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D. C. Cook Nuclear Power Plant Bridgman, Michigan UFSAR Revision 28.0
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UFSAR Revision 28.0
§ 50.34 Contents of Applications; Technical Information.
(a) Preliminary safety analysis report. Each application for a construction permit shall include a preliminary safety analysis report. The minimum information5 to be included shall consist of the following:
(1) Stationary power reactor applicants for a construction permit who apply on or after January 10, 1997, shall comply with paragraph (a)(1)(ii) of this section. All other applicants for a construction permit shall comply with paragraph (a)(1)(i) of this section.
(i) A description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. Special attention should be directed to the site evaluation factors identified in part 100 of this chapter. The assessment must 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, assuming that the facility will be operated at the ultimate power level which is contemplated by the applicant. With respect to operation at the projected initial power level, the applicant is required to submit information prescribed in paragraphs (a)(2) through (a)(8) of this section, as well as the information required by this paragraph, in support of the application for a construction permit, or a design approval.
(ii) A description and safety assessment of the site and a safety assessment of the facility. It is expected that reactors will reflect through their design, construction and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The following power reactor design characteristics and proposed operation will be taken into consideration by the Commission:
(A) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials; (B) The extent to which generally accepted engineering standards are applied to the design of the reactor; (C) 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; (D) 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 UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
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 release6 from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site meteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with part 100 of this chapter. The evaluation must determine that:
(1) An individual located at any point on the boundary of the exclusion area for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 rem7 total effective dose equivalent (TEDE).
(2) 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 total effective dose equivalent (TEDE);
(E) With respect to operation at the projected initial power level, the applicant is required to submit information prescribed in paragraphs (a)(2) through (a)(8) of this section, as well as the information required by paragraph (a)(1)(i) of this section, in support of the application for a construction permit.
(2) A summary description and discussion of the facility, with special attention to design and operating characteristics, unusual or novel design features, and principal safety considerations.
(3) The preliminary design of the facility including:
(i) The principal design criteria for the facility8. Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum requirements for the principal design criteria for water-cooled 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 for construction permits in establishing principal design criteria for other types of nuclear power units; UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
(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 final design will conform to the design bases with adequate margin for safety.
(4) A preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high point vents following postulated loss-of-coolant accidents must be performed in accordance with the requirements of § 50.46 and § 50.46a of this part for facilities for which construction permits may be issued after December 28, 1974.
(5) An identification and justification for the selection of those variables, conditions, or other items which are determined as the result of preliminary safety analysis and evaluation to be probable subjects of technical specifications for the facility, with special attention given to those items which may significantly influence the final design:
Provided, however, That this requirement is not applicable to an application for a construction permit filed prior to January 16, 1969.
(6) A preliminary plan for the applicant's organization, training of personnel, and conduct of operations.
(7) A description of the quality assurance program to be applied to the design, fabrication, construction, and testing of the structures, systems, and components of the facility. Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," sets forth the requirements for quality assurance programs for nuclear power plants and fuel reprocessing plants. The description of the quality assurance program for a nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of appendix B will be satisfied.
(8) An identification of those structures, systems, or components of the facility, if any, which require research and development to confirm the adequacy of their design; and identification and description of the research and development program which will be conducted to resolve any safety questions associated with such structures, systems or components; and a schedule of the research and development program showing that such UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
safety questions will be resolved at or before the latest date stated in the application for completion of construction of the facility.
(9) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter.
(10) A discussion of the applicant's preliminary plans for coping with emergencies.
Appendix E sets forth items which shall be included in these plans.
(11) On or after February 5, 1979, applicants who apply for construction permits for nuclear power plants to be built on multiunit sites shall identify potential hazards to the structures, systems and components important to safety of operating nuclear facilities from construction activities. A discussion shall also be included of any managerial and administrative controls that will be used during construction to assure the safety of the operating unit.
(12) On or after January 10, 1997, stationary power reactor applicants who apply for a construction permit, as partial conformance to General Design Criterion 2 of appendix A to this part, shall comply with the earthquake engineering criteria in appendix S to this part.
(13) On or after July 13, 2009, stationary power reactor applicants who apply for a construction permit shall submit the information required by 10 CFR 50.150(b) as a part of their preliminary safety analysis report.
UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
§ 50.34 Contents of Applications; Technical Information.
(b) Final safety analysis report. Each application for an operating license shall include a final safety analysis report. The final safety analysis report shall include information 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 shall include the following:
(1) 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 (Reactor Site Criteria) of this chapter.
(2) A description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.
(i) 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 shall be discussed insofar as they are pertinent.
(ii) 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 shall be discussed insofar as they are pertinent.
(3) 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 (Standards for Protection Against Radiation) of this chapter.
(4) A final analysis and evaluation of the design and performance of structures, systems, and components with the objective stated in paragraph (a)(4) ((a)(4) A preliminary analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high point vents following postulated loss-of-coolant accidents must be performed in accordance with the requirements of § 50.46 (Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors) and § 50.46a (Acceptance criteria for reactor coolant system venting systems) of this part for facilities for which construction permits may be issued after December 28, 1974.) of this section and taking into account any pertinent information developed since the submittal of the preliminary safety analysis report. Analysis and evaluation of ECCS cooling performance following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of § 50.46 (Acceptance Criteria For Emergency Core Cooling Systems For Light-Water Nuclear Power Reactors) for facilities for which a license to operate may be issued after December 28, 1974.
(5) A description and evaluation of the results of the applicant's programs, including research and development, if any, to demonstrate that any safety questions identified at the construction permit stage have been resolved.
(6) The following information concerning facility operation:
(i) The applicant's organizational structure, allocations or responsibilities and authorities, and personnel qualifications requirements.
(ii) Managerial and administrative controls to be used to assure safe operation.
Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, sets forth the requirements for such controls for nuclear power plants and fuel reprocessing plants. The information on the controls to be used for a nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of appendix B will be satisfied.
(iii) Plans for preoperational testing and initial operations.
(iv) Plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of structures, systems, and components.
(v) Plans for coping with emergencies, which shall include the items specified in appendix E (Emergency Planning and Preparedness for Production and Utilization Facilities).
(vi) Proposed technical specifications prepared in accordance with the requirements of § 50.36 (Technical Specifications).
UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
(vii) On or after February 5, 1979, applicants who apply for operating licenses for nuclear power plants to be operated on multiunit sites shall include an evaluation of the potential hazards to the structures, systems, and components important to safety of operating units resulting from construction activities, as well as a description of the managerial and administrative controls to be used to provide assurance that the limiting conditions for operation are not exceeded as a result of construction activities at the multiunit sites.
(7) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter.
(8) A description and plans for implementation of an operator requalification program.
The operator requalification program must as a minimum, meet the requirements for those programs contained in § 55.59 (Requalification) of part 55 of this chapter.
(9) 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 § 50.61 (Fracture toughness requirements for protection against pressurized thermal shock events) (b)(1) ((b) Requirements. (1) For each pressurized water nuclear power reactor for which an operating license has been issued under this part or a combined license issued under Part 52 of this chapter, other than a nuclear power reactor facility for which the certification required under § 50.82(a)(1)
(Termination of License) has been submitted, the licensee shall have projected values of RTPTS or RTMAX-X, accepted by the NRC, for each reactor vessel beltline material.
For pressurized water nuclear power reactors for which a construction permit was issued under this part before February 3, 2010 and whose reactor vessel was designed and fabricated to the 1998 Edition or earlier of the ASME Code, the projected values must be in accordance with this section or § 50.61a (Definitions). For pressurized water nuclear power reactors for which a construction permit is issued under this part after February 3, 2010 and whose reactor vessel is designed and fabricated to an ASME Code after the 1998 Edition, or for which a combined license is issued under Part 52, the projected values must be in accordance with this section. When determining compliance with this section, the assessment of RTPTS must use the calculation procedures described in paragraph (c)(1) and perform the evaluations described in paragraphs (c)(2) and (c)(3) of this section. The assessment must specify the bases for the projected value of RTPTS for each vessel beltline material, including the assumptions regarding core loading patterns, and must specify the copper and nickel contents and the fluence value used in the calculation for each beltline material. This assessment must be updated whenever there is a significant 2 change in projected values of RTPTS, or upon request for a change in the expiration date for operation of the facility.) and (b)(2) ((b)Requirements (2) The UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
pressurized thermal shock (PTS) screening criterion is 270 °F for plates, forgings, and axial weld materials, and 300 °F for circumferential weld materials. For the purpose of comparison with this criterion, the value of RTPTS for the reactor vessel must be evaluated according to the procedures of paragraph (c) of this section, for each weld and plate, or forging, in the reactor vessel beltline. RTPTS must be determined for each vessel beltline material using the EOL fluence for that material.).
(10) On or after January 10, 1997, stationary power reactor applicants who apply for an operating license, as partial conformance to General Design Criterion 2 of appendix A (General Design Criterion 2 of Appendix A to part 50 of this chapter requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions.)
to this part, shall comply with the earthquake engineering criteria of appendix S to this part. However, for those operating license applicants and holders whose construction permit was issued before January 10, 1997, the earthquake engineering criteria in Section VI (Application to Engineering Design) of appendix A to part 100 of this chapter continues to apply.
(11) On or after January 10, 1997, stationary power reactor applicants who apply for an operating license, shall provide a description and safety assessment of the site and of the facility as in § 50.34(a)(1)(ii). However, for either an operating license applicant or holder whose construction permit was issued before January 10, 1997, the reactor site criteria in part 100 of this chapter and the seismic and geologic siting criteria in appendix A to part 100 (Seismic and Geologic Siting Criteria for Nuclear Plants) of this chapter continues to apply.
(12) On or after July 13, 2009, stationary power reactor applicants who apply for an operating license which is subject to 10 CFR 50.150(a) shall submit the information required by 10 CFR 50.150(b) as a part of their final safety analysis report.
UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
§ 50.71 Maintenance of Records, Making of Reports.
(a) Each licensee, including each holder of a construction permit or early site permit, shall maintain all records and make all reports, in connection with the activity, as may be required by the conditions of the license or permit or by the regulations, and orders of the Commission in effectuating the purposes of the Act, including Section 105 of the Act, and the Energy Reorganization Act of 1974, as amended. Reports must be submitted in accordance with § 50.4 or 10 CFR 52.3, as applicable.
(b) With respect to any production or utilization facility of a type described in § 50.21(b) or 50.22, or a testing facility, each licensee and each holder of a construction permit shall submit its annual financial report, including the certified financial statements, to the Commission, as specified in § 50.4, upon issuance of the report. However, licensees and holders of a construction permit who submit a Form 10-Q with the Securities and Exchange Commission or a Form 1 with the Federal Energy Regulatory Commission, need not submit the annual financial report or the certified financial statement under this paragraph.
(c) Records that are required by the regulations in this part or part 52 of this chapter, by license condition, or by technical specifications must be retained for the period specified by the appropriate regulation, license condition, or technical specification. If a retention period is not otherwise specified, these records must be retained until the Commission terminates the facility license or, in the case of an early site permit, until the permit expires.
(d)
(1) Records which must be maintained under this part or part 52 of this chapter may be the original or a reproduced copy or microform if the reproduced copy or microform is duly authenticated by authorized personnel and the microform is capable of producing a clear and legible copy after storage for the period specified by Commission regulations.
The record may also be stored in electronic media with the capability of producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications, must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with, and loss of records.
(2) If there is a conflict between the Commission's regulations in this part, license condition, or technical specification, or other written Commission approval or authorization pertaining to the retention period for the same type of record, the retention period specified in the regulations in this part for such records shall apply unless the Commission, pursuant to § 50.12 of this part, has granted a specific exemption from the record retention requirements specified in the regulations in this part.
UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
(e) Each person licensed to operate a nuclear power reactor under the provisions of § 50.21 or § 50.22, and each applicant for a combined license under part 52 of this chapter, shall update periodically, as provided in paragraphs (e) (3) and (4) of this section, the final safety analysis report (FSAR) originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the applicant or licensee or prepared by the applicant or licensee pursuant to Commission requirement since the submittal of the original FSAR, or as appropriate, the last update to the FSAR under this section. The submittal shall include the effects1 of all changes made in the facility or procedures as described in the FSAR; all safety analyses and evaluations performed by the applicant or licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment in accordance with § 50.59(c)(2) or, in the case of a license that references a certified design, in accordance with § 52.98(c) of this chapter; and all analyses of new safety issues performed by or on behalf of the applicant or licensee at Commission request. The updated information shall be appropriately located within the update to the FSAR.
(1) The licensee shall submit revisions containing updated information to the Commission, as specified in § 50.4, on a replacement-page basis that is accompanied by a list which identifies the current pages of the FSAR following page replacement.
(2) The submittal shall include (i) a certification by a duly authorized officer of the licensee that either the information accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement, or that no such changes were made; and (ii) an identification of changes made under the provisions of § 50.59 but not previously submitted to the Commission.
(3)
(i) A revision of the original FSAR containing those original pages that are still applicable plus new replacement pages shall be filed within 24 months of either July 22, 1980, or the date of issuance of the operating license, whichever is later, and shall bring the FSAR up to date as of a maximum of 6 months prior to the date of filing the revision.
(ii) Not less than 15 days before § 50.71(e) becomes effective, the Director of the Office of Nuclear Reactor Regulation shall notify by letter the licensees of those nuclear power plants initially subject to the NRC's systematic evaluation program that they need not comply with the provisions of this section while the program is being conducted at their plant. The Director of the Office of Nuclear Reactor Regulation will notify by letter the licensee of each nuclear power plant being UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
evaluated when the systematic evaluation program has been completed. Within 24 months after receipt of this notification, the licensee shall file a complete FSAR which is up to date as of a maximum of 6 months prior to the date of filing the revision.
(iii) During the period from the docketing of an application for a combined license under subpart C of part 52 of this chapter until the Commission makes the finding under § 52.103(g) of this chapter, the update to the FSAR must be submitted annually.
(4) Subsequent revisions must be filed annually or 6 months after each refueling outage provided the interval between successive updates does not exceed 24 months. The revisions must reflect all changes up to a maximum of 6 months prior to the date of filling. For nuclear power reactor facilities that have submitted the certifications required by § 50.82(a)(1), subsequent revisions must be filed every 24 months.
(5) Each replacement page shall include both a change indicator for the area changed, e.g., a bold line vertically drawn in the margin adjacent to the portion actually changed, and a page change identification (date of change or change number or both).
(6) The updated FSAR shall be retained by the licensee until the Commission terminates their license.
UFSAR Revision 28.0 Part 50 - Domestic Licensing of Production and Utilization Facilities
NEI 98-03 [Revision 1]
Guidelines for Updating Final Safety Analysis Reports UFSAR Revision 28.0
June 1999 UFSAR Revision 28.0
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UFSAR Revision 28.0
NEI 98-03 (Rev. 1)
June 1999 Nuclear Energy Institute, 1776 I Street N.W., Suite 400, Washington D.C. (202.739.8000)
NEI 98-03 [Revision 1]
Nuclear Energy Institute Guidelines for Updating Final Safety Analysis Reports June 1999 UFSAR Revision 28.0
NEI 98-03 (Rev. 1)
June 1999 Nuclear Energy Institute, 1776 I Street N.W., Suite 400, Washington D.C. (202.739.8000)
ACKNOWLEDGMENTS NEI appreciates the invaluable assistance of the FSAR Task Force and Regulatory Process Working Group toward development of this guideline.
NOTICE Neither NEI, nor any of its employees, members, supporting organizations, contractors, or consultants make any warranty, expressed or implied, or assume any legal responsibility for the accuracy or completeness of, or assume any liability for damages resulting from any use of, any information apparatus, methods, or process disclosed in this report or that such may not infringe privately owned rights.
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TABLE OF CONTENTS 1 INTRODUCTION............................................................................................................................................1 2 BACKGROUND..............................................................................................................................................1 3 DEFINITIONS.................................................................................................................................................1 4 ROLE OF THE UPDATED FSAR.................................................................................................................. 3 5 SCOPE OF THE UPDATED FSAR.............................................................................................................. 3 6 UPDATING FSARs TO MEET 10 CFR 50.71(e).................................................................................... 6 7 FREQUENCY OF PERIODIC UPDATES................................................................................................... 10 8 TREATMENT OF LONG-TERM TEMPORARY MODIFICATIONS......................................................11 9 TREATMENT OF DISCREPANCIES BETWEEN THE FACILITY AND THE UPDATED FSAR........12 APPENDIX A: MODIFYING THE UPDATED FSAR APPENDIX B: 10 CFR 50.71(e)
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1 INTRODUCTION The purpose of this document is to provide licensees with guidance for updating final safety analysis reports (FSARs) consistent with the requirements of 10 CFR 50.71(e), the FSAR update rule. Guidance is also provided in Appendix A for making voluntary modifications to updated FSARs (UFSARs) (i.e., removal, reformatting and simplification of information, as appropriate) to improve their focus, clarity and maintainability. Figure 1 (page 5) depicts the overall process for updating and modifying the UFSAR.
2 BACKGROUND FSARs originally served as the principal reference document in support of Part 50 license applications. The original FSAR described methods for conforming with applicable NRC regulations and contains the technical information required by 10 CFR 50.34(b),
including information that describes the facility, presents the design bases and the limits on its operation, and presents the safety analyses of the structures, systems and components and of the facility as a whole. In 1980, the NRC issued the FSAR update rule, 10 CFR 50.71(e), which requires licensees to update their FSARs periodically to assure that the information provided is the latest material developed.
Inspections in 1996-97 by the NRC and licensees identified numerous discrepancies between UFSAR information and the actual plant design and operation. These findings have raised questions about possible noncompliance with 10 CFR 50.71(e). The industry has developed this guidance in recognition of the importance of the UFSAR, the need to comply with 10 CFR 50.71(e) update requirements, and the need for UFSARs to be consistent with the plant design and operation.
3 DEFINITIONS 3.1 COMMISSION REQUIREMENTS Commission requirements include regulations, license conditions, technical specifications and orders.
3.2 DESIGN BASES Design bases are information that identifies the specific functions to be performed by a structure, system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values may be (1) restraints derived from generally accepted state-of-the-art practices for achieving functional goals or (2) requirements derived from analysis (based on calculations and/or experiments) of the effects of a postulated accident for which a structure, system or component must meet its functional goals. (10 CFR 50.2).
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Further discussion and examples of design bases are provided in NEI 97-04, Design Bases Program Guidelines.
3.3 HISTORICAL INFORMATION Historical information is that which was provided in the original FSAR to meet the requirements of 10 CFR 50.34(b) and meets one or more of the following criteria:
I information that was accurate at the time the plant was originally licensed, but is not intended or expected to be updated for the life of the plant I
information that is not affected by changes to the plant or its operation I
information that does not change with time.
3.4 OBSOLETE INFORMATION Obsolete information is information about (1) equipment that has been removed from the plant, (2) organizations, programs or procedures that are no longer in effect and do not meet the definition of historical information, or (3) design information, evaluations and other UFSAR description that no longer apply to the facility.
3.5 ORIGINAL FSAR The original FSAR is the FSAR submitted with the application for the operating license, as amended and supplemented, and reviewed by the NRC in granting the initial license to operate the facility. Note that for early licensees, the Final Hazards Summary Report performed the role of the FSAR in the licensing process.
3.6 SAFETY ANALYSES Safety analyses are analyses performed pursuant to Commission requirement to demonstrate the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in 10 CFR 50.34(a)(1) or 10 CFR 100.11.
Safety analyses are required to be presented in the UFSAR per 10 CFR 50.34(b) or 10 CFR 50.71(e) and include, but are not limited to, the accident analyses typically presented in Chapter 14 or 15 of the UFSAR.
3.7 UFSAR DESCRIPTION UFSAR description includes text, tables, diagrams, etc., that provide an understanding of the design bases, safety analyses and facility operation under conditions of normal operation, anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant is designed to function.
3.8 UPDATED FSAR The updated FSAR (UFSAR) is the original FSAR as updated per the requirements of UFSAR Revision 28.0
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4 ROLE OF THE UPDATED FSAR UFSARs provide a description of each plant and, per the Supplementary Information for the FSAR update rule, serve as a reference document to be used for recurring safety analyses performed by licensees, the Commission, and other interested parties. The UFSAR is used by the NRC in its regulatory oversight of a nuclear power plant, including its use as a reference for evaluating license amendment requests and in the preparation for and conduct of inspection activities. For licensees, portions of the UFSAR are used as a reference in evaluating changes to the facility and procedures under the 10 CFR 50.59 change process. The UFSAR also serves to provide the general public a description of the plant and its operation.
5 SCOPE OF THE UPDATED FSAR 10 CFR 50.34(b) defined the scope of information required to be submitted in original FSARs and, by extension, the scope of UFSARs as they exist today. While original FSARs expanded greatly over the years as increasingly detailed information was required of new licensees, the scope given by 10 CFR 50.34(b) provided the common baseline for all original FSARs.
In addition to the scope of information contained in the original FSAR, the scope of todays UFSARs includes information added per the FSAR update rule1. The update rule requires licensees to update their UFSARs to reflect new Commission requirements and the effects of changes to the facility and procedures, safety evaluations and analyses of new safety issues requested by the Commission.
Just as the scope of the original FSAR was determined by the requirements of 10 CFR 50.34(b), it follows that the scope of information subsequently added to the original FSAR through the update process should be guided by the requirements used to establish the content of the original FSAR.
10 CFR 50.34(b) contains the following statement of general intent concerning the required content of FSARs submitted as part of original license applications:
The FSAR shall include information that describes the facility, presents the design bases and the limits on its operation, and presents the safety analyses of the structures, systems and components and of the facility as a whole.
Subsections (1) through (9) of 10 CFR 50.34(b) further define or amplify this statement of general intent. Certain information required to be included in original FSARs is now 1 The scope of the UFSAR also may be affected by the other NRC requirements, such as 10 CFR 54.21(d). This rule requires licensees to supplement their UFSARs as part of the technical information submitted with license renewal applications.
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controlled in separate licensee documents in accordance with other NRC regulations. For example, the plant technical specifications establish the limits on facility operation, including safety limits; limiting safety system settings; and limiting conditions for operation for structures, systems and components. The technical specifications were required as part of the original FSAR under 10 CFR 50.34(b)(6)(vi), but are now controlled separately from the UFSAR per 10 CFR 50.36.
Based on analysis of 10 CFR 50.34(b), UFSAR updates should contain the following basic types of information concerning new requirements and information developed since the UFSAR was last updated that are required to be reflected in the UFSAR under 10 CFR 50.71(e):
I new or modified design bases I
summary of new or modified safety analyses I
UFSAR description sufficient to permit understanding of new or modified design bases, safety analyses, and facility operation (as defined in Section 3.7).
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Figure 1 PROCESS FOR UPDATING AND MODIFYING THE UFSAR UFSAR Update Process (Per 10 CFR 50.71(e))
UFSAR Modification Process (See Appendix A)
Current UFSAR Content New Requirements, Plant/Procedure Changes, Safety Evaluations, New Issues Apply UFSAR Content Update Criteria Apply UFSAR Modification Guidance Reformat UFSAR Information Simplify/Remove UFSAR Information Update UFSAR Information Submit UFSAR Update to the NRC Input Evaluation Changes Reporting Identify Information Removed and Basis for Removal UFSAR Revision 28.0
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6 UPDATING FSARS TO MEET 10 CFR 50.71(e) 6.1 WHAT THE REGULATIONS REQUIRE 10 CFR 50.71(e) requires licensees to periodically update their UFSARs to assure they remain up-to-date such that they accurately reflect the plant design and operation. Per 10 CFR 50.71(e)(4), the UFSAR is required to reflect changes up to a maximum of six months prior to the date that the last update was submitted to the NRC. The 10 CFR 50.71(e) requirements concerning the content of updates are as follows2:
Each person licensed to operate a nuclear power reactor pursuant to the provisions of § 50.21 or § 50.22 of this part shall update periodically, as provided in paragraphs (e)(3) and (4) of this section, the final safety analysis report (FSAR) originally submitted as part of the application for the operating license, to assure that the information included in the FSAR contains the latest material developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last updated FSAR. The updated FSAR shall be revised to include the effects of: all changes made in the facility or procedures as described in the FSAR; all safety evaluations performed by the licensee either in support of requested license amendments or in support of conclusions that changes did not involve an unreviewed safety question; and all analyses of new safety issues performed by or on behalf of the licensee at Commission request. The updated information shall be appropriately located within the FSAR.
The rule does not require that licensees review all the information contained in the UFSAR for each periodic update. Rather, the intent of the rule is that licensees update only those portions that have been affected by licensee activities since the previous update. Per the Supplementary Information provided with the 1980 FSAR update rule, Submittal of updated FSAR pages does not constitute a licensing action but is only intended to provide information. It is not intended for the purpose of re-reviewing plants.... The material submitted may be reviewed by the NRC staff but will not be formally approved.
The rule specifies the types of new information that must be evaluated to determine if the UFSAR must be updated to reflect the new information, i.e., new requirements, changes to the facility or procedures, including supporting safety evaluations, and NRC-requested analyses. The following subsections provide guidance for implementing the requirements.
2 In addition to the update requirements of 10 CFR 50.71(e), the rule also includes certain administrative and reporting requirements. The full text of 10 CFR 50.71(e) is provided in Appendix B to this report.
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6.1.1 New Regulatory Requirements UFSARs must be updated to reflect changes to the facility resulting from new or amended requirements, e.g., Appendix R, the Station Blackout rule (10 CFR 50.63), the Anticipated Transient Without Scram (ATWS) rule (10 CFR 50.62), or plant-specific orders. As a result of such new requirements, the following information must be incorporated in the UFSAR, as applicable:
I new or modified design bases I
summary of new or modified safety analyses I
appropriate UFSAR description as defined in Section 3.7 of this guideline.
If a new NRC requirement does not result in these types of information, the UFSAR does not need to be updated to reflect the new requirement.
6.1.2 Changes to the Facility or Procedures The UFSAR must be updated to reflect the following effects, as applicable, of changes implemented under 10 CFR 50.90 or 10 CFR 50.59, including supporting safety evaluations:
I a change requires update of the existing UFSAR information, including changes to existing design bases, safety analyses or description of existing structures, systems, components or functions described in the UFSAR I
a change results in the removal from the plant of SSCs described in the UFSAR or the elimination of functions or procedures described in the UFSAR I
a change or supporting safety evaluation results in new design bases or safety analyses, or associated description, that must be included in the UFSAR.
If a change or supporting safety evaluation does not affect existing UFSAR information and does not result in new design bases, safety analyses or UFSAR description, the UFSAR does not need to be updated to reflect the change.
6.1.3 Analyses of New Safety Issues Licensees should evaluate the effects of analyses or similar evaluations performed by licensees in response to plant-specific NRC requests or NRC generic letters or bulletins.
NRC-requested analyses and evaluations must be reflected in UFSAR updates only if, on the basis of the results of the requested analysis or evaluation, the licensee determines that the existing design bases, safety analyses or UFSAR description are either not accurate or not bounding or both. The existing design bases, safety analyses and UFSAR description must be updated to reflect the new information, as appropriate.
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If the NRC-requested analyses or evaluations do not cause any of these effects, no change to the UFSAR is required.
6.1.4 Update Process Considerations Licensees should establish a process to identify the types of new information that must be evaluated to determine if the UFSAR must be updated to reflect the new information. To be consistent with the requirements of the FSAR update rule, the process should include sufficient administrative controls to identify information and analyses submitted pursuant to Commission requirements; changes to the facility or procedures; safety evaluations; and analyses of new safety issues performed at Commission request.
In general, controls sufficient to identify information pursuant to Commission requirement should focus on changes to NRC regulations, license conditions, orders and technical specifications. The controls for identifying changes to the facility or procedures and safety evaluations should be integrated with existing licensee administrative controls for implementing design and procedure changes, including the process used by licensees in preparing, reviewing and approving 10 CFR 50.59 evaluations. The controls for identifying analyses of new safety issues performed at Commission request should focus on NRC bulletins, generic letters and analogous plant-specific communications, including NRC requests pursuant to 10 CFR 50.54(f).
6.2 LEVEL OF DETAIL FOR FSAR UPDATES While not explicitly addressing the level of detail required for FSARs, 10 CFR 50.34(b)(2) required that original FSARs include:
description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.
In addition, the Supplementary Information provided with the 1980 FSAR update rule stated: The level of detail to be maintained in the UFSAR should be at least the same as originally provided. Thus, existing UFSAR information of a similar nature may provide a guide for determining the level of detail for new information to be included in UFSAR updates. However, the primary consideration in determining the level of detail for new information is whether updated information is sufficient to permit understanding of new or modified safety analyses, design bases and facility operation.
6.3 EXAMPLES The following examples illustrate the application of the UFSAR update guidance.
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CASE 1: The licensee action is not in response to a new Commission requirement, does not involve a change, safety evaluation or analysis of a new issue, and does not affect existing UFSAR information. Therefore, no update to the UFSAR is required by 10 CFR 50.71(e).
Example Generic Letter 96-01, Testing of Safety-Related Circuits, requested licensees to conduct a review of Logic System Testing to ensure that all elements of the logic circuits were being adequately tested and met the technical specification surveillance requirement for adequate logic system functional testing.
This generic correspondence did not constitute a new regulatory requirement and did not request a new analysis. Provided the licensee response to Generic Letter 96-01 did not result in a change to the facility or actions that affected existing UFSAR information, no change to the UFSAR is required.
CASE 2: The licensee action responds to a new Commission requirement or involves a change, safety evaluation or analysis of a new issue, and update of the UFSAR is required to change existing information.
Example A change to the safety injection system was initiated to address an operability concern identified in NRC Bulletin 88-04, Potential for Safety-Related Pump Loss. An evaluation of safety injection pump minimum-flow lines resulted in an increase in the recommended minimum-flow rate to preclude hydraulic instability at low flow conditions and assure pump operability. As a result of this evaluation, the orifices in the safety injection recirculation lines were modified to provide for increased minimum-flow rate for the pumps.
Unlike the generic letter in the example of Case 1, NRC Bulletin 88-04 requested that licensees evaluate safety-related pump performance under minimum flow conditions, and the licensee evaluation resulted in a change to the safety injection recirculation lines. Because sufficient minimum-flow is necessary to ensure the system is able to perform its intended safety function, the UFSAR description associated with the safety injection system should be modified to include a discussion of the minimum-flow function as it relates to maintaining operability of the safety injection pumps. In some cases, this may entail adding UFSAR discussion of the minimum-flow function where none previously existed.
If licensee evaluations requested by Bulletin 88-04 determined the existing minimum-flow design to be acceptable, no change to the UFSAR is required.
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Example 10 CFR 50.62 (the ATWS rule) required the installation of a new mitigation system specific to the type of plant (Westinghouse, Combustion Engineering, etc.). In response to the ATWS rule, the licensee installed new equipment in the facility. An evaluation was performed in accordance with the guidance in Section 6 to determine if update of the UFSAR was required. Because ATWS constitutes new Commission requirements for the plant, the design bases and associated description of the new ATWS equipment should be added to the UFSAR.
CASE 4: The licensee action responds to a new Commission requirement or involves a change, safety evaluation or analysis of a new issue, and update of the UFSAR is not required.
Example The NRC issued a new requirement, 10 CFR Part 26, requiring licensees to implement a Fitness for Duty Program (FFD). An evaluation was performed in accordance with the guidance in Section 6 to determine if update of the UFSAR was required. The FFD program did not result in new or modified safety analyses or design bases. Provided that the UFSAR does not contain security-related information affected by FFD program implementation, no change or addition to the UFSAR is required as a result of the new requirement.
7 FREQUENCY OF REQUIRED UPDATES As required by 10 CFR 50.71(e)(4), licensees are required to submit a periodic UFSAR update annually or within six months after each refueling outage provided the interval between successive updates does not exceed 24 months. Licensees may request an exemption from this requirement from the NRC. For example, the NRC has granted exemptions allowing licensees to submit a single, combined periodic update for multi-unit plants.
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TREATMENT OF LONG-TERM TEMPORARY MODIFICATIONS The UFSAR is revised periodically to assure that the information reflects the latest material developed.
Nevertheless, at any given time there may be a number of temporary plant and/or procedure changes in effect to support corrective actions or other plant activity. Temporary changes in support of plant operations should be restored to the normal plant condition, e.g., consistent with the UFSAR, in a timely manner. For temporary conditions involving safety-related equipment, timely restoration is required by 10 CFR 50, Appendix B. Per Generic Letter 91-18, Revision 1, temporary conditions subject to Appendix B that exist longer than the next refueling outage are to be explicitly justified as part of tracking documentation.
Temporary changes generally should not be reflected in UFSAR updates. Because UFSAR information may lag the current plant status by up to 30 months, the UFSAR is an inefficient vehicle for documenting temporary conditions. This would cause licensees to needlessly revise information in the UFSAR that would shortly revert to its prior condition. The result would be a UFSAR that described temporary modifications that are no longer installed, and the UFSAR would not reflect their removal until the next periodic update.
Temporary changes are administratively controlled separately from the UFSAR, and the current status of each is tracked to completion. Tracking documentation ensures that plant staff can determine the current plant status to support ongoing plant operations, including evaluations performed under 10 CFR 50.59. For temporary changes subject to 10 CFR 50.59, evaluations are performed and a summary report is submitted to the NRC in accordance with 10 CFR 50.59(b).
In general, UFSARs should not duplicate the licensee's tracking and reporting of temporary changes. However, the licensee should reflect in periodic UFSAR updates temporary modifications meeting both of the following criteria:
I The temporary modification is expected to be in place throughout the next required periodic UFSAR update cycle3, or no schedule for removal has been established; and I
The licensee determines based on the update guidance in Section 6 that the temporary modification should be reflected in the next required UFSAR update.
Temporary modifications reflected in the UFSAR should be clearly identified as such to distinguish temporary conditions from the permanent plant design and normal operation.
Consistent with licensee configuration control procedures, there may be temporary modifications reflected in the UFSAR that are not reflected in other permanent plant documentation.
3 A periodic update cycle is the period between the cutoff dates for new information for successive required UFSAR updates, i.e., from six months (maximum) prior to submittal of one update until six months (maximum) prior to the next.
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Examples 1.
A temporary modification was installed for six months to defeat an alarm that is explicitly discussed in the UFSAR. While the temporary modification affected information contained in the UFSAR, it would not be included in the periodic update because the alarm is expected to be restored to service before the end of the next required periodic UFSAR update cycle.
2.
Temporary cables for an intercom system have been routed through one of the safety-related battery rooms, and the permanent installation is not planned for more than two years. Based on this schedule, and the schedule for the next UFSAR update, the temporary modification is expected to be in place until after the next UFSAR update cycle. Therefore, the temporary modification should be evaluated per Section 6 of the guideline for inclusion in the next required UFSAR update. Because the modification does not affect existing UFSAR information, and does not result in new safety analyses, design bases or UFSAR description, this modification would not be reflected in the next required UFSAR update.
3.
A temporary modification was installed for a safety injection accumulator makeup water pump, and the permanent resolution of the issue will not be implemented for at least two more years. Based on this schedule, and the schedule for the next UFSAR update, the temporary modification is expected to be in place until after the next UFSAR update cycle and should be evaluated per Section 6 of the guidance for inclusion in the next required update. Because the modification affects the existing description of the makeup function for the safety injection accumulators, the UFSAR should be modified to reflect the temporary modification as part of the next required update.
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TREATMENT OF DISCREPANCIES BETWEEN THE FACILITY AND THE UPDATED FSAR If the licensee discovers a discrepancy between the facility and its description in the UFSAR, the licensee should address the discrepancy in accordance with its corrective actions program under 10 CFR 50, Appendix B. If evaluation of the discrepancy results in the identification of a nonconforming or degraded plant condition that may impact the operability of the associated structures, systems and components, the nonconforming or degraded condition should be addressed in accordance with Generic Letter 91-18, Revision 14.
If evaluation of the discrepancy determines that the UFSAR is incorrect, a correction should be initiated in accordance with licensee procedures for inclusion in the next UFSAR update.
4 Licensees also should evaluate nonconforming or degraded conditions for reportability pursuant to NRC requirements.
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June 1999 Appendix A Page 1 APPENDIX A: MODIFYING THE UPDATED FSAR A1 INTRODUCTION As discussed in this guideline, 10 CFR 50.71(e) requires that changes and certain new information be incorporated in periodic updates to the UFSAR. As provided in this appendix, the licensee also may initiate voluntary modifications to the UFSAR unrelated to plant changes or required updates under 10 CFR 50.71(e)to improve its focus, clarity and maintainability. The following sections provide guidance for reformatting, simplifying and removing existing UFSAR information. While not discussed in this document, licensees also may add information that goes beyond regulatory requirements and guidance to facilitate use of the UFSAR by plant staff or for other purposes.
A2 CONTROLLING MODIFICATIONS TO THE UPDATED FSAR As discussed in the following sections, three types of modifications may be made to the information in the UFSAR: reformatting, simplification and removal. UFSAR modifications discussed in Sections A3 through A5 that are not the result of changes to the plant or procedures do not require evaluation under 10 CFR 50.59, but they should be administratively controlled through a process that has the following attributes:
I The licensee process controls what and how information is reformatted, simplified or removed from the UFSAR.
I The licensee process ensures that the UFSAR continues to contain the necessary scope of information as discussed in Section 5 of this guideline.
I As discussed in Section A6, the NRC should be informed of information removed from the UFSAR and the basis for the licensees determination that such information may be removed. This information should be specifically identified to the NRC as part of required UFSAR updates, i.e., in addition to the changed pages and a list of effective pages currently required by 10 CFR 50.71(e).
I It is the intent of this guideline to help licensees remove unimportant information from UFSARs such as excessive detail, obsolete information, or redundant information. This guideline is not intended to be used to remove important information from UFSARs about features or functions of SSCs that insights from operating experience or probabilistic risk assessments indicate are risk-significant.
The intent that risk-significant information be retained does not preclude removal of obsolete or redundant information, or excessive detail concerning the design or operation of risk-significant SSCs, provided that the action is consistent with the guidance in this Appendix.
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June 1999 Appendix A Page 2 A3 REFORMATTING OF UPDATED FSAR INFORMATION Neither 10 CFR 50.34(b) nor 10 CFR 50.71(e) contain requirements on the format of FSARs. Thus the format of the UFSAR is at the option of the licensee, and the licensee may change the format of the UFSAR provided the content of the UFSAR is maintained consistent with these regulations, regulatory guidance committed to by the licensee (e.g.,
Regulatory Guide 1.70), and this guideline. For example, a licensee may elect to reformat the UFSAR to more clearly identify the design bases as defined in 10 CFR 50.2.
Historical information provided in the original FSAR may have become out-of-date and is not expected to be used to support current or future plant operations or regulatory activities. Accordingly, it may be appropriate to reformat such information to distinguish it from UFSAR information actively maintained by licensees to describe the updated plant design and operation.
By definition, reformatting UFSAR informationsuch as designating certain information as historical or relocating historical information to an appendixdoes not remove that information from the UFSAR. As such, changes the licensee initiates to historical information constitute changes to the UFSAR that must be reported to the NRC per 10 CFR 50.71(e).
Absent an NRC requirement, licensees need not update historical information in UFSARs to reflect minor changes in population data or other such changes in the site environment.
However, licensees should evaluate potentially significant changes in the site environs, e.g., a new natural gas line within the site boundary or a major new industrial facility near the plant site, to determine if notification of NRC and appropriate update of the UFSAR are required. For example, 10 CFR 50.9 requires licensees to notify the Commission of information identified by the applicant or licensee as having for the regulated activity a significant implication for public health and safety or common defense and security.
Because changes to historical information as defined in this guideline are generally not expected or required (except possibly to reflect a significant change in the site environs, as discussed above), licensee update of such information under 10 CFR 50.71(e) is not expected.
The following are examples of historical information:
I Description of pre-service inspections I
Description of preoperational tests I
Description of start-up tests I
Description of station organization for initial licensing I
Comparative plant data provided to support original plant licensing UFSAR Revision 28.0
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Industry or other data obtained to support or develop the original plant design bases, including that relating to natural or man-made phenomena such as geography, meteorology, hydrology, geology, seismology, population density and nearby facilities (typically in Chapter 2 of the UFSAR)5 I
Lists of references, figures and submittals relevant only to the original licensing proceeding I
Description of original factory testing of plant equipment, e.g., emergency diesel generators.
Licensees may reformat such historical information by either of the following, or equivalent, methods:
I Qualifying information may be designated as historical via clear annotation in the UFSAR.
I Historical information may be relocated to separate volumes or to specially designated appendices of the UFSAR.
Reformatting of UFSAR information should be controlled in accordance with Section A2.
A4 SIMPLIFYING UPDATED FSAR INFORMATION Licensees may elect to simplify information contained in the UFSAR to improve its focus, clarity and maintainability. As discussed in the subsections below, licensees may simplify UFSAR information by removing excessive detail and by using references to other documents where appropriate.
A4.1 REMOVING EXCESSIVE DETAIL UFSARs contain the scope of information required for the original FSAR by 10 CFR 50.34(b) and the additions to that scope required by 10 CFR 50.71(e). Later license applicants included significantly more detailed information in original FSARs than did earlier applicants. More recent FSARs grew to be 20 to 30+ volumes and may include more detail in certain respects than was absolutely necessary to support NRC safety and licensing reviews.
Removal of excessively detailed text and drawings can improve the focus of UFSARs on significant descriptive, design bases, operational and analytical information that is relevant and useful to support current and future operational and regulatory activities.
Detailed text and drawings may be removed from the UFSAR to the extent that the 5 While data and information supporting the original plant design bases for natural and man-made phenomena may be designated as historical, the associated design bases themselves should not. This is because the original design bases continue to be part of the overall design bases for the facility, and new information may warrant their update.
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The following types of excessively detailed textual information may be removed from UFSARs, except as indicated by applicable regulatory guidance or NRC Safety Evaluation Reports:
I Descriptive information that is not important to providing an understanding of the plant's design and operation from either a general or system functional perspective, e.g., component model numbers I
Design information that is not important to the description of the facility or presentation of its safety analysis and design bases, e.g., component details such as specific motor horsepower ratings for MOVs I
Design information that, if changed during the life of the plant, would have no impact on the ability of plant systems, structures and components described in the UFSAR to perform their design basis function(s), e.g., specific HVAC equipment capacity and flow rate information for structures that do not contain equipment that performs design basis functions I
Analytical information, e.g., detailed calculations, that is not important to providing an understanding of the safety analysis methodology, input assumptions and results, and/or compliance with relevant regulatory and industry standards.
Removal of excessively detailed information from the UFSAR should be controlled in accordance with Section A2, including reporting to NRC as discussed in Section A6.
A4.2 REPLACING DETAILED DRAWINGS WITH SIMPLIFIED SCHEMATICS Detailed drawings, such as piping and instrumentation diagrams (P&IDs), typically contain engineering and component information that goes beyond that appropriate to complement the textual descriptions in the UFSAR and beyond that necessary to aid in understanding of the system design and principal functions. Examples of such information contained in detailed drawings include pipe line numbers, vents and drains, etc.
Simplified schematics may be substituted for detailed drawings under either of the following conditions: (1) the original FSAR contained simplified schematics that the licensee had later replaced with P&IDs or other detailed drawings as a matter of convenience, or (2) the original FSAR included detailed drawings, but simplified schematics will be substituted such that they will not result in removal of information required to be in the UFSAR.
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June 1999 Appendix A Page 5 In the first case, licensees may substitute simplified schematics for detailed drawings because simplified schematics were provided in the original FSAR. Returning simplified schematics to the UFSAR would be consistent with the intent of the FSAR update rule that the level of detail of the UFSAR should be at least the same as that provided in the original FSAR.
In the second case, the licensee would need to ensure that simplified drawings together with associated UFSAR text continue to provide sufficient understanding of design bases, safety analyses and facility operation. For example, if the licensee determines that design bases or safety analyses information is contained in detailed drawings that is not conveyed by text, tables or other means in the UFSAR, the licensee should incorporate the information into the simplified schematic or other UFSAR information so that the UFSAR continues to contain all necessary information. Substitution for detailed drawings as described in this paragraph should be controlled in accordance with Section A2, including reporting to NRC as discussed in Section A6.
In some cases, UFSAR drawings may contain little or no information that is necessary or important to provide sufficient understanding of a facility's design bases, safety analyses or operation. Such drawings may be eliminated completely from UFSARs provided both of the following conditions are met:
I Inclusion of the drawing in the UFSAR must not be part of an existing licensee commitment to the NRC (e.g., to Regulatory Guide 1.70)
I Existing UFSAR text, tables, and other information provide sufficient understanding of a facility's design bases, safety analyses and operation or the licensee supplements the existing information to compensate for the removal of the drawing.
Drawings should generally not be removed from the UFSAR where they are helpful in understanding the textual description of the design or function of important structures, systems and components. When removing drawings from the UFSAR that are to be maintained as part of other design documents, licensees should consider providing a reference in the UFSAR to the location of the drawing.
A4.3 REFERENCING OTHER DOCUMENTS IN UPDATED FSARS When assessing the presentation of existing UFSAR information (or evaluating information to be added), there may be instances when the information exists in a separate source document and it is preferable to reference, rather than duplicate, all or part of the source document in the UFSAR. Referencing, rather than duplicating, information in the UFSAR can simplify the presentation and maintenance of UFSAR information and, in some cases, avoid the need for duplicative reporting of changes to the NRC.
There are two basic ways licensees can reference other documents in the UFSAR depending on the nature of the document and the purpose of the reference. Each is discussed below.
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June 1999 Appendix A Page 6 General References. General references are not considered part of the UFSAR, but are intended to provide background information or additional detail that the reader may refer to in order to learn more about particular material presented in the UFSAR. These may be texts, environmental studies or technical reports, as well as licensee-controlled documents such as operating or maintenance procedures, calculation manuals, etc.
References to such information may be located at specific points in the UFSAR, or they may be listed at the end of UFSAR chapters or in introductory sections.
Licensees may wish to remove excessively detailed, duplicate UFSAR information that is controlled in a separate licensee source document. In some cases, it may be appropriate to provide a brief summary of the detail being removed and/or a general reference to the controlling document as an aid to the reader. Unless the referenced source document is incorporated by reference (as discussed below), referenced information is not part of the UFSAR and would not be subject to 10 CFR 50.71(e), except as specifically committed to by licensees. Replacement of detailed information with a brief summary and/or reference constitutes removal of UFSAR information that must be controlled consistent with guidance in Section A2 and reported to NRC as discussed in Section A6.
Incorporation by Reference. Incorporation by reference refers to a method by which all or part of a separate source document can be made part of the UFSAR without duplicating the desired information in the UFSAR. Information that is appropriate to include in the UFSAR that is also part of a separate licensee-controlled document or technical report may be incorporated in the UFSAR by appropriate reference to that information. By relying on information incorporated by reference, licensees may simplify their UFSARs by removing information that is duplicated in separate, controlling program documents such as the Emergency Plan, Offsite Dose Calculation Manual, Fire Protection Plan and Fire Hazards Analysis Report, Security Plan, Environmental Protection Plan and Quality Assurance Plan.
Considerations when incorporating by reference include the following:
I Licensees should clearly identify in the UFSAR text the document or portion thereof to be incorporated, and state that the document or portion thereof is incorporated by reference in the UFSAR. For example, one option would be to locate in Chapter One of the UFSAR a single section or table that maintains the list of all documents considered incorporated. References should be as clear and specific as possible to avoid misunderstandings about the extent of information incorporated by reference and thus considered part of the UFSAR.
I For information to be incorporated by reference, the information must be publicly available (i.e., it must have been submitted to the NRC) unless there exists an explicit NRC requirement to maintain the information on site. Furthermore, information incorporated by reference into the UFSAR is subject to the update and reporting requirements of 10 CFR 50.71(e) and change controls of 10 CFR 50.59 unless separate NRC change control requirements apply (e.g., 10 CFR 50.54(a)).
Because documents incorporated by reference in UFSARs are subject to the requirements of 10 CFR 50.71(e) and 10 CFR 50.59 (except where separate NRC requirements apply), licensees should ensure that such documents are being maintained in accordance with these requirements. Documents incorporated by UFSAR Revision 28.0
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Licensees may control the Technical Requirements Manual and similar licensee controlled documents in either of the following ways:
I The TRM or other licensee controlled document is explicitly incorporated by reference into the UFSAR. Under this approach, the referenced document is subject to the change control requirements of 10 CFR 50.59 and the update/reporting requirements of 10 CFR 50.71(e), e.g., periodic submittal of change pages, etc.
I The TRM or other licensee controlled document is treated in a manner consistent with procedures fully or partially described in the UFSAR. Under this approach, the referenced document is maintained on-site in accordance with licensee administrative processes, and changes are evaluated using 10 CFR 50.59.
A5 REMOVING UNNECESSARY INFORMATION FROM UPDATED FSARS Licensees may remove obsolete and redundant information and commitments from UFSARs. When removing information as described in this section, licensees should follow the guidance in Section A2 for controlling modifications to the UFSAR and Section A6 for reporting to the NRC.
Obsolete Information. Licensees should remove UFSAR information, as appropriate, in connection with removal of SSCs from the plant or elimination of functions or procedures described in the UFSAR. However, licensee review of UFSAR information may identify where this has not occurred, or where removal of UFSAR information in connection with a change was incomplete. In general, licensees should remove from UFSARs description of equipment that is no longer installed in the plant; organizations, programs or procedures that are no longer in effect; and design information, evaluations or other description that no longer apply to the facility. The exception to this guidance is that programmatic information that was explicitly required under 10 CFR 50.34(b) to be included in original FSARs, e.g., plans for preoperational testing and initial operations, may not be removed from the UFSAR. Such information is considered historical; licensees may opt to reformat this information in accordance with Section A3.
Where the presence of obsolete information indicates a discrepancy between the UFSAR and the actual plant design or operation, the discrepancy should be evaluated in accordance with Section 9 of this guideline.
Organizations, programs and procedures no longer in effect are considered obsolete, as opposed to historical, if they were instituted and included in the UFSAR subsequent to initial plant licensing.
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June 1999 Appendix A Page 8 If equipment has been retired in place (equipment that is no longer in service but has not been physically removed from the plant), functional descriptions in the UFSAR that no longer apply to the equipment are considered obsolete information. To accurately reflect the condition of the plant, physical descriptions of equipment retired in place (e.g.,
component and location) in the form of text and/or drawings should be retained in the UFSAR.
Redundant Information. Licensees may remove duplicate information from the UFSAR. If some or all of the duplicated information is important to facilitate understanding of multiple sections of the UFSAR, the licensee should retain appropriate duplicate information where it is needed. Alternatively, the licensee may remove duplicate information and provide a reference to the location in the UFSAR where the information is to be retained.
Commitments. Some licensees may have incorporated specific commitments made to the NRC within the UFSAR. Consistent with Commission guidance6, licensees may remove from the UFSAR commitments that are not integral to required UFSAR information, i.e.,
design bases, safety analyses and associated description. Removal from the UFSAR does not change the status or nature of commitments to the NRC. NEI 99-00, Guideline for Managing NRC Commitment Changes,7 provides guidance for making changes to NRC commitments.
Licensees should ensure that NRC commitments removed from the UFSAR are included in licensee commitment management or corrective action programs as appropriate. If the licensee committed to the NRC to incorporate a commitment in the UFSAR, then the licensee should inform the NRC of its removal from the UFSAR, consistent with the licensees commitment change process.
A6 REPORTING TO THE NRC INFORMATION REMOVED FROM THE UFSAR Information removed from the UFSAR should be specifically identified to the NRC as part of required UFSAR updates. A brief description of the information removed and the basis for its removal should be provided. This information should not be incorporated in the UFSAR but should be provided in addition to the changed pages and a list of effective pages currently required by 10 CFR 50.71(e).
The following are examples of description suitable for notifying the NRC that information was removed from the UFSAR:
6 In a Staff Requirements Memorandum dated May 20, 1997, the Commission directed the NRC staff to formulate an approach to FSAR updates that would allow obsolete or less meaningful information and commitments to be readily removed from the FSAR.
7 NEI 99-00 was in final draft form at time of publication of this document.
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removed model number information for components of the Reactor Equipment Cooling System previously contained in UFSAR Section XYZ on the basis that this was excessively detailed information I
replaced the P&ID for the auxiliary feedwater system with a simplified schematic.
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June 1999 Appendix B Page 1 APPENDIX B: 10 CFR 50.71(e)
Note: The changes identified below to 10 CFR 50.71(e) are expected to be approved by the Commission in connection with rulemaking to amend 10 CFR 50.59 after publication of this document.
Each person licensed to operate a nuclear power reactor pursuant to the provisions of § 50.21 or § 50.22 of this part shall update periodically, as provided in paragraphs (e)(3) and (4) of this section, the final safety analysis report (FSAR) originally submitted as part of the application for the operating license, to assure that the information included in the FSAR contains the latest material developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last updated FSAR. The updated FSAR shall be revised to include the effects1 of: all changes made in the facility or procedures as described in the FSAR; all safety analyses and evaluations performed by the licensee either in support of requestedapproved license amendments or in support of conclusions that changes did not involve an unreviewed safety question require a license amendment in accordance with § 50.59(c)(2) of this part; and all analyses of new safety issues performed by or on behalf of the licensee at Commission request. The updated information shall be appropriately located within the update to the FSAR.
(1) The licensee shall submit revisions containing updated information to the Commission, as specified in § 50.4, on a replacement-page basis that is accompanied by a list which identifies the current pages of the FSAR following page replacement.
(2)
The submittal shall include (i) a certification by a duly authorized officer of the licensee that either the information accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirement, or that no such changes were made; and (ii) an identification of changes made under the provisions of § 50.59 but not previously submitted to the Commission.
(3)
(i) A revision of the original FSAR containing those original pages that are still applicable plus new replacement pages shall be filed within 24 months of either July 22, 1980, or the date of issuance of the operating license, whichever is later, and shall bring the FSAR up to date as of a maximum of 6 months prior to the date of filing the revision.
(ii) Not less than 15 days before §50.71(e) becomes effective, the Director of the Office of Nuclear Reactor Regulation shall notify by letter the licensees of those nuclear power plants initially subject to the NRC's systematic evaluation program that they need not comply with the provisions of this section while the program is being conducted at their plant. The Director of the Office of Nuclear Reactor Regulation will notify by letter the licensee of each nuclear power plant being evaluated when the systematic evaluation program has been completed. Within 24 months after receipt of this notification, the licensee shall file a complete FSAR which is up to date as of a maximum of 6 months prior to the date of filing the revision.
(4) Subsequent revisions must be filed annually or 6 months after each refueling outage provided the interval between successive updates does not exceed 24 months. The revisions must reflect all changes up to a maximum of 6 months prior to the date of filling. For nuclear power reactor facilities 1 Effects of changes includes appropriate revisions of descriptions in the FSAR such that the FSAR (as updated) is complete and accurate.
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June 1999 Appendix B Page 2 that have submitted the certifications required by §50.82(a)(1), subsequent revisions must be filed every 24 months.
(5) Each replacement page shall include both a change indicator for the area changed, e.g., a bold line vertically drawn in the margin adjacent to the portion actually changed, and a page change identification (date of change or change number or both).
(6)
The updated FSAR shall be retained by the licensee until the Commission terminates their license.
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i of ii UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 1 -
SUMMARY
................................................................................................................... 1 CHAPTER 2 - SITE AND ENVIRONMENT.............................................................................................. 2 CHAPTER 3 (U1) - REACTOR UNIT ONE............................................................................................ 3 CHAPTER 3 (U2) - REACTOR UNIT TWO............................................................................................ 4 CHAPTER 4 - REACTOR COOLANT SYSTEM....................................................................................... 5 CHAPTER 5 - CONTAINMENT SYSTEM INTRODUCTION........................................................................ 6 CHAPTER 6 - ENGINEERED SAFETY FEATURES................................................................................. 7 CHAPTER 7 - INSTRUMENTATION AND CONTROLS............................................................................. 8 CHAPTER 8 - ELECTRICAL SYSTEMS................................................................................................ 9 CHAPTER 9 - AUXILIARY AND EMERGENCY SYSTEMS...................................................................... 10 CHAPTER 10 - STEAM AND POWER CONVERSION SYSTEM.............................................................. 11 CHAPTER 11 - WASTE DISPOSAL AND RADIATION PROTECTION SYSTEM......................................... 12 CHAPTER 12 - CONDUCT OF OPERATIONS...................................................................................... 13 CHAPTER 13 - INITIAL TEST AND OPERATIONS [HISTORICAL].......................................................... 14 CHAPTER 14 (U1) - SAFETY ANALYSIS........................................................................................... 15
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ii of ii UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 14 (U2) - SAFETY ANALYSIS........................................................................................... 17 CHAPTER 15 - SUPPLEMENT FOR RENEWED OPERATING LICENSES................................................. 19
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UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 1 1.0 Introduction and Summary Engineering Configuration Programs 1.1 Plant Site Summary Engineering Mechanical Design 1.2 Design Highlights Engineering Mechanical Design 1.3 Summary Plant Description Engineering Mechanical Design 1.4 Plant Specific Design Criteria Engineering I&C Design & Mechanical Design 1.5 Plant Operation Engineering Configuration Programs 1.6 Research and Development Requirements Engineering Configuration Programs 1.7 Quality Assurance Performance Assurance Performance Compliance Manager 1.8 Identification of Contractors Engineering Mechanical Design 1.9 Facility Safety Conclusions Engineering Mechanical Design 1 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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2 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 2 - SITE AND ENVIRONMENT UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 2 2.1 Site Description Engineering Structural Design 2.2 Meteorology Plant Manager Environmental 2.3 Geology Engineering Structural Design 2.4 Hydrology Engineering Structural Design 2.5 Engineering Engineering Structural Design 2.6 Limnology and Ecology Plant Manager Environmental 2.7 Radiological Environmental Monitoring Program Plant Manager Environmental 2.8 Plant Design Bases Dependent Upon Site & Environs Characteristics Engineering Structural Design 2.9 Plant Design Criteria for Structures & Equipment Engineering Structural Design 2.10 Conclusions Engineering Structural Design 2 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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3 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 3 (U1) - REACTOR UNIT ONE UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 3 3.1 Summary Description Engineering Fuel Section 3.2 Mechanical Design Engineering Fuel Section 3.3 Nuclear Design Engineering Nuclear Safety Analysis 3.4 Thermal and Hydraulic Design Engineering Nuclear Safety Analysis 3.5 Westinghouse Reload Fuel Engineering Fuel Section 3.5.1 Fuel Mechanical Design Engineering Fuel Section 3.5.2 Nuclear Design Engineering Fuel Section 3.5.3 Thermal and Hydraulic Design for Cores Containing 15x15 Upgrade Fuel Engineering Nuclear Safety Analysis 3 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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4 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 3 (U2) - REACTOR UNIT TWO UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 4 3.1 Summary Description Engineering Fuel Section 3.2 Mechanical Design Engineering Fuel Section 3.3 Nuclear Design Engineering Nuclear Safety Analysis 3.4 Thermal and Hydraulic Design Engineering Nuclear Safety Analysis 4 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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5 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 4 - REACTOR COOLANT SYSTEM UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 5 4.1 Design Bases Engineering Mechanical Design 4.2 System Design and Operation Engineering Mechanical Design 4.3 System Design Evaluation Engineering Mechanical Design 4.4 Safety Limits and Conditions Engineering Mechanical Design 4.5 Testing and Inspection Engineering Mechanical Design 5 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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6 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 5 - CONTAINMENT SYSTEM INTRODUCTION UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 6 5.1 General Design Criteria Engineering Structural Design 5.2 Containment Structure Engineering Structural Design 5.3 Ice Condenser Engineering Mechanical Design 5.4 Containment Isolation System Engineering Mechanical Design 5.5 Containment Ventilation System Engineering Mechanical Design 5.6 Containment Penetration and Weld Channel Pressurization Engineering Structural Design 5.7 Containment Structure, Inspection and Test Engineering Structural Design 5.8 Distributed Ignition System Engineering Electrical Design 6 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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7 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 6 - ENGINEERED SAFETY FEATURES UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 7 6.1 Criteria Engineering Mechanical Design 6.2 Emergency Core Cooling System Engineering Mechanical Design 6.3 Containment Spray System Engineering Mechanical Design 7 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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8 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 7 - INSTRUMENTATION AND CONTROLS UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 8 7.1 General Design Criteria Engineering I & C Design 7.2 Protective Systems Engineering I & C Design 7.2.1 Protective Systems Engineering I & C Design 7.2.2 Protective Systems Design Engineering I & C Design 7.2.3 Protective Systems Evaluation Engineering I & C Design 7.3 Control Systems Engineering I & C Design 7.4 Nuclear Instrumentation Engineering I & C Design 7.5 Engineered Safety Features Instrumentation Engineering I & C Design 7.6 In-Core Instrumentation Engineering I & C Design 7.7 Operating Control Stations Engineering I & C Design 7.8 Post Accident Monitoring Instrumentation Engineering I & C Design 8 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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9 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 8 - ELECTRICAL SYSTEMS UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 9 8.1 Design Bases Engineering Electrical Design 8.2 Network Interconnections Engineering Electrical Design 8.3 Station Service Systems Engineering Electrical Design 8.3.1 4160 Volt System Engineering Electrical Design 8.3.2 Low Voltage Power Systems Engineering Electrical Design 8.3.3 120 Volt AC Vital Instrument Bus System Engineering Electrical Design 8.3.4 250 Volt DC System Engineering Electrical Design 8.3.5 250 Volt DC Battery N System Engineering Electrical Design 8.3.6 Lighting System Engineering Electrical Design 8.4 Emergency Power System Engineering Electrical Design 8.5 Design Evaluation Engineering Electrical Design 8.6 Test & Inspections Engineering Electrical Design 8.7 Station Blackout Engineering Electrical Design 9 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 10 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 9 - AUXILIARY AND EMERGENCY SYSTEMS UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 10 9.1 General Design Criteria Engineering Mechanical Design 9.2 Chemical and Volume Control System Engineering Mechanical Design 9.3 Residual Heat Removal System Engineering Mechanical Design 9.4 Spent Fuel Pit Cooling System Engineering Mechanical Design 9.5 Component Cooling System Engineering Mechanical Design 9.6 Sampling Systems Engineering Mechanical Design 9.7 Reactor Components and Fuel Handling System Engineering Mechanical Design 9.8 Facility Service Systems Engineering Mechanical Design 9.8.1 Fire Protection System Engineering Mechanical Design 9.8.2 Compressed Air System Engineering Mechanical Design 9.8.3 Service Water Systems Engineering Mechanical Design 9.9 Auxiliary Building Ventilation System Engineering Mechanical Design 9.10 Control Room Ventilation System Engineering Mechanical Design 10 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 11 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 10 - STEAM AND POWER CONVERSION SYSTEM UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 11 10.1 General Description Engineering Mechanical Design 10.2 Main Steam System Engineering Mechanical Design 10.3 Turbine Generator Engineering Mechanical Design 10.4 Main Condensers Engineering Mechanical Design 10.5 Condensate and Feedwater System Engineering Mechanical Design 10.6 Circulation Water System Engineering Mechanical Design 10.7 Turbine Aux. Cooling Water System Engineering Mechanical Design 10.8 Service Water System Engineering Mechanical Design 10.9 Make-up Water and Primary Water Systems Engineering Mechanical Design 10.10 Chemical Feed Sub-System Engineering Mechanical Design 10.11 Secondary Vent and Drain System Engineering Mechanical Design 11 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 12 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 11 - WASTE DISPOSAL AND RADIATION PROTECTION SYSTEM UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 12 11.1 Waste Disposal System Engineering Mechanical Design 11.2 Plant Radiation Shielding Engineering Mechanical Design 11.3 Radiation Monitoring System Engineering I&C Design &
Mechanical Design 11.4 Plant Health Physics Program Plant Manager Radiation Protection 11.5 Steam Generator Blowdown Treatment System Engineering Mechanical Design 11.6 Radioactive Materials Safety Plant Manager Radiation Protection 12 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 13 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 12 - CONDUCT OF OPERATIONS UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 13 12.1 Organization and Responsibility Plant Manager Operations 12.2 License Operator Requalification Program Plant Manager Operations 12.2.1 Control of Heavy Loads Engineering Configuration Programs 12.3 Emergency Plan Engineering Configuration Programs 12.4 Records Engineering Configuration Programs 12.5 Review and Audit of Operations Engineering Configuration Programs 12.6 Nuclear Design and Support Capability Engineering Configuration Programs 12.7 Written Procedures Engineering Configuration Programs 13 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 14 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 13 - INITIAL TEST AND OPERATIONS [HISTORICAL]
UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 14 13.1 Tests Prior to Reactor Fueling Engineering Configuration Programs 13.2 Final Station Preparation Engineering Configuration Programs 13.3 Initial Testing of the Operating Reactor Engineering Configuration Programs 13.4 Operating Restrictions Engineering Configuration Programs 14 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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TCO Page: 15 of 19 UFSAR TECHNICAL CONTENT OWNER (TCO) ASSIGNMENT LIST CHAPTER 14 (U1) - SAFETY ANALYSIS UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 15 14 (U1)
SAFETY ANALYSIS CHAPTER 14 UNIT 1 14.1 Core and Coolant Boundary Protection Analysis Engineering Nuclear Safety Analysis 14.2 Standby Safeguards Analysis Engineering Nuclear Safety Analysis 14.2.1 Fuel Handling Engineering Nuclear Safety Analysis 14.2.2 Accidental Release of Radioactive Liquids Engineering Nuclear Safety Analysis 14.2.3 Accidental Waste Gas Release Engineering Nuclear Safety Analysis 14.2.4 Steam Generator Tube Rupture Engineering Nuclear Safety Analysis 14.2.5 Rupture of a Steam Pipe Engineering Nuclear Safety Analysis 14.2.6 Rupture of a Control Rod Drive Mechanism (CRDM) Housing (RCCA Ejection)
Engineering Nuclear Safety Analysis 14.2.7 Environmental Consequences Following Secondary System Accidents Engineering Nuclear Safety Analysis 14.2.8 Major Rupture of a Main Feedwater Pipe Engineering Nuclear Safety Analysis 14.3 Reactor Coolant System Pipe Rupture (Loss of Coolant Accident)
Engineering Nuclear Safety Analysis 14.4 Environmental Qualification Analyses Engineering Nuclear Safety Analysis 15 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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SAFETY ANALYSIS CHAPTER 14 UNIT 1 14.4.1 Basis of Discussion Engineering Nuclear Safety Analysis 14.4.2 Postulated Pipe Failure Analysis Outside Containment Engineering Nuclear Safety Analysis 14.4.3 Analysis of Emergency Conditions Engineering Nuclear Safety Analysis 14.4.4 Stress Calculations Engineering Nuclear Safety Analysis 14.4.5 Description of Pipe Whip Analysis Engineering Nuclear Safety Analysis 14.4.6 Compartment Pressures and Temperatures Engineering Nuclear Safety Analysis 14.4.7 Description of Jet Impingement Load Analysis Engineering Nuclear Safety Analysis 14.4.8 Containment Integrity Engineering Nuclear Safety Analysis 14.4.9 Plant Modifications Engineering Nuclear Safety Analysis 14.4.10 Environment Engineering Nuclear Safety Analysis 14.4.11 Electrical Equipment Environmental Qualification Engineering Nuclear Safety Analysis 14.4.11.2.6 Radiation Engineering Nuclear Safety Analysis 14.A Radiation Sources (Appendix 14A)
Engineering Nuclear Safety Analysis 14.G Core and Internals Integrity Analysis Engineering Nuclear Safety Analysis
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SAFETY ANALYSIS CHAPTER 14 UNIT 2 14.1 Core and Coolant Boundary Protection Analysis Engineering Nuclear Safety Analysis 14.2 Standby Safeguards Analysis Engineering Nuclear Safety Analysis 14.2.
Radiological Consequences of Fuel Handling Accident Engineering Nuclear Safety Analysis 14.2.2.
Postulated Radioactive Releases Due to Liquid Containing Tank Failure Engineering Nuclear Safety Analysis 14.2.3 Accidental Waste Gas Release Engineering Nuclear Safety Analysis 14.4 Steam Generator Tube Rupture Engineering Nuclear Safety Analysis 14.2.5 Rupture of Steamline (Steamline Break)
Engineering Nuclear Safety Analysis 14.2.6 Rupture of Control Rod Drive Mechanism (CRDM) Housing (RCCA Ejection)
Engineering Nuclear Safety Analysis 14.2.7 Secondary Systems Accident Environmental Consequences Engineering Nuclear Safety Analysis 14.2.8 Major Rupture of Main Feedwater Pipe (Feedline Break)
Engineering Nuclear Safety Analysis 14.3 Reactor Coolant System Pipe Rupture (Loss of Coolant Accident)
Engineering Nuclear Safety Analysis 14.3.1 Large Break Loss-of-Coolant Accident Analyses Engineering Nuclear Safety Analysis 16 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.
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SAFETY ANALYSIS CHAPTER 14 UNIT 2 14.3.2 Loss of Reactor Coolant from Small Ruptured Pipes OR From Cracks in Large Pipes which Actuates the ECCS Engineering Nuclear Safety Analysis 14.3.3 Core and Internals Integrity Analysis Engineering Nuclear Safety Analysis 14.3.4 Containment Integrity Analysis Engineering Nuclear Safety Analysis 14.3.5 Radiological Consequences of a Loss of Coolant Accident and Other Events Considered in Safety Analysis Engineering Nuclear Safety Analysis 14.3.6 Hydrogen in the Unit 2 Containment after a Loss-of-Coolant Accident Engineering Nuclear Safety Analysis 14.3.7 Long Term Cooling Engineering Nuclear Safety Analysis 14.3.8 Nitrogen Blanketing Engineering Nuclear Safety Analysis 14.3.9 Containment Sump Inventory Analysis Engineering Nuclear Safety Analysis 14.4 Environmental Qualification Analyses Engineering Nuclear Safety Analysis 14.4.1 Basis of Discussion Engineering Nuclear Safety Analysis 14.4.2 Postulated Pipe Failure Analysis Outside Containment Engineering Nuclear Safety Analysis 14.4.3 Safe Shutdown Following Pipe Rupture Engineering Nuclear Safety Analysis 14.4.4 Stress Calculations Engineering Nuclear Safety Analysis 14.4.5 Description of Pipe Whip Analysis Engineering Nuclear Safety Analysis 14.4.6 Compartment Pressure and Temperatures Engineering Nuclear Safety Analysis 14.4.7 Description of Jet Impingement Load Analysis Engineering Nuclear Safety Analysis
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SAFETY ANALYSIS CHAPTER 14 UNIT 2 14.4.8 Concrete Structure Evaluation Engineering Nuclear Safety Analysis 14.4.9 Plant Modifications Engineering Nuclear Safety Analysis 14.4.10 Environment Engineering Nuclear Safety Analysis 14.4.11 Electrical Equipment Environmental Qualification Engineering Nuclear Safety Analysis 14.A Radiation Sources (Appendix 14A)
Engineering Nuclear Safety Analysis CHAPTER 15 - SUPPLEMENT FOR RENEWED OPERATING LICENSES UFSAR Section Chapter / Section Title Responsible Organization Responsible Group 17 15.1 Aging Management Programs And Activities Engineering Engineering Programs 15.2 Evaluation Of Time-Limited Aging Analyses Engineering Engineering Programs 17 Specific TCO assignments are made by the appropriate manager from the organization listed under the Responsible Organization section.