[Federal Register Volume 67, Number 135 (Monday, July 15, 2002)]
[Notices]
[Pages 46542-46547]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-17649]


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NUCLEAR REGULATORY COMMISSION


Notice of Opportunity To Comment on Model Safety Evaluation on 
Technical Specification Improvement To Revise the Completion Time From 
1 Hour To 24 Hours for Condition B of Technical Specification 3.5.1, 
``Accumulators,'' and Its Associated Bases, 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 the completion time from 1 hour to 24 
hours for Condition B of Technical Specification (TS) 3.5.1, 
``Accumulators,'' and its associated Bases. 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 
revise the completion time from 1 hour to 24 hours for Condition B of 
TS 3.5.1, ``Accumulators,'' and its associated Bases. Licensees of 
nuclear power reactors to which the models apply could request 
amendments confirming the applicability of the SE and NSHC 
determination to their reactors. The NRC staff is requesting comments 
on the model SE and model NSHC determination prior to announcing their 
availability for referencing in license amendment applications.

DATES: The comment period expires August 14, 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, MD.
    Comments may be submitted by electronic mail to [email protected].

FOR FURTHER INFORMATION CONTACT: Girija Shukla, Project Manager, Mail 
Stop: O-7E1, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, telephone 301-415-8439.

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. This is accomplished 
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 following a preliminary 
assessment by the NRC staff and finding that the change will likely be 
offered for adoption by licensees. This notice is soliciting comment on 
a proposed change to the STS that modifies requirements regarding 
missed surveillances. 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 to proceed with announcing the availability of 
the change for proposed adoption by licensees. Those licensees opting 
to apply for the subject change to technical specifications 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 would be processed and noticed in accordance 
with applicable rules and NRC procedures.
    This notice involves the revision of the accumulators completion 
time from 1 hour to 24 hours in TSs. This proposed change was proposed 
for incorporation into the STSs by all Owners Groups participants in 
the Technical Specification Task Force (TSTF) and is designated TSTF-
370. TSTF-370 can be viewed on the NRC's Web page at http://www.nrc.gov/reactors/operating/licensing/techspecs/.

Applicability

    This proposed change to modify TS to revise the accumulators 
completion time from 1 hour to 24 hours is applicable to all 
Westinghouse nuclear steam supply system (NSSS) plants regardless of 
plant vintage and number of loops.
    The CLIIP does not prevent licensees from requesting an alternative 
approach or proposing the changes without the attached model SE and the 
NSHC. Variations from the approach recommended in this notice may, 
however, 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. 
Following the staff's evaluation of comments received as a result of 
this notice, the staff may reconsider the proposed change or may 
proceed with announcing the availability of the change in a subsequent 
notice (perhaps with some changes to the safety evaluation or 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 will submit an 
application in

[[Page 46543]]

accordance with applicable rules and other regulatory requirements. The 
staff will in turn issue for each application a notice of consideration 
of issuance of amendment to facility operating license(s), a proposed 
no significant hazards consideration determination, and an opportunity 
for a hearing. A notice of issuance of an amendment to operating 
license(s) will also be issued to announce the revision to the 
completion time for Condition B of TS 3.5.1, ``Accumulators,'' and its 
associated Bases for each plant that applies for and 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-370, Risk-Informed Evaluation of an Extension 
to Accumulator Completion Times for Westinghouse Plants

1.0  INTRODUCTION

    The Nuclear Energy Institute (NEI) Technical Specification Task 
Force (TSTF) has proposed a generic change to the standard technical 
specifications (STSs) (NUREG-1431) on behalf of the industry. This 
proposed generic technical specifications (TSs) change, identified by 
TSTF-370, will revise the completion time (CT) from 1 hour to 24 hours 
for Condition B of Technical Specification (TS) 3.5.1, 
``Accumulators,'' and its associated Bases. Condition B of TS 3.5.1 
currently specifies a CT of one hour to restore a reactor coolant 
system (RCS) accumulator to operable status when declared inoperable 
due to any reason except not being within the required boron 
concentration range.

2.0  BACKGROUND

    Topical Report WCAP-15049, ``Risk-Informed Evaluation of an 
Extension to Accumulator Completion Times,'' was submitted to the NRC 
on August 20, 1998, and approved in the NRC letter dated February 19, 
1999. The WCAP evaluates the risk associated with extending the 
accumulator CT from 1 hour to 24 hours for reasons other than boron 
concentration out of specification.
    Wolf Creek was the lead plant for the Westinghouse Owners Group 
(WOG) program and received plant specific approval for changes to the 
TSs on April 27, 1999 (License Amendment No. 124). In the NRC letter of 
February 19, 1999, the staff indicated that it will not repeat its 
review of the matters described in Topical Report WCAP-15049 when the 
report appears as a reference in license applications, except to ensure 
that the material presented applies to the specified plants involved.
    The proposed change revises the CT from 1 hour to 24 hours for 
Condition B of TS 3.5.1, ``Accumulators,'' and its associated Bases. 
Condition B of TS 3.5.1 currently specifies a CT of one hour to restore 
a RCS accumulator to operable status when declared inoperable due to 
any reason except not being within the required boron concentration 
range.

3.0  EVALUATION

Deterministic Evaluation

    The purpose of the emergency core cooling system (ECCS) 
accumulators is to supply water to the reactor vessel during the 
blowdown phase of a loss-of-coolant accident (LOCA). The accumulators 
are large volume tanks, filled with borated water and pressurized with 
nitrogen. The cover-pressure is less than that of the reactor coolant 
system so that following an accident, when the reactor coolant system 
pressure decreases below tank pressure, the accumulators inject the 
borated water into the RCS cold legs. The current deterministic safety 
analysis has not been changed, and thus the limiting condition of 
operation (LCO), i.e., the lowest functional capability required for 
safe operation continues to be:

``LCO 3.5.1, [Four] ECCS accumulators shall be operable.
Applicability: Modes 1 and 2, Mode 3 with RCS pressure  
[1000] psig.''
Where the bracketed information is nominal, and is subject to 
substitution of plant specific values.

    Under Actions, TSs allow for limited deviations from the LCO. 
Historically, these Actions and associated CTs have been set using 
judgement and are not part of the deterministic safety analysis 
discussed above. Currently, the TS allows for one accumulator to be 
inoperable for one hour for reasons other than boron concentration not 
within limits during Modes 1, 2, and in Mode 3 with pressurizer 
pressure  a plant specific pressure. The WCAP, as well as 
this TSTF, proposes to increase this CT to 24 hours. The proposed CT of 
24 hours is an extension of the current ACTION statement. CTs are by 
their nature determined by conditions of risk and the impact of the 
proposed change on risk is reviewed in the following section.

Risk Evaluation

    A three-tiered approach, consistent with RG 1.177,\1\ was used by 
the staff to evaluate the risk associated with the proposed accumulator 
CT, or allowed outage time (AOT), extension from 1 hour to 24 hours. 
The need for the proposed change was that the current one-hour CT would 
be insufficient in most cases for licensees to take a reasonable action 
when an accumulator was found to be inoperable.
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    \1\ RG 1.177, ``An Approach for Plant-Specific, Risk-Informed 
Decisionmaking: Technical Specifications,'' September 1998.
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Tier 1: Quality of Probabilistic Risk Assessment (PRA) and Risk Impact
    Westinghouse used a reasonable approach to assess the risk impact 
of the proposed accumulator CT extension. The approach is generally 
consistent with the intent of the applicable NRC RGs 1.174 \2\ and 
1.177. The quantitative risk measures addressed in the topical report 
included the change in core damage frequency (CDF) and incremental 
conditional core damage probability (ICCDP \3\) for a single CT. The 
change in large early release frequency (LERF) and incremental 
conditional large early probability (ICLERP\4\) for a single CT was 
qualitatively addressed. Representative calculations were performed to 
determine the risk impact of the proposed change. Various accumulator 
success criteria were considered in these calculations to encompass the 
whole spectrum of Westinghouse plants, e.g., two-, three- and four-loop 
plants. A reasonable effort was also made to address the differences in 
other components of risk analysis such as initiating event (IE) 
frequency and accumulator unavailability among Westinghouse plants.
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    \2\ RG 1.174, ``An Approach for Using Probabilistic Risk 
Assessment in Risk-Informed Decisions on Plant-Specific Changes to 
the Licensing Basis,'' July 1998.
    \3\ ICCDP = [(conditional CDF with the subject equipment out-of-
service)--(baseline CDF with nominal expected equipment 
unavailabilities) x (duration of single CT under consideration)].
    \4\ ICLERP = [(conditional LERF with the subject equipment out-
of-service)--(baseline LERF with nominal expected equipment 
unavailabilities) x (duration of single CT under consideration)].
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    Westinghouse considered a comprehensive range of IEs in the risk 
analysis. LOCAs in all sizes--large, medium and small--were included, 
and reactor vessel failure and interfacing system LOCA were also 
considered. Modeling of accumulators for mitigation of events other 
than large, medium and small LOCAs was identified to have insignificant 
risk impact; therefore, the analysis was performed only on

[[Page 46544]]

accumulator injection in response to large, medium and small LOCA 
events.
    The success criteria considered are summarized as follows:

------------------------------------------------------------------------
         LOCA category            No. of loops       Success criteria
------------------------------------------------------------------------
Large..........................               4  3 accumulators to 3 of
                                                  3 intact loops (3/3);
                                                  2 accumulators to 2 of
                                                  3 intact loops (2/3);
                                                  no accumulators
                                                  required (0/3).
                                              3  2 accumulators to 2 of
                                                  2 intact loops (2/2);
                                                  1 accumulator to 1 of
                                                  2 intact loops (1/2);
                                                  no accumulators
                                                  required (0/2).
Medium and Small...............               2  1 accumulator to 1 of 1
                                                  intact loop (1/1); no
                                                  accumulators required
                                                  (0/1).
                                              4  3 accumulators to 3 of
                                                  3 intact loops (3/3).
                                              3  2 accumulators to 2 of
                                                  2 intact loops (2/2).
                                              2  1 accumulator to 1 of 1
                                                  intact loop (1/1).
------------------------------------------------------------------------

    The success criteria considered in this analysis were comprehensive 
and considered conservative in many cases. For example, many plants 
indicated the accumulator success criteria for medium and small LOCA 
events resulted from their role in an alternate success path, in which 
high pressure injection (HPI) had already failed. Additionally, the 
staff's review of a number of the original individual plant 
examinations (IPEs) indicated that no accumulator was needed at all for 
many medium LOCA sequences and for most of small LOCA sequences.
    The fault trees that model accumulator unavailabilities were 
evaluated. The assumptions made in the fault tree modeling were 
detailed and were found to be reasonable. For example, the model 
assumed that the total CT would be used for each corrective 
maintenance, and this was considered conservative. A comprehensive list 
of failure mechanisms was considered, and potential common cause 
failures for check valves and motor-operated valves were also included. 
Westinghouse used the Multiple Greek Letter technique to determine the 
common cause failure contributions to the accumulator injection 
failure.
    The component failure rates were taken from the Advanced Light 
Water Utility Requirements Document.\5\ Accumulator unavailabilities 
due to boron concentration out of limit and due to other reasons were 
calculated based on a survey of a number of Westinghouse plants. The 
values for component failure rates and accumulator unavailabilities 
were within reasonable range. The common cause factors used were also 
comparable to those used in other PRAs. The accumulator fault trees 
were quantified using the WesSAGE computer code. The code provided 
information on the unavailability and cutsets related to the component 
failures and maintenance activities modeled in the fault trees. A 
separate hand calculation was used to determine the unavailability due 
to potential common cause failures. Evaluation of some of the cutsets 
provided in the topical report did not reveal any unexpected results.
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    \5\ ``Advanced Light Water Utility Requirements Document,'' 
Volume II, ALWR Evolutionary Plant, Chapter 1, Appendix A, PRA Key 
Assumptions and Ground Rules, Rev. 5, Issued December 1992.
---------------------------------------------------------------------------

    The staff examined the accident sequence identification for each 
LOCA category. The probability of the sequence leading to core damage 
involving accumulator failure is summarized for each LOCA category as 
follows:
    Large LOCA: (Large LOCA IE frequency) x (accumulator 
unavailability).
    Medium LOCA: (Medium LOCA IE frequency) x (unavailability of HPI) x 
(accumulator unavailability).
    Small LOCA: (Small LOCA IE frequency) x (unavailability of HPI) x 
(accumulator unavailability).
    The LOCA IE frequencies used for WCAP-15049 are summarized below. 
Also listed are the LOCA frequencies used in NUREG/CR-4550 \6\ (the 
NUREG-1150 study) for pressurized water reactors (PWRs) and those in 
the original IPEs.
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    \6\ NUREG/CR-4550, ``Analysis of Core Damage Frequency: Internal 
Events Methodology,'' Vol. 1, Rev. 1, January 1990.

----------------------------------------------------------------------------------------------------------------
                                           WCAP-15049              NUREG-1150          IPE Average (High; Low)
----------------------------------------------------------------------------------------------------------------
Large LOCA.........................  3 x -4/yr.............  5 x 10-4/yr...........  3.3 x 10-4/yr (5 x 10-4/yr;
                                                                                      1 x 10-5/yr).
Medium LOCA........................  8 x 10-4/yr...........  1 x 10-3/yr...........  7.9 x 10-4/yr (2.6 x 10-3/
                                                                                      yr; 1 x 10-4/yr).
Small LOCA.........................  7 x 10-3/yr...........  1 x 10-3/yr...........  8.9 x 10-3/yr (2.9 x 10-2/
                                                                                      yr; 3.7 x 10-4/yr).
----------------------------------------------------------------------------------------------------------------

    Westinghouse indicated that the IE frequencies for WCAP-15049 were 
based on the plant-specific information contained in the Westinghouse 
Owners Group (WOG) PSA Comparison Database, which documented the PRA 
modeling methods and results of the updated PRAs for Westinghouse 
plants. The mean IE frequencies were used for the risk analysis. These 
were comparable to the values used for the NUREG-1150 study and the 
average values in the original IPEs. The staff also found that the IE 
frequency values in high range among the original IPEs were not much 
higher than those used for this topical report. The HPI unavailability 
values used were 7 x 10-3 and 1 x 10-3/yr for 
medium and small LOCA events, respectively. The staff's examination 
revealed that the HPI unavailability values were generally comparable 
to those used in other PRAs, and were generally conservative.
    The risk measures calculated to determine the impact on plant risk 
were based on three different cases. The risk measures considered in 
each case included the impact on CDF and ICCDP for a single CT, and the 
impact on LERF and ICLERP for a single CT were qualitatively 
considered. The three cases considered were:
    Design basis case. This case required accumulator injection only 
for mitigation of large LOCA events (3/3 for 4-loop, 2/2 for 3-loop, 
and 1/1 for 2-loop).
    Case 1. This case credited realistic accumulator success criteria 
(2/3 for 4-loop, 1/2 for 3-loop, and 0/1 for 2-loop) for large LOCA 
events and credited the use of accumulators in responding to medium and 
small LOCA events (3/3, 2/

[[Page 46545]]

2, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) following 
failure of HPI.
    Case 2. This case credited more realistic improved accumulator 
success criteria (no accumulator required) for large LOCA events and 
credited the use of accumulators in responding to medium and small LOCA 
events (3/3, 2/2, and 1/1 for 4-loop, 3-loop, and 2-loop, respectively) 
following failure of HPI.
    The results were summarized as follows:

----------------------------------------------------------------------------------------------------------------
                                    LOCA CDF (/yr)      LOCA CDF (/yr)
              Case                     (Current)          (Proposed)           [Delta]CDF            ICCDP
----------------------------------------------------------------------------------------------------------------
4-loop Design Basis.............  6.93 x 10-7.......  9.24 x 10-7.......  2.31 x 10-7.......  8.20 x 10-7
4-loop Case 1...................  6.23 x 10-8.......  7.77 x 10-8.......  1.54 x 10-8.......  5.53 x 10-8
4-loop Case 2...................  4.57 x 10-8.......  6.09 x 10-8.......  1.52 x 10-8.......  5.41 x 10-8
3-loop Design Basis.............  4.62 x 10-7.......  6.18 x 10-7.......  1.56 x 10-7.......  8.21 x 10-7
3-loop Case 1...................  4.27 x 10-8.......  5.31 x 10-8.......  1.04 x 10-8.......  5.48 x 10-8
3-loop Case 2...................  3.05 x 10-8.......  4.08 x 10-8.......  1.03 x 10-8.......  5.42 x 10-8
2-loop Design Basis.............  2.31 x 10-7.......  3.09 x 10-7.......  7.80 x 10-8.......  8.21 x 10-7
2-loop Case 1...................  1.52 x 10-8.......  2.04 x 10-8.......  5.20 x 10-9.......  5.42 x 10-8
2-loop Case 2...................  1.52 x 10-8.......  2.04 x 10-8.......  5.20 x 10-9.......  5.42 x 10-8
----------------------------------------------------------------------------------------------------------------

    For both realistic cases, the [Delta]CDFs and ICCDPs were very 
small for 2-loop, 3-loop, and 4-loop plants, and were much below the 
numerical guidelines in the RGs 1.174 and 1.177. The staff also noted 
that the values were considered still bounding in the sense that the 
risk analysis used a multitude of conservative assumptions and data in 
the modeling. For many Westinghouse plants, the realistic impact on 
risk would be much smaller than the values above.
    A set of sensitivity cases were also calculated using higher IE 
frequencies for small and medium LOCAs. The results of the sensitivity 
calculations did not cause the overall risk impact to increase 
significantly.
    Westinghouse indicated that accumulator success or failure has no 
direct impact on the containment performance, and that the LERF would 
therefore increase only in direct proportion to the increased CDF due 
to accumulator failures. Westinghouse concluded that, since the impact 
on CDF was small, the impact on LERF would also be small. The staff 
found the Westinghouse argument to be acceptable; therefore, the impact 
on LERF and ICLERP for a single CT was very small.
    One of the potential benefits of the proposed extended CT was the 
averted risk associated with avoiding a forced plant shutdown and 
startup. The risk associated with a forced plant shutdown and ensuing 
startup due to the inflexibility in current TS could be significant in 
comparison with the risk increase due to the proposed accumulator CT 
increase.
    Based on the staff's Tier 1 review, the quality of risk analysis 
used to calculate the risk impact of the proposed accumulator CT 
extension was reasonable and generally conservative. It was also found 
that the risk impact of the proposed change was below the staff 
guidelines in RGs 1.174 and 1.177.
Tier 2 and 3: Configuration Risk Control
    Tier 2 of RG 1.177 addresses the need to preclude potentially high 
risk configurations which could result if certain equipment is taken 
out-of-service during implementation of the proposed TS change (in this 
case accumulator CT). If such configurations are identified, the 
licensee should also identify appropriate measures to avoid them.
    The accumulators are always needed to mitigate large size LOCAs. 
Large LOCAs require accumulators to inject as analyzed under Tier 1 in 
order to avoid core damage. This means that if a large LOCA occurs 
without the accumulator function, the core will be damaged 
independently of whether other systems, such as HPI, function properly 
or not. However, the probability that a large LOCA occurs in the 24-
hour CT is extremely small (in the order of 1E-7 or less). Furthermore, 
no compensatory or other measures are possible. Due to the negligible 
risk increase associated with this scenario and the fact that there are 
no measures to take once a large LOCA occurs, no ``high risk'' 
configurations are associated with this scenario.
    In general, medium LOCAs do not require accumulators if at least 
one HPI train is available. This means that if a medium LOCA occurs 
when minimum accumulator functionality is unavailable and at the same 
time HPI is unavailable, the core will be damaged. However, the 
probability that a medium LOCA occurs in the 24-hour CT and at the same 
time both trains of HPI are unavailable is extremely small (in the 
order of 1E-8 or less), because it is assumed that the plant is not 
operating at power with both HPI trains out-of-service. This assumption 
is based on current STS that limit operation at power with no HPI 
capability. Therefore, no Tier 2 restrictions beyond those currently in 
the STS are deemed necessary.
    Tier 3 calls for a program to identify ``risk significant'' 
configurations beyond those identified in Tier 2 resulting from 
maintenance or other operational activities and take appropriate 
compensatory measures to avoid such configurations. Because the 
accumulator sequence modeling is relatively independent of that for 
other systems, the Tier 2 analysis by itself is sufficient.
    Furthermore, 10 CFR 50.65(a)(4) (Maintenance Rule) requires that 
licensees assess the risk any time maintenance is being considered on 
safety-related equipment. This requirement serves the objectives of 
Tier 3.
    In summary, the Tier 2 evaluation did not identify the need for any 
additional constraints or compensatory actions that, if implemented, 
would avoid or reduce the probability of a risk-significant 
configuration. The current TS provisions were found to be sufficient to 
address the Tier 2 issue. Because the accumulator sequence modeling is 
relatively independent of that for other systems and the implementation 
of the Maintenance Rule, the staff concluded that application of Tier 3 
to the proposed accumulator CT was not necessary.
    The NRC staff finds that the proposed changes will allow safe 
operation with the changes in CT from 1 hour to 24 hours for Condition 
B of TS LCO 3.5.1, ``Accumulators,'' and its associated Bases. The NRC 
staff also finds that the proposed changes are consistent with the 
incremental conditional core damage probabilities calculated in WCAP-
15049 for the accumulator allowed outage time increase and meet the 
criterion of 5E-07 in RGs 1.174 and 1.177. The analysis and acceptance 
provided in this SE, as demonstrated by WCAP-15049, covers all 
Westinghouse NSSS plants regardless of plant vintage and number of 
loops. The NRC staff, therefore, concludes that the proposed

[[Page 46546]]

TSTF-370, Revision 0 changes are acceptable.

5.0  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].

6.0  ENVIRONMENTAL CONSIDERATION

    The amendment changes a requirement with respect to installation or 
use of a facility component located within the restricted area as 
defined in 10 CFR part 20. The NRC staff has determined that the 
amendment involves 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 amendment involves no 
significant hazards consideration and there has been no public comment 
on such finding ( FR ). Accordingly, the amendment meets the 
eligibility criteria for categorical exclusion set forth in 10 CFR 
51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact 
statement or environmental assessment need be prepared in connection 
with the issuance of the amendment. se

7.0  CONCLUSION

    The Commission has concluded, based on 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 
amendment 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: The proposed amendment would 
change the technical specifications to revise the completion time (CT) 
from 1 hour to 24 hours for Condition B of TS 3.5.1, ``Accumulators,'' 
and its associated Bases. Condition B of TS 3.5.1 currently specifies a 
CT of one hour to restore a reactor coolant system (RCS) accumulator to 
operable status when declared inoperable due to any reason except not 
being within the required boron concentration range.
    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 basis for the accumulator limiting condition for operation 
(LCO), as discussed in Bases Section 3.5.1, is to ensure that a 
sufficient volume of borated water will be immediately forced into the 
core through each of the cold legs in the event the RCS pressure falls 
below the pressure of the accumulators, thereby providing the initial 
cooling mechanism during large RCS pipe ruptures. As described in 
Section 9.2 of the WCAP-15049, ``Risk-Informed Evaluation of an 
Extension to Accumulator Completion Times,'' evaluation, the proposed 
change will allow plant operation in a configuration outside the design 
basis for up to 24 hours, instead of 1 hour, before being required to 
begin shutdown. The impact of the increase in the accumulator CT on 
core damage frequency for all the cases evaluated in WCAP-15049 is 
within the acceptance limit of 1.0E-06/yr for a total plant core damage 
frequency (CDF) less than 1.0E-03/yr. The incremental conditional core 
damage probabilities calculated in WCAP-15049 for the accumulator CT 
increase meet the criterion of 5E-07 in Regulatory Guides (RG) 1.174 
and 1.177 for all cases except those that are based on design basis 
success criteria. As indicated in WCAP-15049, design basis accumulator 
success criteria are not considered necessary to mitigate large break 
loss-of-coolant accident (LOCA) events, and were only included in the 
WCAP-15049 evaluation as a worst case data point. In addition, WCAP-
15049 states that the NRC has indicated that an incremental conditional 
core damage frequency (ICCDP) greater than 5E-07 does not necessarily 
mean the change is unacceptable.
    The proposed technical specification change does not involve any 
hardware changes nor does it affect the probability of any event 
initiators. There will be no change to normal plant operating 
parameters, engineered safety feature (ESF) actuation setpoints, 
accident mitigation capabilities, accident analysis assumptions or 
inputs.
    Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

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

    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
the proposed change. As described in Section 9.1 of the WCAP-15049 
evaluation, the plant design will not be changed with this proposed 
technical specification CT increase. All safety systems still function 
in the same manner and there is no additional reliance on additional 
systems or procedures. The proposed accumulator CT increase has a very 
small impact on core damage frequency. The WCAP-15049 evaluation 
demonstrates that the small increase in risk due to increasing the 
accumulator allowed outage time (AOT) is within the acceptance criteria 
provided in RGs 1.174 and 1.177. No new accidents or transients can be 
introduced with the requested change and the likelihood of an accident 
or transient is not impacted.
    The malfunction of safety related equipment, assumed to be operable 
in the accident analyses, would not be caused as a result of the 
proposed technical specification change. No new failure mode has been 
created and no new equipment performance burdens are imposed.
    Therefore, this change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

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

    The proposed change does not involve a significant reduction in a 
margin of safety. There will be no change to the departure from 
nucleate boiling ratio (DNBR) correlation limit, the design DNBR 
limits, or the safety analysis DNBR limits.
    The basis for the accumulator LCO, as discussed in Bases Section 
3.5.1, is to ensure that a sufficient volume of borated water will be 
immediately forced into the core through each of the cold legs in the 
event the RCS pressure falls below the pressure of the accumulators, 
thereby providing the initial cooling mechanism during large RCS pipe 
ruptures. As described in Section 9.2 of the WCAP-15049 evaluation, the 
proposed change will allow plant operation in a configuration outside 
the design basis for up to 24 hours, instead of 1 hour, before being 
required to begin shutdown. The impact of this on plant risk was 
evaluated and found to be very small. That is,

[[Page 46547]]

increasing the time the accumulators will be unavailable to respond to 
a large LOCA event, assuming accumulators are needed to mitigate the 
design basis event, has a very small impact on plant risk. Since the 
frequency of a design basis large LOCA (a large LOCA with loss of 
offsite power) would be significantly lower than the large LOCA 
frequency of the WCAP-15049 evaluation, the impact of increasing the 
accumulator CT from 1 hour to 24 hours on plant risk due to a design 
basis large LOCA would be significantly less than the plant risk 
increase presented in the WCAP-15049 evaluation.
    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 9th day of July, 2002.

    For the Nuclear Regulatory Commission.
Robert L. Dennig,
Chief, Technical Specifications Section, Operating Reactor Improvements 
Program, Division of Regulatory Improvement Programs, Office of Nuclear 
Reactor Regulation.
[FR Doc. 02-17649 Filed 7-12-02; 8:45 am]
BILLING CODE 7590-01-P