactorSafet_Course
Comments or criticisms on the enclosed training material are welcome and solicited. We
hope to improve and refine the material and plan to issue a revision in 1995, on the basis of your
comments and experience with the first few course deliveries. We also plan to make this document
available abroadto interested countries and, as is usual atTTC, expect afew foreign attendees atthis
course.
Please direct your comments to the undersigned,
Denwood Ross,
Deputy Director
AEOD
USNRC Technical Training Center viii NUREG/CR-6042
ReactorS_et_ Course_R-800) ReactorSafet_Course
i
ACKNOWLEDGEMENTS
This course covers an extremely wide range of topics. Developing this material required input
from numerous people at the NRC _d elsewhere. In particular, we would like to thank Dr.
Denwood Ross, whose breadth and depth of knowledge concerning the history of reactor safety
was invaluable. Additional information and program guidance was provided by Mark
Cunningham and Lee Abramson of the NRC PRA branch. Other key NRC reviewers included
Ken Raglin, Len Reidinger, Larry Bell, Eric Beekjord, Warren Minners, Jocelyn Mitchell, and
Tom McKenna.
At Sandia, information and insights were provided by Walt von Riesemann, Dana Powers, Susan
Dingman, and Fred Harper. Steve Hodge at Oak Ridge National Laboratory also provided an
excellent review of the material. Finally, we wish to acknowledge the support of Darla Tyree,
Emily Preston, and Ellen Walroth in the preparation of this document.
USNRCTeclmicalTrainingCenter ix NUREG/CR-6042
ReactorSafet Course R-800 ReactorSafe_
English to Metric
Conversion Factors
English Metric
1 Foot .3048 meters
1 Mile 1.6093 kilometers
1 ft. 2 .0929 m 2
t
1 gallon 3.785x10 -3ma
1 ft. 3 .02832 m3
1 Ibm .4536 kg.
1 lbf 4.44822 Newtons
1 psi 6895 pascals
1 Btu 1055 Joules
1 Btu/hr. .2931 watts
1 Btu/hr-ft 2 3.155 watts/m 2
USNRCTechnicalTrainingCenter x NUREG/CR.6042
ReactorSafet_course (R.800) ..... 1. HistoricalPerspective
TABLE OF CONTENTS
1.0 HISTORICAL PERSPECTIVE .................................. 1.0-1
1.0.1 Introduction ........................................ 1.0-1
1.0.2 Learning Objectives for Module 1 .......................... 1.0-1
References for Section 1.0 ..................................... 1.0-5
1.1 1946-1953, Emergence of Safety Strategies .................... 1.1-1
1.1.1 The Atomic Energy Act of 1946 ...................... 1.1-1
1.1.2 Siting .............. ........................... 1.1-1
1.1.3 Containment .................................... 1.1-1
1.1.4 Accident Prevention and Safety Systems ................ 1.1-2
1.1.5 Defense in Depth ................................. 1.1-2
References for Section 1.1 ............................... 1.1-6
1.2 1954-1965 Early Commercial Reactors, Emphasis on
Containment ........................................ 1.2-1
1.2.1 Atomic Energy Act of 1954 ......................... 1.2-1
1.2.2 Early Siting Precedents ............................. 1.2-2
1.2.3 Power Reactor Development Company Construction Permit
Application ..................................... 1.2-4
1.2.4 The Price-Anderson Act and WASH-740 ................ 1.2-3
1.2.5 The First Intervention .............................. 1.2-5
1.2.6 Reactor Site Criteria, 10 CFR 100 ..................... 1.2-5
1.2.7 Credit for Engineered Safety Features .................. 1.2-7
References for Section 1.2 ............................... 1.2-9
1.3 1966-1974 Emphasis on Prevention, Public Debate .............. 1.3-1
1.3.1 Reactor Pressure Vessel Integrity ...................... 1.3-1
1.3.2 The China Syndrome .............................. 1.3-2
1.3.3 The AEC Core Cooling Task Force (CUFF) .............. 1.3-3
1.3.4 General Design Criteria ............................ 1.3-4
1.3.4.1 Criterion 1-Quality Standards and Records ......... 1.3-6
1.3.4.2 Criterion 2-Design Bases for Protection
Against Natural Phenomena ................... 1.3-7
1.3.4.3 Criterion 3-Fire Protection .................... 1.3-7
1.3.4.4 Criterion 4-Environmental and Dynamic Effects
Design Bases ............................. 1.3-8
1.3.4.5 Criterion 5-Sharing of Structures, Systems,
and Components ........................... 1.3-9
1.3.5 The National Environmental Policy Act (NEPA) ........... 1.3-10
1.3.6 Emergency Core Cooling System Rulemaking ............. 1.3-11
1.3.7 The Energy Reorganization Act of 1974 ................. 1.3-13
References for Section 1.3 ............................... 1.3-16
USNRCTechnicalTrainingCenter 1-1 NUREG/CR.6042
