ML20095K405
| ML20095K405 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/20/1995 |
| From: | Denton R BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-M93326, TAC-M93327, NUDOCS 9512280324 | |
| Download: ML20095K405 (100) | |
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Ro:EZT E. DENTON Billimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant Vice President 1650 Calvert Cliffs Parkway Nuclear Energy Lusby, Maryland 20657 j 410 586-2200 Ext.44551.ocal l
410 260-4455 Baltimore {i i
l December 20,1995 1
U. S. Nuclear Regulatory Commission Washington, DC 20555 l ATTENTION: Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant J Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 i Proposed Changes to " Integrated Plant Assessment Methodology" (TAC Nos. M93326 and M93327) l
REFERENCES:
(a) Public Meeting between NRC's and BGE's License Renewal Staffs, dated December 6,1995, Discussions on Responses to a Request for Additional ,
Information (RAI) Concerning the Baltimore Gas and Electric Company )
Report Entitled," Integrated Plant Assessment Methodology" l l
(b) Letter from Mr. R. E. Denton (BGE) to Document Control Desk (NRC),
dated December 15, 1995, " Response to Request for Additional Information (RAI) Concerning the Baltimore Gas and Electric Company R.eport Entitled, " Integrated Plant Assessment Methodology, dated August 18,1995"(TAC Nos. M93326 and M93327)
I At the public meeting held on December 6,1995 (Reference a), Baltimore Gas and Electric Company !
committed to provide a marked-up revision (attached) of the Integrated Plant Assessment Methodology that j incorporates the responses provided in Reference (b). By January 12,1996, we will forward a final version )
of the revised methodology, i 1
l 9512280324 951220 PDR ADOCK 05000317 P pop 0
2%Ceu \
Document Control Desk
~ December 20,1995 Page 2 Should you have further questions regarding this matter, we will be pleased to discuss them with you.
truly yours,
,;pr ikW y or R. E. Denton Vice President - Nuclear Energy RED /JM0/bjd Attachment As Stated cc: (Without Attachment)
D. A. Brune, Esquire J. E. Silberg, Esquire L. B. Marsh, NRC D. G. Mcdonald, Jr., NRC S. F. Newberry, NRC S. A. Reynolds, NRC T. T. Martin, NRC Resident Inspector, NRC R. I. McLean, DNR J. II. Walter, PSC T. Tipton, NEl i
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
1.0 INTRODUCTION
The purpose of this Methodology is to document the plant-specific process used for conducting the Integrated Plant Assessment (IPA) for Aging and the Time-Limited Aging Analysis (TLAA)
Review for the Calvert Cliffs Nuclear Power Plant (CCNPP) in order to produce the information specified in the License Renewal (LR) Rule Section 54.21 (Contents of Application - Technical Infonnation).
During the performance of the IPA process steps described in this methodology, all plant structures and components (SCs) which are subject to aging management review (AMR) are identified. For the identified SCs, justification 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 review process which complements the IPA. This review identifies TLAAs in the CCNPP Current Licensing Basis (CLB) which meet the specific criteria defined in the LR Rule. It also identifies exemptions still in effect which are based on a TLAA. For each of the identified analyses, the review task provides justification that the analysis is valid for the period of extended operations, provides a means for updating 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.
The IPA process for CCNPP has been divided into several distinct tasks. Each of these tasks, as well as the TLAA review task, will be discussed in subsequent sections of this methodology.
The 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 of IPA Methodology.
1.1 Hackcround 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 improve reliability, increase availability, reduce operations and maintenance cost, provide recommendations to the capital improvement plan for the site, prepare License Renewal Applications (LRAs) for both Units, as well as contingency plans for decommissioning.
The LCM Program also coordinates 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 l 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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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 1.2 Methodalogy Summary The BGE IPA methodology is based on the premise that, with the possible exception of the detrimental effects of aging on the functionality of certain systems, structures and components (SSCs) in the period of extended operation, the plant's CLB ensures an adequate level of safety for continued plant operations. Figure 1-1 illustrates the flow path of the BGE IPA, as implemented at CCNPP. The relationship between the IPA and the TLAA review is shown in Figure 1-2.
The Methodology is divided into eight sections. The contents of Sections 2.0 through 8.0 are summarized below.
Section 2.0, IPA Methodology Bases and Definitions, contains the following information:
} Definitions ofimportant terms and acronyms that are integral to the IPA methodology.
> Assumptions and initial conditions on which the IPA methodology is based.
> Source documents which were used to develop the methodology.
Section 3.0, System Level Scoping, describes the scoping steps where SSs that perform specific functions (described in Section 54.4 of the L.R Rule) are identified as the initial scope of equipment, which will be the subject of the IPA for aging.
Section 4.0, Component Level Scoping, 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 functions. This section provides two parallel processes for component level scoping, one used for system components and the other for structural components.
Section 5.0, Pre-Evaluation, describes the various steps which are undertaken to determine which I components are " subject to AMR" in the subsequent task of the IPA.
Section 6.0, AMR, describes how the determination is made that existing, modified or new programs or activities for those SCs subject to AMR adequately manage the effects of aging.
Section 7, Commodity Evaluations, describes alternate IPA process steps used at CCNPP for specific commodity groups.
Section 8.0, ILAA Review, describes the process for selecting TLAAs which need to be addressed for LR and methods for addressing the identified analyses.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY IPA Flow Diagram Pa s sive Function passive or ac tive ?
SSC Scoping System Level & Activ e SC Component subject ,
L ,y,; to periodic
. replacem ent
? U
- SS within scope -
- Intended SC Effects Y88 exclude No functions No of aging a Com ponents by language managed by y that contri. of LR Rule existin g bute to 7 activtles Intended 7 functione. U If Yes Yes y Modify existing programs or im plement new programs to manage the effects of aging.
p SCs NOT SUBJECT TO U k AMR Demonstration that the effects of aging are adequately managed.
Fig ure 1 1 IPA Process and TLAA List of TLAAs, the
$$C whichiney TLAAs identify time-dependent relate to, and the programs analyses which meet the y to manage the aging m criteria of the TLAA considerd in the TLAA or definition in 54.3, the justification that the TLAA g remains valid or could be modified to remain valid for O the period of outended operations.
Summary identify exemplians Description based on a TLAA. U Review CLB _
]
- p
.. p
- LRAt lFSARA
.8"8P*""I '
O h Ak
! k Summary Description Pre-Evaluatio Aging Management Review SSC Functla al Y" Demonstration that the effects of Pass we Existing activities
- un 8s - and > aging are adequately managed.
ed adequate IPA = I,ntend,o,e ,o, iongw.d enects oto, manage agin
,,c, I ? b each SS
- Componenta that No No contribute to Modify existing programe or functions (as implement new programs to -
required only) manage the effects of aging.
p l SC not subject to AMR, l Figure 12 JM Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 2.0 IPA METIIODOLOGY BASES AND OVERVIEW This 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 overview 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 provided in Table 21. Many of the 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 to the Rule. The specific rule section which l is the source of the definition is noted parenthetically for definitions marked with an asterisk.
- 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 this type of change include changes in dimension, ductility, fatigue resistance, fracture toughness, mechanical strength, polymerization, viscosity, and dielectric strength.
- 3. Aging Meehanisms - The physical or chemical processes that result in degradation.
These mechanisms include, but are not limited to, fatigue, erosion, corrosion, erosion / corrosion, wear, thermal embrittlement, radiation embrittlement, 1 microbiologically induced effects, creep, and shrinkage.
- 4. Critical Safety Function (CSF)- A condition or action that prevents core damage or I 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 i 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 Ileat Removal Containment Isolation Containment Environment Control Radiation Control Vital Auxiliaries (VA) 1 The definition of CSF is taken directly from CCNPP O-List documentation which pre-dates the current version of the LR rule.
Therefore. the term " passive" in the CSF definition is not necessarily identical to the temt defined in this methodology and used for convenience in the SOC accompanying 10 CFR Part $4.
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5.(*) Current Licensing Basis (CLB)- The set of NRC requirements applicable to a specific l plant and a licensee's written commitments for assuring compliance with and operation ;
i within applicable NRC requirements, and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are 4
docketed and in effect. The 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; license 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 i
10 CFR 50.71, and the licensee's commitments remaining in effect that were made in docketed licensing correspondence, such as licensee responses to NRC bulletins, generic letters, and enforcement actions, as well as licensee commitments documented in NRC safety evaluations or licensee event reports. [ 54.3]
- 6. Device Type (DT) - A more specific categorization of components according to their function 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 starting point for the grouping process in the AMR task.
Components are grouped by DT as they enter this task. Device types may be divided to form more specific groups if needed, or the DT may define the component group for evaluation. Whenever the LR Rule calls for justifications for SCs, the discussions provided by the BGE IPA process are at the device-type level.
- 7. Equipment Type (ET) - A general categorization of components according to their function 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 concrete / 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. The functional requirements are used during the system- scoping l process 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 term " 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 process.
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 S4 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY the 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 -Ithose functions that are the bases for including SSCs within the l scope of LR. [Q54.4b]
- 13. Licensed Life - The maximum period of operations, in calendar years, as defined by statute. For CCNPP, this period is 40 years.
- 14. Life Cycle Management Evaluation Database (LCMEVAL) _ A computer-based l application which is used to facilitate the component level scoping process for systems.
The LCMEVAL was created, tested and documented, 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 process are extracted one system at a time frcm controlled plant databases, loaded into LCMEVAL, and made available to the evaluator. The LCMEVAL helps to streamline the scoping process by automating key steps and facilitating storage and printing of the results.
15.(*) Long-Lived - Components are considered to be long-lived if they are not subject to periodic replacement based on qualified life _on specified time period er properly justilim! replacement on condition program. [ 54.21(a)(1) cad Statement: of Gonsuleratica (SOC), i.e.,60 FR -t 22178]
- 16. Maintenance Strategy - A philosophy regarding the level and type of maintenance that
, a component will receive throughout its life cycle. An adequate maintenance strategy is denned by the following program attributes:
- a. Discovery - Identi6 cation of perfonnance or condition degradation;
- b. Assessment / analysis - Comparison with criteria or other guidance to determine the degree of the degradation;
- c. Corrective action - Mitigation of the degradation; and
- d. Confirmation / Documentation - VeriGcation 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 NUCLEIS Equipment Iechnical Database (NETD) technical data on equipment for a given system.
18.(*) 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 INTEGRATED PLANT ASSESSMENT METHODOLOGY l
- 19. NUCLEIS Database - A mainframe computer-based information system used to j initiate, 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 is that part of the NUCLEIS information system, indexed by component, which contains information specific to each component.
20.(*) Passive - A function is said to be passive if it is nerformed without moving parts does cc: req 6e me!!ca or a change in configuration or properties in order to perform the function during normal operating conditions or in response to an accident. [o l
54.21(a)(1)].
- 21. Plant Event Evaluations - Pre-existing evaluations which show compliance with regulations concerning fire protection (FP), environmental qualification (EQ), PTS, anticipated transients without scram (ATWS) and station blackout (SBO). These evaluations provide the bases for in-scope determinations under {54.4 Criterion 3.
- 22. Plausible Age-Related Degradation Mechanisms (ARDMs) - (See Aging Mechanisms) An ARDM is considered plausible for a specific component if, when l allowed to continue without any prevention or mitigation measures or enhanced !
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 procedures to a one-time only ;
procedure. l 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 currently in 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 ftmetions.
- 26. Structure - The 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 " structural ,
I compouent" is used for clarity.
i 27.(*) Structures and Components (SCs)- The phrase " structures and components" applies to l matters involving the IPA required by {F4.2'(a) because the AMR required within the IPA should be a component level review uther than a more general system level review.
[ SOC i.e.,80 FR-at 22462] In this Methodology, the term " structural components and l components" (SCs) refers to the component level concept.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 28.(*) Systems, Structures and Components (SSCs)- Throughout these discussions, the term
" systems, structures and components" is used when referring to matters involving the discussions of the overall renewal review, the specific LR scope 2 , TLAA and the LR finding. [ SOC i.e., 80 FR 22462]
29.(*) Structure or Component Subject to Aging Management Review - Structures and components subject to an AMR shall encompass those SCs:
(1) That perform an intended function, as described in f 54.4, 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;-and (4) %:t are ret -bject te replacement bred on a properly,;=tified rep!acemen* cn conditica progre . [ 54.21(a)(1) cad SOC !.e.,50 FR 22178h 30.(*) Systems, Structures, and Components within the Scope of LR - are:
(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) The integrity of the reactor coolant pressure boundary (PB);
(ii) The capability to shut down the reactor and maintain it in a safe l shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents that ,
could result in potential offsite exposure comparable to the l 10 CFR Part 100 guidelines.
(2) All non-safety-related (NSR) SSCs whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (1) (i), (ii), or (iii) of this definition.
(3) All SSCs relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for 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). [ 54.4a], l 31.(*) Time-Limited Aging Analysis (TLAA)- those licensee calculations and analyses that:
(1) Involve SSCs within the scope of LR as delineated in 54.4(a);
2 Note that the CCNPP scoping process is a two-step process with the initial step being conducted at the SSC or system level.
The second step is conducted at the component level and the term SCs applies in this step.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY (2) Consider the effects of aging; (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 ability of the SSCs to perform its intended functions, as delineated in 54.4(b); and (6) Are contained or incorporated by reference in the CLB.
(Q54.3]
Table 2-1 List of Acronyms AFW Auxiliary Feedwater AMR Aging Management Review ARDM Age-Related Degradation Mechanism ATWS Anticipated Transient Without Scram BGE Baltimore Gas and Electric Company j CCNPP Calvert Cliffs Nuclear Power Plant CCW Component Cooling Water 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 FP Fire Protection l FSAR Final Safety Analysis Report ;
GIP Generic implementation Procedure I II/I Seismic two over one design criteria IL Instrument Line l 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 244 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Table 2-1 List of Acronyms PB Pressure Boundary PTS Pressurized Thermal Shock PWSCC Primarv Water Stress Corrosion Cracking SBO Station Blackout SCs Structures and Components SG Steam Generator SOC Statements of Consideration SQUG Seismic Qualification Utility Group SR Safety-Related SS System and Structure SSCs Systems, Structures and Components SVP Seismic Verification Project TLAA Time-Limited Aging Analysis UFSAR Updated Final Safety Analysis Report VA Vital Auxiliary 2.2 Assumptions and Initial Conditions The IPA methodology relies on a number of basic assumptions and initial conditions. They 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 processes 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 steps of the IPA. Sections 1 through 5 of the methodology implement the requirements of 54.21(a)(2) to describe and justify the methods used in 54.21(a)(1).
Sections 6, 7 and 8 go beyond the requirements of 54.21(a)(2) by describing the methods used to perform the AMR and TLAA review. Ilowever, 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.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 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 persomel.
I 2.2.5 During the scoping task, tanks which are included in more than one site documentation 4 system, e.g., both on the site structures list and as a component of a particular system in /
I 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 justification for the safety finding of s54.29, these tasks are performed in accordance l with the BGE quality assurance program. ;
2.2.7 Structural components and components, which contribute to one or more passive functions and are long-lived, require evaluation to demonstrate that the effects of aging are adequately managed.
There are a variety of methods available for managing the effects of aging in order to assure the passive intended function. The 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. This correlation of the effects of aging to the appropriate level of aging management is discussed in detail in Section 6 ,
of this methodology.
2.3 IPA Methodology Overview The IPA methodology describes two scoping tasks, two IPA tasks, and the TLAA review task.
Each is described briefly below.
2.3.1 System Level Scooing 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 Comoonent 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. l Such components are designated as within the scope of LR.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 2.3.3 Pre-Evaluation 4
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:
> Those which contribute only to active functions;
> Those which are replaced based on time or qualified life;-and.and
> Those soecifically excluded by the Rule language in 54.21(aV1Vil
> "c:.: hich are ::p! ::d ~ 'h: beh c' : =cadiec' b=::d progr ~
, (Justmcatica ef'h: dequacy cc :uch a rep! ::m:nt pregre- !: ine!aded * *he brad 4~
The result of this task is the list of all SCs in the given system which will be subject to i AMR.
i 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 presented in this section. All the techniques provide the demonstration necessarvan
- equ've!
- n' !: vel of ==wanee to support the finding of 54.29 with respect to the
~
management of effects of aging.
. I 2.3.5 Commoditv Evaluations i
- Six commodity evaluations are described in Section 7 of the IPA Methodology. These techniques are used for a specific set of components found in a number of systems, but which perform the same or similar functions regardless of their system.
- 2.3.6 ILAA Review l The TLAA Review is described in Section 8 of the IPA methodology. This task searches the CCNPP CLB, independent of the IPA process, to locate issues related to the current operating life of the plant which also meet certain other specified criteria. For the 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 j for convenience.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 2-2 SOURCE DOCUMENTS This list of documents represents the sources used for developing the IPA methodology. This table does not represent all references which might be used in actually performing the tasks described in the methodology. References used in the application of the methodology to a specific system are included in the implementing procedures and in the task-specific results.
1 l 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
- 3. 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities" (routinely updated)
- 4. 10 CFR Part 100, Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"
January 1,1991
- 5. Calvert Cliffs Nuclear Power Plant, Units 1 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 l
Amendment 205 (May 1995) for Unit 1, and Amendment 183 (April 1995) for Unit 2
- 7. CCNPP Design Standard, " Structure and Component Evaluation," (DS-011) Revision 0, June 7,1995 4
- 8. CCNPP Design Standard " Control of Equipment Technical Databases," (DS-032) Revision 0, January 25,1995
- 28. CCNPP System Descriptions:.(various revisions) l 1D9. NRC Regulatory Guide 1.97, " Instrumentation for Light-Water Cooled Nuclear Power Plants to l Assess Plant and Environs Conditions During and Following an Accident," Revision 3 110. CCNPP Plant Drawings (various) l 124. NUREG 1377. " Listing of Nuclear Plant Aging Research Reports," and the reports themselves l 112. Industry Technical Reports on PWR Reactor Vessel, PWR Reactor Vessel Internals, PWR l Containment, PWR Reactor Coolant System, Class 1 Structures and Environmentally-Qualified Cables in Containment 1244 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3.0 SYSTEM LEVEL SCOPING This section describes how all plant SSs are reviewed to determine those that are within the scope of LR. This is accomplished through application of the system-scoping process (Figure 3-1).
Determining which SSCs are within the scope of LR is the first major task described in the IPA methodology. Section Q54.21(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 delineated in f54.4, .
- In other words, the results of the system level and component level scoping tasks are the starting point of the IPA.
System level scoping consists of several activities. Section 3.1 describes how SSs are identified and listed. Section 3.2 describes the development of conceptual boundaries for SSs. Section 3.3 2
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 Identincation of SSs The SS listing for CCNPP is provided in Table 3-1. The 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. This 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 components of associated systems. l 3.2 Denne Conceptual Baundaries This step of the system level scoping process tabulates some basic information about each of the SSs listed in Table 3-1. This information, referred to as the " conceptual boundaries" of the SS, is needed to ensure a consistent understanding of what is meant by each of the SS names in this
! table.
The identification of the SS conceptual boundaries is accomplished by reviewing the CCNPP I Updated Final Safety Analysis Report (UFSAR), Technical Specifications, and System l Descriptions, as well as conducting interviews with experienced plant personnel. For each of the l SSs listed in Table 31, a brief system description is developed and the functional requirements !
are identified. The 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 scoping process, 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 M44 Revision 0 l
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY establish the conceptual boundaries of SSs. The component level scoping task (described in Section 4) develops a detailed list of SS intended functions.
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. s ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METIlODOLOGY System Level y ,lant SSs Scoping Process 1 Define conceptual boundaries and functional requirernents Systems "
Develop screening tools d " Y 54.4(a)(1) 54.4(a)(2) 54.4(a)(3)
Criterion Crtterion Critenon t
DBE Vital FP,EO, Flow Charts Auxiliaries ATWS, SBO, Tool PTS Tools o u o IS SYSI8 No Is systed No la systed is the building a '
"g structure required a- 4- structure required
\gClass 1 structure? by the tool? \ structure required by the tool? by the tool?
Yes Yes Yes Yes k h $ h Add Function to Add Function to Add Function to Add Function to intended Functions Intended Functions- Intended Functions -
Intended Functions->
List List List List List of intended l functions for SSs l
Does the t ystern or structur ~
have an intended Yes function?
SSs within the Scope y, of License Renewal l 1
(No further for these SSs action required Figure 3-1 194 Revision 0 l
,. - - . - - - ~ . . - - ._ . . - - - - . _ _ . - - - -.
ATTACHMENT (1) l 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 In formation
- "
> System or structure name;
> Unit numbei;
> Identification number;
> Brief description.t including major components and system interfaces; l 3
> Source document reference (for the description);
k System or structure functional requirement (s); and
> Source document reference (for each functional requirement).
3.3 Screenina Tools Preparation Screening Tools are created during the scoping process in order to add efficiency to the process 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 process. Each tool is described below. An example of a portion of a screening tool is j provided in Table 3-2.
3.3.1 Tools Addressing @54.4(aV1) and (2) i 10 CFR SU(a)(1) and (2)(referred to as s54.4 Criteria 1 and 2) are addressed together in the System Level Scoping process since both of these criteria were used to establish the CCNPP Q-List documentation.
Q54.4 Criterion 1 1 (1) Safety-related systems, structures and components schich are those relied on to remain functional during andfollo>ving design-basis events [as defined in 10 CFR 50.49 (b)(1)] to ensure thefollowingfunctions --
- (i) The integrity of the reactor coolant pressure boundary; (ii) The capability a ehut 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.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
%54.4 Criterion 2 (2) All nonsafety-related systems, structures and components whose failure could
} prevent satisfactory accomplishment of any of the functions identiped in paragraph (a)(1)(i), (ii) or (iii) of this section (i.e., 54.4).
3.3.1.1 DBE Flow Chart Prenaration ,
l The CCNPP UFSAR Chapter 14 DBE accident analyses listed below are reviewed. This list contains both design basis accidents and anticipated operational occurrences. No i external events are analyzed in Chapter 14 of the CCNPP UFSAR. All structures designed to withstand DBE external events are designated as Class I structures at CCNPP, and Class I structures are included within the scope of LR (Section 3.4.1.2).
Desien Basis Event Chapter 14 Location i
Control Element Assembly (CEA) Withdrawal Event Section 2 i Boron Dilution Event Section 3 Excess Load Event Section 4
) Section 5
! Loss of Load Event l Loss of Feedwater Flow Event Section 6 Excess Feedwater lieat Removal Event Section 7 j
- RCS Depressurization Section 8 Loss of Coolant Flow Event Section 9 Loss of Non Emergency AC Power Section 10 4
Control Element Assembly Drop Event Section i1 Asymmetric Steam Generator (SG) Event Section 12 ,
CEA Ejection Section 13 Steam Line Break Event Section 14 i SG Tube Rupture Event Section 15 Seized Rotor Event Section 16 Loss of Coolant Accident Section 17 Fuel llandling incident Section 18 l Turbine-Generator Overspeed incident Section 19 l l
- Containment Pressure Response Section 20 l
~
liydrogen Accumulation in Containment Section 21 l' Waste Gas Incident Section 22
- ' Section 23 Waste Evaporator Incident Maximum liypothetical Accident Section 24
. Excess Charging Accident Section 25 Feed Line Break Event Section 26 1
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY i
3
- The CCNPP Q-List includes Accident Shutdown Flow Sheets for 17 of the DBEs. Each Accident Shutdown Flow Sheet identifies the CSFs and plant functions supporting CSFs, which are necessary to reach safe shutdown for the DBE identified, maintain fission
- product boundaries, 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 The DBE flow charts are a consolidation of f l'
Q-List Accident Shutdown Flow Sheets and any additional supporting systems identified )
as relied on for that accident in UFSAR Chapter 14. l l For the eight DBEs which are identified in the UFSAR and are not the subject of Q-List j Accident Shutdown Flow Sheets, a DDE flow chart is prepared by the system level scoping process. These DBE Flow Sheets contain the following information depending
- on the reason that no Q-List Accident Shutdown Flow Sheet was prepared (as documented in Q-List documentation).
)
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i 3 The terms *Q-List Accident Shutdown Flow Sheet
- and " Vital Auxiliaries Flow Sheets' are used to refer to documentation which already existed as part of the CCNPP Q-List. The terms "DBE Flow Chart" and Nital Auxiliaries Screening Toor are used to denote the document created during the scoping process to compile the O-List information and other specified information. i 2
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
. Reason Why No Accident Shutdown Information Included 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 required to mitigate the event. passive components are required to mitigate the event.
All components relied on for the event A note stating that all components i are already included in another required to mitigate the event are Accident Flow Sheet. included in another DBE Flow Sheet, and specifying which other DBE(s). l 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 excus 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. The VA screening tool prepared during the system level scoping process duplicates the SSCs listed on the Q List Vital Auxiliaries Flow Sheet using the SS nomenclature shown in Table 3-1.
All systems and ftmetions identified in the DBE flow charts and the VA screening tool are coded (by shading) to identify the source document (s) (i.e., UFSAR, Q-List Manual, or both).
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 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 controlling a similar4 scope of SSCs to that defined by the first two criteria of 54.4. Therefore, the CCNPP scoping process is consistent with the Commission's intent stated in the SOC to the LR Rule.
'4 The CCNPP Q-Ust documentation also establishes controls for PAM (Category 1 and 2) equiprnent. Post-Accident Monitonng equipment satisfies 54.4 Criterion 3. rather than 1 or 2, 2d}44 Revision 0 l
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ATTACHMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT i INTEGRATED PLANT ASSESSMENT METHODOLOGY The Q-List data in the NETD is reviewed to identify items listed as 5049 (items which j must meet the requirements of 10 CFR 50.49). A list of the systems containing
- components 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 I or 2 variables (as defined 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.
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3.3.2.3 PTS Screening Tool Preoaration Since neither CCNPP Unit I nor 2 is expected to require an evaluation in accordance 1 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 l component level scoping process, and the component level scoping results for each
!l 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 process would be modified to require incorporating i the PTS functions relied on in the 1.154 analysis into the PTS Screening Tool. The l 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 i
within the scope of LR.
3.3.2.4 ATWS Screening Tool Preparation j
!. The CCNPP UFSAR is reviewed to identify the system functions that address the j 10 CFR 50.62 requirements on ATWS. An ATWS Screening Tool is developed. The
- identified SS, the tool lists the intended function (s) provided and the appropriate source j documents with the revision number.
I- 3.3.2.5 SBO Screening Tool Preparation The Station Blackout Analysis is reviewed to identify SSs which are relied on during the
" coping duration" phase of an SB0 event. An SBO Screening Tool is prepared which f_ lists the SSs relied on in the Station Blackout Analysis, the function (s) that each j provides, and the appropriate source documents with revision numbers. The power l restoration phase of the Station Blackout Analysis is specifically excluded from review i in this criterion since several success paths for restoring power after an SBO are already j- screened as within the scope of LR due to Criterion 1 (SR).
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY An applicant for LR should rely on the plant's CLB, actual plant-specylc experience, industry-wide operating expertence, as appropriate, and existing engineering evaluations to determine those NSR systems, structures, and components that are the initialfocus of the LR review. (60 FR 22467) 3.3.2 Tools Addressing @54.4(aK3) l l
{54.4 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 regtdationsforfire protection (10 CFR 50.48),
envirormental qualification (10 CFR S0.49), 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 performed to demonstrate compliance with the regulations identified in 54.4(a)(3) (referred to as 54.4 Criterion 3). These evaluations are reviewed to identify SSs that are relied on to mitigaf: 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 ftmetion or support function associated with l 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 54.4 Criterion 3. The analyses and evaluations being reviewed in this step are used to identify intended, NSR functions.
3.3.2.1 FP Screening Tool Prenaration The CCNPP UFSAR, FP Program documentation and the CCNPP Interactive Cable Analysis are reviewed to identify the system functions that address the Commission's regulations on FP and the BGE commitments for implementation of those regulations.
The identified SSCs, their intended function (s), and the appropriate source documents with revision numbers are summarized in the FP Tool. l 3.3.2.2 EO Screening Tool Precaration Two tools are produced for this criterion, the EQ tool and the post-accident monitoring (PAM) tool.
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__. _. m _ __ ___ - - _ ___._ ._m l ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3.4 SS Scoping f The scoping process is implemented for each SS by reviewing each of the screening tools generated in Section 3.3 and developing a System Level Scoping Results Table. (An example page of the System Level Scoping Results Table is shown in Table 3-3.) For the DBE tools and the VA tools, the function (s) being provided are noted on the System Level Scoping Results 1 Table. Since the events summarized by the tools address the requirements of the 54.4 criteria, j inclusion of an SS in a tool indicates that it is within the scope of LR. It is important to note that i
all intended functions are identified for each SS during the scoping process. Identifying only one intended function would be sufficient to make an in-scope determination; however, the list of all intended functions for an SS facilitates the component level scoping task. This step is repeated ;
for each SS so that an in-scope determination is made for each. ;
l l 3.4.1 Criteria 1 and 2 -- SR and SR Sunoort SSs 3.4.1.1 DBE Flow Charts and VA Screening Tool
)
The DBE Cow 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 l
- DBE. The source document references and revision numbers are not included in the l
scoping results table since this information can be found in each DBE flow chart or the VA screening tool, j
! I 3.4.1.2 Class 1 Structures
! For all listed structures, the UFSAR Section 5 and Q-List Design Standard are reviewed
- to determine whether the structure or a portion thereofis designated as SR, Class 1. At l
CCNPP, all Class I structures (buildings) are designated as SR; therefore all Class 1
- structures are screened as within the scope of LR. The results of this scoping step are 1 incorporated, along with the appropriate source document references and revision numbers or dates, into the System Level Scoping Results Table for each of the j structures. j i
3.4.2 Criterion 3 -- SSs Relied On in Plant Safetv Evaluations The corresponding 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 . .
< ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
- 4) Any combination of these factors.
i If one of the SSs being screened 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 revision numbers are not included in the scoping results table since this information can be found in each screening tool.
3.5. Esaulta 4
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.
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES 1 Switchyard (500 kV) & Switchyard DC 46 Extraction Steam 2 Electrical 125VDC Distribution 47 Feedwater Heater Drains and Vents 3 Electrical 13kV Transformers & Buses 48 Engineering Safety Feature Actuation 4 Electrical 4 kV Transformers & Buses 49 Simulator Computer 5 Electrical 480V Transformers & Buses 50 Solid Waste Disposal 6 Electrical 480V Motor Control Centers 51 Plant Water 7 Electrical 13kV Unit Buses 52 Safety injection 8 Well and Pretreated Water 53 Plant Drains 9 Intake Structure 55 CEA Drive Mechanism & Electrical 11 Service Water Cooling 56 Reactor Regulating
. 12 Saltwater Cooling 57 Technical Support Center Computer 13 FP 58 Reactor Protective 14 Transformer Deluge 59 Primary Containment 15 Component Cooling Water (CCW) 60 Primary Containment Heating & Ventilation 16 Electrical 250VDC 61 Containment Spray 17 instrument AC 62 Control Boards 18 Vitalinstrument AC 63 Cathodic Protection l
i 19 Compressed Air 64 Reactor Coolant 20 Data Acquisition Computer 65 Seismic 21 Domestic Water 66 Cavity Cooling 4 22 Makeup Demineralizer 67 Spent Fuel Pool Cooling 68 Spent Fuel Storage 23 Diesel Oil 4
24 Emergency Diesel Generator 69 Waste Gas 25 Access Control Area Ventilation 70 Refueling Pool
- 26 Annunciation 71 Liquid Waste 72 Sewage Treatment Plant 27 Auxiliary SGs j 4 28 Auxiliary Steam 73 Hydrogen Recombiner 29 Plant Heating 74 Nitrogen and Hydrogen 30 Control Room Heating, Ventilation 75 Low Voltage DC Control Power
& Air Conditioning 76 Secondary Sample 31 Meteorology Tower & Miscellaneous 77/79 Area / Process Radiation Monitoring ,
Computers 78 Nuclear Instrumentation l 32 Auxiliary Building and Radwaste 80 New Fuel Storage and Elevator l Heating & Ventilation 81 Fuel Handling j 33 Turbine Building Ventilation 83 Main Steam 34 Condensate Precoat Filter 84 Reactor Vessel Internal 35 Chemical Additions-Turbine 85 Plant Access and Surveillance 1 I
36 Auxiliary Feedwater (AFW) 86 Power Plant Security i 37 Demineralized Water and Condensate 87 Unit Transformers 2
Storage 88 Visitor Center Security 38 Sampling System 89 Emergency Operations Facility Security 39 Condensate Polishing Demineralizer 90 Service Building & Outlying Building 41 Chemical and Volume Control Heating, Ventilation & Air Conditioning 42 Circulating Water 91 Lube Oil Storage 43 Condenser Air Removal 92 Gland Steam 44 Condensate 93 Main Turbine 45 Feedwater 94 Plant Computer 2144 Revision 0 l
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 95 Carbon Dioxide 96 Fire and Smoke Detection 97 Lighting and Power Receptacle 98 Main Generator and Excitation i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES (Continued) 99 Cranes / Test Equipment 105 Weight T sting Wire Ropes & Slings (3) l 100 Plant Communications 106 Ladders 1d Gratings (3) 101 Dry FuelStorage 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 ManualValve 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 t Fuel Oil Storage Tank No. 21 Building.
l Hydrogen Storage Pad Modifications 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) d Diesel Generator Building 2 (2) i
- NOTES
- 1. System listing is from Attachment 6 of DS-032, " Control of the Equipment Technical Databases"
- 2. Systems and structures associated with the new diesel generator installation do not become 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 Pa.-t 50 and, therefore, is not in the scope of this LRA.
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ATTACHMENT (1) 1 CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY I
TABLE 3-2 Revision 4 Post-Accident Monitoring Screening Tool (Example)
Reference 1 - Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Uodated Final Safety Analysis Reoort (UFSAR), Section 7.5.8 Reference 2 - Calvert Cliffs Nuclear Power Plant, NUCLEIS Equipment Database a
SYSTEM / SYSTEM STRUCTURE ID No. MONITORING VARIABLE (S)/ FUNCTION (S)
Electrical 125VDC 2 Status of standby power (voltage, current)
Distribution
- Electrical 4kV 4 Status of standby power (voltage, current)
Transformers and Buses Electrical 480V 5 Status of standby power (voltage, current) l
] Transformers and Buses l Service Water 1i + Service water pump status (motor current)
- Containment cooler cooling water flow a
j Saltwater 12 Saltwater pump status (motor current)
Component Cooling 15 CCW heat exchanger outlet temperature Water
- CCW to/from reactor coolant pumps containment isolation j valve position i
. CCW pump discharge pressure (for flow indication)
+ CCW pump status (motor current) l 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 Instrumentation & 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 AFW flow to SGs
. Motor-driven AFW pump status (motor current)
Condensate storage tank 12 level ;
Sampling System 38
- Containment hydrogen concentration 2344 Revision 0 l
ATTACHMENT (1) .
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-3 BGE LCM PROGRAM TABLE 2 SYSTEM LEVEL SCOPING RESULTS (EXAMPLE) Revision 4 CRITERIA 1 & 2 CRITERION 3 Req'd Class i Class I or SR- In Scope System / Structure Unit ID for DBE DBE Plant Function (s) Q or SR-1M 1M Reference PAM FP ATWS SBO PTS EQ YealNo Suntchyard (500 kV) 1&2 1 No None No N/A N/A No No No No No No No and Switchyard DC Electncal 125 VDC 1&2 2 VA VA for Chemical & Volurne Control System No N/A N/A Yes Yes No No No No Yes Distnbution VA for AFW VA for Main Steam VA for Containment Spray VA for Primary Containmerd Heating &
Ventilation VA for Emergency Diesel Generators VA for 4KV Transformers & Buses VA for 480V Motor Control Centers VA for 480V Pus System VA for Vital instrument AC VA for Servi Water VA for CCW VA for Sattwater Cooting VA for Control Room Heating, Ventilation
& Air Conditioning VA for Auxiliary Building & Radwaste Heating & Ventilation VA for RCS VA for Emergency Safety Features Actua-tion System Load Shedding VA for Chemical & Volume Control System (Core Flush)
Electncal 13kV 1&2 3 No FOne No N/A N/A No No No No No No No Transformers and Buses Electncal 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 Safety injection VA for Containment Spray VA for 480V Bus VA for 480V Motor Control Centers VA for Service Water VA for SW Cooling VA for Emergency Safety Features Actua-tion System Load Shedding Electncal 480V 1&2 5 VA VA for CVCS No N/A N/A Yes Yes No No No No Yes 2229 Revision 0 l
. o ATTACliMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 4.0 COMPONENT LEVEL SCOPING Component level scoping is the second and final 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 Q54.4.
The component level scoping process 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 process for systems, which is illustrated in Figure 4-1 and discussed in Section 4.1, or the component level scoping 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 CantpDnent Level Scopine for Systems The component level scoping process for systems is implemented by systematically reviewing the intended functions of the system (determined by the system level scoping process) 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 process for systems is divided into several distinct steps. Each step is discussed below.
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. o ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY Intended functions for the Component system being scoped level Scoping Process for DBE Flow Charts , Systems PAM, SBO, FP, PTS, .
Describe in. tended function ATWS, EQ Screening in more detailif needed.
Tools Other implicit intended functions; e.g., PB,1 E, structural support. "
Consolidate functions to eliminate duplicates u
MEL for the System
/ \
For allintended
[~* functions of the system System Level Scoping l g Results & References List all system Function catalog 01 components which are _
required to perform the Function catalog 02 I*
function or could fail Plant drawings and prevent the __, e
{ function m i a
/
' Function catalog n Q-List documentation ,
Next intended function
\ /
a Operating Instructions l Resort function catalogs by component List of system components and their intended function (s).
Figure 4-1 1143 Revision 0 l
ATTACIIMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METIIODOLOGY Component Level Scoping for Structures ,
oes the structure have "
a stem type components?
{ l Perform component j level scoping using the i (4 - system process for system type Identify structure intended function components.
- Structural support to SR equipment
- Shelter / protection for SR equipment
- Pressure or fission product boundary
- Missile barrier 4 - Class 11/1 support
- Flood protection barrier
- Rated fire barrier !
v Determine generic structural component types in this structure.
' 1 Add unique structural component types.
v Identify structural component types which contribute to each intended function.
I
?
Add supports for large SR equipment to l scoping results. -
v Integrate scoping results for system type and ,
structural type components.
9 List of structural component types and their intended i functions Figure 4-2 3242 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY l 4.1.1 Identification of Detailed System Functions The purpose of this step of the scoping process is to create a detailed list of the intended l functions associated with the system being scoped. The list is compiled in a System 4 Functions Table using the System and Structure Scoping Results, Q List documentation, I
plant drawings, the UFSAR, System Descriptions and other references. It should be
! noted that these intended functions are reauired to be nerformed under a variety of design conditions in accordance with the CLB.
4 The System and Structure Scoping Results contain screening tools which associate intended functions with individual systems. The first substep 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.
i i
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 i' 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 3
being scoped. The following Q-List categories correspond to s54.4 criteria as described
- below
Q-List Flow Sheets - l These How sheets identify components which are relied on to respond to UFSAR 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 l
3 maintain the PB of Duid systems which are not Ossion product boundary Huid 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
- signincant release of radioactive material to the environment. Criteria 1 and 2.
- (Criterion 2 because IE 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 mechanical equipment that is essential to emergency reactor i shutdown, containment isclation, reactor core cooling, and containment and i 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 of instrumentation used to assess the environs and plant conditions during and following an accident. Criterion 3, subset of EQ.
5049 - This category identifies items which are required to be environmentally qualified to the requirements of 10 CFR 50.49. Criterion 3.
CLSI- The category for those SSCs, including their foundations and supports that are designed to remain functional in the safe shutdown earthquake, 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- The 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.
Aller 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. The Q List does not contain information related to several of the regulated events in s54.4 Criterion 3. Therefore, for the categories shown below, no consolidation with Q-List-related functions is possible. The 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 50.18 for FP and safe shutdown after fire.
ATWS - The functions required by 10 CFR 50.62 to provide diverse scram and diverse turbine trip capability during an ATWS 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. The 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 Table in order to be able to trace the steps of the process which created the i table. Parent functions and functions for which no consolidation is possible are assigned I a unique identification number (Function ID) to facilitate subsequent steps in the scoping process. (For the remainder of this methodology, the term " intended function" refers to a i parent function unless otherwise specified.)
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 4.1.2 The MEL To ensure that all components in the plant are scoped with one and only one system, the site MEL is used to provide the equipment list for the component level scoping task for each system. This list is the portion of the NETD which contains all equipment for a given system.
3 In developing the NETD, conventions were established for determining the boundaries between systems. These conventions provided the guidance for determining which system each component in the IPA would be assigned to. Several example conventions
' are listed below. The complete system boundary guidelines are contained in the site design standard for controlling equipment technical databases.
> Heat exchangers are assigned to the load system.
> Electrical components are assigned to load system 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.
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 intended function listed in the System Functions Table or not needed for any of these functions. l l
4.1.3 Develonment of Function Catalogs
,' The next step in the component level scoping proce,s for systems is to determine, for each intended function, which components from the system MEL are needed to perform the function. A list of components for each function is called the function catalog.
4 In order to determine the relationship between a given function and the components contributing to the function, Q List documentation, UFSAR Technical Specifications, I system screening tools and references associated with the screening tools are used.
The active components associated with mitigating the consequences of individual 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 2
Functions Table contains a DBE function or a VA function, the Q-List provides a direct input to the scoping process for determining which components of the given system contribute to 54.4 Criterion 1 and 2.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 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 function for that ponion of the system relied on to mitigate DBEs. Whenever the system function table contains 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 determination is made, based on these Q-List-coded Piping and Instrumentation Drawings, whether the component is within the annotated PB portion of the drawing. If so, the component is included in the PB catalog. Those passive components which perform in exactly the same manner for any intended function are not included in catalogs associated with other functions in order to avoid redundancy.
The Q-List documentation also contains listings which associate specific components to PAM and EQ functions. This listing is used as a direct input to the scoping process whenever PAM or EQ functions are contained in the system function table. Based on this input, a function catalog is created for both PAM and EQ. In order to be more specific regarding which components actually contribute to providing each of the required PAM indications, plant drawings and the BGE UFSAR are consulted. In addition to the component listing, the PAM catalog contains a letter in the notes column to specify which PAM indication is associated with each component.
The Q-List documentation contains a listing which associates specific components to the Class I function. This listing is used as a direct input to the scoping process whenever there is a Class I function in the System Functions Table. Based on this input, a function catalog is created for Class 1. This catalog normally contains electrical panels (EPs) and other enclosure 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 of important events but are not actually associated with any particular DBE in the Q-List 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.
I The NETD contains a field which associates spccific 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:
> The NETD SBO designation is assigned to components mentioned in the Station Blackout Analysis. Other components which must function so that these
" mentioned" components can perform their SBO function are identified and added to the SBO ftmetion 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 I power afler the end of the event. These components do not contribute to any 0
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY SBO functions in the SBO tool, and therefore are not included within the scope of LR. These components are not included in the SBO function catalogs.
When the process is complete, the SBO function catalog or catalogs contain all of the system components which contribute to each intended SBO function.
The equipment in the system MEL which is designated in Q-List documentation as SR category "Q" also requires further analysis during the scoping process. The documentation which supports the classification of these type components is reviewed to determine why the equipment has been designated as SR category Q. If the SR-Q components perform an intended function, 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 create the associated screening tool is consulted, as needed, along with plant drawings to determine exactly which system components contribute to the performance of each listed function. Components which perform exactly the same function to support one of these j criteria as they perform to support a SR function, are not repeated again in these functio'n 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 mus+ . tart tu provide cooling water 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 cf 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 which justify that a component supports a given function.
4.1.4 Generation of Scoping Results Table in the next step of the component level scoping process 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 Component Level Scopine for Sintetures I The component level scoping process described above for systems can also be applied to structures. Ilowever, this process is somewhat different because of the unique features of 3142 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY structures and how they are documented on site. As with systems, the scoping process 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 cornponent scoping are presented below.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY l 4.2.1 Unique Identifiers for Structural Comnonents l The components of.ctructures have not generally been identified and listed in an MEL.
Consequently, the component level scoping for structures cannot use a comprehensive equipment listing as an input.
For certain site structures, such as the containment, specific component types have been identified in the site equipment database. For these structures, a partial MEL is available and the structural component scoping process is divided into two parts:
- 1) The components documented in an MEL for the structure are scoped using the process described in Section 4.1, above, if it is determined that they do not perform a structural-type function. Components such as the containment personnel hatch, the personnel hatch limit switches and the containment penetrations are scoped using this process because they are designated as components of the containment system in the NETD.
- 2) The remaining portions of the structure such as beams, columns and walls are scoped using the process described in this section.
The results are then merged when both procedures are complete to present a combined scoping result for the entire structure.
4.2.2 Eunction Identification The SS scoping process 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 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 component 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 plantstructures 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 ftmetional 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.);
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
- 3. Serve as a PB or a fission product retention barrier to protect public health and safety in the event of any postulated DBEs;
- 4. Serve as a missile barrier (internal or external);
- 5. Provide structural and/or functional support to NSR equipment whose failure !
could directly prevent satisfactory accomplishment of any of the required SR l functions (Example: seismic Category 11 over I design considerations); )
- 6. Provide flood protection barrier (internal5 flooding event); and
- 7. Provide a rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plant.
This listing allows an evaluator with a specific civil engineering background to determine 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, functions 5 and 6 apply to non-Class I structural components which could, if they fail, prevent a SR function from occurring. This 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 review of various source documents. If the structure is a Class 1 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 made based only on the UFSAR and the System and Structure Scoping Results. Therefore, the i
determination of the applicability of these criteria to the structure is deferred until 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. !
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5 Extemat flooding events were considered dunng the design process for CCNPP structures. It was determined that a probable I maximum hurricane would cause the worst-case flooding conditions at the site. The resulting surge and wave action was ,
analyzed as the basis of plant flood protection. The effects of possible wave action were studied using a hydraulic model. l l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 4.23 Structural Comoonent Tvoe Listing for the Structure In the structural component scoping process, components that are structural in nature are i not uniquely identified during the scoping process. For example, each wall in the structure is not identined, named, and listed. Rather than using an MEL of named structural components, the scoping is conducted on a generic listing of structural component types. This generic list was developed by experts in the field of nuclear Class I structures. The generic list started with structural component types contained 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.
Therefore, several fire- and Gooding-related component types were not considered in these reports.)
The evaluator uses this generic component listing and determines which of the component types on the list are actually contained in the structure being scoped. This step is performed by reviewing plant architectural drawings and identifying the specific
- structural types. AJditionally, any structural component types which are unique to the 1 particular structure being scoped, such as the prestressed tendons in the containment and the sluice gates in the intake structure, are noted. These unique structural component types are then added to the list of applicable structural component types. This list serves l as the equivalent of an MEL for structural component scoping task.
4.2.4 Structural Comoonents Which Contribute to Intended Functions This section describes the process used to determine which component types of a structure contribute te the intended functions which the structure performs. For every I
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). The component types which contribute to each intended function are recorded on the " Structural Components Which Contribute to Intended Functions" table.
1 Additionally, the supports for large SR equipment within the structure are identified by l reviewing a listing of the SR equipment install i in the structure that might affect the j
~
design of the structure (such as tanks, heat exchangers, or vessels filled with Guid 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.
1 Q-List documentation and the FL>oding Design Guidelines Manual are reviewed to l determine if structural component types in the structure being scoped are relied on to l contribute to the functions of providing structural and/or functional support to NSR equipment whose failure could direc4y-prevent satisfactory accomplishment of any of l the required SR functions or providing Dood 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 This Structure?"
column of the Structural Intended Functions Table. The components that contribute to 4142 Revision 0 l
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l ATTACHMENT (1) l l CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY these functions are then recorded on the " Structural Components Which Contribute to Intended Functions" table, with a reference to the appropriate intended structure function.
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 component type necessary for an intended function is designated as within the scope of LR.
4.3 Commodity Evaluations that Include Scoping Sections For certain systems or groups of components, 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 commodity 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 steps are performed using the techniques described in Sections 3 and 4. In the other four commodity evaluation processes, the revised approach replaces the component level scoping, pre-evaluation and AMR. Therefore, for the systems covered by these commodity evaluations, the description of the component level scoping is included in Section 7.
TABLE 4-1 l
Scoping Part of Commodity Evaluation Commodity Evaluation?
EPs & Related Equipment No Instrument Lines (ILs) No Cables Yes Cranes and Fuel llandling Equipment Yes Component Supports Yes FP Systems Yes l l
4.4 Results As a result of the component level scoping process, 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 methodology.
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CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METIIODOLOGY 5.0 PRE-EVALUATION l 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.
The Pre-Evaluation task is performed on a system-by-system or structure-by-structure basis I (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 ;
i as it is applied to structures.
The input to this task is the results of the component level scoping step, described in Section 4, for the system being evaluated. These results consist of the intended functions of the system or j 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 determine which SCs in the system or structure must be fuither 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 prc, cess.
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 f54.4, ident# and list those structures and components subject to an AMR. Structures and components subject to an aging management review shall encompass those structures and components --
(i) 7 hat perform an intendedfimction, as described in f54.4 without moving parts or without a change in configuration or properties. 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 I 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 quahfied hfe or specified timeperiod.
Figure 5-1 provides a flow chart of the Pre-Evaluation task.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY Pre-Evaluation Process I
t Functional Scoping !
Results --* For all passive SCs h
/ \
For allintended functions SCs
~*
of the systems subject to replace-ment at set frequency Yes-I or qualified life <40 years?
Does function involve moving parts or change in No Yes configuration or proper-ties of SCs? **
No further IPA SCs No review required. excluded by LR Rule Yes - >
No language?
- u Add SCs to list of No further IPA review passive SCs for the ---
, N,o required for these SCs system i l Add SCs to l
system list of _
i passive,
" long-lived SCs
. l I l All functions l complete? g
\ /
7 l Any passive SCs 4
remaining?
4 1
List of passive SCs for l re-eval Complet the system y
l h SCs Subject g to AMR i
4 5 u
System Agn LRA AM Evaluation Figure 5-1 147-5 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 5.1 Catsgarl2Elatended System Functions as Active or Passiyc The first step of the Pre-Evaluation task is to review the list ofintended functions for the system being evaluated and characterize each as either active or passive. When a function is determined to be passive, all components which contribute to the passive function are categorized as passive components, even though some of these components may also contribute to an active function.
If such components are determined 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 metiemor change in SC l con 0guration 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 FB of a fluid system. l
> Provide structural support or shelter to equipment.
> Provide missile protection.
, > Provide shielding against radiation.
> Provide shielding against high energy line breaks.
> Provide Hood protection.
- > Prevent or isolate faults in an electrical circuit when such protection or isolation does not involve moving partsmetion or a. change in properties or configuration. l (e.g., cable insulation).
i Any function which is determined to be passive is evaluated in Section 5.2-of-the methodology.
5,l.2 Active Functions i
Auive functions require moving partsmetion or a change in SC properties or l configuration to carry out the intended function. For such functions, plant parameters change in a 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: i
> Provide required Dow to a heat exchanger.
> Provide electrical signals to a device. I
> Provide electrical power to a bus or load.
> Provide indication of a plant condition.
> Remove decay heat.
> Provide fault isolation where moving partsmetion or a change in properties or l configuration is involved. (e.g., circuit breakers, fuses) 41 4 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY Active functions 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 Components Are Long-Lived or Short-Lived In this step of the Pre-Evaluation task, all passive SCs are reviewed to determine if they are subject to replacement based on quali6ed life on speci6ed time period or a proper!y justined comlition based replacement prog at. SCs which are not subject to such replacement are classified as long-lived.
The-easeef-uplasement-based on ::pecified :i:ne period in straightforward. Such rReplacement programs may be based on vendor recommendations, plant experience, or any means which establish a specific replacement frequency. Often, replacement based on quali6ed life will also be replacement at a speci6c time period (i.e., the time period dictated by the quali6ed life),
flowever, in some instances the qualified life of an SC may be based on variables other than calendar time. -For-example,-run ti:ne rather than actu-! calendar ti:ne nicy dic ::: replace:nent for-some-ennponente-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.
A related replacement-prograrn :: cne here SC: cre- replaced be:ed c: performance er comlitie: The SOC: accompanying 4he44-Rule :: ate that
' ': ;!::!:!::!: :t ::: g: cr!:::!!y ex !::: camp;, .. ":: :!:::: ;:re
'!:: Cc rejdaeedk::::!: ;;::r,n:rr:: c": ::!!!!: :fre: . ::g:::g , ;;ge:, "'rev w:-
T!:: C::::mi:, '". :t "*: "' ' : pree!::;!: '!:. ::: : .. ::! ;;pp!!:mt-fmm pmvidi::g !!: ::;n:ll1: ;:::ipe:::!:
. !ie: . .:::: "' ::p;:!!::::!: , !!:"' .
rejdaeement-prnmm-k ::! ;:erl:-" . c. : ::":dition-femmive enmponen:;:: ;'" : ": :mumnee-41::::l:::::!: ::dity '" ': " *: - ::'
- !!::;: ri::.!:;fexte "' '!::p: ;::! (60 FR 22 !M ,
1 There-are-instafWes-where-aftdadientieft-of-80-conditNMM4tn-be " sed as4h0 ba !: far-repl%ement of-a-passive,sG-emi-that-sush-replacement-wouki-prestale4he need for-an-AMP. For exantple:
the eoppesiekel-tubes-of r hea*. exchanger-may-have-an-intemled-pre =ure retaining-funetomr This-funetion+,-pas ve rince there are nc :noving parts c: changes-m-eenfiguration-or-properties )
involved-in-performing 4he functice Normally-suehaubes are not replaced based ca-aW's time-period-or-qualified-lif+:--insteadethey are subject-to eddy current-testing-whie4Hlietates whetHubes4nuct be plugged-an444ube plugg:ng li: nit " hich dictate: " hea-the tube bundle 4 mist be+eplaced. Plant experience she"m that4hese heat er.ch^ngers are retubed every 10 !c !5 yea:s:
in caser, r,ueh-as4his one, ".here a plant parameter-for-a-passive SC can be clea:!y !!aked to the ability of the SC to perfor:r it: Stended fune !ca, and "here plan
- cperating experience !.as showtuhat4he ec npeneat-is-replaeed-frequently, the-SC need not be included en the list of SC:
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CALVERT CLIFFS NUCLEAR POWER PLANT i INTEGRATED PLANT ASSESSMENT METIIODOLOGY Other component: subject te cond!!!c nenhoring :ne!ude rubberhynthetw-parts and pc-t specifically-designed and maintained 4er "ce:- Such parts-are periodically monitored and are normally rep!cced several4imes c er il+ norma 14ife ef the p!nn' " hen "cear er other degradation is+bserved. Such SC need not be ineluded en 'he !!" of SCr, r,ubject-to-AMR, The remaining components which contribute to the oassive function will be subject to aging management review unless the comnonent type has been snecifically excluded from the review by the language of the Rule.
ht-these-eases r ptstification-wit! he provided in-the LRA to demonstrate iba! Such SC: cre repleeed4requentlyrami-therefore-rwtaire nc spee!!ic AMR Tab!: 5 ' che"c: the-eriteria-wineh r or these cases eentrolled-plant program dictate-the are-covered ene!>-justifieatic r conditions-whieh-govenHhe-replacement of the SC !!cwever these r programerare-not-dese+ibed hHhe LRA ar-summarized-itHhe FSAR Supplementem!d be required for prograt~ "'%
manage-the effect: " aging re- SC: subje:: +c AMR !nstead, the LRA justifica!!c . vculd contaitHHIemonstratiefHhat-the-eriteria-of-Tab!: 5 ' have been :::isfied for the program-The level-of-control-whieh eust: for such replacemeat-programs and acti"!!!e: under 'he CLR eontinue-inte-the-period-of-extemled-operatica nd is r,ufficien* :o ensure contirued-replacement of-the43G-5.3 Assignment of System Components to Commodity Evaluations As discussed in Section 4.3, there are several categories of equipment which are more efficiently 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 database. As will be discussed in Section 7 of this methodology, the commodity evaluation process for these components covers replaces-all IPA steps, and this pre- l 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 components, the pre-evaluation process 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.
5.3.1 EEs 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 intended function of such panels is to provide structural support to active system components contained 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 process, such panels are excluded from the AMR of their parent system and are instead administratively included with the EPs commodity evaluation. As will be described in Section 7 of this methodology, the commodity 42-M Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIiODOLOGY ,
1 evaluation produces the same results as the AMR process described in Section 6 but the process is adjusted to be more efficient for a particular component type.
5.3.2 Ils and Tubing 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 4 function of maintaining the system PB and most are not subject to periodic replacement.
Consequently, these ILs are subject to AMR. Instrument lines are subject to common environments, are made of common materials and perform the same passive intended function regardless of the system to which they are assigned. Therefore, the BGE IPA process identifies such ILs during the pre-evaluation process and excludes them from the AMR of the parent system. The commodity evaluation of ILs includes lin !ade !)
presswe-retaking pc-tien: of n:tr=en' , =!> s pre =re tran=!!!ers, presswe indica!!cns, ! /e! tran=i::ers, etc.; 2)1)_small bore piping, tubing and fittings from the moff'ast isolation valve connected-to the instrumentsystem piping; and-21) hand valves which are part of the instrumentsma!! hrench lines (such as equalization. instrument isolation and vent valves for pressure differential transmitters): and 31 any_ other comnonents in the instrument line which contribute substantially to maintaining the nressure retaining function of the instrument line.
5.4 Ilow the Pre-Evaluation Process Applies to Structures For plant structures, a modified process is used to determine which SCs are subject to AMR.
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 is clear that all of the intended structural 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 Long-Lived Plant structural components are not normally subject to periodic replacement programs.
Therefore, structural components are considered to be long-lived unless specific justification is provided to the contrary. Such justification would be included in the LRA.
5.4.3 Structures Which are Also Designated as Systems J In two instances, 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 EM Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY listed as individual components of the containment system with unique equipment identifiers.
As discussed in Section 4 cf the "'A 'k:hedo!cgy, the techniques for scoping of a l 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 process 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 The Pre-Evaluation task produces results which serve as input to the AMR task and to specific commodity evaluations. These results and the documentation of the results are discussed below.
5.5.1 fre-Evaluation Results 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 The Pre-Evaluation task produces a list of the SCs which are subject to AMR for inclusion in the LRA. r esyster ecmpenents :=!uded r e- the AMR because e r a rep!ncemen' progran Smed er candit:e , the 'R^ 4" :nclude ,;uvification thc' the program-ha+Mto-frequent ::p! : m:n of the ecmponen*
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY T..A.D I r. f 1 c .
GRITERIA-FGH44h'PLAPC M.r..MT.-ON-GONIi.fT. .. .- in.M,. D.D.A.P,.D...A. A.18 Griterion4 ".:p! ::::::: progre: r 1::::' c= :onditiona>r-performenetw-easure that the SC:
Identifled-aWithin4'.." ..",^,,' ,','f-IME'.. . .,".'.'.".".A....'..#.,^..^ ' ' .,,.....,.......ItI.'..',,'#
^
''''''''''''""^ * * ^ ' ' ' '
inten:'ed :y::::: fanelion(s). F0r example-
,, . : . . : . . ,c ',~,l'....'I.'." ' '. . ^ . test- ti.'. ". '.,"*. ^I
. ~.' .
t t .. m . , . . , .o......y #. . I.....m.
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intemled-systen; function (s)?
'-' ",n:,ed ca-lhe-eendition or performance trait-menitored by this programr-64he con:penent rMa Mat- interWIs -tha'. ". ' '*. . ,^.,".. . '.'"I.
. ..'. . '. ". * ^ ' '.'. ". I. I. I. ". ^. . #. '.}'. ". ',','. .. . '. ". ". .' 8 '. , ym
. '. '. . ". ". W.
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before-ftseentrilMitieft-te-dalend0d ';yste"' fun'J tions !: prevented?
- - liistericallyr-have-aii-maintei ance preventable functional-failures-ef-intended system-funetiens been-de:ee:ed by thewtetivity?
Griterien ' ". . . '.aeement-programs
-. d.. ,, . e,, ....'. .'.: - - -...-. . '. . . *>a .. ..-- .. "., . . , . '." :... ., .. . . .. . . :. . *. -
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eempenee+before4he.effeetsef-agin;; would preveat4my-intended cystem functiens?
Griterien4--Rep! :::nent-programs-based-on-condition-9e-performance - ' ' mplemented-by the-faeility-operatinjHmteedares.
!- !: the nativity- controlled by c r,ite-rev4ew--pecces: hich ineiudes contre!: c.er subsequent
. . . . . ...m rm. . :,. : .
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CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.0 AMR This Section of the IPA Methodology describes how the components 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 determine which adequately manage the effects of aging. Figure 6-1 is a flow chart which represents the AMR process.
The AMR task fulfills the requirements of 10 CFR 54.21(a)(3)of the LR Rule:
For each structure and component identified in paragraph (a)(1) of this section, demonstrate that the effects of aging will be adequately managed so that the intendedfunction@) will be maintained consistent with the CLBfor the period of extended operation.
The input to the AMR task is the list of SCs subject to AMR along with the intended, passive functions for those SCs. The results of this task demonstrate the following for each input SC or group of SCs:
> 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 adequatelyZmanage the effects of aging; or l
> 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 process 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 son.e 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.
Where the approach described in Section 6.2 is followed, several alternatives for managing the aging effects may be viable and it is necessary to select from those alternatives. In addition, technological developments may produce additional viable alternatives in the future for either I See Section 2.1 for the definition of"adeauatelv manaae."
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY AMR Process List of passive, long-lived SCs and their intended functions.
N Can it be 6.1.1 & 6.1.2 SC parl thown that effects of of a complex Yes- aging are being mng'd\Yes-sembly w/o addressing N ARDMs?
No No 4 ,,
Provide documentation IsSC that effects of aging are LRA feplaced frequent A
- being adequately .
based on condrtiorV Yes 3 managed to assure
? 6.1.3 intended functions. #
No l
e is SC long- Y" lived EQ7 List of piausible
-* ARDM/subcomponent combinations 6.2.1 No 6.2.2 t -
i t t c n
Create potential Organize SCs into f [For each plausible \
ARDM List groups of , .w ARDM/subcomponent subcomponents combination
, input from Site Expert Panel Create ARDM matrix -
Assess level of I , concern and severity ofd aging effects
[Forsubcomponent eachAflBMTT ,
6.3
-l combination i l
6.2.3 g, t IsARDM Add ARDM/ Determine the ,
/ plausible based on y,, subcomponent to list of appropriate type of aterial, environmen plausible ones for the aging management _, LRA No & function? system. based on concerns and l effects. Document reasons.
No 1
Document reasons Allplausible ARDM/
t IPA I usfbe All ARDM/subcomponent-
~
- " P "'
m naged?
"*[( Complete combinations complete? \ / (
\ /
Figure 6-1 5275 Revision 0 l l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY ,
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approach. Section 6.3 describes the CCNPP approach for evaluating and selecting aging l manacementfrom-these alternatives durine the IPA nrocess.
l 6.1 Justification that Effects of Agine are Beine Managed Without Specifically Evaluating ARDMs in several instances, a specific evaluation of the ARDMs is not required in order to justify that the effects of aging are being adequately managed by existing plant programs. These approaches are based on the Commission conclusion stated in the SOC accompanying the LR Rule.
As a plant ages, a variety of aging mechanisms are operative, including erosion, corrosion, wear, thermal and radiation embrittlement, microbiologically induced aging effects, creep, shrinkage, andpossibly others yet to be identified orfidly understood. However, the detrimental effects of aging mechanisms can be observed by detrimental changes in the performance characteristics or condition ofsystems, structures, and components if they are properly monitored.
('60 FR 22474)
EcurT4 wee cases are described in this Section. For thrsstwo of these cases, the AMR l 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 othertlhi case, described in Section 6.1.3, the SCs are subject to a TLAA . The resolution ofthe TLAA will be orovided by one of three msthods described in section 8. c, existing CL" progra: = ckeady-managing-the
. effeets of aging-Aw-adefmed4ime-period.
6.1.1 Comnlex 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 tube 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 ability to perform its active intended function. Corrective actions to 11 # Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METIIODOLOGY !
I restore the passive function from its degraded condition are required by the performance )
testing program and by the normal site corrective actica processes.
Because of the readily observable effects of passive function degradation on active performance, a sufficient method of managing the effects of all types of aging could be is 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 Ggontinuing these types of performance and condition monitoring programs would. ensures that the intended functions of the assembly will be adequately managed.
In some cases. the conclusion of the AMR using this approach may be that the discovery technioues available through the performance and condition monitoring orograms are not I
timelv enough to ensure intended functions as reauired by the CLB. For examole. the i discoverv techniques used in a carticular oerformance and condition monitoring program may_nnly. provide reasonable assurance that the intended function can be oerformed under normal loading conditions. Additional evaluation and/or insoection may be reauired to ensure the ability to oerform intended functions under certain more severe loading conditions which are cart of the CCNPP CLB. . In this case. additional evaluations may be performed to demonstrate that ths. aging mechanisms which may affect the ability of SCs to oerform under more severe loading conditions are not plausible for the SCs. Alternatelv. age-related degradation inspections. as described in Section 6.3.3.4. may be nerformed to determine whether there are aging effects of concern for the SCs being evaluated.
1 Because there may not generally be a close tie between degradation of oassive SCs and the active nerformance of a train of eauioment. Tthe oerformance and condition monitoring his AMR technique is used only in selected circumstances. The conditions listed below represent the feHowing-circumstances where this anoroach should be followed rather than using one of the other AMR aonroaches. These conditions do not constitute a cart of the AMR demonstration itself. The demonstration that these conditions are met would not be submitted as oart of the LRA but would be maintained on sitea
> 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 14M Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT ,
INTEGRATED PLANT ASSESSMENT METHODOLOGY
> The complex assembly is covered by the Maintenance Rule.
6.1.2 Comoonent Assemblies Subject to Comolete Refurbishment 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 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 with the CLB.
An example is the main steam isolation valve operator. This assembly contributes primarily to the active function of closing the main steam isolation valve in a specified amount of time. Because the valve operator uses a combination of hydraulic fluid pressure and compressed nitrogen to operate the vr.lve, several components of this operator assembly provide a passive pressure-retaining function. The entire valve operator is removed from the system at regular intervals and refurbished. Some of the pressure-retaining 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 program without considering specific aging mechanisms.
This approach is restricted to refurbishment programs that meet the following criteria:
> 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; f - The passive ftmetion is the pressure-retaining function and is not a fission product boundary function;
> The program requires complete removal of the component assembly from the system;
> The as:,embly components and subcomponents. including nressure boundaries. l are inspected for signs of aging and other degraded conditions; 4
> The refurbishment directs replacement of components and subcomponents that are deteriorated excessively due to aging or other degradation; and MM Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY
> The refurbishment includes nost maintenance testing consistent with current industry practices and the CLE cc=pc= ' :=:=h!y'- 5:=d:d f =:!c= :=
. ._ . _ a. ._.
. . .r. u. r. . . .u. . o _ _ .. . . . . . . . . . . .
6,1 J Long-Lived EO Comoonents Comoonents subiect to EO which have analified lives less than 40 years are short-lived and would be excluded from the aging management review during the ore-evaluation sten of the orocess. _ Components subject to EQ which have aunlined lives of 40 years or greater are subiect to.a time-limited aging analysis (TLAAL The ootions for resolving TLAAs are described in Section 8. Comoleting one of these TLAA ootions for.jang-lived EO canioment will also serve to orovide the reauired IPA demonstration. are'
- !=edy ad:q :::!y m ::g:d re: 'h: eff: :: cf ag! g. '":i: p=g== :===: '_h:: th:
_ r. e . . , _ c. ._. . : _ .. a n .._,.
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, . , . . . . . . . . . . _ n. . e. ..._.a.
. . ,. . . . ,. ,. . . . . .. c - _. ,, ..e. . _. .: _
. ._.: .. ... _ _ z..a f metica, : =:ctd::= 3:' 'h: CLB, :: : y "m: dud g th: qu !! Sed "f: cf th:
ec= pen:::
Some nortions of nassive EO SCs may not be covered by the EO orogram. For examole.
the EO oronram only analifies the creanic material of a solenoid valve. A senarate
~
AMR evaluation using the techniaue described in Section 6.2. g.ill be nerformed to orovide the reauired demonstration for those nortions of nassive EO SCs which are not covered by the EO orogram.
o-icr 'e = eed:rg the que" Sed !!f: ef =y ecmpc== , the EQ p=g== = qui = th:: 'he ec=p==' he == !y=d ic ='=d :h: q=!!E:d !!f: er 'h: th: Ocmp==t 5: =p!=:d.
Th:=fc=, :he ec=h!=:!ca of th: q=?! Sed 'if: =d th: =qu!===' ic ::k: app =pH:::
:!= pdr ic :==d! g thi: q::!!S=:!c- '. d'! 26;= :!y m:=g: 'h: e cr =:: er ;g ng g c equip ==: ===d by '5!: p=g=
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. . u.
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- rc
- he=hj= toc. ci=: 5!: : cmp ==' = =p!=ed 5:=d c its q=" Sed !:fe. "=
- y =mp:=.? A:5 : q='iSed !!f: g:=:= :hr 10 y==, :he EO Preg == i: :=d!::d =
- l 'h ad:q=t: ag=g - =g:==t p=g=- f= LR, d -^ spee!S: :=!=:!= cf ag=g c = h = ic = .
i 1
6.1.4 SCs Subject to Reolacement on Condition d
In the case of certain SCs. an indication of SC condition is used as the basis for reolacement of a passive SC. For examnle. the cooner-nickel tubes of a heat exchanger
. may have an intended oressure-retainine function. This function is oassive since there I are no moving parts or changes in conScuration or oronerties involved in oerforming the function. Such tubes are not replaced based on a soecific time oeriod or cualified life so they would be included in the aging management review. However. they are subject to i >
gddy current testing which dictates when tubes must be plugged and a tube plugging limit which dictates when the tube bundle must be reolaced. Plant exnerience shows that i these heat exchangers are retubed every 10 to 15 years. In cases such as this one. where .
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INTEGRATED PLANT ASSESSMENT METHODOLOGY a niant parameter for a nassive SC is linked to the ability of the SC to nerform its I intended function. and where olant operating experience has shown that the comnonent is renlaced freauently. the condition-based renlacement nrogram would be credited as )
the aging management nrogram for the SCs.
Table 6-1 shows the criteria which are covered in the detailed demonstration for each SC l or group of SCs subject to this AMR method. These detailed results are maintained on site in an auditable format. The instification orovided in the LRA to demonstrate that the efIcsts of aging are adeauately managed would include a summary _.of the detailed justification.
TABLE 6-1 ,
l 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 frecuency interval less than the shortest time between failures of the SC intended function (s)?
> Based on the condition or performance trait monitored by this nrogram. is the comnonent renlaced at intervals that are short relative to the life of the olant?
k llistorically. have all maintenance oreventable functional failures of SC intended functions been detected bv the activity?
Criterion 2 - Replacement procrams 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 narameter to determine the need for replacement of the SC?
> Does the action value or condition orovide an annronriate means of assuring renlacement of the gemnonent before the effects of aging would nrevent any intended system functions?
Criterion 3 - Replacement programs based on condition or nerformance must be implemented by the facility operatine procedurts,
> is the activity controlled by a site review nrocess which includes controls over subseouent revisions?
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY ;
6.2 Performing an AMR by Evaluatine Aging Mechanism 1 In some circumstances, the most efficient manner 8 o t 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.
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 been 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 compiled. 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 evaluations of SCs of this ET. The Potential i
, 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 Service Water System, this ARDM is eliminated as not applicable because the temperatures throughout the Service Water System are too low to warrant consideration of this mechanism. The 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 are divided into two categories: 1) concrete / architectural components; and 2) steel components; and a Potential ARDM List is created for each of these categories.
l 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 s adequate aging management, BGE chose to use a more mechanistic approach due to other benefits derived from performing this approach.
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, INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.2.2 SC Grouning if a system contains several SCs with similar characteristics, the evaluation process can be made more ef6cient by grouping these SCs together for a common evaluation.
All components of systems are classi6ed 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 process.
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. l Design - 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 identi6er.
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 l 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-retaining function but may not have to reposition for any intended function. Therefore, the discs, seats and air operators of the ulves 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.
Again, 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. J After completion of the system Potential ARDM List and after SCs are grouped and subdivided, an ARDM matrix is created and evaluated. The ARDM matrix consists of I
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIiODOLOGY SC group along the other. Each ARDM/subcomponent intersection must be reviewed 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 for the subcomponent. Although material, environment and function are mentioned separately above, when evaluating ARDM plausibility, all of the factors are akmonsidered together. l l
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 4the rationale for designating each ARDM. This information is recorded in evaluation reports and maintained onsite. A list of the notential ARDMs that were evaluated for each groun of SCs in the system is orovided in the LRA.
1 6.3 Methods to Manage the Effects of A<?ine 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 acing management is to manage the effects of aging such that the intended functions are maintamed consistent with the CLit Consequently. each nhase of the maintenance strategv discussed below takes this goal into consideration when determining the adeauacy of an existing or nronosed nrogram or activity. ,
6.3.1 Phases of a Maintenance Strategy An adequate maintenance strategy consists of four phases: Discovery, Assessment /
Analysis, Corrective Action, and Confirmation / Documentation (1) hsatsry - The first phase of a maintenance strategy is identification that 1 detrimental 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 effects of aging are mamfested in degraded performance or condition, an appropriate license renewal review woidd ensure that l licensee programs adequately monitor performance or condition in a manner that allows for timely identification and correction of degraded 1 conditions. (60 FR 22469)
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CALVERT CLIFFS NUCLEA.R POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Aging can be self revealing or identined through speciGe diagnostic techniques. ,
Current Ecxamples of discovery methods include visual observation of external l conditions, eddy current examination for Daws, and ultrasonic testing for detecting wall thinning. As discussed in Section 6.1.1. these discovery methods may reauire augmentation for license renewal to ensure that the effects of aging are discovered in a timel.v manner such that there is reasonable assurance that the CI B will be maintained. Some plant programs may use speci6c 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 managing-aging mechanisms since l 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 speci6c and generic corrective and preventive action. These criteria and guidance will depend on the characteristics of the degradation and the effects on the intended function. For examnie. a safety or safety sunrort system must be canable of nerforming its speciHe safety function for accident prevention and/or mitigation as described in the CLB. Likewise a system nroviding a function for a regulated event must be sapable of performing that function under the conditions described in the CLB evaluation of the regulated event. _The assessment / analysis chase incornorates such reauirements in determining the need for and nature of corrective actions after abnormal or decraded conditions are discovered. One nossible result of such assessment /anaivsis would be to repeat the discovery phase usmg an exnanded samnte size or usina an augmented or imnroved technioue for discovering anEl cuantifying the utent of a narticular aging effect.
(3) Corrective Action - With the degree of degradation known, speci6c corrective l action can be taken to ensure that the equipment performance or condition is restored and the intended function is maintained. Site orocedures currently exist which reouire root cause analysis and actions to orevent recurrence to be included with correctlye actions when annropriate.
(4) Connrmation/ 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.
In combination, these four phases provide a complete maintenance strategy. Sections 1
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.
6.3.2 Site Exnert Panel Innut 4 l
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CALVERT CI.IFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY The selection of the appropriate method for detecting aging efTects is performed through an expert panel review of each plausible ARDM/ subgroup combination. The review is conducted on a system or commodity basis and, typically, consists of following plant representatives:
> The system or commodity aging evaluation engineer;
> The cognizant system engineer; i > 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.
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 system inspection, surveillance, testing and maintenance results. The 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 area specialist brings his expertise in a specialized area (such as non-destructive examination, EQ, chemistry, materials, fatigue) and provides a perspective in determination of program applicability and feasibility.
These individuals also provide technical concurrence that their program methods will effectively detect and monitor the specified aging effects and are presently the preferred methods.
4 The aging management implementation engineer facilitates the panel meetings, provides consistency between system and commodity technical discussions, ensures involvement of the appropriate plant personnel, and ensures closure of open items.
The 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
> llow the effects of the mechanism progress, i If the panel determines that the ARDM occurs and progresses relatively rapidly, then prescriptive plant programs or system modifications may be warranted. One *im-Age:
Idated degradation inspections and/or performance or condition monitoring may be warranted if:
> The mechanism has not been seen yet in operating plants;
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> 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
> The potential consequences to the system intended function are not significant.
6.3.3 Selection of Aging Management Alternatives for Discoverv Once degradation is discovered, the process described in Section 6.3.4 will ensure that the appropriate Assessment / Analysis, Corrective Action, and Confirmation / l Documentation 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 system-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 expert panel approach. Each of the methods can be categorized into one of the following groups.
6.3.3.1 Plant Programs Plant programs are often the most direct and systematic method of detecting and mitigating the effects of aging. They already exist to meet regulatory requirements or recommendations, warranty requirements, or to preserve economic investment based on site 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 determine if they may serve to discover aging effects for the long lived passive components. In some cases, existing condition monitoring or functional testing may be sufficient; existing 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 reauired 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 reoort and LRA documentation.
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CALVERT CLIFFS NUCLEAR POWER PLA.NT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 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 modi 6 cations to an existing program include, but are not limited to, the following:
> Adding components to inspection procedures for specine aging effects;
> Adding specific aging effects mitigation procedures; and
> Tailoring of record keeping and trending requirements.
if no existing pinnt program can be adapted to address the aging effects for the given group of SCs, new programs may need to be implemented.
Some modi 6 cations to existing programs and new programs may be implemented prior to sabmittal or approval of the LRA. Alternately, .the LRA may include a commitment to implement the program or modi 6 cation at an appropriate future date before or, with aooropriate justincation. during the period of extended operation.
Examples of existing plant programs are shown in Table 6-1.
TABLE 6-1 Examples of Existing Plant Programs Maintenance (Preventive) Materials Testing and Evaluation 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 Preservation Electronic Cable Degradation Post-Maintenance Testing Engineering Test Procedures Pressure Test Procedures Surveillance Test Procedures Plant Tours Fatigue Monitoring Protective Coating and Painting Functional Testing System Walkdowns Environmental QualiGcation Thermography inservice Inspection Vibration Monitoring Loose Parts Monitoring Thermal Performance Monitoring l Lube Oil Analysis Operator Rounds 6.3.3.2 Site Issue Reoorting (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;
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- Maintenance planning walkdowns;
> Walkdowns of planned and completed modifications;
'r Fire watches; and
> Personnel safety equipment inspections.
Any observed or suspected condition that requires significant corrective action, whether 4
related to the purpose of a specific activity being performed or not, is documented via an IR. These methods for discoverv are normally complementarv to other. more formal activities such as age-related degradation insnections. If such activities are relied on as the orincioal means of discoverv. anorooriate _iustification would be orovided in the LRA 6.3.3.3 Plant Modifications
- Plant modifications may be appropriate where
e Plant programs cannot effectively discover the effects of aging;
> Experience indicates that the mechanism is occurring; and
> The progression is relatively rapid.
Modifications will occur as part of the normal site modification process which currently )
exists for improving and updating plant response, performance and reliability. l l
Examples of modifications which might result from the aging evaluations include, but are not limited to, the following:
'r Relocation of equipment to a less aggressive environment;
> 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. With instification. this date may btduring the oeriod of extended onerations. l 6.3.3.4 Age-Related Degradation One We insoections l Two distinct cases of age-related degradation insoections are discussed below. Others may also be nossible.
Case 1: Insocction to Suonort 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:
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> A structure may have been built with a concrete mix that provides maximum
- resistance to freeze-thaw.
> A tank may have been built of stainless steel using strict welding controls to minimize theany chance of stress corrosion cracking. l
! > '-" n:n* : p: ned: 'it!.-A!!cy 600 nay have be:n n:: !!:d ic # mize eernwear i In thi.tthse cases, an ene t:ne inspection could be conducted to orovide additional l assurance eenelude that signincant degradation is not occurring or that the rate is l
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 en: !: ne and add!!!cne! inspr :tions would typically be a sta!!ctica!!y l
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representative sample of the population. Whe.e practicable and prudent, the sample wouldwul be biased to focus on bounding or leading components. For example: l 4
> The portion of a structure more likely to experience the ARDM;_or k A statistically representative sample of the valves made of a particular material;
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> Severa! cf +he A!!cy 600 con:penen::: :: predicted ic be n:cre ==:p 'b!: to
- Primary "S::: Str:= Corresie : Creching.
If the sample-inspection indicates little or no degradation, the conclusion could be reached that the decradation will not result in loss of comoonent function during the
~
oeriod of extended ooeration and therefore. no additional aging management activitister orograms would be reauired. ag:ng nechanis:n wc::!d be adequate!y nanaged by *he i en: " ne pe:4!c ror the :c:npenen'. group er ::ructure. Significant degradation, on
- the other hand, would trigger action under the existing corrective action program and the need for additional inspections would be evaluated.
In ::=: Where the sample-inspection demonstrates that there is no significant l degradation and no program is needed to manage the effects of aging, resolution of the ,
aging mechanism would be documented by describing: I
> The en: " n: inspection process and results; and l k Why it is an adequate approach to dispositien the ARDM for the SC group.
Case 2: Insogstion to Validate an ARDM Mitigation Program In other cases. programs may be in olace which prevent or mitigate the effects of aging.
These aging effects could. if left unmanaged. degrade the capability of SCs to oerform their passive intended functions. In these cases. relying uoon the mitigation orogram
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- INTEGRATED PLANT ASSESSMENT METHODOLOGY mav not orovide the necessary level of assurance that the oassive intended function will be maintained during the oeriod of extended oneration. For examole:
i i > An undernround niping system may be wrapoed with a orotective material to orevent contact with moisture and may also be subiect to an imoressed current
$ cathodic orotection system desianed to orevent corrosion. However. because the l'r oininn is buried and the conseauences of failure would be signincant. a decision j- might be made to perform an insnection of a reoresentative samole of the oining i exterior to confirm - that the mitigation measures have been effective in controlline noinn. ,
> A Guid system may be subiect to cheinistry controls which minimize imnurities
. and maintain a basic pH to limit corrosion of carbon steel comoonents.
!' However. because of the large amount of nining and other comoonents subiect to ;
j such treatment throughout the olant and the range of environmental factors. an j insnection of a reoresentative samnle of comoonents gagld be conducted to ,
- confirm that the chemistrv controls in olace have been effective in controlline '
i the effects of anine.
! In these cases. Insoections could be conducted to conGrm that the mitigation orograms are effective in oreventing or mitigating the aging effects which thev were designed to j control.
$ Again. the scone of such inspections would tvolcally be a representative sample of the population of comoonents of concern. Where oracticable and orudent. the samole would be biased to focus on bounding or leading components. For examole: i I
l: k The underground nining system which is closest to the water table and therefore. l 4
most likelv to have been subiected to moisture,
> The oinine system which has exocrienced the worst historv of chemistry transients and/or has the most suscentible locations:
b If these insocctions reveal little or no denradation. the conclusion could be reached that
! the mitigation oronrams are sufHelent to manage the effects of aging during the oeriod of l extended operations. Sinnincant decradation. on the other hand. would trincer action I under the existing corrective action orogram and the need for additional insoections l would be evaluated. .
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Where the insnection demonstrates there is no significant degradation and the existing
, program is adeauate to manage the effects of aging. this would be documented by
! describino:-
l li i l > The attributes of the orouram which orevents or mitigates the aging effect: and I > The insnection orocess and results:
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For both of the cases described above. the insnection techniaue would need to be capable !
of detectine the effects of acine identified by the AMR. Accentance criteria for these
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insoections would be cons!.ctent with current oractices which account for the SC's ability to oerform intended functions in accordance with the CLB.
For both cm. thcA pe-&u! : en: %: inspections described above may be completed before submittal of the LRA. - When such an early inspection detects no signs of significant aging as expected. there is no need to extranolate the results of the insoection.
If. on the other hand. the inspection reveals significant degradation or unexpected conditions. th results would either be conservatively extrapolated through the end of the oeriod of extended coeration or future insoections would be conducted to track the progress of the unexoected degradation. The freauency of such future insocctions would be commensurate with the safety significance of the SCs being insoected as well as consistent with the results discovered during the initial insoection.
Alternately!n ether en:::, the LRA may commit to conduct the en: %: inspection prior to the oeriod of extended ooeration or. with iustification. during the oeriod of extended operation. Ifindustry experience resolves the aging issue in the interim, the commitment to nerform the inspection could be canceled using existing site commitment management orocedures.
6.3.3.5 Industry Ooerating Exoerience Monitoring plant and industry experience provides the orincipal-for discovery means foref unknown and; theorized, and en:erging aging mechanisms. Additionally.
monitoring industrv experience may be included as one feature of a multi-feature aging management approach when noprooriate.
Tne materials used at CCNPP are common to nuclear plants and to many non-nuclear 4
power, cpera&g plants that have longer operating histories. Monitoring plant and industry experience therefore provides timelv information related to reasonable czurance " unknown and theorized these ARDMs so that there is reasonable assurance that such ARDMs wouldwill be discovered before they severely affect 4 intended functions at CCNPP. It also provides assurance that appropriate changes are made to existing programs. l i
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ATI'ACHMENT (1) l CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Industry information is distributed across the nuclear industry via Institute of Nuclear Power Operation's Significant Event Evaluation Information Network program, which is l a small part of Industry's response to NUREG-0737. The plant program for industry experience reviews problems and events across the industry and evaluates the significance and applicability to CCNPP.
Examples ofinformttion that the program captures are:
> Part 21 Notices;
> NRC Bulletins;
> NRC Information Notices;
> NRC Generic Letters;
> Vendor Information Letters;
> Operating Experience Information;
> Significant Event Reports:
> Operations and Maintenance Reminders; and
> Significant Operating Experience Reports.
In some cases, the aging evaluation may be based on ::: rg:ng :nd=trj information l from the nuclear power industry or other industries that indicates unexpected deterioration may occur. Although the aging effects may not have not-been detected yet l 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 l these cases, discovery his already occurred through notification from NRC, Nuclear
[
- Energy Institute, Institute of Nuclear Power Operations, Owners Groups, or vendors.
4 The site issue reoorting IR and corrective action process requires review and evaluation l ,
of the industry experience, and comparison to conditions at CCNPP to determine if l j additional action is needed here. If resolution of the issue is in progress, it will not i necessarily be completed prior to LRA submittal or approval. The site issue reportinglR l 4
and corrective action process ensures that assessment / analysis occurs and appropriate action is taken.
2 For example, a current industry issue is Primary Water Stress Corrosion Cracking ,
(PWSCC) of Alloy 600. BGE has been closelv involved in the industry and owner's groun efforts to resolve Allov 600 issues. BGE has established a multi-disciplined internal working groun to evaluate imolications of alloy 600 aging for CCNPP. 'Ihe working group used B:::d on current industrv knowledge _and, BOE h : deter"ned l f+om material and environmental properties to determine the suscentibility ofthat alloy i 600 pressure boundary comoonents to PWSCCPr'arj "'ater Strez Cc rc:!cn Crack:ng j for._For some comoonents. where PWSCC was determined to be more likelv. more 1
proactive steps have been taken or are being considered. such as reolacement. nickel olating or destructive testing. For-reactor vessel head penetrations.-at CCNpP. the allov 600 working grouo determined that PWSCC -will initiate and propagate much slower than at many other plants. Inspection results from other plants continue to be reviewed i~ by BGE and continue to suggest no immediate concern for CCNPP. Additional plants )
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l CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METIIODOLOGY are planning inspections. At this time, BGE cannot conclude that inspections will be needed at CCNPP. Ilowever, the processes are in place to ensure appropriate future l decisions are made based on accumulated industry knowledge. )
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i- Implementing the A .sessment/ Analysis. Corrective Action and Confirmation /
63.4 Documentation Phases of the Maintenance Stratenv 1
The last three phases of the maintenance strategy are required by the CLB and are
)
< provided by the site IR and corrective action process. Any observed or suspected j condition that requires significant corrective action, whether related to the purpose of the 4
specific activity being performed or not, is documented via an IR Initiation of an IR j causes the degraded condition or performance to be evaluated for immediate personnel-j or nuclear safety concerns, operability concerns, and reportability. The IR is screened
- and classified to ensure that timely corrective action is taken.
i
- Actions necessary to resolve the IR are assigned to the responsible organization. The IR
- remains open until appropriate actions have been completed and documented. For l- significant events and issues, an event investigation and root cause analysis is conducted I
to aid in preventing reoccurrence.
t Therefore there is reasonable assurance that timely discovery of aging issues and effects I will result in f=!y md appropriate action to evaluate, correct, document, and report l them.
l 6.3.5 Avine Manancment for Avine Issues Associated with a Generic Safety Issue (GSI) or i Unresolved Safety issue (USI) l l If there is an outstanding generic issue (GSI or USI) associated with an identified aging I effect or aging management practice. the SOC to the Rule (FR 22484) orovides three
] options (1) If the issue is resolved before LRA submittal. the applicant can incorporate i the resolution into the LRA. (2) An anolicant can iustify that the CLB will be maintained i until a noint in time when one or more reasonable ootions would be available to
- adeauntely manage the effects of aging. (For this alternative. the applicant would have
. to describe how the CLB would be maintained until the chosen ooint in time and cenerally describe the cotions available in the future.) (3) An anolicant could develoo a
. olant soccific orocram that incorocrates a resolution to the aging issue. ,
In determining the appropriate aging management oractice for SCs affected by GSIs and l USIs. these options should be considered throughout the stens of Section 6.3 and one of i the ootions chosen as anorooriate.
r Ear example. the effects of a particular aging mechanism on a soecific material may be designated by the NRC as a GSI. BGE may choose option (2) above to address this issue i in the IPA. Analvsis 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 oarticination in owner's groun activities or other industry activities. including review ofinsocction results from the more limitine olants. could be used to demonstrate i
that the SC_ intended functionLwill be maintaineil consistent with the CLB. Alternate actions could also be develooed as continnencies. deoendine on the results discovered at l the limiting olants. In this manner. the acing issue associated with the GSI could be 1
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- CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY managed for the ourposes of the IPA. Ultimatelv. resolution of the GSI would include actions. if necessarv. which would be implemented under the current licensing basis.
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- 6.4 Plant Program Documentation
- Documentation in the LRA for this step consists of a demonstration that the effects of aging ate adeauately managed as well as a description of the programs and activities which were identified i during the AMR and are relied upon to manage the effects of aging. ^.dditic
- H!y, any pErogram modifications or new programs which need to be implemented in order to adequately manage 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 demonstrationde=iption provided in the LRA and the summary descriotion provided in the FSAR Supplement.
6.5 IPA SUMM ARY
} The completion of the AMR task concludes the IPA required by the LR Rule. This process demonstrates that the effects of aging have been identified and are being or will be adequately managed. The next section of this methodology describes several specific cases where a slightly different process is used to provide the demonstration reauirgsLfor the IPA. ani7 et equ va! :P resa h 4
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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.
4 These 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, but equ:va!:nt, approach.-In-addition-taleseribingahnteps-of-the c!:ern; : prece=, !% ce:W" d- strates
, that each of4he= prec:==: aruquiva!:nt to the pece:= de=nbed-in4h 'ir:t :N =: tion: cf he methode'egyr I
For the purposes of discussion, the six commodity evaluations are divided into two groups: 1) those that at: equ alent-to-end-replace only the AMR step of the IPA (Section 7.1) and 2) those that are-equiv1!:n: to and-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 Cranes and Fuel llandling Equipment IPA 3 Component Supports IPA FP Equipment IPA 1 I
7.1 Commodity Evaluations Which Cover OnlyEquive!::: te the AMR Step l i
4 For the EPs evaluation and the ILs evaluation, the IPA steps 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 steps 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 i commodity AMRs. The system AMR is conducted as described in Section 6. The commodity )
AMRs are conducted as described below.
i 7.1.1 EP Commodity Evaluation 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 I of the system AMR. Ilowever, the expertise of the evaluator and the type of reference I materials and plant documentation needed to perform the AMR for these two types of I 2.14 Revision 0 l
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equipment is substantially different. Furthermore, the AMR of the EPs requires a level l of expertise, reference material and plant documentation similar to that needed for other SCs in electrical distribution and instrumentation systems. Therefore, for efficiency !
reasons, the EPs are removed from the scope of each system AMR and all EPs (electrical l distribution, instrumentation and panels supporting mechanical system operation) are grouped into a common commodity evaluation.
4 The first step of the EP commodity evaluation is to review the scope of all of the pre- l cvaluation results and to include all EPs subject to AMR in thea commodity evaluation, l regardless of the system the panel is assigned to in the site equipment technical database. ]
Performing 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 i some systems, the only components in the system which were subject to AMR were j those included in the scope of the EP comtnodity 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 commodity 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.
e
- r e- *he clier .ng
- :=cr:, 'h: EP cc:r nodity ;va!uatier p ce = i: equ"' !:n' to the
- endard IPA p c = : 1) ":: =cping and pt: ev;!an"e :: done p : the ::endard
. prece=; and 2) The AMR i cenducted p : en: cf'5: methed: dectibed !- th: ::endard 3
pece = The enly d!$:ne: !: t'-: ":!: p c== i: app!!:d 'e equipmen: ":!:hisre deignated c 2 y:::= :- 'he ci:: echni=! databa=, ":!: differ =ce i: = cunted fc- by l
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. ::p ' 'he commodity ev;! ':c" ":!:h sp=iE= 'he =cp: ef 'he 1
cc od!:y ev '"-' ion; cnd
' ^ :::p 'h: pt: =c!ue:icn e ':ich :=ur= :N' eve:y SC subj=t *e ^.MR i
. ,_. .. ma cm .. . ..
- a.. . .. ..a.. ... _. _, . ,m._ . . .A. W- . D.m. , _.... _-. J.: ...,, .M. u.m. : n.. n. .
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. c-d: d !PA p c== d:= : bed i- Sect:en d 6.
. 7.1.2 IL Commodity Evaluation For many fluid systems, the list of SCs subject to AMR includes many pressure-retaining Md-components which are part of small branch ILs. Regardless of which system these ILs are pmLaf =!g=d 'c, they ch::: certain common characteristics are shared with respect to aging management.
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> All consist of piping and!st tubing which contribute to only one passive intended passive-function, i.e., the pressure-retaining boundary of the system;
> All meludeeontain instrumentation which would be affected to some extent by signincant PB leakage; and 4
^4! re des:gned-mascordaneewith-standard practiets cut!!ned i a speci6sation for41*&CCNPP; a:al
> All system piping to which att "" *^ these ILs are attached is also subject to l AMR.
Because of these common characteristics, the BGE IPA process includes an IL commodity.
Again, the scoping and pre-evaluation steps 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 specines the scope of the evaluation. For every Guid system subject to AMR, pre-evaluation results are reviewedm-and-Tubine. Ottings, hand valves and any other in-line comoonents which are associated with the instrument i and contribute substantially to the nressure-retaining function the sy::en pressure- !
retaining-instrumentation-fincluding =cciated valves) isare included in the scope of this i i commodity evaluation.- The list of ther.e-eomponen*:, p!r 'he ~zeeiated tubing and i
6ttings-(which de act F^ve uric,ue-identifer .a the cite equipment database)rfomt-the scopeef-thiwommodity+valuaJefw The-ne*: :ep of the evah:atier estab!! he 'he ocmbinations af materials--and environments-that exist i: the population of H:trwnents, -!ves, tubing-and-fittings-that I are-in-t he-seope-of-this-evalu a t i o n The range of amterials-and-environments-is !
determined-from-a-revww-of-plant des:gn b si mfomtatien such a: the instrumentation ;
specification. Tab!e 2 shows-the-combinations-and-material: . nd 0uid environments I identified-for-IL et CCNPP At this point, one or more of the a generie-AMR methods l I
described in Section 6.1 and 6.2 are evaluation-of- performed on Ils in the scope of this evaluation. materia!: mahnvironmen" is performed !c determine-which ee nbiretions within-the-populatier are subject 'e plausible age related degradation using-tlw-same eriteria described ' Section 6.2. If-plausib!: ^RDM: are discovered fe a generie esmhination er :rsterials-and-environments, 'he equipment ithin the =cp ef-this evaluation-are-rev=.ed M detenr'ne .hieh !L: actually cente - 'her ecmbinatmas- 1 I
Appropriate aging management alternatives are then selected for-the= ARDMs-using the techniques described in Section 6.3.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METilODOLOGY Again, thi+-conmuxlity-appreaeh-proth:ce; resu!!: Sich are equ..n!:n* *-: results pralueesby-the+tandarelPA-procez described !; Sectien: ' f The =cping end pre-evahtatie-steps-are-performed using *he ::andard IPA procez The 4.M R ctep i ;
slightlywlifferen' - that-4he-4va!cetica ef *he effect of aging ' done for generic :
emnbination: cf n":teria!: end envirmm:enM rather 'he", cetual speci!1ed groups of cenpenen" C er .nateria!! eavironmen* co-F"-tion: Sich :: subject !c pleasible ARDMsr-appropriate aging - "agement chemative: re determined end-the SCs: to whieh-these methed: need !c be app!!:d are identified. T1:e resu!!: are *he justification l that-the efte :: ef aging 1 i!! be adequate!y - neged. ThN re=lt-is-presisely the =me a:
that-produeed by *he ::endard !P A precez. Therefore, 'his !L cc- cdity eve'c *ian A
procez ! equivalent ic the : endard !PA procez described in Section: 6.
7.2 Commodity Evaluations Which Cover AllEeuiv !ent ie the Ent:re Scopine and IPA Steps For the cables, structural supports, FP equipment and cranes / fuel handling commodity evaluations, the process described in this section cosersis-equivalent to the component level scoping, the pre-evaluation and the AMR steps. '1:e follo". ng discuzien i!! provide the justilisation-that-the-procez de=ribed is equ'valenHo-ll """A"rd !P ^ process described :"
Seetic. A 6. ;
4 7.2.1 Cables Commodity Evaluation i
i The CCNPP equipment database does not contain specific equipment connectivity for !
individual cables. Instead, a separate Circuit and Raceway database contains i information on cables, their service function (power, control or instrumentation), their materials and their from and to locations. Correlation of cable schemes to individual racewap, equipment and rooms is then possible using the information in this Circuit and ,
Raceway database and design drawings. Because of these differences in site i documentation techniques, the BGE IPA process does not include cables within any of l the system AMRs, but instead evaluates cables as a separate commodity.
7.2.1.1 Elimination of Cables Subiect to Alreadv Adequate!v Managed hv-the EO Program l The cable commodity evaluation process starts with all site cables, regardless of whether they support any of the intended functions described in 54.4. The first screening step in i this process is to set asideeliminate all cables covered by the EQ Program. As Ddiscussedien in Section 6.1.4. SCs subiect to justi!'e th : the EQ program are associated with-is a TLAA that will be evaluated using the orocess described in Section JLafHMiequate-program for mn"Ogmg the Of8'e0:: cf aglng for e!! SC:' '!th!.: the =cp cf ;
i this ;:regra- Therefore, no further review of EQ cables is performed during the cables commodity evaluation. 1 j
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CALVERT CLIFFS NUCLEAR POWER PLANT j INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.2.1.2 AMR for Cables D:!:r 'nctice " ' %::: !: On'y One Pcten!!n! ^.RDM l i For the remaininga!! ner EQ cables, the potential ARDMs which could affect CCNPP cables are considered as discussed in Section 6.2.1, Cables are grouped by common j material characteristics as described in Section 6.2.2 and the potential ARDM(s) are evaluated to determine which are plausible for the gronos of cables as described in
, Section 6.2.3. - At this ooint in the orocess. the comoonent level scooine sten is
~ '
l performed. applying the orincioles described in Section 4. to determine Evhich of' the cables which are subject to olausible ARDMs are within the scone of license renewal.
I The Pre-evaluation step is not performed during this commodity evaluation since all cables are oassive and long-lived.
For those cables subject to .olausible ARDMs which are within the scone of license renewal. aging management alternatives are selected using the orocess described in Section 6.3 _
Dwing the de'.:!np:::n'. cf 'he ec- cd!!y ev 'untic preer:0, n!! cf th::: n:ch: 1!=n: except thern: ! ag:ng ".::: d:!:r :ned ic be "nct pctentic!" for the rea cas specie:d !- Tab!:
3 Herefore, the ren 9nder of-the ::b!:: ec- cdi:y evaluation fccuse en the questien,
, "Are ny of the : b!:: "hich are ' ::cp: orr LR subject to die!:::ric r "u e due to ther:n ! ag:ng c' norn -! sersice :::nper-:w : ! !:2 t'- : 60 years?"
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9 Aging Management Guideline for Electrical Cable and Terminations prepared by Ogden for DOE, Section 4.1.4, p.4-19. ,
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10 System 1000 is a database managed by United Energy Services Corporation under a 10 CFR Part 50, Appendix B program. For mineralinsulated cable, CE Report 93383-CCE.SR80-1 was consulted since no data was found in System 1000 for this material. The System 1000 database contains time to failure versus temperature data for many organic materials. An Arrhenius analysis is used based on this data, to determine the temperature which results in a time to failure of 60 years.
11 This is based on BGE cable design practices using insulated Power Cable Engineers Association Standards, and the fact that Thermolag-type wrappings are not used at Calvert Cliffs.
3086 Revision 0 l
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1 CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY A n.* m m elm m e m .nfsha. n.~,m .m ,, ,t f e . , m . , m t. . . mn.e l ~. . .- .,
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m..~ ,_ .A... c.v. e.t.. ...m... Of. . . ~., e. , v .n f .,. ,. : .g, Therefore, the result of the commodity evaluation is the justification that for all cables 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 d urin e the o. eriod of ex. tended coerations. W. ._.... --"
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M86 Revision 0
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT ,
INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.2.2 Cranes / Fuel llandling Eauinment Commodity Evaluation 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 listed below:
> Spent Fuel Storage
> Refueling Pool
> New Fuel Storage and Elevator
> Fuel fiandling
> Cranes 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 consulted in much the same manner 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 defmed, 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 Any such component 12 or the commodity evaluation of structural sunoortst components are eliminated from the scope of this commodity review. For example, the refueling pool structural concrete, stainless steel liner and the fuel transfer tube are addressed in the AMR of the containment. The spent fuel racks and the spent fuel pool structural concrete and liner are already addressed in the AMR of the Auxiliary Building.
These components are therefore eliminated from the scope of the crane and fuel handling commodity evaluation.
%Rer e! "' nating the i ' ended f"n :!cn'; c-d componen 0 c' ready addrezed by 'he AMR a
of-the en !c';ing 3:ructure (httilding), only-the se!';:ni !!'!N i-tended functica re:nair c.;
- being no ce np!
- te!y addre=ed by 'he enc!ccing stru :ure'r AMR r er nos p!c-:
equipn:ent, :S f"nct:e- !: ec:np!::ely addr:=ed by 'h: con +:--:!cr of the A 1 cc 'he en0le';ing ";truc4ttre end 'he Ocmodity eva!"-tion of cc:npenen* Suppc-t'; (Sectica
' 2.3). He" Over, for crane'; cnd Pue! han'" ng equipn:ent, pe-tion'; cf the ec:npen:n::
12 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.
i 13 Provide structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory accomplishment of any SR functions (referred to a seismic 11 over I or ll/l).
fl286 Revision 0 l i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METilODOLOGY g g n. . . ... .. n. . _ .. ,. .. _ _ . . . . , .. .i. ,. . ,. . . . .. i. cm. ... ,u
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I 1
i I
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4 i
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3.186 Revision 0 l
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. i ATTACHMENT (1) ,
i CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY D o : . ._. .i. ..i m .. m. A . m i .:.,.n i .~. .a . .-~m . . .. m.
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The next step of the commodity evaluation is to determine which portions of the cranes /
- fuelhe=y ! cad handling equipment listed above are subject to AMR. This is i
accomplished by reviewing the hen"y ned ! endling equipment using a orocess similar to Section 5 Pre. evaluation and determining those components and subcomponents which contribute to the intended Hefunctions through moving oartsmetum or a change in configuration or properties. These components and =heen:penent are active and, therefore, are eliminated from the AMRM.
i.
The remaining passive components and = hec:np^nent; are evaluated for the effects of l aging using the techniques described in Section 6.2. Potential ARDM lists are documented for the structural component types. The effects of the potential ARDMs are evaluated to determine if they could prevent the performance of the intended function.
$ The periodic inspections and testing programs for designated heavy load handling )
equipment, as well as other plant programs and activities, are reviewed to determine l whether they adequately manage the effects of the plausible ARDMs. The process l 1 described in Section 6.3 is used to determine the appropriate aging management '
alternatives and these decisions are documented. l i
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- 14 11 is conservatively assumed that no components or subcomponents are replaced based on time or qualified life.
M86 Revision 0 l
l
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ATTACIIMENT (1)
- CALVERT CLIFFS NUCLEAR POWER PLANT j INTEGRATED PLANT ASSESSMENT METIlODOLOGY ;
i deseribed in Seeti: 6.2, .nd aging enagement c!!:rnatives are e.cluated us:ng 'he Seetinn 6.3 precez. Therefore, band en *he above discussion, the erane end fuel hr ad!!ng conunedity evah:a:!r ' . equ valent-to-the-standard !PA p cec = de=ribed-in See:ica A 6.
7.2.3 Comnonent Supports Commodity Evaluation Component supports are associated with equipment in almost every plant system. They perform the same basic function, regardless of the system with which they are associated. For this reason, it was determined that a commodity evaluation of ;
component supports would be more efficient to address these supports than evaluating them as part of the system AMR.
This commodity evaluation begins with the grouoing sten described in Section 6.2.2.
Component suoports are grouoed together by tvoe of support and cauioment sunoorted i as well as by the environment where the supoort is legated. The next step oerformed is l the comoonent level scooine. This sten uses the orincioles described in Section 4 to l
determine which systems ivithin the scone of license renewal contain each of the comoonent supoort tvoes in the identified grouoings. The comoonent sunoort grouos are then evaluated using the ster:s of Section 6.2.1 (Identify Potential ARDMs) and Section 6.2.3 (Create and Resolve the ARDM Matrix 1 Once the olausible aging mechanisms are determined for each comoonent suoport grouo. the steps of Section 6.3 are nerformed to choose appropriate aging management alternatives for adeauatelv managing the effests of aging for these sucoorts.
" 7 plant-pmgrams-govern :nspe; ion-of-eempenen* supper:0 and fern 'he #cendation for-anMeeded agmg :nanagement-progran %e elements of these progra n are deseribesin-the-folkn ng Sections:
l
' 2.3.! Seisnue Verification " reject (SVP)
He SVP-is-implementing-the-requirements of Unresolved Safety-4ssue A 16 to verify the =: :le adeq iaey of mechanica' and-eleetrieal-equipmentr-ineluding equipmen!
supports-and anche age. To :::: the requirements of Unrev:e".d Safety Issue A 16, the I secpe of equiprnent covered to date by the SVu is !!mited in equipment required fn =fe shutdev . fc!!ce ng a =ir":!0 event end !c e!ectrical race".ny suppc-ts#. The =ismie adequacy c:iterie-were-dermesby-the Seimnie-Qualification Utility Occup (SOUG) and I are doeu nented n-the NRC approved Generic 'mplemen"tica "raeedure-(G149-The eriteric .re based e aspectien: e r equipmen+ strue4ura! and - functional condition fol!c ng 19 arong-motion en-thquahe: c+. crer 80-hdustrial facilities. ^: the time of the post en-thquake nspections, *he everage age of the= facilities ta: 22 years, 15 The CCNPP Individual Plant Examination for External Events is essentially " extending" the scope of the original GIP requirements by conducting walkdowns on other equipment to support the seismic aspect of the probabilistic risk assessment. These walkdowns use criteria similar but not identical to the GtP checklists.
3186 Revision 0 l
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY ,
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16 EPRI Report NP-7149-D, " Summary of Seismic Adequacy of 20 Classes of Equipment Required for the Safe Shutdown of Nuclear Plants
- 3686 Revision 0 l
. o ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METilODOLOGY
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ATTACHMENT (1)
- CALVERT CLIFFS NUCLEAR POWER PLANT 4
INTEGRATED PLANT ASSESSMENT METHODOLOGY 1
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M86 Revision 0 l
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CAINERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
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7.2.4 FP Eauinment Commodity Evaluation Over half of the systems which are included in the scope of LR contribute to one or more FP ftmetions. These functions include both Ore suppression / detection functions and functions related to equipment used to demonstrate alternate safe shutdown paths in the event of a severe fire (10 CFR 50 Appendix R). For the vast majority of these systems, l the normal component level scoping process described in Section 4 of this methodology
- is performed. Ilowever, there are seven systems which are in scope for LR primarily because of FP functions 17. For these systems, the alternate scoping process described in Section 7.2.4.1 is used.
I 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 alternate AMR technique is described in Section 7.2.4.2.
J 1
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i 17 l.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.
H286 Revision 0 l l 1
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ATTACHMENT (1) i CALVERT CLIFFS NUCLEAR POWER PLANT
- INTEGRATED PLANT ASSESSMENT METHODOLOGY f 7.2.4.1 Sconing of Systems with PrimarilyM FP Intended Functions The seven systems, which are in scope for LR primarily because of FP functions, are listed below, l
i > Well and Pre-treated water
' > FP
> Plant Heating l
> Condensate
> Plant Drains J k Liquid Waste
> Fire and Smoke Detection l
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 j performed as described in Section 4.1.1 of this methodology. However, aller i performance of this step, the intended functions are reviewed in the pre-evaluation step described in Section 5.1 to determine if the functions should be categorized as active or passive. The subsequent steps of the component level scoping process (review of MEL, l development of function catalogs and generation of scoping results table) are then i 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.
The avoided steps in this modified process are the creation and further consideration of 4
ftmetion catalogs for the active functions. IInd the active function catalogs been created during the comoonent level sconing orocess. the components in these function catalogs would have been excluded from the AMR in Section 5.1 because they contribute to only l 1
active functions. Therefore. -T1his process produces the samea list of SCs subject to l AMR ar/hich M equiv;!:n' to the !!:t "'dch would have been produced by the process l
} described in Sections 4.1 and 5. "ad the :tive f= tion ::: !cg: been created during the i omn;=nen* !:ve! =cping pree =, th: ec nponen" !c 'h = functinn ::".!cg: "ccu!d Fv: l i
been exch:ded fre " 'he A.MR n Sectic 5.! becaux ' hey contribute to only act ve functi=.
l 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-retaining Comoonents l
i 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.
1 I
- 18 See previous footnote.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY The SOC accompanying the LR rule justiDes exclusion of SCs associated with active Dre 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 of performing their intendedfunction. The Commission conchules that it is appropriate to allow license renewal applicants to take creditfor the FPP as an existing program that manages the detrimental effects of aging. The Commission concludes that installed fire protection components that perform active fimctions can be generically excludedfrom an aging management review on the bash ofperformance or condition-monitoring programs afforded by the FPP that are capable of detecting and subsequently mitigating the detrimental effects ofaging. (60 FR 22472)
Although the SOC speciGcally 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 Ere suppression / detection systems, the AMR noolies the nrincioles of Section 6.1.1 and consists of demonstrating that the nerformance and condition monitoring nrograms reauired by the CCNPP FP Program addresses the pressure-retaining portions of these Duid system so that 13. effects of aging are adequately managed.
For the pressure-retaining components in Huid systems credited as alternate safe shutdown equipment for Appendix R, the AMR is performed in accordance with Section 6.2 of this methodology, ex: p: " hen the cendi:!cn: d:=r: bed be!c"> app!y. l
'- c ne enn ,'he cher ::: =fe:hutdc"- fa wtier required cr*he cy:: - 5: fu!!y :::::d during norn -! p!^-: eperatica beenu= the e!!ers. :: =fe hutde" c function i =5ce ned by i:: pe":: productio- functie- ?:+ degridaten-suai !:n' ic prevent a systen from perr ernl ng !!::herm:0 =fe hu de" r"netic: veu!d be de:::Pd and corrected during j nor:na! p!:": cperations. he cite 4R-and-wrrwt!ve ::!cn progran: een be re!!ed upon to de:un:en' nd actre : 'h: degrade"c- to the pe ter p cduction syste:r befer: " -free;3 the cyste:r's ch!!!!y 'c perfern' i'c !!:rnate =fe r'utdes, function I
e s ha e r e e f N s a the servie: ";!:: nd ec npenen' ecc!!ng veter head "-h: during a C: =en;-!c 'h remove:'he nonna! mah up cure:. '":e re :na! =urce cf"'; :: te c'! :h:= h: d in-h:
!: :he den"nera!!::d 2. ::: sy::::r ":: pre =ure re:::":ng SC: cf the de:r:nera!!::d veter sy::::r -: een!ribu:::e h: i- =ded frnetice are eva!ue::d4n :=crdance " 'he pree = de Or hed i- S=:!c ' f 'r'h: nor~;' coure: is rendered "cperch!e by a =rere f:re, the ^.pp= dix R eva!actie credit: the eenden=:: y::=- for pr-! ding t" mehe up =pp!y cf ::::
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, 7.3 Commodity Evaluation Results And Documentation 1
Integrated Plant Assessment documentation for commodity evaluations would-consists of a
=
demonstration that the effects of aging are adeauately managed for the commodity grouns being evaluated and a description of the programs identified during the evaluation which are relied upon to manage the effects of aging. ^ddM^- "y, ny pErogram modifications or new programs which need to be implemented in order 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 programs would also be included in the FSAR Supplement.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 1
8.0 TLAA REVIEW i l
1 This section of the IPA methodology describes the process for reviewing analyses which may I only be valid during the original 40-year license. This task is performed for the entire plant, whereas the Pre-evaluation and AMR steps are performed for each system _aud structure in the scone oflicense renewal.
1 In 10 CFR 54.3, TLAAs are defined as: I 4
Time-limited aging analyses, for the purposes of this part, are those licensee caletdations andanalyses that:
(1) Involve systems, structures, and components within the scope of license l renewal, as delineated in 954.4(a);
(2) Consider the effects ofaging; l I
(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 cc>nclusions or provide the basis for conclusions related to the capability <>f the system, structure, and component to perform its intended famctions, as delineated in f54.4(b); and (6) if re contained or incorporated by reference in the CLB.
The 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 provided in the LRA, as well as a demonstration that one of the following is true for each TLAATLA: l (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 effects of aging on the intendedfunction(s) will be adequately managed l for the period ofextended operation. l The TLAA Review task produces the required list of the TLAAs which are subject to LR review, 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 process.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY TLAA Review Task Electronic Docket Exemptions f
/ \ .
For all TLAAs subject Non-exemption to LR review UFSAR -*
potential TLAAs I
r
/ Industry Codes and ~
Standards Are the is effects of aging y,, _
exemption based adequately on a potential managed?
TW7 ir j Describe TLAA Potential TLAAs yes & indicate aging (including exemptions *- No nianagenient as with potential TLAAs) u described in IPA No is SSC u # covered by CLB Identify SSC which is program which Yes -
updates the subject of TLAA u TLAA?
Exemption not listed in LRA No
' No
- SSC in LR scope?
AND Can
- Potential TLAA relevant TLAA be modi- l to safety determination? fied to be valid through Yes -
eriod of extended "
AND
- Potential TLAA considers the effects operations? Describe TLAA
"~
of aging? & modifications AND to TLAA !
Potential TLAA relates to SSC's ability to N,o j perform intended function Provide other justification f l
? Potential that TLAA is valid for the !
TLAAs not period of extended )
listed in LRA operations Yes y Describe TLAA &
TLAAs subject to LR justification
[AllTLAAsh subject I to LR review I l
complete? TLAA review complete Figure 8-1 2486 Revision 0 l
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CALVERT CLIFFS NUCLEAR POWER PLANT ,
INTEGRATED PLANT ASSESSMENT METIIODOLOGY l Section 54.21(c)(2) of the LR Rule also requires a list of all exemptions granted under 10 CFR 50.12 which are determined to be based on a TLAA. These exemptions must be evaluated and justification provided for the continuation of the exemption during the period of extended operation.
(2) A list must be provided ofplant-specific exemptions granted pursuant to 10 CFR 50.12 andin effect that are based on time-limited aging analyses as defined in 554.3. The applicant shallprovide an evaluation thatjustifies the continuation of these exemptionsfor the period ofextended operation.
The TLA A Review task also fulfills this requirement.
8.1 Identify Analyses to be Included in the Review The first step in the TLAA Review task is a search of the CLB to identify potential TLAAs and exemptions. The CLB search is done by reviewing the CCNPP electronic docket and the UFSAR. The electronic docket contains the complete record of docketed correspondence between the NRC and BGE in an easily accessible computer format. The UFSAR is also searchable in the same format. Potential TLAAs, such as the aging analyses supporting the EQ Program, are identified by phrases indicative of time constraints such as "40 years," "32 EFPY" ,
I
[ effective full power years], and " qualified life." Exemptions are identified by using phrases such as "50.12," and " exemption." Specific examples of potential TLAAs contained in 1 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 determine those that might contain some form of l 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 dependencies are also included on the list of potential )
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 efTects of aging and to determine whether the analyses relate to the ability of the SSC to perform its intended function (s). Pmedia! TLAA: hich =et 'h= few l 2586 Revision 0 l l
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY enteria are
- hen rev::".ed ic deter :ne h ther *he : alysi is gcVerned by : CLB program hi;h ?! update *h: !yi. he EO Progra: :: =ch a progran The potential TLAAs which meet the first four criteria 19r -and-whi:5 dc c' ::::: the ! :: criter:en, are the TLAAs subject to LR review; i.e., those which must be listed in the LRA.
8.3 Disposition of TLAAs Which are Sublect to LR Review l This step in the TLAA Review task compiles the TLAATLA-related information for the LRA.
Because of the deHnition of TLAAfht check perfc ;; d :- Sect!c- S.2 cheve and the requirements of 54.21(c), !! TLA A :"hjee: 'e LR rev:er =: necezarily Eee: SSC: " hich are i- *he =cpe of LR, per j51 A. there is a definite relationship between a TLAA and the IPA results for the same SCs.
8.3.1 Relationshin Between the IPA and TLAAs in some cases. it may be possible to credit the same aging management orograms and activities in the TLAA evaluation as were credited in the IPA. The IPA requires a demonstration that the effects of aging are adequatelv managed for all SCs within the scope of license renewal that are cassive and long lived. 54.21(c) allows three options for addressing TLAAs. one being a demonstration that the effects of aging are adequately managed for the SCs affected by the TLAA. The definition of TLAA orovides that only analyses affectine SCs within the scone oflicense renewal are defined as TLAAs. Therefore. if the IP'A is able to demonstrate that the effects of aging associated with the TLAA are adequately managed.during the oeriod of extended ooerations for a set of SCs. it follows that the requirement under 54.21(c) would also be satisfied. (The reauirements are identical.)
If. on the other hand. certain aging effects associated with a TLAA are difficult or impossible to monitor directly. the IPA crocess may have demonstrated that the effects
- of aging would not orevent the intended ftmetion of the SC using an analytical anproach.
j This anoroach may have involved extending the existing time-related analvsis or 4
substitutine an alternate analysis. to demonstrate that the effects of ac. ine would not I
\
prevent oerfonnance of the intended ftmetion during the oeriod of extended coeration.
In either case. the requirements of 54.21(c) are still satisfied. since 54.21(c) allows ;
, extending the TLAA or lustifying by analysis that the etgrent analysis remains valid for the period of extended coeration.
Therefore, for long-lived components supporting passive functions, the IPA process 4
required by 54.21(a) will have documented that the effects of aging on these SSCs will l be adequately managed. Thus, the only remaining sten is to review thc_lPA results need l 19 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 taarch was performed against the CCNPP l electronic docket and UFSAR implements the criterion that TLAAs be included in or incorporated by reference in l
' the CLB. The time-related queries and the evaluations of codes and standards account for the criterion ihat l TLAAs be related to assumptions regarding the period of the initial license, i.e. 40 years.
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ATTACHMENT (1) 1 CALVERT CLIFFS NUCLEAR POWER PLANT l INTEGRATED PLANT ASSESSMENT METHODOLOGY on!y cheek to ensure that the TLAA evaluation reauirements are metthey c!:0 addre=
the effects ef ag!n;; c=ceinted " the TLAAs.
8.3.2 Methods for Extending or Re-evaluating TLAAs Where the orocess described above chooses to extend an existing analysis or justify that the existine analysis remains valid. the techniaues used to oerform these tasks is specific
~
to each time deoendent issue. Where there is already a widely accented practice (such as 10 CFR 50.61.10 CFR 50.49 or ASME Code) which coverns the TLaA. that orocess is used to re-evaluate or extend the analysiLFor examole.10 CFR 50.61 describes the reauirements associated with Pressurized Thermal Shock. These reauirements would be imolemented to account for PTS during the oeriod of extended ooerations.
Similar to the discussion in Section 6.3.5. if there is an outstanding generic issue associated with the re-analysis orocess. the SOC to the Rule (FR 22484) provides three ootions (1) If the issue is resolved before LRA submittal. the resolution can he incoroorated into the LRA. (2) A iustification can be develooed that the CLB will be
~
maintained until a noint in time when one or more reasonable ootions would be available to adeauntelv manace the effects of acine. For this alternative. a descriotion would he
~ ~
orovided for how th'e CLB would be inaintained until the chosen ooint in time and the ootions available in the future would be described in general terms. (3) A olant soecific orocram could be develooed that incoroorates a resolution to the aging issue.
^: rated above, for SC: :,ubje : te .^""R, the progren: !! ted are these ! ready identified i' the "'A r ce active er shcr: lived SC: ne : 5jee: to AMR, there cre '5 ::
opt +enst l F '" agemen: c+he effee:: cf aging *kving te *he TLAA rus: be deme::: rated; t
> The TL^ ^ nus be c edified 's projee: !!: app!!:abi!!!y to the end of the period of
- extended operationsw 1 l
F JT.tideatiOM th^t 'he TLAA ren'^ n valid fe" the per;cd of extended oper0:!cn *i'U;! he pawided.-
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ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METilODOLOGY 8.4 TLAA Results and D3cumentationS====-v l The results of the TLAA Review task are:
'r The list of TLAAs subject to LR review;
- The hst of exemptions in effect that are based on TLAAs; and l
'r Either:
c The evaluations analyses-which demonstrate _jusufy-that the-TLAAs remains l valid or could be modified to remain valid for the period of extended operation, or i o The demonstration that the effects of aging considered by the TLAAs are being managed.
These results areis4aformanon4s described inch:ded : a pan c!in the LRA. Since the programs l credited in this section will normally be identical to those credited in the IPA, little, if any, new information is expected to be added to the FSAR Supplement. More detailed records of the TLAA Review task are maintained onsite.
I l
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