NUREG-0800

August 2006

USNRC STANDARD REVIEW PLAN

This Standard Review Plan, NUREG-0800, has been prepared to establish criteria that the Office of Nuclear Reactor Regulation 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 the Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition)." Not all sections of the standard

format have a corresponding review plan section. The SRP sections applicable to a combined license application for a new light-water reactor (LWR) will be based on Regulatory Guide DG-1145, "Combined License Applications for Nuclear Power Plants

(LWR Edition)," as superceded by the final guide, until the SRP itself is updated.

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 draft or active 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 # ML062210398.

U.S. NUCLEAR REGULATORY COMMISSION

STANDARD REVIEW PLAN

OFFICE OF NUCLEAR REACTOR REGULATION

14.2.1 GENERIC GUIDELINES FOR EXTENDED POWER UPRATE TESTING PROGRAMS

This Standard Review Plan (SRP) section provides general guidelines for reviewing proposed

extended power uprate (EPU) power ascension testing programs. This review provides

assurance that the proposed testing programs adequately demonstrate that plant structures,

systems, and components (SSCs) important to safety that are affected by the proposed power

uprate will perform satisfactorily in service at the proposed uprated power level.

There are three types of power uprates. Measurement uncertainty recapture power uprates

involve power increases of less than 2 percent, and are achieved by implementing enhanced

techniques for calculating reactor power. Stretch power uprates typically involve power

increases of up to 7 percent and do not generally involve major plant modifications. EPUs

involve power increases greater than stretch power uprates and have been approved for

increases as high as 20 percent. EPUs usually involve significant modifications to major

balance of plant (BOP) equipment. A power uprate is classified as an EPU based on a

combination of the magnitude of the proposed power increase and the plant modifications

necessary to support the requested uprate. This SRP applies only to EPU license amendment

requests (LARs).

REVIEW RESPONSIBILITIES

Primary - Organization responsible for the review of quality assurance

Secondary - Organization responsible for the review of probabilistic risk assessment

Organization responsible for the review of human performance
August 2006 14.2.1 - 2

Additional secondary review responsibilities are detailed in Table 1, “Typical

Steady-State Power Ascension Testing Applicable to Extended Power

Uprates,” to this SRP section

I. AREAS OF REVIEW

The primary reviewer coordinates the review of the overall power uprate testing program. 

Secondary review branches are responsible for reviewing EPU applications to provide

assurance that the licensee has proposed an EPU testing program that in conjunction with plant

operating experience, computer modeling, and analyses, adequately demonstrates that SSCs

will perform satisfactorily in service at the requested increased plant power level. Secondary

review branches will assist in the review of proposed testing plans and acceptance criteria, as

needed. The review of EPU testing programs should be performed in conjunction with staff

reviews of other aspects of the EPU LAR.

II. ACCEPTANCE CRITERIA

The EPU test program acceptance criteria are based on meeting the relevant requirements of

the following Commission regulations:

A. General Design Criterion (GDC) 1, “Quality standards and records,” of 10 CFR Part 50

Appendix A, “General Design Criteria for Nuclear Power Plants,”, insofar as it applies to

the specific plant, requires that SSCs important to safety be tested to quality standards

commensurate with the importance of the safety functions to be performed and that a

quality assurance program be established and implemented to provide reasonable

assurance that these SSCs will satisfactorily perform their safety functions. 

B. Criterion XI, “Test Control,” of 10 CFR Part 50 Appendix B, “Quality Assurance Criteria for

Nuclear Power Plants and Fuel Reprocessing Plants,” insofar as it applies to the specific

plant requires establishment of a test program to assure that testing required to

demonstrate that SSCs will perform satisfactorily in service is identified and performed in

accordance with written test procedures which incorporate the requirements and

acceptance limits contained in applicable design documents.

C. 10 CFR 50.34, “Contents of applications; technical information,” specifies requirements

for the content of the original operating license application, including the requirement that

the final safety analysis report (FSAR) include plans for preoperational testing and initial

operations.

Technical Rationale

This review provides assurance that the proposed EPU test program, along with plant operating

experience, computer modeling, and analyses, adequately demonstrates that SSCs will perform

satisfactorily at EPU conditions. In particular, the EPU application and supporting justification

should provide reasonable assurance that (1) any power-uprate-related modifications to the

facility have been adequately constructed and implemented and (2) the facility can be operated

at the proposed EPU conditions in accordance with design requirements and in a manner that

will not endanger the health and safety of the public.
14.2.1 - 3 August 2006

The following paragraphs describe the technical rationale for application of the above

acceptance criteria to the review of EPU test programs:

A. GDC 1 establishes the testing necessary to demonstrate SSCs important to safety

function in a manner that provides reasonable assurance that the facility can be operated

without undue risk to the health and safety of the public. Specifically, GDC 1 requires that

SSCs important to safety be designed, fabricated, erected, and tested to quality

standards commensurate with the importance of the safety functions to be performed. 

Where generally recognized codes and standards are used, they shall be identified and

evaluated to determine their applicability. Additionally, a quality assurance program

(QAP) shall be established to ensure that SSCs will satisfactorily perform their safety

functions.

Application of GDC 1 to the EPU test program (to the extent that GDC 1 is applicable)

ensures testing is performed, as necessary, to provide assurance that SSCs meet their

design specifications and capabilities. The QAP ensures proper documentation and

traceability that applicable testing was accomplished, and codes and standards were

satisfied. 

Throughout this SRP section, the term “important to safety” is used to refer to SSCs to

which this EPU testing guidance applies. Generic Letter (GL) 84-01, “NRC Use of the

Terms ‘Important to Safety’ and ‘Safety Related,’” indicates that “important to safety”

generally refers to plant equipment needed to meet the provisions of the GDC. However,

as discussed in Section 2.1.5.6 of LIC-100, “Control of Licensing Basis for Operating

Reactors,” the GDC are not applicable to plants with construction permits issued before

May 21, 1971. Each plant licensed before the GDC were formally adopted was evaluated

on a plant-specific basis, determined to meet plant-specific principal design criteria

(PDC), and licensed by the Commission. For the purposes of the EPU test program

review, SSCs should be considered to be important to safety based on the plant-specific

PDC.

B. Criterion XI of Appendix B to 10 CFR Part 50 requires that a test program be established

to assure that all testing required to demonstrate that SSCs will perform satisfactorily in

service is identified and performed in accordance with written test procedures which

incorporate the requirements and acceptance limits contained in applicable design

documents. The test program is required to include, as appropriate, proof tests prior to

installation, preoperational tests, and operational tests of SSCs. Test procedures are

required to include provisions for assuring that all prerequisites for the given test have

been met, that adequate test instrumentation is available and used, and that the test is

performed under suitable environmental conditions. Test results are required to be

documented and evaluated to assure that test requirements have been satisfied.

Application of Criterion XI of 10 CFR Part 50, Appendix B, to the EPU test program will

provide reasonable assurance that SSC capabilities to perform specified functions are not

adversely impacted by increasing the maximum allowed power level. Specifically, the

EPU test program, in conjunction with operating experience and design control measures

(e.g., computer modeling and analyses), provides reasonable assurance that SSCs

important to safety will perform satisfactorily in service following the proposed power

uprate. The EPU test program also assures that deficiencies are identified and corrected
1

Note: The SRP is not a substitute for the NRC’s regulations, and compliance with specific

acceptance criteria is not required.

August 2006 14.2.1 - 4

and that testing activities are conducted in a manner which minimizes operational reliance

on untested safety functions. Regulatory Guide (RG) 1.68, “Initial Test Programs for

Water-Cooled Nuclear Power Plants,” Revision 2, describes the general scope and depth

of initial test programs that the NRC staff found acceptable during the review of original

operating license applications. The SSCs subject to initial testing performed safety

functions that included fission product containment; reactivity monitoring and control

reactor safe shutdown (including maintaining safe shutdown) core cooling accident

prevention and consequence mitigation as specified in the design and credited in safety

analyses.

C. 10 CFR 50.34(b)(6)(iii) requires that each application for a license to operate a facility

include in the FSAR plans for preoperational testing and initial operations. The initial test

program (which includes preoperational testing and testing during initial operation) verifies

that SSCs are capable of performing their safety functions as specified in the design and

credited in safety analyses.

Application of 10 CFR 50.34(b)(6)(iii) to the EPU test program ensures that the licensee

submits adequate information, commitments, and plans demonstrating that the proposed

EPU does not invalidate the testing provisions for the original design basis. The

preoperational testing and the initial startup testing invalidated by the requested increase

in power level are evaluated and repeated as necessary to demonstrate that SSCs are

capable of performing their safety functions. This provides assurance that operation at

the requested higher power level will be within the bounds of the design and safety

analyses and that EPU testing activities will be conducted in a sequence and manner

which minimizes operational reliance on untested SSCs or safety functions. 

III. REVIEW PROCEDURES

The purpose of this review is to obtain reasonable assurance that the proposed EPU testing

program adequately controls the initial power ascension to the proposed EPU power level. The

EPU test program should include steady-state and transient performance testing sufficient to

demonstrate, in conjunction with plant operating experience, computer modeling, and analyses,

that SSCs will perform satisfactorily at the requested power level. The proposed EPU test

program should be based on a systematic review of the initial plant test program to identify initial

licensing power ascension testing that may be invalidated by the requested EPU. Additionally,

the EPU test program should include appropriate testing to demonstrate that EPU-related plant

modifications have been adequately implemented. Specific guidance1

 for assessing licensees’

justifications for not performing certain tests, including large transient tests, is provided in

Section III.C below.
14.2.1 - 5 August 2006

A. Comparison of Proposed EPU Test Program to the Initial Plant Test Program

1. General Discussion

The licensee should compare the proposed EPU testing program to the original test

program carried out during initial plant licensing. The scope of this comparison

should include (1) all initial power ascension tests performed at a power level of

equal to or greater than 80 percent of the original licensed thermal power level; and

(2) initial power ascension tests that were performed at lower power levels if the

EPU would invalidate the test results. The licensee should either repeat the initial

power ascension tests within the scope of this comparison or provide justification

for deviations that are proposed.

2. Specific Acceptance Criteria

The primary reviewer, with assistance as necessary from the secondary review

branches, will determine if the licensee has adequately identified the following in

the EPU LAR:

• all initial power ascension tests performed at a power level of equal to or

greater than 80 percent of the original licensed thermal power level;

• all initial test program tests performed at power levels lower than 80 percent

of the original licensed thermal power level that would be invalidated by the

EPU; and,

• differences between the proposed EPU test program and the portions of the

initial test program identified by the previous criteria.

 The reviewer should refer to the plant-specific testing identified in FSAR Chapter

14.2, “Initial Plant Test Program” (or the equivalent FSAR section for nonstandard

format plants), and startup test reports, if available, to verify that the licensee has

adequately identified the scope of the initial plant test program. Additionally,

Table 1, “Typical Steady-State Power Ascension Testing Applicable to Extended

Power Uprates,” and Table 2, “Typical Transient Testing Applicable to Extended

Power Uprates,” to this SRP section provide a generic summary of power

ascension tests typically performed at or near full power.

The reviewer should verify that the licensee has identified and adequately justified

any proposed differences from or exceptions to the original test program. The

reviewer should refer to Section III.C below for guidance on assessing the

adequacy of justifications for proposed differences and exceptions.
August 2006 14.2.1 - 6

B. Power Ascension Test Considerations for Plant Modifications

1. General Discussion

EPUs may involve significant modifications to major BOP equipment, in addition to

setpoint adjustments and operating parameter changes. Therefore, within its

respective technical area, each secondary review branch will assess if the licensee

adequately evaluated and established appropriate power ascension testing

requirements to address the aggregate impact of EPU plant modifications, setpoint

adjustments, and parameter changes on the dynamic response of the plant

(including primary and secondary plant response, as well as the response of SSCs

important to safety) to anticipated operational occurrences. The objective of this

review is to provide assurance that the nature and extent of plant modifications,

setpoint adjustments, and parameter changes were adequately considered in the

development of the EPU test program.

The power ascension testing described in RG 1.68, Appendix A, Section 5, “Power

Ascension Tests,” focuses on testing that is necessary to demonstrate that the

plant will operate in accordance with design specifications both during normal

steady-state conditions and, to the extent practical, during and following anticipated

operational occurrences. Anticipated operational occurrences are conditions of

normal operation that are expected to occur one or more times during the life of the

plant and include events such as loss of all offsite power, tripping of the main

turbine generator set, and loss of power to all recirculation pumps.

The reviewer is not expected to evaluate the specific component-and system-level

testing provisions for each plant modification, parameter change, or setpoint

adjustment. Based on previous experience, testing required by technical

specifications (TSs) and existing 10 CFR Part 50, Appendix B, QAPs can be relied

upon to adequately demonstrate individual system or component performance

characteristics. Therefore, this review is intended to assure that plant equipment

important to safety that supports functions that rely on the integrated operation of

multiple SSCs following an anticipated operational occurrence are adequately

demonstrated prior to extended operation at the requested EPU power level.

2. Specific Acceptance Criteria

Based on the review of the licensee’s EPU LAR, the reviewer will determine if the

licensee has adequately identified the following:

• plant modifications and setpoint adjustments that are being made and

• changes in primary and secondary system operating parameters (such as

reactor coolant temperature, pressure, and flow rate; main steam and main

feedwater flow rates; and reactor vessel, pressurizer, and steam generator

(SG) levels).

The reviewer should assess if the licensee adequately identified SSCs important to

safety that are affected by EPU-related modifications, setpoint adjustments, and
14.2.1 - 7 August 2006

changes in plant operating parameters. In particular, the licensee should have

considered the impact of first-of-a-kind plant modifications, the introduction of new

system dependencies or interactions, and changes the effects of the steady-state

and transient plant response have on the capability of SSCs important to safety to

perform their specified functions. To assist in this review, Table 2 lists general

acceptance criteria for transient testing for typical anticipated operational

occurrences.

The reviewer should verify that the proposed EPU test program adequately

demonstrates the performance of SSCs important to safety that meet any of the

following criteria: (1) the performance of the SSC is impacted by EPU-related

modifications, (2) the SSC is used to mitigate an anticipated operational occurrence

described in the plant-specific design basis, and (3) performance of the SSC can be

affected by integrated plant operation or transient conditions. Important to safety

SSCs that meet any of these criteria should be subject to appropriate testing

requirements in the EPU test program to confirm that the SSCs will continue to

function in accordance with their design criteria during steady-state and transient

operating conditions, adequately tested by overlapping individual component-or

system-level tests, or adequately justified as an exception to testing in accordance

with the provisions discussed in Section III.C.

C. Justification for Eliminating EPU Power ascension Tests

1. General Discussion

The licensee may establish a test program that does not include some or all of the

tests referred to above in Sections III.A and III.B that would normally be performed,

provided that proposed exceptions are adequately justified in accordance with the

criteria provided in Section III.C.2. Each secondary review branch will verify and

document the adequacy of the licensee’s justification for test exceptions that are

within the branch’s technical area of review.

2. Specific Acceptance Criteria

If the licensee proposes not to perform a test (or tests) that would normally be

included in the EPU test program pursuant to the criteria contained in Sections III.A

and III.B above, the reviewer should provide assurance that sufficient justification

has been provided. Pursuant to 10 CFR Part 50, Appendix B, the proposed EPU

test program must demonstrate that SSCs will perform satisfactorily in service. The

reviewer should consider the following factors when assessing the adequacy of the

licensee’s justification:

a. Power Uprate Operating Experience

If the licensee proposes to omit a test (or tests) from the EPU testing program

based on favorable power uprate operating experience, the reviewer should

determine the applicability of the operating experience to the specific plant

requesting the EPU. The reviewer should consider similarities in plant design

and equipment; uprated power level; plant modifications, setpoint
August 2006 14.2.1 - 8

adjustments and parameter changes; test specifications and methods; and

operating pocedures and EOPs. Additionally, the reviewer should verify that

the licensee adequately considered operating experience of problems,

malfunctions, or other unexpected consequences from previous power

uprates. 

b. Introduction of New Thermal-Hydraulic Phenomena or Identified System

Interactions

The reviewer should provide assurance that the licensee adequately

addressed the effects of any new thermal-hydraulic phenomena or system

interactions that may be introduced as a result of the EPU.

c. Facility Conformance to Limitations Associated With Computer Modeling and

Analytical Methods

If the licensee proposes to omit a test (or tests) from the EPU testing program

based on the results of computer modeling and analytical methods (simply

referred to as analytical methods for the purposes of this discussion), the

reviewer should confirm that plant modifications, setpoint adjustments, and

parameter changes that have been or will be made do not tend to invalidate

the analytical methods being used. In this case, the licensee is expected to

discuss and fully justify the use of analytical methods for this purpose. 

Considerations include, but are not limited to, the nature, capability,

applicability, accuracy, and sensitivity of the analytical methods being used,

including sensitivities associated with extrapolating the use of these methods

to encompass EPU conditions; conformance to any limitations that apply; the

extent that the analytical methods have been updated and corrected to reflect

plant behavior following specific plant transients that have occurred; SSCs

important to safety that are not modeled and why it is not necessary to

include these SSCs in the model to adequately predict transient plant

behavior; the impact of plant modifications, setpoint adjustments, and

parameter changes on the validity, accuracy, and sensitivity of the analytical

methods being used; and measures in the test program that will confirm the

validity, accuracy, and sensitivity of the analytical results. Where existing

information is insufficient to demonstrate the adequacy and applicability of

analytical models and methods that are used for evaluating EPU transient

plant behavior, licensees may be able to demonstrate acceptability by

performing transient tests that are suitable for addressing the specific

analytical shortcomings. Transient tests that are performed for this purpose

need not be limited to the tests in the original startup test program.
14.2.1 - 9 August 2006

d. Plant Operator Familiarization With Facility Operation and Trial Use of

Operating and Emergency Operating Procedures

Plant modifications, setpoint adjustments, and parameter value changes, in

conjunction with the increased decay heat generation associated with higher

power operation, can impact the execution of abnormal plant procedures and

EOPs. For example, the EPU may change the timing and sequence of

significant operator actions that are relied upon in abnormal plant procedures

and EOPs, or it could impact the accident mitigation strategies that are

credited in abnormal or emergency operating procedures.

For each EPU LAR, the technical branch responsible for operator licensing

and human performance should review the impact of the requested power

uprate on operator training and human factors in accordance with separate

SRP guidance. These reviews should include an evaluation of the changes

in operator actions, procedures, and training (including necessary changes to

the control room simulator) resulting from the EPU. The initial test program

objectives, as described in RG 1.68, Revision 2, included plant staff

familiarization with facility operation and trial use of abnormal plant

procedures and EOPs. However, based on previous review experience, it is

not expected that power ascension testing will normally be performed solely

for procedure verification or operator familiarization. Nonetheless, the

possibility exists that the review of operator training (including simulator

training) and human factors aspects of the EPU could indicate a need to

perform power ascension testing for procedure verification or operator

familiarization.

e. Minimal Reductions in the Margin of Safety

The licensee’s justification for omitting a test (or tests) from the test program

may consider the extent of the EPU’s impact on the results of the safety

analyses and may consider such factors as the amount of remaining margin

to SRP acceptance criteria, the sensitivity of the safety analysis results to

changes in analysis assumptions, and the capability of transient testing to

provide useful confirmatory data. To aid in this review, the information

provided in Table 2 to this SRP section includes a reference to the safety

analysis SRP sections related to typical power ascension transient tests, as

applicable.

Although the initial test program objectives, as described in RG 1.68,

Revision 2, included validation of analytical models and verification of

assumptions used for predicting plant response to anticipated transients and

postulated accidents, transient testing is not required expressly to validate

accident analysis codes that are used for EPU license amendment reviews. 

The applicability and use of accident analysis codes for an EPU is reviewed

by the staff in accordance with separate SRP guidance.
August 2006 14.2.1 - 10

f. Guidance Contained in Vendor Topical Reports

The NRC previously reviewed and accepted General Electric (GE) Company

Licensing Topical Report, “Generic Guidelines for General Electric Boiling

Water Reactor Extended Power Uprate” (referred to as ELTR-1), NEDC32424P-A, Class III, February 1999, as a basis for reviewing EPU LARs. 

This topical report provided specific guidance on integrated system transient

testing at EPU conditions. As described in Section 5.11.9.d and Appendix

L.2.4 of ELTR-1, the generator load rejection and the main steam isolation

valve (MSIV) closure tests verify that the plant performance is as predicted

and projected from previous test data.

On March 31, 2003, the NRC approved the use of GE Licensing Topical 

Report NEDC-33004P, “Constant Pressure Power Uprate,” for constant

pressure power uprate (CPPU) EPU LARs. However, as noted in the staff’s

safety evaluation (SE), the staff did not accept the elimination of largetransient testing (e.g., the MSIV closure and turbine generator load rejection

described in NEDC-32424P-A) from the scope of the CPPU test program. 

The staff noted that the need to conduct large-transient testing for a CPPU

would be considered on a plant-specific basis. 

For pressurized water reactors (PWRs), Westinghouse report WCAP-10263,

“A Review Plan for Uprating the Licensed Power of a Pressurized Water

Reactor Plant,” provides limited guidance for power uprate testing. 

Specifically, the document states that the recommended test program for the

nuclear steam supply system and the interfacing BOP systems should be

developed on a plant-specific basis depending on the magnitude of hardware

modifications and the magnitude of the power uprate.

Although the NRC has previously approved certain exceptions to EPU test

program requirements, the applicability of these exceptions to a particular

plant must be considered and justified on a plant-specific basis as discussed

above in Section III.C.2.a.

g. Risk Implications

For cases in which the licensee proposes a risk-informed basis for not

performing certain transient tests, SPSB should be consulted to assist in the

review. Risk-informed justifications for not performing power ascension tests

should be carefully weighed against the potential benefits of performing the

tests. In addition to the risks inherent in initiating a plant transient, the review

should also consider the benefit of identifying potential latent equipment

deficiencies or other plant problems under the controlled circumstances

afforded by the test program. In any case, a risk-informed justification should

not be used as the sole basis for not performing the power ascension tests

that are referred to above in Sections III.A and III.B.
14.2.1 - 11 August 2006

If the licensee provides adequate justification for not performing certain

power ascension tests, the staff may conclude that the EPU test program is

acceptable without the performance of these tests.

D. Evaluate the Adequacy of Proposed Testing Plans

1. General Discussion

The EPU LAR should include plans for the initial approach to the increased EPU

power level and steady-state testing that will be used to verify that the reactor plant

operates within the values of EPU design parameters.

2. Specific Acceptance Criteria

For each EPU power ascension test proposed by the licensee, the staff will review

the test objectives, the summary of prerequisites and test methods, and specific

acceptance criteria to establish that the functional adequacy of SSCs is verified. 

This review assures that the test objectives, test methods, and acceptance criteria

are acceptable and consistent with the design basis for the facility.

Each secondary review branch will review the licensee’s plans for the EPU test

program within the branch’s technical area. The licensee’s EPU test program

should include the following provisions:

• The initial approach to the uprated EPU power level should be incremental

and should include steady-state power hold points to evaluate plant

performance above the original licensed full-power level.

• The licensee should propose appropriate testing and acceptance criteria to

ensure that the plant responds within design predictions. The predicted

responses should be developed using real or expected values of items such

as beginning-of-life core reactivity coefficients, flow rates, pressures,

temperatures, response times of equipment, and the actual status of the

plant, not the values or plant conditions used for conservative evaluations of

postulated accidents.

• Contingency plans should be implemented in the event the plant does not

respond as predicted.

• The test program should be scheduled and sequenced to minimize the time

untested SSCs important to safety are relied upon during operation above the

original licensed full-power level. Safety-related SSCs relied upon during

operation shall be verified to be operable in accordance with existing TS and

QAP requirements.

To assist in this review, Tables 1 and 2 provide generic lists of full-power steadystate and transient tests and related acceptance criteria that are potentially

applicable to an EPU test program.
August 2006 14.2.1 - 12

If a power ascension test is needed to demonstrate that SSCs will perform

satisfactorily in service at EPU conditions, the reviewer should determine if a

license condition is needed to ensure that this testing is performed in a timely and

controlled manner.

IV. EVALUATION FINDINGS

When the EPU LAR review is complete and the reviewer has determined that it is satisfactory

and in accordance with the acceptance criteria in Section II above, a statement similar to the

following should be provided in the staff's safety evaluation.

The NRC staff has reviewed the EPU test program in accordance with SRP

Section 14.2.1. This review included an evaluation of (1) plans for the initial approach to

the proposed maximum licensed thermal power level, including verification of adequate

plant performance; (2) transient testing that is deemed appropriate and necessary to

demonstrate that plant equipment will perform satisfactorily at the proposed increased

maximum licensed thermal power level; (3) proposed exceptions to the power ascension

test referred to in Sections III.A and III.B; and (4) the test program’s conformance with

applicable regulations. For the reasons set forth above, the NRC staff concludes that the

proposed EPU test program provides reasonable assurance that the plant will operate in

accordance with the applicable design criteria and that SSCs affected by the EPU or

modified to support the proposed power increase will perform satisfactorily in service. On

this basis, the NRC staff finds that there is reasonable assurance that the EPU test

program satisfies the applicable requirements discussed in Section II of SRP 14.2.1.

V. IMPLEMENTATION

This SRP section will be used by the NRC staff for recommendations of EPU LARs submitted

pursuant to 10 CFR 50.90. This SRP is not intended to be used in place of plant-specific design

bases to assess the acceptability of an EPU application. The applicability of this SRP is

determined on a plant-specific basis consistent with the design basis of the plant.

In addition, where the NRC has approved a specific methodology (e.g., topical report) for the

type of power uprate being requested, the licensee should follow the format prescribed for that

specific methodology and provide the information called for in that methodology and the NRC's

letter and SE approving the methodology. Except when the applicant proposes an acceptable

alternative method for complying with specified portions of the Commission's regulations, the

method described herein will be used by the NRC staff in evaluating conformance with

Commission regulations.

VI. REFERENCES

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

2. SECY-01-0124, “Power Uprate Application Reviews,” dated July 9, 2001. The related staff

requirements memorandum is dated May 24, 2001.
14.2.1 - 13 August 2006

3. General Electric Company Licensing Topical Report, “Generic Guidelines for General

Electric Boiling Water Reactor Extended Power Uprate,” (ELTR-1), NEDC-32424P-A,

Class III, February 1999.

4. General Electric Company Licensing Topical Report, “Generic Evaluations of General

Electric Boiling Water Reactor Extended Power Uprate,” (ELTR-2), NEDC-32523P-A,

Class III, February 2000, and Supplement 1, Volumes I and II.

5. General Electric Company Licensing Topical Report, “Constant Pressure Power Uprate,”

NEDC-33004P, Revision 1, July 2001.

6. NRC Regulatory Guide 1.68, “Initial Test Programs for Water-Cooled Nuclear Power

Plants,” Revision 2, August 1978.

7. NRR Office Instruction LIC-100, “Control of Licensing Basis for Operating Reactors.”

8. NRR Office Instruction LIC-101, “License Amendment Review Procedures.”

9. NRR Office Instruction LIC-500, “Processing Requests for Reviews of Topical Reports.”

10. Westinghouse WCAP-10263, “A Review Plan for Uprating the Licensed Power of a

Pressurized Water Reactor Power Plant,” January 1983.

11. NRC Inspection Manual, Part 9900, “10 CFR Part 50.59, Changes, Tests, and

Experiments,” Change Notice Number 01-008.

12. NRC Information Notice 2002-26, “Failure of Steam Dryer Cover Plate After a Recent

Power Uprate,” September 11, 2002.

PAPERWORK REDUCTION ACT STATEMENT

The information collections contained in the Standard Review Plan are covered by the

requirements of 10 CFR Part 50.54, which were approved by the Office of Management and

Budget, approval number 3150 - 0011.

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.
August 2006 14.2.1 - 14

Table 1: Typical Steady-State Power Ascension Testing Applicable to Extended Power Uprates

Power Ascension

Test

Reference Recommended Initial

Conditions

Typical Test Acceptance Criteria Primary Technical Review

Branch

Conduct vibration

testing and

monitoring of

reactor vessel

internals and

reactor coolant

system (RCS)

components

RG 1.68, App A

4.s, 5.p

Lowest practical power

level

Reactor vessel and RCS

component vibration

characteristics within design. 

(See NRC Information Notice

2002-26 and RG 1.20) 

RCS components, reactor

vessel internals and piping,

and mechanical engineering

Measure power

reactivity

coefficients

(PWR) or power

vs. flow

characteristics

(BWR)

RG 1.68, App A

5.a

100% of rated thermal

power (RTP)

Characteristics in accordance

with design

reactor systems and core

performance

Steady-state core

performance

RG 1.68, App A

5.b

100% of RTP Characteristics in accordance

with design

reactor systems and core

performance

Control rod

patterns

exchange 

RG 1.68, App A

5.c

Power equal to highest

power level for rod

exchanges allowed at

power

Core limits not exceeded reactor systems and core

performance

Control rod

misalignment

testing

RG 1.68, App A

5.i

100% of RTP

Rod misalignment equal to

or less than TS limits

Demonstrate ability to detect

misalignment

reactor systems and core

performance

Failed fuel

detection system

RG 1.68, App A

5.q

100% of RTP Verify proper operation health physics

Plant process

computer 

RG 1.68, App A

5.r

100% of RTP Inputs and calculation are correct balance-of-plant/

instrumentation and controls 

Calibrate major

plant control

systems 

RG 1.68, App A

5.s

100% of RTP Verify performance reactor systems and core

performance/

balance-of-plant

Main steam and

main feedwater

system operation

RG 1.68, App A

5.v

100% of RTP Operate in accordance with

design performance requirements

balance-of-plant

Shield and

penetration

cooling systems

RG 1.68, App A

5.w

100% of RTP Maintain temperature within

design limits

balance-of-plant

Engineered

Safety Features

auxiliary and

environmental

systems 

RG 1.68, App A

5.x

100% of RTP Capable of performing design

functions

balance-of-plant

Calibrate systems

used to

determine reactor

thermal power

RG 1.68, App A

5.y

100% of RTP Verify performance instrumentation and controls 

Chemical and

radiochemical

control systems

RG 1.68, App A

5.a.a

100% of RTP Control systems function in

accordance with design 

health physics

Sample RCSs

and secondary

coolant systems

RG 1.68, App A

5.a.a

100% of RTP Chemistry limits are not

exceeded

chemical engineering
Power Ascension

Test

Reference Recommended Initial

Conditions

Typical Test Acceptance Criteria Primary Technical Review

Branch

14.2.1 - 15 August 2006

Radiation surveys RG 1.68, App A

5.b.b

100% of RTP Verify shielding adequacy and

identify 10 CFR Part 20 highradiation zones

health physics

Ventilation

systems

(including primary

containment and

steamline tunnel)

RG 1.68, App A

4.j and 5.f.f

100% of RTP Maintain service areas within

design limits

containment and ventilation

Acceptability of

reactor internals,

piping, and

component

movement,

vibrations, and

expansions

RG 1.68, App A

1.a.1, 1.a.3, 1.e.,

and 5.o.o

Lowest practical power

level

Parameters within design values RCS components, reactor

vessel internals and piping,

and mechanical engineering
August 2006 14.2.1 - 16

Table 2: Typical Transient Testing Applicable to Extended Power Uprates

Transient Test Reference Typical Reactor Plant Initial

Conditions

Typical Transient Test Acceptance Criteria Associated 

Standard Review Plan

Accident Analyses Section 

(as applicable)

Dynamic response

of plant to loss of

feedwater flow

RG 1.68,

App A,

Section 5

(Introduction)

Plant performance in accordance with design 15.2.7 Loss of Normal

Feedwater Flow

Relief valve testing RG 1.68,

App A

4.p and 5.t

Inspection

Procedure (IP)

72510

Reactor power level at

predetermined power level

plateaus

All relief valves set in auto

Individual valve functional tests

at prescribed power level

plateaus

Individual valve capacity tests at

low power (25% of RTP) using

bypass valve movement or

turbine generator output as a

measurement variable

Relief valve rating at a specified pressure

setting

Delay time between the signal initiating relief

valve opening and the start of motion

Opening stroke time of the main valve disc and

distance

Closing stroke time of the main valve piston

following release of the pneumatically operated

mechanical push rod

15.1.2 Inadvertent

Opening of an SG

Relief or Safety

Valve 

15.6.1 Inadvertent

Opening of a PWR

Pressurizer

Pressure Relief

Valve or a BWR

Pressure Relief

Valve 

Dynamic response

of plant to design

load swings

RG 1.68,

App A

5.h.h

100% of RTP Performance in accordance with design

Dynamic response

of plant to limiting

reactor coolant

pump trips or

closure of RCS flow

control valves

(reactor coolant

recirculation pump

trip test)

RG 1.68,

App A

5.i.i

IP 72512

100% of RTP 

Trip from steady-state power

operation

Recording of transients

following trip and during pump

restart

Recording of limiting heat

transfer parameters

Return to two-pump operation

in accord with facility operating

procedures

Trip of a single pump and of

both pumps simultaneously

Performance in accordance with design:

Instrumentation is adjusted to provide an

accurate conversion of individual jet pump )p

values to a summed core flow over the range of

two-pump operations

Recirculation pump instrumentation is

calibrated

Loop flow from single-tap and double-tap

pumps agrees within 3%

Core flow from single-tap and double-tap

pumps agrees within 2%

Individual jet pump flow variation from average

pump flow is limited

15.3.1 (BWR) & 15.3.2 (PWR)

Loss of Forced

Reactor Coolant

Flow Including Trip

of Pump Motor

Dynamic response

of the plant to loss

of feedwater

heaters that results

in most severe

feedwater

temperature

reduction

RG 1.68,

App A

5.k.k

90% of RTP Performance in accordance with design 15.1.1 Decrease in

Feedwater

Temperature
Transient Test Reference Typical Reactor Plant Initial

Conditions

Typical Transient Test Acceptance Criteria Associated 

Standard Review Plan

Accident Analyses Section 

(as applicable)

14.2.1 - 17 August 2006

Dynamic response

of plant to turbine

trip

(turbine trip or

generator trip)

RG 1.68,

App A

5.l.l

IP 72580

IP 72514

Trip from steady-state operation

at greater than 95% of RTP

Initiation of the test by trip of the

main generator output breaker

Recirculation system flow

control mode must be specified

Performance in accordance with design,

including:

Reactor coolant pumps do not trip

Pressurizer spray valve opens and closes at

the specified values

Reactor pressure remains below the setpoint of

the first safety valves; pressurizer safety valves

do not lift or weep

Pressurizer level within prescribed limits

Steam system power actuated pressure relief

valve opens and closes at specified values

Reactor coolant pressure/temperature

relationship remains within defined values

SG level remains within prescribed limits, no

flooding of the steamlines during the transient,

no initiation of emergency core cooling system

and MSIV isolation during the transient

Turbine bypass system operates to maintain

specific pressure (plants with 100% bypass

capability shall remain at power without scram

during the transient)

Plants with select rod insertion shall maintain

power without scram from recirculation pump

overspeed or cold feedwater effect

Reactor protection system (RPS) functions

should be verified

All safety and emergency core cooling systems

such as the RPS, high pressure coolant

injection (HPCI), diesel generators (DGs), and

reactor core isolation cooling (RCIC) function

without manual assistance if called upon

Normal reactor cooling systems should

maintain adequate cooling and prevent

actuation of automatic depressurization

system, even though relief valves may function

to control pressure

Plant electrical loads (transferred as designed)

Turbine overspeed criteria met

15.2.1 Turbine Trip

Dynamic response

of plant to

automatic closure of

all MSIVs

RG 1.68,

App A

5.m.m

 IP 72510

Initial power level of 100% of

RTP

Performance in accordance with design 

Acceptance criteria include MSIV closing time

15.2.4 MSIV Closure

(BWR)
Transient Test Reference Typical Reactor Plant Initial

Conditions

Typical Transient Test Acceptance Criteria Associated 

Standard Review Plan

Accident Analyses Section 

(as applicable)

August 2006 14.2.1 - 18

Dynamic response

of plant for full load

rejection

(loss of Offsite

Power Testing)

RG 1.68,

App A

5.n.n

 IP 72517

IP 72582

100% of RTP with electrical

system aligned for normal fullpower operation and load

rejection method should subject

turbine to maximum credible

overspeed condition

Steady-state plant operations

with greater than 10%

generator output (IP 72517 &

72582).

Trip of the plant with breakers in

specified positions so that plant

loads will be transferred directly

to the DGs following loss of

house power

Recirculation system flow

control mode specified

Performance in accordance with design,

including:

Automatic transfer of plant loads as designed,

automatic start of DGs, automatic load of DGs

in the specified sequence

Reactor pressure remains below the first safety

valve setting. Pressurizer safety valves do not

lift

All safety systems such as RPS, HPCI, DGs,

and RCIC function without manual assistance

Normal reactor cooling systems should

maintain adequate core temperatures, and

prevent actuation of the Automatic

Depressurization System; however selected

relief valves may function to control pressure

Turbine bypass system operates to maintain

specified pressure value

Steam system power-actuated pressure relief

valves open and close at specified value

Pressurizer spray valves open and close at

specified values.

Reactor coolant temperature/pressure

relationship remains within prescribed values

Pressurizer level is maintained within

prescribed limits

SG level remains within prescribed limits

15.2.6 Loss of

Nonemergency AC

Power to the

Station Auxiliaries 
BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

NRC FORM 335

(9-2004)

NRCMD 3.7

U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

 (Assigned by NRC, Add Vol., Supp., Rev.,

 and Addendum Numbers, if any.)

NUREG-0800

3. DATE REPORT PUBLISHED

MONTH

August

YEAR

2006

4. FIN OR GRANT NUMBER

2. TITLE AND SUBTITLE

NUREG-0800, Section 14.2.1, "Generic Guidelines for Extended Power Uprate Testing

Programs," of the Standard Review Plan for the Review of Safety Analysis Reports for Nuclear

Power Plants, LWR Edition.

5. AUTHOR(S)

Quality and Vendor Branch

6. TYPE OF REPORT

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

Division of Component Integrity

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

Division of New Reactor Licensing

provide name and mailing address.)

and mailing address.)

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Initial issuance of Standard Review Plan section on extended power uprate test program. This superceeds draft Rev. 0, dated

December 2002.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

NRC, Standard Review Plan, SRP, NUREG-0800, Extended Power Uprate Test Program 14. SECURITY CLASSIFICATION

13. AVAILABILITY STATEMENT

unlimited

(This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Office of Nuclear Reactor Regulation

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Washington, DC 20555-0001