# ![](media/image1.png){width="2.7090277777777776in" height="0.8in"}

#  {#section-1 .Title1}

# Nuclear  {#nuclear .Title1}

#  Safety  {#safety .Title1}

#  Specialist  {#specialist .Title1}

#  {#section-2 .Title1}

#  {#section-3 .Title1}

# Qualification Standard Reference Guide {#qualification-standard-reference-guide .Title1}

[September 2005]{.smallcaps}

[PURPOSE 1](#purpose)

[SCOPE 1](#scope)

[1. Nuclear safety specialists shall demonstrate a working-level
knowledge of the fission process. 2](#__RefHeading___Toc158688630)

[2. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the various methods to reduce exposure.
3](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-various-methods-to-reduce-exposure.)

[3. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of criticality control, safety parameters, alarm systems, and
poisons.
4](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-criticality-control-safety-parameters-alarm-systems-and-poisons.)

[4. Nuclear safety specialists shall demonstrate a working-level
knowledge of terminology used in nuclear safety analysis.
8](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-terminology-used-in-nuclear-safety-analysis.)

[5. Nuclear safety specialists shall demonstrate a working-level
knowledge of the principle hazard and accident analysis methods.
16](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-the-principle-hazard-and-accident-analysis-methods.)

[6. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of terminology associated with probabilistic risk assessment
(PRA) techniques.
21](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-terminology-associated-with-probabilistic-risk-assessment-pra-techniques.)

[7. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of basic heating, ventilation, air conditioning system (HVAC),
and filtration system construction, operation, and application.
24](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-basic-heating-ventilation-air-conditioning-system-hvac-and-filtration-system-construction-operation-and-application.)

[8. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of process instrumentation principles of operation as applied
to nuclear safety-related systems.
27](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-process-instrumentation-principles-of-operation-as-applied-to-nuclear-safety-related-systems.)

[9. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of piping and instrumentation drawings (P&ID).
30](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-piping-and-instrumentation-drawings-pid.)

[10. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of electrical diagrams and schematics.
30](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-electrical-diagrams-and-schematics.)

[11. Nuclear safety specialists shall demonstrate a familiarity level
knowledge of electrical logic diagrams.
31](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-electrical-logic-diagrams.)

[12. Nuclear safety specialists shall demonstrate a working level
knowledge of radioactivity and transformation mechanisms.
32](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-radioactivity-and-transformation-mechanisms.)

[13. Nuclear safety specialists shall demonstrate a working level
knowledge of principles and concepts for internal and external dosimetry
and dose consequences.
33](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-principles-and-concepts-for-internal-and-external-dosimetry-and-dose-consequences.)

[14. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the biological effects of radiation.
34](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-biological-effects-of-radiation.)

[15. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the principles and use of radiological instrumentation and
radiological monitoring/survey practices.
35](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-principles-and-use-of-radiological-instrumentation-and-radiological-monitoringsurvey-practices.)

[16. Nuclear safety specialists shall demonstrate a working-level
knowledge of 10 CFR 830.204, Documented Safety Analysis, and DOE Guide
421.1-2, with respect to its impact on Department nuclear safety.
38](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-10-cfr-830.204-documented-safety-analysis-and-doe-guide-421.1-2-with-respect-to-its-impact-on-department-nuclear-safety.)

[17. Nuclear safety specialists shall demonstrate a working-level
knowledge of the safety basis requirements for environmental restoration
and decommissioning activities.
43](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-the-safety-basis-requirements-for-environmental-restoration-and-decommissioning-activities.)

[18. Nuclear safety specialists shall demonstrate a working-level
knowledge of 10 CFR 830.207, DOE Approval of Safety Basis and
DOE-STD-1104, with respect to its impact on department nuclear safety.
45](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-10-cfr-830.207-doe-approval-of-safety-basis-and-doe-std-1104-with-respect-to-its-impact-on-department-nuclear-safety.)

[19. Nuclear safety specialists shall demonstrate a working-level
knowledge of 10 CFR 830.206, Preliminary Documented Safety Analysis,
with respect to its impact on department nuclear safety.
47](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-10-cfr-830.206-preliminary-documented-safety-analysis-with-respect-to-its-impact-on-department-nuclear-safety.)

[20. Nuclear safety specialists shall demonstrate a working-level
knowledge of 10 CFR 830.202, Safety Basis and DOE-STD-1027, with respect
to its impact on department nuclear safety.
48](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-10-cfr-830.202-safety-basis-and-doe-std-1027-with-respect-to-its-impact-on-department-nuclear-safety.)

[21. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of Department of Energy (DOE) O 420.1A, Facility Safety, DOE G
420.1-1, Nonreactor Nuclear Safety Design Criteria and Explosive Safety
Criteria Guide, for use with DOE O 420.1, Facility Safety, and
DOE-STD-1020, Natural Phenomena Hazards Design and Evaluation Criteria
for Department of Energy Facilities, with respect to its impact on
Department nuclear safety.
50](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-department-of-energy-doe-o-420.1a-facility-safety-doe-g-420.1-1-nonreactor-nuclear-safety-design-criteria-and-explosive-safety-criteria-guide-for-use-with-doe-o-420.1-facility-safety-and-doe-std-1020-natural-phenomena-hazards-design-and-evaluation-criteria-for-department-of-energy-facilities-with-respect-to-its-impact-on-department-nuclear-safety.)

[22. Nuclear safety specialists shall demonstrate a working-level
knowledge of the Technical Safety Requirements as described in 10 CFR
830.205, Technical Safety Requirements and DOE Guide 423.1-1, with
respect to its impact on Department nuclear safety.
57](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-the-technical-safety-requirements-as-described-in-10-cfr-830.205-technical-safety-requirements-and-doe-guide-423.1-1-with-respect-to-its-impact-on-department-nuclear-safety.)

[23. Nuclear safety specialists shall demonstrate a working-level
knowledge of 10 CFR 830.203, Unreviewed Safety Question Process and DOE
Guide 423.1-1, with respect to its impact on Department nuclear safety.
64](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-10-cfr-830.203-unreviewed-safety-question-process-and-doe-guide-423.1-1-with-respect-to-its-impact-on-department-nuclear-safety.)

[24. Nuclear safety specialists shall demonstrate a familiarity level
knowledge of the functional interfaces between safety system software
components and the system level design.
69](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-functional-interfaces-between-safety-system-software-components-and-the-system-level-design.)

[25. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the relationships between the problems being addressed by
safety analysis and design codes, the design requirements for the codes,
and the components of the codes.
70](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-relationships-between-the-problems-being-addressed-by-safety-analysis-and-design-codes-the-design-requirements-for-the-codes-and-the-components-of-the-codes.)

[26. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of Department of Energy (DOE) Policy 450.4, Safety Management
System Policy, and Policy 450.5, Line Environment, Safety and Health
Oversight, as applied to nuclear safety.
70](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-department-of-energy-doe-policy-450.4-safety-management-system-policy-and-policy-450.5-line-environment-safety-and-health-oversight-as-applied-to-nuclear-safety.)

[27. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the following criticality safety-related American National
Standards Institute/American Nuclear Society (ANSI/ANS) standards:
72](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-following-criticality-safety-related-american-national-standards-instituteamerican-nuclear-society-ansians-standards)

[28. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the following Department of Energy (DOE) Orders, Technical
Standards, and Notice:
74](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-following-department-of-energy-doe-orders-technical-standards-and-notice)

[29. Nuclear safety specialists shall demonstrate a familiarity-level
knowledge of the Price-Anderson Amendments Act of 1988 and its
relationship to subparts A and B of 10 CFR 830.
78](#nuclear-safety-specialists-shall-demonstrate-a-familiarity-level-knowledge-of-the-price-anderson-amendments-act-of-1988-and-its-relationship-to-subparts-a-and-b-of-10-cfr-830.)

[30. Nuclear safety specialists shall demonstrate a working-level
knowledge of the requirements in Department of Energy (DOE) Technical
Standard DOE-STD-3009-94, Preparation Guide for U.S. Department of
Energy Nonreactor Nuclear Facility Safety Analysis Reports and
DOE-STD-3010, Airborne Release Fractions/Rates and Respirable Fractions
for Nonreactor Nuclear Facilities.
79](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-the-requirements-in-department-of-energy-doe-technical-standard-doe-std-3009-94-preparation-guide-for-u.s.-department-of-energy-nonreactor-nuclear-facility-safety-analysis-reports-and-doe-std-3010-airborne-release-fractionsrates-and-respirable-fractions-for-nonreactor-nuclear-facilities.)

[31. Nuclear safety specialists shall demonstrate a working-level
knowledge of NSTP-2003-1 that describes the relationship between human
factors/human performance and institutional programs that support the
safety analysis.
91](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-nstp-2003-1-that-describes-the-relationship-between-human-factorshuman-performance-and-institutional-programs-that-support-the-safety-analysis.)

[32. Nuclear safety specialists shall demonstrate a working-level
knowledge of assessment techniques (such as the planning and use of
observations, interviews, and document reviews) to assess facility
performance, report results of assessments, and follow up n actions
taken as the result of assessments.
93](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-assessment-techniques-such-as-the-planning-and-use-of-observations-interviews-and-document-reviews-to-assess-facility-performance-report-results-of-assessments-and-follow-up-on-actions-taken-as-the-result-of-assessments.)

[33. Nuclear safety specialists shall demonstrate a working-level
knowledge of the Department of Energy (DOE)/facility contract provisions
necessary to provide oversight of a contractor's operations.
97](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-the-department-of-energy-doefacility-contract-provisions-necessary-to-provide-oversight-of-a-contractors-operations.)

[34. Nuclear safety specialists shall demonstrate a working-level
knowledge of problem analysis principles and the techniques necessary to
identify problems, potential causes, and corrective action(s) associated
with nuclear safety issues at DOE Defense Nuclear Facilities.
99](#nuclear-safety-specialists-shall-demonstrate-a-working-level-knowledge-of-problem-analysis-principles-and-the-techniques-necessary-to-identify-problems-potential-causes-and-corrective-actions-associated-with-nuclear-safety-issues-at-doe-defense-nuclear-facilities.)

[Selected Bibliography and Suggested Reading
103](#selected-bibliography-and-suggested-reading)

# PURPOSE

The purpose of this reference guide is to provide a document that
contains the information required for a National Nuclear Security
Administration (NNSA) technical employee to successfully complete the
Nuclear Safety Specialist (NSS) Functional Area Qualification Standard.
In some cases, information essential to meeting the qualification
requirements is provided. Some competency statements require extensive
knowledge or skill development. Reproducing all the required information
for those statements in this document is not practical. In those
instances, references are included to guide the candidate to additional
resources.

# SCOPE

This reference guide has been developed to address the competency
statements in the October 2001 edition of DOE-STD-1183-2004, Nuclear
Safety Specialist Functional Area Qualification Standard. Competency
statements and supporting knowledge and/or skill statements from the
qualification standard are shown in contrasting bold type, while the
corresponding information associated with each statement is provided
below it. The qualification standard for the NSS contains 34 competency
statements. Some of the competency statements contain performance-based
requirements. These statements will not be covered in this reference
guide.

Every effort has been made to provide the most current information and
references available as of September 2005. However, the candidate is
advised to verify the applicability of the information provided.

Please direct your questions or comments related to this document to
Gary Gilliland, Training and Development Department at, 845-5689.

**TECHNICAL COMPETENCIES**

## Nuclear safety specialists shall demonstrate a working-level knowledge of the fission process.

a)  Define the following terms:

-   Excitation energy

-   Critical energy

-   Fissile material

-   Fissionable material

-   Fertile material

### Excitation energy

Excitation energy is the measure of how far the energy level of a
nucleus is above its ground state.

### Critical Energy

Critical energy is the minimum excitation energy required for fission to
occur.

### Fissile Material

Fissile material is composed of nuclides for which fission is possible
with neutrons of any energy level.

### Fissionable material

Fissionable material is composed of nuclides for which fission with
neutrons is possible.

### Fertile material

Fertile material is material that can undergo transmutation to become
fissile material.

b)  Describe the curve of binding energy per nucleon vs. mass number and
    give a qualitative description of the reasons for its shape.

As the number of particles in a nucleus increases, the total binding
energy also increases. The rate of increase, however, is not uniform.
This lack of uniformity results in a variation in the amount of binding
energy associated with each nucleon within the nucleus. This variation
in the binding energy per nucleon (BE/A) is easily seen when the average
BE/A is plotted versus atomic mass number. As the atomic mass number
increases, the BE/A decreases for A \> 60. The BE/A curve reaches a
maximum value of 8.79 MeV at A = 56 and decreases to about 7.6 MeV for A
= 238. The general shape of the BE/A curve can be explained using the
general properties of nuclear forces. The nucleus is held together by
very short-range attractive forces that exist between nucleons. On the
other hand, the nucleus is being forced apart by long range repulsive
electrostatic (coulomb) forces that exist between all the protons in the
nucleus. As the atomic number and the atomic mass number increase, the
repulsive electrostatic forces within the nucleus increase due to the
greater number of protons in the heavy elements. To overcome this
increased repulsion, the proportion of neutrons in the nucleus must
increase to maintain stability. This increase in the neutron-to-proton
ratio only partially compensates for the growing proton-proton repulsive
force in the heavier, naturally occurring elements. Because the
repulsive forces are increasing, less energy must be supplied, on the
average, to remove a nucleon from the nucleus. The BE/A has decreased.
The BE/A of a nucleus is an indication of its degree of stability.
Generally, the more stable nuclides have higher BE/A than the less
stable ones. The increase in the BE/A as the atomic mass number
decreases from 260 to 60 is the primary reason for the energy liberation
in the fission process. In addition, the increase in the BE/A as the
atomic mass number increases from 1 to 60 is the reason for the energy
liberation in the fusion process, which is the opposite reaction of
fission.

c)  Explain why only the heaviest nuclei are easily fissioned.

The heaviest nuclei require only a small distortion from a spherical
shape (small energy addition) for the relatively large coulomb forces
that are forcing the two halves of the nucleus apart to overcome the
attractive nuclear forces holding the two halves together. Consequently,
the heaviest nuclei are easily fissionable compared to lighter nuclei.

d)  Explain why uranium-235 fissions with thermal neutrons and
    uranium-238 fissions only with fast neutrons.

Uranium-235 fissions with thermal neutrons because the binding energy
released by the absorption of a neutron is greater than the critical
energy for fission. The binding energy released by uranium-238 absorbing
a neutron is less than the critical energy for fission, so additional
kinetic energy must be possessed by the neutron for fission to be
possible.

e)  Characterize the fission products in terms of mass groupings and
    radioactivity.

Fission products have some general characteristics in common. They
generally decay by emission. The most common mass numbers are grouped
near 95 and 140.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the various methods to reduce exposure.

a)  Describe aspects of dose reduction:

-   Time

-   Distance

-   Shielding

-   Inverse square law

-   ALARA

### Time

Minimizing time in an area where a radiation dose exists minimizes
exposure.

### Distance

Maximizing distance from a radiation source helps minimize exposure.

### Shielding

Maximizing shielding will reduce exposure from a radiation source.

### Inverse square law

Inverse square law applies to any point source that spreads its
influence equally in every direction and which doesn't have a limit to
its range. A point radiation source can be characterized in terms of
this law whether you are talking about
[roentgen](http://hyperphysics.phy-astr.gsu.edu/hbase/nuclear/radrisk.html#c4)s,
rads, or rem. All measures of exposure will drop off by inverse square
law, meaning the further the distance from the point source, the less
the intensity of exposure.

### ALARA (As Low As is Reasonably Achievable) 

ALARA is an approach in managing and controlling exposures and releases
of radioactive material to the environment so that the levels are as low
as is reasonable, taking into account social, technical, economic,
practical, and public policy considerations.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of criticality control, safety parameters, alarm systems, and poisons.

a)  Discuss the effects and applications of the following factors
    relevant to criticality safety of operations:

-   Mass

-   Shape

-   Interaction

-   Separation

-   Geometry

-   Moderation

-   Reflection

-   Concentration

-   Volume

-   Density

-   Neutron absorbers

-   Heterogeneity

-   Enrichment

### Mass

The minimum masses required to attain a critical state are 500 grams of
Pu and 800 grams of U.

### Shape

Limiting the size and configuration of vessels and equivalent thickness
restrictions on tanks has also been a common practice in criticality
control.

### Interaction

When more than one can, vessel, etc., containing fissile material is
present in an area, the index of criticality (the multiplication factor)
for the interacting system will be greater than if the units were
isolated. That is, a collection of units, such as a storage array, could
become critical even though the individual units would be far
sub-critical in isolation.

### Separation

Separation is the sorting of nuclear material based on atomic weight and
mass. This process can be performed to reclaim spent fuel or to separate
materials that are mined and collected in mining operations.

### Geometry

The configuration of fissionable material within its container is
referred to as its geometry. Containers are designed to minimize the
chance of material being able to configure itself into a geometry that
would allow for criticality.

### Moderation

Moderation is the slowing down of neutrons from their fast-birth
energies. The fission likelihood or cross-section is several hundred
times greater for slow neutrons than fast ones. Hydrogenous materials
such as water, oil, and cutting fluids are excellent moderators.

### Reflection

Reflection is the bouncing of neutrons back into a region containing
fissile material by collisions with atoms in surrounding materials. For
example, a thin layer of cutting fluid on a fissile part being machined
represents almost no reflection. Immersing that same part in a bucket of
water, such that there are a few inches of water surrounding it, could
reduce the amount of fissile material required for criticality by about
a factor of two.

### Concentration

In solutions, the concentration of fissile material has a large impact
on the critical mass and critical volume. For example, the minimum
critical masses of U and Pu are about 800 and 500 grams (g),
respectively, at solution concentrations of 30 to 40 g/liter. Below 5
g/liter, infinite volumes cannot be made critical.

### Volume

Container volumes of less than six liters are often used as an aid to
criticality control, particularly for solutions and loose powders which
could accidentally become flooded.

### Density

Density is a measure of the spacing between atoms. Obviously, the
greater the spacing, i.e., the lower the density, the greater neutron
leakage will be. However, for low density powders and cutting chips
which may have a critical mass in excess of 100 kilograms fissile when
dry, the critical mass may drop to about one kilogram during accidental
flooding scenarios.

### Neutron Absorbers

Neutron absorbers refer to other materials that play a dominant role in
absorbing neutrons but do not fission. Boron, cadmium, and gadolinium
are examples. Neutron poisons are most common in solution processing
such as in a boron-epoxy loaded stirrer rod in a precipitation vessel.

### Heterogeneity

Heterogeneity refers to an alloy or solution containing different
fissionable materials. Heterogeneous materials typically have a higher
critical mass than do homogenous materials.

### Enrichment

Enrichment refers to the fissile content of uranium, the most common
being 93 percent, U(93), in which 93 out of 100 atoms are of mass 235,
the remaining 7 being of mass 238. For plutonium, the convention is the
opposite; the enrichment is the non-fissile (Pu) content. Most plutonium
is about 94 percent Pu-239 and 6 percent Pu-240.

f)  Discuss the influence of the presence of non-fissionable materials
    mixed with, or in contact with, fissionable material on nuclear
    criticality safety.

Non--fissionable materials affect the ability of the fissionable
material to reach potential configurations that could lead to a
criticality. The greater the amount of non-fissionable material present,
the lesser the chance of a criticality for the fissionable material.

g)  Discuss the concept of contingencies for checking the validity of
    criticality safety limits.

Process designs shall incorporate sufficient factors of safety to
require at least two unlikely, independent, and concurrent changes in
process conditions before a criticality accident is possible. Protection
shall be provided by either (1) the control of two independent process
parameters, or (2) a system of multiple controls on a single process
parameter. The number of controls required upon a single controlled
process parameter shall be based upon control reliability and any
features that mitigate the consequences of control failure. In all
cases, no single credible event or failure shall result in the potential
for a criticality accident.

h)  Define the following terms:

-   Criticality accident

-   Minimum accident of concern

-   Process area

### Criticality Accident

The release of energy as the result of accidentally producing a
self-sustaining or divergent fission chain reaction is referred to as a
criticality accident.

### Minimum Accident of Concern

Minimum accident of concern refers to the smallest accident a
criticality alarm system is required to detect. A criticality accident
such as this would produce approximately 1015 fissions in its first
minute and provide an absorbed dose in free air of 20 rad at 2 meters if
no significant shielding were present.

### Process Area

A process area is any area involved in the chemical processing,
mechanical processing, handling, or storage of fissionable materials.

i)  Discuss the general principles associated with the use of
    criticality alarm systems including the following:

-   Installation

-   Coverage

-   Detection

-   Alarms

-   Dependability

### Installation

The installation of an alarm system implies a nontrivial risk of a
criticality accident and emergency plans in the event of alarm actuation
shall be maintained.

### Coverage

Where inventory of fissionable materials in individual, unrelated areas
exceeds 700g of U‑235, 520g of U-233, 450g of Pu-239, or 450g of any
combination of these three isotopes, the need for criticality alarm
system coverage shall be evaluated.

### Detection

Where criticality alarm coverage is required, a means shall be provided
to detect excessive amounts or intensities of radiation and to signal
personnel evacuation.

### Alarms

When implemented, criticality alarms shall be for immediate evacuation
purposes only and of sufficient volume and coverage to be adequately
heard in all areas that are to be evacuated. They shall be capable of
automatic actuation and set high enough to minimize actuation by sources
other than criticality, but low enough to detect the minimum accident of
concern. Upon actuation, the signal shall be of the lock-in type and
shall not reset even if the radiation levels fall below the alarm
setpoint. A corresponding evacuation alarm shall be signaled immediately
following detection of a criticality accident; in addition, the
evacuation alarm shall be capable of manual initiation.

### Dependability

Consideration shall be given to the detection of false alarms. To
achieve this, methods such as the use of concurrent response of two or
more detectors or reliable single channel detectors shall be used. A
means (that will not cause an evacuation) to test the response and
performance of the alarm system shall be provided. Upon completion of
testing, the system shall be immediately returned to normal operating
condition. Where operations continue in process areas during a loss of
normal power, emergency power supplies shall be provided for alarm
systems or they shall be monitored by portable alarms. Detectors shall
not fail to actuate when subject to a radiation field of at least 10
rad/second.

j)  Describe the use of neutron poisons.

Neutron poisons are non-fissionable atoms that are effective at
capturing or absorbing neutrons. Neutrons captured or absorbed by
poisons are no longer available to cause fission.

k)  Define the following terms:

-   Burnable poison

-   Non-burnable poison

-   Chemical shim

### Burnable poison

A burnable poison is a material that loses its neutron absorption
characteristics over time due to neutron absorption.

### Non-burnable poison

A non-burnable poison is a material that has relatively constant neutron
absorption characteristics over core life.

### Chemical shim

A chemical shim is a soluble neutron poison that is circulated in the
coolant during normal operation.

l)  Explain the purpose and use of Raschig Rings as a neutron poison.

Borosilicate-glass Raschig Rings contain a form of boron, a neutron
poison. By placing several of the small glass rings (3 cm in height, 3
cm outside diameter, and 0.5 cm thickness) in a container of soluble
fissile material, the borosilicate glass absorbs neutrons, but most
significantly, the surface area of the fissile solution increases,
increasing the probability of neutron leakage into the rings and
diminishing the likelihood of a criticality accident.

## Nuclear safety specialists shall demonstrate a working-level knowledge of terminology used in nuclear safety analysis.

a)  Define the following accident related terms:

-   Accident

-   Authorization agreement

-   Authorization basis

-   Beyond design basis accident

-   Design basis

-   Design basis accidents

-   Evaluation guidelines

-   Safety basis

-   Safety analysis

-   Consequence

-   Frequency

-   Risk

-   External event

-   Internal event

### Accident

An accident is an unplanned event or sequence of events that result in
undesirable consequences.

### Authorization Agreement

An authorization agreement is an agreement between an organization and
DOE on responsibilities for a facility or area.

### Authorization Basis

The authorization basis is comprised of those aspects of the facility
design basis and operational requirements relied upon by DOE to
authorize operation. The authorization basis is described in documents
such as the facility safety analysis report and other safety analysis
documentation.

### Beyond Design Basis Accident

A beyond design basis accident is an accident of the same type as a
design basis accident, but which is defined by parameters that exceed in
severity the parameters defined for the design basis accident. The same
correlation applies to beyond derivative design basis accidents with
regard to derivative design basis accidents.

### Design Basis

The design basis is the set of requirements that bound the design of
structures, systems, and components within a facility. Areas included
are seismic qualifications, fire protection, and safe shutdown.

### Design Basis Accident (DBA)

The design basis accident is a postulated accident that establishes
design and performance requirements for systems, structures, and
components that are important to safety.

### Evaluation Guidelines

The evaluation guidelines establish the hazardous material dose/exposure
values that the safety analysis is evaluated against. Theoretically,
individual doses/exposures exceeding the evaluation guideline should not
occur at a given point inside the evaluation area. Offsite evaluation
guidelines are established for the purpose of identifying and evaluating
safety-class structures, systems, and components.

### Safety Basis

The safety basis is comprised of the information relating to the control
of hazards at a nuclear facility upon which DOE depends for its
conclusion that activities at the facility can be conducted safely.

### Safety Analysis

A safety analysis is a process for systematically identifying hazards of
a DOE operation by analyzing and evaluating potential accidents and
their risks as well as the adequacy of hazard control measures.

### Consequence

A consequence is the direct effect of an event, incident, or accident.
It is expressed as a health effect (e.g., death, injury, or exposure),
property loss, environmental effect, evacuation, or quantity spilled.

### Frequency

Frequency is the ratio of the number of observations in a statistical
category to the total number of observations.

### Risk

Risk is the quantitative or qualitative expression of possible loss that
considers both the probability that a hazardous event will cause harm
and the consequences of that event.

### External Event

An external event is an event unconnected with the operation of a
facility or activity which could have an effect on the safety of the
facility or activity.

### Internal Event

An internal event is an event that is part of the operation of a
facility or activity which could have an effect on the safety of the
facility or activity.

m)  Define the following hazard related terms:

-   Hazard

-   Hazard categorization

-   Hazard category 1

-   Hazard category 2

-   Hazard category 3

-   Hazardous material

### Hazard

A hazard is the existence of a source of danger which could cause
illness, injury, or death to personnel or damage to a facility or the
environment.

### Hazard Categorization

DOE has established a process to classify facilities into hazard
categories. DOE-STD-1027 contains the flow charts which, when followed,
determine the hazard classification.

### Hazard Category 1

The classification of hazard category 1 is designated when a hazard
analysis shows the potential for significant offsite consequences. This
classification is primarily reserved for reactors capable of operating
at a steady state power level greater than 20 megawatts (Class A
reactor). A facility may also be classified as hazard category 1 if the
Program Secretarial Officer (PSO) deems it necessary. This designation
would be due to the level of off-site emergency planning required to
permit operation.

### Hazard Category 2

The classification of hazard category 2 is designated when a hazard
analysis shows the potential for significant onsite consequences. A
release of materials from a hazard category 2 facility would result in
exposures greater than that of 1 rem at 100 meters. Thus, emergency
planning requirements are very important. The potential for a nuclear
criticality accident at a facility will also earn it a hazard category 2
classification.

### Hazard Category 3

The classification of hazard category 3 is designated when a hazard
analysis shows the potential for only significant localized
consequences. Minimum threshold values for a hazard category 3 facility
would result in less than a 10 rem exposure at 30 meters over a 24-hour
period. The consequences of a localized release of materials would be in
excess of state and federal reporting requirements.

### Hazardous Material

A hazardous material is a substance or material that has been determined
by the Secretary of Transportation to be capable of posing an
unreasonable risk to health, safety, and property when transported in
commerce, and which has been so designated.

n)  Define the following safety control related terms:

-   Limiting conditions for operations (LCO)

-   Limiting control settings (LCS)

-   Safety limits (SL)

### Limiting Conditions for Operations (LCO)

LCOs are the limits that represent the lowest functional capability or
performance level of one or more safety-related items required for the
safe operation of a facility.

### Limiting Control Settings (LCS)

LCSs are the settings on process variables associated with safety
systems that control the facility function and will prevent exceeding
the associated safety limits.

### Safety Limits (SL)

SLs are limits on process variables associated with those passive
physical barriers that are necessary for the intended facility function
and that are required to guard against the uncontrolled release of
hazardous materials.

o)  Differentiate between the following categories of individuals who
    may be affected by an accident at a Department nuclear facility:

-   Off-site individual

-   On-site individual

-   Public

-   Worker

### Off-Site Individual

An off-site individual is any member of the public or any other
individual who may not be normally assigned to or located within the
subject site boundaries, and therefore would likely not be trained on
site specific hazards. The offsite area consists of the area beyond the
site boundaries.

### On-site Individual

An on-site individual is any member of the public or any other
individual who may be normally or temporarily located within the subject
site boundaries. This individual may not be trained on the presence of
site-specific hazards. The on-site area includes any area that has been
established as a national defense area or national security area.

### Public

Public means the surrounding population of individuals not normally
assigned to, or not having responsibilities in any way directly related
to, the construction, operation, maintenance, or administration of a
site or any facilities within a site. This population is generally (but
not always) located off-site (e.g., tours, roads, business, etc., may
require the public to travel through or enter into a site).

### Worker

A worker is an individual who by nature of employment has direct
responsibilities related to the construction, operation, maintenance, or
administration of any site or a facility within a site. This includes
those workers who may not have direct responsibilities for a given site
but may have responsibilities that result in being located within the
site boundary. Consequently, workers in general are normally associated
with being on-site, but may be either on-site or off-site.

p)  Differentiate between the function of structures, systems, and
    components in the following classifications:

-   Safety-class structures systems, and components (SSCs)

-   Safety-significant structures, systems and components (SSCs)

-   Defense in depth

### Safety-Class Structures, Systems, and Components (SSCs)

SSCs are designated as safety class if failure could adversely affect
the environment or safety and health of the public as identified by the
safety analysis.

### Safety-Significant Structures, Systems, and Components (SSCs)

SSCs that are not designated as safety-class, but whose preventive or
mitigative function is a major contributor to defense in depth (i.e.,
prevention of uncontrolled material releases) and/or worker safety as
determined from hazard analysis are considered safety significant. As a
general rule of thumb, safety-significant SSC designations based on
worker safety are limited to SSCs whose failure is estimated to result
in an acute worker fatality or serious injuries to workers.
Considerations should be based on engineering judgment of possible
effects and the potential added value of safety-significant SSC
designation.

### Defense in Depth

Defense in depth classification consists of two components:

-   Equipment and administrative features that provide preventive or
    mitigative functions so that multiple features are relied on for
    prevention or mitigation to a degree proportional to the hazard
    potential

-   Integrated safety management programs that control and discipline
    operations

q)  Differentiate between the function and contents of the following
    documents:

-   Documented safety analysis

-   Preliminary documented safety analysis

-   Safety analysis report

-   Basis for interim operations

-   Technical safety requirements

-   Preliminary hazards analysis

### Documented Safety Analysis (DSA)

A DSA reviews the extent to which a nuclear facility can be operated
with respect to workers, the public, and the environment, including a
description of the conditions, safe boundaries, and hazard controls that
provide the basis for ensuring safety.

### Preliminary Documented Safety Analysis (PDSA)

A PDSA is a document that is prepared in connection with the design and
construction of a new DOE nuclear facility or a major modification to a
DOE nuclear facility that provides a reasonable basis for the conclusion
that the nuclear facility can be operated.

### Safety Analysis Report (SAR)

A SAR documents that a nuclear facility can be constructed, operated,
maintained, shut down, and decommissioned safely and in compliance with
applicable laws and regulations.

### Basis for Interim Operations (BIO)

The BIO is the documented establishment of allowable operations for a
current facility and its operational controls until a fully compliant
document is developed and approved by DOE.

### Technical Safety Requirements (TSRs)

TSRs are comprised of the limits, controls, and related actions that
establish the specific parameters and requisite actions for the safe
operation of a nuclear facility.

### Preliminary Hazards Analysis (PHA)

A PHA represents a level of analytical detail appropriate for
preparation of safety documents for facilities.

r)  Differentiate between the controls which have the following
    designations:

-   Mitigating controls

-   Preventive controls

-   Administrative controls

-   Design features

-   Passive controls

-   Active controls

-   Safety SSCs

### Mitigating Controls

Mitigating controls are SSCs that serve to mitigate the consequences of
a release of radioactive or other hazardous materials in an accident
scenario.

### Preventive Controls

Preventive controls are any SSCs that serve to prevent the release of
radioactive or other hazardous materials in an accident scenario.

### Administrative Controls (ACs)

ACs are the provisions relating to organization and management,
procedures, recordkeeping, assessment, and reporting necessary to ensure
the safe operation of a facility.

### Design Features

The design features of a nuclear facility are those features specified
in the TSRs that would have a significant effect on safe operation if
altered or modified.

### Passive controls

Passive controls are nuclear criticality safety control methods that do
not require human intervention or electrical or mechanical reaction to
off specification conditions. These means of control take advantage of
natural forces, such as gravity, physical chemistry limitations, and
inherent physical characteristics, such as rigidity and structural
integrity of cylindrical geometries, and limited compressibility of
solids. These means of control include devices to prevent unsafe
accumulations of fissionable material within a unit, such as siphon
breaks, filters, and pipe blanks between process vessels, and spacing
devices such as birdcages, racks, and stanchions between containers, as
well as fixed neutron poisons, either within vessels (such as raschig
rings) or between containers.

### Active controls

Active controls are the active means for ensuring nuclear criticality
safety control methods. These means of control include active
electrical, mechanical, and hydraulic hardware that sense a process
variable important to nuclear criticality safety and provide automatic
action to secure the system to a safe condition without requiring human
intervention.

### Safety SSCs

Safety SSCs are both safety-class SSCs and safety-significant SSCs.

s)  Differentiate between the following types of facilities:

-   Nuclear facility

-   Non-reactor nuclear facility

-   Radiological facilities

### Nuclear Facility

A nuclear facility is a facility that conducts activities or operations
that involve radioactive and/or fissionable materials in such form and
quantity that a nuclear hazard potentially exists to the employees or
the general public. Included are reactors and accelerators.

### Non-Reactor Nuclear Facility

Non-reactor nuclear facilities are nuclear facilities excluding reactors
and accelerators.

### Radiological facilities

Radiological facilities are facilities that do not meet or exceed
category 3 threshold criteria but still possess some amount of
radioactive material.

t)  Differentiate between the following chemical terms:

-   Temporary Emergency Exposure Limit-1

-   Temporary Emergency Exposure Limit-2

-   Temporary Emergency Exposure Limit-3

-   Emergency Response Planning Guide-1

-   Emergency Response Planning Guide-2

-   Emergency Response Planning Guide-3

### Temporary Emergency Exposure Limit (TEEL)-1

The maximum concentration in air below which it is believed nearly all
individuals could be exposed without experiencing other than mild
transient adverse health effects or perceiving a clearly defined
objectionable odor.

### Temporary Emergency Exposure Limit (TEEL)-2

The maximum concentration in air below which it is believed nearly all
individuals could be exposed without experiencing or developing
irreversible or other serious health effects or symptoms that could
impair their abilities to take protective action.

### Temporary Emergency Exposure Limit (TEEL)-3

The maximum concentration in air below which it is believed nearly all
individuals could be exposed without experiencing or developing
life-threatening health effects.

### Emergency Response Planning Guide (ERPG)-1

The maximum concentration in air below which it is believed nearly all
individuals could be exposed for up to one hour without experiencing
other than mild transient adverse health effects or perceiving a clearly
defined objectionable odor.

### Emergency Response Planning Guide (ERPG)-2

The maximum concentration in air below which it is believed nearly all
individuals could be exposed for up to one hour without experiencing or
developing irreversible or other serious health effects or symptoms that
could impair their abilities to take protective action.

### Emergency Response Planning Guide (ERPG)-3

The maximum concentration in air below which it is believed nearly all
individuals could be exposed for up to one hour without experiencing or
developing life-threatening health effects.

u)  Identify the types of chemical or toxicological hazards that may be
    found in nuclear facilities.

Any chemical or toxicological hazard that is included in the DSA may be
present.

## Nuclear safety specialists shall demonstrate a working-level knowledge of the principle hazard and accident analysis methods.

a)  Identify and discuss the use of different methods for qualitative
    hazard analysis. Identify specific strengths and weaknesses with the
    various methods.

The two primary methods of performing a qualitative hazard analysis are
process hazard analysis (PHA) and failure modes and effects analysis
(FMEA). The basic outputs, however, remain unchanged. The PHA approach
uses a check list that identifies all hazards and then determines all
possible incidents that may happen. The FMEA approach identifies
equipment and then proceeds to determine all possible failures or errors
that may occur. PHA is an easier system to use for someone who does not
have a full understanding of all operations within a facility, such as
an outside consultant. The weakness of the PHA approach is in its
generalities. FMEA is a more thorough approach but requires a much
greater knowledge of operations of that facility/organization. Hazard
evaluation presents potential accidents in terms of hazards, energy
sources, causes, preventive and mitigative features, consequence
estimates, and frequency estimates.

v)  Discuss the methods used to identify and categorize the hazards
    associated with Department nuclear safety analysis.

The results of the hazard analysis provide a comprehensive evaluation of
the complete DSA accident spectrum. This evaluation will be essentially
qualitative in that its aim is to produce a well-reasoned and clear
assessment of facility hazards and their associated controls. The focus
of hazard analysis is on the completeness of consideration given to the
accident spectrum, as opposed to a formalized definition of accident
sequences and assumptions. Summary discussion of methodology is
appropriate, but detailed bases for judgment and any simple mathematical
estimates used in the hazard analysis to guide the judgments of the
analysis for specific accident scenarios are not required to be formally
documented in the DSA.

The results of the final hazard categorization activity are specified in
DOE-STD-1027. This standard includes the facility hazard categorization
and, where segmentation has been employed, the segment boundaries and
individual segment classifications.

w)  Identify and discuss the methods used to determine and analyze
    failure modes of SSCs, ACs, and control programs.

Summary methods germane to safety typically include

-   a general overview of safety in terms of SSCs and administrative
    features

-   identification of any safety-significant SSCs

-   identification of any safety management programs that will be
    assigned TSR coverage in the form of ACs for adequate safety

A general prioritization of the features needs to be included and
expressed in terms of the magnitude of a process hazard. A summary-level
discussion is required, not a detailed discussion or defense of the
prioritization logic. The discussion should exam the raw information in
the hazard evaluation tables and distill it into a clear overview of
safety features at the facility. The use of relevant accident examples
can frame and focus the discussion. Identify structures, systems, and
components as safety-significant SSCs where appropriate. As a general
rule of thumb, safety-significant SSC designations based on safety are
limited to those systems, structures, or components whose failure is
estimated to result in a prompt worker fatality, serious injuries to
workers, or significant radiological or chemical exposures to workers.
Categorize administrative features in terms of the programmatic
elements. Provide documented evidence that safety features are an
integral part of facility design and operation, that basic facility
operations to ensure safety are adequate, and that workers are protected
by a number of means including programs described as ACs.

x)  Identify and discuss methods available to reviewers to determine if
    a hazard analysis has omitted important accident vulnerabilities.

The developments of DBAs along with identification of any other possible
accidents are reviewed to determine if all hazards that may cause an
accident are addressed. In DOE-STD-3009, a flow chart shows that an
iterative process is used.

y)  Identify and discuss the relationship between hazard analysis and
    the postulation of accidents for quantitative consequence analysis
    in Documented Safety Analyses for DOE nuclear facilities. Describe
    what factors govern the choice of an accident warranting further
    consequence analysis.

The hazard analysis identifies DBAs that will be used in the accident
analysis. These accidents represent a complete set of bounding
conditions. The identification of DBAs results from the hazard
evaluation ranking of the complete spectrum of facility accidents. The
approach used at any specific facility is based on the detail needed for
a given facility and the experience of the analysts. This involves
ranking risk for accidents by consequence and frequency. The ranking
schemes are designed to separate the lower risk accidents that are
adequately assessed by hazard evaluation from higher risk accidents that
may warrant additional quantitative analysis if the phenomena involved
are not simplistic. Rankings use a bin system. For example, frequency
bins should typically cover two orders of magnitude. Although the
exercise of binning is essentially qualitative, analysts often use a
simple numerical basis for judgments to provide consistency. Another
methodology would be to use a summary of historical data. Likewise,
before beginning the evaluation, a conservative Gaussian plume
estimation of the amount of material needed outside the building to
cause a certain dose might be performed to aid in defining thresholds of
significance. However the ranking of frequency and consequence into such
broad categories is more of a qualitative than a quantitative exercise.
An important factor in estimating binning thresholds for consequences is
to tie the thresholds to the evaluation guideline so that accidents that
could challenge the guideline are correctly identified for formal
accident analysis. The binning requirement of this subsection does not
preclude the use of other sorting mechanisms in addition to risk sorting
if an analyst finds such mechanisms useful. This accident selection
activity identifies the process and criteria used to select the unique
and representative potential accidents to be included in accident
analysis. Unique accidents are those with sufficiently high-risk
estimates that individual examination is needed. Representative
accidents bound a number of similar accidents of lesser risk.
Representative accidents are examined to the extent they are not bounded
by unique accidents. In any case, at least one bounding accident from
each of the major types determined from the hazard analysis should be
selected unless the bounding consequences are "low." Accidents are
identified and listed by accident category and type.

z)  Identify and discuss essential elements of deterministic and
    probabilistic risk assessment techniques.

### Probabilistic Risk Assessment (PRA)

PRA is the use of probabilistic-based accident and systems modeling
techniques to quantitatively analyze, study, define, and prioritize
credible accident scenarios. The process is normally broken into three
levels:

-   Level I, estimation of accident frequencies

-   Level II, transport of radionuclides into the environment

-   Level III, estimation of effects of radionuclides in the environment
    on the surrounding population

In general the Level I process consists of the following elements:

-   Accident identification and screening to identify credible accidents
    and eliminate incredible accidents. Credible events are normally
    associated with "initiators," or events which could lead to the
    onset of an accident. Therefore, one of the products of this effort
    would be the identification of a set of initiators.

-   Development of accident progression event tree "top events" where
    success or failure of defense-in-depth SSCs or barriers, operator
    actions crucial to SSC or barrier integrity, and phenomenological
    events are modeled to define accident "end states." Trees must be
    developed for each initiator or for groups of similar initiators
    identified above.

-   Quantification of event tree "top events." This process often
    involves the development of detailed fault trees to determine top
    event failure probability. Fault trees generally reduce SSC success
    criteria to combinations of individual sub-component failures
    through the use of Boolean algebra.

-   Quantification of initiating events where the frequency of
    occurrence is determined for each initiator event of interest.

-   Accident sequence event quantification, which is the process of
    determining the sum total failure frequency for all credible
    accident sequences by combining the initiator frequency with
    subsequent failures of defense-in-depth SSCs, barriers, operator
    actions or phenomenological events as defined in the accident
    progression event tree.

-   Analysis and documentation of results.

### Deterministic Risk Assessment

Deterministic risk assessment is the use of computer-based integrated
modeling numerical methods for simultaneous solution of multiple
equations to study accident progression and related response phenomenon.
This process is often used to quantify complex phenomenological events
such as transport of radionuclides upon release.

a)  Given an accident source term of radionuclide/hazardous chemical
    release, discuss the factors that should be considered in selection
    of an appropriate computer code for off-site transport and
    deposition.

### Source Term Methods

Source term estimation is the process of determining the amount and the
types of radioactive isotopes released during a nuclear accident. This
is generally part of the front-end work performed in support of a Level
II PRA, where to estimate a source term, the environmental conditions
surrounding and evolving during an accident are studied using computer
codes and laboratory experiments. Using this information, estimation of
release fractions of nuclides (isotopes) are made based upon the
estimated or known source isotopic content. Some of the factors that
influence isotope release include

-   temperature

-   presence of a fire

-   thermal-hydraulics

-   explosions

-   breach of mechanical barriers to release such as cladding or sealed
    containers

### Environmental Transport Methods

The environmental transport method estimates the ways in which those
release fractions determined previously are transported from the source
to the environment and is typical of a Level III analysis. Of specific
interest are the multiple paths that releases may take into the
environment, namely, airborne, surface waterborne, and subterranean
hydrology paths. For most radiological accidents, the airborne path is
the most significant and is therefore the basis for the majority of
transport studies/models. The main factors significantly affecting
transport include:

-   the chemical state of the radionuclide (solid, liquid, or gas);

-   the release path (airborne or waterborne);

-   the initial energy of release (explosive, fire, steam effects);

-   manmade barriers/mitigative measures (confinements, buildings,
    filters, spray removal systems, etc.);

-   dispersion or plume effects;

-   weather effects including prevailing winds, rain, snow, and
    inversions.

Also, software approved by DOE or software that can be verified to be
correct and accurate in accordance with DOE software QA procedures may
be used to determine off-site transportation and deposition models.

b)  Discuss the physics of fires and explosions as the means of
    generating airborne plumes of hazardous materials and damaging
    barriers to releases. Also describe how the physics affects the
    quantities or rates at which hazardous materials may become airborne
    as a result of spills, evaporation, entrainment, fires and other
    accidents.

Appendix-A of DOE-STD-3009-94 gives definitions to be used for the
physics of an accident.

c)  Discuss the phenomena and modeling of airborne dispersion of toxic
    materials addressing weather effects, turbulent mixing, mixing
    heights, plume temperature, evolution and potential settling or
    plate out of particulates and aerosols, precipitation, building
    wake, and surface roughness effects.

The phenomena and modeling of airborne dispersion is addressed in
Appendix-A of DOE-STD-3009. The application of the different inputs is
discussed in DOE-STD-3009. The standard discusses the use of the most
conservative approach, which is the application of the worst possible
conditions to maximize the effects of a plume.

d)  Discuss the mechanisms involved in the damage caused by extreme
    natural phenomena including: hurricanes, tornadoes, ice storms,
    wind, flood, earthquakes and wild fires.

The mechanisms involved in the damage caused by external forces include
the acceleration of a facility or system from events like earthquakes;
mechanical damages resulting from tornados, hurricanes, ice storms, and
wind; erosion caused by flooding or hurricanes; and thermal damage
stemming from fires or ice storms.

e)  Define and discuss the following terms:

-   Chi/Q

-   Dose conversion factor

-   Breathing rate

-   Aerodynamic equivalent diameter

-   Solubility class

-   Population dose

### Chi/Q

Chi/Q is a sampling distribution methodology.

### Dose Conversion Factor

The dose conversion factor is used to convert radionuclide type to
equivalent exposure values.

### Breathing Rate

The breathing rate is the rate of respiration based on physical activity
being performed.

### Aerodynamic Equivalent Diameter

The aerodynamic equivalent diameter is the diameter of a sphere of
density 1 g/cm^3^ that exhibits the same terminal velocity as the
particle in question.

### Solubility Class

Solubility class relates to the ability for a given chemical or nuclide
to be suspended into solution.

### Population Dose

If all the individuals in the population of interest are exposed to the
same concentrations of a substance, then the estimates of individual
dose represent the average dose rate for members of the population and
the population dose is the average dose times the number of persons in
the population.

f)  Given a source term, determine dose consequences applying Chi/Q,
    dose conversion factor, breathing rate and specific activity as
    applicable.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

g)  Given a simple accident scenario, demonstrate knowledge by
    constructing a simple neutral gas dispersion and heavy gas
    dispersion. Estimate consequences using an accident modeling code
    including hand calculations and explain the assumptions, inputs and
    results.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

h)  Discuss the process for evaluating assumptions made for scenarios
    being modeled.

All assumptions made in the accident analysis are to be validated as
part of the accident analysis activity. For example, if an operator is
supposed to push button Z to stop an accident progression, the accident
analysis needs to make it clear that the operator can actually do so.
Making it clear may simply involve noting there are no physical
phenomena associated with the accident that would preclude the operator
from doing so. Likewise, basic assurance must be provided that equipment
relied upon in unusual or severe environments will function. This
assurance does not constitute the need for or expectation of full,
formal environmental qualification. The above guidance is not meant to
imply that the DSA must contain detailed validations for all
assumptions. The DSA needs to present information at a level that is
considered sufficient for review and approval of the DSA. Referencing an
auditable trail of information as part of the controlled supporting
documentation is acceptable.

i)  Discuss the methods used in the calculation of criticality
    accidents.

Various calculational methods are used depending on the complexity of
the problem being evaluated. The methodology and approach to determining
criticality accidents require the ability to perform some or all of the
following tasks:

-   Identify and discuss the application of several common hand
    calculation methods.

-   Select a single hand-calculation technique (buckling method, solid
    angle, or area density) and prepare an example of its use.

-   Develop an input model for one of the criticality safety codes
    (i.e., MONK, VIM, KENO/SCALE, MCNP, DANTSYS, ANISN, and COG).

-   Describe how cross section data impact Monte Carlo and deterministic
    codes.

-   Describe the importance of validation of computer codes and how it
    is accomplished.

-   Describe the methodology supporting Monte Carlo codes and
    deterministic codes.

-   Describe pitfalls of Monte Carlo calculations.

-   Discuss the strengths and weaknesses of Monte Carlo and Discrete
    Ordinants codes.

-   The diffusion theory model is not strictly valid for treating
    fissile systems in which neutron absorption, voids, and/or material
    boundaries are present. In the context of these limitations,
    identify a fissile system for which a diffusion theory solution
    would be adequate.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of terminology associated with probabilistic risk assessment (PRA) techniques.

a)  Identify the strengths and weaknesses of probabilistic risk
    assessment for safety design and regulatory decision-making.

Probabilistic risk assessment has limitations. It may sometimes exclude
or not adequately quantify potentially important risk factors, such as
very-low-frequency accident initiators, various failures derived from a
common event, physical processes resulting from several low-frequency
failures, or long-term health effects from potentially toxic materials.
Furthermore, because a risk assessment often deals with low-frequency
but high-consequence accident risks, there is considerable potential for
its results to be misunderstood.

The most useful aspects of the probabilistic risk assessment are not
exclusively the risk numbers that are generated; they are also the
insight gained by a systematic and methodical consideration of what can
go wrong with a system. A procedural analysis helps us to understand the
likely vulnerabilities of the system, the threats they pose, and the
measures that could be applied to mitigate or prevent them.
![](media/image2.png){width="8.333333333333333e-3in"
height="8.333333333333333e-3in"}Risk assessment is a particularly
powerful tool when there is only a limited set of alternatives for risk
evaluation. Thus, meaningful insights can be obtained by a risk
assessment without depending on the accuracy of an "actual risk" value;
such values are notoriously difficult to ascertain.

j)  Define the following terms with respect to reliability engineering
    and probabilistic risk assessments:

-   Probability

-   Reliability

-   Availability

-   Unavailability

-   Risk

-   Safety

-   Accident sequence

-   Dominant contributors

-   Minimal cut set

### Probability

Probability is represented as a number in a scale from 0 to 1 which
expresses the likelihood that an event will occur.

### Reliability

Reliability is the probability that an item is able to perform a
required function under stated conditions for a stated period of time.

### Availability

Availability is the fraction of the total time that a device or system
is able to perform its required function.

### Unavailability

Unavailability is the probability of the system being in a failed state
at any given time.

### Risk

Risk is a combined effect of the probability of the occurrence of an
undesired event and the magnitude of the event.

### Safety

Safety is the freedom from unacceptable risk or personal harm.

### Accident Sequence

An accident sequence is a specific unplanned sequence of events that has
a specific undesirable consequence.

### Dominant Contributors

Dominant contributors are inputs that have the greatest effect on a
given set of inputs.

### Minimal Cut Set

A minimal cut set is a cut set such that if any term is removed from the
list the system will no longer fail.

k)  Define the following terms and differentiate between the associated
    processes:

-   Event tree

-   Fault tree

-   Failure modes and effects analysis (FMEA)

### Event Tree

An event tree is a method for illustrating the sequence of outcomes that
may arise after the occurrence of a selected initial event.

### Fault Tree

A fault tree is a method of representing the logical combinations of
various system states which lead to a particular failure event.

### Failure Modes and Effects Analysis (FMEA)

FMEA is the study of the potential failures that might occur in any part
of a system to determine the probable effect of each on all other parts
of the system and on probable operational success.

l)  Discuss how probabilistic risk assessment methods can help in
    understanding accident scenarios.

Acceptance criteria have been inextricably linked to accident scenarios
that are prescribed in some manner. The results of quantitative
probabilistic risk assessments, principally those of nuclear power or
production facilities, are sometimes compared to another type of risk
acceptance criteria referred to as safety goals. PRAs are fundamentally
different analytical methods from deterministic accident scenarios and
produce a different type of product. For example, in PRAs the failure of
a safety feature to perform an intended function is always postulated,
irrespective of the safety classification of the feature. A conceptually
different approach is needed for safety analysis of existing facilities,
where an analysis of the safety of the "facility as is" is performed.
The primary objective of the analytical process must then turn to the
identification of needed controls and their potential inadequacies, and
the corresponding corrective or compensatory measures. For these
existing facilities, safety assurance is provided through an aggressive
approach based on a comprehensive analysis of all hazards leading to the
release of radiological or toxicological material, and ensuring that the
controls identified against each hazard are relevant, specific, and
effective.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of basic heating, ventilation, air conditioning system (HVAC), and filtration system construction, operation, and application.

a)  Given engineering diagrams of a heating, ventilation, air
    conditioning system, identify the following components and discuss
    their purposes:

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(a) **Blowers**

(b) **Fans**

(c) **Dampers**

(d) **Chillers**

(e) **Filters**

(f) **HEPA filters**

(g) **Heat exchangers**

(h) **Scrubbers**

(i) **Hoods**

(j) **Glove boxes**

(k) **Flow, pressure, temperature, current, level, voltage and position
    indicators, recorders and controllers**

> ![](media/image3.wmf) ![](media/image4.wmf) ![](media/image5.wmf)
> ![](media/image6.wmf) ![](media/image7.wmf) ![](media/image8.wmf)
>
> Blower and fans circulate air through ventilation system.
>
> Dampers direct and restrict flow through a ventilation system.
>
> Chillers remove heat from air allowing it to enter a space at a lower
> temperature.
>
> Filters are used to remove any particulates in the air; HEPA filters
> work with an efficiency level greater than 99 percent.
>
> ![](media/image9.wmf) ![](media/image10.wmf) ![](media/image11.wmf)
> ![](media/image12.wmf)
>
> Heat exchangers add or remove heat from air to allow air to condition
> space.
>
> Scrubbers are used to remove particulates in air systems in a
> localized area.
>
> Hoods are used to contain contaminates within a given boundary by
> using a high velocity of air to flow and remove any airborne hazards
> prior to leaving hood space.
>
> The glovebox isolates contaminates from the space where they are
> located.
>
> ![](media/image13.wmf)
>
> Indicators are used to give a positive indication of system parameters
> and to allow for warnings of unstable conditions.

m)  Discuss the relationships between the following in heating,
    ventilation, and air conditioning systems:

-   Supply ventilation

-   Flow

-   Exhaust ventilation

### Supply Ventilation

The supply ventilation is the air delivered to the conditioned space and
used for ventilation, heating, cooling, humidification, or
dehumidification.

### Flow

The flow is the path in a ventilation system through which air travels
and includes the route for the supply air and return air and the path
through the conditioned space.

### Exhaust Ventilation

The exhaust ventilation is the air removed from the conditioned space
that was used for ventilation, heating, cooling, humidification, or
dehumidification.

n)  Describe the purpose of the heating, ventilation, and air
    conditioning system in the following applications:

-   Hoods

-   Glove boxes

-   Hot cells

-   Confinement systems

-   HEPA filtration

### Hoods

Hoods maintain a negative pressure area for storage of toxic chemicals.

### Glove boxes

Glove boxes contain hazardous material in an enclosed environment which
prevents the material from spreading to areas that may affect workers or
the public.

### Hot Cells

Hot wells are areas used to remotely handle highly radioactive or
contaminated equipment. This type of facility allows for all operations
to be performed at a safe distance with adequate protection for workers.

### Confinement Systems

Confinement systems identify and describe the set of SSCs that perform
confinement functions such as process vessels, glove boxes, ventilation
systems, and facility walls.

### HEPA Filtration

HEPA filtration is the use of high-efficiency particulate air filtration
with an efficiency of 99.97 percent.

o)  Discuss the reason for and safety significance of the following
    system parameters:

-   Positive vs. negative system pressure

-   Differential pressure across filters

-   Differential pressure across components

### Positive vs. Negative System Pressure

Negative pressure systems minimize the spread of contamination or
contaminants in the event of failure of a system, while a positive
pressure system would help increase the spread of contamination or
contaminants in the event of a failure.

### Differential Pressure across Filters

As differential pressure across a filter increases, it indicates an
increase in the amount of material contained in the filters. For
radioactive systems, this may lead to increased radiation levels above
what is prescribed in the DSA.

### Differential Pressure across Components

As differential pressure across a component increases, it indicates an
increase in the amount of material settling out in that component. For
radioactive systems, this may lead to increased radiation levels above
what is prescribed in the DSA.

p)  Discuss the potential hazards and failure modes (to equipment and
    personnel) associated with the use of heating, ventilation, and air
    conditioning systems and components within nuclear safety-related
    systems.

During the design of the heating, ventilation, and air conditioning
system, a careful analysis of the failure modes of each component and
the effect of a failure on the operation of the system should be
undertaken. One of the more popular analysis methods is called a fault
tree analysis. This method may be used on entire systems, subsystems,
and individual components. For example, if attempting to do a fault tree
analysis on a room pressure control system, choose the component you
wish to analyze, such as a through-the-wall velocity sensor. List this
at the top of the tree. Next consider all of its failure modes and list
them underneath. Next, for the most serious failure modes, consider all
the affects on controlled components and other coupled and decoupled
systems and list them underneath that mode. Repeat this process until
all the paths are complete. It may be necessary to do all the branches
of the tree to discover which paths represent the worst scenarios. When
all the failure modes of all the components, subsystems, and systems
have been completed, you can make some strategic design modifications to
eliminate or ameliorate the most serious scenarios by installing more
reliable components in key locations or installing redundant controls or
systems to provide backup and thus truncate the tree.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of process instrumentation principles of operation as applied to nuclear safety-related systems.

a)  Explain the process-related reason for measuring temperature,
    pressure, flow, and fluid level.

Temperature, pressure, flow, and fluid levels are monitored in processes
to provide indication of potential failures. Monitoring provides an
indication of failure prior to having equipment damaged or destroyed or
possibly releasing material to the environment. The long-term tracking
of operations allows for trend analysis of potential equipment failures.

q)  For the temperature detection devices listed, explain how the
    instrument provides an output representative of the temperature
    being measured:

-   Thermocouple (TC)

-   Resistance temperature detector (RTD)

### Thermocouple

Thermocouples will cause an electric current to flow in the attached
circuit when subjected to changes in temperature. The amount of current
that will be produced is dependent on the temperature difference between
the measurement and reference junction, the characteristics of the two
metals used, and the characteristics of the attached circuit. Heating
the measuring junction of the thermocouple produces a voltage which is
greater than the voltage across the reference junction. The difference
between the two voltages is proportional to the difference in
temperature and can be measured on the voltmeter (in millivolts). For
ease of operator use, some voltmeters are set up to automatically
display temperature through use of electronic circuitry. Other
applications provide only the millivolt readout. To convert the
millivolt reading to its corresponding temperature, the applicable
conversion chart supplied by the manufacturer needs to be used.

### Resistance Temperature Detector (RTD)

The RTD incorporates pure metals or certain alloys that increase in
resistance as temperature increases and, conversely, decrease in
resistance as temperature decreases. RTDs act somewhat like an
electrical transducer, converting changes in temperature to voltage
signals by the measurement of resistance. The metals that are best
suited for use as RTD sensors are pure, of uniform quality, stable
within a given range of temperature, and able to give reproducible
resistance-temperature readings. Only a few metals have the properties
necessary for use in RTD elements. RTD elements are normally constructed
of platinum, copper, or nickel. These metals are best suited for RTD
applications because of their linear resistance-temperature
characteristics, their high coefficient of resistance, and their ability
to withstand repeated temperature cycles. The coefficient of resistance
is the change in resistance per degree change in temperature, usually
expressed as a percentage per degree of temperature. The material used
must be capable of being drawn into fine wire so that the element can be
easily constructed.

RTD elements are usually long, spring-like wires surrounded by an
insulator and enclosed in a sheath of metal. This particular design has
a platinum element that is surrounded by a porcelain insulator. The
insulator prevents a short circuit between the wire and the metal
sheath. Inconel, a nickel-iron-chromium alloy, is normally used in
manufacturing the RTD sheath because of its inherent corrosion
resistance. When placed in a liquid or gas medium, the Inconel sheath
quickly reaches the temperature of the medium. The change in temperature
will cause the platinum wire to heat or cool, resulting in a
proportional change in resistance. This change in resistance is then
measured by a precision resistance measuring device that is calibrated
to give the proper temperature reading.

r)  For the pressure detection devices listed, explain how the
    instrument provides an output representative of the pressure being
    measured:

-   Magnehelic differential pressure device

-   Photohelic differential pressure device

### Magnehelic Differential Pressure Device

A magnehelic differential pressure device transmits the effects of
changes in air pressure from a diaphragm to an indicating pointer by
means of magnetic linkage and without the use of gears or other direct
mechanical linkages. This system avoids wear and physical contact, which
might destroy the accuracy and sensitivity of the instrument. Because it
uses no fluid, it eliminates evaporation, freezing, or toxicity
problems. It also ensures inertia-free, drift-free pointer movement and
offers unusually high resistance to shock and vibration. The gauges on
this device are also unharmed by pressure surges and ambient temperature
fluctuations. The device quickly provides gas pressures---positive,
negative, or differential. Because operation does not depend on gravity,
you can either face- or flush-mount the device in any position.

### Photohelic Differential Pressure Device

Photohelic gauges use a precise pressure switch that lets you control
low and high gas pressures. Knob controls adjust dual set points, which
allow for variable deadband control. These set points contain photocells
that actuate double pull double throw (DPDT) relays when pressures reach
a preset upper or lower limit, and they can be set up to control
corrective damper action. As the pressure changes in response to damper
motion, and the indicator returns to the null band, damper motion is
halted.

s)  For the position detection devices listed, explain how the detector
    provides an output representative of the position being represented:

-   Limit switches

-   Potentiometer

-   Linear variable differential transformer types

### Limit Switches

A limit switch is a mechanical device which can be used to determine the
physical position of equipment. The limit switch gives on/off output
that corresponds to valve position. Normally, limit switches are used to
provide full-open or full-shut indications. Many limit switches are of
the push-button variety. When the valve extension comes in contact with
the limit switch, the switch depresses to complete, or turn on, the
electrical circuit. As the valve extension moves away from the limit
switch, spring pressure opens the switch, turning off the circuit.

### Potentiometer

A potentiometer provides an accurate indication of position throughout
the travel of a valve or control rod. The extension is physically
attached to a variable resistor. As the extension moves up or down, the
resistance of the attached circuit changes, changing the amount of
current flow in the circuit. The amount of current is proportional to
the valve position. Potentiometer valve position indicator failures are
normally electrical in nature. An electrical short or open will cause
the indication to fail at one extreme or the other. If an increase or
decrease in the potentiometer resistance occurs, erratic indicated valve
position occurs.

### Linear Variable Differential Transformer Types

A linear variable differential transformer (LVDT) is a device which
provides accurate position indication throughout the range of valve or
control rod travel. The extension valve shaft, or control rod, is made
of a metal suitable for acting as the movable core of a transformer.
Moving the extension between the primary and secondary windings of a
transformer causes the inductance between the two windings to vary,
thereby varying the output voltage proportional to the position of the
valve or control rod extension. If the open and shut position is all
that is desired, two small secondary coils could be utilized at each end
of the extension's travel. As a rule, failures are limited to rare
electrical faults which cause erratic or erroneous indications. An open
primary winding will cause the indication to fail to some predetermined
value equal to zero differential voltage. This normally corresponds to
mid-stroke of the valve. A failure of either secondary winding will
cause either full-open or full-closed indications.

t)  Referring to a piping and instrumentation drawing (P&ID) containing
    temperature, pressure, level, flow, or position detection
    components, explain their function in the designated system and
    relationship to system safety.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

u)  Discuss the importance of safety and process instrumentation to
    nuclear safety including redundancy and calibration requirements.

Equipment used for monitoring system operations that are important to
safety should have a redundancy built into the system or should be built
in such a way as to maintain a higher level of reliability. Also, the
equipment must provide a high level of reliability in calibration
operations and performance to guarantee that all indications relating to
safety are accurate and indicate true conditions.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of piping and instrumentation drawings (P&ID).

a)  Given a piping and instrumentation drawing, identify/interpret the
    symbols used for system components including the following at a
    minimum:

-   Valves

-   Pumps

-   Heat exchangers

-   Filters/strainers

-   Fans

-   Compressors

-   Instruments

-   Indicators

-   Controllers

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

v)  Identify how valve conditions (open/closed) are depicted.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

w)  Determine and follow system flowpath(s).

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

x)  Discuss the role of piping and instrumentation diagrams relative to
    identification of failure modes and mapping fault propagation
    through networks to support the identification of accident
    vulnerabilities.

P&IDs are usually designed to present functional information about a
system or component. As a rule, P&IDs do not have a drawing scale and
present only the relationship or sequence between components. These
drawings only present information on how a system functions, not the
actual physical relationships. These drawings allow for easy
identification of important equipment and also allow an individual to
trace component failure throughout the system to readily identify how
faults propagate and to identify weakness in system components essential
for safety.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of electrical diagrams and schematics.

a)  Given a system diagram, identify/interpret the following symbols:

-   **Motors**

-   **Controllers**

-   **Breakers**

-   **Generators**

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-   Batteries

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

y)  Given the appropriate diagram, state the condition
    (energized/de-energized) in which all electrical devices are shown,
    unless otherwise noted on the diagram.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

z)  Given a system diagram, identify the power sources and/or loads and
    their status.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

a)  Discuss the role of electrical one-line diagrams for identifying
    failure modes and for mapping fault propagation through networks to
    support the identification of accident vulnerabilities.

Electrical diagrams are usually designed to present functional
information about a system or component. As a rule electrical diagrams
do not have a drawing scale and present only the relationship or
sequence between components. These drawings only present information on
how a system functions, not the actual physical relationships. These
drawings allow for easy identification of important equipment and also
allow an individual to trace component failure throughout the system to
readily identify how faults propagate and to identify weakness in system
components essential for safety.

## Nuclear safety specialists shall demonstrate a familiarity level knowledge of electrical logic diagrams.

a)  Given a logic diagram, identify/interpret the symbols used on logic
    diagrams to represent the components.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

b)  Identify the symbols used to denote a logical "1" (high/on) and a
    logical "0" (low/off) as used in logic diagrams.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

c)  Given a logic diagram and appropriate information, determine the
    output of each component and the logic circuit.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

d)  Given a logic diagram, identify three different trip settings and
    trace the resulting actions should the trip occur.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

e)  Discuss the role of control logic diagrams in identifying failure
    modes and for mapping fault propagation through networks to support
    the identification of accident vulnerabilities.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working level knowledge of radioactivity and transformation mechanisms.

a)  Define the following term:

-   Activation

### Activation

Activation is the process by which materials become radioactive.

f)  Given the "Chart of Nuclides," trace the decay chain for a specified
    nuclide.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

g)  Given either half-life or the radioactive decay constant, solve
    radioactive decay problems.

The following equation is applied:

A~t~ = Ao e^-λt^

A~t~ = activity present at time t

A~o~ = activity initially present

λ = decay constant (time)^-1^

t = time

h)  Using the specific activity or decay constant of an isotope, convert
    between mass quantities and curies.

The relationship between activity, number of atoms, and decay constant
is shown below:

A = λ N

where:

A = Activity of the nuclide (disintegrations/second)

λ = decay constant of the nuclide (second)^-1^

N = Number of atoms of the nuclide in the sample

The unit of measure for mass is the atomic mass unit (amu). One atomic
mass unit is equal to 1.66 x 10^-24^ grams. A *curie* is a unit of
measure of the rate of radioactive decay equal to 3.7 x 10^10^
disintegrations per second. This is approximately equivalent to the
number of disintegrations that one gram of radium-226 will undergo in
one second.

## Nuclear safety specialists shall demonstrate a working level knowledge of principles and concepts for internal and external dosimetry and dose consequences.

a)  Define the following terms:

-   Committed effective dose equivalent

-   Total effective dose equivalent

-   Whole body

-   Derived air concentrations (DAC)

-   Annual limit of intake (ALI)

-   Weighting factors

-   Stochastic effects

-   Non-stochastic (deterministic) effects

### Committed Effective Dose Equivalent

The committed effective dose equivalent is the dose equivalent
calculated to be received by a tissue or organ over a 50-year period
after the intake of a radionuclide into the body.

### Total Effective Dose Equivalent

The total effective dose equivalent is the sum of the effective dose
equivalent and the committed effective dose equivalent.

### Whole Body

Whole body refers to a dose resulting from exposing the entire body to
radiation.

### Derived Air Concentration (DAC)

The DAC is the airborne concentration that equals the ALI divided by the
volume of air breathed by an average worker for a working year of 2000
hours.

### Annual Limit of Intake (ALI)

The ALI is the derived limit for the permissible amount of radioactive
material taken into the body of an adult radiation worker by inhalation
or ingestion in a year.

### Weighting Factors

Generally, weighting factors refers to a method of attaching different
importance values to different items or characteristics; in the context
of radiation protection, the proportion of the risk of effects resulting
from irradiation of a particular organ or tissue to the total risk of
effects when the whole body is irradiated uniformly.

### Stochastic Effects

Stochastic effects are malignant and hereditary diseases for which the
probability of an effect occurring, rather than its severity, is
regarded as a function of dose without a threshold for radiation
protection purposes.

### Non-Stochastic (Deterministic) Effects

Non-stochastic (deterministic) effects are effects due to radiation
exposure for which the severity varies with the dose and for which a
threshold normally exists.

i)  Discuss the conservatisms of International Commission on
    Radiological Protection (ICRP) Publications 26, 30, 60, 68, 71, and
    72 as they relate to dose conversion factors and consequence
    analysis in hazard categorization and accident analysis.

A Review of ICRP Publications 26, 30, 60, 68, 71, and 72 is required.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the biological effects of radiation.

a)  Describe the effects of radiation exposure on the cellular level
    including:

-   Direct effects

-   Indirect effects

### Direct Effects

If radiation interacts with the atoms of the DNA molecule, or some other
cellular component critical to the survival of the cell, it is referred
to as a direct effect. Such an interaction may affect the ability of the
cell to reproduce and, thus, survive. If enough atoms are affected such
that the chromosomes do not replicate properly, or there is a
significant alteration in the information carried by the DNA molecule,
then the cell may be destroyed by "direct" interference with its
life-sustaining system.

### Indirect Effects

If a cell is exposed to radiation, the probability of the radiation
interacting with the DNA molecule is very small since these critical
components make up such a small part of the cell. However, each cell,
just as is the case for the human body, is mostly water. Therefore,
there is a much higher probability of radiation interacting with the
water that makes up most of the cell's volume. When radiation interacts
with water, it may break the bonds that hold the water molecule
together, producing fragments such as hydrogen (H) and hydroxyls (OH).
These fragments may recombine or may interact with other fragments or
ions to form compounds, such as water, which would not harm the cell.
However, they could combine to form toxic substances, such as hydrogen
peroxide (H2O2), which can contribute to the destruction of the cell.

j)  Describe the regulatory limits established by EPA federal guidance
    reports No.11 and 13 and subsequent DOE radiological evaluation
    guideline and EPA protective actions guides for nuclear accidents.

The documents may be found at
<http://www.epa.gov/radiation/docs/federal/520-1-88-020.pdf> and
<http://www.epa.gov/radiation/docs/federal/402-r-99-001.pdf>.

k)  Identify and discuss the range of doses above which one may expect
    acute radiation illness and early fatalities.

Acute doses of radiation can cause a pattern of clearly identifiable
symptoms (syndromes). These conditions are referred to in general as
Acute Radiation Syndrome. Radiation sickness symptoms are apparent
following acute doses greater than or equal to100 rad. Acute whole body
doses of greater than or equal to 450 rad may result in a statistical
expectation that 50 percent of the population exposed will die within 60
days without medical attention. As in most illnesses, the specific
symptoms, the therapy that a doctor might prescribe, and the prospects
for recovery vary from one person to another and are generally dependent
on the age and general health of the individual.

Blood-forming organ (bone marrow) syndrome (an exposure greater than
100 rad) is characterized by damage to cells that divide at the most
rapid pace (such as bone marrow, the spleen, and lymphatic tissue).
Symptoms include internal bleeding, fatigue, bacterial infections, and
fever.

Gastrointestinal tract syndrome (an exposure greater than 1000 rad) is
characterized by damage to cells that divide less rapidly (such as the
linings of the stomach and intestines). Symptoms include nausea,
vomiting, diarrhea, dehydration, electrolytic imbalance, loss of
digestion ability, bleeding ulcers, and the symptoms of blood-forming
organ syndrome.

Central nervous system syndrome (an exposure greater than 5000 rad) is
characterized by damage to cells that do not reproduce such as nerve
cells. Symptoms include loss of coordination, confusion, coma,
convulsions, shock, and the symptoms of the blood forming organ and
gastrointestinal tract syndromes. Scientists now have evidence that
death under these conditions is not caused by actual radiation damage to
the nervous system, but rather from complications caused by internal
bleeding and fluid and pressure build-up on the brain. Other effects
from an acute dose include the following:

-   Exposure to 200 to 300 rad to the skin can result in the reddening
    of the skin (erythema), similar to mild sunburn, and may result in
    hair loss due to damage to hair follicles.

-   Exposure to 125 to 200 rad to the ovaries can result in prolonged or
    permanent suppression of menstruation in about fifty percent of
    women.

-   Exposure to 600 rad to the ovaries or testicles can result in
    permanent sterilization.

-   Exposure to 50 rad to the thyroid gland can result in benign
    (non-cancerous) tumors.

As a group, the effects caused by acute doses are called deterministic.
Broadly speaking, this means that the severity of the effect is
determined by the amount of dose received. Deterministic effects usually
have some threshold level below which the effect will probably not
occur, but above which the effect is expected. When the dose is above
the threshold, the severity of the effect will increase as the dose
increases.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the principles and use of radiological instrumentation and radiological monitoring/survey practices.

a)  Discuss the purpose, principles of detection and operation, and
    field application of the following:

-   Continuous air monitors (CAM) including tritium alarms

-   Area radiation monitors (ARM)

-   Criticality detection/alarm systems

-   Process radiation monitors

### Continuous Air Monitors (CAM) including Tritium Alarms

The requirements of air monitoring programs should be established to
ensure that airborne radioactivity monitoring is performed at a
frequency consistent with the existing and potential hazards and
activities planned in the area. The selection of air monitoring
equipment should be based on the specific job being monitored. Air
monitoring equipment includes portable and fixed air sampling equipment
and CAMs. Air sampling equipment shall be used where an individual is
likely to receive an annual exposure of 40 or more derived air
concentration (DAC) hours. This intake generally represents a committed
effective dose equivalent to an individual of approximately 100
millirem. Samples shall also be taken as necessary to characterize the
hazard in areas where respiratory protection devices have been
prescribed for protection against airborne radionuclides. Air samples
should be adequate to evaluate the concentrations of airborne
radioactive materials at the individual's work locations. Real-time (or
continuous) air monitors are used to provide early warning to
individuals of events that could lead to substantial unplanned exposures
to airborne radioactivity. Such exposures could result from a breakdown
of engineered controls or improper establishment of boundaries during
work that creates airborne radioactivity. Real-time air monitoring is
necessary to detect and provide warning of airborne radioactivity
concentrations that warrant immediate action to terminate inhalation of
airborne radioactive material. Air sampling equipment should be
positioned to measure air concentrations to which individuals are
exposed. If this cannot be achieved, a program of personal
breathing-zone air sampling should be initiated. Air monitoring
equipment shall be routinely calibrated and maintained on an established
frequency. Air monitoring equipment should be calibrated at least once
each year. CAMs should be capable of measuring 1 DAC when averaged over
8 hours (8 DAC-hours) under laboratory conditions. Real-time air
monitoring equipment required from above should have alarm capability
and sufficient sensitivity to alert individuals that immediate action is
necessary to minimize or terminate inhalation exposures. A technical
basis document should be developed for the airborne radioactivity
monitoring program. The technical basis document should provide the
basis for air monitor selection, placement, and operation. The proper
operation of CAMs should be verified daily by performing an operational
check. Operational checks should include positive air-flow indication,
non-zero response to background activity, and internal check sources or
60 Hz electronic checks when available. Real-time air monitoring
equipment operation should be verified weekly by checking for instrument
response with a check source or with ambient levels of radon daughters.

### Area Radiation Monitors (ARMs)

ARMs should be installed in frequently occupied locations with the
potential for unexpected increases in dose rates and in remote locations
where there is a need for local indication of dose rates prior to
personnel entry. ARMs should not be substituted for radiation exposure
surveys in characterizing a workplace. The need for and placement of
ARMs should be documented and assessed when changes to facilities,
systems, or equipment occur. ARMs should be tested at least quarterly to
verify audible alarm system operability, audibility under ambient
working conditions, and operability of visual alarms when so equipped.
If installed instrumentation is removed from service for maintenance or
calibration, a radiation monitoring program providing similar detection
capability should be maintained, consistent with the potential for
unexpected increases in radiation dose rates. Where an ARM is
incorporated into a safety interlock system, the circuitry should be
such that a failure of the monitor either prevents entry into the area
or prevents operation of the radiation producing device. If the
circuitry is required to ensure compliance with the high radiation area
access control requirements, then the circuitry shall be fail-safe.

### Criticality Detection/Alarm Systems

Criteria for criticality detection/alarm systems are listed below:

-   Criticality alarm systems shall be designed to detect immediately
    the minimum accident of concern. The minimum accident may be assumed
    to deliver the equivalent of an absorbed dose in free air of 20 rad
    at a distance of 2 meters from the reacting material within 60
    seconds.

-   Systems shall be designed so that instrument response and alarm
    latching shall occur as a result of radiation transients of
    1-millisecond duration.

-   The alarm signal shall be for evacuation purposes only, and of
    sufficient volume and coverage to be heard in all areas that are to
    be evacuated. Very high audio background noise in some areas may
    require that the alarm be supplemented with visual signals; however,
    high background noise is a dangerous situation that should be
    prevented by design.

-   Instrument response to radiation shall be calibrated periodically to
    confirm the continuing performance of the instrument. The
    calibration interval may be determined on the basis of experience
    but shall be no less frequent than annually. Tests should be
    performed at least monthly and the results of testing should be
    documented.

-   The nuclear criticality safety program shall be evaluated and
    documented and shall include

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-   an assessment of the need for criticality accident detection devices
    and alarm systems

-   installation of such equipment where total risk to personnel will be
    reduced

The need for criticality alarm systems shall be evaluated for all
activities in which the inventory of fissionable material in individual
unrelated work areas exceeds the design limits. If the fissionable
material mass exceeds the limits and the probability of criticality is
greater than 10-6 per year, a criticality alarm system shall be provided
to cover occupied areas in which the expected dose exceeds 12 rad in
free air. The criticality alarm system should include a criticality
detection device and a personnel evacuation alarm. Reasonable grounds
for incredibility may be presented on the basis of commonly accepted
engineering judgment. The alarm signal shall be for immediate evacuation
purposes only and of sufficient volume and coverage to be heard in all
areas that are to be evacuated. The alarm trip point shall be set low
enough to detect the minimum accident of concern. The minimum accident
of concern may be assumed to deliver the equivalent of an absorbed dose
in free air of 20 rad at a distance of 2 meters from the reacting
material within 60 seconds. The alarm signal shall activate promptly
when the dose rate at the detectors equals or exceeds a value equivalent
to 20 rad/minute at 2 meters from the reacting material. A visible or
audible warning signal shall be provided at a normally occupied location
to indicate system malfunction or loss of primary power. Each alarm
system should be tested at least once every three months. An evacuation
drill shall be conducted at least annually. Criticality alarm systems
may consist of one to several detectors per unit. In multi-detector
units, at least two detectors shall be at the alarm level before
initiating the alarm; in redundant systems, failure of any single
channel shall lead into the trip state.

### Process Radiation Monitors

Process radiation monitors are monitors that are used just like the
ARMs, but which are applied for a certain process or operation.

## Nuclear safety specialists shall demonstrate a working-level knowledge of 10 CFR 830.204, Documented Safety Analysis, and DOE Guide 421.1-2, with respect to its impact on Department nuclear safety.

a)  Discuss the basic purposes and objectives of a Documented Safety
    Analysis.

The purpose of the DSA is to establish safety basis requirements for
hazard category 1, 2, and 3 DOE nuclear facilities. The documented
safety analysis for a hazard category 1, 2, or 3 DOE nuclear facility
must, as appropriate for the complexities and hazards associated with
the facility

-   describe the facility (including the design of SSCs) and the work to
    be performed;

-   provide a systematic identification of both natural and man-made
    hazards associated with the facility;

-   evaluate normal, abnormal, and accident conditions, including
    consideration of natural and man-made external events,
    identification of energy sources or processes that might contribute
    to the generation or uncontrolled release of radioactive and other
    hazardous materials, and consideration of the need for analysis of
    accidents which may be beyond the design basis of the facility;

-   derive the hazard controls necessary to ensure adequate protection
    of workers, the public, and the environment; demonstrate the
    adequacy of these controls to eliminate, limit, or mitigate
    identified hazards; and define the process for both maintaining the
    hazard controls current at all times and controlling their use;

-   define the characteristics of the safety management programs
    necessary to ensure the safe operation of the facility, including
    (where applicable) quality assurance, procedures, maintenance,
    personnel training, conduct of operations, emergency preparedness,
    fire protection, waste management, and radiation protection;

-   with respect to a nonreactor nuclear facility with fissionable
    material in a form and amount sufficient to pose a potential for
    criticality, define a criticality safety program that

-   ensures that operations with fissionable material remain subcritical
    under all normal and credible abnormal conditions

-   identifies applicable nuclear criticality safety standards

-   describes how the program meets applicable nuclear criticality
    safety standards

l)  Describe the responsibilities of contractors for the development and
    maintenance of a Documented Safety Analysis.

In establishing the safety basis for a hazard category 1, 2, or 3 DOE
nuclear facility, the contractor responsible for the facility must

-   define the scope of the work to be performed;

-   identify and analyze the hazards associated with the work;

-   categorize the facility consistent with DOE-STD-1027-92;

-   prepare a documented safety analysis for the facility;

-   establish the hazard controls upon which the contractor will rely to
    ensure adequate protection of workers, the public, and the
    environment.

In maintaining the safety basis for a hazard category 1, 2, or 3 DOE
nuclear facility, the contractor responsible for the facility must

-   update the safety basis to keep it current and to reflect changes in
    the facility, the work, and the hazards as they are analyzed in the
    documented safety analysis;

-   annually submit to DOE either the updated documented safety analysis
    for approval or a letter stating that there have been no changes in
    the documented safety analysis since the prior submission;

-   incorporate in the safety basis any changes, conditions, or hazard
    controls directed by DOE.

m)  Define the following terms and discuss the purpose of each:

-   Design basis

-   Engineered design features

-   Safety analysis

-   Safety basis

-   Basis for interim operation

-   Transportation safety document

-   Safety analysis report for packaging

-   Health and safety plan

-   Hazards analysis report

### Design Basis

The design basis is comprised of the design inputs, the design
constraints, and the design analysis and calculations. It includes
topical areas such as seismic qualification, fire protection, and safe
shutdown. It encompasses consideration of such factors as plant
availability, plant efficiency, costs, and maintainability, and that
subset that relates to safety and the authorization basis.

### Engineered Design Feature

Engineered design features are the safety design limits on engineered
controls, either passive or active, and the bases placed on equipment
designs or operations to ensure safe conditions under all normal,
abnormal, and accident conditions.

### Safety Analysis

A safety analysis reviews the extent to which a nuclear facility can be
operated with respect to workers, the public, and the environment,
including a description of the conditions, safe boundaries, and hazard
controls that provide the basis for ensuring safety.

### Safety Basis

The safety basis is comprised of the documented safety analysis and
hazard controls that provide reasonable assurance that a DOE nuclear
facility can be operated safely in a manner that adequately protects
workers, the public, and the environment.

### Basis for Interim Operation

The basis for interim operations is the documented establishment of
allowable operations for a current facility and its operational controls
until a fully compliant document is developed and approved by DOE.

### Transportation Safety Document

The transportation safety document describes the methodology and
compliance process to meet equivalent safety for any deviation from the
hazardous materials regulations.

### Safety Analysis Report for Packaging

The safety analysis report for packaging is the document that
demonstrates that the package will satisfy the standards of 10 CFR 71
and any other standards that the Headquarters certifying official may
determine applicable for granting a certificate of compliance. Each
certificate shall be issued for a five-year period, after which the
certificate must be renewed by the Headquarters certifying official.

### Health and Safety Plan (HASP)

For all facilities having or dealing with hazardous waste, a HASP shall
be developed. It shall, at a minimum, cover each site, task, operation,
activity, or project, and address the following elements:

-   Health and safety risk or hazard identification for each site task,
    operation, activity, or project defined in the workplan (All
    chemical, radiological, and physical hazards, e.g., noise and heat,
    must be identified in the HASP. Actions taken to mitigate risks or
    hazards shall be identified. This risk or hazard identification
    should have a common basis with, or integrate the results of, other
    documents such as a SAR.)

-   Required training for employees for each site task, operation,
    activity, or project and hazard

-   Personnel protective equipment to be used by workers for each site
    task and operation

-   Employee medical surveillance requirements

-   Frequency and types of required personnel monitoring (including air,
    noise, radiation, heat, stress, etc.) including the methods of
    maintenance and calibration of monitoring and sampling equipment to
    be used

-   Site control measures including work zones, buddy systems, security,
    communications, and safety work practices

-   Decontamination procedures

-   Emergency response plans for safe and effective responses to
    emergencies

-   Confined space entry procedures

-   Programs to protect against and mitigate spills

### Hazards Analysis Report (HAR)

A HAR documents the hazard analysis (HA) of an operation and associated
activities. At a minimum, the HAR shall include the following:

-   An executive summary that provides an overview of the HAR and its
    main conclusions

-   An introduction that provides a discussion of the objectives, the
    scope of the analysis, the operations conducted, and the limitations
    and assumptions employed in the HA

-   A description of the nuclear explosive and its intrinsic hazards

-   A description of the nuclear explosives operations (NEO) and the
    facility(ies) where the operation is to be conducted (The discussion
    should focus on the facility and nuclear explosive configurations
    and processes, including equipment and tooling. The discussion
    should also address whether interfaces between the operation and
    facility have safety implications. Generic safety controls utilized
    during the operation should be discussed.)

-   A discussion of the methodology used to conduct the HA and derive
    safety controls and safety requirements

-   A summary of the identification of hazards and potential hazard
    scenarios under normal and abnormal conditions considering both
    internal and external environments for each step in the operations

n)  Describe the different requirements for the scope and content of
    each type of documented safety analysis and discuss the general
    content of each as well as the required sections of each.

The different types of safety analyses, the applicability of each, and
the required sections are described in 10 CFR 835 appendix A to subpart
B table-2. The URL is included below:

-   <http://www.eh.doe.gov/enforce/rands/10cfr830.pdf>

The content requirements for each type of DSA are contained in the
following documents. The document URLs are also included below:

-   DOE-STD-3009
    <http://www.eh.doe.gov/techstds/standard/std3009/std3009_cn2.pdf>

-   DOE-STD-3011
    <http://www.eh.doe.gov/techstds/standard/std3011/std3011.html>

-   DOE-STD-1120
    <http://www.eh.doe.gov/techstds/standard/std1120/s1120v1.pdf>

-   29 CFR 1910.120
    <http://www.access.gpo.gov/nara/cfr/waisidx_04/29cfrv5_04.html>

-   DOE-STD-3016
    <http://www.eh.doe.gov/techstds/standard/std3016/dpstd301699.pdf>

-   DOE-O-460.1A
    <http://www.directives.doe.gov/pdfs/doe/doetext/neword/460/o4601a.pdf>

-   [DOE-O-461.1](http://www.directives.doe.gov/pdfs/doe/doetext/neword/460/o4601a.pdf)
    <http://www.directives.doe.gov/pdfs/doe/doetext/neword/461/o4611.pdf>

-   DOE-M-461.1-1
    <http://www.directives.doe.gov/pdfs/doe/doetext/neword/461/m4611-1.pdf>

o)  Discuss the approval requirements for the documented safety analysis
    for new facilities and subsequent changes to the documented safety
    analysis.

DOE evaluates the DSA by considering the extent to which the DSA (1)
adequately addresses the criteria set forth in 10 CFR 830.202 and 10 CFR
830.204, and (2) satisfies the provisions of the methodology used to
prepare the DSA. DSA review and approval focuses on the adequacy of the
following approval bases:

-   Base information

-   Hazard and accident analyses

-   Safety SSCs

-   Derivation of TSRs

-   Safety management program characteristics

Once technical justification exists to support conclusions that the DSA
adequately describes how the facility is satisfactory with respect to
the approval bases, the DSA may generally be considered adequate. These
approval bases also form the foundation for documenting DSA approval in
a safety evaluation report (SER). Any changes to the DSA are documented
in the unreviewed safety question (USQ) for the facility and reviewed by
DOE for the above mentioned requirements.

p)  Define who approves facility operations prior to achieving
    Documented Safety Analysis upgrade approval.

The USQ rule has a primary role in preserving the DOE safety basis for
each nuclear facility. The concept of the USQ allows contractors to make
physical and procedural changes and to conduct tests and experiments
without prior DOE approval, as long as these changes do not affect the
safety basis of the facility. If the changes do affect the safety basis,
the DOE must approve the USQ.

q)  Discuss the provisions for deviations, temporary and permanent
    exemptions from the 10 CFR 830.204, and safe harbor methodologies.

Provisions for deviations, exemptions, or safe harbor methodologies are
made for the following activities:

-   Activities that are regulated through a license by the Nuclear
    Regulatory Commission (NRC) or a State under an Agreement with the
    NRC, including activities certified by the NRC under section 1701 of
    the Atomic Energy Act (Act);

-   Activities conducted under the authority of the Director, Naval
    Nuclear Propulsion, pursuant to Executive Order 12344, as set forth
    in Public Law 106--65;

-   Transportation activities which are regulated by the Department of
    Transportation;

-   Activities conducted under the Nuclear Waste Policy Act of 1982, as
    amended, and any facility identified under section 202(5) of the
    Energy Reorganization Act of 1974, as amended;

-   Activities related to the launch approval and actual launch of
    nuclear energy systems into space.

r)  Discuss the application of the graded approach relative to the DSA
    development.

Application of the graded approach ensures that the level of analysis,
documentation, and actions used to comply with a requirement is
commensurate with

-   the relative importance to safety, safeguards, and security;

-   the magnitude of any hazards involved;

-   the life cycle stage of a facility;

-   the programmatic mission of a facility;

-   the particular characteristics of a facility;

-   the relative importance of radiological and nonradiological hazards;

-   any other relevant factor from the rule.

## Nuclear safety specialists shall demonstrate a working-level knowledge of the safety basis requirements for environmental restoration and decommissioning activities.

a)  Discuss the application of DOE-STD-1120 and 29 CFR 1910.120 to
    decommissioning and certain environmental restoration activities.

The environmental, safety, and health (ES&H) principles presented in
DOE-STD-1120 and 29 CFR 1910 apply to all phases of facility
disposition. These phases comprise deactivation, decommissioning, and
any long-term surveillance and maintenance (S&M) activities that are
conducted prior to each of these phases. The guidance contained within
DOE-STD-1120 and 29 CFR 1910 is not intended for non-facility
environmental restoration activities or for material stabilization
activities conducted as part of the operations phase of the facility
life-cycle. DOE-STD-1120 and 29 CFR 1910 are intended for use by
facility disposition project teams implementing the DOE Work Smart
Standards (WSS) process, as well as for use by those teams following a
compliance-based approach using a Standards/Requirements Identification
Document (S/RID) process. All provisions of DOE-STD-1120 and 29 CFR 1910
cover any treatment, storage, or disposal (TSD) operation regulated by
other rules such as state law or RCRA. If an area is used primarily for
treatment, storage, or disposal, any emergency response operations in
that area shall ensure compliance with all parts of DOE-STD-1120 and 29
CFR 1910.120.

s)  Discuss the content of a safety basis HASP and how it can be used in
    a dynamic project, including management of hazard controls.

A HASP is a "living document" to be modified and updated as new
regulations and other requirements are issued. Guiding the development
of this document has been the principle that an effective and
high-quality HASP must provide

-   a clear chain of command for safety and health activities

-   accountability for safety and health performance

-   well defined Headquarters' expectations regarding safety and health

-   well defined task and operational hazards/risks

-   comprehensive hazard prevention and control methods

-   recordkeeping requirements to track program progress

As this document was developed, the working group applied the following
assumptions:

-   It is a generic document to be adapted on a site-by-site basis.

-   An adequate site characterization and comprehensive workplan exists
    at each field site.

-   Operations analysis and risk identification methods are adequate.

-   A written site safety and health program exists as required in 29
    CFR 1910.120 (b).

-   Written site Standard Operating Procedures exist.

-   Site-specific information is available.

The site-specific HASP should include a description of the site that the
HASP will encompass and its applicability to operations. In developing
this description, the preparer should include

-   a brief description

-   background information (e.g., site history, prior site activities)

-   known site contamination

-   a synopsis of site characterization

-   site operations to be performed

t)  Discuss the provisions of 29 CFR 1926.65 and 1910.120 for Technical
    Safety Requirements.

The areas of engineering controls, work practices, personnel protective
equipment (PPE) for substances, and applicable training for these rules
are already developed in some context for use as TSRs for hazardous
material or waste facilities.

u)  Describe the application of DOE-STD-3011, Guidance for Preparation
    of Basis for Interim Operations (BIO) Documents, to nuclear
    facilities safety basis.

The DOE-STD-30011 is applicable to the following:

-   The deactivation of a nuclear facility

-   A nuclear facility with a limited operational life

-   The transition surveillance and maintenance of a nuclear facility

### Deactivation

Deactivation is the process of placing a facility in a stable and known
condition and includes the removal of hazardous and radioactive
materials.

### Nuclear Facility with a Limited Operational Life

A nuclear facility with a limited operational life is a facility for
which there is a short remaining operational period before ending the
facility's mission and initiating deactivation and decommissioning, and
for which there are no intended additional missions other than cleanup.

### Transition Surveillance and Maintenance Activities

Transition surveillance and maintenance activities are activities
conducted when a facility is not operating, or during deactivation,
decontamination, and decommissioning operations when surveillance and
maintenance are the predominant activities being conducted at the
facility. These activities are necessary for satisfactory containment of
hazardous materials and protection of workers, the public, and the
environment. These activities include providing periodic inspections,
maintenance of SSCs, and actions to prevent the alteration of hazardous
materials to an unsafe state.

v)  Discuss the limitations associated with the safety basis for
    "certain environmental restoration activities."

Environmental restoration is the process by which contaminated sites and
facilities are identified and characterized and by which existing
contamination is contained, or removed and disposed of. According to 10
CFR 830, the approved safe harbor methods for dealing with environmental
restoration activities are the method in DOE-STD--1120--98 or successor
document or the method using the provisions in 29 CFR 1910.120 (or 29
CFR 1926.65 for construction activities) for developing a safety and
health program and a site-specific health and safety plan.

w)  Review a chapter of an environmental restoration/decommissioning
    activity safety basis.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

x)  Discuss the unique considerations with respect to control selection
    for waste management activities.

Controls used in waste management operations are developed out of the
Code of Federal Regulations. The engineered controls are addressed in 29
CFR 1910.120 and 29 CFR 1926.65. Controls for waste management
activities are addressed in the rules.

## Nuclear safety specialists shall demonstrate a working-level knowledge of 10 CFR 830.207, DOE Approval of Safety Basis and DOE-STD-1104, with respect to its impact on department nuclear safety.

a)  Describe the basic purpose and contents of a Safety Evaluation
    Report.

The purpose of the SER is to allow DOE to review the safety basis
developed by the contractor and to approve operation of the facility to
the standards set in the DSA and TSRs. The contents of a SER are listed
below:

-   Title page

-   Signature page

-   Executive summary

-   Review process

-   Base information

-   Hazard and accident analysis

-   Safety SSCs

-   Derivation of TSRs

-   Safety management program characteristics

-   TSRs

-   Records

y)  Describe the bases for approval contained in a Safety Evaluation
    Report.

DSA review and approval focuses on the adequacy of the following
approval bases. These approval bases also form the foundation for
documenting DSA approval in a SER:

-   Base information

-   Hazard and accident analyses

-   Safety SSCs

-   Derivation of TSRs

-   Safety management program characteristics

Base information is the first of the approval bases that should be
reviewed and encompasses elements of DSA preparation, completeness, and
general content. Base information is not reviewed for accuracy in and of
itself, but for sufficiency to allow assessment of the other approval
bases that rely on this information. The review for sufficiency can
range from a simple screening effort to more detailed discussions,
depending on the complexity of the DSA.

Insufficient or incomplete base information in a DSA may prevent further
review of the DSA. Reviewers should require resolution of major
discrepancies in base information before evaluation of the more specific
aspects of the safety basis proceeds. It is for this reason that the SER
need only provide a brief statement as to the adequacy of base
information. The approval bases for the TSR document are the TSR
provisions. These TSR provisions may be design features, safety limits,
operating limits, surveillance requirements, or administrative controls.
The approval bases for a TSR document include a disciplined analysis and
tracing of commitments to hazard controls through appropriate provisions
that implement these controls in a TSR document. In some cases the
specific treatment of safety controls in the TSR is committed to in the
DSA; in other cases, it is a judgment call as to the appropriate TSR
treatment. Determining the adequacy of the TSR provisions generally
entails being able to conclude that

-   hazard controls that are discussed in the DSA are faithfully
    translated into TSR provisions

-   the TSR provisions are appropriate and consistent with the DSA

The sources of information in a DSA regarding these provisions are the
hazards analysis, including a description of hazard controls; the
description of safety SSCs, the classification of these SSCs as safety
class, safety significant, or other important defense-in-depth SSCs, the
description of the functional requirements for the safety SSCs, and the
derivation of TSRs section; and the descriptions of the safety
management programs.

z)  Prepare a Safety Evaluation Report consistent with DOE-STD-1104 for
    a safety basis amendment or annual update.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working-level knowledge of 10 CFR 830.206, Preliminary Documented Safety Analysis, with respect to its impact on department nuclear safety.

a)  Describe the application of the requirements of DOE O 420.1A and its
    guidance to the development process for the Preliminary Documented
    Safety Analyses.

Non-reactor nuclear facilities shall be designed with the objective of
providing multiple layers of protection to prevent or mitigate the
unintended release of radioactive materials to the environment. Defense
in depth shall include

-   minimization of material at risk

-   the use of conservative design margins and quality assurance

-   the use of successive physical barriers for protection against the
    release of radioactivity

-   the provision of multiple means to ensure critical safety functions

-   the use of equipment and ACs which restrict deviations from normal
    operations and provide for recovery from accidents to achieve a safe
    condition

-   means to monitor accident releases required for emergency responses

-   the provision of emergency plans for minimizing the effects of an
    accident

Facilities shall be sited and designed in a manner that provides
adequate protection for the health and safety of the public and workers,
including those at adjacent facilities, from the effects of potential
facility accidents involving the release of radioactive materials. All
nuclear facilities with uncontained radioactive materials shall have
means to confine them. Such confinement will act to minimize the spread
of radioactive materials and the release of radioactive materials in
facility effluents during normal operations and potential accidents. For
a specific nuclear facility, the number and arrangement of confinement
barriers and their required characteristics shall be determined on a
case-by-case basis. Factors that shall be considered in confinement
system design shall include type, quantity, form, and conditions for
dispersing the material. Engineering evaluations, trade-offs, and
experience shall be used to develop practical designs that achieve
confinement system objectives. The adequacy of confinement systems to
effectively perform the required functions shall be documented and
accepted through the SAR. Facilities shall be designed to facilitate
safe deactivation, decommissioning, and decontamination at end of life.
Facilities shall be designed to facilitate inspections, testing,
maintenance, and repair and replacement of safety SSCs as part of an
overall reliability, availability, and maintainability program. The
objective is that the facility can be maintained in a safe state,
including during these operations, and in keeping with the ALARA
principle for occupational radiation exposure. Facilities shall be
designed to keep occupational radiation exposure within statutory limits
and incorporate ALARA principles in design, including design provisions
to facilitate decontamination during the operational period.

Facility process systems shall be designed to minimize the production of
wastes and minimize the mixing of radioactive and non-radioactive
wastes. Safety SSCs identified in accordance with this section shall,
commensurate with the importance of the safety functions performed, be
designed so that they can perform their safety functions when called
upon to operate, and under a quality assurance program that satisfies 10
CFR 830.120.

a)  Describe the sequencing of the Preliminary Documented Safety
    Analysis relative to design, procurement, construction and operation
    of new facilities.

The sequence for developing a PDSA is to prepare a PDSA for the facility
and then obtain DOE approval of the nuclear safety design criteria used
in preparing the PDSA. The PDSA must be approved before the contractor
can procure materials or components or begin construction. The DOE may
authorize the contractor to perform limited procurement and construction
activities without approval of a PDSA if DOE determines that the
activities are not detrimental to public health and safety and are in
the best interests of DOE.

b)  Describe the circumstances when a Preliminary Documented Safety
    Analysis must be prepared.

If construction begins after December 11, 2000, the contractor
responsible for a hazard category 1, 2, or 3 new DOE nuclear facility or
a major modification to a hazard category 1, 2, or 3 DOE nuclear
facility must prepare a PDSA for the facility.

c)  Describe the relationship between the PDSA and the design process.

The design process and PDSA are interrelated. They feed each other, with
changes in one having a direct influence on the other, so the PDSA and
design are "living documents" changing with the project.

d)  Perform a detailed review of a PDSA chapter.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working-level knowledge of 10 CFR 830.202, Safety Basis and DOE-STD-1027, with respect to its impact on department nuclear safety.

a)  Describe when a contractor must establish a safety basis for a
    facility.

If a contractor is responsible for a hazard category 1, 2, or 3 DOE
nuclear facility, the contractor must establish and maintain the safety
basis for the facility.

e)  Describe the requirements for the safety basis.

The safety basis must

-   define the scope of the work to be performed;

-   identify and analyze the hazards associated with the work;

-   categorize the facility consistent with DOE--STD--1027--92;

-   prepare a documented safety analysis for the facility;

-   establish the hazard controls upon which the contractor will rely to
    ensure adequate protection of workers, the public, and the
    environment.

f)  Describe the requirements the contractor must perform to maintain
    the safety basis.

In maintaining the safety basis for a hazard category 1, 2, or 3 DOE
nuclear facility, the contractor responsible for the facility must:

-   update the safety basis to keep it current and to reflect changes in
    the facility, the work, and the hazards as they are analyzed in the
    documented safety analysis;

-   annually submit to DOE either the updated documented safety analysis
    for approval or a letter stating that there have been no changes in
    the documented safety analysis since the prior submission;

-   incorporate in the safety basis any changes, conditions, or hazard
    controls directed by DOE.

g)  Discuss the purpose and determine the hazard categorization of an
    operating nuclear facility.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

h)  Describe the exclusion types for radionuclides associated with
    hazard categorization determination.

Sealed radioactive sources that are engineered to pass the special form
testing specified by the Department of Transportation (DOT) in 49 CFR
173.469 or testing specified by American National Standards Institute
(ANSI) N43.6, Sealed Radioactive Sources, may be excluded from summation
of a facility's radioactive inventory. The facility must have
documentation that the source or prototypes of the source have been
tested and have passed the tests specified by DOT or ANSI. Facilities
must also have in place a source control policy that complies with DOE
Notice 5400.9, Sealed Source Control Policy, and the source control
policy specified in Article 431 of the DOE RadCon Manual. Should a
sealed radioactive source fail, as indicated by an increase in the
removable activity, the source shall be removed from service and handled
in accordance with the source control policy established for the
facility. Hazardous materials used in exempted, commercially available
products should not be considered part of a facility's inventory. These
materials include timepieces, illumination devices, thermostats,
electron tubes, microwave receiver tubes, etc. Additionally, material
contained in DOT Type B shipping containers may also be excluded from
summation of a facility's radioactive inventory if the certificates of
compliance are kept current and the materials stored are authorized by
the certificate. However, Type B containers without overpack should have
heat protection provided by the facility's fire suppression system.
These exclusions do not apply to fissile material in the determination
of hazard category 2 status relative to criticality.

i)  Describe the differences between initial and final hazard
    categorizations and where these designations occur in the DSA
    development process.

The initial hazard categorization is specified in the preliminary
assessment of facility hazards. Once a hazards analysis has been
performed, the hazard categorization can be finalized. The final
categorization is based on an "unmitigated release" of available
hazardous material. For the purposes of hazard categorization,
"unmitigated" is meant to consider material quantity, form, location,
dispersibility, and interaction with available energy sources, but not
to consider safety features which will prevent or mitigate a release.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of Department of Energy (DOE) O 420.1A, Facility Safety, DOE G 420.1-1, Nonreactor Nuclear Safety Design Criteria and Explosive Safety Criteria Guide, for use with DOE O 420.1, Facility Safety, and DOE-STD-1020, Natural Phenomena Hazards Design and Evaluation Criteria for Department of Energy Facilities, with respect to its impact on Department nuclear safety.

a)  Discuss the purpose and policy associated with DOE O 420.1A,
    Facility Safety.

The purpose of this Order is to establish facility safety requirements
for DOE/NNSA for

-   nuclear safety design

-   criticality safety

-   fire protection

-   natural phenomena hazards mitigation

-   a system engineer program

This Order applies to all DOE/NNSA facilities with responsibility for
DOE-owned or DOE-leased facilities as follows:

-   All DOE nuclear and non-nuclear facilities

-   All DOE nonreactor nuclear facilities that are classified as hazard
    category 1, 2, or 3

-   All DOE explosives facilities

-   All DOE hazard category 1, 2, and 3 nuclear facilities

j)  Discuss the role of department nuclear safety specialists with
    respect to the implementation of the requirements of DOE Order
    420.1A, Facility Safety.

Nuclear safety specialists as directed by the cognizant Secretarial
Officer, shall ensure that the requirements falling within the scope of
this order are in compliance with this order. Nuclear safety specialists
are responsible for determining whether adequate protection can most
effectively be achieved by continuing to operate under the terms of
existing contracts requiring compliance with old orders or by modifying
the contract to incorporate the requirements of revised orders. Nuclear
safety specialists are responsible for determining that implementation
of new rule or order requirements will provide adequate protection prior
to requesting contract modification to drop old order requirements from
a contract.

k)  Discuss the department policy and objectives with respect to
    safety-class and safety-significant criteria.

Some principles that should be incorporated in a functional
classification process are listed below:

-   Protection of the public is contributed to by all facets of safety
    in design, including safety-class SSCs as well as safety-significant
    SSCs, and, in many DOE cases, by remote sitting. The expectation is
    that public design basis accident dose consequences would generally
    be a small fraction of the evaluation guideline.

-   Protection of the public is predominant in safety design; protection
    of workers is no less important. However, the degree of protection
    for facility workers achievable by SSCs is limited. Other factors
    such as disciplined conduct of operations, training, and safety
    management programs are no less important in assuring worker safety.

In prioritization of items for a facility safety strategy

-   minimization of hazardous materials (material at risk) is the first
    priority

-   safety SSCs are preferred over ACs

-   passive SSCs are preferred over active SSCs

-   preventative controls are preferred over mitigative controls

-   facility safety SSCs are preferred over personal protective
    equipment

-   controls closest to the hazard may provide protection to both
    workers and the public

-   controls that are effective for multiple hazards can be resource
    effective

l)  Discuss the facility and activity applicability of DOE O 420.1A,
    with respect to implementation associated with the design of nuclear
    facilities.

DOE O 420.1A requirements apply to the design and construction of new
hazard category 1, 2, and 3 DOE nonreactor nuclear facilities and to the
design and construction of modifications to existing DOE hazard category
1, 2, and 3 non-reactor nuclear facilities when the proposed
modifications significantly degrades the approved safety basis for the
facility. Modifications to facility design and construction during the
design and construction phase shall conform to the requirements for new
facilities. If explosives are also present in the facility, the design
will also consider the requirements of TM5-1300, Structures to Resist
the Effects of Accidental Explosions.

m)  Identify and discuss the use of DOE standards for seismic safety.

Executive Order 12699 establishes the minimum seismic requirements for
new Federal buildings. The National Earthquake Hazards Reduction Program
(NEHRP) updates the provisions required to meet these requirements every
3 years. The Interagency Committee on Seismic Safety in Construction
(ICSSC) compares model building codes with the NEHRP provisions.
Designers must consider the NEHRP provisions and ICSSC comparisons to
ensure the use of the proper model building code in their design and
evaluation. Currently the International Building Code (IBC) 2000 and
American Society of Civil Engineers (ASCE) 7-98 meet the requirements of
the NEHRP provisions. While using the IBC 2000 or successor documents,
designers must consider the Seismic Use Group and Seismic Design
Category.

Evaluations of existing SSCs must follow, or at least be measured
against, the natural phenomena hazards (NPH) criteria provided in
DOE-STD-1020. For SSCs not meeting these criteria and which cannot be
easily remedied, budgets and schedule for required strengthening must be
established on a prioritized basis. In addition, it is consistent with
the intent of the federal program developed by the ICSSC. When upgrading
becomes necessary, the design should be based on the current design
criteria in the standard for the new facility. DOE G 420.1-2 provides
guidance for facilities with a remaining service life of less than
5 years.

n)  Define PC-1, PC-2, PC-3 and PC-4 and their relationship to nuclear
    facility design and the DSA.

  --------------- ----------------------- -----------------------------------
  **Performance   **Performance Goal      **NPH Performance Goal Annual
  Category (PC)** Description**           Probability of Exceeding Acceptable
                                          Behavior Limits, P~F~**

  1               Maintain Occupant       10^-3^ of the onset of SSC damage
                  Safety                  to the extent that occupants are
                                          endangered

  2               Occupant Safety,        5x10^-4^ of SSC damage to the
                  Continued Operation     extent that the component cannot
                  with Minimum            perform its function
                  Interruption            

  3               Occupant Safety,        10^-4^ of SSC damage to the extent
                  Continued Operation,    that the component cannot perform
                  Hazard Confinement      its function

  4               Occupant Safety,        10^-5^ of SSC damage to the extent
                  Continued Operation,    that the component cannot perform
                  Confidence of Hazard    its function
                  Confinement             
  --------------- ----------------------- -----------------------------------

The design and evaluation criteria for SSCs in Performance Categories 1
and 2 are similar to those given in model building codes. Performance
Category 1 criteria include no extra conservatism against natural
phenomena hazards beyond that in model building codes that include
earthquake, wind, and flood considerations. Performance Category 2
criteria are intended to maintain the capacity to function and to keep
the SSC operational in the event of natural phenomena hazards. Model
building codes would treat hospitals, fire and police stations, and
other emergency-handling facilities in a similar manner to DOE-STD-1020
Performance Category 2 NPH design and evaluation criteria. Performance
Category 3 and 4 SSCs handle significant amounts of hazardous materials
or have significant programmatic impact. Damage to these SSCs could
potentially endanger worker and public safety and the environment or
interrupt a significant mission. As a result, it is very important for
these SSCs to continue to function in the event of natural phenomena
hazards, such that the hazardous materials may be controlled and
confined. For these categories, there must be a very small likelihood of
damage due to natural phenomena hazards. DOE-STD-1020 NPH criteria for
Performance Category 3 and higher SSCs are more conservative than
requirements found in model building codes and are similar to Department
of Defense criteria for high risk buildings and NRC criteria for various
applications.

o)  Discuss aspects of fire protection and fire hazards analysis and
    their relationship to nuclear facility design and the DSA.

The DOE and its contractors possess a variety of processes for analyzing
fire hazards at a facility, activity, and job. However, the outcome and
assumptions of these processes have not always been consistent for
similar types of hazards within the safety analysis and the fire hazard
analysis. Although the safety analysis and the fire hazard analysis are
driven by different DOE Orders and requirements, these analyses should
not be entirely independent and their preparation should be integrated
to ensure consistency of assumptions, consequences, design
considerations, and other controls. Under the DOE policy to implement an
integrated safety management system, identification of hazards must be
evaluated and agreed upon to ensure that the public, the workers, and
the environment are protected from adverse consequences. If integrated
safety management is not incorporated into the preparation of the safety
analysis and the fire hazard analysis, inconsistencies between
assumptions, consequences, design considerations, and controls may occur
that affect safety. In accordance with the integrated safety management
system approach for having a uniform and consistent process, a method
has been suggested by the DOE for determining when fire hazard analyses
and safety analyses are required.

p)  Identify and discuss the strengths and weaknesses of methods
    utilized to analyze the initiation and propagation of fires, and of
    their potential release of hazardous materials.

Fire hazards analyses (FHAs) must be developed for all nuclear
facilities, significant new facilities, and facilities that represent
unique or significant fire safety risks. The FHA shall be developed
using a graded approach. The conclusions of the FHA shall be
incorporated in the SAR Accident Analysis and shall be integrated into
design basis and beyond design basis accident conditions.

DOE Elements and contractors shall develop, implement, and maintain a
comprehensive fire protection program for facilities that includes

-   a reliable water supply of adequate capacity for fire suppression;

-   noncombustible or fire-resistive construction, where appropriate;

-   complete fire-rated barriers that are commensurate with the fire
    hazard to isolate hazardous occupancies and to minimize fire spread
    and loss potential consistent with defined limits as established by
    DOE;

-   automatic fire extinguishing systems throughout all significant
    facilities and in all areas subject to loss of safety-class systems,
    significant life-safety hazards, unacceptable program interruption,
    or fire loss potential in excess of defined limits;

-   redundant fire protection systems in areas where safety-class
    systems are vulnerable to fire damage and where no redundant safety
    capability exists outside of the fire area;

-   in new facilities, redundant safety-class systems in separate fire
    areas;

-   redundant fire protection systems provided in areas where the
    maximum possible fire loss (MPFL) exceeds limits established by DOE;

-   a means to summon the fire department in the event of a fire, such
    as a fire alarm signaling system;

-   a means to notify and evacuate building occupants in the event of a
    fire, such as a fire detection or fire alarm system and illuminated,
    protected egress paths;

-   physical access and appropriate equipment to facilitate effective
    intervention by the fire department, such as an interior standpipe
    system(s) in multi-story or large facilities with complex
    configurations;

-   a means to prevent the accidental release of significant quantities
    of contaminated products of combustion and fire fighting water to
    the environment, such as ventilation control and filter systems and
    curbs and dikes.

Such features would only be necessary if required by the FHA or SAR in
conjunction with other facility or site environmental protection
measures.

Fire and related hazards that are unique to DOE and are not addressed by
industry codes and standards shall be protected by isolation,
segregation, or use of special fire control systems, such as inert gas
or explosion suppression, as determined by the FHA. Fire protection
systems shall be designed such that their inadvertent operation,
inactivation, or failure of structural stability will not result in the
loss of vital safety functions or inoperability of safety-class systems
as determined by the SAR.

q)  Identify and discuss the methods used to determine the seismic
    hazard level to be used in design.

Seismic design or evaluation of PC-1 and PC-2 SSCs is based on model
building code seismic provisions. In these criteria, the current version
of the International Building Code shall be followed. All of the IBC
2000 seismic provisions shall be followed for PC-1 and PC-2 SSCs. Load
combinations to be used for PC-1 and PC-2 will be based on the
provisions in the IBC 2000. Use of site-specific data will be limited
per provisions of IBC 2000. Post-Northridge earthquake SAC
recommendations should be taken into consideration for steel structures.

For systems and components, seismic design forces are accounted as per
the IBC 2000 provisions. If a recent site-specific seismic hazard
assessment is available, it can be used subject to limitations imposed
in the IBC 2000. For evaluation of SSCs using site-specific hazard
analysis, the design shall be based on 5 percent critical damping as
recommended by the IBC 2000. Final earthquake loads are subject to
approval by DOE. A quality assurance program consistent with model
building code requirements shall be implemented for SSCs in Performance
Categories 1 and 2. In addition, peer review shall be conducted for
Performance Category 2 SSCs.

r)  Identify and discuss the methods used to assess the structural
    response of structures and determine whether safety systems may be
    expected to remain functional following an earthquake of postulated
    intensity.

For PC-2 and lower SSCs, the seismic design and evaluation criteria
employ the IBC 2000 seismic provisions with the exception that
site-specific information may be used if appropriate to define the
earthquake input excitation used to establish seismic loadings. The
maximum ground acceleration and ground response spectra determined in
the manner defined in IBC 2000 are used in the appropriate terms of the
IBC 2000 equation for base shear. Use of site-specific earthquake ground
motion data is considered to be preferable to the general seismic
zonation maps from the IBC 2000 with limitations imposed in the IBC 2000
Code. IBC provisions require a static or dynamic analysis approach in
which loadings are scaled to the base shear equation value. Elastically
computed seismic response is reduced by "R" values ranging from 1¼ to 8
as a means of accounting for inelastic energy absorption capability in
the IBC provisions and by these criteria for PC 2 and lower SSCs. This
reduced seismic response is combined with non-seismic concurrent loads
and then compared to code allowable response limits (or code ultimate
limits combined with code specified load factors). Normally, relative
seismic anchor motion (SAM) is not considered explicitly by model
building code seismic provisions. The International Building Code (IBC
2000) has been followed for PC-1 and PC-2 because it is the only the
current model building code meeting NEHRP provisions. The ICSSC has
concluded that the following seismic provisions are substantially
equivalent:

-   International Building Code, IBC 2000

-   1997 NEHRP Recommended Provisions

-   ASCE 7-98

The other model building codes may be followed provided site-specific
ground motion data is incorporated into the development of earthquake
loading similar to the manner described in this document for the IBC
2000. For PC-3 and PC-4 SSCs, the seismic design and evaluation criteria
specify that seismic evaluation be accomplished by dynamic analysis. The
recommended approach is to perform an elastic response spectrum dynamic
analysis to evaluate elastic seismic demand on SSCs. However, inelastic
energy absorption capability is recognized by permitting limited
inelastic behavior. By these provisions, inelastic energy absorption
capacity of structures is accounted for by the parameter, "*Fµ."*
Elastically computed seismic response is reduced by *Fµ* values ranging
from 1 to 3 as a means of accounting for inelastic energy absorption
capability for more hazardous facilities. *Fµ* values are much lower
than R values, increasing the risk reduction ratio. By these provisions,
only the element forces due to earthquake loading are reduced by *Fµ*.
The same *Fµ* values are specified for both PC-3 and PC-4. In order to
achieve different risk reduction ratios appropriate for the different
performance categories, the reduced seismic response is multiplied by a
seismic scale factor. Different seismic scale factors (SF) are specified
for PC-3 and PC-4. The resulting scaled inelastic seismic response is
combined with non-seismic concurrent loads and then compared to code
ultimate response limits. The design detailing provisions from the ACI
349, which provide ductility, toughness, and redundancy, are also
required such that SSCs can fully realize potential inelastic energy
absorption capability. Also, explicit consideration of relative SAM
effects is required for PC-3 and higher. For PC-3 or higher, the dynamic
analysis based deterministic seismic acceptance criteria specified
herein are independent of both the desired risk reduction ratio and the
performance category specified, other than for the seismic scale factor,
SF. Thus, the deterministic acceptance criteria herein may be used over
a wide range of applications including special situations where the
desired seismic performance goal differs.

For PC-2 and lower SSCs, see IBC provisions. For PC-3 and higher SSCs

-   perform dynamic analysis considering the mass and stiffness
    distribution of the structure;

-   perform an elastic response spectrum analysis, but to permit limited
    inelastic behavior. Elastic seismic response is reduced by the
    factor, *Fµ*, to obtain inelastic seismic demand. Explicitly account
    for relative displacement effects, where applicable.

s)  Identify the methods for evaluating the tolerance of structures and
    systems for natural phenomenon.

The methods for evaluating the tolerance of structures and systems are
presented in order of increasing rigor or decreasing conservatism.
Following the "graded approach" philosophy, the method to be used should
be selected on the basis of cost-benefit considerations. The end result
of this step will be a preliminary performance category list for the
NPH-related safety system components of the facility.

### Method I: System-by-System Categorization

In this method, the performance category of a given component of a
safety system will be the same as the highest performance category
applicable to any component of the system guidelines, irrespective of
the safety function of the particular component. Thus, all the
components of a system will have the same performance category. This is
the most conservative of the three methods and the only applicable
method if the safety function of only the entire system is known, and
information on the safety function of individual components of the
system is unavailable.

### Method II: Segment-by-Segment Categorization

A segment of a system is defined here as a group of components that
perform a safety function. It may also be defined as a group of
components belonging to a system that are physically located in a
certain area, room, or building, or whose functions are such that it may
be more cost-effective, or programmatically more desirable, to design or
procure them using the same NPH design criterion. In Method II, all
components belonging to a segment of an NPH-related safety system are
placed in the highest performance category of any component in that
segment.

### Method III: Component-by-Component Categorization

In this method, each component of the NPH-related safety system is
categorized individually. Thus, each component is categorized based on
its own function, and not lumped with the entire segment or system.
Method III is the most rigorous of the three methods, but the other two
methods are more conservative. Use of this method can minimize the
number of higher category components, and thereby reduce the components
that must be designed for higher level NPH design loads. However, in
order to accomplish this, detailed system safety and accident analyses
will be needed.

The data and information necessary to categorize these components may be
partially available in the facility hazard categorization documents or
facility safety and accident analysis reports. If the available data are
not adequate, additional data can be obtained cost-effectively through a
facility walkdown and by interviewing facility operations and
maintenance personnel. It may also be necessary to review design
documents and operation manuals and procedures.

Components, failures of which can cause these situations, can be
identified most effectively by a facility walkdown or by a careful
evaluation of building layout plans and drawings. A convenient and
efficient method of categorizing a large number of these types of SSCs
would be to group them on the basis of their NPH related characteristics
or locations.

Following the "graded approach" philosophy, the rigor with which such
failure potentials are to be determined should depend on the safety
significance and the preliminary performance category of the target, the
hazard category of the facility, and the relative costs of various
methods of determining target failure potentials. In the following
paragraphs, two methods of assessing target failure potentials are
presented in order of decreasing rigor.

### Systematic Analysis Method

Target failure potentials can be assessed using a qualitative but
systematic approach by constructing a fault-tree of the scenario. If
justifiable from cost-benefit considerations, this may be a desirable
method when necessary data for fault tree construction are available.
Generally, it should be used when the failure of the target is dependent
on a large and complex chain of events that may follow the failure of
the source, or to qualify a large system in its entirety.
Component-by-component application of this method is unlikely to be
cost-effective.

### Approximate Method

In this method, the effects of source failure on a target are modeled
approximately but rationally, considering possible scenarios identified
by a facility walkdown. Even though such models are approximate, their
evaluation provides qualitative information that is often adequate for
the purpose.

The consideration of adverse effects of system interaction of one
component or system (source) on another (target) is very important in
determining the performance categories of SSCs. Adverse interaction
effects can be different for different systems. Examples of typical
adverse interaction effects are listed below:

-   Structural failure and falling. Inadequately designed, inadequately
    anchored, and unanchored components may fail, slide, and/or topple
    and fall on or bump into other components that are not designed to
    withstand such interaction effects.

-   Proximity and impact. Adjacent components may impact each other,
    causing damage if the clearance between them is inadequate for
    NPH-induced deflections. Such adverse interaction may occur even if
    the deflection of the source is within its design limits.

-   Differential displacement. A target distribution system may span
    between different structural systems (source). Differential
    displacements between the structural systems may be within
    acceptable behavior limits for the structures, but may still affect
    the distribution system adversely.

-   Mechanical or electrical failure. The failure of a source mechanical
    or electrical component may impair the safety function of another
    component or system.

## Nuclear safety specialists shall demonstrate a working-level knowledge of the Technical Safety Requirements as described in 10 CFR 830.205, Technical Safety Requirements and DOE Guide 423.1-1, with respect to its impact on Department nuclear safety.

a)  Discuss the purpose of Technical Safety Requirements.

TSRs identify the limitations of each DOE owned, contractor operated
nuclear facility based on the documented safety analysis and any
additional safety requirements established for the facility.

t)  Describe the responsibilities of contractors authorized to operate
    defense nuclear facilities for Technical Safety Requirements.

Contractors, in the preparation of DSAs, identify how the safety
requirements of the safety management rule apply to a specific facility,
and describe how the contractor undertakes to design, build, and operate
the facility to be in conformance with the applicable statutes and DOE
rules and directives to ensure facility safety. The analysis of
operations and accidents defines the limits of safe operations,
identifies the required performance of safety-class and
safety-significant SSCs, and describes any ACs or procedures that are
necessary to meet the specific safety criteria for the facility.

u)  Define the following terms and discuss the purpose of each:

-   Safety limits

-   Operating limits

-   Limiting control settings

-   Limiting conditions for operation

-   Surveillance requirements

-   Administrative controls

### Safety Limits (SLs)

SLs are the limits on process variables associated with those
safety-class physical barriers, generally passive, that are necessary
for the intended facility functions and which are required to guard
against the uncontrolled release of radioactive materials.

### Operating Limits (OLs)

OLs are the limits required to ensure the safe operation of a nuclear
facility, including limiting control settings and limiting conditions
for operation.

### Limiting Control Settings (LCSs)

LCSs are boundaries on safety systems that regulate process variables to
prevent exceeding a safety limit.

### Limiting Conditions of Operation (LCOs)

LCOs are the limits that represent the lowest functional capability or
performance level of safety SSCs required for safe operations.

### Surveillance Requirements (SRs)

SRs are requirements relating to testing, calibration, or inspection to
ensure the necessary operability and quality of safety SSCs and their
support systems required for safe operations.

### Administrative Controls

ACs are provisions relating to organization and management procedures,
record keeping, assessment, and reporting necessary to ensure the safe
operation of a facility.

v)  Discuss the margin of safety in a TSR.

TSRs set forth the minimum acceptable limits for operation under normal
and specified failure conditions; they ensure that the available
equipment and initial conditions meet the assumptions in the accident
analysis. The bases for TSRs define the acceptance limits from which
margins of safety may be determined.

w)  Describe the general content of each of the following sections of
    the Technical Safety Requirements:

-   Use and application

-   Basis

-   Design features

### Use and Application

The use and application section of the TSR should contain basic
information and instructions for using and applying the TSR. The
following elements should be addressed under separate headings in this
section:

-   Definitions. Provide an alphabetical list of terms used throughout
    the TSR and their corresponding definitions. Include a note on the
    first page of the list stating that defined terms appear in
    uppercase type throughout the TSR.

-   Operational modes (reactors). Define the operational modes for
    reactor facilities and state that in the interest of uniformity, the
    operational conditions or modes listed are preferred and an attempt
    should be made to fit each reactor facility into this scheme.

-   Operational modes (nonreactor nuclear facilities). Define the
    operational modes for nonreactor facilities and state that in the
    interest of uniformity, the operational conditions and modes listed
    are preferred and an attempt should be made to fit each nonreactor
    nuclear facility into this scheme. If, however, a nonreactor nuclear
    facility cannot be made to fit, modes may be defined as needed,
    provided the definitions are clear and there are definite lines of
    demarcation between modes.

-   Frequency notation. The frequency notations, as used in the
    surveillances and elsewhere, should be defined when included in the
    TSR.

### Basis

The TSR basis provides summary statements of the reasons for the
selection of each specific SL, OL, and SR. The bases show how the
numerical values, conditions, surveillances, and action statements
fulfill the purpose derived from the safety documentation. Included in
the basis should also be a description of the safety functions that each
safety system provides and identification of what is included in each
safety system.

### Design Features

This section describes those design features that, if altered or
modified, would have a significant effect on the safe operation of the
facility. The important attributes of the passive design features that
are taken credit for in the accident analysis should be described
completely. These design features are normally passive characteristics
of the facility not subject to change by operations personnel, e.g.,
shielding, structural walls, relative locations of major components,
installed poisons, or special materials. Active safety features are
normally described in the DSA and are the subject of the various TSRs,
so they are not normally described in the design features section. All
changes or modifications that impact the safety basis of the facility
are subject to the USQ process.

x)  Discuss the definition and implementation principles for the term
    OPERABILITY as used in a Technical Safety Requirement.

Operability embodies the principle that a system, subsystem, train,
component, or device can perform its safety function(s) only if all
necessary support systems are capable of performing their related
support functions. A system or component can be degraded but still
operable if it remains capable of performing its required safety
function at the level assumed in the accident analysis. If systems,
components, or equipment are observed to be functioning, but under
stress, then judgment must be used concerning a declaration of
inoperability. General principles of operability should be followed in
generating LCOs:

-   General principle 1. A system is considered operable as long as
    there exists assurances that it is capable of performing its
    specified safety function(s).

-   General principle 2. A system can perform its specified safety
    function(s) only when all of its necessary support systems are
    capable of performing their related support functions.

-   General principle 3. When all systems designed to perform a certain
    safety function are not capable of performing that safety function,
    a loss of function condition exists.

-   General principle 4. When a system is determined to be incapable of
    performing its intended safety function(s), the declaration of
    inoperability should be immediate.

y)  Discuss the relationship of functional requirements and performance
    criteria to the Technical Safety Requirements.

TSRs define the performance requirements of SSCs and identify the safety
management programs used by personnel to ensure safety. TSRs are aimed
at confirming the ability of the SSCs and personnel to perform their
intended safety functions under normal, abnormal, and accident
conditions. These requirements are identified through hazard analysis of
the activities to be performed and identification of the potential
sources of safety issues. Safety analyses to identify and analyze a set
of bounding accidents that take into account all potential causes of
releases of radioactivity also contribute to development of TSRs.
Through analyses of the encompassing bounding accidents, the necessary
safety systems and accident mitigating systems are identified and their
characteristics are defined. Flowing from the analyses is information
that provides the bases for controls, limits, and conditions for
operation, known as TSRs. The content of the DSA must remain valid so
that the safety basis of the facility, as implemented in operations
through the TSR, remains valid. Therefore, there is a commitment to the
process of unreviewed safety questions regarding any proposed change to
the facility or its operations as described in the DSA. Likewise, all
changes to the TSR bases presented in the DSA should be incorporated
into the TSRs to ensure the information contained therein reflects the
current safety basis of the facility. Any proposed revision to a TSR
should be examined to ensure the basis for the change is supported in
the DSA. The TSR rule requires that such revisions be submitted to DOE
for review with the basis for the proposed change. The change to the TSR
must be approved by DOE before it is implemented.

z)  Discuss the conditions that constitute a violation of the Technical
    Safety Requirements and state the reporting requirements should a
    violation occur.

Although the TSR elements have an importance hierarchy, a TSR violation
can occur for each type of TSR. Violations of a TSR occur as a result of
the following four circumstances:

-   Exceeding an SL

-   Failure to complete an action statement within the required time
    limit following exceeding an LCS or failing to comply with an LCO

-   Failure to perform a surveillance within the required time limit

-   Failure to comply with an AC statement

Failure to comply with an AC statement is a TSR violation when either
the AC is directly violated, as would be the case with not meeting
minimum staffing requirements for example, or the intent of a referenced
program is not fulfilled. To qualify as a TSR violation, the failure to
meet the intent of the referenced program would need to be significant
enough to render the DSA summary invalid. TSR violations involving SLs
require the facility to begin immediately to go to the most stable, safe
condition attainable, including total shutdown. A grace period is
sometimes provided to perform a missed surveillance to provide time for
the performance of the missed surveillance, thereby avoiding the need
for a facility to take immediate, possibly unnecessary corrective
action. Entering the grace period remains a TSR violation even though an
immediate corrective action may not be required.

Reporting of all TSR violations should be made in accordance with the
provisions of DOE O 232.1A, Occurrence Reporting and Processing of
Operations Information. The reporting of violations on ACs can involve
judgment since the details of programs, like a program for criticality
control, do not appear directly as a TSR, and some program requirements
are more important than others. Violations of controls identified in the
accident or criticality scenarios in the DSA should be reported as if
they were TSR violations. To ensure consideration for mitigation in
potential enforcement actions, identified TSR violations should be
evaluated for voluntary reporting to the DOE Noncompliance Tracking
System.

a)  Discuss the requirements for administrative control of the Technical
    Safety Requirements.

ACs are the provisions relating to organization and management,
procedures, record keeping, reviews, and audits necessary to ensure safe
operation of the facility. ACs may include reporting deviations from
TSRs, staffing requirements for facility positions important to safe
operation of the facility, ACs of the criticality safety program, and
commitments to safety management programs important to worker safety. In
general, the ACs should document all those administrative functions that
are required to meet facility safety criteria as identified in the DSA,
including commitments to safety management programs. It is expected that
the ACs will be tailored to the facility activities and the hazards
identified in the DSA. This tailoring should be a direct result of the
DSA, but it may also result from institutional requirements that address
many facilities. As a general practice, safety controls for individual
accident scenarios based on engineered SSCs are preferred to ACs because
they are usually more reliable and more predictable. The tendency to use
ACs as an expedient alternative to an LCO or LCS should be avoided when
possible. Efforts should be made to use engineered SSCs whenever
possible for controlling the likelihood and consequences of accidents.
ACs should be considered for defense in depth rather than the primary or
redundant controls. While ACs may be acceptable for ensuring safe
operation, their generally lower reliability, compared with engineered
controls, should be evaluated carefully when choosing safety measures
for long-term hazardous activities. Human actions, taken either in
response to an event or taken proactively to establish desired
conditions, are subject to errors of omission or commission. Sets of ACs
are prone to common cause failure. The following attributes, which can
be tailored as appropriate, can increase reliability:

-   Use of reader/worker/checker systems

-   Independent verification

-   Positive feedback systems

-   Human factor analysis

-   Operator training and certification

-   Continuing training and requalification

-   Abnormal event response drills

-   Ergonomic considerations in procedures

When invoking ACs for control of accident scenarios, the preceding
attributes, appropriate to the consequences of the accidents they are
intended to prevent, should be considered and also invoked.

b)  Discuss the possible source documents that may be used in developing
    technical safety requirements. Discuss the role of documented safety
    analyses in selecting technical safety requirements and the
    respective flowdown.

The primary source document is the DSA for that facility. Other inputs
come from existing technical safety or operational safety documents that
may exist for the facility. Also, DOE-STD-3009-94, Preparation Guide for
U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis
Reports, along with DOE G 423.1-1 Implementation Guide for Use in
Developing Technical Safety Requirements, may be used in developing
TSRs.

The documented safety analysis identifies hazards that need to be
analyzed. This hazard analysis creates accident scenarios that identify
accident conditions and risks. Controls are developed to mitigate
against the derived accidents. These controls are what the TSRs identify
as important to safety. The safety-significant and safety-class controls
have prescribed TSRs that are used to ensure they are able to meet their
given requirements as part of the safety envelope.

c)  Differentiate between the following facility designations:

-   Category A reactor facility

-   Category B reactor facility

### Category A Reactor Facility

Category A reactor facilities are production, test, and research
reactors designated by DOE based on power level (e.g., design thermal
power rating of 20 megawatts steady state and higher), potential fission
product inventory, and experimental capability.

### Category B Reactor Facilities

Category B reactor facilities are test and research reactors designated
by DOE based on power level (e.g., design thermal power rating of less
than 20 megawatts steady state), potential fission product inventory,
and experimental capability.

d)  Discuss the requirements for emergency actions that depart from the
    approved Technical Safety Requirements.

In an emergency, if a situation develops that is not addressed by the
TSR, site personnel are expected to use their training and expertise to
take actions to correct or mitigate the situation. Also, site personnel
may take actions that depart from the requirements of a TSR provided (a)
an emergency situation exists, (b) these actions are needed immediately
to protect workers, the public, or the environment from imminent and
significant harm, and (c) no action consistent with the TSR is
immediately apparent. Such action must be approved by a certified
operator for reactor facilities or by a person in authority as
designated in the TSRs for nonreactor nuclear facilities. If emergency
action is taken, both a verbal notification should be made to the
responsible head of the field element and a written report made to the
cognizant Secretarial Officer (CSO) within 24 hours.

e)  Discuss the provisions a contractor may follow to develop
    alternatives to Technical Safety Requirements for environmental
    restoration activities.

A contractor for an environmental restoration activity may follow the
provisions of 29 CFR 1910.120 or 1926.65 for construction activities to
develop the appropriate hazard controls (rather than TSR guidance)
provided the activity involves either (1) work not done within a
permanent structure, or (2) the decommissioning of a facility with only
low-level residual fixed radioactivity. Implicit in this guidance is an
understanding that reasonable efforts to remove radioactive systems,
components, and stored materials has been completed and that the work
does not prudently require the use of active safety systems or
components designed to prevent or mitigate the accidental release of
hazardous radioactive materials. DOE-STD-1120-98 also provides guidance
that should be considered in the development of alternatives to TSRs.

f)  Discuss the requirements for the contractor to maintain the
    Technical Safety Requirements current.

As with all hazard controls, TSRs must be kept current and reflect
changes in the facility, the work, and the hazards as they are analyzed
in the documented safety analysis. In addition, DOE expects a contractor
to maintain TSRs, and other hazard controls as appropriate, as
controlled documents with an authorized users list.

g)  Discuss the application of the graded approach relative to Technical
    Safety Requirements.

The graded approach is not directly applicable to the TSRs required by
10 CFR 830.205. However, the graded approach is specified for DSAs
required by 10 CFR 830.204. Thus, the level of detail in the DSA, and
the number of safety parameters identified in the DSA section deriving
the TSRs, will have a direct effect on the number and type of resulting
TSRs.

h)  Perform a review of a safety-class or safety-significant SSC
    including walking down the associated surveillance requirements and
    LCO/LCS.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working-level knowledge of 10 CFR 830.203, Unreviewed Safety Question Process and DOE Guide 423.1-1, with respect to its impact on Department nuclear safety.

a)  Discuss the purpose of the Unreviewed Safety Question process.

The unreviewed safety question (USQ) process allows contractors to make
physical and procedural changes and to conduct tests and experiments
without prior DOE approval if the proposed change can be accommodated
within the existing safety basis.

i)  Discuss the reasons for performing an Unreviewed Safety Question
    determination.

A USQ is expected to be implemented using contractor procedures that
ensure that proposed changes to physical characteristics or operating
procedures are adequately evaluated relative to the approved safety
basis, and that those proposed changes determined to involve USQs are
brought to the attention of DOE for review and approval before changes
are made. A proposed change or test involves a USQ if

-   the probability or consequences of an accident or malfunction of
    equipment important to safety could be increased

-   the possibility of a different type of accident than previously
    evaluated in the DSA could be introduced

-   margins of safety could be reduced

j)  Define the following terms:

-   Discrepant as-found condition

-   Potential inadequacy in the safety analysis

-   Proposed change

### Discrepant As-Found Condition

A discrepant as-found condition is a situation where the actual physical
configuration of equipment important to safety in the facility does not
match the DSA. This may result from an error in the DSA or an error in
the facility configuration.

### Potential Inadequacy in the Safety Analysis (PISA)

A PISA is a condition in which the safety basis may be inadequate or the
physical condition may not be accurate because the safety analysis may
not match the current physical configuration of the facility or the
safety analysis may be inappropriate or may contain errors.

### Proposed Change

A proposed change is a suggested modification to the configuration of a
facility.

k)  Define the conditions for an Unreviewed Safety Question.

The conditions that require a USQ are listed below:

-   Temporary or permanent changes in the facility

-   Temporary or permanent changes in the procedures

-   Tests or experiments not described in the existing DSAs

-   Discovery of potential inadequacies in the existing safety analyses

l)  Describe the responsibilities of contractors authorized to operate
    defense nuclear facilities for the performance of safety
    evaluations.

The contractor responsible for a hazard category 1, 2, or 3 DOE nuclear
facility must obtain approval from DOE for the methodology used to
prepare the documented safety analysis for the facility unless the
contractor uses a methodology approved by DOE.

The documented safety analysis for a hazard category 1, 2, or 3 DOE
nuclear facility must, as appropriate for the complexities and hazards
associated with the facility

-   describe the facility (including the design of safety SSCs) and the
    work to be performed;

-   provide a systematic identification of both natural and man-made
    hazards associated with the facility;

-   evaluate normal, abnormal, and accident conditions, including
    consideration of natural and man-made external events,
    identification of energy sources or processes that might contribute
    to the generation or uncontrolled release of radioactive and other
    hazardous materials, and consideration of the need for analysis of
    accidents which may be beyond the design basis of the facility;

-   derive the hazard controls necessary to ensure adequate protection
    of workers, the public, and the environment; demonstrate the
    adequacy of these controls to eliminate, limit, or mitigate
    identified hazards; and define the process for both maintaining the
    hazard controls current at all times and controlling their use;

-   define the characteristics of the safety management programs
    necessary to ensure the safe operation of the facility, including
    (where applicable) quality assurance, procedures, maintenance,
    personnel training, conduct of operations, emergency preparedness,
    fire protection, waste management, and radiation protection.

With respect to a nonreactor nuclear facility with fissionable material
in a form and amount sufficient to pose a potential for criticality, the
DSA must define a criticality safety program that

-   ensures that operations with fissionable material remain subcritical
    under all normal and credible abnormal conditions

-   identifies applicable nuclear criticality safety standards

-   describes how the program meets applicable nuclear criticality
    safety standards

m)  Describe the actions to be taken by a contractor upon identifying
    information that indicates a potential inadequacy of safety analyses
    or a possible reduction in the margin of safety as defined in the
    Technical Safety Requirements.

Because an inadequacy in the safety analyses has the potential to call
into question information relied on for authorization of operations, DOE
requires the contractor to do the following:

-   Take appropriate action to place or maintain the facility in a safe
    condition.

-   Expeditiously notify DOE when the information is discovered.

-   Perform a USQ determination and submit the results promptly.

-   Complete an evaluation of the safety of the situation and submit it
    to DOE before removing any operational restrictions implemented to
    compensate for the analytical discrepancy.

The third bulleted item, above, means that the USQ determination should
be prepared promptly and the results submitted promptly. This is also
intended to mean that the time frame after initial notification of DOE
until submittal of the USQ determination results should be on the order
of hours or days, not weeks or months. If a USQ is determined to be
present, the evaluation of the safety of the situation will require not
only DOE's review but also its approval of resulting changes before any
operational restrictions are removed.

n)  Discuss the qualification and training requirements for personnel
    who perform safety evaluations.

Implementing procedures should establish the personnel training and
qualifications needed to perform the USQ process. These include required
educational background, years and/or types of work experience, knowledge
of the facility, understanding of DOE requirements related to the
facility safety basis (including the USQ process), and familiarity with
the facility-specific safety basis. All personnel responsible for
preparing, reviewing, or approving USQ documents should receive training
on the application of Section 830.203, including any facility-specific
procedures. The recommended interval for retraining is every two years.
The contractor should maintain a list of those personnel who are
currently qualified to perform the USQ process.

o)  Discuss the actions to be taken if it is determined that a Potential
    Inadequacy in the Safety Analysis is involved.

If the safety analysis is found to have a potential inadequacy, the
following actions must be taken:

-   Take appropriate action to place or maintain the facility in a safe
    condition.

-   Expeditiously notify DOE when the information is discovered.

-   Perform a USQ determination and submit the results promptly.

-   Complete an evaluation of the safety of the situation and submit it
    to DOE before removing any operational restrictions implemented to
    compensate for the analytical discrepancy.

p)  Discuss the following terms as they apply to unreviewed safety
    questions:

-   Categorical exclusions

-   Prior unreviewed safety question determinations

-   Inconsequential changes

-   Margin of safety

-   Design/evaluation basis accidents

-   Important to safety

-   Safety basis

-   Restoration modification

-   Evaluation of safety

-   Unreviewed safety question

-   Justification for continued operations

### Categorical Exclusions

A categorical exclusion is valid when answers to the seven questions
listed in section 3.3 of DOE G 424.1-1A, Implementation Guide for Use in
Addressing Unreviewed Safety Question Requirements, would be "no" for
every credible variation within the category.

### Prior Unreviewed Safety Question Determinations

Prior USQ determinations are USQs performed on facilities that were not
previously part of the review process.

### Inconsequential Changes

Inconsequential changes are changes made to a facility or procedure that
do not affect the safety basis.

### Margin of Safety

A margin of safety is defined by the range between two conditions. The
first is the most adverse condition estimated or calculated in safety
analyses to occur from an operational upset or family of related upsets.
The second condition is the worst-case value known to be safe, from an
engineering perspective. This value would be expected to be related to
the condition at which some accident prevention or mitigation action is
taken in response to the upset or accident, not the actual predicted
failure point of some component.

### Design/Evaluation Basis Accidents

Design/evaluation basis accidents are postulated accidents that evaluate
normal, abnormal, and accident conditions that might be associated with
the generation or release of radioactive or other hazardous materials.

### Important to Safety

Important to safety, with reference to SSCs, means those engineered
features whose function is to prevent or mitigate category 2 event
sequences.

### Safety Basis

The safety basis is comprised of the documented safety analysis and
hazard controls that provide reasonable assurance that a DOE nuclear
facility can be operated safely in a manner that adequately protects
workers, the public, and the environment.

### Restoration Modification

Restoration modification refers to the restoration or modification of
any equipment that may affect the safety basis.

### Evaluation of Safety

Evaluation of safety is the review of conditions that affect the safety
of a system, facility, or component, and includes an evaluation of all
components important to safety.

### Unreviewed Safety Question

A USQ exists in a situation where the following criteria are present:

-   The probability of the occurrence or the consequences of an accident
    or the malfunction of equipment important to safety previously
    evaluated in the documented safety analysis could be increased.

-   The possibility of an accident or a malfunction of a different type
    than any evaluated previously in the documented safety analysis
    could be created.

-   A margin of safety could be reduced.

-   The documented safety analysis may not be bounding or may be
    otherwise inadequate.

### Justification for Continued Operations

The justification for continued operations is the proposal to continue
operations at a nuclear facility during long-term modifications or
during decommissioning operations.

q)  Discuss the responsibilities of the contractor associated with
    Unreviewed Safety Question summaries and the USQ procedure.

Contractors should develop procedures that provide detailed guidance for
the performance of the USQ process, including procedures for screening
and for developing the USQ determinations. The procedures should:

-   define the purpose of the process;

-   set forth the process applicability;

-   provide definitions of appropriate terms, include screening
    criteria, as appropriate, and the basis for their application;

-   include detailed guidance on what must be considered and evaluated
    when performing or reviewing a USQ determination;

-   define the required qualifications and responsibilities of personnel
    performing and reviewing USQ determinations;

-   include documentation requirements for each USQ determination.

r)  Describe DOE's responsibilities when not agreeing with a negative
    determination.

    DOE can declare that a USQ exists as part of its oversight
    responsibility of the USQ process. Such a declaration might result
    from a disagreement with a contractor's negative USQ determination
    or might result from a condition for which the contractor has not
    done a USQ determination. When DOE declares a USQ, it is because DOE
    believes it should be involved in the review and approval of the
    situation to fulfill its responsibilities.

s)  Discuss why the application of the graded approach does not apply to
    the USQ process.

The graded approach is applied to the USQ process only indirectly.
During the hazards analysis and safety analysis processes, equipment is
classified as safety-class, safety-significant, "other" safety, or
non-safety. Once these classifications are established, they are used to
indicate how much effort should be applied. However, no steps of the USQ
process can be eliminated based on such grading. In some situations,
attempts to apply the graded approach to the USQ process have resulted
in inadequate USQ determinations. The graded approach may give a rough
indication of how much justification or basis information should be
provided when explaining the answers to each of the seven USQ
determination criteria. More elaborate and thorough basis information
would be expected for changes to safety equipment than for non-safety
equipment. In any case, the justification for the answers to the USQ
determination criteria needs to be defensible.

t)  Review a USQ Determination including walking down the proposed
    change/potential inadequacy.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a familiarity level knowledge of the functional interfaces between safety system software components and the system level design.

a)  Identify how system-level requirements are established and then
    assigned to hardware, software, and human components of a digital
    instrumentation and control system.

A review of Software Requirements Engineering, edited by Thayer &
Dorfman, IEEE Computer Society Press, 1997, should be conducted to
determine requirements per DOE N 203.1.

u)  Identify the typical requirements that define functional interfaces
    between safety system software components and the system-level
    design, as described in standards such as ANSI/IEEE 830, IEEE Guide
    to Software Requirements Specifications, and IEEE 7-4.3.2, Standard
    Criteria for Digital Computers in Safety Systems of Nuclear Power
    Generating Stations. Identify where this information is documented.

A review of ANSI/IEEE 830 and IEEE 7-4.3.2 is required to compare how
they integrate with DOE N 203.1

v)  Identify the specific records that must be maintained and the
    requirements for maintaining these records to document the
    development of safety system software.

The records management system should have schedules for records
retention and disposition in accordance with the requirements of DOE O
200.1, Information Management Program.

w)  Review a development project for safety system software. Explain how
    the functional interfaces between components and the system-level
    design were established and controlled.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the relationships between the problems being addressed by safety analysis and design codes, the design requirements for the codes, and the components of the codes.

a)  Identify how functional requirements and applicability of safety
    analysis and design computer codes are defined, documented, and
    controlled relative to modeling and data assumptions, design
    constraints, sizing and timing conditions, and input/output
    parameters.

Software used for safety analysis should come from the approved toolbox
of software or DOE must approve a new software type if used. The
software must meet the criteria of DOE G 200.1-1, Software Engineering
Methodology.

x)  Review a development project for safety analysis or design software.
    Explain how the problem being addressed by the software was
    translated into functional requirements, how the requirements were
    established and controlled, and how the code was reconciled with the
    original problem.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

y)  Discuss the DOE toolbox codes (reference
    http://tis.eh.doe.gov/techstds/toolbox_codes.html), their strengths,
    weaknesses and other factors governing their appropriate use and the
    applicable DOE standards and guides for modeling their phenomena.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of Department of Energy (DOE) Policy 450.4, Safety Management System Policy, and Policy 450.5, Line Environment, Safety and Health Oversight, as applied to nuclear safety.

a)  Discuss the fundamentals of Integrated Safety Management and their
    direct application to nuclear safety.

Following are the six components of the DOE safety management system:

-   Objective

-   Guiding Principles

-   Core Functions

-   Mechanisms

-   Responsibilities

-   Implementation

The purpose of DOE Policy 450.5 is to set forth the Department's
expectations for DOE line management environment, safety, and health
(ES&H) oversight and for the use of contractor self-assessment programs
as the cornerstone for this oversight. An effective and efficient
oversight program can be realized when a vigorous contractor
self-assessment program is in place, similar to those used in successful
companies. DOE line oversight and contractor self-assessments together
ensure that field elements and contractors are adequately implementing
the DOE Safety Management System. As a complement to DOE line oversight,
the Department's Office of Environment, Safety, and Health (EH) provides
safety policy, enforcement, and independent internal oversight
functions.

z)  Describe the key elements of an effective contractor self assessment
    nuclear safety program.

A robust, rigorous, and credible contractor ES&H self-assessment program
linked to the DOE safety management system is in place, which includes
elements that address the following:

-   Performance measures and performance indicators

-   Line and independent evaluations

-   Compliance with applicable requirements (rules, regulatory
    standards, contract terms)

-   Data collection, analysis, and corrective actions

-   Continuous feedback and performance improvement

    The results and conclusions of the contractor self-assessments are
    available to DOE.

a)  Discuss the following nuclear safety assessments/surveillance
    activities:

-   Determination of assessment/surveillance requirements

-   Operation/area/site office and contractor notification

-   Assessment/surveillance agenda

### Determination of Assessment/Surveillance Requirements

A determination of assessment/surveillance requirements is a periodic,
value-added appraisal of sufficient frequency and duration to confirm
the contractor's safe performance of work and the effectiveness of the
self-assessment program. A cost-effective appraisal meeting the intent
of this policy might need to be no more than 2 weeks in duration and no
more than once a year at each site.

### Operation/Area/Site Office and Contractor Notification

DOE uses the analysis of contractor self assessment results, performance
measures, and operational awareness as input to scoping the annual
appraisal. Appraisals by non-line organizations, such as EH, or external
organizations, such as the Environmental Protection Agency and state
agencies, are fully considered and not ordinarily duplicated. The
appraisals are conducted primarily by DOE employees. Issues identified
but unresolved during a periodic appraisal are referred to local DOE
personnel for further examination.

### Assessment/Surveillance Agenda

The scope of periodic appraisals, including additional areas of review,
is determined by field elements with input from Headquarters and the
contractor. DOE uses the analysis of contractor self assessment results,
performance measures, and operational awareness, as input to scoping the
annual appraisal.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the following criticality safety-related American National Standards Institute/American Nuclear Society (ANSI/ANS) standards:

-   ANSI/ANS-8.1, Nuclear Criticality Safety in Operations with
    Fissionable Materials Outside Reactors

-   ANS-8.3 (ANSI N-16.2), Criticality Accident Alarm System

-   ANS-8.5 (ANSI N-16.4), Use of Borosilicate-Glass Raschig Rings as a
    Neutron Absorber in Solutions of Fissile Material

-   ANSI/ANS-8.7, Guide for Nuclear Criticality Safety in the Storage of
    Fissile Materials

-   ANS-8.15, Nuclear Criticality Control of Special Actinide Elements

-   ANS-8.19, Administrative Practices for Nuclear Criticality Safety

a)  Describe the contents, requirements, and relationship among the
    above American National Standards Institute/American Nuclear Society
    Standards.

A review of applicable ANSI and ANS standards should be completed to
fully comprehend all requirements.

b)  Discuss the applicability of the above American National Standards
    Institute/‌American Nuclear Society Standards to the Department
    facilities and processes.

A review of the ANSI and ANS standards along with a review of DOE O
420.1A, DOE-STD-3009, DOE-STD-1027, and DOE-STD-3007 is required to
determine the applicability of the standards to Department facilities
and processes.

c)  Discuss the role of the Department nuclear safety specialists in
    implementing the requirements of these Standards.

The nuclear safety specialist provides technical oversight and
independent review of the facility implementing the above mentioned
standards.

d)  Define the following terms associated with nuclear criticality
    safety:

-   Criticality incident

-   Double contingency principle

-   Geometry control

-   Nuclear criticality safety

-   Significant quantity of fissionable material

-   Temporary exemption

### Criticality Incident

Criticality incident refers to the release of energy as a result of
accidentally producing a self sustaining or divergent nuclear fission
chain reaction.

### Double Contingency Principle

The double contingency principle requires that process designs shall
incorporate sufficient factors of safety to require at least two
unlikely, independent, and concurrent changes in process conditions
before a criticality accident is possible. In all cases, no single
failure shall result in the potential for a criticality accident.

### Geometry Control

Geometry control refers to the passive engineering control by which
container equipment design limits fissionable material dimensions and
the spacing between adjacent equipment. Reliance on geometry control
shall be the first priority.

### Nuclear Criticality Safety

Nuclear criticality safety is the prevention or termination of
inadvertent nuclear criticality and protection against injury or damage
due to an accidental nuclear criticality. Generally, prevention is
preferred.

### Significant Quantity of Fissionable Material

Significant quantity of fissionable material refers to the minimum mass
of fissionable material for which control of at least one parameter is
required to ensure subcriticality under all normal and credible abnormal
conditions.

### Temporary Exemption

A temporary exemption is a short term release from a criticality safety
requirement.

e)  Discuss the Management and Operating (M&O) Contractor
    responsibilities for the following in relation to criticality safety
    activities:

-   Criticality safety evaluations

-   Monitoring

-   Surveillance

-   Transportation

-   Storage

### Criticality Safety Evaluations

M&O contractors responsible for nuclear facilities are required to
perform and document detailed nuclear criticality safety analyses. The
facility SAR shall contain a criticality safety section including the
following:

-   Detailed descriptions of all equipment and facilities in which a
    criticality hazard may exist

-   Descriptions of all the fissionable materials and processes to be
    used including maximum quantities, and collection, handling, and
    transportation procedures

-   Analysis of all criticality scenarios and their safety control
    parameters

-   Descriptions of all criticality alarm systems (CAS) and emergency
    procedures

### Monitoring

M&O contractors shall establish monitoring programs to prevent
accumulation of fissionable materials in process and storage equipment,
pipes, and ventilation systems.

### Surveillance

M&O contractors must perform reviews and implement corrective measures
that shall be in place in order to avoid any possible criticality
hazards.

### Transportation

Requirements for transportation in DOE Order 5480.24 shall be applicable
when materials are transferred from one operation to another within a
facility, and from one on-site location to another. Contractors must
follow an approved on-site transportation safety manual.

### Storage

Requirements in DOE O 461.1 shall be complied with regarding off-site
shipment of fissionable material. DOE O 460.1 shall apply for the
storage and safe transportation of weapon components and special
assemblies shipped in national defense.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the following Department of Energy (DOE) Orders, Technical Standards, and Notice:

-   DOE-STD-3011, Guidance for Preparation of Basis for Interim
    Operation (BIO) Documents

-   DOE-STD-3014, Accident Analysis for Aircraft Crash into Hazardous
    Facilities

-   US NRC Guide 1.70, Standard Format and Content of Safety Analysis
    Reports of Nuclear Power Plants

-   29 CFR 1910.120, Safety and Health Programs, Work Plans, Health and
    Safety Plan

-   DOE-STD-1163, Integration of Multiple Hazard Analysis Requirements
    and Activities

-   DOE-STD-3016, Hazards Analysis Reports for Nuclear Explosive
    Operations

-   DOE O 460.1B, Packaging and Transportation Safety

-   DOE Guide 460.1-1, Implementation Guide for Use with DOE Order
    460.1A,

-   Packaging and Transportation Safety

-   DOE O 461.1, Packaging and Transportation of Materials of National
    Security Interest

-   DOE Manual 461.1-1, Packaging and Transfer of Materials of National
    Security Interest Manual

-   Secretary of Energy Notice (SEN) SEN-35-91, Nuclear Safety Policy

a)  Describe the contents, requirements, and relationship between the
    above technical standards, and Secretary of Energy notice.

It is DOE policy that the general public be protected such that no
individual bears significant additional risk to health and safety from
the operation of a DOE nuclear facility above the risks to which members
of the general population are normally exposed. The purpose of the
documents listed above is to establish the basic nuclear safety policy
from which specific safety rules, orders, standards, and other
requirements shall follow. DOE facilities will be designed, constructed,
operated, and decommissioned to assure the protection of the public,
workers, and the environment.

The scientific, engineering, and technical knowledge and ability of DOE
and its contractor personnel directly determine the Department's ability
to achieve its safety policy. Potential hazards associated with nuclear
operations dictate that DOE and contractor personnel possess technical
competence, commitment, discipline, and high standards of professional
and personal excellence. Organizations responsible for DOE nuclear
activities shall have the ability to recruit, train, and qualify
personnel who possess these attributes for job specific positions.

Adherence to appropriate national and international standards in the
design, construction, operation, and decommissioning of DOE's nuclear
facilities and activities is necessary for the successful implementation
of the Department's nuclear safety policy. Standards proven through
years of experience and accepted by professional and technical societies
shall be used wherever applicable. Yet no matter how carefully conceived
and properly developed, technical standards cannot address all
eventualities.

Therefore, DOE contractors must critically assess the standards in use
at DOE facilities to assure that they remain consistent with the latest
information arising from operational experience and developments in
science and engineering. Where standards do not exist or where existing
standards do not suffice, appropriate DOE nuclear safety standards shall
be developed and adopted.

f)  Describe the role of nuclear safety specialists with respect to the
    requirements in these orders, standards, and Secretary of Energy
    notice.

In striving to reach safety goals, DOE nuclear facilities and activities
shall be designed, constructed, operated, and decommissioned with a)
appropriate barriers to prevent or minimize potential radioactive
releases, b) engineered safety features to minimize potential releases,
and c) procedural controls to mitigate the effects of potential
releases. These goals shall be addressed for both new and existing
facilities. Proposed modifications to existing facilities to achieve
these goals shall be prioritized along with other proposed modifications
based on their safety significance. DOE shall pursue the evolution of
additional potential safety goals for plant and co-located workers to
support enhanced safe operations of its facilities. The Nuclear Safety
Specialist ensures that all safety considerations taken are adequate to
meet DOE goals.

g)  Determine whether aircraft crashes pose an acceptable or
    unacceptable hazard to safety of nuclear facilities.

    Note: This is a performance-based competency. The ability to
    complete this competency must be observed by a qualifying official.

h)  Discuss the phenomena of aircraft crashes as a mechanism for
    releasing toxic materials.

For information on this topic, review section D.2 of
<http://www.eh.doe.gov/techstds/standard/std3014/std3014.pdf>.

i)  Discuss the phenomena to which packaging is designed to withstand
    transportation accidents and the relationship to accident severity.

Packaging is designed to withstand the conditions listed below.

*Heat*

Packaging must withstand an ambient temperature of 38°C (100°F) in still
air, and insulation according to the following table:

**Insulation Data**

  -------------- ----------------------------- -------------------------------------
  **Form and                                   **Total insulation for a 12-hour
  location of                                  period (g cal/cm^2^)**
  surface**                                    

  Flat surfaces                                
  transported                                  
  horizontally                                 

                 Base                          None

                 Other surfaces                800

  Flat surfaces                                200
  not                                          
  transported                                  
  horizontally                                 

  Curved                                       400
  surfaces                                     
  -------------- ----------------------------- -------------------------------------

### 

### Cold

Packaging must withstand an ambient temperature of -40°C (-40°F) in
still air and shade.

### Reduced External Pressure

Packaging must withstand a reduced external pressure of 25 kPa (3.5
lbf/in^2^) absolute.

### Increased External Pressure

Packaging must withstand an increased external pressure of 140 kPa (20
lbf/in^2^) absolute.

### Vibration

Packaging must withstand vibration normally incident to transport.

### Water Spray

Packaging must withstand a water spray that simulates exposure to
rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour.

### Free Drop

Packaging must withstand between 1.5 and 2.5 hours after the conclusion
of the water spray test, a free drop through the distance specified
below onto a flat, essentially unyielding, horizontal surface, striking
the surface in a position for which maximum damage is expected.

**Criteria for Free Drop Test (Weight/Distance)**

  ---------------------- ------------------------ ------------- ------------
  **Package weight**                              **Free drop   
                                                  distance**    

  **Kilograms**          **(Pounds)**             **Meters**    **(Feet)**

  Less than 5,000        (Less than 11,000)       1.2           \(4\)

  5,000 to 10,000        (11,000 to 22,000)       0.9           \(3\)

  10,000 to 15,000       (22,000 to 33,100)       0.6           \(2\)

  More than 15,000       (More than 33,100)       0.3           \(1\)
  ---------------------- ------------------------ ------------- ------------

### Corner Drop

Packaging must withstand a free drop onto each corner of the package in
succession, or in the case of a cylindrical package onto each quarter of
each rim, from a height of 0.3 m (1 ft) onto a flat, essentially
unyielding, horizontal surface. This test applies only to fiberboard,
wood, or fissile material rectangular packages not exceeding 50 kg (110
lbs) and fiberboard, wood, or fissile material cylindrical packages not
exceeding 100 kg (220 lbs).

### Compression

For packages weighing up to 5000 kg (11,000 lbs), the package must be
subjected, for a period of 24 hours, to a compressive load applied
uniformly to the top and bottom of the package in the position in which
the package would normally be transported. The compressive load must be
the greater of the following:

-   the equivalent of 5 times the weight of the package

-   the equivalent of 13 kPa (2 lbf/in^2^) multiplied by the vertically
    projected area of the package

### Penetration

Packaging must withstand the impact of the hemispherical end of a
vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs)
mass, dropped from a height of 1 m (40 in) onto the exposed surface of
the package that is expected to be most vulnerable to puncture. The long
axis of the cylinder must be perpendicular to the package surface.

Tests on packaging for hypothetical accident severity must be conducted
as detailed below.

### Free Drop

Packaging must withstand a free drop through a distance of 9 m (30 ft)
onto a flat, essentially unyielding, horizontal surface, striking the
surface in a position for which maximum damage is expected.

### Crush

Packaging must withstand subjection to a dynamic crush test by
positioning the specimen on a flat, essentially unyielding horizontal
surface so as to suffer maximum damage by the drop of a 500-kg (1100-lb)
mass from 9 m (30 ft) onto the specimen. The mass must consist of a
solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a
horizontal attitude. The crush test is required only when the specimen
has a mass not greater than 500 kg (1100 lb), an overall density not
greater than 1000 kg/m^3^ (62.4 lb/ft^3^) based on external dimension,
and radioactive contents greater than 1000 A~2~ not as special form
radioactive material. For packages containing fissile material, the
radioactive contents greater than 1000 A~2~ criterion does not apply.

### Puncture

Packaging must withstand a free drop through a distance of 1 m (40 in)
in a position for which maximum damage is expected, onto the upper end
of a solid, vertical, cylindrical, mild steel bar mounted on an
essentially unyielding, horizontal surface. The bar must be 15 cm (6 in)
in diameter, with the top horizontal and its edge rounded to a radius of
not more than 6 mm (0.25 in), and of a length as to cause maximum damage
to the package, but not less than 20 cm (8 in) long. The long axis of
the bar must be vertical.

### Thermal

Packaging must withstand being fully engulfed, except for a simple
support system, in a hydrocarbon fuel/air fire of sufficient extent, and
in sufficiently quiescent ambient conditions, to provide an average
emissivity coefficient of at least 0.9, with an average flame
temperature of at least 800°C (1475°F) for a period of 30 minutes, or
any other thermal test that provides the equivalent total heat input to
the package and which provides a time averaged environmental temperature
of 800°C. The fuel source must extend horizontally at least 1 m (40 in),
but may not extend more than 3 m (10 ft), beyond any external surface of
the specimen, and the specimen must be positioned 1 m (40 in) above the
surface of the fuel source. For purposes of calculation, the surface
absorptivity coefficient must be either that value which the package may
be expected to possess if exposed to the fire specified or 0.8,
whichever is greater; and the convective coefficient must be that value
which may be demonstrated to exist if the package were exposed to the
fire specified. Artificial cooling may not be applied after cessation of
external heat input, and any combustion of materials of construction,
must be allowed to proceed until it terminates naturally.

### Immersion of Fissile Material

For fissile material subject to 10 CFR 71.55, in those cases where water
inleakage has not been assumed for criticality analysis, package must
withstand immersion under a head of water of at least 0.9 m (3 ft) in
the attitude for which maximum leakage is expected.

### Immersion of All Packages

A separate, undamaged specimen must be subjected to water pressure
equivalent to immersion under a head of water of at least 15 m (50 ft).
For test purposes, an external pressure of water of 150 kPa (21.7
lbf/in^2^) gauge is considered to meet these conditions.

## Nuclear safety specialists shall demonstrate a familiarity-level knowledge of the Price-Anderson Amendments Act of 1988 and its relationship to subparts A and B of 10 CFR 830.

a)  Describe the purpose and scope of the Price-Anderson Amendment Act.

The Price Anderson Amendment Act (PAAA) of 1988 provides incentives for
DOE contractors to abide by nuclear facility safety and operations
requirements as set forth by DOE in such documents as 10 CFR 830. The 10
CFR 830 regulatory requirements are intended to revise and supplement
the existing provisions in DOE Orders for nuclear safety and will
provide a more direct means to implement the civil penalty provisions of
the PAAA. The PAAA provides a reimbursement incentive to DOE contractors
(and their subcontractors and suppliers) for conducting activities that
involve source, special nuclear, or byproduct material in response to
public liability penalties associated with the consequences of those
activities.

j)  Discuss the Act's applicability to the Department's nuclear safety
    activities.

The intent of this act is that it will enhance compliance with such
rules as 10 CFR 830, including programs that contribute to an effective
and safe work environment at nuclear facilities.

k)  Describe the indemnity that DOE offers to contractors.

The PAAA also made DOE contractors (and their subcontractors and
suppliers) subject to civil penalties for violations of DOE rules and
regulations or orders related to nuclear safety.

l)  Discuss the requirements associated with the topics below, as they
    are affected by the rulemaking aspect of the Price-Anderson
    Amendment Act (PAAA):

-   Quality assurance requirements

-   Safety basis requirements

### Quality Assurance

DOE-STD-3009 helps identify quality assurance requirements to ensure
compliance with the PAAA.

### Safety Basis Requirements

The PAAA is the rule that drives the requirements for creating a safety
basis. As the PAAA is updated, the requirements for the safety basis may
need to be updated.

m)  Discuss the role of Department nuclear safety specialists with
    respect to implementing the requirements of the Price-Anderson
    Amendment Act.

The nuclear safety specialist ensures that all aspects of 10 CFR 830 are
implemented at any given site. The nuclear safety specialist performs
independent reviews and audits to determine that the contractor is
implementing all aspects of the act.

## Nuclear safety specialists shall demonstrate a working-level knowledge of the requirements in Department of Energy (DOE) Technical Standard DOE-STD-3009-94, Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports and DOE-STD-3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities.

a)  Discuss the conceptual basis and process for preparation of a
    facility/activity documented safety analysis.

The safety management programmatic requirements identified in 10 CFR 830
form the boundaries within which the safety analysis is performed and
represent the means of assuring safe operation of the facility. Hazard
analysis and accident analysis are performed to identify specific
controls and improvements that feed back into overall safety management.
Consequence and likelihood estimates obtained from this process also
form the bases for grading the level of detail and control needed in
specific programs. The result is documentation of the safety basis that
emphasizes the controls needed to maintain safe operation of a facility.

The level of detail provided in the DSA depends on numerous factors.
Applying the guidance for the graded approach in DOE-STD-3009 will
assist the preparer in establishing an acceptable level of detail. The
foundation for effectively preparing a DSA is the assembly and
integration of an experienced preparation team. The size and makeup of
the team depend on the magnitude and type of facility hazards and the
complexity of the processes that are required to be addressed in the
DSA. In determining the makeup of the preparation team, careful
consideration should be given to the key hazard analysis activity. In
general, the safety analysis base team should include, at a minimum,
individuals experienced in process hazard and accident analyses,
facility systems engineers, and process operators. Individuals with
experience in specific subject matter such as nuclear criticality,
radiological safety, fire safety, chemical safety, or process operations
may be needed in the hazard analysis on a regular or as needed basis.
Such individuals will typically be necessary in the development of
programmatic DSA chapters as well. Consistent, accurate exchange of
information among the team members is at least as important as the
makeup of the team itself. This can be assured through meaningful
integration of the required tasks.

Once team makeup is determined, base information needed to support DSA
development is gathered. Maximum advantage should be taken of pertinent
existing safety analyses and design information (i.e., requirements and
their bases) that are immediately available, or can be retrieved through
reasonable efforts. Other information arises from existing sources such
as process hazards analyses (PrHAs), fire hazards analyses (FHAs),
explosive safety analyses, health and safety plans (HASPs),
environmental impact statements (EISs), etc. The need for additional or
specific information becomes apparent throughout the hazard analysis
process. The remaining key steps for efficient completion of the safety
analysis and the DSA development process are listed below:

-   Identify the DSA project functions using project information and
    ensure the team matches the functions that are required.

-   Perform hazard analysis to provide facility hazard classification,
    evaluate worker safety and defense in depth, and identify unique and
    representative accidents to be carried forward to accident analysis.
    Safety-significant SSCs and TSRs are designated in hazard analysis
    as well.

-   Perform an accident analysis and assess the results to identify any
    safety-class SSCs and accident specific TSRs based on comparison of
    accident consequences to the Evaluation Guideline.

-   Develop the chapters for the DSA by providing information necessary
    to support the results of the safety analysis. These chapters detail
    the results of the analysis, describe the facility and the safety
    SSCs, and the safety management programs that relate to the facility
    safety basis.

n)  Discuss the following in relation to the preparation of the
    Documented Safety Analysis:

-   Worker safety

-   Defense-in-depth

-   Programmatic commitments

-   Technical safety requirements (TSRs)

-   Structures, systems, and components (SSCs)

-   Hazard analysis

-   Accident analysis

-   Application of the graded approach

-   Safe harbor methods

### Worker Safety

Workers, typically those in close proximity to operations, are the
population principally at risk from potential consequences associated
with hazard category 2 and 3 facilities. The DOE recognizes, via 10 CFR
830, the importance of including worker safety in safety analyses by
specifically noting the worker as a population of concern. Developing a
conceptual basis for the methodology used in 10 CFR 830 requires
answering the fundamental question of how worker safety is most
appropriately addressed in the DSA. The Occupational Health and Safety
Administration (OSHA) has published 29 CFR 1910.119, Process Safety
Management of Highly Hazardous Chemicals. The purpose of this regulation
is defined by OSHA in summary fashion as, "employees have been and
continue to be exposed to the hazards of toxicity, fires, and explosions
from catastrophic releases of highly hazardous chemicals in their
workplaces. The requirements in this standard are intended to eliminate
or mitigate the consequences of such releases." Many of the topics
requiring coverage in this federal regulation, such as design codes and
standards, process hazard analysis, human factors, training, etc., are
directly parallel to the requirements in 10 CFR 830. DOE O 440.1 and the
OSHA standard address the issue of worker safety from process accidents
by requiring the performance of hazards analyses for processes
(exclusive of standard industrial hazards) in conjunction with
implementation of basic safety programs that discipline operations and
ensure judgments made in hazard analyses are supported by actual
operating conditions. These requirements effectively integrate programs
and analyses into an overall safety management structure without
requiring quantitative risk assessment. This integration and the basic
concepts of process safety management (PSM) described by OSHA
regulations and the manuals and codes of practice described in DOE O
440.1 are philosophically accepted as appropriate for DSAs. This
standard effectively merges PSM principles with traditional DSA
precepts.

### Defense in Depth

Defense in depth as an approach to facility safety has extensive
precedent in nuclear safety philosophy. It builds in layers of defense
against release of hazardous materials so that no one layer by itself,
no matter how good, is completely relied upon. To compensate for
potential human and mechanical failures, defense in depth is based on
several layers of protection with successive barriers to prevent the
release of hazardous material to the environment. This approach includes
protection of the barriers to avert damage to the plant and to the
barriers themselves. It includes further measures to protect the public,
workers, and the environment from harm in case these barriers are not
fully effective. The defense-in-depth philosophy is a fundamental
approach to hazard control for nonreactor nuclear facilities even though
they do not possess the catastrophic accident potential associated with
nuclear power plants. In keeping with the graded-approach concept, no
requirement to demonstrate a generic, minimum number of layers of
defense in depth is imposed. However, defining defense in depth as it
exists at a given facility is crucial for determining a safety basis.
Operators of DOE facilities need to use the rigorous application of
defense-in-depth thinking in their designs and operations. Such an
approach is representative of industrial operations with an effective
commitment to public and worker safety and the minimization of
environmental releases. For high-hazard operations, there are typically
multiple layers of defense in depth. The inner layer of defense in depth
relies upon a high level of design quality so that important SSCs will
perform their required functions with high reliability and high
tolerance against degradation. The inner layer also relies on competent
operating personnel who are well trained in operations and maintenance
procedures. Competent personnel translate into fewer malfunctions,
failures, or errors and, thus, minimize challenges to the next layer of
defense. In the event that the inner layer of defense in depth is
compromised from either equipment malfunction (from whatever cause) or
operator error and there is a progression from the normal to an abnormal
range of operation, the next layer of defense in depth is relied upon.
It can consist of (1) automatic systems, or (2) means to alert the
operator to take action or manually activate systems that correct the
abnormal situation and halt the progression of events toward a serious
accident. Mitigation of the consequences of accidents is provided in the
outer layer of defense in depth. Passive, automatically or manually
activated features (e.g., containment or confinement system, deluge
systems, filtered exhaust), and/or safety management programs (i.e.,
emergency response) minimize consequences in the event that all other
layers have been breached. The contribution of emergency response
actions to minimizing consequences of a given accident cannot be
neglected as they represent a truly final measure of protection for
releases that cannot be prevented. Structures, systems, or components
that are major contributors to defense in depth are designated as
safety-significant SSCs.

### Programmatic Commitments

Program commitments (e.g., radiation protection, maintenance, quality
assurance) encompass a large number of details that are more
appropriately covered in specific program documents (e.g., plans and
procedures) external to the DSA. The cumulative effect of these details,
however, are recognized as being important to facility safety, which is
the rationale for a top-level program commitment becoming part of the
safety basis. The importance of the program commitments, which can be
incorporated in TSRs as administrative controls, cannot be
overestimated. The safety basis, however, includes only the top-level
summary of program elements, not the details of the program or its
governing documents. Inspection discrepancies in a program would not
constitute violation of the safety basis unless the discrepancies were
so gross as to render premises of the summary invalid. By virtue of
application of the graded approach, the majority of the engineered
features in a facility will not be identified in the categories of
safety-class or safety-significant SSCs even though they may perform
some safety functions. However, such controls noted as a barrier or
preventive or mitigative feature in the hazard and accident analyses
must not be ignored in managing operations. Such a gross discrepancy
would violate the safety basis documented in the DSA even if the
controls are not designated safety-class or safety-significant, because
programmatic commitments extend to these SSCs as well. At a minimum, all
aspects of defense in depth identified must be covered within the
relevant safety management programs (e.g., maintenance, quality
assurance) committed to in the DSA. The details of that coverage,
however, are developed in the maintenance program as opposed to in the
DSA. Facility operators are expected to have noted the relative
significance of these engineered features and have provided for them in
programs, in keeping with standard industrial practice, based on the
importance of the equipment. It is the fact of coverage that is relevant
to the facility safety basis. The details of this programmatic coverage
(i.e., exact type of maintenance items and associated periodicities) are
not developed in or part of the DSA. One overall commitment made in a
DSA is that the contractor will not change the facility configuration
underlying the documented safety basis without implementing and
completing the USQ process. However, situations do occur where a USQ
process is not necessary. For example, a stipulation to have a radiation
protection program in the administrative control section of the TSR is a
commitment; however, changes to specific program provisions do not
require going through the USQ process. DOE facilities that use and rely
on site-wide, safety support services, organizations, and procedures,
may summarize the applicable site-wide documentation provided its
interface with the facility is made clear.

### Technical Safety Requirements

Descriptions result in the definition of functional requirements and
associated performance criteria used to derive TSRs. TSRs are safety
controls developed in accordance with the precepts of 10 CFR 830. TSR
commitments encompass the following. TSRs comprise: (1) SLs, (2)
operational limits consisting of LCSs and LCOs and associated SRs, (3)
ACs, (4) use and application provisions, (5) design features, and (6)
bases appendix. Based on the results of hazard and accident analysis,
TSRs are designated for (1) safety-class SSCs and controls established
on the basis of application of the evaluation guideline, (2)
safety-significant SSCs, (3) defense in depth in accordance with the
screening criteria of DOE G 423.1-1, and (4) safety management programs
for defense in depth or worker safety. The bases appendix provides the
linkage to the DSA. It is important to develop TSRs judiciously. TSRs
should not be used as a vehicle to cover the many procedural and
programmatic controls inherent in any operation. Excessive use of TSR
limits to manage operations will result in distortion of the regulatory
structure DOE is attempting to develop and will dilute the emphasis
intended for the most critical controls. SLs should be limited in number
and designated with caution. They should, however, receive coverage in
the administrative control section of TSRs at a minimum. Judgment should
be used to determine what controls warrant use of operational limits.
When TSR administrative controls are used for purposes other than
generic coverage of safety management programs, descriptions should be
sufficiently detailed that a basic understanding is provided of what is
controlled and why. Beyond safety-significant SSCs designated for worker
safety and their associated TSR coverage, additional worker safety
issues should be covered in TSRs only by administrative controls on
overall safety management programs.

### Structures, Systems, and Components 

Descriptions result in the definition of functional requirements and
associated performance criteria used to derive SSCs. The rule defines
safety-class designation for SSCs that are established on the basis of
application of the evaluation guidelines. This designation carries with
it the most stringent requirements (e.g., enhanced inspection, testing
and maintenance, and special instrumentation and control systems).
Appendix A of DOE-STD-3009 provides guidance for implementing the
evaluation guideline to classify SSCs as safety-class SSCs. This
category of SSCs is provided to ensure that important SSCs will be given
adequate attention in the DSA and facility operations programs.
Safety-significant SSCs are those of particular importance to defense in
depth or worker safety as determined in hazard analysis. Control of such
SSCs does not require meeting the level of stringency associated with
safety-class SSCs. The evaluation guideline is not used for designating
safety-significant SSCs. Safety-class SSCs are designated to address
public risk, which makes a dose guideline at the site boundary a useful
tool. Safety-significant SSCs address risk for all individuals within
the site boundary as well as additional defense in depth for the public,
making a dose guideline at any one point an artificial distinction
distorting the process of systematically evaluating SSCs. TSRs covering
SSCs ensuring defense in depth should generally correlate with
safety-significant SSC designation for defense in depth, but exact
one-to-one correlation is not required.

### Hazard Analysis

The initial analytical effort for all facilities is a hazard analysis
that systematically identifies facility hazards and accident potentials
through hazard identification and hazard evaluation. The focus of the
hazard analysis is on thoroughness and requires evaluation of the
complete spectrum of hazards and accidents. This largely qualitative
effort forms the basis for the entire safety analysis effort, including
specifically addressing defense in depth and protection of workers and
the environment. Basic industrial methods for hazard analysis, its
interface with more structured quantitative evaluations, and the basis
for both have been described in references such as the American
Institute of Chemical Engineers Guidelines for Hazard Evaluation
Procedures (1992). These guidelines have been accepted by OSHA as the
standard for analytical adequacy in characterizing commercial chemical
processes that perform the same type of unit operations conducted at DOE
nonreactor nuclear facilities. The largely qualitative techniques
described in the above reference on hazard analysis provide
methodologies for comprehensive definition of the accident spectrum for
workers and the public. The basic identification of hazards inherent in
the process provides a broad, initial basis for identification of safety
programs needed (e.g., radiation protection, hazardous chemical
protection). The hazard analysis then moves beyond basic hazard
identification to evaluation of the expected consequences and estimation
of likelihood of accidents, an activity that in no way connotes the
level of effort of a probabilistic or quantitative risk assessment.
Throughout the evaluation process, preventive and mitigative SSCs and
pertinent elements of programmatic controls are identified. This
identification also establishes functional requirements for SSCs, which
will subsequently delineate the technical information (i.e., response
parameters), needed to establish performance criteria. The DSA
summarizes these requirements and criteria for safety-class and
safety-significant SSCs only. Refinement of the information obtained in
hazard evaluation leads to overall definition of defense in depth,
worker safety, and environmental protection. The most significant
aspects of defense in depth and worker safety are subject to designation
as safety-significant SSCs and coverage by TSRs. Other items noted are
encompassed by the details of safety management programs (e.g.,
procedures, training, maintenance, quality assurance), which can be
captured in top-level fashion in TSR administrative controls. The hazard
evaluation conducted to assess the accident spectrum associated with
hazards germane to the DSA indicates the adequacy of programmatic
efforts and provides input to programmatic activities whose discipline
provides a significant margin of safety. The process outlined above is
self grading for analytical effort. Analytical effort can be limited to
a simple, resource efficient hazard analysis geared to facility needs,
unless events are noted that are of sufficient complexity to require
more detailed, quantitative evaluations to understand the basis for
safety assurance. Implicit in this methodology is the statement of
DOE-STD-1027 that the largely qualitative level of effort in hazard
analysis is appropriate and sufficient for accident analysis of Hazard
Category 3 facilities. It is again noted that the hazard analysis effort
is not a quantitative risk assessment. Preparers cannot expect the level
of detail associated with a quantitative risk assessment in a hazard
analysis, as the hazard analysis is focused on systematically assessing
what can go wrong in a facility as opposed to deriving mathematical
expressions of risk. The final purpose of hazard analysis is to identify
a limited subset of accidents to be carried forward to accident
analysis. Identification of DBAs in safety analysis and use of DBAs is
appropriate in defining a facility safety basis derivative DBAs.

### Accident Analysis

The complete spectrum of accidents is examined in hazard analysis. A
limited subset of accidents, (i.e., DBAs and derivative DBAs) that bound
"the envelope of accident conditions to which the operation could be
subjected" is carried forward to accident analysis where safety-class
SSCs are designated by comparison of accident consequences to the
evaluation guideline. These scenarios are the accidents requiring formal
definition. Information obtained from specific accidents or
representative accidents enveloping many small accidents are used to
specify functional requirements for safety-class SSCs. An accident
analysis is performed for the bounding accidents. Accident analysis in
this standard refers to the formal quantification (i.e., all assumptions
identified and justified and individual computations presented or
summarized) of accident consequences. The general binning estimates used
in hazard analysis are adequate and representative of the level of
effort desired for frequency determination. Accordingly, accident
analysis need only document the basis used in hazard analysis for
assigning accident likelihood to two orders-of-magnitude bins. The
quantified consequences are compared to the numerical evaluation
guideline for the purpose of identifying safety-class SSCs and any
accident-specific assumptions requiring coverage by TSRs.

### Application of the Graded Approach

The use of a graded approach for the effort expended in safety analysis
and the level of detail presented in associated documentation is
prescribed in 10 CFR 830. The graded approach applied to DSA preparation
and updates is intended to produce cost-efficient safety analysis and
DSA content that provide adequate assurance to the DOE that a facility
has acceptable safety provisions without providing unnecessary
information. As described in 10 CFR 830, the graded approach adjusts the
magnitude of the preparation effort to the characteristics of the
subject facility based on seven factors:

-   The relative importance to safety, safeguards, and security

-   The magnitude of any hazard involved

-   The life-cycle stage of a facility

-   The programmatic mission of a facility

-   The particular characteristics of a facility

-   The relative importance of radiological and non-radiological hazards

-   Any other relevant factor

The rule provides for developing the DSA based on judgment of the
facility in relation to these seven factors. For example, simple hazard
category 3 facilities or facilities that have a short operational life
may only require a limited but adequate analysis documented to a level
less than that required for a hazard category 2 facility. In addition,
facilities with short operational lives should consider the
appropriateness of using DOE-STD-3011 to meet the requirements of 10 CFR
830. On the opposite end of the spectrum, a complex hazard category 1
facility that is just going into operation requires extensive analysis
and highly detailed documentation. The application of the graded
approach may allow for much simpler analysis and documentation for some
of these facilities. For facilities of little hazard, or hazards at the
hazard category 3 level, for which only a modest reduction of risk is
required, the DSA may be simple and short. In such cases all of the
topics for the DSA listed in this standard may not be necessary, and
with proper technical bases, some topics may be omitted or reduced in
the detail that would otherwise be required of hazard category 1 or 2
facilities. Thus, with application of the graded approach, DSAs for
hazard category 3 facilities or facilities with short operational lives
will normally require more simplified DSA analysis and documentation.

### Safe Harbor Methods

Safe harbor methodologies are methodologies for DSA development that
have already been approved by DOE for use in specific circumstances.
Table 2 of appendix A to subpart B of 10 CFR 830 identifies the list of
safe harbor methodologies. Any other methods not listed in this table
require DOE approval before they can be used.

o)  Discuss the relationship between the safe harbor methods for a DSA
    in 10 CFR 830 Appendix A and the requirements for a DSA in 10 CFR
    830.204 with regard to completeness.

The documents identified in Appendix A of 10 CFR 830 are the documents
that DOE has identified as being able to meet the criteria for
completeness in regards to completing a DSA. Section 830.204 of 10 CFR
830 specifies the DSA requirements as listed below:

-   Describe the facility (including the design of safety SSCs) and the
    work to be performed.

-   Provide a systematic identification of both natural and man-made
    hazards associated with the facility.

-   Evaluate normal, abnormal, and accident conditions, including
    consideration of natural and man-made external events,
    identification of energy sources or processes that might contribute
    to the generation or uncontrolled release of radioactive and other
    hazardous materials, and consideration of the need for analysis of
    accidents which may be beyond the design basis of the facility.

-   Derive the hazard controls necessary to ensure adequate protection
    of workers, the public, and the environment, demonstrate the
    adequacy of these controls to eliminate, limit, or mitigate
    identified hazards, and define the process for maintaining the
    hazard controls current at all times and for controlling their use.

-   Define the characteristics of the safety management programs
    necessary to ensure the safe operation of the facility, including
    quality assurance, procedures, maintenance, personnel training,
    conduct of operations, emergency preparedness, fire protection,
    waste management, and radiation protection.

With respect to a nonreactor nuclear facility with fissionable material
in a form and amount sufficient to pose a potential for criticality,
define a criticality safety program that:

-   ensures that operations with fissionable material remain subcritical
    under all normal and credible abnormal conditions

-   identifies applicable nuclear criticality safety standards

-   describes how the program meets applicable nuclear criticality
    safety standards

p)  Describe the objectives of requiring accident analyses in safety
    basis documents.

The accident analysis is used to quantify the hazard analysis and to bin
accidents by order of magnitude. This is done to have relative values to
compare to the evaluation guidelines that define limits allowable for
both workers and the public.

q)  Identify and discuss the use of the source term five factor formula
    in accident analyses.

The source term five-factor formula is identified below:

Source Term = MAR \* DR \*ARF \* RF \* LPF

where:

MAR = material at risk (curies or grams)

DR = damage ratio

ARF = airborne release fraction (or airborne release rate for continuous
release)

RF = respirable fraction

LPF = leakpath factor

### Material at Risk (MAR)

The MAR is the amount of radionuclides (in grams or curies of activity
for each radionuclide) available to be acted on by a given physical
stress. For facilities, processes, and activities, the MAR is a value
representing some maximum quantity of radionuclide present or reasonably
anticipated for the process or structure being analyzed. Different MARs
may be assigned for different accidents as it is only necessary to
define the material in those discrete physical locations that are
exposed to a given stress. For example, a spill may involve only the
contents of a tank in one glovebox. Conversely, a seismic event may
involve all of the material in a building.

### Damage Ratio (DR)

The DR is the fraction of the MAR actually impacted by the
accident-generated conditions. A degree of interdependence exists
between the definitions of MAR and DR. If it is predetermined that
certain types of material would not be affected by a given accident,
some analysts will exclude this material from the MAR. The DR is
estimated based upon engineering analysis of the response of structural
materials and materials of construction for containment to the type and
level of stress/force generated by the event. Standard engineering
approximations are typically used. These approximations often include a
degree of conservatism due to simplification of phenomena to obtain a
useable model, but the purpose of the approximation is to obtain, to the
degree possible, a realistic understanding of potential effects.

### Airborne Release Fraction (ARF)

The ARF is the coefficient used to estimate the amount of a radioactive
material suspended in air as an aerosol and thus available for transport
due to physical stresses from a specific accident. For discrete events,
the ARF is a fraction of the material affected. For mechanisms that
continuously act to suspend radionuclides (e.g., aerodynamic
entrainment/resuspension), a release rate is required to estimate the
potential airborne release from postulated accident conditions.
Generally, accident release rates (ARRs) are based upon measurements
over some extended period to encompass most release situations for a
particular mechanism. The rates are average rates for the broad spectrum
of situations and, as such, the most typically meaningful time unit to
reflect average conditions is 1 hour. There is evidence that in some
situations (e.g., aerodynamic entrainment of sparse powder deposits on a
heterogeneous surface), the rate of release is not uniform with time.
Even in the situations where the rates are relatively uniform, the
source is depleted by the removal of particles from the surface by
aerodynamic forces, and the amount of material airborne decreases with
time unless the source is continuously replenished. The ARFs are based
primarily upon experimentally measured values for the specific material
(e.g., plutonium, uranium, mixed fission products) or surrogates
subjected to the particular type of stress under controlled conditions.
Attention is given to the parameters, if known, that may have a
significant influence upon suspension by the specific mechanism and the
uncertainty in the measurement as indicated by the variability of the
results.

### Respirable Fraction (RF)

The RF is the fraction of airborne radionuclides as particles that can
be transported through air and inhaled into the human respiratory system
and is commonly assumed to include particles 10-µm Aerodynamic
Equivalent Diameter (AED) and less. Other definitions of "respirable
particles" have been presented by various groups at different times. The
British Medical Research Council adopted a definition in 1952
classifying particles with a terminal velocity equal to that of a 5-μm
diameter as "respirable dust." The U.S. Atomic Energy Commission defined
"respirable dust" as insoluble particles that are part of inhaled dust
which penetrate to the non-ciliated portions of the gas exchange region,
and with a 50 percent respirable cut-size of 3.5-μm AED. The American
Conference of Governmental Industrial Hygienists has adopted a
definition that is almost identical, differing only in the 2 μm fraction
allowed. The U.S. Environmental Protection Agency defines "inhalable
dust" (particles penetrating the upper respiratory airway and entering
the thorax) with a 50 percent cutoff at 15-μm AED. The International
Standards Organization---Europe defines "inhalable dust" as particles
entering the nasal or oral passages with a 50 percent cut-size of 10-μm
AED. Accordingly, use of a 10-μm AED cut-size for respirable particles
is considered conservative, and may even be overly conservative since
the mass is a cube function of particle diameter. The size of a particle
is a function of the measurement technique used. If the method used is
optical/electron microscopy or spectrometry, particle size is a
projected diameter measured by the plane that intercepts the
light/electron beam or reflection from light scattered by the particle.
The size represents the two-dimensional area intercepting the beam and,
as with all projections of three dimensions into two, can result in
considerable distortion. Projected diameter approximates the Geometric
Diameter (Dg). Dg is also approximated by sieving where the size
measurement is termed geometric/linear/least linear diameter. The
measurement represents the smallest dimension of the particle that will
pass through the openings in the sieve.

### Leakpath Factor (LPF)

The LPF is the fraction of the radionuclides in the aerosol transported
through some confinement deposition of filtration mechanism. There can
be many LPFs for some accident conditions (e.g., the fraction
transported from the package, such as a shipping container, to the cell
or enclosure; the fraction leaked from the enclosure, cell, or glovebox
to the operating area around the enclosure or room; the fraction leaked
from the room to the building-atmosphere interface). Where multiple
leakpaths are involved, their cumulative effect is often expressed as
one value that is the product of all leakpath multiples. The LPF is a
calculated or standard value based upon (1) established relationships
between the size of the particulate material, airborne transport
mechanisms, and losses by deposition mechanisms, or (2) specified
filtration efficiencies.

r)  Given an accident scenario, determine a reasonably bounding estimate
    of the airborne release fraction (ARF) and respirable fraction (RF),
    material at risk (MAR), leak path factor (LPF) and damage ratio (DR)
    to determine the product (MAR x ARF x RF x DR x LPF).

    Note: This is a performance-based competency. The ability to
    complete this competency must be observed by a qualifying official.

s)  Identify and discuss the methods, conventions and data sources used
    in developing estimates of the five factors for use in accident
    analyses.

In most cases, the ARFs and RFs for conditions bounded by the
experimental parameters can be defined to one significant digit. The
parameter definition process has focused on estimating reasonable
bounding values because of the limited quantity and variability of the
data. The use of the word "reasonable" is an acknowledgement that the
only definitive bounds are ARFs and RFs of 1.0, which can always be
postulated if enough synchronous, extreme localized conditions are
assumed. Such extreme synchronicity is neither an expected condition nor
a practically useful model of reality. The NRC has commented on this
subject in a survey of nonreactor nuclear facility release potential and
historical experience. A key element in defining bounding values from
the data understands the physical entrainment mechanisms at work, their
potential variability, and the inherent limits of such mechanisms.
Median and average values are estimated for some data. These estimates
are made solely for the purpose of providing perspective on potential
conservatism and should not be used as a basis for an ARF statistical
distribution. The generation and suspension of particles is the result
of the interaction of multiple physiochemical variables that have not
been completely characterized as the majority of the experiments
performed were designed in an attempt to reflect reasonably bounding
conditions for specific industrial situations of concern. Accordingly,
the data obtained are more accurately characterized as selected points
from multiple distributions against multiple parameters than as
different values from a common distribution. Even if this point is
neglected, there are still practically intractable problems in
attempting to generate statistical distributions. While the data are
presumed to be bounding for the purpose intended, it is largely unknown
whether the data values are truly 90th percentile, 99th percentile,
99.9th percentile, etc. Further, in many cases it is considered likely
that accident specific ARFs are actually distributed in a highly
irregular manner (i.e., multi-modal or truncated distributions).
Assuming a typical distribution (i.e., log-normal, Poisson) using
standard deviations will produce seriously distorted values that may
have little or nothing to do with reality. The available data do,
however, cover a range of conditions that typify the energy sources
associated with nonreactor nuclear facilities. The data cover a more
complete range of phenomenological concerns than the data upon which
nuclear reactor source terms have been estimated. In general, scaling
effects, while not to be trivialized, are less of an issue with this
data than with comparable reactor source term estimation data. The NRC
has already accepted a considerable amount of this information as a
basis for source term estimation in NUREG-1320, Nuclear Fuel Cycle
Facility Accident Analysis Handbook. DOE-HDBK-3010-94 serves a similar
function for DOE. The values for ARFs and RFs taken from experimental
measurements are reasonably well defined. It is noted, however, that
they are dependent upon the types and levels of stress imposed, the
initial state (physical form, chemical composition, particle size
distribution, degree of dispersion of the material-of-concern), and the
response of the material-of-concern and other materials present. The
applicability of the experimental conditions to the complete spectrum of
processes and potential accident conditions is, however, uncertain. For
this reason, the discussions of the data have tried to indicate where
typical industrial accident phenomena (e.g., fires, explosions, spills)
are considered covered by the data. The estimates of ARFs and RFs
applicable to various accident-generated mechanisms for the suspension
of radioactive materials are based upon experimental data for specific
types and levels of stresses/ force. Care must be used in applying the
ARFs/ARRs to ensure that the values chosen truly reflect the type and
level of stress/force postulated for the event. For instance, the
suspension of powder from a surface (commonly termed resuspension) is
not applicable to situations where the powder is dropped into flowing
gas in a dispersed fashion. Before the ARFs and RFs presented can be
properly applied, the conditions imposed and the response of critical
items must be evaluated. Many standard methods are applicable (e.g., the
rupture pressure of tanks and piping based upon the material of
construction, the thickness, the temperature and pressure). In other
cases (e.g., the blast energy from the deflagration of flammable gas and
oxidant mixtures in the free volume above the materials), however,
standard engineering calculational methods are not available and
interpretation of information and data (e.g., the fraction of the heat
of combustion of reactants that translates into the shock wave) is
required. Once the forces and conditions imposed upon the material for
dispersion/fragmentation and suspension are identified, the applicable
ARF and RF can be selected. In most cases, precise correspondence
between the event conditions and experimental conditions during the
measurement of the ARFs and RFs is not found. For conservative analysis,
the data are applicable if the measurement conditions exceed those
calculated for the event (e.g., if the fall distances for spilled
powders or liquids with characteristics like the materials used in the
experiments are equal to or less than the experimental distance of 3 m).
In most cases, extrapolation beyond the experimental data is valid for a
limited range beyond the maximum (a factor of 2 to 5 dependent on the
slope of the experimental data and the range of conditions covered in
the experimental study) imposed in the experimental study. Models are
available for the calculation of ARFs and RFs for some phenomena (e.g.,
free-fall spill of powders and liquids - Ballinger et. al., 1988; PULF
formula for fragmentation by brittle fracture - Sandia, 1987). Care
should be used in any extrapolation, however, to avoid producing
obviously inappropriate answers. This caution is particularly apt if
calculations are being used to influence facility or process design. A
final emphasis is necessary regarding application of this data. As
developed for the NRC and DOE, it has never been intended to be used as
absolute proof of anything. Special attention has been given to
understanding suspension phenomena, ranges of relevant parameters
covered in experimental studies, artifacts or limitations of the data
that may have been induced by experimental conditions, and possible
effects of relevant parameters that may not have been controlled or
monitored. As noted, this has resulted in development of bounding ARFs
and RFs. The purpose of developing these values was (1) to better
understand the potential bounding hazards presented by nonreactor
nuclear facilities, and (2) to provide information to support general
bases of decision making. The first purpose has been fulfilled by use of
this information in the Defense Programs Safety Survey Report. This use
has supported previous NRC estimates that the bounding consequence
potential for nonreactor nuclear facilities is significantly less than
commercial nuclear power plants, or large commercial chemical plants as
well. In domino fashion, this conclusion has reemphasized the use of the
term "general bases" in the second purpose.

t)  Identify and discuss methods/codes used to determine the
    environmental dispersion and delivered doses from accidental
    releases of hazardous materials.

  ------------------------------ ----------------------------------------
  **Type of Facility Analyzed**  **Method of Analysis**

  Low-Complexity Operations      Use Hazard Analysis

  Single-Failure                 Use Failure Modes and Effects Analysis
  Electro-Mechanical Systems     (FMEA)

  Systems with Redundant         Use Event Tree Analysis (ET)
  Barriers or Requiring Multiple 
  Failures                       

  Large, Moderately Complex      Use Fault Tree Analysis (FT)
  Processes                      

  Complex Fluid Processes        Use Hazard and Operability Studies
                                 (HAZOP)

  High-Complexity Facilities     Use Integrated Event Tree and Fault Tree
                                 Techniques (ETs/FTs)
  ------------------------------ ----------------------------------------

To date, six codes have been identified as toolbox codes:
[ALOHA](http://www.eh.doe.gov/sqa/central_ registry/ALOHA/aloha.htm),
[CFAST](http://www.eh.doe.gov/sqa/central_ registry/CFAST/cfast.htm),
[EPICODE](http://www.eh.doe.gov/sqa/central_ registry/EPIcode/EPI.htm),
[GENII](http://www.eh.doe.gov/sqa/central_ registry/GENII/gen2.htm),
[MACCS2](http://www.eh.doe.gov/sqa/central_ registry/MACCS2/maccs2.htm),
and
[MELCOR](http://www.eh.doe.gov/sqa/central_ registry/MELCOR/melcor.htm).
A general description of each code is provided along with the respective
code owner. The link Software Quality Assurance Central Registry
provides a complete description of the software.

u)  Discuss the effect of prevailing weather, building wake effects, and
    plume buoyancy upon the magnitude and distribution of doses from
    hazardous releases into the atmosphere.

Each code type takes into account the way values are affected by
weather, building configuration, and their distribution in an accident.

v)  Identify and discuss the treatment of uncertainty and the realistic
    effects in accident analyses.

To analyze a "worst case," the uncertainty and realistic effects do not
directly affect the analysis; the overall worst case is used, which
lacks any real credibility of being a realistic approach.

w)  Identify the purpose and relationship between Chapters 3, 4, and 5
    and the technical safety requirements of the documented safety
    analysis.

Chapter 3 of the DSA is the hazard analysis and accident analysis
section, and chapter 4 is the Safety SSC section, which identifies
controls that are derived from the hazard and accident analysis. Chapter
5 is the TSR, which identifies the SSCs that need to provide the
greatest protection and the requirements for these controls. They are
all linked from need to creation to implementation and monitoring.

x)  Complete a review of a hazards analysis including walking down the
    scope of work area.

    Note: This is a performance-based competency. The ability to
    complete this competency must be observed by a qualifying official.

y)  Complete a review of an accident analysis including walking down the
    accident scenario.

    Note: This is a performance-based competency. The ability to
    complete this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working-level knowledge of NSTP-2003-1 that describes the relationship between human factors/human performance and institutional programs that support the safety analysis.

a)  Identify and discuss aspects of person-machine interface that can
    degrade or enhance the safety performance of personnel.

The principals of human performance improvement are listed below:

-   Even people working their best may make mistakes.

-   Error-likely situations are predictable, manageable, and
    preventable.

-   Individual behavior is influenced by organizational processes and
    values.

-   People achieve high levels of performance based largely on the
    encouragement and reinforcement received from leaders, peers, and
    subordinates.

-   Events can be avoided by understanding the reasons mistakes occur
    and applying the lessons learned from past events.

z)  Identify and discuss how written procedures are conducive to
    reliable or unreliable performance of activities important to
    safety.

Written procedures are conducive to performance of activities important
to safety when they are developed using the following objectives:

-   Address procedures with a clear, logical sequence of tasks that make
    them understandable to the user. Procedures must provide all the
    information individuals need to perform the task.

-   Communicate policies for procedure use and adherence. Guidance for
    use of specific procedures should identify those tasks or evolutions
    for which continuous in-hand use or strict adherence is required.

-   Verify the integrity of defenses, especially for tasks important for
    nuclear safety. The number and strength of defenses, such as
    multiple safeguards, equipment trains, physical barriers,
    supervision, and procedures, are related to the potential
    safety-related consequences of errors.

-   Alert workers and supervisors to key tasks/decision points. Elevate
    the attention of individuals by incorporating appropriate cues.

-   Incorporate appropriate information into procedures from applicable
    source documents, such as the plant design documents and
    plant-specific safety analysis.

-   Detail prerequisites and initial conditions. Carefully consider the
    location of this information in the procedures so that the intent of
    each procedure is understood.

-   Define nomenclature used in the procedures.

-   Describe only one action in each procedure step.

-   Provide individual signoffs for selected critical steps. One signoff
    should apply to only one action. Separate signoffs should be
    provided for independent verifications.

a)  Identify and discuss how personnel training programs can be
    conducive to safety or prone to error.

All personnel should be trained on proper implementation of all ACs, not
just for those that are deemed necessary for job requirements. There
shall be a training implementation matrix to ensure training is adequate
and complete. If training is not updated and monitored, training may be
missed by employees and any updates to TSRs may be overlooked.

b)  Identify and discuss how staffing and qualification of operational
    personnel are conducive to safe versus unsafe operations.

To ensure safe operations, all personnel must be able to perform their
assigned tasks and be trained on all TSR requirements. The training must
include updates/changes to TSRs along with addressing any additional TSR
requirements. Training should be adequate to ensure complete compliance
with TSRs to minimize any unsafe operations due to performance of TSRs.

c)  Identify and discuss the influence of management and organizational
    factors upon safety of operations.

Management and organizational factors must foster a work environment
that promotes excellence. The following way to conduct business can be
implemented to improve, or to minimize the degradation of, the
human/machine interface:

-   Communicate accurately and frequently.

-   Anticipate error-likely situations.

-   Check others' work and monitor progress among each other as work is
    performed.

-   Practice critical thinking.

-   Follow approved procedures.

-   Stop the task and collaborate with others when unfamiliar or
    unanticipated conditions arise.

d)  Identify and discuss the methods used to estimate the probability of
    significant mistakes made by personnel and the relationship to
    probabilistic risk assessment.

The following references describe human factors in risk assessment:

-   Excellence in Human Performance, INPO, September 1997.

-   Putting the Human into Hazard Assessment, Helen Rycraft, BNFL, a
    paper presented at the 2003 annual meeting of the Energy Facilities
    Contractors Group (EFCOG) Safety Analysis Working Group, Salt Lake
    City, June 2003.

e)  Identify and discuss the methods for assessing the reliability of
    administrative controls contained in TSRs and facility programs.

The following references provide guidance on the development of
effective and reliable ACs:

-   Excellence in Human Performance, INPO, September 1997.

-   Guidelines for the Conduct of Operations at Nuclear Power Plant
    Stations, Institute of nuclear power operations, INPO 01-002, May
    2001.

## Nuclear safety specialists shall demonstrate a working-level knowledge of assessment techniques (such as the planning and use of observations, interviews, and document reviews) to assess facility performance, report results of assessments, and follow up on actions taken as the result of assessments.

a)  Describe the role of nuclear safety specialists in the assessment of
    government owned contractor operated (GOCO) facilities.

The role of nuclear safety personnel is to ensure that safety
requirements necessary for a practical safeguarding of applicable DOE
facilities and personnel are being adequately implemented. Practical
safe work procedures include training of skilled and unskilled personnel
who work in an area where work is performed with nuclear material. For
exposure to nuclear material, safety measures shall protect personnel
against both normal operations and emergency situations. In addition,
only qualified persons who are capable of working safely in nuclear
facilities and are familiar with the proper use of special precautionary
techniques, personal protective equipment, TSRs, and operating
procedures may perform work with nuclear material. Nuclear safety
personnel perform DOE line management oversight of their assigned
facilities to ensure that

-   the contractor is operating facilities safely and efficiently (i.e.,
    within the boundaries of those controls invoked in the facility
    authorization basis);

-   the contractor's management system is effectively controlling
    conduct of operation as related to nuclear safety;

-   effective lines of communication between DOE and its operating
    contractors are maintained during periods of normal operation, and
    following reportable events, in accordance with DOE Orders and
    requirements.

f)  Describe the assessment requirements and limitations associated with
    the interface with contractor employees.

As assessment requirements and limitations associated with the interface
of nuclear safety personnel and contractor employees vary from site to
site, the local qualifying official will evaluate the completion of this
competency.

g)  Discuss the essential elements of a performance-based assessment
    including:

-   Investigation

-   Fact finding

-   Exit interview

-   Reporting

-   Follow-up

-   Closure

### Investigation

It is important to begin the investigation as soon as an assessment is
called for to ensure that data is not lost. The information that should
be collected consists of conditions before, during, and after operation
of the facility; personnel involvement; environmental factors; and other
information having relevance to the operation of the facility.

### Fact Finding

Once all the data has been collected, the data should be verified to
ensure accuracy. The investigation may be enhanced if some physical
evidence is retained. Establishing a quarantine area, or the tagging and
segregation of pieces and material, should be performed for failed
equipment or components. The basic need is to determine the direct,
contributing, and root causes so that effective corrective actions can
be taken that will prevent recurrence. Some areas to be considered when
determining what information is needed include

-   activities related to the operations of the facility

-   initial or recurring problems

-   hardware (equipment) or software (programmatic-type issues)
    associated with the facility

-   recent administrative program or equipment changes

-   physical environment or circumstances

Some methods of gathering information include conducting interviews and
collecting statements. Interviews must be factual. Preparing questions
before the interview is essential to ensure that all necessary
information is obtained. Interviews should be conducted, preferably in
person, with those people who are most familiar with the system.
Individual statements could be obtained if time or the number of
personnel involved makes interviewing impractical. Interviews can be
documented using any format desired by the interviewer. Consider
conducting a walk-through of the system or facility as part of the
interview if time permits.

### Exit Interview

Assessments can gain value from an exit interview. This interview is
used primarily to present the assessment summary and provide the
assessed organization an opportunity to verify the factual accuracy of
assessment results. To facilitate this, assessors should be prepared to
provide detailed supporting information for those results (ideally, a
draft assessment report should be available at this time). This
interview also offers an opportunity for the assessed organization to
present its management position and any plans for addressing the
results. Reasonable time should be allowed to discuss any concerns, but
this interview should not be used to argue the assessment agenda or
methodology.

### Reporting

Review of reports and documents helps develop the foundation for
identifying weaknesses and areas that are of concern to an auditor.

Review relevant documents or portions of documents as necessary and
reference their use in support of facility operation. Record appropriate
dates and times associated with the occurrence on the documents
reviewed. Examples of documents include the following:

-   Operating logs

-   Correspondence

-   Inspection/surveillance records

-   Maintenance records

-   Meeting minutes

-   Computer process data

-   Procedures and instructions

-   Vendor manuals

-   Drawings and specifications

-   Functional retest specification and results

-   Equipment history records

-   Design basis information

-   Safety analysis report (SAR)/technical specifications

-   Related quality control evaluation reports

-   Operational safety requirements

-   Safety performance measurement system/occurrence reporting and
    processing system (SPMS/ORPS) reports

-   Radiological surveys

-   Trend charts and graphs

-   Facility parameter readings

-   Sample analyses and results (chemistry, radiological, air, etc.)

-   Work orders

### Follow-Up

After a reasonable period of time has elapsed, follow-up activities
should be performed to verify the effectiveness of the corrective
actions and how they were implemented. The verification should, at a
minimum, sample the corrective actions to determine whether the
problem/issue to be addressed has been resolved. The organization's
reporting systems (e.g., noncompliance tracking system, occurrence
reporting and processing system, external oversight reports and
regulatory violations, performance indicators) should be reviewed for
evidence of the problem (or a similar problem) recurring. The same
techniques used to conduct assessments may be used for verifying
corrective actions; however, there are several common ways to verify the
implementation of corrective actions, including the following:

-   Reassessment of the deficient areas

-   Review of new or revised quality-affecting documents such as
    manuals, procedures, and training records

-   Verification during the next scheduled assessment

-   Verification by conducting a surveillance covering the areas of
    concern

### Closure

Contractors send a letter to the directives management group (DMG)
requesting closure and stating that the corrective actions in the
implementation plan have been completed. Any ongoing activities
specified in the letter must be noted. The DMG coordinates approval of
the closure with the appropriate division of primary interest and the
contracting officer's representative.

Management responsible for the activities assessed is also responsible
for the development of effective corrective action of the problem areas
or deficiencies discovered during the assessment. At a minimum, the
corrective action should address

-   measures to correct each deficiency

-   identification of all root causes for significant deficiencies

-   determination of the existence of similar deficiencies

-   corrective actions to preclude recurrence of like or similar
    deficiencies

-   assignment of corrective action responsibility

-   completion dates for each corrective action

For independent assessments, the proposed corrective action should be
reviewed for concurrence by the assessment team leader. This will help
ensure that the planned actions will be effective in resolving the
problem areas and deficiencies reported by the assessment team.

h)  Describe the following assessment methods and the advantages or
    limitations of each method:

-   Document review

-   Observation

-   Interviews

### Document Review

Document review is used extensively during an assessment to substantiate
the information obtained during interviews and observation. During the
course of an assessment, questions may arise concerning what is heard
and seen. The review of documents, including logs, procedures, work
orders, and other data provides a method for answering these questions
and validating the assessment results. The drawback of document review
is that the accuracy of the records cannot be ascertained by review
alone. This technique should be combined with interviews, observation,
inspection, and/or performance testing to complete the picture of
performance. Records and documents should be selected carefully to
ensure they adequately characterize the program, system, or process
being assessed.

### Observation

Observation, the viewing of actual work activities, is often considered
the most effective technique for determining whether performance is
adequate. Assessors should understand the effect their presence has on
the person being observed and convey an attitude that is helpful,
constructive, positive, and unbiased. The primary goal during
observation is to obtain the most complete picture possible of the
performance, which should then be put into perspective relative to the
overall program, system, or process.

### Interviews

Interviews provide a means to verify the results of observation,
document review, inspection, and performance testing. In addition,
interviews allow the responsible person to explain and clarify those
results. The interview helps to eliminate misunderstandings about
program implementation and provides a venue where apparent conflicts or
recent changes can be discussed and the organization and program
expectations can be described. Tools developed during assessment
planning are used to prepare for the interview. Assessors should also
prepare questions in advance to keep the interview focused.

i)  Describe the action to be taken if the contractor challenges the
    assessment findings and explain how such challenges can be avoided.

Disputes over the assessment findings, the corrective action plan or its
implementation (such as timeliness or adequacy) must be resolved at the
lowest possible organizational level. The organization that disagrees
with the disposition of a given issue may elevate the dispute for timely
resolution. The organization that disagrees with the disposition of a
given issue must elevate the dispute in a step-wise manner through the
management hierarchy. The dispute must be raised via a deliberate and
timely dispute resolution process that provides each party with equal
opportunity for input and a subsequent opportunity to appeal decisions
up to the Secretary of Energy, if necessary.

## Nuclear safety specialists shall demonstrate a working-level knowledge of the Department of Energy (DOE)/facility contract provisions necessary to provide oversight of a contractor's operations.

a)  Describe the role of nuclear safety specialists in contractor
    oversight.

The nuclear safety specialist is an outside auditor and provides
regulatory oversight for the contractor facility. The nuclear safety
specialist approves the facility safety basis and ensures compliance
with the safety basis.

j)  Compare and contrast the following:

-   The Department of Energy's expectations of a Management and
    Operating (M&O) contractor

-   A Management and Operating contractor's expectations of the
    Department of Energy

The expectations that the M&O contractor and DOE have of each other are
identified in the contract that the M&O contractor and DOE agree upon.
The contract sets forth what DOE expects in regard to nuclear safety as
part of their oversight. The contractor gets the structure outlining how
DOE oversight is given and maintained at a facility.

k)  Identify the key elements and features of an effective Department of
    Energy and Management and Operating contractor relationship.

Agreements in Principle (AIPs) outline the general points of
understanding or boundaries between DOE and other parties, including M&O
contractors. AIPs set forth the macro-level agreements, and lay out the
foundation for more specific detailed negotiations on the finer points
of the covenant. AIPs are often precursors to memorandums of
understanding (MOUs), memorandums of agreement (MOAs), or actual
contracts.

An MOU is a written agreement broadly stating basic understandings of
tasks and describing a method for performing these tasks between the
Department and other Governmental entities, the private sector, and
educational institutions. A MOU is not a binding contract, nor can it be
used to obligate, transfer, or commit funds. Interagency Agreements
(IA), not MOUs, are used for monetary exchanges between DOE and other
federal agencies.

An MOA is an explicit agreement between DOE and its contractors in the
management and operations of a facility, site, or organization.

l)  Describe the responsibility nuclear safety specialists have
    associated with contractor compliance under the Price-Anderson
    Amendments Act.

Nuclear safety personnel should play an active role in the negotiation
and approval of management agreements that affect nuclear facilities.
Additional responsibilities include

-   reviewing applicable agreements to ensure the safety basis is not
    altered;

-   if a safety consideration is involved, reviewing the proposed
    agreement and making recommendations to the approval authority on
    the necessary requirements;

-   supplying information on applicable laws, regulations, and Orders to
    those negotiating contract requirements.

m)  Describe the role of nuclear safety specialists in the
    cost-plus-award fee process.

The nuclear safety specialist ensures that the contractor meets all the
requirements of the contract. They ensure that all nuclear safety issues
are addressed to the satisfaction of the DOE.

n)  Explain the responsibilities of nuclear safety specialists for DOE O
    442.1A, Employee Concerns Program, and the identification,
    reporting, reviewing, and documentation of employee concerns.

The nuclear safety specialist is to look into any concerns brought by
DOE employees or contractors in regard to employee, public, and
environmental safety. Any issue addressing safety from nuclear material
will be investigated thoroughly and any issues, concerns, or problems
identified will be dealt with immediately and the cognizant DOE
officials will be notified.

o)  Describe the differing professional opinions process used in your
    office.

    Note: This is a performance-based competency. The ability to
    complete this competency must be observed by a qualifying official.

## Nuclear safety specialists shall demonstrate a working-level knowledge of problem analysis principles and the techniques necessary to identify problems, potential causes, and corrective action(s) associated with nuclear safety issues at DOE Defense Nuclear Facilities.

a)  Describe and explain the application of problem analysis techniques
    including the following:

-   Root cause analysis

-   Causal factor analysis

-   Change analysis

-   Barrier analysis

-   Management oversight risk tree analysis

### Root Cause Analysis

Root cause analysis is an analysis method that includes the following
basic steps:

-   Identify the problem. Remember that actuation of a protective system
    constitutes the occurrence but is not the real problem; the
    unwanted, unplanned condition or action that resulted in actuation
    is the problem to be solved. For an example, dust in the air
    actuates a false fire alarm. In this case, the occurrence is the
    actuation of an engineered safety feature. The smoke detector and
    alarm functioned as intended; the problem to be solved is the dust
    in the air, not the false fire alarm. Another example is when an
    operator follows a defective procedure and causes an occurrence. The
    real problem is the defective procedure; the operator has not
    committed an error. However, if the operator had been correctly
    trained to perform the task and, therefore, could reasonably have
    been expected to detect the defect in the procedure, then a
    personnel problem may also exist.

-   Determine the significance of the problem. Were the consequences
    severe? Could they be next time? How likely is recurrence? Is the
    occurrence symptomatic of poor attitude, a safety-culture problem,
    or other widespread program deficiency? Base the level of effort of
    subsequent steps of your assessment upon the estimation of the level
    of significance.

-   Identify the causes immediately preceding and surrounding the
    problem.

-   Identify the reasons why the causes in the preceding identification
    step existed, working your way back to the root cause (the
    fundamental reason that, if corrected, will prevent recurrence of
    this and similar occurrences throughout the facility and other
    facilities under your control). This root cause is the stopping
    point in the assessment of causal factors. It is the place where,
    with appropriate corrective action, the problem will be eliminated
    and will not recur.

### Causal Factor Analysis

Causal factor analysis is used for multi-faceted problems or long,
complex causal factor chains. Cause and effects diagrams describe the
time sequence of a series of tasks and/or actions and the surrounding
conditions leading to an event. The event line is a time sequence of
actions or happenings while the conditions are anything that shapes the
outcome and ranges from physical conditions (such as an open valve or
noise) to attitude or safety culture. The events and conditions as given
on a chart describe a causal factor chain.

### Change Analysis

Change Analysis looks at a problem by analyzing the deviation between
what is expected and what actually happened. The evaluator essentially
asks what differences occurred to make the outcome of this task or
activity different from all the other times this task or activity was
successfully completed. This technique consists of asking the questions:
What? When? Where? Who? How? Answering these questions should provide
direction toward answering the root-cause determination question: Why?
Primary and secondary questions included within each category will
provide the prompting necessary to thoroughly answer the overall
question. Some of the questions will not be applicable to any given
condition. Some amount of redundancy exists in the questions to ensure
that all items are addressed. Several key elements to be addressed
include the following:

-   Consider the event containing the undesirable consequences.

-   Consider a comparable activity that did not have the undesirable
    consequences.

-   Compare the condition containing the undesirable consequences with
    the reference activity.

-   Set down all known differences whether they appear to be relevant or
    not.

-   Analyze the differences for their effects in producing the
    undesirable consequences. This must be done with careful attention
    to detail, ensuring that obscure and indirect relationships are
    identified (e.g., a change in color or finish may change the heat
    transfer parameters and consequently affect system temperature).

-   Integrate information into the investigative process relevant to the
    causes of, or the contributors to, the undesirable consequences.

Change analysis is a good technique to use whenever the causes of the
condition are obscure, you do not know where to start, or you suspect a
change may have contributed to the condition. Not recognizing the
compounding of change (e.g., a change made five years previously
combined with a change made recently) is a potential shortcoming of
change analysis. Not recognizing the introduction of gradual change as
compared with immediate change also is possible. This technique may be
adequate to determine the root cause of a relatively simple condition.
In general, though, it is not thorough enough to determine all the
causes of more complex conditions.

### Barrier Analysis

There are many things that should be addressed during the performance of
a Barrier Analysis. The questions listed below are designed to aid in
determining what barrier failed, thus resulting in the occurrence.

-   What barriers existed between the second, third, etc.,
    condition/situation and the second, third, etc., problem?

-   If there were barriers, did they perform their functions? How?

-   Did the presence of any barriers mitigate or increase the occurrence
    severity? Why?

-   Were any barriers not functioning as designed? Why?

-   Was the barrier design adequate? How?

-   Were there any barriers in the condition/situation source(s)? Did
    they fail? Why?

-   Were there any barriers on the affected component(s)? Did they fail?
    Why?

-   Were the barriers adequately maintained?

-   Were the barriers inspected prior to expected use?

-   Were any unwanted energies present? Why?

-   Is the affected system/component designed to withstand the
    condition/situation without the barriers? How?

-   What design changes could have prevented the unwanted flow of
    energy? Why?

-   What operating changes could have prevented the unwanted flow of
    energy? Why?

-   What maintenance changes could have prevented the unwanted flow of
    energy? Why?

-   Could the unwanted energy have been deflected or evaded? How?

-   What other controls are the barriers subject to? How?

-   Was this event foreseen by the designers, operators, maintainers,
    anyone?

-   Is it possible to have foreseen the occurrence? How?

-   Is it practical to have taken further steps to have reduced the risk
    of the occurrence?

-   Can this reasoning be extended to other similar systems/components?

-   Were adequate human factors considered in the design of the
    equipment?

-   What additional human factors could be added? Should be added?

-   Is the system/component user friendly?

-   Is the system/component adequately labeled for ease of operation?

-   Is there sufficient technical information for operating the
    component properly? How do you know?

-   Is there sufficient technical information for maintaining the
    component properly? How do you know?

-   Did the environment mitigate or increase the severity of the
    occurrence? How?

-   What changes were made to the system/component immediately after the
    occurrence?

-   What changes are planned to be made? What changes might be made?

-   Have these changes been properly and adequately analyzed for effect?

-   What related changes to operations and maintenance have to be made
    now?

-   Are expected changes cost effective? Why? How do you know?

-   What would you have done differently to have prevented the
    occurrence, disregarding all economic considerations (as regards
    operation, maintenance, and design)?

-   What would you have done differently to have prevented the
    occurrence, considering all economic concerns (as regards operation,
    maintenance, and design)?

Barrier analysis is a systematic process that can be used to identify
physical, administrative, and procedural barriers or controls that
should have prevented the occurrence. This technique should be used to
determine why these barriers or controls failed and what is needed to
prevent recurrence.

### Management Oversight Risk Tree Analysis

Management oversight risk tree (MORT) analysis is used to prevent
oversight of the identification of causal factors. It lists, on the left
side of the tree, specific factors relating to the occurrence, and on
the right side of the tree, the management deficiencies that permit
specific factors to exist. The management factors all support each of
the specific barrier/\
control factors. Included is a set of questions to be asked for each of
the factors on the tree. As such, it is useful in preventing oversight
and ensuring that all potential causal factors are considered. It is
especially useful when there is a shortage of experts to ask the right
questions. However, because each of the management factors may apply to
the specific barrier/control factors, the direct linkage or relationship
is not shown but is left up to the analyst. For this reason, events and
causal factor analysis and MORT should be used together for serious
occurrences: one to show the relationship, the other to prevent
oversight.

p)  Describe the following types of investigations and discuss an
    example of the application of each:

-   Type A

-   Type B

A complete description of each type of investigation can be found in
[DOE G
225.1A-1](http://www.directives.doe.gov/cgi-bin/explhcgi?qry1399419052;doe-83).

q)  Compare and contrast immediate, short term, and long term actions
    taken as the result of problem identification or an occurrence.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

r)  Given a nuclear safety event and/or occurrence data, apply problem
    analysis techniques and identify the problems and how they might
    have been avoided.

Note: This is a performance-based competency. The ability to complete
this competency must be observed by a qualifying official.

s)  Describe various data gathering techniques and the use of
    trending/history when analyzing problems.

Methods used in data gathering at facilities may include those listed
below.

### Walkthroughs

A walkthrough of a facility is performed to observe operations, system
configurations, and physical conditions.

### Interviews

Conducting interviews of management, operators, technicians, and
engineers at a facility gives an analyst an idea of the overall
operations and level of knowledge at the facility.

### Document Review

A review of documentation is performed to ensure that all requirements
are being met. This review encompasses all requirements for the safety
basis.

Trending/history is used to quantify accident scenarios, calculate
failure rates, or identify faulty equipment due to a history of high
failure.

# Selected Bibliography and Suggested Reading

10 CFR 261 "Identification and Listing of Hazardous Waste." July 1,
2004.

10 CFR 830 "Nuclear Safety Management*.*" January 10, 2001.

10 CFR 835 "Occupational Radiation Protection." January 1, 2004.

29 CFR 1910 "Occupational Safety and Health Standards." July 1, 2004.

29 CFR 1910.35 "Compliance with NFPA 101-2000, Life Safety Code."
November 7, 2002.

29 CFR 1910.120 "Hazardous Waste Operations and Emergency Response*.*"
July 1, 2004.

29 CFR 1960 "Basic Program Elements for Federal Employee Occupational
Safety and Health Related Matters." July 1, 2004.

48 CFR 970.5204-2 "Laws, Regulations, and DOE Directives." October 1,
2004.

48 CFR 970.5223-1 "Integration of Environment, Safety, and Health into
Work Planning and Execution." October 1, 2004.

Defense Nuclear Facilities Safety Board. *Recommendation 92-2.* 1992.

ISO 14001 *Environmental Management Systems.* International Organization
for Standardization, 1996.

Office of Management and Budget. OMB Circular A-119, *Federal
Participation in the Development and Use of Voluntary Consensus
Standards and in Conformity Assessment Activities.* February 10, 1998.

U.S. Congress. *Clean Air Act.* 1993.

U.S. Congress. *Clean Water Act.* 1977.

U.S. Congress. *Comprehensive Environmental Response, Compensation, and
Liability Act-Superfund Act.* 1980.

U.S. Congress. *Emergency Planning and Community Right-to-Know Act.*
1986.

U.S. Congress. *Federal Facilities Compliance Act.* 1992.

U.S. Congress. *National Environmental Policy Act.* 1969.

U.S. Congress. *National Technology Transfer and Advancement Act.* 1995.

U.S. Congress. *Price-Anderson Amendments Act*. 1988.

U.S. Congress. *Resource Conservation and Recovery Act.* 1976.

U.S. Congress. *Superfund Amendment Reauthorization Act.* 1974.

U.S. Congress. *Toxic Substances Control Act.* 1976.

U.S. Department of Energy. DOE Guide 420.1-1, *Nonreactor Nuclear Safety
Design Criteria and Explosive Safety Criteria Guide for Use with DOE O
420.1 Facility Safety.* March 28, 2000.

U.S. Department of Energy. DOE Guide 421.1-2, *Implementation Guide for
Use in Developing Documented Safety Analyses to Meet Subpart B of 10 CFR
830.* October 24, 2001.

U.S. Department of Energy. DOE Guide 423.1-1, *Implementation Guide for
Use in Developing Technical Safety Requirements.* October 24, 2001.

U.S. Department of Energy. DOE Guide 424.1-1, *Implementation Guide for
Use in Addressing Unreviewed Safety Question Requirements.* October 24,
2001.

U.S. Department of Energy. DOE Guide 450.3-1, *Documentation for Work
Smart Standards Applications: Characteristics and Considerations.*
February 1, 1997.

U.S. Department of Energy. DOE Manual 231.1-2, *Occurrence Reporting and
Processing of Operations Information*. August 19, 2003.

U.S. Department of Energy. DOE Order 414.1B, *Quality Assurance*. April
29, 2004.

U.S. Department of Energy. DOE Order 420.1A, *Facility Safety.* May 20,
2002.

U.S. Department of Energy. DOE Order 433.1, *Maintenance Management
Program for DOE Nuclear Facilities.* June 1, 2001.

U.S. Department of Energy. DOE Order 435.1, *Radioactive Waste
Management.* August 28, 2001.

U.S. Department of Energy. DOE Order 451.1B, *National Environmental
Policy Act Compliance Program.* September 28, 2001.

U.S. Department of Energy. DOE Order 460.2A, *Departmental Materials
Transportation and Packaging Management.* December 22, 2004

U.S. Department of Energy. DOE Order 461.1A, *Packaging and Transfer or
Transportation of Materials of National Security Interest.* April 26,
2004

U.S. Department of Energy. DOE Order 5490.19, *Conduct of Operations for
DOE Facilities.* October 23, 2001.

U.S. Department of Energy. DOE Policy 411.1, *Safety Management
Functions, Responsibilities, and Authorities.* January 28, 1997.

U.S. Department of Energy. DOE Policy 450.4, *Safety Management System
Policy.* October 15, 1996.

U.S. Department of Energy. DOE-HDBK-1010-92, *DOE Fundamentals Handbook,
Classical Physics.* April 1999.

U.S. Department of Energy. DOE-HDBK-1013/2-92, *DOE Fundamentals
Handbook, Instrumentation and Control,* volume 2 of 2. April 1999.

U.S. Department of Energy. DOE-HDBK-1016/1-93, *DOE Fundamentals
Handbook, Engineering Symbology, Prints, and Drawings,* volume 1 of 2.
April 1999.

U.S. Department of Energy. DOE-HDBK-1019/1-93, *DOE Fundamentals
Handbook, Nuclear Physics and Reactor Theory,* volume 1 of 2. April
1999.

U.S. Department of Energy. DOE-HDBK-1019/2-93, *DOE Fundamentals
Handbook, Nuclear Physics and Reactor Theory,* volume 2 of 2. April
1999.

U.S. Department of Energy. DOE-STD-1020-2002, *Natural Phenomena Hazards
Design and Evaluation Criteria for Department of Energy Facilities*.
January 2002.

U.S. Department of Energy. DOE-STD-1027-92, *Hazard Categorization and
Accident Analysis Techniques for Compliance with DOE Order 5480.23,
Nuclear Safety Analysis Reports.* December 2002.

U.S. Department of Energy. DOE-STD-1104-96, *Review and Approval of
Nuclear Facility Safety Basis Documents (Documented Safety Analysis and
Technical Safety Requirements).* May 2002.

U.S. Department of Energy. DOE-STD-1098-99, *Radiological Control.* June
2004.

U.S. Department of Energy. DOE-STD-1183-2001, *Nuclear Safety Standard.*
April 2004.

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