40006 Federal Register / Vol. 74, No. 152 / Monday, August 10, 2009 / Proposed Rules 

NUCLEAR REGULATORY 

COMMISSION 

10 CFR Parts 50 and 52 

[NRC–2004–0006] 

RIN 3150–AH29 

Risk-Informed Changes to Loss-ofCoolant Accident Technical 

Requirements 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Supplemental proposed rule. 

SUMMARY: The Nuclear Regulatory 

Commission (NRC) is proposing to 

amend its regulations that govern 

domestic licensing of production and 

utilization facilities and licenses, 

certifications, and approvals for nuclear 

power plants to allow current and 

certain future power reactor licensees 

and applicants to choose to implement 

a risk-informed alternative to the 

current requirements for analyzing the 

performance of emergency core cooling 

systems (ECCS) during loss-of-coolant 

accidents (LOCAs). The proposed 

amendments would also establish 

procedures and acceptance criteria for 

evaluating certain changes in plant 

design and operation based upon the 

results of the new analyses of ECCS 

performance. 

DATES: Submit comments on this 

supplemental proposed rule by 

September 24, 2009. Submit comments 

specific to the information collections 

aspects of this supplemental proposed 

rule by September 9, 2009. Comments 

received after the above dates will be 

considered if it is practical to do so, but 

assurance of consideration cannot be 

given to comments received after these 

dates. 

ADDRESSES: You may submit comments 

by any one of the following methods. 

Comments submitted in writing or in 

electronic form will be made available 

for public inspection. Because your 

comments will not be edited to remove 

any identifying or contact information, 

the NRC cautions you against including 

any information in your submission that 

you do not want to be publicly 

disclosed. You may submit comments 

on the information collections by the 

methods indicated in the Paperwork 

Reduction Act Statement of this 

document. 

Federal e Rulemaking Portal: Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

NRC–2004–0006. Address questions 

about NRC dockets to Carol Gallagher, 

(301) 415–5905; e-mail 

Carol.Gallagher@nrc.gov. 

Mail comments to: Secretary, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001, Attn: 

Rulemakings and Adjudications Staff. 

E-mail comments to: 

Rulemaking.Comments@nrc.gov. If you 

do not receive a reply e-mail confirming 

that we have received your comments, 

contact us directly at (301) 415–1966. 

Hand deliver comments to: 11555 

Rockville Pike, Rockville, Maryland 

20852, between 7:30 a.m. and 4:15 p.m. 

during Federal workdays. (Telephone 

(301) 415–1966). 

Fax comments to: Secretary, U.S. 

Nuclear Regulatory Commission at (301) 

415–1101. 

You can access publicly available 

documents related to this document 

using the following methods: 

NRC’s Public Document Room (PDR): 

The public may examine publicly 

available documents at the NRC’s PDR, 

Public File Area O–F21, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland. The PDR 

reproduction contractor will copy 

documents for a fee. 

NRC’s Agencywide Document Access 

and Management System (ADAMS): 

Publicly available documents created or 

received at the NRC are available 

electronically at the NRC’s Electronic 

Reading Room at http://www.nrc.gov/ 

reading-rm/adams.html. From this page, 

the public can gain entry into ADAMS, 

which provides text and image files of 

NRC’s public documents. If you do not 

have access to ADAMS or if there are 

problems in accessing the documents 

located in ADAMS, contact the NRC’s 

PDR reference staff at 1–800–397–4209, 

or (301) 415–4737, or by e-mail to 

PDR.Resource@nrc.gov. 

FOR FURTHER INFORMATION CONTACT: 

Richard Dudley, Office of Nuclear 

Reactor Regulation, U.S. Nuclear 

Regulatory Commission, Washington, 

DC 20555–0001; telephone (301) 415– 

1116; e-mail: richard.dudley@nrc.gov. 

SUPPLEMENTARY INFORMATION: 

Table of Contents 

I. Background 

II. Rulemaking Initiation 

III. Description of Proposed Rule 

IV. Discussion on Public Comments 

A. Comments on Selection of the TBS 

B. Comments on Seismic Considerations 

Related to the TBS 

C. Comments on Thermal-Hydraulic 

Analysis 

D. Comments Related to Probabilistic Risk 

Assessment 

E. Comments Related to Applicability of 

the Backfit Rule 

F. Comments on Topics Requested by the 

NRC 

V. Revised Proposed Rule 

A. Overview 

B. Determination of the Transition Break 

Size 

C. Evaluation of the Plant-Specific 

Applicability of the Transition Break 

Size 

D. Alternative ECCS Analysis 

Requirements and Acceptance Criteria 

E. Risk-Informed Changes to the Facility, 

Technical Specifications, and Procedures 

F. Operational Requirements 

G. Reporting Requirements 

H. Documentation Requirements 

I. Submittal and Review of Applications 

J. Applicability to New Reactor Designs 

VI. Specific Topics Identified for Public 

Comments 

VII. Petition for Rulemaking, PRM–50–75 

VIII. Section-by-Section Analysis of Changes 

IX. Criminal Penalties 

X. Compatibility of Agreement State 

Regulations 

XI. Availability of Documents 

XII. Plain Language 

XIII. Voluntary Consensus Standards 

XIV. Finding of No Significant 

Environmental Impact: Environmental 

Assessment 

XV. Paperwork Reduction Act Statement 

XVI. Regulatory Analysis 

XVII. Regulatory Flexibility Certification 

XVIII. Backfit Analysis 

I. Background 

During the last few years, the NRC has 

had numerous initiatives underway to 

make improvements in its regulatory 

requirements that would reflect current 

knowledge about reactor risk. The 

overall objectives of risk-informed 

modifications to reactor regulations 

include: 

(1) Enhancing safety by focusing NRC 

and licensee resources in areas 

commensurate with their importance to 

health and safety; 

(2) Providing NRC with the 

framework to use risk information to 

take action in reactor regulatory matters, 

and 

(3) Allowing use of risk information to 

provide flexibility in plant operation 

and design, which can result in 

reduction of burden without 

compromising safety, improvements in 

safety, or both. 

The Commission published a Policy 

Statement on the Use of Probabilistic 

Risk Assessment (PRA) on August 16, 

1995 (60 FR 42622). In the policy 

statement, the Commission stated that 

the use of PRA technology should be 

increased in all regulatory matters to the 

extent supported by the state-of-the-art 

in PRA methods and data, and in a 

manner that complements the 

deterministic approach and that 

supports the NRC’s defense-in-depth 

philosophy. PRA evaluations in support 

of regulatory decisions should be as 

realistic as practicable and appropriate 

supporting data should be publicly 

available. The policy statement also 


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1All utilities licensed to operate commercial 

nuclear power plants in the United States are 

members of NEI. 

stated that, in making regulatory 

judgments, the Commission’s safety 

goals for nuclear power reactors and 

subsidiary numerical objectives (on core 

damage frequency and containment 

performance) should be used with 

appropriate consideration of 

uncertainties. 

To implement the policy statement, 

the NRC developed guidance on the use 

of risk information for reactor license 

amendments and issued Regulatory 

Guide (RG) 1.174, ‘‘An Approach for 

Using Probabilistic Risk Assessments in 

Risk-Informed Decisions on Plant 

Specific Changes to the Licensing 

Basis,’’ (ADAMS Accession No. 

ML023240437). This RG provided 

guidance on an acceptable approach to 

risk-informed decision-making 

consistent with the Commission’s 

policy, including a set of key principles. 

These principles include: 

(1) Being consistent with the defensein-depth philosophy; 

(2) Maintaining sufficient safety 

margins; 

(3) Allowing only changes that result 

in no more than a small increase in core 

damage frequency or risk (consistent 

with the intent of the Commission’s 

Safety Goal Policy Statement); and 

(4) Incorporating monitoring and 

performance measurement strategies. 

Regulatory Guide 1.174 further 

clarifies that in implementing these 

principles, the NRC expects that all 

safety impacts of the proposed change 

are evaluated in an integrated manner as 

part of an overall risk management 

approach in which the licensee is using 

risk analysis to improve operational and 

engineering decisions broadly by 

identifying and taking advantage of 

opportunities to reduce risk; and not 

just to eliminate requirements that a 

licensee sees as burdensome or 

undesirable. 

II. Rulemaking Initiation 

The process described in RG 1.174 is 

applicable to changes to plant licensing 

bases. As NRC experience with the 

process and applications grew, the NRC 

recognized that further development of 

risk-informed regulation would require 

making changes to the regulations. In 

June 1999, the Commission decided to 

implement risk-informed changes to the 

technical requirements of Part 50. The 

first risk-informed revision to the 

technical requirements of Part 50 

consisted of changes to the combustible 

gas control requirements in Title 10 of 

the Code of Federal Regulations (10 

CFR) Section 50.44 (68 FR 54123; 

September 16, 2003). Other riskinformed regulations promulgated by 

the NRC include § 50.48(c) on fire 

protection (69 FR 33550; June 16, 2004), 

§ 50.69 on special treatment 

requirements for systems, structures, 

and components (69 FR 68047; Nov. 22, 

2004), and § 50.61 on fracture toughness 

requirements for protection against 

pressurized thermal shock events. 

The NRC also decided to examine the 

ECCS requirements for large break 

LOCAs. A number of possible changes 

were considered, including changes to 

General Design Criterion (GDC) 35 and 

changes to § 50.46 acceptance criteria, 

evaluation models, and functional 

reliability requirements. The NRC also 

proposed to refine previous estimates of 

LOCA frequency for various sizes of 

LOCAs to more accurately reflect the 

current state of knowledge with respect 

to the mechanisms and likelihood of 

primary coolant system rupture. During 

public meetings, industry 

representatives expressed interest in a 

number of possible changes to licensed 

power reactors resulting from 

redefinition of the large break LOCA. 

These include: containment spray 

system setpoint changes; fuel 

management improvements; 

optimization of plant modifications and 

operator actions to address postulated 

sump blockage issues; power uprates; 

and changes to the required number of 

accumulators, diesel start times, 

sequencing of equipment, and valve 

stroke times. 

The Staff Requirements Memorandum 

(SRM), of March 31, 2003, 

(ML030910476), on SECY–02–0057, 

‘‘Update to SECY–01–0133, ‘Fourth 

Status Report on Study of Risk-Informed 

Changes to the Technical Requirements 

of 10 CFR part 50 (Option 3) and 

Recommendations on Risk-Informed 

Changes to 10 CFR 50.46 (ECCS 

Acceptance Criteria)’ ’’ (ML020660607), 

approved most of the NRC staff 

recommendations related to possible 

changes to LOCA requirements and also 

directed the NRC staff to prepare a 

proposed rule that would provide a riskinformed alternative maximum break 

size. The NRC began to prepare a 

proposed rule responsive to the SRM 

direction. However, after holding two 

public meetings, the NRC found that 

there were differences between stated 

Commission and industry interests. 

To reach a common understanding 

about the objectives of the LOCA 

redefinition rulemaking, the NRC staff 

requested additional direction and 

guidance from the Commission in 

SECY–04–0037, ‘‘Issues Related to 

Proposed Rulemaking to Risk-Inform 

Requirements Related to Large Break 

Loss-of-Coolant Accident (LOCA) Break 

Size and Plans for Rulemaking on LOCA 

with Coincident Loss-of-Offsite Power,’’ 

(March 3, 2004; ML040490133). The 

Commission provided direction in a 

SRM dated July 1, 2004, 

(ML041830412). The Commission stated 

that the NRC staff should determine an 

appropriate risk-informed alternative 

break size and that breaks larger than 

this size should be removed from the 

design basis event category. The 

Commission indicated that the proposed 

rule should be structured to allow 

operational as well as design changes 

and should include requirements for 

licensees to maintain capability to 

mitigate the full spectrum of LOCAs up 

to the double-ended guillotine break 

(DEGB) of the largest reactor coolant 

system (RCS) pipe. The Commission 

stated that the mitigation capabilities for 

beyond design-basis events should be 

controlled by NRC requirements 

commensurate with the safety 

significance of these capabilities. The 

Commission also stated that LOCA 

frequencies should be periodically 

reevaluated and should increases in 

frequency require licensees to restore 

the facility to its original design basis or 

make other compensating changes, the 

backfit rule (10 CFR 50.109) would not 

apply. 

On March 29, 2005, in SECY–05– 

0052, ‘‘Proposed Rulemaking for ‘RiskInformed Changes to Loss-of-Coolant 

Accident Technical Requirements,’ ’’ the 

NRC staff provided a proposed rule to 

the Commission for its consideration. In 

an SRM on July 29, 2005, the 

Commission directed the NRC staff to 

publish the proposed rule for public 

comment after making certain changes. 

The most significant change requested 

by the Commission was to require that 

after implementing the alternative 

§ 50.46a requirements, all subsequent 

plant changes made by a licensee would 

be evaluated by the licensee’s riskinformed process to ensure that they 

met all of the requirements in § 50.46a. 

Another change requested by the 

Commission was to address the issue of 

seismic loading of degraded piping 

during very large earthquakes and to 

solicit public comments on the subject. 

On November 7, 2005, (70 FR 67598), 

the proposed rule was published in the 

Federal Register (FR) with a comment 

period of 90 days. On December 6, 2005, 

the Nuclear Energy Institute 1 (NEI) 

requested that the comment period be 

extended for 30 additional days. NEI 

stated that additional time was needed 

to prepare high quality comments that 

reflected an industry consensus 

perspective. On December 20, 2005, the 


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2 Different TBSs for pressurized water reactors 

and boiling water reactors would be established due 

to the differences in design and operation between 

those two types of reactors. 

3The scope of changes subject to the change 

criteria in § 50.46a(f) of the proposed rule would be 

greater than the changes currently subject to 

§ 50.59, which applies only to changes to ‘‘the 

facility as described in the FSAR.’’ The change 

criteria in the proposed rule would apply to all 

facility and procedure changes, regardless of 

whether they are described in the Final Safety 

Analysis Report (FSAR). 

4Requirements for license amendments are 

specified in §§ 50.90, 50.91 and 50.92. They include 

public notice of all amendment requests in the 

Federal Register and an opportunity for affected 

persons to request a hearing. In implementing 

license amendments, the NRC typically prepares an 

appropriate environmental analysis and a detailed 

NRC technical evaluation to ensure that the facility 

will continue to provide adequate protection of 

public health and safety and common defense and 

security after the amendment is implemented. 

5Requirements in § 50.59 establish a screening 

process that licensees may use to determine 

whether facility changes require prior review and 

approval by the NRC. Licensees may make changes 

meeting the § 50.59 requirements without 

requesting NRC approval of a license amendment 

under § 50.90. 

Westinghouse Owners Group (WOG) 

submitted a letter endorsing the NEI 

extension request. On January 18, 2006, 

the NRC extended the comment period 

by 30 days to expire on March 8, 2006. 

As directed by the Commission in its 

SRM on SECY–05–0052, the NRC staff 

addressed the seismic issue by 

preparing a report entitled ‘‘Seismic 

Considerations for the Transition Break 

Size’’ (ML053470439). This report was 

posted on the NRC’s rulemaking Web 

site and a notice of its availability and 

opportunity for public comment was 

published in the FR on December 20, 

2005, (70 FR 75501). A public workshop 

was held on February 16, 2006, to 

ensure that stakeholders understood the 

NRC’s intent and interpretation of the 

proposed rule and two public meetings 

were held on June 28, 2006, and August 

17, 2006, to discuss public comments 

received on the proposed rule. 

After evaluating all written public 

comments and comments received at 

the public meetings, the NRC completed 

draft final rule language that addressed 

nearly all commenters’ concerns. On 

October 31 and November 1, 2006, the 

NRC staff met with the Advisory 

Committee on Reactor Safeguards 

(ACRS) to discuss the draft final rule. In 

a letter dated November 16, 2006, 

(ML063190465) the ACRS provided its 

evaluation of the draft final rule. In its 

November 16, 2006, letter to the 

Commission, the ACRS recommended 

that the rule not be issued in its current 

form. The ACRS recommended 

numerous changes to the rule, primarily 

to increase the defense-in-depth 

provided for large pipe breaks. The NRC 

staff evaluated the ACRS 

recommendations, and in SECY–07– 

0082, ‘‘Rulemaking to Make RiskInformed Changes to Loss-of-Coolant 

Accident Technical Requirements’’; 10 

CFR 50.46a ‘‘Alternative Acceptance 

Criteria for Emergency Core Cooling 

Systems for Light-Water Nuclear Power 

Reactors,’’ (May 16, 2007) sought 

additional guidance from the 

Commission on the priority of the rule 

and on the issues raised by the ACRS. 

In its August 10, 2007, SRM 

(ML072220595) in response to SECY– 

07–0082, the Commission approved 

NRC staff recommendations for a 

revised priority and approach for 

addressing the ACRS concerns and 

completing the final rule. On April 1, 

2008, the NRC staff provided the 

Commission with its planned schedule 

(ML080370355) for completing the rule. 

As the NRC staff proceeded to modify 

the rule in response to the ACRS 

recommendations and the Commission’s 

direction, numerous substantive 

changes were made to the requirements 

in the draft final rule. After 

consideration of the extent of these 

changes, the NRC has decided to 

provide another opportunity for public 

comment focusing on the revised 

proposed rule, in order to provide 

public stakeholders with another 

opportunity to review and comment on 

the new language. Because of the 

interrelated nature of the regulatory 

requirements, the NRC is republishing 

the entire 10 CFR 50.46a proposed rule 

to allow public comments on the 

changed requirements and on other 

closely-related regulatory provisions. 

III. Description of November 2005 

Proposed Rule 

The proposed rule published on 

November 7, 2005, (70 FR 67598) would 

divide the current spectrum of LOCA 

break sizes into two regions. The 

division between the two regions is 

delineated by a ‘‘transition break size’’ 

(TBS). 2 The first region includes small 

size breaks up to and including the TBS. 

The second region includes breaks 

larger than the TBS up to and including 

the DEGB of the largest RCS pipe. Break 

area associated with the TBS is not 

based upon a double-ended offset break. 

Rather, it is based upon the inside area 

of a single-sided circular pipe break. 

Pipe breaks in the smaller break size 

region are considered more likely than 

pipe breaks in the larger break size 

region. Consequently, each break size 

region is subject to different ECCS 

requirements, commensurate with 

likelihood of the break. LOCAs in the 

smaller break size region must be 

analyzed by the methods, assumptions, 

and criteria currently used for LOCA 

analysis; accidents in the larger break 

size region will be analyzed by less 

conservative assumptions based on their 

lower likelihood. Although LOCAs for 

break sizes larger than the transition 

break would become ‘‘beyond designbasis accidents,’’ the proposed rule 

would require licensees to maintain the 

ability to mitigate all LOCAs up to and 

including the DEGB of the largest RCS 

pipe during all operating configurations. 

Licensees who perform LOCA 

analyses using the risk-informed 

alternative requirements could find that 

their plant designs are no longer limited 

by certain parameters associated with 

previous DEGB analyses. Reducing the 

DEGB limitations could enable some 

licensees to propose a wide scope of 

design or operational changes up to the 

point of being limited by some other 

parameter associated with any of the 

required accident analyses. Potential 

design changes include modification of 

containment spray designs, modifying 

core peaking factors, modifying 

setpoints on accumulators or removing 

some from service, eliminating fast 

starting of one or more emergency diesel 

generators, increasing power, etc. Some 

of these design and operational changes 

could increase plant safety because a 

licensee could modify its systems to 

better mitigate the more likely smallbreak LOCAs. Other design changes, 

such as increasing power, could cause 

increases in plant risk. Accordingly, the 

risk-informed § 50.46a option would 

establish risk acceptance criteria to 

ensure the risk acceptability of all 

subsequent facility changes. The 

proposed rule required that all future 

facility changes 3 made by licensees 

after adopting § 50.46a be evaluated by 

a risk-informed integrated safety 

performance (RISP) assessment process 

that has been reviewed and approved by 

the NRC via the routine process for 

license amendments.4 The RISP 

assessment process would ensure that 

the cumulative effect of all plant 

changes involved acceptable changes in 

risk and was consistent with other 

criteria from RG 1.174 to ensure 

adequate defense-in-depth, safety 

margins and performance measurement. 

Licensees with an approved RISP 

assessment process could make certain 

facility changes without NRC review if 

they met § 50.59 5 and § 50.46a 

requirements, including the criterion 

that risk increases cannot exceed a 

‘‘minimal’’ level. Licensees could make 

other facility changes after NRC 

approval if they met the § 50.90 

requirements for license amendments 

and the criteria in § 50.46a, including 


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the criterion that total cumulative risk 

increase cannot exceed a ‘‘small’’ 

threshold. Potential impacts of the plant 

changes on facility security would be 

evaluated as part of the license 

amendment review process. 

The NRC would periodically evaluate 

LOCA frequency information. Should 

estimated LOCA frequencies 

significantly increase such that the risk 

associated with pipe breaks larger than 

the TBS is unacceptable, the NRC would 

undertake rulemaking (or issue orders, if 

appropriate) to change the TBS. In such 

a case, the backfit rule (10 CFR 50.109) 

would not apply. If previous plant 

changes were invalidated because of a 

change to the TBS, licensees would 

have to modify or restore components or 

systems as necessary so that the facility 

would continue to comply with § 50.46a 

acceptance criteria. The backfit rule (10 

CFR 50.109) would also not apply to 

these licensee actions. 

IV. Discussion of Public Comments 

The NRC received comments on the 

proposed rule from six nuclear power 

plant licensees, four nuclear industry 

organizations, two reactor vendors, and 

an NRC employee. The comments 

provided by NEI were specifically 

endorsed by the WOG, the Boiling 

Water Reactors Owners Group 

(BWROG), and three nuclear power 

plant licensees. The NRC considered all 

comments in formulating the revised 

proposed rule language. The NRC also 

received comments from a nuclear 

engineering professor on the expert 

elicitation process for determining the 

relationship between pipe break 

frequency and pipe size that was used 

as the baseline for selecting the 

transition break size. Although these 

comments were submitted for NUREG– 

1829 (Draft Report), ‘‘Estimating Loss-ofCoolant Accident (LOCA) Frequencies 

Through the Elicitation Process’’ 

(ML051520574), they were also 

considered in the development of the 

§ 50.46a final rule. 

Comments and other publicly 

available documents related to this 

rulemaking may be viewed 

electronically on the public computers 

located at the NRC’s Public Document 

Room (PDR), Public File Area O–F21, 

One White Flint North, 11555 Rockville 

Pike, Rockville, Maryland. Selected 

documents, including comments, may 

be viewed and downloaded 

electronically via the Federal e 

Rulemaking Portal. Go to http:// 

www.regulations.gov and search for 

documents filed under Docket ID NRC– 

2004–0006. 

Comments addressed six different 

general topics: selection of the TBS, the 

effect of seismic considerations on the 

TBS, thermal-hydraulic ECCS analyses, 

probabilistic risk analysis, applicability 

of the backfit rule, and comments on 

questions posed by the Commission. 

The comments are discussed below by 

topic area. 

A. Comments on Selection of the TBS 

Comment. NEI stated that the TBS 

proposed for boiling water reactors 

(BWRs) is overly conservative and may 

unnecessarily limit or preclude benefits 

for BWRs. They suggested that the 

specified piping for the BWR TBS 

should be equivalent to the 16-inch 

schedule 80 piping in the shutdown 

cooling suction line inside containment. 

The BWROG supported a reduced TBS 

for BWRs consistent with the 95th 

percentile TBS noted from the expert 

elicitation (i.e., without additional 

conservatisms). 

NRC response. The proposed TBS for 

BWRs is currently based on the crosssectional area of the larger of either the 

shutdown cooling residual heat removal 

(RHR) or feedwater pipes which are 

connected to the RCS inside 

containment. These pipe sizes are 

generally in the 18′′ to 24′′ range, and are 

similar in size to the 95th percentile 

estimates from the expert elicitation 

process results for BWRs at a 10¥5 per 

year frequency. (It should be noted that 

the NRC also considered uncertainties 

in the estimates based on analysis 

sensitivities of the expert elicitation 

results, such as the method of 

aggregating the individual frequency 

estimates. The 95th percentile estimate 

of BWR break size diameter for the 

geometric mean aggregation method is 

approximately 13 inches, and the 

corresponding break size for the 

arithmetic mean aggregation method is 

approximately 20 inches.) The actual 

plant pipe sizes were used as a logical 

selection criterion; because for a given 

size break, it is more likely that a break 

will be circumferentially oriented (i.e., a 

complete severance of the pipe). The 

NRC selected the TBS by considering 

the actual size of the attached piping, 

rather than by selecting a single break 

size value which would conservatively 

bound all plant configurations. For 

BWRs, the pipes connecting to the RCS, 

other than the largest reactor 

recirculation piping or main steam line 

piping, are the feedwater and RHR 

piping. Also, these pipes are large 

enough so that a single-ended break of 

one of them will generally bound the 

total cross-sectional discharge area for a 

double-sided break in smaller size 

feedwater or recirculation pipes. For 

these reasons, the NRC continues to 

believe that the TBS for BWRs should be 

based on the cross-sectional area of the 

larger of either the feedwater or RHR 

lines inside containment. No changes to 

the BWR TBS have been made in the 

revised proposed rule. 

Comment. The Nuclear Energy 

Institute, the Westinghouse Owners 

Group (WOG) and a reactor licensee 

stated that for pressurized-water 

reactors (PWRs) with large piping 

connected to both the hot and cold legs, 

the TBS for the hot leg should be based 

on the largest connecting hot leg pipe, 

and the TBS for the cold leg should be 

based on the largest connecting cold leg 

pipe. These are logical break sizes and 

avoid the arbitrary nature of the size of 

the connecting pipe on the hot leg being 

also applied to breaks on the cold leg. 

If no attached piping is connected to the 

cold leg, the cold leg TBS should be the 

same as the hot leg TBS. The WOG 

stated that the NRC and the industry 

should take the opportunity of this rule 

change to determine the appropriate 

transition break size and not settle for a 

rule that is needlessly conservative. 

Because the rulemaking cannot easily be 

changed in the future as new 

information becomes available, the TBS 

should be based on sound technical 

facts and expert opinions with some 

margin for uncertainties and unknowns 

that could show up in the future and 

erode margins. It is not appropriate to 

set the TBS on the basis of where the 

most benefit would be realized because 

this may change tomorrow and there 

will be no easy recourse. The WOG also 

said that the Commissioners have 

recommended a design basis LOCA cutoff frequency of 10¥5 per reactor year, 

which corresponds to a break size of 

about a three or four-inch diameter 

effective break (for PWRs). The WOG 

believes that selecting a TBS equal to 

the largest attached piping (8- to 12-inch 

diameter break) is very conservative. 

However, the WOG has conducted 

thermal-hydraulic and risk analyses that 

show that there are substantial potential 

benefits for PWR plants even with this 

larger TBS. The WOG agreed that setting 

the transition break size at the sizes of 

the piping attached to the RCS loop is 

reasonable because it will provide 

significant benefit while providing 

substantial margin to account for 

uncertainties or any new information 

that may become available on break size 

vs. frequency. The requirement that 

plants must still be able to mitigate 

breaks larger than the TBS provides 

even more margin. 

NRC response. In developing the basis 

for the PWR TBS, the NRC not only 

used the mean break frequency 

estimates from the expert elicitation but 

also included additional allowances for 


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various uncertainties. To address 

uncertainties in the elicitation process, 

the 95th percentile estimates of break 

size diameter were used. Further, the 

methods of aggregating the individual 

frequency estimates were evaluated for 

sensitivities. For PWRs, the break size at 

a 10¥5 per year frequency using the 

geometric mean method is 

approximately 6 inches, and the 

corresponding break size for the 

arithmetic mean method is 

approximately 10 inches. This is similar 

in size to the cross-sectional area of the 

largest pipe attached to the main reactor 

coolant loop on which the TBS is 

ultimately based. The largest attached 

piping in PWRs is generally in the 12- 

to 14-inch nominal pipe size range (with 

inside diameters corresponding to 10.1 

to 11.2 inches), and typically 

corresponds to the surge line which is 

attached to the hot leg. However, on 

some Combustion Engineering and 

Babcock and Wilcox plants, the largest 

attached pipes may be the RHR, safety 

injection, or core flood lines, which may 

not be similarly attached to the hot leg. 

However, as stated in the statement of 

considerations for the initial proposed 

rule (see 70 FR at 67603–67606), the 

NRC selected only one size which 

would uniformly apply for all locations 

in the RCS piping, because the expert 

elicitation did not provide sufficient 

detail to distinguish the hot leg from the 

cold leg break frequencies. The 

commenters did not provide additional 

information or technical data that 

justifies different break frequencies or 

use of a smaller TBS on the cold leg 

piping. Thus, no changes to the PWR 

TBS were made in the revised proposed 

rule. 

B. Comments on Seismic Considerations 

Related to the TBS 

The TBS specified by the NRC in the 

November 7, 2005, proposed rule did 

not include an adjustment to address 

the effects of seismically-induced 

LOCAs. (See 70 FR at 67604.) On 

December 20, 2005, the NRC released a 

report discussing seismic considerations 

for the transition break size (‘‘Seismic 

Considerations for the Transition Break 

Size’’, December 2006; ML053470439). 

The NRC requested specific public 

comments on the effects of pipe 

degradation on seismically-induced 

LOCA frequencies and the potential for 

affecting the selection of the TBS. These 

public comments were considered in 

the final, published report (NUREG– 

1903, ‘‘Seismic Considerations for the 

Transition Break Size’’, February 2008; 

ML080880140). 

Comment. NEI, WOG, BWROG, and a 

reactor licensee all commented that the 

proposed TBS need not be further 

adjusted due to seismic considerations. 

NEI indicated that the NRC’s December 

20, 2005, report demonstrates that the 

seismically-induced LOCA frequency 

contribution is less than the 10¥5 per 

reactor year guideline used by the NRC 

in determining the TBS. NEI further 

commented that median seismic 

capacities for both the primary piping 

system and primary system components 

are higher than most other safety related 

power plant components within the 

nuclear power plant. Because of these 

relative capacities, NEI said the seismic 

risk from very large, low probability 

earthquakes would be controlled by 

consequential safety component failure. 

In addition, NEI stated that the creation 

of the TBS by itself does not produce a 

physical change in the plant that would 

result in an appreciable change in 

seismic risk. The WOG, the BWROG, 

and a reactor licensee endorsed the NEI 

comments. WOG included an additional 

comment which stated that the NRC’s 

December report indicated that seismic 

loading will only have a small (10 per 

cent) effect on the LOCA frequencies 

estimated by the NRC expert panel 

(NUREG–1829, Draft report, June 2005) 

and that effect is well within the 

uncertainty bounds of the frequency 

estimate of the panel. Furthermore the 

NRC has already included a very 

substantial margin above the break size 

that would correspond to a LOCA 

frequency of 10¥5 per reactor year. 

Therefore, seismic effects should not 

change the transition break size. 

NRC Response. The NRC agrees with 

the commenters’ conclusion that the 

TBS defined in the proposed rule need 

not be adjusted further to account for 

the effects of seismically induced 

LOCAs in piping greater than the TBS. 

In reaching its conclusion the NRC 

considered the comments received as 

well as historical information related to 

piping degradation and the potential for 

the presence of cracks sufficiently large 

that pipe failure would be expected 

under loads associated with rare (10¥5

per year) earthquakes. 

The NRC report NUREG–1903, 

‘‘Seismic Considerations for the 

Transition Break Size’’ (February 2008; 

ML080880140) considered the potential 

contribution from two mechanisms: 

direct piping failures and indirect 

failures. Direct failures are those pipe 

ruptures that result when the combined 

earthquake loadings and normal stresses 

exceed the strength of the pipe. The 

report concluded that direct failures 

from earthquakes with return 

frequencies of 10¥5 per year and 10¥6

per year would not be expected unless 

cracks on the order of 30 percent 

through-wall and approximately 145 

degrees around the piping 

circumference were present at the time 

of the earthquake. The NRC reviewed its 

experience with flaws in reactor coolant 

system piping to assess whether cracks 

of this magnitude have ever been found 

in RCS main loop piping, or if other 

information suggests that cracks of this 

magnitude are likely. The NRC 

considered both fabrication induced 

flaws and service induced flaws. No 

large fabrication flaws have ever been 

reported. If large fabrication flaws were 

present and were not detected by the 

initial fabrication inspections and 

subsequent in-service inspections, it 

would be expected that some would 

have grown through-wall over time as a 

result of fatigue or other mechanisms 

and would have been discovered 

through leakage. This has not been 

observed even though most plants have 

been in operation for more than 20 

years. 

With respect to service induced flaws, 

the NRC also considered the potential 

for known degradation mechanisms to 

induce cracks of the critical size. For 

BWRs, intergranular stress corrosion 

cracking (IGSCC) is the only mechanism 

that has been shown to produce large 

cracks. However, regulatory and 

industry programs have been in place 

for many years to specifically address 

this mechanism and as a result, IGSCC 

is being effectively managed. In PWRs, 

a number of partly through-wall flaws 

and a small number of through-wall 

flaws have been discovered and have 

been attributed to primary water stress 

corrosion cracking (PWSCC). To date, 

all flaws discovered were considerably 

smaller than flaws that would lead to 

failure under 10¥5 and 10¥6 per year 

earthquake loadings. PWR plant owners 

have established programs to address 

PWSCC in susceptible reactor coolant 

system piping welds. They are 

inspecting these welds more frequently 

and, in most cases, are applying 

mitigation techniques to manage 

PWSCC. The NRC is working with the 

American Society of Mechanical 

Engineers (ASME) to establish a 

regulatory framework for improved 

inspection and mitigation of PWSCC in 

these welds. The NRC expects that these 

measures will ensure that PWSCC will 

be effectively managed. As a result of 

the above considerations, the NRC 

considers the likelihood of flaws large 

enough to fail under 10¥5 and 10¥6 per 

year earthquake loadings to be 

sufficiently low that the TBS need not 

be modified to address seismically 

induced direct failures. 

Indirect failures are primary system 

pipe ruptures that are a consequence of 


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failures in non-primary system 

components or structural support 

failures (such as a steam generator 

support). Structural support failures 

could then cause displacements in 

components that stress the piping and 

result in pipe failure. The NRC 

performed studies on two plants to 

estimate the conditional pipe failure 

probability due to structural support 

failure given a low return frequency 

earthquake (10¥5 to 10¥6 per year). The 

results indicated that the conditional 

failure probability was on the order of 

0.1. These studies used seismic hazard 

curves from NUREG–1488, ‘‘Revised 

Livermore Seismic Hazard Estimates for 

Sixty-Nine Nuclear Power Plant Sites 

East of the Rocky Mountains,’’ (April 

1994; ML052640591). More recent 

indirect failure studies were completed 

by the Electric Power Research Institute 

(EPRI) on three plants using updated 

seismic hazard estimates. The updated 

seismic hazard increases the peak 

ground acceleration at some sites. The 

highest pipe failure probability 

calculated for the three plants in the 

industry analyses was 6 × 10¥6 per year. 

Although the EPRI failure probability 

was higher than either of the two cases 

calculated by the NRC, the result is still 

lower than the TBS selection guideline 

of 10¥5 per reactor year. The NRC noted 

in its report that indirect failure 

analyses are highly plant-specific. 

Therefore it is possible that example 

plants assessed in the NRC and EPRI 

analyses are not limiting for all plants. 

The NRC has considered the 

importance of indirect failures on the 

selection of the TBS. For the cases 

considered in both the EPRI and NRC 

studies, the likelihood of indirectly 

induced piping failures resulting from 

major component support failures is less 

than 10¥5 per reactor year, the 

frequency criterion used to select the 

TBS. Also, as noted in the public 

comments, the median seismic 

capacities for both the primary piping 

system and primary system components 

are typically higher than other safety 

related components within the nuclear 

power plant. Because of these relative 

capacities, it is expected that a seismic 

event of sufficient magnitude to cause 

consequential failure within the primary 

system would also induce failure of 

components in multiple trains of 

mitigation systems, or even induce 

multiple RCS pipe breaks. 

Consequently, the risk contribution 

from seismically induced indirect 

failures is expected to depend more 

heavily on the relative fragilities of 

plant components and systems than the 

size of the TBS. Therefore, adjustment 

to the TBS for seismically induced 

indirect LOCAs is also not warranted. 

Comment. In the proposed rule, the 

NRC stated that the final rule might 

include requirements for licensees to 

perform plant-specific assessments of 

seismically-induced pipe breaks and, if 

necessary, implement augmented inservice inspection plans before 

implementing the alternative ECCS 

requirements. NEI, WOG, BWROG, and 

a reactor licensee all commented that 

plant specific assessments should not be 

required to demonstrate that the 

seismically induced pipe breaks do not 

significantly affect the likelihood of 

pipe breaks larger than the TBS. NEI 

indicated that the NRC’s December 20, 

2005 report, ‘‘Seismic Considerations 

for the Transition Break Size’’ 

demonstrates that the seismically 

induced LOCA frequency contribution 

is less than the 10¥5 per reactor year 

guideline limit used by the NRC in 

determining the TBS. NEI further 

commented that indirect LOCA seismic 

studies had been performed by EPRI for 

a limited number of plants using more 

recent seismic hazard estimates than 

those used in the NRC’s December 

study. The EPRI study estimated that 

the indirect LOCA probability was less 

than 10¥5 per year for the plants 

examined. The EPRI study found that 

although the latest seismic hazard has 

increased for some parts of the central 

and eastern United States, there are 

several mitigating phenomena that have 

been established within the new plant 

seismic program which tend to counter 

much of that increase. NEI also stated 

that for a risk informed application, the 

change in risk should be the primary 

metric for decision making. The change 

in risk relative to seismic events is 

estimated to be negligible based upon 

the fact that the TBS threshold does not 

directly impact either the seismic 

hazard or the plant seismic fragilities. 

The WOG, the BWROG, and a licensee 

all endorsed the NEI comments. WOG 

included an additional comment which 

stated that the NRC’s December report 

indicated that seismic loading will only 

have a small (∼10 percent) effect on the 

LOCA frequencies estimated by the NRC 

expert panel (NUREG–1829 Draft 

Report, June 2005) and that effect is well 

within the uncertainty bounds of the 

frequency estimate of the panel. A 

reactor licensee had an additional 

comment that plant specific assessments 

to determine the frequency of 

seismically induced pipe breaks would 

be very difficult to complete. The 

licensee said that because pipe 

inspection and repair are such an 

integral part of plant operations, after a 

plant seismic assessment was 

completed, its conclusions would then 

be prejudiced by implementation of 

piping inspection and repair programs. 

The commenter did not explain in detail 

how the results would be prejudiced. 

The commenter also suggested that 

more technically valid piping failure 

probabilities might be obtainable 

through an extensive research program, 

but noted it is questionable whether this 

would provide additional risk insights. 

NRC response. The NRC disagrees 

with the commenters that plant specific 

assessments of seismically induced pipe 

breaks are not necessary before 

implementing the alternative ECCS 

requirements. As discussed in the 

previous comment, although seismic 

considerations do not significantly 

affect TBS selection, the generic nature 

of the seismic risk studies requires an 

applicant to demonstrate that these 

studies are applicable to its plant and 

site. 

The NUREG–1903 study did 

generically conclude (based on 

operating experience, probabilistic risk 

assessment insights, experimental 

testing, and analysis) that the likelihood 

of seismic-induced unflawed piping 

failure was much less than 10¥5 per 

year. However, a general conclusion 

about the likelihood of seismic-induced 

flawed piping failure could not be 

reached for all plants. Twenty-six plantspecific calculations were conducted in 

NUREG–1903 using available seismic 

hazard assessments for plants east of the 

Rocky Mountains (i.e., from NUREG– 

1488; April, 1994) and piping stress and 

material information obtained from 

historical leak-before-break 

applications. These calculations 

indicated that extremely large 

circumferential flaws (i.e., greater than 

30 percent of the piping wall thickness 

for a flaw approximately 145 degrees 

around the piping circumference) would 

be required before failure would occur 

due to earthquakes with a return 

frequency of 10¥5 or 10¥6 per year. 

However, the plant-specific conditions 

used in the calculations were not 

chosen to bound conditions at all 

nuclear power plants. Additionally, 

some plants may have updated seismic 

hazard, piping stress, material property, 

or other information used in the flawed 

piping evaluation. Thus, the NUREG– 

1903 results may not be applicable to 

every plant. 

The ACRS, in its letter dated 

November 16, 2006 (ML063190465), 

also noted that seismic hazards are very 

plant specific. The ACRS further 

recommended that licensees who adopt 

§ 50.46a should demonstrate that the 

results developed by the NRC bound the 


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likelihood of seismically induced failure 

at their plants. The Committee further 

stated that licensees may have to 

perform additional calculations to 

demonstrate a comparable robustness of 

flawed piping. The ACRS 

recommendations are consistent with 

the limitations of the NUREG–1903 

study as noted above. 

It would also be inconsistent with the 

Commission’s intent to allow the 

relaxation of ECCS requirements at a 

plant with a seismically induced large 

break LOCA frequency greater than the 

10¥5 per reactor year criteria used for 

selecting the TBS in the proposed rule. 

Because seismic analyses and, in 

particular, indirect failure estimates are 

highly plant and site specific (as noted 

in NUREG–1903 and in ACRS 

comments), the NRC believes that it is 

necessary for a licensee to demonstrate 

that its seismic LOCA frequency is 

sufficiently low before implementation 

of the alternative ECCS requirements. 

Depending upon the results of the plant 

specific assessment, it may be necessary 

to implement augmented in-service 

inspection plans. As discussed below in 

Section V.C. of this document, the NRC 

is currently preparing guidance for 

conducting these plant-specific 

assessments (‘‘Plant-Specific 

Applicability of 10 CFR 50.46 Technical 

Basis,’’ February 2009; ML090350757). 

C. Comments on Thermal-Hydraulic 

Analysis 

Comment. Both NEI and WOG 

recommended that the proposed new 

reporting requirement in 

§ 50.46a(g)(1)(i) of a 0.4 percent change 

in oxidation as the threshold for 

reporting a change, or the sum of 

changes, in calculated clad oxidation be 

changed from 0.4 percent to 2.0 percent. 

WOG noted that the rationale for 

selecting 0.4 percent is that it is the 

same, on a percentage basis, as the 

existing peak cladding temperature 

(PCT) change reporting requirement. 

WOG also stated that this rationale is 

only true if one considers the range of 

interest of PCT as 0 to 2200 degrees 

Fahrenheit (°F) [(50 °F/2200 °F) × (17 

percent) = 0.4 percent]. If instead, one 

considers the range of interest of PCT as 

1700–2200 °F or 1800–2200 °F, from the 

perspective of transient oxide build-up, 

this same rationale gives a significance 

threshold of 1.7 or 2.1 percent. On this 

basis, WOG recommended that the 

significance threshold for changes in 

oxidation be revised to 2.0 percent. 

WOG also noted that changes in 

oxidation are much more difficult to 

estimate than changes in peak cladding 

temperature because oxidation is an 

integrated parameter based on the 

temperature transient versus time, 

whereas PCT is a point value. If the 

significance threshold for oxidation is 

not adjusted as recommended above, it 

is anticipated that the new oxidation 

reporting requirement will require more 

frequent re-analyses than the current 

regulations require, with no 

commensurate benefit to the public 

health and safety. 

NRC response. The basis for the 0.4 

per year oxidation change is that the 

ratio of the reporting threshold value to 

the change in oxidation from a 

‘‘normal’’ operating level of 4 percent 

(based on a twice-burned oxidation 

thickness of 65 μ for Zircalloy-4) to a 

maximum level of 17 percent should be 

the same as the ratio of the reporting 

threshold value to the change from the 

normal operating cladding temperature 

of 600 °F to the allowed PCT of 2200 °F. 

On that basis the oxidation change of 

0.4 percent was chosen. The trend 

toward thinner cladding material raises 

the initial oxidation percentage even 

closer to the maximum local oxidation 

limit and reduces the margin for change 

in predicted oxidation. 

Additionally, the NRC agrees with the 

WOG comment that calculating 

oxidation is more time-consuming than 

calculating PCT. However, the NRC 

believes WOG is incorrect in stating that 

not reducing the significance threshold 

for reporting changes in calculated 

oxidation will cause the need for 

performing additional oxidation 

calculations. The significance threshold 

for reporting to the NRC only affects the 

frequency of reporting and has no effect 

on the need to do reanalysis. Reanalysis 

is necessary when licensees discover 

errors or make changes to analytical 

codes. 

The Commission has directed the 

NRC staff to revise the ECCS acceptance 

criteria in § 50.46(b) to account for new 

experimental data on cladding ductility 

and to allow for the use of advanced 

cladding alloys. The NRC will soon 

issue an Advance Notice of Proposed 

Rulemaking (ANPR) seeking public 

comments on a planned regulatory 

approach. The NRC expects that this 

rulemaking (Docket ID NRC–2008–0332) 

will establish new cladding 

embrittlement acceptance criteria in 

§ 50.46(b) for design basis LOCAs. As 

these new acceptance criteria are being 

established, the NRC will also make 

conforming changes to § 50.46a as 

necessary for both below and above TBS 

breaks. As a consequence, the NRC now 

believes that the need for a reporting 

requirement in § 50.46a associated with 

errors or changes in ECCS analysis 

methodology would be more 

appropriately addressed in the ongoing 

§ 50.46(b) proceeding. Accordingly, the 

changes to the oxidation reporting 

requirements in the initial proposed 

rule have been removed from the 

revised proposed rule. 

Comment. Framatome commented 

that the analysis or case requirements in 

§ 50.46a(e)(2) for beyond the transition 

break size evaluations are excessive. 

The desire for this portion of the 

regulation is to establish in a reasonable 

way that the plant remains able to 

mitigate a large break LOCA. It is 

unnecessary and inconsistent to elevate 

the consideration of break size effects 

beyond that of other portions or aspects 

of the evaluation that are to be treated 

as reasonable values. Under the 

proposed rule language, a full § 50.46 

evaluation will be required for breaks of 

area less than the TBS. The results for 

these analyses can be extended to the 

smaller break sizes in the greater than 

TBS spectrum with assurance. 

Combining a reasonable selection of 

discharge coefficient (0.6) with the use 

of the 1994 ANS decay heat standard 

would roughly equate a 14-inch 

schedule 160 pipe area (0.7 ft 2), treated 

as below the TBS, with a 1.4 ft 2 break, 

treated as a beyond TBS break. 

Similarly, at the upper end of the break 

spectrum, what used to be considered as 

an 8 to 9 ft 2 break of the cold leg will 

be the equivalent of a historical 5 ft 2

break. The requirement to perform 

sensitivity studies to identify a worst 

case break between these two limits 

seems unwarranted. It would be 

reasonable to just perform the full 

double area break or at most that break 

and one intermediate break. The only 

sensitivity required should be relative to 

break location. Historically, break 

location can have a substantial 

influence on the calculated results. This 

should be resolved prior to the greater 

than TBS calculation either by 

sensitivity studies or by reference to 

appropriate historical analyses. The 

concern can be allayed by either altering 

the rule so that the identification of the 

most severe break size is not required or 

by inserting the concept of reasonable 

confidence that breaks within the 

beyond TBS spectrum will not pose 

consequences substantially more severe 

than those of the calculations 

performed. 

The WOG stated that for NRCapproved best-estimate or Appendix K 

evaluation models, the requirement for 

analyzing a spectrum of break sizes is 

unwarranted. The BWROG said that the 

requirement to re-validate over 30 years 

of experience with performing large 

break LOCA analysis to confirm ‘‘for a 

number of postulated LOCAs of 

different sizes and locations * * * that 


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the most severe postulated LOCAs 

* * * are analyzed’’ is unnecessarily 

burdensome and appears to serve no 

specific technical need. Current bestestimate large break LOCA models, 

which are benchmarked to testing data, 

have yielded no insights that would 

invalidate the previous analytical 

experience and knowledge. WOG 

concluded that this provision in the rule 

language should be removed. 

NRC response. The NRC disagrees 

with the commenters on the need for 

analyzing a spectrum of break sizes. The 

proposed rule language was selected 

because there are two peak cladding 

temperatures, one that occurs below the 

TBS and one that occurs above the TBS. 

The peak above the TBS may not occur 

for the DEGB, but rather, for a break area 

in the range of 0.6 to 0.8 times the DEGB 

area. Because there can be a fairly large 

temperature difference between that 

break and the DEGB, use of the DEGB 

could be non-conservative. The NRC 

also believes that the language of the 

rule provides considerable flexibility in 

implementation (relative to the stated 

comments) because the requirement is 

to analyze a ‘‘number of postulated 

LOCAs * * * sufficient to provide 

assurance that the most severe LOCAs 

* * * are analyzed’’. The use of 

historical analyses is not precluded. No 

changes were made in the revised 

proposed rule. 

Comment. NEI commented that in 

§ 50.46a(e)(2) on ECCS analysis 

methods, one requirement is that 

‘‘comparisons to applicable 

experimental data must be made.’’ NEI 

stated that other approaches such as 

comparison of results to accepted 

analysis techniques or to textbook 

approaches are also appropriate and 

suggested that the requirement be 

reworded to state that ‘‘sufficient 

justification’’ must be provided. 

NRC response. The NRC disagrees 

with this commenter. Computer code-tocode comparisons are not adequate 

because all codes have uncertainty in 

their results. Only code-to-data 

comparisons can be used to accurately 

assess code uncertainties. Similarly, 

computer code results cannot be 

validated by comparison to ‘‘textbook 

approaches’’ because no simple 

textbook approaches exist for modeling 

the highly complex thermal-hydraulic 

phenomena associated with pipe break 

analyses. No changes were made in the 

revised proposed rule. 

Comment. WOG submitted four 

options for how to perform ECCS 

analysis in the beyond-TBS region to 

assist the NRC staff in developing the 

regulatory guide for implementing the 

§ 50.46a rule. 

NRC Response. The NRC will evaluate 

the WOG ECCS analysis options and 

will provide additional implementation 

guidance in the associated regulatory 

guide. 

Comment. The BWROG stated that it 

supports applying the requirements of 

§ 50.46a(b)(1) to reactors with MOX 

[mixed oxide] fuel. 

NRC response. The proposed § 50.46a 

is intended to be an alternative to the 

current ECCS requirements in § 50.46. 

Because § 50.46 does not address the 

use of mixed oxide fuel, the NRC 

believes that the commenter’s proposal 

is beyond the scope of this rulemaking. 

The NRC did not make changes in the 

revised proposed rule to address MOX 

fuel. 

Comment. Proposed § 50.46a(e)(2): 

The following sentence should be 

moved from its current location to just 

in front of the sentence beginning, 

‘‘These calculations * * *’’: ‘‘The 

evaluation must be performed for a 

number of postulated LOCAs of 

different sizes and locations sufficient to 

provide assurance that the most severe 

postulated LOCAs larger than the TBS 

up to the double-ended rupture of the 

largest pipe in the reactor coolant 

system are analyzed.’’ This relocated 

sentence should begin a new paragraph. 

These changes will properly group the 

more detailed analysis requirements. 

NRC response. The NRC agrees that 

movement of the noted sentence 

improves the rule presentation. In the 

revised proposed rule, this sentence has 

been relocated as the commenter 

suggested, but the structure of 

§ 50.46a(e)(2) was not modified. 

Comment. In proposed § 50.46a(e)(2), 

the NRC should clarify the requirements 

for licensee documentation to be 

maintained onsite versus generic 

documentation in or supporting a 

licensing topical report. 

NRC response. In the revised 

proposed rule, the NRC modified 

§ 50.46a(e) to require that analysis 

methods for all LOCAs ‘‘must be 

approved for use by the NRC. Appendix 

K, Part II, to 10 CFR Part 50, sets forth 

the documentation requirements for 

evaluation models.’’ Thus, the 

documentation requirements for 

analysis methods used for breaks larger 

than the TBS are the same as for 

analysis methods used for breaks 

smaller than the TBS. The purpose of 

this change is to increase confidence in 

the ability to mitigate breaks greater 

than the TBS, as recommended by the 

Advisory Committee on Reactor 

Safeguards. 

Comment. In proposed § 50.46a(e)(2), 

the NRC states that these calculations 

[for breaks larger than the TBS] may 

take credit for the availability of offsite 

power and do not require the 

assumption of a single failure. It should 

also be noted that availability of 

equipment is not limited to safetyrelated equipment. 

NRC response. The NRC agrees that 

the suggested language is more 

descriptive and has incorporated the 

change into that last sentence of 

§ 50.46a(e)(2). 

Comment. For PWR LOCAs below 

and above the TBS, the mitigating 

systems and equipment are the same for 

the full spectrum of LOCAs. Although 

non-safety LOCA mitigation systems/ 

components may be applicable in the 

context of BWR LOCA analysis, this is 

not the case for PWRs. If this element of 

the proposed regulation (allowing the 

use of non-safety grade systems) is 

intended to address a situation that is 

only applicable to BWRs, then it should 

not be required for PWRs. 

NRC response. The element of the 

proposed regulation—allowing the use 

of non-safety grade systems—noted by 

the commenter is not intended to 

address a situation that is only 

applicable to BWRs. Although PWR 

plants may not currently have nonsafety systems that could be credited for 

LOCA mitigation (for breaks larger than 

the TBS), modifications could be made 

in the future that facilitate use of nonsafety systems. The revised proposed 

rule would relax existing § 50.46 

requirements to allow ECCS analyses of 

breaks larger than the TBS to take credit 

for both safety-grade and non-safetygrade equipment if such equipment 

exists, is maintained available and 

reliable, and is capable of being 

powered by an on-site source of 

electrical power. 

Comment. The WOG commented that 

the rule should not contain a 

requirement for licensees to submit 

beyond TBS thermal-hydraulic analyses 

to the NRC for approval. One reactor 

licensee commented that the proposed 

rule states that licensees will not be 

required to submit their beyond-TBS 

analysis method or application to the 

NRC for review and approval; instead, 

the NRC intends to maintain regulatory 

oversight of these analyses by 

inspection. That licensee said that 

although not requiring NRC review and 

approval has the appearance of a benefit 

to the licensees, it actually introduces a 

risk of a regulatory crisis should an 

inspection identify a deficiency in the 

beyond-TBS analysis method following 

implementation. Such an identified 

deficiency could result in a 

consequence such as the regulator 

imposing restrictions on reactor 

operation. This risk is greater than for 


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6The scope of changes subject to the change 

criteria in § 50.46a(f) of the proposed rule would be 

greater than the changes currently subject to 

§ 50.59, which applies only to changes to ‘‘the 

facility as described in the FSAR.’’ The change 

criteria in the proposed rule would apply to all 

facility and procedure changes, regardless of 

whether they are described in the FSAR. 

7Requirements for license amendments are 

specified in §§ 50.90, 50.91 and 50.92. They include 

public notice of all amendment requests in the 

Federal Register and an opportunity for affected 

persons to request a hearing. In implementing 

license amendments, the NRC typically prepares an 

appropriate environmental analysis and a detailed 

NRC technical evaluation to ensure that the facility 

will continue to provide adequate protection of 

public health and safety and common defense and 

security after the amendment is implemented. 

the current situation where LOCA 

evaluation models and applications are 

pre-approved by the NRC. It would be 

preferable that NRC review and 

approval of § 50.46a applications be 

obtained prior to implementation to 

avoid such a regulatory crisis. This 

commenter proposed that the NRC agree 

to perform a pre-approval of a licensee’s 

beyond-TBS analysis method and 

application if requested by a licensee. 

NRC response. The NRC has changed 

the proposed rule to require NRC review 

and approval of analysis methods used 

to evaluate plant response to LOCAs 

larger than the transition break size. The 

purpose of this change is to increase 

confidence in the ability to mitigate 

breaks greater than the TBS, as 

recommended by the ACRS. 

Comment. NEI, a reactor vendor, and 

a reactor licensee requested that M5 

cladding (M5) be specified as an 

approved fuel cladding material in 

existing § 50.46(a) and in proposed 

§ 50.46a(b)(1) to avoid the need for 

requesting an exemption to allow its 

use. The reactor vendor stated that 

because M5 is currently being used in 

11 nuclear power reactors of varying 

designs across the United States, it is 

obvious that M5 is an acceptable and 

desirable cladding material. The 

BWROG stated that § 50.46a should be 

made available to reactors with alternate 

cladding materials. 

NRC response. As previously 

discussed, the Commission directed the 

NRC staff to initiate a separate 

rulemaking effort to amend § 50.46(b) to 

address the use of advanced cladding 

alloys. The NRC is considering cladding 

specific issues in that proceeding and 

will also incorporate appropriate 

conforming changes to § 50.46a. The 

NRC is working to revise the ECCS 

acceptance criteria in § 50.46(b) to 

account for new experimental data on 

cladding ductility and to facilitate the 

licensing review of advanced cladding 

alloys such as M5. The NRC plans to 

issue an ANPR during the summer of 

2009 to solicit public comments on a 

planned regulatory approach. In the 

interim, the NRC will continue to 

evaluate the use of cladding materials 

other than Zircalloy or ZIRLO on a caseby-case basis. 

D. Comments Related to Probabilistic 

Risk Assessment 

1. Summary 

The initial proposed rule required 

that all future facility changes 6 made by 

licensees after adopting § 50.46a be 

evaluated by a risk-informed integrated 

safety performance (RISP) assessment 

process that has been reviewed and 

approved by the NRC via the routine 

process for license amendments.7 (See 

70 FR 67612–67615.) Most of the 

commenters on the proposed rule stated 

that current regulatory processes that 

control changes to the facility are 

adequate and therefore, there is no need 

for the RISP change control process. In 

comments generally supported by all 

nuclear industry commenters, NEI 

argued that the controls on the existing 

licensing basis make it virtually 

impossible to make significant adverse 

changes to the risk profile of the plant 

without being required to submit a 

license amendment request for prior 

NRC review and approval. NEI 

concluded that the only item that might 

be missing from the current framework 

that would provide additional assurance 

that the licensee is appropriately 

maintaining the risk profile of the 

facility after adoption of § 50.46a would 

be a requirement that the licensee 

periodically assess the cumulative 

impact of facility changes to the risk 

profile. 

Industry commenters also considered 

the proposed rule’s unbounded scope of 

the facility changes requiring a RISP 

assessment to be an unnecessary burden 

and some argued that this requirement 

is potentially adverse to safety. In this 

regard, the commenters said that 

because most facility changes have no 

material safety significance, requiring a 

RISP assessment of facility changes 

beyond even the criteria established in 

current regulations, such as in § 50.59, 

would add a wide range of activities and 

components to the licensing basis that 

were never reviewed or ever intended to 

be reviewed by the NRC. Thus, licensees 

would be forced to divert valuable 

resources from monitoring plant safety 

to tracking a multitude of items that 

have no safety or risk significance. A 

few commenters recognized that most 

facility changes could be dispositioned 

with a qualitative RISP assessment but 

argued that there would still be cost 

associated with the performance and 

documentation of the assessment. 

All commenters stated that the rule 

should not include the operational 

restriction that all allowable at-power 

configurations be demonstrated to meet 

the ECCS acceptance criteria. The 

suggested alternatives ranged from 

reducing the restrictions and placing 

them under licensee control to 

eliminating them entirely. The 

commenters argued that: 

(1) Existing plant configuration 

control programs, including technical 

specifications and implementation of 

the maintenance rule, provide sufficient 

controls to ensure that implementation 

of § 50.46a will not lead to plant 

operation in high risk configurations; 

(2) Because of the low frequency of 

breaks greater than the TBS there should 

be a minimum of associated operating 

restrictions; 

(3) Any operating restrictions for 

breaks larger than the TBS need to be 

commensurate with risk contribution of 

these larger break sizes; and 

(4) Operating restrictions would 

remove or reduce any potential benefit 

that licensees might gain from the 

adoption of § 50.46a. 

NRC summary response. The NRC 

believes that a risk-informed change 

process is a necessary component of this 

rule because this rule would permit 

changes to facility design bases that 

would not be allowed under current 

regulations. The current regulatory 

processes that control facility changes 

are not adequate to control riskinformed plant changes that would be 

allowed under § 50.46a. However, the 

NRC has modified the risk-informed 

change process considerably by 

reducing the scope of facility changes 

for which a risk assessment is required. 

The NRC considered requiring all 

facility changes to be evaluated as risk 

informed changes and permitting 

licensees to make all facility changes, 

with some exceptions, that satisfy the 

criteria in § 50.59 or other NRC 

regulations without prior NRC review 

and approval. The ACRS commented 

that requiring the change in risk from all 

facility changes to be compared to the 

acceptable risk increase criteria was a 

significant departure from RG 1.174 

guidance and other past risk-informed 

applications. The ACRS recommended 

that this proposal be reviewed for its 

implications. 

Instead of requiring risk assessment of 

all future facility changes, the revised 

proposed rule would require risk 

assessments for only those facility 

changes enabled by the new ECCS 

requirements for pipe breaks greater 

than the TBS. This change would 


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reduce unnecessary burden and bring 

the change control process into 

conformance with RG 1.174 and other 

risk-informed rules and licensing 

actions. Two previous risk-informed 

regulations promulgated by the NRC 

(i.e., §§ 50.69 and 50.48(c)) have 

included similar requirements related to 

the use of PRA and risk-informed 

principles to demonstrate the 

acceptability of facility changes enabled 

by new, risk-informed regulations before 

being implemented by licensees. 

The revised proposed rule defines 

facility changes enabled by § 50.46a as 

changes to the facility, technical 

specifications, and procedures that 

satisfy the revised ECCS analysis 

requirements in § 50.46a but do not 

satisfy the ECCS analysis requirements 

in § 50.46. A risk-informed analysis, 

consistent with that described in RG 

1.174, shall be applied to facility 

changes enabled by the rule. The riskinformed framework established in RG 

1.174 permits licensees to propose 

several individual changes to a facility’s 

licensing basis that have been evaluated 

and will be implemented in an 

integrated fashion. Some facility 

changes proposed by licensees may not 

be enabled by the rule but may lead to 

a risk decrease. RG 1.174 permits 

integrated (bundled) changes in risk to 

be compared to the acceptance 

guidelines from RG 1.174 in order to 

encourage changes that reduce risk. The 

NRC has retained this guidance in 

§ 50.46a(f)(2)(iv) which would permit 

the change in risk from changes enabled 

by the rule to be combined with the 

change in risk from other plant changes 

unrelated to the rule for the purpose of 

demonstrating that the change in risk 

from all changes made under the rule 

meets the acceptance criteria. 

In addition to reducing the scope of 

facility changes to which the riskinformed change process must be 

applied, the NRC has discarded the 

acronym ‘‘RISP’’ in favor of the simpler 

‘‘risk-informed’’ label because the 

elements and processes described by the 

RISP are the elements and processes 

that make up a risk-informed 

evaluation. 

The NRC considered whether to 

simplify the risk-assessment process 

further by relying primarily on current 

regulations to identify which facility 

changes a licensee must submit for prior 

NRC review and approval. The ACRS 

commented that the NRC should use 

risk criteria to determine whether a 

licensee should submit a change 

enabled by the rule for review and 

approval. Subsequently, the NRC 

retained the criteria specifying the 

maximum risk increase for a change that 

a licensee may make without prior NRC 

review and approval. This requirement 

frees licensees and the NRC from the 

burden of evaluating and accounting for 

the many individual facility changes 

that do not have a significant impact on 

risk while retaining NRC review and 

approval for changes that might pose a 

safety concern. 

In response to comments received on 

the operational restrictions in the 

proposed rule, the NRC has decided that 

restrictions must remain on plant 

operation in configurations where it has 

not been demonstrated that breaks larger 

than the TBS can be mitigated, but the 

restrictions will be modified. The 

proposed rule prohibited at-power 

operation in any configuration without 

the demonstrated ability to mitigate a 

LOCA larger than the TBS. The revised 

proposed rule would restrict at-power 

operation in such a configuration to not 

exceed a total of fourteen days in any 12 

month period. Rather than requiring 

licensees to use risk methods to 

determine how long such operation 

would be permitted, what actions would 

be required, and how the controls 

would be implemented, in the 

republished proposed rule the NRC is 

specifying a time limit that simplifies 

implementation without sacrificing 

flexibility and introducing unnecessary 

burden. The NRC believes it is unlikely 

that licensees would experience 

circumstances when they would 

consider operating in such a condition 

for more than fourteen days but feels 

that maintaining the restriction is 

necessary. 

Although the LOCA frequencies on 

which the TBS are founded indicate that 

the expected frequency of breaks larger 

than the TBS is low, these frequencies 

are estimates derived from an expert 

elicitation process. The NRC has 

addressed the associated uncertainty, in 

part, by incorporating other elements 

into the selection of the TBS while 

recognizing that facility changes 

permitted by the rule could reduce the 

capability to mitigate some accidents 

that would currently be mitigated. The 

NRC concluded that the consequence of 

a challenge to the facility from an 

unmitigated break larger than the TBS is 

severe enough to warrant some 

confidence that the break could be 

mitigated. 

Although the NRC currently has no 

guidance explicitly applicable to 

determine an acceptable time interval 

for operation without mitigation 

capability for a beyond-TBS LOCA, 

some related guidance is available. 

Previously, the NRC determined that 

events having at least a 10¥7 probability 

per year should generally be taken into 

consideration in facility design. This 

approach is reflected in NUREG–0800, 

‘‘Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear 

Power Plants.’’ Events taken into 

consideration in facility design are 

design basis events and must meet the 

regulations specifying the required 

ability to mitigate the event. This 

guideline indicates that events with a 

frequency less than 10¥7 per year need 

not be considered in facility design. 

Applying this criterion to develop an 

acceptable time interval during which a 

beyond-TBS LOCA might not be 

successfully mitigated yields about 4 

days per year. Regulatory Guide 1.177, 

‘‘An Approach for Plant-Specific RiskInformed Decisionmaking; Technical 

Specifications,’’ provides risk guidelines 

that are routinely used to judge the 

acceptability of time intervals that 

safety-related equipment can be 

unavailable. Applying the RG 1.177 

criterion yields about 18 days. Neither 

of these guidelines is fully applicable to 

this configuration. The 10¥7 annual 

probability was developed to identify 

events external to the plant that need 

not be included in the design basis and 

is not specifically applicable to internal 

events such as LOCAs. Regulatory 

Guide 1.177 guidelines are normally 

applied to an operating configuration 

when mitigation capability would still 

be available although a single failure 

might fail that capability. Nevertheless, 

they provide an indication that an 

acceptable period of time should be 

measured in days. 

The NRC chose fourteen days as the 

appropriate limit on how long a plant 

can operate in a configuration not 

demonstrated to meet the ECCS 

acceptance criteria for LOCA break sizes 

larger than the TBS. The NRC believes 

that fourteen days should be sufficient 

to allow completion of on-line 

maintenance activities relied on to 

ensure high reliability for safety systems 

while providing adequate protection of 

public health and safety, consistent with 

the low frequency of these LOCAs. The 

NRC believes that a longer time period 

for operating in such a plant condition 

would not be consistent with its stated 

goal of retaining the ability to 

successfully mitigate the full spectrum 

of LOCAs and would not adequately 

address uncertainties in the evaluation 

used to select the TBS. Conversely, a 

shorter time period could lead to 

significant burden to the industry with 

no clear safety benefits and, if 

maintenance activities were adversely 

affected, a possible reduction in safety. 

Therefore, the NRC will limit the 

allowed time period for operation in an 


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unanalyzed condition to fourteen days 

to ensure that mitigation capability is 

maintained except for occasional, brief 

periods necessary to perform online 

maintenance of mitigation structures, 

systems and components. 

The NRC concludes that the fourteen 

day operational restriction would 

protect public health and safety, provide 

adequate time for licensees to perform 

beneficial maintenance activities, be 

commensurate with the safety 

significance of LOCAs with a break size 

larger than the TBS and be consistent 

with the Commission’s intent that 

mitigation capability be retained for the 

full spectrum of LOCA events 

‘‘commensurate with the safety 

significance of these capabilities.’’ 

The NRC agrees with commenters that 

operational restrictions could reduce the 

benefits that may be derived from 

adopting § 50.46a, but the NRC believes 

that this reduction in benefits is 

necessary and prudent to ensure that 

some capability to successfully mitigate 

LOCAs larger than the TBS is retained 

consistent with the risk of these events. 

As an example, because the new 

§ 50.46a ECCS analysis requirements 

provide relief from the single failure 

criterion for pipe breaks larger than the 

TBS, they could permit a facility to 

increase power to the extent that flow 

from both low pressure safety injection 

trains would be required to fully 

mitigate beyond-TBS breaks. However, 

the operational restriction in the renoticed proposed rule would require 

that such a facility reduce power to a 

level where injection from one train is 

sufficient to mitigate beyond-TBS breaks 

if the second train is inoperable or is 

removed from service for preventative 

maintenance for longer than fourteen 

days. The plant would be permitted to 

operate at the increased power level at 

all other times. 

2. Discussion of Specific Comments 

Comment. The RISP process would be 

an extreme regulatory burden on 

licensees and the NRC to implement. 

Five reactor licensees said they would 

not implement the proposed rule 

because of excessive burden. 

NRC response. The NRC disagrees 

with the commenters that the burden to 

develop and implement a risk-informed 

evaluation process as described in the 

initial proposed rule is an extreme 

regulatory burden because many 

elements of a risk-informed evaluation 

process should already exist at power 

reactors. However, as discussed above, 

the NRC has substantially reduced the 

scope of facility changes requiring a 

risk-informed evaluation. The revised 

proposed rule now would require a riskinformed evaluation as described in RG 

1.174 which is consistent with the riskinformed evaluations required by other 

risk-informed applications and 

regulations. The NRC believes that the 

burden associated with implementing a 

risk-informed evaluation program 

would be offset by the flexibility 

provided by the new ECCS analysis 

requirements that will permit facility 

changes that were not permitted by the 

previous ECCS analysis requirements. 

Comment. The risk-informed 

evaluation process emphasizes 

insignificant facility changes. The 

proposed change control requirements 

would require the NRC to be in the 

business of individually reviewing a 

myriad of insignificant facility changes. 

The risk acceptance criteria for allowing 

minimal risk changes appear to be 

contrary to the stated goal of enhancing 

safety. It seems illogical to adopt more 

restrictive requirements on safeguards 

for beyond design basis events than 

exist for design basis events. 

NRC response. The NRC disagrees 

that the proposed rule’s requirements 

would lead to the NRC individually 

reviewing insignificant facility changes. 

Facility changes that are enabled by the 

new ECCS requirements may lead to a 

wide range of estimated increases in 

risk, from immeasurably small to very 

large. The NRC has established an 

acceptance criterion that specifies the 

total amount of risk increase that would 

be considered acceptable from changes 

made under this rule. The revised 

proposed rule also includes a provision 

that prior NRC review is not required for 

individual facility changes that cause no 

more than a minimal increase in risk 

when compared to the overall plant risk 

profile. As discussed below, the NRC 

would consider any increase that is less 

than ten percent of the total acceptable 

risk increase to be minimal. The revised 

proposed rule includes these criteria to 

prevent NRC review of insignificant 

changes while retaining the capability to 

review facility changes that might pose 

a safety concern before implementation. 

Comment. The scope of the required 

PRA is excessive. One commenter stated 

that the PRA scope requirements of 

§ 50.46a(f)(4)(i) in the proposed rule 

appear excessive and should instead use 

text from NRC policy regarding PRA 

scope requirements relative to an 

application, i.e. ‘‘* * * the PRA scope 

is such that all operational modes and 

initiating events that could change the 

regulatory decision substantially are 

included in the model quantitatively.’’ 

Another commenter stated that 

requirements for PRA should not be 

prescribed in the rule. Standards and 

processes exist to establish requirements 

for PRA technical adequacy (e.g., RG 

1.174, RG 1.200, ASME PRA standard). 

A peer-reviewed internal events PRA 

that meets RG 1.200 should be sufficient 

for § 50.46a implementation. A final 

commenter stated that a requirement for 

shutdown PRAs is not appropriate 

because of the low risk associated with 

shutdown configurations at BWRs. 

Requirements for seismic PRAs are also 

inappropriate because these constitute a 

typically small fraction of the overall 

risk for most plants. 

NRC response. The NRC does not 

agree with commenters that the scope of 

the required PRA is excessive and has 

made no changes to the PRA 

requirements in the revised proposed 

rule. Further, the NRC believes that the 

proposed rule language regarding PRA 

scope requirements provided by one of 

the above commenters is consistent with 

the language in both the proposed and 

the revised proposed rules. Thus, the 

commenter’s text was not incorporated 

into the revised proposed rule. 

The required overall characteristics of 

the PRA (and the non-PRA risk 

assessment) are included in the rule 

because these characteristics have been 

determined to be necessary to support 

decision making and inclusion of the 

characteristics in the rule provides 

clarity and predictability. The revised 

proposed rule does not prescribe how it 

will be determined whether a licensee’s 

risk-assessment complies with these 

characteristics. The process to evaluate 

the suitability of each licensees’ risk 

assessment will be described in the 

regulatory guide associated with this 

rule. This process will include staffendorsed industry standards and the 

peer review process currently used by 

the NRC to evaluate the technical 

adequacy of PRAs supporting license 

amendment requests. 

Comment. The requirement to update 

the PRA at a frequency no less often 

than once every two refueling cycles is 

potentially burdensome. An alternative 

would be to require that after every 

second refueling cycle, that the need for 

a PRA update is assessed and that 

appropriate action be initiated. 

NRC response. The commenter’s 

suggestion that the need for a PRA 

update be first assessed and appropriate 

action then be taken is consistent with 

the revised proposed rule. Section 

50.46a(f)(2)(iv) would require that the 

PRA reasonably represent the current 

configuration of the plant. If a PRA 

continues to reasonably represent the 

configuration of the plant after a 

periodic review, the update requirement 

could be satisfied with a simple 

conclusion that changes to the PRA are 

not needed. The NRC believes that an 


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update interval no longer than two 

operating cycles is not unduly 

burdensome; thus, the PRA update 

periodicity was not changed in the 

revised proposed rule. 

Comment. The description of the riskinformed process should not be 

included in the application for a license 

amendment to implement § 50.46a. NEI 

provided complete alternative rule 

language in its comments. At the June 

28, 2006, public meeting to clarify the 

comments, NEI emphasized that the 

proposed rule provided in their 

comments did not require that the RISP 

process be submitted for review because 

they felt that such a review was 

unnecessary. Although this comment 

was not formally submitted, several 

other participants at the June 2006 

public meeting agreed with this 

comment. 

NRC response. The NRC disagrees 

with the comment that a description of 

a licensee’s risk-informed assessment 

process need not be submitted for NRC 

review as part of the licensee’s 

application to adopt § 50.46a. However, 

the NRC believes that the amount and 

complexity of the process description 

that must be submitted will vary 

appropriately depending on which, and 

how many, facility changes enabled by 

the rule a licensee chooses to make. 

As discussed, the NRC has revised the 

proposed rule by reducing the 

requirement that all future facility 

changes be evaluated using a riskinformed evaluation to only requiring 

that facility changes enabled by the rule 

be evaluated. Licensees who make 

limited facility changes under the rule, 

may chose to not submit a request to 

make future facility changes enabled by 

the rule without prior NRC approval as 

would be permitted in paragraph 

(c)(1)(iv). Licensees who make one or 

more risk-informed submittals without 

requesting the authority permitted 

under § 50.46a(c)(1)(iv) would only 

need to demonstrate that the process 

used to evaluate the specific change(s) 

described in each submittal provides 

confidence that the requirements of 

§ 50.46a(f)(2) are satisfied. The content 

of these submittals is expected to be 

similar to, and consistent with, riskinformed license amendment requests 

currently accepted for review by the 

NRC. 

A licensee requesting authority to 

make future changes without NRC 

review as permitted by § 50.46a(c)(1)(iv) 

must submit for NRC review and 

approval additional information, i.e., 

the licensee’s process including its risk 

assessment models and methods that 

will be used for making future riskinformed changes. Section 

50.46a(c)(3)(iii) provides that the NRC 

may approve an application if, in part, 

the licensee’s risk-informed evaluation 

process is adequate for determining 

whether the acceptance criteria in 

§ 50.46a(f) have been met. As described 

in RG 1.174, the technical acceptability 

of a PRA should be commensurate with 

the application for which it is intended; 

the level of detail required of the PRA 

should be sufficient to model the impact 

of the proposed change; and the effects 

of the changes should be appropriately 

accounted for. A licensee’s submittal to 

make future changes must provide 

sufficient information on both the risk 

assessment models and how future 

changes will be reflected in these 

models, to allow the NRC to conclude 

that the requirement in § 50.46a(c)(3)(iii) 

is met. 

Comment. Requirements on late 

containment failure should be removed. 

It is inappropriate to require licensees to 

retain a level of mitigation for late 

containment failure and late 

radiological releases, because these 

releases constitute a very small fraction 

of overall plant risk. Therefore, these 

references should be removed. 

NRC response. The NRC is proposing 

changes in the revised proposed rule 

that would make this topic moot. The 

commenter was remarking on the 

parenthetical ‘‘(early and late)’’ that was 

added to the containment related 

defense in depth element described in 

RG 1.174 when three of the elements 

were incorporated as acceptance criteria 

in the proposed rule. The NRC has 

removed the defense-in-depth 

acceptance criteria in the revised 

proposed rule, including the reference 

to early and late containment failures, 

but has retained the general criterion 

that defense-in-depth be maintained. 

The NRC will continue to follow the 

guidelines in RG 1.174 to address 

defense-in-depth when evaluating 

whether a licensee has satisfied the rule 

criterion that defense-in-depth has been 

maintained. The RG 1.174 guidelines for 

defense-in-depth in risk-informed 

applications have been used 

successfully by the NRC for more than 

a decade and do not need further 

clarification through rulemaking. 

Retaining the defense-in-depth 

guidelines in a regulatory guide instead 

of promulgating acceptance criteria in 

the rule would also allow the NRC to 

more effectively update its guidance as 

new information becomes available or if 

the Commission changes its policy. 

Comment. Section 50.46a(f)(4) 

contradicts § 50.46a(f)(5). One 

commenter stated that § 50.46a(f)(4) 

implies that only a PRA meeting the 

requirements of the following 

paragraphs may be used in the riskinformed assessment. This was seen as 

contradictory to § 50.46a(f)(5), which 

allows non-PRA risk assessment 

methods. 

NRC response. The NRC disagrees 

that the rule language is contradictory. 

The relevant phrase in § 50.46a(f)(4) 

states that ‘‘* * * to the extent that a 

PRA is used in the risk-informed 

assessment, it must * * *,’’ meet the 

following PRA requirements. If a PRA 

need not be used according to 

§ 50.46a(f)(1)(i) and (f)(2)(ii), and a PRA 

is not used, then non-PRA risk 

assessment methods that satisfy the 

requirements in § 50.46a(f)(5) may be 

used. No changes were made in the 

revised proposed rule. 

Comment. Performance monitoring is 

already covered by Appendix B to Part 

50. One commenter stated that the 

proposed requirement for a monitoring 

program designed to detect and prevent 

degradation of systems, structures, and 

components (SSCs) before plant safety is 

compromised is unnecessary. The 

commenter stated that 10 CFR Part 50, 

Appendix B, Criterion XVI for corrective 

action already contains this 

requirement. 

NRC response. The NRC does not 

agree. Appendix B to 10 CFR Part 50 

applies to safety-related SSCs and 

activities. The risk-informed decision 

process includes risk models that 

consider a much broader set of 

accidents and can credit a larger set of 

equipment and actions to mitigate these 

accidents than the set of safety-related 

equipment or actions. The NRC believes 

that performance measurement is an 

important part of risk-informed decision 

making that must be applied 

irrespective of the classification of an 

SSC or activity as ‘‘safety-related.’’ The 

performance monitoring requirement 

remains in the revised proposed rule. 

Comment. Power uprates and 

relaxation of the single failure criteria 

for breaks larger than a TBS LOCA 

could result in a situation when all 

emergency power supplies are needed 

to successfully mitigate a break larger 

than the TBS when accompanied by a 

loss-of-offsite power. The potential 

consequences of relying on the 

availability of offsite power supply in a 

deregulated environment or a 

requirement to have both divisions of 

onsite power available (without single 

failure capability) to mitigate the 

uprated reactor accident would not 

appear to be offset by any compensatory 

factors. 

NRC response. The NRC agrees that 

licensees who adopt § 50.46a could 

potentially make changes to the facility 

such that all emergency onsite power 


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supplies were required to demonstrate 

successful mitigation of a break larger 

than the TBS when accompanied by a 

loss-of-offsite power. Such an operating 

configuration would not be permitted by 

the current regulations. Licensees who 

adopt § 50.46a would have the 

flexibility to make facility changes that 

would not normally be permitted by 

current ECCS regulations but must 

comply with all the requirements of 

§ 50.46a. One requirement is to 

demonstrate that all changes made 

under the rule meet the risk acceptance 

criteria in § 50.46a(f) before the facility 

change may be implemented. Another 

requirement is that the change in risk 

from all changes to the facility must be 

periodically assessed and steps must be 

taken if the result exceeds the 

acceptance criteria in § 50.46a(f)(2). If 

changes to the plant-specific emergency 

power configuration and/or grid 

reliability over time result in risk 

increases exceeding the acceptance 

criteria, the plant changes that would 

permit this operating configuration may 

not be implemented, or other steps must 

be taken to reduce overall facility risk. 

However, in response to the ACRS 

recommendation in the November 16, 

2006, letter from Graham Wallis to 

Chairman Dale E. Klein, 

(ML063190465), to increase the level of 

defense-in-depth provided by the rule 

for mitigating LOCAs larger than the 

TBS, the NRC has modified the revised 

proposed rule with respect to the 

availability of onsite electrical power. 

The NRC has added the requirement 

that all equipment needed to mitigate 

pipe breaks larger that the TBS must be 

designed so that onsite power can be 

provided to the equipment. Onsite 

power may be provided automatically or 

as the result of manual actions taken by 

facility staff within a time frame that 

provides mitigation of damage and 

accident consequences. Although the 

ECCS analyses for pipe breaks larger 

than the TBS may still assume the 

availability of offsite power, the 

availability of onsite power to the 

necessary equipment provides 

additional defense-in-depth for 

postulated large break accidents. 

E. Comments Related to the 

Applicability of the Backfit Rule 

Comment. Commenters stated that the 

proposed rule provision limiting the 

applicability of the backfit rule is 

unnecessary. These commenters stated 

that the rule requires maintaining a 

mitigation capability up to the largest 

LOCA, regardless of the size of the TBS. 

The NRC should either apply the backfit 

rule to future changes in the TBS, or 

define a set of criteria defining how and 

when the NRC would determine that the 

TBS is no longer acceptable. Licensees 

should be provided with a great deal of 

latitude on achieving compliance 

following any change in the TBS, with 

the goal being that risk requirements are 

achieved with a reasonable mix of 

prevention and mitigation. 

NRC response. The NRC disagrees, for 

the most part, with the comments on 

this question. Because the estimated low 

LOCA frequency and corresponding low 

risk of large LOCAs is necessary to 

maintain assurance of public health and 

safety with this risk-informed 

regulation, the NRC believes that the 

exclusion of TBS changes from the 

backfit rule must be maintained in case 

future changes in estimated LOCA 

frequency require changes to the TBS. 

With respect to a commenter’s 

argument about the continuing 

regulatory requirement for LOCA 

mitigative capability beyond the TBS, 

the NRC notes that even though 

mitigative capability is retained, the 

proposed beyond-TBS mitigative 

capability is reduced, as compared to 

the capability required under the 

current ECCS rule. In developing the 

proposed rule, the NRC recognized the 

open-ended nature of the backfit 

exclusion. The NRC attempted to 

develop criteria for assessing whether 

new information mandates a change to 

the TBS. Unfortunately, the NRC was 

unable to develop relatively clear 

criteria and it was concluded that 

adoption of generalized criteria for 

constraining the NRC in future changes 

to the TBS would not prove useful or 

practical. Thus, the proposed rule did 

not set forth proposed criteria for 

assessing whether new information 

mandates a change to the TBS. The NRC 

notes that no commenter suggested any 

criteria for assessing the need for, or 

desirability of, changes to the TBS based 

upon new information. 

The NRC agrees that the proposed 

amendment should provide licensees 

with substantial flexibility to determine 

the manner in which they would come 

back into compliance with applicable 

regulatory requirements following any 

future change in the TBS. Licensees 

who must take actions to come back into 

compliance need not return the plant to 

the precise conditions and 

circumstances in effect immediately 

before implementation of the § 50.46a 

regulation. Rather, licensees should be 

afforded the flexibility of deciding what 

actions to implement to comply with a 

revised TBS. Further, as one of the 

commenters suggests, the overall goal of 

any actions taken to restore compliance 

is to achieve a reasonable mix of 

prevention and mitigation. The NRC 

will consider making this clear in 

implementing guidance. For these 

reasons, the NRC has decided to adopt 

the exclusion of future TBS changes 

from the backfit rule by retaining the 

provisions of proposed §§ 50.46a(m) and 

50.109(b)(2) in the revised proposed 

rule. 

Comment. Proposed §§ 50.109(b)(2) 

and 50.46a(d)(5) should not be adopted, 

and any changes to the TBS should be 

accomplished by rulemaking, and 

evaluated under the backfit rule. 

Excluding future changes to the TBS 

from compliance with the backfit rule 

would defeat the goal of regulatory 

stability embodied in the backfit rule 

and may result in changes that are not 

cost-justified. 

NRC response. The NRC disagrees 

with the comment that the NRC’s three 

reasons for excepting TBS changes and 

any consequent licensee reanalyses and 

changes from the backfit rule do not 

address how the objectives of the backfit 

rule are met. On the contrary, the NRC’s 

first reason (consideration of costs and 

benefits in a regulatory analysis) and the 

third reason (flexibility may reduce 

impacts of changes in the TBS) directly 

address the underlying objectives of the 

backfit rule. In addition, the second 

reason (application of the backfit rule 

favors incremental increases in risk) is 

relevant to the backfit rule’s ‘‘substantial 

increase in protection’’ criterion. A 

backfitting standard that limits increases 

in protection to public health and safety 

or common defense and security to 

those which are both substantial and 

cost-justified, but ignores (or allows) 

incremental decreases in protection 

without restriction does not seem to be 

a justifiable regulatory approach. Hence, 

the NRC believes that adoption of 

criteria to control these incremental 

decreases is justifiable and appropriate, 

even if inconsistent with the objective of 

regulatory stability, which is, arguably, 

the primary objective of the backfit rule. 

Finally, the NRC agrees that the goal 

of regulatory stability is not negated by 

the fact that a licensee’s decision to 

comply with § 50.46a rule would be 

optional or voluntary. On the contrary, 

the NRC believes that regulatory 

stability should be an important factor 

in developing a rule. However, the NRC 

disagrees with the commenter’s implicit 

assertion that, absent consideration 

under the backfit rule, regulatory 

stability would not be appropriately 

considered in any future revisions to the 

TBS. As the NRC stated in the statement 

of considerations in the proposed rule, 

a regulatory analysis would be required 

for any revision to the TBS. (See 70 FR 

67617–67618.) This regulatory tool 

provides an appropriate means of 


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ensuring that regulatory stability is 

considered by the NRC when 

determining whether to revise the TBS. 

Comment. The NRC should not adopt 

the backfitting exclusion provision in 

§ 50.46a(d), which would require that 

any facility changes made necessary by 

the maintenance and upgrading of risk 

assessments, would not be deemed to be 

backfitting. 

NRC response. The NRC disagrees 

with this comment, which was part of 

a broader comment opposing the 

proposed rule’s provision excluding 

from backfit consideration changes to a 

plant and its procedures that are 

necessitated by any future TBS changes 

mandated by the NRC (see the 

immediately-preceding comment 

analysis). The commenter did not 

provide a separate basis supporting its 

position that licensee changes 

necessitated by the periodic risk 

assessment maintenance and upgrading 

(as contrasted with NRC-mandated TBS 

changes) should be subject to backfitting 

consideration. 

The NRC believes that the policy and 

regulatory considerations with respect 

to backfitting of changes stemming from 

future TBS changes are irrelevant to the 

policy and regulatory considerations 

with respect to backfitting of changes 

required to maintain compliance with 

updated risk analyses. The NRC regards 

plant changes necessitated by periodic 

risk assessments under § 50.46a to be 

analogous (from a backfitting 

standpoint) to the 120-month updating 

of inservice inspection (ISI) and 

inservice testing (IST) under § 50.55a(f) 

and (g). Under those provisions, a 

licensee must update its ISI and IST 

program every 120 months to the latest 

version of the ASME Code in effect 12 

months before the beginning of the next 

inspection interval. The NRC has stated 

that the 120-month updating does not 

constitute backfitting, in part because 

the regulatory requirement for updating 

is known to the operating license 

applicant before it receives its license, 

which addresses the policy of regulatory 

stability and predictability embodied in 

the backfit rule. See 69 FR 58804, 58817 

(third column) (October 1, 2004); 67 FR 

60520, 60536–60537 (September 26, 

2002). This logic also applies to the 

periodic risk assessment maintenance 

and upgrading under § 50.46a(d)(4) and 

any necessary licensee actions necessary 

to maintain compliance with the 

relevant 50.46a acceptance criteria. The 

NRC also notes that § 50.46a does not 

prescribe any specific manner or 

approach for achieving compliance 

following the periodic risk assessment 

maintenance and upgrading under 

§ 50.46a(d)(4); this performance-based 

approach to regulation affords the 

licensee substantial flexibility and gives 

the licensee control over how best to 

achieve compliance. This further tends 

to reduce the impact of § 50.46a(d)(4) on 

licensees, which is an implicit objective 

of the backfit rule. For these reasons, the 

NRC declines to adopt the commenter’s 

recommendation. 

Comment. The fact that the proposed 

rule provides an alternative or voluntary 

approach for LOCA analysis does not 

negate either the backfit rule itself or the 

policy of regulatory stability. 

NRC response. The NRC disagrees 

with the comment. As discussed 

elsewhere in the backfitting discussion, 

the backfit rule’s protections apply only 

when the NRC is imposing (directly or 

indirectly) a change to the activities 

authorized by a license; it does not 

apply when the NRC is providing a 

regulatory approach as an alternative to 

compliance with an existing regulatory 

requirement. As a general matter, the 

regulatory stability and predictability 

afforded to a licensee by the backfit rule 

applies to the scope of activities 

approved by the license. If a licensee 

seeks a change to its licensing basis— 

which is what a transition to a voluntary 

alternative is—the licensee is seeking to 

do something that is not within the 

scope of activities authorized by its 

license. It is the NRC’s view that, in 

such a circumstance, the licensee has no 

reasonable expectation that the NRC’s 

criteria for judging the acceptability of 

that proposed change remains the same 

as the criteria used by the NRC in 

judging the original license application. 

Thus, the protections of the backfit rule 

do not apply either when a licensee 

seeks a voluntary change to its licensing 

basis, or when the NRC develops a 

voluntary alternative. 

Comment. The NRC set forth three 

justifications for excepting TBS changes 

from backfitting protection: the 

consideration of alternatives will occur 

in the required regulatory analysis; 

application of the backfitting rule 

effectively favors increases in risk; and 

the flexibility provided by the rule will 

tend to reduce the burden of any 

changes in the TBS. However, even if 

these justifications are true, they do not 

address how the objective of the backfit 

rule will be met or that this objective 

does not apply. 

NRC response. The NRC disagrees in 

part with this comment. The NRC views 

the backfit rule as having three 

underlying objectives: regulatory 

stability and predictability for a 

licensee; reasoned agency 

decisionmaking (that NRC’s decision to 

impose a backfit is assessed against 

rational criteria); and transparency of 

agency decisionmaking (that the reasons 

for the NRC’s determination on the 

overall backfiting criteria are publicly 

available). The second and third 

objectives would be met if the NRC 

imposes future TBS changes by 

rulemaking (which is by far the most 

likely course), inasmuch as such a 

rulemaking must include preparation of 

a regulatory analysis. A regulatory 

analysis which is performed in 

accordance with the NRC’s ‘‘Regulatory 

Analysis Guidelines’’, NUREG/BR– 

0058, Revision 4 (2004), provides for a 

disciplined agency decisionmaking 

process. The draft regulatory analysis is 

published and made available for public 

comment as part of the proposed rule. 

The final regulatory analysis, which 

addresses public comments, is also 

made available to the public as part of 

the final rulemaking. Hence, the NRC 

believes that the backfit rule’s objectives 

of reasoned decisionmaking and 

transparency of agency decisionmaking 

will be satisfied by any rulemaking 

changes to the TBS. With respect to the 

first objective of the backfit rule, the 

NRC recognizes that exclusion of future 

changes to the TBS from the backfit rule 

could lead to reduced regulatory 

stability and predictability because 

neither the adequate protection, 

compliance, or substantial safety 

increase criteria would be binding as 

checks against unwarranted agency 

action. However, the NRC believes that 

this is offset to some extent by two 

factors. First, by explicitly excluding 

future TBS changes and necessary 

changes from the backfit rule, licensees 

who choose to adopt § 50.46a are aware 

that the NRC may revise the TBS in the 

future (the argument here is similar to 

the Commission’s determination that 

the backfit rule does not apply to 

rulemakings endorsing more recent 

editions and addenda of the ASME Code 

for mandatory use in the 120-month 

interval process for ISI and IST in 

§§ 50.55a(f) and (g)). Second, the NRC 

acknowledges that plant-specific orders 

imposing TBS changes would not 

necessarily meet all of the backfit rule 

objectives. However, the NRC’s internal 

process governing the development and 

issuance of orders should, at minimum, 

result in reasoned decisionmaking. 

Moreover, as is the case with 

rulemaking changes to the TBS, 

regulatory predictability for changes to 

the TBS by order is addressed somewhat 

by explicitly stating in both §§ 50.109 

and 50.46a that the backfit rule does not 

apply if a revised TBS is imposed by 

order. These provisions provide notice 

to licensees considering adoption of 

§ 50.46a of the special backfitting 


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process under § 50.46a. Licensees 

contemplating adoption of § 50.46a may 

then factor this limited exclusion from 

the backfit rule into their decision 

whether to adopt § 50.46a. 

Comment. The Commission-proposed 

exclusion of TBS changes from 

backfitting protection would leave 

licensees who voluntarily adopt 

§ 50.46a without recourse to a backfit 

appeal process. 

NRC response. The NRC disagrees 

with the comment. Licensees who adopt 

§ 50.46a would continue to have access 

to the backfitting appeals process with 

respect to licensee-claims of backfit for 

all matters other than those attributable 

to TBS changes. 

Further, affected licensees would 

have an opportunity to raise concerns 

about the cost and expected benefits of 

proposed TBS changes, whether the 

TBS changes are imposed by rulemaking 

or by order. If the TBS were 

accomplished through rulemaking, all 

licensees would have an opportunity to 

comment on the proposed rule, 

including the associated regulatory 

analysis. By contrast, if the NRC 

imposes a TBS change by order, the 

affected licensee would have an 

opportunity to request a hearing on the 

order. During this hearing any issues 

could be raised on costs and benefits for 

the TBS change as applied to that 

licensee. Although these opportunities 

do not constitute, strictly speaking, a 

backfit appeal process, the NRC believes 

that they are the functional equivalent 

of a backfit appeal process. 

Finally, as noted earlier, it is the 

NRC’s expectation that should it 

mandate a change in the TBS, that 

licensees would have substantial 

discretion and flexibility with respect to 

how they would address that TBS 

change. Accordingly, the NRC sees no 

additional benefit from providing a 

licensee with a plant-specific backfitting 

appeal process related to TBS changes 

in addition to the public comment and 

hearing opportunities already provided 

for by law. 

F. Comments on Topics Requested by 

the Commission 

In the initial proposed rule, the NRC 

identified 16 significant topics 

associated with the proposal and invited 

the public to submit specific comments 

on those issues. (See 70 FR 6718— 

6719.) 

NRC Topic 1. In proposed § 50.46a(b), 

the NRC specifically precludes the 

application of the § 50.46a alternative 

requirements to future reactors. 

However, future light water reactors 

might benefit from § 50.46a. The NRC 

requests specific public comments 

regarding whether § 50.46a should be 

made available to future light water 

reactors. 

Comments. Framatome commented 

that § 50.46a should be available to 

nuclear power plants licensed after the 

publication of the rule that are of similar 

design to the current generation of 

operating BWRs and PWRs. Framatome 

stated that the advanced LWR designs 

previously certified (ABWR, System 

80+, AP 600, AP 1000), under design 

certification review (ESBWR) and in the 

pre-review process (US EPR), all fit into 

this category and can realize benefits 

from § 50.46a. However, for § 50.46a to 

apply to a new design, the NRC must 

first make a determination that the 

design is substantially similar to 

currently operating LWRs. The 

applicability to the new design of the 

frequency of pipe rupture versus break 

size curves used as a basis for 

establishing the TBS in § 50.46a must be 

established. The WOG stated that future 

PWRs and BWRs operating with 

materials, pressures and temperatures 

similar to operating LWRs should be 

able to use § 50.46a because there is no 

technical reason that new plants should 

have to meet outdated requirements for 

which existing plants can opt out. The 

BWROG and three other commenters 

also stated that § 50.46a should be made 

available to future light water reactors. 

NRC response. The NRC agrees with 

the commenters who stated that there 

are no technical reasons which prevent 

the new § 50.46a regulations from being 

applied to new light water reactor 

designs that are similar in nature (with 

respect to design and expected LOCA 

pipe break frequency) to current 

operating reactors. However, it would be 

difficult to apply the new regulation to 

certified reactor designs which have 

already received NRC approval. These 

design approvals were completed as 

rulemaking activities for the particular 

standardized design as of the date of the 

application, as amended. Changes may 

not be made to these designs unless the 

designers choose to resubmit the 

designs for reevaluation and reopen the 

design approval/rulemaking process to 

address § 50.46a. Moreover, it is not 

clear that these changes could be made 

under the special backfitting criteria in 

§ 52.63, because it does not appear that 

there is an issue related to adequate 

protection, compliance with 

requirements in effect at the time of 

certification, reduction of unnecessary 

burden, providing detailed design 

information, correcting material errors 

in the certification information, 

increasing standardization, or providing 

a substantial increase in overall safety, 

reliability, or security. 

Three new standardized LWR designs 

and one resubmitted LWR design are 

now being considered by the NRC. 

Although the NRC has not performed a 

detailed analysis of these new designs 

in the manner done for establishing the 

technical basis of this rule for existing 

designs, the frequency of large LOCAs at 

these facilities could be as low as it is 

at current LWRs. Thus, it may be 

appropriate to apply the alternative 

§ 50.46a requirements to these future 

designs. Accordingly, the revised 

proposed rule has been modified to 

apply to new reactor designs, e.g. 

facilities other than those which are 

currently licensed to operate. 

Applicants for design certification or 

combined licenses, holders of combined 

licenses under Part 52, or future 

licensees of operating new light-water 

reactors who wish to apply § 50.46a 

must submit an analysis for NRC 

approval, demonstrating why it would 

be appropriate to apply the alternative 

ECCS requirements and what the 

appropriate TBS would be for the new 

design to meet the intent of § 50.46a. 

In its analysis, the applicant, holder, 

or licensee must demonstrate that the 

proposed reactor facility is similar to 

reactors licensed before the effective 

date of the rule. In addressing similarity 

of the proposed reactor design to current 

reactor designs licensed before the 

effective date of the rule, the applicant, 

holder, or licensee would need to 

address design, construction and 

fabrication, and operational factors that 

include, but are not limited to: 

(1) The similarity of the piping 

materials of construction and 

construction techniques for new 

reactors to those in the currently 

operating fleet; 

(2) The similarity of service 

conditions and operational programs 

(e.g., in-service inspection and testing, 

leak detection, quality assurance etc.) 

for new reactors to those for operating 

plants; 

(3) The similarity of piping design, 

e.g. pipe sizes and pipe configuration, 

for new reactors to those found in 

operating plants; 

(4) Adherence to existing regulatory 

requirements, regulatory guidance, and 

industry programs related to mitigation 

and control of age-related degradation 

(e.g., aging management, fatigue 

monitoring, water chemistry, stress 

corrosion cracking mitigation etc.); and 

(5) Any plant-specific attributes that 

may increase LOCA frequencies 

compared to the generic results in 

NUREG–1829 and NUREG–1903. 

The analysis must also include a 

recommendation for an appropriate TBS 

and a justification that the 


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recommended TBS is consistent with 

the technical basis for this proposed 

rule. For new reactor designs that 

employ design features that effectively 

increase the break size, via opening of 

specially designed valves, to rapidly 

depressurize the reactor coolant system 

during any size loss of coolant accident, 

justification of the relevance of a TBS 

would be necessary. The methodology 

used to determine the proposed TBS 

should be described in the justification. 

Based on information currently 

available, new reactor designs may have 

similar piping materials, similar service 

conditions and operational programs, 

similar piping designs, and similar 

mitigation and control of age-related 

degradation programs to those found in 

currently operating plants. Therefore, 

based on information currently 

available, the NRC envisions that the 

TBS defined in the revised proposed 

rule could be applicable to the new 

reactor designs. 

In addition, a holder of an operating 

or combined license for a plant with a 

currently approved standard design 

could adopt § 50.46a if the design is 

demonstrated, by satisfying the five 

criteria above, to be similar to the 

designs of plants licensed before the 

effective date of the rule and the TBS 

proposed by the licensee is found 

acceptable by the NRC. 

In the revised proposed rule language 

and elsewhere in this document, 

whenever the NRC refers to similarity of 

the designs of new reactors to the 

designs of current operating reactors, 

the NRC intends for ‘‘design’’ to be 

broadly interpreted to encompass 

design, construction and fabrication, 

and operational factors that should be 

addressed, at a minimum, by 

considering the five similarity factors 

indentified above. 

NRC Topic 2. The TBS specified by 

the NRC in the proposed rule does not 

include an adjustment to address the 

effects of seismically-induced LOCAs. 

NRC is currently performing work to 

obtain better estimates of the likelihood 

of seismically-induced LOCAs larger 

than the TBS. By limiting the extent of 

degradation of reactor coolant system 

piping, the likelihood of seismicallyinduced LOCAs may not affect the basis 

for selecting the proposed TBS. 

However, if the results of the ongoing 

work indicate that seismic events could 

have a significant effect on overall 

LOCA frequencies, the NRC may need to 

develop a new TBS. To facilitate public 

comment on this issue, a report from 

this evaluation will be posted on the 

NRC rulemaking Web site at http:// 

ruleforum.llnl.gov before the end of the 

comment period. Stakeholders should 

periodically check the NRC rulemaking 

Web site for this information. [The NRC 

published the report on December 20, 

2005 (70 FR 75501; ML053470439).] The 

NRC requests specific public comments 

on the effects of pipe degradation on 

seismically-induced LOCA frequencies 

and the potential for affecting the 

selection of the TBS. The NRC also 

requests public comments on the results 

of the NRC evaluation that will be made 

available during the comment period. 

NRC response. Comments received on 

this topic were previously discussed in 

Section IV.B. of this document, 

‘‘Comments on Seismic Considerations 

Related to the TBS.’’ Because this topic 

was identified for public comment in 

the initial proposed rule, the NRC 

completed and published the study on 

the risks associated with seismically 

induced LOCAs larger than the TBS 

(NUREG–1903, ‘‘Seismic Considerations 

for the Transition Break Size’’ February 

2008; ML080880140). The NRC 

considered the public comments 

received on seismic considerations in 

the final version of NUREG–1903. As 

previously discussed in Section IV.B of 

this document, the NRC has concluded 

that no adjustment to the TBS is needed 

to account for seismically-induced 

LOCAs. 

NRC Topic 3. Depending on the 

outcome of an ongoing NRC study, the 

final rule could include requirements 

for licensees to perform plant-specific 

assessments of seismically-induced pipe 

breaks. These assessments would need 

to consider piping degradation that 

would not be prejudiced by 

implementation of the licensee’s 

inspection and repair programs. The 

assessments would have to demonstrate 

that reactor coolant system piping will 

withstand earthquakes such that the 

seismic contribution to the overall 

frequency of pipe breaks larger than the 

TBS is insignificant. The NRC requests 

specific public comments on this and 

any other potential options and 

approaches to address this issue. 

NRC response. After this topic was 

identified, the NRC completed and 

published the study on the risks 

associated with seismically-induced 

LOCAs larger than the TBS (NUREG– 

1903, ‘‘Seismic Considerations for the 

Transition Break Size’’ February 2008; 

ML080880140). Comments received on 

this topic were previously addressed in 

Section IV.B of this document, 

‘‘Comments on Seismic Considerations 

Related to the TBS.’’ The NRC has 

concluded that applicants wishing to 

implement the alternative ECCS 

requirements should conduct a plantspecific assessment of the risk 

associated with seismically-induced 

failures of flawed piping. The NRC is 

currently preparing guidance for 

conducting these plant-specific 

assessments (‘‘Plant-Specific 

Applicability of 10 CFR 50.46 Technical 

Basis’’ February 2009; ML090350757). 

NRC Topic 4. The ACRS noted that ‘‘a 

better quantitative understanding of the 

possible benefits of a smaller break size 

is needed before finalizing the selection 

of the transition break size.’’ The TBS to 

be included in the final rule should be 

selected to maximize the potential 

safety improvements. Thus, the NRC is 

soliciting comments on the relationship 

between the size of the TBS and 

potential safety improvements that 

might be made possible by reducing the 

maximum design-basis accident break 

size. 

NRC response. No comments were 

received which specifically addressed 

the relationship between the size of the 

TBS and potential safety improvements 

that might be made possible by reducing 

the maximum design-basis accident 

break size. However, the WOG stated, 

‘‘It is not appropriate to set the TBS on 

the basis of where the most benefit is, 

as this may change tomorrow and there 

will be no easy recourse.’’ This 

comment and other related issues were 

previously discussed in Section III.A of 

this document, ‘‘Comments on Selection 

of the TBS’’. The NRC made no changes 

to the size of the TBS in the revised 

proposed rule. 

NRC Topic 5. Proposed § 50.46a 

includes an integrated, risk-informed 

change process to allow for changes to 

the facility following reanalysis of 

beyond design basis LOCAs larger than 

the TBS. However, because the current 

regulations in 10 CFR part 50 already 

have requirements addressing changes 

to the facility (§§ 50.59 and 50.90), it 

might be more efficient to include the 

integrated, risk-informed change (RISP) 

requirements for plants that use § 50.46a 

under these existing change processes. 

The NRC solicits specific public 

comments on whether to revise existing 

§§ 50.59 and 50.90 to accommodate the 

requirements for making facility 

changes under § 50.46a. 

Comments. Three commenters 

responded directly to this question. One 

stated that §§ 50.59 and 50.90 should 

not be revised to accommodate the 

requirements for making plant changes 

under § 50.46a. Another stated that 

§ 50.59 requirements could be 

augmented to address the risk 

evaluations but that the augmentation 

was not necessary. The third commenter 

stated that §§ 50.59 and 50.90 should 

contain change requirements for 

§ 50.46a but that these requirements 


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should not be the RISP requirements 

included in the proposed rule. 

NRC response. The NRC is not 

changing §§ 50.59 and 50.90 to include 

integrated, risk-informed change 

requirements. The NRC has modified 

the risk-informed change control 

process to apply only to facility changes 

made under the rule, i.e., facility 

changes enabled by the rule as well as 

other facility changes unrelated to the 

rule but bundled together by the 

licensee for estimating the change in 

risk. Other facility changes would be 

unrelated insofar as the basis of the 

changes and NRC approval, when 

necessary, will rely on regulations, 

guidelines, or facility priorities that do 

not depend on the new TBS. The NRC 

changed the process to more closely 

follow the process described in RG 

1.174, which has been used successfully 

for a wide variety of risk-informed 

applications. The NRC has concluded 

that this risk-informed change control 

process can be used to successfully and 

safely implement facility changes 

enabled by the new TBS LOCA in the 

§ 50.46a final rule. 

NRC Topic 6. The proposed rule 

would rely on risk information. The 

NRC has included specifically 

applicable PRA quality and scope 

requirements in the proposed rule. 

However, there are other NRC 

regulations that also rely on risk 

information (e.g. the maintenance rule 

in § 50.65 and § 50.69 pertaining to 

alternative special treatment 

requirements). Consistent with the 

Commission policy on a phased 

approach to PRA quality, it might be 

more efficient and effective to describe 

PRA requirements (e.g., contents, scope, 

reporting, changes, etc.) in one location 

in the regulations so that the PRA 

requirements would be consistent 

among all regulations. The NRC is 

seeking specific public comments on 

whether it would be better to 

consolidate all PRA requirements into a 

single location in the regulations so that 

they were consistent for all applications 

or to locate them separately with the 

specific regulatory applications that 

they support. 

Comments. Five commenters 

recommended that it would be 

preferable to collect all PRA 

requirements in a single location in the 

regulations, but they all also stated that 

it would be premature to use the 

§ 50.46a rulemaking to combine PRA 

requirements at the present time. Some 

commenters argued that different 

applications have different requirements 

for the supporting PRA analyses and 

cautioned that PRA requirements 

should not be based on the most 

demanding application. 

NRC response. The NRC takes note of 

the recommendation that PRA 

requirements be eventually collected 

into a single location in the regulations. 

The NRC agrees that the § 50.46a 

rulemaking is not the appropriate 

vehicle to achieve this regulatory 

change. The NRC will include PRA 

requirements adequate to support this 

rulemaking in the § 50.46a rule. After 

the NRC develops broad-based PRA 

requirements suitable for use on a 

generic basis in different applications, 

the NRC will be able to codify these 

generic PRA requirements in a single 

regulatory location and could remove 

the § 50.46a specific PRA requirements 

(or limit them to existing licensees 

approved under § 50.46a to avoid 

backfitting). 

NRC Topic 7. Proposed § 50.46a 

would include the requirement that all 

allowable at-power operating 

configurations be included in the 

analysis of LOCAs larger than the TBS 

and demonstrated to meet the ECCS 

acceptance criteria. Historically, 

operational restrictions have not been 

contained in § 50.46 but were controlled 

through other requirements (e.g., 

technical specifications and 

maintenance rule requirements). It 

might be more practical to control the 

availability of equipment credited in the 

beyond design-basis LOCA analyses in a 

manner more consistent with other 

operational restrictions. As a result, the 

NRC is soliciting public comments on 

the most effective means for 

implementing appropriate operational 

restrictions and controlling equipment 

availability to ensure that ECCS 

acceptance criteria are continually met 

for beyond design-basis LOCAs. 

Comment. As previously discussed, 

all commenters stated that the NRC 

should not include the operational 

restriction that all allowable at-power 

operating configurations be 

demonstrated to meet the ECCS 

acceptance criteria. Several commenters 

proposed alternatives ranging from 

placing limits that might be required in 

licensee-controlled documentation to 

eliminating all operational restrictions 

associated with breaks greater than the 

TBS. Most commenters stated that 

operational restrictions negated the 

relief from the requirement to assume 

the worst single failure during the 

evaluation of beyond TBS breaks. 

NRC response. As discussed in 

Section III.D of this document, the NRC 

has decided that operational restrictions 

must be retained if it cannot be 

demonstrated in the analysis of LOCAs 

larger that the TBS that the ECCS 

acceptance criteria are met, but the 

restrictions would be reduced. The 

proposed rule prohibited at-power 

operation in a configuration without the 

demonstrated ability to mitigate a LOCA 

larger than the TBS. The revised 

proposed rule would require that atpower operation in such a configuration 

shall not exceed a total of fourteen days 

in any 12-month period. The NRC 

believes that this change will satisfy the 

Commission’s intention that mitigative 

capability be maintained for all breaks 

up to the double-ended rupture of the 

largest reactor coolant pipe and still 

allow a reasonable amount of time for 

licensees to make corrective actions 

needed to restore the plant to a fully 

analyzed configuration. 

NRC Topic 8. Given the Commission’s 

intent (see SRM for SECY–04–0037) that 

facility changes made possible by this 

proposed rule should be constrained in 

areas where the current design 

requirements ‘‘contribute significantly 

to the ‘built-in capability’ of the plant to 

resist security threats,’’ the NRC seeks 

examples on either side of this 

threshold (facility changes allowed 

versus facility changes prohibited), and 

additionally any examples of facility 

changes made possible by § 50.46a that 

could enhance plant security and 

defense against radiological sabotage or 

attack. The NRC also solicits comments 

on whether the proposed § 50.46a rule 

should explicitly include a requirement 

to maintain plant security when making 

facility changes under § 50.46a or 

otherwise rely on a separate rulemaking 

now being considered by the NRC to 

more globally address safety and 

security requirements when making 

facility changes under §§ 50.59 and 

50.90. Any examples of facility changes 

that involve safeguards information 

should be marked and submitted using 

the appropriate procedures. 

Comments. On the first question 

regarding examples of facility changes 

that should or should not be constrained 

in areas where the current design 

requirements ‘‘contribute significantly 

to the ‘built-in capability’ of the plant to 

resist security threats,’’ NEI said that the 

proposed rule would not enable facility 

changes that reduce plant safety margins 

as well as the capacity to deal with 

security threats. NEI stated that the 

opposite is true because the proposed 

rule would increase the safety focus on 

risk-significant events and mitigating 

equipment, and improve the reliability 

and availability of this equipment by 

removing excessive conservatism from 

the design basis. 

On the second question as to whether 

the § 50.46a rule should contain a 

security requirement, NEI said that 


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existing change control requirements in 

the regulations preclude significant 

reductions in safety or security. The 

BWROG supported the NEI position on 

this issue. The WOG stated that the 

security-related aspects of facility 

changes that might be enabled by this 

rule change should be addressed in the 

evaluation of those specific facility 

changes. The WOG also stated that the 

changes to § 50.46a should not be tied 

to security issues. Making a ‘‘security 

connection’’ to this proposed 

amendment would introduce needless 

complications and be 

counterproductive. Issues related to 

preserving ‘‘built-in capability’’ of the 

plant to resist threats should be 

addressed centrally in a single location 

within the regulations. Maintaining all 

requirements related to security in one 

place, either in the regulations or in 

Commission policy, is the most 

appropriate way to avoid conflicting 

information and enhance the ease of 

change. Progress Energy stated that 

consideration for security concerns 

should be included in the consideration 

of safety concerns to avoid possible 

negative effects caused by these 

sometimes competing objectives. 

However, to simplify the processes and 

maintain consistency, the safety and 

security interface should be addressed 

globally by a separate rulemaking. 

NRC response. The NRC agrees with 

commenters that security requirements 

should be addressed by regulations 

separate from those in § 50.46a. The 

NRC is not adding security requirements 

to proposed § 50.46a. Security 

requirements will continue to be 

addressed by overall security 

requirements located elsewhere in the 

regulations. Specifically, 10 CFR 73.58, 

‘‘Safety/security Interface Requirements 

for Nuclear Power Reactors’’ of the new 

Power Reactor Security Rule (74 FR 

13926; March 27, 2009), requires 

licensees to communicate plans for 

proposed plant changes that could 

impact plant security to security 

personnel who are qualified to analyze 

and identify potentially adverse impacts 

that the changes may have on safety 

and/or security programs. After security 

personnel analyze the changes for 

potential impacts, the regulation 

requires the licensee to take appropriate 

actions to mitigate the security impacts. 

NRC Topic 9. Given the potential 

impact to the licensee (because the 

backfit rule would not apply) of the 

NRC’s periodic re-evaluation of 

estimated LOCA frequencies which 

could cause the NRC to increase the 

TBS, should the proposed rule require 

licensees to maintain the capability to 

bring the plant into compliance with an 

increased transition break size (TBS), 

within a reasonable period of time? 

Comments. NEI, the BWROG, and the 

WOG commented that licensees should 

be provided with a great deal of latitude 

on achieving compliance following any 

change in the TBS, with the goal being 

that risk requirements are achieved with 

a reasonable mix of prevention and 

mitigation. 

NRC response. The NRC agrees with 

commenters that the § 50.46a rule 

should provide licensees with 

substantial flexibility to determine how 

they will come back into compliance 

with applicable regulatory requirements 

following any future change in the TBS. 

Licensees who must take actions to 

come back into compliance need not 

return the plant to the precise 

conditions and circumstances in effect 

immediately before implementation of 

§ 50.46a. Rather, licensees would be 

afforded the flexibility of deciding what 

actions they will implement to bring 

about compliance under any revised 

TBS. Further, as one of the commenters 

suggests, the overall goal of any actions 

taken to restore compliance is to achieve 

a reasonable mix of prevention and 

mitigation. 

NRC Topic 10. Is the proposed rule 

sufficiently clear as to be ‘‘inspectable?’’ 

That is, does the rule language lend 

itself to timely and objective NRC 

conclusions regarding whether or not a 

licensee is in compliance with the rule, 

given all the facts? In particular, are the 

proposed requirements for PRA quality 

sufficient in this regard? 

Comment. On the question of whether 

the proposed rule is clear enough to be 

inspectable, NEI was particularly 

concerned that the operational 

restrictions would conflict with the 

existing technical specifications. The 

BWROG supported the NEI position on 

this topic. 

NRC response. To reduce potential 

conflict between plant technical 

specifications and the operability 

requirements in § 50.46a, the NRC has 

also modified operability requirements 

to allow limited operation (for no more 

than a total of fourteen days in any 12- 

month period) in configurations where 

mitigation of LOCAs larger that the TBS 

has not been demonstrated. A detailed 

discussion on the basis for this new 

provision is provided below in Section 

V.F of this document, Operational 

Requirements. 

Comment. NEI stated that the rule 

would be difficult to inspect because it 

overlaps so many existing regulatory 

requirements. The WOG stated that the 

risk-informed aspects of the proposed 

rule, including the PRA quality 

requirements, should rely on the 

guidance of RG 1.174 and RG 1.200. The 

WOG stated that proposed § 50.46a 

should require no more ‘‘inspectability’’ 

than any other performance-based riskinformed application. Another 

commenter stated that the NRC should 

clarify certain aspects of the proposed 

rule and that the rule appropriately 

includes language like ‘‘reasonable 

balance’’ that requires a knowledgeable 

individual to exercise judgment which 

should be informed by appropriate 

regulatory guidance documents. 

NRC response. The NRC has modified 

the proposed rule to provide greater 

operational flexibility and reduce the 

potential for conflict with plant 

technical specification requirements 

that might cause ‘‘inspectability’’ 

problems. Although the WOG stated 

that the proposed rule would not have 

inspectability problems if it relied on 

the guidance in RG 1.174 and RG 1.200, 

the NRC notes that inspectors may not 

inspect licensees for compliance with 

regulatory guides because these guides 

are not regulatory requirements. The 

NRC has incorporated the important 

aspects of RG 1.174 and PRA quality 

guidance into the revised proposed rule 

itself so that inspectors would have a 

clear indication of the § 50.46a 

requirements. Specific inspection 

guidance will be developed as necessary 

after the final rule is published. 

NRC Topic 11. Proposed § 50.46a 

would impose no limitations on 

‘‘bundling’’ of different facility changes 

together in a single application. Facility 

changes which would increase plant 

risk substantially or create risk outliers 

could be grouped with other facility 

changes which would reduce risk so 

that the net change would meet the risk 

acceptance criteria. Are the net change 

in risk acceptance criteria in the 

proposed rule adequate or should some 

additional limitations be imposed to 

avoid allowing facility changes which 

are known to increase plant risk? 

Comments. Several commenters said 

that ‘‘bundling’’ is essential for meeting 

the objectives of this proposed rule 

which concerns overall plant risk. 

Bundling provides licensee management 

with the necessary flexibility to 

reallocate resources for implementation 

of the alternative requirements. The RG 

1.174 criteria related to bundling 

(combined change request in RG 1.174) 

are sufficient and no additional criteria 

or restrictions on bundling should be 

imposed by this proposed rule. 

NRC response. The NRC agrees that 

bundling of facility changes is desirable 

because it appropriately permits 

licensees to credit risk beneficial facility 

changes and encourages licensees to 

identify and implement facility changes 


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that decrease risk. The NRC also agrees 

that the guidelines on combined 

changes in RG 1.174 are sufficient to 

avoid facility changes which would 

unacceptably increase plant risk. 

NRC Topic 12. Is there an alternative 

to tracking the cumulative risk increases 

associated with facility changes made 

after implementing § 50.46a that is 

sufficient to provide reasonable 

assurance of protection to public health 

and safety and common defense and 

security? 

Comments. Four of the commenters 

who responded to the question stated 

that tracking cumulative risk increases 

was reasonable but they appeared to 

define cumulative tracking differently 

than as specified in the requirements of 

the proposed rule. NEI, whose 

comments were generally endorsed by 

most of the 12 commenters, 

recommended rule text stating ‘‘[t]he 

licensee shall periodically assess the 

cumulative effect of changes to the plant 

design configuration and update as 

necessary, the PRA and other risk 

analyses.’’ After discussing this 

proposed text at the June 28, 2006, 

public meeting, the NRC determined 

that the recommendation equated 

tracking cumulative risk increases with 

periodically updating the PRA and 

estimating the latest core damage 

frequency (CDF) and large early release 

frequency (LERF) using the updated 

PRA. NEI intended for these latest risk 

estimates themselves to represent the 

assessment of the cumulative increase. 

However, the proposed rule required 

that some previous estimates of CDF 

and LERF be subtracted from the latest 

estimates to obtain the amount by which 

the CDF and LERF has increased. One 

of the four commenters added that 

tracking the cumulative risk increase (as 

intended by the NRC in the proposed 

rule) was not necessary because the 

threshold for risk increase is low 

enough so that the cumulative effect is 

not significant. A fifth commenter 

argued that tracking cumulative risk 

should not be required by the rule 

because compliance with the guidance 

in RG 1.174 should be sufficient to 

ensure that cumulative risk does not 

impact the health and safety of the 

public. 

NRC response. The NRC has retained 

the requirement to track the total risk 

increases in CDF and LERF made under 

the proposed rule and has retained the 

definition of risk ‘‘increase’’ as being the 

amount by which risk increases. RG 

1.174 provides guidance on judging the 

acceptability of proposed facility 

changes based primarily on the amount 

by which the facility changes increase 

CDF and LERF. The NRC has clarified 

what it has concluded must be tracked 

in § 50.46a(f)(2)(iv) utilizing the 

requirement for tracking the cumulative 

effect on risk of changes made under the 

NFPA–805 standard which was 

incorporated by reference into § 50.48(c) 

(see, 69 FR 33536; June 16, 2004). By 

utilizing the same language in both 

rules, the NRC intends that the 

implementation of both rules would be 

consistent. 

The NRC has concluded that the 

alternative proposed by the commenters 

(i.e. to track cumulative risk by simply 

updating the PRA) is not acceptable 

because the latest estimates of CDF and 

LERF alone provide insufficient 

information to be used in the riskinformed framework contained in RG 

1.174. Two other commenters argued 

that risk tracking is not needed because 

controls external to proposed § 50.46a 

(e.g., in RG 1.174) would ensure that the 

cumulative effect would not be 

significant. The commenters provided 

no basis for their assertions that controls 

external to the rule would keep 

increases in risk small enough to ensure 

protection of public health and safety. 

RG 1.174 does discuss tracking changes 

in cumulative risk, but regulatory guides 

are not enforceable requirements. The 

NRC has determined that it is necessary 

to establish a regulatory requirement to 

track the cumulative risk increases from 

all changes made under this proposed 

rule. The NRC continues to believe that 

risk tracking as described in the 

proposed rule is needed to ensure that 

facility changes permitted by the revised 

ECCS analyses under § 50.46a do not 

result in greater increases in risk than 

were intended by the Commission. 

NRC Topic 13. The NRC requested 

specific public comments on the 

acceptability of applying the change in 

risk acceptance guidelines in RG 1.174 

to the total cumulative change in risk 

from all changes in the plant after 

adoption of § 50.46a. Should other risk 

guidelines be used and, if so, what 

guidelines should be used? 

Comments. As discussed, four 

commenters proposed tracking 

cumulative risk increases by 

periodically updating the PRA, 

estimating the latest CDF and LERF 

using the updated PRA, and equating 

these latest estimates with tracking the 

cumulative risk increase. Applying this 

definition for tracking cumulative risk 

increase, these commenters concluded 

that the change in risk acceptance 

guidelines should not be applied to the 

total cumulative change in risk which 

would not, under their proposals, be 

estimated. 

In general, most commenters’ either 

explicitly or implicitly recommended 

that the rule should not include the 

acceptance criteria that ‘‘the total 

increases in CDF and LERF should be 

small and the overall risk should remain 

small.’’ Proposals for alternatives varied. 

NEI’s proposed rule text did not include 

acceptance criteria related to increases 

in CDF and LERF. Instead, NEI 

proposed requiring the licensee to 

report the results of the updated PRA 

and other risk analyses to the NRC. One 

commenter argued that for facility 

changes enabled by the new § 50.46a, 

compliance with RG 1.174 should be 

sufficient. Two commenters stated that 

risk tracking accomplished by updating 

the PRA and estimating the latest CDF 

and LERF can be used to ensure that the 

total risk as well as the risk from 

specific initiators or classes of accidents 

is not increasing. 

NRC response. The NRC has retained 

the requirement in the revised proposed 

rule that the total change in risk from 

facility changes, measured as the 

amount by which CDF and LERF (or 

LRF for new reactors) increase, be 

tracked and compared to the RG 1.174 

acceptance criteria. However, the NRC 

has reduced the scope of facility 

changes that must be tracked from all 

changes to only those changes made to 

the plant under § 50.46a. 

Implementation of all RG 1.174 

guidelines can only be achieved using a 

process that includes an estimate of the 

cumulative change in risk. Also, 

consistent with the Commission’s 

direction in the SRM for SECY–07– 

0082, the NRC has reduced the size of 

an acceptable risk increase from ‘‘small’’ 

to ‘‘very small’’. The revised proposed 

rule would continue to use the 

quantitative guidelines in RG 1.174. 

NEI’s proposal for reporting the latest 

estimates of CDF and LERF to the NRC 

after each periodic assessment would 

not be useful because the NRC has no 

criteria for determining which CDF and 

LERF values would be acceptable. It 

would be a lengthy process to establish 

such acceptance criteria. Lack of 

acceptance criteria against which the 

latest CDF and LERF can be compared 

will result in different stakeholders 

applying different criteria to judge the 

acceptability of the results most likely 

leading to different conclusions. 

The NRC believes that the two 

comments proposing that the total CDF 

as well as the CDF from specific 

initiators or class of accidents could be 

tracked to ensure that risk from these 

scenarios is not increasing would satisfy 

the requirement that the total increase 

in risk remains very small provided that 

the appropriate initiators or class of 

accident is identified (and including 

LERF or LRF). The commenters did not 


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appear to be proposing that such a 

constraint be included in the rule, 

instead they were only making 

observations on what would be possible. 

Nevertheless, in an SRM on August 10, 

2007, the Commission concluded that 

only a very small increase in risk is 

acceptable when implemented 

according to the requirements in this 

rule. Requiring that there be no risk 

increase, as hypothesized by the 

commenters, is more restrictive than the 

criteria in the revised proposed rule. 

Although the revised proposed rule 

would permit licensees to make plant 

changes that result in very small risk 

increases, the NRC requests stakeholder 

comments on whether any increase in 

risk should be allowed. Instead of the 

risk acceptance criteria allowing very 

small risk increases, should the 

acceptance criteria in the final rule 

require that the net effect of plant 

changes made under § 50.46a be risk 

neutral or risk beneficial? The NRC 

requests stakeholders to provide 

comments on the use of risk acceptance 

criteria that would not allow a 

cumulative increase in risk for plant 

changes made under § 50.46a. 

NRC Topic 14. After approval to 

implement § 50.46a, the proposed rule 

would require tracking risk associated 

with all proposed facility changes but 

would not require a licensee to include 

risk increases caused by previous riskinformed facility changes that were 

implemented before § 50.46a was 

adopted. Licensees who adopt § 50.46a 

before implementing other riskinformed applications would have a 

smaller risk increase ‘‘available’’ 

compared to licensees who have already 

incorporated some risk-informed facility 

changes into their overall plant risk 

before adopting § 50.46a. The NRC 

requests specific public comments on 

whether this potential inconsistency 

should be addressed and, if so, how? 

Comments. Three commenters stated 

that these potential inconsistencies in 

acceptable risk increases should be 

addressed by deleting the requirement 

that the cumulative risk increase be 

tracked and compared to the RG 1.174 

acceptance guidelines. The commenters 

argued that licensees and the NRC have 

effectively managed incremental risk 

without the need for this structure and 

that any facility changes that seek to 

apply the revised design bases should 

be evaluated using the same methods 

proven effective in the past. A fourth 

commenter agreed with the others but 

proposed that inconsistencies among 

licensees created by the order of 

implementing risk-informed 

applications could be resolved by 

allowing a licensee to reestablish the 

baseline and removing some facility 

changes from tracking. 

NRC response. The NRC is proposing 

additional changes in the revised 

proposed rule that would make this 

topic moot. The proposed rule would 

have required tracking total risk from all 

facility changes. This requirement 

reflected a difficulty uniquely 

associated with comparing the total risk 

increases from all facility changes to the 

acceptance criteria. The revised 

proposed rule would only require that 

facility changes made under the rule be 

tracked. Other risk-informed facility 

changes referred to in Topic 14 would 

no longer be included in this change in 

risk estimate and therefore, the 

acceptability of those facility changes 

will be independent of facility changes 

made under this rule (aside from the 

indirect affect these facility changes 

have on the plant’s risk profile). 

NRC Topic 15. Proposed § 50.46a 

would require licensees to report every 

24 months all ‘‘minimal’’ risk facility 

changes made under § 50.46a(f)(1) 

without NRC review. Are there less 

burdensome or more effective ways of 

ensuring that the cumulative impact of 

an unbounded number of ‘‘minimal’’ 

changes remains inconsequential? 

Comments. Several commenters 

stated that the § 50.46a(g)(3) report 

summarizing minimal risk changes 

every 24 months is redundant to reports 

required under § 50.59(d)(2) as well as 

§ 50.71(e). Thus, § 50.46a(g)(3) should 

be deleted. The requirement needlessly 

focuses licensee and NRC resources 

directly on a large set of information 

that by its very definition has no safety 

or risk significance. 

NRC response. The NRC agrees with 

the commenters that the reporting 

requirements in proposed § 50.46a(g)(3) 

could be redundant to other reporting 

requirements for some facility changes 

because some changes made under the 

new rule might be reportable under both 

§ 50.59 and § 50.46a(g)(3). The NRC has 

determined that breaks larger than the 

TBS should be removed from the design 

basis event category. Therefore, the NRC 

believes that some facility changes that 

may be made under the new rule would 

no longer be reportable under § 50.59 

because the change would no longer 

affect design basis events. The NRC is 

proposing to reduce the scope of facility 

changes that need to be evaluated under 

the new provision, from all changes 

made to the facility after adoption of the 

rule to only facility changes that are 

made under the new rule. This change 

would reduce the number of potentially 

redundant reports. 

To avoid the possibility that 

potentially risk-significant changes are 

not reported, the NRC has concluded 

that all facility changes made under the 

new rule should be reported because the 

NRC will rely on the risk evaluation to 

prevent facility changes that might not 

be protective of public health and 

safety. Therefore, the NRC has retained 

the reporting requirements in 

§ 50.46a(g)(3) because these 

requirements would ensure the 

reporting of all potentially risksignificant facility changes made under 

the proposed rule. 

NRC Topic 16. Should the § 50.46a 

rule itself include high-level criteria and 

requirements for the risk evaluation 

process and acceptance criteria 

described in RG 1.174? If these criteria 

were included in the regulatory guide 

only, and not in § 50.46a, how could the 

NRC take enforcement action for 

licensees who failed to meet the 

acceptance criteria? 

Comments. Four commenters stated 

that proposed § 50.46a rule should not 

contain the high-level criteria and 

requirements for the risk evaluation 

process and acceptance criteria 

described in RG 1.174. These 

commenters did not specifically 

propose how the NRC could take 

enforcement action to ensure 

compliance with the criteria, but instead 

asserted that regulatory guidance 

documents and inspection guidelines 

are the appropriate places for the risk 

acceptance criteria. 

NRC response. The NRC does not 

agree with the commenters. The 

proposed rule would have to contain 

high-level requirements for the risk 

evaluation and acceptance criteria to 

establish the legally enforceable 

alternative regulatory requirements 

needed to ensure adequate protection of 

public health and safety in a manner 

which maximizes regulatory 

predictability and stability. The NRC 

believes that proposed § 50.46a should 

build upon NRC and industry 

experience with the key principles of 

risk-informed decision making set forth 

in RG 1.174, but notes that RG 1.174 

only contains guidance, not 

requirements. To be enforceable, 

proposed § 50.46a must contain and 

does contain high-level requirements 

relating to risk, defense-in-depth, safety 

margins, risk, and performance 

measurement. Specific, detailed 

guidance on how to meet the high-level 

requirements will be set forth in 

regulatory guidance and inspection 

guidelines, as appropriate. 


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V. Revised Proposed Rule 

A. Overview 

The NRC’s revised proposed rule 

would establish an alternative set of 

risk-informed requirements with which 

licensees may choose to comply in lieu 

of meeting the current emergency core 

cooling system requirements in 10 CFR 

50.46. Using the alternative ECCS 

requirements would provide some 

licensees with opportunities to change 

other aspects of facility design. 

As was the case in the initial 

proposed rule, the revised proposed rule 

divides the current spectrum of LOCA 

break sizes into two regions. The 

division between the two regions is 

delineated by the TBS. The first region 

includes small size breaks up to and 

including the TBS. The second region 

includes breaks larger than the TBS up 

to and including the DEGB of the largest 

RCS pipe. Break area for the TBS is not 

based on a double-ended offset break. 

Rather, it is based on the inside area of 

a single-sided circular pipe break. Pipe 

breaks in the smaller break size region 

are considered more likely than pipe 

breaks in the larger break size region. 

Consequently, each break size region 

will be subject to different ECCS 

requirements, commensurate with 

likelihood of the break. LOCAs in the 

smaller break size region must be 

analyzed by the same conservative 

methods, assumptions, and criteria 

currently used for LOCA analysis. 

Accidents in the larger break size region 

may be analyzed using more realistic 

methods and assumptions based on 

their lower likelihood. Although LOCAs 

for break sizes larger than the transition 

break would become ‘‘beyond designbasis accidents,’’ the revised proposed 

rule would require that licensees 

maintain the ability to mitigate all 

LOCAs up to and including the DEGB 

of the largest RCS pipe. However, 

mitigation analyses for LOCAs larger 

than the TBS need not assume the lossof-offsite power or the occurrence of a 

single failure. 

Licensees who perform LOCA 

analyses using the risk-informed 

alternative requirements may find that 

their plant designs are no longer limited 

by certain parameters associated with 

previous DEGB analyses. Reducing the 

DEGB limitations could enable licensees 

to propose a wide scope of design or 

operational changes up to the point of 

being limited by some other parameter 

associated with any of the other 

required accident analyses. Potential 

design changes include modification of 

containment spray designs, modifying 

core peaking factors, modifying 

setpoints on accumulators or removing 

some from service, eliminating fast 

starting of one or more emergency diesel 

generators, and increasing power, etc. 

Some of these design and operational 

changes could increase plant safety 

because a licensee could modify its 

systems to better mitigate the more 

likely LOCAs. Other changes, such as 

increasing power, could increase overall 

risk to the public. The risk-informed 

§ 50.46a option would include risk 

acceptance criteria for evaluating future 

design changes to ensure that any risk 

increases are acceptably small. These 

acceptance criteria would be consistent 

with the guidelines for risk-informed 

license amendments in RG 1.174 and 

would ensure both the acceptability of 

the changes from a risk perspective and 

the maintenance of sufficient defensein-depth, safety margins, and 

performance monitoring. The 

requirements for the risk-informed 

evaluation process are discussed in 

detail in Section V.E of this document. 

The NRC will periodically evaluate 

LOCA frequency information. Should 

estimated LOCA frequencies increase 

causing a significant increase in the risk 

associated with breaks larger than the 

TBS, the NRC would undertake 

rulemaking (or issue orders, if 

appropriate) to change the TBS. In such 

a case, the backfit rule (10 CFR 50.109) 

will not apply. If previous plant changes 

are invalidated because of a change to 

the TBS, licensees would have to 

modify or restore components or 

systems as necessary so that the facility 

would continue to comply with § 50.46a 

acceptance criteria. The backfit rule (10 

CFR 50.109) also would not apply in 

these cases. 

Changes consist of a new § 50.46a and 

conforming changes to existing §§ 50.34, 

50.46, 50.46a (redesignated as § 50.46b), 

50.109, 10 CFR Part 50, Appendix A, 

General Design Criteria 17, 35, 38, 41, 

44 and 50, and §§ 52.47, 52.79, 52.137, 

and 52.157. 

B. Determination of the Transition 

Break Size 

To help establish the TBS, the NRC 

developed pipe break frequencies as a 

function of break size using an expert 

opinion elicitation process for 

degradation-related pipe breaks in 

typical BWR and PWR reactor coolant 

systems (NUREG–1829; ‘‘Estimating 

Loss-of-Coolant Accident (LOCA) 

Frequencies through the Elicitation 

Process’’ March 2008; ML082250436). 

The elicitation process is used for 

quantifying phenomenological 

knowledge when data or modeling 

approaches are insufficient. The 

elicitation focused solely on 

determining event frequencies that 

initiate unisolable primary system side 

failures related to material degradation. 

A baseline TBS was established from 

the expert elicitation results for each 

reactor type (i.e., PWR and BWR) that 

corresponded to a break frequency of 

once per 100,000 reactor years (1 × 10¥5

or 10¥5 per reactor year). The NRC then 

considered uncertainty in the elicitation 

process, other potential mechanisms 

that could cause passive component 

failure that were not explicitly 

considered in the expert elicitation 

process, and the higher susceptibility to 

rupture/failure of specific locations in 

the reactor coolant system (RCS); 

adjusting the TBS upwards to account 

for these factors. Other mechanisms that 

contribute to the overall LOCA 

frequency include LOCAs resulting from 

failures of non-passive components and 

LOCAs resulting from low probability 

events (earthquakes of magnitude larger 

than the safe shutdown earthquake and 

dropped heavy loads). These LOCAs 

have a strong dependency on plantspecific factors. 

LOCAs caused by failure of nonpassive components, such as stuck-open 

valves and blown out seals or gaskets 

have a greater frequency of occurrence 

than LOCAs resulting from the failure of 

passive components. LOCAs resulting 

from the failure of non-passive 

components would be small-break 

LOCAs, when considering the size of 

the opening that could result should 

components fail open or blow out (e.g., 

safety valves, pump seals). LOCAs 

resulting from stuck-open valves are 

limited by the size of the auxiliary pipe. 

In some PWRs, there are large loop 

isolation valves in the hot and cold leg 

piping. However, a complete failure of 

the valve stem packing is not expected 

to result in a large flow area, because the 

valves are back-seated in the open 

configuration. Based on these 

considerations, non-passive LOCAs are 

relatively small in size and are bounded 

by the selected TBS. 

LOCAs could also be caused by 

dropping heavy loads that could cause 

a breach of the RCS piping. During 

power operation, personnel entry into 

the containment is typically infrequent 

and of short duration. The lifting of 

heavy loads that if dropped would have 

the potential to cause a LOCA or 

damage safety-related equipment is 

typically performed while the plant is 

shutdown. The majority of heavy loads 

are lifted during refueling evolutions 

when the primary system is 

depressurized, further reducing the risk 

of a LOCA and a loss of core cooling. If 

loads are lifted during power operation, 

they would not be loads similar to the 

heavy loads lifted during plant 


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shutdown, e.g., vessel heads and reactor 

internals. In addition, the RCS is 

inherently protected by surrounding 

concrete walls, floors, missile shields, 

and biological shielding. Thus, the 

contribution of heavy load drops to 

overall LOCA frequency is not 

considered to be significant and would 

not affect the TBS. 

Seismically-induced LOCA break 

frequencies can vary greatly from plant 

to plant because of factors such as site 

seismicity, seismic design 

considerations, and plant-specific 

layout and spatial configurations. 

Seismic break frequencies are also 

affected by the amount of pipe 

degradation occurring prior to 

postulated seismic events. Seismic PRA 

insights have been accumulated from 

the NRC Seismic Safety Margins 

Research Program and the Individual 

Plant Examination of External Events 

submittals. Based on these studies, 

piping and other passive RCS 

components generally exhibit high 

seismic capacities and, therefore, are not 

significant risk contributors. However, 

these studies did not explicitly consider 

the effect of degraded component 

performance on the risk contributions. 

Therefore, the NRC conducted a study 

to evaluate the seismic performance of 

undegraded and degraded passive 

system components (NUREG–1903, 

‘‘Seismic Considerations for the 

Transition Break Size,’’ February 2008; 

ML080880140). This effort examined 

operating experience, seismic PRA 

insights, and models to evaluate the 

failure likelihood of undegraded and 

degraded piping. The operating 

experience review considered passive 

component failures that have occurred 

as a result of strong motion earthquakes 

in nuclear and fossil power plants as 

well as other industrial facilities. No 

catastrophic failures of large pipes 

resulting from earthquakes between 0.2g 

and 0.5g peak ground acceleration have 

occurred in power plants. However, 

piping degradation could increase the 

LOCA frequency associated with 

seismically-induced piping failures. The 

NUREG–1903 report evaluated seismic 

loadings on degraded piping and 

concluded that a very large, pre-existing 

crack on the order of 30 percent 

through-wall and 145 degrees around 

the piping circumference would have to 

be present during a 10¥5 or 10¥6 per 

year earthquake in order for pipe failure 

to occur. The NRC concluded that the 

likelihood of flaws large enough to fail 

during a seismic event is sufficiently 

low that the TBS need not be modified 

to address seismically-induced direct 

piping failures. In reaching its 

conclusion, the NRC considered the 

comments received as well as historical 

information related to piping 

degradation and the potential for the 

presence of cracks sufficiently large that 

pipe failure would be expected under 

loads associated with rare (10¥5 per 

year) earthquakes. 

Indirect failures are primary system 

ruptures that are a consequence of 

failures in nonprimary system 

components or structural support 

failures (such as a steam generator 

support). Structural support failures 

could then cause displacements in 

components that stress and in turn, fail 

the piping. The NRC performed studies 

on two plants to estimate the 

conditional pipe failure probability due 

to structural support failure given a low 

return frequency earthquake (10¥5 to 

10¥6 per year). The results indicated 

that the conditional probability was on 

the order of 0.1. These studies used 

seismic hazard curves from NUREG– 

1488 (NUREG–1488, ‘‘Revised 

Livermore Seismic Hazard Estimates for 

Sixty-Nine Nuclear Power Plant Sites 

East of the Rocky Mountains, April 

1994; ML052640591). More recent 

studies were completed by EPRI on 

three plants using updated seismic 

hazard estimates. The updated seismic 

hazard increases the peak ground 

acceleration at some sites. The highest 

pipe failure probability calculated for 

the three plants in the industry analyses 

was 6 × 10¥6 per year. The NRC noted 

in its report that indirect failure 

analyses are highly plant-specific. 

Therefore, it is possible that example 

plants assessed in the NRC and EPRI 

analyses are not limiting for all plants. 

The NRC has considered the 

importance of indirect failures on the 

selection of the TBS. For the cases 

considered in both the EPRI and NRC 

studies, the likelihood of indirectly 

induced piping failures resulting from 

major component support failures is less 

than 10¥5 per reactor year, the 

frequency criterion used to select the 

TBS. Also, as noted in the public 

comments, the median seismic 

capacities for both the primary piping 

system and primary system components 

are typically higher than other safety 

related components within the nuclear 

power plant. Because of these relative 

capacities, it is expected that a seismic 

event of sufficient magnitude to cause 

consequential failure within the primary 

system would also induce failure of 

components in multiple trains of 

mitigation systems, or even induce 

multiple RCS pipe breaks. 

Consequently, the risk contribution 

from seismically induced indirect 

failures is expected to depend more 

heavily on the relative fragilities of 

plant components and systems than the 

size of the TBS. Therefore, the NRC 

believes that adjustment to the TBS for 

seismically induced indirect LOCAs is 

also not warranted. 

The final consideration in selecting 

the TBS was actual piping system 

design (e.g., sizes) and operating 

experience. For example, due to 

configuration and operating 

environment, certain piping is 

considered to be more susceptible than 

other piping in the same size range. For 

PWRs, the range of pipe break sizes 

determined from the various 

aggregations of expert opinion was 6 to 

10 inches in diameter (i.e., inside 

dimension) for the 95th percentile. This 

is only slightly smaller than the PWR 

surge lines, which are attached to the 

RCS main loop piping and are typically 

12- to 14-inch diameter Schedule 160 

piping (i.e., 10.1 to 11.2 inch inside 

diameter piping). The RCS main loop 

piping is in the range of 30 inches in 

diameter and has substantially thicker 

walls than the surge lines. The expert 

elicitation panel concluded that this 

main loop piping is much less likely to 

break than other RCS piping. The 

shutdown cooling lines and safety 

injection lines may also be 12- to 14- 

inch diameter Schedule 160 piping and 

are likewise connected to the RCS. The 

difference in diameter and thickness of 

the reactor coolant piping and the 

piping connected to it forms a 

reasonable line of demarcation to define 

the TBS. Therefore, to capture the surge, 

shutdown cooling, and safety injection 

lines in the range of piping considered 

to be equal to or less than the TBS, the 

NRC specified the TBS for PWRs as the 

cross-sectional flow area of the largest 

piping attached to the RCS main loop. 

For BWRs, the arithmetic and 

geometric means of the break sizes 

having approximately a 95th percentile 

probability of 10¥5 per year ranged from 

values of approximately 13 inches to 20 

inches equivalent diameter. The 

information gathered from the 

elicitation for BWRs showed that the 

estimated frequency of pipe breaks 

dropped markedly for break sizes 

beyond the range of approximately 18 to 

20 inches. After evaluating BWR 

designs, it was determined that typical 

residual heat removal piping connected 

to the recirculation loop piping and 

feedwater piping is about 18 to 24 

inches in diameter. These pipe sizes are 

consistent with break sizes beyond 

which the pipe break frequency is 

expected to decrease markedly below 

10¥5 per year. It was also recognized 

that the sizes of attached pipes vary 

somewhat among plants. Thus, for 


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BWRs, the TBS is specified as the crosssectional flow area of the larger of either 

the feedwater or the RHR piping inside 

primary containment. 

Because the effects of TBS breaks on 

core cooling vary with the break 

location, the NRC evaluated whether the 

frequency of TBS breaks varies with 

location and whether TBS breaks 

should, therefore, vary in size with 

location. In PWRs, the pressurizer surge 

line is only connected to one hot leg and 

the pipes attached to the cold legs are 

generally smaller than the surge line. 

The cold legs (including the 

intermediate legs) operate at slightly 

cooler temperatures. Thermallyactivated degradation mechanisms 

would be expected to progress more 

slowly in the cold leg than in the hot 

leg. Therefore, the NRC evaluated 

whether it may be appropriate to specify 

a TBS for the cold leg that would be 

smaller than the size of the surge line. 

The frequency of occurrence of a break 

of a given size is composed of both the 

frequency of a completely severed pipe 

of that size (a complete circumferential 

break) plus the frequency of a partial 

break of that size in an equal or larger 

size pipe (a partial circumferential or 

longitudinal break). Therefore, the NRC 

evaluated an option where the TBS for 

the hot and cold legs would be 

distinctly different and would be 

composed of two components: (1) 

Complete breaks of the pipes attached to 

the hot or cold legs at the limiting 

locations within each attached pipe, and 

(2) partial breaks of a constant size, as 

appropriate for either the hot or cold 

leg, at the limiting locations within the 

hot or cold legs. The NRC attempted to 

estimate the appropriate size of the 

partial break component for the TBS by 

reviewing the expert elicitation results 

to determine the frequencies of 

occurrence of partial breaks within hot 

and cold legs that would be equivalent 

to the frequency of a complete surge line 

break. The NRC found that frequencies 

of occurrence of partial breaks of a given 

size are generally lower for the cold leg 

than for the hot leg. However, other than 

this general trend, the elicitation results 

do not contain sufficient information to 

adequately quantify differences among 

the hot leg, cold leg, and surge line pipe 

break frequencies. Because it was not 

possible to establish a smaller partial 

break TBS criterion in the hot or cold 

legs, the NRC concluded that the TBS 

associated with partial breaks in the hot 

and cold legs should remain equivalent 

in size to the internal cross sectional 

area of the surge line. Similarly, the 

elicitation results do not contain 

sufficient detail to quantify break 

frequency differences among the BWR 

recirculation, residual heat removal, and 

feedwater system piping. Thus, a 

smaller partial break TBS criterion also 

could not be established for BWR 

recirculation piping. 

The NRC also evaluated whether TBS 

breaks should be analyzed as singleended or double-ended breaks. To 

address this issue, the NRC reviewed 

the expert elicitation process and the 

guidance given to the experts in 

developing their frequency estimates. 

The NRC concluded that the expert 

elicitation LOCA frequency estimates 

correspond to a break area having an 

equivalent circular diameter at each 

break size. This correspondence is 

representative of a single-ended break. 

Additionally, the experts based their 

estimates on knowledge of postulated 

failure mechanisms in pressure 

boundary components and not on the 

flow rates emanating from the breaks. 

The flow rates are governed by the break 

location and system configuration 

which determines whether reactor 

coolant will be discharged from both 

ends of the break. 

The current design basis analysis for 

light water reactors requires analysis of 

a DEGB of the largest pipe in the RCS. 

Under the proposed rule, all breaks up 

to and including the TBS would be 

analyzed under existing requirements. A 

possible reason for specifying the TBS 

for PWRs as double-ended could be that 

a complete break of the pressurizer 

surge line would result in reactor 

coolant exiting both ends of the break. 

Although this occurs initially during a 

LOCA, core cooling requirements are 

dominated by the flow rate of coolant 

exiting from the hot leg side of the 

break, with much less contribution from 

the flow rate of coolant exiting from the 

pressurizer side. Therefore, specifying 

the TBS break as an area equivalent to 

a double-ended break of the surge line 

would be overly conservative. For 

BWRs, the effect of a double-ended 

break area is also considered to be 

overly conservative. The selected TBS 

for BWRs is based on the larger of the 

residual heat removal or main feedwater 

lines attached to the main recirculation 

piping. A single-ended break in these 

lines would bound double-ended breaks 

of the smaller lines in the reactor 

recirculation and feedwater system. 

Therefore, the NRC concluded that 

treating the TBS as a single-ended break 

reasonably characterizes the expert 

elicitation results and represents the 

flow rates associated with postulated 

pipe breaks within the RCS. 

For the TBS to remain valid at a 

particular facility, future plant 

modifications must not significantly 

increase the LOCA pipe break frequency 

estimates generated during the expert 

elicitation and used as the basis for the 

TBS. For example, the expert elicitation 

panel did not consider the effects of 

power uprates in deriving the break 

frequency estimates. The expert 

elicitation panel assumed that future 

plant operating characteristics would 

remain consistent with past operating 

practices. The NRC recognizes that 

significant plant changes may change 

plant performance and relevant 

operating characteristics to a degree that 

they might impact future LOCA 

frequencies. The NRC will expect 

applicants for plant changes under 

revised proposed § 50.46a to 

demonstrate that those changes do not 

significantly increase break frequencies. 

As discussed in Section V.C. of this 

document, the NRC is currently 

preparing guidance for applicants to use 

to demonstrate that proposed plant 

changes do not undermine the § 50.46a 

technical basis (‘‘Plant-Specific 

Applicability of 10 CFR 50.46 Technical 

Basis’’ February 2009; ML090350757). 

The baseline TBS was adjusted 

upward to account for uncertainties and 

failure mechanisms leading to pipe 

rupture that were not considered in the 

expert elicitation process. As the NRC 

obtains additional information that may 

tend to reduce those uncertainties or 

allow for more structured consideration 

of degradation mechanisms, the NRC 

will assess whether the TBS (as defined 

in § 50.46a) should be adjusted, and 

may initiate rulemaking to revise the 

TBS definition to account for this new 

information. The NRC will also 

continue to assess the failure precursors 

that might be indicative of an increase 

in pipe break frequencies in BWR and 

PWR plants to establish whether the 

TBS would need to be adjusted. 

However, these TBS values are within 

the range supported by the expert 

elicitation estimates when considering 

the uncertainty inherent in processing 

the degradation-related frequency 

estimates. In addition, the NRC believes 

that the TBS definitions in the proposed 

rule would provide necessary 

conservatism to compensate for possible 

future increases in break frequencies. 

The NRC expects that the TBS values 

would result in regulatory stability 

because future LOCA frequency 

reevaluations are less likely to make it 

necessary for the NRC to change the 

TBS and cause licensees to undo plant 

modifications made after implementing 

§ 50.46a. 


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C. Evaluation of the Plant-Specific 

Applicability of the Transition Break 

Size 

As discussed in Section V.B. of this 

document, the NRC has published two 

reports, NUREG–1829 (ML082250436), 

and NUREG–1903 (ML080880140) that 

form part of the technical basis used to 

select the TBS for BWR and PWR plants. 

NUREG–1829 used expert elicitation to 

develop generic LOCA frequency 

estimates of passive system failure as a 

function of break size for both BWR and 

PWR plants and considered normal 

operational loading and transients 

expected over a 60-year plant life. 

NUREG–1903 assessed the likelihood 

that rare seismic events would induce 

primary system failures larger than the 

postulated TBS. NUREG–1903 evaluated 

both direct failures of flawed and 

unflawed primary system pressure 

boundary components and indirect 

failures of nonprimary system 

components and supports that could 

lead to primary system failures. Because 

these studies were not intended to 

develop bounding estimates, unique 

plant attributes may result in plantspecific LOCA frequencies due to 

normal operational and/or seismic 

loading that are greater than reported in 

either NUREG–1829 or NUREG–1903. 

Consequently, the NRC has included a 

requirement that applicants wishing to 

implement § 50.46a conduct an 

evaluation to demonstrate that the 

results in NUREG–1829 and NUREG– 

1903 are applicable to their individual 

plants. 

The NRC is preparing guidance for 

conducting the plant specific review to 

demonstrate the applicability of both 

the NUREG–1829 and NUREG–1903 

results. The scope of this applicability 

guidance would be limited to primary 

system piping and other primary 

pressure boundary components that are 

large enough to result in LOCA break 

sizes larger than the TBS. This guidance 

is applicable to aspects of the facility 

design affecting compliance with ECCS 

requirements and would not pertain to 

design-bases or operational procedures 

associated with other aspects of the 

facility licensing basis. 

The plant applicability evaluation 

would require that § 50.46a applicants 

first demonstrate that the applicable 

systems in the plant adhere to the 

current licensing basis. Additionally, 

the evaluation would require that 

licensees consider the effects of unique, 

plant-specific attributes on the generic 

LOCA frequencies developed in 

NUREG–1829. The licensee would also 

evaluate the effect of proposed plant 

changes on both direct and indirect 

system failures to demonstrate that 

NUREG–1829 results remain applicable 

after the proposed changes have been 

implemented. After a licensee is 

approved to implement revised 

proposed § 50.46a requirements, it 

would also be necessary to evaluate the 

effect of future proposed plant changes 

to demonstrate that NUREG–1829 

results remain applicable after enacting 

the proposed changes. 

An evaluation framework is also 

provided for determining the 

applicability of the NUREG–1903 

assessment of direct piping failures. 

This framework identifies the aspects 

that applicants would consider in a 

plant-specific analysis, provides several 

options for conducting the analysis, and 

describes a systematic approach 

associated with each option. One 

important step is to determine whether 

the NUREG–1903 results can be used 

directly or if a plant-specific analysis is 

required to determine the limiting flaw 

sizes under rare seismic loading. 

NUREG–1903 also addressed indirect 

piping failures caused by rare seismic 

loading. However, the risk of indirect 

failure is highly plant-specific and 

NUREG–1903 only considered the risks 

associated with two different plants. 

Consequently, the limited analysis of 

indirect piping failures does not provide 

a sufficient technical basis for allowing 

generic changes to the seismic design, 

testing, analysis, qualification, and 

maintenance requirements associated 

with any component under § 50.46a. 

Any proposed changes to these criteria 

would be justified using a plant-specific 

analysis to assess the change in risk 

associated with seismically induced 

failures of the relevant component and/ 

or system that results from the proposed 

plant changes. After receiving approval 

to implement revised proposed § 50.46a 

requirements, it would also be necessary 

for licensees to demonstrate that the 

NUREG–1903 results remain applicable 

after implementing proposed changes. 

More specific details on how to 

conduct these applicability reviews are 

available in a white paper entitled, 

‘‘Plant-Specific Applicability of the 10 

CFR 50.46 Technical Basis’’ February 

2009 (ML090350757). Commenters on 

this revised proposed rule may review 

this white paper to get a better 

understanding of the scope of the 

evaluation being considered by the 

NRC. 

D. Alternative ECCS Analysis 

Requirements and Acceptance Criteria 

The revised proposed rule would 

require licensees to analyze ECCS 

cooling performance for breaks up to 

and including a double-ended rupture 

of the largest pipe in the RCS. These 

analyses would have to be performed by 

methods acceptable to the NRC and 

must demonstrate that ECCS cooling 

performance conforms to the acceptance 

criteria set forth in the rule. For breaks 

at or below the TBS, § 50.46a(e)(1) 

would specify requirements identical to 

the existing ECCS analysis requirements 

set forth in § 50.46. However, 

commensurate with the lower 

probability of breaks larger than the 

TBS, § 50.46a(e)(2) of the revised 

proposed rule specifies less 

conservatism for the analyses and 

associated acceptance criteria for breaks 

larger than the TBS. LOCA analyses for 

break sizes equal to or smaller than the 

TBS would be applied to all locations in 

the RCS to find the limiting break 

location. LOCA analyses for break sizes 

larger than the TBS (but using the more 

realistic analysis requirements) would 

also be applied to all locations in the 

RCS to find the limiting break size and 

location. This analytical approach is 

consistent with current NRC regulatory 

positions and industry practice. 

1. Acceptable Methodologies and 

Analysis Assumptions 

Under existing § 50.46 requirements, 

prior NRC approval is required for ECCS 

evaluation models. Acceptable 

evaluation models are currently of two 

types; those that realistically describe 

the behavior of the RCS during a LOCA, 

and those that conform with the 

required and acceptable features 

specified in Appendix K to Part 50. 

Appendix K evaluation models 

incorporate conservatism as a means to 

justify that the acceptance criteria are 

satisfied by an ECCS design. In contrast, 

the realistic or best-estimate models 

attempt to accurately simulate the 

expected phenomena. As a result, 

comparisons to applicable experimental 

data must be made and uncertainty in 

the evaluation model and inputs must 

be identified and assessed. This is 

necessary so that the uncertainty in the 

results can be estimated so that when 

the calculated ECCS cooling 

performance is compared to the 

acceptance criteria, there is a high level 

of probability that the criteria would not 

be exceeded. Appendix K, Part II, 

contains the documentation 

requirements for evaluation models. All 

of these existing requirements are 

included in § 50.46a(e)(1) of the revised 

proposed rule for breaks at or below the 

TBS. 

As currently required under § 50.46, 

the ECCS analysis performed with a 

model other than one based on 

Appendix K must demonstrate with a 

high level of probability that the 


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acceptance criteria will not be exceeded. 

The position taken in RG 1.157 has been 

that 95 percent probability constitutes 

an acceptably high probability. Section 

50.46a(e)(1) of the revised proposed rule 

would retain the high level of 

probability as the statistical acceptance 

criterion. 

Revised proposed §§ 50.46a(e)(1) and 

(e)(2) would require that the worst break 

size and location be calculated 

separately for breaks at or below the 

TBS and for breaks larger than the TBS 

up to and including a double-ended 

rupture of the largest pipe in the RCS. 

Different methodologies, analytical 

assumptions, and acceptance criteria 

may be used for each break size region. 

Consistent with current § 50.46 

requirements, licensees would be 

required to analyze breaks at or below 

the TBS by assuming the worst single 

failure concurrent with a loss-of-offsite 

power, limiting operating conditions, 

and only crediting safety systems. For 

breaks larger than the TBS, licensees 

may take credit for operation of any 

equipment supported by availability 

data provided that onsite power (either 

safety or non-safety) can be reliably 

provided to that equipment through 

manual actions within a reasonable time 

after a loss of offsite power. All nonsafety equipment that is credited for 

analyses of breaks larger than the TBS 

would have to be identified as such and 

listed in the plant technical 

specifications. Analyses of breaks larger 

than the TBS could assume nominal 

operating conditions rather than 

technical specification limits. This 

would also include combining actual 

fuel burnup in decay heat predictions 

with the corresponding operating 

peaking factors at the appropriate time 

in the fuel cycle. The assumptions of 

loss-of-offsite power and the worst 

single failure would not be required 

because breaks larger than the TBS are 

very unlikely; therefore, less margin 

would be needed in the analysis of 

breaks in this region. A capability to 

provide onsite power to non-safety 

equipment in a reasonable time 

following a loss of offsite power (e.g. 

approximately 30 minutes) is a defensein-depth consideration for severe 

accident management. 

2. Acceptance Criteria 

ECCS acceptance criteria in 

§ 50.46a(e)(3) for breaks at or below the 

TBS would be the same as those 

currently required in § 50.46. Therefore, 

licensees would be required to use an 

approved methodology to demonstrate 

that the following acceptance criteria 

are met for the limiting LOCA at or 

below the TBS: 

• PCT less than 2200 °F; 

• Maximum local cladding oxidation 

(MLO) less than 17 percent; 

• Maximum hydrogen production— 

core wide cladding oxidation less than 

one percent; 

• Maintenance of coolable geometry; 

and 

• Maintenance of long-term cooling. 

Commensurate with the lower 

probability of occurrence, the 

acceptance criteria in § 50.46a(e)(4) for 

breaks larger than the TBS would be less 

prescriptive: 

• Maintenance of coolable geometry, 

and 

• Maintenance of long-term cooling. 

The revised proposed rule would 

allow licensees flexibility in 

establishing appropriate metrics and 

quantitative acceptance criteria for 

maintenance of coolable geometry. A 

licensee’s metrics and acceptance 

criteria must realistically demonstrate 

that coolable core geometry and longterm cooling will be maintained. Unless 

data or other valid justification criteria 

are provided, licensees should use 2200 

°F and 17 percent for the limits on PCT 

and MLO, respectively, as metrics and 

quantitative acceptance criteria for 

meeting the rule. Other less 

conservative criteria would be 

acceptable if properly justified by 

licensees. 

However, the NRC acknowledges that 

it would be expensive and timeconsuming for industry to develop the 

necessary experimental and analytical 

data to justify alternative acceptance 

criteria as a surrogate for demonstrating 

coolable geometry. Because of the 

difficulty in demonstrating alternative 

metrics, the NRC is requesting 

stakeholder comments on whether the 

final § 50.46a rule should retain the 

coolable geometry criterion for beyondTBS breaks. Retaining coolable 

geometry would give licensees the 

option to demonstrate alternative 

coolable geometry metrics or use the 

current metric (2200 °F PCT and 17 

percent MLO). If the NRC removed the 

coolable geometry criterion, the beyondTBS acceptance criteria would be the 

same as the acceptance criteria for TBS 

and smaller breaks (2200 °F PCT and 17 

percent MLO). The NRC will evaluate 

stakeholder comments on this question 

before deciding which beyond-TBS 

acceptance criteria to include in the 

final rule. 

As previously discussed in Section 

IV.C of this document, the NRC is 

working to revise the ECCS acceptance 

criteria in § 50.46(b) to account for new 

experimental data on cladding ductility 

and to allow for the use of advanced 

cladding alloys. The NRC will soon 

issue an ANPR seeking public 

comments on a planned regulatory 

approach. The NRC expects that this 

rulemaking (Docket ID NRC–2008–0332) 

will establish new cladding 

embrittlement acceptance criteria in 

§ 50.46(b) for design basis LOCAs. As 

these new acceptance criteria are 

established, the NRC will also make 

conforming changes to § 50.46a as 

necessary for both below and above TBS 

breaks. 

3. Restriction of Reactor Operation 

Section 50.46a(e)(5) would allow the 

Director of the Office of Nuclear Reactor 

Regulation to impose restrictions on 

reactor operation if it is determined that 

the evaluations of ECCS cooling 

performance are not consistent with the 

requirements for evaluation models and 

analysis methods specified in revised 

proposed § 50.46a(e)(1) through (e)(4). 

Non-compliance may be due to factors 

such as lack of a sufficient data base 

upon which to assess model 

uncertainty, use of a model outside the 

range of an appropriate data base, 

models inconsistent with the 

requirements of Appendix K of Part 50, 

or phenomena unknown at the time of 

approval of the methodology. Lack of 

compliance with methodological 

requirements would not necessarily 

result in failure to meet the acceptance 

criteria of revised proposed 

§§ 50.46a(e)(3) and (e)(4), but, rather, 

would provide results that could not be 

relied upon to demonstrate compliance 

with the appropriate acceptance criteria. 

Thus, depending upon the specific 

circumstances, it might be necessary for 

the NRC to impose restrictions on 

operation until these issues are 

resolved. This requirement is included 

in the revised proposed rule for 

consistency with the current ECCS 

regulations, because it is comparable to 

existing § 50.46(a)(2). 

E. Risk-Informed Changes to the 

Facility, Technical Specifications, or 

Procedures 

Licensees who adopt § 50.46a would 

use a risk-informed evaluation process 

to demonstrate, before implementation, 

that facility changes will satisfy the riskinformed acceptance criteria in revised 

proposed § 50.46a(f). Changes that must 

be evaluated are specified in revised 

proposed § 50.46a(d)(3) and would 

include all ‘‘enabled’’ changes that 

satisfy the alternative ECCS analysis 

requirements in § 50.46a but do not 

satisfy the current ECCS analysis 

requirements in § 50.46. Also, changes 

in risk from facility changes not enabled 

by the alternative ECCS requirements 

could be combined with changes in risk 


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from facility changes enabled by 

§ 50.46a if the licensee chooses to 

combine the changes in its application 

of the risk-informed change process 

defined in the rule. In this case, the 

changes made under § 50.46a would 

include those enabled by § 50.46a and 

those not enabled by § 50.46a but 

included in the risk-informed 

application. 

Licensees would be required to 

periodically maintain and upgrade the 

PRA used in the risk assessments and 

ensure that over time all changes made 

under § 50.46a continue to meet the 

risk-informed acceptance criteria. If 

necessary, revised proposed 

§ 50.46a(g)(2) would require the licensee 

to propose steps and a schedule to bring 

the facility back into compliance with 

the acceptance criteria in 

§ 50.46a(f)(2)(ii) or § 50.46a(f)(2)(iii), as 

applicable. 

The risk-informed evaluation would 

be required to demonstrate that 

increases in plant risk (if any) meet 

appropriate risk acceptance criteria, 

defense-in-depth is maintained, 

adequate safety margins are maintained, 

and adequate performance-measurement 

programs are implemented. The NRC 

believes that all changes to a plant, its 

technical specifications, or its 

procedures which are based upon the 

analyses of ECCS performance 

permitted under § 50.46a(e)(2)—with 

the exception of those changes 

permitted under § 50.46a(f)(1)—must be 

reviewed and approved by the NRC for 

two reasons. First, a wide range of 

changes could be implemented under 

§ 50.46a, which, if improperly 

implemented by licensees, could result 

in significant adverse impacts on public 

health and safety or common defense 

and security. NRC review and approval 

would provide verification that a 

licensee has properly evaluated each 

proposed change against the acceptance 

criteria in § 50.46a. Second, changes 

involving technical specifications must 

receive NRC review and approval in the 

form of a license amendment, as 

required by the Atomic Energy Act of 

1954, as amended. Accordingly, the 

NRC’s revised proposed rule would 

require NRC review and approval of all 

changes initiated under § 50.46a(f)(2). 

1. Requirements for the Risk-Informed 

Evaluation 

The revised proposed rule is based 

upon the regulatory premise that the 

acceptability of all licensee-initiated 

changes made under the rule should be 

judged in a risk-informed manner. The 

risk-informed assessment process must 

include methods for evaluating 

compliance with the risk criteria, 

defense-in-depth criteria, safety margin 

criteria, and performance measurement 

criteria in § 50.46a(f). These attributes 

have been identified by the Commission 

as a necessary set of risk evaluation 

tools to ensure that changes to the 

facility do not endanger public health 

and safety. 

Compliance with the risk criteria 

plays a key role in the regulatory 

structure of the proposed rule. A riskassessment must be used to determine 

the change in risk associated with 

facility changes. Inasmuch as PRA 

methodologies are generally recognized 

as the best current approach for 

conducting risk assessments suitable for 

making decisions in areas of potential 

safety significance, § 50.46a(f)(4) of the 

revised proposed rule would require 

that a technically adequate PRA be used 

in demonstrating compliance with the 

requirements of § 50.46a that would 

affect the regulatory decision in a 

substantive manner. However, the NRC 

recognizes that non-quantitative PRA 

assessment methodologies and 

approaches could also be used to 

complement or supplement the 

quantitative aspects of a PRA, especially 

when performance of a quantitative PRA 

methodology of the level needed to 

support a particular decision is not 

justifiable because the safety 

significance of the decision does not 

warrant the level of technical 

sophistication inherent in a PRA. 

Accordingly, § 50.46a(f)(5) is written to 

recognize that non quantitative risk 

assessment may also be utilized. 

a. Probabilistic Risk Assessment 

Requirements 

Sections 50.46a(f)(4)(i) through (iv) set 

forth the four general attributes of an 

acceptable PRA for the purposes of this 

rule. Section 50.46a(f)(4)(i) would 

require that the PRA address initiating 

events from internal and external 

sources, and for all modes of operation, 

including low power and shutdown, 

that would affect the regulatory decision 

in a substantial manner. Failure to 

consider sources of risk from internal 

and external events, or from anticipated 

operating modes, could result in an 

inaccurate characterization of the level 

of risk associated with a plant change. 

Therefore, initiating events from 

internal and external sources and during 

all modes of operation would have to be 

considered by the PRA when the change 

in risk would affect the regulatory 

decision, in order to ensure that the 

effect on risk from licensee-initiated 

changes is adequately characterized in a 

manner sufficient to support a 

technically defensible determination of 

the level of risk. 

Section 50.46a(f)(4)(ii) states that the 

PRA must reasonably represent the 

current configuration and operating 

practices at the plant. A plant’s risk may 

vary as plant configuration and/or plant 

procedures change. Failure to update 

the PRA based upon these configuration 

or procedure changes may result in 

inaccurate or invalid PRA results. 

Accordingly, to ensure that estimates of 

risk adequately reflect the facility for 

which a decision must be made, the rule 

would require that the PRA address 

current plant configuration and 

operating practices. 

Section 50.46a(f)(4)(iii) would require 

that the PRA have ‘‘sufficient technical 

adequacy’’ including consideration of 

uncertainty, as well as a sufficient level 

of detail to provide confidence that the 

calculated risk and the changes in risk 

adequately reflect the proposed facility 

change. The revised proposed rule 

would require the PRA to consider 

uncertainty because the decision maker 

must understand the limitations of the 

particular PRA that was performed to 

ensure that the decision is robust and 

accommodates relevant uncertainties. 

With respect to level of detail, failure to 

model the plant (or relevant portion of 

the plant) at the appropriate level of 

detail may result in calculated risk 

values that do not appropriately capture 

the risk significance of the proposed 

change. 

Finally, § 50.46a(f)(4)(iv) would 

require that, to the extent that the PRA 

is used, the PRA must meet NRCapproved industry standards. The NRC 

has prepared a regulatory guide (RG 

1.200) on determining the technical 

adequacy of PRA results for riskinformed activities. As one step in the 

assurance of technical quality, the PRA 

would be subjected to a peer review 

process assessed against an industry 

standard or set of acceptance criteria 

that is endorsed by the NRC. Industry 

standards for all initiators and operating 

modes are under development but not 

yet complete. The NRC will develop 

review guidelines that endorse criteria 

for considering the sufficiency of a PRA 

peer review process for this application 

in § 50.46(c) if this guidance becomes 

necessary before industry standards 

have been completed and endorsed in 

RG 1.200. 

b. Requirements for Risk Assessments 

Other Than PRA 

Risk assessment need not always be 

performed using PRA. The rule 

explicitly recognizes the possibility of 

using risk assessment methods other 

than PRA to demonstrate compliance 

with various acceptance criteria in the 

rule. However, as with PRA 


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methodologies, the NRC believes that 

minimum quality requirements for 

PRAs and risk assessments used by a 

licensee in implementing the rule must 

be established. Accordingly, 

§ 50.46a(f)(5) would establish the 

minimum requirement for risk 

assessment methodologies other than 

PRA. The NRC believes that this 

requirement provides flexibility to 

licensees to use the non-PRA risk 

methodology (or combination of 

different methodologies) when these 

methodologies produce results that are 

sufficient upon which to base decisions 

that the various acceptance criteria in 

the proposed rule have been met. 

2. Aggregation of Plant Changes When 

Evaluating Changes in Risk 

Licensees often make changes to the 

facility, technical specifications, and 

procedures. Some changes that the 

licensees could make after adopting this 

rule would not have been permitted 

without the new § 50.46a (related or 

enabled changes). Other changes would 

be unrelated insofar as the basis of the 

changes and NRC approval, when 

necessary, will rely on regulations, 

guidelines, or facility priorities that do 

not depend on the new ECCS 

requirements in Section 50.46a. 

Unrelated changes will indirectly 

influence the change in risk of the 

§ 50.46a related changes insofar as they 

change the risk profile of the facility. If 

unrelated changes are combined with 

related changes in determining the 

§ 50.46a change in risk estimates 

(bundling), the result will normally be 

different than if the unrelated changes 

are considered as part of the baseline 

risk associated with the current design 

and operation of the facility. If bundling 

is permitted, a licensee could 

implement facility changes that would 

decrease risk to offset increased risk 

from § 50.46a enabled changes. These 

changes would increase the safety of the 

facility and are expected to result in a 

reallocation of resources to areas where 

safety can be improved. Current NRC 

practice, consistent with RG 1.174, is to 

compare the total or cumulative risk 

increase from all related changes, and 

only related changes, to the acceptance 

guidelines. RG 1.174 does, however, 

permit bundling changes (referred to as 

combined changes in RG 1.174) and 

provides additional acceptance 

guidelines that must be met when 

permitting unrelated plant changes that 

might decrease risk to be combined 

together with a group of related changes 

in a change in risk estimate that would 

be compared to the acceptance 

guidelines. 

The NRC believes that allowing 

bundling of unrelated changes into the 

§ 50.46a change in risk estimates will 

encourage licensees to use risk-informed 

methods to take advantage of 

opportunities to reduce risk, and not 

just eliminate requirements that a 

licensee deems as undesirable. 

However, in some situations, bundling 

could mask the creation of significant 

risk outliers. To ensure that outliers are 

not created, and that the additional 

guidelines in RG 1.174 are appropriately 

applied, the rule would not permit 

bundling of changes without previous 

review and approval. Therefore, the 

revised, proposed § 50.46a(f)(2)(iv) 

would allow changes not enabled by 

§ 50.46a to be combined with changes 

enabled by § 50.46a in the calculation of 

the change in risk when a licensee 

submits an application for a change 

under 50.90. 

3. NRC Approval of a Licensee Process 

for Making Changes to a Licensee’s 

Facility or Procedures Without NRC 

Review and Approval 

As a general matter, the licensee must 

obtain NRC review and approval 

(through a license amendment 

application) for any changes to the 

facility, technical specifications, or 

procedures that may be implemented 

under this section. However, the NRC 

believes that there is a subset of plant 

and procedure changes that would be 

made possible by § 50.46a involving 

minimal changes in risk which also 

have no significant impact upon 

defense-in-depth capabilities. Prior NRC 

review and approval of these changes on 

an individual basis would be 

unnecessary if the NRC has previously 

concluded that the licensee has an 

adequate technical process for 

appropriately identifying this subset of 

changes. In the NRC’s view, plant 

changes which involve minimal changes 

in risk and have no significant impact 

upon defense-in-depth (and do not 

involve a change to the license), by 

definition, do not result in significant 

issues involving public health and 

safety or common defense and security. 

Expending licensee resources to 

prepare an application for approval of 

plant changes involving minimal 

changes in risk and NRC resources to 

review and approve these applications 

is not an efficient use of resources. 

Rather, the NRC believes that if it 

reviews and approves in advance the 

licensee’s processes (including the 

adequacy of the licensee’s PRA and 

other risk assessment methods) and 

criteria for identifying changes which 

are both minimal from a risk standpoint 

and do not significantly affect defensein-depth or plant physical security, then 

there is no need to review and approve 

each of the changes individually. 

Further, the NRC believes that these 

minimal changes are unlikely to impact 

the built-in capability of the facility to 

resist security threats. Accordingly, the 

NRC has proposed an approach in 

§ 50.46a(f)(1) allowing a licensee to 

obtain ‘‘pre-approval’’ of a process for 

identifying minimal plant and 

procedure changes made possible under 

§ 50.46a. 

The revised proposed § 50.46a(f)(1) 

states that a licensee may make changes 

based upon the provisions of this 

section without prior review and 

approval if the stated requirements in 

paragraphs (f)(1) and (f)(3) of this 

section are met. The revised proposed 

rule also states that the provisions of 

§ 50.59 would apply. Licensees with a 

pre-approved change process would be 

allowed to make facility changes 

without NRC approval if they met 

§ 50.59 and § 50.46a requirements. 

Compliance with the § 50.59 

requirements is necessary to ensure that 

facility changes made without NRC 

approval do not result in plant 

conditions that could impact public 

health and safety. Compliance with the 

§ 50.46a(f) requirements for risk 

assessments is required to ensure that 

facility changes result in acceptable 

changes in risk, adequate defense-indepth, that safety margins will be 

maintained, and that adequate 

performance-measurement programs are 

implemented. 

4. Risk Acceptance Criteria for Plant 

Changes 

Sections 50.46a(f)(2)(ii) and (f)(2)(iii) 

would require that the total increases in 

risk are very small and that the overall 

plant risk remains small. Two sets of 

metrics are used to measure risk 

depending on when the applicant’s 

operating license was issued. For 

reactors licensed before the effective 

date of the rule, § 50.46a(f)(2)(ii) would 

apply and CDF and LERF would be 

used. For new reactors licensed after the 

effective date of the rule, 

§ 50.46a(f)(2)(iii) would apply and CDF 

and large release frequency (LRF) are 

used. The NRC believes that this 

requirement is a necessary element for 

ensuring that changes which would be 

permitted by the revised § 50.46a ECCS 

analyses do not result in a greater 

change in risk than intended by the 

Commission. 

a. Risk Estimate 

To satisfy the Commission’s 

requirements in §§ 50.46a(f)(2)(ii) and 

(f)(2)(iii) that the total increases in risk 


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are very small would require that the 

change in risk for each facility change 

be evaluated and shown to meet the 

acceptance guidelines. If a series of 

changes are made over time, 

§ 50.46a(f)(2)(iv) would require that 

cumulative effect of these changes be 

evaluated and shown to meet the 

acceptance criteria. Section 

50.46a(f)(2)(iv) would also permit 

changes in risk from facility changes not 

enabled by § 50.46a to be combined by 

the licensee with facility changes that 

are enabled by this section for the 

purposes of meeting the acceptance 

guidelines. The total change in risk from 

all facility changes made under the rule 

after the adoption of § 50.46a must be 

evaluated and compared to the ‘‘very 

small’’ acceptance criterion before each 

change requiring a risk-informed 

evaluation and after the periodic PRA 

maintenance and upgrading. Requiring 

that the total change in risk from all 

facility changes made under the rule 

after the adoption of § 50.46a be 

compared to the § 50.46a acceptance 

criteria instead of allowing the changes 

in risk to be partitioned and 

individually compared to the 

acceptance criteria would ensure that 

the total risk increase of all changes, as 

they are implemented over time, would 

not constitute more than a very small 

increase in risk. If the total increase in 

the applicable risk metrics were not 

compared to the acceptance criteria, a 

number changes where every individual 

change’s risk increase is kept below the 

proposed rule’s risk acceptance criteria 

could, considered cumulatively, result 

in a significant increase in risk. A 

significant increase would not satisfy 

the Commission’s criteria that the 

overall plant risk remains small. Also, 

comparing the risk increase from each 

change to the acceptance criteria 

independently of all previous changes 

would render the use of the ‘‘very 

small’’ criterion inadequate to monitor 

and control increases in risk from a 

series of plant changes implemented 

over time. 

Comparing the total risk increase to 

the risk increase criterion, and allowing 

bundling of unrelated changes in the 

change in risk estimate, will support the 

NRC’s philosophy that, consistent with 

the principles of risk-informed 

integrated decision making, licensees 

should have a risk management 

philosophy in which risk insights are 

not just used to systematically increase 

risk, but also to help reduce risk where 

appropriate and where it is shown to be 

cost effective. 

b. Acceptance Criteria 

In § 50.46a(f)(2)(ii), CDF and LERF are 

used as surrogates for early and latent 

health effects, which are used in the 

Commission’s Policy Statement on 

Safety Goals (51 FR 30028; August 4, 

1986). The NRC has used CDF and LERF 

in making regulatory decisions for over 

20 years. The NRC endorsed the use of 

CDF and LERF as appropriate measures 

for evaluating risk and ensuring safety 

in nuclear power plants when it 

adopted RG 1.174 in 1997. After the 

adoption of RG 1.174, the NRC has had 

eleven years of experience in applying 

risk-informed regulation to support a 

variety of applications, including 

amending facility procedures and 

programs (e.g., IST and ISI programs), 

amending facility operating licenses 

(e.g., power up-rates, license renewals, 

and changes to the FSAR), and 

amending technical specifications. On 

the basis of this experience, for current 

operating reactors, the NRC has 

determined that CDF and LERF are 

acceptable measures for evaluating 

changes in risk as the result of changes 

to a facility, technical specifications, 

and procedures, with the exception of 

certain changes that affect containment 

performance but do not affect CDF or 

LERF. Changes that affect containment 

performance are considered as part of 

the defense-in-depth evaluation. 

For new reactors, CDF and LRF 

(instead of LERF) would apply as 

indicated in § 50.46a(f)(2)(iii). For new 

reactor licensing the Commission has 

established a goal based on LRF (see 

SRM on SECY–89–102— 

Implementation of the Safety Goals, 

June 15, 1990; and SRM on SECY–90– 

016—Evolutionary Light Water Reactor 

(LWR) Certification Issues and Their 

Relationship to Current Regulatory 

Requirements, June 26, 1990). 

The Commission has concluded that 

changes under this rule should be 

restricted to very small risk increases. 

As discussed in RG 1.174, a very small 

risk increase is independent of a plant’s 

overall risk as measured by the current 

CDF and LERF. Increases in CDF of 

10¥6 per reactor year or less, and 

increases in LERF of 10¥7 per reactor 

year or less are very small risk increases 

for existing reactor facilities. 

For new reactors, the same CDF 

metric is used and the same definition 

of very small increase (i.e., less than 

10¥6 per reactor year) would be used. 

The revised proposed rule uses LRF 

instead of LERF as a metric for new 

reactors. RG 1.174 provides no 

guidelines for LRF. The Commission has 

approved the overall mean frequency of 

a large release of radioactive material to 

the environment (LRF) to be less than 

10¥6 per reactor year. The revised 

proposed rule requires the total increase 

in LRF to be no more than very small. 

The NRC proposes that increases in LRF 

of 10¥8 per reactor year or less are very 

small risk increases for new reactors. 

Because of the difference between the 

LERF acceptance criteria for existing 

reactors and the LRF acceptance criteria 

for new reactors, the NRC is seeking 

specific public comments on this topic. 

Additional background information on 

how the NRC is addressing this issue 

and how the NRC is soliciting public 

input on this topic in this revised 

proposed rule and in other regulatory 

areas is provided in Section J.2. of this 

document. 

After adopting RG 1.174 in 1997, the 

NRC has applied the quantitative 

change in risk guidelines to individual 

plant changes and to sequences of plant 

changes implemented over time. The 

NRC has found these guidelines and the 

CDF and LERF values (when used 

together with the defense in depth, 

safety monitoring, and performance 

measurement criteria) are capable of 

differentiating between changes, and 

sequences of changes, that are not 

expected to endanger public health and 

safety from those that might. The NRC 

believes that applying the LRF guideline 

for determining very small risk 

increases would also be protective of 

public health and safety. 

Section 50.46a(f)(1) would permit 

licensees to make changes under this 

provision without prior review and 

approval if the changes involve minimal 

increases in risk which also have no 

significant impact upon defense-indepth capabilities. A minimal risk 

increase is one which, when considered 

qualitatively by itself or in combination 

with all other minimal increases, would 

never become significant. Logically, a 

minimal increase is less than the very 

small increase in CDF and in LERF, and 

was chosen as an increase of less than 

10¥7 per reactor year for CDF and an 

increase in LERF of less than 10¥8 per 

reactor year. Similarly, for new reactor 

licensing, an increase in LRF less than 

10¥9 per reactor year is a minimal 

increase. Although ten of these changes 

could cause the combination of minimal 

increases to exceed the very small 

criteria, the NRC believes that most of 

these changes will have a much smaller 

(and, in some cases, an unmeasurable) 

increase in risk. Regardless of whether 

a licensee makes changes under 

§ 50.46a(f)(1) instead of § 50.46a(f)(2), 

the total cumulative risk including all 

the individually minimal risk increases 

as well as any increases approved by the 

NRC under § 50.46a(f)(2), would have to 


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be considered in the periodic reporting 

required by § 50.46a(g)(2). If a licensee 

implements an unexpectedly large 

number of minimal risk changes, the 

periodic reporting requirements in 

§ 50.46a(g)(2) would provide adequate 

notice to ensure that the NRC is aware 

of potentially significant changes (or 

any collective impact), so that the NRC 

may undertake additional oversight 

actions as deemed necessary and 

appropriate. 

Additionally, although the revised 

proposed rule would permit licensees to 

make plant changes that result in very 

small risk increases, the NRC is 

requesting stakeholder comments on 

whether the rule should allow plant 

changes that increase risk at all. Instead 

of the risk acceptance criteria allowing 

very small risk increases, should the 

risk acceptance criteria in final rule 

require that the net effect of plant 

changes made under § 50.46a be risk 

neutral or risk beneficial? The NRC 

requests stakeholders to provide 

comments on the use of risk acceptance 

criteria that would not allow a 

cumulative increase in risk for plant 

changes made under § 50.46a. 

5. Defense-in-Depth 

Section 50.46a(f)(3)(i) would require 

that the risk-informed evaluation 

demonstrate that defense-in-depth is 

maintained. Defense-in-depth is an 

element of the NRC’s safety philosophy 

that employs successive measures to 

prevent accidents or mitigate damage if 

a malfunction, accident, or naturally 

caused event occurs at a nuclear facility. 

As conceived and implemented by the 

NRC, defense-in-depth provides 

redundancy in addition to a multiple 

barrier approach against fission product 

releases. Defense-in-depth continues to 

be an effective way to account for 

uncertainties in equipment and human 

performance. The NRC has determined 

that retention of adequate defense-indepth must be ensured in all riskinformed regulatory activities. 

6. Safety Margins 

Section 50.46a(f)(3)(ii) would require 

that adequate safety margins be retained 

to account for uncertainties. These 

uncertainties include phenomenology, 

modeling, and how the plant was 

constructed or is operated. The NRC’s 

concern is that plant changes could 

inappropriately reduce safety margins, 

resulting in an unacceptable increase in 

risk or challenge to plant SSCs. This 

provision would ensure that an 

adequate safety margin exists to account 

for these uncertainties, such that there 

are no unacceptable results or 

consequences (e.g., structural failure) if 

an acceptance criterion or limit is 

exceeded. 

7. Performance Measuring Programs 

Section 50.46a(f)(3)(iii) would require 

that adequate performance measurement 

programs and feedback strategies be 

implemented to ensure that the riskinformed evaluation continues to reflect 

actual plant design and operation. The 

risk-informed evaluation includes the 

risk assessment, maintenance of 

defense-in-depth, and adequacy of 

safety margins. Results from 

implementation of monitoring and 

feedback strategies can provide an early 

indication of unanticipated degradation 

of performance of plant elements that 

may invalidate the demonstration by the 

risk-informed evaluation that the change 

satisfied all the acceptance criteria. This 

section would require that the 

monitoring programs be designed to 

detect degradation of SSCs before plant 

safety is compromised. Permitting 

degradation to advance until plant 

safety could be compromised would be 

inconsistent with the NRC’s regulatory 

responsibility of protecting public 

safety. The NRC expects that licensees 

will integrate existing programs for 

monitoring equipment performance and 

other operating experience on their site 

and throughout industry with the 

performance measuring programs 

required by this section. 

F. Operational Requirements 

The revised proposed rule includes 

five specific operational requirements 

that apply to licensees who are 

approved to implement § 50.46a. These 

requirements are set forth in § 50.46a(d) 

and would remain in effect as long as 

the facility is subject to the § 50.46a 

alternative ECCS requirements until 

such time as the licensee permanently 

ceases operations by submitting the 

decommissioning certifications required 

under § 50.82(a). They are: 

1. Maintain ECCS models and/or 

analysis methods that demonstrate 

compliance with the ECCS acceptance 

criteria. 

2. Maintain reactor coolant leak 

detection equipment available at the 

facility and identify, monitor, and 

quantify leakage to ensure that adverse 

safety consequences do not result from 

leakage from piping and components 

larger than the transition break size. 

3. Perform a risk-informed evaluation 

for each potentially risk-significant 

change (or group of changes) to the 

facility enabled by § 50.46a. 

4. Periodically assess the cumulative 

effect of changes to the plant, 

operational practices, equipment 

performance, and plant operational 

experience. 

5. Do not operate the plant for more 

than fourteen days in any 12 month 

period in an at-power operating 

configuration that has not been 

demonstrated to meet the ECCS 

acceptance criteria for breaks larger than 

the TBS. 

Each of the five operational 

requirements is discussed in detail 

below. 

1. Maintain ECCS models and/or 

analysis methods that demonstrate 

compliance with the ECCS acceptance 

criteria. 

Calculated results of licensee ECCS 

models and/or analysis methods must 

demonstrate compliance with the ECCS 

acceptance criteria throughout the 

operating lifetime of the plant. 

Licensees must also update ECCS 

models and/or analysis methods by 

modifying them as needed to address 

any plant design changes affecting ECCS 

performance during this time period. 

2. Maintain reactor coolant leak 

detection equipment available at the 

facility and identify, monitor, and 

quantify leakage to ensure that adverse 

safety consequences do not result from 

leakage from piping and components 

larger than the transition break size. 

In a Staff Requirements Memorandum 

dated August 10, 2007, responding to 

SECY–07–0082—‘‘Rulemaking To Make 

Risk Informed Changes to Loss-ofCoolant Accident Technical 

Requirements; 10 CFR 50.46a, 

‘Alternative Acceptance Criteria for 

Emergency Core Cooling Systems for 

Light-Water Nuclear Power Reactors’ ’’, 

the Commission directed the NRC staff 

to evaluate various approaches for 

enhancing the 10 CFR 50.46a rule with 

requirements for improved leak 

detection methods. This SRM also 

directed the NRC staff to ‘‘strengthen the 

assurance of defense-in-depth [provided 

by the § 50.46a rule] for breaks beyond 

the transition break size (TBS).’’ 

In response to a recommendation 

made by the Davis-Besse Lessons 

Learned Task Force (DBLLTF), (see 

memorandum from Arthur T. Howell to 

William F. Kane, ‘‘Degradation of the 

Davis-Besse Nuclear Power Station 

Reactor Pressure Vessel Head LessonsLearned Report; September 30, 2002; 

ADAMS Accession No. ML022740211) 

the NRC evaluated whether it should 

impose new requirements on licensees 

in the areas of tighter reactor coolant 

leakage limits and new leakage 

monitoring requirements. Specifically, 

the DBLLTF Recommendation 3.1.5(1) 

said that the NRC should determine 

whether PWR plants should install online enhanced leakage detection systems 


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on critical plant components which 

would be capable of detecting leakage 

rates of significantly less than 1 gallon 

per minute. 

The evaluation identified techniques 

that could improve localized leak 

detection and on-line monitoring and 

several areas of possible improvements 

to leakage detection requirements that 

could provide increased confidence that 

plants are not operated at power with 

reactor coolant pressure boundary 

leakage. Although the NRC concluded 

that there was not a sufficient basis to 

require reduced technical specification 

leakage for existing licensees, the NRC 

recommended updating Regulatory 

Guide 1.45 on leak detection. This RG 

was revised in 2008. 

RG 1.45, Revision 1 incorporates 

progress in reactor coolant pressure 

boundary leakage detection technology; 

addresses the effect on radiation 

monitoring, and, subsequently, on leak 

detection from reduced activity levels of 

coolant resulting from improved fuel 

integrity; and incorporates lessons 

learned from operating experience. The 

title of the Regulatory Guide 1.45, 

Revision 1, has been changed from 

‘‘Reactor Coolant Pressure Boundary 

Leakage Detection Systems’’ to 

‘‘Guidance on Monitoring and 

Responding to Reactor Coolant System 

Leakage,’’ to reflect its broader scope. 

Revision 1 provides detailed guidance 

for timely detection and location of 

leaks, continuous monitoring, 

quantifying and trending of leak rates, 

assessing safety significance, and 

specifying plant actions following 

confirmation of an adverse trend in 

unidentified leak rate. Revision 1 

describes acceptable leakage detection 

systems and methods, using riskinformed and performance-based 

criteria to the extent practical. It retains 

the recommendations for monitoring of 

sump level or flow, airborne particulate 

activity, and condensate flow rate from 

air coolers. Other supplementary 

detection methods are recommended for 

use where and when appropriate. 

Paragraph 50.46a(d)(2) in the revised 

proposed rule contains new enhanced 

leak detection requirements. Enhanced 

leak detection is expected to provide 

increased defense-in-depth against large 

pipe breaks for licensees who 

implement the alternative ECCS rule. 

The NRC has concluded that 

implementing the guidance in 

Regulatory Guide 1.45, Revision 1, by 

licensees choosing to comply with 10 

CFR 50.46a will result in improved 

monitoring and response to leaks in the 

reactor coolant system and will provide 

an acceptable method to satisfy the 

requirements of Section 50.46a(d)(2). 

3. Perform a risk-informed evaluation 

for each change (or group of changes) to 

the facility enabled by § 50.46a. 

In addition to meeting all other 

applicable requirements, a riskinformed evaluation required by 

§ 50.46a(d)(3) would have to be 

performed for changes enabled by 

§ 50.46a. If a licensee has a change 

methodology that was submitted under 

§ 50.46a(f)(1) and approved by the NRC, 

that licensee could make some changes 

without NRC approval, if the acceptance 

criteria in § 50.46a(f)(1) are met. 

Otherwise, the licensee would be 

required to submit the results of its riskinformed evaluation for prior NRC 

review and approval in a license 

amendment request subject to the 

requirements of § 50.90. The licensee 

would have to retain the results of all 

risk-informed evaluations made under 

§ 50.46a(f)(1) and periodically submit a 

summary of the results to the NRC as 

required under § 50.46a(g)(3). 

4. Periodically assess the cumulative 

effect of changes to the facility. 

Key components of risk-informed 

regulation are the monitoring of changes 

in plant risk and feedback to the risk 

assessment and/or plant design 

activities and processes which are the 

subject of the risk assessment. Section 

50.46a(d)(4) would require that after 

adopting § 50.46a, a licensee would be 

required to periodically maintain and 

upgrade the risk assessments (both PRA 

and non-PRA) required under 

§§ 50.46a(f)(4) and (f)(5). In particular, it 

is necessary that the PRA be maintained 

to reflect all plant changes; such as 

modifications, procedure changes, or 

changes in plant performance data. This 

maintenance enables the licensee to 

demonstrate that the total increases in 

CDF and LERF (or LRF for new reactors) 

after adopting § 50.46a continue to meet 

the acceptance criteria in § 50.46a(f)(2). 

The risk assessments would have to 

continue to meet the minimum quality 

requirements in §§ 50.46a(f)(4) and (f)(5) 

to support reasoned decision making 

under the rule. 

The revised proposed rule would 

specify that the maintenance and 

upgrading be conducted periodically 

‘‘but no less often than once every two 

refueling outages.’’ The NRC believes 

that this is an appropriate period 

because the uncertainty of risk changes 

occurring during the two refueling 

outage period is tolerable and unlikely 

to result in high risk situations 

developing as a result of the 

implementation of plant changes. The 

NRC’s determination is based upon the 

stringent acceptance criteria governing 

changes made under § 50.46a, as well as 

the existing deterministic criteria in the 

substantive technical requirements in 

Part 50 and the criteria utilized in 

determining the acceptability of plant 

changes. The updating period specified 

in the rule is also comparable to other 

NRC requirements governing updating 

and reporting of safety information, e.g., 

§§ 50.59, 50.71(e). 

If the assessment of the cumulative 

effect of changes made under the rule 

demonstrates that the acceptance 

criteria in § 50.46a(f)(2) are not met, 

§ 50.46a(g)(2) would require the licensee 

to develop steps and a schedule to bring 

the facility design and operation back 

into compliance with the acceptance 

criteria. These actions may include (but 

are not limited to) corrections to the risk 

analyses to demonstrate compliance, 

implementation of facility changes to 

offset adverse changes in risk, or 

reversal of changes previously made 

under the provisions of § 50.46a(f). The 

NRC believes that this requirement 

provides appropriate flexibility for the 

licensee to determine the actions 

necessary to ensure continued 

compliance with the § 50.46a(f) 

acceptance criteria, and is consistent 

with the concept of performance-based 

regulation. 

5. Do not operate the plant for more 

than a total of fourteen days in any 12 

month period in an operating 

configuration that has not been 

demonstrated to meet the ECCS 

acceptance criteria for breaks larger than 

the TBS. 

As previously discussed in the 

supplementary information of this 

document, the NRC has included 

restrictions in the revised proposed rule 

on plant operation in configurations 

where licensees have not demonstrated 

that LOCAs larger that the TBS will be 

mitigated. The initial proposed rule 

(November 2005) would have 

completely prohibited at-power 

operation in any configuration without 

the demonstrated ability to mitigate a 

beyond-TBS LOCA. The revised 

proposed rule would restrict operation 

in such a configuration to not exceed 

fourteen days in any twelve month 

period. The NRC believes it is unlikely 

that licensees will experience 

circumstances where they would 

consider operating in such a condition 

for more than fourteen days, but has 

concluded that the establishing a limit 

on the allowable time is necessary to 

support the defense-in-depth 

philosophy. Even though the LOCA 

frequencies on which the TBS is 

founded indicate that the expected 

frequency of breaks larger than the TBS 

is low, the restriction is needed because 

there are large uncertainties associated 

with these frequency estimates. The 


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Commission concluded that the 

consequences of a challenge to the 

facility from an unmitigated break larger 

than the TBS are severe enough to 

warrant some confidence that the break 

could be mitigated. Thus the revised 

proposed rule will limit the allowed 

time period for operation in an 

unanalyzed condition to fourteen days 

in any twelve month period to ensure 

that mitigation capability is maintained 

except for occasional brief periods long 

enough to perform online maintenance 

of mitigation structures, systems and 

components. 

G. Reporting Requirements 

1. ECCS Analysis Reporting 

Requirements 

Section 50.46a(g)(1) sets forth 

reporting requirements with respect to 

changes or errors in LOCA evaluation 

models. For each change to or error 

discovered in an ECCS evaluation 

model or analysis method or in the 

application of such a model that affects 

the calculated results, the licensee shall 

report the nature of the change or error 

and its estimated effect on the limiting 

ECCS analysis to the NRC at least 

annually as specified in § 50.4. If the 

change or error is significant, the 

licensee shall provide this report within 

30 days and include with the report a 

proposed schedule for providing a 

reanalysis or taking other action as may 

be needed to show compliance with 

§ 50.46a requirements. The 30 day 

period ensures sufficient time for the 

licensee to complete its evaluation and 

explanation of the changes and 

determine the course of action necessary 

to address compliance issues. For breaks 

smaller than the TBS a significant 

change is one which results in a 

calculated peak fuel cladding 

temperature different by more than 50 

degrees Fahrenheit from the 

temperature calculated for the limiting 

transient using the last acceptable 

model, or is a cumulation of changes 

and errors such that the sum of the 

absolute magnitudes of the respective 

temperature changes is greater than 50 

degrees Fahrenheit. This requirement is 

the same as in § 50.46. The NRC will 

also apply these reporting criteria to 

LOCAs involving pipe breaks larger 

than the TBS unless a specific 

alternative is proposed by a licensee and 

is approved by the NRC. 

2. Risk Assessment Reporting 

Requirements 

Section 50.46a(g)(2) would set forth 

reporting requirements with respect to 

the PRA maintenance and upgrading 

that would be required by § 50.46a(d)(4). 

When updating and upgrading the PRA, 

§ 50.46a(g)(2) would require the licensee 

to report changes to the NRC within 60 

days if the acceptance criteria in 

§§ 50.46a(f)(2)(ii) or (f)(2)(iii) (for new 

reactors) are exceeded. This provision 

would also require the report to include 

a schedule for implementation of any 

corrective actions necessary to bring 

plant operation or design back into 

compliance with the acceptance criteria. 

The 60-day period would ensure 

sufficient time for the licensee to 

complete its evaluation and explanation 

of the changes and determine the course 

of action necessary to address adverse 

changes in risk, while not unduly 

delaying the report to the NRC and 

thereby delaying NRC oversight. The 

NRC believes it should be informed of 

the licensee’s implementation schedule 

so the NRC can ensure that the licensee 

takes corrective action on a timely basis, 

consistent with the safety significance of 

the change. 

Section 50.46a(g)(3) would require 

periodic reports of changes that required 

a risk-informed evaluation under 

§ 50.46a(d)(3) and were implemented 

without prior NRC approval under 

paragraph (f)(1) of this section. This 

process is comparable in many respects 

to the § 50.59 process which requires 

similar reports. 

H. Documentation Requirements 

Section 50.46a(h) of the revised 

proposed rule would require that 

licensees maintain records sufficient to 

demonstrate compliance with § 50.46a 

requirements. When making plant 

changes under § 50.46a(f) and when 

updating its PRA and/or other risk 

assessments, licensees would be 

required to document the bases for 

concluding that the acceptance criteria 

in §§ 50.46a(f)(1) and (f)(2) are satisfied 

and that they continue to be satisfied 

throughout the operating lifetime of the 

facility. Licensees are also required 

under Part II of Appendix K to Part 50 

to document the bases of evaluation 

models used to perform ECCS 

calculations. Licensees would also be 

required to document the time spent in 

an operating configuration not 

demonstrated to meet the ECCS 

acceptance criteria in § 50.46a(c)(3) to 

demonstrate compliance with the 

fourteen days in any twelve month 

period limit in paragraph (d)(5) of this 

section. This documentation could be 

reviewed during NRC inspections and/ 

or audits to ensure that the risk criteria 

in § 50.46a(f) would be satisfied. 

I. Submittal and Review of Applications 

1. Initial Application for Implementing 

Alternative § 50.46a Requirements 

When a licensee first applies to adopt 

the alternative § 50.46a requirements, 

that licensee must submit an application 

under § 50.90 for NRC review and 

approval of a license amendment 

request. The initial application must 

contain the information as specified in 

§§ 50.46a(c)(1)(i) through (v). This 

includes information related to the 

applicability to the facility of the 

NUREG–1829 and NUREG–1903 results; 

information identifying the ECCS 

analysis methods to be used; 

information describing the licensee’s 

risk-informed evaluation process; 

information describing the licensee’s 

proposed process for making riskinformed changes without prior NRC 

approval (if the licensee is seeking 

approval of such a process); and 

information describing non safety 

equipment to be credited for compliance 

with the ECCS acceptance criteria in 

§ 50.46a(e). A licensee’s initial change 

from its existing ECCS analysis need not 

be reviewed by the licensee under the 

provisions of § 50.59. Because the rule 

requires NRC review and approval of 

the initial license amendment 

application for compliance with the 

alternative § 50.46a requirements, there 

is no purpose served by also requiring 

licensees to perform a § 50.59 

evaluation, because § 50.59 is a process 

to determine the need for prior NRC 

approval of a change to a facility or its 

procedures as described in the FSAR. 

After the § 50.46a evaluation models 

and initial ECCS LOCA analyses are 

established by approval of the license 

amendment implementing § 50.46a, 

subsequent changes to ECCS analyses 

would be controlled by the existing 

process in § 50.59 (which provides 

criteria for determining which changes 

are within the licensee’s authority) and 

the requirements in § 50.46a(g) for 

reporting when changes to evaluation 

models and analysis methods (whether 

from correction of errors or changes) is 

significant. 

The initial application may request 

one or more facility changes. The initial 

application may also include a request 

for NRC approval of a process for 

evaluating the acceptability of future 

changes enabled by § 50.46a using the 

provisions in paragraph (f)(1) of this 

section. If approval of a process for 

evaluating future changes is requested, 

the application must include the 

information described in 

§ 50.46a(c)(1)(iv). Otherwise, this 

information would not need to be 

submitted in the initial application. 


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2. Subsequent Applications for Plant 

Changes Under § 50.46a 

After NRC approval of a licensee’s 

initial license amendment application 

addressing ECCS analyses and the riskinformed evaluation processes, 

licensees may submit individual license 

amendment applications for plant 

changes under § 50.90. These individual 

license amendment applications must 

contain: 

a. The information required by 

§ 50.90; 

b. Information from the risk-informed 

evaluation demonstrating that the risk 

criteria, defense-in-depth criteria, safety 

margins, and performance monitoring 

criteria in §§ 50.46a(f)(2) and (f)(3) are 

met; 

c. Information demonstrating that the 

ECCS acceptance criteria in 

§§ 50.46a(e)(3) and (e)(4) are met; and 

d. Information demonstrating that the 

proposed change will not increase the 

LOCA frequency of the facility by an 

amount that would invalidate the 

applicability to the facility of the 

generic NUREG–1829 and NUREG–1903 

reports. 

After reviewing the individual plant 

change license amendment application, 

the NRC may approve the change if it 

complies with the above criteria and all 

other applicable NRC regulations, 

including requirements for plant 

physical security. The NRC would 

evaluate potential impacts of the 

proposed change on facility security to 

ensure that the change does not 

significantly reduce the ‘‘built-in 

capability’’ of the plant to resist security 

threats, thus ensuring that the change is 

not inimical to the common defense and 

security and provides adequate 

protection to public health and safety. 

Licensees who have not submitted a 

request for NRC approval of a process 

for evaluating the acceptability of future 

changes enabled by § 50.46a using the 

provisions in paragraph (f)(1) of that 

section may do so at any time by 

submitting the information described in 

paragraph (c)(1)(iv). 

J. Applicability to New Reactor Designs 

As previously discussed under NRC 

Topic 1, the NRC has evaluated public 

comments and agrees with commenters 

who stated that there are no technical 

reasons which prevent the revised 

proposed § 50.46a regulations from 

being applied to new light water reactor 

designs that are similar in nature (with 

respect to design and expected LOCA 

pipe break frequency) to current 

operating reactors. 

1. Similarity of New Reactor Designs to 

Existing Reactor Designs 

There are several new LWR designs 

for which the NRC expects that the 

frequency of large LOCAs could be as 

low as it is at current LWRs. Thus, it 

could be appropriate to allow applicants 

to apply the § 50.46a requirements to 

these future designs. Accordingly, the 

revised proposed rule has been 

modified to apply to new LWR reactor 

designs; i.e. facilities other than those 

which are currently licensed to operate. 

Applicants for design certification or 

combined licenses, holders of combined 

licenses under 10 CFR part 52, or future 

licensees of operating light-water 

reactors who wish to apply § 50.46a 

must submit an analysis for NRC 

approval demonstrating why it would 

be appropriate to apply the alternative 

ECCS requirements and what the 

appropriate transition break size (TBS) 

would be in order for the new design to 

meet the intent of the § 50.46a rule. 

In its analysis, the applicant, holder, 

or licensee must demonstrate that the 

proposed reactor facility is similar to 

reactors licensed before the effective 

date of the rule. In addressing similarity 

of the proposed design to reactors 

licensed before the effective date of rule, 

the applicant, holder, or licensee would 

need to address design, construction 

and fabrication, and operational factors 

that include, but are not limited to: 

(1) The similarity of the piping 

materials of construction and 

construction techniques for new 

reactors to those in the currently 

operating fleet; 

(2) The similarity of service 

conditions and operational programs 

(e.g., in-service inspection and testing, 

leak detection, quality assurance etc.) 

for new reactors to those for operating 

plants; 

(3) The similarity of piping design, 

e.g. pipe sizes and pipe configuration, 

for new reactors to those found in 

operating plants; 

(4) Adherence to existing regulatory 

requirements, regulatory guidance, and 

industry programs related to mitigation 

and control of age-related degradation 

(e.g., aging management, fatigue 

monitoring, water chemistry, stress 

corrosion cracking mitigation etc.); and 

(5) Any plant-specific attributes that 

may increase LOCA frequencies 

compared to the generic results in 

NUREG–1829 and NUREG–1903. 

The analysis must also include a 

recommendation for an appropriate TBS 

and a justification that the 

recommended TBS is consistent with 

the technical basis for this proposed 

rule. For those new reactor designs that 

employ design features that effectively 

increase the break size via opening of 

specially designed valves to rapidly 

depressurize the reactor coolant system 

during any size loss of coolant accident, 

justification of the relevance of a TBS 

would also be necessary. The 

methodology used to determine the 

proposed TBS should be described in 

the justification. 

Based on information currently 

available, new reactor designs may have 

similar piping materials, similar service 

conditions and operational programs, 

similar piping designs, and similar 

mitigation and control of age-related 

degradation programs to those found in 

currently operating plants. Therefore, 

the TBS defined in the proposed rule for 

currently operating reactors could 

potentially be applicable to some new 

reactor designs. 

In addition, after obtaining an 

operating or combined license for a 

plant with a currently-approved 

standard design, a licensee could adopt 

§ 50.46a if the design is demonstrated to 

be similar to the designs of plants 

licensed before the effective date of the 

rule (by evaluating the criteria above) 

and the TBS proposed by the licensee is 

found acceptable by the NRC. 

2. NRC Request for Public Comments on 

the Use of Large Release Frequency 

(LRF) as the Risk Acceptance Criteria 

Metric for New Reactors 

Regulatory Guide 1.174, ‘‘An 

Approach for Using Probabilistic Risk 

Assessment in Risk Informed Decisions 

on Plant Specific Changes to the 

Licensing Basis,’’ was originally issued 

in July 1998. This RG provides guidance 

for a multitude of risk-informed 

applications and improves consistency 

in regulatory decisions in areas where 

the results of risk analyses are used to 

help justify regulatory action. The guide 

is the foundation for many other riskinformed programs (e.g., inservice 

testing, inservice inspection of piping) 

at the agency. 

Regulatory Guide 1.174 describes five 

key principles of the risk-informed, 

integrated decision making process. In 

Principle 4—When proposed changes 

result in an increase in core damage 

frequency or risk, the increases should 

be small and consistent with the intent 

of the Commission’s Safety Goal Policy 

Statement—the regulatory guide 

presents quantitative guidelines for 

acceptably small increases in CDF and 

LERF, as depicted in Figures 3 and 4 of 

the guide. The magnitude of acceptably 

small increases varies stepwise with the 

baseline CDF and LERF. A small 

increase up to 10¥5 per reactor year for 

CDF and 10¥6 per reactor year for LERF 


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are normally acceptable until the 

baseline risk increases to reference 

values of approximately 10¥4 per 

reactor year and 10¥5 per reactor year 

for CDF and LERF respectively. Plants 

with baseline CDF and LERF which 

exceed the reference values, or with 

baseline risks that are not known with 

precision, would normally be limited to 

very small risk increases of up to 10¥6

per reactor year and 10¥7 per reactor 

year for CDF and LERF, respectively. 

Before RG 1.174 was issued, the 

Commission’s SRM dated June 26, 1990, 

prepared in response to SECY–90–016, 

‘‘Evolutionary Light Water Reactor 

Certification Issues and their 

Relationships to Current Regulatory 

Requirements,’’ established a goal for 

large release frequency (LRF) of less 

than 10¥6 per reactor year for new 

reactor design certification and 

licensing. These goals are discussed 

further in Standard Review Plan 

(NUREG–0800) Chapter 19, and RG 

1.206 ‘‘Combined License Applications 

for Nuclear Power Plants’’ Section 

C.I.19. 

In light of this difference in the risk 

metrics used for currently operating 

reactors (LERF) and new reactors (LRF), 

the NRC is seeking public comments on 

whether LRF should be the metric of 

concern in lieu of LERF for new reactor 

applicants (or licensees) implementing 

the § 50.46a alternative ECCS 

requirements. Because the LRF goal for 

new reactors is a decade lower than the 

10¥5 per reactor year LERF reference 

value above which a facility would be 

limited to very small increases, should 

the definition of what constitutes ‘‘very 

small increase’’ and ‘‘minimal increase’’ 

for LRF (for new reactors) be a full 

decade lower than those defined for 

LERF (for existing reactors) or should 

the definition be based on relative 

change in LRF? 

The NRC has previously sought 

stakeholder input on the issue of risk 

metrics for new light-water reactors. A 

memorandum dated February 12, 2009, 

from R. W. Borchardt, Executive 

Director for Operations, to the 

Commissioners, ‘‘Alternative Risk 

Metrics for New Light-Water Reactor 

Risk-Informed Applications’’ (Adams 

Accession No. ML090160008), provides 

a discussion of the issues. The white 

paper attached to that memorandum 

presents a full discussion of the issues 

and options for applying or modifying 

the current set of reactor risk metrics to 

new reactors. The paper discusses the 

issues posed by the lower risk estimates 

of new reactors in risk-informed 

applications, including changes to the 

licensing basis and the reactor oversight 

process, and describes the advantages 

and disadvantages of each option. 

On February 18, 2009, the NRC held 

a public meeting with stakeholders on 

the topic of risk metrics for new lightwater reactors (see meeting summary; 

Adams Accession No. ML090570356). 

Additionally, both the NRC and 

industry representatives provided a 

briefing on the topic at the April 3, 

2009, meeting of the ACRS. 

As discussed in these documents, the 

NRC is considering several options 

regarding risk metrics for new reactor 

risk-informed applications. The options 

include applying the existing operating 

reactor acceptance guidelines to new 

reactors, using new guidelines and 

thresholds for new reactors, or 

postponing any significant change to the 

process and evaluating new reactors on 

a case-by-case basis for an indeterminate 

period. As described in the NEI paper, 

‘‘Risk Metrics for Operating New 

Reactors’’ (ML090900674; March 27, 

2009), NEI has expressed its preference 

for applying the existing operating 

reactor acceptance guidelines to new 

reactors (which is referred to as Option 

1 in the NRC white paper). 

As part of the public comment 

process for this revised proposed rule, 

public stakeholders are invited to 

comment on the use of any of the 

alternative risk metric approaches for 

determining compliance with the risk 

acceptance criteria in § 50.46a. 

VI. Specific Topics Indentified for 

Public Comment 

The NRC seeks specific public 

comments on three topics. These issues 

were discussed previously in this 

document, but are summarized again 

here to assist commenters. 

1. Although the revised proposed rule 

would permit licensees to make plant 

changes that result in very small risk 

increases, the NRC is requesting 

stakeholder comments on whether the 

rule should allow plant changes that 

increase risk at all. Instead of the risk 

acceptance criteria allowing very small 

risk increases, should the risk 

acceptance criteria in final rule require 

that the net effect of plant changes made 

under § 50.46a be risk neutral or risk 

beneficial? The NRC requests 

stakeholders to provide comments on 

the use of risk acceptance criteria that 

would not allow a cumulative increase 

in risk for plant changes made under 

§ 50.46a. (See Section V.E.4.b of this 

document.) 

2. Because of the difference in the risk 

acceptance criteria metrics used for 

currently operating reactors (LERF) and 

new reactors (LRF), the NRC is seeking 

public comments on whether LRF 

should be the metric of concern in lieu 

of LERF for new reactor applicants (or 

licensees) implementing the § 50.46a 

alternative ECCS requirements. Because 

the LRF goal for new reactors is a 

decade lower than the 10¥5 per reactor 

year LERF reference value above which 

a facility would be limited to very small 

increases, should the definition of what 

constitutes ‘‘very small increase’’ and 

‘‘minimal increase’’ for LRF (for new 

reactors) be a full decade lower than 

those defined for LERF (for existing 

reactors) or should the definition be 

based on relative change in LRF? (See 

Section V.J of this document.) 

3. In § 50.46a(e)(4)(i) of the revised 

proposed rule the NRC proposes 

coolable core geometry as a high level 

performance-based ECCS analysis 

acceptance criterion for beyond-TBS 

LOCAs. Applicants would be allowed to 

justify appropriate metrics to 

demonstrate coolable geometry or use 

the current metrics (2200 °F PCT and 17 

percent MLO). However, the NRC 

acknowledges that it would be 

expensive and time-consuming for 

industry to develop the necessary 

experimental and analytical data to 

justify alternative acceptance criteria as 

a surrogate for demonstrating coolable 

geometry. Because of the difficulty in 

demonstrating alternative metrics, the 

NRC is requesting stakeholder 

comments on whether the final § 50.46a 

rule should retain the coolable geometry 

criterion for beyond-TBS breaks. 

Retaining coolable geometry would give 

licensees the option to demonstrate 

alternative coolable geometry metrics or 

use the current metric (2200 °F PCT and 

17 percent MLO). If the NRC removed 

the coolable geometry criterion, the 

beyond-TBS acceptance criteria would 

be the same as the acceptance criteria 

for TBS and smaller breaks (2200 °F 

PCT and 17 percent MLO). The NRC 

will evaluate stakeholder comments on 

this question before deciding which 

beyond-TBS acceptance criteria to 

include in the final rule. (See Section 

V.D.2 of this document.) 

VII. Petition for Rulemaking, PRM–50– 

75 

In February 2002, the Nuclear Energy 

Institute submitted a petition for 

rulemaking (PRM–50–75) requesting the 

NRC to revise ECCS requirements by 

redefining the large break LOCA 

(ML020630082). Notice of that petition 

was published in the Federal Register 

for public comment on April 8, 2002 (67 

FR 16654). The petition requested the 

NRC to amend § 50.46 and Appendices 

A and K of Part 50 to allow licensees to 

use as an alternative to the doubleended rupture of the largest pipe in the 


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RCS, a maximum LOCA break size of 

‘‘up to and including an alternate 

maximum break size that is approved by 

the Director of the Office of Nuclear 

Reactor Regulation.’’ Seventeen sets of 

comments were received, mostly from 

the power reactor industry in favor of 

granting the petition. A few 

stakeholders were concerned about 

potential impacts on defense-in-depth 

or safety margins if significant changes 

were made to reactor designs based 

upon use of a smaller break size. The 

NRC considered the public comments, 

evaluated the petition, and published a 

notice in the Federal Register resolving 

the petition and closing the PRM–50–75 

docket. (See 73 FR 66000; November 6, 

2008.) The NRC concluded that the 

issue raised by the petitioner should be 

considered in the rulemaking process. 

Documents related to the resolution of 

PRM–50–75 are available at http:// 

www.regulations.gov under docket ID: 

NRC–2002–0018. The NRC is addressing 

the issues raised by the petitioner and 

stakeholders in this rulemaking. 

VIII. Section-by-Section Analysis of 

Changes 

A. Section 50.34—Contents of 

Application; Technical Information 

Paragraph (a)(4)(i) of this section 

would specify that § 50.46a contains 

alternative ECCS requirements that 

licensees could choose to apply to 

reactors whose construction permits 

were issued before the effective date of 

the rule. This section also states that 

applicants for construction permits for 

facilities which may be issued after the 

effective date of the rule could also 

choose to apply the § 50.46a alternative 

ECCS requirements to preliminary 

analysis and evaluation of the design if 

the applicant demonstrates that the 

facility is similar to the designs of 

facilities licensed before the effective 

date of the rule. 

Paragraph (a)(4)(ii) would specify that 

applicants for construction permits for 

facilities which may be issued after the 

effective date of the rule who have not 

demonstrated that the facility is similar 

to the designs of facilities licensed 

before the effective date of the rule may 

not apply the § 50.46a alternative ECCS 

requirements in the preliminary 

analysis and evaluation of the design. 

Paragraph (b)(4)(i) of this section 

would specify that applicants for 

operating licenses for facilities which 

may be issued before the effective date 

of the rule could choose to apply the 

§ 50.46a alternative ECCS requirements 

in the final analysis and evaluation of 

the design. This section also states that 

applicants for operating licenses for 

facilities which may be issued after the 

effective date of the rule could also 

choose to apply the § 50.46a alternative 

ECCS requirements to final analysis and 

evaluation of the design if the applicant 

demonstrates that the facility is similar 

to the designs of facilities licensed 

before the effective date of the rule. 

Paragraph (b)(4)(ii) would specify that 

applicants for operating licenses for 

facilities which may be issued after the 

effective date of the rule who have not 

demonstrated that the design is similar 

to the designs of facilities licensed 

before the effective date of the rule may 

not apply the § 50.46a alternative ECCS 

requirements in the final analysis and 

evaluation of the design. 

B. Section 50.46—Acceptance Criteria 

for Emergency Core Cooling Systems for 

Light-Water Nuclear Power Plants 

Paragraph (a) of this section would 

specify that emergency core cooling 

systems of BWRs and PWRs licensed 

before the effective date of the rule must 

be designed under § 50.46 or § 50.46a. 

Paragraph (a) would also specify that 

emergency core cooling systems of 

BWRs and PWRs licensed after the 

effective date of the rule could also 

choose to comply with the § 50.46a 

alternative ECCS requirements if the 

applicant or licensee demonstrates that 

the design is similar to the designs of 

LWR facilities licensed before the 

effective date of the rule. 

C. Existing Section 50.46a—Acceptance 

Criteria for Reactor Coolant System 

Venting Systems, Is Administratively 

Redesignated as Section 50.46b 

D. Section 50.46a—Alternative 

Acceptance Criteria for Emergency Core 

Cooling Systems for Light-Water 

Reactors 

Paragraph (a) of this section would 

provide definitions for terms used in 

other parts of this section. The 

definition of evaluation model in 

§ 50.46a(a)(2) is the same as in § 50.46. 

The definition of loss-of-coolant 

accidents in § 50.46a(a)(3) is based on 

the existing definition in § 50.46 but has 

been modified to indicate that pipe 

breaks larger than the TBS are beyond 

design-basis accidents. 

The new definitions are: 

(1) Changes enabled by this section, 

which means changes to the facility, 

technical specifications, or procedures 

that comply with § 50.46a but do not 

comply with § 50.46; 

(4) Operating configuration, which is 

used in § 50.46a(d)(5) to specify plant 

equipment availability conditions that 

must be analyzed for conformance with 

acceptance criteria; and 

(5) Transition break size (TBS), which 

is used to distinguish between 

requirements applicable to pipe breaks 

at or below this size from those 

applicable to pipe breaks above this 

size. 

Paragraph (b) would provide the 

applicability and scope of the 

requirements of this section. Proposed 

§ 50.46a would apply to currently 

licensed light-water nuclear power 

reactors (licensed before the effective 

date of the rule). Proposed § 50.46a 

would also apply to LWRs licensed after 

the effective date of the rule which have 

been demonstrated to be similar to the 

designs of LWR facilities licensed before 

the effective date of the rule. Its 

requirements would be in addition to 

any other requirements applicable to 

ECCS set forth in 10 CFR 50, with the 

exception of § 50.46. 

Paragraph (c)(1) would specify the 

contents of initial licensee applications 

for implementing the alternative ECCS 

requirements in § 50.46a. Paragraph 

(c)(1)(i) would require that an 

application contain a written evaluation 

demonstrating applicability of the 

results in NUREG–1829 and NUREG– 

1903 to the licensee’s facility. However, 

if the facility differs significantly from 

the facilities analyzed in NUREG–1903, 

the application must contain a plant 

specific analysis demonstrating that the 

risk of seismically-induced LOCAs 

larger than the TBS is comparable to or 

less than the seismically-induced LOCA 

risk associated with the NUREG–1903 

results. Paragraph (c)(1)(ii) would 

require identification of the NRCapproved analysis methods to be used to 

comply with the ECCS analysis 

requirements and acceptance criteria in 

paragraph (e). Paragraph (c)(1)(iii) 

would require a description of the riskinformed evaluation process used to 

determine whether proposed changes to 

the facility meet the requirements for 

risk-informed evaluations in paragraph 

(f). Paragraph (c)(1)(iv) would require 

licensees who wish to make changes 

enabled by § 50.46a without prior NRC 

approval to submit a description of the 

risk-informed evaluation process and 

the PRA or non-PRA risk-assessment 

methods to be used to determine the 

acceptability of such changes. The 

licensee’s process must be capable of 

demonstrating that all of the acceptance 

criteria in paragraph (f) will be met for 

each change. Paragraph (c)(1)(v) would 

require licensees who wish to adopt the 

alternative ECCS requirements in 

§ 50.46a to submit a description of all 

non safety equipment to be relied on to 

mitigate the consequences of a LOCA 

larger than the TBS. 


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Paragraph (c)(2) states that applicants 

for a construction permit, operating 

license, design approval, design 

certification, manufacturing license, or 

combined license seeking to implement 

the requirements of this section shall, in 

addition to the information that would 

be required by paragraph (c)(1) of this 

section, submit an analysis 

demonstrating why the proposed reactor 

design is similar to the designs of 

currently operating reactors. 

Paragraph (c)(3) specifies the 

acceptance criteria for approval of 

applications to comply with § 50.46a. 

Paragraph (c)(3)(i) would require the 

evaluation submitted under paragraph 

(c)(1)(i) to demonstrate that the NUREG– 

1829 results are applicable to the 

facility, and the risk of seismicallyinduced LOCAs larger than the TBS is 

comparable to or less than the 

seismically-induced LOCA risk 

associated with the NUREG–1903 

results. Paragraph (c)(3)(ii) would 

require that the method(s) for 

demonstrating compliance with the 

ECCS acceptance criteria in paragraphs 

(e)(3) and (e)(4) of this section meet the 

requirements in paragraphs (e)(1) and 

(e)(2). Paragraph (c)(3)(iii) would require 

that the risk-informed evaluation 

process the licensee proposes to use for 

making changes enabled by this section 

be adequate for determining whether the 

acceptance criteria in paragraph (f) of 

this section have been met. Paragraph 

(c)(3)(iv) would require that all non 

safety equipment credited for 

demonstrating compliance with the 

ECCS acceptance criteria is identified 

and listed as such in plant Technical 

Specifications. Paragraph (c)(3)(v) 

would require that the reactor design for 

all applicants other than those holding 

operating licenses issued before the 

effective date of the rule be similar to 

the designs of current operating reactors 

and the applicant’s proposed TBS is 

consistent with the technical basis for 

Section 50.46a. 

Paragraph (d) specifies the 

requirements with which licensees 

would be required to comply during 

facility operation after implementing 

§ 50.46a. 

Paragraph (d)(1) would require that 

the ECCS models be maintained to 

comply with the ECCS acceptance 

criteria in paragraphs (e)(1) and (e)(2) of 

this section. 

Paragraph (d)(2) would require that 

the licensee maintain leak detection 

equipment available at the facility and 

identify, monitor, and quantify leakage 

to reduce the likelihood of a LOCA 

larger than the TBS. 

Paragraph (d)(3) would require that 

changes to the facility, technical 

specifications, or procedures enabled by 

§ 50.46a be evaluated by a risk-informed 

evaluation process which demonstrates 

that acceptance criteria in § 50.46a(f) are 

met. 

Paragraph (d)(4), would require 

licensees to maintain and upgrade its 

PRA analyses no less often than once 

every 2 refueling outages. Maintaining a 

PRA involves the update of PRA models 

to reflect facility changes such as plant 

modifications, procedure changes, or 

changes in plant performance data. 

Upgrading a PRA involves incorporating 

into the PRA models a new 

methodology or significant changes in 

scope or capability that impact the 

significant accident sequences. Risk 

assessments would be required to 

continue to meet the quality 

requirements in §§ 50.46a(f)(4) and 

(f)(5). Licensees would be required to 

take action to ensure that facility design 

and operation continue to be consistent 

with the risk assessment assumptions 

used to meet the acceptance criteria in 

§§ 50.46a(f)(2) or (f)(3). Any necessary 

changes to the facility caused by 

maintaining or upgrading risk 

assessments would not be deemed 

backfitting. 

Paragraph (d)(5) would require 

licensees to control plant operation to 

ensure that for LOCAs larger than the 

TBS, operation in a plant operating 

configuration not demonstrated to meet 

the acceptance criteria in paragraph 

(e)(4) would not exceed a total of 

fourteen days in any 12 month period. 

Paragraph (d)(6) would require 

licensees to perform an evaluation to 

determine the effect of all planned 

facility changes and would prohibit 

licensees from implementing any 

facility change that would invalidate the 

evaluation performed pursuant to 

§ 50.46a(c)(1)(i) demonstrating the 

applicability to the licensee’s facility of 

the generic results in NUREG–1829 and 

NUREG–1903. 

Paragraph (e) would provide the ECCS 

evaluation model requirements, analysis 

requirements, and acceptance criteria 

for the two LOCA break size regions. 

Paragraph (e)(1) would specify model 

and analysis requirements for breaks 

smaller than or equal to the TBS. These 

requirements are the same as the current 

requirements for LOCA analysis models 

in existing § 50.46. 

Paragraph (e)(2) would specify model 

and analysis requirements for breaks 

larger than the TBS. Methods for 

evaluating ECCS cooling performance 

for breaks larger than the TBS must be 

approved by the NRC. However the 

analysis for breaks larger than the TBS 

may be performed using more realistic 

analysis inputs and assumptions than 

those required for breaks smaller than or 

equal to the TBS. Analysis of breaks 

larger than the TBS need not assume a 

coincident single failure of mitigation 

equipment or loss of offsite power. Nonsafety grade equipment may also be 

credited in analyses of breaks larger 

than the TBS provided that onsite 

power can supplied to that equipment 

in a reasonable time in the event offsite 

power is lost. 

Paragraph (e)(3) would provide ECCS 

acceptance criteria for LOCAs smaller 

than or equal to the TBS. The criteria 

specified would be the same as the 

current requirements in § 50.46(b). 

Paragraph (e)(4) would provide ECCS 

acceptance criteria for LOCAs larger 

than the TBS. These acceptance criteria 

would be based on maintaining a 

coolable geometry in the core and 

demonstrating long term cooling 

capability and are less prescriptive than 

the criteria presently used for LOCA 

analysis. 

Paragraph (e)(5) would provide that 

the Director of the Office of Nuclear 

Reactor Regulation may impose 

restrictions on reactor operation if ECCS 

requirements are not met. This 

paragraph would be added to be 

consistent with existing § 50.46 which 

also contains this requirement. 

Paragraph (f) would provide 

requirements for implementing changes 

to the facility, technical specifications, 

and procedures under § 50.46a. 

Paragraph (f)(1) would specify that 

licensees may make changes without 

NRC approval if: 

(i) The changes are permitted under 

§ 50.59; 

(ii) A risk-informed evaluation 

process has been submitted by the 

licensee and reviewed and approved by 

the NRC under § 50.46a(c)(1)(iv); and 

(iii) The change does not invalidate 

the evaluation performed under 

§ 50.46a(c)(1)(i) of the applicability of 

the results in NUREG–1829 and 

NUREG–1903 to the licensee’s facility. 

Paragraph (f)(2) would state that for 

plant changes not permitted under 

paragraph (f)(1), licensees must submit 

an application for a license amendment 

under § 50.90. The application must 

contain: 

(i) The information required under 

§ 50.90; 

(ii) For reactors licensed before the 

effective date of the rule, information 

from the risk-informed evaluation 

demonstrating that the total increases in 

core damage frequency and large early 

release frequency are very small and the 

overall risk remains small, and that the 

risk-informed change criteria in 

paragraph (f)(3) are met; 


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(iii) For all applicants other than 

those holding operating licenses issued 

before the effective date of the rule, 

information from the risk-informed 

evaluation demonstrating that the total 

increases in core damage frequency and 

large release frequency are very small, 

the overall risk remains small, and the 

criteria in paragraph (f)(3) of this section 

are met; 

(iv) An evaluation of the cumulative 

effect of previous changes that have 

increased risk but have met the 

acceptance criteria. If more than one 

plant change is combined, including 

plant changes not enabled by § 50.46a, 

into a group for the purposes of 

evaluating acceptable risk increases, the 

evaluation of each individual change 

shall be performed along with the 

evaluation of combined changes; 

(v) Information demonstrating that the 

ECCS analysis acceptance criteria in 

paragraphs (e)(3) and (e)(4) are met; and 

(vi) Information demonstrating that 

the proposed change will not increase 

the LOCA frequency of the facility 

(including the frequency of seismicallyinduced LOCAs) by an amount that 

would invalidate the applicability to the 

facility of the generic seismic studies 

(NUREG–1829, ‘‘Estimating Loss-ofCoolant Accident (LOCA) Frequencies 

through the Elicitation Process’’, March 

2008 and NUREG–1903, ‘‘Seismic 

Considerations for the Transition Break 

Size’’, February 2008) that support the 

technical basis for § 50.46a. 

Paragraph (f)(3) would specify 

requirements for all plant changes. 

Paragraph (f)(3)(i) would require that 

defense-in-depth is maintained. 

Paragraph (f)(3)(ii) would require that 

adequate safety margins are maintained. 

Paragraph (f)(3)(iii) would require that 

adequate performance-measurement 

programs will be implemented. 

Paragraph (f)(3)(iii) provides criteria on 

the specific attributes required to meet 

the performance measurement 

requirements. 

Paragraph (f)(2) does not require use 

of PRA in assessing risks associated 

with the proposed changes. To the 

extent that PRA is used, paragraph (f)(4) 

of the revised proposed rule would 

identify specific technical requirements 

for the risk-informed assessment. 

(i) Address initiating events from 

sources both internal and external to the 

plant and for all modes of operation, 

including low power and shutdown 

modes, that would affect the regulatory 

decision in a substantial manner; 

(ii) Reasonably represent the current 

configuration and operating practices at 

the plant; 

(iii) Have sufficient technical 

adequacy (including consideration of 

uncertainty) and level of detail to 

provide confidence that the total risk 

estimate and the change in total risk 

estimate adequately reflect the plant and 

the effect of the proposed change on 

risk; and 

(iv) Be determined, through peer 

review, to meet industry standards for 

PRA quality that have been endorsed by 

NRC. 

Paragraph (f)(5) would require that to 

the extent that risk assessment methods 

other than PRA are used to develop 

quantitative or qualitative estimates of 

changes to risk in the risk-informed 

evaluation, an integrated, systematic 

process must be used. All aspects of the 

analyses must reasonably reflect the 

current plant configuration and 

operating practices, and applicable 

plant and industry operating 

experience. 

Paragraph (g) would provide the 

requirements for making reports to the 

NRC. 

Paragraph (g)(1) would require 

reporting of all errors or changes to 

ECCS analyses at least annually as 

specified in § 50.4. For significant 

changes or errors, licensees would be 

required to report within 30 days 

including a schedule for reanalysis or 

other action as needed to show 

compliance with ECCS requirements. 

Under paragraph (g)(1)(i), for LOCAs 

involving pipe breaks equal to or 

smaller than the TBS, significant 

changes would be defined as a change 

in peak cladding temperature of greater 

than 50 °F. Under paragraph (g)(1)(ii), 

for LOCAs involving pipe breaks larger 

than the TBS, a significant change 

would be defined as one resulting in a 

significant reduction in the capability to 

meet the ECCS acceptance criteria in 

§ 50.46a(e)(4). 

Paragraph (g)(2) would set forth 

reporting requirements with respect to 

the PRA maintenance and upgrading 

that would be required by § 50.46a(d)(4). 

When maintaining and upgrading the 

PRA, § 50.46a(g)(2) would require the 

licensee to report changes to the NRC 

within 60 days if the acceptance criteria 

in §§ 50.46a(f)(2)(ii) or (f)(2)(iii) (for new 

reactors) are exceeded. This provision 

would also require the report to include 

a schedule for implementation of any 

corrective actions necessary to bring 

plant operation or design back into 

compliance with the acceptance criteria. 

Paragraph (g)(3) would contain 

reporting requirements for plant 

changes made under § 50.46a(f)(1) 

involving minimal risk. A short 

description of these changes would be 

reported every 24 months. 

Paragraph (h) would provide 

documentation requirements for plant 

changes. Following implementation of 

§ 50.46a, licensees would be required to 

maintain records sufficient to 

demonstrate compliance with the 

requirements in § 50.46a and § 50.71. 

Paragraphs (i) through (l) would be 

reserved for future use. 

Paragraph (m) would provide that 

changes made by the NRC to the TBS 

and all changes required to return a 

facility to compliance with the 

acceptance criteria after a change in the 

TBS are not deemed to be backfitting 

under 10 CFR 50.109. 

E. Section 50.109—Backfitting 

This section would be modified to 

provide that changes made by the NRC 

to the TBS and changes made by 

licensees to continue to comply with 

§ 50.46a are not deemed to be 

backfitting under 10 CFR 50.109. 

F. Appendix A to Part 50—General 

Design Criteria for Nuclear Power Plants 

Five of the general design criteria 

contained in Appendix A would be 

modified to remove the requirement to 

assume a single failure and a loss-ofoffsite power in the systems subject to 

these criteria for pipe breaks larger than 

the TBS up to and including the DEGB 

of the largest RCS pipe for those plants 

implementing § 50.46a. The specific 

criteria are: GDC 17, Electrical power 

systems, GDC 35, Emergency core 

cooling, GDC 38, Containment heat 

removal, GDC 41, Containment 

atmosphere cleanup, and GDC 44, 

Cooling water systems. General Design 

Criterion 50, Containment design basis, 

would also be modified to specify that 

for plants under § 50.46a, leak tight 

containment capability should be 

maintained for ‘‘realistically’’ calculated 

temperatures and pressures for LOCAs 

larger than the TBS. 

G. Section 52.47—Contents of 

Applications; Technical Information 

Paragraph (a)(4) of this section would 

be amended to specify the technical 

information to be submitted in an 

application for a standard design 

certification for a nuclear power facility 

filed separately from the filing of an 

application for a construction permit or 

combined license for such a facility. 

New paragraph (a)(4)(i) would to 

specify that analyses of emergency core 

cooling systems and the need for high 

point vents for standard designs 

certified after the effective date of the 

§ 50.46a rule must be performed under 

the requirements of either § 50.46 or 

§ 50.46a (for ECCS performance) and 

§ 50.46b (for reactor coolant system high 

point vents) if the standard design is 

demonstrated to be similar to the 


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designs of reactors licensed before the 

effective date of § 50.46a. 

New paragraph (a)(4)(ii) would 

specify that analyses of emergency core 

cooling systems and the need for high 

point vents for standard designs 

certified after the effective date of the 

§ 50.46a rule must be performed under 

the requirements of § 50.46 (for ECCS 

performance) and § 50.46b (for reactor 

coolant system high point vents) if the 

standard design is not demonstrated to 

be similar to the designs of reactors 

licensed before the effective date of 

§ 50.46a. 

H. Section 52.79—Contents of 

Applications; Technical Information in 

Final Safety Analysis Report 

In this section paragraph (a)(5) would 

be amended to specify the technical 

information to be submitted in the final 

safety analysis report for an application 

for a combined license for a nuclear 

power facility. 

New paragraph (a)(5)(i) would specify 

that analyses of emergency core cooling 

systems and the need for high point 

vents for plants licensed after the 

effective date of the § 50.46a rule must 

be performed under the requirements of 

either § 50.46 or § 50.46a (for ECCS 

performance) and § 50.46b (for reactor 

coolant system high point vents) if the 

design is demonstrated to be similar to 

the designs of reactors licensed before 

the effective date of § 50.46a. 

New paragraph (a)(5)(ii) would 

specify that analyses of emergency core 

cooling systems and the need for high 

point vents for plants licensed after the 

effective date of the § 50.46a rule must 

be performed under the requirements of 

§ 50.46 (for ECCS performance) and 

§ 50.46b (for reactor coolant system high 

point vents) if the design is not 

demonstrated to be similar to the 

designs of reactors licensed before the 

effective date of § 50.46a. 

I. Section 52.137—Contents of 

Applications; Technical Information 

Paragraph (a)(4) of this section would 

be amended to specify the technical 

information to be submitted in an 

application for approval of a standard 

design for a nuclear power facility. 

New paragraph (a)(4)(i) would specify 

that analyses of emergency core cooling 

systems and the need for high point 

vents for designs approved after the 

effective date of the § 50.46a rule must 

be performed under the requirements of 

either § 50.46 or § 50.46a (for ECCS 

performance) and § 50.46b (for reactor 

coolant system high point vents) if the 

design is demonstrated to be similar to 

the designs of reactors licensed before 

the effective date of § 50.46a. 

New paragraph (a)(4)(ii) would 

specify that analyses of emergency core 

cooling systems and the need for high 

point vents for designs approved after 

the effective date of the § 50.46a rule 

must be performed under the 

requirements of § 50.46 (for ECCS 

performance) and § 50.46b (for reactor 

coolant system high point vents) if the 

design is not demonstrated to be similar 

to the designs of reactors licensed before 

the effective date of § 50.46a. 

J. Section 52.157—Contents of 

Applications; Technical Information in 

Final Safety Analysis Report 

Paragraph (f)(1) of this section would 

be amended to specify the technical 

information to be submitted in the final 

safety analysis report for an application 

for issuance of a license authorizing 

manufacture of nuclear power reactors 

to be installed at sites not identified in 

the manufacturing license application. 

New paragraph (f)(1)(i) would specify 

that analyses of emergency core cooling 

systems and the need for high point 

vents for a license authorizing 

manufacture of nuclear power reactors 

issued after the effective date of the 

§ 50.46a rule must be performed under 

the requirements of either § 50.46 or 

§ 50.46a (for ECCS performance) and 

§ 50.46b (for reactor coolant system high 

point vents) if the design is 

demonstrated to be similar to the 

designs of reactors licensed before the 

effective date of § 50.46a. 

New paragraph (f)(1)(ii) would specify 

that analyses of emergency core cooling 

systems and the need for high point 

vents for a license authorizing 

manufacture of nuclear power reactors 

issued after the effective date of the 

§ 50.46a rule must be performed under 

the requirements of § 50.46 (for ECCS 

performance) and § 50.46b (for reactor 

coolant system high point vents) if the 

design is not demonstrated to be similar 

to the designs of reactors licensed before 

the effective date of § 50.46a. 

IX. Criminal Penalties 

For the purposes of Section 223 of the 

Atomic Energy Act (AEA), as amended, 

the NRC is issuing the proposed rule to 

amend § 50.46, add § 50.46a, redesignate 

existing § 50.46a as § 50.46b and amend 

§§ 52.47, 52.79, 52.137, and 52.157 

under one or more of sections 161b, 

161i, or 161o of the AEA. Willful 

violations of the rule would be subject 

to criminal enforcement. Criminal 

penalties, as they apply to regulations in 

Part 50, are discussed in § 50.111 and as 

they apply to the regulations in Part 52, 

are discussed in § 52.303. 

X. Compatibility of Agreement State 

Regulations 

Under the ‘‘Policy Statement on 

Adequacy and Compatibility of 

Agreement States 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 

NRC, it may wish to inform its licensees 

of certain requirements via a mechanism 

that is consistent with the particular 

State’s administrative procedure laws, 

but does not confer regulatory authority 

on the State. 

XI. Availability of Documents 

Comments and other publicly 

available documents related to this 

rulemaking may be viewed 

electronically on the public computers 

located at the NRC’s Public Document 

Room (PDR), O1 F21, One White Flint 

North, 11555 Rockville Pike, Rockville, 

Maryland. The PDR reproduction 

contractor will copy documents for a 

fee. 

Publicly available documents are 

available electronically at the NRC’s 

Electronic Reading Room at http:// 

www.nrc.gov/reading-rm/adams.html. 

From this site, the public can gain entry 

into the NRC’s Agencywide Document 

Access and Management System 

(ADAMS), which provides text and 

image files of NRC’s public documents. 

If you do not have access to ADAMS or 

if there are problems in accessing the 

documents located in ADAMS, contact 

the NRC Public Document Room (PDR) 

Reference staff at 1–800–397–4209, 

(301) 415–4737 or by e-mail to 

pdr@nrc.gov. The NRC is making the 

documents identified below available to 

interested persons through one or more 

of the following methods as indicated. 

Public Document Room (PDR). The 

NRC Public Document Room is located 

at Public File Area O–F21, One White 

Flint North, 11555 Rockville Pike, 

Rockville, Maryland. 

Federal eRulemaking Portal. Go to 

http://www.regulations.gov and search 

for documents filed under Docket ID 

NRC–2004–0006. Address questions 

about NRC dockets to Carol Gallagher 

(301) 415–5905; e-mail 

Carol.Gallager@nrc.gov. 

NRC’s Electronic Reading Room 

(ERR). The NRC’s public electronic 


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reading room is located at http:// 

www.nrc.gov/reading-rm.html. 

Document PDR Web Err (Adams) 

Initial Proposed Rule (70 FR 67598) ....................................................................................... X NRC–2004–0006 ....... ML091060434 

NRC Report—Seismic Considerations for the Transition Break Size (December 2006) ........ X NRC–2004–0006 ....... ML053470439 

Letter from Graham B. Wallis (ACRS) to Dale E. Klein, ‘‘Draft Final Rule To Risk-Inform 10 

CFR 50.46, ‘Acceptance Criteria For Emergency Core Cooling Systems For Light-Water 

Nuclear Power Reactors’ ’’ (November 16, 2006).

X X ................................ ML063190465 

SECY–07–0082—Rulemaking to Make Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements; 10 CFR 50.46a ‘‘Alternative Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors,’’ (May 16, 2007).X X ................................ ML070180692 

Commission SRM on SECY–07–0082 (August 10, 2007) ...................................................... X X ................................ ML072220595 

Memorandum from Luis A. Reyes to NRC Commissioners, ‘‘Plans And Schedule For The 

Rulemaking On Risk-Informed Changes To Loss-of-Coolant Accident Technical Requirements (April 1, 2008).

X X ................................ ML080370355 

NUREG–1488—Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear 

Power Plant Sites East of the Rocky Mountains (April 1994).

X X ................................ ML052640591 

NUREG–1829—Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the 

Elicitation Process (Draft Report; June 2005).

X X ................................ ML051520574 

NUREG–1829—Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the 

Elicitation Process (Final Report; March 2008).

X X ................................ ML082250436 

NUREG–1903—Seismic Considerations for the Transition Break Size (February 2008) ....... X X ................................ ML080880140 

NRC White Paper—Plant-Specific Applicability of 10 CFR 50.46a Technical Basis (February 2009).X X ................................ ML090350757 

Memorandum from Arthur T. Howell to William F. Kane, ‘‘Degradation of the Davis-Besse 

Nuclear Power Station Reactor Pressure Vessel Head Lessons-Learned Report’’; (September 30, 2002).

X X ................................ ML022740211 

Regulatory Analysis .................................................................................................................. X X ................................ ML091050748 

XII. Plain Language 

The Presidential memorandum dated 

June 1, 1998, entitled ‘‘Plain Language 

in Government Writing’’ directed that 

the Government’s writing be in plain 

language. This memorandum was 

published on June 10, 1998 (63 FR 

31883). The NRC requests comments on 

the proposed rule specifically with 

respect to the clarity and reflectiveness 

of the language used. Comments should 

be sent to the address listed under the 

ADDRESSES caption of the preamble. 

XIII. Voluntary Consensus Standards 

The National Technology Transfer 

and Advancement Act of 1995, Public 

Law 104–113, requires that Federal 

agencies use technical standards that are 

developed or adopted by voluntary 

consensus standards bodies unless 

using such a standard is inconsistent 

with applicable law or is otherwise 

impractical. In this proposed rule, the 

NRC proposes to use the following 

Government-unique standard: 10 CFR 

50.46a. The NRC notes the ongoing 

development of voluntary consensus 

standards on PRAs, such as the ASME/ 

ANS RA–Sa–2009 consensus standard 

on Probabilistic Risk Assessment for 

Nuclear Power Plant Applications. The 

Government standards would allow the 

use of voluntary consensus standards, 

but would not require their use. The 

NRC does not believe that these other 

standards are sufficient to specify the 

necessary requirements for licensees 

who wish to modify plant ECCS 

analysis methods and nuclear power 

reactor designs based on the results of 

probabilistic risk analysis. The NRC is 

not aware of any voluntary consensus 

standard addressing risk-informed ECCS 

design and consequent changes in a 

light-water power reactor facility, 

technical specifications, or procedures 

that could be used instead of the 

proposed Government-unique standard. 

The NRC will consider using a 

voluntary consensus standard if an 

appropriate standard is identified. If a 

voluntary consensus standard is 

identified for consideration, the 

submittal should explain how the 

voluntary consensus standard is 

comparable and why it should be used 

instead of the proposed Governmentunique standard. 

XIV. Finding of No Significant 

Environmental Impact: Environmental 

Assessment 

The NRC 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, 

therefore, an environmental impact 

statement is not required. The basis for 

this determination is as follows: 

This action stems from the NRC’s 

ongoing efforts to risk-inform its 

regulations. If adopted, the proposed 

rule would establish a voluntary 

alternative set of risk-informed 

requirements for emergency core 

cooling systems. The alternative 

requirements are less stringent in the 

area of large break loss-of-coolant 

accidents (LOCAs). Using the alternative 

ECCS requirements will provide some 

licensees with opportunities to change 

various aspects of plant design to 

increase operational flexibility, increase 

power, or decrease costs. Licensee 

actions taken under the proposed rule 

could either decrease the probability of 

an accident or increase the probability 

of an accident by a very small amount. 

Mitigation of LOCAs of all sizes would 

still be required but with less 

redundancy and margin for the larger, 

low probability breaks. Increases in risk, 

if any, would be required to be very 

small so that adequate assurance of 

public health and safety is maintained. 

When considered together, the net effect 

of the licensee actions is expected to 

have an insignificant effect on accident 

probability. 

Thus, the proposed action would not 

significantly increase the probability or 

consequences of an accident, when 

considered in a risk-informed manner. 

No changes would be made in the types 

or quantities of radiological effluents 

that may be released offsite, and there 

is no significant increase in public 

radiation exposure because there is no 

change to facility operations that could 

create a new or significantly affect a 


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previously analyzed accident or release 

path. 

With regard to non-radiological 

impacts, no changes would be made to 

non-radiological plant effluents and 

there would be no changes in activities 

that would adversely affect the 

environment. Therefore, there are no 

significant non-radiological impacts 

associated with the proposed action. 

The primary alternative would be the 

no action alternative. The no action 

alternative, at worst, would result in no 

changes to current levels of safety, risk, 

or environmental impact. The no action 

alternative would also prevent licensees 

from making certain plant modifications 

that could be implemented under the 

proposed rule that could increase plant 

safety, increase operational flexibility, 

or decrease costs. The no action 

alternative would also maintain existing 

regulatory burdens for which there 

could be little or no safety, risk, or 

environmental benefits. 

The determination of this 

environmental assessment is that there 

will be no significant offsite impact to 

the public from this action. However, 

public stakeholders should note that the 

NRC is seeking public participation on 

this assessment. Comments on any 

aspect of the environmental assessment 

may be submitted to the NRC as 

indicated under the ADDRESSES heading 

of this document. 

The NRC has sent a copy of the 

environmental assessment and this 

proposed rule to every State Liaison 

Officer and requested their comments 

on the environmental assessment. 

XV. Paperwork Reduction Act 

Statement 

This proposed rule amends 

information collection requirements 

contained in 10 CFR part 50 that are 

subject to the Paperwork Reduction Act 

of 1995 (44 U.S.C. 3501 et seq). These 

information collection requirements 

have been submitted to the Office of 

Management and Budget (OMB) for 

approval. Existing requirements were 

approved by the Office of Management 

and Budget, control number 3150–0011. 

Type of submission: Revision. 

The title of the information collection: 

10 CFR part 50—Domestic Licensing of 

Production and Utilization Facilities. 

The form number if applicable: Not 

applicable. 

How often the collection is required: 

Annually. 

Who will be required or asked to 

report: Licensees authorized to operate 

a nuclear power reactor or applicants for 

standard design certifications, combined 

licenses, standard design approvals or 

manufacturing licenses who have been 

approved to implement the riskinformed alternative requirements in 10 

CFR 50.46a for analyzing the 

performance of emergency core cooling 

systems during loss-of-coolant 

accidents. 

An estimate of the number of annual 

responses: 12. 

The estimated number of annual 

respondents: 6. 

An estimate of the total number of 

hours needed annually to complete the 

requirement or request: 53,388 hours 

total, including 48,000 hours for 

reporting (an average of 8,000 hours per 

respondent) + 5,388 hours 

recordkeeping (an average of 898 hours 

per recordkeeper). 

Abstract: The Nuclear Regulatory 

Commission (NRC) proposes to amend 

its regulations to permit applicants for 

and/or holders of power reactor 

operating licenses, standard design 

certifications, combined licenses, 

standard design approvals or 

manufacturing licenses to choose to 

implement a risk-informed alternative to 

the current requirements for analyzing 

the performance of emergency core 

cooling systems (ECCS) during loss-ofcoolant accidents (LOCAs). In addition, 

the proposed rule would establish 

procedures and criteria for making 

changes in plant design and procedures 

based upon the results of the new 

analyses of ECCS performance during 

LOCAs. A licensee or applicant 

choosing to use the provisions of 

Section 50.46a would be required to 

submit a license amendment request 

with the required information, using the 

existing processes in Section 50.34 and 

Section 50.90. 

The U.S. Nuclear Regulatory 

Commission 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 OMB 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 20852. The 

OMB clearance package and rule are 

available at the NRC worldwide Web 

site: http://www.nrc.gov/public-involve/ 

doc-comment/omb/index.html for 60 

days after the signature date of this 

notice. 

Send comments on any aspect of 

these proposed information collections, 

including suggestions for reducing the 

burden and on the above issues, by 

September 9, 2009 to the Records and 

FOIA/Privacy Services Branch (T–5 

F53), U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, or by Internet electronic mail to 

INFOCOLLECTS.Resource@NRC.gov 

and to the Desk Officer, Christine Kymn, 

Office of Information and Regulatory 

Affairs, NEOB–10202, (3150–0011), 

Office of Management and Budget, 

Washington, DC 20503. Comments on 

the proposed information collection 

may also be submitted via the Federal 

eRulemaking Portal http:// 

www.regulations.gov, docket # NRC– 

2004–0006. 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. You 

may also e-mail comments to 

Christine_J._Kymn@omb.eop.gov or 

comment by telephone at (202) 395– 

4638. 

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 OMB control 

number. 

XVI. Regulatory Analysis 

The NRC has prepared a draft 

regulatory analysis on this proposed 

regulation. The analysis examines the 

costs and benefits of the alternatives 

considered by the NRC. The NRC 

requests public comment on the draft 

regulatory analysis. Availability of the 

regulatory analysis is provided in 

Section X of this document. Comments 

on the draft analysis may be submitted 

to the NRC as indicated under the 

ADDRESSES heading of this document. 

XVII. 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 


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the size standards established by the 

NRC (10 CFR 2.810). 

XVIII. Backfit Analysis 

The NRC has determined that the 

proposed rule generally does not 

constitute backfitting as defined in the 

backfit rule, 10 CFR 50.109(a)(1), and 

that three provisions of the proposed 

rule effectively excluding certain 

actions from the purview of the backfit 

rule, viz., § 50.109(b)(2); § 50.46a(d)(4), 

and § 50.46a(m), are appropriate. The 

basis for each of these determinations 

follows. 

The NRC has determined that the 

proposed rule does not constitute 

backfitting because it provides a 

voluntary alternative to the existing 

requirements in 10 CFR 50.46 for 

evaluating the performance of an ECCS 

for light-water nuclear power plants. A 

licensee may decide to either comply 

with the requirements of § 50.46a, or to 

continue to comply with the existing 

licensing basis of their plant with 

respect to ECCS analyses. Therefore, the 

backfit rule does not require the 

preparation of a backfit analysis for the 

proposed rule. 

As discussed in Section V.B of this 

document, the NRC may undertake 

future rulemaking to revise the TBS 

based upon re-evaluations of LOCA 

frequencies occurring after the effective 

date of a final rule. A proposed 

amendment to the backfit rule, 

§ 50.109(b)(2), would provide that future 

changes to the TBS would not be subject 

to the backfit rule. The NRC has 

determined that there is no statutory bar 

to the adoption of such a provision. The 

NRC also believes that the proposed 

exclusion of such rulemakings from the 

backfit rule is appropriate. The NRC 

intends to revise the TBS in § 50.46a 

rarely and only if necessary based upon 

public health and safety and/or common 

defense and security considerations. 

The NRC also does not regard the 

proposed exclusion as allowing the NRC 

to adopt cost-unjustified changes to the 

TBS. The NRC prepares a regulatory 

analysis for each substantive regulatory 

action which identifies the regulatory 

objectives of the proposed action, and 

evaluates the costs and benefits of 

proposed alternatives for achieving 

those regulatory objectives. The NRC 

has also adopted guidelines governing 

treatment of individual requirements in 

a regulatory analysis (69 FR 29187; May 

21, 2004). The NRC believes that a 

regulatory analysis performed in 

accordance with these guidelines will 

be effective in identifying unjustified 

regulatory proposals. In addition, this 

revised proposed rulemaking as applied 

to licensees who have not yet 

transferred to § 50.46a would not 

constitute backfitting for those 

licensees, inasmuch as the backfit rule 

does not protect a future applicant who 

has no reasonable expectation that 

requirements will remain static. The 

policies underlying the backfit rule 

apply only to licensees who have 

already received regulatory approval. 

Accordingly, the NRC concludes that 

the proposed exclusion in § 50.109(b)(2) 

of future changes to the TBS from the 

requirements of the backfit rule is 

appropriate. 

As discussed in Section V.E of this 

document, § 50.46a(d)(4) would require 

that a PRA used to demonstrate 

compliance with the risk acceptance 

criteria in § 50.46a(f)(1) or (f)(2) be 

periodically re-evaluated and updated, 

and that the licensee implement 

changes to the facility and procedures as 

necessary to ensure that the acceptance 

criteria continue to be met. To ensure 

that such a re-evaluation and updating 

of the PRA and any necessary changes 

to a facility and its procedures under 

§ 50.46a(d)(4) are not considered 

backfitting, § 50.46a(d)(4) would 

provide that such a re-evaluation, 

updating, and changes are not deemed 

to be backfitting. The NRC believes that 

this exclusion from the backfit rule is 

appropriate, inasmuch as application of 

the backfit rule in this context would 

effectively favor increases in risk. This 

is because most facility and procedure 

changes involve an up-front cost to 

implement a change which must be 

recovered over the remaining operating 

life of the facility in order to be 

considered cost-effective. For example, 

assume that after a change is 

implemented, subsequent PRA analyses 

suggest that the change should be 

‘‘rescinded’’ (either the hardware is 

restored to the original configuration or 

the new configuration is not credited in 

design bases analyses) in order to 

maintain the assumed risk level. The 

cost/benefit determination of the 

second, ‘‘restoring’’ change must 

address the unrecovered cost of the first 

change and the cost of the second, 

‘‘restoring’’ change. In most cases, 

application of cost/benefit analyses in 

evaluating the second, ‘‘restoring’’ 

change would skew the decision-making 

in favor of accepting the existing plant 

with the higher risk. Accumulation of 

these incremental increases in risk does 

not appear to be an appropriate 

regulatory approach. Accordingly, the 

NRC concludes that the backfitting 

exclusion in § 50.46a(d)(4) is 

appropriate. 

Section 50.46a(m) would provide that 

if the NRC changes the TBS specified in 

§ 50.46a, licensees who have evaluated 

their ECCS under § 50.46a shall 

undertake additional actions to ensure 

that the relevant acceptance criteria for 

ECCS performance are met with the new 

TBSs, and that these licensee actions are 

not to be considered backfitting. 

Consequently, the NRC may require 

licensees to take action under 

§ 50.46a(m) without consideration of the 

backfit rule. The NRC has determined 

that there is no statutory bar to the 

adoption of this provision, and that the 

proposed provision represents a 

justified departure from the principles 

underlying the backfit rule. First, the 

NRC’s decision on this matter 

recognizes that any future rulemaking to 

alter the TBS will require preparation of 

a regulatory analysis. As discussed, the 

regulatory analysis will ordinarily 

include a cost/benefit analysis 

addressing whether the costs of the TBS 

redefinition are justified in view of the 

benefits attributable to the redefinition. 

Second, the licensee has substantial 

flexibility under the proposed rule to 

determine the actions (reanalysis, 

procedure and operational changes, 

design-related changes, or a 

combination thereof) necessary to 

demonstrate compliance with the 

relevant ECCS acceptance criteria. The 

performance-based approach of the 

revised proposed rule lends substantial 

flexibility to the licensee and may tend 

to reduce the burden associated with 

changes in the TBS. Accordingly, the 

NRC concludes that the backfitting 

exclusion in § 50.46a(m) is appropriate. 

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. 

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; 

and 5 U.S.C. 553; the NRC is proposing 


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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 part 50 

continues to read as follows: 

Authority: Secs. 102, 103, 104, 105, 161, 

182, 183, 186, 189, 68 Stat. 936, 937, 938, 

948, 953, 954, 955, 956, as amended, sec. 

234, 83 Stat. 444, as amended (42 U.S.C. 

2132, 2133, 2134, 2135, 2201, 2232, 2233, 

2236, 2239, 2282); secs. 201, as amended, 

202, 206, 88 Stat. 1242, as amended, 1244, 

1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 

112 Stat. 2750 (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, 92 Stat. 2951 

as amended by Pub. L. 102–486, sec. 2902, 

106 Stat. 3123 (42 U.S.C. 5841). Section 50.10 

also issued under secs. 101, 185, 68 Stat. 955, 

as amended (42 U.S.C. 2131, 2235); sec. 102, 

Pub. L. 91–190, 83 Stat. 853 (42 U.S.C. 4332). 

Sections 50.13, 50.54(dd), and 50.103 also 

issued under sec. 108, 68 Stat. 939, as 

amended (42 U.S.C. 2138). 

Sections 50.23, 50.35, 50.55, and 50.56 also 

issued under sec. 185, 68 Stat. 955 (42 U.S.C. 

2235). Sections 50.33a, 50.55a and Appendix 

Q also issued under sec. 102, Pub. L. 91–190, 

83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 

and 50.54 also issued under sec. 204, 88 Stat. 

1245 (42 U.S.C. 5844). Sections 50.58, 50.91, 

and 50.92 also issued under Pub. L. 97–415, 

96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 

also issued under sec. 122, 68 Stat. 939 (42 

U.S.C. 2152). Sections 50.80–50.81 also 

issued under sec. 184, 68 Stat. 954, as 

amended (42 U.S.C. 2234). Appendix F also 

issued under sec. 187, 68 Stat. 955 (42 U.S.C. 

2237) 

2. In § 50.34, paragraphs (a)(4) and 

(b)(4) are revised to read as follows: 

§ 50.34 Contents of application; technical 

information. 

(a) * * * 

(4) A preliminary analysis and 

evaluation of the design and 

performance of structures, systems, and 

components of the facility with the 

objective of assessing the risk to public 

health and safety resulting from 

operation of the facility and including 

determination of the margins of safety 

during normal operations and transient 

conditions anticipated during the life of 

the facility, and the adequacy of 

structures, systems, and components 

provided for the prevention of accidents 

and the mitigation of the consequences 

of accidents. 

(i) Analysis and evaluation of ECCS 

cooling performance and the need for 

high point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

either § 50.46 or § 50.46a, and § 50.46b 

for facilities whose operating licenses 

were issued after December 28, 1974, 

but before [EFFECTIVE DATE OF 

RULE], and for facilities for which 

construction permits may be issued after 

[EFFECTIVE DATE OF RULE] and are 

demonstrated under § 50.46a(c)(2) to 

have designs that are similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE]. 

(ii) Analysis and evaluation of ECCS 

cooling performance and the need for 

high point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

§ 50.46 and § 50.46b for facilities for 

which construction permits may be 

issued after [EFFECTIVE DATE OF 

RULE] and are not demonstrated under 

§ 50.46a(c)(2) to have designs that are 

similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]. 

* * * * * 

(b) * * * 

(4) A final analysis and evaluation of 

the design and performance of 

structures, systems, and components 

with the objective stated in paragraph 

(a)(4) of this section and taking into 

account any pertinent information 

developed since the submittal of the 

preliminary safety analysis report. 

(i) Analysis and evaluation of ECCS 

cooling performance following 

postulated LOCAs must be performed 

under the requirements of either § 50.46 

or § 50.46a, and § 50.46b for facilities 

whose operating licenses were issued 

after December 28, 1974, but before 

[EFFECTIVE DATE OF RULE], and for 

facilities whose operating licenses are 

issued after [EFFECTIVE DATE OF 

RULE] and are demonstrated under 

§ 50.46a(c)(2) to have designs that are 

similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]. 

(ii) Analysis and evaluation of ECCS 

cooling performance following 

postulated LOCAs must be performed 

under the requirements of §§ 50.46 and 

50.46b for facilities whose operating 

licenses are issued after [EFFECTIVE 

DATE OF RULE] and are not 

demonstrated under § 50.46a(c)(2) to 

have designs that are similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE]. 

* * * * * 

3. In § 50.46, paragraph (a) is 

amended by adding an introductory 

paragraph and revising paragraph 

(a)(1)(i) to read as follows: 

§ 50.46 Acceptance criteria for emergency 

core cooling systems for light-water nuclear 

power plants. 

(a) Each boiling or pressurized lightwater nuclear power reactor fueled with 

uranium oxide pellets within 

cylindrical zircalloy or ZIRLO cladding 

must be provided with an emergency 

core cooling system (ECCS). The ECCS 

system must be designed under the 

requirements of this section or § 50.46a 

for facilities whose operating licenses 

were issued before [EFFECTIVE DATE 

OF RULE]; for facilities whose operating 

licenses, combined licenses under part 

52 of this chapter, or manufacturing 

licenses under part 52 of this chapter 

are issued after [EFFECTIVE DATE OF 

RULE] and are demonstrated under 

§ 50.46a(c)(2) to have designs that are 

similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]; and for design approvals and 

design certifications under part 52 of 

this chapter issued after [EFFECTIVE 

DATE OF RULE] that are demonstrated 

under § 50.46a(c)(2) to have designs that 

are similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]. The ECCS system must be 

designed under the requirements of this 

section for facilities whose operating 

licenses, combined licenses under part 

52 of this chapter, or manufacturing 

licenses under part 52 of this chapter 

are issued after [EFFECTIVE DATE OF 

RULE] and are not demonstrated under 

§ 50.46a(c)(2) to have designs that are 

similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]; and for design approvals and 

design certifications under part 52 of 

this chapter that are not demonstrated 

under § 50.46a(c)(2) to have designs that 

are similar to the designs of reactors 

licensed before [EFFECTIVE DATE OF 

RULE]. 

(1)(i) The ECCS system must be 

designed so that its calculated cooling 

performance following postulated 

LOCAs conforms to the criteria set forth 

in paragraph (b) of this section. ECCS 

cooling performance must be calculated 

in accordance with an acceptable 

evaluation model and must be 

calculated for a number of postulated 

LOCAs of different sizes, locations, and 

other properties sufficient to provide 

assurance that the most severe 

postulated LOCAs are calculated. 

Except as provided in paragraph 

(a)(1)(ii) of this section, the evaluation 

model must include sufficient 

supporting justification to show that the 

analytical technique realistically 

describes the behavior of the reactor 

system during a LOCA. Comparisons to 

applicable experimental data must be 

made and uncertainties in the analysis 

method and inputs must be identified 

and assessed so that the uncertainty in 

the calculated results can be estimated. 

This uncertainty must be accounted for, 


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so that, when the calculated ECCS 

cooling performance is compared to the 

criteria set forth in paragraph (b) of this 

section, there is a high level of 

probability that the criteria would not 

be exceeded. Appendix K, Part II 

Required Documentation, sets forth the 

documentation requirements for each 

evaluation model. This section does not 

apply to a nuclear power reactor facility 

for which the certifications required 

under § 50.82(a)(1) have been submitted. 

* * * * * 

4. Section 50.46a is redesignated as 

§ 50.46b, and a new § 50.46a is added to 

read as follows: 

§ 50.46a Alternative acceptance criteria for 

emergency core cooling systems for lightwater nuclear power reactors. 

(a) Definitions. For the purposes of 

this section: 

(1) Changes enabled by this section 

means changes to the facility, technical 

specifications, and procedures that 

satisfy the alternative ECCS analysis 

requirements under this section but do 

not satisfy the ECCS requirements under 

10 CFR 50.46. 

(2) Evaluation model means the 

calculational framework for evaluating 

the behavior of the reactor system 

during a postulated design-basis loss-ofcoolant accident (LOCA). It includes 

one or more computer programs and all 

other information necessary for 

application of the calculational 

framework to a specific LOCA, such as 

mathematical models used, assumptions 

included in the programs, procedure for 

treating the program input and output 

information, specification of those 

portions of analysis not included in 

computer programs, values of 

parameters, and all other information 

necessary to specify the calculational 

procedure. 

(3) Loss-of-coolant accidents (LOCAs) 

means the hypothetical accidents that 

would result from the loss of reactor 

coolant, at a rate in excess of the 

capability of the reactor coolant makeup 

system, from breaks in pipes in the 

reactor coolant pressure boundary up to 

and including a break equivalent in size 

to the double-ended rupture of the 

largest pipe in the reactor coolant 

system. LOCAs involving breaks at or 

below the transition break size (TBS) are 

design-basis accidents. LOCAs 

involving breaks larger than the TBS are 

beyond design-basis accidents. 

(4) Operating configuration means 

those plant characteristics, such as 

power level, equipment unavailability 

(including unavailability caused by 

corrective and preventive maintenance), 

and equipment capability that affect 

plant response to a LOCA. 

(5) Transition break size (TBS) for 

reactors licensed before [EFFECTIVE 

DATE OF RULE] is a break area equal 

to the cross-sectional flow area of the 

inside diameter of the largest piping 

attached to the reactor coolant system 

for a pressurized water reactor, or the 

larger of the feedwater line inside 

containment or the residual heat 

removal line inside containment for a 

boiling water reactor. For reactors 

licensed after [EFFECTIVE DATE OF 

RULE], the TBS will be determined on 

a plant-specific basis. 

(b) Applicability and scope. 

(1) The requirements of this section 

may be applied to each boiling or 

pressurized light-water nuclear power 

reactor fueled with uranium oxide 

pellets within cylindrical zircalloy or 

ZIRLO cladding whose operating license 

was issued prior to [EFFECTIVE DATE 

OF RULE]; to each boiling or 

pressurized light-water nuclear power 

reactor fueled with uranium oxide 

pellets within cylindrical zircalloy or 

ZIRLO cladding whose operating 

license, combined license under part 52 

of this chapter or manufacturing license 

under part 52 of this chapter is issued 

after [EFFECTIVE DATE OF RULE] and 

whose design is demonstrated under 

§ 50.46a(c)(2) to be similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE]; and to 

each boiling or pressurized light-water 

nuclear power reactor fueled with 

uranium oxide pellets within 

cylindrical zircalloy or ZIRLO cladding 

whose design approval or design 

certification under part 52 of this 

chapter is demonstrated under 

§ 50.46a(c)(2) to be similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE]. The 

requirements of this section do not 

apply to a reactor for which the 

certification required under § 50.82(a)(1) 

has been submitted. 

(2) The requirements of this section 

are in addition to any other 

requirements applicable to ECCS set 

forth in this part, with the exception of 

§ 50.46. The criteria set forth in 

paragraphs (e)(3) and (e)(4) of this 

section, with cooling performance 

calculated in accordance with an 

acceptable evaluation model or analysis 

method under paragraphs (e)(1) and 

(e)(2) of this section, are in 

implementation of the general 

requirements with respect to ECCS 

cooling performance design set forth in 

this part, including in particular 

Criterion 35 of Appendix A to this part. 

(c) Application. (1) A licensee of a 

facility seeking to implement this 

section shall submit an application for 

a license amendment under § 50.90 that 

contains the following information: 

(i) A written evaluation demonstrating 

applicability of the results in NUREG– 

1829, ‘‘Estimating Loss-of-Coolant 

Accident (LOCA) Frequencies through 

the Elicitation Process’’; March 2008 

and NUREG–1903, ‘‘Seismic 

Considerations for the Transition Break 

Size’’; February 2008’’ to the licensee’s 

facility. As part of this evaluation, the 

application must contain a plant 

specific analysis demonstrating that the 

risk of seismically-induced LOCAs 

larger than the TBS is comparable to or 

less than the seismically-induced LOCA 

risk associated with the NUREG–1903 

results. 

(ii) Identification of the approved 

analysis method(s) for demonstrating 

compliance with the ECCS criteria in 

paragraph (e) of this section. 

(iii) A description of the risk-informed 

evaluation process used in evaluating 

whether proposed changes to the facility 

meet the requirements in paragraph (f) 

of this section. 

(iv) A licensee who wishes to make 

changes enabled by this section without 

prior NRC review and approval must 

submit for NRC approval a process to be 

used for evaluating the acceptability of 

these changes; including: 

(A) A description of the approach, 

methods, and decisionmaking process to 

be used for evaluating compliance with 

the acceptance criteria in paragraphs 

(f)(1), (f)(2), and (f)(3) of this section, 

and 

(B) A description of the licensee’s 

PRA model and non-PRA risk 

assessment methods to be used for 

demonstrating compliance with 

paragraphs (f)(4) and (f)(5) of this 

section. 

(v) A description of non safety 

equipment that is credited for 

demonstrating compliance with the 

ECCS acceptance criteria in paragraph 

(e) of this section. 

(2) An applicant for a construction 

permit, operating license, design 

approval, design certification, 

manufacturing license, or combined 

license seeking to implement the 

requirements of this section shall, in 

addition to the information required by 

paragraph (c)(1) of this section, submit 

an analysis demonstrating why the 

proposed reactor design is similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE] such that 

the provisions of this section may 

properly apply. The analysis must also 

include a recommendation for an 

appropriate TBS and a justification that 

the recommended TBS is consistent 

with the technical basis for this section. 


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(3) Acceptance criteria. The NRC may 

approve an application to use this 

section if: 

(i) The evaluation submitted under 

paragraph (c)(1)(i) of this section 

demonstrates that the NUREG–1829 

results are applicable to the facility, and 

the risk of seismically-induced LOCAs 

larger than the TBS is comparable to or 

less than the seismically-induced LOCA 

risk associated with the NUREG–1903 

results; 

(ii) The method(s) for demonstrating 

compliance with the ECCS acceptance 

criteria in paragraphs (e)(3) and (e)(4) of 

this section meet the requirements in 

paragraphs (e)(1) and (e)(2) of this 

section; 

(iii) The risk-informed evaluation 

process the licensee proposes to use for 

making changes enabled by this section 

is adequate for determining whether the 

acceptance criteria in paragraph (f) of 

this section have been met; and 

(iv) Non safety equipment that is 

credited for demonstrating compliance 

with the ECCS acceptance criteria in 

paragraph (e) of this section is identified 

in plant Technical Specifications. 

(v) For all applicants other than those 

holding operating licenses issued before 

[EFFECTIVE DATE OF RULE], the 

proposed reactor design is similar to the 

designs of reactors licensed before 

[EFFECTIVE DATE OF RULE] and the 

applicant’s proposed TBS is consistent 

with the technical basis of this section. 

(d) Requirements during operation. A 

licensee whose application under 

paragraph (c) of this section is approved 

by the NRC shall comply with the 

following requirements as long as the 

facility is subject to the requirements in 

this section until the licensee submits 

the certifications required by § 50.82(a): 

(1) The licensee shall maintain ECCS 

model(s) and/or analysis method(s) 

meeting the requirements in paragraphs 

(e)(1) and (e)(2) of this section; 

(2) The licensee shall have leak 

detection systems available at the 

facility and shall implement actions as 

necessary to identify, monitor and 

quantify leakage to ensure that adverse 

safety consequences do not result from 

primary pressure boundary leakage from 

piping and components that are larger 

than the transition break size. 

(3) A change enabled by this section 

must, in addition to meeting other 

applicable NRC requirements, be 

evaluated by a risk-informed evaluation 

demonstrating that the acceptance 

criteria in paragraph (f) of this section 

are met. 

(4) The licensee shall periodically 

maintain and upgrade, as necessary, its 

risk assessments to meet the 

requirements in paragraph (f)(4) and 

(f)(5) of this section. The maintenance 

and upgrading shall be consistent with 

NRC-endorsed consensus standards on 

PRA and must be completed in a timely 

manner, but no less often than once 

every two refueling outages. Based upon 

a re-evaluation of the risk assessments 

after the periodic maintenance and 

upgrading are completed, the licensee 

shall take appropriate action to ensure 

that the acceptance criteria in 

paragraphs (f)(2) or (f)(3) of this section, 

as applicable, are met. The PRA 

maintenance and upgrading required by 

this section, and any necessary changes 

to the facility, technical specifications 

and procedures as a result of this reevaluation, shall not be deemed to be 

backfitting under any provision of this 

chapter. 

(5) For LOCAs larger than the TBS, 

operation in a plant operating 

configuration not demonstrated to meet 

the acceptance criteria in paragraph 

(e)(4) of this section may not exceed a 

total of fourteen days in any 12 month 

period. 

(6) The licensee shall perform an 

evaluation to determine the effect of all 

planned facility changes and shall not 

implement any facility change that 

would invalidate the evaluation 

performed pursuant to § 50.46a(c)(1)(i) 

demonstrating the applicability to the 

licensee’s facility of the generic results 

in NUREG–1829 and NUREG–1903. 

(e) ECCS Performance. Each nuclear 

power reactor subject to this section 

must be provided with an ECCS that 

must be designed so that its calculated 

cooling performance following 

postulated LOCAs conforms to the 

criteria set forth in this section. The 

evaluation models for LOCAs must meet 

the criteria in this paragraph, and must 

be approved for use by the NRC. 

Appendix K, Part II, to 10 CFR Part 50, 

sets forth the documentation 

requirements for evaluation models. 

(1) ECCS evaluation for LOCAs 

involving breaks at or below the TBS. 

ECCS cooling performance at or below 

the TBS must be calculated in 

accordance with an evaluation model 

that meets the requirements of either 

section I to Appendix K of this part, or 

the following requirements, and must 

demonstrate that the acceptance criteria 

in paragraph (e)(3) of this section are 

satisfied. The evaluation model must be 

used for a number of postulated LOCAs 

of different sizes, locations, and other 

properties sufficient to provide 

assurance that the most severe 

postulated LOCAs involving breaks at or 

below the TBS are analyzed. The 

evaluation model must include 

sufficient supporting justification to 

show that the analytical technique 

realistically describes the behavior of 

the reactor system during a LOCA. 

Comparisons to applicable experimental 

data must be made and uncertainties in 

the analysis method and inputs must be 

identified and assessed so that the 

uncertainty in the calculated results can 

be estimated. This uncertainty must be 

accounted for, so that when the 

calculated ECCS cooling performance is 

compared to the criteria set forth in 

paragraph (e)(3) of this section, there is 

a high level of probability that the 

criteria would not be exceeded. 

(2) ECCS analyses for LOCAs 

involving breaks larger than the TBS. 

ECCS cooling performance for LOCAs 

involving breaks larger than the TBS 

must be calculated in accordance with 

an evaluation model that meets the 

requirements of either section I to 

Appendix K of this part, or the 

following requirements, and must 

demonstrate that the acceptance criteria 

in paragraph (e)(4) of this section are 

satisfied. The evaluation model must 

include sufficient supporting 

justification to show that the analytical 

technique realistically describes the 

behavior of the reactor system during a 

LOCA. Comparisons to applicable 

experimental data must be made and 

uncertainties in the analysis method 

and inputs must be identified and 

assessed so that the uncertainty in the 

calculated results can be estimated. This 

uncertainty must be accounted for, so 

that when the calculated ECCS cooling 

performance is compared to the criteria 

set forth in paragraph (e)(4) of this 

section, there is a high level of 

probability that the criteria would not 

be exceeded. The evaluation model 

must be used for a number of postulated 

LOCAs of different sizes, locations, and 

other properties sufficient to provide 

assurance that the most severe 

postulated LOCAs larger than the TBS 

up to the double-ended rupture of the 

largest pipe in the reactor coolant 

system are analyzed. These calculations 

may take credit for the availability of 

offsite power and do not require the 

assumption of a single failure. Realistic 

initial conditions and availability of 

safety-related or non safety-related 

equipment may be assumed if supported 

by plant-specific data or analysis, and 

provided that onsite power can be 

readily provided through simple manual 

actions to equipment that is credited in 

the analysis. 

(3) Acceptance criteria for LOCAs 

involving breaks at or below the TBS. 

The following acceptance criteria must 

be used in determining the acceptability 

of ECCS cooling performance: 

(i) Peak cladding temperature. The 

calculated maximum fuel element 


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cladding temperature must not exceed 

2200 °F. 

(ii) Maximum cladding oxidation. The 

calculated total oxidation of the 

cladding must not at any location 

exceed 0.17 times the total cladding 

thickness before oxidation. As used in 

this paragraph, total oxidation means 

the total thickness of cladding metal 

that would be locally converted to oxide 

if all the oxygen absorbed by and 

reacted with the cladding locally were 

converted to stoichiometric zirconium 

dioxide. If cladding rupture is 

calculated to occur, the inside surfaces 

of the cladding must be included in the 

oxidation, beginning at the calculated 

time of rupture. Cladding thickness 

before oxidation means the radial 

distance from inside to outside the 

cladding, after any calculated rupture or 

swelling has occurred but before 

significant oxidation. Where the 

calculated conditions of transient 

pressure and temperature lead to a 

prediction of cladding swelling, with or 

without cladding rupture, the 

unoxidized cladding thickness must be 

defined as the cladding cross-sectional 

area, taken at a horizontal plane at the 

elevation of the rupture, if it occurs, or 

at the elevation of the highest cladding 

temperature if no rupture is calculated 

to occur, divided by the average 

circumference at that elevation. For 

ruptured cladding the circumference 

does not include the rupture opening. 

(iii) Maximum hydrogen generation. 

The calculated total amount of hydrogen 

generated from the chemical reaction of 

the cladding with water or steam must 

not exceed 0.01 times the hypothetical 

amount that would be generated if all of 

the metal in the cladding cylinders 

surrounding the fuel, excluding the 

cladding surrounding the plenum 

volume, were to react. 

(iv) Coolable geometry. Calculated 

changes in core geometry must be such 

that the core remains amenable to 

cooling. 

(v) Long term cooling. After any 

calculated successful initial operation of 

the ECCS, the calculated core 

temperature must be maintained at an 

acceptably low value and decay heat 

must be removed for the extended 

period of time required by the longlived radioactivity remaining in the 

core. 

(4) Acceptance criteria for LOCAs 

involving breaks larger than the TBS. 

The following acceptance criteria must 

be used in determining the acceptability 

of ECCS cooling performance: 

(i) Coolable geometry. Calculated 

changes in core geometry must be such 

that the core remains amenable to 

cooling. 

(ii) Long term cooling. After any 

calculated successful initial operation of 

the ECCS, the calculated core 

temperature must be maintained at an 

acceptably low value and decay heat 

must be removed for the extended 

period of time required by the longlived radioactivity remaining in the 

core. 

(5) Imposition of restrictions. The 

Director of the Office of Nuclear Reactor 

Regulation may impose restrictions on 

reactor operation if it is found that the 

evaluations of ECCS cooling 

performance submitted are not 

consistent with paragraph (e) of this 

section. 

(f) Changes to facility, technical 

specifications, or procedures. A licensee 

who wishes to make changes to the 

facility or procedures or to the technical 

specifications enabled by this rule shall 

perform a risk-informed evaluation. 

(1) The licensee may make such 

changes without prior NRC approval if: 

(i) The change is permitted under 

§ 50.59, 

(ii) The risk informed evaluation 

process described in paragraph (c)(1)(iii) 

of this section demonstrates that any 

increases in the estimated risk are 

minimal compared to the overall plant 

risk profile, and the criteria in 

paragraph (f)(3) of this section are met, 

and 

(iii) The change does not invalidate 

the evaluation performed pursuant to 

paragraph (c)(1)(i) of the applicability of 

the results in NUREG–1829 and 

NUREG–1903 to the licensee’s facility. 

(2) For implementing changes which 

are not permitted under paragraph (f)(1) 

of this section, the licensee must submit 

an application for license amendment 

under § 50.90. The application must 

contain: 

(i) The information required under 

§ 50.90; 

(ii) For applicants whose operating 

licenses were issued before [EFFECTIVE 

DATE OF RULE], information from the 

risk-informed evaluation demonstrating 

that the total increases in core damage 

frequency and large early release 

frequency are very small and the overall 

risk remains small, and the criteria in 

paragraph (f)(3) of this section are met; 

(iii) For applicants whose operating 

licenses were not issued before 

[EFFECTIVE DATE OF RULE], 

information from the risk-informed 

evaluation demonstrating that the total 

increases in core damage frequency and 

large release frequency are very small 

and the overall risk remains small, and 

the criteria in paragraph (f)(3) of this 

section are met; 

(iv) If previous changes have been 

made under § 50.46a, information from 

the risk-informed evaluation on the 

cumulative effect on risk of the 

proposed change and all previous 

changes made under this section. If 

more than one plant change is 

combined; including plant changes not 

enabled by this section, into a group for 

the purposes of evaluating acceptable 

risk increases; the evaluation of each 

individual change shall be performed 

along with the evaluation of combined 

changes; and 

(v) Information demonstrating that the 

criteria in paragraphs (e)(3) and (e)(4) of 

this section are met. 

(vi) Information demonstrating that 

the proposed change will not increase 

the LOCA frequency of the facility 

(including the frequency of seismicallyinduced LOCAs) by an amount that 

would invalidate the applicability to the 

facility of the generic studies (NUREG– 

1829, ‘‘Estimating Loss-of-Coolant 

Accident (LOCA) Frequencies through 

the Elicitation Process’’, March 2008 

and NUREG–1903, ‘‘Seismic 

Considerations for the Transition Break 

Size’’, February 2008’’) that support the 

technical basis for this section. 

(3) All changes enabled by this rule 

must meet the following criteria: 

(i) Adequate defense in depth is 

maintained; 

(ii) Adequate safety margins are 

retained to account for uncertainties; 

and 

(iii) Adequate performancemeasurement programs are 

implemented to ensure the riskinformed evaluation continues to reflect 

actual plant design and operation. These 

programs shall be designed to detect 

degradation of the system, structure or 

component before plant safety is 

compromised, provide feedback of 

information and timely corrective 

actions, and monitor systems, structures 

or components at a level commensurate 

with their safety significance. 

(4) Requirements for risk 

assessment—PRA. Whenever a PRA is 

used in the risk-informed evaluation, 

the PRA must, with respect to the area 

of evaluation which is the subject of the 

PRA: 

(i) Address initiating events from 

sources both internal and external to the 

plant and for all modes of operation, 

including low power and shutdown 

modes, that would affect the regulatory 

decision in a substantial manner; 

(ii) Reasonably represent the current 

configuration and operating practices at 

the plant; 

(iii) Have sufficient technical 

adequacy (including consideration of 

uncertainty) and level of detail to 

provide confidence that the total risk 

estimate and the change in total risk 


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estimate adequately reflect the plant and 

the effect of the proposed change on 

risk; and 

(iv) Be determined, through peer 

review, to meet industry standards for 

PRA quality that have been endorsed by 

the NRC. 

(5) Requirements for risk assessment 

other than PRA. Whenever risk 

assessment methods other than PRAs 

are used to develop quantitative or 

qualitative estimates of changes to risk 

in the risk-informed evaluation, an 

integrated, systematic process must be 

used. All aspects of the analyses must 

reasonably reflect the current plant 

configuration and operating practices, 

and applicable plant and industry 

operating experience. 

(g) Reporting. (1) Each licensee shall 

estimate the effect of any change to or 

error in evaluation models or analysis 

methods or in the application of such 

models or methods to determine if the 

change or error is significant. For each 

change to or error discovered in an 

ECCS evaluation model or analysis 

method or in the application of such a 

model that affects the calculated results, 

the licensee shall report the nature of 

the change or error and its estimated 

effect on the limiting ECCS analysis to 

the Commission at least annually as 

specified in § 50.4. If the change or error 

is significant, the licensee shall provide 

this report within 30 days and include 

with the report a proposed schedule for 

providing a reanalysis or taking other 

action as may be needed to show 

compliance with § 50.46a requirements. 

This schedule may be developed using 

an integrated scheduling system 

previously approved for the facility by 

the NRC. For those facilities not using 

an NRC-approved integrated scheduling 

system, a schedule will be established 

by the NRC staff within 60 days of 

receipt of the proposed schedule. Any 

change or error correction that results in 

a calculated ECCS performance that 

does not conform to the criteria set forth 

in paragraphs (e)(3) or (e)(4) of this 

section is a reportable event as 

described in §§ 50.55(e), 50.72 and 

50.73. The licensee shall propose 

immediate steps to demonstrate 

compliance or bring plant design or 

operation into compliance with § 50.46a 

requirements. For the purpose of this 

paragraph, a significant change or error 

is: 

(i) For LOCAs involving pipe breaks 

at or below the TBS, one which results 

either in a calculated peak fuel cladding 

temperature different by more than 50 

°F from the temperature calculated for 

the limiting transient using the last 

acceptable model, or is a cumulation of 

changes and errors such that the sum of 

the absolute magnitudes of the 

respective temperature changes is 

greater than 50 °F; or 

(ii) For LOCAs involving pipe breaks 

larger than the TBS, one which results 

in a significant reduction in the 

capability to meet the requirements of 

paragraph (e)(4) of this section. 

(2) As part of the PRA maintenance 

and upgrading under paragraph (d)(4) of 

this section, the licensee shall report to 

the NRC if the re-evaluation results in 

exceeding the acceptance criteria in 

paragraphs (f)(1) or (f)(2) of this section, 

as applicable. The report must be filed 

with the NRC no more than 60 days 

after completing the PRA re-evaluation. 

The report must describe and explain 

the changes in the PRA modeling, plant 

design, or plant operation that led to the 

increase(s) in risk, and must include a 

description of and implementation 

schedule for any corrective actions 

required under paragraph (d)(4) of this 

section. 

(3) Every 24 months, the licensee 

shall submit, as specified in § 50.4, a 

short description of each change 

involving minimal changes in risk made 

under paragraph (f)(1) of this section 

after the last report and a brief summary 

of the basis for the licensee’s 

determination pursuant to 

§ 50.46a(f)(2)(vi) that the change does 

not invalidate the applicability 

evaluation made under § 50.46a(c)(1)(i). 

(h) Documentation. Following 

implementation of the § 50.46a 

requirements, the licensee shall 

maintain records sufficient to 

demonstrate compliance with the 

requirements in this section in 

accordance with § 50.71. 

(i) through (l)—[RESERVED] 

(m) Changes to TBS. If the NRC 

increases the TBS specified in this 

section applicable to a licensee’s 

nuclear power plant, each licensee 

subject to this section shall perform the 

evaluations required by paragraphs 

(e)(1) and (e)(2) of this section and 

reconfirm compliance with the 

acceptance criteria in paragraphs (e)(3) 

and (e)(4) of this section. If the licensee 

cannot demonstrate compliance with 

the acceptance criteria, then the licensee 

shall change its facility, technical 

specifications or procedures so that the 

acceptance criteria are met. The 

evaluation required by this paragraph, 

and any necessary changes to the 

facility, technical specifications or 

procedures as the result of this 

evaluation, must not be deemed to be 

backfitting under any provision of this 

chapter. 

5. In § 50.109, paragraph (b) is revised 

to read as follows: 

§ 50.109 Backfitting. 

* * * * * 

(b) Paragraph (a)(3) of this section 

shall not apply to: 

(1) Backfits imposed prior to October 

21, 1985; and 

(2) Any changes made to the TBS 

specified in § 50.46a or as otherwise 

applied to a licensee. 

* * * * * 

6. In Appendix A to 10 CFR Part 50, 

under the heading, ‘‘CRITERIA,’’ 

Criterion 17, 35, 38, 41, 44, and 50 are 

revised to read as follows: 

APPENDIX A TO PART 50—GENERAL 

DESIGN CRITERIA FOR NUCLEAR 

POWER PLANTS 

* * * * * 

CRITERIA 

* * * * * 

Criterion 17—Electrical power systems. An 

on-site electric power system and an offsite 

electric power system shall be provided to 

permit functioning of structures, systems, 

and components important to safety. The 

safety function for each system (assuming the 

other system is not functioning) shall be to 

provide sufficient capacity and capability to 

assure that (1) specified acceptable fuel 

design limits and design conditions of the 

reactor coolant pressure boundary are not 

exceeded as a result of anticipated 

operational occurrences and (2) the core is 

cooled and containment integrity and other 

vital functions are maintained in the event of 

postulated accidents. 

The onsite electric power supplies, 

including the batteries, and the onsite 

electrical distribution system, shall have 

sufficient independence, redundancy, and 

testability to perform their safety functions 

assuming a single failure, except for loss of 

coolant accidents involving pipe breaks 

larger than the transition break size under 

§ 50.46a, where a single failure of the onsite 

power supplies and electrical distribution 

system need not be assumed for plants under 

§ 50.46a. For those pipe breaks only, neither 

a single failure nor the unavailability of 

offsite power need be assumed. 

Electric power from the transmission 

network to the onsite electric distribution 

system shall be supplied by two physically 

independent circuits (not necessarily on 

separate rights of way) designed and located 

so as to minimize to the extent practical the 

likelihood of their simultaneous failure 

under operating and postulated accident 

conditions. A switchyard common to both 

circuits is acceptable. Each of these circuits 

shall be designed to be available in sufficient 

time following a loss of all onsite alternating 

current power supplies and the other offsite 

electric power circuit, to assure that specified 

acceptable fuel design limits and design 

conditions of the reactor coolant pressure 

boundary are not exceeded. One of these 

circuits shall be designed to be available 

within a few seconds following a LOCA to 

assure that core cooling, containment 

integrity, and other vital safety functions are 

maintained. 


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Federal Register / Vol. 74, No. 152 / Monday, August 10, 2009 / Proposed Rules 40051 

Provisions shall be included to minimize 

the probability of losing electric power from 

any of the remaining supplies as a result of, 

or coincident with, the loss of power 

generated by the nuclear power unit, the loss 

of power from the transmission network, or 

the loss of power from the onsite electric 

power supplies. 

* * * * * 

Criterion 35—Emergency core cooling. A 

system to provide abundant emergency core 

cooling shall be provided. The system safety 

function shall be to transfer heat from the 

reactor core following any loss of reactor 

coolant at a rate such that (1) fuel and clad 

damage that could interfere with continued 

effective core cooling is prevented and (2) 

clad metal-water reaction is limited to 

negligible amounts. 

Suitable redundancy in components and 

features, and suitable interconnections, leak 

detection, isolation, and containment 

capabilities shall be provided to assure that 

for onsite electric power system operation 

(assuming offsite power is not available) and 

for offsite electric power system operation 

(assuming onsite power is not available) the 

system safety function can be accomplished, 

assuming a single failure, except for loss of 

coolant accidents involving pipe breaks 

larger than the transition break size under 

§ 50.46a. For those pipe breaks only, neither 

a single failure nor the unavailability of 

offsite power need be assumed. 

* * * * * 

Criterion 38—Containment heat removal. 

A system to remove heat from the reactor 

containment shall be provided. The system 

safety function shall be to reduce rapidly, 

consistent with the functioning of other 

associated systems, the containment pressure 

and temperature following any LOCA and 

maintain them at acceptably low levels. 

Suitable redundancy in components and 

features, and suitable interconnections, leak 

detection, isolation, and containment 

capabilities shall be provided to assure that 

for onsite electric power system operation 

(assuming offsite power is not available) and 

for offsite electric power system operation 

(assuming onsite power is not available) the 

system safety function can be accomplished, 

assuming a single failure, except for analysis 

of loss of coolant accidents involving pipe 

breaks larger than the transition break size 

under § 50.46a. For those pipe breaks only, 

neither a single failure nor the unavailability 

of offsite power need be assumed. 

* * * * * 

Criterion 41—Containment atmosphere 

cleanup. Systems to control fission products, 

hydrogen, oxygen, and other substances 

which may be released into the reactor 

containment shall be provided as necessary 

to reduce, consistent with the functioning of 

other associated systems, the concentration 

and quality of fission products released to the 

environment following postulated accidents, 

and to control the concentration of hydrogen 

or oxygen and other substances in the 

containment atmosphere following 

postulated accidents to assure that 

containment integrity is maintained. 

Each system shall have suitable 

redundancy in components and features, and 

suitable interconnections, leak detection, 

isolation, and containment capabilities to 

assure that for onsite electric power system 

operation (assuming offsite power is not 

available) and for offsite electric power 

system operation (assuming onsite power is 

not available) its safety function can be 

accomplished, assuming a single failure, 

except for analysis of loss of coolant 

accidents involving pipe breaks larger than 

the transition break size under § 50.46a. For 

those pipe breaks only, neither a single 

failure nor the unavailability of offsite power 

need be assumed. 

* * * * * 

Criterion 44—Cooling water. A system to 

transfer heat from structures, systems, and 

components important to safety, to an 

ultimate heat sink shall be provided. The 

system safety function shall be to transfer the 

combined heat load of these structures, 

systems, and components under normal 

operating and accident conditions. 

Suitable redundancy in components and 

features, and suitable interconnections, leak 

detection, and isolation capabilities shall be 

provided to assure that for onsite electric 

power system operation (assuming offsite 

power is not available) and for offsite electric 

power system operation (assuming onsite 

power is not available) the system safety 

function can be accomplished, assuming a 

single failure, except for analysis of loss of 

coolant accidents involving pipe breaks 

larger than the transition break size under 

§ 50.46a. For those pipe breaks only, neither 

a single failure nor the unavailability of 

offsite power need be assumed. 

* * * * * 

Criterion 50—Containment design basis. 

The reactor containment structure, including 

access openings, penetrations, and the 

containment heat removal system shall be 

designed so that the containment structure 

and its internal compartments can 

accommodate, without exceeding the design 

leakage rate and with sufficient margin, the 

calculated pressure and temperature 

conditions resulting from any loss-of-coolant 

accident. This margin shall reflect 

consideration of (1) the effects of potential 

energy sources which have not been included 

in the determination of the peak conditions, 

such as energy in steam generators and as 

required by § 50.44 energy from metal-water 

and other chemical reactions that may result 

from degradation but not total failure of 

emergency core cooling functioning, (2) the 

limited experience and experimental data 

available for defining accident phenomena 

and containment responses, and (3) the 

conservatism of the calculational model and 

input parameters. 

For licensees voluntarily choosing to 

comply with § 50.46a, the structural and leak 

tight integrity of the reactor containment 

structure, including access openings, 

penetrations, and its internal compartments, 

shall be maintained for realistically 

calculated pressure and temperature 

conditions resulting from any loss of coolant 

accident larger than the transition break size. 

* * * * * 

PART 52—LICENSES, 

CERTIFICATIONS AND APPROVALS 

FOR NUCLEAR POWER PLANTS 

7. The authority citation for part 52 

continues to read as follows: 

Authority: Secs. 103, 104, 161, 182, 183, 

185, 186, 189, 68 Stat. 936, 948, 953, 954, 

955, 956, as amended, sec. 234, 83 Stat. 444, 

as amended (42 U.S.C. 2133, 2201, 2232, 

2233, 2235, 2236, 2239, 2282); secs. 201, 202, 

206, 88 Stat. 1242, 1244, 1246, as amended 

(42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 

Stat. 2750 (44 U.S.C. 3504 note); Energy 

Policy Act of 2005, Pub. L. No. 109–58, 119 

Stat. 594 (2005), secs. 147 and 149 of the 

Atomic Energy Act. 

8. In § 52.47, paragraph (a)(4) is 

revised to read as follows: 

§ 52.47 Contents of applications; technical 

information 

(a) * * * 

(4) An analysis and evaluation of the 

design and performance of structures, 

systems, and components with the 

objective of assessing the risk to public 

health and safety resulting from 

operation of the facility and including 

determination of the margins of safety 

during normal operations and transient 

conditions anticipated during the life of 

the facility, and the adequacy of 

structures, systems, and components 

provided for the prevention of accidents 

and the mitigation of the consequences 

of accidents. 

(i) Analysis and evaluation of 

emergency core cooling system (ECCS) 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents may be 

performed under the requirements of 

either § 50.46 or § 50.46a and § 50.46b of 

this chapter for designs certified after 

[EFFECTIVE DATE OF RULE] and 

demonstrated under § 50.46a(c)(2) of 

this chapter to be similar to reactor 

designs licensed before [EFFECTIVE 

DATE OF RULE], or 

(ii) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

§§ 50.46 and 50.46b of this chapter for 

designs that are not demonstrated under 

§ 50.46a(c)(2) of this chapter to be 

similar to reactor designs licensed 

before [EFFECTIVE DATE OF RULE]. 

* * * * * 

9. In § 52.79, paragraph (a)(5) is 

revised to read as follows: 

§ 52.79 Contents of applications; technical 

information in final safety analysis report. 

(a) * * * 

(5) An analysis and evaluation of the 

design and performance of structures, 

systems, and components with the 


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40052 Federal Register / Vol. 74, No. 152 / Monday, August 10, 2009 / Proposed Rules 

objective of assessing the risk to public 

health and safety resulting from 

operation of the facility and including 

determination of the margins of safety 

during normal operations and transient 

conditions anticipated during the life of 

the facility, and the adequacy of 

structures, systems, and components 

provided for the prevention of accidents 

and the mitigation of the consequences 

of accidents. 

(i) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

either § 50.46 or § 50.46a and § 50.46b of 

this chapter for facilities licensed after 

[EFFECTIVE DATE OF RULE] and 

demonstrated under § 50.46a(c)(2) of 

this chapter to be similar to reactor 

designs licensed before [EFFECTIVE 

DATE OF RULE], or 

(ii) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

§§ 50.46 and 50.46b of this chapter for 

facilities licensed after [EFFECTIVE 

DATE OF RULE] and not demonstrated 

under § 50.46a(c)(2) of this chapter to be 

similar to reactor designs licensed 

before [EFFECTIVE DATE OF RULE]. 

* * * * * 

10. In § 52.137, paragraph (a)(4) is 

revised to read as follows: 

§ 52.137 Contents of applications; 

technical information. 

(a) * * * 

(4) An analysis and evaluation of the 

design and performance of SSCs with 

the objective of assessing the risk to 

public health and safety resulting from 

operation of the facility and including 

determination of the margins of safety 

during normal operations and transient 

conditions anticipated during the life of 

the facility, and the adequacy of SSCs 

provided for the prevention of accidents 

and the mitigation of the consequences 

of accidents. 

(i) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

either § 50.46 or § 50.46a and § 50.46b of 

this chapter for designs approved after 

[EFFECTIVE DATE OF RULE] and 

demonstrated under § 50.46a(c)(2) of 

this chapter to be similar to reactor 

designs licensed before [EFFECTIVE 

DATE OF RULE], or 

(ii) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

§§ 50.46 and 50.46b of this chapter for 

designs that are not demonstrated under 

§ 50.46a(c)(2) of this chapter to be 

similar to reactor designs licensed 

before [EFFECTIVE DATE OF RULE]. 

* * * * * 

11. In § 52.157, paragraph (f)(1) is 

revised to read as follows: 

§ 52.157 Contents of applications; 

technical information in final safety analysis 

report. 

(f) * * * 

(1) An analysis and evaluation of the 

design and performance of structures, 

systems, and components with the 

objective of assessing the risk to public 

health and safety resulting from 

operation of the facility and including 

determination of the margins of safety 

during normal operations and transient 

conditions anticipated during the life of 

the facility, and the adequacy of 

structures, systems, and components 

provided for the prevention of accidents 

and the mitigation of the consequences 

of accidents. 

(i) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

either § 50.46 or § 50.46a and § 50.46b of 

this chapter for facilities licensed after 

[EFFECTIVE DATE OF RULE] and 

demonstrated under § 50.46a(c)(2) to be 

similar to reactor designs licensed 

before [EFFECTIVE DATE OF RULE], or 

(ii) Analysis and evaluation of ECCS 

cooling performance and the need for 

high-point vents following postulated 

loss-of-coolant accidents must be 

performed under the requirements of 

§§ 50.46 and 50.46b of this chapter for 

facilities licensed after [EFFECTIVE 

DATE OF RULE] and not demonstrated 

under § 50.46a(c)(2) of this chapter to be 

similar to reactor designs licensed 

before [EFFECTIVE DATE OF RULE]. 

* * * * * 

Dated at Rockville, Maryland, this 6th day 

of July 2009. 

For the Nuclear Regulatory Commission. 

Bruce S. Mallett, 

Acting Executive Director for Operations. 

[FR Doc. E9–18547 Filed 8–7–09; 8:45 am] 

BILLING CODE 7590–01–P 


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