Requirements of Part 50 and Its Subsections

Part 50 — Domestic Licensing of Production and Utilization Facilities

The regulations in this part are promulgated by the Nuclear Regulatory Commission pursuant to the Atomic Energy Act of 1954, and the Energy Reorganization Act of 1974, to provide for the licensing of production (reprocessing facility) and utilization facilities (reactors). This part also gives notice to all persons who knowingly provide to any licensee, applicant, contractor, or subcontractor, components, equipment, materials, or other goods or services, that relate to a licensee’s or applicant’s activities subject to this part, that they may be individually subject to NRC enforcement action for violation. The following are some of the technical subsections and appendices of this part:

Part 50.34 — Contents of Applications; Technical Information

Each application for a construction permit is required to include a preliminary safety analysis report (PSAR). The PSAR is required to include, as a minimum, the following technical information related to the proposed reactor facility:

  • • assessment of the site on which the nuclear reactor facility is to be located, including site evaluation factors identified in Part 100 of Chapter 1 (10 CFR Part 100);
  • • assessment to contain an analysis and evaluation of the major structures, systems, and components of the facility which bear significantly on the acceptability of the site under the site evaluation factors identified in Part 100 of this chapter;
  • • the extent to which generally accepted engineering standards are applied to the design of the reactor, and its safety-related SSCs;
  • • 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;
  • • the evaluation must determine that the dose limits prescribed in 10 CFR Part 100.11(a) are met;
  • • the safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the radiological consequences of accidents. In performing this assessment, an applicant shall assume a fission product release from the core into the containment assuming that the facility is operated at the ultimate power level contemplated.

Based on the comments and questions raised during the review of a PSAR, the applicant is required to develop a final safety analysis report (FSAR) that would include information describing the facility, the design bases and the limits on its operation, and describes a safety analysis of the structures, systems, and components and of the facility as a whole. The FSAR is required to address the following (a partial listing of significant requirements):

  • • all current information, such as the results of environmental and meteorological monitoring programs, which has been developed since issuance of the construction permit, relating to site evaluation factors identified in Part 100 of this chapter;
  • • a description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification upon which such requirements have been established;
  • • For nuclear reactors, such items as the reactor core, reactor coolant system, instrumentation and control systems, electrical systems, containment system, other engineered safety features, auxiliary and emergency systems, power conversion systems, radioactive waste handling systems, and fuel handling systems need to be discussed to the extent they are pertinent;
  • • for facilities other than nuclear reactors, such items as the chemical, physical, metallurgical, or nuclear process to be performed, instrumentation and control systems, ventilation and filter systems, electrical systems, auxiliary and emergency systems, and radioactive waste handling systems need to be discussed, to the extent they are pertinent;
  • • 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;
  • • analysis and evaluation of ECCS cooling performance following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of Section 50.46 for facilities for which a license to operate may be issued.

The above listing describes the essential technical information needed in the PSAR and FSAR. In addition, the regulation requires complete information regarding the operation of the facility; plans for preoperation and inservice testing including the acceptance criteria, description of accidents considered, and the final technical specifications (see Subsection 50.36).

Part 50.36 — Technical Specifications (TS)

Each applicant for a license authorizing operation of a production (reprocessing) or utilization facility (reactor) is required to include in his/her application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, needs to be included in the application, but does not become part of the technical specifications.

To aid the licensees and applicants to develop plant-specific technical specification, NRC has developed standard technical specification (STS) corresponding to each Nuclear Steam Supply System vendor. Thus, NUREG-1430 contains STS for Babcock & Wilcox Plants, NUREG-1431 for Westinghouse Plants, NUREG-1432 for Combustion Engineering Plants, NUREG-1433 for General Electric BWR 4 Loop Plants, and NUREG-1434 for BWR 6 Loop Plants.

Part 50.44 — Combustible Gas Control for Nuclear Power Reactors

This regulation principally applies to pressure suppression containment structures. However, as the containments of BWR MK I and BWR Mk II (see Chapter 1 of this book for description of these containments) are inserted during operation, the BWR Mark III containments, and PWR Ice Condenser containments are affected by this regulation. In case of a loss of coolant accident (LOCA), there is likelihood for generation of combustible gases (e.g., hydrogen), which could result in hydrogen detonation. Means (hydrogen recombiners, igniters) have to be provided to alleviate the possibility of detonation.

Part 50.46 — Acceptance criteria for Emergency Core Cooling Systems (ECCS) for LWRs

The following is a brief description of the intent of this regulation:

Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (ECCS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in this section of the regulation. ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated. In general, the evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident. Comparisons to applicable experimental data must be made and uncertainties in the analysis method and inputs must be identified and assessed so that the uncertainty in the calculated results can be estimated. This uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is compared to the criteria set forth in this section of the regulation, there is a high level of confidence that the criteria would not be exceeded.

Appendix K, Part II Required Documentation, sets forth the documentation requirements for each evaluation model.

Part 50.48 — Fire Protection

This regulation requires the licensees of operating plants to have fire protection plan. This fire protection plan must (i) describe the overall fire protection program for the facility; (ii) identify the various positions within the licensee’s organization that are responsible for the program; (iii) state the authorities that are delegated to each of these positions to implement those responsibilities; and (iv) outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage.

Additionally, the plan must also describe specific features necessary to implement the program such as, (i) administrative controls and personnel requirements for fire prevention and manual fire suppression activities; (ii) automatic and manually operated fire detection and suppression systems; and (iii) the means to limit fire damage to structures, systems, or components important to safety so that the capability to shut down the plant safely is ensured.

Part 50.55a — Codes and Standards

This regulation requires the applicants of construction permit and licensees of operating reactors to properly use the industry Codes and Standards related to design, construction, inspection, quality assurance, preoperational testing, inservice inspections and testing of safety-related structures, systems, and components. The requirements of the codes and standards are incorporated by reference in this regulation. The regulation also includes modifications and limitations of the requirements of the referenced codes and standards.

By general reference to ASME Code, Subsection NE (2007) [4] of Section III of the ASME Boiler & Pressure Vessel Code is used for design, construction, inspection, and preoperational testing of steel containments. However, Subsection NE does not explicitly describe the loads, load combinations, and acceptance criteria for steel containments. Thus, subsection NE is supplemented with the staff position in Regulatory Guide 1.57 (2007) [5]. Also, Subsections IWE (2007) [6] and IWL (2007) [7] for inservice inspection requirements of steel and concrete containments are incorporated by reference in Subsection 50.55a.

Part 50.60 — Acceptance Criteria for Fracture Prevention Measures for LWRs for Normal Operation

This section provides reference to Appendices G and H of this part of the regulation. The Appendices, essentially, require the information related to fracture toughness and material surveillance program.

Part 50.65 — Requirements for Monitoring the Effectiveness of Maintenance at NPPs

The requirements of this section are applicable during all conditions of plant operation, including normal shutdown operations. The requirements of this section are applicable to all safety-related structures, systems, and components (SSCs), as well as to some essential SSCs characterized as none-safety as defined in this regulation. Ashar and Bagchi [8] discuss how the regulation is implemented for structures of nuclear facilities.

Part 50.69 — Risk-Informed Categorization and Treatment of SSCs for Components for Power Reactors

The categorization of NPP structures, systems and components has been based on deterministic classification of NRC’s Regulatory Guide (RG) 1.26 (2007) [9], and RG 1.29 (2007) [10]. In 1990s, NRC initiated a program for using risk based or risk-informed criteria for a number of requirements, including that for categorization of SSC. As of June 2011, the use of these criteria is optional. Under this regulation, SSCs are categorized in four categories:

  • • Risk-Informed Safety Class (RISC)-1 structures, systems, and components (SSCs) means safety-related SSCs that perform safety significant functions.
  • • Risk-Informed Safety Class (RISC)-2 structures, systems and components (SSCs) means nonsafety-related SSCs that perform safety significant functions.
  • • Risk-Informed Safety Class (RISC)-3 structures, systems and components (SSCs) means safety-related SSCs that perform low safety significant functions.
  • • Risk-Informed Safety Class (RISC)-4 structures, systems and components means nonsafety- related SSCs that perform low safety significant functions.

Safety significant function means a function whose loss could result in a significant adverse effect on defense-in-depth, safety margin, or risk.

Part 50.72 — Immediate Notification Requirements for Operating NPPs

By this regulation, all licensees of operating reactors are required to notify the NRC Operations Center via the Emergency Notification System. The regulation requires the licenses to follow the categorization of 1-hour report, 4-hour report, and 8-hour report, as specified in the Regulation. The licensee shall also notify the NRC immediately after notification of the appropriate State or local agencies and not later than 1 hour after the time the licensee declares one of the Emergency Classes.

Eight-hour reports: If an event or a finding is not reported under other paragraphs of this regulation, the licensee is required to notify the NRC as soon as practical and in all cases within 8 hours of the occurrence of any of the following:

Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers (e.g., Containment), being seriously degraded; or the nuclear power plant being in an unanalyzed condition that could significantly degrade plant safety.

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