ML090020097
| ML090020097 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/29/2008 |
| From: | Flaherty M Calvert Cliffs, Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML090020097 (199) | |
Text
Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 Constellation Energy0 Nuclear Generation Group December 29, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:
SUBJECT:
Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Submittal of Fourth Ten-Year Interval Inservice Inspection Program Plan and Request for Approval of Alternatives (Relief Requests) to American Society of Mechanical Engineers Code,Section XI (a)
Letter from Mr. R. J. Laufer (NRC) to Mr. P. E. Katz (BGE), dated March 6, 2003, "Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 -
Alternative to Inservice Inspection Requirement for Replacement Steam Generator Girth Welds"
REFERENCES:
(b)
Letter from Mr. T. L. Tate (NRC) to Mr. M. K. Nazar (Indiana Michigan Power Company), dated September 28, 2007, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Risk-Informed Safety-Based Inservice Inspection Program for Class 1 and 2 Piping Welds" Pursuant to 10 CFR 50.55a(g)(5)(i), Calvert Cliffs Nuclear Power Plant, Inc., the licensee for Calvert Cliffs Nuclear Power Plant (Calvert Cliffs), submits its proposed Fourth Ten-Year Interval Inservice Inspection (ISI) Program Plan. This ISI plan begins October 10, 2009 for each unit and will end on June 30, 2019. A copy of Calvert Cliffs ISI plan for both units is provided in Attachment (1).
In addition to the ISI program submittal, four necessary requests for relief from the requirements of American Society of Mechanical Engineers (ASME) Code,Section XI are submitted in Attachments (2) through (5) for your review and approval. All of the relief requests are submitted under the provision of 10 CFR 50.55a(a)(3)(i) as alternatives that provide an acceptable level of quality and safety. Relief Requests ISI-04-01 and ISI-04-02 (Attachments 2 and 3) are new requests based on ASME Code Case N-747 and N-753, respectively. Relief Request ISI-04-03 (Attachment 4) was previously approved for Calvert Cliffs third ten-year ISI interval in Reference (a). Relief Request ISI-04-04 (Attachment 5) requests approval for implementation of a risk-informed/safety based inservice inspection program for Class I and 2 piping based on ASME Code Case N-716. Relief Request ISI-04-04 is similar to the relief request approved for Donald C. Cook Nuclear Plant in Reference (b).
Ac o-rT
Document Control Desk December 29, 2008 Page 2 Calvert Cliffs requests approval of these relief requests prior to the start of the fourth ten-year ISI interval, which begins October 10, 2009. These relief requests are for the duration of the Fourth Ten-Year Interval ISI.
Should you have questions regarding this matter, please contact Mr. Jay S. Gaines at (410) 495-5219.
Very truly yours, Mark D. Flaherty Manager - Engineering Services MDF/KLG/bjd Attachments:
(1)
Fourth Interval Inservice Inspection Program Plan for Calvert Cliffs Nuclear Power Plant Units I and 2, Revision 0 (2)
(3)
(4)
ASME Code, ASME Code, ASME Code,
Enclosures:
Section XI Relief Request -- ISI-04-01 Section XI Relief Request -- ISI-04-02 Section XI Relief Request -- ISI-04-03 (1) Replacement Steam Generator Weld Locations (2) CCNPP Code Interpretation Request Letter (3) ASME Code Interpretation Letter Section XI Relief Request -- ISI-04-04 (1) Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions (5)
ASME Code,
Enclosure:
cc:
D. V. Pickett, NRC S. J. Collins, NRC Resident Inspector, NRC S. Gray, DNR
ATTACHMENT (1)
FOURTH INTERVAL INSERVICE INSPECTION PROGRAM PLAN FOR CALVERT CLIFFS NUCLEAR POWER PLANT UNITS 1 AND 2, REVISION 0 Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program FOURTH INTERVAL INSERVICE INSPECTION PROGRAM PLAN FOR CALVERT CLIFFS NUCLEAR POWER PLANT UNITS 1 AND 2 Revision 0 Prepared for Constellation Energy by lSaa CO Ino.
Calvert Cliffs Units I & 2 Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 1.0 ASME SECTION XI INSERVICE INSPECTION PROGRAM 1.1 Purpose This plan's objective is to provide a traceable link between the governing code requirements and the implementing procedures in order to ensure that Regulatory and ASME Section Xl Code requirements for the inservice inspection of safety related systems, components and structures are being fulfilled.
These instructions provide the necessary guidance for the Corporate & Station Engineering Programs Unit (EPU) personnel at Constellation Energy's Calvert Cliffs Nuclear Power Plant (CCNPP) Units 1 and 2 to ensure the following:
- 1) Conformance to Title 10, Section 50.55a of the Code of Federal Regulations (10CFR50.55a).
- 2) Conformance to the 2004 Edition of Section Xl of the American Society of Mechanical Engineers (AMSE) Boiler and Pressure Vessel Code
- 3) Conformance to Calvert Cliffs Nuclear Power Plant Technical Requirements Manual (TRM) 15.4.3.
- 4) Conformance to Constellation Energy Corporation and Station policies, practices and procedures.
- 5) The necessary technical content is included in Calvert Cliffs Units 1 and 2 Inservice Inspection Program and implementing procedures.
- 6) The proper ASME Section Xl Code required examinations, tests and administrative procedures are implemented.
- 7) The proper ASME Code request for alternatives and relief requests are submitted to and approved by the regulatory authority.
- 8) Component repair and replacement activities are performed in accordance with ASME Section Xl Code requirements.
- 9) The proper examination, test and repair and replacement records and reports are maintained and submitted.
Figure 1 shows how this program plan effectively functions as a central source to help ensure all regulatory and Constellation Energy requirements are incorporated into the Inservice Inspection Program. This program plan provides a useful aid in program self assessments, procedure preparation and/or revision, 'management quick reference and program familiarization.
Revision 0 2
Prepared for Constellation Energy by k~idI~mI f*t3 riho.
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Figure 1 Compliance Completed Prepared for Revision 0
. 3
Calvert Cliffs Units I & 2 Constellation Energy, ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 1.2 General The CalvertCliffs Nuclear Power Plant Units 1 and 2 (CCNPP 1 & 2) Fourth Interval Inservice Inspection Program is established in accordance with Title 10 Code of Federal Regulations Part 50.55a (10CFR50.55a). This program has been developed to comply with the American Society of Mechanical Engineers (ASME) Section Xl, Rules for Inservice Inspection of Nuclear Power Plant Components and implements the requirements of Technical Requirements Manual (TRM) 15.4.3, Inservice Inspection Program.
CCNPP 1 and 2 received their Construction Permits July 7, 1969. CCNPP 1 and 2 were designed, fabricated and erected in accordance with USAS B31.7, 1969 Edition for nuclear piping, and USAS B31.1.0 - 1967 Edition for non-nuclear piping. The Nuclear Steam Supply System (NSSS) was furnished by Combustion Engineering. The Architect Engineer for CCNPP 1 and 2 during construction was Bechtel Corporation.
Nuclear Regulatory Commission (NRC) Regulatory Guide 1.26 and 10CFR50.2 were used to determine the ASME Code classification for Class 1, 2 and 3 Systems, Structures, and Components.
1.3 Inspection Intervals The Operating Licenses for CCNPP 1 and 2, Docket Nos. 50-317 and 50-318, respectively, were issued on July 31, 1974 and November 11, 1976, respectively. The commercial operation dates for CCNPP 1 and 2 were May 8, 1975, and April 1, 1977, respectively. The First Ten-Year inspection interval ended March 31, 1987. The Second Ten-Year Inspection Interval was scheduled to conclude on April 1, 1997, however, this interval was extended from April 1, 1997 to June 30, 1998 for Unit 1 and June 30, 1999 for Unit 2, as allowed by a Nuclear Regulatory Commission letter dated May 2, 1997, "Notification of Extension of the Inservice Inspection and Inservice Test Program -
Second Ten-Year Interval Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (TAC Nos.
M98251 and M98252)." The Second Ten-Year Inspection Interval for Unit 1 was further extended from June 30, 1998 to June 30, 1999, to coincide with Unit 2, as allowed by a Nuclear Regulatory Commission letter dated August 19, 1998, "Request for Extension of the Second Ten-Year Inservice Interval - Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC No. MA 1680)."
Revision 0 4
Prepared for Constellation Energy by ISaui
- at M IrIo.
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program The Third Inspection Interval for CCNPP 1 and 2 was scheduled to end June 30, 2009, but, was extended to October 9, 2009 as allowed by IWA-2430(d)(1).
This extension allows adoption of the 2004 Edition of Section XI for the Fourth Inspection Interval. The start date for the Fourth Inspection Interval for CCNPP 1 and 2 is October 10, 2009. The Fourth Inspection Interval is divided into inspection periods consistent with Table IWB-2412-1. The inspection periods are scheduled as follows:
1st Period: From October 10, 2009 to October 9, 2012 2nd Period: From October 10, 2012 to October 9, 2016 3rd Period: From October 10, 2016 to June 30, 2019 (3 Years)
(4 Years)
(3 Years)
In accordance with ASME Section Xl, IWB-2412(b) and IWA-2430, that portion of an inspection interval described as an inspection period may be decreased or extended by as much as 1 year to enable inspections to coincide with a plant refueling outage.
However, the adjustments must not cause successive intervals to be altered by more than 1 year. Outages during the fourth inservice inspection interval are scheduled as follows:
CCNPP Unit 1 1st Period 2 nd Period 3 rd Period Refueling Outages Refueling Outages Refueling Outages 1RFO19 1RFO20 1RFO21 1RF022 1RF023 CCNPPUnit2 2 1st Period 2 nd Period 3 rd Period Refueling Outages Refueling Outages Refueling Outages 2RF018 2RF019 2RF020 2RF021 2RF022 1.4 ASME Section XI Code of Record for the Fourth Inservice Inspection Interval 10CFR50.55a requires that inservice inspection of components and system pressure tests conducted during successive 120 month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference 12 months prior to the start of the 120 month inspection interval. On October 10, 2008, the latest edition of the ASME Code Section Xl accepted by the NRC was the 2004 Edition with no Addenda.
Revision 0 5
Prepared for Constellation Energy by 2
/k~uI IE*Q*1m hIro
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program The ISI Program for ASME Class 1, 2, and 3 components for the Fourth Inspection Interval have been developed using the ASME Code, Section Xl, 2004 Edition; except where specific written alternatives from Code requirements has been requested by Constellation Energy and granted by the NRC.
1.5 ASME Section XI Inservice Inspection Program Description The ASME Section XI Inservice Inspection Program for the Fourth Inspection Interval is comprised of the following individual programs:
The ASME Section Xl and Augmented ISI Program for pressure retaining components and their supports, The ASME Section Xl System Pressure Test Program, The ASME Section Xl Containment Inspection Program, and The ASME Section Xl Repair and Replacement Program The programs are separate documents or procedures, each bearing the title of the program. The ISI Programs are administered through MN-3-110, "Inservice Inspection of ASME Section Xl Components".
For the purpose of clarification, the following definitions for terms used in the ASME Section Xl Inservice Inspection program are:
PROGRAM SCHEDULE The term "program" is used to provide the structure and methods for accomplishment of the object and is to be considered synonymous with the term "plan" as used in ASME Section XI, 2004 Edition, Article IWA-2400.
The term "schedule" is used to define the detailed scheme of performing the required inspections over the ten-year interval.
1.5.1 Ten-Year Inservice Inspection (ISI) Proaram Description The Ten-Year Inservice Inspection (ISI) Program details CCNPP compliance with ASME Code, Section Xl 2004, Edition, Articles IWA, IWB, IWC, IWD, and IWF for examination of Class 1, Class 2, and Class 3 welds and supports. This document defines the Class 1, 2, and 3, components and the Code required examinations for each ASME Section XI examination category, and the augmented inspection scope.
The purpose of the Ten-Year ISI Program is to periodically perform nondestructive examination of ASME Class 1, 2, or 3 safety related components and supports in order to identify the presence of any service related degradation.
Revision 0 6
Prepared for Constellation Energy by
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program The Administrative procedures and Inspection Schedule described in the Ten-Year ISI Program, combined with applicable CCNPP and approved vendor procedures, constitute the ISI portion of the Ten-Year ISI Program required by TRM 15.4.3.
For convenience in implementation of commitments, the Ten-Year ISI Program also contains information applicable to augmented examinations which are performed to fulfill 10CFR50.55a Augmented Examination requirements or to address industry operating experience. These augmented examinations are not required by the ASME Code and are not included in the summary totals for the specific examination category.
The Ten-Year ISI Program schedule is contained in the IDDEAL Software Suite ScheduleWorks computer database. This information is retained at the plant site and is available for review.
1.5.2 Repair and Replacement Program Description ASME Section Xl rules for repairs and replacements are not contained in this plan.
MN-3-120, ASME Section Xl Repair and Replacement Program is the implementing procedure which describes the implementation process for the repair and replacement provisions of ASME Code, Section Xl, Articles IWA-4000, IWB-4000, IWC-4000, IWD-4000, IWE-4000, IWF-4000, and IWL-4000.
1.5.3 Snubber Program Description The Snubber Program rules are not contained in this plan.
EN-1-124, Snubber Program, satisfies the requirements established in TRM 15.7.2 for the inspection and testing of safety-related snubbers.
1.5.4 Pressure Test Program Description ASME Section Xl rules for pressure testing are not contained in this plan. MN XXX, ASME Section Xl Periodic Pressure Testing, satisfies the requirements established in TRM 15.4.3 and the ASME Code, Section Xl, Articles IWA-5000, IWB-5000, IWC-5000, and IWD-5000.
1.5.5 Containment Inspection Program Description ASME Section Xl rules for CC and MC components are not contained in this plan.
ISI-PROG-PLAN, ASME Section Xl, IWE/IWL, CCNPP Units 1 and 2, implements the requirements of the ASME Code, Section Xl Articles IWE and IWL (Class MC and CC components).
Revision 0 7
Prepared for Constellation Energy by Akkfltml G:orwpi Ino.
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 1.5.6 Pump and Valve Program Description The Inservice Testing (IST) Program requirements are not contained in this plan.
EN-4-102, ASME Pump and Valve Inservice Testing Program Requirements implements the requirements of the ASME OM Code. This program is designed to assess the operational readiness of ASME Class 1, 2, or 3 pumps and valves and to provide a means for early detection of degradation in either the mechanical or hydraulic characteristics of the components.
1.5.7 Steam Generator Tube Surveillance Program Description ASME Section Xl rules for Steam Generator Tube Inspections are not contained in this plan.
EN-4-106, Steam Generator Surveillance Program, implements the technical and administrative requirements of Technical Specification 5.5.9.
1.6 Administrative Controls The ASME Section XI Inservice Inspection (ISI) Plan is developed and maintained by the ISI Program Owner assigned to Corporate Engineering Programs.
Administrative Procedure MN-3-110, Inservice Inspection of ASME Section Xl Components, is the plant procedure that establishes administrative controls for the ASME Section Xl ISI program. The procedure also describes additional ASME Section Xl ISI Programs at Calvert Cliffs, delineates the functional responsibilities of personnel relative to the administration and implementation of the ASME Section XI ISI Programs, and lists the procedures necessary to implement these programs.
Preparation and changes to the ASME Section Xl ISI Program Plan are the responsibility of the respective Program Owner. MN-3-312, Inservice Inspection Plans and Owner's Activity Report, provides the methodology for review, comment, revision, and approval of programs.
Distribution of the ASME Section XI ISI Program Plan is controlled through a standardized distribution process, defined per CNG-PR-2.01-1000, Document Control Program.
Requests to amend the inspection requirements for the Fourth Inspection Interval, as defined by ASME Code, Section Xl, 2004 Edition, are developed and processed in accordance with MN-3-110.
Copies of the requests and the NRC Safety Evaluation Reports are identified in Section 3.0.
Revision 0 8
Prepared for Constellation Energy by
- kdma, conoqm*.
ho..
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 2.0 ASME CODE CASE APPLICABILITY This section contains ASME Code Cases applicable to the CCNPP Units 1 and 2 Fourth Inspection Interval.
2.1 Adoption of Code Cases All Code Cases adopted for ASME Section XI activities for use during the Fourth Interval are listed in Tables 2.2.1, 2.3.1, and 2.4.1. The use of Code Cases is in accordance with ASME Section Xl, IWA-2440, 10 CFR 50.55a, and Regulatory Guide 1.147. As permitted by ASME Section and Regulatory Guide 1.147 or 10 CFR 50.55a, ASME Section XI Code Cases may be adopted and used as described below:
2.1.1 Adoption of Code Cases Listed for Generic Use in Regulatory Guide 1.147 Code cases that are listed for generic use in the latest revision of Regulatory Guide 1.147 may be included in the ISI program provided any additional provisions specified in the Regulatory Guide are also incorporated.
Table 2.2.1 identifies those code cases approved for generic use.
2.1.2 Adoption of Code Cases Not Approved in Regulatory Guide 1.147 Certain Code cases that have been approved by the ASME Board of Nuclear Codes and Standards may not have been reviewed and approved by the NRC Staff for generic use and listed in Regulatory Guide 1.147.
Use of such Code Cases may be requested in the form of a "Request for Alternative" in accordance with 10 CFR 50.55a(a)(3). Once approved, these Requests for Alternatives will be available for use until such time that the Code cases are adopted into Regulatory Guide 1.147, at which time compliance with the provisions contained in the Regulatory Guide is required.
Table 2.3.1 identifies those code cases that have been requested through Requests for Alternatives. For convenience to the user of this ISI Program, the appropriate internal correspondence number is provided to assist in retrieval from Document Control. All other Requests for Alternatives and Relief Requests (those not associated with NRC approval of Code cases are addressed in Section 3.
Revision 0 9
Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 2.1.3 Adoption of Code Cases Mandated by 10 CFR 50.55a The NRC may require the licensee to follow an augmented inservice inspection program for systems and components for which the Commission deems that added assurance of structural reliability is necessary.
Many times these "Augmented Requirements" will be contained in code cases that ASME has approved.
The NRC may mandate its use and add conditions it believes are necessary via 10 CFR 50.55a. Table 2.4.1 identifies those code cases Mandated by 10 CFR50.55a.
2.1.4 Use of Annulled Code Cases As permitted by Regulatory Guide 1.147, code cases that have been adopted for use in the current interval that are subsequently annulled by ASME, may be used for the remainder of the interval.
2.1.5 Code Case Revisions As permitted by Regulatory Guide 1.147, activities performed to a specific revision of an approved code case need not be changed when a subsequent revision of the code case is listed as the approved version in the Regulatory Guide. An exception to this provision would be the inclusion of a limitation or condition on the use of the code case which is necessary to enhance safety.
2.1.6 Adoption of Code Cases Issued Subsequent to Filing the Inservice Inspection Plan Code cases issued by ASME subsequent to filing the Inservice Inspection Plan with the NRC may be incorporated within the provisions of paragraphs 2.1.1 or 2.1.2 by revision to this ISI Plan. Any subsequent code cases should be added to this program document (in Table 3.4-1) at the next document revision.
2.1.7 Non Inservice Inspection Code Cases Only Code Cases applicable to nondestructive examination requirements for Class 1, 2, and 3 components and component supports are addressed in Table 3.4-1.
Code Cases applicable to System Pressure Testing, Containment Inservice Inspection and Repair/Replacement Activities are addressed in their respective programs.
2.1.8 Code Cases not approved for use by the NRC Certain Code cases that have been approved by the ASME Board of Nuclear Codes and Standards may not be approved by the NRC Staff for generic use and listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use.
Revision 0 10 Prepared for Constellation Energy by llclmil *na
- .pu Ino.
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program However, the NRC may approve their use in specific cases. Code cases listed in the Regulatory Guide will not be used at CCNPP without a approved Request for Alternative in accordance with 10,CFR 50.55a(a)(3).
2.2 Reaulatory Guide 1.147, Revision 15 Approved Code Cases Table 2.2.1 - Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Limitations Number Alternative Examination N-460 Coverage for Class 1 and 2 None Welds Alternative Requirements for N-526 Successive Inspections of Class None 1 and 2 Vessels Repair/Replacement Activity N-532-4 Documentation Requirements None and Inservice Summary Report Preparation and Submission To achieve consistency with the 10 CFR 50.55a rule change published September 22, 1999 (64 FR 51370),
incorporating Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," to Section XI, add the following to the Alternative Methods -
specimen requirements:
N-552 Qualification for Nozzle Inside Radius Section from the Outside "At least 50 percent of the flaws in the Surface demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks. Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches."
Add to detection criteria, "The number of false calls must not exceed three."
Revision 0 11 Prepared for Constellation Energy by I2dkaIl C~r Ino.
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program Table 2.2.1 - Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Limitations Number Alternative Additional N-586-1 Examination Requirements for None Classes 1, 2, and 3 Piping, Components, and Supports Alternative Examination N-593 Requirements for Steam None Generator Nozzle to Vessel Welds Ultrasonic Examination of Penetration Nozzles in Vessels, N-613-1 Examination Category B-D, Item None Nos. B3.10 and B3.90, Reactor Nozzle-to-Vessel Welds, Figs.
IWB-2500-7(a), (b), and (c)
N-624 Successive Inspections None Chemical ranges of the calibration block may vary from the materials specification if (1) it is within the Alternative Calibration Block chemical range of the component N-639 Material specification to be inspected, and (2) the phase and grain shape are maintained in the same ranges produced by the thermal process required by the material specification.
Alternative Pressure-N-641 Temperature Relationship and None Low Temperature Overpressure Protection System Requirements Revision 0' 12 Prepared for Constellation Energy by l[Klilr"*,ntla;.
kIr'l
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 2.2.1 - Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Limitations Number In place of a UT examination, licensees may perform a visual examination with enhanced magnification that has a resolution sensitivity to detect a 1 -mil width wire or crack, utilizing the allowable flaw length criteria of Table Alternative Requirements for IWB-3512-1 with limiting assumptions on the flaw aspect ratio. The N-648-1 Inner Radius Examination of on t
f aspe io. The Ils Reactor Vessel Nozzles provisions of Table IWB-2500-1, Class 1Examination Category B-D, continue to apply except that, in place of examination volumes, the surfaces to be examined are the external surfaces shown in the figures applicable to this table (the external surface is from point M to point N in the figure).
Alternative Requirements for N-665 Beam Angle Measurements None Using Refracted Longitudinal Wave Search Units Alternative Requirements for N-686-1 Visual Examinations, VT-1, VT-2, None H
and VT-3 Revision 0 13 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval inservice Inspection Program 2.3 Code Cases Approved Throuqh Request for Alternatives The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 15 and require a request for alternative prior to implementation. See Section 3.0 of this plan for the applicable requests.
Table 2.3.1 - Code Cases Adopted Via NRC Approved Requests Code Request for' Case Title Alternative No.
Number N-747 Reactor Vessel Head-to-Flange Weld Examinations ISI-04-01 N-753 Vision Tests ISI-04-02 N-716 Alternative Piping Classification and Examination ISI-04-04 Requirements 2.4 Code Cases Adopted Via 10 CFR 50.55a The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 15, but, are mandated in 10 CFR 50.55a as augmented requirements.
Table 2.4.1-Code Cases Adopted Via 10 CFR 50.55a Code Case Title Notes Number Additional Examinations for PWR Pressure Implemented in the N-722 Retaining Welds in Class 1 Components Fabricated Pressure Test With Alloy 600/82/182 Materials Program Conditions specified Alternative Examination Requirements for PWR in paragraphs N-729-1 Reactor Vessel Upper Heads With Nozzles Having (g)(6)(ii)(D)(2)
Pressure-Retaining Partial-Penetration Welds through (6) of 10 I_
ICFR 50.55a Revision 0 14 Prepared for Constellation Energy by i~k~lcdlalm
- n~l
- Ino,
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 3.0 RELIEF REQUESTS The 2004 Edition of ASME Section XI was written to provide requirements for defense in-depth inspections for the nuclear industry. However, not all requirements are applicable or possible to be performed at every plant. Therefore, Constellation has reviewed the requirements contained in the 2004 Edition of Section XI and determined where those requirements would not be viable at Calvert Cliffs Nuclear Plant.
10 CFR 50.55a provides three options to submit these determinations to the staff for review and approval.
3.1 Request For Alternatives that Provide an Acceptable Level of Quality and Safety 10 CFR 50.55a(a)(3)(i) allows alternatives to 10 CFR 50.55a(g), when authorized by the NRC, if the proposed alternatives would provide an acceptable level of quality and safety.
In cases where Constellation proposes alternatives to the ASME Section Xl requirements that would provide an acceptable level of quality and safety, a Request for Alternative, as allowed by 10CFR50.55a(a)(3)(i) will be submitted to the NRC.
3.2 Request For Alternatives Required Due to Burden 10 CFR 50.55a(a)(3)(ii) allows alternatives to 10 CFR 50.55a(g), when authorized by the NRC, if compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. In cases where Constellation proposes alternatives to ASME Section XI when compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, a Request for Alternative as allowed by 10CFR50.55a(a)(3)(ii), will be submitted to the NRC.
3.3 Relief Request Required due to lmpracticality or Limited Examinations 10 CFR 50.55(a)(g)(iii) allows relief to be submitted In cases where the ASME Section Xl requirements for inservice inspection are considered impractical.
In cases where the ASME Section Xl requirements for inservice inspection are considered impractical, Constellation will notify the NRC and submit information to support the determination, as required by 10 CFR50.55a(g)(5)(iii). The submittal of this information will be referred to as a Request for Relief. Per 10 CFR50.55a paragraph (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate Requests for Relief per Paragraph (g)(5) and
"...may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility".
Revision 0 15 Prepared for Constellation Energy by Zkk*mlm Gowpa
- buo,
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program In the event that the entire examination volume or surface (as defined in the ASME Code) cannot be examined due to interference by another component or part geometry, then, in accordance with Code Case N-460, a reduction in examination volume or area is acceptable if the reduction is less than 10%.
In the event that the reduction in examination volume or area is 10% or greater, a request for relief will be submitted. NRC Information Notice 98-42 provides additional guidance that all ASME Section Xl examinations should meet the examination coverage criteria established in Code Case N-460.
Therefore, the guidance included in NRC Information Notice 98-42 will be followed by Constellation when determining whether to prepare a relief request or apply the criteria of Code Case N-460 for examinations where less than 100% coverage of any Section Xl examination is obtained.
3.4 Table of Requests Table 3.4-1 contains an index of Requests For Alternatives and Requests For Relief written in accordance with 10 CFR50.55a (a)(3) and (g)(5).
The applicable Constellation submittal and NRC Safety Evaluation Report (SER) correspondence numbers are also included in Table 3.4-1 for each request for alternative and request for relief.
Note that only Requests for Alternatives or Requests for Relief applicable to nondestructive examination requirements for Class 1, 2, and 3 components and component supports are addressed in Table 3.4-1.
Requests for Alternatives or Requests for Relief applicable to System Pressure Testing, Containment Inservice Inspection and Repair/Replacement Activities are addressed in their respective programs.
Revision 0 16 Prepared for Constellation Energy by llk 0-
, 13a--.
(l Ln. kM
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 3.4-1 Calvert Cliffs Unitsl&2 Fourth Interval Relief Requests(1 )
,Relief Constellation R
st Relief Request Description Correspondence NRC Ser Correspondence Modified Request to Use Code Case N-747, ISI-04-01 Reactor Vessel Head-to-Flange XXXXXXX Waiting NRC SER N/A Weld Examinations Request to Use Code Case N-753, XXXXXXX Waiting NRC SER N/A Vision Tests ISI-04-03 Request to Classify S/G Weld as XXXXXXX Waiting NRC SER N/A Non Gross Structural Discontinuity Request to Use Code Case N-716, ISI-04-04 Alternative Piping Classification XXXXXXX Waiting NRC SER N/A and Examination Requirements Note(l): ISI-20 and ISI-21 are third interval relief request that effect the forth Plan for a discussion of these relief requests.
interval. See the Third Interval Inservice Inspection Revision 0 17 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 Constellation Energy-ASME Section Xl Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 4.0 TEN-YEAR INSERVICE INSPECTION PROGRAM 4.1 General The Ten-Year ISI Program is administered by CCNPP procedure MN-3-110. This procedure delineates the functional responsibilities of personnel relative to the administration and implementation of the Ten-Year ISI Program.
MN-3-312 and the Inservice Inspection Plans are the procedures used to implement the Ten-Year ISI reporting requirements.
The Ten-Year ISI Program details Constellation Energy's compliance with ASME Code,Section XI 2004, Edition, Articles IWA, IWB, IWC, IWD, and IWF for examination of Class 1, Class 2, and Class 3 components and their supports. This document also defines the Class 1, 2, and 3 systems, components and Code classification boundaries, the required examinations for each ASME Section Xl examination category, and the augmented inspection scope.
The administrative procedures and inspection schedule described in the Ten-Year ISI Program, combined with applicable Constellation Energy and approved vendor procedures, constitute the Ten-Year ISI Program required by TRM 15.4.3.
For convenience in implementation of commitments, the Ten-Year Program also contains information applicable to augmented examinations. These examinations are performed to fulfill 10 CFR50.55a Augmented Examination requirements, or are performed to address industry operating experience. These examinations are not required by Code and as such, are not included in the summary totals for the specific examination category.
The purpose of the Ten-Year ISI Program is to periodically perform nondestructive examination of ASME Class 1, 2, or 3 safety related components and supports in order to identify the presence of any service related degradation.
The Ten-Year ISI inspection schedule is contained in an IDDEAL Software Suite, ScheduleWorks computer database. This information is retained at the plant site and is available for review upon request.
4.2 Developmental References 4.2.1 Title 10, Code of Federal Regulations, Part 50.
4.2.2 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section Xl, 2004 Edition no addenda.
4.2.3 Regulatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity.
Revision 0 18 Prepared for Constellation Energy by
& lAlkkhmil GlommpMls* Ino.
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program 4.2.4 4.2.5 4.2.6 4.2.7 4.2.8 4.2.9 4.2.10 Regulatory Guide 1.147, Code Case Applicability.
Regulatory Guide 1.26, Quality Group Classifications.
Calvert Cliffs Unit 1 and 2 Technical Requirements Manual Calvert Cliffs Unit 1 and 2 Technical Specifications MN-3-110, Inservice Inspection of ASME Section Xl Components MN-3-312, Inservice Inspection Plans and Owner's Activity Report MN-3-105, Qualification of Nondestructive Examination Personnel and Procedures Revision 0 19 Prepared for Constellation Energy by lad 1hrKI iD.In
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant.
Fourth Interval Inservice Inspection Program 5.0 Augmented Inservice Inspection Requirements Augmented inservice inspection requirements are those examinations that are specified by documents other than the ASME Section Xl Code.
Typically, these augmented examinations are at the request of the Nuclear Regulatory Commission through such mechanisms as Bulletins, Notices and Regulatory Guides. The augmented examinations addressed by the CCNPP ISI Program during the fourth inspection interval are as follows:
5.1 RI-BER
Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping (Examination of High Energy Line Break Piping)
Source Document: NUREG-0800, Standard Review Plan Associated Document:
Section 3.6.2, CCNPP UFSAR, Appendix 10A, Technical Requirements Manual 15.4.3.2, EPRI TR 1006837, Application of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER)
Programs.
Purpose:
The purpose of this augmented program is to perform examinations on piping subject to High Energy Line Break (HELB) analysis criteria.
Those examinations previously performed on HELB piping per NUREG-0800, Section 3.6.2 will be maintained and performed under the risk-informed break exclusion region (RI-BER) application during the current interval.
During previous intervals, examinations were required on 100% of the subject welds per NUREG-0800, Section 3.6.2, Branch Technical Position MEB 3-1, Paragraph B.l.b(7).
Scope: Augmented Inservice Inspection Program for Main Steam and Main Feedwater Piping. The unencapsulated welds greater than four inches in nominal diameter in the main steam and main feedwater piping runs located outside of Containment and traversing safety-related areas or located in compartments adjoining safety-related areas.
Method: Ultrasonic examinations will be performed per the RI-BER application, except as restricted by part geometry or access.
Industry Code or Standards: ASME Code Section Xl, 2004 Edition Revision 0 20 Prepared for Constellation Energy by l~k;Idml *noBmp~a. bIn.
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program Frequency: Ultrasonic examinations will be performed once per 10 year interval per the RI-BER application. The welds to be examined during each inspection period shall be selected to provide a representative sample of the conditions of the welds. If these examinations reveal unacceptable structural defects in one or more welds, an additional 1/3 of the welds shall be examined and the inspection schedule for the repaired welds shall revert back as if a new interval had begun. If additional unacceptable defects are detected in the second sampling, the remainder of the welds shall also be examined.
Acceptance Criteria or Standard: Flaws detected during examination shall be evaluated by comparing the examination results to the acceptance standards established in ASME Section XI, IWB-3514.
Requlatory Basis: The examination criteria of NUREG-0800, Section 3.6.2 are superseded by the RI-BER application.
5.2 RG 1.14 RCP FLYWHEEL: Reactor Coolant Pump Flywheel Integrity Source Document: NRC Regulatory Guide 1.14 Associated Document: Technical Specification 5.5.7
Purpose:
The reactor coolant pump (RCP) motor flywheels are examined due to a concern about high-energy missiles inside containment that could potentially damage, and cause the simultaneous failure of, multiple trains of redundant safety-related systems.
Scope: The scope includes the examination of all four RCP flywheels.
Method: Surface and volumetric examinations of all four RCP flywheels shall be conducted in accordance with CCNPP Technical Specification 5.5.7. These examinations are to be performed to the extent possible through the access ports in the motor housings without disassembly of the motors.
Industry Code or Standards: ASME Code Section Xl, 2004 Edition.
Frequency: All areas of high stress concentration (i.e., keyway and bore) in all 4 RCP flywheels shall be volumetrically examined approximately every three years, and all exposed surfaces shall be surface examined and the entire flywheel shall be volumetrically examined once every 10-Year ISI Interval.
Acceptance Criteria or Standard: Any flaws detected during examination shall be forwarded to Constellation Engineering for resolution.
Regulatory Basis: The regulatory basis for this augmented examination program is NRC Regulatory Guide 1.14.
Revision 0 21 Prepared for Constellation Energy by
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program
5.3 MRP-139
Primary System Piping Butt Weld Inspection and Evaluation Guideline Source Document: Materials Reliability Project MRP-139 Associated Document: NEI 03-08 letter
Purpose:
Components in the primary coolant system containing Alloy 600 material and 82/182 dissimilar metal butt welds may be susceptible to degradation cause by primary water stress corrosion cracking (PWSCC).
MRP-139 is the industry standard for examination of primary system components which may be susceptible to PWSCC corrosion.
Scope: The scope includes those primary coolant system components containing Alloy 600 material and 82/182 dissimilar metal butt welds.
Method: Surface and volumetric examination methods are required and are based on the component material, its operating temperature, stress reduction methods utilized, (if applicable) and the component repair history and are assigned by Tables 6.1 and 6.2 of MRP-139.
Industry Code or Standards: Materials Reliability Project MRP-1 39.
Frequency: Examination frequencies are based on the component material, its operating temperature, stress reduction methods utilized, (if applicable) and the component repair history and are assigned by Tables 6.1 and 6.2 of MRP-1 39.
Acceptance Criteria or Standard: ASME Section XI, IWB-3500.
Regulatory Basis: Nuclear Energy Institute (NEI) 03-08 Initiative.
5.4 Code Case N-722: Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials Source Document: 10 CFR 50.55a Associated Document: ASME Code Case N-722
Purpose:
Components in the primary coolant system containing Alloy 600 material and 82/182 dissimilar metal butt welds may be susceptible to degradation cause by primary water stress corrosion cracking (PWSCC). Code Case N-722 is the first ASME Code Case developed to find small amounts of leakage prior to pipe failure.
Revision 0 22 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program Scope: The scope includes those primary coolant system components containing Alloy 600 material and 82/182 dissimilar metal butt welds. The inspection requirements of ASME Code Case N-722 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/182 materials that have been mitigated by weld overlay or stress improvement.
Method: Bare metal visual examination (VE) with insulation removed. Alternatively, the VE may be performed with insulation in place using remote visual inspection equipment that provides resolution of the component metal surface equivalent to a bare-metal direct VE.
The VE may be performed when the system or component is depressurized. An ultrasonic examination performed from the component inside or outside surface in accordance with the requirements of Table IWB-2500-1 and Appendix VIII shall be acceptable in lieu of the VE.
Industry Code or Standards: ASME Code Case N-722 Frequency: Examination frequencies are based on the component item number listed in Table 1 of Code Case N-722. The frequency ranges from once each refueling outage to once each inspection interval.
Acceptance Criteria or Standard: ASME Section X1, IWB-3522.
Regulatory Basis: 10 CFR 50.55a(g)(6)(ii)(E) 5.5 Code Case N-729-1: Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds Source Document: 10 CFR 50.55a Associated Document: ASME Code Case N-729-1
Purpose:
Components in the primary coolant system containing Alloy 600 material and 82/182 dissimilar metal butt welds may be susceptible to degradation cause by primary water stress corrosion cracking (PWSCC). Code Case N-729-1 is the ASME Code Case developed to inspect reactor vessel heads to locate PWSCC.
Scope: The scope includes PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration.
Revision 0 23 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Method: There are two methods required depending on location.
- 1. Bare metal visual examination (VE) of the entire surface of the head, including the intersection of each nozzle with the head, with insulation removed. Alternatively, the VE may be performed with insulation in place using remote visual inspection equipment that provides resolution of the component metal surface equivalent to a bare-metal direct VE.
The VE may be performed when the system or component is depressurized.
- 2. A ultrasonic and/or surface examination of the nozzles and partial penetration welds using personnel, procedures, and equipment that have been qualified as required by 10 CFR 50.55a(g)(ii)(D).
Industry Code or Standards: ASME Code Case N-729-1 Frequency: Examination frequencies are based on the component item number listed in Table 1 of Code Case N-722. The frequency ranges from once every 5 years or 3 refueling outages to once each inspection interval.
Acceptance Criteria or Standard: ASME Section Xl, IWB-3120, -3130, or -3140, as applicable.
Regulatory Basis: 10 CFR 50.55a(g)(6)(ii)(D) 5.6 OWNER ELECTED EXAMINATIONS FOR INTERNAL COMMITMENTS There are no Owner Elected examination commitments for CCNPP at this time in the ISI Plan.
5.7 LICENSE RENEWAL EXAMINATIONS FOR AGING MANAGEMENT COMMITMENTS Calvert Cliff's License Renewal Application credits the ISI Program for aging management in the following sections: Reactor Vessel Internals, Reactor Coolant System, Reactor Pressure Vessel and Component Supports.
Revision 0 24 Prepared for Constellation Energy by
~kktmmI
~
-I--
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 6.0 ASME SYSTEMS & EXAMINATION BOUNDARIES 6.1 This section defines those systems that are designated as ASME Class 1, 2, or 3 and provides justification for their inclusion or exclusion within the Fourth Ten-Year Inspection Interval NDE Program.
Several of Calvert Cliffs Units 1 & 2 systems or portions of systems are excluded from system and component NDE as allowed by Articles IWB-1200, IWC-1200, and IWD-1200. However, these portions of systems are not excluded from the pressure testing requirements of ASME Code, Section Xl, except as allowed by Articles IWA-5000, IWB-5000, IWC-5000, and IWD-5000.
System information for the NDE Program is provided in Table 6.1, Calvert Cliffs Unit 1 ASME Code Class Systems (Table 6.2, Calvert Cliffs Unit 2 ASME Code Class Systems). The 2004 Edition of the ASME Code, Section Xl defines the inspection requirements for each of the ASME Code Classes within the fourth inspection interval, which began on October 10, 2009.
6.2 Per IWA-1400(a) of the 2004 Edition of Section Xl, it is the owner's responsibility to determine the appropriate Code Classes for each component and to identify the system boundaries subject to inspection.
IWA-1400(a), footnote 1, states that classification criteria are specified in 10 CFR50. This reference is to footnote 9 of 10 CFR50.55a which references Regulatory Guide 1.26 and Section 3.2.2 of NUREG-0800.
6.3 The component classifications of the ASME Code (Class 1, 2, or 3) determine the rules and requirements for inspection and define the Section XI examination boundaries.
Because early vintage nuclear plants were designed and constructed before Section III of the ASME Boiler and Pressure Vessel Code was incorporated into 10 CFR50.55a, the ASME Section Xl Code classifications for ISI may differ from the original design classifications. Therefore, while the ASME Code classifications determine the rules for repairs and replacements and the component inspection requirements, repairs and replacements are generally performed to meet the specifications of the original design code.
6.4 The 2008 Code of Federal Regulations provides criteria for the classification of Quality Group A components.
In previous issues of 10 CFR, this criterion was provided in section 50.2(v). Regulatory Guide 1.26, Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants, provides criteria for the classification of Quality Group B, C, and D components.
Regulatory Guide 1.26 was used for ASME Code, Section Xl component classification at Calvert Cliffs Units 1 and 2.
Revision 0 25 Prepared for Constellation Energy by Itap
~
'o*0io.fo
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program 6.5 The ASME Code Class 1, 2, and 3 systems required to be examined in accordance with ASME Code, Section Xl, 2004 Edition, are identified on isometric drawings. These are Plant Controlled drawings. Section 11 contains the listing of the Ten-Year ISI Program and the Component Support Program Isometric Drawings.
6.6 A review of ISI boundaries for application of the ASME Code,Section XI, was conducted by Constellation Energy, Inc. to assure proper Class 1, 2, and 3 boundaries prior to updating the Inservice Inspection Program to the 2004 Edition of ASME Code, Section Xl Code for the Fourth Inspection Interval.
Table 6.1 Calvert Cliffs Unit 1 ASME Code Class Systems ASME System P&ID #
Code Inspection Requirements
_Class IWB-2000, 5000 Reactor Coolant 60729-SH 1, 2 1, 2, 3 IWC-2000, 5000 CLASS 3 EXEMPT IWB-2000, 5000 Chemical and Volume Control 60730-SH 1, 2, 3 1, 2, 3 IWC-2000, 5000 IWD-2000, 5000 IWB-2000, 5000 Safety Injection 60731-SH 1,2, 3 1,2 IWC-2000, 5000 IWB-2000, 5000 IWC-5000 Containment Spray 60731-SH 1, 2, 3 2
CAS2DEXM CLASS 2 NDE EXEMPT IWC-1221 (d)
IWD-2000, 5000 Post Accident Sampling 60724-SH 1, 2, 3 3
CLASS 2 NDE EXEMPT IWC-1222(a)
Reactor Coolant & Waste RectrCoolasnt &astn 60724-SH 1, 2, 3 1
IWB-2000, 5000 Processing Sampling Revision 0 26 Prepared for Constellation Energy by at d-io G i--.
-- -t
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 6.1 Calvert Cliffs Unit 1 ASME Code Class-Systems ASME System P&ID #
Code Inspection Requirements Class Feedwater 60702-SH 1, 4 2, NNC IWC-2000, 5000 Auxiliary Feedwater 60583-SH 1 2, 3 IWC-2000, 5000 IWD-2000, 5000 Main Steam 60700-SH 1 2
IWC-2000, 5000 IWD-2000, 5000 Condensate System 60717-SH 1 3CLASS 3 NDE EXEMPT IWC-2000, 5000 CLASS 2 NDE EXEMPT Component Cooling Water 60710-SH 1,2, 3 2, 3 2WC-12EXEc P
IWC-1222(c)
IWD-2000, 5000 IWC-2000, 5000 CLASS 2 NDE EXEMPT Service Water Cooling 60706 2, 3 2WC-12EXEc P
IWC-1222(c)
IWD-2000, 5000 60740-SH 1 IWC-2000, 5000 Steam Line Drainage 62740-SH 1 2
CLASS 2 NDE EXEMPT IWC-1222(c) 60712-SH 3, 5, IWC-2000, 5000 Compressed Air 7
2 CLASS 2 NDE EXEMPT IWC-1222(c).
IWC-2000, 5000 Ventilation 60723-SH 1, 2, 4 2
CLASS 2 NDE EXEMPT
_1
_IWC-1222(c)
Revision 0 1 27 Prepared for Constellation Energy by
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 6.1
-Calvert Cliffs Unit I*i ASME Code Class-Systems___________________
ASME System.,
-P&ID #
'Code' Inspection Requirements C lass IWC-2000, 5000 CLASS 2 NDE EXEMPT Nitrogen Generating &
60726 2
IWC-1222(a)
Blanketing CLASS 2 SPT EXEMPT APPENDIX J IWA-5110(C)
IWC-2000, 5000 CLASS 2 NDE EXEMPT Fire Protection 60714-SH 2 12 IWC-1222(c)
CLASS 2 SPT EXEMPT
__APPENDIX J IWA-51 10(C)
IWC-2000, 5000 Spent Fuel Cooling 60716 2,3 CLASS 2 NDE EXEMPT IWC-1222(c)
IWD-2000, 5000 IWC-2000, 5000 CLASS 2 NDE EXEMPT Radiation Monitoring 60738-SH 1 2
IWC-1222(a)
CLASS 2 SPT EXEMPT APPENDIX J IWA-5110(C)
Containment Charcoal Filter IWC-2000, 5000 C m ral 60711 2
CLASS 2 NDE EXEMPT Spray IWC-1222(a)
Circulating Salt Water Cooling 60708 SH-1,2 3
IWD-2000, 5000 IWC-2000, 5000 CLASS 2 NDE EXEMPT Plant Water 60746-SH 2, 3 2
IWC-1222(c)
CLASS 2 SPT EXEMPT APPENDIX J IWA-5110(C)
Revision 0 28 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program Table 6.1 Calvert Cliffs Unit I ASME Code Class Systems ASME System P&ID #
Code Inspection Requirements Class IWC-2000, 5000 Reactor Coolant Waste 60734-SH 1, 2 2
CLASS 2 NDE EXEMPT Processing 60743-SH 32 2WC-2EXEMP IWC-1222(a)
Chemical Addition 60741-SH 3 3
IWD-2000, 5000 IWC-2000, 5000 Gas Analysis 60744-SH 1, 2 2
CLASS 2 NDE EXEMPT IWC-1222(a)
Steam Generator Blowdown 60761 2
IWC-2000, 5000 Recovery Revision 0 29 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 6.2 Calvert Cliffs Unit 2 ASME Code Class Systems ASME System P&ID #
Code Inspection Requirements Class IWB-2000, 5000 Reactor Coolant 62729-SH 1, 2 1, 2, 3 IWC-2000, 5000 CLASS 3 EXEMPT IWB-2000, 5000 Chemical and Volume Control 62730-SH 1, 2, 3 1,2, 3 IWC-2000, 5000 IWD-2000, 5000 Safety Injection 62731-SH 1,2, 3 1,2 IWB-2000, 5000 IWC-2000, 5000 IWB-2000, 5000 Containment Spray 62731-SH 1, 2, 3 2IWC-5000 CLASS 2 NDE EXEMPT IWC-1 221(d)
IWD-2000, 5000 Post Accident Sampling 60724-SH 1, 3 3
CLASS 2 NDE EXEMPT IWC-1222(a)
Reactor Coolant & Waste Patr Cooesng Samln 60724-SH 1, 3 1
IWB-2000, 5000 Processing Sampling Feedwater 62702-SH 4 2, NNC IWC-2000, 5000 Auxiliary Feedwater 62583 2,
IWC-2000, 5000 IWD-2000, 5000 Main Steam 60700-SH 1 2
IWC-2000, 5000 IWD-2000, 5000 Condensate System 60717-SH 1 3
CLASS 3 NDE EXEMPT Revision 0 30 Prepared for Constellation Energy by lkcdlmIll~pgl bing*;.
n
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program Table 6.2 Calvert Cliffs Unit'2 ASME Code Class Systems_
Code Inspection Requirements Class IWC-2000, 5000 CLASS 2 NDE EXEMPT Component Cooling Water 62710-SH 1,2, 3 2,3 C
WCSS222EcM IWC-1 222(c)
IWD-2000, 5000 IWC-2000, 5000
~CLASS 2NDE EXEMPT Service Water Cooling 60706 2, 3 2WC-12EXEc P
IWC-1222(c)
IWD-2000, 5000 IWC-2000, 5000 Steam Line Drainage 62740-SH 1 2
CLASS 2 NDE EXEMPT IWC-1222(c)
IWC-2000, 5000 Compressed Air 62712-SH 3, 5 2
CLASS 2 NDE EXEMPT IWC-1222(c)
IWC-2000, 5000 Ventilation 60723-SH 1, 2, 4 2
CLASS 2 NDE EXEMPT IWC-1222(c)
IWC-2000, 5000 CLASS 2 NDE EXEMPT Nitrogen Generating &
60726 2
IWC-1222(a)
Blanketing CLASS 2 SPT EXEMPT APPENDIX J IWA-5110(C)
IWC-2000, 5000 CLASS 2 NDE EXEMPT Fire Protection 60714-SH 2 2
IWC-1222(c)
CLASS 2 SPT EXEMPT APPENDIX J IWA-5110(C)
Revision 0 31 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program Table 6.2 Calvert Cliffs Unit 2" ASME Code Class Systems ASME System P&ID #
Code Inspection Requirements
___r_
Class IWC-2000, 5000 CLASS 2 NDE EXEMPT Spent Fuel Cooling 60716 2, 3 2WC-12EXEMP IWC-1222(c)
IWD-2000, 5000 IWC-2000, 5000 CLASS 2 NDE EXEMPT Radiation Monitoring 60738-SH 1 2
IWC-1222(a)
CLASS 2 SPT EXEMPT APPENDIX J IWA-51 10(C)
Containment Charcoal Filter IWC-2000, 5000 62711 2
CLASS 2 NDE EXEMPT Spray IWC-1222(a)
Circulating Salt Water Cooling 62708 SH-1, 2 3
IWD-2000, 5000 IWC-2000, 5000 CLASS 2 NDE EXEMPT Plant Water 60746-SH 2, 3 2
IWC-1222(c)
CLASS 2 SPT EXEMPT APPENDIX J IWA-51 10(C) e 60734-SH 1, 2 IWC-2000, 5000 Reactor Coolant Waste 6073-SH 2
CLASS 2 NDE EXEMPT Processing 60743-SH 3 IWC-1222(a)
Chemical Addition 60741-SH 3 3
IWD-2000, 5000 IWC-2000, 5000 Gas Analysis 60744-SH 1, 2 2
CLASS 2 NDE EXEMPT "IWC-1222(a)
Steam Generator Blowdown 62749 2
IWC-2000, 5000 Recovery Revision 0 32 Prepared for Constellation Energy by Ic~ll*ilos
- bIro.
Calvert Cliffs Units 1 & 2 Constellation Energy ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.0 APPLICATION CRITERIA AND CODE COMPLIANCE 7.1 ASME Section XI The following provides a summary of the application of ASME Code, Section Xl, 2004 Edition to the Calvert Cliffs Nuclear Power Plant Units 1 & 2, Ten-Year Program for the Fourth Inspection Interval.
The application and distribution of examinations for this interval is based upon utilizing Inspection Program B as defined by Articles IWB-2412, IWC-2412, and IWD-2412 of Section Xl.
The results of this application are summarized by ASME Category and Item number and are contained within Table 6.1. These tables only contain those ASME Item numbers that are relevant to CCNPP.
7.1.1 EXAMINATION CATEGORY B-A - PRESSURE RETAINING WELDS IN REACTOR VESSEL Reactor vessel examinations were scheduled on the reactor pressure vessels to meet the 2004 Edition of the ASME Code,Section XI as required by 10 CFR50.55a.
Article IWB-2420 in the 2004 Edition of ASME Code, Section Xl requires that the sequence of component examinations be repeated during each successive inspection interval, to the extent practical.
Table IWB-2500-1, Examination Category B-A, note 4 specifies least 50% of the shell to flange weld will be examined by the end of the first inspection period, however, Request for Alternative ISI-020 was submitted to extend the inspection interval to twenty years for these welds. All of the reactor vessel shell welds will be examined near the end of the inspection interval.
Future examinations will be performed in accordance with Appendix VIII of the ASME Section Xl Code. This meets reactor vessel examination requirements in the 2004 Edition of Section XI as modified by Request for Alternative ISI-020.
7.1.2 EXAMINATION CATEGORY B-B - PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS The 2004 Edition of Section XI, Examination Category B-B requires that pressurizer and primary side of steam generators be examined during the inspection interval. The examinations may be limited to one of a group of vessels with similar function.
Revision 0 33 Prepared for Constellation Energy by kIdmlm f'rnuptla* ino,
Calvert Cliffs Units 1 & 2 Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program STEAM GENERATORS There is Class 1 Examination Category B-B, welds on each steam generator.
Table IWB-2500-1, Category B-B, Note (1) allows the examination be limited to one vessel among a group of vessels performing a similar function or 50%. There is one tube sheet to head welds on each S/G. One of two tube sheet to head welds is selected for examination, thus meeting the one vessel among a group of vessels performing a similar function or 50% requirement.
PRESSURIZER Section Xl requires examination of both shell to head welds and one foot of one intersecting longitudinal weld per head.
The CCNPP pressurizer has four longitudinal welds or two per head. Both of the shell to head welds are selected meeting the 100% requirement and one intersecting long seam weld per head is selected meeting the 50% requirement.
This meets the Examination Category B-B examination requirements in the 2004 Edition of Section XI.
7.1.3 EXAMINATION CATEGORY B-D - FULL PENETRATION WELDS OF NOZZLES IN VESSELS This category applies to the reactor vessel, pressurizer, and steam generators.
ASME Code, Section Xl does not allow the deferral of these examinations, except for nozzle to reactor vessel welds. Table IWB-2500-1, Category B-D, Note 3, allows partial deferral if the examinations are conducted from inside the component and the nozzle weld is examined by straight beam UT from the nozzle bore, the examinations required to be conducted from the shell inside diameter may be performed at or near the end of the interval. The reactor vessel nozzle weld examinations will be deferred until the end of the interval as allowed by Request for Alternative ISI-020, which was submitted to extend the inspection interval to twenty years for these welds.
All of the pressurizer and steam generator nozzle to vessel welds and nozzle inner radii are selected for examination meeting the 100% examination requirement.
This meets the Examination Category B-D examination requirements in the 2004 Edition of Section Xl as modified by Request for Alternative ISI-020.
Revision 0 34 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section Xl Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.4 EXAMINATION CATEGORY B-E - PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS Examination Category B-E was removed from the Code in the 1992 Addenda.
However, Code Case N-729-1 uses Examination Category B-E for its examination requirements. Code Case N-729-1 has been mandated by 10 CFR 50.55a. This Code case requires examination of Reactor Vessel Upper Heads.
The heads were replaced in the third interval with a head and nozzles containing PWSCC resistant material. Based on the materials, the reactor vessel head will be visually examined every other refueling outage and the nozzles and partial-penetration welds in the head will be volumetrically examined once this interval.
This meets the Code Case N-729-1 Examination Category B-E examination requirements as mandated in 10 CFR 50.55a.
7.1.5 EXAMINATION CATEGORY B-F - PRESSURE RETAINING DISSIMILAR METAL WELDS IN VESSEL NOZZLES This category addresses Nozzle-to-Safe End Welds and Piping Welds. CCNPP has developed a Code Case N-716 RI-ISI program. All Examination Category B-F Welds have been re-categorized as R-A welds in accordance with Code Case N-716.
Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-04-04.
Therefore no examinations will be performed per Examination Category B-F.
7.1.6 EXAMINATION CATEGORY B-G PRESSURE RETAINING BOLTING, GREATER THAN 2" IN DIAMETER Components with bolting greater than 2" in diameter include the reactor vessel, steam generators, and reactor coolant pumps. Examinations will be conducted as required in Table IWB-2500-1 in the 2004 Edition of Section X1.
This bolting will only be examined if one of the reactor coolant pumps is disassembled. This is the same requirement as for Examination Category B-L-2 and will be met during the repair replacement process.
All other B-G-1 components are scheduled for examination.
This meets the Examination Category B-G-1 examination requirements in the 2004 Edition of Section Xl.
Revision 0 35 Prepared for Constellation Energy by 40 apo.bo
Calvert Cliffs Units 1 & 2 Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.7 EXAMINATION CATEGORY B-G PRESSURE RETAINING BOLTING, 2" AND LESS IN DIAMETER This category includes relief valve bolting, flange bolting in piping, reactor coolant pump seal bolting, valve bolting in the safety injection, chemical and volume control, shutdown cooling, and reactor coolant systems.
Examinations will be conducted as required in Table IWB-2500-1 in the 2004 Edition of Section XI. This bolting will only be examined if associated connections are disassembled.
For components other than piping, bolting will be required only when the component is examined under Examination Category B-B, B-L-2, or B-M-2. This group of bolting is grouped with the components by the "Program Type".
All B-G-2 bolting examinations will be scheduled during the repair replacement process.
This meets the Examination Category B-G-2 examination requirements in the 2004 Edition of Section XI.
7.1.8 EXAMINATION CATEGORY B-J - PRESSURE RETAINING WELDS IN PIPING This category addresses piping welds. CCNPP Units 1 and 2 have developed a Code Case N-716 RI-ISI program.
All Examination Category B-J Welds have been re-categorized, as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-04-04.
Therefore no examinations will be performed per Examination Category B-J.
7.1.9 EXAMINATION CATEGORY B-K - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category B-K of the ASME Code Section Xl, 2004 Edition requires examination of Integral Attachments.
For vessel attachments Note 4 allows for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination.
For single vessels, only one welded attachment shall be selected for examination.
Each Unit has three RPV Attachments, one attachment weld on each Steam Generator and one PZR attachment weld.
One of the three RPV attachment welds, one of the two S/G attachment welds, and the one PZR attachment weld are selected for examination or 50% of all vessel attachment welds.
For piping pumps and valves, inspection of 10% of the total population of integral welded attachments is required. 10% of all piping welded attachments and 10% of all pump welded attachments are selected for examination.
This meets the Examination Category B-K examination requirements in the 2004 Edition of Section Xl.
Revision 0 36 Prepared for Constellation IEnergy by ltkd-aItl Ca bnimIno.
Calvert Cliffs Units 1 & 2 Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.10 EXAMINATION CATEGORIES B-L-1 AND B-L PRESSURE RETAINING WELDS IN PUMP CASINGS, AND PUMP CASINGS This category involves reactor coolant pumps and requires volumetric examination on pump casing welds and visual examination of pump internals when disassembled. The reactor coolant pumps are the only Code Class 1 pumps.
B-L-1 requires examination of one pump in each group of pumps performing similar functions in the system.
One of the four Reactor Coolant Pumps is selected for examination meeting the 25% requirement.
B-L-2 requires examination of one pump in each group of pumps performing similar functions in the system. However, this examination is only required when a pump is dissembled for maintenance, repair, or volumetric examination. No Reactor Coolant Pumps have been selected. This requirement will be met during the repair replacement process.
This meets the Examination Categories B-L-1 and B-L-2 examination requirements in the 2004 Edition of Section XI.
7.1.11 EXAMINATION CATEGORIES B-M-1 AND B-M PRESSURE RETAINING WELDS IN VALVE BODIES, AND VALVE BODIES This category only involves Reactor Coolant (RCS), Shutdown Cooling (SDC), and Safety Injection (SI) valves. Examinations will be conducted as required in Table IWB-2500-1. There are a total of twenty valves in Unit 1 and sixteen in Unit 2 that fall into this examination category. There are considered to be six groups in Unit 1 and four groups in Unit 2 of valves as defined by program type in the IDDEAL Software Suite, ScheduleWorks computer program.
B-M-1 requires examination of one valve in each group of valves that are of the same size, constructural design, and manufacturing method, and that performs similar functions in the system. One of the two Pressurizer Spray valves contains two body welds which are selected for examination meeting the 50% requirement.
B-M-2 requires examination of one valve in each group of valves that are of the same size, constructural design, and manufacturing method, and that performs similar functions in the system. However, this examination is only required when a valve is dissembled for maintenance, repair, or volumetric examination. No valve body internal surfaces have been selected. This requirement will be met during the repair replacement process.
This meets the Examination Categories B-M-1 and B-M-2 examination requirements in the 2004 Edition of Section XI.
Revision 0 37 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.12 EXAMINATION CATEGORY B-N INTERIOR OF REACTOR VESSEL These examinations will be conducted each inspection period.
This meets the Examination Category B-N-1 examination requirements in the 2004 Edition of Section X1.
7.1.13 EXAMINATION CATEGORY B-N WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSEL These examinations will be deferred until the third period as allowed in Table lWB-2500-1. This meets the Examination Category B-N-2 examination requirements in the 2004 Edition of Section X1.
The reactor vessel core support structures and interior attachments are discussed in Request for Alternative ISI-021, which was submitted to extend the inspection interval to twenty years for these welds.
7.1.14 EXAMINATION CATEGORY B-N REMOVABLE CORE SUPPORT STRUCTURES These examinations will be conducted during the third period to coincide with the 10 year inspection of the reactor vessel. This meets the Examination Category B-N-1 examination requirements in the 2004 Edition of Section X1.
The reactor vessel core removable core support structures are discussed in Request for Alternative ISI-021, which was submitted to extend the inspection interval to twenty years for these surfaces.
7.1.15 EXAMINATION CATEGORY B-O - PRESSURE RETAINING WELDS IN CONTROL ROD HOUSING AND INSTRUMENT NOZZLE HOUSINGS The ASME Code, Section Xl, 2004 Edition, requires volumetric or surface examination of 10% of peripheral control rod drive housings during the inspection interval. There are 28 peripheral Control Rod Element Drives (CEDMs) on the reactor vessel head. To meet the Code requirements the four welds on three of the CEDMS have been selected.
This meets the Examination Category B-O examination requirements in the 2004 Edition of Section XI.
7.1.16 EXAMINATION CATEGORY B-P - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at CCNPP meets the requirements of ASME Code, Section Xl, 2004 Edition for Class 1 systems. Details of the component listing are contained in the Pressure Test Program Plan.
Revision 0 38 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.17 EXAMINATION CATEGORY B-Q - STEAM GENERATOR TUBING The Steam Generator Tube inspection program at CCNPP is governed by CCNPP Improved Technical Specification 5.5.9.
7.1.18 EXAMINATION CATEGORY C-A - PRESSURE RETAINING WELDS IN PRESSURE VESSELS This category applies to the steam generators and shutdown cooling heat exchangers. Note (2) in Table IWC-2500-1, Examination Category C-A only requires examination of welds at gross structural discontinuities, such as, junctions between different thickness, shell to flange welds, cylindrical shell to conical shell, and head to shell welds. Steam Generator Weld five is not considered to be a gross structural discontinuity.
The NRC Staff approved this position in third interval relief request ISI-019.
This position has been submitted in the fourth inspection interval in Request for Alternative ISI-04-03. Note (3) states that "In the case of multiple vessels of similar design, size, and service, the required examinations may be limited to one vessel or distributed among the vessels or 50%." All the vessel welds on one S/G and one tubesheet-to-shell and one shell circumferential weld on a SCHE are selected for examination.
This meets the Examination Category C-A examination requirements in the 2004 Edition of Section XI.
7.1.19 EXAMINATION CATEGORY C-B - PRESSURE RETAINING NOZZLE WELDS IN VESSELS This category applies to steam generators and shutdown cooling heat exchangers.
Note (1) in Table IWC-2500-1, Category C-B, excludes manways and handholes.
Note (3) requires that nozzles selected initially for examination shall be reexamined over the service life of the component. Note (5) allows that in the case of multiple vessels of similar design, size, and service the required examinations may be limited to one vessel or distributed among the vessels. All the vessel nozzle welds and inner radii on one S/G and both nozzle welds on a SCHE are selected for examination The Nozzles on the SCHE are < NPS 12 and therefore, the Nozzle Inner Radius do not require examination in accordance with the examination figure.
This meets the Examination Category C-B examination requirements in the 2004 Edition of Section XI.
Revision 0 39 Prepared for Constellation Energy by ldOciiml *nou~m.;
Inm.
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.20 EXAMINATION CATEGORY C-C - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category C-C of the ASME Code Section Xl, 2004 Edition requires examination of Integral Attachments.
For vessel attachments Note 4 allows for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination.
For single vessels, only one welded attachment shall be selected for examination.
Each unit has eight attachment weld on each Steam Generator and one attachment weld on each SCHE. One of the sixteen S/G attachment welds and one of the two SCHE attachment welds are selected for examination or 11% of all vessel attachment welds.
For piping pumps and valves, inspection of 10% of the total population of integral welded attachments is required.
10% of all piping welded attachments are selected for examination.
This meets the Examination Category C-C examination requirements in the 2004 Edition of Section Xl.
7.1.21 EXAMINATION CATEGORY C-F-i-PRESSURE RETAINING WELDS IN AUSTENITIC STAINLESS STEEL OR HIGH ALLOY PIPING This category addresses Class 2 piping welds. CCNPP Unit 1 has developed a Code Case N-716 RI-ISI program.
All Examination Category C-F-1 welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-04-04.
Therefore no examinations will be performed per Examination Category C-F-I.
7.1.22 EXAMINATION CATEGORY C-F PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL PIPING This category addresses Class 2 piping welds. CCNPP Unit 1 has developed a Code Case N-716 RI-ISI program. All Examination Category C-F-2 Welds have been re-categorized as R-A welds in accordance with Code Case N-716. Code Case N-716 has been submitted to the NRC via Request for Alternative ISI-04-04.
Therefore no examinations will be performed per Examination Category C-F-2.
Revision 0 40 Prepared for Constellation Energy by
Calvert Cliffs Units I & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.23 EXAMINATION CATEGORY C-G - PRESSURE RETAINING WELDS IN PUMPS AND VALVES In case of multiple pumps and valves of similar design, size, function and service in a system, Examination Category C-G requires weld examination my be limited to all the welds in one pump or valve. Each Unit has two valve groups with 2 valves each that have one valve body weld.
One valve from each group is selected for examination, meeting the 50% requirement.
7.1.24 EXAMINATION CATEGORY C-H - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at CCNPP meets the requirements of ASME Code, Section Xl, 2004 Edition for Class 2 systems. Details of the component listing are contained in the Pressure Test Program Plan.
7.1.25 EXAMINATION CATEGORY D-A - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category D-A of the ASME Code Section XI, 2004 Edition requires examination of Integral Attachments.
For vessel attachments Note 3 allows for multiple vessels of similar design, function, and service, only the welded attachment of one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination.
Unit 1 has four multiple vessels and Unit 2 has two multiple vessels that list all of the attachments in one component. Each unit also has one multiple vessel that has 2 attachment components. Unit 1 has a single and Unit 2 has four vessels with one integral attachment. One integral attachment is selected on one of the vessels for each group of vessels.
In addition, two of the SCHE integral attachments and the one integral attachment on each of the single vessels are selected for examination. Seven of the thirteen vessel attachment welds on Unit 1 are selected for examination or 52% of all vessel attachment welds. Eight of the twelve vessel attachment welds on Unit 2 are selected for examination or 64% of all vessel attachment welds.
For piping pumps and valves, inspection of 10% of the total population of integral welded attachments is required.
10% of all piping welded attachments are selected for examination.
This meets the Examination Category D-A examination requirements in the 2004 Edition of Section XI.
Revision 0 41 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program 7.1.26 EXAMINATION CATEGORY D-B - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at CCNPP meets the requirements of ASME Code, Section Xl, 2004 Edition for Class 3 systems. Details of the component listing are contained in the Pressure Test Program Plan.
7.1.27 EXAMINATION CATEGORY F-A - SUPPORTS Examination Category F-A of the ASME Code Section Xl, 2004 Edition requires 25% of Class 1 Piping Supports, 15% of Class 2 Piping Supports, and 10% of Class 3 Piping Supports to be examined during the inspection interval.
For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. The supports have been separated by type as defined in Note 1 to Examination Category F-A. This type has been added to the Item number to clearly identify each support by type. Twenty-five percent (25%) of the Class 1 supports have been selected and are prorated by type and system.
Fifteen percent (15%) of the Class 2 supports have been selected and are prorated by type and system. Ten percent (10%) of the Class 3 supports have been selected and are prorated by type and system. For supports other than piping supports the components have been are scheduled as follows: Unit 1 scheduled the PZR, nine of eighteen S/G, one of the two SCHEs, the SWHE, one of four SRWHX, one of two SWHT, one of two CCWHE, the CCWHT, one of two DACHE, one of two L.O.
Coolers, one of two JWC, one of four, SCHE, six of twenty-four RCPs. Twenty-six of the sixty-five non piping supports are selected for examination or 40% of all non piping supports. Unit 2 scheduled the PZR, nine of eighteen S/G, three of the six SCHEs, one of two SWHE, one of four SRWHX, one of two SWHT, one of two CCWHE, the CCWHT, the DACHE, the L.O. Cooler, the JWC, six of twenty-four RCPs. Twenty-seven of the sixty-three Unit 2 non piping supports are selected for examination or 42.8% of all Unit 2 non piping supports.
This meets the Examination Category F-A examination requirements in the 2004 Edition of Section XI.
7.1.28 EXAMINATION CATEGORY R-A The alternative Code Case N-716, RIS-B Program for piping as described in Relief Request ISI-04-04.
The RIS-B Program has been substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. These welds are selected as provided in Revision 0 42 Prepared for Constellation Energy by kkSaId I InbM
Calvert Cliffs Units 1 & 2 Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program Structural Integrity Associates, Inc.
(SI)
Calculation 0800144.301, N-716 Evaluation of CCNPP Units 1 and 2.
7.2 Augmented Pro-gram 7.2.1 RG 1.14 Reactor Coolant Pump Flywheel Inspections The Reactor Coolant Pump Motor flywheels will be inspected as required by CCNPP Technical Specification 5.5.7. Inservice inspection of each reactor coolant pump flywheel shall be performed, at least once every ten years with the high stress keyways at approximately 3 and 1/3 years intervals. The RCP Flywheels are scheduled each period to meet these augmented requirements.
7.22 MRP-139 Category C Welds Welds of Non-Resistant Material that have been mitigated by SI have a MRP-139 requirement of 50% of each mitigation type within next 6 years; if no indication, continue with existing Code examination program or approved.alternative.
CCNPP Unit 1 has 4 and Unit 2 has 6 welds that are Category C welds. Two welds in Unit 1 and three in Unit 2 are Code Item Number R1.20. Two welds in Unit 1 and three in Unit 2 have multiple Code Item Numbers R1.11 and R1.15.
One weld in Unit 1 and two in Unit 2 of R1.11 and R1.15 and one weld of R1.20 are selected for examination under the MRP Category C Augmented Program.
This meets the Augmented Program MRP Category C examination requirements.
7.2.3 MRP-139 Category E Welds Cold Leg Welds >4 inches of Non-Resistant Material that have not been Mitigated by SI have a MRP-139 requirement of 100% examination each 6 years.- Each unit has nineteen welds that are Category E welds that are either Code Item Number R1.11 or R1.20.
These welds are selected for examination under the MRP Category E Augmented Program. Note that eighteen of the nineteen Unit 1 welds and thirteen of the nineteen Unit 2 welds are also MRP Category K welds.
This.meets the Augmented Program MRP Category E examination requirements.
7.2.4 MRP-139 Cateqory F Welds Welds of Non-Resistant Material that are cracked and have been reinforced by full structural weld overlay have a MRP-1 39 requirement of once in the next 5 years; if no additional indications/growth, continue with existing Code examination program for unflawed condition or approved alternative. CCNPP Unit 2 has two welds that Revision 0 43 Prepared for Constellation Energy by
Calvert Cliffs Units 1 & 2 0
Constellation Energy-ASME Section XI Calvert Cliffs Nuclear Power Plant Fourth Interval Inservice Inspection Program are Category F welds that are Code Item Number R1.20. Both of these welds are selected for examination under the MRP Category G Augmented Program.
This meets the Augmented Program MRP Category F examination requirements.
7.2.5 MRP-139 Category G Welds Welds of Non-Resistant Material that are cracked and have been mitigated by stress improvement have a MRP-139 requirement of 100% each two Refueling Outages. If no additional indications/growth after the 2 nd examination (4 th RFO),
continue with existing Code examination program for unflawed condition or approved alternative. CCNPP Unit 1 has 3 welds that are Category G welds two that are Code Item Number R1.20, one that has multiple Code Item Numbers R1.11 and R1.15. All three of these welds are selected for examination under the MRP Category G Augmented Program.
This meets the Augmented Program MRP Category G examination requirements.
7.2.6 MRP - 139 Category K Welds Reactor vessel (RV) cold leg welds of Non-Resistant Material have a MRP-139 requirement of 100% at least once every three RFOs (not counting RFOs when weld is examined volumetrically as one of the three) or until mitigated or replaced.
Alternatively, for the RV cold leg, or inlet nozzles ONLY, use deterministic analysis as a basis to allow these nozzle welds to be visually examined once per interval.
This option can only be exercised AFTER welds have been UT-examined and fully meet the conditions for being defined as Category E.
In RFOs where a UT is performed from the OD, a visual examination is credited. If the UT is performed from the ID, a visual examination may be credited if the 90% examination volume identified in section 5.1.5 was obtained. CCNPP Unit 1 has nineteen welds and Unit 2 has thirteen welds that are Category K welds that are either Code Item Number R1.11 or R1.20. These welds are selected for examination under the MRP Category K Augmented Program.
Note that eighteen Unit 1 welds and thirteen Unit 2 welds of the nineteen welds are also MRP Category E welds.
'this meets the Augmented Program MRP Category K examination requirements.
7.2.7 RI-BER The purpose of this augmented program is to perform examinations on piping subject to High Energy Line Break (HELB) analysis criteria. Those examinations previously performed on HELB piping per NUREG-0800, Section 3.6.2 will be maintained and performed under the risk-informed break exclusion region (RI-Revision 0 44 Prepared for Constellation Energy by 2r~llSl~I £i~no t ull In
Constellation Energy*
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program BER) application during the current interval.
These welds are selected as provided in EPRI TR 1006837, Application of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs.
Revision 0 45 Prepared for Constellation Energy by
Constellation Energy'
,Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 1. Code Category Summary N bo Required to NumberEto NumberNto Number to NumbNumfber tNuber tNubert ItmEa ubro.be Examined Examination Number to be be.be.eI Cagory Nmer Description Etho Components in Percentage Examined in Examined Examined Examined
,atgr Number Method During Item Exam CoItem No.
IntDring Required Interval in First in Second in Third Interval Period Period Period B-A Pressure Retaining Welds in Reactor Vessel Reactor Vessel B-A B1.11 Circumferential Shell Volumetric 3
3 100%
All Welds 0
0 3
_ Welds B-A BI.12 Reactor Vessel B-A B1.12 Lon Sell Volumetric 9
9 100%
All Welds 0
0 9
Longitudinal Shell Welds Reactor Vessel Accessible B-A B1.21 Circumferential Head Volumetric 1
1 100%
Length of All 0
0 1
Welds Welds Reactor Vessel Accessible B-A B1.22 Meridional Head Welds Volumetric 6
6 100%
Length of All 0
0 6
Welds B-A B1.30 Reactor Vessel Shell-to-Volumetric 1
1 100%
Weld 0
0 1
Flange Weld B-A B1.40 Reactor Vessel Head-to-Volumetric 1
1 100%
Weld 0
0 1
1 Flange Weld and Surface Category Total 21 21 0
0 21 B-B Pressure Retaining Welds in Vessels Other Than Reactor Vessels Pressurizer Shell-to-B-B B2.11 Head Welds Volumetric 2
2 100%
Both welds 1
0 1
Circumferential Pressurizer Shell-to-1 ft (300 mm)
B-BHead Welds Longitudinal Volumetric 4
2 50%
of one weld 1
0 1
Hd 1
WeldsLongit l
1 1
1 per head (2) I I
Revision 0 44 Prepared for Constellation Energy by
Constellation Energy-
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary
'Number to Number to Number to Number of Required to Examination Number to be be be be Item Exam be Examined Category Description Components in During Percentage Examined in Examined Examined Examined
- Number De nMethod Item No.
Required Interval in First in Second in Third Interval Period Period Period Steam Generators B-B B2.40 (Primary Side)
Volumetric 2
1 50%
Weld(l) 0 1
0 Tubesheet-to-Head Weld Category Total 8
5 2
1 2
Note 1: The examination may be limited to one vessel among the group of vessels performing a similar function. (Ref. Table IWB-Notes for Cat. B-B 2500-1, Examination Category B-B, Note 1)
Note 2: The Pressurizer has 2 heads with 2 welds each. Only one weld per head is required, therefore 50%
B-D Full Penetration Welded Nozzles in Vessels B-D B3.100 Reactor Vessel Nozzle Volumetric 6
6 100%
Same as 1st 0
6 Inside Radius Section Interval B-D B3.110 Pressurizer Nozzle-to-Volumetric 4
4 100%
Same as 1st 1
3 0
Vessel Welds Interval B-D B3.120 Pressurizer Nozzle Volumetric 4
4 100%
All nozzles 1
3 0
Inside Radius Section or visual Steam Generators Same as 1st B-D B3.130 (Primary Side) Nozzle-to-Volumetric 6
6 100%
Interval 6
0 0
Vessel Welds Steam Generators B-D B3.140 (Primary Side) Nozzle or visual 6
6 100%
All nozzles 6
0 0
Inside Radius Section B3.90 Reactor Vessel Nozzle-Volumetric 6
6 100%
Same as 1st 0
0 6
-to-Vessel Welds I
I I
I Interval Category Total 32 32 14 6
12 Revision 0 45 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Caegry Item Examriptioxaminxa Itemts n
During Percentage Examined in Examined Examined Examined SItem NO.
Required Interval in First in Second in Third Interval Period Period Period B-E PWR Reactor Pressure Vessel Upper Head Every third Reactor Vessel Head refueling and Nozzle partial outage or 5 B-E B4.30 Penetration Welds of Visual, VE 1
1 100%
calendar 1
0 0
PWSCC Resistant
- years, Material whichever is less All nozzles, Reactor Vessel Head not to exceed Nozzles and Partial
.one B-E B4.40 Penetration Welds of VOL &/or 1
100%
inspection 0
0 1
NozlsanrPriainterval PWSCC Resistant Material (nominally 10 Calendar years)
Category Total 2
2 0
1 B-G-1 Pressure Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter B-G-1 B6.10 Reactor Vessel Closure Same as for 0
0 B-G-1 B1 Head Nuts Visual, VT-i 100%
1st interval Revision 0 46 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 1 Code Category Summary Nume oNumber to Number to Number to Number of Required to Examination Number to be be be be ate Item Description Exam Components in be Examined Percentage Examined in Examined Examined Examined CoItem No.
During Required Interval in First in Second in Third C
Period Period Period Steam Generators Same as for B-G-1 B6.100 Flange Surface when Visual, VT-1 4
0 0%
1st interval(l) 0 0
0 connection disassembled B-G-1 B6.110 Steam Generators Nuts, Visual, VT-1 4
4 100%
Same as for 0
2 2
Bushings, and Washers 1_st interval B-G-1 B6.180 Pumps Bolts and Studs Volumetric 4
0 0%
(1)(2) 0 0
0 Pumps Flange Surface B-G-1 B6.190 when connection Visual, VT-1 4
0 0%
(1)(2) 0 0
0 disassembled Reactor Vessel Closure Same as for B-G-1 B6.20 Studs 1st interval 1
0 0
Pumps Nuts, Bushings, Visual, VT-I 4
0 0%
(1)(2) 0 0
0 B-G-1 B6.200 and Washers Vsa__40%()200 B-G-1 B6.40 Reactor Vessel Threads Volumetric 1
1 100%
Same as for 0
0 1
in Flange 1st interval B-G-1 B6.50 Reactor Vessel Closure Visual, VT-1 1
1 100%
Same as for 1
0 0
Washers, Bushings 1st interval 1
B-G-1 B6.90 Steam Generators Bolts Volumetric 4
4 100%
Same as for 0
2 2
_____and Studs 1st interval Category Total 28 12 3
4 5
Note 1: Not Required unless disassembled Notes for Cat.B-G-1 Note 2: For heat exchangers, piping, pumps, and valves, examinations are limited to components selection for examination under Examination Categories B-B, B-J, B-L-2, and B-M-2.
(Ref. Table IWB-2500-1, Examination Category B-G-1, Note 3)
Revision 0 47 Prepared for Constellation Energy by
Constellation Energy"
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit i Code Category Summary Number to Number to Number to Item Number of Required to Examination Number to be be be be Category Number Description Exam Components in Examined Percentage Examined in Examined Examined Examined Item No.
IntDring Required Interval in First in Second in Third
,Interval
_Period Period Period B-G-2 Pressure Retaining Bolting, 2in. (50 mm) and Less in Diameter Same as for B-G-2 B7.20 Pressurizer Bolts, Studs, Visual, VT-1 9
0 0%
1st interval 0
0 0
and Nuts (1)(3)
Same as for BG2 B.0 Piping Bolts, Studs, andSaesfo Pipin Bolts Nuts Visual, VT-1 6
0 0%
1st interval 0
0 0
(1)(2)(3)
Pumps Bolts, Studs, and Same as for B-G-2 B7.60 Nuds Visual, VT-1 4
0 0%
1st interval 0
0 0
(1)(2)(3)(4)
Valves Bolts, Studs, and Same as for B-G-2 B7.70 Nuts Visual, VT-1 32 0
0%
1st interval 0
0 0
Nuts_
(1)(2)(3)(5)
Category Total 51 0
0 0
0 Note 1: Not required unless disassembled Note 2: For heat exchangers, piping, pumps, and valves, examinations are limited to components selection for examination under Notes for Cat.B-G-2 Examination Categories B-B, B-J, B-L-2, and B-M-2. (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 2)
Note 3: Examination is only required once per interval Note 4: Only one pump of each group of pumps is required as outlined in B-L-2 Note 5: Only one valve of each group of valves is required as outlined in B-M-2 Revision 0 48 Prepared for Constellation Energy by
Constellation Energy
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary RqietoNumber to Number to Number to Number of Required to Examination Number to be be be be Item DExam Nube Examined Category Description Exam in bPercentage Examined in Examined Examined Examined Method op During Required Interval in First in Second in Third NumbrMehd Item No.
Interval IPeriod Period Period B-K Welded Attachments for Vessels, Piping, Pumps, and Valves Pressure Vessels Same as for B-K B10.10 Pesue V eses Surface 6
3 50%
1st interval 1
2 0
Welded Attachments (1)(2)
B-K B10.20 Piping Welded Surface 45 5
10%
Same as for 2
0 3
Attachments 1st interval B-K B10.30 Pumps Welded Surface 16 2
10%
Same as for 2
0 0
I Attachments I
I I
1st interval Category Total 67 10 5
2 3
Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels tB-K shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination.
Note 2: There are three welded attachments on the RPV, 1 on PZR and 1 on each S/G for a total of 6. Must examine 1 on each vessel or multiple vessels, therefore 3 are required for 50%
B-L-1 Pressure Retaining Welds in Pump Casings P
p PSame as for B-L-1 B12.10 Pumps Pump Casing Visual, VT-1 4
1 25%
first interval(l 1
0 0
Weld (B--1)(2)
Category Total 4
1 1
0 0
Notes for Cat. B-L-1 Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.
Note 2: There are four Reactor Coolant Pumps. Each pump has one casing weld, therefore 1 weld of 4 is required for 25%
Revision 0 49 Prepared for Constellation Energy by
Constellation Energy"
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary i
I Numberto Numberto Numberto Number of Required to Examination Number to be be be be Category Number Description Method Components in During Percentage Examined in Examined Examined Examined N
Item No.
Durval Required Interval in First in Second in Third
_nterval Period Period Period B-L-2 Pump Casings p
P p
CSame as for B-L-2 B12.20 Pumps Pump Casing Visual, VT-3 4
0 0%
first interval(I) 0 0
0 (B-L-2)(2)
Category Total 4
0 0
0 0
Notes for Cat. B-L-2 Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.
Note 2: Not required unless disassembled B-M-1 Pressure Retaining Welds in Valve Bodies Valves, Less than NPS 4 Same as for B-M-1 B12.30 (DN 100) Valve Body Surface 4
2 50%
first interval 0
2 0
Welds (B-M-1)
(1)(2)(3)
Category Total 4
2 2
0 Note 1: Examination is limited to at least one valve in each group of valves that are of the same size, constructural design, Notes for Cat.
manufacturing method, and that perform similar functions in the system.
Note 2: There is one valve group that has two valve body welds on each valve. Each valve has two body welds, therefore 2 welds of 4 are required for 50%
Note 3: Since both scheduled welds are in the same valve body they will be examined in the same outage.
Revision 0 50 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number Of Required to Examination Number to be be be be Category Description Components in During Percentage Examined in Examined Examined Examined Caeoy Item DsrpinExam Copnet beEaiein___________
Number Method Drn Item No.
Required Interval in First in Second in Third Int
_Period Period Period B-M-2 Valve Bodies Valve Body, Exceeding Same as for B-M-2 B12.50 NPS 4 (DN 100) (B-M-2) Visual, VT-3 14 0
0%
first interval 0
0 0
(1)(2)
Category Total 14 0
0 0
0 Note 1: Examination is limited to at least one valve in each group of valves that are of the same size, constructural design, B-M-2 manufacturing method, and that perform similar functions in the system.
Note 2: Not required unless disassembled B-N-1 Interior of Reactor Vessel Ieao V l VEach B-N-1 B13.10 Reactor Vessel, Vessel Visual, VT-3 1
1 100%
inspection 1
1 1
Interior (B-N-1) period Category Total 11111 B-N-2 Welded Core Support Structures and Interior Attachments to Reactor Vessels Reactor Vessel (PWR)
Interior AttachmentsSaesfo B-N-2 B1 3.50 IneirAtcmns Visual, VT-I 1
1 100%
Same as for 0
0 1
Within Beltline Region 1st interval (B-N-2)
Reactor Vessel (PWR)
B-N-2 B13.60 IneirAtcmns.
Visual, VT-3 1
1 100%
Same as for 0
0 1
Interior Attachments 1aest intrva Beyond Beltline Region V-stinterval (B-N-2)
I I
I I
I I
I Revision 0 51 Prepared for Constellation Energy by AN" -ýM
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary I
t Numberto Numberto Numberto I
E Number of Required to Examination Number to be be be be, Category Number Description Mex Components in be Examined Percentage Examined in Examined Examined Examined Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period Category Total 2
2 0
0 2
B-N-3 Removable Core Support Structures Reactor Vessel (PWR)
Same as for B-N-3 B13.70 Core Support Structure Visual, VT-3 1
1 100%
1st interval 0
1 (B-N-3)
Category Total 1
1 0
0 1
B-O Pressure Retaining Welds in Control Rod Drive and Instrument Nozzle Housings 10%
Reactor Vessel (PWR)
Volumetric peripheral B-O B14.20 Welds in Control Rod 112 12 10%
CRD 0
0 12 Drive CRD Housing housings (1)
Category Total 112 12 0
0 12 Notes for Cat. B-0 Note 1: There are 28 Peripheral CEDMs, 10% of 28 is 3. There are 4 welds per CEDM for a total of 112 welds. Therefore, 12 welds are selected out of 112 for 10%
Revision 0 52 Prepared for Constellation Energy by kft"-U--ýM
Constellation Energy*
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary r
TNumber to Number to Number to Numbe Required to Examination Number to be be be be Category Item Description Exam Components in be Examined Percentage Examined in Examined Examined Examined Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period C-A Pressure Retaining Welds in Pressure Vessels Pressure Vessels Shell Volumetric Each C-A C1.10 Circumferential Welds (5) 8 4
50%
inspection 2
1 1
interval(1 )(2)
Pressure Vessels Head Volumetric Each C-A C1.20 Circumferential Welds (5) 50%
inspection 1)0 0
interval(1 )(3)
Pressure Vessels Volumetric Each C-A Cl.30 Tubesheet-to-Shell Weld (5) 4 2
50%
inspection 0
1 1
Tubeshee-to-Shel Weld
- 5)
__________interval(1
)(4)
Category Total 14 7
3 2
2 Note 1: The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref. Table IWC-2500-1, Examination'Category C-A, Note 3)
Note 2: There are three circumferential shell welds on both S/G & 1 circumferential shell weld on both SCHE, requiring 3 S/G and 1 SCHE shell welds be scheduled or 4/8 shell welds are required to be examined for 50%
Note 3: There is one circumferential head weld on both S/G, therefore, 1/2 head welds are required to be examined or 50%
Note 4: There is one tubesheet-to-shell weld on both S/G & one tubesheet-to-shell weld on both SCHE, therefore 2/4 tubesheet-to-shell welds are required to be examined for 50%
Revision 0 53 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Required to Examination Number to be be be be Item...
Number oE be Examined IcaExaminatimn NumbinedtExamin Category Number Description Method Components in During Percentage Examined in Examined Examined Examined NumbrNMehodteemao.Required Interval in First in Second in Third Interval Period Period Period C-B Pressure Retaining Nozzle Welds in Vessels Nozzles Without Reinforcing Plate in Vessels > 1/2in. (13mm)
Surface and Each C-B C2.21 Nominal Thickness volumetric 8
4 50%
inspection 1
1 2
Nozzle-to-Shell (Nozzle interval(1)(2) to Head or Nozzle to Nozzle) Weld Nozzles Without Reinforcing Plate in Each C-B 02.22 Vessels > 1/2 in. (13mm)
Volumetric 8
2 25%
inspection 1
1 0
Nominal Thickness Nozzle Inside Radius interval(I)
Section Category Total 16 6
2 2
2 Note 1: The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref. Table IWC-2500-1, Examination Category C-B, Note 4)
Notes for Cat. C-B Note 2: There are two nozzles on both S/G and two nozzles on both SDC HX, therefore 4/8 nozzles are required to be examined or 50%
Note 3: There are two nozzle inner radii on both S/G, however, the two nozzle inner radii on both SDC are not required <12",
therefore 2/8 inner radii are required to be examined or 25%
Revision 0 54 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Item Description Exam ompen Enin Cate Number
__ho_
Com s
m be Examined Percentage Examined in Examined Examined Examined goy N m e ehd Item No.
During Required Interval in First in Second in Third Interval Period Period Period C-C Welded Attachments for Vessels, Piping, Pumps, and Valves Each identified
-*occurrence Pressure Vessels C-CWelded Attachments()
Surface 18 2
11%
and each 0
1 1
inspection interval (1)(2)
(3)
Each identified C-C C3.20 Piping Welded Surface 294 30 10%
occurrence 10 11 9
Attachments(1) and each inspection interval (1)(3)
Category Total 312 32 10 12 10 Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. (Ref.
Table IWC-2500-1, Examination Category C-C, Note 4)
Notes for Cat. C-C Note 2: There are eight Welded Attachments on both S/G and one welded attachment on both SCHE, therefore 2/18 welded attachments are required to be examined for 11%
Note 3: Examination is required whenever component support member deformation is identified. (Ref. Table IWC-2500-1, Examination Category C-C, Note 6)
Revision 0*
55 Prepared for Constellation Energy by
Constellation Energy
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Category Nmer Description E
Components in be Examined Percentage Examined in Examined Examined Examined tNumber Method Item NO.
During Required Interval in First in Second in Third
.Interal Period Period Period C-G Pressure Retaining Welds in Pumps and Valves Valves Valve Body T
Each C-G C6.20 Welds(2)
Surface 4
2 50%
inspection 0
1 1
Welds(2)interval(1
)(2)
Category Total 4
2 0
1 1
Note 1: Examination is limited to all the welds in one valve in each group of valves that are of similar size, design, function, and Notes for Cat. C-G I service. (Ref. Table IWC-2500-1, Examination Category C-G, Note 1 N Note 2: There are two valve groups with two valves each, therefore both welds on two of the four valves are required to be examined for 50%
D-A Welded Attachments for Vessels, Piping, Pumps, and Valves Each identified Pressure Vessels occurrence D-A D1.10 Welded Attachments Visual, VT-1 13 7
52%
and each 2
1 4
inspection interval (1)(2)(3)
Each identified D-A D1.20 Piping Welded Visual, VT-1 548 55 10%
occurrence 9
25 22 Attachments and each inspection interval(3)
Revision 0 56 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary RNumber to Number to Number to Required to Examination Number to be be be be SItem Exam Number of.be ExaminedExmnto ubrbe ebee Category ImDescription Eumbners on Percentage Examined in Examined Examined Examined eNumber Method omponents in During Required Interval in First in Second in Third I
oePeriod Period.
Period Category Total 561 62 11 26 26 Note 1: For multiple vessels of similar design, function, and service, the welded attachments of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. (Ref. Table IWC-2500-1, Examination Category D-A, Note 3)
Notes for Cat. D-A Note 2: There are four groups of multiple vessels (each with two vessels), one single vessel, and two similar vessels with two entries each, therefore, 7/13 are required to be examined for 53%
Note 3: Examination is required whenever component support member deformation is identified. (Ref. Table IWD-2500-1, Examination Category D-A, Note 4)
F-A Supports Each F-A FI.10A Class 1 Piping Supports-Visual, VT-3 27 7
(1) inspection 2
1 4
One Directional interval Each F-A F.10B Class 1 Piping Supportsa-Visual, VT-3 47 12 (1) inspection
.4 5
3 Multi-directional interval Each F-A F1.10C Class 1 Piping Supports-Visual, VT-3 84 21 (1) inspection 8
5 8
Thermal Movement interval Each F-A F1.10 Total Class 1 Piping
- Visual, 158 40 25%
inspection 14 11 15 Supports VT-3 interval Each F-A FI.20A Class 2 Piping Supports-Visual, VT-3 187 29 (1) inspection 12 12 4
One Directional interval Revision 0 57 Prepared for Constellation Energy by
Constellation Energy, Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Category Item Description Mexm Components in be Examined Percentage Examined in Examined Examined Examined o tem No.
During Required Interval in First in Second in Third IntervalPeriod Period Period Each F-A F1.20B Class 2 Piping Supports - Visual, VT-3 269 41 (1) inspection 16 18 6
Multi-directional interval Each F-A F1.20C Class 2 Piping Supports - Visual, VT-3 97 15 (1) inspection 5
5 5
Thermal Movement interval Each F-A F1.20 Total Class 2 Piping
- Visual, 553 83 15%
inspection 33 35 15 Supports VTr-3 interval Class 3 Piping Supports -
Each F-A FI.30A One Diping Visual, VT-3 214 22 (1) inspection 7
7 7
One Directional interval Each F-A F1.30B Class 3 Piping Supports - Visual, VT-3 580 58 (1) inspection 14 23 21 Multi-directional interval Each F-A F1.30C Class 3 Piping Supports - Visual, VT-3 53 6
(1) inspection 2
1 3
Thermal Movement interval Each F-A F1.30 Class 3 Piping
- Visual, 847 85 10%
inspection 23 31 31 Supports VT-3 interval Supports other than Each F-A F1.40 Piping Supports (Class
- Visual, 65 26 40%
inspection 2
9 15 VT-3 1,2,3, and MC) interval(2)(3)
Category Total 1623 234 72 86 76 Revision 0 58 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Item Exam Nubro be Examined Category Num Description m
Components in Puring percentage Examined in Examined. Examined Examined Number Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period Note 1: The total percentage sample shall be comprised of supports form each system, where the individual sample sizes are proportional to the total number of non-exempt supports of each type and function within each system. (Ref. Table IWF-2500-1., Examination Category F-A, Note 2)
Note 2: For multiple components other than piping, within a system of similar design, function, and service, the supports of on!y one of the multiple components are required to be examined. (Ref. Table IWF-2500-1, Examination Category F-A, Note 3)
Note 3: One PZR, nine of eighteen S/G, one of the two SCHEs; the SWHE, one of four SRWHX, one of two SWHT, one of two CCWHE, the CCWHT, one of two DACHE, one of two L.O. Coolers, one of two JWC, two of four SCHE, six of twenty-four RCPs. Therefore 26/65 component supports are required to be examined for 40%.
R-A Risk Informed Piping Welds N-716 LSS Elements that R-A-RO.00 have not been Evaluated 1336 0
0%
0 0
0 for Degradation R-A
.R1.1 1 N-716 Elements Subject Volumetric 55 21 38%
Element (1) 15 3
3
_-A_
R1.11to Thermal Fatigue Volumetri 55_21_38%
Element_(1)_15_3_
N-716 Elements Subject R-A R1.11-15 to Thermal Fatigue and Volumetric 3
2 66%
Element (1) 2 0
0 PWSCC N-716 Elements Subject to Intergranular or R-A R1.16 Transgranular Stress Volumetric 4
1 25%
Element (1) 0 0
1 Corrosion Cracking (IGSCC, TGSCC)
N-716 Elements not R-A R1.20 Subject to a Damage Volumetric 570 54 9.47%
Element (1) 18 13 23 Mechanism I
Revision 0 59 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary RuetNumber to Numberto Numberto Number of Required to Examination Number to be be be be Category Number Description Exam Components in be Examined Percentage Examined in Examined Examined Examined o M tem No.
During Required Interval in First in Second in Third nerva Period Period Period N-716 Elements not R-A R1.20S Subject to a Damage Visual, VT-2 94 1
1%
Element (1) 0 0
1 Mechanism I
I I
I I
I I
_I Category Total 2062 79 35 16 28 Notes for Cat. R-A Note 1: Percentages were determined using SI Calc 0800144.301 Revision 0.
MRP-139C Mitigated Welds of Non-Resistant Material MRP 139 Category C Welds that are in the RI-R-A R1.11-15 ISI Program as Volumetric 2
1 50%
1 0
0 Susceptible to Thermal Fatigue and PWSCC MRP 139 Category C R-A R1.20 Welds that are in the RI-Volumetric 2
1 50%
1 0
0 ISI Program as no DM Category Total 4
1 2
0 0
Notes for MRP-Note 1: Welds of Non-Resistant Material that have been mitigated by SI have a MRP-1 39 requirement of 50% of each mitigation type 139C within next 6 years MRP-139E Non Mitigated Welds of Non-Resistant Material MRP 139 Category E Welds that are in the RI-Volumetric R-A R1.11 ISI Program as (MRP 139 2
4 (1) 2 0
2 Susceptible to Thermal Cat E)
Fatigue Revision 0 60 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be
. be Category Iter Description exam Components in During Percentage Examined in Examined Examined Examined Number Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period MRP 139 Category E Volumetric R-A R1.20 Welds that are in the RI-(MRP 139 17 33 (1) 16 1
16 ISI Program as no DM Cat E)_
Category Total 19 37 (1) 18 1
18 Notes for MRP-Note 1: Cold Leg Welds >4 inches of Non-Resistant Material that have not been Mitigated by SI have a MRP-1 39 139E requirement of 100% examination each 6 years MRP-139G Welds of Non-Resistant Material that are cracked and have been Mitigated by SI MRP 139 Category G Welds that are in the RI-R-A R1.11-15 ISI Program as Volumetric 1
3 (1) 1 1
1 Susceptible to Thermal Fatigue and PWSCC MRP 139 Category G R-A R1.20 Welds that are in the RI-Volumetric 2
6 (1) 2 2
2 1 ISI Program as no DM Category Total 3
9 (1) 3 3
3 Note 1: Welds of Non-Resistant Material that are cracked and have been Mitigated by SI have a MRP-139 requirement of 100% each 139G for M2 Refueling Outages. If no additional indications/growth after the 2nd examination (4th RFO), continue with existing Code 1 39G examination program for unflawed condition or approved alternative MRP-139K Cold Leg Welds of Non-Resistant Material Revision 0 61 Prepared for Constellation Energy by
Constellation Energy
,Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit I Code Category Summary Item Desc o
red to Number to Number to Number to Ite Exam Nube of be Required t Examination Numberto be be
.be be Category.
Numberpo M
od pbe Examined Percentage Examined in Examined Examined Examined Exam Components in During Required Interval in First in Second in Third CateoryNumbr Dscrptio Mehod temNo.
Interval Period Period Period MRP 139 Category K Welds that are in the RI-Volumetric R-A R1.11 ISI Program as (MRP 139 2
2 (1) 0 2
0 Susceptible to Thermal Cat K)
Fatigue MRP 139 Category K Volumetric R-A R1.20 Welds that are in the Rl-(MRP 139 17 18 (1) 1 16 1
Category Total 19 20 (1) 1 18 1
Notes for MRP-Note 1: Cold Leg Welds of Non-Resistant Material have a MRP-1 39 requirement of 100% at least once every three (3) RFOs (not 139K counting RFOs when weld is examined volumetrically as one of the three) or until mitigated or replaced.
RG 1.14 RCP FLYWHEEL RCP RG1.14 RCP Flywheels UT, ECT 4
4 100%
4 4
4 Category Total 4
12 (1) 4 4
4 NotefRG114 Note 1: Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every ten years with the high stress keyways at approximately 3 and 1/3 years intervals RI-BER R-A R1.20 Volumetric 139 3
2%
1 1
1 Category Total 139 3
(1) 1 1
1 Notes for RI-BER Notel: These welds are selected as provided in EPRI TR 1006837 Revision 0 62 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Numnber~to
"~~ ~ ~~~
Ul
"-"Rqied tol I.
be____*
Number of Requir Examination Number to be be be be Ct, Itemn ecito Exam o*be Examitnumeeodubr oNmbrt bgory e
DescriptionN Metho Components in Dunnamed Percentage Examined in Examined Examined Examinred:
Category N
rtItem No.
.Required Interval in First in Second inThird
.. Interval Period Period Period B-A Pressure Retaining Welds in Reactor Vessel Reactor Vessel B-A B1.11 Circumferential Shell Volumetric 3
3 100%
All Welds 0
0 3
Welds B-A B1.12 Reactor Vessel Volumetric 9
100%
All Welds 0
0 Longitudinal Shell Welds Reactor Vessel Accessible B-A B1.21 Circumferential Head Volumetric 1
1 100%
Length of All 0
0 1
Welds Welds SReactor Vessel Accessible B-A B1.22 Rer Vessel s
Volumetric 6
6 100%
Length of All 0
0 6
Meridional Head Welds Welds B-A B1.30 Reactor Vessel Shell-to-Volumetric 1
1 100%
Weld 0
0 1
Flange Weld B-A B1.40 Reactor Vessel Head-to-Volumetric 1
1 100%
Weld 0
0 1
Flange Weld and Surface Category Total 21 21 0
0 21 B-B Pressure Retaining Welds in Vessels Other Than Reactor Vessels Pressurizer Shell-to-B-B B2.11 Head Welds Volumetric 2
2 100%
Both welds 1
0 1
Circumferential Revision 0 63 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Item Number of Required to Examination Number to be be be be Category Number Description Method Components in be Examined Percentage Examined in Examined Examined Examined ExaM Item N
During Required Interval in First in Second in Third interval Period Period Period 1 ft (300 mm)
B-B B2.12 Pressurizer Shell-to-1 f one mm)
Head Welds Longitudinal Volumetric 4
2 50%
of one weld 1
0 1
___________per head (2)
Steam Generators B-B B2.40 (Primary Side)
Volumetric 2
1 50%
Weld(l) 0 1
0 Tubesheet-to-Head Weld Category Total 8
5 2
1 2
Note 1 :The examination may be limited to one vessel among the group of vessels performing a similar function. (Ref. Table IWB-Notes for Cat. B-B 2500-1, Examination Category B-B, Note 1)
Note 2:The Pressurizer has 2 heads with 2 welds each. Only one weld per head is required, therefore 50%
B-D Full Penetration Welded Nozzles in Vessels B-D B3.100 Reactor Vessel Nozzle Volumetric 6
6 100%
Same as 1st 0
0 6
Inside Radius Section Interval B-D B3.1 10 Pressurizer Nozzle-to-Volumetric 4
4 100%
Same as 1st 2
1 1
B-D_
B1 Vessel Welds Interval B-D B3.120 Pressurizer Nozzle Volumetric 4
4 100%
All nozzles 2
1 1
B3.120Inside Radius Section or visual 4
40Aloe 21 Steam Generators Same as 1st B-D B3.130 (Primary Side) Nozzle-to-Volumetric 6
6 100%
Interval 1
5 0
Vessel Welds Steam Generators B-D B3.140 (Primary Side) Nozzle or visual 6
6 100%
All nozzles 1
5 0
Inside Radius Section o vsa Reactor Vessel Nozzle-Volumetric 6
100%
Same as 1st 0
0 6
B-D B3.90 to-Vessel Welds Interval I
Revision 0 64 Prepared for Constellation Energy by
Constellation Energy*
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary r
tNumber to Number to Number to Ii Number of Reqired to Examination Number to be be be be Category Number Description Method Components in be Examined Percentage Examined in Examined Examined Examined Iemb EMet Duterig Required Interval in First in Second in Third Interval Period Period Period Category Total 32 32 6
12 14 B-E PWR Reactor Pressure Vessel Upper Head Every third Reactor Vessel Head refueling and Nozzle partial outage or 5 B-E B4.30 Penetration Welds of Visual, VE 1
1 100%
calendar 1
0 0
PWSCC Resistant
- years, Material whichever is less--
All nozzles, Reactor Vessel Head not to exceed Nozzles and Partial one B-E B4.40 Penetration Welds of VOL &Ior 1
1 100%
inspection 0
0 1
Peerto ed fSur interval PWSCC Resistant (nominallya1 Material (nominally 10 Calendar years)
Category Total 2
2 1
0 1
B-G-1 Pressure Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter Reactor Vessel Closure Same as for 1
0 B-G-1 B6.10 Head Nuts Visual, VT-i 1
1 100%
1st interval 10 Revision 0 65 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Category Number Description Method Components in During Percentage Examined in Examined Examined Examined CoItempon During Required Interval in First in Second in Third Interval Period Period Period Steam Generators Same as for B-G-1 B6.100 Flange Surface when Visual, VT-1 4
0 0%
1st interval(I) 0 0
0 connection disassembled Steam Generators Nuts, Same as for B-G-1Bushings, and Washers ual, VT-4 4
100%
1st interval 0
4_0 B-G-1 B6.180 Pumps Bolts and Studs Volumetric 4
0 0%
(1)(2) 0 0
0 Pumps Flange Surface B-G-1 B6.190 when connection Visual, VT-1 4
0 0%
(1)(2) 0 0
0 disassembled Reactor Vessel Closure Volumetric 1
1 100%
Same as for 1
0 0
B-G-1 B6.20 Studs 1st interval B-G-1 B6.200 Pumps Nuts, Bushings, Visual, VT-1 4
0 0%
(1)(2) 0 0
0 and Washers B-G-1 B6.40 Reactor Vessel Threads Volumetric 1
1 100%
Same as for 0
0 1
in Flange 1st interval B-G-1 B6.50 Reactor Vessel Closure Same as for 0
1 0
B-G-1 B6.50 Washers, Bushings Visual, VT-i 1
1 100%
1st interval Steam Generators Bolts Sameasfor1.0 2
2 BG1 B690 and Studs Volumetric 4
4 100%
1st interval "0
2 1_2 Category Total 28 12 2
7 3
Note 1: Not Required unless disassembled Note 2: For heat exchangers, piping, pumps, and valves, examinations are limited to components selection for examination under Notes for Cat.B-G-1 Examination Categories B-B, B-J, B-L-2, and B-M-2:
(Ref. Table IWB-2500-1, Examination Category B-G-1, Note 3)
Revision 0 66 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Ctgr ItmDsrpinExam
'Cmoet Nme oNme oNumber to Category Item Description Exams in be Examined Percentage Examined in Examined Examined Examined Number Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period B-G-2 Pressure Retaining Bolting, 2in. (50 mm) and Less in Diameter Same as for B-G-2 B7.20 Pressurizer Bolts, Studs, Visual, VT-1 9
0 0%
1st interval 0
0 0
and Nuts (1)(3)
Piping Bolts, Studs, and Same as for B-G-2 B7.50.
Nuts Visual, VT-1 6
0 0%
1st interval 0
0 0
(1)(2)(3)
Same as for B-G-2 B7.60 Pumps Bolts, Studs, and Visual, VT-1 4
0 0%
1st interval 0
0 0
(1)(2)(3)(4)
Valves Bolts, Studs, and Same as for B-G-2 B7.70 Nuts Visual, VT-I 32 0
- 0%
1st interval 0
0 0
Nuts_
(1)(2)(3)(5)
Category Total 51 0
0 0
0 Note 1: Not required unless disassembled Note 2: For heat exchangers, piping, pumps, and valves, examinations are limited to components selection for examination under Examination Categories B-B, B-J, B-L-2, and B-M-2. (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 2)
Notes for Cat.B-G-2 Note 3: Examination is only required once per interval Note 4: Only one pump of each group of pumps is required as outlined in B-L-2 Note 5: Only one valve of each group of valves is required as outlined in B-M-2 Revision 0 67 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Item Exam
.be Examie Category Number Description em Components in b mined Percentage Examined in Examined Examined Examined NDsrpi Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period B-K Welded Attachments for Vessels, Piping, Pumps, and Valves Pressure Vessels Same as for B-K B10.10 Welded Attachments Surface 6
3 50%
1st interval 1
0 2
Welded__Attachments_
(1)(2)
B-K B10.20 Piping Welded Surface 9
1 10%
Same as for 0
1 0
Attachments 1st interval B-K B1.30Pumps Welded Same as for 1
0 1
uAttachments Surface 16 2
10%
1stinterval 1
0_1 Category Total 31 6
2 1
3 Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels Notes for Cat. B-K Ishall be selected for examination. For single vessels, only one welded attachment shall be selected for examination.
Note 2: There are three welded attachments on the RPV, 1 on PZR and 1 on each S/G for a total of 6. Must examine 1 on each vessel or multiple vessels, therefore 3 are required for 50%
B-L-1 Pressure Retaining Welds in Pump Casings Same as for B-L-1 B12.10 Pumps Pump Casing Visual, VT-1 4
1 25%
first interval(I) 0 0
1 Welds (B-L-l)
_(2)
Category Total 4
1 0
0 1
Notes for Cat. B-L-1 Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.
Note 2: There are four Reactor Coolant Pumps. Each pump has one casing weld, therefore 1 weld of 4 is required for 25%
Revision 0 68 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Re e Jo Numberto Number to Numberto I
Number of Required to Examination Number to be be be be C a e o y I t e m D s r p i nE x a m
.op n n s i b e E x a m i n e d Category M
Components in Durn Percentage Examined in Examined Examined Examined Number DescripnMethod Item No.
g Required Interval in First in Second in Third Interval Period Period Period B-L-2 Pump Casings p Pm Same as for B-L-2 B12.20 Pumps Pump Casing Visual, VT-3 4
0 0%
first interval(I) 0 0
0 (B-L-2)(2)
Category Total 4
0 0
0 0
Notes for Cat. B-L-2 Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.
NtfoCtB I Note 2: Not required unless disassembled B-M-1 Pressure Retaining Welds in Valve Bodies Valves, Less than NPS 4 Same as for B-M-1 B12.30 (DN 100) Valve Body Surface 4
2 50%
first interval 0
0 2
_Welds (B-M-1)
(1)(2)(3)
Category Total 4
2 1
0 0
2 Note 1: Examination is limited to at least one valve in each group of valves that are of the same size, constructural design, manufacturing method, and that perform similar functions in the system.
Note 2: There is one valve group that has two valve body welds on each valve. Each valve has two body welds, therefore 2 welds of 4 are required for 50%
Note 3; Since both scheduled welds are in the same valve body they will be examined in the same outage.
B12.50 Valve Body, Exceeding Visual, VT-3 NPS 4 (DN 100) (B-M-2) 14 0
Same as for 0%
first interval (1)(2) 0 0
1 0
Revision 0 69 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Numberto Numberto Numberto Item Exa Number of Required to Examination Number to be be be be Category Number Description Method Components in Examined Percentage Examined in Examined Examined Examined Item No.
During Required Interval in First in Second in Third Interval Period Period Period Category Total 14 0
0 0
0 Note 1: Examination is limited to at least one valve in each group of valves that are of the same size, constructural design, B-M-2 manufacturing method, and that perform similar functions in the system.
Note 2: Not required unless disassembled B-N-1 Interior of Reactor Vessel Each B-N-1 B13.10 Reactor Vessel, Vessel Visual, VT-3 1
3 100%
inspection 1
1 1
Interior (B-N-I)
_period Category Total 1
3 1
1 1
B-N-2 Welded Core Support Structures and Interior Attachments to Reactor Vessels Reactor Vessel (PWR)
B-N-2 B13.50 Interior Attachments Visual v-r 1
1 100%
Same as for 0
0 1
Within Beltline Region 1st interval (B-N-2)
Reactor Vessel (PWR)
B-N-2Interior Attachments Visual, VT-3 1
1 100%
Same as for 0
0 1
Beyond Beltline Region 1stinterval (B-N-2)
Category Total 2
2 0
0 2
Revision 0 70 Prepared for Constellation Energy by
Constellation Energy, Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary R
e to INumberto Numberto Numberto Number of Required to Examination Number to be be be be Category Item Description Meth Components in Examined Percentage Examined in Examined Examined Examined ExaMe Co moItem No.
During Required Interval in First in Second in Third Interval Period Period Period B-N-3 Removable Core Support Structures Reactor Vessel (PWR)
Same as for B-N-3 B13.70 Core Support Structure Visual, VT-3 1
1 100%
1st interval 0
0 1
(B-N-3)
Category Total I
1 0
0 1
B-O Pressure Retaining Welds in Control Rod Drive and Instrument Nozzle Housings 10%
Reactor Vessel (PWR)
Volumetric peripheral B-O B14.20 Welds in Control Rod 112 12 10%
CRD 0
0 12 Drive CRD Housing housings (1)
Category Total 112 12 0
0 12 Notes for Cat Note 1: There are 28 Peripheral CEDMs, 10% of 28 is 3. There are 4 welds per CEDM for a total of 112 welds. Therefore, 12 welds are selected out of 112 for 10%
C-A Pressure Retaining Welds in Pressure Vessels Pressure Vessels Shell Volumetric Each C-A C1.10 Circumferential Welds (5) 8 4
50%
inspection 3
2 0
interval(1 ) (2)
Each C-A Cl.20 Pressure Vessels Head Volumetric 2
50%
inspection 0
Circumferential Welds (5) interval(1 )(3)
Revision 0 71 Prepared for Constellation Energy by
Constellation Energy-
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to dNumber oft Examination Number to be be be be Category m
Description m
Components in During Percentage Examined in Examined Examined Examined Item No.
During Required Interval in First in Second in Third Interval Period Period Period C-A C1.30 Pressure Vessels Volumetric 4
2 50%
inspection 0
0 1
Tubesheet-to-Shell Weld (5) 50%
inspection I
I interval(1 )(4)
Category Total 14 7
3 2
2 Note 1: The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref. Table Notes for Cat. C-A N IWC-2500-1, Examination Category C-A, Note 3)
Note 2: There are three circumferential shell welds on both S/G & 1 circumferential shell weld on both SCHE, requiring 3 S/G and 1 SCHE shell welds be scheduled or 4/8 shell welds are required to be examined for 50%
Note 3: There is one circumferential head weld on both S/G, therefore, 1/2 head welds are required to be examined or 50%
Note 4: There is one tubesheet-to-shell weld on both S/G & one tubesheet-to-shell weld on both SCHE, therefore 2/4 tubesheet-to-shell welds are required to be examined for 50%
C-B Pressure Retaining Nozzle Welds in Vessels Nozzles Without Reinforcing Plate in Vessels> 1/2in. (13mm)
Surface and Each C-B C2.21 Nominal Thickness volumetric 8
4 50%
inspection 1
1 2
Nozzle-to-Shell (Nozzle interval(1 )(2) to Head or Nozzle to
(
Nozzle) Weld Nozzles Without Reinforcing Plate in Each C-B C2.22 Vessels > 1/2 in. (13mm)
Volumetric 8
2 25%
inspection 01 1
1 Nominal Thickness Nozzle Inside Radius interval(I)
I ___ I ____
_ ISection I
I I
Revision 0 72 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Required to Number to Number to Number to Item Exa Number of Requined Examination Number to be be be be Catemont Exam be Examined Percentage Examined in Examined Examined Examined CoItem No.
During Required Interval in First in Second in Third Interval Period Period Period Category Total 16 6
1 2
3 Note 1: The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref. Table Notes for Cat. C-B IWC-2500-1, Examination Category C-B, Note 4)
Note 2: There are two nozzles on both S/G and two nozzles on both SDC HX, therefore 4/8 nozzles are required to be examined or 50%
Note 3: There are two nozzle inner radii on both S/G, however, the two nozzle inner radii on both SDC are not required <12",
therefore 2/8 inner radii are required to be examined or 25%
C-C Welded Attachments for Vessels, Piping, Pumps, and Valves Each identified occurrence CC C10 Pressure Vessels ocrec C-C C3.10 Welded Attachments(l)
Surface 18 2
11%
and each 0
0 2
inspection interval (1)(2)
Each identified C-C C3.20 Piping Welded Surface 238 24 10%
occurrence 9
8 8
Attachments(1) and each inspection interval (1) (3)
Category Total 312 26 9
8 10 Revision 0 73 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2-Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Category Number Exam e
be Examined Percentage Examined in Examined Examined Examined Components iN During Required Interval in First in Second in Third Interval Period Period Period Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. (Ref.
Table IWC-2500-1, Examination Category C-C, Note 4)
Notes for Cat. C-C Note 2: There are eight Welded Attachments on both S/G and one welded attachment on both SCHE, therefore 2/18 welded attachments are required to be examined for 11%
Note 3: Examination is required whenever component support member deformation is identified. (Ref. Table IWC-2500-1, Examination Category C-C, Note 6)
C-G Pressure Retaining Welds in Pumps and Valves Each C-G C6.20 Valves Valve Body Surface 4
2 50%
inspection 0
1 1
Welds(2) interval(1 )(2)
Category Total 4
2 0
1 1
Note 1: Examination is limited to all the welds in one valve in each group of valves that are of similar size, design, function, and Notes for Cat. C-G I service. (Ref. Table IWC-2500-1, Examination Category C-G, Note 1 Note 2: There are two valve groups with two valves each, therefore both welds on two of the four valves are required to be examined for 50%
D-A Welded Attachments for Vessels, Piping, Pumps, and Valves Revision 0 74 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary e to Number to Number to Number to Item DceNumber of Requiredxto Examination Number to be be be be Ca r
Nmb Description Exme C
o t
be Examined Percentage Examined in Examined Examined Examined tegory Number Method Item No.
During Required Interval in First in Second in Third Period Period Period Each identified Pressure Vessels occurrence D-AWelded Attachments Visual, VT-1 12 7
52%
and each 1
1 6
inspection interval (1)(2)(3)
Each identified D-A D1.20 Piping Welded Visual, VT-1 489 49 10%
occurrence 9
18 22 Attachments and each inspection interval(3)
Category Total 561 56 10 19 28 Note 1: For multiple vessels of similar design, function, and service, the welded attachments of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. (Ref. Table IWC-2500-1, Examination Category D-A, Note 3)
Notes for Cat. D-A Note 2: There are four groups of multiple vessels (each with two vessels), one single vessel, and two similar vessels with two entries each, therefore, 7/13 are required to be examined for 53%
Note 3: Examination is required whenever component support member deformation is identified. (Ref. Table IWD-2500-1, Examination Category D-A, Note 4)
F-A Supports Each F-A F1.10A Class 1 Piping Supports - Visual, VT-3 46 12 (1) inspection 4
3 5
One Directional interval Revision 0 75 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary R
itNumberto Numberto Numberto Number of Required to Examination Number to be be be be Item Exam.beEaid Category Number Description Components in be Examined Percentage Examined in Examined Examined Examined Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period Each F-A F1.10B Class 1 Piping Supports-Visual, VT-3 50 13 (1) inspection 3
1 9
Multi-directional interval Each F-A F1.10C Ciass 1 Piping Supports-Visual, VT-3 74 19 (1) inspection 7
9 2
Thermal Movement interval Each F-A F1.10 Total Class I Piping
- Visual, 170 43 25%
inspection 14 13 16 Supports VT-3 interval Class 2 Piping Supports -
Each F-A F1.20A One Diping Visual, VT-3 204 31 (1) inspection 5
15 11 One Directional interval Each F-A FI.20B Class 2 Piping Supports-Visual, VT-3 259 39 (1) inspection 13 13 13 Multi-directional interval Each F-A F.20 Class 2 Piping Supports - Visual, VT-3 96 15 (1) inspection 5
2 7
F-A F1.20C Thermal Movement interval Each F-A F1.20 Total Class 2 Piping
- Visual, 559 84 15%
inspection 23 30 31 Supports VT-3 interval Each F-A F1.30A Class 3 Piping Supports - Visual, VT-3 269 27 (1) inspection 9
10 8
One Directional interval Each F-A F1.30B Class 3 Piping Supports-Visual, VT-3 533 54 (1) inspection 10 24 20 Multi-directional interval Revision 0 76 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section Xl Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Numberto Numberto Numberto Number Required to Examination Number to be be be be Item Exam N
be Examined Category Components in During Percentage Examined in Examined Examined Examined Number Description Method Item No.
During Required Interval in First in Second in Third Interval Period Period Period Each F-A F1.30C Class 3 Piping Supports - Visual, VT-3 70 7
(1) inspection 4
1 2
Thermal Movement interval Each F-A F1.30 Class 3 Piping
- Visual, 872 88 10%
inspection 23 35 30 Supports VT-~3 interval Supports other than Visual Each F-A F1.40 Piping Supports (Class Viul 63 26 41%
inspection 0
8 18 11,2,3, and MC)
I interval(2)(3)_
Category Total 1664 241 60 86 95 Note 1: The total percentage sample shall be comprised of supports form each system, where the individual sample sizes are proportional to the total number of non-exempt supports of each type and function within each system. (Ref. Table IWF-2500-1, Examination Category F-A, Note 2)
Note 2: For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. (Ref. Table IWF-2500-1, Examination Category F-A, Note 3)
Note 3: One PZR, nine of eighteen S/G, one of the two SCHEs, the SWHE, one of four SRWHX, one of two SWHT, one of two CCWHE, the CCWHT, one of two DACHE, one of two L.O. Coolers, one of two JWC, two of four SCHE, six of twenty-four RCPs. Therefore 26/65 component supports are required to be examined for 40%.
R-A Risk Informed Piping Welds N-716 LSS Elements tha R-A RO.00 have not been Evaluated for Degradation R-A R1. 1 1N-716 Elements Subject to Thermal Fatigue Revision 0 77 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Number Of Examination Number to be be be be Category Item Description Exam Components in be Examined Percentage Examined in Examined Examined Examined Item No.
Required Interval in First in Second in Third Interval Period Period Period N-716 Elements Subject R-A R1. 11-15 to Thermal Fatigue and Volumetric 3
3 100%
Element (1) 2 0
1 PWSCC N-716 Elements Subject to Intergranular or R-A R1.16 Transgranular Stress Volumetric 6
2 25%
Element (1) 2 0
0 Corrosion Cracking (IGSCC, TGSCC)
N-716 Elements not R-A R1.20 Subject to a Damage Volumetric 612 52 8.19%
Element (1) 20 22 10 Mechanism N-716 Elements not R-A R1.20S Subject to a Damage Visual, VT-2 13 0
0%
Element (1) 0 0
0 Mechanism Category Total 2072 72 29 27 16 Notes for Cat. R-A Note 1: Percentages were determined using SI Calc 0800144.301 Revision 0.
MRP-139C Mitigated Welds of Non-Resistant Material MRP 139 Category C Welds that are in the RI-R-A R1.11-15 ISI Program as Volumetric 3
2 (1) 2 0
0 Susceptible to Thermal
_Fatigue and PWSCC MRP 139 Category C R-A R1.20 Welds that are in the RI-Volumetric 3
1 (1) 1 0
0 ISI Program as no DM Revision 0 78 Prepared for Constellation Energy by
Constellation Energy*
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Number of Required to Examination Number to be be be be Category Item Description Exam Components in be Examined Percentage Examined in Examined Examined Examined Number Method Item ono.
During Required Interval in First in Second in Third SInterval Period Period Period Category Total 6
3 1
0 0
Notes for MRP-Note 1: Welds of Non-Resistant Material that have been mitigated by SI have a MRP-1 39 requirement of 50% of each mitigation type 139C within next 6 years MRP-139E Non Mitigated Welds of Non-Resistant Material MRP 139 Category E Welds that are in the RI-Volumetric R-A R1.11 ISI Program as (MRP 139 2
4 (1) 1 1
2 Susceptible to Thermal Cat E)
Fatigue MRP 139 Category E Volumetric R-A R1.20 Welds that are in the RI-(MRP 139 17 26 (1) 2 15 9
1_
I Category Total 19 37 (1) 18 1
18 Notes for MRP-Note 1: Cold Leg Welds >4 inches of Non-Resistant Material that have not been Mitigated by SI have a MRP-139 139E requirement of 100% examination each 6 years MRP-139F Welds of Non-Resistant Material that are Cracked and have a Weld Overlay MRP 139 Category F R-A R1.20 Welds that are in the RI-Volumetric 2
0 (1) 0 0
0 ISI Program as no DM Category Total 20 (1) 00 0
Revision 0 79 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code Category Summary Number to Number to Number to Item Exam Number of Required to Examination Number to be be be be Category N
ber Description Metd Components in be Examined Percentage Examined in Examined Examined Examined Item No.
During Required Interval in First in Second in Third Interval Period Period Period Notes for MRP-Note 1: Welds of Non-Resistant Material that are cracked and have a weld overlay have a MRP-1 39 requirement Once in the next 5 years; if no additional indications/growth, continue with existing Code examination program for unflawed condition or yer;i1oadtinl3dctos/9whFotne iheitn od xmnto rogado nlwdcniino 139F approved alternative. These two welds are scheduled for examination in the 3rd interval.
MRP-139K Cold Leg Welds of Non-Resistant Material MRP 139 Category K Welds that are in the RI-Volumetric R-A R1.11 ISI Program as (MRP 139 2
2 (1) 0 2
0 Susceptible to Thermal Cat K)
Fatigue MRP 139 Category K Volumetric R-A R1.20 Welds that are in the Rl-(MRP 139 11 11 (1) 0 9
I A
Category Total 13 13 (1) 0 11 2
Notes for MRP-Note 1: Cold Leg Welds of Non-Resistant Material have a MRP-1 39 requirement of 100% at least once every three (3) RFOs (not 139K counting RFOs when weld is examined volumetrically as one of the three) or until mitigated or replaced.
RG 1.14 RCP FLYWHEEL RCP RG1.14 RCP Flywheels UT, ECT 4
4 (1) 4 4
4 Category Total 4
12 (1) 4 4
4 Notes for RG 1. 14 Note 1: Inservice inspection of each reactor coolant pump flywheel shall be performed at least once every ten years with the high stress keyways at approximately 3 and 1/3 years intervals Revision 0 80 Prepared for Constellation Energy by
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Fourth Interval Inservice Inspection Program CCNPP Unit 2 Code. Category Summary u
d toNumberto Numberto Numberto Number of Required to Examination Number to be be be be Category Item DescriptionExa Components in be Examined Percentage Examined in Examined Examined Examined at orCNumber Method Item NO.
During Required Interval in First in Second in Third Interval eNnrPeriod Period Period RI-BER R-A R1.20 Volumetric 154 2
1%
1 1
0 Category Total 154 2
(1) 1 1
0 Notes for RI-BER Notel: These welds are selected as provided in EPRI TR 1006837 Revision 0 81 Prepared for Constellation Energy by
SConstellation Energy-Calvert Cliffs Units I & 2 ConsrtelfcleatPon erg PASME Section XI Calvert Cliffs Nuclear Power Plant Inservice Inspection Program 8.0 LISTING OF DEFINITIONS 8.1 Authorized Inspection Agency - an organization that is empowered by an Enforcement Authority to provide inspection personnel and services as required by ASME Section Xl.
8.2 Authorized Nuclear Inservice Inspector - a person who is employed and has been qualified by an Authorized Inspection Agency to verify that examinations, tests, and repairs are performed in accordance with the rules and requirements of ASME Section XI.
8.3 Authorized Nuclear Inservice Inspector Supervisor - a person who is employed by an Authorized Inspection Agency to supervise Authorized Nuclear Inservice Inspectors and who is qualified as an Authorized Nuclear Inservice Inspector.
8.4 Component an item in a nuclear power plant such as a vessel, concrete containment, pump, valve, piping system, or component support.
8.5 Component ID - Unique Plant Identifier for the ScheduleWorks database 8.6 Enforcement Authority - a regional or local governing body, such as State or Municipality of the United States or a Province of Canada, empowered to enact and enforce Boiler and Pressure Vessel Code legislation.
8.7 Engineering Evaluation - an evaluation of indications that exceed allowable acceptance standards to determine if the margins required by the Design Specifications and Construction Code are maintained.
8.8 Evaluation - the process of determining the significance of examination or test results, including the comparison of examination or test results with applicable acceptance criteria or previous results.
8.9 Examination - denotes the performance of nondestructive testing and visual observation such as volumetric examinations (radiography, ultrasonic and eddy current), surface examinations (liquid penetrant or magnetic particle), and visual examinations (VT-1, VT-2, and VT-3).
8.10 Examination Category - a grouping of items to be examined or tested.
8.11 Fabrication - actions such as forming, machining, assembling, welding, brazing, heat treating, examination, testing, inspection, and certification, but excluding design, required to manufacture components, parts or appurtenances.
8.12 General Corrosion - an approximately uniform wastage of a surface of a component, through chemical or electrochemical action, free of deep pits or cracks.
Revision 0 82 Prepared for Constellation Energy by
a ECalvert Cliffs Units 1 & 2 Constellation EnrPlagy ASME Section XI Calvert Cliffs Nuclear Power Plant Inservice Inspection Program 8.13 Hold Time - the time after pressurization to test conditions before the visual examinations commence.
8.14 Inservice Inspection - methods and actions for assuring the structural and pressure-retaining integrity of safety-related nuclear power plant components in accordance with the rules of Section XI of the ASME Code.
8.15 Inservice Inspection Program Owner - individual in the Constellation Generation Group's Corporate Engineering Programs Unit responsible for development and oversight of station activities required to implement the inservice Inspection Program.
8.16 Inservice Life - the period of time from the initial use of an item until its retirement from service.
8.17 Inservice Test - a special test procedure for obtaining, through measurement or observation, information to determine the operational readiness of a system or component.
8.18 Inspection - verification of the performance of examinations and tests by an Inspector.
8.19 Inspection Program - the plan and schedule for performing examinations or tests as required by Section Xl of the ASME Code.
8.20 Inspector - an Authorized Nuclear Inservice inspector.
8.21 Nondestructive Examination
- an examination by the visual, surface, or volumetric method.
8.22 Normal Plant Operating Conditions - the operating conditions during reactor startup, operation at power, hot standby, and reactor cooldown to cold shutdown conditions. Test conditions are excluded.
8.23 Normal Plant Operation - the conditions of startup, hot standby, operation within the normal power range, and cooldown and shutdown of the power plant.
8.24 Open Ended - a condition of piping or tubing that permits free discharge to the atmosphere or containment atmosphere.
8.25 Operating Convenience - a provision to facilitate plant operation but not required to perform a specific function in shutting down a reactor to cold shutdown condition or in mitigating the consequences of an accident.
8.26 Operational Readiness - the ability of a component or system to perform its intended function when required.
Revision 0 83 Prepared for Constellation Energy by dikk~l CoraptwIno.
Constellation Energy-Calvert Cliffs Units 1 & 2 0 Consertelfcleaton er PASME Section XI Calvert Cliffs Nuclear Power Plant Inservice Inspection Program 8.27 Original Construction Code - the original Codes(s) Edition and Addenda under which the component was constructed and installed.
8.28 Owner - the organization legally responsible for the operation, maintenance, safety, and power generation of the nuclear power plant, (i.e., Constellation Energy).
8.29 Piping System - a functional nuclear power plant system with boundaries defined by a plant Design Specification, and/or System Flow Diagrams.
8.30 Pressure Test Program Owner -
a Station Engineer or individual with equivalent qualifications responsible for the development and overall implementation of the system pressure test program as required by ASME Section Xl.
8.31 Regulatory Authority - a federal government agency, such as the United States Nuclear Regulatory Commission, that is empowered to issue and enforce regulations affecting the design, construction, and operation of nuclear power plants.
8.32 Relevant Condition - a condition observed during a visual examination that requires supplemental examination, corrective measure, repair, replacement, or analytical evaluation.
8.33 Summary Number - Unique ISI designator for the ScheduleWorks database.
8:34 System Hydrostatic Test Boundary - the boundary subject to test pressurization during a system hydrostatic test shall be defined by the system boundary (or each portion of the boundary) within which the components have the same minimum required classification and are designed to the same primary pressure rating as governed by the system function and the internal fluid operating conditions, respectively.
8.35 System Leakage Test Boundary - the boundary subject to test pressurization during a system leakage test shall extend to the pressure retaining components within the system boundary pressurized during normal plant operation, or during a system operability test.
8.36 System Pressure Test - a test in which the pressure retaining components within each system boundary is subject to system pressure under which visual examination VT-2 is performed to detect leakage. The test may be conducted in conjunction with one or more of the following system tests:
" System Leakage Test - a system leakage test conducted with the system normal operating conditions, or during a system operability test.
- System Hydrostatic Test - a system hydrostatic test conducted during a plant shutdown at a pressure above nominal operating pressure or system pressure for which overpressure protection is provided.
Revision 0 84 Prepared for Constellation Energy by Arkkkmla Cowp ho.
Consteilation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program
- System Pneumatic Test - a system pneumatic test conducted in lieu of a hydrostatic pressure test for components within the scope of IWC and IWD.
8.37 Test - a procedure to obtain information through measurement or observation to determine the operational readiness of a component or system while under controlled conditions.
8.38 Verify - to determine that a particular action has been performed in accordance with the rules and requirements of Section XA of the ASME Code either by witnessing the action or by reviewing records.
8.39 Visual Examination VT the VT-2 Visual Examination is conducted in accordance with ASME Section Xl, IWA-2212 to determine the presence of leakage from pressure retaining components.
Revision 0 85 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program 9.0 LISTING OF ASME CALIBRATION BLOCKS
,CCNP ASME Calibration Blocks Block Type Description Material Thick Dia.
Drawing 5" VESSEL CALIBRATION 0D-2929-056 CC-01 N/A BOKN/A 5"
0 RED BLOCK REV D CC-02 N/A 7" VESSEL CALIBRATION N/A 7"
21 D-2929-011 BLOCK REV G 9" BLOCK VESSEL 0D-2929-012 CALIBRATION BLOCK REV F 11" VESSEL UT CALIBRATION 0-2929-237 CC-04 N/A BLOCK N/A 11",
0 REV B HEAD FLANGE TO HEAD UT 0D-2929-188 CALIBRATION BLOCK REV B CC-06 N/A PRESSURIZER SURGE LINE ASTM A-351-65 GRADE CF8M/SS 1.260 0
D-2929-047 CALIBRATION BLOCK REV E 3" COLD LEG CALIBRATION C-2929-057 CC-07 N/A BOKM-1806-4 3"
0 RE0 BLOCK REV D CC-08 N/A 3 3/4" UT CALIBRATION BLOCK SA-516-70/CS 3 3/4" 0
N/A 4 3/4" HOT LEG CALIBRATION C2929-197 CC-09 N/A BOKM-1804-1 4 3/4" 0
REC BLOCK REV C CALIBRATION BLOCK FOR C-2929-090 CC-10 N/A REACTOR COOLANT PUMP SS 3 1/4" 10 REV 0 SAFE-END WELDS CC-11 N/A REGENERATIVE HEAT SA 182 F304 SMS/SS
.875" 8
C-2929-040 EXCHANGER REV B C-2929-075 CC-12 N/A 2" PIPE CALIBRATION BLOCK ASTM A376/SS
.344 2
REV0C REV C 2 1/2" PIPE CALIBRATION C-2929-074 CC-13 N/A BOKSA 376/SS
.375 0
REC BLOCK REV C 0-2929-073 CC-14 N/A 3" PIPE CALIBRATION BLOCK SA 376/SS
.438 3
REV90 REV D 4" PIPE UT CALIBRAITION D-2929-072 CC-15 N/A BOKSA 376 GR 316/SS
.43 RE4 BLOCK REV E 6" SCHEDULE 120 UT 0-2929-060 CC-16 N/A CABTNLO A-376-64 GR 316/SS
.570 6
REV0C CALIBRATION BLOCK REV C 0-2929-069 CC-17 N/A 12" PIPE CALIBRATION BLOCK ASTM A376/SS 1.080 12 REV90 REV D CC-18 N/A 14" PIPE CALIBRATION BLOCK ASTM 376 TYPE 316 WITH 1 3/16" 14 D-2929-068 SUPPL. S-2, 486/SS REV B CC-19 N/A 4" PIPE CALIBRATION BLOCK ASTM 1-106-68 GRADE B/CS
.244 4
C-2929-071 REV B 0-2929-070 CC-20 N/A 6" PIPE CALIBRATION BLOCK ASTM Al 06/CS
.310" 6
REV0C IREV C C-2929-076 CC-21 N/A 6" PIPE CALIBRATION BLOCK CARBON STEEL
.560 6
REV90 REV D 6" MAINSTEAM PIPE C-2929-066 CC-22 N/A CALIBRATION BLOCK ASTM A-155, CL 1, KC-70/CS
.984 0
REV C Revision 0 86 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program CCNP ASME
_Calibration
- Blocks, Block Type Description Material Thick Dia.
Drawing CC-23 N/A 36" CALIBRATION BLOCK ASTM A155 GRADE MCF 70 2.160 36 D-2929-065 CLASS 1/CS REV C C-2929-067 CC-24 N/A 16" PIPE CALIBRATION BLOCK AST A-106C/CS
.633 0
REV0E REV E C-2929-077 CC-25 N/A 24" PIPE CALIBRATION BLOCK CARBON STEEL 1.253" 24 REV0B REV B 4" PRESSURIZER RELIEF C-2929-143 CC-26 N/A NOZZLE TO SAFE END N/A
.645" 4
REV B CALIBRATION BLOCK REV B CLOSURE HEAD STUD C-2929-036 CALIBRATION BLOCK REV F CC-28 N/A CLOSURE HEAD NUT N/A 7"
0 B-2929-038 CALIBRATION BLOCK (OLD)
REV A CC-29 N/A CLOSURE HEAD NUT N/A 7"
0 C-2929-141 CALIBRATION BLOCK (NEW)
REV A REACTOR COOLANT PUMP CC-30 N/A STUD CALBAT BLOC N/A 4.750 0
N/A STUD CALIBRATION BLOCK VERTICAL SUPPORT C-2929-046 CC-31 N/A ATTACHMENT LUG ASTM A-316/SS 4 1/8" 0
REV B CALIBRATION BLOCK 6" BONNET VALVE FLANGE-B-2929-049 CC-32 N/A BODY WELD CALIBRATON N/A 4 13/16 0
REV C BLOCK CC-33 N/A 3" UT CALIBRATION BLOCK N/A 3"
0 C-2929-196 REV B CC-34 N/A 5" CALIBRATION BLOCK SA-515-70 5"
0 N/A REACTOR COOLANT PUMP NUT C-2929-227 UT CALIBRATION BLOCK REV A PUMP STUDY UT CALIBRATION D-2929-238 CC-37 N/A BOKN/A 4 1/8 0
REB BLOCK REV B 34" PIPE UT CALIBRATION D-2929-226 CC-38 N/A BOKASTM A155 KC65 CLI 1.060 0
RV BLOCK REV A NOZZLE TO SHELL UT D-2929-229 CC-39 N/A CALIBRATION BLOCK SA 508, CL II 6"
0 REV A CC-40 N/A INLET AND OUTLET NOZZLE A-508, CLASS 2
.7" 0
D-2929-30 CC-40 N/A_______
INNER RADIUS A-508,_CLASS_2_7"_0_D-2929-30 CC-41 N/A OD INNER RADIUS D-2929-236 CALIBRATION BLOCK SA507, CLASS II 7"
0 REV C CC-42 N/A 6" PIPE UT CALIBRATION ASTM A-376 TYPE 304
.432" 0
D-2929-244 BLOCK I
CC-43 N/A 6" PIPE UT CALIBRATION ASTM A-312 TP 304/SS
.280 0
D-2929-246 BLOCK CC-44 N/A 6" PIPE UT CALIBRATION ASTM A-312 TYPE 304/SS
.134 0
D-2929-245 BLOCK CC-45 N/A 8" PIPE UT CALIBRATION ASTM A-312 TYPE 304/SS
.322" 0
D-2929-242 BLOCK 8" PIPE LIT CALIBRATION D22
-4 CC-46 N/A BLOCK ASTM-1 312, TP-304/SS
.148" 0
D-99-4 Revision 0 87 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program CCNPASME Calibration Blocks Block Type Description' Material Thick Dia.
Drawing CC-47 N/A 10" PIPE UT CALIBRATION ASTM A-312, TYP 304/SS
.365 10 D-2929-241 BLOCK CC-48 N/A 10" PIPE UT CALIBRATION ASTM A-312 TYPE 304/SS
.250 0
D-2929-247 BLOCK CC-49 N/A 10" PIPE UT CALIBRATION ASTM A-312 TP 304/SS
.165 0
D-2929-239 BLOCK CC-50 N/A 12" PIPE UT CALIBRATION ASTM 1-312 TUP 304
.250 0
0-2929-248
__________BLOCK_____
CC-51 N/A 14" PIPE UT CALIBARTION ASTM A-312 TP 304
.250 0
D-2929-240 BLOCK CC-52 N/A 18" PIPE UT CALIBRATION ASTM 1-312, TP-304
.250 0
D-2929-249 BLOCK CC-53 N/A 24 BIELTCALBATO ASTM A-312, TP-304
.3750 0
D--2929-250 24" PIPE UT CALIBRATION D-2929-250 CC-53 N/A BLOCK ASTM A-312, TP304 0
REV A CC-54 N/A 7" VESSEL UT CALIBRATION ASTM A533 GR.B N/A 0
D-2929-254 BLOCK CC-55 N/A 5" VESSEL UT CALIBRATION SA 533 GR. B 5"
0 D-2929-255 BLOCK SHUTDOWN COOLING HEAT D-2929-252 CC-56 N/A EXCHANGER UT CALIBRATION SA516 GR70 1.125" 0
REV B BLOCK CC-57 N/A LETDOWN HEAT EXCHANGER ASTM A-240 TP 304
.800" 0
N/A IUT CALIBRATION BLOCK CC-58 N/A 4" PIPE UT CALIBRATION ASTM A-106 GR B
.337 0
D-2929-251A BLOCK CC-59 N/A 12" PIPE UT CAIBRATION A-376, TP-316 1.125 0
D-2929-253
______________BLOCK CC-60 N/A
-14" PIPE UT CALIBRATION ASTM A-376, TP 316 1.250 0
D-2929-256 BLOCK I
REV A 12" PIPE UT CALIBRATION 0D-2929-258 CC-61 N/A BLOCK ASTM A106, GR B
.375 0
REV A CC-6 N/ARPV NOZZLE PIPE WELD CC-62 N/A
_NOZZLEPMOCKUP ASME SA508, CL 2 8"
0 D-2929-600 CC-63 N/A NOZZLE TO PIPE WELDED ASTM A-508 CLASS 1 4.5" 0
E-2929-601 MOCKUP CC-64 N/A 4" PIPE UT CALIBRATION ASTM A376 GR. 304/SS
.337 4
D2929-602 BLOCK
~4" PIPE UT CALIBRATION CC-65 N/A BLOCK ASTM A376, GR 304/SS
.237 0
D-2929-603 3" PIPE UT CALIBRATION 0D-2929-604 CC-66 N/A BLOCK ASTM A376, GR.304/SS
.300 0
REV A VESSEL FLANGE TO SHELL UT 0D-2929-605 CC-67 N/A CALIBRATION BLOCK SA 508,.CL2 14" 0
REV A SUPPORT SKIRT UT CC-68 N/A CALIBRATION BLOCK ASTM 4516, GR. 70 6"
0 0-2929-606 CC-69 N/A RPV FLANGE THREAD UT SA 508, CL2 7.030 0
0-2929-607 CALIBRATION BLOCK S5,2.0D 9-Revision 0 88 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNP ASME Calibration Blocks Block Type Description,,Material Thick Dia.
Drawing CS side 1.7"/SS E-2929-61 0 CC-70 N/A NOZZLE TO SAFE-END MOCKUP N/A 0
side REV C 1.32" FW HEATER COVER PLATE UT D-2929-611 CALIBRATION BLOCK REV A CC-72 N/A FW HEATER NOZZLE UT ASTM A105 2.812" 0
D-2929-612 CALIBRATION BLOCK SAFETY AND RELIEF NOZZLE-D-2929-613 CC-73 N/A TO-SAFE END MOCKUP ASTM 1-508 CLASS II 1.312 0
REV B ELBOW-SAFE END-PIPE D-2929-614 CC-74 N/A MOCKUP ASTM 1351 CF8M 3.35 0
REV B CC-75 N/A PUMP STUD UT CALIBRATION ASME SA540, GRB 23 4.75" 0
D-2929-618 CC-75 N/A BLOCK REV A CLOSURE HEAD STUD ASME SA-540 GRADE B24, Length 83D7863 REV CC-76 N/A CALIBRATION BLOCK CLASS 3 56.250" 0
2 CC-77 N/A 4' 8" X 1 1/4" DIAMETER STUD CARBON STEEL 1 1/4" 0
CB-01-46 CC-78 N/A 4' 8" X 1 1/4" DIAMETER STUD CARBON STEEL 1 1/4" 0
CB-01-47 ALLOY 600 GENERIC 83C6905 CC-79 N/A CALIBRATION BLOCK ALLOY 600SB-166 1' &.5" 0
REV.0 CC-80 PDI CARBON STEEL ALTERNATIVE A516-70 STEEL 2.25" 0
CB02112 ALTERNATIVE BOLOCK PDI CC-81 ALTERNATIVE 304 SS ALTERNATIVE BLOCK
- 304 SS 2.25" 0
CB02113 CAL BLOCK PDI CC-82 ALTERNATIVE 316 SS ALTERNATIVE BLOCK 316SS 2.25" 0
CB02114 BLOCK CC-83 STUD UT CALIBRATION STANDARD SA 193 GR. B7 Length 2.25 N/A PRIMARY MANWAY STUD 22.0" CC-84 N/A.
WELD 22 CIRCUMFERENTIAL SA-508 CL 3A/CLADDING 7.941 0
781:1D416 SEAM BLOCK INCONNEL-152 & SS-308 CC-85 N/A W22 BACKCLADDING & H/HOLE SA-508,CL 3A/INCONNEL-152 2.284 0
7811 C407 BLOCK OVERLAY CC-86 N/A 4"-6" CIRCUMFERENTIAL SEAM SA 508, 5"
0 78110406 CC-86 N/A_______
BLOCK TUBESHEET OVERLAY RADIUS CC-87 N/A BLOCK A508, GR2 5.5-6.5 0
781 1C107 CC-88 N/A PRIMARY NOZZLE INNER SA-508, GR 3 CL 1 6"
0 SWRI-D-70433 CC-88 N/A_______
RADIUS BLOCK CC-89 N/A TUBESHEET OVERLAY FLAT SA-508, CL2/INCONNEL 52 1.25" 0
7811 C404 FACE BLOCK OVERLAY I
I I
CC-90 N/A 2"-4" CIRCUMFERENTIAL SEAM SA508,CL3A 3.025" 0
78110405 CC-90 N/A BLOCK SA50,ICLA_3.25"__781D40 Revision 0 89 Prepared for Constellation Energy by jlrdki.l CbrMnpla; Irl
Constellation Energy-Calvert Cliffs Units I & 2 Calvert Cliffs Nuclear Power Plant Inservice Inspection Program 10.0 RECORDS AND REPORTS 10.1 Records and reports for the Inservice Inspection Program, outage examination schedules, examination
- results, procedures, certifications,
- test, repairs, and replacements are maintained in accordance with Constellation Energy procedures, and meet the requirements of ASME,Section XI, Article IWA-6000 and Code Case N-532-4.
10.2 The ISI summary reports are prepared in accordance with Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report. The summary report for steam generator tube inspections is created by the implementation of EN-4-106, Steam Generator Surveillance Program, in accordance with the requirements contained in CCNPP Technical Specification 5.6.9.
10.3 The CCNPP Repair/Replacement Program shall be prepared in accordance with IWA-4150.
Upon completion of all activities associated with the Repair/Replacement Program, Form NIS-2A shall be prepared and maintained by CCNPP.
10.4 The Owner's Activity Report, Form OAR-1 shall be processed within 90 calendar days of the completion of each refueling outage. As a minimum it shall contain the following:
- A listing of items with flaws or relevant conditions that exceed the acceptance criteria and the required evaluation to determine acceptability for continued service shall be provided in the format of Table 1 of Code Case N-532-4.
- An abstract for repair/replacement activities that were required due to an item containing a flaw or relevant condition that exceed acceptance criteria of Section XI, Division 1 shall be provided in the format of Table 2 of Code Case N-532-4. This information is required even if'the discovery of the flaw or relevant condition that necessitated the repair/replacement activity did not result from an examination or test required by Section XI, Division 1. If the acceptance criteria for a particular item is not specified in Section XI, Division 1, the provisions of IWA-3100(b) shall be used to determine which repair/replacement activities are required to be included in the abstract.
If no items met the criteria of (a) or (b), the term "None" shall be recorded on the applicable table.
If there are multiple inspection plans with different intervals, periods, Editions, or Addenda, they shall be identified on Form OAR-I.
The completed Form OAR'-1 shall be submitted to the regulatory and enforcement authorities having jurisdiction at the plant site.
Revision 0 90 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program 11.0 INSERVICE INSPECTION DATA MANAGEMENT SOFTWARE 11.1 The Inservice Inspection Data Management software is comprised of a series of program modules assembled in a comprehensive software package entitled the Iddeal Software Suite. This software suite is NETWORK based within the Constellation Energy system and is accessible by any computer within Constellation Energy.
11.2 Access to the Iddeal Software Suite is obtained through the Windows Citrix TM icon entitled the "Iddeal Software Suite." The software has limited access to protect the contents of the databases. Access to the software is limited through the use of a user defined password and Active Directory membership. The following programs comprise the IDDEAL SOFTWARE SUITE:
IDDEAL - IDDEAL is used to track and progress completion of outage examinations. Generates various reports to status examinations.
- SCHEDULEWORKS - Maintains the complete ISI NDE database. Outage schedules are assembled and progressed for interval and period statistics. The 90 day report is generated from this software.
CERTWORKS - Maintains and tracks the personnel certifications of inspectors.
EQUIPWORKS - Maintains and tracks the NDE equipment certifications.
- SNUBWORKS - Used to track all work associated with plant snubbers.
11.3 The Program Owner or designate have DATABASE ADMINISTRATOR authority for the software.
All security access and levels of access to the Iddeal Software Suite are provided by these administrators.
11.4 All software users, aside from the ADMINISTRATORS, will ACCESS" or "MODIFY ACCESS" based their security administrators.
be limited to,"READ ONLY level assignment by the Prepared for Constellation Energy by l ;=rwQQpI;;.
kIn.
Revision 0 91
Constellation Energy*
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program 12.0 COMPONENT & ISOMETRIC DRAWINGS 12.1 The Component and Isometric Drawings Index identifies the welds and components required to be examined under the rules of ASME Code,Section XI, 2004 Edition. Other Plant Controlled drawings, marked-up sketches, and the Iddeal Software Suite, ScheduleWorks database will be used when additional information is required.
The piping integral attachment welds for examination categories B-K-1 and C-C are identified on the corresponding Component Support drawings.
The Code boundaries shown on these sketches were obtained from the plant controlled P&ID drawings.
Revision 0 92 Prepared for Constellation Energy by tdm
.Ia
Constellation Energy#
Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI ISI Drawing Design Reference Drawing Drawing No.
Description Design Drawing Description A-1 Reactor Pressure Vessel Welds 12017-01 Pressure Vessel Welding and Machining RPV Closure Head - Meridianal & Circumferential A-2 Welds 12017-0016 Meridianal & Circumferential Welds A-3 Cross Section Views Of Pressurizer 12019-15 Vessel Assembly & Final Machining A-4 Typical Steam Generator Tube Sheet & Lower Head 12010-0088 S.G. Tube Sheet & Lower Head Steam Generator Nos. 11 & 12 High Pressure Head A-5 Assembly 12010-0088 S.G. Tube Sheet & Lower Head A-6 Reactor Coolant System Layout 12024-0014 General Arrangement Plan Piping A-7 42-1n. Reactor Coolant Line 42-RC-1 1 12024-04 Piping Details & Assembly A-7 42-1n. Reactor Coolant Line 42-RC-1 1 12024-05 Nozzle Details A-8 30-In. Reactor Coolant Line 30-RC-11A 12024-0006 Nozzle Details Piping A-8 30-In. Reactor Coolant Line 30-RC-11A 12024-0009 Piping Details & Assembly Piping A-8 30-In. Reactor Coolant Line 30-RC-1 1A 12024-04 Piping Details & Assembly A-8 30-In. Reactor Coolant Line 30-RC-1 1A 12024-08 Piping Details & Assembly A-9 30-In. Reactor Coolant Line 30-RC-1 1 B 12024-04 Piping Details & Assembly A-9 30-In. Reactor Coolant Line 30-RC-1 1 B 12024-08 Piping Details & Assembly A-10 42-1n. Reactor Coolant Line 42-RC-1 2 12024-04 Piping Details & Assembly Revision 0 93 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I
_Component and Isometric Drawings Index isl ISIISI 1SIg Design Design Reference Drawing Drawing No.
Description Design Drawing No.
Description A-10 42-In. Reactor Coolant Line 42-RC-12 12024-05 Nozzle Details A-11 30-In.Reactor Coolant Line 30-RC-12A 12024-04 Piping Details & Assembly A-11 30-In.Reactor Coolant Line 30-RC-12A 12024-08 Piping Details & Assembly A-12 30-In.Reactor Coolant Line 30-RC-12B 12024-0009 Piping Details & Assembly Piping A-12 30-In.Reactor Coolant Line 30-RC-12B 12024-04 Piping Details & Assembly A-12 30-In.Reactor Coolant Line 30-RC-12B 12024-08 Piping Details & Assembly A-13 12-In. Pressurizer Surge Line 12024-0009 Piping Details & Assembly Piping Piping Isometric Containment Structure 12" Surge Line From Pressurizer To Hot Leg A-13 12-In. Pressurizer Surge Line 60798 Steam Generator.
12-and 14-In. Shutdown Cooling Lines 12-SC-1004 A-14 and 14-SC-1004 91101 Shutdown Cooling Piping Inside Containment 12-and 14-In. Shutdown Cooling Lines 12-SC-1004 A-14 and 14-SC-1004 91146 Reactor Shutdown Cooling 12-In. Safety Injection Line 12-SI-1009 and 6-In. Safety Safety Injection Piping System Tank No. 11 A, A-15 Injection Line 6-SI-1001 91100SH0001 Leg No. 11A 12-In. Safety Injection Line 12-SI-1010 and 6-In. Safety Safety Injection Piping System Tank No. 11B, A-16 Injection Line 6-SI-1002 91100SH0002 Leg No. 11B 12-in. Safety Injection Line 12-SI-101 1 and 6-In. Safety Safety Injection Piping System Tank No. 12A, A-17 Injection Line 6-SI-1003 91100SH0003 Leg No. 12A 12-In. Safety Injection Line 12-SI-1012 and 6-In. Safety Safety Injection Piping System Tank No. 12B, A-18 Injection Line 6-SI-1004 91100SH0004 Leg No. 12B Revision 0 94 Prepared for Constellation Energy by
Constellation Energy-
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1
~Comnonent and Isometric Drawinas Index ComnonentRfernc DrawingrcDrwiasIne ISI ISI Drawing Design Drawing No.
Design Reference Drawing Drawing No.
Description Description 4-In. Pressurizer Spray Line 4-PS-1003 and 4-PS-A-19 1003M 12600A-0750SH0001 Pressurizer Spray Valve Relocation - Unit 1 4-In. Pressurizer Spray Line 4-PS-1003 and 4-PS-A-19 1003M 12600A-0750SH0002 Pressurizer Spray Valve Relocation - Unit 1 4-In. Pressurizer Spray Line 4-PS-1003 and 4-PS-A-19 1003M 91099SH0002 Pressure Spray System 3-In. Pressurizer Spray Line 3-PS-1001 and 3-PS-A-20 1001M 91099SH0001 Pressure Spray System 3-In. Pressurizer Spray Line 3-PS-1001 and 3-PS-A-20 1001M 91099SH0002 Pressure Spray System 3-In. Pressurizer Spray Line 3-PS-1002 and 3-PS-A-21 1002M 91099SH0001 Pressure Spray System 3-In. Pressurizer Spray Line 3-PS-1002 and 3-PS-A-21 1002M 91099SH0002 Pressure Spray System 2 1/2 -In. Safety and Relief Lines 2 2 SR-1 003 and 2 1/2-SR-1007; and 4-In. Safety and Relief Lines 4-SR-1001 Pressurizer Relief Valve Piping Inside A-22 and 4-SR-1005 91098SH0002 Containment 2 1/2 -In. Safety and Relief Lines 2 2 SR-1 004 and 2 1/2A-SR-1008; and 4-In. Safety and Relief Lines 4-SR-1002 Pressurizer Relief Valve Piping Inside A-23 and 4-SR-1006 91098SH0001 Containment Regen. Heat Exch. Inlet From Letdown Loop A-24 2-In. Letdown Lines 2-LD-1 004 FSK-MP-0455 12A CC-3 & CC-15 Regen. Heat Exch. Inlet From Letdown Loop A-24 2-In. Letdown Lines 2-LD-1004 FSK-MP-456 12a Unit 1 CC-3 (Nuclear CL I)
Regen. Heat Exch. Inlet From Letdown Loop A-24 2-In. Letdown Lines 2-LD-1004 FSK-MP-456-H 12a Unit 1 CC-3 (Nuclear CL I)
Regenerative Heat Exchanger No.11 To Auxiliary Spray CC-5 Nuclear Class I, CC-7 A-25 2-In. Charging Line 2-CV-1003 FSK-MP-0490 Nuclear Class II Revision 0 95 Prepared for Constellation Energy by
Constellation Energy-
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1
_Component and Isometric Drawings Index ISI ISI Drawing No.
Design Reference Drawing
- Drawing No.
Description Design Drawing Description Regenerative Heat Exchanger No.11 To Charging Line Loop 12B CC-5 Nuclear Class I A-26 2-In. Charging Line 2-CV-1004 FSK-MP-0491 CC-7 Nuclear Class II Regenerative Heat Exchanger No. 11 To Charging Loop Line 11 A-Unit 1 CC-5 Nuclear A-27 2-In. Charging Line 2-CV-1005 and 2-CV-1006 FSK-MP-0492 Class I CC-7 Nuclear Class II Primary Loop Drains Steam Gen. #11 CC-9 &
A-28 2-In. Drain Line 2-DR-1003 FSK-MP-575 HC-2 Primary Loop Drains Steam Gen. #11 CC-9 &
A-28 2-In. Drain Line 2-DR-1003 FSK-MP-575-H-1 HC-2 Primary Loop Drains Steam Gen. #11 CC-9 &
A-29 2-In. Drain Line 2-DR-1004 FSK-MP-1053 HC-2 Primary Loop Drains Steam Gen. #11 CC-9 &
A-29 2-In. Drain Line 2-DR-1004 FSK-MP-1053-H HC-2 Primary Loop Drains - Stm. Gen. #12 CC-9 &
A-30 2-In. Drain Line 2-DR-1005 FSK-MP-576 HC-2 Primary Loop Drains - Stm. Gen. #12 CC-9 &
A-30 2-In. Drain Line 2-DR-1005 FSK-MP-576-H-1 HC-2 Primary Loop Drains - Stm. Gen. #12 CC-9 &
A-31 2-In. Drain Line 2-DR-1006 FSK-MP-1006 HC-2 Primary Loop Drains - Stm. Gen. #12 CC-9 &
A-31 2-In. Drain Line 2-DR-1006
-FSK-MP-1006-H HC-2 A-32 2-In. Drain Line 2-DR-1007 FSK-MP-597 Reactor Coolant System Drains CC-9 & HC-2 A-33 Typical Reactor Coolant Pump Body N/A Reactor Coolant Pump Body (Typ.)
B-1 Steam Generator Outline N/A Steam Generator Outline B-2 Steam Generator Upper Vessel 91483 Steam Generator Snubber Identification Revision 0 96 Prepared for Constellation Energy by
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI ISI Drawing No,,
Design Reference Drawing Design Drawing No.
Drawing No.
Description Description B-3 Shutdown Cooling Heat Exchanger Outline 12015-12 Shutdown Cooling Heat Exchanger B-4 Regenerative Heat Exchanger Weld Identification 12108-01 Regenerative Heat Exchanger B-5 24-1n. Safety Injection Line 24-SI-1201 91097SH0012 Shutdown Heat Exchanger Piping B-6 24-1n. Safety Injection Line 24-SI-1202 91097SH0013 Shutdown Heat Exchanger Piping B-7 18-1n. Safety Injection Line 18-SI-1203 91097SH0012 Shutdown Heat Exchanger Piping B-7 18-1n. Safety Injection Line 18-SI-1203 91160 From Refueling Water Tank B-7 18-In. Safety Injection Line 18-SI-1203 91175 Refueling Water Supply B-7 18-1n. Safety Injection Line 18-SI-1203 91203 From Refueling Water Tank B-8 18-In. Safety Injection Line 18-S1-1204 91097SH0013 Shutdown Heat Exchanger Piping B-8 18-1n. Safety Injection Line 18-SI-1204 91165 From Refueling Water Tank B-9 14-1n. Safety Injection Line 14-SI-1201 91097SH001 1 Shutdown Heat Exchanger Piping B-9 14-1n. Safety Injection Line 14-SI-1201 91097SH0013 Shutdown Heat Exchanger Piping B-10 14-1n. Safety Injection Line 14-SI-1202 91097SH001 1 Shutdown Heat Exchanger Piping B-10 14-1n. Safety Injection Line 14-SI-1202 91097SH0012 Shutdown Heat Exchanger Piping B-11 14-1n. Safety Injection Line 14-SI-1201 & 14-SI-1202 91097SH0011 Shutdown Heat Exchanger Piping Revision 0 97 Prepared for Constellation Energy by
Constellation Energy Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI ISI Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description.
B-12 14-1n. Safety Injection Line 14-SI-1210 91097SH0012 Shutdown Heat Exchanger Piping B-13 14-1n. Safety Injection Line 14-SI-1211 91097SH0013 Shutdown Heat Exchanger Piping B-14 12-1n. Safety Injection Line 12-SI-1213 91097SH0001 Shutdown Heat Exchanger Piping B-15 12-1n. Safety Injection Line 12-SI-1214 91097SH0001 Shutdown Heat Exchanger Piping B-15 12-In. Safety Injection Line 12-SI-1214 91097SH0026 Shutdown Heat Exchanger Piping Component Cooling Pump Room Piping Safety B-15 12-In. Safety Injection Line 12-SI-1214 91109SH0001 Injection System B-15 12-1n. Safety Injection Line 12-SI-1214 91168 L.P. Safety Injection B-16 12-1n. Safety Injection Line 12-SI-1216 91097SH0006 Shutdown Heat Exchanger Piping Component Cooling Pump Room Piping B-16 12-1n. Safety Injection Line 12-SI-1216 91110 Shutdown Cooling System B-17 10-In. Safety Injection Line 10-SI-1202 91097SH0026 Shutdown Heat Exchanger Piping B-18 10-In. Safety Injection Line 10-SI-1203 12530A-0010 Area #16, #12 LPSI Pump Discharge B-18 10-In. Safety Injection Line 10-SI-1203 91097SH0014 Shutdown Heat Exchanger Piping B-18 10-In. Safety Injection Line 10-SI-1203 91097SH0026 Shutdown Heat Exchanger Piping B-19 10-In. Safety Injection Line 10-SI-1206 91097SH0012 Shutdown Heat Exchanger Piping B-20 10-In. Safety Injection Line 10-SI-1207 91097SH0013 Shutdown Heat Exchanger Piping Revision 0 98 Prepared for Constellation Energy by
Constellation Energy'
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1
_Component and Isometric Drawings Index ISI Dai Design DDesign Reference Drawing Drawing No.
Description Design Drawing No.
Description B-21 8-In. Safety Injection Line 8-SI-1207 91097SH0015 Shutdown Heat Exchanger Piping B-22 8-In. Safety Injection Line 8-SI-1208 91097SH0015 Shutdown Heat Exchanger Piping Component Cooling Pump Room Piping B-23 8-In. Safety Injection Line 8-SI-1218 91110 Shutdown Cooling System B-24 8-In. Safety Injection Line 8-SI-1220 91154 Safety Injection Pump Discharge B-24 8-In. Safety Injection Line 8-SI-1220 91168 L.P. Safety Injection B-25 8-In. Safety Injection Line 8-SI-1221 91168 L.P. Safety Injection B-26 6-In. Safety Injection Line 6-SI-1201 91097SH0017 Shutdown Heat Exchanger Piping B-26 6-In. Safety Injection Line 6-SI-1201 91126 H.P. Safety Injection B-27 6-In. Safety Injection Line 6-SI-1203 91097SH0018 Shutdown Heat Exchanger Piping B-28 6-In. Safety Injection Line 6-SI-1204 91097SH0016 Shutdown Heat Exchanger Piping B-28 6-In. Safety Injection Line 6-SI-1204 91169 L.P. Safety Injection B-29 6-In. Safety Injection Line 6-SI-1205 91143 L.P. Safety Injection System B-29 6-In. Safety Injection Line 6-SI-1205 91168 L.P. Safety Injection B-30 6-In. Safety Injection Line 6-SI-1206 91097SH0015 Shutdown Heat Exchanger Piping B-31 6-In. Safety Injection Line 6-SI-1207 91154 Safety Injection Pump Discharge Revision 0 99 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I
_Component and Isometric Drawings Index ISI ISI Drawing Design Reference Drawing Drawing No.
Description Design iDrawing No.
Description B-32 6-In. Safety Injection Line 6-SI-1208 91144 L.P. Safety Injection System B-32 6-In. Safety Injection Line 6-SI-1208 91168 L.P. Safety Injection B-33 6-In. Safety Injection Line 6-SI-1209 91097SH0017 Shutdown Heat Exchanger Piping Safety Injection Piping System Tank No. 11A, B-34 6-In. Safety Injection Line 6-SI-1210 91100SH0001 Leg No. 11A B-34 6-In. Safety Injection Line 6-SI-1210 91143 L.P. Safety Injection System B-35 6-In. Safety Injection Line 6-SI-1211 91097SH0015 Shutdown Heat Exchanger Piping Safety Injection Piping System Tank No. 11 B, B-36 6-In. Safety Injection Line 6-SI-1212 91100SH0002 Leg No. 11B B-36 6-In. Safety Injection Line 6-SI-1212 91144 L.P. Safety Injection System Safety Injection Piping System Tank No. 12A, B-37 6-In. Safety Injection Line 6-SI-1213 91100SH0003 Leg No. 12A B-37 6-In. Safety Injection Line 6-SI-1213 91154 Safety Injection Pump Discharge Safety Injection Piping System Tank No. 12B, B-38 6-In. Safety Injection Line 6-SI-1214 91100SH0004 Leg No. 12B B-38 6-In. Safety Injection Line 6-SI-1214 91154 Safety Injection Pump Discharge B-39 6-In. Safety Injection Line 6-SI-1216 91154 Safety Injection Pump Discharge Component Cooling Pump Room Piping B-40 6-In. Safety Injection Line 6-SI-1217 91110 Shutdown Cooling System B-41 6-In. Safety Injection Line 6-SI-1219 91097SH0018 Shutdown Heat Exchanger Piping Revision 0 100 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1
_Component and Isometric Drawings Index ISI ISI Drawing Design Drawing No.
Design Reference Drawing Drawing No.
Description Description B-42 4-In. Safety Injection Line 4-SI-1204 91097SH0002 Shutdown Heat Exchanger Piping B-42 4-In. Safety Injection Line 4-SI-1204 91097SH0015 Shutdown Heat Exchanger Piping B-43 4-In. Safety Injection Line 4-SI-1205 91097SH0006 Shutdown Heat Exchanger Piping B-43 4-In. Safety Injection Line 4-SI-1205 91097SH0O18 Shutdown Heat Exchanger Piping B-44 4-In. Safety Injection Line 4-SI-1206 91097SH0010 Shutdown Heat Exchanger Piping B-45 4-In. Safety Injection Line 4-SI-1207 91126 H.P. Safety Injection B-46 4-In. Safety Injection Line 4-SI-1208 91097SH0016 Shutdown Heat Exchanger Piping B-46 4-In. Safety Injection Line 4-SI-1208 91097SH0017 Shutdown Heat Exchanger Piping B-47 4-In. Safety Injection Line 4-SI-1209 91169 L.P. Safety Injection B-48 4-In. Safety Injection Line 4-SI-1 210 91153 Safety Injection System B-48 4-In. Safety Injection Line 4-SI-1210 91169 L.P. Safety Injection B-49 4-In. Safety Injection Line 4-SI-1211 91126 H.P. Safety Injection B-49 4-In. Safety Injection Line 4-SI-1211 91152 Safety Injection System B-50 3-In. Safety Injection Line 3-SI-1201 91097SH0017 Shutdown Heat Exchanger Piping B-51 3-In. Safety Injection Line 3-SI-1202 91097SH0017 Shutdown Heat Exchanger Piping Revision 0 101 Prepared for Constellation Energy by i*kl"
Constellation Energy
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI 1SI Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description--
B-52 3-In. Safety Injection Line 3-SI-1203 91097SH0016 Shutdown Heat Exchanger Piping B-53 3-In. Safety Injection Line 3-SI-1214 12550A-07 Area #26 H.P. Safety Injection B-53 3-In. Safety Injection Line 3-SI-1214 FSK-MP-744 Safety Injection To Primary Loop CC-13 B-54 3-In. Safety Injection Line 3-SI-1215 12550A-07 Area #26 H.P. Safety Injection Safety Injection - Unable To Read Drawing B-54 3-In. Safety Injection Line 3-SI-1215 FSK-MP-743 Title Safety Injection - Unable To Read Drawing B-54 3-In. Safety Injection Line 3-SI-1215 FSK-MP-743-H Title B-55 3-In. Safety Injection Line 3-SI-1220 12536A-09 Area #19, Safety Injection Safety Injection System - Unit No. 1 CC-6, B-55 3-In. Safety Injection Line 3-SI-1220 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-55 3-In. Safety Injection Line 3-SI-1220 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
B-56 3-In. Safety Injection Line 3-SI-1221 12536A-08 Area #19, Safety Injection Safety Injection To Primary Loop CC-6 & CC-B-57 2-In. Safety Injection Line 2-SI-1202 FSK-MP-750 13 (Nuclear CL II)
Safety Injection To Primary Loop CC-6 & CC-B-57 2-In. Safety Injection Line 2-SI-1202 FSK-MP-750-H 13 (Nuclear CL II)
Charging Pump 11, 12 & 13 Discharge To B-58 2-In. Safety Injection Line 2-SI-1203 FSK-MP-1 668 HPSI HDR CC-6 Charging Pump Discharge To HPSI Header B-58 2-In. Safety Injection Line 2-SI-1203 FSK-MP-432 CC-6 & CC-7 (Nuclear CL II)
Safety Injection - Unable To Read Drawing B-59 2-In. Safety Injection Line 2-SI-1204 FSK-MP-743 Title Revision 0 102 Prepared for Constellation Energy by
Constellation Energy
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI ISI Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description Safety Injection - Unable To Read Drawing B-59 2-In. Safety Injection Line 2-SI-1204 FSK-MP-743-H Title Safety Injection System - Unit No. 1 CC-6, B-60 2-In. Safety Injection Line 2-SI-1205 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-60 2-In. Safety Injection Line 2-SI-1205 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-61 2-In. Safety Injection Line 2-SI-1206 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
Safety Injection - Unable To Read Drawing B-62 2-In. Safety Injection Line 2-SI-1207 FSK-MP-743-H Title B-63 2-In. Safety Injection Line 2-SI-1208 FSK-MP-744 Safety Injection To Primary Loop CC-13 Safety Injection To Primary Loop DC-1 & CC-B-63 2-In. Safety Injection Line 2-SI-1208 FSK-MP-747 13 (Nuclear CL II)
Safety Injection - Unable To Read Drawing B-64 2-In. Safety Injection Line 2-SI-1209 FSK-MP-743 Title Safety Injection - Unable To Read Drawing B-64 2-In. Safety Injection Line 2-SI-1209 FSK-MP-743-H Title Safety Injection System - Unit No. 1 CC-6, B-65 2-In. Safety Injection Line 2-SI-1210 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-65 2-In. Safety Injection Line 2-SI-1210 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-66 2-In. Safety Injection Line 2-SI-1211 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-66 2-In. Safety Injection Line 2-SI-1211 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
Safety Injection - Unable To Read Drawing B-67 2-In. Safety Injection Line 2-SI-1212 FSK-MP-743-H Title Revision 0 103 Prepared for Constellation Energy by
Constellation Energy
- Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1 Component and Isometric Drawings Index ISI ISI Drawing Design Reference Drawing Drawing No.
Description.
Design Drawing No.
Description Safety Injection System - Unit No. 1 CC-6, B-68 2-In. Safety Injection Line 2-SI-1213 FSK-MP-698-H-1 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-69 2-In. Safety Injection Line 2-SI-1214 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection To Primary Loop DC-1 & CC-B-70 2-In. Safety Injection Line 2-SI-1215 FSK-MP-747 13 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-71 2-In. Safety Injection Line 2-SI-1216 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection System - Unit No. 1 CC-6, B-72 2-In. Safety Injection Line 2-SI-1217 FSK-MP-698 CC-13, Dc-1 (Nuclear CL II)
Safety Injection - Unable To Read Drawing B-73 2-In. Safety Injection Line 2-SI-1218 FSK-MP-743 Title B-74 2-In. Safety Injection Line 2-SI-1219 FSK-MP-744 Safety Injection To Primary Loop CC-13 Containment Spray & Safety Injection Piping B-75 2-In. Safety Injection Line 2-SI-1221 FSK-MP-1783 DC-2 Containment Spray & Safety Injection Piping B-76 2-In. Safety Injection Line 2-SI-1222 FSK-MP-1783 DC-2 Containment Spray & Safety Injection Piping B-77 2-In. Safety Injection Line 2-SI-1223 FSK-MP-119 DC-2 Containment Spray & Safety Injection Piping B-78 2-In. Safety Injection Line 2-SI-1224 FSK-MP-103 GC-7 & DC-2 Containment Spray & Safety Injection Piping B-78 2-In. Safety Injection Line 2-SI-1224 FSK-MP-119 DC-2 Containment Spray & Safety Injection Piping B-79 2-In. Safety Injection Line 2-SI-1225 FSK-MP-1782 DC-2 & GC-7 Containment Spray & Safety Injection Piping B-79 2-In. Safety Injection Line 2-SI-1225 FSK-MP-1783 DC-2 Revision 0 104 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 1 Component and Isometric Drawings Index ISI 1
aig Design Drawing No.
Design Reference Drawing Drawing No.
Description Description Charging Pump Discharge To HPSI Header B-80 2-In. Safety Injection Line 2-SI-1226 FSK-MP-432 CC-6 & CC-7 (Nuclear CL II)
B-81 10-In. Containment Spray Line 10-CS-1202 91097SH0030 Shutdown Heat Exchanger Piping B-82 10-In. Containment Spray Line 10-CS-1204 91097SH0005 Shutdown Heat Exchanger Piping B-83 10-In. Containment Spray Line 10-CS-1218 91097SH0001 Shutdown Heat Exchanger Piping B-83 10-In. Containment Spray Line 10-CS-1218 91097SH0030 Shutdown Heat Exchanger Piping B-84 8-In. Containment Spray Line 8-CS-1203 91097SH0005 Shutdown Heat Exchanger Piping B-85 8-In. Containment Spray Line 8-CS-1204 91097SH0030 Shutdown Heat Exchanger Piping B-86 (Fig. B-86A) 8-In. Containment Spray Line 8-CS-1205 91097SH0002 Shutdown Heat Exchanger Piping B-86 (Fig. B-86A) 8-In. Containment Spray Line 8-CS-1205 91145 Containment Spray B-86 (Fig. B-86A) 8-In. Containment Spray Line 8-CS-1205 91172 Containment Spray B-87 Spray System Header No. 2 Inside (Fig. B-86B) 8-In. Containment Spray Line 8-CS-1205 91096 Containment B-87 (Fig. B-86B) 8-In. Containment Spray Line 8-CS-1205 91145 Containment Spray B-88 (Fig. B-87A) 8-In. Containment Spray Line 8-CS-1206 91097SH0006 Shutdown Heat Exchanger Piping B-88 (Fig. B-87A) 8-In. Containment Spray Line 8-CS-1206 91173 Containment Spray B-89 Spray System Header No. 1 Inside (Fig. B-87B) 8-In. Containment Spray Line 8-CS-1206 91095 Containment Revision 0 105 Prepared for Constellation Energy by
Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI DSI Drawing Design Reference Drawing Drawing No.
Description Design Drawing No.
Description B-89 (Fig. B-87B) 8-In. Containment Spray Line 8-CS-1206 91162 Containment Spray B-89 (Fig. B-87B) 8-In. Containment Spray Line 8-CS-1206 91173 Containment Spray B-90 (Fig. B-88) 14-1n. Shutdown Cooling Line 14-SC-1203 91097SH001 1 Shutdown Heat Exchanger Piping B-90 (Fig. B-88) 14-1n. Shutdown Cooling Line 14-SC-1203 91146 Reactor Shutdown Cooling B-90 (Fig. B-88) 14-1n. Shutdown Cooling Line 14-SC-1203 91166 Shut Down Cooling B-91 (Fig. B-89) 14-1n. Shutdown Cooling Line 14-SC-1204 91097SH0011 Shutdown Heat Exchanger Piping B-92 (Fig. B-90) 12-1n. Shutdown Cooling Line 12-SC-1208 91097SH0002 Shutdown Heat Exchanger Piping B-92 (Fig. B-90) 12-1n. Shutdown Cooling Line 12-SC-1208 91097SH0006 Shutdown Heat Exchanger Piping B-93 (Fig. B-91) 12-In. Shutdown Cooling Line 12-SC-1213 91097SH0001 Shutdown Heat Exchanger Piping IB-94 (Fig. B-92) 12-In. Shutdown Cooling Line 12-SC-1215 91097SH0002 Shutdown Heat Exchanger Piping B-94 "Component Cooling Pump Room Piping Safety (Fig. B-92) 12-In. Shutdown Cooling Line 12-SC-1215 91109SH0001 Injection System B-95 (Fig. B-93) 10-In. Shutdown Cooling Line 12-SC-1214 91097SH0001 Shutdown Heat Exchanger Piping B-95 (Fig. B-93) 10-In. Shutdown Cooling Line 12-SC-1214 91097SH0005 Shutdown Heat Exchanger Piping B-96 (Fig. B-94) 8-In. Shutdown Cooling Line 12-SC-1206 91097SH0011 Shutdown Heat Exchanger Piping Revision 0 106 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program CCNPP Unit I Comnonent and Isometric Drawinas Index ISI DSI Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description B-97 (Fig. B-95) 8-In. Shutdown Cooling Line 12-SC-1207 91097SH0006 Shutdown Heat Exchanger Piping B-98 (Fig. B-96) 8-In. Shutdown Cooling Line 12-SC-1209 91097SH0002 Shutdown Heat Exchanger Piping B-99 (Fig. B-97) 8-In. Shutdown Cooling Line 12-SC-1210 91097SH0002 Shutdown Heat Exchanger Piping B-100 Main Steam Piping Stress Isometric Auxiliary (Fig. B-98) 36-1n. Main Steam Line 36-MS-1201 60345SH0001 Building B-101 Main Steam Piping Stress Isometric Auxiliary (Fig. B-99) 36-In. Main Steam Line 36-MS-1202 60345SH0001 Building B-102 Main Steam Piping Stress Isometric Auxiliary (Fig. B-100) 34-In. Main Steam Line 34-MS-1201 60345SH0001 Building B-102 (Fig. B-100) 34-In. Main Steam Line 34-MS-1201 91309 Main Steam Encapsulation B-103 Main Steam Piping Stress Isometric Auxiliary (Fig. B-101) 34-In. Main Steam Line 34-MS-1202 60345SH0001 Building B-103 (Fig. B-101) 34-1n. Main Steam Line 34-MS-1202 91309 Main Steam Encapsulation B-104 Main Steam Piping Stress Isometric Auxiliary (Fig. B-102) 34-1n. Main Steam Line 34-MS-1204 60345SH0001 Building B-105 Main Steam Piping Stress Isometric Auxiliary (Fig. B-103) 34-1n. Main Steam Line 34-MS-1205 60345SH0001 Building B-106 (Fig. B-104) 6-In. Main Steam Line 6-MS-1207 B-107 (Fig. B-105) 6-In. Main Steam Line 6-MS-1208 B-108 (Fig. B-106) 6-In. Main Steam Line 6-MS-1237 Revision 0 107 Prepared for Constellation Energy by
0 Constellation Energy" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit I Component and Isometric Drawings Index ISI ISIg Design Design Reference Drawing Drawing No.
Description Design Drawing No.
Description B-109 (Fig. B-107) 6-In. Main Steam Line 6-MS-1238 B-110 (Fig. B-108) 16-1n. Feed Water Line 16-FW-1201 91093 Feedwater-Unit No. 1 Inside Containment B-110 (Fig. B-108) 16-1n. Feed Water Line 16-FW-1201 91310SH0001 Isometric Feedwater Aux. Bldg.
B-111 (Fig. B-109) 16-1n. Feed Water Line 16-FW-1202 91093 Feedwater-Unit No. 1 Inside Containment B-111 (Fig. B-109) 16-1n. Feed Water Line 16-FW-1202 91310SH0001 Isometric Feedwater.Aux. Bldg.
B-112 (Fig. B-110) 16-1n. Feed Water Line 16-FW-1218 91093 Feedwater-Unit No. 1 Inside Containment B-113 (Fig. B-111) 16-1n. Feed Water Line 16-FW-1 219 91093 Feedwater-Unit No. 1 Inside Containment B-114 (Fig. B-112)
Valve Body Welds N/A Typical Valve Body-To-Flange Weld B-115 4-in. Auxiliary Feedwater 4-AF-1014 91103 Auxiliary Feedwater Inside Containment B-1 16 4-in. Auxiliary Feedwater 4-AF-1 015 91103 Auxiliary Feedwater Inside Containment B-117 4-in. Auxiliary Feedwater 4-AF-1 004 91140 Aux. Feedwater B-118 4-in. Auxiliary Feedwater 4-AF-1 003 91140 Aux. Feedwater B-119 4-in. Auxiliary Feedwater 4-AF-1010 60556 Piping Isometric Auxiliary Feedwater System B-120 4-in. Auxiliary Feedwater4-AF-1011 60556 Piping Isometric Auxiliary Feedwater System Revision 0 108 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2
_Component and Isometric Drawings Index ISI Drawing Design Drawing No.
Design Reference Drawing Drawing No.
Description Description A-1 Reactor Pressure Vessel Weld Identification 13017-02 Vessel "As-Built" Dimensions A-2 Identifications 12017-0016 PWR RPV Closure Head Meridianal & Circumferential Weld Closure Head Forming And Welding 172" l.D.
A-3 Cross Section Views of Pressurizer 12019-0015 Vessel Assembly & Final Machining A-4 Typical Steam Generator Tube Sheet & Lower heard 12010-0088 Generator Steam Generator Nos. 21 & 22 High Pressure Head High Pressure Head Details And Assembly A-5 Assembly 12010-0044 Steam Generator A-6 Reactor Coolant System Layout 12024-42 General Arrangement Plan A-7 42-in. Reactor Coolant Line 42-RC-21 12024-04 Piping Details & Assembly A-7 42-in. Reactor Coolant Line 42-RC-21 12024-05 Nozzle Details Piping A-7 42-in. Reactor Coolant Line 42-RC-21 12024-0006 Nozzle Details Piping A-7 42-in. Reactor Coolant Line 42-RC-21 62729SH001 Reactor Coolant System A-8 42-in. Reactor Coolant Line 42-RC-22 12024-0006 Nozzle Details Piping A-8 42-in. Reactor Coolant Line 42-RC-22 12024-04 Piping Details & Assembly A-8 42-in. Reactor Coolant Line 42-RC-22 12024-05 Nozzle Details Piping A-8 42-in. Reactor Coolant Line 42-RC-22 62729SH001 Reactor Coolant System Revision 0 109 Prepared for Constellation Energy by
Constellation Energy-
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2
_Component and Isometric Drawings Index ISI3 ISID Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description A-9 30-in. Reactor Coolant Line 30-RC-21A 12024-0009 Piping Details & Assembly Piping \\
A-9 30-in. Reactor Coolant Line 30-RC-21A 12024-04 Piping Details & Assembly A-9 30-in. Reactor Coolant Line 30-RC-21A 12024-08 Piping Details & Assembly A-1 0 30-in. Reactor Coolant Line 30-RC-21 B 12024-04 Piping Details & Assembly A-1 0 30-in. Reactor Coolant Line 30-RC-21 B 12024-08 Piping Details & Assembly A-11 30-in. Reactor Coolant Line 30-RC-22A 12024-04 Piping Details & Assembly A-11 30-in. Reactor Coolant Line 30-RC-22A 12024-08 Piping Details & Assembly A-12 30-in. Reactor Coolant Line 30-RC-22B 12024-04 Piping Details & Assembly A-12 30-in. Reactor Coolant Line 30-RC-22B 12024-08 Piping Details & Assembly A-13 12-in. Pressurizer Surge Line 12024-0009 Piping Details & Assembly Piping A-13 12-in. Pressurizer Surge Line 60798 A-13 12-in. Pressurizer Surge Line 12024-0042 General Arrangement Plan A-13 12-in. Pressurizer Surge Line 62729SH001 Reactor Coolant System 12-and 14-in. Shutdown Cooling Lines 12-SC-2004 and A-14 14-SC-2004 91319 Shutdown Cooling Piping Inside Containment 12-and 14-in. Shutdown Cooling Lines 12-SC-2004 and A-14 14-SC-2004 91320 Reactor Shutdown Cooling Revision 0 110 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISI Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing No.
Description 12-in. Safety Injection Line 12-SI-2009 and 6-in. Safety Safety Injection Piping System (Tank No. 21B A-15 Injection Line 6-SI-2001 91298SH0002 Leg 21 B) 12-in. Safety Injection Line 12-SI-2010 and 6-in. Safety Safety Injection Piping System (Tank No. 21A A-16 Injection Line 6-SI-2002 91298SH0001 Leg 21A) 12-in. Safety Injection Line 12-SI-2011 and 6-in. Safety Safety Injection Piping System (Tank No. 22B A-17 Injection Line 6-SI-2003 91298SH0004 Leg 22B) 12-in. Safety Injection Line 12-SI-2012 and 6-in. Safety Safety Injection Piping System (Tank No. 22A A-1 8 Injection Line 6-SI-2004C 91298SH0003 Leg 22A)
A-1 9 4-in. Pressurizer Spray Line 4-PS-2003 91305SH0001 Pressurizer Spray System Unit No. 2 A-20 3-in. Pressurizer Spray Line 3-PS-2001 91305SH0001 Pressurizer Spray System Unit No. 2 A-20 3-in. Pressurizer Spray Line 3-PS-2001 91305SH0002 Pressurizer Spray System Unit No. 2 A-21 3-in. Pressurizer Spray Line 3-PS-2002 91305SH0001 Pressurizer Spray System Unit No. 2 A-21 3-in. Pressurizer Spray Line 3-PS-2002 91305SH0002 Pressurizer Spray System Unit No. 2 4-in. Safety and Relief Lines 4-SR-2001 and 4-SR-2005; 2.5-in. Safety and Relief Lines 2.5-SR-2003 and 2.5-SR-A-22 2007 91317SH0002 Pressurizer Relief Valve Piping Inside CTMT 2 4-in. Safety and Relief Lines 4-SR-2002 and 4-SR-2006; 2.5-in. Safety and Relief Lines 2.5-SR-2004 and 2.5-SR-A-23 2008 Regenerative Ht. Exch. #21 To Charging Line Loop Unit 2 CC-5 Nuclear CL I & CC-7 Nuclear A-24 2-in. Charging Lines 2-CV-2005 and 2-CV-2006 FSK-MP-3108-H CL II Regenerative Heat Exchanger No. 21 To Auxiliary Spray CC-5 Nuclear Class I CC-7 A-25 2-in. Charging Lines 2-CV-2018 FSK-MP-3106 Nuclear Class II Revision 0 ill Prepared for Constellation Energy by
Constellation Energy
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISI Drawing Design Reference Drawing Drawing No.
Description Design Drawing No.
Description Regenerative Heat Exchanger To Charging Loop Line CC-5 Nuclear Class I CC-7 Nuclear A-26 2-in. Charging Lines 2-CV-2021 FSK-MP-3107 Class II Primary Loop Drain - Steam Gen. 21A CC A-27 2-in. Drain Line 2-DR-2003 FSK-MP-3189 Nuclear CL I & HC-2-Nuclear CL II Primary Loop Drains Steam Gen. #21 (Nuclear A-28 2-in. Drain Line 2-DR-2004 FSK-MP-3141 CL I) CC-9 & HC-2 (Nuclear CL II)
Primary Loop Drain Steam Generator No. 22A A-29 2-in. Drain Line 2-DR-2005 FSK-MP-3194
& N.D.T. Relief Protection Primary Loop Drains Steam Gen. #22 (Nuclear A-30 2-in. Drain Line 2-DR-2006 FSK-MP-3140-H CL I) CC-9 & HC-2 (Nuclear CL II)
A-31 2-in. Drain Line 2-DR-2007 FSK-MP-3148 Reactor Coolant System Drains Regenerative Heat Exchanger Inlet From Letdown Loop 22A CC-3 (Nuclear CL I) & CC-A-32 2-in. Letdown Line 2-LD-2004 FSK-MP-3142-E 15 (Nuclear CL II)
A-33 Typical Reactor Coolant Pump Body N/A Reactor Coolant Pump Body (Typ.)
B-1 Steam Generator Outline 91484 Steam Generator Outline B-2 Steam Generator Upper Vessel - Layout of Supports 13010-0003 B-3 Shutdown Cooling Heat Exchanger Outline 62710SH0002 Shutdown Cooling Heat Exchanger Outline B-3 Shutdown Cooling Heat Exchanger Outline 62731SHOO03 Shutdown Cooling Heat Exchanger Outline Regenerative Heat Exchanger Weld Identification Line B-4 2-LD-2004 12108-01 Regenerative Heat Exchanger Weld Identification Line Regenerative Heat Exch. Inlet From Letdown B-4 2-LD-2004 FSK-MP-3142 Loop 22A CC-3 (Nuc CL I) & CC-15 (Nuc CL II)
Revision 0 112 Prepared for Constellation Energy by
Constellation Energy' Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2-
.Component andIsometric Drawings Index
- ISI ISI Drawing Di r
nNDesign Reference Drawing Drawing-No.
Description
_____DesgnDawin Description B-5 24-in. Safety Injection Line 24-SI-2001 91097SH0019 Shutdown Heat Exchanger Piping B-6 24-in. Safety Injection Line 24-SI-2002 91097SH0020 Shutdown Heat Exchanger Piping B-7 18-in. Safety Injection Line 18-SI-2003 91097SH0019 Shutdown Heat Exchanger Piping B-7 18-in. Safety Injection Line 18-SI-2003 91206 From Refueling Water Tank No. 21 B-7.
18-in. Safety Injection Line 18-SI-2003 91388 From Refueling Water Tank B-7 18-in. Safety Injection Line 18-SI-2003 91389 Refueling Water System B-8 18-in. Safety Injection Line 18-SI-2004 91097SH0020 Shutdown Heat Exchanger Piping Safety Injection Piping From The Refueling B-8 18-in. Safety Injection Line 18-SI-2004 91360 Water-Tank B-9 14-in. Safety Injection Line 14-SI-2001 91097SH0020 Shutdown Heat Exchanger Piping B-10 14-in. Safety Injection Line 14-SI-2002, 91097SH0019 Shutdown Heat Exchanger Piping B-10 14-in. Safety Injection Line 14-SI-2002 91577 Shutdown Heat Exchanger Piping B-11 14-in. Safety Injection Line 14-SI-2004 91577 Shutdown Heat Exchanger Piping B-12 14-in. Safety Injection Line 14-SI-2010 91097SH0019 Shutdown Heat Exchanger Piping B-13 14-in. Safety Injection Line 14-SI-2011 91097SH0020 Shutdown Heat Exchanger Piping
,L B-14 12-in. Safety Injection Line 12-SI-2004A 91097SH0028 Shutdown Heat Exchanger Piping Revision 0 113 Prepared for Constellation Energy by
Constellation Energy-
'Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI IS1 Drawing No.
Design Reference Drawing Drawing No.
Description Design Drawing Description B-14 12-in. Safety Injection Line 12-SI-2004A 91097SH0029 Shutdown Heat Exchanger Piping Comp. Cooling Pump Room Piping Safety B-14 12-in. Safety Injection Line 12-SI-2004A 91378 Injection System B-14 12-in. Safety Injection Line 12-SI-2004A 91383 L.P. Safety Injection B-15 12-in. Safety Injection Line 12-SI-2013 91097SH0028 Shutdown Heat Exchanger Piping B-16 12-in. Safety Injection Line 12-SI-1216 (2016) 91097SH0006 Shutdown Heat Exchanger Piping Component Cooling Pump Room Piping B-16 12-in. Safety Injection Line 12-SI-1216 (2016) 91110 Shutdown Cooling System B-17 10-in. Safety Injection Line 10-SI-2002 91097SH0029 Shutdown Heat Exchanger Piping B-18 10-in. Safety Injection Line 10-SI-2003 91097SH0021 Shutdown Heat Exchanger Piping B-18 10-in. Safety Injection Line 10-SI-2003 91097SH0029 Shutdown Heat Exchanger Piping B-19 10-in. Safety Injection Line 10-SI-2006 91097SH0019 Shutdown Heat Exchanger Piping B-20 10-in. Safety Injection Line 10-SI-2007 91097SH0020 Shutdown Heat Exchanger Piping B-21 8-in. Safety Injection Line 8-SI-2007 91097SH0023 Shutdown Heat Exchanger Piping B-22 8-in. Safety Injection Line 8-SI-2008 91097SH0023 Shutdown Heat Exchanger Piping Comp. Cooling Pump Room Piping Shutdown B-23 8-in. Safety Injection Line 8-SI-2018 91381 Cooling System B-24 8-in. Safety Injection Line 8-SI-2020 91195 Safety Injection Pump Discharge Revision 0 114 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section Xl Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index iS ISIDg Design Design Reference Drawing Drawing No.
Description Design Drawing No.
Description B-24 8-in. Safety Injection Line 8-SI-2020 91383 L.P. Safety Injection B-25 8-in. Safety Injection Line 8-SI-2021 91383 L.P. Safety Injection B-26 6-in. Safety Injection Line 6-SI-2001A 91097SH0024 Shutdown Heat Exchanger Piping B-26 6-in. Safety Injection Line 6-SI-2001A 91370 H.P. Safety Injection B-27 6-in. Safety Injection Line 6-SI-2001B 91097SH0023 Shutdown Heat Exchanger Piping B-28 6-in. Safety Injection Line 6-SI-2003A 91097SH0025 Shutdown Heat Exchanger Piping B-29 6-in. Safety Injection Line 6-SI-2004A 91097SH0022 Shutdown Heat Exchanger Piping B-29 6-in. Safety Injection Line 6-SI-2004A 91384 H.P. Safety Injection B-30 6-in. Safety Injection Line 6-SI-2005 91382 L.P. Safety Injection System B-30 6-in. Safety Injection Line 6-SI-2005 91383 L.P. Safety Injection System B-31 6-in. Safety Injection Line 6-SI-2006 91097SH0023 Shutdown Heat Exchanger Piping B-32 6-in. Safety Injection Line 6-SI-2006A 91195 Safety Injection Pump Discharge B-33 6-in. Safety Injection Line 6-SI-2007 91195 Safety Injection Pump Discharge B-34 6-in. Safety Injection Line 6-SI-2008 91383 L.P. Safety Injection System B-34 6-in. Safety Injection Line 6-SI-2008 91390 L.P. Safety Injection System
.Revision 0 115 Prepared for Constellation Energy by
Constellation Energy
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISIg Design Design Reference Drawing Drawing No.
Description Design Drawing No.
Description B-35 6-in. Safety Injection Line 6-SI-2009 91097SH0024 Shutdown Heat Exchanger Piping B-36 6-in. Safety Injection Line 6-SI-2009A 91097SH0025 Shutdown Heat Exchanger Piping Comp. Cooling Pump Room Piping Shutdown B-37 6-in. Safety Injection Line 6-SI-2017 91381 Cooling Sys.
Salt Water Return From ECCS Pump Room B-38 6-in. Safety Injection Line 6-SI-2201 91282 Cooler No. 11 Safety Injection Piping System (Tank No. 21B B-38 6-in. Safety Injection Line 6-SI-2201 91298SH0002 Leg 21 B)
B-38 6-in. Safety Injection Line 6-SI-2201 91382 L.P. Safety Injection System Safety Injection Piping System (Tank No.21A B-39 6-in. Safety Injection Line 6-SI-2202 91298SH0001 Leg 21A)
B-39 6-in. Safety Injection Line 6-SI-2202 91390 L.P. Safety Injection System B-40 6-in. Safety Injection Line 6-SI-2203 91195 Safety Injection Pump Discharge Safety Injection Piping System (Tank No. 22B B-40 6-in. Safety Injection Line 6-SI-2203 91298SH0004 Leg 22B)
B-41 6-in. Safety Injection Line 6-SI-2204 91195 Safety Injection Pump Discharge Safety Injection Piping System (Tank No. 22A B-41 6-in. Safety Injection Line 6-SI-2204 91298SH0003 Leg 22A)
B-42 6-in. Safety Injection Line 6-SI-2004 91097SH0023 Shutdown Heat Exchanger Piping B-42 6-in. Safety Injection Line 6-SI-2004 91380 Shutdown Heat Exchanger Piping B-43 4-in. Safety Injection Line 4-SI-2005 91097SH0022 Shutdown Heat Exchanger Piping Revision 0 116 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2
_Component and Isometric Drawings Index IS 1SI Drawing Design Drawing No.
Design Reference Drawing Drawing No.
Description DesignDrawing Description B-43 4-in. Safety Injection Line 4-SI-2005 91097SH0024 Shutdown Heat Exchanger Piping B-44 4-in. Safety Injection Line 4-SI-2005A 91097SH0025 Shutdown Heat Exchanger Piping Safety Injection Recirculation To Refueling B-45 4-in. Safety Injection Line 4-SI-2006 91385 Water Tank Unit 2 B-46 4-in. Safety Injection Line 4-SI-2007 91370 H.P. Safety Injection B-47 4-in. Safety Injection Line 4-SI-2008 91193 Safety Injection System B-47 4-in. Safety Injection Line 4-SI-2008 91370 H.P. Safety Injection B-48 4-in. Safety Injection Line 4-SI-2009 91384 H.P. Safety Injection B-49 4-in. Safety Injection Line 4-SI-2010 91194 Safety Injection System B-49 4-in. Safety Injection Line 4-SI-2010 91384 H.P. Safety Injection B-50 3-in. Safety Injection Line 3-SI-2001 91097SH0024 Shutdown Heat Exchanger Piping B-51 3-in. Safety Injection Line 3-SI-2002 91097SH0024 Shutdown Heat Exchanger Piping B-52 3-in. Safety Injection Line 3-SI-2003 91097SH0022 Shutdown Heat Exchanger Piping B-53 3-in. Safety Injection Line 3-SI-2014 91382 L.P. Safety Injection System B-54 3-in. Safety Injection Line 3-SI-2015 91390 L.P. Safety Injection System Safety Injection CC-6, CC-13 & DC-1 (Nuclear B-55 3-in. Safety Injection Line 3-SI-2016 FSK-MP-3175 Class II)
J Revision 0 117 Prepared for Constellation Energy by
Constellation Energy-
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISI Drawing Design Drawing No.
Design Reference Drawing Drawing No.
Description DesignDrawing Description B-56 3-in. Safety Injection Line 3-SI-2017 FSK-MP-3174 Safety Injection CC-6, CC-13 & DC-1 H.P. Safety Injection System CC-6 & CC-13 B-57 2-in. Safety Injection Line 2-SI-2002 FSK-MP-3806 (Nuclear Class II)
Charging Pump Discharge To HPSI Header B-58 2-in. Safety Injection Line 2-SI-2003 FSK-MP-3540H CC-7, CC-6 (Nuclear CL II)
Charging Pump Discharge To HPSI Header B-58 2-in. Safety Injection Line 2-SI-2003 FSK-MP-3541 H CC-7, CC-6 (Nuclear CL II)
H.P. Safety Injection System CC-6 (Nuclear, B-59 2-in. Safety Injection Line 2-SI-2004 FSK-MP-3809 Class II)
Safety Injection CC-6, CC-13 & DC-1 (Nuclear B-60 2-in. Safety Injection Line 2-SI-2005 FSK-MP-3175 Class II)
B-61 2-in. Safety Injection Line 2-SI-2006 FSK-MP-3174 Safety Injection CC-6, CC-13 & DC-1 B-62 2-in. Safety Injection Line 2-SI-2008 FSK-MP-3174 Safety Injection CC-6, CC-13 & DC-1 Charging Pump Discharge To HPSI Header B-63 2-in. Safety Injection Line 2-SI-2010 FSK-MP-3540H CC-7, CC-6 (Nuclear CL II)
H.P. Safety Injection System CC-13 (Nuclear B-64 2-in. Safety Injection Line 2-SI-2012 FSK-MP-3810 Class II)
H.P. Safety Injection System CC-6 & CC-13 B-65 2-in. Safety Injection Line 2-SI-2015 FSK-MP-3806 (Nuclear Class II)
H.P. Safety Injection System CC-13 (Nuclear B-65 2-in. Safety Injection Line 2-SI-2015 FSK-MP-3807 Class II)
Safety Injection CC-6, CC-13 & DC-1 (Nuclear B-66 2-in. Safety Injection Line 2-SI-2017 FSK-MP-3175 Class II)
B-67 2-in. Safety Injection Line 2-SI-2018 91390 L.P. Safety Injection System H.P. Safety Injection System CC-13 (Nuclear B-67 2-in. Safety Injection Line 2-SI-2018 FSK-MP-3810 Class II)
Revision 0 118 Prepared for Constellation Energy by
Constellation Energy-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISIg Design Design Reference Drawing Drawing No.
Description Design Drawing No.
Description B-68 2-in. Safety Injection Line 2-SI-2019 91382 L.P. Safety Injection System H.P. Safety Injection System CC-6 & CC-13 B-68 2-in. Safety Injection Line 2-SI-2019 FSK-MP-3806 (Nuclear Class II)
B-69 2-in. Safety Injection Line 2-SI-2020 FSK-MP-3174 Safety Injection CC-6, CC-13 & DC-1 Safety Injection CC-6, CC-13 & DC-1 (Nuclear B-70 2-in. Safety Injection Line 2-SI-2021 FSK-MP-3175 Class II)
H.P. Safety Injection System DC-1 (Nuclear B-71 2-in. Safety Injection Line 2-SI-2035 FSK-MP-3808 Class II)
H.P. Safety Injection System DC-1 (Nuclear B-72 2-in. Safety Injection Line 2-SI-2036 FSK-MP-3811-H Class II)
Safety Injection CC-6, CC-13 & DC-1 (Nuclear B-73 2-in. Safety Injection Line 2-SI-2037 FSK-MP-3175 Class II)
B-74 2-in. Safety Injection Line 2-SI-2038 FSK-MP-3174 Safety Injection CC-6, CC-13 & DC-1 B-75 2-in. Safety Injection Line 2-SI-2041 FSK-MP-117-H-2 Containment Spray & Safety Injection Piping B-76 2-in. Safety Injection Line 2-SI-2042 FSK-MP-1 17-H-2 Containment Spray & Safety Injection Piping Containment Spray & Safety Injection Piping B-77 2-in. Safety Injection Line 2-SI-2043 FSK-MP-120 DC-2(Nuclear CL II)
Containment Spray & Safety Injection Piping B-78 2-in. Safety Injection Line 2-SI-2044 FSK-MP-120-HA DC-2 (Nuclear CL II)
B-79 2-in. Safety Injection Line 2-SI-2045 FSK-MP-117-H-2 Containment Spray & Safety Injection Piping Containment Spray & Safety Injection Piping B-79 2-in. Safety Injection Line 2-SI-2045 FSK-MP-118-H GC-7 & DC-2 B-80 10-in. Containment Spray Line 10-CS-2002 91097SH0028 Shutdown Heat Exchanger Piping Revision 0 119 Prepared for Constellation Energy by
Constellation Energy"
-Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2
~Component and Isometric Drawinas Index ISI ISI Drawing Design No.
Design Reference Drawing Drawing No.
Description Drawing Description B-81 10-in. Containment Spray Line 10-CS-2004 91097SH0027 Shutdown Heat Exchanger Piping B-82 10-in. Containment Spray Line 1O-CS-2018 91097SH0028 Shutdown Heat Exchanger Piping B-83 8-in. Containment Spray Line 8-CS-2003 91097SH0027 Shutdown Heat Exchanger Piping B-84 8-in. Containment Spray Line 8-CS-2004 91097SH0028 Shutdown Heat Exchanger Piping B-85 (Fig. B-85A) 8-in. Containment SprayLine 8-CS-2005 91180 Containment Spray B-85 (Fig. B-85A) 8-in. Containment Spray Line 8-CS-2005 91380 Shutdown Heat Exchanger Piping B-85 (Fig. B-85A) 8-in. Containment Spray Line 8-CS-2005 91386 Containment Spray B-86 (Fig. B-85B) 8-in. Containment Spray Line 8-CS-2005 91180 Containment Spray B-87 (Fig. B-86A) 8-in. Containment Spray Line 8-CS-2006 91188 Containment Spray B-87 (Fig. B-86A) 8-in. Containment Spray Line 8-CS-2006 91373 Containment Spray Discharge B-87 (Fig. B-86A) 8-in. Containment Spray Line 8-CS-2006 91379 Shutdown Heat Exchanger Piping B-88 Spray System Header No. 1 Inside (Fig. B-86B) 8-in. Containment Spray Line 8-CS-2006 91372 Containment B-88 (Fig. B-86B) 8-in. Containment Spray Line 8-CS-2006 91373 Containment Spray Discharge B-89 14-in. Shutdown Cooling Line 14-SC-2003 and 12-in.
(Fig. B-87)
Shutdown Cooling Line 12-SC-2003 91358 Shutdown Cooling B-89 14-in. Shutdown Cooling Line 14-SC-2003 and 12-in.
(Fig. B-87)
Shutdown Cooling Line 12-SC-2003 91320 Reactor Shutdown Cooling Revision 0 120 Prepared for Constellation Energy by
Constellation Energy"
.Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit2
_Component and Isometric Drawings Index ISI ISI Drawing DaDesign Reference Drawing Drawing No.
Description Design Drawing No.
Description B-89 14-in. Shutdown Cooling Line 14-SC-2003 and 12-in.
(Fig. B-8 7)
Shutdown Cooling Line 12-SC-2003 91577 Shutdown Heat Exchanger Piping B-90 (Fig. B-88) 14-in. Shutdown Cooling Line 14-SC-2004A 91577 Shutdown Heat Exchanger Piping B-91 (Fig. B-89) 12-in. Shutdown Cooling Line 12-SC-2008 91379 Shutdown Heat Exchanger Piping B-91 (Fig. B-89) 12-in. Shutdown Cooling Line 12-SC-2008 91380 Shutdown Heat Exchanger Piping B-92 (Fig. B-90) 12-in. Shutdown Cooling Line 12-SC-2013 91097SH0028 Shutdown Heat Exchanger Piping B-93 Comp. Cooling Pump Room Piping Safety (Fig. B-91) 12-in. Shutdown Cooling Line 12-SC-2015 91378 Injection System B-93 (Fig. B-91) 12-in. Shutdown Cooling Line 12-SC-2013 91380 Shutdown Heat Exchanger Piping B-94 (Fig. B92) 10-in. Shutdown Cooling Line 10-SC-2014 91097SH0027 Shutdown Heat Exchanger Piping B -94 (Fig. B-92) 10-in. Shutdown Cooling Line 10-SC-2014 91097SH0028 Shutdown Heat Exchanger Piping B-95 (Fig. B-93) 8-in. Shutdown Cooling Line 8-SC-2006 91577 Shutdown Heat Exchanger Piping B-96 (Fig. B-94) 8-in. Shutdown Cooling Line 8-SC-2007 91379 Shutdown Heat Exchanger Piping B-97 (Fig. B-95) 8-in. Shutdown Cooling Line 8-SC-2009 91380 Shutdown Heat Exchanger Piping B-98 (Fig. B-96) 8-in. Shutdown Cooling Line 8-SC-2010 91380 Shutdown Heat Exchanger Piping B-99 (Fig. B-97) 36-in. Main Steam Line 36-MS-2001 62345SH0001 Main Steam System Isometric Revision 0 121 Prepared for Constellation Energy by
Constellation Energ" Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units I & 2 ASME Section XI Inservice Inspection Program CCNPP.Unit 2
_in Component and Isometric Drawings Index S.,1 Drawing Drsign.Drwingo Design Reference-Drawing Drawing No.
Description,
__Description B-99 (Fig. B-97) 36-in. Main Steam Line 36-MS-2001 91590 Main Steam Encapsulation Unit #2 B-100 (Fig. B-98) 36-in. Main Steam Line 36-MS-2002 62345SH0001 Main Steam System Isometric B-100 (Fig. B-98) 36-in. Main Steam Line 36-MS-2002 91590 Main Steam Encapsulation Unit #2 B-101 (Fig. B-99) 34-in. Main Steam Line 34-MS-2001 62345SH0001 Main Steam System Isometric B-101 (Fig. B-99) 34-in. Main Steam Line 34-MS-2001 91590 Main Steam Encapsulation Unit #2 B-102 (Fig. B-100) 34-in. Main Steam Line 34-MS-2002 B-103 (Fig. B-101) 34-in. Main Steam Line 34-MS-2004 B-104 (Fig. B-102) 34-in. Main Steam Line 34-MS-2005 B-1 05-(Fig. B-103) 6-in. Main Steam Line 6-MS-2007 B-106 (Fig. B-104) 6-in. Main Steam Line 6-MS-2008 B-107
,(Fig. B-105) 6-in. Main Steam Line 6-MS-2037 B-108 (Fig. B-106) 6-in. Main Steam Line 6-MS-2038 B-109 (Fig. B-107) 16-in. Feedwater Line 16-FW-2001 B-l10 (Fig. B-108) 16-in. Feedwater Line 16-FW-2002 Revision 0 122 Prepared for Constellation Energy by
Constellation Energyf Calvert Cliffs Nuclear Power Plant Calvert Cliffs Units 1 & 2 ASME Section XI Inservice Inspection Program CCNPP Unit 2 Component and Isometric Drawings Index ISI ISI DrawingDrawing No.
Design Reference Drawing Drawing No.
Description Description B-111 (Fig. B-109) 16-in. Feedwater Line 16-FW-2018 B-112 (Fig. B-110) 16-in. Feedwater Line 16-FW-2019 B-112 4-in. Auxiliary Feedwater 4-AF-2014 91369SH001 Auxiliary Feedwater Inside Containment B-113 (Fi.
B-11)
Typical Valve Body-to-Flange Weld B-1 14 4-in. Auxiliary Feedwater 4-AF-2015 91369SH002 Auxiliary Feedwater Inside Containment B-115 4-in. Auxiliary Feedwater 4-AF-2014 91368 Auxiliary Feedwater B-1 16 4-in. Auxiliary Feedwater 4-AF-2003 91368 Auxiliary Feedwater B-1 17 4-in. Auxiliary Feedwater 4-AF-2011 62577 Piping Isometric Auxiliary Feedwater System B-118 4-in. Auxiliary Feedwater 4-AF-2010 62577 Piping Isometric Auxiliary Feedwater System Revision 0 123 Prepared for Constellation Energy by
ATTACHMENT (2)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-01 Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ATTACHMENT (2)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-01 I.
COMPONENTS Code Class:
1
References:
American Society of Mechanical Engineers (ASME)Section XI, 2004 Edition Code Case N-747 Technical Basis for Reactor Vessel Head-to-Flange Weld Examinations as Prescribed in ASME Case N-747, dated November 8, 2005 Examination Category:
B-A Item Number:
B 1.40
==
Description:==
Alternative Requirements for Examination of the Reactor Vessel Head-to-Flange Weld Unit/Inspection Calvert Cliffs Unit I & 2 - Fourth 10-Year Interval Applicability:
II.
CODE REQUIREMENTS Table IWB-2500-1, Category B-A, Item Number B1.40 requires a volumetric and a surface examination to be performed once per interval on the reactor vessel head-to-flange weld.
The examination includes essentially 100% of the weld length.
III. PROPOSED ALTERNATIVE Pursuant to 10 CFR 50.55a(a)(3)(i), Calvert Cliffs requests authorization to utilize the alternative requirements in ASME Code Case N-747 in lieu of the requirements of Table IWB-2500-1, Examination Category B-A, Item Number B 1.40.
IV. BASIS FOR PROPOSED ALTERNATIVE The alternative examination requirements in Code Case N-747 provide an option to reduce undue burden and worker radiation exposure, while maintaining plant safety.
Specifically, it provides alternative requirements for the reactor vessel head-to-flange weld to be inspected by surface examination once each ten-year inspection interval, using the current surface examination area shown in Figure IWB 2500-5. This alternative requirement may only be implemented after the weld has received at least one inservice volumetric examination, which may be performed as part of the preservice inspection, with no service-induced flaws having been identified.
The basis for elimination of the concurrent surface and volumetric examination requirement for the head-to-flange weld is rooted in nearly 40 years of service experience for this weld. The technical bases for the alternative criteria of Code Case N-747 are provided in the associated White Paper for the action entitled, "Technical Basis for Reactor Vessel Head-to-Flange Weld Examinations as Prescribed in ASME Case N-747," dated November 8, 2005. This White Paper evaluated a number of factors including component geometry, associated stresses, fracture toughness, fatigue considerations, corrosion and industry experience with examinations. Based on this evaluation, the White Paper concluded that only a surface examination should be required for the reactor vessel head-to-flange weld provided no defects had been detected during any preservice or inservice examinations.
In addition, the Examination Category B-P pressure tests and visual examinations normally conducted in conjunction with refueling outages will also continue.
I
ATTACHMENT (2)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-01 The reactor vessel head-to-flange weld is not a dissimilar metal or Alloy 600 weld, and is a full penetration design. In addition, there have been no defects detected on this weld during pre-service or inservice examinations.
It is therefore concluded that the concurrent volumetric and surface examination requirement may be eliminated for the reactor vessel head-to-flange weld, and that the outer surface examination discussed above provides an acceptable level of quality and safety.
V.
CONCLUSION 10 CFR 50.55a(a)(3) states:
"Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate. The applicant shall demonstrate that:
(i)
The proposed alternatives would provide an acceptable level of quality and safety, or (ii)
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
As discussed in Section IV above, the proposed alternative requirements in Code Case N-747 provide an acceptable level of quality and safety to the requirements in ASME Section XI, 2004 Edition, Table IWB-2500-1, Category B-A, Item Number B13.40. Therefore, Calvert Cliffs requests authorization to perform the requested alternative to the ASME Code requirement pursuant to 10 CFR 50.55a(a)(3)(i).
2L
ATTACHMENT (3)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-02 Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ATTACHMENT (3)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-02 I.
COMPONENTS Code Class:
Not Applicable
References:
American Society of Mechanical Engineers (ASME)Section XI, 2004 Edition Code Case N-753 Examination Category:
Not Applicable Item Numbers:
Not Applicable
==
Description:==
Alternative Requirements to the Visual Acuity Demonstration Requirements of IWA-2321(a)
Unit/Inspection Calvert Cliffs Unit 1 & 2 - Fourth 10-Year Interval Applicability:
II.
CODE REQUIREMENTS Paragraph IWA-2321 (a) of ASME Code Section XI, requires that nondestructive examination (NDE) personnel be administered the following vision tests annually: "Personnel shall demonstrate natural or corrected near-distance acuity of 20/25 or greater Snellen fraction, with at least one eye, by reading words or identifying characters on a near-distance test chart, such as a Jaeger chart, that meets the requirements of IWA-2322. Equivalent measures of near-distance acuity may be used. In addition, personnel performing VT-2 or VT-3 visual examinations shall demonstrate natural or corrected far-distance acuity of 20/30 or greater Snellen fraction or equivalent with at least one eye."
III. PROPOSED ALTERNATIVE Pursuant to 10 CFR 50.55a(a)(3)(i), Calvert Cliffs requests authorization to utilize the alternative requirements in ASME Code Case N-753 in lieu of the requirements of IWA-232 I(a).
IV. BASIS FOR PROPOSED ALTERNATIVE Code Case N-753 provides an alternative to the visual acuity demonstration requirements of IWA-2321(a) that will allow the testing to be administered and documented by an Optometrist, Ophthalmologist, or other health care professional who administers vision tests.
The visual acuity testing for NDE personnel performing ASME Section XI examinations is required to be administered annually. In addition to this vision testing, which is typically administered by utility personnel, many NDE personnel also have annual visual acuity testing in conjunction with routine eye examinations administered by an Optometrist, an Ophthalmologist, or other health care professional who administers vision tests.
Optometrists, Ophthalmologists, and other health care professionals who administer vision tests are typically educated and experienced in the proper techniques for vision testing, such as the Snellen fraction or Jaeger chart methods required by Section XI. This training and expertise provides a sound level of confidence that the visual acuity testing administered will be a reliable indicator that the tested NDE personnel can satisfactorily perform Section XI non-destructive examinations.
The testing performed by Optometrists, Ophthalmologists, and other health care professionals who administer vision tests will satisfy IWA-2321(a) requirements, including documentation which I
ATTACHMENT (3)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-02 details the tests performed, compliance with IWA-2321(a) criteria and the date the testing was administered.
The use of Code Case N-753 alternative requirements allows the flexibility for utilities to accept visual acuity testing performed by outside health care professionals in lieu of the visual acuity testing performed by in-house personnel. In many instances this flexibility will eliminate duplicate testing and thus provide a reduction in the costs and manpower associated with qualifying NDE personnel.
Because Code Case N-753 does not change the qualification criteria in IWA-2321(a), the implementation of the included alternative requirements does not affect the level of quality or safety provided by NDE personnel.
V.
CONCLUSION 10 CFR 50.55a(a)(3) states:
"Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate. The applicant shall demonstrate that:
(i)
The proposed alternatives would provide an acceptable level of quality and safety, or (ii)
Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
As discussed in Section IV above, the proposed alternative requirements in Code Case N-753 provide an acceptable level of quality and safety to the requirements in ASME Section XI, 2004 Edition, IWA-2321(a).
Therefore, Calvert Cliffs requests authorization to perform the requested alternative to the Code requirement pursuant to 10 CFR 50.55a(a)(3)(i).
2
ATTACHMENT (4)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-03 Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ATTACHMENT (4)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-03 I.
COMPONENTS Code Class:
2
References:
American Society of Mechanical Engineers (ASME)Section XI, 2004 Edition Section III Interpretation ofNB-3213.2 Examination Category:
C-A Item Number:
C1.10
==
Description:==
Shell Circumferential Welds Unit/Inspection Calvert Cliffs Unit 1 & 2 - Fourth 10-Year Interval Applicability:
II. CODE REQUIREMENTS Table IWC-2500-1 of ASME Code Section XI, Examination Category C-A, Item Numbers C1.10, requires in part that shell circumferential welds in pressure vessels be volumetrically examined when the welds are located at a gross structural discontinuity as defined by ASME Section III, Paragraph NB-3213.2. Examples are junctions between shells of different thicknesses, cylindrical shell-to-conical shell junctions, shell (or head)-to-flange welds, and head-to-shell welds.
III. PROPOSED ALTERNATIVE Pursuant to 10 CFR 50.55a(a)(3)(i), Calvert Cliffs proposes to continue to not classify the steam generator closure girth weld as a structural discontinuity and therefore not subject to volumetric examination.
IV. BASIS FOR PROPOSED ALTERNATIVE Calvert Cliffs submitted this request during the Third Ten-Year Inservice Inspection (ISI) interval for Units 1 and 2 in Reference (1).
The Nuclear Regulatory Commission (NRC) approved the alternative in Reference (2).
During the Third Ten-Year ISI interval the Calvert Cliffs Steam Generator Replacement Project
'replaced the steam generator lower assembly section containing the steam generator tubes and completely refurbished the original steam drum in accordance with ASME Boiler and Pressure Vessel Code Section III, "Rules for Construction of Nuclear Facility Components" 1989 Edition no Addenda and ASME Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" 1998 Edition no Addenda. Both sections were joined by the closure girth weld. The secondary side of the steam generator (both the original Combustion Engineering and the replacement Babcock & Wilcox Canada sections) is classified per the requirements of the ASME Section XI, Class 2 for the purpose of ISI but was constructed in accordance with ASME Section 111, Class I requirements.
As such, a 'stress and fatigue analysis of the secondary side has been performed which determined the predicted maximum stress intensity ranges and cumulative usage factors at specific junctions throughout the vessel shell. The junctions evaluated included the closure girth weld and other shell circumferential welds currently categorized as ISI welds.
In lieu of categorizing the closure girth weld as an ISI weld solely due to the weld being classified by definition as a gross structural discontinuity (since the weld is a junction between shells of different thicknesses), Calvert Cliffs proposed to utilize the stress analysis to show that susceptibility of this weld to fatigue cracking is significantly less than the steam generator welds currently in the ISI.
I
ATTACHMENT (4)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-03 program. Therefore adding the closure girth weld to the ISI program for the replacement steam generators provides no added value in monitoring and maintaining the structural integrity of the vessel.
In support of this effort, we have reviewed the applicable sections of various editions of ASME Section III, Section VIII, and Section XI Codes to determine the basis for the current definition of a gross structural discontinuity.
In addition, a Code Interpretation from the ASME Section III Committee clarifying the definition of a gross structural discontinuity (Enclosures 2 and 3) was received. As documented in the following section, based on this Code Interpretation, the low stress intensity and usage factor values determined by the stress and fatigue analysis do not support classifying the closure girth weld as a gross structural discontinuity.
Calvert Cliffs' ISI program for the secondary side of the steam generators currently includes the following circumferential welds (Enclosure 1):
- Head Circumferential Weld
" Upper Steam Drum Shell-To-Transition Cone Weld
" Tubesheet-to-Shell Weld The head circumferential weld and the upper steam drum shell-to-transition cone weld are welds on the original steam drum section, which was re-installed. These two circumferential welds have been subjected to three ten-year ISI interval examinations.
The tubesheet-to-shell weld is part of the replacement lower assembly and therefore a new weld in the ISI program.
The stress and fatigue analysis performed for the replacement steam generators evaluated the entire vessel for a design life of 40 years taking into account the operating history of the steam drum section prior to replacement. A summary of the stress analysis is tabulated below:
Range of Stress Allowable Stress Fatigue Fatigue Junction Intensity (ksi)
Intensity (ksi)
Usage Usage Factor (PL +Pb+Q) 3Sm 1.5Sm Factor Limit Head Circumferential Weld 20.3 80.1 40.1 0.04 1.0 Tubesheet-to-Shell Weld 71.6 90.0 45.0 0.03 1.0 Upper Steam Drum-to-Transition Cone Weld 36.0 80.1 40.1 0.02 1.0 Replacement Steam Generator Closure Girth Weld 26.0 80.1 40.1 0.002 1.0 The data tabulated above shows that the susceptibility of the closure girth weld to fatigue cracking is very low in comparison to the other three circumferential welds listed that are currently in the ISI program. Of particular note is the comparison between the upper steam drum shell-to-transition cone weld and the closure girth weld. Per ASME Section XI, Table IWC-2500-1, the upper steam drum shell-to-transition cone weld is also an ISI weld solely due to the weld being classified as a gross structural discontinuity since this weld is a cylindrical shell-to-conical shell junction. The upper steam drum shell-to-transition cone weld has both a higher stress intensity range and fatigue usage factor than the closure girth weld. The upper steam drum shell-to-transition cone weld is part of the original steam drum and has undergone two ten-year ISI examinations with no flaws detected. This weld will remain an ISI weld. Based on the stress analysis performed for the replacement steam generator, the probability of the upper steam drum shell-to-transition cone weld developing a fatigue crack is significantly higher than the closure girth weld. Therefore, subjecting the closure girth to 2
ATTACHMENT (4)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-03 future volumetric inspections will not provide any added value in monitoring the structural integrity of the steam generators.
V.
CONCLUSION 10 CFR 50.55a(a)(3) states:
"Proposed alternatives to the requirements of paragraphs (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation or Director, Office of New Reactors, as appropriate. The applicant shall demonstrate that:
(i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety."
As discussed in Section IV above, the proposed alternative requirements provide an acceptable level of quality and safety to the requirements in ASME Section XI, 2004 Edition, Examination Category C-A, Item Number C1.10. Therefore, Calvert Cliffs requests authorization to perform the requested alternative to the Code requirement pursuant to 10 CFR 50.5 5a(a)(3)(i).
REFERENCES (1)
Letter from Mr. P. E. Katz (CCNPP) to Document Control Desk (NRC), dated October 22, 2002, ASME Section XI Relief Request to Use an Alternative to the Inservice Inspection Requirement for Replacement Steam Generator Girth Welds (2)
Letter from Mr. R. J. Laufer (NRC) to Mr. P. E. Katz (CCNPP), dated March 6, 2003, Alternative to Inservice Inspection Requirement for Replacement Steam Generator Girth Welds ENCLOSURES (1)
Replacement Steam Generator Weld Locations (2)
CCNPP Code Interpretation Request Letter (3)
ASME Code Interpretation Letter 3
ENCLOSURE (1)
Replacement Steam Generator Weld Locations Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ENCLOSURE (1)
HEAD CIRCUMFERENTIAL WELD
&.-T rRNS IT 10 N fCISýOEIANTýE~bGEN ERATGR TUSESHEET-TO-SHELL WELD FIGURE 1.
REPLACE MENT STEAM GENERATOR WELD LOCATIONS
ENCLOSURE (2)
CCNPP Code Interpretation Request Letter Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ENCLOSURE (2)
Osatwrt Cliffs tHucIarn Po~wer PI&a1 I eu U&M~ klr;:5 rawirry Ctnslalla*',, Gonarallon Grv*Jp. LLO Lmsy. Mzvr~tlsnd 2DS5"7 Steam Generator Project October 8, 2002 SGP-PM.02-186 Secretary ASME Boiler and Pressure Vessel Committee ASME International Three Park Ave.
New York, NY 10016-5990
Subject:
Section Il1. Interpretation of NB-321 3.2 Dear Sic We are In the final stages of preparing for our upcoming steam generator replacement project and need to finalize the closure girth weld joint configuration for the two-piece vessel assembly In the field.Section XI requires some welds to be uttrasonicaliy examined, If it Is required that a circumferential weld at a tapered transition be ultrasonically examined, a different weld joint configuration Will be used tbarn if ultrasonic examlnation Is not required. To determine where examinations are to be performed,Section XI refers to Section IIi for the definition of gross structural discontinuity.
Section XI, Table IWC-2500-1, Category C-A requires examination of circumferential shell welds at gross structural discontinuities, as defined in Section III, NB-3213.2. Unfortunately, the definition for gross structural discontinuity In NB-3213.2 is unclear and Is inconsistent with the definitlon for gross structural discontinuity in B31.7, Section ViII, Division 2, 4-112(b.) and Section Viii, Division 3, KD-210(b). The original definition of gross structural discontinuity In the 1968 Edition of Section III is consistent*with the current definitions in Section Vill, Divisions 2 and 3. We feel the definitions in all three books should be essentially the same because the stress analysis methodologies In the three books are identical.
When the original Section III definitlon (or the Section ViII, Division 2 and 3 definqiton) for gross structural discontinuity Is used, it seems obvious that a structural discontinuity is considered
.gross' only when associated with stresses higher than the basic allowable primary stresses (primary membrane plus primary bending). Table NB-3217-1, and Fig. NB-3222-1 seem to support this conclusion. Further confirmation of this conclusion is the fact that F-102.2 in the B31.7 Code states that, 'The C factors in Equations (10), (11), and (12) are gross structural discontinuitles." In 831.7 and Section lit, Equation (10) addresses primary plus secondary stress Intensity range; Equation (11) addresses peak stress Intensity range; and Equation (12) addresses simplified elastic-plastic discontinuity analysis. These three equations all deal with stresses that are much higher than the allowable stress values for primary stress. From this faMt, It seems apparent that structural discontinuities are considered "gross' only when associated with stresses higher then the primary stress alloviable values.
I
ENCLOSURE (2)
NB-3361 is also related to the issue of the Code definition of gross structural discontinuity. The first sentence states, 'in general, a tapered transition section as shown in Fig. NBS-3361-1 which Is a type of gross structural discontinuity (NB-3213.2) shall be provided at joints of Categories A and B between sections that differ in thickness by more than one-fourth the thickness of the thinner section." We have calculated the stresses at the transition joint of our vessel and the sum of primary plus secondary plus peak is less than 1.5S.
Based on the above, our questions are as follows:
Question (1): is a structural discontinuity considered to be a 'gross structural discontinuity' when the membrane stress Intensity does not exceed 1.1 Sm and the membrane plus bending stress Intensity does not exceed 1.5 Sm?
Reply (I); No.
Question (2): Does the first sentence of NB-3361 classify a tapered transition in thickness as a
- gross structural discontinuity"?
Reply (2): No. The definition used for classifying whether a discontinuity is a 'gross structural discontinuityr is given in NS-32I3,2. The tapered transition referenced in NB-3361 is a type of structural discontinuity, but whether or not It is a "gross structural discontinuity" depends on the level of stress calculated at the joint.
We would appreciate an answer as soon as possible as we are now preparing our mock-up and welding processes. There is an urgent need for a quick response if at all possible.
Sincerely, J. R. Dalrymple Project Manager Steam Generator Project JRDIsmp cvc T. L Kewerth JOý0. Calle File e1 g]SGPrA02-185.doc 2
ENCLOSURE (3)
ASME Code Interpretation Letter Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
ENCLOSURE (3) codn~~ uSgrs ASME Intoamal~onamI USA Qetolxr 18, 2002 stellcailon Oiinlms tup,LO 1650 Cavcrt 0L3 Pnkwuy LmbyMaryland 20437 Sui~cct ASM! Sectmal1,N-flt 1te&~ac~a: our lelir (W4r Qqtcber 2, 2002 ASbM nlo 0:)4102-413 Our xz
~m~ndiu*
df Lbs pw~eet& It Y=
vir airY. Wn u ele a
ekn Fig NB-3m-I) Wbcas rb nmsraIteyde dnoe IS~(HIl0 ndW znr,*ne plus bnidhig stmktes I~aSy doe not exts 1.5 S. (Fig. MD-1221 -1)7 Rcply(Q No QIIWUfca P2; owet be frst Oak=Otere a36 oflWI 3leh a apre~d tostoi= In atitian asOPta Nw-..
tnier Reply (2)ý
'No. Te&fiosd!rrw~r ttcadsotn I
'r~
tcn bubifi dsattinity, but sletbcr or tat It is a Vmuriucatetn diaeozdtlty dopeodz on the level ofstews cvteihy ttthc ol 1-212-59148$13 AIM ~iptf tme VVI h NWAaft~ Cit VAp~ttaNWata
$kM Wh'ýflt 6uvfth Mbta tnk;* w" ltb rkg to 1 ktswraPjh ýd The American Socelty ni MnfthArdstFnlnaw
ATTACHMENT (5)
ASME CODE, SECTION XI RELIEF REQUEST -- ISI-04-04 Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
Application of ASME Code Case N-716 RISK-INFORMED / SAFETY-BASED INSERVICE INSPECTION PROGRAM PLAN CAL VERT CLIFFS NUCLEAR POWER PLANT UNITS 1 AND 2 REVISION 0 PREPARED BY:
Name Richard Fougerousse Signature Date 11/25/08 REVIEWED BY:
Jim Moody Name Signature p7w4,-+-
Date 11/25/08 Page 1
CONSTELLATION NUCLEAR CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS I AND 2 REQUEST FOR ALTERNATIVE ISI-04-04 Application of ASME Code Case N-716 RISK-INFORMED / SAFETY-BASED INSERVICE INSPECTION PROGRAM PLAN Table of Contents
- 1.
Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PRA Quality
- 2.
Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs
- 3.
Risk-Informed / Safety-Based ISI Process 3.1 Safety Significance Determination 3.2 Failure Potential Assessment 3.3 Element and NDE Selection 3.3.1 Additional Examinations 3.3.2 Program Relief Requests 3.4 Risk Impact Assessment 3.4.1 Quantitative Analysis 3.4.2 Defense-in-Depth
- 4.
Implementation and Monitoring Program
- 5.
Proposed ISI Program Plan Change
- 6.
References/Documentation ATTACHMENT A - CCNPP PRA Quality Review Page 2
CONSTELLATION NUCLEAR CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS 1 AND 2 REQUEST FOR ALTERNATIVE ISI-04-04
- 1.
INTRODUCTION Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (CCNPP) are currently in the third inservice inspection (ISI) interval as defined by the American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel Section XI Code for Inspection Program B.
CCNPP plans to implement a risk-informed / safety-based inservice inspection (RISB) program beginning with the first inspection period of the fourth ISI interval. The fourth interval will commence October 10, 2009 for CCNPP Units 1 and 2.
The ASME Section XI code of record for the thirdlSI interval at CCNPP is the 1998 Edition for Examination Category B-F, B-J, C-F-i, and C-F-2 Class 1 and 2 piping components. The ASME Section XI code of record for the fourth ISI interval at CCNPP will be the 2004 Edition for these welds.
The objective of this submittal is to request the use of the RISB process for the inservice inspection of Class 1 and 2 piping. The RISB process used in this submittal is based upon ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division 1, which is founded in large part on the RI-ISI process as described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, Revised Risk-Informed Inservice Inspection Evaluation Procedure.
1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," and Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping. Additional information is provided in Section 3.4.2 relative to defense-in-depth.
1.2 Probabilistic Safety Assessment (PSA) Quality The CCNPP PRA has been demonstrated to be adequate for this application. The history and development of the PRA is described in further detail in Attachment A. As described in Attachment A, the CCPRA Revision 0 internal -events model has been reviewed as part of the Combustion Engineering Owners Group (CEOG) Peer Review Process in November 2001. This review was also conducted against the draft ASME PRA Standard, Revision 14A.
This was accomplished by using the ASME Standard supporting requirements as sub-tier criteria.
The team was provided with a self-assessment of the CCPRA to the draft ASME PRA requirements. The team used this self-assessment to aid in gauging the technical adequacy of the CCPRA. All eleven technical elements were found to "Meet" requirements. The CCNPP PRA, including the Page 3
internal events flooding evaluation are being further upgraded to meet the documentation requirements of the ASME Standard and Regulatory Guide 1.200 and any future PRA modeling changes will be considered as part of the N716 living program review.
- 2.
PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAMS 2.1 ASME Section XI ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 currently contain requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components, except as amended by application of a risk-informed inservice inspection (RI-ISI) program based on ASME Code Case N-578 (Relief Request No. RR-RI-ISI-2) that was approved for use at CCNPP by the NRC on April 16, 2003..
The alternative RISB Program for piping is described in Code Case N-716. The RIS_B Program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section XI Code will be unaffected.
2.2 Augmented Programs The impact of the RISB application on the various plant augmented inspection programs listed below were considered.
This section documents only those plant augmented inspection programs that address common piping with the RISB application scope (e.g., Class 1 and 2 piping).
The original plant augmented inspection program for high-energy line breaks outside containment, previously implemented in accordance with NUREG-0800, Section 3.6.2 and Technical Requirements Manual 15.4.3.2 will be revised prior to the start of the fourth ISI interval in accordance with the risk-informed break exclusion region methodology (RI-BER) described in EPRI Report 1006937, Extension of EPRI Risk Informed IS! Methodology to Break Exclusion Region Programs. EPRI Report 1006937 was approved by the NRC in 2002.
" The plant augmented inspection programs previously implemented in response to NRC Bulletins 88-08, Thermal Stresses in Piping Connected. to Reactor Coolant Systems, and 88-11, Pressurizer Surge Line Thermal Stratification, were subsumed by the RI-ISI Program since the thermal fatigue concerns addressed by these bulletins were explicitly considered in the application of the RI-ISI process. Since the RI-ISI and RISB degradation mechanism criterion is identical, these plant augmented inspection programs are subsumed by the new RIS_B Program.
The plant augmented inspection program for flow accelerated corrosion (FAC) per GL 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning, is relied-upon to manage this damage mechanism but is not otherwise affected or changed by the RIS_B Program.
Page 4
A plant augmented inspection program is being implemented at CCNPP in response to MRP-139, Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines. The requirements of MRP-139 will be used for the inspection and management of PWSCC susceptible welds and will supplement the RISB Program selection process. The RIS_B Program will not be used to eliminate any MRP-139 requirements.
CCNPP has conducted an evaluation in accordance with MRP-146, Materials Reliability Program: Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines, and these results have been incorporated into the RIS_B Program.
- 3.
RISK-INFORMED / SAFETY-BASED ISI PROCESS The process used to develop the RIS_B Program conformed to the methodology described in Code Case N-716 and consisted of the following steps:
" Safety Significance Determination Failure Potential Assessment Element and NDE Selection Risk Impact Assessment Implementation Program Feedback Loop 3.1 Safety Significance Determination The systems assessed in the RIS_B Program are provided in Tables 3.1-1 and 3.1-2 for
- Units 1 and 2, respectively. The piping and instrumentation diagrams and additional plant information including the existing plant ISI Program were used to define the piping system boundaries.
Per Code Case N-716 requirements, piping welds are assigned safety-significance categories, which are used to determine the treatment requirements.
High safety-significant (HSS) welds are determined in accordance with the requirements below.
Low safety-significant (LSS) welds include all other Class 2, 3, or Non-Class welds.
(1)
Class 1 portions of the reactor coolant pressure boundary (RCPB), except as provided in 10 CFR 50.55a(c)(2)(i) and (c)(2)(ii);
(2)
Applicable portions of the shutdown cooling pressure boundary function. That is, Class 1 and 2 welds of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:
(a)
As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to Page 5
the containment penetration, whichever encompasses the larger number of welds; or (b)
Other systems or portions of systems from the RPV to the second isolation valve (i.e., farthest from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; (3)
That portion of the Class 2 feedwater system [> 4 inch nominal pipe size (NPS)] of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve; (4)
Piping within the break exclusion region (> NPS 4) for high-energy piping systems as defined by the Owner. This may include Class 3 or Non-Class piping; and (5)
Any piping segment whose contribution to CDF is greater than 1E-06 (and per NRC feedback on the Grand Gulf and DC Cook RISB pilot applications 1E-07 for LERF) based upon a plant-specific PSA of pressure boundary failures (e.g., pipe whip, jet impingement, spray, inventory losses). This may include Class 3 or Non-Class piping.
3.2 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant-specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in EPRI TR-112657 (i.e., the EPRI RI-ISI methodology),
with the exception of the deviation discussed below.
Table 3.2 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative.
A deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for CCNPP. Table 3-16 of EPRI TR-112657 contains criteria for assessing the potential for thermal stratification, cycling, and striping (TASCS).
Key attributes for horizontal or slightly sloped piping greater than NPS 1 include:
- 1.
The potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids; or
- 2.
The potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids; or
- 3.
The potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid; or.
- 4.
The potential exists for two phase (steam/water) flow; or
- 5.
The potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow; AND Page 6
AT >50'F, AND Richardson Number > 4 (this value predicts the potential buoyancy of a stratified flow)
These criteria, based on meeting a high cycle fatigue endurance limit with the actual AT assumed equal to the greatest potential AT for the transient, will identify locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology that would allow consideration of fatigue severity is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below.
Turbulent Penetration TASCS Turbulent penetration typically occurs in lines connected to piping containing hot flowing fluid.
In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ATs can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration.
For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will keep the line filled with hot water. If there is no potential for in-leakage towards the hot fluid source from the outboard end of the line, this will result in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore, TASCS is not considered for these configurations.
Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.
Low flow TASCS In some situations, the transient startup of a system (e.g., shutdown cooling suction piping) creates the potential for fluid stratification as flow is established.
In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.
Page 7
Valve leakage TASCS Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference.
However, since this is generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected.
Convection Heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected.
In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for considering cycle severity. The above criteria have previously been submitted by EPRI to the NRC for generic approval [letters dated February 28, 2001 and March 28, 2001, from P.J.
O'Regan (EPRI) to Dr. B. Sheron (USNRC), Extension of Risk-Informed Inservice Inspection Methodology]. The methodology used in the CCNPP RISB application for assessing TASCS potential conforms to these updated criteria. Final materials reliability program (MRP) guidance on the subject of TASCS has been incorporated into the CCNPP RISB application in accordance with MRP-146.
It should be noted that the NRC has granted approval for RI-ISI relief requests incorporating these TASCS criteria at several facilities, including Comanche Peak (NRC letter dated September 28, 2001) and South Texas Project (NRC letter dated March 5, 2002).
3.3 Element and NDE Selection Code Case N-716 and lessons learned from the Grand Gulf and DC Cook RISB pilot applications provide criteria for identifying the number and location of required examinations.
Ten percent of the HSS welds shall be selected for examination as follows:
(1)
Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements:
(a)
A minimum of 25% of the population identified as susceptible to each degradation mechanism and degradation mechanism combination shall be selected.
(b)
If the examinations selected above exceed 10% of the total number of HSS welds, the examinations may be reduced by prorating among each degradation mechanism and degradation mechanism combination, to the extent practical, such that at least 10% of the HSS population is inspected.
(c)
If the examinations selected above are not at least 10% of the HSS weld population, additional welds shall be selected so that the total number selected for examination is at least 10%.
Page 8
(2)
At least 10% of the RCPB weldsshall be selected.
(3)
For the RCPB, at least two-thirds of the examinations shall be located between the inside first isolation valve (IFIV) (i.e., isolation valve closest to the RPV) and the RPV.
(4)
A minimum of 10% of the welds in that portion of the RCPB that lies outside containment (OC) (e.g., portions of the main feedwater system in BWRs) shall be selected.
(5)
A minimum of 10% of the welds within the break exclusion region (BER) shall be selected.
Currently, there are one hundred thirty nine (Unit 1) and one hundred thirty two (Unit 2)
BER program welds at CCNPP. These will be examined in accordance with the RI-BER Program during the fourth ISI interval.
In contrast to a number of RI-ISI Program applications where the percentage of Class 1 piping locations selected for examination has fallen substantially below 10%, Code Case N-716 mandates that 10% be chosen. A brief summary is provided below, and the results of the selections are presented in Tables 3.3-1 and 3.3-2 for Units 1 and 2, respectively. Section 4 of EPRI TR-112657 was used as guidance in determining the examination requirements for these locations.
Unit Class 1 Welds(1 )
Class 2 Welds(2)
NNS Welds(3)
All Piping Welds (4)
Total SSelected Toal ced Total Selected Total Selected 1
475 56 1574 14 13 9
2062 79 2
443 51 1617 17 12 4
2072 72 Notes (1) Includes all Category B-F and B-J locations. All 475 Unit 1 Class 1 piping weld locations and 443 Unit 2 Class 1 piping weld locations are HSS.
(2) Includes all Category C-F-1 and C-F-2 locations. Of the 1574 Unit 1 Class 2 piping weld locations, 238 are HSS and the remaining 1336 are LSS. Of the 1617 Unit 2 Class 2 piping weld locations, 223 are HSS and the remaining 1394 are LSS.
(3) These 13 Unit 1 and 12 Unit 2 non-nuclear safety (NNS) piping weld locations are HSS.
(4) Regardless of safety significance, Class 1 and 2 in-scope piping components will continue to be pressure tested as required by the ASME Section XI Program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RISB Program.
3.3.1 Additional Examinations If the flaw is original construction or otherwise acceptable, Code rules do not require any additional inspections. Any unacceptable flaw will be evaluated per the requirements of ASME Code Section XI, IWB-3500 and/or IWB-3600.
As part of performing evaluation to IWB-3600, the degradation mechanism that is responsible for the flaw will be determined and accounted for in the evaluation.
The process for ordinary flaws is to perform the evaluation using ASME Section XI.
If the flaw meets the criteria, then it is noted and appropriate successive Page 9
examinations scheduled.
If the nature and type of the flaw is service-induced, then similar systems or trains will be examined. If the flaw is found unacceptable for continued operation, it will be repaired in accordance with IWA-4000 and/or applicable ASME Section XI Code Cases. The need for extensive root cause analysis beyond that required for IWB-3600 evaluation will be dependent on practical considerations (i.e., the practicality of performing additional NDE or removing the flaw for further evaluation during the outage). The NRC is involved in the process at several points. For preemptive weld overlays, a relief request in accordance with 10 CFR 50.55a(a)(3) is usually required for design and installation. Should a flaw be discovered during an examination, a notification in accordance with 10 CFR 50.72 or 10 CFR 50.73 may be required.
IWB-3600 requires the evaluation to be submitted to the NRC. Finally, the evaluation will be documented in the corrective action program and the Owner submittals required by Section XI.
The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions.
Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include HSS elements up to a number equivalent to the number of elements required to be inspected during the current outage.
If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined during the current outage.
No additional examinations need be performed if there are no additional elements identified as being susceptible to the same root cause conditions.
3.3.2 Program Relief Requests An attempt has been made to select RISB locations for examination such that a minimum of >90% coverage (i.e., Code Case N-460 criteria) is attainable.
However, some limitations will not be known until the examination is performed since some locations may be examined for the first time by the specified techniques.
In instances where locations at the time of the examination fail to meet the >90%
coverage requirement, the process outlined in 10 CFR 50.55a will be followed.
Consistent with previously approved RI-ISI submittals, CCNPP will calculate coverage and use additional examinations or techniques in the same manner it has for traditional Section Xi examinations. Experience has shown this process to be weld-specific (e.g., joint configuration). As such, the effect on risk, if any, will not be known until that time.
Relief requests will be submitted per the guidance of 10CFR50.55a(g)(5)(iv) within one (1) year after the end of the interval.
No CCNPP relief requests are being withdrawn due to the RISB application.
Page 10
3.4 Risk Impact Assessment The RISB Program development has been conducted in accordance with Regulatory Guide 1.174 and the requirements of Code Case N-716, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.
This evaluation categorized segments as high safety significant or low safety significant in accordance with Code Case N-716, and then determined what inspection changes are proposed for each system. The changes include changing the number and location of inspections and in many cases improving the effectiveness of the inspection to account for the findings of the RIS_B degradation mechanism assessment.
For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and-will be focused to enhance the. probability of detection (POD) during the inspection process.
3.4.1 Quantitative Analysis Code Case N-716 has adopted the EPRI TR-1 12657 process for risk impact analyses whereby limits are imposed to ensure that the change in risk of implementing the RIS_B Program meets the requirements of Regulatory Guides 1.174 and 1.178. The EPRI criterion requires that the cumulative change in CDF and LERF be less than 1 E-07 and 1 E-08 per year per system, respectively.
For LSS welds, CCDP and CLERP values of 1E-4 and 1E-5 are generally conservatively used, unless pipe segments in the plant internal flooding study are found with higher values. For the CCNPP RISB application, CCDP and CLERP values of 1E-3 and 1E-4 have been used for LSS welds to bound plant internal flooding study results. The 1E-3 and IE-4 values used for CCDP and CLERP was determined based on results from the plant internal flooding study for RWST piping and have been conservatively applied as an upper bound for all LSS welds.
With respect to assigning failure potential for LSS piping, the criteria are defined by Table 3 of the Code Case. That is, those locations identified as susceptible to FAC (or another mechanism and also susceptible to water hammer) are assigned a high failure potential. Those locations susceptible to thermal fatigue, erosion-cavitation, corrosion or stress corrosion cracking are assigned to a medium failure potential and those locations that are identified as not susceptible to degradation are assigned a low failure potential.
In order to streamline the risk impact assessment, a review was conducted to verify that the LSS piping was not susceptible to FAC or water hammer. This review was conducted similar to that done for a traditional RI-ISI application.
Thus, the High failure potential category is not applicable to LSS piping, In lieu of conducting a formal degradation mechanism evaluation for all LSS piping (e.g.
to determine if thermal fatigue is applicable), these locations were conservatively Page 11
assigned to the Medium failure potential ("Assume Medium" in Tables 3.4-1 and 3.4-2) for use in the change-in-risk assessment.
Experience with previous industry RI-ISI applications shows this to be conservative.
CCNPP has conducted a risk impact analysis per the requirements of Section 5 of Code Case N-716 that is consistent with the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-1 12657. The analysis estimates the net change in risk due to the positive and negative influences of adding and removing locations from the inspection program.
The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) values used to assess risk impact were estimated based on pipe break location.
Based on ihese estimated values, a corresponding consequence rank was assigned per the requirements of EPRI TR-1 12657 and upper bound threshold values were used as provided below.
Consistent with the EPRI risk-informed methodology, the upper bound for all break locations that fall within the high consequence rank range was based on the highest CCDP value obtained (e.g., Large LOCA for CCNPP).
Page 12
CCDP and CLERP Values Based on Break Location Break Location Estimated Consequence Upper Bound Designation CCDP CLERP Rank CCDP CLERP LOCA 6.5E-03 6.5E-04 HIGH 6.5E-03 6.5E-04 RCPB pipe breaks that result in a loss of coolant accident - The highest CCDP for Large LOCA was used (0.1 margin used for CLERP)
ILOCA) 1.3E-05 1.3E-06 MEDIUM 1.OE-04 1.0E-05 RCPB pipe breaks that result in an isolable LOCA - Calculated based on Large LOCA CCDP of 6.5E-03 and valve fail to close probability of 2E-3 (0.1 margin used for CLERP)
PLOCA 6.5E-06 6.5E-07 MEDIUM 1.0E-04 1.0E-05 RCPB pipe breaks that result in a potential LOCA - Calculated based on Large LOCA CCDP of 6.5E-3 and valve rupture probability of -1E-3 (0.1 margin used for CLERP)
PPLOCA 6.5E-09 6.5E-10 MEDIUM(1 )
1.0E-04 1.0E-05 Class 2 pipe breaks that result in a potential LOCA - Calculated based on Large LOCA CCDP of 6.5E-3 and valve rupture probability for 2 valves of -1E-6 (0.1 margin used for CLERP)
PLOCA - OC 1.OOE-03 1.00E-03 HIGH 6.5E-03 6.5E-04 RCPB pipe breaks that result in a potential LOCA outside containment - Calculated based on valve rupture probability of -1E-3 FWl 5.OE-04 5.OE-05 HIGH 6.5E-03 6.5E-04 Class 2 pipe breaks that occur in feedwater piping between steam generator and first isolation valve inside containment - CCDP for large steam line break inside containment (SLBI) used (0.1 margin used for CLERP)
FW2 5.3E-06 5.3E-07 MEDIUM 1.OE-04 1.OE-05 Class 2 pipe breaks that occur in feedwater piping between first isolation valve inside containment and containment penetration -
CCDP for loss of feedwater (LOFW) used (0.1 margin used for CLERP)
FW3 5.OE-07 5.OE-08 MEDIUM(1 )
1.OE-04 1.OE-05 Class 2 pipe breaks that occur in feedwater piping outside containment between containment penetration and isolation valve - Calculated based on CCDP for large steam line break upstream of MSIV (LSLBU) of 5E-4 and valve failure probability of -1 E-3 (0.1 margin used for CLERP)
FW4 5.0E-10 5.0E-11 MEDIUM"1 )
1.OE-04 1.OE-05 Non Code Class pipe breaks that occur in feedwater piping outside containment upstream of the outer isolation valve - Calculated based on CCDP for large steam line break upstream of MSIV (LSLBU) of 5E-4 and valve failure probability for 2 valves of -1 E-6 (0.1 margin used for CLERP)
MS1 5.OE-04 5.OE-05 HIGH 6.5E-03 6.5E-04 Class 2 pipe breaks that occur in main steam piping outside containment between containment penetration and MSIV - CCDP for large steam line break upstream of MSIV (LSLBU) used (0.1 margin used for CLERP)
MS2 3.OE-04 3.OE-05 HIGH 6.5E-03 6.5E-04 Non Code Class pipe breaks that occur in main steam piping outside containment downstream of MSIV - CCDP for large steam line break downstream of MSIV (LSLBD) used (0.1 margin used for CLERP)
LSS 1.OE-03 1.OE-04 HIGH 1.OE-03 1.OE-04 Class 2 pipe breaks that occur in the remaining system piping designated as low safety significant - CCDP conservatively estimated based on internal flooding study results for RWST piping (0.1 margin used for CLERP)
Note
- 1. Although the calculated CCDP and CLERP values for the PPLOCA, FW3 and FW4 break locations fall in the "Low" consequence rank range, a "Medium" consequence rank is, conservatively used for risk impact.
Page 13 --
The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as x, and is expected to have a value less than 1 E-
- 08. Piping locations identified as medium failure potential have a likelihood of 20x,. These PBF likelihoods are consistent with References 9 and 14 of EPRI TR-1 12657. In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RISB approach.
Tables 3.4-1 and 3.4-2 present summaries of the RIS_B Program versus ASME Section XI program requirements on a per system basis for Units 1 and 2, respectively, as indicated below.
Unit 1 - 1998 Edition of ASME Code was used for the selection of Category B-F, B-J, C-F-1 and C-F-2 piping welds for the Third Interval IS[ Program.
Unit 2 - 1974 Edition of ASME Code with Summer 1975 Addenda was used for the selection of Category B-F and B-J piping welds for the Second Interval ISI Program; Code Case N-408 was used for the selection of Category C-F-1 and C-F-2 piping welds for the Second Interval ISI Program; no Unit 2 selections were made for the Third Interval prior to the development of the N-578 RI-ISI Program.
The presence of PWSCC was adjusted for in the quantitative analysis by excluding its impact on the failure potential rank. The exclusion of the impact of PWSCC on the failure potential rank and therefore in the determination of the change in risk is appropriate, because PWSCC is a damage mechanism managed by a separate, independent plant augmented inspection program. The RISB Program credits and relies upon this plant augmented inspection program to manage this damage mechanism. The plant PWSCC Program will continue to determine where and when examinations shall be performed. Hence, since the number of PWSCC examination locations remains the same "before" and "after" and no delta exist, there is no need to include the impact of PWSCC in the performance of the risk impact analysis.
Page 14 -
As indicated in the following tables, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RISB Program, and satisfies the acceptance criteria of Regulatory Guide 1.174 and Code Case N-716.
CCNPP Unit 1 Risk Impact Results Systemz' ARCDF Results ARLERF Results w/ POD w/o POD w/ POD w/o POD RC
-1.23E-08
-2.37E-09
-1.23E-09
-2.37E-10 CVC
-7.09E-09
-3.97E-09
-7.09E-10
-3.97E-16 SI 5.35E-09 5.35E-09 5.35E-10 5.35E-10 FW
-4.45E-09
-2.37E-09
-4.45E-10
-2.37E-10 MS 2.43E-10 2.43E-10 2.43E-11 2.43E-11 CS O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 TOTAL
-1.82E-08
-3.12E-09
-1.82E-09
-3.12E-10 Note (1) Systems are described in Table 3.1-1.
CCNPP Unit 2 Risk Impact Results System~')
ARCDF Results ARLERF Results w/POD w/o POD w/IPOD-]
w/o POD RC
-9.26E-09
-1.46E-09
-9.26E-10
-1.46E-10 CVC
-3.74E-09
-2.18E-09
-3.74E-10
-2.18E-10 SI 5.33E-09 5.33E-09 5.33E-10 5.33E-10 FW
-4.52E-09
-2.44E-09
-4.52E-10
-2.44E-10 MS 4.98E-10 4.98E-10 4.98E-11 4.98E-11 CS O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 TOTAL
-1.17E-08
-2.50E-10
-1.17E-09
-2.50E-11 Note (1) Systems are described in Table 3.1-2.
Page 15
3.4.2 Defense-in-Depth The intent of the inspections mandated by ASME Section Xl for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for selecting inspection locations is based upon structural discontinuity and-stress analysis results.
As depicted in ASME White Paper 92-01-01 Rev. 1, Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds, this method has been ineffective in identifying leaks or failures.
EPRI TR-1 12657 and Code Case N-716 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data.
This process has two key independent ingredients; that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense-in-depth is maintained.
First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased.
Secondly, a generic assessment of high-consequence sites has been determined by Code Case N-716 supplemented by plant-specific evaluations thereby requiring a minimum threshold of inspection for important piping whose failure would result in a LOCA or BER break. Finally, Code Case N-716 requires that any piping on a plant-specific basis that has a contribution to CDF of greater than 1 E-06 (or 1 E-07 for LERF) be included in the scope of the application. CCNPP did not identify any such piping.
All locations within the Class 1, 2, and 3 pressure boundaries will continue to be.
pressure tested in accordance with the Code, regardless of its safety significance.
- 4.
IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RISB Program, procedures that comply with the guidelines described in Code Case N-716 will be prepared to implement and monitor the program.
The new program will be implemented at the beginning of the fourth ISI interval.
No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.
The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RIS_B process, as appropriate.
Page 16
The monitoring and corrective action program will contain the following elements:
A.
Identify B.
Characterize C.
(1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans D.
Decide E.
Implement F.
Monitor G.
Trend The RIS_B Program is a living program requiring feedback of new relevant information to ensure the appropriate identification of HSS piping locations. As a minimum, this review will be conducted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant-specific feedback.
For preservice examinations, CCNPP will follow the rules contained in Section 3.0 of N-716.
Welds classified HSS require preservice inspection. The examination volumes, techniques, and procedures shall be in accordance with Table 1. Welds classified as LSS do not require preservice inspection.
- 5.
PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the RISB Program and ASME Section Xl program requirements for in-scope piping is provided in Tables 5-1 and 5-2 for Units 1 and 2, respectively, as indicated below.
Unit 1 - 1998 Edition of ASME Code was used for the selection of Category B-F, B-J, C-F-1 and C-F-2 piping welds for the Third Interval ISI Program.
Unit 2 - 1974 Edition of ASME Code with Summer 1975 Addenda was used for the selection of Category B-F and B-J piping welds for the Second Interval ISI Program; Code Case N-408 was used for the selection of Category C-F-1 and C-F-2 piping welds for the Second Interval ISI Program; no Unit 2 selections were made for the Third Interval prior to the development of the N-578 RI-ISI Program.
CCNPP will implement the new program at the beginning of the fourth ISI interval.
Page 17
- 6.
REFERENCESIDOCUMENTATION USNRC Safety Evaluation for Calvert Cliffs Nuclear Plant, Unit Nos. 1 and 2 - American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) -
Relief for Risk-Informed Inservice Inspection of Piping, dated April 16, 2003 EPRI Report 1006937, Extension of EPRI Risk Informed ISI Methodology to Break Exclusion Region Programs EPRI TR-112657, Revised Risk-Informed Inservice Inspection Evaluation Procedure, Rev. B-A ASME Code Case N-716, Alternative Piping Classification and Examination Requirements,Section XI Division I Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping USNRC Safety Evaluation for Grand Gulf Nuclear Station Unit 1, Request for Alternative GG-ISI-002-Implement Risk-Informed ISI based on ASME Code Case N-716, dated September 21, 2007 USNRC Safety Evaluation for DC Cook 'Nuclear Plant, Units 1 and 2, Risk-Informed Safety-Based ISI program for Class 1 and 2 Piping Welds, dated September 28, 2007 Supporting Onsite Documentation SI Calc 0800144.301, N-716 Evaluation of CCNPP Units I and 2, Revision 0 Page 18
List of Acronyms BER CC CCDP CDF CLERP CS CVC DMs ECSCC E-C:
FAC FW FW1 FW2 FW3 FW4 HSS IGSCC ILOCA LERF LOCA LSS MIC MS MS1 MS2 MSIV PIT PLOCA PLOCA -OC PPLOCA POD PWR: FW PWSCC RC RCPB RCPBIFIV RCPBOc RI-ISI RIS_B SDC SI SXl TASCS TGSCC TT Vol/Sur Sur Break Exclusion Region Crevice Corrosion Condition Core Damage Probability Core Damage Frequency Condition Large Early Release Probability Containment Spray System Chemical & Volume Control System Degradation Mechanisms External Chloride Stress Corrosion Cracking Erosion-Cavitation Flow-Accelerated Corrosion Feedwater System Feedwater piping between steam generator and first isolation valve inside containment Feedwater piping between first isolation valve inside containment and containment penetration Feedwater piping outside containment between containment penetration and isolation valve Feedwater piping outside containment upstream of the outer isolation valve High Safety Significant Intergranular Stress Corrosion Cracking Isolable Loss of Coolant Accident Large Early Release Frequency Loss of Coolant Accident Low Safety Significant Microbiologically-Influenced Corrosion Main Steam System Main Steam piping outside containment between containment penetration and MSIV Main Steam piping outside containment downstream of MSIV Main Steam Isolation Valve Pitting
-Potential Loss of Coolant Accident Potential Loss of Coolant Accident - Outside Containment Potential Loss of Coolant Accident (2 Valves)
Probability of Detection Pressurized Water Reactor: Feedwater Primary Water Stress Corrosion Cracking Reactor Coolant System Reactor Coolant Pressure Boundary Reactor Coolant Pressure Boundary Inside First Isolation Valve Reactor Coolant Pressure Boundary Outside Containment Risk-Informed Inservice Inspection Risk-Informed / Safety-Based Inservice Inspection Shutdown Cooling Safety Injection System Section Xl Thermal Stratification, Cycling, and Striping Transgranular Stress Corrosion Cracking Thermal Transients Volumetric and Surface Surface Page 19
Table 3.1-1 N-716 Safety Significance Determination for Unit 1 N-716 Safety Significance Determination Safety Significance
System Description
>1 E-6cDF H
-Count RCPB J SDC PWR: FW j BER j >1E7 High Low RC - Reactor Coolant 241 V
CVC - Chemical & Volume Control 123 V
/
SI - Safety Injection 111 i
V V
57 V
V 1088 V
V 4
V V
MS - Main Steam 135 V
V 57 V
CS - Containment Spray 191
SUMMARY
RESULTS FOR ALL SYSTEMS 111 V
V V
364 V
V 57 V
V 55 V
V 139 V
V 1336 V
TOTALS 2062 726 1336 Page 20
Table 3.1-2 N-716 Safety Significance Determination for Unit 2 N-716 Safety Significance Determination Safety Significance
System Description
Weld D
>1 E-67cDF High Low RC - Reactor Coolant 233 V
CVC - Chemical & Volume Control 103 V
V SI - Safety Injection 107 V
V 57 V
V 1159 V
V V
V 46 V
V 2
- V V
MS - Main Steam 128 V
V 48 V
CS - Containment Spray 187 V
SUMMARY
RESULTS FOR ALL SYSTEMS 107 V
VV 336 57 V
V 2
V" V"
V 46 V
V 130 V
V 1394 V
TOTALS 2072 678 1394 Page 21
Table 3.2-1 Failure Potential Assessment Summary for Unit 1 Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive TASCS T TT IGSCC TGSCC ECSCC PWSCC MIC PIT Cc E-C FAC RC V
V V
CVC SI(2)
FW, MS(2)
CS(2)
Notes
- 1. Systems are described in Table 3.1-1.
- 2.
A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the CS system in its entirety, as well as portions of the SI and MS systems.
Page 22
Table 3.2-2 Failure Potential Assessment Summary for Unit 2 Thermal Fatigue Stress Corrosion Cracking Localized Corrosion Flow Sensitive I
TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RCc CVC Si(2)
FW MS(2)
CS(2)
Notes
- 1.
Systems are described in Table 3.1-2.
- 2.
A degradation mechanism assessment was not performed on low safety significant piping segments. This includes the CS system in its entirety, as well as portions of the SI and MS systems.
Page 23
Table 3.3-1 N-716 Element Selections for Unit I Systemr()
Selections HSS DMs RCPB RCPB IFNV RCPBOc BER RC Required 25 of 241 TASCS, TT 4 of 13 25 of 241 17 n/a n/6 TT, (PWSCC) 1 of 3 TASCS 5 of 17 TT 2 of 7 None (PWSCC) 1 of 3 Made 29 TASCS, TT 4
29 29 n/a n/a TT, (PWSCC) 3 TASCS 5
TT 2
None (PWSCC) 2 CVC Required 13 of 123 TT 3 of 10 13 of 123 9
n/a n/a Made 13 TT 6
13 9
n/a n/a SI Required 17 of 168 IGSCC 1 of 4 12 of 111 8
1 of 5 n/a None (PWSCC) 1 of 1 Made 17 IGSCC 1
14 12 1
n/a None (PWSCC) 1 FW Required 6 of 59 TASCS, TT 2 of 8 n/a n/a n/a 1 of 4 Made 6
,TASCS, TT 4
n/a n/a n/a 1
MS Required 14 of 135 n/a n/a n/a n/a 14 of 135 Made 14 n/a n/a n/a n/a 14 CS Required n/a n/a n/a n/a n/a n/a Made n/a n/a n/a n/a n/a n/a TOTAL Made 79 28 56 50 1
15 Note
- 1.
Systems are described in Table 3.1-1.
Page 24
Table 3.3-2 N-716 Element Selections for Unit 2 System(1)
Selections HSS DMs RCPB RCPBIFIV RCPBOc BER RC Required 24 of 233 TASCS, TT 4 of 16 24 of 233 16 n/a n/a TT, (PWSCC) 1 of 3 TASCS 2 of 8 TT 2of5 None (PWSCC) 1 of 3 Made 26 TASCS, TT 4
26 26 n/a n/a TT, (PWSCC) 3 TASCS 2
TT 2
None (PWSCC) 2 CVC Required 11 of 103 TT 2 of 7 11 of 103 8
n/a n/a Made 11 TT 3
11 10 n/a n/a SI Required 17 of 164 IGSCC 2 of 6 11 of 107 8
1 of 5 n/a None (PWSCC) 1 of 1
.Made 17 IGSCC 2
14 11 1
n/a None (PWSCC) 1 FW Required 5 of 50 TASCS, TT 2 of 8 n/a n/a n/a 1 of 4 Made 5
TASCS, TT 4
n/a n/a n/a I
MS Required 13 of 128 n/a n/a n/a n/a 13 of 128 Made 13 n/a n/a n/a n/a 13 CS Required n/a n/a n/a n/a n/a n/a Made n/a n/a n/a n/a n/a n/a TOTAL Made 72 23 51 47 1
14 Note
- 1. Systems are described in Table 3.1-2.
Page 25
Table 3.4-1 Risk Impact Analysis Results for Unit 1 S
Safety B
L (2)
Failure Potential Inspections CDF Impact LERF Impact m
Significance BDMs Rank 3 SXI40)
RISB Delta wI POD w/o POD wI POD wlo POD RC High LOCA TASCS, TT Medium 3
4 1
-3.51 E-09
-6.50E-10
-3.51 E-10
-6.50E-11 RC High LOCA TT, (PWSCC)
Medium (Medium) 2 3
1
-2.73E-09
-6.50E-10
-2.73E-10
-6.50E-11 RC High LOCA TASCS Medium 2
5 3
-5.07E-09
-1.95E-09
-5.07E-10
-1.95E-10 RC High LOCA TT Medium 3
2
-1
-1.17E-09 6.50E-10
-1.17E-10 6.50E-11 RC High LOCA None (PWSCC)
Low (Medium) 2 2
0 0.00E+00 0.00E+00 0.00E+00 0.00E+00 RC High LOCA None Low 20 13
-7 2.28E-10 2.28E-10 2.28E-11 2.28E-11 RC High PLOCA None Low 0
0 0
0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 TOTAL
-1.23E-08
-2.37E-09
-1.23E-09
-2.37E-1 0 CVC High LOCA TT Medium 0
6 6
-7.02E-09
-3.90E-09
-7.02E-10
-3.90E-10 CVC High LOCA None Low 0
2 2
-6.50E-11
-6.50E-11
-6.50E-12
-6.50E-12 CVC High ILOCA None Low 0
4 4
-2.OOE-12
-2.00E-12
-2.OOE-13
-2.OOE-13 TOTAL
-7.09E-09
-3.97E-09
-7.09E-10
-3.97E-10 SI High PLOCA IGSCC Medium 1
1 0
0.OOE+00 0.OOE+00 O.OOE+00 0.00E+00 SI High LOCA None (PWSCC)
Low (Medium) 1 1
0 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 SI High LOCA None Low 6
11 5
-1.63E-10
-1.63E-10
-1.63E-11
-1.63E-11 SI High PLOCA None Low 18 0
-18 9.OOE-12 9.OOE-12 9.OOE-13 9.OOE-13 SI High PPLOCA None Low 3
3 0
0.00E+00 0.OOE+00 0.OOE+00 0.00E+00 SI High PLOCA - OC None Low 1
1 0
O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 SI Low LSS N/A Assume Medium 55 0
-55 5.50E-09 5.50E-09 5.50E-10 5.50E-10 TOTAL 5.35E-09 5.35E-09 5.35E-10 5.35E-10 Page 26
Table 3.4-1 (Cont'd)
Risk Impact Analysis Results for Unit 1 eafet Break Location(2 )
Failure Potential Inspections CDF Impact LERF Impact Significance DMs Rank(3 )
SXI( 4 )
RISB [ Delta w/ POD w/o POD w/ POD w/o POD FW High FW1 TASCS, TT Medium 0
4 4
-4.68E-09
-2.60E-09
-4.68E-10
-2.60E-10 FW High FW1 None Low 7
0
-7 2.28E-10 2.28E-10 2.28E-11 2.28E-11 FW High FW2 None Low 3
1
-2 1.OOE-12 1.OOE-12 1.00E-13 1.00E-13 FW High FW3 None Low 0
0 0
0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 FW High FW4 None Low 0
1 1
-5.OOE-13
-5.OOE-13
-5.OOE-14
-5.OOE-14 TOTAL
-4.45E-09
-2.37E-09
-4.45E-10
-2.37E-10 MS High MS1 None Low 3
6 3
-9.75E-11
-9.75E-11
-9.75E-12
-9.75E-12 MS High MS2 None Low 0
8 8
-2.60E-10
-2.60E-10
-2.60E-11
-2.60E-11 MS Low LSS N/A Assume Medium 6
0
-6 6.00E-10 6.OOE-10 6.OOE-1 1 6.00E-11 TOTAL 2.43E-10 2.43E-10 2.43E-11 2.43E-11 CS Low LSS N/A Assume Medium 0
0 0
O.00E+00 O.OOE+00 0.00E+00 0.00E+00 TOTAL O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 GRAND
-1.82E-08
-3.12E-09
-1.82E-09
-3.12E-10 TOTAL Notes
- 1.
Systems are described in Table 3.1-1.
- 2.
The "LSS" break location designation in Table 3.4-1 is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).
- 3.
The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]
- 4.
Only those ASME Section XI Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.
Page 27
Table 3.4-2 Risk Impact Analysis Results for Unit 2 System0)
Safety Break Location( 2 )
Failure Potential Inspections CDF Impact LERF Impact significance DMs Rank (3 )
SXJ 4)
RISB Delta w/ POD w/o POD W1 POD w/o POD RC High LOCA TASCS, TT Medium 3
4 1
-3.51E-09
-6.50E-10
-3.51E-10
-6.50E-11 RC High LOCA TT, (PWSCC)
Medium (Medium) 3 3
0
-2.34E-09 0.00E+00
-2.34E-10 0.00E+00 RC High LOCA TASCS Medium 0
2 2
-2.34E-09
-1.30E-09
-2.34E-10
-1.30E-10 RC High LOCA TT Medium 2
2 0
-1.56E-09 0.OOE+00
-1.56E-10 0.00E+00 RC High LOCA None (PWSCC)
Low (Medium) 2 2
0 0.00E+00 0.00E+00 0.OOE+00 0.OOE+00 RC High LOCA None Low
.28 13
-15 4.88E-10 4.88E-10 4.88E-11 4.88E-11 RC High PLOCA None Low
.0 0
0 0.OOE+00 0.OOE+00 0.00E+00 0.00E+00 TOTAL
-9.26E-09
-1.46E-09
-9.26E-10
-1.46E-10 CVC High LOCA TT Medium 0
3 3
-3.51E-09
-1.95E-09
-3.51E-10
-1.95E-10 CVC High LOCA None Low 0
7 7
-2.28E-10
-2.28E-10
-2.28E-11
-2.28E-11 CVC High ILOCA None Low 0
1 1
-5.00E-13
-5.OOE-13
-5.00E-14
-5.00E-14 TOTAL
-3.74E-09
-2.18E-09
-3.74E-10
-2.18E-10 SI High PLOCA IGSCC Medium 1
2 1
-1.O0E-11
-1.OOE-11
-1.OOE-12
-1.00E-12 SI High LOCA None (PWSCC)
Low (Medium) 1 1
0 0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 SI High LOCA None Low 11 10
-1 3.25E-11 3.25E-11 3.25E-12 3.25E-12 SI High PLOCA None Low 6
0
-6 3.OOE-12 3.OOE-12 3.OOE-13 3.OOE-13 SI High PPLOCA None Low 11 3
-8 4.OOE-12 4.OOE-12 4.OOE-13 4.OOE-13 SI High PLOCA - OC None Low 1
1 0
0.OOE+00 0.OOE+00 0.OOE+00 0.00E+00 SI Low LSS N/A Assume Medium 53 0
-53 5.30E-09 5.30E-09 5.30E-10 5.30E-10 TOTAL 5.33E-09 5.33E-09 5.33E-10 5.33E-10 Page 28
Table 3.4-2 (Cont'd)
Risk Impact Analysis Results for Unit 2 System" Safety -
Break Location (2)
Failure Potential
{
Inspections CDF Impact LERF Impact Significance DIVs Rank(')
SXI(4)
RISB IDelta w/ POD w/o POD w/ POD w/o POD FW High FW1 TASCS, TT Medium 0
4 4
-4.68E-09
-2.60E-09
-4.68E-10
-2.60E-10 FW High FW1 None Low 5
0
-5 1.63E-10 1.63E-10 1.63E-11 1.63E-11 FW High FW2 None Low 2
0
-2 1.00E-12 1.00E-12 1.00E-13 1.00E-13 FW High FW3 None Low 1
0
-1 5.00E-13 5.OOE-13 5.00E-14 5.OOE-14 FW High FW4 None Low 0
1 1
-5.00E-13
-5.OOE-13
-5.00E-14
-5.00E-14 TOTAL
-4.52E-09
-2.44E-09
-4.52E-10
-2.44E-10 MS High MS1 None Low 16 10
-6 1.95E-10 1.95E-10 1.95E-11 1.95E-11 MS High MS2 None Low 0
3 3
-9.75E-11
-9.75E-11
-9.75E-12
-9.75E-12 MS Low LSS N/A Assume Medium 4
0
-4 4.00E-10 4.00E-10 4.00E-11 4.00E-11 TOTAL 4.98E-1 0 4.98E-10 4.98E-1 I 4.98E-1 1 CS Low LSS N/A Assume Medium 0
0 0
0.OOE+00 0.OOE+00 0.00E+00 0.OOE+00 TOTAL 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 GRAND
-1.17E-08
-2.50E-10
-1.17E-09
-2.50E-1I TOTAL Notes
- 1.
Systems are described in Table 3.1-2.
- 2.
The "LSS" break location designation in Table 3.4-1 is used to identify those Code Class 2 locations that are not HSS because they do not meet any of the five HSS criteria of Section 2(a) of N-716 (e.g., not part of the BER scope).
- 3.
The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low' dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]
- 4.
Only those ASME Section XI Code inspection locations that received a volumetric examination in addition to a surface examination are included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.
Page 29
Table 5-1 Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 1 Systemn11 Safety Significance Break Location Failure Potential Code Weld Section XI Code Case N-716 High Low DMs J
Rank(')
Category Count Vol/Sur Sur Only RISB I OtherS)
RC V
LOcA TASCS, TT Medium B-J 13 3
0 4
B-F 2
2 0
2 RC V
Medium (Medium)
B.J DMW 1
0 0
1 RC LOCA TASCS Medium B-J 17 2
5 5
RC LOCA TT Medium B-J 7
3 0
2 B-F 2
2 0
Low (Medium)
B.jDMW 1
0 0
0 B-JDMw 13 0
0 13 RC V
LOCA None Low" B-J 170 20 13 0
RC
/
PLOCA None Low B-J 15 0
3 0
B-J DMw 2
0 0
2 CVC V"
LOCA TT Medium B-J 8
0 2
4 B-J DMw 1
0 1
1 CVC V
12 1 +
-V -
CVC V
ILOCA None Low B-J 46 0
23 4
1 0
1 SI V
Low (Medium)
B-JDMw 1
1 0
1 B-J 4
1 0
4 SI V
LOCA None Low B-J 24 5
0 7
SI V
PLOCA None Low B-J 73 18 0
0 SV PPLOCA None Low C-F-1 57 3
0 3
SI V
PLOCA - OC None Low B-J 5
1 0
1 S
V LSS N/A Assume Medium C-F-1 1088 55 42 0
Page 30
Table 5-1 (Cont'd)
Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 1 Safety Significance Failure Potential Cd Wed Section XI Code Case N-716
- System(1 Break LocationCoe Wl High Low-ý DMs Rank(2 )
Category Count Vol/Sur Sur Only RISB Othe FW V.
FW1 TASCS, TT Medium C-F-2 8
0 0
4 FW V
FW1 None Low C-F-2 24 7
0 0
FW FW2 None Low C-F-2 8
3 0
1 FW FW3 None Low C-F-2 15 0
0 0
FW V
FW4 None Low NNS 4
0 0
1 MS MS1 None Low C-F-2 126 3
9 6
MS MS2 None Low NNS 9
0 0
8 MS LSS N/A Assume Medium C-F-2 57 6
0 0
CS V
LSS N/A Assume Medium C-F-1 191 0
0 0
Notes
- 1.
Systems are described in Table 3.1-1.
- 2.
The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]
- 3.
The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BWR to be credited toward the 10% requirement. This option is not applicable for the CCNPP RISB application. The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals.
Page 31
Table 5-2 Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 2 Safety Significance Failure Potential Code Weld Section XI Code Case N-716
{System(1 Bra octo High Low DMs Rank(2)
Category Count Vol/Sur Sur Only RISB OtherP3 )
RC V
LOCA TASCS,TT Medium B-J 16 3
0 4
B-F 2
2 0
2 RC V
Medium (Medium)
B-JDMW 1
1 0
1 RC V
LOCA TASCS Medium B-J 8
0 5
2 RC V
LOCA TT Medium B-J 5
2 0
2 B-F 2
2 0
2 RC V
Low (Medium)
B-J 0MW 1
0 1
0 B-J DMw 13
-8 5
13 RC V
LOCA None Low B-J 167 20 23 0
RC V
PLOCA None Low B-J 18 0
5 0
B-J DMw 2
0 2
2 CVC V
LOCA TT Medium B-J 5
0 3
1 B-JD
~
1 0
1 1
CVC V
LOCA None Low B-J 45 0
2 6
CVC V
ILOCA None Low B-J 50 0
9 1
SI V
PLOCA IGSCC Medium B-J 6
1 0
2 Si V
Low (Medium)
B-J DMw 1
1 0
1 B-J DMw 4
4 0
4 Si V
LOCA None Low B-J 23 7
0 6
SI v
PLOCA None Low B-J 68 6
0 0
SI V
PPLOCA None Low C-F-1 57 11 0
3 SI V
PLOCA - OC None Low B-J 5
1 0
1 SI V
LSS N/A Assume Medium C-F-1 1159 53 56 0
Page 32
Table 5-2 (Cont'd)
Inspection Location Selection Comparison Between ASME Section XI Code and Code Case N-716 for Unit 2 ste Safety Significance BreakLocation Failure Potential Code Weld Section XI Code Case N-716 High Low DMs RankO2)
Category Count Vol/Sur tSur Only RISB I OtherP3)
FW 1"
FW1 TASCS,TT Medium C-F-2 8
0 0
4 FW
/
FW1 None Low C-F-2 21 5
0 0
FW FW2 None Low C-F-2 6
2 0
0 FW FW3 None Low C-F-2 13 1
0 0
FW FW4 None Low NNS 2
0 0
1 MS MS1 None Low C-F-2 118 16 0
10 MS.
MS2 None Low NNS 10 0
0 3
MS V
LSS N/A Assume Medium C-F-2 48 4
0 0
CS V
LSS N/A Assume Medium C-F-1 187 0
0 0
Notes
- 1. Systems are described in Table 3.1-2.
- 2.
The failure potential rank for high safety significant (HSS) locations is assigned as "High", "Medium" or "Low" dependent upon potential susceptibly to the various types of degradation mechanisms. [Note: LSS locations were conservatively assumed to be a rank of Medium (i.e., "Assume Medium").]
- 3.
The column labeled "Other" is generally used to identify plant augmented inspection program locations credited per Section 4 of Code Case N-716. Code Case N-716 allows the existing plant augmented inspection program for IGSCC (Categories B through G) in a BVWR to be credited toward the 10% requirement. This option is not applicable for the CCNPP RISB application. The "Other" column has been retained in this table solely for uniformity purposes with other RISB application template submittals.
Page 33
ENCLOSURE (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Calvert Cliffs Nuclear Power Plant, Inc.
December 29, 2008
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Introduction Constellation Energy Group (CEG) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating CEG nuclear generation sites.
This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews.
The following information describes this approach as it applies to the CCNPP PRA.
PRA Maintenance and Update The CEG risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the CEG risk management program, which consists of a governing procedure CNG-CM-2.01, Probabilistic Risk Assessment (PRA) Directive, and subordinate implementation procedures.
CEG procedure CNG-CM-1.01-3003, "Probabilistic Risk Assessment Configuration Control" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating CEG nuclear generation sites. CNG-CM-1.01-3004, "PRA Process for Internal Evaluations", includes the process to meet the overall CEG risk management program, including CNG-CM-1.01-3003, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operational experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:
9 Design changes and procedure changes are reviewed for their impact on the PRA model.
PRA screens are required for all design and procedure changes.
New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon percentage changes in PRA failure rates or frequencies.
This has recently been noted as an area for improvement in our configuration control procedure. This will be revised to require a data update on a periodic frequency as well.
In addition to these activities, CEG risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:
Documentation of the PRA model, PRA products, and bases documents.
The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.
Guidelines for updating the full power, internal events PRA models for CEG nuclear generation sites.
Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65(a)(4)).
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions PRA Self Assessment and Peer Review Several assessments of technical capability have been made, and continue to be planned, for the CCNPP PRA models. These assessments are as follows:
An independent PRA peer review was conducted under the auspices of the Westinghouse Owners Group in November of 2001, following the Industry PRA Peer Review process [Reference 1] at that time. This peer review included an assessment of the PRA model maintenance and update, process.
The CCPRA Revision 0 internal events model has been reviewed as part of the Combustion Engineering Owners Group (CEOG) Peer Review Process in November 2001. The peer review team consisted of five full-time members.
Four members were utility personnel and the remaining member was the team leader/facilitator from Westinghouse.
Two of the reviewers had participated in previous CEOG peer reviews. The reviewers had a combined PRA experience of approximately seventy-five years. It should also be noted that CCNPP PRA personnel have participated in every CEOG sponsored peer review and through this process have gained considerable insight and experience as to the technical requirements for developing an effective PRA. The peer review was performed using the existing version of NEI 00-02.
The sections where open gaps from Appendix B of RG 1.200 Revision 1 are included in Open Items noted in Attachment 1. Note that the model at this time was in RISKMAN (Large Event Tree) software.
A recent conversion has been completed to CAFTA software. A detailed review of the CAFTA model results was performed to validate the CAFTA model and to address any differences in the results.
Note.
The results of the evaluations in this submittal use the worst case values from the RISKAAN and CAFTA models to ensure the analysis is conservative.
The peer review addressed eleven technical elements. The team uses a checklist for each technical element as a framework to evaluate the scope, comprehensiveness, completeness and fidelity of the PRA being reviewed. Each item on the checklist is evaluated with a grade:
Exceeds Meets Marginal Inadequate Items evaluated other than meets are documented with Facts and Observations (F&Os). Each F&O is provided with a level of significance (A: Extremely Important, B: Important, C: Desirable, D: Minor, 0: Observation, S: Superior).
At our request, the team also evaluated the CCPRA against the draft ASME PRA Standard, Revision 14A. This was accomplished by using the ASME Standard supporting requirements as sub-tier criteria.
The team was provided with a self-assessment of the CCPRA to the draft ASME PRA requirements.
The team used this self-assessment to aid in gauging the technical adequacy of the CCPRA. This peer review included an assessment of the PRA model maintenance and update process.
The process reviewed at this time was a plant specific process, not the corporate process noted above in CNG-CM-1.01-3003.
2
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions A summary of the disposition of 2001 Industry PRA Peer Review facts and observations (F&Os) for the CCNPP models was documented WCAP-15801, "Calvert Cliffs Units 1 and 2: Probabilistic Risk Assessment Peer Review Report" (Reference 5). As noted in that document, there were no significance level A F&Os from the peer review, and all, with one exception, significance level B F&Os were addressed and closed out.
The exception was regarding interfacing system LOCAs.
The peer team finding was related to the documentation level of this initiator. The frequencies compare well with NUREG/CR-5750 values. CCNPP plans to update this analysis as PWROG guidance is developed in 2009 along with an overall documentation update for the CAFTA conversion. At our request, the team also evaluated the CCPRA against the draft ASME PRA Standard, Revision 14A. This was accomplished by using the ASME Standard supporting requirements as sub-tier criteria. The team was provided with a self-assessment of the CCPRA to the draft ASME PRA requirements. The team used this self-assessment to aid in gauging the technical adequacy of the CCPRA.
All eleven technical elements were found to "Meet" requirements.
The review team found several strengths.
Below is an excerpt from the peer review report.
"The review found several strengths. The Calvert Cliffs PRA staff is knowledgeable and is committed to improving the Calvert Cliffs PRA. Calvert Cliff's management is also committed to improving the quality of the PRA, as witnessed by the resources dedicated to the maintenance and application of the PRA. Throughout the peer review, the Calvert Cliffs PRA staff was cooperative and helpful. The documentation supporting the PRA is the most extensive and thorough of all PRAs reviewed to date and was readily available to the review team, which greatly facilitated the review. Prior to the arrival of the peer review team, the Calvert Cliffs PRA staff performed a self-assessment against the NEI criteria using the peer review checklists and a self-assessment against the requirements presented in draft 14A of the ASME PRA standard. The results of these self-assessments were documented with an annotated checklist that cross-correlated both sets of requirements. This self-assessment appears to have been thorough and objective.
The Calvert cliffs peer review also included a review of the Calvert Cliffs PRA against the ASME High Level Requirements (HLRs) for 9 of 10 key PRA areas. Internal Flooding is considered as not being assessed since five of the seven Internal Flooding HLRs were not reviewed in sufficient depth to evaluate the compliance. Although the LERF element was reviewed, the documentation HLR for LERF was not assessed due to a delay in locating the MAAP analyses.
Overall, the Calvert Cliffs PRA met with the requirements of 45 of 46 assessed ASME PRA Standard HLRs for a Category II PRA. (Note: Compliance with an HLR does not imply 100%
compliance with all Supporting Requirements (SRs) for that HLR.) One HLR was not met due to the lack of uncertainty analyses.
Thus, it is concluded that the Calvert Cliffs PRA is sufficient to support Category II risk-informed applications with support from deterministic analyses to address any individual weaknesses that may impact the specific analyses."
A total of forty-three F&Os were identified. No "A" level significant issues were identified. The issues that were identified are focused on localized areas of the PRA.
3
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions An estimated core damage frequency reduction of five to ten percent is expected as a result of incorporating the significant issues identified by this review. These issues are discussed below.
Of the 213 sub-elements reviewed, eighty-seven percent were evaluated as meeting or exceeding the requirements.
Three sub-elements were graded as "Inadequate." All three are associated with the lack of an uncertainty analysis. The need to perform this analysis has been previously identified. The uncertainty analysis for this proposed application is not judged necessary or to have an impact on the analysis.
A grade of "Exceeds" was assigned to four sub-elements.
IE-16 DA-15 ST-4 L2-14 Traceable initiating event documentation Documentation of failure probabilities that do not fit into the basic event database Reactor pressure vessel failure modeled appropriately Containment capability is analyzed under severe accident conditions Although Internal Flooding was not reviewed in detail the self assessment for flooding showed that we met all major elements for Category It. Subsequent reviews by SAIC of the Flood model, regarding an update to newerEPRI guidance, was that the flood model was of high quality.
The below table summarizes the results:
Peer Review Sub-element Grades
.....e
,Number ofrSub-elemehts.
Percent Exceeds 4
2%
Meets 181 85%
Marginal 25 12%
Inadequate 3
1%
A total of forty-three Facts and Observations (F&Os) were written.
Facts and Observations Types Type D6Dsciti6ijin Nurmiber A
Extremely Important 0
B Important 11 C -
Desirable 18 D
Minor I
0 Observation 13 S
Superior 1
Total 43 4
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Peer Review "B" Level Facts & Observations A summary of the eleven "B" level F&Os, with'status and impact on N-716 is provided below:
IE-01 The CCPRA uses a SGTR initiating event frequency based on NUREG/CR-5750 which covers industry experience up to 1995. A CE Owners Group(CEOG) document issued in June 2000 has a new SGTR frequency that is about twenty percent lower.
CCNPP Response: CCNPP PRA has been updated with NUREG/CR-6928 as the prior for SGTR initiating event frequency. (NUREG/CR-5750 data updated in NUREG/CR-6928).
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
IE-02 The CCPRA does not explicitly model the spurious RCP seal failure and spurious primary safety relief valve opening as initiating events.
CCNPP Response: These have been added explicitly into the CCNPP PRA.
N-716 Evaluation of CCNPP Units I and 2 Impact: No impact as the issue addressed in the model used in application.
IE-05 The CCPRA uses a draft reference for the bases of its interfacing LOCA initiating events. A CCPRA open item (CRMP 199) had been previously issued in June 2000 to require an update of this analysis. The frequencies used for these initiators are consistent with typical industry values so this should not have a significant impact.
CCNPP Response: As stated by the reviewers, this issue is not expected to cause any significant impact. The completion of the documentation for the interfacing LOCA has been prioritized after the other more significant issues. The issue had been previously captured in our configuration control program as open item (CRMP 199). Reviews of the current analysis show it to be reasonable. This analysis will be updated in the, update of ASME compliance for the CAFTA Fault Trees developed. This is scheduled in 2009.
N-716 Evaluation of CCNPP Units 1 and 2 Impact:
The results appear reasonable when compared with NRC guidance in NUREG/CR-5750 and there is no indication that updates will change the conclusions on risk categories for piping that is related to potential ISLOCAs.
However, CEG will review the results of any updates to determine if any changes are required in piping risk categories.
IE-07 The failure mode associated with the spurious actuation of the over-current or 86-lockout device is not considered as a potential initiating event. This failure is also not considered in the accident mitigation portion of the model.
CCNPP Response: This issue has been addressed in the CCNPP PRA.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
5
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions AS-01 The CCPRA models single and multi-tube SGTRs.
For multi-tube SGTRs, no recovery is currently credited. Crediting recovery could reduce the calculated CDF by fifteen percent.
CCNPP Response: This is the most significant finding of this peer review. The CCPRA models both single tube and multi-tube SGTRs. For multi-tube ruptures, we have taken a conservative approach pending time to more thoroughly address this issue. The issue was captured as an open item (CRMP 285 and 336).
However, the perspective of the reviewers helped clarify the appropriate approach for modeling this issue. The improved modeling was incorporated into the CCPRA Revision I model.
N-716 Evaluation of CCNPP Units I and 2 Impact: No impact as the issue addressed in the model used in application.
AS-04 This finding recommends the incorporation of the new RCP seal LOCA model. This model is currently under review by the NRC. Implementation guidance is also under development by the CEOG.
CCNPP Response: The new RCP Seal model approved by the NRC (WCAP-16175) has been implemented.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
SY-02 The failure of the time delay relay associated with the steam admission valves to the TD AFW pump turbines is not included in the CCPRA. The relay failure could result in the over speed failure of the TD AFW pump.
CCNPP Response: Failure Mode is added to the model.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
SY-08 On loss of component cooling, the CCPRA credits the possibility of failure of the RCP motor before RCP seal failure in order to prevent a seal failure. If the motor is stopped by failure, the likelihood of seal failure is significantly reduced.
CCNPP Response: RCP Motor is no longer credited as a beneficial failure.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
DA-01 NUREG/CR-5497 provides the latest industry data source for common cause failure events. The CCPRA has adopted this data for a very limited scope of components (MSSVs and the PORVs).
CCNPP Response: The use of NUREG/CR-5497 and updates via INEL has been incorporated into the PRA Model.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
6
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions HR-01 The CCPRA quantified the pre-initiator human actions using a failure rate based on NUREG/CR-1278.
The values were treated as mean values.
However, the values in NUREG/CR-1278 are median values.
CCNPP Response: Updated mean values incorporated into the CCNPP PRA model N-716 Evaluation of CCNPP Units I and 2 Impact: No impact as the issue addressed in the model used in application.
HR-03 There are several human actions that use Unit-2 to help mitigate Unit-I. No example of an analysis could be found that considered the impact of the operating status (operation, outage, and startup) on the human actions.
CCNPP Response:
A review of actions which could be impacted by a dual unit trip was completed. These actions are degraded on dual unit trips to account for additional resources and other relevant shaping factors.
N-716 Evaluation of CCNPP Units I and 2 Impact: No impact as the issue addressed in the model used in application.
During 2005, the CCNPP PRA model results were evaluated in the WOG PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.
Results of this cross-comparison are presented in WCAP-16464 [Reference 4]. Noted in this document was the fact that, after allowing for plant-specific features, there are no MSPI cross-comparison outliers for CCNPP. There was an initial issue on asymmetries between HPSI pumps which was resolved and documented (See Table 4.2-2 of Reference 4).
A gap analysis for the CCNPP PRA internal events model is currently being performed by a group of vendors. This gap analysis is using the available version of the ASME PRA Standard [Reference 6] and Regulatory Guide 1.200, Revision 1 [Reference 3]. The gap analysis is focusing on the newly developed CAFTA model.
The documentation for the RISKMAN version of the model is sufficiently well developed at this time to meet category II for documentation and plans are in place for an update in 2009 for the CAFTA model to RG 1.200, Revision 1, Category II. The comments on the documentation for the RISKMAN model were generally positive and the concerns noted were in the layout of the documentation versus the technical content. The gap assessment will be complete early in 2009.
7
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Changes since Revision 0 (Peer Reviewed)
The table below lists the revisions made to the CCNPP PRA Model since the CEOG Peer Review.
Table I CCPRA Development History Mb-del unit-. Descitio-n
-Unit-2 De~sbipion..
D. aDte Revision 4 Internal Events and High Winds RISKMAN Model converted into Same as Unit 1 11/2008 (CAFTA)
CAFTA Internal and External Events Updated RCP Seal LOCA model to WCAP-16175 Revisions 0
Impact of dual unit trips Thru 2 & 3 on Human Actions.
Same as Unit 1 7/2006 Updated common cause factors. More realistic modeling of Switchgear HVAC Internal and External Events -
Revision 1 Internal and External Events development based on over 200 10/2002 identified changes between Unit-I and 2.
Changes between Revision 0 (Internal Events Peer Reviewed) and Current Model There are a number of changes between the Revision 0 model and the Revision 1 model. The majority of the substantive internal events changes are directly a result of the internal events peer review.
The internal Revision 0 CDF was 4.OE-5. The internal events for Revision 1 CDF was 3.4E-5. Revision 2 CDF for internal events was 2.75E-05. The final draft version in RISKMAN 3a had an internal event CDF of 4.7E-05. However, there was significant conservative binning of the human actions impacted on dual unit trips, which was not addressed in RISKMAN. The binning in the CAFTA Revision 4 model is much more realistic. Internal Events CDF in CAFTA is 2.3E-05 (including high winds events) at IE-12 truncation.
NOTE: The results used in this application are based on Version 3A.
The results have been cross checked with the CAFTA Revision 4 model. The CAFTA results were not used (except in the case where they were more conservative and bounding) in the base document as the change to CAFTA may be considered a major revision and a gap assessment and Peer Review for ASME is needed to complete the revision for License Application purposes.
8
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Revision 1 Model Improvements Improved Human Action Methodology The current CCPRA human action methodology was refined to address several identified issues. The base methodology remains the same. These issues included:
V Better Treatment of Peer Checking
" Better Treatment of Degradation due to entry into the Functional Recovery EOP and for fire conditions V
Clarified Performance Shaping Factors Increased Credit for OC DG on Safety Related DG Failures and maintenance More realistic modeling of OC DG use for recovery upon failure or unavailability of a SR DG.
Incorporated PORV Logic Modification This modification prevents the failure of two 120VAC vital busses from causing a spurious opening of the PORVs and removes the 125VDC dependency to open the PORV for once-through-core-cooling (OTCC). The modification also added a key-locked switch to the control panel to enable the initiating of OTCC. Top Event PO was added as a plant-condition flag to represent the installation status of this modification.
The modification is installed in Unit-I (2002 Refueling Outage) and Unit-2 (2003 Refueling Outage). A second PORV top event (Top Event PP) was also added to better address the status of whether both PORVs are open to support pressure relief or decay heat removal.
Improved Multiple SGTR Modeling Improved multiple SGTR modeling per the recommendation of the Peer Review team. As long as the steam generators are isolable (Top Event SQ), multiple tube ruptures (Top Event WS) are considered recoverable. This improvement addresses the peer review F&O "AS-Ol." See Section 6.1.11.2 Peer Review "B" Level Fact and Observations for a discussion and resolution of this issue.
Improved Containment Modeling Restructured top events and initiating events such that containment breaches greater than two inches are considered "large."
This change impacts penetrations less than two inches (Top Event SH),
greater than two inches (Top Event SI), the Hydrogen Purge line (Top Event SG) and the containment sump line (Top Event SR). The definition of LOCA initiating events less than two inches (VV 1) and greater than two inches (1VV2) also changed.
These changes make the CCPRA LERF modeling more consistent with industry practice.
Improved AFW Flow Control and Make-up Modeling The CCPRA Revision 0 model addressed all AFW flow control conditions using a half dozen split fractions. The human actions used to represent these conditions addressed both early and late flow control conditions, local and remote control and the status of indication. The approach to modeling the AFW flow control was revised to better address the impact of these various conditions. This 9
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions approach allowed for the better matching of the interviewed action with the accident scenario. The new human action methodology was used for these actions.
The CCPRA Revision 1 model also improves the link between steam generator overfill recovery actions and the AFW flow path top event (Top Event Fl). If the operator fails to align AFW Pump 23 to Unit-I (Top Event OB) or to start AFW Pump 13 (Top Event FH) then it is unlikely that he will be able to perform the complicated actions involved with recovering from a steam generator overfill.
Improved Long-Term Condensate Functionality The human actions associated with long-term condensate inventory (Top Event F3) were re-interviewed and restructured. The interview process discovered a built-in peer checking process for dual unit trips, like LOOPs, in that both units monitor the shared condensate storage tank. Other insights were also gained during the interview process. This resulted in a general improvement in the modeling of the maintenance of long-term availability.
Improved ESFAS Actuation Modeling ESFAS actuation channel 15V/28V power supply tops (Top Events JI and J2) were added. These power supplies were previously modeled within each ESFAS model top event. The new top events allow the model to effectively account for the potential of a common mode failure of all ESFAS top events.
Improved HPSI Pump Modeling Top Events HD, HE & HJ now represent the Human Action to throttle HPSI flow (for 11, 12 and 13 HPSI pumps). Failure to throttle results in HPSI Pump failure after the start of recirculation.
In Revision 0, the failure of these human actions was considered a short-term failure of the HPSI pumps.
In Revision 1, this is considered to fail the HPSI pumps only after the start of re-circulation, which occurs later in the accident. This is primarily a Level 2 impact.
Updated High Wind Model On April 28th, 2002 a significant tornado struck Southern Maryland. Weather experts determined that the tornado briefly reached F4 in La Plata, Maryland. The tornado continued across Charles and Calvert Counties passing north of CCNPP. The high wind analysis did not consider the probability of F4 or F5 tornadoes. The High Wind initiating frequencies were updated to reflect the recent tornado experience.
In addition, the model structure was changed to reflect the impact associated with a larger footprint tornado.
6.1.11.5 CCNPP Unit-2 Revision I Model
Background
Construction began on CCNPP, Units I and 2 in July 1969. The Nuclear Steam Supply Systems (NSSS) for both units are supplied by Combustion Engineering.
The NSSS encompasses the Reactor Vessel, Steam Generators, Reactor Coolant Pumps, Pressurizer and Emergency Core Cooling Systems.
Bechtel Corporation designed the balance-of-plant systems and was 10
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions responsible for overall construction. Unit-I began commercial operation in May 1975 and Unit-2 on April 1977.
The Main Turbine Generator and the Steam Generator Feed Pumps for each unit are supplied by different vendors, General Electric Company for Unit-I and Westinghouse Electric Corporation for Unit-2. Although these components are different, they have almost exactly the same technical parameters and accomplish the same function.
A key difference between the units occurred with the addition of two new diesel generators. The IA DG and OC DG, added in the late 1990's, are both air-cooled diesel generators. The three other site diesels are Service Water (SRW) cooled. DG IA supports a Unit-I bus while DG OC is a station blackout (SBO) diesel that can be aligned to any Unit-1 or Unit-2 4kV ESF bus. This configuration resulted in Unit-i having one SRW cooled DG and one air-cooled DG. Unit-2 has two SRW cooled DGs.
PRA Model The CCNPP Unit-2 Revision I model is the first Unit-2 model that quantifies both internal and external events. Note that the difference in the DG configuration, including the addition of the OC DG, occurred after the IPE Submittal.
The Unit-2 model starts with the Unit-I model. Differences between the units were then identified and incorporated as changes to the Unit-I event-tree rules to create the Unit-2 event-tree rules. Over two hundred differences are modeled between the Unit-I and Unit-2 models. A simplified approach was used in the development of the Unit-2 fire model in that cable routing between the units was generally considered the same.
Two significant differences between the units are described below:
Diesel Generator Configuration Although the Unit-i and Unit-2 models contain the functions of all four DGs and the OC DG, the alignment of these diesels from the individual unit perspective is different. Therefore, the Unit-2 model reflects that the dedicated. DGs are service-water dependent. The Unit-I model has one air-cooled DG (IA DG) and one SRW dependent DG.
Turbine Building Service Water Configuration Both units have an interface between the safety-related SRW headers located in the Auxiliary Building and the non-safety-related SRW sub-system located in the Turbine Building (Turbine Building SRW). In both units, the Auxiliary Building headers supply cooling to the Turbine Building loads. Two redundant sets of isolation valves are able to isolate the Turbine Building loads from the safety-related SRW headers on a SIAS condition.
In Unit-I, the two SRW Headers remain separate after entering the Turbine Building until just prior to returning to the Auxiliary Building. The instrument air compressors are the critical components cooled by SRW Header 11.
SRW Header 12., provides cooling to the steam generator feeder pump and condensate booster pump lube oil coolers. Therefore, the loss of SRW Header 11 will result in the loss of the instrument air compressors.
11
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions The Unit-2 SRW Headers combine into a common header immediately after entering the Turbine Building. A loss of a single SRW header does not result in the loss of the instrument air compressors.
Revisions 2 through 3a Changes Noted below are the improvements included in the CCPRA RISKMAN model through Revision 3a. Note that the WOG's (Westinghouse Owner's Group) Mitigating Systems Performance Index Cross Comparison (Reference 4) was performed with Revision 2 of the plant model. The Calvert values were within a factor of three of the median value for each MSPI component importance criteria evaluated.
Updated RCP Seal LOCA model to WCAP-16175 Implemented the NRC approved RCP Seal Model documented in WCAP-16175.
Impact of dual unit trips on Human Actions.
The Peer Review Finding HR-03 (noted previously) discussed the impact of the status of the opposite unit. Human actions which were considered to be impacted by this status were adjusted on initiators which included a trip of the opposite unit.
Update of Common Cause Factors to NRC published values Relevant common cause factors from NUREG/CR-5497 or updates from INEL were implemented into the plant model More realistic modeling of Switchgear HVAC Updated room heat up rates were the basis for improved and more realistic modeling of impacts of SWGR HVAC failures and available times for recovery.
CAFTA Revision 4 The Calvert Internal Events and High Winds models were converted into CAFTA software. A detailed review of the model included the following:
Comparison of the fault tree logic with the current RISKMAN Event Trees Detailed cutest review of dominant cutsets for each initiator. This included determining why there were differences and documenting a model change or ajustification that the CAFTA model was correct.
" A comparison of MSPI component FV's in the MSPI Basis Document and CAFTA results Validation and Verification of component RAWs in EOOS. This included a comparison with previous model results for on-line risk assessment in the QSS Evaluator (RISKMAN Cutsets).
All deltas were justified or plant model changes made.
" As part of the PRA model update in 2009, the gap analysis will be updated to reflect pertinent changes to both the PRA Standard and Regulatory Guide 1.200.
12
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Other Relevant CCPRA Open Items Issues requiring action are entered into the CCPRA Configuration Risk Management Program (CRMP) database as a CRMP Issue. These maintain the current list of issues where there are gaps that require closure. Issues are prioritized as to their potential impact on the calculated risk as follows:
A Potential changes of five percent or more to CDF or LERF B
Potential changes of one percent or more to CDF or LERF C
Potential changes that enhance or have limited sequence impact D
Documentation issues The open CRMP Issues were reviewed to identify those that could have a potential impact on the proposed change to the diesel required action completion time extension.
The CRMP issues which remain open and could impact this analysis are discussed below.
Open "A" CRMP Issues CRMP 285 Provide a Basis for the likelihood of a Multi-Tube SGTR This modeling was revised per recommendations from the peer review to reduce the likelihood of multiple tube failing on a SGTR initiating event. This is based on the historical data from SGTR's in the industry. No other detailed information is available at this time on multi-tube failures.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: No impact as the issue addressed in the model used in application.
CRMP 586 Add Improve the Model of Diesel Generator (DG) Operation without Cooling The Emergency Diesel Generators are assumed to fail in less then 10 minutes on a loss of cooling water. Thus, on Spurious Safety Systems actuations the DGs with external cooling water supports (1B, 2A, 2B) are failed. There is some uncertainty on the timing of these failing and credit could potentially be taken a recovery of cooling prior to failure. However, the detailed evaluation is not complete to determine the feasibility for crediting recoveries.
N-716 Evaluation of CCNPP Units 1 and 2 Impact: This may make some of the results for CCDPs and Floods slightly conservative (only if the DGs can run for longer then 10 minutes without cooling)
Relevant "B" CRMP Issues CRMP 505 Investigate improving the HPSI injection header MOV throttling success criteria CCNPP currently assumes that if on a RAS any HPSI header MOV fails to throttle, or Operations fails to throttle these MOVs, that the HPSI pump will fail due to lack of NPSH.
This is a conservative assumption based on EOP procedural steps.
13
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions N-716 Evaluation of CCNPP Units 1 and 2 Impact:
If the HPSI pump would not fail in these cases the CCDP for all LOCAs would decrease.
Of the remaining gaps, those pertaining to the internal flooding analysis are currently being addressed via a flood model upgrade that is expected to be completed by July of 2009. Those pertaining to the LERF analysis will be addressed during the upgrade for all internal events in 2009. The other remaining gaps will be reviewed for consideration during the 2009 model update process but are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. The remaining gaps are documented in the CRMP database so that they can be tracked and their potential impacts accounted for in applications where appropriate. shows the gaps relevant to Appendix B of Regulatory Guide 1.200, Revision 1
[Reference 3] when comparing to the ASME PRA Standard [Reference 6] as well as significant gaps in the CRMP database.
PRA Configuration Control The CCPRA configuration is procedurally controlled by CNG-CM-1.01-3003, "PRA Configuration Control". This procedure provides the control and processes for maintaining the CCPRA consistent with the as-operated, as-built plant.
Design changes and procedure changes are monitored for impact on the PRA. Issues requiring action are entered into the CCPRA Configuration Risk Management Program (CRMP) database as a CRMP Issue.
These issues are prioritized in accordance with their significance for implementation into future PRA updates. Significant CRMP Issues that are the result of errors are entered into the site corrective action program.
CNG-CM-1.01-3004, PRA Process for Internal Evaluations provides the guidance for documentation of RG 1.200 compliance.
This includes all internal events (including Flood Modeling) and Fire PRA compliance documentation.
General Conclusion Regarding PRA Capability for Risk-Informed In-Service Inspection The CEG PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in this risk-informed process. In the risk-informed inservice inspection (RI-ISI) program at CCNPP, the EPRI Risk Informed ISI methodology (Reference 8) is used to define alternative inservice inspection requirements. Plant-specific PRA-derived risk significance information is used during the RI-ISI plan development to support the consequence assessment, risk ranking, element selection and delta risk evaluation steps.
The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental components of the EPRI methodology. First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RI-ISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results. Further, the LSS classification were conservatively binned as High Risk. None of the Medium Risk break locations challenged the High classification (Highest was 1.3E-05).
14
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions Consequence Results Binning Groups
.Consequence Category CCDP Range CLERP Range High CCDP > 1 E-4 CLERP >.lE-5 Medium 1E-6 < CCDP < 1E-4 IE-7 < CLERP < 1E-5 Low CCDP < 1E-6 CLERP < 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result. Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology. Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.
Secondly, the impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix.
CONSEQUENCES OF PIPE RUPTURE POTENTIAL FOR IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY PIPE RUPTURE ANDLARGE EARLYRELEASE PROBABILITY PER DEGRADATION MECHANISM SCREENING CRITERIA NONE LOW MEDIUM HIGH HIGH LOW HIGHfNI H IIGHI.
FLOW ACCELERATED CORROSION Category 7 Category 5 Caery3 Cate-gry I MEDIUM LOW LOW MEDIUM; 111H.1 OTHER DEGRADATION MECHANISMS Category 7 Category 6 Ciategory 5,ý (tel'ory 2
LOW LOW LOW LOW MIEDIUMNI NO DEGRADATIONMECHANISMS Category 7 Category 7 Category 6
-Category 4j[
Thirdly, the EPRI RI-ISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspection population rather than masking other important components.
That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population. These three facets of the methodology reduce the importance and influence of PRA on the final list of candidate welds.
15
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions The limited manner of PRA involvement in the RI-ISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174.
Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:
"There are, however, some applications that, because of the nature of the proposed change, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.
An example is risk-informed inservice inspection (RI-ISI). In this application, risk significance was used as one criterion for selecting pipe segments to be periodically examined for cracking.
During the staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary. Therefore, the staff review of plant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability."
Conclusion Regarding PRA Capability for Risk-Informed ISI The Calvert PRA models continue to be suitable for use in the risk informed in-service inspection application. This conclusion is based on:
The PRA maintenance and update processes in place, The PRA technical capability evaluations that have been performed and are being planned, and The RI-ISI process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RI-ISI process to PRA attribute capability beyond ASME PRA Standard Capability Category I.
As the PRA analyses continues to be improved during the 10-year interval these results will be reviewed to determine which, if any, would merit RI-ISI-specific sensitivity studies.
PRA Quality References
- 1. NEI-00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," Rev. A3.
- 2.
Deleted
- 3.
U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 1, January 2007.
4, WCAP-16464-NP, "Westinghouse Owner's Group Mitigating Systems Performance Index Cross Comparison," Revision 0, August 2005.
5, WCAP-15801, Revision 0, "Calvert Cliffs Units 1 and 2 PRA Peer Review Report", April 2002.
6, American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sb-2005, New York, New York, December 2005.
7, Deleted
- 8.
Revised Risk-Informed Inservice Inspection Evaluation Procedure, EPRI TR-1 12657, Revision B-A, December 1999.
9, Deleted 16
Enclosure (1)
Summary.Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions ATTACHMENT 1 -Status of Open Gaps to Capability Category II of the ASME PRA Standard' L Jit~ j7 7~
~
j~
4p~ijc~ejSR~I ~uriieht Status /c m L C
~
uniintoentII7 IE-0 The CCPRA uses a draft reference for the bases of its interfacing LOCA initiating events. A CCPRA open item (CRMP 199) had been previously issued in June 2000 to require an update of this analysis.
The frequencies used for these initiators are consistent with typical industry values so this should not have a significant impact.
IE-C12 As stated by the reviewers, this issue is not expected to cause any significant impact. The completion of the documentation for the interfacing LOCA has been prioritized after the other more significant issues. The issue had been previously captured in our configuration control program as open item (CRMP 199). Reviews of the current analysis show it to be reasonable and that it meets the general criteria for Category I. This analysis will be updated in the update of ASME compliance for the CAFTA Fault Trees developed. This is scheduled in 2009.
The results appear reasonable when compared with NRC guidance in NUREG/CR-5750 and there is no indication that updates will change the conclusions on risk categories for piping that is related to potential ISLOCAs. However, CEG will review the results of any updates to determine if any changes are required in piping risk categories.
CRMP 586 Add Improve the Model of Diesel Generator (DG)
SY-A 17, SY-A 19, Operation without Cooling SY-B7 The Emergency Diesel Generators are assumed to fail in less then 10 minutes on a loss of cooling water.
Thus, on Spurious Safety Systems actuations the DGs with external cooling water supports (IB, 2A, 2B) are failed. There is some uncertainty on the timing of these failing and credit could potentially be taken a recovery of cooling prior to failure.
However, the detailed evaluation is not complete to determine the feasibility for crediting recoveries.
This may make some of the results for CCDPs and Floods slightly conservative (only if the DGs can run for longer then 10 minutes without cooling)
The gap analysis is conducted independently of RI-ISI and is based on comparing the PRA model against the supporting requirements of ASME PRA standard at capability category II. Many of the identified gaps are not applicable to RI-ISI since in general capability category I is sufficient. For completeness, all current gaps are identified in.
17
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use in Risk-Informed Inservice Inspection Program Licensing Actions
..1Title, *DqescriptLon of-Gap, Applicab*ieSR.,..CuiStat us nAetphcaImpoSRsrtaciie Rt1-iS 7
CRMP Investigate improving the HPSI injection header SY-A20, SY-B7 CCNPP currently assumes that if on If the HPSI pump would not fail in 505 MOV throttling success criteria, a RAS any HPSI header MOV fails these cases the CCDP for all LOCAs to throttle, or Operations fails to would decrease. Potential minor throttle these MOVs, that the HPSI decrease in Flood CCDPs.
pump will fail due to lack of NPSH.
This is a conservative assumption based on EOP procedural steps.
CRMP Interview Plant Operations Personnel on potential IE-A6 Documentation of Operations Additional general transient events 746 Initiating Events interviews for potential IE's is not will not impact the N-716 code case I_________________________________
___________available or not performed.
analysis._
CRMP Documentation of Uncertainty Analyses IE-D3, AS-C3, Documentation of Uncertainty Will not directly impact the N-716 747 SC-C3, SY-C3, analysis requires upgrade in IE's and analysis outside of insights on inputs HR-13, DA-E3, Quantification to meet Category II and assumptions and parametric and IF-F3, QU-E1, epistemic uncertainty.
QU-E2, QU-E3, QU-E4, LE-G4 CRMP Additional thermal hydraulic analyses for accident AS-A9, HR-G4 Limited set of TH analyses require Results for CCDP could decrease as 748 sequence are needed to increase realism of accident some conservatism in accident more detailed TH cases will allow sequence results.
sequence evaluations (Success less conservative binning.
Criteria)
CRMP Accident Sequence Progression and CAFTA Model AS-C2 The current CAFTA Conversion is This may revise the CCDPs for 749 Conversion essentially a duplication of the CAFTA results, but worst case from RISKMAN Large Event Tree, Small CAFTA and RISKMAN used for this Fault Tree approach. Reviewers application.
have had concerns that this approach is not typical in the industry and difficult to verify AS. Consider revising to a Small Event Tree, Large Fault Tree approach.
CRMP Documentation of System Walkdowns with System SY-A4 Documentation and performance of Minimal change to analyses 750 Engineers and Operations updated walkdowns with engineering expected, meets Category I.
and operations is needed to meet
___ __ _ __ _ _ __Category I! of this element.
18
Enclosure (1)
Summary Statement of CCNPP PRA Model Capability for Use.in Risk-Informed Inservice Inspection Program Licensing Actions Til eci'lojfq)-p Pjeabic S$~s 0Curent Stains/ f(omment1i Thaportaace to~ RI-ISI CRMP More complete documentation on system boundaries SY-A8 Documentation issue to clarify Documentation issue only.
751 is needed boundaries between systems. As the CCNPP model has been noted by reviewers to be very detailed there are few actual model changes required due to this documentation
.....u g a e...
up g ra d e........
CRMP Flood Initiating Events are binned in overly IF-D3 Example - the Truck Bay Loading There are no impacts on code case 752 conservative groups area IE includes various system N-716, ISI, as none of the flood lEs sources that include impact of loss of exceed IE-06 in CDF. This will all of the source systems. Includes decrease CDF of resultant JEs CCW as one source - so RCP Seal developed.
LOCA challenged in each case.
CRMP Flood IE frequencies currently use pipe segment IF-D5a Current review by vendors shows Expect a decrease in flood IE 753 frequencies. More recent EPRI publications use pipe that the cases reviewed would have a frequencies. Currently the most risk length frequency development.
lower frequency using the updated significant flood lEs are methodology from EPRI. Current conservatively binned (See IE-D3 plan is to update all significant flood comment) no IE's are expected to lEs to the new methodology in the increase to CDF above IE-06 (if first two quarters of 2009.
increase at all).
CRMP Consider state of the art additional recoveries in the LE-C3, LE-C9b grade the Level II model to add lWill potentially decrease LERF 754 Level II model recent PWROG research on operator results.
actions to prevent large early release.
19