View source for L-17-183, 10 CFR 50.55a Request to Apply Alternative Examination Requirements to Certain ASME Section Xl Class 1 Components (Request IR-056. Revision 2)

{{Adams
| number = ML17272A093
| issue date = 09/29/2017
| title = 10 CFR 50.55a Request to Apply Alternative Examination Requirements to Certain ASME Section Xl Class 1 Components (Request IR-056. Revision 2)
| author name = Hamilton D
| author affiliation = FirstEnergy Nuclear Operating Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000440
| license number = NPF-058
| contact person = 
| case reference number = L-17-183
| document type = Letter type:L
| page count = 19
}}

=Text=
{{#Wiki_filter:FENOC Perry Nuclear Power Plant P.O. Box 97 10 Center Road
,- i(-     a/  /   ^7*   *  ^                                                                    Perry, Ohio 44081 FirstEnergy Nuclear Operating Company                                                                   y David B. Hamilton                                                                                      440-280-5382 Vice President September 29, 2017 L-17-183                                                               10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

==SUBJECT:==

Perry Nuclear Power Plant Docket No. 50-440, License No. NPF-58 10 CFR 50.55a Request to Apply Alternative Examination Requirements to Certain ASME Section XL Class 1 Components (Request IR-056. Revision 2)
Pursuant to 10 CFR 50.55a(z)(1), FirstEnergy Nuclear Operating Company (FENOC) requests Nuclear Regulatory Commission approval of a request to apply alternative examination methods to certain American Society of Mechanical Engineers Section XI, Class 1 components at the Perry Nuclear Power Plant (PNPP).
The enclosed proposed alternative would be implemented during the PNPP third inservice inspection interval. FENOC requests approval of the alternative by October 31, 2018 to support the PNPP spring 2019 refueling outage.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at (330) 315-6810.

==Enclosure:==

Perry Nuclear Power Plant, 10 CFR 50.55a Request IR-056, Revision 2 cc:    NRC Region III Administrator NRC Resident Inspector NRC Project Manager

Enclosure L-17-183 Perry Nuclear Power Plant, 10 CFR 50.55a Request IR-056, Revision 2 (17 pages follow)

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-056, Revision 2 Page 1 of 12 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)
            -Alternative Provides Acceptable Level of Quality and Safety-
: 1. American Society of Mechanical Engineers (ASME) Code Components Affected ASME Boiler and Pressure Vessel Code (ASME Code) Section XI, Class 1, Examination Category B-N-1 (Interior of Reactor Vessel), and B-N-2 (Welded Core Support Structures and Interior Attachments to Reactor Vessels), Item Numbers:
* B13.10-Vessel Interior
* B13.20 - Vessel Interior Attachments within Beltline Region
* B13.30 - Interior Attachments beyond Beltline Region
* B13.40 - Core Support Structure Table 1 provides a detailed list of components associated with each item number.

===2. Applicable Code Edition and Addenda===
ASME Code, Section XI, 2001 Edition through the 2003 Addenda.
: 3. Applicable Code Requirements ASME Code, Section XI, Paragraph IWB-2500(a) states in part that:
Components shall be examined and tested as specified in Table IWB-2500-1
["Examination Categories"].
Table IWB-2500-1 specifies, in part, the following visual examinations.
B13.10    Examine accessible areas of the reactor vessel interior each inspection period by the VT-3 method (B-N-1).
B13.20    Examine accessible interior attachment welds within the beltline region each inspection interval by the VT-1 method (B-N-2).
B13.30    Examine accessible interior attachment welds beyond the beltline region each inspection interval by the VT-3 method (B-N-2).
B13.40    Examine accessible surfaces of the core support structure each inspection interval by the VT-3 method (B-N-2).

Perry Nuclear Power Plant IR-056, Revision 2 Page 2 of 12
: 4. Reason for Request A wealth of inspection data has been gathered during inspections across the BWR industry. Based on the gathered inspection data, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) has developed Inspection and Evaluation (I&E) guidelines and has recommended aggressive specific inspection by Boiling Water Reactor (BWR) operators to completely identify material condition issues with BWR components. The I&E guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanisms, and require re-examination at conservative intervals. In contrast, the ASME Code inspection requirements were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience.
As an alternative to the ASME Code inspection requirements, use of BWRVIP I&E inspection guidelines will avoid unnecessary inspections, while reducing radiological dose.
: 5. Proposed Alternative and Basis for Use FirstEnergy Nuclear Operating Company (FENOC) proposes to apply the BWRVIP I&E guidelines listed below to affected ASME Code components identified in Table 1, in lieu of the requirements of ASME Code, Section XI, Paragraph IWB-2500(a) and Table IWB-2500-1, including the examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting.
Not all of the components addressed by these guidelines are ASME Code components. The particular guidelines that are applicable to the subject ASME Code components are:
BWRVIP-03, "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines" BWRVIP-18, Revision 2-A, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines" BWRVIP-26-A, "BWR Top Guide Inspection and Flaw Evaluation Guidelines" BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines" BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines" BWRVIP-41, Revision 3, "BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines"

Perry Nuclear Power Plant IR-056, Revision 2 Page 3 of 12 BWRVIP-42, Revision 1," BWR Vessel and Internals Project, Low Pressure Coolant Injection (LPCI) Coupling Inspection and Flaw Evaluation Guidelines" BWRVIP-47-A, "BWR Lower Plenum Inspection and Flaw Evaluation Guidelines" BWRVIP-48-A, "Vessel ID [Internal Diameter] Attachment Weld Inspection and Flaw Evaluation Guidelines" BWRVIP-76, Revision 1-A, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" (see Note)
BWRVIP-94NP, Revision 2, "BWR Vessel and Internals Project Program Implementation Guide" BWRVIP-100, Revision 1-A, "Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds" BWRVIP-138, Revision 1-A, "Updated Jet Pump Beam Inspection and Flaw Evaluation" BWRVIP-180, "BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines" BWRVIP-183-A, "Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines" Note: If flaw evaluations are required for BWRVIP-76 examinations, the fracture toughness values of BWRVIP-100 will be utilized.
The BWRVIP executive committee periodically revises the BWRVIP Guidelines to include enhancements in inspection techniques and flaw evaluation methodologies.
Where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline that forms the safety basis for the NRC-authorized proposed alternative to the requirements of 10 CFR 50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for relief has been approved.
Any deviations from the referenced BWRVIP guidelines for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process. Currently, PNPP does not have any deviations from the subject guidelines.
Implementation of the proposed alternative actions of this relief request will be subject to inspection by an Authorized Inspection Agency.
The BWRVIP provides BWR Vessel and Internals Inspection Summaries to the NRC periodically. Reference 19 is the BWR Vessel and Internals Inspection Summary transmitted to the NRC that includes PNPP. This summary provides, on a

Perry Nuclear Power Plant IR-056, Revision 2 Page 4 of 12 component-by-component basis, the inspection methods utilized, the inspection dates, and the results of the inspections through the spring 2015 outage. This summary also contains the identified corrective actions. Corrective actions and inspections performed prior to the BWRVIP were implemented to the requirements of ASME Section XI, as applicable. The inspection summary that includes the information for the latest PNPP outage in spring 2017 has not yet been assembled and transmitted to the NRC by the BWRVIP.
BWRVIP guidelines have been written to examine the safety significant reactor vessel internal components and evaluate the examination results for these components using appropriate methods and re-examination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations.
The NRC has agreed with the BWRVIP approach in principle and has issued safety evaluations for these guidelines (see References 4 through 18 below). A comparison of ASME Code examination requirements and BWRVIP examination requirements is provided in Attachment 1.
In support of this request, the following information addresses the inspection of furnace-sensitized stainless steel and Alloy 182 welds, specific areas subject to nozzle cracking, and hydrogen water chemistry (HWC) effectiveness.
Furnace-sensitized stainless steel and Alloy 182 welds:
Furnace-sensitized stainless steel vessel attachment welds are inspected as required by ASME Section XI and BWRVIP applicable guidelines. The sensitization status of the steam dryer and the feedwater support brackets has not been determined, and as such, they are assumed to be furnace-sensitized.
With regard to examination category B-N-1, there are no Alloy 182 welds. With regard to examination category B-N-2, the following locations have Alloy 182 welds:
* Shroud support (Weld H9)
* Shroud support legs (Weld H12)
* Access hole cover The shroud support and shroud support leg inspections are specified in BWRVIP-38 and the access hole cover inspection is specified in BWRVIP-180.
No additional augmented inspections are performed on the Alloy 182 welds outside of that defined in BWRVIP-38 and BWRVIP-180. There has been no cracking identified in the Alloy 182 welds at PNPP.
Feedwater nozzle and control rod drive (CRD) return line nozzle:
NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking: Resolution of Generic Technical Activity A-10 (Technical Report)," is implemented in both the inservice inspection (ISI) and BWRVIP

Perry Nuclear Power Plant IR-056, Revision 2 Page 5 of 12 programs as augmented inspections. In regard to the ISI program, the requirements are limited to the six (6) feedwater nozzle inner radii. The CRD return line at the PNPP has been cut and capped and, as such, is not subject to these examinations. The feedwater sparger flow holes and welds in the tees and arms are examined per NUREG-0619. Vessel attachment welds are examined by utilizing the requirements of BWRVIP-48-A.
Effectiveness of hvdroqen water chemistry (HWC) in coniunction with noble metal chemical addition (NMCA):
The PNPP utilizes the General Electric (GE) Online NobleChem' (OLNC) process as the current means of NMCA. This process replaced the previous NobleChem process and is used in conjunction with HWC for the purpose of mitigating intergranular stress corrosion cracking (IGSCC) in the reactor internals. IGSCC is considered mitigated when the local electrochemical corrosion potential (ECP) is reduced below -230 millivolts (mV) standard hydrogen electrode (SHE).
The ECP for Cycle 16 at PNPP is as follows:
ECP for Cycle 16 600
                                                                                                                    ; 3/4/17 22*6, 413.9 Shutdown for 1R16 l 1/13/1611:03, 339.5
                                             ; Hydrogen Benchmark
                                             ;          Test 400
                                                                           ! 2/13/16 13:55, 329.9       2/28/1719:40,140.9!
                                                                           '*    Startup from           {      HWC Trip        j 4/24/15 7:45, 145.5                                         Forced Outage StartupfromlRlS                                                               1 2/25/17 7:39, 126.9 ;
j      HWC Trip 200
              -2G0 2/1/1612:48, -444.4 Startup from
                                        ;   Forced Outage j 1/23/16 22:04. -492.9 i i 2/8/16 19:22, -407.2 ;
              -400               :         Scram           \
Scram L.,                . I
                                                                    ^^'l"'1"           ' f     * ~
               -5O0 J>

Perry Nuclear Power Plant IR-056, Revision 2 Page 6 of 12 The ECP for Cycle 17 (the current cycle of operation) at PNPP is as follows:
ECP for Cycle 17 300.0 j 7/19/17 18:21, 1*4.8 4/3/17 10:3*. 211.8                           I      HWCTrip 203.0        Startup from 1R16 100.0
               -2D0.G l 5/21/17 14:45, -537.S I RWCUB pump 005,
               -303.0 i no ECP measurement.
i   7/22/17 9:30,-405.2   j j     ECP remains j Start of RWCU A Pump \
I ~-5Q0mV(SHE)with                         i After Pump Repiacefnenti i hydrogen in service.
               -4OD.0
               -5GG.0 !
               -soc.o '
3/15/2C17 4/14/2017         5/16/2017         6/13/2017 7/13/2017          8/12/2017       9/11/2017 The molar ratio value at PNPP is calculated using BWRVIP-202, "BWR Vessel and Internals Application (BWRVIA) for Radiolysis and ECP Analysis Version 3.1." This is an effective tool for estimation of ECP at specific locations and for demonstrating that sufficient hydrogen is being injected to maintain the molar ratio of hydrogen to oxidants greater than 2.                     Three values are developed at the beginning of an operating cycle: BOC (beginning of cycle), MOC (middle of cycle), and EOC (end of cycle). The values selected are from the upper downcomer location, which is considered the most conservative location by the BWRVIP.

Perry Nuclear Power Plant IR-056, Revision 2 Page 7 of 12 As projected by the BWRVIA model at downcomer, S1 carryunder, the Cycle 16 molar ratio values applicable to the PNPP are:
* BOC: 5.2
* MOC: 5.2
* End of reactivity (before coast down starts): 4.7
* EOC: (coast down): 2.3 As projected by the BWRVIA model at downcomer, S1 carryunder, the Cycle 17 molar ratio values applicable to the PNPP are:
* BOC: 5.2
* MOC: 5.1
* End of reactivity (before coast down starts): 4.8
* EOC: (coast down): 3.1 The needed catalyst loading is > 0.1 ug/cm2. The catalyst loading for the previous two operating cycles at PNPP was as follows:
* Cycle 15 average catalyst loading from in-vessel artifact scraping was 0.85 ug/cm2.
* Cycle 16 average catalyst loading from in-vessel artifact scraping was 0.97 ug/cm2.
HWC availability for the prior and current PNPP operating cycles is as follows:
* Cycle 16 HWC availability was 99.4 percent.
* Cycle 17 HWC availability is currently at 99.4 percent through the end of August 2017.
Summary The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than "all surfaces." The BWRVIP examination methods (including an enhanced visual examination VT-1
[EVT-1] or ultrasonic test [UT]) are superior to the ASME Code required VT-3 for flaw detection and characterization. In most cases, the BWRVIP examination frequency is equivalent to or more frequent than the examination frequency required by the ASME Code. In cases where the BWRVIP examination frequency is less frequent than required by the ASME Code, the BWRVIP examinations are performed in a more comprehensive manner and focus on the areas most vulnerable. Therefore, the superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency, or with a less frequent examination frequency but with those examinations being performed in a more comprehensive manner, and using comparable flaw evaluation criteria, results in the

Perry Nuclear Power Plant IR-056, Revision 2 Page 8 of 12 BWRVIP criteria providing a level of quality and safety that is equivalent or superior to that provided by the ASME Code requirements.
Therefore, use of these BWRVIP guidelines as an alternative to the ASME Code requirements provides an acceptable level of quality and safety.
: 6. Duration of Proposed Alternative The alternative is requested for the balance of the third 10-year inservice inspection interval (which will expire May 17, 2019).
: 7. Precedents Revision 1 of this relief request, which requested similar relief for Perry's third 10-year inspection interval, was approved by the NRC (Reference 1). Revision 2 of this relief request is seeking the approval of the use of additional BWRVIP guidelines, as well as the approval of the use of subsequently revised guidelines.
Similar relief requests throughout the industry have been submitted and approved by the NRC. An example is the request for the Clinton Power Station (CPS), which was approved by the NRC on March 10, 2016 (References 2 and 3). Both CPS and PNPP are General Electric Company BWR plants with design designated as BWR/6.
Differences between the CPS relief request (Reference 2) and the proposed relief request for PNPP are explained as follows.
* Some of the revision levels and approvals designated for the BWRVIP guidelines in the PNPP request are different, specifically BWRVIP-18, -76, and -183. This is attributed to the fact that these guidelines have since been revised or were approved (reflecting the -A designation) or both.
* The PNPP request identifies BWRVIP-100, whereas the CPS request does not. The PNPP list on page 3 identifies BWRVIP-100 since BWRVIP-76 references BWRVIP-100 for flaw evaluations.
* Some of the component information provided in Table 1 of the PNPP request differs from the information provided in the CPS request. These differences are due to plant design differences and the way in which components are considered (for example, component considered part of the welded core shroud versus a beltline weld).
The differences cited above are considered minor and do not invalidate the CPS example as a similar precedent.

Perry Nuclear Power Plant IR-056, Revision 2 Page 9 of 12
: 8. References
: 1. Letter NRC to FirstEnergy Nuclear Operating Company, "Perry Nuclear Power Plant, Unit No. 1, RE: Safety Evaluation in Support of 10 CFR 50.55a Requests for the Third 10-Year In-service Inspection Interval (TAC Nos. ME5373, ME5376, ME5377, ME5379, and ME5380)," dated January 31, 2012 (ADAMS Accession Number ML120180372).
: 2. Letter Exelon Generation Company to NRC, "Submittal of Relief Requests I3R-1O,1ISI-OO4, and 2ISI-013 Concerning Use of the BWRVIP Guideline in Lieu of Specific ASME Code Requirements," dated April 10, 2015 (ADAMS Accession No. ML15100A228).
: 3. Letter NRC to Exelon Generation Company, "Clinton Power Station, Unit 1 - Use of BWRVIP Guidelines in Lieu of Specific ASME Code Requirements (CAC No.
MF6115)," dated March 10, 2016 (ADAMS Accession No. ML16012A344).
: 4. Letter NRC to BWRVIP, "Final Safety Evaluation for Electric Power Research Institute Topical Report 'BWRVIP-18, Revision 2: Boiling Water Reactor Vessel and Internals Project, Boiling Water Reactor Vessel Core Spray Internals Inspection and Flaw Evaluation Guidelines (TAC No. MF8809),'" dated February 22, 2016 (ADAMS Accession No. ML16011A199).
: 5. Letter NRC to BWRVIP, "Final Safety Evaluation of BWRVIP Vessel and Internals Project, BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines (BWRVIP-25)", EPRI Report TR-107284, December 1996 (TAC NO. M97802), dated December 19, 1999.
: 6. Letter NRC to BWRVIP, "NRC Approval Letter of BWRVIP-26-A, 'BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines,1" dated September 9, 2005.
: 7. Letter NRC to BWRVIP, "Proprietary version of NRC Staff Review of BWRVIP-27-A, 'BWR Standby Liquid Control System/Core Plate AP Inspection and Flaw Evaluation Guidelines,"1 dated June 10, 2004.
: 8. Letter NRC to BWRVIP, "Final Safety Evaluation of the 'BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38),' EPRI Report TR-108823 (TAC NO. M99638)" dated July 24, 2000.
: 9. "BWRVIP-41, Revision 3, BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines," EPRI Technical Report 1021000, dated September, 2010.

Perry Nuclear Power Plant IR-056, Revision 2 Page 10 of 12 1O."BWRVIP-42, Revision 1, BWR Vessel and Internals Project, LPCI Coupling Inspection and Flaw Evaluation Guidelines," dated June, 2010.
: 11. Letter NRC to BWRVIP, "NRC Approval Letter of BWRVIP-47-A, 'BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines,1" dated September 9, 2005.
: 12. Letter NRC to BWRVIP, "NRC Approval Letter of BWRVIP-48-A, 'BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines,'" dated July 25, 2005.
: 13. Letter NRC to BWRVIP, "Final Safety Evaluations of the Boiling Water Reactor Vessel and Internals Project 76, Rev. 1-A Topical Report, Boiling Water Reactor Core Shroud Inspection and Flaw Evaluation Guidelines (TAC No. ME8317),"
dated November 12, 2014.
: 14. Letter from Chairman, BWR Vessel and Internals Project to NRC, "Project No.
704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2),"
dated September 22, 2011 (ADAMS Accession No. ML11271A058).
: 15. Letter NRC to BWRVIP, "Final Proprietary Safety Evaluation for Electric Power Research Institute Topical Report BWRVIP-100, Revision 1, "BWRVIP Vessel and Internals Project: Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds" (TAC No. ME8329)," dated April 12,2016.
: 16. Letter NRC to BWRVIP, "Electric Power Research Institute Final Safety Evaluation for Technical Report 1016574 "BWRVIP-138, Revision 1-A: BWR
[Boiling Water Reactor] Vessel and Internals Project 'Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines'" (TAC No. ME2191)," dated May 14, 2012.
17."BWRVIP-180: BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines," dated November 2007.
: 18. NRC to BWRVIP, "Final Safety Evaluation for Electric Power Research Institute Topical Report BWRVIP-183-A, "BWR Vessel and Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines'" (TAC No. ME2178),"
dated December 31, 2015.
: 19. Letter 2016-059 from BWRVIP to NRC, "Project No. 704 - BWR Vessel and Internals Inspection Summaries for Spring 2015 Outages." Dated May 23, 2016 (ADAMS Accession No. ML16152A162).

Perry Nuclear Power Plant 10 CFR 50.55a Request IR-056, Revision 2 Page 11 of 12 TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements for BWR/6 <1>
ASME Item No.
ASME Exam      ASME    ASME         Applicable BWRVIP           BWRVIP Exam              BWRVIP Table          Components                                                                                                                     BWRVIP Frequency Scope      Exam  Frequency           Document                   Scope                  Exam IWB-2500-1 Accessible Each    BWRVIP-18-2-A, 26-A, 38, 41, Overview examinations of components during BWRVIP examinations are B13.10     Reactor Vessel Interior     Areas       VT-3 period      42, 47-A, 48-A, 76-1-A             performed to satisfy Code VT-3 inspection requirements.
(Non-specific)
Jet pump riser brace Jet Pump Riser Brace                                                                                              EVT-1              25% every 6 years Each                                     bracket welds Accessible                            BWRVIP-48-A B13.20                                             VT-1  10-year RPV Surveillance        welds                                 Table 3.2          Surveillance sample interval                                                              VT-1            Each 10-Year interval Sample Holder                                                                   holder bracket welds Core spray primary Core Spray piping                                                                                                              100% every four refueling and supplemental            EVT-1 bracket                                                                                                                             cycles bracket Feedwater sparger Feedwater sparger bracket attachment           EVT-1            Each 10-Year interval bracket                                 Each Accessible                           BWRVIP-48-A                   welds B13.30                                             VT-3  10-year Guide Rod support        welds                                 Table 3.2            Guide rod support interval                                                              VT-3            Each 10-Year interval bracket                                                                          bracket welds Steam Dryer hold down                                                                  Steam Dryer hold VT-3            Each 10-Year interval bracket                                                                      down bracket welds Steam Dryer support                                                                Steam Dryer support EVT-1            Each 10-Year interval bracket                                                                          bracket welds

Perry Nuclear Power Plant IR-056, Revision 2 Page 12 of 12 TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements with BWRVIP Guidance Requirements for BWR/6(1)
ASME Item No.
ASME Exam         ASME          ASME            Applicable BWRVIP                 BWRVIP Exam             BWRVIP Table              Components                                                                                                                                               BWRVIP Frequency Scope          Exam      Frequency                Document                         Scope                Exam IWB-2500-1 BWRVIP-38,                                                           Based on as-found conditions, Accessible                                                                     Welds H8 and H9(2)       EVT-1 or UT Shroud Support Plate                                                                  3.2.2,                                                           to a maximum 6 years for one Surfaces Figures 3-4, 3-5                                                        side EVT-1,10 years for UT Accessible                     Each                                                                    Per BWRVIP-Surfaces        VT-3        10-year                                                                  38 NRC SER (beneath core                  Interval              BWRVIP-38,                Welds H10.H11 and           (7/24/00),
Shroud Support Legs                                                                                                                                            When accessible plate; rarely                                           3.2.3                          H12              inspect with accessible)                                                                                             appropriate method (4)
BWRVIP-76-1-A,                                                         Based on as-found conditions, Shroud Horizontal                                                                                             Welds H1- H7 2.2                                           EVT-1 or UT      to a maximum 6 years for one welds                                                                                                   as applicable Figure 2-2(3)                                                         side EVT-1, 10 years for UT Maximum 6 years for one-sided EVT-1, 10 years for UT; only required when BWRVIP-76-1-A, 2.3, 3-3,            Vertical and Ring Shroud Vertical welds                                                                                                                 EVT-1 or UT       horizontal welds are found to Figures 2-4, 3-2, 3-3            Segment Welds B13.40                                                                                                                                                                     contain flaws exceeding certain limits or the shroud is a repaired shroud Each                                                                                             Per repair designer Accessible                                        BWRVIP-76-1-A, Shroud Repairs (3)                             VT-3        10-year                                            Tie-Rod Repair             VT-3              recommendations per Surfaces                                              3.5, 3.6 Interval                                                                                              BWRVIP-76 BWRVIP-26-A Top Guide and Top 3.2                   Top Guide Studs             VT-3              Each 10-year Interval Guide Grid Table 3-2 BWRVIP-25 Core Support Plate                                                                     3.2                    None for BWR/6              N/A                        N/A Table 3.2 EVT-1 of BWRVIP-47-A                 CRGT Body Welds Control Rod Guide                                                                                                                     body welds      10% of the CRGT Assemblies 3.2                   and Fuel Support Tubes (CRGTs)                                                                                                                      and VT-3 of               within 12 years Table 3.3                   Pins and Lugs pins and lugs NOTES:
D   This Table provides only an overview of the requirements. For more details, refer to ASME Code, Section XI, Table IWB-2500-1, and the appropriate BWRVIP document.
: 2)  For Perry this results in a requirement of 10 percent of the weld length. However, for H9 essentially 100 percent of the weld length was ultrasonically examined.
: 3)  Perry's shroud is a Category B un-repaired shroud.
: 4)  When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection of these welds. Until such time, and as committed to in BWRVIP-47-A, Section 3.2.5, visual inspections of the lower plenum area (which includes the shroud support legs) will be performed to the extent practical when access is made available through non-routine refueling outage activities (for example, jet pump disassembly).

IR-056, Revision 2, Attachment 1 COMPARISON OF ASME CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Page 1 of 5 The following paragraphs provide a comparison of the examination requirements in ASME Code, Section XI, examination Table IWB-2500-1, Item Nos. B13.10, B13.20, B13.30, and B13.40, to the examination requirements in the BWRVIP guidelines.
Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.
ASME Code, Section XI. Requirement B13.10 - Reactor Vessel Interior Accessible Areas (B-N-1)
The ASME Code requires a VT-3 examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately three years, during the first inspection interval, and each period during each successive 10-year inspection interval.
Typically, these examinations are performed every other refueling outage of the Inspection Interval, or three times during the 10-year interval. This examination requirement is a non-specific requirement that is a departure from the traditional Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts; loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products; wear; and structural degradation.
Portions of the various examinations required by the applicable BWRVIP guidelines require examination of accessible areas of the reactor vessel during each refueling outage. Examination of core spray piping and spargers (BWRVIP-18-R2-A), top guide (BWRVIP-26-A), jet pump welds and components (BWRVIP-41), interior attachments (BWRVIP-48-A), core shroud welds (BWRVIP-76-R1-A), shroud support (BWRVIP-38),
low pressure coolant injection couplings (BWRVIP-42), and lower plenum components (BWRVIP-47-A) provides such access. Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 examination of these areas or spaces as the specific weld or component examinations are performed. This provides an equivalent method of visual examination on a more frequent basis than that required by the ASME Code. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements. Therefore, the specified BWRVIP guideline requirements meet or exceed the ASME Code requirements (including method and frequency requirements) for examination of the interior of the reactor vessel. Accordingly, these BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject ASME Code requirements.

IR-056, Revision 2, Attachment 1 COMPARISON OF ASME CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Page 2 of 5 ASME Code. Section XI. Requirement - B13.20 - Interior Attachments within Beltline Region (B-N-2)
The ASME Code requires a VT-1 examination of accessible reactor interior surface attachment welds within the beltline each 10-year interval. In the boiling water reactor, this includes the jet pump riser brace welds-to-reactor vessel wall and the surveillance specimen support bracket welds-to-reactor vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the surveillance specimen support bracket welds, and requires an EVT-1 examination of the remaining attachment welds in the beltline region in the first 12 years, and then 25 percent during each subsequent 6 years.
The jet pump riser brace examination requirements are provided below to show a comparison between the ASME Code and BWRVIP examination requirements.
* The ASME Code requires a 100 percent VT-1 examination of the jet pump riser brace-to-reactor vessel wall pad welds each 10-year interval.
* BWRVIP-41 requires an EVT-1 examination of the jet pump riser brace-to-reactor vessel wall pad welds the first 12 years and then 25 percent during each subsequent 6 years.
* BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined.
The ASME Code VT-1 examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP enhanced VT-1 (EVT-1) is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and intergranular stress corrosion cracking (IGSCC), the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT-1 examination.
The ASME Code visual examination method requires (depending on applicable ASME Code Edition) that a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires the same 0.044 inch resolution on the examination surface, and additionally the performance of a cleaning assessment and cleaning as necessary. While the jet pump riser brace configuration varies depending on the vessel manufacturer, BWRVIP-48-A includes diagrams and prescribes examination for each configuration.

IR-056, Revision 2, Attachment 1 COMPARISON OF ASME CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Page 3 of 5 The calibration standards used for BWRVIP EVT-1 exams utilize the same ASME Code characters, thus assuring at least equivalent resolution compared to the ASME Code requirements. Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced flaw detection capability of an EVT-1 with a less frequent examination schedule provides an acceptable level of quality and safety to that provided by the ASME Code.
ASME Code, Section XI, Requirement - B13.30 - Interior Attachments Beyond Beltline Region (B-N-2)
The ASME Code requires a VT-3 examination of accessible reactor vessel interior surface attachment welds beyond the beltline each 10-year interval. In the boiling water reactor, this includes the core spray piping support bracket welds-to-reactor vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support and hold down bracket welds-to-reactor vessel wall, and the guide rod support bracket weld-to-reactor vessel wall. BWRVIP-48-A requires as a minimum the same VT-3 examination method as the ASME Code for some of the interior attachment welds beyond the beltline region, and in some cases specifies an EVT-1 for these welds. For those interior attachment welds that have the same VT-3 method of examination, the same scope of examination (accessible welds), the same examination frequency (each 10-year interval) and the same ASME Code flaw evaluation criteria are used.
Therefore, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by the ASME Code.
For the core spray support bracket attachment welds and the feedwater sparger support bracket attachment welds, the BWRVIP guidelines require an EVT-1 examination at the same frequency as the ASME Code, or at a more frequent rate. Therefore, the BWRVIP enhanced examination requirements provide the same level of quality and safety compared to that provided by the ASME Code.
The core spray piping bracket-to-reactor vessel attachment weld is used as an example for comparison between the ASME Code and BWRVIP examination requirements as discussed below.
The ASME Code examination requirement is a VT-3 examination of each weld every 10 years. The BWRVIP-48-A examination requirement is an EVT-1 for the core spray piping bracket attachment welds with each weld examined every four cycles (eight years for a two year fuel cycle).
The BWRVIP-48-A examination method EVT-1 has a superior flaw detection and sizing capability, the examination frequency is greater than the ASME Code requirements, and the same flaw evaluation criteria are used.

IR-056, Revision 2, Attachment 1 COMPARISON OF ASME CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Page 4 of 5 The ASME Code VT-3 examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An EVT-1 examination is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, the relevant degradation mechanisms for BWR reactor vessel internal attachments.
Therefore, because the EVT-1 examination method provides the same examination scope (accessible welds), an increased examination frequency (8 years instead of 10 years) in some cases, and the same flaw evaluation criteria as the ASME Code, the level of quality and safety provided by the BWRVIP criteria meets or exceeds that provided by the ASME Code requirements.
ASME Code. Section XI. Requirement - B13.40 - Core Support Structure (B-N-2)
The ASME Code requires a VT-3 examination of accessible surfaces of the integrally welded core support structure each 10-year interval. In a General Electric Company boiling water reactor designated BWR/6, the welded core support structure has primarily been considered the shroud itself and the shroud support structure, including the shroud support plate (annulus floor) the shroud support ring, the shroud support welds, and the shroud support legs (if accessible). Historically, this requirement has been interpreted and satisfied differently across the industry. Category B-N-2 is titled, "Integrally Welded Core Support Structures and Interior Attachments to Reactor Vessels." However, since the title for Item No. B13.40 simply states, "Core Support Structure," some plants, including Perry, have also applied the examination requirements to other core support structures such as the control rod guide tubes, core plate, and top guide assembly. The proposed alternate examinations replace this ASME Code requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.
* The ASME Code requires a VT-3 of accessible surfaces each 10-year interval.
* The BWRVIP requires, as a minimum, the same examination method (VT-3) as the ASME Code for integrally welded core support structures, and for specific areas, it requires either an enhanced visual examination technique (EVT-1) or ultrasonic examination (UT).
BWRVIP recommended examinations of core support structures are focused on the known susceptible areas of these structures, including the welds and associated weld heat affected zones. As a minimum, the same or superior visual examination technique is required for examination at the same frequency as the ASME Code examination requirements. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at a frequency identical to the ASME Code requirement.

IR-056, Revision 2, Attachment 1 COMPARISON OF ASME CODE EXAMINATION REQUIREMENTS TO BWRVIP EXAMINATION REQUIREMENTS Page 5 of 5 The BWRVIP guidelines require an EVT-1 or UT of core support structures. The core shroud, shroud support plate, and top guide grid are used as examples for comparison between the ASME Code and BWRVIP examination requirements as shown below.
Comparison of BWR core shroud examination and flaw evaluation requirements:
* The ASME Code requires a VT-3 examination of accessible surfaces every 10 years.
* BWRVIP-76-R1-A requires an EVT-1 examination from the inside and outside surface, where accessible, or UT examination of select circumferential welds that have not been structurally replaced with a shroud repair, at a calculated "end of interval" that will vary depending upon the amount of flaws present, but not to exceed 10 years.
Comparison of BWR shroud support inspection and flaw evaluation requirements:
* The ASME Code requires a VT-3 examination of accessible surfaces every 10 years.
* BWRVIP-38 requires examinations of the support plate to shroud weld (H8) and support plate to reactor vessel weld (H9). Examination coverage is required to be 100 percent flaw tolerance or 10 percent of the weld length, whichever is greater. Examinations are to be performed by EVT-1 or UT from the annulus or UT from the reactor pressure vessel outside surface. Re-inspection depends upon the amount of flaws present, but not to exceed six years for EVT-1 or 10 years for UT.
Comparison of BWR shroud top guide grid inspection and flaw evaluation requirements:
* The ASME Code requires a VT-3 examination of accessible surfaces every 10 years.
* For BWR/6 plants, which have top guide grids that are fabricated from two solid plates that are welded together with a machined out grid, BWRVIP-183-A requires EVT-1 or UT examinations of the rim areas containing the weld and heat affected zone from the top surface of the top guide and two cells in the same plane/axis as the weld every six years. The regions of the grid beam cells to be inspected are the bottom 2 inches of the interior side surfaces.}}