ML12093A486
| ML12093A486 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/29/2010 |
| From: | NRC/OGC |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML12093A482 | List: |
| References | |
| RAS 22204, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, MRP-227-A | |
| Download: ML12093A486 (76) | |
Text
NRC000114E Resubmitted: April 2, 2012 RAI Set 4- Final Responses: 10/29/2010 remaining core barrel welds and the non-cast lower support column bodies in Westinghouse plants. The original degradation effect of concern was cracking from SCC (because of the residual stresses in the welds), but cracking from IASCC - in particular for the circumferential weld connecting the upper and lower core barrel sections that is located adjacent to the core active region - was also a potential concern. This potential concern about IASCC cracking did not extend to the upper core barrel flange weld, which is outside the core active region with relatively low irradiation exposure. However, the results of the functionality analysis showed that irradiation-induced stress relaxation limited the IASCC ratio to 0.41, well below the threshold of concern. Therefore, because of its greater thickness and relatively residual and operating stresses, the upper core barrel flange weld was designated as the Primary component, with the remaining core barrel welds (circumferential and axial) were relatively less affected, becoming designated as Expansion components.
Since these welds are part of a core support structure, they are all subject to ASME Code Section XI Examination Category B-N-3 visual (VT-3) inspections. However, in order to determine the adequacy of those visual examinations for detecting SCC, the experts reviewed available information on the flaw tolerance of structures similar to the core barrel that were known to have reduced fracture toughness from neutron irradiation exposure, including but not limited to the information contained in MRP-2 10, "Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Components, June 2007. These and other flaw tolerance calculations have been based on lower bound fracture toughness information and simple geometries - such as a through-wall crack in a flat plate - with remote tensile stress treated parametrically. The experts shared their flaw tolerance information, which showed that critical flaw lengths were two inches or greater for remote tensile stresses of the order of 30 ksi and through-wall flaws. In addition, the experts examined the available information from the functionality analyses on the irradiation-induced stress relaxation at all elevations for the core barrel welds. Some experts argued for looking beyond through-wall flaws to examine critical flaw lengths for part-through flaws. In such cases, surface-breaking flaws greater than five inches in length and extending to over ten inches in length, depending upon flaw depth, were needed to reach critical flaw length. However, the consensus of the experts was to use a conservative expansion criterion of a two-inch-long flaw length for a surface-breaking detected flaw. The experts also debated the need for increasing the rigor of the examination from a VT-3 to a VT- 1 and perhaps even to a EVT- 1 examination. However, expert judgments on the potential crack-opening surface displacement for a two-inch-long, surface-breaking flaw led to a consensus that the character recognition requirements for the VT-1 examination would be sufficient to ensure detection and length sizing. Finally, the experts debated the need for altering the ASME Code frequency of examination from ten years to some shorter period.
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RAI Set 4- Final Responses: 10/29/2010 Information from the functionality analyses showed that the conservatisms embedded in the high-leakage and low-leakage core histories, plus the conservatism in the lower-bound flaw tolerance calculations, plus the conservatism in the expansion criterion relative to critical flaw lengths, plus the conservatism of the VT-I examination were sufficient to warrant continuing the existing ASME Code inspection periodicity.
Another example, the Westinghouse core barrel, is discussed below. The reasoning for linking SCC and IASCC in the primary and expansion strategy for the Westinghouse core barrel is outlined in Section 4.2.2 of MRP-232. The following paragraphs are excerpted from that document.
The aging degradationmechanisms identifiedfor the core barrelstructure are listed in Table 4-8. Due to the large size of the core barreland the significanceof the welds, sections of the core barrelwere originallylisted as separatecomponents. Although this division was helpful in the identificationof aging degradationissues,for the evaluationof the core barrel,the welded structure will be consideredas a single assembly consistent with the approachusedfor CE-designedplants.
By the conventions used in this program,IASCC is defined as theform of SCC that is observed in materialswith neutronfluences greaterthan 3 dpa. Because the core barrelcontains both irradiatedand unirradiated welds, the core barrelassembly was screened in for both SCC and IASCC.
The welds in the core barrelwere originallyidentified as potentially susceptible to SCC due to the residualstresses producedby welding in conjunction with deadweight loads and operationalstresses.
Analysis indicates that irradiation-induced stress relaxation reduces the weld residual stresses below the threshold for LASCC in the section of the core barrel immediately adjacent to the core. However, for core barrel welds outside the active core region, there is no mechanism for stress relaxation. Due to the relatively low potential for reaching or exceeding the IASCC susceptibility ratio, the core barrel welds are not considered to be a lead item for IASCC.
The lack of any known predictive model (or data) for SCC in non-irradiated stainless steels in PWR environments makes it difficult to provide an analysis that eliminates the concern for SCC.
The potential for large residual stresses in the unirradiated core barrel welds make them a potential lead component for SCC. Under normal operating conditions the upper flange weld is expected to experience the highest stress. Given the critical structural role of the core barrel, periodic inspection for cracking of the high stress weld is recommended.
Similar types of discussion could be added for each and every Primary to Expansion link.
However, the expert elicitation process does not generally lend itself to a detailed narration of the discussion and decision-making process. The notes that were included in "Letter to Reactor Internals Focus Group from MRP,
Subject:
Minutes of the Expert PanelMeetings on Expansion Criteriafor ReactorInternals I&E Guidelines, MRP 2008-72
RAI Set 4- Final Responses: 10/29/2010 036 (via email), June 12, 2008)" give the results and some of the reasoning that took place within the expert panels. In conclusion it can be stated that in no case is a component item in the expansion group predicted to experience the aging effect sooner or at a faster rate than its linked Primary component.
RAI 4-27: MRP-227 and supporting reports do not clearly document how the consideration of degradation mechanisms associated with weld heat-affected zones, weld repair, and variability in welding processes and parameters was addressed in the susceptibility evaluation. Please provide an overview of how these issues were evaluated to determine the final AMP recommendations for welded components and also provide specific examples to illustrate the impact of these issues on the final inspection requirements.
Response: The treatment of welds, including weld heat-affected zones, weld repair, and variability in welding processes and parameters, was not treated identically in the two susceptibility screening efforts, as documented in MRP-189 and MRP-191.
The original version of MRP-189 screened for multi-pass welds, without consideration of welding process or welding geometry, primarily looking for SCC susceptibility.
However, Revision 1 of MRP- 189 contained an extensive update of the original MRP-189, with the specific intent of addressing multi-pass welds, their heat-affected zones, the variability in welding processes, and any influence of weld geometry. Table 3-2 is similar to the table in the original MRP- 189, showing only whether or not the particular internals component contained a multi-pass weld, implying heat-affected zones that were considered to be susceptible to SCC. Revision 1 of MRP-189 contains Section 3.3 and Table 3-3 that address such items as welding process and welding geometry.
Generally, the susceptibility evaluations documented in MRP-191 for Combustion Engineering and Westinghouse internals only divided welds into a separate category for evaluation when the screening criteria of MRP-175 were also separated. For example, the second paragraph in Section 3.2 of MRP- 191 cites austenitic stainless steel welds, especially those with less than 5% ferrite content, highly-constrained welds, and parts with > 20% cold work as items for which SCC could be an issue. Then, Table 3-1 of MRP- 191 specifically identifies welds separately from other high effective stress locations for SCC screening criteria, while Table 3-2 of MRP-191 implies that welds are included for IASCC through the criterion that all components with effective stresses above 30 ksi were screened in. Table 3-5 in MRP-191 identifies welds separately for thermal aging embrittlement criteria, while Tables 3-6 and 3-7 lump austenitic welds and austenitic base metal in the same category. In other words, welds and their heat-affected zones were treated separately where screening criteria for susceptibility supported such 73
RAI Set 4 - Final Responses: 10/29/2010 distinctions. As a result, welds and the volume of material near welds are called out for special attention when susceptibility is so indicated. MRP- 191 did not explicitly address variability in welding processes and parameters.
The potential for weld repair (grinding out a defect found in a weld during either pre-service or in-service examination, and re-welding) was treated in both MRP-190 and MRP- 191 through the explicit conservative inclusion of welds for SCC or the implicit conservative inclusion of high effective stress locations for IASCC. Such conservative inclusions of weld locations avoided the need for an exhaustive review of component fabrications records, which may or may not have included the level of detail that would have been required to identify specific component locations of concern. The more efficient approach was to assume that all components that were judged to be heavily deformed or welded during manufacture were initially screened in for SCC, regardless of stress level, with a relatively similar conservative approach (see Figure 5-1 in MRP- 191) used for initial screening for IASCC. In that way, any potential weld repairs would be captured in the initial screening.
The potential for repair welding of non-welded material, such as to repair porosity in a stainless steel casting was considered outside the scope of the screening exercise.
The MRP- 189, Revision 1, process and the MRP- 191 process both led to robust and defendable recommendations for specific weld and heat-affected zone inspection requirements.
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APPENDIX A - PROPOSED CHANGES to MRP-227-Rev. 0 As discussed in the meeting between the NRC and the MRP/Industry on 10/14/2010, the MRP is proposing some changes to MRP-227-Rev. 0 to the NRC for incorporation into MRP-227-A.
The MRP committees have concurred with these changes. All proposed changes are listed below.
- 1) The MRP proposes to elevate requirement "7.6 Aging Management Program Results Requirement" in Section 7 from "Good practice" to "Needed" and to change the text of this requirement to:
"Needed: Each commercial U.S. PWR unit shall provide a summary report of all inspections and monitoring, items requiringevaluation, and new repairsto the MRP ProgramManagerwithin 120 days of the completion of an outage during which PWR internals within the scope of MRP-227 are examined."
- 2) The MRP proposes to add a new requirement to Section 7 and to add the following text to Section 7:
"7.7 Evaluation Requirement Needed: If an engineeringevaluation is used to disposition an examination result that does not meet the examination acceptance criteriain Section 5, this engineering evaluation shall be conducted in accordance with an NRC-approved evaluation methodology."
- 3) The MRP proposes to add the following minimum coverage requirements for some of the Primary components.
- For Table 4-1 (B&W Primary components), notes to the table indicated in quotes will be added to the following components:
o For Upper core bolts and their locking devices, Lower core bolts and their locking devices, Baffle-to-former bolts, Locking devices including locking welds, or baffle-to-former bolts and internal baffle-to-baffle bolts:
"A minimum of 75% of the total population (examined + unexamined),
including coverage consistent with the Expansion criteria in Table 5-1, must be examined for inspection credit."
- For Table 4-2 (CE Primary components), notes to the table indicated in quotes will be added to the following components:
o For Core shroud bolts:
"A minimum of 75% of the total population (examined + unexamined),
including coverage consistent with the Expansion criteria in Table 5-2, must be examined for inspection credit."
o For Upper (core support barrel) flange weld:
"A minimum of 75% of the total weld length (examined + unexamined),
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including coverage consistent with the Expansion criteria in Table 5-2, must be examined from either the inner or the outer diameter for inspection credit."
For Table 4-3 (Westinghouse Primary components), notes to the table indicated in quotes will be added to the following components:
o For Baffle-edge bolts, Baffle-former bolts:
"A minimum of 75% of the total population (examined + unexamined),
including coverage consistent with the Expansion criteria in Table 5-3, must be examined for inspection credit."
o For Upper core barrel flange weld:
"A minimum of 75% of the total weld length (examined + unexamined),
including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or the outer diameter for inspection credit."
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- 4) The MRP proposes the following changes to the B&W tables in MRP-227-Rev. 0. Note that these tables do not include all of the proposed changes to the B&W Tables (see point 3 above). A set of tables with all the changes combined could be provided to the NRC later if necessary.
Change a:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of a records review and accessibility evaluation, it was determined that the "CSS vent valve disc shaft or hinge pin" was inaccessible. This component is listed as part of the Core Support Shield Assembly in Table 4-1 of MRP-227, Rev. 0 as shown below.
ttemAppicablit itan Aplicablity Effct Mecanis)
Efect (echansm) Epanson ink (Nte ExansionUnk Not 2)
- 2) Examination (Note 2) Method/Frequency Examintion Coverage Core Support Shield Assembly Cracking (TE), including Visual (VT-3) examination during 100% of accessible surfaces CSS vent valve top retaining rng the detection of surface the next 10-year ISI. (See BAW-2248A, page 4.3 and 055 vent valve bottom retaining ring All plants irregularities, such as None Table 4-1.)
damaged, fractured CSS vent valve disc shaft or hinge pin material, or missing Subsequent examinations on the (Note 1) items 10-year ISI interval. See Figures 4-10 and 4-11 In order to reflect the actual generic condition and to clarify the requirements, MRP proposes that the following two rows be inserted into MRP-227-A and the existing row (shown above) be deleted. Also, Figure 4-10 will be revised and be replaced for clarity.
Item Applicability Effect (Mechanism) Expansion Link (Note 2) Exmnato2)hd~rqec Examination __Coverage Core Support Shield Assembly Cracking (TE), including Visual (VT-3) examination during 100% of accessible surfaces the detection of surface the next 10-year ISI. (See BAW-2248A, page 4.3 and CSS vent valve top retaining ring irregularities, such as None Table 4-1.)
All plants daagdofatue CSS vent valve bottom retaining ring damaged, fractured material, or missing Subsequent examinations on the (Note 1) items 10-year ISI interval. See Figures 4-10 and 4-11 Core Support Shield Assembly No examination requirements. Inaccessible.
CSS vent valve disc shaft or hinge pin All plants Cracking (TE) None (Note 1) Justify by evaluation or by See Figure 4-10.
replacement.
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Chancqe b:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that the "effect (mechanism)" column only identifies the "effect (mechanism)" for the UCB bolts and does not Clearly describe the "effect (mechanism)" for the associated locking devices. Also, the associated locking devices were omitted from the "Expansion Link" items. In addition, the locking devices were omitted from the "Examination Coverage" column. This component is listed as part of the Core Support Shield Assembly in Table 4-1 of MRP-227, Rev. 0 as shown below.
item Applicability Effect (Mechanism) Expansion Link (Note 2) (Note 2)
Examination Method/Frequency Examination Coverage Volumetric examination (UT) of the LCB (Note 3) bolts within two refueling outages from 1/1/2006 or next 10-year ISI interval, whichever is first.
UTS, LTS, and FD bolts Core Support Shield Assembly 100% of accessible bolts.
Subsequent examination to be Upper core barrel (UCB) bolts and their All plants Cracking (SCC) SSHT bolts (CR-3 and DB determined after evaluating the locking devices only) baseline results. See Figure 4-7 Lower grid shock pad bolts Visual (VT-3) examination of bolt (TMI-1 only) locking devices on the 10-year ISi Iinterval.
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 4-1 of MRP-227-A, as shown below.
Item Applicability Effect (Mechanism) Expansion Unk (Note 2) Examination Method/Frequency Examination Coverage (Note 2)
LCB and their locking Volumetric examination (UT) of the devices (Note 3) bolts within two refueling outages Bolt: Cracking (SCC) UTS LTS and FD bolts from 1/1/2006 or next 10-year ISI and their locking devices interval, whichever is first.
Core Support Shield Assembly Locking Devices: Loss of 100% of accessible bolts and Upper core barrel (UCB) bolts and their All plants material, damaged, SSHT bolts and their Subsequent examination to be locking devices.
locking devices distorted, or missing locking devices (CR-3 and determined after evaluating the locking devices (Wear or DB only) baseline results. See Figure 4-7.
Fatigue damage by failed bolts) shock pad bolts Visual (VT-3) examination of bolt Lower grid locking devices on the 10-year ISI and their locking devices interval.
(TMI-1 only) Inevl A4
Change c:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, itwas determined that the "effect (mechanism)" column only identifies the "effect (mechanism)" for the LCB bolts and does not clearly describe the "effect (mechanism)" for the associated locking devices. Also, the associated locking devices were omitted from the "Expansion Link" items. In addition, the locking devices were omitted from the "Examination Coverage" column. This component is listed as part of the Core Support Shield Assembly in Table 4-1 of MRP-227, Rev. 0 as shown below.
ItmApplicability Efet(Mechanism) Expansion Link (Note 2) Examination MsthodfFrequency Examination Coverage
__________________________________(Note 2)
Volumetric examination (UT) of the bolts during the next 10-year ISI UTS, LTS, and FD bolts interval from 1/1/2006.
Core Barrel Assembly SSHT bolts (CR-3 and DB Subsequent examination to be 100% of accessible bolts Lower core barrel (LCB) bolts and their All plants Cracking (SCC) only) determined after evaluating the locking devices baseline results. See Figure 4-8 Lower grid shock pad bolts (TMI-1 only) Visual (VT-3) examination of bolt locking devices on the 10-year ISI I_ I_ I_ I interval.
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 4-1 of MRP-227-A, as shown below.
RtOM Applicabfility Effect (Mechanism)
Expansion Unit (Note 2) Exainaion~i (Note 2)
MethodFrequteny Examination Coverage UTS, LTS, and FD bolts Volumetric examination (UT) of the and their locking devices bolts during the next 10-year ISI Bolt: Cracking (SCC) interval from 1/1/2006.
ySSHT bolts and their 100% of accessible bolts and Core Barrel Assembly Locking Devices: Loss of locking devices (CR-3 and Subsequent examination to be locking devices.
Lower core barrel (LCB) bolts and their All plants material, damaged, DB only) determined after evaluating the locking devices distorted, or missing baseline results.
locking devices (Wear or See Figure 4-8.
Fatigue damage by failed bolts) Lower grid shock pad bolts Visual (VT-3) examination of bolt and their locking devices locking devices on the 10-year ISI (TMI-1 I only) I interval. I A5
Change d:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, each of the noted mechanisms in the "Effect (Mechanism)"
column for the baffle-to-former bolts in Table 4-1 of MRP-227 Rev. 0 do not result in cracking. AREVA proposes that this column be modified to correctly reflect the effects and age-related degradation mechanisms for this component. This component is listed as part of the Core Barrel Assembly in Table 4-1 of MRP-227, Rev. 0 as shown below.
item Applicability Effect (Mechanism) Expansion Link (Note 2) Examination (Note 2) MethodlFrequency E ination Coverage Baseline volumetric examination (UT) no later than two refueling 100% of accessible bolts Core Barrel Assembly Cracking (IASCC, IE, Baffle-to-baffle bolts, outages from the beginning of the All plants IC/ISR/Fatigue/Wear, Baffle-to-former bolts Overload) Core barrel-to-former bolts license renewal period with subsequent examination after 10 See Figure 4-2 to 15 additional years.
In order to reflect the correct effects and age-related degradation mechanisms for this component, MRP proposes that the following modification be made and note added into Table 4-1 of MRP-227-A to the "Effect (Mechanism)" column shown below.
Item Applicability Effect (Mechanism) Expansion Unk (Note 2) Extaminaton M lFrequenc Examination Coverage (Note 2)
Cracking (IASCC, IE, Baseline volumetric examination (UT) no later than two refueling 100% of accessible bolts.
Core Barrel Assembly Overload) Baffle-to-baffle bolts, outages from the beginning of the All plants Baffle-to-former bolts Core barrel-to-former bolts license renewal period with (Note 4) subsequent examination after 10 See Figure 4-2.
(Note__4)to 15 additional years.
Notes:
- 1. A verification of the operation of each vent valve shall also be performed through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred. Examine the valves for evidence of scratches, pitting, embedded particles, variation in coloration of the seating surfaces, cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve inservice test programs (see AREVA doc. BAW-2248A, page 4.3 and Table 4-1).
- 2. Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1.
- 3. Expansion to LCB applies if the required Primary examination of LCB has not been performed as scheduled in this table.
- 4. The primary aging degradation mechanisms for loss of joint tightness for this item are ICand ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to IC/lSR. Bolt stress relaxation cannot be readily inspected by NDE. Only bolt cracking is inspected by UT inspection in this table. The effect of loss of joint tightness on the functionality will be addressed by analysis of the core barrel assembly.
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Change e:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that the "effect (mechanism)" column only identifies the "effect (mechanism)" for the UTS bolts and SSHT studs/nuts or bolts and does not clearly describe the "effect (mechanism)" for the associated locking devices. Also, the associated locking devices were omitted from the "Item" and "Primary Link" columns. In addition, the locking devices were omitted from the "Examination Method" and "Examination Coverage" columns. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
tem Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly All plants Upper thermal shield bolts (UTS) 100% of accessible bolts Core Barrel Assembly Cracking (SCC) UCB and LCB bolts Volumetric examination (UT)
Surveillance specimen holder tube CR-3, DB See Figure 4-7 (SSHT) studs/nuts (CR-3) or bolts (DB)
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 4-4 of MRP-227-A.
ROM Applicability Effect (Mechanism) Primary ink (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly Bolt: Cracking (SCC)
Upper thermal shield bolts (UTS) and All plants their locking devices Bolt: Volumetric examination (UT). 100% of accessible bolts and Locking Devices: Loss of UCB and LCB bolts and locking devices.
Core Barrel Assembly material, damaged, their locking devices distorted, or missing Locking Devices: Visual (VT-3)
Surveillance specimen holder tube CR-3, DB locking devices (Wear or examination. See Figure 4-7.
(SSHT) studs/nuts (CR-3) or bolts (DB) Fatigue damage by and their locking devices failed bolts)
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Change f:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that the words-no examination requirements-were omitted from the "Examination Method" column for the core barrel cylinder and former plates items. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
Item Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly Inaccessible.
Core barrel cylinder (including vertical Cracingdeticlu Justify by evaluation or by and circumferential seam welds) All plants cracking readily detectable Baffle plates SeepFiguren4-Former plates I In order to reflect this, MRP proposes that the following wording to be inserted into the "Examination Method" column of Table 4-4 of MRP-227-A.
Item Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly No examination requirements, Inaccessible.
Core barrel cylinder (including vertical Cracking (IE), including Corebanrurentl cylider (cludig vAll plants readily detectable Baffle plates and circumferential seam welds) cracking Justify by evaluation or by See Figure 4-2.
Former plates replacement.
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As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, each of the noted mechanisms in the "Effect (Mechanism)"
column for the baffle-to-baffle bolts and core barrel-to-former bolts in Table 4-4 of MRP-227 Rev. 0 do not result in cracking. AREVA proposes that this column be modified to correctly reflect the effects and age-related degradation mechanisms for this component. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
Itm pliailt Effect (Mechmnism~) Primary Link (Note 1) EaiainMto Nt )EaiainCvrg Internal baffle-to-baffle bolts: N/A No examination requirements, Justify by evaluation or by See Figure 4-Core Barrel Assembly Cracking (IASCC, IE, replacement. SeeFigure_4-2 Baffle-to-baffle bolts All plants IC/ISR/Fatigue/Wear, Baffle-to-former bolts External baffle-to-baffle bolts, Core barrel-to-former bolts Overload) Barrel-to-former bolts: Inaccessible No examination requirements, Justify by evaluation or by See Figure 4-2 replacement.
In order to reflect the correct effects and age-related degradation mechanisms for this component, MRP proposes that the following modification be made and note added into Table 4-4 of MRP-227-A to the "Effect (Mechanism)" column shown below.
Item Applcabiy Effect (Mechanism) Primary nk (Note 1) . Ech aminatism.
Internal baffle-to-baffle bolts:
No examination requirements. N/A.
Justify by evaluation or by See Figure 4-2.
Core Barrel Assembly Cracking (IASCC, IE, replacement.
Baffle-to-baffle bolts All plants Overload) Baffle-to-former bolts External baffle-to-baffle bolts, Core barrel-to-former bolts (Note 2) Barrel-to-former bolts:
Inaccessible.
No examination requirements.
See Figure 4-2.
Justify by evaluation or by replacement.
Note:
- 1. Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-1.
The primary aging degradation mechanisms for loss ofjoint tightness for these items are IC and ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to IC/ISR. Bolt stress relaxation cannot be readily inspected by NDE. Only bolt cracking is inspected by UT inspection in this table. The effect of loss of closure integrity on the functionality will be addressed by analysis of the core barrel assembly.
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Change h:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, itwas determined that the words-no examination requirements-were omitted from the "Examination Method" column for the extemal baffle-to-baffle bolts locking devices, including locking welds and for the core barrel-to-former bolts locking devices, including locking welds. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
Item Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly Locking devices, including Inaccessible.
Locking devices, including locking All plants Cracking (IASCC, IE) locking welds, of baffle-to- Justify by evaluation or by welds, for the external baffle-to-baffle former bolts or internal replacement.
bolts and core barrel-to-former bolts baffle-to-baffle bolts See Figure 4-2.
In order to reflect this, MRP proposes that the following wording to be inserted into the "Examination Method" column of Table 4-4 of MRP-227-A.
Item Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Core Barrel Assembly Locking devices, including No examination requirements. Inaccessible.
Locking devices, including locking All plants Cracking (IASCC, IE) locking welds, of baffle-to-welds, for the external baffle-to-baffle former bolts or internal bolts and core barrel-to-former bolts baffle-to-baffle bolts replacement.
by Figuro 4-2.
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Chancqe i:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that the "effect (mechanism)" column only identifies the "effect (mechanism)" for the lower grid shock pad bolts and does not clearly describe the "effect (mechanism)" for the associated locking devices. Also, the associated locking devices were omitted from the "Item" and "Primary Link" columns. In addition, the locking devices were omitted from the "Examination Method" and "Examination Coverage" columns. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
mi Effect (Mechanism) Primary nk (Note 1) Examination Method (Not 1) Exunlnation Coverage Lower Grid Assembly 100% of accessible bolts TMI-1 Cracking (SCC) UCB and LCB bolts Volumetric examination (UT)
Lower grid shock pad bolts See Figure 4-4 In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 4-4 of MRP-227-A.
kmer Applcbl,111t Effect (mehanim) Prmay Link (Note 1) Exmination Method (NoeW 1) Examination Coverage Bolt: Cracking (SCC)
Bolt: Volumetric examination (UT). 100% of accessible bolts and Lower Grid Assembly Locking Devices: Loss of UCB and LCB bolts and locking devices.
Lower grid shock pad bolts and their TMI-1 material, damaged, their locking devices locking devices distorted, or missing Locking Devices: Visual (VT-3) locking devices (Wear or examination. See Figure 4-4.
Fatigue damage by I__ _ _ I___
failed bolts) I I I All
Change i:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that the "effect (mechanism)" column only identifies the "effect (mechanism)" for the LTS bolts and flow distributor bolts and does not clearly describe the "effect (mechanism)" for the associated locking devices. Also, the associated locking devices were omitted from the "Item" and "Primary Link" columns. In addition, the locking devices were omitted from the "Examination Method" and "Examination Coverage" columns. This component is listed as part of the Core Barrel Assembly in Table 4-4 of MRP-227, Rev. 0 as shown below.
Item Appicabiilty Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Lower Grid Assembly Lower thermal shield bolts (LTS) 100% of accessible bolts All plants Cracking (SCC) UCB and LCB bolts Volumetric examination (UT)
Flow Distributor Assembly See Figure 4-8 Flow distributor bolts (FD)
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 4-4 of MRP-227-A.
Item Applicability Effect (Mechanism) Primary Unk (Note 1) Examination Method (Note 1) Examination Coverage Lower Grid Assembly Bolt: Cracking (SCC)
Lower thermal shield bolts (LTS) and Bolt: Volumetric examination (UT). 100% of accessible bolts and their locking devices Locking Devices: Loss of UCB and LCB bolts and locking devices.
All plants material, damaged, their locking devices Flow Distributor Assembly distorted, or missing Locking Devices: Visual (VT-3)
Flow distributor bolts (FD) and their locking devices (Wear or examination. See Figure 4-8.
locking devices Fatigue damage by failed bolts)
A12
Change k:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of a records review and accessibility evaluation, it was determined that the "CSS vent valve disc shaft or hinge pin" was inaccessible. This component is listed as part of the Core Support Shield Assembly in Table 5-1 of MRP-227, Rev. 0 as shown below.
MmApiaiiy Eaixpnation Accepace ion Epans Uk xasinCiei Additional Examnation 4
. .... .C.. . e... 1) _. ... . .. .. .Acceptance
. C rL..
Core Support Shield Assembly Visual (VT-3) examination.
CSS vent valve top retaining ring SS vent valve bottom retaining ring ll plants The specific relevant condition None N/A N/A is evidence of damaged or CSS vent valve disc shaft or hinge pin fractured material, and missing In order to reflect the actual generic condition and to clarify the requirements, MRP proposes that the following two rows be inserted into MRP-227-A and the existing row (shown above) be deleted.
iter Aplicbilty xamnatonccptance (Nt 1)
-xasoUn~)Epnso
__e rt AdditonalExamheuion A__ - Cr Visual (VT-3) examination.
Core Support Shield Assembly CSS vent valve top retaining ring All plants The specific relevant condition None N/A. N/A.
CSS vent valve bottom retaining ring is evidence of damaged or fractured material, and missing Inaccessible.
Core Support Shield Assembly All plants None N/A. N/A.
CSS vent valve disc shaft or hinge pin Justify by evaluation or replacement.
A13
Change 1:
As noted in the meeting between the NRC and MRP/lndustry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, itwas determined that associated locking devices were omitted from the "Expansion Unk(s)" column for each of the expansion link bolts. This component is listed as part of the Core Support Shield Assembly in Table 5-1 of MRP-227, Rev. 0 as shown below.
I R Examination Acceptance Criteria (Note 1) L Expansion Unk(s)
E C Expansion Criteria
- 1) Confirmed unacceptable indications Additional Examlinton Acceptance Criteria exceeding 10% of the UCB bolts shall require that the UT examination be expanded by the completion of the next refueling outage to include:
For all plants
- 1) Volumetric (UT) examination 100% of the accessible UTS bolts, 100% of the of the UCB bolts. accessible LTS bolts, 100% of the accessible FD bolts, LCB (Note 2)
Additionally for TMI-1 1) The examination The examination acceptance UT examination to include 100% of the acceptance criteria for the UT criteria for the UT of the UCB UTS, LTS, and accessible lower grid shock pad bolts, of the expansion bolting shall bolts shall be established as FD bolts be established as part of the part of the examination technical Core Support Shield Additionally for CR-3 and DB examination technical justification.
Assembly justification.
UT examination to include 100% of the All plants SSHT bolts Upper core barrel (UCB) bolts accessible SSHT bolts.
(CR-3 and DB and their locking devices 2) Visual (VT-3) examination of only) 2) The specific relevant the UCB bolt locking devices.
condition for the expansion of
- 2) Confirmed evidence of relevant conditions the VT-3 locking devices is exceeding 10% of the UCB bolt locking devices evidence of broken or missing Lower grid shall require that the VT-3 examination be The specific relevant condition bolt locking devices.
shock pad bolts expanded by the completion of the next for the VT-3 of the UCB bolt (TMI-1 only) refueling outage to include:
locking devices is evidence of broken or missing bolt locking For all plants devices.
100% of the accessible UTS, LTS, and FD bolt locking devices, Additionally for TMI-1 100% of the accessible lower grid shock pad bolt locking devices, A14
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 5-1 of MRP-227-A.
ftem Apliailt ~Exmnto
@ Accpac Epni o ExpwansonCieiaAdtonlEamnto I I Crtei (Nt ) n)I __ __eptance I A__ Criteri
- 1) Confirmed unacceptable indications exceeding 10% of the UCB bolts shall require that the UT examination be expanded by the completion of the next refueling outage to include:
For all plants
- 1) Volumetric (UT) LCB and their 100% of the accessible UTS bolts, 100% of the examination of the UCB locking devices accessible LTS bolts, 100% of the accessible FD bolts. (Note 2) bolts, Additionally for TMI-1 UT examination to include 100% of the accessible 1) The examination acceptance The examination acceptance UTS, LTS, and lower grid shock pad bolts, criteria for the UT of the criteria for the UT of the UCB FD bolts and expansion bolting shall be bolts shall be established as their locking Core Support Shield part of the examination devices I Additionally for CR-3 and DB established as part of the examination technical Assembly technical justification. UT examination to include 100% of the accessible justification.
SSHT bolts.
Upper core barrel (UCB) All plants SSHT bolts and bolts and their locking 2) Visual (VT-3) examination their locking devices of the UCB bolt locking 2) The specific relevant devices (CR-3 2) Confirmed evidence of relevant conditions condition for the expansion of devices. and DB only) exceeding 10% of the UCB bolt locking devices the VT-3 locking devices is shall require that the VT-3 examination be evidence of broken or missing expanded by the completion of the next refueling bolt locking devices.
The specific relevant Lower grid shock outage to include:
condition for the VT-3 of the pad bolts and For all olants UCB bolt locking devices is their locking evidence of broken or devices (TMI-1 100% of the accessible UTS, LTS, and FD bolt missing bolt locking devices. only) locking devices, Additionally for TMI-1 100% of the accessible lower grid shock pad bolt locking devices, Additionally for CR-3 and DB A15
Change m:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that associated locking devices were omitted from the "Expansion Link(s)" column for each of the expansion link bolts. This component is listed as part of the Core Barrel Assembly in Table 5-1 of MRP-227, Rev. 0 as shown below.
Item Applicability ExCaination Acceptance Expansion Unk(s) Expansion Criteria Additional Examination Criteria (Note 1) AncepxpnsoCriteria
- 1) Confirmed unacceptable indications exceeding 10% of the LCB bolts shall require that the UT examination be expanded by the completion of the next refueling outage to include:
- 1) Volumetric (UT) examination For all plants of the LCB bolts.
100% of the accessible UTS bolts, 100% of the accessible LTS 1) The examination bolts, 100% of the accessible FD acceptance criteria for the UT The examination acceptance bolts, of the expansion bolting shall criteria for the LIT of the LCB UTS, LTS, and FD be established as part of the bolts part ofshall the be establishedtechnical examination as bolts Additionally for TMI-1 examination technical Core Barrei Assembly justification. 100% of the accessible lower grid justification.
CorerBarrel bAselm(LB)bolts tSSHT justi . bolts (CR-3 and shock pad bolts, Lower Gore barrel (LOB) bolts and their All plants DB only) Additionally for CR-3 and DB locking devices 2) Visual (VT-3) examination of the LCB bolt locking devices. 100% of the accessible SSHT 2) The specific relevant Lower grid shock pad bolts by the completion of the condition for the expansion of bolts (TMI-1 only) next refueling outage. the VT-3 of the locking The specific relevant condition devices is evidence of broken for the VT-3 of the LCB bolt or missing bolt locking locking devices is evidence of 2) Confirmed evidence of relevant devices.
broken or missing bolt locking conditions exceeding 10% of the devices. LCB bolt locking devices shall require that the VT-3 examination be expanded by the completion of the next refueling outage to include:
For all plants I___
n% nf thp. q r.Pq.ihlP. I ITS A16
In order to reflect these above omissions, MRP proposes that the additional wording be inserted into Table 5-1 of MRP-227-A.
- 1) Confirmed unacceptable indications exceeding 10% of the LCB bolts shall require that the UT examination be expanded by the completion of the next refueling outage to include:
- 1) Volumetric (UT) examination For all plants of the LOB bolts. 100% of the accessible UTS bolts, 100% of the accessible LTS 1) The examination UTS, LTS, and FD bolts, 100% of the accessible FD acceptance criteria for the UT The examination acceptance bolts and their locking bolts, of the expansion bolting shall criteria for the UT of the LCB devices be established as part of the bolts shall be established as Additionally for TMI-1 examination technical part of the examination technical 100% of the accessible lower grid justification.
Core Barrel Assembly justification. SSHT bolts and their shock pad bolts, Lower core barrel (LCB) bolts and their All plants locking devices (CR-3 locking devices 2) Visual (VT-3) examination of and DB only) Additionally for CR-3 and DB the LCB bolt locking devices. 100% of the accessible SSHT 2) The specific relevant bolts by the completion of the condition for the expansion of Lower grid shock pad next refueling outage. the VT-3 of the locking bolts and their locking devices is evidence of broken The specific relevant condition devices (TMI-1 only) or missing bolt locking for the VT-3 of the LCB bolt locking devices is evidence of 2) Confirmed evidence of relevant devices.
broken or missing bolt locking conditions exceeding 10% of the devices. LCB bolt locking devices shall require that the VT-3 examination be expanded by the completion of the next refueling outage to include:
For all plants I ion% nf th, . nr.rq.c.*.ihlP I ITS A17
Change n:
As noted in the meeting between the NRC and MRP/Industry on 10/14/2010, AREVA is working with its owners to support implementation of MRP-227. As part of this effort, it was determined that an incorrect wording ("former plate" in lieu of "baffle plate") was erroneously used in the "Expansion Criteria" column for the Core Barrel Assembly Baffle-to-former bolts component. This component is listed as part of the Core Barrel Assembly in Table 5-1 of MRP-227, Rev. 0 as shown below.
Item Applicability Examination Acceptance Criteria (Note 1) Expansion Unk(s) Expansion Criteria Additional Examination Acceptance Criteria Confirmed unacceptable indications in greater than or equal to 5% (or 43) of the baffle-Baseline volumetric (UT) to-former bolts, provided that examination of the baffle-to- none of the unacceptable bolts former bolts. are on former elevations 3, 4, and 5, or greater than 25% of the bolts Baffle-to-baffle bolts, on a single former plate, shall Core Barrel Assembly Arequire an evaluation of the All plants The examination acceptance Care barrel-to-former internal baffle-to-baffle bolts for N/A Baffle-to-former bolts criteria for the UT of the baffle- bolts the purpose of determining to-former bolts shall be whether to examine or replace the established as part of the internal baffle-to-baffle bolts. The examination technical evaluation may include external justification. baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining whether to replace them.
A18
In order to reflect correction of this wording, MRP proposes that the modified wording be inserted into Table 5-1 of MRP-227-A.
RonApliability Examination Acceptance Exason s Exaso Crtei Additional Examination
_____Criteria (Note 1) _ _ __Acceptence Criteria Confirmed unacceptable indications in greater than or equal to 5% (or 43) of the baffle-Baseline volumetric (UT) to-former bolts, provided that examination of the baffle-to- none of the unacceptable bolts former bolts. are on former elevations 3, 4, and 5, or greater than 25% of the bolts Baffle-to-baffle bolts, on a single baffle plate, shall Core Barrel Assembly require an evaluation of the All plants The examination acceptance Core barrel-to-former internal baffle-to-baffle bolts for N/A Baffle-to-former bolts criteria for the UT of the baffle- bolts the purpose of determining to-former bolts shall be whether to examine or replace the established as part of the internal baffle-to-baffle bolts. The examination technical evaluation may include external justification, baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining whether to replace them.
A19
- 5) The MRP proposes the following changes to the CE and Westinghouse tables in MRP-227-Rev. 0. Changes are indicated in track changes as well as bar on the left side of each row with changes. Note that these tables do not include all of the proposed changes to the CE and Westinghouse Tables (see point 3). A set of tables with all the changes combined could be provided to the NRC later if necessary.
a) The MRP proposes to use the additions in the Effect (Mechanism) column in Tables 4-2, 4-3, 4-5, 4-6, 4-8 and 4-9 as shown in the tables below.
b) The MRP proposes to clarify some of the 4-2 CE table entries for TLAA/fatigue analysis by replacing the words "plant-specific fatigue analysis" with the words "evaluation to determine the potential location and extent of fatigue cracking" as shown in the tables below.
c) The MRP proposes to replace the title in column 4 "Primary Link" with "Reference" for Tables 4-8 and 4-9 as shown in the tables below. The MRP proposes to make the Westinghouse "Remaining core barrel welds" consistent between Tables 3-3, 4-3, 4-6 and 5-3 as shown in the tables below.
d) The MRP proposes to delete the sentence "Replacement of 304 springs by 403 springs is required when the spring stiffness is determined to relax beyond design tolerance" in the Westinghouse Plants Primary Components Table (Table 4-1) for the Alignment and Interfacing Components Internals hold down spring item.
e) The MRP proposes to delete the text "or as supported by plant-specific justification" for the core-shroud bolts item in CE Table 4-2 and the baffle-former bolts item in Westinghouse Table 4-3.
A20
Table 4-2 CE Plants Primary Components Alibili Effect Expansion Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency (Note 1) Examination Coverage Core Shroud Bolted plant Cracking (IASCC, Core support column Baseline volumetric (UT) 100% of accessible boltsT-er-as Assembly (Bolted) designs Fatigue) bolts, Barrel-shroud examination between 25 and 35 supp...ed by plant spe.ifie Core shroud bolts Aging bolts EFPY, with subsequent jutiifieetien.: Heads are Management (IE examination after 10 to 15 accessible from the core side.
and ISR) additional EFPY to confmn UT accessibility may be stability of bolting pattern. affected by complexity of Re-examination for head and locking device high-leakage core designs designs.
requires continuing inspections See Figure 4-24.
on a ten-year interval.
Core Shroud Plant designs Cracking (IASCC) Remaining axial welds Enhanced visual (EVT-1) Axial and horizontal weld Assembly (Welded) with core Aging examination no later than seams at the core shroud Core shroud plate- shrouds Management (IE) 2 refueling outages from the re-entrant comers as visible former plate weld assembled in beginning of the license from the core side of the two vertical renewal period and subsequent shroud, within six inches of sections examination on a ten-year central flange and horizontal interval. stiffeners.
See Figures 4-12 and 4-14.
Core Shroud Plant designs Cracking (IASCC) Remaining axial welds, Enhanced visual (EVT-1) Axial weld seams at the core Assembly (Welded) with core Aging ribs and rings examination no later than shroud re-entrant comers, at Shroud plates shrouds Management (IE) 2 refueling outages from the the core mid-plane (+ three assembled beginning of the license feet in height) as visible from with full- renewal period and subsequent the core side of the shroud.
height shroud examination on a ten-year See Figure 4-13.
plates interval.
A21
Table 4-2 CE Plants Primary Components Aliblity Effect Expansion Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency (Note 1) Examination Coverage Core Shroud Bolted plant Distortion (Void None Visual (VT-3) examination no Core side surfaces as Assembly (Bolted) designs Swelling) later than 2 refueling outages indicated.
Assembly including: from the beginning of the See Figures 4-25 and 4-26.
" Abnormal license renewal period.
interaction with Subsequent examinations on a fuel assemblies ten-year interval.
" Gaps along high fluence shroud plate joints
" Vertical displacement of shroud plates near high fluence joint Aging Management (IE)
Core Shroud Plant designs Distortion (Void None Visual (VT-1) examination no If a gap exists, make three to Assembly (Welded) with core Swelling), as later than 2 refueling outages five measurements of gap Assembly shrouds evidenced by from the beginning of the opening from the core side at assembled in separation between license renewal period, the core shroud re-entrant two vertical the upper and Subsequent examinations on a comers. Then, evaluate the sections lower core shroud ten-year interval, swelling on a plant-specific segments basis to determine frequency Aging and method for additional Management (IE) examinations.
See Figures 4-12 and 4-14.
Core Support All plants Cracking (SCC) Remaining core barrel Enhanced visual (EVT-1) 100% of the accessible Barrel Assembly assembly welds, core examination no later than surfaces of the upper flange Upper (core support support column welds two refueling outages from the weld.
barrel) flange weld beginning of the license See Figure 4-15.
renewal period. Subsequent examinations on a ten-year interval.
A22
Table 4-2 CE Plants Primary Components Alibili Effect Expansion Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency (Note 1) Examination Coverage Core Support All plants Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Barrel Assembly demonstrated by time-limited defined by plant speeifie Lower flange weld aging analysis (TLAA), aal:ysisevaluation to
.igu.e enhanced visual (EVT-1) determine the potential examination, no later than location and extent of fatigue 2 refueling outages from the cracking.
beginning of the license See Figure 4-15.
renewal period. Subsequent examination on a ten-year interval.
Lower Support All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Structure a core support Aging demonstrated by time-limited defined by plant speeifie Core support plate plate Management (IE) aging analysis (TLAA), fatigue-analysisevaluation to enhanced visual (EVT- 1) determine the potential examination, no later than location and extent of fatigue 2 refueling outages from the cracking.
beginning of the license See Figure 4-16.
renewal period. Subsequent examination on a ten-year interval.
Upper Internals All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Assembly core shrouds demonstrated by time-limited defined by plant speeifie Fuel alignment plate assembled aging analysis (TLAA), fatigue- analysisevaluation to with full- enhanced visual (EVT- 1) determine the potential height shroud examination, no later than location and extent of fatigue plates 2 refueling outages from the cracking.
beginning of the license See Figure 4-17.
renewal period. Subsequent examination on a ten-year interval.
A23
Table 4-2 CE Plants Primary Components Alibili Effect Expansion Examination Item Applicability (Mechanism) Link (Note 1) Method/Frequency (Note 1) Examination Coverage Control Element All plants with Cracking (SCC, Remaining instrument Visual (VT-3) examination, no 100% of tubes in peripheral Assembly instrument Fatigue) that guide tubes within the later than 2 refueling outages CEA shroud assemblies (i.e.,
Instrument guide guide tubes in results in missing CEA shroud assemblies from the beginning of the those adjacent to the perimeter tubes the CEA supports or license renewal period. of the fuel alignment plate).
shroud separation at the Subsequent examination on a See Figure 4-18.
assembly welded joint ten-year interval.
between the tubes Plant-specific component and supports integrity assessments may be required if degradation is detected and remedial action is needed.
Lower Support All plants with Cracking (Fatigue) None Enhanced visual (EVT-1) Examine beam-to-beam welds, Structure core shrouds that results in a examination, no later than in the axial elevation from the Deep beams assembled detectable surface- 2 refueling outages from the beam top surface to 4 inches with full- breaking indication beginning of the license below.
height shroud in the welds or renewal period. Subsequent See Figure 4-19.
plates beams examination on a ten-year Agin interval, if adequacy of Management (OE) remaining fatigue life cannot be demonstrated.
Note: 1. Examination acceptance criteria and expansion criteria for the CE components are in Table 5-2.
A24
Table 4-3 Westinghouse Plants Primary Components Alibili Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Control Rod Guide All plants Loss of Material None Visual (VT-3) examination no 20% examination of the Tube Assembly (Wear) later than 2 refueling outages number of CRGT assemblies, Guide plates (cards) from the beginning of the with all guide cards within license renewal period, and no each selected CRGT assembly earlier than two refueling examined.
outages prior to the start of the See Figure 4-20.
license renewal period.
Subsequent examinations are required on a ten-year interval.
Control Rod Guide All plants Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)
Tube Assembly Fatigue) instrumentation (BMI) examination to determine the CRGT lower flange weld Lower flange welds Aging column bodies, Lower presence of crack-like surface surfaces and adjacent base Management (IE support column bodies flaws in flange welds no later metal.
and TE) (cast) than 2 refueling outages from See Figure 4-21.
the beginning of the license renewal period and subsequent examination on a ten-year interval.
Core Barrel All plants Cracking (SCC) Remaining core barrel Periodic enhanced visual 100% of one side of the Assembly welds, Lower support (EVT-1) examination, no later accessible surfaces of the Upper core barrel column bodies (non than 2 refueling outages from selected weld and adjacent flange weld cast) the beginning of the license base metal.
renewal period and subsequent See Figure 4-22.
examination on a ten-year interval.
A25
Table 4-3 Westinghouse Plants Primary Components Alibility Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Baffle-Former All plants with Cracking (IASCC, None Visual (VT-3) examination, Bolts and locking devices on Assembly baffle-edge Fatigue) that with baseline examination high fluence seams. 100% of Baffle-edge bolts bolts results in between 20 and 40 EFPY and components accessible from
" Lost or broken subsequent examinations on a core side.
locking devices ten-year interval. See Figure 4-23.
" Failed or missing bolts
- Protrusion of bolt heads Aging Management (IE and ISR)
Baffle-Former All plants Cracking (IASCC, Lower support column Baseline volumetric (UT) 100% of accessible bolts. or-as Assembly Fatigue) bolts, Barrel-former examination between 25 and 35 supp....d byaplan z,....ic Baffle-former bolts Aging bolts EFPY, with subsequent j .siea..*i*
... -Heads accessible Management (lE examination after 10 to 15 from the core side. UT and ISR) additional EFPY to confirm accessibility may be affected stability of bolting pattern. Re- by complexity of head and examination for high-leakage locking device designs.
core designs requires See Figures 4-23 and 4-24.
continuing examinations on a ten-year interval.
A26
Table 4-3 Westinghouse Plants Primary Components Item Alibilit Applicability (Mechanism)
Effect Expansion (Note 1)Link Examination(Note 1)
Method/Frequency Examination Coverage Baffle-Former All plants Distortion (Void None Visual (VT-3) examination to Core side surface as indicated.
Assembly Swelling), or check for evidence of See Figures 4-24, 4-25, 4-26 Assembly Cracking (IASCC) distortion, with baseline and 4-27.
(Includes: bBBaffle that results in: examination between 20 and 40 gftlates, I&Baffle- e Abnormal EFPY and subsequent Eedge Rbolts.-A-se, interaction with examinations on a ten-year and indirect effects of fuel assemblies interval.
void swelling in e Gaps along high fFormer plates). fluence baffle joint e Vertical displacement of baffle plates near high fluence joint
- Broken or damaged edge bolt locking systems along high fluence baffle joints Alignment and All plants with Distortion (Loss of None Direct measurement of spring Measurements should be taken Interfacing 304 stainless Load) height within three cycles of the at several points around the Components steel hold Note: This beginning of the license circumference of the spring, Internals hold down down springs mechanism was not renewal period. If the first set with a statistically adequate spring strictly identified in of measurements is not number of measurements at the original list of sufficient to determine life, each point to minimize age-related spring height measurements uncertainty. R, ........ of degradation must be taken during the next 301 spr.ngs by 103 springs is mechanisms [7]. two outages, in order to required when the spring extrapolate the expected spring StiffneSS is dterMiRned to re!ax height to 60 years. beyond design t-elrmanc.
See Figure 4-28.
A27
Table 4-3 Westinghouse Plants Primary Components Alibi Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Thermal Shield All plants with Cracking (Fatigue) None Visual (VT-3) no later than 2 100% of thermal shield Assembly thermal shields or Loss of Material refueling outages from the flexures.
Thermal shield (Wear) that results beginning of the license See Figures 4-29 and 4-36.
flexures in thermal shield renewal period. Subsequent flexures excessive examinations on a ten-year wear, fracture, or interval.
complete separation Note:
- 1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.
A28
Table 4-5 CE Plants Expansion Components Effect Primary Link Examination Method Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Shroud Bolted plant Cracking (IASCC, Core shroud bolts Volumetric (UT) examination, 100% (or as supported by plant-Assembly (Bolted) designs Fatigue) with initial and subsequent specific justification) of barrel-Barrel-shroud bolts Aging examination frequencies shroud and guide lug insert Management (lE dependent on the results of core bolts with neutron fluence and ISR) shroud bolt examinations, exposures > 3 displacements per atom (dpa).
See Westinghouse design Figure 4-23.
Core Support Barrel All plants Cracking (SCC, Upper (core support Enhanced visual (EVT-1) 100% of accessible welds and Assembly Fatigue) barrel) flange weld examination, with initial and adjacent base metal.
Lower core barrel subsequent examinations See Figure 4-15.
flange dependent on the results of the upper (core support barrel) flange weld examinations.
Core Support Barrel All plants Cracking (SCC) Upper (core support Enhanced visual (EVT-1) 100% of one side of the Assembly Aging barrel) flange weld examination, with initial and accessible weld and adjacent Remaining core barrel Management (IE) subsequent examinations base metal surfaces for the weld assembly welds dependent on the results of core with the highest calculated barrel assembly upper flange operating stress.
weld examinations. See Figure 4-15.
Lower Support All plants Cracking (SCC, Upper (core support Visual (VT-3) examination, Examination coverage Structure except those IASCC, Fatigue) barrel) flange weld with initial and subsequent determined by plant-specific Core support column with core including damaged examinations based on plant analysis.
welds shrouds or fractured evaluation of SCC See Figures 4-16 and 4-31.
assembled material susceptibility and with full- Agin demonstration of remaining height shroud Management (IE) fatigue life.
plates A29
Table 4-5 CE Plants Expansion Components Effect Primary Link Examination Method Item Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage Core Shroud Bolted plant Cracking (IASCC, Core shroud bolts Ultrasonic (UT) examination, 100% (or as supported by Assembly (Bolted) designs Fatigue) with initial and subsequent plant-specific analysis) of core Core support column Aging examination frequencies support column bolts with bolts Management (IE) dependent on the results of core neutron fluence exposures shroud bolt examinations. >3 dpa.
See Figures 4-16 and 4-33.
Core Shroud Plant designs Cracking (IASCC) Core shroud plate- Enhanced visual (EVT- 1) Axial weld seams other than the Assembly (Welded) with core former plate weld examination, with initial and core shroud re-entrant comer Remaining axial welds shrouds subsequent examination welds at the core mid-plane.
assembled in frequencies dependent on the See Figure 4-12.
two vertical results of the core shroud weld sections examinations.
Core Shroud Plant designs Cracking (IASCC) Shroud plates of Enhanced visual (EVT- 1) Axial weld seams other than Assembly (Welded) with core Agin welded core shroud examination, with initial and core shroud re-entrant comer Remaining axial welds shrouds Management (IE) assemblies subsequent examination welds at the core mid-plane, assembled frequencies dependent on the plus ribs and rings.
Ribs and rings with full- results of the core shroud weld See Figure 4-13.
height shroud examinations.
plates Control Element All plants Cracking (SCC, Peripheral instrument Visual (VT-3) examination, 100% of tubes in CEA shroud Assembly with Fatigue) that guide tubes within with initial and subsequent assemblies.
Remaining instrument instrument results in missing the CEA shroud examinations dependent on the See Figure 4-18.
guide tubes guide the CEAtubes in supports or assemblies results of the instrument guide separation at the tubes examinations.
shroud welded joint assembly between the tubes and supports.
Note: 1. Examination acceptance criteria and expansion criteria for the CE components are in Table 5-2.
A30
Table 4-6 Westinghouse Plants Expansion Components Item Applicability Effect Primary Link Examination Method Examination Coverage (Mechanism) (Note 1) (Note 1) Examination Coverage Core Barrel All plants Cracking (IASCC, Baffle-former bolts Volumetric (UT) examination, 100% of accessible bolts.
Assembly Fatigue) with initial and subsequent Accessibility may be limited by Barrel-former bolts Agin examinations dependent upon presence of thermal shields or Management (IE, results of baffle-former bolt neutron pads.
Void Swelling and examinations. See Figure 4-23.
ISR)
Lower Support All plants Cracking (IASCC, Baffle-former bolts Volumetric (UT) examination, 100% of accessible bolts or as Assembly Fatigue) with initial and subsequent supported by plant-specific Lower support column ASging examinations dependent on justification.
bolts Management (LE results of baffle-former bolt See Figures 4-32 and 4-33.
and ISR) examinations.
Core Barrel All plants Cracking (SCC, Upper core barrel Enhanced visual (EVT-1) 100% of one side of the Assembly Fatigue) flange weld examination, with initial accessible surfaces of the Remaining Welds Agin examination and re- selected weld and adjacent base (Core barrel flanges, Management (lE of examination frequency metal.
core barrel outlet lower sections) dependent on the examination See Figure 4-22 nozzles),OwOr-eOOF results for upper core barrel baF-_l-fan-_ _ _ __d flange.
Lower Support All plants Cracking (IASCC) Upper core barrel Enhanced visual (EVT- 1) 100% of accessible surfaces.
Assembly Aging flange weld examination, with initial See Figure 4-34.
Lower support column Management (IE) examination and re-bodies examination frequency (non cast) dependent on the examination results for upper core barrel flange weld.
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Table 4-6 Westinghouse Plants Expansion Components Item Applicability Effect Primary Link Examination Method Examination Coverage Itmpliailt (Mechanism) (Note 1) (Note 1)
Lower Support All plants Cracking (IASCC) Control rod guide Visual (EVT-1) examination. 100% of accessible support Assembly including the tube (CRGT) lower columns.
Lower support column detection of flanges See Figure 4-34.
bodies fractured support (cast) columns Aging Management (IE)
Bottom-Mounted All plants Cracking (Fatigue) Control rod guide Visual (VT-3) examination of 100% of BMI column bodies Instrumentation including the tube (CRGT) lower BMI column bodies as for which difficulty is detected System detection of flanges indicated by difficulty of during flux thimble Bottom-Mounted completely insertion/withdrawal of flux insertion/withdrawal Instrumentation (BMI) fractured column thimbles. Flux thimble See Figures 4-35.
column bodies bodies insertion/withdrawal to be Aging monitored at each inspection Management (IE) interval.
Note:
- 1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.
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Table 4-8 CE Plants Existing Programs Components Item Ite Aplicbilty Applicability (Mechanism)
Effect LinkReference eFerenc Examination Method Examination Coverage Core Shroud All plants Loss of material ASME Code Section Visual (VT-3) examination, First 10-year ISI after 40 years Assembly (Wear) XI general condition examination of operation, and at each Guide lugs Aging for detection of excessive or subsequent inspection interval.
Guide lug inserts and Management (ISR) asymmetrical wear.
bolts Lower Support All plants Cracking (SCC, ASME Code Section Visual (VT-3) examination to Accessible surfaces at specified Structure with core IASCC, Fatigue) XI detect severed fuel alignment frequency.
Fuel alignment pins shrouds Aging pins, missing locking tabs, or assembled Management (IE excessive wear on the fuel with full- and ISR) alignment pin nose or flange.
height shroud plates Lower Support All plants Loss of material ASME Code Section Visual (VT-3) examination. Accessible surfaces at specified Structure with core (Wear) XI frequency.
Fuel alignment pins shroud Aging assembled in Management (1E two vertical and ISR) sections Core Barrel All plants Loss of material ASME Code Section Visual (VT-3) examination. Area of the upper flange Assembly (Wear) XI potentially susceptible to wear.
Upper flange A33
Table 4-9 Westinghouse Plants Existing Programs Components Effect E ...... .n Item Applicability (Mechanism) inkReference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Section XI Visual (VT-3) examination to All accessible surfaces at Core barrel flange (Wear) determine general condition for specified frequency.
excessive wear.
Upper Internals Assembly All plants Cracking (SCC, ASME Code Section XI Visual (VT-3) examination. All accessible surfaces at Upper support ring or skirt Fatigue) specified frequency.
Lower Internals Assembly All plants Cracking ASME Code Section XI Visual (VT-3) examination of All accessible surfaces at Lower core plate (IASCC, the lower core plates to detect specified frequency.
XL lower core plate (Note Fatigue) evidence of distortion and/or
- 1) Agin_ loss of bolt integrity.
Management (1E)
Lower Internals Assembly All plants Loss of material ASME Code Section XI Visual (VT-3) examination. All accessible surfaces at Lower core plate (Wear) specified frequency.
XL lower core plate (Note 1)
Bottom Mounted All plants Loss of material NUREG-1801 Surface (ET) examination. Eddy current surface Instrumentation System (Wear) Rev. 1 examination as defined in Flux thimble tubes plant response to IEB 88-09.
Alignment and Interfacing All plants Loss of material ASME Code Section XI Visual (VT-3) examination. All accessible surfaces at Components (Wear) specified frequency.
Clevis insert bolts (Note 2)
Alignment and Interfacing All plants Loss of material ASME Code Section XI Visual (VT-3) examination. All accessible surfaces at Components (Wear) specified frequency.
Upper core plate alignment pins Notes:
- 1. XL = "Extra Long" referring to Westinghouse plants with 14-foot cores.
- 2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.
A34
- 6) The MRP proposes to add the following reference to Section 8 of MRP-227-Rev. 0: "[26].
WCAP-1 7096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements - Revision 2", December 2009."
- 7) The MRP proposes to replace the current Appendix A in MRP-227-Rev. 0 called Aging Management Program Attributes by the EPRI DRAFT Input (12-01-09): New Appendix A to MRP-227 A called Operating Experience Summary provided in letter MRP 2009-091
(
Subject:
Transmittal of Initial Draft Material to Support NRC Update of NUREG 1801, "Generic Aging Lessons Leamed Report" (GALL)) sent to the NRC in December 2009.
- 8) The MRP proposes to replace the words in last paragraph Section 7.1 of MRP-227 with clarifying words about NEI-03-08. Specifically the words "Addendum D to NEI 03-08 [1]"
with the following: "Addendum E to NEI 03-08, Revision 2". Reference 1 will also be updated to reflect NEI 03-08, Revision 2, January 2010.
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MRP-227 Roadmap October 29, 2010 Appendix B MRP-227 Roadmap BI
MRP-227 Roadmap October 29, 2010 The following road map is intended to provide information to NRC staff that will facilitate their review of MRP-227. The goal is not to tell the technical story in a different fashion, but rather to provide an overview of the steps involved in development of MRP-227 and point the staff to the appropriate supporting documents. In preparing this roadmap, no new information has been provided. Everything noted in this roadmap has been excerpted from other references previously provided to the NRC staff as part of the MRP-227 review and RAI process.
The Materials Reliability Program (MRP) has developed inspection and evaluation (I&E) guidelines for managing long-term aging of pressurized water reactor (PWR) reactor internals.
Specifically, the guidelines are applicable to reactor internal structural components; they do not address fuel assemblies, reactivity control assemblies, or welded attachments to the reactor vessel.
The program to develop these guidelines has been underway for almost a decade, organized around a framework and strategy for managing effects of aging in PWR internals, dependent on a substantial database of material data and supporting evaluation results. The goal of this development was primarily to support license renewal, but the guidelines support reactor internals aging management for the current license period as well.
It is important to recognize that this effort relied on the previous work in MRP-205 (Issue Management Tables). These tables identified all safety significant issues for all PWR primary loop and internals components. Further, only two components were identified during the initial screening (step 1) that had any safety consequences that were dispositioned in the development of MRP-227; as explained in this roadmap.
The guidelines are applicable to nuclear steam supply system (NSSS) vendor Babcock &
Wilcox-designed (B&W), Combustion Engineering-designed (CE) and Westinghouse-designed (W) PWR internals. The guidelines are based on a broad set of assumptions about nuclear unit operation, which encompass the range of current unit conditions for the U.S. fleet of PWRs. The aging management strategy reports, MRP-231 for B&W and MRP-232 for CE and W, provide the basis for these guidelines. The functional evaluations, including the screening and the Failure Modes, Effects and Criticality Analysis (FMECA), that support the guidelines were based on representative B&W, W and CE PWR reactor vessel internals configurations, existing analyses, inspections, and operational histories, which were generally conservative, but not necessarily bounding in every parameter.
These guidelines do not reduce, alter, or otherwise affect current American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI or unit-specific licensing inservice inspection requirements. The guidelines do not replace the current licensing basis for the current and extended license periods, which have been reviewed and approved by the US NRC on a plant-specific basis based on NUREG- 1800 and NUREG- 1801.
The goal is to ensure the long-term safety, integrity, and reliability of PWR internals using proven and familiar methods for inspection, monitoring, surveillance, and reporting.
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MRP-227 Roadmap October 29, 2010 An experienced team consisting of utility, NSSS vendor and EPRI experts, representing a broad spectrum of reactor design, operations, and materials expertise, worked on the project. The team reviewed available data and industry experience on materials aging to develop a systematic approach for identifying and prioritizing inspection requirements for internals. The process used to develop the MRP-227 recommendations may be described in terms of the following sequence of steps:
Step 1 - Identify PWR internals components, materials, and environments Step 2 - Identify degradation screening criteria Step 3 - Characterize components and screen for degradation (A, non-A)
Step 4 - FMECA Review Step 5 - Severity categorization (A, B, C)
Step 6 - Engineering Evaluation and Assessment' Step 7 - Categorize for Inspection (Primary, Expansion, Existing, No Additional Measures) and Aging Management Strategy Step 8 - Preparation of MRP-227 I&E Guidelines The processing of the reactor internals components through these eight steps is outlined in the following paragraphs. The screening and categorization processes for B&W components is are contained described in MRP-189 Rev. 1, MRP-190, and MRP-231. The screening and categorization processes forO and the W and CE internals are described in MRP-191 and MRP-232.
In addition to the documents specifically focused on PWR reactor internals, two other resources were utilized - the Materials Degradation Matrix (MDM) and the PWR Issue Management Tables (IMTs) that are compiled in MRP-205, rRev. 1. The MDM was first issued in 2004. It documents all known relevant/plausible degradation mechanisms and materials, including welds, in the primary loop and reactor internals for BWRs and PWRsS. This document was developed with the support of domestic and international experts from NSSS vendors, national laboratories, utilities and consultants. (It is worth noting that NRC conducted a similar activity that is documented in their Expert Panel Report on Proactive Materials Degradation Assessment NUREG/CR-6923. It reached essentially the same conclusions.) The PWR IMTs used the information from the MDM and assessed, at a component level the consequences of failure, as well as inspection, mitigation and repair technology associated with that component. The MDM and IMTs are maintained as "living documents" and updated periodically.
Key to the development of MRP-205 was the extensive efforts by the NSSS vendors, key utility personnel and supporting experts to identify the failure consequences at a component level. This work is described in MRP-157 for B&W plants and in MRP-156 for W and CE plants. These documents were used extensively in the overall development of MRP-227.
1Step 6 has previously been identified as a "Functionality Evaluation" or "Functionality Assessment" in each of the reference documents, for which the chosen words unfortunately are now felt It was determined that these terms mayto have been somewhat misleading. It has been renamed herein as Engineering Evaluation and Assessment to more closely describe for clarification of the work that has actually been performed.
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MRP-227 Roadmap October 29, 2010 Finally, the following is a list of key assumptions or premises used in the development of MRP-227.
- 1. The 1995 Statements of Consideration related to the revised License Renewal Rule (60 FR 22488) address the relationship of license renewal to plant licensing bases. In amending the "first principle of license renewal", the SOC states:
" Thefirst principleof license renewal was that, with the exception of age-related degradationunique to license renewal andpossibly afew other issues relatedto safety only during the periodof extended operationof nuclearpower plants, the regulatoryprocess is adequate to ensure that the licensing bases of all currently operatingplantsprovides and maintains an acceptable level of safety so that operationwill not be inimical to public health and safety or common defense and security."
The 1995 SOC also states:
"An applicantfor license renewal should rely on the plant's CLB, actualplant-specific experience, industry-wide operatingexperience, as appropriate,and existing engineering evaluations to determine those nonsafety-relatedsystems, structures, and components that are the initialfocus of the license renewal review. Considerationof hypotheticalfailures that could resultfrom system interdependenciesthat are not part of the CLB and that have not been previously experiencedis not required.
Therefore, when considering aging management, only the CLB need be considered.
Hypothetical failures associated with system interdependencies are not required to be considered in demonstrating adequate aging management. Therefore, the escalation effects were not directly considered in the FMECA process, nor were they required to be considered.
- 2. Inservice inspection and testing requirements of the ASME Boiler and Pressure Vessel Code (Section XI) and other operating experience (OE) related requirements, when combined with existing regulations, have been adequate to demonstrate continued safe operation and component integrity through 40 years of operation with existing programs.
- 3. Components not subject to significant aging-related degradation will continue to be managed by the existing programs that are in place (e.g.Section XI and other OE-related requirements), as appropriate. Simply stated, when MRP-227 concludes "No Additional Measures" are needed, it means that no new actions are needed for that component for the renewal period.
- 4. The Aging Management Review (AMR) topical reports prepared for B&W, CE and Westinghouse plants during the license renewal process were a basis for the work performed for MRP-227 (BAW-2248A, WCAP-14577-RI-A and CE NPSD-1216).
- 5. The supporting documents for the Issue Management Tables (MRP-205) were another basis for this work. These tables identified all safety significant issues for all PWR primary loop and internals components.
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MRP-227 Roadmap October 29, 2010
- 6. The level of analysis and evaluation detail is consistent with the guidance for Systems Structures and Components (SSC) covered in the license renewal Standard Review Plan (NUREG-1800) and in the GALL (NUREG-180 1).
- 7. Consistent with the License Renewal Rule, the current design bases are considered adequate.
In the extended operating period, for passive long-lived components, components are screened to determine if they are subject to degradation associated with aging.
- 8. Components were designed, manufactured, installed and inspected to accepted regulatory standards. In light of the positive operating experience, there is additional validation that the manufacturing and construction processes were adequate.
- 9. MRP-227 is a living document, which will be periodically updated to reflect both positive and potentially negative information from inspection results obtained by a series of plants entering the period of extended operation.
1.0 Step 1. Identify PWR internals components, materials, and environments The first step of the process was to identify the PWR internals components and items within the scope of the program on a generic basis. The starting point for the listing of reactor internals components was the IMTs published in MRP-156 and MRP-157 and other existing reports that provided information beneficial to screening. This initial list was augmented to provide additional clarification for plant-to-plant variations in design and materials.
1.1 B&W AREVA began with a review of BAW-2248A for the seven B&W-design operating units.
BAW-2248A is a B&WOG topical report that contains a technical evaluation of aging effects related to B&W PWR internals component items. It was provided to the NRC staff to demonstrate that the effects of aging during the period of extended operation for B&W PWR internals can be adequately managed. The evaluation applies to the following units:
" Arkansas Nuclear One, Unit 1 (ANO-1)
" Oconee Nuclear Station, Units 1, 2, and 3 (ONS-1, -2, -3)
- Three Mile Island, Unit 1 (TMI-1)
The staff provided a review of the topical report (BAW-2248) against the requirements in 10CFR54 and issued a Safety Evaluation Report (SER) in 1999, which resulted in issuance of BAW-2248A in March 2000. Since that time, the B&WOG has disbanded and EPRI, through the MRP, has continued the investigation on potential aging effects and establishment of monitoring and inspection programs for PWR internals component items. (Note: This was contained in BAW-2248A as applicant action item 4.) This The MRP work expanded the effort on a generic basis for all seven operating B&W-design units. Therefore, the MRP work includes not only the five units above, but it now includes the following additional units:
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MRP-227 Roadmap October 29, 2010
- Crystal River, Unit 3 (CR-3)
- Davis-Besse, Unit 1 (DB-1)
As part of the MRP effort to identify the PWR internals components and items for all of the B&W design units, MRP-157 was used as the starting point and a review of original B&W design drawings was also performed. The MRP-157 report (Table 4-14) contains the listing of B&W PWR internals components and items, which was developed from the original B&WOG report (BAW-2248A) and augmented through personal knowledge and additional record searching for the remaining units not included in the B&WOG report. This effort encompasses each of the components and items in BAW-2248A and MRP-157, and identified a few more items than contained in BAW-2248A and MRP-157. In addition, the MRP effort reviewed and evaluated weld locations associated with all identified internals components. These Therefore,are included in MRP- 189, particularly the weld locations ( MRP- 189 Rev. 1 contains the complete listing of components and items that was used in this step to be used in development of the MRP-227 I&E guidelines).
1.2 CE & W The complete list of 120 Westinghouse reactor internals components considered in the development of the MRP-227 recommendations is provided in MRP-191 Table 4-4. The NRC has previously accepted the list of 24 structures and components provided in WCAP-14577-RI-A as an acceptable basis for the scope of an aging management review of Westinghouse reactor internals. The list of components developed under the MRP efforts encompasses the same scope as the previous aging management review, but includesadds additional detail and specificity to aid in the aging assessment.
The CE reactor internal component list was also based on the IMT presented in MRP-156. The complete list of 79 CE internals components considered in the development of the MRP-227 recommendations is provided in MRP-191 Table 4-5.
2.0 Step 2. Identify degradation screening criteria The second step of the process was to develop and apply screening criteria to identify those PWR internals component items for which the effects of age-related degradation on functionality during the license renewal term may be significant. The screening criteria definition agreed upon by the industry expert panel for the MRP is as follows:
- Screening Value - the level of susceptibility when an aging effect may be significant with respect to continued functionality or safety The screening value was chosen to be sufficiently conservative such that potential component items could be selected for further evaluation of the effects of aging degradation on functionality.
Eight degradation mechanisms are currently considered relevant when assessing material aging in reactor internals (see Section 1.4 of MRP- 175). Those degradation mechanisms are:
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MRP-227 Roadmap October 29, 2010 Stress Corrosion Cracking (SCC),
Irradiation Assisted Stress Corrosion Cracking (IASCC),
Wear, Fatigue, Thermal Embrittlement, Irradiation Embrittlement, Void Swelling, and Irradiation Induced Stress Relaxation/Creep.
Development and justification of the screening criteria required knowledge of the specific aging mechanisms and their effects, some engineering judgment, extensive test data, and the use of empirical extrapolation where test data were lacking. The screening criteria used to identify components potentially susceptible to these eight mechanisms and the basis for the screening values is described in detail in MRP-175.
3.0 Step 3. Characterize components and screen for degradation (A, non-A)
The third step in the process is to evaluate the components identified in Step I against the screening criteria developed in Step 2 and documented in MRP-175.
3.1 B&W Tables 3-2 and 3-3 in Section 3 of MRP-189 Rev. 1 contain the results of the initial screening efforts. It should be noted that thermal stress relaxation of austenitic stainless steel bolting was removed as an aging degradation mechanism for the screening process in MRP- 189 Rev. 1 as a result of industry discussions and the justification provided in Appendix B of MRP- 191. Wear and fatigue that may be related to thermal stress relaxation were likewise removed from consideration for such bolting.
Because of the lack of specific ASME design rules for core support structures at the time of design and construction,Section III of the ASME Code was used as a guideline for the design criteria for the PWR internals in operating B&W units. As noted in BAW-2248A (see cChapter 2 of the report), the qualification of the internals was accomplished by both analytical and test methods. Thus, values of calculated stress, fatigue usage factors, etc. for many of the PWR internals components and items are not available nor were they required at the time of design.
Through the expert panel approach, estimates of potential stress, fatigue usage, etc. were made and used for many of the component items during the screening process. Specific stress inputs were only used for screening a limited number of components (MRP-189 Rev. 1 Table 3-2) from existing stress calculations at the time of screening. The loading sources considered in the stress values are discussed in Response to RAI 4-1. For a few items, a review of available records (stress calculation reports, unit-specific analyses, etc.) was performed that was able to identify the various values provided in MRP-189 Rev. 1 Table 3-2 (see Sections 3.2 and 3.3 of MRP-189 Rev. 1).
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MRP-227 Roadmap October 29, 2010 Table 1 provides the screening parameters for the representative components 2 from each category that are selected for this roadmap discussion, along with the screening results for each of the aging mechanisms and the initial screening category assigned to each component.
Of the B&W RV internals components that were screened-in as "Non-A" in Step 3, 47 components were placed in the "No additional measures" category by Steps 4, 5, 6, 7, and 8. The B&W RV internals was not designed to the ASME Section III, Subsection NG, and no core support structure or internals structure designations were specified by B&W during the design.
However, the safety significance of the RV internals components was evaluated for the MRP-157 report and for MRP-190. The safety significance of these 47 components is summarized below.
FMECA Safety Consequence:
Of the 47 components,
- Two have a FMECA safety consequence metric of"2".
- 44 have a FMECA safety consequence of metric of "I"
- Safety consequence for one component (the upper grid assembly rib section) was not evaluated by FMECA as the CUF value used for screening-in fatigue was from the 205-FA design and was considered incorrect for the B&W 177-FA design by the FMECA panel. [Note: This component has an IMT safety consequence of"G" in MRP-157. See below.]
MRP-190 (FMECA) safety consequences metrics:
- 1. Safe: no or minor hazard condition exists
- 2. Marginal: safe shutdown is possible (though with reduced margins to adequately cool the core and/or successfully insert the control rods); localized fuel assembly damage
- 3. Severe: safe shutdown is possible (though with very reduced margins to adequately cool the core and/or successfully insert the control rods); core damage (multiple damaged fuel assemblies)
- 4. Critical: safe shutdown is not possible (margins to adequately cool the core and/or successfully insert control rods are totally eroded); extensive core damage IMT Safety Consequence Of the 47 components, e Five have IMT safety consequence metrics of "G and F"
- 23 have an IMT safety consequence metric of"G"
- 19 have no IMT safety consequence MRP-157 (IMT) consequences of failure metrics:
2 Note: Each of the steps contains information and/or tables that refer to specific tables or sections in the reference documents for the B&W design. A complete listing of components for the B&W design can be found in these tables or sections in the reference documents from which these representative components have been selected for the discussions in this roadmap.
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MRP-227 Roadmap October 29, 2010 (A) Precludes the ability to reach safe shutdown (B) Causes a design basis accident (C) Causes significant onsite and/or offsite exposure (D) Jeopardizes personnel safety (E) Breaches reactor coolant pressure boundary (F) Breaches fuel cladding (G) Causes a significant economic impact Therefore, in summary, of the 47 components placed in the "No additional measures" category, none are considered to have any safety related consequence in the event of loss of function from any age-related degradation mechanism.
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MRP-227 Roadmap October 29, 2010 Table 1 Screening Parameters, Screening Results for Each Aging Mechanism and Initial Screening Category forSelected B&W RI Components (extracted from Tables 3-2 and 3-3 of MRP-189 Rev. 1)
__ 1 ER i _i 0 0 -
v30 <0.1 CRGT C0ntrol Assume Assume Rod Guide 605 < 5E18 <0.01 <30 No No <0.1 A A A Not A A A A A Not A Sectors CSS Vent Valve TopandBottom 605 < 5E18 <0.01 9.8 No No Ac Assume A A A A A Not A A A >
A A Not A Retaining Rings <0.1 C605 < 5E18 <0.01 No No sume A A A A A NotA A A Not A Disc Tubes <30 <0.1 CSS Vent Valve AssAssume
< 5E18 <0.01 9.8 No No Asum A A A A A Not A A A Not A Disc Pin or HingeShaft 605 <0.1 Corine rre 620 5.0E+21 7.5 1.0 No Yes 0.21 Not A A A A Not A A Not A A Not A Baffle Plates 646 6.4E+22 96 <20 No No <0.1 A NotA A A A A Not A NotA NotA Former Plates 647 5.OE+22 75 <20 No No <0.1 A Not A A A A A Not A Not A Not A Core Barrel-to- Assume Assume Former Plate 633 1.5E+22 22.5 <30 No No <0.1 A A A A A A Not A Not A Not A Dowels < N o A A A A A A N o NIIA BIO
MRP-227 Roadmap October 29, 2010
-ponC -r - ---- - -
Sco >I A l 0a E. .2 a~ 0..a FLowe Disridbutorm Assume t560 5.E+18 0.008 82 No No Not A A A A A A A A Not0 (FD) Bolts <0.1 Bll
MRP-227 Roadmap October 29, 2010 3.2 CE & WW&CE Design representative values of the key screening parameters for each reactor internals component in the CE and W fleet were required to complete the screening evaluation. A detailed analysis to generate specific values for either the CE or W design was not performed as part of the MRP project. Representative values, meant to be limiting values for the fleet were determined from existing design basis analysis wherever possible. When hard numbers were not available, teams of reactor internals engineering experts were assembled to provide conservative estimates or to determine if there was any potential for the component to exceed the screening criteria. In all cases, the component condition was conservatively estimated. The process used by Westinghouse to determine these values is described in the following subsections. From this information, the team assessed the data for each component and reached consensus on representative values to use in the screening. This process was published in Section 4 of MRP- 191. The component conditions as determined by the teams of experts are provided in MRP-191 Table A-1.
The screening process simply compared the estimated component conditions to the MRP-175 screening levels. Based on this screening process, 48 of the 120 Westinghouse components and 8 of the 79 CE components were identified with no potential aging considering each of the degradation mechanisms. The components with no screened-in aging degradation mechanisms are identified in MRP-191 Table 6-5 and Table 6-6 for W and CE components respectively.
These components, which are listed in Table 2 and Table 3 of this roadmap document were tentatively placed in Category A, pending review by the FMECA panel in the following step of the assessment process.
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MRP-227 Roadmap October 29, 2010 Table 2 Westinghouse Components with No Screened-In Degradation Mechanisms (Data extracted from MRP-191 Table 6-5)
Assembly Sub-Assembly Component Material IMTof Conseq.
Failure Upper Internals Control Rod Guide Tube Anti-rotation 304SS G Assembly Assemblies and Flow studs and Downcomers nuts Bolts 316 SS NONE Flexureless 304 SS G inserts Housing 304 SS G plates Inserts 304SS N/A Lock bars 304 SS NONE Support pin 304 SS NONE cover plates Support pin 316SS NONE cover plate cap screws Support pin 304SS NONE cover plate locking caps and tie straps Support pin X-750 NONE nuts Support pin 316SS NONE nuts Water flow 304SS N/A I
slot liiaments 4. 4.
Upper Instrumentation Bolting 316SS NONE Conduit and Supports Brackets, 304SS NONE clamps, terminal blocks, and conduit straps I__________
Conduit seal assembly-I tubesheets 304SS NONE B13
MRP-227 Roadmap October 29, 2010 Assembly Sub-Assembly Component Material IMT Conseq.
__ __ __ __ _ I_ I___________ __ _ _ I__ _ I__ of F allure Conduit seal 304SS NONE assembly-tubes Conduits 304 SS NONE Flange bases 304 SS NONE Locking caps 304 SS NONE Support 304SS NONE tubes Upper Plenum UHI flow 304 SS G columns Upper Support Column Adapters 304 SS G Assemblies Column 304SS G bodies Flanges 304SS G Lock keys 304 SS G Nuts 304SS G Upper Support Plate Bolts 316 SS NONE Assembly Upper Support Plate Flange 304 SS N/A Assembly Lock keys 316SS NONE Ribs 304SS G Upper 304SS G SUDDort Dlate Lower Internals Bottom Mounted BMI column 304L SS NONE Assembly Instrumentation (BMI) lock caps Column Assemblies Diffuser Plate Diffuser plate 304SS NONE Head Cooling Spray Head cooling 304 SS NONE Nozzles spray nozzles Lower Support Column Lower 304SS G Assemblies support column nuts Lower 304SS G support column sleeves B14
MRP-227 Roadmap October 29, 2010 Assembly r
Sub-Assembly I Component I
II Material jIofIMT ofie Conseq.
Failure Lower Support Casting or Lower 304 SS A, G Forging support forging Radial Support Keys Radial 304 SS G support key lock keys Secondary Core Support SCS bolts 316 SS NONE (SCS) Assembly SCS energy 304SS NONE absorber SCS guide 304SS NONE post SCS housing 304SS NONE SCS lock 304SS NONE keys Interfacing Interfacing Components Clevis insert Alloy 600 G Components lock keys Clevis insert 316 SS G lock keys Head and 316SS NONE vessel alignment pin bolts Head and 304L SS NONE vessel alignment pin lock cups Head and 304SS NONE vessel alignment pins IMT Consequence of Failure - G: Causes significant economic impact A: Precludes a safe shutdown B15
MRP-227 Roadmap October 29, 2010 Table 3 CE Components with No Screened-In Degradation Mechanisms (Data extracted from MRP-191 Table 6-6)
Assembly/ Component Material IMT Conseq. Of Sub-Assembly Failure Upper Internals Assembly Control rod 316 SS N/A shroud-bolts GSSS studs 316SS N/A GSSS spherical UNS N/A washer sets S21800 Flange block A286 SS N/A shear pins Control Element Assembly Shim bolts 316 SS N/A (CEA)-Shroud Assemblies Core Support Barrel Core barrel 316SS N/A Assembly snubber lug bolts Core barrel A286 SS N/A snubber lug bolts Alignment key 304 SS NONE dowel pins I I 4.0 Step 4. Failure Modes, Effects and Criticality Analysis (FMECA)
The fourth step in the process was to perform a Failure Modes, Effects and Criticality Analysis (FMECA). While the specific approach used by AREVA for the B&W units varied with that used by Westinghouse for the CE and W units, the principles employed were similar and produced conservative results. It is important to note that items that were screened as "A" in step 3 above (i.e. - no augmented aging management needed) were re-assessed and this confirmed that the original screening was valid. A summary of each approach is described below. The details of the approaches are described in MRP-190 for the B&W units and MRP-191 for the CE and W units.
4.1 B&W The objective of the FMECA, described in detail in MRP-190, is to provide a systematic, qualitative review of the B&W-designed PWR internals to identify combinations of internals component items and age-related degradation mechanisms that potentially result in degradation leading to significant risk. The FMECA is used to examine the susceptibility, and safety and economic consequences of identified internals component item/age-related degradation mechanism combinations. For those items screened as "A" (in Step 3 above), the FMECA team provided verification that there were "no credible degradation mechanisms" associated with these items.
The FMECA approach uses inductive reasoning to ensure that the potential failure of each component item is analyzed to determine the results or effects thereof on the system and to classify each potential failure mode according to its severity.
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MRP-227 Roadmap October 29, 2010 Each failure mode (i.e., aging effect) was judged on its importance to risk, based on the susceptibility (likelihood of the degradation mechanism) and severity of consequences. For this FMECA, consequences were examined from two perspectives: safety and economic. The FMECA report developed a risk matrix to correlate the consequence severity of a particular age-related degradation mechanism with the susceptibility of that particular mechanism occurring.
Different risk bands were used within the matrix to categorize the level of risk of a particular component item/degradation mechanism pair, and provide guidance on the strategies that should be developed to reduce the corresponding risk and a basis for ranking and categorization. This "risk metric" is not to be confused with risk in a probabilistic risk assessment, for which the metrics of core damage frequency and large early release frequency are typically used.
The criticality metrics of a particular component item failure are evaluated qualitatively by assessing both the susceptibility to an age-related degradation mechanism and subsequent effect, and the severity of the consequences (see Figure 4-1 of MRP-189 Rev. 1). For this FMECA, two types of consequences are considered: safety and economic. When considered together, the criticality metrics represent the risk due to the failure of a particular component item. The criticality metrics are fully described in both MRP-189 Rev. 1 and MRP-190 (also see Step 5 below).
4.2 W and CE & W A FMECA was conducted to evaluate the likelihood and severity of damage associated with the identified degradation mechanism. The Westinghouse FMECA team was asked to review and concur with information for all 120 identified reactor internals components. Similarly the CE FMECA team was asked to review and concur with information for all 79 identified components.
While the screening process evaluated only the potential susceptibility of the component to the eight identified aging degradation mechanisms, the FMECA panel considered both the susceptibility and the potential safety consequences of degradation.
The Westinghouse FMECA process and results are described in MRP-191 and summarized in the following sub-sections. The discussion record of the FMECA expert panel meetings is considered Westinghouse proprietary, but can be made available for NRC review.
4.2.1 FMECA Review of Components with No Identified Degradation Mechanism The evaluation team was charged to review the results for the 48 Westinghouse and 8 CE components with no identified degradation mechanisms. The panel was asked to concur with these screening results or to recommend reinstating the component for further evaluation. The panel concluded that the application of the screening process was extremely conservative and there was no need to reinstate additional components for further evaluation.
The FMECA panel was also asked to review the 48 Westinghouse and 8 CE components with no identified degradation mechanism and determine that there was "No need to assess damage probability". As part of this process, the FMECA panel reviewed the consequences of failure conclusions from the MRP Issue Management Table (IMT) as described in MRP-156. These B17
MRP-227 Roadmap October 29, 2010 IMT consequences are noted in Table 2 and Table 3. The IMT treats consideration of the probability of degradation and the consequences of failure as completely independent phenomena.
4.2.2 Westinghouse NSSS Of the 48 Westinghouse components considered, the only component with potential safety-related consequence of failure identified in the IMT was the lower core support forging. (The cast stainless steel version of this component was screened-in due to thermal embrittlement concerns.) Loss of support due to catastrophic failure of this structure could preclude safe shut down of the reactor. THowever, the FMECA panel could not identify any potential cause or mode of catastrophic failure that would require aging management of this large forging. The inspection required for non-age related degradation of this component is specified in ASME Section XI. Therefore the lower support forging was not reinstated for additional evaluation.
There were no potential safety-related concerns ("Precludes safe shutdown" or "Breaches fuel cladding") identified in the IMT for the remaining 47 Westinghouse components. Potential economic consequences of failure were noted in 17 of the remaining components. The FMECA panel concurred with this conclusion and concluded that there was no need to include these components in the aging management strategy because there are no safety implications to failure and the economic consequences of unanticipated failure are not severe enough to justify the expenditure of resources to manage such low probabilities of occurrence.
4.2.3 CE It is difficult to produce a one-to-one correspondence between the CE reactor internals component list in MRP-156 and the list in MRP-227 because additional detail has been added to facilitate the evaluations in MRP-227. However a thorough review showed there are no potential safety related concerns identified for the CE reactor internals components listed in Table 3.
4.2.4 FMECA Review of W and CE Components with One or More Identified Degradation Mechanisms The FMECA process was employed to assess the likelihood of failure and the likelihood of damage in the remaining 72 Westinghouse and 71 CE components. The FMECA process is described in detail in Section 6 of MRP- 191. Additionally it is noted that the members of the FMECA were consistent for all discussions for a given NSSS design.
The FMECA process was conducted on a component-by-component basis and the FMECA categorization was based on the cumulative effects of all eight degradation mechanisms in each component. Potential susceptibility to multiple degradation modes was one of the factors considered by the FMECA panel.
The FMECA panel findings for the Westinghouse reactor internals are provided in Table 6-5 and CE reactor internals in Table 6-6 of MRP-191. The FMECA panel discussions included evaluation of design and analysis data and are therefore considered to be Westinghouse B18
MRP-227 Roadmap October 29, 2010 proprietary. The FMECA panel findings are also included on the lists of potentially susceptible components in each degradation mechanism series. It should be noted that the FMECA ranking is conservatively based on the cumulative effect of all degradation modes and may not be an indicator of a specific single degradation mode.
5.0 Step 5. Severity Categorization (A, B, C)
The fifth step of the process was to use the results of the FMECA to categorize each of the component items into the categories A, B, and C. As was the case with the FMECA, the severity categorization processes used by AREVA and Westinghouse varied in their specific steps but accomplished the intended goal. All of the reactor internals were placed into one of three categories based on the significance and severity of the potential degradation. A summary of each approach is described below. The details of the approaches and results are described in MRP- 189 Rev. 1 and MRP- 190 for the B&W units and MRP- 191 for the CE and W units.
The FMECA panels for both AREVA and Westinghouse agreed that the "A" (or Category A) events are deemed so improbable (very, very low likelihood of occurrence) that even if a Level B, C, or D event were to occur, the risk impact would not be significant.
5.1 B&W Categorization of PWR internals was subsequently performed, based on the screening criteria and the likelihood and severity of safety consequences, into categories that range from those components for which these issues are insignificant (Category A) to those components that are potentially moderately significant (Category B) to those components that are potentially significantly affected (Category C). This is detailed in MRP-189 Rev. 1 and MRP-190.
The criticality metrics used in the AREVA FMECA are as follows:
5.1.1 Susceptibility The susceptibility metric is a qualitative assessment of the likelihood (expressed as a probability or frequency) that an age-related degradation mechanism might occur, given the existing environmental conditions (e.g., temperature, pressure, fluence, etc.), material properties (type of metal, stress-strain), etc. occurring over the life of a nuclear power unit (up to 60 calendar years, considering license renewal). The susceptibility is unrelated to the consequences, e.g., the component item failure or loss of function. The susceptibility qualitative metric was determined as a result of the expert panel meeting. This criticality metric uses an A, B, C, D scale (increasing frequency).
A - Improbable: not likely to occur (Category A from the initial screening performed in Chapter 3 is synonymous with this susceptibility metric; the Category A results were reviewed by the FMECA expert panel)
B - Unexpected: not very likely to occur, though possible; conditions are such that the age-related degradation mechanism is not expected to occur very often B19
MRP-227 Roadmap October 29, 2010 C - Infrequent: likely to occur, conditions are such that the age-related degradation mechanism is expected to occur occasionally D - Anticipated: very likely to occur; conditions are such that the age-related degradation mechanism is expected to occur B/I - The susceptibility is sometimes modified with an "I" to indicate an improbable occurrence over the 60-year time period being considered. For example: B/I indicates an unexpected, but possible, degradation mechanism whose initiation results in a certain state that is not credible (or improbable), e.g., SCC crack leading to a 360 degree weld crack. To carefully distinguish between the different types of likelihood, it is possible (B) to have SCC cracking around a weld, but improbable (I) that such as crack would grow around the weld to the critical crack size needed to fail the weld.
Component item/degradation mechanism pairs identified as improbable are not explicitly evaluated for consequences. However, there are a number of combinations that while identified as improbable will either result in severe consequences, affect the ability to cope with a LOCA, or will require the successful "operation" of the guide lugs.
Accordingly, while not classified into a specific risk band, these items, as noted in the footnotes of Table 4-1 (MRP- 189 Rev. 1) should never be removed from the current ASME inspection requirements (VT-3).
5.1.2 Severity of Consequences Severity classifications are assigned to provide a qualitative measure of the potential consequence resulting from a component item failure. For those component item/age-related degradation mechanism pairs for which the susceptibility metric was assigned an "A," i.e.,
"Category A," there was no subsequent evaluation of the consequence due to the very low (i.e.,
improbable) event frequency. For the PWR internals FMECA, two aspects of consequences are considered: safety and economic. Thus, there are two columns in the FMECA for which qualitative metrics are assigned. The two sets of severity of consequence qualitative metrics were determined as a result of the expert panel meeting. These criticality metrics use a 1, 2, 3, 4 scale (increasing severity).
For severity of consequences (safety), the qualitative metric has been defined as:
- 1. Safe: no or minor hazard condition exists
- 2. Marginal: safe shutdown is possible (though with reduced margins to adequately cool the core and/or successfully insert the control rods); localized fuel assembly damage
- 3. Severe: safe shutdown is possible (though with very reduced margins to adequately cool the core and/or successfully insert the control rods); core damage (multiple damaged fuel assemblies)
- 4. Critical: safe shutdown is not possible (margins to adequately cool the core and/or successfully insert control rods are totally eroded); extensive core damage B20
MRP-227 Roadmap October 29, 2010 The safety consequence metric assigned will be the highest value, i.e., bounding consequence, for normal operation or design basis event (transient, LOCA, seismic) when the failure mode is not detectable. Typically, the safety consequences were estimated to be the same for normal operation and a design basis event (when the failure mode is not detectable). Note that there were no severity of consequences (safety) identified with a metric of 4.
For severity of consequences (economic), the qualitative metric has been defined as:
- 1. No or trivial cost
- 2. Cost that can be generally handled within the existing unit budget and resources (order of millions of dollars)
- 3. Cost that exceeds the normal unit budget and resources (order of tens of million dollars)
- 4. Cost that potentially affects the utility's overall financial health (order of hundreds of million dollars)
Note that the economic consequences assume that the failure mode is discovered through some means, e.g., unit inspection, notification of discovery at another unit site, etc. This is also conservative when assessing the risk. Note that the severity of consequences (economic) metric was not used in assignment of the preliminary Category A. B. and C items.
Based upon the FMECA results, the PWR internals that were potentially the most affected were placed into Category C, while the components that are potentially only moderately affected were placed into Category B. In addition, the FMECA process determined that some components not initially Category A were sufficiently unaffected by consequences to be subsequently placed into Category A.
The risk matrix in MRP- 189 Rev. 1 (Figure 4-1) does not include a column for the susceptibility metric value of "A" because, as noted in MRP- 190 (Section 3.2), the "A" (or Category A) events are deemed so improbable (very, very low likelihood of occurrence) that the safety severity of consequence metric was not evaluated, implying that even if there was an adverse consequence, the risk impact would be insignificant. However, to clarify how component items were categorized, the Figure 1 below provides a correlation to the risk matrix (Figure 4-1 of MRP-189 Rev. 1) and also includes a column for Category A items:
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MRP-227 Roadmap October 29, 2010 Increasing Susceptibility from A to D A B C D
.7 I A A B B 2
U 3
4 Figure 1: Consequence vs. Susceptibility for Ranking *Note: There are no component items in the B&W-design internal with an assigned safety consequence metric equal to 4; therefore, the last row of this figure is not applicable to the MRP effort.
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MRP-227 Roadmap October 29, 2010 The initial Category A, B, and C results for selected B&W components are provided in Table 4.
Table 4 Initial Category A, B and C Results for Selected B&W Components (Extracted from Tables 4-1 and 4-2, MRP-189 Rev. 1)
Component Safety Economic A,B,C(MRP189)
Band Band Rev. I CRGT Spacer Castings I III B CRGT Control Rod Guide Tubes II Ill B CRGT Control Rod Guide Sectors II Ill B CSS Vent Valve Top and Bottom Retaining I III B Rings CSS Vent Valve Disc I Ill B CSS Vent Valve Disc Shaft or Hinge Pin I III B I II Core Barrel Cylinder B I Ill Baffle Plates II Ill C II II Ill Ill Former Plates II Ill C Ill Ill II II Core Barrel-to-Former Plate Dowels B Lower Grid Support Post Cap Screw I IB II Ill Flow Distributor (FD) Bolts C IV V B23
MRP-227 Roadmap October 29, 2010 Component Degradation Safety Economic A, B, C Mechanism Band Band (MRP189 Rev. 1)
CRGT Spacer Castings TE I III B CRGT Control Rod Guide Tubes Wear II III B CRGT Control Rod Guide Sectors Wear II i1l B CSS Vent Valve Top and Bottom Retaining TE III B Rings CSS Vent Valve Disc TE I III B CSS Vent Valve Disc Shaft or Hinge Pin TE I III B SCC I II Core Barrel Cylinder B IE I III IASCC III III Baffle Plates IE II III C VS II II IASCC III III Former Plates IE II Ill C VS III III IE II II Core Barrel-to-Former Plate Dowels B VS I I Fatigue I I Lower Grid Support Post Cap Screw IE I I B Wear I I Flow Distributor (FD) Bolts SCC IV V C It is also interesting to compare the IMT (MRP-157) results to the FMECA results. For each component item that constitutes part of the PWR internals, consequences of failure evaluations were performed in the IMT considering each of the applicable degradation mechanisms (without regard for existing mitigation strategies). This includes following the logical path from component failure to safe shutdown. The consequences evaluation is considered to be reality-based not design-based, so these evaluations are not related to the design bases of the B&W units.
Scenarios that rely on a sequence of low probability events reach to get a failure may be documented as such and the failure evaluation terminated. Systems that must operate correctly to satisfy the defined failure sequence are identified. It is also noted that the evaluations do not consider electrical system failures due to component item degradation (e.g., RCS instrumentation). The expert panel participants are listed in the IMT and represent a broad scope B24
MRP-227 Roadmap October 29, 2010 of expertise in the design and operation of the B&W units. In the IMT, the general approach used in the consequences of failure evaluations was as follows:
" For each component item, consequences of failure evaluations were performed considering all of the applicable degradation mechanisms identified by the MDM. The evaluations assume that the unit is initially at full power steady-state conditions.
Assuming failure while the unit is at other Level A service conditions impacts the availability of various systems, the unit conditions, and therefore the sequence of events to safe shutdown.
" Level A conditions other than full power, as well as Level B, C, and D conditions are considered coincident with component degradation that does not require unit shutdown during normal operations. These coincident conditions are not rigorously treated, but are discussed from the perspective of their potential contribution to adverse consequences.
[For clarification, this means that service level events (Levels B, C, and D) were not superimposed along with gross failure from aging degradation of the component or item under consideration. This is a similar approach to that used in Chapter 15 of the FSAR.]
" The evaluations consider the functions that the component item supports and the impact that the degradation might have on the ability of the reactor vessel internals to continue performing thoseat functions. For instance, through-wall cracking, significant wear (at a location of contact or close tolerance), or embrittlement, could compromise the structural integrity of a component item, so each is considered in the evaluations. If different degradation mechanisms lead to different results, then each is treated individually.
Multiple degradation sites are not considered because common mode and/or cascading failures are not in the scope of the project. Loose parts were generically evaluated as well.
The following consequences of failure were evaluated:
A. Precludes the ability to reach safe shutdown B. Causes a design basis accident C. Causes significant onsite and/or offsite exposure D. Jeopardizes personnel safety E. Breaches reactor coolant pressure boundary F. Breaches fuel cladding G. Causes a significant economic impact As shown in Table 4-14 of the IMT (MRP-157), none of the safety-related consequences of failure (items A-E) were determined to be applicable (similar to the FMECA results) and only consequences of failure items F and G were determined to be applicable to the B&W PWR internals. However, it should be noted that there were differences between the consequence evaluations performed in the IMT and the FMECA. An explanation of the differences is provided in Appendix B of MRP- 190.
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MRP-227 Roadmap October 29, 2010 5.2 CE & W All of the reactor internals were placed into one of three categories based on the significance and severity of the potential degradation. These three categories were:
Category A: Component items for which aging degradation significance is minimal orand aging effects are below the screening criteria.
Category B: Component items above screening levels but are not "lead" component items and aging degradation significance is moderate.
Category C: "Lead" component items for which aging degradation significance is high or moderate and aging effects are above screening levels.
5.2.1 Components Placed in Category A Based on FMECA After review and confirmation by the FMECA panel, all of the components that were not identified in the screening process for potential susceptibility to any of the eight degradation mechanisms were retained as originally placed in Category A.
The FMECA panel also observed that, due to the conservative nature of the screening process, many components that had been identified for potential degradation were known to not be susceptible to degradation. The most obvious example of the conservative nature of the process was that the surveillance capsule components were identified for irradiation embrittlement because the screening process attributed the peak core barrel fluence to all of the potential attachments. However the FMECA panel observed that the surveillance capsules contain dosimetry packages and the fluences were known to be well below the threshold for irradiation embrittlement.
To more accurately reflect the degradation potential for the components and account for the overly conservative nature of the screening process, the FMECA panel recommended that components with low failure likelihood and either low or medium damage likelihood, especially where the potential for any damage was considered to be readily detectable and manageable in attaining a safe operational state, be moved to Category A. Components with low failure likelihood and high damage likelihood were not considered as candidates to be moved to Category A under any conditions. These criteria are illustrated in Figure 2. By definition, all components with potential safety concerns were classified as high damage likelihood. Therefore, no components with identified safety concerns were affected by this re-classification.
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MRP-227 Roadmap October 29, 2010 I
Failure Consequence (Damage Likelihood)
Likelihood Low Medium High High 2 3 3 Medium 1 2 3 Low 1 1 2
-, ~Cate gory A ___
None I 0 0 0 Figure 2 FMECA Criteria for Aging Significance Table The 41 Westinghouse components with one or more identified degradation mechanisms that were moved to Category A based on the FMECA results are listed in Table 5. The 48 CE components moved to Category A based on the FMECA are listed in Table 6. The FMECA panel identified 27 Westinghouse and 27 CE components with low failure probability and low damage consequence. There were an additional 14 Westinghouse and 21 CE components with low failure probability and medium damage consequence. Although the FMECA panel identified a potential economic consequence of failure in the components with medium likelihood of damage, the low failure probability resulted in minimal risk to plant operation.
Therefore these 14 Westinghouse and 21 CE components were also placed in Category A.
Application of the FMECA process to the Lower Core Plate Fuel Alignment Pin Bolts is provided in Example 1.
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MRP-227 Roadmap October 29, 2010 Example 1: Lower Core Plate Fuel Alignment Pin Bolts Placed In Category A Based on FMECA Original screening results: MRP-191 Table 5-1
- IASCC, Wear, Fatigue, Irradiation Embrittlement, Void Swelling, Irradiation Induced Stress Relaxation/Creep Functional
Description:
MRP-191 Section C.2.1
- The LCP is bolted at the periphery to a ring welded to the ID of the core barrel. The span of the plate is supported by lower support columns that are attached at their lower end to the lower support plate. At the center, a removable plate is provided for access to the vessel lower head region.
FMECA
Conclusion:
MRP-191 Table 6-5
- Low Failure Probability, Low Consequence
- Screening process overestimated fluence because it assumed components attached to LCP saw same peak fluence. These bolts are located on periphery.
- No history of failures
- Bolts are redundant fasteners.
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MRP-227 Roadmap October 29, 2010 Table 5. Westinghouse Components Moved to Category A Based on FMECA Process (Data extracted from MRP-191 Table 6-5)
IMT Screened-in Likelihood Likelihood Assembly Sub-Assembly Component Material Conseq. of Degradation of Failure of Damage Failure Mechanisms L, M, H L, M, H Upper Control Rod Enclosure pins 304 SS NONE SCC, Wear L M Internals Guide Tube Assembly Assemblies and Flow Downcomers Upper guide tube 304 SS NONE SCC, Wear L M enclosures Flanges-intermediate 304 SS G SCC, Fatigue L M Flanges-intermediate CF8 G SCC, Fatigue, TE L M Flanges-lower 304 SS G SCC, Fatigue L M Guide tube support 316 SS NONE Wear, Fatigue, ISR L M pins Mixing Devices Mixing devices CF8 NONE SCC, TE, ISR L L Upper Core Plate Fuel alignment pins 316 SS NONE Wear L L and Fuel Alignment Pins Upper core plate 304 SS A, G Wear, Fatigue L M Upper Plenum UHI flow column CF8 G TE, IE L L bases Upper Support Bolts 316 SS G Wear, Fatigue, ISR L M Column Assemblies Column bases CF8 G SCC, TE, IE L M Extension tubes 304 SS G SCC L M Upper Support Deep beam ribs 304 SS G SCC L M Plate Assembly
_ Deep beam stiffeners 304SS G SCC L M B29
MRP-227 Roadmap October 29, 2010 IMT Screened-in Likelihood Likelihood Assembly Sub-Assembly Component Material Conseq. of Degradation of Failure of Damage Failure Mechanisms L, M, H L, M, H Inverted top hat (ITH) 304 SS N/A SCC, Fatigue L M flange Inverted top hat (ITH) 304 SS N/A SCC L M upper support plate Lower Baffle and Baffle bolting lock 304SS NONE IASCC, IE, VS L L Internals Former Assembly bars Assembly Bottom Mounted BMI column bolts 316 SS NONE Fatigue L L Instrumentation (BMI) Column Assemblies BMI column 304 SS G IASCC, IE, VS L L extension bars BMI column nuts 304 SS NONE IASCC,Wear, Fatigue, L L IE, VS, ISR Irradiation Irradiation specimen 304 SS NONE Wear, IE L L Specimen Guides guides Irradiation specimen 316 SS NONE IASCC, Wear, Fatigue, L L guide bolts IE, ISR Irradiation specimen 304L SS NONE IE L L guide lock caps Specimen plugs 304SS NONE IE L L Lower Core Plate Fuel alignment pins 316 SS NONE IASCC, Wear,IE, VS L L and Fuel Alignment Pins LCP-fuel alignment 316 SS NONE IASCC, Wear, Fatigue, L L pin bolts IE, VS, ISR LCP-fuel alignment 304L SS NONE IASCC, IE, VS L L p in loc k ca ps I I I I I _I B30
MRP-227 Roadmap October 29, 2010 IMT Screened-in Likelihood Likelihood Assembly Sub-Assembly Component Material Conseq. of Degradation of Failure of Damage I I I Failure Mechanisms L, M, H L, M, H Neutron Neutron panel bolts 316 SS NONE IASCC, Wear, Fatigue, L L Panels/Thermal IE, ISR Shield Neutron panel lock 304 SS NONE IE L L caps Thermal shield bolts 316 SS NONE IASCC, Wear, Fatigue, L L IE, ISR Thermal shield 316 SS NONE IE L L dowels Thermal shield or 304 SS G IE L L neutron panels Radial Support Radial support key 304 SS G Wear L L Keys bolts Radial Support Radial support keys 304 SS G SCC, Wear L L Keys Secondary Core SCS base plate 304 SS NONE SCC L L Support (SCS)
Assembly Interfacing Interfacing Clevis inserts Alloy 600 G Wear L L Components Components Clevis inserts 304 SS G Wear L L Clevis inserts Stellite G Wear L L Internals hold-down 304 SS G Wear L L spring Internals hold-down 403 SS G Wear, TE L L spring I IIIII IMT Consequence of Failure - G: Causes significant economic impact A: Precludes a safe shutdown B31
MRP-227 Roadmap October 29, 2010 Table 6. CE Components Moved to Category A Based on FMECA Process (Data extracted from MRP-191 Table 6-6)
Likelihood Likelihood SbAssembly/
Assembly/ Component Material IMTfFiueMcaim Conseq. Screened-in Degradation _t_
of Failure of Damage L,M,H L,M,H Upper Internals Upper guide 304 SS G SCC L M Assembly structure support plate Upper guide 304 SS G SCC, Wear L M structure support flange-upper Upper guide 304 SS G SCC L M structure support flange-lower Cylindrical skirt 304SS G SCC L M Grid plate 304SS G SCC L M Control rod 304 SS N/A SCC L M shroud-grid ring Control rod 304SS N/A SCC L M shroud-grid beams Control rod 304SS N/A SCC L M shroud-cross braces GSSS guide 304SS N/A SCC L M structure plate GSSS support 304SS N/A SCC L M cylinder Flange blocks 304 SS N/A Wear L L B32
MRP-227 Roadmap October 29, 2010 Likelihood Likelihood Assembly/ Component Material IMT Conseq. Screened-in Degradation of Failure of Damage Sub-Assembly of Failure Mechanisms L,M,H L,M,H Flange block 410 SS N/A TE L L bolts RVLMS support 304 SS N/A SCC, Wear, Fatigue L L structure tubes Fuel bundle 316 SS N/A Wear, Fatigue, ISR L L guide pins Fuel bundle 304 SS N/A Wear, Fatigue, ISR L L guide pin nuts Hold down ring 403 SS/ G Wear, TE L L F6NM Belleville Alloy 718 N/A Wear L L washer Lower Support Core support 316 SS N/A IASCC, Wear, Fatigue, IE, L L Structure plate bolts ISR Core support 304SS N/A IE L L plate dowel pins Anchor block 316 SS N/A Wear, Fatigue, IE, ISR L L bolts Anchor block 304 SS N/A IE L L dowel pins Fuel alignment 304 SS NONE IE L M pins Core support 304 SS A, G SCC, Wear L L beams Bottom plate 304SS N/A SCC L L ICI support 304SS N/A SCC L L columns B33
MRP-227 Roadmap October 29, 2010 Ase y SUkelihood Ukelihood Sub-Assembly Component Material IMTof Failureq Screened-inismsDegradation of Failure of Damage L,M,H L,M,H Control Element CEA shrouds 304 SS G SCC L M Assembly (CEA)-Shroud Assemblies CEA shrouds CPF8/CF8 G SCC, TE L M CEA shroud 304SS G SCC L M bases CEA shroud CF8 G SCC, TE L M bases CEA shroud 304 SS G SCC L M extension shaft guides Modified CEA CF8 G SCC, TE L M shroud extension shaft guides Internal/external 304 SS NONE SCC L M spanner nuts CEA shroud A286 SS NONE Wear, Fatigue, ISR L M bolts CEA shroud tie 304SS N/A SCC L M rods Snubber blocks 304SS N/A SCC L L Snubber shims XM-29 N/A Wear L L Core Support Core barrel 304, 321 G SCC, Wear L L Barrel snubber lugs or 348 SS Assembly Alignment keys A286 SS NONE Wear L L Alignment keys 304 SS NONE Wear L L Core barrel 304SS G SCC, Wear L M outlet nozzles B34
MRP-227 Roadmap October 29, 2010 Likelihood Likelihood Assembly/ Component Material IMT Conseq. Screened-in Degradation of Failure of Damage Sub-Assembly of Failure Mechanisms L,M,H L,M,H Thermal shield 304 SS G SCC L L Thermal shield 304 SS NONE Wear L L support pins Core Shroud Guide lugs 304 or NONE SCC L L Assembly 348 SS Guide lug 304, 321 NONE Wear L L inserts or 348 SS In-Core ICI guide tubes 316 SS NONE SCC, IE L L Instrumentation (ICI)
ICI nozzle 304SS G SCC L L support plate ICI thimble 304 SS G SCC, Wear L L support plate ICI thimble 304SS NONE SCC, Wear L L tubes-upper B35
MRP-227 Roadmap October 29, 2010 5.2.2 Components Placed in Categories B and C The remaining 31 Westinghouse and 23 CE "non-Category A" components were evaluated and placed in Category B or Category C based on the FMECA results and analysis using the Category definitions. Each component was assigned a FMECA aging significance grouping based on the FMECA categories as indicated in Figure 2.
Two exceptions were noted to the components identified by the screening and FMECA process.
First, it was observed that the X-750 flexures in Westinghouse plants were obsolete due to plant modifications to resolve the aging concerns. These flexures were removed from subsequent consideration. Second, it was noted that the Zr-4 thimble tubes in the CE In-Core Instrumentation system were known to be subject to an irradiation growth phenomenon that was not addressed as one of the eight degradation modes. These thimble tubes were automatically placed in Category C.
Of the remaining components, 12 Westinghouse and 13 CE components ranked as medium failure likelihood and low failure consequence were automatically placed in Category B.
Evaluations of the impact of each of the identified degradation mechanisms were used to rank the significance of the remaining 19 Westinghouse and 9 CE components. Based on that ranking, 12 Westinghouse components were identified as Category C and an additional 6 Westinghouse components were added to the Category B list. A total of 6 CE components (including the Zr-4 thimble tubes mentioned above) were identified as Category C, with the remaining 4 components added to Category B.
There were two additional exceptions to this categorization process discussed in Section 7.2 of MRP- 191:
- 1. The Westinghouse lower support casting, had been identified as a FMECA Group 2 component based on the consequences of an assumedfailure.However, consistentwith the MRP-134 definitions, this component was placed into CategoryA after consideration of the very low probabilityof degradationand consequence due to the identified thermal embrittlement degradationmechanism.
- 2. The otherOne exception is the internalshold down springfabricatedfrom 304 SS.
Thermal "ratcheting' leading to permanent deformation, is not one of the explicitly characterizeddegradationmechanismsfrom MRP-175 but may occur in this component and reduce the spring hold-downforce over time. This particularphenomenon was assessed to have a moderate likelihood of occurrence;hence, it was assigned to Category B to warrantattention during the development of Inspection andEvaluation (I&E) guidelines.
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MRP-227 Roadmap October 29, 2010 The final list of 31 Westinghouse and 23 CE Category B and Category C items is provided in MRP-191 Tables 7-2 and 7-3. This information is summarized here in Tables 6 and 7. This list of Category B and C Components is carried forward into MRP-227 Tables 3-2 and 3-3. The aging management strategy for the reactor internals is built around examination of these items.
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