An International Comparison of the Impact
of Safety Regulation on LWR Performance
by
Steven Craig Anderson
MIT-EL 87-006
June 1987
AN INTERNATIONAL COMPARISON OF THE IMPACT
OF SAFETY REGULATION ON LWR PERFORMANCE
by
STEVEN CRAIG ANDERSON
B.S. Mech. Eng., Cornell University
(1985)
SUBMITTED IN PARTIAL FULFILLMENT
OF THE REQUIREMENTS FOR THE
DEGREE OF
MASTER OF SCIENCE
IN TECHNOLOGY AND POLICY
at the
MASSACHUSETTS INSTITUTE OF TECHNOLOGY
June 1987
0
Massachusetts Institute of Technology
Signature of Author
Technology and Policy Program
May 8, 1987
Certified by
Professor Kent F. Hansen
Department of Nuclear Engineering
Thesis Supervisor
Accepted by
Professor Richard L. de Neufville
Chairman, Technology and Policy Program
AN INTERNATIONAL COMPARISON OF THE IMPACT
OF SAFETY REGULATION ON LWR PERFORMANCE
by
STEVEN CRAIG ANDERSON
Submitted to the Technology and Policy Program
on May 8, 1987 in partial fulfillment of the
requirements for the Degree of Master of Science in
Technology and Policy
ABSTRACT
During the decade 1975-1984,
the US nuclear power
industry achieved a lower level of reactor performance
than that realized in many other Western nations.
Previous work suggested that international differences
much
of
the
regulation
account
for
in
safety
discrepancy. US annual regulatory losses averaged over
10% during the ten-year study period.
The present
investigation compares nuclear safety regulation in
France, Sweden, and Switzerland with that in the United
States 1) to determine whether greater regulatory
stringency was indeed responsible for poorer US plant
performance, and
2) to examine key international
differences in the the division and coordination of
responsibility between safety regulators and nuclear
and
solving
technical
utilities
for
recognizing
problems.
Analysis of the US data revealed that, on
average, over 90% of US regulatory outages were
of
the
following:
technical
attributed
to
one
specification limiting conditions of operation or NRCrequired inspections or NRC-required modifications. It
was found that the European nations experienced the
same variety of technical problems seen in the United
Furthermore, the scope and stringency of
States.
It
European and US safety regulation are comparable.
was found that inconsistencies in outage reporting
practices account for much of the discrepancy in
regulatory loss between the United States and the other
Therefore, it is concluded that safety
nations.
regulation is not the primary cause of differences in
reactor performance observed between the United States
and other nations.
Thesis Supervisor:
Title:
Dr. Kent F. Hansen
Professor of Nuclear Engineering
ACKNOWLEDGEMENTS
deserve
organizations
and
individuals
Many
acknowledgement and sincere thanks for the support and
assistance generously provided throughout the various
phases of this project.
Financially, this research was supported by the
US Department of Energy, the Westinghouse Electric
Corporation, the Center for Energy Policy Research at
MIT, and the Technische Universitit Berlin.
and
industry
nuclear
with
interviews
The
regulatory officials were central to the investigation;
many people gave their time in scheduling meetings,
speaking with us, and reviewing drafts of this paper.
the
for
individuals
these
to
indebted
are
We
to
thanks
Special
information and comment provided.
Olofsson
Mr. Pierre Lienart of EdF and Mr. Bengt-G6ran
including
contributions,
many
their
for
KSU
of
coordinating our visits to their respective countries.
Much of this project was completed while studying
at the Technische Universitit Berlin (TUB) in early
Financial support for my visit was provided by a
1987.
(Airlift
Stiftung Luftbrckendank
from the
grant
Special thanks to Frau Roswitha
Gratitude Foundation).
Paul-Walz in the Foreign Students' Office at TUB for
her kind and cheerful assistance with the logistics of
my stay.
The Division for Energy Engineering and Resource
Economics at TUB was my academic home while in Berlin.
I am grateful to Professor Dietmar Winje, for his
personal assistance and contributions to the research.
Thanks also to my colleague Dipl.-Ing. Ulrich Lorenz
interesting
many
and
assistance
daily
his
for
The Institute's technical assistants,
discussions.
provided
and student employees
staff,
secretarial
invaluable and friendly aid with telephone calls,
All these
mailings, and German computer software.
professionally
Berlin
to
visit
my
made
people
successful and personally most enjoyable.
At MIT, many thanks to my advisor, Professor Kent
guidance
and
comments
valuable
his
for
Hansen
Professor
writing and research.
throughout this
Richard Lester also contributed importantly to early
discussions for the project.
Finally, I would like to thank Jeff Jacob and
Rachel Sheffet in Boston and my parents, Carolyn and
Jerry Anderson in Washington, DC, for their personal
assistance in running errands, making inquiries, and
Their
waiting in line on my behalf while I was away.
iii
efforts
and
their
regular
appreciated.
S. C.
A.
May 1987
Berlin (West)
iv
letters
were
greatly
To my parents, for their
love, patience, and support
throughout my life
TABLE OF CONTENTS
Section
Abstract
. .
.
Acknowledgements
.
.
.
.
...
Table of Contents . .
.
.
.
.
.
.
List of Tables and Figures
1.0
2.0
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
.
ii
iii
.
.
.
.
.
. .
.
.
.
.
.
.
.
vi
S. viii
Introduction
. . . . . . . . . . . . .
1.1 Background . . . . . . . . . . . .
1.2 The Importance of Good Performance
1.3 Measuring Nuclear Power Plant
Performance . . . . . . . . . .
. . . .
1.4 Outline of the Investigation
10
Regulatory Impact
on US Plant Performance
2.1
2.2
2.3
2.4
3.0
.
.
. . . .
11
Method of Analysis
. . . . . .
Regulatory Outage Classification
2.2.1 Technical Specifications/
LCO Violations . . . . .
2.2.2 Inspections . . . . . . .
2.2.3 Modifications . . . . . .
Plant Systems Most Affected
by Regulation . . . . . . . . .
Conclusions
. . . . . . . . . .
. .. .
. . . .
11
12
.
.
.
.
17
17
18
.
.
.
.
.
.
.
.
19
22
S .
.
.
23
.
.
.
.
.
.
.
.
.
23
23
25
30
. . . .
34
.
.
38
.
38
US Data --
. .
.
.
.
Source and
France
. . . . . . . . .
3.1 Organizations Influencing
Performance and Safety
3.1.1
Safety Regulators
3.2
3.3
3.4
.
.
.
.
.
.
.
.
3.1.2 Utilities . . . . . . . .
3.1.3
Technical Advisors
. . .
Organizational Relationships
in Practice ..
... ..........
Outage Reporting
and Classification
. . . . . .
3.3.1 General Principles
and Examples . . . . . .
3.3.2 Technical Specifications/
LCO Violations . . ....
3.3.3
Inspections . . . . . . .
3.3.4 Modifications
......
Comparisons with US Experience
.
.
.
.
.
.
.
.
...
.
.
.
.
39
.
.
.
.
.
.
.
.
41
43
.
44
. ..
4.0
5.0
Sweden . . . . . . . . . . . . . . . . . . .
4.1 Organizations Influencing
Performance and Safety .
... . .
.
4.1.1 Safety Regulators ........ .
4.1.2 Utilities . . . .
...........
4.2 Organizational Relationships
in Practice . ....
.. . . . . . . . .
4.3 Outage Reporting
and Classification . . . .
... . . .
4.3.1 General Principles
and Examples ........
.. .
4.3.2 Technical Specifications/
LCO Violations . . . . . . . . .
4.3.3 Inspections . . . . . .
.. . . .
4.3.4 Modifications ..........
4.4 Comparisons with US Experience . . . . .
Switzerland ...
. . .
.. . . . . . . . .
5.1 Organizations Influencing
Performance and Safety . . . . . . .
5.1.1 Safety Regulators . . . . . . .
5.1.2 Utilities .
.....
.....
5.2 Organizational Relationships
in Practice .....
.....
...
5.3 Outage Reporting
and Classification .
.. . . . . . .
5.3.1 General Principles
and Examples . ....
.. ...
5.3.2 Technical Specifications/
LCO Violations . . ......
5.3.3 Inspections . ...
....
. . .
5.3.4 Modifications .
..
.....
5.4 Comparisons with US Experience .....
50
50
51
54
55
59
59
62
65
65
67
70
.
.
.
70
70
73
.
74
.
83
83
.
.
84
85
86
88
6.0
Conclusion .
...
..
............
91
6.1 Data Reclassification
and Conclusions . ....
.. .... . . . 91
6.2 Further Work . ...
....
. . . . . . . 101
7.0
References
8.0
Appendices . . ....
................
103
8.1 OPEC-2 Cause Code List .......
103
8.2 Glossary of Abbreviations Used . ... . . 116
. . . . . . . . . . . . . . .
vii
102
LIST OF TABLES AND FIGURES
Table
1.1
Average Annual Regulatory Loss
.
.
.
.
3
Performance Indices Used in
the Study . . . . . . . . . . .
.
.
9
for Six Nations, 1975-1984
Table
Table
1.2
2.1
Average Annual US Regulatory
Capacity Loss, 1975-1984, for PWRs,
excluding TMI - Classification by
regulatory outage cause . .
Table
2.2
.
....
15
Average Annual US Regulatory
Capacity Loss, 1975-1984,
for BWRs - Classification by
regulatory outage cause . .
Table
2.3
. ...
16
Average Annual US Regulatory
Capacity Loss, 1975-1984, for PWRs,
excluding TMI - Classification by
plant system
Table
2.4
.
..
. ....
.
.
.
20
Average Annual US Regulatory
Capacity Loss, 1975-1984,
for BWRs - Classification by
.
.
21
Organizational Structure
of SCSIN
. . . . . . . .
.
24
Organizational Structure
of IPSN . . . . . . . . .
.
31
plant system
Figure 3.1
Figure 3.2
Figure 4.1
Figure 5.1
Table
6.1
. .
.
.
.
Organizational Structure
of SKI
.
.
.
. .
Organizational Structure
of HSK
. . . . . . . . .
52
.
72
Reclassified Average Annual
US Regulatory Capacity Loss,
1975-1984, for PWRs,
excluding TMI - Classification by
regulatory outage cause .
Table
6.2
....
Reclassified Average Annual
US Regulatory Capacity Loss,
1975-1984, for BWRs Classification by
regulatory outage cause . ...
viii
.
96
.
97
1.0
Introduction
i1.IBackround
Most light water reactors (LWRs) used around the
world for electric power production share substantially
the same design and technology.
As a result, the same
types of technical problems with these power plants are
experienced internationally.
In light of this, it is
curious
of
varies
that the performance
widely from nation to nation.
States,
inferior
for
instance,
to
nations.
from
these nuclear plants
that
performance
achieved
This performance
varying
technical
human
by
is
many
discrepancy
influences
factors.
In
For
the United
significantly
other
likely
rather
example,
Western
arises
than
nuclear
from
plant
management policies and regulatory organizations assume
a widely different
Through
an
character
international
from
nation to
comparison
of
nation.
such
human
structures, insight may be gained into the underlying
causes of poor plant performance in the United States.
Earlier work comparing
the
Federal Republic
of
Germany with the United States identified regulation as
a
chief
between
contributor
to
these
nations.'
two
the
performance
discrepancy
Another
previous
investigation 2 revealed that the United States reports
values of
significantly higher
than
loss*
regulatory
Table 1.1 on page
are observed in many other nations.
3 gives data for countries previously investigated for
the ten-year period 1975-1984.
tempting
infer
to
than
requirements
of
set
other
the
in
regulators
the
do
is
safety
US
the
as
stringent
more
a
enforces
regulator,
NRC,
the
that
1.1, it
in Table
From the figures presented
Some US utilities believe this to be true,
nations.
blaming burdensome regulation as a significant cause of
Before reaching this
poor nuclear plant performance.
conclusion, however, one must verify that all nations
studied have
used
same
the
regulatory
definition of
loss in their reported statistics.
This
investigation attempts
differences
international
as
losses,
such
centers
around
nations
in
constitute
those
the
defining
a
in
regulatory
account for
observed
Table
question
what
to
of
does
loss.
regulatory
in
The
1.1.
inquiry
among
consistency
and
what
Without
the
does
a
not
uniform
definition for these losses, one is left comparing the
*For the uninitiated, an "annual loss" in the
vernacular of performance statistics may be crudely
defined as the total fraction of time that plants are
Regulatory losses (i.e., losses
shut down each year.
due to the requirements of a regulatory body) are one
detailed
More
loss.
annual
this
of
component
information on performance statistics is given in
Plant
Power
Nuclear
"Measuring
1.3,
Section
Performance."
TABLE 1.1:
Average Annual Regulatory Loss,
for Six Nations, 1975-1984
(in percent)
PWRs
BWRs
Federal Republic
of Germany
0.9
11.3
France
0.0
NA
Japan
0.0
0.0
Sweden
4.4
0.7
Switzerland
0.0
0.0
Nation
United States
10.9
10.4
(PWR = pressurized water reactor;
BWR = boiling water reactor;
NA
= not applicable -- France had no BWRs included in
the study)
Source:
Wilson, pp. 290-92.
from
the
of
data
consistency
Once
1.1.
Table
be drawn
may not
inferences
reasonable
incomparable;
is
assured, the relative impact of safety regulation upon
plant performance in these nations will be clear.
An
investigation
earlier
consistent
applied
outage classification criteria to data from the Federal
Republic
of
Germany
and
Upon
States.
United
the
reclassification, the rather surprising result was that
German
regulatory
BWRs)
exceeded
compared
an
on
conventions
in
footing. * , 3
equal
will
investigation
analyze
France,
United
the
in
those
plants,
(for all
losses
outage
Sweden,
and
compare the practice of these nations
United States.
PWRs
States,
The
and
when
present
classification
Switzerland
and
to that of
the
As indicated in the opening paragraph,
the key difference among these nations is probably not
reactor technology, but rather the responses of nuclear
utilities and safety regulators to the complexities and
problems of this technology.
Thus,
two
avenues
of
inquiry
(somewhat
intertwined) are proposed:
results for regulatory
* Citing Hulkower's
from all US and German plants:
As stated: US 10.7% ; FRG 4.4%
After reclassification: US 8.7% ; FRG 10.3%.
4
loss
o
the
Examine
nature
of
outage
in
classification
France, Sweden, and Switzerland vis A vis that in the
United States.
Examine the relationships among the relevant actors
o
in
safety
nation's nuclear
each
regulatory system,
i.e., utilities, regulators, central government, and
public intervenors.
The
environment
organizational
study
of
outage
is
In
classification.
in
the
particular,
the
significant
division of responsibility between utilities and their
will
regulators
safety
influence
rationale
the
for
outage classification in each nation.
With an
understanding
in the
conventions
three
of
classification
outage
European
nations,
comparison may be made with US practices.
process,
a proper
Through this
the unique burden of safety regulation on US
nuclear reactor performance will be revealed.
1.2
The Importance of Good Performance
Nuclear
matter
of
substantial
power
jargon
costs,
plant
and
performance
statistics.
monetary
because of poor performance.
and
is
not
Indeed,
otherwise,
just
there
a
are
incurred
The most tangible and immediate cost is the cost
Aside from hydropower, nuclear
of replacement power.
are
plants
the
for a
expensive baseload units
least
utility to operate (note that the low expense partially
explains
are
baseloading).4
the
meet
to
available
not
When nuclear
facilities
be
must
power
demand,
obtained from higher cost plants, usually oil-, coal-,
but
monetary cost, less
Another
or gas-fired.
just as real,
immediate
is that new plants will have to be
built sooner if the performance of existing plants
Conversely, as Wilson implies, 5 a
worse than expected.
substantial
improvement in performance could
forestall new plant construction.
reserve
margins
is
demand
as
Faced with narrowing
increases
are
expected, many US utilities
actually
than
faster
keenly interested
in
any alternatives to building new plants.
poor performance, primarily
A potential cost of
that
is
economic in nature,
a great number
of
plant
start-ups and shutdowns could hasten the deterioration
This is not to say that plants
of many plant systems.
having such an operating history are unsafe,
some
safety
margins
cyclical stresses
rather,
is
components
that
will
on
may
the
the
be
well
be
by
the
The
surest
effect,
lives
of
affected
plant.
service
narrowed
although
shortened;
more
frequent
replacements will be required, thus raising maintenance
costs.
Furthermore, the entire plant's lifetime may be
becomes
mothballing
abbreviated
if
alternative
than
continued operation
a less
expensive
with
extensive,
ongoing component replacement.
is
cost of poor performance
A non-monetary
the
increase in radiation exposure for utility maintenance
A
personnel.
inspection or maintenance
precautions
safety
by
received
Nonetheless,
tasks
are
taken
people
the
such
fewer
involve
outages
conducted within
many
Naturally,
building.
containment
of
significant number
to
radiological
outages
will
dose
the
minimize
performing
the
the
work.
less
mean
radiation exposure for utility maintenance staff.
1.3
Measuring Nuclear Power Plant Performance
A variety of criteria are employed from nation to
nation to gauge nuclear power plant performance around
the
world.
classes:
1)
These
may
load
availability related.
be
factor
broadly
divided
related,
and
into
2)
two
energy
For this investigation, two such
indices are important in representing the data of the
four countries studied.
The capacity factor (CF) is a
common load-related criterion; the energy availability
factor
(EAF) is a criterion of the second type.
are defined below:
These
NEG
CF
= -----------NER * PH
(1.1)
;
NEG
(----- + EEDH)
NER
EAF = ------------------
;
(1.2)
PH
where
CF
= Capacity factor
EAF = Energy availability factor
EEDH = Equivalent economic derating hours: the total
equivalent full power hours lost for economic
reasons, e.g., load following, fuel
conservation, coastdown to refueling
NEG = Net electrical generation (MW)
NER = Net electrical rating (MW)
PH
= Period hours: for annual CF or EAF, the number
hours in a year
(Any internally consistent set of units is acceptable.)
In nations that baseload their nuclear reactors,
the
economic
losses
negligible or zero.
(EEDH
thirds
France
of
nuclear
fraction,
nuclear
is
the
some
exception.
supply
electricity
With
this
This
policy
Over
two-
comes
from
substantial
load following must
plants.
are
This is, in fact, true in most
a notable
French
sources.
1.2)
equation
In this case, the CF and the EAF
values are very close.
nations;
in
nuclear
be practiced with
results
in
a
4.6
percentage point difference between CF and EAF in the
performance statistics of France.6
For this investigation, the choice of performance
indices for each nation is shown in Table 1.2 below.
Performance Indices Used in the Study
TABLE 1.2:
Country
Performance Index
France
Energy Availability
Sweden
Capacity
Switzerland
Capacity
United States
Capacity
countries
For
capacity and
the
not
the
this
parameter
EAF
thus
(only
required
actually
is
most
the
expresses
plant, whether
is obtainable from the
electricity
demand.
between
EAF is the criterion of choice.
implies,
name
energy that
discrepancy
energy availability is significant
France in this study),
As
the
where
accurately
to
or
meet
assesses
the
capabilities of power plants, as it distinguishes true
performance
potential
from
economic and demand factors.
the
In
external
of
effects
countries where
the
two indices are substantially identical, the choice of
statistics was made according to the
availability and
completeness of data.
Throughout
"regulatory
country,
the
losses."
this
term
availability loss,
report,
If
refers
references
applied
to
to
either
are
a
made
to
particular
capacity
or
When
not
according to Table 1.2.
applied to a specific country, regulatory loss refers
to
the
(negative) regulatory
impact
on performance,
regardless of the means of measurement.
1.4
Outline of the Investigation
Chapter
of
nature
outages
regulatory
including
purposes
2 outlines the
of
data,
For
particular.
in
comparison,
international
the US
chapter
this
identifies the major causes of US regulatory loss and
the plant systems affected by these losses.
The next three chapters, 3, 4, and 5, discuss the
results
Switzerland,
respectively,
and
industry
each
utility
France,
with
regulatory
and
nature
the
officials
Outlined
for
behavior
of
relevant to nuclear safety,
safety regulators and utilities.
and
Sweden,
representatives.
are
country
organizations
in
interviews
of
principally
Outage classification
Finally, comparisons are drawn with
is also analyzed.
the US situation.
Conclusions
particular,
US
are
presented
regulatory
in
are
outages
according to European conventions.
Chapter
6.
In
reclassified
Recalculated values
of US loss are then compared to the figures for other
nations.
and
A summary discussion is given of regulatory
organizational
differences
States and the European nations.
between
United
Some recommendations
for further work conclude the report.
10
the
2.0
Reulatory Impact on US Plant Performance
2.1
US Data --
The
of
portion
Source and Method of Analysis
source
the
of
US
data
Operating
for
Plant
study
this
Evaluation
was
a
-
2
Code
OPEC-2 is maintained by the S. M.
(OPEC-2) database.
Stoller Corporation for the Institute of Nuclear Power
Operations
(INPO).
commercial LWRs
The
database
in the United States
larger than 400
All events at these plants which cause outages or
MWe.
which are otherwise significant are
2.*
all
incorporates
included in OPEC-
The data used for this study were all those events
(a total of 37,492) occurring in the decade 1975-1984.
Central to the OPEC-2 database is the descriptive
numerical coding for each event, hereafter referred to
as the cause code.
The cause code is a fifteen-digit
cipher which describes each event with respect to the
affected,
plant
hardware
event
(both physical
safety
system
considerations.
Appendix 8.1
together
influences
operation,
A cause
describe
and
code
various
each
on
miscellaneous
list
the
particulars of
and regulatory),
showing the
fully
external
is
other
supplied
in
coding levels which
event.
For
this
outages, significant events
addition to
*In
include: major repair or maintenance, safety system
actuation or failure, and any event contributing to the
critical path of an outage.
investigation,
the database management software dBASE
III by Ashton-Tate was used.
Regqulatoy Outage Classification
2.2
As the database manager, the Stoller Corporation
uses
data
from
several
organizations,
including
utilities, equipment vendors, and the NRC in generating
OPEC-2.
Among
these
sources,
classification
schemes
consistency
the database,
in
classification
outages
the
is
criteria
in
a
in
variety
use.
Stoller
have
some
ensure
applies uniform
distingushing
influence
outage
To
from purely technical problems.
utilities
of
on
regulatory
Thus,
OPEC-2
while
outage
classification, Stoller has the last word.
In the
Cause Code
List on page
110
of
Section
8.1, Stoller lists the outage causes that it considers
to be regulatory under the heading, "NRC Originated."
Previous investigators had added some categories* also
thought
be
to
additions are
37,492
US
selected
regulatory
preserved here.
events
5,105
to
from
as
grouping;7
this
Out
1975-1984,
regulatory
of
this
the
their
total
of
investigation
events
for
further
analysis.
*"Fuel
Fuel
Limits
pages
List.
103
and Core -- Safety Restrictions" and "BWR
These are found on
-MCPR, MAPLHGR."
and 110,
respectively,
of
the
Cause
Code
The first major
the
determination of
regulatory loss.
step of this
the most
investigation was
significant
causes
of
Within the regulatory loss category,
further subdivisions may be made according to the exact
cause of the outage, as listed in Section 8.1.
United
States,
safety
limits
regulatory
of
the
outages
technical
are
attributed
specifications
known as limiting conditions of operation
required
inspections
or
modifications,
less frequently observed causes.
In the
(also
(LCOs)),
and
to
to
to
other
These different types
of regulatory outages may be distinguished by sorting
the database
according
to
the regulatory information
contained in the cause code for each event.
In
provided,
their
addition
OPEC-2
urgency.
to
the
regulatory
distinguishes
The
outages
elementary
"forced" and "scheduled."*
classification
according
categories
used
to
are
For regulatory outages, the
relative amounts of forced and scheduled outages serve
as one indicator of the stringency or inflexibility of
regulation.
The
various
causes
of
regulatory
outages
are
listed in descending order of significance in Table 2.1
for
PWRs
(excluding TMI)
and
Table
2.2
for BWRs
on
*In this work, forced outages are those that
could not be postponed beyond the next weekend.
Scheduled outages
could be
postponed beyond the
weekend, but perhaps not until the next seasonal lowload period.
This distinction is a simplification of
that used by the OPEC-2 database.
13
pages
15
and 16
respectively.*
The figures
in each
table are the average annual capacity factors lost due
to regulation during 1975-1984 (in percent).
Note that
the majority of regulatory outages for both plant types
(91.9%
of
attributed
PWR
to
loss
and
the same
94.4%
of
BWR
three causes:
inspections, and modifications."
loss)
can
be
LCO violations,
The nature of each
of these major causes is now explored in more detail.
*Data for PWRs are reported "excluding TMI,"
i.e., not including the lost capacity from the two
Three Mile Island plants. This distinction removes the
distortion of the data due to the prolonged shutdown at
Outages at other plants brought
these two plants.
about by regulatory directives issued in response to
the accident (e.g., "TMI modifications") remain as a
part of these PWR statistics.
category,
"Combination"
that
the
**Note
or
violations
or
LCO
inspections
comprising
modifications in combination with a non-regulatory
cause is included in these percentages.
TABLE 2.1:
Average Annual US Regulatory Capacity Loss,
1975-1984, for PWRs, excluding TMI
(in percent)
Classification by regulatory outage cause
__ICapacit
Outage Cause
Loss (%)
NRC-originated inspections
3.47
NRC-originated modifications
2.33
LCO violations
1.86
NRC licensing proceedings & hearings
0.61
Fuel and core safety restrictions
0.04
Combination
0.04
Unavailability of safety-related
equipment
0.03
TOTAL
8.38%
(Combination category comprises inspections or LCO
violations or modifications in combination with a nonregulatory cause.)
Source:
OPEC-2 Database
15
TABLE 2.2:
Average Annual US Regulatory Capacity Loss,
1975-1984, for BWRs
(in percent)
Classification by regulatory outage cause
Outage Cause
___
Capacity Loss (%)
NRC-originated modifications
4.45
Combination
2.86
NRC-originated inspections
1.61
LCO violations
0.86
Fuel and core safety restrictions
0.32
NRC licensing proceedings & hearings
0.16
BWR fuel limits, i.e., MCPR, MAPLHGR
0.08
Unavailability of safety-related
equipment
0.02
10.36%
TOTAL
(Combination category comprises inspections or LCO
violations or modifications in combination with a nonregulatory cause.)
Source:
OPEC-2 Database
16
2.2.1
Technical Specifications/LCO Violations
The limiting conditions of
operation
one part of the technical specifications.
implies,
(LCOs) are
As the name
they are standing safety limits that
plant operation.
In the United States,
govern
the LCOs
(and
the rest of the technical specification, as well) are
formulated by the utility with input from the equipment
vendor.
Prior to initial plant start-up, the NRC must
approve
the
plant.
Subsequently,
technical
specifications
if an
LCO is
for
the
exceeded
entire
at
any
time, the plant is legally required to shut down.
In
such cases where an LCO causes an outage, the OPEC-2
database attributes the outage to regulation.
2.2.2
Inspections
Inspections can be motivated
ways:
of
by the NRC in
two
1) through surveillance requirements, also part
the
technical
specifications,
which
stipulate
a
certain inspection schedule for critical plant systems,
and 2) through inspection/enforcement bulletins (IEBs),
NRC
orders
requiring
including inspections.
these
two
types
plants
take
action,
often
Note a key distinction between
of
Surveillance requirements
plant, and
to
regulatory
inspections.
are standing rules for each
are in effect from day to day.
IEBs, in
contrast, are ad hoc responses by the NRC to problems
brought to its attention.
Surveillance
requirements
in
the
United
States
sometimes stipulate that the surveillance interval is
to be variable, depending on the
number of
defective
components encountered. .As an example, suppose that a
plant's pipe supports are inspected, and none are found
defective.
In this case,
repeated for
however,
one year.
the
next
the inspection might not be
If one support is defective,
inspection
might
occur
in
three
months; if two are defective, monthly inspections might
be required, and so forth.
Modifications
2.2.3
In the OPEC-2 database, modifications ordered by
the NRC are divided into two classes: 1) modifications
due
to
a malfunction
and
deficiency,
2)
restrictive
criteria.
categories
have
or
a
construction
simplicity,
lumped
been
design
to
due
modifications
For
or
more
these
two
under
together
"modifications," since both embrace the same corrective
for
albeit
measure,
made in response
are
modifications
different
reasons.
Some
to IEBs,
although
the majority are due to other regulatory measures.
Of
particular
1975-1984 were
Mile
Island
regulation.
many
the
(TMI)
interest during
effects
in
1979
of
the
upon
the
study period
accident
US
nuclear
In the two years following the
inspections
at Three
safety
accident,
and modifications were motivated by
18
the NRC through IEBs and other means.
were
responsible
for
a
substantial
regulatory capacity loss in the years
These measures
portion
of
1979 and
the
1980.
Though the data used are not sufficient to establish a
causal
link
outages,
between TMI
many
of
the
problems contributing
and
the
entire
required
increase
changes
in
addressed
to the TMI accident.
In
fact,
many measures implemented at the plants were referred
to as "TMI modifications."
2.3
Plant Systems Most Affected by_Regulation
Table 2.3
(for PWRs,
excluding TMI)
on page
20
and Table 2.4 (for BWRs) on page 21 present the plant
systems responsible for regulatory losses in descending
order of
are
the
significance.
average
regulation
Again, the figures
annual capacity factors
during
1975-1984
(in
presented
lost due
percent).
to
Steam
generators are the components with the most associated
losses
for
containments
PWRs,
are
while
reactor
significant
for
coolant
both
systems
plant
and
types.
These three plant systems account for 70.6% of PWR and
79.2% of BWR regulatory losses.
Note also that economic losses are
in the US statistics.
indeed small
No economic losses are observed
for PWRs and they appear only in the twelfth rank for
BWRs.
This confirms
the
assertion made in
the
last
chapter that the difference between capacity factor and
19
TABLE 2.3:
Rank
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
19
20
21
22
23
Average Annual US Regulatory Capacity Loss,
1975-1984, for PWRs, excluding TMI
(in percent)
Classification by plant system
Plant System
Steam Generators
Containment System
Reactor Coolant System
Condensate/Feedwater System
Core Cooling, Safety Injection
Fuel and Core
Undefined Failure
Refueling and Maintenance
Structural/Intersystem Problems
Turbine
Chemical and Volume Control
Reactor Trip System
Electrical Systems
Circulating/Service Water
Auxiliary Systems
Condenser
Component Cooling Water
Main Steam System
Thermal Efficiency Losses
Start-up, Operator Training
Left Over
Generator
Utility Grid (Noneconomic)
TOTAL
Source:
OPEC-2 Database
_Caaity
Loss
2.74
2.39
0.79
0.48
0.40
0.36
0.36
0.23
0.23
0.12
0.07
0.07
0.05
0.04
0.03
0.01
0.01
<0.01
<0.01
<0.01
<0.01
<0.01
<0.01
8.38%
()
TABLE 2.4:
Rank
1
2
3
4
5
6
7
8
9
10
11
12
13
14
15
16
17
18
Source:
Average Annual US Regulatory Capacity Loss,
1975-1984, for BWRs
(in percent)
Classification by plant system
Plant System
Capacity Loss
Containment System
Reactor Coolant System
Fuel and Core
Core Cooling, Safety Injection
Undefined Failure
Structural/Intersystem Problems
Circulating/Service Water
Electrical Systems
Chemical and Volume Control
Turbine
Refueling and Maintenance
Economic
.Condenser
Reactor Trip System
Auxiliary Systems
Start-up, Operator Training
Condensate/Feedwater System
Main Steam System
4.18
4.03
0.94
0.45
0.14
0.13
0.13
0.09
0.08
0.07
0.05
0.03
0.01
<0.01
<0.01
<0.01
<0.01
<0.01
TOTAL
10.36%
OPEC-2 Database
21
(W
energy availability in the United States is negligible.
2.4
Conclusions
This chapter has established two conclusions on
the
nature
of
US
regulatory
losses
that
will
be
important in the rest of this investigation:
o
The
regulatory
majority
of
US
instruments
loss
are
responsible
LCOs
of
for
the
the
technical
specifications, inspections, and modifications.
o
The plant components contributing most significantly
to US regulatory losses are steam generators, reactor
coolant systems, and containments.
These
observations,
in
combination
with
additional
information on the character of US regulatory outages,
indicate
an
appropriate
focus
remainder of this investigation.
classification
strategies
of
practices
the French,
and
of
inquiry
for
the
From here, the outage
problem
Swedish,
management
and Swiss nuclear
industries may be compared with the US experience.
3.0
France
3.-Organizations Influencing..erformance and Safety
In France,
three
safety of
to the
contributions
organizations make significant
industry.
the nuclear
First, the safety regulator is the Central Service for
the Safety of Nuclear Installations
the
Ministry
de
Electricite
national
Third,
of
France
utility
the
(IPSN) is
Industry
a part of
Research.
Second,
(EdF) is
the
government-owned
operates
all
nuclear
which
Protection
and
(SCSIN),
and
an independent
Nuclear
Safety
Institute
advisory body that provides
expert technical support to both EdF and SCSIN.
close
working
plants.
relationship
is
maintained
A very
among
the
fulfill
two
three bodies.
3..1
aSfet..ty Regulaors
The
duties:
nuclear
SCSIN
was
created
1) to act as
power,
and
2)
in
1973
the official state advocate of
to
ensure
public and the natural environment.
diagram
is
comprises
directors,
given
in Figure
three
expert
four divisions
secretariat.
to
3.1
on
groups,
the
safety
of
the
An organizational
page
24.8
eight
(not shown),
SCSIN
regional
and a general
Minister of Industry
and International Trade
The Protection
and
Nuclear Safety
Institute
The General Director
of Industry
Independent
Technical
Advice
Regional Industry
and
Research Directors
The Central Service
for the Safety of
Nuclear Installations
E XPERT
The Standing
Group in Charge
of Nuclear
Reactors
FIGURE 3.1:
GROUPS
The Standing Group
in Charge of
Installations Other
Than Reactors
The Standing Nuclear
Section of the
Central Commission
of Pressure Equipment
Organizational Structure of SCSIN
Source:
Reference 8.
3.1.2
Utilities
Electricit6 de France, as
utility,
electric
generates
power.
industries
approximately
The
for
the national electric
remainder
internal
90%
is
of
France's
generated by
consumption.
began on EdF's first PWR in 1969.
some
Construction
As of 1985, 37,000
MWe of nuclear capacity were on line, amounting to 65%
of France's electricity supply.
EdF
resources
look
to
maintains
within
SCSIN
a
substantial
base
its organization,
for
technical
and
of
technical
thus does
assistance.
Since
not
EdF
itself is responsible for plant construction as well as
operation, it
is
in the best position to provide for
the safety of the plant, literally from the ground up.
Therefore, EdF maintains a very capable technical staff
to oversee all aspects of safety, in both construction
and operation.
of
technical
At its headquarters, there is a group
experts
that
strive
to
identify
the
underlying causes of current technical problems, and to
recommend appropriate action.
In
addition,
EdF
maintains
a direct
technical
liaison with the NRC in the United States, as well as
with the regulatory bodies of other nations.
EdF has
access to a French database on worldwide nuclear plant
outages, which is used to augment the substantial plant
experience data
from French
plants.
The OECD plant
database and other resources are also available to EdF,
such as those of the IAEA.
Interestingly, EdF was its own regulator for the
its
of
operation
were
reactors
The plants
existed.
any
before
built
EdF's
authority
constructed by
a
then operated by EdF.
and
analytical
gas
the high degree of trust
Such a situation illustrates
placed in
regulatory
were actually
firm, and
national engineering
early
The
plants.
earliest
technical
capacities,
a positive and professional
which today translates to
relationship with SCSIN.
EdF has nearly always enjoyed an excellent public
commitment
twofold
Its
image.
to
safety
(France
containment has won the utility much support.
has
had
typically
place
to
work,
and
of
some
attracting
trouble
cost
electricity
in
Also important, EdF is considered a
Western Europe.)
prestigious
lowest
the
cost
and
the
the
has
company
most
no
talented
engineering graduates.
So far, EdF has not been content to rest on its
Its policies and practices have continued to
laurels.
stress and achieve safe, economical operation.
of
the
French
high
plants,
experience
pursued
degree
EdF
analysis.
with
input
of
has
design
much
This
from
to
standardization
gain
program
operating,
26
Because
from
is
among
operating
aggressively
management,
and
and
characteristics
evaluation
in
helps
problems
of
plant
generating
the next generation of
design of
the
for
guidelines
Their
personnel.
construction
plants.
As
observed,
EdF
is
committed
to
economical
Characteristically, EdF responds
operating practices.
to problems in a plant with a "temporary fix," a remedy
that safely suffices until the next refueling outage,
not necessarily a repair acceptable in the long term.
In this way, technical resources may be brought to bear
on
the
repairs
comprehensive
later
may
the
with
conjunction
more
the
time-consuming,
be
conducted
in
outage,
refueling
annual
and
controlled,
unhurried,
Also,
way.
organized
an
in
problem
significantly increasing the availability factor.
There
are
cases,
of
course,
where
the
problem
cannot wait until refueling, and interim measures are
For the examples of valve replacement
unsatisfactory.
or steam generator leakage, the start of the outage may
be able to be delayed a few days to a week, in order to
coincide
with
electricity
on,
among
other
repair
The
demand period.
other
things,
work
the
that
such
technical problems;
hardware
with
a
decision will
historical
problem and the season of the year.
note
or
lower
depend
trend of
the
It is important to
problems
are
considered
the regulator is never blamed for
27
any resulting outage.
This point will be
elaborated
later.
The primary concern of most of EdF's policies is
safety.
As
an example,
consider
the
complete plant
evaluation/inspection performed on each power
station
upon completion of its first year in service.
This is
the
same
comprehensive
performed
every
elsewhere.
ten
inspection
operating
that
years
is
in
commonly
France
and
EdF believes that no other nation's nuclear
industry
performs
the
same
complete
one-year
evaluation.
EdF
that
is
Reactor
safety
maintains
an
independent
the equivalent
Safeguards
committee
often the
of
the
(ACRS) in
addresses
safety
committee
Advisory Committee on
the
United
current
States.
plant
problems,
same issues which the regulator,
studying.
An
example
of
the
The
SCSIN,
committee's
is
action
occurred in response to a report of ruptured tube guide
pins,
first from one plant, then from a second.
The
problem was especially vexing, because the plants were
of
different
analyzed
the
generations.
failures
EdF's
safety
and recommended
to
committee
SCSIN
that
staggered replacement be accomplished during refueling.
SCSIN accepted the committee's proposition.
Another
response to
example
of
an
problems caused
28
EdF
by
initiative
the
came
in
severe winter of
1986-1987.
Several plants experienced instrumentation
problems due to the cold weather.
were
not
very
disturbing
to
severe
EdF
in
as
These difficulties
themselves,
but
they
a possible indication
were
that
the
effects of a cold winter on its plants were not well
understood.
Accordingly,
EdF
instituted
a
comprehensive review of the impact of cold weather on
many aspects of plant operation.
not
only the
designed
This action addressed
instrumentation problems,
to
foresee
and
prevent
but
other
was
also
malfunctions
induced by extreme weather.
To conclude this look at the nature of EdF, the
utility's
coherent
outage
philosophy is noted.
management
and
safety
Consistent with the emphasis on
the technical nature of nuclear safety, EdF maintains
excellent
engineering
organization.
Also,
resources
when
in
technical
its
own
problems
cause
plants to shut down, the outages are blamed on faulty
equipment rather than on a capricious regulator or an
unreasonably stringent specification.
technology
extends
to
the
close,
relationships with SCSIN and IPSN.
these
bodies,
engineers
non-technical bureaucrats
Finally, despite
SCSIN
oversight
capacities
of
do
IPSN,
cooperative
In discussions with
talking;
lawyers
do not play pivotal
the checks
and
the
This emphasis on
the
roles.
and balances afforded by
independent
responsibility
29
and
for
technical
plant
safety
rests foremost with the plant operator.
policies
have
served both
EdF
and
To date, these
the
nation
quite
well.
3.1.3
Technical Advisors
The
Protection
(IPSN) is
one of
and
nine
Nuclear
Safety
institutes
Atomic Energy Commission (CEA).
within
nuclear
within
industry,
its
its
the
French
An organization chart
is given in Figure 3.2 on page 31.9
responsibilities
Institute
IPSN has diverse
mandate,
major
but
for
contributions
the
lie
in
research and development, reactor safety, and radiation
protection.
IPSN is the main technical support for the
regulatory
body, SCSIN,
and
also
has daily
contacts
with EdF.
IPSN
was
created
by
ministerial
decree
in
November 1976 as a focus for CEA's efforts in radiation
protection,
nuclear
safety,
and
safeguards.
The
organization serves the nuclear industry in particular,
but
also
provides
Ministries of
Affairs
and
Environment.
technical
assistance
to
the
Industry and Research, Health, Internal
Decentralization,
IPSN employs close to
Transport,
and
1500 people;
its
1986 budget was approximately FF 1 billion.'0
IPSN
advise
is
indeed
well-qualified
others
in the
field
30
of
and
equipped
reactor safety.
to
Four
General.
Management
IPSN Director
Nuclear Safety
Public Relations
-----------------------------------
Services Office
Deputy
Director
and Communication
Director
of
o
'Research
Health Protection
Safety Analysis
Safeguards
Division
Division
Division
I
L
II
Technical Protection
Safety Research
Division
Division
Central
Decommissioning
Division
FIGURE 3.2:
Organizational Structure of IPSN
Source: Reference 9.
research
and
reactors
facilities
(including
many
experimental
critical
facilities
and
test
and
test
loops) are utilized by IPSN in its research programs.
More
300
than
analysis
technical
studies
for
These
Research.
perform
specialists
Ministry
the
activities
safety
of
Industry
and
provide
detailed
and
sophisticated technical knowledge that is vital for the
drafting and implementation of appropriate regulations.
Nuclear safety
regulation
takes
three
different
forms in France:1 '
o
Ministerial orders and decrees
o
Technical
specifications
to
ministerial
support
recommendations
o
Guidelines from component vendors
IPSN is intimately involved in preparing both technical
A part of
specifications and vendor guidelines.
technical
specifications
is referred
to
as
the
the Basic
SCSIN prepares these rules based
Safety Rules (RFS).
on the outcome of research programs conducted by IPSN
specialists.
The vendor guidelines include the Design
and Construction Rules
SCSIN
approval
evaluates
makes
a
by
(RCC),
the
the prudency of
recommendation
which are submitted for
plant
the
to
32
IPSN
vendor.
proposed standards
SCSIN.
The
RCC,
then
and
in
are
particular,
They
guidelines.
prescriptive
not
allow plant-specific, innovative solutions, and permit
plants to keep pace with technological change.
EdF
and
IPSN
are
operators
safety.
reactor
for
responsible
fundamentally
plant
that
agree
Accordingly, safety analysis is chiefly based on EdF's
Operating experience data from
studies and research.
similar
in
useful
safety
addressing
The high degree of plant standardization in
questions.
however,
efforts,
maintaining
its
valuable.
data resource especially
this
France makes
EdF's
also
is
plants
own
do
not
research
discourage
program
to
IPSN
from
check
and
substantiate the utility's work.
IPSN's
Much
of
of
scope
its
addresses
work
is
truly
PWR
power
research
impressive.
plants
in
Included in these efforts are research on
particular.
the behavior of structures and components, reliability
and
of
probability
failure
factors
human
considerations.
studies
of
area of
interest gained additional
the
Three
studies,
Mile
Island
of
in PWRs receives much attention.
various
behavior
the
IPSN
test
facilities
This
last
significance after
accident.
the thermohydraulics
special
and
analyses,
Concerning
two-phase
risk
transients
Other research at the
includes
work
on
fuel
(including conditions of major fuel damage),
behavior
of
cesium
and
iodine
aerosols
under
containment
various
containments.
of
venting
filtered
Finally,
joint
research
IPSN
and many
between
are underway
programs
the
and
conditions,
Western
nations, as well as the Soviet Union.12
Organizational Relationships in Practice
3.2
The chain of responsibility for reactor safety in
EdF
France begins with each plant's operating staff.
as
only
then
regulator
closest
the next
a whole has
by
SCSIN,
itself
has
competent
people,
technical
access to the exhaustive
also has
The
regulator.
safety
the
followed
oversight,
and
IPSN.
resources of
EdF and the individual plants, however, have the most
detailed,
plant-specific
safety
Hence,
information.
the technical opinion of EdF is believed and respected
by
the
other bodies
This
industry.
involved
trust
in
French
the
professional,
a
facilitates
nuclear
constructive, and technical dialog among all parties.
There are several examples of SCSIN finding fault
with EdF's procedures and standards, but even in these
cases, SCSIN is not considered to have caused any plant
outages
that
may
have
resulted.
This
is
largely
because of the mutual respect felt by EdF and SCSIN for
their respective
promptly
informs
roles in the nuclear
SCSIN
any
of
industry.
problems,
and
EdF
readily
takes the initiative in proposing solutions, preventing
the
necessity
of regulatory
34
intervention.
There
are
legal requirements for EdF to notify SCSIN immediately
of unusual events at power plants; however, EdF appears
to be voluntarily more forthright than required by law.
After notification, EdF typically presents an informal
proposal for remediation of the problem to SCSIN.
action
is
opinions
intended
and
to
to
encourage
prepare
the
the
way
airing
for
This
of
all
building
a
consensus.
Differences of opinion between EdF and SCSIN are
actually
rather
common.
established between
these conflicts.
the
A
two
joint
committee
organizations
to
is
resolve
This group performs its function well
(and usually peacefully), as no stalemates or arguments
remain
over
the agreements
forged by this
committee.
As a result, EdF has never been seen by the public as
challenging SCSIN's policies or procedures, or as being
generally
unruly,
argumentative,
and
uncooperative.
Thus, the participatory process of conflict resolution
in
a
professional
atmosphere
minimizes
any
later
dissent.
An example of
wrote
a proposal
this
to
process may be
SCSIN
requirements
for
independently
reviewed
safety
it,
revising
system
and
some
the
cited.
EdF
surveillance
testing.
IPSN
relatively minor
differences of opinion surfaced among the three bodies.
The collaborative committee was successful, however, in
then
was
proposal
The
disagreement.
the
resolving
accepted with minor amendments.
and
groups,
advocacy
citizens,
Private
governmental bodies have never intervened substantially
in
the
individuals
private
action, either
ever
have
EdF or
SCSIN.
motivated
an
of
decisions
and
actions
No
SCSIN
SCSIN is
directly or indirectly.
not
legally required to act upon or even to acknowledge any
the legislature,
Committee.
This Committee is
usually occurs
this
and
memoranda
free
than
through
rather
calls
there
In response to the
Parliamentary
some
were
conditions
regarding
discussions
to ask questions,
such as
telephone
accident,
supplied.
power
French
at
EdF, which was
Some information was asked of
plants.
its
by
via informal means,
formal hearings or investigations.
Chernobyl
no
the Parliamentary Energy
overseer in
but
been
affairs
EdF's
of
made
inquiries
protracted
have
There
public.
the
from
petition
Parliament seemed satisfied, as no further
action was taken.
EdF
and
IPSN
relationship.
encompasses
and
IPSN's
SCSIN.
EdF
The
technical
advice
IPSN,
scope
have
appear
to
of
their
discussions
comment
and
When there is
and
also
on
sound
communication
reactor
on EdF's
a
safety,
proposals
to
a difference of opinion between
SCSIN
must
choose
between
the
recommendations, or create its own compromise.
EdF and
IPSN disagree fairly often in this process, so SCSIN's
judgment is frequently necessary.
One avenue through which the personnel of EdF and
IPSN
have
extensive
liaison engineer
contact
is
through
from IPSN who is
the
on-site
present during the
start-up of any plant.
This liaison is available for
technical
does
duties.
support
and
not
perform
inspection
The purpose of having the engineer on-site is
to give IPSN a firsthand knowledge of activities at the
site,
not
to
analyze
understand
the
plant
considers
IPSN
deficiencies.
the
depth
and
for
possible
it
important
of
the
variety
to
technical
problems facing EdF, as well as the utility's solution
strategies.
during
There is, of course, much more happening
start-up
oversee;
preparation
all
therefore,
one
activities
person
are
can
assigned
In this way, the liaison can concentrate
priorities.
effort on
than
the most
critical
aspects
of
the
start-up
of
start-up
period.
IPSN does conduct
activities.
Yet,
its
own analysis
is
EdF's
it
responsibility
to
recognize any difficulties and to report them promptly
to
IPSN
so
punctually.
prepares
that
its
analysis
can
be
performed
With input from the liaison engineer, IPSN
a status report
for
SCSIN
and
an
advisory
paper for EdF.
These official documents are preceded
by much informal discussion amongst the three bodies.
Thus, the content of the formal reports is no surprise
to anyone, and rarely creates any controversy.
Another example of EdF/IPSN cooperation includes
a
joint
research
effort
extending
plant
licensing
analyses to include beyond-design-basis accidents.
1979-1980, the
In
two organizations adopted the practice
of routinely planning for these accidents.
They view
such collaborative research as important, not only for
enhancing
cooperation
duplication of
but
effort.
also
for
preventing
In the same spirit, even the
plant vendor Framatome joins EdF and CEA in dividing
and coordinating the research agenda.
3.3_ -OutageReportina
3.3.1
and Classification
General Principles and Exampmles
France
reports
no
regulatory
reactors experience the same sorts
those
in
the
United States,
but
losses.
French
of difficulties as
the
French industry
classifies these problems differently from US industry.
There
is,
in
France,
no
motivated by a regulator.
such
thing
as
an
outage
The various requirements of
SCSIN are automatically assumed to be reasonable, just,
and
appropriate.
When
some
component
of
equipment violates one of SCSIN's requirements,
plant
it is
equipment
the
is
that
resulting outage, not the requirement.
EdF representative,
any
for
accountable
held
According to an
labeling outages as regulatory is
deemed "unwise," and is hence not practiced.
Alternatively, human error can cause outages and
is
thus
example,
also cited by
EdF
tests
required
some
cause.
For
delicate
and
as an outage
are
very
sensitive, offering many opportunities for human error
or misjudgment to cause a reactor trip.
Any resulting
outages are, however, attributed to human error, rather
than to the set of regulations prescribing the testing.
Another cause of outages reported by EdF pertains
to
of
violations
axial
offset
are
These
limits.
treated as "administrative" limits; the plant operator
manages the reactor so as to stay within them.
are
viewed
by
some
in
EdF
as
These
regulatory
causing
outages, but these outages are never reported as such.
3.3.2
Technical_._Specifications/LCO Violations
Technical specifications are written by EdF.
most
before
cases,
their
their
content
formal
is
issuance.
discussed
Thus,
In
with
SCSIN
the
final
documents are no surprise to the regulator, and contain
standards upon which all parties have agreed.
SCSIN is
free to ask questions or propose modifications to
the
specifications.
Any
such changes
are
effected
in
a
joint EdF/SCSIN committee.
The baselines for establishing
were
specifications
technical
modeled after
the US reactor vendor, Westinghouse.
built before
the first French
those of
All French plants
1982 were constructed by Framatome
(the
French nuclear vendor),
using what was substantially a
Westinghouse
Even
design.
today,
French
technical
specifications are quite similar to those for US PWRs.
EdF personnel were confident that the stringency of the
two nations'
specifications
is most often comparable,
with French standards more stringent in some areas.
Technical specifications are implemented through
EdF policies and observed by the plant operators.
Each
critical
parameters
indicative of the reactor's physical state.
If an LCO
operator
monitors
trends
in
the
by
less
drastic means, the operator shuts down the plant.
Many
imminent
appears
violation
and
unavoidable
times, however, the LCOs are never closely approached
is because EdF observes a set of
This
in operation.
operating specifications that are often more stringent
than
the
LCOs
specifications.
in
the
technical
EdF adopts these conservative policies
in the interest of
years.
contained
In order
achieving plant lifetimes of
to
realize
reliable
operation
30-40
over
this period of time, EdF believes that both operating
and
maintenance
practices
must
be
painstaking
and
exacting, erring only on the side of conservatism and
prudency.
One notable case may be cited of a disagreement
between
EdF
and
SCSIN
over
technical specifications.
to
the
French
border
the
implementation
In a power plant quite close
with
the
Federal
Republic
Germany, SCSIN ordered a shutdown before
closely approached.
location of
but
did
outage
was
delay
of
any LCO was
This was done due to the sensitive
the plant.
not
of
in
EdF disagreed with the order,
obeying
considered
a
the
directive.
technical
This
specification
violation, as SCSIN's judgment is never blamed for an
outage.
Thus, SCSIN merely acts to call attention to
objective
conditions
in the plant.
Such
conditions,
when evaluated using standards and criteria accepted by
EdF and SCSIN, may necessitate a shutdown.
3.3.3
Inspections
During regular plant operation,
present.
no inspector is
Frequent plant visits are preferred instead,
customarily two per month.
The date of the inspection
and the agenda for the visit are established in advance
to
ensure
that the
plant
staff
and
other
technical
resources are available for discussions and analysis.
Nearly
all
accomplished
in
particularly
the
appreciate the
note
that
routine
inspections
conjunction
annual
with
in
France
other
refueling
are
outages,
outage.
To
success of French inspection policies,
the equivalent energy
availability lost in
1986 due to routine inspections of French 900 MW PWRs
was
only
0.08%.
(1975-1984) of
For
comparison,
the
average
annual capacity loss due to
"regulatory" inspections
for US
PWRs
value
so-called
(excepting TMI)
was 3.47%, over 43 times greater.
One explanation for the discrepancy in inspection
outages
lies
in
the
nature
surveillance requirements.
of
the
two
nations'
Unlike the United States,
France does not have a variable surveillance interval
that depends on the number of component malfunctions.
Furthermore,
France
has
inspection/enforcement
apparently
more
analogous
bulletins.
formidable
requirements, standards
no lower.
nothing
Yet,
set
of
US
to
the
US
despite
an
regulatory
demanded in French plants are
For example,
recall
that a comprehensive
ten-year inspection is conducted in each French plant
after only one year of operation.
This inspection is
conducted as a matter of policy by EdF;
result of any regulatory directive.
42
it is not the
3.3.4
Modifications
France
did
not
make
as
many
modifications
in
response to the TMI accident as did some other nations.
Some modifications that were typical reactions to TMI
in other countries had fortuitously been effected by
France
prior
to
the
accident.
implemented in France after
For
those
changes
the accident, a concerted
effort was made by EdF and SCSIN to schedule the work
during
annual
devotes
refueling
many
outages.
resources
to
In
outage
planned
outages.
Many
of
EdF
planning
scheduling, to minimize forced outages
of
general,
and
and extensions
EdF's
post-TMI
modifications addressed human factors concerns in the
control room.
EdF representatives could not recall any post-TMI
modifications that were too urgent to be delayed until
refueling.
There were a few plants where cracks were
discovered in steam generator outlet piping, but this
was
With
probably
not
TMI-motivated,
certain time restrictions,
devise
its
cases,
the
refueling.
own
schedule
work
for
according
to
SCSIN allowed EdF
remediation.
necessitated
EdF.
shutdowns
In
to
some
prior
to
Nonetheless, the outage cause reported was
pipe cracking rather than a regulatory order.
It
is
likely that 1) the high degree of standardization among
French
electric
plants,
and
utility,
2)
EdF,
the
were
43
existence
significant
of
only
one
reasons
why
SCSIN
allowed
EdF
such
scheduling repairs.
discretion
and
Most any problem is
freedom
in
likely to be
generic in the plants, as all share the same types of
components.
Furthermore, EdF's standards and policies
apply to all French plants; these standards are wellunderstood by SCSIN.
3.4
Comparisons with US Experience
As
noted
in
specifications,
specifications
the
EdF
more
established LCOs.
section
observes
a
stringent
than
on
set
technical
of
the
operating
officially
These company standards are adopted
voluntarily, without regulatory pressure; EdF considers
them one element of prudent engineering practice.
This
is not the case in the United States.
Regarding
inspectors
at
France
inspections,
the
no
plants,
has
variable
no
resident
surveillance
intervals, and nothing analogous to an IEB, unlike the
United
States.
SCSIN,
French
between
Yet,
inspection
the
efforts of
requirements
are
EdF
similar
and
to
EdF appears to compensate
those in the United States.
for the less demanding regulatory surveillance with its
own
policies.
For
comprehensive ten-year
example,
performing
the
inspection after only one year
of operation is an EdF practice not seen in the United
States.
44
SCSIN never
forced a modification outage
upon
EdF.
At most, SCSIN's requested changes were discussed
with
the
utility,
regulator's
impositions
and
implemented
satisfaction.
have
by
Indeed,
ever been made
EdF
no
to
regulatory
in France;
conflicts
between the regulator and utility are always
via
technical
technology
discussions.
frequently
In
plays
a
the
resolved
the
United
muted
role
States,
in
such
dialogs, where legal concerns dominate the agenda.
It is apparent that while French regulations are
in
some
makes
areas
up
the
practices.
stringent
usually
As
observed,
the
EdF's
considered
if
an
with
these
that
SCSIN
standards,
a
than
regulation
policies
Even
EdF's
intrusive
difference
than
outages.
beyond
less
technical
US
its
statutes,
voluntary
are
Thus
more
it
responsible
requirement should
any
safety
frequently
itself.
are
EdF
consequent
difficulty
and
is
for
reach
outage
is
never
a
regulatory imposition.
Four fundamental differences between
the US
and
French nuclear industries should be borne in mind when
making regulatory comparisons:
o
Plant age.
Western
average,
French plants
Europe)
than
are
US
(indeed, those in all of
significantly
plants.
45
Many
younger,
of
the
on
most
troublesome
problems
(steam
generator
leaks,
for
instance) worsen with age.
o
Plant standardization.
With the tremendous degree
of standardization observed in France, data from the
detailed inspection of one plant may be statistically
extrapolated
to
all
specific features.
be
other
generic.
implies
The
the plants.
that
resources
concentrated on these
yet probably
barring
plant-
Hence, detailed inspections may
distributed among all
standardization
plants,
most
of
every problem
EdF
difficulties
widespread)
Furthermore,
rather
may
thus
is
be
(relatively few,
than on
a host
of
local problems at each plant.
o
Utility diversity.
are well-known
EdF is a monolith.
to SCSIN and to
IPSN.
Its policies
Furthermore,
these policies are applied consistently to all plants
(this is
turn).
SCSIN
partly
facilitated
by
standardization,
Thus, with a minimum of effort and inquiries,
may
be
confident. of
understanding
problem at any plant will be attacked.
not
be
in
said
of
the
NRC's
position
how
any
The same may
in
the
United
States.
o
Utility size.
EdF's
size is an asset in achieving
and maintaining a high level of technical competence.
In the
United
utilities
States,
with
there
varying
46
are
sizes
dozens
and
of
nuclear
technical
capacities.
Some small utilities do have excellent
technical resources, but in general, small size and
meager
budgets
are tough obstacles
to
establishing
and maintaining high-caliber technical abilities.
There remain significant differences between the
industries of
the two nations, however,
plausibly
explained
instance,
the
by
many
the
factors
positive
that are
not
above.
characteristics
For
of
the
EdF/SCSIN relationship in France form a stark contrast
with the nature of the typical utility/NRC relationship
in
the
United
professional,
States.
mutually
In
the
respectful,
oriented environment prevails.
the
atmosphere
is
French
instead
and
industry,
a
technically
In the United States,
characterized by mistrust,
litigation, and poor plant performance.
The French utility, EdF, is technically excellent
in its own right, independent of
SCSIN or IPSN.
The
utility is treated as a technical peer of the regulator
or IPSN.
Because of this common level of competence, a
permanent technical dialog exists among the three, one
that
evolves
to
address
new
issues
as
they
Again, the US situation is sharply different.
utilities have
NRC is
reactor
arise.
As many
only mediocre technical abilities,
the
relied upon for the analysis and assurance of
safety.
technical
The
analysis
NRC
also
functions
houses
under
regulatory
one
roof.
and
In
France,
these
roles
are
filled
by
separate
organizations, SCSIN and IPSN respectively.
Technical
analysis
is
much
nontechnical
less
factors
likely
when
to
the
be
two
tainted
functions
by
are
separated.
IPSN and SCSIN officials shared opinions of the
US
nuclear
relation
industry,
to
that
utilities
of
and
France.
NRC
Regarding
alike,
in
utilities,
several representatives perceived that the US utilities
were
generally
more
reticent
than
EdF
to
accept
responsibility for the technical state of their plants.
Additionally, utility management in the United States
was
thought
to
suffer
from
inadequate
organization.
Regarding outage reporting practices, one EdF official
felt
that
stringent
US
utilities
regulation
readily
for
what
technical problems in France.
blamed
would
excessively
be
considered
Addressing and avoiding
such technical problems at the plants would be simply
part
and
parcel
of
EdF's
voluntary
engineering
practices.
Concerning the NRC, EdF representatives perceived
US
regulation
prescriptive.
as
"going
It
was
prescription
may
convenient)
when
operated
a wide
by
be
too
also
recognized
required
dealing
variety
far,"
(or at
with
of
or
being
that
least
nonstandard
utilities.
too
greater
be
more
plants
Whatever
technical expertise that does exist at nuclear plants,
however,
tradition
is
of
rarely
the
sought by
resident
the NRC.
inspector
Lastly,
at
US
the
plants
epitomizes the mistrust between regulator and utility,
and points to the dearth of a professional relationship
between them.
49
4.0
Sweden
4.1
Organizations Influencing Performance and Safety
The Swedish Nuclear Power
Inspectorate
(SKI) is
Sweden's nuclear safety regulatory authority, analogous
to the NRC in the United States.
Other administrative
and
are
scientific
bodies
in Sweden
responsible
for
related nuclear safety issues, namely:1 3
o
National Institute of Radiation Protection
Established
by
the
Radiation
works closely with SKI
Protection
(SSI).
Act,
to ensure safe conditions
SSI
at
all nuclear facilities.
o
Conducts testing
Swedish Plant Inspectorate (SA).
of pressure vessels at nuclear installations.
o
National
Board
for
Spent
Nuclear
Fuel
(NAK).
Oversees technical research and coordinates financing
for the handling and disposal of spent fuel.
o
National
Created
by
Environmental
the
Protection
Environment
Board
Protection
(SNV).
Act,
SNV
monitors non-nuclear disturbances in the vicinity of
nuclear facilities.
o
National Board of Occupational
National
Electrical
Safety and Health,
Inspectorate.
These
bodies
exercise the same surveillance duties at nuclear and
non-nuclear installations.
In addition, each of the four municipalities having a
nuclear power station nearby maintains a local
committee.
proposed
safety
These groups keep informed of current and
nuclear
regulations
at
committees
also
safety
their
and
radiation
respective
perform
power
public
protection
plants.
The
information
and
emergency planning duties.
4.1.1
Safety Regulators
The
principal
regulator,
are
functions
nuclear
of
SKI,
facility
the
safety
licensing
and
oversight, the promotion of safety, and the supervision
of
handling
and
storage
of
fissionable
nuclear
SKI and the utilities both stress, however,
material.
that primary and direct responsibility for plant safety
lies
with
the
Supplementary
technical
and
utility
duties
research
communicating
and
of
operating
SKI
include
the
coordinating
development in nuclear
nuclear
safety
plant.
information
safety
to
the
public.
An organization chart is given for SKI in Figure
4.1 on page 52. 1 4
government;
The SKI Board is appointed by the
the Director General serves
as its chair.
Reporting to the Board are the two technical Offices,
I
Reactor Safety Committee
BOARD
'H
Safeguards Committee
I-
DIRECTOR GENERAL
I
Information
Secretariat
Research Committee
Office of Inspection
Office of Regulation
Barseback
Safety Assessment
Forsmark
Nuclear Waste
Oskarshamn
Analysis
Ringhals
Research
Department
of
Administration
Nuclear Materials
FIGURE 4.1:
Organizational Structure of SKI
Source:
Reference 14.
the Office of Inspection and the Office of Regulation,
as well as the Administration and Information sections.
The key functional distinction between the two offices
is that the Office of Regulation is responsible for the
text
and
Office
these
of
specifications
Inspection is
regulations.
of
regulations,
while
the
concerned with enforcement of
The
Inspection Office
liaison inspector for each plant.
assigns
one
This official is not
an inspector in residence at the site
(there are none);
instead, frequent visits are made.
Also
indicated
in
the
chart,
three
advisory
committees are a part of SKI: 1 5
o
The Reactor Safety Committee keeps informed of SKI's
supervisory activities
and provides technical
advice
on reactor safety and licensing matters.
o
The Safeguards Committee advises SKI on safeguarding
nuclear material,
including measures to combat theft
and sabotage committed against a nuclear facility or
a transport vehicle.
o
The
Research
Committee
research projects
proposes
and
and is available as
evaluates
an advisor
to
the Research Division.
SKI
power
is
financed
utilities
and
with
with
monies
funds
from
the
allocated
nuclear
by
the
government.
The total budget for budget year 1984/85
was SEK 25,000,000
(roughly $3 million at that time).
An additional SEK 44,400,000
at
that
time)
research.1
was
(approximately $5 million
allocated
for
nuclear
safety
s
SKI
currently
professionals
and
has
25
about
85
supporting
employees:
personnel.
60
All
employees work at a single location in Stockholm.
4.1.2
Utilities
Sweden has
plants:
three
owning nuclear power
the state-owned Swedish State Power Board, and
(at least
Kraftgrupp AB,
1980,
four utilities
partly
partly) private
Sydkraft AB,
in
response
utilities,
and OKG
to
the
Forsmarks
Aktiebolag.
Three
Mile
In
Island
accident, the four nuclear utilities formed the Nuclear
Safety Board of the
1987,
RKS
merged
Swedish Utilities
with
the
(RKS).
personnel
As
of
training
organization run by the utilities to form the Nuclear
Training and
the
Safety Center
utilities'
safety
(KSU),
which today houses
collaboration
efforts
and
training programs.
KSU both coordinates the in-house safety efforts
of the individual utilities and conducts its own safety
projects,
utilities.
drawing
on
the
Much
of
KSU's
combined
54
resources
attention
is
of
devoted
the
to
managing
the
operating
data from
experience
feedback
program,
domestic and
whereby
foreign plants
are
collected, analyzed, and disseminated in useful form to
the
nuclear
provide
utilities.
each
plant
Experience
with
feedback
a relevant
set
of
aims
to
data
on
technical disturbances that can serve as an information
resource in
anticipating and
addition,
KSU
and
planning
and
investment
the
resolving
utilities
in
problems.
emphasize
high
In
outage
quality,
high
availability measures.
In discussions
with
SKI,
the
safety regulator,
and KSU, the utilities' own safety organization, both
groups characterized the regulator/utility relationship
as
positive,
technically
oriented,
and
cooperative
rather than adversarial.
The two bodies seem jointly
committed
viable
Sweden.
the
to
a safe
and
nuclear
industry
in
To this end, SKI takes pains to give utilities
maximum
regulatory
possible
actions,
advance
so
notice
that each
plant
of
upcoming
may
schedule
outages most efficiently.
There was no instance of
extreme that it forced a plant
noted,
however,
(usually
that
refueling)
plant
had
a regulatory order so
to cold shutdown.
start-up
sometimes
from
been
regulatory action (see next paragraph).
55
KSU
a shutdown
delayed
by
KSU was asked
SKI
how
was
stringency
to
likely
of regulations,
changes
implement
the
in
for example, modifications
arising from revisions in computer
seismic
codes for
KSU felt that the utilities would be allowed
analysis.
a reasonably long time (i.e., until the next refueling)
within
which
to schedule
with
contrasts
criteria,
design
seismic
NRC's
the
the
necessary
the
of
handling
in
revisions
This
work.
of
issue
were
which
sufficient cause for near-immediate shutdowns.
In the opinion of KSU officials, the most extreme
regulatory
action
taken
by
SKI
occurred
the
upon
discovery of cracked piping at the Ringhals 1 plant in
August 1986.
defect
This
was uncovered in
week of the refueling outage.
the
last
As a result, Ringhals 1
was down for four more weeks for inspection and welding
work.
SKI also issued a statement giving notice that
all BWRs
would be inspected within a short period of
time.
Another anecdote
illustrates
the
usual
outcome
when there is a technical difference of opinion between
In the Oskarshamn 3 and Forsmark 3
a utility and SKI.
plants, the main steam isolation valves had been newly
repaired.
intervals
Inspections of the valves were to occur at
ranging
from weekly
to
every three
weeks.
The respective utilities felt this inspection frequency
to
be
unnecessarily
conservative;
they
requested
a
change in inspection frequency, decreasing it to every
four weeks.
SKI did not permit this change.
At other
times, ring inspections in generators and requirements
for
auxiliary
power
supplies
have
elicited
utility
concerned
utility/regulator disagreements.
In
expressed
all
these
its
dissent
eventually
protests
occur
in
cases,
complied
politically to
numerous
with
by utilities
Sweden,
in
SKI's
over
as
it
be seen
anxieties,
can
wishes.
would
be
with
very
do
not
damaging
SKI.
(see p. 67),
but
Vehement
regulation
the industry's
already tenuous at best.
nuclear power
discussions,
safety
arguing
current moratorium in Sweden
post-Chernobyl
the
With
the
as well as
position
is
Thus, the precariousness of
create a rather
placid
regulatory
environment.
The
availability
enabling more
of
sensitive
new
analysis
monitoring
or
techniques
more
accurate
modeling have not forced the regulator's hand, causing
SKI to shut down plants until the new methods may be
implemented.
In
reality,
many
developed
voluntarily
by
This
is,
might
work
as
one
the
new
techniques
utilities
expect,
are
themselves.
enthusiastically
encouraged by SKI, but SKI has never forced new methods
on
a
utility.
With
the
foresight
and
initiative
generally demonstrated by the utilities, SKI may never
have had the opportunity to do so.
Indeed, the Swedish utilities frequently lead SKI
in recognizing and addressing technical problems.
The
individual plants do seem justified in maintaining (and
SKI
agrees)
expert,
that
and
that
each
primary
with the operator.
utilities
Just as
to
be
it is
plant
is
safety
its
own
technical
responsibility
lies
There are strong incentives for the
alert,
unwise to
competent,
and
cooperative.
appear an adversary of
SKI,
likewise, the utilities cannot afford to let SKI lead
them around, scolding and cajoling them, presenting the
regulatory hoops through which the utilities must jump.
One
of
the
KSU
staff
utility/regulator
observed
relationship
that
the
in
typical
Sweden
was
characterized by "more dialog than directives."
Representatives of
KSU were
asked if
there was
ever any complaint from SKI of untimely notification of
the regulatory authority regarding events
No
significant
cases
reporting
routine
operating
information
nuclear utility.
is
to
be
utilities'
incorporated
of
followed
in
were
recalled.
The
includes
daily
Sweden
transmitted
to
SKI
from
each
Any deviation from normal conditions
reported;
response
in
this
at a plant.
the
requirements
to
most
such
governing
the
deviations
are
standard operating
proced.ures
of
each
that
plant.
SKI
extreme
does
have
measures
the
be
authority
taken
to
require
(including
plant
shutdown), but as noted above, such extraordinary cases
have not occurred.
4.3
Outage Reporting and Classification
4.3.1
General Principles and Examples
Sweden
reports
Nationally,
the
figures
about
were
few
highest
12%
regulatory
regulatory
for
PWRs
1982, and 1983) and 4% for BWRs
years
had
Reactor
an
insignificant
technology
is
quite
capacity
(in the
of
similar
loss
years
(in 1976).1
amount
losses.
7
1975,
All other
these
in
losses.
the
United
States and Sweden; one would expect the same technical
problems
in
criteria
for
the
two
nations,
but
classifying
the
otherwise.
Thus,
regulatory
or
ascertain
what
does
and
perhaps
resulting
it
does
was
not
different
outages
as
necessary
to
constitute
a
regulatory loss in Sweden.
In
a
discussion
indicated
that
preheater
sections
imposed loss.
with
required
has
This
KSU,
redesign
been
its
of
for
most of
the
activity was
the
absence
tasks
of
steam
considered
many
large
required by the
a
generator
regulator-
responsible
high capacity losses in 1982 and 1983.
given
representatives
for
the
One explanation
losses
was
that
regulator could be
accomplished during the annual refueling outage.
There
is a cooperative effort between SKI
and the utilities
to
work
schedule
time.
nearly
Such was
the
hydrodynamic
load
example
a
of
Oskarshamn
all
regulatory
case with
I plant.
due
for
Another
outage
A recent
this
modifications
calculations
regulatory
during
BWRs.
occurred
refueling
at
to
the
period
was
extended for replacement of the core grid when the grid
support
bolts
cracked.
Yet
another
case
of
a
regulatory loss occurred at Ringhals 2 with a charging
pump problem.
Noting
all
representatives
these
offered
examples,
one
a
definition
regulatory loss in Sweden.
such
losses
arise
difference of
general
of
the
KSU
of
a
He proposed that nearly all
from
an
unresolved
technical
opinion between SKI and a utility.
If
SKI insisted on its viewpoint in these situations, the
utility was very likely to label any resulting outage
as
regulatory.
statement,
Most
although
of
some
those present
preferred
accepted this
a more
inclusive
reason
for
definition.
As
mentioned
above,
one
the
infrequency of SKI-motivated losses is that much of the
work that SKI might require is scheduled in conjunction
with other outages, particularly refueling.
maintain
a
"stop
list"
of
pending
Utilities
repairs
and
preventive maintenance, which assigns priority to these
tasks for
unless
the next forced or scheduled outage.
required
SKI's
significantly
extends
work
an
forces
outage,
shutdown
the
the
Thus,
outage
is
or
not
considered regulatory.
As an example of an outage
not
considered
plant.
year
consider
It has been operating at
when
cracks
were
found
the
Ringhals
80% power since
in
the
steam
2
last
generator
The steam generators had already been scheduled
tubes.
for
regulatory,
(a derating, actually)
in 1989,
replacement
interested
in
prolonging
however,
their
so
life
the utility
until
was
time.
that
The utility found (and SKI agreed) that by lowering the
plant's power just 20%, pressure and temperature would
drop
such that crack growth essentially
stopped.
SKI
probably would have prohibited operation if the cracks
continued to grow, but it was essentially the utility's
Thus, this derating was
decision to derate the plant.
not
treated
a
as
regulatory
It
imposition.
was
mentioned that, in the case of an SKI-ordered derating,
the lower power level is adopted as a new baseline for
calculating capacity losses.
the
in
deratings
data,
so
This practice masks these
these
must
treated
be
independently when seeking regulatory losses.
At
was
the
found
Ringhals
to
have
1 plant,
some
intergranular
61
four-inch
stress
piping
corrosion
cracking
(IGSCC).
regulatory
This
loss,
but no
might
have
resulted
shutdown was
in
required.
repairs were accomplished while operating.
a
The
In the wake
of this discovery, no other plants were forced down for
inspection or testing.
There are numerous cases in the
United States where problems at one plant have prompted
the
NRC
to
call
for
inspections
at
others,
often
necessitating shutdowns.
Next,
the
stringency
of
different
facets
of
Swedish safety regulation will be evaluated: technical
specifications, inspections, and modifications.
4.3.2
Technical Specifications/LCO Violations
Swedish technical
specifications are drafted by
the utility, drawing upon the data
the
plant's
approved by
resolve
vendor.
SKI.
The
A
significant
official approval.
and experience of
specifications
joint committee
differences
of
is
must
be
convened
opinion
prior
to
to
In this way, there is much informal
contact and negotiation between utility and regulator.
These
processes
make
the
official
specifications
by
questions
interpretation
of
SKI
very
approval
of
straightforward.
and
understanding
the
Some
may
remain, but these are normally inconsequential for the
plant's operation.
62
Swedish
technical
varieties.
operation,
There
and
are
there
specifications
come
specifications
for
is
also
a
in
two
standard
special
set
of
specifications that apply during refueling outages.
In
addition, there are rules and procedures separate from
either
set
of
technical
specifications
that
embodied in the Surveillance Test Book (STB).
are
The STB
contains requirements that are more detailed and plantspecific than
the technical specifications.
Criteria
for the performance of individual plant components are
enumerated in this document.
to the
equal
The STB is not ancillary
technical specifications;
weight
and
both documents carry
standing.
Like
the
technical
specifications, the STB can be revised and reworked by
a joint SKI/utility committee.
conservatively
are
specifications
Technical
in
practice.
applied
Utilities
rather
customarily
monitor critical parameters of the reactor's operation
to
ensure
LCOs.
an
that the
plant
remains
safely
within
the
If a trend in one of these parameters indicates
impending LCO
violation,
the
utility
itself
will
shut the plant down if no lesser corrective action is
effective.
is
already
shutdown.
Alternatively, SKI may see that the .plant
in
violation
The latter
is
of
the
much
interest)
to
remain
and
order
less frequent,
utilities view it as good practice
best
LCO
as
a
the
(and in their own
within
technical
specifications
and within the
STB.
of the
strictures
Such utility-motivated shutdowns are not classified as
regulatory losses, whereas the rare SKI imposition is
considered a regulatory loss.
KSU representatives were asked
specifications
technical
of
if violations
deterministic
are clear-cut,
events, or rather, if judgment and discretion are often
in
applied
that
some
establishing
negotiation
responded
not
or
there
a
of
violation
in
when
Usually, however,
specifications.
discussion, as
whether
over
occurs
down
shut
would
plant
They
violations.
is not much
the specifications include restrictions
on the time allowed until shutdown or repair that are
observed without question.
Technical
analogous
process
changes
be
changed
to
the
process
of
their
which
are
then must be
actually quite
approved by
the
only
participants
proposed revisions
public
a
initial
hearings
in
Such
Swedish
the
It is noteworthy
discussion
of
utility, i.e.,
no
the
are SKI and the
are held.
SKI.
especially in
common,
early years of a plant's operation.
that
via
The utility desiring the change proposes a
formation.
new text
may
specifications
citizens
must go
directly to the government with grievances; this avenue
of dissent has not been used frequently.
64
Inspections
4.3.3
SKI does not
plants.
employ resident inspectors at
the
Creating such a position might be considered
an anti-cooperative gesture by KSU;
indeed, some KSU
personnel saw such inspectors in the United States as a
manifestation of
the distrust between the NRC and the
At the Forsmark plants in Sweden, the
US utilities.
SKI inspector visits about every two weeks.
is
customarily
inspections
are
announced
a
in
rarity.
The visit
advance;
According
surprise
to
KSU,
a
"continuous dialog" exists between each utility and its
SKI inspector.
SKI itself has daily contact with these
inspectors.
Plant
components
constantly
evolving
are
inspected
hierarchy
of
according
emphasis.
to
a
Those
systems exhibiting the most problems will receive the
greatest
amount
of
scrutiny.
It
appears
that
the
amount of surveillance and testing done in Sweden and
the
United
States
is
commensurate,
performing occasionally more.
not the
with
Sweden
The major difference is
amount of testing and inspection, but
rather
how systems are selected for testing in each country.
4.3.4
Modifications
KSU
personnel
required due
to
could
recall
no
modifications
construction or design deficiencies.
65
SKI
has,
however,
issued
more
restrictive
which required plant modifications.
example
of
this
action
was
in
criteria
The most notable
response
to
the
TMI
accident in the United States.
SKI devised an "action program" (with some input
from
the utilities)
in
certain
comprising
Ringhals
2 was
actually
conducting
accident.
SKI's
the wake
required
the
only
TMI
of the
In
backfits.
PWR
in
refueling
at
operation;
the
time
for
three
In
weeks.
1979,
it
was
of
the
extended the
required modifications
refueling outage
incident
addition,
SKI
added mandatory safety reporting requirements for all
report
analysis
as-operated safety
Periodically, an
nuclear plants.
would
be
as
required,
well
as
a
comprehensive ten-year safety report after each decade
of operation.
A second
although
government,
commission was
study,
which
utilities,
major
SKI
universities,
collaborators.
conduct
completed
This
and
study
in
from
involved.
the
A
a reactor
safety
1980.
The
early
SKI
was
came
TMI
also
was
assembled to
was
to
reaction
were
one
among
component
the
of
widespread discussions on the safety of nuclear power
after TMI.
Because of the attention now focused on the
industry, the construction of the Forsmark plants 1, 2,
and 3, Oskarshamn 3, and Ringhals 3 and 4 was delayed
at least one year.
In March 1980,
shortly after the
reactor safety
study was released, a public referendum was held on the
future
of
nuclear
power
in
Sweden.
People
voted
in
favor of phasing out nuclear power completely by 2010.
KSU representatives felt that the TMI
primary
motivator
of
the
accident was the
referendum
and
the
major
determinant of its outcome.
4.4
Comparisons with US .Experience
The discrepancy in the amount of regulatory loss
reported by Sweden and the United States is due almost
entirely
to
the outage
each nation.
Some differences
regulation;
inspectors
similar
classification
for
exist in
and nothing analogous
technical
requirements
are
differences
appear,
in the
Swedish
by
of
no
resident
to an IEB.
However,
specifications
in place
the means
has
Sweden
instance,
scheme used
and
inspection
two nations.
regulations
are
Where
usually
stricter, as with their conservative application of the
LCOs.
More information on technical specifications is
given below.
were
most
difference
for
Furthermore, the same technical problems
troublesome
is
that when
technical
inspections),
for
both
outages are
specification
Swedish
countries.
utilities
67
incurred
violations
hold
the
The
key
(whether
or
for
hardware
accountable,
problem.
the
labeling
the
difficulty
as
a technical
In the United States, it is the regulation or
regulator
outage.
that
Hence,
is
far
considered
more
the
cause
regulatory
of
the
outages
are
reported in US performance statistics than in those of
Sweden.
Comparing technical specifications in Sweden with
those
in
Efforts
that
the United
are
made
are
in
States,
some differences
Sweden
to
site-specific
and
create
plant
specifications
plant-specific.
specifications tend to be more generic.
Swedish
appear.
specifications
US
For the early
(drafted in
1973-1974),
the baseline data was taken from US vendor information
(i.e.,
from
Westinghouse).
Specifications
used
by
ASEA-ATOM, the Swedish vendor of most later plants, are
somewhat
more
States.
In
stringent
than
particular, BWR
those
in
the
specifications
United
are
more
demanding concerning water chemistry requirements.
addition,
Swedish
technical
specifications
In
stress
functional requirements of plant systems, e.g., leakage
rates,
rather
than
prescriptive
standards,
e.g.,
specifications on pipe or vessel integrity.
Representatives
regulations
Sweden.
were
This
far
of
more
observation
KSU
were
voluminous
must
be
certain
that
US
than
those
of
interpreted
with
care, because any difference in regulatory scope may be
due
to significant differences between
utilities.
technically
On
balance,
competent,
Swedish
take
Swedish and US
utilities
greater
are
more
initiative
with
safety measures, and set higher internal standards than
do US utilities.
Hence, a less
prescriptive and less
intrusive set of regulations may be justified for more
vigilant utilities, as those in Sweden appear to be.
Unique among the European nations studied, Sweden
did
with
report
the
sections.
some
regulatory
redesign
This
was
of
loss.
steam
classified
It
was
associated
generator
preheater
as
regulatory
because
there was an unresolved technical difference of opinion
between SKI and the utilities over the shutdowns.
69
5.0
Switzerland
5.1
Organizations Influencing Performance and Safety
5.1.1
Safety Regulators
The
the
nuclear safety regulator
Swiss
in Switzerland
Federal Nuclear Safety Inspectorate
within the
Swiss Federal Office of
Energy.
is
(HSK),
Broadly,
its responsibilities are nuclear safety and radiation
protection.
functions
Similar to the NRC,
performed
by
HSK
the major technical
are
conducting
safety
analyses for proposed power plants, writing technical
reviews,
and performing
surveillance
and
backfitting
analyses for existing plants.
Unlike the NRC, however,
HSK
This
has
no legal
section.
is
so because
HSK
operates in a less politically charged environment.
A
separate section of the Office of Energy assumes legal
responsibilities,
permits.
such
as
the
,granting
of
plant
This section is also the legal advisor for
HSK.
The Office of
grants
general,
three
Energy
different
construction,
in the
permits
and
Federal government
to
nuclear
plants:
operating.
HSK's
responsibility includes any decision for a plant within
the
purview
example,
of
when
developed in
one
the
of
Swiss
the
three
seismic
1979, all plants had
70
permits.
risk
charts
As
an
were
to be requalified.
In particular, the anchoring of
analyzed
and
required
no
modified
change
if
in
electrical panels was
necessary.
the
operating
This
action
permit,
so
HSK
managed the recertification.
Currently,
HSK
employs
professional and supporting.
a total
of
54
people,
The 1983 budget
(latest
figures available) was approximately SF 5 million.
additional
SF
10
million was
research contracts.
allocated
for
An
external
See Figure 5.1 on page 72
for a
diagram of the structure of HSK.18
The
Swiss
Supervision
quality.
for
Pressure
Vessel
(SVDB) has authority over pressure vessel
The Corporation has conventional and nuclear
divisions;
resident
Corporation
the
nuclear
inspector
is
section works
present
at
under HSK.
the plants,
only a
part-time liaison who performs periodic checks.
has
been
appointing
some
a
support
full-time
from
resident
plant.
71
the
government
inspector
No
at
There
for
each
Personnel and
Organization
of Nuclear
Installations
Section
I
Secretarial
Service
Director
Administrative
Service
Assistant
Staff
I
Reactor Safety Division
|
- -Radiation
Protection Division
Accident Consequences
Reactor Technology
Section
and Emergency
Planning Section
Systems and Electrical
Engineering Section
Plant
Coordinators
Radiological
Surveillance Section
Mechanical and Civil
Engineering Section
FIGURE 5.1:
Organizational Structure of HSK
Source:
~"'"~' ~'~'~''
"-c~
Radioactive Waste
Management Section
Reference 18.
5.1.2
Utilities
The Swiss nuclear utilities
staffs.
but
employ rather small
The technical departments are small as well,
highly
dedicated
and
competent.
Each
plant
maintains a skilled engineering staff on-site; there is
no
separate
technical
headquarters.
There
is
much
emphasis on hiring quality workers, whether engineers
or laborers,
phase.
rate
and on doing
so
early
in
the
start-up
This practice, in combination with a turnover
of
only
dedicated
a
and
few
percent
professional
each
staff
year,
creates
thoroughly
a
familiar
with the plant.
The utilities emphasize quality and redundancy in
the entire plant, in both the nuclear and balance of
plant
fractions.
HSK
supports
this
policy.
The
utilities feel that non-nuclear induced stresses on a
plant (scrams caused by turbine trips, for example) are
just as
severe as many nuclear events, from a safety
standpoint.
events;
Hence,
quality
pains
assurance
permeate the entire plant.
are
taken
measures
to
and
avoid
such
redundancy
This is not tantamount to
introducing greater complexity into the plants.
On the
contrary, utilities strive to keep
simple
the plants
and transparent.
While quality is a priority, the utilities manage
the plants to maximize their availability according to:
73
Availability = F(Reliability, Maintenance)
This equation is applied to safety equipment as well as
The relation implies
to the power production system.
that
quality
high
scheduled
outage
of
amount
This is
time
forced
done up
are
outages
to
measures
(reliability)
adopted
to
requiring
decrease
the
(maintenance) experienced.
the point that
the forced outage
time saved equals the scheduled outage time invested.
The mathematical relation is always tempered by safety
considerations;
reliability
is
felt
be intimately
to
connected with safety.
In
planning
outages,
scheduled
the
utilities
budget the time to prevent hurried maintenance. Efforts
are made not to overwork the employees.
This keeps the
staff alert, and minimizes carelessness.
reserved
for
unanticipated
work
during
Also, time is
the
outage,
which always seems to arise.
5.2 ..Organizational Relationshipsin Practice
The
professional
staffs
of
both
HSK
and
the
utilities employ engineers and scientists exclusively.
Thus, their
discussions are
than legally oriented.
tough."
always technically rather
The dialog is "fair, open, and
Even
in
the
relatively
small
Swiss
industry, considerable diversity exists.
nuclear
Out of
the
five nuclear utilities which own at least a share of a
nuclear plant, a range of technical opinions is usually
found.
By
national
association
reactor
themselves,
safety
the
utilities
to discuss
and
have
technical
to. promote
formed
aspects
excellence.
a
of
Their
differences of opinion are often lessened through these
meetings.
As
the
regulator,
HSK
adopts
a
more
conservative position than most utilities, but the gap
is
rarely
variety
large.
of
possibly
Swiss
the
Diversity
nuclear
is
plant
also
found
designs.
two Beznau plants,
no
two
in
the
Excepting
of
the
five
plants operating in Switzerland are markedly similar.
Despite
such
diversity
in
both
utilities
and
plant hardware, HSK refrains from drafting prescriptive
requirements.
oriented,
instead
Rather, HSK's standards are functionally
requiring
of
a
certain
specifying the
and systems.
use
level
of
of
certain
performance
technologies
Furthermore, HSK strives to be consistent
in interacting with all the nation's utilities.
HSK
does
not
divide
geographically or
the
industry
otherwise, most
of
Since
into
regions,
HSK's
officials
are well-acquainted with key personnel at each plant.
Of course, the small size of the Swiss nuclear industry
significantly
reduces
the
partitioning or layering.
need
for
administrative
Such policies of HSK have created a peaceful and
cooperative
environment
Switzerland.
resolution
for
the
nuclear
industry
Nevertheless,
opportunities
disagreements
and
of
the
for
the
amendment
of
requirements have been provided by policy and law.
example,
if
a
utility
believes
that
a
in
For
certain
inspection requirement is no longer reasonable, it may
propose and defend a new standard in meetings with HSK.
The
final
decision on
HSK's alone.
such
a proposal,
however,
is
Even if the proposal is rejected by HSK,
there
remains
exact
timing
Swiss
utilities
some
of
latitude
any
have
outages
for
negotiation
that
availed
may be
on
the
required.
themselves
of
this
proposal and review process numerous times, often with
success.
This
demonstrates
that
HSK
and
utility
perceptions of safe and prudent industry practice are
largely in agreement.
Before passing judgment on a controversial issue,
HSK is required to hear the arguments of the utility.
If
a utility
is
not
satisfied with HSK's
subsequent
findings, it may appeal to the Federal government for
another decision.
None, however, have ever resorted to
this avenue of redress.
The harmonious professional relationship between
the regulator and each utility is created and supported
by their close and regular communication.
76
Each plant
is
required
HSK.
unusual
Moreover,
so
immediately,
events
that HSK
is
report
operating
a monthly
submit
to
are
to
reported
be
apprised of
to
the
plants'
Potentially dangerous conditions at plants
condition.
close to international borders also entail notification
of
other
nations.
Also
meetings
between HSK and
discuss
both
general
of
the
importance
staff
concerns
are
of
and
periodic
each plant
to
plant-specific
matters.
Swiss
The
problems
of
response
to
plant
escalating
widely
the
experienced
and
costs
extended
construction periods serves as an excellent example of
In Switzerland as around
industry-wide collaboration.
the world, there was great interest in containing plant
capital costs and shortening construction times without
compromising safety or availability.
task
force
suppliers,
authorities
worked
this end,
of
representatives
comprising
engineering
To
firms,
on
these
and
under
sponsorship of the Federal Office of Energy.
force met
frequently
from
1983
to
1986.
plant
regulatory
the
issues
a
the
The task
The
final
report* issued in January 1986 presented many promising
recommendations
for
reducing
costs
and
construction
times, and for backfitting existing plants.
*The report was entitled Proj ek tabwicklung . und
(PQS),
or
bei _Kernkraftwerken
Qualit.tssicherung
for
Nuclear
Assurance
Quality
Pro jpect Management and
Power Plants.
Inspection practices
same
collaborative,
are
inspectors
in Switzerland reflect
the
spirit.
No
HSK
There
are
cooperative
at
resident
plants.
the
designated liaison personnel for each plant within HSK,
who
periodic
make
procedures.
of
checks
conditions
plant
and
Unannounced inpections are permitted, but
are unusual, as both parties prefer that the technical
agenda for each inspection be established in advance.
As
state of
utilities
for
the
technical
the plant, many inspections are conducted on
responsible
for
its
plant' is
Each
initiative.
own
their
responsible
are
own
safety;
HSK
primarily
oversees
the
utilities and ensures that these responsibilities are
carried out.
If HSK were dissatisfied with the quality
or extent of a plant's particular testing or inspection
procedure, the regulator would require that the routine
be repeated.
In sensitive
cases, such tests might be
supervised or assisted by an HSK (or an HSK-appointed)
expert.
of
In general, however, HSK does not keep abreast
the many detailed questions requiring attention in
each plant.
Thus, it is the utilities that must be the
first to recognize and address technical problems.
Some
specific
examples
batteries
for
regulator/utility
At the Leibstadt plant for
interactions may be given.
instance,
of
powering
emergency
equipment
were found to be defective; replacements were ordered.
78
Because
these
earthquakes
batteries
of
were
required
a certain intensity,
would not be easy to acquire.
to
withstand
the replacements
In fact, it would be one
month before the equipment qualified for the task would
Because the plant could not legally operate
arrive.
without such back-up power, the plant staff asked HSK
for permission to operate using ordinary batteries (not
earthquake-proven) for the 30 days until the new ones
HSK
arrived.
this
allowed
modification of
standard
procedures.
At
the
Leibstadt
emergency
heat
installed,
separate
emergency
core
and G6sgen
a special
(SEHR)
had
system
removal
from
cooling
plants,
the
low
standard
pressure
For
(ECCS).
system
been
these
plants, permission was requested to perform maintenance
on
the
The
ECCS.
disabling
of
the
work would
ECCS
during
the
require
temporary
operation
the
the
of
plant.
Ordinarily, such an action would be imprudent.
Because
of
granted.
the
Some
specifications
the
redundancy, however,
minor
were
amendments
was
the
technical
These
examples
to
required.
request
illustrate HSK's general flexibility in rulemaking and
receptivity to utility requests.
Conversely, utilities appear to have an open ear
for
regulatory
system
at
proposals.
Leibstadt,
this
In
the case
system
was
of
the
actually
SEHR
first
Most agreed with HSK
proposed by HSK to the utilities.
that such a capability was a sound idea and should be
As demonstrated above, such
installed in their plants.
by
accompanied
often
are
requirements
extra
compensating regulatory flexibility.
A
a
of
case
construction
considerable
a
a
utility's
the
In
delay.
in
resulted
ultimately
impasse
This
plant.
and
the Leibstadt
problems with
design
utility concerned
HSK
between
disagreement
opinion, certain regulatory rules were, at first, not
sufficiently explicit regarding standards of design and
The
review.
procedures
control
also
utility
by
was
underway,
to
prior
HSK
observed
document
the
questioned
construction.
construction
Once
requirements
standards
more explicit
became
led
established in
to
some
a 1973
and
HSK's
demanding;
The
redesign.
safety
design
basis
contract between Leibstadt
the supplier was no longer adequate.
new
and
The utility also
objected that some equipment orders for the plant had
been significantly delayed by HSK design reviews.
In
the utility's perspective, HSK was motivated to become
more
demanding
by
political
external
including some from neighboring nations.
and
discovered
plant's
design
more
detailed
information
and construction.
pressures,
HSK learned
about
The regulator
the
took
with
issue
many
characteristics,
nonetheless
the
revealed
newly
thus causing the delay.
attributed
than
rather
of
viewing
the
it
as
delay
to
design
The utility
political
result
a direct
forces
of
HSK's
actions.
HSK's views on the
are somewhat different.
Leibstadt construction delay
According to
the regulators,
the plant incorporated a new containment design which
In addition, the design of
had not been used before.
the nuclear island was new, both to the supplier and to
the
regulator.
The
utility* began plant
construction
without awareness of the potential problems that these
new
designs
could
present.
officials
Regulatory
objected that the utility did not allow them sufficient
time
to
check
the
with construction.
proposed
design
before
proceeding
As a result, HSK's standards were
still evolving while the plant was being built.
In addition to its
relationship with utilities,
HSK is affected by governmental actions and, to a more
limited extent, by the public.
their
actions
scrutinized by
HSK officials can have
a political
proceeding.
Hearings to gather information are fairly common.
For
these meetings, HSK staff may be required to testify or
to
supply
written
statements.
Dismissals
or
other
extreme or capricious actions, however, are very rare.
(MPs) can request that HSK
Members of Parliament
provide answers to specific questions on plant safety.
Some HSK staff members usually travel to Bern to assist
the Energy Minister in answering the questions of the
The government also controls HSK's budget.
Parliament.
This power has not been used as a tool of anti-nuclear
sentiment, as has happened occasionally in the United
In general, MPs say privately to-HSK that they
States.
favor
more
money
for
the
regulatory
As
body.
the
regulatory budget has not been prominent on the public
budget
agenda,
An
discussed.
have
increases
in
increase
not
personnel
been
formally
was
recently
requested by HSK, which it received.
The Swiss public has one main avenue by which it
may require HSK action.
for
especially
radius
of
those
a nuclear
This
people
opportunity is intended
within
living
a certain
(including, incidentally,
plant
those German citizens close to a Swiss plant.)
people
may
affecting
accompanied
contained
by
safety
additional
and
in
HSK
of
answers
health
their
information
Hence,
require
safety.
the
reports,
permit
is
These
to
questions
The
technical
application,
usually
adequate.
safety analysis motivated by public
hearings has not caused delays in plant construction or
start-up.
5.3
Outage_ Reporting and Classification
The Swiss industry reports no regulatory outages.
Similar to the practice of France and Sweden, problems
that
arise,
whether
LCO
violations
or
needed
modifications, are classified as technical difficulties
and never as regulatory outages.
An illustration of outage reporting may be cited.
At Leibstadt, there was a scram due to operator error.
This
caused
a
human error.
short
outage,
which was
attributed
to
Upon start-up, a safety relief valve and
discharge line were found leaking.
The
leak
was
not
an
urgent
Leibstadt chose to defer the repair.
acceptable
to
HSK.
An
safety
issue;
This decision was
inspection,
however,
was
requested by HSK one week prior to the repair in order
to
identify
the
extent
with
this
the
of
leak's precise
the work
wish,
but
location
required.
would
have
Leibstadt
performed
without a preliminary inspection.
nor
the
additional
inspection
considered a regulatory loss.
as
technical
issues
for
Neither
time,
gauge
complied
the
repair
the repair
however,
was
Rather, they were viewed
which
the
responsible, and classified accordingly.
83
and to
utility
was
At the Leibstadt plant in 1986,
was lost due to testing.
the
MSIV
and
the
only 0.17
EFPH
The tests conducted were of
turbine
inlet
valve.
At
first,
testing the turbine inlet valve required reducing the
plant's power to 90%.
loss,
the
Even with this relatively small
utility was
able
still further in later tests.
learning
the
plant's
to reduce
the
lost
This was accomplished by
sensitivity
to
scrams,
test
by
In subsequent turbine valve tests, the utility
test.
only needed ta lower power to about 98%.
1986,
EFPH
Leibstadt
had
only
1.66%
For the year
capacity lost
due to
forced outages.
TechnicalSpecifications/LCO Violations
5.3.2
Swiss technical specifications, especially those
developed for newer plants,
in
the
United
are very similar to those
Swiss
States.
LCOs
basically
were
derived from US values and remain quite similar to them
Older
today.
specifications,
while
also
using
US
vendor information as a baseline, are tailored to the
designs
individual
of
the
earlier
Swiss
plants.
Important differences in the regulation of old and new
plants
plants.
arise
This
from
greater
feature has
redundancy
in
lengthened
the
the
newer
inspection
interval and the allowable inoperability time for many
components.
Swiss
similar
to
stringency
is
difference
that
the
the
US
Swiss
also
are
requirements
surveillance
The
rules.
chief
do
requirements
of
not
incorporate the variable surveillance interval found in
the US documents.
Both plant operators and regulators
see some countervailing considerations that limit both
and stringency of surveillance requirements.
the scope
First, there is the economic incentive to keep capacity
due
losses
reasonably
inspections
to
Second,
low.
inspection time is minimized to prevent plant employees
the
conducting
radiation
dose than
is
more
receiving
from
checks
absolutely
of
a
Last,
necessary.
components are assigned strict inspection priorities to
prevent
the
diversion
technical
of
attention
expertise from the most important systems.
and
Since the
employees' time is a scarce resource, it must first be
concentrated
on
the
weakest
and
most
vulnerable
systems.
HSK has very rarely asked for mid-cycle shutdowns
for inspections.
The few that have been ordered were
to verify the adequacy of earlier repair work performed
during the annual refueling outage.
an outage
occurred at the Mfihleberg
1986.
An
inspection
of
the
recirculation piping was required.
be replaced
in the refueling outage
85
An example of such
plant in
stainless
January
steel
This piping was to
later
that year;
thus, the utility was likely not inclined to check the
piping on its own initiative, in the
of trouble.
absence of
signs
HSK allowed the work to be scheduled at
the utility's convenience.
Officially, this inspection
was motivated by a temporary change
specifications;
here,
implementation of
this
was
in the
technical
tantamount
to
the
additional safety precautions.
The
inspection
revealed
intact;
further corrective measures were necessary
no
that
the
earlier
repairs
were
prior to pipe replacement.
5.3.4
Modifications
When a new problem is discovered at a plant, the
principle of
governs the
HSK has
utility responsibility
for
responses of both HSK and the
the legal
right to
shut the
to problems,
the utility
strict
time
limits,
demands
it.
is not
unless
Rather,
the
management
a measured, careful
stand behind in confidence.
solution
strategies,
utilities.
plant down,
such extreme action has never occurred.
formulate
plant safety
but
In responding
normally
faced with
situation's
takes
response
the
urgency
time
that they
to
can
The utility's package of
including
its
proposed
modifications, is presented to HSK for discussion and
approval.
Utilities that take the initiative in presenting
appropriate
solutions
are
in
86
a
better
position
to
control the modifications required of their plants.
By
taking the lead in problem resolution, it is easier to
the
resist
of
introduction
experimental
unproven
procedures and hardware or unnecessary complexity by a
This is not to
well-meaning but meddlesome regulator.
On the contrary, much of
say that HSK is meddlesome.
the regulatory burden has been lifted from HSK by the
competence and
technical
utilities'
vigorous
efforts
toward trouble-free plants.
actions
HSK's
and
policies
consistent
are
and
If a problem
harmonious with the utilities' behavior.
is brought to its attention (by an event at a nuclear
plant elsewhere in the world, for example),
ask each
HSK would
utility to render an opinion explaining how
development
the
foreign
HSK
and the utilities
relevant
is
informed of
keep
nuclear events and research.
carefully
followed
the
its
to
plants.
international
For example, both groups
international
problems
with
stress corrosion cracking and reactor trip systems.
The primary response to
the TMI
accident by HSK
was the establishment of an independent internal study
group to examine critically various TMI-related issues.
In
particular,
problem of
the
study
group
hydrogen formation
filtered containment
and
venting for
was
to
address
the
the possibility of
pressure
relief.
A
group
expert
of
outside
specialists
also
HSK was
of
convened for consultation on these topics.
Remarkably,
outages
no
were
experienced
in
There was, however, a
Switzerland as a result of TMI.
delay in the initial start-up of one plant after the
No modifications or other delays related to
accident.
start of
commercial
TMI caused
any outages after the
operation.
HSK required some modifications, similar to
the
new US
requirements,
operating plants
for
which
Again, these changes were
arose from the TMI accident.
effected during the annual refueling outage, causing no
further shutdowns or deratings.
5.4
Comparisons_ with US Experience
comparing
When
US
and
Swiss
the
nuclear
industries, one must keep in mind that the Swiss system
is an order of magnitude smaller in terms of the number
This difference in scale accounts for many
of plants.
For example, it
disparities between the two nations.
is
much
treatment
easier
of
the HSK
for
Swiss utilities
likewise in the United States.
regions
of
the
NRC
may
consistent
to be
than
for
the NRC
in
to
its
do
The five administrative
encourage
the
same
close
relationship between regulator and utility as exists in
a smaller system.
could
lead
to
This division of
interregional
the NRC, however,
inconsistencies
which
confound the uniform implementation of regulation on a
national scale.
it
industry,
certain
Also
is
critical
because of
infeasible
test
for
procedures
the size
the
NRC
the US
of
to
oversee
at plants.
HSK
is
able to send support personnel or an expert consultant
if requested for especially sensitive work.
Despite these differences in size, the Swiss and
US
industries
nuclear
share
the
characteristic
of
HSK is able to
having a wide variety of plant designs.
respond effectively to this diversity with functionally
oriented
regulations;
NRC
requirements,
in
contrast,
are generally far more prescriptive.
Swiss LCOs were modeled after those in the United
States; they remain substantially similar today.
Like
considers
the
and
France
Sweden,
Switzerland
Thus,
observation of
the LCOs merely good practice.
violations are
technical problems in Switzerland, not
regulatory outages as in the United States.
Inspection requirements in Switzerland include no
provision
for
a
variable
resident inspectors, or IEBs.
surveillance
interval,
Furthermore, where the
redundancy of safety systems exceeds that in the United
States, longer times of inoperability are permitted in
Switzerland.
Perhaps as a result, very few mid-cycle
inspections have ever been required.
89
Modifications in Switzerland (including those for
were
TMI)
outages.
of
a
accomplished
without
incurring
any
forced
This contrasts sharply with the US experience
substantial
number
of
forced
outages
modifications, especially in the wake of TMI.
90
for
6.0
Conclusion
6.1
Data Reclassification and Conclusions
In
the
differences
practices
final
last
in
were
revealed
the
European outage
data
will
be
outage
significant
classification
and discussed.
central
Now,
question
be put to the test.
investigation will
US
chapters,
European and US
chapter,
about
three
in the
of
this
Knowing more
classification conventions,
reclassified
according
to
the
European
This adjustment will yield a more accurate
standards.
picture of the burden of US regulation relative to that
in the other nations studied.
As
expected,
because
of
highly
similar
technologies, the plant systems most troublesome in the
United States were also those causing the most outages
in
Europe.
More
importantly,
the
same
"regulatory"
constructs and practices observed in the United States
(e.g.,
technical
specifications,
surveillance
requirements, and required modifications) were in place
Furthermore, the stringency of
in Europe.
European requirements
however,
between
the
and
and
The difference,
is comparable.
US
the US
European
systems
is
fourfold:
1)
In
the
European
operating standards
nations
studied,
the
voluntary
of
the
utilities
and practices
at
apparently
are
least
Thus, a shutdown may
regulations in those nations.
be
a plant
when
required
established
limits
limits
those
by
imposed
operating
these
exceeds
before
practice,
industrial
by
safety
as
stringent
as
regulation
external
are
reached.
The European nations consider capacity lost due to
2)
operating limits as a forced outage, rather than as a
regulatory outage, as in the United States.
US,
In the
3)
losses
imposes
safety regulation
on
utilities, through forced outages in particular, in a
manner in which European regulation does not.
4)
Frequent discussions and interactions occur between
and
utilities
between
place
takes
in
regulators
characterized by
professionalism
regulator/utility
relationship
or
litigious
antagonistic
and
people
technical
and
has
atmosphere
dialog
This
Europe.
The
respect.
none
seen
is
of
the
in
the
a
key
United States.
The
third
point
above
highlights
distinction that must be made here between regulatory
behavior, on the one hand, and outage classification on
the other.
of
all
Regarding scheduled outages, the regulators
four nations
resulting
in
in the
study have
scheduled outages.
Only
issued
orders
the
United
in
(and Sweden, to a much lesser extent),
States
outages
these
have
classified
been
actually
however,
as
Thus, largely similar regulatory behavior
regulatory.*
has met with distinct
on both sides of the Atlantic**
classification methodologies for resulting outages.
In
reclassifying the US regulatory outage data, scheduled
outages
will
totals
to
be
make
from
subtracted
the
the regulatory
conform
data
closely
more
loss
to
European conventions.
For
forced
outages,
the
picture
is
different.
This investigation found no evidence that the European
utilities studied had ever been faced with a regulatory
order requiring a forced outage.
the
reason
why
no
reported in France
forced
This may indeed be
regulatory
outages
are
or Switzerland, and only a few in
*Sweden's regulatory losses, as noted in the
technical
of
a
result
as
come
chapter,
fourth
differences of opinion between the regulator and a
are not
losses
scheduled
The Swedish
utility.
and
readily
as
quite
regulatory
as
classified
In
arbitrarily as are some US scheduled losses.
Sweden, a bona fide disagreement must exist; in the
(an IEB, for
any regulatory order
United States,
example) regardless of utility opinion, may cause a
regulatory outage.
not
is
nations
among
**Regulatory behavior
completely comparable for scheduled outages, although
less international variation is observed 'than for
The NRC often stipulates a shorter
forced outages.
time horizon for outage planning than do Europe's
Refer to the discussion following Tables
regulators.
6.1 and 6.2 for further information.
93
Again, in contrast, the United States reports
Sweden.*
a
the
Here,
outage
in
differences
transatlantic
outages.
forced
regulatory
of
amount
substantial
classification may be indicative of true disparities in
regulatory behavior.
how
the
The data do not show conclusively
would
European utilities
respond
In particular,
negotiable demand for a forced outage.
it
how
unclear
is
resulting
any
a non-
to
outage
would
be
will
be
classified.
assumed
the
that
uncertainties,
these
of
Because
US
outages
forced
it
from
result
regulatory actions that are not observed in the three
European
nations.
recasting
in
Thus,
the
US
data,
forced outages alone will be treated as unique and bona
fide
of
sources
regulatory
loss.
In
reclassification will observe this rule.
the
regulatory
particular, the
loss unique
to
the
general,
the
Namely, only
United
States
(in
forced outage component of regulatory
loss) will be included in the reclassification.
A different rule will be used for LCO violations
and outages due to the unavailability of safety-related
equipment.
For
these,
neither
scheduled nor
forced
*Concerted efforts are made by the regulator and
utilities in Sweden to schedule needed repairs at a
This policy is
time that is best for the utility.
apparently successful in achieving far more scheduled
Many of the extensive
than forced regulatory outages.
discussions between SKI and the utilities would be preempted by forced outages.
94
outages will be included in the reclassification.
is
because
the
three
European
nations,
This
without
exception, treat safety readiness and the observance of
LCOs as an integral part of industry practice, rather
than as burdensome impositions by the regulator.
The reclassified US data appear in Table 6.1 (for
PWRs, excluding TMI) and Table 6.2
96 and 97,
respectively.
These
(for BWRs) on pages
include the
original
data presented in Tables 2.1 and 2.2, respectively, for
comparison.
reflected
The downward revisions discussed above are
in
the
reclassified
data.
Note
that
the
scheduled outage component is also subtracted from the
lesser
outage
categories
--
licensing,
safety restrictions, BWR fuel limits.
fuel
and
core
These causes of
regulatory loss were not addressed in this study due to
their small
effect.
It is assumed, however, that as
with the major outage causes, only the forced outages
are unique to the United States.
Therefore, only this
component is included.
From Tables 6.1 and 6.2, it can be concluded that
international differences in safety regulation are not
the
primary
between
source
the
the United States
studied.
The
loss
comparable
are
studied,
of
from
and the European
recalculated
Table
to
1.1.
performance
totals
losses
The
for US
in
other
data
difference
countries
regulatory
countries
represent
an
TABLE 6.1:
Reclassified Average Annual US Regulatory
Capacity Loss,
1975-1984, for PWRs, excluding TMI
(in percent)
Classification by regulatory outage cause
Outage Cause
Capacit y Loss (%)
(as originally
(as
stated)
reclassified)
NRC-originated
inspections
3.47
1.15
NRC-originated
modifications
2.33
0.63
LCO violations
1.86
0.0
NRC licensing proceedings
& hearings
0.61
0.24
Fuel and core
safety restrictions
0.04
0.02
Combination
0.04
0.02
Unavailability of
safety-related equipment
0.03
0.0
TOTAL
8.38%
2.06%
(Combination category comprises inspections or LCO
violations or modifications in combination with a nonregulatory cause.)
Source:
OPEC-2 Database
TABLE 6.2:
Reclassified Average Annual US Regulatory
Capacity Loss,
1975-1984, for BWRs
(in percent)
Classification by regulatory outage cause
Outage Cause
(as originally
stated)
Loss () .Capacity
(as
reclassified)
NRC-originated
modifications
4.45
0.41
Combination
2.86
0.25
NRC-originated
inspections
1.61
0.53
LCO violations
0.86
0.0
Fuel and core
safety restrictions
0.32
0.31
NRC licensing proceedings
& hearings
0.16
0.03
BWR fuel limits,
i.e., MCPR, MAPLHGR
0.08
0.05
Unavailability of
safety-related equipment
0.02
0.0
TOTAL
10.36%
1.58%
(Combination category comprises inspections or LCO
violations or modifications in combination with a nonregulatory cause.)
Source:
OPEC-2 Database
estimate
of
interpreted
US
in
regulatory
light
of
losses,
and
two
caveats
the
should
be
discussed
below.
By
excluding
reclassified
data,
all
scheduled
the
amount
regulatory loss is understated.
losses
of
bona
in
the
fide
US
Some of the scheduled
losses arise from impositions by the NRC that would not
occur in the other nations studied.
are not
still,
as urgent as
US
utilities
those
are
These impositions
eliciting forced outages;
not
given
the
freedom
schedule their work optimally in these cases.
data available
in the OPEC-2 database
discriminate between short-term
days
ahead of
time)
and
The US
do not usually
scheduling
long-term
to
(e.g.,
scheduling
ten
(e.g.,
during the annual refueling outage, perhaps ten months
away).
Without such a distinction in the US data, it
is not
possible to more closely
observe
the
European
outage classification conventions.
A second,
the
and
reclassified US
countervailing
data
should
qualification
also
be
remembered.
The retention of all forced outages in Tables 6.1
6.2
results
losses.
in
an
overstatement
of
for
US
and
regulatory
It is improbable that the European utilities
would respond with a regulatory classification for all
forced
outages.
Their propensity
to
take
technical
responsibility for the plant would likely lead them to
believe that some of the forced outages are recommended
by good engineering practice.
No
source
and
of
the
further
the
conclusions
may be
drawn about
performance discrepancy between
United
States.
Some
the
Europe
generalizations
about
European utilities' attitudes and behavior may be made,
however.
European utilities
are
clearly responsible
for:
o
Maintaining
a
competent
in-house
technical
capability
o
Proper systematic monitoring and management of plant
operation to anticipate and prevent problems
o
Taking the initiative in proposing plans of action
to the regulators as problems do arise
o
Encouraging
a
detailed
understanding
of
the
operation and characteristics of individual plants
o
Advancing the development and adoption of
improved
safety systems at their plants.
The
cornerstone of
the European
approach
to
nuclear
safety is that plant safety is first and foremost the
responsibility of
each utility.
This tenet
explains
much of the Europeans' regulatory policy and industry
practice.
99
Accordingly,
European
utilities
leadership and demonstrate a high
exercise
level of
technical
competence in anticipating, recognizing, and preventing
technical problems.
standards,
Despite
difficulties
the
with
utilities'
the
plant
do
exacting
surface
occasionally, in Europe as in the United States.
such problems
When
arise, European utilities take vigorous
initiatives in formulating solutions, cooperating with
the safety regulators in a collaborative, constructive,
and technical dialog.
As
with
practices
of
US
US
investigation,
plant
performance
utilities
however,
it
vary
is
levels,
the
From
this
widely.
apparent
that many
US
utilities do not embrace the same practices, policies,
and attitudes
as do European utilities.
At the same
time, one must recognize the structural disincentives
existing
in
the
United States
positive behavior.
forces
which discourage
more
Economic, regulatory, and political
external
to
the
utility
all
contribute
negatively to the environment of the nuclear industry.
Nonetheless,
utilities
it
could
industry, and
some
is
elements
clear
benefit
the
of
from
study
themselves,
public by
the
this
judiciously
style
counterparts.
100
of
the
that
US
nuclear
assimilating
their
European
Further Work
6.2
This
investigation
demonstrated
that
the
plant
performance differential between Europe and the United
States
arises
regulation.
from
As
sources
other
indicated by some
of
than
the
safety
descriptive
passages in this report, a logical next step would be
an
international
practices.
determine
Differences
if
they
discrepancy.
many
At
aspects
correlate
comparison
of
utility
here
explain
should
the
management
performance
policies
be
with
and
probed
to
performance
observed
in-depth
best, an
utility
good
of
investigation
might
attempt
certain
of
to
management
styles and policies.
As a second layer of complexity, further studies
could
bring
other
consideration.
organizations
and
influences
into
For example, attention could be focused
on the effects of the legal environment in which the US
industry
operates,
and
the
influence
of
organized
public opposition in the form of intervenor groups.
Also of
economic
interest
regulation
on
are
the
influences
complex
performance.
Here,
topics worth examining include economic
to
good
operating
performance,
budget
the
effects
restrictions
on
of
plant
of
specific
disincentives
capital
and
quality
and
performance, and the sometimes contradictory objectives
of economic and safety regulation.
101
7.0
References
IK. F. Hansen and D. K. Winje, Dispgrities in
Nuclear Power Plant Performance in the United States
and the Federal.. Reublic of Germany (Cambridge, MA:
MIT-EL 84-018, 1984).
2 Christopher
T. Wilson, A Numerical Comparison of
International Light Water Reactor Performance 1975-1984
(Cambridge, MA: MIT-EL 86-007, 1986), pp. 290-92.
3
Seth David Hulkower, The Effects of Regulation
on the Performance of Nuclear Power in the United
States and The Federal Republic of Germany (Cambridge,
MA: MIT-EL 86-008, 1986), p. 75.
4 Ibid.,
5 Wilson,
p. 2.
p. 18.
6Ibid., p. 37.
7 Hulkower,
pp. 17-18.
8 E.
Beckjord, et.al.; International Comparison of
LWR Performance (Cambridge, MA: MIT-EL 87-004, 1987);
p. 3-29.
9 Comissariat
(CEA); The
g l'Energie Atomique
Safety
Institute
IPSN
Protection
and
Nuclear
(Fontenay-aux-Roses, France: CEA); p. 18.
'0 Ibid., p. 2.
'iIbid., p. 3.
12Ibid.,
pp. 4-5.
13Swedish Nuclear Power Inspectorate (SKI), A
Presentation of Our Activities (Stockholm: SKI), p. 15.
14Ibid., pp. 4-5.
5
sIbid.,
p. 3.
16Ibid.
17Wilson, pp. 293, 295.
1eSwiss Federal Department of Transportation,
Nlear
ederal~.......
Swiss ......
F...
and. Energy;
Communications,
_Organization
Tasks__and
_(SK),__
Safety Insectorate
(Wurenlingen, Switzerland: HSK, 1984); p. 5.
102
Revision 10
IDate - 4/84Q
OPEC II CAUSE CODES
:)IOPEC II C'AUSE CODES, (Cont.)
First
Level
Secorl
Level
Thiull
Level
Third
Level
6.
7.
Notes I. Code check valves "I" in the Third Level.
S.
First
Level
Second
Level
01
Undelined Failure
02
Fuel and Core
0)
19
I - Miscella-eous
Miceallaneous fuel (such as PWIt preconditioning)
I.
6 - Core/luel problems
I)lrnahle poison problems (e.g., IPRA vibration BrAW
2.
-1973)
Fuel failhres/lRCS activity (PWR)
3.
Foreign oblict
4.
3. IIWIR l'CIOM (Event No. mnust fAllow event tillis is asoci.Ited willth
7 - Operational Restrictinn
Poison curtalin clhanges (I)WR)
t.
Control rod repatch (PWR esp. tB&W)
6.
Restrictions
8 - Mechanical
Increase core DI)/P (NHOH addition)
I.
9
(crud accumulation)
Poison curtain vibrations (1WR-VYPII-1973-191)
2.
191)-1976)
LPRM vibrations (IWR-9
3.
Fuel failure - offgas limits ( Wi)
4.
Fuel deislfications
5.
10
Control Rod Guide Trube nut (RlW 1981)
6.
10
Control Rod (;uide Tube (CE 1977)
7,
9 - Safety Restrictions
soak)
fuel
(includes
changes
rod
OWR control
2.
ECCS peaking factor (PWR)
3.
4. EOL scram reactlvity/rod worth restrictions (includes
shutdown margin)
Core tilt/Xenon restriction (out of Ilux band)
3.
BWR thermal limlits (includes "rod limited")
6.
Thermal power restriction
7.
Reactivity coellicient (e.g., mod.temp. coell.)
8.
Reactor Coolant System
2 - Pumps
Reactor conlant/recirc pumps and motors (except motor
2.
oil cooler-036))
3 Piping/Tanks
2.
Auxiliary piping (I-inch or less, vents.,drains)
Main process piping (include IIPI nozzle/sale end crack
1.
- l&W 1982 and :1i'l thernal sleeve crack)
8.
Flanges, ,nanways, littings
5.
Supports, solcbbers
'a - Valves
9.
2.
1.
Strainers, filters
Core spray piping - IGSCC (no sale-end problemns-03)9/other
Io .zlc prohlemls-0 371)
Rrrirc valve bypass piping (I1WR beginning in 1974 I(;SCC)
Sale.end problem (1WR - IGSCC)
3.
9.
IAuxiliary piping valves
On-oil valves (stop, stop-check, isolation,)
Control valves
Relief valves (excluding primary system relief valves)
Main steam isolation valve (BWR) (lor PWR MSIV-98t46)
Primary system relief valves (PWR pressuri..er; 3WR
main steam) RV, SRV, ARV. PORV, 0PS
Pressurizer spray valves
Recirc. flow control
2.
Flw
S.
i.
6.
7.
S - Instmusents A Contlrols
Irnperathre (I WRI-tnalln steamlhigh teanperatllre)
4
Pressure & Overpressiure protection system (O15)
Level
Nulear (ex-core detectors) (in-core/es-core calibrations)
In-core detectors (TIP, LPRM, APRM, APDM, all IRWR
muclear monitors) (no RM, LPRM vibration)l
S.
Loose part monitorluiclear noise systemos/RllS (B& W)
9.
Integrated control system (ICS) (IlE W)
6 - Ileat ExchangersFanls
2.
Control rod drive cooling lain and heat exchanger
3.
Reactor coolant lpuinp .motor-oil cooler
.
I.
5.
6.
7.
7 - Reactor Vessel ' Internals
I.
2.
3.
4.
5.
6.
7.
R.
9.
3 - Control
tod
2.
).
to.
6.
7.
.9.
Other (including CSS Sparger)
10
Flanges/Seal Ring
Vessel nozzles (excluli0ng sale-end problesm - 03)9) ther,nal stress (see also 01I - IWR CRl norzles)
Internals (e.g., holddown spring - CEI thermal shieldW)
leedwater spargers (IWRs - 1972-1976)
let purips ( ITR-)3 191)-1915)
Sperimnen holders S IT (IRAW 1976)
Control rod guIde tllhbe pin (Westinghouse 1982)
10
Control rod nozzles thermnal stress (WR
WI
see also
9373)
10
A\seimblies
Motor drives/magletic jack drives (PWR)
llydraulic drive (IWR)
Scrami mechanism * RWR accumulator & pilot valve
# A FWS IC (see also I293)
Control roxIs (IIWR -1971- 1974 inverted CR: cracking)
MI.Gsts/ logic/rela.ys/power supply/seqiuence cotroller I
Control rod positiin ind.icatiosn
'- ' ~ I' ~'" ' *. ~..'..L'
CII-~- _I L . -l -CII1 - -~
~--UI.
--- - C.L-II I_ _ .- I-- I-i-. -I_~-~UJILII L Cll C l II --
OPEC II CAUSE COlES, (Cont.)
OPEC II CAUSE COI)ES. (Cont.)
First
Level
04
05
Second
Level
Third
Level
9 - Water Quality/Chemistry
RCS chemistry
2.
RCS chenical clean
3.
4.
RCS boron concentration
Steam Generators
) - lPiing 5G nozzle problern 1979 - see 9'739)
Auxiliary piping (l-inch or less, vents, drains)
2.
U.
Flanges rnanways, littings, handholes
Snubher s/Suppor ts
5.
#6- Valves (except blowdown system valves)
Auxiliary piping valves
2.
3.
Off-on valves
4.
Control valves
5.
Relief va!ves
S- Instruments and Controls
4.
Presscare
5.
Level
3.
Loose Parts Monitor
6 - Bllowdown System
2.
Supports
3.
Piping
4.
Valves
I.
Instruments & controls for blowdown system
7 - Steam Generator Tubes
2.
Caustic attack
3.
thinning
4.
Denting (eddy cuorreot testing & tube support plate deterioration)
8 - Other Components
2.
Spargers
3.
Clad separation
I.
Moisture separatori/moisture carryover
6.
Internal tube supports (excluding denting problem)
7.
Clad prohlemns (other than separation)
9 - Chemistry Modifications
2.
Chemistry changeover (AYT-phosphate changeover)
3.
Sludge lancing
4.
Circulation nodiclications
5.
Chemical rleanimg
6.
Boric acid soak
Cemicnlal & Volsme Control Systemn/RX Water Cleanp System
2 - Pumps
1.
Charging (l s W Ma;keuq Water 'umpus)
Roric 4cid I rainsler Plump
1 - P'ipig/T miks
2.
Au illiry piping,
tI.
Mlain p ecsf
f
piltPl,
la,7.-*s, ftinll- .
5.
First
Level
Second
Level
Third
Level
6.
7.
.
Strainers, filters
tank (IllT, hAT or BAST, VCT)
I)WST, Primary Water Makeup Tank
2.
3.
4.
Auxiliary pipinK valves
On-ofl valves
Contral valves
5.
Relief valves
11- Valves
5 - Instrumentation & Controls
2. Flow
Temperature (mcluding heat tracing circuits)
3.
4.
Pressure
5.
Level
6.
Chemistry
ioration
7.
6 - leat Exchangers
Regenerative
2.
Nonregenerative
3.
Excess letdown
4.
RCP seal return
1.
7 - Chemical Processing Equipment
Evaporatorlconcentra tor
2.
Gas stripper
1.
I)emineralizers (BWR - makeup system)Xsee 071)
4.
9 - Chemistry
2.
3.
06
Boron concentration
Boron straltlication in tank
Condenser
3 - Pipes
- Valves
4
5 - Conlenser
Vacuum ltC
7 - Tlbe
8 - Loss of Vacnn and Back Pressure Limits (see 1674)
9 - Components
Shell/casing
2.
Air ejector
3.
Waterbox (e.g., fouling see also 1674)
4.
5.
Ballies
Stakitg (Palisades, PR-2)
6.
QI
Condeiiate/Fre(dwa er_/Auxillary Feedwater/Mak eup Water Sys tems
2 - I'manps A P oump lrives (elaborate in comminents)
2.
Feedwater (I:W)
3.
Cotmensate/booster pumps
4.
Ileater drain
C:enicaln.ile.innera liter
S.
Other
6.
7.
Ausiliary feecldwater
I - l'iliaK/ I .ank
_~ I Ir I * )I _~~.
OPEC It CAUSE CODES, (Cont.)
First
Level
Second
Level
Third
Level
4
5
6
7
9
0
First
Level
2.
3.
Auxiliary piping
Main process piping
is.
Flanges, littings
5.
6.
Supports, snubbhhers
Strainers, fIlters
Extraction steamo system/heater drains/HDT
De-mineralizer system
FW&S nozzlesI safety portion of FW piping (INEE
Rulletin 79-13)
8.
9.
3PEC II CAUSE CODES, (Cont.)
Main Steam Sstem
3-7lipin T.Tnks
Auxiliary piping
2.
Maii, Process piping
3.
Flanges, fittings
Supports, snolbbers
Strainers, filters
6.
' - Valves
2.
Auxiliary piplg valves
On-ofil
3.
4.
Control
5.
Reliel (SRV & steam dump valve for PWRs) (MSRV
for IWRl-0447)
Mais stealn isolation valves (PWR) and reverse checks
(WSIV for IhiWsI-O)'036)
Main te;ain byp ass (to condeinser) * IIWR dumllp to c:omIdlellser hot well
Third
Level
5-
lnstrolamellts & Controls
2.
Flow
1.
Temperature
1.
Pressure
5.
Level
6.
Moisture separator reheater
7.
Steam bypais/steaim dump
7 - Moistlre Separator/Reheater ( UR)
2.
Drains/carryover/water level control problems
1.
Tubesltube support problems (includin K tube leaks)
I.
M\Sl relief valve problems
5.
Other MSR valve problens
6.
MSI steam supply valve
- Valves
Auxiliary piping valves
2.
3.
On-offI
'.
Control
Relief
5.
FW reg. valves
6.
Extraction steam
7.
Demineralizer systern
3.
9.
FW isolation valves
- Instrument and Controls
2.
Flow
Temperature
3.
Pressure
4.
Level (lncluditgFW heater level controls)
5.
Chemistry (e.g., chloride monlitor)
6.
7. . FW flow control (except r&W ICS-0359) (S/G high or
low trips)
- Hleat Exchangers
Feedwater heaters
2.
Other (include FWP gland seal condenser)
3.
- Denineralizers
Capacity limitations
2.
- FW Chemistry
Second
Level
19
Turbine
3 - Piping
2.
3.
Auxiliary piping
Crossover piping
I - Valves
2.
1.
.
3.
5 - Instruments
2.
3.
4.
S.
7 - Components
2.
3.
4.
Auxiliary piping valves
Intercept/stop
Control/thrott le/governor valves
Combined intercept valves (CIV)
& Controls
Instruments
ElIC/supervisory system/ (GE-permanent magnet)
Pressure Regulator ()WR-IPR EPR * MPR)
Turbine Protective Devices (overspeed test)
Shaft/blades
Bearings
Gland seals
5. Turning gear
Casing
6.
Turbine Balancing
7.
8 - Oil System (dlo not include bearing or EIIC problems)
10
Generator
5 - hnstrusoieoits & Coltrols
2.
Instruments
I.
Logic/controls (iicluiding under frequency relay)
4.
Core monitor (i,.aiza.lion detector)
5.
Voltage regulator
7 - Auxiliary Systeens
2.
Ii Cooling
1.
t120 Cooling
4.
Ol (Ihf bhearilrs, control, seals)
1.
fls ,lct/leais/lml
1S 1,h:l ,.loling
r
3 - Componl ets
2.
Exciter (perman.rent Imalnet except it%GE -EIIC)
OPEC II CAUSE COI)ES, (Cont.)
OPEC II CAUSE CODES, (Cont.)
First
Level
Second
Level
Third
Level
3.
4.
1.
6.
7.
3.
II
1
Rotor
Stator
Shaft
Bearings
Bshing
Turbine-generator -exciter shaft :ouplieg
Electrical Systens
7 - Trasnsforners
Main
2.
3.
Other (starteep, statimin auxiliary)
S - Switchgear/bses (except instrunent & safeguards bises)
9 - Safety-related equ ipment
Ininterruptable power supply (DC power systemrn 125 VAC
2.
initrument buses, inverters, MG sets, relays)
Emergency diesels (including oiutput breakers)
3.
4t. Essential/vital saleguards buses (saleguards other electrical)
Elc:trical connectors
5.
Gas tourbines
6.
of RTS or spurious trips not caused by actual
Reactor TipSystem (Only failueres
2
trip parameters or detector/transmlitter failures)
2
6 - RCS Input Channels andi Other Chaunnels
RICP breaker
2.
6
Mode switch
3.
7 - Reactor coolant system (RCS) Inpuit channels (continued)
instru-nentation
Nuclear
2.
RCS water level
3.
Reactor pressure
4.
RCS coolant flow
3.
RCS temperature
6.
Pressurizer pressure
7.
Delta tenp./low )NIR trip (CE - Thermal Margin * Low P)
8.
Under voltagelnder frequency trip
9.
8 - Secondary system inputs
2.
ruirbhne/generator inpults
Feedwatersteam flow mismatch and low steam generater
1.
4
level (steam &.feed suepture control system)
Main steam isolation valve closure
4.
5. Main steam activity r temnperalture
Main steam presure
6.
7.
Feelwater Ilow
Stean generator level
3.
Main steam flow (higlh or low)
9.
Other
channellcoumponensts
9drywell
2.
(Cowtainaetl
3.
4.
1.
Safety injectin siginal
RCS leak dleterlctiol
Cowtainmeneut high pressire
First
Second
Level
Level
Third
Level
6.
7.
8.
9.
I
Auxlitary 5yste!_s
I - OfI ,a.s
Logic/relays/pernlissives
Manual
Reactor trip breakers (see also 0184)
Rod drop slgnal
systeens (AOG)IVentilation
I.
2.
3.
Systems
Of Igps (exclhding recombiners)
Reconhiner (IWR)
Plane Vent/Filter System (PWR see also 2233), (SI3GTS
(BWR), (SLCRC-BV)
8 - Other systemns
2.
Instrument/service air or nitrogen
1.
Radioactsve waste (RMCII) * RISM5
(area and process
monitoring systevms)
4.
Process conputer/RWM/rod block/DDPS (at TP + SL)
5.
Auxiliary boiler
6.
Fire Protection Systemn (including fire barriers)
7.
Meteorological instruments
S.
Seismic instru nents
It
RefuelioggMaintenance
6 - Care physics tests
7 - Refueling
I - Reflieling equipment problems
9 - Maintenance
I
Utility
6
7
9
16
Circulating Water/Service Water System
2 - Pumps
2.
Circulating water pump
3.
Service water Iump (V - River Water System)
4.
Cooling tower circulation pinmp
5.
River Water Pumps (Farley) Aux R WS (RV)
I - Piping
2.
Auxiliary piping
).
Main process piping
4.
Flanges. manways, fittings
5.
Supports, snutbbers
6.
Strainers, lill rs, fish & trash rake (see also 1672)
4 - Valves
2.
Auxiliary piping
3. On-ofl
4.
Control
S - Instruments & controls
Grid (Noneconomic)
- Other offil-site grid problems (grid maintenance)
- Loss of load/load rejection
- Loss of off-site power or of f-site caused under-voltage
condition or other electrical disturbance
OPEC II CAUSE COI)ES, (Cont.)
First
Level
Second
Level
Third
Level
Flow
2.
Temperature (for circ. water temperature limits - 1674)
3.
4.
Pressure
5.
Level
6.
Chemistry
6 - Heat exchangers
Cooling towers (no CT temp lim-1674/ultimate heat
2.
sink/saleguards C r-2 62)
3.
rube/shell
4.
Intake heaters
7 - Intakes/discharg.s (spray-pond problems)
2.
Foulinglicing
3.
Structural failure
4.
Cooling water temperature/design insufficiency/
high temperature
5.
Excessive fish kill
Discharge diffuser
4.
EPA discharpe limit
7.
9 - Water treatment
17
Thermal Efficiency Losses
19
e yniection System
Core Cool in/a!
2 - Pumlps (containment spray pump - see 2223)
2.
High pressure core injection
3.
Low pressure core injection (except RHR pumps)
4.
Residual heat removal (RHR) (no IIWIR-see LPCI)
Rfecirculation (I/ORSP - BV, 5, NA)
J.
RCIC (11WR)
6.
7.
Core Spray Pump (LPCS)
3 - Piping/tanks
2.
Auxiliary piping
Main process piping
3.
4.
Flanges, fittings
5.
Supports, simbhlbers
Strainers, filters
6.
RWST & CST (1I&W - iWSTI BWR - DWST see 0538)
7.
l'Accumulator (Core Flood Tanks, IT, U
8.
Accumoulator (Core Flood Tanks, SIT, 'JIll)
9.
4 - Valves
2.
Auxiliary piping valves
On-oil valves
3.
I.
Contral valves
Reliel valves
6.
Check valves
7.
1%ILC explosive valve
OPEC II CAUSE COIDES, (Cont.)
First Second
LevelLevelLevel
Third
I - Instrirnents -i controls
2.
Flow
Temperatures
3.
'i.
Pressure
5.
Level
6.
Safety injection actuation (logic circuitry/actuators)
7.
LPCI Loop Selection Logic
6 - Ilea exchangers
2.
RIllR heat exchanger
3.
Is)lation Condenser (IWR)
9 - Chelnistry
20
Initial Plant Startutp/Oerator Training
7 - Sta.rtup testing
3 - Power isension/reduced power operation
9 - Operator Training/Enrgerncy Plan Testing
21
Paired Unit Inpact
22
Containment System (shield building)
2 - Puitlps
2.
I)rywell pump
3.
Containment building (quench) spray pump (no IIORSP-1925Xprimarily on BWR-2 units)
3 - Piping
2.
3.
4.
5.
6.
Auxiliary piping
Main process piping
Flanges, fittings
Supports, snubbers, high energy line problems, general
snubber inspection
Strainers, filters
4 - Valves
Other valves (e.g., BWR vacuum breaker)
Auxiliary piping valves
On-oil valves
Control valves
5. Relief valves (include PWR vacuum relief)
6.
Containment building (Cl) isolation valves (except MSIV
& FW IV)
7.
CI purge/lexhaut valves (torus vent valve)
S.
DiWR MSIV leakage control svstem valves
9.
Vacullm reliel (IWR)
5 - Instrnuments & controls
2.
Flow
3.
Temperature
4.
Pressutre
$.
Level
I.
2.
3.
4.
OPEC II CAUSE COI)ES, (Cont.)
First
Level
Second
Level
Third
Level
6.
7.
S.
CB Isol actiuation (VIAS)
CB spray achtation (CIlA)
G(.s anl.lyzer (i WR - Containment Inerting System)
(PWIR 11-2 analyzerXsee also 2237)
1
6- UIsat exchansgers
I.
i)rywell cooling
2.
Ice cooldensers
Recirculaton lasi coolers (see also 2283)
4.
Casing Cooling System (NA - ssbatlnoslJleric)
7 - Containment strs:tures
2.
Torus
3.
Penetrationss
4.
Containment leakage
IDrywell
5.
Liner (e.g., WCPS)
6.
7.
Airlocks finchdlng airlock seal leakage)
S.
Integrity (ILRiT-pp i rests - 5th Tier-01) App 1 Test
or unknown
$ - Fansl/air filters (Plant vent filters & SBGTS- I)7)
2.
Recirculation air fans & dampers-PWRs (B il s-drywell
coolers)
3.
CB ventilation fasts and ducts, and outside filters s (1
2
pusge (see A.do 137) & 2263)
I.
CB charcoal filters (inside CB)
J.
Vacuum pimips (sIb-atmospheric contalnohents)
6.
Penetration cooling fans
7.
IIZ recomsbiner (PWRs)/blowers (see also 1372)
23
Component Cooling Water
2 - Plumps
2.
Component cooling water pump
2.
3. '
4.
Auxiliary piping
Main process piping
Flanges, fittings
Supports, sauliers
Strainers, filters
3 - Piping
5.
6.
4 - Valves
2.
3.
Auxiliary piping valves
On-oll valves
Control valves
5. Relie valves
5 - Instruments & controls
2.
Flow
3.
Tellperasturr
4.
Pressisre
5.
Level
6 - Heat exchangers
2.
Cooling towers, ulltinate hat sink CT (no tnormal CT- 1662)
3.
ThelNell
4.
OPEC 11CAUSE COoES, (Cont.)
First
Level
24
%S.cond
Level
Third
Level
Problems
Structtsres !*nter_.system
7 - 'trltuctlrns
ContrAl hulding (e.g., Trojan 1978 problem)
2.
Auxmliac y bll g. (e.g., Salem 1980)
3.
Mae steam tiunnel (see also 1285 or 1173)
4.
S - Electric:al
Cable routmng
2.
Cable splices and electrical connectors
3.
Browns Ferry lire
4.
San Onolre I cable fire
$.
25
Economicr
7 -Fulel "c onomlic
Coast to relueling/fuel depletion
2.
3. Fuel conservation
- ;rid ecoiomiic
2.
Low-system demandspinning reserve
3. Load following
99
Left Over
9
9
10
10
I0
10
I)
FOIJR
FOIURTI-I LEVELs PRORLEM MOIE/EXTENT
r-i
LEVEL: PROBLEM MOI)E/EXTENT, (Cont.)
we r
00
01
No failure, abnormality or malfunction
Unknown failure or problem undefined by below categories
Active Conponents (Pumps,Valves 5_nubbers Motors_e..)
(or jist says failutre)
to operate, start or ranm
toral
Hf.ilhare
50
Failhie of automatic functioning but not mnamial
Si
52 Failore of control (operates but erratic or incorrect control)
Mispostioning/inisaligniing/spuious closure ('ASIV)/misoperation (e.g., erroneous
S)
start of pump). Note: component still workimg but did wrong thing
Performance degradation (operates out of spec)
54
Flow performance degradation/llimits
5s
adation/lini ts
Level dtegr
56
Time perlormance degradation/himits
57
Chemical degradation/limits
53
Radioactivity limits
19
Noncurtailing degradation (operable bu)t nondisabling problem present)
60
seat leakage
Valve
61
ruction inadequacy
Structral/esig
62
Passive System (Pressure Boundary_ Pipes Strainers)
6 -- otal Asilr-e Icomplete loss of flowi)
Performance degradati-on
66
Flow performnance degradation/limits (fouling/ice formati n)
67
Level perlormnaice degradation/limits
68
Time performance degradation/limits
69
Chemical degradation/limits (water chemistry)
70
Radioactivity limits
71
Noncurtailing degradation (operable but nondisabling problem present)
72
Unspecified leak
73
711
Defective weld
Crack - no leak (stated)
75
Crack - leak
76
Tube Ieak
77
External leak from seal, gasket, packing, etc.
78
Valve seat leakage
61
Structural/design inadequacy/construc tion deficiency
79
A tive ElectricalCompo!qen s Generators. Relays, Breakers, Switch, Detectors,etc.)
i0 Total fai re to operate or perform (or listsays failure)
Failire of automatic functioning but not nanlisal
S
Failire of control (operates bit erratic or incorrect contral)
32
Level, #53))
48 Misl sitianinu/Mis~peration (see mIth
Per ,frinance degradation (operates ouit of spec:)
83
Current degradatioln/limits
3i
1 ile performance rdegradatioii/limits
85
Voltage degradation/limits (lIV) (sre also #93)
86
5etpoint drift
38
Noncur talling degradation (operable with nondisabling problem present)
87
89
Spilrios trip
49
Str: tural/llesign/construcltion iinadeqiai.cy
Transformer)
CCPo
Passive Electrical SysteFslBuses. CCaCles,
Total hI.ijlrlsorts, laults, grounds, .arching, current interrupts fron blown
90
fuse resllting iincomplete loss of power)
Performance legradation
91
Cue rent degr idation/limits
92
Voltage degradatio/llimits (IV, vollage spike) (see also 086)
73
Noncurtalllng ,legadatilin (e.g., cracks in insulation-corroded contacts)
)4
Strlctr.lilde;igin'/constrc ti o ialadclqsacy
95
Old Fourth Level Coding (may be encountered in some pre-1978 data)
Leakage/Crack
External leakage (packmng/seals/gaskets)
04
Unspocilied leaikage,
I's
.
Tiube leakage (Il exchangerlcondenser)
26
External l-akage - cracks
3I
Delfective weld
32
Valve seat Ieakage
18
Activity
05
1ligh
06
I ligh
07
Iligh
08
High
gaseous
liquiid
process
area
(external to plant)
(in plant)
Components
Valve operator
16
Valve component
17
Valve seat leakage
IS
Relief valve lilting/failure to seat
21
19
Local control (pump to valve)
20
22
23
24
Pump components
PIimp drive
Pump bearing
Plnp \/(; sets
13
Oil seal
Other
Water chemistry
02
13 Water tesnperature
Folihng - healt echan.tgerlintake/discharges, etc.
27
Structural fI.ihlre
29
I)efectve weld
12
25
Elc,:riral ponetr ition welds
1
Of1g:s e.xplosion
FIFT1H LEVELs EXTERNAL CAUSE OF EVENT (Cont.)
FIFTH LEVELs EXTERNAL CAUSE OF EVENT
00
01
Not specifically specified as an external cause
Extrnal cause-not covered below
NRC Orl!inated
02 -egIla-tory/Operational limit (Safety Limit of T.S.) (use only if outaKe results)
Regulatory reqelire'nent to inspect for possible dleficiency
0)
)49 Regulatory require nent to nodiy equipment due to snalhmction or construlction/design deficiencies
IReg\uatory requirement to modify equipment dse to more restrictive criteria
)5
29
NRC licensing proceedings and hearings
UInavailsbility of safety-related equipment (tse "92" if limit or restriction
19
is known)
Plant
06
07
08
0)9
50
26
52
53
Originated External Causes
Festing (use For power red. or S.I). fOr test, not trips or failure during test)
Testing error
these Include procedural inadequacies
Maintenance error
Operator error
Personnel involvement sispected to have precipitated protblen/failure
Noen-NRC precipitated derating (discretionary derating-no equipment lailure)
No (ailhselproblean - ptirely preventive maintenance or inspec:tion
No mentios of problem (possible preventive maintenance or NRC required
nod - no degradation stated)
External Eqilpve ent Malfunction
upply
" "aster
I
.i""I'incTiono I
12
Mallriction of local instrument air/service air supply/cooling air
supply
oil
of
local
Malleuction
13
Mallnction of local electrical supply
It
Fuel supply
17
Local controllinstrumentatin (part of component package)
18
15
Other auxiliary system malfunction (e.g. cooling)
BWR
30
31
32
3))
14
35
36
17
40
41
42
43)
44
Fuel Limits
Ci seq. or pattern change ( every 5 EFPW)
CR Ad). Extensios, etc.
Periodic reduction for testing at CR adj. frequency (in BWR-TV test-/llmonth)
(no mention of CR adjustment)
Periodic redction for load following at CR adj. frequency
Pet indic reduction for unknown cause at CR adj. freqeinccy
Periodic reduction for other reason at Clt adjisst frequency (no nmentills of
CR)
Weekly reduction for testing
Weekly reduction Ifor other reason (suspected testinR)
MCIR
MAI'LIIGR
General thermal limit or comb. of 41) & 41
Persistent undefined nIessing arornd between 80% & 100%
Load drop (suspect fuel thermal limit)
Cracking
71
72
73
Vibration induced crack
SCC indulcedreack
Thermal latigue induced crack
Other
4 Pl'ant internal or externial environmnental effects (lightning, icing)
Fires
25
61
Bearing malfmction
42
Pump or valve-drive mechanism
Combination
91
Cosniations:
62 (drive) & other Sth-level coding
92
"
07, 0, 09, 50 (error) & other 5th-level coding
93
02, 01, e04,05 (NRC) & other th-level coding
9
"
3X, 4X. (BWR fuel) & other Sth-level coding
95
"
61 (learing) & other 5th-levelcodling
98
Multiple combination of 91, 92, 9), 94
99
Other level 5 colnhintation
SIXTtl LEVEL: METHOI OF SHUTDOWN
NOTES:
Use coding of methld of shutdown reported in Gray Books (except for
Classilication 4 and 9).
I.
2.
I
2
3
4
5
6
7
8
9
2
I
4
S
6
7
8
9
2.
-to shutdown).
3.
3.
Classifications marked with an asterisk are different than those in the
Gray Books.
4.
4.
The number of startups equals the number of outages of Classification I,
2. and 1.
SEVEN r1t LEVEL: SIUTDOWN PARTICULARS
0
NOrES:
I.
Rather than using the Gray Book Classification 4, a consistent coding
of thile method of sltdown shoul he used for each outage if it extends
into mdultiple mothly periods (unless thile status c:hanges - e.g., reduc:tion
Not an outage or reduction (work which is done with unit operating and no
derating - e.g., LERS which occur during operations but cause no shutdown)
Manual (includes manual shutdowns with a unit trip at low power
Manual scram
Automatic scram (Including test performed with intent of tripping unit)
Turbine trip with no reactor trip* (reactor trip is assumed unless told differently)
Load reductions
Continuationl of other outage (no interstitial electrical generation)*
Noncurtalling or concurrent work or inspection within an outage* (including
all ROs during an outage)
Related transient (te., second failure that follows from events of an earlier
failure or ahbnornal functioning during the e.rlier failure)*
Unknown"
9
!I;HMfl
No extra information or no shutdown
cause)
51 signal generated (only oie per 51 generated--with most immediate
Inllowng this
There are related trasients or signlirant suhsequeit events
event (:nust always precede event with S it (6th level)
2*5 7
ionltest lailure (not sit operaoutage/incident results fromna testlitspecti
rithe
tiona.l failure)
S:heduled test performed with intent of tripping the Itrbine or reactor (~ee
0 1, Level (,)
3 62
3,5
3.2
00
01
99
02
03
LEVEL: SIGNIFICANT TRANSIENT INIT.\TOR
Significant transient initiating events or precursors should be denoted
in this code classification.
Initiating events are to be coded (i.e., are significant) if they result in
or would have reslted in a trip or runback of the unit (e.g., loss of feedwater flow should he coded tmily if tme reactor trips or a Ieedwater pump
trips) except for tlhose transientts where no trip is expected (no operatiots
out of spec withot a resultioiy, trip should be coded here).
Such signqificant transients identilied in WfSlI 1400 are identified by
in ast.-risk(').
Multiple significant transient initiators in a single event should be denoted
hy a new code classification.
No significant transient initiator
IUnknowsn or uncertain significant transient initiator
Unclassified as yet
Reactor trip (no other initiating events noted) * spurious reactor trips
Generator trip
Turbine/Generator Transients
05
rucrbine trip (verspeed trip)P
06
Turhine trip (other)*
07
Iligh steam flow" (no RV or Dump V stuck open) (see 77, 31)
08
Low steam flow*
42
Loss of load with subsequent loss of off-site power
43 Excessive load increase with subsequent loss of off-site power
44
Loss of load
45
Excess load increase"
46
Furbine trip with lailsre of generator breaker to open (failure to relay auxiliary
loads to off-site power)"
47
Other
Electrical Power Transients
40
Loss of power on an auxiliary bus (6.9 kv, 480 v, 120 v)
41
Loss of AC power from off-site network' (station blackout) (include partial
blackout)
42 Loss of load with subsequent loss of ofI-site power
43
Excessive load increase with subsequent loss of off-site power
41
Loss of load'
45
Excess load increase*
46 Turbine trip with lailure of generator breaker to open (failure to relay auxiliary
oadls
to off-site power)"
47
Other
ieedlwater, Conldensate. Circulating Water and CVC System Transient
II
Loss of FW ow - F W pump or punp (ls
sie problem'
12
Loss of FW flow - malfunction of VW flow con"rol' (including valve failure
.A.l low V; l'vel)
II
I;Icrease of FW flow - malfunction of FW flow control" (including valve failure
h
.1id higlSC level)
IS
Loss I ,cotlensate pulips'
EIGMIII
EIGIIHTH LEVELs SIGNIFICANT TRANSIENT INITATOR, (Cont.)
Fredwater Condensate.Circulating~Water and CVC System Transient, (Cott.)
tmittrip a rmh.s':kl
i
6_ Lou of circulatig arpnsps (msing
Loss o conderser vacutum
17
change)
step
(60o"
heating'
FW
Loss ol
13
CVCS or other tmalIflin:tion resulting in RCS boron dilution (no trip necess.sry)
19
Radiation release Iromln CVCS or other system in ausxiliar y buildin K (no trip
91
necessary)
Reactor Coolant PunmpReclrculation I'urmp Transiens
caused by strings ol inactive RC or recirculatitn loop
FTrp
i
Itecirculation flow control f.ailore - decreasing flow (IIWR)"
52
Recirculatin flow control failure - increasing flow (IWR)*
53
RCII trip or malfunction' (including sthaft break and partial loss of Ilow)
5.
RCP sei.,re'
15
7
ItCP seal lailure
Reactor Coolant System Pressure and Temperature Transients or SI or RX I
epressurization of primary system (no leaks) no trip necessary)
S
Inalvcrtent
56
Inadvertent overpressurization of prinary system (no trip necessary)
57
Excessive cooldown or heatup rate (no trip necessary)
58
Iligh or Low RCS Tenop. (no trip necessary) or T&P lincuations
59
Pressuse and/or level fluctuations inclhding bubble in RCS (no
60
trip necessary)
Inadvertent SI or spurlon SI signals (IlPCI pump start - OWT)*
65
Accidental depressurization of the main steam system (e.g., stuck open RV)
66
(see 34e)(no trip necessary)
Reactor trip (no other initiating events noted) * spurious reactor trips
02
Control Rod Transients
conrolie; d (r Improper) rod (assembly or bank) withdrawal at power'
61 tl
Uncontrolled rod withdlrawal during startups
62
Control rod assenmbly drop or misalignment' (no trip necessary)
63
Rod ejection or CRI)M Iouusing rupture
76
Valve Malfunction Transients
inadvertent opening of a tuerbine stean
of control resulting ilt
Malfutirf
31
bypass valve* (see 07,77) (or stllck open bypass valves)
Closure of MSIVQ
32
Spurious operling of SlV (RCS for PWR, MS for GWR)P (no trip necessary)
33
Stuck open 5/RV (see 66)
34
Spurinus opening of S/G PORV (I'WR)O (no trip necessary)
Is
t ystem/Steam Leak (LOCA .and Steam lireak)
Reactor Coollant
LargE tiCS leak ifroum equivaienl 6-incih diameter Ile or larger)l
71
leak'
IICS
Small
72
Stle:k open 5/11V
34
SG ittte leak* (P'WI) (see 9th L 051)
73
74
ICP seal lailure
75
IICS systemr Imndary valve failre* (leak past valve - see 072) (9L I042)
('tIl onecharnisun lousing rupture or C1 ejectixo
16
linehoreak' (allsi.es) (see 07,11)
77
Ste.tian
7I
FW pipe break' (all siz.s)
46
07
0O
LEVEL: SIGNIFICANT TRANSIENT INITATOR, (Cont.)
Arcidental depressurizati'n of the main steam system (e.g., sturk open RV)
(u) trip ,e-e-sary)
lligl st Dil flow' (no RV or Dulmp V stuck open) (see 77,31)
Low stedan flow*
Other Transient Initiatom s
ri(- trip necessiury)
miu
h e
li.flueling ac:cident (e.g., dropped fuel assembly or dropped RV component)
36
(I.) trill ncr(ess.ry)
Shipping fNel pool accident (e.q., cracked pool liner) (no trip necessary)
37
Slpling c.ask storage tank release (no trip es-esary)
88
Waste gas itoi age tank release (no trip necessaty)
39
Lititd waste storage tank release (no trip necessary)
90
Raliation release fro:n CVCS or other systen in auxiliary building or outside
91
CI1 (except 39 and 90) (no trip necessary)
Environmental (tora.ldoes, Ilooding, lire, earthqllake) (no trip necessary)
92
Contral room uninhabitability (9th Level - 194) (no trip necessary)
93
Loss of Service Water System (9th Level - 041)
91
Loss of Istru'nent Air Systems (9th Level - 101)
95
Loss ofl itIR flow (no trip cessary)
96
NINTH LEVEL: SIGNIFICANT SAFETY SYSTEM PERFORMANCE
NOTESs
I.
NINTHll LEVEL: SIGNIFICANT SAFETY SYSTEM PERFORMANCE. (Cont)
A failure unless otherwise specified is defined as follows:
a.
b.
System failure of a systein to meet minimum design requiremaerts (e.g., miniulumo flow through proper paths).
Component failure - failure of the component to perform
as intended or meet specifications. Examples:
(I)
(ii)
iii)
= Reactor TripSysten (or ROS)
Syst-em l;ailue (failure of more than 2 full-length CR to insert)
Uninserted CR following trip
ItR5 i&C or logic unconservative malfunction (no spurious trip - or tripped
channel) (refer to LER Prompt Il, 30 DIay Report 01) (including detector)
035 CRD or CRD system problem (no dropped CRi)
RS
0l
0)2
031
Valve failure - failure to operate, or in wrong position,
or leakage past valve
Tank laihlre/incapacitation ; leak, or level, or concentration out of sl)ec (inconsistent with specifications)
Pump falure - fahilre to start, keep rimnning, or deliver
design flow
SI or ESF Actuation
036 System failure
037 Logic or IC unconservative malfunction (including detectors)
2.
If there is a component fallhre that results in a system failure as well,
the outage should be coded as a system failure.
3.
Multiple failures or failures of safety systems not covered by the following
coding classifications should be given a new code number.
4.
Most LERs that are codled should have an entry in this level.
000
001
No unconservative failure in safety system
Unknown
999
008
009
Unclassified
No lailure - procedural inadequacy
No failure - QA inadequacy
8
Unanticipated or Common Mode SafetlEjvent
Oil Nonconservative errors in SAR accident analyses or Tech Spec bases
012 Commnon mnode incapacitating of safeguards equipment (Oescribe in comments)
'13 Disagreement with predicted value of reactivity balance (LER Prompt Report 14)
011 Structural inadequacy potentially affecting salety
2
013 Piping inadequacy potentially lfectillg safety
012 Snubher olt of compliance
016 Electrical cables - potentIal inadequacy affecting safety
2
017 Other potential inadequacy affecting safety
018 TMI modifications
7
EP = Electric Power
021 System failure (insufficient AC or DC power to safeguard buses to operate
mninium ESS features)
022 Diesel generator lailure (failure to start on demand, while running, to load
on bus, or of a suppor t system that would incapacitate the DG)
023 Gas turbine failure/hydro unit failure (Ocosiee)
024 DC power supply faihlre (battery, inverter, etc.)
027 Safegiards electrical failure (other than bus)
028 Offsite power sour:e problem
025 Safeguards mlsfii ilre
026 Instrurnent bs failure
028 Other oflsite problems (no Oconce Ihyhk uInit)
1
10/
Containment and Secondary Containment Problems
040 Other Cl1 lailuse - not listed below ( e.g., low torus DP)
]
041 Large leak ( Xcfm) in airlocks, penetrations, etc. + system failures in CIAS
039 Other containment subsystem system failure
042 Small leak ( Xcm) in airlocks, penetrations, etc.
043 Containment hIalation System I&C, logic, actuating circuitry failure
044 Colltainment isolation valve failure and leaks (including ventilation valves)
015 Containment v.nt/pup ge - Standby Gas Treatment System (SIGTS) failure
046 Hlydrogen reconbiner failure
047 Weld Channel Pressurization System component failure (PWR)
048 Containment recirculation fan/cooler/filter system component failure (PWR)
Drywell Cooler (D0W1)
049 Post-accident pressure relief system component (PWR) + Vacuum breaker (BWR)
012 Sioubber or piping support out of compliance of failure
0)3
tSIV I.ilre (includling leaks)
057 BWR MSIV leakage control system (valves & controls)
954 FW isolation valve failure
055 Ci%vacuum pump failure
2
261j Contain.nent Depressurization Actuation IC failure (PWR)
507 AD)S acteuation problems (BWR)
056 Cl) Inerting System component failure (include gas/l2 analyzer)
2
NOTE:
Safley
061
062
(see
06)
64
2
2
9
9
065
For CB Spray see 30Osfor OWRs and 260s for PWRs
3
ys m loumdarX Abnormal Degradation (LER Reportable Item Prompt I),
I fuel cladding
In reactor c:oolant pressure boundary (including RCS - not CO - isolation valve
Ie.k)
alve) In primary containment
In other c:ontainer of radioactivity
I.oose prt i.n Rt:S
IQ
Loose part in Stean Generator
10
Saet Stem Coolant System (PWR
- IIiSW)
01"item failulre
072 (IRS/IIIPSW pump failire
073 Ileat exluanger failtire
074 System valve lailure
07
systei hNC failuiew
076, O)tlh .r Ialilar
CRS or Component Cooling Water; BWR
.~4
r
~ ----
-- ---
----- -
- -(l~-W1---- -l -I
_
NINTH LEVELs SIGNII:ICANT SAFETY SYSTEM I'ERFOIRMANCE, (Cont)
Emnerency Service Water System (Level 8-94)
(NSRW)(IIV - River water pumps)
System fail;re
Emergency service water pump failure
0i1
082
05
System valve fallure
o084QSystem l&C lfailure
lal,.re
085 Other
ECIR (lIPRSJ(lE nerg.ency Coolant
Ultimate (Safegards) leat Sink
System valve failure
System hIAC failrwe
Other failure
092
09)
094
Service !j.t eos
i0--Str'i;a,,it Air System coempinent lailure (Level 3-95)
102 Fire Protection System component flallre
103 Failed luel detector failure
104 CCit lamitahility systems component failure (Level s-9))
125 Itadiation Moiloring System component failure
106 Filter Exhaust System outside ol C(l - IIVAC (IWR area coolers)
045 SIIGTS (ltWRs)
Othler
1i20Turbine
ecirculation lor Snall IBreak LOCA)
2)2
253
214
Syslei valve failure
IU.C latilre
Olher faimare
C51
(Collainmnent Spray Injecttil System) (CSS)
262
261
264
265
Ccotainment slway pump failre
System valve failure
System tAC failre (include CIA)
Other lailure
CSRS (Crntainment Recirculation System) (I/ORSP'
272
213
214
215
Containent recirculation pump failure
System valve faihure
System I&(C ailure
Other failure
stop valve (TSV) or CV does not close or slow response
SISA I Sliwn lkoxidelldr.azime Addition)
ESS (EMERGENCY SAFEGIJARDS SYSfE~LS)
PWR
ECI (Emerfenr.y Coolant nLe!!cio,)
i.;iia
m of aruumator, LPIS, SIMS, etc.)
Si i--ie,
Rtelueling Water Storage Tank (RWSI) failurefincapacitationl
IHeat tracing system comptinent laihlre
RCS leak dete,:tio system
212
213
214
LPSI Failures
i~
220
222
223
224
225
cc
lator failureincapacitatiin
1liR pump lailure
LP pump failre
System valve failure
I&C failure
Other failure
11I1S
239
2)31
232
213)
214
2)1
216
217
238
239
Failures
Charging pump failure
lP pnp failure
Boron Injectius Tank (I I) lailre/incapacitation
oric Arld Tank (IBAT) lailtrrelincapacitation
System valve failute
I&C failhre
Other component failure of II'IS
Other component lallure of CVCS or Boron Addition System
tipper head injection system valves
Upper head injection accumnulator and other
-i
LPRS (Low Pressure Itecirculation System)
241
System inltun*
2'12
243
2,4
%ystein valve lailere
INC fhailnme
'1tt:erI.ihl.e-
282
23)
234
235
NAOtI or thydrazine) tank failhrelincapacilation
System valve failwre
SystemSIC failure
Other failkre
AF:WS (A.xillary Feedwater System)
ai7
SI Fyi;iter
214) I hlspiecilied AF W pump fallHre
212 Small AFW pump failure
291 Laspe AFW pump lallte
2 '9
Cnlensase Storage rank(CST)
295 Spslem valve lailune
296 Sysitem SAC failure
2'11 Othlr failere
Saley
1i1
102
1111
)l
0
10)
11
312
306
1O)
18
l
110
ailure
adIl Relic Valves (Slit)
i
pressurizer- cde safetly valve flailre to open
lItS i ossuleer code safety valve failre to reseat
IiCS pressurizer SIl valve fallere to s)pea
iCS pressurizer code S/it valve failure to rese.at
ltCS presslts er Overpressure Protective System failure
IiCS pressurizer safety other problem
1C%pressurizer relief valve other problem
\1S safety valve failore to open
MS45
salty valve lailure t, reseat
MS relief valve failure to open
MS tille! valve failkre to reseal
M SRV other prolemn
R I t System
SI "'i i
'yste,. fail-ire
lnm p l.ilute
221 ltll
it) 11Iit v.lve 1.1iime
21 till, s' I. so.1I
,.
It
, II*f
t.t*
l I
s1.,.
ESS (EMERGENCY SAFEGIJARIDS SYSTEMS), Cont.
ECI (Emerncy Coolant In .tiof)
411 System ailrie ;not listeCd below
1'12 Condensate Storage Tank (CST) failure/incapacitation
413 MIJfl-OSWT
t'14 Leak detection system component problem
FW (Feedwater System
491 .System lailre
492 Feedwater Ipump failure
49)
Condenlsate pump failure
,.94 System valve faillre
95 System I&C failure
496 Other failure
LPCIS + RHR Failures (LPCI/RIIR/VCCS)
-21 Pump faIiare
422 System valve failure
423 System I&C failure
424 Other failure
425 RIHR heat exchanger problem
S/R (Safety and Relief Valves) o *DS
501 MS safety valve failure to open
502 MS safety valve failure to reseat
503 MS S/11 valve failure to open - mranual operation
I04 MS S/Ri valve failure to open - autonatic operation
B WR
CSIS
411
432
4)3
4 34
Failures (CS)
(.P) core spray pump lailure
System valve failire
System I&C failure
Other failutre
IIPCIS Failures
:ore injection pump failure
iii iP
442 System valve failure
443 System 14C failure
444 Other failure
Standby LL!iid Poison Injection Systenm
1 SLPII system lailure
452 SLPI tank failure
451 SLPI pump failure
4 54 Explosive plug valve failure
455 System I&C failIre
456 Other failure
RCIC
461
462
463
464
465
Failures , Isolation Condenser System Failures
RCIC pump faihlure
Isolation condenser failure
System valve failire
Systeih I&C failure
Other failure
VS (Vapor Suppression) 'ystem
481 Systemn ailiEre
482 Torus support cracks
483 Other ltotus problem
484t
Torus DP problem
485 Vacuum breaker prohl.e.n (di ywell to torus)
',36 Other VS p)ro,)lfem
4'7 Vacum rhl.l pr'ohlom (C(: to I xl ) inc'he prrmimk.iv- proble'sui
105
506
507
'MS S/R valve failure to reseat
M'S S/R valve other performance failure
AD)S actuation problems
Other
520 Component failure in dump to condenser Hotwell System
Coltainlwent Spra
ystemn (BWR-2s)
5)0 CSS System failure
5)1 CSS pump lailure
5)2 CSS valve failure
53)
CSS heat exchanger failure
531 CSS I*C failuare
s15 Other CS5 c:omponent failure
8...2
Glossary of _Abbreviations Used
ACRS
Advisory Committee on Reactor Safeguards
BWR
Boiling Water Reactor
CEA
French Atomic Energy Commission
(Commissariat A l'gnergie Atomique)
CF
Capacity Factor
EAF
Energy Availability Factor
ECCS
Emergency Core Cooling System
EdF
Electricit6 de France
EFPH
Equivalent Full Power Hours
FF
French Francs
FRG
Federal Republic of Germany
HSK
Swiss Federal Nuclear Safety Inspectorate
(Hauptabteilung fir die Sicherheit der
Kernanlagen)
IAEA
International Atomic Energy Agency
IEB
Inspection/Enforcement Bulletin (US)
IPSN
Protection and Nuclear Safety Institute
(Institut de Protection et de Suret6
Nucleaire)
KSU
KirnkraftSikerhet och Utbildning
(Nuclear Training and'Safety Center)
LCO
Limiting Condition of Operation
LWR
Light Water Reactor
MAPLHGR
Maximum Average Planar Linear
Heat Generation Rate
MCPR
Minimum Critical Power Ratio
MIT-EL
Massachusetts Institute of Technology Energy Laboratory
MP
Member of Parliament
MSIV
Main Steam Isolation Valve
MW
Megawatts
116
MWe
Megawatts (electric)
NRC
Nuclear Regulatory Commission (US)
OECD
Organisation for Economic Cooperation
and Development
OPEC-2
Operating Plant Evaluation Code - 2
PWR
Pressurized Water Reactor
RCC
Design and Construction Rules
RKS
Nuclear Safety Board of the Swedish Utilities
(RAdet f6r KdrnkraftS&kerhet)
SCSIN
Central Service for the Safety of Nuclear
Installations
(Service Central de Suret6
des Installations Nucl6aires)
SEK
Swedish Kronor
SF
Swiss Francs
SKI
Swedish Nuclear Power Inspectorate
(Statens K&rnkraftinspektion)
STB
Surveillance Test Book (Sweden)
TMI
Three Mile Island
US
United States
117
(France)