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| {{Adams | | {{Adams |
| | number = ML013100305 | | | number = ML13350A262 |
| | issue date = 11/30/2001 | | | issue date = 07/31/1978 |
| | title = (Revision 2), Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants | | | title = Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, for Comment |
| | author name = | | | author name = |
| | author affiliation = NRC/RES | | | author affiliation = NRC/RES |
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| | docket = | | | docket = |
| | license number = | | | license number = |
| | contact person = Graves H (301)415-5880 | | | contact person = |
| | case reference number = DG-1100
| | | document report number = RG-1.143 |
| | document report number = RG-1.143, Revision 2 | |
| | document type = Regulatory Guide | | | document type = Regulatory Guide |
| | page count = 28 | | | page count = 6 |
| }} | | }} |
| {{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION Revision 2 November 2001 REGULATORY | | {{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION |
| GUIDE
| | July 1978 |
| OFFICE OF NUCLEAR REGULATORY RESEARCH
| | 00 |
| REGULATORY GUIDE 1.143 (Draft was issued as DG-1100)
| | REGULATORY GUIDE. |
| DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT
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| SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN
| | OFFICE OF STANDARDS DEVELOPMENT |
| LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
| | REGULATORY GUIDE 1.143 DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND |
| | COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS |
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| ==A. INTRODUCTION== | | ==A. INTRODUCTION== |
| This regulatory guide has been revised to provide guidance to licensees and applicants on methods acceptable to the staff for complying with the NRC's regulations in the design, construction, installation, and testing the structures, systems, and components of radioactive waste management facilities in light- water-reactor nuclear power plants.
| | Paragraph (a) of k 50.34, "Contents of applica- tions; technical information," |
| | of 10 CFR Part 50, |
| | "Domestic Licensing of Production and Utilization Facilities," requires that each application for a con- struction permit include a preliminary safety analysis report. Part of the information required is a prelimi- nary design of-the facility, including among other things the principal design criteria for the facility. |
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| | Appendix A, "General Design Criteria for Nuclear Power Plants," |
| | to 10 CFR Part 50 establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants. |
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| In 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," § 50.34,
| | Criterion I, "Quality Standards and Records," of A'ppendix A requires that structures, systems, and components important to safety be designed, fabri- cated, erected, and tested to quality standards com- mensurate with the importance to safety of the s function to be performed. Criterion 2, "Design ses for Protection Against Natural Phenome,: |
| "Contents of Applications; Technical Information," requires that each application for a construction permit include a preliminary safety analysis report. Part of the information required is related to quality assurance and the preliminary design of the facility, including, among other things, the principal design criteria for the facility. Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 establishes overall quality assurance requirements for structures, systems, and components important to safety. Appendix A, ''General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 establishes minimum requirements for the principal design criteria for light-water-cooled nuclear power plants. | | " o |
| | - |
| | pendix A requires, among other thin |
| | "t st st P |
| | tures. systems. and components impo" |
| | t to fety be designed to withstand the effect |
| | ,ural phenomena such as earthquakes without lo- f capa- bility to perform their ,4,t* functions and that the design bases for these truc~rs, systems, and com- ponents reflect th hi1p' |
| | ce "the safety functions to be perform t. c¶-eri060, "Control of Releases of RadioapiivlMate~ils to the Environment-," of Appen A rAA |
| | irs ts at the nuclear power unit de- sign in ma' |
| | control suitably the release of radioactiv naterials in gaseous and liquid effluents and to handi# radioactive solid waste produced during normal reactor operation, including anticipated opera- tional occurrences. |
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| Criterion 1, "Quality Standards and Records,'' of Appendix A requires that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards Regulatory guides are issued to describe and make available to the public such information as methods acceptable to the NRC staff for implementing specific parts of the NRCs regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed by the NRC staff in its review of applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit or license by the Commission.
| | .This guide furnishes design guidance acceptable to the NRC staff relating to seismic and quality group classification and quality assurance provisions for radioactive waste management systems, structures, and components. Further, it describes provisions for controlling releases of liquids containing radioactive materials, e.g., spills or tank overflows, from all plant systems outside reactor containalog. |
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| This guide was issued after consideration of comments received from the public. Comments and suggestions for improvements in these guides are encouraged at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be submitted to the Rules and Directives Branch, ADM, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
| | B. DISCU ,$Sll3I |
| | *. |
| | One aspect of nuclear. 0o l |
| | -cration is [he control and manage g, geous. and solid radioactive wast g' crated as a byprod- uct of nuclear |
| | . |
| | ,ose of this guide is to provide i ma d |
| | eria that will provide rea- sonable St components and structures use the. |
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| Regulatory guides are issued in ten broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental and Siting;
| | ive waste management and steam en' |
| 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General.
| | b down systems are designed. con- c dh,'stalled, and tested on a level commensu- he need to protect the health and safety of pu lic and plant operating personnel. It sets forth m nimum staff recommendations and is not intended to prohibit the implementation of more rigorous de- sign considerations, codes, standards, or quality as- surance measures. |
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| Single copies of regulatory guides (which may be reproduced) may be obtained free of charge by writing the Distribution Services Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to DISTRIBUTION@NRC.GOV. Electronic copies of this guide and other recently issued guides are available on the internet at NRCs home page at <WWW.NRC.GOV> in the Reference Library under Regulatory Guides. This guide is also in the Electronic Reading Room through NRCs home page, Accession Number ML013100305.
| | Working Group ANS-55, Radioactive Waste Sys- tems, of Subcommittee ANS-50, Nuclear Power Plant System Engineering, of the American Nuclear Society Standards Committee has developed stand- ards that establish requirements and provide recom- mendations for the design, construction, and per- formance of BWR (ANSI N197-1976) and PWR |
| | (ANSI N199-1976) liquid radioactive waste process- ing systems. Standards for gaseous and solid radioac- tive waste processing systems are being developed. |
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| commensurate with the importance to safety of the safety function to be performed and that a quality assurance program be established and implemented in order to provide adequate assurance that these structures, systems, and components will satisfactorily perform their safety function.
| | I Radioactive waste, as used in this guide. means those liq- uids. gases, or solids containing radioactive materials thatl hy design or operating practice will be processed prior itt final dis- position. |
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| Criterion 2, "Design Bases for Protection Against Natural Phenomena," of Appendix A
| | USNRC REGULATORY GUIDES |
| requires, among other things, that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes, tornados, or flooding without loss of capability to perform their safety functions. The design bases for these structures, systems, and components are to reflect the importance of the safety functions to be performed.
| | Comment% should be sent to the Secretay of the Co.mmnin. US. Nutii t-q., |
| | Regulatory Guides are issued to describe and make available to khe Public method, latory Commission. Waihington, D.C. |
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| Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, of 10 CFR Part 50
| | 20555. Attention Docketnrg. a |
| states general design requirements for the implementation of General Design Criterion 2.
| | "'.,r . |
| | accepttable to the NRC nuatl of implementing specific parts of the Commission's Branch. |
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| Criterion 60, "Control of Releases of Radioactive Materials to the Environment,'' of Appendix A
| | regultations, to delineate techniques used by the &talf in evaluating specific problems The guide are issued in the followmng ten broad dniori$ |
| requires that the nuclear power unit design include means to suitably control the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid waste produced during normal reactor operation, including anticipated operational occurrences. The release of radioactive materials from external man-induced events and design basis accidents must also be controlled.
| | or postulated accidents, or to Provide guidance to applicants. Regulatory Guide: |
| | are not slubstitutes for regul-tlions, and compliance with them is not required. |
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| This regulatory guide is being revised to provide design guidance acceptable to the NRC
| | ===1. Power Reactors === |
| staff in regard to natural phenomena hazards, internal and external man-induced hazards, and quality group classification and quality assurance provisions for radioactive waste management systems, structures, and components.1 Further, it describes provisions for mitigating design basis accidents and controlling releases of liquids containing radioactive materials, e.g., spills or tank overflows, from all plant systems outside reactor containment.
| | 6. Producft Methods and solutions different from those set Out in the guides will be acce't. |
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| Licensees and applicants may propose means other than those specified by the provisions of the Regulatory Position of this guide for meeting applicable regulations. No new requirements are being imposed by this regulatory guide. Implementation of this guidance by licensees will be on a strictly voluntary basis.
| | 2. Research and Test Reactors |
| | 7. Trarrsaotlation able if they provide a basis for the findingst requiite to the issuance at continuance |
| | 3. Fuels and Materials Facilities |
| | 8. Occupational Health of a permit at license by the Commission. |
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| The information collections contained in this regulatory guide are covered by the requirements in 10 CFR Part 50, which were approved by the Office Management and Budget, approval number 3150-0011. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
| | 4. Environmental and Siting |
| | 9. Antitrusl Re.,ew |
| | 5.Vleterits alnd Piasti Pratect'On tO. Generat Comments and suggestionr% for irrlprovements in these guides are encouraged at all times, and guides will be" revised, as appropriate, to accommodate comments and Reu its for single Copies 01 issued guides iwhtch may be lsi'tuwceet orfoJ r tiltQ' |
| | to reflect nevs informatlon or experience. |
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| 1 Adams et al, "Re-evaluation of Regulatory Guidance Provided in Regulatory Guides 1.142 and 1.142, NUREG/CR-5733, August 1999. Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC
| | Howsever, comments on this guide.if mini on an automatic dtobution list tfo sinrlle cur.e, of tilt.'re quide% in r.l N |
| 20402-9328 (telephone (202)512-1800); or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road, Springfield, VA 22161; (telephone (703)487-4650; <http://www.ntis.gov/ordernow>. Copies are available for inspection or copying for a fee from the NRC Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or (800)397-4209; fax (301)415-3548; email is PDR@NRC.GOV.
| | ,eceived within about t rc, m.nths alter its issuance, wil t be particularty useful in divisionst should be made in writing to the U.S. Nuclear Regtulatory Crniin,,sJw. |
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| 1.143-2
| | evalual ing the need lor an early revision. |
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| ==B. DISCUSSION==
| | Washlington, D.C. |
| One aspect of nuclear power plant operation is the control and management of liquid, gaseous, and solid radioactive waste2 (radwaste) generated as a byproduct of nuclear power. The purpose of this guide is to provide information and criteria that will provide reasonable assurance that components and structures used in the radioactive waste management and steam generator blowdown systems are designed, constructed, installed, and tested on a level commensurate with the need to protect the health and safety of the public and plant operating personnel. It sets forth minimum staff recommendations and is not intended to prohibit the implementation of more rigorous design considerations, codes, standards, or quality assurance measures.
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| ANSI/ANS Standards 55.1-1992, Solid Radioactive Waste Processing System for Light Water Cooled Reactor Plants,3 55.4-1993, Gaseous Radioactive Waste Processing Systems for Light Water Plants,3 and 55.6-1993, Liquid Radioactive Waste Processing Systems for Light Water Reactor Plants,3 have been reviewed for applicability to this guide. These ANSI/ANS
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| Standards provide a wider range of guidance than that provided in Sections 11.2, Liquid Waste Management System; 11.3, Gaseous Waste Management System; and 11.4, Solid Waste Management System, of Chapter 11, Radioactive Waste Management, of NUREG-0800,
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| Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants.4 As appropriate, guidance from the ANSI/ANS standards has been incorporated by reference.
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| For the purposes of this guide, the radwaste systems are considered to begin at the interface valves in each line from other systems provided for collecting wastes that may contain radioactive materials and to include related instrumentation and control systems. The radwaste system terminates at the point of controlled discharge to the environment, at the point of recycle to the primary or secondary water system storage tanks, or at the point of storage of packaged solid wastes.
| | Attentiton D..ectot. Division osf Dociument Ciill, |
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| The steam generator blowdown system begins at, but does not include, the outermost containment isolation valve on the blowdown line. It terminates at the point of controlled discharge to the environment, at that point of interface with other liquid systems, or at the point of recycle back to the secondary system. For design purposes, portions of radwaste systems that interface with other systems are considered to be in the system with more rigorous requirements.
| | These standards provide more detailed guidance with regard to the specific requirements of the radioactive waste processing system than are presented in this guide. It is expected that these standards will be en- dorsed separately to be used in conjunction with this guide or that reference to applicable sections ma-- be used in future revisions to this guide. |
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| Except as noted, this guide does not apply to the reactor water cleanup system, the condensate cleanup system, the chemical and volume control system, the reactor coolant and auxiliary building equipment drain tanks, the sumps and floor drains provided for collecting liquid wastes, the boron recovery system, equipment used to prepare solid waste solidification agents, the building ventilation systems (heating, ventilating, and air conditioning), instrumentation and
| | For the purpose of this guide. the radwaste systems are considered to begin at the interface valve(s) in each line from other systems provided for collecting wastes that may contain radioactive materials and to include related instrumentation and control systems. |
| 2 Radioactive waste, as used in this guide, means liquids, gases, or solids that contain radioactive materials that by design or operating practice will be processed prior to final disposition.
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| 3 Copies may be obtained from the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60525.
| | The radwaste system terminates at the point of con- trolled discharge to the environment. at the point of recycle back to storage for reuse in the reactor, or at the point of storage of packaged solid wastes prior to shipment offsite to a licensed burial ground. The steam generator blowdown system begins at, but does not include, the outermost containment isolation valve on the blowdown line. It terminates at the point of controlled discharge to the environment, at the point of interface with other liquid systems, or at the point of recycle back to the secondary systems. Ex- cept as noted, this guide does not apply to the reactor water cleanup system, the condensate cleanup sys- tent. the chemical and volume control system, the reactor coolant and auxiliary building equipment drain tanks, the sumps and floor drains provided for. |
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| 4 Copies of sections of NUREG 0800 are available by email to DISTRIBUTION@NRC.GOV or by fax to (301)415-2289.
| | collecting liquid wastes, the boron recovery system. |
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| | equipment used to prepare solid waste solidification agents, the building ventilation systems (heating. |
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| sampling systems beyond the first root valve, or the chemical fume hood exhaust systems. In addition, this guide does not apply to the main condenser circulating or component cooling water systems, the spent fuel handing and storage systems, or the fuel pool water cleanup system.
| | ventilating, and air conditioning). or the chemical fume hood exhaust systems. |
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| The design and construction of radioactive waste management and steam generator blowdown systems should provide assurance that radiation exposures to operating personnel and to the general public are as low as is reasonably achievable. One aspect of this consideration is ensuring that these systems are designed to quality standards that enhance system reliability, operability, and availability. In developing this design guidance, the NRC staff has considered designs and concepts submitted in license applications and resulting operating system histories. It has also been guided by industry practices and the cost of design features, taking into account the potential impact on the health and safety of operating personnel and the general public. | | The design and construction of radioactive waste management and steam generator blowdown systems should provide assurance that radiation exposures to operating personnel and to the general public are as low as is reasonably achievable. One aspect of this consideration is ensuring that these systems are de- signed to quality standards that enhance system relia- bility. operability, and availability. In development of this design guidance. the NRC staff has considered designs and concepts submitted in license applica- tions and resulting operating system histories. It has also been guided by industry practices and the cost of design features, taking into account the potential im- pact on the health and safety of operating personnel and the general public. |
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| ==C. REGULATORY POSITION== | | ==C. REGULATORY POSITION== |
| 1. SYSTEMS HANDLING RADIOACTIVE MATERIALS IN LIQUIDS | | 1. Systems Handling Radioactive Materials in Liquids I.1 The liquid radwaste treatment system includ- ing the steam generator blowdown system downstream of the second containment isolation valve should meet the following criteria: |
| 1.1 Liquid Radwaste Treatment System The liquid radwaste treatment system, including the steam generator blowdown system, downstream of the outer-most containment isolation valve should meet the following criteria.
| | I.1. 1 These systems should he designed and tested to requirements set forth in the codes and standards listed in Table I supplemented by the provi- sions in I. 1.2 and in regulatory position 4 of this guide. |
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| 1.1.1 The structures, systems, and components (SSCs) of the liquid radwaste treatment system should be designed and tested to requirements set forth in the codes and standards listed in Table 1 of this guide, supplemented by Regulatory Positions 1.1.2 and 1.1.3 of this guide.
| | I. .2 Materials for pressure-retaining compo- nents should conform to the requirements of the spec- ifications for materials listed in Section 1I of the ASME Boiler and Pressure Vessel Code.- except that malleable, wrought. or cast iron materials and plastic pipe should not be used. Materials should be compat- ible with the chemical. physical. and radioactive en- vironment of specific applications. Manufacturers' |
| | material certificates of compliance with material specifications. such as those contained in the codes referenced in Table I . may he provided in lieu of cer- tified material test reports. |
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| 1.1.2 Materials for pressure-retaining components, excluding HVAC duct and fire protection piping, should conform to the requirements of the specifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code,5 except that malleable, wrought, or cast iron materials and plastic pipe should not be used. Materials should be compatible with the chemical, physical, and radioactive environment of specific applications during normal conditions and anticipated operational occurrences. Manufacturers' material certificates of compliance with material specifications such as those contained in the codes referenced in the materials column of Table 1 may be provided in lieu of certified material test reports. | | 1.1 .3 Foundations and walls of structures that house the liquid radwaste system should be designed to the seismic criteria described in regulatory position |
| | 5 of this guide to a height sufficient to contain the maximum liquid inventory expected to be in the building. |
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| 1.1.3 Foundations and walls of structures that house the liquid radwaste system should be designed to the natural phenomena and internal and external man-induced hazards criteria described in Regulatory Position 6 of this guide to a height sufficient to contain the maximum liquid inventory expected to be in the building.
| | I. I.4 Equipment and components used to col- lect. process, and store liquid radioactive waste need not be designed to the seismic criteria given in regu- latory position 5 of this guide. |
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| 5 Copies may be obtained from the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th Street, New York, NY 10017.
| | 1.2 All tanks located outside reactor containment and containing radioactive materials in liquids should be designed to prevent uncontrolled releases of radioactive materials due to spillage (in buildings or from outdoor tanks). The following design features should be included for tanks that may contain radioactive materials: |
| | 1.2.1 All tanks inside and outside the plant. in- cluding the condensate storage tanks, should have provisions to monitor liquid levels.. Potential over- flow conditions should actuate alarms both locally and in the control room. |
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| 1.143-4 | | 1.2.2 All tank overflows and drains and sample lines should be routed to the liquid radwaste treat- ment system.- |
| | 1.2.3 Indoor tanks should have curbs or elevated thresholds with floor drains routed to the liquid rad- waste treatment system.3 |
| | 1.2.4 The design should include provisions to prevent leakage from entering unmonitored systems and ductwork in the area. |
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| 1.2 SSCs Outside Containment that Contain Radioactive Liquids All SSCs located outside the reactor containment that contain radioactive materials in liquid form should be classified as described in Regulatory Position 5 and designed in accordance with Regulatory Position 6. In addition, any such component should be designed to prevent uncontrolled releases of radioactive materials caused by spillage in buildings or from outdoor components. The following design features should be included for such components and should meet the criteria contained in Sections 5.2, 5.3, and 5.4 of ANSI/ANS 55.1-1992.
| | 2 Copies may he obtained from the American Society of Mechanical Engineers. United Engineering Center. 345 East 47th Street. New York. New York 10017. |
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| 1.2.1. All tanks inside and outside the plant, including the condensate storage tanks, should have provisions to monitor liquid levels. Designated high-liquid-level conditions should actuate alarms both locally and in the control room.
| | Retention by an intermediate sump or drain tank designed for handling radioactive materials and having provisions for muting 1o the liquid radwaste system is acceptable. |
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| 1.2.2. All radwaste tanks, overflows, drains, and sample lines should be routed to the liquid radwaste treatment system. Retention by an intermediate sump or drain tank that is designed for handling radioactive materials and that has provisions for routing to the liquid radwaste system is acceptable. | | 0 |
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| 1.2.3. Indoor radwaste tanks should have curbs or elevated thresholds with floor drains routed to the liquid radwaste treatment system. Retention by an intermediate sump or drain tank that is designed for handling radioactive materials and that has provisions for routing to the liquid radwaste system is acceptable. | | 1.2.5 Outdoor tanks should have a dike or reten- tion pond capable of preventing runoff in the event of |
| | 'a tank overflow and should have provisions for sam- pling collected liquids and routing them to the liquid radwaste treatment system. |
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| 1.2.4. The design should include provisions to prevent leakage from entering unmonitored and nonradioactive systems and ductwork in the area.
| | 2. Gaseous Radwasie Systems |
| | 2. I The gaseous radwaste treatment system4 should meet the following criteria: |
| | 2. I.I The systemns should he designed and tested to requirements set forth in the codes and standards listed in Table I supplemented by the provisions noted in 2.1.2 and in regulatory position 4 of this guide. |
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| 1.2.5. Outdoor tanks should have a dike or retention pond capable of preventing runoff in the event of a tank overflow and should have provisions for sampling collected liquids and routing them to the liquid radwaste treatment system. | | 2.1.2 Materials for pressure-retaining compo- nents should conform to the requirements of the spec- ifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code 2 except that malleable. wrought.. or cast iron materials and plastic pipe should not be used. Materials should he compat- ible with the chemical, physical, and radioactive en- vironment of specific applications. Manufacturers material certificates of compliance with material specifications, such ais those contained in the codes referenced in Table I, may be provided in lieu of cer- tified materials test reports, |
| | 2.1.3 Those portions of the gaseous radwaste treatment system that are intended to store or delay the release of gaseous radioactive waste. including portions of structures housing these systems. should be designed to the seismic design criteria given in regulatory position 5 of this guide. For the systems that normally operate at pressures above 1.5 atmos- pheres (absolute). these criteria should apply to isola- tion valves, equipment. interconnecting piping. and components located between the upstream and downstream valves used to isolate these components from the rest of the system (e.g.. waste gas storage tanks in the PWR) and to the building housing this equipment. For systems that operate near ambient pressure and retain gases on charcoal adsorbers. these criteria should apply to the tank support elements (e.g.. charcoal delay tanks in a BWR) and the build- ing housing the tanks. |
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| 2. GASEOUS RADWASTE SYSTEMS
| | 3. Solid Radwaste System |
| For a BWR, the gaseous radwaste system includes the system provided for treatment of normal offgas releases from the main condenser vacuum system beginning at the point of discharge from the condenser air removal equipment; for a PWR the gaseous radwaste system includes the system provided for the treatment of gases stripped from the primary coolant.
| | 3.1 The solid radwaste system consists of slurry waste collection and settling tanks, spent resin stor- age tanks, phase separators, and components and subsystems used to solidify radwastes prior to offsite shipment. The solid radwaste handling and treatment system should meet the following criteria: |
| | " For a RWR this includes the system provided for treatment of normal offgas releases from the main condenser vacuum sys- tem beginning at the point of discharge from the condenser air removal equipment; for a PWR this includes the system provided for the treatment of gases stripped front the primary coolant. |
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| 2.1 The SSCs of the gaseous radwaste treatment system should be designed and tested to requirements set forth in the codes and standards listed in Table 1 supplemented by Regulatory Positions 2.2 and 2.3 of this guide.
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| 2.2 Materials for pressure-retaining components, excluding HVAC duct and fire protection piping, should conform to the requirements of the specifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code, except that malleable, wrought, or cast iron materials and plastic pipe should not be used. Materials should be compatible with the
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| chemical, physical, and radioactive environment of specific applications during normal conditions and anticipated operational occurrences. If the potential for an explosive mixture of hydrogen and oxygen exists, adequate provisions should be made to preclude buildup of explosive mixtures, or the system should be designed to withstand the effects of an explosion. Manufacturers' material certificates of compliance with material specifications such as those contained in the codes referenced in the materials column of Table 1 may be provided in lieu of certified materials test reports.
| | 1 The system should be designed and tested t) the requirements set forth in the codes and stand- ards listed in Table I supplemented by the provisions noted in 3.1.2 and in regulatory position 4 of this guide. |
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| 2.3 The portions of the gaseous radwaste treatment system that are intended to store or delay the release of gaseous radioactive waste, including portions of structures housing these systems, should be classified as described in Regulatory Position 5 and designed in accordance with Regulatory Position 6. | | 3.1 .2 Materials for pressure-retaining conmpo- nents should conform to the requirements of the spec- ifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code- except that tualleable. wrought. or cast iron materials and plastic pipe should not be used. Materials shoulh be compalt- ible with the chemical. physical, and radioactive en- vironment of specific applications. NIanufacturers" |
| | material certificates of cotmpliarnce with material specifications. such ats those contained in the codes referenced in Table I , tna. vbe provided in lieu of cer- tified materials test reports. |
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| 3. SOLID RADWASTE SYSTEM | | 3.1.3 Foundations and adjacent walls of struc- lures that house the solid radwaste system should be designed to the seismic criteria given in regulatory position 5 of this guide to a heighl sufficient to conl- cain the maximum liquid inventory expected to be in the building. |
| The solid radwaste system consists of slurry waste collection and settling tanks, spent resin storage tanks, phase separators, and components and subsystems used to dewater or solidify radwastes prior to storage or offsite shipment.
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| 3.1 The SSCs of the solid radwaste treatment system should be designed and tested to the requirements set forth in the codes and standards listed in Table 1 supplemented by Regulatory Positions 3.2 and 3.3 of this guide. | | 3.1.4 Equipment and components used to col- lect, process. or store solid radwasles need not be de- signed to seismic criteria given in regulatory position |
| | 5 of this guide. |
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| 3.2 Materials for pressure-retaining components, excluding HVAC duct and fire protection piping, should conform to the requirements of the specifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code, except that malleable, wrought, or cast iron materials and plastic pipe should not be used. Materials should be compatible with the chemical, physical, and radioactive environment of specific applications during normal conditions and anticipated operational occurrences. Manufacturers' material certificates of compliance with material specifications such as those contained in the codes referenced in the materials column of Table 1 may be provided in lieu of certified materials test reports.
| | 4. Additional Design, Construction, and Testing Criteria In addition to the requirements inherent in the codes and standards listed in Table I, the following criteria, as a minimum, should be implemented for components and systems considered in this guide: |
| | 4.1 The quality assurance provisions described in regulatory position 6 of this guide should be applied. |
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| 3.3 Foundations and adjacent walls of structures that house the solid radwaste system should be designed to the natural phenomena and internal and external man-induced hazards guidance given in Regulatory Position 6 of this guide to a height sufficient to contain the maximum liquid inventory expected to be in the building.
| | 4.2 Process piping systems include the first root valve on sample and instrument lines. Pressure- retaining components of process syslems should use welded construction to the maximum practicable ex- tent. Flanged joints or suitable rapid disconnect fit- tings should be used only where maintenance or op- erational requirements clearly indicate that such con- struction is preferable. Screwed connections in which threads provide the only seal should not be used ex- cept for instrumentation connections where welded connections are not suitable. Process lines should not be less than 3/4 inch (nominal I.D.). Screwed con- nections backed up by seil welding, mechanical joints, or socket welding may be used on lines 3/4 inch or larger but less than 2-1/2 inches (nominal I.D.). For lines 2-1/2 inches and above, pipe welds should be of the butt-joint type. Nonconsumable backing rings should not be used in lines carrying re- sins or other particulate material. All welding con- stituting the pressure boundary of pressure-retaining |
| | 0 |
| | 1.143-3 |
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| 3.4 Equipment and components used to collect, process, or store solid radwastes need not be designed to the seismic guidance in Regulatory Position 6 of this guide. | | components should be performed in accordance with ASME Boiler and Pressure Vessel Code Section IX.-2 |
| | 4.3 Piping systems should be hydrostatically tested in (heir entirety except at atmospheric tank connec- tions where no isolation valves exist. Pressure testing should be performed on as large a portion of the in- place systems as practicable. Testing of piping sys- tems should be. performed in accordance with appli- cable ASME or ANSI codes, but in no case at less than 75 psig. The test pressure should be held for a minimum of 30 minutes with no leakage indicated. |
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| 4. ADDITIONAL DESIGN, CONSTRUCTION, AND TESTING | | 4.4 Testing provisions should be incorporated to enable periodic evaluation of the operability and re- quired functional performance of active components of the system. |
| In addition to the requirements inherent in the codes and standards listed in Table 1, the following, as a minimum, should be applicable to SSCs listed in Regulatory Position 6 of this guide.
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| 1.143-6
| | 5. |
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| 4.1 Radioactive waste management SSCs should be designed to control leakage and facilitate access, operation, inspection, testing, and maintenance in order to maintain radiation exposures to operating and maintenance personnel as low as is reasonably achievable. Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable, provides guidance that is acceptable to the NRC staff on this subject.
| | Seismic Design for Radwaste Management Systems and Structures Housing Radwaste Management Systems |
| | 5.1 Gaseous Radwaste Management Systems' |
| | 5. 1.1 For the evaluation of the gaseous radwaste system described in regulatory position 2.1.3. a simplified seismic analysis procedure to determine seismic loads may be used. The simplified procedure consists of considering the system as a single- degree-of-freedom system and picking up a seismic response value from applicable flnor response spectra, after the fundamental frequct-c. of the sys- tem. is determined. The floor response spectra should be obtained analytically (regulatory position 5.2) |
| | from the application of the Regulatory Guide 1.60 de- sign response spectra normalized to the maximum ground acceleration for the operating basis earth- quake (OBE), as established in the application, at the foundation of the building housing the gaseous rad- waste system. More detailed guidance can be found in Regulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components." |
| | 5. 1.2 The allowable stresses to be used for steel system support elements should be those given in |
| | .Specification for the Design, Fabrication and Erec- tion of Structural Steel fot Buildings," adopted in February 1969.' The one-third allowable stress in- crease provisions f6r combinations involving earth- quake loads, indicated in Section 1.5.6 of the specifi- cation* should be included. For design of concrete structures, use of ACI 349-761 as endorsed in Regu- latory Guide 1.142, "Safety-Related Concrete Struc- tures for Nuclear Power Plants (Other Than Reactor Vessels and Containments),'" is acceptable. |
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| 4.2 The quality assurance provisions described in Regulatory Position 7 of this guide should be applied.
| | 5 For those systems that require seismic capabilities, as indi- cated in regulatory position 2.1.3. |
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| 4.3 Pressure-retaining components of process systems should use welded construction to the maximum practicable extent. Process systems include the first root valve on sample and instrument lines. Flanged joints or suitable rapid-disconnect fittings should be used only where maintenance or operational requirements clearly indicate such construction is preferable. Screwed connections in which threads provide the only seal should not be used except for instrumentation and cast pump body drain and vent connections where welded connections are not suitable.
| | ' Copies may be obtained from the American Institute of Steel Construction, Inc., 101 Park Avenue, New York, New York t017. |
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| Process lines should not be less than 3/4 inch (nominal). Screwed connections backed up by seal welding, mechanical joints, or socket welding may be used on lines 3/4 inches or larger but less than 2-1/2 inches. For lines 2-1/2 inches and above, pipe welds should be of the butt-joint type.
| | 5.1.3 The construction and inspection require- ments for the support elements should comply with those stipulated in AISC or ACI Codes as appro- priate. |
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| Nonconsumable backing rings should not be used in lines carrying resins or other particulate material. All welding constituting the pressure boundary of pressure-retaining components should be performed in accordance with ASME Boiler and Pressure Vessel Code Section IX.
| | 5.2 Buildings Flousing Radwaste Systems |
| | 5.2.1 Input motion at the foundation of the building housing the radwaste systems should be de- fined. This motion should bedefined by normalizing the Regulator)y Guide 1.60 spectra to the maximum ground acceleration selected for the plant OBE. A |
| | simplified analysis should be performed to determine appropriate seismic loads and floor response spectra pertinent to the location of the system, i.e., an analy- sis of the building by a several-degrees-of-freedom mathematical model and the use of an approximate method to generate the floor response spectra for radwaste systems and the seismic loads for the build- ings. No time history analysis is required. |
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| 4.4 Piping systems should be hydrostatically tested in their entirety except (1) at atmospheric tanks where no isolation valves exist, (2) when such testing would damage equipment, and (3) when such testing could seriously interfere with other system or component testing. For (2) and (3), pneumatic testing should be performed. Pressure testing should be performed on as large a portion of the in-place systems as practicable. Testing of piping systems should be performed in accordance with applicable ASME or ANSI codes listed in Table 1.
| | 5.2.2 The simplified method for determining seismic loads for the building consists of (a) calculat- ing the first several modal frequencies and participa- tion factors for the building. (b) determining modal seismic loads using regulatory position 5.2.1 input spectra, and (c) combining modal seismic loads in one of the ways described in Regulatory Guide 1.92. |
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| 4.5 Inspection and testing provisions should be incorporated to enable periodic evaluation of the operability and required functional performance of active components of the system.
| | "Combining Modal Responses and Spatial Compo- nents in Seismic Response Analysis.'" |
| | 5.2.3 With regard to generation of floor re- sponse spectra for radwaste systems, simplified methods that give approximate floor response spectra without need for performing a time history analysis may be used. |
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| 5. CLASSIFICATION OF RADWASTE SYSTEMS FOR DESIG | | 5.2.4 The load factors and load combinations to be used for the building should be those given in ACI 349-76 1 as endorsed in Regulatory Guide 1. 142. |
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| ==N. PURPOSE==
| | The allowable stresses for steel components should be those given in the AISC Manual. (See regulatory position 5.1.2.) |
| S
| | 5.2.5 The construction and inspection require- ments for the building elements should comply with those stipulated in the AISC or ACi Code as appro- priate. |
| There are three safety classes, or classifications, for radwaste management facilities: RW-
| |
| IIa (High Hazard), RW-IIb (Hazardous), and RW-IIc (Non-Safety).1 RW-IIa is the most stringent class and RW-IIc is the least stringent. These classifications were developed primarily for natural phenomena and man-induced hazard design. The radiological release criteria (500 millirem at the unprotected area boundary and 5 rem to facility personnel within the protected boundary) was selected to be consistent with the criteria of 10 CFR Part 20, Standards for Protection Against Radiation. This safety classification is applied to SSCs as follows.
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| 5.1 For a given structure housing radwaste processing systems or components, if the total design basis unmitigated radiological release (considering the maximum inventory) at the | | 5.2.6 The foundation media of structures hous- ing the radwaste systems should be selected and de- signed to prevent liquefaction from the effects of the maximum ground acceleration selected for the plant OBE. |
| 1.143-7
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| boundary of the unprotected area is greater than 500 millirem per year or the maximum unmitigated exposure to site personnel within the protected area is greater than 5 rem per year, the external structures are classified as RW-IIa.
| | 5.3 In lieu of the criteria and procedures defined above, optional shield structures constructed around and supporting the radwaste systems may be erected to protect the radwaste systems from effects of hous- ing structural failure. If this option is adopted, the |
| | ' Copies may be obtained from the American Concrete Insti- tute, P.O. Box 19150, Redford Station. Delroit, Michigan |
| | 48219. |
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| 5.2 For a given structure housing radwaste processing systems or components, if the total design basis unmitigated radiological release (considering the maximum inventory) at the boundary of the unprotected area is less than 500 millirem per year and the maximum unmitigated exposure to site personnel within the protected area is less than 5 rem per year, the external structure is classified as RWE-IIb.
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| | 1.143-4 |
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| 5.3 Any systems or components in a RW-IIa facility (see Regulatory Position 5.1) that store, process, or handle radioactive waste in excess of the A1 quantities given in Appendix A,
| | I.9 procedures described in. regulatory position 5.2 need only be applied to the shield structures while treating the rest of the housing structures as non-seismic Category I. |
| Determination of A1 and A2, to 10 CFR Part 71, Packaging and Transportation of Radioactive Material, are classified as RW-IIa. These systems or components that process radioactive waste in excess of the A2 quantities but less than the A1 quantities given in Appendix A to 10 CFR Part
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| 71 are classified as RW-IIb. All other components are classified as RW-IIc. This classification may be modified for specific radwaste components.
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| 5.4 Any systems or components in a RW-IIb structure (see Regulatory Position 5.2)
| | 6. Quality Assurance for Radwaste Management Systenms Since the impact of these systems on safety is lim- ited, a quality assurance program corresponding to the full extent of Appendix B to 10 CFR Part 50 is not required. However. to ensure that systems will perform their intended function. a quality assurance program sufficient to ensure that all design. construc- tion. and testing provisions are met should be estab- lished and documented. The following quality ass'ur- ance program is acceptable to the NRC staff. It is reprinted by permission of the American Nuclear So- ciety from ANSI N199-1976, "Liquid Radioactive Waste Processing System for Pressurized Water Reactor Plants...s |
| that are used to store or process specified radioactive waste in excess of the A1 quantities given in Appendix A to 10 CFR Part 71 are classified as RW-IIb. All other systems or components are classified as RW-IIc.
| | "4.2.3 Quality Control. The design, procure- ment. fabrication and construction activities shall conform to the quality control provisions of the codes and standards specified herein. In addition, or where not covered by the referenced codes and standards. |
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| The unprotected area boundary mentioned in Regulatory Position 5.1 is shown in Figure 1.
| | the following quality control features shall be estab- lished. |
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| A flowchart of the Safety Classification Process is shown in Figure 2. The classifications discussed in Regulatory Positions 5.1 through 5.4 are not intended to apply to radwaste storage facilities if they do not contain any systems or components that exceed the quantities specified in Regulatory Positions 5.1 through 5.4.
| | "4.2.3.1 System Designer and Procurer |
| | "(I) Design and Procurement Document Control-Design and procurement documents shall be independently verified for cotiformance to the re- quirements of this standard by individual(s) within the design organization who are not the originators of the document. Changes to these documents shall be verified or controlled to maintain conformance to this standard. |
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| 6. NATURAL PHENOMENA AND MAN-INDUCED HAZARDS DESIGN FOR
| | "(2) Control of Purchased Material. Equip- ment and Services-Measures to ensure that suppliers of material, equipment and construction services are capable of supplying these items to the quality speci- fied in the procurement documents shall be estab- lished. This may be done by an evaluation or a sur- vey of the suppliers' products and facilities. |
| RADWASTE MANAGEMENT SYSTEMS AND STRUCTURES
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| 6.1 General Design Criteria Solid, liquid, and gaseous radwaste SSCs described in Regulatory Positions 1, 2, and 3 for natural phenomena and internal and external man-induced hazards should be evaluated as put forth in this position.
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| 6.1.1. The natural phenomena and internal and external man-induced hazards demand definitions are as given in Table 2.
| | "(3) |
| | Instructions shall be provided in pro- curement documents to control the handling, storage, shipping and preservation of material and equipment to prevent damage. deterioration or reduction of cleanness. |
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| 6.1.2. The natural phenomena and internal and external man-induced hazards design load combinations are as given in Table 3.
| | K Copies may he obtained from American Nuclear Society. |
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| 1.143-8
| | 555 North Kensington Avenue. L.a Grange Park, Illinois 60525. |
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| 6.1.3. The natural phenomena and internal and external man-induced hazards should meet capacity criteria in Table 4.
| | "4.2.3.2 System Constructor |
| | -(0I) Inspection. In addition to required code inspections a program for inspection of activities af- fecting quality shall be established and executed by. |
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| 6.1.4. The acceptability evaluation should be based on the requirements of the codes and standards given in Table 1, using the capacity criteria in Table 4.
| | or for. the organization performing the activity to ver- ify conformance with the documented instructions. |
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| 6.2 Buildings Housing Radwaste Systems
| | procedures, and drawings for accomplishing the ac- tivity. This shall include the visual inspection of components prior to installation for confornmance with procurement documents and the visual inspection of items and systems following installation, cleanness and passivation (where applied). |
| 6.2.1 Regardless of its safety classification, the foundation and walls up to the spill height of the building housing the radwaste systems should be designed to the criteria of Tables 1, 2, 3, and 4.
| | "'(2) Inspection. Test and Operating Status. |
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| For classifications RW-IIb and RW-IIc, all SSCs should be designed at least for seismic base shear requirements of the Standard Uniform Building Code(UBC), 1997. The guidance of Volume 2 of the UBC 1997 and American Society of Civil Engineers ASCE 7-95, "Minimum Design Loads for Buildings and Other Structures," should be used as noted in Table 2 of this regulatory guide.
| | Measures should be established to provide for the identification of items which have satisfactorily passed required inspections and tests. |
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| 6.2.2. In lieu of the criteria and procedures referenced in this Regulatory Position 6, optional shield structures constructed around and supporting the radwaste systems may be erected to protect the radwaste systems from the effects of failure of the housing structure. If this option is adopted, Regulatory Position 6.2.1 need only be applied to the shield structures.
| | -(3) Identification and Corrective Action for Items of Nonconformance. Measures should he estab- lishe't #o identify items of nonconformance with re- gard to the requirements of the procuremcntit docu- ments or applicable codes and standards and to iden- tifv the action taken to correct such items. |
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| 7. QUALITY ASSURANCE FOR RADWASTE MANAGEMENT SYSTEMS
| | In Section 4.2.3.2(3). *ilems of nonconf'ormance" |
| Since the impact of these systems on safety is limited, the extent of control required by Appendix B to 10 CFR Part 50 is similarly limited. To ensure that systems will perform their intended functions, a quality assurance program sufficient to ensure that all design, construction, and testing provisions are met should be established and documented. A quality assurance program acceptable to the NRC staff is presented in ANSI/ANS-55.6-1993, "Liquid Radioactive Waste Processing System for Pressurized Water Reactor Plants."
| | should 5e interpreted to include failut.1 |
| Section 4.3, "Quality Assurance," of ANSI/ANS 55.6-1993 provides quality assurance guidance that is acceptable to the NRC staff for the system designer and procurer and for the system constructor. The design, procurement, fabrication, and construction activities should conform to the quality control provisions of the codes and standards specified in Table 1 of this guide. In addition, or when not covered by the referenced codes and standards, sufficient records should be maintained to furnish evidence that quality assurance measures are being implemented.
| | , 111:- afunc- tions, deficiencies, deviations, and defective material and equipment. |
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| The records should include results of reviews and inspections and should be identifiable and retrievable. | | Sufficient records should be maintained to furnish evidence that the measures identified above are being implemented. The records should include results of reviews and inspections and should be identifiable and retrievable. |
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| 1.143-9
| | * |
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| ==D. IMPLEMENTATION== | | ==D. IMPLEMENTATION== |
| The purpose of this section is to provide information to licensees and applicants regarding the NRC staffs plans for using this regulatory guide. | | The purpose of this section is to provide informa- tion to applicants regarding the NRC staff"s plans for using this regulatory guide. |
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| Except in those cases in which the applicant proposes an acceptable alternative method for complying with the specified portions of the NRCs regulations, the method described in this guide reflecting public comments will be used in the evaluation of an applicant's design, construction, installation, and testing of radioactive waste management facilities, and in the evaluation of structures, systems, and components in light-water-cooled nuclear power plants.
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| Current licensees may, at their option, comply with the guidance in this regulatory guide.
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| 1.143-10
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| Table 1 - Codes and Standards for the Design of SSC in Radwaste Facilities1 Component Design and Construction Materials Welding Inspection and Testing Structures - Concrete ACI-318 or ACI 3492,3,4 ACI-318 or ACI 349 ACI-318 or ACI 349 ACI-318 or ACI 349 Structures-Steel (Hot Rolled) AISC-ASD or AISC LFRD or AISC ASTM-A36 AWS-D1.1 AISC Standards and AWS
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| N-690(S327) 2,4 Standards Structures-Steel (Cold Formed) AISI SG-673 ASTM-A500 AWS-D1.3, D9.1 AISC Standards and AWS
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| Standards Piping and Valves ANSI/ASME B31.3 5,6 ASME-Sec. II7 ASME, Sec. IX ANSI/ASME B31.3 Atmospheric Tanks API-650 ASME Sec. II ASME, Sec. IX API-620
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| Tanks (0-15 psig) API-620 ASME Sec. II ASME, Sec. IX API-650
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| Pressure Vessels and Tanks (>15 ASME BPVC Div. 1 or Div. 2 ASME Sec. II ASME, Sec. IX ASME Section VIII, Div. 1 psig) or 2 Pumps API-610; API-674; API-675; ASTM A571- ASME, Sec. IX ASME BPVC Code Section ASME BPVC Section VIII, Div. 1 84(1997) or ASME III, Class 38 or Div. 2 Sec. II
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| Heat Exchangers TEMA STD, 8th Edition; ASME ASTM B359-98 or ASME, Sec. IX ASME Section VIII, Div. 1 BPVC Section VIII Div. 1 or Div. 2 ASME Sec. II or 2 HVAC Systems SMACNA Stds. 6,9 ASTM F856-97 AWS-D1.1, D1.3, SMACNA Stds ASTM C1290-00 D9.1 Conduit and Cable Trays NEMA TC2-1998 ASTM B633-98, AWS-D1.1, D1.3, NEMA TC2-1998, NEMA VE1-1998 A123/A123M-01 D9.1 NEMA VE1-1998 NEMA TC2, VE1 Fire Protection Systems NFPA-136,10 ; NFPA-14 ASTM-A795 AWS-D1.1, D1.3, NFPA-13 D9.1, D10.9 Flexible Hoses and Hose ANSI/ANS-40.37 ANSI/ANS-40.37 ANSI/ANS-40.37 ANSI/ANS-40.37 Connections for MRWP11 Footnotes for Table 1:
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| 1 For a comprehensive lists of codes and standards referenced in Tables 1-4, see Appendix A to this regulatory guide.
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| 2 Applicable to structure enclosing or supporting pressurized gaseous waste or liquid waste systems up to spill height. Also applicable to solid waste facility foundations slab and connected wall or column sections up to a height of 10 feet.
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| 3 Appropriate load combinations and capacity criteria for component designs are specified in Table 3 of this regulatory guide.
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| 4 Class RW-IIa Structures are to use ACI-349 and/or AISC N-690(S327) as applicable.
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| 5 Class RW-IIa and RW-IIb Piping Systems are to be designed as category "M" systems.
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| 6 Classes RW-IIa, RW-IIb, and RW-IIc are discussed in Regulatory Position 5 of this regulatory guide.
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| 7 ASME BPVC Section II required for Pressure Retaining Components.
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| 8 ASME Code Stamp, material traceability, and the quality assurance criteria of ASME BPVC, Section III, Div. 1, Article NCA are not required. Therefore, these components are not classified as ASME Code Class 3.
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| 9 Class RW-IIa and RW-IIb HVAC systems are to use SMACNA "Seismic Restraint Manual Guides for Mechanical Systems."
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| 10 Class RW-IIa and RW-IIb Fire Protection Systems are to be designed to NFPA-13, Section 4-14.4.3.
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| 11 Flexible hoses should only be used in conjunction with Moblie Radwaste Processing Systems (MRWP).
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| Table 2 - Natural Phenomena and Internal/External Man-Induced Hazard Design Criteria for Safety Classification Loading Classification RW-IIa RW-IIb RW-IIc (High Hazard) (Hazardous) (Non-Safety)
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| Earthquake OBE or 1/2 SSE ASCE 7-95, Category III1 ASCE 7-95, Category II1 or UBC 97, Category 22 UBC-97, Category 42 Wind ASCE 7-95, Category III1 ASCE 7-95, Category III1 ASCE 7-95, Category II1 Tornado ANS 2.3 at a Probability of 1 x Not Required Not Required
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| 10-5 /yr or three-fifths of Criteria in Regulatory Guide 1.76, Table 1.
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| Tornado Missile A. 75 lbs, 3 in. nominal Not Required Not Required from SRP Section 3.5 diameter sch. 40 pipe.
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| Maximum velocity 0.4 x max. wind speed horizontal and 0.28 times max. wind speed vertical direction.3 B. Automobile wt. 4000 lbs with frontal area of 20.0 sq.
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| ft. traveling horizontally at
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| 0.2 times maximum wind speed horizontally and 0.14 times maximum wind speed up to a height of 35 ft above grade.4 Flood Regulatory Guide 1.59, one-half ASCE 7-95 ASCE 7-95 of the PMF.5 Precipitation (6) ANS 2.8 at probability of 1 x ASCE 7-95, Category III1 ASCE 7-95, Category II1 (Rain, Snow) 10-3/yr or Regulatory Guide
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| 1.59, one-half precipitation specific for the PMF.5 Accidental Explosion To be evaluated on a case-by- Not Required Not Required Fixed Facility case basis, plant-specific definition.
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| Accidental Explosion See Regulatory Guide 1.91. Not Required Not Required Transportation Vehicle Malevolent Vehicle Regulatory Guide 5.68 or plant- Not Required Not Required Assault specific definition.
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| Small Aircraft Crash Plant-specific definition Not Required Not Required Footnotes for Table 2:
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| 1 ASCE 7-95, Table 1-1. 4 Impact-type missile.
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| 2 UBC-97, Table 16-k. 5 PMF = Probable Maximum Flood.
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| 3 Penetrating-type missile. 4 Resistance to lightening strike should also be included in the design.
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| 1.143-13
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| Table 3 - Design Load Combinations System, Structure, Service Levels SSC Safety Class Component (SSC)
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| RW - IIa RW- IIb RW- IIc External Structures A (Normal) D + L + To D + L + To D + L + To (Concrete, Steel, Component Support Structures1) B (Severe; Upset) D + L + To D + L + Tb D + L + Tb D + L + To + Eo D + L + To + Eo D + L + To + Eo D + L + To + W + R D + L + To + W + R D + L + To + W + R
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| D + L + To + F D + L + To + F D + L + To + F
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| External Conduits and Cable Trays D (Abnormal Extreme; D + L + To + Wt Not required Not required Faulted) D + L + To + Vm D + L + To + Ac D + L + Ta + AD
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| D + L + Ta + A
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| Internal Structures A (Normal) D + L + To D + L + To D + L + To (Concrete, Steel Component Support Structures1) B (Severe, Upset) D + L + Tb D + L + Tb D + L + Tb D + L + To + Eo D + L + To + Eo D + L + To + Eo Internal Conduit and Cable D + L + To + F D + L + To + F D + L + To + F
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| Trays D (Abnormal Extreme; D + L + Ta + AD N/R N/R
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| Faulted) D + L + Ta + A
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| Pressure Retaining A (Normal) Pd + D + Dm Pd + D + DM Pd + D + Dm Components2 (Piping, To To To Valves, Pressure Vessels, Atmosphere, Tanks, 0-15 B (Severe, Upset) Po + D + Dm + Eo Po + D + Dm + Eo Po + D + Dm + Eo psig Tanks, Pumps Heat Po + D + Dm + W + R Po + D + Dm + W + R Po + D + Dm + W + R
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| Exchangers) P + D + Dm + F P + D + Dm + F P + D + Dm + F
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| HVAC Systems Tb Tb Tb Fire Protection Systems D (Abnormal, Extreme P + D + Dm + Wt N/R N/R
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| Faulted) Po + D + Dm + Ym Po + D + Dm + Ac Pa + D + Dm + AD
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| Pa + D + Dm + A
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| Nomenclature:
| |
| D = Dead Loads Wt = Tornado Loads Including Missile Effects L = Live loads Vm = Malevolent Vehicle Assault Loads To = Normal Operating Thermal Expansion Loads Ac = Aircraft Crash Loads Tb = Upset Thermal Expansion Loads AD = Design Basis Accident Loads Ta = Accident Thermal Loads A = Other Accident Loads Eo = OBE or 1/2 SSE Seismic Loads Pd = Design Pressure Eo = Seismic Loads per Table 2 For RW-IIb Components Pb = Maximum Upset Pressure Eo = Seismic Loads per Table 2 For RW-IIc Components Po = Normal Operating Pressure W = Wind Load Pa = Applicable Accident Pressure R = Precipitation Loads (Rain, Snow) DM = Design Mechanical Loads F = Flood Loadings Footnotes:
| |
| 1 Component support structures include supporting elements for piping, tanks, vessels pumps, heat exchangers, conduits, cable trays, HVAC systems, fire protection systems, etc.
| |
| | |
| 2 For most pressure-retaining components, primary and secondary stresses are evaluated separately to separate criteria. The design code of record is the controlling document in the establishment of the primary and secondary stress combination and evaluation methods.
| |
| | |
| 1.143-14
| |
| | |
| Table 4 - SSC Design Capacity Criteria Code or Standard Service Capacity Criteria Level RW-IIa RW-IIb RW-IIc ACI-349 A, B, D Load Factors and Capacity N/A N/A
| |
| Criteria per ACI-349 as modified by Regulatory Guide 1.142 ACI-318 A, B, D Load Factors and capacity criteria Load factors and capacity Load factors and capacity per ACI-349 as modified by criteria per ACI-349 as criteria per ACI-349 as Regulatory Guide 1.142. All modified by Regulatory modified by Guide 1.142.
| |
| | |
| other design per ACI-318 criteria Guide 1.142. All other All other design per ACI-
| |
| design per ACI-318 318 criteria.
| |
| | |
| criteria.
| |
| | |
| AISC-N690 A Capacity criteria Table Q 1.5.7.1 Capacity criteria Table Q Capacity criteria Table Q
| |
| for normal loads. 1.5.7.1 for normal loads. 1.5.7.1 for normal loads B Capacity criteria 1.33 times that Capacity criteria 1.33 Capacity criteria 1.33 for Level A loads times that for Level A times that for Level A
| |
| loads loads D Capacity criteria per Table N/R N/R
| |
| Q.1.5.7.1 for Abnormal Extreme Loads AISC-ASD A Capacity Criteria per Capacity Criteria per Capacity Criteria per Specification for Structural Steel Specification for Specification for Buildings Allowable Stress Structural Steel Structural Steel Design and Plastic Design, Part Buildings Allowable Buildings Allowable
| |
| 5,"chapters A-M
| |
| Stress Design and Stress Design and Plastic Design, Part Plastic Design, Part
| |
| 5,"chapters A-M 5,"chapters A-M
| |
| B Capacity Criteria 1.33 times Capacity Criteria 1.33 Capacity Criteria 1.33 that for Level A loads. times that for level A times that for level A
| |
| loads. loads.
| |
| | |
| D Capacity Criteria per N/R N/R
| |
| Specification for Structural Steel Buildings Allowable Stress Design and Plastic Design, Part 5,"chapters A
| |
| and N
| |
| 1.143-15
| |
| | |
| Table 4 - SSC Design Capacity Criteria (continued)
| |
| Code or Standard Service Capacity Criteria Level RW-IIa RW-IIb RW-IIc AISC LRFD A,B,D Load factors and capacities Not required Not required per LRFD specifications for structural steel buildings.
| |
| | |
| AISI CFSDM A Capacity criteria per Capacity criteria per Capacity criteria per Specification for the Design Specification for the Specification for the of Cold Formed Steel Design of Cold Formed Design of Cold Formed Structural Members Steel Structural Steel Structural Members Members B Capacity criteria 1.33 times Capacity criteria 1.33 Capacity criteria 1.33 Level A times Level A times Level A
| |
| D Capacity criteria 1.6 times Not required Not required Level A
| |
| ANSI/ASME B31.3 A B31.3 Design Load B31.3 Design Load B31.3 Design Load Capacities Capacities Capacities B B31.3 Occasional Load B31.3 Occasional Load B31.3 Occasional Load Capacities Capacities Capacities D 1.8 Times B31.3 Occasional Not required Not required Load Capacities ASME BPVC, A ASME BPVC, Section VIII, ASME BPVC, Section ASME BPVC, Section Section VIII, Div. 1 or Div. 1 or Div. 2 Design VIII, Div. 1 or Div. 2 VIII, Div. 1 or Div. 2 Div. 2 Capacities Design Capacities Design Capacities B Capacity criteria 1.2 Times Capacity criteria 1.2 Capacity criteria 1.2 Level A criteria Times Level A criteria Times Level A criteria D Capacity criteria 1.8 times Not required Not required Level A criteria SMACNA Stds.(1) A SMACNA Design Criteria SMACNA Design SMACNA Design Criteria Criteria B SMACNA Design Criteria SMACNA Design SMACNA Design Criteria Criteria D 1. Duct support members to Not required Not required meet capacity criteria for AISI
| |
| SG-673 or AISC-ASD for Level D Loads.
| |
| | |
| 2. Ducting stresses to be less than the material yield stress and limited to 2/3 critical buckling.
| |
| | |
| NFPA-131 A NFPA Design Criteria NFPA Design Criteria NFPA Design Criteria
| |
| 1.143-16
| |
| | |
| Table 4 - SSC Design Capacity Criteria (continued)
| |
| Code or Standard Service Capacity Criteria Level RW-IIa RW-IIb RW-IIc B NFPA Design Criteria for NFPA Design Criteria NFPA Design Criteria Earthquake and Wind Loads for Earthquake and for Earthquake and Wind Loads Wind Loads D 3. Support members to meet N/R N/R
| |
| capacity criteria for AISI SG-
| |
| 673 or AISC-ASD for Level D Loads
| |
| 4. Piping Stresses to meet the B31.3 Level D Capacity Criteria ANSI/NEMA STDS A ANSI/NEMA Design Criteria ANSI/NEMA Design ANSI/NEMA Design (Cable Trays/Conduit) for Normal Loads Criteria for Normal Criteria for Normal Loads Loads B ANSI/NEMA Design Criteria ANSI/NEMA Design ANSI/NEMA Design for Wind and Seismic Loads Criteria for Wind and Criteria for Wind and Seismic Loads Seismic Loads D 5. Support members to meet Not required Not required capacity criteria for AISI-
| |
| CFSDM or AISC-ASD for Level D Load
| |
| 6. Trays and members to meet the capacity criteria for AISI-CFSDM for Level D
| |
| Loads Pumps A For Design Criteria, API 610, For Design Criteria, For Design Criteria, (API Series Stds) API 674, API 675 API 610, API 674, API API 610, API 674, API
| |
| 675 675 B ASME QME-1 1997 ASME QME-1 1997 ASME QME-1 1997 D ASME QME-1 1997 Not required Not required Tanks A API-620, API-650 API-620, API-650 API-620, API-650
| |
| B Capacity Criteria per ASME- Capacity Criteria per Capacity Criteria per BPVC - Section III, NC-3800, ASME-BPVC - ASME-BPVC - Section NC-3900 for Level B loads. Section III, NC-3800, III, NC-3800, NC-3900
| |
| All other Design per API NC-3900 for Level B for Level B loads. All Criteria. loads. All other other Design per API
| |
| Design per API Criteria.
| |
| | |
| Criteria.
| |
| | |
| 1.143-17
| |
| | |
| Table 4 - SSC Design Capacity Criteria (continued)
| |
| Code or Standard Service Capacity Criteria Level RW-IIa RW-IIb RW-IIc D Capacity Criteria per ASME- Not required Not required BPVC - Section III, NC-3800,
| |
| NC-3900 Level D Loads. All other Design per API Criteria.
| |
| | |
| Footnotes for Table 4:
| |
| 1 For Level A and B Loads, the Design Criteria is primarily a design by rule approach versus a specific analysis criteria.
| |
| | |
| 1.143-18
| |
| | |
| Site Boundary or Boundary of the Owner Controlled (Unprotected) Area Access Control Radwaste Facility Admin Bldg Protected Area Boundary Unprotected (Secured Area) Area Figure 1 - Informational Schematic Describing Protected and Unprotected Areas
| |
| 1.143-19
| |
| | |
| Is the Unmitigated Release
| |
| @ the Protected Area Boundary > 500mrem or YES Unmitigated Exposure NO
| |
| For Site Personnel Inside the Protected Area
| |
| > 5 rem
| |
| ?
| |
| Classify Classify Overall Facility Overall Facility as RW IIa as RW IIb Evaluate SSC Evaluate SSC
| |
| in the Facility in the Facility Does the Subject Does the Subject SSC Contain Radioacitive SSC Contain YES Quantites NO YES Radioacitive NO
| |
| > A1 ? Quantites
| |
| > A1 ?
| |
| Classify SSC Classify SSC
| |
| as RW IIb as RW IIc Classify SSC
| |
| as RW IIa Does the Subject SSC Contain Radioacitive Quantites YES > A2 ?
| |
| NO
| |
| Classify SSC Classify SSC
| |
| as RW IIb as RW IIc Figure 2 - Flowchart of Safety Classification Process Appendix A
| |
| 1.143-20
| |
| | |
| INDUSTRY CODES AND STANDARDS
| |
| American Concrete Institute, ACI-318, "Building Code Requirements for Reinforced Concrete (ACI 318-89, Revised 1999), 1999.
| |
| | |
| American Concrete Institute, ACI-349, Code Requirements for Nuclear Safety Related Concrete Structures," 1997.
| |
| | |
| American Institute of Steel Construction, N690 (S327), Nuclear Facilities, Steel Safety-Related Structures For Design and Fabrication, 1984.
| |
| | |
| American Institute of Steel Construction, Manual of Steel Construction Load and Resistance Factor Design, Volumes I and II, 2nd Edition, 1994.
| |
| | |
| American Institute of Steel Construction, Specifications for Structural Steel Buildings, Manual of Steel Construction, 2nd Edition, 1995.
| |
| | |
| American Institute of Steel Construction, "Specifications for Structural Steel Buildings, Allowable Stress Design and Plastic Design, Manual of Steel Construction, 9th Edition, 1993.
| |
| | |
| American Iron and Steel Institute, SG-673, Specification for the Design of Cold-Formed Steel Structural Members, August 1986 with December 1989 Addendum.
| |
| | |
| American Nuclear Society, Standard for Estimating Tornado and Extreme Wind Characteristics at Nuclear Power Sites, ANSI/ANS 2.3-1983.
| |
| | |
| American Nuclear Society, Determining Design Basis Flooding at Power Reactor Sites, ANSI/ANS 2.8-1992.
| |
| | |
| American Nuclear Society, "Mobile Radioactive Waste Processing Systems," ANSI/ANS 40.37-
| |
| 1993.
| |
| | |
| American Nuclear Society, "Solid Radioactive Waste Processing System for Light-Water- Cooled Reactor Plants," ANSI/ANS-55.1-1992.
| |
| | |
| American Nuclear Society, "Gaseous Radioactive Waste Processing Systems for Light Water Reactor Plants," ANSI/ANS-55.4-1993.
| |
| | |
| American Nuclear Society, "Liquid Radioactive Waste Processing System for Light Water Reactor Plants," ANSI/ANS 55.6-1993.
| |
| | |
| American Petroleum Institute, 610, Centrifugal Pumps for Petroleum, Heavy Duty Chemical, and Gas Industry Services, 1995.
| |
| | |
| 1.143-21
| |
| | |
| American Petroleum Institute, 620, Design and Construction of Large, Welded, Low-Pressure Storage Tanks, 1990.
| |
| | |
| American Petroleum Institute, 650, Welded Steel Tanks for Oil Storage, 1998.
| |
| | |
| American Petroleum Institute, 674, Positive Displacement Pumps-Reciprocating, 1995.
| |
| | |
| American Petroleum Institute, 675, Positive Displacement Pumps-Controlled Volume, 1994.
| |
| | |
| American Society of Civil Engineers, 7-95, "Minimum Design Loads for Buildings and Other Structures," 1995.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section II,
| |
| Material Specification, 1999.
| |
| | |
| American Society of Mechanical Engineers, Boildr and Pressure Vessel Code, Section III,
| |
| Rules for Construction of Nuclear Power Plant Components, Division 1, Subsection ND Class
| |
| 3 Components, July 1998 with July 1999 Addenda.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section VIII,
| |
| Pressure Vessels, 1999.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section VIII,
| |
| Rules for Construction of Pressure Vessel, Division 1, July 1998 with July 1999 Addenda.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section VIII,
| |
| Rules for Construction of Pressure Vessel, Division 2, Alternative Rules, July 1998 with July
| |
| 1999 Addenda.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section IX,
| |
| Welding and Brazing Qualification, 1999.
| |
| | |
| American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, B31.3, Process Piping, 1999.
| |
| | |
| American Society of Mechanical Engineers, QME-1-1997, Qualification of Active Mechanical Equipment Used in Nuclear Power Plants, December 31, 1997.
| |
| | |
| American Society for Testing & Materials, A36-00, Standard Specification for Carbon Structural Steel, 2000.
| |
| | |
| American Society for Testing & Materials, A123/A123M-01, Standard Specification for Zinc (Hot-Dip Galvanized) Coatings on Iron and Steel Products, 2001.
| |
| | |
| 1.143-22
| |
| | |
| American Society for Testing & Materials, A500-99, Standard Specification for Cold-Formed Welded and Seamless Carbon Steel Structural Tubing in Rounds and Shapes, 1999.
| |
| | |
| American Society for Testing & Materials, A571-84 (1997, Standard Specification for Austenitic Ductile Iron Castings for Pressure-Containing Parts Suitable for Low-Temperature Service, 1997.
| |
| | |
| American Society for Testing & Materials, A795-97, Standard Specification for Black and Hot- Dipped Zinc-Coated Welded and Seamless Steel Pipe for Fire Protection Use, 1997.
| |
| | |
| American Society for Testing & Materials, B359-98, Standard Specification for Copper and Copper-Alloy Seamless Condenser and Heat Exchanger Tubes With Integral Fins, 1998.
| |
| | |
| American Society for Testing & Materials, B633-98, Standard Specification for Electrodeposited Coatings of Zinc on Iron and Steel, 1998.
| |
| | |
| American Society for Testing & Materials, C1290-00, Standard Specification for Flexible Fibrous Glass Blanket Insulation Used to Externally Insulate HVAC Ducts, 2000.
| |
| | |
| American Society for Testing & Materials, F856-97, Standard Practice for Mechanical Symbols, Shipboard Heating, Ventilation, and Air Conditioning (HVAC), 1997.
| |
| | |
| American Welding Society, D1.1, Structural Welding Code-Steel, 17th Edition, 2000.
| |
| | |
| American Welding Society, D1.3, Structural Welding Code-Sheet Steel, 1998.
| |
| | |
| American Welding Society, D9.1, Sheet Metal Welding Code, 1990.
| |
| | |
| American Welding Society, D10.9, Specification for Qualification of Welding Procedures and Welders for Piping and Tubing, 1980.
| |
| | |
| International Conference of Building Officials, Uniform Building Code, 1997.
| |
| | |
| National Electrical Manufacturers Association, Publication Number TC2, Electrical Polyvinyl Chloride(PVC) Tubing and Conduit, 1998.
| |
| | |
| National Electrical Manufacturers Association, Publication Number VE1, Metal Cable Tray Systems, 1996.
| |
| | |
| National Fire Protection Association, NFPA 13, Installation of Sprinkler Systems, 1999.
| |
| | |
| National Fire Protection Association, NFPA 14, Standard for the Installation of Standpipe Fire Protection, Private Hydrant, and Hose Systems, 2000.
| |
| | |
| Sheet Metal and Air Conditioners Contractor National Association, Seismic Restraint Manual Guides for Mechanical Systems, 2nd Edition, 1998.
| |
| | |
| 1.143-23
| |
| | |
| Standard Uniform Building Code, International Conference of Building Officials, 1997.
| |
| | |
| Tubular Exchanger Manufacturers Association, Standards of the Tubular Exchanger Manufacturers Association, Eighth Edition, 2000.
| |
| | |
| The Codes and Standards are available from:
| |
| American Concrete Institute (ACI), Box 19150, Redford Station, Detroit, MI 48219.
| |
| | |
| American Institute of Steel Construction (AISC), One E. Wacker Drive, Suite 3100, Chicago, IL
| |
| 60601-2001.
| |
| | |
| American Iron and Steel Institute (AISI),1101 17th Street, NW, Washington, DC 20036.
| |
| | |
| American Nuclear Society (ANS), 555 N. Kensington Avenue, La Grange Park, IL 60525.
| |
| | |
| American Petroleum Institute (API), 1220 L Street, NW, Washington, DC 20005.
| |
| | |
| American Society of Mechanical Engineers (ASME), 345 East 47th Street, New York, NY 10017.
| |
| | |
| American Society for Testing & Materials (ASTM), 100 Barr Harbor Drive, West Conshohocken, PA 19428-2959.
| |
| | |
| American Welding Society (AWS), 550 NW LeJeune Road, Miami, FL 33126.
| |
| | |
| International Conference of Building Officials, 5360 Workman Mill Road, Whittier, CA 90601-
| |
| 2798. (www.icbo.org)
| |
| National Electrical Manufacturers Association (NEMA), 1300 N. 17th Street, Rosslyn, VA 22209.
| |
| | |
| National Fire Protection Association (NFPA), Inc., Battery March Park, Quincy, MA 02269.
| |
| | |
| Sheet Metal and Air Conditioners Contractor National Association (SMACNA), 4201 Lafayette Center Drive, Chantilly, VA 20153-1230.
| |
| | |
| Tubular Exchanger Manufacturers Association (TEMA), 25 N. Broadway, Tarrytown, NY 10591.
| |
| | |
| 1.143-24
| |
| | |
| REGULATORY ANALYSIS
| |
| | |
| ===1. STATEMENT OF PROBLEM===
| |
| Revision 1 of Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, was issued in October 1979. This guide provided design guidance acceptable to the NRC
| |
| staff related to seismic and quality group classification and quality assurance provisions for radioactive waste management structures, systems, and components. Further, it describes provisions for controlling releases of liquids containing radioactive materials, e.g., spills or tank overflows, from all plant systems outside reactor containment. Regulatory Guide 1.143 encompassed the design of buildings, structures, systems, and components and referred to several design and construction codes and standards, such as American National Standards Institute (ANSI)
| |
| N197-1976, ANSI N199-1976, American Nuclear Society (ANS) ANS 55.1-1979, ANS 55.4-1979, American Concrete Institute ACI-318-1977, and American Institute of Steel Construction AISC-
| |
| 1969.
| |
| | |
| These references are now obsolete or have been superseded by newer ANSI and ANS
| |
| radioactive waste facility design standards. ANS has since issued ANS-55.1-92, ANS-55.4-93, and ANS-55.6-93, which are the industry consensus standards currently applicable to the overall design of radioactive waste facilities. In addition, several other referenced codes such as Building Code and Commentary, ACI-318-77; or Specification for the Design, Fabrication and Erection of Structural Steel for Buildings, AISC-1969, have been updated and modified since Revision 1 of Regulatory Guide 1.143 was issued. Also, there has been increased understanding of, and corresponding changes in relation to, radiation exposure and monitoring and quality assurance needs for the design and construction of radioactive waste facilities and the associated systems, structures, and components.
| |
| | |
| The Operating Basis Earthquake (OBE), as was used in Revision 1 of Regulatory Guide
| |
| 1.143 as the design basis, creates further difficulties. In 1997, the NRC staff revised 10 CFR
| |
| 100.23 and added Appendix S to 10 CFR Part 50 that essentially state that, if the review level earthquake (OBE) is defined as less than 1/3 of the safe-shutdown earthquake (SSE), no explicit design analysis for the OBE level earthquake will be required. In other words, the revised criteria have effectively eliminated the OBE as a design basis seismic event. In recent staff licensing actions, the Standard (Advanced) Reactor Designs used only a SSE event as the design basis, consistent with the methodology in the recent revision of 10 CFR 100.23 and the addition of Appendix S to 10 CFR Part 50. Thus, Revision 1 of Regulatory Guide 1.143 was almost not usable for standard reactor designs.
| |
| | |
| The staff maintains that recommendations based on the latest editions of the design and construction Standards and Codes mentioned above and references to current quality assurance standards and NRC regulations provide a means to achieve better evaluation of radioactive waste management systems, structures, and components installed in light water-cooled nuclear power plants.
| |
| | |
| 1.143-25
| |
| | |
| ===2. OBJECTIVE===
| |
| The objective of the regulatory action is to update NRC guidance on the design, construction, and quality assurance of radioactive waste management systems, structures, and components installed in light-water-cooled nuclear power plants.
| |
| | |
| 3. ALTERNATIVES AND CONSEQUENCES OF PROPOSED ACTION
| |
| 3.1 Alternative 1 - Do Not Revise Regulatory Guide 1.143 If Regulatory Guide 1.143 were not revised, licensees would continue to rely on the current version of Regulatory Guide 1.143 with references from the late 1960s and mid-1970s. The staff acknowledges that many licensees who are presently involved in the design of radioactive waste management systems, structures, and components installed in light-water-cooled nuclear power plants, as a matter of practice, already rely on more recent editions of ANSI and ANS radioactive waste facility design standards and ACI and AISC codes.
| |
| | |
| 3.2 Alternative 2 - Update Regulatory Guide 1.143 The NRC staff has identified the following consequences associated with adopting Alternative 2.
| |
| | |
| 3.2.1 Licensees will use the latest consensus standards available, thereby improving design, evaluation, and quality assurance of radioactive waste management systems, structures, and components. The staff views the latest standards as improved because they incorporate the latest technology and knowledge on the subject.
| |
| | |
| 3.2.2 Regulatory efficiency will be improved by reducing uncertainty as to what is acceptable and by encouraging consistency in the design, evaluation, and quality assurance of radioactive waste management systems, structures, and components. The benefits to both the NRC
| |
| and industry will be to the extent this occurs. An updated regulatory guide would facilitate NRC
| |
| review because licensee submittals should be more predictable and consistent analytically.
| |
|
| |
|
| Similarly, licensees adherence to the latest consensus standards should benefit licensees by reducing the likelihood for follow-up questions and possible revisions to licensees plans.
| | This guide reflects current NRC staff practice. |
|
| |
|
| 3.2.3 An updated regulatory guide could result in cost savings for both the NRC and industry. From the NRCs perspective, relative to the baseline, NRC will incur one-time incremental costs to develop the regulatory guide for public comment and to finalize the regulatory guide. However, the NRC should also realize cost savings associated with the review of licensee submittals. In the staffs view, the continuous and on-going cost savings associated with these reviews should more than off-set this one-time cost.
| | Therefore, except in those cases in which the appli- cant proposes an acceptable alternative method for complying with specified portions of the Commis- sion's regulations. the method described herein is being and will continue to be used in evaluation of submittals in connection with applications for operat- ing licenses, construction permits, or amendments thereto until this guide is revised as a result of suggestions from the public or additional staff review. |
|
| |
|
| On balance, it is expected that industry would realize a net savings, as their one-time incremental cost to review and comment on a revised regulatory guide would be more than
| | 0 |
| 1.143-26
| | 1.143-5 |
|
| |
|
| compensated for by the efficiencies (e.g., reduced follow-up questions and revisions) associated with each licensee submittal.
| | TABLE 1 EQUIPMENT CODES |
| | FIQUI PMENT |
| | CODES |
| | Detsigni and |
| | 'a hr ica Iio)n |
| | "'elder Qualification M~aterials' |
| | and Procedures Inspection and Testing Prcssutie Vesisel |
| | 0-I ni VSphri 'Ianks Ili ping.- and V'alve.- |
| | ASMI- Code Sc*tion VIII\\ I)v I |
| | ;-\\SME' |
| | Section III. Clam, 3. kr API 6501. or A.W\\VWA 1I) -I W.. |
| | AS.ME' Scctioh III. |
|
| |
|
| 3.2.4 The use of industry consensus standards that are already being used by licensees would enhance the continued use of the guidance contained in ANS-55.1-92, ANS-55.4-93, and ANS-55.6-93, thereby avoiding costs related to a new agency-prepared standard. This approach would also comply with the Commissions directive that standards developed by consensus bodies be utilized per Public Law 104-113, National Technology and Transfer Act of 1995. | | Class 3. or API 620' |
| | .\\S.IE Code Section VIII, Div. I |
| | AN.I TE.MA |
| | ANSI t131.1 ASME Code Section II |
| | ASME Code: |
| | Section II |
| | ASMI" code: |
| | Section II |
| | ASMNE Code Section II |
| | AS'IM an ASME Code Section II |
| | AS.IE Code Section II or Manufacturers Standard ASME Code Section IX |
| | ASNIIE Co'de Section IX |
| | A.SMIE Code Section IX |
| | ASNII: Code Section IX |
| | AS.ME Co'de Section IX |
| | ASNIF Code Section IX |
| | (a.s required) |
| | ASNME Code SectionVIll. D)iv. I |
| | AS\\IE Code, sect ion Ill. |
|
| |
|
| ===4. CONCLUSION===
| | CiGas 3. or API 650. |
| Based on this regulatory analysis, it is recommended that the NRC revise Regulatory Guide
| |
| 1.143. The staff concludes that the proposed action will reduce unnecessary burden on both the NRC and its licensees, and it will result in an improved process for the design, evaluation, and quality assurance of radioactive waste management systems, structures, and components.
| |
|
| |
|
| Furthermore, the staff sees no adverse effects associated with a revision to Regulatory Guide 1.143.
| | or AWWA D-I1002 ASME Codc' Section Ill. |
|
| |
|
| BACKFIT ANALYSIS
| | Cklss 3. or API 6202 ASNSMI F Code Seclton VIII. Div. I |
| The regulatory guide does not require a backfit analysis as described in 10 CFR 50.109(c)
| | ANSI 1331. 1 ASMEN |
| because it does not impose a new or amended provision in the NRCs rules or a regulatory staff position interpreting the NRCs rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require the modification or addition to systems, structures, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant may select a preferred method for achieving compliance with a license or the rules or the orders of the Commission as described in 10
| | Section Ill Clans 3: or Hydraulic Insltimue Stantjdard |
| CFR 50.109(a)(7). This regulatory guide provides an opportunity to use industry-developed standards if that is the method preferred by the licensee or applicant.
| | 0 |
| | I Fih..i .Io fLIII11:,:Cd 11II%l~ic o |
| | l nk ma -% ! |
| | '.CL,01 Ill iWird.inic., |
| | %kith aippr.priaiv. :,IIcIls tif SL'ctjim; |
| | 10. ... SNst Bo~iler and t'res.urc |
| | .\\S.\\tt (*odc. lo~r app. c~zit..ti:, uliei loaitiie miilltme uit. |
|
| |
|
| 1.143-27}} | | rI'i |
| | ~r~'' |
| | )ii OCRt'r |
| | 1 r |
| | ihr~mid A.t\\lhIIF |
| | ciI r' |
| | .mitd iit! |
| | M31061 |
| | ~ |
| | c |
| | 'fC |
| | .Alll i . |
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Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, for Comment| ML13350A262 |
| Person / Time |
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| Issue date: |
07/31/1978 |
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| From: |
Office of Nuclear Regulatory Research |
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| To: |
|
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| References |
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| RG-1.143 |
| Download: ML13350A262 (6) |
|
U.S. NUCLEAR REGULATORY COMMISSION
July 1978
00
REGULATORY GUIDE.
OFFICE OF STANDARDS DEVELOPMENT
REGULATORY GUIDE 1.143 DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT SYSTEMS, STRUCTURES, AND
COMPONENTS INSTALLED IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
A. INTRODUCTION
Paragraph (a) of k 50.34, "Contents of applica- tions; technical information,"
of 10 CFR Part 50,
"Domestic Licensing of Production and Utilization Facilities," requires that each application for a con- struction permit include a preliminary safety analysis report. Part of the information required is a prelimi- nary design of-the facility, including among other things the principal design criteria for the facility.
Appendix A, "General Design Criteria for Nuclear Power Plants,"
to 10 CFR Part 50 establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants.
Criterion I, "Quality Standards and Records," of A'ppendix A requires that structures, systems, and components important to safety be designed, fabri- cated, erected, and tested to quality standards com- mensurate with the importance to safety of the s function to be performed. Criterion 2, "Design ses for Protection Against Natural Phenome,:
" o
-
pendix A requires, among other thin
"t st st P
tures. systems. and components impo"
t to fety be designed to withstand the effect
,ural phenomena such as earthquakes without lo- f capa- bility to perform their ,4,t* functions and that the design bases for these truc~rs, systems, and com- ponents reflect th hi1p'
ce "the safety functions to be perform t. c¶-eri060, "Control of Releases of RadioapiivlMate~ils to the Environment-," of Appen A rAA
irs ts at the nuclear power unit de- sign in ma'
control suitably the release of radioactiv naterials in gaseous and liquid effluents and to handi# radioactive solid waste produced during normal reactor operation, including anticipated opera- tional occurrences.
.This guide furnishes design guidance acceptable to the NRC staff relating to seismic and quality group classification and quality assurance provisions for radioactive waste management systems, structures, and components. Further, it describes provisions for controlling releases of liquids containing radioactive materials, e.g., spills or tank overflows, from all plant systems outside reactor containalog.
B. DISCU ,$Sll3I
One aspect of nuclear. 0o l
-cration is [he control and manage g, geous. and solid radioactive wast g' crated as a byprod- uct of nuclear
.
,ose of this guide is to provide i ma d
eria that will provide rea- sonable St components and structures use the.
ive waste management and steam en'
b down systems are designed. con- c dh,'stalled, and tested on a level commensu- he need to protect the health and safety of pu lic and plant operating personnel. It sets forth m nimum staff recommendations and is not intended to prohibit the implementation of more rigorous de- sign considerations, codes, standards, or quality as- surance measures.
Working Group ANS-55, Radioactive Waste Sys- tems, of Subcommittee ANS-50, Nuclear Power Plant System Engineering, of the American Nuclear Society Standards Committee has developed stand- ards that establish requirements and provide recom- mendations for the design, construction, and per- formance of BWR (ANSI N197-1976) and PWR
(ANSI N199-1976) liquid radioactive waste process- ing systems. Standards for gaseous and solid radioac- tive waste processing systems are being developed.
I Radioactive waste, as used in this guide. means those liq- uids. gases, or solids containing radioactive materials thatl hy design or operating practice will be processed prior itt final dis- position.
USNRC REGULATORY GUIDES
Comment% should be sent to the Secretay of the Co.mmnin. US. Nutii t-q.,
Regulatory Guides are issued to describe and make available to khe Public method, latory Commission. Waihington, D.C.
20555. Attention Docketnrg. a
"'.,r .
accepttable to the NRC nuatl of implementing specific parts of the Commission's Branch.
regultations, to delineate techniques used by the &talf in evaluating specific problems The guide are issued in the followmng ten broad dniori$
or postulated accidents, or to Provide guidance to applicants. Regulatory Guide:
are not slubstitutes for regul-tlions, and compliance with them is not required.
1. Power Reactors
6. Producft Methods and solutions different from those set Out in the guides will be acce't.
2. Research and Test Reactors
7. Trarrsaotlation able if they provide a basis for the findingst requiite to the issuance at continuance
3. Fuels and Materials Facilities
8. Occupational Health of a permit at license by the Commission.
4. Environmental and Siting
9. Antitrusl Re.,ew
5.Vleterits alnd Piasti Pratect'On tO. Generat Comments and suggestionr% for irrlprovements in these guides are encouraged at all times, and guides will be" revised, as appropriate, to accommodate comments and Reu its for single Copies 01 issued guides iwhtch may be lsi'tuwceet orfoJ r tiltQ'
to reflect nevs informatlon or experience.
Howsever, comments on this guide.if mini on an automatic dtobution list tfo sinrlle cur.e, of tilt.'re quide% in r.l N
,eceived within about t rc, m.nths alter its issuance, wil t be particularty useful in divisionst should be made in writing to the U.S. Nuclear Regtulatory Crniin,,sJw.
evalual ing the need lor an early revision.
Washlington, D.C.
?M055.
Attentiton D..ectot. Division osf Dociument Ciill,
These standards provide more detailed guidance with regard to the specific requirements of the radioactive waste processing system than are presented in this guide. It is expected that these standards will be en- dorsed separately to be used in conjunction with this guide or that reference to applicable sections ma-- be used in future revisions to this guide.
For the purpose of this guide. the radwaste systems are considered to begin at the interface valve(s) in each line from other systems provided for collecting wastes that may contain radioactive materials and to include related instrumentation and control systems.
The radwaste system terminates at the point of con- trolled discharge to the environment. at the point of recycle back to storage for reuse in the reactor, or at the point of storage of packaged solid wastes prior to shipment offsite to a licensed burial ground. The steam generator blowdown system begins at, but does not include, the outermost containment isolation valve on the blowdown line. It terminates at the point of controlled discharge to the environment, at the point of interface with other liquid systems, or at the point of recycle back to the secondary systems. Ex- cept as noted, this guide does not apply to the reactor water cleanup system, the condensate cleanup sys- tent. the chemical and volume control system, the reactor coolant and auxiliary building equipment drain tanks, the sumps and floor drains provided for.
collecting liquid wastes, the boron recovery system.
equipment used to prepare solid waste solidification agents, the building ventilation systems (heating.
ventilating, and air conditioning). or the chemical fume hood exhaust systems.
The design and construction of radioactive waste management and steam generator blowdown systems should provide assurance that radiation exposures to operating personnel and to the general public are as low as is reasonably achievable. One aspect of this consideration is ensuring that these systems are de- signed to quality standards that enhance system relia- bility. operability, and availability. In development of this design guidance. the NRC staff has considered designs and concepts submitted in license applica- tions and resulting operating system histories. It has also been guided by industry practices and the cost of design features, taking into account the potential im- pact on the health and safety of operating personnel and the general public.
C. REGULATORY POSITION
1. Systems Handling Radioactive Materials in Liquids I.1 The liquid radwaste treatment system includ- ing the steam generator blowdown system downstream of the second containment isolation valve should meet the following criteria:
I.1. 1 These systems should he designed and tested to requirements set forth in the codes and standards listed in Table I supplemented by the provi- sions in I. 1.2 and in regulatory position 4 of this guide.
I. .2 Materials for pressure-retaining compo- nents should conform to the requirements of the spec- ifications for materials listed in Section 1I of the ASME Boiler and Pressure Vessel Code.- except that malleable, wrought. or cast iron materials and plastic pipe should not be used. Materials should be compat- ible with the chemical. physical. and radioactive en- vironment of specific applications. Manufacturers'
material certificates of compliance with material specifications. such as those contained in the codes referenced in Table I . may he provided in lieu of cer- tified material test reports.
1.1 .3 Foundations and walls of structures that house the liquid radwaste system should be designed to the seismic criteria described in regulatory position
5 of this guide to a height sufficient to contain the maximum liquid inventory expected to be in the building.
I. I.4 Equipment and components used to col- lect. process, and store liquid radioactive waste need not be designed to the seismic criteria given in regu- latory position 5 of this guide.
1.2 All tanks located outside reactor containment and containing radioactive materials in liquids should be designed to prevent uncontrolled releases of radioactive materials due to spillage (in buildings or from outdoor tanks). The following design features should be included for tanks that may contain radioactive materials:
1.2.1 All tanks inside and outside the plant. in- cluding the condensate storage tanks, should have provisions to monitor liquid levels.. Potential over- flow conditions should actuate alarms both locally and in the control room.
1.2.2 All tank overflows and drains and sample lines should be routed to the liquid radwaste treat- ment system.-
1.2.3 Indoor tanks should have curbs or elevated thresholds with floor drains routed to the liquid rad- waste treatment system.3
1.2.4 The design should include provisions to prevent leakage from entering unmonitored systems and ductwork in the area.
2 Copies may he obtained from the American Society of Mechanical Engineers. United Engineering Center. 345 East 47th Street. New York. New York 10017.
Retention by an intermediate sump or drain tank designed for handling radioactive materials and having provisions for muting 1o the liquid radwaste system is acceptable.
0
I
4
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1.143-2
1.2.5 Outdoor tanks should have a dike or reten- tion pond capable of preventing runoff in the event of
'a tank overflow and should have provisions for sam- pling collected liquids and routing them to the liquid radwaste treatment system.
2. Gaseous Radwasie Systems
2. I The gaseous radwaste treatment system4 should meet the following criteria:
2. I.I The systemns should he designed and tested to requirements set forth in the codes and standards listed in Table I supplemented by the provisions noted in 2.1.2 and in regulatory position 4 of this guide.
2.1.2 Materials for pressure-retaining compo- nents should conform to the requirements of the spec- ifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code 2 except that malleable. wrought.. or cast iron materials and plastic pipe should not be used. Materials should he compat- ible with the chemical, physical, and radioactive en- vironment of specific applications. Manufacturers material certificates of compliance with material specifications, such ais those contained in the codes referenced in Table I, may be provided in lieu of cer- tified materials test reports,
2.1.3 Those portions of the gaseous radwaste treatment system that are intended to store or delay the release of gaseous radioactive waste. including portions of structures housing these systems. should be designed to the seismic design criteria given in regulatory position 5 of this guide. For the systems that normally operate at pressures above 1.5 atmos- pheres (absolute). these criteria should apply to isola- tion valves, equipment. interconnecting piping. and components located between the upstream and downstream valves used to isolate these components from the rest of the system (e.g.. waste gas storage tanks in the PWR) and to the building housing this equipment. For systems that operate near ambient pressure and retain gases on charcoal adsorbers. these criteria should apply to the tank support elements (e.g.. charcoal delay tanks in a BWR) and the build- ing housing the tanks.
3. Solid Radwaste System
3.1 The solid radwaste system consists of slurry waste collection and settling tanks, spent resin stor- age tanks, phase separators, and components and subsystems used to solidify radwastes prior to offsite shipment. The solid radwaste handling and treatment system should meet the following criteria:
" For a RWR this includes the system provided for treatment of normal offgas releases from the main condenser vacuum sys- tem beginning at the point of discharge from the condenser air removal equipment; for a PWR this includes the system provided for the treatment of gases stripped front the primary coolant.
t.
I3.
1 The system should be designed and tested t) the requirements set forth in the codes and stand- ards listed in Table I supplemented by the provisions noted in 3.1.2 and in regulatory position 4 of this guide.
3.1 .2 Materials for pressure-retaining conmpo- nents should conform to the requirements of the spec- ifications for materials listed in Section II of the ASME Boiler and Pressure Vessel Code- except that tualleable. wrought. or cast iron materials and plastic pipe should not be used. Materials shoulh be compalt- ible with the chemical. physical, and radioactive en- vironment of specific applications. NIanufacturers"
material certificates of cotmpliarnce with material specifications. such ats those contained in the codes referenced in Table I , tna. vbe provided in lieu of cer- tified materials test reports.
3.1.3 Foundations and adjacent walls of struc- lures that house the solid radwaste system should be designed to the seismic criteria given in regulatory position 5 of this guide to a heighl sufficient to conl- cain the maximum liquid inventory expected to be in the building.
3.1.4 Equipment and components used to col- lect, process. or store solid radwasles need not be de- signed to seismic criteria given in regulatory position
5 of this guide.
4. Additional Design, Construction, and Testing Criteria In addition to the requirements inherent in the codes and standards listed in Table I, the following criteria, as a minimum, should be implemented for components and systems considered in this guide:
4.1 The quality assurance provisions described in regulatory position 6 of this guide should be applied.
4.2 Process piping systems include the first root valve on sample and instrument lines. Pressure- retaining components of process syslems should use welded construction to the maximum practicable ex- tent. Flanged joints or suitable rapid disconnect fit- tings should be used only where maintenance or op- erational requirements clearly indicate that such con- struction is preferable. Screwed connections in which threads provide the only seal should not be used ex- cept for instrumentation connections where welded connections are not suitable. Process lines should not be less than 3/4 inch (nominal I.D.). Screwed con- nections backed up by seil welding, mechanical joints, or socket welding may be used on lines 3/4 inch or larger but less than 2-1/2 inches (nominal I.D.). For lines 2-1/2 inches and above, pipe welds should be of the butt-joint type. Nonconsumable backing rings should not be used in lines carrying re- sins or other particulate material. All welding con- stituting the pressure boundary of pressure-retaining
0
1.143-3
components should be performed in accordance with ASME Boiler and Pressure Vessel Code Section IX.-2
4.3 Piping systems should be hydrostatically tested in (heir entirety except at atmospheric tank connec- tions where no isolation valves exist. Pressure testing should be performed on as large a portion of the in- place systems as practicable. Testing of piping sys- tems should be. performed in accordance with appli- cable ASME or ANSI codes, but in no case at less than 75 psig. The test pressure should be held for a minimum of 30 minutes with no leakage indicated.
4.4 Testing provisions should be incorporated to enable periodic evaluation of the operability and re- quired functional performance of active components of the system.
5.
Seismic Design for Radwaste Management Systems and Structures Housing Radwaste Management Systems
5.1 Gaseous Radwaste Management Systems'
5. 1.1 For the evaluation of the gaseous radwaste system described in regulatory position 2.1.3. a simplified seismic analysis procedure to determine seismic loads may be used. The simplified procedure consists of considering the system as a single- degree-of-freedom system and picking up a seismic response value from applicable flnor response spectra, after the fundamental frequct-c. of the sys- tem. is determined. The floor response spectra should be obtained analytically (regulatory position 5.2)
from the application of the Regulatory Guide 1.60 de- sign response spectra normalized to the maximum ground acceleration for the operating basis earth- quake (OBE), as established in the application, at the foundation of the building housing the gaseous rad- waste system. More detailed guidance can be found in Regulatory Guide 1.122, "Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components."
5. 1.2 The allowable stresses to be used for steel system support elements should be those given in
.Specification for the Design, Fabrication and Erec- tion of Structural Steel fot Buildings," adopted in February 1969.' The one-third allowable stress in- crease provisions f6r combinations involving earth- quake loads, indicated in Section 1.5.6 of the specifi- cation* should be included. For design of concrete structures, use of ACI 349-761 as endorsed in Regu- latory Guide 1.142, "Safety-Related Concrete Struc- tures for Nuclear Power Plants (Other Than Reactor Vessels and Containments),'" is acceptable.
5 For those systems that require seismic capabilities, as indi- cated in regulatory position 2.1.3.
' Copies may be obtained from the American Institute of Steel Construction, Inc., 101 Park Avenue, New York, New York t017.
5.1.3 The construction and inspection require- ments for the support elements should comply with those stipulated in AISC or ACI Codes as appro- priate.
5.2 Buildings Flousing Radwaste Systems
5.2.1 Input motion at the foundation of the building housing the radwaste systems should be de- fined. This motion should bedefined by normalizing the Regulator)y Guide 1.60 spectra to the maximum ground acceleration selected for the plant OBE. A
simplified analysis should be performed to determine appropriate seismic loads and floor response spectra pertinent to the location of the system, i.e., an analy- sis of the building by a several-degrees-of-freedom mathematical model and the use of an approximate method to generate the floor response spectra for radwaste systems and the seismic loads for the build- ings. No time history analysis is required.
5.2.2 The simplified method for determining seismic loads for the building consists of (a) calculat- ing the first several modal frequencies and participa- tion factors for the building. (b) determining modal seismic loads using regulatory position 5.2.1 input spectra, and (c) combining modal seismic loads in one of the ways described in Regulatory Guide 1.92.
"Combining Modal Responses and Spatial Compo- nents in Seismic Response Analysis.'"
5.2.3 With regard to generation of floor re- sponse spectra for radwaste systems, simplified methods that give approximate floor response spectra without need for performing a time history analysis may be used.
5.2.4 The load factors and load combinations to be used for the building should be those given in ACI 349-76 1 as endorsed in Regulatory Guide 1. 142.
The allowable stresses for steel components should be those given in the AISC Manual. (See regulatory position 5.1.2.)
5.2.5 The construction and inspection require- ments for the building elements should comply with those stipulated in the AISC or ACi Code as appro- priate.
5.2.6 The foundation media of structures hous- ing the radwaste systems should be selected and de- signed to prevent liquefaction from the effects of the maximum ground acceleration selected for the plant OBE.
5.3 In lieu of the criteria and procedures defined above, optional shield structures constructed around and supporting the radwaste systems may be erected to protect the radwaste systems from effects of hous- ing structural failure. If this option is adopted, the
' Copies may be obtained from the American Concrete Insti- tute, P.O. Box 19150, Redford Station. Delroit, Michigan
48219.
.1
0
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1.143-4
I.9 procedures described in. regulatory position 5.2 need only be applied to the shield structures while treating the rest of the housing structures as non-seismic Category I.
6. Quality Assurance for Radwaste Management Systenms Since the impact of these systems on safety is lim- ited, a quality assurance program corresponding to the full extent of Appendix B to 10 CFR Part 50 is not required. However. to ensure that systems will perform their intended function. a quality assurance program sufficient to ensure that all design. construc- tion. and testing provisions are met should be estab- lished and documented. The following quality ass'ur- ance program is acceptable to the NRC staff. It is reprinted by permission of the American Nuclear So- ciety from ANSI N199-1976, "Liquid Radioactive Waste Processing System for Pressurized Water Reactor Plants...s
"4.2.3 Quality Control. The design, procure- ment. fabrication and construction activities shall conform to the quality control provisions of the codes and standards specified herein. In addition, or where not covered by the referenced codes and standards.
the following quality control features shall be estab- lished.
"4.2.3.1 System Designer and Procurer
"(I) Design and Procurement Document Control-Design and procurement documents shall be independently verified for cotiformance to the re- quirements of this standard by individual(s) within the design organization who are not the originators of the document. Changes to these documents shall be verified or controlled to maintain conformance to this standard.
"(2) Control of Purchased Material. Equip- ment and Services-Measures to ensure that suppliers of material, equipment and construction services are capable of supplying these items to the quality speci- fied in the procurement documents shall be estab- lished. This may be done by an evaluation or a sur- vey of the suppliers' products and facilities.
"(3)
Instructions shall be provided in pro- curement documents to control the handling, storage, shipping and preservation of material and equipment to prevent damage. deterioration or reduction of cleanness.
K Copies may he obtained from American Nuclear Society.
555 North Kensington Avenue. L.a Grange Park, Illinois 60525.
"4.2.3.2 System Constructor
-(0I) Inspection. In addition to required code inspections a program for inspection of activities af- fecting quality shall be established and executed by.
or for. the organization performing the activity to ver- ify conformance with the documented instructions.
procedures, and drawings for accomplishing the ac- tivity. This shall include the visual inspection of components prior to installation for confornmance with procurement documents and the visual inspection of items and systems following installation, cleanness and passivation (where applied).
"'(2) Inspection. Test and Operating Status.
Measures should be established to provide for the identification of items which have satisfactorily passed required inspections and tests.
-(3) Identification and Corrective Action for Items of Nonconformance. Measures should he estab- lishe't #o identify items of nonconformance with re- gard to the requirements of the procuremcntit docu- ments or applicable codes and standards and to iden- tifv the action taken to correct such items.
In Section 4.2.3.2(3). *ilems of nonconf'ormance"
should 5e interpreted to include failut.1
, 111:- afunc- tions, deficiencies, deviations, and defective material and equipment.
Sufficient records should be maintained to furnish evidence that the measures identified above are being implemented. The records should include results of reviews and inspections and should be identifiable and retrievable.
D. IMPLEMENTATION
The purpose of this section is to provide informa- tion to applicants regarding the NRC staff"s plans for using this regulatory guide.
This guide reflects current NRC staff practice.
Therefore, except in those cases in which the appli- cant proposes an acceptable alternative method for complying with specified portions of the Commis- sion's regulations. the method described herein is being and will continue to be used in evaluation of submittals in connection with applications for operat- ing licenses, construction permits, or amendments thereto until this guide is revised as a result of suggestions from the public or additional staff review.
0
1.143-5
TABLE 1 EQUIPMENT CODES
FIQUI PMENT
CODES
Detsigni and
'a hr ica Iio)n
"'elder Qualification M~aterials'
and Procedures Inspection and Testing Prcssutie Vesisel
0-I ni VSphri 'Ianks Ili ping.- and V'alve.-
ASMI- Code Sc*tion VIII\\ I)v I
- -\\SME'
Section III. Clam, 3. kr API 6501. or A.W\\VWA 1I) -I W..
AS.ME' Scctioh III.
Class 3. or API 620'
.\\S.IE Code Section VIII, Div. I
AN.I TE.MA
ANSI t131.1 ASME Code Section II
ASME Code:
Section II
ASMI" code:
Section II
ASMNE Code Section II
AS'IM an ASME Code Section II
AS.IE Code Section II or Manufacturers Standard ASME Code Section IX
ASNIIE Co'de Section IX
A.SMIE Code Section IX
ASNII: Code Section IX
AS.ME Co'de Section IX
ASNIF Code Section IX
(a.s required)
ASNME Code SectionVIll. D)iv. I
AS\\IE Code, sect ion Ill.
CiGas 3. or API 650.
or AWWA D-I1002 ASME Codc' Section Ill.
Cklss 3. or API 6202 ASNSMI F Code Seclton VIII. Div. I
ANSI 1331. 1 ASMEN
Section Ill Clans 3: or Hydraulic Insltimue Stantjdard
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