New Reactors:

Striving for

Enhanced Safety
An NRC new reactor vendor inspector observes ultrasonic testing 

inside a steam generator.

An NRC inspector reviews a shipment of materials at the Vogtle site 

in Georgia.

NRC inspectors visit Japan Steel Works to review quality controls for 

important plant components.

Front cover, left: NRC geologists and inspectors examine the 

excavation of the V.C. Summer Unit 2 foundation in South Carolina.

Front cover, right: Work in progress on the “nuclear island” at the 

Vogtle Unit 3 site in Georgia. (Courtesy: Southern Company, Inc.)

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information collections were approved by the Office of 

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Contents

Introduction ........................................................... 1

Safety goals for nuclear power plants................... 1

Fundamental policies for new reactors................. 2

New reactor specifics............................................. 4

Implementing the policies..................................... 6

Conclusion............................................................. 8

References ............................................................. 9

Introduction

The U.S. Nuclear Regulatory Commission 

(NRC) was created by Congress in 1974 as 

an independent agency. The NRC regulates 

the Nation’s civilian use of nuclear materials, 

including nuclear power plants, to ensure adequate 

protection of public health and safety, to promote 

the common defense and security, and to protect 

the environment.

Hope Creek Nuclear Power Plant in New Jersey

All 104 operating nuclear power plants in the 

United States were subject to a rigorous licensing 

process, described in Title 10 of the Code of 

Federal Regulations (10 CFR), Part 50, 

“Domestic Licensing of Production and Utilization 

Facilities.” The NRC approved construction of 

these plants between 1964 and 1978 and granted 

the most recent operating license in 1996. These 

plants continue to operate safely today and are 

constantly monitored through inspections and 

licensing reviews.

Following the licensing of these reactors, the 

NRC implemented a strategy for licensing 

the next generation of nuclear power plants, 

linking the foundation of safety provided by the 

NRC’s regulations with the industry’s desire for 

continuous improvement. This brochure describes 

key NRC policies, rules, and guidance that 

contribute to enhanced safety for new reactors 

being proposed today.

Safety goals for nuclear 

power plants

In 1986, the Commission published its policy 

statement on safety goals for the operation of all 

nuclear power plants. This statement, developed 

with extensive public input following the 1979 

accident at Three Mile Island (TMI), established 

goals to define an acceptable level of radiological 

risk for nuclear power plants.

The policy statement provides two high-level, 

qualitative goals:

Individual members of the public should 

be provided a level of protection from 

the consequences of nuclear power plant 

operation such that individuals bear no 

significant additional risk to life and health.

Societal risks to life and health from nuclear 

power plant operation should be comparable 

to or less than the risks of generating 

electricity by viable competing technologies 

and should not be a significant addition to 

other societal risks.
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In addition, the statement translates these goals 

into quantitative objectives related to the risk of 

accidental or cancer-related death.

As noted in the discussion supporting the policy 

statement, these goals represent a low level of risk 

to public health and safety that the industry should 

strive to meet. 

Fundamental policies for 

new reactors

In the 1980s, the NRC also began to develop 

its strategy for licensing the next generation of 

reactors. This strategy benefited from years of 

experience with the operating reactors, insights 

gained from detailed assessments of those plants, 

and research results.

Through a series of policy statements and 

interactions with staff, the Commission 

developed fundamental policies for new reactors 

that considered this wide experience base and 

improved technology.

Protection from severe accidents

Nuclear power plant applicants must demonstrate 

that their facility can withstand a broad range of 

abnormal conditions and accidents—“design-basis 

accidents”—without releasing harmful amounts 

of radioactive material. More serious accidents, 

termed “severe accidents,” are expected to occur 

much less frequently, but they could have more 

significant consequences. In the 1985 policy 

statement on severe accidents, the Commission 

presented its policy for reducing the likelihood and 

mitigating the consequences of these types 

of accidents.

Owners of operating nuclear power plants had 

made numerous improvements to the design and 

operation of their plants to address severe accident 

issues. Many of these updates were identified 

during followup to the TMI accident or through 

ongoing evaluation of plant operations. At the 

time the policy statement was issued, therefore, 

the Commission felt confident it did not need to 

require further generic actions for operating 

power plants.

Control room response to the TMI accident

For new plants, the Commission had more to say 

in the policy statement:

...this should not be viewed as implying a 

Commission policy that safety improvements 

in new plant designs should not be actively 

sought. The Commission fully expects 

that vendors engaged in designing new 

standard (or custom) plants will achieve a 

higher standard of severe accident safety 

performance than their prior designs.

Specifically, the Commission encouraged vendors 

to design plants that were simpler and gave 

operators more time to react to failures or 

unexpected conditions. 

In addition, the severe accident policy statement 

established specific criteria for new designs. 

Applicants submitting new designs for NRC 

approval could address severe accidents acceptably 

if they did the following:

• complied with the current regulations, 

including those added in response to the 

TMI accident
3

• resolved specific safety issues, with a 

special focus on reliability of decay heat 

removal and electrical power

• completed a probabilistic risk assessment 

(PRA)1 and considered severe 

accident vulnerabilities

The phases or “levels” of a PRA

In completing these steps, applicants would 

consider a range of alternatives in addressing 

safety issues and reducing risk from severe 

accidents. The staff would reach a conclusion on 

the acceptability of the design through a review 

of the applicant’s traditional engineering analysis, 

complemented by insights from the PRA.

1

 PRAs estimate risk using calculations related to what can 

go wrong, how likely that occurrence is, and what the 

consequences would be. In the nuclear industry, PRAs 

are often used to understand strengths and weaknesses of 

the design and operation of a nuclear power plant.

Advanced reactor designs

In 1986, the Commission issued its policy 

statement on the regulation of advanced nuclear 

power plants. In keeping with the “advanced” 

designation, the Commission stated its 

expectation that these reactors “would provide 

more margin prior to exceeding safety limits and/

or utilize simplified, inherent, passive, or other 

innovative means to reliably accomplish their 

safety functions.”

The policy statement went on to list attributes that 

should be considered in advanced designs, such as 

the following:

• simplified safety systems that would need 

fewer operator actions

• reliable equipment that would reduce the 

need to activate safety systems

• easily maintainable equipment that would 

expose plant personnel to less radiation

Nuclear power plant workers wearing protective clothing 

to reduce radioactive contamination
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Which kind of reactor?

The different terms for types of reactors can be 

confusing. New reactors are often considered to be 

any reactors envisioned in addition to the fleet of 

104 currently operating reactors.

The current fleet consists mainly of large reactors 

that use regular water (“light” water, as opposed to 

“heavy” water that has a different type of hydrogen 

than commonly found in nature) for both cooling 

the core and facilitating the nuclear reaction. These 

reactors are called large light-water reactors.

Reactors can be categorized by the types of systems 

they use. Evolutionary designs, as with earlier large 

light-water reactor designs, have “active” safety 

systems powered by alternating current (ac). Passive

designs rely on physical phenomena such as gravity 

and natural cooling, as well as longer lasting batteries 

rather than ac-powered systems, for a specified time 

(e.g., 72 hours) following an accident.

When these policies were being developed in the 

1980s, advanced was the term used for reactors 

significantly different from those that were being 

constructed or operated. At the time, such designs 

included high-temperature gas-cooled reactors, 

liquid-metal-cooled reactors, and light-water reactors 

of innovative design. Passive designs such as the 

Westinghouse AP600 (and later AP1000) were 

suitably innovative to be categorized as “advanced.”

Today, the passive large light-water reactor designs 

are generally grouped with other evolutionary large 

light-water reactors and termed simply new reactors. 

These designs include the AP1000, the General 

Electric Hitachi Economic and Simplified BoilingWater Reactor (ESBWR—also a passive design) 

and Advanced Boiling-Water Reactor (ABWR), 

the AREVA U.S. Evolutionary Power Reactor 

(U.S. EPR), and the Mitsubishi U.S. Advanced 

Pressurized-Water Reactor (US-APWR).

In the NRC’s Office of New Reactors, the Advanced 

Reactor Program manages licensing activities 

for three categories of advanced reactors: small 

modular light-water reactors, high-temperature 

gas-cooled reactors, and liquid-metal-cooled reactors.

Regardless of design or terminology, all of these 

designs receive a full safety review by the NRC 

before being licensed for operation in this country.

New reactor specifics

In addition to these general policies, the 

Commission realized the value of enhancements 

in certain technical areas. Through a series of 

policy papers, the Commission directed the staff 

to impose certain new features and programs on 

proposed designs.

Enhancements for new 

reactors

In January 1989, the NRC staff informed the 

Commission of the planned approach to ongoing 

reviews of evolutionary new reactor designs. 

The staff identified several issues that should be 

resolved for these designs, such as fire protection, 

station blackout, and maintenance of 

design reliability.

A year later, the NRC staff presented the 

Commission with recommendations on a number 

of issues fundamental to agency decisions on 

evolutionary designs. The Commission approved 

most of these recommendations, sanctioning a 

number of key enhancements for new reactors.

The AP1000 reactor design (Courtesy: Westinghouse)
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These enhancements include the following:

• a diverse system to ensure shutdown of 

the reactor even if the main system fails

• design features to ensure high reliability of 

the shutdown decay heat removal system

• an alternate ac power source to mitigate 

station blackout

• additional requirements for fire protection, 

including ensuring safe shutdown without 

operator intervention, even if a fire 

disables all equipment in one fire area

• methods for core debris spreading and 

cooling if a severe accident were to occur

• a depressurization system and reactor 

cavity design features to contain ejected 

core debris if a severe accident were 

to occur

The ESBWR reactor design

A few years later, as the NRC began to review 

passive designs, the staff requested Commission 

direction on extending certain requirements to 

passive designs, as well as direction on 

additional enhancements.

In response, the Commission extended a number 

of its previous positions, such as those on 

enhanced fire protection, core debris coolability, 

and high pressure core melt ejection, to passive 

designs. The Commission also approved new 

positions on topics such as common-cause failures 

of instrumentation and control systems and steam 

generator tube ruptures.

In response to these Commission policies, new 

reactor applicants included a number of additional 

features in the designs they submitted to the NRC. 

Design-specific enhancements

The Commission has also made policy decisions 

on design-specific enhancements to ensure 

protection of the public and the environment.

For example, the Westinghouse AP600 design 

originally relied only on natural processes to 

remove radioactive material from the air in 

containment during a severe accident. Though 

these natural removal processes remove sufficient 

radioactive material during design-basis accidents, 

the staff determined that an active system would 

give the staff greater confidence in the design’s 

response to severe accidents.

As a result, the Commission approved the staff’s 

recommendation that the AP600 be required to 

add a containment spray system. This system was 

included in the certified design specifically to 

address severe accidents and was not credited to 

mitigate any design-basis accident.

Non-light-water reactors

In 2003, the Commission clarified its expectations 

for enhanced safety relative to reactor designs 

that cool the core using an inert gas or liquid 

metal instead of water. Similar to the policy 

for light-water reactors, reactor designers are 

expected to propose designs with enhanced 

safety characteristics; as needed, the NRC staff 
6

will recommend additional enhancements in 

areas of high uncertainty, subject to Commission 

endorsement. These enhancements could include 

additional design features or testing, as well as 

additional review or oversight by the NRC.

Implementing the policies

The NRC has implemented the Commission’s 

policies on new reactor safety through rules, 

guidance, staff reviews, and inspection.

10 CFR Part 52

The Commission approved issuance of 

10 CFR Part 52, “Licenses, Certifications, and 

Approvals for Nuclear Power Plants,” in 1989, 

with a major update in 2007. The regulations in 

10 CFR Part 52 provide for standardized reviews 

of designs, sites, and license applications for 

new nuclear power plants. These requirements 

formalize the criteria for new designs identified in 

the policy statement on severe accidents.

In 10 CFR Part 52, the NRC also included further 

requirements related to severe accidents. Design 

certification applicants must provide a description 

and analysis of design features that prevent or 

mitigate specific severe accident phenomena 

(e.g., steam explosion, containment bypass). They 

must also provide an environmental report that 

addresses the costs and benefits of adding more 

severe accident mitigation features.

In addition, the rule includes the expectation 

that new reactors will reflect an extremely low 

probability of accidents that could result in the 

release of significant quantities of 

radioactive material.

Certified designs

When the NRC certifies a new reactor design, 

it documents the certified portion of the design 

in an appendix to 10 CFR Part 52. This process 

means that the important features of these designs 

are recorded as rules that must be followed by 

applicants who wish to build that design.

The ABWR reactor design

Vendors’ consideration of the policies described 

above resulted in certified designs that reflect an 

enhanced level of safety compared with currently 

operating reactors. Because the Commission’s 

intention was to maintain this level of safety 

throughout plant operation, the design certification 

rules include a statement that the Commission will 

deny exemptions from the certified design that 

would “result in a significant decrease in the level 

of safety otherwise provided by the design.” In the 

documentation that accompanied the first design 

certifications, the Commission emphasized its 

expectation that the industry would cooperate with 

the NRC to maintain this level of enhanced safety.

The design certification rules also include a 

special change process to ensure that the features 

developed to protect against severe accidents 

would be maintained. The NRC staff must review 

proposed changes that affect design features 

that deal with “ex-vessel” accidents, where core 

material leaves the reactor vessel, if there would 
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be a substantial increase in the probability or 

consequences of these types of accidents.

Aircraft impact rule

Another rule that implements the Commission’s 

expectation of enhanced safety is 10 CFR 50.150, 

“Aircraft Impact Assessment,” issued in 2009. 

This rule requires new reactor applicants to 

assess what would happen if a large commercial 

aircraft hit their plant and to show that harmful 

amounts of radioactive materials from the reactor 

core and spent fuel would not be released to the 

environment.

The supplementary information that accompanied 

the final rule explicitly states that the rule 

“provides an enhanced level of protection beyond 

that which is provided by the existing adequate 

protection requirements.” The discussion refers 

to the policy statements on severe accidents and 

advanced reactors, emphasizing that–

This regulatory approach has demonstrated 

its success, as all designs subsequently 

submitted to and certified by the Commission 

represent substantial improvement in safety 

for operational events and accidents. The final 

aircraft impact rule will further enhance the 

safety of new nuclear power plants for aircraft 

impacts and is consistent with these policy 

statements.

Risk insights

As discussed in the policy statement on severe 

accidents and codified in 10 CFR Part 52, new 

reactor applicants must develop a PRA and use 

insights from the PRA to reduce the risk from 

severe accidents. In response, applicants have 

reduced the risk of their reactors through a number 

of design changes.

For example, the ABWR includes a number 

of design and procedure modifications that the 

applicant chose to make based on risk insights. 

These changes included the following:

• an ac-independent means of spraying 

water in containment

• a containment overpressure 

protection system

• a type of concrete that would limit 

production of noncondensible gases 

during a severe accident

The U.S. EPR containment building 

(Courtesy: AREVA NP)

More recently, the AP1000 design incorporated 

multiple enhancements based on PRA insights, 

including the following:

• changing the type or normal position of 

certain valves to improve reliability

• modifying the core design to reduce the 

risk from potential failures to shut down 

the reactor
8

• selecting an optimal design for an 

accumulator tank that would quickly 

replenish water during an accident

Applicants’ risk assessments and related design 

changes, therefore, have contributed to the 

enhanced safety of new reactors. 

Staff reviews and inspections

Before the NRC certifies a design or licenses 

a plant to operate, the NRC staff completes a 

thorough technical review to ensure that the plant 

will be safe and secure and meet all relevant 

requirements. In this review, the staff uses 

guidance documents that provide additional detail 

on how to implement the NRC’s requirements.

Many of these guidance documents refer directly 

to policies and regulations described in this 

brochure. There is guidance for applicants on how 

to implement enhancements in fire protection and 

PRA, to name just two. There is also guidance for 

NRC staff on how to perform the safety review. 

For example, the guidance for reviewing PRA 

information acknowledges that designs may 

incorporate features intended to make the plant 

safer and easier to operate. Accordingly, the 

staff determines whether the design represents a 

reduction in risk compared to existing plants.

NRC inspection team in front of an 11,300-ton press at 

Creusot Forge in France

The NRC is also implementing a stringent 

construction inspection program to ensure that the 

facility is constructed in the way it was licensed. 

Resident inspectors will oversee daily activities 

at each new reactor construction site. Additional 

regional inspectors and technical staff will 

participate when needed. Using a risk-informed 

approach, the NRC determines how serious an 

inspection issue is and follows up with more 

inspections or other penalties when warranted. 

These activities ensure that the Commission’s 

requirements and expectations related to new 

reactor safety and security are met.

The US-APWR reactor design (Courtesy: Mitsubishi)

Conclusion

The NRC is confident that the current generation 

of nuclear power plants is operating safely. This 

conclusion is supported by rigorous licensing 

reviews, improvements that have resulted in 

reduced radiological risk, and daily inspections. 

New reactor designers have used technology 

improvements, operational experience, and 
9

insights from PRAs to make their designs 

even safer. The Commission has consistently 

encouraged these objectives and, when needed, 

imposed additional requirements to enhance safety.

As a result, new reactor designs benefit from 

significant enhancements in safety. The NRC’s 

reviews of these designs continue to support its 

mission of protecting public health and safety and 

the environment.

References

1. U.S. Code of Federal Regulations, “Domestic Licensing 

of Production and Utilization Facilities,” Part 50, 

Chapter I, Title 10, “Energy.”

2. “Safety Goals for the Operations of Nuclear Power 

Plants; Policy Statement; Republication,” Volume 51 of 

the Federal Register (FR), page 30,028, 

August 21, 1986.

3. “Policy Statement on Severe Reactor Accidents 

Regarding Future Designs and Existing Plants,” 

50 FR 32138, August 8, 1985.

4. “Regulation of Advanced Nuclear Power Plants; 

Statement of Policy,” 51 FR 24643, July 8, 1986, 

revised in 59 FR 35461, July 12, 1994.

5. SECY-89-013, “Design Requirements Related to the 

Evolutionary Advanced Light Water Reactors,” 

January 19, 1989.

6. SECY-89-102, “Implementation of Safety Goal Policy,” 

March 30, 1989, and associated Staff Requirements 

Memorandum (SRM), June 15, 1990.

7. SECY-90-016, “Evolutionary Light Water Reactor 

(LWR) Certification Issues and Their Relationship to 

Current Regulatory Requirements,” January 12, 1990, 

and associated SRM, June 26, 1990.

8. SECY-93-087, “Policy, Technical, and Licensing 

Issues Pertaining to Evolutionary and Advanced LightWater Reactor (ALWR) Designs,” April 2, 1993, and 

associated SRM, July 21, 1993.

9. NUREG-1503, “Final Safety Evaluation Report Related 

to the Certification of the Advanced Boiling Water 

Reactor Design,” July 1994.

10. SECY-97-044, “Policy and Key Technical Issues 

Pertaining to the Westinghouse AP600 Standardized 

Passive Reactor Design,” February 18, 1997, and 

associated SRM, June 30, 1997.

11. SECY-03-0047, “Policy Issues Related to Licensing 

Non-Light-Water Reactor Designs,” March 28, 2003, 

and associated SRM, June 26, 2003.

12. “Early Site Permits; Standard Design Certifications; and 

Combined Licenses for Nuclear Power Reactors 

[10 CFR Part 52],” 54 FR 15372, April 18, 1989.

13. “Licenses, Certifications, and Approvals for Nuclear 

Power Plants [Revised 10 CFR Part 52],” 72 FR 49352, 

August 28, 2007.

14. NUREG-1793, “Final Safety Evaluation Report Related 

to Certification of the AP1000 Standard Design,” 

September 2004.

15. Regulatory Guide 1.206, “Combined License 

Applications for Nuclear Power Plants (LWR Edition),” 

June 2007.

16. NUREG-0800, “Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear Power Plants: 

LWR Edition,” Section 19.0, “Probabilistic Risk 

Assessment and Severe Accident Evaluation for New 

Reactors,” June 2007.

17. “Policy Statement on the Regulation of Advanced 

Reactors,” 73 FR 60612, October 14, 2008.

18. NUREG-0800, Section 9.5.11, “Fire Protection 

Program,” March 2009.

19. “Consideration of Aircraft Impacts for New Nuclear 

Power Reactors,” 74 FR 28112, June 12, 2009.

20. Regulatory Guide 1.189, “Fire Protection for Nuclear 

Power Plants,” October 2009.

21. SECY-10-0121, “Modifying the Risk-Informed 

Guidance for New Reactors,” September 14, 2010, and 

associated SRM, March 2, 2011.

22. DC/COL-ISG-018, “Interim Staff Guidance on 

Standard Review Plan, Section 17.4, Reliability 

Assurance Program,” March 2011.

23. NUREG/BR-0314, Revision 2, “Protecting Our Nation: 

A Report of the U.S. Nuclear Regulatory Commission,” 

June 2011.
NUREG/BR-0356

November 2011