RESULTS FROM NONDESTRUCTIVE EXAMINATION OF PWR VESSEL INTERNALS. Jack Spanner, EPRI, USA

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1 More Info at Open Access Database Introduction RESULTS FROM NONDESTRUCTIVE EXAMINATION OF PWR VESSEL INTERNALS Jack Spanner, EPRI, USA This paper describes the initial results obtained to date from examining pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection and evaluation guidelines. These examinations have been performed at seven nuclear plants. The objective of this aging management program for internals is to provide utilities a systematic approach to monitor potential degradation of internals components, support power up-rate activities, and meet license renewal requirements. An accompanying inspection standard for PWR internals is implemented with the guideline. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE systems used to perform those inspections. The Electric Power Research Institute s Materials Reliability Program (MRP) has developed an Inspection and Evaluation (I&E) Guideline (MRP-227-A) [1] to meet the requirements for aging management of pressurized water reactor vessel internals and license renewal activities. To support the guidelines MRP has also developed an Inspection Standard (MRP-228, Rev 1) [2] that provides the inspection requirements. Generally, visual examinations will be used to detect evidence of age related degradation mechanisms for all components, except ultrasonics are used to volumetrically examine some of the bolts. During development of the I&E Guidelines several components were identified as being potential areas to consider for inspection. Most of these components are not typically included in a nuclear in-service inspection program with the pressure boundary components. MRP-227 addresses the following damage mechanisms: Stress Corrosion Cracking (Cracking) Irradiation-Assisted Stress Corrosion Cracking (Cracking) Wear (Loss of Material) Fatigue (Cracking) Thermal Aging Embrittlement (Loss of Toughness > Cracking) Irradiation Embrittlement (Loss of Toughness > Cracking) Void Swelling and Irradiation Growth (Dimensional Change/Distortion > Cracking) Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation- Enhanced Creep (Loss of Preload > Wear/Cracking) NDE System Qualification The process for NDE system qualification described herein is based on the use of technical justifications as described in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section V, Article 14 [3] and is similar to the original version of the European Network for Inspection Qualification (ENIQ). The Technical Justification is a detailed explanation of the procedure including the method, and any laboratory or field experience that support the procedure capabilities. The Technical Justification provides the technical basis and rationale for the qualification. This qualification process builds on the field experience and previous work performed by the inspection vendors and other inspection organizations to develop the examination systems already in use today throughout industry and does not mandate any new performance demonstrations. Where there is a need to simulate performance specific examinations for the purpose of preparing the Technical Justifications required by the Inspection Standard, or for other reasons, such as to qualify improved techniques, new or existing vendor mockups may be used or, where mockups of such components are not otherwise available, mockups developed by the MRP may be used. The Technical Justification is a required written report providing the detailed explanation of the examination process, including the theory of the examination technique as applied to reactor internals inspections, the essential variables of the procedure, other influential parameters important to the overall performance of the examination system, and field experience and/or mockup demonstrations supporting the capabilities of the NDE system. 1044

2 The Technical Justification includes a description of the active or anticipated material damage mechanisms and the NDE evaluation process used to interpret the results of the applied examination technique. Prior to use of an NDE system for reactor internals inspections, the Technical Justification report shall be reviewed and approved by the utility. Visual Inspection of Reactor Pressure Vessel (RPV) Internals, Components The Inspection Standard is to be used by pressurized water reactor utilities when performing visual examination (Enhanced VT-1 (EVT-1), detailed VT-1 and general VT-3) of reactor vessel (RPV) internals, components, and associated repairs to meet the recommendations set forth in applicable Materials Reliability Program documents. As used in this paper, the EVT-1 and VT-1 techniques are used for detecting surface imperfections and flaws such as cracks, wear, erosion, and corrosion. Specifically, EVT-1 is used for the detection of surface breaking flaws, whereas VT-1 is used to detect surface conditions such as gaps. VT-3 is used for detecting general degradation conditions. ASME Section XI [4] describes VT-1 and VT-3 while MRP-228 provides the requirements for EVT-1. The EVT-1 and VT-1 need to resolve the same.044 inch (1 mm) characters. However, the EVT-1 requires.5 inch/sec (13 mm/sec) maximum scan speed, cleaning assessment and a viewing angle within 30 degrees perpendicular to the surface. Utilities shall also conduct site specific training for all personnel evaluating inspection data. The training shall include utility-specific procedural requirements, configuration details, previous inspection results, operation of inspection equipment, specific outage inspection scope, and any other pertinent information related to inspection, evaluation and reporting, as applicable. The training is to be conducted prior to inspections for each refueling outage. Personnel evaluating inspection data shall have a minimum of ten (10) hours of work time experience performing EVT- 1, VT-1 and/or VT-3 remote visual inspections of reactor internals Ultrasonic Examination of Bolting in Reactor Pressure Vessel Internals This section describes requirements and recommendations for the performance of ultrasonic examination of bolts in reactor vessel internal components. UT Technical Justifications and Demonstrations Ultrasonic examination systems are required to be qualified by a Technical Justification (TJ) as described earlier. The organization that prepared the procedure will also be responsible for preparing the Technical Justification. The utilities are responsible for approving the Technical Justification. Equipment Requirements The equipment included in the UT procedure as an essential variable or range of variables as described in this section must be used for the examinations and be included in the technical justification and demonstration, as applicable (Figure 1). Examination coverage should be reported as the number or percentage of the bolts that were successfully examined out of the population of bolts that were scheduled for examination. Figure 1 UT Inspection of B-F Bolt Mockup 1045

3 Application The accompanying Inspection Standard is intended for the use of individual plant owners in preparing inspection procedures and qualifying NDE systems used to perform inspections needed for their PWR internals aging management programs. As a companion to the I & E Guidelines of MRP-227, this standard is required for compliance with one of the mandatory requirements of MRP-227. These requirements are applicable to internals in the three Nuclear Steam Supply System pressurized water reactor designs currently operating in the United States: Babcock & Wilcox (B&W), Combustion Engineering (CE), and Westinghouse Nuclear, respectively. PWR Internals Inspection Results Seven PWR units have completed or partially completed inspections of their internals in accordance with MRP-227. Some owners decided to perform the inspections during one, two or three refueling outages so some of the owners that decided to inspect over multiple outages have not completed their inspections yet. The inspections so far have essentially met the minimum extent of inspections with very few detections of flawed components. Almost all of the detections have been located in bolting, including the locking devices such as locking bars and welds, The following list summarizes the results of the inspections conducted to date: Unit 1: o Each CRGT Guide Card is acceptable until the next scheduled inspection required by MRP-227-A based upon wear projections performed o Observed volumetric Guide Card wear ranging from 21%-51%. Unit 2 : o Baffle to Former Bolts Replace or UT of plant specific minimum bolt pattern or alternate bolts as selected by analysis; 1 bolt (or 1%) identified with a relevant UT indication o Guide Card Wear 100% VT-3 Inspection; no appreciable wear seen o CRGT Flange Welds 100% EVT-1 Inspection; no relevant indications identified 1046

4 o Core Barrel Upper Flange Weld 100% EVT-1 Inspection; no relevant indications identified o Thermal Shield Flexures 100% VT-3; no relevant conditions o Core Barrel Baffle Edge Bolts & Seams 100% VT-3 (all 16 high fluence seams for full length); no relevant conditions o Baffle Former Assembly 100% VT-3; no relevant conditions Unit 3: o Only one UT detection of a likely flaw in one bolt. o VT-3 identified two bolts deformed at the hex head points o VT-3 also identified one baffle to former bolt with one missing locking bar weld o VT-3 also identified one edge bolt with one missing locking bar weld Unit 4: o All 1088 baffle to former bolts were UT and VT-3 examined. All accessible 936 baffle edge bolts were VT-3 examined. o Baffle former bolt UT detected indications in two bolts located at the head to shank region of the bolt. o With only two assumed failed bolts out of 1088, there is significant margin above the industry evaluation guideline for reactor internals Unit 5: o No relevant indications Unit 6: o UT 5 lower core barrel (LCB) with crack like indications o VT 1 LCB locking device with a missing weld on one side and a small weld on the other. o The coverage for the VT-3 examination of 7 of the LCB bolts and their locking devices was 40% to 50% due to the partial obstruction from the stand, although 100% of the accessible surfaces were examined. o UT 1 flow distributor bolt with crack like indication. o 860 out of 864 baffle-to-former bolts o 4 baffle-to-former bolts were un-inspectable due to large welds on locking bars. 1047

5 Unit 7 o Guide Cards - 93 of 297 guide cards with recordable indications (wear) were identified. o Based on Operation Curve Method 2 CRGTs yellow in ~ 7 EFPY, 3 CRGTs yellow in 17 to 20 EFPY, and 28 yellow in 20+ EFPY o Based on Operation Curve Method 1 CRGT red in ~ 7 EFPY and 32 CRGTs red in 20+ EFPY o CRGT Flange Welds o 110 of 118 (93%) accessible welds in 24 Peripheral CRGTs References Note: Missed welds attributed to interference from RV Flange Protector Ring o CB Lower Girth Weld o (Covered by thermal shield) o 78% of exterior surface o Weld features not visible o CB Lower Flange Weld o 92% of exterior surface (RV Flange Protector Ring interference) o Weld features not visible 1. Materials Reliability Program: PWR Internals Inspection and Evaluation Guidelines (MRP- 227-A), Electric Power Research Institute, Palo Alto, CA, Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228, Rev 1), EPRI, Palo Alto, CA ASME Boiler and Pressure Vessel Code, Section V, Nondestructive Examination, American Society of Mechanical Engineers, New York, NY, 2007 Edition through 2008 Addenda 4. ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection, American Society of Mechanical Engineers, New York, NY, 2004 Edition through 2008 Addenda 1048