ML20112C024
| ML20112C024 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/20/1996 |
| From: | Doroshuk B, Tilden B, Tucker R BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20112B955 | List: |
| References | |
| NUDOCS 9605240124 | |
| Download: ML20112C024 (100) | |
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- O Calvert Chffs NuclearPower Plant i
- License RenewalProject l l
i l Aging Management Review Report i !
for the l 1
Containment Structure (System 059) lO
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! Revision 3 May, 1996 1
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Prepared bye, Date: #!II j B.M. Tilden ;
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Reviewed by:
R.L. Tucker Date: 8 7!k Approved by: ) Date: 8 10! %
- d. W. Doroshuk i
!O 9605240124 960522 l l
i f- LIFE CYCLE MANAOEMENTUNIT l
C FINAL REPORT CONTAINMENT STRUCTURE AGING MANAGEMENT REVIEW RESULTS TABLE OF CONTENTS l
Section Page Number f
t' TABLE OF CONTENTS l LISTOF A'ITACIIMENTS iii iv LIST OF APPENDICES l
LIST OF TABLES v vi LIST OF EFFECTIVE PAGES l
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1.0 INTRODUCTION
1-1 1.1 Containment Structure Description 1-1 i
1.1.1 0;ntainment Structure LCM Description 1-1 l
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! U l.1.2 Containment Structure LCM Boundary 1-1 1.1.3 Containment Structure Intended Functions 1-2 1.2 Evaluation Methods 1-3 1.3 Containment Stnicture Specific Definitions 1-3 1.4 Containment Structure Specific References 1-4 2.0 STRUCTURAL COMPONENTS WITHIN THE SCOPE OF LICENSE RENEWA-L 2-1 STRUCTURAL COMPONENTS PRE-EVALUATION 3-1 3.0 l
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FINAL REPORT CONTAINMENT STRUCTURE AGING MANAGEMENT REVIEW RESULTS l TABLE OF CONTENTS Section Page Number 4.0 STRUCTURAL COMPONENTS AGING EFFECTS EVALUATION 4-1 4.1 Evaluation 41 4.2 Aging Mechanisms 4-1 4.2.1 Potential Aging Mechanisms 4-1 -
! 4.2.2 Component Grouping 4-2 4.2.3 Plausible Aging Mechanisms 4-2 4.2.4 Aging Management Program Identification 4-3 4.2.5 Aging Management Recommendations 4-3 O
.V l 5.0 PROGRAM EVALUATION - 5-1 5.1 Program AdequacyEvaluation 5-1 <
l 5.2 Structural Components Subject to Adequate Programs 5-1 5.2.1 Existing Programs 5-1 !
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l 5.2.2 Modified Existing Programs 5-2 l
5.2.3 New Programs 5-2 i
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FINAL REPORT CONTAINMENT STRUCTURE AGING MANAGEMENT l
! REVIEW RESULTS LIST OF ATTACHMENTS ,
I l Attachment 1 Potential Aging Mechanisms Applicable to Stmetural Components Attachment 2 Plausible Aging Mechanisms Applicable to Structural Components Attachment 3 Structural Component - Aging Mechanism Matrix Codes Attachment 4 Summary of Aging Management Review Results Attachment 5 Adequate Program Evaluation Attachment 6 Program / Activity Modifications l
l Attachment 7 Walkdown Report - Examination of the Unit 1 Containment Structure l '
l Calvert Cliffs Nuclear Power Plant l
Attachment 8 Attributes in New Program l
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! - LIFE CYCLE MANAGEMENT UNIT FINAL REPORT *
! CONTAINMENT STRUCTURE AGING MANAGEMENT REVIEW RESULTS LIST OF APPENDICES
! Appendix A Freeze-Thaw i
l Appendix B Leaching ofCalcium Hydroxide Appendix C Aggressive Chemicals f
l Appendix D Reaction with Aggregates i Appendix E Corrosion in Embedded Steel /Rebar Appendix F Creep Appendix G Shrinkage Appendix H Abrasion and Cavitation
! AppendixI Cracking of Masonry Block Walls Appendix J Settlement -
Appendix K Corrosion of Steel Appendix L Corrosion of Liner Appendix M Corrosion of Tendons Appendix N Prestress Losses l
Appendix 0 Weathering Appendix R Elevated Temperature Appendix S Irradiation Appendix T Fatigue l
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LIFE CYCLE MANAGEMENT UNIT ID V FINAL REPORT CONTAINMENT STRUCTURE AGING MANAGEMENT REVIEW RESULTS LIST OF TABLES Table Ilde Page Number l
l 1-1 Containment Structure Specific References 1-5 21 Containment Structural Components Within the Scope of License Renewal 2-2 4-1 List of Potential Aging Mechanisms for Containment Structural Components 4-5 4-2 Containment Structural Aging Effects Summary 4-6 I
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FINAL REPORT '
l CONTAINMENT STRUCTURE AGING MANAGEMENT j REVIEW RESULTS I
l LIST OF EFFECTIVE PAGES l
l Revision Pages Summary of Change
! O All Initial revision prepared using LCM-10S, Revision 0.
1 All -Implemented new steps added to LCM-10S to accomplish program evealuation.
2 All Changes made to reflect disposition of Technical Problem Reports written against Revision 0 and to concet transcription errors between the results and the final report sections.
3 All Wording changes to make the language in the final report sections more consistent with the language used in the Integrated Plant Assessment Methodology. Also, technical changes regarding the aging management strategy used to address degradation effects associated with corrosion in structural steel. j I (h O
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1.0 INTRODUCTION
1.1 CONTAINMENT STRUCTURE DESCRIPTION nis section describes the scope and boundaries of the Containment Structure (System 059) l as it was evaluated. Section 1.1.1 provides a brief synopsis of the Containment Structure as l described in existing plant documentation. The Containment Structure boundary is defined l in Section 1 Q to clarify the port'ons of the structure considered in this evaluation.
Section 1.13 is a detailed breakdowa of the Containment Structure intended functions for license renewal and is provided as a basis for component scoping and the identification of component-specific functions.
1.1.1 Containment Structum LCM Description i The Containment Stmeture is a Class I stmeture, housing the reactor and other NSSS components. The Containment Structure consists of a reinforced concrete cylinder and a shallow domed roof which rests on a reinforced concrete foundation
! slab. He concrete cylirder and dome have a post-tensioned contraction design.
Attached to the inside eithe Containment Structure is a carbon steel liner. Here are three personnel and equipment access openings in the Containment Structure: a
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~ two-door personnel lock, a large diameter single door equipment hatch, and a two-l door personnel escape hatch. _
i The Containment Structure has numerous penetrations for piping and electrical connections. These penetrations are leak tight, inerted assemblies, welded to the Containment liner. A fuel transfer tube penetration in the Containment Structure is l provided to permit fuel movement between the refueling pool in the Containment l Structure and the spent fuel pool in the Auxiliary Building.
Two sumps are provided in the Containment Structure floor; a normal sump and an emergency sump.
1.1.2 Containment Structure LCM Boundarv i
The Containment Structure and its structural components provide support and
! shelter to safety related and non-safety related equipment inside the Containment Structure. The LCM boundary addressed by this scoping and evaluation included all in-Containment structural components serving such functions and components comprising the pressure boundary but did not include commodity items such as pipe supports and snubbers. Structural components within this Containment Structure include supports for the following major device types:
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l Accumulators (ACCUMU), boilers (BOILER), compressors (COMP), fans (FAN), l' j' heat exchangers (HX), motors (MO'IOR), motor control centers (MCC), ,
penetrations (PEN), pumps (PUMP), vessels (VESSEL), hydrogen recombiners :
(RECOMB), and screens (SCREEN). ;
Also included in the Containment Structure boundary are structural or functional l supports for non-safety related equipment of the above device types. During an. [
abnormal event such as a seismic event, failure of these non-safety-related equipment supports must not adversely affect the operability of other safety related l components. l l
l- Per the BGE Integrated Plant Assessment Methodology, Containment components l which have unique identifiers in the NUCLEIS Equipment Technical database ,
(such as doors and penetrations) were evaluated using the Aging Management Review procedure for systems. De results of this task are documented and a
- separate AMR Repost entitled " Aging Management Review. Report for the l Containment System." {
l 1.13 Containment Structure Intended Functions
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- A detailed'revidw of the Contain' ment Structuie intended functions was~ completed ,
L during the scoping process described in'the BGE Integrated Plant Assessment Methodology. He following functions for the Containment Structure were identified as structural intended functions on Table IS of " Component Level Scoping Results for the Primary Containment Structure (Sys. 059)."
1.13.1 Function LR-S-1 1 I Provides structural and/or functional support to safety-related equipment.
1.13.2 Function LR-S-2 Provides shelter / protection to safety-related equipment. :
1.133 Function LR-S-3 Serves as a pressure boundary or fission product retention barrier to protect public health and safety in the event of any postulated DBEs.
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LIFE CYCLE MANAGEMENT UNIT 1.1.3.4 Function LR-S-4 5
Serves as a missile barrier (internal or external).
1.13.5 Function LR-S-5 Provides structural and/or functional support to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions. ,
1.13.6 Function LR-S-6 Provides flood protection barrier (internal flood event). i 1.13.7 Function LR-S-7 Provides rated fire barriers to confine or retard a fire from spreading to or from adjacent areas of the plant.
1.2 EVALUATION METHODS l b
- Containment structural components within the' scope ~pf license' ' renewal were evaluated in accordance with BGE procedure EN-1-305,3 Revision 0, " Component Aging Management Review Procedure for Structures." The results of these evaluations are summarized in Sections 3.0 through 5.0.
13 CONTAINMENT STRUCTURE SPECIFIC DEFINITIONS This section provides the definitions for any specific terms unique to the Containment stnictural component level evaluation. i h Definition None N/A 2
Revision 0, I and 2 were done to LCM-10S. EN-1-305 is a new version of LCM-10S which updated procedure format and termnology only.
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LIFE CYCI E MANAGEMENT UNIT I.4 CONTAINMENT STRUCTURE SPECIFIC REFERENCES References utilized in the completion of the Containment structural component level evaluation are listed in Table 1-1. Drawings and procedures used as source documents in the evaluation were i taken at the revision level of record at the start of the task for Revision 1. "Ihe update performed in Revision 2 of this report incorporated several TPRs. The update performed in Revision 3 was performed to address a new strategy for the aging management of corrosion effects of structural steel. Only references affected by the Revision 2 and 3 update have been updated.
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LIFE CYCLE MANAGEMENT UNIT Table 1-1 Containment Structure Specific References Document ID. Document Tide Revision No. Dgg IyJag UFSAR Calvert Cliffs Nuclear Power P' ant Units I and 2 Updated Final 14 1992 Report Safety Analysis Report Technical Calvert ClifB Nuclear Power Plant, Units I and 2, Technical 182 9/27B3 Report Specification Specification 159 9/27/93
- C-W 1evel Scoping Results for Prunary Contamment i 1996 Report Structure (Sys.059)
EPRI RP-2643-27 Class I Structures License Renewal Indust y Report - 12/91 Report NUMARC 90 01 Pressurized Water Reactor Contamment Structures License 1 9/91 Report Rentwel trestry Report
- Exanination of the Unit I Containment Structure - Calvert Cliffs
- 8/92 Report Nuclear Power Plant
- Mather, B.,"How to Make Concrete that will be immune to the - 11/89 Paper effects of freezing and thawing," ACI Fall Convention, San Diego d
- Troxell, O.E., Davis,11E,, and Kelly, J.W., " Composition and 2 Edition 1968 Text Properties off',ncretc," McGraw Hill "StandardSpecificationforConcrete Aggregates," Amerm.an - 1982 Spec ASTM C33-82
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Society ofTesting and Materials ._ _ ,
- Civil and Structural Design Criteria for Calvert Cliffs Nuclear ~ 0 8/2/91 Guide Power Plant Unit No. I and 2, by Bechtel Power Corp.
Specification for Fumishing and Delivery ofConcrete Calvert 8 4/70 Spec 6750-C Cliffs Nuclear Power Plant Unit No. I and,2 ACI 318-63 " Building Code Requirenmnts for Reinforced Concretc," American - 1963 Code Concrete Institute
" Guide to Durable Concretc," American Concrete Institute - 1%7 Standard ACI 201.2R 67
- " Concrete Manual," U.S. Department of the interior 8* Edition 1975 Code Specification for Furnishing and Installation of Piezameter - Calvert 0 11/73 Spec 6750 C-23E Cliffs Nuclear Power Plant Unit No. I and 2
" Potential Reactivity of Aggregates (Chemical Method)," American - 1966 Code ASTM C-28946 Society ofTesting and Materials
" Petrographic Examination of Aggregates for Concretc," American - 1965 Code ASul C-295-65 Society ofTesting and Materials
- Letter from Charles County Sand & Gravel Co. to Bechtel Corp.
- 6/30/72 Letter
- Skoulikidas, T., Tsakopoulos, A., and Moropoulos, T., - - Paper
" Accelerated Rebar Corrosion When Connected to Lightning Conductors and Protection of Rebars with Needles Diodes Using Atmospheric Electricity," in Publication ASTM STP 906,
" Corrosion Effects of Strzy Currents and the Techniques for Evaluating Corrosion of Rebars in Concrete" w
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LIFE CYCI2 MANAGEMENT UNIT (n) v Table 1-1 Containinent Structure Specific References Document ID. Document Title Revkinn No. Dgig Iygg ACI-209R-82 " Prediction of Creep Shrinkage, and Temperature Effects in - 1982 Standard Concrete Structu:es," American Concrete Institute
- " Design and Control ofConcrete Mixtures," Portland Cement 11* Edition 7/68 Guide Association IAEA-1ECDOC-670 " Pilot Studies on Management of Aging of Nuclear Power Plant - 10/92 Report Components," International Atomic Energy Agency MN-3-100 Painting and Other Protective Coatings 2 4/96 Proc TRD-A-1000 Coating Application Performance Standard 8 8/91 Spec 6750-A-24 Specification for Painting and Special Coatings 12 10As2 Spec 6750-C-31 Specification for Furnishing, Detailing, Painting, and Delivering 2 500 Spec Containment and Auxiliary Build'mg Structural Steel ACI 215R-74
- Consideration for Design of Concrete Structures Subjected to - 1986 Standard Fatigue Imading," American Concrete Institute
- " Specification for the Design, Fabrication, and Erection of -
1963 Spec A i Structural Steel for Duildings," American Institute of Steel Construction (b -
Brockengrough, R.L., and Johnson, B G., " Steel Design Manual,*_ -
5n4 Text United States Steel Corporation
- " Specification and Load Data for Wej-Its," Vendor's product - 1977 Catalog catalog by Wej-It Corp., Drownfictd, CO.
6750-C-16 Specification for Fumishing, Fabricating, Delivering and Erection 8 5UI Spec of the Containment Structure Liner Plate and Accessory Steel 6750-C-28 Specification for Stainless Steel Liner Plate and Spent Fuel Pool 5 603 Spec Bulkhead Gate M-665-l/2
- Containment Liner Plate Surveillance," Surveillance Test - - Procedure Procedure (STP)
NUREG-0797 Safety Evaluation Report Related to the Operation of Comanche - 7/81 SER Peak Steam Electric Station, Units I and 2 ANSI /ANS-6.4 ' Guidelines on the Nuclear Analysis and Design cf Concrete -
1985 Code Radiation Shielding for Nuclear Power Plants," American Nuclear Standard
- Iliisdorf, II.R., Kropp, J., and Koch, II.J., "1he Effects ot' Nuclear - 1978 Paper Radiation on the Mechanical Properties of Concrete," Douglas Mclienry Intemational Symposiuia on Concrete and Concrete Structures, American Concrete Institute Publication SP-55 NUREG/CR4652, Naus, DJ.,' Concrete Component Aging and its Significance - 9/86 Paper ORNUTM-10059 Relative to Life Extension of Nuclear Power Plants," Oak Ridge National Laboratory, Oak Ridge, TN (O
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Table 1-1 Containment Structure Specific References DocumentID. Document Title Revision No. Dgg Iygg i
ACI 34945 " Code? , for Nuclear Safety Related Concrete - 1985 Code Structures," Amerzen Concrete Institute
- EQ Design Manual. Calvert Oiffs Nuclear Power Plant 17 1992 Guide M463-1 and Contamment Tendon Surveillance, Calvert Cliffs Nuclear Power 6 10/85 Procedure M463-2 Plant 4 9/85 -
- "Prestressing Report Contamment Structure," Calvert Cliffs - - Report Nuclear Power Plant, Unit 2 ,
ASME SectionIll, " Code for Concrete Reactor Vessels and Contamments," American - 1986 Code l-Division 2 Society of Mechanical Engineers Boiler and Pressure Vessel Code l
60 340-E, " Reactor Cooling System," Calvert Cliffs Nuclear Power Plant 10 2n5 Drawing )
60 341-E,and 7 1In0 60-342-E 6 7n3 60-346-E and " Encapsulation Details Main Steam System," Sheets I and 2, 7 1004 Drawing ,
62-346-E Calvert Cliffs Nuclear Power Plant Unia 1 and 2 4 6/80 j 60-235-E and " Component Cooling Water System," Calvert Cliffs Nuclear Power 37 1/93 Drawing 62-235-E Plant Units 1 and 2 28 1/92 l
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l l 2.0 STRUCTURAL COMPONENTS WITHIN THE SCOPE OF LICENSE RENEWAL The Containment structural components were scoped in accordance with the process described in the BGE Integrated Plant Assessment Mede4 ology. The Containment l Structure was scoped using procedure LCM-1IS for structural wmponents. The purpose l
of component scoping is to identify all structural components trot provide one of the
- structure's intended functions identified in Section 1.13. These structeral components are designated as components within the scope oflicense renewal.
i As a result of the scoping,31 structural component types were identified as providing one ,
of the structure's intended functions listed in Section 1.13. A summary of the scoping result is in Table 2-1.
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I n LIFE CYCLE MANAGEMENT UNIT k) l l Table 2-1 Con e=Inment Structural Comnonents Within the Scone of License Renewal I l
i i STRUCTURAL COMPONENT TYPE INTENDED FUNCTION (S) i Concrete Column LR-S-1 and 5 i Concrete Beams LR-S-1 and 5
( Concrete Slabs and Equipment Pads LR-S-1 and 5 Elevated Floor Slab LR-S-1 and 5 Cast-in-Place Anchors LR-S-1 and 5
! Grout LR-S-1 and 5 Concrete Dome LR-S-1,2,3,4,5, and 7 Concrete Containment Wall LR-S-1,2,3,4,5,6. and 7 l Concrete Basemat LR-S-1,2,3,4,5,6, and 7 l l Primary Shield Wall LR-S-1,2, and 4 Secondary Shield Wall LR-S-1,2, and 4
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l Refueling Pool (Concrete) LR-S-1 and 6 Removable Missile Shield LR-S-4 l
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V Steel Column LR-S-1 and 5 :
Steel Beams LR-S-1 and 5 f 1
Baseplates LR-S-1 and 5 Floor Framing LR-S-1 and 5 l Steel Bracings LR-S-1 and 5 l Platform Hangers LR-S-1 and 5 4 Decking LR-S-1 and 5 Floor Grating LR-S-1 and 5 j Checkered Plates LR-S-1 and 5 Post-Tensioning System LR-S-1,2,3, and 4 Crane Girder LR-S-5 Containment Liner LR-S-3 Basemat Liner LR-S-3 l LR-S-3 j l Refueling Pool (Liner)
LR-S-1 !
l Post-Installed Anchors Lubrite Plates LR-S-1 and 5 l Coating LR-S-1,2,3,4,5, and 7 !
! Panitions and Ceilings LR-S-7 O
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>O 3.0 STRUCTURAL COMPONENTS PRE-EVALUATION Per the BGE Integrated Plant Assessment Methodology, the pre-evaluation task is not conducted on stmetures. Structural components are assumed to be passive and long-lived and therefore subject to an Aging Management Review. Consequently, Table 2-1 also represents a list of structural component types. subject to Aging Management Review.
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4.0 STRUCTURAL COMPONENTS AGING EFFECTS EVALUATION l 4.1 EVALUATION l
l The evaluation of the Containment structural components within the scope of license l renewal was completed in ac,ordance with BGE procedure, " Component Aging l Management Review for S'.ructures," EN-1-305, Revision 0. 'Ihis procedure evaluated all l 31 component types idep'.lfied in Section 2.1. The evaluation accomplished the following:
(1) Identified PO'JNTIAL aging mechanisms for each structural component type.
! (2) Identified FLAUSIBLE component aging mechanisms for each structural j component type or specific components within the component type based on the -
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- environmental conditions
- material of construction
- impact on intended functions p) (3) Developed attributes for programs to manage the effects of aging from those aging l Q- ~
mechanisms identified as PLAUSIBLE.
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_ l (4) Evaluated program adequacy to demonstrate that the effects of aging will be I managed so that the intended function (s) will be maintained for the period of j extended operation. '
These steps are discussed in greater detail in the sections that follow.
4.2 AGING MECHANISMS 4.2.1 Potential Aping Mechanisms l
j This step of the aging evaluation identifies aging mechanisms that are considered i to be POTENTIAL for a given component type. An aging mechanism is considered POTENTIAL for a structural component if the evaluation concludes that the aging mechanism could occur in generic applications of the structural l component type throughout the plant due to susceptible materials of construction l and conducive environmental service conditions.
l l A comprehensive list of 18 aging mechanisms was developed that may be applicable to structural component types. This was based on the EPRI industry reports prepared for the PWR containment structure and Class I structures.
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Other references used to prepare this list include the following: I
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. NRC NPAR Reports
. IAEA Reports l
+ DOE Reports 7 The list of aging mechanisms and materials they affect are shown in Table 4-1.
ne specific description of each is provided in Attachment 1 of procedure EN ,
l 305 or is described in detail in Section 1.0 of the corresponding appendices (A through T) in the aging management review results. ,
Each aging mechanism was evaluated for applicability (i.e., POTENTIAL) to the structural component type based on its material. of construction and the 4 environmental conditions where the component type could be located. His approach ensures all the components within a component type will be evaluated if the potential of degradation exists.
ne results of the stmetural component type POTENTIAL scoping of the component list of aging mechanisms are presented in Table 4-1.
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l 4.2.2 Comr== ant Grounina The grouping of structural components which are within the scope of license j
renewal is primarily based on their materials and their special functions, if any, that contribute to safety, or in the opinion of the evaluator, warrant special l attention. The componerits are grouped into four categories:
(1) Concrete (including reinforcing steel)
(2) Structural steel (3) Architectural items such as doors, roofing materials, and protective coatings (4) Additional components that may have an unique function in the structure 4.2.3 Plannible Aoine Mechanla==
1 The identification of PLAUSIBLE aging mechanisms is accomplished through a careful review of the POTENTIAL aging mechanism list, the development of which is discussed in Section 4.2.1. A potential aging mechanism is considered plausible if when it is allowed to continue without any additional preventative oc
! mitigative measures, the aging mechanism would result in the structural AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE 4-2 REVISION 3
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l j component not being able to perform its intended function. An aging mechanism is also considered plausible if there is insufficient evidence to conclude that future ;
degradation will have no impact on the intended functions of the Containment j l stmetumi components. De plausibility determination is made through a careful L consideration of all the factors required to allow the aging degradation to occur. In l particular, the aging mechanism is scoped for plausibility on the basis of:
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. Materialofconstruction
. Environmental service conditions
. Design and constmetion considerations ;
- Impact on intended functions -
-. Physical conditions of the component i ,
ne results of the aging mechanism plausibility scopmg is an agmg mechanism-component matrix listing the aging mechanism and its disposition. The aging mechanism matrix developed for each structural component type is included in Attachment 3 in the evaluation results. ;
Aging mechanisms determined to be PLAUSIBLE are provided specific aging 3 management recommendations to mitigate the effects of the aging mechanism.
[V Table 4-2 summarizes the results of the plausibility determination and recommendations for the Containment Structure, j 4.2.4 Aging Management Program Identification Once plausible aging mechanisms have been identified, the evaluation is continued to determine whether existing plant programs adequately address the effects of aging for the period of extended operations. If existing programs would not manage the effects of aging during the period of extended operation, a one-time inspection could be condacted, modifications could be made to existing programs, or new programs could be initiated to adequately manage the effects of aging.
This evaluation did not include a determination of whether recommended changes <
to existing programs or new programs would actually be implemented or which programs would be included in the FSAR Supplement.
4.2.5 Aoine Ma==ee===t Recom=r '=F===
The evaluation of all structural component types in the Containment Structure identified a total of 16 aging mechanisms that have the POTENTIAL to degrade these components. A detailed review of the specific component intended functions, material of construction and its basis of design and construction identified PLAUSIBLE component aging mechanisms as shown in the second column of Table 4-2. In some cases, the conclusion that the aging mechanism is
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PLAUSIBLE was made because the condition of the component was not available or could not be readily verified due to lack of accessibility.
Recommended aging management activities include actions to perform condition assessment, to verify conditions conducive to degradation do not exist, and to develop inspection and monitoring programs to ensure degradation can be detected and corrective actions can be taken.
The following is a summary of the recommendations:
Outside Containment (1) Sample the water quality of groundwater using the existing groundwater monitoring wells. If samples fail to confirm that groundwater quality precludes degradation of below grade concrete, take additiothal cc: ective action such as including below grade concrete in an age-related degradation inspection program.
(2) Continue the existing tendon surveillance program to monitor the condition and performance of the tendon system.
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V (3) Update the time-limited aging analysis for allowable prestMss losses of the post-tensioned tendon system to reflect the period of extended operations.
If necessary, retension any tendons to account for predicted prestress losses during the period of extended operations.
Inside Containment (1) Continue to perform visual inspections of coated surfaces of structural steel and the corrosion liner in accessible areas.
(2) Develop an age-related degradation inspection program for coated surface of structural steel and the containment liner that are not readily accessible.
(3) Continue to monitor leakage from the refueling pool to ensure that no corrosion mechanisms have degraded the refueling pool liner weld and heat affected regions.
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Table 4-1 List of Potential Aging Mach ==ta== for Contain=*=t Structural Comnonents Potential to Affect l Aging Mech == lam Descrintion Cont =Inment Structure? Materials Affected Freeze-Thaw Yes Concrete Leaching of Calcium Hydroxide Yes Concrete Aggressive Chemicals Yes Concrete Reaction with Aggregates Yes Concrete Corrosion in Embedded Steel /Rebar Yes Steel, Concrete Creep Yes Concrete Shrinkage Yes Concrete, Partitions and Ceilings
- Abrasion and Cavitation No Concrete
! Cracking of Masonry Block Walls No* Block Walls Settlement Yes Structure ;
s Corrosion in Steel Yes Steel CorrosioTin Liner Yes _ Steel Liners (Carbon i and Stainless) l Corrosion in Tendons Yes Steel Prestressing Losses Yes Steel Weathering Yes Coating, Partitions and Ceilings Elevated Temperature Yes Concrete, Coating Irradiation Yes Concrete, Steel, Coating, Partitions and Ceilings -
Fatigue Yes Concrete
- There are no block walls inside the Containment l
l C
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AGING MANAGEMENT REVIEW RESULTS FINAL REPORT l
CONTAINMENT STRUCTURE 4-5 REVISION 3
O O C LIFE CYCLE MANAGEMENT UNIT Table 4-2 Containment Stnrture Aging Effects Summary STRUCITRAL PLAUSIBLE AGINGMECHANISM COMPONENTS RECOMMENDATION REMARKS j Concrete Columns None None Seejustificationin Appendices DandT.
Concrete Beams None None Seejustification in Appendices D and T. [
t Ground floor slabs a Equipment None None Seejustification in Appendices D and T.
Pads !
Elevated Floor Slabs None Norm Seejustificanon in.b.h D and T.
Cast-in-place Anchors Conosion in steel See r.w.-~L for "Ssect Columns
- Seejustificsion in Appendix K.
l Grout None None None.
f Post-installed Anchors Conosion in steel See r - --- - ? '-- for
- Steel Columns
- Secjustification in Appendix K.
All exposed surfaces of structural steel w..w e are w m/. by a g .
SW Coh h u'ml protectne coatmg. For acassibic areas, significant mating degradation i and/or the presence of conosion will be identified, an issue report wntsen, and conective action taken through the following custmg site programs; MN-3-100. Painting and Other Protectrve Contmgs Program.
Q1 2-100, Issue Repo: ting For those structural steel w..w e not readily accessible, significant coating degradation and/or the presence ofarrosion will be determined utilizing an age related degradation inspedion.
SteelBeams Corrosion in steel See sm-~h for
- Steel Columns
- Seejustification in Appendix K.
Baseptsres Corrosion in steel See isw....~h for " Steel Columns
- Seejustifiction in Appendix K.
Floor Framing Conosionin steel See w....~h for
- Steel Columns
- Seejustificsson in Appendix K.
Steel Bracings Conosion in steel See r - --- -- -- fo'r
- Steel Columns
- Seejustification in Appendix K.
Platform Hangers Corrosion in steel Seec -- -- A-i for
- Steel Columns
- Saajustification in Appendix K.
AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE 4-6 REVISION 3
mY LICE CYCLE MANAGEMENT UNIT Table 4-2 Contninment Structure Aging Effects Summary STPUCTURAL PLAUSIBLE AGING MECILtNISM COMPONENTS RECOMMENDATION REMARKS Decking Conosion in steel See mmm Jation for
- Steel Columns
- Secjustificanon in Appendix K.
Floor Grating Conosion in steel Seer - - ta'-- for' Steel Columns
- Seejustificanon in Appendix K.
Checkered Plate Corrosion in steel See im,----- - ! r-- for
- Steel Cohrswis* Seejustificationin Apperulix K.
Coatings (including galvanizing None None SeeJ. C 4- in AppendicesOandS.
material)
Post-tensioning System Conosionin tendons The tendon surveillance program STP-M663-1/2 should be contmoed Seejustificaten in Appendices M and N.
Prestressinglosses tivoughout the Heense renewal period to monitor the condition and m6.-me of the tendon system.
11Ecurrent prestresslosses which were specifically predicted for a service life of 40 years should be updated to reflect the period of exter&d operations. Ifnecessary, tendons shcuid be retensioned to reflect the predicted prestress losses.
Crane Girder Conesion in steel See m - - tr-- for"SteelColumns* Secjustificationin Appendix K.
Concrete Dome None ,
None Seejustification in Appendices A, B, C, D and E.
Concrete Containment Wall Aggressive chemicals See nom Jat for
- Concrete Basemat* aging mechanisms are plausible only for the below Corrosionin embedded steel /rebar grade portion of the Contamment wall.
Seejustification in Appendices A, B, C, D, E and R.
- **" " Secjustificationin Appendices B,C,D,E J,and S.
Concrete Basemat Aggressive chemicals Conosion in embedded steci/rebar
@ die 3.m. ---, for waer W W M data can k used ta evaluate the impact of chemical anack on the exterior surfaces of exposed om,_.
AGING MANAGEMENT REVIEW RESULTS FINAL REPORT i
CONTAINMENT STRUCTURE l4-7 REVISION 3
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- LIFE CYCLE MANAGEMENT UNIT i
T a N e 4-2 '
Contninment Structure Aoine Effects hmmary a
s STRUCIURAL PLAUSIBLE AGINGMECHANISM RECOMMENDATION REMARKS COMPONEN13 All u,~ad surfaces of the contamment leer are covered by a 3,, gf" .
g g.gg, protective coatmg. For accessible areas, segmficant coating degradation and/or the presena ofcorrosion will be identified, an issue report ,
writsen, and corrective action taken through the fbliowing existing site Programs. :
MN-3-100, Pantag and Other Protective Coategs Prograrn.
QL 2-100, Issue Reportmg For portions of the contemment liner not readily -Ne, significant l coating degradation an&or the presence of cormsion will be determeed j
~
utilizing an age relased 4 ~ --inspection.
Basemat Liner Corrosioninliner See; "" in Appendix L j Pnmary Shield Wall None None Seejustification in Appendices D, R, S, and T. !
None None Seejustification in.",, " - -- D. S, and T.
Secondary Shield Wall i
Contmue to inomsorleakage fiern the refueling pool to ensure that no Seejustificsson in.'m "- _- L ;
Refueling Pool (Liner) Corrosioninliner 2 corrosion mechanisms have degraded the liner welds and heat affecsed regions.
Refueling Pool (Concrete) None None . Seejustification in Appendix D.
Removable Missile Shield None i None SeejumArar==iin Appendix D. ;
Conosion in steel See recommendaten ihr *Seect Columns" Seejuuttile=ana in ?_,, K. i Lubrite Plate None None Seejustificssmain ?_ ' -- G O, and S. _
Partitions and Ceilings t
AGING MANAGEMENT REVIEW RESULTS FINAL REPORT l
CONTAINMENT STRUCTURE ,
'4-8 REVISION 3
. _ . . _ _ _ _ . . _ . . _ . _ _ _ _ . . . _ . _.__________________.___._______m_______.______,r- m . ____ --_ , . - - . , - . . - - - - . _ , - , - . - t w .-
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LIFE CYCLEMANAGEMENT UNIT C
5.0 PROGRAM EVALUATION 5.1 PROGRAM ADEQUACY EVALUATION l 1
Program adequacy evaluations were completed in accordance with EN-1 305, Revision 0, for those programs or aging management alternatives developed to address PLAUSIBLE aging mechanisms. 'Ihe evaluation of programs or aging management alternatives considered the following criteria as a means of establishing the adequacy of specific CCNPP programs:
- 1. Adequate programs must ensure management of the effects of aging for those ,
structural components subject to plausible aging mechanisms. l
\
- 2. Adequate programs must contain acceptance criteria against which the need for corrective action will be evaluated. and ensure that timely corrective action will be taken when these acceptance criteria are not met. j
- 3. Adequate programs rnu:;t be implemented by the facility operating procedures and reviewed by the onsite review committee.
O The results of the program adequacy evaluations are provided in Section 5.2.
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5.2 STRUCTURAL COMPONENTS SUBJECT TO ADEQUATE PROGRAMS 5.2.1 Fristing Procrama The program evaluation task reviewed all existing CCNPP programs that were established to monitor, inspect, and repair Containment Structure structural components that are degraded by identified plausible aging mechanisms.
Components that can be managed by an existing program are as follows:
All structural steel comooncnts in accessible areas. MN-3-100 in combination with QL-2-100 for identifying, documenting, and correcting significant coating degradation are adequate for managing the effects of corrosion.
Accessible nortions of the containment liner. MN-3-100 in combination with QL-2-100 for identifying, documenting, and correcting significant coating degradation are adequate for managing the effects of corrosion.
Refueline cool stainless steel liner. Monitoring refueling pool leakage per established system summary and improvement plan performance indicators is adequate to manage the effects of corrosion mechanisms.
O AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE 5-1 REVISION 3
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( LIFE CYCLEMANAGEMENT LINTT 5.2.2 ' Modified ExistinFProFrams :
His section provides the summary results for those structural components that were determined to have an existing CCNPP Program / Activity that with modification would become an adequate program to manage the effects of aging during the period of extended operations. He evaluation started from evaluating i
structural component types and applicable aging mechanisms and has focused on specific components or locations. Components that can be managed by modifying an existing program or activity are as follows: 1 Prestress tendanc ne tendon surveillance inspection procedure should be revised to include a lift-off force versus time curve for 60 years. Based on additional surveillance testing results and the updated (60-year) lift-otf force versus time curve, retensioning of selected tendons may be necessary to meet the lift-off requirements in the 60-year curve prior to the period of extended operations. !
i 1
5.2.3 New Programs 1 This section provides the summary results for those structural components that were determined to require a new CCNPP Program / Activity to be created as an adequate program to manage the effects of aging during the renewal period.
J Components that can be managed by the creation of such new programs or activities include the following:
Below nrade nortion of Containment wall: An investigative program to test the water quality of the groundwater should be developed to determine if there is any possibility of aggressive chemical attack on Containment wall. Inspection of the exterior, below grade surfaces and additional excavation and testing may be necessary if results from the investigative tests are not favorable.
Bacamat An investigative program to test the water quality of the groundwater should be developed to determine if there is any possibility of aggressive chemical attack on Containment basemat. Inspection of the exterior, below grade surfaces and additional excavation and testing may be necessary if results from the investigative tests are not favorable.
Basemat Liner. An investigative program to test the water quality of the groundwater should be developed to determine whether there is any possibility of corrosion of the basemat liner should any groundwater come into contact with t' e liner.
Non-Accessible Structural Steel: An age related degradation inspection, as defined in the BGE Integrated Plant Assessment Methodology, should be conducted for p structural steel components that are not readily accessible. The ARDI Program V
AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE 5-2 REVISION 3
M CYCLEMANAGEMENT UNIT must provide requirements for identification of a representative sample of components for inspection, the inspection sample size, appropriate inspection techniques, and requirements for reporting of results and corrective actions.
! Non-Acceccible Portions of the Cantninment I .iner An age related degradation l inspection, as defined in the BGE Integrated Plant Assessment Methodology, l should be conducted for portions of the containment !iner that are not readily accessible. 'Ihe ARDI Program must provide requirements for identification of a i representative sample ofcomponents for inspection, the inspection sample size,
- appropriate inspection techniques, and requirements for reporting of results and corrective actions, m
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0a AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE 5-3 REVISION 3
I I
LIFE CYCLE MANAGEMENT List of Attachments and Appendices 1 For the Containment Structure Aging Management Review Total Pares Attachment 1, Potential Aging Mechanisms Applicable to Structural Components 3 l
Attachment 2, Plausible Aging Mechanisms Applicable to Stmetural Components 3 !
l Attachment 3, Structural Components Aging Mechanism Matrix Codes 3 Attachment 4, Aging Management Rctiew Results 3 Attachment 5, Adequate Program Evaluation 17 Attachment 6, Program / Activity (PA) Modifications 2 Attachment 7, Walkdown Report - Examination of Condensate Storage Tank No.12 Enclosure I Attachment 8, Attributes in New Program 6 Appendices I
Appendix A - Freeze-lhaw 6 Appendix B - Leaching of Calcium Hydroxide 7 Appendix C - Aggressive Chemicals 4 Appendix D - Reactions with Aggregates 5 l [3 Appendix E - Corrosion of Embedded Steel /Rebar 5 l
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Appendix F - Creep Appendix G - Shrinkage 4 Appendix H - Abrasion and Cavitation 2 Appendix I - Cracking of Masonry Block Walls 2 Appendix J - Settlement 3 Appendix K - Corrosion of Steel 4 f Appendix L - Corrosion of Liner 6 Appendix M - Corrosion of Tendons 3 Appendix N - Prestress Losses 3 Appendix 0 - Weathering 3 Appendix P - Not Used 0 Appendix Q - Not Used 0 Appendix R - Elevated Temperature 4 Appendix S - Irradiation 4 Appendix T - Fatigue 8 l i
Og AGING MANAGEMENT REVIEW RESULTS FINAL REPORT CONTAINMENT STRUCTURE REVISION 3
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- Attachment 1 1
Potential Aging Mechanisms Applicable to Structural Components !
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O O O l ATTACHMENT 1: POTENTIAL AGING MECHANISMS APPLICABLE TO STRUCTURAL COMPONENTS REVISION: 3 DATE: Mav1996 STRUCTURE NAME: Containment , tructure SYSTEM NUMBER: 059 Sheet 2 of 3 STRUCTURAL POTENTIAL AGING MECHANISMS APPLICABLE TO CONCRETE / ARCH. COMPONENTS REMARKS COMPONENTS A B C D E F G H '
J R S T O 4 - - - - NA - - - 4 - LR Functions LR-S-1, S Concrete Columns - - -
4 - - - -
NA - - - 4 - LR Ftmetions LR-S-1,5 Concrete Beams - - -
4 NA - - - 4 - LR Ftmetions LR-S-1, S Concrete Stabs & Equip. Pads - - - - - - -
4 - - - - NA - - - 4 - LR Functions LR-S-1,5 Elsvtted Floor Slabs - - -
4 4 4 4 4 - - -
NA - - - - - LR Functions LR-S-1 through 5,7 Concrete Dome 4 4 ! 4 4 - - - NA - 4 - - - LR Functions LR-S-1 through 7 Concrete Containment Walt 4 4 4 4 - - -
NA i - 4 - LR Functions LR-S-1 through 7 Concrete Basemat -
4 - - - NA - 4 4 4 - LR Functions LR-S-1,2,4 Primarv Shield Wall - - - -
4 - NA - - 4 4 - LR Functions LR-S-1,2,4 Srcondary Shield Walls - - - - - -
4 - -
NA - - - - - LR Functions LR-S-1,6 Re. fueling Pool (Concrete) - - - - -
Rwmovable Missile Shield - - - 4 - - - -
NA - - - - - LR Function LR-S-4
- - - - - NA - - - - - LR Functions LR-S-1,5 Grout - - -
- - - - NA - - 4 - 4 LR Functions LR-S-1 through 5,7 Coatino - - - -
Parti'- h and Ceilings - - - - - - 4 - NA - - 4 - 4 LR Function LR-S-7 Legend: A Freeze-thaw G Shrinkage M Corrosion in tendons S trradiation B Leaching of calcium hydroxide H Abrasion and cavitation N Prestressing losses T Fatigue C Aggressive chemicals 1 Cracking of masonry block walls O Weathering U (Not Used)
D Reaction with aggregates J Settlement P (Not Used) V (Not Used)
E Corrosion in embedded steet/rebar K Corrosion in steel '
Q (Not Used) NA Not applicable F Creep L Corrosion in Liner R Elevated temperature - Not potential
l l ATTACHMENT 1: POTENTIAL AGING MECHANISMS APPLICABLE TO STRUCTURAL COMPONENTS
, t DATE: May1996 REVISION: 3 STRUCTURE NAME: Containment Structure SYSTEM NUMBER: 059 Sheet 3 of 3 [
STRUCTURAL POTENTIAL AGING MECHANISMS APPLICABLE TO STEEL COMPONENTS REMARKS COMPONENTS K L M N R S T [
4 - - 4 LR Functions LR-S-1,5 r Stesi Columns - - -
4 - - - - 4 LR Func6una LR-S-1,5 I Steal Beams -
4 - - - - - 4 LR Functions LR-S-1,5 Bauplates Floor Framing 4 - - - - - 4 LR Functions LR-S-1. 5 4 4 LR Functions LR-S-1,5 Stael Bracings - - - - -
4 4 L" Fur:ctiorrs m-S-1,5 l Platform Hangers - - - - -
4 - t - 4 LR Functions LR-S-1,5 Dscking - - -
4 LR Functions LR-S-1,5 Floor Grating - - - - - -
4 - - - - LR Functions LR-S-1,5 I Checkered Plates - -
Post-Tensioning System - - 4 4 - - - LR Functions LR-S-1 through 4 Crtne Girder 4 - - - - - 4 LR Function LR-S-5 Containment Liner - 4 - - - - - LR Function LR-S-3 I
Bassment Liner - 3 - - - - - LR Function LR-S-3 Refueling Pool (Liner) - 4 - - - - - LR Function LR-S-3 i Cast-in-place Anchors 4 - - - - - - LR Functions LR-S-1,5 Post-installed Anchors 4 - - - - - - LR Function LR-S-1 4 - - - - LR Function LR-S-1, 5 Lubrite Plates - -
l t.
Legend: A Freeze-thaw G Shrinkage M Corrosion in tendons S trradiation B Leaching of calcium hydroxide H Abrasion and cavitation N Prestressing losses T Fatigue C Aggressive chemicals I Cracking of masonry block walls O Weathering U (Not Used)
D Reaction with aggregates J Settlement P (Not Used) V (Not Used)
E Corrosion in embedded steet/rebar .K Corrosion in steel Q (Not Used) NA Not applicable F Creep L Corrosion in Liner R Elevated temperature - Not potential
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1 Attachment 2 Plausible Aging Mechanisms Applicable to Structural Components l
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fh (3 (3 Q N.Y ATTACHMENT 2: PLAUSIBLE AGING MECHANISMS APPLICABLE YO STRUCTURAL COMPONENTS DATE: May1996 REVISION: 3 SYSTEM NUMBER: 059 Sheet 2_ u! 3 STRUCTURE NAME: Containment structure STRUCTURAL PLAUSIBLE AGING MECHANISMS APPLICABLE TO CONCRETE / ARCH. COMPONENTS REMARKS COMPONENTS (SEE ATTACHMENT 4 FOR JUSTIFICATION)
A B C D E F G H I J R S T O 104 - - - - NA - - - 109 -
Concrete Columns - - -
104 - - - - NA - - -
109 -
Concrete Beams - -
Concrete Stabs & Equip. Pads - - -
104 - - - - NA - - -
109 -
- 104 - - - - NA - - - 109 -
Elevated Floor Slabs - -
Concrete Dome 101 102 103 104 105 - - - NA - - - - -
101 102 PA 104 PB - - - NA - 107 - - -
Concrete Containment Wall 102 PA 104 PB - - -
NA 106 108 - -
Concrete Basemat -
Primary Shield Wall - - - 104 - - - - NA - 107 108 109 -
Secondary Shield Walls - - - 104 - - - - NA - - 108 109 -
Rifueling Pool (Concrete) - - -
104 - - - - NA - - - - -
Removable Missile Shield - - -
104 - - - - NA - - - - -
Grout - - - - - - - - NA - - - - -
Cotting - - - - - - - -
NA - - 108 - 110 Ptrtitions and Ceilings - - - - - -
111 -
NA - -
108 - 110 i
Legend: A Freeze-thaw G Shrinkage M Corrosion in tendons S trradiation B Leaching of calcium hydroxide H Abrasion and cavitation N Prestressing losses T Fatigue C Aggressive chemicals I Cracking of masonry block wa!!s O WeatheMg U (Not Used)
D Reaction with aggregates J Settlement P (Not Used) V (Not Used)
E Corrosion in embedded steet/rebar K Corrosion in steel Q (Not Used) NA ARDM not applicable F Creep L Corrosion in Liner R Etavated temperature - ARDM not potential
T
- ATTACHMENT 2: PLAUStBLE AGING MECHANISMS APPLICABLE TO STRUCTURAL COMPONENTS DATE: Mav1996 REVISION:s _
STRUCTURE NAME: Containment Structurn_ SYSTEM NUMBER: 059 Sheet 3 of 3 STRUCTURAL PLAUSIBLE AGING MECHANISMS APPLICABLE TO STEEL COMPONENTS REMARKS COMPONENTS (SEE ATTACHMENT 4 FOR JUSTIFICATION)
K L M N R S T Stast Columns PD - - - - - 109 Stool Beams PD - - - - -
109 Bisa Plates PD - - - - -
109 Floor Framing PD - - - - - 109 Stest Bracings PD - - - - - 109 Plitform Hangers PD - - - - -
109 Oscking PD - - - - -
109 Floor Grating PD - - - - - -
Checkered Plates PD - - - - - -
Post-Tensioning System - - PF PG - - -
Crsne Girder PD - - - - - 109 Containment Liner - PE - - - - -
Bassment Liner - PE - - - - -
Refueling Pool (Liner) -
PE - - - - -
Cast-in-place Anchors PD - - - - - -
Post-Installed Anchors PD - - - - - -
Lubrite Plates PD - - - - - -
{
Legend:- A Freeze-thaw G Shrinkage M Corrosion in tendons S Irradiation B Leaching of calcium hydroxide H Abrasion and cavitation N Prestressing losses T Fatigue C Aggressive chemicals I Cracking of masonry block walls O Weathering U (Not Used)
D Reaction with aggregates J Settlement P (Not Used) V (Not Used)
E Corrosion in embedded steet/rebar K Corrosion in steel O (Not Used) NA ARDM not applicable F Creep L Corrosion in Liner R Elevated temperature - ARDM not potential
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! Attachment 3 4
f Structural Component - Aging Mechanism Matrix Codes 4
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ATTACHMENT 3 STRUCTURAL COMPONENT- AGING MECHANISM MATRIX CODES :
Revision: 3 Date: May 1996 STRUCTURE NAME: Primary Containment Structure SYSTEM NUMBER: 059 Sheet 2 of 3 CODE JUSTIFICATION REMARKS 101 See Appendix A 102 See Appendix B ,
103 See Appendix C 104 See Appendix D 105 See Appendix E l
106 See Appendix J 107 See Appendix R 108 See Appendix S 109 See Appendix T --
110- See Appendix 0 111 See Appendix G l
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ATTACHMENT 3 STRUCTURAL COMPONENT - AGING MECHANISM MATRIX CODES Revision: 3 Date: May 1996 STRUCTURE NAME: Primary Containment Structure SYSTEM NUMBER: 059 Sheet 3 of 3 CODE JUSTIFICATION REMARKS PA See Appendix C -
l PB See Appendix E !
PC Not Used PD See Appendix K PE See Appendix L See Appendix M ;
PF 1
PG See Appendix N l i O !
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i Attachment 4 s
i Summary of Aging Management Review Results 4
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l Attachment 4
SUMMARY
OF AGING MANAGEMENT REVIEW RESULTS May 1996 Revision: 3 STRUCTURE /SYSTEMNUMBER- 059 STRUCTURENAME: Containment COMPONENTS AFFECTED CONCRETE / ARCH. STEEL PROGRAM / COMMENT AGING MECIIANISMS None None Not Needed Freeze 'Ihaw None None Not Needed Leaching of Ca(OH)2 '
Aggressive Chemicals 1. Below grade portion of None None existing. Need to investigate containment wall' ,
ground water.
- 2. Basemat ,
None None Not Needed Reaction with Aggregates
- 1. Below grade portion of None None existing. Verify water qualityof Corrosion of Embedded containment wall
- ground water.
Steel /Rebar
- 2. Basemat None None Not Needed Creep None None Net Needed Shrinkage Abrasion / Cavitation None None Not Needed.
None None Not Needed.
Cracking of Masonry Block Walls Settlement None None Not Needed I
None All structural steel members MN-3-100,QL-2-100, ARDI.
Corrosion in Steel
. Sheet 2_ ofl
O O O Attachment 4
SUMMARY
OF AGING MANAGEMENT REVIEW RESULTS Revision: 3 May 1996 STRUCTURE /SYSTEMNUMBER: 059 STKUCTURENAME: Containment COMPONENTS AFFECFED AGING MECHANISMS CONCRETE / ARCH. STEEL PROGRAM / COMMENT Corrosion in Liner None Ia In-containment dome and 1. MN-3-100,QI 2-100, ARDI.
wallliner
- 2. Exteriorbasemat 2. None existing. Need to investigate ground water.
3 Sensitized zone of the 3. PEG-19,QI 2-100.
refueling canalliner Corrosion in Tendons None Pishca tendons STP-M-663-1/2 Prestressing Losses None Piestressed tendons STP-M-663-1/2 Weathering None None NotNeeded Elevated Temperature None None NotNeeded Irradiation None None NotNeeded Fatigue None -
None Not Needed Sheet 1oQ
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Attachment 5 i
Adequate Program Evaluations i O .
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l Attachment 5 ADEQUATE PROGRAM EVALUATION REVISION: 3 May 1996 STRUCTURE / SYSTEM NUMBER: 052 STRUCTURE NAME: Containment STRUCTURAL COMPONENT DESCRIPTION: Accaccible nortions ofIn-Containment Wall / Dome LincI AGING MECHANISM DESCRIPTION: Corrosion of liner 4
CCNPP PA or Task ID: MN-3-100. OL-2-100 t
Criteria 1: Adequate programs must ensure mitigation of the effects of age-related degradation for the SSCs identified as within the scope of license renewal.
DISCOVERY DESCRIPTION / BASIS:
4
- 1. Is there a frequency interval in the PA or Task?
^
YES X NO 1
Basis: MN-3-100 requires an inspection of all coated surfaces inside the containment at the
/ beginning of each refueling and maintenance outage to verify the condition of contings.
Deficiencies are documented and reported via 01,2-100 for prioritiration and corrective action.
- 2. Is the frequency interval consistent with industry standards, industry experience, experience unique to Calvert Cliffs, or vendors' rec'ommendations?
YES X NO Basis: The everv-maior-outage insocction interval is a common industry practice and is considered accentable for areas in the containment.
- 3. Will the PA or Task be applicable to all stmetural components under the same component type?
YES X NO Basis: All coated surfaces that are readily accessible are visually inspected during the MN 100 activity.
1 Sheet _2. of 12
1 i
, ( Attachment 5 - Adequate Program Evaluation (continued)
REVISION: 3 May 1996 ARDM DESCRIPTION: Corrosion ofliner CCNPP PA or TASK ID: MN-3-100. OL-2-100 4
Criteria 2: Adequate programs must contain acceptance criteria against which the need for corrective action will be evaluated, and ensure that timely corrective action will be taken when these acceptance criteria are not met.
a ASSESSMENT / ANALYSIS / CORRECTIVE ACTION DESCRIPTION / BASIS: ,
1
- 1. Does the PA or Task have an action or alert value or condition parameter to determine the need for corrective action? I
- YM X NO Basis
- There is no quantitative alert value to determine the r.pd for corrective action. The need is based on the iudgement of the insnector and the safety significance of the structural comnonent needing re-conting. The procedure prioriti7es painting categories which dictate the urgency of corrective actions.
- 2. Does the action value or condition provide sufficient indication of degradation to ensure that 1 4
O' there will not be a functional failure prior to the next PA or Task?
YES X NO Basis: MN-3-100 performs a thorough inspection of accessible portions of the containment liner coated surfaces. Indications of deterioration of contings nre documented.
prioritized and corrected well before corrosion of the liner could impact the intended function of the liner even under design loading conditions.
- 3. Will the action value or condition parameter remain the same during the renewal period?
YES X NO Basis: The corrective actions and condition parameters prescribed in MN-3-100 are based on the surface condition of the coated comnonent. This approach does not need to be revised during the renewal neriod.
Sheet _3_ of _12.
I Attachment 5 - Adequate Fivareiii Evaluation (continued)
REVISION: 3 May 1996 ARDM DESCRIPTION: Corra=8aa of Haar CCNPP PA or TASK ID: MN-3-100. OL-2-100
- 4. Does the PA or task ensure that corrective action is taken?
YES X NO Basis: Procedure MN-3-100 assigns priorities to determine the need for corrective painting.
preventive painting. apnearance painting. or no painting required. To ensure proner appliention and qualified proteeive ennting is used the appropriate conting application oerformance standard is invoked in the procedure.
'i.
Does the PA or Task ensure that the corrective action is appropriately scheduled?
YES X NO Basis: MN-3-100 implements corrective action via O1,2-100 which assigns a due date for corrective action to occur. The completion date is driven by engineering indgement based on the condition of the degraded confina and its contribution to the comnonent's
~ ' '
intended function.
Criteria 3: Adequate programs must be implemented by the facility operating procedures and reviewed by the onsite review conunittee.
CONFI(MATION/ DOCUMENTATION DESCRIPTION / BASIS:
- 1. Does the PA or task have a review / approval process?
YES X NO Basis: The procedure reauires signatures from the appropriate level of supervision (i.e.
POSRC. Manager of Calvert Cliffs Nuclear Power Plant. and GSOA) after it is submitted by the resoonsible engineer.
- 2. Does the PA or task have a change / revision process?
YES X NO Basis: MN-3-100 and OL-2-100 are controlled by the site orocedure for orecaring and revising procedures.
O Sheet _4 of _11.
_~ _ _ _ .__ _. _
q Q Attachment 5 ADEQUATE PROGRAM EVALUATION l
Revision: 3 May 1996 STRUCTURE / SYSTEM NUMBER:31% STRUCTURENAME: Containment l l
STRUCTURAL COMPONENT DESCRIPTION: Prestressed Tendons AGING MECHANISM DESCRIPTION: Prestress losses CCNPP PA or Task ID: STP-M-663-1/2 Criteria 1: Adequate programs must ensure mitigation of the effects of age-related degradation for the SSCs identified as within the scope oflicense renewal.
DISCOVERY DESCRIPTION / BASIS: l
- 1. Is there a frequency interval in the PA or Task? )
YES .2L. N O ___ l l
l Basis: Both the Unit I and Unit 2 nrocedures are imolemented in accordance with the
['
' freauency intervals soecified in olant technical soecification section 4.6.1.6.1. l l
1
- 2. Is the frequency interval consistent with industry standarbs, industry experience, experience unique to Calvert Cliffs, or vendors' recommendations?
YES.X_ NO _
Basis: The freauency interval is in accordance with the requirements in Position C.I.2 in Regulatorv Guide 1.35. This interval was develooed during the containment design based on the credicted degradation rate for the time deoendent tendon force / condition losses.
- 3. Will the PA or Task be applicable to all structural components under the same component type?
YES .X_ NO _
Basis: The orocedure is anolicable to tendons in the Unit 1 system except those which have l been exemoted from insocction as documented in Apoendix SA of the FSAR. The crocedure would be extended to Unit 2 tendons should the Unit 1 results warrant.
O Sheet .i_ of _11
_. .---_..m. ~ _ _ ._ ___ _ _ . . _ . _ _ _ . _ . _ . _ ._
i Attachment 5 - Adequate Program Evaluation (continued) j Revision 3 May 1996 '
2 ARDM DESCRIPTION: Prestress I maan i
CCNPP PA or TASK ID: STP-M-663-1/2 Criteria 2: Adequate programs must contain acceptance criteria against which the need for corrective action will be evaluated, and ensure that timely corrective action will be taken when these acceptance criteria are not met.
ASSESSMENT / ANALYSIS / CORRECTIVE ACTION DESCRIPTION / BASIS:
i 1. Does the PA or Task have an action or alert value or condition parameter to determine the need 4 for corrective action?
i
) YES.X NO __ l l
Basis: Ibe tendon surveillance orocadure provides accantance criteria for niesumss level and tendon system comnonent physical conditions over a 40-year oneratine life. Thaca !
action values include tendon lift-off forces as a measure of nreshess level. strenath
~
l testino of tendon wires for ohvsical condition chemical facting of chanthing filler for eresse nronerties to a==ure continued protection of the tendon wires and extensive visual insnection for broken wires and corrosion levels. The nrocedure and plant Technical Snecification 4.6.1.6.1 also comniv with Position 7 in Rennlatnrv Guide 1.35 to evaluate
~
j D the insnection results.
_ 1 1 2. Does the action value or condition provide sufficient indication of degradation to ensure that j there will not be a functional failure prior to the next PA or Task?
YES.X_ NO ._
Basis: Expected lift-off forces incornorated in the procedure were develoned for the tendon system based on a service life of 40 years. The prestress force data and physical condition data obtained in each surveillance will be evaluated in accordance with guidance in Position 7 of Regulatorv Guide 1.35 such that the integrity of the orestressed tendon system is ensured orior to the next PA. The prestressed tension system is a ,
i passive. and highly redundant system. A recearch ofindustrv data (IFRs. SOERs) renorted verv few incidents of random malfunction of the tendons or its components.
Tine hiah redundancy orovides additional assurance that individual tendon reduced nerformance has a nealigible effect on the functional canability (prestress level) and the system is reliable. Therefore. criteria nronosed in the Regulatorv Guide and adonted by Calvert Cliffs will ensure the tendon system (ar.d subseauently. the containment) will nerform its intended functions at the time of the insnection and as nroiected through the time interval to the next surveillance.
Sheet _fi._ of _.11
Attachment 5 - Adequate Program Evaluation (continued)
- Revision 3 May 1996 l ARDM DESCRIPTION: Prestress Immes CCNPP PA or TASKID: STP-M-663-1/2
- 3. Will the action value or condition parameter remain the same during the renewal period ?
i YES._ NO X (See item I in Attachment 6)
J Basis: Action values and condition narameters will remain the emme during the renewal neriod
- to maintain the containment boundarv nrestress level and accantable phvcical conditions to ensure the intenrity of the tendan svatem comonnants. It should be nntad that i
psiucss losses are a time limited aging analysis and are reflected in the curves of ,
exnected Lift-off Force versus Time in the plant tachnical snecification and the procedure. These curves were develoned during initist plant licanna ===nming a service life of the tendon svetem for 40 years. These curves will need to be re-evaluatad to e=tablish the oredicted prestress levels during the renewal neriod. However. the minimum nrestress level and ohvsical candition reanirements will remmin the mmme.
l
- 4. Does the PA or Task ensure that corrective action is taken? '
i a
i 4 YES .2L. NO _
Basis: The orocedure and the Technical Snecification (4.6.1.6.1) comply with the guidance for
~
insnection and evaluation ofinanection results inRegulatnry Guide 1.35. The steps adonted will ensure tendons. or arouns of tendons'not meeting the predetermined lift-off l forces or ohvsical condition assessment will be further evaluated and annropriate i corrective actions nrescribed and imnlemented. Anv abnormal occurrence will also have to meet renortable reauirements 'to ensure nroner corrective actions and documentation as reauired by evaluation.
[ 5. Does the PA or Task ensure that the corrective action is appropriately scheduled?
YES .2L. NO _
Basis: If a low lift-off force or degraded ohvsical condition is considered evidence of nossible abnormal denradation of the component. FFD must be notified of this condition and
- Shift SunervIsor must enter a 90 day action statement.
a i
Sheet .2._ of._12
, sO
[Q Revision 3 Attachment 5 - Adequate Pmgram Evaluation (continued)
May 1996 ARDM DESCRIPTION: Prestress Losses l
CCNPP PA or TASKID: STP-M-663-1/2 l Criteria 3: Adequate programs must be implemented by the facility operating procedures and l reviewed by the onsite review committee.
CONFIRMATION / DOCUMENTATION DESCRIPTION / BASIS:
l 1. Does the PA or Task have a review / approval process?
YES X_ NO _
Basis: This orocedure has a review /anoroval orocess ner EN-4-104.
l l
- 2. Does the PA or Task have a change / revision process?
YES .X__ NO _
Basis: This orocedure has a chance / revision orocess ner EN-4-104.
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Sheet .B_ of 3
Attachment 5 O]
g ADEQUATE PROGRAM EVALUATION Revision: 3 May 1996 STRUCTURE / SYSTEM NUMBER: 059 STRUCTURE NAME: Containment STRUCTURAL COMPONENT DESCRIPTION: All accessible internal structural steel members i
AGING MECHANISM DESCRIPTION: Corrosion of steel CCNPP PA or Task ID: MN-3-100. OL-2-100 Criteria 1: Adequate programs must ensure mitigation of the effects of age-related degradation for the SSCs identified as within the scope oflicense renewal.
DISCOVERY DESCRIPTION / BASIS:
- 1. Is there a frequency interval in the PA or Task? l YES .X_ NO _. i
/~') Basis: MN-3-100 reauires an insnection of all coated surfaces inside the containment at the
(/ beginning of each refueling and maintenance outage to verify the condition of contings.
j Deficiencies are documented and reported via OI,2-100 for prioritiration and corrective action.
- 2. Is the frequency interval consistent with industry standards, industry experience, experience unique to Calvert Cliffs, or vendors' reco'mmendations?
YES.X NO ._
Basis: The everv-maior-outage insnection interval is a common industry oractice and is considered acceptable for areas in the containment.
- 3. Will the PA or Task be applicable to all structural components under the same component type?
YES .2L NO _.
Basis: All coated surfaces that are readily accessible are visually insnected during the MN l 100 activity. I Sheet .2_ of.12
l t
l Attachment 5 - Adequate Program Evaluation (continued)
Revision 3 May 1996 ARDM DESCRIPTION: Corrosion of Steel i CCNPP PA or TASKID: STP-M-663-1/2
! Criteria 2: Adequate programs must contain acceptance criteria against which the need for
( corrective action will be evaluated, and ensure that timely corrective action will be taken when these acceptance criteria are not met. )
, ASSESSMENT / ANALYSIS / CORRECTIVE ACTION DESCRIPTION / BASIS: l
! i l
l 1. Does the PA or Task have an action or alert value or condition parameter to determine the need i for corrective action?
i YES .X_ NO _. l 1
l Basis: There is no quantitative alert value to determine the need for corrective actinnThe need is based on the iudgement of the inanector and the safety signifiennee of the structursl comnonent needine re-cnntina. The prneadure prioritiram naintine entaoories which dictate the urgency of corrective actions.
, 2. Does the action value or condition provide sufficient indication of degradation to ensure that l there will not be a functional failure prior to the next PA or Task?
- ~
YES .X._ NO__ .
i i
! Basis: MN-3-100 nerforms a thorough insnection of the ennted surfacae of accaccible i
structural steel comnonents inside containment. Indientions of deterioration of continos 7
are documented. nrioritized and' corrected well before corrosion of the liner could ,
imnact the intended function of the liner even under decian londina conditions.
i
- 3. Will the action value or condition parameter remain the same during the renewal period ?
l YES _X_ NO _
l Basis: The corrective actions and condition parameters prescribed in MN-3-100 are based on i
the surface condition of the coated comoonent. This approach does not need to be revised during the renewal neriod.
I 1
i O Sheet _LQ_ of 12
- , ~ ~ - - . - . . . _ _ - . - - - . . -_. . - _.-..-. -.--- . . . - _.
b Revision 3 Attachment 5 - Adequate Program Evaluation (continued)
May 1996 ARDM DESCRIPTION: Carronian of Steel CCNPP PA or TASK ID: STP-M-663-1/2
- 4. Does the PA or Task ensure that corrective action is taken?
YES.X_ NO___
i Basis: Procedure MN-3-100 nacione priorities to determine the naad for corrective nnintino.
preventive paintina. annenrnnce naintino. or no naintine reanired. To enenre nroner annliention and annlified protective enntina is need. the anninnrinte cnntina annlientinn performance standard is invoked in the procedure
- 5. Does the PA or Task ensure that the corrective action is appropriately scheduled?
l Ys X NO l
Basis: MN-3-100 imnlements corrective action via OI,2-100 which ammions a due date for corrective action to occur. The completion date is driven by enoineerina indoement based on the condition of the degraded enntina and its contribution to the enmnonent's intended function.
Criteria 3: Adequate programs must be implemented by the facility operating procedures and
~
reviewed by the onsite review committee. .
CONFIRMATION / DOCUMENTATION DESCRIPTION / BASIS:
- 1. Does the PA or Task have a review / approval process?
YES X_ NO _ ;
I Basis: The procedure reauires sianatures from the aporonrinte level of suoervision (i.L -
POSRC. Manager of Calvert Cliffs Nuclear Power Plant. and GSOA) after it is submitted by the resnonsible engineer.
- 2. Does the PA or Task have a change / revision process?
YES _.X. NO _
Basis: MN-3-100 nnd 01,2-100 are controlled by the site procedure for preparing and revisine nrocedures.
l Sheet _LL of 12 l
Attachment 5 ADEQUATE PROGRAM EVALUATION l Revision: 3 May 1996 l STRUCTURE / SYSTEM NUMBER:_012 STRUCTURE NAME: Containment STRUCTURAL COMPONENT DESCRIPTION: Prestressed Tendons -
l ARDM DESCRIPTION: Corrosion in tendons l l l l CCNPP PA or Task ID: STP-M-663-1/2 I
I Criteria 1: Adequate programs must ensure mitigation of the effects of age-related degradation for the SSCs identified as within the scope oflicense renewal.
l l DISCOVERY DESCRIPTION / BASIS:
i
- 1. Is there a frequency interval in the PA or Task?
l l YES .X_ NO _
f t f ( Basis: Both the Unit I and Unit 2 nrocedures are imnlemented in accordance with the freauency intervals soecified in olant technical soecification section 4.6.1.6.1.
- 2. Is the frequency interval consistent with industry standards, industry experience, experience unique to Calvert Cliffs, or vendors' recommendations?
YES_X_ NO _
Basis: The freauency interval is in accordance with the reauirements in Position C.I.2 in Regulatory Guide 1.35. This interval was develooed during the containment design
- based on the industrv's exoerience with the corrosion resistant oronerties of the material
! and the environment it is exoosed to.
1
- 3. Will the PA or Task be applicable to all structural components under the same component type?
YES .X_ NO __
Basis: The orocedure is anolicable to all tendons in the system excent those which have been exemoted from insoection as documented in Annendix SA of the FSAR.
I e
- O Sheet _12_ of_12
Attachment 5 - Adequate Program Evaluation (continued) J Revision 3 May 1996 ARDM DESCRIPTION: Corrosion of Steel j CCNPP PA orTASK ID: STP-M-663-1/2 i
Criteria 2: Adequate programs must contain acceptance criteria against which the need for corrective action will be evaluated, and ensure that timely corrective action will be taken when these acceptance criteria are not met.
ASSESSMENT / ANALYSIS / CORRECTIVE ACTION DESCRIPTION / BASIS:
I 1. Does the PA or Task have an action or alert value or condition parameter to determine the need l for corrective action?
! l l
YES .2L. NO _
Basis: The surveillance test orocedures containe guidance. criteria and/or refers to anorooriate standards which serve as accentance criteria for the corrosion insnection.
[
l
! 2. Does the action value or condition provide sufficient indication of degradation to ensure that there will not be a fimetional failure prior to the next PA or Task?
YES .X_ NO __
~
Basis: Theirocedures in both units mandate visual inanection of the lendons for indications of corrosion and laboratory testing for sheathing filler to maintain its orotective fimetion. l Historic data from Calvert Cliffs and from the industry reoorted very few incidents of malfunction of the system or its comnonents. which orovides additional accurance that
~
such systems are reliable. Therefore. criteria oronosed in the Regulatorv Guide and ,
imnlemented in the Calvert Cliffs tendon surveillance orocedures will ensure the tendon system (and subseauently. the containment) will oerform its intended functions at the time of the insnection and as oroiected through the time interval to the next surveillance.
^
- 3. Will the action value or condition parameter remain the same during the renewal period 7 l
l YES ._X N O __
- i t
Basis: Historic data from Calvert Cliffs and from the industry reported very few incidents of malfunction of the system or its comoonents. this provides additional assurance that the methodolouv has been successful in maintaining the reliability of the system. Since the
~'
surveillance orocedures and the accentance criteria in the orocedures are to ensure the availability and the reliability of the system. this accentance criteria should not be I
changed during the renewal oeriod.
[
l lO Sheet _13_. of__11 1
l Attachment 5 - Adequate Program Evaluation (continued)
Revision 3 May 1996 !
ARDM DESCRIPTION: Corrosion of Steel CCNPP PA or TASK ID: STP-M-663-1/2
- 4. Does the PA or Task ensure that corrective action is taken?
YES .X_. NO _
Basis: The surveillance test orocedures reauire that any insoection results determined to be ,
unsatisfactory be reoorted to the STE for investigation and to ensure all reporting l reauirements are met.
- 5. Does the PA or Task ensure that the corrective action is appropriately scheduled?
YES .X_ NO _
Basis: All corrective actions must meet reoorting reauirements soecified in Section 4.6.1.6.4 of both Units 1 and 2.
Criteria 3: Adequate programs must be implemented by the facility operating procedures and reviewed by the onsite review committee.
O C/ CONFIRMATION / DOCUMENTATION DESCRIPTION / BASIS:
- 1. Does the PA or Task have a review / approval process?
YES _.X_ NO _ ,
Basis: This orocedure has a review /anoroval orocess per EN-4-104.
- 2. Does the PA or Task have a change / revision process?
YES .X__ NO _
Basis: This orocedure has a change / revision orocess ner EN-4-104.
()
O Sheet _L4_ of _12
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Attachment 5 ADEQUATE PROGRAM EVALUATION Revision: 3 May 1996 l STRUCTURE / SYSTEM NUMBER:_012 STRUCTURE NAME: _ Containment STRUCTURAL COMPONENT DESCRIPTION: Refueling Pool Liner
- ARDM DESCRIPTION
- Corrosion in liners CCNPP PA or Task ID: PEG-19. System Summary and Imorovement Plana (Indicator for Refueling Pool Irakaoe) and OL-2-100 Issue Reoorting l
Criteria 1: Adequate programs must ensure mitigation of the effects of age-related degradation l for the SSCs identified as within the scope oflicense renewal.
l DISCOVERY DESCRIPTION / BASIS:
- 1. Is there a frequency interval in the PA or Task?
YES_X_ NO _
l (d' j Basis: The Refueling Pool SSIP contains an indicator to. monitor nool structural leakaue when the cool is filled. SSIP indicators must be reviewed and undated on a set freauenev.
! 2. Is the frequency interval consistent with industry standards, industry experience, experience
! unique to Calvert Cliffs, or vendors' reco'mmendations?
l YES _X__ NO __
l Basis: The refueling nool is only filled for a short time oeriod each refueling cycle. The
~
leakage is monitored frecuently during this neriod and reoorted as an SSIP indicator on a neriodic basis. This interval is considered to be sufficient to detect and trend any corrosion of the cool liner.
l l 3. Will the PA or Task be applicable to all structural components under the same component type?
l YES .2L. NO _
l Basis: The orocedure is anolicable to the refueling nools in both units.
i C) Sheet _11. of _12
! A l
Attachment 5 - Adequate Program Evaluation (contir.ued)
Revision 3 May 1996 ARDM DESCRIPTION: Corrosion in liners
! CCNPP PA orTASK ID: System Summary Performance Indicator for Refueling Pool Ienknee Criteria 2: Adequate programs must contain acceptance criteria against which the need for l corrective action will be evaluated, and ensure that timely corrective action will be taken when these acceptance criteria are not met.
t ASSESSMENT / ANALYSIS / CORRECTIVE ACTION DESCRIPTION / BASIS:
1
- 1. Does the PA or Task have an action or alert value or condition parameter to determine the need for corrective action?
I YES .2L_ NO _
Basis: The Refueling Pool SSIP reautres assignment of a grade for structural leakage. If Itakage were to increase. the lower grade assigned for the Refueling Pool would reauire an imnrovement olan to determine the cause and correct the degrading trend.
- 2. Does the action value or condition provide sufficient indication of degradation to ensure that there will not be a functional failure prior to the next PA or Task?
YES .2L. .
NO _
i Basis: Th Refueling Pool Liner is a fluid retaining boundarv oniv. It does not orovide any structural supnort function. Structural integrity is maintained by the surrounding concrete. Therefore. leakage detection and trending is an accentable techniaue for ensuring that the refueling pool liner is canable of nerforming its intended function under ,
all design conditions reauired by the CLB. The surrounding concrete has been evaluated l l for the effects of borated water leakage and a determination made that minor leakage l l will not affect the structural integrity ftmetion of the concrete including embedded rebar.
- 3. Will the action value or condition parameter remain the same during the renewal period ?
YES _2L. NO __
l Basis: There will continue to be a SSIP indicator to detect and trend any refueling pool leakage.
- 4. Does the PA or Task ensure that corrective action is taken?
YES .2L_ NO _
Basis: The SSIP nrocedure reauires a system imnrovement nian for any renorted grade below a specified level. System imnrovement plans receive annropriate sunervisory attention on a periodic basis in order to ensure that the degrading trend is corrected.
)
l Sheet _16_ of _.12 l
("
\ Revision 3 Attachment 5 - Adequate Program Evaluation (continued)
May 1996 ARDM DESCRIPTION: Corrosion in liners CCNPP PA or TASK ID: System Summary Performance Indicator for Refueling Pool Lenhage
- 5. Does the PA or Task ensure that the corrective action is appropriately scheduled?
YES _X NO B, asis: Any corrective action required would be imolemented via OL-2-100 which contains a process for orioritizing and scheduling corrective action.
Criteria 3: Adequate programs must be implemented by the facility operating procedures and reviewed by the onsite review contmittee.
CONFIRMATION / DOCUMENTATION DESCRIPTION / BASIS:
- 1. Does the PA or Task have a review / approval process?
YES X._ NO Basis: The SSIP procedure requires approval from the noprooriate level of supervision.
d 2. Does the PA or Task have a' char.gdr6 vision process? .
YES .X NO _
Basis: The SSIP procedure is controlled bv approoriate orocesses to ensure changes do not invalidate the bases for the orocedure.
O V Sint 17 of__ll i
._ _ - - __ _ __ _ _ ~ . _ _ _ - _ _ . _ _. _ . _ . . _
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! Attachment 6 4
i J
! Program / Activity (PA) Modification 1
I i
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l Sheet 1 of 2 l
O O O Attachment 6 PROGRAM / ACTIVITY (PA) MODIFICATION Revision: 3 May 1996 STRUCTURE / SYSTEM NUMBER: 059 STRUCTURE NAME: Containment STRUCTURAL COMPONENT: Prestressed tendons AGING MECHANISM: Prestressed losses CCNPP PA OR TASK ID: STP-M-663-1/2 DOCUMENTI NEW/ REVISED CORRECTIVE SUBTASK PRESENT TASK APPROACH ACTION / RECOMMENDATION
- 1. Force versus Time The existing lift-oft force curves in the 1) Re-evaluate the existing curves to reflect the required 1) curves for predicted procedure and the plant technical prestress levels for the renewal pened. Based on the prestress loss specification were developed for 40 surveillance data and test results of the wires, develop the years operation. 60-year curve that best projects the remaining senice life of the tendons.
Basis:
The predicted tendon lift-off forces during the renewal period are not available from the existing curves.
1 Sheet 2 of 2
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Attachment 7 Walkdown Report - Examination of the Containment Structure l
l The examination of the Containment Structure is documented in a separate report >
entitled ' Examination of the Unit 1 Containment Structure - Calvert Cliffs Nuclear Power Plant." The report contains photographs and is complemented by video tape. :
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Sheet I of 1
!o
O Attachment 8 Attributes in New Program O _
e O ss > > < <1-
i Attachment 8 ATTRIBUTESIN NEW PROGRAM r Revision: 3 May 1996 l
STRUCTURE / SYSTEM NUMBER:._Qji9_
STRUCTURE NAME: Cantainmant STRUCTURAL COMPONENT DESCRIPTION: Below nrade nortion of containment wall and .
cnncrete h==ammt l
AGING MECHANISM: Ayer===ive chemie=le APPLICABLE APPENDIX: A_n nandiv C BACKGROUND: The intandad fi=^tinna of the enntainment wall and the h==amat is to nrovide I
, annoort protactinn and shelter to safety relatad and nnn. safety relatad l eoninment incida the enntainment Chemical at*=ck is nianeihle only if the w= tar i chemistry of aroundwater has bernme sianificantly more aparessive than was !
oriainally anticinataA i RECOMMENDED ATTRIBUTES: Since dearadation of the below nrade nortinn of the containment wall and O h==amat would be n1=ncible if the water chemistry has barnme more neareceive. ~
1 i
k Ihc. proposed pranram will begin with inventicative t==ke followed by corrective ;
action if naca==arv. The reenmmended annrnach is: .
- 1. Rectnre the aroundwater observation wells installed during initial plant L construction for sampling purnoses.
- 2. Secure namnles of the aroundwater for water chemistry facting. If the water chemistrv meets the oriainal clesian reauirements (Cl ions < 500 n.om. SO.,
ions < 1500 nomt no further action is nacaccarv. !
- 3. If the water chemistry tests conclude that the concrete comnonents are l I
beine denradad by chemical anents. the levels of chemical concentration
~
~
will need to be accecced to determine the anoropriate corrective action. 1 I
BASIS: Bacance of the desian and construction of the containment wall and basemat.
and the knowleAae of the water chemistn durina the denian of the niant. it is '
unlikely that chemical attack to concrete is a maior concern.
! l a i i !
a i J i Sheet .2_ of.fi.
I
__ _ - _ __ __, - ________i
- - - -..~ - _
f Attachment 8 i ATTRIBUTES IN NEW PROGRAM Revision: 3 May 1996 i STRUCTURE / SYSTEM NUMBER:li9_ ;
STRUCTURE NAME: Containment ,
L i
STRUCTURAL COMPONENT DESCRIPTION: Fvterior (surface) ha=amat liner i
AGING MECHANISM: Corrosion ofliner i APPLICABLE APPENDIX: Annandiv L BACKGROUND: The intandad fimetian of the ha=amat linar is to nrovide a lanktioht barrier to ,
minimize the rela ==a of radinactivity in the event of a da= ion hacia accidant 'Ihe
~
i only nn==ihility that the arterior hamamat liner can be daaraAad by corrnminn is +
from avnnmure to undararmmd watar which mav lankt hrough cnnstructinn iointa l and cracks of the ennerete kanamat- l RECOMMENDED <
ATTRIBUTES: Since denradatinn of the exterior h==amat liner would be plan =ible if the water chemistry has beenme more acera==ive the nronnuesi program will begin with
} invaatiaative tanke followed by corrective action if nara===rv. The J
recommended anornach is: _
- l. Restore the aroundwatar observation wells installed during initial niant ggnstruction for camnling purnoses. ]
- 2. Secure camnles of the aroundwatar for water chemistry tectina If the water chemistrv meets the orioinal demian reauirements ( nH > 4.0.
Cl ions < S00 nom. SOg ions < 1500 npm ). no further action is _j i
necessary. ;
- 3. If the water chemistry tests conclude that the concrete components are beine denradad by chemical noents the levels of chemical concentration ;
will'need to be ==caccad to determine the enernariate corrective action.
BASIS: Becanca of the design and construction of the exterior 1 acamat liner. and the knowledoe of the watar chemistrv during the design of the olant. it is unlikely that corrosion of the basemat liner is a maior concern.
k Sheet _1_ of.fi.
- _ . _ _ . . ..._m _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ .. _ _ ._. __ . _ _ -
e Attachment 8 A'ITRIBUTES IN NEW PROGRAM Revision: 3 May 1996 1
l STRUCTURE / SYSTEM NUMBER:_Di9_
t STRUCTURE NAME: Cnntainment l STRUCTURAL COMPONENT DESCRIPTION: Below grada nortian of containment wall and cantainment hamamat 1
AGING MECHANISM: Corrosion of FmbaAAad Steel /Rebars APPLICABLE APPENDIX: Apnendir E i
l BACKGROUND: The intended fimetion of the containment wall and the hecammt is to provide l l sunnort. nrotactinn and shelter to mafety relataA and nnn. safety related l eaninment inside the cantninmant Corrnminn of below armA. nonvinn of I
containment wall and cnnerete haemmat is nianeihle only if thev are avnosed to
' ' ~
l an aparettive environment on a enntinual hacie -l t
l RECOMMENDED ATTRIBUTES: Since degradation of the below grade nostion of the containment wall and the j bacemat would have been plancible only if the watar chamistry has become more ~
I_ corrosive. the nronosed program will benin with inve=tiaative tacks followed by ;
corrective action if neceuarv. The racam'mendad anprnach is::
i l 1. Restore the aroundwater observation wells installed durine initial niant
~
~
construction for'campling purnoses.
- 2. Secure namnles of the groundwater for water chemistry testing. If the - ,
mter chemirtrv meets the original desian reauirements ( nH > 4.0.
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,_Cl jops < 500 nnm. SO; ions < 1500 nom i no further action is *
! accessarv. l l
- 3. If the water chemistry tests concluda that the concrete components are beine denraded by chemical neents. the levels of chemical concentration
~ ~
l will need to be assessed to determine the annronriate corrective action. ,
I i
BASIS: Becance of the desian and construction of the containment basemat and below 1
nrade nortions of the containment walls and the knowledge of the water
~
chemistrv durinn the desien of the niant. it is unlikely that corrosion of embedded steel /rebar is a maior concern.
t i.
1 i Sheet A_ of.fi.
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e f Attachment 8 ATTRIBUTES IN NEW PROGRAM Revision: 3 May 1996 ATTRIBUTES IN NEW PROGRAM (continued)
STRUCTURE / SYSTEM NUMBER: 059 STRUCTURENAME: Cantainment l STRUCTURAL COMPONENT DESCRIPTION: Non-accessible structural steel Corrosion of Steel ARDM DESCRIPTION:
I
! Annandiv K APPLICABLE APPENDIX:
BACKGROUND: Safety rel=*ad struc*nral etaal in the Cantninment is covered with an annronriata ~
nrotactive enatino Corrnainn of structurni =*aal can oniv neene if th-=-~
nrotactive enatinen have been daaradad Aoine manmaamant of denr= dad
~
enatina ennditinne on meca==ible'gtructural Aaal in the Containment is seenm'nliched thronoh the enmhinatinn of eviatina niant nronrama. However.
~
I structural steel enmrmnente not randily neca==ihte'reauire mEditional maina management.
RECOMMENDED
, ATTRIBUTES: An noe relatad denradatior. inenaction (ARDn nronram as daceribed in the BGE i
!- Intenr=*ad PInnt Ac=a== ment Mathadolony shouldb'e imnlementad to address -
.)
J corrosion of nnn meca==ible structural steel enmannante which 2nnnort the intended functions of the Containment The ARDI Program must consist of the j followine 1
- 1. Identification of non-accemmihle locations.
- 1. Selection of repracantative structural steel comnonents for insnection.
i
- 2. Development of an in=nection namnie size. ;
- 3. Use of Annronrinte insnection technioues. ,
- 4. Reauiremente for renorting of results and corrective actions if aging l
concerns are identified.
BASIS: The ARDI Pronram will ensure that degradaA conditions due to corrosion of steel are identified and corrected such that non-accessible structural steel comnonents of the Containment will be canahle of nerforminn their intended f functions under all design conditions reauired by the current licensing basis.
k Sheet .1. of Jz.
Attachment 8 i ATTRIBUTESIN NEW PROGRAM i Revision: 3 May 1996 ATTRIBUTES IN NEW PROGRAM (continued)
STRUCTURE / SYSTEM NUMBER: 0$9 i
l STRUCTURENAME: Containment
! STRUCTURAL COMPONENT DESCRIPTION: Non ace,==ible nodinns of the Containment Lact ARDM DESCRIPTION: Corrosion ofI iner l APPLICABLE APPENDIX: Appendix L l BACKGROUND: All ernosed surfae>= of the Cantainment liner are covered with an =r.7.capdate ,
protective enatina. Corrocian of the liner can only occur if tha=> protective enatinos have been degradad Aoine mannaement of denraded enatino conditions on the arraceible nortions of the carbon steel liner in the Containment is acenmnlicheA thronah the combination of eviatina niant nronramm. However.
oortions of the liner which are not randily accaccible reauire additional moina j
'mananement. l l f'\ l l C RECOMMENDED An noe related denradation insne tion (ARDI) nronram as described in the BGE l ATTRIBUTES:
l Intearatad Plant Inceccment !Aethodolony should be imnlemented to address corrosion of non-accessihie nortions of the containment liner. The ARDI -
Program must consist of the following-
- 1. Identification of non-accessible locations.
l
- 1. Selection of reoresentative locations of the liner for insnection.
- 2. Develonment of an insnection namnle size. .
- 3. Use of nonrooriate insoection techniaues.
- 4. Reauirements for renorting of results and corrective actions if aging concerns are identified.
BASIS: The ARDI Pronram will ensure that degraded conditions due to corrosion of the liner are identified and corrected such that non-accessible nortions of the Containment liner will be canable of nerforming their intended functions under all design conditions reauired by the current licensing basis.
O Sheet _6_ of _fi.
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l O aeesaoix ^- easeze-Ta^w I
l 1.0 MECHANISM DESCRIPTION' ,
1 1
Repeated cycles of freezing and thawing can alter both the mechanical properties and physical form of the concrete, thus affecting the structural integrity of the component.
The freeze-thaw phenomenon oct.urs when water freezes within the concrete's pores, creating hydraulic pressure. This pressure either increases the size of the cavity or forces water out of the cavity into surrounding voids.
Freeze-thaw damage is characterized by scaling, cracking, and spalling. Scaling or )
surface flaking occurs in the presence of moisture and is aggravated by the use of deicing j salts. Cracks or spalling occurs when volds are already filled with water, and freezing causes pressure to increase. In extreme cases of freeze-thaw damage, the cover over reinforcing steel is reduced, and the reinforcing steel is eventually exposed to accelerated !
corrosion. Concrete is vulnerable to the expansive effects of the resulting corrosion products, thereby weakening the cencrete's resistance to further attack by aggressive emironments. l l
To minimize the adverse effects of freeze-thaw, three factors must be considered in the design and placement of concrete:2 The cement paste must have an entrained air system with an appropriate void (O/ spacing factor.
The aggregate must be of a sufficiedtly high quality to resist scaling.
The in-place concrete must be allowed to mature sufliciently before exposure to cyclic freezing and thawing.
As shown in Figure A-1, the optimal air content range extends from 3 to 6 percent based on the nominal maximum size of coarse aggregate.'
2.0 EVALUATION -
2.1 Conditions According to Specification ASTM C33-82, " Standard Specification for Concrete ,
i Aggregates,"4 the CCNPP site is located in the geographic region subject to severe weathering conditions. As stated in CCNPP's " Civil and Structural Design Criteria,"5 the j frost penetration depth is 20 to 22 inches.
i S/7/96 EA-1 Revision 3
l 1
l Freeze-Thaw l I
t
.2.2 Potential Aging Mechanism Determination Freeze-thaw is a potential aging mechanism for the following concrete stmetural. ;
- components of containment because they are exposed to outside cold weather
t !
Concrete dome LR functions LR-S-1 through 5, and 7 ;
l Concrete containment wall above the frost line (which is 2 feet below grade) Ut functions LR-S 1 through 7 where:
e LR-S 1: Provides stmetural and/or functional support to safety-related l
equipment, t
LR-S-2: Provides shelter / protection to safety-related equipment.
j LR-S-3: Serves as a pressure boundary or fission product retention barrier to protect public health and safety in the event of any postulated DBEs.
i LR-S-4: Serves as a missile barrier (internal or external).
(
LR S-5: Provides structur'al and/or functional' support' to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
LR-S-6: Provides flood protection barrier (internal flooding event).
LR S-7: Provides rated fire barriers to confine or retard a fire from spreading to or from adjacent areas of the plant.
Other concrete structural components are either below the frost line or are located inside the containment building. Therefore, freeze-thaw is not a potential aging mechanism for all other structural components. ;
2.3 Impact on Intended Functions l
If the effects of freeze-thaw were not considered in the original design or are allowed to degrade the above structural components unmitigated for an extended period of time, this aging mechanism could alTect all the intended functions of components listed in Section l 2.2.
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l 5/7/96 m A-2 Revision 3 l
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Freeze-Thaw i
.2.4 Design and Construction Considerations CCNPP concrete derign specification No. 6750-C-9' specifies:
i 9.3.1 The Portland cement concrete furnished, unless otherwise specified herein, shall conform to ASTAf C-94 Specification for Ready Mix Concrete, ACI 318-63 Building Code Requirementsfor Relitforced Concrete, ACI 301-66 Standard Specifications for Structural Concrete for Building, and ACI Manual of Concute inspection.
10.1.2.2 All aggregate shall conform to ASTAf Designation C33. ...
Section 10.1.16 of ASTM Designation C33-67 specifies that:
1 Procedures for making freezing and thawing tests of concrete are l described in ASTAf Method C290, "Testfor Resistance of Concrete Specimens to Rapid Freezing and nawing in Water," and in ASTM Method C291, " Resistance ofConcrete Specimens to Rapid Freezing in Air and Thawing in Water."
I
' pI t Both ASTM Methods C290 and C291 cover the method for determining the resistance of concrete specimens to rapidly repeated cycles of freezing and thawing in the laboratory.
l l Design specification No. 6750-C-9 for CCNPP also specifies: i 10.4.2.1 The Subcontractor shall specify the air entraining agent he proposes to use: It shall be in accordance with ASTM C-260, l l
capable ofentraining 3-5% air, be completely water soluble, and be l completely dissolved when it enters the batch. De Subcontractor shallgive 30 days advance notice ofthe type ofAEA he proposes to use. l i i I l I ACI 318' and its relevant ACI standards and ASTM specifications provide the physical
- property requirements of aggregate and air entraining admixtures, chemical and physical l requirements of air-entraining cements, and proportioning of concrete including I
containing entrained air to maximize the concrete resistance to freeze-thaw action.
2.5 Plausibility Determination Based on the discussion on Section 2.4, concrete used for the containment dome and containment wall was designed and constmeted in accordance with the requirements specified in ACI-318 and its relevant ACI standards and ASTM specifications. Those requirements satisfy the attributes discussed in Section 1.0 that maximize concrete's resistance to freeze-thaw action. In addition, a walkdown of the Unit I containment conducted during 1992' documented no evidence of damage from freeze-thaw. Therefore, p freeze-thaw is not a plausible aging mechanism for the containment dome and
'd containment wall.
5/7/96 E A-3 Revision 3 l
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. l l p) Freeze-Thaw !
- (w/
I I
24 Existing Programs 1 l There are no existing programs at CCNPP that are designed specifically to identify or to l repair freeze-thaw damage. Since freeze-thaw is not a plausible aging mechanism that l could degrade the containment structural components, no management program is
- necessary.
l i
3.0 CONCLUSION
The CCNPP site is located in the geographic region subject to severe weathering <
conditions. Although freeze-thaw cycles can degrade concrete components that are exposed to cold temperatures and in constant contact with moisture, these components were constructed with concrete designed to maximize its resistance to freeze-thaw cycles.
A walkdown inspection of the Unit I containment structure performed in 1992 found no indication of freeze-thaw effect on the concrete structure. This finding substantiated further the conclusion that freeze-thaw is not a plausible aging mechanism for the structural components of the containment.
4.0 RECOMMENDATION Freeze-titaw is not a plausible aging mechanism for any concrete structural components of the containment. No further evaluation or recommendation is required.
5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991,
- 2. Mather, B., "How to Make Concrete that Will Be Immune to the Effects of .
Freezing and Thawing," ACI Fall Convention, San Diego, November 1989. ;
i
- 3. " Design and Control of Concrete Mixtures," Portland Cement Association, Thirteenth Edition. ,
- 4. " Standard Specification for Concrete Aggregates," American Society of Testing and Materials, ASTM C33-82.
- 5. Civil and Structural Design Criteria for Calvert Cliffs Nuclear Power Plant Unit No. I and 2, by Bechtel Power Corporation, Revision 0, August 2,1991.
l
- 6. " Specification for Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear l
Power Plant Unit No. I and 2," CCNPP's Design Specification No. 6750-C-9, i Revision 8, April 1970, u
l \
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I S/7/96 m A-4 Revision 3 l
l
- .. . . - _ . - . - . - . - - . - - - - . ~ - .- .- . - . . _ . .
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Freeze-Thaw l >
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! 7. " Building Code Requirements for Reinforced Concrete," American Concrete Institute, ACI 318-63.
- 8. " Examination of the Unit 1 Containment Structure - Calvert Cliffs Nuclear Power Plant," August 1992.
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5/7/96 M A-5 Revision 3 l -
- - . . . . . . _ . . . - . .~ .. - -
l Freeze-Thaw I
F-% pement 020 Feense-thew cydes: 300 OJ g - Specimens: 3 s 3 allk-in,
{ eeno ete pelems Cement: Type L 517 su per ceyd OJ S Slump: 2-31n.
{
OJ 4 - k 1 .
O.12 -
!( r{-h monimum-size aggregate OJO -
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0.04 -
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OD2 - -
. . . . i O 6 8 80 42 64 O 2 4 l
- Air content, percent l
Effects of AirContent on Durability, Compassive Stangth, and Requind Water Content of Concrete
( (Soutre: Refennce 3) 5/7/96 Figure A-1 E A-6 Revision 3
A- .-a.m m l
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l
,t APPENDIX B - LEACHING OF CALCIUM HYDROXlDE _
l 1.0 MECHANISM DESCRIPTION 1 Water, either from rain or melting snow, that contains small amounts of calcium ions can readily dissolve calcium corapounds in concrete when it passes thmugh cracks, inadequately prepared construction joints, or areas inadequately consolidated during placing. The most readily soluble calcium compound is calcium hydroxide (lime). The aggressiveness or affinity of water to leach calcium hydroxide depends on its dissolved salt content and its temperature.
- Since leaching occurs when water passes through the concrete, structures that are subject to flowing liquid, ponding, or hydraulic pressure are more susceptible to degradation by leaching than those structures that water merely passes over.
Leaching of calcium hydroxide is visible on concrete surfaces that have dried.
The leachate is almost colorless until carbon dioxide is absorbed and the material l dries as a white deposit. The white deposit is a product of water, free lime from the concrete, and carbon dioxide that has been absorbed from the air.
! When calcium hydroxide is leached away, other cementitious constituents become exposed to chemical decomposition, eventually leaving behind silica and alumina gels with little or no strength.2 Leaching over a long period of time increases the porosity and permeability of concrete, making it more susceptible to p) other forms of aggressive attack and reducing the strength of concrete. Leaching l
- ( also lowers the pH of concrete and threatens the integrity of the exterior
~
protective oxide fin of rebar. -
i Resistance to leaching and efflorescence can be enhanced by using concrete with low permeability. A dense concrete with a suitable cement content that has been
! well cured is less susceptible to calcium hydroxide loss from percolating water because of its low permeability and low absorption rate. The design attributes to enhance water-tightness include low water-to-cement ratio, smaller coarse aggregate, long curing periods, entrained air, and thorough consolide.'.ion?
Figures B-1 and B-2 show the impact on permeability due to wat.':-to-cement ratio, aggregate size, and curing time.
l 2.0 EVALUATION 2.1 Conditions The contahunent wall and the dome are exposed to the outside environment and are expected to have rainwater passing over the exterior surface. The I containment dome is provided with a roof drainage system to prevent ponding.
The containment basemat is in contact with underground water. A permanent dewatering system was installed during construction to maintain a stable groundwater table at El.10'-0", which is the elevation at the top of the basemat. l f'~% l S/7/96 m B-1 Revision 3
i 1
Leaching of Calcium Hydroxide i
i 2.2 Potential Aging Mechanism Deteruni== tion Leaching of calcium hydroxide is a potential aging mechamsm for the following structural components of contamment because they coidd be exposed to flowing liquid, ponding, or hydraulic pressure:
Concrete dome LR functions LR-S-1 through 5, and / J Concrete containment LR functions LR-S-1 through 7 wall Concrete basemat LR functions LR-S-1 through 7 where:
LR-S-1: Provides structural and/or functional support to safety-related equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
LR-S-3: Serves as a pressure boundary or fission product retention barrier to protect public health and safety in the event of any postulated DBEs._
LR-S-4: Serves as a missile barrieh (internal or external).
LR-S-5: Provides structural and/or functional support to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
l LR-S-6: Provides flood protection barrier (internal flooding event).
! LR-S-7: Provides rated fire barriers to confine or retard a fire from -
spreading to or from adjacent areas of the plant.
Leaching of calcium hydroxide is not a potential aging mechanism for other structural components of the containment because they are inside the containment building.
2.3 Impact on Intended Functions if the effects ofleaching of calcium hydroxide were not considered in the original design or are allowed to degrade the above structural components unmitigated for an extended period of time, this aging mechanism could affect all the intended functions of components listed in Section 2.2.
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S!7/96 m B-2 Revision 3
Leaching of Calcium Hydroxide (Oj l
2.4 Design and Constniction Considerations Leaching attack can be nunimized by providing a low-perineability concrete mix design during construction. CCNPP concrete design specification No. 6750-C-94 specifies:
9.3.1 The Portland cement concrete furnished, unless otherwise speciped herein, shall conform to ASTM C-94 Spectpcation for Ready Mix Concrete, ACI 318-63 Building Code Requirements for Reinforced Concrete, ACI 301-66 Standard Specifcations for Structural Concrete for Building, ar.d AC: Manual of Concrete i inspection.
12.1 Conaete Quality 12.1.1.1 Portland cement shall conform to ASTM Designation C-94-67, Alternate No.1 and ACI 301-66.
l
! 12.1.2.1 Concrete shall meet thefollowing requirements:
Nominal
% 28-Day Simnp at Simnp Mashman 1 Class Strength Point of Totenmce Aggregate lise and Loation (psi) Placement (in.} 75tze (in.)
C-2 4,000 2 1% 1-% in. Containment Base Slab and Other Structural Concrete C Grout 4,000 - - M4 Containment Joints D-1 S,000 3 1% N4 in. Walls and Slabsless than 12' thick and Congested Rebar i D-2 S,000 2 1% 1-% in. Containment Walls and Dome and
Other Structural Concrete .
l D Grout 5,000 - - M4 Construction Joints Dry Pack 4,000 0 - M4 As Directed Tremie 4,000 6 -
N4 in. As Directed l Concrete 12.1.S Mix Design i
s Sm96 m B-3 Revision 3
Leaching of Calcium Hydroxide 12.1.S.1 The Constructor shall retain an -rraam: Testing Laboratory, at his own cost, to design and test initial concrete mixes.
The initial mixes shall be designed in accordance with ACI ;
Standards 613 and 301 to produce a required strength of15 percent i over specifed strength for reinforced concrete at 28 days and 25 percent over specifed strengthfor post-tensioned concrete at 28 days for each class of concrete with slump and sneximum sizes of ,
aggregate as specifed in the Classifantion Tabh (Section 12.1.2).-
12.1.5.2 The Constructor shallfurnish the Subcontractor with mix designs one month prior to the manufacture of concrete. Furnishing mix designs shall not relieve the Subcontractor of his responsibility for compliance with the provisions of the Specifcation. Where necessary, the Constructor shallincrease or decrease cementfactors as deemed necessaryfor design mixes using statistical methods descnimi
! in the ACI 214-65for the particular class of concrete. An increesc in the water-cement ratio of a mix design or a decrease in its cement
- quantity shall constitute a new mix design and the provisions of
! Section 12.1.S.1 of this Specifcation shall apply. Calcium chloride l_ shall not be used.
T l j 2.5 Plausibility Determination l
Based on the discussion in Section 2.1, the contamment dome and contamment wall are exposed to water passing over the surface. The containment basemat is located below the designed underground water table and may be subjected to some hydraulic pressure. ,However, as discussed in Section 2.4, concrete used far .
the containment was designed in accordance with ACI 3185 and its relevant ACI l
standards and ASTM specifications to maxunize resistance to leaching of ce',cium l hydmxide. A walkdown6 in 1992 observed only slight traces of leacb%g on the i contamment dome and wall and were judged to have no adverse impact on the -
l integrity of these components. Therefore, leaching of calcium hydroxide is not a l
plausible aging mechanism for the containment dome, the containment wall, and .
l the concrete basemat.
1 l
2.6 Existing Progmms l There are no existing programs at CCNPP that ere designed specifically to l identify or to repair damage to concrete due to leacidng of calcium hydmxide. ;
! Since teaching of calcium hydroxide is not a plausible aging mechanism that !
could degrade the containment structural components, no management pmgram is necessary.
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Sf7/96 E B-4 Revision 3
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Leaching of Calcium Hydroxide !
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3.0 CONCLUSION
i !
The containment dome and containment wall surfams in CCNPP are exposed to -l water. No ponding or hydraulic pressum will form to leach the calcium i
- hydroxide. Although the containment basemat could be subjected to hydraulic
) pressure due to undergmund water, the concrete mix was designed for low ,
permeability and high compressive strength which pmvide the best protection 4 against leaching. i 1 5 6
This conclusion is supported by a 1992 walkdown inspection during which only >
minor traces of leaching marks were detected in various amas of the containment
> dome and wall. These indications were judged to have ne impact on containment integrity. Themfore, leaching of calcium hydroxide is not a plausible aging mechanism for any concmte structural components of the containment.
4.0 RECOMMENDATION :
Leaching of calcium hydroxide is not a plausible aging mechankm for any [
concrete structural components of the containment. No further evaluation or j recommendation is required.
4 3
i
- - . . . _ i
5.0 REFERENCES
- l
- 1. " Class 1 Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991. ,
- 2. Troxell, G. E., Davis, H. E., and Kelly, J. W., Composition and Properties of Concrete, Second Edition, McGraw Hill,1968. i
- 3. " Guide to Durable Concrete," American Concrete Institute, j ACI-201.2R-67. .
- 4. " Specification for Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear Power Plant Unit No.1 and 2," CCNPP's Design Specification No. 6750-C-9, Revision 8, April 1970.
- 5. " Building Code Requirements for Reinforced Concrete," American Concrete Institute, ACI 318-63.
- 6. " Examination of the Unit 1 Containment Structure - Calvert Cliffs Nuclear Power Plant," August 1992.
j 7. Concrete Manual, Eighth Edition, U.S. Department of the Interior,1975.
7 i
l 5/7N6 m B-5 Revision 3
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Leaching Of Calcium Hydroxide i
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0.5 0.6 0.7 0,8 0.9 WATER - CEMENT RATIO BY WEIGHT 1
4 1
I Relationship Between Coefficient of i
Penneability and Watecto-Cement Ratio j (Soutre: Reference 7) 1 27/96 Figure B-1a B-6 Revision 3
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Leaching of Calcium Hydroxide I
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- . - - - - - - _ - - _ - ~ - - - - _ _ . . - - - - - - - - -
A (vi APPENDIX C - AGGRESSIVE CHEMICALS l
l 1.0 MECHANISM DESCRIPTION' 1
Concrete, being highly alkaline (pli > 12.5), is vulnerable to degradation by strong acids.
Acid attack can increase porosity and permeability of concrete, reduce its alkaline nature at the surface of the attack, reduce strength, and render the concrete subject to further l deterioration. Portland cement concrete is not acid-resistant, although varying degrees of resistance can be achieved depending on the materials used and the attention to placing, consolidating, and curing. No Portland cement concrete, regardless of its composition, will i withstand exposure to highly acidic fluids for long periods.
Below grade, sulfate solutions of sodium, potassium, and magnesium sometimes found in groundwater may attack concrete, often in combination with chlorides. He exposed surfaces of structures located near industrial plants are vulnerable to industnal pollution from the l sulfur-based acid rain and are subject to deterioration. Sulfate attack produces significant expansive stresses within the concrete, leading to cracking, spalling, and strength loss. Once i
established, these conditions allow further exposure to aggressive chemicals. Groundwater chemicals can also damage foundation concrete. A dense concrete with low permeability may provide an acceptable degree of protection against mild acid attack. Any factors that l tend to improve the compressive strength of the concrete will have a beneficial effect on low l
permeability. Herefore, the better the quality of the constituent material, the less permeable l _
the concrete. Low water-to-cement ratio, smaller aggregate, long curing period, entrained air, l and thorough consolidation all contribute to watestightness.
~
l Concrete thus constructed has a low permeability and effective protection against sulfate and chloride attack. Minimum degradation threshold limits for concrete have been established at
, 500 ppm chloride or 1,500 ppm sulfates. He use of an appropriate cement type (e.g., ASTM l C150, Type II) and pozzolan (e.g., fly ash) also increases sulfate resistance.
l 2.0 EVALUATION l
2.1 Conditions l
l The only significant inventory of aggressive chemicals stored inside the containment is l borated water, and it is primarily in safety-related systems such as the primary coolant system, l safety injection system, and chemical volume control system. Because of the safety significance of these systems, undetected leakages of borated water for an extended period of time cannot occur. Herefore, the containment's interior surface and all internal stmetural ,
! components are not exposed to the risk of aggressive chemicals.
There is no heavy industry near the CCNPP site that could release aggressive chemicals to the
! atmosphere. Ilowever, the containment dome and the above-grade portion of the containment wall are exposed to an environment containing chloride ions due to the containmenfs proximity to the Chesapeake Bay.
He outside, below-grade surface of the containment is exposed to scil and groundwater. De a potential for degradation by aggressive chemicals depends on the quality of concrete, the chemical composition of soil and groundwater, and the level of groundwater in relation to
' below-grade portions of the structure.
5/7/96 m C-1 Revision 3
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y Aggressive Chemicals i
2.2 Potential Aging Mechanism Determination l l
Attack by aggressive chemicals is a potential aging mechanism for the following concrete structural components of containment because they are exposed to outside environment:
Concrete dome LR functions LR-S-1 through 5, and 7 I
i Concrete containment wall LR functions LR-S-1 through 7 l l l
Concrete basemat LR functions LR-S-1 through 7 ;
l where:
LR-S-1: Provides stmetural and/or functional support to safety-related equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
LR-S-3: Serves as a pressure boundary or fission product retention barrier to .
protect public health and safety in the event of any postulated DBEs. {
?fM
() LR-S-4: Serves as a missile barrier (intemal or extemal).
LR-S-5: Provides structural and/or. functional support to no safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
l LR-S-6: Provides flood protection barrier (intemal flooding event).
i l LR-S-7: Provides rated fire barriers to confine or retard a fire from spreading to or
[ from adjacent areas of the plant.
! Other concrete structural components are located inside the containment building; therefore, . j l attack by aggressive chemicals is not a potential aging mechanism.
2.3 Impact on Intended Functions If the effects of attack by aggressive chemicals were not considered in the original design or are allowed to degrade the above structural components unmitigated for an extended period of time, this aging mechanism could affect all the intended functions of components listed in Section 2.2.
2.4 Design and Construction Considerations The containment was constructed with concrete that complies with CCNPPs design 2
specification No. 6750-C-9 to assure low permeability. Another design consideration was the use of prestressed tendons to minimize crack development in the concrete. These b properties provide the best protection against chemical attacks.
5/7/96 a C-2 Revision 3
1 Aggressive Chemicals 2.5 Plausibility Determination l
l Based on the discussion in Sections 2.1 and 2.4, attack by aggressive chemicals is not a plausible aging mechanism for the containment dome and the containment wall above grade.
Because chemical contents of groundwater are not known, attack by aggressive chemicals to the below-grade portion of the concrete containment wall and the concrete basemat is a plausible aging mechanism.
2.6 Existing Programs here are no existing programs at CCNPP that are designed specifically to identify or to mpair damage to concrete due to aggressive chemicals. Since attack by aggressive chemicals is not a plausible aging mechanism for all concrete components inside containment, the containment dome, and the containment wall above grade, no management program is needed l for these components.
3.0 CONCLUSION
N Attack by aggressive chemicals is not plausible for the containment dome- and the
\ containment wall above grade because concrete with low permeability and prestressed tendons, which minimize concrete cracking, wem used in construction of the containment dome and wall. Additionally, there is no heavy industry near the CCNPP site to release aggressive chemicals. Attack by aggressive chemicals is also not plausible for concrete components inside the containment because excessive leakages of borated water inside the l
containment cannot occur. .
The below-grade portion of the containment wall and the concrete basemat are exposed to I groundwater. Because the quality of groundwater is not known, degradation due to l aggressive chemicals is plausible.
4.0 RECOMMENDATION -
l j During initial plant construction, groundwater observation wells were installed to monitor the fluctuation of the groundwater table, and samples were taken for groundwater quality testing.'
Although the wells are still in place, the monitoring activities have been discontinued. It is recommended that the groundwater water quality be tested using these wells. This data can be used to evaluate the effects of chemical attacks on the exterior surface of the containment structure below grade.
5.0 REFERENCES
! 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27,
- December 1991.
J 3
l S/7/96 m C-3 Revision 3 I
- , - , ~ . . . - -.
4 i
i ) Aggressive Chemicals i
I.
i 2. " Specification for Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear 2
Power Plant Unit No. I and 2," CCNPP's Design Specification No. 6750-C-9, i Revision 8, April 1970.
- 3. " Specification for Furnishing and Installation of Piezometers - Calvert Cliffs Nuclear Power Plant Unit No. I and 2," CCNPPs Design Specification No. 6750- l j C-23E, Revision 0, November 1973. l A
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q C, APPENDlX D - REACTIONS WITH AGGREGATES 1.0 MECHANISM DESCRIPTION' Certain mineral constituents of all aggregates react with chemical compounds that compose the Portland cement, most notably alkalis. Alkalis may also be introduced from admixtures, salt-contaminated aggregates, and penenden by seawster or solutions of deicing salt.
However, it is only when the expansive reaction products become extensive and cause cracking of concrete that aggregate reactivity is considered a deleterious reaction.
Three principal deleterious reactions between aggregates and alkalis have been identified as alkali-aggregate, cement-aggregate, and expansive alkali-carbonate reactions.
Alkali-aggregate reaction, more properly designated as alkali-silica reaction, involves aggregates that contain silica and alkaline solutions. All silica minerals have the potential to react with alkaline solution, but the degree of reaction and ultimate damage incurred can vary significantly. Alkali-silica traction can cause expansion and severe cracking of concrete structures. Reactive materials in the presence of potassium, sodium, and calcium oxides derived from the cement react to form solids, which car. expand upon exposure to water. :
Cement-aggregate reaction occurs when the alkalis in cement and some siliceous constituents of the aggregates react. His reaction is complicated by environmental conditions that produce O
()
high concrete shrinkage and alkali concentrations on the surface due to drying. Sand-gravel aggregates from some river systems in the Midwestem United States have been involved in deteriorated concrete attributable to this reaction..
Expansive alkali-carbonate reaction occurs between certain carbonate aggregates and alkalis, and produces expansion and cracking. Certain limestone aggregates, usually dolomitic, have been reported as reactive.
Aggregates that react with alkalis can ca .tse expansion of varying severity, even to the extent of producing cracking of the concrete and resulting loss of strength and durability if the expansion is severe. He cracking is irrr gular and has been referred to as map cracking.
Moisture must be available for chemical reactions between aggregates and alkalis to occur. .
Consequently, areas that are either consistently wet or attemately wet and dry are susceptible to deterioration given the presence cf potentially reactive aggregates, ne deleterious effects of reactive aggregates are best avoided by using aggregates from sources that have a proven record of service. If such records are unavailable, aggregates should be examined petrograpnically to identify potentially reactive constituents. Chemical reactions of aggregates for both fast and slow reaction rates were recognized as early as 1940.
He method to identify the reactive constituents in concrete aggregates was first published in ASTM C-289, " Potential Reactivity of Aggregates (Chemical Method)"2 and ASTM C-295,
" Petrographic Examina6on of Aggregates for Concrete"' in 1952 and 1954, respectively.
Both standards provide guidance for selecting aggregates and cements to avoid alkali-aggregate reactions.
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( Reactions with Aggregates 2.0 EVALUATION 2.1 Conditions The aggregates used in the concrete of the CCNPP contamment came from sites in Charles County, Maryland', which is not in the geographic regions known to yield aggregates suspected of or known to cause aggregate reaction.
2.2 Potential Aging Mechanism Determination Reaction with aggregates is a potential aging mechanism for the following concrete structural l components ifreactive aggregates were used in the concrete structure construction:
Concrete columns LR functions LR-S-1,5 Concrete beams LR functions LR-S-1,5 Ground slab and equipment pads LR functions LR-S-1,5 Elevated floor slabs LR functions LR-S-1,5 L _.
Concrete dome L.R functions LR-S-1 through 5, and 7 Concrete containment wall LR functions LR-S-1 through 7 Concrete basemat. LR functions LR-S-1 through 7 Primary shield walls LR functions LR-S-1,2,4 Secondary shield walls LR functions LR-S-1,2,4 Refuelingpool LR functions LR-S-1,6 -
Removable missile shield LR function LR-S-4 where: l LR-S-1: Provides structural and/or functional support to safety-related equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
LF S3 Serves as a pressure boundary or fission product retention barrier to pro. public health and safety in the event of any postulated DBEs.
LR-S-4: Serves as a missile barrier (internal or extemal).
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Reactions with Aggregates i
LR-S-5: Provides structural and/or functonal support to non-safety-related l eW whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
LR-S-6: Provides flood protection barner (internal Goodmg event).
LR-S-7: Provides rated fire barriers to confine or retani a fire from spreadmg to or from adjacent areas ofthe plant.'
2.3 Impact on Intended Fumetions If the effects of reaction with aggregates were not considered in the onginal design or are allowed to degrade the above structural ---i==:= unnutigated for an extended period of l l' time, this aging # "== could affect all the intended functions of wuqw_ s listed in l Section 2.2.
! 2.4 Design and Constreetion Considerations t
All aggregates used in construction of the CCNPP contamment structure were investigated, l tested, and examined based on the following specifications. t 8
CCNPP'sdesignspecificationNo.6750-C-9 specifies: !
~
- 10.1.1.1 Cement shall be Portland cement,1ppe 11 conforming *to ASTM Designation C-150, . . . The cement shall not contain more than
! 0.60 percent by weight ofalkalies calculated as Na;0plus 0.658 K2 0.
Only one brand ofcement shall be usedfor all work.
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' 15.2.3.1 The Bilder, at his expense, shall retain an approved ,
independent testing laboratory to sample and test aggregates and the aggregate source in accordance with methods as specified in :
ASTM Designation C-33. Acceptability ofaggregate andsource ,
'u shall be based on thefollowing ASTMtests:
MethodofTest ASTMDesignadon ;
I PotentialReactivity C-289 ,
15.2.3.4 Upon award ofthe subcontract, the Subcontractor shall submit for petrographic analysis, in accordance with ASTM Designation C-295, a 5-poundsample ofquarried material, or of '
l alluvial, 2-U2 pounds each of sand and coarse material which has been certified as sampled at the proposed aggregate source
- by an approvedtestinglaboratory.
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e .._,,__- .y.,. , , , ,, , ,,_-..,m- __- -
I Reactions with Aggregates 13.2.3.6 . . . Aggregates will be tested during theprogress ofthe work . . .The following user tests will be performed on every i 4,000 tons ofaggregates deliveredto theJobsite: ,
MediedofTest ASTMDes4gnaden PotentialReactivity C-289 I i l Both ASTM C289 and C295 provide guidance for selecting aggregates and cements to avoid :
alkali-aggregate reactions, and both standards were specified for use in CCNPPs concrete !
specification. The aggregates used in the containment concrete were specifically inva*ig=M
( tested, and examined in accordance with the ASTM specifications to determine potential for i l reactivity with alkalis.
2.5 Plausibility Determination t
Based on the discussion in Section 2A, the aggregates used in CCNPPs contamment concrete ;
i were specifically investigated, tested, and examined in tswd s with the pertinent ASTM :
- l. specifications to minimize the potential for reactivity with alkalis. 'Ihis conclusion is l supported by a 1992 walkdown inspection report
- that h=na'*~l no indications of concrete '
l
- j. g damage due to this mechanism. For these reasons, reactions with aggregates will not degrade ;
i any concrete components of the contamment and will have no adverse impact on the intended .
functions of these concrete structural coinponents. Therefore, reaction with aggregates is not a f ;
I plausible aging mechanism for any concrete stnictural w-:-wats of the CCNPP ;
f containment.
l 2.6 Existing Programs j l'
There are no existing programs at CCNPP that are designed specifically to identify or to repair j l damage incurred by reaction with aggregates. Since reaction with aggregates is not a plausible ;
l aging mechanism that could degrade the containment structural components, no management ;
program is necessary. .
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3.0 CONCLUSION
Since the potential effects of aggregate reactions on all concrete components were well known l l and understood, measures to avoid using reactive aggregates were implemented for CCNPP in l design specification No. 6750-C-9. The aggregates used in the containment concrete were ,
l specifically investigated, tested, and examined in accordance with applicable ASTM !
specifications to mimmize any reactivity of aggregates with alkalis.
- 4.0 RECOMMENDATION Reaction with aggregates is not a plausible aging mechanism for any concrete component of the CCNPP containment and requires no further evaluation or recommendation.
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l Reactions with Aggregates l
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5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. " Potential Reactivity of Aggregates (Chemical Method)," American Society of Testing and Materials, ASTM C-289-66.
- 3. " Petrographic Examination of Aggregates for Concrete," American Society of Testing and Materials, ASTM C-295-65.
- 4. Letter from Charles County Sand & Gravel Co., Inc. to Bechtel Corporation, June 30,1972.
- 5. " Specification f,r Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear Power Plant Uni.. No. I and No. 2," Design Specification No. 6750-C-9, Revision 8, April 1970. 7
- 6. " Examination of the Unit 1 Containment Structure - Calvert Cliffs Nuclear Power Plant," August 1992.
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l O ^eesaoix e- coaaosion or e=seooeo sreeaesaa 1.0 MECHANISM DESCRIPTION'
! The environments that induce corrosion of reinforcing steel, embedded steel, and cast-in-place anchor bolts are similar. Therefore, this appendix is applicable to all structural components that are either part of or comprise these three component types.
Concrete's high alkalinity (pH > 12.5) provides an environment around embedded i
steel /rebar and protects them from corrosion. If the pH is lowered (e.g., to 10 or less),
corrosion may occur. However, the corrosion rate is still insignificant until a pH of 4.0 is l
i reached. A reduction in pH can be caused by the teaching of alkaline products through cracks, the entry of acidic materials, or carbonation. Chlorides can be present in constituent materials of the original concrete mix (i.e., cement, aggregates, admixtures, and water), or they may be introduced emironmentally. The severity of corrosion is influenced by the properties and type of cement and aggregates as well as the concrete moisture content.
Galvanized decking and galvanized embedments are used in some structures. Since galvanizing material is not considered a dissimilar metal, its application will not aggravate corrosion of the structure.
Studies have also been conducted to determine the effects of stray c!ectrical currents on l t [~]
reinforcing steel. Lightning conductors exchange electrons with the atmosphere and, if v/ connected to reinforcing steel, may accelerate the corrosion process. However, while stray electrical currents can aggravate active corrosion, they are not age-related*.
Corrosion products have a volume greater than the original metal. The presence of corrosion products on embedded steel or rebar subjects the concrete to tensile stress that eventually causes hairline cracking, rust staining, spalling, and more severe cracking.
These actions will expose more embedded steel /rebar to a potentially corrosive emironment and cause further deterioration in the concrete. A loss of bond between the concrete and embedded steel /rebar will eventually occur, along with a reduction in steel cross section. Rebar corrosion can cause deterioration of concrete from a series of hairline cracking, rust staining, spalling, and more severe cracking. These conditions can -
ultimately impair structural integrity.
The degree to which concrete will provide satisfactory protection for embedded steel /rebar depends in most instances on the quality of the concrete and the depth of concrete cover over the steel. The permeability of the concrete is also a major factor affecting corrosion resistance. Concrete oflow permeability contains less water under a given exposure and, hence, is more likely to han lower electrical conductivity and better resistance to corrosion. Such concrete also resists absorption of salts and their penetration into the embedded steel and prmides a barrier to oxygen, an essential element of the corrosien process. Low water-to-cement ratios and adequate air entrainment increase resistance to water penetration and thereby provide greater resistance to corrosion.
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Corrosion of Embedded Steel /Rebar 2.0 EVALUATION At CCNPP, embedded steel has been used in composite structural members or as anchorages of concrete surface attachments. Liner plate anchorages, either steel studs or structural shapu, used in the containment liner, the refueling canal, and the spent fuel pool are also considered as embedded steel. Reinforcing steel (rebar) and cast-in-place anchors are both teated as embedded steel in the evaluation of corrosion effects, because I the environment and the technical basis for their corrosion induction are similar. The l
%se plates under the xtumns or those used as part of attachments to the concrete surface are treated as .tructural steel, and the evaluation of their corrosion effects is addressed in Appendix K. Eccause the design and inspection requirements of liner plates differ 1 significantly from those of structural steel, the corrosion effect on liner plates is discussed )
separately in Appendix L.
2.1 Conditions !
l The only significant inventory of agernsive chemicals stored inside the containment is borated water, and it is primarily in safety related systems such as the primary coolant !
system, safety injection system, and chemical volume control system. Because of the safety significance of these systems, undetected leakages of borated water for an extended period of time cannot occur. Therefore, the containment's interior surface and all internal structural components are not exposed to the risk of aggressive chemicals.
_ 1 The primary area of concern is the exterior surface of the containment where moisture and l oxygen may have access to the embedded steel and rebars. Chlorides in the atmosphere j from the Chesapeake Bay could gain access to the steel. However, only the above-grade portion of the containment is exposed to this emironment. The below-grade exterior surface could be exposed to groundwater on a more or less continuous basis. A dewatering system, installed during construction, would maintain a stable groundwater level at El.+10.0 ft, which is the same elevation as that of the basemat's top surface.
However, there is no known program to determine if the dewatering system continues to perform its function after construction.
2.2 Potential aging mechanism Determination Corrosion of embedded steel /rebar is a potential aging mechanism for the following structural components of containment because they are exposed to the outside emironment and could be subjected to corrosive emironmeats: l Concrete dome LR functions LR-S-1 through 5, and 7 Concrete containment wall LR functions LR-S-1 through 7 Concrete basemat LR functions LR-S-1 through 7 1
where:
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Corrosion of Embedded Steel /Rebar l A
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IR-S-1: Provides structural and/or functional support to safety-related l
equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
)
LR-S-3: Servei as a pressure boundary or fission product retention barrier to protect public health and safety in the event of any postulated DBEs.
LR-S-4: Serves as a missile barrier (internal or external).
LR-S-5: Provides structural and/or functional support to non-safety-related l equipment whose failure could directly prevent satisfactory accomplishment of l any of the required safety-related functions. j LR-S 6: Provides flood protection barrier (internal flooding event).
LR-S-7: Provides rated fire barriers to confine or retard a fire from spreading or from adjacent areas of the plant.
Other concrete structural components are located inside the containment building; therefore, corrosion of embedded steel /rebar is not a potential aging mechanism.
2.3 Impact on Intended Functions, _ ,
If the effects of corrosion of embedded steellrebar were not considered in the original design or are allowed to degrade the above structural components unmitigated for an extended per'od of time, this aging mechanism could affect all the intended functions of components listed in Section 2.2.
2.4 Design and Constmetion Considerations The containment structure was constructed with concrete that complies with - CCNPP's 8
design specification No. 6750-C-9, which adheres to the relevant ACI Codes and ASTM specifications for a concrete structure of low permeability. Also proper concrete covers were specified in accordance witn ACI 318 Code to effectively prohibit exposure of embedded steel /rebar to the corrosive environment. Another design consideration was the use of prestressed tendons to minimize crack development in the concrete containment.
d In addition, CCNPP's UFSAR specifies the rebar requirement to minimize concrete crack l' development as follows:
O.23 percent reinforcing shall be provided at the tension face for small members; O.2 percent, for medium size members; 0.13 percent,for large members.
A minimum of 0.2 percent bonded reinforcing steel is provided in two perpendicular directions on the exteriorfaces of the wall and domeforproper crack control.
50N6 E E-3 Revision 3
Corrosion of Embedded Steet/Robar !
During initial plant construction, a cathodic protection system was installed at the CCNPP site to adtigate steel corrosion,' including the rebars in the containment wall and concrete basemat of the containment, j l
2.5 Plausibility Determination Based on the discussion in Sections 2.1 and 2.4, corrosion is not a plausible aging mech ==i== for *=h* Mad steel /rebar in the containment dome, the above-grade portion of mainin= ant wall, and all components inside containment. This conclusion is supponed s
by a 1992 walkdown inspection report that d-:-:"===ted no indications of damage to concrete due to corrosion of embedded steel /rebar. l As discussed in Section 2.1, only the below-grade portion of the containment wall and the concrete basemat could be exposed to an aggressive environment on a continuous basis and could be susceptible to embedded steel /rebar corrosion. Because the chemical quality !
of the underground water is not known, corrosion of embedded steel /rebar is a plausible aging mechanism for the below-grade portion of the concrete containment wall and the concrete basemat. l 2.6 Existlag Programs l i There are no existing programs at CCNPP that are designed specifically to identify or to repair damage of the concrete structure due to corrosion of embedded steel /rebar.
3.0 CONCLUSION
Based on the discussion in Sections 2.1 and 2.4, corrosion of embedded steel /rebar is not a plausible aging mechanism for concrete components inside the containment, the containment dome, and the above-grade portion of the containment wall. No further evaluation is required for these concrete structural components.
Because the quality of the groundwater is not known,' corrosion of embedded steel /rebar is ,
a plausible aging mechanism for the below-grade portion of the containment wall and the concrete basemat.
4.0 RECOMMENDATION During initial plant construction, groundwater observation wells were installed to monitor j the fluctuation of the groundwater table, and samples were taken for groundwater quality testing.' Although the wells are still in place, the monitoring activities have been _j discontinued. It is recommended that the groundwater water quality be tested using these wells. This data can be used to evaluate the effects of chemical attacks on the below-grade portion of the containment structure due to aggressive groundwater exposure.
Win 6 a E-4 Revision 3
l Corrosion of Embedded Steel /Rebar
5.0 REFERENCES
l 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. Skoulikidas, T., Tsakopoulos, A., and Moropoulos, T., " Accelerated Rebar Corrosion When Connected to Lightning Conductors and Protection of Rebars with Needle Diodes Using Atmosphere Electricity," in Publication ASTM STP 906, " Corrosion Effect of Stray Currents and the Techniques for Evaluating Corrosion of Rebars in Concrete."
- 3. " Specification for Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear Power Plant Unit No. I and 2," CCNPP's Design Specification No. 6750-C-9, Revision 8, April 1970.
- 4. "Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR)," Baltimore Gas and Electric Co.
- 5. "Exan,inction of the Unit 1 Containment Structure - Cr.tvert Cliffs Nuclear Power Plant," August 1992.
- 6. " Specification for Furnishing and Installation of Piezometers - Calvert Cliffs l Nuclear Power Plant Unit No. I and 2,".CCNPP's Design Specification.No., ,, _ . , _ '
l l 6750-C-23E, Revision 0, November 1973.
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O Areewoix e-casse 1.0 MECHANISM DESCRIITION' Creep is defined as the time-dependent increase of strain in hardened concrete that has been subjected to sustained stress. The sustained stress results from the dead load and live load of the stmeture and from temperature effects. Creep defonnation is a function ofloading history, environment, and material properties of the concrete. The time-dependent deformation of concrete under compressive load consists of strain resulting from progressive cracking at the aggregate-cement paste interface, from moisture exchange with the atmosphere, and from moisture movement within the concrete.
The effects of temperatures on creep are not linear. At 122 *F, creep strain is about two to three times as great as at room temperature (68 - 75 *F.) But from 122 *F to 212 *F, creep strain continues to increase four to six times that experienced at room temperatures While l
little is known about creep rate beyond 212 *F, the maximum creep rate may have occurred between 122 *F and 176 *F.'
Creep is not visible because micro-cracking occurs at the aggregate cement-paste interface.
The defonnation resulting from cracking and from moisture exchange with the atmosphere is not recoverable. Creep defonnation can generally be characterized as follows:
h V
- Increased water-to-cement ratio results in increased creep magnitude.
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Increased aggregate-txement ratio results in increased creep magnitude for a-l given volume ofconcrete.
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Creep deformation is approximately proportional to the applied load for a level not exceeding abou; 40% to 60% of the ultimate strength of concrete. )
l Concrete age at application of load affects creep (i.e., the older the concrete, the l less the creep). l Creep increases with increased temperature.
Aggregate with a high modulus of elasticity and low porosity will minimize creep.
l Creep-induced concrete cracks are typically not large enough to result in concrete deterioration or in exposure of the reinforcing steel to environmental stressors. Cracks of this magnitude do not reduce the concrete's compressive strength. Creep is significant when new concrete is subjected to load and decreases exponentially with time. Any degradation is noticeable in the first few years of plant life. According to ACI 209R-82,' 78% of creep i
occurs within the first year,93% within 10 years,95% within 20 years, and %% within 30 years. At any given stress, high-strength concretes show less creep than low-strength l
j concretes.
ACI 209R-82 provides guidance for predicting creep in concrete stmetures. Prestressed concrete structures may be subject to more pronounced creep and relaxation effects, O
d particularly in combination with elevated temperatures. Its effect is reflected in terms of prestressing loss in tendons, which is discussed in Appendix N.
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i Creep i
l 2.0 EVALUATION 2.1 Conditions There is no condition in CCNPP that could aggravate the effect of concrete creep initiated right after concrete constmetion. Most of the concrete creep will have occurred well before the time of a license renewal application. Herefore, creep of concrete structural components should not be regarded as an aging mechanism for license gneScal. .
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2.2 Potential Aging Mechanism Determination Creep is not a potential aging mechanism for any contamment concrete stmetural components I because creep proceeds at a decreasing rate with age and is not expected to continue after 40 years.
2.3 Impact on Intended Functions I
Since creep is not a potential aging mechanism, it will not affect the intended functions of any
, containment structural components.
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! (/ 2.4 Design and Construction Considerations i l At CCNPP, all reinforced concrete components, except the containment dome and the cylinder, were designed based on the working stress design method. The induced stresses are much lower than the ultimate strength of the concrete, which is specified as f. = 5,000 psi for all concrete structural components except for containment basemat. The conta'mment basemat i was constructed of concrete with f, = 4,000 psi'. However, the containment basemat is l
subject to low forces during normal plant operation condition. Therefore, creep in all concrete l- components is minimal because of the low compressive stresses in concrete and the use of high-strength concrete. Besides, creep proceeds at a decreasing rate with age; normally, % %
! of creep has occurred within 30 years. Herefore, creep is not expected to continue during the license renewal period.
l 251 Plausibility Determination Not applicable.
2.6 Existing Programs Not applicable.
3.0 CONCLUSION
Most of the concrete creep occurred well before the time of license renewal application.
- ,O Therefore, creep of concrete structural components should not be regarded as an aging mechanism for license renewal.
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Creep
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j 4.0 RECOMMENDATION Not applicable.
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5.0 REFERENCES
l i
l 1. " Class 1 Structures License Renewal lodustry Report," EPRI's Project RP-2643-27, December 1991.
- 2. " Prediction of Creep, Shnnkage, and Temperature Effects in Concrete Structures,"
American Concrete Institute, ACI 209R-82.
- 3. " Specification for Furnishing and Delivery of Concrete - Calvert Cliffs Nuclear l
l Power Plant Unit No. I and 2," CCNPP's Design Specification No. 6750-C-9, l Revision 8, April 1970.
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I APPENDIX G - SHRINKAGE l 1.0 MECHANISM DESCRIPTION '
l.1 Concrete A workable concrete mix typically contains more water than is needed to offset the effects of hydration. When concrete is exposed to air, large portions of the free water evaporate. As water evaporates, capillary tension develops in the water armaining in the concrete while the concrete drie and shrinks in volume. Should these stresses exceed the tensile strength of the concrete, a crack forms. Initial shrmkage occurs during curing and continues months after placement. Subsequent drying and shrmkage occurs in concrete that is not continuously wet 2
or submerged. According to ACI 209R-82,91% of the shrmkage occurs during the first year,98% in 5 years, and 100% in 20 years.
Excessive shrinkage causes cracking of the concrete surfaces, which provides a means for aggressive elements to make contact with the embedded steel /rebar, thus promoting the possibility of corrosion. The aging mechanism due to corrosion of embedded steel /rebar is discussed in Appendix E.
He other effect of concrete shrmkage is the prestress losses in tendons of the prestressed concrete Containment Stn cture, which is discussed in Appendix N.
O V 1.2 Marinite XL Board
~
he Marinite XL board is fabricated from calcium silicate with inert fillers and reinforcing agents. Shrmkage of Marinite XL board material may occur due to exposure to elevated j temperatures.
2.0 EVALUATION l
2.1 Conditions here is no condition in CCNPP that could aggravate the effect of concrete shrinkage initiated right after concrete construction. Most of the concrete shrinkage will have occurred well before the time of a license renewal application. Therefore, shrinkage of concrete structural components should not be regarded as an aging mechanism for license renewal.
He Marinite XL boards are located in the Containment Structure. The normal environmental conditions (outside the primary shield) are as follows * -
Temperature 120'F (average limit) i llumidity 70 % (max.)
! Radiation 0.35E6 rads (40 years integrated gamma dose under normal operating l
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'I Shrinkage i
2.2 Potential Aging Mechanism Determination l Shrinkage is not a potential aging mechanism for any Containment Structure concrete '
structural components because shrinkage in concrete proceeds at a decreasing rate with age . _
and is not expected to continue after 40 years.
i Shrmkage is a potential aging mechanism for the following components based on environmental conditions inside the Containment Structure i i
e Partitions and Ceilings (Mannite XL board) LR function LR S-7 !
i
. where: ;
LR-S-7: Provides rated fire bemers to confine or retard a fire from spreadmg to or . .;
from adjacent areas of the plant. ;
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2.3 Impact on Intended Functions e Since shnnkage is not a potential aging mechanism, it will not affect the intended functions of
( . any Containment Structure concrete structural components. ;
Degradation due to shnnkage could result in the .Marinite XL board being unable to perform its intended function ofserving as a fire barrier 2.4 Design and Construction Considerations - l Concrete }
l Since shnnkage can be minimized by keeping the water content of the paste as low as !
possible, the use oflow slump concrete is a major factor in controlling shnnkage ', As stated J in paragraph 12.1.2.1 of CCNPP design specification No. 6750-C-9,' a low slump of 2 inches ,
was specified for all concrete used in the CCNPP Containment Structure. ;
'Ihe development of concrete cracking due to shnnkage can also be minimized by providing I
adequate reinforcing steel. For this purpose, CCNPP has specifed the minimum rebar requirement in Section 5.1.2.5 of the UFSAR: '
O.25 percent reinforcing shall be provided at the tensionfacefor small members; 0.2 percent.for medium size members; and 0.15 percent,for large members.
A minimum of 0.2 percent bonded reinforcing steel is provided in two \
perpendicular directions on the exteriorfaces ofthe wall anddomefor 'l proper crack control.
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Since low slump concrete is used at CCNPP to minimize concrete cracks from shrinkage and ,
additional rebars are used to mitigate crack propagation, shrinkage of any concrete )
!' component of the Containment Structure is mimmal Marinite XL Board i ne Marinite XL board is used as a fire separation barrier on cable trays within the Containment Structure. De board is designed to meet Appendix R cnteria. He Marinite j XL board is used only on cable trays outside the primary shield in the Containment Structure'. He Marinite XL boani is fabricated from calcium silicate with inert fillers and reinforcing ag,ents. Shrinkage of Marinite XL board matenal begins to occur at temperatures l' above 300*F' . He board is required to maintain its integrity during normal environmental conditions within the Containment Structcre in order to provide the barrier function.
- j 2.5 Plausibility Determination Based on the discussion in Sections 2.1 and 2.4, no architectural components of the Containment Structure are subject to conditions higher than their design threshold. Also, i shrinkage is not a potential aging mechanism for structural components of the Containment Structure. Herefore, shrinkage is not a plausible aging mechanism for any structural or architectural components of the CCNPP Containment Structure.
- 2.6 Existing Programs _ . . ._ _ . . . _ _ _ , ,
Here are no existing programs at CCNPP design'ed to identify damages to components of the Containment Structure due to shrinkage. However, since this is not a plausible aging mechanism that could degrade these components, no program is necessary.
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3.0 CONCLUSION
Shnnkage in concrete is not a long-term aging mechanism and is not expected to continue after 40 years during the license renewal period. .
All Marinite XL board in the Containment Structure are exposed to the normal .
l environmental conditions. As indicated in Section 2.0 above, the temperature level inside the Containment Structure are predicted to be below the degradation threshold for the Marinite l XL board. Derefore, shrinkage is not a plausible age-related degradation mechanism for the ]
Marinite XL board in the Containment Structure.
4.0 RECOMMENDATION Shnnkage in concrete is not a long-term aging mechanism and is not expected to continue after 40 years during the license renewal period.
f Shnnkage of Marinite XL board is not a plausible aging mechardsm for the Containment Structure. No further evaluation or recommendation is required.
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5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
! 2. " Prediction of Creep, Shrmkage, and Temperature Effects in Concrete Structures,"
American Concrete Institute, ACI 209R-82
- 3. Design and Control of Concrete Mixtwes, lith Edition, Portland Cement Association, July 1%8.
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- 4. " Specification for Furnishing md Delivery of Concrete - Calvert Cliffs Nuclear i Power Plant Unit No. I and 2," CCNPP's Design Specification No. 6750-C-9, j
Revision 8, April 1970. i l S. "Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Updated Final Safety Analysis l Report (UFSAR)," Baltimore Gas and Electric Co.
l m 6. Manville Products Catalog IND-33610-80 (page 3).
!U l 7. Drawing 62-150-E,. Revision . 6,... Appendix .R . Separation . . Requirements, Containment El. S'-0" -
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- 8. "EQ Design Manual - Calvert Cliffs Nuclear Power Plant, Unit No. I and 2,"
Baltimore Gas and Electric Co.
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APPENDIX H - ABRASION AND CAVITATION 1.0 MECHANISM DESCRIPTION' As water moves over concrete surfaces, it can cany abrasive materials or it can create a i negative pressure (vacuum) that can cause abrasion and cavitation. If significant amounts of a concrete are removed by either of these processes, pitting or aggregate exposure occurs due to loss of cement paste. These degradations are readily detected by visual examination in l accessible locations.
Abrasion and cavitation occur only in concrete structures that are continuously exposed to flowing water. Cavitation damage is not common if velocities are less than 40 fps. In closed conduits, however, degradation due to cavitation can occur at velocity as low as 25 fps when abrupt changes in slope or curvature exist.
4 2.0 EVALUATION 2.1 Conditions Neither the containment building nor its structural components are exposed to continuously
- flowing water.
2.2 Potential Aging Mechanism Determination Attack by abrasion and cavitation is not a potential aging mechanism for the structural components of containment because the CCNPP containment is not exposed to continuously flowing watei, 2.3 Impact on Intended Functio'ns Not applicable.
2.4 Design and Construction Considerations Not applicable.
2.5 Plausibility Determination Not applicable.
2.6 Existing Programs Not applicable.
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i (g Abrasion and Cavitation
3.0 CONCLUSION
The CCNPP containment is not exposed to continuously flowing water. Therefore, abrasion and cavitation are not a potential aging mechanism for any structural components of the containment.
l 4.0 RECOMMENDATION Not applicable.
5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
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(J) APPENDIX l - CRACKING OF MASONRY BLOCK WALLS L0 MECHANISM DESCRIPTION' Masonry blocks walls can be designed as structural or shield walls. Masonry block wall cells may or may not contain reinforcing steel to provide structural strength for the wall. The extent of grouted cells varies with the specific design requirements for a bearing wall.
Some age-related degradation mechanisms that affect masonry block walls are the same as those that affect reinforced concrete walls. The potential for embedded steel and reinforced steel corrosion in block walls is similar to that of reinforced concrete.
Masonry block walls are vulnerable to unique age-related degradation mechanisms. Any restraint imposed on a masonry block wall that will prevent the wall from free expansion or contraction will induce stresses within the wall. Restraint against expansion results in small stresses depending on the strength of the block wall materials and thus rarely causes degradation of the concrete block wall. Moreover, expansion of the wall is offset by shrmkage from carbonation and drying. Restraint against free contraction causes tensile stresses within the wall. If these stresses exceed the tensile strength of the unit, the bond strength between the mortar and the unit, or the shearing strength of the horizontal mortar joint, cracks occur to relieve the stresses. Expansion or contraction of masonry block walls may be caused by changes in temperature, changes in moisture content of the constituent materials, carbonation, rN and movement of adjacent stmetural components (e.g., supporting floor or foundations). 1 Shrinkage due to moisture loss is among the principal causes of volume changes in masonry block walls. Factors affecting the drying shrinkage are the type of aggregate used, the method )
of curing, and the method of storage. Units made with sand and gravel aggregate will normally exhibit the least shrinkage; those with pumice, the highest. The difference between the moisture content of the masonry units during construction and the building in use will determine the amount of shrinkage that occurs. High-pressure steam curing and proper drying of concrete masonry units reduce the potential shrinkage of the walls. ;
If proper isolation is not provided at the joint between the masonry block wall and the !
supporting structural components (e.g., floor slabs or beams), long-term creep and variation in stiffness of the supporting components can also cause cracking.
Durability of the masonry mortar used at the block joints may affect the long-term structural integrity of the masonry block wall. Although aggressive environments and the use of unsound materials may contribute to the deterioration of mortar joints, most degradation results from water entering the concrete masonry and freezing.
The mechanisms cited above which cause cracking of concrete block walls are age-related.
Although they are ongoing processes throughout a plant's life, most cracking occurs in the early stages ofplant operation.
2.0 EVALUATION 2.1 Potential Aging Mechanism Determination A
There is no masonry block wall in the containment system. Therefore, this aging mechanism does not apply to the containment.
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Cracking of Masonry Block Walls 1
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Not applicable.
2.3 Design Considerations Not applicable.
2.4 Impact on Intended Functions >
Not applicable.
2.5 Plausibility Determination Not applicable.
2.6 Existing Programs Not applicable.
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3.0 CONCLUSION
I Cracking of masoray block walls is not a plausible degradation mechanism for CCNPP's l containment.
4.0 RECOMMENDATION Not applicable.
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5.0 REFERENCES
Not applicable.
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!p t s APPENDIX J - SETTLEMENT l
l 1.0 MECHANISM DESCRIPTION' All structures settle during constmetion and for months after constmetion. The amount of settlement depends on the physical properties of foundation material. These properties range from rock (with little or no settlement likely) to compacted soil (with some settlement expected). Settlement may occur during the design life from changes in environmental conditions, such as lowering of the groundwater table. Settlement can occur in two stages:
clastic expansion and time-dependent settlement. Elastic expansion of the confmed soil occurs i due to excavation unloading and results in a slightly upward movement. During construction, l the soil moves downward as load is applied. ~1his clastic movement should be small and is l complete when constmetion is completed. It has no effect on the stmeture and is not 2
l considered an aging mechanism . He excavation unloading and structural loading cause a
- small change in the void ratio of the soil. His change results in a very small amount of time-l dependent settlement. The settlement rate will decline after completion of construction.
l l Settlement of structures is usually small and is typically determined by survey. Concrete and
! steel stmetural members can be affected by differential settlement between supporting l foundations, within a building, or between buildings. Severe settlement can cause l misali,unent of equipment and lead to overstress conditions within the structure' When
( buildings experience significant settlement, cracks in stmetural members, differential elevations of supporting members bridging between buildings, or both may be visibly j
(3 f
%)
) detected. >
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t 2.0 EVALUATION l
2.1 Conditions' ,
i The basemat elevation of the containment at CCNPP is approximately 70 feet below the average ground elevation. The basemat is situated on Miocene soil, which is exceptionally dense and will support heavy foundation loads. He major soil types are sandy silts, silty sands, and slightly clayed sands. He ultimate bearing capacity of the foundation strata is in excess of 80,000 psf, and the allowable bearing capacity is 15,000 psf. However, the design ,
, bearing pressure of the basemat is 8,000 psf. The soil bearing pressure was about the same as
! the overburden removed due to excavation. i 2.2 Potential Aging Mechanism Determination !
Settlement is a potential aging mechanism for all structural components in the containment structure. Since the concrete basemat is the only structural component directly supported by the soil media, and also for convenience of discussion, only the concrete basemat is identified as the structural component subject to the aging mechanism due to settlement.
Concrete basernat LR-S-1 through 7 C')
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LR-S-1: Provides structural and/or functional support to safety-related equipment.
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! LR-S-2: Provides shelter / protection to safety-related equipment. l l
LR-S-3: Serves as a pressure boundary or fission product retention barrier to protect l public health and safety in the event of any postulated DBEs.
i LR-S-4: Serves as a missile barrier (intemal or external).
LR-S-5: Provides structural and/or functional support to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the requhi safety-related functions. ]
LR-S-6: Provides flood protection barrier (internal flooding event). i i
LR-S-7: Provides rated fire barriers to confine or retard a fire from spreading to or l l from adjacent areas of the plant.
2.3 Impact on Intended Functions If the effects of settlement were not considered in the original design or are allowed to degrade i the above structural component unmitigated for an extended period of time, this aging j mechanism could affect intended functions LR-S-1,3, and 5 'of the concrete basemat. !
2.4 Design and Construction Considerations
! In addition to soil bearing capacity, settlement of the containment basemat was also investigated in the design of the contaimnent. A maximum post-construction settlement of 1/2 2
inch was predicted in the original containment design . S nce the concrete basemat is a rigid foundation and is situated on a exceptionally dense soil, the containment structure tends to l
uniformly settle as a rigid body. Most of the predicted 1/2 inch settlement is in terms of .
uniform settlement, which has no adverse effect on the stmetural components of the containment. A small fraction of the 1/2 inch settlement will be in terms of differential settlement. It is so small that the effect on the structural component is negligible.
ne excavation for the containment building was below the groundwater table. A dewatering system was installed during plant construction to maintain the groundwater table at El.10'-0".
His groundwater table level was considered in the original design of all underground structures.' )
2.5 Plausibility Determination Based on the discussion in Sections 2.1 and 2.4, the soil type at the CCNPP containment is exceptionally dense, and the design bearing pressure is about the same as that of the removed overburden and is much smaller than the allowable bearing capacity. As discussed in Section 2.4, the predicted settlement is small and the differential settlement is negligible. A l
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dewatenng system was installed to minimize the fluctuation of groundwater table, thus providmg stable geological conditions of the plant site. 'Iherefore, vestlement is not u - ;
plausible aging mechamsm for any structural wo.,,c-.;. of the mntainment 2.6 - Existing Programs I
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- l. There are no existing programs at CCNPP that are designed specifically to identify or to repair ;
damage to concrete mcurred by settlement. Since this is not a plausible aging mechanism that I could degrade the contamment structural &=, no management program is necessasy. ;
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3.0 CONCLUSION
CCNPP's containment is situated on Miocene soil, which is -W 81y dense and will - ,
support heavy foundation loads. Addnionally, the stmetural load on the containment basemat 1 is about the same as the removed overbunien weight. 'therefore, the soil bearing stress is well below its ultunate bearmg capacity, and the long-term settlement is predicted to be only 1/2 inch.' In addnion, the settlement rate declined aAer completion of w, ;..4;cs, and the '
groundwater table is mamtamed by the L_..g system. Long-term settlement is not ,
expected to contmue aAer 40 years. Therefore, settlement is not a plausible aging mechanism {
for the structural 9-- ; = #= of the containment.
b G '1 4.0 RECOMMENDATION _
l- S** lament is not a plausible aging mechanism for the conen:te basemat of the containment and l requires no further evaluation or recommendation.
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5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. "Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Updated Final Safety Analysis Report (UFSAR)," Baltimore Gas and Electric Co.
- 3. " Pilot Studies on Management of Aging of Nuclear Power Plant Components,"
International Atomic Energy Agency,IAEA-TECDOC-670, October 1992.
- 4. Civil and Structural Design Criteria for Calvert Cliffs Nuclear Power Plant, Units I and 2, by Bechtel Power Corporation, Revision 0, August 2,1991.
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l G]/ APPENDIX K - CORROSION OF STEEL ;
1.0 MECHANISM DESCRIPTION' l
Steel corrodes in the presence of moisture and oxygen as a result of electrochemical reactions.
Initially, the exposed steel surface reacts with oxygen and moisture to form an oxide film as rust. Once the protective oxide film has been formed and ifit is not disturbed by erosion, alternating wetting and drying, or other surface actions, the oxidation rate will diminish rapidly with time. Chlorides, either from seawater, the atmosphere, or groundwater, increase 1 the rate of corrosion by increasing the electrochemical activity. If steel is in contact with I
another metal that is more noble in the galvanic series, corrosion may accelerate.
In some cases, corrosion of stnictural steel in contact with water may be microbiologically induced due to the presence of certain organisms, which is sometimes referred to as microbiologically influenced corrosion (MIC). Rese organisms, which include microscopic forms such as bacteria and macroscopic types such as algae and barnacles, may influence corrosion on steel under broad ranges of pressure, temperature, humidity, and pH. MIC effects on carbon steel may result in rr.ndom pitting and general corrosion.
He rate of steel corrosion depends on site-specific environmental conditions and measures taken to prevent corrosion. A steel structure surface subjected to attemately wet and dry conditions corrodes faster than one exposed to continuously wet conditions. Atmospheric
/ corrosion proceeds much more rapidly in areas where the atmosphere is chemically polluted C)/ by vapors of sulfur oxides and similar substances. Steel will cormde much faster in the vicinity of seawater because of sodium chloride in the atmosphere. He corrosion ~ rate of steel usually increases with rising temperatures.
Corrosion products such as hydrated oxides of iron (rust) form on exposed, unprotected surfaces of the steel ar.d are easily visible. 'Ihe affected surface may degrade such that visible i
perforation may occur. In the case of exposed surfaces of structural steel with protective coatings, corrosion may cause the protective coatings to lose their ability to adhere to the corroding surface. In this case, damage to the coatings can be visually detected well in advance of significant degradation.
l 2.0 EVALUATION 2.1 Conditions Steel can corrode in the presence of moisture and oxygen as a result of electrochemical reactions, especially in areas where there is an inadequate drainage system. In containment, structural steel components vulnerable to corrosion are the steel members such as base plates and brackets that are not readily accessible for visual inspection and that can form pockets to harbor liquids.
2.2 Potential Aging Mechanism Determination l
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! - Corrosion is a potential aging mechanism for the following containmeni structural steel (d components because conditions conducive to steel corrosion discussed in Sections 1.0 and 2.1 exist:
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T Corrosion of Steel Steelcolumns LR functions LR-S-1,5 l
Steelbeams 1.R functions LR-S-1,5 l
Base plates LR functions LR-S-1,5 ;
Floor framing LR functions LR-S-1,5 1
Steelbracing LR functions LR-S-1,5 l
Platform hangers LR functions LR S-1,5 Decking LR functions.LR-S-1,5 i
Floor grating LR functions LR-S-1,5 1 Checkered plates LR functions LR-S-1,5
- Cast-in-place anchors LR functions LR-S-1,5 !
Post-installed anchors LR function LR-S-1 1
Crane girder - LR function LR-S-5 l l
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Lubrite plates LR function LR-S-1,5 where:
LR-S 1: Provides structural and/or functional support to safety-related equipment.
LR-S-5: Provides structural and/or functional support to non-safety-r: lated equipment whose failure could directly prevent satisfactory accomplishment of ,
any of the required safety-related functions.
2.3 Impact on Intended Functions If corrosion of steel is allowed to degrade the above structural steel components unmitigated
- for an extended period of time, this aging mechanism could affect all intended functions of '
components listed in Section 2.2.
2.4 Design and Construction Considerations Since corrosion was considered a potential degradation mechanism for all structural steel components of the containment, its effects were considered in the original design. As a result, all exposed structural steel surfaces in the containment except grating, checkered plates, and
} metal decking, which are galvanized steel, were shop-painted or field-painted during the
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constmetion phase in accordance with CCNPPs design specifications No. 6750-C-31' and No.6750 A 24'.
Maintenance of protective coatings on CCNPPs equipment and structures follows the requirements specified in Calvert Cliffs Procedum MN-3-100'. . His program sets forth procedural controls that comply with 10 CFR Part 50, Appendix B and satisfy the protective coating requirements in Regulatory Guide 1.54 which endorses ANSI N101.4-1972. His procedure requires inspection of all painted plant areas inside containment. Application of coatings at CCNPP follows standard procedures specified in the applicable design standard'.
Dis standard specifies material based on normal and post-accident ensironmental conditions i
- and addresses the refurbishment of exi> ting coated surfaces and the need for new coating and recoating. I
- Galvanic material on steel components is one of the protective coating systems. Maintenance ;
- . of galvanic coatings is covered under the same maintenance pmgram as paint and other l protective coatings.
. . l ne post-installed anchon used for the horizontal tie rod anchorages of reactor coolant pump l restraints are Wej-It' concrete anchors. Wej-It anchor bolts are made of cold-rolled, high strength steel having a rust-resistant zinc coating with a final clear acetate coating that -
p exceeds U. S. govemment specifications and the ASDi B-ll7 salt spmy test.' ,
2.5 Plausibility Determination _ _ . . . - . .
l Based on the discussion in Sections 2.1,2.3 and 2.4, corrosion could affect the intended functions of all structural steel members and is, therefore, a plausible aging mechanism for all steel components listed in Section 2.2.
2.6 Existing Programs A coating program, MN-3-100' is implemented at CCNPP for the exposed surfaces of structural steel components inside the containment. Conditions adverse to quality (such as degraded paint or corrosion) are reported in an Issue Repo't under Qle2-100' De coatings .
program provides the administrative control over how corrective actions are performed. De combination of these existing plant programs will ensure that corrosion effects on accessible structural steel are adequately managed.
Dese programs do not provide for the evaluation of the coating condition on structural steel !
components that are not normally accessible. An age related degradation inspection program as defined in the BGE Integrated Plant Assessment Methodology is necessary to address.the aging efrects of the non-accessible structural steel components
3.0 CONCLUSION
j All exposed structural steel surfaces in the containment are covered by a protective coating.
All these protective coatings are maintained per Calvert Cliffs Procedure MN-3100'. His
. procedure requires an inspection at the beginning of each major outage of all coated plant l
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l r Corrosion of Steel areas inside containment. In areas accessible for coating 'n5+:tks damage to coatmg can be detected visually well in advance of degradation due to corrosion of the structural steel.
Aging management of degraded coatmg conditions on accessible structural steel in the Containment Structure is accomplished through the combination of existing plant programs.
However, structural steel components not readily accessible require additional aging management.
l 4.0 RECOMMENDATION i
l Coatings on structural steel in accessible areas are adequately managed by existing plant programs. A new program utilizing an age related degradation 'n5+:tkn should be developed to address degradation of coatings on structural steel components that are not normally accessible l l
5.0 REFERENCES
- 1. " cms 1 Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. " Specification for Furnishing, Detailing, Painting, and Delivering Containment
- and Auxiliary Building Structural Stect," CCNPP's Design Specification No. 6750-l C-31, Revision 2, May 1970.
- 3. " Specification for Painting and Special Coatings," CCNPP's Design Specification No. 6750-A-24, Revision 12, October 1982.
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- 4. " Protective Coating and Painting Program," CCNPPs Procedure MN-3-100. .
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- 5. " Coating Application Performance Standard," TRD-A-1000, Calvert Cliffs Nuclear Power Plant, Unit No. I and 2, Revision 8, August 1991.
- 6. " Specification and Load Data for Wej-Its," Vendor's product catalog by Wej-It Corp., Brownfield, Colorado.
- 7. " Issue Reporting and Assessment", Calvert Cliffs Nuclear Power Plant Administrative Procedure QL-2-100, Revision 4. Date 1/2/96 j
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I hr~ APPENDIX L- CORROSION OF LINER 1.0 MECHANISM DESCRIPTION'#
1.1 Carbon Steel Liner Carbon steel liner corrosion can be either galvanic or electrochemical. Electrochemical corrosion of carbon steel is caused by exposure to aggressive aqdeous solutions, which is described in Appendix K, " Corrosion of Steel."
l Galvanic cormsion occurs only in the presence of electrolyte when the electrical potential difference between dissimilar metals placed in contact with each other results in the flow of electrons between them. The less resistant metal becomes the anode in this couple and is subject to corrosion, while the more resistant metal becomes the cathode and corrodes very little, if at all. The rate of galvanic corrosion is a function of the potential difference between ,
the metals and the geometric relationship of the metals. Galvanic corrosion reduces the j thickness of the anode metal.
Liner corrosion reduces liner plate thickness. Excessive reduction in thickness compromises the pressure retention capability of the liner. Co:Toded surfaces of the liner could result in separation of the protective coatings from the steel surface, and coating degradation becomes apparent.
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U l.2 Stainless Steel Line_r ne stainless steel liner is prone to stress cormsion cracking (SCC), which is defined as cracking under the combined actions of corrosion and tensile stresses. He phenomenon of SCC can result in fracture of the metal. He stresses may be either applied (extemal) or residual (intemal). He stress corrosion cracks themselves may be either transgranular or ,
intergranular, depending on the metal and the corrosive agent. As is normal in all cracking, i the crackr are perpendicular to the tensile stress. Usur.lly there is little or no obvious visual I evidence of corrosion. The three principal factors necessary to initiate stress cormsion cracking are tensile stresses, corrosive environment, and susceptible material. The tensile stresses necessary to cause SCC must be at or near the material's yield point. This is I facilitated when the material is substantially cold worked, contains residual stress from welding, or is subjected to significant applied loads. Different corrosive environments induce different levels of SCC on various materials. With respect to material susceptibility, austenitic stainless steels, such es SA 240 Type 304, are prone to SCC, particularly when sensitization is present as in heat-affected zones and at creviced geometries.
In a sensitized condition, Type 304 stainless steel may develop intergranular stress corrosion cracking (IGSCC). The heat-affected zones of welds in Type 304 stainless steel are potential sites for IGSCC. IGSCC occurs when changes in the microstructure take place due to the welding heat, rendering the heat-affected zones " sensitized", and when high residual stresses occur in and around the welds. The degree of sensitization depends on the metal's composition. For example, sensitization usually occurs when Cr in boundaries combines with carbon. A low carbon content stainless steel, such as Type 304L, is relatively immune to n IGSCC in the fuel pool environments. His is because the low carbon content (0.03 percent maximum) of Type 304L results in sensitization levels during welding so low that its heat-affected zones are resistant to IGSCC in the fuel pool environments.
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2.0 EVALUATION ne containment /basemat liners and the refueling pool liner at CCNPP are ASTM A36 I carbon steel and SA 240 Type 304 stainless steel, respectively.8# nese liners were constructed from a series ofindividual steel plates welded together. Both the plate material and the welds are subject to the same potential degradation mechanisms. De significance of potential degradation of the liners is considered to apply equally to the plate material and the welds.'
Neither the carbon steel liner nor the stainless steel liner has dissimilar metals; therefore, they are not subject to galvanic corrosion.
2.1 Conditions 2.1.1 Containment Walland Dome Liner l The containment wall and the containment dome are 3' 9" and 3'-3" thick, respectively, and are subject to compressive stress due to dead weight and prestress load under normal plant
! ,o operating conditions. His configuration minimizes cracks in the concrete that allow
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penetration of moisture, oxygen, and chlorides, which cause corrosion degradation.
nerefore, the containment liner from the concrete side is not exposed to aggressive chemicals from the outside environment, such as acid rain, salt-containing atmospheres, and groundwater. De interior surfaces of the containment wall and dome liners are exposed to the containment intemal environment, and corrosion of these surfaces could occur in the presence of moisture and oxygen as a result of electrochemical reactions unless the existing coating is maintained by an effective coating management program.
2.1.2 Basemat Liner ne top surface of the basemat liner is protected from equipment and corrosive agents by an 18-inch-thick concrete cover . De bottom surface of the basemat liner is susceptible to .
potential contact with underground water. The primary paths of ingress for these fluids are the construction joints in the concrete basemat underneath the liner.
2.1.3 Refueling Poot Liner SA-240 Type 304 stainless steel used in the refueling pool is resistant to electrochemical corrosion in the refueling / spent fuel pool environments. The corrosion rate of this steel ranges from 0.05 mil in 100 years (virtually no corrosion) to less than 0.01 mil per year in a borated fuel pool water environment.' Herefore, the electrochemical corrosion is negligible and is not a potential aging mechanism for the stainless steel liner.
l He stainless steel liner in the refueling pool is not a load-bearing stnictural component. He induced strains in the liner, resulting from conformation to deformation of the concrete wall of the pool, are negligible under normal plant operating conditions. The liner is not exposed N. /
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to corrosive environmental conditions under normal operstmg conditions, nerefore, the !
conditions for SCC to occur do not exist for the stainless steel Laer in the refueling pool.
ne heat-affected zoles of welds at the stainless steel liner are potential sites for i
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" sensitization." Sensitized Type 304 stainless steel is susceptible to IGSCC in boric acid solution.' Degradation of the stainless steel liner due to IGSCC in the pool is typically evidenced by leakage and detected by observation of an increased amount of pool water ;
leakage. !
2.2 Potential Aging Mechanism Determination Corrosion is a potential aging mechanism for the following structural components of containment because conditions exist that are conducive to corrosion of liner plates, as L discussed in Section 1.0: ,
Containmentliner LR function LR-S-3 (internalsurface)
- Ba:,emat liner LR function LR-S-3 (externalsurface) !
Refueling pool LR function LR-S-3 :
(stainless steelliner) _ . _
where: >
LR-S-3: Serves as a pressure boundary (including fluid retaining boundary) or ,
fission product retention barrier to protect public health and safety in the event of any postulated DBEs.
2.3 Impact on Intended Functions L If the effects of corrosion ofliner were not considered in the original design or are allowed to ,
degrade the above structural components unmitigated for an extended period of time, this I
aging mechanism could affect the intended function of components listed in Section 2.2.
2.4 Design and Construction Considerations ne containment liner and the basemat liner were not designed as load bearing stn etural components. Deir primary function is to maintain a leaktight barrier to minimize the telease of radioactive nuclides to the atmosphere in the event of a postulated DBE inside the containment. They were designed to conform to the deformation of the containment concrete. Under normal operating conditions, the strain imposed on the liner plate is much l
less than the yield strain of the liner plate. For corrosion protection, the inside face of the containment liner was covered with a protective coating during the construction stage.
l ne welds for each section of the basemat liner plate were covered with test channels, and a test pipe was provided for each continuous segment of the leak chase channels. He tops or
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( ^; Corrosion of Liner i v the pipes are I to 1.5 inches above the top of the concrete base slab and are sealed with caps.
Rese pipes were initially used to test the leaktightness of the basemat liner, and they can also be used to detect the leakage of the basemat liner, ne n: fueling pool liner was not designed to carry any design loads and was designed as a leaktight barrier.' Under normal operatmg conditions, the imposed strain on the liner due to conforming to concrete deformation is very small and is negligible because the stresses in refueling pool concrete components are minimal.
2.5 Plausibility Determination i
Based on the discussion in Sections 2.1.1 and 2.3, corrosion is a plausible aging mechanism for the liner of the containment dome and wall.
He inside surface of the basemat liner is covered with an 18-inch-thick concrete slab, and the containment atmosphere is not corrosive. Herefore, corrosion of this surface of the basemat liner is not a plausible aging mechanism. Although the bottom surface of the basemat liner is covered with at least 7 feet of concrete mat, it does have the potential to be exposed to underground water. Since the quality of underground water is unknown, corrosion of the bottom surface of the basemat liner is considered a plausible aging mechanism until it is shown that the groundwater will not degrade the basemat concrete or embedded rebar.
io\
Based on the discussion in Section 2.13, corrosion in sensitized zones of the refueling pool _ _.. ____
liner due to IGSCC is a plausible aging mechanism.
2.6 Existing Programs ne inside surface of the cohtainment liner is covered by a protective coating. Accessible portions of this liner are currently maintained under CCNPP Procedure MN-3-100' and QIe ,
2-100' . For inaccessible portions, an age-related degradation inspection is needed to manage l the effects ofcorrosion.
Corrosion of the refueling pool liner is managed by a System Summary and Improvement Plan (PEG-19') performance indicator which measures arid trends structural leakage from the refueling pool any time the pool is filled.
There are no existing programs at CCNPP that are designed specifically to identify conosion of the exterior surfaces of the basemat liner.
3.0 CONCLUSION
ne exposed surface of the containment liner is susceptible to degradation due to j
electrochemical corrosion. Therefore, conosion of the containment liner is a plausible aging mechanism. The exposed surface of the liner is covered by a protective coating that is currently maintained under CCNPP procedure MN-3-100. An age-related degradation inspection will be used to address corrosion of areas of the containment liner that are not
/'s normally accessible.
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Corrosion of Liner ne possibility of conesion also exists for the bottom surface of the basemat liner due to the potential exposure to underground water leaking through the constructionjoints/ cracks of the concrete basanat undemeath the liner, herefore, corrosion is a plausible aging mechanism for the h=ane liner but this aging can be man 6ged by sampling the quality of the ground water and > - d-5 that the groundwater will not cause detenoration of the basemat liner should they come into contact.
l For the stainless steel liner of the refueling pool, degradation due to IGSCC of heat-affected l zones of welds could cause the liner to leak, herefore, IGSCC of the stainless steel liner is a
- l. plausible aging mechanism for the refueling pool liner. Because the liner only serves as a j fluid retaining boundary and does not provide a structural integnty function, measunng and i trending leakage from the refueling pool (when filled) is an effective aging management technique.
4.0 RECOMMENDATION
- Based on the above discussion, the following recommendations are made:
Dunng initial plant constraction, groundwater observation wells were installed to monitor the fluctuation of the yvw4 water table, and samples were taken for yv.4-e. quality testmg. Although the wells are still in place, the monitoring activities have been discontinued. It is recommended that the groundwater weer l
! quality be tested using these wells. .This data can be used to evaluate the effects of. _ _ _.
chemical attacks on the below-gradepution of the containment structure due to l aggressive groundwater exposure A age-related degradation inspection should be used to verify the condition of the containment liner in areas that are not readily accessible.
5.0 REFERENCES
- 1. "Pressunzed Water Reactor Containment Structures License Renewal Industry Report," NUMARC Report 90-0l, Revision 1, September 1991. .
l 2. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, i December 1991.
L f 3. " Specification for Fumishing, Fabricating, Delivering and Erection of the Containment Structure Liner Plate and Accessory Steel," CCNPP's Design l
Specification No. 6750-C-16, Revision 8, May 1971.
- 4. " Specification for Stainless Steel Liner Plate and Spent Fuel Pool Bulkhead Gate,"
CCNPP's Design Specification No. 6750-C-28, Revision 5, Juz: 1973.
- 5. " Safety Evaluation Report Related to the Operation of Comanche Peak Steam Electric Station, Units 1 and 2," NUREG-0797, July 1981.
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- 6. " Painting and Other Protect e Coatings", CCNPPs Procedure MN-3-100*
Revision 2
- 7. " Issue Reporting and Assessment", Calvert Cliffs Nuclear Power Plant Administrative Procedure QI 2-100, Revision 4. Date 1/2/96 8 " System Summary and Improvement Plans", Plant Engineering Section Guideline PEG-19, Revision 0 l
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(n) v APPENDIX M - CORROSION OF TENDONS l
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1.0 MECHANISM DESCRIPTION' When corrosion of prestressing tendons occurs, it is generally in the form of localized corrosion. Most conosion-related failures of prestressing tendons have been attributed to pitting, stress corrosion, hydrogen embrittlement, or some combination of these.
Pitting is a highly localized form of corrosion. The primary parameter affecting its occunence and rate is the environment surrounding the metal. The presence of halide ions, particularly chloride ions, is associated with pitting conosion.
Stress corrosion resuhs from the simultaneous presence of a conducive environment, a susceptible material, and tensile stress. He environmental factors known to contribute to stress conosion cracking (SCC) in carbon steels are hydrogen sulfide, ammonia, nitrate solutions, and seawater. Prestressing tendon anchor heads, which are constmcted of a high strength, low alloy steel bolting material, are vulnerable to SCC.
Hydrogen embrittlement (technically, not a form of corrosion) occurs when hydrogen atoms, produced by corrosion or excessive cathodic protection potential, enter the metal latice.
Ilydrogen produced by corrosion is not usually sufficient to result in hydrogen embrittlement of carbon steel. Cathodic polarization is the usual method by which tids hydrogen is n produced. The interaction between the dissolved hydrogen atoms and the metal atoms results in a loss of ductility manifested as brittle fracture.
Conosion of prestressing wires causes cracking or a reduction in the wires' cross-sectional area. In either case, the prestressing forces applied to the concrete are reduced. If the prestressing forces are reduced below the design level, a reduction in design margin results.
2.0 EVALUATION 2.1 Conditions The potential for corrosion of tendons . a considered in the initial design of the prestressed .
tendon system of CCNPP's contairaent structure, nerefore, a petroleum-based grease product VISCONORUST 2090P' was used in the tendon sheathing to protect the tendon.2 Based on 1,3 ,5 , and 10-year tendon surveillances of Unit I containment, a few minor rust spots on the tendon bearing plates and a pitting spot on one prestressing wire were noted. The rust spots on the tendon bearing plates were evaluated and determined to have no effect on the plates' performance. Grease samples taken during each surveillance were laboratory tested, and all met the acceptance criteria specified in the surveillance procedure. The wire with l
pitting was removed for laboratory testing which showed that the wire met the original specification requirements.
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l 2.2 Potential Aging Mechanism Determination l Corrosion is a potential aging mechanism for the post-tensioning system, including the % inch diameter prestressing wires, the anchor heads, the shims, and the bearing plates, because they could be exposed to corrosive environment as described in Section 1.0.
Post-tensioning system LR-S-1 thmugh 4 where:
LR-S-1: Provides structural and/or functional support to safety-related equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
LR-S-3: Serves as a pressure boundary or fission product retention barrier to l protect public health and safety in the event of any postulated DBEs.
I LR-S-4: Serves as a missile barrier (internal or external).
2.3 Impact on Intended Functions
,m If the effects of corrosion are allowed to degrade the above structural component unmitigated (V) for an extended period of time, this aging mechanism could affect intended function LR-S-3 of -
the post-tensioning system. Intended functions LR-S-1,2, and 4 will n'ot be affect'ed because these functions will be served by the containment wall.
2.4 Design and Construction Considerations The Calvert Cliffs post-tensioning system is a BBRV system furnished by Prcxon Corporation.2 There are a total of 875 tendons including 204 dome tendons, 467 hoop tendons, and 204 vertical tendons.' Each tendon consists of 90 %-inch-diameter wires (ASTM A-421-65T),2 anchor heads, and 2 sets of shims. The tendon sheathing system consists of spirally wound carbon steel tubing connecting to a trumplate (bearing plate and trumpet) at each end. The bearing plates were fabricated from steel plate conforming with ASTM A-6-66 and the trumpets from AISI C1010-C1020 material. VISCONORUST 2090P' is used in the tendon sheathing to protect the tendon from corrosion.'
The tendon system is currently monitored by a surveillance program in accordance with CCNPP's procedures STP-M-663-1/2' for meeting the inspection requirements of Regulatory Guide 1.35.
2.5 Plausibility Determination Based on the discussion in Sections 2.1 through 2.4, corrosion is a plausible aging mechanism j
for the post-tentioning system that includes the %-inch-diameter prestressing wires, anchor heads, shims, and bearing plates.
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r I Corrosion of Tendons l
2.6 Existing Programs ne tendon system in CCNPP contamment is tested and monitored by tendon surveillance program STP-M-663-1/2.' The surveillance test was performed in 1,3, and 5 years after the structural integrity test, and every 5 years thereafter.
3.0 CONCLUSION
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Corrosion of tendons was considered in the initial design of the prestressed tendon system and is periodically monitored as part of the tendon surveillance program Corrosion inspection of the tendon system is performed in accordance with CCNPP's procedures STP-M-663-1/2 for meeting the inspection requirements of Regulatory Guide 1.35.
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4.0 RECOMMENDATION 1
I The surveillance program should be continued throughout the license renewal period to l monitor the condition and performance of the tendon system a
O REFERENCES U -
5.0
- 1. " Pressurized Water Reactor Cdntainment Structures License ~ Renewal Industry
~'
~l Report," NUMARC Report 90-1, Revision 1, September 1991. j
- 2. "Calvert Cliffs Nuclear Power Plant, Units I and 2, Updated Final Safety Analysis i Report (UFSAR),". Baltimore Gas and Electric Co.
- 3. "Prestressing Report - Containment Structure, Calvert Cliffs Nuclear Power Plant, Unit 2," Baltimore Gas and Electric Company.
- 4. " Containment Tendon Surveillance," STP-M-663-1/2, Calvert Cliffs Nuclear ,
Power Plant, Units I and 2, Baltimore Gas and Electric Co. I l
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(m) v APPENDIX N - PF.ESTRESS LOSSES 1.0 MECIIANISM DESCRIPTION' As the plant ages, tendons that were prestressed during construction tend to lose tension.
Termed prestress losses, these reductions in stress are not readily observable. Several factors i contribute to prestress losses: I Stress relaxation of prestressing wires Shrinkage, creep, and clastic deformation of concrete j Anchorage seatinglosses Tendon friction Reduction in wire cross section due to corrosion With the exception of corrosion-induced wire cross-sectional loss, predictions of prestress losses were calculated during design to ensure the containment can maintain its pressure capacity under postulated DBE inside the containment.
!Q t 2.0 EVALUATION
\
2.1 Conditions The conditions and performance of the Calvert Cliffs tendon system are monitored by lift-off tests in accordance with CCNPP's surveillance test procedures STP-M-663-1/2.2 The procedures were prepared to ineet requirements in Regulatory Guide 1.35. l The prestressed tendons in CCNPP Unit I containment have been tested in 1,3 ,5 , and 10-year surveillance following the testing procedure and acceptance criteria specified in Test Procedure STP-M-663-1.2 Based on the test and investigation results, no evidence of prestress loss exceeding the technical specification requirements has been observed.
2.2 Potential Aging Mechanism Determination l l Prestress losses are a potential aging mechanism for the post-tensioning system because stress l I
l losses with age are a common phenomenon in a typical prestressed tendon.
Post-tensioning system LR-S-1 through 4 where:
LR-S-1: Provides structural and/or functional support to safety-related equipment.
LR-S-2: Provides shelter / protection to safety-related equipment.
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I L. Prestress Losses l l l LR-S-3: Serves as a pressure boundary or fission product retention barrier to protect public health and safety in the event of any postulated DBEs. i LR-S-4: Serves as a missile bamer (intemal or extemal).
2.3 Impact on Intended Fanetions 1
If prestress losses are allowed to degrade the above structural component unmitigated for an extended period of time, this aging whanism could affect intended function LR-S-3 of the post-tensioning system. Intended functions LR-S 1,2, and 4 will not be affected because these .
i - functions will be served by the contamment wall. !
2.4 Design and Construction Considerations l l
De Calvert Cliffs post-tensioning system is a BBRV system fumished by Prescon Corporation. Dere are a total of 875 tendons including 204 donw tendons,467 hoop tendons, 8
and 204 vertical tendons Each tendon consists of 90 %-inch <hameter wires (ASTM A-421- l 65T),2 anchor heads, and 2 sets of shims. De tendon sheathmg system consists of spirally I
- wcund carbon steel tubing enaaming to a trumplate (beanng plate and trumpet) at each end.
I ne bearing plates were fabncated from steel plate conforming with ASTM A-6-66 and the trumpets from AISI C1010-C1020 material. VISCONORUST 2090P' is used in the tendon
[. sheathing to protect the tendon from co Tosion.'
In accordance with Regulatory Guide 135, liA-off tests of tendons are performed penodically.
Prestress force measurements from the liA-off' test are w...pi M with the tmet prediction of prestress losses which are calculated and considered in the design of prestressed l concrete contamments. However, the existing prediction of prestress losses considered a service life of 40 years.s herefore, the prediction must be updated to reflect the proposed
- term oflicense renewal and to provide the basis for liA-off test comparisons during the license-renewal period.
l 2.5 Plausibility Determination Based on the discussion in Sections 2.1 through 2.4, prestress loss is a plausible aging !
mechanism for the prestressing wires of the post-tensioning system. l 2.6 Existing Programs The tendon system in CCNPP containment is tested and monitored by tendon surveillance program STP-M-663-1/2.8 He surveillance test was performed in 1,3, and 5 years aAer the structural integrity test, and every 5 years thereaRer.
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' Prestress Losses i
3.0 CONCLUSION
Prestress losses in tendons were considered in the initial design of the prestress tendon system and are periodically monitored as pan of the tendon surveillance program during the current
, license priod. Prestressing force in tendons is monitored by CCNPP's procedures STP-M-
) 663-1/2 for meetmg the lift-off force requirements of Regulatory Guide 1.35. He current prestress losses were predicted for 40-year service life. He tendon liRoff forces measured in each of the prior surveillance tests all met the requirements specified in Section 3.6 of the CCNPP Technical Specification.8 4.0 RECOMMENDATION The current prestress losses that were specifically predicted for a service life of 40 years should be updated to reflect the anticipated prestress losses during the license renewal period.
Retensioning of tendons, if determined to be necessary based on tendon surveillance data and the updated prestress losses, should be performed.
5.0 REFERENCES
V 1.
Pressurized Water Reactor Containment Structures License Renewal Industry ~ ~ ~ " ~ ~ ~~~
i Report," NUMARC Repon 90-1, Rev.ision'1, September 1991. "~
- 2. " Containment Tendon Surveillance," STP-M-663-lC, Calven Cliffs Nuclear Power Plant, Units I and 2, Baltimore Gas and Electric Co.
_ . . . . . . ..____.._.3_ ._ . "Pmstmssg Repnri, ContainmentSrmeturr CalvertClifik.Nmlearlower.. Plack . .. . . . _ ..q Unit 2," Baltimore Gas and Electric Co. ;
1
- 4. "Calvert Cliffs Nuclear Power Plant, Units I and 2, Updated Final Safety Analysis Report (UFSAR)," Baltimore Gas and Electric Co.
- 5. "Calven Chffs Nuclear Power Plant, Units 1 and 2, Technical Specification,"
Baltimore Gas and Electric Co.
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O APPENDlX O - WEATHERING 1.0 MECHANISM DESCRIPTION 1.1 Concrete Components and structures that are located in an environment that is exposed to ambient conditions are susceptible to degradation due to weathering. Aging mechanisms associated witt. weathering include exposure to sunlight (ultraviolet exposure), changes in humidity, ozone cycles, temperature and pressure fluctuations, and snow, rain, or ice. The effects of weathering on most materials are evidenced by a decrease in elasticity, an increase in hardness, and shnnkage.
1.2 Marinite XL Board The Marinite XL board is fabricated from calcium silicate with inert fillers and reinforcing agents. Weathering of Marinite XL board material may occur due to exposure to temperature and humidity changes.
2.0 EVALUATION 2.1 Conditions According to Specification ASTM C33-82," Standard Specification forConcretaAggregates,".._ __ . . _ _ .
' the CCNPP site is located in the geegraphic region subject to severe weathering conditions. .
All outdoor components will experience the extreme temperature ranges, rain, snow, and changes in humidity expected at the CCNPP site.
The Marinite XL boards are located in the Containment Structure. The normal
~enviromnematconditions'(outside diepranary sideid) ate as followsh Temperature 120*F (average limit)
Humidity 70 % (max.)
Radiation 0.35E6 rads (40 years integrated gamma dose under normal operating .
conditions) 2.2 Potential Aging Mechanism Determination Degradation by weathering is a potential aging mechanism for the following Containment Structure architectural components because they are exposed to temperature and humidity fluctuations similar to outdoor conditions:
- protective coating (exterior applications) LR functions LR-S-1,2,3,4,5,7
. partitions and ceilings (Marinite XL board) LR function LR-S- 7 l
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. Weathering
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where:
LR-S-1: Providesstructuraland/orfunctionalsupporttosafety-selatedeqmpment.
LR-S-2: Provides shelter / protection to safety-related equipment.
LR-S-3: Serves as a pressure boundary or fission product retention barrier to protect {
,- public health and safety in the event of any postulated DBEs. j l i l LR-S-4: Serves as a missile barrier (intemal or external).
l l LR-S-5: Provides structural and/or functional support to non-safety-related equipment l whose failure could directly prevent ud6'4 y accomphshment of any of the required safety-related fnact=nt l
I LR-S-7: Provides rated fire barriers to confine or retard a fire from spreading to or from adjimit areas of the plant.
l' l' 2.3 _ Impact on Intended Functions l Weathering of the inside& d= t Stmeture protective coatings could result in a loss of their functions protecting structural 9+;-:--H- from any corrosive environment inside the l
j: Contamment Structure. However, coatmgs inside the Containment Structure are only safety-related because of the potential impEit of coatmg failure on the operation of the emergency sump in the event of a design basis accident inside the Contamment Structure herefore, l
degradation due to weathering effects will not pievent these coatings from performing their safety-related functions.
I
_ _ - . __._._ Weathering.could impact .the_, intended function.,o_f the. Marinite XL . board if the. aging, , , , , .,!
l mechanism caused degradation of the material such that it was not capable of performing its' fire barrier function.
!' 2.4 Design and Construction Considerations Since weathering has no impact on the intended functions of protective coating, no further discussion of CCNPP's design and construction considerations is necessary in regard to protective coating. :
I The Marinite XL board is used as a fire separation barrier on cable trays within the i
Containment Structure ne board is designed to meet Appendix R criteria and is required to maintain its integrity during normal environmental conditions within the Containment Structure in order to provide the barrier function. The Marinite XL board is used only on l 8 cable trays outside the primary shield in the Contamment Structure . The Marinite XL board is fabricated from calcium silicate with inert fillers and reinforcing agents.
As noted in Section 2.1 above, the Marinite XL boards are exposed to a normal average temperature of 120*F and to a normal maximum relative humidity of 70%. The Marinite XL board is not affected by moisture or high humidity in accordance with the j manufacturer's literature * *. The manufacturer has indicated that Marinite XL board has been used in moist-cure ovens with normal temperatures ranging between 200*F and 250*F i
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and with steam injection to maintain a wet atmosphere. After 10' years of continuous operation, no harmful effects were identified on the Marinite'.
Since the environment in the Containment Structure is basically ambient conditions for l many living and working environments, and are much less severe than a moist-cure oven, l i
no deterioration of the Marinite would be plausible under the conditions occurnng in the Containment Structure.
2.5 Plausibility Determination Based on the discussion in Sections 2.1 and 2.4, edwilig is not a plausible aging mechanism for any architectural 3-- ;-:-wts in the CCNPP Containment Structure. l 2.6 Existing Prograans Since ederlig is nm a plautible aging mechanism, no program is needed to control this degradation mechan'.sm to maintain the intended functions of the Cmtauunent Structure architectural compos ents. l i
3.0 CONCLUSION
1 Based on this evaluation, weathenng is not a plausible aging mechanism because no intended functioiIs of the Containment Structure architectural components are affect x! by this aging mechanism.
4.0 RECOMMENDATION .
l architectural components. No further action is required.
5.0 REFERENCES
- 1. " Standard Specification for Concrete Aggregates," American Society of Testing and Materials, ASTM C33-82.
- 2. Component Level Scoping Results for the Primary Containment Structure (Sys.
059).
- 3. "EQ Design Manual - Calvert Cliffs Nuclear Power Plant, Unit No. I and 2,"
Baltimore Gas and Electric Co.
- 4. Manville Products Catalog IND-33610-80 (page 3),
i
- 5. BGE Drawing 62-150-E, Rev. 6, Appendx R Separation Rqmts, Ctmt. El. 5'-0".
l
( 6. Manville Products Catalog R-34 3-87 (pages 2 and 5).
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v APPENDIX R - ELEVATED TEMPERATURE 1.0 MECHANISM DESCRIPTION' During normal plant operation, solar heat load and equipment heat loads contribute to an increase in temperature of the intemal environment of a structure. Of all structural components in a structure, only components made of concrete material are potentially affected within the temperature range in which the structure will experience during normal plant operating conditions. As a result of elevated temperature, compressive strength, tensile strength, and the modulus of elasticity of concrete could be reduced by greater than 10 percent in the temperature range of I80 to 200 *F. Long-term exposure to high temperatures (> 300
- F) may cause surface scaling and crackieg. Otherwise, there is no visible physical ni:mifestaticit of concrete degradation due to erposure to elevated temperature.
2 ASME Code ,Section III, Division 2 indicates that as long as concrete temperatures do not exceed 150 *F, aging due to elevated temperature exposure is not significant. Localized hot spots are limited in area and do not exceed 200 *F by design. ACI-349' allows local area temperatures to reach 200 *F before special provisions are required.
2.0 EVALUATION
/"N 2.1 Conditions
( )
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Section 5.2.1 of Baltimore Gas and Electric Company's EQ Manual states:
Ambient air bulk temperatures imide containment are limited to an average of120 *F byplant technicalspecification section 3/4.6.1.5. A 120 *F normal service air ambient temperature is usedfor evaluating component normal service aging unless otherwise specified Note that this wlue does not include process heating or energi ation l temperature rise efect. . specfies that the maximum normal ambient l temperaturefor the pressuri:er house is 140 *F. The primary loop RTDs which are located in the containment pump bays, experience higher temperatures, assumed to be 160 *F based on unverified temperaturemonitoring. .
In addition, Section 5.1.4.4 of BG&E's UFSAR 5states:
The main high-temperature piping consists of two penetrations for feedwater, two penetrationsfor main steam, two penetrationsfor steam generator blowdown, onefor the reactor coolant letdown line and one for the reactor coolant sampling line. These have a maximum operating temperature range between 435 *F and 537 *F. Thermal insulation is provided on the outside diameter of each line and separate coolant circulation, with instrumentation suitable for flow monitoring, is provided in the air gap betueen the insulation and the
, penetration liner sleeve. The combination ofinsulation and coolant l circulation is designed to restrict the maximum temperature rise in the
% concrete to 150 *F.
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Elevated Temperature As stated above, the inside surface of the primary shield wall is subject to ==*=iaaA intemal heat buildup, and the concrete around the eight pipe penetrations (two main steam lines, two feedwater lines, two steam generator blowdown lines, one reactor coolant letdown line, and one reactor coolant sample line) is also subject to extended local heatup ,
2.2 , Potential Aging Mechanism Determination .
Elevated temperature is a potential aging mechak u for the following concrete structural '
components of containment because they could be exposed to temperatures higher than the degradation threshold of elevated temperature for concrete (150 *F):
- Concrete containment wall LR functions LR-S-1 through 7 Primary shield wall LR functions LR-S-1,2,4 where:
LR-S-1: Provides structural and/or finv+ianal support to safety-related equipment.
LR S-2 : Provides shelter / protection to safety-related equipment.
LR-S-3 : Serves as a pressure boundary or fission product rgention barrier to.
protect public health and safety in the< vent of any postulated DBEs.
LR-S-4 : Serves as a missile barrier (internal or external).
LR-S-5 : Provides structural and/or functional support to non-safety-tated equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
LR-S-6 : Provides flood protection barrier (internal flooding event).
LR-S-7 : Provides rated fire barrier to confme or retard a fire from spreading to or from adjacent areas of the plant.
Other structural components are not exposed to temperatures higher than the degradation threshold of elevated temperature for concrete. Therefore, elevated temperature is not a potential aging mechanism for these components.
2.3 Impact on Intended Functions If the effects of elevated temperature are allowed to degrade the above structural components unmitigated for an extended period of time, this aging mechanism could affect all intended functions of components listed in Section 2.2.
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Elevated Temperature 1
2.4 Design and Construction Considerations 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 containment building is limited to 150 *F.7 The primary shield wall is subjected to sustained internal heat buildup.
However, thermal insulation and an af.r-cooling system are provided at the inner surface of the reactor cavity wall to maintain the concrete temperature at or below 150 *F."
l The concrete around the eight hot pipe penetrations (two main steam lines, two feedwater lines, two steam generator blowdown lines, one reactor coolant letdown line, and one reactor coolant sample line) is also subject to ext:nded high temperature or local heatup. However, a cooling system combining insulation and coolant circulation was implemented to restrict the maximum temperature in the concrete to 150 *F.ss.io nese temperatures are below the degradation thresholds of elevated temperature for concrete.
A higher temperature of 160 *F noted in Section 2.1 is limited to a localized area of the containment pump bays. Herefore, it will not degrade the concrete in this area.
p 2.5 Plausibility Determination Q Based on the discussion in Sections 2.1 and 2.4, no structural components are exposed to temperatures higher than the degradation threshold of elevated temperature for concrete.
Therefore, elevated temperature is not a plausible aging mechanism for any structural components of the CCNPP containment.
2.6 Existing Programs Although there is no existing program to monitor the temperature profiles for the surfaces of the eight high-temperature containment penetrations listed in Section 2.4 or the surface of the primary shield wall, the original design recognized the potential of elevated temperatures inside the primary shield wall and in the vicinity of the hot piping penetrations in the containment wall. Consequently, insulation and cooling system were installed to maintain their surface temperatures at or below 150 *F as documented in the FSAR and other design documents. Therefore, no program is needed to manage this aging mechanism.
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3.0 CONCLUSION
The prunary shield wall is subject to sustained internal heat buildup, and the concrete around the eight hot pipe penetrations is also subject to extended local heatup. However, protective measurec have been implemented during design and construction stages to mitigate the j temperature in these structural components to within 150 *F. No other structural components are exposed to elevated temperature due to a technical specification limit of maintaining the containment bulk temperature at 120 *F.
O (O
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't
- / Elevated Temperature i
l Therefore, elevated temperature is not a plausible aging mechamsm for any structural components of the containment.
4.0 RECOMMENDATION '
Elevated temperature is not a plausible aging mechanism for any structural components of the
! containment. Therefore, no further evaluation or rammma-4= tion is necessary.
l l
5.0 REFERENCES
- 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. -"Code for Concrete Reactor Vessels and Containments," ASME Boiler and Pressure Vessel Code,Section III, Division 2,1986.
- 3. " Code Requirements for Nuclear Safety Related Concrete Structures," Anwrican Concrete Institute, ACI 349-85.
[~' '4. "EQ Design Manual - Calvert Cliffs Nuclear Power Plant, Unit No. I and 2,"
Baltimore Gas and Electric Co.
. (
t i _
- 5. "Calvert Cliffs Nuclear Power Plant, Units I and 2, Updated Final Safety Analysis
- Report (UFSAR)," Baltimore Gas and Electric Co.
- 6. Technical Specification Manual, Calvent Cliffs Nuclear Power Plant, Unit i .
Ur.it 2.
. 7. Civil and Structural Design Criteria for Calven Cliffs Nuclear Power Plant, Unit No. I and 2, by Bechtel Power Corporation, Revision 0, August 2,1991.
- 8. ' " Reactor Cooling System - Calvert Cliffs Nuclear Power Plant," BG&E's Design Drawings 60-340-E,60-341-E, and 60-342-E.
- 9. " Encapsulation Details - Main Steam System, Calvert Cliffs Nuclear Power Plant, Units I and 2," BGAE Design Drawings 60-346-E Sheet 2 (Unit 1) and 62-346-E Sheet 2 (Unit 2).
- 10. " Piping and Instrument Diagram - Component Cooling System, Calvert Cliffs Nuclear Power Plant, Units 1 and 2," BG&E Design Drawings 60-235-E (Unit 1),
and 62-235-E (Unit 2).
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i A
V APPENDlX S -IRRADIATION 1.0 MECHANISM DESCRIPTIONU#3 1.1 Concrete Concrete components in a nuclear pIwer plant exposed to excessive neutron or gamma radiation (incident flux > 10 MeV/cm'-sec)I could be impaired due to aggregate growth, decomposition of water or thermal warming of concrete. As the temperature of concrete increases and free water within the concrete evaporates, the structural characteristics of concrete are degraded. With the water loss, concrete can experience a decrease in its compressive, tensile, and bonding strengths, and in its modulus of elasticity. However, this loss of free water which results in a small decrease in concrete density will have little effect on concrete's gamma attenuation propenies unless water loss is significant, depleting the presence of hydrogen atoms which contribute to concrete's shielding characteristics of fast neutrons. Typically, gamma radiation affects the cement paste portion of the concrete, producing heat and causing water migration.
Existing experimental data provide some general information on the impact of direct radiation on the mechanical properties of concreteHl. The average concrete sample does not begin to experience a compressive or tensile strength loss until exposure exceeds a neutron fluence of 10 neutrons /cm'.
The experimental dataHIindicate minimal compressive loss for exposure up to 5x10 neutrons /cm'.
1.2 Reinforcing Steel, Structural Steel, and Liner Steel degradation due to neutron irradiation is caused by the displacement of atoms from their normal lattice positions to form both interstices and Vacancies. The effect of this mechanism is to increase the yield strength, decrease the ultimate tensile ductility, and increase the ductile-to-brittle transition temperature. These defects on a macroscopic level produce what is referred to as radiation-induced embrittlement, which is encountered in the design and operation of reactor pressure vessels. By comparing the currently available stress-strain curves for unirradiated and irradiated mild steel, a reduction in ductility of rebar subjected to high radiation exposure (> 10 neutrons /cm') is indicated.[5]
1.3 Tendon l
~
The effects of irradiation on prestressing wires in tendons are the same as those described for reinforcing steel with regard to the effects on yield strength and the modulus of elasticity. For !
prestressing wires, radiation exposure will cause a decrease in the expected relaxation. The grease used in the tendon sheaths loses viscosity due to gamma radiation.D3 l 1.4 Maranite XL Board The Maranite XL board is fabricated from calcium silicate with inert fillers and reinforcing agents.I'l Silicates as a class of material are capable of absorbing up to lx10" Rads before any significant damage occurs.
(s l
i l
! I I
s irradiation I i t
i 1
2.0 EVALUATION J 1
2.1 Conditions :
l Several in scope structural components are in high radiation areas inside the containment, in particular l the primary shield wall which is subject to high neutron dose. Some components receive a higher i gamma dose than neutron dose. As noted in Section 1.0, the radiation degradation thresholds for each j constituent are: :
i Concrete 10" neutrons /cm' l I Steel > 10" neutrons /cm' s
Tendon Wires . 4x10 neutrons /cm'I'I ,
Tendon Grease Ix10 rad l 1
Maranite XL board Ix10" rad 2.2 Potential Aging Mechanism Determination l l
l.
O Irradiation is a potential aging mechanism for the following structuml and architectural components of f - the containment. .
. coating )
i
. concrete basemat
. pr mary shield wall l
. secondary shield wall
. partitions and ceilings (Maranite XL board) 2.3 Impact on Intended Functions if the effects of irradiation are allowed to degrade the above structural and architectural components
, unmitigated for an extended period of time, this aging mechanism could affect all their intended
! functions.
2.4 Design and Construction Consideration BG&E's EQ Design Manualt71specifies the following 40-year normal service doses for use in environmental qualMcation evaluations:
General areas 0.35x10' rads l 1
' 3.5x10' rads Containment pump bays I
i 5/7/96 S-2 Revision 3 i
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(m) Irradiation i q.J \
Inside primary shield wall 3.5x10' rads 2.5x10 neutron /cm2 Based on the above, the allowable 60-year service radiation doses are:
Generalareas 0.53x10' rads Containment pump bays 5.25x10' rads Inside primary shield wall 5.25x10' rads 3.75x10 neutron /cm' As indicated above, the allowable 60-year radiation doses of neutron and gamma radiation incurred by the structural and architectural components are less than the irradiation degradation threshold for each constituent of all structural and architectural components. Therefore, irradiation is not a plausible age-related degradation mechanism for any structural or architectural component of the containment.
2.5 Plausibility Determination Based on the discussion in Sections 2.1 and 2.4, no structural or architectural components of (m)
'd containment are exposed to radiation higher than their design threshold. Herefore, irradiation is not a plausible aging mechanism for any structural or architectural components of the.CCNPP containment.
2.6 Existing Programs here are no existing programs at CCNPP designed to identify damages to structural or architectural components of the containment due'to radiation. Ilowever, since this is not a plausible aging mechanism that could degrade these components, no future program is necessary.
3.0 CONCLUSION
All structural and architectural components in the containment are exposed to neutron and gamma radiation. As indicated in Section 2.0 above, the neutron fluence level and the maximum integrated gamma does inside the containment are predicted to be below the degradation threshold for each constituent of all structural and architectural components. Derefore, irradiation is not a plausible age-related degradation mechanism for the structural or architectural components of the containment. ,
i RECOMMENDATIONS 4.0
)
i Irradiation is not a plausible aging mechanism for the concrete structural or architectural components in containment. No further evaluation or recommendation is required.
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S/7/96 S-3 Revision 3 i
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l Irradiation l
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5.0 REFERENCES
i 1. " Class I Structures License Renewal Industry Report," EPRI's Project RP-2643-27, l December 1991.
- 2. " Pressurized Water Reactor Containment Structures License Renewal Industry l I Report," NUMARC Report 90-1, Revision 1, September,1991.
l t
- 3. " Guidelines on the Nuclear Analysis and Design of Concrete Radiation Shielding for !
I Nuclear Power Plants", American Nuclear Standard ANSI /ANS-6.4
- 4. Hilsdorf, H.R., Kropp, J., and Koch, II.J., "1he Effects of Nuclear Radiation on the ;
Mechanical Pmperties of Concrete," Douglas McHenry International Symposium on ,
I Concrete and Concrete Structures, American Concrete Institute Publication SP-55, l
1978
- 5. Naus, DJ., " Concrete Component Aging and its Significance Relative to Life Extension of Nuclear Power Plants," NUREG/CR-4652, ORNUTM-10059, Oak Ridge National Laboratory, Oak Ridge, Tenn., September 1986 j
- 6. " Code Requirements for Nuclear Safety Related Concrete Structures," ACI 349-85, American Concrete Institute, Detroit, Michigan ,
- 7. "EQ Design Manual - Calvert CYiffs-Nuclear Power Plant, Unit No. I and 2,"
Baltimore Gas and Electric Co. ;
l
- 8. Manville Products Catalog IND-33610-80 (page 3).
O l l
l 5/7/96 S-4 Revision 3
m (v) APPENDlX T- FATIGUE 1.0 MECHANISM DESCRIPTION' Fatigue is a common degradation of structural members produced by periodic or cyclic loadings that are less than the maximum allowable static loading. Fatigue results in progressive, localized damages to structural materials.
Two types of fatigue exist for structural components. He first mechanism, sometimes referred to as low-cycle fatigue, is low frequency (<100 cycles for concrete structures and <1 x 10' for steel structures) of high-level repeated loads due to abnormal events such as SSE or strong winds. Structures exposed to such events must be thoroughly evaluated by analysis or by inspection or both after occurrence. De fatigue degradation caused by such loading may not occur or may occur only a few times during the service life of a structure. Herefore, low-cycle fatigue is not age-related and is not a license renewal issue, ne other fatigue mechanism is high frequency of low-level, repeated loads such as equipment vibration. Referred to as high-cycle fatigue, it is an age-related degradation mechanism.
1.1 Concrete' l
De fatigue strength of concrete structures Las become a concem due to the widespread l
[] adoption of ultimate strength design procedures and the use of high-strength materials that require concrete structural members to perform satisfactorily under high-stress levels.
! L/
Repeated loading causes cracking in~ component materials of a member and alters its statif"-' ' ~
load-carrying characteristics.
Fatigue strength of plain concrete is essentially the same whether the mode of loading is tension, compression, or flexure, ne stress-to-fatigue life relationship can be represented by an S-N curve as shown in Figure T-1, where S represents the maximum stress in the cycle and N represents the number of cycles required to produce failure. A series of specimen testing determines fatigue behavior, and the results are plotted on a log-scale. At a given number of service cycles (N) the material has a defined allowable fatigue strength. Review of S-N curves of plain concrete beams in ACI report 215R-74' indicates the following:
Fatigue strength ofconcrete decreases with the increasing number ofcycles. The S-N curvesfor concrete are approximately linear beturen 10' and 10' cycles.
This indicates that there is no limiting value ofstress below which thefatigue hfe willbe infinite.
A decrease ofthe range between marimum and minimum load results in increased l
l fatigue strengthfor a given number ofcycles. When the minimum and maximum l loads are equal, the strength ofthe specimen corresponds to the static strength of coner ete determined under normal test conditiom.
l Thefatigue strength ofplain concretefor a hfe of 10 million cyclesfor tension, compression, orflexure is roughly about $$ percent ofits static strength.
Fatigue fracture of concrete is characterized by considerably larger strains and cracking as b compared with fracture of concrete under static loading.
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Smgg a T-1 Revision 3 i
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l.
vp Fatigue 1
Fatigue failure of reinforcing steel has not been as signifwant a factor in its application as for reinforcement in concrete stmeture. There have been few documented cases of reinforcing .
i !
I fatigue failures in the concrete industry. ACI repost 215R 74' notes that the lowest stress l- range known to have caused a fatigue failure of a straight hot-rolled deformed bar embedded
! in a concrete beam is 21 ksi. This failure occurred aRer 1.25x10' cycles of loading on a concrete beam containing a No.11, Grade 60 rebar, when the minimum stress level was 17.5 ksi.
1.2 Steel' I
Fatigue of steel structures may cause progressive degradation and is initiated by plasti:
deformation within a localized region of the structure. A nonuniform distribution of stresses l through a cross-sectiou may cause a stress level to exceed the yield point within a small area and cause plastic movement after the number of stress reversal cycles reaches the material's endurance limit. This is the maximum stress to which the steel can be subjected for a given service life. Such conditions will eventually produce a minute crack. The localized plastic movement further aggravates the nonuniform stress distribution, and further plastic movement causes the crack to grow.
L I
The fatigue behavior of steel structures strongly depends on their surface conditions (e.g.,
whether they are polished or in an as-received condition). The fatigue strength of structural steel components is generally represented by a modified Goodman diagram as shown in. _
._j Figures T-2 and T-3, which is generated from the S-N curves. The fatigue strength of structural steel decreases as the number of cycles increases until the fatigue limit is reached.
If the maximum stress does not exceed the fatigue limit, an unlimited number of stress cycles 1 can be applied at that stress ratio without causing failure. I 2.0 EVALUATION i
2.1 Conditions Some of the internal structural components of containment are subject to high cycle, low-l level repeated load, such as equipment vibration load, during normal plant operation. Ihe containment wall and containment dome were designed for abnormal events such as seismic and hurricane loads that are regarded as low cyclic load condition. Such loads may not occur or may occur for a very shoit duration only a few times during the service life of the containment. Therefore, the fatigue damage of the containment dome and wall is not age- 1 related.
l 5/7/96 a T-2 Revision 3
.- - _ ~ . . . . - . - - . - - . - - .- - -. ~ . . - . - - - . . . .-. ... -- - ..-
P l
O v
Fatigue !
2.2 Potential Aging Mechanism Determination ,
- Fatigue is a potential aging mechanism for the following structural components of the containment because they could exponence high frequency oflow-level, repeated loads such -
as equipment viharionload:
Concrete columns LR functions LR-S-1,5 Concrete beams LR functions LR-S-1,5 Ground slab and equipment pads LR functions LR-S-1,5 Elevated floor slab LR functions LR-S-1,5 Primary shield wall LR functions LR-S-1,5 Secondary shield walls LR functions LR-S-1,5 Steelcolumns LR functions LR-S-1,5 Steelbeams LR functions LR-S-1,5 L- . . _ , . . . . . . . . _.
Crane girder LR function LR-S-5 Base plates LR functions LR-S-1,5 Floor framing LR functions LR-S-1,5 Steelbracings LR functions LR-S-1,5 Platform hangers LR functions LR-S-1,5 Decking LR functions LR-S-1,5 where:
LR-S-1: Provides structural and/or functional support to safety-related equipment.
LR-S-5: Provides structural and/or functional support to non-safety-related equipment whose failure could directly prevent satisfactory accomplishment of any of the required safety-related functions.
Fatigue is not a potential degradation mechanism for other containment structural components l because they are not subject to the high frequency oflow level, repeated loads.
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S/7/96 a T-3 Revision 3
_ _ _ . _ . _ _ . . _ _ _ _ _ . _ _ . __. . - _ _ . . . ._ _____m _ .
l
- N Fatigue 1
- 23 Impact on Intended Functions
- If the effects of fatigue were not considered in the original design or are allowed to degrade the above stmetural components unmitigated for an extended penod of time, this aging mechanism could affect all intended functions of components listed in Section 2.2.
2.4 Design and Construction Considerations All internal concrete components of the CCNPP containment were designed in accordance
' with ACI-318-63,3d he design code' limited the maximum permissible design stress level to l 4 less than 50 pen:ent of static strength, which is less than the fatigue strength of concrete (55 -l 1
percent of static strength). In addition, actual concrete stresses mduced by cyclic loads during l
normal plant operation, such as those from machine vibration, are a small portion of the combined stresses resulting from static and dynamic loads. His means that the stress range !
i (magnitude of stress fluctuation) is also small and within the limit that yields extremely long i fatigue life (> 10' cycles, which is equivalent to infinite life), as shown in Figure T-1. !
4 All structural steel components in the containment were designed in accordance with American Institute of Steel Constmetion (AISC-1963) specification. # For the design of steel members and connections subject to repeated variation oflive load stress, this specification' requires that consideration be given to the number of stress cycles, the expected range of l -
stress, and the type and location of a member or detail.' For life cycks of more than 2x10' !
loading, the maximum stress may.not. exceed two-thirds of the basic allowable stress provided . I L '
i in Sections 1.5 and 1.6 of the AISC specification,' which is equivalent to 40 percent of the material yield strength.
ASTM A 36 carbon steel _is typically used for all structural steel components in the l~
l containment.' As shown in the fatigue strength curves in Figures T 2 and T 3, the fatigue limit for as-received A-36 steel is about 20 ksi at a life cycle of approximately 2x10', which is about 55 percent of the material yield strength. De maximum design stresses of all steel components were limited to 40 percent of material yield strength and are less than the material fatigue limit. Again, the actual steel stresses induced by cyclic loads are small
' portion of the combined stresses resulting from static and dynamic loads. .
2.5 Plausibility Determination
' Based on the discussion in Section 2.4, fatigue will not degrade the structural components listed in Section 2.2. Therefore, fatigue is not a plausible aging mechanism for any structural components of the containment, i
2.6 Existing Programs !
Here are no existing programs at CCNPP that are designed specifically to identify or to repair the damage to structural steel components due to fatigue. Since fatigue is not a plausible aging mechanism that could degrade the containment structural components, no
! management program is necessary.
4 4 f 4
a T-4 Revision 3 5/7/96
O Fatigue Q
3.0 - CONCLUSION ,
Some concrete components in the containment of CCNPP are subject to high cycles oflow-' i level repeated load. nese components were designed in accordance with ACI 318-63',
which limits the maximum design stress to less than 50 percent of the static stress of the concrete. He concrete fati extremely high cycles (>10'gue cycles) of loadmgstength Derefore, is about fatigue 55degrade will not percentanyof its static s concrete components in the containment and requires no further evaluation.
Steel components in the containment subject to high-cycle (>10' cycles) loading conditions .
were designed in accordance with the AISC-63 specification.' he maximum stress in steel components and connections is smaller than the fatigue limit of steel. Fatigue degradation will have no adverse effects on the continued safety function performance during the license renewal term and requires no further evaluation for all structural steel components in the containment.
4.0 RECOMMENDATION Fatigue is not a plausible aging mechanism for the structural components in the containment ;
building. Herefore, no further evaluation or recommendation is necessary.
I b
5.0 REFERENCES
- 1. " Class I Structures License Renewa Industry Report," EPRI's Project RP-2643-27, December 1991.
- 2. " Consideration for Design of Concrete Stnictures Subjected to Fatigue Loading,"
American Concrete Institute, ACI 215R-74,1986. .
- 3. " Building Code Requirements for Reinforced Concrete," American Concrete Institute, ACI 318-63.
- 4. Civil and Structural Design Criteria for Calvert Cliffs Nuclear Power Plant, Unit l
No. I and 2, by Bechtel Power Corporation, Revision 0, August 2,1991.
- 5. " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," American Institute of Steel Construction,1%3.
- 6. Brockengrough, R.L., and Johnson, B.G., Steel Design Manual, United States Steel Corporation.
4 i
5/7/96 E T-5 Revision 3
Fatigue
)
1.0 ' ' ' ' ' ' -
m
[s in$ man *0'I0 e% % P=80%
0.8 - N s Ne. P=50%(avg.)b s% % *% %
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N % %
0.6 -
S
= 0.15 ,7.f, %s ~
x Smax Probobility fr of Failure 0.4 -
O.2 -
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g f f f f I o io 102 105 104 105 10' 107 Cycles io Failure,N l l
Fatigue Strength of Plain Concrete Beams (Source: Reference 2) 5/7/96 Figure T-1 m T-6 Revision 3
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Fatigue Strength of As-Received A36 Structural Carbon Steel (Source: Reference 6) 5/7/96 Figure T-2 e T-7 Revision 3
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USS "T.1" STSEL i YatLD STRANSTN R e .1 180' =RmD R*1 3
- 1 N'
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Sea . 633 TRI.T8N STREL
.. YsELD POINT i ,
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Y15LD POINT i
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.S m.la . . > - . . ., . .
100 80 SO 40 30 0 30 40 60 80 100
- STRESSES ISI. hst
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Fatigue Strength of Transversely' Groove-Welded Structural Steel Plates at 2X10 Stress Cycles l (Source: Reference 6)
O S/7/96 Figure T-3 E T-8 Ravision 3
_