[Federal Register Volume 67, Number 149 (Friday, August 2, 2002)]
[Notices]
[Pages 50475-50485]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-19538]


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NUCLEAR REGULATORY COMMISSION


Notice of Opportunity To Comment on Model Safety Evaluation on 
Technical Specification Improvement To Modify Requirements Regarding 
Mode Change Limitations Using the Consolidated Line Item Improvement 
Process

AGENCY: Nuclear Regulatory Commission.

ACTION: Request for comment.

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SUMMARY: Notice is hereby given that the staff of the Nuclear 
Regulatory Commission (NRC) has prepared a model safety evaluation (SE) 
relating to the modification of requirements regarding technical 
specifications (TS) mode change limitations. The NRC staff has also 
prepared a model no significant hazards consideration (NSHC) 
determination relating to this matter. The purpose of these models is 
to permit the NRC to efficiently process amendments that propose to 
modify requirements that limit changing operational modes. Licensees of 
nuclear power reactors to which the models apply could then request 
amendments, confirming the applicability of the SE and NSHC 
determination to their reactors. The NRC staff is requesting comment on 
the model SE and model NSHC determination prior to announcing their 
availability for referencing in license amendment applications.

DATES: The comment period expires September 3, 2002. Comments received 
after this date will be considered if it is practical to do so, but the 
Commission is able to ensure consideration only for comments received 
on or before this date.

ADDRESSES: Comments may be submitted either electronically or via U.S. 
mail.
    Submit written comments to Chief, Rules and Directives Branch, 
Division of Administrative Services, Office of Administration, Mail 
Stop: T-6 D59, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001. Hand deliver comments to: 11545 Rockville Pike, Rockville, 
Maryland, between 7:45 a.m. and 4:15 p.m. on Federal workdays. Copies 
of comments received may be examined at the NRC's Public Document Room, 
11555 Rockville Pike (Room O-1F21), Rockville, Maryland. Comments may 
be submitted by electronic mail to [email protected].

FOR FURTHER INFORMATION CONTACT: Robert Dennig, Mail Stop: O-12H4, 
Division of Regulatory Improvement Programs, Office of Nuclear Reactor 
Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001, telephone 301-415-1156.

SUPPLEMENTARY INFORMATION:

Background

    Regulatory Issue Summary 2000-06, ``Consolidated Line Item 
Improvement Process for Adopting Standard Technical Specification 
Changes for Power Reactors,'' was issued on March 20, 2000. The 
consolidated line item improvement process (CLIIP) is intended to 
improve the efficiency of NRC licensing processes, by processing 
proposed changes to the standard technical specifications (STS) in a 
manner that supports subsequent license amendment applications. The 
CLIIP includes an opportunity for the public to comment on proposed 
changes to the STS after a preliminary assessment by the NRC staff and 
finding that the change will likely be offered for adoption by 
licensees. This notice solicits comment on a proposed change to the STS 
that modifies requirements for mode change limitations. The CLIIP 
directs the NRC staff to evaluate any comments received for a proposed 
change to the STS and to either reconsider the change or announce the 
availability of the change for adoption by licensees. Licensees opting 
to apply for this TS change are responsible for reviewing the staff's 
evaluation, referencing the applicable technical justifications, and 
providing any necessary plant-specific information. Each amendment 
application made in response to the notice of availability will be 
processed and noticed in accordance with applicable rules and NRC 
procedures.
    This notice involves the modification of TS requirements regarding 
mode change limitations. This change was proposed for incorporation 
into the standard technical specifications by the Owners Groups 
participants in the Technical Specification Task Force (TSTF) and is 
designated TSTF-359. TSTF-359 can be viewed on the NRC's Web page at 
http://www.nrc.gov/reactors/operating/licensing/techspecs.html.

[[Page 50476]]

Applicability

    This proposal to modify technical specification requirements for 
mode change limitations is applicable to all licensees who have adopted 
or will adopt, in conjunction with the proposed change, technical 
specification requirements for a Bases control program consistent with 
the TS Bases Control Program described in Section 5.5 of the applicable 
vendor's STS.
    To efficiently process the incoming license amendment applications, 
the staff requests that each licensee applying for the changes proposed 
in TSTF-359 include Bases for the proposed TS consistent with the Bases 
proposed in TSTF-359. In addition, licensees that have not adopted 
requirements for a Bases control program by converting to the improved 
STS or by other means, are requested to include the requirements for a 
Bases control program consistent with the STS in their application for 
the proposed change. The need for a Bases control program stems from 
the need for adequate regulatory control of some key elements of the 
proposal that are contained in the proposed Bases for LCO 3.0.4 and SR 
3.0.4. The staff is requesting that the Bases be included with the 
proposed license amendments in this case because the changes to the TS 
and the changes to the associated Bases form an integral change to a 
plant's licensing bases. To ensure that the overall change, including 
the Bases, includes appropriate regulatory controls, the staff plans to 
condition the issuance of each license amendment on the licensee's 
incorporation of the changes into the Bases document and on requiring 
the licensee to control the changes in accordance with the Bases 
Control Program. The CLIIP does not prevent licensees from requesting 
an alternative approach or proposing the changes without the requested 
Bases and Bases control program. However, deviations from the approach 
recommended in this notice may require additional review by the NRC 
staff and may increase the time and resources needed for the review.

Public Notices

    This notice requests comments from interested members of the public 
within 30 days of the date of publication in the Federal Register. 
After evaluating the comments received as a result of this notice, the 
staff will either reconsider the proposed change or announce the 
availability of the change in a subsequent notice (perhaps with some 
changes to the safety evaluation or the proposed no significant hazards 
consideration determination as a result of public comments). If the 
staff announces the availability of the change, licensees wishing to 
adopt the change must submit an application in accordance with 
applicable rules and other regulatory requirements. For each 
application the staff will publish a notice of consideration of 
issuance of amendment to facility operating licenses, a proposed no 
significant hazards consideration determination, and a notice of 
opportunity for a hearing. The staff will also publish a notice of 
issuance of an amendment to operating license to announce the 
modification of requirements for mode change limitations for each plant 
that receives the requested change.

Proposed Safety Evaluation

U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor 
Regulation, Consolidated Line Item Improvement, Technical Specification 
Task Force (TSTF) Change TSTF-359, Changes to Limiting Condition for 
Operation 3.0.4 and Surveillance Requirement 3.0.4 Regarding Mode 
Change Limitations

1.0  Introduction

    On March 9, 2001, the Nuclear Energy Institute (NEI) Risk Informed 
Technical Specifications Task Force (RITSTF) submitted a proposed 
change, TSTF-359, Revision 5, to the standard technical specifications 
(STS) (NUREGs 1430-1434) on behalf of the industry (TSTF-359 Revisions 
1 through 4 were internal NEI iterations). TSTF-359, Revision 5, is a 
proposal to change the STS Limiting Condition for Operation (LCO) 3.0.4 
and Surveillance Requirement (SR) 3.0.4 requirements regarding mode 
change limitations. The proposed change would modify LCO 3.0.4 and SR 
3.0.4 by risk informing limitations on entering the mode of 
applicability of a LCO.
    At the July 31, 2001, NRC/RITSTF meeting, the staff provided verbal 
comments, questions and requests for additional information (RAIs) 
pertaining to TSTF-359, Revision 5. In response to the staff RAIs and 
questions, the RITSTF submitted TSTF-359, Revision 6, on February 22, 
2002. In a letter of April 26, 2002, the staff suggested specific 
changes that were needed, and after further discussions, the RITSTF 
submitted the final TSTF-359, Revision 7, on July 17, 2002. This 
proposal is one of the industry's initiatives under the risk-informed 
technical specifications program. These initiatives are intended to 
maintain or improve safety while reducing unnecessary burden and to 
make technical specification requirements consistent with the 
Commission's other risk-informed regulatory requirements, in particular 
the maintenance rule.
    The current technical specifications (TS) specify that a nuclear 
power plant cannot go to higher modes of operation \1\ (i.e., move 
towards power operation) unless all TS systems, normally required for 
the higher mode, are operable. This limitation is included (with 
several exceptions for some plants) in LCO 3.0.4 and SR 3.0.4. LCO 
3.0.4 and SR 3.0.4 in the STS currently state in part that when an LCO 
or SR is not met, ``entry into a MODE or other specified condition in 
the applicability shall not be made except when the associated actions 
to be entered permit continued operation in the MODE or other specified 
condition in the applicability for an unlimited period of time.'' The 
industry believes that this requirement is unnecessarily restrictive 
and can unduly delay plant startup while considerable resources are 
being used to resolve startup issues that are risk insignificant or low 
risk. A maintenance activity that takes longer than planned can delay a 
mode change and adversely impact a utility's orderly plant startup and 
return to power operation. The objective of the proposed change is to 
provide additional operational flexibility without compromising plant 
safety.
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    \1\ MODE numbers decrease in the transition ``up to a higher 
mode of operation''; power operation is MODE 1.
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    The proposed changes to LCO 3.0.4 and SR 3.0.4 would allow, for 
systems and components, mode changes into a TS condition that has a 
specific required action and completion time. The licensee will utilize 
the LCO 3.0.4 or SR 3.0.4 allowance only when they determine that there 
is a high likelihood that the LCO will be satisfied within the LCO 
completion time (CT), after the mode change. In addition, the LCO 3.0.4 
and SR 3.0.4 allowances can be applied to values and parameters in 
specifications when explicitly stated in the TS (non-system/component 
TS such as: Reactor Coolant System Specific Activity). These changes 
are in addition to the current mode change allowance when a required 
action has an indefinite completion time. The LCO 3.0.4 and SR 3.0.4 
mode change allowances are not permitted for the systems and components 
(termed ``higher risk'') listed in Section 3.1.1, ``Identification of 
Risk Important TS Systems and Components,'' for the modes specified. 
Two examples are: (1) Westinghouse plants cannot transition from Mode 5 
to Mode 4 without a High Head Safety Injection System train operable; 
and, (2)

[[Page 50477]]

Westinghouse plants cannot transition up into any mode with an 
inoperable required emergency diesel generator.

2.0  Regulatory Evaluation

    In 10 CFR 50.36, the Commission established its regulatory 
requirements related to the content of TS. Pursuant to 10 CFR 50.36, TS 
are required to include items in the following five specific categories 
related to station operation: (1) Safety limits, limiting safety system 
settings, and limiting control settings; (2) limiting conditions for 
operation (LCOs); (3) surveillance requirements (SRs); (4) design 
features; and (5) administrative controls. The rule does not specify 
the particular requirements to be included in a plant's TS. As stated 
in 10 CFR 50.36(c)(2)(i), the ``Limiting conditions for operation are 
the lowest functional capability or performance levels of equipment 
required for safe operation of the facility. When a limiting condition 
for operation of a nuclear reactor is not met, the licensee shall shut 
down the reactor or follow any remedial action permitted by the 
technical specification * * *'' By convention, the LCOs are contained 
in Sections 3.1 through 3.10 of the TS. TS Section 3.0, on ``LCO and SR 
Applicability,'' provide details or ground rules for complying with the 
LCOs. LCO 3.0.4 and SR 3.0.4 address requirements for LCO compliance 
when transitioning between modes of operation.
    Technical specifications have taken advantage of risk technology as 
experience and capability have increased. Since the mid-1980's, the NRC 
has been reviewing and granting improvements to technical 
specifications that are based, at least in part, on probabilistic risk 
assessment (PRA) insights. In its final policy statement on technical 
specification improvements of July 22, 1993, the Commission stated that 
it expects that licensees will utilize any plant specific PRA or risk 
survey in preparing their technical specification related submittals. 
In evaluating these submittals, the staff applies the guidance in RG 
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis,'' 
dated July 1998 and in RG 1.177, ``An Approach for Plant-Specific, 
Risk-Informed Decisionmaking: Technical Specifications,'' dated August 
1998. The staff has appropriately adapted this guidance to assess the 
acceptability of upward mode changes with equipment inoperable. This 
review had the following objectives:
     To ensure that the plant risk does not increase 
unacceptably during the actual implementation of the proposed change 
(e.g., when the plant enters a higher mode while an LCO is not met). 
This risk increase is referred to as ``temporary.''
     To compare and assess the risk impact of the proposed 
change to the acceptance guidelines of the Commission's Safety Goal 
Policy Statement, as documented in RG 1.174. The risk impact, which is 
measured by the average yearly risk increase associated with the 
change, aims at minimizing the ``cumulative'' risk associated with the 
proposed change so that the plant's average baseline risk is maintained 
within a minimal range.
     To assess the licensee's ability to identify risk 
significant configurations resulting from maintenance or other 
operational activities and take appropriate compensatory measures to 
avoid such configurations.
    The staff reviewed the reliance on 10 CFR 50.65(a)(4) for the non-
higher risk systems and components, and related guidance to assess and 
manage the risk of upward mode changes. The Commission has found that 
compliance with 10 CFR 50.65(a)(4) satisfies the configuration risk 
management objectives of RG 1.177 for technical specification 
surveillance interval and completion time extensions. Reliance on 10 
CFR 50.65(a)(4) processes was also found adequate for managing risk of 
missed surveillances as described in the Federal Register on September 
28, 2001 (66 FR 49714).
    The staff review also had the objective of ensuring that existing 
inspection programs have the necessary controls in place to allow NRC 
staff to oversee the implementation of the proposed change, reliance on 
10 CFR 50.65(a)(4), and the ability to adequately assess the licensee's 
performance associated with risk assessments. The review encompassed 
inspection procedures (i.e., NRC Inspection Procedure 62709 (12/28/00), 
``Configuration Risk Assessment and Risk Management Process,'' and NRC 
Inspection Procedure 71111.13 (1/17/02), ``Maintenance Risk Assessments 
and Emergent Work Control''), the significance determination process 
(SDP) (i.e., draft ``Maintenance Risk Assessment and Risk Management 
Significance Determination Process''), enforcement guidance (i.e., 
draft Enforcement Manual Section 8.1.11, ``Actions Involving the 
Maintenance Rule''), and the associated reactor oversight process.

2.1  Proposed Change to LCO 3.0.4 and SR 3.0.4

    Currently LCO 3.0.4 does not allow entrance into a higher mode (or 
other specified condition) in the applicability when an LCO is not met, 
except when the associated actions to be entered permit continued 
operation in that mode or condition indefinitely or a specific 
exception is granted. Similarly, when an LCO's surveillances have not 
been met within their specified frequency, entry into a higher mode (or 
other specified condition) is not allowed by SR 3.0.4. The current STS 
\2\ LCO 3.0.4 reads:

    \2\ Plant specific wording for current equivalent LCO 3.0.4 is 
similar to current STS LCO 3.0.4 wording.
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    ``When an LCO is not met, entry into a MODE or other specified 
condition in the Applicability shall not be made except when the 
associated ACTIONS to be entered permit continued operation in the MODE 
or other specified condition in the Applicability for an unlimited 
period of time. This Specification shall not prevent changes in MODES 
or other specified conditions in the Applicability that are required to 
comply with ACTIONS or that are part of a shutdown of the unit.
    Exceptions to this Specification are stated in the individual 
Specifications. These exceptions allow entry into MODES or other 
specified conditions in the Applicability when the associated ACTIONS 
to be entered allow unit operation in the MODE or other specified 
condition in the Applicability only for a limited period of time.
    LCO 3.0.4 is only applicable for entry into a MODE or other 
specified conditions in the Applicability in [MODES 1, 2, 3, and 4 {for 
PWRs}][MODES 1, 2, and 3 {for BWRs}].''
    The revised LCO 3.0.4 will read:

    ``When an LCO is not met, entry into a MODE or other specified 
condition in the Applicability shall only be made
    (a) when the associated Actions to be entered permit continued 
operation in that MODE or other specified condition in the 
Applicability for an unlimited period of time, or
    (b) after performance of a risk assessment addressing inoperable 
systems and components, consideration of the results, determination 
of the acceptability of entering the MODE or other specified 
condition in the Applicability, and establishment of risk management 
actions, if appropriate; exceptions to this Specification are stated 
in the individual Specifications, or
    (c) when an allowance is stated in the individual value or 
parameter Specification.'' This Specification shall not prevent 
changes in MODES or other specified conditions in

[[Page 50478]]

the Applicability that are required to comply with ACTIONS or that 
are part of a shutdown of the unit.
    LCO 3.0.4 is only applicable for entry into a MODE or other 
specified conditions in the Applicability in [MODES 1, 2, 3, and 4 
{for PWRs}][MODES 1, 2, and 3 {for BWRs}].''

    The current STS \3\ SR 3.0.4 reads:

    \3\ Plant specific wording for current equivalent SR 3.0.4 is 
similar to current STS SR 3.0.4 wording.
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    ``Entry into a MODE or other specified condition in the 
Applicability of an LCO shall not be made unless the LCO's 
Surveillances have been met within their specified frequency. This 
provision shall not prevent entry into MODES or other specified 
conditions in the Applicability that are required to comply with 
ACTIONS or that are part of a shutdown of the unit.
    SR 3.0.4 is only applicable for entry into a MODE or other 
specified conditions in the Applicability in [MODES 1, 2, 3, and 4 
{for PWRs}][MODES 1, 2, and 3 {for BWRs}].''

    The revised SR 3.0.4 will conform to the changes to LCO 3.0.4 and 
read:

    ``Entry into a MODE or other specified condition in the 
Applicability of an LCO shall not be made unless the LCO's 
Surveillances have been met within their specified frequency. When 
an LCO is not met due to a Surveillance not having been met, entry 
into a MODE or other specified condition in the Applicability shall 
only be made
    (a) when the associated Actions to be entered permit continued 
operation in that MODE or other specified condition in the 
Applicability for an unlimited period of time, or
    (b) after performance of a risk assessment addressing inoperable 
systems and components, consideration of the results, determination 
of the acceptability of entering the MODE or other specified 
condition in the Applicability, and establishment of risk management 
actions, if appropriate; exceptions to this Specification are stated 
in the individual Specifications, or
    (c) when an allowance is stated in the individual value or 
parameter Specification.
    This provision shall not prevent entry into MODES or other 
specified conditions in the Applicability that are required to 
comply with ACTIONS or that are part of a shutdown of the unit.
    SR 3.0.4 is only applicable for entry into a MODE or other 
specified conditions in the Applicability in [MODES 1, 2, 3, and 4 
{for PWRs}][MODES 1, 2, and 3 {for BWRs}].''

    The proposed LCO 3.0.4(a) retains the current allowance for when 
the required actions allow indefinite operation. The proposed LCO 
3.0.4(b) and SR 3.0.4(b) allow entering modes or other specified 
conditions in the applicability except when higher risk systems and 
components (listed in section 3.1.1), for the mode being entered, are 
inoperable. The decision for entering a higher mode or condition in the 
applicability of the LCO will be made by plant management after the 
required risk assessment has been performed and requisite risk 
management actions established, through the program established to 
implement 10 CFR 50.65(a)(4). Entry into the modes or other specified 
conditions in the applicability of the TS shall be for no more than the 
duration of the applicable required actions completion time or until 
the LCO is met. Current notes in individual specifications that 
permitted mode changes are now encompassed by LCO 3.0.4(b) and can be 
removed. Notes that prohibit mode changes under LCO 3.0.4(b) must be 
added (i.e., for higher risk systems and components). The proposed LCO 
3.0.4(b) and SR 3.0.4(b) allowances can involve multiple components in 
a single LCO or in multiple LCOs; however, use of the LCO 3.0.4(b) and 
SR 3.0.4(b) provisions are always contingent upon completion of a 10 
CFR 50.65(a)(4) based risk assessment.
    LCO 3.0.4 or SR 3.0.4 allowances related to values and parameters 
of TS are not typically addressed by LCO 3.0.4(b) or SR 3.0.4(b) risk 
assessments, and are therefore addressed by a new LCO 3.0.4(c) and SR 
3.0.4(c). LCO 3.0.4(c) and SR 3.0.4(c) refer to allowances already in 
the TS and annotated in the individual TS. LCO 3.0.4(c) and SR 3.0.4(c) 
also allow for entry into the modes or other specified conditions in 
the applicability of a TS for no more than the duration of the 
applicable required actions completion time or until the LCO is met. 
Examples of LCO 3.0.4 and SR 3.0.4 utilization of required actions and 
completion times are provided in Appendix A for clarification.

3.0  Technical Evaluation

    During the development of the current STS, improvements were made 
to LCO 3.0.4, such as clarifying its applicability with respect to 
plant shutdowns, cold shutdown mode and refueling mode. In addition, 
during the STS development, almost all the LCOs with completion times 
greater than or equal to 30 days, and many LCOs with completion times 
greater than or equal to 7 days, were given individual LCO 3.0.4 
exceptions. During some conversions to the STS, individual plants 
provided acceptable justifications for other LCO 3.0.4 exceptions. All 
of these specific LCO 3.0.4 exceptions allow entry into a mode or other 
specified condition in the TS applicability while relying on the TS 
required actions and associated completion times. The proposed change 
under evaluation would provide standardization and consistency to the 
use and application of LCO 3.0.4 and SR 3.0.4, both internal to and 
between each of the specifications and STS NUREGs. This proposed change 
will also ensure consistency through the utilization of appropriate 
levels of risk assessment of plant configurations for application of 
LCO 3.0.4 and SR 3.0.4. However, nothing in this safety evaluation 
should be interpreted as encouraging upward mode transition with 
inoperable equipment. Good practice should dictate that such 
transitions should normally be initiated only when all required 
equipment is operable and that mode transition with inoperable 
equipment should be the exception rather than the rule.
    The current LCO 3.0.4(a) and SR 3.0.4(a) allowances are retained in 
the proposal and do not represent a change in risk from the current 
situation. The LCO 3.0.4(b) and SR 3.0.4(b) allowances apply to systems 
and components, and require a risk assessment prior to utilization to 
ensure an acceptable level of safety is maintained. The LCO 3.0.4(c) 
and SR 3.0.4(c) allowances apply to parameters and values which have 
been previously approved by the NRC in a plants specific TS. The 
licensee will provide in their TS Bases a discussion and list of each 
NRC approved LCO 3.0.4(c) and SR 3.0.4(c) specific value and parameter 
allowances. The Bases of LCO 3.0.4 and SR 3.0.4 will be revised to 
explain the new allowances and their utilization.
    The staff did a qualitative assessment of the risk impact of the 
proposed change in LCO 3.0.4(b) and SR 3.0.4(b) allowances by 
evaluating how the licensee's implementation of the proposed risk-
informed approach is expected to meet the requirements of the 
applicable RGs. The staff referred to the guidance provided in RG 
1.174, ``An Approach for Using Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific Changes to the Licensing Basis,'' 
and in RG 1.177, ``An approach for Plant-Specific, Risk-Informed 
Decsionmaking: Technical Specifications.'' RG 1.177 provides the 
staff's recommendations on utilizing risk-information to assess the 
impact of proposed changes to nuclear power plant technical 
specifications on the risk associated with plant operation. Although RG 
1.177 does not specifically address the type of generic change in this 
proposal, the staff considered the approach documented in RG 1.177 in 
evaluating the risk information provided in support of the proposed 
change in LCO 3.0.4 and SR 3.0.4.
    The staff's evaluation of how the implementation of the proposed 
risk-informed approach, used to justify LCO 3.0.4(b) and SR 3.0.4(b) 
allowances, agrees with the objectives of the

[[Page 50479]]

guidance outlined in RG 1.177 is discussed in Section 3.1. Oversight of 
the risk-informed approach associated with the LCO 3.0.4(b) and SR 
3.0.4(b) allowances is discussed in Section 3.2.

3.1  Evaluation of Risk Management

    Both the temporary and cumulative risk of the proposed change are 
adequately limited. The temporary risk is limited by the exclusion of 
higher risk systems and components, and completion time limits 
contained in technical specifications (Section 3.1.1). The cumulative 
risk is limited by the temporary risk limitations and by the expected 
low frequency of the proposed mode changes with inoperable equipment 
(Section 3.1.2). NRC oversight of a licensee's implementation of 10 CFR 
50.65(a)(4) as applied to the proposed change provides adequate 
assurance of the licensee's ability to use the LCO 3.0.4(b) or SR 
3.0.4(b) provisions under appropriate circumstances, i.e., to identify 
risk-significant configurations when entering a higher mode or 
condition in the applicability of an LCO (Section 3.1.3).

3.1.1 Temporary Risk Increases

    RG 1.177 proposes the incremental conditional core damage 
probability (ICCDP) and the incremental conditional large early release 
probability (ICLERP) as appropriate measures of the increase in 
probability of core damage and large early release, respectively, 
during the period of implementation of a proposed TS change. In 
addition, RG 1.177 stresses the need to preclude potentially high risk 
configurations introduced by the proposed change. The ICCDP associated 
with any specified plant condition, such as the condition introduced by 
entering a higher mode with plant equipment inoperable, is expressed by 
the following equation:

ICCDP = R d = (R1-Ro) d    (1)

where

R = the conditional risk increase, in terms of core damage 
frequency (CDF), caused by the specified condition
d = the duration of the specified plant condition
R1 = the plant CDF with the specified condition permanently 
present
Ro = the plant CDF without the specified condition

    The same expression can be used for ICLERP by substituting the 
measure of risk, i.e., large early release frequency (LERF) for CDF. 
The magnitude of the ICCDP and ICLERP values associated with plant 
conditions applicable to LCO 3.0.4(b) and SR 3.0.4(b) allowances can be 
managed by controlling the conditional risk increase, R (in 
terms of both CDF and LERF) and the duration, d, of such conditions. 
The following sections discuss how the key elements of the proposed 
risk-informed approach, used to justify LCO 3.0.4(b) and SR 3.0.4(b) 
allowances, are expected to limit R and d and, thus, prevent 
any significant temporary risk increases.
    Identification of Risk Important TS Systems and Components. A major 
element that limits the risk of the proposed mode change flexibility is 
the exclusion of certain systems and associated LCOs for the mode 
change allowance. Technical specifications allow operation in Mode 1 
(power operation) with specified levels of inoperability for specified 
times. This provides a benchmark of currently acceptable risk against 
which to measure any incremental risk inherent in the proposed LCO 
3.0.4(b) and SR 3.0.4(b). If a system inoperability accrues risk at a 
higher rate in one or more of the transition modes than it would in 
Mode 1, then an upward transition into that mode should not be allowed 
without demonstration of a high degree of experience and sophistication 
in risk management. However, the risk management process evaluated in 
Section 3.1.3 is adequate if high risk systems/components are excluded 
from the scope of LCO 3.0.4(b) and SR 3.0.4(b).
    The importance of most TS systems in mitigating accidents increases 
as power increases. However, some TS systems are relatively more 
important during lower power and shutdown operations, because:
     certain events are peculiar to modes of plant operation 
other than power operation,
     certain events are more probable at modes of plant 
operation other than power operation,
     some modes of plant operation have less mitigation system 
capability than power operation.
    The risk information submitted in support of the proposed changes 
to LCO 3.0.4 and SR 3.0.4 includes qualitative risk assessments 
performed by each owners group to identify higher risk systems and 
components at the various modes of operation, including transitions 
between modes, as the plant moves upward from the refueling mode of 
operation toward power operation. The owners groups' generic 
qualitative risk assessments are included as attachments to TSTF-359, 
Revision 7. Each of the owners groups' generic qualitative risk 
assessments discuss the technical approach used and the systems/
components subsequently determined to be of higher risk significance; 
the systems/components not to be granted the LCO 3.0.4 or SR 3.0.4 
allowances for the various modes listed. The owners groups generic 
qualitative risk assessments are:
     BWR Owners' Group Risk-informed Technical Specification 
Committee, ``Technical Justification to Support Risk-informed 
Improvements to Technical Specification Mode Restraints for BWR 
Plants,'' General Electric Company GE-NE A13-00464 (Rev[2]).
     ``B&W Owners Group Qualitative Risk Assessment for 
Increased Flexibility in MODE Restraints,'' Framatome Technologies BAW-
2383.
     Combustion Engineering Owners Group (CEOG) Task 1181, 
``Qualitative Risk Assessment for Relaxation of Mode Entry 
Restraints,'' CE Nuclear Power LLC, CE NPSD-1207 (Rev[0]).
     ``WOG Qualitative Risk Assessment Supporting Increased 
Flexibility in MODE Restraints.''
    Following interactions with the staff, all owners groups used the 
same systematic approach in their qualitative risk assessments to 
identify the higher risk systems in the STS, consisting of the 
following steps:
     identification of plant conditions (i.e., plant parameters 
and availability of key mitigation systems) associated with changes in 
plant operating modes while returning to power.
     identification of key activities that have the potential 
to impact risk and which are in progress during transitions between 
modes while the plant is returning to power.
     identification of applicable accident initiating events 
for each mode or other specified condition in the applicability.
     identification of the higher risk systems and components 
by combining the information in the first three steps (qualitative risk 
assessment).
    The risk assessments properly used the results and insights from 
previous deterministic and probabilistic studies to systematically 
search for plant conditions in which certain key plant components are 
more important in mitigating accidents than at power operation (Mode 
1). This search was systematic, taking the following factors into 
account for the various stages of returning the plant to power:
     the status of accident mitigation and normally operating 
systems.
     the status of key plant parameters such as reactor coolant 
system pressure.
     the key activities that are in progress during transitions 
between modes which have the potential to impact risk (e.g. the 
transfer from

[[Page 50480]]

auxiliary to main feedwater at some PWR plants when Mode 1 is entered).
     the applicable accident initiating events for each mode of 
plant operation.
     design and operational differences among plants or groups 
of plants.
    The following systems and components were identified by each of the 
four owners groups as higher risk systems and components, when the 
plant is entering a new mode.

            Boiling Water Reactor Owners Group (BWROG) Plants
------------------------------------------------------------------------
            System                     BWR type           Entering mode
------------------------------------------------------------------------
High Pressure Coolant           BWR 3 & 4.............  2, 1
 Injection (HPCI) System.
High Pressure Core Spray        BWR 5 & 6.............  2, 1
 (HPCS).
Reactor Core Isolation Cooling  BWR 3, 4, 5 & 6.......  2, 1
 (RCIC) System.
Isolation Condenser Diesel      BWR 2.................  2, 1
 Generators (including other.
Emergency/Shutdown AC Power     All...................  All
 Supplies).
Hardened Wetwell Vent System..  BWR 2, 3 & 4 with Mark  3, 2, 1
                                 I Containment.
Residual Heat Removal System..  All...................  4
------------------------------------------------------------------------


----------------------------------------------------------------------------------------------------------------
                        System                                                Entering Mode
----------------------------------------------------------------------------------------------------------------
     Babcock & Wilcox Owners Group (B&WOG) Plants
 
Emergency Diesel Generators (EDG) & Hydro-Electric      5, 4, 3, 2, 1
 Units for Oconee.
Emergency Feedwater (EFW) System......................  1
Decay Heat Removal (DHR) System.......................  5, 4
 
   Combustion Engineering Owners Group (CEOG) Plants
 
Emergency Diesel Generators (EDGs)....................  5, 4, 3, 2, 1
Auxiliary Feedwater/Emergency Feedwater (AFW/EFW)       4, 3, 2, 1
 System.
High Pressure Safety Injection (HPSI) System..........  4, 3 (below 1700 psia)
LTOP/PORVs (when used for Low Temperature Overpressure  5, 4 (below set temperature)
 Protection (LTOP)).
Shutdown Cooling System (Low Pressure Safety Injection  5
 (LPSI) pumps).
 
        Westinghouse Owners Group (WOG) Plants
 
Emergency Diesel Generators (EDGs)....................  5, 4, 3, 2, 1
Auxiliary Feedwater (AFW) System (for plants depending  4, 3, 2, 1
 on AFW for startup).
High Head Safety Injection System.....................  4
Cold Overpressure Protection System...................  5, 4
Residual Heat Removal (RHR) System....................  5
----------------------------------------------------------------------------------------------------------------

    If a licensee identifies a higher risk system for only some of the 
modes of applicability, the TS for that system would be modified by a 
Note that reads, for example, ``LCO 3.0.4(b) is not applicable when 
entering MODE 1 from MODE 2.'' Systems identified as higher risk for 
modes outside the applicability of LCO 3.0.4 and SR 3.0.4 (Modes 5 and 
6 for PWRs, and Modes 4 and 5 for BWRs), are also to be excluded from 
transitioning up to the mode of higher risk, however, those systems 
will be addressed by administrative controls.
    In summary, the staff's review of the owners groups qualitative 
risk assessments finds that they are of adequate quality to support the 
application (i.e., they identify the higher risk systems and 
components) associated with entering higher modes of plant operation 
with equipment inoperable while returning to power.
    [Plant Specific changes will be described here.]
    Limited Time in TS Required Actions. Any temporary risk increase 
will be limited by, among other factors, duration constraints imposed 
by the TS CTs of the inoperable systems. For the systems and components 
which are not higher risk, any temporary risk increase associated with 
the proposed allowance will be smaller than what is considered 
acceptable when the same systems and components are inoperable at 
power. This is due to the fact that CTs associated with the majority of 
TS systems and components were developed for power operation and pose a 
smaller plant risk for action statement entries initiated or occurring 
at lower modes of operation as compared to power operation.
    The LCO 3.0.4(b) or SR 3.0.4(b) allowance will be used only when 
the licensee determines that there is a high likelihood that the LCO 
will be satisfied following the mode change. This will minimize the 
likelihood of additional temporary risk increases associated with the 
need to exit a mode due to failure to restore the unavailable equipment 
within the CT. As discussed in Section 3.2, the revised reactor 
oversight process monitors unplanned power changes as a performance 
indicator. The reactor oversight process thus discourages licensees 
from entering a mode or other specified condition in the applicability 
of an LCO, and moving up in power, when there is a likelihood that the 
mode would have to be subsequently exited due to failure to restore the 
unavailable equipment within the CT.
3.1.2  Cumulative Risk Increases
    The cumulative risk impact of the change to allow the plant to 
enter a higher mode of operation with one or more safety-related 
components unavailable (as proposed here), is measured by the average 
yearly risk increase associated with the change. In general, this 
cumulative risk increase is assessed in terms of both CDF and LERF 
(i.e., DCDF and LERF, respectively). The increase in 
CDF due to the proposed change is expressed by the following equation, 
which integrates the risk impact from all expected specified conditions 
(i.e., all expected plant conditions caused by mode changes with 
various TS systems and components unavailable).


[[Page 50481]]


CDF = (CDFi) =  
ICCDPi fi (2)


where

CDFi = the CDF increase due to specified condition 
i
ICCDPi = the ICCDP associated with specified condition i
fi = the average yearly frequency of occurrence of specified 
condition i

    A similar expression can be used for LERF by substituting 
the measure of risk, i.e., LERF for CDF. The magnitude of the 
CDF and LERF values associated with plant conditions 
applicable to LCO 3.0.4(b) or SR 3.0.4(b) allowances can be managed by 
controlling the temporary risk increases, in terms of both CDF and LERF 
(i.e., ICCDP and ICLERP), and the frequency (f), of each of such 
conditions. In addition to the points made in the previous section 
regarding temporary risk increases, the following points put into 
perspective how the key elements of the proposed risk-informed 
approach, used to justify an LCO 3.0.4(b) or SR 3.0.4(b) allowance, are 
expected to prevent significant cumulative risk increases by limiting 
the frequency of its use:
     The frequency of risk significant conditions will be 
limited by not providing the LCO 3.0.4(b) and SR 3.0.4(b) allowances to 
the higher risk systems and components.
     The frequency of risk significant conditions will be 
limited by the requirement to assess the likelihood that the LCO will 
be satisfied following the mode change. In addition, the reactor 
oversight process discourages licensees from entering a mode or other 
specified condition in the applicability of an LCO and moving up in 
power when it is likely that the mode would have to be subsequently 
exited due to failure to restore the unavailable equipment within the 
completion time.
     The frequency of risk significant conditions is limited by 
the fact that such conditions can occur only when the plant is 
returning to power following shutdown, i.e., during a small fraction of 
time per year (data over the past five years indicates that the plants 
are averaging 2.1 startups per year).
    The addition of the proposed LCO 3.0.4(b) or SR 3.0.4(b) allowances 
to the plant maintenance activities is not expected to change the 
plant's average (cumulative) risk significantly.
3.1.3  Risk Assessment and Risk Management of Mode Changes
    With all safety systems and components operable, a plant can 
transition up in mode to power operation. With one or more system(s) or 
component(s) inoperable, this change permits a plant to transition up 
in mode to power operation if the inoperable system(s) or component(s) 
are not in the pre-analyzed higher risk category, a 10 CFR 50.65(a)(4) 
based risk assessment is performed prior to the mode transition, and 
the requisite risk management actions are taken. The proposed TS Bases 
state, ``When an LCO is not met, LCO 3.0.4 also allows entering MODES 
or other specified conditions in the Applicability following assessment 
of the risk impact and determination that the impact can be managed. 
The risk assessment may use quantitative, qualitative, or blended 
approaches, and the risk assessment will be conducted using the plant 
program, procedures, and criteria in place to implement 10 CFR 
50.65(a)(4), which requires that risk impacts of maintenance activities 
to be assessed and managed.'' It should be noted that, the risk 
assessment, for the purposes of LCO 3.0.4(b) and SR 3.0.4(b), must take 
into account all inoperable TS equipment regardless whether the 
equipment is included in the licensee's normal 10 CFR 50.65(a)(4) risk 
assessment scope. The risk assessments will be conducted using the 
procedures and guidance endorsed by Regulatory Guide 1.182, ``Assessing 
and Managing Risk Before Maintenance Activities at Nuclear Power 
Plants.'' The results of the risk assessment shall be considered in 
determining the acceptability of entering the MODE or other specified 
condition in the Applicability, and any corresponding risk management 
actions. * * * A risk assessment and establishment of risk management 
actions, as appropriate, are required for determination of acceptable 
risk for entering MODES or other specified conditions in the 
Applicability when an LCO is not met. Elements of acceptable risk 
assessment and risk management actions are included in Section 11 of 
NUMARC 93-01 ``Assessment of Risk Resulting from Performance of 
Maintenance Activities,'' as endorsed by RG 1.182 which addresses 
general guidance for conduct of the risk assessment, quantitative and 
qualitative guidelines for establishing risk management actions, and 
example risk management actions. These risk management actions include 
actions to plan and conduct other activities in a manner that controls 
overall risk, increased risk awareness by shift and management 
personnel, actions to reduce the duration of the conditions, actions to 
minimize the magnitude of risk increases (establishment of backup 
success paths or compensatory measures), and determination that the 
proposed MODE change is acceptable.
    The guidance references state that a licensee's risk assessment 
process should be sufficiently robust and comprehensive to assess risk 
associated with maintenance activities during power operating, low 
power and shutdown conditions (all modes of operation), including 
changes in plant conditions. NUMARC 93-01 states that the risk 
assessment should include consideration of: the degree of redundancy 
available for performance of the safety function(s) served by the out 
of service equipment; the duration of the out of service condition; 
component and system dependencies that are affected; the risk impact of 
performing the maintenance during shutdown versus at power; and, the 
impact of mode transition risk. For power operation, key plant safety 
functions are those that ensure the integrity of the reactor coolant 
pressure boundary, ensure the capability to shut down and maintain the 
reactor in safe shutdown condition, and ensure the capability to 
prevent or mitigate the consequences of accidents that could result in 
potentially significant offsite exposures.
    While the inoperabilities permitted by the completion times of 
technical specification required actions take into consideration the 
safety significance and redundancy of the system or components within 
the scope of an LCO, the completion times generally do not address or 
consider concurrent system or component inoperabilities in multiple 
LCOs. Therefore, the performance of the 10 CFR 50.65(a)(4) risk 
assessment which looks at the entire plant configuration is essential 
(and required) prior to changing operational mode. The 10 CFR 
50.65(a)(4) risk assessment will confirm (or reject) the 
appropriateness of transitioning up in mode given the actual status of 
plant safety equipment.
    The risk impact on the plant condition of invoking an LCO 3.0.4(b) 
or SR 3.0.4(b) allowance will be assessed and managed through the 
program established to implement 10 CFR 50.65(a)(4). This program is 
consistent with RG 1.177 and RG 1.174 in its approach. The Maintenance 
Rule implementation guidance addresses controlling temporary risk 
increases resulting from maintenance activities. This guidance, 
consistent with guidance in RG 1.177, establishes action thresholds 
based on qualitative and

[[Page 50482]]

quantitative considerations and risk management actions. Significant 
temporary risk increases following an LCO 3.0.4(b) or SR 3.0.4(b) 
allowance are unlikely to occur unless:
     High risk configurations are allowed (e.g., certain 
combinations of multiple component outages), or
     Risk management of plant operation activities is 
inadequate.
    The requirements associated with the proposed change are 
established to ensure that such conditions will not occur.
    The thresholds of the cumulative (aggregate) risk impacts, assessed 
pursuant to 10 CFR 50.65(a)(4) and the associated implementation 
guidance, are based on the permanent change guidelines in NRC RG 1.174. 
Therefore, licensees will manage the risk exercising LCO 3.0.4 or SR 
3.0.4 in conjunction with the risk from other concurrent plant 
activities to ensure that any increase, in terms of core damage 
frequency (CDF) and large early release frequency (LERF) will be small 
and consistent with the Commission's Safety Goal Policy Statement.

3.2  Oversight

    The reactor oversight process (ROP) provides a means for assessing 
the licensee's performance in the application of the proposed mode 
change flexibility. The adequacy of the licensee's assessment and 
management of maintenance-related risk is addressed by existing 
inspection programs and guidance for 50.65(a)(4). Although the current 
versions of that guidance do not specifically address application of 
the licensee's (a)(4) program to support risk-informed technical 
specifications, it is expected that in most cases, risk assessment and 
management associated with risk-informed technical specifications would 
be required by (a)(4) anyway.
    Adoption of the proposed change will make failure to assess and 
manage the risk of an upward mode change with inoperable equipment 
covered by technical specifications, prior to commencing such a mode 
change, a violation of technical specifications. Further, as explained 
above in general, under most foreseeable circumstances, such a change 
in configuration would also require a risk assessment under 10 CFR 
50.65(a)(4). Inoperable systems or components will necessitate 
maintenance to restore them to operability, and hence a 10 CFR 
50.65(a)(4) risk assessment would be performed prior to the performance 
of those maintenance actions (except for immediate plant stabilization 
and restoration actions if necessary). Further, before altering the 
plant's configuration, including plant configuration changes associated 
with mode changes, the licensee must update the existing (a)(4) risk 
assessment to reflect those changes.
    The Federal Register Notice issuing a revision to the maintenance 
rule, 10 CFR 50.65, (Federal Register, Vol 64 No 137, Monday, July 19, 
1999, pg 38553), along with NRC Inspection Procedure 71111.13, and 
Section 11, dated February 22, 2000, ``Assessment of Risk Resulting 
from Performance of Maintenance Activities,'' of NUMARC 93-01, all 
indicate that to determine the safety impact of a change in plant 
conditions during maintenance, a risk assessment must be performed 
before changing plant conditions. The Bases for the proposed TS change 
mandate that the risk assessment and management of upward mode changes 
will be conducted under the licensee's program and process for meeting 
10 CFR 50.65(a)(4). Oversight of licensee performance in assessing and 
managing the risk of plant maintenance activities is conducted 
principally by inspection in accordance with Reactor Oversight Program 
Baseline Inspection Procedure (IP) 71111.13, ``Maintenance Risk 
Assessment and Emergent Work Control.'' Supplemental IP 62709, 
``Configuration Risk Assessment and Risk Management Process,'' is 
utilized to evaluate the licensee's process, when necessary. Appendix B 
of this SE presents excerpts from IP 71111.13 and IP 62709 that provide 
evidence of how the oversight of licensee risk assessment and risk 
management activities is accomplished.
    The ROP is described in overview in NUREG-1649, Rev 3, ``Reactor 
Oversight Process,'' and in detail in the NRC Inspection Manual. 
Inspection Procedure 71111.13 requires verification of performance of 
risk assessments when they are required by 10 CFR 50.65(a)(4) and in 
accordance with licensee procedures. The procedure also requires 
verification of the adequacy of those risk assessments and verification 
of effective implementation of licensee-prescribed risk management 
actions. The rule itself requires such assessment and management of 
risk prior to maintenance activities, including preventive maintenance, 
surveillance and testing, (and promptly for emergent work) during all 
modes of plant operation. The guidance documents for both industry 
implementation of (a)(4) and NRC oversight of that implementation 
indicate that changes in plant configuration (which would include mode 
changes) in support of maintenance activities must be taken into 
account in the risk assessment and management process. Revisions to NRC 
inspection guidance and licensee implementation procedures will be 
needed to address oversight of risk assessment and management required 
by TS in support of mode changes that are not already required under 
the circumstances by (a)(4). This consideration provides performance-
based regulatory oversight of the use of the proposed flexibility, and 
a disincentive to use the flexibility without the requisite care in 
planning.
    In addition, the staff is in the process of developing detailed 
significance determination process (SDP) guidance for use in assessing 
inspection findings related to 10 CFR 50.65(a)(4). This guidance was 
issued in draft for comment and is expected to become final in Fall 
2002. The ROP considers inspection findings and performance indicators 
in evaluating licensee ability to operate safely. The SDP is used to 
determine the significance of inspection findings related to licensee 
assessment and management of the risk associated with performing 
maintenance activities under all plant operating or shutdown 
conditions. Unplanned reactor shutdowns (automatic and manual) and 
unplanned power changes are two of the Reactor Safety Performance 
Indicators that the ROP utilizes to assess licensee performance and 
inform the public. Thus, the ROP provides a disincentive to entering a 
mode or other specified condition in the applicability of an LCO and 
moving up in power, when there is a significant likelihood that the 
mode would have to be subsequently exited due to failure to restore the 
unavailable equipment within the completion time.

3.3  Summary

    The industry, through the Nuclear Energy Institute (NEI) Risk 
Informed Technical Specifications Task Force (RITSTF), has submitted a 
proposed technical specification (TS) change to allow entry into a 
higher mode of operation, or other specified condition in the TS 
applicability, while relying on the TS conditions, and associated 
required actions and completion times, provided a risk assessment is 
performed to confirm the acceptability of that action. The proposal 
revises standard technical specification (STS) LCO 3.0.4 and SR 3.0.4, 
and their application to the TS. New paragraphs (a), (b), and (c) are 
proposed for LCO 3.0.4 and SR 3.0.4.
    The proposed LCO 3.0.4(a) and SR 3.0.4(a) retain the current 
allowance, permitting the mode change when the TS required actions 
allow indefinite operation.

[[Page 50483]]

    Proposed LCO 3.0.4(b) and SR 3.0.4(b) is the change to allow entry 
into a higher mode of operation, or other specified condition in the TS 
applicability, while relying on the TS conditions and associated 
required actions and completion times, provided a risk assessment is 
performed to confirm the acceptability of that action for the existing 
plant configuration. The staff review finds that the process proposed 
by industry for assessing and managing risk during the implementation 
of the proposed LCO 3.0.4(b) and SR 3.0.4(b) allowances, meets 
Commission guidance for technical specification changes. Key elements 
of this process are listed below.
     A risk assessment shall be performed before any LCO 
3.0.4(b) or SR 3.0.4(b) allowance is invoked.
     The risk impact on the plant condition of invoking an LCO 
3.0.4(b) or SR 3.0.4(b) allowance will be assessed and managed through 
the program established to implement 10 CFR 50.65(a)(4) and the 
associated guidance in RG 1.182. Allowing entry into a higher mode or 
condition in the applicability of an LCO after an 10 CFR 50.65(a)(4) 
based risk assessment and appropriate risk management actions are taken 
for the existing plant configuration will ensure that plant safety is 
maintained.
     The LCO 3.0.4(b) or SR 3.0.4(b) allowance will be used 
only when the licensee determines that there is a high likelihood that 
the LCO will be satisfied within the required action's completion time.
     TS systems and components which may be of higher risk 
during mode changes have been identified generically by each owner's 
group for each plant operational mode or condition. Licensees will 
identify such plant specific systems and components in the individual 
plant TS. The proposed LCO 3.0.4(b) and SR 3.0.4(b) allowances do not 
apply to these systems and components for the mode or condition in the 
applicability of an LCO at which they are of higher risk.
     Plants adopting LCO 3.0.4(b) and SR 3.0.4(b) will ensure 
that plant procedures in place to implement 10 CFR 50.65(a)(4) address 
the situation where entering a mode or other specified condition in the 
applicability is contemplated with plant equipment inoperable. Such 
plant procedures typically follow the guidance in NUMARC 93-01, Section 
11, as revised in February 2000 and endorsed by NRC RG 1.182.
    The NRC's reactor oversight process provides the framework for 
inspectors and other staff to oversee the implementation of 10 CFR 
50.65(a)(4) requirements at a specific plant and assess the licensee's 
actions and performance.
    The LCO 3.0.4(b) and SR 3.0.4(b) allowance does not apply to values 
and parameters of the technical specifications that have their own 
respective LCOs (e.g., Reactor Coolant System Specific Activity), but 
instead those values and parameters are addressed by LCO 3.0.4(c) and 
SR 3.0.4(c). The TS values and parameters for which mode transition 
allowances apply, will have a note that states LCO 3.0.4(c) or SR 
3.0.4(c) is applicable.
    The objective of the proposed change is to provide additional 
operational flexibility without compromising plant safety.

4.0  State Consultation

    In accordance with the Commission's regulations, the [ ] State 
official was notified of the proposed issuance of the amendment. The 
State official had [(1) no comments or (2) the following comments--with 
subsequent disposition by the staff].

5.0  Environmental Consideration

    The amendments change a requirement with respect to the 
installation or use of a facility component located within the 
restricted area as defined in 10 CFR part 20 and change surveillance 
requirements. [For licensees adding a Bases Control Program: The 
amendment also changes record keeping, reporting, or administrative 
procedures or requirements.] The NRC staff has determined that the 
amendments involve no significant increase in the amounts and no 
significant change in the types of any effluents that may be released 
offsite, and that there is no significant increase in individual or 
cumulative occupational radiation exposure. The Commission has 
previously issued a proposed finding that the amendments involve no 
significant hazards considerations, and there has been no public 
comment on the finding. Accordingly, the amendments meet the 
eligibility criteria for categorical exclusion set forth in 10 CFR 
51.22(c)(9) [and (c)(10)]. Pursuant to 10 CFR 51.22(b), no 
environmental impact statement or environmental assessment need be 
prepared in connection with the issuance of the amendments.

6.0  Conclusion

    The Commission has concluded, on the basis of the considerations 
discussed above, that (1) there is reasonable assurance that the health 
and safety of the public will not be endangered by operation in the 
proposed manner, (2) such activities will be conducted in compliance 
with the Commission's regulations, and (3) the issuance of the 
amendments will not be inimical to the common defense and security or 
to the health and safety of the public.

Proposed No Significant Hazards Consideration Determination

    Description of Amendment Request: A change is proposed to the 
standard technical specifications (STS) (NUREGs 1430 through 1434) and 
plant specific technical specifications (TS), to allow entry into a 
mode or other specified condition in the applicability of a TS, while 
in a condition statement and the associated required actions of the TS, 
provided the licensee performs a risk assessment and manages risk 
consistent with the program in place for complying with the 
requirements of 10 CFR 50.65(a)(4). LCO 3.0.4 and SR 3.0.4 exceptions 
in individual TS would be eliminated, and SR 3.0.4 revised to reflect 
the LCO 3.0.4 allowance.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a TS 
condition and the associated required actions is not an initiator of 
any accident previously evaluated. Therefore, the probability of an 
accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the consequences 
of an accident while entering and relying on the required actions while 
starting in a condition of applicability of the TS. Therefore, the 
consequences of an accident previously evaluated are not significantly 
affected by this change. The addition of a requirement to assess and 
manage the risk introduced by this change will further minimize 
possible concerns. Therefore, this change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.

[[Page 50484]]

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From any Previously Evaluated

    The proposed change does not involve a physical alteration of the 
plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the applicability 
of a TS, while in a TS condition statement and the associated required 
actions of the TS, will not introduce new failure modes or effects and 
will not, in the absence of other unrelated failures, lead to an 
accident whose consequences exceed the consequences of accidents 
previously evaluated. The addition of a requirement to assess and 
manage the risk introduced by this change will further minimize 
possible concerns. Thus, this change does not create the possibility of 
a new or different kind of accident from an accident previously 
evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS allow 
operation of the plant without the full complement of equipment through 
the conditions for not meeting the TS Limiting Conditions for Operation 
(LCO). The risk associated with this allowance is managed by the 
imposition of required actions that must be performed within the 
prescribed completion times. The net effect of being in a TS condition 
on the margin of safety is not considered significant. The proposed 
change does not alter the required actions or completion times of the 
TS. The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's performance 
of a risk assessment and the management of plant risk. The change also 
eliminates current allowances for utilizing required actions and 
completion times in similar circumstances, without assessing and 
managing risk. The net change to the margin of safety is insignificant. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.
    Based upon the reasoning presented above and the previous 
discussion of the amendment request, the requested change does not 
involve a significant hazards consideration.

    Dated at Rockville, Maryland, this 26th day of July, 2002.

    For the Nuclear Regulatory Commission.
Robert L. Dennig,
Section Chief, Technical Specifications Section, Operating Improvements 
Branch, Division of Regulatory Improvement Programs, Office of Nuclear 
Reactor Regulation.

Appendix A

LCO 3.0.4  Examples

    Example 1, LCO 3.0.4(a), (NUREG-1431): The plant is in Mode 3 
ready to go to Mode 1, power operation, with one power range neutron 
flux channel inoperable. LCO 3.3.1, ``Reactor Trip System (RTS) 
Instrumentation,'' Table 3.3.1-1, Function 2.a., requires four power 
range neutron flux-high channels to be operable, and the 
applicability is Modes 1 and 2. With one power range neutron flux-
high channel inoperable, Condition D, Required Actions D.1.1 and 
D.1.2 require the inoperable channel to be placed in trip within 6 
hours and reduce thermal power to  75% RTP within 12 
hours; or, Required Actions D.2.1 and D.2.2 require placing the 
inoperable channel in trip within 6 hours and verifying QPTR is 
within limits (performance of SR 3.2.4.2) once per 12 hours. 
Verifying QPTR is within limits is only required if the power range 
neutron flux input to QPTR is inoperable. The plant can proceed to 
Mode 2 (or further, i.e., Mode 1) as long as the Required Actions of 
Condition D are met. If the plant has proceeded to Mode 2 (or 
further, i.e., Mode 1) and the Required Actions of Condition D have 
not been met, the plant must be placed in Mode 3. No risk assessment 
is required because the allowance of LCO 3.0.4(a) applies. However, 
risk assessment may be required by 10 CFR 50.65(a)(4).
    Example 2, LCO 3.0.4(b), (NUREG-1431): The plant is in Mode 5 
ready to go to Mode 1, power operation, with one component cooling 
water (CCW) train inoperable. LCO 3.7.7, ``Component Cooling Water 
(CCW),'' requires two CCW trains to be operable and the 
applicability is Modes 1, 2, 3, and 4. With one CCW train inoperable 
Required Action A.1 of LCO 3.7.7 requires the inoperable CCW train 
to be restored and the completion time is 72 hours. There is also a 
note applied to Required Action A.1 that requires entry into 
applicable Conditions and Required Actions of LCO 3.4.6, ``RCS 
Loops--MODE 4,'' for residual heat removal loops made inoperable by 
CCW. If a residual heat removal loop is being used to comply with 
LCO 3.4.6, and that loop is made inoperable by the inoperable CCW 
train, the completion times for the applicable conditions and 
required actions of LCO 3.4.6 may be more restrictive than those of 
LCO 3.7.7. The plant can proceed to Mode 4 if there is reasonable 
assurance that the inoperable CCW train can be restored to operable 
status within the applicable completion time, and a risk assessment 
has been performed and requisite risk management actions have been 
implemented. If the plant has proceeded to Mode 4 (or further, i.e., 
Mode 3, 2, or 1) and the inoperable CCW train has not been restored 
within the required completion time, the plant must return to Mode 
5. Note that if two trains of CCW are inoperable, the plant cannot 
proceed to Mode 4 because LCO 3.7.7 does not contain a condition for 
two inoperable CCW trains.
    Example 3, LCO 3.0.4(b), (NUREG-1431): The plant is in Mode 5 
ready to go to Mode 1, power operation (with steam generators 
operable). In Case 1, one required Atmospheric Dump Valve (ADV) line 
is inoperable. In Case 2, two or three required ADV lines are 
inoperable.
    Case 1--LCO 3.7.4, ``Atmospheric Dump Valves (ADVs),'' requires 
three ADV lines to be operable and the Applicability is Modes 1, 2, 
and 3 and Mode 4 when steam generator is relied upon for heat 
removal. With one required ADV line inoperable Required Action A.1 
requires the required ADV line to be restored with a Completion Time 
of seven days. The plant can proceed to Mode 4 (when steam 
generator(s) are relied on for heat removal) provided there is 
reasonable assurance that the required ADV line can be restored 
within 7 days, and a risk assessment has been performed and 
requisite risk management actions have been implemented. If the 
plant has proceeded to Mode 4 (or further, i.e., Mode 3, 2, or 1) 
and the required ADV is not restored within 7 days, the plant must 
return to Mode 5 (if steam generator(s) are being used for heat 
removal) or Mode 4 where steam generators are not being used for 
heat removal, as applicable.
    Case 2--With two or three required ADV lines inoperable, 
Condition B, Required Action B.1 requires restoration of all but one 
of the required ADV lines within a Completion Time of 24 hours. The 
plant can proceed to Mode 4 (when steam generators are relied on for 
heat removal) provided there is reasonable assurance that the 
required ADV lines will be restored, and a risk assessment has been 
performed and requisite risk management actions have been 
implemented. After the plant has restored all but one of the 
required ADV lines to operability within 24 hours, the final 
required ADV line must be restored within seven days from the time 
of entry into Mode 4. If the plant has proceeded to Mode 4 (or 
further, i.e., Mode 3, 2 or 1) and the required ADV lines have not 
been restored within the applicable completion time, the plant must 
return to Mode 5 or Mode 4 (where steam generators are not relied on 
for heat removal).

Appendix B

Reactor Oversight Process, Inspection Procedures 71111.13 and 62709 
Excerpts

Inspection Procedure (IP) 71111.13, ``Maintenance Risk Assessment 
and Emergent Work Control''

IP 71111.13-02, Inspection Requirements, 02.01, Risk Assessment and 
Management of Risk

    a. Risk Assessment Performance. Verify performance of risk 
assessments when required by 10 CFR 50.65(a)(4) and in accordance 
with licensee procedures, prior to changes in plant configuration 
for maintenance activities, including preventive maintenance, 
surveillance and testing, (and promptly for emergent work) during 
all

[[Page 50485]]

modes of plant operation. Verify risk assessment performance for 
configuration changes involving structures, systems or components * 
* *
    b. Risk Assessment Adequacy. Verify the accuracy and 
completeness of the information considered in the risk assessment. 
Verify the appropriate use of the risk assessment tool, i.e., that 
the licensee uses it in a manner consistent with (1) its 
capabilities and limitations, (2) plant conditions and evolutions, 
(3) external events and containment status, and (4) licensee 
procedures. * * *
    c. Risk Management. Verify that the licensee recognizes, and/or 
enters as applicable, the appropriate licensee-established risk 
category or band according to risk assessment results and licensee 
procedures. Verify that normal work controls or risk management 
actions as required are promptly and effectively implemented 
commensurate with the risk band in effect and in accordance with 
licensee procedures. Verify that the key safety functions for the 
plant mode of operation are preserved. * * *

IP 71111.13, Appendix A, Risk Assessment Performance Verification Phase

    ``Determine if a Risk Assessment (RA) was required using the 
following criteria:
    1. When required. RAs are required by (a)(4) prior to 
maintenance-related plant configuration changes and are normally 
performed for scheduled maintenance. However, emergent conditions, 
such as external events or SSC failures or degraded performance in 
service or during testing, may require actions prior to performing 
an RA, or could invalidate the existing RA. In this case, the RA 
should be performed (or reevaluated) to address the changed plant 
conditions. The industry guidance, revised Section 11 of NUMARC 
93001, as endorsed by RG 1.182, states that if the plant 
configuration is restored prior to conducting or reevaluating the 
RA, the RA need not be conducted, or reevaluated if already 
performed. Nevertheless, to the extent practicable and commensurate 
with safety, the licensee should perform or reevaluate the RA before 
changing the plant configuration further, but in any case, promptly 
and to the extent practicable concurrently with, but without 
delaying, plant stabilization and restoration. Note that licensee 
deviation from work schedules and work plans, just as emergent work 
can, may invalidate risk assessments prepared for the maintenance 
period (e.g., the common 12-week rolling schedule).
    2. Operating Modes When RA Required. RAs are required by (a)(4) 
for maintenance activities performed during all modes of plant 
operation and transitions between modes. For (a)(4) purposes, at 
power means normal steaming (Mode 1) and startup (Mode 2). Shutdown 
means hot standby (Mode 3 in a pressurized water reactor (PWR) 
only), hot shutdown (Mode 3 in a boiling water reactor, Mode 4-PWR), 
cold shutdown (Mode 5), and refueling (Mode 6). Plants without a 
shutdown probabilistic risk assessment (PRA) must still assess 
shutdown maintenance risk by some means, typically an expert panel 
using a qualitative (key safety function) or blended qualitative/
quantitative approach. * * *''

Supplemental IP 62709, ``Configuration Risk Assessment and Risk 
Management Process'

IP62709

    An appropriate assessment would include a review of the current 
configuration of the plant and the plant configuration expected 
during the planned maintenance activity. Assessing the current plant 
configuration as well expected changes to plant configuration due to 
the planned maintenance activities is intended to insure that the 
plant is not inadvertently placed in risk-significant 
configurations. * * * Furthermore, assessing the degree of safety 
function degradation requires that there be an understanding of the 
impact of maintenance activities on the capability of the plant to 
prevent or mitigate accidents and transients, as well as the 
potential impact of external conditions (e.g., inclement weather, 
electrical grid instability, flooding or seismic events) on plant 
maintenance configurations. The assessments may range from 
deterministic judgments to the use of an on-line PSA tool. * * * The 
process for performing these safety assessments should be scrutable 
and repeatable. Known limitations in the assessment process should 
be described in the licensee's Maintenance Rule program 
documentation. The licensee's process should be sufficiently robust 
and comprehensive to assess maintenance activities during power 
operating conditions and low power and shutdown conditions. The 
sophistication of the assessment(s) for evaluating the risk of a 
maintenance configuration should be commensurate with the complexity 
of the configuration.
    IP 62709, 02.02 Configuration Risk Assessments: Determine if the 
licensee has adequately assessed the overall effect on the 
performance of safety functions when SSCs are removed from service 
for surveillance or maintenance activities. Obtain plant operating/
maintenance records for at least two or three monthly periods of 
high maintenance activities during power operation with a particular 
focus on periods when trains of components were removed from service 
or when components of different trains were out of service 
simultaneously for surveillance or maintenance. In the case of plant 
shutdown conditions, select two or three weekly periods of plant 
outage surveillance or maintenance activities with a particular 
focus on periods of reduced reactor coolant system inventory, 
reduced shutdown cooling availability, or reduced electrical 
availability. Evaluate the results of the licensee's safety 
assessments of those time periods, and verify the licensee's safety 
assessments encompassed all the SSCs that have significant impact on 
public health and safety. If the licensee had not kept records of 
prior assessment results, * * * consider performing independent 
assessments of current maintenance activities.
    IP 62709, 02.03 Risk Management: Determine if a licensee is 
using a reasonable approach to manage risk of the planned 
configurations when SSCs are removed from service for surveillance 
or maintenance activities. On the basis of licensee's safety 
assessments of those selected maintenance configurations, either 
during power operation or shutdown conditions, verify that the 
licensee has process controls in place that ensure risk management 
actions would be implemented for plant maintenance configurations 
with risk increases that exceed risk management thresholds.''

[FR Doc. 02-19538 Filed 8-1-02; 8:45 am]
BILLING CODE 7590-01-P