ML21006A335

DG 1381 for Rev 0 of RG 1.244 - Control of Heavy Loads for Nuclear Facilities

ML21006A335
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Issue date:	04/01/2021
From:	Steve Jones Office of Nuclear Regulatory Research
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ML21006A313	List: ML21006A335 ML21006A336 ML21006A337 ML21139A118
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DG-1381 RG 1.244
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U.S. NUCLEAR REGULATORY COMMISSION DRAFT REGULATORY GUIDE DG-1381 Proposed new Regulatory Guide 1.244 Issue Date: April 2021 Technical Lead: Steven R. Jones CONTROL OF HEAVY LOADS AT NUCLEAR FACILITIES A. INTRODUCTION Purpose This regulatory guide (RG) describes an approach that is acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) to meet regulatory requirements for control of heavy loads at nuclear facilities, specifically to provide appropriate protection against equipment failure that could result in a heavy load drop. It endorses, with clarifications, the American Society of Mechanical Engineers (ASME) Standard (Std.) NML-1, Rules for the Movement of Loads Using Overhead Handling Equipment in Nuclear Facilities (Ref. 1) and ASME Std. NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) (Ref. 2). It also endorses, with clarifications, Chapters 1-3 of ASME Std. BTH-1, Design of Below-the-Hook Lifting Devices (Ref. 3). The NRC staff expects endorsement of these consensus standards to provide safety and efficiency benefits.

Applicability This RG applies to applicants and licensees subject to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities" (Ref. 4), all applicants and licensees for a power reactor combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 5), and applicants and licensees for an independent spent fuel storage installation (ISFSI) under 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste (Ref. 6). This RG also applies, in part, to applicants and holders of standard design certifications, standard design approvals, or manufacturing licenses for power reactors under 10 CFR Part 52.

Applicable Regulations The regulations in 10 CFR Part 50 establish criteria for the licensing of production and utilization facilities, including nuclear power plants. The specific sections applicable to this RG include the following:

This RG is being issued in draft form to involve the public in the development of regulatory guidance in this area. It has not received final staff review or approval and does not represent an NRC final staff position. Public comments are being solicited on this DG and its associated regulatory analysis. Comments should be accompanied by appropriate supporting data. Comments may be submitted through the Federal rulemaking Web site, http://www.regulations.gov, by searching for draft regulatory guide DG-1381. Alternatively, comments may be submitted to the Office of Administration, Mailstop: TWFN 7A-06M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, ATTN:

Program Management, Announcements and Editing Staff. Comments must be submitted by the date indicated in the Federal Register notice.

Electronic copies of this DG and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at https://nrcweb.nrc.gov/reading-rm/doc-collections/reg-guides/. The DG is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML21006A335. The regulatory analysis may be found in ADAMS under Accession No. ML21006A337.

• 10 CFR 50.34, Contents of application; technical information, requires, in part, that power reactor license applicants include in the safety analysis report: the design of the facility (including the principle design criteria, the design bases, and the relationship of the design bases to the principle design criteria); a description and analysis of auxiliary and fuel handling systems insofar as they are pertinent to showing that safety functions will be accomplished, and a description of plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of SSCs. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.

• Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50 establishes minimum requirements for the principal design criteria for nuclear power plants. The general design criteria (GDC) in 10 CFR Part 50, Appendix A, applicable to this RG include the following:

o GDC 1, Quality standards and records, requires structures, systems, and components (SSCs) important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

Where generally recognized codes and standards are used, they shall be identified and evaluated to determine their applicability, adequacy, and sufficiency and shall be supplemented or modified as necessary to assure a quality product in keeping with the required safety function.

o GDC 2, Design bases for protection against natural phenomena, requires that SSCs important to safety be designed to withstand the effects of natural phenomena such as earthquakes, with appropriate consideration in the design basis of (1) the most severe natural phenomena historically reported, (2) combinations of normal and accident conditions with the effects of the natural phenomena, and (3) the importance of the safety functions threatened by the natural phenomena.

o GDC 4, Environmental and dynamic effects design bases, requires appropriate protection for SSCs important to safety against dynamic effects, including the effects of missiles (e.g., falling heavy loads) that may result from equipment failures.

The regulations in 10 CFR Part 52 establish criteria for the issuance of early site permits, standard design certifications, combined licenses, standard design approvals, and manufacturing licenses for nuclear power facilities. To that end, the specific 10 CFR Part 52 regulations applicable to this RG include the following:

• 10 CFR 52.47, Contents of applications; technical information; 10 CFR 52.137, Contents of applications; technical information; and 10 CFR52.157, Contents of applications; technical information in final safety analysis report; , require, in part, that applicants for nuclear power facility standard design certifications, standard design approvals, and manufacturing licenses, respectively, include in the safety analysis report: the design of the facility (including the principle design criteria, the design bases, and the relationship of the design bases to the principle design criteria); and a description and analysis of auxiliary and fuel handling systems insofar as they are pertinent to showing that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.

• 10 CFR 52.79, Contents of applications; technical information in final safety analysis report, requires, in part, that power reactor combined license applicants include in the safety DG - 1381, Page 2 of 16

analysis report: the design of the facility (including the principle design criteria, the design bases, and the relationship of the design bases to the principle design criteria); a description and analysis of auxiliary and fuel handling systems insofar as they are pertinent to showing that safety functions will be accomplished, and a description of plans for conduct of normal operations, including maintenance, surveillance, and periodic testing of SSCs. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations.

The regulations in 10 CFR Part 72 establish criteria for the issuance of specific or general licenses to receive, transfer, and possess power reactor spent fuel, power reactor-related greater than Class C (GTCC) waste, and other radioactive materials associated with spent fuel storage in an ISFSI. To that end, the specific 10 CFR Part 72 regulations applicable to this RG include the following:

• 10 CFR72.24, Contents of application: Technical information, requires, in part, that applicants for ISFSIs include in the application: the design of the facility (including design criteria, the design bases, materials of construction, and applicable codes and standards); an evaluation of the design of SSCs important to safety with regard to their performance in preventing or mitigating accidents; and a description of the quality assurance program that satisfies the requirements of 10 CFR Part 72, Subpart G, Quality Assurance, applied to design, fabrication, construction, testing, and operation of SSCs important to safety.

• 10 CFR 72.122, Overall requirements, requires, in part, that SSCs important to safety must be designed to withstand postulated accidents, such as handling system component failures, and appropriate consideration of the effects of natural phenomena.

Related Guidance

• RG 1.13, Spent Fuel Storage Facility Design Basis (Ref. 7), provides guidance specifying that cranes capable of carrying heavy loads should be prevented, preferably by design rather than by interlocks, from moving over the spent fuel storage pool.

• NUREG-0612, Control of Heavy Loads at Nuclear Power Plants (Ref. 8), provides criteria for protection of critical SSCs from the effects of heavy load handling system failures and specifies good standard industrial practices for the handling of heavy loads.

• NUREG-0554, Single Failure-Proof Cranes for Nuclear Power Plants (Ref. 9), provides technical guidance for the design, fabrication, installation, and testing of overhead cranes with the ability to withstand credible component failures, natural phenomena, and operator errors while maintaining control of the suspended load.

Purpose of Regulatory Guides The NRC issues RGs to describe methods that are acceptable to the staff for implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific issues or postulated events, and to describe information that the staff needs in its review of applications for permits and licenses. Regulatory guides are not NRC regulations and compliance with them is not required. Methods and solutions that differ from those set forth in RGs are acceptable if supported by a basis for the issuance or continuance of a permit or license by the Commission.

Paperwork Reduction Act DG - 1381, Page 3 of 16

This RG provides voluntary guidance for implementing the mandatory information collections in 10 CFR Parts 50, 52, and 70 that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et.

seq.). These information collections were approved by the Office of Management and Budget (OMB),

approval numbers 3150-0011, 3150-0151 and 3150-0132. Send comments regarding this information collection to the FOIA, Library, and Information Collections Branch (T6-A10M), U.S. Nuclear Regulatory Commission, Washington, DC 20555 0001, or by e-mail to Infocollects.Resource@nrc.gov,,

and to the OMB reviewer at: OMB Office of Information and Regulatory Affairs (3150-0011, 3150-0151, and 3150-0132), Attn: Desk Officer for the Nuclear Regulatory Commission, 725 17th Street, NW Washington, DC20503; e- mail: oira_submission@omb.eop.gov.

Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless the document requesting or requiring the collection displays a currently valid OMB control number.

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B. DISCUSSION Reason for Issuance The NRC staff is issuing the RG for the purpose of endorsing the following three ASME Standards as stated:

• ASME NML-1 provides programmatic guidance and is endorsed with clarifications that specify information needed to demonstrate that safety functions would be appropriately protected against handling system equipment failures. NRC endorsement of ASME Std. NML-1 in this RG provides an update to the guidance in the NRC technical report NUREG-0612.

• ASME NOG-1 is the crane design standard and is endorsed with clarifications that specify information needed related to application of quality assurance measures and the interface of the crane with other structures. ASME Std. NOG-1 is used for the design of overhead cranes with multiple girders and top running trollies using wire rope hoists. NRC endorsement of ASME Std.

NOG-1 in this RG provides an update to the guidance in NRC technical report NUREG-0554.

• ASME BTH-1 provides criteria for the design of special lifting devices and load lifting attachments. The staff endorses this standard in part, limiting the endorsed scope to that for mechanical special lifting devices conforming to Design Categories B and C as defined in the standard and excluding sections that address specialized equipment (e.g., electrical components, vacuum lifting devices, and electromagnetic lifting devices). This specialized equipment would not be acceptable for the principle safety function of retaining the suspended load, but specialized equipment may be mounted on a special lifting device for other purposes.

The use of consensus standards where available is consistent with Commission policy and provides updated information reflecting operating experience and risk-informed considerations. The existing technical reports, NUREG-0612, issued August 1980, and NUREG-0554, issued May 1979, are outdated and do not reflect the current risk-informed perspective on heavy load handling activities.

Background

The staff recognizes that safe control of heavy load handling activities in nuclear facilities may be accomplished in several ways. The preferred method, as suggested in RG 1.13, is to design the layout of the facility so that overhead lifting equipment cannot operate over or near SSCs essential to accomplishment of fundamental safety functions.1 However, light-water reactor design and operation involve certain load handling activities, such as transfer of irradiated fuel from storage pools to dry storage and removal of the reactor vessel head and internal structures in support of refueling, which, if the load were dropped, could challenge the performance of safety functions. Other reactor types may also require load handling activities for continued operation that could similarly challenge safety functions. In addition, existing facilities were constructed without full consideration of load handling activities that could challenge safety functions in other ways. For these handling evolutions, the staff has accepted other methods of providing reasonable assurance that key safety functions would be accomplished in the event of failures affecting load handling equipment.

1 For the purpose of this RG, the fundamental safety functions are (1) control of nuclear reactivity, (2) adequate removal of heat from the reactor and from stored irradiated fuel, (3) appropriate confinement of radioactive material, and (4) maintenance of adequate shielding against radiation.

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The staff described these methods in NUREG-0612, which provides for the use of one of the following methods in areas where a handling system failure could challenge execution of key safety functions: use of a highly reliable handling system, verification by analysis that safety functions would be accomplished in the event of a load drop, or the use of electrical interlocks or mechanical stops to prevent the motion of loads over SSCs necessary to perform key safety functions. The guidelines in NUREG-0612 establish the following criteria to demonstrate by analysis of load drop consequences that safety functions had been accomplished:

• Releases of radioactive material result in doses well within regulatory limits.

• Damage does not result in a fuel configuration with an effective neutron multiplication factor greater than 0.95.

• Resulting leakage from the reactor vessel and spent fuel pool is within makeup capabilities.

• At least one train of equipment performing essential safety functions would be undamaged.

In addition, the guidelines of NUREG-0612 provide practices addressing the conduct of normal load handling operations, including maintenance and testing. These practices include (1) defined load paths, (2) load handling procedures, (3) trained crane operators, (4) special lifting devices designed and maintained to appropriate standards, (5) appropriately selected standard lifting devices, (6) inspection, testing, and maintenance of the overhead crane to appropriate standards, and (7) overhead cranes designed to appropriate minimum standards.

In 2019, ASME issued Std. NML-1 to provide guidelines for conducting lifting and handling operations at nuclear facilities using overhead handling systems. Compared to NUREG-0612 guidelines, the standard covers a broader scope in terms of the types of overhead handling systems and the safety significance of the load handling activities. The NUREG-0612 guidelines applied only to inherently stable overhead cranes with top running trolleys and wire rope hoists used in areas around stored nuclear fuel or over SSCs essential to achieve and maintain safe shutdown. The guidelines of Std. NML-1 apply to a variety of additional overhead handling systems, including gantries with hydraulic jacking towers commonly used for handling of dry spent fuel storage cask system components, underhung hoists suspended from monorails or jib cranes commonly used for heavy-component movement inside nuclear facility buildings, and mobile cranes with extendable booms for movement of heavy components above and around smaller structures (e.g., service water pump replacement through the roof of a water intake structure).

The guidelines of Std. NML-1 use a risk-informed approach by qualitatively considering the probability and consequences of a load handling event during a planned lift to accommodate variations in safety significance and the associated controls applied to manage the risk of the lift. The probability of a load handling event is established by evaluating the handling system design, rigging configuration, and load characteristics. The potential consequences of a load handling event are classified considering the properties of the lifted load and the SSCs in the vicinity of the load path. The classifications are standard lift, special lift, and critical lift, with nuclear safety critical lift a subset of critical lift. Nuclear safety critical lifts encompass lifts that could challenge the continued performance of fundamental safety functions and are comparable to the scope of lifts the staff considered in developing NUREG-0612.

Maintenance activities in which a load handling event could reduce the redundancy of equipment available to perform a fundamental safety function but could not directly prevent the accomplishment of the function may be classified as special lifts.

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The guidelines of ASME Std. NML-1 address the plans for conduct of normal handling system operations, including maintenance, surveillance, and periodic testing of handling system SSCs. These plans for normal operations include lift procedures; crane operator and rigger qualifications; crane inspection, testing, and maintenance; and selection and use of lifting devices not specially designed.

These guidelines have been expanded to address the variety of load handling systems that may be used for safety-significant load handling evolutions. These guidelines provide an acceptable means of defining plans for conduct of normal heavy load handling activities consistent with applicable regulations.

The guidelines of ASME Std. NML-1 also establish criteria for protection of safety functions during nuclear safety critical lifts. These criteria include control of load motion by design and interlocks, enhanced safety handling systems, and engineering controls that include analysis of a postulated load drop. The enhanced safety handling system includes a crane designed as single failure proof, such as an overhead crane designed to meet the Type I criteria of ASME Std. NOG-1, and lifting devices, which may include special lifting devices designed to ASME Std. BTH-1 for recurrent lifts. The acceptance criteria for load drop analyses remain the same as those in the NUREG-0612 guidelines in all material aspects.

The ASME Std. NML-1 guidelines for establishment of an enhanced safety handling systems using an overhead crane designed to meet the Type I criteria of ASME Std. NOG-1 in its entirety constitute an complete, acceptable method of evaluation to demonstrate that safety functions would be accomplished with appropriate consideration of the effects of natural phenomena or credible equipment failures. The ASME Std. NML-1 guidelines for enhanced safety handling systems using cranes of other designs, controlled ranges of motion, or to establish engineering controls derived from analysis of postulated load drops provide an acceptable approach to develop a method of evaluation demonstrating that safety functions would be accomplished.

The NRC staff defined criteria for cranes used as part of a highly reliable handling system in NUREG-0554. This NUREG provides technical guidance for the design, fabrication, installation, and testing of overhead cranes with the ability to withstand credible component failures, natural phenomena, and operator errors while maintaining control of the suspended load. The specifications for a Type I crane in ASME Std. NOG-1 provide a more comprehensive standard for design, fabrication, and preoperational testing of these cranes that better incorporates knowledge from crane manufacturers, crane users, and the NRC staff, among other stakeholders, as part of the consensus standard development process.

Many operating reactor licensees have committed to using the guidelines of NUREG-0612 for heavy load handling activities. Section 2-6, Nuclear Safety Critical Lifts, paragraph (b), of ASME Std. NML-1 states that facilities with a control of the heavy load handling program described in the facility safety analysis report may continue to handle nuclear safety critical lifts in a manner consistent with the control of heavy loads program. Nonmandatory Appendix A to ASME Std. NML-1 includes a conformance matrix comparing the guidelines of NUREG-0612 with the guidelines from ASME Std. NML-1. Similarly, nonmandatory Appendix C to ASME Std. NOG-1 includes a conformance matrix comparing the guidelines of NUREG-0554 with the overhead crane design specifications in ASME Std. NOG-1. These appendices serve as an aid to licensees that desire to modify their licensing basis to use ASME Std. NML-1 and/or ASME Std. NOG-1 for the control of heavy loads pursuant to the requirements of 10 CFR 50.59, Changes, tests and experiments. However, consistent with NRC backfit and issue finality policies, the continued use of existing heavy load handling programs or development of new heavy load handling programs based on the provisions of NUREG-0612 and NUREG-0554 remains acceptable.

Similarly, the licenses for ISFSIs (Specific or General) have invoked the guidelines of NUREG-0612 for the handling of loaded canisters. These license requirements typically specify the use of handling systems including cranes that satisfy the criteria of NUREG-0554 unless the loaded canister and any surrounding overpack has been designed to withstand postulated drops that encompass proposed handling DG - 1381, Page 7 of 16

operations. As stated above, the continued use of existing heavy load handling programs or development of new heavy load handling programs based on the provisions of NUREG-0612 and NUREG-0554 remains acceptable.

The staff has issued modifications and clarifications to regulatory guidance related to control of heavy load handling activities in Regulatory Issue Summary (RIS) 05-25, Clarification of NRC Guidelines for Control of Heavy Loads (Ref. 10); RIS-05-025, Supplement 1, Clarification of NRC Guidelines for Control of Heavy Loads (Ref. 11); and RIS-08-28, Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts (Ref. 12). In RIS-05-025, the staff discusses the relationship between heavy load handling guidelines and nuclear power plant operating experience. The NRC issued Supplement 1 to RIS-05-025 to notify stakeholders of modifications to heavy load handling regulatory guidance for NRC staff that were derived from the operating experience.

Separately, the NRC issued RIS-08-028 to endorse initiative guidelines developed by the Nuclear Energy Institute (NEI), which were intended for voluntary implementation by licensees to ensure that heavy loads lifts continue to be conducted safely and that the plant licensing bases accurately reflect facility heavy load handling practices. These guidance documents continue to be relevant to those facilities that have incorporated the regulatory information into their facility licensing basis. However, ASME Std. NML-1 included changes to heavy load handling guidance derived from operating experience and an allowance for incorporation of existing heavy load handling program elements adopted through implementation of the NEI initiative. Therefore, licensees that choose to adopt the guidelines of ASME Std. NML-1 have implicitly adopted the guidance associated with the NUREGs and generic communications discussed above and further consideration of their content during license applications and other change processes is not necessary.

Consideration of International Standards The International Atomic Energy Agency (IAEA) works with member states and other partners to promote the safe, secure, and peaceful use of nuclear technologies. The IAEA develops Safety Requirements and Safety Guides for protecting people and the environment from harmful effects of ionizing radiation. This system of safety fundamentals, safety requirements, safety guides, and other relevant reports, reflects an international perspective on what constitutes a high level of safety. To inform its development of this RG, the NRC considered IAEA Safety Requirements and Safety Guides pursuant to the Commissions International Policy Statement (Ref. 13), and Management Directive and Handbook 6.6, Regulatory Guides (Ref. 14).

The following IAEA Safety Requirements were considered in the development/ update of this Regulatory Guide:

• IAEA Specific Safety Requirements (SSR)-2/1, Safety of Nuclear Power Plants: Design (Ref. 15), addresses design considerations for overhead lifting equipment in nuclear power plants in Section 6, Design of Specific Plant Systems, as Requirement 76: Overhead Lifting Equipment, and associated paragraph 6.55. This RG incorporates similar design guidelines and is consistent with the fundamental safety principles in IAEA SSR-2/1.

Documents Discussed in Staff Regulatory Guidance This RG endorses the use of one or more codes or standards developed by external organizations, and other third party guidance documents. These codes, standards and third party guidance documents may contain references to other codes, standards or third party guidance documents (secondary references). If a secondary reference has itself been incorporated by reference into NRC regulations as a requirement, then licensees and applicants must comply with that standard as set forth in the regulation. If DG - 1381, Page 8 of 16

the secondary reference has been endorsed in a RG as an acceptable approach for meeting an NRC requirement, then the standard constitutes a method acceptable to the NRC staff for meeting that regulatory requirement as described in the specific RG. If the secondary reference has neither been incorporated by reference into NRC regulations nor endorsed in a RG, then the secondary reference is neither a legally-binding requirement nor a generic NRC approved acceptable approach for meeting an NRC requirement. However, licensees and applicants may consider and use the information in the secondary reference, if appropriately justified, consistent with current regulatory practice, and consistent with applicable NRC requirements.

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C. STAFF REGULATORY GUIDANCE The NRC staff is issuing this RG for the purpose of endorsing the following three ASME Standards, with the clarifications stated below:

• ASME Std. NML-1, Rules for the Movement of Loads Using Overhead Handling Equipment in Nuclear Facilities

• ASME Std. NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder)

• ASME Std. BTH-1, Design of Below-the-Hook Lifting Devices, in part, only Chapters 1 through 3.

Regulatory Position C.1

1. This RG endorses ASME Std. NML-1-2019 as acceptable guidance for establishing a nuclear facility heavy load handling program, subject to the following clarifications:

a. In regard to Section 2-6.1(c)(1) of ASME Std. NML-1, if handling-system-controlled range of motion is credited to ensure that essential safety functions would be maintained, then the evaluation of handling system controls for range of motion should include the following considerations:

(1) Boundaries of the range of motion should include the following specific factors:

• a margin for tipping equal to the height of the load when the characteristics of planned handling system lifts include a load with a center of gravity higher than the width of its base in its planned handling orientation; and

• exclusion of components from within the range of motion when secondary effects from that components failure (e.g., flooding from pressurized piping that could credibly cause failure of equipment outside the boundary) would prevent performance of an essential safety function; and (2) administrative measures should prevent unintended disabling of the range of motion controls when they are not a permanent feature of the design (e.g., key-operated bypass of electrical motion interlocks).

b. In regard to Section 2-6.1(c)(2) of ASME Std. NML-1, if enhanced handling system reliability is credited to ensure essential safety functions would be maintained, then the handling system design should meet the crane and lifting device standards by conforming to the following guidelines:

(1) The crane should meet the endorsed standard as described in Regulatory Position C.2, unless one of the exceptions defined in (2) applies.

(2) If the proposed nuclear safety critical lift is

• outside of nuclear power plant structures (e.g., operations related to an independent spent fuel storage facility),

• involves an infrequent major component replacement, or DG - 1381, Page 10 of 16

• inside nuclear power plant structures with inadequate space or strength to accommodate a crane conforming to Regulatory Position C.2, then the alternative lifting system designs should be consistent with the critical lift guidelines of Section 4-1.1, Overhead Crane, of ASME Std. NML-1, and the design should satisfy the following criteria:

(a) meet applicable national consensus standard(s) for the type(s) of lifting systems used, (b) apply quality assurance in design, fabrication, installation, and initial testing, (c) apply conservative design criteria to the structural elements essential to support the load, (d) apply conservative design criteria to the mechanical components essential to stopping or holding the load, (e) include redundancy in the design of mechanical components essential to stopping or holding the load that are subject to fatigue or wear, (f) use fail-safe electrical systems and components when failure of the electrical system or component could affect the ability to stop or hold the load, (g) Apply appropriate quality assurance measures to electrical components intended to detect equipment failures and that actuate following those equipment failures to stop or hold the load.

(3) Special lifting devices and load attachment points should meet the endorsed standard as described in Regulatory Position C.3.

(4) Standard lifting devices (e.g., slings) and lifting hardware (e.g., shackles) should meet the selection criteria and use restrictions specified in Sections 5-1.2.1 and 5-2 of ASME Std. NML-1.

c. In regard to Section 2-6.1(c)(3) of ASME Std. NML-1, the following clarifications apply to analyses of postulated load drops with the potential to damage multiple irradiated fuel assemblies:

(1) The analysis of radiological consequences should conform with appropriate guidance from RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Ref. 16).

(2) The consequences of the postulated drop should not result in deformation or repositioning of the fuel pins that would preclude modeling of the fuel as a regular array for evaluation of nuclear reactivity.

d. In regard to Section 2-6.2, Movement of Loads over Irradiated Fuel, of ASME Std. NML-1, the staff endorses the use of an enhanced reliability handling system for handling heavy loads directly over irradiated fuel that satisfies Regulatory Position C.1.b as the preferred means of ensuring safety functions would be accomplished, or, as a secondary option, use of engineering controls as described in Section 2-6.1(c)(3) of ASME Std. NML-1 and clarified by Regulatory Position C.1.c.

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Regulatory Position C.2

2. This RG endorses ASME Std. NOG-1-2020 criteria for a Type I crane in its entirety as acceptable guidance for a crane with multiple girders, a top running trolley, and a wire rope hoist used as part of an enhanced reliability handling system, subject to the following clarifications:

a. A quality assurance program equivalent to the applicants or licensees quality assurance program should be applied to the design, fabrication, installation, and initial testing of the crane. The scope of the quality assurance program should encompass those components defined in Section 2000, Quality Assurance, of ASME Std. NOG-1 for a Type I crane.

b. The crane runway is outside the scope of ASME NOG-1, and therefore, the adequacy of the runway structure to support the transmitted load for the load combinations included in the crane design should be confirmed, consistent with the facility design basis for structures important to safety.

Regulatory Position C.3

3. This RG endorses the guidelines of Chapters 1 through 3 of ASME Std. BTH-1-2017 for design and fabrication of special lifting devices and load attachment points that form a mechanical load path from the load to the crane hook (or other designed attachment point) and are used as part of an enhanced reliability handling system. The specific design criteria are subject to the following clarifications:

a. Design Category B applies to lifting devices configured to provide two independent load paths between the load and the crane attachment points in a manner such that each load path can independently hold the load in a stable configuration.

b. Design Category C applies to lifting devices configured to provide a single load path between the load and the crane attachment point.

c. Load attachment points for both standard lifting devices and special lifting devices should be designed to Design Category B criteria for two independent load paths and Design Category C criteria for single load path configurations.

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D. IMPLEMENTATION The NRC staff may use this regulatory guide as a reference in its regulatory processes, such as licensing, inspection, or enforcement. However, the NRC staff does not intend to use the guidance in this regulatory guide to support NRC staff actions in a manner that would constitute backfitting as that term is defined in 10 CFR 50.109, Backfitting; 10 CFR 70.76, Backfitting, 10 CFR 72.62, Backfitting;and as described in NRC Management Directive 8.4, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, (Ref. 17), nor does the NRC staff intend to use the guidance to affect the issue finality of an approval under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. The staff also does not intend to use the guidance to support NRC staff actions in a manner that constitutes forward fitting as that term is defined and described in Management Directive 8.4. If a licensee believes that the NRC is using this regulatory guide in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfitting or forward fitting appeal with the NRC in accordance with the process in Management Directive 8.4.

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REFERENCES2 2 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html The documents can also be viewed online or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD. For problems with ADAMS, contact the PDR staff at 301-415-4737 or (800) 397-4209; fax (301) 415-3548; or e-mail pdr.resource@nrc.gov.

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1 American Society of Mechanical Engineers (ASME), Standard NML-1-2019, Rules for the Movement of Loads Using Overhead Handling Equipment in Nuclear Facilities, New York, NY, June 28, 2019.

2 ASME, Standard NOG-1-2020, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), New York, NY, December 4, 2020.

3 ASME, Standard BTH-1-2017, Design of Below-the-Hook Lifting Devices, New York, NY, March 15, 2017.

4 U.S. Code of Federal Regulations (CFR), Domestic licensing of production and utilization facilities, Part 50, Chapter 1, Title 10, Energy.

5 CFR, Licenses, certifications, and approvals for nuclear power plants, Part 52, Chapter 1, Title 10, Energy.

6 CFR, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste, Part 72, Chapter 1, Title 10, Energy.

7 U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Storage Facility Design Basis, Regulatory Guide (RG) 1.13.

8 NRC, Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36, NUREG-0612, Agencywide Documents Access and Management System (ADAMS)

Accession No. ML070250180.

9 NRC, Single-Failure-Proof Cranes for Nuclear Power Plants, NUREG-0554, ADAMS Accession No. ML110450636.

10. NRC, Regulatory Issue Summary (RIS)05-025, Clarification of NRC Guidelines for Control of Heavy Loads, October 31, 2005. ADAMS Accession No. ML052340485.

11 NRC, RIS 05-025, Supplement 1,Clarification of NRC Guidelines for Control of Heavy Loads, May 29, 2007. ADAMS Accession No. ML071210434.

12 NRC, RIS 08-28, Endorsement of Nuclear Energy Institute Guidance for Reactor Vessel Head Heavy Load Lifts, December 1, 2008. (ADAMS Accession No. ML082460291

13. NRC, Nuclear Regulatory Commission International Policy Statement, Federal Register, Vol. 79, No. 132, July 10, 2014, pp. 39415-39418.

14 NRC, Management Directive (MD) 6.6, Regulatory Guides, Washington, DC, May 2, 2016 (ADAMS Accession No. ML18073A170).

15 International Atomic Energy Agency (IAEA), Safety of Nuclear Power Plants: Design, Specific Safety Requirements (SSR) 2/1, Rev. 1, Vienna, Austria, February 2016.

16 NRC, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, RG 1.183.

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17 NRC, Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests, Management Directive 8.4, September 20, 2019, ADAMS Accession No. ML18093B087.

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