ML061630245
| ML061630245 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/29/2006 |
| From: | Martin R E NRC/NRR/ADRO/DORL/LPLC |
| To: | Summer H L Southern Nuclear Operating Co |
| Martin R E, NRR/DORL, 415-1493 | |
| Shared Package | |
| ML061910064 | List: |
| References | |
| TAC MD0256, TAC MD0257 | |
| Download: ML061630245 (24) | |
Text
June 29, 2006 Mr. H. L. Summer, Jr.
Vice President - Farley Project
Southern Nuclear Operating
Company, Inc.
Post Office Box 1295
Birmingham, AL 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENTS REGARDING INOPERABILITY OF SNUBBERS (TAC NOS.
MD0256 AND MD0257)
Dear Mr. Summer:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 173 to Renewed Facility Operating License No. NPF-2 and Amendment No. 166 to Renewed Facility Operating
License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Unit Nos. 1 and 2. The
amendments consist of changes to the Technical Specifications (TSs) in response to your
application dated February 17, 2006.
The amendments modify TS requirements for inoperable snubbers by adding Limiting Condition for Operation 3.0.8 to be consistent with the provisions of Industry/TS Task Force (TSTF)
change TSTF-372, Revision 4. The availability of TSTF-372 for adoption by licensees was
published in the Federal Register on May 4, 2005 (70 FR 23252).
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.Sincerely,/RA/Robert E. Martin, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing
Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosures:
- 1. Amendment No. 173 to NPF-2
- 2. Amendment No. 166 to NPF-8
- 3. Safety Evaluation cc w/encl: See next page June 29, 2006 Mr. H. L. Summer, Jr.
Vice President - Farley Project
Southern Nuclear Operating
Company, Inc.
Post Office Box 1295
Birmingham, AL 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NOS. 1 AND 2 ISSUANCE OF AMENDMENTS REGARDING INOPERABILITY OF SNUBBERS (TAC NOS.
MD0256 AND MD0257)
Dear Mr. Summer:
The Nuclear Regulatory Commission has issued the enclosed Amendment No. 173 to Renewed Facility Operating License No. NPF-2 and Amendment No. 166 to Renewed Facility Operating
License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Unit Nos. 1 and 2. The
amendments consist of changes to the Technical Specifications (TSs) in response to your
application dated February 17, 2006.
The amendments modify TS requirements for inoperable snubbers by adding Limiting Condition for Operation 3.0.8 to be consistent with the provisions of Industry/TS Task Force (TSTF)
change TSTF-372, Revision 4. The availability of TSTF-372 for adoption by licensees was
published in the Federal Register on May 4, 2005 (70 FR 23252).
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.Sincerely,/RA/Robert E. Martin, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing
Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosures:
- 1. Amendment No. 173 to NPF-2
- 2. Amendment No. 166 to NPF-8
- 3. Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PublicRidsAcrsAcnwMailCenter RidsNrrPMSLingam LPL2-1 R/FG. Hill, OiS (4 hard copies)RidsNrrDorlDpr (BSingal)
RidsNrrDorlLpl2-1(EMarinos)RidsNrrDirsItsb(TKobetz)RidsOgcRp RidsNrrPMRMartin(hard copy)T. Tjader, NRRRidsRgn2MailCenter(SShaeffer)
RidsNrrLARSola(hard copy)Package No. ML061910064Amendment No. ML061630245 Tech Spec No. ML061910406OFFICENRR/LPL2-1/PMNRR/LPL2-1/PMNRR/LPL2-1/LANRR/TSBNRR/LPL2-1/BCNAMESLingamRMartinRSolaTTjaderEMarinos DATE06/29/066/29/066/30/066/26/067/5/06 OFFICIAL RECORD COPY SOUTHERN NUCLEAR OPERATING COMPANY, INC.
ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 173 Renewed License No. NPF-21.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Sout hern Nuclear Operating Company, Inc.(Southern Nuclear), dated February 17, 2006, complies with the standards and
requirements of the Atomic Energy Act of 1954, as amended (the Act), and the
Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;C.There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the
public, and (ii) that such activities will be conducted in compliance with the
Commission's regulations;D.The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.2.Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of
Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows: (2)Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 173, are hereby incorporated in the license. Southern
Nuclear shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/Evangelos C. Marinos, Chief Plant Licensing Branch II-1
Division of Operating Reactor Licensing
Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-2
and the Technical Specifications Date of Issuance: June 29, 2006 ATTACHMENT TO LICENSE AMENDMENT NO. 173 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 AND LICENSE AMENDMENT NO. 166 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace page 4 of Renewed Facility Operating License No. NPF-2 with the attached page 4.
Replace page 3 of Renewed Facility Operating License No. NPF-8 with the attached page 3.
Replace the following pages of the License and Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and
contain marginal lines indicating the areas of change.
Remove Insert License Pages License PagesLicense No. NPF-2, page 4License No. NPF-2, page 4License No. NPF-8, page 4License No. NPF-8, page 4 TSs Pages TSs Pages3.0-13.0-13.0-23.0-2 3.0-33.0-3 3.0-43.0-4 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 166 Renewed License No. NPF-81.The Nuclear Regulatory Commission (the Commission) has found that:A.The application for amendment by Sout hern Nuclear Operating Company, Inc.(Southern Nuclear), dated February 17, 2006, complies with the standards and
requirements of the Atomic Energy Act of 1954, as amended (the Act), and the
Commission's rules and regulations set forth in 10 CFR Chapter I;B.The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;C.There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the
public, and (ii) that such activities will be conducted in compliance with the
Commission's regulations;D.The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.2.Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of
Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows: (2)Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 166, are hereby incorporated in the license. Southern
Nuclear shall operate the facility in accordance with the Technical Specifications.3.This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/Evangelos C. Marinos, Chief Plant Licensing Branch II-1
Division of Operating Reactor Licensing
Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-8
and the Technical Specifications Date of Issuance: June 29, 2006 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 173 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 166 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated February 17, 2006 (Agencywi de Documents Access and Management System Accession No. ML060520069), Southern Nuclear Operating Company, Inc. (the licensee)
proposed license amendments to change the Technical Specifications (TSs) for the Joseph M.
Farley Nuclear Plant, Unit Nos. 1 and 2.
The requested changes would modify TS r equirements for inoperable snubbers by adding Limiting Condition for Operation (LCO) 3.0.8 to be consistent with the provisions of Industry/TS
Task Force (TSTF) change TSTF-372, Revision 4. The availability of TSTF-372 for adoption by
licensees was published in the Federal Register on May 4, 2005 (70 FR 23252).
On April 23, 2004, the Nuclear Energy Institute (NEI) Risk Informed Technical Specifications Task Force (RITSTF) submitted a proposed change, TSTF-372, Revision 4, to the standard
technical specifications (STS) (NUREGs 1430-1434) on behalf of the industry (TSTF-372, Revisions 1 through 3 were prior draft iterations). TSTF-372, Revision 4, is a proposal to add
an STS Limiting Condition for Operation 3.0.8, allowing a delay time for entering a supported
system TS, when the inoperability is due solely to an inoperable snubber, if risk is assessed and
managed. The postulated seismic event requiring snubbers is a low-probability occurrence and
the overall TS system safety function would still be available for the vast majority of anticipated challenges.
This proposal is one of the industry's initiatives being developed under the risk-informed TS program. These initiatives are intended to maintain or improve safety through the incorporation of risk assessment and management techniques in the TSs, while reducing unnecessary
burden and making TS requirements consistent with the Nuclear Regulatory Commission's (NRC's) other risk-informed regulatory requirements, in particular the Maintenance Rule.
The proposed change adds a new LCO, LCO 3.0.8, to the TSs. LCO 3.0.8 allows licensees to delay declaring an LCO not met for equipment, supported by snubbers unable to perform their associated support functions, when risk is assessed and managed. This new LCO 3.0.8 states:
When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for
this reason if risk is assessed and managed, and:a.The snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem
supported system or are associated with a single train or subsystem supported
system and are able to perform their associated support function within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />;
orb.The snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or
subsystem supported system and are able to perform their associated support
function within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not
met.The proposed TS change is described in Sections 1.0 and 2.0. The technical evaluation and approach used to assess its risk impact is discussed in Section 3.0. The results and insights of
the risk assessment are presented and discussed in Section 3.1. Section 3.2 summarizes the
NRC staff's conclusions from the review of the proposed TS change.
2.0 REGULATORY EVALUATION
In Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36, the Commission established its regulatory requirements related to the content of the TSs. Pursuant
to 10 CFR 50.36, TSs 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 ...." TS Section 3.0, on "LCO and SR Applicability,"
provides details or ground rules for complying with the LCOs.
Snubbers are chosen in lieu of rigid supports in areas where restricting thermal growth during normal operation would induce excessive stresses in the piping nozzles or other equipment.
Although they are classified as component standar d supports, they are not designed to provide any transmission of force during normal plant operations. However, in the presence of dynamic
transient loadings, which are induced by seismic events as well as by plant accidents and
transients, a snubber functions as a rigid support. The location and size of the snubbers are
determined by stress analysis based on different combinations of load conditions, depending on the design classification of the particular piping.
Prior to the conversion to the improved STS, TS requirements applied directly to snubbers.
These requirements included:*A requirement that snubbers be functional and in service when the supported equipment is required to be operable,*A requirement that snubber removal for testing be done only during plant shutdown,
- A requirement that snubber removal for testing be done on a one-at-a-time basis when supported equipment is required to be operable during shutdown,*A requirement to repair or replace within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> any snubbers, found to be inoperable during operation in Modes 1 through 4, to avoid declaring any supported equipment
inoperable,*A requirement that each snubber be demonstrated operable by periodic visual inspections, and*A requirement to perform functional tests on a representative sample of at least 10 percent of plant snubbers, at least once every 18 months during shutdown.
In the late 1980s, a joint initiative of the NRC and industry was undertaken to improve the STS.
This effort identified the snubbers as candidates for relocation to a licensee-controlled
document based on the fact that the TS requirements for snubbers did not meet any of the four
criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the improved STS. The NRC approved the
relocation without placing any restriction on the use of the relocated requirements. However, this relocation resulted in different interpretations between the NRC and the industry regarding
its implementation. The NRC has stated, that since snubbers are supporting safety equipment
that is in the TS, the definition of OPERABILITY must be used to immediately evaluate
equipment supported by a removed snubber and, if found inoperable, the appropriate TS
required actions must be entered. This interpretation has in practice eliminated the 72-hour
delay to enter the actions for the supported equipment that existed prior to the conversion to the
improved STS (the only exception is if the supported system has been analyzed and determined
to be OPERABLE without the snubber). The industry has argued that since the NRC approved
the relocation without placing any restriction on the use of the relocated requirements, the
licensee-controlled document requirements for snubbers should be invoked before the
supported system's TS requirements become applic able. The industry's interpretation would, in effect, restore the 72-hour delay to enter the actions for the supported equipment that existed
prior to the conversion to the improved STS. The industry's proposal would allow a time delay
for all conditions, including snubber removal for testing at power. The option to relocate the
snubbers to a licensee-controlled document, as part of the conversion to improved STS, has
resulted in non-uniform and inconsistent treatment of snubbers. On the one hand, plants that
have relocated snubbers from their TS are allowed to change the TS requirements for snubbers
under the auspices of 10 CFR 50.59, but they are not allowed a 72-hour delay before they enter
the actions for the supported equipment. On the other hand, plants that have not converted to
improved STS have retained the 72-hour delay if snubbers are found to be inoperable, but they
are not allowed to use 10 CFR 50.59 to change TS requirements for snubbers. It should also be noted that a few plants that converted to the improved STS chose not to relocate the snubbers to a licensee-controlled document and, thus, retained the 72-hour delay. In addition, it
is important to note that unlike plants that have not relocated, plants that have relocated can
perform functional tests on the snubbers at power (as long as they enter the actions for the
supported equipment) and at the same time can reduce the testing frequency (as compared to
plants that have not relocated) if it is justified by 10 CFR 50.59 assessments. Some potential
undesirable consequences of this inconsistent treatment of snubbers are:*Performance of testing during crowded time period windows when the supported system is inoperable with the potential to reduce the snubber testing to a minimum since the
snubber requirements that have been relocated from TS are controlled by the licensee,*Performance of testing during crowded windows when the supported system is inoperable with the potential to increase the unavailability of safety systems, and*Performance of testing and maintenance on snubbers affecting multiple trains of the same supported system during the 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> allotted before entering MODE 3 under
To remove the inconsistency in the treatment of snubbers among plants, the TSTF proposed a risk-informed TS change that introduces a delay time before entering the actions for the
supported equipment, when one or more snubbers are found inoperable or removed for testing, if risk is assessed and managed. Such a delay time will provide needed flexibility in the
performance of maintenance and testing during power operation and at the same time will
enhance overall plant safety by:*Avoiding unnecessary unscheduled plant shutdowns and, thus, minimizing plant transition and realignment risks,*Avoiding reduced snubber testing, and thus, increasing the availability of snubbers to perform their supporting function,*Performing most of the required testing and maintenance during the delay time when the supported system is available to mitigate most challenges and, thus, avoiding increases
in safety system unavailability, and*Providing explicit risk-informed guidance in areas where guidance currently does not exist, such as the treatment of snubbers impacting more than one redundant train of a
supported system.
3.0 TECHNICAL EVALUATION
The industry submitted TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers," in support of the proposed TS change. This submittal (Ref. 1) documents a risk-
informed analysis of the proposed TS change. Probabilistic risk assessment (PRA) results and
insights are used, in combination with deterministic and defense-in-depth arguments, to identify
and justify delay times for entering the actions for the supported equipment associated with
inoperable snubbers at nuclear power plants. This is in accordance with guidance provided in
Regulatory Guides (RGs) 1.174 and 1.177 (Refs. 2 and 3, respectively).
The risk impact associated with the proposed delay times for entering the TS actions for the supported equipment can be assessed using the same approach as for allowed completion time (CT) extensions. Therefore, the risk assessment was performed following the three-tiered
approach recommended in RG 1.177 for evaluating proposed extensions in currently allowed
CTs:*The first tier involves the assessment of the change in plant risk due to the proposed TS change. Such risk change is expressed (1) by the change in the average yearly core damage frequency (CDF) and the average yearly large early release frequency (LERF) and (2) by the incremental conditional core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP). The assessedCDF and LERF values are compared to acceptance guidelines, consistent with the Commission's Safety Goal Policy Statement as documented in RG 1.174, so that the
plant's average baseline risk is maintained within a minimal range. The assessed
ICCDP and ICLERP values are compared to acceptance guidelines provided in
RG 1.177, which aim at ensuring that the plant risk does not increase unacceptably
during the period the equipment is taken out of service.*The second tier involves the identification of potentially high-risk configurations that could exist if equipment in addition to those associated with the change were to be taken
out of service simultaneously, or other risk-significant operational factors such as
concurrent equipment testing were also involved. The objective is to ensure that
appropriate restrictions are in place to avoid any potential high-risk configurations.*The third tier involves the establishment of an overall configuration risk management program (CRMP) to ensure that potentially risk-significant configurations resulting from
maintenance and other operational activities are identified. The objective of the CRMP
is to manage configuration-specific risk by appropriate scheduling of plant activities
and/or appropriate compensatory measures.
A simplified bounding risk assessment was performed to justify the proposed addition of LCO 3.0.8 to the TS. This approach was necessitated by (1) the general nature of the proposed
TS changes (i.e., they apply to all plants and are associated with an undetermined number of
snubbers that are not able to perform their function), (2) the lack of detailed engineering
analyses that establish the relationship between earthquake level and supported system pipe
failure probability when one or more snubbers are inoperable, and (3) the lack of seismic risk
assessment models for most plants. The simplified risk assessment is based on the following
major assumptions, which the NRC staff finds acceptable, as discussed below:*The accident sequences contributing to the risk increase associated with the proposed TS changes are assumed to be initiated by a seismically-induced loss-of-offsite-power (LOOP) event with concurrent loss of all safety system trains supported by the out-of-service snubbers. In the case of snubbers associated with more than one train (or
subsystem) of the same system, it is assumed that all affected trains (or subsystems) of
the supported system are failed. This assumption was introduced to allow the
performance of a simple bounding risk assessment approach with application to all
plants. This approach was selected due to the lack of detailed plant-specific seismic risk
assessments for most plants and the lack of fragility data for piping when one or more
supporting snubbers are inoperable.
- The LOOP event is assumed to occur due to the seismically-induced failure of the ceramic insulators used in the power distribution systems. These ceramic insulators
have a high confidence (95 percent) of low probability (5 percent) of failure (HCLPF) of
about 0.1g, expressed in terms of peak ground acceleration. Thus, a magnitude 0.1g
earthquake is conservatively assumed to have 5 percent probability of causing a LOOP
initiating event. The fact that no LOOP events caused by higher magnitude earthquakes
were considered is justified because (1) the frequency of earthquakes decreases with
increasing magnitude and (2) historical data (References 4 and 5) indicate that the mean
seismic capacity of ceramic insulators (used in seismic PRAs), in terms of peak ground
acceleration, is about 0.3g, which is significantly higher than the 0.1g HCLPF value.
Therefore, the simplified analysis, even though it does not consider LOOP events
caused by earthquakes of magnitude higher than 0.1g, bounds a detailed analysis which
would use mean seismic failure probabilities (fragilities) for the ceramic insulators.*Analytical and experimental results obtained in the mid-eighties as part of the industry's "Snubber Reduction Program" (References 4 and
- 6) indicated that piping systems have large margins against seismic stress. The assumption that a magnitude 0.1g
earthquake would cause the failure of all safety syst em trains supported by the out-of-service snubbers is very conservative because safety piping systems could withstand
much higher seismic stresses even when one or more supporting snubbers are out of
service. The actual piping failure probability is a function of the stress allowable and the
number of snubbers removed for maintenance or testing. Since the licensee-controlled
testing is done on only a small (about 10 percent) representative sample of the total
snubber population, typically only a few snubber s supporting a given safety system are out for testing at a time. Furthermore, since the testing of snubbers is a planned activity, licensees have flexibility in selecting a sample set of snubbers for testing from a much
larger population by conducting configuration-specific engineering and/or risk
assessments. Such a selection of snubbers for testing provides confidence that the
supported systems would perform their functions in the presence of a design-basis
earthquake and other dynamic loads and, in any case, the risk impact of the activity will remain within the limits of acceptability defined in risk-informed RGs 1.174 and 1.177.*The analysis assumes that one train (or subsystem) of all safety systems is unavailable during snubber testing or maintenance (an entire system is assumed unavailable if a
removed snubber is associated with both trains of a two-train system). This is a very
conservative assumption for the case of corrective maintenance since it is unlikely that a
visual inspection will reveal that one or more snubbers across all supported systems are
inoperable. This assumption is also conservative for the case of the licensee-controlled
testing of snubbers since such testing is performed only on a small representative sample.*In general, no credit is taken for recovery actions and alternative means of performing a function, such as the function performed by a system assumed failed (e.g., when LCO 3.0.8b applies). However, most plants have reliable alternative means of
performing certain critical functions. For example, feed and bleed (F&B) can be used to
remove heat in most pressurized water reactors (PWRs) when auxiliary feedwater (AFW), the most important system in mitigating LOOP accidents, is unavailable.
Similarly, if high pressure makeup (e.g., reactor core isolation cooling) and heat removal
capability (e.g., suppression pool cooling) are unavailable in boiling water reactors (BWRs), reactor depressurization in conjunction with low pressure makeup (e.g., low pressure coolant injection) and heat removal capability (e.g., shutdown cooling) can be
used to cool the core. A 10 percent failure probability for recovery actions to provide
core cooling using alternative means is assumed for Diablo Canyon, the only West
Coast PWR plant with F&B capability, when a snubber impacting more than one train of
the AFW system (i.e., when LCO 3.0.8b is applicable) is out of service. This failure
probability value is significantly higher than the value of 2.2E-2 used in Diablo Canyon's
PRA. Furthermore, Diablo Canyon has analyzed the impact of a single limiting snubber
failure, and concluded that no single snubber failure would impact two trains of AFW.
No credit for recovery actions to provide core cooling using alternative means is
necessary for West Coast PWR plants with no F&B capability because it has been
determined that there is no single snubber whose non-functionality would disable two
trains of AFW in a seismic event of magnitude up to the plant's safe shutdown
earthquake (SSE). It should be noted that a similar credit could have been applied to
most Central and Eastern U.S. plants but this was not necessary to demonstrate the low
risk impact of the proposed TS change due to the lower earthquake frequencies at
Central and Eastern U.S. plants as compared to West Coast plants.*The earthquake frequency at the 0.1g level was assumed to be 1E-3/year for Central and Eastern U.S. plants and 1E-1/year for West Coast plants. Each of these two values
envelop the range of earthquake frequency values at the 0.1g level, for Eastern U.S. and
West Coast sites, respectively (References 5 and 7).*The risk impact associated with non-LOOP accident sequences (e.g., seismically initiated loss-of-coolant-accident (LOCA) or anticipated-transient-without-scram (ATWS)
sequences) was not assessed. However, this risk impact is small compared to the risk
impact associated with the LOOP accident sequences modeled in the simplified
bounding risk assessment. Non-LOOP accident sequences, due to the ruggedness of
nuclear power plant designs, require seismically-induced failures that occur at
earthquake levels above 0.3g. Thus, the frequency of earthquakes initiating non-LOOP
accident sequences is much smaller than the frequency of seismically-initiated LOOP
events. Furthermore, because of the conservative assumption made for LOOP
sequences that a 0.1g level earthquake would fail all piping associated with inoperable
snubbers, non-LOOP sequences would not include any more failures associated with
inoperable snubbers than LOOP sequences. Therefore, the risk impact of inoperable
snubbers associated with non-LOOP accident sequences is small compared to the risk
impact associated with the LOOP accident sequences modeled in the simplified
bounding risk assessment.*The risk impact of dynamic loadings other than seismic loads is not assessed. These shock-type loads include thrust loads, blowdown loads, water hammer loads, steam
hammer loads, LOCA loads and pipe rupture loads. However, there are some important
distinctions between non-seismic (shock-type) loads and seismic loads which indicate
that, in general, the risk impact of the out-of-service snubbers is smaller for non-seismic
loads than for seismic loads. First, while a seismic load affects the entire plant, the
impact of a non-seismic load is localized to a certain system or area of the plant.
Second, although non-seismic shock loads may be higher in total force and the impact
could be as much or more than seismic loads, generally they are of much shorter
duration than seismic loads. Third, the impact of non-seismic loads is more plant-specific; thus, harder to analyze generically, than for seismic loads. For these reasons, licensees will be required to confirm every time LCO 3.0.8 is used, that at least
one train of each system that is supported by the inoperable snubber(s) would remain
capable of performing their required safety or support functions for postulated design
loads other than seismic loads.
3.1 Risk Assessment Results and Insights
The results and insights from the implementation of the three-tiered approach of RG 1.177 to support the proposed addition of LCO 3.0.8 to the TS are summarized and evaluated in the
following sections 3.1.1 to 3.1.3.
3.1.1 Risk Impact
The bounding risk assessment approach, discussed in section 3.0, was implemented generically for all U.S. operating nuclear power plants. Risk assessments were performed for
two categories of plants, Central and East Coast plants and West Coast plants, based on
historical seismic hazard curves (earthquake frequencies and associated magnitudes). The first
category, Central and East Coast plants, includes the vast majority of the U.S. nuclear power
plant population (Reference 7). For each category of plants, two risk assessments were
performed:*The first risk assessment applies to cases where all inoperable snubbers are associated with only one train (or subsystem) of the impacted safety systems. It was conservatively
assumed that a single train (or subsystem) of each safety system is unavailable. It was
also assumed that the probability of non-mitigation using the unaffected redundant trains (or subsystems) is 2 percent. This is a cons ervative value given that for core damage to occur under those conditions, two or more failures are required.*The second risk assessment applies to the case where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety
systems. It was assumed in this bounding analysis that all safety systems are unavailable to mitigate the accident, except for West Coast PWR plants. Credit for using
F&B to provide core cooling is taken for plants having F&B capability (e.g., Diablo
Canyon) when a snubber impacting more than one train of the AFW system is
inoperable. Credit for one AFW train to provide core cooling is taken for West Coast
PWR plants with no F&B capability (e.g., San Onofre) because it has been determined
that there is no single snubber whose non-functionality would disable two trains of AFW
in a seismic event of magnitude up to the plant's SSE.
The results of the performed risk assessments, in terms of core damage and large early release risk impacts, are summarized in Table 1. The first row lists the conditional risk increase, in
terms of core damage frequency (CDF), R CDF , caused by the out-of-service snubbers (as assumed in the bounding analysis). The second and third rows list the incremental conditional
core damage probability (ICCDP) and the incremental conditional large early release probability (ICLERP) values, respectively. The ICCDP for the case where all inoperable snubbers are
associated with only one train (or subsystem) of the supported safety systems, was obtained by
multiplying the corresponding R CDF value by the time fraction of the proposed 72-hour delay to enter the actions for the supported equipment. The ICCDP for the case where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety system, was obtained by multiplying the corresponding R CDF value by the time fraction of the proposed 12- hour delay to enter the actions for the supported equipment. The ICLERP values
were obtained by multiplying the corresponding ICCDP values by 0.1 (i.e., by assuming that the
ICLERP value is an order of magnitude less than the ICCDP). This assumption is conservative
since containment bypass scenarios, such as steam generator tube rupture accidents and interfacing system loss-of-coolant accidents, would not be uniquely affected by the out-of-
service snubbers. Finally, the fourth and fifth rows list the assessed CDF and LERF values, respectively. These values were obtained by dividing the corresponding ICCDP and ICLERP
values by 1.5 (i.e., by assuming that the snubbers are tested every 18 months, as was the case
before the snubbers were relocated to a licensee-controlled document). This assumption is
reasonable because (1) it is not expected that licensees would test the snubbers more often
than what used to be required by the TS, and (2) testing of snubbers is associated with higher
risk impact than the average corrective maintenance of snubbers found inoperable by visual
inspection (testing is expected to involve significantly more snubbers out of service than
corrective maintenance). The assessed CDF and LERF values are compared to acceptance guidelines consistent with the Commission's Safety Goal Policy Statement as documented in
RG 1.174, so that the plant's average baseline risk is maintained within a minimal range. This
comparison indicates that the addition of LCO 3.0.8 to the existing TS would have an
insignificant risk impact.
Table 1.--Bounding Risk Assessment Results for Snubbers Impacting a Single Train and Multiple Trains of a Supported System
Central and east West coast coast plants Plants
Single MultipleSingle Multiple
train train train train
R CDF/yr.................... 1E-6 5E-6 1E-4 5E-4 ICCDP............................. 8E-9 7E-9 8E-7 7E-7 ICLERP............................ 8E-10 7E-10 8E-8 7E-8CDF/yr..................... 5E-9 5E-9 5E-7 5E-7LERF/yr.................... 5E-10 5E-10 5E-8 5E-8 --
The assessed CDF and LERF values meet the acceptance criteria of 1E-6/year and 1E-7/year, respectively, based on guidance provided in RG 1.174. This conclusion is true
without taking any credit for the removal of potential undesirable consequences associated with
the current inconsistent treatment of snubbers (e.g., reduced snubber testing frequency, increased safety system unavailability and treat ment of snubbers impacting multiple trains) discussed in Section 1 above, and given the bounding nature of the risk assessment.
The assessed ICCDP and ICLERP values are compared to acceptance guidelines provided in RG 1.177, which aim at ensuring that the plant risk does not increase unacceptably during the
period the equipment is taken out of service. This comparison indicates that the addition of
LCO 3.0.8 to the existing TS meets the RG 1.177 numerical guidelines of 5E-7 for ICCDP and
5E-8 for ICLERP. The small deviations shown for West Coast plants are acceptable because of
the bounding nature of the risk assessments, as discussed in section 2.
The risk assessment results of Table 1 are also compared to guidance provided in the revised section 11 of NUMARC 93-01, Revision 2 (Reference 8), endorsed by RG 1.182 (Reference 9),
for implementing the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65.
Such guidance is summarized in Table 2. Guidance regarding the acceptability of conditional
risk increase in terms of CDF (i.e., R CDF) for a planned configuration is provided. This guidance states that a specific configuration that is associated with a CDF higher than
1E-3/year should not be entered voluntarily. Since the assessed conditional risk increase,R CDF , is significantly less than 1E-3/year, plant configurations including out-of-service snubbers and other equipment may be entered voluntarily if supported by the results of the risk
assessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TS. Table 2.--Guidance for Implementing 10 CFR 50.65(a)(4)
R CDF Guidance
Greater than 1E-3/year................. Configuration should not normally be entered voluntarily.
ICCDP Guidance ICLERP Greater than 1E-5.....Configuration should not normally be entered voluntarily. Great er than 1E-6. 1E-6 to 1E-5..............Assess non- quantifiable factors.
Establish risk management actions..
Less than1E-6..........Normal work controls..
Less than 1E-7.
Guidance regarding the acceptability of ICCDP and ICLERP values for a specific planned configuration and the establishment of risk m anagement actions is also provided in NUMARC 93-01. This guidance, as shown in Table 2, states that a specific plant configuration that is
associated with ICCDP and ICLERP values below 1E-6 and 1E-7, respectively, is considered to
require "normal work controls." Table 1 shows that for the majority of plants (i.e., for all plants in
the Central and East Coast category) the conservatively assessed ICCDP and ICLERP values
are over an order of magnitude less than what is recommended as the threshold for the "normal
work controls" region. For West Coast plants, the conservatively assessed ICCDP and ICLERP
values are still within the "normal work controls" region. Thus, the risk contribution from out of
service snubbers is within the normal range of maintenance activities carried out at a plant.
Therefore, plant configurations involving out of service snubbers and other equipment may be
entered voluntarily if supported by the result s of the risk assessment required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TS. However, this simplified bounding analysis indicates
that for West Coast plants the provisions of LCO 3.0.8 must be used cautiously and in
conjunction with appropriate management actions, especially when equipment other than
snubbers is also inoperable, based on the results of configuration-specific risk assessments
required by 10 CFR 50.65(a)(4), by LCO 3.0.8, or by other TS.
The staff finds that the risk assessment results support the proposed addition of LCO 3.0.8 to the TS. The risk increases associated with this TS change will be insignificant based on
guidance provided in RGs 1.174 and 1.177 and within the range of risks associated with normal
maintenance activities. In addition, LCO 3.0.8 will remove potential undesirable consequences
stemming from the current inconsistent treatment of snubbers in the TS, such as reduced
frequency of snubber testing, increased safety system unavailability and the treatment of
snubbers impacting multiple trains.
3.1.2 Identification
of High-Risk Configurations
The second tier of the three-tiered approach recommended in RG 1.177 involves the identification of potentially high-risk configurations that could exist if equipment, in addition to
that associated with the TS change, were to be taken out of service simultaneously. Insights from the risk assessments, in conjunction with important assumptions made in the analysis and defense-in-depth considerations, were used to identify such configurations. To avoid these
potentially high-risk configurations, specific restrictions to the implementation of the proposed
TS changes were identified.
For cases where all inoperable snubbers are associated with only one train (or subsystem) of the impacted systems (i.e., when LCO 3.0.8a applies), it was assumed in the analysis that there will be unaffected redundant trains (or subsystems) available to mitigate the seismically initiated
LOOP accident sequences. This assumption implies that there will be at least one success
path available when LCO 3.0.8a applies. Therefore, potentially high-risk configurations can be
avoided by ensuring that such a success path exists when LCO 3.0.8a applies. Based on a
review of the accident sequences that contribute to the risk increase associated with LCO 3.0.8a, as modeled by the simplified bounding analysis (i.e., accident sequences initiated by a
seismically-induced LOOP event with concurrent loss of all safety system trains supported by the out of service snubbers), the following restrictions were identified to prevent potentially
high-risk configurations:*For PWR plants, at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable
snubber(s), must be available when LCO 3.0.8a is used*For BWR plants, one of the following two means of heat removal must be available when LCO 3.0.8a is used:
At least one high pressure makeup path (e.g., using high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC) or equivalent) and heat
removal capability (e.g., suppression pool cooling), including a minimum set of
supporting equipment required for success, not associated with the inoperable
snubber(s), or At least one low pressure makeup path (e.g., low pressure coolant injection (LPCI) or containment spray (CS)) and heat removal capability (e.g., suppression
pool cooling or shutdown cooling), including a minimum set of supporting
equipment required for success, not associated with the inoperable snubber(s).
For cases where one or more of the inoperable snubbers are associated with multiple trains (or subsystems) of the same safety system (i.e., when LCO 3.0.8b applies), it was assumed in the
bounding analysis that all safety systems are unav ailable to mitigate the accident, except for West Coast plants. Credit for using F&B to provide core cooling is taken for plants having F&B
capability (e.g., Diablo Canyon) when a snubber impacting more than one train of the AFW
system is inoperable. Credit for one AFW train to provide core cooling is taken for West Coast
PWR plants with no F&B capability (e.g., San Onofre) because it has been determined that
there is no single snubber whose non-functionality would disable more than one train of AFW in
a seismic event of magnitude up to the plant's SSE. Based on a review of the accident
sequences that contribute to the risk increase associated with LCO 3.0.8b (as modeled by the
simplified bounding analysis) and defense-in-depth considerations, the following restrictions
were identified to prevent potentially high-risk configurations:*LCO 3.0.8b cannot be used at West Coast PWR plants with no F&B capability when a snubber whose non-functionality would disable more than one train of AFW in a seismic event of magnitude up to the plant's SSE is inoperable (it should be noted, however, that
based on information provided by the industry, there is no plant that falls in this
category).*When LCO 3.0.8b is used at PWR plants, at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the
inoperable snubber(s), or some alternative means of core cooling (e.g., F&B, fire water
system or "aggressive secondary cooldown" using the steam generators) must be
available.*When LCO 3.0.8b is used at BWR plants, it must be verified that at least one success path exists, using equipment not associated with the inoperable snubber(s), to provide
makeup and core cooling needed to mitigate LOOP accident sequences.
3.1.3 Configuration
Risk Management
The third tier of the three-tiered approach recommended in RG 1.177 involves the establishment of an overall configuration risk management progr am (CRMP) to ensure that potentially risk-significant configurations resulting from maintenance and other operational activities are
identified. The objective of the CRMP is to manage configuration-specific risk by appropriate
scheduling of plant activities and/or appropriate compensatory measures. This objective is met
by licensee programs to comply with the requirements of paragraph (a)(4) of the Maintenance
Rule (10 CFR 50.65) to assess and manage risk resulting from maintenance activities, and by
the TS requiring risk assessments and management using (a)(4) processes if no maintenance
is in progress. These programs can support licensee decision making regarding the appropriate
actions to manage risk whenever a risk- informed TS is entered. Since the 10 CFR 50.65(a)(4)
guidance, the revised (May 2000) Section 11 of NUMARC 93-01, does not currently address
seismic risk, licensees adopting this change must ensure that the proposed LCO 3.0.8 is
considered with respect to other plant maintenance activities and integrated into the existing 10
CFR 50.65(a)(4) process whether the process is invoked by a TS or (a)(4) itself.
3.2 Summary
and Conclusions
The option to relocate the snubbers to a licensee-controlled document, as part of the conversion to Improved STS, has resulted in non-uniform and inconsistent treatment of snubbers. Some
potential undesirable consequences of this inconsistent treatment of snubbers are:*Performance of testing during crowded windows when the supported system is inoperable with the potential to reduce the snubber testing to a minimum since the
relocated snubber requirements are controlled by the licensee.*Performance of testing during crowded windows when the supported system is inoperable with the potential to increase the unavailability of safety systems.*Performance of testing and maintenance on snubbers affecting multiple trains of the same supported system during the 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> allotted before entering MODE 3 under
To remove the inconsistency among plants in the treatment of snubbers, licensees are proposing a risk-informed TS change which introduces a delay time before entering the actions
for the supported equipment when one or more snubbers are found inoperable or removed for
testing. Such a delay time will provide needed flexibility in the performance of maintenance and
testing during power operation and at the same time will enhance overall plant safety by (1)
avoiding unnecessary unscheduled plant shutdowns, thus, minimizing plant transition and
realignment risks; (2) avoiding reduced snubber testing, thus, increasing the availability of
snubbers to perform their supporting function; (3) performing most of the required testing and
maintenance during the delay time when the supported system is available to mitigate most
challenges, thus, avoiding increases in safety system unavailability; and (4) providing explicit risk-informed guidance in areas where guidance curr ently does not exist, such as the treatment of snubbers impacting more than one redundant train of a supported system.
The risk impact of the proposed TS changes was assessed following the three-tiered approach recommended in RG 1.177. A simplified bounding risk assessment was performed to justify the
proposed TS changes. This bounding assessment assumes that the risk increase associated
with the proposed addition of LCO 3.0.8 to the TS is associated with accident sequences
initiated by a seismically-induced LOOP event with concurrent loss of all safety system trains supported by the out-of-service snubbers. In the case of snubbers associated with more than
one train, it is assumed that all affected trains of the supported system are failed. This
assumption was introduced to allow the performance of a simple bounding risk assessment
approach with application to all plants and was selected due to the lack of detailed plant-specific
seismic risk assessments for most plants and the lack of fragility data for piping when one or
more supporting snubbers are inoperable. The impact from the addition of the proposed LCO 3.0.8 to the TS on defense-in-depth was also evaluated in conjunction with the risk assessment
results.Based on this integrated evaluation, the staff concludes that the proposed addition of LCO 3.0.8 to the TS would lead to insignificant risk increases, if any. Indeed, this conclusion is true
without taking any credit for the removal of potential undesirable consequences associated with
the current inconsistent treatment of snubbers, such as the effects of avoiding a potential
reduction in the snubber testing frequency and increased safety system unavailability.
Consistent with the NRC staff's approval and inherent in the implementation of TSTF-372, licensees interested in implementing LCO 3.0.8 must, as applicable, operate in accordance with
the following stipulations:1.Appropriate plant procedures and administrat ive controls will be used to implement the following Tier 2 Restrictions.(a)At least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable
snubber(s), must be available when LCO 3.0.8a is used at PWR plants.(b)At least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable
snubber(s), or some alternative means of core cooling (e.g., F&B, fire
water system or "aggressive secondary cooldown" using the steam
generators) must be available when LCO 3.0.8b is used at PWR plants. (c)LCO 3.0.8b cannot be used by West Coast PWR plants with no F&B capability when a snubber, whose non-functionality would disable more
than one train of AFW in a seismic event of magnitude up to the plant's
SSE, is inoperable.(d)BWR plants must verify, every time the provisions of LCO 3.0.8 are used, that at least one success path, involving equipment not associated with
the inoperable snubber(s), exists to provide makeup and core cooling
needed to mitigate LOOP accident sequences.(e)Every time the provisions of LCO 3.0.8 are used licensees will be required to confirm that at least one train (or subsystem) of systems
supported by the inoperable snubbers would remain capable of
performing their required safety or support functions for postulated design
loads other than seismic loads. LCO 3.0.8 does not apply to non-seismic
snubbers. In addition, a record of the design function of the inoperable
snubber (i.e., seismic vs. non-seismic), implementation of any applicable
Tier 2 restrictions, and the associated plant configuration shall be
available on a recoverable basis for NRC staff inspection.2.Should licensees implement the provisions of LCO 3.0.8 for snubbers, which include delay times to enter the actions for the supported equipment when one or more
snubbers are out of service for maintenance or testing, it must be done in accordance
with an overall CRMP to ensure that potentially risk-significant configurations resulting
from maintenance and other operational activities are identified and avoided, as
discussed in the proposed TS Bases. This objective is met by licensee programs to
comply with the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65, to assess and manage risk resulting from maintenance activities or when
this process is invoked by LCO 3.0.8 or other TS. These programs can support licensee
decision making regarding the appropriate actions to manage risk whenever a risk-
informed TS is entered. Since the 10 CFR 50.65(a)(4) guidance, the revised (May 2000)
Section 11 of NUMARC 93-01, does not currently address seismic risk, licensees
adopting this change must ensure that the proposed LCO 3.0.8 is considered in
conjunction with other plant maintenance activities and integrated into the existing
10 CFR 50.65(a)(4) process. In the absence of a detailed seismic PRA, a bounding risk
assessment, such as utilized in this Safety Evaluation, shall be followed.
The licensee included in its application the revised TS Bases to be implemented with the TS change. The NRC staff finds that the TS Bases Control Program is the appropriate process for
updating the affected TS Bases pages and has, therefore, not included the affected Bases page
with this amendment.
4.0 STATE CONSULTATION
In accordance with the Nuclear Regulatory Commission's regulations, the Alabama State official was notified of the proposed issuance of the amendments. The State official had no comments.
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. 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 Nuclear
Regulatory Commission has previously issued a proposed finding that the amendments involve no-significant-hazards consideration, and there has been no public comment on the finding
(71 FR 23960; April 25, 2006). Accordingly, the amendments meet the eligibility criteria for
categorical exclusion set forth in 10 CFR 51.22©)(9). Pursuant to 10 CFR 51.22(b) no
environmental impact statement or environm ental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Nuclear Regulatory Commission has concluded, on the basis of 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.
7.0 REFERENCES
1.TSTF-372, Revision 4, "Addition of LCO 3.0.8, Inoperability of Snubbers," April 23, 2004.
2.Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decision Making on Plant Specific Changes to the Licensing Basis," USNRC,
August 1998.3.Regulatory Guide 1.177, "An Approach for Plant Specific Risk- Informed Decision Making: Technical Specifications," USNRC, August 1998.4.Budnitz, R. J. et al., "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants," NUREG/CR-4334, Lawrence Livermore National Laboratory, July 1985.5.Advanced Light Water Reactor Utility Requirements Document, Volume 2, ALWR Evolutionary Plant, PRA Key Assumptions and Gr oundrules, Electric Power Research Institute, August 1990.6.Bier V. M. et al., "Development and Application of a Comprehensive Framework for Assessing Alternative Approaches to Snubber Reduction," International Topical
Conference on Probabilistic Safety Asse ssment and Risk Management PSA '87, Swiss Federal Institute of Technology, Zurich, August 30-September 4, 1987.7.NUREG-1488, "Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains," April 1994.8.NEI, Revised Section 11 of Revision 2 of NUMARC 93-01, May 2000.
9.Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants," May 2000.
Principal Contributor: T. R. Tjader
Date: June 29, 2006 Joseph M. Farley Nuclear Plant, Units 1 & 2 cc:
Mr. J. R. Johnson General Manager
Southern Nuclear Operating Company, Inc.
P.O. Box 470
Ashford, AL 36312 Mr. B. D. McKinney, Licensing Manager Southern Nuclear Operating Company, Inc.
P.O. Box 1295
Birmingham, AL 35201-1295 Mr. M. Stanford Blanton Balch and Bingham Law Firm
P.O. Box 306
1710 Sixth Avenue North
Birmingham, AL 35201Mr. J. Gasser Executive Vice President
Southern Nuclear Operating Company, Inc.
P.O. Box 1295
Birmingham, AL 35201 State Health Officer Alabama Department of Public Health
434 Monroe St.
Montgomery, AL 36130-1701 Chairman Houston County Commission
P.O. Box 6406
Dothan, AL 36302 Resident Inspector U.S. Nuclear Regulatory Commission
7388 N. State Highway 95
Columbia, AL 36319