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§ 52.47 Contents of applications; technical information.The application must contain a level
of design information sufficient to
enable the Commission to judge the
applicant's proposed means of assuring
that construction conforms to the design
and to reach a final conclusion on all
safety questions associated with the
design before the certification is
granted. The information submitted for
a design certification must include
performance requirements and design
information sufficiently detailed to
permit the preparation of acceptance
and inspection requirements by the
NRC, and procurement specifications and construction and installation
specifications by an applicant. The
Commission will require, before design
certification, that information normally
contained in certain procurement
specifications and construction and
installation specifications be completed (a) The application must contain a final safety analysis report (FSAR) that describes the facility, presents the design bases and the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as a whole, and must include the following information: (1) The site parameters postulated for the design, and an analysis and evaluation of the design in terms of those site parameters; (2) A description and analysis of the
structures, systems, and components
(SSCs) of the facility, with emphasis
upon performance requirements, the
bases, with technical justification
therefor, upon which these
requirements have been established, and
the evaluations required to show that
safety functions will be accomplished. It
is expected that the standard plant will
reflect through its design, construction,
and operation an extremely low
probability for accidents that could
result in the release of significant
quantities of radioactive fission
products. The description shall be
sufficient to permit understanding of the
system designs and their relationship to
the safety evaluations. Such items as the
reactor core, reactor coolant system,
instrumentation and control systems,
electrical systems, containment system, (i) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials; (ii) The extent to which generally accepted engineering standards are applied to the design of the reactor; (iii) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; and (iv) The safety features that are to be
engineered into the facility and those
barriers that must be breached as a (A) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem 4 total effective dose equivalent (TEDE); (B) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a radiation dose in excess of 25 rem TEDE; (3) The design of the facility including: (i) The principal design criteria for the facility. Appendix A to 10 CFR part 50, general design criteria (GDC), establishes minimum requirements for the principal design criteria for watercooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units; (ii) The design bases and the relation of the design bases to the principal design criteria; (iii) Information relative to materials of construction, general arrangement, and approximate dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with an adequate margin for safety; (4) An analysis and evaluation of the
design and performance of structures,
systems, and components with the (5) The kinds and quantities of radioactive materials expected to be produced in the operation and the means for controlling and limiting radioactive effluents and radiation exposures within the limits set forth in part 20 of this chapter; (6) The information required by § 20.1406 of this chapter; (7) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter; (8) The information necessary to demonstrate compliance with any technically relevant portions of the Three Mile Island requirements set forth in 10 CFR 50.34(f), except paragraphs (f)(1)(xii), (f)(2)(ix), and (f)(3)(v); (9) For applications for light-watercooled
nuclear power plants, an
evaluation of the standard plant design
against the Standard Review Plan (SRP)
revision in effect 6 months before the
docket date of the application. The
evaluation required by this section shall
include an identification and
description of all differences in design
features, analytical techniques, and
procedural measures proposed for the
design and those corresponding features, techniques, and measures
given in the SRP acceptance criteria.
Where a difference exists, the evaluation
shall discuss how the proposed
alternative provides an acceptable
method of complying with the
Commission's regulations, or portions (10) The information with respect to the design of equipment to maintain control over radioactive materials in gaseous and liquid effluents produced during normal reactor operations described in 10 CFR 50.34a(e); (11) Proposed technical specifications prepared in accordance with the requirements of §§ 50.36 and 50.36a of this chapter; (12) An analysis and description of the equipment and systems for combustible gas control as required by 10 CFR 50.44; (13) The list of electric equipment important to safety that is required by 10 CFR 50.49(d); (14) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in 10 CFR 50.60 and 50.61; (15) Information demonstrating how the applicant will comply with requirements for reduction of risk from anticipated transients without scram events in § 50.62; (16) A coping analysis, and any design features necessary to address station blackout, as required by 10 CFR 50.63; (17) Information demonstrating how
the applicant will comply with
requirements for criticality accidents in (18) A description and analysis of the
fire protection design features for the
standard plant necessary to comply with (19) A description of the quality
assurance program applied to the design
of the structures, systems, and
components of the facility. Appendix B
to 10 CFR part 50, "Quality Assurance
Criteria for Nuclear Power Plants and
Fuel Reprocessing Plants," sets forth the
requirements for quality assurance
programs for nuclear power plants. The (20) The information necessary to demonstrate that the standard plant complies with the earthquake engineering criteria in 10 CFR part 50, appendix S; (21) Proposed technical resolutions of those Unresolved Safety Issues and medium- and high-priority generic safety issues which are identified in the version of NUREG–0933 current on the date up to 6 months before the docket date of the application and which are technically relevant to the design; (22) The information necessary to
demonstrate how operating experience
insights have been incorporated into the (23) For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass; (24) A representative conceptual
design for those portions of the plant for
which the application does not seek (25) The interface requirements to be
met by those portions of the plant for
which the application does not seek (26) Justification that compliance with
the interface requirements of paragraph
(a)(25) of this section is verifiable (27) A description of the designspecific probabilistic risk assessment (PRA) and its results. (b) The application must also contain: (1) The proposed inspections, tests,
analyses, and acceptance criteria that
are necessary and sufficient to provide (2) An environmental report as required by 10 CFR 51.55. (c) This paragraph applies, according to its provisions, to particular applications: (1) An application for certification of
a nuclear power reactor design that is an evolutionary change from light-water (2) An application for certification of
a nuclear power reactor design that
differs significantly from the light-water (3) An application for certification of
a modular nuclear power reactor design
must describe and analyze the possible [68 FR 54142, Sept. 16, 2003; 72 FR 49526, Aug. 28, 2007] 3 The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events. These accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products. 4 A whole body dose of 25 rem has been stated
to correspond numerically to the once in a lifetime
accidental or emergency dose for radiation workers
which, according to NCRP recommendations at the
time could be disregarded in the determination of
their radiation exposure status (see NBS Handbook
69 dated June 5, 1959). However, its use is not
intended to imply that this number constitutes an
acceptable limit for an emergency dose to the public
under accident conditions. This dose value has
been set forth in this section as a reference value,
which can be used in the evaluation of plant design
features with respect to postulated reactor
accidents, to assure that these designs provide |
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