ML15335A303

Official Exhibit - NRC00208B-00-BD01 - MRP Letter 2014-09, Biennial Report of Recent MRP-227-A Reactor Internals Inspection Results (May 12, 2014)

ML15335A303
Person / Time
Site:	Indian Point
Issue date:	05/12/2014
From:	NRC/OGC
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28143, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15335A303 (31)

Text

United States Nuclear Regulatory Commission Official Hearing Exhibit NRC00208B In the Matter of: Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3) Submitted: August 10, 2015 ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NRC00208B-00-BD01 Identified: 11/5/2015 Admitted: 11/5/2015 Withdrawn:

Rejected: Stricken:

Other:

ENCLOSURE TO EPRI LETTER MRP-2014-009 INDIVIDUAL UTILITY REPORTS OF MRP-227-A RELATED INSPECTION RESULTS

Plant: Ginna Nuclear Power Plant Utility: Constellation Energy Nuclear Group Date of Exams: May 2011 Plant Age: 41.6 (years) / 33.51 EFPY Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Control Rod Guide Loss of Material None Visual (VT-3) 20% examination 100% of all Initial on site results Tube Assembly (Wear) examination, of the number of Guide Cards were Sat.

Guide plates (cards) CRGT assemblies, along with with all guide the Aggressive Guide cards within each continuous Card wear or selected CRGT section due Backside wear was assembly to not observed at any examined, accessibility measured Guide during Split Tube locations.

Pin replacement Westinghouse activities evaluation was to, "Inspect after an additional 35 EFPYs or as required per MRP-227".

Control Rod Guide Cracking (SCC, Bottom-mounted Enhanced 100% of outer 100% of all No Recordable Tube Assembly Fatigue) instrumentation visual (EVT-1) (accessible) Lower Indications.

Lower flange welds (BMI) column examination to CRGT lower Flanges bodies, determine the flange weld welds were Lower support presence of surfaces and inspected column bodies (cast) crack-like adjacent base due to surface flaws metal. accessibility in flange during Split welds. Pin replacement activities Access to 100% of the Guide Cards and Lower Flange welds was only achievable due to the Split Pin replacement activities.

1

Ginna Nuclear Power Plant

.1 Effect Expansion Link Examination Coverage Examination Item EfcExasoLnk Method (Mechanism) (Note 1) (Note 1) Coverage Achieved Findings Core Barrel Cracking (SCC) Lower support Enhanced 100% of one side (100% of the No Recordable Assembly column bodies (non visual (EVT-1) of the accessible required Indications.

Upper core barrel cast) examination, surfaces of the coverage) of flange weld selected weld and the OD weld adjacent base and partial metal. coverage of the ID weld of the Upper Core Barrel Flange weld per MRP-227 Rev 0.

Core Barrel Cracking (SCC, None Enhanced 100% of one side N/A per No Recordable Assembly IASCC, Fatigue) visual (EVT-1) of the accessible MRP-227 Indications.

Upper and lower Aging examination, surfaces of the Rev 0, core barrel cylinder Management (IE) selected weld and VT-3 per ISI girth welds adjacent base Program, metal. Examination Category B-N-3 Core Barrel Cracking (SCC, None Enhanced 100% of one side N/A per No Recordable Assembly Fatigue) visual (EVT-1) of the accessible MRP-227 Indications.

Lower core barrel examination. surfaces of the Rev 0, flange weld selected weld and VT-3 per ISI adjacent base Program, metal. Examination Category I_ _ I_

_B-N-3 I The ID was more accessible due to Baffle Former Bolt inspection and replacement activities; however, the ID surface was ground flush and very non-contrasting. It was difficult for the remote camera to focus on the area of interest, i.e., weld surface, HAZ and adjacent base material. We could not assure 100% coverage due to the camera's inability to maintain focus on these surfaces. The OD required multiple scans using an ROV to assure coverage and EVT-1 required scanning speed limits <1/2"/sec.

The "Core Barrel Assembly Upper and lower core barrel cylinder girth welds" and the "Core Barrel Assembly Lower core barrel flange weld" items were not inspected EVT-1 due to the MRP-227 Rev 0 requirements. VT-3s were performed per the ISI Program.

2

Ginna Nuclear Power Plant Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Baffle-Former Cracking (IASCC, None Visual (VT-3) Bolts and locking 100% No Recordable Assembly Fatigue) that examination, devices on high Indications.

Baffle-edge bolts results in fluence seams.

" Lost or broken 100% of locking devices components

" Failed or accessible from missing bolts core side.

" Protrusion of bolt heads Although accessibility was not an issue, multiple color cameras were damaged during the inspections inside the Core Barrel due to high Radiation fields. A black and white camera was eventually used to complete these exams.

3

'4

Ginna Nuclear Power Plant EffectExamination Effect Expansion Link metion Examination Coverage Examination Item (Mechanism) (Note 1) Method Coverage Achieved Findings (Mchnsm (oe )(Note 1)1 Baffle-Former Cracking (IASCC, Lower support Volumetric 100% of Plant Specific Assembly Fatigue) column bolts, (UT) accessible bolts or Justification Baffle-former bolts Barrel-former bolts examination, as supported by per MRP-227 plant-specific Rev 0 justification. Heads accessible from 1-UT of 56 1-All Acceptable the core side. UT bolts replaced accessibility may in 1999..

be affected by complexity of head 2-Replaced 25 2-All Acceptable and locking device of the designs. originally planned replacement of 182 Bolts.

3-UT from the 3-All Acceptable back side of 24 removed bolts.

4-UT of 99 4-98 Acceptable Head to Shank 1 crack like region of the indication.

in-place Acceptable by original bolts revised pattern in the analysis minimum bolting pattern.

4

Ginna Nuclear Power Plant Item Effect Expansion Link Examination Method Examination Coverage Examination (Mechanism) (Note 1) (Note 1) Coverage Achieved Findings Baffle-Former Distortion (Void None Visual (VT-3) Core side surface 100% No Recordable Assembly Swelling), or examination to as indicated. Indications.

Assembly Cracking (IASCC) check for (Includes: Baffle that results in evidence of plates, baffle edge

• Abnormal distortion.

bolts and indirect interaction with effects of void fuel assemblies swelling in former e Gaps along plates) high fluence baffle joint

• Vertical displacement of baffle plates near high fluence joint

• Broken or damaged edge bolt locking systems along high fluence baffle joint Although accessibility was not an issue, multiple color cameras were damaged during the inspections inside the Core Barrel due to high Radiation fields. A black and white camera was eventually used to complete these exams.

Alignment and Distortion (Loss None Direct Measurements N/A, The N/A Interfacing of Load) measurement should be taken at Ginna Components of spring several points Internals Internals hold down Note: This height. around the hold down spring mechanism was circumference of spring is 410 not strictly the spring, with a Stainless identified in the statistically Steel.

original list of adequate number age-related of measurements degradation at each point to mechanisms [7]. minimize uncertainty.

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Ginna Nuclear Power Plant Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Thermal Shield Cracking None Visual (VT-3). 100% of thermal 100% No Recordable Assembly (Fatigue) shield flexures. Indications.

Thermal shield or Loss of flexures Material (Wear) that results in thermal shield flexures excessive wear, fracture, or complete separation No issues were experienced during these examinations. The Thermal Shield Flexures are accessible, in fact have previously been included in the ISI 10 Vessel Exams.

Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227 rev 0.

6

Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants Plant Name: Kewanee Power Station Utility: _Dominion Date of Exams:_4-13-12 to 4-14-12 Plant Age: 38 -(years) /_31.1 EFPY Primary Components: The only MRP-227-A inspections performed at KPS during the KR32 RFO are the Control Rod Guide Tube Assembly Guide plates (cards)

-_0L1l ,1 " I11*

I , L JaJ.YLdLUIt: UIILII Lllt:

next scheduled inspection required by MRP-227-A based upon wear projections performed in accordance with WCAP-17451-P and WCAP-17562-P. The inspection results are documented in WCAP-17598-P.

Observed volumetric GC wear range from 21%-51%.

7

j Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants A Plant Name: Surrv Unit 1 Utility: Dominion Generation Date of Exams: May 13h - 18 ,2012 Plant Age: 40 (May 26, 1972) (years) / 29.6 EFPY Primary Components Comments: The following CRGTs were examined per the requirements of MRP - 227 - A: B8 (Bank D) Near 'B' Hot Leg Nozzle, B6 (Bank A), C7 (Shutdown Bank A), D6 (Bank B), Ell (Shutdown Bank B), G9 (Shutdown Bank B), H10 (Bank C), K14 (Bank A) Near 'A' Hot Leg Nozzle, K12 (Bank B), P8 (Bank B). Bank refers to control or shutdown bank group (typ). Of the CRGTs examined, four were common to H.B. Robinson: B8, B6, K14, and P8. No findings to report - all examinations were satisfactory, and ligament wear was < 5%.

Surry is 3 loop design with 15 x 15 fuel. It operates as a "low flow" plant. The inspected guide tubes were installed in 1984 (at 6.8 EFPY) as replacements for the original tubes. The accumulated EFPY on the guide tubes was 22.8 EFPY. The wear rate is therefore approximately 0.22 %/EFPY. Extrapolation of the recorded results to the current plant age of 29.6 EFPY would not change the conclusion that the long term expected guide card wear for Surry Unit I is very small.

8

Surry Unit 1

/An accessiDie weias were examinea, and the MRP-227-A inspection requirement was satisfied. No relevant indications were noted.

L;omments:

Specific Guide Tubes inspected: B -6, B -8, B- 10, D-4, D-6, D- 10, D - 12, F-2, F-4, F - 12, F- 14, H -2, H - 14, K-2, K-4, K- 12, K-14, M-4, M-6, M-10, M-12, P-6, P-8, and P-10.

The continuous section of Surry's guide tubes is of the "open" design and is not enclosed by a shroud. Therefore, the welds of the continuous section "sheaths" were inspected.

Each guide tube sheath has a "U"shaped weldment, the three legs of which were considered to be three separate welds for inspection and coverage calculation purposes. The lower or end leg of each weldment was typically more accessible for an EVT-1 inspection than the two side legs. A greater number of welds could be inspected on the upper flange of the continuous section as compared to the lower. Some of the lower flanges were obstructed by core exit thermocouple mixer tubes.

An additional 42 welds were attempted but could not be inspected to the EVT-1 standard; these are recorded as "best effort" examinations. No relevant conditions were identified among these 42 best effort examinations.

The inspected guide tubes were installed in 1984 (at 6.8 EFPY) as replacements for the original tubes. The accumulated EFPY on the guide tubes was 22.8 EFPY. The degradation mechanisms of concern in this inspection are fatigue (high cycle) and SCC. For constant parameters conducive to degradation both these mechanisms would be expected to have observable effects at a relatively early stage in plant life. Since the replacement guide tubes have accumulated 22.8 EFPY (77%) of a total 29.6 EFPY plant operation, these examinations are considered valid for concluding that the linked Expansion components are also similarly free of significant degradation. This conclusion is also reinforced by the core barrel upper flange weld (original equipment) inspection, and is similarly free of SCC indications.

9

Surrv Unit 1 The EVT - 1 examination of the one upper core barrel weld was SAT; the weld was examined from the ID surface as required per the MRP guidance. There were no indications.

The exam fulfilled the requirements of MRP-227-A.

It is not feasible to inspect the OD of this weld when the core barrel is removed because it sits above the water and must be heavily shielded.

Examination has not N/A been done yet.

Examination has not N/A been done yet.

10

Surry Unit 1 IComments:

During 1R23 (Fall 2010), One baffle edge bolt was found to 936 accessible baffle - have a missing weld on one side of its edge bolts received VT - lock bar. This was attributed to a 3 inspection, fabrication error, not aging effects, and is acceptable for continued safe operation.

Comments: These exams were completed before EPRI had pi - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

During 1R23 (Fall 2010), The most significant examination 1088 baffle - to - former result was detection of a likely flaw in bolts received UT one bolt, identified as "C113". The examination and the depth of the flaw in C1 13 is not locking bars received VT quantified; however it is located at the

- 3 examinations. head to shank region of the bolt. This condition was found acceptable by a bolting pattern analysis. Also, some channels of the UT signals showed a significant "back wall" reflection from the end of the bolt, so the bolt is not completely severed. The lock bar for this bolt has no relevant conditions and is considered capable of Iperforming I its retention function.

Comments: Four other relevant conditions were found on other bolts. Two bolt heads were deformed at the points of its hex head to the extent that the UT probe was slightly displaced from full contact with the bolt head. Because of this, the back wall signal, although strong, was slightly outside its parameters, and the bolts were classified as "non-inspectable". Nevertheless, review of the UT signals showed the strong back wall signal with no intervening indications of flaws between it and the bolt head. It was thus concluded that these bolts were not flawed. The bolt deformation is attributed to original manufacturing, since the bolts are both located at an inner corner adjacent to plate number 1 and have limited accessibility. The other two relevant conditions were one missing lock bar weld out of two required. One baffle-former bolt and one edge bolt had this condition. This condition is also attributed to original fabrication error. The integrity of the lock bar with a single weld is considered adequate in view of the many years of service with this condition.

These exams were completed before EPRI had prepared a MRP - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

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Surrv Unit 1

.4

%-UIIIIIIUIILMI.

Direct measurement of The estimated average spring height the Reactor Vessel HDS was 3.6312+/-0.0002 inches. This was completed during result is greater than the minimum 1R24 - measurements requirement of 3.610 inches; were taken at 8 locations therefore, the final result is in the same general acceptable. This result confirms areas as the as - built adequate hold down capability measurements. Unit l's through at least 60 total years of HDS is 304 - SS. reactor operation.

Comments: Because Unit 1 has an austenitic (304 - SS) stainless steel hold down spring, measurements of its relaxation was required per the MRP-227-A guidance. Per analysis, considering Surry Unit 1 as-built dimensions and the required hold down force for design conditions, the minimum acceptable height for assuring a minimum for 60 year service life was computed.

Spring height measurements were taken at eight locations (every 450) with three individual measurements taken at each location.

The estimated average spring height was 3.6312+/-0.0002 inches. This result is greater than the minimum requirement; therefore, the final result is acceptable. This result confirms adequate hold down capability through at least 60 total years of reactor operation, and no further measurements are required.

N/A - examination N/A - examination scheduled for scheduled for 1R25 - Fall 1 R25 - Fall of 2013.

of 2013.

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Surry Unit 1

• ,UlI II I lllIIL.-,

Notes to Westinghouse Primary Components Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% of the total identified sample population must be examined

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined from either the inner or outer diameter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

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Surry Unit 1 Expansion Components I Exminan Method Required Examination Coverage Examination Coverage Achieved Findings (Note 1)

Upper Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Upper core plate Comments:

Lower Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support forging or castings See Figure 4-33 of MRP-227-A.

Comments:

Core Barrel Volumetric 100% of accessible bolts. N/A N/A Assembly examination (UT) Accessibility may be Barrel-former bolts limited by presence of thermal shields or neutron pads (Note 2).

See Figure 4-23 of MRP-227-A.

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Surry Unit 1 S.Exmaon Coverage Achieved I Findings (Note 1)

Comments:

Lower Support Volumetric 100% of accessible bolts N/A N/A Assembly examination (UT) or as supported by plant-Lower support specific justification (Note column bolts 2).

See Figures 4-32 and 4-33 of MRP-227-A.

Comments:

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Core barrel outlet selected weld and nozzle welds adjacent base metal (Note 2).

See Figure 4-22 of MRP-227-A.

Comments:

15

Surry Unit 1 Item Examination Meho Required Examination Coverage Examination Coverage Achieved Findings (Note 1)

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Upper and lower core selected weld and barrel cylinder axial adjacent base metal (Note welds 2).

See Figure 4-22 of MRP-227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(non cast) 227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) support columns (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(cast) 227-A.

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Surry Unit 1 Comments:

Bottom Mounted Visual examination 100% of BMI column N/A N/A Instrumentation (VT-3) bodies for which difficulty System is detected during flux Bottom-mounted thimble instrumentation (BMI) insertion/withdrawal.

column bodies See Figure 4-35 of MRP-227-A.

Comments:

Notes to Westinghouse Expansion Component Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

V 17

Surry Unit 1 varlAsml iuleamnto Existing Programs Components

-ebre lne (-73 odtriesrae lcesbeNA- examinatio tseiid schedule 1R25 - Fall of 2013.

for NA- exami sche 1R25 - Fall of 2013.

for L;o mments:

Daccesible N/A - examination scheduled for N/A - examination scheduled for tVT-3) satspecified 1R25 - Fall of 2013. 1R25 - Fall of 2013.

)er sppor ringor fequency.

~oi mments:

All le N/A - examination scheduled for N/A - examination scheduled for

ebl eaintinofth uracsat seied 1R25 -Fall of 2013. 1R25 -Fallof 2013.

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Surry Unit 1 Comments:

N/A - examination scheduled for N/A - examination scheduled for 1R25 - Fall of 2013. 1R25 - Fall of 2013.

Comments:

Be 1R24 (May 2012) 49 of 50 flux No issues were noted.

Instumetatin eamiatio (E) eamintio asthimble tubes for SPS Unit 1. 1 Systm deinedin pant flux thimble tube is capped -

Fluxthibletube repone toIEB88- removed from service (tube F-4).

Comments:

The following flux thimble tubes were replaced prior to the May 2012 inspection: B10, D12, E5, F2, HI, H13, J3, J12, L4, L5, L14, N12, &

R8.

Flux Thimble tube B7 had 43% OUTER WALL DEGRADATION.

Flux Thimble tube Ell had 10% INNER WALL DEGRADATION.

N/A - examination scheduled for N/A - examination scheduled for 1R25 - Fall of 2013. 1R25 - Fall of 2013.

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Surry Unit 1 Comments:

N/A - examination scheduled for N/A - examination scheduled for sufacesaspecified 1R25 - Fall of 2013. 1R25 - Fall of 2013.

Comments:

Notes to Westinghouse Existing Programs Components Table:

1. XL = "Extra Long" referring to Westinghouse plants with 14-foot cores.

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Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants Plant Name: Surry Unit 2 Utility: Dominion Generation Date of Exams: Nov 13 - 18'h, 2012 Plant Age: 40 (Jan 29, 1973) (years) / 30.04 EFPY Primary Components Comments: The following CRGTs were examined per the requirements of MRP - 227 - A: C7 (Shutdown Bank A) Near 'C' Cold Leg Nozzle, D4 (Bank C) Near 'C' Cold Leg Nozzle, F10 (Bank D) Control Bank Near 'B' Cold Leg Nozzle, G9 (Shutdown Bank B) Shutdown Bank Near Core Center, J7 (Shutdown Bank B) Shutdown Bank Near Core Center, J13 (Shutdown Bank A) Shutdown Bank Near 'B' Hot Leg, K4 (Bank B) Control Bank B Near 'A' Hot Leg Nozzle [RHR Supply], LIl (Shutdown Bank B) Shutdown Bank Near 'B' Hot Leg Nozzle, M10 (Bank B) Control Bank B Between 'A' Cold Leg Nozzle and 'B' Hot Leg Nozzle, N9 (Shutdown Bank A) Near 'A' Cold Leg.

Bank refers to control or shutdown bank group (typ). Of the CRGTs examined, one was common to Surry Unit 1. No findings to report -

all examinations were satisfactory, and ligament wear was < 10% (<5% by volume). These CRGTs are different than the ones inspected for Surry Unit I in that they have 9 card levels instead of 8; the 20% sample included most CRGTs that were identified by Westinghouse as having high wear based on an FME inspection in 2005.

Surry is 3 loop design with 15 x 15 fuel. It operates as a "low flow" plant, and it has operated with a "very low leakage" core since 1984.

The inspected guide tubes are original except for the split pins with 9 guide cards per tube; the split pins were replaced in 2005. Since the replacement split pins were functionally equivalent to the original ones, the replacements should not affect guide card wear rates.

The accumulated EFPY on the guide tubes was 30.22 EFPY.

The guide card listed below had wear volume equal to approximately 5%: N 7 - Ligament C. The rest had wear < 5%; therefore, all guide cards examined fell into a wear level of GREEN described in WCAP - 17562 - P, Rev. 0.

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Surry Unit 2 Accessible Lower Flange All accessible welds were examined, Welds (Upper and Lower) and the MRP-227-A inspection 24 outer CRGT requirement was satisfied. No assemblies were relevant indications were noted.

inspected during the 1R24 outage. 272 welds achieved some level of EVT-1 access. Of these, some did not have 100%

coverage. The average coverage for the 272 welds was 95.7% which exceeds the minimum 75% requirement of MRP-227-A.

Comments:

Specific Guide Tubes inspected: B-6, B -8, B- 10, C-7, C-9, D-4, D-6, D- 10, D- 12, E-5, E- 11, F-2, F-4, F- 12, F- 14, G-3, G-13, H-2, H-14, J-3, J-13, K-2, K-4, K-12, K-14, L-5, L-11, M-4, M-6, M-10, M-12, N-9, P-6, P-8, and P-10.

The continuous section of Surry's guide tubes is of the "open" design and is not enclosed by a shroud. Therefore, the welds of the continuous section "sheaths" were inspected.

Each guide tube sheath has a "U"shaped weldment, the three legs of which were considered to be three separate welds for inspection and coverage calculation purposes. The lower or end leg of each weldment was typically more accessible for an EVT-1 inspection than the two side legs. A greater number of welds could be inspected on the upper flange of the continuous section as compared to the lower. Some of the lower flanges were obstructed by core exit thermocouple mixer tubes.

An additional 21 welds were attempted but could not be inspected to the EVT-1 standard; these are recorded as "best effort" examinations. No relevant conditions were identified among these 21 best effort examinations.

The inspected guide tubes are the original tubes. The accumulated EFPY on the guide tubes was 30.04 EFPY. The degradation mechanisms of concern in this inspection are fatigue (high cycle) and SCC. For constant parameters conducive to degradation both these mechanisms would be expected to have observable effects at a relatively early stage in plant life.

22

Surry Unit 2 100% of the accessible The EVT - 1 examination of the one weld received an EVT - 1 upper core barrel weld was SAT; the examination, weld was examined from the ID surface as required per the MRP guidance. There were no indications.

The exam fulfilled the requirements of MRP-227-A.

Comments:

It is not feasible to inspect the OD of this weld when the core barrel is removed because it sits above the water and must be heavily shielded. There were no reportable indications observed on the Upper Core Barrel Flange Weld, but a rub - mark was observed below the Upper Core Barrel Flange Weld in the 900 - 1800 Quadrant.

Examination has not N/A been done yet.

These exams are scheduled for 2R25.

Comments: Scheduled during 21125- Spring 2014.

23

Surry Unit 2 During 2R23 (Spring All 936 accessible edge bolts 2011), 936 accessible received the required VT - 3 baffle - edge bolts examinations. No issues were note; received VT - 3 all examinations were acceptable.

inspection.

Comments: These exams were completed before EPRI had prepared a - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

24

Surry Unit 2 LJUIIuIn /O kopring i ne most s51n1,u.dnIL xýdIIrnIIdLIUrl 2011), 1088 baffle - to - results were detection of likely flaws former bolts received UT in two bolts, identified as Baffle Plate examination and the 22 Bolt "G63" and Baffle Plate 42 Bolt locking bars received VT "A125". The depth of the flaw in G63

- 3 examinations, and A125 were not quantified; however, each flaw was located at the head to shank region of the bolt.

Also, some channels of the UT signals showed a "back wall" reflection from the end of each bolt, so each bolt is not completely severed. The lock bar for each of these bolts has no relevant conditions and is considered capable of 1performing its retention function.

t.omments: oee aoove.

These exams were completed before EPRI had prepared a MRP - 227 report ing template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

The inspection for gaps and distortion due to void swelling has not been done yet. The edge bolt inspection is reported above.

These exams are scheduled for 2R25.

25

Surry Unit 2 Comments: Because Unit 2 has an austenitic (304 - SS) stainless steel hold down spring, measurements of its relaxation was required per the MRP-227-A guidance. Per analysis, considering Surry Unit 2 as-built dimensions and the required hold down force for design conditions, the minimum acceptable height for assuring a minimum for 60 year service life was computed.

Spring height measurements were taken at eight locations (every 450) with three individual measurements taken at each location.

The estimated average spring height was 3.6384+/-0.0001 inches. This result is greater than the minimum requirement; therefore, the final result is acceptable. This result confirms adequate hold down capability through at least 60 total years of reactor operation, and no further measurements are required.

N/A - examination N/A - examination scheduled for scheduled for 2R25 - 2R25 - Spring of 2014.

Spring of 2014.

Notes to Westinghouse Primary Components Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% of the total identified sample population must be examined.

26

Surry Unit 2

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined from either the inner or outer diameter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

27

Surry Unit 2 V

Expansion Components

.Requied Examination Coverage Examination iteEamiaton etodCoverage Achieved Findings (Note 1)

Upper Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Upper core plate Comments:

Lower Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support forging or castings See Figure 4-33 of MRP-227-A.

Comments:

Core Barrel Volumetric 100% of accessible bolts. N/A N/A Assembly examination (UT) Accessibility may be Barrel-former bolts limited by presence of thermal shields or neutron pads (Note 2).

See Figure 4-23 of MRP-227-A.

28

Surry Unit 2 L.omments:

Lower Support Volumetric 100% of accessible bolts N/A N/A Assembly examination (UT) or as supported by plant-Lower support specific justification (Note column bolts 2).

See Figures 4-32 and 4-33 of MRP-227-A.

Comments:

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Core barrel outlet selected weld and nozzle welds adjacent base metal (Note 2).

See Figure 4-22 of MRP-227-A.

Comments:

a 29

Surry Unit 2 Examination Method Required Exmnto Coverage Examination

........ Item _...Coverage Achieved Findings (Note 1)

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Upper and lower core selected weld and barrel cylinder axial adjacent base metal (Note welds 2).

See Figure 4-22 of MRP-227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(non cast) 227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) support columns (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(cast) 227-A.

30

United States Nuclear Regulatory Commission Official Hearing Exhibit NRC00208B In the Matter of: Entergy Nuclear Operations, Inc.

(Indian Point Nuclear Generating Units 2 and 3) Submitted: August 10, 2015 ASLBP #: 07-858-03-LR-BD01 Docket #: 05000247 l 05000286 Exhibit #: NRC00208B-00-BD01 Identified: 11/5/2015 Admitted: 11/5/2015 Withdrawn:

Rejected: Stricken:

Other:

ENCLOSURE TO EPRI LETTER MRP-2014-009 INDIVIDUAL UTILITY REPORTS OF MRP-227-A RELATED INSPECTION RESULTS

Plant: Ginna Nuclear Power Plant Utility: Constellation Energy Nuclear Group Date of Exams: May 2011 Plant Age: 41.6 (years) / 33.51 EFPY Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Control Rod Guide Loss of Material None Visual (VT-3) 20% examination 100% of all Initial on site results Tube Assembly (Wear) examination, of the number of Guide Cards were Sat.

Guide plates (cards) CRGT assemblies, along with with all guide the Aggressive Guide cards within each continuous Card wear or selected CRGT section due Backside wear was assembly to not observed at any examined, accessibility measured Guide during Split Tube locations.

Pin replacement Westinghouse activities evaluation was to, "Inspect after an additional 35 EFPYs or as required per MRP-227".

Control Rod Guide Cracking (SCC, Bottom-mounted Enhanced 100% of outer 100% of all No Recordable Tube Assembly Fatigue) instrumentation visual (EVT-1) (accessible) Lower Indications.

Lower flange welds (BMI) column examination to CRGT lower Flanges bodies, determine the flange weld welds were Lower support presence of surfaces and inspected column bodies (cast) crack-like adjacent base due to surface flaws metal. accessibility in flange during Split welds. Pin replacement activities Access to 100% of the Guide Cards and Lower Flange welds was only achievable due to the Split Pin replacement activities.

1

Ginna Nuclear Power Plant

.1 Effect Expansion Link Examination Coverage Examination Item EfcExasoLnk Method (Mechanism) (Note 1) (Note 1) Coverage Achieved Findings Core Barrel Cracking (SCC) Lower support Enhanced 100% of one side (100% of the No Recordable Assembly column bodies (non visual (EVT-1) of the accessible required Indications.

Upper core barrel cast) examination, surfaces of the coverage) of flange weld selected weld and the OD weld adjacent base and partial metal. coverage of the ID weld of the Upper Core Barrel Flange weld per MRP-227 Rev 0.

Core Barrel Cracking (SCC, None Enhanced 100% of one side N/A per No Recordable Assembly IASCC, Fatigue) visual (EVT-1) of the accessible MRP-227 Indications.

Upper and lower Aging examination, surfaces of the Rev 0, core barrel cylinder Management (IE) selected weld and VT-3 per ISI girth welds adjacent base Program, metal. Examination Category B-N-3 Core Barrel Cracking (SCC, None Enhanced 100% of one side N/A per No Recordable Assembly Fatigue) visual (EVT-1) of the accessible MRP-227 Indications.

Lower core barrel examination. surfaces of the Rev 0, flange weld selected weld and VT-3 per ISI adjacent base Program, metal. Examination Category I_ _ I_

_B-N-3 I The ID was more accessible due to Baffle Former Bolt inspection and replacement activities; however, the ID surface was ground flush and very non-contrasting. It was difficult for the remote camera to focus on the area of interest, i.e., weld surface, HAZ and adjacent base material. We could not assure 100% coverage due to the camera's inability to maintain focus on these surfaces. The OD required multiple scans using an ROV to assure coverage and EVT-1 required scanning speed limits <1/2"/sec.

The "Core Barrel Assembly Upper and lower core barrel cylinder girth welds" and the "Core Barrel Assembly Lower core barrel flange weld" items were not inspected EVT-1 due to the MRP-227 Rev 0 requirements. VT-3s were performed per the ISI Program.

2

Ginna Nuclear Power Plant Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Baffle-Former Cracking (IASCC, None Visual (VT-3) Bolts and locking 100% No Recordable Assembly Fatigue) that examination, devices on high Indications.

Baffle-edge bolts results in fluence seams.

" Lost or broken 100% of locking devices components

" Failed or accessible from missing bolts core side.

" Protrusion of bolt heads Although accessibility was not an issue, multiple color cameras were damaged during the inspections inside the Core Barrel due to high Radiation fields. A black and white camera was eventually used to complete these exams.

3

'4

Ginna Nuclear Power Plant EffectExamination Effect Expansion Link metion Examination Coverage Examination Item (Mechanism) (Note 1) Method Coverage Achieved Findings (Mchnsm (oe )(Note 1)1 Baffle-Former Cracking (IASCC, Lower support Volumetric 100% of Plant Specific Assembly Fatigue) column bolts, (UT) accessible bolts or Justification Baffle-former bolts Barrel-former bolts examination, as supported by per MRP-227 plant-specific Rev 0 justification. Heads accessible from 1-UT of 56 1-All Acceptable the core side. UT bolts replaced accessibility may in 1999..

be affected by complexity of head 2-Replaced 25 2-All Acceptable and locking device of the designs. originally planned replacement of 182 Bolts.

3-UT from the 3-All Acceptable back side of 24 removed bolts.

4-UT of 99 4-98 Acceptable Head to Shank 1 crack like region of the indication.

in-place Acceptable by original bolts revised pattern in the analysis minimum bolting pattern.

4

Ginna Nuclear Power Plant Item Effect Expansion Link Examination Method Examination Coverage Examination (Mechanism) (Note 1) (Note 1) Coverage Achieved Findings Baffle-Former Distortion (Void None Visual (VT-3) Core side surface 100% No Recordable Assembly Swelling), or examination to as indicated. Indications.

Assembly Cracking (IASCC) check for (Includes: Baffle that results in evidence of plates, baffle edge

• Abnormal distortion.

bolts and indirect interaction with effects of void fuel assemblies swelling in former e Gaps along plates) high fluence baffle joint

• Vertical displacement of baffle plates near high fluence joint

• Broken or damaged edge bolt locking systems along high fluence baffle joint Although accessibility was not an issue, multiple color cameras were damaged during the inspections inside the Core Barrel due to high Radiation fields. A black and white camera was eventually used to complete these exams.

Alignment and Distortion (Loss None Direct Measurements N/A, The N/A Interfacing of Load) measurement should be taken at Ginna Components of spring several points Internals Internals hold down Note: This height. around the hold down spring mechanism was circumference of spring is 410 not strictly the spring, with a Stainless identified in the statistically Steel.

original list of adequate number age-related of measurements degradation at each point to mechanisms [7]. minimize uncertainty.

5

Ginna Nuclear Power Plant Effect Expansion Link Examination Coverage Examination Item EfcExasoLik (Mechanism) Method (Note 1) (Note 1) Coverage Achieved Findings Thermal Shield Cracking None Visual (VT-3). 100% of thermal 100% No Recordable Assembly (Fatigue) shield flexures. Indications.

Thermal shield or Loss of flexures Material (Wear) that results in thermal shield flexures excessive wear, fracture, or complete separation No issues were experienced during these examinations. The Thermal Shield Flexures are accessible, in fact have previously been included in the ISI 10 Vessel Exams.

Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227 rev 0.

6

Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants Plant Name: Kewanee Power Station Utility: _Dominion Date of Exams:_4-13-12 to 4-14-12 Plant Age: 38 -(years) /_31.1 EFPY Primary Components: The only MRP-227-A inspections performed at KPS during the KR32 RFO are the Control Rod Guide Tube Assembly Guide plates (cards)

-_0L1l ,1 " I11*

I , L JaJ.YLdLUIt: UIILII Lllt:

next scheduled inspection required by MRP-227-A based upon wear projections performed in accordance with WCAP-17451-P and WCAP-17562-P. The inspection results are documented in WCAP-17598-P.

Observed volumetric GC wear range from 21%-51%.

7

j Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants A Plant Name: Surrv Unit 1 Utility: Dominion Generation Date of Exams: May 13h - 18 ,2012 Plant Age: 40 (May 26, 1972) (years) / 29.6 EFPY Primary Components Comments: The following CRGTs were examined per the requirements of MRP - 227 - A: B8 (Bank D) Near 'B' Hot Leg Nozzle, B6 (Bank A), C7 (Shutdown Bank A), D6 (Bank B), Ell (Shutdown Bank B), G9 (Shutdown Bank B), H10 (Bank C), K14 (Bank A) Near 'A' Hot Leg Nozzle, K12 (Bank B), P8 (Bank B). Bank refers to control or shutdown bank group (typ). Of the CRGTs examined, four were common to H.B. Robinson: B8, B6, K14, and P8. No findings to report - all examinations were satisfactory, and ligament wear was < 5%.

Surry is 3 loop design with 15 x 15 fuel. It operates as a "low flow" plant. The inspected guide tubes were installed in 1984 (at 6.8 EFPY) as replacements for the original tubes. The accumulated EFPY on the guide tubes was 22.8 EFPY. The wear rate is therefore approximately 0.22 %/EFPY. Extrapolation of the recorded results to the current plant age of 29.6 EFPY would not change the conclusion that the long term expected guide card wear for Surry Unit I is very small.

8

Surry Unit 1

/An accessiDie weias were examinea, and the MRP-227-A inspection requirement was satisfied. No relevant indications were noted.

L;omments:

Specific Guide Tubes inspected: B -6, B -8, B- 10, D-4, D-6, D- 10, D - 12, F-2, F-4, F - 12, F- 14, H -2, H - 14, K-2, K-4, K- 12, K-14, M-4, M-6, M-10, M-12, P-6, P-8, and P-10.

The continuous section of Surry's guide tubes is of the "open" design and is not enclosed by a shroud. Therefore, the welds of the continuous section "sheaths" were inspected.

Each guide tube sheath has a "U"shaped weldment, the three legs of which were considered to be three separate welds for inspection and coverage calculation purposes. The lower or end leg of each weldment was typically more accessible for an EVT-1 inspection than the two side legs. A greater number of welds could be inspected on the upper flange of the continuous section as compared to the lower. Some of the lower flanges were obstructed by core exit thermocouple mixer tubes.

An additional 42 welds were attempted but could not be inspected to the EVT-1 standard; these are recorded as "best effort" examinations. No relevant conditions were identified among these 42 best effort examinations.

The inspected guide tubes were installed in 1984 (at 6.8 EFPY) as replacements for the original tubes. The accumulated EFPY on the guide tubes was 22.8 EFPY. The degradation mechanisms of concern in this inspection are fatigue (high cycle) and SCC. For constant parameters conducive to degradation both these mechanisms would be expected to have observable effects at a relatively early stage in plant life. Since the replacement guide tubes have accumulated 22.8 EFPY (77%) of a total 29.6 EFPY plant operation, these examinations are considered valid for concluding that the linked Expansion components are also similarly free of significant degradation. This conclusion is also reinforced by the core barrel upper flange weld (original equipment) inspection, and is similarly free of SCC indications.

9

Surrv Unit 1 The EVT - 1 examination of the one upper core barrel weld was SAT; the weld was examined from the ID surface as required per the MRP guidance. There were no indications.

The exam fulfilled the requirements of MRP-227-A.

It is not feasible to inspect the OD of this weld when the core barrel is removed because it sits above the water and must be heavily shielded.

Examination has not N/A been done yet.

Examination has not N/A been done yet.

10

Surry Unit 1 IComments:

During 1R23 (Fall 2010), One baffle edge bolt was found to 936 accessible baffle - have a missing weld on one side of its edge bolts received VT - lock bar. This was attributed to a 3 inspection, fabrication error, not aging effects, and is acceptable for continued safe operation.

Comments: These exams were completed before EPRI had pi - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

During 1R23 (Fall 2010), The most significant examination 1088 baffle - to - former result was detection of a likely flaw in bolts received UT one bolt, identified as "C113". The examination and the depth of the flaw in C1 13 is not locking bars received VT quantified; however it is located at the

- 3 examinations. head to shank region of the bolt. This condition was found acceptable by a bolting pattern analysis. Also, some channels of the UT signals showed a significant "back wall" reflection from the end of the bolt, so the bolt is not completely severed. The lock bar for this bolt has no relevant conditions and is considered capable of Iperforming I its retention function.

Comments: Four other relevant conditions were found on other bolts. Two bolt heads were deformed at the points of its hex head to the extent that the UT probe was slightly displaced from full contact with the bolt head. Because of this, the back wall signal, although strong, was slightly outside its parameters, and the bolts were classified as "non-inspectable". Nevertheless, review of the UT signals showed the strong back wall signal with no intervening indications of flaws between it and the bolt head. It was thus concluded that these bolts were not flawed. The bolt deformation is attributed to original manufacturing, since the bolts are both located at an inner corner adjacent to plate number 1 and have limited accessibility. The other two relevant conditions were one missing lock bar weld out of two required. One baffle-former bolt and one edge bolt had this condition. This condition is also attributed to original fabrication error. The integrity of the lock bar with a single weld is considered adequate in view of the many years of service with this condition.

These exams were completed before EPRI had prepared a MRP - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

11

Surrv Unit 1

.4

%-UIIIIIIUIILMI.

Direct measurement of The estimated average spring height the Reactor Vessel HDS was 3.6312+/-0.0002 inches. This was completed during result is greater than the minimum 1R24 - measurements requirement of 3.610 inches; were taken at 8 locations therefore, the final result is in the same general acceptable. This result confirms areas as the as - built adequate hold down capability measurements. Unit l's through at least 60 total years of HDS is 304 - SS. reactor operation.

Comments: Because Unit 1 has an austenitic (304 - SS) stainless steel hold down spring, measurements of its relaxation was required per the MRP-227-A guidance. Per analysis, considering Surry Unit 1 as-built dimensions and the required hold down force for design conditions, the minimum acceptable height for assuring a minimum for 60 year service life was computed.

Spring height measurements were taken at eight locations (every 450) with three individual measurements taken at each location.

The estimated average spring height was 3.6312+/-0.0002 inches. This result is greater than the minimum requirement; therefore, the final result is acceptable. This result confirms adequate hold down capability through at least 60 total years of reactor operation, and no further measurements are required.

N/A - examination N/A - examination scheduled for scheduled for 1R25 - Fall 1 R25 - Fall of 2013.

of 2013.

12

Surry Unit 1

• ,UlI II I lllIIL.-,

Notes to Westinghouse Primary Components Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% of the total identified sample population must be examined

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined from either the inner or outer diameter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

13

Surry Unit 1 Expansion Components I Exminan Method Required Examination Coverage Examination Coverage Achieved Findings (Note 1)

Upper Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Upper core plate Comments:

Lower Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support forging or castings See Figure 4-33 of MRP-227-A.

Comments:

Core Barrel Volumetric 100% of accessible bolts. N/A N/A Assembly examination (UT) Accessibility may be Barrel-former bolts limited by presence of thermal shields or neutron pads (Note 2).

See Figure 4-23 of MRP-227-A.

14

Surry Unit 1 S.Exmaon Coverage Achieved I Findings (Note 1)

Comments:

Lower Support Volumetric 100% of accessible bolts N/A N/A Assembly examination (UT) or as supported by plant-Lower support specific justification (Note column bolts 2).

See Figures 4-32 and 4-33 of MRP-227-A.

Comments:

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Core barrel outlet selected weld and nozzle welds adjacent base metal (Note 2).

See Figure 4-22 of MRP-227-A.

Comments:

15

Surry Unit 1 Item Examination Meho Required Examination Coverage Examination Coverage Achieved Findings (Note 1)

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Upper and lower core selected weld and barrel cylinder axial adjacent base metal (Note welds 2).

See Figure 4-22 of MRP-227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(non cast) 227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) support columns (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(cast) 227-A.

16

Surry Unit 1 Comments:

Bottom Mounted Visual examination 100% of BMI column N/A N/A Instrumentation (VT-3) bodies for which difficulty System is detected during flux Bottom-mounted thimble instrumentation (BMI) insertion/withdrawal.

column bodies See Figure 4-35 of MRP-227-A.

Comments:

Notes to Westinghouse Expansion Component Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

V 17

Surry Unit 1 varlAsml iuleamnto Existing Programs Components

-ebre lne (-73 odtriesrae lcesbeNA- examinatio tseiid schedule 1R25 - Fall of 2013.

for NA- exami sche 1R25 - Fall of 2013.

for L;o mments:

Daccesible N/A - examination scheduled for N/A - examination scheduled for tVT-3) satspecified 1R25 - Fall of 2013. 1R25 - Fall of 2013.

)er sppor ringor fequency.

~oi mments:

All le N/A - examination scheduled for N/A - examination scheduled for

ebl eaintinofth uracsat seied 1R25 -Fall of 2013. 1R25 -Fallof 2013.

18

Surry Unit 1 Comments:

N/A - examination scheduled for N/A - examination scheduled for 1R25 - Fall of 2013. 1R25 - Fall of 2013.

Comments:

Be 1R24 (May 2012) 49 of 50 flux No issues were noted.

Instumetatin eamiatio (E) eamintio asthimble tubes for SPS Unit 1. 1 Systm deinedin pant flux thimble tube is capped -

Fluxthibletube repone toIEB88- removed from service (tube F-4).

Comments:

The following flux thimble tubes were replaced prior to the May 2012 inspection: B10, D12, E5, F2, HI, H13, J3, J12, L4, L5, L14, N12, &

R8.

Flux Thimble tube B7 had 43% OUTER WALL DEGRADATION.

Flux Thimble tube Ell had 10% INNER WALL DEGRADATION.

N/A - examination scheduled for N/A - examination scheduled for 1R25 - Fall of 2013. 1R25 - Fall of 2013.

19

Surry Unit 1 Comments:

N/A - examination scheduled for N/A - examination scheduled for sufacesaspecified 1R25 - Fall of 2013. 1R25 - Fall of 2013.

Comments:

Notes to Westinghouse Existing Programs Components Table:

1. XL = "Extra Long" referring to Westinghouse plants with 14-foot cores.

20

Tables for Reporting MRP-227-A Inspection Results for Westinghouse Plants Plant Name: Surry Unit 2 Utility: Dominion Generation Date of Exams: Nov 13 - 18'h, 2012 Plant Age: 40 (Jan 29, 1973) (years) / 30.04 EFPY Primary Components Comments: The following CRGTs were examined per the requirements of MRP - 227 - A: C7 (Shutdown Bank A) Near 'C' Cold Leg Nozzle, D4 (Bank C) Near 'C' Cold Leg Nozzle, F10 (Bank D) Control Bank Near 'B' Cold Leg Nozzle, G9 (Shutdown Bank B) Shutdown Bank Near Core Center, J7 (Shutdown Bank B) Shutdown Bank Near Core Center, J13 (Shutdown Bank A) Shutdown Bank Near 'B' Hot Leg, K4 (Bank B) Control Bank B Near 'A' Hot Leg Nozzle [RHR Supply], LIl (Shutdown Bank B) Shutdown Bank Near 'B' Hot Leg Nozzle, M10 (Bank B) Control Bank B Between 'A' Cold Leg Nozzle and 'B' Hot Leg Nozzle, N9 (Shutdown Bank A) Near 'A' Cold Leg.

Bank refers to control or shutdown bank group (typ). Of the CRGTs examined, one was common to Surry Unit 1. No findings to report -

all examinations were satisfactory, and ligament wear was < 10% (<5% by volume). These CRGTs are different than the ones inspected for Surry Unit I in that they have 9 card levels instead of 8; the 20% sample included most CRGTs that were identified by Westinghouse as having high wear based on an FME inspection in 2005.

Surry is 3 loop design with 15 x 15 fuel. It operates as a "low flow" plant, and it has operated with a "very low leakage" core since 1984.

The inspected guide tubes are original except for the split pins with 9 guide cards per tube; the split pins were replaced in 2005. Since the replacement split pins were functionally equivalent to the original ones, the replacements should not affect guide card wear rates.

The accumulated EFPY on the guide tubes was 30.22 EFPY.

The guide card listed below had wear volume equal to approximately 5%: N 7 - Ligament C. The rest had wear < 5%; therefore, all guide cards examined fell into a wear level of GREEN described in WCAP - 17562 - P, Rev. 0.

21

Surry Unit 2 Accessible Lower Flange All accessible welds were examined, Welds (Upper and Lower) and the MRP-227-A inspection 24 outer CRGT requirement was satisfied. No assemblies were relevant indications were noted.

inspected during the 1R24 outage. 272 welds achieved some level of EVT-1 access. Of these, some did not have 100%

coverage. The average coverage for the 272 welds was 95.7% which exceeds the minimum 75% requirement of MRP-227-A.

Comments:

Specific Guide Tubes inspected: B-6, B -8, B- 10, C-7, C-9, D-4, D-6, D- 10, D- 12, E-5, E- 11, F-2, F-4, F- 12, F- 14, G-3, G-13, H-2, H-14, J-3, J-13, K-2, K-4, K-12, K-14, L-5, L-11, M-4, M-6, M-10, M-12, N-9, P-6, P-8, and P-10.

The continuous section of Surry's guide tubes is of the "open" design and is not enclosed by a shroud. Therefore, the welds of the continuous section "sheaths" were inspected.

Each guide tube sheath has a "U"shaped weldment, the three legs of which were considered to be three separate welds for inspection and coverage calculation purposes. The lower or end leg of each weldment was typically more accessible for an EVT-1 inspection than the two side legs. A greater number of welds could be inspected on the upper flange of the continuous section as compared to the lower. Some of the lower flanges were obstructed by core exit thermocouple mixer tubes.

An additional 21 welds were attempted but could not be inspected to the EVT-1 standard; these are recorded as "best effort" examinations. No relevant conditions were identified among these 21 best effort examinations.

The inspected guide tubes are the original tubes. The accumulated EFPY on the guide tubes was 30.04 EFPY. The degradation mechanisms of concern in this inspection are fatigue (high cycle) and SCC. For constant parameters conducive to degradation both these mechanisms would be expected to have observable effects at a relatively early stage in plant life.

22

Surry Unit 2 100% of the accessible The EVT - 1 examination of the one weld received an EVT - 1 upper core barrel weld was SAT; the examination, weld was examined from the ID surface as required per the MRP guidance. There were no indications.

The exam fulfilled the requirements of MRP-227-A.

Comments:

It is not feasible to inspect the OD of this weld when the core barrel is removed because it sits above the water and must be heavily shielded. There were no reportable indications observed on the Upper Core Barrel Flange Weld, but a rub - mark was observed below the Upper Core Barrel Flange Weld in the 900 - 1800 Quadrant.

Examination has not N/A been done yet.

These exams are scheduled for 2R25.

Comments: Scheduled during 21125- Spring 2014.

23

Surry Unit 2 During 2R23 (Spring All 936 accessible edge bolts 2011), 936 accessible received the required VT - 3 baffle - edge bolts examinations. No issues were note; received VT - 3 all examinations were acceptable.

inspection.

Comments: These exams were completed before EPRI had prepared a - 227 reporting template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

24

Surry Unit 2 LJUIIuIn /O kopring i ne most s51n1,u.dnIL xýdIIrnIIdLIUrl 2011), 1088 baffle - to - results were detection of likely flaws former bolts received UT in two bolts, identified as Baffle Plate examination and the 22 Bolt "G63" and Baffle Plate 42 Bolt locking bars received VT "A125". The depth of the flaw in G63

- 3 examinations, and A125 were not quantified; however, each flaw was located at the head to shank region of the bolt.

Also, some channels of the UT signals showed a "back wall" reflection from the end of each bolt, so each bolt is not completely severed. The lock bar for each of these bolts has no relevant conditions and is considered capable of 1performing its retention function.

t.omments: oee aoove.

These exams were completed before EPRI had prepared a MRP - 227 report ing template. EPRI was notified of these data through a detailed narrative summary submitted to the MRP.

The inspection for gaps and distortion due to void swelling has not been done yet. The edge bolt inspection is reported above.

These exams are scheduled for 2R25.

25

Surry Unit 2 Comments: Because Unit 2 has an austenitic (304 - SS) stainless steel hold down spring, measurements of its relaxation was required per the MRP-227-A guidance. Per analysis, considering Surry Unit 2 as-built dimensions and the required hold down force for design conditions, the minimum acceptable height for assuring a minimum for 60 year service life was computed.

Spring height measurements were taken at eight locations (every 450) with three individual measurements taken at each location.

The estimated average spring height was 3.6384+/-0.0001 inches. This result is greater than the minimum requirement; therefore, the final result is acceptable. This result confirms adequate hold down capability through at least 60 total years of reactor operation, and no further measurements are required.

N/A - examination N/A - examination scheduled for scheduled for 2R25 - 2R25 - Spring of 2014.

Spring of 2014.

Notes to Westinghouse Primary Components Table:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3 of MRP-227-A.

2. A minimum of 75% of the total identified sample population must be examined.

26

Surry Unit 2

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3 of MRP-227-A, must be examined from either the inner or outer diameter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

27

Surry Unit 2 V

Expansion Components

.Requied Examination Coverage Examination iteEamiaton etodCoverage Achieved Findings (Note 1)

Upper Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Upper core plate Comments:

Lower Internals Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support forging or castings See Figure 4-33 of MRP-227-A.

Comments:

Core Barrel Volumetric 100% of accessible bolts. N/A N/A Assembly examination (UT) Accessibility may be Barrel-former bolts limited by presence of thermal shields or neutron pads (Note 2).

See Figure 4-23 of MRP-227-A.

28

Surry Unit 2 L.omments:

Lower Support Volumetric 100% of accessible bolts N/A N/A Assembly examination (UT) or as supported by plant-Lower support specific justification (Note column bolts 2).

See Figures 4-32 and 4-33 of MRP-227-A.

Comments:

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Core barrel outlet selected weld and nozzle welds adjacent base metal (Note 2).

See Figure 4-22 of MRP-227-A.

Comments:

a 29

Surry Unit 2 Examination Method Required Exmnto Coverage Examination

........ Item _...Coverage Achieved Findings (Note 1)

Core Barrel Enhanced visual 100% of one side of the N/A N/A Assembly examination (EVT-1) accessible surfaces of the Upper and lower core selected weld and barrel cylinder axial adjacent base metal (Note welds 2).

See Figure 4-22 of MRP-227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) surfaces (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(non cast) 227-A.

Comments:

Lower Support Enhanced visual 100% of accessible N/A N/A Assembly examination (EVT-1) support columns (Note 2).

Lower support column bodies See Figure 4-34 of MRP-(cast) 227-A.

30