ML17304A064

Comment (4) of Nima Ashkeboussi Regarding Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores

ML17304A064
Person / Time
Site:	Nuclear Energy Institute
Issue date:	10/20/2017
From:	Ashkeboussi N Nuclear Energy Institute
To:	Cindy Bladey Rules, Announcements, and Directives Branch
References
82FR40173 00004, DG-3053, NRC-2017-0183
Download: ML17304A064 (7)

Text

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PUBLIC SUBMISSION As of: 10/25/17 4:37 PM Received: October 20, 2017 Status: Pending Post Tracking No. lkl-8zbv-fch7 Page 1of1 Comments Due: October 23, 2017 Submission Type: Web

. Docket: NRC-2017-0183 Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores Comment On: NRC-2017-0183-0001 Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores; Draft Regufatory Guide for Comment Document: NRC-2017-0183-DRAFT-0005 Comment on FR Doc # 2017-17934

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Submitter Information Name: Nima Ashkeboussi Submitter's ~epresentative: Lana Dargan Organization: Nuclear Energy Institute.

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General Comment (fcJ.rJ< 4?J/,;/3 See attached file(s)

© Attachments 10-20-17 _NRC_NEI_Industry Comments on NRC Draft Regulatory Guide DG-3053 Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores + Attachment SUNSI Review Complete Template = ADM - 013 E-RIDS= ADM-03

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https://www.fdms.gov/fdms/getc6ntent?objectld=0900006482bed668&format=xml&showorig=false 10/25/2017

NIMA ASHKEBOUSSI Director, Fuel Cycle Programs 1201 F Street, NW, Suite 1100 Washington, DC 20004 P: 202.739.8022 nxa@nei.org nei.org October 20, 2017 Ms. Cindy Bladey Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Submitted vfa Regulations.gov NUCLEAR ENERGY INSTITUTE

Subject:

Industry Comments on NRC Draft Regulatory Guide DG-3053, "Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores" (Docket ID:,NRC-2017-'0183)

Project Number: 689

Dear Ms. Bladey:

The Nuclear Energy Institute (NEI) 1, on behalf of our fuel cycle facility members, appreciates the opportunity to provide industry comments on DG-3053, "Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores." We also appreciated staff's update on this draft regulatory guide (DG) during the public meeting on September 26, 2017. We are pleased to provide our written comments in this letter and the attachment.

We understand that the purpose of this DG (which will be issued as Revision 3 to RG 3.71) is to endorse several American National Standards Institute/American Nuclear Society (ANSI/ANS)-8 nuclear criticality safety standards that tiave been added, reaffirmed, or revised (some containing certain exceptions and clarifications), and make other changes that have occurred since the last revision of RG 3.71 in 20102* The revision would also endorse International Organization for Standardization (ISO) Standard 7753:1987, "Nuclear Energy - Performance and Testing Requirements for Criticality Detection and Alarm Systems."

Two of the guides (ANSI/ANS-8.10-1983 and ANSI/ANS-8.23-2007) that were previously endorsed in RG 3.71, Rev. 2 are now endorsed with added exceptions. These exceptions appear to be an imposition of either new or different regulatory staff positions. The problem is compounded with the changes made to NUREG-1520, Rev. 2 "Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility." We encourage NRC staff to review Section 5.4.3. Li' (Use of Industry-Standards) of NUREG-1520, _

1 The Nuclear Energy Institute (NEI) is the organization responsible for establishing unified industry policy on matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEI's members include.all entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect/engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations and entities involved in the nuclear energy industry.

2 NEI submitted a comment letter to the NRC on DG-3030 (Revision 2) dated September 23, 2010.

NUCLEAR. CLEAN AIR ENERGY

Ms. Cindy Bladey October.20, 2017 Page 2 I

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in light of the changes made to DG-3053. Furthermore, we would also like to refer the NRC to NEI's comr:nent letter3 and attachment on Draft NUREG-1520, Rev. 2, which addresses several relevant items in Chapter 5, Nuclear Criticality Safety.

Furthermore, under Section D (Implementation), the last paragraph on~Page 11 states: "If an existing licensee voluntarily seeks a license amendment or change... then the staff may request that the licensee either follow the guidance in this regulatory guide or provide an equivalent alternative process that demonstrates compliance with the. underlying NRC regulatory requirements. This is not considered backfitting... " It would be helpful to gain additional context from NRC staff as to what level of justification would be required, in outlining "an equivalent alternative process that demon~rates compliance."

. *Additionally, it is industry's understanding that, pursuant to the NRCs current position regarding so-called

. '~forward,fitting/ the application of Revision 3 would fall* outsi,de th.e definition of backfitting only in situations where it: (1) relates directly to a license~'s voluntary request,* and (2) *is an ess.ential

• consideration in the NRC staff's determination.of the. acce:pfability of the licensee's.voh.1ntary request<t Iri.

qther.words, it is.our understanding that.i..m~el~ted licensing,actioris*or arn~ndmentsshould.n'ot'garne~-rtew commitments regarding th~ new referenced standards :in :Revision** 31. rior sl\\oulq it garn~f: 'subsequent **RAis related to criticality safety, as discussed during the September 2£?;" ?Ol7. puf?lic r-neetirig.*. :

The last sentence on Page 11, as referenced above states: "This is not *considered backfitting as defined in 10 CFR 70.76 or 10 CFR 72.62." This paragraph notes that license. amendments or:changes would not cons~itute a backfit, but the DG is silent on license renewals. This omission leaves. considerable uncertainty as to NRC's expectations for a license renewal. For example, an attempt to endorse a particular standard's.

current version may be problematic during a Part 70 license renewal, in which several decades are likely to pass between renewals. This places an unnecessary burden on the licensee to perform significant gap analyses, demonstrate or justify exceptions to revised standards, or potentially make significant licensed program upgrades.

The applicability of DG-3053 is also proposed to be expanded to include fuel cycle transportation (10 CFR Part 71 - Packaging and Transportation of Radioactive Material). However, this proposed scope expansion could impart unintended burdens on fissile package certificate holders when a renewal is pursued, without a clear regulatory basis, articulated benefit, or safety concern. This too appears to be an imposition of a regulatory.staff position that is either new or different from a previous NRC staff position. In addition, the

. fis~ile package ~equirements in 10 CFR 71.55.and R:G 7.9, "Standard Forl'T1at and Content of Part 71.

• :packages fcir:*Radioactive: Material" have not changed* and there is no equivalent ~'Use of Industry Standard" *

~tatement (similar to Section 5.4.3.1.1 in NUREG-1520, Rev. 2)JorPart71 fissile:pa.ckage reviews, as* there is for the reviews of applications to construct, modify o~ operate Part lQ nuclear fuel cycle. facilities. As a.

result, it is unclear ho.vii NRC plans to expand the* applicabilicy of DG~ 305lto Part. 71 *reviews, unless through..

inappropriate use of the RAI process..

3 NEI letter to Ms. Cindy Bladey, Docket NRC-2012-0220, dated November 3, 2014.

4 Letter from S.G. Burns (NRC) to E.C. Ginsberg (NEI), July 14, 2010, at FN 2.

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Ms. Cindy Bladey October 20, 2017 Page 3 Thank you again for the opportunity to comment on DG-3053 and the related public meeting on September 26, 2017. We look forward to seeing how these comments are addressed in the final guidance. If you have any questions, please contact me or Hilary Lane (202-739-8148; hml@nei.org).

Sincerely, Jl?.,

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Nima Ashkeboussi Attachment c:

• craig Erlanger, NMSS, NRC.

Margie Kotzalas, NMSS, NRC Christopher Tripp, NMSS, NRC

ATTACHMENT Industry Comments on Draft Regulatory Guide DG-3053, "Nuclear Criticality Safety Standards for Nuclear Materials Outside Reactor Cores"

1) General Comment:

a. 'This DG endorses and aims to harmonize both ANSI/ANS-8 standards and, for the first time, standards developed by ISO,* per* routine reviews by th<;! ANSI and ISO.

• standard working group.committees; Industry believes that *harmonization *of these standards through guidance document,s,is misplaced as an NRC goal, and consideration should be given to the removal ofthis.ISO standard.. H-owever, should ISO standards remain, DG-3053 should'ad9ress the acceptability of a licensee using a combin~tion of both stand~rds. For instance, clarify i.f it is acceptable for a licensee_ to endorse all ANSI/ANS-8 standards exceptfor ANSI/ANS-8.3-1997, "Criticality Accident Alarm System," and alternatively endorse ISO Standard 7753:1987, "Nuclear Energy-Performance and Testing Requirements for Criticality Detection and Alarm Systems." Otherwise, without further clarification, this DG endorses two different standards for criticality accident alarm systems (2.b ANSI/ANS--8.3 and 2.h ISO 7753:1987), which is not necessary and will certainly cause confusion as to which takes precedent.

b. As stated in the Federal Register Notice (FRN), "the proposed revision would provide methods that are acceptable to the NRC staff... " and the RG provides one method by which a licensee may demonstrate compliance, and alternative methods (to demonstrate compliance) with appropriate justification may be deemed acceptable by the NRC. DG-3053 incorporates several "shall" statements, understood to denote

• a requirement of the standard, if and only *if the licensee.commits to that particular standard. Alternatively, if the standard is not-accepted by the licensee, and the licensee choses an alternate method, "shall" statements would not apply.

2) Specific Comments:

a. As noted during the September 261 2017 public meeting regarding DG-3053~ the -

ANSI/ ANS standards are continually being revised and re-affirmed by several working groups approximately every 5 years. However, this is problematic ih tha.t the effective dates of issuance are frequently changing.and endorsing the current versions of the 18 standards in DG-3053 is not practical. :For example; several on the list provided in DG-3053 (issued August 2017), Section C.1 (Page 6) are already outdated and have the following errors:

i. The referenced citations for ANSI/ANS-8.12 are incorrect-it is now R2016.

ii. The referenced citations forANSI/ANS-8.14 are incorrect-it is now R2016.

iii. The referenced citations for ANSI/ANS-8.17 are incorrect-it is now R2014.

iv. The referenced citations for ANSI/ANS-8.20 are incorrect-it is now_ R2015.

v. The referenced citations for ANSI/ANS-8.22 are incorrect-it is now R2016.

vi. The referenced citations for ANSI/ANS-8.26 are,incorrect-it is now R2016.

1

Furthermore, NRC stated that RG 3.71 is ultimately a "living document," that will be revised on a "regular basis." While we applaud the NRC for periodically updating its guidance documents to ensure that endorsements of standards remain current, we recommend revisions occur at a frequency of a minimum of every 5 years, in recognizing the Cumulative Effects ofRegulation on the fuel cycle facility community.

• Furthermore, a minimum of 5 years between revisions will ensure a more comprehensive review, versus piecemeal changes. We do note.that the.last'revision

.. was 7 years ago, which we -feel is an appropriate timeframe between revisions..

b. Section C.2.b, paragraph 2 (Page 7) states: "Section 4.2.2 of the standard states

• that a criticality alarm system is not required in ar:eas:.whete personnelwould be subject to. an excessive radiation dose.~,. We believe, there *js *an.error iri this sentence.. _

and the correct sentence would read "... where*p~rsonn~! would not be subject to an.

excessive radiation dose."

-c.. The ANSI/ANS-8.3 exceptions endorsement in DG-3053 added the sentence: "A clarification is that 10 CFR 70.24 requires placement of detectors in areas where threshold quantities of special nuclear material are present, but that audible or visual alarms may be located in areas where immediate evacuation is determined to be necessary based on the potential for an excessive dose.". Excessive radiation dose is defined in ANSI/ANS-8.3-1997 R2012 as "any dose to personnel corresponding to an absorbed dose from neutrons and gamma rays equal to or greater than 0.12 Gy (12 rad) in free air." However, 10 CFR 70.24(a)(1) requires a monitoring system capable of detecting a 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 one. minute. The added sentence in the exceptions end_orsement of ANSI/ANS-~.3 :is.notcl~ar Clr:Jd appears to change the. regulatory

- requirementfrom detecting 20 rads within one. minute to a systeni capable of detecting 12 rad. Industry recommends using the' current Reg Guide ~.71, Revision 2 monitoring system exception endorsements forANSI/ANS-8.3..

d. The ~xceptions endorsement of Section_ ci.c~. (Page 8), is, not :cle~r a~d a*ppear to be.

misaligned with the regulatory definitions in-10 CFR 70;61(c): The ANSI/ANS standard states shielding and confinement should be applied such that the total effective dose to any individual outside the shielded and confined area will not exceed 10 rem and that the total effective dose to an individual outside the restricted area will not exceed 0.5 rem. The DG then states that the dose limits in ANSI/ANS-8.10 are more conservative than the performance requirements in 10 CFR 70.61 and are applicable (i.e., 10 rem versus 25 rem for workers outside the shielded and confined area or 0.5 rem versus 5. rem to an individual outside the controlled area). As currently written, this DG endorsement appears to lower the regulatory definition of an intermediate consequence event described in 10 CFR 70.61(c) and NUREG-1520.

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e. The exceptions endorsement of Section C.2.d (Page 8) states: "... 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 iri which it istb be stored." The following new sen.tence states: "Alternatively, licensees an.d 'applicantsniay perform a. mislo~q analysis, along with additiohaladministrative loading.proc;:edures to.reduc~the *

' likelihoqd of a !Tlisload, in,lieu.of a *qua11titative measurement~'... '..

section4.10.of'ANSI/ANs-8:11a11ows both-physicai':*~easure~ent5:pr:appropriate.. :

• analysis and ve~il'ication to determ!n'e the appr9priate f~el burnt:ip credit as follows:.. '

"In performing: tbe criticality safety evaluatibn, ~~e f~el: chqra~~ri~ti~~* ~~~.~., * ~~teri~1 * *.

compositions, geometry, temperature* that.affect reactivity shall be chosen from the range of credible values such that the maximum neutron multiplication factor of the system is obtained. Credit may be taken for fu~I burnup by establishing a maximum fuel unit reactivity and assuring that each fuel unit has a reactivity no greater than the maximum established reactivity. Assurance that the reactivity limit is not exceeded may be provided by (1) a measurement that can be related to the reactivity or (2) an analysis and verification of the exposure history of each fuel unit."

NRC currently accepts analytical methods to determine burnup credit in criticality safety analyse~ in the following documents:

NUREG/CR 6801 (dated March*2003), "Recommendations for Addressing AXial Burnup in PWR Burnup Credit Analysis" Nl,JREG/CR 7109 (dated April 201?), "An Approach for Validating Actinide and

• fissic>'nProduct Burnup*Credit Criticality Safety:Analyses-Criticality

• Predictions" *

• . NEI 12~16, Rev 1 (dated.April 2014)*, "Guidan~e for Performing Criticality,

. :~Analyses of F_uel Storage_ at Light *WaterReactor'~awer Plarit:S".

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The DG-3053 endorsement of,this. partiqlla~-~tlindar(j.d~es* riot p'rovjd~*the:'fle~ipility*

to_ use currently acceptable practices to use: a burnup" ar:ialysis. ni'ethocl and' appears' -

to be narrowly focused on loading spent.fuel casks~

f.

Section C.2.f (Page 8) states: "Section 4.1 of the standard requires that verification..

of the computer code system be completed prior to validation. A clarification is that provisions should be made for routine (e.g., annual) reverification, and not merely before validation." It is not clear why annual reverification is required if there are no changes to the computer code system (i.e., method, hardware, software, or operating system). Industry suggests either removing this new statement, or that reverification should be performed only when there is a change to the configuration of the computer code system (for example, when a patch is installed to fix a glitch in the code, reverification should be performed). Performing reverification annually provides no added value when the software hasn't changed, and appears to be an imposition of a regulatory staff position that is either new or different from a previous NRC staff position. Currently, the computer code system is verified prior to use on the :production workstation used to* perform said transport calculations.

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