ML20087K792
| ML20087K792 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 08/18/1995 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
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| References | |
| NUDOCS 9508240149 | |
| Download: ML20087K792 (138) | |
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CALVERTfCLIFFS-' NUCLEAR POWER PLANT i
INTEGRATEDiPLANT ASSESSMENTL 51ETHODOLOGYe:
i Baltimore Gas and Electric Company August 18,1995 Revision 0 Dg82 149 950828 g
OCK 05o00337 PDR
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METJIODOLOGY TABLE OF CONTENTS bpg w CTIQ h d h hu % m a h we m
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1.0 I NT R O D U CTI O N.........
........................................ 1 1.1 Background...............................................................................................................I 1,2 Meth o d ol o gy S u mm a ry............................................................................................. 2 2.0 1NTEGRATED PLANT ASSESSMENT MET 110DOLOGY........................................ 4 4
BASES AND OVERVIEW i
2.1 Definitions..........................................................................................................4 l
2.2 Assumptions and Initial Conditions.................................................................................... 10 3.0 SYSTE M L EVE L SCO P I N G........................................................................................... I 4 3.1 Identification of Systems and St ru ct u res............................................................................ 14 3.2 Define Con cept u al B o un d aries...............................................................................................
16 3.3 Sc reenin g Tools Prep a ratio n........................................................................................... 16 3.4 Systems an d St ru ct u res S c opin g........................................................................................ 2 2 3.5 Results..............................................................................................................23 4.0 CO M PON ENT LEVE L S CO PING................................................................................... 28 4.1 Component Level Scoping fo r Systems................................................................................ 28 4.2 Component Level Seopin g fo r St ruct ures............................................................................ 35 4.3 Commodity Evaluations that Include Scoping Sections..................................................... 39 4.4 Results...............................................................................................................................39 i
5.0 PR E. E VA L UA TI O N S................................................................................................. 40 5.1 Categorize Intended System Functions as Active or Passive................................................ 42 5.2 Determine Whether Components Are Long. Lived or Short. Lived.............................. 43 5.3 Assignment of System Components to Commodity Evaluations............................................ 44 5.4 How the Pre. Evaluation Process Applies to Structures...................................................... 45 5.5 Pre-Evalu ation Results an d Document ation......................................................................... 46 i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE OF CONTENTS y - - gy y
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k w a S E C TI O N w a u n sua m m a s m a u ;a n w a s w a s w a n,7 yammmaPAGE 6.0 A G I NG M A NA G EM E NT R EVI EW..........................................................................
48 6.1 Justification that Effects of Aging are Being Managed Without Specifically.................. 50 Evaluating Age.Related Degradation Mechanisms 6.2 Performing an A sing Management Review....................................................
. 52 6.3 Methods to Manage the Effects of Aging.........................................................
55 6.4 Plant Program Do cumen t atio n............................................................................
62 6.4 Integrated Plant Assessment Summary................................................................
.... 62 7.0 COMMODITY APPROACllES TO AGING MANAGEMENT REVIEW.
....... 63 7.1 Commodity Evaluations Equivalent to the Aging Management Review Step.........
63 7.2 Commodity Evaluations Equivalent to the Entire Integrated Plant Assessment.................... 67 7.3 Commodity Evaluation Results and Documentation............................................
80 8.0 TIM E. LIMITED AGING ANALYSIS REVI EW..............................................
.. 8I 8.1 Identify Analyses to be Ineluded in the Review...................................................
.... 83 8.2 Review of Potential Time. Limited Aging Analyses.....................................
83 8.3 Disposition of Time. Limited Aging Analyses Which are Subject to License.................. 84 Renewal Review 8.4 Summary.......................................................................................................................85 ii Revision 0 J
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CALVERT CLIFFS NUCLEAR POWER PLANT.
INTEGRATED PLANT ASSESSMENT METHODOLOGY
1.0 INTRODUCTION
ne purpose of this Methodology is to dow.i.c.it the plant-specific process used for conducting the Integrated Plant Assessment (IPA) for Agmg 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.
Informatson).
During the performance of the IPA process steps described in this whadalogy, all plant structures -
and cu.i.pcz.c..ts (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 miended functions of these SCs are adequately managed (see definitions).
In addstion to the IPA process, this methodology describes the TLAA review process which complements the IPA. His review identifies TLAAs in the CCNPP Current Licensing Basis (CLB) which meet the specific criteria defined in the LR Rule. It also identifies 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.
He 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 ofIPA Methodology.
1.1 Backeround 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 deternune 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
- " ^== (LRAs) for both Units, as well as contingency plans for decc i..issioning. The LCM Pregnun also coordinates site activities regarding reactor vessel issues (including pressurized thermal shock [ PTS]) and provides input to corporate Genera. ion Planning and Accounting offices for strategic generation planning. Additional services govemed 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 actisities to focus attention on material conditions and aging management. The LCM Program involves all five Nuclear Energy Division departments and a number of other BGE disisions.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 1.2 Methodolony Summary
'Ihe 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, stmetures 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 Methodoloav 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 Scooina, describes the scoping steps where SSs that perform specific functions (described in Section 54.4 of the LR Rule) are identified as the initial scope of equipment, which will be the subject of the IPA for aging.
Section 4.0, Component Level Scopina, 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.
l Section 5.0, Pre-Evaluation, describes the various steps which are undertaken to determine which '
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, TLAA 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 Passive Function possive or active?
p SSC Scoping SSC System Level &
Active Component subject L,,,g to periodic replacement
- SS within I
U scope
- Intended SSC Effects functione Yes re placed of aging No No
- Components frequently on mensgod by that sentrb condition existing bute to
?
ectivties required 1
functions.
1P l'
Yes Yes qr Modify existing programe or implement new progrents to manage the effects of aging.
p k
SSCs NOT SUBJECT TO l'
AMR Describe programs to manage the effects of aging during the Figure k1 period of extended operations.
IPA Process and TLAA List of TLAAs.the SSCe which tney TLAAS
^"
ves raiat* ** aad ** Pres'a=s identify time-dependent analyses u,
to manage the aging analyses which meet the y
already required 3
considers in the TLAA or critena of the TLA A to be update the justification that the TLAA j
definiten in 64.3.
er CLB remains vahd ercould be modified to remain vahd for Yes the per6od of extended Ne JL operations.
I Summary enury exempeone TLAA does not need Descripton sed on a W.
to be addreseed n.wi g
in th. LR A.
~e-*ta~
CLS
~
O l.eU*93'*$
j??*F$tof S itJLA R A F8 Aft?.
j j
)M-ShpitfBt il i
AL t
t I
t Summary S&C Functonal Pre-Evaluatio Aging Evaluat6en Description I
ScopenS
$$Cs y,,
y, Describe programs to P8888
Exiseng activities
- W W Sh IPA
. Required
- ad adequate to manage manage the effects of aging during l'a8 4d eflecte of egin the period of extended operations.
functions for
?
?
each ss iL
- Components that Ne contribute to funct6ons {ee teodify eulsung programs or required only) implement new progress te
- * "* 8' '"* '** 88 *8 '"8 -
1r l
SC not subject to AMR.
l Figute 1-2 3
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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 dermed below and the meaning of acronyms is provided in Table 2-1. Many of the following defmitions, identified by *, are taken from the LR Rule, Sections 54.3, 54.4, 54.21, and 54.31. The specific rule section which 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 Mechanisms - 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, microbiologically induced effects, creep, and shrinkage.
4.
Critical Safety Function (CSF) - A condition or action that prevents core damage or minimizes radiation release to the public. A CSF may be fulfilled through automatic or manual actuation of a system or systems, from passivei system performance, from inherent plant design, or from operator action while following recovery guidelines set down in procedures. The seven CSFs include:
Reactivity Control Reactor Coolant System (RCS) Pressure and Inventory Control RCS Ileat Removal Containment Isolation Containment Emironment Control Radiation Control Vital Auxiliaries (VA) 1 The definition of CSF is taken directly from CCNPP Q-Ust 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 term defined in this methodology and used for convenience in the SOC accompanying 10 CFR Part 54.
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-CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 5.(A)
Current Licensing Basis (CLB) - The set of NRC requirements applicable to a specific plant and a hcensee's wntsen commitments for assuring compliance with and operation within applicable NRC requirements, and the plant-specific design basis (including all modificatens and additions to such comnutments over the life of the license) that are L.
dacimted and in effect.- The CLB includes the. NRC - regulations contamed in
.10 CFR Parts 2,19, 20, 21, 30, 40, 50, 51, 54, 55, 70, 72, 73,100, and apa 4 >=
thereto; orders; license conditions; -=atiaa=; and technical specifications, it also includes the plant-specific design basis information defined in 10 CFR 50.2, as.
^
dc-r==wi in the most recent Final Safety' Analysis Report.(FSAR) as required by 10 CFR 50.71, and the licensee's commitments remaining in effect that were.made in-docketed licensing correspa-We, 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. [6 54.3]
6.
Device Type (DT) - A more specific categorization of canna-ats M.g to their functon 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. Cwi-:= 3 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-t3pe level.
7.
Equipment Type (ET) - A general categorization of components according to their functon and design. Examples of specific ETs are valve, piping, instrument, etc.- For those SCs subject to AMR, the list of age-related degradation a haaisms (ARDMs) which needs to be addressed is developed for each ET. Structural components are categorized into genenc groupings of concrete / architectural and steel components.
8.
Extended Operations, Period of-The additional amount of time beyond the expiration of the current operatmg 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 functon 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. 'lhe functional requirements are used during the system-scoping process to establish conceptual boundaries so that when a _ detailed function is detennined to be an intended function, the evaluator will know which SS to associate the function with. 'Ihe 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 654.21(a) for LR have been identified and that the effects of aging on the '
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INTEGRATED PLANT ASSESSMENT-METHODOLOGY.
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functionality of such SCs will be managed to maintain the CLB, such that there is an j
acceptable level of safety during the period of extended operationsc [{54.3]
.i
?
12.(*) Intended Function - those functions that are the bases for including SSCs within the scopeofLR. [l54.4b]
j i
13.
Licensed Life - De maximum period of operations, in calendar > cars, as defined by statute. For CCNPP, this period is 40 years.
j 14.
Life Cycle ' Management. Evaluation Database ' (LCMEVAL) A. computer-based i
application which is used to facilitate the component level scoping process for systems
-i
' he 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 from
.[
controlled plant databases, loaded into LCMEVAL, and made available to the evaluator.
l De LCMEVAL helps to streamline the scoping process by automating key steps'and ^
t facilitating storage and printing of the results.
j 15.(a) Long-Lived - Components are considered to be long-lived if they are not subject to l
periodic replacement based on qualified life, specified time period or properly justified replacement on condition program. [{54.21(a)(1) and Statements of Consideration (SOC),
i.e.,60 FR at 22478]
I 16.
Maintenance Strategy - A philosophy regarding the level and type of maintenance that a cc..w.t will' receive throughout its life cycle. An adequate maintenance strategy is.
l defined by the following program attributes:
-l
(
a.
Discovery - Identification of performance or condition degradation;.
l b.
Assessment / analysis - Comparison with criteria or other guidance to deternune i
the degree of the degradation;~
t c.
Corrective action - Mitigation of the degradation; and i
d.
Confirmation / Documentation - Verification and ha-tation that the intended j
function was restored from its degraded condition as a result of the corrective i
action.
l-l 17.
Master Equipment List (MEL) - A compilation of the NUCLEIS Equipment Iechnical Database (NETD) technical data on equipment for a given system.
j j
18.(a) Nuclear Power Plant'-' A commercial nuclear power facility of a type' described in 10 CFR 50.21(b) or 50.22. [i 54.3]
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' 19.
- NUCLEIS Database - A mainframe computer-based information system used to initiate, plan, schedule, track and provide a history of maintenance for all plant components.
NETD is an acronym used to denote the EUCLEIS Equipment Iechnical Database, whichi is that part of the NUCLEIS information system, indexed by component, which contains mformation specific to each component.
20.(*) Passive - A function is said to be passive if it does not require motion or a change in configuration or properties in order to perform the' function during normal operatung
=w%s or in response to an accident. [i 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).
"Ihese 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 allowed to continue
- without any prevention or mitigation measures or mh-~1 monitoring techniques, it could not be shown that the component would maintain its capability to perfonn 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.
24.(a) Renewal Term - The period of time that is the sum of the additional amount of time beyond the expiration of the operating license (not to exceed 20 years) that is requested in the renewal application plus the remaining' number of years on the operating license currentlyin effect. [l54.31(b)]
25.
Screening Tool - A summary of source document (s) compiled through the research of an -
event / topic which contains lists of responding SSCs and their intended functions.
26.
Structure - The term structure, when used as a stand-alone term in this malvvialogy, refers to a building. When a component of a structure is referred to, the term 'itructural -
component"is used for clarity.
27.(*) Structures and Components (SCs) 'Ihe phase 'htructures and.x-g-:
=ts" applies to -
matters involving the IPA required by (54.21(a) because the AMR required within the IPA should be a component level review rather than a more general system level review
[ SOC i.e., 80 FR at 22462] In this Methodology, the term 'htructural components and components" (SCs) refers to the component level concept.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 28.(*) Systems, Structures and Components (SSCs) - Throughout these discussions, the term
" systems, structures and components" is used when referring to matters invohing the
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discussions of the overall renewal review, the specific LR scope, 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 perfonn an intended function as described in {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 (3) nat are not subject to replacement based on a properly justified replacement on i
condition program. [Q54.21(a)(1) and SOC i.e.,60 FR 22478].
30.(a) 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 (i)
He integrity of the reactor coolant pressure boundary (PB);
i (ii)
He capability to shut down the reactor and maintain it in a safe shutdown condition; c,r (iii)
The capability to prevent or mitigate the consequences of accidents that could mult in potential offsite exposure comparable to the 10 CFR Pan 100 guidelines.
(2)
All non-safety-related (NSR) SSCs whose failure could prevent satisfactory accomplishment of any of the fimetions identified in paragraphs (1) (i), (ii), or (iii) of this defmition.
(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].
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 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)
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.
[{54.3]
Table 2-1 List of Aeronyms AFW Auxiliary Fecdwater AMR Aging Management Review j
ARDM Age-Related Degradation Mechanism ATWS Anticipated Transient Without Scram BGE Baltimore Gas and Electric Company 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 Erwironmental Qualification ET Equipment Type FP Fire Protection FSAR Final Safety Analysis Report GIP Generic Implementation Procedure II/I Seismic two over one design criteria IL Instrument Line IPA Integrated Plant Assessment IR Issue Report LCM Life Cycle Management LCMEVAL Life Cycle Management Evaluation Database LR License Renewal LRA License Renewal Application MEL Master Equipment List 9
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' Table 2-1 List of Acronyms NETD NUCLEIS Equipment Technical Database NSR Non-Safety-Related PAM Post-Accident Monitoring PB Pressure Boundary PTS Pressurized ucrmal Shock SBO Station Blackout SCs Stmetures and Components SG Steam Generator j
SOC Statements of Consideration.
j SQUG Seismic Qualification Utility Group j
SR Safety-Related 4
i 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 Auxiliarv i
2.2 Assumptions and Initial Conditions
'I 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
%c criteria underlying the system level and component level scoping processes are identical.
2.2.3 He 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 j
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. However, the description of these methods should facilitate a better understanding of the results produced by these tasks. The results will be documented in the LRA and FSAR Supplement.
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' 2.2.4' 'Ibe IPA methodology is designed to make maxunum use of existing BGE programs,.
system and equipment lists, documents, and databases to reduce duplication of effort and produce implementation results which reference equipment nomenclature already familiar to site personnel.
j
- 1 2.2.5 During the scoping task, tanks which are included in more than one sitehimentation
-I system, e.g., both on the site structures list and as a coinponent of a particular system in an -
MEL, are included ' nly as co. w.c.es of a system during the IPA process, 1
o
-2.2.6 Because the tasks described in this methodology are essential for providmg the justification for the safety finding of $54.29, these tasks are performed in acwid.nce with the BGE
"]
quality assurance program 2.2.7 Structural en=pa==ts and components, which contribute to one or more passive functions and are long-lived, require evaluation to demonstrate that the effects' of aging are 1
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 i
concern associated with degraded equipment condition.. 'Ihis 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 Methodoloav 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 Scooina 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.
j 2.3.2 Comoonent level Scopina 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 perfonn its intended functions. 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-Evalugigi 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 fbrtherIPA resiew:
Those which contribute only to active functions; Those which are replaced based on time or qualified life; and Those which are replaced on the basis of a condition-based program.
(Justification of the adequacy of such a replacement program is included in the LRA.)
The result of this task is the list of all SCs in the given system which will be subject to AMR.
2.3.4 AMR The AMR task (Section 6) demonstrates that the effects of aging are adequately managed (see Definitions).
Several different techniques for developing this justification are presented in this section. All the techniques provide an equivalent level of assurance to support the finding of $54.29 with respect to the management of effects of aging.
2.3.5 Commodity Evaluations 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 TLAA Review 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 whhh also meet certain other specified criteria. For the identified TLAA, the justificatbn 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 l
is not technically part of the IPA, but its description is included in the IPA Methodology for convenience.
12 Revision 0 L
ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 2-2 SOURCE DOCUMENTS His 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.
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 Pan 100, Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants,"
January 1,1991 1
5.
Calvert Cliffs Nuclear Power Plant, Units I and 2, Updated Final Safety Analysis Report, Revision 17, November 1994 6.
Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Technical Specifications Manual, through Amendment 205 (May 1995) for Unit 1, and Amendment 183 (April 1995) for Unit 2 7.
CCNPP Design Standard, ' Structure and Component Evaluation," (DS-011) Resision 0, June 7,1995 8.
CCNPP Design Standard ' Control of Equipment Technical Databases,"(DS-032) Revision 0, January 25,1995 8.
CCNPP System Descriptions. (various revisions) 9.
NRC Regulatory Guide 1.97, " Instrumentation for Light-Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Resision 3 10.
CCNPP Plant Drawings (various) j 11.
NUREG-1377 " Listing of Nuclear Plant Aging Research Reports," and the reports themselves 12.
Industry Technical Reports on PWR Reactor Vessel, PWR Reactor Vessel Internals, PWR Containment, PWR Reactor Coolant System, Class 1 Structures and Environmentally-Qualified Cables in Containment 13 Resision 0
l ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY i
3.D SYSTEM LEVEL SCOPING l
nis section describes how all plant SSs are reviewed to determine those that are within the scope of LR. His is accomplished through application of the system-scoping process (Figure 3-1).
Determining which SSCs are within the scope of LR is the first major task described in the IPA methodologv. Section (54.21(a)(1) of the LR Rule states that the IPA must be conducted -
l For those systems structures and components within the scope of this part, as delineated in f54.4,...
1 In other words, the results of the system level and component level scoping tasks are the starting point of the IPA.
l t'
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 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.
l 3.1 Identification of SSs t
The SS listing for CCNPP is provided in Table 3-1. The CCNPP De gn Standard for " Control of the Equipment Technical Databases," (See Table 2-1, Reference 8) was sed 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 stmetures 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.
i 14 Resision 0
ATTACliMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT t
INTEGRATED PLANT ASSESSMENT METIIODOLOGY System Level All plant SSs Scoping Process t
Derine conceptual boundaries and functi nalrequirements Structures Systems o
Develop screening tools o
f 54.4(a)(1) 54.4(a)(2) 54.4(a)(3)
Criterion Criterion Criterion DBE Vital FP,EQ, Flow Charts Auxiliaries ATWS, SBO, Tool PTS Tools v
v v
fstructure required is systew is systew is sntew "g
structure reqQed structure required n
C W 1 h ue?
by the tool?
by the tool?
by the tool?
Yes Yes Yes Yes 1
A 1
1 Add Function to Add Function to Add Function to Add Function to Intended Functions-intended Functons-Intended Functions-Intended Functions-.
Ust List List Ust l
I I
I Ust of intended r
functions for SSs I
oes ystem or structu
~
have an intended Y
function?
SSs within the Scope of License Renewal 1
No further action required for these SSs Figure 3-1 i
t 15 Revision 0
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 3.2 -
Define Concentual Boundaries j
His step of the system level scoping process tabulates some basic information about each of the -
SSs listed in Table 3-1. His information, referred to as the 'tonceptual boundaries"of the SS, is.
needed to ensure a consistent understanding of what is meant by each of the SS names in this table.
De 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 SSs listed in Table 3-1, a brief system description is developed and the functional requirements are
. identified. The description includes a listing of the major components and major rystem 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 e
intended function is identified, the conceptual boundaries allow the evaluator to determine which SS the intended function should be associated with. The list of functional requirements does not represent a detailed list of intended functions, but it is sufficient to establish the conceptual boundaries of SSs. He component level scoping task (described in Section 4) develops a detailed list of SS intended functions.
j He following information is compiled for each SS and entered into a table designated as Table 1,
" System / Structure Information":
1 System or structure name; Unit number; Identification number, Brief description; including major components and system interfaces; Source document reference (for the description); --
System or structure functional requirement (s); and Source document reference (for each functional requirement).
3.3 Screenine 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. He tool contains a list of SSCs which respond to the event and their intended functions.
De source documents identified in this section are reviewed against the 554.4 criteria contained in the LR Rule. For each criterion, appropriate information is taken from the source documents and -
j summarized in one or more screening tools. The tools are then used to complete the screening i
process.. Each tool is described below. An example of a portion of a screening tool is provided in
]
Table 3-2.
i i
l 1
i 16 Revision 0
.~
l,
' ATTACHMENT (1)
~l i
CALVERT CLIFFS NUCLEAR POWER PLANT i
INTEGRATED PLANT ASSESSMENT METHODOLOGY
- l 3.3.1 Tools Addressina 654.4(a)(1) and (2) l l
' 10 CFR 54.4(a)(1) and (2) (referred to as {54.4 Criteria 1 and 2) are addressed together in
]
~
the System Imel Scoping process since both of these criteria were used to establish the
.CCNPP Q-List documentation.
554.4 Criterion 1 l
o (1) Safety-related systems, structures and components which are those relied on to remain functional during andfollowing design-basis events [as defned in 10 CFR 50.49 (b)(1)] to ensure thefollowingfunctions -
d (t) :
The integrity ofthe reactor coola :tpressure boundary; (ii)
The capability to shut down the reactor and maintain it in a safe shutdown condition; or (III)
The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure
~
comparable to the 10 CFR Part 100 guidelines.
554.4 Criterion 2 i
c (2) All nonsafety-related systems, structures and components whose failure could prevent satisfactory accomplishment of any of the functions identiped In
.l paragraph (a)(1)(t), (ii) or (iii) ofthis section (i.e., f54.4).
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 j
external events are analyzed in Chapter 14 of the CCNPP UFSAR. All structures
)
designed to withstand DBE external events are designated as Class 1. structures at CCNPP, and Class 1 structures are included within the scope of LR (Section 3.4.1.2).
17 Revision 0
i ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Desian Basis Event Chaoter 14 Location Control Element Assembly (CEA) Withdrawsl Event Section 2 Boron Dilution Event Section 3 Excess Load Event Section 4 Loss of Load Event Section 5 Loss of Feedwater Flow Event Section 6 Excess Feedwater Heat Removal Event Section 7 RCS Depressurization Section 8 Loss of Coolant Flow Event Section 9 Loss ofNon-Emergency AC Power Section 10 Control Element Assembly Drop Event Section 11 Asymmetric Steam Generator (SG) Event Section 12 CEA Ejection Section 13 Steam Line Break Event Section 14 SG Tube Rupture Event Section 15 Seized Rotor Event Section 16 Loss of Coolant Accident Section 17 Fuel Handling Incident Section 18 Turbine-Generator Overspeed Incident Section 19 Containment Pressure Response Section 20 Hydrogen Accumulation in Containment Section 21 Waste Gas Incident Section 22 Waste Evaporator Incident Section 23 Maximum Hypothetical Accident Section 24 Excess Charging Accident Section 25 Feed Line Break Event Section 26 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 Q-List Accident Shutdown Flow Sheets and any additional supporting systen's identified as relied on for that accident in UFSAR Chapter 14.
For the eight DBEs which are identified in the UFSAR and are not the subject of Q-List Accident Shutdown Flow Sheets, a DBE flow chart is prepared 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).
3 The terms *Q-L'at Accident Shutdown Flow Sheet" and " Vital Auxiliaries Flow Sheets" are used to refer to documentation which already exleted as part of the CCNPP Q-List. The terms *DBE Flow Chart" and " Vital Auxuiaries Screening Tool" are used to derde the document created during the scoping process to compile the Q-List information and other specified information.
18 Revision 0
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY Reason WhfNoWecidest Shutdown " ?Infojmation Includedin Scoping 1
- Flow Sheet is in the Q-List
- Results DBE Flow Chart-'
- No active components are relied on to Passive components which mitigate mitigate the event.
the DBE.
No active or Passive components are A note stating that no active or passive required to mitigate the event.
components are required to mitigate the event.
All components relied on for the event A note stating that all components are already included in another Accident required to mitigate the event are Flow Sheet.
included in another DBE Flow Sheet, and specifying which other DBE(s).
The DBE flow charts for the remaining 17 DBEs identify the systems and the functions provided by each of these systems in order to upport the CSFs necessary to reach safe shutdown for the specific DBE, maintain the fission product barriers, and prevent offsite releases in excess of established guidelines.
Q-List documentation also contains a specific flow sheet for VAs. Electric power distribution; control air; cooling water; and heating, ventilation, and air conditioning functions for the SR equipment required to respond to each DBE are annotated in the-corresponding Q-List Accident Shutdown Flow Sheet. The Q-List Vital Auxiliaries Flow Sheet is a compilation of the systems performing these VA functions for all of the Q-List Accident Shutdown Flow Sheets. 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 functions 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 1
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 i
controlling a similar scope of SSCs to that defined by the first two criteria of {54.4.
'l 4
Therefore, the CCNPP scoping process is consistent with the Commission's intent stated in the SOC to the LR Rule.
1 I
i 4
The CCNPP Q-List documentation also establishes controls for PAM (Category 1 and 2) equipment. Post-Accident Monitoring
(~
equipment satistes {54.4 Crttenon 3, rather than 1 or 2.
1 19 Revision 0 l
L
- g ATTACHMENT (1)
/...
p CALVERT CLIFFS NUCLEAR' POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY
' An applicant for LR should rely on the plant's CLB, actual plant-specifc experience, industry-wide operating experience, as appropriate, and existing engineering enluations to determine. those NSR systems, structures, and
. components that are the initialfocus ofthe LR review. (60 FR 22467) 3.3.2 Tools Addressina 654.4(aV3i
$54.4 Criterion 3 (3)
- All systems, structures and components relied on in safety analyses or plant enluations to perform a function that demonstrates compliance with the Commission's regulationsforfire protection (10 CFR 50.48),
environmental qualspcation (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62),
andstation blackout (10 CFR 50.63).
Plant evaluations have been performed to L. =+.te compliance with the regulations identified in 654.4(a)(3) (referred to as 054.4 Criterion 3). 'Ihese evaluations are reviewed to identify SSs that are relied on to mitigate the subject plant event as well as any systems or structures whose failure would result in failure of other equipment to mitigate the particular event. As was the case for Criteria 1 and 2, an SS is listed as within the scope of LR, when the mitigation function or support function associated with it is credited in the -
analysis or evaluation.' Mentioning an SS in the analysis or evaluation does not necessarily indicate that the SS contributes to an intended function.
Additionally, if the SS function is identical to a SR function (as identified in the Q-Lim),
then the function need not be repeated on the tools addressing 054.4 Criterion 3. ;'Ihe analyses and evaluations being reviewed in this step are used to identify intended, NSR functions.
3.3.2.1 FP Screenine Tool Preparation
'Ihe 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.
3.3.2.2 EO Scre.mina Tool Preparation Two tools are produced for this criterion, the EQ tool and the post-accident monitoring (PAM) tool.
20 Revision 0
]
L j
ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 1he Q-List data in the NETD is reviewed to identify items listed as 5049 (items which must meet the requirements of 10 CFR 50.49). A list of the systems containing components designated as EQ is prepared with the Q-List revision number (or date,' as appropriate) provided as a reference.
The CCNPP UFSAR is resiewed to identify the systems containing components required for PAM category 1 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.
3.3.2.3 PTS Screenine Tool Preparation Since neither CCNPP Unit I nor 2 is expected to require an evaluation in accordance with I
Regulatory Guide 1.154 in order to satisfy 10 CFR 50.61 requirements, no equipment is included within the scope of LR due to the PTS Rule. The PTS Screening Tool is provided in the System Level Scoping Results, but this tool merely notes that no SSCs are relied on for this event. Additionally, the System Level Scoping Results, the component level scoping process, and the component level scoping results for each system include the contingency to implement a PTS scoping criterion, but the results indicate no PTS-related SSCs. If a Regulatory Guide 1.154 evaluation is required at some point in the future, the scoping process would be modified to require incorporating the PTS functions relied on in the 1.154 analysis into the PTS Screening Tool. The Regulatory Guide 1.154 analysis would also trigger an update to the system level and component level scoping results to include the SSCs associated with the 1.154 functions within the scope of LR.
3.3.2.4 ATWS Screenine Tool Preparation The CCNPP UFSAR is reviewed to identify the system functions that address the 10 CFR 50.62 requirements on ATWS. An ATWS Screening Tool is developed. The tool lists the SSCs which are relied on in response to an ATWS event. For each identified SS, the tool lists the intended function (s) provided and the appropriate source documents with the revision number.
I 3.3.2.5 SBO Screenine Tool Preparalign The Station Blackout Analysis is resiewed to identify SSs which are relied on during the
" coping duration" phase of an SBO event. An SBO Screening Tool is prepared which lists
'I the SSs relied on in the Station Blackout Analysis, the function (s) that each prmides, and the appropriate source documents with revision numbers. The power restoration phase of the Station Blackout Analysis is specifically excluded from review in this criterion since several success paths for restoring power after an SBO are already screened as within the scope of LR due to Criterion 1 (SR).
4 21 Revision 0
ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 3.4 SS Sconine ne 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 Table. Since the events summarized by the tools address the requirements of the 654.4 criteria, inclusion of an SS in a tool indicates that it is within the scope of LR. It is important to note that all intended functions are identified for each SS during the scoping 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.
3.4.1 Criteria 1 and 2 - SR and SR Support SSs 3.4.1.1 DBE Flow Charts and VA Screeninn Tool The DBE flow charts and the VA screening tool, (see Section 3.3.1.1), are used to identify those SSs whose functions support the CSFs for a DBE, or whose failure would prevent performance of the CSFs. Systems and structures listed in one or more of the DBE flow charts or the VA screening tool are included in the System Level Scoping Results Table under Criteria 1 and 2. For each SS listed in the results table, all applicable DBEs are identified along with the functions that the SS provides for each DBE. The source j
document references and revision numbers are not included in the scoping results table since this information can be found in each DBE flow chart or the VA screening tool.
3.4.1.2 Class 1 Structures For all listed structures, the UFSAR Section 5 and Q-List Design Standard are resiewed to determine whether the structure or a portion thereof is designated as SR, Class 1. At CCNPP, all Class I structures (buildings) are designated as SR; therefore all Class 1 structures are screened as within the scope of LR. The results of this scoping step are incorporated, along with the appropriate source document references and revision numbers or dates, into the System Level Scoping Results Table for each of the structures.
3.4.2 Criterion 3 - SSs Relied On in Plant Safety Evaluations ne 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; 2)
Rose which contain components identified as EQ or PAM; 3)
Those whose functions are relied on in plant event evaluations for ATWS, SBO, and PTS; or 22 Revision 0
ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 4)
Any combination of these factors.
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. 'Ihe 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 Results 3
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.
f i
)
1 1
23 Revision 0
ATTACilMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES 1
Switchyard (500 kV) & Switchyard DC 48 Engineering Safety Feature Actuation 2
Electrical 125VDC Distribution 49 Simulator Computer 3
Electrical 13kV Transformers & Buses 50 Solid Waste Disposal 4
Electrical 4 kV Transformers & Buses 51 Plant Water 5
Electrical 480V Transformers & Buses 52 Safety injection 6
Electrical 480V Motor Control Centers 53 Plant Drains 7
Electrical 13kV Unit Buses 55 CEA Drive Mechanism & Electrical 8
Well and Pretreated Water 56 Reactor Regulating l
9 Intake Structure 57 Technical Support Center Computer j
11 Service Water Cooling 58 Reactor Protective 12 Saltwater Cooling 59 Primary Containment 13 FP 60 Primary Containment Heating & Ventilation 14 Transformer Deluge 61 Containment Spray 15 Component Cooling Water (CCW) 62 Control Boards 16 Electrical 250VDC 63 Cathodic Protection 17 Instrument AC 64 Reactor Coolant 18 VitalInstrument AC 65 Seismic 19 Compressed Air 66 Cavity Cooling 20 Data Acquisition Computer 67 Spent Fuel Pool Cooling 21 Domestic Water 68 Spent Fuel Storage 22 Makeup Demineralizer 69 Waste Gas 23 Diesel Oil 70 Refueling Pool 24 Emergency Diesel Generator 71 Liquid Waste 25 Access Control Area Ventilation 72 Sewage Treatment Plant 26 Annunciation 73 Hydrogen Recombiner 27 Auxiliary SGs 74 Nitrogen and Hydrogen 28 Auxiliary Steam 75 Low Voltage DC Control Power 29 Plant Heating 76 Secondary Sample 30 Control Room Heating, Ventilation 77/79 Area / Process Radiation Monitoring
& Air Conditioning 78 Nuclear instrumentation 31 Meteorology Tower & Miscellaneous 80 New Fuel Storage and Elevator Computers 81 Fuel Handling 32 Auxiliary Building and Radwaste 83 Main Steam Heating & Ventilation 84 Reactor Vessel intemal 33 Turbine Building Ventilation 85 Plant Access and Surveillance 34 Condensate Precoat Filter 86 Power Plant Security 35 Chemical Additions-Turbine 87 Unit Transformers 36 Auxiliary Feedwater (AFW) 88 Visitor Center Security 37 Domineralized Water and Condensate 89 Emergency Operatior.s Facility Security Storage 90 Service Building & Outlying Building 38 Sampling System Heating, Ventilation & Air Conditioning 39 Condensate Polishing Demineralizer 91 Lube Oil Storage 41 Chemical and Volume Control 92 Gland Steam 42 Circulating Water 93 Main Turbine 43 Condenser Air Removal 94 Plant Computer 44 Condensate 95 Carbon Dioxide 45 Feedwater 96 Fire and Smoke Detection 46 Extraction Steam 97 Lighting and Power Receptacle 47 Feedwater Heater Drains and Vents 98 Main Generator and Excitation i
1 24 Revision 0
. ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOIDGY '
TABLE 3-1 CCNPP SYSTEMS AND STRUCTURES (Continued) 99 Cranes / Test Equipment 105 Weight Testing Wire Ropes & Slings (3) 100 Plant Communications 106 Ladders and Gratings (3) 101 DryFuelStorage 107 Roads 102 Plant Areas 108 Docks and Marine Related Structures 103 Emergency Diesel Generator Building 109 Shop Equipment (3)-
Heating, Ventilation & Air Conditioning (2) 110 Manual Valve Components (3) 104 Lubrication 111 Materials Processing Facility (3)
AdditionalStructures Auxiliary Building.
Condensate Storage Tank No.12 Enclosure Domestic Water Treatment Plant Engine Generator House Equipment Hatch Access Building. No.1 Equipment Hatch Access Building, No. 2 FP Pump House Fuel Assemblies Fuel Oil Storage Tank No. 21 Building.
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 Turt>ine Building Waste Water Treatment Building.
Well Observation Building Well Water Pump House Independent Spent Fuel Storage installation (4)
Diesel Generator Building 1 (2)
Diesel Generator Building 2 (2)
NOTES:
1.
System listing is from Attachment 6 of DS-032, Control of the Equipment Technical Databases" 2.
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 Part O l
and, therefore, is not in the scope of this LRA.
)
1 25 Revision 0
ATTACHMENT (1) j i
CALVERT CLIFFS NUCLEAR POWER PLANT I
INTEGRATED PLANT ASSESSMENT METHODOLOGY
.i TABLE 3-2 Revision 4 Post-Accident M'onitoring Screening Tool (Example) i Reference 1 -
Calvent Cliffs Nuclear Power Plant, Units 1 & 2, Uodated Final Safety Analysis Reoort (UFSAR), Section 7.5.8 i
Reference 2 -
Calvert Cliffs Nuclear Power Plant, NUCLEIS Equipment Database
]
l i
SYSTEM /
SYSTEM STRUCTURE ID No.
MONITORING VARIABLE (S)/ FUNCTION (S) i Electrical 125VDC 2
. Status of standby power (voltage, current) i Distribution Electrical 4kV 4
Status of standby power (voltage, current)
Transformers and Buses Electrical 480V 5
Status of standby power (voltage, current) i Transformers and Buses Service Water 11 Service water pump status (motor current)
'i a Containment cooler cooling water flow Saltwater 12 Saltwater pump status (motor current)
Component Cooling Water 15 CCW heat exchanger outlet temperature
. CCW to/from reactor coolant pumps containment isolation valve position CCW pump discharge pressure (for flow indication) j CCW pump status (motor current)
VitalInstrument AC 18
. Status of standby power (voltage)
Compressed Air 19 Instrument air containment isolation valve position indication Data Acquisition 20 Provide fault protection for Instrumentation & Controls Computer loops 1
Emergency Diesel 24 Status of standby power (voltage, current, VAR, frequency)
Generator Auxiliary Building &
32 Fuel pool exhaust fan damper position Radwaste Heating &
Ventilation AFW 36
- Motor-driven AFW pump status (motor current)
Condensate storage tank 12 level Sampling System 38 Containment hydrogen concentration 26 Revision 0
RsTTACHMENT (1)_
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 3-3 BGE LCM PROGRAM TABLE 2 SYSTEM LEVEL SCOP 180G REstJLTS (EXAMPLE)
Revistop 4 CRITERIA 1 & 2 CRITERtON 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 iM Reference PAM FP ATWS 580 PTS EQ Yesseo S.a.;.,e; (500 kV) 1&2 1
No None No N/A N/A No No No No No No No and S ^^ ^.1 DC Electrical 125 VDC 1&2 2
VA VA for Chemical & Volume Control System No N/A N/A Yes Yes No No No No Yes Drotritnd6cn VA for AFW VA for Mein Steam VA for Containment Spray l
VA for Pnmery Containment Hesting &
Vertiistion VA for Emergency Diesel Generators VA for 4KVTransformers & Buses VA for 480V Motor Control Centers i
VA for 480V Bus System 1
VA forVitalInstrumerd AC VA for Service Water VA for CCW VA lbr Seitwater Cooling VA for Control Room Hesting, Ventilation
& Air Condtioning i
VA for Auxiliary Building & Redweste Heating & Ventilation l
VA for RCS VA for Emergency Se%ty Features Actus-tion System Lead Shedding VA for Chemical & Volume Control System.
(Core Flush)
Electrical 13kV 1&2 3
No None No N/A N/A No No No No No No No Transformers and Buses Electncel 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 inged6on vA for Containment Spray VA for 480V Bus VA for 480V Motor Control Centers VA for Serv 6ce Water VA for SW Cooling VA for Emergency Safety Features Actus-tion Syulem Load Shedding Electncel 480V 1&2 5
No N/A N/A Yes Yes No No No No Yes 27 Revision 0
l ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT l
INTEGRATED PLANT ASSESSMENT METIIODOLOGY j
j 4.0 COMPONENT LEVEL SCOPING I
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 $54.4.
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 Component Level Seonine 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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)
Component Intended functions for the system being scoped Level Scoping Process for Systems DBE Flow Charts PAM, SBO, FP, PTS,
. Describe intended function ATWS, EQ Screening in more detailif needed.
Tools Other implicit intended functions; e g., PB,1E, structural support.
Consolidate functons to eliminate duplicates v
[
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MEL for the System For allintended
~" functions of the system System Level Scoping u
Results & References List all system Function catalog 01 components which are required to perform the
~
Function catalog 02 function or could fall Plant drawings and prevent the a
y function a
e U
Q-Ust documentation Function catalog n Next intended function
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l Operating Instructions l
l Resort function catalogs by component y
List of system j
components and si theirintended
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function (s),
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vseca m asessesa Figure 4-1 29 Resision 0 i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIf0DOLOGY Component Level Scoping l
for Structures oes the structure have tem type components
{
Perform component level scoping using the l
Q*,
system process for system type identify structure intended function components.
- Structural support to SR equipment
- Shelter / protection for SR equipment
- Pressure or fission product bour dary
- Missile barrier
- Class 11/1 support
- Flood protection barrier
- Rated fire barrier v
Determine ceaeric structural component types in this structure.
v Add unique structural component types.
v Identify structural component types which contribute to each intended function.
v Add supports for large SR equipment to scoping results.
j v
l Integrate scoping results for system type and structural type components.
i 1P List of structural y
component types j
and their intended Q
f Figure 4-2 functions l
, g.
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1 ATTACHMENT- (1) q
-l CALVERT. CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSF.SSMENT METHODOLOGY-4.1.1 Identification of Detailed System Functions 1
. De purpose of this step of the scoping process is to create a detailed list of the intended functions associated with the system being scoped De list is compiled in a System; 1
~
Functions Table using the System and Structure Scoping Results, Q-List &==- 2-4 j
plant drawings, the UFSAR, System Descriptions and other references.
i ne System and Structure Scoping Results contain screening tools which associate l
im functions with individual systems. The first substep of creatmg tha detailed l
function list is to review all of the screemns tools and, in the System Functions Table, j
record the intended functions of the system being scoped.
l The CCNPP Q-List Design Standard (Table 2-1 Reference 8) is the site reference which i
governs what components are controlled as SR, SR support, or other miscellaneous category equipment.
To ensure consistency with the Q-List documentation, the i
LCMEVAL software application is used to compile a listing of all Q-List categories which are associated with any components in the system being scoped (Q-List Criteria listing).
This listing represents the Q-List related functions associated.with the system being.
j scoped The following Q-List categories correspond to (54.4 criteria as described below:
i Q-List Flow Sheets -
These flow sheets identify components which are relied on to respond to UFSAR
]
Chapter 14 DBEs or serve as VA to SR equipment.l Criteria 1 and 2.
PB-The categocy of PB mechanical items which maintain the systczn PB of the RCS, maintain the radiological boundary to preyc at exceeding 10 CFR Part 100 limits, l
or maintain safety system boundary to limit system leakage. l Criteria 1 and 2.
i (Criterion 2 because PB includes the components needed to maintain the PB of fluid systems which are not fission product 'ooundary fluid systems.)
IE-The category of electrical equipment and systems that are essential to emergency 1
reactor shutdown, containment isolation, reactor core cooling, and contamment and reactor heat removal,.or otherwise are essential in preventing significant release of radioactive material to the environment. Criteria 1 and 2. - (Criterion 2 because 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-ne category of mechanical equipment that is essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment'and reactor heat removal, or otherwise are essential in preventing significant release of radioactive material to the emironment.. Criterion 1.
PAM - Post-accident monitoring category ofinstrumentation used to assess the emirons.
j and plant conditions during and following an accident. Criterion 3, subset of -
1 EQ.
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INTEGRATED PLANT ASSESSMENT METIIODOLOGY
$049 -
His 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.
After producing the Q-List Criteria Listing for the system being scoped, this list is consolidated with the functions already listed in the System Functions Table to finalize the detailed functions listing for the system. The Q-List does not contain information related f
to several of the regulated events in (54.4 Criterion 3. Herefore, for the categories shown below, no consolidation with Q-List-related functions is possible.
He associated screening tools and their references are used to validate the detailed system function (s) for these criteria.
FP -
The functions required by 10 CFR 50.48 for FP and safe shutdown after fire.
ATWS - The functions required by 10 CFR 50.62 to provide diven 1 :/am and diverse turbine trip capability during an ATWS event.
PTS-The functions required by 10 CFR 50.61 to proside protection during a PTS event.
%c fmal step of intended function identification is to eliminate redundant functions.
Functions enveloped by another function or identical to another function are consolidated.
He 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 cmated the table. Parent functions and functions for which no consolidation is possible are assigned a unique identification number (Function ID) to facilitate subsequent steps in the scoping process.
(For the remainder of this methodology, the term " intended function" refers to a parent function unless otherwise specified.)
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I 4.1.2
'!he MEL 1
To ensure that all co.v.ye.aas in the plant are scoped with one and only one system, the i
site MEL is used to provide the equipment list for the component level scoping task for -
1 each system. 'Ihis list is the portion of the NETD which contains all equipment for a given system.
3 In developing the NETD, conventions were established for determming the boundaries
- between systems. These conventions provided the guidance for determuung which system each ec.r.isa; in the IPA would be assigned to. Several example conventions are listed -
below. The complete system boundary guidelines are contained in the site design standani i
for controlling equipment technical databases I
Heat exchangers are assigned to the load system.
Electrical components are assigned to load system from the load side of the circuit breaker.
2 Sensors are assigned to the system in which they sense. Actuators are assigned to
. the system in which the actuation takes place.
Transfonr. :: e,rc assigned to the lower voltage system.
'l i
As cach scoping task is begun, the LCMEVAL software application is loaded from the ~
l 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 l
listed in the System Functions Table or not needed for any of these functions.
]
1 4.1.3 Develooment of Function Catalons The next step in the component level scoping process for systems is to determine, for each intended function, which components from the system MEL are needed to perform the l
functioc. A list of components for each function is called the function catalog.
I l
In order to determine the relationship between a given function and the components contributing to the function, Q-List documentation, UFSAR, Technical Specifications, system screening tools and references associated with the screening tools are usedi The active components associated with mitigating the consequences e mdividual DBEs or providing VA functions to SR equipment are listed in the plant Q-List documentation along with a refereme to their safety function (s). Consequently, whenever a System Functions Table contahs 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 154.4 Criterion I and 2.
.l 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 33 Revision 0 i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY function for that ponion of the system relied on to mitigate DBEs. Whenever the system j
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 ponion 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 eatalogs associated with other functions in order to avoid redundancy.
He Q-List documentation also contains listings which associate specific components to PAM and EQ functions. His listing is used as a direct input to the scoping 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. His catalog nonnally contains electrical panels (EPs) and other enclosure devices which contain SR equipment but have no explicit active safety function.
Many clectrical 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 suppon of a number ofimponant events but are not actually associated with any panicular DBE in the Q-List documentation. When a system contains components that are SR and designated only as 1E or 1M, a separate function catalog is created to contain these components.
The NETD contains a field which associates specific components with the Station Blackout Analysis. This SBO designation is used as an input to scoping for SBO and further review is conducted during the IPA process as described below:
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 j
to the SBO function catalogs.
Much of the equipment mentioned in the Station Blackout Analysis is mentioned i
because it is secured at the start of an SBO event or is used when restoring power after the end of the event. These components do not contribute to any SBO 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.
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1 When the process is complete, the SBO function catalog or catalogs contain all of the system components which contribute to each intended SBO function.
l ne 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 i
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 i
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 criteria as they perform to support a SR function, are not repeated again in these function catalogs to avoid redundancy. For example, if a pump is required to start during a severe l
fire to ensure plant shutdown and the same pump must start to provide cooling water to SR l
equipment to mitigate the consequences of a DBE, that pump would not be repeated in the FP function catalog.
All of the function catalogs discussed above are created using the LCMEVAL software system which contains data loaded directly from a controlled site database (NETD) where possible. For the functions where no source of direct component data is available in software format, the individual components are entered one at a time into the function catalog. The software ensures that only valid components (i.e., in the MEL for the system j
being scoped) are added to function catalogs. It also facilitates the recording of reference documents which justify that a component supports a given function.
l 4.1.4 Generation offcooina Results Table In the next step of the component level scoping process for systems, the function catalogs that were developed in Section 4.13 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 l
scope of LR by the LCMEVAL software system. The table ofin-scope components and the intended functions that they contribute to is designated as the Component Level Scoping Results Table.
4.2 Component Level Seonine for Structures l
The component level scoping process described above for systems can also be applied to structures. However, this process is somewhat different because of the unique features of l
structures and how they are documented on site. As with systems, the scoping process is implemented by d, xnnining which structural components are required for the performance of the intended functions of the structure. Details of the methodology implementing the structural component scoping are presented below.
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i CALVERT CLIFFS NUCLEAR' POWER PLANT.
INTEGRATED PLANT ASSESSMENT METHODOLOGY 4.2.1 Umque Identifiers for Components
,j 1he components of structures have not scnerally been identified and listed in an MEL.
Cca i.a:ly, the component level scoping for structures cannot use a comprehensive :
.I equipment listing as an input.'
j For certain site structures, such 'as the contamment,' specific w..wst types have been f
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:
j 1)
"Ihe 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 contamment personnel hatch, the personnel hatch limit switches and the' contamment
.j penetrations are scoped using this process because they are designated as 1
cc.r.pcwa of the containment system in the NETD.
l 2) 1he 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.
j I
4.2.2 Function Identification 1
1 The SS scoping process identifies some structures as within the scope of LR because they -
I are designed to Class I criteria or because they are required for DBE purposes Unlike the ~
scoping n:sdts for systems, the Class 1 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 l
determine specific structure design basis information such as which external ' events the j'
structure is designed to withstand, and which stmetural 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 structures determined that a structure, or components of the structure are designed to perform one or more of the following functions in support of the.
{54.4 criteria:
1.
Provide structural and/or functional support to SR equipment; 2.
Provide shelter / protection to SR equipment.. (This function includes radiation protection. for EQ equipment and high energy line break-related protection equipment.);
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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 functions (Example: seismic Category II over I design considerations);
6.
Provide flood protection barrier (internals flooding event); and 7.
Provide a ratd fire barrier to confme 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 stmetural 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.
He applicability of each fimetion to the structure is detennined by a resiew of various source documents. If the structure is a Class I structure, the UFSAR and the System and Structure Scoping Results must be referenced to determine which of functions 1-4 apply.
He 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 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.
5 External nooding events were considered during the design process for CCNPP structures. It was deterrnined that a probable maxirnum hurricane would cause the worst-case flooding conditions at the site. The resulting surge and wave action was analyzed q
as the basis of plant flood protection. The effects of possible wave action were studied using a hydraulic rnodel.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 4.2.3 Structural Component Tvoc Listina for the Structure In the structural component scoping process, components that are structural in nature are not uniquely identified during the scoping process. For example, each wall in the structure is not identified, named, and listed. Rather than using an MEL of named structural components, the scoping is conducted on a generic listing of structural component types.
his generic list was developed by experts in the field of nuclear Class I structures. The generic list started with structural component types 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 flooding-related component types were not considered in these reports.)
He evaluator uses this generic component listing and determines which of the component types on the list are actually contained in the structure being scoped. His step is performed by reviewing plant architectural drawings and identifying the specific structural types. Additionally, any structural component types which are unique to the particular structure being scoped, such as the prestressed tendons in the containment and the sluice gates in the intake structure, are noted. These unique structural component types are then added to the list of applicable structural component types. This list serves as the equivalent of an MEL for structural component scoping task.
4.2.4 Structural Components Which Contribute to Intended Functions 3
This section describes the process used to determine which component types of a structure contribute to the intended functions which the structure performs. For every function i
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 i
and Structure Scoping Results (including documents referenced by these results). He component types which contribute to each intended function are recorded on the
" Structural Components Which Contribute to Intended Functions" table.
Additionally, the supports for large SR equipment within the structure are identified by resiewing a listing of the SR equipment installed in the structure that might affect the design of the structure (such as tanks, heat exchangers, or vessels filled with fluid and pumps which require a pedestal as a foundation.). These SR equipment supports are also included in the " Structural Components Which Contribute to Intended Functions" table.
Q-List documentation and the Flooding Design Ouidelines Manual are resiewed to determine if structural component types in the structure being scoped are relied on to contribute to the functions of providing structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory accomplishment of any of the required SR functions or providing flood protection barriers. If structural component types in the structure being scoped are determined to contribute to these functions, then this information is captured by recording "YES" in the " Applicable to This Structure?"
column of the Structural Intended Functions Table. The components that contribute to 38 Resision 0
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ATTACHMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY these functions are then recorded on the "Strr:tural 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 Scopine 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:
l' 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.
i j
TABLE 4-1 Scoping Part ofi
/ Conunoditp Evaluat' ion
' Commodity Evaluation?
j EPs & Related Equipment No Instrument Lines (ILs)
No Cables Yes Cranes and Fuel Handling Equipment Yes Component Supports Yes FP Systems Yes 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 I
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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J CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 5.0 PRE-EVALUATION
'Ihis 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 (except for equipment covered by the commodity evaluations which replace the entire IPA process, as described in Section 4.3). The description provided in Sections 5.1 through 5.3 of the methodology applies primarily to systems. Section 5.4 describes the differences in the process as it is applied to structures.
The input to this task is the results of the component level scoping step, described in Section 4, for the system being evaluated. These results consist of the intended functions of the system or structure being evaluated and a designation of which portions of the system or stmeture 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 stmeture must be further evaluated for the effects of aging. The SCs or groups of SCs determined not to be subject to AMR require no further evaluation in the IPA process.
The output of the Pre-Evaluation task is the list of SCs which need to be evaluated further for the efrects of aging in the AMR task.
The Pre-Evaluation task is governed by (54.21(a)(1) of the LR Rule.
54.2)(a)(1) For those systems and structures within the scope of this part, as delineated in $$4.4, identify and list those structures and components subject to an AAfR. Structures and components subject to an aging management review shall encompass those structures and components -
(i)
That perform an intended function, as described in 554.4 withoat movingparts or without a change in configuration orproperties. These stro:c~ res and components include, but are not limited to, pressur o retaeng boundaries, component supports, reactor coolant pressure boaubries, the reactor vessel, core support structures, containment, wsmic 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 toperiodic 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 1
/
\\
i Functional Scoping
--+
For all passive SCs Results
/
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SCs For allintended functions of the systems subject to replace-ment at set frequency Yes-or qualified hfe<4 years?
Does No further IPA review function No required for these SCs involve motion or chang
'n properties or configur ["}
E0" Of i
SCs SCs include No further IPA subject to regular justification No review required.
y, eplacement basfe in LRA.
No Y"
n conditj n
Add SCs to list of Does the condition passive SCs for the system ased replacemen s program assure intended No function l
v e
h*
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/
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Add SCs to system list i
Yes of passive, LL SCs 1
1 SCs Subject List of passive SCs for
/
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to Aging 4-the system Any passive SCs
+
remaining?
Management l
Review m
o re-eval Complet t
u System Commodity Aging Aging Managemeni Evaluation Review Figure 5-1 41 Revision 0
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 5.1 Catenorize Intended System Functions as Active or Passive 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 ftmetion 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.
J 5.1.1 Passive Functions i
Passive functions are those which require no motion or change in SC configuration or propenies to carry out the requirements of the ftmetion, 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 PB of a fluid system.
I Provide structural support or shelter to equipment.
i Provide missile protection.
Provide shielding against radiation.
Proside shielding against high energy line breaks.
Provide flood protection.
Prevent or isolate faults in an electrical circuit when such protection or isolation j
does not involve motion or change in properties or configuration. (e.g., cable j
insulation).
1 Any function which is detennined to be passive is evaluated in Section 5.2 of the methodology.
5.1.2 Active Functions Active functions require motion or a change in SC propenies or 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:
Provide required flow to a heat exchanger.
Provide electrical signals to a desice.
Provide electrical power to a bus or load.
Provide indication of a plant condition.
Remove decay heat.
Provide fault isolation where motion or a change in properties or configuration is involved. (e.g., circuit breakers, fuses)
Active functions require no funher evaluation in the IPA process. Any components which contribute to active intended functions would not be included in the list of SCs subject to 42 Revision 0
ATTACIIMENT (1)
CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY AMR, unless warranted by their contribution to other intended functions which are passive.
5.2 Determine Whether Components Are Lone-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 qualified life, specified time period or a properly justified condition-based replacement program. SCs which are not subject to such replacement are classified as long-lived.
The case of replacement based on a specified time period is straightforward. Such replacement programs may be based on vendor recommendations, plant experience, or any means which establish a specific replacement frequency. Often, replacement based on qualified life will also be replacement at a specific time period (i.e., the time period dictated by the qualified life). However, in some instances the qualified life of an SC may be based on variables other than calendar time.
For example, run time rather than actual calendar time may dictate replacement for some components. 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 program is one where SCs are replaced based on performance or condition.
'Ihe SOCs accompanying the LR Rule state that -
. the Commission has decided not to generically exclude components that are replaced based on performance or conditionfrom an aging management review.
The Commission does not intend to preclude a license renewal applicantfrom providing site specipc justification in a license renewal application that a replacement program based on performance or condition for a passive component provides reasonable assurance thatfunctionality will be maintained in the period ofextended operations. (60 FR 22478)
There are instances where an indication of SC condition can be used as the basis for replacement of a passive SC and that such replacement would preclude the need for an AMR. For example, the copper-nickel tubes of a heat exchanger may have an intended pressure-retaining function. This function is passive since there are no moving parts or changes in configuration or properties involved in performing the function. Normally such tubes are not replaced based on a specific time period or qualified life. Instead, they are subject to eddy current testing which dictates when tubes must be plugged and a tube plugging limit which dictates when the tube bundle must be replaced.
Plant experience shows that these heat exchangers are retubed every 10 to 15 years. In cases such as this one, where a plant parameter for a passive SC can be clearly linked to the ability of the SC to perform its intended function, and where plant operating experience has shown that the component is replaced frequently, the SC need not be included on the list of SCs subject to AMR.
Other components subject to condition monitoring include rubber / synthetic parts and parts specifically designed and maintained for wear. Such parts are periodically monitored and are normally replaced several times over the normal life of the plant when wear or other degradation is observed. Such SCs need not be included on the list of SCs subject to AMR.
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In these cases, justification will be provided in the LRA to A maa=* rate that such SCs are replaced j
frequently, and therefore require no specific AMR. Table 5-1 shows the criteria which are covered 1
in each justification. For these cases, controlled plant programs dictate the conditions which govern the rep %=~* of the SC. However, these programs are not described in the LRA or summarized in the FSAR Supplement, as would be required for programs which manage the effects of aging for SCs subject to AMR. Instead, the LRA justification' would contain a demonstration that the criteria of Table 5-1 have been satisfied for the program. ~ 'Ihe level of -
control which exists for such replacement programs and activities under the CLB will continue into i
the period of extw peration and is sufficient to ensure continued replacement of the SC.
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l 5.3 Auk m.: of System Cor=:s to Commodity Evaluations 1
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 l
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 c+-- -iity'-
evaluation process for these components replaces all IPA steps, and this pre-evaluation discussion would not apply to such components. For the EP and IL commodities, some or all of the components are assigned equipment identifiers in the CCNPP equipment database. For these l
components, the preevaluation 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 j
the eu....odity evaluation of the appropriate commodity group. These two cases are discussed below.
L_
r 5.3.1 Eh j
Electrical panels are assigned to a number of systems in the CCNPP equipment database i
because they are functionally related to the system components. In all cases, the passive i
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.
L At this point in the pre-evaluation process, such panels are. excluded from the AMR of I
their parent system and are instead administratively included with the EPs cc....odity cvaluation. As will be described in Section 7 of this methodology, the commodity
[
t 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 JLs and Tubinn Many fluid systems contain a number of small ILs which are part of the systems' pressure-
[
retaining boundary. Such small branch lines contribute to the passive intended function of 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 l
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-I cnvironments, are made of common materials and perform the same passive intended j
function reganiless 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 j
AMR of the parent system. 'Ihe w....adity_ evaluadon of ILs includes: 1) pressure- -
j retaining portions ofinstruments, such as pressure transmitters, pressure indications, level O
transmitters, etc.; 2) small bore piping, tubing and fittings' from'the first isolation valve -
l cWM to the system piping; and 3) hand valves which are part of the small branch lines l
(such as equalization and vent valves for pressure differential transmitters).
1 5.4 How the Pre-Evaluation Process Anolies to Structures
.l l
For plant structures, a modified process is used to determine which SCs are subject to AMR.~
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l 5.4.1 Passive Versus Active ~
Section 4 of the IPA Methodology describes the seven intended structural functions which 1
j may cause a structure to be included within the scope of LR per 654.4 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 Lone-Lived Plant structural w..yor.cr.ts are not nonnally 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.
1 5.4.3 Structures Which are Also Deci==*M as Systems In two instances, plant structures are also characterized as systems in the CCNPP site docuraentation system and system-type components are associated with these " systems."
For example, the primary containment structure is also designated as the contamment system. All penetration seals, as well as several position switches and access doors, are -
listed as individual components of the containment system with unique equipment identifiers.
As discussed in Section 4 of the IPA Methodology, the techniques for scoping of a structure as well u those for scoping a system are applied to such a structure. Two distinct sets of scoping results are produced-one for the system w5---- ts 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
.i structural scoping results, pre-evaluation steps would not be performed for the reasons j
described in Sections 5.4.1 and 5.4.2.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 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.
t 5.5.1 Pre-Evaluation Results Section 5 identifies the SCs which are subject to AMR. 'Ihis 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 Docurpentation The Pre-Evaluation task produces a list of the SCs which are subject to AMR for inclusion t
in the LRA. For system components excluded from the AMR because of a replacement program based on condition, the LRA will includejustification that the program has led to i
frequent replacement of the component.
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TABLE 5-1 y
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 intended system function (s). For example -
Is the discovery activity frequency interval less than the shortest time between failures ofintended l
system function (s)?
Based on the condition or performance trait monitored by this program, is the component replaced at intervals that are short relative to the life of the plant and is the component replaced before its i
contribution to intended system functions is prevented?
Historically, have all maintenance preventable functional failures ofintended system functions been detected by the activity?
j Criterion 2-Replacement programs based on condition or performance must contain appropriate acceptance criteria which ensure timely replacement of the SCs.
Does the activity have an action or alert value or condition parameter to determine the need for h
replacement of the SC7 Does the action value or condition provide an appropriate means of assuring replacement of the component before the effects of aging would prevent any intended system functions?
i Criterion 3 - Replacement programs based on condition or performance must be implemented by the facility operating procedures.
Is the activity controlled by a site review process which includes controls over subsequent -
revisions?
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' 6.0 AMR l
Dis Section of the IPA Methodology describes how the' components which were determined m l
Secten 5 to be subject to AMR are evaluated for the effects of age-related degradation. It also -
I
- describes the' approach used to identify and evaluate aging management alternatives to determme j
which adequately manage the effects of aging. Figure 6-1 is a flow chart which represents the A M R process.
De AMR task fulfills the requirements of 10 CFR 54.21(a)(3)of the LR Rule:
For each structure and component idents)ed in paragraph (a)(1) of this section.-
demonstrate that the efects of aging will be adequately managed so that the intendedfunction(s) will be maintained consistent with the CLBfor the period of extended operation.
l De input to the AMR task is the list of SCs subject to AMR along with_ the intended, passive 1
functions for those SCs. The results of this task demonstrate the following for each input SC or -
group of SCs:
1 Management of the effects of aging is not required because these effects are not detrimental to the ability of the SC to perfonn its intended function consistent with the' CLB; Existing programs or activities will adequately manage the effects of aging; or New programs or activities or the modifications to existing programs or actisities 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. De process described in this Section applies to SCs of both systems and structures with very few exceptions. nese exceptions are described in the steps where they.
occur.
De AMR can be performed in one of two general ways. In some circumstances, it is possible to demonstrate that existing plant programs adequately manage the effects of aging without an explicit evaluation of the aging mechanisms. His 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 attematives.,in addition, technological developments may produce additional viable alternatives in the future for either approach. Section 6.3 describes the CCNPP approach for evaluating and selecting from these alternatives.
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ATTACHMENT (1) 1 t-CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY AMR Process usiofpass~e. w.
Eved SCs and their I
intended functions.
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6.1.1 & 6.1.2 Can it be j
is SC part hown that effects o of a complex Yes aging are being mng'd Yes-1 assembly?
w/o addressing i
ARDMs?
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Provide documentation No that effects of aging are LRA
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4 No D
being managed to l
assure intended 0*13 j
isSC l
hved EQ7 No Listof plausible d
ARDM/subcomponent
- *D'"** ""
6.2.1 6.2.2 u
o Create potential Organize SCs into ARDM 1.ist groups of
/ For eac plasible \\
subcomponents
+ ARDM/suhiw =nt l
combination i
input from Site Create ARMD matrix
=
Assess level of f
concem and severity of e
/ For each ARDM/ N aging effects i
shii,~.snt 6.3 combination 6.2.3 y
isARDM Add ARDM/
Determine the plausible based on subcomponent to list of appropriate type of y,,,
terlaf, environrnent plausible ones for the aging management LRA
& function?
system.
based on concems and No effects. Document reasons.
No
+
Document reasons j
"[,
~
All plausible ARDM/
ypg u
g, ana d?
Complete Alf ARDM/subcomponent
\\
/
combinations complete?
\\
/
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 6.1 Justification that Effects of Anine are Beine Manneed Without SDecifically Evaluatine ARDMs
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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 ofaging mechanisms are operative, including erosion.
corrosion, wear, thermal and radiation embrittlement, microbiologically induced aging efects, creep, shrinkage, andpossibly others yet to be identified orfully understood. However, the detrimental efects of aging mechanisms can be observed by detrimental changes in the performance characteristics or condition of systems, structures, and components if they are properly monitored.
l (60 FR 22474) v
'Ihree cases are described in this Section. For two of these cases, the AMR demonstrates that the effects of aging on the passive function would be reflected in a change in one or more monitored performance or condition characteristics of the SCs. Therefore, by adequately monitoring these performance or condition characteristics, the effects of aging on the passive intended function are i
also adequately managed. In the third case, described in Section 6.1.3, an existing CLB program is already managing the effects of aging for a defined time period.
6.1.1 Complex Assemblies Whose Only Passive Function is Closelv Linked to Active Performance For some complex assemblies of SCs, the principal intended function is an active function.
Some of their components are subject to AMR because the components contribute to a passive pressure-retaining function to support the active functions of the entire assembly.
j l
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 3
function. However, there would be a readily observable affect on the diesel generator performance if the pressure-retaining components deteriorated significantly. For example, significant cooling water or lube oil piping leakage would result in increased bearing temperatures, and significant starting air leakage would affect diesel start times.
Additionally, experience has shown that even minor leakage from any of these supporting subsystems is observed by operators conducting routine testing well before they result in j
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 restore the passive i
function from its degraded condition are required by the performance testing program and 1
by the normal site corrective action 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 is to subject the assembly of components to a rigorous performance and condition monitoring program.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY In the cited example, the diesel generator support systems are subject to sun'eillance requirements to demonstrate operability in accordance with the Technical Specifications and to a comprehensive reliability program required by other regulations. Continuing these types of performance and condition monitoring programs ensures that the intended functions of the assembly will be adequately managed.
'Ihis AMR technique is used only in the following circumstances:
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 The complex assembly is covered by the Maintenance Rule.
6.1.2 Comoonent Assemblies Subiect to Complete 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 refurbislunent 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 speciflad amount of time. Because the valve operator uses a combination of hydraulic fluid pressure and compressed nitrogen to operate the valve, several components of this operator 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 intenal.
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.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY This approach is restricted to refurbishment programs that meet the following criteria:
He refurbishment is conducted at regular intervals on a complex assembly of components where the pressure-retaining functioc only directly supports the active intended function of the assembly; He passive function 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 assembly components and subcomponents are inspected for signs of aging and other degraded conditions; The refurbishment directs replacement of components and subcomponents that are deteriorated excessively due to aging or other degradation; and The component assembly's intended functions are tested after the refurbishment.
6.L 3 Lonn-Lived EO Components Components subject to EQ are already adequately managed for the effects of aging. This program ensures that the effects of aging will not prevent the qualified component from performing its intended function, in accordance with the CLB, at any time during the qualified life of the component.
Prior to exceeding the qualified life of any component, the EQ program requires that the component be reanalyzed to extend the qualified life or that the component be replaced.
Therefore, the combination of the qualified life and the requirement to take appropriate action prior to exceeding this qualification will adequately manage the effects of aging on equipment covered by this program.
Any component in the scope of LR which has a qualified life ofless than 40 years would not be subject to AMR since this component is replaced based on its qualified life. For any cotaponent with a qualified life greater than 40 years, the EQ Pmgram is credited as the adequate aging management program for LR, with no specific evaluation of aging mechanisms.
6.2 Performine an AMR by Evaluatine Acine Mechanisms 7
in some circumstances, the most efficient manner to show that the effects of aging are being adequately managed is to evaluate the effects of specific aging mechanisms on the intended 7
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 52 Revision 0 l
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y d
functmas and to
=%.te that those effects are being managed This' Section describes this
)
method ofperformmg an AMR.
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6.2.1 Cr=*ia= a Pne=*i=1 ARDM List I
The first stcp of the specific evaluation of ARDMs is to determine which ARDMs must be '
l 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 m an agmg evaluation and the ET has not been evaluated as 1
part of a previous evaluation, a new Potential ARDM List is created Industry documents
~
are revmwed 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 l
is compiled. De 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. His Potential ARDM List is the list of ARDMs that will be i
considered for subsequent evaluations of SCs of this ET. De Potential ARDM List is updated as each SC of the same ET is evaluated.
{
He 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 j
imtial list for the ET for piping. However, when finalizing the Potential ARDM List for -
the Service Water System, this ARDM is climinated as not applicable because the temperatures throughout the Service Water System are too low to warrant consideration of
. this M"sm. %c basis for marking an ARDM as not potential is recorded on the
-l Potential ARDM List for the system.
i i
Structural components are not associated with a particular ET in the site equipment
'l database, and therefore a modification to this step is needed for structural components.
Instead of creating the Potential ARDM List for each ET, structural canaan =* types j
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.
6.2.2 SC Groupine If a system contains several SCs with similar characteristics, the evaluation process can be made more efficient by grouping these SCs together for a common evaluation.
[
All components of systems are classified in the site equipment database with a particular '
~
DT code. Examples of such DTs are hand valves, check valves, pressure transmitters and -
heat exchangers. The DT can be further divided to facilitate the evaluation process. For
. adequate aging management, BGE chose to use a more mechanistic approach due to other benefits derived from performing this approach.
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example,'if the check valves.of a particular system are made of two distinctly different matenals, 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 r/hich controls valves in the system would be in another.
'I i
External Enviiu. ist - All system underground piping could be included in one group,.
while the aboveground piping would be in another.
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 identifier.
Groups'may be further subdivided into the individual subcomponents which make up the components in the group if this facilitates the subsequent evaluation.
If certain-subcu yersts are not required for the SC to perform its intended, passive function, they J
'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 3
any intended function. Therefore, the discs, seats and air operators of the valves in this group would not be subject to AMR because they do not contribute to an intended passive !
L 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 anm-ws.
I Again, because of site documentation differences for structural components, the structural component type is used to establish the initial level of grouping in 1he same manner as DT -
is used for system components.
6.2.3 CIrate and Resolve the ARDM Matrix'.
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 all potential ARDMs along one' axis and all remaining subcomponents for a particular SC 5
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 subco.v.yorst.
Although material, environment and function are mentioned separately above, when evaluating ARDM plausibility, all of the factors are also considered together.
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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. The rationale for designating each ARDM is recorded in evaluation reports and maintained onsite.
6.3 Methods to Mannee the Effects of Arine
'Ihis 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.
6.3.1 Phases of a Maintenance Straterv An adequate maintenance strategy consists of four phases: Discovery, Assessment /
Analysis, Corrective Action, and Confirmation / Documentation (1)
Discovery - The first phase of a maintenance strategy is identification that detrimental effects of aging are or could be cecurring. As stated in the SOC for the LR Rule:
The Commission believes that, regardless of the specyic aging mechanisms, only age-related degradation that leads to degraded performance or condition (i.e. detrimental efects) during the period of extended operation is ofprincipal concernfor license renewal. Because the detrimental efects ofaging are manifested in degradedperformance or condition, an appropriate license renewal review would ensure that licensee programs adequately monitor performance or condition in a manner that allowsfor timely identification and correction ofdegraded i
conditions. (60 FR 22469)
Aging can be self revealing or identified through specific diagnostic techniques.
Examples of discovery methods include visual observation of external conditions, eddy current examination for flaws, and ultrasonie testing for detecting wall thinning. Some plant programs may use specific detection techniques to detect and monitos 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 managing aging since other plants may discover aging effects before CCNPP.
(2)
Assessment / Analysis - Once performance or condition degradation is discovered, its progress must be compared to criteria or other guidance to determine the degree
)
of the degradation and the need for specific and generic corrective and preventive action. These criteria and guidance will depend on the characteristics of the degradation and the effects on the intended function.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY ne aging management implementation engineer facilitates the panel meetings, prosides consistency between system and commodity technical discussions, ensures involvement of the appropriate plant personnel, and ensures closure of open items.
He panel as a team determines the appropriate methods to manage the effects of aging for the given system or commodity considering two main factors:
The likelihood the ARDM will occur for the specific application; and How the effects of the mechanism progress.
If the panel determines that the ARDM occurs and progresses relatively rapidly, then prescriptive plant programs or system modifications may be warranted.
One-time inspections and/or performance or condition monitoring may be warranted if:
The mechanism has not been seen yet in operating plants; Present knowledge indicates progression is gradual; and The known characteristics of the ARDM indicate a potentially severe impact on the system intended function.
Continuing to monitor and evaluate industry experience may be appropriate if:
There is little or no experience with a panicular mechanism occurring for the system environment; Current knowledge indicates the ARDM progresses relatively slowly; and The potential consequences to the system intended functian are not significant.
6.3.3 Selection of Acine Manacement Alternatives for Discovery Once degradation is discovered, the process described in Section 6.3.4 will ensure that the appropriate Assessment / Analysis, Corrective Action, and Confirmation / 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 Program _t Plant r r ums are often the most direct and systematic method of detecting and mitigating the efim vf aging.
Hey already exist to meet regulatory requirements or recomme.: aaons, warranty requirements, or to preserve economic investment based on 57 Revision 0
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l l(3 Corrective Action - With the degree of degradation known, specific corrective action can be taken to ensure that the equipment performance or condition is l
t restored and the intended function is maintained I
t (4)
Confirmation /Dec- '=* ion - After the corrective action is performed, post.
J maintenance verification or testing confirms that maintenance was performed p
correctly and the equipment is capable of performing its intended function.; The l
^
corrective action and testing are documented as part of plant records for future i
i reference.
l In combination, these four phases provide a complete maintenance strategy Sections 6.3.2 j
and 6.3.3 describe how discovery activities are identified and selected. Section 6.3.4_-
i describes how the latter 3 phases are implemented.
f 6.3.2 Site Exocrt PanelInput De selection of the appropriate method for detecting aging effects is performed through an expert panel review of each plausible ARDM/ subgroup combination. The review is conducted on a system or commodity basis and, typically, consists of following plant mpresentatives:
He system or commodity aging evaluation engineer;
%e cognizant system engineer; j
Appropriate plant program managers / technical area specialists; and 1
The aging management implementation engineer.
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Each member brings specific focus and talent to the expert panel.
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.I ne aging evaluation engineer presents-the results of the system aging' evaluations j
highlighting the intended functions of the systems, the components subject to AMR, and -
the plausible aging effects. He aging evaluation engineer also proposes the methods by.
which the effects of aging can be managed.
]
1 The system engineer brings his knowledge of the system and functional requirements, f
knowledge of the plant and industry experience with the system, and familiarity with j
system inspection, surveillance, testing and maintenan::e results. The system engineer also j
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 ceacurrence that their program methods will effectively detect and ~
c monitor the specifier' aging' effects and are presently the preferred methods.
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site experience. They are typically selected as the method of discovering aging when they exist and can discover the effects of the plausible mechanism.
s 1
The plant programs applicable to the system are identified and reviewed to determine if
-i 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.-
3 Existing plant programs can also be modified to ensure the discovery phase of the mamtenance strategy is adequate for the period of extended operation.' Examples of j
modifications to an existing program include, but are not limited to, the following:
i Adding components to inspection procedures for specific aging effects; Adding specific aging effects mitigation procedures; and Tailoring of record keeping and trending requirements.
i If no existing plant program can be adapted to address the aging effects for the given.
group of SCs, new programs may need to be implemented.-
Some modifications to existing programs and new programs may be implemented prior to '
l submittal or approval of the LRA. Alternately, the LRA may include a commitment to
.,l implement the program or modification at an appropriate future date before or during the.
period of atended 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 j
Maintenance Standards Program Performance Evaluation Program Check Valve Reliability Performance Evaluation Program (Operations)
]
Eddy Current Testing Plant Lay-up and Equipment Preservation i
Electronic Cable Degradation Post-Maintenance Testing Engineering Test Procedures PressureTest Procedures-Surveillance Test Procedures Plant Tours Fatigue Monitoring Protective Coating and Painting FunctionalTesting System Walkdowns 4
Environmental Qualification Thermography Inservice Inspection Vibration Monitoring loose Parts Monitoring Thermal Performance Monitoring Lube Oil Analysis Operator Rounds j
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6.3.3.2 Site Issue Reportina (IR) and Corrective Action Pronram 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 actisities are:
I Plant tours by supenisors and managers-Management and supenisoryjob observations; Maintenance planning walkdowns;
-l Walkdowns of planned and completed modifications; Fire watches; and; Personnel safety equipment inspections.
1 Any observed or suspected condition that requires significant corrective action, whether i
related to the purpose of a specific activity being performed or not, is documented via an 1
IR.
6.3.3.3 Plant Modifications i
Plant modifications may be appropriate where:
Plant programs cannot effectively discover the effects of aging;
- i 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.
Examples of modifications which might result from the aging evaluations include, but are not limited to, the following:
Relocation of equipment to a less aggressive emironment; Change of material to improve resistance to the aging mechanism; and Change in the equipment operation.
Modifications to plant equipment may be implemented prior to submittal of the LRA.
Alternately, the LRA~ may commit to implement a modification at an appropriate future -
date.
6.3.3.4 One-Time Insocctions 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 any chance of stress corrosion cracking.
Instrument taps made with Alloy 600 may have been installed to minimize corrosion.
In these cases, a one-time inspection could be conducted to conclude that significant degradation is not occurring or that the rate is sufficiently slow to preclude concern during the period of extended operation. Alternatively, the inspection might conclude that additional inspections are needed during the period of extended operation.
The scope of such one-time and additional inspections would typically be a statistically representative sample of the population. Where practicable and prudent, the sample will be biased to focus on bounding or leading components. For example:
The portion of a structure more likely to experience the ARDM; A statistically representative sample of the valves made of a particular material; or Several of the Alloy 600 components that are predicted to be more susceptible to Primary Water Stress Corrosion Cracking.
If the sample indicates little or no degradation, the aging mechanism would be adequately managed by the one-time inspection for the component group or structure. 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 cases where the sample demonstrates there is no significant degradation and no program is needed to manage the effects of aging, resolution of the aging mechanism would be documented by describing:
The one-time inspection process; and Why it is an adequate ari, mach to disposition the ARDM for the SC group.
A particular one-time inspection may be completed before submittal of the LRA. In other cases, the LRA may commit to conduct the one-time inspection. If industry experience resolves the aging issue in the interim, the inspection could be canceled.
6.3.3.5 Industry Operatinn Experience Monitoring plant and industry experience provides for discovery of unknown, theorized, and emerging aging mechanisms. The materials used at CCNPP are common to nuclear plants and to many non-nuclear, operating plants that have longer operating histories.
Monitoring plant and industry experience therefore provides reasonable assurance that these ARDMs will be discovered before they severely affect intended functions at CCNPP.
It also provides assurance that appropriate changes are made to existing programs.
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- Industry information is. distributed across the' nuclear industry via Institute of Nuclear Power Operation's Significant Event Evaluation Information Network program which is a 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.
Exainples ofinformation that the program captures are:
1 7
Part 21 Notices; L
NRC Bulletins; NRC Information Notices; NRC Generic Letters; VendorInformation Letters; Operating Experience Information; 5
Significant Event Reports; Operations and Maintenance Reminders; and Significant Operating Experience Repons.
In some cases, the aging evaluation may be based on emerging industry information from the nuclear power industry or other industries that indicates -axpa +~i deterioration may l
occur. Although the aging effects have not been detected yet at CCI!PP w most other plants with similar equipment, similarities in materials and environments make it possible for the aging effects to occur at Calvert Cliffs. In these cases, discovery has already occurred through notification from NRC, Nuclear Energy Institute, Institute of Nuclear '
Power Operations, Owners Groups, or vendors.
The IR and corrective action process requires review and evaluation of the industry experience, and comparison to conditions at CCNPP to determine if additional action is needed here. If resolution of the issue is in progress, it will not necessarily be completed prior to LRA submittal or approval. The site IR and corrective action process ensures that assessment / analysis occurs and appropriate action is taken.
For example, a current industry issue is Primary Water Stress Corrosion Cracking of Alloy 600. Based on current knowledge, BGE has determined from' material and environmental properties that Primary Water Stress Corrosion Cracking'for reactor. vessel.
head penetrations at CCNPP will initiate and propagate much slower than at many other plants. Inspection results from other plants suggest no immediate concem for CCNPP.
Additional plants are planning inspections. - At this time, BGE cannot conclude that.
inspections will be needed However, the processes are in place to ensure appropriate future decisions are made based on accumulated industry knowledge.
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CALVERT CLIFFS NUCLEAR POWER PLANT LNTEGRATED PLANT ASSESSMENT METHODOLOGY 6.3.4 Imolementina the Assessment. Corrective Action and Confirmation Phases of the Maintenance Stratenv De last three phases of the maintenance strategy are required by the CLB and are provided by the site IR and corrective action process. Any obsened or suspected condition that requires significant corrective action, whether related to the purpose of the specific activity being performed or not, is documented via an IR. Initiation of an IR causes the degraded condition or performance to be evaluated for immediate personnel or nuclear safety concerns, operability concerns, and reportability. The IR is screened and classified to ensure that timely corrective action is taken.
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 significant events and issues, an event im'estigation and root cause analysis is conducted to aid in preventing reoccurrence.
Therefore there is reasonable assurance that timely discovery of aging issues and effects will result in timely and appropriate action to evaluate, correct, document, and report them.
6.4 Plant Prorram Documentation Documentation in the LRA for this step consists of a description of the programs and activities which were identified during the AMR and are relied upon to manage the effects of aging.
Additionally, any program 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 description presided in the
-l LRA and the summary prosided in the FSAR Supplement.
i 6.5 IPA
SUMMARY
l The completion of the AMR task concludes the IPA required by the LR Rule. D as 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 arrive at equivalent results.
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' 7.0 COMMODITY APPROACHES TO AMR As discussed briefly in Section 1 and 4 of this methodology, the approach described in the first six
'l sections of the methodology was followed for all plant SSCs with only a few exceptions.' 'Ihese six" exceptions are described in this section.
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'Ihe intent of a commodity evaluation is identical to the normal IPA approach, i.e.; to ter.asi.w.te 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 equivalent, approach; In
-j addition to describing the steps of the alternate process, this section demonstrates that each of these i
processes are equivalent to the process described in the first six sections of the methodology, i
For the purposes of discussion, the six commodity evaluations are divided into'two groups: 1) those i
that are equivalent to and replace only the AMR step of the IPA (Section 7.1) and 2) those that are..
i equivalent to and replace the entire IPA process (Section 7.2). Table 7-1 shows the sixto iwsiity j
evaluations and which belong to each of the categories described above.
i TABLE 7-1 l
.d
, :x x Equivalent to Entire IPA ora l
qCa===~4ty Evaluation,
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lust AMR7J I
EPs AMR ILs AMR Cables IPA Cranes and Fuel Handling Equipment IPA Component Supports IPA FP Equipment IPA 7.1 Qgnodity Evaluations Eauivalent to the AMR Sten 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 c,nc 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 M-rsdity AMRs. The system AMR is conducted as described in Section 6.' The commodity '
AMRs are conducted as described below.
7.1.1 EP Commodity Evaluation For many fluid systems, the list of SCs subject to AMR includes many pressure-retaining
-l fluid system components and a relatively few EPs which provide structural support to active electrical equipment. All of these components could have been evaluated as part of the system AMR. However, the expertise of the evaluator and the type of reference materials and plant documentation needed to perform the AMR for these two types of equipment is substantially different. Furthermore, the AMR of the EPs requires a level of i
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY expertise, reference material and plant documentation similar to that needed for other SCs in electrical distribution and instrumentation systems. Therefore, for efficiency reasons, f.I the EPs are removed from the scope of each system AMR and all EPs (electrical distribution, instrumentation and panels supporting mechanical system operation) are grouped into a common commodity evaluation.
The first step of the EP commodity evaluation is to review the scope of all of the pre-.
evaluation results and to include all EPs subject to AMR in a commodity evaluation, regardless of the system the panel is assigned to in the site equipment technical database.'
Performmg this step maintains the link between the scoping and pre-evaluation results, which are done system-by-system, and the scope of the commodity evaluation. For some,
systems, the only components in the system which were subject to AMR were those included in the scope of the EP commodity 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.
For the following reasons, the EP commodity evaluation process is equivalent to the standard IPA process: 1) The scoping and pre-evaluation are donc per the standard j
process; and 2) The AMR is conducted per one of the methods described in the standard 1
process. The only difference is that this process is applied to equipment which is not I
designated as a system in the site technical database. This difference is accounted for by two factors -
An extra step in the commodity evaluation which specifies the scope of the commodity evaluation; and A step in the pre-evaluation which ensures that eveiy SC subject to AMR is targeted for either the system AMR or a commodity evaluation.-
Therefore, the EP commodity evaluation produces a result that is equivalent to the standard IPA process described in Section 4 - 6.
7.1.2 IL Commodi.trEvaluation For many fluid systems, the list of SCs subject to AMR includes many pressure-retaining fluid components which are past of small branch ILs. Regardless of which system these Its are assigned to, they share certain common characteristics with respect to aging management.
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1 CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY All consist of piping and tubing which contribute to only one intaahd passive 3
function, i.e., the pressure-retaining boundary of the system;
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All contain instrumentation which would be affected to some extent by significant PB leakage; b
All are designed in accordance with standard practices outlined in a specification
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forILs at CCNPP; and
.i All system piping attached to these Its is also subject to AMIL l
- I Because of these common characteristics, the BGE IPA process includes an IL cc.i i.cdity.
Again, the scoping and pre-evaluation steps of the IPA are performed using the IPA approach described in Sections 3 - 5. During the Pm evaluation task, the IL components are separated from the remainder of the system pressure-retaining boundary and are-targeted for a m --:-iity evaluation. Similar to the EP commodity evaluation, the first -
step of the IL commodity evaluation specifies the ' scope of the evaluation._ For every fluid 2 system subject to AMR, pre-evaluation results 'are reviewed and the system pressum-retaining instmmentation (including associated valves) is included in the scope of this :
i
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evaluation. The list of these components, plus the associated tubing and fittings (which do not have unique identifiers in the site equipment database), form the scope of this--
commodity evaluation.
The next step of the evaluation establishes the combinations of materials and environments that exist in the population ofinstruments, valves, tubing and fittings that are in the scope of this evaluation. The range of materials and environments is determined from a resiew of plant design basis infonnation such as the instrumentation specification. Table 7-2e shows the combinations and materials and fluid environments identified for Its at CCNPP.'
At this point, a generic evaluation of materials and environments is performed to determine which combinations within the population are subject to plausible age-related degradation using the same criteria described in Section 6.2.L If plausible ARDMs are discovered for a j
generic combination of materials and environments, the equipment within the scope of this j
evaluation are reviewed to determine which'ILs actually contain these combinations.
.j Appropriate aging management alternatives are then selected for these ARDMs using the techniques described in Section 6.3.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY TABLE 7-2 MATERIALS AND FLUID ENVIRONMENTS FOR INSTRUMENT LINES.
Macednis Steel Mesul and 8
Be'7 -
Bruns Bronze /
Bans-N B25 M
Memel Tenen Peevelmha Steel stainless sestdens 304 SS 316 SS 17-7 PH SS Held W
Brass SW Sorel Steet Environnwnes Acid w Caustic X
Air X~
X X
X X
X Ammonia X
Borated Water X
X X
X X
X X
Carbon Dioxide X
FuelOil X
X X
Fyrequel 220 X
Ilydrazine X
Ilydrogen X
X Gas X
X Lube Oil X
X X
Nitrogen X
X X
Oil X
X
'X i
Sahwaser X
X X
X X
Steam X
X X
X Water X
X X
X X.
X X
X X
X X
X 66 Revisiori 0 m
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.m -.
m m...m
- -.. a--
m 2
mm 1
e m
um..
=-er--en 2.
. ev a
11--
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY Again, this commodity approach produces results which are equivalent to results produced by the standard IPA process described in Sections 4 - 6. The scoping and pre-evaluation steps are performed using the standard IPA process. The AMR step is slightly different in that the evaluation of the effects of aging is done for generic combinations of materials and environments rather than actual specified groups of components.
For material /
environment combinations which are subject to plausible ARDMs, appropriate aging management attematives are determined and the SCs to which these methods need to be applied are identified. The results are the justification that the effects of aging will be adequately managed. This result is precisely the same as that produced by the standard IPA process. Herefore, this IL commodity evaluation process is equivalent to the standard IPA process described in Sections 4 - 6.
7.2 Commodity Evaluations Eauivalent to the Entire IPA For the cables, stmetural supports, FP equipment and cranes / fuel handling commodity evaluations, the process described in this section is equivalent to the component level scoping, the pre-evaluation and the AMR steps. The following discussion will provide the justification that the process descriled is equivalent to the standard IPA process described in Sections 4 - 6.
7.2.1 Cables Commodity Evaluation The CCNPP equipment database does not contain specific equipment connectivity for individual cables. Instead, a separate Circuit and Raceway database contains information on cables, their service function (power, control or instrumentation), their materials and their from and to locations. Correlation of cable schemes to individual raceways, equipment and rooms is then possible using the information in this Circuit and Raceway database and design drawings. Because of these differences in site documentation techniques, the BGE IPA process does not include cables within any of the system AMRs, but instead evaluates cables as a separate commodity.
7.2.1.1 Elimination of Cables Alreadv Adeauatelv Managed by the EO Prozranj 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 this process is to climinate all cables covered by the EQ Program. Discussion in Section 6.1.4 justifies that the EQ program is an adequate program for managing the effects of aging for all SCs within the scope of this program. Herefore, no further review of EQ cables is performed during the cables commodity enluation.
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- For all non EQ cables, the potential ARDMs which could affect CCNPP cables are considered. During the development of the commodity evaluation process, all of these mechanisms except thermal aging were determined to be ' hot potential"for the reasons i
specified in Table 7-3.
Therefore, the remainder of the cables commodity evaluation focuses on the queston, "Are any of the cables which are in-scope for LR' subject to i
dielectric failure due to thermal aging at normal senice temperatures in less than 60 years?"
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CALVERT CLIFFS NUCLEAR POWER PLANT LNTEGRATED PLANT ASSESSMENT METHODOLOGY Table 7-3 AGING MECIIANISM JUSTIFICATION WHY NOT POTENTIAL -
Radiation Stress For polymers ased in fabricating cables installed at CCNPP, the radiation threshold above which changes in 8
mechanical ani electrical properties becomes noticeable is 1E06 Rads. His would require that a cable be exposed to an average 2 R/hr dose for 60 years. BGE EQ Cable Reports record the maximum non-accident plant dose to be 0.35E6 Rads over 40 years. Extrapolating to 60 years results in a dose of 0.525E6 Rads. Since this value is less than the stated threshold the effects of radiation on non-EQ cables at CCNPP owr 60 years is considered insignificant.
Mechanical Stress and Installation BGE cable installation damage is precluded by installation standard practices which include limitations on cable Damage pulling tension and bend radius. It is not BGE's practice to oversize conduits for pulling of additional cables through occupied conduits. Cables are tested after installation and before operation. Installation damage-induced failures generally occur within a short time after the damaged cable is energized. Therefore, aging penalties to account for potential installation damage are considered unnecessary for cables installed at CCNPP. Mechanical stress due to forces associated with electrical faults are not considered to be a concern at Cahert Cliffs. His is due to the fast action of circuit protective devices at high currents (which cause large magnetic forces) and the fact that all cables are fully supported.
Electrical Stress Normal electrical (voltage) stress is not significant due to the overrating of cable insulation. For example, 600V and 1000V cabL is operated on a 480V system. Similarly, SkV cable and 15kV rated cables are operated on 4160V and 13.8kV systems, respectively. Overvoltages are limited in duration and amplitude due to the fact that the only portions of the plant distribution system exposed to lightning are the primaries of the 500-13kV service transformers, and these are protected by arrestors. All other portions are metal enclosed, underground or indoors.
Water Trecing His phenomenon is limited to High and Medium voltage Cross-Linked Polyethylene insulated cable in a wet environment. Electrical stress is not sufficient to create water trees in cables operated below 4 kV. All cables used in 4 kV and 13 kV service at CCNPP are Ethylene Propylene Rubber. Therefore, water treeing is not an aging concern at CCNPP.
8 Aging Management Guideline for Electrical Cable and Terminations prepared by Ogden for DOE, Section 4.1.4, p.4-19.
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7.2.1.3 Grounina of Cables and Calculation of Service Limitina Tm.iewMures j
i normal aging is a function ofinsulation material and service temperaturef The cables are -
l designated with a cable code which depends upon cable characteristics, such as insulation i'
and jacket material, number ~ and size of conductors,' application (power, ' control, instrumentation), etc. De next step in the cable evaluation process is to group the ' cable i
codes acw.t.;-g to insulation material. (Note that no credit is taken for cable jackets in the -
1 cable evaluaten to add additional conservatism to the approach.) A 60-year service limiting ts..r.&ure is then determined for each group based on infonnation contained in -
.l 8
the System 1000 industry material database. This. step includes all scheduled cables whether they are "in-scope for LR," or not.
l 7.2.1.4 Comoarison of Service Limitinn TcmimrMures to Actual Service Temocratures The next step in the commodity evaluation process is to determine service temperatures for
-l the cables by considering the ambient temperature of the spaces containing the cables and l
any ohmic heatmg effects.
The cable installation standard requires' r, epi &ing q
instrumentation cable from power and control cables. Instrumentation cables ' arc not subject to any significant ohmic heating since they are operated well below their ampacity limits and are separated from other cabling which may serve as heat sources. The highest (non-accident) annual temperature of any area that contains cabling is the Main Steam :
j; Penetration Room Its levelized annual temperature is 160*F.
Therefore, if.
j instrumentation cable has a service limiting temperature of 160*F or higher, then the cable R
is not expected to have dielectric failure during 60 years of service at CCNPP.
H Power and control cabling has an insulation temperature rating of 194*F (90 C), or higher.
He combination of ambient space temperature plus ohmic heating effects are not expected to exceed 194*F at any time during normal operations 10. Derefore, if the 60-year service :
1 limiting temperature exceeds 194*F (90*C), then no further evaluation for LR is required since such cable is not expected to experience dielectric failure during 60 years of service.
7.2.1.5 Determination of Which of the Limitina Cables are Within the Scope of LR.
Those instrumentation cables with a 60-year service limiting temp:rature less than 160 F, and those power and control cables with a 60-year senice limiting temperature less than 194*F (90 C), could potentially be subject to excessive dielectric degradation during the period of extended operations. The next step of the commodity evaluation is to determme w
9 System 1000 is a database managed by United Energy Services Corporation under a 10 CFR Part 50, Appendix B program. For mineral insulated 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 snelysis is used, based on this data, to determine the temperature which results in a time to failure of 60 years.
10 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.
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I CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY which of the limiting cables are within the scope of LR to determine which cables require L
. aging management action. This final screen is performed by applying the following rules:
Any cable associated with a SR load or a load whose failure could prevent n'
operation of a SR function is within the scope of LR.
I Any cable associated with equipment relied upon for response to the regulated l events listed in (54.4(a)(3) is within the scope of. LR if the plant-specific evaluaten for these events requires such cables to supply power to the load as part of the event response For example, cables supplying power to a load which is.
turned off during the response to an SBO would not be included within the scope L.
of LR. Cables providing diverse ' scram or diverse turbine trip signals in.
l accordance with the ATWS Rule would be within the scope of LR.
7.2.1.6 Deternunation of Ada==e Aninn M=aa-ament Practices.
l Dis final step of the cables commodity evaluation uses the techniques discussed in Section 6.3. Aging management alternatives are identified for the limiting cable within the scope
.)
7 of LR. Input from the site expert panel is obtamed, and methods which will adequately' manage the effects of thermal aging are developed and documented. Aging management practices being considered include temperature monitoring of the hottest cables using the.
60-year service limiting temperature or the insulation rated temperature as critical alent values.
q 7.2.1.7 Justification that the Cables Commodity Evaluation Process is Eauivalent to the Senad d 1 9
IPA Process He process described above for evaluating cables performs all of the steps included in the standard IPA process However, these steps are performed in a different order.. The scoping step is deferred until after plausible aging mechanisms are identified for specific -
groups of cables. He pre evaluation step is trivial since all non-EQ cables are passive and long-lived and, therefore, all are determined to be subject to AMR.' Therefore, the process begins with the method of performing an AMR discussed in Section 6.2. ' He AMR demanMrates that the effects of aging would not prevent the intended function for a :
number of cable groups. For the remaining groups, aging management alternatives are i
identified. He scope of the required aging management program is then detemuned by -
performing the. component level scoping step on only those cables which need to be managed for the effects of aging.
Herefore, 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. His result g.
is identical to the result produced by the standard IPA process described in Sections 4 - 6.
Herefore, these processes are equivalent.
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7.2.2 Cr==am/ Fuel P=~ma-Eauinment Cum. - e Ev=L 'ian Jl v
De 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 1
instead of being addressed individually in the standard IPA process. He five systems are listed below-Spent Fuel Storage j
Refueling Pool' New Fuel Storage and Elevator FuelHandling j
Cranes 1
De first step of this commodity evaluation is to determine which components in these -
1 systems contribute to the intended functions. The UFSAR and Q-List documentation is
-a 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 j
therefore within the scope of LR.
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Once the u-------ts within the scope of LR are defmed, the next step is to determme
-i which of these components have already been addressed for their intended, structural type -
function as part of the building which houses the component 11. Any such components are l
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 l
4 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. Dese components -
are therefore eliminated from the scope of the crane and fuel handling. commodity j
' evaluation.
After eliminating the intended functions and ew-- =ts already addressed by the AMR of the enclosing strubture (building), only.the seismic II/112 intended function remains as l
being not completely addressed by the enclosing structure's 'AMR.-
For most plant j
_ equipment, this function is completely addressed by the comb' ation of the AMR of the m
enclosing structure and the commodity evaluation of component supports (Section 7.2.3).-
However, for cranes and fuel handling equipment, portions'of the components other than i
the component supports and stmetural foundations could contribute to the II/I function if such equipment is used to liA and carry heavy loads over'SR equipment. Therefore, further review of this function is conducted in this w.v r.adity evaluation.
11 Secause the scoping process for structures addresses all structural support functions for equipment housed by the structure, it is expected that the majority of these components would have already been addressed;-
however, this step of the commodity evaluation is intended to confirm the process.
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Provide structural and/or functional support to NSR equipment whose failure could directly prevent satisfactory
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12 accomplishment of any SR functions (referred to a seismic ll over I or M).
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' Only designated load handling equipment at CCNPP is allowed to lift heavy loads over SR '
3 equipment. His designated load handling equipment is controlled in accordance with -
NUREG-0612." Control of Heavy Ioads," which was issued to hmmt the results of Unresolved Safety' Issue A-36. ; his NUREG provides a set of guidelines intended to 1
minimize the passibility ofload drops on safe shutdown equipment or decay heat removal equipment.
He control of heavy loads is addressed in the CCNPP UFSAR in Section 5.7 which lists the crancs subject b the guidelines of NUREG-0612. ~ nis section of the UFSAR concludes that the remaining overhead handling systems are excluded from NUREG-0612 -
controls because (1) lift points and safe shutdown equipment are sufficiently separated; or1 (2) the largest load lifted is not a heavy load. Wrefore, only the designated heavy load handling equipment requires AMR to address the II/I function (beyond the AMR of the enclosing structure and the commodity evaluation of the structural supports).
-l h next step of the commodity evaluation is to determine which portions of the heavy load handling equipment listed above are subject to AMR. This is accomplished by reviewmg.
the heavy load handling equipment and determining those components and submig.a.ts which contribute to the intended II/I function through motion or a change in configuration or properties. These components and subm.g. cats are active and, therefore,'are 13 eliminated from the AMR,
he remaining passive components and subcomponents are evaluated for the effects of '
aging using the techmques 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. & periodic inspections and testing programs for designated heavy load handling equipment, as well as other plant programs and activities, are reviewed to determme whether they adequately manage the effects of the plausible ARDMs. The process described in Section 6.3 is used to determine the appropriate aging management alternatives and these decisions are documented.
T5e crane and fuel handling commodity evaluation is equivalent to the standard IPA
- socess described in Sections 4 - 6. The component level scoping task is performed by.
d reviewing the same documents that would be reviewed during the component level scoping ;
task described in Section 4. Eliminating components already covered by other~ structural evaluations avoids duplication of effort. This step has no impact on the completeness of the IPA results. Eliminating components and subcomponents which contribute to intended j
functions through motion or a change in configuration or properties is consistent with the' i
process described in Section 5, Pre-evaluation. Finally, the remaining heavy load handling equipment is evaluated using the same process as that described in Section 6.2, and ag^mg j
management alternatives are evaluated using the Section 6.3 process. Herefore, based on '
t 13 it is conservatively assumed that no components or subcomponents are replaced based on time or qualified life.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY the above discussion, the crane and fuel handling commodity evaluation is equivalent to the standard IPA process described in Sections 4 - 6.
j 7.2.3. Component Sunoorts 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.
1
. For this reason, it was determined that a commodity evaluation of component supports'.
l would be more efficient to address these supports than evaluating them as past of the.
t system AMR.
q Two plant programs govem inspoetion of component supports and form the foundation for any needed aging management program. 'Ihe elemerits of these programs are described in the following Sections 7.2.3.1 Seismic Verification Proicct (SVP)
The SVP is implementing the requirements of. Unresolved Safety Issue 'A-46 to verify the f
seismic adequacy of mechanical and clectrical equipment, including equipment supports and anchorage. To meet the requirements of Unreviewd Safety Issue A-46, the scope of equipment covered to date by the SVP is limited to equipment required for safe shutdown 14 following a seismic event and to electrical raceway supports The seismic adequacy _
criteria were dermed by the Scismic Qualification Utility Group-(SQUG) and are
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documented in the NRC-approved Generic Implementation Procedure (GIP). The criteria are based on inspections of equipment structural and functional condition'following i
19 strong-motion carthquakes at over 80 industrial facilities. At the time of the post-earthquake inspections, the average age of these facilities was 22 years, including at least '
11 facilities (or units within a facility) over 40 years 15. 'Ihe SVP equipment walkdowns have been conducted by SQUG-trained Seismic Capability Engineers, who evaluated and i
documented the condition of each equipment item in accordance with the GIP'w31kdown checklists. 'Ihese walkdown checklists require evaluations of equipment anchorage and support load path, including assessments of concrete condition and any other factors that -
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might lessen the seismic adequacy of the equipment's support. The w31kdown checklists also require a vicinity check to ensure that NSR equipment installed near.the equipment '
j being evaluated is adequately secured and does not pose a credible threat to the equipment.
being evaluated. "Ihe component support commodity evaluation relics on these walkdowns i
as a key element for managing the aging of structural supports for all equipment in the j
scope of the SVP.
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' 14 ' The CCNPP Individual Plant Examination for Extemal Events is essentially " extending" the scope of the original GlP regulroments by conducting walkdowns on other equipment to support the seismic aspect of the i
probabilistic risk assessment. These walkdowns use criteria similar but not identical to the GIP checklists.
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15L EPRI Report NP-7149-D," Summary of Seismic Adequacy of 20 Classes of Equipment Required for the Safe
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Shutdown of Nuclear Plants" l
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CALVERT CLIFFS NUCLEAR POWER PLANT ~
j INTEGRATED PLANT. ASSESSMENT METHODOLOGY Since the SQUG walkdowns under the SVP are a one time' occurrence, additional discovery and assessment activities are required in order to rely on this program as an on-gomg activity to manage the_' effects of aging for the LR term. ' Calvert Cliffs intends to l
commit to the GIP as an altemate method to verify the seismic =A~==ey of new and L
rep!=eanent equipment. -When the GIP procedure is'adaptat the associated process requires maintaining the availability of SQUG expertise after the A-46 walkdowns are l
complete. With SQUG expertise available onsite, any cumycr.c;;t support types' that are i
subject to ARDMs could be re-evaluated as appropriate. Comnutment to the GIP ensures -
i adequate assessment of any degraded conditions that are discovered during the period of.
j extended operations.
System engineers frequently conduct general walkdowns'en their systems in accorder.cc '
l with site guidelines. The site IR and corrective action process requires all plant peisc..r.cl -
l (including system engineers) to formally document any discrepancy they observe in the l
plant, including any potential structural support deficiency. Because of these required walkdowns, operator rounds, and other walkdowns performed by personnel familiar with -
l the plant, significant deterioration of the material condition of the structural supports would be discovered and reported for evaluation by qualified individuals.
7.2.3.2 Comoonent Supports in Senice Insocction f
i ne Component Support Insenice Inspection (ISI). Program' requires inspection of
.l equipment supports for components such as piping that are subject to ASME Code-requirements. - The Code requirements are specified in ASME Section XI, Article IWF.-
The scope of the Component Support ISI Program includes all ASME Class I,2 and 3
.i piping supports except those excluded by Article IWF 1230. Supports are inspected at -
regular frequencies on a prescribed sampling of all piping supports as specified in l
Table IWF-2500-1. He Component Support ISI Program requires inspection of supports
.j for the effects of aging, contains acceptance criteria and requires specific actions when the -
i criteria are not met. Therefore, it would serve as the aging management program for all' l
supports in the ISI Program.
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7.2.3.3 Determinine the Scope Differences Between Component Support Programs and LR j
l Based on the above discussion, the first step of the commodity evaluation of component supports is designed to determine the scope differences between these two programs, and l
the component supports within the scope of LR.; his step reviews the requirements for j
including supports in these programs and compares them to the {54.4 requirements for-l inclusion within the scope of LR. Scoping requirements which are common to LR and to l
either one of these programs would narrow the evaluation for the applicable em-:=t
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supports to a deternunation of what, if any, modifications would be needed to the program
- l to credit the program as an' adequate' aging management program. Any LR scoping requirements which do not correspond to requirements for inclusion in ISI or SVP would define the scope of component supports which may have to be addressed by modifying the scope of one of the above programs or by a different program. Such program changes would only be necessary ifit is determined that the effects of aging would be detrimental to the intended function of the component supports.
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CALVERT CLIFFS NUCLEAR POWER PLANT
.O INTEGRATED PLANT ASSESSMENT METHODOLOGY-7.2.3.4 E-A
== What F=h====4 If Anv. are Na~L d to Justify that the SVP Pronram Adequately Manages the EKects of Agmg
'Ihe next step of this task is an evaluation of the effects of aging on the various matenals
'which are used for component supports and the various environments in which they ase -
found in the plant. The observed condition of the component supports during the SVP walkdowns is used as one of the inputs during this review of aging effects. If any.
oombination of matenals and environments is susceptible to significant aging effects then -
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further aging management is needed for LR. 'Ihe threshold for. significance in this -
evaluation takes into consideration the observed condition of the. equipment supports during the initial SVP walkdown as a key factor in determining whether a more focused ~
aging management program is needed. Component supports which were found to be in good condition after almost 20 years of plant operation are not expected to be subject to
'j any new aging during the LR period and, therefore, could be effectively managed by the routine walkdown and IR process. Component supports found to be degraded during the SVP walkdowns may, in addition to outlier resolution activities required under'the GIP, require additional aging management activities at some point in the future.-
7.2.3.5 Determinina Enhancements. If Anv. Needed to Justify that the ISI' Pronram Maan=*dy' Manares the Effects of Anina Similar to the step described in 7.2.3.4, the ISI program is reviewed to ensure that it adequately manages the effects of aging. Again this step includes a review of the aging effects of the vanous materials used for ow =r.t supports covered by this program, and
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the vanous environments where these supports are located Again, any material and environment combinaten which is particularly susceptible to significant aging effects would be singled out for potential enhancements to the ISI program -The threshold for "significant aging effects" would again consider the results of past inspections to deternune whether ISI has proven to be adequate in identifying and resolving aging issues for component supports in the past.
7.2.3.6 Determinina Anorooriate Anina Mananement for Component Sunoorts Not Covered by ISI or SVP If any component supports are determined not to be covered by ISI or the'SVP (including Individual Plant Examination for External Events walkdowns, see footnote 7), the-commodity evaluation will determine the most appropriate method to manage the effects of aging for these supports. Several possibilities are listed below and others could be developed during the performance of the evaluation:
Add the component supports to an augmented ISI program Add the component supports to the scope of equipment covered by the SVP.
Demonstrate that the aging effects on the component support are bounded by supports already included in the scope of one of the credited aging management programs.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METIIODOLOGY 7.2.3.7 Component Suonort Commodity Process Eauivalent to Standard IPA Procem The commodity evaluation of component supports produces the same results as the standard IPA process described in Section 4 - 6. The major difference in this commodity evaluation is the initial premise that the ultimate aging management program for these supports will be based on one of two existing programs that perform inspection of -
component supports. Components in the scope of LR but not covered by the SVP or ISI Program are identified. The ' identified" component supports may be added to either program. Optionally, the effect of aging on the supports will be assessed to identify plausible aging mechanisms and appropriate aging management alternatives.
The next two steps of the evaluation are focused on providing the justification that the two programs do adequately address the effects of aging and/or identifying any enhancements needed to the programs. He only difference between this step and Section 6.3 is that the review in the commodity evaluation only considers these two programs, where the 6.3 review could credit a broader range of site activities. His difference does not affect the technical results in that justification of program adequacy is still prosided.
Based on the above discussion, the conclusion can be reached that the justifications provided by this commodity evaluation are equivalent to those produced under the standard IPA process described in Sections 4 - 6.
7.2.4 FP Eauipment Commodity Evaluation Over half of the systems which are included in the scope of LR contribute to one or more FP functions. These functions include both fire suppression / detection functions and j
functions related to equipment used to demonstrate alternate safe shutdown paths in the event of a severe fire (Appendix R). For the vast majority of these systems, the normal component level scoping process described in Section 4 of this methodology is performed.
However, there are seven systems which are in scope for LR primarily because of FP functions 18. For these systems, the alternate scoping process described in Section 7.2.4.1 is used.
Some passive intended FP functions are performed by fluid systems which are not SR.
For the SCs which are subject to AMR only because of such passive intended functions, an altemate AMR technique is described in Section 7.2.4.2.
16 i 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.
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.2.4.1 Scopine of Systems with Primarilvi7 FP Intended Functions The seven systems, which are in scope for LR primarily because of FP functions, are listed below.
Well and Pre-treated water FP Plant Heating Condensate Plant Drains Liquid Waste Fire and Smoke Detection Due to similarity of function, and the fact that most of the FP intended functions are active, an alternate approach is used for conducting the component level scoping of these systems. For these seven systems, identification of detailed system functions is performed as described in Section 4.1.1 of this methodology. However, after performance of this step, the intended functions are reviewed 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, development of function catalogs and generation of scoping results table) are then conducted on only the passive intended functions of the system and the remainder of the pre-evaluation (short-lived versus long-lived) is completed on only these scoping results.
He avoided steps in this modified process are the creation and further consideration of function catalogs for the active functions. His process produces a list of SCs subject to AMR which is equivalent to the list which would be produced by the process described in Sections 4.1 and 5. Had the active function catalogs been created during the component level scoping process, the components in these function catalogs would have been excluded from the AMR in Section 5.1 because they contribute to only active functions.
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 Compon.tnu 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.
I ne SOC accompanying the LR mle justifies exclusion of SCs associated with active fire
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suppression / detection fimetions from the scope of AMR based on the plant's FP Program.
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17 See previous footnote.
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The FPP { Fire Protection' Pro 4 ram] is part of the 'CLB and: contains l
maintenance and testing criteria thatprovide reasonable assurance thatfre; protection systems, structures and components are capable ofperforming k
their intendedfunction.' The Commission concludes that it is appwpriate to
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allow license renewal applicants to take creditfor the FPP as an existing _
progmm that manages the detrimental efects of aging 3 The Commission i
concludes that installed pre protection components that perform activeJ
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functions can be generically excludedfrom an aging management review on the basis ofperformance or condition-monitoring programs aforded by the r
FPP that; are capable of detecting and subsequently ' mitigating the detrimental efects ofaging. (60 FR 22472)
[l Although the SOC specifically refers only to SCs which contribute to active functions, the.
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justification could apply equally to ' installed FP components that' perform passive -
Wm." *lherefere, for the fire suppression / detection systems, the AMR consists of i..si.hding that the CCNPP FP Program addresses the pressure-retaining portions of I
these fluid system so that the effects of aging are adequately managed For the pressure-retaining components in fluid systems credited as altemate safe shutdown equipment for Appendix R, the AMR is performed in accordance with Section 6 of this' methodology, except when the conditions described below apply.
In some cases, the alternate safe shutdown function required of the system is fully tested during normal plant operation because the alternate safe shutdown function is~ subsumed by its power production function.' Any degradation sufficient to prevent a system from performing its alternate safe shutdown function would be detected and corrected during normal plant operations.. The site IR and corrective action program can be relied upon to document and correct the degradation to the power production system before it affects the system's ability to perform its alternate safe shutdown furetion.
a An example is the condensate system's intended function of providing make-up water to'.
the service water and component cooling water head tanks during a fire scenario that.
removes the normal make-up source. The normal source of water to fill these head tanks is the demineralized water system. The pressure-retaining SCs of the domineralized water 1
system that contribute to its intended function are evaluated in accordance with the process described in Section 6. If the normal source is rendered inoperable by a severe fire, thei Appendix R evaluation credits the condensate system for providing this make-up supply of.
water.
The pressure-retaming SCs of a large portion of the condensate system contribute to this attemate safe shutdown function. However, it is not conceivable that the condensate '
system would ever be capable of performing its normal power production function, but i
incapable of supplying a small amount of water to the head _ tanks.
'Iherefore,-
documentation and correction of a degraded power production function under the site IR-and corrective action program would assure that the effects of aging would not impact ths intended alternate safe shutdown function.
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l CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 7.2.4.3 FP Commodity Evaluation is Eauivalent to Standard IPA
- %e commodity evaluation described above produces results which are equivalent to the results produced by the standard IPA process described in Sections 4 - 6. While an interim' i
list of SCs in the scope of LR is not produced, the modified scoping and pre-evaluation -
process produces the identical list of SCs subject to AMR as the standard process, i
The co....adity evaluation of FP SCs applies the same cces described in Section 6.1 j
where degrad=*=i of passive SCs may be readily managed by active performance and
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' condition monitonng The focus of this approach is the justification that the CCNPP FP Program ensures both the active and passive fire suppression / detection functions through H
' maintenance and system rr.onitoring. His justification will demonstrate that the effects of aging are adequately managed. This result is equivalent to that produced in Section 6 of j
this methodology.
l De AMR approach described for certain attemate safe shutdown SCs allows for the demaaetration that maintenance of the altemate safe shutdown pressure-retaining boundary is subsumed by maintenance of the plant power production function.
With' this demaaeration, the approach would then conclude that the effects of aging can~ be
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adequately managed by the normal site IR and corrective action program. Therefore, this method provides justifications and conclusions equivalent to the Section 6' AMR process.
l In all other cases, the standard AMR process is followed for the SCs associated with l
pressure-retaining FP functions. Therefore, following this commodity approach for FP' equipment will produce an equivalent level of documentation to justify that the ' effects of.
aging will be managed for the period of extended operations.
7.3 Commodity Evaluation Results And Documentation Integrated Plant Assessment documentation for commodity evaluations would consist of a description of the programs identified during the evaluation which are relied upon to manage the.
cffects of aging. Additionally, any program modifications or new programs which need to be-implemented in order to adequately manage the effects.of aging for the period. of avta=!~8 1
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 8.0 TLAA REVIEW This section of the IPA methodology describes the process for reviewing analyses which may 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.
In 10 CFR 54.3, TLAAs are defined as:
Time-limited aging analyses, for the purposes of this part, are those licensee calculations andanalyses that:
(1) Involve systems, structures, and components within the scope oflicense renewal, as delineatedin 554.4(a);
(2) Consider the effects ofaging; (3) Involve time-limited assumptions defined by the current operating term, for example, 40 years;
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(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 capability of the system, structure, and component to perform its intended Jimctions, as delineated in f54.4(b); and i
(6) Are 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 ')rovides 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 TLA:
(t) The analyses remain validfor the period ofextended operation; (ii) The analyses have been projected to the end of the period of extended operation; or (iii) The efects ofaging on the intendedfunction(s) will be adequately managed for the period ofextended operation.
The TLAA Review task produces the required list of the TLAAs which are subject to LR resiew, and demonstrates that these analyses will meet one of the three conditions listed above. Figure 8-1 is a flow diagram which shows the TLAA review process.
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TLAA Review Task Electronic Docket Exemptions Nomexempton '
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potential TLAAs TLAAs subject to LR review industry Codes and Standards is
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exemption based 7
3 on a potential For all TLAAs subject TLAA?
to LR review PotentialTLAAs Yes (including exemptions with potential TLAAs) u Are the LRA No effects of aging
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adequately Identify SSC which is managed?
subject of TLAA v
Describe TLAA Exemption not
& Programs listed in LRA 7
LR scope?
no AND Can
- Potential TLAA relevant TLAA be modi-to safety determination?
g, fied to be valid through Yes,
AND eriod of extended
- Potential TLAA considers the effects operations?
Describe TLAA of aging?
& modifications AND to TLAA
- Potential TLAA relates to SSC's abihty to perform intended function Provide otherjustifcation
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Potential that TLAA is valid for the -
l TLAAs not period of extended listed in LRA operations Yes Describe TLAA &
ls SSC justification covered by CLB
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program which Yes updates potential subject l
TLAA?
to LR review complete?
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CALVERT CLIFFS NUCLEAR POWER PLANT INTEGRATED PLANT ASSESSMENT METHODOLOGY 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 and in efect that are based on time-limited aging analyses as defined in 551.3. The applicant shallprovide an evaluation thatjustifies the continuation ofthese exemptionsfor the period ofextended operation.
The TLAA Review task also fulfills this requirement.
8.1 Identify Analyses to be included in the Review he first step in the TLAA Review task is a search of the CLB to identify potential TLAAs and exemptions. The CLB search is done by reviewing the CCNPP electronic docket and the UFSAR.
He 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"[ effective full power years),
and " licensed life." Exemptions are identified by using phrases such as "50.12," and " exemption."
Specific examples of potential TLAAs contained in regulatory literature such as SECY 94-140 are reviewed in advance of the electronic search to help focus the search for potential TLAAs.
The potential TLAAs identified above are supplemented by a further search of the electronic docket. Codes and standards which govern design of SSCs at nuclear power plants were reviewed as part of a joint industry effort to determine those that might contain some form of TLAA. An additional search of the CCNPP clectronic docket and UFSAR is perfonned 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 TLAAs.
Exemptions that are based on time limited aging analyses, the potential TLAAs identified through time related queries and the potential TLAAs identified through codes / standards queries comprise the complete set of po;ential TLAAs identified in this step.
8.2 Review of Potential TLA As The potential TLAAs are reviewed to determine if they affect an SSC in the IPA scope, to determine whether the analyses are relevant to a safety determination, to determine whether the analyses consider the effects of aging and to determine whether the analyses relate to the ability of the SSC to perform its intended function (*).
Potential TLAAs which meet these four 1
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l criteria 18 are then reviewed to determine whether the analysis is governed by a CLB program j
which will update the analysis. He EQ Program is such a program. The potential TLAAs which 2
meet the first four criteria, and which do not meet the last criterion, are the TLAAs subject to LR review; i.e., those which must be listed in the LRA.
8.3 Disposition of TLAAs Which are Subiect to LR Review i
This step in the TLAA Review task compiles the TLA-related information for the LRA. Because l
of the first check performed in Section 8.2 above, all TLAAs subject to LR review must necessarily affect SSCs which are in the scope of LR, per (54,4. Herefore, for long-lived components supporting passive functions, the IPA process required by 654.21(a) will have documented that the effects of aging on these SSCs will be adequately managed. Thus, the IPA results need only check to ensure thr.t they also address the effects of aging associated with the TLAAs As noted above, for SCs subject to AMR, the programs listed are those already identified in the
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IPA. For active or short-lived SCs not subject to AMR, there are three options:
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Management of the effects of aging relating to the TLAAs must be demonstrated;
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The TLAA must be modified to project its applicability to the end of the period of l
extended operation; or l
l Justification that the TLAA remains valid for the period of extended operation must be i
provided.
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18 The definition of a TLAA contains six criteria. The two criteria not addressed in this step were already I
addressed in the initial search technique. The fact that the electronic search was performed against the l
CCNPP electronic docket and FSAR implements the criterion that TLAAs be included in or incorporated by
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reference in the CLB. The time-related queries and the evaluations of codes and standards account for the criterion that TLAAs be related to assumptions regarding the period of the initial license, i.e.,40 years.
84 Resision 0
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q ATTACHMENT (1)
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CALVERT CLIFFS NUCLEAR POWER' PLANT
. INTEGRATED PLANT ASSESSMENT METHODOLOGY I
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Summarv Ihe results of the TLAA Review task are:
The list of TLAAs subject to LR review;
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l The exemptions in effect that are based on TLAAs; and l
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c The analyses which justify that the TLAA remains valid or could be modified to f
remain valid for the period of extended operation, or o
The demonstration that the effects of aging considered by the TLAAs are being managed l
This infonnation is included as a past of the LRA. Since the programs 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.
85 Revision 0
8 ATTACIIMENT (2) i 1
i 10 CFR.PART 54 SAMPLE. RESULTS' i
FOR
.CALVERT CLIFFS NUCLEAR POWER PLANT Baltimore Gas and Electric Company August 18,1995 Revision 0
1 1
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BALTIMORE CAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS 1
FOR 1
CALVERT CLIFFS NUCLEAR POWER PLANT l
i TABLE OF CONTENTS l
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G E N E RA L I N FO RM ATI O N....................................................................................................
2 TECI I N I CA L I N FO RM ATI O N..............................................................................................
3 U FS A R S U P P L E M E NT............................................................................................................
e 4
TECI I NI CA L S P E CI FI CATI O N S t.....
t 5
EN VI R O N M E N TA L I N FO RM ATI O N...................................................................................
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APPENDICES A
TECIINICAL INFORMATION 4
l B
UFSAR SUPPLEMENT i
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l C
TECIINICAL SPECIFICATIONS i
I D
ENVIRONMENTAL INFORMATION i
l 1
i 4
I
BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT 1.0 GENERAL INFORMATION Baltimore Gas and Electric Company (BGE) hereby makes application, pursuant to the prosisions of Title 10 of the Code of Federal Regulations, Part 54 (10 CFR Pan 54) for the renewal of the operating license for Calvert Cliffs Nuclear Power Plant (CCNPP) Unit 1 issued pursuant to Section 103 of the Atomic Energy Act of 1954, as amended, and Title 11 of the Energy Reorganization Act of 1974. This application for the renewal of the operating license for CCNPP Unit I contains information pursuant to the provisions of 10 CFR Part 54 and contains the following pans:
a.
the general information which is set out herein; b.
the technical information required by 10 CFR Part 54, which is described in the following sections of this application; c.
any Technical Specification changes or additions necessary to manage the effects of aging during the period of extended. operation and justification for those changes; and d.
the environmental linformation~ required'by(10 CFRLPArt 54' which3sncontained in an appendix to this application?
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Baltimore Gas and Electric Company 1.2 Address of Annlicant Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203-1475 1.3 Description of Business of Anoliennt Baltimore Gas and Electric Company is an investor-owned utility engaged primarily in the business of producing and selling electricity and purchasing and selling natural gas. The Company, which was the first gas utility and one of the first electric utilities in the United States, serves an area which includes Baltimore City and all or part of nine Central Maryland counties. The area served i
with electricity approximates 2,300 square miles with over one million customers, while the area served with gas includes 610 square miles with about 550,000 customers.
To senice this area, the Company operates 10 electric generating plants in Central Maryland, including CCNPP. The Company also maintains shared ownership of generating facilities in Pennsylvania (Keystone, Conemaugh and Safe Harbor). In addition, Company is also a member of 1
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1 BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT the Pennsylvania-New Jersey-Maryland Interconnection, a power pool of eight Mid-Atlantic companies that provides reliability and the opportunity for bulk power sales.
The Company obtains substantially all of the natural gas it sells through purchases from pipeline suppliers and natural gas producers. In addition, the Company acts as a broker for large commercial and industrial customers, which requires locating, buying and transmitting gas.
Constellation IIoldings, Inc., a wholly-owned subsidiary, directs the Company's diversification efforts. This corporation holds the stock of three other companies engaged in such diversified activities as real estate development, energy and environmental projects, and investments and financial senices.
Baltimore Gas and Electric Company Home Products and Services, Inc., a wholly-owneJ subsidiary, provides revenue growth opportunities in markets relating to core gas d cl%tric businesses. This corporation operates 11 retail stores throughout the service territory marketing home appliances, electronics, and design and installation senices.
1.4 Leral Status and Organi7ation :
.4 Baltimore Gas and Electric Company is a public~ utility incorporated under the laws. of the State of Maryland with its' principal o'ffice. located in'Baltiinore,. Maryland atithe"adnessistated above.
Baltimore Gas and Electric Company is not foreign owned, controlled or dominated. Baltimore Gas and Electric Company makes this application on its own behalf and is not acting as an agent or representative of any other person.
The names and business addresses of BGE's directors and principal officers, all of whom are citizens of the United States, are as follows:
Directors:
Christian 11. Poindexter Chairman & Chief Executive Officer Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 II. Furlong Baldwin Chairman of the Board & Chief Executive OfIlcer Mercantile Bankshares Corporation PO Box 1477 Baltimore, MD 21203 Beverly B. Byron 4000 Cathedral Avenue Washington, DC 20016 2
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CALTIMORE GAS & ELECTRIC COMPANY j
10 CFR PART 54 SAMPLE RESULTS i
-FOR-j CALVERT CLIFFS NUCLEAR POWER PLANT J. Owen Cole Chairman of the Executive Committee i
First Maryland Bancorp j
PO Box 1596 Baltimore,MD 21203 Dan A. Colussy Chainnan, President & Chief Executive Officer UNC Incorporated 175 Admiral Cochrane Drive j
Annapolis, MD 21401-7394 Edward A. Crooke President & Chief Operating Officer Baltimore Gas and Electric Company P. O. Box 1475 Baltimore. MD 21203-1475 James R. Curtiss, Esquire Partner Winston & Strawn p%
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Washington,' DC ~20005-3502.f Jerome W. Geckle -
Chairman of the Board (Retire'd)(
sPlfH Corporation.
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l PO Box 305 Maryland Line, MD 21105 Martin L. Grass President & Chief Operating Officer Rite Aid Corporation P. O. Box 3165 Harrisburg,PA 17105 Freeman A. Hrabowski,111 President University of Maryland Baltimore County 5401 Wilkens Avenue Catonsville,MD 21228 Nancy Lampton Chairman & Chief Executive Officer American Life and Accident Insurance Company ofKentucky 3 Riverfront Plaza Louisville, KY 40202 George V. McGowan Chainnan of the Executive Committee Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 i
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CALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT -
George L. Russell, Jr.
Partner Piper & Marbury 1100 Charles Center South 36 South Charles Street Baltimore, MD 21201 Michael D. Sullivan Lombardi Research Group, LLC 106 Old Court Road, Suite 303 Baltimore,MD 21208 Officers:
Christian H. Poindexter Chairman of the Board & Chief Executive OfYicer Baltimore Gas and Electric Company P. O. Box 1475 g-
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m Baltimore, MD 21203-1475 George C. Creel Senior Vice President, Generation.
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Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Thomas F. Brady Vice President, Customer Service Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 lierbert D. Coss, Jr.
Vice President, Gas Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Robert E. Denton Vice President, Nuclear Energy Baltimore Gas and Electric Company 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702 4
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CALTIMIRE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS
-FOR CALVERT CLIFFS NUCLEAR POWER PLANT Carserlo Doyle Vice President Electric Interconnection and Transmission Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Jon M. Files Vice President Management Services Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Ronald W. Lowman Vice President, Fossil Energy Baltimore Gas and Electric Company Fort Smallwood Road Complex 1000 Brandon Shores Road Baltimore, MD 21226 G. Dowell SchwaKJr'.i
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S Charles H. Shivery Vice President Finance and Accounting, Chief Financial Officer and Secretary Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Joseph A.Tiernan Vice President, Corporate Affairs Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Stephen F. Wood Vice President Marketing and Sales Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 Richard M. Bange, Jr.
Controller and Assistant Secretary Baltimore Gas and Electric Company l
P. O. Box 1475
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BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT I
Lynne H. Church Treasurer and Assistant Secretary Baltimore Gas and Electric Company P. O. Box 1475
-l Baltimore, MD 21203-1475 j
Thomas E. Ruszin, Jr.
Assistant Treasurer Baltimore Gas and Electric Company P. O. Box 1475
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Baltimore, MD 21203-1475 1.5 CInss and Period of License Applied For i
The Company requests renewal of the Class 104(b) operating license for CCNPP Unit 1 (license number DPR-53) for a period of 20 years beyond the expiration of the current license. The Company also requests renewal for the necessary source, special nuclear material and by-product licenses as may be necessary for the continued operation of the plant.
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Igg The nuclear statibn?knownf as! CCNPPfUnits 1 Land 2,iis;l'ocated onitheiwest! snore of the Chesapeake Bay in Cahiert County, Maryland, some'45 miles"sodtisast of Washingtoh, DC, and 3
60 miles south of;BiltimorcJ{0peratiostof the tvinLCombustion Engineering press 6rized-water reactors results m an approximate net electncal output of 845 megawatts for each reactor. Details j
concerning the plant and the site are contained in the Updated Final Safety Analysis Report for Calvert Clifts Nuclear Power Plant Units 1 and 2.
1.6 Construction Dates The Company does not propose to construct or alter a production or utilization facility with respect to this application.
1.7 Regulatory Acencies The Public Senice Commission of Maryland has jurisdiction over the rates and services provided by the Company's utility operations. Their address is:
Public Senice Commission of Maryland 6 St. Paul Centre Baltimore, MD 21202-6806 l
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BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54
.MPLE RESULTS a
CALVERT CLIFFS }\\, CLEAR POWER PLANT Local news publications which circulate in the area around CCNPP and which are considered appropriate to give reasonable notice of the application are:
Calvert Independent Newspaper P. O. Box 910 Prince Frederick, MD 20678 Calvert County Recorder P. O. Box F Prince Frederick, MD 20678 Enterprise Newspaper P. O. Box 700 Lexington Park, MD 20653 The Star-Democrat 29088 Airport Drive Easton, MD 21601 The Daily Banner 1000 Goodwill Road P. O. Box 580 P"
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1.8 Indemnity Agreement;i % x s^
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Attached as Enclosure 1 to this license application are the conforming changes to the indemnity agreement (10 CFR Part 140, Appendix B), which account for the expiration term of the proposed license.
1.9 Communientions All communications to the applicant pertaining to this application should be sent to:
Robert E. Denton Vice President, Nuclear Energy Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702 In addition, it is requested that copies be sent to the Company's General Counsel and Washington counsel:
D. A. Brune, Esquire General Counsel Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, MD 21203-1475 7
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BALTIMORE GAS & ELECTRIC COh1PANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT -
Jay E. Silberg, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, NW Washington, DC 20037 b
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BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT 2.0 TECIINICAL INFORMATION The Atomic Energy Act of 1954 limits the duration of the operating license for nuclear power plants to a maximum of 40 years; however, it does allow for the renewal of the license. License renewal is the terminology used to describe the regulatory requirements, technical scope, and environmental evaluations necessary to obtain approval for the continuation of power plant operation beyond the initial 40-year license period. Two principal regulations govem license renewal: Parts 51 and 54 of Title 10 to the Code of Federal Regulations. Part 51 addresses the requirements for the environmental aspects of license renewal and Part 54 addresses the remaining requirements.
The focus of Part 54 is to ensure that the intended functions of structures and components are maintained in the extended operating period by demonstrating that the effects of aging on long-lived, passive structures and components are being adequately managed. The technical information required by 10 CFR Part 54 provides this demonstration. 10 CFR 54.21 requires that an application contain an Integrated Plant Assessment (IPA), any changes to the current licensing basis made during Nuclear Regulatory Commission (NRC) review of the application, an evaluation of time-limited aging analyses, and an Updated Final Safety Analysis Report (UFSAR) supplement. Appendix A of this application contains the integrated plant assessment, the chan?.es to the currcint licensing basis, and the: tinic-limit'ed ~ aging' analyses; (including exemptions grantec snder 10 CFR 50.12).
9
BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT 3.0 UFSAR SUPPLEMENT 10 CFR 54.21(d) requires that a supplement to the UFSAR be provided as part of the application. This supplement contains a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analysis for the period of extended operation. The contents of the UFSAR supplement are based on the material found in Appendix A. In some instances, summary descriptions of programs and activities already exist in the Calvert Cliffs UFSAR. Those descriptions will be referenced in the supplement. After the license is granted, the UFSAR supplement will be incorporated into the Calvert Cliffs UFSAR as Section 1.9.
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BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLUTS NUCLEAR POWER PLANT 4.0 TECIINICAL SPECIFICATIONS i
Changes to the Technical Specifications or additions necessary to manage the effects of aging during the period of extended operation must be identified in the license renewal application (10 CFR 54.22).
Justification for the change or addition must be included in the application as well. Once the license is issued by the NRC, the proposed changes to the Technical Specifications will be incorporated into the plant Technical Specifications issued along with the license.
Any changes or additions to the Technical Specifications are identified and discussed in Appendix C. No Technical Specification changes or additions have be n identified at this time.
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1 BALTIMORE GAS & ELECTRIC COMPANY 10 CFR PART 54 SAMPLE RESULTS FOR CALVERT CLIFFS NUCLEAR POWER PLANT 5.0 ENVIRONMENTAL INFORMATION 10 CFR 54.23 requires that an environmental report that meets the requirements of 10 CFR Part $1, Subpart A be submitted with the application. Once the license is issued, the environmental information submitted with this application becomes historical information subject to being updated in accordance with 10 CFR Part 51. The environmental report is located in Appendix D.
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APPENDIX A f
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- TECHNICAL INFORMATION i
l Baltimore Gas and Electric Company 10 CFR Part 54 Sample Results for Calvert Cliffs Nuclear Power Plant Units 1 and 2 i-e
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TABLE OF CONTENTS i
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1.0 INTRODUCTION
I 1.1 Scope..................................................................................................................................
j 1.2 Cu rre nt Lice nsin g B asis Ch an ges..............................................................................................
I 1.3 Time. Limit ed A gi n g A n alyses...................................................................................................
l 1.4 Exemptions..................................................................................................................................
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1 2.0 I NTEG RATE D PL ANT ASS ESS M ENT M ETI10 D O LO G Y...............................................
2.1 Time-L i m i t ed A gi n g A n alys i s.....................................................................................................
i 3.0 CONTAINMENT........................................................................................................................
3.1 Co n t sin m en t S t ru c t u re............................................................................................................
I 4.0 R EA CTO R CO O LANT S YSTE M...........................................................................................
4.1 R e a ct o r Co ol a n t S y s t e m..............................................................................................................
4.2 ReaetorVessel.............................................................................................................................
4.3 Rea c t o r Ves s el I n t e rn al s.........................................................................................................
5.0 M E C l I A N I CA L S Y ST E M S...................................................................................................
l 5.1 Co n t ain m e n t I s o l a t i o n..................................................................................................................
5.2 N u c l e a r S u p p o rt Sys t e m s.............................................................................................................
5.3 DieselSystems............................................................................................................................
5.4 Co o li n g W a t e r S ys t e m s................................................................................................................
5.5 IIcatin g, Ventilatin g, Air Conditionin g Systems....................................................
l 5.6 FireProtection...........................................................................................................................
5.7 O t h e r Fl ui d Sy s t em s.................................................................................................................
5.8 Fee d wat e r/S t cam Syst em s.......................................................................................................
i 1
10 CFR Part 54 Sample Results i
Calvert Clifts Nuclear Power Plant
1 APPENDIX A TECIINICAL INFORMATION
1.0 INTRODUCTION
The Atomic Energy Act of 1954 limits the duration of the operating licenses for nuclear power plants to a i
maximum of 40 years; however, it does allow for their renewal. License renewal is the terminology used to describe the regulatory requirements, technical scope, and environmental evaluations necessary to obtain approval for continuation of power plant operations beyond the initial 40-year license. Two principal regulations govern this process: Parts 51 and 54 of Title 10 of the Code of Federal Regulations. Part 51 addresses the requirements for the environmental aspects of license renewal, and Part 54 addresses the remaining requirements. This appendix addresses the technical requirements outlined in Part $4.
1.1 Scope This appendix will address the requirements of 10 CFR 54.21(a)-(c). These sections of Part 54 require that information pertaining to the integrated plant. assessment, current licensing basis changes, and time-limited aging analyses be addrbssed isthe applicati6n;.
1.2 Current Licensing BEsis Cliangesi i
As required by 10 'CFR 5411(bkcurrent licensing basis' changes, which hhve a" material effect on the content of this application, will be identified at least annually while the application is under review. Those changes will be identified in Table 1-1. Changes will be made as amendments to the license application.
1.3 Time-Limited Aging Analyses 10 CFR 54.21(c) requires that a list of time-limited aging analyses be provided in the application. In addition, the application must discuss the validity of the analyses and the adequate management of the cffects of aging for the period of extended operation. The results of the evaluation of the time-limited aging analyses are provided in Table 1-2 (IJTER), Section 2.1 provides more information concerning the i
validity of the analyses and the adequacy of the aging management programs (IJ TER).
1.4 Exemptions A list of plant-specific exemptions which are based on time-limited aging analyses and are still in effect is required by 10 CFR 54.21(c)(2). An evaluation must be provided for each such exemption which justifies the continuation of that exemption for the period of extended operation. No exemptions have been identified which are still in effect and are based on time-limited aging analyses.
10 CFR Part $4 Sample Results 1-1 Calvert Cliffs Nuclear Power Plant Revision 0
APPENDIX A
-TECIINICAL INFORMATION TABLEl-1 i
CURRENT LICENSING BASIS CITANGES 1
i Revision 0 This is the initial issue of the application. No licensing basis changes are identified which could materially I
affect the application.
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10 CFR Part 54 Sample Results 1-2 Calvert Cliffs Nuclear Power Plant Revision 0
APPENDIX A TECIINICAL INFORMATION 2.0 INTEGRATED PLANT ASSESSMENT METilODOLOGY The IPA for CCNPP consists of two tasks: 1)detennining which structures, systems and components (SSCs) are subject to aging management review; and 2) performing that review.
During the scoping process, plant SSCs and their functions are reviewed to identify components which are within the scope oflicense renewal. The use of the plant's integrated electronic system of information provides continuity at system interfaces to ensure that all appropriate components are reviewed. The fo owing criteria, as defined in 10 CFR 54.4, are used to identify SSCs which are within the scope of n
license renewal:
Safety-related SSCs which are relied upon to remain functional during and following design basis events to ensure:
Th6 integrity of the reastor coolant pressure bouhd$ryi a.
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The capability..to shut'down the reactor and maintain 'it in a' safe. shutdown condition; or ThE cap 5i1[ty to prevent ~or mit'igat'e'the consequences of' accident's"that could result in
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c.
potential ofTsite exposures that are comparable to the guideline exposures of 10 CFR Part 100.
Non-safety-related SSCs whose failure could prevent satisfactory accomplishment of any of the functions identified above.
Structures, systems and components relied on in safety analyses or plant evaluations to perform a e
function that demonstrates compliance with NRC regulations for:
a.
Fire Protection (10 CFR 50.48);
b.
Environmental Qualification (10 CFR 50.49);
c.
Pressurized Thennal Shock (10 CFR 50.61);
d.
Anticipated Transients without Scram (10 CFR 50.62); and c.
Station Blackout (10 CFR 50.63).
Scoping is first conducted at the system and structure level to identify structures and systems within the scope of license renewal and their intended functions. Next, the components of scoped structures and systems are reviewed to detennine which components contribute to the intended functions identified during the system level scoping process. The results of this step is a listing of plant SSCs within the scope of license renewal and the intended functions which need to be preserved during the period of extended operation.
10 CFR Part 54 Sample Results 2-1 Calvert Cliffs Nuclear Power Plant Revision 0
APPENDIX A TECilNICAL INFORMATION The IPA is conducted in two parts. First, the SSCs within the scope of license renewal are reviewed to determine which are subject to an aging management review. Any structures and components which only contribute to active functions, or which are replaced based on a qualified life, or specific time period, are excluded from an aging management review. The remaining structures and components are evaluated for j
the effects of aging to ensure that plant programs and activities will manage these efTects so that the intended functions will be preserved.
A more detailed description of these processes is documented in Calvert Chyfs Nuclear Power Plant integrated Plant Assessment Methodology, which has been submitted to the NRC.
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i 10 CFR Part 54 Sample Results 2-2 Calven Cliffs Nuclear Power Plant Revision 0
i AFPENDIX A
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TECIINICAL INFORMATION 3.0 CONTAINMENT i
3.1 Containment Structure The Containment Structure components were scoped in accordance with the process described in Section 2.0. As a result of this scoping,19 containment structural component types were determined to require funher aging management. A list of these component types is shown in Table 3-1.
Functions Within the Scope of License Renewal The Containment Structure is a seismic Class I building whose structural components proside support and shelter to safety-related and non-safety-related equipment inside the Containment (see UFSAR Section 5-1). The components addressed by the evaluation for this structure included all Containment structural, components sening such. functions; and components comprising the pressure boundaiyLbutLdid not' include' items such asLpipe supports, which are addressed in a separate evaluationiStrudural components sithin this system boundary inclu'de supports for the steam generatorsc In general, Containment Structure components provide seven functions within the scope oflicense renewal:
a.
Provide structural and/or functional support to safety-related equipment.
1 b.
Provide structural and/or functional suppon to non-safety-related equipment where failure of thi' equipment could directly prevent satisfactory accomplishment of safety-related i
funnons.
l c.
Provide shelter or protection to safety-related equipment, including radiation shielding for equipment qualification and high-energy line break protection.
d.
Serve as a pressure boundary or fission product barrier to protect public health and safety in the event of any postulated design basis events inside the containment.
c.
Serves as a barrier against internal or external missiles.
f.
Provide a protective barrier for internal flood event.
g.
Provide a rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plant.
Management of Component Aging ne effects of aging considered for the Containment Structure included: freeze-thaw, leaching of calcium hydroxide, aggressive chemical attacks, reaction with aggregates, corrosion in embedded steel /rebar, creep, shrinkage, settlement, corrosion, prestressing losses, weathering, elevated temperature, irradiation, and fatigue. These types of aging can be broken into those affecting reinforced concrete structures and those affecting exposed steel structures.
10 CFR Part 54 Sample Results 3-1 Calvert Cliffs Nuclear Power Plant Resision 0
APPENDIX A TECIINICAL INFORMATION Reinforced Concrgle Structures The concrete used in the containment dome and walls was designed and constructed in accordance with the requirements specified in American Concrete Institute (ACI) standards and American Society for Testing Materials (ASTM) specifications as identified in the UFSAR. These standards and specifications provide the physical property requirements of the aggregate and air-entraining admixtures, chemical and physical requirements of air-entraining cements, and proportioning of concrete including entrained air to maximize the concrete resistance to freeze-thaw action, leaching of calcium hydroxide, reactivity with alkalis, and reactivity with aggregates. In addition, these specifications assure that the concrete is of low permeability which provides resistance against attacks by aggressive chemicals. A walkdown inspection of the Unit 1 Containment Structure in 1992 (Reference 1) found no conditions which would suggest that unexpected or non-repairable deterioration had occurred. No management programs are deemed necessary for these aging effects.
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Protective soatingfarciused ori the containment exteriorFand interiorbThh intended function of y
inside-containment coatind $ tol protect sirdctura[codiponentsiffom?ang corrosive environment.
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Interior coating fhfure can potentially impact"ths cheration of slid emergen'cy. sump in the event of
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a design basis a6ciden't inside the containment #Although Calvert Cliffs Nuclear Power Plant site is located in a geographic region subject to severe weathering conditions, the weather only affects exterior surface coatings and does not affect internal coatings. Therefore, no management programs are necessary for weathering effects of coatings within the Containment Structure.
I Creep, shrinkage, and settlement are mechanisms whose potential for causing age-related degradation diminishes with time (References 2 and 3). In the cases of creep and shrinkage, the use of low slump concrete was incorporated into the design to minimize susceptibility to these effects. In addressing settlement, it should be noted that the containment basemat is situated on
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Miocene soil, which is exceptionally dense and easily supports heavy foundation loads. The design pressure of the soil was set as 8,000 psf, and the soil bearing pressure was about the same as the overburden removed as a result of excavation. In addition, a dewatering system is installed to minimize the fluctuations in the groundwater table that might contribute to settlement; thus providing stable geological conditions at the plant site. Long-term settlement is not expected to continue after 40 years. Because the effects of creep, shrinkage, and settlement are not expected to continue into the period of license renewal, no management program is necessary for these aging mechanisms.
The ambient bulk temperature inside the containment during normal plant operation is limited to 120'F, and the concrete surface temperature for the design of the containment is limited to 150'F.
Although the primary shield wall is subjected to sustained internal heat buildup, themtal insulation and an air-cooling system are provided at the inner surface of the reactor cavity wall to maintain the concrete temperature at or below 150'F. The reactor coolant pump bays, however, experience higher local temperatures which are assumed to be 160 F (Reference 4). American Society of Mechanical Engineers (ASME), Section 111, Division 2 (Reference 5) indicates that as long as concretc/ grout temperatures do not exceed 150 F, aging due to elevated temperature exposure is not significant. In addition, ACl-349-85 (Reference 6) allows local area temperatures to reach 200 F before special provisions are required. Thus, because the bulk and localized temperatures are below the thresholds of elevated temperature as stated in References 4, 5, and 6, concrete 10 CFR Part 54 Sample Results 3-2 Calvert Cliffs Nuclear Poner Plant Revision 0
APPENDIX A 1
TECIINICAL INFORMATION degradation caused by elevated temperatures is not expected to affect the function of the concrete containment structures. No aging management program is necessary for elevated temperatures.
Concrete fatigue strength is about 55 percent of its static strength at extremely high cycles of loading. Internal concrete components of the Containment Structure were designed in accordance with ACI-318-63 (Reference 7). The design code limits the maximum permissible design stress level to less than 50 percent of static strength, which is less than the fatigue strength of concrete.
Therefore, fatigue will not degrade any concrete components in the containment, and an aging management program for concrete fatigue is not required.
Irradiation over a specific threshold may degrade concrete due to aggregate growth, decomposition of water, or thermal warming. However, 40-year normal service doses from Reference 4 cxtrapolated to 60 years remain below the specified irradiation threshold. No aging management program is necessary for.moni.toring irradiation and irradiation mduced effects.
Steel Stmet'urefand_QLmagnqntd y,
l'l s
The only signficant'inventprp..of a'ciEichhemihalinormally[lycated[in$idelthe containment is i
borated waterJwhich'is primarily in safety-related systems.EBecause only a timited amount of t
leakage is allowed by the Technical Specifications for these systems, undetected leakage of borated water for an extended period of time should not occur. Thus, the containment's interior surface and all internal stmetural components have only minimal, localized exposure to aggressive chemicals. Embedded stecurebar can be exposed to the outside environment, if concrete coverage is insufficient. However, the concrete used for the Containment Structure complies with the relevant ACI standards and ASTM specifications for low permeability, and proper concrete cover was specified and installed in accordance with ACI-318-63 (Reference 7) to effectively prohibit exposure of embedded steeVrebar to potentially corrosive environments (i.e., the outside environment). In addition, rebar requirements minimize concrete crack development which, in turn, reduces the likelihood of exposure of embedded steeUrebar to the outside environment. A walkdown inspection performed in 1992 (Reference 1) confirmed no damage to the visible portion j
of the Containment Structure at CCNPP caused by corrosion of embedded steeVrebar.
Steel exposed to moisture and oxygen can corrode. However, corrosion was considered in the original design of structural steel components used in the containment. As a result, all exposed anchors and structural steel surfaces in the containment except grating, checkered plates, and metal decking (which are all galvanized steel) are shop-or field-painted during the construction phase.
Application of coatings at CCNPP follows standard procedures which also address the refurbishment of existing coated surfaces and the need for new coating and recoating. In addition, these items are visually inspected in accordance with existing plant procedures. No additional program for steel components in containment is needed for those items that are accessible to visual inspection. Modifications to current programs will be made to include a representative sample of items that cannot be readily visually inspected.
Irradiation beyond a certain threshold can affect the reinforcing steel, structural steel, liners, and tendons of the Containment Structure reducing the ductility of the metal. However, the 40-year normal service doses from Reference 4 cxtrapolated to 60 years remain below the specified 10 CFR Part 54 Sample Results 3-3 Calvert Cliffs Nuc1 car Power Plant Revision 0
APPENDIX A TECliNICAL INFORMATION irradiation threshold. No management program is necessary for irradiation and irradiation induced effects.
Fatigue occurs as a result of periodic or cyclic loadings that are less than the maximum allowable static loading and results in progressive, localized damage to structural materials.
Steel components in the containment subject to high-cycle loading were designed in accordance with American Institute of Steel Construction, Inc. (AISC) specifications (Reference 8) to address high-cycle loading. The maximum stress in steel components and connections is less than the fatigue limit of steel. Therefore, no program to manage the effects of fatigue is required.
Actions to specifically address managing the effects of aging on a particular component type are described in the following paragraphs:
Internal Structural Steel}jembers (e c.. Grating. Plates. and Metal _ Decking) and Lubrite Plates a
Some struckurd(astsel}medershvithin the. Cdntainment Structurel(such{asirating, plat metal decking)ffidjubritiplateise n'of rbdil l accessible forNisifal; inspection. A strategy will be developdf foFselecting aispreientative'lample of th6se:componentseEnisting plant procedures for visual MaiingGinspeetionhan'dicercetiyermeasur's5willkise?ihddifidd to include this e
representative sample of components. Lubrite plates at the steam generator supports will be inspected to determine their condition ofthe steam generators are replaced. The modifications to the existing visual inspection program will ensure that the intended function of all components not madily accessible is not impaired.
Containment Liner The containment dome and wall liner is covered by existing programs requiring coating visual inspections. Ilowever, for liner locations that are not readily accessible for visual inspections, a one-time inspection will be performed.
Refueline Pool Liner Leakage from the refueling pool is currently monitored. Continued monitoring of leakage will ensure that the intended function of the refueling pool liner is not impaired.
Below-Grade Portion of Containment Walls and Basemat 1
Chemical attack of the below-grade portions of the Containment Structure walls and basemat is j
only possible if groundwater chemical composition is significantly more aggressive than it was found to be during pre-construction testing. Groundwater observation wells installed during initial plant construction will allow investigation of the ground water chemistry (if needed) to ensure that groundwater composition is not significantly changed. This will ensure that the intended function of the below-grade portions of the Containment Structure walls and basemat is not impaired.
10 CFR Part $4 Sample Results 3-4 Calvert Cliffs Nuclear Power Plant Revision 0
APPENDIX A TECIINICAL INFORMATION Basemat Liner If groundwater investigation suggests that acid attack on the exterior of the basemat liner is occurring, evaluation of the basemat liner would be performed in conjunction with Appendix J testing. This will ensure that the intended function of the basemat liner is not impaired.
ContainmentPrestressed Tendons Technical Specification 4.6.1.6.1 addresses checking containment tendon lift-off force. The associated frequency interval was developed during the containment design and was based on nuclear industry experience, considering the corrosion resistant properties of the tendon material and the environment to which this material is subjected. Plant procedures provide acceptance criteria for the prestress !evel and t' ndon system component physical conditions over the 40-year e
operating life.x Prestress. lcsses are time-dependent and are reficcted in the curves of expected lift-off force ydrsus timefj. Hbwever[these curves areleurrently onip established foi 40 years and will be re-evaluateditoicstabl.ishethefprbdibted prestress levels 91uring theiliceiise renewal period.
Proceaures "also niandate the use"of visdal inspections of the tend 6ns to detect' indications of tendon corrosion aEd *the..useT6filsb8 rat $rpt estinglto: ensure 7tliatothilslienthing[fdler is sufficient to t
maintain itsIfundtionMThhunininimn prsstressflev010and the%hysicullcofidition requirements i
currently established will not change. Inspections performed in accordance with existing plant procedures (using modified lift-off versus time curves) will be adequate to detect and correct loss of prestress and corrosion in containment tendons.
References l.
Eramination of the Unit 1 Containment Structure - Calvert Chfs Nuclear Power Plant, August 1992 2.
Prediction ofCreep, Shrinkage, and Temperature Effects in Concrete Structures (ACI-209R-82),
American Concrete Institute,1982 3.
Calvert Clifs Nuclear Power Plant, Units 1 and 2, Updated Final Safety Analysis Report, Baltimore Gas and Electric Company 4.
EQ Design Manual-Calvert Chfs Nuclear Power Plant Unit No.1 and 2, Baltimore Gas and Electric Company 5.
Codefor Concrete Reactor Vessels and Containments, ASME Boiler and Pressure Vessel Code,Section III, Division 2,1986 6.
Code Requirements for Nuclear Safety Related Concrete Structures (ACI-349-85), American Concrete Institute,1985 7.
Building Code Requirements for Reinforced Concrete (ACI-318-63),
American Concrete Institute,1963 10 CFR Part 54 Sample Results 3-5 Calvert Cliffs Nuclear Power Plant Revision 0
APPENDIX A TECIINICAL INFORMATION J
8.
Calvert Chfs Nuclear Power Plant, Units 1 and 2, Design Criteria Manual, Baltimore Gas and Electric Company TABLE 3-1 CONTAINMENT STRUCTURE COMPONENTS REQUIRING r
)
Concrete Containment Walls Concrete Basemat Steel Columns Steel Beams Lubrite Plates._,,
, Baseplates Floor Framingi.!
Steel Blacings, Platform Hangersi
)Deckingj Floor Grating;.
? Checkered Plates i:.
Post-Tensioning Systeth '
- iCrane Girder vg Containment Linerb L Basemat Liner 4 - ~
e t
Refueling Pool (Liner)
Post-Installed Anchors Cast-in-Place Anchors s
1 i
l 10 CFR Part 54 Sample Results 3-6 Calven Cliffs Nuclear Power Plant Revision 0 l
r APPENDIX A TECIINICAL INFORMATION 4.0 REACTOR COOLANT SYSTEM 4.3 Reactor Vessel Internals l
He reactor vessel internals were determined to be within the scope oflicense renewal using the process I
described in Section 2.0. Since all of the system components are considered to be passive and long-lived, an aging management review was performed on all system components. A list of the evaluated components is shown in Table 4-1. Two of the reactor vessel internals components, the flow baffle and core support barrel snubber / snubber bolts, have been included in the IPA evaluation for the reactor vessel, and, therefore, are addressed in Section 4.2.
Functions Within the Scope of License Renewal He components;of the reactor vessel internals are divided into three major groups (see UFSAR 1
Section 3.3.3)i.the: core l support; barrel,;the lowericore support structure (including the core f
shroud), and the upper guide structure (incl.uding the conirol blement assembly (CEA) shrouds and the in-core instrumentatfori gu_ide tubes).~ IUpdated Final Safety: Analysis: Report Figures 3.1-1, 3.3-1,3.3-6,i 3.3-11,3.3-13, and 3.3-14 depict the components of the reactor vessel internals.
The major support member of the reactor vessel internals is the core support assembly which consists of the core support barrel, the lower support structure, and the core shroud. The core support assembly is supported at its upper end by a flange which rests on a ledge in the reactor vessel flange. The lower flange of the core support barrel supports and positions the lower support structure. The lower support structure provides support for the core by means of a core support plate supported by columns resting on beam assemblies. The core support plate provides support and orientation for the fuel assemblics. The core shroud, which provides lateral support for the fuel assemblics, is also supported by the core support plate. The lower end of the core support barrel is restrained radially by the core support barrel snubbers.
The upper guide structure assembly consists of the upper support plate, the CEA shrouds, a fuel assembly alignment plate and an expansion compensating ring. The upper guide structure assembly aligns and laterally supports the upper end of the fuel assemblics, maintains the CEA spacing, prevents fuel assemblics from being lified out of position during a design basis accident
]
and protects the CEAs from the efTect of coolant cross flow in the upper plenum.
'l The reactor vessel internals provide four functions within the scope oflicense renewal a.
Support and orient the reactor core fuel assemblics and control element assemblies.
b.
Absorb dynamic loads and transmit them to the reactor vessel flange.
c.
Direct coolant flow through the core.
d.
Support and orient in-core instrumentation.
10 CFR Part 54 Sample Results 4-1 Calvert Cliffs Nuc! car Power Plant i
i APPENDIX A 1
TECIINICAL INFORMATION Management of Component Aging j
\\
The aging effects considered for the reactor vessel intemals included: Eeneral corrosion, pitting, irradiation-assisted stress corrosion cracking (IASCC),
stress corrosion cracking (SCC)/intergranular stress corrosion cracking /intergranular attack, microbiologically induced corrosion, crosion, crosion/ corrosion, neutron embrittlement, hydrogen danage, low-and high-cycle fatigue, oxidation, wear, creep, and stress relaxation.
i i
Strict controls on water chemistry for the reactor coolant, in conjunction with the carefully selected materials used in the Reactor Coolant System, prevent certain degradation mechanisms from affecting components of the reactor vessel internals. Degradation mechanisms not considered credible based on coolant chemistry and reactor vessel intemals component material included:
general corrosion, pitting / crevice corrosion, microbiologically-induced corrosion, hydrogen damage, and. exidation (References 1 and 2).mln addition, tensile stresses in many components are j
not sufficien'ltolinitiateiSCCiandlintersranularStressfcorrosionfcrasklidg and, for these t
components;Jno emanagemen4 program
- 1 j s {necessary] for? these} degradation mechanisms i
(Reference 1).1.. "
i
~
m n
Normal operating' pressureffluid velocitiesf arid partic$ late levhls' found inf he Reactor Coolant t
System are not sufficient to produce crosion and erosion / corrosion effects for the materials in the reactor vessel internals (References 1,2, and 3). The operating temperature of the system is not sufficient to induce creep for any material used in this system or to induce thermal aging effects for most of the materials used in the system (References 1, 2, and 3). Aging management for those materials for which thermal aging is a plausible degradation mechanism is discussed in the sections to follow. Thus, in general, no management program is necessary for erosion, erosion / corrosion, creep, and thermal aging of the reactor vessel internals.
1 For components that do not rely on preloading for their functionality, no program for stress relaxation is necessary. Ilowever, stress relaxation is addressed for those components which depend on preloading. Programs for these components are discussca in sections to follow.
j Components that are not subject to relative motion associated with sliding, flow induced vibration, loss of clamping force, or from thermal effects are not subject to degradation due to wear. Thus, for components not subject to relative motion, no management program is necessary for wear.
Aging management for components that are subject to relative motion are addressed in sections to follow.
i One of the series ofindustry reports developed by the Nuclear Management and Resources Council and a subsequent summary document of the technical information contained in this series of industry reports (References 1 and 3) discusses degradation mechanisms and their significance as applied to the reactor vessel internals. In general, the programs for the management of the various types of degradation, discussed below, are in agreement with the conclusions documented in Reference 3.
Actions to specifically address managing the effects of aging on a particular component type (or groups of component types) are described in the following paragraphs.
Since detailed fluence levels for each component were not determined, all reactor vessel internal 1
components within the scope of license renewal were considered to be potentially susceptible to 10 CFR Part 54 Sampic Results 4-2 Calvert Cliffs Nuclear Power Plant i
j,.
l APPENDIX A I
TECIINICAL INFORMATION aging due to neutron embrittlement and IASCC. Neutron embrittlement manifests itself as loose, missing, cracked or fractured parts, bolting, or fasteners. Calvert Cliffs Nuclear Power Plant procedures implement ASME Section XI Inservice Inspection (ISI) VT3 inspections (per B-N-3 of subsection IWB) for accessible reactor vessel internals components.Section XI ISI will detect the precursors of failure from neutron embrittlement. To datefthere has been no evidence that IASCC has actually occurred in pressurized-water reactors (PWRs). However, CCNPP will use the Section XI Isis to monitor reactor vessel internals for signs of IASCC, In addition, industry activities will be monitored to determine whether additional, more focused activities are needed based upon industry experience and research. Thus, neutron embrittlement and IASCC will be effectively managed for all reactor vessel internals for the period of license renewal.
Lower Suopon Structure Beam Assembliss and Incore Instrumentation (ICD Thimble Suopm1 Elaig Neutron embMEmentEdIA dC Are thd onikagingimechanisms cons'dere[ plausible for these i
component groupsEAi discussed in the previous section,LASME Section XI ISI VT-3 inspections and monitoring industry activities with'resjsect to'IASCC will_ be employed lto manage these aging mechanism 6 Thus, the intended functions' of the lo ver support structure beam assemblics and the.
ICI thimble' support plate will not be impaired.
Core Support Barrel (CSBL CSB Alienment Kev. Upper Guide Structure Support Plate. CEA Shroud F,xIgnsion Guide. and Fuel Alignment Pin _s Wear is a plausible degradation mechanism for components that are in close proximity and that may come into direct contact during vibration. The core support barrel upper flange, CSB alignment key, upper guide structure support plate, CEA shroud extension guide, and fuel alignment pins have the potential for wear. The remaining core support barrel (i.e., excluding the upper flange) components are not subject to wear. Since wear is the normal focus ofISI, existing procedures implementing ASME Section XI ISI VT-3 inspections will ensure that accessible components are visually inspected to detect indications of wear. Thus, existing procedures will ensure that the intended function of the core support barrel, CSB alignment key, upper guide structure support plate, CEA shroud extension guide, and fuel alignment pins will not be impaired.
Core Support Plats Low-cycle fatigue has been determined to be a plausible age related degradation mechanism for the core support plate. Baltimore Gas and Electric Company's fatigue monitoring program for Reactor Coolant System components determines " critical components" based on the necessity of their function for safe plant shutdown and upon their expected thermal fatigue accumulation.
Controlling transients for fatigue accumulation were determined, and, based on operating procedures and stress reports, fatigue accumulation equations for each of the " critical components" were developed. A one-time analysis will be performed to demonstrate that low-cycle fatigue of the core support plate is insignificant when compared with the " critical components" monitored by the fatigue monitoring program. Continued monitoring of these
" critical components" will ensure that fatigue in other portions of the system bounded by these
" critical components" is also adequately managed. Therefore, the intended function of the core support plate will not be impaired.
10 CFR Part 54 Sample Results 4-3 Calvert Cliffs Nuclear Power Plant
APPENDIX A TECIINICAL INFORMATION Core Suonort Columns Cast austinctic stainless steel (CASS) is used in some reactor vessel internals components (such as the core support columns) and is subject to thermal aging for the temperature range experienced by the system. Thermal aging may manifest itselfin the form of crack initiation under certain loading conditions. He need to manage this effect depends on the delta ferrite content of the CASS material. If the delta ferrite content of CASS components is less than the limit expressed in Reference 1, thermal aging is not a significant age related degradation mechanism. If the delta ferrite content of the core support columns is confirmed to be within this limit, aging management for thermal aging would consist of continued monitoring ofindustiy experience and research.
De fatigue monitoring program in conjunction with a one-time analysis, as discussed in the " Core Support Plate" section, will also manage the effects of low-cycle fatigue for the core support columns. Thus,.the p.rograms. outlined above,will ensure that;the; intended function of the core support colu'mns is not.. impaired;
~q L %A3 NO; "s
Core Shroud TiiRgd 7 [t <
ik j ew h
3 For the core'shfoud tie r$ds"(nith.hs' exception of thE cme shroud ~tieTod lock straps, for which 4
stress relaxation is not a concern), low-cycle fatigue and stress relaxation are considered to be plausible aging mechanisms. As discussed in the " Core Support Plate" section, low-cycle fatigue will be managed by the existing fatigue monitoring program at CCNPP in conjunction with a one-time analysis.
As the reactor vessel internals age, preloaded components could undergo stress relaxation.
Although, stress relaxation has not yet been experienced in the range of temperatures present for PWR internals components, EPRI NP-5461 (Reference 4) documents that in-pile results at 550 F show that stress relaxation can occur in bolts made of 304 stainless steel at neutron fluence levels of 6 x 10" neutrons /cm. The core shroud tie rod environment meets these conditions.
2 The level of concern from loss of prestress is low for the core shroud tie rod components because the core shroud and upper guide structure are in compression from the expansion compensation ring and the vessel head. Any relaxation of this compression will be detected and managed by inspection of the expansion compensation ring and interfacing components.
In addition, the rods are in an area of relatively low coolant flow. Hus, loss of preload in these components is not likely to cause excessive vibration or fatigue for fasteners.
Calvert Cliffs Nuclear Power Plant will continue to monitor indust:y activities in this area. If industry experience demonstrates the need, more focused activities will be initiated. Thus, the intended function of the core shroud tic rods will not be impaired.
10 CFR Part 54 Sample Results 4-4 Calvert Clifts Nuclear Power Plant i
APPENDIX A TECIINICAL INFORMATION Exoansion Cornpensation Rinn Stress relaxation and wear have been determined to be plausible age-related degradation mechanisms for the expansion compensation ring. As discussed in the " Core Shroud Tie Rod" section, stress relaxation has not yet been experienced in the range of temperatures present for PWR internals components. ASME XI ISIS will continue to be perfomied in accordance with existing plant procedures. As discussed in the "CSB, CSB Alignment Key, Upper Guide Structure Support Plate, CEA Shroud Extension Guide, and Fuel Alignment Pins" section, the ISI program will be effective in managing wear of the reactor vessel internals. Thus, the programs outlined above will ensure that the intended function of the expansion compensation ring will not be impaired.
Core Shroud
^ PWM dT:Tn9 s.
.;:y 7,
g :. N"m ; m y
=q
! :3. - ~.:. - m:::f:Jt The socketiheadRcap? screws Jxoff theEcore!shroudKhhve,lbeen; fabricated! using iron-based superalloy Ai 6, which isisusb6ptible to SCCiExisting procudures implementing ASME XI ISI 28 casure that these(com;ionernsjare'visdall/irisisscted to.dstect any loose parts Or gross indications of structural fastener failure' thafmaylindicate! SCC l LPrevious inspections 4f the core shroud at CCNPP hafc found rio ind' cation"of SCC?
1.
n
~
i Stress relaxation is a plausible age-related degradation mechanism for the socket head cap screws, the hex head nuts, and the hex head heavy nuts. As previously discussed in the " Core Shroud Tie Rod" section, stress relaxation is addressed by ASME XI ISIS of other components and by the continued monitoring ofindustry activities in this area.
Low-cycle fatigue is possible for core shroud components with the exception of the head cap screws, hex head nuts, and hex head heavy nuts. Actions as described in the " Core Support Plate" section (one-time analysis and continuation of the fatigue monitoring program) will be taken to address low-cycle fatigue for these components. Thus, the programs outlined above will ensure that the intended function of the core shroud is not impaired.
Fuel Alinnment Plate / Guide Luc and Inserts For the fuel alignment plate socket head cap screws (which are fabricated using A-286), SCC and stress relaxation are considered to be plausible aging mechanisms, and for the, remaining components of the fuel alignment plate, wear is considered to be a plausible aging mechanism. As discussed in the " Core Shroud" section, SCC is addressed by existing procedures implementing ASME XI ISIS. As discussed in the " Core Shroud Tie Rod" section, stress relaxation will be addressed by ASME XI ISIS of other components and by monitoring industry activities within this area. As discussed in the "CSB, CSB Alignment Key, Upper Guide Structure Support Plate, and CEA Shroud Extension Guide" section, wear is adequately managed by the existing ISI program.
Thus, the programs outlined above will ensure that the intended function of the fuel alignment plate is not impaired.
10 CFR Part 54 Sample Results 45 Calvert Clifts Nuclear Power Plant
APPENDIX A TECilNICAL INFORMATION CEA Shroud and Bolts Because the components comprising the CEA shroud and bolts are constructed of various materials, the plausible age-related degradation mechanisms are addressed by component, rather than for the CEA shroud as a whole.
Stress corrosion cracking and stress relaxation are considered as plausible age-related degradation mechanisms for the CEA shroud socket head cap screws, which are fabricated using A-286. Stress corrosion cracking will be addressed in the same manner as described for the core shroud (using ASME XI Isis). Stress relaxation will be addressed by continuing to monitor industry aethities and to initiate focused actions ifindustry experience demonstrates such a need.
Stress relaxation is considered to be a plausible age-related degradation mechanism for the CEA shroud intemal and _expmal spanner nut and tabs The concem from. loss of prestress on these r
fasteners is low because the bore shroud and '6pper guide structure are under compression from the expansion dompensation ring and vessel head, andf thus,2 excessive vibrationTor fatigue for these fastencrs isln6t likely to'oc'ctirt ' s~ discuss'ed previously?ASME XI ISIS Lof other components and A
monitoring Indu~strylacthities iri this area (And initiating other actions if dethmihed to be necessary i
by industry'expericisce) will satisfactorily address stress relax:itionE High-cycle fatigue is considered to be a plausible age-related degradation mechanism for the CEA shroud retension pin /retension block / positioning pin. The nature of high-cycle fatigue is such that it is normally detected early in plant life, making it highly unlikely that components subject to high-cycle fatigue will fail during the period of extended operation. In addition, the consequences of such a failure, ifit were to occur, would be low, as the failure would manifest itself as a loose part in the exit region of the vessel. Because of the low probability and consequence of a failure, documenting and investigating any loose parts discovered during nonnal refueling operations is considered to be satisfactory in addressing the effects of high-cycle fatigue.
Wear and high-cycle fatigue are considered to be plausible age-related degradation mechanisms for the CEA shroud suppon/ guide support / channel support and the CEA shroud base / shroud / lower shroud / transition plate / flow channel / guide / instrument tube / insert / channel / channel cap.
As previously discussed in the "CSB, CSB Alignment Key, Upper Guide Structure Support Plate, and CEA Shroud Extension Guide" section, wear, which is the focus of the existing ISI program, will j
be adequately managed by that program for the period of extended operation. As discussed previously, high-cycle fatigue will be managed by documenting and investigating any loose parts discovered during refueling operations.
Wear, high-cycic fatigue, and thermal aging are considered to be plausible age-related degradation mechanisms f'.,r the CEA shroud tube which is fabricated from cast austenitic stainless steel. As i
presiously discussed in the "CSB, CSB Alignment Key, Upper Guide Structure Support Plate, and CEA Shroud Extension Guide" section, wear, which is the focus of the existing ISI program, will be adequately managed by that program for the period of extended operation. As discussed previously, high-cycle fatigue will be managed by documenting and investigating any loose parts discovered during refueling operations. As discussed in the " Core Support Columns" section, the need to manage thermal aging depends upon the delta ferrite content of the CASS material of the components. If the delta ferrite content is confirmed to be within the limit stated in Reference 1,
'l 10 CFR Part 54 Sample Results 4-6 Calvert Cliffs Nuclear Power Plant
APPENDIX A TECilNICAL INFORMATION aging management for thermal aging would consist of continued monitoring ofindustry experience and research.
Hus, the intended function of the CEA shroud and bolts will not be impaired.
References 1.
Pressurized Water Reactor Pressure VesselInternals License RenewalIndustry Report, Nuclear Management and Resources Council, Report Number 90-05, December 1992 2.
Pressurized Water Reactor Aging Degradation Study, Phase 1,
NUREGICR-6048, September 1993 c.-
gw.,.e v..-
..>-m 3.
Summary of Technical Information"and Agreements from lluclear Management and Resources Councilindusirj Reports Addressin$ License Renewal, drkft NUREGICRi????, November 1994 p.---
n e 3
j n
ComponenE'}lthEstimation.n,.#:OlJYR StreictusalJMaterials Degradation Mechanisms, NP-54 E.. f.:
. ~, id,..~.:
.....L
.a 4.
Electric Power Research Institute -
s u=
l 1
i 10 CFR Part 54 Sample Results 4-7 Calvert Clifts Nuclear Power Plant
[-
i APPENDIX A y
TECIINICAL INFORMATION TABLE 4-1 i
REACTOR VESSEL INTERNALS COMPONENTS RECEIVING AN AGING MANAGEMENT REVIEW Flow Baffle
- Core Support Barrel Snubber
- Core Support Barrel Alignment Key Lower Support Structure Beam Assembly Core Support Barrel Core Support Plate Fuel Alignment Pins Core Support Columns.,
Core ShroudTie Rod (.
4 ICI. Sap' ort Platef p j,ly,
~
p u
4 Exparision_ Coispe$sdiion;Ringh ' j{,
.j~, -
Upper Guide Structure S0pport Platd V M CEA Shroud Extension Shaft Guides !
s a
CEA Shroud and Bolts Fuel Alignment Plate Core Shroud The noted device types are included in the evaluation for the reactor vessel.
10 CFR Part 54 Sample Results 48 Calvert ClitTs Nuclear Power Plant
APPENDIX A TECIINICAL INFORMATION 5.0 MECIIANICAL SYSTEMS 1
1 5.4 Cooling Water Systems 5.4.3 Saltwater System The Saltwater System components were scoped in accordance with the process described in Section 2.0.
Two major categories of components were evaluated with respect to management of component aging. The largest category is Saltwater System pressure ' boundary components. Passive compressed air components supporting the operation of the Saltwater System and which are included in the Saltwater System boundary constitute the other category. As a result of this scoping,18 component types were determined to require a component aging management review and are identified in Table 5-1.
Functions Within the Scope of License Renewal.
The Saltwater SystemLfuncti_ons to removeLheat;from thej Component: Cooling Water (CCW)
System heat exchangers[thelService Watdr (SRW) System heat exchangers ind the Emergency Core Cooling' System pump room'ai(coolsrs And td transfer that heat to the Ch6sapeake Bay. The Saltwater System consists of two subsystenis per units Eacii subsystem functions independently to provide reliable and redundant heat removal from the system loads. Updated Final Safety Analysis Report Section 9.5 describes the Saltwater System.
In derming the Saltwater System boundaries, several components not normally considered part of the system are included in the system evaluation: Senice Water System, Component Cooling Water System heat exchangers and the Emergency Core Cooling System Pump Room Air Coolers.
The Saltwater System provides one function within the scope oflicense renewal:
a.
Maintain the Saltwater System pressure boundary.
Management of Component Aging Aging effects considered to be plausible for the Saltwater System include: corrosion (e.g., pitting, general, crevice and galvanic); crosion (e.g., cavitation, crosion corrosion and particulate wear);
degradation ofliner materials; wear; fouling; and fatigue. The programs credited for managing the effects of Saltwater System aging will identify and correct the effects of aging so that the intended functions of the structures or components are ensured.
Visual examinations are used to manage aging on multiple items below. They are effective in the identification of aging associated with the Saltwater System. Through-wall failures of pressure boundary components typically manifest themselves as small pits on the inside of the pipe and as pinholes on the outside. The Saltwater System is unlagged and exposed (except the underground piping), which facilitates visual inspections. Leaks are easily detected visually at the system's l
operating pressure. Since these effects of aging can be detected early, catastrophic failure of the Saltwater System is not expected. Actions to specifically address managing the effects of aging on f'
particular component types are described in the following paragraphs.
l 10 CFR Part 54 Sample Results 5-1 Calvert Clifts Nuclear Power Plant
t.
APPENDIX A TECIINICAL INFORMATION Cement Mortar-and Rubber-Lined Pioine and Valves Calvert Cliffs large bore Saltwater System piping is either carbon steel rubber-lined pipe or cast iron concrete mortar-lined pipe. The cement mortar-lined pipe design has provided 20 years of experience at Calvert Cliffs. Since no piping failures due to the exterior environment have been identified, management of component aging is focused upon the internal lining and piping environment.
Brough-wall leakage of the cement mortar-lined piping have been experienced in the above-ground portions of the Saltwater System and have resulted in replacement of the above-ground, cement mortar-lined piping with rubber-lined piping. Rubber lining was selected for use for the above-ground piping primarily due to positive plant experience in repair work on the Saltwater System. Evaluations demonstrated that the failures of the above ground piping were due to failures at field-welded joints with. mortar repairs and also to the extraordinary dynamic hydraulic loading from throttled valjes ivithin the above-ground portionsa The cement mortar-lined, cast iron piping used for bufied sortions bf the Saltwater System has. ndt' experienced these faihires. Inspections of I
the undergfoun.d7pipingshoWed[the lining 'tolberinl good l shape.s ;Thus[no replacement was necessary for the cement inortar-lined piping used for buried. portions of the Saltwater System.
k w & a ba m ms au Wear erosion and chemical attack of the lining, as well as corrosion of the piping and valves (if lining or coating deterioration occurs), have been determined to be the applicable aging mechanisms for these components. Ensuring that the piping linings are not breached prevents corrosive attacks of the interior metal surface oflined Saltwater System piping and valves. He program to inspect large-bore Saltwater System piping lining will detect aging of the lining. In addition, system pressure tests and leakage inspection will continue to be performed to ensure the intended function of the cement mortar-and rubber-lined Saltwater piping and valves is not impaired.
Saran-and Kvnar-Lined Pinina and Valves Saran-and kynar-lined piping and valves have established a good history of oniy limited leakage concerns. However, wear erosion and chemical attack of the lining as well as corrosion of the piping (if lining or coating deteriorates) have been detennined to be the applicable aging mechanisms. Maintaining the lining of these components will prevent corrosion of the internal surfaces of the pressure boundary components. Visual inspection of Saltwater System components will monitor the condition of the piping / valve lining. External inspections are donc during routine walkdowns while intemal inspections can be donc during maintenance periods. In addition, system pressure tests and leakage inspections are performed to detect signs of aging degradation in the piping and valves. The programs outlined above will ensure that the intended function of the saran-and kynar-lined piping arid valves is not impaired.
Copper-Nickel Pinine q
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Copper-Nickel (Cu-Ni) pipmg exists in one small piping run of the Saltwater System. This piping material was selected for use because of its intrinsic saltwater corrosion resistance. Selective teaching and pitting have been determined to be the applicable aging mechanisms for this material.
i System pressure tests and leakage inspections are periodically performed on this piping and will 10 CFR Part 54 Sample Results 5-2 Calvert Cliffs Nuclear Power Plant
APPENDIX A TECIINICAL INFORMATION ensure that the intended function of the Cu-Ni piping portions of the Saltwater System is not impaired.
Unlined Carbon Steel / Cast Iron Valves and Strainers Corrosion and flow-related erosion mechanisms have been determir.ed to be the applicable aging mechanisms for these components. Existing procedures institute ' system pressure tests and leakage inspections on these components. These tests and inspections will detect signs of aging and will ensure that the intended function of these components is not impaired.
Unlined Brass. Bronze. and Nickel-Based Allov Relief and Isolation Valves. and Automatic Vents Corrosion has been determined to be the aging mechanism applicable to brass, bronze, and nickel-
. relief and isolation valves, and automatic vents Copper-and nickel-based alloys have based alloy $d for saltnater applications due to their corrosion resistance iri this s
been select components $e# subject tol leakage l testing and pressure festi.ng4]n< addit /sn, rellef valves are bench-tested penodically[and areMvisijallyl inspected during{that processy These tcsts and inspections hill detect signs of agirig and willinsure ths the intended function)f these components is not impaiicd.0 :
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Comoressed Air System Control and ILand Valves and Accumulators Hand and control valves and accumulators are Compressed Air System components that support the operation of the Saltwater System (see UFSAR Sections 9.5.2.3 and 9.10). Excessive entrained moisture and particulates in the process air could result in corrosion of these components.
Thus, precluding excessive moisture and particulates in the process air is an effective method for managing corrosion of these components. Existing plant procedures ensure that process air from the Compressed Air System is clean and dry, in accordance with industry standards. Therefore, no 1
additional management program is necessary for these components. The program outlined above will ensure that the intended function of these components is not impaired.
I Stainless Steel Control Valves. Hand Valves. Flow Restriction Orifices. and Hermowells
%ese components are subject to a flowing saltwater environment, and the following aging mechanisms apply: corrosion and flow-related erosion. Stainless steel alloys were selected for this application because of their corrosion resistance in the saltwater environment. Inspections performed to assess large-bore Saltwater System piping also address these components. System pressure tests and leakage inspections are also helpful in detecting aging degradation. Inspection of theflow restricting orifice wordd idennfy degradation that may lead to changes in the bore j
diameter. These inspections and tests will ensure that the intended function of these components is not impaired.
Stainless Steel Instmment Isolation Valves. Vent and Drain Valves. and Pressure Transmitters These components are subject to a stagnant saltwater emironment, and corrosion has been determined to be the applicable aging mechanism. Stainless steel alloys were selected for this application because of their corrosion resistance in the saltwater environment. Inspections 10 CFR Part 54 Sample Results 5-3 Calvert Cliffs Nuclear Power Plant
APPENDIX A TECilNICAL INFORMATION performed to assess large-bore Saltwater System piping also addressed these components. System pressure tests and leakage inspections will also be helpful in detecting aging degradation. These inspections and tests will ensure that the intended function of these components is not impaired.
Pumpj!
Corrosion and flow-related erosion have been determined to be the applicable aging mechanisms for the pressure boundary portions of the pumps in the system. A review of maintenance records indicated that pump and pump component failures had previously occurred and required an extensive pump restoration. The restoration included a pump overhaul, casing replacement and coating of pump internal ccoonents. Existing procedures for inspecting pump casings and internals will ensure that the canded function of the Saltwater System pumps is not impaired.
Therefore, no additional management program is necessary for these components.
Heat Excha' eers'!
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Corrosion, ddw-relatederosion, and linin' jeterioration of'thnharinel heads lave been determined g
to be the applicable aging mechanisnis for the heat exchingers. - Existing plant ll rocedures measure i
leakage through the heat exchangers.C Degradation'due to aging that could~ result in a transfer of fluid between the tube and shell sides of the heat exchanger (loss of pressure boundary) would be identified during these measurements. Walkdowns and visual inspections performed on these components could also detect signs of aging. Visual inspections of the channel head liner to detect degradation of the lining are performed. Thus, the actions outlined above ensure that the intended function of the heat exchangers will not be impaired.
Solenoid Valve and Current to Pneumatic Devicgi These components are exposed to compressed air internally. They were reviewed to determine whether aging efTects could adversely affect functions within the scope of license renewal (i.e.,
pressure boundary function). No plausible aging mechanisms were identified for these device types.
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10 CFR Part 54 Sample Results 5-4 Calver, ClitTs Nuclear Power Plant
APPENDIX A TECilNICAL INFORMATION TABLE 5-1 SALTWATER SYSTEM COMPONENTS REQUIRING AGING MANAGEMENT Cement Mortar-Lined Piping Rubber-Lined Piping Saran-Lined Piping Kynar-Lined Piping Check Valve Control Valve Hand Valve Heat Exchanger Current / Pneumatic Device Motor Operated Valve Temperature Indicator Pressure Test Point '
Pressure Switch Pressure Transmitter Reliefyalve,,,,_,,,,
Solenoid yalve, r,m,.
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l UPDATED FINAL SAFETY ANA' LYSIS REPORT
' SUPPLEMENT Baltimore Gas and Electric Company 10 CFR Part 54 Sample Results for Calvert Cliffs Nuclear Power Plant Units 1 and 2
APPENDIX D UFSAR SUPPLEMENT INTRODUCTION 10 CFR 54.21(d) requires that the UFSAR include a summary description of the programs and activities needed to manage the effects of aging and the evaluation of time-limited aging analysis for the period of extended operation. This is done by means of a UFSAR Supplement contained as part of the application.
Only those programs and activities which manage aging for the period of extended operation and which are relied upon in Appendix A, Technical Information, are included here. The format of the Supplement will allow its incorporation into the UFSAR as Section 1.9 once the license is granted. Changes to this Appendix will be controlled under the same plant processes which control the current UFSAR. Changes will be made as amendments to the license application.
The attached UFSAR Supplement discusses only those programs associated with the pilot systems.
Appropriate additional program descriptions will be provided in the final application.
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Calvert Cliffs Nuclear Power Plant
APPENDIX B UFSAR SUPPLEMENT
'1.9 1,1 CENSE RENEWAL 1.9.1 GENERAL For the license renewal process, plant procedures were used to develop a list of plant SSCs within
' the scope of license renewal as indicated in 10 CFR Part 54. From that list, structures and 3
components subject to an aging management review were identified. Existing plant programs and activities were reviewed to ensure that the effects of plausible aging mechanisms identified for these structures and components were sufficiently managed for the period of extended operation.
As a result of this review, the need to modify existing programs or to develop new programs to manage aging effects was identified.
10 CFR Part 54 requires that a license renewal application be accompanied by a UFSAR supplement that includes a summary description of the programs and activities for managing the effects of aging aiid fof the"evaluatioriof tim 6 limited analfses for~the period ofixtended operation.
If, based on chahse[iEth@ldtjlicensind bssisjofplahi"$$dificalisiis$ny structures and components arnsuliis5ist)ttp/deteimineditb lbeJsubjectito %nfaginfhsnahement review, a description of!how;tliefeffectsTofJsging will?be!nianaged mustLbs included in the UFSAR.
Subsequent chingeftojhelaging nianagdmenti frogramsjcanLbelmadelinLabcordance with the l
existing regulations under l_0LCFR 50.59t Section 1.9.2 provides a summary description of the unmodified, modified, and created programs and activities credited for managing the effects of aging for the period oflicense renewal.
1.9.2 PROGRAMS FOR MANAGING TIIE EFFECTS OF AGING 1.9.2.1 Existing Plant Programs and Activities The following programs are presently in place and will manage the effects of aging where credited. Some existing programs may be included in the Section 1.9.2.2 (Modified Plant Programs and Activities) where specific components or groups of components require specific adjustments in order to manage aging effects for the period of extended operation.
1 Comoressed Air System Air Ouality Moisture and particulate levels in the Compressed Air System are monitored to ensure that the air is clean and dry in accordance with industry standards. Maintaining the process air in the Compressed Air System within established guidelines is credited for addressing aging management for Compressed Air System components including components interfacing with other systems.
Heat Exchanger Performance Testing Procedures exist which test the thermal performance of the heat exchangers. Results of these tests can detect transfer of fluid between the tube and shell sides of the heat exchanger indicative of a loss of saltwater pressure boundary.
j 10 CFR Part 54 Sample Results 2
Cahcrt Cliffs Nuclear Power Plant
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APPENDIX B UFSAR SUPPLEMENT Reactor Coolant Water Chemistry Control Procedures exist that establish administrative controls and requirements for system cleanliness. In addition, the Technical Specifications specify the limits within which Reactor Coolant System chemistry must be maintained. Maintaining Reactor Coolant System chemistry within the limits specified in plant procedures and specifications is the basis for excluding certain aging mechanisms from consideration for license renewal.
Refuelina Pool Liner Leakage Monitoring I
Leakage from the refueling pool liner is currently monitored. Monitoring leakage from the refueling pool liner las been credited as the method to manage aging of the refueling canal liner.
System'Walkdowns and Iri5scligna w
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AdminhmtihihNe21ufes rek$irejeri$dN d$ikSwns andllrishecti6nslof plant systems.
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These Wdks$vnk;a$d'insl sctiods iliclEdelvisukciamiristions..;'oflsyjtem components to i
provide earl (identificationiof minor system degradation resulting fronfaging prior to the t
loss of;fuMtionE x1 '
1.9.2.2 Modified Plant Programs and Activities The following modifications to existing programs were determined to be necessary to address specific aging mechanisms or to address aging of specific components.
Fatigue Monitorine Procram l
4 To control the efTects of fatigue on plant equipment, this program monitors specific plant transients and tests and collects data from the plant computer and/or operating logs. De resulting information is used to track the number of events and the effects of the transients and/or tests for fatigue sensitive areas of the plant.
A one-time analysis will be performed, if necessary, to demonstrate that low-cycle fatigue of applicable reactor vessel internals is bounded by the " critical components" monitored by the fatigue monitoring program.
Inservice Insoection (ASME XI) -
This program ensures ASME Class 1,2 and 3 components are inspected as required by 10 CFR 50.55a and ASME Section XI. The type of inspections, the required frequency for the inspections, the components to be inspected and the acceptance criteria are outlined i
in ASME Section XI. The program will be modified to include a formal listing of components to be visually inspected.
10 CFR Part 54 Sample Results 3
Calvert Cliffs Nuclear Power Plant
APPENDIX B UFSAR SUPPLEMENT Protective Coatina and Paintina Visual Insnections Application of coatings at CCNPP follows standard procedures that also address the refurbishment of existing coated surfaces and the need for new coating and recoating.' In addition, readily accessible items are visually inspected. A strategy will be developed for selecting a representative sample of components within the Containment Structure (including portions of the containment liner) that are not readily accessible. A one-time inspection of these areas will be performed and appropriate follow-up actions will be taken. Lubrite plates at the steam generator supports will be inspected ifthe opportunity arises (e.g., steam genu mr replacement) to determine their condition.
Tendon Surveillances Procedures, implemented in accordance with the Technical Specifications, exist for checkingitendonliiftsff forceEandEphysicalTcon~ditions:Theselprocedures also provide accept:inde9ifArinfoWthMinstressil'dYeCandit$ddneysienUcomponent physical conditinskPiisir'si fosses *ar6 timei leiseridhni,[hhd curves 6f[ espe'cted lift-off force 2
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versus!tiine;.(arciesfablished[forlafpeiiod; ofi40Tpear's? EThefexisting curves will be
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I re-evaldathd to sflect the requir$d p~restress lhels fdr the$riod nf license renewal based on test':in'djufveillancidaiki
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m Monitorbe Industry initiatives Some aging mechanisms considered plausible have not yet been observed in operating PWRs anJ, therefore, no industry consensus on appropriate degradation management progranv. has been reached. In these cases (e.g., irradiation assisted stress corrosion cracking,, stress relaxation, and thermal aging), industry initiatives will continue to be f
monitored and evaluated.
'j 1.9.2.3 Plant Programs and Activities Created to Manage Aging Effects In cases where existing programs were not in place or could not be modified to address specific types of aging or aging of specific components, new programs are being developed to address aging management for the period oflicense renewal.
I Groundwater Chemistry Investigations Groundwater observation wells installed during initial plant construction will allow investigation of the ground water chemistry to determine whether groundwater chemical composition is significantly more aggressive that it was found to be during pre-construction testing. One-time sampling will determine if groundwater chemistry has changed significantly and will provide the basis for additional action, if needed.
10 CFR Part 54 Sample Results 4
Calvert Clifts Nuclear Power Plant
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APPENDIX B i
UFSAR SUPPLEMENT Containment Basemat Liner L.cakane Testina If groundwater chemistry investigations suggest that an acid attack of the containment basemat liner is occurring, an evaluation of the basemat liner will % performed in i
conjunction with Appendix J testing.
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APPENDIX C
- TECHNICAL SPECIFICATIONS-Baltimore Gas and Electric Company 10 CFR Part 54 Sample Results for Calvert Cliffs Nuclear Power Plant Units I and 2
1 APPENDIX C TECIINICAL SPECIFICATIONS INTRODUCTION 10 CFR 54.22 requires that the application identify and justify any Technical Specification changes needed to manage the effects of aging during the extended operating period. No Technical Specification changes have been identified which are necessary for the management of aging during the extended operating period.
Also, no additions to the Technical Specifications have been identified at this time.
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10 CFR Part 54 Sample Results 1
Calven Cliffs Nuclear Power Plant 1
I APPENDIX D P
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. ENVIRONMENTAL INFORMATION:
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Calvert Cliffs Nuclear Power Plant Units I and 2
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TECIINICAL INFORMATION TABLE OF CONTENTS p, r sm,y, _. _._..,,,-, s w.
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6.0 ELECTRICAL / INSTRUMENT & CONTROLS SYSTEMS.............................................
6.1 Elect ri c al/I n s t ru men t Pa nel s......................................................................................................
6.2 Cables........................................................................................................................................
7.0 STR U CT U RA L/CI VI L........................................................................................................
7.1 A u xili a ry B uil d i n g......................................................................................................................
7.2 IntakeStructure...........................................................................................................................
7.3 Fuel Oil S t o ra ge Tan k E n cl o s ure.................................................................................................
1 7.4 Con den s at e S t o ra ge Tank E n cl o s u re...........................................................................................
7.5 A u x ili ary Fee d w a t er Pu m p R o o m...............................................................................................
7.6 Co m p o n e n t S u p p o ris...............................................................................................................
7.7 Cranes / Fuel I l an diin g E q uip m en t...............................................................................................
1 8.0 SYSTEMS WITI I A CTI VE FUN CTI ONS...........................................................................
8.1 CirculatingWater........................................................................................................................
8.2 MainTurhine...............................................................................................................................
8.3 CavityCooling.........................................................................................................................
8.4 Cont rol Element D rive Mech anism s..................................... _...........................................
l 8.5 FireDetection...............................................................................................................................
i 8.6 Annunciation................................................................................................................................
8.7 Te c hn ical S u pp o rt Cen t e r Co m p u t e r..........................................................................................
8.8 Communications..........................................................................................................................
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APPENDIX D ENVIRONMENTAL INFORMATION Environmental information supporting the extended operating period will be supplied later.
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