ML050610258

From kanterella
Jump to navigation Jump to search
Sher Bahadur Memo Request to Waive CRGR Review of Proposed Revision to Regulatory Guide 3.71 Nuclear Criticality Safety Standards for Fuels and Materials Facilities
ML050610258
Person / Time
Issue date: 03/25/2005
From: Strosnider J
Office of Nuclear Material Safety and Safeguards
To: Bahadur S
NRC/CRGR
References
DG-3023, RG-3.071, Rev 01
Download: ML050610258 (32)
v • d • e


Text

March 25, 2005 MEMORANDUM TO: Sher Bahadur, Chairman Committee To Review Generic Requirements FROM: Jack R. Strosnider, Director /RA/

Office of Nuclear Material Safety and Safeguards

SUBJECT:

REQUEST TO WAIVE COMMITTEE TO REVIEW GENERIC REQUIREMENTS REVIEW OF PROPOSED REVISION TO REGULATORY GUIDE 3.71, NUCLEAR CRITICALITY SAFETY STANDARDS FOR FUELS AND MATERIALS FACILITIES The staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS), has prepared a draft revision of Regulatory Guide (RG) 3.71, which is temporarily identified as Draft Regulatory Guide DG-3023. For your information, Attachment 1 to this memorandum is a copy of Draft Regulatory Guide DG-3023, Attachment 2 briefly describes the changes, and Attachment 3 shows the differences between the current and revised versions of the guide in redlining and strikethrough, and Attachment 4 provides the NMSS answers to the Committee To Review Generic Requirements (CRGR) Appendix C questions.

Prior to 1998, the NRC issued a variety of regulatory guides to endorse most of the Series 8 nuclear criticality safety standards promulgated by the American National Standards Institute and the American Nuclear Society (ANSI/ANS). The original publication of RG 3.71 (in 1998) consolidated and superseded the other regulatory guides (which the NRC subsequently withdrew) without altering any existing licensing commitments or introducing new requirements.

Specifically, the NRC withdrew RG 3.1-1987, RG 3.4-1986, RG 3.43-1979, RG 3.45-1989, RG 3.47-1981, RG 3.57-1986, RG 3.58-1986, RG 3.68-1994, RG 3.70-1997, and RG 8.12-1988.

With this memorandum, the NMSS staff is asking the (CRGR) to waive its review of the revised guide. In support of this request, we note that the guidance contained in the document does not involve a backfit. As indicated in Attachment 2, the revisions involve (1) endorsing one new standard developed since the NRC originally published RG 3.71 in 1998, (2) endorsing six newer versions of standards currently in RG 3.71, (3) clarifying one current endorsement to reflect revisions to the related standard since 1998, (4) clarifying endorsement of one standard that ANSI/ANS withdrew after 1998, and (5) incorporating editorial changes to reflect the NRCs current regulatory guide format.

If you have any questions regarding the revision of RG 3.71, please contact Harry Felsher at (301) 415-5521 or HDF@nrc.gov.

Attachments: 1. Draft Regulatory Guide DG-3023

2. Description of Changes to Regulatory Guide 3.71
3. Redline Strikeout Comparison of NRC Regulatory Guide 3.71
4. Appendix C to CRGR Charter

If you have any questions regarding the revision of RG 3.71, please contact Harry Felsher at (301) 415-5521 or HDF@nrc.gov.

Attachments: 1. Draft Regulatory Guide DG-3023

2. Description of Changes to Regulatory Guide 3.71
3. Redline Strikeout Comparison of NRC Regulatory Guide 3.71
4. Appendix C to CRGR Charter DISTRIBUTION:

FCSS r/f NMSS r/f ML050610258

  • See Previous Concurrence INDICATE IN BOX: E=COPY W/ATT/ENCL; C=COPY W/O ATT/ENCL; N=NO COPY OFFICE TSG E TSG E Tech Ed C FCSS NMSS NAME HFelsher
  • MGalloway
  • PGarrity
  • RPierson JStrosnider DATE 03/02/05 03/08/05 03/03/05 03/24/05 03/25/05 U.S. NUCLEAR REGULATORY COMMISSION March 2005 OFFICE OF NUCLEAR REGULATORY RESEARCH Division 3 DRAFT REGULATORY GUIDE

Contact:

H.D. Felsher, (301) 415-5521 DRAFT REGULATORY GUIDE DG-3023 (Proposed Revision 1 of Regulatory Guide 3.71, dated August 1998)

NUCLEAR CRITICALITY SAFETY STANDARDS FOR FUELS AND MATERIAL FACILITIES A. INTRODUCTION This revised regulatory guide provides licensees and applicants with updated guidance concerning criticality safety standards that the U.S. Nuclear Regulatory Commission (NRC) has endorsed for use with nuclear fuels and material facilities. As such, this guide describes methods that the NRC staff considers acceptable for complying with the NRCs regulations in Title 10, Parts 70 and 76, of the Code of Federal Regulations (10 CFR Parts 70 and 76).

In 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, Section 70.20, General License To Own Special Nuclear Material, defines a specific license to acquire, deliver, receive, possess, use, transfer, import, or export special nuclear material. According to 10 CFR 70.22, Contents of Applications, each application for such a license must contain proposed procedures to avoid nuclear criticality accidents. In 10 CFR Part 76, Certification of Gaseous Diffusion Plants, Section 76.87, Technical Safety Requirements, states that the technical safety requirements should reference procedures and equipment that are applicable to criticality prevention.

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory position in this area. It has not received staff review or approval and does not represent an official NRC staff position.

Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Comments may be submitted electronically through the NRCs interactive rulemaking Web page at http://www.nrc.gov/what-we-do/regulatory/rulemaking.html.

Copies of comments received may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. Comments will be most helpful if received by April 11, 2005.

Requests for single copies of draft or active regulatory guides (which may be reproduced) or for placement on an automatic distribution list for single copies of future draft guides in specific divisions should be made to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this draft regulatory guide are available through the NRCs interactive rulemaking Web page (see above); the NRCs public Web site under Draft Regulatory Guides in the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/; and the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML050390450. Note, however, that the NRC has temporarily limited public access to ADAMS so that the agency can complete security reviews of publicly available documents and remove potentially sensitive information. Please check the NRCs Web site for updates concerning the resumption of public access to ADAMS.

The NRC staff has developed this regulatory guide to provide guidance on complying with these portions of the NRCs regulations by describing procedures for preventing nuclear criticality accidents in operations that involve handling, processing, storing, and/or transporting special nuclear material at fuel and material facilities. This regulatory guide endorses specific nuclear criticality safety standards developed by the American Nuclear Societys Standards Subcommittee 8 (ANS-8), Operations with Fissionable Materials Outside Reactors. This guide is not intended to be used by nuclear reactor licensees.

Regulatory guides are issued to describe to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations, and compliance with regulatory guides is not required. Regulatory guides are issued in draft form to solicit public comment and involve the public in developing the agencys regulatory positions.

Draft regulatory guides have not received complete staff review; therefore, they do not represent official NRC staff positions.

This draft regulatory guide contains information collections that are covered by the requirements of 10 CFR Part 70, which the Office of Management and Budget (OMB) approved under OMB control number 3150-0009. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid OMB control number. However, this draft regulatory guide contains additional information collections that are covered by the requirements of 10 CFR 76.8, which apply to a wholly-owned instrumentality of the United States and affect fewer than ten respondents. As a result, OMB clearance is not required pursuant to the Paperwork Reduction Act (44 U.S.C. 3501, et seq.).

B. DISCUSSION The NRC initially issued Regulatory Guide 3.71 in 1998 to provide guidance concerning procedures that the staff considered acceptable for complying with the agencys regulatory requirements in 10 CFR 70.20, 70.22 and 76.87. Toward that end, the original guide endorsed specific safety standards that ANS-8 developed to provide guidance, criteria, and best practices for use in preventing and mitigating criticality accidents during operations that involve handling, processing, storing, and/or transporting special nuclear material at fuel and material facilities.

The original guide also took exceptions to certain portions of individual ANSI/ANS-8 standards.

In addition, the original guide consolidated and replaced a number of earlier NRC regulatory guides, thereby providing all of the relevant guidance in a single document.

The ANS-8 standards endorsed in Regulatory Guide 3.71 were approved by the American Nuclear Societys Consensus Committee N16 on Nuclear Criticality Safety, as well as the American National Standards Institute (ANSI). Nonetheless, each ANSI/ANS-8 standard is reviewed every 5 years by a working group of expert practitioners in the area so that it can be revised, reaffirmed, or withdrawn, as appropriate to reflect the current state of the art. (This time can be extended to as long as 10 years or more under special circumstances.) New standards are also added when the need arises. Since the timing and issuance of individual standards is independent of the other standards, the list of current standards and their respective dates of issuance is constantly changing.

As a result, since the NRC published Regulatory Guide 3.71 in 1998, several ANSI/ANS-8 nuclear criticality safety standards have been added, reaffirmed, revised, or withdrawn.

Consequently, the NRC staff has decided to update this guide to clarify which standards the agency endorses and to clearly state exceptions to individual standards. This proposed revision does not change any of the guidance provided in Regulatory Guide 3.71; rather, it provides guidance concerning changes that have occurred since the NRC published the original guide in 1998. For completeness, this guide restates the endorsements and exceptions stated in Regulatory Guide 3.71, as applicable, while identifying endorsements of or exceptions to new or modified standards. Since the ANSI/ANS-8 standards are constantly being issued, revised, reaffirmed, or withdrawn, the NRC staff plans to revise this guide on a regular basis.

C. REGULATORY POSITION The ANSI/ANS-8 nuclear criticality safety standards provide procedures and methodologies that the NRC staff considers generally acceptable for use in preventing and mitigating nuclear criticality accidents. However, use of the ANSI/ANS-8 nuclear criticality safety standards is not a substitute for detailed nuclear criticality safety analyses for specific operations.

In addition, inclusion of a reference to another standard in an endorsed standard does not imply NRC endorsement of the referenced standard.

The NRC staff will follow the requirements denoted in the ANSI/ANS-8 standards. The word shall in an ANSI/ANS-8 standard denotes a requirement; the word should denotes a recommendation; and the word may denotes permission (neither a requirement nor a recommendation). When a licensee or applicant commits to an ANSI/ANS-8 standard cited in this regulatory guide, the licensee or applicant must perform all operations in accordance with the requirements stated in that standard, but not necessarily with its recommendations. Licensees or applicants may follow the recommendations given in the ANSI/ANS-8 standards, unless an exception is stated in this regulatory guide, otherwise specified in 10 CFR Parts 70 or 76, or addressed by other acceptable methods.

1. ANSI/ANS-8 Nuclear Criticality Standards Endorsed by the NRC The NRC endorses the following ANSI/ANS-8 nuclear criticality safety standards:
  • ANSI/ANS-8.5-1996 (Reaffirmed in 2002), Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material
  • ANSI/ANS-8.6-1983 (Reaffirmed in 2001), Safety in Conducting Subcritical Neutron-Multiplication Measurements In Situ
  • ANSI/ANS-8.12-1987 (Reaffirmed in 2002), Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors
  • ANSI/ANS-8.14-2004, Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors
  • ANSI/ANS-8.15-1981 (Reaffirmed in 1995), Nuclear Criticality Control of Special Actinide Elements
  • ANSI/ANS-8.21-1995 (Reaffirmed in 2001), Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors
  • ANSI/ANS-8.22-1997, Nuclear Criticality Safety Based on Limiting and Controlling Moderators
2. ANSI/ANS-8 Nuclear Criticality Standards Endorsed by the NRC with Exceptions The NRC endorses the following ANSI/ANS-8 nuclear criticality safety standards, but takes exception to certain sections, as follows:
  • ANSI/ANS-8.1-1998, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors The guidance on validating calculational methods for nuclear criticality safety, as specified in ANSI/ANS-8.1-1998, provides a procedure that is acceptable to the NRC staff for establishing the validity and applicability of calculational methods used in assessing nuclear criticality safety. However, it is not sufficient to merely refer to this standard in describing the validation of a method. Rather, a licensee or applicant should provide the details of validation (as stated in Section 4.3.6 of the standard) to (1) demonstrate the adequacy of the margins of subcriticality relative to the bias and criticality parameters, (2) demonstrate that the calculations embrace the range of variables to which the method will be applied, and (3) demonstrate the trends in the bias upon which the licensee or applicant will base the extension of the area of applicability. In addition, the details of validation should state computer codes used, operations, recipes for choosing code options (where applicable), cross-section sets, and any numerical parameters necessary to describe the input.
  • ANSI/ANS-8.3-1997 (Reaffirmed in 2003), Criticality Accident Alarm System The guidance on criticality accident alarm systems, as specified in ANSI/ANS-8.3-1997 (reaffirmed in 2003), is generally acceptable to the NRC staff. An exception is that 10 CFR 70.24, Criticality Accident Requirements, requires criticality alarm systems in each area in which special nuclear material is handled, used, or stored, whereas Section 4.2.1 of the standard merely requires an evaluation for such areas. Another exception is that 10 CFR 70.24 and 10 CFR 76.89, Criticality Accident Requirements, require that each area must be covered by two detectors, whereas Section 4.5.1 of the standard permits coverage by a single reliable detector. Finally, 10 CFR 70.24 and 10 CFR 76.89 require a monitoring system capable of detecting a nuclear criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of 2 meters from the reacting material within 1 minute.
  • ANSI/ANS-8.10-1983 (Reaffirmed in 1999), Criteria for Nuclear Criticality Safety Controls in Operations With Shielding and Confinement The guidance on using shielding and confinement as a nuclear criticality safety control, as specified in ANSI/ANS-8.10-1983 (reaffirmed in 1999), is generally accepted by the NRC staff. An exception to Section 4.2.1 of the standard is the assumption that the radiation source strengths and releases from a nuclear criticality accident arise from an excursion occurring in an unfavorable geometry containing a solution of 400 g/L of uranium enriched in U-235. The excursion produces an initial burst of 1x1018 fissions in 0.5 second, followed successively at 10-minute intervals by 47 bursts of 1.9x1017 fissions, for a total of 1x1019 fissions in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The excursion is assumed to be terminated by evaporation of 100 liters of the solution. Licensees and applicants may use a less-conservative nuclear criticality accident condition if detailed analyses of credible nuclear criticality accidents are performed and shown to be applicable to the conditions being evaluated.
  • ANSI/ANS-8.17-1984 (Reaffirmed in 1997), Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors The general safety criteria and criteria to establish subcriticality, as specified in ANSI/ANS-8.17-1984 (reaffirmed in 1997), provide guidance that is acceptable to the NRC staff for preventing nuclear criticality accidents in handling, storing, and transporting fuel assemblies at fuel and material facilities. The only exception is that licensees and applicants may take credit for fuel burnup only when the amount of burnup is confirmed by physical measurements that are appropriate for each type of fuel assembly in the environment in which it is to be stored.
3. ANSI/ANS-8 Nuclear Criticality Standards Withdrawn by the NRC The NRC has withdrawn its endorsement of ANSI/ANS-8.9-1987 (Reaffirmed in 1995),

Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials. This standard, which was listed in Regulatory Guide 3.71-1998, has subsequently been withdrawn by ANS (i.e., it is a historical standard). Although the NRC has withdrawn its endorsement of the standard, there is nothing technically wrong with this standard and it would be acceptable for licensees and applicants to use it.

D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staffs plans for using this draft regulatory guide. No backfitting is intended or approved in connection with the issuance of this guide.

The NRC has issued this draft guide to encourage public participation in its development.

Except when an applicant or licensee proposes or has previously established an acceptable alternative method for complying with specified portions of the NRCs regulations, the methods to be described in the active guide will reflect public comments and will be used in evaluating submittals in connection with license applications submitted under 10 CFR Part 70, Domestic Licensing of Special Nuclear Material, and 10 CFR Part 76, Certification of Gaseous Diffusion Plants.

REGULATORY ANALYSIS The NRC published a draft regulatory analysis when the agency issued the original draft of this guide (as DG-3013) for public comment in January 1998. Since that time, several ANSI/ANS 8 nuclear criticality safety standards have been added, reaffirmed, revised, or withdrawn. Consequently, the NRC staff has decided to update this guide to clarify which standards the agency endorses and to clearly state exceptions to individual standards. However, this proposed revision does not change any of the guidance provided in Regulatory Guide 3.71; rather, it provides guidance concerning changes that have occurred since the NRC published the original guide 1998. Consequently, the NRC staff has not prepared a separate regulatory analysis for this revised regulatory guide. The original regulatory analysis for DG-3013 is available for inspection or copying (for a fee) in the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland. The PDRs mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548; and by email to PDR@nrc.gov.

BACKFIT ANALYSIS This proposed revision of Regulatory Guide 3.71 does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

Brief Description of the Changes to Regulatory Guide 3.71 The staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Material Safety and Safeguards (NMSS), has prepared a draft revision of Regulatory Guide (RG) 3.71, which is temporarily identified as Draft Regulatory Guide DG-3023. This revision involves the following specific changes:

  • endorsing one new Series 8 nuclear criticality safety standard that the American National Standards Institute and the American Nuclear Society (ANSI/ANS) developed since the NRC originally published RG 3.71 in 1998
  • endorsing six newer versions of standards currently in RG 3.71
  • clarifying one current endorsement to reflect revisions to the related standard since 1998
  • clarifying endorsement of one standard that ANSI/ANS withdrew after 1998
  • incorporating editorial changes to reflect the NRCs current regulatory guide format The remainder of this attachment briefly describes the nature and intent of these five changes.

(1) Endorsing one new standard developed since the NRC published RG 3.71 in 1998:

  • ANSI/ANS-8.14-2004, Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors, is not addressed in the current version of RG 3.71. This standard was developed to address the need to provide information about soluble neutron absorbers that is more detailed than the information provided in the general nuclear criticality safety standard ANSI/ANS-8.1. The working group that developed ANSI/ANS-8.14 included representatives from the NRC, NRC licensees, the U.S. Department of Energy (DOE), DOE contractors, the nuclear industry, and unaffiliated members of the nuclear criticality safety community.

Thus, ANSI/ANS-8.14 reflects input from all potential users of the standard.

(2) Endorsing six newer versions of standards currently in NRC RG 3.71:

  • ANSI/ANS-8.5-1996 (Reaffirmed in 2002), Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material, was reaffirmed after the NRC published the current version of RG 3.71. The working group did not make any technical changes to the standard.
  • ANSI/ANS-8.6-1983 (Reaffirmed in 2001), Safety in Conducting Subcritical Neutron-Multiplication Measurements In Situ, was reaffirmed after the NRC published the current version of RG 3.71. The working group did not make any technical changes to the standard.
  • ANSI/ANS-8.7-1998, Nuclear Criticality Safety in the Storage of Fissile Materials, was revised after the NRC published the current version of RG 3.71. The revised standard includes several textual enhancements and tabulated changes resulting from the working groups confirmatory evaluations, which revealed that uncertainties associated with the calculated values were larger than previously evaluated. Therefore, the working group removed Table 5.12 and portions of Tables 5.2, 5.5,and 5.6. Thus, the working group made appropriate technical changes to the standard.
  • ANSI/ANS-8.12-1987 (Reaffirmed in 2002), Nuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors, was reaffirmed after the NRC published the current version of RG 3.71. The working group did not make any technical changes to the standard.
  • ANSI/ANS-8.20-1991 (Reaffirmed in 1999), Nuclear Criticality Safety Training,was reaffirmed after the NRC published the current version of RG 3.71.

The working group did not make any technical changes to the standard.

  • ANSI/ANS-8.21-1995 (Reaffirmed in 2001), Use of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors, was reaffirmed after the NRC published the current version of RG 3.71. The working group did not make any technical changes to the standard.

(3) Clarifying one current endorsement to reflect revisions to the related standard since 1998:

  • ANSI/ANS-8.1-1998, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, was revised after the NRC published the current version of RG 3.71. The working group made changes in the validation section, but did not alter the intent of the previous version. Instead, those changes provided clarification and amplification. Thus, the working group made appropriate changes to the standard. Nonetheless, given the extent of the changes in the validation section and the date of the standard, the NRC staff had to modify its earlier exception to the standard. However, there is no change in the intent of the NRCs previous endorsement with exception.

(4) Clarifying endorsement of one standard that ANSI/ANS withdrew after 1998:

  • ANSI/ANS-8.9-1987 (Reaffirmed in 1995), Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials, is addressed in the current version of RG 3.71. However, ANSI/ANS has since withdrawn this standard (i.e., it is a historical standard). Although the NRC staff has withdrawn its endorsement of the standard, there is nothing technically wrong with this standard, and it remains acceptable for use by licensees and applicants.

(5) Incorporating editorial changes to reflect the NRCs current regulatory guide format:

  • Draft Regulatory Guide DG-3023 was prepared in consultation with a staff member from the NRCs Office of Nuclear Regulatory Research (RES). That RES staff member is responsible for formatting regulatory guides according to agency standards, is an experienced NRC technical editor, and has concurred on the package.

Redline/Strikeout Comparison of current/revised NRC Regulatory Guide 3.71 WordPerfect Document Compare Summary Original document: P:\nrc-rg-3-71.wpd Revised document: @PFDesktop\:MyComputer\P:\draft-rg-3.71.wpd Deletions are shown with the following attributes and color:

Strikeout, Blue RGB(0,0,255).

Deleted text is shown as full text.

Insertions are shown with the following attributes and color:

Double Underline, Redline, Red RGB(255,0,0).

The document was marked with 167 Deletions, 174 Insertions, 0 Moves.

(Draft was issued as DG-3013, Published 1/98)

August 1998 A. IntroductionDRAFT REGULATORY GUIDE DG-3023 (Proposed Revision 1 of Regulatory Guide 3.71, dated August 1998)

NUCLEAR CRITICALITY SAFETY STANDARDS FOR FUELS AND MATERIAL FACILITIES A. INTRODUCTION This revised regulatory guide provides licensees and applicants with updated guidance concerning criticality safety standards that the U.S. Nuclear Regulatory Commission (NRC) has endorsed for use with nuclear fuels and material facilities. As such, this guide describes methods that the NRC staff considers acceptable for complying with the NRCs regulations in Title 10, Parts 70 and 76, of the Code of Federal Regulations (10 CFR Parts 70 and 76).

In 10 CFR Part 70, "DomesticDomestic Licensing of Special Nuclear Material," Section 70.20, "GeneralGeneral License To Own Special Nuclear Material," defines a specific license to acquire, deliver, receive, possess, use, transfer, import, or export special nuclear material.

According to 10 CFR 70.22, "ContentsContents of Applications," each application for such a license must contain proposed procedures to avoid nuclear criticality accidents. In 10 CFR Part 76, "CertificationCertification of Gaseous Diffusion Plants," Section 76.87, "TechnicalTechnical Safety Requirements," states that the technical safety requirements should reference procedures and equipment that are applicable to criticality prevention.

ThisThe NRC staff has developed this regulatory guide has been developed to provide guidance on complying with these portions of the NRC's regulations by describing procedures for preventing nuclear criticality accidents in operations involvingthat involve handling, processing, storing, andand/or transporting special nuclear material at fuels and material facilities. This regulatory guide endorses specific ANSI/ANS-8 nuclear criticality safety standards for these purposes. This guide also consolidates and replaces the guidance from a number of regulatory guides, thereby withdrawing those regulatory guides.developed by the American Nuclear Societys Standards Subcommittee 8 (ANS-8), Operations with Fissionable Materials Outside Reactors. This guide is not intended to be used by nuclear reactor licensees.

The Regulatory guides are issued to describe to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations, and compliance with regulatory guides is not required. Regulatory guides are issued in draft form to solicit public comment and involve the public in developing the agencys regulatory positions.

Draft regulatory guides have not received complete staff review; therefore, they do not represent official NRC staff positions.

This draft regulatory guide contains information collections contained in this regulatory guidethat are covered by the requirements of 10 CFR Part 70, which were approved by the Office of Management and Budget, approval (OMB) approved under OMB control number 3150-0009, and 10 CFR Part 76. The NRC may notneither conduct or sponsornor sponsor, and a person is not required to respond to, a collection of information unless itan information collection request or requirement unless the requesting document displays a currently valid OMB control number.

B. Discussion The American Nuclear Society's Standards Subcommittee 8, "Operations with Fissionable Materials Outside Reactors" (ANS-8), has developed national standards for the prevention and mitigation of However, this draft regulatory guide contains additional information collections that are covered by the requirements of 10 CFR 76.8, which apply to a wholly-owned instrumentality of the United States and affect fewer than ten respondents. As a result, OMB clearance is not required pursuant to the Paperwork Reduction Act (44 U.S.C. 3501, et seq.).

B. DISCUSSION The NRC initially issued Regulatory Guide 3.71 in 1998 to provide guidance concerning procedures that the staff considered acceptable for complying with the agencys regulatory requirements in 10 CFR 70.20, 70.22 and 76.87. Toward that end, the original guide endorsed specific safety standards that ANS-8 developed to provide guidance, criteria, and best practices for use in preventing and mitigating criticality accidents duringduring operations that involve handling, processing, storing, andand/or transporting special nuclear materials at fuels and material facilities. These national standards have beenThe original guide also took exceptions to certain portions of individual ANSI/ANS-8 standards. In addition, the original guide consolidated and replaced a number of earlier NRC regulatory guides, thereby providing all of the relevant guidance in a single document.

The ANS-8 standards endorsed in Regulatory Guide 3.71 were approved by the American Nuclear SocietySocietys Consensus Committee N16 on Nuclear Criticality Safety and by, as well as the American National Standards Institute (ANSI). Nonetheless, each ANSI/ANS-8 standard is reviewed every 5 years by a working group of expert practitioners in the area so that it can be revised, reaffirmed, or withdrawn, as appropriate to reflect the current state of the art.

(This time can be extended to as long as 10 years or more under special circumstances.) New standards are also added when the need arises. Since the timing and issuance of individual standards is independent of the other standards, the list of current standards and their respective dates of issuance is constantly changing.

As a result, since the NRC published Regulatory Guide 3.71 in 1998, several ANSI/ANS-8 nuclear criticality safety standards have been added, reaffirmed, revised, or withdrawn.

Consequently, the NRC staff has decided to update this guide to clarify which standards the agency endorses and to clearly state exceptions to individual standards. This proposed revision does not change any of the guidance provided in Regulatory Guide 3.71; rather, it provides guidance concerning changes that have occurred since the NRC published the original guide in 1998. For completeness, this guide restates the endorsements and exceptions stated in Regulatory Guide 3.71, as applicable, while identifying endorsements of or exceptions to new or modified standards. Since the ANSI/ANS-8 standards are constantly being issued, revised, reaffirmed, or withdrawn, the NRC staff plans to revise this guide on a regular basis.

C. REGULATORY POSITION The ANSI/ANS-8 nuclear criticality safety standards provide guidance and criteria on good practices forprocedures and methodologies that the NRC staff considers generally acceptable for use in preventing and mitigating nuclear criticality accidents. However, use of the ANSI/ANS-8 nuclear criticality safety standards is not a substitute for detailed nuclear criticality safety at fuels and material facilities.

The ANSI/ANS-8 national standards list additional documents as references. The specific applicability or acceptability of these listed documents will be covered separately in otheranalyses for specific operations. In addition, inclusion of a reference to another standard in an endorsed standard does not imply NRC endorsement of the referenced standard.

The NRC staff will follow the requirements denoted in the ANSI/ANS-8 standards. The word shall in an ANSI/ANS-8 standard denotes a requirement; the word should denotes a recommendation; and the word may denotes permission (neither a requirement Nor a recommendation). When a licensee or applicant commits to an ANSI/ANS-8 standard cited in this regulatory guides, where appropriate.

C. Regulatory Position Theguide, the licensee or applicant must perform all operations in accordance with the requirements stated in that standard, but not necessarily with its recommendations. Licensees or applicants may follow the recommendations given in the ANSI/ANS-8 standards, unless an exception is stated in this regulatory guide, otherwise specified in 10 CFR Parts 70 or 76, or addressed by other acceptable methods.

1. ANSI/ANS-8 Nuclear Criticality Standards Endorsed by the NRC The NRC endorses the following ANSI/ANS-8 nuclear criticality safety standard documents describe procedures and recommendations that should be followed to prevent and mitigate nuclear criticality accidents.

ANSI/ANS-8.1-1983 (Reaffirmed in 1988), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors" ANSI/ANS-8.3-1997, "Criticality Accident Alarm System" standards:

  • ANSI/ANS-8.5-1996 (Reaffirmed in 2002), "UseUse of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material"
  • ANSI/ANS-8.6-1983 (Reaffirmed in 19952001), "SafetySafety in Conducting Subcritical Neutron--Multiplication Measurements In Situ"
  • ANSI/ANS-8.7-1975 (Reaffirmed in 1987)7-1998, "Guide for NuclearNuclear Criticality Safety in the Storage of Fissile Materials"

ANSI/ANS-8.9-1987(Reaffirmed in 1995), "Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials"<

ANSI/ANS-8.10-1983 (Reaffirmed in 1988), "Criteria for Nuclear Criticality Safety Controls in Operations With Shielding and Confinement" Materials

  • ANSI/ANS-8.12-1987 (Reaffirmed in 19932002), "NuclearNuclear Criticality Control and Safety of Plutonium-Uranium Fuel Mixtures Outside Reactors" Reactors
  • ANSI/ANS-8.14-2004, Use of Soluble Neutron Absorbers in Nuclear Facilities Outside Reactors
  • ANSI/ANS-8.15-1981 (Reaffirmed in 1995), "NuclearNuclear Criticality Control of Special Actinide Elements" ANSI/ANS-8.17-1984 (Reaffirmed in 1997), "Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors" ANSI/ANS-8. 19-1996, "AdministrativeElements
  • ANSI/ANS-8.21-1995 (Reaffirmed in 2001), "UseUse of Fixed Neutron Absorbers in Nuclear Facilities Outside Reactors"
  • ANSI/ANS-8.22-1997, "NuclearNuclear Criticality Safety Based on Limiting and Controlling Moderators"

The methods described in theResponse

2. ANSI/ANS-8 Nuclear Criticality Standards Endorsed by the NRC with Exceptions The NRC endorses the following ANSI/ANS-8 nuclear criticality safety standards have been applied in a number of specific cases during reviews and selected licensing actions. These methods reflect the latest general NRC approach to nuclear criticality safety in operations involving handling, processing, storing, and transporting special nuclear material at fuels and material facilities.

Most of the ANSI/ANS-8 nuclear criticality safety standards have been endorsed by NRC in other regulatory guides. This regulatory guide consolidates and replaces the following regulatory guides without altering any existing licensing commitments nor introducing any new requirements. These regulatory guides are therefore being withdrawn.

Regulatory Guide 3.1, "Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material" (Revision 2, September 1987)

Regulatory Guide 3.4, "Nuclear, but takes exception to certain sections, as follows:

  • ANSI/ANS-8.1-1998, Nuclear Criticality Safety in Operations with Fissionable Materials at Fuels and Materials Facilities" (Revision 2, March 1986)

Regulatory Guide 3.43, "Nuclear Criticality Safety in the Storage of Fissile Materials" (Revision 1, April 1979)

Regulatory Guide 3.45, "Nuclear Criticality Safety for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials" (Revision 1, April 1989)

Regulatory Guide 3.47, "Nuclear Criticality Control and Safety of Homogeneous Plutonium-Uranium Fuel Mixtures Outside Reactors" (July 1981)

Regulatory Guide 3.57, "Administrative Practices for Nuclear Criticality Safety at Fuels and Materials Facilities" (October 1986)

Regulatory Guide 3.58, "Criticality Safety for Handling, Storing, and Transporting LWR Fuel at Fuels and Materials Facilities" (October 1986)

Regulatory Guide 3.68, "Nuclear Criticality Safety Training" (April 1994)

Regulatory Guide 3.70, "Use of Fixed Neutron Absorbers at Fuels and Materials Facilities" (August 1997)

Regulatory Guide 8.12, "Criticality Accident Alarm Systems" (Revision 2, October 1988)

The ANSI/ANS-8 nuclear criticality safety standards listed above provide procedures and methodology generally acceptable to the NRC staff for the prevention and mitigation of nuclear criticality accidents. However, use of the ANSI/ANS-8 nuclear criticality safety standards is not a substitute for detailed nuclear criticality safety analyses for specific operations. Exceptions to some of the ANSI/ANS-8 nuclear criticality safety standards are as follows.

The guidelines forReactors The guidance on validating calculational methods for nuclear criticality safety contained, as specified in ANSI/ANS-8.1-1983 (Reaffirmed in 1988), "Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors," provide a procedure acceptable to the NRC1-1998, provides a procedure that is acceptable to the NRC staff for establishing the validity and applicability of calculational methods used in assessing nuclear criticality safety. However, it willis not be sufficient to merely to refer to this guidestandard in describing the validation of a method. TRather, a licensee or applicant should provide the details of validation (as stated in Section 4.3.6 of the standard should be provided to demonstrate) to (1) demonstrate the adequacy of the margins of subcriticality relative to the bias and criticality parameters, to(2) demonstrate that the calculations embrace the range of variables to which the method will be applied, and to demonstrateand (3) demonstrate the trends in the bias upon which the licensee or applicant will base the extension of the area of applicability will be based.

. In addition, the details of validation should state computer codes used, operations, recipes for choosing code options (where applicable), cross-section sets, and any numerical parameters necessary to describe the input.

  • ANSI/ANS-8.3-1997 (Reaffirmed in 2003), Criticality Accident Alarm System The guidance on criticality accident alarm systems contained, as specified in ANSI/ANS-8.3- 1997, "Criticality Accident Alarm System3-1997 (reaffirmed in 2003)," is generally acceptable to the NRC staff. An exception is that criticality alarm systems are required by 10 CFR 70.24,"Criticality Criticality Accident Requirements," requires criticality alarm systems in each area in which special nuclear material is handled, used, or stored, while Sectionwhereas Section 4.2.1 of the standard merely requires an evaluation for such areas. Another exception is that each area is required by 10 CFR 70.24 and 10 CFR 76.89, "CriticalityCriticality Accident Requirements," to require that each area must be covered by two detectors, whereas Section 4.5.1 of the standard permits coverage by a single reliable detector. Finally, 10 CFR 70.24 and 10 CFR 76.89 require a monitoring system capable of detecting a nuclear criticality that produces an absorbed dose in soft tissue of 20 rads of combined neutron and gamma radiation at an unshielded distance of 2 meters from the reacting material within 1 minute is required by 10 CFR 70.24 and 10 CFR 76.89.
  • ANSI/ANS-8.10-1983 (Reaffirmed in 1999), Criteria for Nuclear Criticality Safety Controls in Operations With Shielding and Confinement The guidance on using shielding and confinement as a nuclear criticality safety control contained, as specified in ANSI/ANS-8.10-1983 (Reaffirmed in 1988), "Criteria for Nuclear Criticality Safety Controls in Operations with Shielding and Confinement,"reaffirmed in 1999), is generally accepted by the NRC staff. An exception to Section 4.2.1 of the standard is the assumption that the radiation source strengths and releases from a nuclear criticality accident are assumed to beise from an excursion occurring in an unfavorable geometry containing a solution of 400 g/lL of uranium enriched in U--235.

The excursion produces an initial burst of 1E+18 1x1018 fissions in 0.5 second, followed successively at 10 -minute intervals by 47 bursts of 1.9E+ 179x1017 fissions, for a total of 1E+191x1019 fissions in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The excursion is assumed to be terminated by evaporation of 100 liters of the solution. ALicensees and applicants may use a less

-conservative nuclear criticality accident condition may be used if detailed analyses of credible nuclear criticality accidents are performed and shown to be applicable to the conditions being evaluated.

  • ANSI/ANS-8.17-1984 (Reaffirmed in 1997), Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors The general safety criteria and criteria to establish subcriticality contained in ANSI/ANS-8.

17-1997, "Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors,", as specified in ANSI/ANS-8.17-1984 (reaffirmed in 1997),

provide guidance that is acceptable to the NRC staff for preventing nuclear criticality accidents in handling, storing, and transporting fuel assemblies at fuels and material facilities. The only exception is that licensees and applicants may take credit for fuel burnup may be taken only when the amount of burnup is confirmed by physical measurements that are appropriate for each type of fuel assembly in the environment in which it is to be stored.

The NRC staff will follow the requirements as denoted in the ANSI/ANS-8 national standards. The word "shall" in a standard denotes a requirement, the word "should" denotes a recommendation, and the word "may" denotes permission, neither a requirement nor a recommendation. When an applicant or licensee commits to the ANSI/ANS-8 national standards cited in this regulatory guide, all operations must be performed in accordance with the requirements stated in the national standards but not necessarily with its recommendations.

Recommendations given in the ANSI/ ANS-8 national standards may be followed unless an exception is stated in this regulatory guide or otherwise specified in 10 CFR Part 70 or Part 76, or addressed by other acceptable methods.

D. Implementation3. ANSI/ANS-8 Nuclear Criticality Standards Withdrawn by the NRC The NRC has withdrawn its endorsement of ANSI/ANS-8.9-1987 (Reaffirmed in 1995),

Nuclear Criticality Safety Criteria for Steel-Pipe Intersections Containing Aqueous Solutions of Fissile Materials. This standard, which was listed in Regulatory Guide 3.71-1998, has subsequently been withdrawn by ANS (i.e., it is a historical standard). Although the NRC has withdrawn its endorsement of the standard, there is nothing technically wrong with this standard and it would be acceptable for licensees and applicants to use it.

D. IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this draft regulatory guide.

Except in those cases in which No backfitting is intended or approved in connection with the issuance of this guide.

The NRC has issued this draft guide to encourage public participation in its development.

Except when an applicant or licensee proposes or has previously established an acceptable alternative method for complying with the specified portions of the NRC's regulations, the methods to be described in thisthe active guide will reflect public comments and will be used in the evaluation of in evaluating submittals in connection with license applications submitted under 10 CFR Part 70, "DomesticDomestic Licensing of Special Nuclear Material," and 10 CFR Part 76, "CertificationCertification of Gaseous Diffusion Plants."

E. Regulatory Analysis REGULATORY ANALYSIS AThe NRC published a draft regulatory analysis was published withwhen the agency issued the original draft of this guide when it was published(as DG-3013) for public comment (Task DG-3013, January 1998). No changes were necessary, soin January 1998. Since that time, several ANSI/ANS-8 nuclear criticality safety standards have been added, reaffirmed, revised, or withdrawn. Consequently, the NRC staff has decided to update this guide to clarify which standards the agency endorses and to clearly state exceptions to individual standards.

However, this proposed revision does not change any of the guidance provided in Regulatory Guide 3.71; rather, it provides guidance concerning changes that have occurred since the NRC published the original guide 1998. Consequently, the NRC staff has not prepared a separate regulatory analysis has not been prepared for Rthis revised regulatory Gguide 3.71. A copy of tThe draftoriginal regulatory analysis for DG-3013 is available for inspection or copying (for a fee) in the NRC's Public Document Room at 2120 L Street NW.(PDR), which is located at 11555 Rockville Pike, Rockville, Maryland. The PDRs mailing address is USNRC PDR, Washington, DC, under Task DG-3013.

20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4205, by fax at (301) 415-3548; and by email to PDR@nrc.gov.

BACKFIT ANALYSIS This proposed revision of Regulatory Guide 3.71 does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

APPENDIX C TO THE COMMITTEE TO REVIEW GENERIC REQUIREMENTS (CRGR) CHARTER (I) The proposed generic requirement or staff position as it is proposed to be issued for public comments.

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory position in this area. It has not received staff review or approval and does not represent an official NRC staff position.

Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory analysis or value/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Comments may be submitted electronically through the NRCs interactive rulemaking Web page at http://www.nrc.gov/what-we-do/regulatory/rulemaking.html. Copies of comments received may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. Comments will be most helpful if received by April 11, 2005.

(ii) Draft papers or other documents supporting the requirements or staff positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any Committee member may request the CRGR staff to obtain a copy of any reference material for his or her use.)

There are none. This proposed RG neither increases, reduces, nor modifies existing requirements or staff positions. Rather, it updates the NRCs current guidance to reflect changes to the standards since the previous RG was published.

(iii) Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would modify requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirements or staff positions.

This proposed RG neither increases, reduces, nor modifies existing requirements or staff positions. Rather, it updates the NRCs current guidance to reflect changes to the standards since the previous RG was published.

(iv) The proposed method of implementation and resource implications, along with the concurrence (and any comments) of OGC on the method proposed, and the concurrence of all affected program offices or an explanation of any non-concurrences.

The method of implementation will be the proposed RG. All affected program offices have concurred herein. The NRCs Office of the General Counsel (OGC) has no legal objection to the proposed RG.

(v) Regulatory analyses generally conforming to the directives and guidance of NUREG/BR-0058 and NUREG/BR-0184, as applicable. (This does not apply for backfits that ensure compliance or ensure, define, or re-define adequate protection.

For power reactors, a documented evaluation is required as discussed under item (ix) of this Appendix.)

The NRC published a draft regulatory analysis when the agency issued the original draft of this guide (as DG-3013) for public comment in January 1998. Since that time, several ANSI/ANS-8 nuclear criticality safety standards have been added, reaffirmed, revised, or withdrawn. Consequently, the NRC staff has decided to update this guide to clarify which standards the agency endorses and to clearly state exceptions to individual standards.

However, this proposed revision does not change any of the guidance provided in Regulatory Guide 3.71; rather, it provides guidance concerning changes that have occurred since the NRC published the original guide 1998. Consequently, the NRC staff has not prepared a separate regulatory analysis for this revised regulatory guide. The original regulatory analysis for DG-3013 is available for inspection or copying (for a fee) in the NRCs Public Document Room (PDR).

(vi) Identification of the category of power reactors to which the generic requirement or staff position is to apply (that is, whether it is only applicable to future plants, operating plants, all pressurized water reactors (PWRs), all boiling water reactors (BWRs), specific nuclear steam supply system (NSSS) vendor types, plants of specific vintage, gaseous diffusion plants (GDPs), etc.).

The proposed revision to the RG applies to fuels and materials facilities, specifically to those facilities licensed or certified under either 10 CFR Part 70 or 10 CFR Part 76.

(vii) For proposed backfits, other than either the compliance or the adequate protection backfits, a backfit analysis as defined in the Backfit Rule (10 CFR 50.109 for power reactors and 10 CFR 76.76 for the GDPs) should be performed. The backfit analysis shall include, for each category of nuclear power reactor or nuclear materials facility or activity, an evaluation which demonstrates how the proposed action should be prioritized and scheduled in light of other ongoing regulatory activities. The backfit analysis shall document for consideration pertinent information available concerning any of the following factors, as appropriate, and any other information, which is relevant and material to the proposed action:

(a) Statement of the specific objectives that the proposed action is intended to achieve; (b) General description of the activity that the licensee or applicant would be required to perform in order to complete the action; (c) Potential change in the risk to the public from the accidental offsite release of radioactive material; (d) Potential impact on radiological exposure of facility employees and other onsite workers; (e) Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay; (f) The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing regulatory requirements and staff positions; (g) The estimated resource burden on the NRC associated with the proposed action and the availability of such resources; (h) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action; (I) Whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis; (j) For both rulemaking actions and proposed generic correspondence, staff evaluation of comments received as a result of the notice and comment process; (k) How the action should be prioritized and scheduled in light of other ongoing regulatory activities. The following information may be appropriate in this regard:

1. The proposed priority or schedule,
2. A summary of the current backlog of existing requirements awaiting implementation,
3. An assessment of whether implementation of existing requirements should be deferred as a result, and
4. Any other information that may be considered appropriate with regard to priority, schedule, or cumulative impact. For example, could implementation be delayed pending public comment?

This item is not applicable because the proposed revision of the RG does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

(viii) For each proposed backfit analyzed pursuant to 10 CFR 50.109(a)(2),

10 CFR 72.62(c), or 10 CFR 76.76(a)(3), (i.e., for backfits other than either adequate protection backfits or compliance backfits), the proposing office director's determination, together with the rationale for the determination based on the consideration of the previous paragraphs (I) through (vii), that (a) a substantial increase in the overall protection of public health and safety or the common defense and security will be derived from the proposal; and (b) the direct and indirect costs of implementation for the facilities affected are justified in view of this increased protection.

This item is not applicable because the proposed revision of the RG does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

(ix) For adequate protection or compliance backfits affecting power reactors, evaluated pursuant to 10 CFR 50.109(a)(4) (or analogous provisions in 10 CFR 72.62 or 10 CFR 76.76, as appropriate),

(i) A documented evaluation consisting of:

(1) the objectives of the modification (2) the reasons for the modification (3) if the compliance exception is invoked, (A) the requirements (e.g., Commission regulation, license condition, order) or written licensee commitments, for which compliance is sought.

(B) an assessment of risk/safety implications of not requiring licensees to immediately restore compliance, and the basis for determination that a reasonable concession could be allowed to defer restoration of compliance at a later time (e.g., next refueling outage).

(C) demonstrated consideration of other possible alternatives and rationale for rejecting them in favor of compliance backfitting.

(D) evaluation from cost-benefit considerations (not a full-blown regulatory analysis) and a rationale for compliance exception.

(4) If the adequate protection exception is invoked, the basis for concluding that the matter to be addressed involves adequate protection, and why current requirements (e.g., Commission regulation, license condition, order) or written licensee commitments do not provide adequate protection.

(b) In addition, for actions that were immediately effective (and therefore issued without prior CRGR review as discussed in Section III of the CRGR Charter),

the evaluation shall document the safety significance and appropriateness of the action taken and (if applicable) consideration of how costs contributed to selecting the solution among various acceptable alternatives.

This item is not applicable because the proposed revision of the RG does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

(x) For each request for information from power reactor licensees under 10 CFR 50.54(f), which is for purposes other than to verify compliance with the facility's licensing basis, an evaluation that includes at least the following elements:

(a) A problem statement that describes the need for the information in terms of potential safety benefit.

(b) The licensee actions required and the cost to develop a response to the information request.

(c) An anticipated schedule for NRC use of the information.

(d) A statement affirming that the request does not impose new requirements on the licensee, other than submittal of the requested the information.

(e) The proposing office director's determination that the burden to be imposed on the respondents is justified in view of the potential safety significance of the issue to be addressed in the requested information.

Under the provisions of 10 CFR 50.54(f), unless the request for information is for the purpose of verifying compliance with the licensing basis of a facility, the EDO shall approve the staff's justification. Additional guidance for preparing this evaluation is provided in Section 5.4 of NUREG/BR-0058, Revision 2.

Include an analogous evaluation addressing items (a) through (e) for each information request directed to the licensees of the selected nuclear materials facilities or referred to in Section III of the CRGR Charter.

This proposed RG neither increases, reduces, nor modifies existing requirements or staff positions. Rather, it updates the NRCs current guidance to reflect changes to the standards since the previous RG was published.

(xi) For each proposed power reactor backfit analyzed pursuant to 10 CFR 50.109(a)(2)

(i.e., backfits other than either adequate protection or compliance backfits), an assessment of how the proposed action relates to the Commission's Safety Goal Policy Statement.

This item is not applicable because the proposed revision of the RG does not require a backfit analysis, as described in 10 CFR 70.76(b) and 10 CFR 76.76(b), because it does not impose a new or amended provision in the Commissions rules or a regulatory staff position interpreting the Commissions rules that is either new or different from a previous applicable staff position. In addition, this regulatory guide does not require modification or addition to structures, systems, components, or design of a facility or the procedures or organization required to design, construct, or operate a facility. Rather, a licensee or applicant is free to select a preferred method for achieving compliance with a license or the rules or orders of the Commission, as described in 10 CFR Parts 70 and 76.

Retrieved from "https://kanterella.com/w/index.php?title=ML050610258&oldid=2595626"