SECR

"Eouq.•UNITED NUCLEAR REGULATORY STATESCOMMISSION

•°'•°WASHINGTON, D.C. 20555-0001

August 27, 2015

:tETARY

COMMISSION VOTING RECORD

DECISION ITEM: SECY-15-0065

TITLE: PROPOSED RULEMAKING: MITIGATION OF BEYONDDESIGN-BASIS EVENTS (RIN 3150-AJ49)

The Commission acted on the subject paper as recorded in the Staff Requirements

Memorandum (SRM) of August 27, 2015.

This Record contains a summary of voting on this matter together with the individual vote

sheets, views and comments of the Commission.

Secretary of the Commission

Enclosures:

1. Voting Summary

2. Commissioner Vote Sheets

cc: Chairman Burns

Commissioner Svinicki

Commissioner Ostendorff

Commissioner Baran

OGC

EDO

PDR
VOTING SUMMARY - SECY-15-0065

RECORDED VOTES

CHRM. BURNS

COMR. SVINICKI

COMR. OSTENDORFE

COMR. BARAN

NOT

APRVD DISAPRVD ABSTAIN PARTICIP COMMENTS

X X X

X X x

X X X

DATE

7/28/15

7/24/15

7/10/15

X X 8/4/15
1

F

NOTATION VOTE

RESPONSE SHEET

TO: Annette Vietti-Cook, Secretary

FROM: Chairman Burns

SUBJECT: SECY-15-0065: PROPOSED RULEMAKING:

MITIGATION OF BEYOND-DESIGN-BASIS EVENTS

(RIN 3150-AJ49)

Approved X Disapproved X Abstain ___

Not Participating ___

COMMENTS: Below ____Attached X None ___

Entered in "STARS" Ys Yes XNX No
Chairman Burns Comments on SECY-15-0065 Proposed Rulemaking: Mitigation of Beyond-Design-Basis Events

The proposed rulemaking presented in SECY-15-0065 represents the culmination of several

years of hard work by the NRC staff to address some of the most significant lessons from the

2011 accident at the Fukushima Dal-ichi nuclear power plant in Japan. I commend the staff for

the outstanding job they did to bring together the many complex pieces that make up the

Mitigation of Beyond-Design-Basis Events rulemaking.

I approve publication of the proposed rule notice in the Federal/Register, subject to the attached

edits and to removal of provisions requiring Severe Accident Management Guidelines (SAMGs)

and provisions for separate design requirements for new reactor applicants.

SAMGs

As the staff pointed out in SECY-1 5-0065, SAMGs are currently in place at all operating power

reactor sites as a voluntary industry initiative. Efforts to develop SAMGs began in the

mid-to-late 1980s when the NRC and the nuclear power industry identified a need to consider

plant conditions that could lead to a severe accident. These efforts led to the nuclear industry

voluntarily initiating a coordinated program on accident management in 1990. In 1994, the NRC

accepted the industry's approach for the management of severe accidents, including licensee

regulatory commitments to implement plant-specific SAMGs. The NRC assessed the ongoing

implementation of SAMGs in the 1997-1998 time frame and concluded that the results of the

voluntary initiative achieved the NRC's overall objectives established for accident management.

Following the events at Fukushima Dai-ichi in 2011, the NRC again inspected the

implementation, training, and maintenance of licensee SAMGs. The inspectors observed

inconsistent implementation of SAMGs and attributed it to the voluntary nature of this initiative.

Based in part on these inspection findings, the NRC's Near-Term Task Force (NTTF)

recommended in their July 2011 report that the NRC require licensees to integrate onsite

emergency response capabilities, including SAMGs. However, the NTTF also noted, "The

results of the SAMG inspection do not indicate, nor does the Task Force conclude that, the

SAMGs would not have been effective if needed."

The staff cited two principal factors that caused it to recommend that the Commission

reconsider its view of imposing SAMG requirements: (1) a greater appreciation of external

event uncertainty and the consequences that can occur as a result of an inadequate facility

design basis for external events and (2) that the SAMGs voluntary initiative was not entirely

successful.

While I agree that the accident at Fukushima has given us a greater appreciation of external

event uncertainty and consequences, it has also spurned us to action to take significant steps to

address beyond-design-basis external events. The nuclear industry's efforts to address the

requirements imposed by NRC Order EA-1 2-049, 'Order to Modifying Licenses with Regard to

Requirements for Mitigation Strategies for Beyond-Design-Basis External Events," (Mitigation

Strategies Order) are substantially improving the ability of the operating fleet to withstand

external events.

While there is no doubt that some licensees were not giving adequate attention to maintaining

SAMGs when the last inspections were performed in 2011, those discrepancies have since

2
been rectified. In addition, in the four years since the NTTF issued its report, the industry has undertaken significant efforts to update its guidelines for SAMGs to incorporate lessons learned

from the Fukushima Dai-ichi accident and include experience gained since the 1990s. The

industry has also developed guidance for plant-specific SAMG development, implementation,

maintenance, training, command and control, and drills and exercises. Therefore, I do not view

the assumption that today's plant SAMGs are in a similar state to those of 2011 as valid. Also

noteworthy is the industry's commitment, as explained in a presentation to the NRC's Advisory

Committee on Reactor Safeguards (ACRS) on April 9, 2015, and reiterated in a May 11, 2015,

letter from the Nuclear Energy Institute 1, the industry supports submittal of a docketed

commitment by each site that would ensure: (1) maintenance of SAMG strategies, (2)

integration with Emergency Operating Procedures and other guideline sets, (3) timely

incorporation of Owners Group revisions to SAMGs, and (4) establishment of configuration

controls. After careful review of the staff's draft rulemaking package and numerous meetings on

this topic, the ACRS noted:

At issue is not whether SAMGs should be developed, implemented, trained upon,

and exercised, but whether these activities are maintained and monitored

through regulatory requirements or by a strengthened voluntary initiative. Given

the extremely low likelihood that an event will lead to the use of SAMGs,

regulatory requirements should not impose unnecessary burden or divert

attention from more important safety objectives.

The staff concluded that the costs of imposing SAMGs by rule are justified in view of the

quantitative and qualitative benefits of SAMGs, mainly because SAMGs provide "substantial

defense-in-depth benefits associated with their use following the onset of core damage." For

the reasons cited above, I agree with Commissioner Ostendorff that the additional defense-indepth that would be gained from making the SAMGs a regulatory requirement rather than a

voluntary initiative does not provide a sufficient basis to support this provision of the proposed

rule. To be clear, I do not take issue with the staffs use of the defense-in-depth argument, per

se, but rather, with the significant weight they gave to a SAMG requirement (over that of a

voluntary initiative) for increasing defense in depth. Nevertheless, given the past issues with

industry implementation of SAMGs, I agree with Commissioner Ostendorff that the staff should

update the Reactor Oversight Process to provide for periodic oversight of industry's

implementation of SAMGs.

New Reactor Desi~qn Requirements

The staff is proposing to require applicants for new nuclear power plant designs to include

design features sufficient to enhance coping durations and minimize reliance on human actions

to maintain or restore key safety functions during specified beyond-design-basis conditions.

The staff states that such capability would maximize reliance on installed design features to

maintain or restore the key safety functions. The staff relies heavily on the Commission's Policy

Statement on the Regulation of Advanced Reactors as a basis for its proposal.

I appreciate the efforts of the staff member who submitted a non-concurrence on this portion of

the proposed rule and believe that he has raised some valid points. However, the main reason

for my disapproval of the new reactor design provisions relates to past Commission guidance on

1 A. R. Pietrangelo, Nuclear Energy Institute, letter to Mark Satorius, U.S. Nuclear Regulatory

Commission, May 11, 2015, on Use of Qualitative Factors in Regulatory Decision Making.

3
this topic. I agree with Commissioner Ostendorff that Commission direction in the SRM for SECY-1 1-01 24, "Recommended Actions to Be Taken without Delay from the Near-Term Task

Force Report," is instructive here. There, the Commission encouraged the staff to craft

recommendations that continue to realize the strengths of a performance-based system as a

guiding principle and noted that approaches should be flexible and able to accommodate a

diverse range of circumstances and conditions. Most pertinent to today's decision is the

following Commission statement:

In consideration of events beyond the design basis, a regulatory approach

founded on performance-based requirements will foster development of the most

effective and efficient, site-specific mitigation strategies, similar to how the

agency approached the approval of licensee response strategies for the "loss of

large area" event under its B.5.b program.

The Commission did not distinguish a separate standard for new reactor designs, and,

therefore, it can be assumed that the Commission meant for the performance-based standard to

apply to new designs as well as to the existing operating fleet. In addition, the staff did not

propose and the Commission did not approve a different standard when addressing mitigation

strategies requirements for any of the three new reactor combined licenses issued since

imposition of the Mitigation Strategies Order. For these reasons, I agree with Commissioner

Ostendorif that a more flexible approach for new reactor applicants is preferred over the

additional design requirements proposed by the staff. The Advanced Reactor Policy Statement

itself provides sufficient encouragement for new reactor designers to seek ways to reduce

reliance on operator actions and provide longer time constants for decision making during

accidents, while retaining the flexibility in implementation options sought by the Commission.

Given the significant revisions to the proposed rulemaking that would be required to address

removal of requirements imposing SAMGS and new reactor design requirements, the revised

proposed rulemaking package should be provided to the Commission for information at least 10

days before submittal to the Office of the Federal Register for publication.

Both Commissioners Ostendorff and Svinicki raised the issue of our path forward on resolution

of the remaining Tier 2 and Tier 3 Fukushima lessons learned actions in their votes on this

paper. I agree with my colleagues that this is an appropriate time to scrutinize the remaining

Tier 2 and Tier 3 actions, given the significant milestone that issuance of this proposed rule will

represent. I support directing the staff to provide the Commission a plan and schedule for

resolving all remaining Tier 2 and Tier 3 actions within 90 days after issuance of the staff

requirements memorandum for SECY-15-0065.

I would like to close by reemphasizing my gratitude for the extraordinary job that the staff has

done to prepare this proposed rulemaking and to seek external stakeholder input and ACRS

review at multiple points along the way. The staff is to be congratulated for its hard work and

thoughtful presentation of the many complex issues embodied in the proposed rulemaking for

the mitigation of beyond-design-basis events.

2015

4
B. NRC Near-Term Task Force.

C. Implementation of the NTTF Recommendations.

D. Consolidation of Regulatory Efforts.

E. Public Involvement.

Ill. Petitions for Rulemaking.

IV. Discussion.

A. Rulemaking Objectives.

B. Rulemaking Scope.

C. Proposed Rule @gm,,,.=•',ti,.

D. Proposed Rule Regulatory Bases.

V. Section-by-Section Analysis.

VI. Regulatory Oversight of Severe Accident Management Guidelines.

VII. Specific Requests for Comments.

VIII. Regulatory Flexibility Certification.

IX. Availability of Regulatory Analysis.

X. Availability of Guidance.

XI. Backfitting and Issue Finality.

XlI. Cumulative Effects of Regulation.

XIII. Plain Writing.

XIV. Environmental Assessment and Proposed Finding of No Significant

Environmental Impact.

XV. Paperwork Reduction Act Statement.

XVI. Criminal Penalties.

XVII. Coordination with NRC Agreement States.

XVIII. Compatibility of Agreement State Regulations.

7
compliance with the proposed requirement to have mitigation strategies and guidelines for

maintaining or restoring core cooling and containment capabilities.

This proposed rule treats the EDMG and SAMG requirements in a manner similar to the

requirements for FSGs. For a licensee who has § 50.82(a)(1) or § 52.110(a) certifications

docketed at the NRC, the lack of fuel in their reactor core and the absence of challenges to the

containment would render unnecessary EDMGs for core cooling and containment capabilities

and SAMGs related to fuel in the reactor vessel. This licensee would not need to comply with

any f~h4"eqireent in hispropsedrulrather, the licensee would be required to comply t.,011neS

with the proposed requirement to have EDMGs or SAMGs as based on the presence of fuel in

the•-etG•.v&e

Once the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certifications, that

licensee would not need to comply with the requirements proposed by this rule that the

equipment relied on for the mitigation strategies include reliable means to remotely monitor

wide-range spent fuel pool levels to support effective prioritization of event mitigation and

recovery actions. These proposed requirements are based on the requirements in Order

EA-12-051. This order requires a reliable means of remotely monitoring wide-range SFP levels

to support effective prioritization of event mitigation and recovery actions in the event of a

beyond-design-basis external event with the potential to challenge both the reactor and SFP.

The NRC has also recinded Order EA-1 2-051 for the Shutdown NPP Group mentioned

previously. These rescissions were based, in part, on the NRC's conclusions that once a

licensee certifies the permanent removal of the fuel from its reactor vessel, the safety of the fuel

in the SEP becomes the primary safety function for site personnel. In the event of a challenge

to the safety of fuel stored in the SEP, decision-makers would not have to prioritize actions and

the focus of the staff would be the SFP condition. Thus, once fuel is permanently removed from

40
2. In § 50.8, paragraph (b) is revised to read as follows:

§ 50.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in §§50.30,

50.33, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54,

50.55, 50.55a, 50.59, 50.60, 50.61, 50.61a, 50.62, 50.63, 50.64, 50.65, 50.66, 50.68, 50.69,

50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, 50.150, 50.155, and

appendices A, B, E, G, H, I, J, K, M, N, 0, Q, R, and S to this part.

3. In §50.34:

a. revise paragraphs (a)(13) and add paragraph (a)(14);

b. revise paragraph (b)(12) and add paragraph (b)(13); and

c. revise paragraph (i).

The revisions and additions read as follows:

§ 50.34 Contents of applications; technical information.

(a) * * *

(13) On or after July 13, 2009, power reactor applicants who apply for a construction

permit shall submit the information required by 10 CFR 50.150(b) as a part of their preliminary

safety analysis report.

(14) On or after [EFFECTIVE DATE], power reactor applicants applying for a

construction permit Rme'-submit a preliminary description of the design features included in the

plant design under § 50.155(d), and an explanation of how those design features comply with

the requirements of § 50.155(d).

146
(b) * * *

(12) On or after July 13, 2009, power reactor applicants who apply for an operating

license which is subject to 10 CFR 50.150(a) shall submit the information required by 10 CFR

50.150(b) as a part of their final safety analysis report.

(13) On or after(EFFECTIVE DATE], power reactor applicants applying for an

operating license mts submit a final description of the design features included in the plant

design under § 50.155(d), and an explanation of how those design features comply with the

requirements of § 50.155(d).

• * * * *

(i) Mitigation of beyond-design-basis events.

Each application for a power reactor operating license under this part must include the

applicant's plans for implementing the requirements of § 50.155 and 10 CER part 50, Appendix

E, Section VII, including a schedule for achieving full compliance with these requirements. The

application must also include a description of:

(1) the integrated response capability required by § 50.155(b);

(2) the equipment upon which the strategies and guidelines required by § 50.155(b)(1)

rely, including the planned locations of the equipment and how the equipment and SSCs meet

the design requirements of § 50.155(c); and

(3) the strategies and guidelines required by § 50.155(b)(2).

• * * * *

4. In § 50.54:

a. remove paragraph (hh)(2);

147
(3) When the NRC has docketed the certifications described in § 50.82(a)(1) or

§ 52.110(a) of this chapter, submitted by a licensee subject to the requirements of this section

and section VII of appendix E to 10 CFR part 50, then that licensee shall comply with the

requirements of §§ 50.155(b)-(f) associated with maintaining or restoring secondary

containment capabilities, if applicable, and spent fuel pool cooling capabilities, but not with

§ 50.155(c)(4) and section VII of appendix E to 10 CFR part 50, for the unit described in the

§ 50.82(a)(1) or § 52.110(a) certifications until the spent fuel pool(s) is empty of all irradiated

fuel.

(i) Holders of operating licenses or combined licenses for which the NRC has docketed

the certifications described in § 50.82(a)(1) or § 52.110(a) of this chapter need not meet the

requirements of this section except for paragraph (b)(2) if the licensee performs and retains an

analysis demonstrating that the decay heat of the fuel in the spent fuel pool removed solely by

heating and boiling of water within the spent fuel pool and the boil-off period provides sufficient

time for the licensee to obtain off-site resources to sustain the spent fuel pool cooling function

indefinitely.

(ii) Dominion Nuclear Connecticut, Inc. (Millstone Power Station Unit 1) is not subject to

the requirements of this section.

(4) The requirements in paragraph (d) of this section apply to applicants for:

(i) Construction permits for nuclear power reactors under this part;

(ii) Operating licenses for nuclear power reactors under this part for which a construction

permit was issued after [EFFECTIVE DATE];

(iii) Standard design certifications under part 52 of this chapter;

(iv) Standard design approvals under part 52 of this chapter;

(v) Combined licenses under part 52 of this chapter that do not reference a standard

design certification, standard design approval, or manufacturing license; and

149
TO:

FROM:

SUBJECT:

NOTATION VOTE

RESPONSE SHEET

Annette Vietti-Cook, Secretary

COMMISSIONER SVINICKI

SECY-15-0065: PROPOSED RULEMAKING:

MITIGATION OF BEYOND-DESIGN-BASIS EVENTS

(RIN 3150-AJ49)

Approved XX In Part

Not Participating___

COMMENTS: Below

Disapproved XX In Part Abstain

Attached XX None

~Z14J15

DAT

Entered Yes •/No in "STARS" YsN
Commissioner Svinicki's Comments on SECY-15-0065

Proposed Rulemaking: Mitigation of Beyond-Design-Basis Events

I approve for publication in the Federal Register the draft proposed rule notice (Enclosure 2 to

SECY-15-0065) subject to the attached edits and any further conforming changes necessary to

align the draft notice with the specific disapprovals delineated in this vote. Specifically, I

disapprove the staff's proposal to impose requirements in the rule for Severe Accident

Management Guidelines (SAMGs), and I disapprove the staff's proposed rule provisions related

to design features for new reactor applicants.

Laying aside these discrete departures from the staff's proposal, it is important to note that the

rulemaking package as a whole embodies the fruits of so many individual staff efforts and,

collectively, many thousands of hours of labor conducted over the past years. The proposed

rule is a culmination of the agency's effort tO respond thoughtfully to the complex challenge of

very low probability but potentially high consequence events by finding that set of measures that

will enhance national resiliency but do so without seeking extremes. I commend these efforts

and all who contributed, and I support fully the preponderance of the proposed rule. It is also

important to note that the bulk of these measures is in place or underway - providing this

enhanced resiliency now, or soon - as this rulemaking to codify the implementation takes place.

With respect to those two elements I disapprove, I cannot support the staff's proposed new

treatment of SAMGs because it is fundamentally at odds with agency policy and practice. It

ignores Commission policy on the function of voluntary industry initiatives, which the

Commission has repeatedly validated as filling a narrow but key role in our regulatory

framework. The staff's underlying regulatory analysis adopts cost benefit analysis techniques

inconsistent with coherent practice and defying longstanding Commission policy (recently

reinforced) that determining regulatory outcomes with a singular reliance on qualitative factors

runs the risk of amounting to little more than "a thumb on the scale." Circumstances justifying

such a thing are extraordinary and their occurrence exceedingly rare. Such circumstances do

not exist here. As the staff notes, available risk information indicates that SAMGs have only a

small safety benefit. The amorphous, non-quantified benefit of "further enhancing defense-indepth" cannot serve as a magic cure-all that sweeps away all deficits in our quantitative cost

benefit analysis, lest we end up where Commissioner Magwood cautioned we would - able to

justify the imposition of absolutely any safety requirement at all.

With regard to the provisions related to design features for new reactor applicants, i agree with

the non-concurring individual who found the proposed requirements arbitrary and ambiguous.

Moreover, I find the "parachuting in" of these proposed new requirements deeply out of process

with respect to agency policy development. It is neither an outgrowth of the Commission's

Policy Statement on the Regulation of Advanced Reactors, as the staff asserts, nor an evolution

of any other post-Fukushima development. I disapprove these proposed provisions and agree

with Commissioner Ostendorff that both operating and new reactors are accommodated fully

and sufficiently within the regulatory framework provided by the proposed Mitigation of BeyondDesign-Basis Events rulemaking.

The draft regulatory analysis for the proposed rule (Enclosure 3) should be revised and

republished in a form to contain only those regulatory provisions supported by the Commission

in its decision on the proposed rule. The statement on page 67 that, after Fukushima, "the NRC

mindset changed" should be struck as inconsistent with agency actions and Commission
2

decisions since the Japan events, which have reaffirmed the efficacy of NRC's longstanding regulatory framework for dispositioning issues arising from those events. The draft

supplemental regulatory analysis addressing design features for new reactors (Enclosure 4)

should be revoked.

Finally, I share Commissioner Ostendorff's observation that the publication of the proposed rule

on Mitigation of Beyond-Design-Basis Events marks a significant milestone in the agency's

regulatory response to the Fukushima event. In light of this, as he notes, it is appropriate to

assess the planned path forward on remaining open items in Tiers 2 and 3. I approve his

proposal that the staff be directed to provide to the Commission, no later than 60 days after the

issuance of the staff requirements memorandum on this paper, a plan and schedule for

resolving all remaining Tier 2 and Tier 3 actions.

Further, by virtue of the evaluation contained in SECY-1 5-0057, "Seventh 6-Month Status

Update on Response to Lessons Learned from Japan's March 11, 2011, Great Tohoku

Earthquake and Subsequent Tsunami," stating that the staff "remains confident that the

[emergency preparedness] EP programs in support of nuclear power plants provide an

adequate level of protection of the public health and safety" and has concluded that calculated

dose estimates "do not appear to call the EPZ or [potassium iodide] KI approach into question,"

these Tier 3 items are closed.
[7590-01 -P]

NUCLEAR REGULATORY COMMISSION

10 CFR Part 50

[Docket Nos. PRM-50-97 and PRM-50-98; NRC-2011-0189 and NRC-2014-0240]

RIN 3150-AJ49

Mitigation of Beyond-Design-Basis Events

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its

regulations that establish regulatory requirements for nuclear power reactor applicants and

licensees to mitigate beyond-design-basis events. The NRC is proposing to make generically

applicable requirements in Commission orders for mitigation of beyond-design-basis events and

for reliable spent fuel pool instrumentation. The rule proposes to establish regulatory

rngrequirements for command and control, drills, training

and change control. The rule would set requirements for enhanced onsite emergency response

capabilitie •,he; . Fi#r;•,S•;•a•• nally, the proposed rule

would address a number of petitions for rulemaking (PRMs) submitted Ih

March 2011 Fukushima Dai-ichi event.. This rulemaking is applicable to power reactor

-licensees; power reactor construction permit, design certification, design approval, and license

applicants; and decommissioning power reactor licensees. This rulemaking combines two NRC
activities for which documents have been published in the Federal Register - Onsite

Emergency Response Capabilities (RIN 3150-AJ1 1; NRC-2012-0031) and Station Blackout

Mitigation Strategies (RIN 3150-AJ08; NRC-201 1-0299). The new identification numbers for

this consolidated rule are RIN 3150-AJ49 and NRC-2014-0240.

DATES: Submit comments by [INSERT DATE 75 DAYS FROM DATE OF PUBLICATION].

Comments received after this date will be considered if it is practical to do so, but the

Commission is able to ensure consideration only for comments received before this date. A

public meeting will be held during the public comment period; refer to the NRC's public meeting

schedule on the NRC Web site, www.nrc..qov or directly at http://meetinos.nrc.gov/pmnslmtaq.

ADDRESSES: You may submit comments by any of the following methods (unless this

document describes a different method for submitting comments on a specific subject):

* Federal Rulemaking Web Site: Go to http://www.reaqulations..qov and search for

Docket ID NRC-2014-0240. Address questions about NRC dockets to Carol

Gallagher; telephone: 301-415-3463; e-mail: Carol.Gallaoqher~nrc.,Qov. For

technical questions contact the individuals listed in the FOR FURTHER

INFORMATION CONTACT section of this document.

• E-mail comments to: Rulemaking.Comments~nrc.gov. If you do not receive an

automatic e-mail reply confirming receipt, then contact us at 301-415-1677.

* Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at 301-415-

1101.

* Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington,

DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.

2
* Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 20852,

between 7:30 a~m. and 4:15p.m. (Eastern Time) Federal workdays; telephone: 301-

415-1677.

For additional direction on obtaining information and submitting comments, see "Obtaining

Information and Submitting Comments"' in the SUPPLEMENTARY INFORMATION section of

this document.

FOR FURTHER INFORMATION CONTACT: Timothy Reed, Office of Nuclear Reactor

Regulation, telephone: 301-415-1462, e-mail: Timothy.Reed(~nrc..qov:;rcBomn Ofc

of Nuclear Reactor Regulation, telephone: 301-415-2963, e-mail: Eric.Bowman•,nrc.,qov: U.S.

Nuclear Regulatory Commission, Washington DC 20555-0001.

SUPPLEMENTARY INFORMATION:

EXECUTIVE SUMMARY:

A. Need for the Regulatory Action

The U.S. Nuclear Regulatory Commission (NRC) is proposing to amend its regulations to

establish regulatory requirements for nuclear power reactor applicants and licensees to mitigate

beyond-design-basis-events. The NRC is proposing a rule that would make Commission Order

EA-12-049 and Order EA-12-051 generically applica l•,,stablish regulatory requirements for

~~~~~~~~~~~ a ami. A~~l ~d~O OM~ ~pr fan integrated response Capability

including supporting requirements for command and control, drills, training and change cont[•r'

include requirements for enhanced onsite emergency response caailt

ruiiil~l,•lii i UIii-~r 1 IIi.IydGhI ;.tlratzg•io-. for n•'w, roto ad address a number of

petitions for rulemaking (PRMs) submitted ,,, ,-''-:,fr,,•-,'=' the March 2011 Fukushima Dai3
ichi event. This rulemaking would be applicable to operating power reactor licensees; power

reactor construction permit, dci3,c. * ,, t,,,d• .. license applicants; and

decommissioning power reactor licensees. The NRC is conducting this rulemaking to amend

the regulations to reflect requirements imposed on current licensees by order and to reflect the

lessons learned from the Fukushima accident.

B. Major Provisions

Major provisions of this proposed rule include amendments or additions to parts 50 and 52 of

Title 10 of the Code of Federal Regulations (10 CFR) that would:

* Revise the 10 CFR par~~ a~~ ontent of application" requirements to

reflect the additional information that would be required for applications.

* Add proposed § 50.155, which contains beyond-design-basis mitigation

requirements that would make Orders EA-12-049 and EA-12-051 generically

applicable; requires an integrated response capability for beyond-design-basis

events that includes the integration of~-uideline sets with the existing

emergency operating procedures; training requirements; drills or exercise

requirements; and change control requirements• =,"•d ,t--bliohcc rcquir..c. 11~,• L

* Revise 10 CER part 50, appendix E to include enhanced capabilities for

assessing the impact and release of radioactive materials for multi-unit events;

to remove references to specific technology for each licensee's emergency

response data system; to include enhanced capabilities for onsite and offsite

communications; and to add staffing analysis requirements to address multi-unit

events.

4
C. Costs and Benefits

The NRC prepared a draft regulatory analysis to determine the expected quantitative

costs and benefits of the proposed rule, as well as qualitative factors to be considered in the

NRC's rulemaking decision. The draft analysis demonstrates that the proposed rule is justified.

The draft analysis examines the benefits and costs of the proposed rule requirements relative to

the baseline (i.e., no action alternative). Additionally, the draft analysis estimates the historical

costs incurred as a result of implementation of Order EA-12-049, Order EA-12-051, and related

industry initiatives. The proposed rule encompasses provisions that fall n•tc. t:,z =,-,r--u. :

4 ~~ithin the scope set forth in Order EA-12-049 or Order EA-12-051, as well as related

industry initiative•and 1l5i-poviblu11• ;,soc•iateo wi•bL~ Iv ,•ldu•,--,y, " ~

initiatives. As a result of the proposed rule, the NRC estimates that the industry as a whole

would incur a total one-time cost of $30 million, followed by an annual cost of $2.4 million. The

total present value of these costs is $58 million (using a 7 percent discount rate) and $72 million

(using a 3 percent discount rate) over a 63-year period. The average power reactor site would

incur a one-time cost Of approximately $510,000, followed by an annual cost of approximately

$42,000.

The proposed rule would result in a total one-time cost to the NRC of $1.1 million to

complete the rulemaking (i.e., complete the proposed rule, analyze public comments, hold

public meeting(s), and develop the final rule and regulatory guidance) and to oversee\

impemnttio SMGreatd o te rqureent• hi on-tmeco Wudb t olw ed byanr&

annua costof aproxiately$170000 fr SAM-relted rquireents
defense-in-depth benefits associated with their use following the onset of core damage. These

benefits are discussed in detail in appendix A to the draft regulatory analysis. Based on the I

NRC's assessment of the costs and benefits of the proposed rule, the NRC has concluded that /

the proposed rule is justified. For more information, please see the draft regulatory analysis /

(AgencywideM 154A2Document Access and Management. System (ADAMS) Accessionl~ No. ,/

The NRC epared a drsuppleme al regulatory alsi to d.etermine the e ected

quantitatieosts and benes of the pr osed requireent for new ators to inc d desi

featur in the plantdign sufficieto enhance ping duratoand minimizeliance o

h lan actionst maintain or lre t core cong, contain ent, and spen uel pool coling

capbiltie n et urng dd lss f ac pwerc crret ithe er a loss of •7rmal

. acces to the ultimaheat sink orr passive re ctor designs, oss of normalccess to7• t

normal heat sin The supple ntal regulat analysis con udes that the (oposed/

reqrUimree ms are justified d consiste with the Coin sion's expecta ns for ne reactors.

No. ML15069A278).

TABLE OF CONTENTS:

I. Obtaining Information and Submitting Comments.

A. Obtaining Information.

B. Submitting Comments.

I1. Background.

A. Fukushima Dai-ichi.

6
'I'

B. NRC Near-Term Task Force.

C. Implementation of the NTTF Recommendations.

0. Consolidation of Regulatory Efforts.

E. Public Involvement.

Ill. Petitions for Rulemaking.

IV• Discussion.

A. Rulemaking Objectives.

B. Rulemaking Scope.

C. Proposed Rule Organization.

0. Proposed Rule Regulatory Bases.

V. Section-by-Section Analysis.

•. Specific Requests for Comments.

Regulatory Flexibility Certification.

•. Availability of Regulatory Analysis.

Y. Availability of Guidance.

•• Backfitting and Issue Finality.

.•. Cumulative Effects of Regulation.

P. Plain Writing.

fi Environmental Assessment •nd Proposed Finding of No Significant

Environmental Impact.

•. Paperwork Reduction Act Statement.

. Criminal Penalties.

j9ti. Coordination with NRC Agreement States.

)#rl. Compatibility of Agreement State Regulations.

J 7
4|| ,tX Voluntary Consensus Standards.

)O )a,9(, " Public Meeting.

•. Availability of Documents.

I. Obtaining Information and Submitting Comments.

A. Obtaining Information.

Please refer to Docket ID NRC-2014-0240 when contacting the U.S. Nuclear Regulatory

Commission (NRC) about the availability of information for this action. You may obtain publiclyavailable information related to this action by any of the following methods:

* Federal Rulemaking Web Site: Go to http://www.re~ulations..qov and search for

Docket ID NRC-2014-0240.

* NRC's Agencywide Documents Access and Management System (ADAMS):

You may obtain publicly-available documents online in the ADAMS Public

Documents collection at http://www.nrc.aov/readino -rm/adams.html. To begin the

search, select "ADAMS Public Documents" and then select "Begqin Web-based

ADAMS Search." For problems with ADAMS, please contact the NRC's Public

Document Room (PDR) reference staff at 1-800-397-4209, 301-415-4737, or by email to pdr.resource•,nrc..qov. For the convenience of the reader, instructions about

obtaining materials referenced in this document are provided in the "Availability of

Documents"' section.

* NRC's PDR: You may examine and purchase copies of public documents at the NRC's

PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, Rockville, Maryland

20852.

8
B. Submitting Comments.

Please include Docket ID NRC-2014-0240 in the subject line of your comment

submission.

The NRC cautions you not to include identifying or contact information in comment

submissions that you do not want to be publicly disclosed in your comment submission. The

NRC will post all comment submissions at http://www.requlations.qov as well as enter the

comment submissions into ADAMS, and the NRC does not routinely edit comment submissions

to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the

NRC, then you should inform those persons not to include identifying or contact information that

they do not want to be publicly disclosed in their comment submission. Your request should

state that the NRC does not routinely edit comment submissions to remove such information

before making the comment submissions available to the public or entering the comment into

ADAMS.

You may submit comments on the information collections by the methods indicated in

the Paperwork Reduction Act Statement.

II. Background.

A. Fukushima Dai-ichi.

At 2:46 p.m. Japan standard time on March 11, 2011, the Great East Japan Earthquake,

rated a magnitude 9.0, occurred at a depth of approximately 25 kilometers, 130 kilometers east

of Sendai and 372 kilometers northeast of Tokyo off the coast of Honshu Island. This

earthquake resulted in the automatic shutdown of I luclear power plants (NPPs) at four sites

along the, northeast coast of Japan including±,te hr~e~fecors at the Fukushima Dai-ichi NPP

9
(the three remaining plants were in outages). The earthquake precipitated a large tsunami that

is estimated to have exceeded 14 meters in height at the Fukushima Dai-ichi NPP. The

earthquake and tsunami produced widespread devastation across northeastern Japan, resulting

in approximately 25,000 people dead or missing, displacing many tens of thousands of people,

and significantly impacting the infrastructure and industry in the northeastern coastal areas of

Japan.

The earthquake and tsunami disabled the majority of the external and internal electrical

power systems at the Fukushima Dai-ichi NPP, leaving it with only a few hours' worth of battery

power. Since a NPP licensee typically relies on electrical power to keep its reactor core and

spent fuel pool (SFP) cool, this loss of internal and external power was a significant challenge to

operators at Fukushima Dai-ichi. In addition, the combination of severeevents challenged the

implementation of emergency plans and procedures.

B. NRC Near-Term Task Force.

The NRC Chairman's tasking memorandum, COMGBJ-1 1-0002, "NRC Actions

Following the Events in Japan," established a senior-level task force referred to as the

"Near-Term Task Force" (NTTF) to conduct a systematic and methodical review of NRC

regulations and processes to determine if the agency should make safety improvements in light

of the events in Japan. On July 12, 2011, the NRC staff provided the Commission with the

report of the NTTF (NTTF Report) as an enclosure to SECY-1 1-0093, "Near-Term Report and

Recommendations for Agency Actions Following the Events in Japan." The NTTF concluded

that continued U.S. plant operation and NRC licensing activities present no imminent risk to

public health and safety. While the NTTF also concluded that the current regulatory system has

served the NRC and the public well, it found that enhancements to safety and emergency

preparedness are warranted and made a dozen general recommendations for Commission

10
consideration. In examining the Fukushima Dai-ichi accident for insights for reactors in the

United States, the NTTF addressed protecting against accidents resulting from natural

phenomena, mitigating the consequences of such accidents, and ensuring emergency

preparedness. The NTTF found that the Commission's longstanding defense-in-depth

philosophy, supported and modified as necessary by state-of-the-art probabilistic risk

assessment techniques, should continue to serve as the primary organizing principle of its

regulatory framework. The NTTF concluded that the application of the defense-in-depth

philosophy could be strengthened by including explicit requirements for beyond-design-basis

events.

In response to the NTTF Report, the Commission directed the NRC staff to engage with

stakeholders to review and assess the NTTF recommendations in a comprehensive and holistic

manner and to provide the Commission with fully-informed options and recommendations. The

Commission's Staff Requirements Memorandum (SRM)-SECY-1 1-0093 provided that direction

and specifically directed the NRC staff to pursue recommendation 1 of the NTTF Report

independent of the activities associated with the review of the remaining recommendations.

The NTTF's recommendation 1 was to establish a logical, systematic, and coherent regulatory

framework for adequate protection that appropriately balances defense-in-depth and risk

considerations. This recommendlation included steps for the establishment of a Commission

policy statement for a risk-informed defense-in-depth framework including extended designbasis requirements and the initiation of rulemaking to implement that framework. The results of

the NRC staff work on NTTF recommendation 1 were provided to the Commission in

SECY-1 3-01 32, "Plan for Updating the U.S. Nuclear Regulatory Commission's Cost Benefit

Guidance," and dispositioned by the Commission in SRM-SECY-1 3-01 32, which specifically

disapproved the establishment of a design-basis extension category of events and associated

regulatory requirementsl but allowed for~keevaluation, as appropriate, in the context of the
Commission direction on the proposed policy statement for a long-term Risk Management

Regulatory Framework. That work is outside of the scope of this rulemaking. li Com•;•,'-i

C. Implementation of the NTTF Recommendations.

Following the issuance of the NTTF Report, the NRC staff provided the Commission with

recommendations for near-term action in SECY-1 1-0124, "Recommended Actions to be Taken

Without Delay from the Near-Term Task Force Report," dated September 9, 2011. The

suggested near-term actions addressed several NTTF recommendations associated with this

rulemakir•)iluding NTTF recommendations 4, 8, and 9.3. In SRM-SECY-1 1-0124, dated

October 18, 2011, the Commission directed the NRC staff to, among other things: initiate a

rulemaking to address NTTF recommendation 4, Station Blackout (SBO) regulatory actions, as

an Advance Notice of Proposed Rulemaking (ANPR); designate the SBO rulemaking associated

with NTTF recommendation 4 as a high priority rulemaking; craft recommendations that

continue to realize the strengths of a performance-based system as a guiding principle; and

consider approaches that are flexible and able to accommodate a diverse range of

circumstances and conditions. As discussed more fully in later portions of this proposed rule,

the regulatory actions associated with NTTF recommendation.4 evolved substantially from this

early Commission direction, and included issuance of Order EA-12-049 that, as implemented,

ultimately addressed all of NTTF recommendation 4 as well as other recommendations.

In SECY-1 1-0137, "Prioritization of Recommended Actions to Be Taken in Response to

Fukushima Lessons Learned," dated October 3, 2011, the NRC staff, based on its assessment

of the NTTF recommendations, proposed to the Commission a three-tiered prioritization for

implementing regulatory actions stemming from the NTTF recommendations. The Tier 1

recommendations were those actions having the greatest safety benefit that could be

implemented without unnecessary delay. The Tier 2 recommendations were those actions that

12
needed further technical assessment or critical skill sets to implement, and the Tier 3

recommendations were longer-term actions that depended on the completion of a shorter-term

action or needed additional study to support a regulatory action. On December 15, 2011, the

Commission approved the staff's recommended prioritization in SRM-SECY-1 1-0137.

The NTTF recommendations that form the basis of this rulemaking activity are:

* NTTF recommendation 4: strengthen SB0 mitigation capability at all operating

and new reactors for design-basis and beyond-design-basis external events;

* NTTF recommendation 7: enhance spent fuel pool makeup capability and

instrumentation for the spent fuel pool;

* NTTF recommendation 8: strengthen and integrate onsite emergency response

capabilities such as emergency operating procedures (EOPs), Severe Accident

Management Guidelines (SAMGs), and extensive damage mitigation guidelines

(EDMGs);

* NTTF recommendation 9: require that facility emergency plans address staffing,

dose assessment capability, communications, training and exercises, and

equipment and facilities for prolonged station blackout, multi-unit events, or both;

* NTTF recommendation 10: pursue additional emergency protection topics related

to multi-unit events and prolonged station blackout, including command and

control structure and the qualifications of decision makers; and

* NTTF recommendation 11: pursue emergency management topics related to

decision making, radiation monitoring, and public education, including the ability

to deliver equipment to the site with degraded offsite infrastructure.

In response to input received from stakeholders, the NRC accelerated the schedule

originally proposed in SECY-11-0137. On February 17, 2012, the NRC staff recommended in

13
SECY-12-0025, 'Proposed Orders and Requests for Information in Response to Lessons

Learned From Japan's March 11, 2011, Great T~hoku Earthquake and Tsunami," that the

Commission issue orders and requests for information.

To address Tier 1 NTTF recommendation 4, the NRC issued Order EA-12-049 on March

12, 2012, requiring all U.S. nuclear power plant licensees to implement strategies that would

allow them to cope without their permanent electrical power sources for an indefinite period of

time. These strategies would provide additional capability to maintain or restore reactor core

and spent fuel cooling, as well as protect the reactor containment. This order also addressed:

portions of NTTF recommendation 9 to require that facility emergency plans address prolonged

station blackouts and multi-unit events; portions of NTTF recommendation 10 to pursue

additional emergency protection topics related to multi-unit events and prolonged station

blackout; and portions of NTTF recommendation 11 to pursue emergency procedure topics

related to decision-making, radiation monitoring, and public education.

To address Tier 1 NTTF recommendation 7, the NRC issued Order EA-12-051 on March

12, 2012, requiring all U.S. nuclear power plant licensees to have a reliable indication of the

water level in associated spent fuel storage pools.

To address Tier 1 NTTF recommendation 8, the NRC issued an ANPR on April 18, 2012

(77 FR 23161), to engage stakeholders in rulemaking activities associated with the methodology

for integration of onsite emergency response processes, procedures, training and exercises.

D. Consolidation of Regulatory Efforts.

While developing the NTTF rulemakings, the NRC staff recognized that efficiencies

could be gained by consolidating the rulemaking efforts due to the inter-relationships among the

proposed changes. The NRC staff recommended to the Commission in COMSECY-1 3-0002,

"Consolidation of Japan Lessons Learned Near-Term Task Force Recommendations 4 and 7

14
Regulatory Activities," COMSECY-13-0010, 'Schedule and Plans for Tier 2 Order on

Emergency Preparedness for Japan Lessons Learned," and SECY-14-0046, "Fifth 6-Month

Status Update on Response to Lessons Learned From Japan's March 11, 2011, Great Tohoku

Earthquake and Subsequent Tsunami," the consolidation of rulemaking activities that address

NTTF recommendations 4, 7, 8, portions of 9, 10.2, and 11.1. Section II.B of this document

contains a more complete discussion of the scope of NTTF recommendations addressed by this

proposed rule. The Commission approved these consolidations in the associated SRMs. These

consolidations were intended to:

1. Align the proposed regulatory framework with ongoing industry implementation efforts to

produce a more coherent and understandable regulatory framework. Given the

complexity of these requirements and their associated implementation, the NRC

concluded that this is an important objective for the regulatory framework.

2. Reduce the potential for inconsistencies and complexities between the related

rulemaking actions that could occur if the efforts remained as separate rulemakings.

3. Facilitate better understanding of the proposed requirements for both internal and

external stakeholders, and thereby lessen the impact on internal and external

stakeholders who would otherwise need to review and comment on multiple rulemakings

while cross-referencing both proposed rules and sets of guidance documents.

E. Public Involvement.

This proposed rule consolidates two previous rulemaking efforts: the Station Blackout

Mitigation Strategies rulemaking, directed by SRM-COMSECY-13-0002, and the Onsite

Emergency Response Capabilities rulemaking, which implemented NTTF recommendation 8.

Both regulatory efforts offered extensive external stakeholder involvement opportunities,

including public meetings, ANPRs issued for public comment, and draft regulatory basis

15
documents issued for public comment. The major opportunities for stakeholder involvement

were:

1. Station Blackout ANPR (77 FR 16175; March 20, 2012);

2. Onsite Emergency Response Capabilities ANPR (77 FR.231 61; April 18, 2012);

3. Station Blackout Mitigation Strategies draft regulatory basis and draft rule concepts

(78 FR 21275; April 10, 2013). The final Station Blackout Mitigation Strategies regulatory

basis was subsequently issued on July 23, 2013 (78 FR 44035); and

4. Onsite Emergency Response Capabilities draft regulatory basis (78 FR 1154; January 8,

2013). The final Onsite Emergency Response Capabilities regulatory basis, with

preliminary proposed rule language, was subsequently issued on October 25, 2013

(78 FR 63901).

The NRC described in each final regulatory basis document how it considered

stakeholder feedback in developing the respective final regulatory basis, including consideration

of ANPR comments and draft regulatory basis document comments. Section 5 of the Station

Blackout Mitigation Strategies regulatory basis document includes a discussion of stakeholder

feedback used to develop the final regulatory basis. Appendix B to the Onsite Emergency

Response Capabilities regulatory basis includes a discussion of stakeholder feedback used to

develop that final regulatory basis.

The public has had multiple opportunities to engage in these regulatory efforts. Most

noteworthy were the following:

1. Preliminary proposed rule language for Onsite Emergency Response Capabilities

made available to the public on November 15, 2013 (78 FR 68774).

2. Consolidated rulemaking proof of concept language made available to the public on

February 21, 2014.

16
3. Preliminary proposed rule language for Mitigation of Beyond-Design-Basis Events

rulemaking made available to the public on August 15, 2014.

4. Preliminary proposed rule language for Mitigation of Beyond-Design-Basis Events

rulemaking made available to the public on November 13, 2014 and December 8,

2014 to support public discussion with the Advisory Committee on Reactor

Safeguards (ACRS).

The NRC staff has had numerous interactions with the ACRS, and in all cases these

were public meetings, including the following:

1. The ACRS Plant Operations and Fire Protection subcommittee met on February 6,

2013 to discuss the Onsite Emergency Response Capabilities regulatory basis.

2. The ACRS Regulatory Policies and Practices subcommittee met on December 5,

2013 and April 23, 2013 to discuss the Station Blackout Mitigation Strategies

regulatory, basis.

3. The ACRS full committee met on June 5, 2013 to discuss the Station Blackout

Mitigation Strategies regulatory basis.

4. The ACRS Fukushima subcommittee met on June 23, 2014 to discuss consolidation

of Station Blackout Mitigation Strategies and Onsite Emergency Response

Capabilities rulemakings.

5. The ACRS full committee met on July 10, 2014 to discuss consolidation of Station

Blackout Mitigation Strategies and Onsite Emergency Response Capabilities

rulemakings.

6. The ACRS Fukushima subcommittee met on November 21, 2014 to discuss

preliminary proposed Mitigation of Beyond-Design-Basis Events rulemaking language.

7. The ACRS Fukushima full committee met on December 4, 2014 to discuss preliminary

proposed Mitigation of Beyond-Design-Basis Events rulemaking language.

17
The NRC held many additional public meetings that have supported the development of

thsproposed rule. Not withstanding these efforts to engage the public during the preparation of I

this proposed rule, the Commission is committed to the rigors of the notice-and-comment

process enacted by the Administrative Procedures Act, and is providing members of the public a

75-day comment period on the requirements NRC is proposing today.

III. Petitions for Rulemaking.

During development of this proposed rule, the NRC gave consideration to the issues

raised in six petitions for rulemaking (PRMs) submitted to the NRC, five from the National

Resources Defense Council Inc. (NRDC) (PRM-50-97, PRM-50-98, PRM-50-1 00, PRM-50-1 01,

and PRM-50-1 02), and one submitted by Mr. Thomas Popik (PRM-50-96). The petitions filed by

NRDC use the NTTF Report as the sole basis for the PRMs. The NTTF recommendations that

the NRDC PRMs rely upon are: 4.1, 7.5, 8.4, 9.1, and 9.2. This proposed rule addresses each

of these recommendations, and therefore it would resolve the issues raised by the PRMs. The

NRDC petitions were dated July 26, 2011, and docketed by the NRC on July 28, 2011. The

NRC published a notice of receipt in the Federal Register on September 20, 2011 (76 FR

58165), and did not ask for public comment at that time.

In PRM-50-97 (NRC-2011-0189), the NRDC requested emergency preparedness

enhancements for prolonged Station blackouts in the areas of communications ability,

Emergency Response Data System (ERDS) capability, training and exercises and equipment

and facilities (NTTF recommendation 9.2). The NRC determined that the issues raised in this

PRM should be considered in the NRC's rulemaking process. The NRC's consideration of the

issues raised in PRM-50-97 are reflected in the proposed provisions in § 50.155(,) and •, and

the proposed amendments to appendix E in both section VI and in new section VII,

"Communications and Staffing Requirements for the Mitigation of Beyond Design Basis Events."

18
The NRC concludes that consideration of the PRM issues, as discussed herein, will address

PRM-50-97. The NRC is closing the docket for this petition and intends to take final action on

this petition in the Federal Register notice the NRC issues for the final Mitigation of BeyondDesign-Basis Events rule.

ln PRM-50-98 (NRC-201 1-0189), the NRDC requested emergency preparedness

enhancements for multi-unit events in the areas of personnel staffing, dose assessment

capability, training and exercises, and equipment and facilities (NTTF recommendation 9.1).

The NRC determined that the issues raised in this PRM should be considered in the NRC's

rulemaking process. The NRC's consideration of the issues raised in PRM-50-98 are reflected

in the proposed provisions § 50.155(b)(5), (•, and •;and the proposed amendment to

appendix E in section IV as well as the addition of a new section VII. The NRC concludes that

consideration of the PRM issues, as discussed herein, will address PRM-50-98. The NRC is

closing the docket for this petition and intends to take final action on this petition in the Federal

Register notice the NRC issues for the final Mitigation of Beyond-Design-Basis Events rule.

In PRM-50-1 00, the NRDC requested enhancement of spent fuel pool makeup capability

and instrumentation for the spent fuel pool (NTTF recommendation 7.5). The NRC determined

that the issues raised in this PRM should be considered in the NRC's rulemaking process, and

the NRC published a document in the Federal Register with this determination on July 23, 2013

(78 FR 44034). The NRC's consideration of the issues raised in PRM-50-1 00 are reflected in

the proposed provisions § 50.155(b)(1) and (c)(4). The proposed rule would make generically

applicable NRC's Order EA-12-051, "Spent Fuel Pool Instrumentation." The NRC concludes

that consideration of the PRM issues, as discussed herein, addresses PRM-50-100. The NRC

has already closed the docket for this petition and intends to take final action on this petition in

the Federal Register notice the NRC issues for the final Mitigation of Beyond-Design-Basis

Events rule.

19
In PRM-50-101, the NRDC requested that § 50.63, "Loss of all alternating current

power," (sic) be revised to establish a minimum coping time of 8 hours for a loss of all

alternating current (ac) power, establish the equipment, procedures, and training necessary to

implement an extended loss of ac power (72 hours) for core and spent fuel pool cooling and for

reactor coolant system and primarycontainment integrity as needed, and preplan/prestage

offsite resources to support uninterrupted core and spent fuel pool cooling and reactor coolant

system and containment integrity as needed (NTTF recommendation 4.1). The NRC

determined that the issues raised in this PRM should be considered in the NRC's rulemaking

process, and the NRC published a document in the Federal Register with this determination on

March 21, 2012, (77 FR 16483). The NRC's consideration of the issues raised in PRM-50-101

is reflected in the proposed provisions § 50.155(b)(1), (c), ),(f), a~.(?he NRC concludes

that consideration of the PRM issues, as discussed herein, addresses PRM-50-1 01. The NRC

has already closed the docket for this petition and intends to take final action on this petition in

the Federal Register notice the NRC issues for the final Mitigation of Beyond-Design-Basis

Events rule.

In PRM-50-1 02, the NRDC requested more realistic, hands-on training and exercises on

SAMGs and EDMGs for licensee staff expected to implement those guideline sets and make

decisions during emergencies (NTTF recommendation 8.4). The NRC determined that the

issues raised in this PRM should be considered in the NRC's rulemaking process, and the NRC

published a document in the Federal Register with this determination on April 27, 2012

(77 FR 25104). The NRC's consideration of the issues raised in PRM-50-102 are reflected in

the proposed provisions § 50.1 55 and •9. The NRC concludes that consideration of the PRM

issues, as discussed herein, addresses PRM-50-102. The NRC has already closed the docket

for this petition and intends to take final action on this petition in the Federal Register notice the

NRC issues for the final Mitigation of Beyond-Design-Basis Events rule.

20
In PRM-50-96, Mr. Thomas Popik requested that the NRC amend its regulations to require

facilities licensed by the NRC to assure long-term cooling and unattended water makeup of

spent fuel pools in the event of geomagnetic storms caused by solar storms resulting in longterm losses of power. The NRC determined that the issues raised in this PRM should be

considered in the NRC's rulemaking process and the NRC published a document in the Federal

Register with this determination on December 18, 2012 (77 FR 74788). In that Federal Register

document, the NRC also closed the docket for this petition. Specifically, the NRC indicated that

it would monitor the progress of the mitigation strategies rulemaking to determine whether the

requirements established would address, in whole or in part, the issues raised in the PRM. In

this context, the proposed requirements in § 50.155(b)(1) and (c) and the associated draft

regulatory guidance should address in part the issues raised because these actions would

establish offsite assistance to support maintenance of the key functions (including both reactor

and spent fuel pool cooling) following an extended loss of ac power that has been postulated for

geomagnetic events. Additional consideration of these issues will result from NRC's

participation in the interagency task force developing a National Space Weather Strategy and

the associated action plan. Both the strategy and action plan are expected to be completed in

2015. When the National plans are completed, the NRC will reevaluate the need for additional

actions to address the impact of geomagnetic storms on nuclear power plants within the overall

context of the National Space Weather Strategy and action plan.

IV. Discussion.

A. Rulemaking Objectives.

The regulatory objectives of this rulemaking are to: 1) make the requirements in Order

EA-12-049 and Order EA-12-051 generically applicable, giving consideration to lessons learned

21
from implementation of the orders; 2) establish new requirements for an integrated response

caaiiy~sals new requirements for actions that are related to

onsite emergency response; rrvI ") 'e~rmrt r -- L~a[iy •raiegie• ['u, iw o• t

•.• an address issues raised by PRMs that were submitted 4 •-fe~~f'he

March 2011 Fukushima Dai-ichi event.

1. Make the requirements in Order EA-12-049 and Order EA-12-051 generically

applicable, giving consideration to lessons learned from implementation of the orders.

An objective of this rulemaking is to place the requirements in Order EA-1 2-049 and

Order EA-12-051 into the NRC's regulations so that they apply to all current and future power

reactor applicants, and to provide regulatory clarity and stability to power reactor licensees. In

making the requirements of Order EA-1 2-049 generically-applicable, this proposed rule would

also consider the reevaluated hazard information developed in response to the March 12, 2012,

NRC letter issued under § 50.54(f) as part of providing reasonable protection for mitigation

strategies equipment against external flooding or seismic hazards. Because these orders were

issued to current licensees, the requirements of these orders would not apply to future

licensees. In the absence of this proposed rule, these requirements would need to be

implemented for new reactor applicants or licensees through additional orders or license

codiios(as was done for the Vogtle Electric Generating Plant, Units 3 and~ d Virgil C./

Summer Nuclear Station, Units 2 and 3 combined licenses (COLs), respectively). As part of the

rulemaking, the NRC considered stakeholder feedback and lessons-learned from the

implementation of the orders, including any challenges or unintended consequences associated

with implementation. The NRC reflected this stakeholder input in the draft regulatory guidance

for this proposed rule.

22
2. Establish new requirements for an integrated response capability t

An objective of this rulemaking is to establish requirements for an integrated response

capability for beyond-design-basis events that would integrate existing strategies and guidelines

(implemented through guideline sets) with the existing EOPs. This would include guideline sets

that implement the requirements of current § 50.54(hh)(2) and Order EA-12-049.

alz, 1 .".ldz =i-lnootc that irp!.. ILS~1Z3 h~ '"r i ~ '" ifdwa

Tiprposed rule would require sufficient staffing, command and control, training,

drills, and change control to support the integrated response capability.

3. Establish new requirements for actions that are related to onsite emergency

response.

An objective of this rulemaking is to establish requirements for onsite emergency

response capabilities being implemented in conjunction with the implementation of Order

EA-12-049. This proposed rule contains new requirements for staffing and communications

assessment, and clarifies requirements for multi-source dose assessment.

4. Pr ide requirments f gitigating rategies for ew reactor designs.

An o ectiveo isj rulem ing is to e /' blish requirents for appliasfr ne reactor

desig to Pro'e mitig~atin trategies f beyond-desn-basis eventshis Propo 'ed rul

c tains ruirementsf includingi sign featureniohe plant desi sufficie°o enha: e

c op' durations aj minimize/riance on hu n actions to m/ itain or retore core oling,

c-ontainmentd spent, fue ool cooling c pabilities durnn n extend• loss of all c power

(ELAP)c.F •escuurrentwwitho itth~errraallOCssonormal access heatul~tim e heat sink o for passive

23
5. Address a number of PRMs submittedteMrh21 /

Fukushima Dai-ichi event.

An objective of this rulemaking is to address the five PRMs filed by the NRDC that raise

issues that pertain to the technical objectives of this rulemaking. The petitions rely solely on

the NTTF Report, and request that the NRC undertake rulemaking in a number of areas that are

addressed by this proposed rule. This proposed rule also addresses, in part, the PRM

submitted by Mr. Thomas Popik.

B. Rulemaking Scope.

The scope of this rulemaking, described in terms of the relationship to var'ious NTTF

recommendations that provided the regulatory impetus for this proposed rule, includes:

1. All the requirements that were within the scope of Station Blackout Mitigation

Strategies rulemaking. These requirements address NTTF recommendations 4 and 7. This

aspect of this proposed rule will also address NTTF 11.1 regarding onsite emergency resources

to support multi-unit events with station blackout, including the need to deliver equipment to the

site despite degraded offsite infrastructure. This provision currently is being implemented

through Order EA-12-049.

2. All the requirements that were within the scope of the Onsite Emergency

Response Capabilities rulemaking. These requirements address NTTF recommendation 8, as

directed by SRM-SECY-1 1-0137. This aspect of this proposed rule also addresses command

and control issues in NTTF recommendation 10.2.

3. Numerous requirements regarding onsite emergency response actions are

included in this proposed rule. These emergency response actions currently are being

implemented by Order EA-12-049; in addition, NRC staff has developed draft guidance to

24
support the emergency response aspect of this proposed rule. The specific regulatory actions

related to emergency response in this proposed rule and the associated NTTF

recommendations are:

a. Staffing and communications: addresses NTTF recommendation 9.3; also

discussed in NTTF recommendations 9.1 and 9.2. These regulatory issues currently are being

implemented through Order EA-12-049. The proposed requirements also address supporting

facilities and equipment, as discussed in the same NTTF recommendations.

b. Multi-source term dose assessment: addresses NTTF recommendation 9.3; also

discussed in NTTF recommendation 9.1. This regulatory issue is being implemented voluntarily

by industry. l'• ; e Q 4

c. Training and exercis 4: addresses NTTF recommendation 9.3; also discussed in

NTTF recommendations 9.1 and 9.2. These regulatory issues currently are being implemented

through Order EA-12-049.

•Accordingly, this rulemaking will address all th~recommendations in NTTF

recommendations 4, 7, 8, 9.1, 9.2, 9.3 (with one exception - EROS modernization is addressed,

but maintenance of EROS capability throughout the accident is not addressed), 10.2, and 11.1.

This rulemaking also addresses NTTF recommendation, 9.4: modernize EROS. This

action differs from the other regulatory actions because EROS is not an essential component of

a licensee's capability to mitigate a beyond-design-basis external event. However, ERDS is an

important form of communication between the licensee and the NRC. Modernization of EROS

has been completed voluntarily by industry; therefore, NRC has included amendments to the

technology-specific references in 10 CFR part 50, appendix E, section VI, "Emergency

Response Data System," in this proposed rule.

I

,j .. ... St£: ceuz~i~ h plt,~ cg

25

; 6'..

.nt fzr applic'" fn np ulear
Unlike the requirements for the mitigation of beyond-design-basis external events

imposed by Order EA-1 2-049, and requirements that address the loss of large areas of the plant

due to explosions and fire in current § 50.54(hh)(2) (NRC is proposing in this rule to move these

requirements to a new section), SAMGs are not an NRC requirement imposed on licensees.

Nevertheless, SAMGs are well established guidance documents that have been developed by

the nuclear power industry with substantial NRC involvement, have been implemented by every

operating nuclear power reactor licensee for decades, and are the subject of a license condition

for combined licenses. Following the Three Mile Island (TMI) accident, the nuclear power

industry revised its emergency response procedures to be symptom-based, and as a result,

developed EOPs. In the mid-to-late 1980s, the NRC and the nuclear power industry identified a

need to consider plant conditions that could lead to a severe accident. These efforts led to the

nuclear industry voluntarily initiating a coordinated program on accident management in 1990.

Section 5 of Nuclear Energy Institute (NEI) 91-04 (formerly NUMARC 91-04), Revision 1,

"Severe Accident Closure Guidelines," describes the elements of the industry's severe accident

management closure actions. The program involves the development of: 1) a structured

method by which utilities could systematically evaluate and enhance their ability to deal with

potential severe accidents, 2) vendor-specific SAMGs for use by licensees in developing plantspecific SAMGs, and 3) guidance and material to support utility activities related to training for

severe accidents. In 1992, the Electric Power Research Institute (EPRI) developed the SAMG

26
Technical Basis Report (TBR). Volume one of this report covers general actions that could be

taken to manage a severe accident (referred to as SAMG candidate high level actions) and their

effects, and volume two is a detailed report on the physics of accident progression. By letter

dated June 20, 1994, the NRC accepted the industry's approach for mitigating the

consequences of severe accidents, including licensee regulatory commitments to implement

plant-specific SAMGs, using the guidance developed in section 5 of NEI 91-04, Revision 1, by

December 31, 1998.

The NRC assessed the ongoing implementation of SAMGs at a select number of plants

during the 1997-1998.time frame as discussed in SECY-97-132, "Status of the Integration Plan

for Closure of Severe Accident Issues and the Status of Severe Accident Research," and

SECY-98-1 31, "Status of the Integration Plan for Closure of Severe Accident Issues and the

Status of Severe Accident Research," and concluded that the results of the voluntary initiative

achieved the NRC's overall objectives established for accident management in SECY-89-012,

"Staff Plans for Accident Management Regulatory and Research Programs." In 2012, EPRI

revised the TBR to account for the initial lessons learned from the Fukushima Dai-ichi accidents,

as well as enhanced understanding of severe accident behavior gained from additional research

and analyses performed since the original report was published.

•Following the events at Fukushima Dai-ichi, the NRC again inspected the

implementation, ongoing training, and maintenance of licensee SAMGs at all power reactor

sites, except those that had permanently ceased operation, through performance of Temporary

Instruction (TI)-251 5/184, "Availability and Readiness Inspection of Severe Accident

Management Guidelines (SAMGs)." The NRC found that some licensees had not maintained

the SAMGs in accordance with the latest revisions of the applicable industry generic technical

guidelines nor conducted training in a consistent and systematic manner. The NRC inspectors

27
attributed the inconsistent implementation and training on SAMGs to the voluntary nature of this

initiative.

Based in part on the findings of the inspections described above, the NTTF

recommended that the NRC require licensees to integrate onsite emergency response

capabilities, including SAMGs. R ~a~d[a 'h it "ccr oul rcui licnrzccc to

Order requirements, which were justified as necessary for adequate protection under § 50.109,

SAMGs do not involve adequate protection. Because the imposition of SAMGs also would not

be necessary to bring licensees into compliance with an existing NRC requirement, a SAMGs

requirement would have to be justified under § 50.109 as a cost-justified, substantial increase in

protection of the public health and safety or common defense and security.

As part of the effort to develop the backfitting justification for imposition of SAMG

requirements, the NRC sought to make use of any applicable quantified risk information that

might provide risk insights to inform the justification. In this regard, the NRC looked at its recent

technical analysis 1 performed in support of the Containment Protection and Release Reduction

(CPRR) rulemaking regulatory basis2. This analysis is relevant because it examined regulatory

alternatives that would be implemented after core damage to determine whether any of the

contemplated approaches can be justified under the NRC's backfitting provisions. In this

respect, the risk insights stemming from this work might have relevance to NRC's consideration

of SAMG requirements where the safety benefits would occur after core damage. The NRC

also considered other post-Fukushima regulatory efforts (e.g., the safety benefits due to

1 The technical risk insights were presented to the ACRS. Reliability and PRA, and Fukushima subcommittees on

August 22, 2014, and to the ACRS Reliability and PRA subcommittee on November 19, 2014. This footnote is

informational only; it does not imply advisory committee endorsement of the technical analysis.

2 Refer to the draft regulatory basis for Containment Protection and Release Reduction.

28
implementation of Order EA-12-049 mitigation strategies, which result in a reduction in core

damage frequency) within this technical analysis. The NRC acknowledges that the work to

support the CPRR rulemaking was not conducted to provide a complete quantitative measure of

the possible safety benefits of SAMG requirements, particularly with regard to how SAMGs

might benefit maintenance of containment integrity or support more informed protective action

recommendlations by the emergency response organization following core damage. However,

this technical analysis work does provide valuable risk insights that the NRC concluded were

important to fully inform the decision on this matter, and that additionally influenced the NRC's

development of the proposed SAMG framework.

The CPRR technical analysis includes a screening for a conservative high estimate of

frequency-weighted individual latent cancer fatality risk. This screening analysis combined the

highest ELAP frequency among all boiling water reactors (BWRs) with Mark I or Mark II

containments, a success probability in the FLEX equipment3 of 0.6 per demand following core

melt, the highest conditional individual latent cancer fatality (ILCF) risk among all BWRs with

Mark I or Mark II containments, and a worst case re-habitability assumption. This yields a

conservative high estimate of frequency-weighted individual latent cancer fatality risk of

approximately 7x1 0. per reactor year. This combination of assumptions does not exist at any

BWR with a Mark I or Mark II containment. This conservative estimate of the risk can be viewed

as the maximum possible risk that could be removed or reduced through regulatory action (i.e.,

the CPRR technical analysis examines a range of post-core damage regulatory actions for

BWRs with Mark I or Mark II containments to identify whether any of these proposals might

result in a safety benefit large enough to be justified under the Commission's backfitting

requirements). This estimate is compared against the quantitative health objective, which is a

3 Refer to NEI 12-06, Revision 0, 'Diverse and Flexible Coping Strategies (FLEX) implementation Guide," for a

description of industry-developed guidance on FLEX strategies and equipment.

29
quantitative measure that equates to 1/10 of 1 percent of the ILCF risk and relates to the

Commission's Safety Goal Policy. This quantitative metric for the individual latent cancer fatality

risk is approximately 2x10.6 per reactor year. This technical work shows that the risk is well

below a level that equates to 1/10 of 1 percent of the surrounding population's latent cancer

fatality risk. This result also means, that, from a quantitative standpoint, achieving risk

reductions that might satisfy backfitting requirements is/Jinlikely. More refined risk estimates

from the same work (i.e., which remove the worst case assumptions and instead use

assumptions specific to each power reactor), push this potential risk benefit significantly lower,

by approximately two orders of magnitude. This result demonstrates the benefits of the NRC's

regulations to both effectively keep the frequency of core damage very low at BWRs with Mark I

and II containments, and to ensure through emergency preparedness requirements that the

surrounding population is adequately protected. Those general attributes of the NRC's

regulations that result in this risk insight (i.e., requirements that resulted in reduced core

damage frequencies and effective emergency preparedness requirements) apply to all power

reactor designs. The NRC has not performed a comprehensive quantitative analysis of the

potential safety benefits of SAMG requirements for all types of reactors. However, the general

risk insights obtained from the CPRR work align well with NUREG-1 935, "State-of-the-Art

Reactor Consequence Analyses (SOARCA) Report," (November 2012), which shows very low

levels of risk (e.g., individual early fatality risk is essentially zero)af•'iCF risk is thousands of

times lower than the NRC Safety G9'•lnd millions of times lower than the general cancer fatility

risk in the United States from all causes). As such, the available risk insights point to the likely

outcome that a comprehensive quantitative analysis, where the proposed regulatory action is

intended to provide its safety benefit in the post-core damage environment (as is the case for

use of SAMGs), would not demonstrate a substantial safety benefit. In addition, for the specific

case of the SAMG requirements in this proposed rule, the proposed regulatory action's benefit

30
must also recognize that imposing SAMG requirements must be compared with the current

regulatory state, e.g., SAMGs exist and are voluntarily in use under an industry initiative.

perspective, *chspecifically recognizes the unquantified benefits of SA s in suPPort of

containment integri yparticularly for beyond-design-basis external ents for which

uncertai icore

damage lat is

informec ,ulId:

Along with its quantitative analysis, the staff prepared a SAMG backfit /

1 analysis that relied on qualitative factors, such as defense-in-depth. The NRC - Commission found that this analysis was inconsistent with the

review • Commission's direction in the staff requirements memorandum (SRM) for ;s, or

SECY-14-0087, "Qualitative Consideration of Factors in the Development. velop

licensee of Regulatory Analyses and Backfit Analyses." In that SRM, the

the SM Commission did "not authorize an expansion of the consideration of

qualitative factors in regulatory analyses and backfit analyses."

Fukushi Consequently, SAMGs will be implemented through docketed

commitments from each nuclear power plant licensee..

initiative, by iing SAMGs be maintained. Specifically, this would mea that the SAMGs

would be faintained within the plant configuration management system and re c~t generic

indu• updates and improvements. .

/ 3. Result in the integration of the SAMGs with other guideline sets and with th%

, symptom-based EOPs.". •• •"

The NRC concludes this proposed regulatory structure will result in updated SAMGs that

1) reflect facility configuration, 2) include generic industry guidance, and 3) would be available

for use if a severe accident occurs. -A u,; ,, , ,l• 3AtMC; ............. . n dpth~

•As discussed in the "Regulatory Oversight of Severe Accident Management

31

J
Guidelines" section of this document, the NRC's role regarding the regulatory oversight of

SAMGs would be to ensure this proposed regulatory structure is in place through inspection.

Scope of Procedure and Guideline Integration

This rulemaking limits the scope of the integrated response capability to tb:guideline

sets. This proposed rule includes these new provisions:

1. § 50.155(b)(1), resulting from Order EA-12-049, and addressing beyond-designbasis external events; these requirements are those that the NRC termed in previous regulatory

basis interactions as "Station BlaCkout Mitigation Strategies." The nuclear industry refers to

these as "FLEX Support Guidelines" (FSGs).

2. § 50.155(b)(2) (current § 50.54(hh)(2)). These requirements are defined in

NEI 06-12, Revision 2, "B.5.b Phase 2 & 3 Submittal Guideline," as a subset of the strategies

and guidelines for addressing the loss of large areas of the plant due to explosions and fires and

are termed "Extensive Damage Mitigation Guidelines." The NRC proposes to expand the scope

of the generic term "EDMGs" to include all of the strategies and guidelines used to implement

§ 50.54(hh)(2).

3.- § 50n 5(b)(3, -,o, ,,ify,,,g urntvlnar ni rh initiative i.A

• :rnvioiiJ.h, in t"his doc-, ment, typialyrofolrzd tG B ci;•,tho "SA..I•" or"Sevr "~in .

The NRC is proposing this integrated response capability structure to avoid

unnecessarily revisiting the existing symptom-based EOPs that were developed following the

TMI accident. The NRC has determined that current regulations addressing EOPs, which

include the quality assurance requirements of criterion V, "instructions, Procedures, and

Drawings," and criterion VI, "Document Control," in appendix B to 10 CFR part 50, and the

administrative controls section of the technical specifications for each plant as well as the

32

1
guidance provided in regulatory guides and technical reports (e.g., NUREG-0660, "NRC Action

Plan Developed as a Result of the TMI-2 Accident," issued May 1980; NUREG-0737,

"Clarification of TMI Action Plan Requirements," issued November 1980; and NUREG-071 1,

"Human Factors Engineering Program Review Model," issued November 2012) provide

sufficient regulation and control of the EOPs to provide reasonable assurance of adequate

protection of public health and safety. In addition, the EOPs are the subject of a national

consensus standard (American National Standards Institute/American Nuclear Society 3.2

1994, "Administrative Controls and Quality Assurance for the.Operational Phase of Nuclear

Power Plants.") In order to avoid the unnecessary regulatory burden that would result by

restructuring the EOPs, proposed § 50.1 55(b)(4) would require that the FSGsAEDMGs ,R•-

•SAMbe integrated with the EOPs, rather than moving the requirements for EOPs to

§ 50.155.

Guideline Sets Excluded From the Proposed Rule

During the development of this proposed rule, other guideline sets were considered for

inclusion within the integrated response capability. The guideline sets considered included fire

response procedures, alarm response procedures (ARPs), and abnormal operating procedures

(AO Ps).

Similar to the EOPs, ARPs and AOPs are subject to existing NRC regulations (e.g., 10

CFR part 50, appendix B, criteria V and VI) that adequately ensure integration with other

procedure sets in use at power reactors. These procedures have been used by operating

power reactor licensees in actual and simulated events for many years; any further integration

effort to address potential issues would likely have already been identified and corrected by

existing processes (or will be identified and corrected under the quality assurance program).

33
The issue of whether to include fire response procedures in the scope of proposed

§ 50.155(b) was initially raised as recommendation 1 .g. by the ACRS in its letter to the thenChairman Jaczko dated October 13, 2011, "Initial ACRS Review of: (1) the NRC Near-Term

Task Force Report on Fukushima and (2) Staff's Recommended Actions to be Taken Without

Delay." That letter expressed the ACRS view that:

[The] efforts to integrate the onsite emergency response caPabilities should be

expanded to include the plant fire response procedures. These procedures

provide operator guidance for coping with fires that are beyond a plant's original

design basis. Some plant-specific fire response procedures instruct operators to

manually de-energize major electrical buses and realign fluid systems in

configurations that may not be consistent with the guidance or expectations in

the EOPs. Experience from actual fire events has shown that parallel execution

of fire procedures, Abnormal Operating Procedures (AOPs), and EOPs can be

difficult and can introduce operational complexity. Therefore, these procedures

should also be included in the comprehensive efforts to better coordinate and

integrate operator responses during challenging plant conditions.

This recommendation was reiterated in the ACRS letter of November 8, 2011, "ACRS

Review of Staff's Prioritization of Recommended Actions to Be Taken in Response to

Fukushima Lessons Learned (SECY-1 1-01 37)."

In SECY-12-0025, enclosure 3, the NRC documented the formal process used in

evaluating additional recommendations that were made by the ACRS as follows:

The staff developed a process to disposition all additional issues, including

recommendations by the ACRS. All issues are reviewed by a panel of seniorlevel advisors from different NRC program offices. The panel determines

whether each issue represents a valid safety concern, and whether there is a

clear nexus to the Fukushima Dai-ichi accident. If neither criterion is met, or only

one criterion is met, the panel chooses to either disposition the issue with no

action, or direct it to one of the NRC's existing regulatory processes (e.g., generic

issue process). If both criteria are met, the issue is forwarded for further

consideration by the cognizant technical staff in the appropriate NRC line

organization. Should the issue go forward, the cognizant technical staff is tasked

with developing a proposal for Steering Committee (SC) disposition. The SC

may elect to take no further action, disposition the issue using an existing NRC

process, or prioritize the issue as a Tier 1, 2, or 3 item under the Japan LessonsLearned Program.

34
By letter dated February 27, 2012, the NRC responded to the ACRS recommendations

of October 13, 2011 and November 8, 2011, discussing the disposition of ACRS

recommendation 1l.g. as follows:

The NRC staff evaluated how to appropriately integrate the fire response

procedure into a licensee's onsite emergency response capabilities and

determined that the fire response procedures would be best considered with the

agency's Tier 3 actions associated with NTTF Recommendation 3.

This disposition of the ACRS recommendation also was documented in SECY-1 2-0025.

In its letter of March 13, 2012, the ACRS acknowledged that the formal screening process used

by the NRC for additional recommendations was acceptable, but nevertheless expressed the

view that "[i]rntegration of the fire response procedures presents similar challenges to those

faced by integration of the SAMGs and EOMGs with the EQPs. Hence, this integration effort

should be addressed as part of NTTF Recommendation 8 rather than as a seismic-induced-fire

issue under NTTF Recommendation 3."

Recognizing the continued ACRS interest in the integration of fire response procedures

with onsite emergency actions and the existence of an additional program of work to be taken

up on the ACRS recommendation, the NRC has concluded that the reasoning underlying the

initial prioritization of ACRS recommendation 1 .g was sound and it would be inappropriate to

include fire response procedure integration within this rulemaking effort. The NRC offers the

following reasons for the exclusion of firefighting strategies and procedures from the scope of

integration in this rulemaking:

1. The NRC-required fire protection program is designed to function autonomously

from other ongoing activities and is implemented by a fire brigade that is manned in all modes of

operation and is well-trained. Firefighting activities are led by personnel knowledgeable of

overall plant operations, including the equipment necessary for safe shutdown of the plant.

35
These personnel communicate with the main control room in order to prioritize and deconflict

activities.

2. Comprehensive firefighting strategies and implementing procedures have been

developed for each area of the plant and fire brigade qualified individuals participate in drills on

a quarterly basis to demonstrate proficiency with the use of these strategies and procedures in

the context of concurrent use of other, non-integrated procedures throughout the plant.

3. EOPs, EDMGs, FSGs, and SAMGs account for equipment lost due to concurrent

fires during events by providing alternate methods to accomplish the functions the equipment

was to. have performed.

C. Proposed Rule Construct

To accomplish NRC's rulemaking objectives in a manner consistent with the described

scope, this proposed rule has beern based on these precepts:

1. The central requirement will be an integrated response capability that includes

currently existing procedures and guideline sets adwta~A~"dditional requirements

will support this integrated response capability.

2. The mitigation strategies under Order EA-1 2-049 established the basic

framework for broader capability to mitigate beyond-design-basis external events that impact an

entire reactor site. This framework includes: supporting drills, training, change control, staffing,

communications capability, multi-source term assessment capability, and command and control.

-3-. h',;.y•dLIion of Whis up~ibiiiy irilo NI-U'S requirements hlas Deei, •xtzd--d tdP.• /

;.viud • post-core-damage part of acuidui .tigo... 1 (.z., . Crlb.~ ~

As a result, the proposed new § 50.155 is structured to have:

1. Integrated response requirements in paragraph (b).

36
2. Supporting equipment requirements in paragraph (c) that include equipment

required by both Order EA-12-049 and Order EA-12-051.

3. External hazard equipment protection requirements in paragraph (c) that reflect

the hazard information developed under the § 50.54(f) letter of March 12, 2012.

A. Doign fzourc ... o,"u'i'o"men for .... cnt for•-,.- ,,c,3; now. or3C,- d.ocig,-,. in ppr•Lgr"op•I•_

D(,•. Supporting training, drills, and change control requirements in paragraphs •) (i

$• f/ Requirements that relate to enhanced onsite emergency response capabilities

located in appendix E to 10 CFR part 50, to include a new section VII.

o ./ Implementation requirements that establish compliance deadlines in paragraph

f)i

(h).

8.--• Cu•,irtng new reactor appiication requirements (under eitner U E ot5

pdIL"o• roce•) iii Lh• appropriate content oT appiicuai~iu,,• tcc.

The proposed rule is applicable to both current licensees and new applicants (under

either 10 CFR part 50 or part 52) as established by proposed paragraph (a). Additionally, the

proposed rule contains provisions to facilitate power reactor decommissioning.

0. Proposed Rule Regulatory Bases.

Applicability

The proposed rule would apply, in whole or in part, to applicants for and holders of a

construction permit or operating license for a nuclear power reactor under 10 CFR part 50, or

combined license under 10 CFR part 5:a•, dl..jL z tnad ~~g ct...~i~

c..t.'nd,•rd dcaig a ,pol•'. And rngnufanh i~rinp li~it- "nder• 10.....rt5

37 1
This proposed rule would not apply to holders of an operating license for a non-power

reactor under 10 CFR part 50. Non-power reactor licensees would not be subject to this

proposed rule because non-power reactors pose lower radiological risks to the public from

accidents than do power reactors because: 1) the core radionuclide inventories in non-power

reactors are lower than in power reactors as a result of their lower power levels and often

shorter operating cycle lengths; and 2) non-power reactors have lower decay heat associated

with a lower risk of core melt and fission product release in a loss-of-coolant accident than

power reactors.

A holder of a general or specific 10 CFR part 72 independent spent fuel storage

installation (ISFSI) license would not be subject to this proposed rule for the ISFSI, because the

decay heat load of the irradiated fuel would be sufficiently low prior to movement to dry cask

storage that it could be air-cooled. This would meet the proposed sunsetting criteria (discussed

later in this section of this document,).

The GE Morris facility in Illinois consists of a spent fuel pool licensed under

10 CFR part 72 as an ISESI and consequently would not need to comply with this proposed

rule. The NRC considered including the GE Morris facility within the scope of this rule but found

that the age (and corresponding low decay heat load) of the fuel in the facility made it

unnecessary. The GE Morris facility also would meet this proposed rule's sunsetting criteria.

While the proposed rule would leave in force the requirements of the current § 50.54(hh)(2),

those requirements are not applicable to GE Morris due to its status as a non-part 50 licensee.

In the course of the development and implementation of the guidance and strategies required by

the current § 50.54(hh)(2), the NRC evaluated whether additional mitigation strategies were

warranted at GE Morris and concluded that no mitigating strategies were warranted beyond

existing measures, due to the extended decay time since the last criticality of the fuel stored

there, the resulting low decay heat levels, and the assessment that a gravity drain of the GE

38
Morris SEP is not possible due to the low permeability of the surrounding rock and the high level

of upper strata groundwater.

This proposed rule would establish a "sunsetting" or phased removal of requirements for

licensees of decommissioning power reactors. Licensees would not need to meet requirements

that relate to the reactor source term and associated fission product barriers once all fuel has

been permanently removed from the reactor vessel and placed in the spent fuel pool. This

proposed rule would require secondary containment for reactor designs that employ this feature

as a fission product barrier for the spent fuel pool source term.

Once the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certification of

permanent removal of fuel from the reactor vessel and certification of permanent cessation of

operations, that licensee would not be subject to requirements to have mitigation strategies and

guidelines for maintaining or restoring core cooling and containment capabilities. As discussed

previously, these proposed requirements are based on Order EA-12-049. The licensees for the

Kewaunee Power Station, Crystal River Unit 3 Nuclear Generating Plant, San Onofre Nuclear

Generating Station Units 2 and 3, and Vermont Yankee Nuclear Power Sta~p~t~bnitted

§ 50.82(a)(1) certifications after issuance of Order EA-12-049; the NRC ha~kre:inde Order

EA-12-049 to this group of NPP licensees (Shutdown NPP Group). These rescissions were

based on the NRC's conclusion that the lack of fuel in the licensee's reactor core and the

absence of challenges to the containment rendered unnecessary the development of guidance

and strategies to maintain or restore core cooling and containment capabilities. Consistent with

these rescissions, the NRC proposes to relieve licensees in decommissioning from the

requirement to comply with proPOsed requirements tO have mitigation strategies and guidance

to maintain or restore core cooling and containment capabilities. Moreover, these licensees

would not need to comply with any of the requirements in this proposed rule based on

39
compliance with the proposed requirement to have mitigation strategies and guidelines for

maintaining or restoring core cooling and containment capabilities.

This proposed rule treats the EDMG aR,,6M~requrements in a manner Similar to the

requirements for FSGs. For a licensee who has § 50.82(a)(1) or § 52.110(a) certifications

docketed at the NRC, the lack of fuel in their reactor core and the absence of challenges to the

containment would render unnecessary EDMGs for core cooling and containment capabilities

md AM2~rz~td ,. felinth r.co "e~This licensee would not need to comply with

any of the requirements in this proposed rule; rather, the licensee would be required to comply

with the proposed requirement to have EDMGs erBf~ased on the presence of fuel in J

the reactor vessel.

Once the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certifications, that

licensee would not need to comply with the requirements proposed by this rule that the

equipment relied on for the mitigation strategies include reliable means to remotely monitor

wide-range spent fuel pool levels to support effective prioritization of event mitigation and

recovery actions. These proposed requirements are based on the requirements in Order

EA-12-051. This order requires a reliable means of remotely monitoring wide-range SEP levels

to support effective prioritization of event mitigation and recovery actions in the event of a

beyond-design-basis ext a, vent with the potential to challenge both the reactor and SFP.

The NRC has also rkied Order EA-12-051 for the Shutdown NPP Group mentioned

previously. These rescissions were based, in part, on the NRC's conclusions that once a

licensee certifies the permanent removal of the fuel from its reactor vessel, the safety of the fuel

in the SFP becomes the primary safety function for site personnel. In the event of a challenge

to the safety of fuel stored in the SFP, decision-makers .would not have to prioritize actions and

the focus of the staff would be the SEP condition. Thus, once fuel is permanently removed from

40
the reactor vessel, the basis for the Order EA-1 2-051 would no longer apply. Consistent with

the NRC order rescissions, the NRC proposes to no longer require licensees in

decommissioning to have a reliable means to remotely monitor wide-range spent fuel pool

levels to support effective prioritization of event mitigation and recovery actions in the event of a

beyond-design-basis external event with the potential to challenge both the reactor and SEP.

Once the NRC has docketed a licensee's § 50.82(a)(1) or § 52.110(a) certifications, that

licensee would not need to comply with the requirements in proposed Section VII,

"Communications and Staffing Requirements for the Mitigation of Beyond Design Basis Events,"

in 10 CFR part 50, appendix E. These proposed requirements are based on the

March 12, 2012, § 50.54(f) letters that requested operating power reactor licensees to perform,

among other things, emergency preparedness communication and staffing evaluations for

prolonged loss of power events consistent with NTTF recommendation 9.3. Once the licensees

for the Shutdown NPP Group were no longer operating power reactors, they informed the NRC

that they would no longer proceed with implementing recommendation 9.3. In response to the

filings, the NRC determined that, for beyond-design-basis external events challenging the safety

of the spent fuel at the Shutdown NPP Group:

recovery and mitigation actions could be completed over a long period of time

due to the slow progression of any accident as a result of the very low decay

heat levels present in the pool within a few months following permanent

shutdown of the reactor. Thus, spent fuel pool beyond design basis accident

scenarios at decommissioning reactor sites do not require the enhanced

communication and staffing that may be necessary for the reactor-centered

events the 50.54(f) letter addresses.4

Order EA-12-049 also required power reactor licensees to have certain spent fuel pool

cooling capabilities. In the rescission letters to the licensees for the Shutdown NPP Group, the

4 See the "Availability of Documents" section of this document for the NRC letters to the licensees for Kewaunee

Power Station, Crystal River Unit 3 Nuclear Generating Plant, San Onofre Nuclear Generating Station Units 2 and 3,

and Vermont Yankee Nuclear Power Station.

41
NRC determined that, due to the passage of time~fuel's low decay heat and the long time to boil

off the water inventory in the spent fuel pool obviated the need for the Shutdown NPP Group

licensees to have guidance and strategies necessary for compliance with Order EA-1 2-049.

The rescission of Order EA-12-049 for those licensees eliminated the requirement for them to

comply with the Order's requirements concerning beyond-design-basis event strategies and

guidelines for spent fuel pool cooling capabilities. Consistent with the basis for the Order

rescissions, licensees in decommissioning could be relieved from the proposed requirements

concerning beyond-design-basis event strategies and guidelines for spent fuel pool cooling

capabilities and any related requirements. These licensees would have to perform and retain

an analysis demonstrating that sufficient time has passed since the fuel within the spent fuel

pool was last irradiated such that the fuel's low decay heat and boil-off period provide sufficient

time for the licensee to obtain offsite resources to sustain the spent fuel pool cooling function

indefinitely. Similarly, the proposed rule would "lluv ftcC;,MC •m'! _, ,iuu;-,~i t,, 9 r,

-1ee~~~the requirements for EDMGs in place. Licensees could make use of the equipment

in place for EDMGs should that equipment be available, recognizing that the protection for that

equipment is against the hazards posed by events that result in losses of large areas of the

plant due to fires or explosions rather than beyond-design-basis external events resulting from

natural phenomena. If the EDMG equipment is not available, the offsite resources would be

used by the licensee for only onsite emergency response (i.e., spent fuel pool cooling). This

proposed amendment would not impact any commitments licensees have made regarding

exemptions from offsite emergency planning requirements, which consider a beyond-designbasis event that could result in a zirconium cladding fire due to a loss of SEP inventory and do

not consider offsite resources in mitigation strategies.

The NRC proposes-to maintain the EDMGs requirement, because an event for which

EDMGs would be required is not based on the condition of the fuel, but may instead result from

42
aircraft impact and a beyond-design-basis security event which could introduce kinetic energy

into the spent fuel pool independent from the decay heat of the fuel. These types of events and

their potential consequences were considered as a part of the rulemaking dated March 7, 2009,

on Power Reactor Security Requirements (74 FR 13926). In the course of that rulemaking, the

NRC took into account stakeholder input and determined that it would be inappropriate to apply

the EDMG requirements to permanently shutdown and defueled reactors where the fuel was

removed from the site Or moved to an ISFSI. However the resulting rule was written to remove

the EDMG requirements once the certifications of permanent cessation of operations and

removal of fuel from the reactor vessel were submitted rather than upon removal of fuel from the

SFP. The NRC proposes to correct this error from the 2009 final rule in the proposed rule as

explained in the "EDMGs" portion of this section.

The NRC proposes to exclude from the proposed requirements the licensee for Millstone

Power Station Unit 1,. Dominion Nuclear Connecticut, Inc. Dominion Nuclear Connecticut, Inc.

is also the licensee for Millstone Power Station Units 2 and 3, but this exclusion would apply to

Dominion Nuclear Connecticut, Inc. in its capacity as licensee for only Unit 1, which is not

operating but has irradiated fuel in its spent fuel pool and satisfies the proposed criteria for not

having to comply with the proposed rule except for the EDMG requirements. In the course of

the development and •implementation of' the guidance and strategies required by current

§ 50.54(hh)(2), the NRC evaluated whether additional mitigation strategies were warranted at

Millstone Power Station Unit 1 and concluded that no mitigating strategies were warranted

beyond existing measures, principally due to the extended decay time since the last criticality

there on November 4, 1995, and the resulting low decay heat levels allowing sufficient time for

the use of existing strategies augmented by mitigation strategies existing in 2005. The

exclusion for Millstone Power Station Unit 1 in this proposed rule is based upon that conclusion-,

43
recognizing that additional mitigating capabilities will be present due to the implementation of

the § 50.54(hh)(2) strategies at the collocated Millstone Power Station Units 2 and 3.

In contrast to Millstone Power Station Unit 1, the Shutdown NPP Group licensees were

issued license conditions for the mitigating strategies corresponding to the § 50•54(hh)(2)

strategies. These license conditions are condition 2.C.(10) to Renewed Operating License No.

DPR-43 for Kewaunee Power Station, condition 2.C.(14) to Facility Operating License No.

DPR-72 for Crystal River Unit 3 Nuclear Generating Plant, condition 2.C.(26) to Facility

Operating License NPF-1 0 for San Onofre Nuclear Generating Station Unit 2, condition 2.C.(27)

to Facility Operating License NPF-15 for San Onofre Nuclear Generating Station Unit 3, and

condition 3.N to Renewed Operating License No. DPR-28 for Vermont Yankee Nuclear Power

Station. Those licensees and future power reactor licensees that enter decommissioning would

have the burden to show that operation in a decommissioning status with irradiated fuel in the

spent fuel pool without the EDMG license condition or the proposed requirement to comply with

the proposed EDMG requirement would provide adequate protection of public health and

safety.

Integrated Response Capability

Each applicable applicant or licensee would be required to develop, implement, and

maintain an integrated response capability that includes FSGs, EDMGs, SA~'Ps,(

sufficient staffing, and a supporting organizational structure with defined roles, responsibilities,

and authorities for directing and performing these strategies, guidelines, and procedures.

As discussed in the NTTF Report, EOPs have long been part of the NRC's safety

requirements. The NRC regulations address them through the quality assurance requirements

of criterion V and criterion VI in appendix B to 10 CFR part 50, and in the administrative controls

section of the technical specifications for each plant. Following the accident at TMI Unit 2,

44
EOPs were upgraded to address human factors considerations in order to improve human

reliability including the operator's ability to mitigate the consequences of a broad range of

initiating events and subsequent multiple failures without the need to diagnose specific events.

In other words, EOPs were modified from their previous event-driven nature to be symptombased. Numerous subsequent regulatory guides (RGs) and technical reports (e.g., NUREG0660, NUREG-0737, and NUREG-071 1) also address EOPs. In addition, the EOPs are the

subject of a national consensus standard (American National Standards Institute/American

Nuclear Society 3.2-2012, "Administrative Controls and Quality Assurance for the Operational

Phase of Nuclear Power Plants"). The subject matter for the initial and requalification training,

written exam, and operating test for reactor operators and senior reactor operators also

includes the EOPs. While implementing EOPs, the event command and control functions

remain in the control room under the direction of the senior licensed operator on shift.

The NTTF Report also discusses the development of SAMGs by the nuclear industry,,

durnn e 1980s and 1990s in response to the TMI accident and follow-up activities. T Je

follow-up activ nlded extensive research and study (including several pro ilistic risk

assessments) on severe idns and severe accident phenomena. D ending upon the ;..

Simplementation of owners group g anc by particular licensee'he SAMGs were written for,

Suse by either plant technical support staff, l oaenthe plant's Technical Support

o perators to manage accident seqe sta progress beyn e ont where EOPs and other

Splant procedures are applia and useful. The EOPs typically cover a ~ens to the point of

45
3i••f•.rN;•uii•wi•. L~IdL inciupe imminent or actual signiTIcaniLu dIIa'- t• th... f'=ucl,' in' ther •~trn c

The nuclear industry developed EDMGs following the terrorist events of September 11,

2001, in response to security advisories, orders, and license conditions issued by the NRC that

required licensees to develop and implement guidance and strategies intended to maintain or

restore core cooling and containment and spent fuel pool cooling capabilities under the

circumstances associated with the loss of large areas of the plant due to fire or explosion. The

EDMGs further extend the range of initiating events and plant damage states for which

strategies and guidelines are available for use by operators to include the loss of large areas of

the plant and a subsequent impairment of the operability and functionality of structures, systems

and components that are within that area. NEI 06-12, "B.5.b Phase 2&3 Submittal Guideline,"

Revision 2, December 2006 (the NRC-endorsed guidance for the requirements associated with

EDMGs) provides that the EDMGs "must be interfaced with. existing SAMGs so that potential

competing considerations associated with implementing these and other strategies are

appropriately addressed."

Based upon these considerations, the NTTF recommended that the NRC require

licensees to further integrate EOPs, SAMGs and EDMGs, including a clarification of transition

points, command and control, decision making, and rigorous training that includes conditions

'that are as close to real accident conditions as feasible.

Subsequent to issuance of the NTTF Report, the range of initiating events and plant

damage states for which strategies and guidelines are available for use by operators was further

extended through the development of mitigating strategies for beyond-design-basis external

events in response to Order EA-12-049. The development and implementation of this set of

strategies and guidelines was accomplished With the knowledge of the existence of the other

NTTF recommendations and took them into account to the extent practical. In order to provide

46
better integration with the EOPs e sutnStrategies and guidelines (FSGs)

leave the designation of command and control and decision-making functions within the EOPs®;

-e-.•• .• •. sr/omne in the NTTF Report, this proposed rule would

require that EDMGs, , ,and FSGs be integrated with EOPs, consistent with the

expectation that EOPs remain the central element of a licensee's initial response capability.

In establishing a requirement for a response capability that encompasses the use of

EOPs,• DM"'MGs, and FSGs, the NRC considered the fact that these strategies,

guidelines and procedures were, and are currently being, developed at separate times over a

period of several decades and that the associated efforts •have been focused on responding to

different types of initiating events and plant damage states. -A-t-uyh• iiuot .cftz Ldwi,•

.... As a result, these strategies, guidelines and

procedures may not properly reflect consideration of the interfaces (e.g., procedure transitions),

dependencies (e.g., reliance on common systems or resources) and interactions (e.g.,

alignment of response strategies) among strategies, guidelines and procedures that may be

used in combination, either consecutively or concurrently, to mitigate a design-basis or beyonddesign-basis event.

Additionally, the NRC considered that these strategies, guidelines and procedures are

not used by a single licensee organizational unit but will Often require coordination and transfer

of responsibilities amongst licensee organizational units. For example, •,;tluuyli r-QP dLu,,ti-i -,

-a UhI~Ui~ed by .iiembers ot tne main control room lMu;R) compie~iii,,Ptf ',ion.-,d opor~tors, '•-.

-directicn' f.,,, SAivik, actions may begin in thle MLAt bui. L•.,,I;;.;I Lo ai1d te c-"ondu;tcd pr:,c'vI',&4&-l•

eEDMGs may be implemented under conditions of loss oft'•CV .)a nd

therefore initiated and directed by knowledgeable and available site personnel until coordination

and augmentation efforts enable transition to a more stable command and control structure.

47
The mitigation strategies for extreme external events, though initiated by the MCR complement

of licensed operators, may require coordination with and augmentation by offsite organizations.

Further, and as noted previously, there are potential accident scenarios in which a licensee

might employ strategies from more than one of these strategies, guidelines and procedures

during its response to an accident. One plausible sequence is for an initial response to be

under the EOPs, supplemented by actions under the FSGs, and ultimately transition to actions

under the SAMGS) Such an accident progression would engage and require the coordination ofj

multiple licensee organizational units.

In light of the preceding considerations, the proposed rule would require that the

mitigating strategies, guidelines and procedures, staffing, and supporting organizational

structure be developed, implemented, and maintained such that they function as an "integrated"

response capability. The intent is to ensure that applicants and licensees establish and

maintain a functional capability to produce a coordinated and logical response under a wide

range of accident conditions. The intent is not to require physical integration (e.g., organizations

need not be merged and strategies, guidelines and procedures need not be combined), but

rather to require a functional integration of the elements of the response capability. To achieve

this functional integration, the NRC expects that applicants and licensees will have addressed

the interfaces, dependencies, and interactions among the elements of their response capability

such that elements work together to support effective performance under the full range of

accident conditions. For example, functional integration of the strategies, guidelines and

procedures would ensure that transition points are explicitly identified and conflicts between

strategies are eliminated to the extent practical. Functional integration of response

organizations would ensure that organizations working together to use these strategies,

guidelines, and procedures (e.g., to coordinate actions or provide support) have clearly defined

48
lines of communication between the organizations, as well as clearly defined authorities and

responsibilities relative to each other, such that there are no gaps or conflicts.

The proposed requirements for FSGs would make generically-applicable requirements

previously imposed on licensees by Order EA-12-049 and for Virgil C. Summer Nuclear Station

Units 2 and 3 by license condition as described in Memorandum and Order CL1-12-095 . These

proposed requirements would provide additional defense-in-depth measures that increase the

capability of nuclear power plant licensees to mitigate consequences of beyond-design-basis

external events. Consistent with Order EA-12-049 and associated license conditions, these

proposed provisions would be made generically-applicable in recognition that beyond-designbasis events have an associated significant uncertainty, and that the NRC concluded additional

measures were warranted in light of this uncertainty.

The proposed FSG strategies and guideline' requirements are intended to mitigate

consequences of beyond-design-basis external events from natural phenomenon that result in

an ELAP concurrent with either a loss of normal access to the ultimate heat sink, or for passive

reactor designs, a loss of normal access to the normal heat sink. Recognizing that beyonddesign-basis external events are fundamentally unbounded, and that these events can result in

a multitude of damage states and associated accident conditions, a significant regulatory

challenge is developing bounded requirements that meaningfully address the regulatory issue.

From a practical standpoint, development of mitigation strategies requires that there be some

definition (or boundary conditions established) for an onsite damage state for which the

strategies would then address and thereby provide an additional capability to mitigate beyonddesign-basis external event conditions that might occur. The damage state should ideally be

representative of a large number of potential damage states that might occur as a result of

s Summer, CL1-12-09, 75 NRC at 440, and the V.C. Summer Unit 2 license, License No. NPF-93, Condition 2.D.(13)

and V.C. Summer Unit 3 license, License No. NPF-94, Condition 2.D.(13).

49
extreme external events, and it should present an immediate challenge to the key safety

functions, so that the resultant strategies actually improve safety. The assumed damage state

for this proposed rule is the same as that assumed to implement the requirements of

EA-1 2-049, attachment 2 for currently operating power reactors: an ELAP condition concurrent

with loss of normal access to the ultimate heat sink (LUlHS). This assumed damage state is

effective at immediately challenging the key safety functions following a beyond-design-basis

external event (i.e., core cooling, containment and spent fuel pool cooling). Requiring strategies

to maintain or restore these key functions under such circumstances would result in an

additional mitigation capability consistent with the Commission's objective when it issued Order

EA-1 2-049.

The proposed rule would not be prescriptive in terms Of the specific set of initial and

boundary conditions assumed for the ELAP and LUHS condition, recognizing that the damage

state for current operating reactors, defined in more detail in draft regulatory guidance for this

proposed rule (DG)-1 301, "Flexible Mitigation Strategies for Beyond-Design-Basis Events,"

reflects current operating power reactor designs and the reliance of those designs on ac power,

while the assumed damage state for a future design may be different depending upon the

design features. Specifically, this damage state was implemented through the assumption of

the ELAP to the onsite emergency ac buses, but did allow for ac power from the inverters to be

assumed available in order to establish event sequence and the associated times for when

mitigation actions would be assumed to be required. To address the Order EA-i2-049

requirement for an actual loss of all ac power, including ac power from the batteries (through

inverters), contingencies are included in the mitigation strategies to enable actions to be taken

under those circumstances (e.g., sending operators to immediately take manual control over a

non ac-powered core cooling pump). As such, this proposed provision is meant to make

generically-applicable the current implementation under EA-12-049 (i.e., there is no intent to

50
¢

either relax or impose new requirements), and be performance-based to allow some flexibility

for future designs. As an example, some reactor designs (e.g., Westinghouse APIO000 and

General Electric Economic Simplified Boiling Water Reactor (ESBWR)) use passive safety

systems to meet NRC requirements for maintaining key safety functions. The inherent design of

those passive safety systems makes certain assumptions, such as loss of access to the ultimate

heat sink, not credible. Accordingly, the assumed condition for the FSG requirements for

passive reactors is the loss of normal access to the normal heat sink, discussed further below.

Nevertheless, in this proposed rule the NRC is requiring that the strategies and guidelines be

capable of implementation during a loss of all ac power.

Regarding the assumed LUHS for combined licenses or applications referencing the

API1000 or the ESBWR designs, the assumption was modified to be a loss of normal access to

the normal heat sink (see attachment 3 to Order EA -1 2-049, Summer, CL1-12-09, 75 NRC at

440, the V.C. Summer Unit 2 license, License No. NPF-93, Condition 2.D.(13), and the V.C.

Summer Unit 3 license, License No. NPF-94, Condition 2.D.(13)). This modified language

reflects the passive design features of the AP1 000 and the •SW that jA

provide core cooling, containment, and spent fuel cooling capabilities for 72 hours without I

reliance on ac power. These features do not rely on access to any external water sources for

the first 72 hours because the containment vessel and the passive containment cooling system

serve as the safety-related ultimate heat sink for the AP1000 design and the isolation condenser

system serves as the safety-related ultimate heat sink for the ESBWR design.

As discussed previously, the range of beyond-design-basis external events is

unbounded. These proposed provisions are not intended, and should not be understood to

me@, at the mitigation strategies can adequately address all postulated beyond-design-basis

external events. It is always possible to postulate a more severe event, with greater damages

and for which the mitigation strategies may not be able to maintain or restore the functional

51
capabilities (e.g., meteorite impact). Instead, the proposed requirements provide additionalmitigation capability in light of uncertainties associated with external events, consistent with the

NRC'.s regulatory objective when it issued Order EA-12-049.

This proposed rule would require that the FSGs be capable of being implemented sitewide. This recognizes that severe external events are likely to impact the entire reactor site,

and for multi-unit sites, damage all the power reactor units on the site. This requirement means

that there needs to be sufficient equipment and supporting staff to enable the core cooling,

containment, and spent fuel pool cooling functions to be maintained or restored for all the power

reactor units on the site. This is a distinguishing characteristic of this set of mitigating strategies

from those that currently exist for § 50. 54(hh)(2), for which the damage state was a more

limited, albeit large area of a single plant, reflecting the hazards for which that set of strategies

was developed.

The NRC gave consideration to whether there should be changes made to § 50.63 to

link those requirements with this proposed rule. This consideration stemmed from

recommendation 4.1 of the NTTF Report to "initiate rulemaking to revise 10 CFR 50.63" and the

understanding that the proposed rule could result in an increased station blackout coping

capability, in addition to the regulatory objective of the proposed provisions, which is to provide

additional beyond-design-basis external event mitigation. Because of the substantive

differences between the requirements of § 50.63 for licensees to be able to withstand and

recover from a station blackout and the proposed requirements, the NRC determined that such

a linkage was not necessary and could lead to regulatory confusion.

The principal regulatory objective of § 50.63 was to establish station blackout coping

durations for a specific scenario (i.e., loss-of-offsite power coincident with a failure of both trains

of emergency onsite ac power, typically, the failure of multiple emergency diesel generator In

meeting this regulatory objective, the NRC recognized that there would be safety benefits

52
accrued through the provision of an alternate ac source diverse from the emergency diesel

generators and therefore defined such a source in § 50.2. In furtherance of this alternative

means to comply with § 50.63, the NRC also defined the event a licensee must withstand and

recover from as a station blackout rather than a loss of all ac power. A station blackout allows

for continued availability of ac power to buses fed by station batteries through inverters or by

alternate ac sources. This proposed rule would provide an additional capability to mitigate

beyond-design-basis external events. Because the condition assumed for the mitigation

strategies to establish the additional mitigation capability includes an ELAP, which is more

conservative than a station blackout as defined in § 50.2, there can be a direct relationship

between the two different sets of requirements with regard to the actual implementation at the

facility. Specifically, implementation of the proposed mitigation strategies links into the station

blackout procedures (e.g., the applicable strategies would be implemented to maintain or

restore the key safety functions when the EOPs reach a "response not obtained" juncture).

Step-by-step procedures are not necessary for many aspects of the proposed mitigating

strategies and guidelines. Rather, the strategies and guidelines should be flexible, and thus

enable plant personnel to adapt them to the conditions that result from the beyond-design-basis

external event. The proposed provisions typically would result in strategies and guidelines that

use both installed and portable equipment, instead of only relying on installed ac power sources

(with the exception of protected battery power) to maintain or restore core cooling, containment,

and spent fuel pool cooling capabilities. By using equipment that is separate from the normal

installed ac-powered equipment, the strategies and guidelines have a diverse attribute. By

having available multiple sets of portable equipment that can be deployed and used in multiple

ways depending on the circumstances of the event, operators are able to implement strategies

and guidelines that are flexible and adaptable.

53
The proposed mitigation strategies requirements are both performance-based and

functionally-based. •e rgormance-based requirements~r r~ oo ni•1~ that the new

requirements would provide most benefit to future reactors whose designs could differ

significantly from current power reactor designs and as such, use of more prescriptive

requirements could be problematic and create unnecessary regulatory impact and need for

exemptions. Use of functionally-based requirements results from the need to have

requirements that can address a wide range of damage states that might exist following beyonddesign-basis external events. Maintaining or restoring three key functions (core cooling,|

containme~,•d spent fuel pool cooling) supports maintenance of the fission product barriers

(i.e., fuel clad, reactor coolant pressure boundary, and containment) and results in an effective

means to mitigate these events, while remaining flexible such that the strategies and guidelines

can be adapted to the damage state that occurs. Functionally-based requirements also result in

strategies that align well with the symptom-based procedures used by power reactors to

respond to accidents. Accordingly, Order EA-12-049 contained requirements for a threephased approach for current operating reactors. This proposed rule does not specify a number

of phases; instead, the NRC is proposing higher level, performance-based requirements

consistent with this discussion.

The NRC gave consideration to incorporating into the proposed rule a requirement that

licensees be capable of implementing the strategies and guidelines "whenever there is

irradiated fuel in the reactor vessel or spent fuel pool." This provision would have been a means

of making generically-applicable the requirement from Order EA-12-049 that licensees be

capable of implementing the strategies and guidelines "in all modes." The NRC considers the

terminology "whenever there is irradiated fuel in the reactor vessel or spent fuel pool" would be

a better means to address the/r'rder requirement since it does not use technical specification

type language (i.e., mode~~hich would not be in effect when a licensee offloads the fuel into)

LI/ 54
the spent fuel pool during an outage. The NRC concluded that the use of the phrases

"whenever there is irradiated fuel in the reactor vessel or spent fuel pool" or "in all modes" is not

necessary because the proposed applicability provisions would ensure that licensees will be

required to have mitigation strategies for beyond-design-basis external events for the various

configurations that can exist for the reactor and spent fuel pools throughout the operational,

refueling and decommissioning phases.

The mitigation strategies and guidelines implemented under NRC Order EA-12-049

assume a demanding condition that maximizes decay heat that would need to be removed from

the reactor core and spent fuel pool source terms on site. This implementation results in a more

restrictive timeline (i.e., mitigation actions required earlier following the event to take action to

maintain or restore cooling to these source terms) and a greater resulting additional capability.

These assumed at-power conditions are 100 days at 100 percent power prior to the event for

the reactor core as was used for § 50.63. This assumption establishes a conservative decay

heat for the reactor source term. The assumed spent fuel pool conditions include the design

basis heat load for the spent fuel pool, typically a full core offload following a refueling outage.

This establishes a conservative heat load for the spent fuel pool. The NRC recognizes th s a

practical re~ijt)ese conditions would not exist simultaneously. The NRC considers the

development of timelines for the proposed mitigating strategies using the maximum heat load

for either the reactor core or the spent fuel pool to be appropriate. While establishing the

capability to mitigate the maximum heat load for both simultaneously would be compliant with

the proposed requirements, it would not be necessary.

The NRC recognizes the difficulty of developing engineered strategies for the

extraordinarily large number of possible plant and equipment configurations that might exist

under shutdown conditions (i.e., at shutdown when equipment may be removed from service,

when there is ongoing maintenance and repairs or refueling operations, or modifications are

55
being implemented). The proposed requirements mean that licensees should be cognizant of

such configurations, equipment availability, and decay heat states that could present greater

challenges under these conditions, and design mitigation strategies that they can be

implemented under such circumstances.

The NRC considered requiring the strategies to be developed considering the need to

plan for delays in the receipt of offsite resources as a result of damage to the transportation

infrastructure. While severe events could damage local infrastructure, and could create

challenges with regard to the delivery of offsite resources, the NRC concluded that having this

level of specificity in the proposed provisions would not be necessary. lnste.This proposed

rule contains provisions that are more performance-based, requiring continued maintenance or

restoration of the functional capabilities until acquisition of offsite assistance and resources.

Potential delays and other challenges presented by extreme events that affect acquisition and

use of offsite resources would be addressed by licensee programs that implement the proposed

provisions.

Order EA-12-049 included a requirement that licensees develop guidance and strategies

to obtain "sufficient offsite resources to sustain [the functions of core cooling, containment, and

spent fuel pool cooling] indefinitely." The NRC considered using this language in the proposed

rule, but concluded that this would be better phrased as "indefinitely, or until sufficient site

functional capabilities can be maintained without the need for the mitigation strategies." The

NRC concluded that this phrase better communicates the existence of a transition from the use

of the mitigating strategies to recovery operations.

The NRC recognizes that the use of the proposed mitigating strategies would potentially

require departure from a license condition or a technical specification (contained in a license

issued under 10 CFR part 50 or 52) and could be considered a procedurelization of the

allowance provided under § 50.54(x). Given that the initiation of the use of these strategies may

56
be included in emergency operating procedures or other procedures, which might be considered

procedures described in the final safety analysis report (as updated), there is an interaction with

the provisions of § 50.59(c)(1) regarding the need to obtain a license amendment in order to

make the necessary change to those procedures. The NRC considered including provisions in

the proposed rule s pecificaf., allow departures from license conditions or technical)

specifications in order to clarify this situation, but found these provisions unnecessary. For

holders of operating licenses under 10 CFR part 50 and combined licenses under 10 CFR part

52 that were subject to Order EA-1 2-049, the provisions of that Order provided more specific

criteria for making the necessary changes than § 50.59, making that section inapplicable as set

forth in § 50.59(c)(4). Those criteria included the provision of an overall integrated plan to the

NRC for review. Similar criteria were included in license conditions for the combined licenses

for Virgil C. Summer Nuclear Station, Units 2 and 3.

EDMGs

The NRC proposes to move the EDMGs requirement currently in § 50.54(hh)(2) to a

new mitigation of beyond-design-basis events section of 10 CFR part 50. In addition to moving

the text, the NRC proposes to make a few editorial changes. The wording used to describe

these requirements has evolved from "guidance and strategies," in Interim Compensatory

Measures Order EA-02-026, dated February 25, 2002, to "strategies," in the corresponding

license conditions, to "guidance and strategies," in § 50.54(hh)(2), to its proposed form

"strategies and guidelines." The word "guidelines" was chosen rather than "guidance" to better

reflect the nature of the instructions that could be developed as appropriate by a licensee and to

avoid confusion with the term "regulatory guidance." The word "strategies" is used in the

proposed rule to reflect its meaning, "plans of action." The resulting plans of action could

include plant procedures, methods, or other guideline documents, as deemed appropriate by

57
the licensee during the development of these strategies. These plans of action would also

include the arrangements made with offsite responders for support during an actual event. No

substantive change to the requirements is intended by this proposed change in the wording.

Applicability of the requirements of § 50.54(hh)(2) is currently governed by

§ 50.54(hh)(3), which makes these requirements inapplicable following the submittal of the

certifications required under § 50.82(a) or § 52.1 10(a)(1). As discussed in the statement of

considerations for the Power Reactor Security Rulemaking (74 FR 13926), the NRC believes

that it would be inappropriate for the requirements for EDMGs to apply to a permanently

shutdown, defueled reactor, where the fuel was removed from the site or moved to an ISFSI.

The NRC proposes to require EDMGs for a licensee with permanently shutdown defueled

reactors, but with irradiated fuel still in its spent fuel pool, because the licensee must be able to

implement effective mitigation measures for large fires and explosions that could impact the

spent fuel pool while it contains irradiated fuel. The difference between the proposed rule and

§ 50.54(hh)(3) would correct the wording of the latter provision to implement the sunsetting of

the associated requirement as was intended by the Commission in 2009. This change would

not constitute backfitting for currently operating reactors because the proposed change

concerns decommissioning reactors. The proposed change would not constitute backfitting for

currently decommissioning reactors because the EDMGs are also required by the licensees'

license conditions that were made generically applicable through the Power Reactor Security

Rulemaking and remain in effect.

Nu ar plants redesigned suh t design sis events andccidents/t t res

chalges to, n"a, or shutdownant opera*s should no I to signiict damage tfue•"-

58t
or ccident results in conditions in which the emergency operating procedures are u uccessful

in pre nting the conditions in which imminent or actual significant fuel damage/ y occur, then

the emergncy operating procedures and other procedures or guidelines (e.g.EDMGs, AOPs

(for spent fuerool events, which were added to SAMGs following the Fukhima accident))

provide transition iteria to implement the SAMGs./

The NRC prop, se to require strategies and guidelines for se by decision-making

personnel under conditiorin which significant damage to fuel* the reactor vessel or spent fuel

pool is occurring or is immineh Although there is not a sndard definition for "significant

damage" as related to fuel in the actor vessel or sp t fuel pool, in this context it means that

fex istencaed ofran inadequate cooicndtnfor ~theactore coeirspn fuel poolgsuh hat ts oulrdin

lhea ToI- significnt fedaan .in tryFksia•i ciet.Ti conditinwoley beplnt-peific flor reexamped low

reactorsevessel wciet hstatergeeineWs orehighsar corextemperl uresforpressurizedSwater

The pro sedr es uleg woud reuire licnfusee w keto "ma y ii~sintaina the Pategea ndgidelns,

which mea that licensees would be required to update their SAMGs c-sistent with the_

genertechnical guideline documents applicable to their facility or ensure th * ie-specific

59
congu on mana enprr 0 m in order toeure that, th S G refle ang the

supporting the,,rategies and uidelines in •fporated int he SAMGs.*

Integration with EO~s•,••

In developing a proposed requirement for the integration of FSGs/EDMGs, a A/&• 

with the EOPs, the NRC considered their differences in content and the standards for usage

applied to them. The EOPs are a specific and prescribed set of instructions implemented in

accordance with exacting standards for usage and adherence (e.g., step-by-step .sequential

performance, concurrent execution of multiple sections) that operators and plant staff are

required to follow when performing a specific task or addressing plant conditions. When

implementing procedures, each step is to be performed as prescribed, with rare exceptions.

The strategies and guidelines that would be required differ from EOPs primarily in terms of the

level of detail to which they are written and expectations regarding usage. These strategies and

guidelines may be a less prescriptive set of instructions not subject to the same constraints

imposed by standards of usage for procedure implementation (e.g., may not be followed in a

step-by-step manner). This is because of: 1) the large number of possible event initiators, plant

configurations, and sequences; and 2) the high degree of uncertainties in event progression and

consequences. The strategies and guidelines can take the form of high level plans that identify

and describe potential, previously evaluated, success paths for addressing specific conditions

such as loss of core cooling. As a result, strategies and guidelines provide operators and plant

staff the information and latitude to respond as necessary to unpredictable and dynamic

situations, allowing them to adapt to the actual conditions and damage states without the

60
burden of detailed procedures and the challenge of determining which procedure may be

applicable and effective under the uncertain conditions of a beyond design basis accident.

Given these differences in content and standards for usage, the intent of the proposed

rule is not to require conformance of the strategies and guidelines to the level of detail and

standards of usage for EOPs, or consolidation of the strategies, guidelines and procedures into

a single set of instructions, but rather, as previously described, to require functional integration

of strategies and guidelines with the EOPs. The objective is for the strategies, procedures, and

guidelines to retain or employ the characteristics that support their effective use under the range

of conditions to which they are each intended to apply while ensuring that the strategies and

guidelines, in conjunction with the EOPs, constitute a useable and cohesive set of instructions

for mitigating the Consequences of a wide range of initiating events and plant damage states.

To achieve this functional integration, the NRC expects that applicants and licensees will have

addressed the interfaces, dependencies, and interactions among the strategies and guidelines

that would be required under the proposed rule and the EOPs, such that they can be

implemented in concert with each other, as necessary, to effectively use available plant

resources and direct a logical and coordinated response to a wide range of accident conditions.

In keeping with the basis for a functional integrati )n of the strategies and guidelines with

EOPs, the proposed rule would require that the FSGO5sI•DMGs, a•dSA"'be integrated'"with

the Emergency Operating Procedures (EOPs)." This proposed language is intended to

communicate the NRC's expectation that the EOPs retain their role as the primary means of

directing emergency operations and that the strategies and guidelines that would be required

under the proposed rule would be integrated with EOPs to support their implementation or

augment them where their implementation is not successful in preventing significant fuel

damage.

61
The NRC considered establishing specific criteria for the integration of the strategies and

guidelines with EOPs but opted to specify• only a high level requirement to allow applicants and

licensees flexibility in the means by which they achieve the functional integration described

•previously. Common means that licensees and applicants will use to achieve functional

integration might include the following:

1. Strategies, guidelines, and procedures have clearly defined transitions (e.g., entry

and exit conditions with distinct pointers) from one strategy, guideline, or procedure to another.

2. Individuals are cued by the document or trained to know when transitions between

the strategies, guidelines, and procedures result in corresponding changes in the associated

standards for usage (e.g., when transitioning from EOPs to SAM, the operator is able to

recognize the transition from a step-by-step procedure to a flexible guideline set where it is

permissible to deviate from the order or method of accomplishing the steps).

3. Licensees establish expectations (e.g., through standards' for usage) pertaining to the

parallel use of strategies, guidelines, and procedures. Plant personnel using different

strategies, guidelines, and procedures concurrently understand which is the controlling

procedure and therefore which actions take precedence.

4. Licensees identify and resolve conflicts between the strategies, guidelines and

procedures.

5. Licensees identify competing considerations when using the strategies, guidelines

and procedures and eliminate or address them in guidance.

6. Licensees control the development and maintenance of their content and format in

accordance with human factors standards and guidelines (e.g., writer's guides) that recognize

and address the interfaces between them in order to achieve compatibility of the strategies,

guidelines, and procedures.

62
Staffing

•The NRC proposes to require licensees to provide the staffing ne~cessary for having an

integrated response capability to support implementation of the FSG,•lEDMGs, afr&AM

To be effective, staffing for an expanded response capability should include the trained and

qualified individuals who would be relied upon to analyze, recommend, authorize, and

implement the mitigating strategies. The staffing must directly support the assessment and

implementation of a range of mitigation strategies intended to maintain or restore the functions

of core cooling, containment, and spent fuel pool cooling.

The staffing analyses required by proposed appendix E, section VII should determine

when personnel performing expanded response functions should report to the site, within a

timeframe sufficient to support implementation of the strategies that are not assigned to the onshift staff. This will ensure that the functions of core cooling, containment, and spent fuel pool

cooling are continuously maintained or are promptly restored.

The NRC has endorsed the industry guidance for conducting staffing analyses,

NEI 10-05, "Assessment of On-Shift Emergency Response Organization Staffing and

Capabilities," Revision 0, and NEI 12-01, "Guideline for Assessing Beyond Design Basis

Accident Response Staffing and Communications Capabilities," Revision 0, and the NRC has

issued Interim Staff Guidance (ISG), NSIR/DPR-ISG-01, "Emergency Planning for Nuclear

Power Plants," that provides the requisite details in determining the staffing levels and for which

posiions aswellas~ ich beyond design basis external events the applicants and licensees j

should evaluate.

The recommended minimum positions and staffing levels for emergency plans were

initially provided in NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and

Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear

Power Plants." Following the September 11, 2001 events, the NRC issued Enhancements to

63
Emergency Preparedness Regulations (EP final rule) (76 FR 72560) to amend 10 CFR part 50,

appendix E, to address, in part, concerns about the assignment of tasks or responsibilities to

on-shift emergency response organization (ERO) personnel that would potentially overburden

them and prevent the timely performance of their functions under the emergency plan.

Licensees must have enough on-shift staff to perform specified tasks in Various functional areas

of emergency response 24 hours a day, 7 days a week. The proposed rule addresses the

staffing requirements for the expanded response capabilities for on-shift response an• the ERO

The proposed rule would require adequate staffing to implement the FSGsjiEDMGs,

SAM vith the EOPs without requiring further analysis to supplement analyses that were

completed as a result of post-Fukushima orders or the EP final rule. Staffing levels should be

established to ensure that if strategies are executed there would be no delays in completing

them caused by the lack of qualified personnel. The NRC expects that the use of drills, existing

training analyses and other methods would verify sufficient staffing levels.

Command and Control

The NRC proposes to require licensees to have a supporting organizational structure

with defined roles, responsibilities, and authorities for directing and performing the FG• ¢j

EDMGs,..Ri6M<he objective is to ensure that licensees address the organizational

implications of: (1) implementing the FSGs; and (2) integrating the FS s~DMGs, aR4,•r

with the EOPs such that organizational units responsible for on-site accident mitigation (e.g.,

main control room, emergency operations facility, and technical support center staff) can

support a coordinated implementation of these procedures and guidelines under the challenging

conditions presented by beyond-design-basis events.

Additional requirements currently exist in 10 CFR part 50, appendix E, section IV.A for

the inclusion within the emergency plan of a description of the organization for coping with

64
radiological emergencies, including definition of authorities, responsibilities, and duties of

individuals assigned to the licensee's emergency organization and the means for notification of

such individuals in the event of an emergency. These requirements provide the command and

control structure fore use in the execution of the emergency plan. The current 10 CFR part 50, ,

appendix E, secti' n'.A.2.a, and 5.•urther require that the emergency plan include 1) a

detailed description of the authorities, responsibilities, and duties of the individual(s) who will

take charge during an emergency; 2) plant staff emergency assignments, authorities,

responsibilities, and duties of an onsite emergency coordinator who shall be in charge of the

exchange of information with offsite authorities responsible for coordinating and implementing

offsite emergency measures; and 3) the identification, by position and function to be performed,

of other employees of the licensee with special qualifications for coping with emergency

conditions that may arise.

The need for defined command and control structures and responsibilities for use in

beyond-design-basis conditions was recognized in the course of the development of the

guidance and strategies for the current § 50.54(hh)(2). As stated in the industry's guidance

document for that set of requirements, NEI 06-12, "B.5.b Phase 2 & 3 Submittal Guideline,"

Revision 2, "Experience with large scale incidents has shown that command and control

execution can be a key factor to mitigation success." The guidance and strategies developed

for that effort include an EDMG for initial response to provide a bridge between normal

operational command and control and the command and control that is provided by the ERO in

the event that the normal command and control structure is disabled. The NRC considers that

the actions taken in the development of the EDMG for initial response for the guidance and

strategies for the current § 50.54(hh)(2) would continue to be adequate for compliance with the

proposed rule for EDMGs following the proposed movement of those requirements.

65
Th onocc •,,ev,.sion,, nul 1 .,.' y ldc,, rJz 1-b v , gil ,",bl - -(jui'

os thteexitngcomad Ad ulbo"

All previous requirements did not specify a command and control structure for a

•multi-unit event that includes the potential need for acquisition of offsite assistance to support

onsite event mitigation. Additionally, these requirements were not understood to require such a

response since they preceded the Fukushima event and the regulatory actions that stemmed

from that event. As a practical matter, the current command and control structures, including

any changes that resulted from the implementation of Order EA-12-049 requirements, are

expected to be sufficient to ensure that the functional objectives of the proposed rule are

achieved. Accordir~t~e NRC recognizes that this new requirement may not be necessary /

and is requesting stakeholder feedback on this issue (refer to section VII of this notice).

Equipment

The NRC proposes to have requirements for licensee equipment, including

instrumentation, that is relied upon for use in '" .. .. -... •'•'-

This rulemaking does not propose to modify the regulatory treatment of equipment relied upon

for the EDMGs currently required by § 50.54(hh)(2). The regulatory treatment of that equipment

will remain as it is described in the endorsed guidance document for those strategies and

eqimpmeentatin ad upainte6 popf ne SAMGs as as deci-akino rd•ameor feruhoosng

b~etwlee~enpr-lanne/d rm oi, nses rfS ig on existing saadClyi, qipet S••••percifically wit~h regrd :tog

66
paThprae propse inuale potetild sevee acientenvironments.ic Aslsuchuimement atiofOndof

rEquire9,ataments would includechs ofbetesiate asflosumto andes calcuateions muto d.aetemn

operuator caactionst asdwells dciso-akienge lmts ande actingontalevelsntionandy FPR hasin

deelpabltedt al untsc cal basise doumjentt thes TRfOrders.hc"poiesxesietchia

basi infor stinfry thisdpranch afNI1-0,andinomtonrelaed to pla ntei status asesutidncluin

tReA-u-irematacmetso 2itiatind 3,whchraeades asr feollos:'Teseg-BsiEternateie musnts, inclhaved

specifications for licensee provision of a spare capability in order to assure the reliability and

67
availability of the equipment required to provide the capacity and capability requirements of the

Order. This spare capability was also referred to within the guidance as an "N+I" capability,

where "N" is the number of power reactor units on a site. The NRC considered including

requirements similar to the spare capability specification of NEI 12-06 in-the proposed rule but

determined that such an inclusion would be too prescriptive and could result in the need to grant

exemptions for alternate approaches that provide an effective and efficient means to provide the

required capability of the Order. One example of this is in the area of flexible hos ~r which a

strict application of the sparing guidance could necessitate provision of spare hose or cable

lengths sufficient to replace the longest run of hoses when significant operating experience with

similar hoses for fire protection does not show a failure rate that would support this as a need.

•The development of the mitigating strategies in response to Order EA-12-049 relied

upon a variety of initial and boundary conditions that were provided in the regulatory guidance of

JLD-ISG-2012-01, Revision 0, and NEI 12-06, Revision 0. These initial~and boundary

conditions followed the philosophy of the basis for imposition of the requirements of Order EA12-049, which was to require additional defense-in-depth measures to provide continued

reasonable assurance of adequate protection of public health and safety. As a result, the

industry response to.Order EA-12-049 includes diverse and flexible means of accomplishing

safety functions rather than providing an additional further hardened train of safety equipment.

These requirements and conditions included the acknowledgement that, due to the fact that

initiation of an event requiring use of the strategies would include multiple failures of safetyrelated structures, systems, and components (SSCs), it is inappropriate to postulate further

failures that are not consequential to the initiating event. As a result, the NRC has determined

that the conditions to which the instrumentation relied on for the mitigating strategies would be

exposed do not include conditions stemming from fuel damage, but instead are limited as

described previously..The NRC has determined that it should not be necessary for the

68
instrumentation to be designed specifically for use in the mitigating strategies and guidelines,

but instead it would be necessary that the design and associated functional performance be

sufficient to meet the demands of those strategies.

The underlying proposed requirements are for events that are not included in the design

basis events as that term is used in the § 50.2 definition of safety-related SSCs. Because of

this, reliance on equipment for use in the related strategies would not result in the applicability of

10 CFR part 50, appendix A, General Design Criterion (GDC)-2, "Design bases for protection

against natural phenomena," or the principal design criterion (PDC) applicable to a plant's

operating license if issued prior to GDC-2. The proposed rule would require reasonable

protection for the equipment relied on for the mitigation strategies to a hazard level as severe as

that originally determined for the facility under GDC-2 or the applicable PDC unless the

reevaluated hazards stemming from the March 12, 2012, NRC letter issued under § 50.54(f), as

assessed by the NRC show that increased protection is necessary. The March 12, 2012, NRC

letter requested information on licensees' seismic and flooding hazards; licensees and the NRC

are currently scheduled to complete most of the work on the flooding reevaluations prior to the

anticipated effective date of this proposed rule. The NRC notes that there are some licensees

whose licensing bases include requirements for protection from natural phenomena beyond

those established at the original licensing (e.g., North Anna Power Station for the seismic

hazard), but anticipates that these different hazard levels will be captured in the reevaluation of

external hazards under the March 12, 2012, NRC letter.

As discussed in COMSECY-14-0037, "Integration of Mitigating Strategies for BeyondDesign-Basis External Events and The Reevaluation of Flooding Hazards," (ADAMS Accession

No. ML14309A256) and its associated SRM, the requirements of Order EA-12-049 were

imposed in parallel with the agency's March 12, 2012, requests for information on the

reevaluation of external hazards. As a result, Order EA-1 2-049 included a requirement in both

69
Attachment 2 and 3 for licensees to provide reasonable protection for equipment associated

with the required mitigating strategies from external events without specific reference to the

necessary level of protection. The appropriate level of protection from external hazards,

particularly flooding, was the subject of discussion in the course of NRC-held public meetings

leading up to the issuance of JLD-ISG-2012-01 and its endorsement of the industry guidance for

Order EA-12-049, NEI 12-06. Section 6.2.3.1 of NEI 12-06 specifies that the level of protection

for flooding should be "the flood elevation from the most recent site flood analysis. The

evaluation to determine the elevation for storage should be informed by flood analysis

applicable to the site from early site permits, combined license applications, and/or contiguous

licensed sites." The choice of this hazard level was driven by the recognition th~hile the

flooding hazard reevaluations by holders of operating licenses and construction permits may not

be complete in advance of the development and implementation of the mitigating strategies,

,tf~normation available from flood analyses for nearby sites could be taken into account in

choosing the appropriate level in order to avoid the need for rework or modification of the

strategies. Many licensees took the former approach, using their best estimates of potential

hazard levels and providing additional margin to the current licensing basis. (See, e.g., the

description of the flooding strategies for Fort Calhoun Station on page B-43 et seq., of Omaha

Public Power District's Overall Integrated Plan (Redacted) in Response to March 12, 2012,

Order EA-12-049.)

In COMSECY-14-0037, the NRC staff requested that the Commission affirm that:

1) licensees for operating nuclear power plants need to address the reevaluated flooding

hazards within their mitigating strategies for beyond-design-basis external events; 2) licensees

for operating nuclear power plants may need to address some specific flooding scenarios that

could significantly damage the power plant site by developing targeted or scenario-specific

mitigating strategies, possibly including unconventional measures, to prevent fuel damage in

70
reactor cores or spent fuel pools; and 3) the NRC staff should revise the flooding assessments

and integrate the decision-making into the development and implementation of mitigating

strategies in accordance with Order EA-12-049 and this rulemaking. These principles reflect the

NEI 12-06 reference to the "most recent flood analysis" discussed above and the documentation

by licensees in their overall integrated plans for the mitigating strategies t[ ltthe time of their

submi4A•'flood and seismic reevaluations pursuant to the § 50.54(f) letter of March 12, 2012,(

are not completed and therefore not assumed in this submittal. As the reevaluations are

completed, appropriate issues will be entered into the corrective action system and addressed

on a schedule commensurate with other licensing bases changes." In SRM-COMSECY-14-

0037, the Commission approved the first two items recommended by the NRC staff, regarding

the need for operating nuclear power plant licensees to address the reevaluated flood hazards

within the mitigating strategies and the potential for using targeted or scenario specific mitigating

strategies. The Commission did not approve the third recommendation, but that

recommendation is outside the scope of this rulemaking effort. The NRC drafted the proposed

rule to reflect this direction and in recognition of the fact that the wording of Order EA-1 2-049

and its associated guidance did not make clear that the mitigating strategies equipment would

require protection to the reevaluated hazard levels resulting from the § 50.54(f) request for

information of March 12, 2012.

Because the events for which the proposed mitigating strategies are to be used are

outside the scope of the design basis events considered in establishing the basis for the design

of the facility, equipment that is relied upon for those mitigating strategies may not fall within• the

scope of § 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear

power plants." Nevertheless, the NRC proposes that such equipment should receive adequate

maintenance in order to assure that it is capable of fulfilling its intended function when called

upon.

71
The NRC proposes to require licensees to have a means to remotely monitor wide-range

SEP level as a part of the equipment relied upon to support the FSGs. This provision would

make generically-applicable the requirements imposed by Order EA-12-051. The NRC

considered including the detailed req~uirements from Order EA-12-051 within the proposed rule,

but determined that the more performance-based approach taken with the proposed rule would

better enable an applicant for a new reactor license or design certification to provide innovative

solutions to address the need to effectively prioritize event mitigation and recovery actions

between the source term contained in the reactor vessel and that contained within the spent fuel

pool.

maintain or restore co oling, containment, and spent fuel pool cooling capabiit dur~ing an]

ELAP concurrent with either os of normal access to the ultimate heat sin r, for passive .

licensees, to develop, implement, and mainai d ceand strategies to maintainOr restore'

core cooling, containment, and spen fpool cooling cap ltes (ile., key safety functions) /

following abeyond-design-basixternal event. Based on thelC'sassessment of new

insights from the evensFukushima _Dai-ichi, it concluded th~at thes .dditiona, req~uirements

must be imposed~f licensees or constr~uction permi~t ho~lders to° Increase th..capability of ....

nuclear _por plants_ to mitigatte..beyond-design-basis external, event, and tat seaditional

requiments are needed to provide adeuate, protection to public, healt and...... safety. T
NEI 12-0 has focused on operating reactors and reflects that operating reactors 'e

constrained b existing SS~s and plant layouts. However, without such constrnts, new

reactor applicants aqve an opportunity to incorporate enhanced capability f mitigating

strategies into the des eA

New reactor applica s an address major elements of Orde A1 2-049 by

incorporating into the plant desi those design features for miti ating strategies that provide

enhanced capability to ensure that k sa3fety functions are ainained or restored. Such

design features should reduce reliance o and simplify tmanual actions necessary to

maintain or restore key safety functions, and 1wmr time to assess plant conditions and

prolong the use of installed equipment, as comp• d to currently operating reactors. This

approach is consistent with the Policy State •nt on Regulation of Advanced Reactors

(73 FR 60612; October 14, 2008), in w the Commissi ncouraged vendors to include

certain features into the plant desig ,and the most effective o portunity to do so is during the

design of those SS~s. \

During the design of e SS~s that perform key safety functio , n applicant would\

have an opportunity to*fcorporate additional capability to mitigate beyon design-basis external/

events. That is, rat er than relying heavily on portable (flexible) equipment to itigate these

events as with ~rrently operating reactors, new reactor applicants would be abl toreduce or

delay relia, e on portable equipment and offsite resources through engineered desi n features.

New rector applicants would also have greater flexibility to include these design featu s at the

tdes, n• stage without the constraints on operating7 reactors. As a result, design features fo new
inctorpoaemh designs ol cnapabilitiesation assuraegoes byclsing ocntaihenaptegrity, riandit,/

Thex Cmi of sslity also has uo dreted o the iiatinew sreactorsinopres ceanaft

enhaneentsgthator aprovdditach safestybmarging atendtv defesegn -depthirecogizngs fo :the

opportunitsies cnitha ar t wvilablhe rNgR te'ds p ig stp a gtie. For example, in0.50 S ArMScralm-ac 6,

"EvolutionaryqLight Waer reactor (plR)Criicantsion passu soprr andaTheirsRelainshi tCrrn t

icreuapo eura ements,"h desgaabltied J 2 oasun e c190rte comisolng aprcovtnednthiegNrCity affd

position msi~lohsdrce that adrs new reactors 06 huhntaioncorof~ analernteaciouceo

apoednhaneenRstafs thatewrats psto provideaionlsftymrn hydfeseica--separtio frecfgirigte

oprortecitionsequipment toaensure h saf te sudowin ctange achi edassmping theMlSsof9a-fire

ara ha ate eypnfiearirtome 50.48 "Fi eiProtelation shptoCrrn

nucoletoary po r htWaer eactor ppicnt i Clu erinthefaiictyeign thIcpbiisomtiaesh

Rgltry equiremeitiga," duringn perid/wenmcndiions the 190 createdeb exeNal the evetaoudfhnde

orposeven tacesto pewroeactedon-sitddes potal equthstaipmnto and asscitednt acsonnection

Appoinse woul bexrequied topingcorpoateldes.iignfatures tomanain oRM retore keymisasity
T urkey oint nuclear plant site during hurricane Andrew in 1992, and during hurrica/ Katrina in

2005."Aitholh the effects can last for several days, they should not preclude a ess//to staged

equipment. The esign features should result in a plant design that is better quipped for an

ELPconditionadt us haealonger period of time until the facility wo l~dnedtrlyo

portable equipment and o site resources to maintain or restore key fety functions./

The benefit of being a le to cope with an ELPwt eu d rlacnoeao

actions was recognized by the NFReport based on the re e•w of insights from the

Fukushima Dai-ichi accident. Greate line on design etures that include well thought-out

human-machine interfaces would reduce r ance on d simplify the manual actions necessary

to maintain or restore key safety functions. Fu , reducing reliance on human actions would

also reduce the potential for huma'n failures d, ring •essful, adverse conditions.

While the requirement to incorpor design fea res would result inanipoe

capability to mitigate an ELAP, which •ould be caused by aeyond-design-basis external event,

the NRC does not intend for a lic see to rely solely on these edign features. In light of the

unknwn evertyduration, oher aspects of beyond-design-basi events that could result in

an ELAP, it is important f/ea licensee to have diverse, flexible strategi o addressing such

events that include d ign features as one aspect of those strategies. In o •er words, including

these design fea/t res would not preclude the need for portable equipment an offsite resources

in a licensee' Smitigation strategies. Flexible strategies would also allow for conti encies if the

designfeures are not available or otherwise not performing as expected. Therefor while the

Sdesiger of key safety systems would be required to incorporate design features to bette%

e ble the li~censee to mitigate these events, a licensee would need to ensure its mitigation /

strteges lsoinclude diversity and flexibility through portable equipment and offsite resources. •

'......_ .75,
sprt key safety functions are being designed (such as during the development of dsg

certifict application) and not when including such design features could impo( a significant

burden on lic sees or applicants and potentially decrease standardization chas a

comind lcese •aplication referencing a certified design). The NRC intproposing to

require licensed nude- ,power plants, certified designs, or license a ictosreferencing an,

approved design (e.g., a s~ign certification) to incorporate suc/ dein features because the

potentially significant modificatnstto the• design have not en justified under the NRC's backfit

rule at 10 CFR 50.109 or the issue alit provisionso1 F part 52.

As a result, the NRC is proposin tat ne nuclear power reactor designs should be

capable of maintaining or restoring key safe unLctions of core cooling, containment, and spent

•fuel pool cooling capabilities using desi fetue(i.e., installed SSCs) for near-term coping for

ras long as practical following initiat nof an ELAP eve . The extended coping duration would

provide a greater amount oftie to plan and deploy portab on-site equipment for long-term

coping.

An NRC sta nember did not agree with proposing requirem t or new reactor

designs to inco orate design features as part of this rule and submitted a n-concurrence. In

accordancwith the NRC's non-concurrence process, NRC management and st consdered,'

/ re udn"NncnurneN ponded to, the staff member's -25-0.and concerns, and documnentation of the non-concurrei can/

Training

The NRC anticipates that mitigation of the effects of beyond-design-basis events using

the proposed strategies and guidelines would be principally accomplished through manual

76
actions rather than automated plant responses for currently operating reactors. No o•tr "

wol ols ~au ui iiiriual •iiuns due to IncuIpu•-,tuII Gf d•;§nfati :' 111plifig th )I

Additionally, the NRC notes that it anticipates that instructions provided for event mitigation may

be largely provided as high level strategies and guidelines rather than step-by-step procedures.

The use of strategies and guidelines supports the ability to adapt the mitigation measures to the

specific plant damage and operational conditions presented by the event. However, effective

use of this flexibility will depend upon the knowledge and abilities of personnel to select _

appropriate strategies or guidelines from a range of options and implement mitigation measures

using equipment or methods that may differ from those employed for normal operation or

design-basis event response. As a result, the NRC considers personnel training and

qualification necessary to ensure that individuals will be capable of effectively performing their

roles and responsibilities in accordance with the strategies and guidelines that would be

required by this proposed rule.

The NRC acknowledges that licensee training programs, such as those required for

licensed operators under 10 CFR part 55, "Operators' Licenses," the programs for plant

personnel specified under § 50.120, "Training and Qualification of Nuclear Power Plant

Personnel," and the training for emergency response personnel required by 10 CFR part 50,

appendix E, section IV.F, "Training,' would likely provide for many of the knowledge and abilities

required for performing activities in accordance with the strategies and guidelines that would be

required by this proposed rule. Nevertheless, as noted previously, the NRC anticipates that

these strategies and guidelines may use new methods or equipment that require knowledge and

abilities not currently addressed under existing training programs a~ a res re may be )

gaps in these training programs that must be addressed to support effective use of the

strategies and guidelines. Accordingly, this proposed rule would further require that licensees

77
provide for the training of personnel using a systems approach to training as defined in § 55.4

(the Systems Approach to Training (SAT) process), except for elements already covered under

other NRC regulations 6. The SAT process, which is acceptable for meeting training

requirements under 10 CFR part 55 and § 50.120, would also be appropriate for licensee

identification and resolution of any current gaps or future modifications to personnel training that

may be necessary to provide for the training of personnel performing activities in accordance

with the mitigating strategies and guidelines that would be required by this proposed rule. The

NRC recognizes that there are other training programs that are currently acceptable for meeting

other regulatory required training (e.g., 10 CFR part 50, appendix E, section IV.F) that do not

use the SAT process. In light of the existence of these training programs, which have been

found acceptable for more frequently occurring design-basis events, the NRC has determined

that these training programs can meet the needs for common elements with beyond-designbasis event mitigation. Therefore, the NRC would not require licensees to revise these training

programs to use the SAT process to meet the proposed requirements. Licensees would be

required to use the SAT process for newly identified training requirements supporting the

effective use of the strategies and guidelines that would be required by this proposed rule.

By using the SAT process, licensees would identify and train on any additional tasks that

would be necessary to implement the strategies and guidelines for the mitigation of beyonddesign-basis events as defined in this proposed rule. The additional tasks identified would be

incorporated into the training program to ensure appropriate training would be administered for

each qualified individual designated to implement the strategies and guidelines required by this

proposed rule.

6 This definition of a systems approach to training (SAT), is a training program that includes the following five

elements: 1) systematic analysis of the jobs to be performed; 2) learning objectives derived from the analysis which

describe desired performance after training; 3) training design and implementation based on the learning objectives;

4) evaluation of trainee mastery of the objectives during training; and 5) evaluation and revision of the training based

on the performance of trained personnel in the job setting.

78
The new training needs that result from identification of anY new tasks should be

evaluated by licensees to determine whether each of their plant referenced simulators can be

used to allow personnel to simulate their actions and resulting plant response. The NRC would

not require licensees to update their plant referenced simulators to model conditions that may

exist during significant damage to fuel in the reactor vessel or spent fuel pool. +.ee''r,4...

eng i unoci referenced simultorJ ~rcuap~ is o table moe f U th aimu,•to c1~iiywudL

• ti ,-fo ieans to suppert •he llmn•& iItranlng afesignea to Impl•iii ..... miian ...

ct~c~ana guiaellnes requireo Dy mhis proposeD ruti_.--.

the written and Opera*tg xams, for operators and senior operators wou include SAMGs as aI

required topic for training. ingseveral public meetings the nuer power industry strongly

displacement of higher priority trainin pics, such as EOPs. Ad oaly, the industry stated /

that revising §§ 55.41, 55.43, an5545 to include SAMGs as required tr in°fr newly__

licensed operators would gatively impact the overall length of the operator lice *nhtaining\

program, which is c ~ently an average of 22 months in length. The NRC is proposing t equire

training based. the SA.process instead of. r~evising.§.§ 55.41, 55.43, and 55.45 because th%

SAT pro ss is we~ll-est~abli~sh~ed and required fo~r initi~al licensing of operators under 10 CF-R part

•55a to train and qualify individuals identified in § 50.120.
adiinlkno wldg and abilities retlostat aeued to sever cideithat o ol d bpera serlctedsn frtexan

wnteiiiloertrlcni'ritten examinationsan potios optheoating test.hee.e wBedge and aiitiesare

rlatedsn wrto1)then transintions aro eEO t as o e n itheS Ms rando2mte iescuthorityrduring

ipeettoofte s.Teenwknowledge and abilities e • araosfreahvno~yto e NC poonsistent d the

nelydeveionad knowledge and abilities ceatal ogsvreai isi that lcudb for slce theAdacdolnte

Rer plant design.

Change Control

The proposed requirements address beyond-design-basis events, and as such, currently

existing change control processes do not address all aspects of a contemplated change,

including most notably § 50.59. As such, the proposed change control provision is intended to

supplement the existing change control processes and focus on the beyond-design-basis

aspects of the proposed change.

This proposed rule would not contain criteria typically included in other change control

processes that are used as a threshold for determining when a licensee needs to seek NRC

review and approval prior to implementing the proposed change. Instead, the proposed

provisions would require that the evaluations of the proposed change reach a conclusion that all

new requirements continue to be met and that this evaluation is documented and maintained to

support NRC inspection.

80
Proposed changes that remain consistent with regulatory guidance would be

acceptable, since such changes would ensure continued compliance with the proposed

provisions in this rulemaking. The NRC recognizes that the proposed change control provisions

may result in licensees seeking NRC review and approval of proposed changes that do not

follow current regulatory guidance for this proposed rulemaking potentially through a license

amendment or through NRC review of new or revised regulatory guidance. Accordingly, the

NRC is requesting stakeholder feedback on this issue to determine whether there is a better

regulatory approach for change control (refer to the "Specific Requests for Comments" section

of this document).

During public discussions before issuance of this proposed rule, there was a suggestion

that the NRC should consider a provision to allow a licensee to request NRC review of a

proposed change, and that if the NRC did not act upon the request for a suggested time period

(e.g., 180 days) that the request be considered "acceptable." The NRC did not include this

"negative consent" type of approval process in the proposed rule and instead the proposed

change control process places the responsibility on the licensees to ensure that proposed

changes result in continued compliance with the proposed rule provisions, or are otherwise

submitted to the NRC following the § 50.12 exemption process. The NRC expects to obtain

stakeholder feedback on this issue and will consider that feedback when developing the final

rule provisions.

A licensee may intend to change its facility, procedures, or guideline sets to revise some

aspect of beyond design basis mitigation (i.e., governed by the proposed provisions of this

rulemaking), and the same change can impact multiple aspects of the facility (i.e., impact

"design basis" aspects of the facility and be subject to other regulations and change control

processes). As previously discussed, the NRC anticipates that a licensee would ensure that a

proposed changed is consistent with endorsed guidance to ensure continued compliance with

81
the proposed provisions. This same change could also impact safety-related structures,

systems, and components, either directly (e.g., a proposed change that impacts a physical

connection of mitigation strategies equipment to a safety-related component or system) or

indirectly (e.g., a proposed change that involves the physical location of mitigation equipment in

the vicinity of safety-related equipment that presents a potential for adverse physical/spatial

interactions with safety-related components). As such, § 50.59 would need to be applied to

evaluate the proposed change for any potential impacts to safety-related SSCs.

Additionally, proposed changes can impact numerous aspects of the facility beyond the

safety-related impacts, including implementation of fire protection requirements, security

requirements, emergency preparedness requirements, or safety/security interface requirements.

Accordingly, it would be necessary for a licensee to ensure that all applicable change control

provisions are used to judge the acceptability of facility changes including, for example, change

control requirements for fire protection, security, and emergency preparedness. Additionally,

recognizing the nature of mitigation strategies and the reliance on human actions, it is also

necessary to ensure that the proposed changes satisfy the safety/security interface

requirements of § 73.58. It is the obligation of the licensee to comply with all applicable

requirements, and as such, the proposed change control provisions could be viewed as

unnecessary. However recognizing the potential complexity of proposed facility changes and

the complexity of existing regulato~ry requirements that govern change control, the NRC

concluded that adding the proposed change control provision, for the purposes of regulatory

clarity, was warranted.

A •znt, 1 e~pe oT change control proces ulithnt~ -- • i• 1.lib'•"

--r1 fJuuds a process for cnlanges [u d,,d d~pdI,,I [iui-ii ', u,,, i ,• •tr;lS&g..o t;' ', 11 1 If c n,,

'-ert~ iu,,, inlclualng a similar change control prou•, vwcro to be =_,_,bject t I •piuoe

82
chagecn'feturoi proequTheent NCpansoad lcnsew a reeechange controlrtisionteftr designprosdc

cetifictiogn reaules subjet toe proposed design features requirment, (incluingaendments~d

holdersth refere•ncingsct~nVII thetdesign certificationtqus adprture fromthe desg fetrepse in

thane reeeceddeintcertificeNCpa nstion e dd new change ctolpro ision w oul re uiure thatif the"n 

appliicant or rlenseeschanc th s oe d deinfrain reqtuired b § e7amen)to beinclud eding thendet

finao aft anay reotofAR the o design standardting cer)tific atbioe lthenste applicants ord

olcense allensurenthat te changed featreifconiusto seartisefyo the design bfetuves i

cieinthe r pef oposed design at feric urZ-esnwcag • requirement.ifth rvso o 

Imp/ementation

The NRC proposes a compliance schedule of four years following the effective date of

the rule. This proposed rule does not include any special provision for a holder of a COL as of

the effective date of the rule for which the Commission has not made the finding required under

§ 52.103(g) (i.e., a COL holder still in the construction phase). The NRC considers the duration

of four years prior to compliance with the requirements of the proposed rule to be acceptable

because the majority of these requirements have been previously implemented under Orders

EA-12-049 and Order EA-12-051 or § 50.54(hh)(2), or are in response to the § 50.54(f) requests

for information issued March 12, 2012. •SAMG re uirements, substantial

reslofterct ntnf _.~t •nte induIqtry to i'dt thl e t_ genqeric SAMG g"id3ncc.

83
Regulatory Basis for New Emergency Response Capability Requirements

A significant objective of this rulemaking is to make the requirements that were

previously imposed under Order EA-1 2-049 generically applicable. As an implicit part of the

implementation of Order EA-12-049, additional emergency response capabilities were included

to address a beyond-design-basis external event that impacts multiple power reactor units, and

potentially multiple source terms, on the site. In all caT_•tese additional proposed revisions

are considered to be necessary to effectively mitigate such an event, consistent with the NRC's

intent in issua•"eeOrder EA-12-049. These proposed requirements were not explicitly

addressed in the previous regulatory bases documents issued for the two rulemakings. This

section discusses the basis for these proposed emergency response capability provisions.

The March 12, 2012, § 50.54(f) letters (i.e., Request for Information Pursuant to Title 10

of the Code of Federal Regulations 50.54(f)) requested information from the licensees that, in

part, was intended to verify the adequacy of emergency planning to address what was then

termed prolonged SBO7 and multi-unit events. The accident at Fukushima highlighted the need

to determine and implement the required staff to fill all necessary positions responding to multiunit events. Additionally, NRC recognizes that the communication equipment relied upon to

coordinate the event response during an ELAP should be powered and maintained.

1. Onsite and offsite communications capability

This proposed rule would require additional communications capabilities for events that

result in extended loss of ac power onsite, or potential destruction of offsite communications

infrastructure. Because of the destruction to communications capability that occurred at

Fukushima, the NRC would propose requirements for licensees to provide a greater capability

? While the letter .made use of the term "prolonged SBO," the request for information was for a loss of all alternating current

power, which was subsequently termed an ELAP. The phrase "prolonged SBO" is retained here to accurately reflect the

wording used in the letter.

84
to communicate with onsite staff to support mitigation of the event, and to support offsite

communications to gain any additional support or to perform emergency preparedness

functions. The proposed requirements would support effective implementation of the FSGs and

were included as part of the implementation of Order EA-1 2-049.

2. Staffing assessment i

This proposed rule would require an assessment that is considered essential j6'"effective

implementation of the FSGs. This assessment matches the one that was conducted under the

March 12, 2012, request for information that was developed to align with the requirements

included in Order EA-12-049 (i.e., the staffing analysis specifically considered the staffing needs

for implementing Order EA-12-049); licensees would not be required to repeat the staffing

analysis. A lesson-learned from the Fukushima event is that there are increased staffing

demands following a beyond-design-basis external event, and this coupled with the subsequent

NRC requirements issued in Order EA-12-049 required the staffing analysis to provide a level of

assurance that the FSGs can be implemented. This provision would then sLport the proposed J

requirements of the rule to have sufficient staffing to implement the FSGs!4EDMGs, •ud-'AMI-- J

in conjunction with the EOPs.

3. Change control

The NRC would not require a power reactor applicant or licensee to address or

implement the proposed communications and staffing analysis requirements through the

licensee's or applicant's emergency plan or maintain the capabilities as a part of the emergency

preparedness program. This approach would allow for site-specific flexibility in implementation.

Therefore, the requirements of maintaining the communications and staffing analysis in an

effective emergency plan and controlling changes to it under § 50.54(q) would not apply when

implementation of the requirements is not in the emergency plan, but in all cases, the change

control process of the proposed rule would apply. However, if an applicant or a licensee

85
incorporates the communications and staffing analysis into the emergency preparedness

program through the emergency plan or emergency plan implementing procedures, the

requirements of § 50.54(q) would apply.

4. Multi-source term assessment capability

This proposed rule would require licensees to have a means for determining the

magnitude of, and for continually assessing the impact of, the release of radioactive materials,.

including from all reactor core and spent fuel pool sources. A lesson learned from the

Fukushima Dai-ichi event is that there is a potential for a beyond-design-basis external event to

result in multiple source terms from multiple release points, and under such a situation,

additional capabilities are necessary to support development of appropriate protective action

recommendations. In COMSECY-1 3-0010, "Schedule and Plans for Tier 2 Order on

Emergency Preparedness for Japan Lessons Learned," dated March 27, 2013, the NRC staff

informed the Commission that licensees would provide information about their current multiple

source assessment capability, or a schedule for implementing such a capability, and that

associated implementation would occur by the end of calendar year 2014. Licensee

implementation of the multiple source assessment capability will be verified by inspection under

TI-251 5/191, "Inspection of the Licensee's Responses to Mitigation Strategies Order EA-1 2-049,

Spent Fuel Pool Instrumentation Order EA-12-051 and Emergency Preparedness Information

Requested in NRC March 12, 2012." The NRC has been working with the industry and

stakeholders through public meetings to review and provide feedback on NEI 13-06,

"Enhancements to Emergency Response Capabilities for Beyond Design Basis Accidents and

Events," Revision 0, which, in part, would provide licensees with guidance on implementing a

multiple source dose assessment capability.

The capability should be available to support responses during events both within and

beyond the plant design basis. Also, the licensee should discuss the site's multi-unit and

86
multiple source dose assessment capability with the offsite response organizations, particularly,

with the agencies that are responsible for making decisions on public protective action

recommendations. Agreement on the methods and results will avoid unnecessary delays during

the event in making the public protective action decisions, public notification, and the

implementation of protective actions.

5. Technology-neutral Emergency Response Data System

The proposed requirements of 10 CFR part 50, appendix E, section VI, for the

Emergency Response Data System (EROS) would reflect the use of up-to-date technologies

and remain technology-neutral so that the equipment supplied by NRC would continue to be

replaced as needed, without the need for future rulemaking because equipment becomes

obsolete. In 2005, the NRC initiated a comprehensive, multi-year effort to modernize all aspects

of the ERDS, including the hardware and software that constitute the ERDS infrastructure at

NRC headquarters, as well as the technology used to transmit data from licensed power reactor

facilities. As described in NRC Regulatory Issue Summary 2009-1 3, "Emergency Response

Data System Upgrade From Modem to Virtual Private Network Appliance," the NRC engaged

licensees in a program that replaced the existing modems used to transmit EROS data with

Virtual Private Network (VPN) devices. The licensees now have less burdensome testing

requirements, faster data transmission rates, and increased system security.

V. Section-by-Section Analysis.

Proposed § 50. 8 Information Collection Requirements: 0MB Approval

* This section, which lists all information collections in 10 CFR part 50 that have been

approved by the Office of Management and Budget (0MB), is revised by adding a reference to

§ 50.155, the mitigation of beyond-design-basis events rule. As discussed in the "Paperwork

Reduction Act Statement" section of this document, the 0MB has approved the information

87
collection and reporting requirements in the final mitigation of beyond-design-basis events rule.

No specific requirement or prohibition is imposed on applicants or licensees in this section.

Proposed § 50. 34 Contents of Applications; Technical In formation

Section 50.34 identifies the technical information that must be proVided in applications

for construction permits and operating licenses. Paragraphs (a) and (b) of this section identify

the information to be submitted as part of the preliminary or final safety analysis report,

respectively. New paragraph (i) of this section would identify information to be submitted as part

of an operating license application, but not necessarily included in the final safety analysis

report.

The NRC is proposing an administrative change to § 50.34(a)(13) and (b)(12) to remove

the word "stationary" from the requirement for power reactor applicants who apply for a

construction permit or operating license, respectively. Section 50.34(a)(13) and 50.34(b)(12)

were added to the regulations in 2009 to reflect the requirements of § 50.150(b) regarding the

inclusion of information within the preliminary or final safety analysis reports for applicants

subject to § 50.150. Section 50.34(a)(13) and (b)(12) were inadvertently limited to "stationary

power reactors," matching the wording of § 50.34(a)(1), (a)(1 2), (b)(1 0), and (b)(1 1), which

pertain to seismic risk hazards for stationary power reactors. The NRC does not intend to

change the meaning of this requirement by removing the word "stationary" from these

requirements. This change is intended to ensure consistency in describing the types of

applications to which the requirements apply.

Prpsd§ 53.

3 4 (d)( 4) would require applicants for a cons~rucuiu, J,,it t u •ubmit'a.

prlm dl y aescription of the design features Included in tre plali,'dc~ig3n in thz ,•-Tcfllf1

88
50.-4),3 ... .- ,o.... rzquir ppic ....... nt. s anper2ting liconct =cptfhub•=fne 

Proposed § 50.34(i) would require each application for an operating license to include

the applicant's plans for implementing the requirements of proposed § 50.155 and

10 CER part 50, appendix E, section VII, including-a schedule for achieving full compliance with

these requirements. This paragraph would also require the application to include a description

of: 1) the integrated response capability that would be required by proposed § 50.155(b); 2) the

equipment upon which the strategies and guidelines that would be required by proposed

§ 50.155(b)(I) rely, including the planned locations of the equipment and how the equipment

land SSCs would meet the design requirements of proposed § 50.155(c); and 3) the strategies

and guidelines that would be required by proposed § 50.155(b)(2).

Proposed § 50.54 Conditions of Licenses

Applicability of the requirements of § 50.54(hh) is currently governed by § 50.54(hh)(3),

which makes these requirements inapplicable to a nuclear power plant for which the

certifications required under § 50.82(a) or § 52.1 10(a)(1) have been submitted. This rulemaking

proposes to renumber § 50.54(hh)(3) to reflect the proposed movement of the requirements

currently within § 50.54(hh)(2) to proposed § 50.155(b)(2). The proposed § 50.54(hh)(2)

includes editorial changes to reflect that the applicability is to the licensee rather than the facility

and to correct the section numbers for the required certifications. Additionally, proposed

§ 50.54(hh)(2) clarifies that the inapplicability is dependent upon the NRC docketing of the

certifications rather than licensee submittal because § 50.82(a)(2) and § 52.110(b) set the

89
docketing of the certifications as the point at which operation of the reactor is no longer

authorized, and fuel cannot be placed in the reactor vessel.

Proposed §50.155(a), "Applicability" d Q e.wti*

Proposed § 50.155(a) would describe which entities would be subject to this proposed

rule. Proposed § 50.155(a)(1) would provide that each holder of an operating license for a

nuclear power reactor under part 50 and each holder of a combined license under part 52 after

the Commission has made the finding under § 52.103(g) that the acceptance criteria have been

met, would be required to comply with the requirements of this proposed rule until the time when

the NRC has docketed the certifications described in § 50.82(a)(1) or § 52.110(a). These

certifications inform the NRC that the licensee has permanently ceased to operate the reactor

and permanently removed all fuel from the reactor vessel. Upon the docketing of the

certifications, by operation of law under § 50.82(a)(2) or § 52.110(b), the licensee's part 50 or 52

license, respectively, no longer authorizes operation of the reactor or emplacement or retention

of fuel in the reactor vessel. At this point, many portions of this proposed rule would not apply to

the licensee because the removal of fuel from the reactor vessel would eliminate the risk of a

reactor-based beyond-design-basis event and the need to prepare to mitigate those events.

Proposed § 50.155(a)(3) would set forth the requirements that would apply to this licensee.

Proposed § 50.1 55(a)(2) would provide that each applicant for an operating license for a

nuclear power reactor under part 50 and each holder of a combined license before the

Commission makes the finding under § 52.103(g) would be required to comply with the

requirements of this proposed rule no later than the date on which the Commission issues the

operating license under § 50.57 or makes the finding under § 52.103(g), respectively. Under

this regulation, oPerating license applicants and CCL holders' would be in compliance with this

90
proposed rule before they begin operating their reactors, thereby providing additional defensein-depth capabilities at the inception of power operations.

Proposed § 50.1 55(a)(3) would address power reactor licensees that permanently stop

operating and defuel their reactors and begin decommissioning the reactors. The proposed

paragraph would provide that when an entity subject to the requirements of proposed § 50.155

submits to the NRC the certifications described in section § 50.82(a)(1) or § 52.110(a), and the

NRC dockets those certifications, then that licensee would be required to comply with the

requirements of proposed § 50.1 55(b)-.,)associated with maintaining or restoring secondary

containment, if applicable, and spent fuel pool cooling capabilities for the reactor described in

the § 50.82(a)(1) or § 52.110(a) certifications, except for the requirements in proposed

§ 50.155(c)(4) and proposed in 10 CFR part 50, appendix E, section VII. In other words, the

licensee could discontinue compliance with the requirements in proposed § 50.155 associated

with maintaining or restoring core cooling or the primary reactor containment functional

capability for the reactor described in the § 50.82(a)(1) or § 52.110(a) Certifications./

Compliance with the requirements of proposed § 50.155(b)-(2 associated with maintaining or (

restoring secondary containment, if applicable, and spent fuel pool cooling capabilities would

continue as long as spent fuel remains in the spent fuel pool(s) associated with the reactor

described in the § 50.82(a)(1) or § 52.110(a) certifications.

Proposed § 50.1 55(a)(3)(i) would discontinue the requirement to comply with proposed

§ 50.155(b)(1) requirements concerning beyond-design-basis event strategies and guidelines

for spent fuel pool cooling capabilities, and any requirements based on compliance with

proposed § 50.155(b)(1), for certain licensees in decommissioning. These licensees would

have to perform and retain an analysis demonstrating that sufficient time has passed since the

fuel within the spent fuel pool was last irradiated such that the fuel's low decay heat and boil-off

period provide sufficient time in an emergency for the licensee to obtain off-site resources to

91
sustain the spent fuel pool cooling function indefinitely and therefore obviate the need to comply

with proposed § 50.1 55(b)(1) using installed or on-site portable equipment.

Proposed § 50.155(a)(3)(i) also would discontinue the requirement to comply with the

remaining provisions of proposed § 50.155 excePt proposed § 50.155(b)(2) when the fuel in the

spent fuel pool reaches the point where beyond-design-basis event strategies and guidelines for

spent fuel cooling capabilities would no longer be needed.

Proposed § 50.155(a)(3)(ii) would exempt the licensee for Millstone Power Station

Unit 1, Dominion Nuclear Connecticut, Inc. from the requirements of proposed § 50.155.

Under proposed § 50.155(a)(3), once a power reactor licensee has permanently stopped

operating and defueled its reactor and has removed all irradiated fuel from the spent fuel pool(s)

associated with the reactor described in the § 50.82(a)(1) or § 52.110(a) certifications, the

licensee could cease compliance with all requirements in proposed § 50.155 for the unit(s)

described in the § 50.82(a)(1) or § 52.110(a) certifications.

Proposed paragraph 50.155(a)(4) would specify the persons to whom the requiee

o 0.155(d) apply. Section 50.155(d) would apply to applicants for construction pe rmits,

operating enses, standard design certifications, standard design approvals mbined

licenses, andcma acturing licenses. The NRC chose the term " cants" so that this

proposed rule would oly ly to a person who hasdsub;ed a nuclear power reactor design •

for NRC approval, but a o yet h•eived NRC proval as of the effective date of this final

rule. As a result, this proposed rule w n ply to any licensee, permit holder or applicant

92
applicati ns requesting approval of the design of the key safety functions. In doi g so, the /

proposed § 0.155(d) would not apply to an application that references a pre/ ously approved

nuclear power p'lnt design. For example, under § 50.155(a)(4)(v), a combied license

application referenci an existing design certification would not be reuired to comply. As a

result, the South Texas R~oject _Units 3 and 4 combined license ap ~lication would not be

required to comply with § 5 .15(d) because it references an ix•sting design certification.

Further, § 50.155(d) wod, apply to applications, for ew standard design certifications.

However, § 50.155(d) would not ap to applicationso r renewals of standard design certifications. For example, the NRC a0okee w applications for renewal of the

Advnce Bolin Waer eacordesign 'e d in appendix A to 10 CER part 52. These

applicants would not be required to comp with 50.155(d).

The NRC also clarifies that, il/e § 50.155( would only apply to certain applicants as

desried n 5015(a)4) fu eoperating and comih d licensees would also be required to

address the remaining req e•Jments under § 50.155 at the- _•propriate time as described in

§ 50.155(a)(2). Furthe , applicants for an operating license or c •bined license would also beb

subject to the info fation collection requirements in the respective c t~ent of applications

sections ame ded in this proposed rule. Therefore, a combined license licant not

referenci ga standard design who is subsequently granted a license, and an erating" license

appl ant, would be subject to all of the proposed requirements under § 50.155, whe as other

/, plicants or licensees would only be subject to certain proposed requirements. •

93
Proposed § 50.155(b), "Integrated response capability"

Proposed paragraph (b) would require that each applicant or licensee develop,

implement, and maintain an integrated response capability that includes: 1) mitigation

strategies for beyond-design-basis external events, 2)•vreaccident r•• gudz',",ic:;

"•extensive damage mitigation guidelines,,~ integration of these strategies and guidelines with

emergency operating procedures,)5 sufficient staffing to support implementation of the

guidelines in conjunction with the EOPs, andy a supporting organizational structure with

defined roles, responsibilities, and authorities for directing and performing these strategies,

guidelines, and procedures. The intent is to require that the operating and combined license

holders described in § 50.155(a) be able to mitigate the consequences of a wide range of

initiating events and plant damage states that can challenge public health and safety.

The specification of strategies, guidelines and procedures for the response capability not

only defines the required scope of the capability but sets forth the expectation that the response

capability must include planned methods for responding that are documented in some form of

written instruction. To serve their function, these strategies, guidelines and procedures must be

acted upon by individuals capable of understanding their appropriate application and

implementing them. Accordingly, proposed § 50.155(b)(,•, in conjunction with proposed [

§ 50.1 55(), would require that the response capability include an adequate number of

personnel with the knowledge and skills to implement the Strategies, guidelines and procedures (

and that the mitigation activities of these individuals be coordinated in accordance with a

defined command and control structure as would be required by proposed § 50.155(b)(~

Proposed § 50.155(b) would specify that the integrated response capability be

"developed, implemented, and maintained." This language reflects NRC consideration that

whereas certain elements of the integrated response capability have been developed and are

currently in place (e.g., the EOPs), other elements (e.g., guidelines to mitigate

94
beyond-design-basis external events) may require additional efforts to complete and integrate.

The term "implement" is used in proposed § 50.155(b) to mean that the integrated response

capability is established and available to respond, if needed (e.g., the licensee has approved the

strategies, guidelines, and procedures for use). The term "maintain" as used in proposed

§ 50.155(b) reflects the NRC's intent that licensees ensure that the integrated response

capability, once established, be preserved consistent with the change control provisions of

proposed § 50.155(g).

Proposed § 50.155(b)(1) would establish requirements for applicants and licensees to

develop, implement and maintain strategies and guidelines to mitigate beyond-design-basis

external events from natural phenomenon that result in an extended loss of ac power concurrent

with either a loss of normal access to the ultimate heat sink or, for passive reactor designs, a

loss of normal access to the normal heat sink. These provisions would require that the

strategies and guidelines be capable of being implemented site-wide and include:

.. Maintaining or restoring core cooling, containment, and spent fuel pool cooling

capabilities; and

ii. Enabling the use and receipt of offsite assistance and resources to support the

continued maintenance of the functional capabilities for core cooling, containment, and spent

fuel pool cooling indefinitely, or until sufficient site functional capabilities can be maintained

without the need for the mitigation strategies.
instry guida e, for use bypplicants an ensees, t t incorporate, esson earne,•d nd

feedbackemminge s~ofrothe implemetion of Ordi~ EA-12-049, csistent vith Co mission

The proposed § 50,155(b)(I) would limit the requirements for mitigation strategies to

addressing "external events from natural phenomeren." This proposed language is meant to(

differentiate these requirements from those that currently exist within § 50.54(hh)(2), which

address beyond-design-basis external events leading to loss of large areas of the plant due to

explosions and fire. This proposed provision also results in the need to have mitigation

equipment be reasonably protected from the effects of external natural phenomenon as

discussed in later portions of this proposed notice.

The proposed requirements to enable "the use and receipt of offsite assistance and

resources to support the continued maintenance of the functional capabilities for core cooling,

containment, and spent fuel pool cooling indefinitely, or until sufficient site functional capabilities

can be maintained without the need for the mitigation strategies" means that licensees would

need to plan for obtaining sufficient resources (e.g., fuel for generators and .pumps, cooling and

makeup water) to continue removing decay heat from the irradiated fuel in the reactor vessel

and spent fuel pool as well as to remove heat from containment as necessary untiy~ilternate

means of removing heat is established. The alternate means of removing heat could be

achieved through repairs to existing SSCs, commissioning of new SSCs, or reduction of decay

heat levels through the passage of time sufficient to allow heat removal through losses to the

ambient environment. More detailed planning for offsite assistance and resources would be

necessary for the initial period following the event; less detailed planning would be necessary as

the event progresses and the licensee can mobilize additional support for recovery.

96
Proposed § 50.155(b)(2) would move requirements for EDMGs that currently exist in

§ 50.54(hh)(2) to proposed § 50.155(b)(2). This move would consolidate the requirements for

beyond-design-basis strategies and guidance into a single section to promote efficiency in their

consideration and allow for better integration. Although the wording of proposed § 50.155(b)(2)

differs from that of § 50.54(hh)(2), no substantive change in the requirements is intended.

The preamble to § 50.155(b)(2) that is contained in § 50. 155(b) is worded so that it

would require that licensees "develop, implement, and maintain" the strategies and guidance

required in § 50.155(b)(2) rather than using the wording of § 50•54(hh)(2) to require that

licensees "develop and implement" the described guidance and strategies. The addition of the

word "maintain" was proposed in order to correct an inconsistency with the wording of

§ 50.54(hh)(1), which was promulgated along with § 50.54(hh)(2) in the Power Reactor Security

Rulemaking, issued on March 27, 2009 (74 FR 13926), and to clarify that the NRC considers

the plain language meaning of the transitive verb "to implement," "to put into effect," as it was

used in the context of § 50.54(hh)(2) as including maintenance of the resulting guidance and

strategies. The wording of the requirement as it was originally issued in the Interim

Compensatory Measures Order, EA-02-026, dated February 25, 2002, was worded to require

licensees to "develop" specific guidance, while the corresponding license conditions imposed by

the conforming license amendment was worded to require each affected licensee to "develop

and maintain" strategies. The NRC believes that the phrase "develop, implement, and maintain"

would provide better clarity of what is necessary for compliance with the requirements without

substantively changing the requirements.

•rn e-n "in "ana ~Ly ana guidelines to mitigate m cne~ ec~3 ~ 1 ~Lj~

• s LO i,,,iiiiiiiroractual slgniltlcantcaamageuuii n eactor vessel orp~tful o

L.i;1 Juu strat-e-gies a-nodguideiines Lu Support aCI~tSifnlIte-nded

97
The phra "'arresting the progression of fuel damage" means stabilizing t odtion of

fuel damage and prey ting further fuel damage. Examples of these actions/euld be restoring

ameans oreorigthe r covse water lee rrefilling the speny uel posuhthat •

sufficient heat can be removed t tabilize conditions and prevent f hrfe degradation. The

phrase 'maximizing the duration for w *hcontainment capab i maintained" means

removing sufficient mass and energy from nt0ainment to 5revent the ultimate failure of the

containment which could lead to an uncontrolle adogical release to the environment. One

example of an action that accomplishes this wo ud b venting the containment 0nly as

necessary to maintain containment pressu, slightly belo the Primary Containment Pressure

Limit (i.e., the pressure limit at whichntainment integrity c be assured assuming maximum

decay heat load) thereby mainta ng the containment integrity ca ability as long as possible

which also serves to minimize radiological releases to the environmen' The phrase "minimizing

radiological releases eans that the strategies and guidelines contain gui ace to initiate•

radiological releas in a manner which has the least impact to public health a d afety yet still

protects aga'st the uncontrolled loss of the containment function. Examples coul include

directio o vent only as necessary to prevent exceeding the contain~ment des~ign pressre imit

Proposed § 50.155(b)(,) would establish requirements for licensees to integrate theJ

strategies and guidelines in (b)(l)•e•" (b)(•) with EOPs. The Commission's intent regarding

integration of strategies, guidelines, and procedures was introduced in the section-by-section

98
analysis of the proposed § 50.155(b) requirement for an integrated response capability and is

described further under "Integration with EOPs" of Section D, Proposed Rule Regulatory Bases.

Proposed § 50.155(b)(% would establish requirements for licensees to provide the

staffing necessary for having an integrated response capability to support implementation of the

strategies and guidelines in proposed (b)(1)-(b)(3). The number and composition of the

response staff should be sufficient to implement mitigation strategies intended to maintain or

restore the functions of core cooling, containment, and spent fuel pool cooling for all affected

units. The word "sufficient" is used in the proposed paragraph to reflect its meaning "adequate."

Proposed § 50.155(b)(• would establish requirements for licensees to have a

supporting organizational structure with defined roles, responsibilities, and authorities for

directing and performing the guidelines in (b)(1)-(b)(3).

Proposed § 50.155(c) Equipment requirements

Proposed § 50.155(c)(1) would require that equipment relied on for the mitigation

strategies of proposed paragraph (b)(1) have sufficient capacity and capability to simultaneously

maintain or restore core cooling, containment, and spent fuel pool capabilities for all the power

reactor units and spent fuel pools within the licensee's site boundary.

The phrase sufficient "capacity and capability" in proposed § 50.1 55(c)(1) means that the

equipment and the instrumentation relied on to support the decision making necessary to

accomplish the associated mitigating strategies of § 50.1 55(b)(1) should have the design

specifications necessary to assure that it will function and provide the requisite plant information

when subjected to the conditions it is expected to be exposed to in the course of the execution

of those mitigating strategies. These design specifications would include appropriate

consideration of environmental conditions that are predicted in the thermal-hydraulic and room

99
heat up analyses used in the development of the mitigating strategies responsive to

§ 50.155(b)(1).

Proposed § 50.155(c)(2) would require reasonable protection of the § 50.155(b)(1)

equipment rather than the treatment of SSCs important to safety under GDC-2, which requires

that those SSCs be designed to withstand the effects of natural phenomena without loss of

capability to perform their safety functions. The phrase "reasonable protection" was initially

proposed in Recommendation 4.2 of the NTTF Report in the context of a proposed NRC Order

to licensees to require "reasonable protection" of equipment required by § 50.54(hh)(2) from the

effects of design-basis external events along with providing additional sets of equipment as an

interim measure during a subsequent rulemaking on prolonged SBO. The NTTF based this

recommendation on the potential usefulness of the EDMGs in circumstances that do not involve

loss of a large area of the plant and explained that reasonable protection from external events

as used in the NTTF Report meant that the equipment must "be stored in existing locations that

are reasonably protected from significant floods and involve robust structures with enhanced

protection from seismic and wind-related events."

The NRC carried forward the use of the phrase "reasonable protection" in

Order EA-12-049 with regard to the protection required for equipment associated with the

mitigation strategies. That Order did not, however, define "reasonable protection." The NRC

guidance in JLD-ISG-2012-01 discussed "reasonable protection" as follows:

Storage locations chosen for the equipment must provide protection from

external events as necessary to allow the equipment to perform its function

without loss of capability. In addition, the licensee must provide a means to bring

the equipment to the connection point under those conditions in time to initiate

the strategy prior to expiration of the estimated capability to maintain core and

spent fuel pool cooling and containment functions in the initial response phase.

In JLD-ISG-2012-01, the NRC endorsed NEI 12-06, Revision 0, as providing an

acceptable method to provide reasonable protection, storage, and deployment of the equipment

100
associated with Order EA-12-049. The NEI 12-06, Revision 0, also omitted a definition for the

phrase "reasonable protection," but did provide guidelines for use by licensees for protecting the

equipment from the hazards that would be commonly applicable: 1) seismic hazards; 2)

flooding hazards; 3) severe storms with high winds; 4) snow, ice and extreme cold; and 5) high

temperatures. These guidelines included the use of structures designed to or evaluated

equivalent to American Society for Civil Engineers (ASCE) Standard 7-10, "Minimum Design

Loads for Buildings and Other Structures," for the seismic and high winds hazards, rather than

requiring the use of a structure that meets the plant's design basis for the Safe Shutdown

Earthquake or high winds hazards including missiles. The NEI 12-06 guidelines also allow

storage of the equipment above the flood elevation from the most recent site flood analysis,

storage within a structure designed to protect the equipment from the flood, or storage below

the flood level if sufficient time will be available and plant procedures would address the need to

relocate the equipment above the flood level based on the timing of the limiting flood

scenario(s). The NEI 12-06 guidelines further provide that multiple sets of equipment may be

stored in diverse locations in order to provide assurance that sufficient equipment will remain

deployable to assure the success of the strategies following an initiating event.. The NRCendorsed guidelines in NEI 12-06 do not consider concurrent, unrelated beyond-design-basis

external events to be within the scope of the initiating events for the mitigating strategies. There

is an assumption of a beyond-design-basis external event that establishes the event conditions

for reasonable protection, and then it is assumed that the event leads to an ELAP and LUHS.

But, for example, there is not an assumption of multiple beyond-design-basis external events

occurring at the same time. As a result, reasonable protection for the purposes of compliance

with Order EA-1 2-049 would allow the provision of specific sets of equipment for specific

hazards with the required protection for those sets of equipment being against the hazard for

which the equipment is intended to be used.

101
The NRC proposes to continue the use of the phrase "reasonable protection" in

proposed § 50.1 55(c)(2) in order to distinguish the character of the required protection of

GDC-2, which requires that SSCs important to safety be designed to withstand the effects of

natural phenomena, from that of proposed § 50.155(c)(2), which would allow damage to or loss

of specific pieces of equipment so long as the capability to use some of the equipment to

accomplish its intended purpose is retained. "Reasonable protection" would also allow for

protection of the equipment using structures that could deform as a result of natural phenomena

so long as the equipment could be deployed from the structure to its place of use.

The remaining portion of proposed § 50.1 55(c)(2) would set the hazard level for which

"reasonable protection" of the equipment must be provided. The hazard level would be the level

determined for the design basis for the facility for protection of safety-related SSCs from the

effects of natural phenomena, or, for the seismic or flooding hazards, the greater of the hazard

level determined for the design basis for the facility and the licensee's reevaluated hazards,

stemming from the March 12, 2012, NRC letter issued under § 50.54(f). The timing for the

proposed requirement for reasonable protection against the reevaluated hazards is set by

§ 50.1 55(c)(2)(i) at two years following the effective date of the proposed rule. Operating power

reactor licensees that were requested to reevaluate their seismic and flooding hazard levels by

the NRC by letter dated March 12, 2012 under 10 CFR 50.54(f) are currently on a submittal and

NRC review schedule to have confirmation of the reevaluated hazard levels by December 2015.

Given that the rulemaking schedule for this proposed rule is to provide the final rule to the

Commission in December 2016, the anticipated effective date of the final rule would be mid-tolate 2017. Requiring compliance within two years following the effective date of the final rule

would allow licensees with a new hazard level the opportunity to take measurements to support

any necessary plant modifications during the first refueling outage following NRC confirmation of

those levels and the opportunity to implement those modifications in a subsequent refueling

102
outage after the effective date of the rule. The NRC is requesting feedback on this proposed

implementation schedule in section VII of this notice.

Proposed paragraph (c)(3) would require that licensees perform adequate maintenance.

on the equipment relied on for the mitigation strategies responsive to proposed paragraph (b)(1)

to assure that the equipment is capable of fulfilling its intended function following a beyonddesign-basis external event. The phrase "adequate maintenance" means sufficient routine

maintenance and testing are performed, reflecting the storage and readiness conditions of the

equipment, for a licensee to conclude that the equipment is capable of performing its function to

a degree that would support the successful execution of the mitigation strategies of paragraph

(b)(1). Provision of "adequate maintenance" also entails the establishment of a system of

programmatic controls for the equipment to limit the quantity of equipment taken out of service

for maintenance and testing in order to limit the unavailability of that equipment appropriately

and to provide assurance that sufficient equipment will remain available to satisfy proposed

paragraph (c)(1).

Proposed paragraph (c)(4) would make generically applicable the requirements of

Order EA-12-051 by requiring that licensees include a reliable means to remotely monitor widerange spent fuel pool levels to support effective prioritization of event mitigation and recovery

actions.
pro sss. This paragraph would require new reactor designs to include design features tha•

provide nhanced capabilities for addressing these events. The term "design features"means

SSCs, mdincng the physical arrangement of such SSCs, and their functional capa ~ilities--key

characteristic of the SSCs that result in their contribution to maintain or restor h key safety

functions. Examles of design features are the inclusion of equipment su s passive or

steam-driven coolin systems that might be supported by dc power. E ,,mples of functional

capabilities of a design •ature are the flow capacity of a pump or th eelectrical capacity of a

power supply. ')up

Proposed § 50.1 §5(d) uId also require that the ap cat include design features in the

plat sffiien dsigt adres teffctsofthe evn . The term "include design features in •

the plant design" means that the plant sign would* dude those design features that support ,.-

the key safety functions. For example, if thd ner determines that the key safety functions

would be more effectively maintained by in dia specific installed component rather than

relying on portable equipment, the pladsgco include a design feature.

Further, § 50.155(d) would /equire design fausthat enhance coping durations and

minimize reliance on human/ions to maintain or restore or~e cooling, containment, and spentfuel pool cooling capabii~s. The phrase "enhance coping du tons' in proposed § 50.155(d)

means that the desi features increase the amount of time that c recooling, containment, and

spent fuel pool oling capabilities can be maintained in the initial Ph e of the event before the•

transition f the permanently installed SS~s to portable equipment. Th term "minimize

relianc°n human actions" means that the design features either obviate or re uce, tO the

extnt practical, the need for operator actions during the initial phase of the event. eTNRC

hos "mnimze"versus "reduce" in paragraph (d) to requieteapicant to design the p n~t to
duigti iinso ioactions tht ) operators would be able to focus more timand

attention on monitoring pln ditions and planning for the potential trninto relying on

Sportable equipment and offsite reso e=s; and 2) there is lower pote l for human failures

during adverse, stressful conditions with the jectv of achi g a lower risk of failure to.

execute the mitigation strategy. .

The condition described in § 50.155d an ELAP curent with either a loss of

normal access to the ultimate heat sin •r, for passive reactor des lgs, a loss of normal access

Proposed § 50. 155(i) Train~ig requirements

Proposed § 50.155(f) would require that each licensee specified in § 50.155(a) provide

for the training and qualification of personnel that perform activities in accordance with the

2

strategies and guidelines identified in § 50.155(b)(1)-(b)(;i).

Prop osed § 50. 155(/ Drills and Exercises(

Proposed § 50.155(2)' would require that each licensee and applicant specified in

§ 50.155(a) conduct drills and exercises for personnel, who would perform activities in

accordance with the strategies and guidelines identified in § 50.155(b)(1)-(b)(:5. The use of

drills and exercises allows demonstration and evaluation of the licensee's capability to execute

105
the in~tegrated response capability required by § 50.155(b) mitigation strategies and guidelines

in light of the specific plant damage and operational conditions presented by an initiating event.

"Integrated" is used to describe the licensee's or applicant's approach to using all tools, spaces,

qualified personnel and resources during a performance enhancing experience to the furthest

extent practical given a set of initiating conditions and within the bounds of a drill or exercise

scenario. When two or more strategies or guidelines in § 50.155(b)(1)-(b)(,) are potentially

useful, "integrated" is meant that transitions to and from one set of strategies or guidelines in

§ 50.1 55(b)(1 )-(b)(•" to another are coordinated.

This proposed rule uses the words "drill" and "exercise" as they are defined in

NUREG-0654/FEMA-REP-1, Revision 1,8 meaning an evaluated performance-enhancing

experience that reasonably simulates the interactions between appropriate centers, work

groups, strike teams, or individuals that would be expected to occur during the event. For the

initial drill or exercise, the licensee would be required to demonstrate its capability to transition

to and Use one or more of the strategies that would be required by § 50.1 55(b)(1 )-(b)(,3) from j

the AOPs or EOPs, whichever would govern for the initiating event and plant degrading

conditions, using the equipment and communication systems used for the EOPs and guidelines.

Propsed§50.155(•(1) would require the initial drill or exercise to be conducted within

12 months prior to the issuance of the first operating license (OL) for the unit described in the

application. This would allow the license applicant to implement any improvements Or corrective

actions identified during the drill or exercise, and allow the Commission to consider the results

of any drill or exercise actions in the decision whether to authorize the OL. Because

§ 50.155(.f)(1) applies only to applicants for operating licenses, it would not apply to holders of1

B Planning Standards N.1 Exercise and N.2 Drills.

106
operating licenses under 10 CFR part 50, who are subject to proposed § 50.155(2)(4), or holders

of combined licenses under 10 CFR part 52, who are subject to proposed § 50.1 55(•)(2) - (4).

e

Proposed § 50.155•t(2) would require the licensee to conduct an initial drill or exercise

that demonstrates the capability to transition from the AOPs or EOPs, use one or more of the

strategies and guidelines in paragraphs (b)(1)-(b)(2) of this section, and use communications

equipment required in 10 CER part 50, appendix E, section VII no more than 12 months before

the date specified for completion of the last inspections, tests, and analyses in the inspections,

tests, analyses, and acceptance criteria (ITAAC) completion schedule as required by § 52.99(a)

for the unit described in the combined license.

This proposed rule would set the completion date for the initial drill or exercise at "no

more than 12 months before the date specified for completion of the last inspections, tests, and

analyses in the ITAAC completion schedule required by § 52.99(a) for the unit described in the

combined license" in order to allow the licensee to implement any improvements or corrective

actions identified during the drill or exercise, and allow the Commission to consider the results

of any drill or exercise actions in the decision whether to make the § 52.103(g) finding. C.

The proposed § 50.155(,f(2) requirement for initial drills or exercises is limited to holders

of combined licenses under 10 CFR part 52 before the Commission has made the finding under

§ 52.103(g). A combined license holder for whom the Commission has already made the

finding under § 52.103(g) as of the effective date of the rule would not be subject to proposed

§ 50.1 55 (2), but would instead be subject to § 50.1 55(,2)(4) for the proposed initial drill

e

requirements. Proposed § 50.155(2(3) would require operating power reactor licensees under

10 CFR part 50 and those under 10 CFR part 52 for whom the Commission has made the

finding under § 52.103(g) to conduct subsequent drills, exercises, or both that collectively

demonstrate a capability to use at least one of the strategies and guidelines in each of proposed 2' §50.155(b)(1)-(b)(i) in succeeding 8-year intervals. This would require that the drills and J

107
exercises performed to demonstrate this capability include transitions from other procedures

and guidelir•/js applicable, and the use of communications equipment that would be required

by proposed 10 CFR part 50, appendix E, section VII. This proposed requirement differs from

_..

the proposed § 50.155(•(1) and (•'2) initial demonstration requirement, in that it would require

licensees to demonstrate a continuing capability, and as such, it is structured to require

licensees to demonstrate at least one of the strategies and guidelines from each of the

guidelines during the 8-year interval.

L

Proposed § 50.1 55(•(4) would require holders of operating licenses or combined

licenses for which the Commission has made the finding under § 52.103(g) to conduct an initial

drill or exercise that demonstrates the capability to transition to and use one or more of the

strategies and guidelines in proposed § 50.155(b)(1)-(b)(,8) and use communications equipment

required in 10 CER part 50, appendix E, section VII. Proposed § 50.155(#')(4) would be

e

equivalent to proposed § 50.155(•(1) and (2) for initial drills or exercises, but would apply to

current licensees. Following this initial drill or exercise, the licensee would be required to

conduct subsequent drills, exercises, or both that collectively demonstrate a capability to use at

least one of the strategies and guidelines in each of proposed § 50.1 55(b)(1 )-(b)(.•) in)

ee

§ 50.1 55(.f)(3) for subsequent drills or exercises, but would apply to current licensees under

10 CFR part 50 and those under 10 CFR part 52 for whom the Commission has made the

finding under § 52.103(g) as of the effective date of the rule.

Proposed § 50. 155(&) Changej Control

Proposed § 50.155(a) would establish requirements that govern changes in the

implementation of the requirements of proposed § 50.155 and 10 CFR part 50, appendix E,

section VII. Prior to implementing a proposed change, proposed § 50.155(A)1) would require.

108
the licensee to perform an evaluation to ensure that the provisions of proposed § 50.155 and

10 CFR part 50, appendix E, section VII, continue to be met. Proposed § 50.155(•)(2) would

require that licensees maintain documentation of the paragraph ($)(1) evaluations until the

requirements of this proposed § 50.155 and 10 CFR part 50, appendix E, section VII, no longer

apply. Finally, proposed § 50.155 (3) would inform licensees that proposed changes must

continue to be subject to all other applicable change control processes.

rrpsd§5u 3•~) would require a licensee who was, as an appl•li,~JJ~At

Proposed § 50.1 55(• would set schedules for compliance for different classes of

licensees depending on the circumstances unique to each class. Paragraphs (•(1) and (2)

would require licensees of operating reactors to comply with all requirements within four years

of the effective date of the rule.

Proposed 10 CFR Part 50, Appendix E, Section I, Introduction

The NRC proposes adding the sentence, "Section VII of this appendix also provides for

'Communications and Staffing Requirements for the Mitigation of Beyond-Design-Basis Events'

that do not need to be contained within a licensee's emergency plan" to the end of paragraph

1.2. The NRC is not proposing to require an applicant or licensee to address or implement the

proposed requirements in Section VII of Appendix E through the applicant's or licensee's

emergency plan or maintain the capabilities as a part of the emergency preparedness program.

This would allow for site-specific flexibility in implementation.

109
Proposed 10 CER Part 50, Appendix E, Section IV.B, Assessment Actions

The NRC proposes adding the phrase, "including from all reactor core and spent fuel

pool sources," into paragraph 8.1 following "determining the magnitude of, and for continually

assessing the impact of, the releases of radioactive materials." This proposed rule would

require all licensees to establish the capability to perform offsite dose assessments during an

event involving concurrent radiological releases from all on-site units and spent fuel pools, and

for multiple release points. The capability would quantify the total releases from the site and

estimate the offsite dose consequences.

Proposed 10 CFR Part 50, Appendix E, Section IV.E, Emergency Facilities and Equipment

The NRC proposes adding the phrase, "including from all reactor core and spent fuel

pool sources," into paragraph E.2 following "equipment for determining the magnitude of, and

for continuously assessing the impact of, the release of radioactive materials to the

environment." The proposed rule would require that equipment used for multi-unit dose

assessment be maintained in a ready state.

Proposed 10 CER Part 50, Appendix E, Section IV, Training

The proposed rule would move the § 50.54(hh)(2) exercise requirement from 10 CFR

part 50, appendix E, section IV.F.2.j to § 50.155(,•'. This move would change the exercise/

requirement to a drill requirement, aligning the requirement with the mitigation strategies drill

requirements described in § 50.155(•.

The proposed rule would also require that periodic opportunities for a performance

enhancing experience should be provided to personnel responsible for performing multi-unit

dose assessment and assessing the results in accordance with the site's emergency plan and

implementing procedures.

110
(

Proposed 10 CFR Part 50, Appendix E, Section VI, Emergency Response Data Systems

The NRC proposes to change its Emergency Response Data Systems regulations to

require the use of technology-neutral equipment. The NRC proposes to restate the

requirements in paragraph 3.c to replace the phrase "onsite modem" with "equipment" and

removing references to a specific "unit" or equipment use.

Proposed 10 CFR Part 50, Appendix E, Section VII, Communications and Staffing

Requirements for the Mitigation of Beyond-Design-Basis Events

Proposed section VII would require power reactor applicants and licensees to conduct a

detailed analysis to provide the basis for the staffing necessary for responding to a beyonddesign-basis external event as described in § 50.155(b)(1) during an extended loss of ac power

(ELAP), and while access to the plant and normal access to the ultimate or normal heat sink are

lost. Additionally, the proposed section VII would require power reactor applicants and

licensees to maintain at least one onsite and one offsite communications system functional

during an ELAP and a loss of the local communication infrastructure.

The current rule in 10 CER part 50, appendix E, section IV.E.9 requires, "At least one

onsite and one offsite communication system; each system shall have a backup power source."

However, the current rule doesn't address an interruption in the offsite communication services.

The proposed rule would require the power reactor applicants and licensees to maintain the

communication capabilities of communication amongst onsite staff and between onsite staff and

offsite personnel in light of the lessons learned at Fukushima Dai-ichi. Furthermore, the

proposed rule would require the power reactor applicants and licensees to submit the staffing

analysis, results and implementation plans to meet the requirements, and the submissions

111
would afford the NRC the opportunity to identifY any common industry implementation problems

and address them in guidance.

The proposed rule would require an applicant for an OL to complete a detailed staffing

analysis at least 2 years before the issuance of the first operating license for full power (one

authorizing operation above 5 percent of rated thermal power). The time frame allows thle

applicant to implement any improvements or corrective actions identified during the analysis,

and the results of any analysis to inform the Commission's decision in authorizing the CL.

The proposed rule would require that an applicant for a COL conduct a detailed staffing

analysis and submit the analysis and results to the NRC 2 years before the date specified for

completion of the last inspections, tests, and analyses in the inspections, tests, analyses, and

acceptance criteria (ITAAC) completion schedule as required by § 52.99(a) for the unit

described in the combined license. The time frame allows the applicant to implement any

staffing and communications system improvements and corrective actions identified during the

analysis, and the results of any analysis to inform the Commission's decision in issuing the

§ 52.103(g) finding.

The proposed rule would provide that when the NRC has docketed the certifications

described in § 50.82(a)(1) or § 52.110(a) for a power reactor licensee, then that licensee would

no longer be subject to section VII of appendix E to 10 CFR part 50 for the unit described in the

§ 50.82(a)(1) or § 52.110(a). certifications.

,•-ctio," 62.17 idzntifizs- th. r..u... t--oc *hni•:: information to he innhiided in an•

--alifcatnon Tor a Slandl. c.in _rt_!i¢finn~ Th9 prc-"p•--e iul wuuldIGv •thi3 zoctio" b•,

cknga new paragraph (a )-(L2j :•ui~. aa --p"'coti'r f.. ......... 3incotiiotir
would py only to standard design certification applications, and would not apply totandard

design certi ations issued before the effective date of this rule. Thus, any stand, rd design

certification app ation that is docketed and is Under review by the NRC as ofh effective date

ofprpoethis proposed§ 5.5()rule oul need to amend its application to include/teinrationreuedb

Proposed § 52. 79 Contents of Apietos Technical In form io in Final Safety Analysis

Report •

Section 52.79 identifies the requir technicanorato to be included in an

application for a combined license. The prop rule ,•-would revise this section by adding a

newpargrah a)(8) eqirig echappli to combined license that is subject to

proposed § 50.155(d) to include a ds tinof the d ignf~veatures included in the plant design

under proposed § 50.155(d) and a exlanation of how tho e design features would comply

with the requirements of prope § 50.155(d). The proposed qirement would not apply to

combined licenses issueefore the effective date of this rule, nor ould it apply to combined

licenses referencingstandard design certification, a standard design proval, orea

des tuig n ation, ;neTu'aCcertif a standard de s•ig ap•prv I i ao pltor, manufa cturing efrncr license, tha "ts te d ockete: d

and un r eview bY the NRC as of the effective date of this proposed rule would nee t e /

a nded to include the information required by proposed § 50.155(d). -

113
Proposed § 52. 80 Contents of Applications; Additional Technical Information

Section 52.80 identifies the required additional technical information to be included in an

application for a combined license. Proposed paragraph (d) would be amended to require a

combined license applicant to include the applicant's plans for implementing the requirements of

proposed § 50.155 and 10 CFR part 50, appendix E, section VII, including a schedule for

achieving full compliance with these requirements. This paragraph would also require the

application to include a description of: 1) the integrated response capability that would be

required by proposed § 50.155(b); 2) the equipment upon which the strategies and guidelines

that would be required by proposed § 50.155(b)(1) rely, including the planned locations of the

equipment and how the equipment and SSCs would meet the design requirements of proposed

§ 50.155(c); and 3) the strategies and guidelines that would be required by proposed

§ 50.155(b)(2).

application standard design approval. The proposed rule would rei his section by

adding a new paragr (a)(7) that would require each appliai or a standard design I

standard design approvlsued before the effective date of this r Tus any standard

design approva a dl cation that is docketed and under review by the NRC as h ffective

date of thire~would need to be amended to include the information required bY prop• d
S in52.157 identifies the required technical information to be included in the fina

Ssafety analysi e~port as part of an application for a manufacturing license. The propose'rule

Swould revise this s tion by adding a new paragraph (f)(33) that would require each p~lication

for a manufacturing lices that is subject to proposed § 50.155(d) to include a d rption of

the design features include *~kn the plant design under proposed § 50.155(d) dan explanation

of how those design features wo d ,comply with the requirements of prop sed• § 50.155(d). J

Thus, the proposed requirement wou not apply to manufacturing lic 'ses issued before the

effective date of this rule, nor would it app• tmanufacturing -.-- lice s applications referencing a i 

standard design certification or standard desigh approval. T s, any manufacturing license

application, not referencing a standard design, that* do eted and under review by the NRC as

of the effective date of this proposed rule would nee to e amended to include the information

required by proposed § 50.155(d). _i

VI euaoyOversig o SevereAccidentMan ement Guideline.

As discussed in section IV/ . of this notice, the NRC's intent f ~regulatory oversight of

the SAMGs is limited to/oe ht through inspection. The NRC cosd d including the

SAMGs as an inspectio, Irprogram element of the baseline inspection progra• of the reactor

oversight process O) under the emergency preparedness cornerstone dunrteI

development the ROP. At that time, the scope of the inspection program elemen consideredI
SChapter ,IMC) 0308, "Reactor Oversight Process Basis Document," Attachment 2, "Tci I

Is for Inspection Program," documents that the NRC "staff concluded that regular i pction

'of SA• was not appropriate because the guidelines are voluntary and have nrguatory

,basis. The mlergency response organization that would implement SAMGs i inspected

through EP [enrgency preparedness] baseline inspection and performa e is covered by two

PIs [performance in cators]."

Subsequent to tl Fukushima Dai-ichi accident, the NRC pfomdan inspection under

TI-25151184, "Availability an Readiness Inspection of Severe/cidn Management

Guidelines (SAMGs)'" in order t deermine that the SAMG were available and how they were

being maintained and to determine tl nature and exte, o•f licensee implementation of SAMG

training and exercises. The results of th*seto conducted under TI-2515/1 84 informed

the NTTF recommendations and the NRC ap ch for the proposed regulatory oversight of

SAMGs. This proposed rule would require,1 enstesto develop, implement, and maintain

SAMGs, and as such, the NRC would n include S kvGs within the scope of the ROP.

The NRC's intent for inspecti n of SAMGs is simi rin nature to the inspections that

were conducted under TI-251511 4 and those that are cond ted as part of the baseline

inspection program for the m lgating/strategies under the curren § 50.54(hh)(2) using

Inspection Procedure (IP /1111 .05T, "Fire Protection (Triennial)," A'IAMS Accession No.

ML12328A158, and I 71111.O5XT,. "Fire Protection - NFPA 805 (Trienn l),"'ADAMS

Accession No. MI 2328A167. The NRC will use the results of the inspectio p~erformed under

TI-251 5/191,"nspection of the Licensee's Responses to Mitigation Strategies Or re EA-1 2-049,

Spent Eu Pool instrumentation Order EA-12-051 and Emergency Preparedness Info ation

Req sted in NRC March 12, 2012," to inform the development of the inspection procedu foer

,AMGs and will ultimately use IMC 0308, Appendix B, "Reactor Oversight Process

li )
V I .• 8eii Requests for Comments.

The NRC is seeking advice and recommendations from the public on the proposed rule.

We are particularly interested in comments and supporting rationale from the public on the

followina:

1. Change Control. The provisions governing change control in proposed § 50.155(•) do

not contain a criterion or a set of criteria that would establish a threshold beyond which

prior NRC review and approval would be necessary to support a proposed change to the

facility impacting the beyond-design-basis aspects of this proposed rulemaking and its

supporting implementation guidance. Tc*R ~fc ~ ~ tasto hen

thtcudb barorjuaging cnarnges tome iaiil - - _itin th,, propo. ,.ed (bjd)(1.•

mii•i~ trntog.ic ...... 'b differ,,'z;. t c•° ,u iui, ,u l, , ,L~i a uvcmi;'c..g

,,o._.AM'gs-of pro pooacd "-(~ .... ui~ "rrsil ... .. :LL.•.^.i h4 hrl~ o

example, a set of criteria that asks whether a proposed facility change adversely impacts

the capability to maintain and restore core cooling, containment, and spent fuel pool

cooling capabilities, in conjunction with a criterion that asks whether the proposed facility

change adversely impacts the supporting equipment requirements in proposed

paragraph (c) might be sufficient for judging whether changes to the facility that impact

the implementation of the mitigation strategies of proposed (b)(1) require prior NRC

review and approval. "H~uw~v~i, Lhls set OT criteria may not b• up, i1,,,l for ,.AMvGs, an6"-•

b•-:8e •-f "rt~i f..c,3 •,,k ,n.,hothr /thoP pr•"~ i avoa

'mplct3 l• •th~ZMbio e perspective ot supportring Inio" " m , ...kng in 5:"

•Wa are stakeholders' views on this proposed change

/

N

117
control structure, and what do stakeholders suggest for revising the change control

process to contain criteria for determining the need for prior NRC review and approval?

2. Application of Other Change Control Processes. Proposed § 50.155(•(3) contains-a

requirement for licensees to use all applicable change control processes for facility

changes, and not simply apply~oroposed paragraph (• (i.e., the proposed change

control process of paragraph.(•) is only applicable to facility changes with respect to

their beyond-design-basis aspects and to the extent that such changes impact

implementation of the requirements of proposed § 50.155 or the proposed

10 CFR part 50, appendix E, section VII) to the exclusion of other change control

processes. This recognizes that facility changes can impact multiple aspects of the plant

having different applicable requirements, and being subject to different change control

requirements. For example, a licensee may want to make a facility change (e.g., a

physical connection device) to support implementation of the beyond-design-basis

external event mitigation strategies, and this change might impact safety-related SSCs.

In addition to applying the new change control provision to ensure beyond-design-basis

aspects of the proposed change result in continued compliance with the new

requirements of this proposed rule, the licensee would also need to apply 10 CER 50.59

to ensure that the facility change does not, due to its impact on safety-related SSCs,

require prior NRC approval. The NRC requests feedback on the need for this proposed

provision, or suggestions on how it might be improved.

3. Reasonable Protection. The proposed rule language contains a requirement in

proposed § 50.155(c)(2) that equipment supporting the proposed mitigation

requirements of paragraph (b)(1) be "reasonably protected" from the effects of natural

phenomenon including both those in the current plant design basis as well as the

118
reevaluated hazards under the March 12, 2012, § 50.54(f) request concerning flooding

and seismic hazards. As a practical matter, implementation of Order EA-12-049 began

before the reevaluated hazard information was available. The NRC recognizes that

licensees were mindful of the hazard information, and attempted to address it during

implementation. The NRC requests feedback concerning any costs and impacts that

licensees would expect to occur as a result of this proposed requirement to include such

things as rework or changes to previously implemented mitigation strategies.

NRC would e stakeholder views on alternative regulatory approaches thatn achieve

the objective (e~. nuig through inspection that SAMGs are update to reflectth

dsractint pofrso-urcesh frma modre importaine as (fonfgrsaft)an hi anyacst/mpat

informationy thpact. wolnbeThe NRC osuggtiputcudacate tesprnting regulatoryanls.

5. New Reactor Requirements. Te proposed rule language inclesa requirement for

Sapplicants for new nude power reactors to include design feature in he plant design

sufficient to enhanccoping durations and minimize reliance on human ions to

maintain or e re core cooling, containment, and spent fuel pool cooling cap iites.

(key saf u functions) tnCi ndu during ahn ELAP uPconcurrent r upstlfsc with loss o°f normal access to e~arqieetwu~ thl•b• 

119
for a new reactor design to have greater reliance on installed equipment, less relianc

on operator action, and thus more time for diagnosis, planning, and preparation fort

trsition to reliance on portable equipment and offsite services. While this requi ement

woul ocus on potential improvements in the plant design, it would not obvia the need

fra licen ee using this design to develop mitigating strategies to depend n portable

equipment a' offsite resources. Further, this requirement would only ppyto

applicants who a ,,seeking approval of the key safety functions, an wudnot apply to

an applicant reeren gi a design where those key safety functio have been previously

approved by the NRC. hat are stakeholders' views on the edfor such a

reqirmetIsthrea etr ay of accomplishing the ormnindunderlying

purpose? Should the requirem 7t be imposed on ohaplicants or licensees (i.e.,

regardless of whether the NRC ha ~reviously approved the design of the key safety

funtins) ad f s, nde watbasis. Wha re stakeholders' views on the costs and

benefits of such a requirement? Please I;vide any available information to support

your comments.

II

I

6. Transition to SAMGs. Propos § 50.155(b)(3) w ,Id require strategies and guidelines

for mitigating the consequ csof events that "progtre a to imminent or actual significant

damage" to fuel in the ecrvselor spent fuel pool. else types of strategies and

guidelines have h~trclybeen referred to as SAMGs. The •RC understands that in

S practice, tranrio to the SAMGs would occur if there are availab instrumentation

readingseg,, core exit temperatures exceed 1200 degrees Fahren it) or there is an

abs c of available instrumentation readings but circumstances indicat tothe

peaosthat actual or imminent damage to fuel, may have occurred. The a ual

condition of the fuel is not known, and as such it is not possible to know if "signific nt

~~~120•...
Li/ Mitigation ofBeyond-Design-Basis EvnsStaffing Analysis. Proposed 10 CFR part 50,

appendix E, section VII, would require an analysis for the staffing necessary to support

mitigation of a beyond-design-basis external event. This requirement would supplement

the separate staffing analysis requirement that already exists in 10 CFR part 50,

appendix E, section IV.A.9. The reason for the two separate staffing analysis

requirements is related to the historical imposition of the requirements for the staffing

analyses in the emergency preparedness rulemaking of 2011 and the March 12, 2012

Request for Information under 10 CFR 50.54(f). The NRC is seeking feedback on

whether it would be more efficient in practice for the two staffing analyses and their

corresponding requirements to be combined, particularly for future reactor applicants.

Would there be any unintended consequences to keeping the analyses separate or

combining them? Is there a better way of achieving the underlying purpose of this

requirement?

,•. Training Requirements. Section 50.155f) of the proposed rule would require licensees

to provide for the training and qualification of personnel that perform activities in

accordance with the strategies and guidelines identified in parag,?phs (b)(1)-(b)(/) (i.e.,

mitigation strategies for beyond-design-basis external events4 extensive damage

mitigation guidelines, a Lo'._'cre ,cdJ 1,t m..nogc,-.,ct g'-idcline) using the SAT

121

(
process as defined in § 55.4. The NRC notes that whereas many individuals at licensee

facilities that would be subject to the proposed rule are trained under the SAT process

(e.g., individuals Specified under § 50.120), some individuals (e.g., firefighting and

emergency preparedness personnel) may be currently trained under programs that are

not required by NRC regulation to use the SAT process (e.g., National Fire Protection

Association standards for training and 10 CER part 50, appendix E). It is not the NRC's

intent to extend the requirement for SAT-based training to the entirety of such programs.

Rather, the intent of the proposed requirement would be to ensure that any training that

is not currently part of existing programs but would be needed for performing activities in

accordance with the strategies and guidelines identified in paragraphs proposed

§ 50"155(b)(1)-(b)••' be identified and provided for in accordance with the SAT process".

The NRC requests comment on potential unintended consequences of the proposed rule

language for programs not currently required to be SAT-based and if unintended

consequences are identified, proposed alternative language for requiring the necessary

amendments to such programs.

Drill or Exercise Frequency. Proposed § 50.155•(3) and § 50.155(•(4) would require I

that following an initial drill or exercise, licensees would be required to conduct

subsequent drills, exercises, or both, that collectively demonstrate a capability to use at

least one of the strategies and guidelines in each of proposed § 50.155(b)(1)-(b)(• in

succeeding 8-year intervals. This would require that the drills or exercises performed to

demonstrate this capability include transitions from other procedures and guidelines as

applicable, and the use of communications equipment that would be required by

proposed 10 CFR part 50, appendix E, section VII, and that licensees shall not exceed 8

years between any consecutive drills or exercises. These requirements would be

122
separate from the 8-year emergency preparedness exercise Cycle requirements in

10 CFR part 50, appendix E, section iV.F. The NRC is seeking feedback on whether the

drill or exercise frequency proposed by § 50.155•(3) and § 5015()is appropriate./

-1 1/. Equipment Requirements. Proposed § 50.155(c)(1) would require the capacity and (

•capability of the equipment relied on for the mitigation strategies required by proposed

§ 50.155 (b)(1) to be sufficient to simultaneously maintain or restore core cooling,

containment, and spent fuel pool cooling capabilities for all the power reactor units within

the site boundary. Addition•)roposed § 50.155(c)(3) would require the equipment

relied on for the mitigation strategies in proposed § 50.1 55(b)(1) to receive adequate

maintenance such that the equipment is capable of fulfilling its intended function. The

intent of these two proposed provisions is to make elements of Order EA-1 2-049

generically-applicable. Order EA-12-049 did not contain a specific maintenance

requirement, but instead contained a performance-based requirement "to develop,

implement and maintain strategies," and failure to perform adequate maintenance would

likely lead to a :failure to meet this more general requirement, which is also contained in

proposed § 50.1 55(b)(1). Additionally, the supporting guidance for the proposed rule for

proposed § 50.155(b)(1) carries forward the same approach that was used for

implementation of Order EA-12-049, and contains a number of programmatic controls

that in an analogous fashion to the maintenance provision in proposed § 50.1 55(c)(3), if

not followed, would likely lead to a loss of equipment capacity and capability and result

in a failure to comply with the proposed § 50.155(b)(1). Therefore, the NRC would like

stakeholder views on the need for a separate maintenance provision• within the proposed

rule.

123
4--- /. Equipment PrtcinImplementation Daln.The NRC is proposing to require

licensees to reasonably protect the equipment relied upon to implement the mitigation

strategies required by proposed § 50.155(b)(1). That equipment would need to be

•reasonably protected from the effects of natural •phenomena that are, at a minimum,

equivalent to the design basis of the facility. The proposed rule would require each

licensee that received the March 12, 2012, NRC letter issued under § 50.54(f) to provide

reasonable protection against that reevaluated seismic or flooding hazard(s) by 730

days following the effective date of the final rule, if the reevaluated hazard exceeds the

design basis of its facility. This is based on the anticipated completion dates for the

licensees' hazard reevaluations and their confirmation by the NRC and the potential

need for planning and implementing modifications during refueling outages. The NRC

recognizes that certain licensees may need input into their analyses of reevaluated

hazards from other government agencies, without any certainty of when that input would

be provided. This reliance on information from other entities could remove from the

licensee's control the ability to comply with the proposed rule by a specific date. The

NRC requests comments on the proposed implementation schedule, including

suggestions for the criteria that licensees would need to satisfy to extend the schedule.

c 2. Methodology for addressing reevaluated hazards. In SRM-COMSECY-14-0037, the

Commission affirmed that: 1) licensees for operating nuclear power plants need to

•address the reevaluated flooding hazards within their mitigating strategies for beyonddesign-basis external events; and 2) licensees for operating nuclear power plants may

need to address some specific flooding scenarios that could significantly damage the

power plant site by developing targeted or scenario-specific mitigating strategies,

possibly including unconventional measures, to prevent fuel damage in reactor cores or

124
spent fuel pools. The NRC is proposing to require licensees for operating nuclear power

plants to address the reevaluated flooding hazard levels by reasonably protecting the

mitigating strategies equipment to those levels if they exceed the design-basis flood

level for the facility. Alternatively, the NRC could: 1) place this requirement within

§ 50.155(b)(1) as a condition the associated strategies and guidelines must be capable

of addressing; or 2) include a separate requirement for targeted or scenario-specific

mitigating strategies as an option to address the reevaluated flooding hazards. The

NRC seeks comment on whether the first of these options would be a better means to

communicate the need for a licensee's strategies and guidelines to be capable of

execution in the context of the new flooding hazard levels than including the requirement

in § 50.155(c)(2). The NRC seeks additional comment on whether it would be

appropriate to allow further flexibility in the licensee's strategies and guidelines by

establishing an alternative means of compliance that does not include the surrogate

condition of a loss of all alternating current power for specific beyond-design-basis

conditions such as the reevaluated flooding hazards. For example, if a licensee could

protect their internal power distribution system and emergency diesel generators from

the reevaluated flooding hazard, it may not be necessary for the licensee to assume the

loss of all alternating current power.

tO 1. Command and Control. Requirements for command and control and organizational •

structures currently exist in numerous locations, including 10 CER part 50, appendix E,

section IV&" well as within the typical administrative controls portions of technical--

specifications for power reactor licensees. These requirements do not plainly limit the

scope of the roles, responsibilities and authorities to events within the design or licensing

basis of the facility, although past NRC practice has been to treat these requirements in

that manner. The proposed rule includes a further requirement on the subject in order to

125
clarify the scope of what is required for organizational structures at power reactor

licensees. Alternatively, the NRC is considering 4-the expansion of scope of regulatory

oversight of the organizational structures would require imposition of a new requirement

or the expansion of scope would be better accomplished by communicating the

understanding that the scope of the existing requirements covers the full spectrum of

events that would be included in this rulemaking. The latter method of accomplishing

this Would have the potential advantage of leaving the requirements for command and

control and organizational structures in a single regulation (i.e., 10 CFR part 50,

appendix E, section IV.A). The NRC seeks stakeholder input on this subject.

3.lt Regulatory Flexibility Certification. (

Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC certifies that this rule will

not, if promulgated, have a significant economic impact on a substantial number of small

entities. This proposed rule affects only the licensing and operation of nuclear power plants.

The companies that own these plants do not fall within the scope of the definition of "small

entities" set forth in the Regulatory Flexibility Act or established in 10 CFR 2.810, "NRC size

standards."

~~a~b~t fRgltr nlss ! 

The NRC has prepared a draft regulatory analysis and a draft supplemental regulatory

analysis on this proposed regulation. The analyses examine the costs and benefits of the

alternatives considered by the NRC. The NRC requests public comment on the draft regulatory

analyses. The draft regulatory analyses are available as indicated in the "Availability of

126•
Documents" section of this document. Comments on the draft analyses may be submitted to

the NRC as indicated under the ADDRESSES section of this document.

2//X. Availability of Guidance.(

(p

The NRC is issuing for comment drafrlatygudneo support the implementation

of the proposed requirements in this rulemaking. You may access information and comment

submissions related to the DGs by searching on http://www.requlations.qov under Docket ID

NRC-201 4-0240.

The DG-1301, "Flexible Mitigation Strategies for Beyond-Design-Basis Events," provides

licensees and applicants with an acceptable method of responding to an ELAP and

demonstrating compliance with the proposed regulations requiring additional defense-in-depth

measures for the mitigation of beyond-design-basis external events.

The DG-1 31 7, "Wide-Range Spent Fuel Pool Level Instrumentation," describes one

method ofam^''v. o÷•='v:aln dn safety enhancements in the form of reliable (

spent fuel pool instrumentation for beyond-design-basis external events.

The DG-1 31 9, "Integrated Response Capabilities for Beyond-Design-Basis Events,"

describes one method the NRC endorses to enhance a site's ability to implement the on-site

emergency preparedness programs and guidelines and better cope with conditions resulting

from a beyond-design-basis external event.

You may submit comments on this draft regulatory guidance by the following methods:

a Federal Rulemaking Web site: Go to http://www.re~qulations.eqov and search for

Docket ID NRC-2014-0240. Address questions about NRC dockets to Carol Gallagher;

telephone: 301-415-3463; e-mail: Carol .Gallaoqher•,nrc.Qov.

127
e Mail comments to: Cindy Bladey, Office of Administration, Mail Stop: OWFN-12-

H08, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

.•B~ackfitting and Issue Finality. I

Proposed Rule

As required by §§ 50.109, 52.63, 52.83, and 52.98, the Commission has completed a

backfit and issue finality analysis for the proposed rule. The Commission finds that the backfi~z'' (

contained in the proposed rule, whcn:r •ii,=d~rud in me aggregate, wu'• c,•io;•t•t -d

indicated in the "Availability of Documents" section of this document.

Draft Regulatory Guidance

The NRC is issuing, for public comment, three DGs that would support implementation

of this proposed rule: DG-1301, "Flexible Mitigation Strategies for Beyond-Design-Basis

Events"; DG-1317, "Wide-Range Spent Fuel Pool Level Instrumentation"; and DG-1319,

"Integrated Response Capabilities for Beyond-Design-Basis Events." These DGs would provide

guidance on the methods acceptable to the NRC for complying with this proposed rule. The

DGs would apply to all current holders of nuclear power plant construction permits and

operating licensees (under 10 CFR part 50 and renewed licenses under 10 CER part 54) and

combined licenses under 10 CFR part 52, and applicants for construction permits; operating

licensees; •tnadd "ppro......... 1 ~u~aunbdladJ~g .~i. liic .. ~IL

n'M..roefer.ef- .. bIdd t~ud slu usyu~i LifiTaItion, sTandard dlesign approval, o, rna"'atri'

i~e., multi-source dose assessment, is considered part of the set of emergency \ preparedness (EP) requirements to provide reasonable assurance of adequate

protection of public health and safety, consistent with the regulatory basis for EP

that has existed for more than three decades.
"•el;s"•,dud ianuTacturlng ies LId ,,• ,•,, , ,c a 3tad d.g zctfzH~ • r

Issuance of the DGs in final form would not constitute backfitting under § 50.109 and

would not otherwise be inconsistent with the issue finality provisions in 10 CFR part 52. As

discussed in the "Implementation" section of each DG, the NRC has no current intention to

impose the-DGs, if finalized, on current holders of a construction permit, operating license, or

combined license.

Applying the DGs, if finalized, to applications for construction permits; operating

licensees; 3t"n'iord "ccign .... fo, in;^^ ... t~ndard• •cinapzl,~bndlcne htd

,.~~ refzrzn',c.z_ -z- ct.3n_•rr rl=ipn certificatinn Qtfrndrd desig C;.'3•povcil, ois=,T~~tL,,',

-3odrd ,dig•,1 ~~ , 1 uvv.iwould not constitute backfitting as defined in § 50.109 or be otherwise

inconsistent with the applicable issue finality provisions in 10 CFR part 52, because such

applicants are not within the scope of entities protected by § 50.109 or the applicable issue

finality provisions in 10 CFR part 52. Neither § 50.109 nor the issue finality provisions under

10 CFR part 52 - with certain exceptions - were intended to apply to every NRC action that

substantially changes the expectations of current and future applicants. The exceptions to this

principle for these DGs are: 1) a combined license applicant that references an already-issued

standard design certification, standard design approval, or manufacturing license, which have

their own specific issue finality provisions; and 2) a manufacturing license applicant that

references an already-issued standard design certification or standard design approval.

However, the proposed operational requirements addressed in the DGs are not within the scope

of issues that may be resolved for design certification or design approval, and therefore are not

afforded issue finality protection under 10 CFR part 52; whereas the design requirements

addressed in the DGs are expected to be resolved during the design phase (e.g., design

129
certification applications) and therefore would be afforded issue finality protection under 10 CFR

part 52. Manufacturing licenses could contain operational matters that would be required by the

proposed rule. If the NRC were to impose the DGs on such a licensee, the NRC would have to

address the criteria for avoiding issue finality as described in the applicable issue finality

provision.

Xl mu lative Effects of Regulation.

The NRC engaged extensively with external stakeholders throughout this rulemaking

and related regulatory activities. Public involvement has included: 1) issuance of two ANPRs

and two draft regulatory basis documents that requested stakeholder feedback; 2) issuance of

conceptual and preliminary proposed rule language in support of public meetings; 3) numerous

public meetings with the ACRS; and 4) many more public meetings that supported both the

development of the draft regulatory basis documents as well as development of the

implementing guidance for the two orders that this rulemaking would make generically

applicable (i.e., Orders EA-12-049 and EA-1!2-051). Section II.E of this notice provides a more

detailed discussion of public involvement.

The NRC is following its CER process with regard to the issuance of draft guidance with

the proposed rule to support more informed external stakeholder feedback. The "Availability of

Guidance" section of this document describes how the public can access the draft guidance for

which the NRC seeks external stakeholder feedback.

Finally, the NRC is requesting CER feedback on the following questions:

~c o t/

1. In light of the current or projected CER challenges, do h proposed rule's compliance

dates provide sufficient time to implement the new proposed requirements, including

changes to programs, procedures, and the facility? Specifically, the current proposed

130
rule would require each holder of an operating license or holder of a combined license

for which the Commission made the finding specified in § 52.103(g) to comply with all

provisions of the proposed rule no later than four years following the effective date of the

rule, unless otherwise specified in proposed § 50.155 or proposed 10 CFR part 50,

/This effort is anticipated to equire refueling tages to support i lllementation. T~h/•

NRC •reests feedback o what this time piod should be, reco°nizing that licen e

resorces to implemen /SAMG requirem nts are likely also°i olved with im mentation/,

of the mitigation strategies for beyond-design-basis exterr l events.

2. If current or projected CER challenges .exist, what should be done to address this

situation? For example if more time is required for implementation of the new

requirements, what period of time would be sufficient?

3. Do other NRC regulatory actions, including the post-Fukushima actions and It )

other actions (e.g., generic communications, license amendment requests, inspection

findings of a generic nature), influence the implementation of the proposed rule's

requirements?

4. Are there unintended consequences associated with implementation of these

requirements, including implementing the requirements as a priority over other facility

modifications that are currently being prioritized and scheduled?

5. Please provide feedback on the NRC's supporting regulatory analyses for this

rulmakng.Of note, the regulatory analys•~ estimat• the cost of implementing both f

131
.- •.... ••,,,=•=•u•erln • =,=,o, o ,vO I u,•,,••.The NRC would(

appreciate feedback regarding those estimates.

The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal agencies to write

documents in a clear, concise, and well-organized manner. The NRC has written this document

to be consistent with the Plain Writing Act as well as the Presidential Memorandum, "Plain

Language in Government Writing," published June 10, 1998 (63 FR 31883). The NRC requests

comment on this document with respect to the clarity and effectiveness of the language used.

.XW Environmental Assessment and Proposed Finding of No Significant

Environmental Impact.

The Commission has determined under the National Environmental Policy Act of 1969,

as amended, and the Commission's regulations in subpart A of 10 CFR part 51, that this rule, if

adopted, would not be a major Federal action significantly affecting the quality of the human

environment, and an environmental impact statement is not required. The basis of this

determination reads as follows: The proposed action will not result in any radiological effluent

impact as it will not change any design basis structures, systems, or components that function

to limit the release of radiological effluents during or after an accident. This proposed rule does

not change the standards and requirements for radiological releases and effluents. None of the

revisions or additions in the proposed rule would affect current occupational or public radiation

exposure. The rule would not cause any significant non-radiological impacts, as it would not

affect any historic sites or any non-radiological plant effluents. The NRC concludes that this

132
proposed rule will not cause any significant radiological or non-radiological impacts on the

human environment.

The determination of this environmental assessment is that there will be no significant

effect on the quality of the human environment from this action. Public stakeholders should

note, however, that comments on any aspect of this environmental assessment may be

submitted to the NRC as indicated in the "Addresses" section of this document. The

environmental assessment is available as indicated under the "Availability of Documents"

section.

The NRC has sent a copy of the environmental assessment and this proposed rule to

every State Liaison Officer and has requested comments.

4~rokRdcinAt / 

This proposed rule contains new or amended information collection requirements that

are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq). This rule has

been submitted to the 0MB for approval of the information collection requirements.

Type of submission, new or revision: New

The title of the information collection: Mitigation of Beyond-Design-Basis Events

Proposed Rule

The form number if applicable: Not applicable

How often the collection is required: Once and annually.

Who will be required or asked to report:' Operating nuclear power reactor sites

(comprised of 64 operating sites), as well as the Pressurized Water Reactor Owners Group and

the Boiling Water Reactor Owners Group.

133
An estimate of the number of annual responses: 91

The estimated number of annual respondents: 66 (64 recordkeepers + 2 owners

groups)

An estimate of the total number of hours needed annually to complete the requirement

or request: 121,909

Abstract: In response to the Great East Japan Earthquake of March 11, 2011, the NRC

is seeking to: 1) permanently write into agency rules the requirements already imposed by the

Order EA-12-049 and Order EA-12-051; 2) •,,caLe i•yi•uy reu•,•,. ,k, ,.•,, Acciu"ednr•

-3

emergency preparedness-related industry initiatives into the regulations;)4• provide

requirements for mitigating strategies for new reactor designs; and' address a number of

petitions for rulemaking (PRMs) submitted ,, the March 2011 Fukushima

Dai-ichi event.

The NRC is seeking public comment on the potential impact of the information

collections contained in this proposed rule and on the following issues:

1. Is the proposed information collection necessary for the proper performance of the

functions of the NRC, including whether the information will have practical utility?

2. Is the estimate of burden accurate?

3. Is there a way to enhance the quality, utility, and clarity of the information to be

collected?

4. How can the burden of the .information collection be minimized, including the use of

automated collection techniques?

A copy of the 0MB clearance package may be viewed free of charge at the NRC Public

Document Room, One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville, MD

134
20852. The 0MB ciearance package and rule are available at the NRC worldwide Web site:

http://www.nrc.,qov/Dublic-involve/doc-comment/omb/index.html for 60 days after the signature

date of this notice.

Send comments on any aspect of these proposed regulations related to information

collections, including suggestions for reducing the burden and and on the previously stated

issues, by [INSERT DATE 30 DAYS AFTER PUBLICATION IN THE FEDERAL REGISTER] to

the Records and FOIA/Privacy Services Branch (T-5 F52), U.S. Nuclear Regulatory

Coinmission, Washington, DC 20555-0001, or by electronic mail to

Infocollects.Resource@NRC.gov and to the Desk Officer, Office of Information and Regulatory

Affairs, NEOB-10202 (3150-0011), Office of Management and Budget, Washington, DC 20503.

Comments on the proposed information collections may also be submitted via the Federal

eRulemaking Portal http://www.re~qulations..qov, Docket No. NRC-2014-0240. Comments

received after this date will be considered if it is practical to do so, but assurance of

consideration cannot be given to comments received after this date.

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for

information or an information collection requirement unless the requesting document displays a

currently valid 0MB control number.

XV(,$Crminal Penalties. (

For the purposes of Section 223 of the Atomic Energy Act of 1954, as amended (AEA),

the NRC is issuing this proposed rule that would amend 10 CFR parts 50 and 52 under one or

more of Sections 161 b, 161 i, or 161o of the AEA. Willful violations of the rule would be subject

135
to criminal enforcement. Criminal penalties as they apply to regulations in 10 CFR parts 50 and

52 are discussed in §§ 50.111 and 52.303.

XVy?"oordination with NRC Agreement States.(

The Agreement States are receiving notification of the publication of this proposed rule.

XVIir'Coptbiiyof Agreement State Regulations.,

Under the "Policy Statement on Adequacy and Compatibility of Agreement State

Programs," approved by the Commission on June 20, 1997, and published in the Federal

register (62 FR 46517; September 3, 1997), this rule is classified as compatibility "NRC."

Compatibility is not required for Category "NRC" regulations. The NRC program elements in this

category are those that relate directly to areas of regulation reserved to the NRC by the AEA or

the provisions of Title 10 of the Code of Federal Regulations, and although an Agreement State

may not adopt program elements reserved to the NRC, it may wish to inform its licensees of

certain requirements via a mechanism that is consistent with a particular State's administrative

procedure laws, but does not confer regulatory authority on the State:

• .Y. 'olountary Consensus Standards.

The National Technology Transfer and Advancement Act of 1995, Pub. L. 104-113,

requires that Federal agencies use technical standards that are developed or adopted by

voluntary consensus standards bodies unless the use of such a standard is inconsistent with

applicable law or otherwise impractical. In this proposed rule, the NRC will add requirements for

the mitigation of beyond-design-basis events. This action does not constitute the establishment

of a standard that contains generally applicable requirements.

) ./3i •Public Meeting. (

136
The NRC will conduct a public meeting on this proposed rule for the purpose of

describing the proposed rule to the public and answering questions from the public on the

proposed rule.

The NRC will publish a notice of the location, time, and agenda of the meeting in the

Federal Register, on Regulations.gov, and on the NRC's public meeting Web site within at least

10 calendar days before the meeting. Stakeholders should monitor the NRC's public meeting

Web site for information about the public meeting at:

httpxllwww. nrc..qovlpublic-involve/public-meetinpqslindex. cfm.

XX). Availability of Documents.

The documents identified in the following table are available to interested persons

through one or more of the following methods, as indicated.

ADAMS

ACCESSION NO. I

Document WEB LINK/I

FEDERAL

REGISTER

______________________________________ CITATION

PrimaryRulemakingDocuments __________

Draft regulatory analysis and backfit and issue finality analysis ML15049A212

Draft supplemental regulatory analysis ML1 5069A278

Environmental assessment MLI15049A21 5.

Draft Regulatory Guides_________

DG-1 301, Flexible Mitigation Strategies for Beyond-Design-Basis ML13168A031

Events

DG-1 31 7, Wide-Range Spent Fuel Pool Level Instrumentation ML14245A454

DG-1 319, Integrated Response Capabilities for Beyond-Design- ML14265A070

Basis Events OhrRfrne

ACRS Transcript--r~k•,e ?1'ull Committee, Discuss preliminary -ML14345A387

Mitigation of Beyond-Design-Basis Events Rulemaking Language,

December 4, 2014

ACRS Transcript-Fukushima Subcommittee, Discuss Preliminary -ML1 4337A67 1

Mitigation of Beyond-Design-Basis Events Rulemaking Language,

November 21, 2014

(

137
ACRS Transcript--Full Committee, Discuss Consolidation of Station ML14223A631

Blackout Mitigation Strategies and Onsite Emergency Response

Capabilities Rulemakin s, July 10, 2014 _________

ACRS Transcript--Full Committee, Discuss the Station Blackout ML1 31 75A344

ACRS Transcript--Joint Fukushima and PRA Subcommi e•_7 ML14265A059

Discuss CPRR Technical Analysis, Au ust 22, 2014

ACRS Transcript--Plant Operations and Fire Protection ML13063A403

Subcommittee, Discuss the Onsite Emergency Response

Capabilities Regulatory Basis, February 6, 2013

ACRS Transcript--Reactor Safeguards Reliability and PRA ML14337A651

Subcommittee, Discuss CPRR Technical Analysis, November 19,

2014

ACRS Transcript--Regulatory Policies and Practices Subcommittee, ML13148A404

Discuss the Station Blackout Mitigation Strategies regulatory basis,

December 5, 2013 and April 23, 2013

American National Standards Institute/American Nuclear Society 3.2- httpj:llwww.ans.orglst

2012, "Administrative Controls and Quality Assurance for the ore/

Operational Phase of Nuclear Power Plants"

CL1-12-09,. South Carolina Electric & Gas Co. and South Carolina ML12090A531

Public Service Authority (Also Referred to as Santee Cooper)

COMGBJ-1 1-0002, "NRC Actions Following the Events in Japan," MLI110800456

March, 21, 2011

COMSECY-13-0002, "Consolidation of Japan Lessons Learned ML1 301 1A037

Near-Term Task Force Recommendations 4 and 7 Regulatory

Activities," January 25, 2013 _________

COMSECY-13-0010, "Schedule and Plans for Tier 2 Order on ML12339A262

Emergency Preparedness for Japan Lessons Learned," dated March

27, 2013

COMSECY-14-0037, "Integration of Mitigating Strategies for Beyond- ML14309A256

Design-Basis External Events and The Reevaluation of Flooding

Hazards," November 21, 2014

Conceptual Consolidated Preliminary Proposed Rule Language for ML14052A057

NTTF Recommendations 4, 7, 8 and 9, February 21, 2014.

Containment Performance and Release Reduction draft regulatory ML15022A214

basis

Crystal River Unit 3, "NRC Response to Duke Energy's Final ML13325A847

Response to The March 2012 Request for Information Letter,"

January, 22, 2014

Crystal River Unit 3, "Rescission of Order EA-12-049, 'Order ML13212A366

Modifying Licenses with Regard to Requirements for Mitigation

Strategies for Beyond Design Basis External Events,"'" August 27,

2013_________ _

k

138
Crystal River Unit 3, Final Response to March 12, 2012 Information ML13274A341

Crystal River Unit 3, "Rescission Of Order EA-12-051, 'Order ML13203A161

Modifying Licenses With Regard To Reliable Spent Fuel Pool

Instrumentation,"i August 27, 2013 _________

Federal Register Notice--Enhancements to Emergency 76 FR 72560

Preparedness Regulations, Final Rule, November 23, 2011

Federal Register Notice--Onsite Emergency Response Capabilities, 78 FR 63901

Regulatory Basis, October 25, 2013

Federal Register Notice--Onsite Emergency Response 77FR 23161

Capabilities, Advanced Notice of Proposed Rulemaking, April 18,

2012

Federal Register Notice--Onsite Emergency ResponseCapabilities, 78 FR 1154

Draft Regulatory Basis, January 8, 2013

Federal Register Notice--Onsite Emergency Response 78 FR 68774

Capabilities, Preliminary Proposed Rule Language, November 15,

2013

Federal Register Notice--Policy Statement on the Regulation of 73 FR 60612

Advanced Reactors, Final Policy Statement, October 14, 2008 _________

Federal Register Notice--Power Reactor Security Requirements, 74 FR 13926

Final Rule, March 27, 2009

Federal Register Notice--PRM-50-1 00, Petition for Rulemaking 78 FR 44034

Submitted by the Natural Resources Defense Council, Inc., July 23,

2013

Federal Register Notice--PRM-50-101, Petition for Rulemaking 77 FR 16483

Submitted by the Natural Resources Defense Council, Inc., March

21, 2012 _ _ _ _ _ _ _ _

Federal Register Notice--PRM-50-1 02, Petition for Rulemaking; 77 FR 25104

Submitted by the Natural Resources Defense Council, Inc., April 27,

2012

Federal Register Notice--PRM-50-96, Long-Term Cooling and 77 FR 74788

Unattended Water Makeup of Spent Fuel Pools, Consideration in

the Rulemaking Process, December 18, 2012

Federal Register Notice--PRM-50-97, PRM-50-98, 76 FR 58165

PRM-50-99, PRM-50-1 00, PRM-50-1 01, PRM-50-1 02, Petitions for

Rulemaking Submitted by the Natural Resources Defense Council,

Inc., Notice of Receipt, September 20, 2011

Federal Register Notice--Statement of Principles and Policy for the 62 FR 46517

Agreement State Program; Policy Statement on Adequacy and

Compatibility of Agreement State Programs, Final Policy Statements,

September 3, 1997

Federal Register Notice--Station Blackout Mitigation Strategies, 78 FR 21275

Draft Regulatory Basis and Draft

Rule Concepts, April 10, 2013

Federal Register Notice--Station Blackout Mitigation Strategies, 78 FR 44035

Regulator Basis, July 23, 2013

139
Federal Register Notice--Station Blackout, Advanced Notice of 77 FR 1 6175

Proposed Rulemaking, March 20, 2012 _________

Interim Staff Guidance, NSIR/DPR-ISG-01, "Emergency Planning for ML113010523

Nuclear Power Plants," November 2011__________

JLD-ISG-2012-01, "Compliance with Order EA-12-049, Order ML12229A166

Modifying Licenses with Regard to Requirements for Mitigation

Strategies for Beyond-Design-Basis External Events," Revision 0,

August 29, 2012

Inspection Manual Chapter (IMC) 0308, "Reactor Oversight Process ML062890421

Basis Document," Attachment 2, "Technical Basis for Inspection

Program," October, 16, 2006

Kewaunee Power Station, 60-Day Response to March 12, 2012 ML13123A004

Information Request Regarding Recommendation 2.1. Seismic

Reevaluations, April 29, 2013

Kewaunee Power Station, Rescission of Order EA-12-049, "Order ML14059A411I

Modifying Licenses with Regard to Requirements for Mitigation

Strategies for Beyond Design Basis External Events," June 10, 2014

Kewaunee Power Station, Response to Request for Relief from ML1 3322B255

Responding Further to the March 2012 Request for Information

Letter for Recommendation 9.3., January 22, 2014

Letter from ACRS to Chairman Jaczko, "Initial ACRS Review of: (1) ML11284A136

The NRC Near-Term Task Force Report on Fukushima and (2)

Staff's Recommended Actions to be Taken Without Delay," October

13, 2011

Letter from ACRS to Mr. R. W. Borchardt, "Response To February ML1 2072A1 97

27, 2012 Letter Regarding Final Disposition Of Fukushima-Related

ACRS Recommendations In Letters Dated October 13, 2011 And

November 8, 2011," March 13, 2012 ________

Letter from R.W. Borchardt to J. Sam Amijo, Chairman ACRS, "Final ML1 2030A1 98

Disposition Of The Advisory Committee On Reactor Safeguards'

Review Of (1) The U.S. Nuclear Regultaory Commission Near-Term

Task Force Report On Fukushima, (2) Staff's Recommended Actions

To Be Taken Without Delay (SECY-1 1-0124), And (3) Staff's

Prioritization Of Recommended Actions To Be Taken In Response

To Fukushima Lessons-Learned" February 27, 2012

NEI 06-12, "B.5.b Phase 2&3 Submittal Guideline," Revision 2, ML070090060

December 2006

NEI 10-05, "Assessment of On-Shift Emergency Response ML1 11751698

Organization Staffing and Capabilities," Revision 0, June 2011

NEI 12-01, "Guideline for Assessing Beyond Design Basis Accident ML12125A412

Response Staffing and Communications Capabilities," Revision 0,

May 2012

NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) ML15027A542

Implementation Guide," Revision 1 Draft C, January 2015

NEI 13-06, "Enhancements to Emergency Response Capabilities for ML14269A230

Beyond Design Basis Accidents and Events," Revision 0, September

2014

140
NEI 14-01, "Emergency Response Procedures and Guidelines for ML14269A236

Beyond Design Basis Events and Severe Accidents," Revision 0,

September 2014

NEI 91-04 (formerly NUMARC 91-04), Severe Accident Issue ML072850981

Closure Guidelines, Revision 1, December 1994

Non-concurrence NCP-201 5-003 ML1 5091A646

NUREG-0654/FEMA-REP-1, "Criteria for Preparation and Evaluation ML040420012

of Radiological Emergency Response Plans and Preparedness in

Support of Nuclear Power Plants," Revision 1, November 1980

NUREG-0660, Volume1 and 2, "NRC Action Plan Developed as a ML072470526 and

Result of the TMI-2 Accident," May 1980 ML072470524

NUREG-0711, "Human Factors Engineering Program Review ML12324A013

Model," Revision 3, November 2012

N UREG-0737, "Clarification of TM I Action Plan Requirements," ML1 02560051

November 1980

NUREG-0737, "Clarification of TM I Action Plan Requirements," MLI102560009

Supplement 1, November 1980

NUREG-1 935, "State-of-the-Art Reactor Consequence Analyses ML12332A057

(SOARCA) Report," November 2012

Omaha Public Power District's Overall Integrated Plan (Redacted) in ML1 311 6A208

Response to March 12, 2012, Order EA-12-049, February 28, 2013 ________

Order EA-02-026, "Order for Interim Safeguards and Security ML02051 0635

Compensatory Measures," February 25, 2002

Order EA-1 2-049, "Issuance of Order to Modify Licenses With ML1 2054A735

Regard to Requirements for Mitigation Strategies for Beyond-DesignBasis External Events," (Mitigating Strategies Order), March 12,

2012

Order EA-12-051, "Order Modifying Licenses with Regard to Reliable ML12056A044

Spent Fuel Pool Instrumentation,"

Preliminary proposed rule language for Mitigation of Beyond-Design- ML14336A641

Basis Events rulemaking made available to the public on November

13, 2014 and December 8, 2014 to support public discussion with

the ACRS.

Preliminary proposed rule language for Mitigation of Beyond-Design- ML14218A253

Basis Events Rulemaking, August 15, 2014.

PRM 50-1 02, "NRDC's Petition For Rulemaking to Require More ML1 1216A242

Realistic Training on Severe Accident Mitigation Guidelines," July 26,

2011__ _ _ _ _ _ _ _

PRM 50-97, "NRDC's Petition For Rulemaking to Require ML1 121 6A237

Emergency Preparedness Enhancements for Prolonged Station

Blackouts," July 26, 2011

PRM-50-100, "NRDC's Petition For Rulemaking to Require ML11216A240

Licensees to Improve Spent Nuclear Fuel Pool Safety,"i July 26, 2014 _________

PRM-50-101, "NRDC's Petition For Rulemaking to Revise 10 C.F.R. ML11216A241

§50.63," July 26, 2011.__________

141
PRM-50-96, "Petition for Rulemaking Submitted by Thomas Popik on MLI110750145

Behalf of the Foundation for Resilient Societies to adopt regulations

that would require facilities licensed by the NRC under 10 CFR Part

50 to assure long-term cooling and unattended water makeup of

spent fuel pools," March 14, 2011

PRM-50-98, "NRDC's Petition For Rulemaking to Require ML1 121 6A238

Emergency Preparedness Enhancements for Multiunit Events," July

26, 2011

Regulatory Issue Summary 2009-1 3, "Emergency Response Data ML092670124

System Upgrade from Modem to Virtual Private Network Appliance,"

September 28, 2009

Request for Information Pursuant to Title 10 of the Code of Federal ML12053A340

Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3,

of the Near-Term Task Force Review of Insights From the

Fukushima Dai-lchi Accident, March 12, 2012

Severe Accident Management Guidance Technical Basis Report, http:llwww.epri.com/a

Volume 1: Candidate High- bstracts/Pa~qes/Prod u

Level Actions and Their Effects. EPRI, Palo Alto, CA: 2012. ctAbstract.aspx?Prod

1025295. uctld= 1025295

Severe Accident Management Guidance Technical Basis Report,

Volume 2: The Physics of

Accident Progression. EPRI, Palo Alto, CA: 2012. 1025295.

San Onofre Nuclear Generating Station Units 2 and 3, "Rescission ML14I113A572

of Order EA-12-049, 'Order Modifying Licenses with Regard to

Requirements for Mitigation Strategies for Beyond Design Basis

External Events,"'" June 30, 2014

San Onofre Nuclear Generating Station Units 2 and 3, "NRC MLI13329A826

Response To Southern California Edison's Final Response to the

March 2012 Request for Information Letter," January 22, 2014 _________

San Onofre Nuclear Generating Station Units 2 and 3, Final ML13276A020

Response to the March 12, 2012 Information Request Regarding

Near-Term Task Force Recommendations 2.1, 2.3, and 9.3 and

Corresponding Commitments San Onofre Nuclear Generating

Station (SONGS) Units 2 and 3, September 30, 2013 _________

San Onofre Nuclear Generating Station Units 2 and 3, "Rescission of ML14I111A069

Order EA-1 2-051, 'Order Modifying Licenses with Regard to Reliable

Spent Fuel Pool Instrumentation,'" June 30, 2014

SECY-1 1-0093, "Near-Term Report and Recommendations for MLI11186A950

Agency Actions Following the Events in Japan," July 12, 2011__________

SECY-1 1-0124, "Recommended Actions to be Taken Without Delay ML1 1245A127

from the Near-Term Task Force Report," September 9, 2011

SECY-1 1-0137, "Prioritization of Recommended Actions to Be Taken ML1 1272A1 11

in Response to Fukushima Lessons Learned," October 3, 2011

SECY-12-0025, "Proposed Orders and Requests for Information in ML12039A103

Response to Lessons Learned From Japan's March 11, 2011, Great

T~hoku Earthquake and Tsunami," February 17, 2012

142
SECY-13-0132, "Plan for Updating the U.S. Nuclear Regulatory ML13274A495

Commission's Cost Benefit Guidance," January 2, 2014

SECY-14-0046, "Fifth 6-Month Status Update on Response to ML14064A523

Lessons Learned From Japan's March 11, 2011i, Great Tohoku

Earthquake and Subsequent Tsunami," April 17, 2014

SECY-89-012, "Staff Plans for Accident Management Regulatory ML12251A414

and Research Programs," Januar 18, 1989

SECY-97-1 32, "Status of the Integration Plan for Closure of Severe ML992930144

Accident Issues and the Status of Severe Accident Research," June

23, 1997

SECY-98-1 31, "Status of the Integration Plan for Closure of Severe ML992880008

Accident Issues and the Status of Severe Accident Research," June

8, 1998

COMSECY-14-0037, "Integration of Mitigating Strategies for Beyond- ML14309A256

Design-Basis External Events and The Reevaluation of Flooding

Hazards"

SRM-COMSECY-1 4-0037, "Integration of Mitigating Strategies for ML1 5089A236

Beyond-Design-Basis External Events and The Reevaluation of

Flooding Hazards"

SRM-COMSECY-13-0002, "Consolidation of Japan Lessons ML13063A548

Learned Near-Term Task Force Recommendations 4 and 7

Regulatory Activities"

SRM-SECY-1 1-0093, "Near-Term Report and Recommendations for MLIl12310021

Agency Actions Following the Events in Japan," August 19, 2011

SRM-SECY-1 1-0137, "Prioritization of Recommended Actions to Be ML1 13490055

Taken in Response to Fukushima Lessons Learned," December 15,

2011

SRM-SECY-1 3-01 32, "U.S. Nuclear Regulatory Commission Staff ML14139A104

Recommendation for the Disposition of Recommendation 1 of the

Near-Term Task Force Report," May 19, 2014

SRM-SECY-201 1-0124, "Recommended Actions to be Taken MLI112911571

Without Delay From the Near-Term Task Force Report," October 18,

2011

SRM-SECY-90-016, "Evolutionary Light Water Reactor Certification ML003707885

Issues and Their Relationship to Current Regulatory Requirements,"

June 26, 1990__________

-Temporary Instruction 2515/191, "Inspection of the Licensee's ML1 4273A444

Responses to Mitigation Strategies Order EA-12-049, Spent Fuel

Pool Instrumentation Order EA-12-051 and Emergency

Preparedness Information Requested in NRC March 12, 2012,"

March 12, 2012

Temporary Instruction 2515/184, "Availability and Readiness MLI11115A053

Inspection of Severe Accident.Management Guidelines (SAMGs)"

April 29, 2011

Vermont Yankee Nuclear Power Station, "Rescission of Order EA- ML14321A685

12-049, 'Order Modifying Licenses with Regard to Requirements for

Mitigation Strategies for Beyond Design Basis External Events,"'"

March 2, 2015

143
Vermont Yankee Nuclear Power Station, "Rescission of Order EA- ML14321A696 12-051, 'Order Modifying Licenses with Regard to Reliable Spent

Fuel Pool Instrumentation,'" March 2, 2015

Throughout the development of this rule, the NRC may post documents related to this

rule, including public comments, on the Federal rulemaking Web site at

http://www.repqulations..qov under Docket ID NRC-2014-0240. The Federal rulemaking Web site

allows you to receive alerts when changes or additions occur in a docket folder. To subscribe:

1) Navigate to the docket folder NRC-2014-0240); 2) click the."Sign up for E-mail Alerts" link;

and 3) enter your e-mail address and select how frequently you would like to receive e-mails

(daily, weekly, or monthly).

List of Subjects

10 CFR Part 50

Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental

relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria,

Reporting and recordkeeping requirements.

10 CFR Part 52

Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early

site permit, Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power

plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of

site, Reporting and recordkeeping requirements, Standard design, Standard design certification,

•Incorporation by reference.

144
For the reasons set out in the preamble and under the authority of the Atomic Energy

Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and.5 U.S.C.

552 and 553, the NRC is proposing to adopt the following amendments to 10 CFR parts 50 and

52.

PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES

1. The authority citation for 10 CFR part 50 continues to read as follows:

Authority: Atomic Energy Act secs. 102, 103, 104, 105, 147, 149, 161, 181, 182, 183,

186, 189, 223, 234 (42 U.S.C. 2132, 2133, 2134, 2135, 2167, 2169, 2201, 2231, 2232, 2233,

2236, 2239, 2273, 2282); Energy Reorganization Act secs. 201, 202, 206 (42 U.S.C. 5841,

5842, 5846); Nuclear Waste Policy Act sec. 306 (42 U.S.C. 10226); Government Paperwork

Elimination Act sec. 1704 (44 U.S.C. 3504 note); Energy Policy Act of 2005, Pub. L. No. 109-58,

119 Stat. 194 (2005). Section 50.7 also issued under Pub. L. 95-601, sec. 10, as amended by

Pub. L. 102-486, sec. 2902 (42 U.S.C. 5851). Section 50.10 also issued under Atomic Energy

Act secs. 101, 185 (42 U.S.C. 2131, 2235); National Environmental Protection Act sec. 102 (42

U.S.C. 4332). Sections 50.13, 50.54(d), 50.54(dd), and 50.103 also issued under Atomic Energy

Act sec. 108 (42 U.S.C. 2138).

Sections 50.23, 50.35, 50.55, and 50.56 also issued under Atomic Energy Act sec. 185

(42 U.S.C. 2235). Appendix Q also issued under National Environmental Protection Act sec.

102 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 20)4 (42 U.S.C. 5844).

Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-415 (42 U.S.C. 2239). Section

50.78 also issued under Atomic Energy Act sec. 122 (42 U.S.C. 2152). Sections 50.80- 50.81

also issued under Atomic Energy Act sec. 184 (42 U.S.C. 2234).

145
2. In § 50.8, paragraph (b) is revised to read as follows:

§ 50.8 Information collection requirements: 0MB approval.

(b) The approved information collection requirements contained in this part appear in §§50.30,

50.33, 50.34, 50.34a, 50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54,

50.55, 50.55a, 50.59, 50.60, 50.61, 50.61 a, 50.62, 50.63, 50.64, 50.65, 50.66, 50.68, 50.69,

50.70, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, 50.150, 50.155, and

appendices A, B, E, G, H, I, J, K, M, N, 0, Q, R, and S to this part.

3. In § 50.34:

a. revise paragraphs (a)(13) and add paragraph (a)(14)i.

b. revise paragraph (b)(12) and add paragraph (b)(13); and

c. revise paragraph (i).

The revisions and additions read as follows:

§ 50.34 Contents of applications; technical information.

(a) * * *

(13) On or after July 13, 2009, power reactor applicants who apply for a construction

permit shall submit the information required by 10 CFR 50.150(b) as a part of their preliminary

safety analysis report.

• (1-4) On, c, gft, [EFFECTIVE= DATE],_, p•; ievdur applicants applying for a

-•c•-n n;.;mt •t L~tc ."q.•--•bll apeiminary e description of the design e~a ure ,

r.u•nt docign. unrder § 50. 155(d), :rnd :n oxple~nat",n c; ,hG~ t,,6t= udt;u, ,atu;ros co•mply ..with• •

146
(b) * * *

(12) On or after July 13, 2009, power reactor applicants who apply for an operating

license which is subject to 10 CFR 50.150(a) shall submit the information required by 10 CFR

50.150(b) as a part of their final safety analysis report.

•prti~ "ir must~b11 ; ,d r;,ld d~u, iptiuri ur he desiyi, rkutuur= iii.Judwd ;i, 1.;, pId-,iL"--

* * * * *

(i) Mitigation of beyond-design-basis events.

Each application for a power reactor operating license under this part must include the

applicant's plans for implementing the requirements of § 50.155 and 10 CFR part 50, Appendix

E, Section VII, including a schedule for achieving full compliance with these requirements. The

application must also include a description of:

(1) the integrated response capability required by § 50.155(b);

(2) the equipment upon which~the strategies and guidelines required by § 50.155(b)(1)

rely, including the planned locations of the equipment and how the equipment and SSCs meet

the design requirements of § 50.155(c); and

(3) the strategies and guidelines required by § 50.155(b)(2).

• * * * *

4. In § 50.54:

a. remove paragraph (hh)(2);

147
b.

C.

redesignate paragraph (hh)(3) as (hh)(2); and

revise paragraph (hh)(2) to read as follows:

§ 50.54 Conditions of licenses.

*k "* *r "* *

(hh) * * * "

(2) This section does not apply to a licensee that has submitted the certifications

required under § 50.82(a)(1) or § 52.110(a) of this chapter once the NRC has docketed those

certifications.

5. Add § 50.155 to read as follows:

§ 50.155 Mitigation of Beyond-Design-Basis Events

(a) Applicability. (

(1) Each holder of an operating license for a nuclear power reactor under this part and

each holder of a combined license under part 52 of this chapter after the Commission has made

the finding under § 52.103(g), before the NRC'sdocketing of the license holder's certifications

described in section § 50.82(a)(1) or § 52.110(a) of this chapter, shall comply with the

requirements of this section and section VII of appendix E to 10 CER part 50.

(2) Each applicant for an operating license for a nuclear power reactor under this part

and each holder of a combined license under part 52 of this chapter before the Commission has

made the finding under § 52.103(g) shall comply with the requirements of this section and

section VII of appendix E to 10 CFR part 50 no later than the date on which the Commission

issues the operating license under § 50.57 or makes the finding under § 52.103(g), respectively.

148
(3) When the NRC has docketed the certifications described in § 50.82(a)(1) or

§ 52.110(a) of this chapter, submitted by a licensee subject to the requirements of this section

and section VII of appendix E to 10 CFR part 50, then that licensee shall comply with the

requirements of §§ 50.155(b)-• associated with maintaining or restoring secondary

containment capabilities, if applicable, and spent fuel pool cooling capabilities, but not with

§ 50.155(c)(4) and section VII of appendix E to 10 CFR part 50, for the unit described in the

§ 50.82(a)(1) or § 52.110(a) certifications until the spent fuel pool(s) is empty of all irradiated

fuel.

(i) Holders of operating licenses or combined licenses for which the NRC has docketed

the certifications described in § 50.82(a)(1) or § 52.110(a) of this chapter need not meet the

requirements of this section except for paragraph (b)(2) if the licensee performs and retains an

analysis demonstrating that the decay heat of the fuel in the spent fuel pool is removed solely by

heating and boiling of water within thie spent fuel pool and the boil-off period provides sufficient

time for the licensee to obtain off-site resources to sustain the spent fuel pool cooling function

indefinitely.

(ii) Dominion Nuclear Connecticut, Inc. (Millstone Power Station Unit 1) is not subject to

the requirements of this section. K -" (4) The require nt in paragra (d) of this section pply, to applicants for:

(i) Cons ction permits for rclear power reactor.8 under this part;

Opeating_ licenses rnclear power re or under this part fo hch a construction

,,lrmit was issued af1e FFECTIVE DATE]; •

" (iii) St a~n rd design certifications nder part 52 of this ch,/ tter;

(iv tandard design approvs under part 52 of tisapter;-

,/(v) Combined licensesnder part 52 of thiscpter that do not referce a standard

design certification, sta r einapoamnfcturirng license Zand

• -149

(

p_-

\
(b) Integrated response capability. Each applicant or licensee shall develop, implement,

and maintain an integrated response capability that includes:

(1) Mitigation Strategies for Beyond-Design-Basis External Events.

Strategies and guidelines to mitigate beyond-design-basis external events from natural

phenomena that result in an extended loss of all ac power concurrent with either a loss of

normal access to the ultimate heat sink or, for passive reactor designs, a loss of normal access

to the normal heat sink. These strategies and guidelines must be capable of being implemented

site-wide and must include:

(i) Maintaining or restoring core cooling, containment, and spent fuel pool cooling

capabilities; and

(ii) The acquisition and use of offsite assistance and resources to support the functions

required by paragraph (b)(1)(i) of this section indefinitely, or until sufficient site functional

capabilities can be maintained without the need for the mitigation strategies.

(2) Extensive Damage Mitigation Guidelines (EDMGs).

Strategies and guidelines to maintain or restore core cooling, containment, and spent

fuel pool cooling capabilities under the circumstances associated with loss of large areas of the

plant due to explosions or fire, to include strategies and guidelines in the following areas:

(i) Firefighting;

(ii) Operations to mitigate fuel damage; and

(iii) Actions to minimize radiological release.

-(=3) Cv•Adn fV~d1Iayhrmeri.Guid~iine~s.• /

150
actios, nded0to: //

(iii) Mir} ize radiologica Peleases.

3• (4) Integration of strategies and guidelines in paragraphs (b)(1) - (b)(3) of this section

with the Emergency Operating Procedures (EOPs).

L.( (• Sufficient staffing to support implementation of the strategies and guidelines in

paragraphs (b)(1) - (b)(3) of this section in conjunction with the EOPs to respond to events.

S A supporting organizational structure with defined roles, responsibilities, and

authorities for directing and performing the strategies and guidelines in paragraphs (b)(1) -

(b)(3) of this section.

(c) Equipment.

(1) The capacity and capability of the equipment relied on for the mitigation strategies

required by paragraph (b)(1) of this section must be sufficient to simultaneously maintain or

restore core cooling, containment, and spent fuel pool cooling capabilities for all the power

reactor units within the site boundary.

(2) The equipment relied on for the mitigation strategies required by paragraph (b)(1) of

this section must be reasonably protected from the effects of natural phenomena that are

equivalent to the design basis of the facility.

(i) By [DATE 730 DAYS AFTER THE EFFECTIVE DATE OF THE RULE], each

licensee that received the March 12, 2012, NRC letter issued under § 50.54(f) concerning

reevaluations of seismic and flooding hazard levels, shall provide reasonable protection against

that reevaluated seismic or flooding hazard(s) if it exceeds the design basis of its facility.

151

i
(3) The equipment relied on for the mitigation strategies in paragraph (b)(1) of this

section must receive adequate maintenance such that the equipment is capable of fulfilling its

intended function.

(4) The equipment relied on for the mitigation strategies in paragraph (b)(1) of this

section must include reliable means to remotely monitor wide-range spent fuel pool levels to

support effective prioritization of event mitigation and recovery actions.

s icient to enhae coping duratio and mini e reliance on hi an actions to i•ntain or ,

for passive reacto dsigns, aos ofnra cest

d (1•Training requirements,.

Each licensee shall provide for the training and qualification of personnel that perform

activities in accordance with the strategies and guidelines identified in paragraphs (b)(1)-(b)(•

of this section. The training and qualification on these activities must be developed using the

systems approach to training as defined in § 55.4 except for elements already covered under

other NRC regulations.

a. Drills and Exercises.

(1) An applicant for an operating license issued under this part shall conduct an initial(

drill or exercise that demonstrates the capability to transition to and use one or more of the

strategies and guidelines in paragraphs (b)(1)-(b)48) of this section and use the communications [!

equipment required in 10 CER part 50, appendix E, section VII no more than 12 months before

issuance of an operating license for the unit described in the license application.

152
(2) A holder of a combined license issued under 10 CFR part 52 before the Commission

has made the finding under § 52.103(g), shall conduct an initial drill or exercise that

demonstrates the capability to transition to and use one or more of the strategies and guidelines

in paragraphs (b)(1)-(b)(,8) of this section and use the communications equipment required in (

10 CFR part 50, appendix E, section VII no more than 12 months before the date specified for

completion of the last inspections, tests, and analyses in the ITAAC completion schedule

required by § 52.99(a) for the unit described in the combined license.

(3) once the Commission issues an operating license to an entity described in eI

paragraph (t•(1) of this section or makes the finding under § 52.103(g) for an entity described in

paragraph ($(2) of this section, the licensee shall conduct subsequent drills, exercises, or both

that collectively demonstrate a capability to use at least one of the strategies and guidelines in

each of paragraphs (b)(1)-(b)(Z) of this section in succeeding 8-year intervals. The drills and

exercises performed to demonstrate this capability must include transitions from other (

procedures and guidelines as applicable, and the use of communications equipment required in

10 CFR part 50, appendix E, section VII. Each licensee shall not exceed 8 years between any

consecutive drills or exercises.

(4) A holder of an operating license issued under this part or a combined license Under

10 CFR part 52 for which the Commission has made the finding specified in § 52.103(g) as of

[EFFECTIVE DATE], shall conduct an initial drill or exercise that demonstrates the capability to

transition to and use one or more of the strategies and guidelines in paragraphs (b)(1)-(b)(• 'of (

this section and use communications equipment required in 10 CER part 50, appendix E,

section VII by [DATE 4 YEARS AFTER EFFECTIVE DATE]. Following this initial drill or

exercise, the licensee shall conduct subsequent drills, exercises, or both that collectively

demonstrate a capability to use at least one of the strategies and guidelines in each of

paragraphs (b)(1)-(b)(• of this section in succeeding 8-year intervals. The drills and exercises

153
performed to demonstrate this capability must include transitions from other procedures and

guidelines as applicable, and the use of communications equipment required in 10 CFR part 50,

appendix E, section VII. Each licensee shall not exceed 8 years between any consecutive drills

or exercises.

.. ' (• Change Control.

(1) A licensee may make changes in the implementation of the requirements in this

section and 10 CFR part 50, appendix E, section VII without NRC approval, provided that before

imPlementing each such change, the licensee performs an evaluation demonstrating that the

provisions of this section and 10 CER part 50, appendix E, section VII continue to be met.

(2) Documentation of all changes, including the evaluation required by paragraph (0'(1)

of this section, shall be maintained until the requirements of this section and section VII of

appendix E to 10 CFR part 50 no longer apply.

(3) Changes in the implementation of requirements in this chapter subject to change

control processes other than paragraph"•of this section and resulting from changes in the

implementation of the requirements in this section and 10 CFR part 50, appendix E, section VII

must be processed via their respective change control processes.

secti,,ce n seeka ;ruof t~u, 1 --e•r r.•rtcr wh, i' •uj• ,prag raph d) o thi•--

•p ud•, ~ J.•, •,,=;-, () u hi -cion, c ten te licensee snall\

--doaigF ,f•tur1 ¢,3,.tifl•c3 t3ca..tisf,'; t~ho dzci"rg,• ob~v dc;,.,d in a.r,"g.rap~h (d) of this c .

-• Imolement ation. (

Unless otherwise specified in this section or 10 CFR part 50, appendix E, section VII:

154
(1) Each holder of an operating license under this part on [EFFECTIVE DATE OF THE

FINAL RULE] shall comply with all the provisions of this section no later than four years

following [EFFECTIVE DATE OF THE FINAL RULE.].

(2) Each holder of a combined license under 10 CFR part 52 for which the Commission

made the finding specified in § 52.103(g) as of [EFFECTIVE DATE OF THE FINAL RULE] shall

comply with all the provisions of this section no later than four years following [EFFECTIVE

DATE OF THE FINAL RULE].

6. In Appendix E to part 50:

a. revise paragraph 1.2;

b. revise paragraphs IV.B.1, IV.E.2, and IV.F.2.j;

c. revise paragraph Vl.3.c; and

d. add Section VII.

The revisions and additions read as follows:

Appendix E to Part 50--Emergency Planning and Preparedness for Production and

Utilization Facilities

2. This appendix establishes minimum requirements for emergency plans for use in

attaining an acceptable state of emergency preparedness. These plans shall be described

generally in the preliminary safety analysis report for a construction permit and submitted as

part of the final safety analysis report for an operating license. These plans, or major features

thereof, may be submitted as part of the site safety analysis report for an early site permit.

155
Section VII of this appendix also provides for "Communications and Staffing Requirements for

the Mitigation of Beyond-Design-Basis Events" that do not need to be contained within a

licensee's emergency plan.

IV. * * *

B. * * *

1. The means to be used for determining the magnitude of, and for continually assessing

the impact of, the release of radioactive materials, including from all reactor core and spent fuel

pool sources, shall be described, including emergency action levels that are to be used as

criteria for determining the need for notification and participation of local and State agencies, the

Commission, and other Federal agencies, and the emergency action levels that are to be used

for determining when and what type of protective measures should be considered within and

outside the site boundary to protect health and safety. The emergency action levels shall be

based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring. By

June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile

action that may adversely affect the nuclear power plant. The ihitial emergency action levels

shall be discussed and agreed on by the applicant or licensee and state and local governmental

authorities, and approved by the NRC. Thereafter, emergency action levels shall be reviewed

with the State and local governmental authorities on an annual basis.

S. * * *

156
2. Equipment for determining the magnitude of and for continuously assessing the

impact of the release of radioactive materials, including from all reactor core and spent fuel pool

sources, to the environment;

j. The exercises conducted under paragraph 2 of this section by nuclear power reactor

licensees must provide the opportunity for the ERO to demonstrate proficiency in the key skills

necessary to implement the principal functional areas of emergency response identified in

paragraph 2.b of this section. Each exercise must provide the opportunity for the ERO to

demonstrate key skills specific to emergency response duties in the control room, T-SC, OSC,•

EOF, and joint information center. Additionally, in each eight calendar year exercise cycle,

nuclear power reactor licensees shall vary the content of scenarios during exercises conducted

under paragraph 2 of this section to provide the opportunity for the ERO to demonstrate

proficiency in the key skills necessary to respond to the following scenario elements: hostile

action directed at the plant site, no radiological release or an unplanned minimal radiological

release that does not require public protective actions, an initial classification of or rapid

escalation to a Site Area Emergency or General Emergency, and integration of offsite resources

with onsite response. The licensee shall maintain a record of exercises conducted during each

eight year exercise cycle that documents the content of scenarios used to comply with the

requirements of this paragraph. Each licensee shall conduct a hostile action exercise for each of

its sites no later than December 31, 2015. The first eight-year exercise cycle for a site will begin

in the calendar year in which the first hostile action exercise is conducted. For a site licensed

157
under Part 52, the first eight-year exercise cycle begins in the calendar year of the initial

exercise required by Section IV.F.2.a.

VI. * * *

c. In the event of a failure of NRC-supplied equipment, a replacement will be furnished

by the NRC for licensee installation.

Section VII. Communications and Staffing Requirements for the Mitigation of Beyond Design

Basis Events

All changes associated with implementation of the requirements in this section are

subject to § 50.155•. The change control provisions of § 50.54(q) do not apply to proposed (

changes associated with implementation of the requirements in this section, unless the

requirements in this section are implemented within the licensee's emergency plan.

1. Each nuclear power reactor applicant or licensee shall perform a detailed analysis

demonstrating that sufficient staff is available to implement the guidelines and strategies to

respond to a beyond design basis external event resulting in impeded access to the nuclear

power plant, an extended loss of ac power sources concurrent with either a loss of normal

access t~i mate heat sink or, for passive reactor designs, a loss of normal access to the(

normal heat sink, and affecting all units on-site.

158
a. An applicant for a power reactor operating license under this part shall perform this

analysis at least 2 years before the issuance of the first operating license for full power (one

authorizing operation above 5 percent of rated thermal power).

b. A holder of a combined license issued under 10 CFR part 52 before the Commission

has made the finding under § 52.103(g) shall perform this analysis and submit it to the NRC

under § 52.3 at least two years before the date specified for completion of the last inspections,

tests, and analyses in the ITAAC completion schedule required by § 52.99(a) for the plant.

c. Each holder of a power reactor operating license or combined license for which the

Commission has made the finding specified in § 52.103(g) as of [EFFECTIVE DATE] shall,

before the NRC's docketing of the license holder's certifications described in section

§ 50.82(a)(1) or § 52.110(a) of this chapter, perform this analysis and submit it to the NRC

under § 50.4 no later than [DATE 365 DAYS AFTER EFFECTIVE DATE].

2. Each nuclear power reactor applicant or Jicensee shall make and describe adequate

provisions for at least one onsite and one offsite communications system capable of remaining

functional during an extended loss of alternating current power including the effects of the loss

of the local communications infrastructure.

a. An applicant for a power reactor operating license under this part shall make these

provisions no later than the issuance of the first operating license for full power (one authorizing

operation above 5 percent of rated thermal power).

b. A holder of a combined license issued under 10 CFR part 52 before the Commission

has made the finding under § 52.103(g) shall make these provisions no later than the date

specified for completion of the last inspections, tests, and analyses in the ITAAC completion

schedule required by § 52.99(a) for the plant.

c. Each holder of a power reactor operating license under this part or a combined

license issued under 10 CFR part 52 for which the Commission has made the finding specified

159
in § 52.103(g) as of [EFFECTIVE DATE] shall, before the NRC's docketing of the license

holder's certifications described in section § 50.82(a)(1) or § 52.110(a) of this chapter, make

these provisions no later than [DATE 365 DAYS AFTER EFFECTIVE DATE].

Part 52 .- LICENSES, CERTIFICATIONS, AND APPROVALS FOR NUCLEAR POWER

PLANTS

7. The authority citation for part 52 continues to read as follows:

Authority: Atomic Energy Act secs. 103, 104, 147, 149, 161,181, 182, 183, 185, 186,

189, 223, 234 (42 U.S.C. 2133, 2201, 2167, 2169, 2232, 2233, 2235, 2236, 2239, 2282);

Energy Reorganization Act secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851);

Government Paperwork Elimination Act sec. 1704 (44 U.S.C. 3504 note); Energy Policy Act of

2005, Pub. L. 109-58, 119 Stat. 594 (2005).

•• • I ---§52.7,ad • nw argrah a)29torea a fl160 •
S52.79Content lictins; technicalinformationnfia analysis report.

(•o 1/. In § 52.80, revise paragraph (d) to read as follows:(

§ 52.80 Contents of applications; additional technical information.

(d)The applicant's plans for implementing the requirements of § 50.155 and 10 CFR part

50, appendix E, section VII, including a schedule for achieving full compliance with these

requirements, and a description of:

(1) the integrated response capability required by § 50.155(b);

(2) the equipment upon which the strategies and guidelines required by § 50.155(b)(1)

re'ly, including the planned locations of the equipment and how the equipment and SSCs meet

the design requirements of § 50.155(c); and

(3) the strategies and guidelines required by § 50.155(b)(2).

•,,Jd \ ,atT\4, . .. ,,,,

161
* *r ' *k

*t *

n application for a standard design approval that is subject

design features included in the plant design under § 50.15•

•wJ~ose design features comply with the requirements of •,

an

12. In § 52.157, paragraph (f)(33) to read

§ 52.157 Contents technical information in final analysis report.

* *

(f) * *

,( ) For an application for a manufacturing license that is subject to § 50.155(i iption of the design features included in the plant design under § 50.155(d) and ai

]nation of how those design features comply with the requirements of § 50.155(d).

162
Dated at Rockville, Maryland, this xxth day of Xxxxx, 2015.

For the Nuclear Regulatory Commission.

Annette L. Vietti-Cook,

Secretary of the Commission.

163
TO:

FROM:

SUBJECT:

NOTATION VOTE

RESPONSE SHEET

Annette Vietti-Cook, Secretary

COMMISSIONER OSTENDORFF

SECY-15-0065: PROPOSED RULEMAKING:

MITIGATION OF BEYOND-DESIGN-BASIS EVENTS

(RIN 3150-AJ49)

Approved X

Not Participating

COMMENTS: Below

Disapproved X Abstain

Attached X None

SIGNATURE

7 I,)/P/f

DATE

Entered in "STARS" Yes NNo
Commissioner Ostendorff's Comments on SECY-15-0065: "Proposed Rulemaking: Mitigation of Beyond-Design-Basis Events"

This package represents a tremendous body of work and I commend the staff for this

accomplishment. I also thank the ACRS for their review of the rulemaking package and their

timely and thoughtful recommendations. Lastly, I would like to recognize the extensive efforts of

licensees and other stakeholders who have participated in development of this proposed rule.

Significant progress has been made by the NRC and the regulated industry in response to the

March 2011 accident at Fukushima Dai-ichi. The NRC's post-Fukushima actions represent a

comprehensive and integrated suite of safety enhancements to address lessons learned from

the accident. The Mitigation-of-Beyond-Design-Basis-Events rulemaking proposed in SECY-1 5-

0065 integrates many of the NRC's post-Fukushima actions into one rulemaking, including: (1)

Near-Term-Task-Force (NTTF) Recommendation 4 regarding enhanced mitigation capability for

station blackout events; (2) NTTF Recommendation 7 regarding enhanced spent fuel pool

safety; (3) NTTF Recommendation 8 regarding enhanced onsite emergency response

capabilities; and (4) actions to enhance emergency preparedness associated with NTTF

Recommendations 9, 10, and 11.

It is important to note that the key safety enhancements described in this draft proposed

rulemaking package are alreadY being realized through licensee implementation of NRC orders

issued after the Fukushima accident, specifically: (1) mitigating strategies as required by Order

EA-12-049; (2) spent fuel pool instrumentation as required by Order EA-12-051; and (3)

emergency response capabilities for multi-unit events as implemented by licensees in

conjunction with Order EA-12-049. Several facilities are already in full compliance with the

requirements of these orders, and the remaining plants are scheduled to be in compliance with

these orders by the end of 2016. The Mitigation-of-Beyond-Design-Basis-Events rulemaking

would codify these requirements to make them generically applicable to current and future

licensees. In parallel, efforts are underway to address venting capability for Mark I and Mark II

boiling water reactor containments as described in SECY-15-0085, and to address reevaluated

seismic and flooding hazards as discussed in SECY-15-0059.

The Mitigation-of-Beyond-Design-Basis-Events rulemaking is a significant step toward bringing

the post-Fukushima actions to closure. I approve publication of the proposed rule with two

exceptions as discussed in more detail below.

Severe Accident Mitigation Guidelines: I disapprove the staff's proposal to impose

requirements for Severe Accident Management Guidelines (SAM Gs).

As discussed in SECY-15-0065, in the mid- to late-i1980s, the NRC and the nuclear power

industry identified a need to consider plant conditions that could lead to a severe accident.

These efforts led to the nuclear industry voluntarily initiating a coordinated program on accident

management in 1990. Subsequently, the NRC accepted the industry's approach for mitigating

the consequences of severe accidents, including licensee regulatory commitments to implement

plant-specific SAMGs. In the late 1990's, the staff also assessed implementation of SAMGs at

a select number of plants and concluded that the results of the voluntary initiative achieved the

•NRC's overall objectives.

Following the Fukushima accident, the NRC took a more comprehensive look at SAMG

implementation under Temporary Instruction (TI) 2515/1 84, "Availability and Readiness

Inspection of Severe Accident Management Guidelines (SAMGs)." While the inspections

Page 1 of 5
verified that all licensees had SAMGs, the inspections also revealed some inconsistencies in SAMG implementation with regard to maintenance of the procedures and training. This led to

the NTTF recommendation that the SAMGs should be a regulatory requirement. This

recommendation was predicated in part on the NTTF's assertion, in the context of NTTF

Recommendation 1, that voluntary industry initiatives should not serve as a substitute for

regulatory requirements.

In SRM-SECY-99-063, "The Use by Industry of Voluntary Initiatives in the Regulatory Process,"

the Commission stated that:

The Commission has approved the staffs recommendation that voluntary

industry initiatives will not be used in lieu of regulatory action where a question of

adequate protection of public health and safety exists. Voluntary industry

initiatives are approved as an appropriate substitute for NRC regulatory action

where the action to be taken is needed to meet existing requirements or for

cases where substantial increase in overall protection can be achieved with costs

of implementation justifying the increased protection. The Commission has

agreed that the current regulatory framework does not preclude voluntary

industry initiatives and existing regulatory processes can be used to support

implementation of voluntary initiatives.

The staff subsequently provided to the Commission SECY-1 3-01 32, "U.S. Nuclear Regulatory

Commission Staff Recommendation for the Disposition of Recommendation 1 of the Near Term

Task Force Report." In SECY-13-0132, the staff proposed re-affirmation of the Commission's

expectation that such initiatives may not be used in lieu of NRC regulatory action on adequate

protection issues, but did not otherwise propose a restriction on the use of voluntary initiatives.

In SRM-SECY-13-0132, the Commission did not object to the staff's proposal, thereby

upholding the policy in SRM-SECY-99-063.

Consistent with this policy, it is not necessary to include the SAMGs as a regulatory requirement

as part of the proposed Mitigation-of-Beyond-Design-Basis-Events rulemaking because they are

not required for adequate protection. As such, it is appropriate for the industry to continue

implementation of the SAMGs as a voluntary initiative. Further, as stated in a May 11, 2015

letter from the Nuclear Energy Institute, industry supports submission of a docketed

commitment by each site to ensure: (1) maintenance of the SAMG strategies; (2) integration

with Emergency Operating Procedures and other guidelines; (3) timely incorporation of Owners

Group revisions; and (4) establishment of configuration controls. As discussed in SECY-13-

0132, docketed commitments ensure that licensee voluntary initiatives are well-documented and

transparent to the public. Further, there is an existing process to track docketed commitments.

NEI-99-04, "Guidelines for Managing NRC Commitments" was found acceptable by the staff, as

discussed in SECY-00-0045. While a voluntary industry initiative is not directly enforceable, the

NRC can provide oversight to verify implementation. As such, licensee failure to meet a formal

commitment could be the basis for a notice of deviation, and any associated finding would be

captured by the Reactor Oversight Process. The NRC should update the Reactor Oversight

Process to explicitly provide periodic oversight of industry's implementation of the SAMGs.

Additionally, if the Commission were to require SAMGs as a regulatory requirement, it would

need to be supported by a robust analysis of the benefits and costs to justify the requirement as

a cost-justified, substantial safety enhancement in accordance with 10 CFR 50.109, the Backfit

Rule. Furthermore, the use of qualitative factors as part of this justification would need to be

Page 2 of5S
consistent with the Commission's poiicy on the use of qualitative factors, recently articulated in SRM-SECY-14-0087. As stated in the qualitative factors SRM:

To ensure that qualitative factors are used in a judicious and disciplined manner,

the revised guidance should continue to encourage quantifying costs to the

extent possible and use qualitative factors to inform decision making, in limited

cases, when quantitative analyses are not possible or practical (i.e., due to lack

of methodologies or data).

The staff's use of qualitative factors in the regulatory analysis of the proposed SAMG

requirement in SECY-1 5-0065 did not fulfill these objectives because consideration of

qualitative factors took precedence over available quantitative information.

As discussed in the staff's regulatory analysis (see SECY-15-0065, Enclosure 3), the available

risk information indicates that SAMGs would have a small safety benefit. Specifically, the staff

referenced technical work showing that the risk of a severe accident is well below a level that

equates to one-tenth of 1 percent of the surrounding population's latent cancer fatality risk and

therefore, from a quantitative standpoint, achieving risk reductions that might satisfy backfitting

requirements would be unlikely. The staff goes on to discuss the consideration of qualitative

factors, such as defense-in-depth that would be gained by requiring SAMGs. The NRC's

regulatory framework has its foundation in the concept of defense-in-depth and I fully appreciate

the need to maintain this foundation. However in this situation, the additional defense-in-depth,

that would be gained from making the SAMGs .a regulatory requirement rather than a voluntary

initiative, does not provide a sufficient basis to support this PrOvision of the proposed

rulemaking.

As we go forward with resolution of remaining post-Fukushima actions and we consider

regulatory requirements in other areas, we must adhere to our regulatory framework and uphold

the principles of good regulation to ensure that new requirements are appropriately justified.

New Reactor Desigqn Requirements: I disapprove imposing additional requirements specific

to new reactor applicants.

I appreciate the views provided in the non-concurrence associated with this provision of the

draft proposed rule. I have carefully considered the pros and cons of the staff's recommended

approach. In doing so, I reflected on prior Commission direction on the post-Fukush~ima actions.

In SRM-SECY-11-01 24, the Commission directed that:

As the staff evaluates Fukushima lessons-learned and proposes modifications to

NRC's regulatory framework, the Commission encourages the staff to craft

recommendations that continue to realize the strengths of a performance-based

system as a guiding principle. In order to be effective, approaches should be

flexible and able to accommodate a diverse range of circumstances and

conditions. In consideration of events beyond the design basis, a regulatory

approach founded on performance-based requirements will foster development

of the most effective and efficient, site-specific mitigation strategies, similar to

how the agency approached the approval of licensee response strategies for the

"loss of large area" event under its B.5.b program.

It is under this backdrop that I disapprove the staff's recommendation to impose additional

requirements on new reactor applicants in favor of allowing a more flexible approach for new

Page 3 of 5
reactor applicants to demonstrate compliance with the Mitigation-of-Beyond-Design-Basis- Events rule. The significant safety enhancements embodied by the proposed Mitigation-ofBeyond-Design-Basis-Events rulemaking provide an appropriate standard for both operating

and new reactors. Absent a compelling basis, I see no reason to impose greater or different

standards for new reactors with regard to Fukushima lessons learned.

I will note further that this provision should have been raised to the Commission as a separate

matter in advance of the draft proposed rulemaking package. All of the other provisions in the

proposed rule are a logical evolution from previous Commission direction on the postFukushima actions. This one, however, is not. If significant technical or policy issues arise

regarding compliance with the proposed Mitigation-of-Beyond-Design-Basis-Events rule during

any design certification review, the staff should raise those issues to the Commission's attention

early in the review process.

Path Forward for Remaining Tier 2 and 3 Actions:

The proposed Mitigation-of-Beyond-Design-Basis-Events rulemaking represents a major step

forward in addressing the lessons learned from the Fukushima accident. This effort is the

culmination of a holistic evaluation of lessons learned with priority given to those actions with

the greatest potential for safety improvement. Given the risk reduction that this proposed

rulemaking and other ongoing regulatory action actions represent, it is time to reflect on the

accomplishments to date and to set a clear path forward to bring the remaining post-Fukushima

actions to a close. At this juncture, it is appropriate to ask what, if any, additional enhancements

are needed. It is time to scrutinize the remaining Tier 2 and Tier 3 post-Fukushima actions in

the context of what has already been achieved, what additional incremental risk reduction is

possible, and whether these actions constitute cost-justified substantial safety enhancements.

In the context of Project AIM and the Cumulative Effects of Regulation initiative, these questions

must be answered before expending significant resources on these efforts.

For example, in SECY-15-0057, "Seventh 6-Month Status Update on Response to Lessons

Learned from Japan's March 11, 2011, Great Tohoku Earthquake and Subsequent Tsunami,"

dated April 9, 2015, and in prior status updates, the staff has repeatedly asserted that:

NRC staff remains confident that the EP programs in support of nuclear power

plants provide an adequate level of protection of the public health and safety and

that appropriate protective actions can and will be taken in the event of a

radiological event. This includes evacuations and the use of potassium iodide

(KI). Available information and studies from the Fukushima accident have not

changed the staff's position. The current size of EPZs [Emergency Planning

Zones] provides reasonable assurance that adequate protective measures can

and will be taken in the event of a radiological emergency at an existing nuclear

power plant .. .. On December 25, 2014, the Japanese Investigation Committee

on the Fukushima Health Survey released results on external radiation exposure

dose estimates and thyroid doses, using survey information from over 531,000

affected people in the Fukushima Prefecture. NRC staff is in process of

obtaining and reviewing this information for insights on the EPZ size and KI

approach within the U.S. The dose estimates do not appear to call the EPZ or KI

approach into question.

In light of the conclusions in SECY-1 5-0057 related to EPZ size and the use of KI, it would

appear from one Commissioner's perspective that this Tier 3 action could be closed.

Page 4 of 5
We must also bring these actions to resolution in a timely manner to uphold our regulatory principle of reliability, wherein our regulations should be promptly, fairly, and decisively

administered. To that end, I request that the staff provide to the Commission by the end of

October 2015 a plan and schedule for resolving all remaining Tier 2 and Tier 3 actions.

Summary:

I approve the proposed rule notice for publication in the Federal Register, upon removal of the

proposed requirements for SAMGs (10 CFR 50.155(b)(3)) and the proposed requirements for

new reactors (10 CFR 50.155(d)). Prior to publication, the staff should remove questions 4, 5

and 6 in Section VII, "Specific Requests for Comments," of the draft FRN and modify Question 1

of Section VII, modify Question 1 in Section XII, "Cumulative Effects of Regulation," of the draft

Federal Register Notice (FRN) to reflect the removal of the SAMGs and new reactor provisions

of the draft proposed rule. The staff should also update the implementation schedule in 10 CFR

155(h), "Implementation," to reflect the revised scope of the proposed rule.

Lastly, in his presentation during the July 9, 2015 Commission meeting on the draft proposed

rule, Mr. David Lochbaum of the Union of Concerned Scientists, suggested that the NRC

"include a concise, plain-English brochure in the rulemaking package to help stakehoiders

assemble the pieces and see the overall picture." I thank Mr. Lochbaum for his suggestion and

agree that the staff should develop a plain-language brochure or other similar communications

tool to facilitate meaningful public engagement on this proposed rule.

Page 5 of 5
1''

NOTATION VOTE

RESPONSE SHEET

TO:

FROM:

Annette Vietti-Cook, Secretary

Commissioner Baran

SUBJECT: SECY-15-0065: PROPOSED RULEMAKING:

MITIGATION OF BEYOND-DESIGN-BASIS EVENTS

(RIN 3150-AJ49)

Approved XX

Not Participating

COMMENTS: Below

Disapproved Abstain

_____Attached XX None

3/, /,f

DATE

Yes XX No __ Entered in "STARS"
Commissioner Baran's Comments on SECY-15-0065, "Proposed Rulemaking: Mitigation of Beyond-Design-Bases Events"

In this paper, the NRC staff asks the Commission to approve a proposed rule to

establish regulatory requirements for nuclear reactor licensees and certain applicants to mitigate

beyond-design-basis events. This rulemaking is a key component of the agency's response to

the March 2011 Fukushima Dai-ichi accident in Japan. Shortly after the accident, the

Commission directed a Near-Term Task Force (NTTF) comprised of agency staff to conduct a

systematic review of NRC processes and regulations to determine whether improvements to

NRC's regulatory framework were needed. The NTTF provided the Commission with a report

that presented twelve overarching recommendations for enhancing nuclear plant safety. The

Commission subsequently approved the prioritization of post-Fukushima actions into three tiers

and directed the NRC staff to strive for completion and implementation of the safety

enhancements within five years.

This proposed rule is the result of extensive work by the NRC staff to address lessons

learned from the Fukushima accident. It responds to NTTF recommendations 4 and 7 by

proposing to make the requirements in previous NRC orders for mitigation of beyond-designbasis events and for reliable spent fuel pool instrumentation generally applicable to all nuclear

plants. It addresses NTTF recommendations 8 and 9 by proposing to require Severe Accident

Management Guidelines (SAMGs) as part of an integrated emergency response capability that

would integrate several nuclear plant emergency procedures. In response to NTTF

recommendations 10 and 11, the proposed rule would set requirements for enhanced onsite

emergency response capabilities. In addition, the proposed rule would reqUire new reactor

designs to incorporate features to enhance coping capability and reduce reliance on operator

manual actions in the event of a severe accident.

I approve the NRC staff's recommendation to publish the proposed rule in the Federal

Register for a 75-day public comment period. There appears to be broad agreement on the

need for the proposed requirements related to mitigating strategies, spent fuel pool•

instrumentation, and enhanced emergency response capabilities. However, some stakeholders

have raised concerns about the proposed requirements for SAMGs and new reactor design

features.

Severe Accident Management Guidelines

I support including a proposed regulatory requirement to develop, maintain, and train on

SAMGs in the proposed rule. Currently, SAMGs are implemented through a voluntary industry

initiative, which was developed to enhance the capabilities of nuclear plant operators to prevent

and mitigate severe accidents following the Three Mile Island accident. In a January 1995 letter

to the Nuclear Energy Institute (NEI), the NRC staff accepted the industry's regulatory

commitment to implement accident management programs at each nuclear plant in lieu of

pursuing regulatory requirements for emergency response improvements. Later that year,

nuclear plant licensees made voluntary docketed regulatory commitments to implement what

are now referred to as SAMGs. 1 These voluntary regulatory commitments included schedules

for implementation, and each licensee agreed to assess current capabilities to respond to

severe accident conditions using NEI guidance approved by NRC. This guidance stated that

licensees would be expected to develop SAMGs, link the SAMGs to the plant's emergency plan,

1 SECY-96-088, "Status of the Integration Plan for Closure of Severe Accident Issues and the Status of

Severe Accident Research" (April 29, 1996).

1
incorporate severe accident material into training programs, and establish a means to consider and adopt new severe accident information. 2 All nuclear plants implemented SAMGs by

December 31, 1998, and the NRC staff initiated a pilot inspection program.3 However, in 2000,

the Reactor Oversight Process was implemented and the NRC staff concluded "that regular

inspection of SAMG was not appropriate because the guidelines are voluntary and have no

regulatory basis."'4 As a result, SAMGs were left to licensees to maintain without NRC

inspections for the next twelve years.

Following Fukushima, at the request of the NTTF, the NRC inspected licensees'

implementation of SAMGs. As described in the proposed rule, "NRC found that some licensees

had not maintained the SAMGs in accordance with the latest revisions of the applicable industry

generic technical guidelines nor conducted training in a consistent and systematic manner."

The NRC inspectors attributed this to "inconsistent implementation and training on SAMGs to

the voluntary nature of this initiative."'5 The actual inspection results were much more specific

than these general conclusions. For example, in Region I, inspectors found that SAMG triggers

at one site had not been revised since initial issuance in 1998 even though they were directly

impacted by license basis changes over the years, such as power uprates.6 At another site in

Region II, eight of the 33 licensee's emergency response organization members were not

qualified on SAMGs and two of the four Site Emergency Directors lacked SAMG training.7 In

Region Ill, another site had not fully implemented the initial owners group SAMGs from the

1990s or any of the subsequent revisions. Inspectors also found that no exercise or drill had

been conducted at the site since 1998 and recurring training on SAMGs for licensee emergency

response organization members did not exist.8 Overall, follow-up inspections resulted in four

green findings being issued to four sites in Regions Ill and IV related to problems with

implementation of the SAMG regulatory commitments. This troubling history was a major

reason the NTTF recommended making SAMGs mandatory rather than continuing to rely on a

volun~tary initiative.

The NRC staff recommends requiring SAMGs in order to ensure that they are

enforceable. In light of the results of the post-Fukushima inspections, the agency cannot be

confident that a continued voluntary approach to SAMGs would provide reasonable assurance

that SAMGs would be maintained and effective at every plant in the United States. As the

Deputy Executive Director for Reactor and Preparedness Programs explained at the July 9,

.2015, Commission meeting on the proposed rule:

It's because we see [SAMGs are] important and it's because we recognize what can

happen if you have an accident and you don't have the ability to deal with [it], make

decisions about what should happen in the plant post-severe accident, it's because of

2 Section 5 of NEI 91-04, Revision 1, "Severe Accident Issue Closure Guidelines."

3 SECY-98-1 31, "Status of the Implementation Plan for Closure of Severe Accident Issues and the Status

of Severe Accident Research" (June 8, 1998).

4 Inspection Manual Chapter 0308, Attachment 2, "Technical Basis for Inspection Program" (October 16,

2006).

STemporary Instruction 2515/184, "Availability and Readiness Inspection of Severe Accident

Management Guidelines (SAMGs)" and Inspection Summary," ADAMS Accession No. ML1 111 5A053.

6 Region I Completion of Temporary Instruction-I184, "Availability and Readiness Inspection of Severe

Accident Management Guidelines (SAMGs)" (May 27, 2011 ).

7 Region II Completion of Temporary Instruction-i184, "Availability and Readiness Inspection of Severe

Accident Management Guidelines (SAMGs)" (June 2, 2011).

8 Region Ill Completion of Temporary Instruction-184, "Availability and Readiness Inspection of Severe

Accident Management Guidelines (SAMGs)" (June 1, 2011).

2
that, that we want a fair amount of assurance that plants will have them, they'll be trained on them, that they'll be maintained.

In its mitigating strategies order, the Commission required, as a matter of adequate

protection, additional onsite and portable equipment to prevent core damage during beyonddesign-basis events.. Though SAMGs focus on post-core damage situations, the NRC staff

expects that those guidelines will rely on this FLEX equipment and accompanying FLEX support

guidelines. In my View, it does not make sense to require FLEX equipment as a matter of

adequate protection but not require the guidelines and integration of procedures necessary to

ensure the effective use of that equipment to halt core damage in certain severe accidents.

The staff's recommendation to include a requirement for SAMGs in the proposed rule is

supported by a detailed regulatory analysis. The staff attempted to quantify some of the

benefits of the proposed requirement and provide a transparent qualitative analysis of other

benefits. The staff found that "because the available quantitative risk information is not a

complete measure of the SAMG safety benefits, the NRC relied on quantitative and qualitative

reasons to conclude that SAMG requirements would result in a substantial additional protection

for public health and safety.'' 9

This use of qualitative factors is consistent with Commission direction in

SRM-SECY-14-0087, which stated that qualitative factors should be used "when quantitative

analyses are not possible or practical." This direction must be understood in the context of what

the agency is trying to accomplish with a regulatory analysis of a proposed rule. A basic tenet

of regulatory analysis is that it should examine all relevant costs and benefits, whether they can

be quantified or not. Each benefit and cost should be examined, if feasible. If a benefit or cost

can be adequately quantified, there is obviously no need to conduct a qualitative analysis of that

benefit or cost. But if a benefit or cost cannot be adequately quantified, it is appropriate and

necessary to provide a qualitative analysis of that benefit or cost. The ability to adequately

quantify one or two benefits (or costs) clearly cannot preclude a qualitative consideration of

other, non-quantifiable benefits (or costs). The alternative is an incomplete or inadequate

examination of the costs and benefits of a proposed regulatory action.

In the case of the proposed requirement for SAMGs, the staff found that "the available

quantitative risk information is not a complete measure of the SAMG safety benefits."'9 At the

July 9, 2015, Commission meeting, the NRC staff further explained the limitations of the SAMG

quantitative analyses:

unfortunately, we did not have a PRA [probabilistic risk assessment], in fact, with a

measure of SAMG benefits. In fact, I think it would take a number of PRAs, and

certainly something a little more expansive than some of the PRAs we have available to

us today. .... So it would have been, in our view, a pretty large undertaking. As a

practical matter, we simply couldn't do it in our schedule,.

In other words, the staff did not have the tools to do a full quantitative analysis of the safety

benefits of SAMGs. The staff therefore turned to a quantitative analysis completed for a

different rulemaking, the Containment Protection and Release Reduction (CPRR) draft

regulatory basis. The staff explained the limits of the CPRR analysis at the Commission

meeting:

9 Regulatory Analysis: Proposed Rulemaking to Address Mitigation of Beyond-Design-Basis Events (April

2015) at 2.

3
We fully understood that they were limited and we were challenged as such. We knew they were limited because they - first of all, they were never intended to measure

SAMGs .... It was limited in terms of this was only Mark 1 and Mark 2 designs, clearly. I

have mentioned they don't look at all of the core damage sequences, so they weren't

complete.

Although the staff found this quantitative analysis helpful in generating risk insights, the staff

understood that the analysis was incomplete and that it needed to supplement that analysis with

a qualitative examination of the safety benefits of requiring SAMGS.

The Advisory Committee on Reactor Safeguards (ACRS) agreed with the staff that the

limited quantitative analysis of SAMGs "is not sufficient to justify broad conclusions about the

quantitative benefitS from SAMGs for a complete spectrum of events across the entire U.S.

operating reactor fleet."1 ° During the June 11, 2015, Commission meeting with the ACRS,

Chairman Stetkar explained that the benefits of SAMGs are very site specific. He stated that

"for some units, there would be more benefit than others," but "we don't know how much that

benefit is in a quantitative sense." He concluded that SAMGs "may very well" be a case where

the staff doesn't currently have the tools to do a full quantitative analysis.

Thus, the NRC staff and the ACRS agree that it was not practical for the staff to

complete an adequate quantitative analysis of the safety benefits of SAMGs. Consistent with

Commission direction and the basic principles of cost-benefit analysis, the staff supplemented

its limited quantitative analysis with a qualitative description of the safety benefits of requiring

SAMGs. This analysis showed that requiring SAMGS provides a cost-justified substantial safety

benefit.

On the other hand, according to the Nuclear Energy Institute's presentation at the July 9,

2015, Commission meeting on the proposed rule, the staff's proposal to require SAMGs would

not require licensees to do anything more than they are already doing voluntarily and would not

result in a regulatory burden. Because requiring SAMGs would provide a substantial safety

benefit at little or no additional cost, it is clear that the proposed rule should include a

requirement for SAMGs. Including the requirement in the proposed rule does not constitute a

final agency decision on whether to impose the regulatory requirement. It merely allows

members of the public to share their views on the proposal.

In the absence of majority Commission support for proposing to require SAMGS, the

proposed rule should at least specifically seek public comment on whether the final rule should

include such a requirement. This would provide stakeholders who did not have an opportunity

to participate in the recent Commission meeting a chance to express their views about whether

SAMGs should be required, whether and how such a requirement is supported by the staff's

regulatory analysis, whether the quantitative and qualitative analyses of the benefits of requiring

SAMGs were adequate, and how a requirement would compare to a voluntary initiative. After a

strong, well-supported staff recommendation to propose a SAMGs requirement, we owe it to the

public to get a broad range of views on the merits of such a requirement.

Even if the proposed rule does not include a requirement for SAMGs, it should continue

to require integration of the FLEX support guidelines, extreme damage mitigating guidelines,

10 ACRS letter to the NRC Chairman, "Draft SECY Paper, "Proposed Rulemaking: Mitigation of BeyondDesign-Basis Events" (April 22, 2015).

4
and voluntarily maintained SAMGs with the existing Emergency Operating Procedures at each plant. In addition, the NRC staff should update the Reactor Oversight Process (ROP) to

explicitly provide for NRC oversight of the voluntary regulatory commitment on SAMGs, at least

once every three-year ROP cycle. NRC guidance should clarify how deviations or performance

deficiencies would be addressed.

New Reactor Designs

I support including in the proposed rule a provision to require applicants for new power

reactor designs to incorporate design features to provide longer coping durations and minimize

reliance on human actions to maintain or restore key safety functions .during beyond-designbasis events causing an extended loss of AC power. Such design features would provide

enhanced margins of safety and allow plant operators more time to diagnose and manage

challenges during a severe accident. As the proposed requirement would only apply to

applicants for new power reactor designs, not to currently operating reactors, certified designs,

or active license applications referencing certified designs, the backfit rule and Part 52 issue

finality provisions do not apply. This approach of establishing additional design requirements for

new reactor applicants to enhance safety is consistent with NRC's approach to the Aircraft

Impact Assessment rule, which required new reactor applicants to perform assessments and

incorporate into their designs the capability to assure core cooling and containment integrity, as

well as spent fuel pool cooling and integrity, in response to beyond-design-basis events. The

staff's proposed approach also follows the Commission's 2008 Policy Statement on the

Regulation of Advanced Reactors, which expressed the Commission's expectation that

"advanced reactors will provide enhanced margins of safety and/or use simplified, inherent,

passive, or other innovative means to accomplish their safety and security functions." 11 The

Policy Statement highlighted "[s]implified safety systems that, where possible, reduce required

operator actions" as one such design feature. The staff's recommended provision and the

Advanced Reactor Policy Statement both recognize that retrofitting reactors after they are built

to enhance safety is often far more expensive than engineering a solution during the design

phase of a project. I agree with the NRC staff that the proposed rule should take the

commonsense approach of improving designs to increase the capability, reliability, and flexibility

of the plant features relied upon to mitigate beyond-design-basis events.

In the absence of majority Commission support for proposing this requirement for new

reactor designs, the proposed rule should at least specifically seek public comment on whether

the final rule should include such a requirement. This would provide stakeholders other than

those who participated in the recent Commission meeting a meaningful opportunity to express

their views on this matter.

Conclusion

I appreciate the staff's hard work on this important post-Fukushima rulemaking to

address several of the recommendations made by the Near Term Task Force four years ago. I

also want to thank the ACRS and the stakeholders who have had an opportunity to participate in

the development of this proposal for their thoughtful insights and recommendations.

I agree with Commissioner Ostendorff that the staff should take the suggestion of David

Lochbaum of the Union of Concerned Scientists to develop a plain-language brochure or other

similar communication tool for inclusion in the proposed rulemaking package describing the

11 73 FR 60612 (October 14, 2008).

5
elements of the proposed rule and placing them in the broader context of the response to the Fukushima accident. This type of document would assist the public in digesting the proposed

rule and facilitate public comment on the proposal.

The NRC staff is currently planning to provide a paper to the Commission in October on

the resolution of the remaining post-Fukushima Tier 2 and Tier 3 actions. I look forward to

reviewing the paper at that time.

6