Evaluation of Basis for Periodic Visual Examination of

© 2018 Electric Power Research Institute, Inc. All rights reserved.

Heather Feldman, Robert Grizzi, and Chris Joffe, Electric Power Research Institute (EPRI)

Glenn White and Markus Burkardt,Dominion Engineering, Inc.

Industry/NRC Materials Programs Technical Information Exchange Meeting, Rockville, MD

May 21-23, 2018

Evaluation of Basis for Periodic Visual Examination of Accessible

Areas of Reactor Vessel Interior per Examination Category B-N-1

of ASME Section XI, Division 1

2© 2018 Electric Power Research Institute, Inc. All rights reserved.

Objective

In mid-2017, EPRI initiated a project to develop a technical basis for optimization of examination frequency. It included the periodic VT-3 visual examinations of the reactor vessel interior specified for ASME Section XI Examination Category B-N-1– Examination Categories B-N-2 and B-N-3 were not included in the

scope of the EPRI assessment


Status

Technical Report is available:– Evaluation of Basis for Periodic Visual Examination of Accessible

Areas of Reactor Vessel Interior per Examination Category B-N-1 of ASME Section XI, Division 1. EPRI, Palo Alto, CA: 2018. 3002012966.

– https://www.epri.com/#/pages/product/000000003002012966/– Could be used to inform utility-developed relief requests


Implementation within ASME Section XI

At May 10, 2018 meeting, ASME Section XI Standards Committee moved Code Action 10-123 to letter ballotCode Action 10-123 would publish a Code Case N-885

providing alternative requirements for Examination Categories B-N-1, B-N-2, and B-N-3– Removes the B-N-1 examination– Reconsolidates B-N-2 and B-N-3 examinations as category B-NCode Action 10-123 references EPRI 3002012966

The proposed code case provides an alternative in which the B-N-1 examination is not performed.


Approach

Review requirements and purpose for Examination Category B-N-1– Historical editions of Section XI– ASME Code InterpretationsReview current industry guidance for in-vessel examinationsReview current industry guidance for foreign material

exclusion and foreign material searchesReview operating experience


Approach (cont’d)

Assess corrosion and cracking of low-alloy steel (LAS) at locations of cracked, damaged, missing, or removed cladding– The B-N-1 examination is also an opportunity as a general area

examination to detect other relevant conditions in the interior of the reactor vessel

– Some plants perform a VT-3 examination of accessible regions of the reactor vessel interior surfaces (i.e., cladding) as a conservative extension of the B-N-1 scope


Outline of EPRI Technical Basis Document1. Introduction

1.1 Background1.2 Objective1.3 Scope1.4 Approach1.5 Report Organization

2. Review of Current Examination Requirements2.1 Examination Category B-N-1, Item B13.10 Examination

Frequency and Scope2.2 VT-3 Visual Examination2.3 Examination Category B-N-1, Item B13.10 Examination

Relevant Conditions and Acceptance Criteria2.4 Review of Prior Editions of Section XI2.5 Review of ASME Code Interpretations2.6 Industry Guidance2.7 Regulatory Positions2.8 Area Accessible for VT-3 Examinations

3. Review of Operating Experience3.1 Sources3.2 Cladding3.3 Cracking of Underlying Low-Alloy Steel3.4 Corrosion of Underlying Low-Alloy Steel3.5 Effects of Debris and Foreign Objects

4. Assessment of Low-Alloy Steel Corrosion at Cracked, Damaged, Missing, or Removed Cladding4.1 PWRs4.2 BWRs

5. Potential for Low-Alloy Steel Cracking at Cracked, Damaged, Missing, or Removed Cladding5.1 Resistance of Low-Alloy Steel to SCC5.2 Industry Assessments of Reactor Vessel Aging Degradation5.3 Industry Assessment of Reactor Vessel Pressurized

Thermal Shock6. Conclusions and Recommendations

6.1 Conclusions6.2 Recommendations for Alternative Examination Requirements

7. References


Industry Guidance and Practices to Address Debris in RV Preventative practices following Foreign Material Exclusion programs and

procedures exist at every plant– Example industry guidance documents published by EPRI and INPO include

EPRI 3002003060, EPRI 1024917, and EPRI 1026776 In practice, routine activities address foreign materials or debris within the

reactor vessel:– Loose or missing parts and debris located above the reactor core tend to accumulate

on top of the fuel, which is observed during the Core Verification activities performed at the end of refueling outages

– Foreign object search and retrieval (FOSAR) is typically performed either prior to fuel movement (reload) or reactor reassembly

– Foreign objects and debris are often identified during maintenance activities and other examinations, including examinations in BWRs of shroud support plate welds H-8 and H-9, as well as jet pump adapter AD-2 welds on the shroud support plate

These activities are performed more frequently than once each Section XI inspection period


Operating Experience and Analyses Debris and foreign material have been identified as leading contributors to

fuel rod failures– FME guidance and work practices have been developed and implemented to

address the concern Cracked, damaged, missing, or intentionally removed cladding has been

found to be inconsequential, with no significant detrimental effects on the underlying LAS:– Severe corrosion rates due to concentrated boric acid or ongoing air in-leakage

are not credible for PWR reactor vessel interior– General corrosion in BWRs is a low concern in both normal water chemistry and

hydrogen water chemistry environments– Observed cracking that propagated through the cladding over long periods to the

underlying LAS has typically arrested upon reaching the LAS– Analyses of corrosion and service-induced fatigue cracking of underlying LAS are

consistent with the favorable operating experience


Relevance to Potential Reactor Vessel DegradationDetection of Adverse Conditions Associated with Cracked, Damaged, or Missing Cladding Numerous examinations and activities other than Examination Category B-N-1

VT-3 are performed in the reactor vessel during refueling outages, such as:– PWR MRP-227 internals examinations (in period of extended operation)– BWRVIP examinations– Other Section XI examinations (B-A, B-D, B-N-2, and B-N-3)– Examinations of reactor vessels performed per 10 CFR 50.61a (the Alternate PTS Rule)– FOSAR– Core Verification activities

When these examinations are performed, there are opportunities for detecting adverse conditions associated with cracked, damaged, missing, or intentionally removed cladding– Includes evidence of damaged cladding (including due to impact, wear, or fretting), LAS

corrosion, or cracking penetrating the cladding (as evidenced by rust bleed-out)– Accessibility to cladding is limited during normal refueling outages, especially for PWRs


Conclusions The purpose of the B-N-1 VT-3 examination is to detect foreign material and

debris in the interior of the reactor vessel The following examinations and maintenance activities are in place to detect

foreign material and debris:– Foreign material and debris examinations are not directly related to vessel integrity and are

routinely performed during Core Verification, FOSAR, and other maintenance activities– EPRI and INPO have published guidance on appropriate practices for detection and removal of

foreign material and debris, including from the interior of the reactor vessel, for example: EPRI 3002003060, EPRI 1024917, and EPRI 1026776

Plant experience and analyses show that LAS degradation at cracked, damaged, or missing cladding is of low concern Other activities within the reactor vessel are opportunities to detect relevant

conditions affecting the cladding, for example:– PWR internals examinations (after period of extended operation), BWRVIP examinations,

examinations of reactor vessels performed per 10 CFR 50.61a, and other Section XI examinations such as Examination Categories B-A, B-D, B-N-2, and B-N-3


Together…Shaping the Future of Electricity


Backup Slides


Industry Guidance and Practices to Address Debris in RV

Debris and foreign material have:– been identified as the leading contributors to fuel rod failures– the potential to cause blockage of flow through the core or blockage of control rods– led to other equipment degradation or inoperability, lost generation, and spread of

high radiation and contamination levels throughout the plant

Industry has developed practices:– to reduce the amount of foreign objects or debris that may be introduced in the

reactor coolant system as a result of human error– for detection and removal of foreign material and debris, including from the RV

interior


Relevance to Potential Reactor Vessel DegradationPotential for LAS Corrosion – Operating Experience

Severe corrosion rates resulting from boric acid corrosion in PWRs can only be caused by conditions not credible for the PWR reactor vessel interior:– Air in-leakage or concentration of boric acid (associated with pressure

boundary leakage or evaporation of a limited volume of liquid)General corrosion in BWRs is of very low concern in both normal

water chemistry and hydrogen water chemistry environmentsLocalized corrosion effects are not expected to be a significant

concern based on operating experience and the likely range of applicable creviced locations


Relevance to Potential Reactor Vessel DegradationPotential for LAS Corrosion – Analysis

The potential for wall thickness reduction due to corrosion of the underlying LAS material was conservatively evaluated considering applicable environments and laboratory data– Separate assessments performed for PWRs and BWRsAny reduction in thickness is expected to be acceptably small

with regard to vessel structural integrity– Based on Section III NB-3200 design rules


Relevance to Potential Reactor Vessel DegradationPotential for LAS CrackingOperating experience:

– Cracking penetrating the cladding has been occasionally detected via rust bleed-out in B-N-1 and other visual examinations

– Cracking that has propagated through the cladding over long periods to the underlying LAS has typically arrested upon reaching the LAS

Analyses:– No structural credit is given to cladding– Analyses show insignificant growth of postulated surface flaws into the

underlying LAS due to service-induced fatigue Consistent with the operating experience

– Reactor vessel integrity assessments, which model the potential for shallow surface cracks in the cladding to result in brittle fracture, do not explicitly credit periodic visual examinations of the cladding


Relevance to Potential Reactor Vessel DegradationAccessibility of Cladding

For PWRs:– Accessibility to cladding is very limited during normal refueling outages– Cladding becomes fully accessible during RV-ISI outages required

nominally every 10 or 20 yearsFor BWRs:

– The steam dryer and moisture separator / shroud head assembly are removed during normal refueling outages, enabling access to the area above the top guide as well as in the annulus area between the shroud and the vessel wall from the top guide down to the shroud support plate

– With NRC approval through the relief request process, the majority of U.S. BWRs credit the numerous and frequent activities within the vessel associated with BWRVIP examinations in lieu of the B-N-1 examination

Documents

Evaluation of Basis for Periodic Visual Examination of