Part 50 section 7 final supporting statement

FINAL SUPPORTING STATEMENT FOR
10 CFR PART 50
STANDARDS FOR LICENSES, CERTIFICATIONS AND REGULATORY APPROVALS
SECTION 7
(50.44(c), Hydrogen Control Requirements;
Appendix K, 50.46, Acceptance Criteria for Emergency Core Cooling Systems (ECCS);
50.47, 50.54(q & t), Appendix E, Emergency Planning;
50.48, Appendix R, Fire Protection;
50.49, Environmental Qualification;
50.120, Training & Qualification of Nuclear Power Plant Personnel;
Appendix J, Primary Reactor Containment Leakage Testing For
Water-Cooled Power Reactors;
Appendix S, 50.54(ff), Earthquake Engineering Criteria for Nuclear Power Plants)
3150-0011
ABSTRACT
Title 10 of the Code of Federal Regulations (10 CFR), Part 50 contains the Nuclear Regulatory
Commission’s (NRC) requirements and provisions for “Standards for Licenses, Certifications and
Regulatory Approvals.
The U.S. Nuclear Regulatory Commission (NRC) is authorized by Congress to have responsibility
and authority for the licensing and regulation of nuclear power plants, research/test facilities, fuel
reprocessing plants and other utilization and production facilities licensed pursuant to the Act. To
meet its responsibilities, the NRC conducts a detailed review of all applications for licenses to
construct and operate such facilities. The purpose of the detailed review is to ensure that the
proposed facilities can be built and operated safely at the proposed locations, and that all
structures, systems and components important to safety will be designed to withstand the effects
of postulated accident conditions, without undue risk to the health and safety of the public.
Under 10 CFR Part 50, before a company can build a nuclear power plant at a particular site, it
must obtain a construction permit from the NRC. Subsequently, the company must obtain an
operating license from the NRC before it can operate the plant. The decision by the NRC as to
whether to approve a company's application for a construction permit or an operating license is
based largely on the NRC staff's detailed review of the information provided by the company as
part of its application. Information provided by the applicant as part of the application is crucial to
the licensing process as it provides the NRC with the information it needs to make a decision with
regard to the proposed plant's impact on the public's health and safety and the environment.
These regulations affect 89 licensees for operating nuclear power plants, 31 non-power
production and utilization facilities, 15 combined operating license holders/applicants and 29 power
plants that are currently being decommissioned. Licensees may voluntarily submit a request for an
exemption to the Commission and maintain a record of that request. Licensees must perform
certain tasks, maintain records and prepare reports to demonstrate their fulfillment of regulatory
requirements. The reporting and recordkeeping requirements pertain to hydrogen control
analyses; licensee emergency preparedness plans; fire protection plans; records documenting
drills and training records for each fire brigade member; list of electric equipment important to

safety; documents related to the establishment, implementation and maintenance of the training
and qualification of nuclear power plant personnel; pre-operational and periodic tests must be
documented in a readily available summary report and made available for inspection upon request
at the nuclear power plant.
This section incorporates the Mitigation of Beyond Design Basis Events (MBDBE) Final Rule,
which was approved by OMB August 2019.
A.

JUSTIFICATION
1.

Need for the Collection of Information
The information is needed in order to determine licensee compliance with the
regulations set forth in CFR 50.44c; 50.46; 50.47; 50.54(q & t). 50.48; 50.49; 50.120
and Appendix S, 50.54(ff). Details of these regulations can be found at the end of this
supporting statement in Appendix A, “Description of Requirements.”

2.

Agency Use of Information
Applicants or licensees requesting approval to construct or operate utilization or
production facilities are required by the Atomic Energy Act of 1954, as amended (the
Act), to provide information and data that the NRC may determine necessary to
ensure the health and safety of the public.
The NRC uses the records and reports required in this part to ascertain that licensees’
licensing the design, construction, operation, and decommissioning of commercial
nuclear power plants and other nuclear facilities programs are adequate to protect
public health and minimize danger to life and property and that licensees’ personnel
are aware of and follow up on the information and steps needed to perform licensed
activities in a safe manner. The reports and recordkeeping requirements allow NRC to
determine whether to take actions, such as to conduct inspections or to alert other
licensees to prevent similar events that may have generic implications. The
emergency plans provide reasonable assurance that adequate protective measures
can and will be taken in the event of a radiological emergency. The emergency
preparedness information submitted by licensees enables the NRC to determine the
adequacy of the emergency plan, in regard to compliance with the emergency
preparedness regulations. This includes whether additional regulatory oversight is
needed. The information is also used to update information in the NRC Emergency
Operations Center used in support of an NRC’s response to an actual emergency,
drill, or exercise.

3.

Reduction of Burden Through Information Technology
There are no legal obstacles to reducing the burden associated with this information
collection. The NRC encourages respondents to use information technology when it
would be beneficial to them. The NRC has issued Guidance for Electronic
Submissions to the NRC which provides direction for the electronic transmission and
submittal of documents to the NRC. Electronic transmission and submittal of

documents can be accomplished via the following avenues: The Electronic Information
Exchange (EIE) process, which is available from the NRC's “Electronic Submittals”
Web page, by Optical Storage Media (OSM) (e.g. CD-ROM, DVD), by facsimile or by
e-mail. It is estimated that approximately 90% of the responses are filed electronically.
4.

Effort to Identify Duplication and Use Similar Information
No sources of similar information are available. There is no duplication of
requirements.

5.

Effort to Reduce Small Business Burden
The regulations cited previously affect both commercial power reactor licensees and
non-power reactor licensees (e.g., research and test reactors operated by colleges
and universities). Appendix E to 10 CFR Part 50 states that Regulatory Guide 2.61 will
be used as guidance for the acceptability of research and test reactor emergency
response plans. Regulatory Guide 2.6 endorses ANSI/ANS-15.16-1982.2 The
American Nuclear Society revised ANSI/ANS-15.16-1982 on September 13, 2008, and
the NRC is pursuing endorsement of ANSI/ANS-15.16-2008 with a revision to
Regulatory Guide 2.6. In addition, NUREG-08493 addresses emergency plans for
research and test reactors. Together, these documents present the non-power reactor
emergency planning and preparedness requirements.
The emergency preparedness record keeping and reporting burden for non-power
reactors is less than for power reactors, because the requirements are based on the
potential risks associated with the specific reactor, and the corresponding need to
protect the health and safety of the public and the environment. Non-power reactors
are much smaller than power reactors, and as such, create a lesser risk from credible
accidents.

6.

Consequences to Federal Program or Policy Activities if the Collection is Not
Conducted or is Conducted Less Frequently
If the information is not collected, NRC will not be able to assess whether licensees
are operating within the specific safety requirements applicable to the licensing and
operating activities for existing nuclear power reactors and research and test reactors.
The information and required frequency from licensees that seek to licensee and
operator nuclear power reactors and research and test reactors is essential to NRC’s

1

Regulatory Guide 2.6, Emergency Planning for Research and Test Reactors, Rev. 1, March
1983.
2

ANSI/ANS-15.16-1982, American National Standard for Emergency Planning for Research
Reactors, October 11, 1982.
3

NUREG-0849, Standard Review Plan for the Review and Evaluation of Emergency Plans for
Research and Test Reactors, October 1983.

determination of whether the applicant has adequate equipment, training, funds and
experience throughout the life of the licensee to protect the public health and safety.
If the information were not collected, or collected less frequently, the NRC could be
unaware for an extended period of time that an existing or revised emergency plan is
no longer adequate to protect the health and safety of the public and the environment.
Without a timely review of information, changes to personnel, procedures, equipment,
or facilities, or failure to maintain an effective emergency plan, could adversely affect
emergency preparedness and response without the NRC imposing required corrective
measures.
7.

Circumstances which Justify Variation From OMB Guidelines
A licensee must submit a report under 10 CFR 50.46(a)(3)(ii) within 30 days of
discovering any significant change or error so that NRC is apprised of significant
safety issues requiring immediate resolution.
Section 50.4(b)(5) requires that written communications for emergency plans and
related submissions, the signed original must be sent to the NRC Document Control
Desk, with one copy to the appropriate Regional Office, and one copy to the
appropriate NRC Resident Inspector (if one has been assigned to the site of the
facility). This is required because the NRC has both a headquarters and regional
offices, and an NRC Resident Inspector located at the site.
Section 50.54(q)(6) requires that licensees retain their emergency plan and each
change that reduces the effectiveness of the plan as a record until the Commission
terminates the reactor license, which is initially issued for 40 years. Section 50.54(t)
requires that the results and recommendations from emergency plan and emergency
preparedness program reviews be retained for five years. This ensures that the plans
will be maintained and will provide appropriate documentation that will support NRC
review.
Licensees must retain the fire protection plan until the NRC terminates the license in
order to ensure the health and safety of the public.
The records required by 10 CFR 50.49(d) and 10 CFR 50.49(j) are required to be
maintained for the life of the component so that the NRC and the licensees can
periodically assess and determine if equipment important to safety at nuclear power
plants meets specified performance requirements.
Rather than requiring records to be routinely submitted to the NRC, 10 CFR 50.120
requires sufficient records to be maintained on-site to permit NRC verification of the
adequacy of the programs. Pursuant to 10 CFR 50.71, program records are to be
retained until termination of the license. Job performance qualifications are to be
retained and maintained for each employee for the duration of employment. These
record retention requirements result in an auditable trail for ensuring that training is
developed, evaluated, and revised based on job performance requirements, and that
power reactor personnel are qualified to perform their jobs.

Leakage test results, implementation plans, and records of the performance-based
testing program must be kept for the operating lifetime of each nuclear plant for
reference purposes.
8.

Consultations Outside the NRC
Opportunity for public comment on the information collection requirements for this
clearance package was published in the Federal Register on February 19, 2021,
(86 FR 10360). Additionally, NRC staff contacted five stakeholders via email. The
stakeholders were new, operating and research and test reactor owner licensee
representatives and interested stakeholders from Duke Energy Progress, LLC, Kairos
Power, Southern Nuclear Operating Co., Washington State University and X-Energy.
The NRC received one out-of-scope comment as a result of the FRN. No additional
responses or comments were received as a result of the FRN or the staff’s direct
solicitation of comment.

9.

Payment or Gift to Respondents
Not applicable.

10.

Confidentiality of Information
Confidential and proprietary information is protected in accordance with NRC
regulations at 10 CFR 9.17(a) and 10 CFR 2.390(b).

11.

Justification for Sensitive Questions
This regulation does not request sensitive information.

12.

Estimated Industry Burden and Burden Hour Cost
Detailed burden estimates are included in the supplemental burden spreadsheet titled,
“Burden worksheet for Section 7, Standards for Licenses, Certifications and
Regulatory Approvals.”
Hours
Reporting
Recordkeeping
TOTAL

Responses
423,951
4,484
413,701
164
837,652
4,648

The total estimated cost for information collection requirements in this section is
estimated to be 837,652 hours at a cost of $233,704,908 (837,652 hours x $279/hr).
Detailed burden estimates are included in the supplemental burden spreadsheet titled,
“Table 1 - Summary of Supporting Statements.” The $279 hourly rate used in the
burden estimates is based on the Nuclear Regulatory Commission’s fee for hourly rates
as noted in 10 CFR 170.20 “Average cost per professional staff-hour.” For more

information on the basis of this rate, see the Revision of Fee Schedules; Fee Recovery
for Fiscal Year 2020 (85 FR 37250, June 19, 2020).
13.

Estimate of Other Additional Costs
The quantity of records to be maintained is roughly proportional to the recordkeeping
burden and therefore can be used to calculate approximate records storage costs.
Based on the number of pages maintained for a typical clearance, the records storage
cost has been determined to be equal to .0004 times the recordkeeping burden cost.
Therefore, the storage cost for this clearance is estimated to be $46,169 (413,701
recordkeeping hours x $279 x .0004).

14.

Estimated Annualized Cost to the Federal Government
The staff has developed estimates of annualized costs to the Federal Government
related to the conduct of this collection of information. These estimates are based on
staff experience and subject matter expertise and include the burden needed to
review, analyze, and process the collected information and any relevant operational
expenses.
The annualized estimated cost to the government is shown on the attached Summary
Table. The annualized cost to the government is estimated to be $5,656,725 (20,275
staff hours x $279/hr) as shown on the attached Summary Table.

15.

Reasons for Changes in Burden or Cost
The burden and number of responses have changed as described in the tables below:
Burden change
2018 estimates
Reporting
Recordkeeping
Third Party
Disclosure
Total

Current
Change
submission
299,065.0
423,951
+124,886
409,758.2
413,701
+3,942.8
0
0
0.00
708,823.2

837,652

+128,828.8

Change in Responses
2018 estimates
Reporting
Recordkeeping
Third Party
Disclosure
Total

Current
submission

Change

4593.70
173.00

4,484
164

4766.70

0.00
4,648

-109.7
-9
0.00
-118.7

The primary reason for the change in the burden estimates associated with these
requirements is due to the one expected submission of an DC application as stated in
the Overview, Section 1. This submission is subject to the requirements related to

Appendix S, and 50.54(ff) earthquake engineering criteria (DC applicant), this burden
is estimated at 155,000 hours per applicant. Other minimal increase changes
occurred in Appendix K requirements for Emergency Core Cooling Systems (ECCS),
to include all applicable licensees.
There was a minimal decrease related to Appendix S, and 50.54(ff) earthquake
engineering criteria (Operating Reactors), through historical review, staff determined
the expected responses could be reduced, resulting in a reduction of responses and
4,500 hours of burden.
Recordkeeping decreases occurred primarily due to the number of licensees subject
to these regulations, these reductions occurred in the areas of 50.44(c), 50.48,
Appendix R, 50.49(d),(f),(j), Appendix S, and 50.54(ff) earthquake engineering criteria
(Operating Reactors), 50.120(b) due to a reduction of sites and 50.48(c) where no
additional licensees are expected to transition to this regulation. These minor
reductions resulted in an overall reduction of 9 recordkeepers.
Recordkeeping decreased for one-time implementation requirements associated with
50.155(a)(1). This section imposed one-time recordkeeping burden to review
procedures, programs, and plans to confirm that they are consistent with the rule
requirements in the Mitigation of Beyond-Design-Basis Events final rule. Only 17
licensees are estimated to complete these requirements during the clearance period
because 38 licensees are estimated to have completed these requirements. This
resulted in a burden reduction of 11,970 hours
16.

Publication for Statistical Use
The information being collected is not expected to be published for statistical use.

17.

Reason for Not Displaying the Expiration Date
The recordkeeping and reporting requirements for this information collection are
associated with regulations and are not submitted on instruments such as forms or
surveys. For this reason, there are no data instruments on which to display an OMB
expiration date. Further, amending the regulatory text of the CFR to display
information that, in an annual publication, could become obsolete would be unduly
burdensome and too difficult to keep current.

18.

Exceptions to the Certification Statement
None.

B.

COLLECTIONS OF INFORMATION EMPLOYING STATISTICAL METHODS
Not applicable.

Appendix A – Description Requirements
Mitigation of Beyond Design Basis Events (MBDBE) Final Rule, approved August 2019
Section 50.34(i) requires applicants for power reactor operating licenses to include plans
for implementing the requirements in 10 CFR 50.155, “Mitigation of Beyond-Design-Basis
Events,” including a schedule for achieving full compliance, a description of the integrated
response capability, and the equipment and location of the equipment upon which the
strategies rely. These requirements were established under Order EA-12-049 and
licensees have already completed this requirement.
Section 50.155 requires licensees to review their previous compliance under Orders EA-12049 and EA-12-051 against the MBDBE rule requirements to confirm their compliance with
the MBDBE rule, as well as to make changes to procedures, programs, and plans to
reference the new MBDBE rule requirements (rather than the Order requirements). Sites
will incur a one-time recordkeeping burden to review procedures, programs, and plans to
confirm that they are consistent with the rule requirements.
Section 50.155(b)(1) requires licensees to develop strategies and guidelines to mitigate
beyond-design-basis external events from natural phenomena that are developed
assuming a loss of all ac power concurrent with either a loss of normal access to the
ultimate heat sink or loss of normal access to the normal heat sink. These requirements
were established under Order EA-12-049 and licensees have already completed this
requirement.
Section 50.155(b)(2) requires licensees to develop strategies and guidelines to maintain or
restore core cooling, containment, and spent fuel pool cooling capabilities under the
circumstances associated with loss of large areas of the plant impacted by the event, due
to explosions or fire. These requirements were established under Order EA-02-026 and
licensees already comply with this requirement.
Section 50.155(f) requires licensees to maintain documentation of changes in the
implementation of the requirements of section 50.155. These requirements were
established under Order EA-12-049.
Standards for Licenses, Certifications and Regulatory Approvals
SECTION VI- (50.44(c), Hydrogen Control Requirements:
10 CFR 50.44(b)(1), (2), (3), and (4)- contain requirements for a mixed atmosphere,
combustible gas control, equipment survivability, and monitoring of hydrogen and oxygen
concentrations during an accident, for currently-licensed reactors. Further, 10 CFR
50.44(b)(5) requires each current holder of an operating license for a boiling water reactor
(BWR) with a Mark III-type of containment or for a pressurized water reactor (PWR) with an
ice condenser-type of containment to perform certain detailed analyses regarding hydrogen
control, structural capability, and equipment survivability. However, as noted above, all of
the requirements have already been met for currently-licensed reactors.

10 CFR 50.44(c)- requires future water-cooled reactor applicants and licensees to:
(1) Mixed Atmosphere: Have a mixed atmosphere during accidents;
(2) Combustible Gas Control: Either have an inerted atmosphere or limit hydrogen
concentrations in containment during and following an accident that releases an equivalent
amount of hydrogen as would be generated from a 100 percent fuel clad-coolant reaction,
uniformly distributed, to less than 10 percent (by volume) and maintain containment
structural integrity and appropriate accident mitigating features.
(3) Equipment Survivability: Containments that do not rely upon an inerted atmosphere to
control combustible gases must be able to establish and maintain safe shutdown and
containment structural integrity with systems and components capable of performing their
functions during and after exposure to the environmental conditions created by the burning
of hydrogen. Environmental conditions caused by local detonations of hydrogen must also
be included, unless such detonations can be shown unlikely to occur. The amount of
hydrogen to be considered must be equivalent to that generated from a fuel clad-coolant
reaction involving 100 percent of the fuel cladding surrounding the active fuel region.
(4) Monitoring: Equipment must be provided for monitoring oxygen in containments that
use an inerted atmosphere for combustible gas control, and for monitoring hydrogen in all
containments. Equipment for monitoring oxygen and hydrogen must be functional, reliable,
and capable of continuously measuring the concentration of the monitored gas in the
containment atmosphere following a significant beyond-design-basis accident for
combustible gas control and accident management, including emergency planning.
(5) Structural Analysis: An applicant must perform an analysis that demonstrates
containment structural integrity. This demonstration must use an analytical technique that
is accepted by the NRC and include sufficient supporting justification to show that the
technique describes the containment response to the structural loads involved. The
analysis must address an accident that releases hydrogen generated from 100 percent fuel
clad-coolant reaction accompanied by hydrogen burning. Systems necessary to ensure
containment integrity must also be demonstrated to perform their function under these
conditions.
10 CFR 50.44(d)- requires future non-water-cooled reactor applicants and licensees and
certain future water-cooled reactor applicants and licensees to provide:
(1) Information addressing whether accidents involving combustible gases are technically
relevant for their design; and,
(2) If accidents involving combustible gases are found to be technically relevant,
information (including a design-specific probabilistic risk assessment) demonstrating that
the safety impacts of combustible gases during design-basis and significant beyonddesign-basis accidents have been addressed to ensure adequate protection of public
health and safety and common defense and security.

SECTION VII- Appendix K, 50.46, ECCS
10 CFR 50.46- provides an alternate method of meeting the 10 CFR 50 Appendix K
requirements for Emergency Core Cooling Systems (ECCS). It permits licensees or
applicants to analyze ECCS performance using realistic calculations. This method of
calculation may remove some operating restrictions and, thus, motivate licensees to submit
realistic analyses for review. This aspect of the rule represents a voluntary information
collection burden to the industry. Realistic analyses are not required of licensees not
electing this option.
10 CFR 50.46(a)(3)(i)- requires that each applicant for, or holder of, an operating license or
construction permit, other than a holder of a license for a reactor facility for which the
certifications required under 10 CFR 50.82(a)(1) have been submitted, shall estimate the
effect of any change to, or error in, an acceptable evaluation model, or in the application of
such a model, to determine if the change or error is significant. For this purpose, a
significant change or error is one which results in a calculated peak fuel cladding
temperature differing by more than 500F from the temperature calculated for the limiting
transient using the last acceptable model, or is a cumulation of changes and errors, such
that the sum of the absolute magnitudes of the respective temperature changes is greater
than 500F.
10 CFR 50.46(a)(3)(ii)- requires that, for each change to, or error discovered in, an
acceptable evaluation model or in the application of such a model that affects the
temperature calculation, the applicant or licensee shall report the nature of the change or
error, and its estimated effect on the limiting ECCS analysis, to the Commission at least
annually. If the change or error is significant, the applicant or licensee shall provide this
report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46
requirements. This schedule may be developed using an integrated scheduling system
previously approved for the facility by the NRC. For those facilities not using an NRCapproved integrated scheduling system, a schedule will be established by the NRC staff
within 60 days of receipt of the proposed schedule. Any change or error correction that
results in a calculated ECCS performance that does not conform to the criteria set forth in
10 CFR 50.46(b) is a reportable event as described in 10 CFR 50.55(e), 10 CFR 50.72 and
10 CFR 50.73. The affected applicant or licensee shall propose immediate steps to
demonstrate compliance or bring plant design or operation into compliance with 10 CFR
50.46 requirements.
The effort associated with the reports required by 10 CFR 50.46 will vary, depending upon
the nature of the ECCS model change or error being addressed. Most of the annual
reports disclose that no changes were made to the ECCS evaluation or convey information
about minor changes. These reports will require little effort to prepare. Other annual
reports may be based on extensive re-analysis of ECCS performance, resulting in a greater
expenditure of effort. To arrive at its estimate of the burden associated with the annual
reports, the staff used its understanding of the types of reports typically submitted and its
experience in the level of effort required to conduct ECCS evaluations.
10 CFR 50, Appendix K.I.A.- offers licensees the option to use a reduced power level
margin for ECCS evaluation or maintain the current margin of 2% power. To use this
option and apply a lower assumed power level, licensees would be required to demonstrate

the uncertainties associated with measuring reactor thermal power. The resulting change
to ECCS evaluation results must be reported per 10 CFR 50.46(a)(3) and filed as a license
amendment.
10 CFR 50, Appendix K.II.1.a.- requires that a description of each evaluation model be
furnished. The description shall be sufficiently complete to permit technical review of the
analytical approach including the equations used, their approximations in difference form,
the assumptions made, and the values of all parameters or the procedure for their
selection, as for example, in accordance with a specified physical law or empirical
correlation.
10 CFR 50, Appendix K.II.1.- requires that a complete listing of each computer program be
furnished to the NRC upon request in the same form as used in the evaluation model (EM).
NRC does not anticipate the need to request such information during this clearance period.
SECTION VIII- 50.47, 50.54(q & t), Appendix E, Emergency Planning
Section 50.47 contains emergency planning standards that must be met in onsite and
offsite emergency plans for a nuclear power reactor. Appendix E to 10 CFR Part 50
specifies the content of emergency plans for production and utilization facilities and
establishes the minimum requirements for emergency plans to provide reasonable
assurance that public health and safety is not endangered by operation of the facility
concerned.
Section 50.54 establishes license conditions for licenses issued by the NRC.
Section 50.54(q) requires nuclear power, research reactor and/or fuel facility licensees to
follow and maintain in effect emergency plans which meet the applicable standards in
10 CFR 50.47 and requirements in Appendix E to 10 CFR Part 50. Section 50.54(q)
authorizes licensees to make changes to their emergency plans without NRC approval
provided the licensee performs and retains an analysis demonstrating that the change(s)
does/do not reduce the effectiveness of the plan and establishes the record keeping and
reporting requirements for changes made to an emergency plan. Changes made to
emergency plans must be submitted to the NRC within 30 days after the change is put into
effect to allow the NRC to review the changes in a timely manner. Without a timely review,
changes to personnel, procedures, equipment, or facilities that could adversely affect
emergency preparedness, including failure to maintain an effective emergency plan, could
exist without being examined by the NRC. The NRC could be unaware of potential
reductions in the adequacy of emergency preparedness for an extended period of time,
such that the revised plans may no longer provide reasonable assurance that adequate
protective measures can and will be taken in the event of a radiological emergency.
Section 50.54(t) requires licensees to provide for the development, revision,
implementation, and maintenance of its emergency preparedness program, and specifies
that all program elements must be periodically reviewed by persons who have no direct
responsibility for the implementation of the program.
Inspection Reporting Requirements for Emergency Preparedness
Inspections are an important element of NRC’s reactor oversight process (ROP), in that
they ensure that licensees continue to meet applicable regulatory requirements. The NRC

evaluates plant performance by analyzing two distinct inputs: (1) inspection findings
resulting from NRC’s inspection program, and (2) performance indicators (PIs) reported by
the licensee. There are three emergency preparedness PIs: (1) drill and exercise
performance, (2) emergency response organization drill and exercise participation, and (3)
alert and notification system reliability. The data which make up the PIs are generated by
the licensees, and reported to the NRC on a quarterly basis.
10 CFR 50.4(b)(5)- (Emergency plan and related submittals)
10 CFR 50.4(b)(5) provides the specific regulatory requirements for the submittal of written
communications associated with emergency plans submitted under 10 CFR 50.34,
changes to an emergency plan maintained under 10 CFR 50.54(q), and emergency
implementing procedures as described in Section V, “Implementing Procedures,” of
Appendix E to 10 CFR Part 50.
SECTION IX- 50.48, Appendix R, Fire Protection
10 CFR 50.48(a)- requires that each operating nuclear power plant have a fire protection
plan that satisfies Criterion 3 of 10 CFR 50 Appendix A. This fire protection plan must
describe the overall fire protection program for the facility, identify the various positions
within the licensee's organization that are responsible for the program, state the authorities
that are delegated to each of these positions to implement those responsibilities, and
outline the plans for fire protection, fire detection and suppression capability, and limitation
of fire damage. The plan must also describe specific features necessary to implement the
program described above, such as administrative controls and personnel requirements for
fire prevention and manual fire suppression activities, automatic and manually operated fire
detection and suppression systems, and the means to limit damage to structures, systems,
and components important to safety so that the capability to safely shut down the plant is
ensured. Licensees shall retain the fire protection plan and each change to the plan as a
record until the Commission terminates the reactor license and shall retain each
superseded revision of the procedures for three years from the date it was superseded.
10 CFR 50.48(b)- requires that plants licensed to operate before January 1, 1979, meet
sections III.G, III.J, and III.O of 10 CFR Part 50, Appendix R; and fire protection features
accepted by the staff in fire protection safety evaluation reports.
10 CFR 50.48(c)- was implemented in 2004 to provide licensees with the option to
transition their fire protection programs to ones based on National Fire Protection
Association Standard NFPA 805, “Performance-Based Standard for Fire Protection for
Light Water Electric Generating Plants, 2001 Edition” [69 FRN 33536, June 16, 2004]. The
NRC has received 32 Letters of Intent covering 47 units. Forty-six units have already
completed the transition. The remaining unit will not be transitioning.
10 CFR 50.48(f)- requires licensees that have submitted 10 CFR 50.82(a)(1) certifications
to maintain a fire protection program to address the potential for fires which could cause the
release or spread of radioactive materials. Several units have recently shutdown or
announced their intention to shutdown in the near future and then move to the
decommissioning process.

10 CFR 50.48(f)(2)- requires that the fire protection program be assessed by the licensee
on a regular basis and revised, as appropriate, during decommissioning.
10 CFR 50.48(f)(3)- permits the licensee to make changes to the fire protection program
without prior NRC approval if the changes do not reduce the effectiveness of fire protection
for facilities, systems, and equipment which could result in a radiological hazard.
10 CFR 50 Appendix R,- "Fire Protection Program for Nuclear Power Facilities Operating
Prior to January 1, 1979," includes sections III.G, III.J, and III.O that are required to be met
by plants licensed to operate before January 1, 1979. Section III.G requires the capability to
safely shut down. III.J requires emergency lighting. III.O requires the reactor coolant pump
oil collection systems.
10 CFR 50 Appendix R,- "Fire Protection Program for Nuclear Power Facilities Operating
Prior to January 1, 1979," also require each nuclear power plant to have a fire brigade
(III.H), documented drills (Section III.I.3.d) and documented training records for each fire
brigade member (Section III.I.4). Portions of 10 CFR 50 Appendix R were backfit to nuclear
power plant licensees, however Sections III.H and III.I were not. All nuclear power plants
have fire brigades, drills and training, but those features are not based on the rule
requirements captioned above. Therefore, there is no burden associated with this section of
10 CFR 50, Appendix R.
Forty-six nuclear units have already transitioned to performance-based fire protection
programs under 10 CFR 50.48(c). These comply with requirements analogous to those
under 10 CFR 50 Appendix R, Section III.G, as part of their new fire protection programs or
provide justification using performance-based methods for other means of complying with
GDC 3 of Appendix A to 10 CFR 50 (Responses on the part of the NFPA 805 licensees can
be assumed to be incorporated into their reporting requirements under Section 50.48(c).)
SECTION X- 50.49, Environmental Qualification:
Recordkeeping Requirements
10 CFR 50.49(a)- requires applicants and licensees of nuclear power plants, other than a
nuclear power plant for which 10 CFR 50.82(a)(1) certifications have been submitted, to
establish a program for qualifying the electric equipment important to safety as defined in
10 CFR 50.49. The current licensees have completed this requirement. Additional
information is expected to be collected from approximately 4 new combined operating
license (COL) holders.
10 CFR 50.49(d)- requires applicants and licensees to prepare a list of electric equipment
important to safety, and include the performance specifications under conditions existing
during and following design basis accidents, the electric characteristics for which
performance under specified conditions can be ensured, and the environmental conditions
in which it must operate. Applicants and licensees must keep the list and information in the
file current. All current licensees have prepared lists of equipment and performance
specifications, and future information collection under this section of the regulation is
required to the degree it is necessary for keeping the information current. New COL
applicants would need to prepare and maintain this list of electrical equipment important to
safety that is covered under this section.

10 CFR 50.49(f)- requires each item of electric equipment important to safety to be
qualified by one of four specified methods, all with a supporting analysis to show that the
equipment to be qualified is acceptable. Licensees have completed this requirement for
existing plant equipment. However, this requirement remains active for qualification of new
equipment installations and for replacement equipment that falls under the scope of this
regulation. The COL applicants would need to qualify each item of electric equipment
important to safety under one of four specified methods, and perform a supporting analysis
to show that the equipment to be qualified is acceptable. All COL applicants are expected
to qualify electrical equipment during the clearance period.
10 CFR 50.49(j)- requires that a record of the qualification, including documentation
required by 10 CFR 50.49(d), be maintained in an auditable form for the entire period
during which the covered item is installed or stored for future use in the nuclear power
plant. This is required to permit verification that each item of electric equipment important
to safety is qualified for its application and meets its specified performance requirements
when it is subjected to the conditions predicted to be present when it must perform its
safety function, up to the end of its qualified life. This requirement would not apply to
COL’s because the plants would be in the initial design phase.
10 CFR 50.49(l)- requires replacement equipment to be qualified in accordance with the
provisions of 10 CFR 50.49 unless there are sound reasons to the contrary. Therefore,
unless there is suitable justification for some alternate course of action, new equipment
installations and replacement equipment that fall under the scope of 10 CFR 50.49 must be
qualified in accordance with 10 CFR 50.49 requirements, including the documentation
requirements of 10 CFR 50.49(d), CFR 50.49(f) and CFR 50.49(j). The licensee must
maintain any justification for an alternative course of action on site, and the justification
must be available for inspection as part of the inspection procedure. This requirement
would not apply to COL’s because the plants would be in the initial design phase.
Reporting Requirements
10 CFR 50.49(h)- requires each licensee to notify the NRC of any significant equipment
qualification problem that may require extension of the completion date, provided pursuant
to 10 CFR 50.49(g), within 60 days of its discovery. Since this requirement has been
completed by all licensees, no further collection of information is required under this section
of the regulation. This requirement would not apply to COL’s because the activity would be
completed as part of the initial design.
10 CFR 50.49(i)- requires applicants for operating licenses granted after February 22,
1983, but prior to November 30, 1985, to perform an analysis to ensure that the plant can
be safely operated pending completion of equipment qualification required by this section.
This requirement is complete and is not applicable to new COLs.
SECTION XXX- 50.120, Training & Qualification of Nuclear Power Point Personnel
Section 306 of the Nuclear Waste Policy Act of 1982, Public Law 97-425, directed the NRC
to “promulgate regulations or other appropriate Commission regulatory guidance for the
training and qualifications of civilian nuclear power plant operators, supervisors, technicians
and other appropriate operating personnel.” In April 1993, the NRC published 10 CFR
50.120.

10 CFR 50.120- requires that each applicant for, and holder of, an operating license for a
nuclear power plant establish, implement, and maintain training programs for 9 categories
of power plant personnel that provide qualified personnel to operate and maintain the
facility in a safe manner in all modes of operation.
10 CFR 50.120(b)(1)- requires that applicants and licensees develop and maintain these
training programs with a “systems approach to training (SAT)” based on job performance
requirements. Section 10 CFR 50.120 builds on existing industry practice related to
training. Training for the personnel covered by 10 CFR 50.120 has already been developed
and implemented by the industry.
10 CFR 50.120(b)(2)- requires power plant applicants and licensees to periodically
evaluate and revise the training programs to reflect industry experience, changes to the
site, procedures, regulations, and quality assurance requirements.
10 CFR 50.120(b)(2)- also requires periodic review of the training programs by licensee
management, and requires licensees and applicants to maintain and keep available for
NRC inspection, materials sufficient to verify the adequacy of the training programs.
Documents related to the establishment, implementation, and maintenance of the training
programs must be kept. Documentation demonstrating the job performance qualifications of
personnel performing in positions covered by 10 CFR 50.120, including contractor
personnel, must be maintained for each employee for the duration of employment.
SECTION XXXI - Appendix J, Containment Leakage
The 10 CFR 50 Appendix J is divided into two options: Option A, Prescriptive
Requirements, and Option B, Performance-Based Requirements. Option B is a
performance-based rule in which the intervals between tests are established, in part, based
on the previous leakage rate performance of the component or system. A licensee may
adopt, on a voluntary basis, either or both of the overall leakage testing requirements (Type
A tests) and the local leakage rate testing requirements (Type B and C tests) of Option B.
In either case, the recordkeeping requirements of Option B must be implemented. The preoperational and periodic Type A, B, and C tests must be documented to show that the
performance criteria for leakage have been met. The comparison to previous results of the
performance of the overall containment system, and of individual components within it,
must be documented to show that the test intervals established for the containment system
and components within it are adequate. These records must be available for inspection at
plant sites, but licensees are not required to submit these results to the Nuclear Regulatory
Commission (NRC).
Neither option of 10 CFR 50 Appendix J contains specific reporting requirements. All
requirements to make reports to the NRC were eliminated from 10 CFR 50 Appendix J (in
what is now known as Option A) in 1995, and Option B, promulgated in 1995, also contains
no reporting requirements, other than referring to the requirements contained in 10 CFR
50.72 and 10 CFR 50.73. For either option, licensees, under 10 CFR 50.72 and 10 CFR
50.73, currently report any instances of leakage exceeding authorized limits in the
Technical Specifications (TS) of the license.
Although there are no specific reporting requirements, each option has recordkeeping
requirements.

OPTION A
10 CFR 50, Appendix J, Section III requires licensees to develop a program consisting of a
schedule for conducting Type A, B and C tests for leak testing the primary reactor
containment and related systems and components penetrating the primary containment
pressure boundary. Since this information is presented in the Final Safety Analysis Report
(FSAR), any burden involved in its preparation is considered under preparation of the
FSAR.
10 CFR 50, Appendix J, Section III.A.6 states that if a licensee's containment does not pass
the Type A test, the test schedule applicable to subsequent Type A tests will be reviewed
and approved by the Commission. No Commission notifications are expected during this
clearance period.
10 CFR 50, Appendix J, Section V.B requires recordkeeping of test results. The preoperational and periodic tests must be documented in a readily available summary report
that will be made available for inspection, upon request, at the nuclear power plant. The
summary report shall include a schematic arrangement of the leakage rate measurement
system, the instrumentation used, the supplemental test method, and the test program
selected as applicable to the pre-operational test, and all the subsequent periodic tests.
The report shall contain an analysis and interpretation of the leakage rate test data for the
Type A test results to the extent necessary to demonstrate the acceptability of the
containment's leakage rate in meeting acceptance criteria.
10 CFR 50, Appendix J. Section V.B. 2
For each periodic test, leakage test results from Type A, B, and C tests shall be included in
the summary report. The summary report shall contain an analysis and interpretation of the
Type A test results and a summary analysis of periodic Type B and Type C tests that were
performed since the last Type A test. Leakage test results from Type A, B, and C tests that
failed to meet the acceptance criteria of Appendix J, Sections III.A.5(b), III.B.3, and III.C.3
shall be included in a separate accompanying summary report that includes an analysis
and interpretation of the test data, the least squares fit analysis of the test data, the
instrumentation error analysis, and the structural conditions of the containment or
components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses of the supplemental verification test employed to demonstrate the
validity of the leakage rate test measurements shall also be included.
OPTION B
10 CFR 50, Appendix J, Section III.A requires that a Type A test be conducted 1) after the
containment system has been completed and is ready for operation and 2) at a periodic
interval based on the historical performance of the overall containment system as a barrier
to fission product releases to reduce the risk from reactor accidents. The test results must
be compared with previous results to examine the performance history of the overall
containment system to limit leakage.
10 CFR 50, Appendix J, Section III.B requires Type B and Type C pneumatic tests to be
conducted (1) prior to initial criticality, and (2) periodically thereafter at intervals based on
the safety significance and historical performance of each boundary and isolation valve to

ensure the integrity of the overall containment system as a barrier to fission product release
to reduce the risk from reactor accidents.
The performance-based testing program must be established which contains a
performance criterion for Type B and C tests, consideration of leakage-rate limits and
factors that affect performance, evaluations of performance, and comparison to previous
test results.
10 CFR 50, Appendix J, Section IV requires that the results of pre-operational and periodic
Type A, B, and C tests must be documented to show that performance criteria for leakage
have been met. The comparison to previous results of the performance of the overall
containment system and of individual components within it must be documented to show
that the test intervals established for the containment system and components within it are
adequate. These records must be available for inspection at plant sites.
10 CFR 50, Appendix J, Section V.A requires that if the requirements for tests in Option B,
Section III.A, or Option B, Section III.B, are implemented, the recordkeeping requirements
in Option B, IV, for these tests must be substituted for the reporting requirements of the
tests contained in Option A.
10 CFR 50, Appendix J, Section V. B. 2 requires that a licensee or applicant for an
operating license may adopt Option B, or parts thereof, by submitting its implementation
plan and request for revision to technical specifications. (Burden for changes to TS is
covered by the Section 2 Supporting Statement.)
10 CFR 50, Appendix J. Section V. B. 3
The regulatory guide or other implementation document used to develop a performancebased leakage program must be included, by general reference, in the plant's TS. The
submittal for TS revisions must contain justification, including supporting analyses, if the
licensee chooses to deviate from methods approved by the Commission and endorsed in a
regulatory guide.
10 CFR 50. Appendix J. Section V. B. 4
The detailed licensee programs for conducting testing under Option B must be available at
the plant site for inspection.
SECTION XXXIII- Appendix S, Earthquake Engineering Criteria
10 CFR 50 Appendix S IV(a)(3) states that if vibratory ground motion exceeds that of the
Operating Basis Earthquake Ground Motion, or if significant plant damage occurs, the
licensee must shut down the nuclear power plant. If systems, structures, or components
necessary for the safe shutdown of the nuclear power plant are not available after the
occurrence of the Operating Basis Earthquake Ground Motion, the licensee must consult
with the Commission and must propose a plan for the timely, safe shutdown of the nuclear
power plant. Both 10 CFR 50 Appendix S IV(a)(3) and 10 CFR 50.54(ff) require that prior
to resuming operations, the licensee must demonstrate to the Commission that no
functional damage has occurred to those features necessary for continued operation
without undue risk to the health and safety of the public and that the licensing basis is
maintained.

GUIDANCE DOCUMENTS FOR INFORMATION COLLECTION REQUIREMENTS
CONTAINED IN
10 CFR PART 50
STANDARDS FOR LICENSES, CERTIFICATIONS AND REGULATORY APPROVALS
SECTION 7
(50.44(c), Hydrogen Control;
Appendix K, 50.46, ECCS;
50.47, 50.54(q & t), Appendix E, Emergency Planning;
50.48, Appendix R, Fire Protection;
50.49, Environmental Qualification;
50.120, Training & Qualification;
Appendix J, Containment Leakage;
Appendix S, Earthquake Engineering Criteria)
3150-0011
Title
Regulatory Guide 2.6, Emergency Planning for
Research and Test Reactors, Rev. 1, March
1983.

Accession number
ML003740234

NUREG-0849, Standard Review Plan for the
Review and Evaluation of Emergency Plans for
Research and Test Reactors, October 1983.

ML062190191

NUREG 0654, Criteria for Preparation and
Evaluation of Radiological Emergency
Response Plans and Preparedness in Support
of Nuclear Power Plants

ML040420012