ML20134Q324
| ML20134Q324 | |
| Person / Time | |
|---|---|
| Issue date: | 02/24/1997 |
| From: | Samson Lee NRC (Affiliation Not Assigned) |
| To: | FRANCE |
| References | |
| NUDOCS 9702260388 | |
| Download: ML20134Q324 (23) | |
Text
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,.y February 24, 1997 SMIRT-97 B.P. N 64' 91193 GIF-SUR-YVETTE CEDEX France l
Dear SMIRT Secretary:
Attached are three manuscripts for the 14th International Conference on:
Structural Mechanics in Reactor Technology (SMIRT-14), Lyon, France, August 17-22, 1997. They are:
" Aging Management of Reactor Coolant System Piping for License Renewal"
" Technical Information from Industry Reports Addressing License Renewal"
" Nuclear Power Plant Generic Aging lessons Learned (GALL)"
SMIRT-14 requests a fee of 1500 French Francs per manuscript.
Ms. Yvette Brown is arranging payment for these three manuscripts and her telephone number is (301)415-6507.
Sincerely,
/s/
Sam Lee Attachments:
As stated i
DISTRIBUTIONS:
(w/ attachments)
Central Files PUBLIC PDLR R/F, 0-11D23 (w/o attachments)
Yvette Brown, T-712 C)3, /
Vicki Yanez, T-6E7 C. Regan, 0-11023 P.T. Kuo, 0-11D23 S. Newberry, 0-11D23 g 4 y C)
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PDLR PDLR:S' ADM/DC/ FAB X-I Slee $ b PTKuo1 YBrowny[6 l
02/24/97 02hT/97 02/19/971 0FFICIAL RECORD COPY
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AGING MANAGEMENT OF REACTOR COOLANT SYSTEM PIPING FOR LICENSE RENEWAL S. Lee and P. T. Kuo Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission, Washington, D.C., USA l
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ABSTRACT The U. S. Nuclear Regulatory Commission (USNRC) has reviewed the Babcock and l
Wilcox Owners Group (B&WOG) topical report addressing the reactor coolant l
system (RCS) piping for license renewal. The USNRC staff found that, in general, the existing plant programs are adequate for managing the effects of j
aging of the RCS piping components. However, there are a few situations where additional aging management programs would be necessary for license renewal.
1.
INTRODUCTION Licenses to operate nuclear power plants 'in the United States are issued by the USNRC for a fixed period of time not to exceed 40 years.
However, these i
licenses may be renewed by the USNRC for an additional period not to exceed 20 years. The revised license renewal rule sets forth the requirements for the renewal of operating licenses for commercial nuclear power plants [1].
Applicants for license renewal are required to perform an assessment for certain-important plant structures and components to ensure that the effects
.of aging will be adequately managed so that their intended functions will be maintained for the period of extended operation.
The B&WOG submitted topical report BAW-2243, " Demonstration of the Management of Aging Effects for the Reactor Coolant System Piping," [2] for USNRC staff review and approval. The topical report evaluated the aging management of the RCS piping for license renewal for their Generic License Renewal Program (GLRP) member plants. The purpose of the topical report is to provide a l
technical evaluation of the effects of aging on the RCS piping and demonstrate that the aging effects for the RCS piping are adequately managed for the l
period of extended operation associated with license renewal.
The USNRC has reviewed and approved the BaWOG topical report which is the first technical assessment of aging management of a plant component for license renewal in accordance with the revised license renewal rule [1].
This 719-2
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l paper provides an' overview of the review performed and identifies those existing programs and the new programs accepted for the period of extended
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operation.
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RCS PIPING COMP 0NENTS AW INTEMED FUNCTIONS j
Babcock and Wilcox (B&W) plants are 2-loop pressurized water-reactors (PWRs).
l The RCS components within the scope of the B&WOG topical report are piping
- components of the RCS within the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI inservice inspection (ISI) program for Class I components [31 The RCS piping components addressed in the topical report are piping, vaive bodies, and bolting.
Figure 1 provides a schematic of the B&W RCS piping within the scope of the topical report [2]. The piping size' ranges-from the 91 cm (36 in) hot leg.to the 1 cm (1/2 in) instrumentation, vent, drain, and sampling lines. All bolting within the scope of the topical report are less than 5 cm (2 in) in diameter. The B&W plant hot leg and cold leg piping are fabricated from carbon steel internally clad with stainless steel. The remaining piping is generally fabricated from stainless steel. The valve bodies are generally fabricated from stainless steel.. The bolting is fabricated from low-alloy steel and stainless steel. There are also Alloy 600 safe-ends.
The intended function of the RCS piping components is the maintenance of the structural integrity of the reactor coolant pressure boundary under normal, upset, emergency, and faulted conditions, in.accordance with the plant's current licensing basis.
3.
APPLICABLE AGING EFFECTS
.The B&WOG' reviewed relevant plant-specific and industry-wide operating experience of the RCS piping components relating to aging. They identified cracking, boric. acid corrosion of carbon steel and low-alloy steel, thermal l
aging embrittlement of cast' stainless steel, and loss 'of bolting preload as applicable aging ' effects. Table I summarizes the applicable aging effects described in the B&WOG topical report that need to be managed for license renewal for the RCS piping components in the B&WOG GLRP member plants.
.The B&WOG also indicated that fatigue needs to be managed for license renewal.
However, they indicated that fatigue is outside the scope of the topical report and will be evaluated on a plant-specific basis for license renewal under the " time-limited aging analysis" provision of the license renewal rule
[1].
4.
AGING MANAGEMENT PROGRAMS FOR LICENSE RENEWAL There are existing aging management programs for the RCS piping components and they are:
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ASME Section XI Class 1 ISI I31 - Plants have ISI programs on 10-year intervals based on the ASME Section XI Code.
The Class 1 ISI program is described in Subsection IWB of Section XI and is divided into
" Examination Categories." The B&WOG will also implement the mandatory Appendices VII and VIII of ASME Section XI for license renewal.
)
Resnonse to Generic Letter 88-05 I41 - PWR licensees have responded to the USNRC's generic letter describing their programs for mitigating the effects of boric acid corrosion of external surfaces of carbon steel reactor coolant pressure boundary components.
Proaram evaluated in Generic Letter 85-20 I51 - B&W plants are implementing a B&WOG task force developed program to manage potential cracking of the B&W high pressure injection nozzle thermal sleeves.
Information resultina from Information Notice 90-10 I61 - PWR licensees have evaluated the information on the potential for primary water stress corrosion cracking of Alloy 600 materials and have considered actions, as appropriate.
Response to Bulletin 82-02 I71 - PWR licensees have responded to the USNRC's bulletin describing their maintenance procedures for threaded 3
fasteners in the components of the reactor coolant pressure boundary.
Technical soecification RCS leakaae limits - Plant technical specifications contain surveillance requirements to monitor and trend l
RCS leakage, specific limits for identified and unidentified RCS leakage, and no leakage from the reactor coolant pressure boundary.
The B&WOG topical report relied on these existing aging management programs to continue to manage aging of the RCS piping components for the period of extended operation.
l The USNRC staff reviewed the specific applicable aging effects against the specific elements of the existing aging management programs. The USNRC staff found that, in general, the existing plant programs are adequate for managing the effects of aging of the RCS piping components to ensure that the intended function of the RCS piping components would be maintained for' license renewal.
l However, there are a few situations where additional aging management programs would be necessary for license renewal. These augmented /new programs are:
Auamented inspection of Allov 82/182 clad hot lea seament - In B&W l
plants, a 24 cm (9-1/2 in) flow meter section of the hot leg is internally clad with Alloy 82/182. Operating experience shows that Alloy 82/132 material may be susceptible to cracking.
Cracking of the cladding could potentially lead to underlying base metal degradation.
l To manage the potential cracking of the Alloy 82/182 cladding for l
license renewal, the B&WOG will perform a one-time volumetric l
(ultrasonic) inspection of the Alloy 82/182 clad flow meter section of the hot leg at or near the end of the current license term.
This one-time-inspection could be performed by the B&WOG at only one selected 719-4 i
site if the B&WOG justifies that the inspection results bound all B&WOG GLRP member plants.
Auomented insoection of small bore oloina - For RCS piping less than 10 cm (4 in), the ASME Section XI ISI is based on a surface examination of the piping outside surface and leakage detection under " Examination Category B-P."
For RCS piping less than or equal to 2.5 cm (1 in), the ASME Section XI ISI program is based solely on leakage detection. A volumetric inspection can usually detect significant cracking originating from the inside surface. However, a crack originating from the inside surface of a pipe but has not penetrated through the pipe wall can not be detected by either surface examination of the outside surface or leakage test.
Piping with a part-through wall crack and therefore, not leaking, may not have the structural integrity to ensure the reactor coolant pressure boun:lai,r function of the piping components for all design loads. Cracked, but not leaking, piping could fail during a design loading condition such as a seismic event.
- Further, aging could be a common cause of degradation of piping in a similar service environment.
Thus, the B&WOG will perform additional sample i
l inspections of small bore, that is, less than 10 cm (4 in), piping for license renewal.
l New orocram to manaae loss of touahness of cast stainless steel - Valve l
bodies fabricated from cast stainless steel may be subject to thermal j
aging resulting in a loss of fracture toughness.
However, there is currently no procedures in ASME Section XI to evaluate flaws in cast stainless steel materials. The B&WOG compared the lower-boe d toughness property for aged cast stainless steel recently developed at the Argonne i
National Laboratory [8] with the toughness used in evaluating submerged arc welds (SAWS) in IWB-3640 of ASME Section XI and found them to be similar. Thus, the B&WOG will use the SAW procedures in ASME Section XI to evaluate flaws found in valve bodies fabricated from cast stainless steel for license renewal.
l Table 2 summarizes the aging management programs for license renewal described in the B&WOG topical report for the RCS piping components in the B&WOG GLRP member plants.
5.
CONCLUSIONS The USNRC has reviewed and approved the B&WOG topical report which is the first technical assessment of aging management of a plant component for i
license renewal in accordance with the revised license renewal rule [1]. The B&WOG topical report addresses the aging management of the RCS piping for license renewal for the B&WOG GLRP member plants [2]. The scope of the topical report covers the piping components of the RCS within the ASME Section XI,ISI program for Class I components, that is, the RCS piping, valve bodies, and bolting.
The intended function of the RCS piping components is to maintain the structural integrity of the reactor coolant pressure boundary under normal, upset, emergency, and faulted conditions.
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l-The USNRC staff found that, in general, the existing plant programs are adequate for managing the effects of aging of the RCS piping components for license renewal. However, there are a few situations where additional aging management programs would be necessary for license renewal and they are:
augmented inspection of the Alloy 82/182 clad flow meter section of the hot leg, augmented inspection of small bore piping, and a new program to manage loss of toughness of cast stainless steel materials. Also, fatigue is outside the scope of the B&WOG topical report and will be evaluated on a plant-specific basis for license renewal. On this basis, the USNRC staff concluded that the B&WOG topical report describes acceptable arograms to manage the i
effects of aging so that the intended function of tie RCS piping components of 1
B&WOG GLRP member plants will be maintained consistent with the plant's j
current licensing basis for the period of extended operation.
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6.
REFERENCES I
l
-1, 1995. Nuclear Power Plant License Renewal; Revisions.
Federal Register 60: 22461-22495.
2.
Grubbs, E.E., & M.A. Rinckel 1995. B/.W-2243, Demonstration of the l
Management of Aging Effects for the Reactor Coolant Systen Piping.
i B&W Nuclear Technologies: Lynchburg, Virginia.
l 3.
1989. ASME Boiler and Pressure Vessel Code.
L The American Society of Mechanical Engineers: New York, N.Y.
4.
1988. Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.
U. S. Nuclear Regulatory Commission:
Washington, D.C.
j 5.
1985. Generic Letter 85-20, Resolution of Generic Issue 69: High Pressure Injection /Make-Up Nozzle Cracking in Babcock and Wilcox Plants.
U. S. Nuclear Regulatory Commission: Washington, D.C.
6.
1990.
Information Notice 90-10, Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600.
U. S. Nuclear Regulatory Commission: Washington, D.C.
7.
1982..Bulletin 82-02, Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants.
1 U. S. Nuclear Regulatory Commission: Washington, D.C.
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Chopra, 0.K., & W.J. Shack 1994. NUREG/CR-6177, Assessment of Thermal Embrittlement of Cast Stainless Steels.
U. S. Nuclear Regulatory Commission: Washington, D.C.
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kEMEND C RCS Hot and Coki Legs E Pressurizer Surge and Spray unos a High Pressure injection E Low Pressure injection, Decay Heat, Core Flood g
B Vents, Drains and incore Monitoring DRAIN 4
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FIGURE 1 B&W REACTOR COOLANT SYSTEM PIPING (2) 719-7 4
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I TABLE I APPLICABLE AGING EFFECTS FOR B&W RCS PIPING COMPONENTS Component Acolicable Aoina Effect I
l Piping Cracking Loss of material (carbon steel external surface) i l
Valve Cracking bodies Loss of fracture toughness (cast stainless steel)
Loss of material (carbon steel external surface)
Bolting Cracking Loss of bolting preload i
Loss of material (low-alloy steel) 4 I
TABLE 2 AGING MANAGEMENT PROGRAMS FOR B&W RCS PIPING COMP 0NENTS FOR LICENSE RENEWAL Comoonent Aoina Manaaement Proaram for Renewal Piping ASME Section XI " Examination Categories B-F, B-J, and B-P" Response to Generic Letter 88-05 on boric acid corrosion Program evaluated in Generic Letter 85-20 on thermal sleeve cracking Information resulting from Information Notice 90-10 on Alloy 600 Technical specification RCS leakage limits Augmented inspection of Alloy 82/182 clad hot leg segment Augmented inspection of small bore piping Valve ASME Section XI " Examination Categories B-M-1, B-M-2, and bodies B-P" Response to Generic Letter 88-05 on boric acid corrosion Technical specification RCS leakage limits New program to manage loss of toughness of cast stainless steel Bolting ASME Section XI " Examination Categories B-G-2 and B-P" Response to Generic Letter 88-05 on boric acid corrosion Response to Bulletin 82-02 on bolting degradation Technical specification RCS leakage limits i
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Technical Information from Industry Reports Addressing License Renewal I
L Christopher M. Regan i
4 United States Nuclear Regulatory Commission, U.S.A.
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1 ABSTRACT The USNRC reviewed and documented, in NUREG-1557, technical information and agreements on aging issues that resulted from the review of nine industry reports submitted l
by NUMARC in the early 1990's. The detrimental effects of aging and acceptable aging management programs are delineated in the text with the exception of 15 issues which have been high-lighted for continued analysis. Both the USNRC and industry have been using this _
report to support preparation and review of license renewal applications.
I 1.
. INTRODUCTION In order - to establish the. United States Nuclear Regulatory Commission (USNRC) understanding of technical issues related to renewal of operating licenses for nuclear power plants, under Title 10, Part 54, of the U. S. Code of Federal Regulations (10 CFR Part 54) the USNRC reviewed and documented the detrimental effects of aging and the aging programs to manage these effects for certain systems, structures and components. The USNRC effort resulted in the publication of NUREG-1557, " Summary of Technical Information and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," in late 1996. This information on the detrimental effects of aging and the pertinent management programs were originally documented in the license renewal industry reports (irs) submitted to the USNRC for review by the Nuclear
'i Utilities and Resources Management Council (NUMARC) beginning in the late 1980's. The purpose of this paper is to summarize the information contained in NUREG-1557 and the notable findings and conclusions reached in the aforementioned document.
2.
REPORT DEVELOPMENT c
l In the late 1980's, NUMARC, now 'the Nuclear Energy Institute (NEI), submitted for USNRC review ten irs addressing aging issues associated with specific structures and components of nuclear power plants [1-10], and one IR addressing the screening methodology for performing an integrated plant assessment (IPA)[ll], under Title 10, Part 54, of the 4
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United States Code of Federal Regulations (10 CFR Part 54). The ten irs for specific structures and components are:
- 1. Pressurized Water Reactor (PWR) Reactor vessel [1]
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- 2. Boiling Water Reactor (BWR) Reactor Vessel [2]
- 3. PWR Containment [3]
- 4. BWR Containment [4]
- 5. PWR Reactor Coolant Pressure Boundary [5]
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- 6. BWR Reactor Coolant Pressure Boundary [6]
- 7. PWR Reactor Vessel Internals [7]
- 8. BWR Reactor Vessel Internals [8]
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- 9. Class I Structures [9]
- 10. Low-Voltage, In-Containment, Environmentally Qualified Cable [10]
The original intent of the irs for specific structures and components was to serve as a referenceable surrogate for carrying out the integrated plant assessment (IPA) requirements of the USNRC license renewal rule,10 CFR Part 54, as published in 1990. The IPA information to be submitted by a prospective applicant was to describe aging affects applicable to the systems, structures, and components within the scope of license renewal for their plant and to describe and justify the programs for managing those effects for a period j
l of extended operation beyond the original 40 year operating license. The detrimental effects of aging on certain systems, structures, and components within nuclear power plants and the suggested programs for managing these detrimental effects of aging had been described in the irs for use by individual utilities seeking to submit an application to the USNRC for a renewed operating license.
In 1992 the USNRC staff and industry resources were redirected and review of the irs was terminated. USNRC and industry efforts were concentrated on revising the license renewal rule to focus on the effects of aging rather than an indeterminable number of aging mechanisms. Nonetheless, it was determined after finalization of the revised license renewal l
rule in 1995, that the effort already expended on review of the irs should be utilized. The technical information and agreements reached at the point of review cessation, therefore, I
were to be incorporated into the draft USNRC standard review plan for license renewal (SRP-LR).
After a hiatus in IR review activities, NUREG-1557, " Summary of Technical Information l
and Agreements from Nuclear Management and Resources Council Industry Reports Addressing License Renewal," was completed. This report summarized the technical i
information and NUMARC/USNRC agreements reached from nine of the ten irs, the cable IR [10] was excluded. The cable IR addresses the issue of environmental qualification (EQ) l of electric equipment, which at the time of NUREG-1557 development, had been superseded by the USNRC EQ action plan to address aging of cables. The technical information and agreements documented in NUREG-1557 represent the status of the USNRC staff's review when the USNRC and industry resources were redirected to address license renewal rule implementation issues.
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3.
TECHNICAL INFORMATION AND AGREEMENT FORMAT NUREG-1557 encompasses nine of the original ten NUMARC irs, the relevant USNRC l
staff review documentation for each IR, and the NUMARC responses / positions taken with respect to the USNRC review. All the relevant documentation has been compiled and is-listed in Appendix A of NUREG-1557. The technical information and agreements from the USNRC review have been compiled into tables, see examples Exhibits I and 2, and are presented in Appendix B of NUREG-1557. Each table consists of seven columns. The l
columns list among other things the aging-related degradation mechanisms (ARDw..,
addressed in the irs and their effects on structures and components. The effects of ARDMs were based primarily on information in the irs. In summary the following "ARDMs" and their corresponding " effects" have been considered to affect structures and components in the Reactor Pressure Vessel (RPV), Reactor Vessel Internals, and the Primary Coolant Pressure Boundary (PCPB).
Aging Mechanism Ae_ine Effects
- 1. Corrosion, Microbiologically Loss of material""
induced corrosion, Boric Acid
- Corrosion
- 2. Creep Change in dimension
- 3. Erosion / Corrosion (E/C)
Wall thinning -
- 4. Fatigue Cumulative fatigue damage
- 5. Stress Corrosion Cracking (SCC)"
Crack initiation and growth (to include:IGSCC, TGSCC, & IASCC)
- 6. Neutron Irradiation Embrittlement
- Loss of fracture toughness
- 7. Stress Relaxation' Loss of preload
- 8. Wear" Attrition
' 9. Thermal Embrittlement ("+""*'
. Loss _of fracture toughness
- Boric acid wastage of external surfaces I
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- Also listed as an Effect:
'Also listed as Irradiation Embrittlement (PWR Vessel Internals)
- Also listed as Thermal Aging (PWR Pressure Vessel) i I
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' The following "ARDMs" and their " effects" have been considered to affect structures and components in the Reactor Containment and Class I Structures-I Anine Mechanism Aning Effects i
Concrete Structures -
- 1. Freeze 'Diaw Scaling, cracking, and spalling l
- 2. Leaching of Calcium Hydroxide Increase of porosity & permeability
- 3. Aggressive Chemical Attack
- Increase of porosity & permeability, cracking, and spalling
- 4. Reaction with Aggregates Expansion and cracking l
- 5. illevated Temperature -
Imss of strength and modulus
- 6. Irradiation of Concrete loss of strength and modulus
- 7. Creep Deformation
- 8. Shrinkage Cracking
- 9. Corrosion Loss of material
- 10. Abrasion and Cavitation Loss of material
- 11. Restraint, Shrinkage, Creep, &
Cracking of masonry walls Aggressive Environment
- 12. Concrete Interaction with Aluminum less of strength
- 13. Cathodic Protection Current Cathodic protection effect on l
bond strength i
Structural Steel & Stainless Steel Liner -
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- 1. Corrosion, Local corrosion, Loss of material Atmospheric corrosion
- 2. Elevated Temperature Loss of strength and modulus f
- 3. Irradiation Loss of fracture toughness I
- 4. Stress Corrosion Cracking Crack initiation and growth Reinforcing Steel (Rebar) -
- 1. Corrosion of Embedded Steel Cracking, spalling, loss of bond,
& Loss of material
- 2. Elevated Temperature Loss of strength and modulus
- 3. Irradiation Loss of strength and modulus Miscellaneous -
- l. Fatigue Cumulative fatigue damage
- 2. Settlement" Cracking, distortion, increase in component stress level
- 3. Mechanical Wear Lockup""
- 4. Strain Aging (of Carbon Steel)+
Loss of fracture toughness
- 5. Loss of Prestress*"
Reduction of design margin
- 6. Corrosion of Steel Piles Loss of material
- 7. Corrosion of Tendons Loss of material l'
- Includes Stainless Steel-Bellows (PWR Containments)
- Also listed as ' Aggressive Chemicals" (PWR Containments) i "Also listed as " Differential Settlement" (BWR Containmetits)
~~Also listed as *Prestress lesses" (BWR Containments)
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Also listed as
- Attrition" (PWR Containments) 721-5
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W$4AWrterW Ofledindetal figformalsart and NUMARC/NRC agreernentsfrorrt PWR catfaunmens strucftses indtastr1/ report I
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- QC ww Ad8md Comment NUMARC/NRC Basis for Moldtentmas ESerto C-
.:e Matertales Number 8' Agreement or Proceaal Aseement or Proposal W
lsmane of Concrete Containments Concrete G-12.
For concrete containment Degradation caused by aggres-Chemical puroesty &
Retnineced/Prestreened S S, structures that meet the be-elve chemical attack is non-i l
Attack permembehty
= Concrete Dome 5 38 to sia requirements, adgreserve sign 1Acent for concrete contain-cracking. &
= Concrete Contasnment Wall S 41 chemscal attack is non-ment structures not esposed to apelling Above crede signinrant ARDM agreessve environment (pH
S00 ppra chloride. sand 1500 ppra sulfatel.aor er eg.
)
posed to ground water that ex-ceeds the pH. chlertde, sulfate J
hmats the exposure is for inter-l mittent penode only Agreeneve increase of Concrete Containments Concrete G 10.
Accessible concrete surtares in cases where containment "i
Chemiral puroeity &
Reinforced /Prentressed G.13 are pertuderally esamined in concrete is esposed to aggres-Attack permembthey,
= Concrete Containment Wall G 1S.
accordance wHh the prore-save groundwater (pH SOO ppm. & sulfate spalling
= Concrete Basemat S 36 leak rate test or in accordance s 1500 ppml. pertader Inspec-Free Standing Steel Containment S 37, with /LaiE Sect, XI. Sutmect.
tion of accese4ble concrete sur-with Flat Bottom & en ice 54i IWLa faces as part of Type A ante-Condenser S 6S grated leak rate test performed l
- Concrete Basemat S 66.
Management for the edects of under Appendts J. IOCFRSO.?
5 69 aggresssve chemical attack of or in accordance with AShtE S 72.
conrrete surfaces that are not Sect. XI. Subsect f%Laexam S 75 perkxhrally exammed due to category L A. & guadelh.es of anaccesashetely requires further ACl2011*
i plant specinc esaluation Funher evaluation for management of inacreasible s
areas is to be jusufled on a plant spectnc has6e i
EXHIBIT 1. Sample Table (B3) from NUREG-1557:PWR Containment Structures IR TaNe M Brnrf summary of terhnactd arlfurmafian and MJMARC/NRC agreernentsfnen UWW tsw4 internois industry report Aging-Related NRC Degrulan=i Agtng Comment NUMARC/NRC Basis los Mwhannarn FNects Componene s Materiale Numbere Agreement or 7 _
Agreement or Pranasal IGSCC Crack Access Hole Cover Alloy 600 02.
GESIL 46251I recommends Recommendauons of GESIL j
inHiduon &
G S.
volumetric inspections, im-462Sil & safety analysts are growth G 9.
piementatmn la plant specific. current & eSecuve inspection G I1
& recommended repair is to - programa for detection & eval.
I G 22.
attach retnforcement hardware usuon repair of access hole G 24 covers '
Fore Shroud Head fiolis 55 G 2 7.
GESIL 4332 recommends UT Recommersia. tons of GE.SIL Allot us)
G 28 esammauen during outages.
4132 & replacement with crevere G 29 tmplementat6on is plant free deangn are current & effec-G 33.
sperinc. & replacement is WHh tive inspection programs for de-j G 34 crevtre free design tection & evaluation replace.
I S3 ment of core shroud head bohs Control thte SS S4 GESIL IS7 routme replace-Routtne replacement & opera-S6 ment 1 operational parameter uonal parameter monitortr( are S 17 momsonna inspection current & eSecttve programs for 4
detection & evaluauon-S 22.
evaluahon & reptarernent S 25 replacement of control blades Control Rat Drive R RDI
%5 5 26 to ASME Sers XD requires von ASME Sect XI. Subecct IWB)
Housing S 29 umenne enam of welds & VT 2 esam categones B O & B P is 5 31 of pressure retaming current & e$ectree program for l
5 32.
bnundary & system leakage &
detert6on & evaluauon repair-i I
5 3M hydro staair tests replacemeni r4 CRD housma Core Sprav 5parger h5 5 49 NRC Hulletin MO 136 recom -
NRC Bulletin MO 136 & safety S 54 mends visual insprt Iaun durtna analysis are erective inspectson i
4%
refurtmg outage *,. an.ilvural programs for detection. evalua-a'. alen A ren.ier unn repair of core sprav sparget Briiermediate Rar$e Wnnort Q 511,404 Res l ' rei om Recommendations of GESIL Source Range Wnnor mends visual insperunn 409 F leakage monttonrg, & re l
llRM /SRMI Dry Tubes leakage mumsonna se placement with crevire free de placement is witti crevu e free sign. are encruve inspection l-dewin & remtant matenal programs for detertton & evalua f
l tion reptarement of dry tutes l
EXHIBIT 2. Sample Table (B8) from NUREG-1557:BWR Vessel Internals IR l
721 4 i-i
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l In addition, the tables list the specific structures and/or components, and their construction materials. For each IR, a complete list of what are defined as structures and components is included in the end of the table (see Exhibit 3).
In general, the irs present only representative examples and do not provide a comprehensive list of the type, grade, and i
specification of materials used for various reactor structures and components but give a good baseline from which comparative reviews by a prospective license renewal applicant can begin. For most irs, only material categories such as stainless steel (SS), cast austenitic stainless steel (CASS), Ni Alloy, or carbon steel (CS), are described. A detailed list of material type and grade is, however, provided in the PWR and BWR reactor vessel irs.
IJST OF BWR CONTAINHEMT COMPONtNTS MARE I STEEL CONTAINMENT MARK I CONCRETE CONTAINMENT MA3UL !!! STEEL CONTAIMMENTS Drywell Interior Surfare Drywell Lmer intenor surfse Containment ShellInternor Surface Drywell Exters..r Sudas e Drywell Lurr Entenor sudse Contamment $ bell Exterk>r Surface Drywell Head Tona i mer intertor Surls e Suppr Chamber Shell Interk>r Sudare Embedded Shell ReE*nn forus Lmer inactu>r surfai e as W.ncriuw Suppr Chamiwr Shell Eslerk r Surfa c Drywell Support Skirt Torus I.uwt Enscrior Surf.u e Itasemat Liner Sami Pvnke I Regkan Laswr Aru hors Laner Arwhors
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Torus Intes sur hurt.u e 17tiwell ( mu rete
( orwrcle Ibsesnat j
Torus insciwr Surfm e as Waterlme Tonn i une ror Cs.eu rete till m Annulus Tunn Fasertor %ud.ne larswrlii usu vete Hemi.en mg Sleri t miwddert ShcIl Negmn inrue Rmg Guder lorus com re se 'tcudon uw 5ere 1 MARK III CONCRETE CONTAIMMENTS
\\ent l u.es
\\rntIuws Contamuwnt brwr imenor Sudace Vene b w Helb,ws
\\ ent 1,mr fiellow s Contamment laner illemt Sudm e teta llender
%enslic+1rrs hoppr t hamler Lmer or Claddmg insenor Surtse Imwou nnwrs.uut he a. ma lh wni unwes arul tirae uw Suppr ( hamter Lmer Exterior Surfse Vent System hupgens
% rus Nvisem Suppon Com rete Coruamment Wall Abswe Grade Torus $camw Hestrames Drsweit licad Coru rete Contammens Wall Delow Grade Tonas Supgert Columns /%a idles MARK II CONCRETE CONTAIMMENTS Cont rete Dorow ECCS Suchon lleader l>rTwell i mer 1merna Surise liasemat Liner Oscan Plant wuh Usu naic 1 CS Surt.e es l>r)well 8 wwr Enterwir hudas e Corwrete thsemal Unrnated Submerged CS Surfscs Nippt i hamler Lmer Interwr Surfm e Luwr Arwhors MARE II STEEL CONTAIMMENTS Suppt a hamtier Luwt insenur burim e at Waecrime Contamment % all Remforting Secci Drywellintermr budse Neppr i hamber Lmer Estenor Surta e Denne Remforrma Secci Ikywell Eaterwu Surfm e lar.cr Aan hors tidsemat Remforring Steel Drywell licad Lmer Hegion Shehled by Ihapheagm i kuer COMMON COMPONENTS Suppr Chamber Entermr Sudm e Contamnwnt Concrete Penetration Sleeves Suppr Chamber Intenne Surfare Concrete contammeni Hemfun mg %Icel Dmmmtlar Metal Welds suppr Chamber Insenor Surface at waterime Drvwell ficad Penetrauon Hellows Region $hicided by Diaphragm Fkmr IMwnromer Pipes ami I4.u mg Personnel Aarlot h Embrdded Shell Regmn Cons rete liasemat Equ6pment Hatthes Sand Ptn het Hegen Hasemas laner CHD Hatch Suppov1 Skart Hairmat Reinforring 5:ect Duwnsmer Pipes and Bracing Prestressuul Tendons and Durs a Orvan Plant wuh Uncoated C5 Sudares Unconted Submerged Cs Sudares EXHIBIT 3. Sample Listing of IR Components: BWR Containment Components The USNRC staff review of each original IR resulted in a set of comments / issues for each IR which are referenced in NUREG-1557 in order to recreate the lineage of the aging issue agreement.
NUMARC/USNRC agreements or proposals on whether a ARDM or j
' ARDM/ component combination is potentially significant, and if it is potentially significant, is given and a brief description of the program that can adequately manage the effects of aging is presented in a similar fashion. The technical basis for these agreements or proposals, including assumptions and references, are also described in briefin NUREG-1557.
l A few examples of the information delineated in the report as aging management programs l
and their bases are given below.
721-7
l e.
1.
For a specific ARDM or ARDM/ component combination, if the effects of aging i
l i
are not potentially significant, "non-significant" is listed in the agreements column.
The technical basis, assumptions, and references for the agreement are presented in column seven. For example, the effect of creep is non-significant for BWR l
primary coolant piping and fittings fabricated from carbon steel (CS) or stainless steel (SS) because the reactor operating temperatures are significantly lower than the temperatures at which creep.is a concern to CS and for SS components. Also, if the effects of aging are not potentially significant when certain bounding conditions are met, then "for components that meet the basis requirements, this ARDM is non-significant" is listed in the agreements column. For example, the effects of freeze thaw is non-significant for Class I concrete structures that meet the following criteria: located in geographic regions of negligible weathering conditions (weathering index < 100 day-inch / year): and if located in severe weathering conditions (weathering index 100-500 day-inch / year) the concrete mix design meets the air content and water-to-cement ration requirements of American j
Concrete Institute (ACI) 318-63 or ACI-349-85.
2.
If a specific ARDM/ component combination is potentially significant and the effects of aging are adequately addressed by current management programs, then a brief description of the program is provided in column six. For example, the program delineated in NUREG-0313, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," July 31, 1977, and implemented through USNRC Generic Letter 88-01, "USNRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," January 25,1988, is a current and adequate program to manage the effects of intergranular stress corrosion cracking (IGSCC) of SS piping and fittings of BWR primary coolant pressure boundary.
3.
If a specific ARDM/ component combination is potentially significant and the j
current programs are not adequate for managing the effects of aging, then column six simply states " current practices to be enhanced, select plant-specific aging management." For these cases, the NUMARC recommended aging management options are described in Chapter 6 of the irs. However, chapter six was not the focus of the USNRC review of the irs.
4 An ARDM or ARDM/ component combination is listed as " unresolved issue" if no agreement was reached between NUMARC and the USNRC staff. For these cases, both the NUMARC and USNRC proposals are briefly described in the agreements column. An example of an unresolved issue is the effects of thermal aging embrittlement on PWR primary coolant system components fabricated from cast austenitic stainless steel (CASS). The NUM ARC proposal considers a ferrite content screening criterion and American Society of Mechanical Engineers (ASME) Code Section XI, Subsection IWB, inspection to be an adequate program for managing the effects of thermal embrittlement.
The USNRC proposal, however, considers that ferrite content criterion is inadequate for screening and VT-3 visual examination is not intended or reliable for detecting tight cracks.
721-8
- __~ _
4.
SUMMARY
AND OBSERVATIONS l
Nine of the ten irs submitted by NUMARC addressing the detrimental effects of aging j
associated with specific structures and components of nuclear power plants were reviewed l
by the USNRC. The technical information and NUMARC/USNRC agreements for each IR have been compiled into tables (Appendix B to NUREG-1557). The information presented in each of the tables includes specific structures and components and their materials of construction; ARDMs and their effects on structures and components; relevant comments of the USNRC staff; and the NUMARC/USNRC agreements or proposals and their technical l
bases, including assumptions and references.
l l
Considerable effort was expended by industry representatives and the USNRC staff to come to agreement on the aging issues pertinent to several major structures and components in nuclear power plants. These aging issues and associated structures and components are of I
prime importance for a prospective applicant aiming to satisfy the requirements of the revised license renewal rule,10 CFR Part 54, published in 1995. NUREG-1557 documents the i
USNRC understanding of the information delineated in the irs and the agreed upon and disagreed upon technical positions from the previous USNRC review effort. Resulting from j
the industry /USNRC exchanges it was determined that fifteen open technical iv.ues existed.
i These open technical issues include the following:
1 1
(1) Fatigue in Metal Components (2) Environmental Qualification of Cables i
(3) Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components (4) Irradiation-Assisted Stress Corr'sion Cracking of Reactor Internals Components o
(5) Stress Relaxation of PWR Internals Components (6) Primary Water Stress Corrosion Cracking of High-Nickel Alloys (7) Stress Corrosion Cracking in PWR Metal Components (8) Neutron Irradiation Embrittlement (Definition of Reactor Vessel Beltline)
(9) Freeze-Thaw Damage in Concrete Structures (Significance of Effects)
(10) Alkali-Aggregate Reactions in Concrete Structures (Significance of Effects)
(11) Differential Settlement of PWR Containments and Class I Structures
-(Significance of Effects)
(12) Reinforcement Corrosion in PWR Concrete Containments (Significance of Effects)
(13) One Time Inspections of Concrete and Steel Structures (14) Ultrasonic Inspection of Pressure Vessels and Components l
(15) Visual Inspection of Components and Structures using certain American Society of Mechanical Engineers Code Acceptance Criteria l
Resolution of these technical issues have been slated to be the focus of USNRC license j
renewal reviews. Resolution of these technical issues, for a prospective license renewal applicant, is paramount to successfully meeting the require.ments of the revised license i
renewal rule,10 CFR Part 54. Several of these issues are also currently under review by the USNRC for applicability to operating reactors for their existing 40 year operating license.
1 721-9
,8 l
In the future, as U.S. utilities begin preparation of an application for license renewal, NUREG-1557 will have established a starting point to begin an aging management analysis.
Documented in NUREG-1557 are the aging management programs deemed acceptable by the USNRC and the technical issues where the USNRC and the industry hold diverse positions.
It is resolution of the open technical issues elucidated by this report that is necessary to provide a stable and predicable regulatory environment for a license renewal applicant. By referencing this publicly available report, a prospective license renewal applicant can become familiar with the level of effort necessary to initiate aging studies of other systems, structures, and components that may also be subject to the requirements of the license renewal rule. NUREG-1557 can be used as a time saving and a cost cutting mechanism by a license renewal applicant. In addition to the benefits for industry utilities, the USNRC is currently considering incorporation of the appropriate technical information and agreements as the basis for a revised draft USNRC SRP-LR. This new SRP-LR will establish the methods and acceptance criteria that meet the requirements of 10 CFR Part 54 that the USNRC staff will utilize to perform a review of future license renewal applications.
5.
CONCLUSIONS Industry submittal and subsequent USNRC teview of nine of the ten irs resulted in the establishment of a clear understanding of iridustry and USNRC positions related to the detrimental effects of aging and the programs necessary for managing them. NUREG-1557 concisely summarizes these agreements and disagreements in a format usable to a prospective i
license renewal applicant. The information delineated may afford a prospective applicant an introduction to the significant aging issues needing resolution. Although several aging issues j
remain unresolved the report contains valuable information that may be referenced by a i
prospective applicant saving them both time and resources. It should also be noted that the USNRC will consider utilizing this report by incorporating resolved aging issues into a draft SRP-LR that will meet the requirement of the revised license renewal rule,10 CFR Part 54.
6.
REFERENCES 1.
Nuclear Management and Resources Council. September 1992. Pressure Water Reactor Vessel License Renewal Industry Report, May 1990. Revision 1. Report Number 90-04.
2.
Nuclear Management and Resources Council. September 1992. Boiling Water Reactor Vessel License Renewal Industry Report, October 1989. Revision 1. Report Number 90-02.
3.
Nuclear Management and Resources Council. September 1991. Pressurized Water Reactor Containment Structures License Renewal Industry Report, August 1989. Revision l
j
- 1. Report Number 90-01.
4.
Nuclear Management and Resources Council. December 1991. Boiling Water Reactor Containments License Renewal Industry Report, July 1990. Revision 1. Report Number 90-10.
721-10
t 5.
Nuclear Management and Resources Council. May 1992. PWR Reactor Coolant System i
License Renewal Industry Report, October 1990. Revision 1. Report Number 90-07.
6.
Nuclear Management and Resources Council. April 1992. BWR Primary Coolant Pressure Boundary License Renewal Industry Report, September 1990. Revision 1. Report Number 90-09.
7.
Nuclear Management and Resources Council. December 1992. Pressurized Water i
Reactor Vessel Internals License Renewal Industry Report, September 1990._ Revision 1.
Report Number 90-05.
t i
8.
Nuclear Management and Resources Council. June 1992. Boiling Water Reactor Vessel l
Internals License Renewal Industry Report, February 1990. Revision 1. Report Number 90-03.
9.
Nuclear Management and Resources Council. December 1991. Class I Structures License Renewal Industry Report, June 1990. Revision 1. Report Number 90-%.
- 10. Nuclear _ Management and Resources Council. March 1993. Low-Voltage, In-Containment,- Environmentally-Qualified Cable License Renewal Industry Report, July 1990.
l Revision 1. Report Number 90-08.
I1. Nuclear Management and Resources Council. October 6,1989. Methodology to Evaluate Plant Equipment for License Renewal.
I L
i 721-11
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' Nuclear Power Plant Generic Aging Lessons Learned (GALL) l Christopher M. Regan l
United States Nuclear Regulatory Commission, U.S.A.
l 1500Fr Christopher M. Regan l
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i ABSTRACT -
The purpose of the generic aging lessons learned paper was to provide a syste'matic review -
of plant aging information to assess materials and component aging issues related to
~
l operation and license renewal of nuclear power plants. The results were documented using a standardized tabular and electronic database format and definitions of aging-related l
l degradation mechanisms and effects. Results reveal that all significant aging issues are I
currently being addressed by the U.S. Nuclear Regulatory Commission regulatory process.
i
- 1.. INTRODUCTION j
l Approximately 110 nuclear electrical power generating plants operating in the United Sates j
of America generated roughly 20% of the nations electrical demand. Some of these plants have been in operation for many years. It is well established that many of the critical L
components in nuclear power plants are subject to time-dependant degradation, or aging, L
as a result of normal plant operations. In recognition of the potentially adverse effects of
-the aging process on plant' safety, the United States Nuclear Regulatory Commission's l
(USNRC) Office of Nuclear Regulatory Research began by establishing the Nuclear Plant Aging Research (NPAR) Program. The principal objective of this program was to develop a basic understanding.of age-related degradation (ARD) processes and their effect on nuclear power plant systems, structures, and components. In addition, the Nuclear Energy L
Institute (NEI), formerly the Nuclear Management and Resources Council (NUMARC),
developed a series of license renewal industry. repota (irs) to support a prospective applicant when submitting an application for a renewed license under the requirements of i
Title 10, Part 54, of the U.S. Code of Federal Regulations (10 CFR Part 54) [1]. The irs '
. describe the NUMARC assessment of plant aging issues and management strategies for several components and structures.
i 720-2
e.
To develop appropriate technical criteria for addressing the aging issues related to nuclear power plant license renewal, the USNRC initiated an activity to assess and integrate the age-related information from all USNRC documentation to include NPAR reports, generic communications, and License Event Reports (LERs) and to use the results of this assessment to supplement and update license renewal guidance previously developed. This activity was called the Generic Aging Lessons Learned (GALL) Program, t'ae results of which have been documented in NUREG/CR-6490 (ANL-96/13), " Nuclear Power Plant Generic Aging lessons Learned (GALL), Main Report and Appendices A and B."
l NUREG/CR-6490 presents the results of the GALL review program. The GALL effort was sponsored by the USNRC with.significant input provided through a joint effort i
involving 12 technical experts from Argonne National Laboratory (ANL) and Idaho National Engineering Laboratory (INEL). ANL reviuved information on mechanical, structural, and thermal-hydraulic components and systems and INEL reviewed information on electrical corr.ponents and systems. The results of these reviews were compiled using a standardized tabular format and standardized definitions of ARD effects. All tabulated review information is contained in Appendices A- (Volume 1) and B (Volume 2) of NUREG/CR-6490. The information is also available'in a computerized data base format based on the software program FoxPro. The data base allows rapid queries and sorts of the large amount of information generated by the review.
V 2.
DESCRIPTION OF REVIEW PROCESS More than 550 documents containing nuclear power plant information were reviewed for
" GALL" information.
The USNRC staff performed searches for current operating i
experience documents covering the 5 year period, 1989-1994, using the USNRC's Nuclear Documents Managements System (NUDOCS). The period preceding 1989 was documented in the NPAR program reports which are also summarized in NUREG/CR-6490. The searches used the following terms: aging, degradation, and failures. A total of 163 NPAR reports, 31 USNRC Generic Letters, 265 USNRC Information Notices, 82 LERs, 5 USNRC Bulletins, and 10 NUMARC Industry Reports (irs) containing mechanical, structural, thermal-hydraulic, and electrical systems and components were reviewed under the GALL program. The results of these reviews were compiled by using a standardized j
tabular format and stand:vdized definitions of ARD and effects. A standardized and consistent set of definition:: and descriptors for all aging mechanisms encountered during the review was developed for the GALL effort. Individual aging mechanisms ar.d effects were identified and defined, and these are listed and described in Table 8 of NUREG/CR-6490. This list will help focus and systematize future reactor aging studies.
The reports, notices, letters, and bulletins revie-vd are listed in Tables 2 though 7 of u
NUREG/CR-6490. The results from each reviewed document are summarized in the GALL tables contained in Appendices A (volume !) and B (volume 2) of the two vo%me 720-3
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NUREG/CR-6490 report. A separate table was prepared for each of the NPAR reports and l
NUMARC irs; findings from the USNRC Generic Letters, USNRC Information Notices, l
USNRC Bulletins, and LERs are tabulated 'oy year in separate tables. All of the GALL table information has also been entered into a FoxPro data base software program that can oe used on IBM PC-compatible sy tems to retrieve and categorize information on structures and components and their related aging effects.
l The information contained in the GALL tables is a summary of that provided in the l
reviewed reports. It was found that not all of the reports, notices, and bulletins reviewed j
l contained relevant information on Age-Related Degradation (ARD) processes. A number l
l of the NPAR reports described programs, methodologies, computer codes, etc.,'for l_
studying and analyzing aging processes in nuclear components, but did not provide detailed
{
l information on the processes themselves. The tables for these reports contain a standard L
statement indicating this fact. Almost ah of the USNRC Generic Letters, USNRC Information Notices, LERs, and USNRC Bulletins reviewed contained detailed information on the failure of specific components, but the failures were sometimes judged not to be aging-related by the reviewer or by the author of the reviewed document. For example, failures caused by improper heat treatment, preexisting defects introduccd during l
manufacturing, or overloads or other abuse during operation were not considered aging-related by the reviewer, even though the failure might not have occurred umil the component had been in service for some time. GALL table entries are not provided for USNRC Generic Letters, USNRC Information Notices, LERs, and USNRC Bulletins judged not to contain detailed information on specific aging effects and their impact on specific plant components. The structure of the information documented in NUREG/CR-6490 and in the database is in such format that enables a researcher to easily access information on a particular aging issue.
In the future, information documented in NUREG/CR-6490 may be used as a reference tool for reviewers performing license renewal application reviews or for the USNRC staff reviewing a licensee submittal. Indeed, a prospective license renewal applicant or nuclear power plant licensee may use this document as one of the sources for determining
(
applicable aging effects to be addressed in a license renewal application.
3.
OBSERVATIONS AND FINDINGS More than 550 documents comprising 163 NPAR reports,31 USNRC Generic Letters,265 USNRC Information Notices, 82 Licensee Event Reports, 5 USNRC Bulletins, and 10
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.c The review reveals (1) that there are no new issues with respect to the components subject to ARD and the degradation mechanisms responsible and (2) that all ongoing significant issues are currently being addressed by the regulatory process. However, (3) the aging of
_ passive components has been high-lighted for continued scrutiny.
Included in NUREG/CR-6490, are recommendations of the referenced report's author for each aging issue and which are contained in Appendices A and B of NUREG/CR-6490.
The GALL effort then added the ANUINEL judgement of the validity of the author's recommendations based on present day understanding of the issue. The characterization-of the issue is noted by one of four possible aging issue current relevance categorization -
indicators with the all-important "may potentially need further evaluation," indicating the possibility of emerging aging issues. This characterization of particular issues, however, appears only a few times and where it does appear the issue is currently being addressed by the regulatory process. A summary of general observations concerning specific aging issues and the components affected are presented below.
3.1 Mechanical, Structural, and Thermal-Hydraulic Components and Systems As expected, corrosion and corrosicn-related processes were the dominant mechanism of age related degradation (ARD) in coolant piping and steam generators. Where high-
. velocity fluids were present in piping, erosion / corrosion was also a significant mechanism.
Additionally for piping, feedwater nozzles, and interfacing tanks and other componenti, nonuniform water temperature fields aggravated by thermal buoyancy can cause large 1
induced structural thermal stresses of either quasi-steady, low-cycle, or thermal shock nature and can lead to cracking or significant structural distortion. These thermal stresses are usually not accounted for in component design and are highly plant and mode-of-l operation-dependant. They can occur under normal or intermittent operation of plant systems and tend to be worse under low flow conditions. For reactor internals, irradiation-assisted stress corrosion cracking was an important source of degradation where high radiation fields were present. - Other forms of corrosion, as well as vibrational fatigue, also contributed to internals degradation.
L Pump-and valve 1 casings were likewise found to be subject to corrosion and erosion / corrosion related degradation. Thermal embrittlement was an important mechanism I
in cast stainless steel pump and valve components. Moving parts in pumps and valves suffered from ARD produced by wear, vibration, fatigue, and erosion / corrosion. Valve and pump seals and other elastomer components were subject to degradation by physical and chemical degradation at elevated temperature and/or prolonged exposure to the service environment.
The principle degradation mechanism affecting concrete structures was ler'hing and breakdown of cement phases under the action of aggressive chemicals, degradation due to -
freeze-thaw cycles, and corrosive attack of the embedded rebar.
The responsible mechanism (s) for some concrete wall cracking was found to be not well understood.
720-7 i
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t Diesel generators, air compressors, and ventilating and air conditioning equipment suffered l
principally from wear, vibration, and fatigue associated with reciprocating motion, as well as corrosion and wear induced by contamination. Heat exchangers and steam generators were subject to contamination and corrosion, as well as biofouling, thermal fatigue, and vibrational fatigue. Vibrational fatigue, wear, and elevated temperature degradation of damping fluids commonly caused degradation in mubbers.
1 Table 1 lists aging issues found to occur almost equally in boiling water reactor (BWR) and
. pressurized water reactor (PWR) plants and tend to center on various forms of corrosion and fatigue. Another important commonality of the components listed in this table is that j
they are all what are termed passive components as described in 10 CFR Part 54 [11].
l This may be of considerable significance because the literature reviewed seems to indicate j
that passive components are not as extensively or thoroughly covered by current plant maintenance procedures.
Furthermore, survei!!ance and monitoring methods and instrumentation and procedures have not been as extensively developed or employed for passive components subjected to the highlighted aging mechanisms, nor are some of the L
passive component aging mechanisms as well understood. Thus, plant life extension by employing component replacement and maintenance could be more tenuous for the passive l
components. Furthermore, passive components tend to be some of the most costly in a plant and are frequently not as easy to replace. For these reasons, the knowledge base for t
predicting relevant aging ' effects behavior and significance, which is essential to the development of robust plans for aging reduction, monitoring procedures,'and maintenance, is very important for passive components.
l 3.2 Electrical Components and Systems Breakers and relays were usually covered together in the same report; the predominant aging-related failure mechanisms were contact wear, sticking linkage, loss of lubrication, or elevated temperature. Normally energized relay coils were frequently mentioned as high failure-rate items because of the insulation breakdown caused by elevated temperature due to self-heating from the continuous current. Breakers are routinely refurbished on periodic schedules. Instrumentation and control (I&C) systems, including breakers and sensors, are made up of many small components that are routinely replaced after a number of years of service, as determined by qualification programs. Thus aging !s controlled by scheduled maintenance and periodic replacement. Redundancy in the Reactor Protection System and L
Engineered Safety Features Actuation Systems allows for taking'a channel out of service for maintenance.
l Degradation of cable insulation and jackets is the sajor effect of cable aging, due primarily L
to radiation and elevated temperature. Despite sizable efforts to develop electrical and mechanical methods of detecting cable insulation degradation, there are no reliable methods L
of detecting degradation of electrical cable insulation in a reactor containment. Electrical j.
parameters, while relatively easy to measure, were found not to give a good indication of mechanical degradation of the cable insulation. The mechanical indentor method was successful only for some of the jacket insulation types.
f 720-8 l
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'On
\\s TABLE 1. Selected Examples ofIssues Significant to Passive Structures & Components o.
Reactor Component Material Degradation Process References l
Type PWR Instrumentation Low-alloy steel Environmentally assisted fatigue.
NUREG/CR-and control rod (A533B, A508)
Appropriate design rules do not yet exist 5490 [2]
drive (CRD) with Type 308 or in the American Society of Mechanical bousmg nozzles 309 SS clad Engineers (ASME) Boiler & Pressure Vessel Code l
BWR Closure Studs A-540, B22, B2, Environmentally assisted fatigue, NUREG/CR-j-
and or B24 fretting, and boric acid corrosion if 5490 [2]
PWR leakage present PWR CRD system Various Dropped or stuck rod due to failure by NUREG/CR-components fatigue, mechanical wear, or stress 5555 [3]
i corrosion crackmg BWR Jet pump & -
Inconel X-750 Cracking and possible failure form NUREGICR-holdown beams vibrational and/or environmentally 5754 [4]
I assisted fatigue and stress corrosion cracking BWR Reactor Various Crack initiation, growth, and possible NUREG/CR-J Internals failure from irradiation-assisted stress 5754 [4]
i corrosion cracking (IASCC) i.
PWR Lower core Type 304 stainless Cracking and possible failure from NUREG/CR-support steel (SS), A-vibrational fatigue and IASCC 6048 [5]
1 structure 286, Inconel X-components 750,' and others BWR Pressure vessel Low-alloy steel Cracking, possibly stress corrosion Information upper head (A533B, A508) cracking (SCC), of weld clad, with Notice (IN) with Type 308 or cracks penetrating into underlying base 90-29 [6]
309 SS clad metal l
BWR Core Shroud Type 304 SS SCC (or IASCC) leads to circumferential IN 93-79 [7],
cracking of core shroud and concems IN 94-42 [8]
about possible structural failure in an accident or seismic event BWR Recirculating Cemented WC in Preferential corrosive dissolution on Ni l
coolant pump Ni binder binder under certain undefined conditions seals leads to excessive seal leakage and possible eventual pump failure
- BWR, All piping and Commonly used large thermally induced stresses, either NUREG/CR-
]
and feedwater materials, low quasi-steady or low-cycle transient -
4731 l
l PWR nozzles and alloy steels thermal fatigue, induced by nonuniform Vols.1 & 2 l
interfacing coolant temperature fields aggravated by
[9]
tanks and thermal-buoyancy-caused stratification l-components under no-flowIlow-flow levels, cause wall cracking / gross abnormal component distortion, usually not accounted for in component design, highly plant and mode-of-operation dependant i
BWR Shielding wall Reinforced Actual process and mechanisms unclear; NUREG/CR-and concrete and concrete shows up as large surface cracks not 4652 [10]
PWR other locations caused by structural loading t
720-9
le.
lY Motors and generators occasionally fail due to bearing wear caused by vibration and winding insulation breakdown from elevated temperatures. Brushes also age due to wear.
' Battery chargers and inverters are small electrical systems made up of many electronic components that, like the instrumentation and control (I&C) system, can be taken out of service for maintenance because of redundancy. Many of the electrical I&C components l
. are included in plant quality assurance (QA) programs that require periodic replacement.
. Inverter failures have caused numerous problems. Many of the electrical.I&C components are included in plant quality assurance programs that require periodic replacement. A more L
detailed analysis may be carried out at a later date to assess the significance of these i
mitigative practices and the aging processes.
.i i
4.
CONCLUSIONS i
i l
'A preliminary assessment of the GALL tables re":als that all significant issues with respect to structure and component aging are currently being addressed by the USNRC regulatory L
process. Nonetheless, the aging of certain structures and components and the resulting l
aging effects, particularly in the category of what are termed passive structures and L
components, have been high-lighted for continued scrutiny and evaluation. Among these
[
issues erosion / corrosion and fatigue of mechanical and structural components.and degradation of electrical insulation of cables and other electrical items are the most notable.
The material documented in NUREG/CR-6490 may be referenced by both' operators of currently licensed nuclear power plants and those wishing to extend their current operating l
license for a additional operating period.
L 5.
REFERENCES l
1.
United States Nuclear-Regulatory Commission. January 1,1996. Code of Federal Regulations, Title 10, Part 54.
l.
2.
Werry, E. V. October 31,1990. NUREG/CR-5490, Regulatory Instrument Review:
Management of Aging of LWR Major Safety Components.
L lt 3.
Gunther, W. and Sullivan, T. March 31,1991. NUREG/CR-5555, Aging Assessment j
of the Westinghouse PWR Control Rod Drive System.
4.
Luk, K. H. September 30,1993.- NUREG/CR-5754, Boiling-Water Reactor Internals
- Aging Degradation Study-Phase I.
5.
Luk, K. H. September 30, 1993. NUREG/CR-6048, Pressurized-Water Reactor Internals Aging Degradation Study-A Phase I Report.
l 6..
United States Nuclear Regulatory Commission. April 30, 1990. Information Notice 90-29, Cracking of Cladding and Its Heat Affected Zone in the Base Metal of Reactor
+
Vessel Head.
4.
i 720-10 t
l
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LIp
,4 7.
United States Nuclear Regulatory Commission. September 30, 1993. Information l
l Notice 93-79, Core Shroud Cracking at Beltline Region Welds in Boiling-Water Reactors.
i 8.
United States Nuclear Regulatory Commission. June 7,1994. Information Notice 94-42, Cracking in the Lower Region of the Core Shroud in Boiling-Water Reactors.
l 9.
Shah, V. N. and MacDonald, P. E. June 30, 1987, Volume 1, and November 30, 1989, Volume 2. NUREG/CR-4731. Residual Life Assessment of Major Light Water l
Reactor Components.
i l
- 10. Naus, D. J. September 30,1986. NUREG/CR4652, Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants 1
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