ML20096D732
| ML20096D732 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 01/11/1996 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20096D714 | List: |
| References | |
| NUDOCS 9601190247 | |
| Download: ML20096D732 (80) | |
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4 CALVERT CLIFFS NUCLEAR POWER PLANT-- -
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~ INTEGRATED PLANT ASSESSMENT l
l METHODOLOGY' i
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Baltimore Gas and Electric Company January 11,1996 Revision 1 9601190247 960111 PDR ADOCK 05000317 P PDR
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE OF CONTENTS
-- . 4mmmmmmmmmmmggwymn-ggSEC110Nmnsassegnamnamummegamaah;4msessmanna nden FAGFs 1.0 I NTRO D UCTIO N.............. ........................... .... ................................................ ........ . 1 1.1 Background.................................................................................................1 l 1.2 Methodology Summary. .... ................ .. ... ...... ............................................... ......... 2 ,
i 2.0 INTEGRATED PLANT ASSESSMENT METHODOLOGY .............. ............................ 4 BASES AND OVERVIEW 2.1 Definitions............................................................................................................4 4 i
2.2 Assumptions and Initial Conditions ................. ............... .... ............................................ 10 2.3 Integrated Plant Assessment Methodology Overview.......... ...................... ................. ...11 ;
t 3.0 S YSTEM LE VE L SCOPI N G . . .................................... .... ................ .......... ............ 14 ,
3.1 Identification of Systems and Structures ....................... ............ ... ................ ... ............ 14 l
'1 3.2 Define Concept ual Boundaries ... ....... .... .............. ................ ........................................ 14 3.3 Sc reening Tools Preparation ................. ................ .......................................................... 16 3.4 Systems and St ruct u res Sc oping .... ....................... ........................ .................................... 21 3.5 Results..............................................................................................................................22 4.0 COM PO NENT L EVE L SCO PI NG............ .... ................................................................. 2 7 4.1 Component Level Scoping for Systems ................. .. .. ...................... .............................. 27 r
4.2 Component Level Scoping for Structures............. ...................... ........................................ 34 4.3 Commodity Evaluations that Include Scoping Sections ....... ................................................ 38 4.4 Results.....................................................................................................................................38 5.0 PRE.E VA L UATI O N ..... .. ............................. ...... .... ......................... ...... ....... .. ........ .. ...... ..... 3 9 5.1 Categorize Intended System Functions as Active or Passive... ............................................ 41 5.2 Determine Whether Components Are Long. Lived or Short. Lived ....................................... 42 5.3 Assignment of System Components to Commodity Evaluations ............................................ 42 5.4 How the Pre-Evaluation Process Applies to Structures ........................... ............................ 43 5.5 Pre. Evaluation Results and Documentation ........................................................................... 44 i Resision I
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE OF CONTENTS i y..
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6.0 A G ING MANA G EM ENT RE VI EW .............. ........... .. ................................................. 45 6.1 Justification that Effects of Aging are Being Managed Without Specifically......................... 47 I 1
Evaluating Age-Related Degradation Mechanisms 6.2 Performing an Aging Management Review by Evaluating Aging Mechanisms ..................... 51 6.3 Methods to Manage the EfTects of A ging. ... .. .. ......... ........ ......................................... 54 6.4 Plant Program Documentation .............. ............. .................................................. ......... 64 6.5 Iategrated Plant Assessment Summary ..................................... ........................................... 64 7.0 COMMODITY APPROACHES TO AGING MANAGEMENT REVIEW.................... 65 7.1 Commodity Evaluations Equivalent to the Aging Management Review Step. ... ................ 65 4
7.2 Commodity Evaluations Which Cover All Scoping and Integrated ...................................... 67 l l
Plant Assessment Tssks 7.3 Commodity Evaluation Results and Documentation ......................................... ...... ......... 72 1
8.0 TIME-LIMITED AGING ANALYSIS REVI EW ................................. .......................... 73 8.1 Identify Analyses to be included in the Review...................................................................... 75 8.2 Review of Potential Time-Limited Aging Analyses ................................................... ........... 75 8.3 Disposition of Time-Limited Aging Analyses Which are Subject to License . ...................... 76 Renewal Review i 1
8.4 Time-Limited Aging Analyses Results and Documentation.................................................... 77 l
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CALVERT CLIFFS NUCLEAR POWER PLANT !
INTEGRATED PLANT ASSESSMENT METHODOLOGY l 1
1.0 INTRODUCTION
I ne purpose of this Methodology is to document the plant-specific process used for conducting the ;
Integrated Plant Assessment (IPA) for Agmg and the Time-Limited Aging Analysis (TLAA) i Renew for the Calvent Cliffs Nuclear Power Plant (CCNPP) in order to produce the information I specified in the License Renewal (LR) Rule Secten 54.21 (Contents of Application - Technical i information).
During the performance of the IPA process described in this methodology, all plant structurcs and components (SCs) which are subject to aging management review (AMR) are identified. For the identified SCs,justificaten is developed that demonstrates that the effects of aging on the intended functions of these SCs are adequately managed (see definitions).
In addition to the IPA process, this methodology describes the TLAA rewew task which complements the IPA. His review identifies TLAAs in the CCNPP Current Licensing Basis (CLB) which meet the specific criteria defined in the LR Rule. It also identifies exemptons still in effect which are based on a TLAA For each of the identified analyses, the renew task provides justification that the analysis is valid for the period of extended operations, provides a means for updatmg the analysis so that it will be valid for the period of extended operation or documents that the aging issue covered by the TLAA is adequately managed.
De IPA process for CCNPP has been d;vided into several distinct tasks. Each of these tasks, as well as the TLAA review task, will be discussed in subsequent sections of this methodology. He purpose of this section of the methodology is to provide general background information regarding the Baltimore Gas & Electric Company (BGE) Life Cycle Management (LCM) Program and to briefly introduce the topics presented in the following sections ofIPA Methodology, 1.1 Backaround i
Baltimore Gas and Electric Company has embarked on a comprehensive, long-term LCM Program for CCNPP, Units I and 2. The LCM Program directly supports BGE's Corporate Operational Strategy of preserving the long-term operation of CCNPP. In this capacity, the LCM Program governs the major evaluations to determine the reconfiguration of systems and structures (SSs) to i improve reliability, increase availability, reduce operations and maintenance cost, provide recc..ws.dations to the capital improvement plan for the site, prepare License Renewal Applications (LRAs) for both Units, as well as contingency plans for decc.wissioning. The LCM Program also coordmates site activities regarding reactor vessel issues (including pressurized thermal shock [ PTS]) and provides input to corporate Generation Planning and Accounting offices for strategic generation planning. Additional services governed by the LCM Program include project management of the 24-month cycle project, the Instrumentation and Controls Upgrade Project and Power Uprate Feasibility Studies.
Because of its role in preserving the long-term operation of CCNPP, the LCM Program has integrated specific design, engineering, operations, and maintenance activities to focus attention on material conditions and aging management. The LCM Program involves all five Nuclear Energy Division departments and a number of other BGE divisions.
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ATTACHMENT (1) 4 CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY i
1.2 Methodolony Summarv
'The BGE IPA methodology is based on the prenuse that, with the possible exception of the !
detnmental effects of aging on the functionality of certam systems, structures and cc.i.pc.sts (SSCs) in the penod of extended operaton, the plant's CLB ensures an adequate level of safety for !'
contmuod plant operatens Figure 1-1 illustrates the flow path of the BGE IPA, as implemented at CCNPP 'The relationship bemw. the IPA and the TLAA revmw is shown in Figure 1-2.
'Ihe Methodology is divided into eight sections, each of which describes a task. The contents of h*iaa 2.0 through 8.0 are summarized below.
Section 2.0, IPA M4=lala-v Bases and Ddaitiana. contams the following information: ;
> Defimtens ofimportant terms and acronyms that are integral to the IPA methodology. .
> Ass'.anptons and initial conditions on which the IPA methodology is based
> Source ?-- - =a which were used to develop the methodology.
i Section 3.0, System level Scooina. describes the scoping task where SSs that perform specific !
functens (described in Section 54.4 of the LR Rule) are identified as the initial scope of ,
equipment, which will be the subject of the IPA for aging. '
Secten 4.0, Component Level Sconina. describes how the SS intended functions are identified in more detail, and how individual components of the SS are evaluated to determine which components contribute to the intended functens 'Ihis section provules two parallel processes for w.iyerst level scoping, one used for system co.i.rersts and the other for structural components.
.lection 5.0, Pre-Ev=batiaa_ describes the task to determine which components are " subject to AMR" in the subsequent task of the IPA. ;
1 Section 6.0, AMR. describes how the determination is made that existing, modified or new l programs or activities for those SCs subject to AMR adequately manage the effects of aging.
Section 7, Commodity Evaluations. describes alternate IPA process task used at CCNPP for specific commodity groups.
Section 8.0, TLAA Review. describes the process for selecting TLAAs which need to be addressed for LR and methods for addressing the identified anal >ws.
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CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY IPA Flow Diagram
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,ig.,e t.t IPA Process and TLAA i
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37 l SC not subject to AMR. l Figure 1-2 1
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CALVERT CLIFFS NUCLEAR POWER PLANT 4 INTEGRATED PLANT ASSESSMENT METHODOLOGY 2.0 IPA METHODOLOGY BASES AND OVERVIEW nis section defines the terms and acronyms (Section 2.1) that are used throughout the methodology. Section 2.2 presents the assumptions and initial conditions on which the IPA methodology is based. Finally, Section 2.3 presents an oversiew of the methodology tasks.
2.1 Definitions .
There are a number of terms and acronyms that are used throughout this methodology. These terms are defined below and the meaning of acronyms is prosided in Table 2-1. Many of the l following definitions, identified by *, are taken from the LR Rule, Sections 54.3, 54.4, 54.21, and 54.31 or from the Statements of Consideration (SOC) to the Rule. He specific rule section which is the source of the definition is noted parenthetically for definitions marked with an asterisk. !
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- 1. Adequately Managed - The effects of aging are adequately managed for a group of SCs if their intended passive functions will be maintained consistent with the CLB during the period of extended operations.
- 2. Age-Related Degradation - A change in SSC performance or physical or chemical properties resulting in whole or part from one or more aging mechanisms. Examples of l this type of change include changes in dimension, ductility, fatigue resistance, fracture toughness, mechanical strength, polymerization, siscosity, and dielectric strength.
- 3. Aging Mechanisms - The physical or chemical processes that result in degradation. These mechanisms include, but are not limited to, fatigue, erosion, corrosion, crosion/ corrosion, !
wear, thermal embrittlement, radiation embrittlement, microbiologically-induced effects, creep, and shrinkage.
- 4. Critical Safety Function (CSF) - A condition or action that prevents core damage or minimizes radiation release to the public. A CSF may be fulfilled through automatic or manual actuation of a system or systems, from passive 1 system performance, from inherent plant design, or from operator action while following recovery guidelines set down in procedures. The seven CSFs include:
Reactivity Control Reactor Coolant System (RCS) Pressure and Inventory Control RCS Heat Removal Containment Isolation Containment Emironment Control Radiation Control Vital Auxiliaries (VA) 1 The defrWtion of CSF is taken directly from CCNPP Q-Ust documentation which pre-dates the current version of the LR rule.
Therefore, the term possive"in the CSF def6rdtion is not necessarily identical to the term def6ned in this J-62gi and used for convenience in the SOC accompanying 10 CFR Part 54.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY l
5.(*) Carrent Licensing Basis (CLB) - De set of NRC requirements applicable to a specific plant and a licensee's written comnutments for assuring compliance with and operation
!. within applicable NRC requirements, and the plant-specific design basis (including all
,. modificatens and additions to such comnutments over the life of the license) that are i
<tackdM and in effect. De CLB includes the NRC regulations contained in 10 CFR Parts 2,19, 20, 21, 30, 40, 50, 51, 54. 55, 70, 72, 73,100, and appendices thereto; orders; heense conditions; exemptions; and technical specifications. It also includes the plant-specific design basis information defined in 10 CFR 50.2, as documented in the most recent Final Safety Analysis Report (FSAR) as required by 10 CFR 50.71, and the licensee's commitments remanung in effect that were made in
<tackdari licensing corr %*, such as licensee responses to NRC bulletins, genenc laters, and sluiwwa actions, as well as licensee comimtments niaeamented in NRC safety evaluations or heensee event reports. [l54.3]
- 6. Device Type (DT) - A more specific categorizatim of components according to their faadina and design. Equipment types (ETs) are broken into a number of DTs. For example, the ET for valves include DTs hand valve, check valve, control valve, and others.
Device types are the startmg point for grouping in the AMR tuk. Components are grouped by DT as they enter this task. Device types may be disided to form more specific groups if needed, or the DT may define the component group for evalustion.. Whenever the LR Rule calls for justifications for SCs, the discussions provided by the BGE IPA process ait at the device-type level.
- 7. Equipment Type (ET) - A general categorization of components according to their faadiaa and design. Examples of specific ETs are valve, piping, instrument, etc. For those SCs subject to AMR, the list of age-related degradation mechanisms (ARDMs) which needs to be addressed is developed for each ET. Structural components are categorized into generic groupings of concretc/ architectural and steel components.
- 8. Extended Operations, Period of- The additional amount of time beyond the expiration of the current operating license that is requested in the renewal application.
- 9. Function Catalog - A Function Catalog for a particular intended function of a system consists of the list of all system components required to support that intended function that are within the boundary of the given system.
- 10. Functional Requirements - The general, high level functions which an SS may be called on to perform. De functional requirements are used during the system scoping task to establish conceptual boundaries so that when a detailed function is determined to be an intended function, the evaluator will know which SS to associate the function with. The tenn " functional requirements" is used to distinguish these high level functions from the detailed intended functions contained in the screening tools and used during the component level scoping task.
11.(*) Integrated Plant Assessment (IPA) - A licensee assessment that demonstrates that a nuclear power plant facility's systems, structures, and components requiring AMR in accordance with {54.21(a) for LR have been identified and that the effects of aging on the 5 Resision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY functionality of such SCs will be managed to maintain the CLB, such that there is an acceptable level of safety during the period of extended operations. [{54.3] .
12.(*) Intended Function 'Ihose functions that are the bases for including SSCs within the scope of LR. [{54.4b]
- 13. Licensed Life 'Ihe maximum period of operations, in calendar years, as defmed by statute. For CCNPP, this period is 40 years.
- 14. Life Cycle Management Evaluation Database (LCMEVAL) - A computer-based application which is used to facilitate the component level scoping task for systems. The LCMEVAL was created, tested and h=aa'a4 in accordance with the BGE Quality ,
Assurance Program for Software Development, to justify its use in the safety-related (SR) scoping tasks. Master Equipment List data, Q-List data, drawing references, and other information useful in the scoping task are extracted one system at a time from controlled '
plant databases, loaded into LCMEVAL, and made available to the evaluator. The LCMEVAL helps to streamline the scoping task by automating key steps and facilitating storage and printing of the results.
15.(a) Long-Lived - Components are considered to be long-lived if tley are not subject to periodic replacement based on qualified life or specified time period. [{54.21(a)(1)]
- 16. Maintenance Strategy - A philosophy regarding the level and type of maintenance that a cumycrsit will receive throughout its life cycle. An adequate maintenance strategy is ;
. defined by the following program attributes: !
- a. Discovery - Identification of performance or condition degradation;
- b. Assessment / analysis - Comparison with criteria or other guidance to determine .
the degree of the degradation; l
- c. Corrective action - Mitigation of the degradation; and
- d. Confirmation / Documentation - Verification and documentation that the intended function was restored from its degraded condition as a result of the corrective action.
- 17. Master Equipment List (MEL)- A compilation of the HUCLEIS Equipment Icchnical l Database (NETD) technical data on equipment for a given system. ,
18.(a) Nuclear Power Plant - A commercial nuclear power facility of a type described in 10 CFR 50.21(b) or 50.22. [{54.3] !
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- CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY
- 19. NUCLEIS Database - A mainframe computer-based information system used to initiate, s
plan, schedule, track and provide a history of maintenance for all plant components.
NETD is an acronym used to denote the NUCLEIS Equipment Iechnical Database, which l is that part of the NUCLEIS information system, indexed by component, which contains ,
J informaten specific to each component. i 20.(*) Passive - A function is said to be passive if it is performed without moving parts or a
- change in configuration or properties in order to perform the function during normal -
operstmg conditions or in response to an accident. [{54.21(a)(1)].
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- 21. Plant Event Evaluations - Pre-existing evaluations which show compliance with i regulations concerning fire protection (FP), environmental qualification (EQ), PTS, 1 anticipated transients without scram (ATWS) and station blackout (SBO). These
! i evaluations provide the bases for in-scope determinations under {54.4 Criterion 3.
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- 22. Plausible Age-Related Degradation Mechanisms (ARDMs) - (See Aging Mechanisms) l An ARDM is considered plausible for a' specific co.rgst if, when allowed to continue without any prevention or mitigation measures or enkned monitoring techniques, it could not be shown that the component would maintain its capability to perform its intended, passive function throughout the period of extended operation.
- 23. Program / Activity (PA) - A group of procedures, formal or informal, that provide ,
reasonable assurance that SSCs are capable of fulfilling their intended functions. This '
may range from a formalized, long established group of produres to a one-time only procedure, j 24.(*) Renewal Term - The period of time that is the sum of the additional amount of time beyond the expiration of the operating license (not to exceed 20 years) that is requested in the renewal application plus the remaining number of years on the operating license currentlyin effect. [{54.31(b)]
- 25. Screening Tool - A summary of source document (s) compiled through the research of an event / topic which contains lists of responding SSCs and their intended functions.
- 26. Structure 'Ihe term structure, when used as a stand-alone term in this methodology, refers to a building. When a component of a structure is referred to, the term 'htructural co.rgst"is used for clarity.
27.(a) Structures and Components (SCs) - The phrase 'ttructures and components" applies to matters involving the IPA required by {54.21(a) because the AMR required within the IPA should be a component level review rather than a more general system level review.
[ SOC i.e., 80 FR 22462) In this Methodology, the term 'htructural components and components" (SCs) refers to the component level concept.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY l 28.(*) Systems, Structures and Components (SSCs) - Throughout these discussions, the term i W, structures and ev...ycs.cr.ts"is used when referring to matters involving the discussions of the overall renewal resiew, the specific LR scope 2 , TLAA and the LR findmg. [ SOC i.e.,80 FR 22462]
29.(a) Structure or Component Subject to Aging Management Review (AMR) - Stmetures and g- r ==n subject to an AMR shall encompass those SCs:
(1) That perform an intended function, as described in 654,4, without moving parts or _
a change in configuration or properties; and (2) That are not subject to replacement based on a qualified life or specified time period. [Q54.21(a)(1)]
30.(a) Systems, Structures, and Components within the Scope of LR - are:
L (1) Safety-related SSCs, which are those relied on to remain functional during and ;
following design basis events (DBEs) [as described in 10 CFR 50.49(b)(1)] to ensure the following functions: ;
(i) 'Ihe integrity of the reactor coolant pressure boundary (PB); i (ii) 'Ihe capability to shut down the reactor and maintain it in a safe shutdown ;
condition; or ;
(iii) The capability to prevent or mitigate the consequences of accidents that ;
could result in potential offsite exposure comparable to the 10 CFR Part 100 guidelines.
(2) All non-safety-related (NSR) SSCs whose failure could prevent satisfactoiy accomplishment of any of the functions identified in paragraphs (1) (i), (ii), or (iii) i of this definition.
(3) All SSCs relied on in safety yses or plant evaluations to perform a l functionthat demonstrates compliance with the Commission's regulations for j FP (10 CFR 50.48), EQ (10 CFR 50.49), PTS (10 CFR 50.61), ATWS <
(10 CFR 50.62), and SBO (10 CFR 50.63). [Q54.4a]
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31.(*) Time-Limited Aging Analysis (TLAA)- those licensee calculations and analyses that:
(1) Involve SSCs within the scope of LR as delineated in 654.4(a);
l (2) Consider the effects of aging l
2 Note that the CCNPP scoping process is a tvestep process with the initial step being conducted at the SSC or system level. The 4 second stop is conducted at the component kvel and the term SCs applies in this step. l l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY (3) Involve time-Imuted assumptions defined by the current operating term, for example,40 years; (4) Were det.....i.M to be relevant by the licensee in making a safety determination; ;
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(5) - Involve conclusions or provide the basis for conclusions related to the ability of )
the SSCs to perform its intended functions, as delineated in {54.4(b); and '
(6) Are contamed or incorporated by reference in the CLB.
[654.3) l l
l Table 2-1 List of Acronyms 1 l AFW Auxiliary Feedwster :
AMR Agmg Management Resiew ;
ARDM Age-Related Degradation Mechanism !
ATWS Antmipated Transient Without Scram I BGE Baltunore Gas and Electric Company ;
CCNPP Calvert Cliffs Nuclear Power Plant .
CCW Component Cooling Water I CEA Control Element Assembly CLB Current Licensing Basis CSF Critical Safety Function :
DBE Design Basis Event DT Device Type EP Electrical Panel EQ Environmental Qualification ET Equipment Type l l
FP Fire Protection FSAR Final Safety Analysis Report IL Instrument Line IPA Integrated Plant Assessment IR Issue Report LCM Life Cycle Management LCMEVAL Life Cycle Management Evaluation Database LR License Renewal LRA License Renewal Application MEL Master Equipment List NETD NUCLEIS Equipment Technical Database NSR Non-Safety Related PAM Post-Accident Monitoring PB Pressurc Boundary PTS Pressurized Thermal Shock PWSCC Primary Water Stress Corrosion Cracking RCS Reactor Coolant System SBO Station Blackout 9 Revision 1 4
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Table 2-1 List of Acronyms SCs Structures and Components SG Steam Generator SOC Statements of Consideration SR Safety-Related SS System and Structure SSCs Systems, Structures and Components TLAA Time-Limited Aging Analysis i
UFSAR Updated Final Safety Analysis Report VA Vital Auxiliary 2.2 Assumntions and Initial Conditions The IPA methodology relies on a number of basic assumptions and initial conditions, ucy include:
2.2.1 The scoping methodology assumes that the most effective approach in scoping SSCs is the use of two levels of scoping, i.e., system level and component level. This segregates SSCs into logical, manageable pieces and is similar to approaches used during design, construction, and operation.
2.2.2 The criteria underlying the system level and component level scoping tasks are identical.
2.2.3 The purpose of the IPA methodology is to provide a basis for the procedures which implement the steps of the scoping task and the IPA. Sections 1 through 5 of the methodology implement the requirements of 654.21(a)(2) to describe and justify the.
methods used in {54.21(a)(1).
Sections 6,7 and 8 go beyond the requirements of f54.21(a)(2) by describing the methods used to perform the AMR and TLAA review. However, the description of these methods should facilitate a better understanding of the results produced by these tasks. The results will be documented in the LRA and FSAR Supplement.
2.2.4 The IPA methodology is designed to make maximum use of existing BGE programs, system and equipment lists, documents, and databases to reduce duplication of effort and produce implementation results which reference equipment nomenclature already familiar to site personnel.
2.2.5 During the scoping task, tanks which are included in more than one site documentation system, e.g., both on the site structures list and as a component of a particular system in an MEL, are included only as components of a system during the IPA process.
2.2.6 Because the tasks described in this methodology are essential for providing the justCeation for the safety finding of {54.29, these tasks are performed in accordance with the BGE quality assurance program.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 2.2,7 Structural cu..@.ciits and components, which contribute to one or more passive functions and are long-lived, require evaluation to A man =* rate that the effects of aging are adequately managed.
Here are a vanety of methods available for managmg the effects of aging in order to assure the passive intended function. He appropriate method for a given situation depends on a number of factors, including the severity of the aging effects and the level of concern associated with degraded equipment condition. His correlation of the effects of agmg to the appropriate level of aging management is discussed in detail in Section 6 of this methodology.
2.3 IPA Methodolorv Overview he IPA methodology describes two scoping tasks, two IPA tasks, and the TLAA review task.
Each is described briefly below.
2.3.1 System Level Scoo'ma System level Scoping (Section 3) establishes boundaries for plant SSs, develops screening tools which capture the $54.4 scoping criteria, and then applies the tools to identify SSs within the scope of LR.
2.3.2 Component Level Scooing Component Level Scoping (Section 4) evaluates the components of SSs within the scope of LR to identify those which are required for the SS to perform its intended functions. Such cosi.posiciits are designated as within the scope of LR.
2.3.3 Pre-Evaluation Pre-evaluation (Section 5) determines which SCs, of those within the scope of LR, are subject to AMR. During the performance of this task, the following categories of SCs are eliminated from further IPA review:
> Bose which contribute only to active functions;
> nose which are replaced based on time or qualified life; and
> Those specifically excluded by the Rule language in {54.21(a)(1)(i).
%c result of this task is the list of all SCs in the given system which will be subject to AMR.
2.3.4 AMR The AMR task (Section 6) demonstrates that the effects of aging are adequately managed (see Definitions). Several different techniques for developing this justification are 11 Revision 1
l ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY I I
presented in this secten All the techniques provide the demonstration necessary to i support the findmg of 154.29 with respect to the management of effects of aging. j i
2.3.5 Commodity Evaluations ;
l Six cu.. ..cdity evaluations are described in Section 7 of the IPA Methodology. These i techniques are used for a specific set of coiripescsts found in a number of systems, but which perform the same or similar functions regardless of their system.
I 2.3.6 TLAA Review l 1
1
'Ihe TLAA Review is described in Section 8 of the IPA methodology. This task searches . i the CCNPP CLB, mdependent of the IPA process, to locate issues related to the current i operating life of the plant which also meet certain other specified criteria. For the j identified TLAA, the justification is provided that the time-limited issue is or will be '
addressed through one of the three approaches specified in {54.21(c). Note that this task !
is not technically part of the IPA, but its description is included in the IPA Methodology for convemence ;
TABLE 2-2 1
SOURCE DOCUMENTS This list of hments represents the sources used for developing the IPA methodology. This table does l not-represent all references which might be used in actually performing the tasks described in the !
ma'hadalagy. References used in the application of the methodology to a specific system are included in !
the implementing procedures and in the task-specific results. j 1
- 1. Life Cycle Management / License Renewal Program Management Plan, Revision 2, April 1992 ;
- 2. 10 CFR Part 54, " Nuclear Power Plant License Renewal, Final Rule," May 8,1995 i l
- 3. 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities" (routinely updated)
- 4. 10 CFR Part 100, Appendix A, "Scismic and Geologic Siting Criteria for Nuclear Power Plants,"
January 1,1991
- 5. Calvert Cliffs Nuclear Power Plant, Units I and 2, Updated Final Safety Analysis Report, Revision 17, November 1994
- 6. Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications Manual, through Amendment 205 (May 1995) for Unit 1, and Amendment 183 (April 1995) for Unit 2
- 7. CCNPP Design Standard, ' Structure and Component Evaluation," (DS-Oll) Resision 0, June 7,1995 12 Revision 1 i 1
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
- 8. CCNPP Design Standard ' Control of Equipment Technical Databases,"(DS-032) Revision 0, January 25,1995
- 9. CCNPP System Descriptions (various revisions)
- 10. NRC Regulatory Guide 1.97, "Instmmentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Revision 3
- 11. CCNPP Plant Drawings (various) '
- 12. NUREG-1377, " Listing of Nuclear Plant Aging Research Reports," and the reports themselves
- 13. Industry Technical Repons on PWR Reactor Vessel, PWR Reactor Vessel Internals, PWR Contamment, PWR Reactor Coolant System, Class 1 Structures and Environmentally-Qualified Cables in Containment l
13 Revision 1
ATTACHMENT (1) i
. CALVERT CLIFFS NUCLEAR POWER PLANT j "jfEGRATED PLANT ASSESSMENT METHODOLOGY 3.0 SYSTEM LEVEL SCOPING nis section describes how all plant SSs are reviewed to determine those that are within the scope of LR. His is accomplished through application of the system scoping task (Figure 3-1).
I=..Lg which SSCs are within the scope of LR is the first major task described in the IPA methodology. Section 54.21(a)(1) of the LR Rule states that the IPA must be conducted -
For those systems structures and components within the scope of this part, as delineatedin f54.4, . .
In other words, the results of the system level and component level scoping tasks are the startmg i
point of the IPA.
System level scoping consists of several activities. Section 3.1 describes how SSs are identified I and listed. Section 3.2 describes the development of conceptual boundaries for SSs. Section 3.3 describes the development of system screening tools. Section 3.4 describes how all in-scope SSs are identified. Section 3.5 describes how the scoping results are documented.
3.1 Identification of SSs De SS hstmg for CCNPP is provided in Table 3-1. De CCNPP Design Standard for " Control of the Equipment Technical Databases," (See Table 2 1, Reference 8) was used to develop the list of systems at CCNPP. His approach ensures that system designations are consistent with those established for current site programs and the MEL. The structures list was obtained through a review of the latest revision to the Plant Property and Building Drawing No. 61-502-E. Tanks identified on this drawing are not included in the list of structures since tanks are included as s,..w.;;;s of associated systems 3.2 Define Concentual Boundaries This step of the system level scoping task tabulates some basic information about each of the SSs l listed in Table 3-1. His information, referred to as the 'tonceptual boundaiies"of the SS, is '
needed to ensure a consistent understanding of what is meant by each of the SS names in this tabic.
He identification of the SS conceptual boundaries is accomplished by resiewing the CCNPP Updated Final Safety Analysis Report (UFSAR), Technical Specifications, and System Descriptions, as well as conducting interviews with experienced plant personnel. For each of the SSs listed in Table 3-1, a brief system description is developed and the functional requirements are identified. De description includes a listing of the major components and major system interfaces for each SS. The functional requirements list includes only the general, high level functions that an SS may be called on to perform. In the follow-on steps of the woping task, whenever an intended function is identified, the conceptual boundaries allow the evaluator to determine which SS the intended function should be associated with. The list of functional requirements does not represent ;
a detailed list of intended functions, but it is sufficient to establish the conceptual boundaries of i SSs. De component level scoping task (described in Section 4) develops a detailed list of SS intended functions.
14 Revision I
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ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT ;
INTEGRATED PLANT ASSESSMENT METHODOLOGY l
System Level . SS, Scoping Process 1 Denne conceptual boundertes and Structures Systems u Develop screerung toons l
P 54.4(e)(1) 54 4(e)(2) $4.4(e)(3)
Criterion Creerion Criterion D8E Vital FP,EQ, Flow Cherts Auxiliaries ATWS, SBO, ,
Tool PTS Tools l
u y u v is the building a '- i structure required g- structure required a structure required Ch 1 h7 i by the tool? by the tool? by the tool? !
Yes Yes Yes y E E d } I Add Function to Add Function to Add Functkn to Add Function to Intended Functions intended Functions -
Intended Functions -
Intended Functions -e W W W Mt )
l l l l 1
Ust ofintended functions for SSs Does the g
ystern or structure 2 have en w.nd.d
- Yes function?
SSs within the Scope ; - ,
of License Renewal L l No further action required for these S$s Figure 3-1 15 Resision 1
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY The following information is compiled for each SS and entered into a table designated as Table 1,
" System / Structure Information:"
> System or structure name;
> Unit number;
> Identification number,
> Brief description, including major components and system interfaces;
> Source document reference (for the description);
> System or structure functional requirement (s); and
> Source h==t reference (for each functional requirement).
3.3 Swa.2: Tools Prenaration Screemng Tools are created during the scoping task in order to add efficiency by allowing the evaluator to review each reference document only once, rather than once for each system. A screening tool is a summary of a source document or documents compiled through research of an event. The tool contains a list of SSCs which respond to the event and their intended functions.
The source documents identified in this section are reviewed against the (54.4 criteria contained in the LR Rule. For each criterion, appropriate information is taken from the source documents and summarized in one or more screening tools. The tools are then used to complete the screening task.
Each tool is described below. An example of a portion of a screening tool is provided in Table 3-2.
3.3.1 Tools Addressina 654.4(a)(1) and (2) 10 CFR 54.4(a)(1) and (2) (referred to as (54.4 Criteria 1 and 2) are addressed together in the System Level Scoping task since both of these criteria were used to establish the CCNPP Q-List documentation.
654.4 Criterion 1 (1) Safety-related systems, structures and components which are those relied on ,
to remain functional during andfollowing design-basis events [as defined in 10 CFR 50.49 (b)(1)] to ensure thefollowingfunctions -
(1) The integrity ofthe reactor coolantpressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (lit) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the 10 CFR Part 100 guidelines.
16 Resision I
l ATTACHMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 554.4 Criterion 2 (2) All nonsafety-related systems, structures and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraph (a)(1)(i), (ii) or (iii) ofthis section (i.e., f54.4). l 3.3.1.1 DBE Flow Chart Preparation The CCNPP UFSAR Chapter 14 DBE accident analyses listed below are resiewed. This list contains both design basis accidents and anticipated operational occurrences. No l extemal events are analyzed in Chapter 14 of the CCNPP UFSAR. All structures l designed to withstand DBE external events are designated as Class I structures at l CCNPP, and Class I structures are included within the scope of LR (Section 3.4.1.2). j l
Desian Basis Event Chaoter 14 Location .
CEA Withdrawal Event Section 2 l Boron Dilution Event Section 3 j Excess Load Event Section 4 Loss of Load Event Section 5 j Loss of Feedwater Flow Event Section 6 l' Excess Feedwater Heat Removal Event Section 7 RCS Depressurization Section 8 Loss of Coolant Flow Event Section 9 !
Loss of Non-Emergency AC Power Section 10 l CEA Drop Event Section 11
. Asymmetric SG Event Section 12 CEA Ejection Section 13 !
Steam Line Break Event Section 14 SG Tube Rupture Event Section 15 Seized Rotor Event Section 16 Loss of Coolant Accident Section 17 Fuel Handling Incident Section 18 Turbineficnerator Overspeed Incident Section 19 Containment Pressure Response Section 20 Hydrogen Accumulation in Containment Section 21 Waste Gas Incident Section 22 Waste Evaporator Incident Section 23 Maximum Hypothetical Accident Section 24 Excess Charging Accident Section 25 Feed Line Break Event Section 26 i
17 Revision 1
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1 ATTACHMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT i INTEGRATED PLANT ASSESSMENT METHODOLOGY l l
3
%e CCNPP Q-List includes Accident Shutdown Flow Sheets for 17 of the DBEs. Each Accident Shutdown Flow Sheet identifies the CSFs and plant fw*ians supporting CSFs, I which are necasary to reach safe shutdown for the DBE identified, maintain fission product boundanes, and prevent offsite releases in excess of established guidelines. These flow sheets also identify the supporting systems (as well as VA systems) which are required to satisfy the associated CSF. He DBE flow charts are a consolidation of Q-List l Accident Shutdown Flow Sheets and any additional supposting systems identified as relied 'j on for that accident in UFSAR Chapter 14.
1 For the eight DBEs which are identified in the UFSAR and are not the subject of Q-List Accident Shutdown Flow Sheets, a DBE flow chart is prepared during the system level scoping task. Dese DBE Flow Sheets contain the followmg information depending on the reason that no Q-List Accident Shutdown Flow Sheet was prepared (as documented in Q-
- List documenution).
Reason WhyNo Accident Shutdown JInformationIncluded in Scoping
' Flow Sheet is in the Q List- 'Results DBE Flow Chart No active components are relied on to Passive components which mitigate mitigate the event. the DBE.
No active or Passive components are A note stating that no active or passive required to mitigate the event. components are required to mitigate the event.
All cu..W.ts relied on for the event A note statmg that all components are already included in another Accident required to mitigate the event are Flow Sheet, included in another DBE Flow Sheet, and specifyin6 which other DBE(s). ,
The DBE flow charts for the remaining 17 DBEs identify the systems and the functions provided by each of these systems in order to support the CSFs necessary to reach safe ;
shutdown for the specific DBE, maintain the fission' product barriers, and prevent offsite releases in excess of established guidelines.
Q-List documentation also contains a specific flow sheet for VAs. Electric power distribution; control air; cooling water; and heating, ventilation, and air conditioning functions for the SR equipment required to respond to each DBE are annotated in the corresponding Q-List Accident Shutdown Flow Sheet. The Q-List Vital Auxiliaries Flow Sheet is a compilation of the systems performing these VA functions for all of the Q-List Accident Shutdown Flow Sheets. De VA screening tool prepared during the system level scoping task duplicates the SSCs listed on the Q-List Vital Auxiliaries Flow Sheet using the SS nomenclature shown in Table 3-1.
3 The terme Q-List Accident Shutdown Flow Shost"end Vital Auxihenes Flow Shoote*ere used to refer to documentation which already emieted as part of the CCNPP Q-Ust. The terme DBE Flow Chart"end Vital Auupieries Screening Tool" ore used to denote j the document created during the scoping process to compile the Q-List information and other specired lnformation. j l
18 Revision 1
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ATTACHMENT (1) i-CALVERT CLIFFS NUCLEAR POWER PLANT
- INTEGRATED PLANT ASSESSMENT METHODOLOGY All systems and fuiam identified in the DBE flow charts and the VA screening tool are l coded (by shadmg) to identify the source document (s) (i.e., UFSAR, Q-List Manual, or l both).
a i
By relying on the Q-List Accident Shutdown Flow Sheets and Vital Auxiliaries Flow-
! Sheets, all SR SSs are identified, as well as all SSs that could fail and prevent the
! functioning of SR SSCs. This identification is not limited to first level, second level or any i specific level of support equipment. Rather, the scoping is performed consistent with the
! CCNPP Q-List Design Standard which was developed with the intent of identifying and j
controlling a similar4 scope of SSCs to that defined by the first two criteria of {54.4.
1herefore, the CCNPP scoping task is consistent with the Commission's intent stated in the SOC to tlic LR Rule.
An applicant for LR should rely on the plant's CLB, actual plant-specifc experience, industry-wide operating experience, as appropriate, and existing engineering evaluations to determine those NSR systems, structures, and components that are the initialfocus ofthe LR review. (60 FR 22467) 3.3.2 Tools Addressina 654.4(aK3) 654.4 Criterion 3 (3) All systems, structures and components relied on in safety analyses or plant evaluations to perform afunction that demonstrates compliance with the Commission's regulationsforfre protection (10 CFR 50.48),
environmental quallycation (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62),
andstation blackout (10 CFR 50.63).
Plant evaluations have been perfonned to demonstrate compliance with the regulations identified in {54.4(a)(3) (referred to as (54.4 Criterion 3). These evaluations are resiewed to identify SSs that are relied on to mitigate the subject plant event as well as any systems or structures whose failure would result in failure of other equipment to mitigate the particular event. As was the case for Criteria 1 and 2, an SS is listed as within the scope of LR when the mitigation function or support function associated with it is credited in the analysis or evaluation. Mentioning an SS in the analysis or evaluation does not necessarily indicate that the SS contributes to an intended function.
Additionally, if the SS function is identical to a SR function (as identified in the Q-List),
then the function need not be repeated on the tools addressing 554.4 Criterion 3. The analyses and evaluations being reviewed in this step are used to identify intended, NSR functions.
4 The CCNPP Q-List documentation eleo a=m%es controis for PAM (Cetegory 1 and 2) equipment Post-Accident Monitoring equipment setienes {54.4 Criterion 3, rather then 1 or 2.
19 Revision I
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3.3.2.1 FP Scimdas Tool Preparation l
- Ihe CCNPP 'UFSAR, FP Program h=entation and the CCNPP Interactive Cable l Analysis are reviewed to idemify the system functens that address the Commission's i regulations on FP and the BGE commitments for implementation of those regulations. The i identified SSCs, their intended function (s), and the appropriate source documents with revision numbers are summarized in the FP Tool.
3.3.2.2 EO Screeninn Tool Preoaration ,
Two tools are produced for this criterion, the EQ tool and the PAM tool.
'Ihe Q-List data in the NETD is reviewed to identify items listed as 5049 (items which must meet the requirements of 10 CFR 50.49). A list of the systems containing 0-;; ==ts designated as EQ is prepared with the Q-List revision number (or date, as appropriate) provided as a reference.
The CCNPP UFSAR is reviewed to identify the systems containing components required for PAM category 1 or 2 variables (as defmed in Regulatory Guide 1.97). A PAM System summary table is prepared. It lists each system which is required for PAM, the variable (s) it monitors, and the appropriate source document and revision.
3.3.2.3 PTS Screeninn Tool Preoaration Since neither CCNPP Unit I nor 2 is expected to require an evaluation in accordance with Regulatory Guide 1.154 in order to satisfy 10 CFR 50.61 requirements, no equipment is included within the scope of LR due to the PTS Rule. The PTS Screening Tool is provided in the System Level Scoping Results, but this tool merely notes that no SSCs are relied on for this event. Additionally, the System Level Scoping Results, the component level scoping task, and the component level scoping results for each system include the contingency to implement a PTS scoping criterion, but the results indicate no PTS-related SSCs. If a Regulatory Guide 1.154 evaluation is required at some point in the future, the scoping task would be modified to require incorporating the PTS functions' relied on in the 1.154 analysis into the PTS Screening Tool. The Regulatory Guide 1.154 analysis would also trigger an update to the system level and component level scoping results to include the SSCs associated with the 1.154 functions within the scope of LR.
3.3.2.4 A'IWS Screeninn Tool Preoaration The CCNPP UFSAR is reviewed to identify the system functions that address the 10 CFR 50.62 requirements on ATWS. An A'IWS Screening Tool is developed. The tool lists the SSCs which are relied on in response to an ATWS event. For each identified SS, the tool lists the intended function (s) provided and the appropriate source documents with the revision number.
20 Revision 1
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ATTACHMENT (1) !
! CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY j J 3.3.2.5 SBO Sciwiu Tool Preoaration ne Staten Blackout Analysis is reviewed to identify SSs which are relied on during the j
" coping duraten" phase of an SBO event. An SBO Screening Tool is prepared which lists 1 the SSs relied on in the Station Blackout Analysis, the function (s) that each provides, and the appropnate source documents with revision numbers. De power restoration phase of l the Station Blackout Analysis is specifically excluded from review in this criterion since l several success paths for restoring power aRer an SBO are already screened as within the l scope of LR due to Criterion 1 (SR). l 1
3.4 SS Scoome <
ne scoping task is implemented for each SS by reviewing each of the screening tools generated in l Section 3.3 and developing a System Level Scoping Results Table. (An example page of the l System Level Scoping Results Table is shown in Table 3-3.) For the DBE tools and the VA tools, j the function (s) being provided are noted on the System Level Scoping Results Table. Since the events summanzed by the tools address the requirements of the {54.4 criteria, inclusion of an SS in a tool indicates that it is within the scope of LR.- It is important to note that all intended functions -
are identified for each SS during the scoping task. Identifying only one intended function would be i sufficient to make an in-scope determination; however, the list of all intended functions for an SS l facilitates the component level scoping task. His step is repeated for each SS so that an in-scope !
determmation is made for each. j 3.4.1 Criteria 1 and 2 - SR and SR Sunoort SSs i l
3.4.1.1 DBE Flow Charts and VA Seiwss Tool l De DBE flow charts and the VA screening tool, (see Section 3.3.1.1), are used to identify those SSs whose functions support the CSFs for a DBE, or whose failure would prevent performance of the CSFs. Systems and structures listed in one or more of the DBE flow charts or the VA screening tool are included in the System Level Scoping Results Table under Criteria 1 and 2. For each SS listed in the results table, all applicable DBEs are l identified along with the functions that the SS provides for each DBE. The source '
document references and revision numbers are not included in the scoping results table since this information can be found in each DBE flow chart or the VA screening tool.
3.4.1.2 Class 1 Structures For all listed structures, the UFSAR Section 5 and Q-List Design Standard are resiewed to determine whether the structure or a portion thereof is designated as SR, Class 1. At CCNPP, all Class I structures (buildings) are designated as SR; therefore all Class I structures are screened as within the scope of LR. The results of this scoping step are incorporated, along with the appropriate source document references and revision numbers or dates, into the System Level Scoping Results Table for each of the structures.
21 Revision 1
l ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3.4.2 - Criterion 3 - SSs Relied On in Plant Safety Evaluations i The correspondmg screening tools (see Section 3.3.2) are used to identify the following SSs:
- 1) Those that perform functions designated as required for FP;
- 4) Any combination of these factors.
If one of the SSs being screen:xt is listed in any of these tools, it satisfies Criterion 3. The results of this scoping step are incorporated into the System Level Scoping Results Table for each of the SSs. The source document references and resision numbers are not included in the scoping results table since this information can be found in each screening tool.
3.5 Results As a result of system level scoping, SSs are assigned to one of two categories: (1) those that are within the scope of LR; and (2) those that are not. Systems and structures that belong to category (1) require further scoping in preparation for the IPA process and proceed to component level scoping, as described in Section 4.0.
l l
l l
22 Revision i
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER Pl. ANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES 1 Switchyard (500 kV) & Switchyard DC 48 Engineering Safety Feature Actuation 2 Electrical 125VDC Distribution 49 Simulator Computer 3 Electrical 13kV Transformers & Buses 50 Solid Waste Disposal 4 Electrical 4 kV Transformers & Buses 51 Plant Water 5 Electrical 480V Transformers & Buses 52 Safety injection 6 Electrica! 480V Motor Control Centers 53 Plant Drains 7 Electrical 13kV Unit Buses 55 CEA Drive Mechanism & Electrical 8 Well and Pretreated Water 56 Reactor Regulating 9 Intake Structure 57 Technical Support Center Computer 11 Service Water Cooling 58 Reactor Protective 12 Saltwater Cooling 59 Primary Containment 13 FP 60 Primary Containment Heating & Ventilation 14 Transformer Deluge 61 Containment Spray 15 CCW 62 Control Boards 16 Electrical 250VDC 63 Cathodic Protection 17 Instrument AC 64 Reactor Coolant 18 VitalInstrument AC 65 Seismic 19 Compressed Air 66 Cavity Cooling 20 Data Acquisition Computer 67 Spent Fuel Pool Cooling l 21 Domestic Water 68 Spent Fuel Storage 22 Makeup Demineralizer 69 Waste Gas 23 Diesel Oil 70 Refueling Pool 24 Emergency Diesel Generator 71 Liquid Waste 25 Access Control Area Ventilation 72 Sewage Treatment Plant 26 Annunciation 73 Hydrogen Recombiner 27 Auxiliary SGs 74 Nitrogen and Hydrogen 28 Auxiliary Steam 75 Low Voltage DC Control Power 29 Plant Heating 76 Secondary Sample 30 Control Room Heating, Ventilation 77/79 Area / Process Radiation Monitoring
& Air Conditioning 78 Nuclearinstrumentation 31 Meteorology Tower & Miscellaneous 80 New Fuel Storage and Elevator Computers 81 Fuel Handling 32 Auxiliary Building and Radwaste 83 Main Steam Heating & Ventilation 84 Reactor Vessel Intemal 33 Turbine Building Ventilation 85 Plant Access and Surveillance 34 Condensate Precoat Filter 86 Power Plant Security i 35 Chemical Additions-Turbine 87 Unit Transformers 36 AFW 88 Visitor Center Security 37 Demineralized Water and Condensate 89 Emergency Operations Facility Security Storage 90 Service Building & Outlying Building 38 Sampling System Heating, Ventilation & Air Conditioning 39 Condensate Polishing Demineralizer 91 Lube Oil Storage 41 Chemical and Volume Control 92 Gland Steam 42 Circulating Water 93 Main Turbine 43 Condenser Air Removal 94 Plant Computer 44 Condensate 95 Carbon Dioxide 45 Feedwater 96 Fire and Smoke Detection 46 Extraction Steam 97 Lighting and Power Receptacle 47 Feedwater Heater Drains and Vents 98 Main Generator and Excitation 23 Resision 1
ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES (Continued) 99 Cranes / Test Equipment 105 Weight Testing Wire Ropes & Slings (3) 100 Plant Communications 106 Ladders and Gratings (3) 101 Dry Fuel Storage 107 Roads 102 Plant Areas 108 Docks and Marine Related Structures 103 Emergency Diesel Generator Building 109 Shop Equipment (3)
Heating, Ventilation & Air Conditioning (2) 110 Manual Valve Components (3) 104 Lubrication 111 Materials Processing Facility (3)
AdditionalStructures Auxiliary Building.
Condensate Storage Tank No.12 Enclosure Domestic Water Treatment Plant Engine Generator House Equipment Hatch Access Building. No.1 Equipment Hatch Access Building. No. 2 FP Pump House Fuel Assemblies Fuel Oil Storage Tank Nc. 21 Building.
Hydroptn Storage Pad ModifN.ations Mechanical Lock-up (No. 3)
Modifications Mechanical Lock-up (No. 4)
Oil Interceptor Pit Service Building [B-3)
South Service Building.
Switchgear Structure Transformer Foundations Turbine Building Waste Water Treatment Building.
Well Observation Building Well Water Pump House Independent Spent Fuel Storage installation (4)
Diesel Generator Building 1 (2)
Diesel Generator Building 2 (2)
NOTES:
- 1. System listing is from Attachment 6 of DS-032, Control of the Equipment Technical Databases"
- 2. Sr4 ems and structures associated with the new diesel generator installation do not t.f come part of the CCNPP licensing basis until after the 1996 refueling outage, and therefore, are not yet included in the scoping results.
- 3. These systems were not included as systems in the LR scoping process because they are portable equipment or because they are already included in other systems.
- 4. The independent Spent Fuel Storage Installation is not licensed under 10 CFR Part 50 and, therefore, is not in the scope of this LRA.
I 24 Revision 1
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-2 Revision 4 Post-Accident Monitoring Screening Tool (Example) l Reference 1 - Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Updated Final Safety l Analysis Reoort (UFSAR), Section 7.5.8 l
[ Reference 2 - Calven Cliffs Nuclear Power Plant, NUCLEIS Equipment Database l
l SYSTEM / SYSTEM STRUCTURE ID No. MONITORING VARIABLE (S)/ FUNCTION (S)
Electrical 125VDC 2
- Status of standby power (voltage, current)
Distribution Electrical 4kV 4 + Status v standby power (voltage, current)
Transformers and Buses Electrical 480V 5 Status of standby power (voltage, current) l Transformers and Buses Senice Water 11 Senice water pump status (motor current) l
. Containment cooler cooling water flow Saltwater 12
- Saltwater pump status (motor current)
CCW 15 CCW heat exchanger outi;; temperature
+ CCW to/from reactor coolant pumps containment isolation valve position CCW pump discharge pressure (for flow indication)
- CCW pump status (motor current)
VitalInstrument AC 18 Status of standby power (voltage)
Compressed Air 19 Instrument air containment isolation valve position indication Data Acquisition 20 Provide fault protection for Instmmentation & Controls Computer loops Emergency Diesel 24
- Status of standby power (voltage, current, VAR, frequency)
Generator Auxiliary Building & 32 Fuel pool exhaust fan damper position Radwaste Heating &
Ventilation AFW 36
. Motor-driven AFW pump status (motor current)
. Condensate storage tank 12 level Sampling System 38 . Containment hydrogen concentration 25 Revision I
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I ATTACHMENT (1) ,
L CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY ;
TABLE 3-3 i eGE I.CM PROGRAM j TABLE 2 ;
9 SYSTEM LEVEL SCOMNG RESul.TS (EXAMPLE) Rowleion 4 CRITERIA 1 & 2 wuizenvis 3 Re(d Cimme i Class I or SR. In Scope l System / Structure Unit D for DOE DOE Plant Punction(s) Q or SR-1M 1M Reference PAM PP ATVE SSO PTS EQ Yeesto l Suelechyard (500 kV) 142 1 No None No WA N/A No No No No No No No j and Switchyard DC -
l Elodncel 125 VDC 142 2 VA VA for Chem 6 col & Volume Centrol System No WA N/A Yes Yes No No No No Yes !
) k
VA for Mein Steam t
- VA for Containment Spray VA for Primary Containmort Hosting &
l l !
vertilation i VA lor Emergency Diesel Generators !
VA for 4KV Transformers & Buess l VA ter 480V Motor Cottrol Corners VA for 480V Bus System E i
i VA forVItalinstrument AC VA for Service Water VA for CCW VA for Selhuster Cooling j VA for Control Room Hosting, Vanillation :
r
& AirCondttoning I
VA for Auxiliary Buleding & Reduumete !
Heating & Ventilation VA for RCS !
, VA for Emergency Selsey Features Actus- t
' t (Core Flush) t l
Electncel 13kV 142 3 No None No N/A N/A No No No No No No No '
Transformers and Buses i Eledncel 4kV 1&2 4 VA VA for AFW No N/A N/A Yes Yes No No No No Yes Transformers and Buses VA for Sology trpction i VA for Coitamment Spray l VA lor 480V Bus -i VA for 480V Motor Control Centers ;
VA for Service Water
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_ _ _ _ . . . _ _ _ . _ _ _ ___ _ _ _ _ _ _ _ _ _ _ . , . - _ - _ - . . . . - . . . _ _ , . _ . . , _ - ~ _ . - - . . . . . - . _ _ ..#
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR. POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY 4.0 COMPONENT LEVEL SCOPING Component level scoping is the second and fmal task needed to determine the scope of SSCs to be ;
addressed by the IPA for aging. The criteria for including components within the scope of LR are the same as those for SSs and are defined in (54.4.
The component level scoping task is conducted one system at a time for each SS designated as within the scope of LR. The scoping is accomplished through application of either the component level scoping task for systems, which is illustrated in Figure 4-1 and discussed in Section 4.1, or the component level scoping task process for structures, illustrated in Figure 4-2 and discussed in Section 4.2. Section 4.3 describes several variations to the standard component level scoping process used in specific instances Section 4.4 describes how the results are documented.
4.1 Component Level Scooine for Systems The component level scoping task for systems is implemented by systematically reviewing the intended functions of the system (determined by the system lael scoping task) to determine which system components contribute to the performance of the functions. Components are designated as within the scope of LR if they are required for their system to perform an intended function.
The component level scoping task for systems is divided into several distinct steps. Each step is discussed below. ,
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY intended functions for the Component
'Y't** *"S " P'd Level Scoping Process for DBE Flow Charts , Systems PAM, SBO, FP, PTS, . Describe intended function ATWS, EQ Screening in more detailif needed.
Tools _
Other implicit intended functions; e.g., PB,1E, structural support. "
Consolidate functions to eliminate duplicates u
MEL for the System For alintended
~~" functions of the system System Level Scoping u Results & References List an system Function catalog 01 components which are _
required to perform the Function catalog 02 function or could fail Plant drawings and prevent the _, a function a e
U Q-List documentation Function catalog n Next intended function
\ / h Operating Instructions f l Resort function
,, catalogs by component List of system 3 components and their intended function (s).
Wrkikwksskidik Figure 4-1 28 Revision 0
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Component Level Scoping for Structures oes the structure have y, em type components
{ ,
Perform component
, level scoping using the 4 system process for system type identify structure intended function components.
- Structural support to SR equipment
- Shelter /prote%on for SR equipment
- Preesure or fission product boundary !
- Miselle barrier I
- Class 11/l support
- Flood protection barrier
- Rated fire barrier v
Determine generic structural component types ;
in this structure.
v Add unique structural component types.
v identify structural component types which contribute to each intended function.
v Add supports for large SR equipment to scoping results.
v Integrate scoping results for system type and ,
structural type components.
v List of structural component types and their intended functions ;
Figure 4-2 nmmewmemswee ~
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 4.1.1 hinmina of D*iled System Functions
'Ihe purpose of this step'of the scoping task is to create a detailed list of the intended functions associated with the system being scoped The list is compiled in a System Functions Table using the System and Structure Scoping Results, Q-List documentation, plant drawings, the UFSAR, System Descriptions and other references. It should be noted that these intended functions are required to be performed under a variety of design conditions in accordance with the CLB.
'Ihe System and Structure Scoping Results contain screening tools which associate intended functions with individual systems The first step of creating the detailed function list is to review all of the screening tools and, in the System Functions Table, record the intended functions of the system being scoped The CCNPP Q-List Design Standard (Table 2-1 Reference 8) is the site reference which governs what components are controlled as SR, SR support, or other miscellaneous category equipment. To ensure consistency with the Q-List documentation, the LCMEVAL software application is used to compile a listing of all Q-List categories which are associated with any components in the system being scoped (Q-List Criteria listing).
This listing represents the Q-List related functions associated with the system being scoped The following Q-List categories correspond to {54.4 criteria as described below:
l Q-List Flow Sheets -
1hese flow sheets identify components which are relied on to respond to UFSAR l Chapter 14 DBEs or serve as VA to SR equipment. Criteria 1 and 2.
PB - The category of PB mechanical items which maintain the system PB of the RCS, maintain the radiological boundary to prevent exceeding 10 CFR Part 100 limits, or maintain safety system boundary to limit system leakage. Criteria 1 and 2.
(Criterion 2 because PB includes the components needed to maintain the PB of fluid systems which are not fission product boundary fluid systems.)
lE - The category of electrical equipment and systems that are essential to emergency !
reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment. Criteria 1 and 2. (Criterion 2 because 1E includes electrical isolation devices whose sole " intended" function is to prevent an electrical fault in a NSR portion of the system from affecting the SR functions of the system.)
IM- The category of mechartcal i equipment that is essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the environment. Criterion 1.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY PAM - Post-accident monitoring category ofinstrumentation used to assess the environs and plant conditions during and following an accident. Criterion 3, subset of EQ.
5049 - His category identifies items which are required to be environmentally qualified to the requirements of 10 CFR 50.49. Critenon 3.
CLSI- The category for those SSCs, including their foundations and supports that are designed to remain functional in the safe shutdown carthquake, as defined in 10 CFR Part 100. Criterion 2. ("CLSl" is the Q-List Manual designation for items referred to as " Seismic Category 1" or " Class 1" elsewhere in this methodology.)
Q- %e category for any item specified by the Q-List Committee as requiring the same level of quality assurance as provided for SR items. (Criterion to be determined during scoping.)
SBO- The category of equipment required to withstand and recover from an SBO event. Criterion 3.
After producing the Q-List Criteria Listing for the system being scoped, this list is consolidated with the functions already listed in the System Functions Table to finalize the detailed functions listing for the system. nc Q-List does not contain information related to several of the regulated events in (54.4 Criterion 3. Herefore, for the categories shown below, no consolidation with Q-List-related functions is possible. He associated screening tools and their references are used to validate the detailed system function (s) for these criteria.
FP- The functions required by 10 CFR 50.48 for FP and safe shutdown after fire.
AMS - The functions required by 10 CFR 50.62 to provide diverse scram and diverse turbine trip capability during an AWS event.
PTS- The functions required by 10 CFR 50.61 to provide protection during a PTS event.
The final step of intended function identification is to eliminate redundant functions.
Functions enveloped by another function or identical to another function are consolidated.
De enveloping function is designated as the " Parent" function, while the enveloped .
function is the " Child" function. The child function is retained on the System Functions l Table in order to be able to trace the steps of the process which created the table. Parent functions and functions for which no consolidation is possible are assigned a unique identification number (Function ID) to facilitate subsequent steps in the scoping task. (For the remainder of this methodology, the term " intended function" refers to a parent function unless otherwise specified.)
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 4.1.2 "Ihe MEL ?
To ensure that all cu.i.pcssts in the plant are scoped with one and only one system, the site MEL is used to provide the equipment list for the cuiipcsst level scoping task for each system. 'Ihis list is the portion of the NETD which contains all equipment for a given system In developing the NETD, conventions were established for determining the boundaries betww. systems 'Ihese conventions provided the guidance for determining which system each co...pc.ws.: in the IPA would be assigned to. Several example conventions are listed below The complete system boundary guidelines are =*=lacA in the site design standard for controlling equipment technical databases
> Heat exchen.ss are assigned to the load system. -
> Electrical components are assigned to load s3*cm from the load side of the circuit breaker.
> Sensors are assigned to the system in which they sense. Actuators are assigned to the system in which the actuation takes place. :
> Transformers are assigned to the lower voltage system. .
i As each scoping task is begun, the LCMEVAL software application is loaded from the NETD with the MEL for the system to be scoped Each of the components on this list >
must be dispositioned during the scoping task as either contributing to an interxicd function listed in the System Functions Table or not needed for any of these functions.
4.1.3 Develooment of Function Catalons
'Ihe next step in the ceiipc.. cat level scoping task for systems is to determine, for each intended function, which components from the system MEL are needed to perform the 1 function. A list of components for each function is called the function catalog.
l l
In order to determine the relationship between a given function and the components contributing to the function, Q-List documentation, UFSAR, Technical Specifications, system screening tools and references associated with the screening tools are used.
The active components associated with mitigating the consequences ofindisidual DBEs or providing VA functions to SR equipment are listed in the plant Q-List documentation along with a reference to their safety function (s). Consequently, whenever a System Functions Table contains a DBE function or a VA function, the Q-List provides a direct input to the scoping task for determining which components of the given system contribute to {54.4 Criterion I and 2.
The Q-List documentation also includes Piping and Instrumentation Drawings which are coded to reflect the portions of each system which passively support the system PB l 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY function for that portion of the system relied on to mitigate DBEs. Whenever the system function table contams DBE functions and the MEL contains mechanical PB components, a PB function catalog is created for the system. For each component in the MEL, a det ..d..idion is made, based on these Q-List-coded Piping and Instrumentation Drawings, whether the w..gsst is within the e~i PB portion of the drawing. If so, the ,
sr.ycz.cs.t is included in the PB catalog. nose passive components which perform in exactly the same manner tbr any intended function are not included in catalogs associated with other functions in order to avoid redundancy. i 1
The Q-List documentation also contains listings which associate specific components to ;
PAM and EQ functions. His listing is used as a direct input to the scoping task whenever :
PAM or EQ functions are contained in the system function table. Based on this input, a l function catalog is created for both PAM and EQ. In order to be more specific regarding j which w..ycssts actually contribute to providing each of the required PAM indications, l plant drawings and the BGE UFSAR are consulted. In addition to the component listing, the PAM catalog contams a letter in the notes column to specify which PAM indication is I associated with each component.
He Q-List <l=maatation contains a listing which associates specific components to the Class 1 function. His listing is used as a direct input to the scoping task whenever there is a Class i function in the System Functions Table. Based on this input, a function catalog is created for Class 1. His catalog normally contains EPs and other enclosure 2 devices which contain SR equipment but have no explicit active safety function.
Many electrical and a few mechanical components are identified in the Q-List Manual as >
IE only or IM only. Such components perform the same function in support of a number ofimportant events but are not actually associated with any particular DBE in the Q-List j documentation. When a system contains components that are SR and designated only as IE or IM, a separate function catalog is created to contain these components.
He NETD contains a field which associates specific components with the Station '
Blackout Analysis. This SBO designation is used as an input to scoping for SBO and further review is conducted during the IPA process as described below:
> De NETD SBO designation is assigned to components mentioned in the Station l Blackout Analysis. Other components which must function so that these l
" mentioned" components can perform their SBO function are identified and added I to the SBO function catalogs. ;
> Much of the equipment mentioned in the Station Blackout Analysis is mentioned because it is secured at the start of an SBO event or is used when restoring power ,
after the end of the event. These components do not contribute to any SBO i functions in the SBO tool, and therefore are not included within the scope of LR. I Dese components are not included in the SBO function catalogs.
When the step is complete, the SBO function catalog or catalogs contain all of the system components which contribute to each intended SBO function.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
'Ihe equipment in the system MEL which is designated in Q-List documentation as SR category "Q" also requires further analysis during the scoping task. 'Ihe documentation which supports the classification of these type cui.pcssts is reviewed to determine why 2
the equipment has been designated as SR category Q. If the SR-Q components perform an intended functon, the components are included in the corresponding function catalog.
Otherwise, the components are categorized as not within the scope of LR.
For the ATWS, PTS and DBE functions contained in the System Functions Table, one ;
function catalog is created for each listed function. The reference information used to I create the associated screening tool is consulted, as needed, along with plant drawings to determme exactly which system components contribute to the perfonnance of each listed function. Cwi ycssts which perform er.actly the same function to support one of these criteria as they perform to support a SR function, are not repeated again in these function catalogs to avoid redundancy For example, if a pump is required to start during a severe fire to ensure plant shutdown and the same pump must start to pro 5ide cooling w~4ter to SR equipment to mitigate the consequences of a DBE, that pump would not be repeated in the FP function catalog.
All of the function catalogs discussed above are created using the LCMEVAL software system which contains data loaded directly from a controlled site database (NETD) where possible. For the functions where no source of direct component data is available in software format, the individual components are entered one at a time into the function catalog. The software ensures that only valid components (i.e., in the MEL for the system being scoped) are added to function catalogs. It also facilitates the recording of reference documents whichjustify that a component supports a given function.
1 4.1.4 Generation of Scopine Results Table In the next step of the component level scoping task for systems, the function catalogs that were developed in Section 4.1.3 are resorted by LCMEVAL to produce a list of system components and the intended functions associated with each component. Components not associated with any intended function are designated as not within the scope of LR by the LCMEVAL software system. The table of in-scope components and the intended functions that they contribute to is designated as the Component Level Scoping Results Table.
4.2 Camr:r= Level Sconine for Structures The component level scoping process described above for systems can also be applied to structures. However, this process is somewhat different because of the unique features of structures and how they are documented on site. As with systems, scoping is implemented by determining which structural components are required for the performance of the intended functions of the structure. Details of the methodology implementing the structural componcat scoping are presented below.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY l i
1 4.2.1 Uniaue Th&rs for Structural Components !
The components of structures have not generally been identified and listed in an MEL.
C=~--a'ly, the cc.r.e .cr.t cr level scoping for structures cannot use a comprehensive equipment listing as an input.
I For certain site structures, such as the contamment, specific ec riporient types have been identified in the site equipment database For these structures, a partial MEL is available and the structural cc.iiper.cr.t scoping task is divided into two parts: i i
- 1) 'Ihe components documented in an MEL for the structure are scoped as described I in Section 4.1, above, ifit is determined that they do not perfonn a structural-type !
function. Examples include components, such as the containment personnel hatch, i the personnel hatch limit switches, and the containment penetrations because they ,
1 are designated as evirycr.cr.ts of the containment system in the NETD.
- 2) The remaming portions of the structure such as beams, columns and walls are scoped per the remaming steps of Section 4.2.
The results are then merged when both procedures are complete to present a combined scoping result for the entire structure.
4.2.2 Function Identification ,
The SS scoping task identifies some structures as within the scope of LR because they are ,
designed to Class I criteria or because they are required for DBE purposes Unlike the j scoping results for systems, the Class I structure in-scope determination does not actually )
reveal a great deal about the intended functions of the structure. Therefore, during the co...yor.cr.t level scoping, the evaluator reviews Chapters 5 and 5A of the UFSAR to determine specific structure design basis information such as which external events the structure is designed to withstand, and which structural components contribute to these intended functions.
By their nature, structures perform mostly passive functions and are constructed in accordance with predetermined design requirements. Therefore, civil engineers experienced with nuclear plant structures determined that a structure, or components of the structure, are designed to perform one or more of the following functions in support of the
$54.4 criteria:
- 1. Provide structural and/or functional support to SR equipment;
- 2. Provide shelter / protection to SR equipment. (This function includes radiation protection for EQ equipment and high energy line break-related protection equipment.)
- 3. Serve as a PB or a fission product retention barrier to protect public health and safety in the event of any postulated DBEs; 35 Revision 0
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INTEGRATED PLANT ASSESSMENT METHODOLOGY ,
- 4. Serve as a missile barrier (intemal or external); l
- 5. Provide structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions (Example: seismic Category II over I design considerations);
- 6. Provide flood protection barrier (internals flooding event); and
- 7. Provide a rated fire barrier to confme or retard a fire from spreading to or from adjacent areas of the plant.
His listing allows an evaluator with a specific civil engineering background to determme which of the generic structure functions apply to the structure being evaluated without being an expert on DBEs.
Functions 1-4 are associated with Class I structures. Class I design requirements are the structure level equivalent of SR components specified in {54.4 Criterion 1. In a similar fashion, fWiaa= 5 and 6 apply to non-Class I structural components which could, if they fail, prevent a SR function from occurring. His is the structural equivalent for {54.4 Criterion 2. Function 7 is the equivalent for the portion of {54.4 Criterion 3 which is applicable to structures.
The applicability of each function to the structure is determined by a resiew of various j source documents. If the structure is a Class I structure, the UFSAR and the System and Structure Scoping Results must be referenced to determine which of functions 1-4 apply.
The applicability of functions 5 and 6 to the structure being scoped cannot be mr.de based only on the UFSAR and the System and Structure Scoping Results. Therefore, the determination of the applicability of these criteria to the structure is deferred until l Section 4.2.4. To determine whether the structure being evaluated performs function 7 (DBE), the System and Structure Scoping Results are consulted.
Regardless of their applicability to the structure being evaluated, the seven functions are assigned generic ID numbers that can be used with any structure being scoped Therefore, the Structure Intended Functions Table has the same basic format for every structure. The functions that apply to the structure are identified by indicating "YES" in the " Applicable to This Structure?" column of the Structure Intended Functions Table.
4.2.3 Structural Co.iiucr.cr.t Tvoe Listina for the Structure In the structural component scoping task, components that are structural in nature are not uniquely identified during scoping. For example, each wall in the structure is not identified, named, and listed. Rather than using an MEL of named structural components, S
Edomal flooding events were considered during the design process for CCNPP structures. It was determined that a probable meidmum hurricane would cause the worst <.aee flooding conditions at the site. The resulting surge and move action was analyzed as the beeis of plant flood protection, The effects of poossbie wave action were studied using a hydraulic model.
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY the scoping is conducted on a generic listing of structural component types. His generic list was developed by experts in the field of nuclear Class I structures. The generic list started with structural component types contamed in the Containment Industry Technical Report and the Class 1 Structures Industry Technical Report Other structural component types were added to the list to ensure completeness (e.g., The Industry Technical Reports considered only SR functions. Herefore, several fire- and flooding-related component types were not considered in these reports.)
ne evaluator uses this generic component listing and deterndnes which of the component types on the list are actually contained in the structure being scoped. This step is perfvi. .ed by reviewmg plant architectural drawings and identifying the specific structural types. Additionally, any structural component types which are unique to the particular structure being scoped, such as the prestressed tendons in the containment and the sluice gates in the intake structure, are noted. Rese unique structural component types are then added to the list of applicable structural component types. His list serves as the equivalent of an MEL for structural component scoping task.
4.2.4 Structural Co.. ucs.crits Which Contribute to Intended Functions This section describes the step used to detennine which component types of a structure contribute to the intended functions which the structure performs. For every function listed in the Structure Intended Functions Table that has a "YES" in its " Applicable to This Structure?" column, a review is made of the UFSAR, the Q-List Manual, or the System and Structure Scoping Results (including documents referenced by these results). %c component types which contribute to each intended function are recorded on the !
" Structural Components Which Contribute to Intended Functions" table. .
Additionally, the supports for large SR equipment within the structure are identified by i reviewing a listing of the SR equipment installed in the structure that might affect the design of the structure (such as tanks, heat exchangers, or vessels filled with fluid and pumps which require a pedestal as a foundation.). These SR equipment supports are also included in the " Structural Components Which Contribute to Intended Functions" table.
i Q-List documentation and the Flooding Design Guidelines Manual are resiewed to I determine if structural component types in the structure being scoped are relied on to i contribute to the functions of providing structural and/or functional support to NSR l equipment whose failure could prevent satisfactory accomplishment of any of the required SR functions or providing flood protection barriers. If structural component types in the structure being scoped are determined to contribute to these functions, then this information is captured by recording "YES" in the " Applicable to his Structure?" column of the Structural Intended Functions Table. The components that contribute to these functions are then recorded on the " Structural Components Which Contribute to Intended Functions" table, with a reference to the appropriate intended structure function.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY When completed, the ' Structural Components which contribute to Intended Functions" table provides the correlation between component types in the structure and their intended function (s). Each ec nponent type necessary for an intended function is designated as within the scope of LR.
4.3 Commodity Evaluations that Include Scooinn Sections For certain systems or groups of cu.nycaents, an alternate IPA process was chosen to accomplish the same results as the process described in the first six sections of this methodology.- Each of these situations, where cawww3ity approaches were chosen, are shown in Table 4-1, and described in more detail in Section 7 of this methodology. For two of the commodity evaluations, the scoping and Pre-Evaluation tasks are performed using the techniques described in Sections 3 and 4. In the other four cu.n.nodity evaluation processes, the alternate process replaces the component level scoping, pre-evaluation and AMR. Therefore, for the systems covered by these commodity evaluatxms, the description of the component level scoping is included in Section 7.
- TABLE 4-1 Scoping Part of Commodity Evaluation; Commodity Evaluation?-
EPs & Related Equipment No Ils No Cables Yes Cranes and Fuel Handling Equipment Yes Component Supports Yes FP Systems Yes 4.4 Results 4
As a result of the component level scoping task, components are assigned to one of two categories:
(1) those that are within the scope of LR; and (2) those that are not. Only components that are within the scope of LR are included in the IPA process. These components proceed to the Pre-Evaluation task introduced in the next section of this nuhodology.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 5.0 PRE-EVALUATION This section describes the Pre-Evaluation task. The purpose of this task is to determine which plant SCs are " subject to AMR" in the IPA process.
& Pre-Evaluation task is performed on a system-by-system or stmeture-by structure basis (except for equipment covered by the commodity evaluations which replace the entire IPA process, as described in Section 4.3). The description provided in Sections 5.1 through 5.3 of the methodology applies primarily to systems Section 5.4 describes the differences in the process as it is applied to structures.
The input to this task is the results of the component level scoping task, described in Section 4, for the system being evaluated. These results consist of the intended functions of the system or structure being evaluated and a designation of which portions of the system or structure contribute to the intended functions. From these inputs, the criteria in the LR Rule for "SCs subject to AMR" are applied to determme which SCs in the system or structure must be further evaluated for the effects of aging. The SCs or groups of SCs determined not to be subject to AMR require no further evaluation in the IPA process.
The output of the Pre-Evaluation task is the list of SCs which need to be evaluated further for the effects of aging in the AMR task.
The Pre-Evaluation task is governed by (54.21(a)(1) of the LR Rule.
54.21(a)(1) For those systems and structures within the scope of this part, as delineated in 554.4, identify and list those structures and components subject to an AhfR. Structures and components subject to an aging management review shall encompass those structures and components -
(i) That perform an intended function, as described in f54.4 without movingparts or without a change in configuration orproperties. These structures and components include, but are not limited to, pressure retaining boundaries, component supports, reactor coolant pressure boundaries, the reactor vessel, core support structures, containment, seismic category 1 structures, electrical cables and connections, and electrical penetrations, excluding but not limited to, pumps (except casing), valves (except body), motors, batteries, relays, breakers, and transistors; and (ii) That are not subject to periodic replacement based on a qualified life or specified timeperiod.
Figure 5-1 provides a flow chart of the Pre-Evaluation task.
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Pre-Evaluation Process i
Functonal Scoping /
Results --+ For al passive SCs
/ \
For alintended functens SCs of the systems subject to replee-ment at set frequency Yes-or quali6ed hfe<40 years?
Does functon invoke oving parts or change in No con 5guration or proper-ties of y,,}
SCs? **
No further IPA SCs No review required. excluded by LR Rule Yee - >
y, language?
+ u Add SCs to list of No further IPA review passive SCs for the No required for these SCs Add SCs to system list of _
passive,
" long-lived SCs i AN functons complete?
{
7 g
\ / Any passive SCs remaining?
' L Ust of passive SCs for J Pre-eval Comp!ete the sptem 2 ,
1 y
l SCs Subject to AMR l
1 1r System Commoe LRA AMR ^b8 <
Evaluation f '
Figure 5-1 40 Resision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT i INTEGRATED PLANT ASSESSMENT METHODOLOGY l l
5.1 Catemorine Intended System Functions as Active or Passive l l
De first step of the Pre-Evaluation task is to review the list of intended functions for the system ,
being evaluated and characterize each as either active or passive. When a function is detemuned to be passive, all components which contribute to the passive function are categorized as passive j w...ycssts, even though some of these cw..pcssts may also contribute to an active function. If l such components are determmed to be subject to AMR, the subsequent AMR task considers only the effects of aging on the passive intended function to which these components contribute. The components' contribution to active functions need not be considered in this evaluation.
5.1.1 Passive Functions Passive functions are those which require no moving parts or change in SC configuration ,
or properties to carry out the requirements of the function. Such functions generally do not result in plant parameters changing in a measurable manner during normal plant operations. Examples of passive functions are listed below:
> Maintain the pressure-retaining boundary of a fluid system.
> Provide structural support or shelter to equipment.
> Provide missile protection.
> Provide shielding against radiation.
> Provide shielding agamst high energy line breaks.
> Provide flood protection.
> Prevent or isolate faults in an electrical circuit when such protection or isolation does not involve moving parts or a change in properties or configuration.
(e.g., cable insulation).
Any function which is determined to be passive is evaluated in Section 5.2.
5.1.2 Active Functions Active functions require moving parts or a change in SC properties or configuration to carry out the intended function. For such functions, plant parameters change in a ,
I measurable manner during normal plant operation. Performance of this equipment may be assessed by observing, measuring or trending these parameters. Examples of active functions are:
> Provide required flow to a heat exchanger.
> Provide electrical signals to a device.
> Provide electrical power to a bus or load.
> Provide indication of a plant condition. ,
> Remove decay heat.
> Provide fault isolation where moving parts or a change in properties or configuration is involved. (e.g., circuit breakers, fuses) 41 Revision 1
. ATTACHMENT (1) 1 CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Active feians require no further evaluation in the IPA process. Any components which contribute to active intended functions would not be included in the list of SCs subject to AMR, unless warranted by their contribution to other intended functions which are passive.
5.2 Determine Whether Comnonents Are Lonn-Lived or Short-Lived In this step of the Pre-Evaluation task, all passive SCs are reviewed to determine if they are subject to rep 1=~=ent based on qualified life or specified time period. Structures and components which 4 are not subject to such replacement are classified as long-lived.
Rep!=~ aaat programs may be based on vendor rew.o.ocr.d.nons, plant experience, or any means which establish a specific replacement frequency. Often, replacement based on qualified life will also be rep!=cament at a specific time period (i.e., the time period dictated by the qualified life).
However, in some mstances the qualified life of an SC may be based on variables other than calendar time. In either case (calendar time replacement or qualified life replacement), the SCs subject to such replacement would not be included in the list of SCs subject to AMR.
'Ihe remaining components which contribute to the passim function will be subject to AMR unless
~
the cuo.yor. crit type has been specifically excluded from the review by the language of the Rule.
5.3 A: "- =e f oSystem C: __xrds to Commodity Evel=*;eas 1
As discussed in Section 43, there are several categories of equipment which are more efficiently I I
evaluated across system boundaries as members of commodity groups. Commodity groups are components which are present in a number of systems, but which perform the same function regardless of the system to which they are assigned. Commodities such as cables were not scoped as part of a specific system because these components are not assigned to systems in the CCNPP equipment dat=h==a As will be discussed in Section 7 of this methodology, the commodity evaluation for these components covers the entire IPA process, and this pre-evaluation discussion would not apply to such components. For the EP and IL commodities, some or all of the components are assigned equipment identifiers in the CCNPP equipment database. For these cu,upor. cats, the Pre-Evaluation task includes an administrative step to remove these components from the scope of the AMR of the assigned system, and to bin these components for the commodity evaluation of the appropriate commodity group. These two cases are discussed below.
53.1 EP_g Electrical panels are assigned to a number of systems in the CCNPP equipment database because they are functionally related to the system components. In all cases, the passive intendad function of such panels is to provide structural support to active system components contamed in the panel and/or to ensure electrical continuity of power, control or instrumentation signals. Electrical panels include switchboards, motor control centers, control panels and instrumentation panels.
At this point in the Pre-Evaluation task, such panels are excluded from the AMR of their parent system and are instead administratively included with the EPs commodity 42 Revision 1 l
- . _ - 1
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY evaluation As will be described in Section 7 of this methodology, the cc.ii.i.adity l cvaluation produces the same results as the AMR task described in Section 6 but the ;
evaluation is adjusted to be more efficient for a particular component type. I t
t 5.3.2 IL: and Tubina l Many fluid systems contain a number of small ILs which are part of the systems' pressure-retaining boundary. Such small branch lines contribute to the passive intended function of mamtaming the system PB and most are not. subject to periodic replacement. ;
Consequently, these Its are subject to AMR. Instrument lines are subject to common f environments, are made of common materials and perform the same passive intended function regardless of the system to which they are assigned. 'Iherefore, the BGE IPA process identifies such ILs during the Pre-Evaluation task and excludes them from the AMR of the parent system. The commodity evaluation of ILs includes: 1) small bore ;
piping, tubing and fittings from the root isolation valve to the instrument; 2) hand valves - t which are part of the instrument lines (such as equalization, instrument isolation and vent valves for pressure differential transmitters); and 3) any other components in the mstrument line which contribute substantially to maintaining the pressure retaining function of the mstrument line. Section 7.1.2 contains a discussion of how this third ,
criterion for inclusion of comyer.ciits in the IL Commodity Evaluation is applied. ;
5.4 How the Pre-Evaluation Task Anolies to Structures ,
For plant structures, a modified task is used to determine which SCs are subject to AMR.
s 5.4.1 Passive Versus Active Section 4 of the IPA Methodology describes the seven intended structural functions which may cause a structure to be included within the scope of LR per (54.4 of the LR Rule.
From reviewing these functions and the description of passive functions in Section 5.1.1, it i is clear that all of the intended stmetural functions are passive. Therefore, the steps of the !
Pre-Evaluation task to characterize functions as active or passive are not needed for :
structures.
5.4.2 Short-Lived Versus Lona-Lived Plant structural components are not normally subject to periodic replacement programs.
'Iherefore, structural components are considered to be long-lived unless specific justification is provided to the contrary. Suchjustification would be included in the LRA.
5.4.3 Structures Which are Also Designated as Systems In two mstances, plant structures are also characterized as systems in the CCNPP site documentation system and system-type components are associated with these " systems."
For example, the primary containment structure is also designated as the containment system. All penetration seals, as well as several position switches and access doors, are 43 Revision 1
. . - _ - - - . - . . _ _ ~ .. . - . . - - . . .
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY listed as individual wipets of the contaitunent system with unique equipment identifiers.
As discussed in Sectmn 4, the techniques for scoping of a structure as well as those for scoping a system are applied to such a structure. Two distinct sets of scoping results are produced- one for the system components and one for the structural components. In this case, the Pre-Evaluation task described in the previous steps of Section 5 would be applied to the system scoping results. For the structural scoping results, Pre-Evaluation steps would not be performed for the reasons described in Sections 5.4.1 and 5.4.2.
5.5 Pre-Evaluation Results and Documentation .
l
%e Pre-Evaluation task produces results which serve as input to the AMR task and to specific )
commodity evaluatens Dese results and the documentation of the results are discussed below.
5.5.1 Pre-Evaluation Results l Section 5 identifies the SCs which are subject to AMR. This list of SCs and their intended passive functions serve as the input to the AMR task described in Section 6. Section 5 also removes certain passive, long-lived SCs from the scope of their parent system AMR, and includes them instead in the commodity evaluation for a specific commodity type.
5.5.2 Pre-Evaluation Documentation i
De Pre-Evaluation task produces a list of the SCs which are subject to AMR for inclusion in the LRA. .
l l
l l
l l
l l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.o AMR This Section of the IPA Methodology describes how the co.r.ycacnts which were determined in Section 5 to be subject to AMR are evaluated for the effects of age-related degradation. It also describes the approach used to identify and evaluate aging management alternatives to determme which adequately manage the effects of aging. Figure 6-1 is a flow chart which represents the i A M R process j The AMR task fulfills the requirements of 10 CFR 54.21(a)(3)of the LR Rule: l 1
For each structure and component identi)ed in paragraph (a)(1) of this section, demonstrate that the efects of aging will be adequately managed so that the intendedfunction(s) will be maintained consistent with the CLBfor the period of i extended operation.
The input to the AMR task is the list of SCs subject to AMR along with the intended, passive i functions for those SCs. The results of this task demonstrate the following for each input SC or group of SCs:
1
> Management of the effects of aging is not required because these effects are not detrimental to the ability of the SC to perform its intended function consistent with the CLB;
> Existing programs or activities will adequately7manage the effects of aging; or
> New programs or activities or the modifications to existing programs or activities will need to be implemented to adequately manage the effects of aging.
Like the Pre-Evaluation task, the AMR task is usually performed on a system-by-system and structure-by-structure basis. The task described in this Section applies to SCs of both systems and structures with very few exceptions. These exceptions are described in the steps where they occur.
The AMR can be performed in one of two general ways. In some circumstances, it is possible to demonstrate that existing plant programs adequately manage the effects of aging without an explicit evaluation of the aging mechanisms. This approach is described in Section 6.1. In other instances; however, it is most efficient to evaluate the effects of specific aging mechanisms on the intended functions. Section 6.2 describes this approach.
7 See Section 2.1 for the definition of " adequately manage."
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY AMR Process List of pesehe,long-Sved SCs and their Intended functions 6.1.1 & 6.1.2 Can m
'1 sSC that effects o of a complex Yes-+ agin0 are being mng'd Yes-w/o addr: 'g
=
ARDMs?
No 4 No Provide documentshon isSC that effects of aging are Yes " : being adequately LRA
'"
- managed to assure 7 6.1.3 Intended funchons #
No is SC long-Rved EQ7 List of pm
-> ARDWo h vW W4 comtmouons 6.2.1 No 6.2.2
+
+ + ,,
Create potential ARDM List Organize SCe into groups of
/ or F each W \
sh, W;e
.e ARDWs h v@ we combination
+ Input from Site l Create ARDM matrix Assess level of
{ concem and severty of :
/ Foreach ARDW X aging effects i shigw; 6.3 combination 6.2.3 , o isARDM Add ARDW Determine the plausible based on y,, shi.yu,,ers to not of appropriate type of orial, environment plausible ones for the agingira ge.rs _, LRA g, & function? system. based on concems and effects. Document reasons.
f ib
+
Document receons M
f AR N u AN plausible ARDW PI' gpg subcomponents -Yes AR ARDWshiWwa
_ g Complete combmations complete? \ /
\ /
Figure 6-1 l 1
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l ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 4
Where the approach described in Section 6.2 is followed, several alternatives for managing the l aging effects may be viabb and it is necessary to select from those alternatives. In addition, I technological developments may produce additional viable alternatives in the future for either I approach. Section 6.3 describes the CCNPP approach for evaluating and selecting aging management alternatives during the IPA process. l l
6.1- Justification that Effects of Arine are Beine Manneed Without Specifically Evaluatine ARDMs In several instances, a specific evaluation of the ARDMs is not required in order to justify that the effects of agmg are being adequately managed by existing plant programs. Hese approaches are based on the Commission conclusion stated in the SOC accompanying the LR Rule.
As aplant ages, a variety ofaging mechanisms are operative, including erosion, corrosion, wear, thermal and radiation embrittlement, microbiologically induced aging efects, creep, shrinkage, andpossibly others yet to be identifed orfully understood. However, the detrimental efects of aging mechanisms can be observed by detrimental changes in the performance characteristics or condition of systems, structures, and components if they are properly monitored.
(60 FR 22474) ,
Four cases are described in this Section. For three of these cases, the AMR demonstrates that the -
effects of aging on the passive function would be reflected in a change in one or more monitored performance or condition characteristics of the SCs. Therefore, by adequately monitoring these performance or condition characteristics, the effects of aging on the passive intended function are also adequately managed. In the other case, described in Section 6.1.3, the SCs are subject to a TLAA. He resolution of the TLAA will be provided by one of three methods described in Section 8.
l 6.1.1 Complex Assemblies Whose Only Passive Function is Closelv Linked to Active Performance For some complex assemblies of SCs, the principal intended function is an active function.
Some of their components are subject to AMR because the components contribute to a passive pressure-retaining function to support the active functions of the entire assembly.
An example is the diesel generator supporting equipment. The pressure-retaining components of the diesel starting air, lube oil, fuel oil, cooling water and scavenging air system are subject to AMR because they contribute to a passive pressure-retaining function. However, there would be a readily observable affect on the diesel generator performance if the pressure-retaining components deteriorated significantly. For example, significant cooling water or lube oil piping leakage would result in increased bearing temperatures, and significant starting air leakage would affect diesel start times.
Additionally, experience has shown that even minor leakage from any of these supporting >
subsystems is observed by operators conducting routine testing well before they result in actual performance degradation. These effects would be observed during routine testing, before the deterioration of the pressure-retaining components could affect the diesel's i
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i ATTACHMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT i INTEGRATED PLANT ASSESSMENT METHODOLOGY 4
a ability to perform its active intended function Corrective actions to restore the passive
- nW from its degraded condition are required by the performance testing program and by the normal site corrective action processes.
Because of the readily observab.e effects of passive function degradation on active performance, a sufficient method of managing the effects of all types of aging could be to subject the assembly of components to a rigorous performance and condition monitoring program In the cited example, the diesel generator support systems are subject to ' ,
surveillance requirements to demonstrate operability in accordance with the Technical Specifications and to a comprehensive reliability program required by other regulations.
The conclusion of the AMR using this technique could be that continuing these types of '
performance and condition monitoring programs would ensure that the intended functions of the assembly will be adequately managed.
In some cases, the conclusion of tl:e MIR using this approach may be that the discovery techniques available through the performance and condition monitoring programs are not -
timely enough to ensure intended functions as required by the CLB. For example, the discovery techniques used in a particular performance and condition monitoring program may only provide reasonable assurance that the intended function can be performed under normal loadmg conditions. Additional evaluation and/or inspection may be required to ensure the ability to perform intended functions under certain more severe loading conditions which are part of the CCNPP CLB In this case, additional evaluations may be performed to demonstrate that the aging mechanisms which may affect the ability of SCs to perform under more severe loadmg conditions are not plausible for the SCs.
Alternately, age-related degradation inspections, as described in Section 6.3.3.4, may be performed to determine whether there are aging effects of concern for the SCs being evaluated.
Because there may not generally be a close tie between degradation of passive SCs and the active performance of a train of equipment, the performance and condition monitoring AMR technique is used only in selected circumstances. The conditions listed below represent the circumstances where this approach should be followed rather than using one of the other AMR approaches. These conditions do not constitute a part of the AMR demonstration itself. De demonstration that these conditions are met would not be submitted as part of the LRA but would be maintained onsite.
> A complex assembly of components where the pressure-retaining function directly supports active performance of the assembly;
> The passive function is the pressure-retaining function and is not a fission product boundary function;
> The active intended functions are performed by redundant trains;
> Performance testing is well documented with verification that corrective actions assure the continued performance of all intended active functions; and 48 Revision 1
! ATTACHMENT (1) >
i CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY j > The complex assembly is covered by the Maintenance Rule.
J i 6.1.2 Component Assemblies Subiect to Comolete Refurbishment a For some complex assemblies of SCs, the entire assembly is subject to a program which requires complete refurbishment at periodic intervals. Components of such assemblies may be subject to AMR because their pressure-retaining function supports the active functions of the entire assembly. Deterioration of the pressure-retaining components j would be discovered and corrected during the refurbishment activities before the
- deterioration could affect the intended function of the assembly in a manner not consistent l with the CLB.
4 An example is the main steam isolation valve operator. This assembly contributes j primarily to the active function of closing the main steam isolation valve in a specified j amount of time. Because the valve operator uses a combination of hydraulic fluid pressure
- and compressed nitrogen to operate the valve, several components of this operator i assembly provide a passive pressure-retaining function. The entire valve operator is j
removed from the system at regular intervals and refurbished. Some of the pressure-retaming components and subcomponents are replaced every refurbishment interval.
Others are inspected and replaced if they meet certain described conditions. The entire
, assembly is re-assembled and tested to ensure satisfactory performance and then re-
) installed in the system. Such a refurbishment program manages all plausible aging effects
! to ensure that the intended function of the valve operator is maintained in accordance with the CLB. Therefore, this program may be credited as an adequate aging management i program without considering specific aging mechanisms. l t
i This approach is restricted to refurbishment programs that meet the following criteria:
]
I
- > The refurbishment is conducted at regular intervals on a complex assembly of '
components where the pressure-retaining function only directly supports the active intended function of the assembly;
{
i )
> The passive function is the pressure-retaining function and is not a fission product '
i boundary function; 4
4
> The program requires complete removal of the component assembly from the system;
> The assembly components and subcomponents, including pressure boundaries, are 3
inspected for signs of aging and other degraded conditions;
> The refurbishment directs replacement of components and subcomponents that are j deteriorated excessively due to aging or other degradation; and f > The refurbishment includes post maintenance testing consistent with current ;
i industry practices and the CLB.
a i 49 Revision 1 J
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.1.3 Lonn-Lived EO Components Components subject to EQ which have qualified lives less than 40 years are short-lived and would be excluded from the AMR during the Pre-Evaluation task. Components subject to EQ which have qualified lives of 40 years or greater are subject to a TLAA.
The options for resoMng TLAAs are described in Section 8. Completing one of these TLAA options for long-lived EQ equipment will also serve to provide the required IPA demonstration.
Some portions of passive EQ SCs may not be covered by the EQ program For example, the EQ program only qualifies the organic material of a solenoid valve. A separate AMR evaluation using the technique described in Section 6.2, will be performed to provide the required demonstration for those portions of passive EQ SCs which are not covered by the EQ program.
6.1.4 SCs Subiect to Reolacement on Condition In the case of certain SCs, an indication of SC condition is used as the basis for replacement of a passive SC. For example, the copper-nickel tubes of a heat exchanger may have an intended pressure-retaining function. This function is passive since there are no moving parts or changes in configuration or properties involved in performing the function. Such tubes are not replaced based on a specific time period or qualified life so they would be included in the AMR. However, they are subject to eddy cur ent testing which dictates when tubes must be plugged and a tube plugging limit which mcutes when the tube bundle must be replaced. Plant experience shows that these heat exchangers are retubed every 10 to 15 years. In cases such as this one, where a plant parameter for a passive SC is linked to the ability of the SC to perform its intended function, and where plant operating experience has shown that the component is replaced frequently, the condition-based replacement program would be credited as the aging management program for the SCs.
Table 6-1 shows the criteria which are covered in the detailed demonstration for each SC or group of SCs subject to this AMR method. These detailed results are maintained onsite in an auditable format. The justification provided in the LRA to demonstrate that the effects of aging are adequately managed would include a summary of the detailed justification.
50 Resision 1
ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT j INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 6-1 CRITERIA FOR REPLACEMENT ON CONDITION PROGRAMS Criterion 1 - Replacement programs based on condition or performance must ensure that the SCs identified as within the scope of LR will be replaced before degradation would result in loss of the SC intended function (s). For example -
> Is the discovery activity frequency interval less than the shortest time between failures of the SC intended function (s)?
> Based on the condition or performance trait monitored by this program, is the component replaced at intervals that are short relative to the life of the plant?
> Historically, have all maintenance preventable functional failures of SC interxled functions been detected by the activity?
Criterion 2- Replacement programs based on condition or performance must contain appropriate acceptance criteria which ensure timely replacement of the SCs.
> Does the activity have an action or alert value or condition parameter to determine the need for replacement of the SC7
> Does the action value or condition provide an appropriate means of assuring replacement of the component before the effects of aging would prevent any intended system functions?
Criterion 3 - Replacement programs based on condition or performance must be implemented by the facility operating procedures.
> Is the activity controlled by a site review process which includes controls over subsequent revisions?
6.2 Performine an AMR by Evaluatine Arine Mechanisms In some circumstances, the most efficient manner 8 to show that the effects of aging are being adequately managed is to evaluate the effects of specific aging mechanisms on the intended functions and to demonstrate that those effects are being managed. This Section describes this method of performing an AMR.
8 Unlike the methods described in Subsection 6.1, this method of performing the AMR could have been used for all SCs subject to AMR. However, this method is not always the most efficient method. For some SCs, even if one of the more efficient methods described in Subsection 6.1 would have been sufficient to demonstrate adequate aging management, BGE chose to we = more mechanistic approach due to other benefits derived from performing this approach.
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 6.2.1 Creating a Potential ARDM List The first step of the specific evaluation of ARDMs is to determine which ARDMs must be evaluated. For system components, the list of such ARDMs is referred to as the " Potential ARDM List" for a given ET.
When an ET is encountered in an aging evaluation and the ET has not beca evaluated as part of a previous evaluation, a new Potential ARDM List is created. Industry documents are reviewed to identify the aging mechanisms which need to be considered. From reference materials, a list of all of the ARDMs which might affect any SC of the given ET is compiicd. The list also includes a discussion of the various stressors which cause or exacerbate the ARDMs. It also includes a list of any characteristics of selected SCs which might prevent the ARDMs. This Potential ARDM List is the list of ARDMs that will be considered for subsequent evalua' ions of SCs of this ET. The Potential ARDM List is updated as each SC of the same ET is evaluated.
The next step is to eliminate those ARDMs which are not applicable to any of the SCs in the system being evaluated. For example, creep is an ARDM which is included on the initial list for the ET for piping. However, when finalizing the Potential ARDM List for the Senice Water System, this ARDM is eliminated as not applicable because the temperatures throughout the Senice Water System are too low to warrant consideration of this mechanism. 'Ihe basis for marking an ARDM as not potential is recorded on the Potential ARDM List for the system.
Structural components are not associated with a particular ET in the site equipment database, and therefore a modification to this step is needed for structural components.
Instead of creating the Potential ARDM List for each ET, structural component types aredivided into two categories: 1) concretc/ architectural components; and 2) steel components; and a Potential ARDM List is created for each of these categories.
6.2.2 SC Grouoina If a system contains several SCs with similar characteristics, the evaluation can be made more efficient by grouping these SCs together for a common evaluation.
All components of systems are classified in the site equipment database with a particular DT code. Examples of such DTs are hand valves, check valves, pressure transmitters and heat exchangers. The DT can be further divided to facilitate the evaluation. For example, if the check valves of a particular system are made of two distinctly different materials, two separate groups may be formed. Other possible examples are listed below:
Internal Environment - All system piping which carries saltwater could be in one group while the instrument air piping which controls valves in the system would be in another.
External Environment - All system underground piping could be included in one group, while the above ground piping would be in another.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
')gign - Other design parameters besides material could be selected as grouping attributes. For example, plate and frame heat exchangers may be grouped separately from shell and tube heat exchangers.
The grouping attributes and the component ids are recorded and each group is assigned a unique identifier.
Groups may be further subdivided into the individual subcomponents which make up the components in the group if this facilitates the subsequent evaluation. If certain subcomponents are not required for the SC to perform its intended, passive function, they are identified and excluded from further evaluation. For example, a group of air-operated valves may have an intended pressure-retammg function but may not have to reposition for any intended function. Therefore, the discs, seats and air operators of the valves in this group would not be subject to AMR because they do not contribute to an intended passive function. Whenever subcomponents are eliminated from further evaluation because they do not contribute to the intended, passive functions, the bases for these decisions are also documented.
Agam, because of site documentation differences for structural components, the structural component type is used to establish the initial level of grouping in the same manner as DT :
is used for system components.
6.2.3 Create and Resolve the ARDM Matrix.
After completion of the system Potential ARDM List and after SCs are grouped ar.d subdivided, an ARDM matrix is created and evaluated. He ARDM matrix consists of all potential ARDMs along one axis and all remaining subcomponents for a particular SC group along the other. Each ARDM/subcomponent intersection must be resiewed during this step.
For each ARDM/subcomponent combination, the following is considered: 1) the material of the subcomponents in the group; 2) the operating environment; and 3) the passive intended functions. If the ARDM does not affect the material, is not perpetuated by the environment or occurs to such a small degree that the intended function is maintained, the ARDM is designated as not plausible far the subcomponent. Although material, environment and function are mentioned separately above, when evaluating ARDM plausibility, all of the factors are considered together.
Integrated Plant Assessment documentation for this step consists of the list of the ARDMs that are plausible for each group of SCs subject to AMR and the rationale for designating each ARDM. His information is recorded in evaluation reports and maintained onsite. A list of the potential ARDMs that were evaluated for each group of SCs in the system is provided in the LRA.
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ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY j 6.3 Methods to Manare the Effects of Arina This Section describes how the aging management methods are chosen and justified for the period of extended operations. Methods chosen for managing the effects of aging will be consistent with site strategies for maintenance of equipment material condition. One of the goals of aging management is to manage the effects of aging such that the intended functions are maintained i consistent with the CLB. Consequently, each phase of the maintenance strategy discussed below takes this goal into consideration when determining the adequacy of an existing or proposed '
program or actisity.
6.3.1 Phases of a Maintenance Stratery An adequate maintenance strategy consists of four phases: Discovery, Assessment / ;
Analysis, Corrective Action, and Confirmation / Documentation (1) Discovery - The first phase of a maintenance strategy is identification that detnmental effects of aging are or could be occurring. As stated in the SOC for the LR Rule:
The Commission believes that, regardless of the specific aging mechanisms, only age-related degradation that leads to degraded performance or condition (i.e. detrimental efects) during the period of extended operation is ofprincipal concernfor license renewal. Because the detrimental efects ofaging are manifested in degradedperformance or condition, an appropriate license renewal review would ensure that licensee programs adequately monitor performance or condition in a manner that allowsfor timely identification and correction ofdegraded conditions. (60 FR 22169) i Aging can be self-revealing or identified through specific diagnostic techniques. j Current examples of discovery methods include visual observation of external I conditions, eddy current examination for flaws, and ultrasonic testing for detecting wall thinning. As discussed in Section 6.1.1, these discovery methods may require augmentation for LR to ensure that the effects of aging are discovered in a timely manner such that there is reasonable assurance that the CLB will be maintained.
Some plant programs may use specific detection techniques to detect and monitor aging while others rely on walkdowns by plant personnel to observe and document degraded conditions or performance. Monitoring and evaluating industry experience also serves as a discovery activity for currently unknown or theorized aging mechanisms since other plants may discover aging effects before CCNPP.
(2) Assessment / Analysis - Once performance or condition degradation is discovered, its progress must be compared to criteria or other guidance to determine the degree of the degradation and the need for specific and generic corrective and preventive action. 'Ihese criteria and guidance will depend on the characteristics of the degradation and the effects on the intended function. For example, a safety or ;
safety support system must be capable of performing its specific safety function I i
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for accident prevention and/or mitigation as described in the CLB. Likewise, a i system provuhng a function for a regulated event must be capable of perfornung l that function under the conditions described in the CLB cvaluation of the regulated l event. The as===aat/ analysis phase incorporates such requirements in A.. .g the need for and nature of corrective actions after abnonnal or !
degraded conditions are discovered. One possible result of such l
=====~*/ analysis would be to repeat the discovery phase using an expanded l sar :le size or using an augme=*~i or improved technique for discovering and quantifying the extent of a particular aging effect. j (3) Corrective Action - With the degree of degradation known, specific corrective -l action can be taken to ensure that the equipment performance or condition is i restored and the intended function is maintained Site procedures currently exist i which require root cause analysis and actions to prevent recurrence to be included i with corrective actions when appropriate.
(4) Confirmation / Documentation - After the corrective action is performed, post-maintenance verification or testing confirms that maintenance was performed correctly and the equipment is capable of performing its intended function. The !
corrective action and testing are documented as part of plant records for future reference. l i
in combination, these four phases provide a complete maintenance strategy. Sections 6.3.2 :
and 6.3.3 describe how discovery activities are identified and selected. Section 6.3.4 describes how the latter 3 phases are implemented. i 6.3.2 Site Expert PanelInnut t
1he selection of the appropriate method for detecting aging effects is performed through an expert panel review of each plausible ARDM/ subgroup combination. The review is .
conducted on a system or ccinirulity basis and, typically, consists of following plant !
representatives-1
> The system or commodity aging evaluation engineer;
> The cognizant system engineer;
> Appropriate plant program managers / technical area specialists; and
> The aging management implementation engineer.
Each member brings specific focus and talent to the expert panel.
The aging evaluation engineer presents the results of the system aging evaluations highlighting the intended functions of the systems, the components subject to AMR, and the plausible aging effects. The aging evaluation engineer also proposes the methods by which the effects of aging can be managed.
i The system engineer brings his knowledge of the system and functional requirements, knowledge of the plant and industry experience with the system, and familiarity with 55 Revision 1
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY I system inspecten, surveillance, testing and maintenance results. He system engineer also provides site technical concurrence to execute the aging management methods for his system under a renewed license.
Each plant program manager / technical a a specialist brings his expertise in a specialized area (such as non-destructive examination, EQ, chunistry, materials, fatigue) and provides a perspective in determinati >n of program applicability and feasibility. These individuals l also provide technical concurrence that their program methods will effectively detect and !
monitor the specified aging effects and are presently the preferred methods. !
De aging management implementation engineer facilitates the panel meetmgs, provides l consistency between system and commodity technical discussions, ensures involvement of j the appropriate plant personnel, and ensures closure of open items.
He panel as a team determines the appropriate methods to manage the effects of aging for ,
the given system or commodity considering two main factors:
> The likelihood the ARDM will occur for the specific application; and
> How the effects of the mechanism progress.
If the panel detemunes that the ARDM occurs and progresses relatively rapidly, tlx:n prescriptive plant programs or system modifications may be warranted. Age-related degradation inspections and/or performance or condition monitoring may be warranted if: I
> The mechanism has not been seen yet in operating plants;
> Present knowledge indicates progression is gradual; and
> The known characteristics of the ARDM indicate a potentially severe impact on the system intended function.
Continuing to monitor and evaluate industry experience may be appropriate if:
> There is little or no experience with a particular mechanism occurring for the system environment;
> Current knowledge indicates the ARDM progresses relatively slowly; and
> Re potential consequences to the system intended function are not significant.
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INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.3.3 kl~ *iaa of Aaian h'---- Alternatives for Discoverv )
1 Once degradation is discovered, the step described in Section 6.3.4 will ensuit that the I appropriate As=- ==dAnalysis, Corrective Action, and Confirmation / Documentation i occur for all SCs. Therefore, for the purposes of the IPA, it is only necessary to establish how the degradation will be discovered on a systen'.-by-system basis.
Appropriate methods for discovering the effects of aging are selected for all of the SCs subject to the AMR based on the expcit panel approach. Each of the methods can be categorized into one of the following groups.
6.3.3.1 Plant Proarams l 1
Plant programs are often the most direct and systematic method of detecting and mitigating l the effects of aging. 'Ihey already exist to meet regulatory requirements or i is <.. == Maas, warranty requirements, or to preserve economic investment based onsite experience. They are typically selected as the method of discovering aging when they exist and can discover the effects of the plausible mechanism.
The plant programs applicable to the system are identified and reviewed to determme if they may serve to discover aging effects for the long lived passive eg = =. In some cases, existing condition monitoring or functional testing may be sufficient; existing l focused inspections may be sufficient in others. Programs adequate to detect or monitor the effects of aging during the period of extended operations are credited without modification Whenever an activity required by an existing industry code such as ASME Section XI is credited as an aging management program, the specific version of the code to which BGE is currently committed should be noted in the AMR report and LRA documentation.
l Existing plant programs can also be modified to ensure the discovery phase of the !
maintenance strategy is adequate for the period of extended operation. Examples of !
modifications to an existing program include, but are not limited to, the following: l a
> Adding components to inspection procedures for specific aging effects; l
> Adding specific aging effects mitigation procedures; and
> Tailoring of record keeping and trending requirements.
If no existing plant program can be adapted to address the aging effects for the given group of SCs, new programs may need to be implemented.
Some modifications to existing programs and new programs may be implemented prior to submittal or approval of the LRA. Alternately, the LRA may include a commitment to implement the program or modification at an appropriate future date before or, with appropriate justification, during the period of extended operation. 1 Examples of existing plant programs are shown in Table 6-1.
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TABLE 6-1 Examples of Existing Plant Programs Maintenance (Preventive) Materials Testing and Evaluetion Maintenance (Corrective) Motor-Operated Valve Program Maintenance Standards Program Performance Evaluation Program Check Valve Reliability Performance Evaluation Program (Operations)
Eddy Current Testing Plant Lay-up and Equipment Presenation Electronic Cable Degradation Post-Maintenance Testing -
Engineering Test Procedures Pressure Test Procedures Surveillance Test Procedures Plant Tours 1 Fatigue Monitoring Protective Coating and Painting )
Fu*1 Testmg System Walkdowns j Environmental Qualification Thermography l Inservice Inspection Vibration Monitoring J Loose Parts Monitoring Thermal Performance Monitoring Lube Oil Analysis Operator Rounds 6.3.3.2 Site Issue Reportina (IR) and Corrective Action Program In cases where the effects of aging are observed in less formal activities or as a result of work in the vicinity, the IR and corrective action program is relied on for discovery.
Examples ofless formal activities are:
> Plant tours by supervisors and managers; ;
> Management and supervisoryjob obsenstions;
> Maintenance planning walkdowns; 4
> Walkdowns of planned and completed nedifications;
> Fir:: watches; and
> Personnel safety equipment inspections.
Any observed or suspected condition that requires significant corrective action, whether related to the purpose of a specific activity being performed or not, is documented via an IR. These methods for discovery are normally complementary to other, mo e formal activities, such as age-related degradation inspections. If such activities are relied on as ,
the principal means of discovery, appropriate justification would be prosided in the LRA. I 6.3.3.3 Plant Modifications
)
Plant modifications may be appropriate where:
> Plant programs cannot effectively discover the effects of aging
> Experience indicates that the mechanism is occurring; and
> The progression is relatively rapid. i 58 Revision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Modifications will occur as part of the normal site modification process which currently exists for improving and updating plant response, performance and reliability.
Examples of modifications which might result from the aging evaluations include, but are not limited to, the following:
> Relocation of equipment to a less aggressive emironment;
> Change of material to improve resistance to the aging mechanism; and
> Change in the equipment operation.
Modifications to plant equipment may be implemented prior to submittal of the LRA.
Alternately, the LRA may commit to implement a modification at an appropriate future date. Withjustification, this date may be during the period of extended operations.
6.3.3.4 Ane-Related Dearadation Inspections Two distinct cases of age-related degradation inspections are discussed below. Others may also be possible.
Case 1: Inspection to Support a Non-Plausible Determination In some cases aging mechanisms are possible but the effects of the aging are expected to have minimal consequences due to the equipment material and operating conditions. For example:
> A structure may have been built with a concrete mix that prosides maximum resistance to freeze-thaw,
> A tank may have been built of stainless steel using strict welding controls to minimize the chance of stress corrosion cracking.
In this case, an inspection could be conducted to provide additional assurance that significant degradation is not occurring or that the rate is sufficiently slow to preclude concern during the period of extended operation. Alternatively, the inspection might conclude that additional inspections are needed during the period of extended operation.
The scope of such inspections would typically be a representative sample of the population. Where practicable and pmdent, the sample vould be biased to focus on bounding or leading components. For example:
> The portion of a structure more likely to experience the ARDM; or
> A statistically representative sample of the valves made of a particular material; If the inspection indicates little or no degradation, the conclusion could be reached that the degradation will not result in loss of component function during the period of extended operation, and therefore, no additional aging management activities or programs would be 59 Revision 1
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required. Significant degradation, on the other hand, would trigger action under the existing corrective action program and the need for additional inspections would be evaluated.
Where the inspection demonstrates that there is no significant degradation and no program is needed to manage the effects of aging, resolution of the aging mechanism would be documented by & scribing:
! > The inspection process and results; and
> Why it is an adequate approach to disposition the ARDM for the SC group.
Case 2: Inspection to Validate an ARDM Mitigation Program In other cases, programs may be in place which prevent or mitigate the effects of aging.
Dese aging effects could, if left unmanaged, degrade the capability of SCs to perform their passive intended functions. In these cases, relying upon the mitigation program may )
not provide the necessary level of assurance that the passive intended function will be maintained during the period of extended operation. For example:
> An underground piping system may be wrapped with a protective material to prevent contact with moisture and may also be subject to 7n impressed current cathodic protection system designed to prevent corrosion. However, because the i piping is buried and the consequences of failure would be significant, a decision might be made to perform an inspection of a representative sample of the piping !
exterior to confirm that the mitigation measures have been effective in controlling aging.
> A fluid system may be subject to chemistry controls which minimize impurities and maintain a basic pH to limit corrosion of carbon steel components. However, because of the large amount of piping and other components subject to such treatment throughout the plant and the range of environmental factors, an I inspection of a representative sample of components could be conducted to I confirm that the chemistry controls in place have been effective in controlling the i effects of aging. !
In these cases, inspections could be conducted to confirm that the mitigation programs are effective in preventing or mitigating the aging effects which they were designed to control.
Again, the scope of such inspections would typically be a representative sample of the population of components of concern. Where practicable and prudent, the sample would be biased to focus on bounding or leading components. For example:
> The underground piping system which is closest to the water table and therefore, most likely to have been subjected to moisture; and 60 Resision 1
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> . The piping system which has experienced the worst history of chemistry transients ,
and/or has the most susceptible locations.
If these inspections reveal little or no degradation, the conclusion could be reached that the mitigation programs are sufficient to manage the effects of aging during the period of ,
extended operatons. Significant degradation, on the other hand, would trigger action under ;
the existing corrective action program and the need for additional inspections would be [
evaluated.
l Where the inspection demonstrates there is no significant degradation and the existing !
program is adequate to manage the effects of aging, this would be documented by ,
describing:
> The attributes of the program which prevents or mitigates the aging effect; and i
> The inspection process and results.
For both of the cases described above, the inspection technique would need to be capable ,
of detectmg the effects of aging identified by the AMR. Acceptance criteria for these i inspections would be consistent with current practices which acccant for the SC's ability ,
to perform intended functions in accordance with the CLB. )
For both cases, the inspections described above may be completed before submittal of the l LRA. When such an early inspection detects no signs of significant aging as expected, {
there is no need to extrapolate the results of the inspection. If, on the other hand, the l inspection reveals significant degradation or unexpected conditions, the results would either be conservatively extrapolated through the end of the period of extended operation or future ia==~@s would be conducted to track the progress of the unexpected degradation.
The frequency of such future inspections would be commensurate with the safety significance of the SCs being inspected, as well as consistent with the results discovered during the initialinspection.
Altemately, the LRA may commit to conduct the inspection prior to the period of extended operation or, with justification, during the period of extended operation. If industry ;
experience resolves the aging issue in the interim, the commitment to perfonn the j inspection could be cancelled using existing site commitment management procedures. i I
6.3.3.5 Industry Operatina Experie_ng Monitoring plant and industry experience provides the principal discovery means for unknown and theorized aging mechanisms. Additionally, monitoring industry experience !'
may be included as one feature of a multi-feature aging management approach when appropriate.
The materials used at CCNPP are common to nuclear plants and to many non-nuclear power plants that have long operating histories. Monitoring plant and industry experience therefore provides timely information related to unknown and theorized ARDMs, so that 61 Revision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY there is reasonable assurance that such ARDMs would be discovered before they severely
' affect intended functons at CCNPP. It also provides assurance that appropriate changes are made to existing programs.
Industry information is distributed across the nuclear industry sia Institute of Nuclear Power Operation's Significant Event Evaluation Information Network program, which is a small part of Industry's response to NUREG-0737. He plant program for industry experience reviews problems and events across the industry and evaluates the significance ,
and applicability to CCNPP. ;
Examples ofinformation that the program captures are:
> Part 21 Notices;
> NRC Bulletins;
> NRC Information Notices;
> NRC Generic Letters;
> Vendor Information Letters; ,
> Operatmg Experience Information; '
> Significant Event Reports;
> Operations and Maintenance Remmders; and
> Significant Operating Experience Reports. ,
In some cases, the aging evaluation may be based on information from the nuclear power industry or other industries that indicates unexpected deterioration may occur. Although the aging effects may not have been detected at CCNPP or most other plants with similar equipment, similarities in materials and environments may make it possible for the aging effects to occur at Calvert Cliffs. In these cases, discovery has already occurred through ;
notification from NRC, Nuclear Energy Institute, Institute of Nuclear Power Operations, Owners Groups, or vendors.
The site issue reporting and corrective action process requires review and evaluation of the industry experience, and comparison to conditions at CCNPP to determine if additional action is needed here. If resolution of the issue is in progress, it will not necessarily be completed prior to LRA submittal or approval. The site issue reporting and corrective action process ensures that assessment / analysis occurs and appropriate action is taken.
For example, PWSCC of Alloy 600 nozzles was an unknown and theorized aging mechanism. As issues related to it were emerging at CCNPP, in 1989, BGE became involved in industry and owner's groups efforts to resolve Alloy 600 issues. Now that it is a current issue, BGE will propose a specific aging management program for Alloy 600 in the renewal application. However, BGE is continuing to follow industry developments of Alloy 600 management program that results in improvements in existing activities at CCNPP.
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INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.3.4 Isaala=a=*3= the A =="- =^/ Analysis. Corrective Adina and Confirm =*inal Documentation Phases of the Mamtenance Strategv 1
"Ihe last three phases of the maintenance strategy are required by the CLB and are l provided by the site IR and corrective action process. Any observed or suspected l condition that requires signincant corrective action, whether related to the purpose of the I specific activity being perferi..cd or not, is documented via an IR. Initiation of an IR causes the degraded condition or performance to be evaluated for immediate personnel or ;
nuclear safety concerns, operability concerns, and reportability. The IR is screened and _j classified to ensure that timely corrective action is taken. '
Actions necessary to resolve the IR are assigned to the responsible organization. The IR remams open until appropriate actions have been completed and documented. For significant events and issues, an event investigation and root cause analysis is conducted to :
aid in preventing reoccurrence Therefore there is reasonable assurance that timely discovery of aging issues and effects will result in appropriate action to evaluate, correct, document, and report them.
6.3.5 Agmg Management _for Anina Issues Au=M='ad with a Generic Safety Issue (GSI) or Unresolved Safety Issue l
If there is an outstandmg generic issue (GSI or Unresolved Safety Issue) associated with j an identified aging effect or aging management practice, the SOC to the Rule (60 FR 22484) provides three options: I) If the issue is resolved before LRA submittal, the applicant can incorporate the resolution into the LRA. 2) An applicant can justify that the CLB will be maintamed until a point in time when one or more reasonable options would be available to adequately manage the effects of aging. (For this alternative, the applicant would have to describe how the CLB would be maintamed until the chosen point in time and generally describe the options available in the future.) 3) An applicant could develop a plant-specific program that incorporates a resolution to the aging issue.
In determmmg the appropriate aging management practice for SCs affected by GSIs and Unresolved Safety Issues, these options should be considered throughout the steps of Section 6.3 and one of the options chosen as appropriate.
For example, the effects of a particular aging mechanism on a specific material may be designated by the NRC as a GSI. Baltimore Gas and Electric Company may choose option 2) above to address this issue in the IPA. Analysis could be used to demonstrate that other plants are more susceptible to the particular aging effects than CCNPP. Based on this analysis, reliance on continued participation in owner's group activities or other industry activities, including review of inspection results from the more limiting plants, could be used to demonstrate that the SC intended functions will be maintained consistent with the CLB Alternate actions could also be developed as contingencies, depending on the results discovered at the limiting plants. In this manner, the aging issue associated with the GSI could be managed for the purposes of the IPA. Ultimately, resolution of the GSI 1 would include actions, if necessary, which would be implemented under the CLB.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.4 Plant Pronram Documentation Documentation in the LRA for this task consists of a demonstration that the effects of aging are adequately managed as well as a description of the programs and activities which were identified during the AMR and are relied upon to manage the effects of aging. Program modifications or new programs which need to be implemented in order to adequately nanage the effects of aging for the period of extended operation would be described briefly. A summary description of these existing programs and activities, program modifications and new programs are included in the FSAR Supplement. Detailed justification of the adequacy of the programs will be maintained onsite to serve as the basis for the demonstration provided in the LRA and the summary description provided in the FSAR Supplement.
6.5 IPA Summary The completion of the AMR task concludes the IPA required by the LR Rule. The IPA process demonstrates that the effects of aging have been identified and are being or will be adequately managed. He next section of this methodology describes several specific cases where a slightly different process is used to provide the demonstration required for the IPA.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.0 COMMODITY APPROACHES TO AMR As discussed briefly in Section I and 4 of this methodology, the approach described in the first six sections of the methodology was followed for all plant SSCs with only a few exceptions. Rese six exceptions are described in this section.
The intent of a commodity evaluation is identical to the normal IPA approach; i.e., to demonstrate that the effects of aging are adequately managed. For each case discussed in this section, increased efficiency was the primary motivation in adopting an alternate approach.
For the purposes of discussion, the six commodity evaluations are divided into two groups: 1) those that replace only the AMR task of the IPA (Section 7.1); and 2) those that replace the entire IPA process (Section 7.2). Table 7-1 shows the six commodity evaluations and which belong to each of the categories described above.
TABLE 7-1 Commodity Evaluation Equivalent to Entire IPA or Just AMR?-
EPs AMR ILs AMR Cables IPA Crancs and Fuel Handling Equipment IPA Component Supports IPA FP Equipment IPA l
l 7.1 Commodity Evaluations Which Cover Only the AMR Task For the EPs evaluation and the ILs evaluation, the tasks of system level scoping, component level scoping and pre-evaluation are performed as described in Sections 3,4 and 5, respectively. The output of these tasks for the many systems which contain one of these two commodities is a list of the SCs subject to AMR. The performance of the AMR is split into the system AMR and commodity AMRs. The system AMR is conducted as described in Section 6. The commodity AMRs are conducted as described below.
I EP Commodity Evaluation 7.1.1 )
For many fluid systems, the list of SCs subject to AMR includes many pressure-retaining fluid system components and a relatively few EPs which provide structural support to active electrical equipment. All of these components could have been evaluated as part of the system AMR. However, the expertise of the evaluator and the type of reference materials and plant documentation needed to perform the AMR for these two types of equipment is substantially different. Furthermore, the AMR of the EPs requires a level of expertise, reference material and plant documentation similar to that needed for other SCs in electrical distribution and instrumentation systems. Therefore, for efliciency reasons, 65 Revision 1
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INTEGRATED PLANT ASSESSMENT METHODOLOGY the EPs are removed from the scope of each system AMR and all EPs (electrical
- distribution, instrumentation and panels supporting mechanical system operation) are grouped into a common commodity evaluation.
'Ihe first step of the EP commodity evaluation is to review the scope of all of the pre- ,
evaluation results and to include all EPs subject to AMR in the comnedity evaluation, regardless of the system the panel is assigned to in the site equipment technical database Performmg this step maintains the link between the scoping and pre evaluation results, which are done system-by-system, and the scope of the commodity evaluation. For some systems, the only components in the system which were subject to AMR were those included in the scope of the EP a- - vHty evaluation. For these systems, no system AMR was performed at all since the EP commodity evaluation addressed all system components requiring an AMR.
After the scope of the co.. ..cdity evaluation is established, the IPA process for conducting an AMR described in Section 6.2 is applied to the newly formed scope of EPs in exactly -
the same manner as it is applied to a plant system. Panels are grouped by common material, function and environment. Potential ARDMs are listed. Age-related degradation mechanisms matrices are created and resolved, and aging management alternatives are evaluated.
7.1.2 IL Commodity Evaluation For many fluid systems, the list of SCs subject to AMR includes many pressure-retaining
)
components which are part of small branch ILs. Regardless of which system these ILs are part of, certain common characteristics are shared with respect to aging management.
l
> All consist of piping and/or tubing which contribute to only one passive intended function, i.e., the pressure-retaining boundary of the system;
> All include instrumentation which would be affected to some extent by significant PB leakage; and
> All system piping to which these Its are attached is also subject to AMR. i I
Because of these common characteristics, the BGE IPA process includes an IL commodity. j 1
Again, the scoping and Pre-Evaluation tasks of the IPA are performed using the IPA !
approach described in Sections 3 - 5. During the Pre-Evaluation task, the IL components are separated from the remainder of the system pressure-retaining boundary and are targeted for a commodity evaluation. Similar to the EP commodity evaluation, the first step of the IL commodity evaluation specifies the scope of the evaluation. For every fluid l system subject to AMR, pre-evaluation results are reviewed. Tubing, fittings, hand valves and any other in-line components which are associated with the instrument and contribute substantially to the pressure-retaining function are included in the scope of this commodity evaluation. A determination has been made in 10 CFR 54.21(a)(1) that certain component types should be excluded from the AMR. Those specifically listed in 10 CFR Part 54 (as 66 Resision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY being excluded from the AMR) include pressure transmitters, pressure indicators and water level transmitters. Based on this guidance in the LR Rule, the contribution of these sm.gorsts to the passive, pressure-retauung function is determined not to be substantial enough to warrant an AMR, and these components are not included in the IL commodity evaluaten Other corspersts with the same characteristics as those listed in l
{54.21(a)(1), but not specifically listed in this section of the Rule (e.g., differential l pressure transmitters and indicators, pressure switches, water level indicators), are also l determined not to be subject to AMR for the same reason A correlation to the generic l exclusion from the AMR for these additional component types will be provided in the IL i Commodity Evaluation LRA Section. This correlation will consist of a discussion of how l these component types have the same characteristics as those listed and excluded from the ;
AMR in {54.21(a)(1) of the LR Rule. I At this point, one or more of the AMR methods described in Section 6.1 and 6.2 are performed on Its in the scope of this evaluation. Appropriate aging management l alternatives are then selected using the techniques described in Section 6.3.
9 Commodity Evaluations Which Cover All Sconine and IPA Tasks '
For cables, structural supports, FP equipment and cranes / fuel handling equipment, the commodity evaluation covers the component level scoping, the pre-evaluation and the AMR tasks 7.2.1 Cables Commodity Evaluation !
The CCNPP equipment database does not contain specific equipment connectivity for individual cables. Instead, a separate Circuit and Raceway database contains information ,
on cables, their service function (power, control or instrumentation), their materials and their from and to locations. Correlation of cable schemes to individual raceways, equipment and rooms is then possible using the information in this Circuit and Raceway database and design drawings. Because of these differences in site documentation ;
techniques, the BGE IPA process does not include cables within any of the system AMRs, but instead evaluates cables as a separate enmnwiity.
I 7.2.1.1 AMR for Cables Subiect to the EO Program )
l The cable commodity evaluation tasks starts with all site cables, regardless of whether they support any of the intended functions described in {54.4. As discussed in Section 6.1.4, SCs subject to the EQ program are associated with a TLAA that will be evaluated as described in Section 8. Therefore, no further review of EQ cables is performed during the cables commodity evaluation.
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! INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.2.1.2 AMR for Non-EO Cables !
I For the remaming cables, the potential ARDMs which could affect CCNPP cables are considered as discussed in Section 6.2.1. Cables are grouped by common material !
characteristics as described in Section 6.2.2 and the potential ARDM(s) are evaluated to deternune whx:h are plausible for the groups of cables as described in Sectxm 6.2.3. At this point, the component level scoping task is performed, applying the principles described l
, in Section 4, to determine which of the cables which are subject to plausible ARDMs are 3 within the scope of LR. The Pre-Evaluation task is not performed during this cowselity !
evaluation since all cables are passive and long-lived. l For those cables subject to plausible ARDMs which are within the scope of LR, aging management alternatives are selected using the steps described in Section 6.3 l
Therefore, the result of the enmmadity evaluation is the justification that for all cables I within the scope of LR, the effects of aging will be adequately managed by plant programs ,
or activities, or the effects will not prevent the intended functions of the cables during the period of extended operations. )
7.2.2 Cranes / Fuel Handlina Eauioment Commodity Evaluation ,
l The system level scoping results identify five systems within the scope of LR which are !
related to cranes and fuel handling. Because the only intended function of these five !
systems are structural in nature, these five systems are included in a commodity evaluation instead of being addressed individually in the standard IPA process. The five systems are i listed below:
> Spent Fuel Storage !
> Refueling Pool
> New Fuel Storage and Elevator 1
> Fuel Handling
> Cranes I I
The first step of this commodity evaluation is to determine which components in these j systems contribute to the intended functions. The UFSAR and Q-List documentation is i consulted as described in Section 4.2 to determine which components of these systems contribute to the intended structural functions and are therefore within the scope of LR.
Once the components within the scope of LR are defined, the next step is to determine which of these components have already been addressed for their intended, structural type function as part of another AMR (e.g. the AMR of the building which houses the component8 or the commodity evaluation of structural supports). Any such components 9 Because the scoping process for structures addresses all structural support functions for equipment housed by the structure, it is expected that the majority of these components would have already been addressed; however, this step of the commodity evaluation is intended to confirm the process.
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i CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3 are eliminated from the scope of this s. .aiity review. For example, the refueling pool
] structural concrete, stainless steel liner and the fuel transfer tube are addressed in the l AMR of the containment. The spent fuel racks and the spent fuel pool structural concrete j and liner are already addressed in the AMR of the Auxiliary Building. These cue. gor. cats 2
are therefore eliminated from the scope of the crane and fuel handling wnosiity l evaluation.
He next step of the commodity evaluation is to determine which portions of the .
cranes / fuel handling equipment listed above are subject to AMR. This is accomplished by ]
( reviewing the equipment using a process similar to Section 5 Pre-evaluation and i determimng those wi.yacents which contribute to the intended functions through moving l
. parts or a change in configuration or properties. %ese components are active and,' j therefore, are elimi=*~i from the AMR10 The remaining passive components are evaluated for the effects of aging using the techniques described in Section 6.2. Potential ARDM lists are documented for the structural E-5 ==t types. The effects of the potential ARDMs are evaluated to j determme if they could prevent the performance of the intended function. The periodic l inspections and testing programs for designated heavy load handling equipment, as well as j other plant programs and activities, are resiewed to determine whether they adequately ;
} manage the effects of the plausible ARDMs. The steps described in Section 6.3 are used
- to determme the appropriate aging management alternatives and these decisions are j documented.
s 7.2.3 Co.ovo.-. Suonorts Co n.rsiity Evaluation Component supports are associated with equipment in almost every plant system. They perfonn the same basic function, regardless of the system with which they are associated.
For this reason, it was determmed that a commodity evaluation of component supports l l
would be more efficient to address these supports than evaluating them as part of the !
system AMR.
4 j This commodity evaluation begins by performing a scoping task similar to the component
- level scoping of structures described in Sections 4.2.3 and 4.2.4. A generic list of component support types is developed by reviewing industry and plant-specific .
1 information, including Seismic Qualification Utility Group guidance. American Society of l
, Mechanical EngineersSection XI Component support inspection documentation and the CCNPP System Level Scoping Results. All component support types which might provide i support to equipment within the scope of LR are identified, except that snubbers are i specifically excluded as active equipment consistent with the guidance provided in the LR ,
- Rule. The Component Level Scoping Results for each system are then resiewed and the
, component support types which provide support for components within the scope of LR '
) 10 it is conservatively assumed that no components or subcomponents are replaced based on time or qualified life.
4 i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY i are deternuned De results of this step is a listing of the components support types I subject to AMR for each system within the scope of LR. I Except for the exclusion of snubbers, the remaining component supports are treated as passive, long-lived structural components and are subjected to tle AMR. No other pre-evaluation type step is performed for this commodity evaluation.
He.AMR of component supports is then conducted using steps similar to those described in Section 6.2. Potential ARDMs are identified per Step 6.2.1, and the ARDM matrix is l resolved as described in Section 6.2.3. (The intent of component grouping, as described in Section 6.2.2, is already accomplished by the selection of component support types during the scoping steps.) Once the plausible aging mechanisms are determined for each component support type, the steps of Section 6.3 are performed to choose appropriate aging management alternatives for adequately managing the effects of aging for these supports.
7.2.4 FP Equioment Commodity Evaluation Over half of the systems which are included in the scope of LR contribute to one or more FP functions. These functions include both fire suppression / detection functions and functions related to equipment used to demonstrate alterr.::te safe shutdown paths in the event of a severe fire (10 CFR Part 50 Appendix R). For the vast majority of these systems, the nonnal component level scoping task described in Section 4 of this methodology is performed. However, there are seven systems which are in scope for LR primarily because of FP functions 11 For these systems, the alternate scoping steps described in Section 7.2.4.1 are used.
Some passive intended FP functions are performed by fluid systems which are not SR.
For the SCs which are subject to AMR only because of such passive intended functions, an attemate AMR technique is described in Section 7.2.4.2.
7.2.4.1 Scooina of Systems with Primarilv12 FP Intended Functions The seven systems, which are in scope for LR primarily because of FP functions, are listed below.
> Well and Pre-Treated Water
> FP
> Plant Heating
> Condensate
> Plant Drains 11 1.e., The only intended functions of three of the seven systems is a FP function. The other four systems have a FP function and a containment isolation function.
12 See previous footnote. ;
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> Liquid Waste l
> Fire and Smoke Detectum Due to similarity of function, and the fact that most of the FP intended functions are active, an alternate approach is used for conducting the component level scoping of these systems For these seven systems, identification of detailed system functions is performed as described in Section 4.1.1 of this methodology. However, after performance of this step, the intended functions are reviewed using the pre-evaluation step described in Section 5.1 to determme if the functions should be categorized as active or passive. The subsequent steps of the component level scoping task (review of MEL, development of function catalogs and generation of scoping results table) are then conducted on only the passive intended functions of the system and the remainder of the pre-evaluation (short-lived versus long-lived) is completed on only these scoping results.
'Ihe avoided steps in this modified task are the creation and further consideration of function catalogs for the active functions. Had the active function catalogs been created during the component level scoping task, the components in these function catalogs would l have been excluded from the AMR in Section 5.1 because they contribute to only active )
functions. Therefore, this task produces the same list of SCs subject to AMR as would j have been produced by the steps described in Sections 4.1 and 5.
For all of the remaining systems and structures with FP functions, the component level scoping is performed as described previously in Section 4.
7.2.4.2 AMR of FP Pressure-Retaininn Components The pressure-retaining SCs of fluid systems, which are in the scope of LR only because of their contribution to a FP intended function, are addressed in this Section.
The SOC accompanying the LR rule justifies exclusion of SCs associated with active fire suppression / detection functions from the scope of AMR based on the plant's FP Program.
The FPP (Fire Protection Program) is part of the CLB and contains maintenance and testing criteria that provide reasonable assurance thatfire protection systems, structures and components are capable ofperforming their intendedfunction. The Commission concludes that it is appropriate to allow license renewal applicants to take creditfor the FPP as an existing ,
program that manages the detrimental efects of aging. The Commission j concludes that installed fre protection components that perform active functions can be generically excludedfrom an aging management review on i the basis ofperformance or condition-monitoring programs aforded by the l FPP that are capable of detecting and subsequently mitigating the detrimental effects ofaging. (60 FR 22472)
Although the SOC specifically refers only to SCs which contribute to active functions, the justification could apply equally to ' installed FP components that perform passive functions." Therefore, for the fire suppression / detection systems, the AMR applies the 71 Revision 1
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, principles of Section 6.1.1 and consists of demnanrating that the performance and l condition monitonng programs required by the CCNPP FP Program address the pressure-i retaming portions of these fluid system so that the effects of aging are adequately j managed.
i i For the pressure-retaming components in fluid systems credited as alternate safe shutdown l' equipment for Appendix R, the AMR is performed in accordance with Section 6.2 of this methodology.
- 7.3 Commodity Evaluation Results And Documentation Integrated Plant Assessment documentation for commodity evaluations consists of a demonstration ;
- - that the effects of aging are adequately managed for the commodity groups being evaluated and a !
- description of the programs identified during the evaluation which are relied upon to manage the effects of aging. Program modifications or new programs which need to be implemented in order 4 to adequately manage the effects of aging for the period of extended operation would be described.
A summary description of the existing programs and activities, program modifications and new i programs would also be included in the FSAR Supplement.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 8.0 TLAA REVIEW nis section of the IPA methodology describes the task for reviewing analyses which may only be valid during the original 40-year license. His task is performed for the entire plant, whereas the Pre-evaluation and AMR tasks are performed for each system and structure in the scope of LR.
I In 10 CFR 54.3, TLAAs are defined as:
Time-limited aging analyses, for the purposes of this part, are those licensee calculations andanalyses that:
l (1) Involve systems, structures, and components within the scope of license '
renewal, as delineatedin f54.4(a);
(2) Consider the efects ofaging; (3) Involve time-limited assumptions defined by the current operating term, for example, 40 years; (4) Were determined to be relevant by the licensee in making a safety determination; ;
(5) Involve conclusions or provide the basis for conclusions related to the l capability of the system, structure, and component to perform its intended functions, as delineated in f54.4(b); and (6) Are contained or incorporated by reference in the CLB.
He SOC accompanying the LR Rule clarifies the definition of TLAA by explaining that an analysis is relevant if it 'provides the basis for the licensee's safety determination and, in the absence of the analysis, the licensee may have reached a different safety conclusion."
(60 FR 22480) The LR Rule requires that a list of TLAAs (as defined above) be prosided in the LRA, as well as a demonstration that one of the following is true for each TLAA:
(I) The analyses remain validfor the period ofextended operation; (ii) The analyses have been projected to the end of the period of extended operation; or (iii) The efects ofaging on the intendedfunction@) will be adequately managed for the period ofextended operation.
The TLAA Review task produces the required list of the TLAAs which are subject to LR resiew, and demonstrates that these analyses will meet one of the three conditions listed above. Figure 8-1 is a flow diagram which shows the TLAA review task.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TLAA Review Task i Exemptions I Electronic Docket :
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/ \ l For al TLAAs subject Non-exemption to LR review UFSAR -o 1 potential TLAAs v
industry Codes and ~~
Standards p. Are the exemption b M M aging y, _
on a potential @ustely TLAA? \ **"*9'd7 N
Describe TLAA 1
Potential TLAAs~ Yes -
& indicate aging Oncluding exemptions e--- No management as
" # covered by CLB Identify SSC which is program which Yes-subject of TLAA updates the
" TLAA7 Exemption not listed in LRA No No
- SSC in .
LR scope?
AND Can
- Potential TLAA relevant TLAA be modi-to safety determination? fied to be valid through Yes-AND riod of extende "
- Potential TLAA considers the effects operations? Describe TLAA of esing? & modifications AND to TLAA
- Potential TLAA relates No to SSC's ability to perform intended function Provide other justMcation
? Potential that TLAA is valid for the ;
TLAAs not period of extended l listed in LRA operations ;
Yee y ,
Describe TLAA & I TLAAs subject to LR justification ;
[ARTLAAsh j subject to LR review I complete? y TLAA review l complete Figure 8-1 74 Revision 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Sectum 54.21(c)(2) of the LR Rule also requires a list of all exemptions granted under 10 CFR 50.12 wiuch are determined to be based on a TLAA Dese exemptions must be evaluated and justification provxiod for the contmustion of the exemption during the period of erW operation (2) A list must be provided ofplant-specipe exemptions granted pursuant to 10 CFR 50.12 and in efect that are based on time-limited aging analyses as depned in $54.3. The applicant shallprovide an evaluation thatjustipes the continuation ofthese exemptionsfor the period ofextended operation. ;
he TLAA Review task also fulfills this requirement i 8.1 Identify Analyses to be Included in the Review ;
He first step in the TLAA Review task is a search of the CLB to identify potential TLAAs and ex ..y;ksis. He CLB search is done by reviewing the CCNPP electronic docket and the UFSAR.
De electronic docket contains the complete record of docketed correspondence between the NRC and BGE in an casily accessible computer format he UFSAR is also scarchable in the same -
format. Potential TLAAs, such as the aging analyses supporting the EQ Program, are identified by ;
phrases uxhcative of time constramts such as "40 years," "32 EFPY"[ effective full power years],
and " qualified life." Exemptions are identified by using phrases such as "50.12," sad "exen; ion."
Specific examples of potential TLAAs contained in regulatory literature such as SECY 94-140 are reviewed in advance of the electronic search to help focus the search for potential TLAAs The potential TLAAs identified above are supplemented by a further search of the electronic .
docket. Codes and standards which govern design of SSCs at nuclear power plants were reviewed as part of a joint industry effort to determme those that might contain some form of TLAA An additional search of the CCNPP electronic docket and UFSAR is performed using this list of codes !
and standards as the input queries. Any commitments to or reliance on one of the codes and standards with potential time Wies are also included on the list of potential TLAAs Exemptions that are based on time limited aging analyses, the potential TLAAs identified through time related queries and the potential TLAAs identified through codes / standards queries comprise !
the complete set of potential TLAAs identified in this step.
8.2 Review of Potential TLAAs The potential TLAAs are reviewed to determine if they affect an SSC in the IPA scope, to determine whether the analyses are relevant to a safety determination, to determine whether the analyses consider the effects of aging and to determine whether the analyses relate to the ability of the SSC to perform its intended function (s). The potential TLAAs which meet the first four criteria 13 are the TLAAs subject to LR review; i.e., those which must be listed in the LRA.
13 The definition of a TLAA contains six criteria. The two criteria not addressed in this step were already addressed in the initial search technique. The fact that the electronic search was performed against the CCNPP electronic docket and UFSAR implements the criterion that TLAAs be included in or incorporated by 75 Revision 1 l
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l 8.3 Disnosition of TLAAs Which are Subiect to LR Review his step in the TLAA Review task compiles the TLAA-related information for the LRA. Because of the definition of TLAA and the requirements of 54.21(c), there is a defmite relationship between a TLAA and the IPA results for the same SCs.
8.3.1 Fah*iaaAa Between the IPA and TLAAs In some cases, it may be possible to credit the same aging management programs and ;
activities in the TLAA evaluation as were credited in the IPA. He IPA requires a
? .,aregion that the effects of aging are adequately managed for all SCs within the 4
scope of LR that are passive and long lived. 54.21(c) allows three options for addressing TLAAs, one being a demanctration that the effects of aging are adequately managed for the SCs affected by the TLAA The definition of TLAA provides that only analyses affecting SCs within the scope of LR are defmed as TLAAs Therefore, if the IPA is able to demonstrate that the effects of aging associated with the TLAA are adequately managed during the period of avtaad~t operations for a set of SCs, it follows that the requirement under 54.21(c) would also be satisfied. (ne requirements are identical.)
If, on the other hand, certain aging effects associated with a TLAA are difficult or .
impossible to monitor directly, the IPA process may have demonstrated that the effects of -
aging would not prevent the intended function of the SC using an analytical approach.
This approach may have involved extending the existing time-related analysis or .
substituting an alternate analysis, to demonstrate that the effects of aging would not i prevent performance of the intended function during the period of extended operation. In cither case, the requirements of 54.21(c) are still satisfied, since 54.21(c) allows extending the TLAA orjustifying by analysis that the current analysis remains valid for the period of extended operation. ,
Therefore, for long-lived components supporting passive functions, the IPA process required by $54.21(a) will have documented that the effects of aging on these SSCs will be adequately managed. Thus, the only remaining step is to review the IPA results to ensure that the TLAA evaluation requirements are met.
8.3.2 Methods for Extending or Re-evaluating TLAAs Where, as a result of the factors discussed above, the decision is made to extend an existing analysis or justify that the existing analysis remains valid, the techniques used to extend orjustify are specific to each time dependent issue. Where there is already a widely accepted practice (such as 10 CFR 50.61,10 CFR 50.49 or ASME Code) which governs the TLAA, that process is used to re-evaluate or extend the analysis. For example, l reference in the CLB. The time-related queries and the evaluations of codes and standards account for the criterion that TLAAs be related to assumptions regarding the period of the initial license, i.e.,40 years.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 10 CFR 50.61 describes the requirements associated with PTS. Rese requirements would be implemented to account for PTS during the period of extended operations.
Similar to the discussion in Section 6.3.5, if there is an outstanding generic issue associated with the m-analysis process, the SOC to 1:e Rule (60 FR 22484) prosides threc !
options: 1) If the issue is resolved before LRA submittal, the resolution can be incorporated into the LRA, 2) A justification can be developed that the CLB will be maintained until a point in time when one or more reasonable options would be available to adequately manage the effects of aging. For this alternative, a description would be provided for how the CLB would be maintained until the chosen point in time and the options available in the future would be described in general terms. 3) A plant-specific program could be developed that incorporates a resolution to the aging issue.
8.4 TLAA Results and Documentation The results of the TLAA Review task are:
> The list of TLAAs subject to LR resiew;
> The list of exemptions in cfTect that are based on TIAAs; and
> Either:
c ne evaluations which demonstrate that TLAAs remain valid or could be modified to remain valid for the period of extended operation, or o The demonstration that the effects of aging considered by the TLAAs are being managed.
Rese results are described in the LRA. Since the programs credited in this section will normally be identical to those credited in the IPA, little, if any, new infonnation is expected to be added to the FSAR Supplement. More detailed records of the TLAA Review task are maintained onsite.
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