Bulletin 79-01B: Supplement No. 2, Environmental Qualification of Class IE Equipment
IEB 79-01B, Supple. 2
NUCLEAR REGULATORY COMMISSION
OFFICE OF INSPECTION AND ENFORCEMENT
WASHINGTON, D.C. 20555
September 30, 1980
IE Supplement No. 2 to Bulletin 79-01B: ENVIRONMENTAL QUALIFICATION OF CLASS
Enclosed are the generic questions and answers which resulted from
NRC/Licensee meetings in NRC Regional Offices during the week of July 14,
1980 regarding environmental qualification of Class 1E equipment in use at
power reactor facilities. These answers address specific questions asked
during the meetings. Due to the generic nature of some of these questions,
the staff is issuing them as a bulletin supplement. The regional meetings
highlighted the fact that in some cases, the scope and depth of the 79-01B
review was not clear to licensees. Therefore, these answers may affect your
79-01B submittal. These submittals are required by a separate order to be
completed by November 1, 1980.
Some answers given in Supplement No. 1 to IEB-79-01B are superseded by these
answers. For example, in Bulletin Supplement No. 1, issued on February 29,
1980, the answer to question No. 5 specified that TMI lessons learned
equipment was not included in the review. However, due to the extension of
the response date from April 14, 1980 to November 1, 1980, this equipment is
now being addressed since its installation is either complete or required
before the issuance of the February 1, 1981 SER. (See Question No. 21 of
No specific response is requested by this Supplement; however, all answers
contained in the enclosure to this Supplement should be carefully reviewed
and considered for applicability in your response to IEB 79-01B.
IE Bulletin No. 79-01B was issued under a blanket GAO clearance (B18O225
(ROO72); clearance expired July 31, 1980) specifically for identified
generic problems. Supplement No. 2 to Bulletin 79-01B is for information,
hence no GAO clearance is required.
1. Generic Questions and Answers
to IEB-79-01B and Memorandum
and Order (CLI-80-21) dated
May 23, 1980
Q.1 Define the scope of review with respect to the June 1982 deadline.
What is required beyond the June 1982 date for qualification?
By June 30, 1982, all safety-related electrical equipment
potentially exposed to a harsh environment in nuclear generating
stations, licensed to operate on or before June 30, 1982, shall be
qualified to either the DOR guidelines or NUREG-0588 (as
applicable). Safety-related electrical equipment are those
required in bringing the plant to a cold shutdown condition and to
mitigate the consequences of the accident. The qualification of
safety-related electrical equipment to function in environmental
extremes, not associated with accident conditions, is the
responsibility of the licensee to evaluate and document in a form
that will be available for the NRC to audit. Qualification to
assure functioning in mild environments must be completed by June
The qualification schedules for consideration of the dynamic
loading of safety-related equipment (electrical and mechanical)
and the environmental qualification review of mechanical equipment
are being developed. It is the intention of the staff to initiate
this effort as soon as possible.
Q.2 Clarify the required submittal dates for ORs, NTOLs, and CPs. What
about OLs whose 100% license is not expected by June 1982?
A.2 The required schedule for submitting information in response to
the Commission Order and Memorandum (CLI-80-21) is provided below.
Plants who have received an operating license, either for full or
limited power operation, are required to meet the schedule for
operating reactors. Plants who have committed to the NRC, to meet
schedules in advance of those provided below are required to meet
that commitment. In all cases, plants are required to have their
equipment fully qualified to the applicable standards either by
June 30, 1980, or by the time the operating license is granted,
whichever comes later.
Operating Reactors and NTOL (operating license expected by
February 1, 1981)
- Submittal to be received no later than November 1, 1980
OLs (operating license expected by June 30, 1982)
- Submittal to be received no later than 4 months prior to
issuance of operating license
OLs and CPs (operating license expected after June 30, 1982)
- Submittal to be received no later than 6 months prior to
issuance of operating license.
- 2 -
Q.3 Define the requirements and applicable criteria for ORs, NTOLs,
and OLs. Specifically address the NTOLs whose CP SER is prior to
July. 1974 and after July 1974. Can a CP whose SER is prior to
1974 use the DOR guidelines?
A.3 Table 1 describes the application of each document. All operating
reactors as of May 23, 1980, will be evaluated against the DOR
guidelines. In cases where the DOR guidelines do not provide
sufficient detail, but NUREG-0588 Category II does, NUREG-0588
will be used.
ORs OLs Cps
DOR GUIDELINES CP SER CP SER
Before 7/1/74 After 7/1/74
USE NUREG-0588 NUREG-0588(CAT.II) NUREG-0588(CAT.I) NUREG-0588(CAT.I)
REPLACEMENT COMPONENTS NEW RULE WHEN
USE NUREG-0588 (CAT.I) IN EFFECT
All plants licensed after May 23, 1980, shall conform to
NUREG-0588. In accordance with Regulatory Guide 1.89, all such
operating licenses for facilities whose construction permit SER is
dated July 1, 1974 or later, are to be reviewed against IEEE Std.
323-1974. Thus, for these licensees, the operating license
applicant is to qualify equipment to the Category I column in
NUREG-0588. For operating licenses issued after May 23, 1980,
whose construction permit SER is dated before July 1, 1974, the
operating license applicant is to qualify equipment to at least
Category II column of NUREG-0588; unless the licensee made
commitment in the construction permit record to use the 1974
standard, or unless the operating licensee application record
indicates that the 1974 standard is to be used, in such cases
Column I of NUREG-0588 is to be used.
While there are differences between the Category II column of
NUREG-0588 and the DOR guidelines, the differences are in details
and in the optional part of the documents. The minimum
requirements set forth by these documents are general and
compatible. Thus, the minimum standards set by either of the two
documents are equally applicable to ORs and NTOLs.
Q.4 Clarify the reporting requirements for LERS with respect to Part
50.55e vs 79-01B.
- 3 -
Are only those items, known to be unqualified, immediately
reportable? Are items, for which there are no data or for which
there are insufficient data, open items to be resolved, but are
not immediately reportable?
A.4 The requirement for reporting in IEB 79-01B does not change the
reporting requirements defined in the license conditions. In
general, CPs should report via 50.55e. Operating plants should use
When a determination has been made that reasonable assurance does
not exist to ensure that the Class IE electrical equipment
component(s) can perform their safety-related function, that is
reportable. Inadequate or no data are factors in this
determination. The time and technical judgements required to make
the determination should be based on the significance of this
specific equipment, components, and the discrepancies.
Q.5 How does the "Q" list review interface with the EQB effort? Can
the NRC provide more specific guidance on how to pick out the
required safety-related equipment?
A.5 The "Q" list provides a source from which the required equipment
may be selected. The information required to be submitted by
November 1, 1980, is for safety-related electrical equipment
potentially exposed to a harsh environment resulting from an
accident. Safety-related equipment are those required to help
bring the plant to cold shutdown and to mitigate the accident
(LOCA, HELB inside or outside-containment). "Mitigate" includes
safety-related functions such as containment isolation, and
prevention of significant release of radioactive material.
In order to "pick out" the safety-related equipment, the licensee
should generate a list of safety functions typically performed by
plant safety systems. Examples are listed in Table II. For each
safety function identified in Table II, list the systems,
subsystems, or components assumed available in the plant FSAR or
emergency procedures to perform that function during a LOCA or any
HELB inside or outside containment. If a plant specific safety
function not listed in Table II is identified, that function and
the corresponding systems or equipment to perform the function
should be added to the licensee's list.
The systems and equipment identified above should be included
regardless of the original classification when the plant received
its operating license; i.e., some control grade equipment will
probably be named in emergency procedures. However, if plant
emergency procedures specify a preferred mode of accident
mitigation involving equipment recognized by the licensee as
unlikely to meet environmental qualification criteria, an
alternate mode of performing the safety function and qualifiable
equipment may be identified. In such cases, the emergency
procedures must clearly indicate how the
- 4 -
operator is to use environmentally qualified safety-related
display instrumentation to diagnose failure to perform such safety
Plant emergency procedures typically include provisions for the
operator to sample or monitor radioactivity levels or combustible
gas levels, to confirm that valves are in the correct position, to
monitor flow or temperature, etc. Some of these functions are
essential for correct operator action, to mitigate accidents, and
prevent radioactive releases. When this is the case, the radiation
sensors, valve position indicators, pressure transmitters,
thermocouples, etc., should be qualified to function in the
relevant accident environment.
Licensees should, therefore, review their emergency procedures to
determine the electrical components needed to perform the
functions of Safety-Related Display Information, Post Accident
Sampling and Monitoring, and Radiation Monitoring. When equipment
implied by the emergency procedures is not listed, justification
must be provided that failure of such equipment would not prevent
accident mitigation or release of radioactivity.
Equipment now indicated in emergency procedures in response to
TMI-2 Lessons Learned should be listed. Equipment which is or will
be installed due to TMI Lessons Learned should be addressed
similar to other existing safety-related equipment (e.g.,
saturation meter, sump level indicators, torus water volume,
The licensee should document anticipated service conditions in
every portion of the plant where the environment could be
influenced by the accident or its consequences. These service
conditions should also be correlated with the safety-related
systems and subsystems identified above. Whenever an item of
safety-related equipment may be located in an environment outside
the range of normal conditions, due to the harsh environment
resulting from the accident, and the equipment is needed to
mitigate the consequences of the accident, place it on the list of
equipment in a potentially hostile environment. Conclusions which
show that equipment is unqualified should include a basis for
continued plant operation.
TYPICAL EQUIPMENT/FUNCTIONS NEEDED FOR
MITIGATION OF A LOCA OR MSLB ACCIDENT
Engineered Safeguards Actuation
Main Feedwater Shutdown and Isolation
- 5 -
Emergency Core Cooling
Containment Heat Removal
Containment Fission Product Removal
Containment Combustible Gas Control
Containment Radiation Monitoring
Control Room Habitability Systems (e.g., HVAC, Radiation Filters)
Ventilation for Areas Containing Safety Equipment
Post Accident Sampling and Monitoring
Safety Related Display Instrumentation
(1) These systems will differ for PWRs and BWRs and for older and newer
plants. In each case, the system features which allow for transfer to
recirculation cooling mode and establishment of long-term cooling with
boron precipitation control are to be considered as part of the system
to be evaluated.
(2) Emergency shutdown systems include those systems used to bring the
plant to a cold shutdown condition following accidents which do not
result in a breach of the reactor coolant pressure boundary together
with a rapid depressurization of the reactor coolant system. Examples
of such systems and equipment are the RHR system, PORVs, RCIC,
pressurizer sprays, chemical and volume control system, and steam dump
(3) More specific identification of these types of equipment can be found
in the plant emergency procedures.
Q.6 NUREG-0588 was issued for comment. Will any changes impact the
requirements established by the Commission memorandum and order?
Will the daughter standards referenced be corrected/changed?
A.6 The requirement established by the Commission memorandum and order
will not change as a result of comments on NUREG-0588. No
substantive changes are anticipated in NUREG-0588 or in referenced
daughter standards. A revision is anticipated, making corrections.
Q.7 Can IEEE Std. 650 (Standards for Qualification of Class IE static
battery chargers and invertors for nuclear power generating
stations) be used for qualifying the balance of plant components
which are not exposed to harsh environments?
A.7 The methods and procedures relating to design stress analysis,
aging of electrical/electronic components and the stress test
identified in this standard are acceptable for qualifying the
balance of plant components which are not exposed to harsh
- 6 -
Q.8 Provide the staff's definition of "central location" for
qualification documentation. What documentation is expected to be
maintained? Will it be acceptable to maintain summary test reports
at the utility central file and provide a reference to the NSSS
Vendor's file for the actual test reports? Does NRC require test
reports to be submitted to support qualification?
A.8 The central location should be at the utilities corporate
headquarters or plant site. Both the DOR guidelines and NUREG-0588
specify that sufficient information must be available to verify
that the safety-related electrical equipment has been qualified in
accordance with the guidance and requirements. Details for the
information and documentation required for type tests, operating
experience, analysis, and extrapolation of test data from
operating experience are provided in Section 5 of NUREG-0588 and
Section 8 of IEEE Std. 323-74.
The staff will accept summary test reports maintained at the
utility's central file which reference the actual test reports and
data available in a single location at the NSSS vendor's facility.
The Licensee/Applicant must make the determination that necessary
information and documentation, to support qualification of
equipment, is in conformance with DOR guidelines and NUREG-0588.
This vendor information file must be maintained current, auditable
and available throughout the life of the referencing plant.
Test reports are not required to be submitted. Test report
references must be included in the plant submittals and these
reports must be available for staff review on demand.
Q.9 The staff was directed to codify, by Technical Specification, some
of the requirements of the Order. Can you give some of the details
of this requirement, how the staff expects to meet this directive
A.9 The staff has proposed to the Commission changes to the Technical
Specifications (e.g., Appendix A Section 6.10 of the license)
which require the establishment and maintenance of a centrally
located file which will contain the information necessary to
verify the qualification adequacy of all safety-related electrical
Q.10 With respect to the NRC data base, how will utilities address and
obtain information from it?
A.10 The industry access method for the data base will be addressed in
the final stages of system development. This information should be
available by mid-1981. Licensees will be informed at that time.
Q.11 How should submittals containing data and qualification
information be submitted? What format should we use if we have
several facilities at different stages (OR, NTOL, CP)?
- 7 -
A.11 The qualification information and data should be submitted with
the appropriate officer's notarized sworn statements. The format
for the data should be in accordance with the format provided in
I&E Bulletin 79-01B or the letters provided to the plants in the
SEP program. Either format is acceptable.
Q.12 Is testing required of equipment which completes its
safety-related function within the first minute(s) of a LOCA or
HELB? (E.g., nuclear instrumentation or other instruments
providing RPS inputs, isolation valves, etc.)
A.12 The staff does not require that the nuclear instrumentation and
its associated components be environmentally qualified for a LOCA
or HELB. The nuclear instrumentation system is used for transient
conditions but is not required for a LOCA or HELB.
The staff does require that equipment designed to perform its
safety-related function within a short time into an event be
qualified for a period of at least 1 hour in excess of the time
assumed in the accident analysis. The staff has in indicated that
time is the most significant factor in terms of the margins
required to provide an acceptable confidence level that a
safety-related function will be completed. Our judgment of at
least 1 hour is based on the acceptance of a type test for a
single unit and the spectrum of accidents (small and large breaks)
bounded by the single test: Also see answer to question 21.
Q.13 Testing is currently being performed on some equipment, and
contracts have been issued for testing additional equipment
specifying conformance to IEEE Std 323-1971. For sequential
testing, how do we factor in aging? If early test failure occurs
due to "non E-Q" mechanisms, can the test be extrapolated using
A.13 Sequential testing requirements are specified in NUREG-0588 and
the DOR guidelines. Licensees must follow the test requirements of
the applicable document.
1. If the test has been completed without aging in sequence,
justification for such a deviation must be submitted.
2. If testing of a given component has been scheduled but not
initiated, the test sequence/program should be modified to
3. Test programs in progress should be evaluated regarding the
ability to comply by incorporating aging in the proper
sequence. These would then fall in the first or second
When a failure occurs due to a non-EQ related mechanism,
acceptability of analysis to extrapolate the test data would be
dependent on several considerations (e.g., the specific function
being demonstrated, the
- 8 -
failure mechanism, when the failure occurred, etc,) may be very
difficult to achieve. If such a failure occurs it may be more
prudent to correct the failure and continue with the test.
Q.14 What is the definition of harsh environment? How are the
environmental profiles defined outside containment?
A.14 Harsh environment is defined by the limiting conditions, as
specified in IE Bulletin 79-01B, resulting from the entire
spectrum of LOCAs HELBs. Specifically, the harsh environment from
a LOCA considers the worst parameters resulting over the spectrum
of postulated break sizes, break locations and single failures.
Similarly, the HELBs inside and outside of containment consider
the spectrum of breaks including main steam and feedwater line
breaks. The parameters to be considered are: temperature,
pressure, humidity, caustic spray, radiation, duration of
exposure, aging and submergence. Mechanical and flow-induced
vibrations and seismic effects will be considered separately.
Environmental profiles for HELB outside of containment have not
been generically established due to the uniqueness of each
facility. Service conditions for areas outside containment exposed
to a HELB must be evaluated on a plant-by-plant basis. Each of
the parameters listed above must be considered. Acceptable
engineering methods should be used for this calculation.
Temperature and pressure history may be available from earlier
HELB evalations. The radiation source terms are discussed under
Question 18 below. Further guidance for selecting the piping
systems and conducting the review are delineated in Regulatory
Guide 1.46 and Standard Review Plans 3.6.1 and 3.6.2.
Q.15 The DOR Guidelines and NUREG-0588 give time and temperature
parameters. Can we use different values of these parameters? Will
plant-specific profiles still be with the guidance provided?
Q.15 For minimum high temperature conditions in pressure-suppression-
type containments, we do not require that 340 F for 6 hours be
used for BWR drywells or that 340 F for 3 hours be used for PWR
ice condenser lower compartments. These values are a screening
device, per the Guidelines, and can be used in lieu of a
plant-specific profile, provided that expected pressure and
humidity conditions as a function of time are accounted for.
In general, the containment temperature and pressure conditions as
a function of time should be based on analyses in the FSAR.
However, these conditions should bound those expected for coolant
and steam line breaks inside the containment with due
consideration of analytical uncertainties. The steam line break
condition should include superheated conditions: the peak
temperature, and subsequent temperature/pressure profile as a
function of time. If containment spray is to be used, the impact
of the spray on required equipment should be accounted for.
- 9 -
The adequacy of a plant-specific profile is dependent on the
assumption and design considerations at the time the profiles were
developed. The DOR guidelines and NUREG-0588 provide guidance and
considerations required to determine if the plant-specific
profiles encompass the LOCA and HELB inside containment.
Q.16 Could you elaborate on what the staff expects with regard to
If parts or subcomponents are purchased from a vendor who does not
have a quality assurance program, can it be qualified to meet IEEE
Std. 323-74 requirements?
A.16 The QA programs should accommodate any increased scope due to the
new environmental qualification documentation requirements.
Procedures incorporated by the licensee for data acquisition
should be documented and available for staff review upon request.
Requirements for QA programs are provided in Part 50, Appendix B,
of the Code of Federal Regulations.
Part 50, Appendix B of the Code of Federal Regulations states that
the applicant/licensee shall be responsible for the establishment
and execution of quality assurance programs. Specifically in
purchasing parts or components, it is the responsibility of the
licensee/applicant to ensure that the applicable quality assurance
procedures for their plant are met.
In determining the qualification status of existing equipment
purchased from a vendor, where a QA program did not exist, the
utility should consider the following:
1. The complexity of design, complexity of manufacturing
process, and end use.
2. Past performance of vendor.
3. Past operating history of products, especially similar
products, made by vendor.
4. Procedures, equipment, and results of environmental
qualification testing relative to those for other equipment
for which a QA program was applied.
Q.17 Define the requirements for "replacement parts." Are they the same
for "spare" parts? Clearly discuss the alternatives for existing
inventories of parts/components. If equipment is ordered to meet
IEEE Std. 323-1974 standard but lead time exceeds June 1982, can
we use IEEE Std. 323-1971 qualified components in the interim?
A.17 The requirements for "replacement" and "spare" parts are the same
for the purposes of complying with the Commission order and
- 10 -
memorandum. After May 1980, all parts used to replace presently
installed parts shall be qualified to Category I of NUREG-0588
"unless there are sound reasons to the contrary." Nonavailability
and/or the fact that the part to be used as a replacement is a
spare part purchased prior to May 23, 1980, and is in stock are
among the factors to be considered in weighing whether there are
"sound reasons to the contrary." All replacement parts shall as a
minimum conform to the requirements described in the answer to
question 3. Justification for deviation from Category I or
NUREG-0588 shall be documented by the licensee and records shall
be available for audit, upon request by the NRC.
Q.18 DOR Guidelines, NUREG-0588 and NUREG-0578, define or give guidance
for calculating radiation source terms. However, since one is more
restrictive than the other, which do we use?
A.18 Both the DOR guidelines and NUREG-0588 are similar in that they
provide the methods for determining the radiation source term when
considering LOCA events inside containment (100% noble gases/50%
iodine/1% particulates). These methods consider the radiation
source term resulting from an event which completely depressurizes
the primary system and releases the source term inventory to the
NUREG-0578 provides the radiation source term to be used for
determining the qualification doses for equipment in close
proximity to recirculating fluid systems inside and outside of
containment as a result of LOCA. This method considers a LOCA
event in which the primary system may not depressurize and the
source term inventory remains in the coolant.
NUREG-0588 also provides the radiation source term to be used for
qualifying equipment following non-LOCA events both inside and
outside containment (10% noble gases/10% iodine/O% particulates).
When developing radiation source terms for equipment
qualification, the licensee must ensure consideration is given to
those events which provide the most bounding conditions. The
following table summarizes these considerations:
LOCA NON-LOCA HELB
Outside Containment NUREG-0578 NUREG-0588
in RCS) in RCS)
- 11 -
Inside Containment Larger of
in containment) in RCS)
Q.19 Can gamma equivalents be used rather than beta exposure for
A.19 Yes. Gamma equivalents may be used when consideration of the
contributions of beta exposure have been included in accordance
with the guidance given in the DOR guidelines and NUREG-0588.
Cobalt 60 is one acceptable gamma radiation source for
environmental qualification of safety-related equipment. Cesium
137 may also be used.
Q.20 If a piece of equipment will become submerged after completing its
required action, must it be qualified for submergence?
A.20 If the equipment (1) meets the guiadance and requirements of the
DOR guidelines or NUREG-0588 for the LOCA and HELB (small and
large breaks) accidents and (2) licensees demonstrate that its
failure will not adversely affect any safety-related function or
mislead the operator after submergence, the equipment could be
considered exempt from that portion (submergence) of
Q.21 What qualification is required of Reactor Pressure Vessel internal
instrumentation (e.g., thermocouples) and new instruments required
as the result of TMI Lessons Learned?
A.21 TMI Lessons Learned instrumentation will be considered in the
February 1, 1981 SER. This equipment is subject to the same
requirements as other safety-related electrical equipment. The
guidance and requirements of NUREG-0588 referenced daughter
standards, and Reg Guides will be used by the staff in assessing
the adequacy of the qualification information. The in-core
environment should consider the worst source term for radiation
effects, the worst humidity for the corresponding temperature, and
high temperatures consistent with that of a damaged core.
Q.22 Is qualification "by use" an acceptable method (e.g., CRDM's in
A.22 Qualification by use has limited application. Often the equipment
has never seen the harsh environment and no conclusions can be
drawn as to its operability in a harsh environment. Some
- 12 -
based on operating experience is a recognized method subject to
the requirements of NUREG-0588 and the Guidelines. Credit can be
taken for the natural aging of the equipment and for the location
of the equipment or other portions of the overall qualification
Q.23 How long should "long term" equipment be qualified for
A.23 "Long term" for the purpose of qualifying equipment for a harsh
environment is variable. A determination of "long term" for
qualification of equipment should be based on the considerations
listed below for each postulated accident scenario. Justification
for the value used should be provided with the equipment
1. The time period over which the equipment is required to bring
the plant to cold shutdown and to mitigate the consequences
of the accident.
2. The ability to change, modify or add equipment during the
course of the accident or in mitigating its effects which
will provide the same safety-related function.
Q.24 Why do we want component surface temperature rather than the bulk
A.24 Temperature measurements are required during the qualification
testing to establish that the component was subjected to the most
severe temperature environment postulated to occur. These
temperature measurements are required to be made as close to the
component surface as practicable to ensure that they are
representative of the environment in which the component is
tested. The surface temperature of the component, although not
specifically required, is considered to be a conservative
measurement of the test temperature environment.
Page Last Reviewed/Updated Friday, May 22, 2015