March 2007

USNRC STANDARD REVIEW PLAN

This Standard Review Plan, NUREG-0800, has been prepared to establish criteria that the U.S. Nuclear Regulatory Commission 

staff responsible for the review of applications to construct and operate nuclear power plants intends to use in evaluating whether

an applicant/licensee meets the NRC's regulations. The Standard Review Plan is not a substitute for the NRC's regulations, and

compliance with it is not required. However, an applicant is required to identify differences between the design features, analytical techniques, and procedural measures proposed for its facility and the SRP acceptance criteria and evaluate how the proposed

alternatives to the SRP acceptance criteria provide an acceptable method of complying with the NRC regulations.

The standard review plan sections are numbered in accordance with corresponding sections in Regulatory Guide 1.70, "Standard

Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)." Not all sections of Regulatory Guide 1.70

have a corresponding review plan section. The SRP sections applicable to a combined license application for a new light-water

reactor (LWR) are based on Regulatory Guide 1.206, "Combined License Applications for Nuclear Power Plants (LWR Edition)."

These documents are made available to the public as part of the NRC's policy to inform the nuclear industry and the general public

of regulatory procedures and policies. Individual sections of NUREG-0800 will be revised periodically, as appropriate, to accommodate comments and to reflect new information and experience. Comments may be submitted electronically by email to

NRR_SRP@nrc.gov.

Requests for single copies of SRP sections (which may be reproduced) should be made to the U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to DISTRIBUTION@nrc.gov. Electronic copies of this section are available through the NRC's public Web site at

http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/, or in the NRC's Agencywide Documents Access and Management System (ADAMS), at http://www.nrc.gov/reading-rm/adams.html, under Accession # ML070230012.

NUREG-0800

 U.S. NUCLEAR REGULATORY COMMISSION

STANDARD REVIEW PLAN

15.0.3 DESIGN BASIS ACCIDENT RADIOLOGICAL CONSEQUENCE ANALYSES FOR

ADVANCED LIGHT WATER REACTORS

REVIEW RESPONSIBILITIES

Primary - Organization responsible for the review of design basis accident radiological

consequence analyses

Secondary - Organization responsible for the review of meteorology

I. AREAS OF REVIEW

Chapter 15 of the SRP discusses the analysis of postulated accidents that could affect the safe

design and siting of an advanced light-water reactor (LWR). The staff reviews information

presented by the applicant for a construction permit (CP), operating license (OL), standard

design certification (DC), early site permit (ESP), or combined operating license (COL)

concerning radiological consequence analyses for postulated design basis accidents. This

SRP section applies to reviews performed for each of these types of applications. The review

covers the following specific areas: 

1. CP, OL, DC or COL Applications. For a CP, OL, DC or COL application, the staff

reviews the radiological consequences of potential design basis accidents (DBAs) in six

parts: (1) review of selected bounding design basis accidents, (2) review of accident

source terms, (3) review of the major structures, systems, and components of the facility

that are intended to mitigate the radiological consequences of a DBA, (4) review of the
15.0.3-2 March 2007

characteristics of fission product releases from the proposed site (for CP, OL and COL

reviews) or reference site (for the DC review) to the environment, (5) review of the

meteorological characteristics of the proposed site for the CP, OL or COL review

(reference site for DC review), and (6) review of the total calculated radiological

consequence dose at the exclusion area boundary (EAB), low population zone (LPZ)

and control room from the bounding DBAs. In support of the SRP Section 13.3

emergency planning review, the staff also reviews the dose analysis performed to

demonstrate technical support center (TSC) habitability. 

The application must contain sufficient nuclear plant design information for the staff to

review in making a determination regarding the acceptability of the proposed site using

the radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1) and

General Design Criterion (GDC) 19.

2. ESP Applications that Reference Standard Reactor Designs Certified by NRC. 

Standard reactor designs are certified with a postulated set of short-term atmospheric

relative concentration (χ/Q) values at an EAB and LPZ in lieu of site-specific

meteorological data and actual distances to the EAB and LPZ. The NRC has

determined, for purposes of the ESP review, that the certified standard reactor designs

meet the radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1),

provided that the site parameters fall within those postulated in the design certification.

3. ESP Applications that Use the Plant Parameter Envelope (PPE) Approach. A PPE is a

set of plant design parameters that are expected to bound the characteristics of a

reactor or reactors that may be constructed at a site, and it serves as a surrogate for

actual reactor design information. The PPE values are selected by the applicant to

bound a range of possible current and future reactor designs. The PPE values and

associated information in the ESP application must contain sufficient information for the

staff to make a determination regarding the acceptability of the proposed site using the

radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1). 

4. ESP Applications that Neither Reference the Standard Reactor Designs Certified by

NRC Nor Use the PPE Approach. Applications may be received that neither reference a

certified design nor use the PPE approach. For example, an application may reference

a “standard” design that is not yet certified, or a custom design. In such cases, the staff

reviews the radiological consequences of potential DBAs in six parts: (1) review of

selected bounding design basis accidents, (2) review of accident source terms,

(3) review of the major structures, systems, and components of the facility that bear

significantly on the acceptability of the site for mitigating the radiological consequences

of a DBA under the radiological consequence evaluation, (4) review of the

characteristics of fission product release from the site to the environment, (5) review of

the meteorological characteristics of the proposed site, and (6) review of the total

calculated radiological consequence dose at the EAB and LPZ from the bounding DBAs. 

The application must contain sufficient nuclear plant design information for the staff to

review in making a determination regarding the acceptability of the proposed site using

the radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1).

5. COL Applications that Reference Standard Reactor Designs Certified by NRC and ESP

Issued by NRC. Should the site characteristic short-term χ/Q values specified in the
1

Additional supporting information of prior DC rules may be found in DCD Tier 2 Section 14.3.

15.0.3-3 March 2007

ESP fall within the postulated short-term χ/Qs for the chosen certified design, the staff

concludes that the COL applicant has satisfied the radiological consequence evaluation

factors identified in 10 CFR 50.34(a)(1). However, the application must contain

sufficient information regarding control room habitability for the staff to make a

determination regarding the acceptability of the proposed control room design using the

radiological dose acceptance criteria specified in GDC 19.

 

6. COL Applications that Reference an ESP Issued by NRC but not a Certified Standard

Reactor Design by NRC. The staff reviews the radiological consequences of potential

DBAs in five parts: (1) review of selected bounding design basis accidents, (2) review of

accident source terms, (3) review of the major structures, systems, and components of

the facility that bear significantly on the acceptability of the site for mitigating the

radiological consequences of a DBA under the radiological consequence evaluation,

(4) review of the characteristics of fission product release from the site to the

environment, and (5) review of the total calculated radiological consequence dose at the

EAB and LPZ from the bounding DBAs to determine whether the applicable regulations

in 10 CFR 50.34(a)(1) and GDC 19 regarding dose consequence evaluation factors

have been met.

7. COL Applications that Reference a Certified Standard Reactor Design by NRC but not

an ESP Issued by NRC. The staff evaluates the site-specific short-term χ/Qs for the

selected site and uses the site-specific χ/Qs and the source term determined in the

certified design to determine whether the applicable regulations in 10 CFR 50.34(a)(1)

regarding dose consequence evaluation factors have been met. The application must

contain sufficient information regarding control room habitability for the staff to make a

determination regarding the acceptability of the proposed control room design using the

radiological dose acceptance criteria specified in GDC 19.

8. COL Action Items and Certification Requirements and Restrictions. For a DC

application, the review will also address COL action items and requirements and

restrictions (e.g., interface requirements and site parameters).

 For a COL application referencing a DC, a COL applicant must address COL action

items (referred to as COL license information in certain DCs) included in the referenced

DC. Additionally, a COL applicant must address requirements and restrictions (e.g.,

interface requirements and site parameters) included in the referenced DC.

Review Interfaces

Other SRP sections interface with this section as follows:

1. For DC applications and COL applications referencing a DC rule or DC application,

review of the site parameters in the Design Control Document (DCD) Tier 1 and Chapter

2 of the DCD Tier 21

 submitted by the applicant is performed under SRP Section 2.0,

“Site Characteristics and Site Parameters.”
15.0.3-4 March 2007

2. Review of the short-term χ/Q values for use in the DBA radiological consequences

analyses is performed under SRP Section 2.3.4, “Short-Term Dispersion Estimates for

Accidental Atmospheric Releases.”

3. Review of the coolant radioactivity source terms for non-LOCA accidents is performed

under SRP Section 11.1, “Source Terms.”

4. Review of the provisions for protection of the control room from radiation and habitability

during an emergency is performed under SRP Section 6.4, “Control Room Habitability

System.” A similar review of TSC habitability is performed in support of SRP

Section 13.3, “Emergency Planning.”

5. Review of the emergency safety features (ESFs) ventilation and filtration systems that

are designed to remove fission products is performed under SRP Section 6.5.1, “ESF

Atmosphere Cleanup Systems.” 

6. Review of the fission product removal capability of containment spray systems is

performed under SRP Section 6.5.2, “Containment Spray as a Fission Product Cleanup

System.”

7. Review of the analysis modeling of fission product removal capability for plant systems

and structures is performed under SRP Section 6.5.3, “Fission Product Control Systems

and Structures.”

8. If a plant design includes an ice condenser in the containment, the review of the fission

product removal capability of the ice condenser is performed under SRP Section 6.5.4, 

“Ice Condenser as a Fission Product Cleanup System.”

9. The review of the fission product removal capability of a BWR pressure suppression pool

is performed under SRP Section 6.5.5, “Pressure Suppression Pool as a Fission Product

Cleanup System.”

10. For review of DC applications, CPs and OLs, and COLs or ESPs referencing an

advanced light-water reactor design, this SRP section supersedes the radiological

analyses, assumptions, acceptance criteria, and methodologies identified in the SRP

sections (with appendices) listed below. Provisions related to the nonradiological

analysis aspects of these SRP sections remain applicable. 

A. Section 15.1.5, “Steam System Piping Failures Inside and Outside of

Containment (PWR)” 

B. Sections 15.3.3-15.3.4, “Reactor Coolant Pump Rotor Seizure and Reactor

Coolant Pump Shaft Break”

C. Section 15.4.8, “Spectrum of Rod Ejection Accidents (PWR)”

D. Section 15.4.9, “Spectrum of Rod Drop Accidents (BWR)”

E. Section 15.6.2, “Radiological Consequences of the Failure of Small Lines

Carrying Primary Coolant Outside Containment”
15.0.3-5 March 2007

F. Section 15.6.3, “Radiological Consequences of Steam Generator Tube Failure

(PWR)”

G. Section 15.6.4, “Radiological Consequences of Main Steam Line Failure Outside

Containment (BWR)”

H. Section 15.6.5, “Loss-of-Coolant Accidents Resulting From Spectrum of

Postulated Piping Breaks Within the Reactor Coolant System Pressure Boundary”

I. Section 15.7.4, “Radiological Consequences of Fuel Handling Accidents”

J. Section 15.7.5, “Spent Fuel Cask Drop Accidents”

The specific acceptance criteria and review procedures are contained in the reference SRP

sections.

II. ACCEPTANCE CRITERIA

Requirements

Acceptance criteria are based on meeting the relevant requirements of the following

Commission regulations:

1. Section 50.34(a)(1) of 10 CFR Part 50, “Contents of applications; technical information,”

as it relates to the evaluation and analysis of the offsite radiological consequences of

postulated accidents with fission product release.

2. General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50, “Control room,” as

it relates to maintaining the control room in a safe condition under accident conditions by

providing adequate protection against radiation.

3. Section 100.21 of 10 CFR Part 100, “Non-seismic siting criteria,” as it relates to the

evaluation and analysis of the radiological consequences of postulated accidents for the

type of facility to be located at the site in support of evaluating the site atmospheric

dispersion characteristics.

4. Paragraph IV.E.8 of Appendix E, to 10 CFR Part 50, “Emergency Planning and

Preparedness for Production and Utilization Facilities,” as it relates to adequate

provisions for an onsite technical support center (TSC) from which effective direction can

be given and effective control can be exercised during an emergency.

SRP Acceptance Criteria

Specific SRP acceptance criteria acceptable to meet the relevant requirements of the NRC’s

regulations identified above are as follows for the review described in this SRP section. The

SRP is not a substitute for the NRC’s regulations, and compliance with it is not required. 

However, an applicant is required to identify differences between the design features, analytical

techniques, and procedural measures proposed for its facility and the SRP acceptance criteria

and evaluate how the proposed alternatives to the SRP acceptance criteria provide acceptable

methods of compliance with the NRC regulations. 
15.0.3-6 March 2007

1. Offsite Radiological Consequences of Postulated Design Basis Accidents. The

acceptance criteria are based on the requirements of 10 CFR 50.34(a)(1) as related to

mitigating the radiological consequences of an accident in accordance with

10 CFR 52.17(a)(1) [early site permits], 10 CFR 52.47(a)(1) [standard design

certifications] and 10 CFR 52.79(b) [combined licenses].

The plant design features intended to mitigate the radiological consequences of

accidents, site atmospheric dispersion characteristics and the distances to the exclusion

area boundary (EAB) and to the low population zone (LPZ) outer boundary are

acceptable if the total calculated radiological consequences for the postulated fission

product release fall within the following exposure acceptance criteria specified in

10 CFR 50.34(a)(1)(ii)(D):

A. An individual located at any point on the boundary of the exclusion area for any

2-hour period following the onset of the postulated fission product release, would

not receive a radiation dose in excess of 25 rem total effective dose equivalent

(TEDE), and

B. An individual located at any point on the outer boundary of the LPZ, who is

exposed to the radioactive cloud resulting from the postulated fission product

release (during the entire period of its passage), would not receive a radiation

dose in excess of 25 rem TEDE.

For CP, OL, DC and COL reviews, the application is acceptable with regard to the

radiological consequences of analyzed DBAs if the calculated TEDEs at the EAB and the

LPZ outer boundary do not exceed the dose acceptance criteria listed in Table 1 below. 

For ESP applications that neither reference the standard reactor designs certified by

NRC nor use the PPE approach, the staff may establish dose acceptance criteria lower

than those stated above for certain DBAs based on the probability of occurrence. 

Examples of such criteria are illustrated in Table 1. 

For COL applications using an ESP with a PPE approach, these acceptance criteria may

be applied at that time. Such applicants bear the burden of ensuring sufficient margin is

provided in the design parameters (for example, PPE values) in the ESP application to

compensate for uncertainty in those parameters. The margin should be large enough

such that the actual design submitted at the COL stage, coupled with the site

characteristics as described in the ESP, will comply with NRC regulations.

2. Control Room Radiological Habitability. The acceptance criterion is based on the

requirements of GDC 19 that mandate a control room design providing adequate

radiation protection to permit access and occupancy of the control room under accident

conditions for the duration of the accident, without personnel receiving radiation

exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for

the duration of the accident. These requirements are incorporated by reference in

10 CFR 52.47(a)(1) [standard design certifications] and 10 CFR 52.79(b) [combined

licenses].

The radiation protection design of the control room is acceptable if the total calculated

radiological consequences for the postulated fission product release fall within the
15.0.3-7 March 2007

exposure acceptance criteria specified in GDC 19 of 5 rem TEDE for the duration of the

accident. 

3. Technical Support Center Radiological Habitability. This acceptance criterion is based

on the requirement of Paragraph IV.E.8 of Appendix E to 10 CFR Part 50 to provide an

onsite TSC from which effective direction can be given and effective control can be

exercised during an emergency. The radiation protection design of the TSC is

acceptable if the total calculated radiological consequences for the postulated fission

product release fall within the exposure acceptance criteria specified for the control room

of 5 rem TEDE for the duration of the accident. 

Table 1

Accident Dose Criteria

Accident or Case

EAB and LPZ

Dose Criteria Analysis Release Duration

LOCA 25 rem TEDE 30 days for all leakage pathways

BWR Main Steam Line Break Instantaneous puff, until MSIV

Fuel Damage or Pre-incident Spike 25 rem TEDE isolation

Equilibrium Iodine Activity 2.5 rem TEDE

BWR Rod Drop Accident 6.3 rem TEDE 24 hours

Small Line Break Accident 2.5 rem TEDE Until isolation, if capable, or until

cold shutdown is established

PWR Steam Generator Tube Rupture Affected SG: time to isolate;

Unaffected SG(s): until cold

shutdown is established Fuel Damage or Pre-incident Spike 25 rem TEDE

Coincident Iodine Spike 2.5 rem TEDE

PWR Main Steam Line Break Until cold shutdown is established

Fuel Damage or Pre-incident Spike 25 rem TEDE

Coincident Iodine Spike 2.5 rem TEDE

PWR Locked Rotor Accident 2.5 rem TEDE Until cold shutdown is established

PWR Rod Ejection Accident 6.3 rem TEDE 30 days for containment leakage

pathway; Until cold shutdown is

established for secondary

pathway

Fuel Handling Accident or Cask Drop 6.3 rem TEDE 2 hours

Technical Rationale

The technical rationale for application of these acceptance criteria to the areas of review

addressed by this SRP section is discussed in the following paragraphs: 

1. Compliance with 10 CFR 50.34(a)(1) ensures that the safety analysis report (SAR)

includes a description and safety assessment of the standard design, custom design
15.0.3-8 March 2007

and/or site on which the facility is to be located. The review performed under this SRP

section ensures that the SAR contains a sufficient description of the design basis

accident radiological consequences analyses that will enable the staff to evaluate the

planned site and provide reasonable assurance that plant design and operation will

reflect site considerations in a manner adequate to minimize the consequences of an

accident.

The dose acceptance criteria in Table 1 of this SRP section are fractions of the

10 CFR 50.34(a)(1) dose reference values for accidents other than the LOCA, as has

been done historically. For events having a moderate frequency of occurrence, any

release of radioactive material must be such that the calculated offsite doses are a small

fraction of the 10 CFR 50.34(a)(1) reference values. A small fraction is defined as less

than 10% of the 10 CFR 50.34(a)(1) reference values, or 2.5 rem TEDE. The plant site

and dose mitigating engineered safety features are acceptable with respect to the

radiological consequences of a postulated control rod drop accident (BWR), control rod

ejection accident (PWR), fuel handling accident or cask drop accident if the calculated

offsite doses are well within the dose reference values in 10 CFR 50.34(a)(1). “Well

within" is defined as 25% of the 10 CFR 50.34(a)(1) reference values, or 6.3 rem TEDE. 

2. Compliance with the radiological provision of GDC 19 provides assurance that control of

the plant is maintained during emergency operation. The applicant is required to

maintain the control room in a safe condition under accident conditions, including loss-ofcoolant accidents, and provide adequate radiation protection to permit access and

occupancy of the control room under accident conditions for the duration of the accident. 

The review performed under this SRP section for CPs, OLs, DCs and COLs determines if

the design of the control room is acceptable with respect to the radiological

consequences of design basis accidents. 

3. 10 CFR 100.21 requires that radiological dose consequences of postulated accidents

meet the requirements of 10 CFR 50.34(a)(1) for the type of facility proposed to be

located at the site. Compliance with 10 CFR Part 100 provides assurance that the

consequences of an accident on the proposed site will be within acceptable levels. The

review performed under this SRP section for CPs, OLs, COLs and ESPs determines if

the site is acceptable with respect to the radiological consequences of design basis

accidents. 

4. Paragraph IV.E.8 of Appendix E, to 10 CFR Part 50, requires that onsite emergency

facilities be provided, from which effective direction can be given and effective control

can be exercised during an emergency. NUREG-0737 III.A.1.2, Emergency Response

Facilities, describes requirements for maintaining emergency facilities in a safe, habitable

condition under accident conditions by providing adequate protection against radiation

and toxic gases. In particular, the TSC should provide the same level of protection

against radiation that the control room provides, for the duration of the event. The

radiological consequences analysis for the TSC is performed under this section to

support the Section 13.3 review for acceptability of the TSC. 

III. REVIEW PROCEDURES

The reviewer will select material from the procedures described below, as may be appropriate

for a particular case.
15.0.3-9 March 2007

These review procedures are based on the identified SRP acceptance criteria. For deviations

from these acceptance criteria, the staff should review the applicant’s evaluation of how the

proposed alternatives provide an acceptable method of complying with the relevant NRC

requirements identified in Subsection II.

For reviews of OL applications, these procedures are used to verify that the data and analyses

remain valid and that the facility’s design specifications are consistent with these data. As

applicable, reviews of OLs and COLs include a determination that the content and intent of

technical specifications related to the plant features intended to mitigate the radiological

consequences of postulated design basis accidents are acceptable and consider any identified

unique conditions.

1. ESP applications that reference standard reactor designs certified by NRC 

A. In the evaluation using Section 2.3.4 of this standard review plan, the staff

reviews the applicant’s meteorological data, inputs, assumptions, and dispersion

model used to estimate the site-specific short-term atmospheric dispersion

estimate (χ/Q) values in the ESP application. 

B. The staff compares the site-specific short-term χ/Q values in the ESP application

with short-term χ/Q values postulated in the reactor design certification. 

C. If the site-specific short-term χ/Q values fall within those postulated in the design

certification, no further radiological consequence evaluation is needed. 

D. If the site-specific short-term χ/Q values exceed those postulated in the design

certification, the staff verifies that the applicant has demonstrated that the

radiological consequences associated with the bounding DBAs using the

applicant’s site-specific short-term χ/Q values meet the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1).

NOTE: At the COL stage, the staff verifies that no changes from the site-specific shortterm χ/Q values specified in the ESP application have occurred due to changes in plant

design, plant location on the site, building orientation, or fission product release points. 

The staff performs independent confirmatory radiological consequence dose calculations

using the site-specific short-term χ/Q values and the source term provided in the certified

reactor design control document to determine the resulting radiological consequences at

the EAB and LPZ for public information and to supplement the design basis.

NOTE: Also the COL stage, the staff determines that the radiation protection design of

the control room is acceptable if the total calculated radiological consequences for the

postulated fission product release fall within the exposure acceptance criteria specified in

GDC 19 of 5 rem TEDE for the duration of the accident. 

2. ESP applications that use the PPE approach

A. The staff reviews the proposed PPE values to determine whether the set of PPE

values is sufficient to enable the staff to conduct its evaluation of the radiological

consequences. The PPE values should be found not unreasonable for
15.0.3-10 March 2007

consideration in the staff’s findings regarding compliance with Subpart A of

10 CFR Part 52.

B. To enable the staff to perform its independent radiological consequence

analyses, the PPE values should include, but are not limited to, the following

design basis accident source term parameters:

i. The isotopic quantities of fission products released in curies to the

environment from the site. 

ii. Rates of fission product release to the environment from the site as a

function of time.

C. The staff reviews the following information if available: (1) the timing and rate of

fission product release from the fuel and (2) the isotopic quantities and the

chemical forms of fission products released from the fuel, following selected

bounding DBAs. This information will help the staff determine whether the

proposed PPE values are not unreasonable. The fission product release rates

should be fractions of fission product inventory in the reactor core at the ultimate

maximum power level. 

D. In the evaluation using Section 2.3.4 of this standard review plan, the staff

reviews the site-specific short-term χ/Q values determined by the applicant and

performs an independent evaluation of atmospheric dispersion.

E. The staff performs independent confirmatory radiological consequence analyses

using the docketed PPE values and the site-specific short-term χ/Q values

provided in the ESP application to determine whether the proposed site meets the

radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1) at

the nearest EAB and LPZ outer boundary as described in Chapter 2 of the site

safety assessment. 

F. For the methodology and assumptions for calculating the radiological

consequences of postulated accidents, the staff will use, where applicable, the

regulatory positions stated in Regulatory Guide 1.183, “Alternative Radiological

Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors”.

NOTE: If a COL application references a certified design and an ESP that referenced a

PPE, the staff reviews (at the COL stage) the site-specific short-term χ/Q values

specified in the ESP to confirm that the site-specific short-term χ/Q values are bounded

by those short-term χ/Q values postulated in the reactor design certification based on the

proposed plant design, the plant location on the site, and the fission product release

points.

In the event that the site-specific short-term χ/Q values exceed the bounds of those

postulated in the referenced design certification, the staff verifies that the COL applicant

has demonstrated that the radiological consequences associated with the bounding

DBAs using its site-specific χ/Q values continue to meet the radiological consequence

evaluation factors of 10 CFR 50.34(a)(1). In addition, the staff determines that the

radiation protection design of the control room is acceptable if the total calculated
15.0.3-11 March 2007

radiological consequences for the postulated fission product release fall within the

exposure acceptance criteria specified in GDC 19 of 5 rem TEDE for the duration of the

accident. 

3. COL applications that reference both an ESP and a standard reactor design certified by

NRC

A. The staff verifies that no changes from the site-specific short-term χ/Q values

specified in the ESP application have occurred due to changes in plant design,

plant location on the site, building orientation, or fission product release points. 

B. Should the site-specific characteristic short-term χ/Q values specified in the ESP

fall within the postulated short-term χ/Qs for the chosen certified design, the staff

concludes that the COL applicant has satisfied the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1). 

C. If the site-specific short-term χ/Q values do not fall within the postulated

short-term χ/Q values for the chosen certified design, the staff reviews the

applicant’s radiological dose calculations and performs independent confirmatory

radiological consequence dose calculations using the site-specific short-term χ/Q

values and the source term provided in the certified reactor design control

document. 

D. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the facility are combined and are compared with

the radiological consequence evaluation factors identified in 10 CFR 50.34(a)(1)

at the nearest EAB and LPZ outer boundary stated in the applicant’s site safety

assessment. 

E. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the facility, including all sources of radiation

exposure to the control room personnel, are combined, and the calculated dose

in the control room is compared with the radiological consequence evaluation

factors identified in GDC 19.

F. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the facility, including all sources of radiation

exposure to the personnel in the technical support center, are combined, and the

calculated dose in the TSC is compared with the radiological consequence

evaluation factors identified for the control room of 5 rem TEDE for the duration

of the accident.

4. CP, OL, DC and COL and ESP applications that neither reference a standard reactor

design certified by NRC nor use the PPE approach

A. The staff reviews the sequences of DBA events as described by the applicant to

ensure that the spectrum of DBAs includes the bounding DBA with respect to 

calculated fission product releases. The spectrum of DBAs has generally been

assumed to reflect a substantial meltdown of the reactor core (a major reactor

accident) with subsequent release of appreciable quantities of fission products to
15.0.3-12 March 2007

the environment. Although the loss-of-coolant (LOCA) is typically the maximum

credible accident associated with the light-water reactor design, the applicant

should consider other accident sequences of greater radiological consequence

for the specific reactor designs selected by the applicants or for reasonably

foreseeable future reactor designs if the applicant has not selected the specific

reactor designs at the time of ESP application.

B. The staff reviews a spectrum of representative DBAs selected and evaluated by

the applicant for determining the bounding DBA radiological consequences. The

selected DBA should cover a spectrum of reactor transients and accidents. 

C. The applicant’s proposed accident source terms are reviewed in the following

areas:

i. Fission product inventory in the reactor core operated at the ultimate

maximum proposed power level with the limiting condition which

maximizes fission product releases. 

ii. Timing and rate of fission product release from the fuel following selected

DBAs. The fission product release rates should be fractions of fission

product inventory in the reactor core based on the maximum full power

operation.

iii. The coolant activity concentration is reviewed under SRP Section 11.1.

The non-LOCA DBA coolant source terms are calculated based on the

coolant activity concentration and include iodine spiking, using the

guidance in RG 1.183. 

iv. The isotopic quantities in curies and the chemical forms of fission

products released to the containment and to the environment. The staff

reviews the modeling of changes in chemical form as the releases are

processed by mitigating systems.

v. Rates of fission product release to the environment from the site during

the entire period of the DBA as a function of time.

D. The staff reviews the fission product distribution, transport, removal, and release

models within and between the major structures and systems, as well as the

engineered safety feature (ESF) components of the facility, that bear significantly

on the acceptability of the site with respect to the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1). The staff reviews the

efficiencies of fission product removal by the ESF systems and components. 

Conditions for credit for fission product removal by ESF systems are discussed

in SRP Sections 6.5.1 - 6.5.5 and control room habitability systems are

discussed in SRP Section 6.4. The review under this SRP section should be

coordinated with the primary review organization for each of the aforementioned

sections. 
15.0.3-13 March 2007

E. Advanced reactor designs may include unique design features and passive

safety systems. The design basis accident radiological consequences analyses

may consider credit for the mitigation capability of the design through natural

fission product removal processes such as diffusiophoresis, thermophoresis and

gravitational settling. The staff’s review of removal through natural fission

product removal processes or for unique features of the design will require

additional information from the applicant to fully explain the process being

credited, the amount of removal being credited (specifically decontamination

factors or coefficients and timing), basis for the proposed values and inputs to

the dose analysis calculation, and the justification for assuming the removal

process is applicable to the design of the plant for the duration of the event. The

staff should determine if a technical assistance contract to assist the NRC staff

should be placed to verify the applicant’s proposed fission product removal

credit. 

F. The staff reviews the points of fission product release from the major structures

and systems, and from the ESF components of the facility.

G. Using Section 2.3.4 of this standard review plan, the staff reviews the sitespecific short-term χ/Q values determined by the applicant, and performs an

independent evaluation as described therein.

H. The staff performs an independent confirmatory radiological consequence

analysis using pertinent information in the applicant’s SAR to determine whether

the proposed site meets the radiological consequence evaluation factors

identified in 10 CFR 50.34(a)(1). For applications other than an ESP, the staff

also determines if the requirements of GDC 19 for maintaining the control room

in a safe condition are met. By performing an independent confirmatory

calculation, the staff will evaluate the reasonableness of the licensee’s analysis

model and results.

I. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the site are combined, and the calculated doses

are compared with the radiological consequence evaluation factors identified in

10 CFR Part 50.34(a)(1) at the nearest EAB and LPZ outer boundary stated in

the applicant’s SAR. 

J. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the site, including all sources of radiation

exposure to the control room personnel, are combined, and the calculated dose

in the control room is compared with the radiological consequence evaluation

factors identified in GDC 19.

K. For each postulated accident, the calculated doses from all postulated fission

product release pathways from the site, including all sources of radiation

exposure to the personnel in the technical support center, are combined, and the

calculated dose in the TSC is compared with the radiological consequence

evaluation factors identified for the control room of 5 rem TEDE for the duration

of the accident.
15.0.3-14 March 2007

L. For the methodology and assumptions for calculating the radiological

consequences, the staff will use the regulatory positions stated in Regulatory

Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design Basis

Accidents at Nuclear Power Reactors,” as applicable to the plant design. 

Additional information on the progression and assumptions for the failure of

small lines carrying coolant outside containment can be found in SRP Section

15.6.2. 

5. Review Procedures Specific to 10 CFR Part 52 Application Type

A. Early Site Permit Reviews: Subpart A to 10 CFR Part 52 specifies the

requirements and procedures applicable to the Commission’s review of an ESP

application for approval of a proposed site. Information required in an ESP

application includes a description of the site characteristics and design

parameters of the proposed site. The scope and level of detail of review of data

parallel that used for a CP review. 

In the absence of certain circumstances, such as a compliance or adequate

protection issue, 10 CFR 52.39 precludes the staff from imposing new site

characteristics, design parameters, or terms and conditions on the ESP at the

COL stage. Accordingly, the reviewer should ensure that all physical attributes

of the site that could affect the design basis of SSCs important to safety are

reflected in the site characteristics, design parameters, or terms and conditions

of the early site permit.

B. Standard Design Certification Reviews: DC applications do not contain general

descriptions of site characteristics because this information is site-specific and

will be addressed by the COL applicant. However, pursuant to 10 CFR

52.47(a)(1), a DC applicant must provide site parameters postulated for the

design. Site parameters associated with this SRP section are reviewed, as

applicable, to verify that: 

a. The postulated site parameters are representative of a reasonable

number of sites that have been or may be considered for a COL

application;

ii. The appropriate site parameters are included as Tier 1 information. This

convention has been used by previous DC applicants. Additional

guidance on site parameters is provided in SRP Section 2.0; 

iii. Pertinent parameters are stated in a site parameters summary table; and

iv. The applicant has provided a basis for each of the site parameters.

C. Combined License Reviews: For a COL application referencing a certified

standard design, the NRC staff reviews that application to ensure that sufficient

information is presented to demonstrate that the characteristics of the site fall

within the site parameters specified in the DC rule. Should the actual site

characteristics not fall within the certified standard design site parameters, the

COL applicant will need to demonstrate by some other means that the proposed
15.0.3-15 March 2007

facility is acceptable at the proposed site. This might be done by re-analyzing or

redesigning the proposed facility.

For a COL application referencing an ESP, NRC staff reviews the application to

ensure the applicant provides sufficient information to demonstrate that the

design of the facility falls within the site characteristics and design parameters

specified in the early site permit as applicable to this SRP section. In

accordance with 10 CFR 52.79(b)(2), should the design of the facility not fall

within the site characteristics and design parameters, the application shall

include a request for a variance from the ESP that complies with the

requirements of 10 CFR 52.39 and 10 CFR 52.93. 

In addition, long-term environmental changes and changes to the region

resulting from human or natural causes may have introduced changes to the site

characteristics that could be relevant to the design basis. In the absence of

certain circumstances, such as a compliance or adequate protection issue, 10

CFR 52.39 precludes the staff from imposing new site characteristics, design

parameters, or terms and conditions on the early site permit at the COL stage. 

Consequently, a COL application referencing an ESP need not include a

re-investigation of the site characteristics that have previously been accepted in

the referenced ESP. However, in accordance with 10 CFR 52.6, “Completeness

and Accuracy of Information,” the applicant or licensee is responsible for

identifying changes of which it is aware, that would satisfy the criteria specified in

10 CFR 52.39. Information provided by the applicant in accordance with

10 CFR 52.6(b) will be addressed by the staff during the review of a COL

application referencing an ESP or a DC.

For a COL application referencing either an ESP or DC or both, the staff should

review the corresponding sections of the ESP and DC FSER to ensure that any

early site permit conditions, restrictions to the DC, or COL action items identified

in the FSERs are appropriately handled in the COL application. 

IV. EVALUATION FINDINGS

The review should document the staff’s evaluation of the applicant’s design basis accident

radiological consequences analyses against the relevant regulatory criteria. The evaluation

should support the staff’s conclusions as to whether the regulations are met. The reviewer

should state what was done to evaluate the applicant’s submittal. The staff’s evaluation may

include verification that the applicant followed applicable regulatory guidance, performance of

independent calculations, and/or validation that the appropriate assumptions were made. The

reviewer may state that certain information provided by the applicant was not considered

essential to the staff’s review and was not reviewed by the staff. While the reviewer may

summarize or quote the information offered by the applicant in support of its application, the

reviewer should clearly articulate the bases for the staff’s acceptance and conclusions.

The reviewer verifies that the applicant has provided sufficient information and that the review

and calculations (if applicable) support conclusions of the following type to be included in the

staff's safety evaluation report. The reviewer also states the bases for those conclusions.
15.0.3-16 March 2007

A conclusion of the following type for the radiological consequence analyses will be included in

Section 15 of the site safety evaluation, standard design safety evaluation, or combined license

safety evaluation:

1. ESP application that references a standard reactor design certified by NRC. 

As set forth above, the staff has reviewed the site-specific short-term atmospheric

dispersion (χ/Q) values at the exclusion area boundary (EAB) and at the boundary of the

low population zone (LPZ) for the proposed site in the early site permit (ESP) application

and has verified that they are within the postulated design basis χ/Q values specified in

the [name of certified reactor design] design control document.

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [name] site, in conjunction with the engineered safety features as described in the

[name] certified standard design, are sufficient to provide reasonable assurance that the

total radiological consequences of the design basis accidents considered in the [name]

certified design will be within the radiological consequence evaluation factors of

10 CFR 50.34(a)(1).

[or:] 

As set forth above, the staff has reviewed the site-specific short-term χ/Q values at the

EAB and at the boundary of the LPZ for the proposed site in the ESP application and

found that they exceed the postulated design basis χ/Q values specified in the [name of

certified reactor design] design control document. However, the staff has verified that

the applicant has demonstrated that the radiological consequences associated with the

bounding DBAs using its site-specific χ/Q values meet the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1).

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [name] site, in conjunction with the engineered safety features as described in the

[name] certified standard design, are sufficient to provide reasonable assurance that the

total radiological consequences of the design basis accidents considered in the [name]

certified design will be within the radiological consequence evaluation factors of

10 CFR 50.34(a)(1).

2. ESP application that uses the PPE approach. As set forth above, the applicant

submitted its radiological consequence analyses using the site-specific short-term χ/Q

values and the plant parameter envelope (PPE) source term values and concluded that

the proposed site meets the radiological consequence evaluation factors identified in

Section 50.34(a)(1). The results of the applicant’s radiological consequence dose

calculation are provided in Table [ ], and the PPE values and the site-specific χ/Q values

used by the applicant and the staff are listed in Tables [ ] through [ ].

The staff reviewed the radiological consequence analyses submitted by the applicant

and finds that the PPE values that are inputs to these analyses are not unreasonable

based on information provided by the applicant, on the staff’s experience in evaluating

similar parameters, and on the staff’s confirmatory investigation and evaluation. 

To verify the applicant’s radiological consequence analyses, the staff performed its

confirmatory radiological consequence dose calculation using the site-specific χ/Q
15.0.3-17 March 2007

values and the PPE source term values provided by the applicant, and the staff finds

that its results are within the radiological consequence evaluation factors identified in

Section 50.34(a)(1). Although the staff performed its independent radiological

consequence dose calculation as a means of confirming the applicant’s results, the

staff’s approval of the ESP is based on the applicant’s analyses. 

Therefore, the staff concludes that the distances to the EAB and the LPZ outer

boundary of the [name] site, in conjunction with the source term and the fission product

release rates from the site to the environment provided by the applicant, are sufficient to

provide reasonable assurance that the total radiological consequences of the design

basis accidents will be within the dose evaluation factors set forth at

10 CFR 50.34(a)(1). This conclusion is subject to confirmation at the combined license

(COL) stage that the relevant design parameters specified by the applicant in the COL

application are bounded by the applicant’s PPE submitted with the ESP application.

3. ESP application that neither references a standard reactor design certified by NRC nor

uses the PPE approach. As set forth above, the applicant has selected and analyzed

the bounding design basis accidents and has determined that the total radiological

consequences of such accidents meet the radiological consequence evaluation factors

identified in 10 CFR 50.34(a)(1). The results of the applicant’s radiological

consequence dose calculation are provided in Table [ ]. 

The staff reviewed the radiological consequence analyses provided by the applicant and

has performed an independent analysis of the radiological consequences of each

design basis accident considered in the application using the site-specific χ/Q values at

the EAB and LPZ proposed in the ESP application. The staff finds that its results are

also within the radiological consequence evaluation factors identified in

10 CFR 50.34(a)(1). Although the staff performed an independent radiological

consequence dose calculation as a means of confirming the licensee’s results, the

staff’s approval of the ESP is based on the applicant’s analyses. Details of the staff’s

analyses are presented in Section [ ] of this safety evaluation report, and the results are

listed in Table [ ].

Therefore, the staff concludes that the distances to the EAB and the LPZ outer

boundary of the [site name] site, in conjunction with the source term and the fission

product release rates from the site to the environment provided by the applicant, are

sufficient to provide reasonable assurance that the total radiological consequences of

the design basis accidents will be within the dose evaluation factors set forth at

10 CFR 50.34(a)(1). This conclusion is based on the staff review of the applicant’s

analysis and on the staff’s independent analysis, which confirms that the calculated total

doses are within the dose evaluation factors set forth at 10 CFR 50.34(a)(1).

4. Standard reactor design certification application. As set forth above, the applicant has

selected and analyzed the bounding design basis accidents and has determined that the

total radiological consequences of such accidents meet the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1) and GDC 19 for the standard reactor

design, considering a reference site. The results of the applicant’s radiological

consequence dose calculation are provided in Table [ ]. 
15.0.3-18 March 2007

The staff reviewed the radiological consequence analyses provided by the applicant and

has performed an independent analysis of the radiological consequences of each

design basis accident considered in the application using the design reference χ/Q

values at the EAB, LPZ, and control room proposed in the application. The staff finds

that its results are also within the radiological consequence evaluation factors identified

in 10 CFR 50.34(a)(1) and GDC 19. Although the staff performed an independent

radiological consequence dose calculation as a means of confirming the licensee’s

results, the staff’s approval of the standard design is based on the applicant’s analyses. 

Details of the staff’s analyses are presented in Section [ ] of this safety evaluation

report, and the results are listed in Table [ ].

The staff performed a similar review of the applicant’s evaluation of the design basis

accident radiological consequences in the technical support center in support of the

emergency planning review. The staff has reasonable assurance that the dose in the

TSC will be within 5 rem TEDE. The details of the staff’s analysis is presented in

Section [ ] of this safety evaluation report, and the results are listed in Table [ ]. 

Therefore, the staff concludes that the plant features intended to mitigate the

radiological consequences of postulated design basis accidents, in conjunction with the

source term and the fission product release rates from the site to the environment

provided by the applicant, are sufficient to provide reasonable assurance that the total

radiological consequences of the design basis accidents will be within the dose

evaluation factors set forth at 10 CFR 50.34(a)(1) and GDC 19. This conclusion is

based on the staff review of the applicant’s analysis and on the staff’s independent

analysis, which confirms that the calculated total doses are within the dose evaluation

factors set forth at 10 CFR 50.34(a)(1) and GDC 19.

5. Combined license application without ESP or certified standard reactor design. As set

forth above, the applicant has selected and analyzed the bounding design basis

accidents and has determined that the total radiological consequences of such

accidents meet the radiological consequence evaluation factors identified in

10 CFR 50.34(a)(1) and GDC 19. The results of the applicant’s radiological

consequence dose calculation are provided in Table [ ]. 

The staff reviewed the radiological consequence analyses provided by the applicant and

has performed an independent analysis of the radiological consequences of each

design basis accident considered in the application using the site-specific χ/Q values at

the EAB, LPZ and control room proposed in the COL application. The staff finds that its

results are also within the radiological consequence evaluation factors identified in

10 CFR 50.34(a)(1) and GDC 19. Although the staff performed an independent

radiological consequence dose calculation as a means of confirming the licensee’s

results, the staff’s approval of the COL is based on the applicant’s analyses. Details of

the staff’s analyses are presented in Section [ ] of this safety evaluation report, and the

results are listed in Table [ ].

The staff performed a similar review of the applicant’s evaluation of the design basis

accident radiological consequences in the technical support center in support of the

emergency planning review. The staff has reasonable assurance that the dose in the

TSC will be within 5 rem TEDE. The details of the staff’s analysis is presented in

Section [ ] of this safety evaluation report, and the results are listed in Table [ ].
15.0.3-19 March 2007

Therefore, the staff concludes that the distances to the EAB and the LPZ outer

boundary of the [site name] site and plant features intended to mitigate the radiological

consequences of postulated design basis accidents, in conjunction with the source term

and the fission product release rates from the site to the environment provided by the

applicant, are sufficient to provide reasonable assurance that the total radiological

consequences of the design basis accidents will be within the dose evaluation factors

set forth at 10 CFR 50.34(a)(1) and GDC 19. This conclusion is based on the staff

review of the applicant’s analysis and on the staff’s independent analysis, which

confirms that the calculated total doses are within the dose evaluation factors set forth

at 10 CFR 50.34(a)(1) and GDC 19.

6. Combined license application with certified standard reactor design. As set forth above,

the staff has reviewed the site-specific short-term atmospheric dispersion (χ/Q) values

at the exclusion area boundary (EAB), at the boundary of the low population zone (LPZ),

in the technical support center (TSC) and in the control room for the proposed site in the

combined operating license (COL) application and has verified that they are within the

design reference set of χ/Q values specified in the [name of certified reactor design]

design control document.

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [name] site and plant features intended to mitigate the radiological consequences of

postulated design basis accidents, in conjunction with the engineered safety features as

described in the [name] certified standard design, are sufficient to provide reasonable

assurance that the total radiological consequences of the design basis accidents

considered in the [name] certified design will be within the radiological consequence

evaluation factors of 10 CFR 50.34(a)(1) and GDC 19. 

[or :] 

As set forth above, the staff has reviewed the site-specific short-term χ/Q values at the

EAB, at the boundary of the LPZ, in the TSC and in the control room for the proposed

site in the COL application and found that [name the χ/Q receptors that are not within

the DCD] exceed the design reference set of χ/Q values specified in the [name of

certified reactor design] design control document. However, the staff has verified that

the applicant has demonstrated that the radiological consequences associated with the

bounding DBAs using its site-specific χ/Q values meet the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1) and GDC 19.

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [site name] site and plant features intended to mitigate the radiological

consequences of postulated design basis accidents, in conjunction with the engineered

safety features as described in the [name] certified standard design, are sufficient to

provide reasonable assurance that the total radiological consequences of the design

basis accidents considered in the [name] certified design will be within the radiological

consequence evaluation factors of 10 CFR 50.34(a)(1) and GDC 19. 
15.0.3-20 March 2007

7. Combined license application with both ESP and certified standard reactor design. As

set forth above, the staff has verified that the site-specific short-term atmospheric

dispersion (χ/Q) values at the exclusion area boundary (EAB) and at the boundary of the

low population zone (LPZ) for the proposed site in the early site permit (ESP) and that

the site-specific short-term χ/Q values for the control room and technical support center

(TSC) are within the design reference set of χ/Q values specified in the [name of

certified reactor design] design control document. Therefore, the applicant has

demonstrated that the radiological consequences associated with the bounding DBAs

using its site-specific χ/Q values meet the radiological consequence evaluation factors

identified in 10 CFR 50.34(a)(1) and GDC 19.

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [site name] site and plant features intended to mitigate the radiological

consequences of postulated design basis accidents, in conjunction with the engineered

safety features as described in the [name] certified standard design, are sufficient to

provide reasonable assurance that the total radiological consequences of the design

basis accidents considered in the [name] certified design will be within the radiological

consequence evaluation factors of 10 CFR 50.34(a)(1) and GDC 19. 

[or:] 

As set forth above, the staff has reviewed the site-specific short-term χ/Q values at the

EAB and at the boundary of the LPZ for the proposed site in the ESP and the 

site-specific short-term χ/Q values for the control room and the technical support center

(TSC) and has found that [name the χ/Q receptors that are not within the DCD] exceed

the design reference set of χ/Q values specified in the [name of certified reactor design]

design control document. However, the staff has verified that the applicant has

demonstrated that the radiological consequences associated with the bounding DBAs

using its site-specific χ/Q values meet the radiological consequence evaluation factors

identified in 10 CFR 50.34(a)(1) and GDC 19.

Therefore, the staff concludes that the distance to the EAB and to the LPZ boundary of

the [name] site, in conjunction with the engineered safety features as described in the

[name] certified standard design, are sufficient to provide reasonable assurance that the

total radiological consequences of the design basis accidents considered in the [name]

certified design will be within the radiological consequence evaluation factors of

10 CFR 50.34(a)(1) and GDC 19.

8. Combined license application with ESP only. As set forth above, the applicant has

selected and analyzed the bounding design basis accidents using the site-specific shortterm atmospheric dispersion (χ/Q) values at the exclusion area boundary (EAB) and at

the boundary of the low population zone (LPZ) for the proposed site in the early site

permit (ESP), as well as site-specific short-term χ/Q values for the control room and

technical support center (TSC), and has determined that the total radiological

consequence of such accidents meets the radiological consequence evaluation factors

identified in 10 CFR 50.34(a)(1) and GDC 19. The results of the applicant’s radiological

consequence dose calculation are provided in Table [ ]. 
15.0.3-21 March 2007

The staff reviewed the radiological consequence analyses provided by the applicant and

has performed an independent analysis of the radiological consequences of each

design basis accident considered in the application using the EAB and LPZ χ/Q values

from the ESP and site-specific χ/Q values at the control room proposed in the COL

application. The staff finds that its results are also within the radiological consequence

evaluation factors identified in 10 CFR 50.34(a)(1) and GDC 19. Although the staff

performed its independent radiological consequence dose calculation as a means of

confirming the licensee’s results, the staff’s approval of the COL is based on the

applicant’s analyses. Details of the staff’s analyses are presented in Section [ ] of this

safety evaluation report, and the results are listed in Table [ ].

The staff performed a similar review of the applicant’s evaluation of the design basis

accident radiological consequences in the technical support center in support of the

emergency planning review. The staff has reasonable assurance that the dose in the

TSC will be within 5 rem TEDE. The details of the staff’s analysis is presented in

Section [ ] of this safety evaluation report, and the results are listed in Table [ ].

Therefore, the staff concludes that the distances to the EAB and the LPZ outer

boundary of the [site name] site and plant features intended to mitigate the radiological

consequences of postulated design basis accidents, in conjunction with the source term

and the fission product release rates from the site to the environment provided by the

applicant, are sufficient to provide reasonable assurance that the total radiological

consequences of the design basis accidents will be within the dose evaluation factors

set forth at 10 CFR 50.34(a)(1) and GDC 19. This conclusion is based on the staff

review of the applicant’s analysis and on the staff’s independent analysis, which

confirms that the calculated total doses are within the dose evaluation factors set forth

at 10 CFR 50.34(a)(1) and GDC 19.

For DC and COL reviews, the findings will also summarize the staff’s evaluation of

requirements and restrictions (e.g., interface requirements and site parameters) and COL

action items relevant to this SRP section.

V. IMPLEMENTATION

The staff will use this SRP section in performing safety evaluations of DC applications and

license applications submitted by applicants pursuant to 10 CFR Part 50 or 10 CFR Part 52. 

Except when the applicant proposes an acceptable alternative method for complying with

specified portions of the Commission’s regulations, the staff will use the method described

herein to evaluate conformance with Commission regulations.

The provisions of this SRP section apply to reviews of applications submitted six months or

more after the date of issuance of this SRP section, unless superseded by a later revision. 

VI. REFERENCES

1. 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities”

2. 10 CFR 50.34, “Contents of applications; technical information”
15.0.3-22 March 2007

3. 10 CFR 50, Appendix A, “General Design Criteria for Nuclear Power Plants”

4. 10 CFR 50, Appendix E, “Emergency Planning and Preparedness for Production and

Utilization Facilities”

5. 10 CR Part 52, “Early Site Permits; Standard Design Certifications; and Combined

Licenses for Nuclear Power Plants”

6. 10 CFR Part 100, “Reactor Site Criteria”

7. Regulatory Guide RG 1.206, “Combined License Applications for Nuclear Power Plants

(LWR Edition),” 2007.

8. Regulatory Guide 1.183, “Alternative Radiological Source Terms for Evaluating Design

Basis Accidents at Nuclear Power Reactors,” July 2000

9. NUREG-1793, “Final Safety Evaluation Report Related to Certification of the AP1000

Standard Design,” September 2004

10. NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants,”

February 1995

11. NRR Review Standard RS-002, “Processing Applications for Early Site Permits,”

May 3, 2004 (ADAMS Accession No. ML040700094

12. SECY-98-154, “Results of the Revised (NUREG-1465) Source Term Re-Baselining for

Operating Reactors,” June 1998

 

PAPERWORK REDUCTION ACT STATEMENT

The information collections contained in the Standard Review Plan are covered by the requirements of 10 CFR Part 50 and

10 CFR Part 52, and were approved by the Office of Management and Budget, approval number 3150-0011 and 3150-0151. 

PUBLIC PROTECTION NOTIFICATION

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information

collection requirement unless the requesting document displays a currently valid OMB control number.