Regulatory Guide 1.143
| ML013100305 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/2001 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| Graves H (301)415-5880 | |
| References | |
| DG-1100 RG-1.143, Revision 2 | |
| Download: ML013100305 (28) | |
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.
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U.S. NUCLEAR REGULATORY COMMISSION Revision 2 November 2001 REGULATORY
GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.143 (Draft was issued as DG-1100)
DESIGN GUIDANCE FOR RADIOACTIVE WASTE MANAGEMENT
SYSTEMS, STRUCTURES, AND COMPONENTS INSTALLED IN
LIGHT-WATER-COOLED NUCLEAR POWER PLANTS
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.
In 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," § 50.34,
"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.
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
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
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.
1.143-2 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.
Criterion 2, "Design Bases for Protection Against Natural Phenomena," of Appendix A
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.
Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, of 10 CFR Part 50
states general design requirements for the implementation of General Design Criterion 2.
Criterion 60, "Control of Releases of Radioactive Materials to the Environment, of Appendix A
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.
This regulatory guide is being revised to provide design guidance acceptable to the NRC
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.
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.
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.
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.
3Copies may be obtained from the American Nuclear Society, 555 North Kensington Avenue, La Grange Park, Illinois 60525.
4Copies of sections of NUREG 0800 are available by email to DISTRIBUTION@NRC.GOV or by fax to (301)415-2289.
1.143-3
B. DISCUSSION
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.
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
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,
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.
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.
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.
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
5 Copies may be obtained from the American Society of Mechanical Engineers, United Engineering Center, 345 East 47th Street, New York, NY 10017.
1.143-4 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.
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.
C. REGULATORY POSITION
1.
SYSTEMS HANDLING RADIOACTIVE MATERIALS IN LIQUIDS
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.
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.
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 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.
1.143-5
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.
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.
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.
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.
1.2.4. The design should include provisions to prevent leakage from entering unmonitored and nonradioactive systems and ductwork in the area.
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.
GASEOUS RADWASTE SYSTEMS
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.
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.
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
1.143-6 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.
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.
SOLID RADWASTE SYSTEM
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.
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.
3.2 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.
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.
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.
4.
ADDITIONAL DESIGN, CONSTRUCTION, AND TESTING
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.
1.143-7
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.
4.2 The quality assurance provisions described in Regulatory Position 7 of this guide should be applied.
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.
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.
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.
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.
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.
5.
CLASSIFICATION OF RADWASTE SYSTEMS FOR DESIG
N. PURPOSE
S
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.
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
1.143-8 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.
5.2 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.
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,
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 71 are classified as RW-IIb. All other components are classified as RW-IIc. This classification may be modified for specific radwaste components.
5.4 Any systems or components in a RW-IIb structure (see Regulatory Position 5.2)
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.
The unprotected area boundary mentioned in Regulatory Position 5.1 is shown in Figure 1.
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.
6.
NATURAL PHENOMENA AND MAN-INDUCED HAZARDS DESIGN FOR
RADWASTE MANAGEMENT SYSTEMS AND STRUCTURES
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.
6.1.1. The natural phenomena and internal and external man-induced hazards demand definitions are as given in Table 2.
6.1.2. The natural phenomena and internal and external man-induced hazards design load combinations are as given in Table 3.
1.143-9
6.1.3. The natural phenomena and internal and external man-induced hazards should meet capacity criteria in Table 4.
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.
6.2 Buildings Housing Radwaste Systems
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.
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.
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.
7.
QUALITY ASSURANCE FOR RADWASTE MANAGEMENT SYSTEMS
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."
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.
The records should include results of reviews and inspections and should be identifiable and retrievable.
1.143-10
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.
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.
Current licensees may, at their option, comply with the guidance in this regulatory guide.
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)
N-690(S327) 2,4 ASTM-A36 AWS-D1.1 AISC Standards and AWS
Standards Structures-Steel (Cold Formed)
AISI SG-673 ASTM-A500
AWS-D1.3, D9.1 AISC Standards and AWS
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
Tanks (0-15 psig)
API-620
ASME Sec. II
ASME, Sec. IX
API-650
Pressure Vessels and Tanks (>15 psig)
ASME BPVC Div. 1 or Div. 2 ASME Sec. II
ASME, Sec. IX
ASME Section VIII, Div. 1 or 2 Pumps API-610; API-674; API-675;
ASME BPVC Section VIII, Div. 1 or Div. 2 ASTM A571-
84(1997) or ASME
Sec. II
ASME, Sec. IX
ASME BPVC Code Section III, Class 38 Heat Exchangers TEMA STD, 8th Edition; ASME
BPVC Section VIII Div. 1 or Div. 2 ASTM B359-98 or ASME Sec. II
ASME, Sec. IX
ASME Section VIII, Div. 1 or 2 HVAC Systems SMACNA Stds. 6,9 ASTM F856-97 ASTM C1290-00
AWS-D1.1, D1.3, D9.1 SMACNA Stds Conduit and Cable Trays NEMA TC2-1998 NEMA VE1-1998 ASTM B633-98, A123/A123M-01 NEMA TC2, VE1 AWS-D1.1, D1.3, D9.1 NEMA TC2-1998, NEMA VE1-1998 Fire Protection Systems NFPA-136,10 ; NFPA-14 ASTM-A795 AWS-D1.1, D1.3, D9.1, D10.9 NFPA-13 Flexible Hoses and Hose Connections for MRWP11 ANSI/ANS-40.37 ANSI/ANS-40.37 ANSI/ANS-40.37 ANSI/ANS-40.37 Footnotes for Table 1:
1 For a comprehensive lists of codes and standards referenced in Tables 1-4, see Appendix A to this regulatory guide.
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.
3 Appropriate load combinations and capacity criteria for component designs are specified in Table 3 of this regulatory guide.
4 Class RW-IIa Structures are to use ACI-349 and/or AISC N-690(S327) as applicable.
5 Class RW-IIa and RW-IIb Piping Systems are to be designed as category "M" systems.
6 Classes RW-IIa, RW-IIb, and RW-IIc are discussed in Regulatory Position 5 of this regulatory guide.
7 ASME BPVC Section II required for Pressure Retaining Components.
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.
9 Class RW-IIa and RW-IIb HVAC systems are to use SMACNA "Seismic Restraint Manual Guides for Mechanical Systems."
10 Class RW-IIa and RW-IIb Fire Protection Systems are to be designed to NFPA-13, Section 4-14.4.3.
11 Flexible hoses should only be used in conjunction with Moblie Radwaste Processing Systems (MRWP).
1.143-13 Table 2 - Natural Phenomena and Internal/External Man-Induced Hazard Design Criteria for Safety Classification Loading Classification RW-IIa (High Hazard)
RW-IIb (Hazardous)
RW-IIc (Non-Safety)
Earthquake OBE or 1/2 SSE
ASCE 7-95, Category III1 or UBC 97, Category 22 ASCE 7-95, Category II1 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
10-5 /yr or three-fifths of Criteria in Regulatory Guide 1.76, Table 1.
Not Required Not Required Tornado Missile from SRP Section 3.5 A. 75 lbs, 3 in. nominal diameter sch. 40 pipe.
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.
ft. traveling horizontally at
0.2 times maximum wind speed horizontally and 0.14 times maximum wind speed up to a height of 35 ft above grade.4 Not Required Not Required Flood Regulatory Guide 1.59, one-half of the PMF.5 ASCE 7-95 ASCE 7-95 Precipitation (6)
(Rain, Snow)
ANS 2.8 at probability of 1 x
10-3/yr or Regulatory Guide 1.59, one-half precipitation specific for the PMF.5 ASCE 7-95, Category III1 ASCE 7-95, Category II1 Accidental Explosion Fixed Facility To be evaluated on a case-by- case basis, plant-specific definition.
Not Required Not Required Accidental Explosion Transportation Vehicle See Regulatory Guide 1.91.
Not Required Not Required Malevolent Vehicle Assault Regulatory Guide 5.68 or plant- specific definition.
Not Required Not Required Small Aircraft Crash Plant-specific definition Not Required Not Required Footnotes for Table 2:
1 ASCE 7-95, Table 1-1.
4 Impact-type missile.
2 UBC-97, Table 16-k.
5 PMF = Probable Maximum Flood.
3 Penetrating-type missile.
4 Resistance to lightening strike should also be included in the design.
1.143-14 Table 3 - Design Load Combinations System, Structure, Component (SSC)
Service Levels SSC Safety Class
RW - IIa RW- IIb RW- IIc External Structures (Concrete, Steel, Component Support Structures1)
External Conduits and Cable Trays A (Normal)
D + L + To D + L + To D + L + To B (Severe; Upset)
D + L + To D + L + To + Eo D + L + To + W + R
D + L + To + F
D + L + Tb D + L + To + Eo D + L + To + W + R
D + L + To + F
D + L + Tb D + L + To + Eo D + L + To + W + R
D + L + To + F
D (Abnormal Extreme;
Faulted)
D + L + To + Wt D + L + To + Vm D + L + To + Ac D + L + Ta + AD
D + L + Ta + A
Not required Not required Internal Structures (Concrete, Steel Component Support Structures1)
Internal Conduit and Cable Trays A (Normal)
D + L + To D + L + To D + L + To B (Severe, Upset)
D + L + Tb D + L + To + Eo D + L + To + F
D + L + Tb D + L + To + Eo D + L + To + F
D + L + Tb D + L + To + Eo D + L + To + F
D (Abnormal Extreme;
Faulted)
D + L + Ta + AD
D + L + Ta + A
N/R
N/R
Pressure Retaining Components2 (Piping, Valves, Pressure Vessels, Atmosphere, Tanks, 0-15 psig Tanks, Pumps Heat Exchangers)
HVAC Systems Fire Protection Systems A (Normal)
Pd + D + Dm To Pd + D + DM
To Pd + D + Dm To B (Severe, Upset)
Po + D + Dm + Eo Po + D + Dm + W + R
P + D + Dm + F
Tb Po + D + Dm + Eo Po + D + Dm + W + R
P + D + Dm + F
Tb Po + D + Dm + Eo Po + D + Dm + W + R
P + D + Dm + F
Tb D (Abnormal, Extreme Faulted)
P + D + Dm + Wt Po + D + Dm + Ym Po + D + Dm + Ac Pa + D + Dm + AD
Pa + D + Dm + A
N/R
N/R
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)
=
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.
1.143-15 Table 4 - SSC Design Capacity Criteria Code or Standard Service Level Capacity Criteria RW-IIa RW-IIb RW-IIc ACI-349 A, B, D
Load Factors and Capacity Criteria per ACI-349 as modified by Regulatory Guide 1.142 N/A
N/A
ACI-318 A, B, D
Load Factors and capacity criteria per ACI-349 as modified by Regulatory Guide 1.142. All other design per ACI-318 criteria Load factors and capacity criteria per ACI-349 as modified by Regulatory Guide 1.142. All other design per ACI-318 criteria.
Load factors and capacity criteria per ACI-349 as modified by Guide 1.142.
All other design per ACI-
318 criteria.
AISC-N690
A
Capacity criteria Table Q 1.5.7.1 for normal loads.
Capacity criteria Table Q
1.5.7.1 for normal loads.
Capacity criteria Table Q
1.5.7.1 for normal loads B
Capacity criteria 1.33 times that for Level A loads Capacity criteria 1.33 times that for Level A
loads Capacity criteria 1.33 times that for Level A
loads D
Capacity criteria per Table Q.1.5.7.1 for Abnormal Extreme Loads N/R
N/R
AISC-ASD
A
Capacity Criteria per Specification for Structural Steel Buildings Allowable Stress Design and Plastic Design, Part
5,"chapters A-M
Capacity Criteria per Specification for Structural Steel Buildings Allowable Stress Design and Plastic Design, Part
5,"chapters A-M
Capacity Criteria per Specification for Structural Steel Buildings Allowable Stress Design and Plastic Design, Part
5,"chapters A-M
B
Capacity Criteria 1.33 times that for Level A loads.
Capacity Criteria 1.33 times that for level A
loads.
Capacity Criteria 1.33 times that for level A
loads.
D
Capacity Criteria per Specification for Structural Steel Buildings Allowable Stress Design and Plastic Design, Part 5,"chapters A
and N
N/R
N/R
1.143-16 Table 4 - SSC Design Capacity Criteria (continued)
Code or Standard Service Level Capacity Criteria RW-IIa RW-IIb RW-IIc AISC LRFD
A,B,D
Load factors and capacities per LRFD specifications for structural steel buildings.
Not required Not required AISI CFSDM
A
Capacity criteria per Specification for the Design of Cold Formed Steel Structural Members Capacity criteria per Specification for the Design of Cold Formed Steel Structural Members Capacity criteria per Specification for the Design of Cold Formed Steel Structural Members B
Capacity criteria 1.33 times Level A
Capacity criteria 1.33 times Level A
Capacity criteria 1.33 times Level A
D
Capacity criteria 1.6 times Level A
Not required Not required ANSI/ASME B31.3 A
B31.3 Design Load Capacities B31.3 Design Load Capacities B31.3 Design Load Capacities B
B31.3 Occasional Load Capacities B31.3 Occasional Load Capacities B31.3 Occasional Load Capacities D
1.8 Times B31.3 Occasional Load Capacities Not required Not required ASME BPVC,
Section VIII, Div. 1 or Div. 2 A
Div. 1 or Div. 2 Design Capacities ASME BPVC,Section VIII, Div. 1 or Div. 2 Design Capacities ASME BPVC,Section VIII, Div. 1 or Div. 2 Design Capacities B
Capacity criteria 1.2 Times Level A criteria Capacity criteria 1.2 Times Level A criteria Capacity criteria 1.2 Times Level A criteria D
Capacity criteria 1.8 times Level A criteria Not required Not required SMACNA Stds.(1)
A
SMACNA Design Criteria SMACNA Design Criteria SMACNA Design Criteria B
SMACNA Design Criteria SMACNA Design Criteria SMACNA Design Criteria D
1. Duct support members to 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.
Not required Not required NFPA-131 A
NFPA Design Criteria NFPA Design Criteria NFPA Design Criteria
Table 4 - SSC Design Capacity Criteria (continued)
Code or Standard Service Level Capacity Criteria RW-IIa RW-IIb RW-IIc
1.143-17 B
NFPA Design Criteria for Earthquake and Wind Loads NFPA Design Criteria for Earthquake and Wind Loads NFPA Design Criteria for Earthquake and Wind Loads D
3. Support members to meet 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 N/R
N/R
ANSI/NEMA STDS
(Cable Trays/Conduit)
A
ANSI/NEMA Design Criteria for Normal Loads ANSI/NEMA Design Criteria for Normal Loads ANSI/NEMA Design Criteria for Normal Loads B
ANSI/NEMA Design Criteria for Wind and Seismic Loads ANSI/NEMA Design Criteria for Wind and Seismic Loads ANSI/NEMA Design Criteria for Wind and Seismic Loads D
5. Support members to meet 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 Not required Not required Pumps (API Series Stds)
A
For Design Criteria, API 610,
API 674, API 675 For Design Criteria, API 610, API 674, API
675 For Design Criteria, API 610, API 674, API
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-
BPVC - Section III, NC-3800,
NC-3900 for Level B loads.
All other Design per API
Criteria.
Capacity Criteria per ASME-BPVC -
Section III, NC-3800,
NC-3900 for Level B
loads. All other Design per API
Criteria.
Capacity Criteria per ASME-BPVC - Section III, NC-3800, NC-3900
for Level B loads. All other Design per API
Criteria.
Table 4 - SSC Design Capacity Criteria (continued)
Code or Standard Service Level Capacity Criteria RW-IIa RW-IIb RW-IIc
1.143-18 D
Capacity Criteria per ASME-
BPVC - Section III, NC-3800,
NC-3900 Level D Loads. All other Design per API Criteria.
Not required Not required 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.
1.143-19 Site Boundary or Boundary of the Owner Controlled (Unprotected) Area Access Control
Radwaste Facility Protected Area Boundary Unprotected (Secured Area) Area Figure 1 - Informational Schematic Describing Protected and Unprotected Areas Admin Bldg
1.143-20
Is the Unmitigated Release
@ the Protected Area Boundary > 500mrem or Unmitigated Exposure For Site Personnel Inside the Protected Area
> 5 rem
?
Classify Overall Facility as RW IIa Classify Overall Facility as RW IIb Does the Subject SSC Contain Radioacitive Quantites
> A1 ?
Classify SSC
as RW IIa Does the Subject SSC Contain Radioacitive Quantites
> A2 ?
Classify SSC
as RW IIc Does the Subject SSC Contain Radioacitive Quantites
> A1 ?
Classify SSC
as RW IIc YES
NO
YES
NO
YES
NO
YES
NO
Evaluate SSC
in the Facility Evaluate SSC
in the Facility Figure 2 - Flowchart of Safety Classification Process Appendix A
1.143-21 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-
1993Property "ANSI code" (as page type) with input value "ANSI/ANS 40.37-</br></br>1993" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
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.
1.143-22 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.
1.143-23 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.
1.143-24 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.
1.143-25 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.
1.143-26
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.
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.
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
1.143-27 compensated for by the efficiencies (e.g., reduced follow-up questions and revisions) associated with each licensee submittal.
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.
4.
CONCLUSION
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.
BACKFIT ANALYSIS
The regulatory guide does not require a backfit analysis as described in 10 CFR 50.109(c)
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 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.