ML20211L293
| ML20211L293 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1997 |
| From: | Kuo P, Samson Lee, Liu W Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1611, NUDOCS 9710100155 | |
| Download: ML20211L293 (63) | |
Text
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l Aging Management of Nuclear l Power Plant Containments for i
i License Renewal i
i j
l D. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation W. C. Liu, P. T Kuo, S. S. Lee 6l hfc7
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AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most oocuments cited in NRC publications will be available from one of the following sources:
1.
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l
NUREG-1611 Aging Management of Nuclear Power Plant Containments for License Renewal e
Manuscript Completed: September 1997 Date Published: September 1997 W. C. Liu, P T. Kuo, S. S. Lee Division of Reactor Program Management Omce of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 p
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ABSTRACT The Nuclear Regulatory Commission (NRC) published its license renewal rule, I
Title 10 of the Code of Federal Regulations (10 CFR) Part 54, on May 8,1995, providing the requirements for renewal-of operating licenses for nuclear power pl ants.
10 CFR 54.21(a)(1)(1) requires an aging management review of containment structures to ensure-that the effects of aging will be managed so that their intended functions will be maintained for the period of extended o)eration. In 1990, the Nuclear Management and Resources Council (NUMARC), now t1e Nuclear Energy Institute (NEI), submitted for NRC review, the industry reports (irs), NUMARC Report 90-01 and NUMARC Report 90-10, addressing aging management issues associated with PWR containments and BWR containments for license renewal, respectively.
-Recently, th9 Commission amended 10 CFR 50.55a to 3romulgate requirements for inservice inspection of containment structures.
Tie final rule on 550.55a,
" Codes and Standards for Nuclear Power Plants; Subsection IWE and Subsection IWL," was published in August 1996. This rule incorporates by reference the 1992 Edition with the 1992 Addenda of Subsectiens IWE and IWL of Section XI, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and l
Pressure Vessel Code addressing the inservice inspection of metal containments / liners and concrete containments, respectively.
The purpose of this report is to reconcile the technical information and agreements resulting from the NUMARC IR reviews and the-inservice inspection requirements of Subsections IWE and IWL as promulgated in 550.55a for license renewal consideration. This report concludes that Subsections IWE and IWL of Section XI, Division-1, of the ASME Code as endorsed in 550.55a are generally consistent with the technical information and agreements reached during the IR reviews. Specific exceptions are identified and additional evaluations and augmented inspection activities for renewal are recommended.
iii NUREG-1611
-=
CONTENTS Page i
ABSTRACT...............................................................
iii ABBREVIATIONS.......................................................... vil-LIST OF TABLES.........................................................
ix 1.
INTRODUCTION.......................................................
I 2.
LICENSE RENEWAL EVALUATION OF AGING MANAGEMEN STRUCTURES..................................T OF CONTAINMENT I
3.
CONCLUSIONS........................................................
2
{
REFERENCES.............................................................
4 TABLES a
?
l.
AGING MANAGEMENT OF PWR CONTAINMENTS FOR LICENSE RENEWAL...........
7 2.
AGING MANAGEMENT OF BWR CONTAINMENTS FOR LICENSE RENEWAL...........
27 APPEMDICES A
IMPLEMENTATION HIGHLIGHTS OF SUBSECTIONS IWE AND 10 CFR 50.55a...................................IWL THROUGH 47 B
LIST OF PWR CONTAINMENT COMPONENTS.................................
49 C
LIST OF BWR CONTAINMENT COMP 0NENTS..................................
50 v
A88REVIATIONS-AC I'-
'American Concrete Institute AISC-American Institute of Steel Construction iARDM -
Age-Related Degradation Mechanism-
'ASME-
-American Society of Mechanical Engineers ASTM
- American Society for Testing and Materials-BWR'
-Boiling Water Reactor CS-Carbon Steel CFRJ Code of Federal. Regulations
-CRD-Control Rod Drive ECCS Emergency Core Cooling System GSI Generic _ Safety Issues
'IR
-Industry Report ISI:
_ Inservice Inspection IWE-Subsection of ASME Code,Section XI, " Rules for Inservice -Inspection of Nuclear Power Plant Components," containing " Requirements for Class MC and Metallic Liners of Class-CC Components of Light-Water Cooled Plants" 11WF Subsection'of ASME Code,-Section XI, " Rules for Inservice _ Inspection of Nuclear Power' Plant Components," containing " Requirements for:
Clr2s 1,_2, 5, and MC Component: Supports of Light-Water Cooled Plants"-
IWL.
Subsection of ASME Code, SectionxXI,J" Rules for Inservice; Inspection of Nuclear Power Plant Components," containing _" Requirements for-Class CC Concrete Components of Light-Water Cooled-Plants" NEI Nuclear. Energy Institute NRC'
-Nuclear Regulatory Commission NUMARC Nuclear Management and Resources Council ppm-Parts per million PWR-Pressurized Water Reactor RG
. Regulatory Guide SS' Stainless Steel
L LIST OF TABLES ix NUREG-1611
-TABLE 1.
AGING MANAGEMENT OF PWR CONTAINMENTS FOR LICENSE RENEWAL l
Page J
4 l
Concrete & Steel Containment One-time Inspection...................
7 1
Concrete Structure - Aging Mechanisms i
Freeze-Thaw...................................................
8 Leaching of Cal cium Hydroxide.................................
9 Aggre s sive Chemi cal Attack....................................
10 Reaction with Aggregates......................................
11-Elevated Temperature..........................................
12 Irradiation of-Concrete.......................................
12 Concrete Interaction with Aluminum............................
13 Structural Steel & Liner - Aging Mechanisms Corrosion.....................................................
14 Elevated Temperature..........................................
15 Irradiation of Steel..........................................
15 Stress Corrosion Cracking.....................................
16 Peinforcing Steel - Aging Mechanism l
Corros i on o f Embedded Steel...................................
17 Reinforcing Steel & Prestressing Tendons - Aging Mechanisms El e v at ed Temp e ra t u re..........................................
18 Irradiation of Steel..........................................
19 Containment Structures & Components - Aging Mechanism Fatigue....................................................... 20 Containment Structure & Its Concrete Basemat - Aging Mechanisms-Settlement....................................................
21 Erosion of Cement.............................................
22 Containment Structure & Its Components - Aging Mechanism Strain Agi ng of Carbon Steel..................................
23
. Concrete Containment Prestressing Tendons - Aging Mechanisms Stress Relaxation of Prestressing Wire, Shrinkage Creep, Anchorage Seating Losses, and Tendon-Friction.................
24 Corrosion of Tendons..........................................
25 Containment Pressure Retaining Components - Aging Mechanism Me ch an i c al W e a r...............................................
26 xi NUREG-1611 s
'l r
\\
Table 2.
AGING MANAGEMENT 0F BWR CONTflNMENTS FOR LICENSE RENEWAL Page-1_
Concrete & Steel Containment One-time Inspection...................
27 Concrete Structure - Aging Mechanisms Freeze-Thaw................................................... 28 Leachi ng of Cal cium Hydroxide................................. 29 Agg re s s i ve C hemi c al At t a c k.................................... 30
-Reaction with Aggregates...................................... 31 Elevated Temperature.......................................... 32 4
Irradiation of Concrete....................................... 32 Structural Steel & Liner - Aging Mechanisms Atmospheric Corrosion......................................... 33 Local Corrosion............................................... 34 Elevated Temperature.......................................... 35 i
. Irradiation of Steel.......................................... 35 Stress Corrosion Cracking.................................... 36 Reinforcing Steel (Rebar) - Aging Mechanism Corrosion of Embedded Steel................................... 37 Reinforcing Steel & Prestressing Tendons - Aging Mechanisms El evat ed Temperature.............,............................ 38
[
Irradiation of Steel.......................................... 39 Containment Structures & Components - Aging Mechanism Fatigue....................................................... 40 4:
E Containment Structure & Its Concrete Basemat - Aging _ Mechanisms Settlement....................................................
41 E r o s i o n o f C e me n t............................................. 42 l
-Containment Structure & Its Components - Aging Mechanism Strain Aging of Carbon Steel.................................. 43 Concrete Containment Prestressing Tendons - Aging Mechanisms Stress Relaxation of Prestressing Wire, Shrinkage Creep, Anchorage. Seating Losses, and Tendon Friction................. 44 Corrosion'of Tendons......................................... 45 Containment Pressure Retaining Components - Aging Mechanism Me c h a n i c al We a r............................................... 46 i
1 i
xiii NUREG-1611
)
1.0 INTRODVCTION Part 54 of 10 CFR, the license renewal rule, was published on May 8, 1995, providing requirements for renewal of operating licenses for nuclear power plants.
10 CFR 54.21(a)(1)(1) requires an aging management review of structures and components within the scope of license renewal to ensure that the effects of aging will be managed so that their intended functions will be maintained for the period of extended operation. Containment structures are subject to this requirement.
Receptly, the Commission amended 10 CFR 50.55a to promulgate requirements for inservice inspection of containment structures. The final rule on 550.55a,
" Codes and Standards for Nuclear Power Plants; Subsection IWE and Subsection IWL," was published on August 8, 1996 (61 FR 41303).
This rule incorporates by reference the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of Section XI, Division 1, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code addressing the inservice inspection of metal containments / liners and concrete containments, respectively (References 1 and 2]. Guidance for implementation of the containment inspection requirements is described in Appendix A of this report.
In 1990, the Nuclear Management and Resources Council (NUMARC), now the Nuclear Energy Institute (NEI), submitted for NRC review, ten industry reports (irs) addressing aging issues associated with specific structures and components of nuclear power plants for license renewal. Of the 10 irs, cne addresses PWR containments, and another addresses BWR containments [ References 3 and 4]. No safety evaluations were developed for the review of these irs.
However, NUREG-1557 provides a brief summary of the technical information and NUMARC/NRC agreements resulting from the review of nine of the 10 1Rs.
The NUMARC Report 90-08, " Low-Voltage, In-Containment, Environmentally-qualified Cable License Renewal Industry Report,"[ Reference 5] is not addressed in this NUREG since the subject is being addressed under GSI-168.
On August 26, 1996, the Commission issued Draft Regulatory Guide DG-1047,
" Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses." for public comment as part of the implementation of 10 CFR Part 54, the license renewal rule. A comment was received concerning whether the NRC staff had any plans to limit the scope of license renewal review for containments since the final rule on 10 CFR 50.55a, which endorses the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL, was published on August 8, 1996. The purpose of this NUREG is to reconcile the technical information and agreements resulting from the NUMARC IR reviews and the inservice inspection requirements of Subsections IWE and IWL as promulgated in 550.55a for license renewal consideration.
2.0 LICENSE RENEWAL EVALUATION OF AGING MANAGEMENT OF CONTAINMENT STRUCTURES The NRC staff reviewed Tables B3 and B4 of NUREG-1557 [ Reference 5] for the-PWR and BWR containments, respectively, to determine if Subsections IWE and IWL inspection requirements are consistent with the technical agreements from the IR reviews. Where NUREG-1557 indicates that an aging effect on specific 1
1 structures is nn-significant, the NRC staff recommends no aging management program.- Where hdREG-1557 indicates that an aging effect is non-significant if certain conditions are met, the NRC staff reviewed Subsections IWE/IWL and
$50.55a to determine if the containment inspection requirements would be adequate to manage that aging effect regardless of whether those conditions are met. Where NUREG-1557 indicates that an aging effect should be managed with specified programs, the NRC staff reviewed Subsections IWE/IWL and
$50.55a requirements to determine if they are adequats to manage that aging effect for the renewal term.
If the NRC staff determined that Subsections IWE/IWL and 650.55a requirements should be augmented to manage a certain aging effect for the renewal term, additional inspections or evaluations are recommended.
The results of the NRC staff evaluation are provided in Tables 1 and 2 of this report for the PWR and BWR containments, respectively.
The PWR and BWR containment structural components evaluated in NUREG-1557 are listed in Appendix B and Appendix C of this report, respectively.
3.0 CONCLUSION
S The NRC staff has reconciled the technical information and agreements from the NUMARC IR reviews and the inspection requirements of Subsections IWE/IWL as promulgated in 650.55a for managing the effects of aging for PWR and BWR 3
containments for the period of extended operation.
The staff found that the requirements of Subsections IWE/IWL and 650.55a will be an effective aging management program for managing the aging effects of containment structures for the period of extended operation, provided that the following additional evaluations and inspections specifically for license renewal are also g
performed:
r a.
Specific requirements contained in Part 54, the license renewal rule, such I
as Part 54.4 for scoping and intended function, and Part 54.21 for evaluating time-limited aging analyses.
b.
ASME Section XI, Appendix VII and Appendix VIII [ References 6 and 7] to be implemented when ultrasonic examinations are utilized for inspection of containments, c.
The following issues, in addition to implementing Subsections IWE/IWL through 550.55a, should be addressed in a license renewal application:
(1)
Management of potential aging effects of structures in inaccessible arcas when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas.
This is discussed in Items 1, 3, 4, 9, and 13 of Table 1 for PWR containments, and in Items 1, 3, 4, 9, and 13 of Table 2 for BWR containments of this report.
(2)
Fatigue associated with containment penetration bellows and penetration sleeves. This is discussed in Item 16 of Table 1 for PWR containments, and in Item 16 of Table 2 for BWR containments of this report.
(3)
Settlement associated with containment concrete basemat bearing on soil or piles, or experiencing significant changes in ground water conditions. This is discussed in Item 17 of Table 1 for PWR containments, and in item 17 of Table 2 for BWR containments of this report.
(4)
Erosion of cement for porous concrete if subfoundation layers of porous concrete are used in the construction of containment concrete basemat with the presence of underground water. This is discussed in Item 18 of Table 1 for PWR containments, and in Item 18 of Table 2 for BWR containments of this report.
(5)
Performance of examinations specified in Examination Category E-B for pressure retaining welds, and Examination Category E-F for pressure retaining dissimilar mets 1 welds of Subsection IWE for license renewal.
1his is discussed in Item 12 of Table 1 for PWR containments, and Item 12 of Table 2 for BWR containments of this i
report.
(6).
Cracking of penetration bellows. This is discussed in Item 12 of Table 1 for PWR containments, and Item 12 of Table 2 for BWR containments of this report.
(7)
Elevated temperature of prestressing tendons for (prestressed) concrete containments. This is discussed in Item 14 of Table 1 for l
PWR containments, and Item 14 of Table 2 for BWR containments of l
this report.
)
l The NRC staff recommends that the requirements of Subsections IWE and IWL through 650.55a, and those items identified in sections 3.a. through 3.c.
1 above be incorporated into the Standard Review Plan for License Renewal (SRP-LR).
ftEFERENCES 4
- 1. Subsection !WE, " Requirements for Class MC and Metallic Liners of Class CC components of Light-Water Cooled Power Plants,"Section XI Division 1, Boi,er and Pressure Vessel Code, American Society of Mechanical Engineers, New York, N.Y., 1992 Edition and 1992 Addenda.
- 2. Subsection (WL, " Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants "Section XI, Division 1. Boiler and Pressure
.VesselCode,AmericanSocletyofMechanicalEngineers,NewYork,N.Y.,
1992 Edition and 1992 Addenda.
- 3. " Pressurized Water Reactor Containment Structures License Renewal Industry Report," NUMARC Report Number 90-01, Revision 1 Nuclear Management and Resource Coun;.i September 1991.
- 4. "BWR Containments, License Renewal Industry Report," NUPARC Report Number 90 10, Revision 1. Nuclear Management and Resource Council, December 1991.
- 5. NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License F.enewal," U.S. Nuclear Regulttory Commission, October 1996.
- 6. Appendix Vll, " Qualification of Nondestructive Examination Personnel for Ultrasonic Examination "Section XI. Division 1, Boiler and Pressure
'essel Code, American Society of Mechanical Engineers, Naw York, N.Y.,
t 1989 Edition.
- 7. Appendix Vlli, " Performance Demonstration for Ultrasonic Examination Systems,"Section XI Dtvision 1, Boiler and Pressure Vessel Code, American Society of hechanical Engineers, New York, N.Y., 1989 Addenda.
- 8. NRC Information Notice 97-10. " Liner Plate Corrosion in Concrete Containments," March 13, 1997.
i
- 9. NRC Information Notice 97-11. " Cement Erosion from Containment i
Subfoundations at Nuclear Power Plants," March 21, 1997.
- 10. NRC Regulatory Guide 1.35, Revision 3. " Inservice Inspnction of Ungrouted Tendons in Prestressed Concrete Containments," July 1990.
- 11. ACI 201.2R-77, " Guide to Durable Concrete," American Concrete Institute.**
j
Fatigue Loading," American Concrete Institute.**
1
- 13. ACI 318, " Building Code Requirements for Reinforced Concrete," American Concrete Institute.**
f NVREG-1611 4
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-em----..,-
- +..
t REFERENCES frontinued)
- 14. ACI 359, " Code for Concrete Reactor Vessel and Containments," American l
Concrete Institute.**
- 15. NRC Information Notice 92-20, " Inadequate local Leak Rate Testing," Harch 3, 1992.
l
- 16. Letter from T. F. Plunkett of the Florida Power and Light Company to l
Stewart D. Ebneter of hRC, dated January 25, 1993.
denotes that the citation is used as a reference to provide only background information.
5 NUREG-1611 I
TABLE 1.
AGING MANAGEMENT OF PWR CONTAINMENTS FOR LICENSE RENEWAL Com)onent,A Mec1anism &ging item Aging Issue and Evaluation, Effects l
01 Concrete & Steel iting: A "one-time inspection for license Conta' nment renewal." NUREG-1557 states that one time l
inspection is an unresolved issue regarding staff request for inspection of concrete Aoina mechanism:
containment & steel containment to assess the current condition of containment and to Not applicable.
provide a baseline information for any future inspections (Page B-28 of NUREG-1557).
Aoina effects:
Recommendation: The issue is resolved with the implementation of IWE/lWL through 150.55a.
General However, specific-recommendations for applicable aging effects are addressed hereinafter in this table.
Discussion: Subsections IWE and IWL require periodic inspection of the containment in accessible areas. These inspections would periodically assess the condition of the containment and each inspection would provide a documented baseline for subsequent inspections. Furthermore, 650.55a(b)(2)(ix)(E) and (b)(2)(x)(A) require an evaluation of inacc6ssible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
However, the management of potential aging effects of inaccessible areas, when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas, is addressed individually for each applicable aging effect (i.e.
Items 1, 3, 4, 9, and 13 of this table). Conditions for such inaccessible areas should be evaluated for license renewal. A program for a one-time inspection may be proposed.
02 concrete Structure litug: NURfG-1557 states that freeze-thaw is non-signif? ant if the following conditions are met: coscrete containment structures Aoina mechanism:
located in a geographic regions of negligible weathering conditions (weathering index <100 freeze-thaw day-inch /yr); and if located in severe (weathering index >500 day-inch /yr)hering or moderate (100-500 day-inch /yr)ix design meets weat conditions with the concrete m Aaina effects:
the air content & water-to-cement ratio Scaling, cracking, requirements of ASTM C260 or equivalently, the
& spalling ASME Sect. Ill, Division 2,her freeze-thaw is paragraph CC 2231.7.1. The issue of whet potentially significant for the concrete containment dome, particularly in severe weathering regions, is identified as unresolved (Page B-29 of NUREG-1557).
RecommendatiRD: The issue is resolved with the implementation of IWL.
Discussion: Freeze-thaw results in scaling, cracking, and spalling. Any freeze-thaw degradation would initially ap) ear in the exposed concrete structure. Su)section IWL, Examination Category L-A, requires periodic visual examination of accessible concrete surfaces and would detect any freeze-thaw damage of the concrete contain:nent, including the dome, regardless of whether the above weathering conditions are met.
1 NUREG-1611 8
03 Concrete Structure issue: NUREG-1557 states that leaching of calcium hydroxide is non-significant for containment concrete structures if the Aaina mechanism:
following conditions are met: concrete structures not exposed to flowing water; and teaching of calcium for concrete structures that are exposed to hydroxide flowing water but are constructed using the guidance of ACI 201.2R-77 to ensure dense, wall-cured concrete with low permeability and Aoina effects:
control cracking through proper arrangement l
and distribution of reinforcement (Page B-29 i
Increase of of NUREG-1557).
porosity &
permeability Recommendation: The issue would be managed i
with the implementation of IWL through l
650.55a. However, the management of potential leaching of calcium hydroxide of inaccessible areas of containment concrete structures when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussian: lWL, Examination Category L-A, requires periodic examination of accassible concrete surfaces and $50.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in
)
degradation to such inaccessible areas.
Regardless of whether the above conditions are i
met, potential leaching of calcium hydroxide would be detected as water stains on accessible surfaces by the IWL visual examination. However, the management of potential leaching of calcium hydroxide of inaccessible areas (e.g., below grade portion of concrete structures with presence of flowing water) when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas, needs to be evaluated on a plant specific ba:is.
I 04 Concrete Structure hsm: NUREG-1557 states that aggressive chemical attack is non-significant for above grade concrete containment structures because Aoina mechanism:
they are not exposed to ground water.
Aggressive chemical attack is non-significant Aggressive chemical for below grade concrete containment attack structures if the following conditions are met containment concrete is not exposed to aggressive ground water (pH <5.5, chloride Aaina effects:
>500 ppm, & sulfate >l500 ppm); or if exposed to ground water that exceeds the pH, chloride, increase of sulfate limits, the exposure is for porosity and intermittent periods only. NUREG-1557-permeability, indicates that inspection of concrete cracking, and containment structure should be in accordance spalling with IWL. NUREG-1557 states that evaluation for management of inaccessible areas of below grade concrete containment structures is to be justified on a plant-specific basis (Page B-30 of NUREG-1557).
Recommenddign: The issue would b, managed with the implementation of IWL through 650.55a. However, the management of potential aggressive chemical attack of inaccessible areas of containment concrete structures when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: Aggressive chemical attack results in increase of porosity and permeability, cracking and spalling. IWL, Examination Category L-A, requires periodic examination of accessible concrete surfaces and 650.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Regardless of whether the above conditions are met, potential aggressive chemical attack would be detected by IWL and 550.55a(b)(2)(ix)(E). However, the management of potential aggressive chemical attack of inaccessible areas when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated.
HUREG-1611 10
l 05 Concrete Structure Ingg: NVREG-1557 states that reaction with aggregates is an unresolved issue. NUREG-1557 indicates that the NRC staff believes tnat A2ina mechanism:
alkaline-aggregate reactions can not be ruled out. Tests involving aggregates alone are not Reaction with sathfactory in predicting aggregate aggregates performance. Alkaline-aggregate reaction may occur after 25 or more years (Page B-31 of Aoina effects:
Recommendation: The issue is resolved with the l
Expansion and implementation of IWL through 650.65a.
cracking Discussion: If alkaline-aggregate reaction occurs, it will manifest itself as spalling and cracking of the surface of the concrete due to expansion because of the chemical reaction. Further, reaction with aggregates in inaccessible areas would also occur in accessible areas because aggregates wers used in construction of both accessible and inaccessible areas. lWL, Examination Category L-A, requires periodic examination of accessible concrete surfaces and 650.55a(b)(2)(ix)(E) hen conditions exist in requires an evaluation of inaccessible areas w accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. lWL and 150.55a(b)(2)(ix)(E) will detect such degradation, 11 NVREG-1611
i
)
06 Concrete Structure hing: NUREG-1557 states that elevated temperature is non-significant for concrete structures if it meets the following Aaina mechanism conditions: concretecontainmentstructugesbe main ained at operating temperatures <6,6 C Elevated and local area temperatures <93 C temperature
- or for concrete structures that experience temperatures greater than the above specified limits, a plant specific Aaina effects
justification should be provided (Page B-32 of NUREG-1557).
Loss of strength &
modulus Recommendation: For concrete containment structures that ex)erience temperatures greater than the a)ove specified limits, a plant specific evaluation is needed.
Discussion: Elevated tem)erature results in loss of concrete strengt1 and modulus which may not be detected by the implementation of IWL and $50.55a modification until the aging effects are so severe as to result in cracking and spalling. Thus, for concrete structures that experience temperatures greater than the above specified limits, a plant specific justification should be provided.
07 Concrete Structure hing: NUREG-1557 states that irradiation of concrete is non-significant for containment Aaina mechanism:
Recommendation: The issue is_non-significant.
Irradiation of concrete Discussion: The neutron fluence levels and maximum integrated gamma doses experienced by containment concrete during the license Aaina effecti:
renewal term is not ex)ected to exceed the levelatwhichmeasura)1edegradationog 2
loss of strength &
concgete strength pro erties occurs (10 n/cm modulus
& 10 rads, respecti ely). Thus the issue is non-significant.
NUREG-1611 12
08 Concrete Structure 111gg: NUREG-1557 states that concrete interaction with aluminum is non-significant for concrete containment structures if Aoina mechanism:
aluminum piping was not used for concreto placement, otherwise any adverse effects of Concrete concrete interactions with aluminum would have interaction with been identified during the initial acceptance aluminum test prior to initial operation (Page B-42 of NUREG-1557).
Aaina effects:
Recommendation: Concrete interaction with aluminum is not an issue for license renewal.
Loss of strength Discussion: Adverse effects of concrete interactions with aluminum would have occurred during the placement of concrete and would have been identified during initial structural acceptance test prior to plant initial operation. Identifit;d concerns would have been evaluated during plant construction for l
appropriate corrective action. Any containment having concrete placed through aluminum pipelines which successfully completed its acceptance tests was not adversely affected by this placing condition. Thus it is not an issue for license renewal.
l 13 NUREG-1611
09 Struct. Steel &
Jngg: NVREG-1557 indicates that corrosion is Liner non-significant for above grade steel liner, steel containment shells (except at the proximity of the ice-condenser), and common Aoina mechanism:
steel components. NUREG-1557 indicater that galvanic corrosion is non-significant for Corrosion penetration bellows if they are protected by shields from corrosive environment. NUREG-1557 indicates that inspection of structural steel Aoina effects:
and liner should be in accordance with IWE.
NUREG-1557 also indicates that evaluation for loss of material menagement of inaccessible areas of below grade structural steel and liner is to be justified on a plant-specific basis (Pages B-37 and B-38 of NUREG-1557).
Recommendation: The issue would be managed with the implementation of IWE through 550.55a. However, the management of potential corrosion of inaccessible areas of structural steel liner, steel containment snells, and common steel components when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: IWE, Examination Categories E-A, E-C, E-0, & E-G, provides periodic examination of accessible areas to uncover evidence of structural degradation and should detect corrosien of structural steel and liner; IWE, Examination Category E-P (Appendix J to 10 CFR 50, Type A-test), recuires a general inspection and an integratec leakage test; and
$50.55a(b)(2)(x)(A) requires an evaluation of acceptabilityofinaccessibleareaswhen conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
c However, the management of-potential corrosion of inaccessible areas of structural steel liner, steel containment shells, and common steel components when conditions in' accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated, i
NUREG-1611 14 1
10 Struct. Steel 4 111gg: NUREG-1557 states that elevated Liner temperature is non-significant for containment structural steel liner, steel containment shells, and common steel components such as Aoina mechanism:
penetration bellows / sleeves, personnel airlock, equipment hatcLes (Page B-33 of Elevated NU9EG-1557).
temperature Recommendation: The issue is non-significant.
Aaina effects:
Discussion: OperatingtemperaturgswithinPWR which are well below the 316lC (600 F) levelcontainment str Loss of strength &
modulus at which the structural integrity of rebar/ concrete combination begins to be significantly affected. Thus the issue is non-significant.
11 Struct. Steel &
11nte: NUREG-1557 states that irradiation of Liner steel is non-significant for containment structural steel liner, containment shells, and common steel components (Page B-35 of l
Aaina mechanism:
Irradiation of Recommendation: The issue is non-significant.
I steel Discussion: The cumulative radiation exposure experienced by concrete containment liners or Aoina effects:
free-standing steel containment shells throughout the license renewal term is Loss of fracture expegtedgobefarbelowthelevelof toughness 2x10 n/cm (>l MeV) which could cause a change in mechanical or physical properties. Thus the issue is non-significant.
15 NUREG-1611
_ _~
12 Struct. Steel &
lssue: NUREG-1557 indicates that SCC is non-List significant for concrete containment st el i
liner, free-standing steel containment shells, and common steel components in the containment Aaina mechanb_g:
environment unless dissimilar metal is used, and in the case of SS bellows assemblies for Struss corrosion CS vent lines or pipe sleeves if the materials cracking (SCC) are protected by shields from corrosive i
environment (Page B-37 of lVREG-1557).
Aoina effects:
Becommendation: This issue would be managed by Examination Categories E-B & E-F of Subsection Crack initiation &
IWE and Appendix J to 10 CFR 50. In addition, growth an augmented VT-1 visual examination of bellows bodies should be performed using enhanced technigdes qualified for detecting stress corrosion cracking in bellows bodies.
Discussion: IWE, Examination Category E-F, provides periodic surface examination of pressure retaining dissimilar metal welds for dissimilar metals and could detect SCC. IWE, Examination Category E-S, provides periodic visual examination of pressure retaining welds for containment penetrations. Also, any leakage associated with the containment shell or steel liner due to through-wall cracks resulting from SCC would be detected by periodic Appendix J 1eak rate test & remains within the limits of plant specifications or Subsection IWE. Although 950.55a indicates that Examination Categories E-B & E-F are optional during the current term of operation, these examinations should be performed for license renewal to demonstrate that no SCC has been initiated. In addition, since occurrences of transgranular stress corrosion cracking have been identified in operating plants on stainless steel bellows (Reference 15), an augmented examination on the surface areas of bellows bodies should be performed so that cracking would be detected.
l 1
NUREG-1611 16 j
13 Reinforcina Steel Issue: NUREG-1557 states that corrosion of (Rebar) embedded steel is non-significant for concrete structures above grade if not exposed to aggressive environment, pH<ll.5 or chlorides f,aina mechanism:
>500 ppm; or if exposed to aggressive _
environment, concrete has relatively high Corrosion of strength (27.6 MPa (4 ksi)), low water-to-embedded steel cement ratio (0.35-0.45), adequate air entrainment (3-6%), low permeability, and designed in accordance with ACI 318 or ASME Aoina effects:
Section 111, Division 2. NUREG-1557 also indicates corrosion of embedded steel for l
Loss of bond & loss concrete structures below grade exposed to l
of material aggressive ground water (pH <5.5, chloride
>500 ppm, & sulfate >l500 apm) should be exhmined in accordance witi IWL and management of inaccessible areas should be justified on a case-by case basis. Also the NRC staff considers that potential degradation due to chloride corrosion (e.g., ground water chemical attack) of PWR containments should be addressed (Page B-36 of NUREG-1557).
l l
Recommendation: The issue would be managed with the implementation of IWL through 650.55a. However, the management of )otential corrosion of inaccessible areas of emaedded steel when conditions in accessible areas may not. indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: IWL, Examination Category l-A, requires periodic examination of accessible concrete surfaces.and $50.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible araas that could indicate the presence of or result in degradation to such inaccessible areas.
Corrosion of embedded steel results in cracking and spalling of concrete and would be detected by inspections, regardless of whether the above conditions are met.
However, the management of potentiel corrosion-of inaccessible areas of emt ided steel, when conditions in accessible areas may not indicate the presence of or result in degradation to ::uch inaccessible areas needs to be evaluated. This would also address the staff's concern on chloride corrosion.
17 NUREG-1611
14 Reinf. Steel l Issue: NUREG-1557 states that elevated liestr. Tendons temperature is non-significant for concrete containment reinforcing steel and for concrete containment prestressing tendons (Page B-32 of Aoina mechanis.10:
Elevated Recommendation: The issue is non-significant, temperature except for prestressed tendons. The tendon surveillance program should be augmented to include additional tendons based on their sun Aaina effects:
exposure or proximity to hot penetrations.
Loss of strength &
Discussion: OperatingtemperaturgswithinPWR containment structures are 4 -66 C fl20-150,F modulus which are well below the 316lC (600 F) level at which the structural integrity of rebar/ concrete combination begins to be significantly affected. Thus the issue is non-significant for reinforcing steel. However, increase in temperature increases prestress loss in presteessed tendons. Prestress losses increasedfrom8%toJ4%whenthetgmperagure was increased from 20 C (68 F) to 32 C (90 F)
[ Reference 16). Thus, tem)eratures due to sun exposure or proximity to act penetrations may increase the prestress loss in tendons. The tendon surveillance program described in Regulatory Guide 1.35 (Reference 10) is based on a small sample size, that is, a 4 percent random sample including a repeat tendon.
Tendons subject to warm temperatures may not be tested because of this small sample size.
The tendon surveillance program should be augmented to include additional tendons.
These additional tendons should be selected based on their sun exposure or proximity to hot penetrations.
NVREG-1611 18
l 15 Reinf. Steel &
lssue: NUREG-1557 states that irradiation of Prestr. Tendqng steel is non-significant for concrete structures reinforcing steel (including basemat reinforcing steel) and concrete Aaina mechanism:
containment prestressing tendons (Page B-34 of Irradiation of steel EtcommendittiSD: The issue is non-significant.
Discussion: The cumulative radiation exposure Aoina effects:
experienced by reinforced concrete containment structures during the license renewa expected to be below the level of 10),n/cmtergis i
Loss of fracture for l
toughness degradation of reinforcing steel, and PWR I
concrete containment prestressing tendons &
corrosion inhibitors will not receive enough radiation exposure during the license renewal tep /cm,- & 10to incur ge related degradation (<4 x 8
10 n rads, respectively). Thus the issue is non-significant.
19 NUREG-1611
16 Containment hug: NUREG-1557 states that fatigue is non-Structures &
significant for containment structures and its Comoonents components, except for the penetration sleeves and bellows. NUREG-1557 also indicates that fatigue is an unresolved issue for concrete Aoina mechanism:
containment penetration sleeves and steel containment penetration bellows and fatigue Fatigue damage may not be detectable by a leak rate test (Pages B-40 & B-41 of NUREG-1557).
Aoina effects:
Recommendation: Fatigue is non-significant for containment structures and its components Cumulative fatigue except for the penetration sleeves and damage bellows.
Fatigue of containment penetration sleever and penetration bellows is a " time-limitH aging analysis" and must be evaluated in acco dance with the license renewal rule,
-10 CFR 54.21(c).
Discustign: Fatigue is non-significant for containment concrete, reinfort g steel, prestressing system components, steel liners, and free-standing steel containments, because theyaredesignedtohayegoodfatigue strength properties (10 cycles) of below yield load in accordance with ASME Section 111. Division 2, or ACl 318, and ACI 215R-74 codes.
Containment )enetration sleeves and penetration )ellows are designed to Section til of the ASME-Code which requires a fatigue analysis based on an assumed riumber of cycles.
This fatigue-analysis is a " time-limited aging analysis" and must be evaluated in accordance with license renewal rule 954.21(c) to ensure that the effects of aging on the intended functions will be adequately managed for the period of extended operation, s
h_
NUREG-1611-20
]
4
17 Containment hEte: NUREG-1557 states that settlement is an Structure & its unresolved issue for containment concrete Concrete basemat basemat for sites with soil, or significant changes in ground water conditions, and the effect of settlement needs to be evaluated Aaina mechanhm:
(Page B-42 of NUREG-1557).
Settlement Recommenda11gn: The issue is resolved by establishing a settlement monitoring program which would ensure that differential Aaina effects:
settlement of containment basemat does not exceed the design criteria for a containment Cracks, distortion, structure and its basemat which is resting on increase in soil or piles, or experiencing significant component stress changes in ground water conditions.
level Discussion: Effects of differential settlement are potentially significant for a containment structure and its concrete basemat that is resting on soil or piles, or experiencing I
significant changes in ground water conditions. Subsection lWL does not address the effects of settlement. Because the effects of settlement could cause cracks and I
distortion of concrete basemat and could I
result in increasing stress levels greater i
than original design basis in the basemat and other parts of the containment structure. The effects of settlement of PWR containments need to be evaluated. However, a settlement monitoring program could ensure that the differential settlement does not exceed the design criteria for the containment structures throughout the license renewal term. A settlement monitoring program should be provided to manage settlement for containment basemat bearing on soll or piles, or experiencing significant changes in ground water conditions for the period of extended operation.
21 NUREG-1611
18 Containment Inug: NRC Information Notice 97-11 indicates Structure & its that erosion of cement from porous concrete Concrete basemat could be potentially significant for porous concrete subfoundations below concrete basemat if subfoundation liyers of porous concrete are Aoina mechanism:
used in the construction of concrete basemat Erosion of cement Recommendation: For those applicable plants, the management of potential erosion of cement Aoina effects:
from porous concrete needs to be justified on a plant-specific basis.
Loss of strength Discussion: 650,55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could-indicate the presence of or result in degradation to such inaccessible areas.
However, the management of potential erosion of cement from porous concrete of inaccessible areas of containment concrete basemat when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated on a plant-specific basis.
1 NUREG-1611 22
19 Containment litug: NUREG-1557 indicates that strain aging itructure & its is non-significant for steel containment Components structures (including common components, such as penetration sleeves, penetration bellows, personnel airlock, and equipment hatches) that Aoina mechanism:
meet the following conditions: Dynamic strain aging is r.on-significant for free standing Strain aging of steel containment structures that do not allow carbon t. teel loads to exceed the elastic limit. Static strain aging is non-significant for free standing steel containment structures that Aoina effects:
were not cold worked; or if cold worked during the forming process, the plates were Loss of fracture normalized or stress relieved or both after toughness forming with minimal
<5%) subsequent cold working (Page B-43 of(NUREG 1557).
Recommendation: This issue is non-significant.
Discussion: Dynamic strain aging is not expected in the carbon steel components of steel cor.tainments during their service life, since the strain:; associated with the design service loads are below the elastic limit of the material. The PWR containment is made from l
low-carbon steel, and the steel is normalized or stress relieved or both following plate rolling. Further, strain aging requires stressingofthematerialtoaboveitsyiejd stregs, and aging at temperatures-above 93 C (200 F). Carboncrelatgd strain aging at 0
temperatures below 93 C (200 F) is normally negligible due to the_ low solubility of carbon in this temperature range. The PWR containment has a maximum tempera ure during normal operation of about 66}C (150'F), and loading conditions do not produce service stresses in l
the range of the material yield strength. Thus strain aging is non-significant for the steel containment structures and its components.
l 23 NUREG-1611
20 Cone. Containment Issue: NUREG-1557 indicates that loss of Prestr. Tendons prestress due to stress relaxation, shrinkage creep, etc., would be a reduction of design margin and could be potentially significant Aoina mechanism:
for prestressing tendons for license renewal (Page B-41 of NUREG-1557).
Stress relaxation of prestressing Recommendation: Loss of prestress for wire, shrinkage prestressing tendons would be managed with the creep, anchorage implementation of Subsection IWL through seating losses, and 550.55a. In addition, the tendon prestress tendon friction evaluation is a " time-limited aging analysis' and must be evaluated in accordance with the license renewal rule, 10 CFR 54.21(c).
Discussion: Subsection IWL, Examination loss of prestress Category L-8, and $50.55a inspections would be able to d:tect potential loss of prestress for prestressing tendons.
For example: IWL-2522 provides examination method for tendon force measurements; IWL-3221 provides acceptance standard for measuring tendon force; 550.55a(b)(2)(ix)(B) states that "when evaluation of consecutive surveillances of prestressing forces for the same tendon...
indicates a trend of prestress loss....,
an evaluation shall be performed;" repair and replacement are addressed in IWL-4000 and IWL-7000, respectively. In addition, the recownendations of Regulatory Guide 1.35
[ Reference 10) were incorporated into the 1992 Edition with the 1992 Addenda of Subsection IWL and 650.55a(b)(2)(ix)(A)-(D), the final rule, dated August 8, 1996 (61 FR 41304). Thus the " loss of prestress" for prestressing tendons would be managed with the implementation of IWL through 950.55a.
Further, the tendon prestress evaluation is a
" time-limited aging analysis" and must be evaluated for renewal to demonstrate that the prestressing force will meet the design requirements at the end of 60 years in accordance with license renewal rule 654.21(c).
l l
NUREG-1611 24 l
21 Conc. Containment Issue: NUREG-1557 s'ites that corrosion of Prestr. Tendons tendons is an unresolved issue in that the NRC staff is concerned that a large amount of grease leakage can degrade concrete strength.
Aoina mechanism:
IWL (1992 Edition with 1992 Addenda) lacks certain criteria contained in RG 1.35. These Corrosion of criteria are addressed in 10 CFR 50.55a final tendons rule, dated August 8, 1996, Section 50.55a(b)(2)(ix)(A)-(D) on issues such as failed wires and tendon sheathing filler Aoina effects:
grease conditions. Also, anchor heads have failed in prestressed concrete containments.
Loss of material NUREG-1557 also states that prestressing tendons and tendon anchorage hardware should be examined in accordance with the provisions of RG 1.35 for prestressed concrete containments (Page B-39 of NVREG-1557).
l Recommer,dation: The issue is resolved with the implementation of Subsection IWL through 550.55a.
Discussion: Subsection IWL, Examination Category L-8, and 950.55a inspections would be able to detect corrosion of prestressing tendons. For example: IWL-2525 provides methods for examination of corrosion protection medium and free water; IWL-3221 provides acceptance standard for corrosion protection medium; IWL-2524 provides visual examination of tendon anchorage areas; IWL-3221 provides acceptance standard for inspection of tendon anchorage areas; 650.55a(b)(2)(ix)(A) states that " grease caps that are accessible must be visually examined to detect grease leakage or grease cap deformations.......
that indicates deter 4+ation of anchorage hardware." In additun, the recommendations of Regulatory Guide 1.35 [ Reference 10) were incorporated into the 1992 Edition with the 1992 Addenda of Subsection IWL and 650.55a(b)(2)(ix)(A)-(D),
the final rule, dated August 8, 1996 (61 FR 41304). Thus the " Corrosion of prestressing tendons" would be managed with the implementation of IWL through 650.55a.
25 NUREG-1611 s
r 22 Containment 111M1: NUREG-1557 does not address mechanical Pressure Retainina wear for PWR containments.
- However, comoonents mechanical wear could be potentially significant for components that are subject to relative sliding or rotating motion and that Aaina mechanism:
are susceptible to fretting and/or lockup.
Mechanical Wear Recommendation: Mechanical wear for PWR containment pressure retaining components would be managed with the implementation of Aoina effects:
Subsection IWE (Examination Categories E-D, E-G, and E-P).
Fretting, Lockup Discussion: NUREG-1557 addresses r.echanical wear issue only for BWR containments. However, mechanical wear should also be addressed for PWR containments. Inspection and mitigation of machanical wear conducted in accordance with the provisions of Subsections IWE & IWF, as applicable, would ensure that the integrity of containment pressure retaining components and their supports is maintained throughout the license renewal term. IWE, Examination Categories E-D & E-G, provides periodic examinations for the pressure retaining components (including airlocks, and equipment hatches). IWE, Examination Category E-P (Appendix J to Part 50. Type B test), would detect local leaks for those components. Thus any potential mechanical-wear degradation would be detected by the implementation of IWE for those pressure retaining components.
(There are no PWR containment pressure retaining component supports listed in Appendix B of this report and therefore, Subsection IWF is not applicable.)
i Evaluation is based on:
(1) the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of Section XI of the ASME Code; (2) the 1989 Edition of Section XI of the ASME Code including Appendix Vil, " Qualification of Nondestructive Examination Personnel for Ultrasonic Examination," and Appendix VIII (1989 Addenda),
" Performance Demonstration for Ultrasonic Examination Systems;"
(S) the final rule on 10 CFR 50.55a, Codes and Standards, dated August 8, 1996 (61 FR 41303); and (4) NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," dated October 1996.
NUREG-1611 26
TABLE 2.
AGING MANAGEMENT OF BWR CONTAINMENTS FOR LICENSE RLNEWAL.
Com)onent, Aging item Mec1anism & Aging issue and Evaluation, Effects l
01 Concrete & Steel hing: A "one-time inspection for license containment reneral." NUREG-1557 states that one time inspection is an unresolved issue regarding staff request for inspection of concrete Aoina mechanism containment & steel containment to assess the current condition of containment and to Not applicable.
provide a baseline information for any future inspectior.s (Page B-28 of NVREG-1557).
Aoina effects:
Recommendation: The issue is resolved with the I
implementation of IWE/lWL through 150.55a.
General However, specific-recommendations for applicable aging effects are addressed hereinafter in this table.
QhtunhD: Subsections IWE and IWL require periodic inspection of the containment in accessible areas. These inspections would periodically assess the condition of the i
containment and each inspection would provide a documented baseline for subsequent inspections,furthermore,650.55a(b)(2)(ix and (b)(2)(x)(A) require an evaluation of )(E) inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
However, the management of potential aging effects of inaccessible areas, when conditions in accessible areas m6y not indicate the presence of or result in degradation to such inaccessible areas, is addressed individ tally for each applicable aging effect (i.e., items 1, 3, 4, 9, and 13 of this table). Conditions for such inaccessible areas should be evaluated for license renewal. A program for a one-time inspection may be proposed.
27 NUREG-1611
02 Concrete Structure hsjg: NUREG-1557 states that freeze-thaw is non-significant for Mark I & Mark 11 concrete containments because they zre protected from Aaina mechanism:
freezing by the secondary containment. Freeze-thaw is non-significant for Mark 111 concrete Freeze-thaw containment components if the following conditions are met: Mark 111 concrete containment components located in a geographic Aoina effects:
regions of negligible weathering conditions (weathering index <100 day-inch /yr); and if Scaling, cracking, located in severe (weathering index >500 day-
& spalling inch /yr) or moderate (100-500 day-inch /yr) weathering conditions with the concrete mix design meets the air content & water-to-cement ratio requirements of ASTM C260-77; or equivalently, the ASME Sect. Ill Division 2, paragraph CC 2231.7.1; or the susceptible surfaces are protected by shielding (Page B-46 of NUREG-1557).
Recommendation: The issue is resolved with the implementation of IWL.
Discussion: Freeze-thaw results in scaling, cracking, and spalling. Any freeze-thaw degradation would initially ap) ear in the exposed concrete structure. Su)section IWL, Examinatten Category L-A, requires periodic visual-examination of accessible concrete surfaces and would detect any freeze-thaw damage of the concrete containment, including the dome, regardless of whether the above-weathering conditions are met.
NUREG-1611 28 l
03 Concr et e_itructure ling: NUREG-1557 states that leaching of calcium hydroxide is non-significant for containment concrete structures if the Aoina mechanism:
following conditions are met: concrete structuns not exposed to flowing water; and teaching of calcium for concrete structures that are exposed to hydroxide flowing water but are constructed using the guidance of ACI 201.2R-77 to ensure dense, well-cured concrete with low permeability and Aoina effects:
control cracking through proper arrangement and distribution of reinforcement (Page B-47 increase of of NUREG-1557).
porosity &
The issue would be managed permeability Ren_maendation:
with the implementation of IWL through 650.55a. However, the management of potential leaching of calcium hydroxide of inaccessible areas of containment concrete structures when l
conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: lWL, Examination Category L-A, requires periodic examination of accessible concrete surfaces and $50.55a(b)(2)(ix)(E) requires an evaluation of inaccessiLle areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Regardless of whether the above conditions are met, potantial leaching of calcium hydroxide would be detected as water stains on accessible surfaces by the IWL visual examination. However, the management of potential leaching of calcium hydroxide of inaccessible areas (e.g., below grade portion of concrete structures with presence of flowing water) when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas, needs to be evaluated on a plant-specific basis.
29 NUREG-1611
04 Concrete Structure hing: NUREG-1557 states that aggressive chemical attack is non-significant for above grade containment concrete structures because Aoina mechanism:
they are not exposed to ground water.
Aggressive chemical attack is non-significant Aggressive chemical for below grade containment concrete attack structures if the following conditions are met: containment concrete is not ex)osed to aggressive ground water (pH <5.5, c11oride Aaina effects:
>500 ppm, & sulfate >l500 ppm); or if expos %d to ground water that exceeds the pH, chloride, increase of sulfate limits, the exposure is for porosity and intermittent periods only. NUREG-1557 permeability, indicates that inspection of containment cracking, and concrete structure should be in accordance spalling with IWL. NUREG-1557 states that evaluati>n for management of inaccessible areas of be. low grade containment concrete structures is to be justified on a plant-specific basis (Page B-48 ofNUREG-1557).
Resommendation: The issue would be managed with the implementation of IWL through 650.55a. However, the management of potential aggressive chemical attack of inaccessible areas of containment concrete structures when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: Aggressive chemical attack results in increase of porosity and permeability, cracking and spalling. IWL, Examination Category L-A, requires periodic examination of accessible concrete surfaces and 550.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Regardless of whether the above conditions are met, potential aggressive chemical attack would be detected by IWL and of potential) aggress)ive chemical attack of$50.55a(b)(2 (ix)(E.
inaccessible areas when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated.
NUREG-1611 30
05 C,oncrete Structure issue: NUREG-1557 states that reaction with aggregates is an unresolved issue. NUREG-1557 indicates that the NRC staff believes thtt Aoina mechanism:
alkaline-aggregate reactions een not be ruled out. Tests involving aggregates alone are not Reaction with satisfactory in predicting aggregate aggregates performance. Alkaline-aggregate reaction may occur after 25 or more years (Page B-49 of NUREG-1557).
Aaina effects:
Recommendation: The issue is resolved with the Expansion and implementation of IWL through 950.55a.
cracking Discussion: If alkaline-aggregate reaction occurs, it will manifest itself as spalling and cracking of the surface of the concrete due to expansion because of the chemical l
reaction. Further, reaction with aggregates in inaccessible areas would also occur in accessible areas because aggregates were used in construction of both accessible and inaccessible areas. lWL, Examination Category L-A, requires periodic examination of accessible concrete surfaces and 550.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. IWL and 550.55a(b)(2)(ix)(E) will detect such degradation.
31 NUREG-1611
06 Concrete Structur_g Issue: NUREG-1557 states that elevated temperature is non-significant for concrete structures if it meets the following Acina mechanism:
conditions: concretecontainmentstructupesbe maingained at operating temperatures <66 C 0
Elevated (150'F); or for concrete structures thatF) and local area temperature (200 temperature experience temperatures greater than the above specified limits, a plant specific 6_aina effects:
justification should be provided (Page B-56 of NUREG-1557).
Loss of strength &
modulus Recommendation: For concrete structures that experience temperatures greater than the above specified limits, a plant specific evaluation should be performed.
Discussion: Elevated temperature results in loss of concrete strength and modulus which may not be detected by the implementation of IWL and 650.55a modification until the aging effects are so severe as to result in cracking and spalling. Thus, for concrete structures that experience temperatures greater than the above specified limits, a plant specific justification should be provided.
07 Concrete Structur_q lssue: NUREG-1557 states that irradiation of concrete is non-significant for containment concrete structures (Page B-57 of NUREG-1557).
Aainc mechanism:
Recommendation: The issue is non-significant.
Irradiation of concrete Discussion: The neutron fluence levels and maximum integrated gamma doses experienced by containment concrete du,ing the license Aaina effects:
renewal term is not expected to excead the level at which measurable degradation o concrete strength propgrties occurs (10['n/cm 2
loss of strength &
modulus neutron radiation & 10 rads gamma radiation, respectively for concrete). Thus the issue is non-significant.
NUREG-1611 32 1
08 Struct. Steel &
Issue: NUREG-1557 states that atmospheric Liner corrosion is non-significent for containment steel components if the following conditions e
are met: (a) containment steel components Aoina mechanisn):
fabricated from stainless steel, or for i
components having intact protective coating, 1
Atmospheric or for components having a corrosion allowance corrosion 21/32 inch. A" ti.nitic SS is corrosion resistant. 1 itmospheric corrosion for i
carbon and le alloy steels without protective Aaina effects:
coatings is less th:n 0.5 mils per year or
<l/3? inches for a 60-year period, and (b) the loss of material examination categories E-A, E-P, & E-C of ASME i
Sect, XI, Subsect. IWE are required to be p
performed in conjunction with 10 CFR 50, Appendix J, Type /. leak rate test (Pages B-50 4
& 51 of NUREG-1557).
Bng;nmendation: This issue would be managed with the implement'. tion of IWE throuah 950.55a.
Discussion: IWE, Examination Categories E-A &
E-C, requires periodic examination of i accessible surfaces for containment steel
' structures & its components; Examination Category E-P (Appendix J to 10 CFR-50, Type A l
test), requires a general inspection and an integrated leakage test; and l
650.55a(b)(2)(x)(A) requires an evaluation of inaccessible area; when conditior.s exist in accessible hreas that could indicate the presence of or result in degradation to such inaccessible areas. Atmospheric corrosion may exist when relevant conditions for coated areas including evidence of flaking, blistering, peeling, discoloration, etc., and L
relevant conditions for uncoated areas including evidence of cracking, discoloration, wear, pitting, excessive corrosion, etc.,
occur. If the examination areas are found to be defective, or to be suspectable, the 5
4 augmented examinations of IWE-1240 will apply j
to ensure that cinimum wall thickness is properly evaluated and maintained in l
accordance with IWE-3512.
h a
33 NVREG-1611
-09 Struct. Steel i Issue: (a) NUREG-1557 indicates that local-Liner corrosion would be managed with the implementation of IWE-1240 for steel containment and its common components. NUREG-Aoina mechanism:
1557 indicates that a plant-specific aging program is required to manage the local Local corrosion corrosion of steel containment inaccessible areas and/or embedded carbon steel containment components (Pages B-53 and B-55 of NUREG-Aoina effects:
1557). (b) NUREG-1557 indicates that local corrosion is non-significant-for concrete Loss of material containment liners and anchors if the following conditions are met: Corrosion of the liner plate is mitigated by protective coatings on the interior surface, and by the alkalint environment between the exterier surface of the liner plate and the concrete structure. Stainless steel is corrosion resistant (Pages B-54 of NUREG-1557).
Recommendation: All access;ble areas would be managed with implementing IWE through 650.55a.
However, the management of potential local corrosion of iniccessible areas of structural steel and liner when conditions in accessiole areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: IWE, Examination Categories E-A, E-C, E-D, and E-G, provides periodic examination of accessible areas to uncover structural degradation. This inspection would detect local corrosion regardless whether the-conditions in (b) are met. IWE-1240 specifies augmented inspections for areas likely to experience accelerated degradation and aging.
550.55a(b)(2)(x)(A) requires an evaluation of acceptability of inaccessible areas when conditions exist in accessible areas that couN indicate the presence of or result in deseadation to such inaccessible areas.
Howev'4, the management of local corrosion of inaccessible areas when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated.
NUREG-1611 34
10 Struct. Steel &
Issue: NUREG-1557 states that elevhted Liner temperature is non-significant for PWR containment structural steel and liner (Page B-33 of NUREG-1557). However, NUREG-1557 does Aoina mechanism:
not address elevated temperature effects on BWR containment steel liner (Page B-56 of Elevate 4 NUREG-1557).
temper.turo Recommandation: The issue is non-significant.
Aaina effects:
Discussion: OperatingtemperatyreswithinBWR containment structures,C (600,6 C (150 F) which are <6 Loss of strength &
are well below the 316 F) level at modulus which the structural integrity of rebar/ concrete combination begins to be significantly affected. Thus the issue is non-significant. This conclusion is also applicable to the BWR containment steel liner.
l i
11 E ruct. Steel &
Issue: NUREG-1557 states that irradiation of Liner steel is non-significant for containmant structural steel & liner (Page B-57 of NUREG-1557).
Aaina mechanism:
Recommendation: The issue is nor.-significant.
Irradiation of l
steel Discussion: The neutron fluence levels &
t maximum integrated gamma doses inr.urred by containment components, includir.g cor.tainment
.Taina effects:
steel & liners throughout the license renewal period are not expected to exceed the level at Loss of fracture whichpeasgrabledegradationoccurs toughness (2x10' n/cm for all components made of carbon f
steel, steinless steel, and liner plate)
Thus I the issue is non-significant.
35 NUREG-1611
1 12 Struct. Steel &
Inga: (a) NUREG-1557 indicates that SCC is
- Lingr, non-significant for containment components, including penetration sleeves, bellows, and vent line bellows if the following conditions Aaina mechanirm:
are met: for austenitic SS containment components that are only excosed to the Stress corrosion containment or reactor building environment or cracking (SCC) their normal operational stress levels are less than materials yield strength or fracture mechanics analysis has established that cracks Aoina effects:
do not propagate; and for high strength bolts if material yield strength is <1034 f1Pa(<150 Crack initiation &
ksi). (b) SCC would be managed for suppression growth chamber shell interior surface by im)lementing 10 CFR 50, Appendix J integrated lea ( rate test to maintain liner integrity (Page B-63 of NUREG-1557).
Recow endation: This issue would be managed by Examination Categories E-C & E-F of Subsection IWE and Appendix J to 10 CFR 50. In addition, an augmented VT-1 visual examination of bellows bodies should be performed using enhanced techniques qualified for detecting stress corrosion cracking in bellows bodies.
Discussion: IWE Examination Cmgory E-F provides periodic surface examination of pressure retaining dissimilar metal welds for dissimilar metals and could detect SCC. IWE, Examination Category E-B, provides periodic visual examination of pressure retaining welds for containment penetrations. Also any leakage associated with the steel liner, including suppr ession pool liner, due to through-wall cracks resulting from SCC would be detected by periodic Appendix J leak rate test & remains within the limits of plant specifications or Subsection IWE. Although 550.55a indicates that Examination fategories E-B & E-F are optional during the current term of operation, these examinations should be performed for license renewal to demonstrate that no SCC has been initiated. In addition, since occurrences af transgranular stress corrosion cracking I have been identified in operating plants on SS oellows (Reference 15], an augmented examination on the surface areas of bellows bodies should be performed so that cracking would be detected.
NUREG-1611 36
13 Reinforcina Steel Issue: NUREG-1557 states that corrosion of (Rebar) embedded steel is non-significant for concrete structures not exposed to aggressive environment (pH<ll.5 or chlorides >500 ppm);
Aaina mechanism:
or for concr?te exposed to aggressive environment but has relatively high strength Corrosion of (27.6 MPa (4 ksi)) and low water-to-cement embedded steel ratio (0.35-0.45), adequate air entrainment (3-6%), low permeability, and are designed in accordance with ACI 318 or ASME Section III, Aoina effects:
Division 2. NUREG-1557 indicates corrosion of embedded steel for concrete structures below Loss of bond & loss grade exposed to aggressive ground water (pH of material
<5.5, chloride >500 ppm, & sulfate >l500 g pm) should be examined in accordance with IWL and management of inaccessible areas should be justified on a case by case basis. Also the NRC staff considers that potential degradation due to chloride corrosion (e.g., ground water chemical attack) of containments should be addressed (Page B-52 of NUREG-1557).
l Recommendatign: The issue would be managed with the implementation of IWL through 650.55a. However, the management of potential I
corrosion of inaccessible areas of embedded steel when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be justified on a plant-specific basis.
Discussion: IWL, Examination Category L-A, requires periodic examination of accessible areas and 650.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Corrosion of embedded steel results in cracking and spalling of concrete and would be detected by inspections, regardless of whether the above conditions are met. However, the management of potential corrosion of inaccessible areas of embedded steel, when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas needs to be evaluated. This would also address the staff's concern on chloride corrosion.
37 NUREG-1611
\\
14 Reinf. Steel 1
.lssue: NUREG-1557 states that elevated Prestr. Tendons temperature is non-significant for concrete containment reinforcing steel and for concrete containment prestressing tendons (Pac
-56 of Aaina mechanism:
Elevated Res.opnendatiqn: The issue is non-significant, temperature except for prestressed tendons. The tendon surveillance program should be augmented to include additional tendons based on their sun Aaino effects:
exposure or proximity to hot penetrations.
Loss of strength &
Discussion: Operatingtemperatyresw containment structures are <66 C(150,ithin BWR modulus F) which are well below the 316 C (600"F) level at which the structural integrity of rebar/ concrete combination begins to be significantly affected. Additionally, concrete containment prestressing tendgns are,F). Thus normally subjected to temperatures <60 C (140 the issue is non-significant for reinforcing steel. However, increase in temperature increases prestress loss in prestressed 1
tendons. Prestress losses increased from 8% to j
14% when the temp 20 C (68 F) to 32,eratupe was increased from 0
0 C (90 F) (Reference 16).
Thus, temperatures due to sun exposure or proximity to hot penetrations may-increase the prestress loss in tendons. The tendon surveillance program described in Regulatory Guide 1.35 (Reference 10] is based on a small j
sample size, that is, a 4 percent random sample including a repeat tendon. Tendons-subject to warm temperatures may not be-tested because of this small sample size. The tendon surveillance program should be augmented to include additional tendons. These additional tendons should be selected based on their sun exposure or proximity to hot penetrations.
NUREG-1611 38
15 Reinf. Steel &
Issue: NUREG-1557 states that irradiation of-Prestr. Tendons steel is non-significant for concrete structures reinforcing steel (including basemat reinforcing steel) and concrete Aoina mechanism:
containment prestressing tendons (Page B-57 of NUREG-1557),
Irradiation of steel Recommendation: The issue is non-significant.
Discussion: The aeutron fluence levels &
Aaina effects:
maximum integrated gamma doses incurred by containment components, including rebars &
Loss of fracture prestressed tendons for both the current and toughness license renewal period are not expected to exceed the level at whichgeasgrable degradation occars, (4x10 n/cm forconcgete 2
containment prestressing tendons; 2 x 10 n/cm for all components made of CS, SS including rebar, and liner). Thus the issue is non-significant.
l l
f 39 NUREG-1611 l
-16 Containment issue: NUREG-1557 states that fatigue is non-Structures &
significant for containment structures and its Comoonen11 components, except for the penetration sleeves and bellows. NUREG-1557 also indicates that fatigue is an unresolved issue for containment Aoina mechanism:
penetration sleeves and penetration bellows and fatigue damage may not be detectable by a Fatigue leak rate test (Pages B-58 & B-59 of NllREG-1557).
Aoina effects:
Recommendation: Fatigue is non-significant for containment structures and its components Cumulative fatigue except for the penetration sleeves and damage bellows. Fatigue of containment penetration sleeves and penetration bellows is a " time-limited aging analysis" and must be evaluated in accordance with license renewal rule 554.21(c).
Discussion: Fatigue is non-significant for containment concrete, reinforcing steel, prestressing system components, steel liners, and free-standing steel containments, because-theyaredesignedtohayegoodfatigue strength properties (10 cycles) of below yield load in accordance with ASME Section III, Division 2, or ACI-318, and ACI 215R-74 codes.
Containment penetration sleeves and penetration behows are designed to Section III of tlm ASME Code which requires a fatigue analysis based on an assumed number of cycles.
This fatigue analysis is a " time-limited aging analysis" and must be evaluated in accordance with license renewal rule 554.21(c) to ensure that the effects of aging on the intended functions will be adequately managed for the period of extended operation.
I NUREG-1611 40
17
@ntainment Issue: NUREG-1557 ir.dicates that for BWR 5tructure & its containment concrete basemat bearing on soil Concrete Basemat or piles, or experiencing significant changes in ground water conditions, a settlement monitoring program is required to ensure that Aaina mechanism:
the dif ferential settlement does not exceed the design criteria for the containment Settlement throughout the license renewal term (Page B-62 of NUREG-1557).
Aoina effects:
Recommendation: The issue would be managed by establishing a settlement monitoring program Cracks, distortion, which would ensun that differential increase in settlement of containment basemat does not component stress exceed the design criteria for a containment level structure and its basemat bearing on soil or piles, or experiencing significant changes in ground water conditions.
Discussion: Effects of differential settlement are potentially significant for a containment structure and its concrete basemat that is resting on soil or piles, or experiencing significant changes in ground water conditions. Subsection IWL does not address the effects of settlement. Because the effects l
of settlement could cause cracks and l
distortion of concrete basemat and could i
result in increasing stress levels greater than the original design basis in the basemat and other parts of the containment structure.
A settlement monitoring program could ensure that the differential settlement does not exceed the design criteria for the containment structures throughout the license renewal term. A settlement monitoring program should be provided to mariage settlement for containment basemat bearing on soil or piles, or experiencing significant changes in ground water conditions for the period of extended operation.
41 NUREG-1611 t
18 Containment issue: NRC Information Notice 97-11 indicates Structure & its that erosion of cement from porous concrete Concrete basemat could be potentially significant for porous concrete subfoundations below concrete basemat if subfoundation layers of porous concrete are Aoino mechanism:
used in the construction of concrete basemat with the presence of underground water.
Erosion of cement Recommendation: For these applicable plants, the management of potential erosion of cement Aoino effects:
from porous concrete needs to be justified on a plant-specific basis.
Loss of strength Dhgs ilsn: 650.55a(b)(2)(ix)(E) requires an evaluation of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessibla sreas.
However, the management of potentioi erosion of cement from porous concrete of inaccessible areas of containment concrete basemat, when conditions in accessible areas may not indicate the presence of or result in degradation to such inaccessible areas, needs to be evaluated on a plant-specific basis.
NUREG-1611 42 l
19 Containment hiuq: NUREG-1557 indicates that strain aging Structure & its is non-significant for steel containment Componenti structures (including common components. such as penetration sleeves, penetration tellows, personnel airlock, equipment hatches, and CRD Aoina mechanism:
hatch), and concrete containment steel components (including vent lines, vent line Strain aging of bellows, and drywell head) that meet the carbon steel following conditions: Dynamic strain aging is i
non-significant for containment steel components that do not allow loads to exceed Aoina effic_ti:
the elastic limit. Static strain aging is non-significant for containment steel components Loss of fracture that were not cold worked; or if cold worked toughness during the forming process, the plates were normalized or stress relieved or both after forming with minimal (<5%) subsequent cold working (Page B-61 of NUREG-1557).
Recommendation: This issue is non-significant.
{1i scussion: Dynamic strain aging is not expected in the carbon steel components of containments during their service life, since the strains associated with the design service loads are below the elastic limit of the material. The BWR containment is made from low-carbon steel, and the steel is normalized l
or stress relieved or both following plate rolling. Further, strain aging requirrs stressing of the material to above its yield stress, and aging at temperatures above 93 C (200 F). Carbon-relatgd strain aging at 0
temperatures below 93 C (200 F) is normally negligible due to the low solubility of carbon
(
in this temperature range. The BWR containment has a maximum temperature during normal g
operation of about 66 C (150 F), and loading conditions do not produce service stresses in the range of the material yield strength. Thus strain aging is non-significant for containme7t steel structure and its steel components.
43 NUR:G-1611 l
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- 20-
, Cone.-Containment Issue: NUREG-1557 indicates that loss of fIcstr. Tendons prestress due to stress relaxation, shrinkage creep, etc., would be a reduction of design margin and could be potentially significant Aoina nechanism:
for_prestressing tendons for license renewal (Page B-62 of NUREG-1557).
-Stress relaxation-of prestressing Recommendation: Loss of prestress for wire, shrinkage prestressing-tendons would be managed with the creep, anchorage implementation of Subsection IWL through seating losses, and 950.55a. In addition, the tendon prestress tendon friction evaluation is a " time-limited aging analysis" and must be evaluated in accordance with the license renewal rule,10 CFR 654.21(c).
Ag.ina effects:
Discussion: Subsection IWL, Examination Loss of prestress Category L-8, and 550.55a inspections would be able to detect. potential loss of prestress for prestressing tendons, for example: IWL-2522 provides examination method for tendon force measurements; IWL-3221 provides acceptance standard for measuring tendon force; 650.55a(b)(2)(fx)(B) states-that "when evaluation of consecutive surveillances of prestressing forces for the same tendon...
Indicates a trend of prestress loss....,
an evaluation shall-be performed;" repair and replacement are addressed in IWL-4000 and IWL-7000, respectively. In addition,-the recommendations of Regulatory Guide 1.35
[ Reference 10] were incorporated into the 1992 Edition with the-1992 Addenda of Subsection-IWL and-650.55a(b)(2)(ix)(A)-(D), the final l
rule, dated August 8, 1996 (61 FR 41304). Thus I
the " loss of prestress" for prestressing tendons would be managed with the implementation of IWL through 550.55a.
Further, the tendon prestress evaluation is a
" time-limited aging analysis" and must be evaluated for. renewal to demonstrate that the prestressing force will meet theLdesign L
requirements at_ the end of 60 years in accordarce with license renewal rule l
654.21(c).
l l
l
(
l NUREG-1611 44 l
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21 Conc. Containment Issue: NUREG-1557-states that corrosion of Prestr. Tendons tendons is an unresolved issue.in that the NRC staff is concerned that a large amount of grease leakage can degrade concrete strength.
Aoina mechanism:
IWL (1992 Edition with 1992 Addenda) lacks
- ertain criteria contained in RG 1.35. These Corrosion of criteria are addressed in 10 CFR 50.55a final tendons rule, dated August 8, 1996, Section 50.55a(b)(2)(ix)(A)-(D) on issues such as failed wires and tendon sheathing filler Aaina effect.1:
grease conditions. Also, anchor heads have failed in prestressed concrete containments.
Loss of material NUREG-1557 states that prestressing tendons and tendon anchorage hardware should be examined-in accordance with the provisions of RG 1.35 for prestressed concrete containments (Page B-55 of NUREG-1557).
Recommendation: The issue is resolved with the l
implementation of Subsection IWL through l
950.55a.
l-Discussion: Subsection IWL, Examination l:
Category L-8, and 650.55a inspections would be able to detect corrosion of prestressing tendons. For example: IWL-2525 provides methods for examination of corrosion protection medium and free water; IWL-3221 provides acceptance standard for corrosion protection medium; IWL-2524 provides visual examination of Nndon anchorage areas; IWL-3221 provides acceptance standard for inspection of tendon anchorage areas; 950.55a(b)(2)(ix)(A) states that " grease caps that are accessible must be visually examined to detect grease leakage or grease cap deformations.......
that indicates deterioration of anchorage hardwarc." In addition, the recomrtandations of Regulatory Guide 1.35 [ Reference 10) were incorporated into the 1992 Edition with the 1992-Addenda of Subsection IWL and 550.55a(b)(2)(ix)(A)-(D),
the final rule, dated: August 8, 1996 (61 FR 41304). Thus the " Corrosion of prestressing tendons" would be managed with the implementation of IWL through 650.55a.
45 NUREG-1611
22 Containment Issue: Mechanical wear could be potentially Pressure Retainina significant for components that are subject to Components relative sliding or rotating motion and that are susceptible to fretting and/or lockup.
(Page B-60 of NUREG-1557).
Aoina mechanism:
Recommendation: Mechanical wear for BWR Mechanical wear containment pressure retaining components &
their supports would be managed with the implem.ntation of Subse tions IWE and IWF.
Aaina effects:
Discussion: Inspection and mitigatio7 of Fretting, Lockup mechanical wear conducted in accordance with the provisions of Subsections IWE & IWF would ensure that the integrity of containment pressure retaining components and their supports is maintained throughout the license renewal term. IWE, Examination Categories E-D
& E-G, provides periodic examinations for the pressure retaining components, (including airlock, equipment hatch, CRD hatch & drywell head). IWE, Examination Category E-P (Appendix J to Part 50, Type B test), would detect local leaks for those components. The supporting components such as downcomer bracing, column &
saddle supports, seismic restraints & vent system supports are considered MC component supports which are periodically examined by Examination Category F-A of Subsection IWF.
Thus any potential mechanical wear degradation would be detected by the implementation of IWE and IWF for those pressure retaining components and their supports.
- Evaluation is based on:
(1) the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of Section XI of the ASME Code; (2) the-1989 Edition of Section.XI of the ASME Code including Appendix VII, " Qualification of Nondestructive Examination Personnel for Ultrasonic Examination," and Appendix VIII (1989 Addenda),
" Performance Demonstration for Ultrasonic Examination Systems;"
(3) the final rule on 10 CFR 50.55a, Codes and Standards, dated August 8, 1996 (61 FR 41303); and (4) NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," dated October 1996.
Note (1) " Concrete interaction with aluminum" is not addressed in NUREG-1557 for BWR containment. However, this item is evaluated and considered not an issue for license renewal for PWR containment. Thus this item is considered not an issue for BWR containment.
NUREG-1611 46
APPENDIX A - IMPLEMENTATION HIGHLIGHTS OF SUBSECTIONS IWE AND IWL THROUGH 10 CFR 50.55a Subsection IWE provides rules for inservice inspection, repair, and replacement of Class MC pressure retaining components and their integral attachments and of metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments in light-water cooled power plants.
Subsection IWL piovides rules for inservice inspection and repair of the reinforced concrete and the post-tensioning systems of Class CC components.
Licensees will be required to incorporate Subsection IWE and Subsection IWL into their inservice inspection (ISI) program.
Licenees will be required to implement the containment examinations in accordance with Subsections IWE and IWL as endorsed by 550.55a by September 9, 2001.
In endorsing Subsections IWE and IWL,10 CFR 50.55a sets forth additional requirements to assure that the critical areas of containments are periodically inspected to detect and take corrective action for defects that could compromise a containment's structural integrity.
These additional requirements include:
(a) Four modifications specified in 550.55a(b)(2)(x) for examination of metal containments and the liners of concrete containments.
These are: (1) Section 50.55a(b)(2)(x)(A) states that the licensee shall evaluate the acceptability f
of inaccessible areas of metal containments and the liners of concrete l
containments (Class MC), v: hen conditions exist in accessible areas that couM t
indicate the presence of or result in degradation to such inaccessible areas; (2) Section 50.55a(b)(2)(x)(B) permits alterrdive lighting and resolution requirements for remote visual examination of the containment; (3)- Section 50.55a(b)(2)(x)(C) makes the examination of pressure retaining welds and pressure retaining dissimilar metal welds optional; and (4) Section 50.55a(b)(2)(x)(0) is added to provide an alternative sempling plan.
(b) Five modifications specified in 950.55a(b)(2)(ix) for examination of concrete containments. These. modifications must be implemented when using Subsection IWL.
Four of these issues are identified in Regulatory Guide 1.35, Revision 3, but are not addressed in the referenced Subsection IWL.
The five modifications are: (1) Section 50.55a(b)(2)(ix)(A) requires that grease caps which are accessible be visually examined to detect-grease leakage or grease cap deformation; (2) Section 50.55a(b)(2)(ix)(B) requires the preparation of an engineering evaluation report when consecutive surveillances indicate a trend of prestress loss to below the minimum prestress requirements; (3)
Section 50.55a(b)(2)(ix)(C) requires that an evaluation be performed for instances of wire failure and slip of wires in anchorages; (4) Section 50.55a(b)(2)(ix)(D) addresses sampled sheathing filler-grease and reportable conditions; and (5) Section SG.55a(b)(2)(ix)(E) requires that licensees evaluate the acceptability cf inaccessible areas of concrete containments when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
(c) One limitation specified in l50.55a(b)(2)(vi) for effective edition and addenda of Subsection IWE and Subsection IWL.
It states that the 1992 Edition with the 1992 Addenda of Subsection IWE.and Subsection IWL shall be used when 4
47 NUREG-1611 n
l
-performing containment examinations as modified and supplemented by the-requirements described in 550.55a(b)(2)(ix) and 550.55a(b)(2)(x),
respectively.
(d) One-clarification 4specified in 550.55a(b)(2)(x)(E).
It states that a general visual examination as required by Subsection IWE shall be performed once each-period.-
~
9 1
l 3.-
t k
NUREG-1611 48
.--rz
APPENDIX B - LIST OF PWR CONTAINMENT COMPONENTS
- 2. CONCRETE CONTAINMENTS (REINFORCED / PRESTRESSED)
Concrete Dome Dome Peinforcing Steel Concrete Containment Wall Above Grade
- . Containment Wall Reinforcing Steel Above Grade Concrete Containment Wall Below Grade Containment Wall Reinforcing Steel Below Grade Concrete Basemat Basemat Reinforcing Steel Containment Liner Interior Surface Containment Liner Above Grade Exterior Surface Containment Liner Below Grade Exterior Surface Basemat Liner Interior Surface l
Basemat Liner Exterior Surface Liner Anchors Above Grade Liner Anchors Below Grade
- 2. FREE-STANDING STEEL CONTAINMENT WITH FLAT BOTTOM & AN ICE CONDENSER Dome Shell Interior Surface Dome Shell Exterior Surface Cylindrical Shell Interior Surface Cylindrical Shell Exterior Surface Embedded Shell Region Concrete Basemat Basemat Reinforcing Steel Basemat Liner Liner Anchors
- 3. FREE-STANDING CYLINDRICAL & SPHERICAL STEEL CONTAINMENT WITH ELLIPTICAL BOTTOM Containment Shell Interior Surface Containment Shell Exterior Surface a
Embedded Shell Region Sand Pocket Region
- 4. CONCRETE CONTAINMENTS PRESTRESSED ONLY Prestressing Tendons
- 5. COMMON COMPONENTS Penetration Sleeves Penetration Bellows Personnel Airlock Equipment Hatches
Reference:
Page B-45 of NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," dated October 1996.
49 NUREG-1611
APPENDIX C - LIST OF BWR CONTAINMENT COMPONENTS I. MARK I CONCRETE CONTAINMENTS
- Drywell Liner Interior Surface Drywell-Liner Exterior Surface Torus Liner Interior Surface Torus Liner Interior Surface at Waterline Torus Liner Exterior Surface Liner Anchors Drywell Concrete I
Torus Concrete Drywell Concrete Reinforcing Steel Torus Concrete Reinforcing Steel Vent Lines Vent Line Bellows Vent Hcaders-Downcomers and Bracing Vent System Supports Drywell Head
- 2. MARK I STEEL CONTAINMENTS Drywell Interior Surface Drywell Exterior Surface J
Drywell Head Embedded Shell Region Drywell Support Skirt Sand Pocket Region Torus Interior Surface Torus Interior Surface at Waterline Torus Exterior Surface Torus Ring Girder Vent Lines Verit Line Bellows Vent Header Downcomers and Bracing
- Drywell Exterior Shell with Compressible Material Vent System Supports Torus Seismic Restraints Torus Support Columns / Saddles ECCS Suction Header Ocean Plant with Uncoated CS Surfaces Uncoated Submerged CS Surfaces
- 3. MARK II CONCRETE CONTAINMENTS Drywell Liner Interior Surface Drywell Liner Exterior Surface Suppression Chamber Liner Interior Surface Suppression Chamber Liner Interior Surface at Waterline Suppression _ Chamber Liner Exterior Surface Liner Anchors Liner Region Shielded by Diaphragm Floor Containment Concrete NUREG-1611 50
- 3. MARK II CONCRETE CONTAINMENTS (Continued)
Concrete Containment Reinforcing Steel Drywell Head Downcomer Pipes and Bracing Concrete Basemat Basemat Liner Basemat Reinforcing Steel Prestressing Tendons and Ducts
- 4. MARK II STEEL CONTAINMENTS Drywell Interior Surface Drywell Exterior Surface Drywell Head Suppression Chamber Interior Surface Suppression Chamber Exterior Surface a
Suppression Chamber Interior Simface at Waterline Region Shielded by Diaphragm Fioor Embedded Shell Pegion Sand Pocket Region Support Skirt Downcomer Pipes and Bracing Drywell Exterior Shell with Compressible Material Ocean Plant with Uncoated CS Surfaces Uncoated Submerged CS Surfaces
- 5. MARK III CONCRETE CONTAINMENTS Containment Liner Interior Surface Containment Liner Exterior Surface Suppression Chamber Liner or Cladding Interior Surface Suppression Chamber Liner Exterior Surface Concrete Containment Wall Above Grade Concrete Containment Wall Below Grade Concrete Dome Basemat Liner Concrete Basemat Liner Anchors Containment Wall Reinforcing Steel Dome Reinforcing Steel a
Basemat Reinforc'ng Steel 6.-MARK III STEEL CONTAINMENTS Containment Shell Interior Surface Containment Shell Exterior Surface Suppression Chamber Shell Interior Surface Suppression Chamber Shell Exterior Surface Basemat Liner Liner Anchors Concrete Basemat Basemat Reinforcing Steel Concrete Fill in Annulus Embedded Shell Region 51 NUREG-1611 l
__m
- 7. CONTAINMENT COMMON COMPONENTS Penetration Bellows Penetration Sleeves Dissimilar Metal Welds Personnel Airlock Equipment Hatches CR0 Hatch
Reference:
Pages B-65, B-53, and B-52 of NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," dated October 1996.
I NUREG-1611 52
NRC FORM 334 U.3, NUCLEAR REoVLATORY CoMMISS60N
- 1. REPORT NUMBE'A QM (Ass 4gned py teRC. Add Vot, Supp., Rev E3E BIBLIOGRAPHIC DATA SHEET
"*^"*"*"""""*" " *"d tsee m.tucacn. on e. cvnse)
- 2. TITLE AND SUBTITLE NUREG-1611 Aging Management of Nuclear Power Plant Containments for Ucense Renewal 3.
DATE REPORT PUBLISHED l
uenm vEAR September 1997
- 5. AUTHOR (S)
- 6. TYPE OF REPORT W. C. Uu, P. T. Kuo, S. S. Lee Regulatory 7.PERcoCOVERED pebasvecease)
- 8. PERFORMING ORoANIZATION NAME ANO ADORESS rr NRc, pave ovman, omce or Rep n, u S. AAcher Reguisaary comsvuon, and madno eness, acorescar.
pave nome end msang eness }
Division of Reactor Program Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington. DC 20555-0001
- 9. SPONSORING ORGANIZATON - NAME AND ADORESS (#NRC, lyse *Se ne as e60=ot #conteckr. pave NRC Dnsen, Orce or Regen, U S Nucasar Regudstry Comnwssen, end mee4 anoes }
S:me as 8. above.
I
- 10. SUEMENTARY NOTES
- 11. AB3 TRACT (200 mada or ess)
In 1990, the Nuclear Management and Resources Council (NUMARC), non the Nuclear Energy Institute (NEI), submitted for NRC review, the industry reports (irs), NUMARC Report 90-01 and NUMARC Report 90-10, addres.ing aging management issues associated with PWR containments and BWR containments for license renewal, respectively. In 1996, the Commission amended 10 CFR 50.55a to promulgate requirements for inservice inspection of containment structures. This rule amendment incorporates by rzference the 1992 Edition with the 1992 Addenda of Subsections IWE and IWL of the ASME Code addressing the inservice inspection of metal containments / liners and concrete containments, respectively.
The purpose of this report is to reconcile the technical information and agreements resulting from the NUMARC IR reviews which are generally described in NUREG-1557 and the inservice inspaction requirements of subsections IWE and IWL as promulgated in $50.55a for license renewal consideration.
This report concludes that Subsections IWE and IWL as endorsed in $50.55a rare generally consistent with tha technical agreements reached during the IR reviews. Specific exceptions are identified and additional svaluations and augmented inspections for renewal are recommended.
- 12. KEY WORoS CESCRIPTORs (Uss words or preses met we asses researchers a locanno me report) 13 AvAusiuTY siArEMENT aging management unlimited lic'nse renewal 14 secuRrrycLAssFicanoN containment structures gas %)
Subsections fWE and IWL unclassified 10 CFR 50.55a F**
- 10 CFR Part 54 unclass4fied
- 15. NUMBER OF PAGEU
- 16. PRICE NRC FORW 335 Q 40)
TNe form was e#ectrorweasy produced by Eine Federet Ferr% inc
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NURE';-M11
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