NUCLEAR ENERGY INSTITUTE 

Anthony R. Pietrangelo 

DIRECTOR, LICENSING 

NUCLEAR GENERATION DIVISION 

November 17, 2000 

Mr. David B. Matthews 

Division of Regulatory Improvement Programs 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555 

PROJECT NUMBER: 689 

Dear Mr. Matthews: 

We are pleased that the Commission has approved Regulatory Guide 1.187 which 

endorses NEI 96-07, Revision 1, Guidelines forlO CFR 50.59 Implementation, without 

exception. The Commission action marks achievement of our shared goal to provide 

licensees with consensus guidelines for implementing 10 CFR 50.59 and thus restore 

regulatory stability in this important area. We commend the Commission and the 

NRC staff for their efforts in this regard. 

The final NEI 96-07, Revision 1, dated November 2000, is attached. The final 

guideline includes two additional sentences in Section 4.1.5 on evaluation of fire 

protection program changes in accordance with the standard license condition. Also, 

the Foreword and Section 1.4 have been clarified to reflect that when completed, 

10 CFR 72.48 guidance will become Appendix B of NEI 96-07, Revision 1. An 

electronic copy is being provided separately to Eileen McKenna. 

We intend to monitor licensee implementation of the revised 10 CFR 50.59 rule and 

guidance. Please contact Russ Bell at 202-739-8087 regarding any significant 

implementation questions that may arise. 

Sincerely, 

Anthony R. Pietrangelo 

Enclosure 

c: Cindy Carpenter, NRC/NRR 

Steve West, NRC/NRR 

Eileen McKenna, NRC/NRR 

Document Control Desk 

S-•FT "J.,% cj1E 4:,, • S• - " C 2C'" @ 37C FHC 202 73; 8000 ý ,X 202 785 40'Q www.nei.org
NEI 96-07, Revision I

Nuclear Energy Institute 

GUIDELINES FOR 10 CFR 50.59 

IMPLEMENTATION

November 2000
ACKNOWLEDGMENTS

In 1996, NSAC-125, Guidelines for 10 CFR 50.59 Safety Evaluations, was 

transformed into NEI 96-07 with minor changes to address specific NRC concerns. 

Much of this long standing industry guidance continues to underlie the revised 

guidance presented in this document. We appreciate EPRI allowing NEI to use 

NSAC-125 in this manner and we recognize the efforts of the individuals who 

contributed to the development of NSAC- 125. 

The revised guidance in this document was developed with the invaluable assistance 

of the 10 CFR 50.59 Task Force and the Regulatory Process Working Group. 

NOTICE 

Neither NEI, nor any of its employees, members, supporting organizations, 

contractors, or consultants make any warranty, expressed or implied, or assume 

any legal responsibility for the accuracy or completeness of, or assume any liability 

for damages resulting from any use of, any information apparatus, methods, or 

process disclosed in this report or that such may not infringe privately owned 

rights.
NEI 96-07, Revision 1 

November 2000 

FOREWORD 

In 1999, the NRC revised its regulation controlling changes, tests and experiments 

performed by nuclear plant licensees-the first changes to 10 CFR 50.59 in more 

than 30 years. The changes were prompted by the need to resolve differences in 

interpretation of the rule's requirements by the industry and the NRC that came in 

clear focus in 1996. These differences existed despite general recognition that 

licensee implementation of 10 CFR 50.59 has been effective in controlling activities 

affecting plant design and operation. The rule changes had two principal objectives, 

both aimed at restoring much-needed regulatory stability to this extensively used 

regulation: 

. Establish clear definitions to promote common understanding of the rule's 

requirements 

. Clarify the criteria for determining when changes, tests and experiments 

require prior NRC approval 

While effective at controlling changes, 10 CFR 50.59 was, at the same time, viewed 

as overly restrictive of licensee changes and unduly burdensome. License 

amendment requests were prepared, submitted and reviewed by the NRC for many 

changes having little or no impact on the plant design or operation. Indeed, some 

beneficial changes were withdrawn by licensees upon determination that the change 

would have to go through the burdensome license amendment process. 

Moreover, substantial resources were expended each year by licensees to process and 

submit to NRC lengthy evaluations for numerous insignificant changes. The 

changes approved by the Commission in 1999 made 10 CFR 50.59 more focused and 

efficient by: 

"* Providing greater flexibility to licensees, primarily by allowing changes 

that have minimal safety impact to be made without prior NRC approval 

"* Clarifying the threshold for "screening out" changes that do not require 

full evaluation under 10 CFR 50.59, primarily by adoption of key 

definitions. 

These changes will conserve both licensee and NRC resources while continuing to 

ensure that significant changes are thoroughly evaluated and approved by the NRC 

as appropriate. 

This document provides guidance for implementing the revised rule. While it 

contains new guidance corresponding to new and revised rule criteria, overall, the 

document reflects a refinement of longstanding industry practice, not a radical new

i
NEI 96-07, Revision 1 

November 2000 

approach. The basic philosophy behind 10 CFR 50.59 implementation and a 

substantial amount of guidance reflected in this document can be traced to 

EPRI/NSAC-125-the original industry guidance document in this area-issued in 

1989. 

Other past guidance related to 10 CFR 50.59, including NRC generic 

communications, was also reviewed and reflected in this document as appropriate. 

The intent is to provide comprehensive guidance that is consistent with the 1999 

changes to 10 CFR 50.59. 

In parallel with the rulemaking to amend 10 CFR 50.59, the NRC made conforming 

changes to the analogous provision in 10 CFR Part 72 for control of changes, tests 

and experiments involving independent fuel storage facilities. The intent of 

conforming 10 CFR 72.48 to the terms of 10 CFR 50.59 was to provide for consistent 

implementation of these two analogous regulations. Accordingly, the guidance 

herein on implementing 10 CFR 50.59 has been used as the basis for guidance on 

implementing 10 CFR 72.48. When completed, 10 CFR 72.48 guidance will become 

Appendix B to NEI 96-07, Revision 1.

ii
NEI 96-07, Revision 1 

November 2000 

TABLE OF CONTENTS 

FOREWORD_-i 

1 INTRODUCTION .1 

2 DEFENSE IN DEPTH DESIGN PHILOSOPHY AND 10 CFR 50.59 .7 

3 DEFINITIONS AND APPLICABILITY OF TERMS . 9 

4 IMPLEMENTATION GUIDANCE _... . .. 23 

4.1 APPLICABILITY ........................................................................................... 23 

4.2 SCREENING ................................................................................................. 29 

4.3 EVALUATION PROCESS .............................................................................. 41 

4.4 APPLYING 10 CFR 50.59 To COMPENSATORY ACTIONS To 

ADDRESS NONCONFORMING OR DEGRADED CONDITIONS ....................... 70 

4.5 DISPOSITION OF 10 CFR 50.59 EVALUATIONS ........................................ 72 

5 DOCUMENTATION AND REPORTING _73 

APPENDICES 

A. 10 CFR 50.59-CHANGES, TESTS, AND EXPERIMENTS. A

B. GUIDELINES FOR 10 CFR 12A8 IMPLEMENTATION [FUTURE) . B-1

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NEI 96-07, Revision 1 

November 2000 

1 INTRODUCTION 

1.1 PURPOSE 

10 CFR 50.59 establishes the conditions under which licensees may make 

changes to the facility or procedures and conduct tests or experiments 

without prior NRC approval. Proposed changes, tests and experiments 

(hereafter referred to collectively as activities) that satisfy the definitions and 

one or more of the criteria in the rule must be reviewed and approved by the 

NRC before implementation. Thus 10 CFR 50.59 provides a threshold for 

regulatory review-not the final determination of safety-for proposed 

activities. 

The purpose of this document is to provide guidance for developing effective 

and consistent 10 CFR 50.59 implementation processes. 

1.2 RELATIONSHIP OF 10 CFR 50.59 TO OTHER REGULATORY REQUIREMENTS 

AND CONTROLS 

As the process for controlling a range of activities affecting equipment and 

procedures at a nuclear power plant, implementation of 10 CFR 50.59 

interfaces with many other regulatory requirements and controls. To 

optimize the use of 10 CFR 50.59, the rule and this guidance should be 

understood in the context of the proper relationship with these other 

regulatory processes. These relationships are described below: 

1.2.1 Relationship of 10 CFR 50.59 to Other Processes That Control 

Licensing Basis Activities 

10 CFR 50.59 focuses on the effects of proposed activities on the safety 

analyses that are contained in the updated FSAR (UFSAR) and are a 

cornerstone of each plant's licensing basis. In addition to 10 CFR 50.59 

control of changes affecting the safety analyses, there are several other 

complementary processes for controlling activities that affect other aspects of 

the licensing basis, including: 

"* Amendments to the operating license (including the technical 

specifications) are sought and obtained under 10 CFR 50.90. 

"* Where changes to the facility or procedures are controlled by more 

specific regulations (e.g., quality assurance, security and emergency 

preparedness program changes controlled under 10 CFR 50.54(a),

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NEI 96-07, Revision 1 

November 2000 

(p) and (q), respectively; Off-site Dose Calculation Manual changes 

controlled by technical specifications), 10 CFR 50.59 states that the 

more specific regulation applies. 

"* Changes that require an exemption from a regulation are processed 

in accordance with 10 CFR 50.12. 

"* Guidance for controlling changes to licensee commitments is 

provided by NEI 99-04, Guideline for Managing NRC Commitment 

Changes. 

"* Where a licensee possesses a license condition that specifically 

permits changes to the NRC-approved fire protection program (i.e., 

has received the standard fire protection license condition contained 

in Generic Letter 86-10), subsequent changes to the fire protection 

program would be controlled under the license condition and not 10 

CFR 50.59. 

"* Maintenance activities, including associated temporary changes, are 

subject to the technical specifications and are assessed and managed 

in accordance with the Maintenance Rule, 10 CFR 50.65; screening 

and evaluation under 10 CFR 50.59 are not required. 

Together with 10 CFR 50.59, these processes form a framework of 

complementary regulatory controls over the licensing basis. To optimize the 

effectiveness of these controls and minimize duplication and undue burden, it 

is important to understand the scope of each process within the regulatory 

framework. This guideline discusses the scope of 10 CFR 50.59 in relation to 

other processes, including circumstances under which different processes, 

e.g., 10 CFR 50.59 and 10 CFR 50.90, should be applied to different aspects 

of an activity. 

In addition to controlling changes to the facility and procedures described in 

the UFSAR under 10 CFR 50.59 as required by the rule, some licensees also 

control changes to other licensing basis information using the 10 CFR 50.59 

process. This may be in accordance with a requirement of the license or 

commitment to the NRC. The technical specifications bases are an example 

of documentation that may be outside the UFSAR but that is controlled via 

10 CFR 50.59 by many licensees.

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NEI 96-07, Revision 1 

November 2000 

1.2.2 Relationship of 10 CFR 50.59 to 10 CFR Part 50, Appendix B 

Prior to the operating license, 10 CFR Part 50, Appendix B, assures that the 

facility design and construction meet applicable requirements, codes and 

standards in accordance with the safety classification of systems, structures 

and components (SSCs). Appendix B design control provisions ensure that 

all changes continue to meet applicable design and quality requirements. 

The design and licensing bases evolve in accordance with Appendix B 

requirements up to the time that an operating license is received, and 10 

CFR 50.59 is not applicable until after that time. Both Appendix B and 10 

CFR 50.59 apply following receipt of an operating license. 

Appendix B also addresses corrective action. The application of 10 CFR 

50.59 to compensatory actions that address degraded and nonconforming 

conditions is described in Section 4.4. 

1.2.3 Relationship of 10 CFR 50.59 to the UFSAR 

10 CFR 50.59 is the process that identifies when a license amendment is 

required prior to implementing changes to the facility or procedures 

described in the UFSAR or tests and experiments not described in the 

UFSAR. As such, it is important that the UFSAR be properly maintained 

and updated in accordance with 10 CFR 50.71(e). Guidance for updating 

UFSARs to reflect activities implemented under 10 CFR 50.59 is provided by 

Regulatory Guide 1.18 1, which endorses NEI 98-03, Revision 1. 

1.2.4 Relationship of 10 CFR 50.59 to 10 CFR 50.2 Design Bases 

10 CFR 50.59 controls changes to both 10 CFR 50.2 design bases and 

supporting design information contained in the UFSAR. In support of 10 

CFR 50.59 implementation, Section 4.3.7 of this guideline defines the design 

basis limits for fission product barriers that are subject to control under 10 

CFR 50.59(c)(2)(vii), and Section 4.3.8 provides guidance on the scope of 

methods of evaluation used in establishing design bases or in the safety 

analyses that are subject to control under 10 CFR 50.59(c)(2)(viii). 

Additional guidance for identifying 10 CFR 50.2 design bases is provided in 

NEI 97-04, Appendix B. 

As discussed in Section 3.3, "design bases functions" (defined in NEI 97-04, 

Appendix B) are a subset of "design functions" for purposes of 10 CFR 50.59 

screening.

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NEI 96-07, Revision 1 

November 2000 

1.3 10 CFR 50.59 PROCESS SUMMARY: 

After determining that a proposed activity is safe and effective through 

appropriate engineering and technical evaluations, the 10 CFR 50.59 process 

is applied to determine if a license amendment is required prior to 

implementation. This process involves the following basic steps as depicted 

in Figure 1: 

z Applicability and Screening: Determine if a 10 CFR 50.59 evaluation 

is required. 

m Evaluation: Apply the eight evaluation criteria of 10 CFR 50.59(c)(2) to 

determine if a license amendment must be obtained from the NRC. 

* Documentation and reporting: Document and report to the NRC 

activities implemented under 10 CFR 50.59. 

Later sections of this document discuss key definitions, provide guidance for 

determining applicability, screening, and performing 10 CFR 50.59 

evaluations, and present examples to illustrate the application of the process.

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NEI 96-07, Revision 1 

November 2000 

Proposed Activity 

Evaluate Activity to Ensure It Is Safe and Effective 

Does the Activity 

Require a Change Yes Request/Obtain a License Amendment 

Sto the Tech Under 10 CFR 50.90 

Specs? 

Screnng 

Section No 

4.1 

Is the Activit AplyOte,10Reuain 50.54 ~escFR eg. 

Controlled by Yes 10 CFR 50.65 

AnotherP 10 CFR 50.55a Regulation ors • 10 CFR 50.46 

hageProcess• 10 CFR 50.12 

10 CFR Part 20 

See Screening: Seto Is 10 CFR 50.59 No Imlmn heAtvt Secti~~n -• Evaluation IpentheAivy 

4.2 Reurd 

See 

Does theproActivityNc, Yes Request/Obtain a License Amendment 

Section Require Prior Under 10 CFR 50.90 

4.3 Annrnvwl?

Figure 1 

10 CFR 50.59 Process

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NEI 96-07, Revision 1 

November 2000 

1.4 APPLICABILITY TO 10 CFR 72.48 

Concurrent with the rulemaking to amend 10 CFR 50.59, the NRC made 

conforming changes to the analogous provisions in 10 CFR 72.48 controlling 

licensee changes, tests and experiments to independent spent fuel storage 

installations (ISFSIs). The provisions of 10 CFR 72.48 were also extended to 

holders of Part 72 Certificates of Compliance. As a result, 10 CFR 72.48 

establishes criteria identical to those in 10 CFR 50.59 under which both an 

ISFSI license holder and a certificate holder may make changes to the facility 

or cask design, make changes to procedures and conduct tests or experiments 

without prior NRC approval. 

The intent of conforming 10 CFR 72.48 to the terms of 10 CFR 50.59 was to 

provide for consistent implementation of these two analogous regulations. 

Consistent with this intent, the guidance herein on implementing 10 CFR 50.59 

has been used as the basis for guidance on implementing 10 CFR 72.48. When 

completed, 10 CFR 72.48 guidance will become Appendix B to NEI 96-07, 

Revision 1. 

1.5 CONTENT OF THIS GUIDANCE DOCUMENT 

The NRC has established requirements for nuclear plant systems, structures 

and components to provide reasonable assurance of adequate protection of 

the public health and safety. Many of these requirements, and descriptions 

of how they are met, are documented in the updated FSAR (UFSAR). 10 

CFR 50.59 allows a licensee to make changes in the facility or procedures as 

described in the UFSAR, and to conduct tests or experiments not described in 

the UFSAR, unless the changes require a change in the technical 

specifications or otherwise require prior NRC approval. In order to perform 

10 CFR 50.59 screenings and evaluations, an understanding of the design 

and licensing basis of the plant and of the specific requirements of the 

regulations is necessary. Individuals performing 10 CFR 50.59 screenings 

and evaluations should also understand the rule and concepts discussed in 

this guidance document. 

In Section 2, the relationship between the design criteria established in 

10 CFR 50, Appendix A, and 10 CFR 50.59 is discussed as background for 

applying the rule. 

Section 3 presents definitions and discussion of key terms used in 10 CFR 

50.59 and this guideline. 

Section 4 discusses the application of the definitions and criteria presented in 

10 CFR 50.59 to the process of changing the plant or procedures and the

6
NEI 96-07, Revision 1 

November 2000 

conduct of tests or experiments. This section includes guidance on the 

applicability requirements for the rule, the screening process for determining 

when a 10 CFR 50.59 evaluation must be performed and the eight evaluation 

criteria for determining if prior NRC approval is required. Examples are 

provided to reinforce the guidance. Guidance is also provided on addressing 

degraded and nonconforming conditions and on dispositioning 10 CFR 50.59 

evaluations. 

Section 5 provides guidance on documenting 10 CFR 50.59 evaluations and 

reporting to NRC. 

Appendix A provides the text of 10 CFR 50.59 as published in the Federal 

Register on October 4, 1999. When completed, Appendix B will provide 

guidance and examples illustrating the application of this guidance to 

changes involving independent spent fuel storage installations and spent fuel 

storage cask designs, per 10 CFR 72.48. 

2.0 DEFENSE IN DEPTH DESIGN PHILOSOPHY AND 10 CFR 50.59 

One objective of Title 10 of the Code of Federal Regulations is to establish 

requirements directed toward protecting the health and safety of the public 

from the uncontrolled release of radioactivity. At the design stage, protection 

of public health and safety is ensured through the design of physical barriers 

to guard against the uncontrolled release of radioactivity. Other sources of 

radioactivity including radwaste systems are included. The defense-in-depth 

philosophy includes reliable design provisions to safely terminate accidents 

and provisions to mitigate the consequences of accidents. The three physical 

barriers that provide defense-in-depth are: 

m Fuel Clad 

m Reactor Coolant System Boundary 

* Containment Boundary 

These barriers perform a health and safety protection function. They are 

designed to reliably fulfill their operational function by meeting all criteria 

and standards applicable to mechanical components, pressure components 

and civil structures. These barriers are protected extensively by inherent 

safety features and through the implementation of engineered safety 

features. The public health and safety protection functions are analytically 

demonstrated and documented in the UFSAR. Analyses summarized in the 

UFSAR demonstrate that under the assumed accident conditions, the 

consequences of accidents challenging the integrity of the barriers will not 

exceed limits based on the criteria established in GDC 19 or the guidelines

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NEI 96-07, Revision 1 

November 2000 

established in 10 CFR 100. Thus, the UFSAR analyses provide the final 

verification of the nuclear safety design phase by documenting plant 

performance in terms of public protection from uncontrolled releases of 

radiation. 10 CFR 50.59 addresses this aspect of design by requiring prior 

NRC approval of proposed activities that, although safe, require a technical 

specification change or meet specific threshold criteria for NRC review. 

This protection philosophy pervades the UFSAR accident analyses and Title 

10 of the CFR. To understand and apply 10 CFR 50.59, it is necessary to 

understand this perspective of maintaining the integrity of the physical 

barriers designed to contain radioactivity. This is because: 

m UFSAR accidents and malfunctions are analyzed in terms of their 

effect on the physical barriers. There is a relationship between barrier 

integrity and dose. 

* The principal "consequences" that the physical barriers are designed to 

preclude is the uncontrolled release of radioactivity. Thus for purposes 

of 10 CFR 50.59, the term "consequences" means dose. 

For many licensees, ANSI standards define categories of accidents or 

malfunctions. For each category a probability (frequency) and a 

corresponding acceptable consequence is given in terms of barrier loss and 

radioactivity release. Consequences resulting from accidents and 

malfunctions are analyzed and documented in the UFSAR and are evaluated 

against dose acceptance limits that vary depending on the event frequency. 

The design effort and the operational controls necessary to ensure the 

required performance of the physical barriers during anticipated operational 

occurrences and postulated accidents are extensive. Because 10 CFR 50.59 

provides a mechanism for determining if NRC approval is needed for 

activities affecting plant design and operation, it is helpful to review briefly 

the requirements and the objectives imposed by the CFR on plant 

construction and operation. The review will define more clearly the extent of 

applicability of 10 CFR 50.59. 

Appendix A to 10 CFR Part 50 provides General Design Criteria for most 

nuclear power plants (for pre-Appendix A plants the criteria are in the 

UFSAR). Section II of Appendix A includes criteria for protection by multiple 

fission product barriers. The criteria establish requirements for inherent 

protection, instrumentation and control, reactor coolant pressure boundary 

and reactor coolant system design, containment design, control rooms, 

electric power systems, and related inspection and testing. All of these

8
NEI 96-07, Revision 1 

November 2000 

requirements concentrate on protecting fission product barriers either 

through inherent or mitigative means. 

Section III of Appendix A establishes extensive requirements on reactor 

protection and reactivity control systems, the objectives again being the 

protection of fission product barriers. With similar intent, Sections IV, V and 

VI provide extensive design, inspection, testing and operational requirements 

for the quality of the reactor coolant pressure boundary, and fluid systems in 

general, reactor containment, and fuel and radioactivity control. These 

requirements ensure inherent and engineered protection of the fission 

product barriers. Introductory statements of Appendix A address the need 

for consideration of a single failure criterion and redundancy, diversity and 

separation of mitigation and protection systems. Section I of Appendix A 

imposes requirements on the quality of implemented protection and the 

conditions under which these systems must function without loss of 

capability to perform their safety functions. These conditions include natural 

phenomena, fire, operational and accident generated environmental 

conditions. 

The implementation of this design philosophy requires extensive accident 

analyses to define the correct relationship among nominal operating 

conditions, limiting conditions for operations and limiting safety systems 

settings to prevent safety limits from being exceeded. The UFSAR presents 

the set of limiting analyses required by NRC. The limiting analyses are used 

to confirm the systems and equipment design, to identify critical setpoints and 

operator actions, and to support the establishment of technical specifications. 

Therefore, the results of the UFSAR accident analyses reflect performance of 

equipment under the conditions specified by NRC regulations or requirements. 

Changes to plant design and operation and conduct of new tests and 

experiments have the potential to affect the probability and consequences of 

accidents, to create new accidents and to impact the integrity of fission product 

barriers. Therefore, these activities are subject to 10 CFR 50.59. 

3.0 DEFINITIONS AND APPLICABILITY OF TERMS 

The following definitions and terms are discussed in this section: 

3.1 10 CFR 50.59 Evaluation 

3.2 Accident Previously Evaluated in the FSAR (as updated) 

3.3 Change

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NEI 96-07, Revision 1 

November 2000 

3.4 Departure from a Method of Evaluation Described in the FSAR (as 

updated) 

3.5 Design Bases (Design Basis) 

3.6 Facility as Described in the FSAR (as updated) 

3.7 Final Safety Analysis Report (as updated) 

3.8 Input Parameters 

3.9 Malfunction of an SSC Important to Safety 

3.10 Methods of Evaluation 

3.11 Procedures as Described in the FSAR (as updated) 

3.12 Safety Analyses 

3.13 Screening 

3.14 Tests or Experiments Not Described in the FSAR (as updated) 

3.1 10 CFR 50.59 EVALUATION 

Definition: 

A 10 CFR 50.59 evaluation is the documented evaluation against the eight 

criteria in 10 CFR 50.59(c)(2) to determine if a proposed change, test or 

experiment requires prior NRC approval via license amendment under 10 

CFR 50.90. 

Discussion 

It is important to establish common terminology for use relative to the 10 

CFR 50.59 process. The definitions of 10 CFR 50.59 Evaluation and 

Screening are intended to clearly distinguish between the process and 

documentation of licensee screenings and the further evaluation that may be 

required of proposed activities against the eight criteria in 10 CFR 

50.59(c)(2). Section 4.3 provides guidance for performing 10 CFR 50.59 

evaluations. The screening process is discussed in Section 4.2. 

The phrase "change made under 10 CFR 50.59" (or equivalent) refers to 

changes subject to the rule (see Section 4.1) that either screened out of the 10 

CFR 50.59 process or did not require prior NRC approval based on the

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NEI 96-07, Revision 1 

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results of a 10 CFR 50.59 evaluation. Similarly, the phrases "10 CFR 50.59 

applies [to an activity]" or "[an activity] is subject to 10 CFR 50.59" mean 

that screening and, if necessary, evaluation are required for the activity. The 

"10 CFR 50.59 process" includes screening, evaluation, documentation and 

reporting to NRC of activities subject to the rule. 

3.2 ACCIDENT PREVIOUSLY EVALUATED IN THE FSAR (AS UPDATED) 

Definition: 

Accident previously evaluated in the FSAR (as updated) means a design 

basis accident or event described in the UFSAR including accidents, such 

as those typically analyzed in Chapters 6 and 15 of the UFSAR, and 

transients and events the facility is required to withstand such as floods, 

fires, earthquakes, other external hazards, anticipated transients without 

scram (ATWS) and station blackout (SBO). 

Discussion: 

The term "accidents" refers to the anticipated (or abnormal) operational 

transients and postulated design basis accidents that are analyzed to 

demonstrate that the facility can be operated without undue risk to the 

health and safety of the public. For purposes of 10 CFR 50.59, the term 

"accidents" encompasses other events for which the plant is required to cope 

and that are described in the UFSAR (e.g., turbine missiles, fire, earthquakes 

and flooding). Note that, although fire is an event for which a plant is 

required to cope and is described in the UFSAR (by reference to the Fire 

Hazards Analysis for some licensees), changes to the fire protection program 

are, for most licensees, governed by requirements other than 10 CFR 50.59, 

as discussed in Section 4.1.5. 

Accidents also include new transients or postulated events added to the 

licensing basis based on new NRC requirements and reflected in the UFSAR 

pursuant to 10 CFR 50.71(e), e.g., ATWS and SBO. 

3.3 CHANGE 

Definition: 

Change means a modification or addition to, or removal from, the facility or 

procedures that affects: (1) a design function, (2) method of performing or 

controlling the function, or (3) an evaluation that demonstrates that intended 

functions will be accomplished.

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NEI 96-07, Revision 1 

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Discussion: 

Additions and removals to the facility or procedures can adversely impact the 

performance of SSCs and the bases for the acceptability of their design and 

operation. Thus the definition of change includes modifications of an 

existing provision (e.g., SSC design requirement, analysis method or 

parameter), additions or removals (physical removals, abandonment or 

nonreliance on a system to meet a requirement) to the facility or procedures. 

The definitions of "change...," "facility..." (see Section 3.6), and 

"procedures..." (see Section 3.11) make clear that 10 CFR 50.59 applies to 

changes to underlying analytical bases for the facility design and operation 

as well as for changes to SSCs and procedures. Thus 10 CFR 50.59 should be 

applied to a change being made to an evaluation for demonstrating adequacy 

of the facility even if no physical change to the facility is involved. Further 

discussion of the terms in this definition is provided as follows: 

Design functions are UFSAR-described design bases functions and other SSC 

functions described in the UFSAR that support or impact design bases 

functions. Implicitly included within the meaning of design function are 

the conditions under which intended functions are required to be 

performed, such as equipment response times, process conditions, 

equipment qualification and single failure. 

Design bases functions are functions performed by systems, structures and 

components (SSCs) that are (1) required by, or otherwise necessary to comply 

with, regulations, license conditions, orders or technical specifications, or (2) 

credited in licensee safety analyses to meet NRC requirements., 

UFSAR description of design functions may identify what SSCs are 

intended to do, when and how design functions are to be performed, and 

under what conditions. Design functions may be performed by safety

related SSCs or nonsafety-related SSCs and include functions that, if not 

performed, would initiate a transient or accident that the plant is 

required to withstand. 

As used above, "credited in the safety analyses" means that, if the SSC 

were not to perform its design bases function in the manner described, the 

assumed initial conditions, mitigative actions or other information in the 

analyses would no longer be within the range evaluated (i.e., the analysis 

results would be called into question). The phrase "support or impact 

design bases functions" refers both to those SSCs needed to support design 

1 Definition of design bases function from revised Appendix B to NEI 97-04 (endorsed by Regulatory 

Guide 1.186).

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NEI 96-07, Revision 1 

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bases functions (cooling, power, environmental control, etc.) and to SSCs 

whose operation or malfunction could adversely affect the performance of 

design bases functions (for instance, control systems and physical 

arrangements). Thus, both safety-related and nonsafety-related SSCs may 

perform design functions. 

Method of performing or controlling a function means how a design function 

is accomplished as credited in the safety analyses, including specific 

operator actions, procedural step or sequence, or whether a specific 

function is to be initiated by manual versus automatic means. For 

example, substituting a manual actuation for automatic would constitute 

a change to the method of performing or controlling the function. 

Evaluation that demonstrates that intended functions will be accomplished 

means the method(s) used to perform the evaluation (as discussed in 

Section 3.10). Example: a thermodynamic calculation that demonstrates 

the emergency core cooling system has sufficient heat removal capacity for 

responding to a postulated accident. 

Temporary Changes 

Temporary changes to the facility or procedures, such as jumpering 

terminals, lifting leads, placing temporary lead shielding on pipes and 

equipment, removal of barriers and use of temporary blocks, bypasses, 

scaffolding and supports, are made to facilitate a range of plant activities and 

are subject to 10 CFR 50.59 as follows: 

* 10 CFR 50.59 should be applied to temporary changes proposed as 

compensatory actions to address degraded or nonconforming conditions as 

discussed in Section 4.4. 

* Other temporary changes to the facility or procedures that are not 

associated with maintenance are subject to 10 CFR 50.59 in the same 

manner as permanent changes, to determine if prior NRC approval is 

required. Screening and, as necessary, evaluation of such temporary 

changes may be considered as part of the screening/evaluation of the 

proposed permanent change. 

Risk impacts of temporary changes associated with maintenance activities 

(i.e., temporary alterations) should be assessed and managed in accordance 

with 10 CFR 50.65(a)(4) and associated guidance, as discussed in Section 

4.1.2. Applying 10 CFR 50.59 to such activities is not required provided that 

temporary alterations are not in effect longer than 90 days at power, and

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affected SSCs are restored to their normal, as-designed condition at the 

conclusion of the maintenance activity. 

3.4 DEPARTURE FROM A METHOD OF EVALUATION DESCRIBED IN THE FSAR (As 

UPDATED) 

Definition: 

Departure from a method of evaluation described in the FSAR (as updated) 

means (i) changing any of the elements of the method described in the FSAR 

(as updated) unless the results of the analysis are conservative or essentially 

the same; or (ii) changing from a method described in the FSAR to another 

method unless that method has been approved by NRC for the intended 

application. 

Discussion: 

The 10 CFR 50.59 definition of "departure ..." provides licensees with 

flexibility to make changes in methods of evaluation that are "conservative" 

or that are not important with respect to demonstrating that SSCs can 

perform their intended design functions. See also the definition and 

discussion of "methods of evaluation" in Section 3.10. Guidance for 

evaluating changes in methods of evaluation under criterion 10 CFR 

50.59(c)(2)(viii) is provided in Section 4.3.8. 

Conservative vs. Nonconservative Evaluation Results 

Gaining margin by revising an element of a method of evaluation is 

considered to be a nonconservative change and thus a departure from a 

method of evaluation for purposes of 10 CFR 50.59. Such departures require 

prior NRC approval of the revised method. In other words, analytical results 

obtained by changing any element of a method are "conservative" relative to 

the previous results, if they are closer to design bases limits or safety 

analyses limits (e.g., applicable acceptance guidelines). For example, a 

change in an element of a method of evaluation that changes the result of a 

containment peak pressure analysis from 45 psig to 48 psig (with design 

basis limit of 50 psig) would be considered a conservative change for purposes 

of 10 CFR 50.59(c)(2)(viii). This is because results closer to limiting values 

are considered conservative in the sense that the new analysis result 

provides less margin to applicable limits for making future physical or 

procedure changes without a license amendment. 

If use of a modified method of evaluation resulted in a change in calculated 

containment peak pressure from 45 psig to 40 psig, this would be

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nonconservative. This is because the change would result in more margin 

being available (to the design basis limit of 50 psig) for a licensee to make 

more significant future changes to the physical plant or procedures. 

"Essentially the Same" 

Licensees may change one or more elements of a method of evaluation such 

that results move in the nonconservative direction without prior NRC 

approval, provided the results are "essentially the same" as the previous 

result. Results are "essentially the same" if they are within the margin of 

error for the type of analysis being performed. Variation in results due to 

routine analysis sensitivities or calculational differences (e.g., rounding 

errors and use of different computational platforms) would typically be 

within the analysis margin of error and thus considered "essentially the 

same." 

"Approved by the NRC for the Intended Application" 

Rather than make a minor change to an existing method of evaluation, a 

licensee may also adopt completely new methodology without prior NRC 

approval provided the new method is approved by the NRC for the intended 

application. A new method is "approved by the NRC for the intended 

application" if it is approved for the type of analysis being conducted and the 

licensee satisfies applicable terms and conditions for its use. Specific 

guidance for making this determination is provided in Section 4.3.8.2. 

3.5 DESIGN BASES (DESIGN BASIS) 

Definition: 

(10 CFR 50.2) Design bases means that information which identifies the 

specific functions to be performed by a structure, system, or component of a 

facility and the specific values or ranges of values chosen for controlling 

parameters as reference bounds for design. These values may be (1) 

restraints derived from generally accepted "state-of-the-art" practices for 

achieving functional goals or (2) requirements derived from analysis (based 

on calculations and/or experiments) of the effects of a postulated accident for 

which a structure, system, or component must meet its functional goals. 

Discussion 

Guidance and examples for identifying 10 CFR 50.2 design bases are 

provided in Appendix B of NEI 97-04, Design Bases Program Guidelines, 

Revision 1, [Month] 2000.

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3.6 FACILITY AS DESCRIBED IN THE FSAR (As UPDATED) 

Definition: 

Facility as described in the final safety analysis report (as updated) means: 

m The structures, systems and components (SSC) that are described in the 

final safety analysis report (FSAR) (as updated), 

* The design and performance requirements for such SSCs described in the 

FSAR (as updated), and 

* The evaluations or methods of evaluation included in the FSAR (as 

updated) for such SSCs, which demonstrate that their intended 

function(s) will be accomplished. 

Discussion: 

The scope of information that is the focus of 10 CFR 50.59 is the information 

presented in the original FSAR to satisfy the requirements of 10 CFR 

50.34(b), as updated per the requirements of 10 CFR 50.71(e) and as 

supplemented pursuant to 10 CFR 54.2 1(d). The definition of "facility as 

described in the FSAR (as updated)" follows from the requirement of 10 CFR 

50.34(b) that the FSAR (and by extension, the UJFSAR) contains "a 

description and analysis of the SSCs of the facility, with emphasis upon 

performance requirements, the bases, with technical justification therefore, 

upon which such requirements have been established, and the evaluations 

required to show that safety functions will be accomplished." 

10 CFR 50.59 screening of facility changes is discussed in Section 4.2.1.1. 

3.7 FINAL SAFETY ANALYSIS REPORT (AS UPDATED) 

Definition: 

Final Safety Analysis Report (as updated) means the Final Safety Analysis 

Report (or Final Hazards Summary Report) submitted in accordance with 10 

CFR 50.34, as amended and supplemented, and as updated per the 

requirements of 10 CFR 50.71(e) or 10 CFR 50.71(f), as applicable. 

Discussion: 

As used throughout this guidance document, U7FSAR is synonymous with 

"FSAR (as updated)." The scope of the UFSAR includes its text, tables, 

diagrams, etc., as well as supplemental information explicitly incorporated by 

reference. References that are merely listed in the UFSAR and documents

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that are not explicitly incorporated by reference are not considered part of 

the UFSAR and therefore are not subject to control under 10 CFR 50.59. 

Per 10 CFR 50.59(c)(4), licensees are not required to apply 10 CFR 50.59 to 

UFSAR information that is subject to other specific change control 

regulations. For example, licensee quality assurance programs, emergency 

plans and security plans are controlled by 10 CFR 50.54(a), (p) and (q), 

respectively. 

Per 10 CFR 50.59(c)(3), the 'FSAR (as updated)," for purposes of 10 CFR 

50.59, also includes UFSAR update pages approved by the licensee for 

incorporation in the UFSAR since the last required update was submitted 

per 10 CFR 50.71(e). The intent of this requirement is to ensure that 

decisions about proposed activities are made with the most complete and 

accurate information available. Pending UFSAR revisions may be relevant 

to a future activity that involves that part of the UFSAR. Therefore, pending 

UFSAR revisions to reflect completed activities that have received final 

approval for incorporation in the next required update should be considered 

as part of the UFSAR for purposes of 10 CFR 50.59 screenings and 

evaluations, as appropriate. Appropriate configuration management 

mechanisms should be in place to identify and assess interactions between 

concurrent changes affecting the same SSCs or the same portion of the 

UFSAR. 

Guidance on the required content of UFSAR updates is provided in 

Regulatory Guide 1.181 and NEI 98-03, Revision 1, Guidelines for Updating 

FSARs, June 1999. 

3.8 INPUT PARAMETERS 

Definition: 

Input parameters are those values derived directly from the physical 

characteristics of SSC or processes in the plant, including flow rates, 

temperatures, pressures, dimensions or measurements (e.g., volume, weight, 

size, etc.), and system response times. 

Discussion: 

The principal intent of this definition is to distinguish methods of evaluation 

from evaluation input parameters. Changes to methods of evaluation 

described in the UFSAR (see Section 3.10) are evaluated under criterion 10 

CFR 50.59(c)(2)(viii), whereas changes to input parameters described in the

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FSAR are considered changes to the facility that would be evaluated under 

the other seven criteria of 10 CFR 50.59(c)(2), but not criterion (c)(2)(viii). 

If a methodology permits the licensee to establish the value of an input 

parameter on the basis of plant-specific considerations, then that value is an 

input to the methodology, not part of the methodology. On the other hand, 

an input parameter is considered to be an element of the methodology if: 

* The method of evaluation includes a methodology describing how to 

select the value of an input parameter to yield adequately conservative 

results. However, if a licensee opts to use a value more conservative 

than that required by the selection method, reduction in that 

conservatism should be evaluated as an input parameter change, not a 

change in methodology. 

* The development or approval of a methodology was predicated on the 

degree of conservatism in a particular input parameter or set of input 

parameters. In other words, if certain elements of a methodology or 

model were accepted on the basis of the conservatism of a selected 

input value, then that input value is considered an element of the 

methodology. 

Examples illustrating the treatment of input parameters are provided in 

Section 4.2.1.3. 

Section 4.3.8 provides guidance and examples to describe the specific 

elements of evaluation methodology that would require evaluation under 10 

CFR 50.59(c)(2)(viii) and to clearly distinguish these from specific types of 

input parameters that are controlled by the other seven criteria of 10 CFR 

50.59(c)(2). 

3.9 MALFUNCTION OF AN SSC IMPORTANT TO SAFETY 

Definition: 

Malfunction of SSCs important to safety means the failure of SSCs to 

perform their intended design functions described in the UFSAR (whether or 

not classified as safety-related in accordance with 10 CFR 50, Appendix B). 

Discussion: 

Guidance and examples for applying this definition are provided in Section 

4.3.

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3.10 METHODS OF EVALUATION 

Definition: 

Methods of evaluation means the calculational framework used for 

evaluating behavior or response of the facility or an SSC. 

Discussion: 

Examples of methods of evaluation are presented below. Changes to such methods 

of evaluation require evaluation under 10 CFR 50.59(c)(2)(viii) only for evaluations 

used either in UFSAR safety analyses or in establishing the design bases, and only 

if the methods are described, outlined or summarized in the UFSAR. Methodology 

changes that are subject to 10 CFR 50.59 include changes to elements of existing 

methods described in the UFSAR and to changes that involve replacement of 

existing methods of evaluation with alternative methodologies.

Elements of Methodology Example

m Data correlations 

m Means of data reduction 

m Physical constants or coefficients 

n Mathematical models 

n Specific limitations of a computer 

program 

* Specified factors to account for 

uncertainty in measurements or data 

* Statistical treatment of results 

* Dose conversion factors and assumed 

source term(s)

m DNBR correlations 

* ASME III and Appendix G 

methods for evaluating reactor 

vessel embrittlement specimens 

m Heat transfer coefficients 

m Decay heat models 

m No voiding in PWR hot legs for 

non-LOCA analyses 

* 120% of 1971 decay heat model 

* Vendor-specific thermal design 

procedure 

n ICRP factors

Methods of evaluation described in the UFSAR subject to criterion 10 CFR 

50.59(c)(2)(viii) are: 

m Methods of evaluation used in analyses that demonstrate that 

design basis limits of fission product barriers are met (i.e., for the 

parameters subject to criterion 10 CFR 50.59(c)(2)(vii)) 

* Methods of evaluation used in UFSAR safety analyses, including 

containment, ECCS and accident analyses typically presented in

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UFSAR Chapters 6 and 15, to demonstrate that consequences of 

accidents do not exceed 10 CFR 100 or 10 CFR 50, Appendix A, 

dose limits 

m Methods of evaluation used in supporting UFSAR analyses that 

demonstrate intended design functions will be accomplished under 

design basis conditions that the plant is required to withstand, 

including natural phenomena, environmental conditions, dynamic 

effects, station blackout and ATWS. 

3.11 PROCEDURES AS DESCRIBED IN THE FSAR (AS UPDATED) 

Definition: 

Procedures as described in the final safety analysis report (as updated) 

means those procedures that contain information described in the FSAR (as 

updated) such as how structures, systems, and components are operated and 

controlled (including assumed operator actions and response times). 

Discussion: 

The scope of information that is the focus of 10 CFR 50.59 is the information 

presented in the original FSAR to satisfy the requirements of 10 CFR 

50.34(b), as updated per the requirements of 10 CFR 50.71(e) and as 

supplemented pursuant to 10 CFR 54.2 1(d). 

For purposes of 10 CFR 50.59, "procedures" are not limited to plant 

procedures specifically identified in the UFSAR (e.g., operating and 

emergency procedures). Procedures include UFSAR descriptions of how 

actions related to system operation are to be performed and controls over the 

performance of design functions. This includes UFSAR descriptions of 

operator action sequencing or response times, certain descriptions (text or 

figure) of SSC operation and operating modes, operational and radiological 

controls, and similar information. If changes to these activities or controls 

are made, such changes are considered changes to procedures described in 

the UFSAR, and the changes are subject to 10 CFR 50.59. 

Even if described in the UFSAR, procedures that do not contain information 

on how SSCs are operated or controlled do not meet the definition of 

"procedures as described in the UFSAR" and are not subject to 10 CFR 50.59. 

Sections 4.1.2 and 4.1.4 identify examples of procedures that are not subject 

to 10 CFR 50.59. 

10 CFR 50.59 screening of procedure changes is discussed in Section 4.2.1.2.

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3.12 SAFETY ANALYSES 

Definition: 

Safety analyses are analyses performed pursuant to NRC requirements to 

demonstrate the integrity of the reactor coolant pressure boundary, the 

capability to shut down the reactor and maintain it in a safe shutdown 

condition, or the capability to prevent or mitigate the consequences of 

accidents that could result in potential offsite exposures comparable to the 

guidelines in 10 CFR 50.34(a)(1) or 10 CFR 100.11. Safety analyses are 

required to be presented in the UFSAR per 10 CFR 50.34(b) and 10 CFR 

50.71(e) and include, but are not limited to, the accident analyses typically 

presented in Chapter 15 of the UFSAR. 

Discussion: 

Safety analyses are those analyses or evaluations that demonstrate that 

acceptance criteria for the facility's capability to withstand or respond to 

postulated events are met. Containment, ECCS and accident analyses 

typically presented in Chapters 6 and 15 of the UFSAR clearly fall within the 

meaning of "safety analyses" as defined above. Also within the meaning of 

this definition for purposes of 50.59 are: 

m Supporting UFSAR analyses that demonstrate that SSC design 

functions will be accomplished as credited in the accident analyses 

m UFSAR analyses of events that the facility is required to withstand 

such as turbine missiles, fires, floods, earthquakes, station blackout 

and ATWS. 

Note that, although fire is an event that a plant is required to withstand and 

for which it has been analyzed accordingly in the UFSAR (by reference to the 

Fire Hazards Analysis for some licensees), changes to the fire protection 

program and associated analyses are (for most licensees) governed by licensee 

requirements other than 10 CFR 50.59, as discussed in Section 4.1.5.

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3.13 SCREENING 

Definition: 

Screening is the process for determining whether a proposed activity requires 

a 10 CFR 50.59 evaluation to be performed. 

Discussion: 

Screening is that part of the 10 CFR 50.59 process that determines whether a 

10 CFR 50.59 evaluation is required prior to implementing a proposed activity. 

The definitions of "change," "facility as described...," "procedures as 

described..." and "test or experiment not described..." constitute criteria for 

the 10 CFR 50.59 screening process. Activities that do not meet these criteria 

are said to "screen out" from further review under 10 CFR 50.59, i.e., may be 

implemented without a 10 CFR 50.59 evaluation. 

Engineering and technical information concerning a proposed activity may be 

used along with other information as the basis for determining if the activity 

screens out or requires a 10 CFR 50.59 evaluation. 

Further discussion and guidance on screening are provided in Section 4.2. 

3.14 TESTS OR EXPERIMENTS NOT DESCRIBED IN THE FSAR (AS UPDATED) 

Definition: 

Tests or experiments not described in the final safety analysis report (as 

updated) means any activity where any structure, system, or component is 

utilized or controlled in a manner which is either: 

* Outside the reference bounds of the design bases as described in the 

UFSAR, or 

m Inconsistent with the analyses or descriptions in the UFSAR. 

Discussion: 

10 CFR 50.59 is applied to tests or experiments not described in the UFSAR. 

The intent of the definition is to ensure that tests or experiments that put the 

facility in a situation that has not previously been evaluated (e.g., 

unanalyzed system alignments) or that could affect the capability of SSCs to 

perform their intended design functions (e.g., high flow rates, high

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temperatures) are evaluated before they are conducted to determine if prior 

NRC approval is required. 

Maintenance-related testing is assessed and managed under 10 CFR 

50.65(a)(4), as discussed in Section 4.1.2. 10 CFR 50.59 screening of tests 

and experiments unrelated to maintenance is discussed in Section 4.2.2. 

Examples of tests unrelated to maintenance and thus subject to 10 CFR 

50.59 include (1) most core physics testing, (2) room heat-up testing to 

validate a design/analysis input, and (3) testing to help determine which 

of two redesign alternatives to pursue. 

4 IMPLEMENTATION GUIDANCE 

Licensees may determine applicability and screen activities to determine if 

10 CFR 50.59 evaluations are required as described in Sections 4.1 and 4.2, 

or equivalent manner. 

4.1 APPLICABILITY 

As stated in Section (b) of 10 CFR 50.59, the rule applies to each holder of a 

license authorizing operation of a production or utilization facility, including 

the holder of a license authorizing operation of a nuclear power reactor that 

has submitted a certification of permanent cessation of operations required 

under 10 CFR 50.82(a)(1) or a reactor licensee whose license has been 

amended to allow possession but not operation of the facility. 

4.1.1 Applicability to Licensee Activities 

10 CFR 50.59 is applicable to tests or experiments not described in the 

UFSAR and to changes to the facility or procedures as described in the 

UFSAR, including changes made in response to new requirements or generic 

communications, except as noted below: 

m Per 10 CFR 50.59(c)(1)(i), proposed activities that require a change to the 

technical specifications must be made via the license amendment process, 

10 CFR 50.90. Aspects of proposed activities that are not directly related 

to the required technical specification change are subject to 10 CFR 50.59. 

* To reduce duplication of effort, 10 CFR 50.59(c)(4) specifically excludes 

from the scope of 10 CFR 50.59 changes to the facility or procedures that 

are controlled by other more specific requirements and criteria 

established by regulation. For example, 10 CFR 50.54, which was

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promulgated after 10 CFR 50.59, specifies criteria and reporting 

requirements for changing quality assurance, physical security and 

emergency plans. 

In addition to 50.90 and 50.54(a), (p) & (q), the following include change 

control requirements that meet the intent of 50.59(c)(4) and may take 

precedence over 50.59 for control of specific changes: 

* 10 CFR 50.65 (Maintenance Rule). See additional discussion in 

Section 4.1.2. 

* 10 CFR Part 50, Appendix B, (Quality Assurance Criteria). See 

additional discussion in Section 4.1.4. 

* Standard FP license condition (if applicable). See additional 

discussion in Section 4.1.5. 

* 10 CFR 50.55a (Codes and Standards) 

* 10 CFR 50.46, (ECCS Rule) 

* 10 CFR 50.12, (Specific Exemptions) 

* 10 CFR Part 20 (Standards for Radiation Protection). 

Activities controlled and implemented under other regulations may require 

related information in the UFSAR to be updated. To the extent the UFSAR 

changes are directly related to the activity implemented via another 

regulation, applying 10 CFR 50.59 is not required. UFSAR changes should be 

identified to the NRC as part of the required UFSAR update, per 10 CFR 

50.71(e). However, there may be certain activities for which a licensee would 

need to apply both the requirements of 10 CFR 50.59 and that of another 

regulation. For example, a modification to a facility involves additional 

components and substantial piping reconfigurations as well as changes to 

protection system setpoints. The protection system setpoints are contained in 

the facility technical specifications. Thus, a license amendment to revise the 

technical specifications under 10 CFR 50.90 is required to implement the 

new system setpoints. 10 CFR 50.59 should be applied to the balance of the 

modification, including impacts on required operator actions. 

4.1.2 Maintenance Activities 

Maintenance activities are activities that restore SSCs to their as-designed 

condition, including activities that implement approved design changes. 

Maintenance activities are not subject to 10 CFR 50.59, but are subject to the 

provisions of 10 CFR 50.65(a)(4) as well as technical specifications. 

Maintenance activities include troubleshooting, calibration, refurbishment, 

maintenance-related testing, identical replacements, housekeeping and

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similar activities that do not permanently alter the design, performance 

requirements, operation or control of SSCs. Maintenance activities also 

include temporary alterations to the facility or procedures that directly relate 

to and are necessary to support the maintenance. Examples of temporary 

alterations that support maintenance include jumpering terminals, lifting 

leads, placing temporary lead shielding on pipes and equipment, removal of 

barriers, and use of temporary blocks, bypasses, scaffolding and supports. 

Licensees should ensure operability in accordance with the technical 

specifications and should assess and manage the risk impact of maintenance 

activities per 10 CFR 50.65(a)(4) and NUMARC 93-0 1, Industry Guidelines for 

Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.2 

In addition to assessments required by 10 CFR 50.65(a)(4), 10 CFR 50.59 

should also be applied in the following cases: 

" A temporary alteration in support of the maintenance is expected to 

be in effect during at-power operations for more than 90 days. In 

this case, 10 CFR 50.59 would be applied to the temporary 

alteration prior to implementation in the same manner as a 

permanent change. 

" The plant is not restored to its original condition upon completion of 

the maintenance activity (e.g., if SSCs are removed, the design, 

function or operation is altered, or if temporary alteration in support 

of the maintenance is not removed). In this case, 10 CFR 50.59 

would be applied to the permanent change to the plant. 

Installation and post-modification testing of approved facility changes are 

indistinguishable, in terms of their risk impact on the plant, from 

maintenance activities that restore SSCs to their as-designed condition. As 

such, installation and testing of approved facility changes are maintenance 

activities that must be assessed and managed in accordance with 10 CFR 

50.65(a)(4). This contrasts with historical practice where 10 CFR 50.59 

reviews addressed the design, installation and post-modification testing of 

proposed facility changes. Going forward, 10 CFR 50.59 will address the 

effect, following implementation, of proposed facility changes to determine if 

prior NRC approval is required; the risk impact of actually implementing the 

change will be assessed and managed per 10 CFR 50.65(a)(4). 

2 Regulatory Guide 1.182, issued June 1, 2000, endorses the industry guidance on 10 CFR 50.65(a)(4) 

provided in Section 11 of NUMARC 93-01, Revision 3, August 2000

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If a temporary alteration necessary to install a facility change is expected to 

be in effect longer than 90 days at power, the required 50.59 review of the 

temporary alteration may be performed as part of the 50.59 review for the 

facility change. 

10 CFR 50.59 should be applied to temporary changes proposed as 

compensatory actions for degraded or nonconforming conditions, as discussed 

in Section 4.4. 

Control of Maintenance Procedures 

Changes to procedures for performing maintenance are made in accordance 

with applicable 10 CFR Part 50, Appendix B, criteria and licensee 

procedures. Licensee processes should ensure that changes to plant 

configurations called for by procedures are consistent with the technical 

specifications. 10 CFR 50.59 does not apply to such changes because, like the 

maintenance activities themselves, changes to procedures for performing 

maintenance do not permanently alter the design, performance 

requirements, operation or control of SSCs. 

Certain maintenance procedures, including those for technical specification 

required surveillance and inspection, may contain important information 

concerning SSC design, performance, operation or control. Examples include 

acceptance criteria for valve stroke times or other SSC function, torque 

values, and types of materials (e.g., gaskets, elastomers, lubricants, etc.). 

Licensee design and/or configuration control processes should ensure that 10 

CFR 50.59 is applied to changes in such information and that maintenance 

procedure changes do not inadvertently alter the design, performance 

requirements, operation or control of SSCs. 

If a change to a maintenance procedure affects information in the UFSAR 

(e.g., a specific test or maintenance frequency), the affected information 

should be updated in accordance with 10 CFR 50.71(e). 

4.1.3 UFSAR Modifications 

Per NEI 98-03 (Revision 1, June 1999), as endorsed by Regulatory Guide 

1.181 (September 1999), modifications to the UFSAR that are not the result 

of activities performed under 10 CFR 50.59 are not subject to control under 

10 CFR 50.59. Such modifications include reformatting and simplification of 

UFSAR information and removal of obsolete or redundant information and 

excessive detail.

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Similarly, 10 CFR 50.59 need not be applied to the following types of 

activities: 

m Editorial changes to the UFSAR (including referenced procedures, 

topical reports, etc.) 

m Clarifications to improve reader understanding 

* Correction of inconsistencies within the UFSAR (e.g., between 

sections) 

m Minor corrections to drawings, e.g., correcting mislabeled valves 

m Similar changes to UFSAR information that do not change the 

meaning or substance of information presented. 

4.1.4 Changes to Procedures Governing the Conduct of Operations 

Even if described in the UFSAR, changes to managerial and administrative 

procedures governing the conduct of facility operations are controlled under 

10 CFR 50, Appendix B, programs and are not subject to control under 10 

CFR 50.59. These include, but are not limited to, procedures in the following 

areas: 

m Operations and work process procedures such as control of 

equipment status (tag outs) 

n Shift staffing and personnel qualifications 

m Changes to position titles 

* Administrative controls for creating or modifying procedures 

m Training programs 

m On-site/off-site safety review committees 

m Plant modification process 

* Calculation process 

Example 

The UFSAR states that the shift supervisor will authorize all radioactive 

liquid releases. This is an administrative requirement on the conduct of 

facility operations. Thus assigning this function to another individual would 

not be subject to 10 CFR 50.59 but would be done in accordance with 10 CFR 

Part 50, Appendix B, criteria and licensee procedures. The licensee would be 

required to reflect the change in the next required update of the UFSAR, per 

10 CFR 50.71(e).

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4.1.5 Changes to Approved Fire Protection Programs 

Most nuclear power plant licenses contain a section on fire protection (FP). 

Originally, these fire protection license conditions varied widely in scope and 

content. These variations created problems for licensees and for NRC 

inspectors in identifying the operative and enforceable fire protection 

requirements at each facility. 

To resolve these problems, the NRC promulgated guidance in Generic Letter 

86-10, "Implementation of Fire Protection Requirements," for licensees to: 

* Incorporate the fire protection program and major commitments 

into the FSAR for the facility, and 

* Amend the operating license to substitute a standard fire protection 

license condition for the previous license condition(s) regarding fire 

protection. 

Under the standard fire protection license condition, licensees may 

(1) Make changes to their approved FP programs without prior NRC 

approval provided that the changes would not adversely affect the 

ability to achieve and maintain safe shutdown in the event of a fire, 

and 

(2) Alter specific features of the approved program provided such 

changes do not otherwise involve a change to the license or 

technical specifications, or require an exemption. 

Adoption of the standard fire protection license condition provided a more 

consistent approach to evaluating changes to the facility, including those 

associated with the fire protection program. Originally, changes to the FP 

program under the FP license condition were also subject to 10 CFR 50.59; 

however, this created confusion as to which regulatory requirement governed 

FP program changes. 

10 CFR 50.59(c)(4) provides that when applicable regulations establish more 

specific criteria for controlling certain changes, 10 CFR 50.59 does not also 

apply. Consistent with this intent, the standard fire protection license 

condition establishes specific criteria for control of fire protection changes 

and falls within the scope of 10 CFR 50.59(c)(4). Thus, applying 10 CFR 

50.59 to fire protection program changes is not required. 

Changes to the fire protection program should be evaluated for impacts on 

other design functions, and 10 CFR 50.59 should be applied to the non-fire

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protection related effects of the change, if any. 

Consistent with current practice, determinations made under the standard 

fire protection license condition should be based on a written evaluation that 

remains available for NRC review for the life of the plant. These written 

evaluations should provide the basis for the licensee's conclusion that 

changes to the fire protection program do not adversely affect the ability to 

achieve and maintain safe shutdown in the event of a fire. An evaluation 

performed in accordance with the license condition should include an 

assessment of the impact of the change on the existing fire hazards analysis 

for the area, as is current practice. The assessment should address the 

effects on combustible loading and distribution and should consider whether 

circuits or components, including associated circuits, for a train of equipment 

needed for safe shutdown could be affected, or whether a new element could 

be introduced into the area. 

Under the standard license condition, approved fire protection program 

documents (e.g., fire hazards analysis) are incorporated in the UJFSAR, and 

as such, changes to this information are subject to 10 CFR 50.71(e) reporting 

requirements. 

4.2 SCREENING 

Once it has been determined that 10 CFR 50.59 is applicable to a proposed 

activity, screening is performed to determine if the activity should be 

evaluated against the evaluation criteria of 10 CFR 50.59(c)(2). 

Engineering, design and other technical information concerning the activity 

and affected SSCs should be used to assess whether the activity is a test or 

experiment not described in the UFSAR or a modification, addition or 

removal (i.e., change) that affects: 

"* A design function of an SSC 

"* A method of performing or controlling the design function, or 

"* An evaluation for demonstrating that intended design functions 

will be accomplished. 

Sections 4.2.1 and 4.2.2 provide guidance and examples for determining 

whether an activity is (1) a change to the facility or procedures as described 

in the UFSAR or (2) a test or experiment not described in the UFSAR. If an 

activity is determined to be neither, then it screens out and may be 

implemented without further evaluation under 10 CFR 50.59. Activities that 

are screened out from further evaluation under 10 CFR 50.59 should be 

documented as discussed in Section 4.2.3.

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Each element of a proposed activity must be screened except in instances 

where linking elements of an activity is appropriate, in which case the linked 

elements can be considered together. A test for linking elements of proposed 

changes is interdependence. 

It is appropriate for discrete elements to be considered together if (1) they are 

interdependent as in the case where a modification to a system or component 

necessitates additional changes to other systems or procedures; or (2) they 

are performed collectively to address a design or operational issue. For 

example, a pump upgrade modification may also necessitate a change to a 

support system, such as cooling water. 

If concurrent changes are being made that are not linked, each must be 

screened separately and independently of each other. 

Activities that screen out may nonetheless require UFSAR information to be 

updated. Licensees should provide updated UFSAR information to the NRC 

in accordance with 10 CFR50.71(e). 

Specific guidance for applying 10 CFR 50.59 to temporary changes proposed 

as compensatory actions for degraded or nonconforming conditions is 

provided in Section 4.4. 

4.2.1 Is the Activity a Change to the Facility or Procedures as Described 

in the UFSAR? 

To determine whether or not a proposed activity affects a design function, 

method of performing or controlling a design function or an evaluation that 

demonstrates that design functions will be accomplished, a thorough 

understanding of the proposed activity is essential. A given activity may 

have both direct and indirect effects that the screening review must consider. 

The following questions illustrate a range of effects that may stem from a 

proposed activity: 

m Does the activity decrease the reliability of an SSC design function, 

including either functions whose failure would initiate a transient/ 

accident or functions that are relied upon for mitigation? 

m Does the activity reduce existing redundancy, diversity or defense

in-depth? 

m Does the activity add or delete an automatic or manual design 

function of the SSC?

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"* Does the activity convert a feature that was automatic to manual or 

vice versa? 

"* Does the activity introduce an unwanted or previously unreviewed 

system or materials interaction? 

"* Does the activity adversely affect the ability or response time to 

perform required actions, e.g., alter equipment access or add steps 

necessary for performing tasks? 

"* Does the activity degrade the seismic or environmental 

qualification of the SSC? 

"* Does the activity adversely affect other units at a multiple unit 

site? 

"* Does the activity affect a method of evaluation used in establishing 

the design bases or in the safety analyses? 

"* For activities affecting SSCs, procedures, or methods of evaluation 

that are not described in the UFSAR, does the change have an 

indirect effect on electrical distribution, structural integrity, 

environmental conditions or other UFSAR-described design 

functions? 

Per the definition of "change" discussed in Section 3.3, 10 CFR 50.59 is 

applicable to additions as well as to changes to and removals from the facility 

or procedures. Additions should be screened for their effects on the existing 

facility and procedures as described in the UFSAR and, if required, a 10 CFR 

50.59 evaluation should be performed. NEI 98-03 provides guidance for 

determining whether additions to the facility and procedures should be 

reflected in the UFSAR per 10 CFR 50.71(e). 

Consistent with historical practice, changes affecting SSCs or functions not 

described in the UFSAR must be screened for their effects (so-called "indirect 

effects") on UFSAR-described design functions. A 10 CFR 50.59 evaluation is 

required when such changes adversely affect a UFSAR-described design 

function, as described below. 

Screening for Adverse Effects 

A 10 CFR 50.59 evaluation is required for changes that adversely affect 

design functions, methods used to perform or control design functions, or 

evaluations that demonstrate that intended design functions will be

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accomplished (i.e., "adverse changes"). Changes that have none of these 

effects, or have positive effects, may be screened out because only adverse 

changes have the potential to increase the likelihood of malfunctions, 

increase consequences, create new accidents or otherwise meet the 10 CFR 

50.59 evaluation criteria.3 

Per the definition of "design function," SSCs may have preventive, as well as 

mitigative, design functions. Adverse changes to either must be screened in. 

Thus a change that decreases the reliability of a function whose failure could 

initiate an accident would be considered to adversely affect a design function 

and would screen in. In this regard, changes that would relax the manner in 

which Code requirements are met for certain SSCs should be screened for 

adverse effects on design function. Similarly, changes that would introduce a 

new type of accident or malfunction would screen in. This reflects an overlap 

between the technical/engineering ("safety") review of the change and 10 

CFR 50.59. This overlap reflects that these considerations are important to 

both the safety and regulatory reviews. 

If a change has both positive and adverse effects, the change should be 

screened in. The 10 CFR 50.59 evaluation should focus on the adverse 

effects. 

The screening process is not concerned with the magnitude of adverse effects 

that are identified. Any change that adversely affects a UFSAR-described 

design function, method of performing or controlling design functions, or 

evaluation that demonstrates that intended design functions will be 

accomplished is screened in. The magnitude of the adverse effect (e.g., is the 

minimal increase standard met?) is the focus of the 10 CFR 50.59 evaluation 

process. 

Screening determinations are made based on the engineering/technical 

information supporting the change. The screening focus on design functions, 

etc., ensures the essential distinction between (1) 10 CFR 50.59 screenings, 

and (2) 10 CFR 50.59 evaluations, which focus on whether changes meet any 

of the eight criteria in 10 CFR 50.59(c)(2). Technical/engineering 

information, e.g., design evaluations, etc., that demonstrates changes have no 

adverse effect on UFSAR-described design functions, methods of performing 

or controlling design functions, or evaluations that demonstrate that 

intended design functions will be accomplished may be used as basis for 

screening out the change. If the effect of a change is such that existing 

safety analyses would no longer be bounding and therefore UFSAR safety 

analyses must be re-run to demonstrate that all required safety functions 

3 Note that as discussed in Section 4.2.1.1, any change that alters a design basis limit for a fission 

product barrier-positively or negatively-is considered adverse and must be screened in.

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and design requirements are met, the change is considered to be adverse and 

must be screened in. The revised safety analyses may be used in support of 

the required 10 CFR 50.59 evaluation of such changes. 

Changes that entail update of safety analyses to reflect improved 

performance, capacity, timing, etc., resulting from a change (beneficial effects 

on design functions) are not considered adverse and need not be screened in, 

even though the change calls for safety analyses to be updated. For example, 

a change that improves the closure time of main control room isolation 

dampers reduces the calculated dose to operators, and UFSAR dose 

consequence analyses are to be updated as a result. In this case, the dose 

analyses are being revised to reflect the lower dose for the main control room, 

not to demonstrate that GDC limits continue to be met. A change that would 

adversely affect the design function of the dampers (post-accident isolation of 

the main control room) and increase the existing calculated dose to operators 

would be considered adverse and would screen in. In this case, the dose 

analyses must be re-run to ensure that GDC limits continue to be met. The 

revised analyses would be used in support of the 10 CFR 50.59 evaluation to 

determine if the increase exceeds the minimal standard and requires prior 

NRC approval. 

To further illustrate the distinction between 10 CFR 50.59 screening and 

evaluation, consider the example of a change to a diesel generator-starting 

relay that delays the diesel start time from 10 seconds to 12 seconds. The 

UFSAR-described design function credited in the ECCS analyses is for the 

diesel to start within 12 seconds. This change would screen out because it is 

apparent that the change will not adversely affect the diesel generator design 

function credited in the ECCS analyses (ECCS analyses remain valid). 

However, a change that would delay the diesel's start time to 13 seconds 

would screen in because the change adversely effects the design function (to 

start in 12 seconds). Such a change would screen in even if 

technical/engineering information supporting the change includes revised 

safety analyses that demonstrate all required safety functions supported by 

the diesel, e.g., core heat removal, containment isolation, containment 

cooling, etc., are satisfied and that applicable dose limits continue to be met. 

While this change may be acceptable with respect to performance of required 

safety functions and meeting design requirements, the analyses necessary to 

demonstrate acceptability are beyond the scope/intent of 10 CFR 50.59 

screening reviews. Thus a 10 CFR 50.59 evaluation would be required. The 

revised safety analyses would be used in support of the 10 CFR 50.59 

evaluation to determine whether any of the evaluation criteria are met such 

that prior NRC approval is required for the change. Additional specific

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guidance for identifying adverse effects due to a procedure or methodology 

change is provided in subsections 4.2.1.2 and 4.2.1.3, respectively. 

4.2.1.1 Screening of Changes to the Facility as Described in the UFSAR 

Screening to determine that a 10 CFR 50.59 evaluation is required is 

straightforward when a change adversely affects an SSC design function, 

method of performing or controlling a design function, or evaluation that 

demonstrates intended design functions will be accomplished as described in 

the UFSAR. 

However, a facility also contains many SSCs not described in the UFSAR. 

These can be components, subcomponents of larger components or even 

entire systems. Changes affecting SSCs that are not explicitly described in 

the UFSAR can have the potential to adversely affect SSC design functions 

that are described and thus may require a 10 CFR 50.59 evaluation. In such 

cases, the approach for determining whether a change involves a change to 

the facility as described in the UFSAR is to consider the larger, UFSAR

described SSC of which the SSC being modified is a part. If for the larger 

SSC, the change adversely affects a UFSAR-described design function, 

method of performing or controlling the design function, or an evaluation 

demonstrating that intended design functions will be accomplished, then a 

10 CFR 50.59 evaluation is required. 

Another important consideration is that a change to nonsafety-related SSCs 

not described in the UFSAR can indirectly affect the capability of SSCs to 

perform their UFSAR-described design function(s). For example, increasing 

the heat load on a non safety-related heat exchanger could compromise the 

cooling system's ability to cool safety-related equipment. 

Seismic qualification, missile protection, flooding protection, fire protection, 

environmental qualification, high energy line break and masonry block walls 

are some of the areas where changes to nonsafety-related SSCs, whether or 

not described in the UFSAR, can affect the UFSAR-described design function 

of SSCs through indirect or secondary effects. 

Equivalent replacement is a type of change to the facility that does not alter 

the design functions of SSCs. Licensee equivalence assessments, e.g., 

consideration of performance/operating characteristics and other factors, may 

thus form the basis for screening determinations that no 10 CFR 50.59 

evaluation is required. 

As discussed in Section 4.2.1, only proposed changes to SSCs that would, 

based on supporting engineering and technical information, have adverse

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effects on design functions require evaluation under 10 CFR 50.59. Changes 

that have positive or no effect on design functions may generally be screened 

out. In addition, any change to a design bases limit for a fission product 

barrier must be considered adverse and screened in. This is because 10 CFR 

50.59(c)(2)(vii) requires prior NRC approval any time a proposed change 

would "exceed or alter" a design bases limit for a fission product barrier. 

The following examples illustrate the 10 CFR 50.59 screening process as 

applied to proposed facility changes: 

Example 1 

A licensee proposes to replace a relay in the overspeed trip circuit of an 

emergency diesel generator with a nonequivalent relay. The relay is not 

described in the UFSAR, but the design functions of the overspeed trip 

circuit and the emergency diesel generator are. Based on 

engineering/technical information supporting the change, the licensee 

determines if replacing the relay would adversely affect the design function 

of either the overspeed trip circuit or EDG. If the licensee concludes that the 

change would not adversely affect the UFSAR-described design function of 

the circuit or EDG, then this determination would form the basis for 

screening out the change, and no 10 CFR 50.59 evaluation would be required. 

Example 2 

A licensee proposes a nonequivalent change to the operator on one of the 

safety injection accumulator isolation valves. The UFSAR describes that 

these isolation valves are open with their circuit breakers open during 

normal operation. These are motor operated, safety-related valves required 

for pressure boundary integrity and to remain open so that flow to the RCS 

will occur during a LOCA as RCS pressure drops below -600 psi. They are 

remotely closed during a normal shutdown so as to not inject when not 

required. Technical/engineering work supporting this change ensures that 

the replacement operator is capable of performing the functions of the 

existing operator and will not adversely affect the connected Class 1E bus or 

diesel. This change would screen out because (1) the valve operator does not 

perform, support or impact the UFSAR-described design function (to ensure 

pressure boundary integrity and remain open when required) that supports 

safety injection performance credited in the safety analyses, and (2) the 

change does not adversely affect other SSC design functions (e.g., of the 

Class 1E bus). 

If the proposed change was to configure the valve as a normally closed valve 

that automatically opens on loss of reactor coolant system pressure, 10 CFR

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50.59 evaluation would be required because the change would adversely 

affect the reliability of the safety injection function as credited in the safety 

analyses. 

Example 3 

A licensee proposes to replace a globe valve with a ball valve in a vent/drami 

application to reduce the propensity of this valve to leak. The UFSAR

described design function of this valve is to maintain the integrity of the 

system boundary when closed. The vent/drain function of the valve does not 

relate to design functions credited in the safety analyses, and the licensee 

has determined that a ball valve is adequate to support the vent/drain 

function and is superior to the globe valve in terms of its isolation function. 

Thus the proposed change affects the design of the existing vent/drain 

valve-not the design function (pressure boundary integrity) that supports 

system performance credited in the safety analyses-and 

evaluation/reporting under 10 CFR 50.59 is not required. The screening 

determination should be documented, and the UFSAR should be updated per 

10 CFR 50.7 1(e) to reflect the change. 

Example 4 

The bolts for retaining a rupture disk are being replaced with bolts of a 

different material and fewer threads, but equivalent load capacity and 

strength, such that the rupture disk will still relieve at the same pressure as 

before the change. Because the replacement bolts are equivalent to the 

original bolts, the design function of the rupture disk (to relieve at a specified 

pressure) is unaffected, and this activity may be screened out as an 

equivalent change. 

4.2.1.2 Screening of Changes to Procedures as Described in the UFSAR 

Changes are "screened in" (i.e., require a 10 CFR 50.59 evaluation) if they 

adversely affect how SSC design functions are performed or controlled 

(including changes to UFSAR-described procedures, assumed operator 

actions and response times). Proposed changes that are determined to have 

positive or no effect on how SSC design functions are performed or controlled 

may be screened out. 

For purposes of 10 CFR 50.59 screening, changes that fundamentally alter 

(replace) the existing means of performing or controlling design functions 

should be conservatively treated as adverse and screened in. Such changes 

include replacement of automatic action by manual action (or vice versa),

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changes to the man-machine interface, changing a valve from 'locked closed" 

to "administratively closed" and similar changes. 

The following examples illustrate the 10 CFR 50.59 screening process as 

applied to proposed changes affecting how SSC design functions are 

performed or controlled: 

Example 1 

Emergency operating procedures include operator actions and response times 

associated with response to design basis events, which are described in the 

UJFSAR, but also address operator actions for severe accident scenarios that 

are outside the design basis and not described in the UFSAR. A change 

would screen out at this step if the change was to those procedures or parts of 

procedures dealing with operator actions during severe accidents. 

Example 2 

If the UFSAR description of the reactor start-up procedure contains eight 

fundamental sequences, the licensee's decision to eliminate one of the 

sequences would screen in. On the other hand, if the licensee consolidated 

the eight fundamental sequences and did not affect the method of controlling 

or performing reactor start-up, the change would screen out. 

Example 3 

The UFSAR states that a particular flow path is isolated by a locked closed 

valve when not in use. A procedure change would remove the lock from this 

valve such that it becomes a normally closed valve. In this case, the design 

function is to remain closed, and the method of performing the design 

function has fundamentally changed from locked closed to administratively 

closed. Thus this change would screen in and require a 10 CFR 50.59 

evaluation to be performed. 

Example 4 

Operations proposes to revise its procedures to change from 8-hour shifts to 

12-hour shifts. This change results in mid-shift rounds being conducted every 

6 hours as opposed to every 4 hours. The UFSAR describes high energy line 

breaks including mitigation criteria. Operator action to detect and terminate 

the line break is described in the UFSAR, which specifically states that 4 

hours is assumed for the pipe break to go undetected before it would be 

identified during operator mid-shift rounds. The change from 4 to 6 hour 

rounds is a change to a procedure as described in the UFSAR that adversely 

affects the timing of operator actions credited in the safety analyses for

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limiting the effects of high energy line breaks. Therefore, this change 

screens in, and a 10 CFR 50.59 evaluation is required. 

4.2.1.3 Screening Changes to UFSAR Methods of Evaluation 

As discussed in Section 3.6, methods of evaluation included in the UFSAR to 

demonstrate that intended SSC design functions will be accomplished are 

considered part of the "facility as described in the UFSAR." Thus use of new 

or revised methods of evaluation (as defined in Section 3.10) is considered to 

be a change that is controlled by 10 CFR 50.59 and needs to be considered as 

part of this screening step. Adverse changes to elements of a method of 

evaluation included in the UFSAR, or use of an alternative method, must be 

evaluated under 10 CFR 50.59(c)(2)(viii) to determine if prior NRC approval 

is required (see Section 4.3.8). Changes to methods of evaluation (only) do 

not require evaluation against the first seven criteria. 

Changes to methods of evaluation not included in the UIFSAR or to 

methodologies included in the UFSAR that are not used in the safety 

analyses or to establish design bases may be screened out. 

Methods of evaluation that may be identified in references listed at the end 

of UFSAR sections or chapters are not subject to control under 10 CFR 50.59 

unless the UFSAR states they were used for specific analyses within the 

scope of 10 CFR 50.59(c)(2)(viii). 

Changes to methods of evaluation included in the UFSAR are considered 

adverse and require evaluation under 10 CFR 50.59 if the changes are 

outside the constraints and limitations associated with use of the method, 

e.g., identified in a topical report and/or SER. If the changes are within 

constraints and limitations associated with use of the method, the change is 

not considered adverse and may be screened out. 

Proposed use of an alternative method is considered an adverse change that 

must be evaluated under 10 CFR 50.59(c)(2)(viii). 

The following examples illustrate the screening of changes to methods of 

evaluation: 

Example 1 

The UFSAR identifies the name of the computer code used for performing 

containment performance analyses, with no further discussion of the methods 

employed within the code for performing those analyses. Changes to the 

computer code may be screened out provided that the changes are within the

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constraints and limitations identified in the associated topical report and SER. 

A change that goes beyond restrictions on the use of the method would be 

considered adverse and evaluated under 10 CFR 50.59(c)(2)(viii) to determine 

if prior NRC approval is required. 

Example 2 

The UFSAR describes the methods used for atmospheric heat transfer and 

containment pressure response calculations contained within the CONTEMPT 

computer code. The code is also used for developing long-term temperature 

profiles (post-recirculation phase of LOCA) for environmental qualification 

through modeling of the residual heat removal system. Neither this 

application of the code nor the analysis method is discussed in the UFSAR. A 

revision to CONTEMPT to incorporate more dynamic modeling of the residual 

heat removal system transfer of heat to the ultimate heat sink would screen 

out because this application of the code is not described in the UFSAR as being 

used in the safety analyses or to establish design bases. Changes to 

CONTEMPT that affect the atmospheric heat transfer or containment pressure 

predictions may not screen out (because the UFSAR describes this application 

in the safety analyses), and may require a 10 CFR 50.59 evaluation. 

Example 3 

The steamline break mass and energy release calculations were originally 

performed at a power level of 105% of the nominal power (plus uncertainties) in 

order to allow margin for a future power up-rate. The utility later decided that 

it would not pursue the power up-rate and wished to use the margin to address 

other equipment qualification issues. The steamline break mass and energy 

release calculations were reanalyzed, using the same methodology, at 100% 

power (plus uncertainties). This change would screen out as a methodology 

change because the proposed activity involved a change to an input parameter 

(% power) and not a methodology change. This change should be screened per 

Section 4.2.1.1 to determine if it constitutes a change to the facility as described 

in the UFSAR that requires evaluation under 10 CFR 50.59(c)(2)(i-vii). 

Example 4 

The LOCA mass and energy release calculations were originally performed at 

a power level of 105% of the nominal power, plus uncertainties. Some of the 

assumptions in the analysis were identified as nonconservative, but the NRC 

concluded in the associated SER that the overall analysis was conservative 

because of the use of the higher initial power. The utility later decided that 

it would not pursue the power up-rate and wished to use the margin to 

address other equipment qualification issues. The LOCA break mass and

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energy release calculations were reanalyzed, using the same methodology, at 

100% power (plus uncertainties). This change would not screen out because 

the proposed activity involved a change to an input parameter that was 

integral to the NRC approval of the methodology. 

Example 5 

Due to fuel management changes, core physics parameters change for a 

particular reload cycle. The topical report and associated SER that describe 

how the core physics parameters are to be calculated explicitly allow use of 

either 2-D or 3-D modeling for the analysis. A change to add or remove 

discretionary conservatism via use of 3-D methods instead of 2-D methods or 

vice-versa would screen out because the change is within the terms and 

conditions of the SER. 

4.2.2 Is the Activity a Test or Experiment Not Described in the UFSAR? 

As discussed in Section 3.14, tests or experiments not described in the 

UFSAR are activities where an SSC is utilized or controlled in a manner that 

is outside the reference bounds of the design for that SSC or inconsistent 

with analyses or description in the UFSAR. 

As discussed in Section 4.1.2, testing associated with maintenance is assessed 

and managed under 10 CFR 50.65(a)(4) and is not subject to 10 CFR 50.59. 

Tests and experiments that are described in the UFSAR may be screened out 

at this step. Tests and experiments that are not described in the UFSAR 

may be screened out provided the test or experiment is bounded by tests and 

experiments that are described. Similarly, tests and experiments not 

described in the UFSAR may be screened out provided that affected SSCs 

will be appropriately isolated from the facility. 

Examples of tests that would "screen in" at this step (assuming they were not 

associated with maintenance or described in the UFSAR) would be: 

1. For BWIRs, hydrogen injection into the reactor coolant system to 

minimize stress corrosion cracking 

2. For BWRs, zinc injection into the reactor coolant system to reduce 

activation 

3. For PWRs, ECCS flow tests that affect the ability to remove decay heat 

4. Operation with fuel demonstration assemblies.

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Examples of tests that would "screen out" would be: 

1. Steam generator moisture carryover tests (provided such testing is 

described in the UFSAR) 

2. Balance-of-plant heat balance test 

3. Information gathering that is nonintrusive to the operation or design 

function of the associated SSC. 

4.2.3 Screening Documentation 

10 CFR 50.59 record-keeping requirements apply to 10 CFR 50.59 

evaluations performed for activities that screened in, not to screening records 

for activities that screened out. However, documentation should be 

maintained in accordance with plant procedures of screenings that conclude 

a proposed activity may be screened out (i.e., that a 10 CFR 50.59 evaluation 

was not required). The basis for the conclusion should be documented to a 

degree commensurate with the safety significance of the change. For 

changes, the documentation should include the basis for determining that 

there would be no adverse effect on design functions, etc. Typically, the 

screening documentation is retained as part of the change package. This 

documentation does not constitute the record of changes required by 10 CFR 

50.59, and thus is not subject to 10 CFR 50.59 documentation and reporting 

requirements. Screening records need not be retained for activities for which 

a 10 CFR 50.59 evaluation was performed or for activities that were never 

implemented. 

4.3 EvALUATION PROCESS 

Once it has been determined that a given activity requires a 10 CFR 50.59 

evaluation, the written evaluation must address the applicable criteria of 10 

CFR 50.59(c)(2). These eight criteria are used to evaluate the effects of 

proposed activities on accidents and malfunctions previously evaluated in the 

UFSAR and their potential to cause accidents or malfunctions whose effects 

are not bounded by previous analyses. 

Criteria (c)(2)(i-vii) are applicable to activities other than changes in methods 

of evaluation. Criterion (c)(2)(viii) is applicable to changes in methods of 

evaluation. Each activity must be evaluated against each applicable criterion.

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If any of the criteria are met, the licensee must apply for and obtain a license 

amendment per 10 CFR 50.90 before implementing the activity. The 

evaluation against each criterion should be appropriately documented as 

discussed in Section 4.5. Subsections 4.3.1 through 4.3.8 provide guidance and 

examples for evaluating proposed activities against the eight criteria. 

Each element of a proposed activity must undergo a 10 CFR 50.59 

evaluation, except in instances where linking elements of an activity is 

appropriate, in which case the linked elements can be evaluated together. A 

test for linking elements of proposed changes is interdependence. 

It is appropriate for discrete elements to be evaluated together if (1) they are 

interdependent as in the case where a modification to a system or component 

necessitates additional changes to other systems or procedures; or (2) they 

are performed collectively to address a design or operational issue. For 

example, a pump upgrade modification may also necessitate a change to a 

support system, such as cooling water. 

If concurrent changes are being made that are not linked, each must be 

evaluated separately and independently of each other. 

The effects of a proposed activity being evaluated under 10 CFR 50.59 should 

be assessed against each of the evaluation criteria separately. For example, 

an increase in frequency/likelihood of occurrence cannot be compensated for 

by additional mitigation of consequences. Evaluations should consider the 

effects of the proposed activity on operator actions. 

Specific guidance for applying 10 CFR 50.59 to temporary changes proposed 

as compensatory actions for degraded or nonconforming conditions is 

provided in Section 4.4. 

4.3.1 Does the Activity Result in More Than a Minimal Increase in the 

Frequency of Occurrence of an Accident? 

In answering this question, the first step is to identify the accidents that 

have been evaluated in the UFSAR that are affected by the proposed activity. 

Then a determination should be made as to whether the frequency of these 

accidents occurring would be more than minimally increased. 

For most licensees, accidents and transients have been divided into 

categories based upon a qualitative assessment of frequency. For example, 

ANSI standards define the following categories for plant conditions for most 

PWRs as follows:

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* Normal Operations - Expected frequently or regularly in the course 

of power operation, refueling, maintenance or maneuvering 

* Incidents of Moderate Frequency - Any one incident expected per 

plant during a calendar year 

* Infrequent Incidents - Any one incident expected per plant during 

plant lifetime 

* Limiting Faults - Not expected to occur but could release significant 

amounts of radioactive material thus requiring protection by design. 

ANSI standards for BWRs have slightly different but equivalent definitions. 

During initial plant licensing, accidents were typically assessed in relative 

frequencies, as described above. Minimal increases in frequency resulting 

from subsequent licensee activities do not significantly change the licensing 

basis of the facility and do not impact the conclusions reached about 

acceptability of the facility design. 

Since accident and transient frequencies were considered in a broad sense as 

described above, a change from one frequency category to a more frequent 

category is clearly an example of a change that results in more than a 

minimal increase in the frequency of occurrence of an accident. 

Changes within a frequency category could also result in more than a 

minimal increase in the frequency of occurrence of an accident. Normally, 

the determination of a frequency increase is based upon a qualitative 

assessment using engineering evaluations consistent with the UFSAR 

analysis assumptions. However, a plant-specific accident frequency 

calculation or PRA may be used to evaluate a proposed activity in a 

quantitative sense. It should be emphasized that PRAs are just one of the 

tools for evaluating the effect of proposed activities, and their use is not 

required to perform 10 CFR 50.59 evaluations. 

Reasonable engineering practices, engineering judgment and PRA 

techniques, as appropriate, should be used in determining whether the 

frequency of occurrence of an accident would more than minimally increase 

as a result of implementing a proposed activity. A large body of knowledge 

has been developed in the area of accident frequency and risk significant 

sequences through plant-specific and generic studies. This knowledge, where 

applicable, should be used in determining what constitutes more than a 

minimal increase in the frequency of occurrence of an accident previously 

evaluated in the UFSAR. The effect of a proposed activity on the frequency

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of an accident must be discernable and attributable to the proposed activity 

in order to exceed the more than minimal increase standard. 

Although this criterion allows minimal increases, licensees must still meet 

applicable regulatory requirements and other acceptance criteria to which 

they are committed (such as contained in regulatory guides and nationally 

recognized industry consensus standards, e.g., the ASMIE B&PV Code and 

IEEE standards). Further, departures from the design, fabrication, 

construction, testing and performance standards as outlined in the General 

Design Criteria (Appendix A to Part 50) are not compatible with a "no more 

than minimal increase" standard. 

Because frequencies of occurrence of natural phenomena were established as 

part of initial licensing and are not expected to change, changes in design 

requirements for earthquakes, tornadoes and other natural phenomena 

should be treated as potentially affecting the likelihood of a malfunction 

rather than the frequency of occurrence of an accident. 

The following are examples where there is not more than a minimal increase 

in the frequency of occurrence of an accident: 

Example 1 

The proposed activity has a negligible effect on the frequency of occurrence of 

an accident. A negligible effect on the frequency of occurrence of an accident 

exists when the change in frequency is so small or the uncertainties in 

determining whether a change in frequency has occurred are such that it 

cannot be reasonably concluded that the frequency has actually changed (i.e., 

there is no clear trend toward increasing the frequency). 

Example 2 

The proposed activity meets applicable NRC requirements as well as the 

design, material and construction standards applicable to the SSC being 

modified. If the proposed activity would not meet applicable requirements 

and standards, the change is considered to involve more than a minimal 

increase in the frequency of occurrence of an accident, and prior NRC 

approval is required. 

Example 3 

The change in frequency of occurrence of an accident is calculated to support 

the evaluation of the proposed activity, and one of the following criteria are 

met:

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"* The increase in the pre-change accident or transient frequency does 

not exceed 10 percent or 

"* The resultant frequency of occurrence remains below 1E-6 per year 

or applicable plant-specific threshold. 

If the proposed activity would not meet either of the above criteria, the 

change is considered to involve more than a minimal increase in the 

frequency of occurrence of an accident, and prior NRC approval is required. 

4.3.2 Does the Activity Result in More Than a Minimal Increase in the 

Likelihood of Occurrence of a Malfunction of an SSC Important to 

Safety? 

The term "malfunction of an SSC important to safety" refers to the failure of 

structures, systems and components (SSCs) to perform their intended design 

functions-including both nonsafety-related and safety-related SSCs. The 

cause and mode of a malfunction should be considered in determining 

whether there is a change in the likelihood of a malfunction. The effect or 

result of a malfunction should be considered in determining whether a 

malfunction with a different result is involved per Section 4.3.6. 

In determining whether there is more than a minimal increase in the 

likelihood of occurrence of a malfunction of an SSC to perform its design 

function as described in the UFSAR, the first step is to determine what SSCs 

are affected by the proposed activity. Next, the effects of the proposed 

activity on the affected SSCs should be determined. This evaluation should 

include both direct and indirect effects. 

Direct effects are those where the proposed activity affects the SSCs (e.g., a 

motor change on a pump). Indirect effects are those where the proposed 

activity affects one SSC and this SSC affects the capability of another SSC to 

perform its UFSAR-described design function. Indirect effects also include 

the effects of proposed activities on the design functions of SSCs credited in 

the safety analyses. The safety analysis assumes certain design functions of 

SSCs in demonstrating the adequacy of design. Thus, certain design 

functions, while not specifically identified in the safety analysis, are credited 

in an indirect sense. 

After determining the effect of the proposed activity on the important to 

safety SSCs, a determination is made of whether the likelihood of a 

malfunction of the important to safety SSCs has increased more than 

minimally. Qualitative engineering judgment and/or an industry precedent 

is typically used to determine if there is more than a minimal increase in the

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likelihood of occurrence of a malfunction. An appropriate calculation can be 

used to demonstrate the change in likelihood in a quantitative sense, if 

available and practical. The effect of a proposed activity on the likelihood of 

malfunction must be discernable and attributable to the proposed activity in 

order to exceed the more than minimal increase standard. A proposed 

activity is considered to have a negligible effect on the likelihood of a 

malfunction when a change in likelihood is so small or the uncertainties in 

determining whether a change in likelihood has occurred are such that it 

cannot be reasonably concluded that the likelihood has actually changed (i.e., 

there is no clear trend toward increasing the likelihood). A proposed activity 

that has a negligible effect satisfies the minimal increase standard. 

Evaluations of a proposed activity for its effect on likelihood of a malfunction 

would be performed at level of detail that is described in the UFSAR. The 

determination of whether the likelihood of malfunction is more than 

minimally increased is made at a level consistent with existing UFSAR

described failure modes and effects analyses. While the evaluation should 

take into account the level that was previously evaluated in terms of 

malfunctions and resulting event initiators or mitigation impacts, it also 

needs to consider the nature of the proposed activity. Thus, for instance, if 

failures were previously postulated on a train level because the trains were 

independent, a proposed activity that introduces a cross-tie or credible 

common mode failure (e.g., as a result of an analog to digital upgrade) should 

be evaluated further to see whether the likelihood of malfunction has been 

increased. 

Changes in design requirements for earthquakes, tornadoes and other 

natural phenomena should be treated as potentially affecting the likelihood 

of malfunction. 

Although this criterion allows minimal increases, licensees must still meet 

applicable regulatory requirements and other acceptance criteria to which 

they are committed (such as contained in regulatory guides and nationally 

recognized industry consensus standards, e.g., the ASME B&PV Code and 

IEEE standards). Further, departures from the design, fabrication, 

construction, testing and performance standards as outlined in the General 

Design Criteria (Appendix A to Part 50) are not compatible with a "no more 

than minimal increase" standard. 

Examples 1-4, below, illustrate cases where there would not be more than a 

minimal increase in the likelihood of occurrence of a malfunction of an SSC 

important to safety:

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Example 1 

The change involves installing additional equipment or devices (e.g., cabling, 

manual valves, protective features) provided all applicable design and 

functional requirements (including applicable codes, standards, etc.) continue 

to be met. For example, adding protective devices to breakers or installing 

an additional drain line (with appropriate isolation capability) would not 

cause more than a minimal increase in the likelihood of malfunction. 

Example 2 

The change involves substitution of one type of component for another of 

similar function, provided all applicable design and functional requirements 

(including applicable codes, standards, etc.) continue to be met and any new 

failure modes are bounded by the existing analysis. 

Example 3 

The change satisfies applicable design bases requirements (e.g., seismic and 

wind loadings, separation criteria, environmental qualification, etc.). 

Example 4 

The change involves a new or modified operator action that supports a design 

function credited in safety analyses provided: 

m The action (including required completion time) is reflected in plant 

procedures and operator training programs 

m The licensee has demonstrated that the action can be completed in 

the time required considering the aggregate affects, such as 

workload or environmental conditions, expected to exist when the 

action is required 

w The evaluation of the change considers the ability to recover from 

credible errors in performance of manual actions and the expected 

time required to make such a recovery 

a The evaluation considers the effect of the change on plant systems. 

Examples 5-8 are cases that would require prior NRC approval because they 

would result in more than a minimal increase in the likelihood of occurrence 

of a malfunction of an SSC important to safety:

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Example 5 

The change would cause design stresses to exceed their code allowables or 

other applicable stress or deformation limit (if any), including vendor

specified stress limits for pump casings that ensure pump functionality. 

Example 6 

The change would reduce system/equipment redundancy, diversity, 

separation or independence. 

Example 7 

The change would (permanently) substitute manual action for automatic 

action for performing UFSAR-described design functions. (Guidance for 

temporary substitution of manual action for automatic action to compensate 

for a degraded/nonconforming condition is provided in NRC Generic Letter 

91-18, Revision 1.) 

Example 8 

The change in likelihood of occurrence of a malfunction is calculated in 

support of the evaluation and increases by more than a factor of two. Note: 

The factor of two should be applied at the component level. Certain changes 

that satisfy the factor of two limit on increasing likelihood of occurrence of 

malfunction may meet one of the other criteria for requiring prior NRC 

approval, e.g., exceed the minimal increase standard for accident/transient 

frequency under criterion 10 CFR 50.59(c)(2)(i). For example, a change that 

increases the likelihood of malfunction of an emergency diesel generator by a 

factor of two may cause more than a 10% increase in the frequency of station 

blackout. 

4.3.3 Does the Activity Result in More Than a Minimal Increase in the 

Consequences of an Accident? 

The UFSAR, based on logic similar to ANSI standards, provides an 

acceptance criterion and frequency relationship for "conditions for design." 

When determining which activities represent "more than a minimal increase 

in consequences" pursuant to 10 CFR 50.59, it must be recognized that 

"consequences" means dose. Therefore, an increase in consequences must 

involve an increase in radiological doses to the public or to control room 

operators. Changes in barrier performance or other outcomes of the proposed 

activity that do not result in increased radiological dose to the public or to

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control room operators are addressed under Section 4.3.7, concerning 

integrity of fission product barriers, or the other criteria of 10 CFR 

50.59(c)(2). 

NRC regulates compliance with the provisions of 10 CFR 50 and 10 CFR 100 

to assure adequate protection of the public health and safety. Activities 

affecting on-site dose consequences that may require prior NRC approval are 

those that impede required actions inside or outside the control room to 

mitigate the consequences of reactor accidents. For changes affecting the 

dose to operators performing required actions outside the control room, an 

increase is considered more than minimal if the resultant "mission dose" 

exceeds applicable GDC 19 criteria. The guidance in the remainder of this 

section applies to evaluation of effects of changes on main control room and 

off-site doses. 

The consequences covered include dose resulting from any accident evaluated 

in the UFSAR. The accidents include those typically covered in UFSAR 

Chapters 6 and 15 and other events for which the plant is designed to cope 

and are described in the UFSAR (e.g., turbine missiles and flooding). The 

consequences referred to in 10 CFR 50.59 do not apply to occupational 

exposures resulting from routine operations, maintenance, testing, etc. 

Occupational doses are controlled and maintained As Low As Reasonably 

Achievable (ALARA) through formal licensee programs. 

10 CFR Part 20 establishes requirements for protection against radiation 

during normal operations, including dose criteria relative to radioactive 

waste handling and effluents. 10 CFR 50.59 accident dose consequence 

criteria and evaluation guidance are not applicable to proposed activities 

governed by 10 CFR Part 20 requirements. 

The dose consequences referred to in 10 CFR 50.59 are those calculated by 

licensees-not the results of independent, confirmatory dose analyses by the 

NRC that may be documented in safety evaluation reports. 

The evaluation should determine the dose that would likely result from 

accidents associated with the proposed activity. If a proposed activity would 

result in more than a minimal increase in dose from the existing calculated 

dose for any accident, then the activity would require prior NRC approval. 

Where a change in consequences is so small or the uncertainties in 

determining whether a change in consequences has occurred are such that it 

cannot be reasonably concluded that the consequences have actually changed 

(i.e., there is no clear trend toward increasing the consequences), the change 

need not be considered an increase in consequences.

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General Design Criterion 19 of Appendix A to 10 CFR 50 requires radiation 

protection to permit access and occupancy of the control room under accident 

conditions without personnel receiving radiation exposure in excess of 5 rem 

whole body, for the duration of the accident. 10 CFR 100 establishes 

requirements for exclusion area and low population zones around the reactor 

so that an individual located at any point on its boundary immediately 

following onset of the postulated fission product release would not receive a 

total radiation dose to the whole body in excess of 25 rem or a total radiation 

dose of 300 rem to the thyroid for iodine exposure. In the Standard Review 

Plan (SRP), NUREG-0800, the NRC established acceptance guidelines for 

certain events that are considered of greater likelihood than the limiting 

accidents. For example, for a steam generator tube rupture, the SRP 

acceptance guideline is that the dose be less than or equal to a small fraction 

(i.e., 10 percent) of the 10 CFR 100 thyroid dose value, or 30 rem. 

Therefore, for a given accident, calculated or bounding dose values for that 

accident would be identified in the UFSAR. These dose values should be 

within the GDC 19 or 10 CFR 100 limits, as applicable, as modified by SRP 

guidelines (e.g., small fraction of 10 CFR 100), as applicable. An increase in 

consequences from a proposed activity is defined to be no more than minimal 

if the increase (1) is less than or equal to 10 percent of the difference between 

the current calculated dose value and the regulatory guideline value (10 CFR 

100 or GDC 19, as applicable), and (2) the increased dose does not exceed the 

current SRP guideline value for the particular design basis event. The 

current calculated dose values are those documented in the most up-to-date 

analyses of record. This approach establishes the current SRP guideline 

values as a basis for minimal increases for all facilities, not just those that 

were specifically licensed against those guidelines 4. 

For some licensees the current calculated dose consequences may already be 

in excess of the SRP guidelines for some events. In such cases, minimal 

increase is defined as less than or equal to 0.1 rem. 

In determining if there is more than a minimal increase in consequences, the 

first step is to determine which accidents evaluated in the LUFSAR may have 

their radiological consequences affected as a direct result of the proposed 

activity. Examples of questions that assist in this determination are: 

(1) Will the proposed activity change, prevent or degrade the 

effectiveness of actions described or assumed in an accident 

discussed in the UFSAR? 

4 For licensees who adopt the alternative source term, evaluations against this criterion should be in 

terms of total effective dose equivalent and the limits established by 10 CFR 50.67 (effective 

January 24, 2000).

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(2) Will the proposed activity alter assumptions previously made in 

evaluating the radiological consequences of an accident described in 

the UFSAR? 

(3) Will the proposed activity play a direct role in mitigating the 

radiological consequences of an accident described in the UFSAR? 

The next step is to determine if the proposed activity does, in fact, increase 

the radiological consequences of any of the accidents evaluated in the 

UFSAR. If it is determined that the proposed activity does have an effect on 

the radiological consequences of any accident analysis described in the 

UFSAR, then either: 

(1) Demonstrate and document that the radiological consequences of the 

accident described in the UFSAR are bounding for the proposed 

activity (e.g., by showing that the results of the UFSAR analysis 

bound those that would be associated with the proposed activity), or 

(2) Revise and document the analysis taking into account the proposed 

activity and determine if more than a minimal increase has occurred 

as described above. 

The following examples illustrate the implementation of this criterion. In 

each example it is assumed that the calculated consequences do not include a 

change in the methodology for calculating the consequences. Changes in 

methodology would need to be separately considered under 10 CFR 

50.59(c)(2)(viii) as discussed in Section 4.3.8. 

Example 1 

The calculated fuel handling accident (FHA) dose is 50 rem to the thyroid at 

the exclusion area boundary. As a result of a proposed change, the calculated 

FHA dose would increase to 70 rem. Ten percent of the difference between 

the calculated value and the regulatory limit is 25 rem [ 10% of (300 rem- 50 

rem)]. The SRP acceptance guideline is 75 rem. Because the calculated 

increase is less than 25 rem and the total is less than the SRP guideline, the 

increase is not more than minimal. 

Example 2 

The calculated dose consequence for a particular steam generator tube 

rupture accident is 25 rem thyroid at the exclusion area boundary. As a 

result of a proposed change, the calculated dose consequence would increase 

to 29 rem thyroid. The increase is not more than minimal because the new

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calculated dose does not exceed the applicable SRP guideline of 30 rem 

thyroid, nor does the incremental change in consequences (4 rem) exceed 10 

percent of the difference between the previous calculated value and the 

regulatory limit of 300 rem thyroid. Ten percent of the difference between 

the regulatory limit (300 rem) and the calculated value (25 rem) is 27.5 rem 

(10% of 275). Since 4 rem is less than 27.5, this change does not cause more 

than a minimal increase in consequences. 

Example 3 

The calculated dose consequence of a fuel handling accident is 25 rem to the 

thyroid at the exclusion area boundary. Because of a proposed change, the 

calculated dose consequence would increase to 65 rem. The SRP guideline 

for this accident is 75 rem and is still met. The incremental increase in dose 

consequence (40 rem), however, exceeds 10 percent of the difference to the 

regulatory limit or 27.5 rem [ 10% of (300 rem - 25 rem)]. Therefore, the 

change results in more than a minimal increase in consequences and thus 

requires prior NRC approval. 

Example 4 

The calculated dose to the control room operators following a loss of coolant 

accident is 4 rem whole body. A change is proposed to the control room 

ventilation system such that the calculated dose would increase to 4.5 rem. 

The regulations dictate that the control room doses are to be controlled to less 

than 5 rem by General Design Criterion 19. Although the new calculated 

dose is less than the regulatory limits, the incremental increase in dose (0.5 

rem) exceeds the value of 10 percent of the difference between the previously 

calculated value and the regulatory value or 0.1 rem [10% of (5 rem - 4 rem)]. 

This change would require prior NRC review because the increase in 

consequences exceeds the minimal standard. 

Example 5 

The existing safety analysis for a fuel handling accident predicts an off-site 

dose to the thyroid of 77 rem. The SRP guideline for this event is 75 rem. A 

proposed change would result in an increase in the calculated dose from 77 to 

77.1 rem. In this case, the proposed change would not cause more than a 

minimal increase in consequences because the new calculated value, even 

though greater than the SRP value, is within the guideline limit of 0.1 rem.

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4.3.4 Does the Activity Result in More Than a Minimal Increase in the 

Consequences of a Malfunction? 

In determining if there is more than a minimal increase in consequences, the 

first step is to determine which malfunctions evaluated in the UFSAR have 

their radiological consequences affected as a result of the proposed activity. 

The next step is to determine if the proposed activity does, in fact, increase 

the radiological consequences and, if so, are they more than minimally 

increased. The guidance for determining whether a proposed activity results 

in more than a minimal increase in the consequences of a malfunction is the 

same as that for accidents. Refer to Section 4.3.3. 

4.3.5 Does the Activity Create a Possibility for an Accident of a Different 

Type? 

The set of accidents that a facility must postulate for purposes of UFSAR 

safety analyses, including LOCA, other pipe ruptures, rod ejection, etc., are 

often referred to as "design basis accidents." The terms accidents and 

transients are often used in regulatory documents (e.g., in Chapter 15 of the 

Standard Review Plan), where transients are viewed as the more likely, low 

consequence events and accidents as less likely but more serious. In the 

context of probabilistic risk assessment, transients are typically viewed as 

initiating events, and accidents as the sequences that result from various 

combinations of plant and safety system response. This criterion deals with 

creating the possibility for accidents of similar frequency and significance to 

those already included in the licensing basis for the facility. Thus, accidents 

that would require multiple independent failures or other circumstances in 

order to "be created" would not meet this criterion. 

Certain accidents are not discussed in the UFSAR because their effects are 

bounded by other related events that are analyzed. For example, a 

postulated pipe break in a small line may not be specifically evaluated in the 

UFSAR because it has been determined to be less limiting than a pipe break 

in a larger line in the same area. Therefore, if a proposed design change 

would introduce a small high energy line break into this area, postulated 

breaks in the smaller line need not be considered an accident of a different 

type. 

The possible accidents of a different type are limited to those that are as 

likely to happen as those previously evaluated in the UFSAR. The accident 

must be credible in the sense of having been created within the range of 

assumptions previously considered in the licensing basis (e.g., random single 

failure, loss of off-site power, etc.). A new initiator of an accident previously 

evaluated in the UFSAR is not a different type of accident. Such a change or

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activity, however, which increases the frequency of an accident previously 

thought to be incredible to the point where it becomes as likely as the 

accidents in the UFSAR, could create the possibility of an accident of a 

different type. For example, there are a number of scenarios, such as 

multiple steam generator tube ruptures, that have been analyzed 

extensively. However, these scenarios are of such low probability that they 

may not have been considered to be part of the design basis. However, if a 

change or activity is proposed such that a scenario such as a multiple steam 

generator tube rupture becomes credible, the change or activity could create 

the possibility of an accident of a different type. In some instances these 

example accidents could already be discussed in the UFSAR. 

In evaluating whether the proposed change or activity creates the possibility 

of an accident of a different type, the first step is to determine the types of 

accidents that have been evaluated in the UFSAR. Accidents of a different 

type are credible accidents that the proposed activity could create that are 

not bounded by UFSAR-evaluated accidents. 

4.3.6 Does the Activity Create a Possibility for a Malfunction of an SSC 

Important to Safety with a Different Result? 

Malfunctions of SSCs are generally postulated as potential single failures to 

evaluate plant performance with the focus being on the result of the 

malfunction rather than the cause or type of malfunction. A malfunction 

that involves an initiator or failure whose effects are not bounded by those 

explicitly described in the UFSAR is a malfunction with a different result. A 

new failure mechanism is not a malfunction with a different result if the 

result or effect is the same as, or is bounded by, that previously evaluated in 

the UFSAR. The following examples illustrate this point: 

"* If a pump is replaced with a new design, there may be a new failure 

mechanism introduced that would cause a failure of the pump to run. 

But if this effect (failure of the pump to run) was previously evaluated 

and bounded, then a malfunction with a different result has not been 

created. 

"* If a feedwater control system is being upgraded from an analog to a 

digital system, new components may be added that could fail in ways 

other than the components in the original design. Provided the end 

result of the component or subsystem failure is the same as, or is 

bounded by, the results of malfunctions currently described in the 

UFSAR (i.e., failure to maximum demand, failure to minimum 

demand, failure as-is, etc.), then this upgrade would not create a 

"malfunction with a different result."

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An example of a change that would create the possibility for a malfunction 

with a different result is a substantial modification or upgrade to control 

station alarms, controls, or displays that are associated with SSCs important 

to safety that creates a new or common cause failure that is not bounded by 

previous analyses or evaluations. 

Certain malfunctions are not explicitly described in the UFSAR because their 

effects are bounded by other malfunctions that are described. For example, 

failure of a lube oil pump to supply oil to a component may not be explicitly 

described because a failure of the supplied component to operate was 

described. 

The possible malfunctions with a different result are limited to those that are 

as likely to happen as those described in the UFSAR. For example, a seismic 

induced failure of a component that has been designed to the appropriate 

seismic criteria will not cause a malfunction with a different result. 

However, a proposed change or activity that increases the likelihood of a 

malfunction previously thought to be incredible to the point where it becomes 

as likely as the malfunctions assumed in the UFSAR could create a possible 

malfunction with a different result. 

In evaluating a proposed activity against this criterion, the types and results 

of failure modes of SSCs that have previously been evaluated in the UFSAR 

and that are affected by the proposed activity should be identified. This 

evaluation should be performed consistent with any failure modes and effects 

analysis (FMEA) described in the UFSAR, recognizing that certain proposed 

activities may require a new FMEA to be performed. Attention must be 

given to whether the malfunction was evaluated in the accident analyses at 

the component level or the overall system level. While the evaluation should 

take into account the level that was previously evaluated in terms of 

malfunctions and resulting event initiators or mitigation impacts, it also 

needs to consider the nature of the proposed activity. Thus, for instance, if 

failures were previously postulated on a train level because the trains were 

independent, a proposed activity that introduces a cross-tie or credible 

common mode failure (e.g., as a result of an analog to digital upgrade) should 

be evaluated further to see whether new outcomes have been introduced. 

Once the malfunctions previously evaluated in the UFSAR and the results of 

these malfunctions have been determined, then the types and results of 

failure modes that the proposed activity could create are identified. 

Comparing the two lists can provide the answer to the criterion question. An 

example that might create a malfunction with a different result could be the 

addition of a normally open vent line in the discharge of an emergency core 

cooling system pump. The different result of a malfunction could be potential 

voiding in the system causing it not to operate properly.

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4.3.7 Does the Activity Result in a Design Basis Limit for a Fission 

Product Barrier Being Exceeded or Altered? 

10 CFR 50.59 evaluation under criterion (c)(2)(vii) focuses on the fission 

product barriers-fuel cladding, reactor coolant system boundary and 

containment-and on the critical design information that supports their 

continued integrity. Guidance for applying this criterion is structured 

around a two-step approach: 

"* Identification of affected design basis limits for a fission product 

barrier 

"* Determination of when those limits are exceeded or altered. 

Identification of affected design basis limits for a fission product barrier 

The first step is to identify the fission product barrier design basis limits, if 

any, that are affected by a proposed activity. Design basis limits for a fission 

product barrier are the controlling numerical values established during the 

licensing review as presented in the UFSAR for any parameter(s) used to 

determine the integrity of the fission product barrier. These limits have 

three key attributes: 

* The parameter is fundamental to the barrier's integrity. Design 

basis limits for fission product barriers establish the reference bounds for 

design of the barriers, as defined in 10 CFR 50.2. They are the limiting 

values for parameters that directly determine the performance of a fission 

product barrier. That is, design bases limits are fundamental to barrier 

integrity and may be thought of as the point at which confidence in the 

barrier begins to decrease. 

For purposes of this evaluation, design bases parameters that are used to 

directly determine fission product barrier integrity should be distinguished 

from subordinate parameters that can indirectly affect fission product 

barrier performance. Indirect effects of changes to subordinate parameters 

are evaluated in terms of their effect on the more fundamental design bases 

parameters/limits that ensure fission product barrier integrity. For 

example, auxiliary feedwater design flow is a subordinate parameter for 

purposes of this evaluation, not a design bases parameter/limit. The 

acceptability of a reduction in AFW design flow would be determined based 

on its effect on design bases limits for the RCS (e.g., RCS pressure).

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m The limit is expressed numerically. Design basis limits are numerical 

values used in the overall design process, not descriptions of functional 

requirements. Design basis limits are typically the numerical event acceptance 

criteria used in the accident analysis methodology. The facility's design and 

operation associated with these parameters as described in the UFSAR will be 

at or below (more conservative than) the design basis limit. 

m The limit is identified in the UFSAR. As required by 10 CFR 50.34(b), 

design basis limits were presented in the original FSAR and continue to 

reside in the UFSAR. They may be located in a vendor topical report that 

is incorporated by reference in the UFSAR. 

Consistent with the discussion of 10 CFR 50.59 applicability in Section 4.1, 

any design basis limit for a fission product barrier that is controlled by 

another, more specific regulation or technical specification would not require 

evaluation under criterion (c)(2)vii. The effect of the proposed activity on 

those parameters would be evaluated in accordance with the more specific 

regulation. Effects (either direct or indirect-see discussion below) on design 

basis parameters covered by another regulation or technical specification 

need not be considered as part of evaluations under this criterion.

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Examples of typical fission product barrier 

following table:

design basis limits are identified in the

Barrier Design Bases Typical Design Basis Limit 

Parameter 

Value corresponding to the 95/95 

Fuel DNBR/MCPR DNB criterion for a given DNB 

Cladding correlation 

Fuel temperature Centerline fuel melting 

temperature 

Linear heat rate Peak linear heat rate (typ. in 

kW/ft) established to ensure clad 

integrity 

Fuel enthalpy Cal/gm associated with dispersion 

Clad strain Internal pressure associated with 

clad liftoff 

Fuel burnup Limit (typ. in MWd/ton) 

established to ensure clad 

integrity 

Clad temperature * 2200 degrees F 

Clad oxidation * 17% local and 1 % overall 

RCS Pressure Designated limit in safety analysis 

Boundary for specific accident 

Stresses * ASME Code compliance for 

normal, upset, faulted, etc., as 

appropriate for accident 

Heat-up/Cool-down* Applicable ASME Code stress 

limits 

Containment Pressure Containment design pressure 

* These parameters are commonly controlled by 10 CFR 50.55a, 10 CFR 50.46 

and/or a specific technical specification and therefore would not be subject to 

10 CFR 50.59.

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The list above may vary slightly for a given facility and/or fuel vendor and 

may include other parameters for specific accidents. For example, 

n PWR licensees may use 100% pressurizer level as a limiting parameter 

to ensure RCS integrity for some accident sequences. 

m A peak containment temperature may be established in the UFSAR as 

an independent limit for ensuring the integrity of the containment. 

If a given facility has these or other parameters incorporated into the UFSAR 

as a design basis limit for a fission product barrier, then changes affecting it 

should be evaluated under this criterion. 

Two of the ways that a licensee can evaluate proposed activities against this 

criterion are as follows. The licensee may identify all design bases 

parameters for fission product barriers and include them explicitly in the 

procedure for performing 10 CFR 50.59 evaluations. Alternatively, the 

effects of a proposed activity could be evaluated first to determine if the 

change affects design bases parameters for fission product barriers. The 

results of these two approaches are equivalent provided the guidance for 

"exceeded or altered" described below is followed. In all cases, the direct and 

indirect effects of proposed activities must be included in the evaluation. 

Exceeded or altered 

A specific proposed activity requires a license amendment if the design basis 

limit for a fission product barrier is "exceeded or altered." The term 

"exceeded" means that as a result of the proposed activity, the facility's 

predicted response would be less conservative than the numerical design 

basis limit identified above. The term "altered" means the design basis limit 

itself is changed. 

The effect of the proposed activity includes both direct and indirect effects. 

Extending the maximum fuel burn-up limits until the fuel rod internal gas 

pressure exceeds the design basis limit is a direct effect that would require a 

license amendment. As discussed earlier, indirect effects provide for another 

parameter or effect to cascade from the proposed activity to the design basis 

limit. For example, reducing the design flow of auxiliary feedwater pumps 

following a loss of main feedwater could reduce the heat transferred from the 

RCS to the steam generators. That effect could increase the RCS 

temperature, which would raise RCS pressure and pressurizer level. The 10 

CFR 50.59(c)(2)(vii) evaluation of this change would focus on whether the 

design basis limit associated with RCS pressure for that accident sequence 

would be exceeded.

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Altering a design basis limit for a fission product barrier is not a routine 

activity, but it can occur. An example of this would be changing the DNBR 

value from the value corresponding to the 95/95 criterion for a given DNB 

correlation, perhaps as a result of a new fuel design being implemented. (A 

new correlation or a new value for the "95/95 DNB criterion" with the same 

fuel type would be evaluated under criterion (c)(2)(viii) of the rule.) Another 

example is redesigning portions of the RCS boundary to no longer comply 

with the code of construction. These are infrequent activities affecting key 

elements of the defense-in-depth philosophy. As such, no distinction has 

been made between a conservative and nonconservative change in these 

limits. In contrast with these examples, altering AFW design flow, or other 

subordinate parameter/limit, is not subject to the "may not be altered" 

criterion because AFW design flow is not a design bases limit for fission 

product barrier integrity. 

Evaluations performed under this criterion may incorporate a number of 

refinements to simplify the review. For example, if an engineering 

evaluation demonstrates that no parameters are affected that have design 

basis limits for fission product barriers associated with them, no 10 CFR 

50.59(c)(2)(vii) evaluation is required. Similarly, most parameters that 

require evaluation under this criterion have calculations or analyses 

supporting the facility's design. If an engineering evaluation demonstrates 

that the analysis presented in the UFSAR remains bounding, then no 10 

CFR 50.59(c)(2)(vii) evaluation is required. When using these techniques, 

both indirect and direct effects must be considered to ensure that important 

interactions are not overlooked. 

Examples illustrating the two-step approach for evaluations under this 

criterion are provided below: 

Example 1 

It is proposed to delay the automatic start of the stand-by condensate booster 

pump to eliminate spurious automatic starts. The proposed change is of 

sufficient magnitude such that it "screens in" as affecting a UFSAR-described 

design function. 

Identification of design basis limits 

The direct effects of a reduction in condensate flow would be reviewed to 

identify potentially affected design basis parameters. In addition, the 

indirect effect on feedwater flow and feedwater pump NPSH of a possible 

transient reduction in condensate flow/pressure would be considered.

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Likewise, consideration of indirect effects would be extended to the reactor or 

steam generator (BIWR or PWR, as applicable). The review concludes that no 

design basis limits are either directly or indirectly affected. 

The change in the frequency of a reactor trip as a result of normal condensate 

system malfunctions would be evaluated under other 10 CFR 50.59 criteria. 

Exceeded or altered 

Since no design basis limits were identified, this element of the evaluation is 

not applicable. 

Example 2 

The heat transfer capability of an RHR heat exchanger tube bundle has 

degraded, and it is proposed to accept the condition "as-is." 

Identification of design basis limits 

The effects of the reduced heat transfer capability would be reviewed. The 

direct effect would include the increased temperature of the suppression pool 

or containment sump [BWR or PWR, as applicable]. The indirect effects 

would include increasing the peak containment post-accident pressure and 

increased enthalpy of ECCS flow. The increased ECCS enthalpy would also 

affect peak clad temperature (PCT). Thus, the proposed activity affects two 

design basis limits: containment pressure and PCT. In this example, the 

design basis limits would most likely serve as the acceptance criteria for the 

two parameters in the LOCA analysis described in the UFSAR. (Most 

licensees use containment design pressure and 2200 degrees F for those 

values.) 

Exceeded or altered 

Any increase in peak containment post-accident pressure would be compared 

to the design basis limit, in this case, containment design pressure. If the 

revised peak post-accident containment pressure exceeded the design basis 

limit, then a license amendment would be required. 

On the other hand, PCT is governed by a more specific regulation, 10 CFR 

50.46. Therefore, the evaluation under this criterion would not address the 

impact on this parameter. Rather, any changes or corrections to an 

acceptable evaluation model or application of such a model that affects the 

PCT calculation would be evaluated per the requirements of 10 CFR 

50.46(3)(ii).

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In this example, the design basis limit for containment pressure is not being 

altered. Therefore, this element of the review is not applicable. 

Example 3 

Recently identified corrosion inside the primary containment has prompted a 

re-evaluation of the existing containment design pressure of 55 psig. This re

evaluation has concluded that a design pressure of 48 psig is the maximum 

supportable. As the final resolution to the degraded containment condition, 

the licensee proposes to reduce the containment design pressure as reflected 

in UFSAR safety analyses from 55 to 48 psig. 

Identification of design basis limit 

The affected parameter is post-accident peak containment pressure. This 

parameter directly affects the containment barrier. Its design basis limit 

from the UFSAR is the existing containment design pressure of 55 psig. 

Exceeded or altered 

The design basis limit itself has been "altered" and thus a license amendment 

is required. The issue of conservative vs. nonconservative is not germane to 

requiring a submittal. That is, prior NRC approval is required regardless of 

direction because this is a fundamental change in the facility's design. 

4.3.8 Does the Activity Result in a Departure from a Method of Evaluation 

Described in the UFSAR Used in Establishing the Design Bases or in 

the Safety Analyses? 

The UFSAR contains design and licensing basis information for a nuclear 

power facility, including description on how regulatory requirements for 

design are met and how the facility responds to various design basis 

accidents and events. Analytical methods are a fundamental part of 

demonstrating how the design meets regulatory requirements and why the 

facility's response to accidents and events is acceptable. As such, in cases 

where the analytical methodology was considered to be an important part of 

the conclusion that the facility met the required design bases, these 

analytical methods were described in the UFSAR and received varying levels 

of NRC review and approval during licensing. 

Because 10 CFR 50.59 provides a process for determining if prior NRC 

approval is required before making changes to the facility as described in the 

UFSAR, changes to the methodologies described in the UFSAR also fall

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under the provisions of the 10 CFR 50.59 process, specifically criterion 

(c)(2)(viii). In general, licensees can make changes to elements of a 

methodology without first obtaining a license amendment if the results are 

essentially the same as, or more conservative than, previous results. 

Similarly, licensees can also use different methods without first obtaining a 

license amendment if those methods have been approved by the NRC for the 

intended application. 

If the proposed activity does not involve a change to a method of evaluation, 

then the 10 CFR 50.59 evaluation should reflect that this criterion is not 

applicable. If the activity involves only a change to a method of evaluation, 

then the 10 CFR 50.59 evaluation should reflect that criteria 10 CFR 

50.59(c)(2)(i-vii) are not applicable. 

The first step in applying this criterion is to identify the methods of 

evaluation that are affected by the change. This is accomplished during 

application of the screening criteria in Section 4.2.1.3. 

Next, the licensee must determine whether the change constitutes a 

departure from a method of evaluation that would require prior NRC 

approval. As discussed further below, for purposes of evaluations under this 

criterion, the following changes are considered a departure from a method of 

evaluation described in the UFSAR: 

. Changes to any element of analysis methodology that yield results that 

are nonconservative or not essentially the same as the results from the 

analyses of record 

. Use of new or different methods of evaluation that are not approved by 

NRC for the intended application. 

By way of contrast, the following changes are not considered departures 

from a method of evaluation described in the UFSAR: 

. Departures from methods of evaluation that are not described, 

outlined or summarized in the UFSAR (such changes may have been 

screened out as discussed in Section 4.2.1.3) 

. Use of a new NRC-approved methodology (e.g., new or upgraded 

computer code) to reduce uncertainty, provide more precise results or 

other reason, provided such use is (a) based on sound engineering 

practice, (b) appropriate for the intended application and (c) within the 

limitations of the applicable SER. The basis for this determination 

should be documented in the licensee evaluation

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Use of a methodology revision that is documented as providing results 

that are essentially the same as, or more conservative than, either the 

previous revision of the same methodology or another methodology 

previously accepted by NRC through issuance of an SER. 

Subsection 4.3.8.1 provides guidance for making changes to one or more 

elements of an existing method of evaluation used to establish the design 

bases or in the safety analyses. Subsection 4.3.8.2 provides guidance for 

adopting an entirely new method of evaluation to replace an existing one. 

Examples illustrating the implementation of this criterion are provided in 

Section 4.3.8.3. 

4.3.8.1 Guidance for Changing One or More Elements of a Method of 

Evaluation 

The definition of "departure ..." provides licensees with the flexibility to 

make changes under 10 CFR 50.59 to methods of evaluation whose results 

are "conservative" or that are not important with respect to the 

demonstrations of performance that the analyses provide. Changes to 

elements of analysis methods that yield conservative results, or results that 

are essentially the same, would not be departures from approved methods. 

Conservative vs. Nonconservative Results 

Gaining margin by changing one or more elements of a method of evaluation 

is considered to be a nonconservative change and thus a departure from a 

method of evaluation for purposes of 10 CFR 50.59. Such departures require 

prior NRC approval of the revised method. Analytical results obtained by 

changing any element of a method are "conservative" relative to the previous 

results, if they are closer to design bases limits or safety analyses limits (e.g., 

applicable acceptance guidelines). For example, a change from 45 psig to 48 

psig in the result of a containment peak pressure analysis (with design basis 

limit of 50 psig) using a revised method of evaluation would be considered a 

conservative change when applying this criterion. In other words, the 

revised method is more conservative if it predicts more severe conditions 

given the same set of inputs. This is because results closer to limiting values 

are considered conservative in the sense that the new analysis result 

provides less margin to applicable limits for making potential physical or 

procedure changes without a license amendment. 

In contrast, if the use of a modified method of evaluation resulted in a change 

in calculated containment peak pressure from 45 psig to 40 psig, this would 

be a nonconservative change. That is because the change would result in

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more margin being available (to the design basis limit of 50 psig) for the 

licensee to make more significant changes to the physical facility or 

procedures. 

"Essentially the Same" 

Licensees may change one or more elements of a method of evaluation such 

that results move in the nonconservative direction without prior NRC 

approval, provided the revised result is "essentially the same" as the previous 

result. Results are "essentially the same" if they are within the margin of 

error for the type of analysis being performed. Variation in results due to 

routine analysis sensitivities or calculational differences (e.g., rounding 

errors and use of different computational platforms) would typically be 

within the analysis margin of error and thus considered "essentially the 

same." For example, when a method is applied using a different 

computational platform (mainframe vs. workstation), results of cases run on 

the two platforms differed by less than 1%, which is the margin of error for 

this type of calculation. Thus the results are essentially the same, and do not 

constitute a departure from a method that requires prior NRC approval. 

The determination of whether a new analysis result would be considered 

"essentially the same" as the previous result can be made through 

benchmarking the revised method to the existing one, or may be apparent 

from the nature of the differences between the methods. When 

benchmarking a revised method to determine how it compares to the 

previous one, the analyses that are done must be for the same set of plant 

conditions to ensure that the results are comparable. Comparison of analysis 

methods should consider both the peak values and time behavior of results, 

and engineering judgment should be applied in determining whether two 

methods yield results that are essentially the same. 

4.3.8.2 Guidance for Changing from One Method of Evaluation to 

Another 

The definition of "departure..." provides licensees with the flexibility to make 

changes under 10 CFR 50.59 from one method of evaluation to another 

provided that the new method is approved by the NRC for the intended 

application. A new method is approved by the NRC for intended application 

if it is approved for the type of analysis being conducted, and applicable 

terms, conditions and limitations for its use are satisfied. 

NRC approval has typically followed one of two paths. Most reactor or fuel 

vendors and several utilities have prepared and obtained NRC approval of 

topical reports that describe methodologies for the performance of a given

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type or class of analysis. Through a safety evaluation report (SER), NRC 

approved the use of the methodologies for a given class of power plants. In 

some cases, the NRC has accorded "generic" approval of analysis 

methodologies. Terms, conditions and limitations relating to the application 

of the methodologies are usually documented in the topical reports, the SER, 

and correspondence between the NRC and the methodology owner that is 

referenced in the SER or associated transmittal letter. 

The second path is the approval of a specific analysis rather than a more 

generic methodology. In these cases, the NRC's approval has typically been 

part of a plant's licensing basis and limited to a given plant design and a 

given application. Again, a thorough understanding of the terms, conditions 

and limitations relating to the application of the methodology is essential. 

This information is usually documented in the original license application or 

license amendment request, the SER, and any correspondence between the 

NRC and the analysis owner that is referenced in the SER or associated 

transmittal letter. 

It is incumbent upon the users of a new methodology-even one generically 

approved by the NRC-to ensure they have a thorough understanding of the 

methodology in question, the terms of its existing application and 

conditions/limitations on its use. A range of considerations is identified below 

that may be applicable to determining whether new methods are technically 

appropriate for the intended application. The licensee should address these 

and similar considerations, as applicable, and document in the 10 CFR 50.59 

evaluation the basis for determining that a method is appropriate and 

approved for the intended application. To obtain an adequate understanding 

of the method and basis for determining it is approved for use in the intended 

application, licensees should consult various sources, as appropriate. These 

include SERs, topical reports, licensee correspondence with the NRC and 

licensee personnel familiar with the existing application of the method. If 

adequate information cannot be found on which to base the intended 

application of the methodology, the method should not be considered "approved 

by the NRC for the intended application." 

The applicable terms and conditions for the use of a methodology are not 

limited to a specific analysis; the qualification of the organization applying the 

methodology is also a consideration. Through Generic Letter 83-11, 

Supplement 1,5 the NRC has established a method by which licensees can 

demonstrate they are generally qualified to perform safety analyses. Licensees 

thus qualified can apply methods that have been reviewed and approved by 

the NRC, or that have been otherwise accepted as part of another plant's 

5 Generic Letter 83-11, Supplement 1, "Licensee Qualification for Performing Safety -Analyses," June 24, 

1999

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licensing basis, without requiring prior NRC approval. Licensees that have 

not satisfied the guidelines of Generic Letter 83-11, Supplement 1, may, of 

course, continue to seek plant-specific approval to use new methods of 

evaluation. 

When considering the application of a methodology, it is necessary to adopt 

the methodology en toto and apply it consistent with applicable terms, 

conditions and limitations. Mixing attributes of new and existing 

methodologies is considered a revision to a methodology and must be 

evaluated as such per the guidance in Section 4.3.8.1. 

Considerations for Determining if New Methods May be Considered "Approved 

by the NRC for the Intended Application" 

The following questions highlight important considerations for determining 

that a particular application of a different method is technically appropriate 

for the intended application, within the bounds of what has been found 

acceptable by NRC, and does not require prior NRC approval. 

"* Is the application of the methodology consistent with the facility's 

licensing basis (e.g., NUREG-0800 or other plant-specific commitments)? 

Will the methodology supersede a methodology addressed by other 

regulations such as 10 CFR 50.46, 10 CFR 50.55a or the plant technical 

specifications (Core Operating Limits Report or Pressure/Temperature 

Limits Report)? Is the methodology consistent with relevant industry 

standards? 

If application of the new methodology requires exemptions from 

regulations or plant-specific commitments, exceptions to relevant industry 

standards and guidelines, or is otherwise inconsistent with a facility's 

licensing basis, then prior NRC approval may be required. The applicable 

change process must be followed to make the plant's licensing basis 

consistent with the requirements of the new methodology. 

"* If a computer code is involved, has the code been installed in accordance 

with applicable software quality assurance requirements? Has the plant

specific model been adequately qualified through benchmark comparisons 

against test data, plant data or approved engineering analyses? Is the 

application consistent with the capabilities and limitations of the 

computer code? Has industry experience with the computer code been 

appropriately considered? 

The computer code installation and plant-specific model qualification are 

not directly transferable from one organization to another. The

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installation and qualification should be in accordance with the licensee's 

quality assurance program. 

"* Is the facility for which the methodology has been approved designed and 

operated in the same manner as the facility to which the methodology is 

to be applied? Is the relevant equipment the same? Does the equipment 

have the same pedigree (e.g., Class 1E, Seismic Category I, etc.)? Are the 

relevant failure modes and effects analyses the same? If the plant is 

designed and operated in a similar, but not identical, manner, the 

following types of considerations should be addressed to assess the 

applicability of the methodology: 

"* How could those differences affect the methodology? 

"* Are additional sensitivity studies required? 

"* Should additional single failure scenarios be considered? 

"* Are analyses of limiting scenarios, effects of equipment failures, 

etc., applicable for the specific plant design? 

"* Can analyses be made while maintaining compliance with both the 

intent and literal definition of the methodology? 

"* Differences in the plant configurations and licensing bases could 

invalidate the application of a particular methodology. For example, the 

licensing basis of older vintage plants may not include an analysis of the 

feedwater line break event that is required in later vintage plants. Some 

plants may be required to postulate a loss of off-site power or a maximum 

break size for certain events; others may have obtained exemptions to 

these requirements from the NRC. Some plants may have pressurizer 

power-operated relief valves that are qualified for water relief; other 

plants do not. Plant specific failure modes and effects analyses may 

reveal new potential single failure scenarios that can not be adequately 

assessed with the original methodology. The existence of these 

differences does not preclude application of a new methodology to a 

facility; however, differences must be identified, understood and the basis 

documented for concluding that the differences are not relevant to 

determining that the new application is technically appropriate.

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4.3.8.3 EXAMPLES 

The following examples illustrate the implementation of this criterion: 

Example 1 

The UFSAR states that a damping value of 0.5 percent is used in the seismic 

analysis of safety-related piping. The licensee wishes to change this value to 

2 percent to reanalyze the seismic loads for the piping. Using a higher 

damping value to represent the response of the piping to the acceleration 

from the postulated earthquake in the analysis would result in lower 

calculated stresses because the increased damping reduces the loads. Since 

this analysis was used in establishing the seismic design bases for the piping, 

and since this is a change to an element of the method that is not 

conservative and is not essentially the same, this change would require prior 

NRC approval under this criterion. 

On the other hand, had NRC approved an alternate method of seismic 

analysis that allowed 2 percent damping provided certain other assumptions 

were made, and the licensee used the complete set of assumptions to perform 

its analysis, then the 2 percent damping under these circumstances would 

not be a departure because this method of evaluation is considered "approved 

by the NRC for the intended application." 

Example 2 

A facility has a design basis containment pressure limit of 50 psig. The 

current worst-case design basis accident calculation results in a peak 

pressure of 45 psig within two minutes. The licensee revises the method of 

evaluation, and the recalculated result is 40 psig. This change would require 

prior NRC approval because the result of the recalculation is not 

conservative. If the licensee used a different method that was approved by 

the NRC and met all the terms and conditions of the method, a recalculated 

result of 40 psig would not require prior NRC approval. 

Example 3 

A licensee revises the seismic analysis described in the UFSAR to include an 

inelastic analysis procedure. This revised method is used to demonstrate 

that cable trays have greater capacity than previously calculated. This 

change would require prior NRC approval as it would not produce results 

that are essentially the same.

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Example 4 

Licensee X has received NRC approval for the use of a method of evaluation 

at Facility A for performing steamline break mass and energy release 

calculations for environmental qualification evaluations. The terms and 

conditions for the use of the method are detailed in the NRC SER. The SER 

also describes limitations associated with the method. Licensee Y wants to 

apply the method at its Facility B. Licensee Y has satisfied the guidelines of 

GL 83-11, Supplement 1. After reviewing the method, approved application, 

SER and related documentation, to verify that applicable terms, conditions 

and limitations are met and to ensure the method is applicable to their type 

of plant, Licensee Y conducts a 10 CFR 50.59 evaluation. Licensee Y 

concludes that the change is not a departure from a method of evaluation 

because it has determined the method is appropriate for the intended 

application, the terms and conditions for its use as specified in the SER have 

been satisfied, and the method has been approved by the NRC. 

Example 5 

The NRC has approved the use of computer code and the associated analysis 

of a steamline break for use in the evaluation of component stresses. A 

licensee uses the same computer code and analysis methodology to replace its 

evaluation of the containment temperature response. This change would 

require prior NRC approval unless the methodology had been previously 

approved for evaluating containment temperature response. 

4.4 APPLYING 10 CFR 50.59 To COMPENSATORY ACTIONS To ADDRESS 

NONCONFORMING OR DEGRADED CONDITIONS 

Three general courses of action are available to licensees to address non

conforming and degraded conditions. Whether or not 10 CFR 50.59 must be 

applied, and the focus of a 10 CFR 50.59 evaluation if one is required, depends 

on the corrective action plan chosen by the licensee, as discussed below: 

m If the licensee intends to restore the SSC back to its as-designed condition 

then this corrective action should be performed in accordance with 10 

CFR 50, Appendix B (i.e., in a timely manner commensurate with safety). 

This activity is not subject to 10 CFR 50.59. 

m If an interim compensatory action is taken to address the condition and 

involves a temporary procedure or facility change, 10 CFR 50.59 should be 

applied to the temporary change. The intent is to determine whether the 

temporary change/compensatory action itself (not the degraded condition) 

impacts other aspects of the facility or procedures described in the

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UFSAR. In considering whether a temporary change impacts other 

aspects of the facility, a licensee should pay particular attention to 

ancillary aspects of the temporary change that result from actions taken 

to directly compensate for the degraded condition. 

a If the licensee corrective action is either to accept the condition "as-is" 

resulting in something different than its as-designed condition, or to 

change the facility or procedures, 10 CFR 50.59 should be applied to the 

corrective action, unless another regulation applies, e.g., 10 CFR 50.55a. 

In these cases, the final corrective action becomes the proposed change 

that would be subject to 10 CFR 50.59. 

In resolving degraded or nonconforming conditions, the need to obtain NRC 

approval for a proposed activity does not affect the licensee's authority to 

operate the plant. The licensee may make mode changes, restart from 

outages, etc., provided that necessary SSCs are operable and the degraded 

condition is not in conflict with the technical specifications or the license. 

The following example illustrates the process for implementing a temporary 

change as a compensatory action to address a degraded/nonconforming 

condition: 

A level transmitter for one Reactor Coolant Pump (RCP) lower oil reservoir 

failed while at power. The transmitter provides an alarm function, but not an 

automatic protective action function. The transmitter and associated alarm 

are described in the UFSAR, as protective features for the RCPs, but no 

technical specification applies. Loss of the transmitter does not result in the 

loss of operability for any technical specification equipment. The transmitter 

fails in a direction resulting in a continuous alarm in the control room. The 

alarm circuitry provides a common alarm for both the upper and lower oil 

reservoir circuits, so transmitter failure causes a hanging alarm and a 

masking of proper operation of the remaining functional transmitter. 

Precautionary measures are taken to monitor lower reservoir oil level as 

outlined in the alarm manual using available alternate means. An interim 

compensatory action is proposed to lift the leads (temporary change) from the 

failed transmitter to restore the alarm function for the remaining functioning 

transmitter. 

Lifting the leads is a compensatory action (temporary change) that is subject 

to 10 CFR 50.59. The 10 CFR 50.59 screening would be applied to the 

temporary change itself (lifted leads), not the degraded condition (failed 

transmitter), to determine its impact on other aspects of the facility described 

in the UFSAR. If screening determines that no other UFSAR-described SSCs 

would be affected by this compensatory action, the temporary change would 

screen out, i.e., not require a 10 CFR 50.59 evaluation.

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4.5 DISPOSITION OF 10 CFR 50.59 EVALUATIONS 

There are two possible conclusions to a 10 CFR 50.59 evaluation: 

(1) The proposed activity may be implemented without prior NRC approval. 

(2) The proposed activity requires prior NRC approval. 

Where an activity requires prior NRC approval, the activity must be 

approved by the NRC via license amendment in accordance with 10 CFR 

50.90 prior to implementation. An activity is considered "implemented" when 

it provides its intended function, that is, when it is placed in service and 

declared operable. Thus, a licensee may design, plan, install and test a 

modification prior to receiving the license amendment to the extent that 

these preliminary activities do not themselves require prior NRC approval 

under 10 CFR 50.59. 

For example, a modification to a facility involved the replacement of a train 

of a safety system with one including diverse primary components (diesel

driven pump vice a motor-driven pump). The installation of the replacement 

train was largely in a new, separate structure. Ultimately the modification 

would require NRC approval because of impacts on the technical 

specifications as well as due to differences in reliability of the replacement 

pump in some situations. There was insufficient time to seek and gain NRC 

approval prior to construction. The licensee prepared a 10 CFR 50.59 

screening to support construction of the separate structure through 

preliminary testing. The limited interfaces with the existing facility were 

assessed and determined to not change the facility or procedures as described 

in the UFSAR. Upon receipt of the license amendment the final tie-in, 

testing and operation were fully authorized. 10 CFR 50.59 should be applied 

to any aspects of the activity not adequately addressed in the license 

amendment request and/or associated safety evaluation report. 

For proposed activities that are determined to require prior NRC approval, 

there are three possible options: 

(1) Cancel the planned activity. 

(2) Redesign the proposed activity so that the it may proceed without prior 

NRC approval. 

(3) Apply for and obtain a license amendment under 10 CFR 50.90 prior to 

implementing the activity. Technical and licensing evaluations 

performed for such activities may be used as part of the basis for 

license amendment requests.

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It is important to remember that determining that a proposed activity 

requires prior NRC approval does not determine whether it is safe. In fact, a 

proposed activity that requires prior NRC approval may significantly 

enhance overall plant safety at the expense of a small adverse impact in a 

specific area. It is the responsibility of the utility to assure that proposed 

activities are safe, and it is the role of the NRC to confirm the safety of those 

activities that are determined to require prior NRC review. 

5.0 DOCUMENTATION AND REPORTING 

10 CFR 50.59(d) requires the following documentation and recordkeeping: 

The licensee shall maintain records of changes in the facility, of changes in 

procedures, and of tests and experiments made pursuant to paragraph (c) of 

this section. These records must include a written evaluation that provides 

the bases for the determination that the change, test or experiment does not 

require a license amendment pursuant to paragraph (c)(2) of this section. 

(1) The licensee shall submit, as specified in § 50.4, a report containing a 

brief description of any changes, tests and experiments, including a 

summary of the evaluation of each. A report must be submitted at 

intervals not to exceed 24 months. 

(2) The records of changes in the facility must be maintained until the 

termination of a license issued pursuant to this part or the termination of 

a license issued pursuant to 10 CFR Part 54, whichever is later. Records 

of changes in procedures and records of tests and experiments must be 

maintained for a period of 5 years. 

The documentation and reporting requirements of 10 CFR 50.59(d) apply to 

activities that require evaluation against the eight criteria of 10 CFR 

50.59(c)(2) and are determined not to require prior NRC approval. That is, 

the phrase in 10 CFR 50.59(d)(1), "made pursuant to paragraph (c)," refers to 

those activities that were evaluated against the eight evaluation criteria 

(because, for example, they affect the facility as described in the UTFSAR), but 

not to those activities or changes that were screened out. Similarly, 

documentation and reporting under 10 CFR 50.59 is not required for 

activities that are canceled or that that are determined to require prior NRC 

approval and are implemented via the license amendment request process. 

Documenting 10 CFR 50.59 Evaluations 

In performing a 10 CFR 50.59 evaluation of a proposed activity, the evaluator 

must address the eight criteria in 10 CFR 50.59(c)(2) to determine if prior

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NRC approval is required. Although the conclusion in each criterion may be 

simply "yes," "no" or "not applicable," there must be an accompanying 

explanation providing adequate basis for the conclusion. Consistent with the 

intent of 10 CFR 50.59, these explanations should be complete in the sense 

that another knowledgeable reviewer could draw the same conclusion. 

Restatement of the criteria in a negative sense or making simple statements 

of conclusion is not sufficient and should be avoided. It is recognized, 

however, that for certain very simple activities, a statement of the conclusion 

with identification of references consulted to support the conclusion would be 

adequate and the 10 CFR 50.59 evaluation could be very brief. 

The importance of the documentation is emphasized by the fact that 

experience and engineering knowledge (other than models and experimental 

data) are often relied upon in determining whether evaluation criteria are 

met. Thus the basis for the engineering judgment and the logic used in the 

determination should be documented to the extent practicable and to a 

degree commensurate with the safety significance and complexity of the 

activity. This type of documentation is of particular importance in areas 

where no established consensus methods are available, such as for software 

reliability, or the use of commercial-grade hardware and software where full 

documentation of the design process is not available. 

Since an important goal of the 10 CFR 50.59 evaluation is completeness, the 

items considered by the evaluator must be clearly stated. 

Each 10 CFR 50.59 evaluation is unique. Although each applicable criterion 

must be addressed, the questions and considerations listed throughout this 

guidance document to assist evaluating the criteria are not requirements for 

all evaluations. Some evaluations may require that none of these questions 

be addressed while others will require additional considerations beyond those 

identified in this guidance. 

When preparing 10 CFR 50.59 evaluations, licensees may combine responses 

to individual criteria or reference other portions of the evaluation. 

As discussed in Section 4.2.3, licensees may elect to use screening criteria to 

limit the number of activities for which written 10 CFR 50.59 evaluations are 

performed. A documentation basis should be maintained for determinations 

that the changes meet the screening criteria, i.e., screen out. This 

documentation does not constitute the record of changes required by 10 CFR 

50.59, and thus is not subject to the recordkeeping requirements of the rule.

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Reporting to NRC 

A summary of 10 CFR 50.59 evaluations for activities implemented under 10 

CFR 50.59 must be provided to NRC. Activities that were screened out, 

canceled or implemented via license amendment need not be included in this 

report. The 10 CFR 50.59 reporting requirement (every 24 months) is 

identical to that for UFSAR updates such that licensees may provide these 

reports to NRC on the same schedule.

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APPENDIX A 

TEXT OF 10 CFR 50.59 

§ 50.59 Changes, tests, and experiments. 

(a) Definitions for the purposes of this section: 

(1) Change means a modification or addition to, or removal from, the facility 

or procedures that affects a design function, method of performing or controlling the 

function, or an evaluation that demonstrates that intended functions will be 

accomplished. 

(2) Departure from a method of evaluation described in the FSAR (as 

updated) used in establishing the design bases or in the safety analyses means (i) 

changing any of the elements of the method described in the FSAR (as updated) 

unless the results of the analysis are conservative or essentially the same; or (ii) 

changing from a method described in the FSAR to another method unless that 

method has been approved by NRC for the intended application. 

(3) Facility as described in the final safety analysis report (as updated) 

means: 

(i) The structures, systems, and components (SSC) that are described 

in the final safety analysis report (FSAR) (as updated), 

(ii) The design and performance requirements for such SSCs described 

in the FSAR (as updated), and 

(ini) The evaluations or methods of evaluation included in the FSAR 

(as updated) for such SSCs which demonstrate that their intended 

function(s) will be accomplished. 

(4) Final Safety Analysis Report (as updated) means the Final Safety 

Analysis Report (or Final Hazards Summary Report) submitted in accordance with 

§ 50.34, as amended and supplemented, and as updated per the requirements of § 

50.71(e) or § 50.71(f), as applicable. 

(5) Procedures as described in the final safety analysis report (as updated) 

means those procedures that contain information described in the FSAR (as 

updated) such as how structures, systems, and components are operated and 

controlled (including assumed operator actions and response times). 

(6) Tests or experiments not described in the final safety analysis report (as 

updated) means any activity where any structure, system, or component is utilized 

or controlled in a manner which is either: 

(i) Outside the reference bounds of the design bases as described in the 

final safety analysis report (as updated) or 

(ii) Inconsistent with the analyses or descriptions in the final safety 

analysis report (as updated).

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(b) Applicability. This section applies to each holder of a license authorizing 

operation of a production or utilization facility, including the holder of a license 

authorizing operation of a nuclear power reactor that has submitted the 

certification of permanent cessation of operations required under § 50.82(a)(1) or a 

reactor licensee whose license has been amended to allow possession but not 

operation of the facility. 

(c)(1) A licensee may make changes in the facility as described in the final safety 

analysis report (as updated), make changes in the procedures as described in the 

final safety analysis report (as updated), and conduct tests or experiments not 

described in the final safety analysis report (as updated) without obtaining a 

license amendment pursuant to § 50.90 only if: 

(i) A change to the technical specifications incorporated in the license 

is not required, and 

(ii) The change, test, or experiment does not meet any of the criteria in 

paragraph (c)(2) of this section. 

(2) A licensee shall obtain a license amendment pursuant to § 50.90 prior to 

implementing a proposed change, test, or experiment if the change, test, or 

experiment would: 

(i) Result in more than a minimal increase in the frequency of 

occurrence of an accident previously evaluated in the final safety 

analysis report (as updated); 

(ii) Result in more than a minimal increase in the likelihood of 

occurrence of a malfunction of a structure, system, or component (SSC) 

important to safety previously evaluated in the final safety analysis 

report (as updated); 

(iii) Result in more than a minimal increase in the consequences of an 

accident previously evaluated in the final safety analysis report (as 

updated); 

(iv) Result in more than a minimal increase in the consequences of a 

malfunction of an SSC important to safety previously evaluated in the 

final safety analysis report (as updated); 

(v) Create a possibility for an accident of a different type than any 

previously evaluated in the final safety analysis report (as updated); 

(vi) Create a possibility for a malfunction of an SSC important to 

safety with a different result than any previously evaluated in the 

final safety analysis report (as updated); 

(vii)Result in a design basis limit for a fission product barrier as 

described in the FSAR (as updated) being exceeded or altered; or 

(viii) Result in a departure from a method of evaluation described in 

the FSAR (as updated) used in establishing the design bases or in the 

safety analyses

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(3) In implementing this paragraph, the FSAR (as updated) is considered to 

include FSAR changes resulting from evaluations performed pursuant to this 

section and analyses performed pursuant to § 50.90 since submittal of the last 

update of the final safety analysis report pursuant to § 50.71 of this part. 

(4) The provisions in this section do not apply to changes to the facility or 

procedures when the applicable regulations establish more specific criteria for 

accomplishing such changes. 

(d)(1)The licensee shall maintain records of changes in the facility, of changes in 

procedures, and of tests and experiments made pursuant to paragraph (c) of this 

section. These records must include a written evaluation which provides the bases 

for the determination that the change, test or experiment does not require a license 

amendment pursuant to paragraph (c)(2) of this section. 

(2) The licensee shall submit, as specified in § 50.4, a report containing a 

brief description of any changes, tests, and experiments, including a summary of 

the evaluation of each. A report must be submitted at intervals not to exceed 24 

months. 

(3) The records of changes in the facility must be maintained until the 

termination of a license issued pursuant to this part or the termination of a license 

issued pursuant to 10 CFR Part 54, whichever is later. Records of changes in 

procedures and records of tests and experiments must be maintained for a period of 

5 years.

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APPENDIX B 

GUIDELINES FOR 10 CFR 72.48 IMPLEMENTATION 

(FUTURE) 

Appendix B is being developed separately to provide guidance and examples for 

applying 10 CFR 72.48 to changes involving independent spent fuel storage 

installations and spent fuel storage cask designs that is analogous to that for 10 

CFR 50.59. This appendix will be the subject of a separate NRC regulatory guide.

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