UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, DC 20555-0001

August 23, 2007

NRC REGULATORY ISSUE SUMMARY 2007-20

IMPLEMENTATION OF PRIMARY-TO-SECONDARY LEAKAGE 

PERFORMANCE CRITERIA

ADDRESSEES

All holders of operating licenses for pressurized-water reactors (PWRs), except those who have

permanently ceased operations and have certified that fuel has been permanently removed

from the reactor vessel.

INTENT

The U.S. Nuclear Regulatory Commission (NRC) is issuing this Regulatory Issue Summary

(RIS) to notify licensees of its regulatory position regarding implementation of the

primary-to-secondary leakage performance criteria described in Technical Specification Task

Force (TSTF), “Steam Generator Tube Integrity.” NRC expects addressees to review this RIS

for applicability to their facilities and consider actions as appropriate. This RIS requires no

action or written response from addressees.

BACKGROUND INFORMATION

The NRC staff, with external stakeholder involvement, developed modified generic Technical

Specifications (TS) for addressing steam generator (SG) tube integrity. The generic changes to

the Standard Technical Specifications were submitted by the TSTF and are designated

TSTF-449. On May 6, 2005 (70 FR 24126), the NRC approved Revision 4 to TSTF-449 and

applied the consolidated line item improvement process to provide an example application for

adoption of TSTF-449 into a licensee’s TSs. The NRC staff’s model safety evaluation on

TSTF-449 was published in the Federal Register on March 2, 2005 (70 FR 10298).

The revised generic TSs are performance-based because they focus on ensuring the

tubes satisfy performance criteria that are commensurate with assurance of adequate tube

integrity. There are three SG performance criteria: structural integrity, accident-induced

leakage, and operational leakage. SG tube integrity is maintained when all three of these

criteria are met.

All PWR plants have submitted applications to adopt SG TSs similar to those in TSTF-449. In

reviewing these applications, there have been numerous interactions between the NRC staff

and licensees concerning the accident-induced leakage performance criteria including its

relationship to the operational leakage performance criterion. During these interactions, it

ML070570297
RIS 2007-20

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appeared that uncertainties existed regarding how to implement the operational and

accident-induced primary-to-secondary leakage criteria. The purpose of this RIS is to clarify

the NRC staff’s position related to implementation of the leakage performance criteria in

TSTF-449 and committed to by licensees.

The operational leakage performance criterion in TSTF-449 is 150 gallons per day (gpd)

through any one SG. This limit is based on operating experience with SG tube degradation

mechanisms that result in primary-to-secondary leakage. This criterion, in conjunction with the

implementation of the SG program described in TSTF-449, which ensures SG tube integrity is

maintained between inspections, is an effective measure for limiting the frequency of SG tube

ruptures.

The accident-induced primary-to-secondary leakage performance criteria consists of two

criteria. The first criterion involves limiting the primary-to-secondary leakage rate for all design

basis accidents, other than a SG tube rupture, to a value less than that assumed in the accident

analyses in terms of total leakage rate for all SGs and leakage rate for an individual SG. Since

this criterion was established as part of the plant licensing basis, applicants for PWR licenses

are required to analyze the consequences of postulated design basis accidents. Typical

accidents analyzed include a locked rotor, control rod ejection, and a main steamline break. 

These analyses consider the potential for primary-to-secondary leakage to exist and the

analyses must show that the offsite radiological doses associated with this leakage does not

exceed the regulatory requirements which include Title 10 of the Code of Federal Regulations

(10 CFR) Part 100, “Reactor Site Criteria,” limits (or some fraction thereof), 10 CFR 50.67,

“Accident Source Term,” and General Design Criterion (GDC) 19, “Control Room,” of 10 CFR

Part 50, Appendix A, “General Design Criteria for Nuclear Power Plants.” The second criterion

of the accident-induced leakage performance criteria involves limiting the primary-to-secondary

leak rate to 1 gallon per minute (gpm) per SG, unless the NRC staff has approved higher limits

for specific types of degradation at specific locations. This second criterion is intended to

ensure that the potential for induced leakage during severe accidents will be maintained at a

level that will not increase risk.

SUMMARY OF ISSUE

The purpose of this RIS is to clarify the NRC staff’s position related to the implementation of the

leakage performance criteria in TSTF-449 and committed to by licensees.

Issue 1: Potential primary-to-secondary leakage for all design basis accidents should not

exceed the value assumed in the accident analyses.

Primary-to-secondary leakage is assumed to occur in several design basis accidents

(e.g., locked rotor or steam line break). The radiological dose consequences associated with

this assumed leakage are evaluated to ensure they remain within regulatory limits (e.g., 10 CFR

Part 100, GDC 19). The accident-induced leakage performance criteria is intended to ensure

the primary-to-secondary leak rate during an accident remains less than or equal to the

primary-to-secondary leak rate assumed in the accident analyses.

Implementing the accident-induced leakage performance criteria requires an analysis of the

condition of the SG tubing during a SG inspection to calculate the magnitude of the
RIS 2007-20

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primary-to-secondary leak rate which could potentially occur for each of the design basis

accidents. These calculated leak rates for each design basis accident should not exceed the

value assumed in the corresponding accident analyses. This applies to all such accidents

except for a SG tube rupture (which is specifically exempted in the performance criteria). For

some accidents, this may involve calculating the total primary-to-secondary leak rate associated

with all of the flaws in a SG for a specific accident and comparing this leak rate to the leak rate

assumed in the accident analyses (i.e., the accident-induced leak rate limit for this accident); for

other accidents it may simply involve comparing the leakage rate observed during normal

operation to the leakage rate assumed in the accident analyses. This latter case would only be

applicable in the event that the loading conditions (e.g., differential pressure or temperature) on

the tubes during the accident are no more severe (from a leakage integrity standpoint) than the

loading conditions on the tubes during normal operation. That is, the normal operating leakage

would bound the leakage expected during the accident.

Issue 2: Accident-induced leakage includes leakage existing prior to the accident.

Accident-induced leakage includes the leakage existing prior to the accident in addition to the

incremental increase in primary-to-secondary leakage induced during the accident. The

accident-induced leak rate is the primary-to-secondary leak rate that would be observed during

an accident. The accident-induced leak rate is not limited to just the increase (or decrease) in

the leak rate as a result of the different loading conditions imposed on the SG tubes during the

postulated accident.

Issue 3: The temperature at which the volumetric primary-to-secondary flow rate

(i.e., leak rate) is evaluated should be consistent with the temperature assumed in the

accident analyses.

The primary-to-secondary leak rate measured during normal operation and assumed in the

accident analyses are in terms of volumetric flow rates (i.e., gpd or gpm). The volume that a

specific amount of water (primary-to-secondary leakage) occupies depends on its temperature. 

The normal operating primary-to-secondary leakage rate is normally calculated at room

temperature conditions in accordance with industry guidelines. However, the

primary-to-secondary leakage rate assumed in the accident analyses may have been assumed

to occur for a different temperature. In addition, at some plants, the primary-to-secondary

leakage rate for a specific accident may have been evaluated at one temperature for assessing

the offsite radiological dose consequences and at another temperature for assessing the

control room radiological dose consequences. In comparing the accident-induced

primary-to-secondary leak rates to their limits, the volumetric flow rate should be determined for

the specific temperature(s) used in the accident analyses. In addition, if the accident analyses

were performed assuming that accident-induced leakage is equal to the normal operating

leakage rate limit for a specific temperature, the measured normal operating

primary-to-secondary leak rate should be adjusted to that temperature for comparison with the

normal operating primary-to-secondary leak rate limit.

This issue is also addressed in NRC Information Notice 97-79, “Potential Inconsistency in the

Assessment of the Radiological Consequences of a Main Steam Line Break Associated with

the Implementation of Steam Generator Tube Voltage-Based Repair Criteria.”
RIS 2007-20

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Issue 4: The assumptions regarding the pre- and post-accident leakage rate must be

satisfied.

The analyses for a design basis accident may assume a pre-accident primary-to-secondary

leak rate. The pre-accident primary-to-secondary leak rate is assumed to establish the

secondary coolant activity. The accident analyses may assume a post-accident leak rate

different from that assumed prior to the accident. To satisfy the accident-induced leakage

performance criteria, the pre- and post-accident primary-to-secondary leak rate must be less

than the corresponding value(s) assumed in the accident analyses for the appropriate

temperature conditions.

Issue 5: The normal operating primary-to-secondary leak rate may need to be kept well

below the normal operating primary-to-secondary leak rate limit to ensure the

accident-induced leakage performance criteria are not exceeded.

The loading conditions on the tubes during an accident may be different than the loading

conditions on the tubes during normal operation. As a result, the primary-to-secondary leak

rate observed during normal operation may change under accident conditions. In some cases,

the primary-to-secondary leak rate may increase as a result of the accident, while in other

cases it may decrease. If the loading conditions during an accident result in an increase in the

primary-to-secondary leak rate (when compared to the normal operating leak rate), it may be

necessary to restrict the normal operating leak rate to less than the normal operating leakage

rate limit. This applies not only to units that assume the primary-to-secondary leak rate

observed during the accident is the same as the normal operating primary-to-secondary leak

rate limit, an assumption that is permitted by the NRC’s Standard Review Plan; but also to other

units since the increase in primary-to-secondary leak rate going from normal operating

conditions to accident conditions can result in significant increases in the leak rate (depending

on the accident). The actual amount that the leak rate may increase is a function of several

factors including the type of flaw that is leaking. For example, the leak rate from a crack may

increase significantly (e.g., by an order of magnitude depending on through-wall crack length)

under accident conditions.

Issue 6: The term “most limiting accident” should be clearly defined when it is used.

The term “most limiting accident” can have various meanings. In some cases, this term is used

to describe the accident that results in the largest primary-to-secondary leak rate (given a

specific set of flaws or conditions (e.g., leak limiting sleeves) in the SG). In other cases, the

term is used to describe the accident that results in the most significant radiological dose

consequences relative to the applicable regulatory limits (e.g., largest dose or closest to the

radiological dose limit). Because the term “most limiting accident” does not have a single

meaning, the staff and licensees should clearly define what is meant by “most limiting accident”

whenever the term is used.
RIS 2007-20

Page 5 of 6

Issue 7: In the event that a primary-to-secondary leak rate is not assumed for each SG,

appropriate controls should be in place to ensure the assumed accident-induced

primary-to-secondary leak rate for all SGs is not exceeded.

In some cases, an accident analyses will be performed using a total primary-to-secondary leak

rate without discerning the leak rate associated with each of the SGs. In this case, there should

be appropriate controls in place to ensure that the total primary-to-secondary leak rate from all

SGs is kept below the primary-to-secondary leak rate assumed in the accident analyses. 

These controls should account for any primary-to-secondary leak rate that may be observed

during normal operation and the possible changes in the normal operating leak rate as a result

of accident loading conditions.

For example, at one plant a total primary-to-secondary leak rate of 1 gpm is assumed in the

radiological dose analyses for a steam line break accident. This plant has two SGs and there

are no assumptions regarding the amount of leakage from each of the SGs. If the projected

accident-induced leak rate from the faulted SG was calculated to be 1 gpm (based on the flaws

in the SG), this would be acceptable provided the normal operating primary-to-secondary leak

rate in the other SG was zero. This example illustrates that if the projected accident-induced

primary-to-secondary leak rate for the faulted SG is allowed to approach the leak rate assumed

in the accident analyses, the operational leak rate for the other SG may need to be kept below

the normal operating primary-to-secondary leak rate limit. This would not be a concern if the

accident analyses (or plant procedures) were to assume the non-faulted SG was leaking at the

normal operating leak rate limit (or if the sum of the projected accident-induced leak rate for the

faulted SG and the normal operating leak rate limit for the non-faulted SG was kept below the

1 gpm accident-induced leak rate limit).

Issue 8: Exclusions to the risk-informed 1 gpm limit on accident-induced leakage are

evaluated on a case-by-case basis.

In adopting TSTF-449, the NRC staff recognized that for certain severe accident sequences

involving high primary side pressure and a depressurized secondary system (“high-dry”

condition), primary-to-secondary leakage may lead to more heating of the leaking tube than

would be the case if the tube were not leaking, thus increasing the potential for failure of that

tube and a consequent large early release. The 1 gpm accident-induced leakage limit on the

total leak rate from each SG during design basis accidents (other than a SG tube rupture)

ensures that the potential for induced leakage during severe accidents will be maintained at a

level that will not increase risk.

Under certain circumstances, the staff has approved exceptions to the 1 gpm accident-induced

leakage limit. These exceptions have been limited to the leakage associated with a specific

degradation mechanism at a specific location; under all circumstances, the calculated total leak

rate (from all sources) must be kept below the primary-to-secondary leak rate assumed in the

accident analyses. The staff evaluates the exceptions to the 1 gpm limit on a case-by-case

basis by considering the risk implications associated with the increased limit. Approval of the

exceptions to the 1 gpm limit are not limited to degradation mechanisms to which an alternate

repair criteria (NRC approved tube repair criteria that may be implemented as an alternative to
RIS 2007-20

Page 6 of 6

the standard depth based repair criteria of 40-percent through-wall), nor does approval of an

alternate tube repair criteria automatically justify an exception to the 1 gpm limit.

BACKFIT DISCUSSION

This RIS requires no action or written response and is, therefore, not a backfit under

10 CFR 50.109. Consequently, the NRC staff did not perform a backfit analysis.

FEDERAL REGISTER NOTIFICATION

A notice of opportunity for public comment on this RIS was not published in the Federal

Register because this RIS is informational and pertains to a staff position that does not

represent a departure from current regulatory requirements and practice.

CONGRESSIONAL REVIEW ACT

The NRC has determined that this action is not subject to the Congressional Review Act.

PAPERWORK REDUCTION ACT STATEMENT

The information collections contained in this RIS are covered by the requirements of 10 CFR

Parts 50 and 100, which were approved by the Office of Management and Budget (OMB),

approval numbers 3150-0011 and 3150-0093.

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.

CONTACT

Please direct any questions about this matter to either of the technical contacts listed below.

/RA by TQuay for/

Michael Case, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

Technical Contacts: Ken Karwoski, NRR Emmett Murphy, NRR

301-415-2752 301-415-2710

 E-mail: kjk1@nrc.gov elm@nrc.gov
RIS 2007-20

Page 6 of 6

the standard depth based repair criteria of 40-percent through-wall), nor does approval of an

alternate tube repair criteria automatically justify an exception to the 1 gpm limit.

BACKFIT DISCUSSION

This RIS requires no action or written response and is, therefore, not a backfit under

10 CFR 50.109. Consequently, the NRC staff did not perform a backfit analysis.

FEDERAL REGISTER NOTIFICATION

A notice of opportunity for public comment on this RIS was not published in the Federal

Register because this RIS is informational and pertains to a staff position that does not

represent a departure from current regulatory requirements and practice.

CONGRESSIONAL REVIEW ACT

The NRC has determined that this action is not subject to the Congressional Review Act.

PAPERWORK REDUCTION ACT STATEMENT

The information collections contained in this RIS are covered by the requirements of 10 CFR

Parts 50 and 100, which were approved by the Office of Management and Budget (OMB),

approval numbers 3150-0011 and 3150-0093.

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.

CONTACT

Please direct any questions about this matter to either of the technical contacts listed below.

/RA by TQuay for/

Michael Case, Director

Division of Policy and Rulemaking

Office of Nuclear Reactor Regulation

Technical Contacts: Ken Karwoski, NRR Emmett Murphy, NRR

301-415-2752 301-415-2710

 E-mail: kjk1@nrc.gov elm@nrc.gov

DISTRIBUTION: RIS File

ML070570297

OFFICE DCI/CSGB Tech. Editor DCI/CSGB:BC NRR/DCI NRR/DCI:DD DRA/AAD:BC NRR/DRA:DD

NAME EMurphy HChang AHiser KKarwoski MEvans MKotzalas CHolden

DATE 3/12/2007 3/2/2007 4/2/2007 4/3/2007 4/17/2007 4/18/2007 4/23/2007

OFFICE DIRS/ITSB NRR/DORL:DD OE OGC/NLO OGC/CRA NRR/PMAS OIS

NAME TKobetz CHaney DSolorio MBaty DReddick LHill CColburn

DATE 05/03/2007 5/24/2007 6/16/2007 7/05/2007 6/28/2007 6/22/2007 7/24/2007

OFFICE DPR/PGCB:LA DPR/PGCB DPR/PGCB:BC NRR/DPR:DD

NAME CHawes AMarkley MMurphy TQuay for MCase

DATE 08/02/07 08/03/07 08/10/07 08/23/07

 OFFICIAL RECORD COPY