ML20236U724
| ML20236U724 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/31/1998 |
| From: | Mary Anderson, Feige E, Porter A IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC (Affiliation Not Assigned) |
| Shared Package | |
| ML20236U697 | List: |
| References | |
| CON-FIN-J-2229 INEEL-EXT-97-01, INEEL-EXT-97-01149, INEEL-EXT-97-1, INEEL-EXT-97-1149, NUDOCS 9807300399 | |
| Download: ML20236U724 (53) | |
Text
{{#Wiki_filter:.. INEEL/ EXT-97-01149 Technical Evaluation Report on the Third 10-Year Interval inservice inspection Program Plan: Northeast Nuclear Energy Company, Millstone Nuclear Power Station, Unit No. 2 Docket Number 50-336 M. T. Anderson, E. J. Feige, A. M. Porter Published March 1998 Idaho National Engineering and Environmental Laboratory Materials Physics Department Lockheed Martin Idaho Technologies Company Idaho Falls, Idaho 83415 Prepared for the Division of Engineering Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 JCN No. J2229 (Task Order TWA A22) 9807300399 980722 ! PDR ADOCK 05000336 P PDR g Ehclosure 2 a
ABSTRACT This report presents the results of the evaluation of the Millstone Nuclear Power Station, Unit 2, Third 10-YearIntervalinservice Inspection Program, Revision 2, submitted by letter dated July 2,1996, including the requests for relief frorn the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements that the licensee has determined to be impractical. The Millstone Nuclear Power Station, Unit 2, Third 10-Year IntervalInservice Inspection Program, Revision 2, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate editionladdenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system o component examination exclusion criteria, and (d) compliance with ISI-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report. 1 l 4 l I l l l l This work was funded under: U.S. Nuclear Regulatory Commission JCN No. J2229, Task Order TWA-A22 Technical Assistance in Support of the NRC Inservice inspection Program
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SUMMARY
The licensee, Northeast Nuclear Energy Company (NNECO), has prepared the Millstone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice inspection Program, Revision 2, to meet the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, except that the extent of examination of Class 1 piping welds has been determined by the 1974 Edition with Addenda through Summer 1975 as permitted by 10 CFR 50.55a(b). The third 10-year interval began December 26,1996, and will end December 25,2006. The Millstone Nuclear Power Station, Unit 2, Third 10-Year IntervalInservice Inspection Program, Revision 2, submitted in a letter dated July 2,1996, was reviewed, as were the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. As a result of this review, two requests for additional information (RAl) were prepared describing the information and/or clarification required from the licensee in order to complete the review. The licensee provided the requested information in submittals dated March 20,1997, and August 28,1997. The review of the Millstone Nuclear Power Station, Unit 2, Third 10-Year Interval Inservice Inspection Program, Revision 2, the licensee's responses to the Nuclear Regulatory Commission's RAls, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code identified no deviations from regulatory requirements or commitments in the Millstone Nuclear Power Station, Unit 2, Third 10-YearIntervalinservice Inspection Program, Revision 2, except:
- Code Case N-547 is not acceptable as written, The volumetric examinations of reactor pressure vessel closure head studs have not been scheduled in accordance with Table IWB-2412.
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CONTENTS A B ST R A C T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii S U MM A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii
- 1. I NT R O D U CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN . . . . . . . . . . . . . . . . . 3 i 2.1 Documents Evalua ted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 !
l 2.2 Compliance with Code Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 i 2.2.1_ Compliance with Applicable Code Editions . . . . . . . . . . . . . . . . . . . . . . . 3 j; 2.2.2 Acceptability of the Examination Sample . . . . . . . . . . . . . . . . . . . . . . . . 5 l 2.2.3 Exemption Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 2.2.4 Augmented Examination Commitments . . . . . . . . . . . . . . . . . . . . . . . . . . i
- 2. 3 C onclusion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6 . . . . . . .
! 3. EVALUATION OF RELIEF REQUESTS .................................7 5 3.1 Cla ss 1 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1 Reactor Pressure Vessel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.1.1.1 Request for Relief RR-89-01, Examination Category B-A, item B1.21, Reactor Pressure Vessel Circumferential Head Welds . . . . 7 3.1.1.2 Request for Relief RR-89-02, Examination Category B-A, item B1.22, Reactor Pressure Vessel Meridional Head Welds . . . . . . . . 8 3.1.1.3 Request for Relief RR 89-03, Examination Category B-A, items B1.11, B1.12, B1.21, and 81.22, Reactor Pressure Vessel Shell l Circumferential, Longitudinal, Bottom Head Circumferential and l Meridional Weld s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . S 3.1.1.4 Request for Relief RR 89-04 (March 20,1997, Revision), Examination Category' B-D, items B3.90 and B3.100, Reactor ! Pressure Vessel Nozzle-to-Shell and Nozzle inner Radius ! Examinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 3.1.1.5 Relief Request RR-89-05 (March 20,1997, Revision), Examination Category B-G-1, item B6.10, Surface Examination of the Reactor Pressure Vessel Closure Head Nuts . . . . . . . . . . . . . . . . . . . . . 12 3.1.1.6 Request for Relief RR-89-07, Examination Category B-0, item B14.10, Control Rod Drive (CRD) Housing Welds ........... 14 I
' 3.1. 2 Pre ssurizer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 5 '
3.1.2.1 Request for Relief RR-89-11, Examination Category B-D, Itam B3.120, Pressurizer Nozzle Inner Radius Examinations . . . . . . .15 3.1.3 Heat Exchangers and Steam Generators ........................'16 3.1.3.1 Request for Relief RR-89-10 (March 20,1997, Revision), Examination Category B-D, item B3.130, Steam Generator Nozzle-to-Shell Weld Examinations . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 !
-3.1.3.2 Request for Relief RR-89-12, Examination Category B-D, item j i
B3.120 Steam Generator Nozzle Inner Radius Sections . . . . . . . 20 l 3.1.3.3 Request for Relief RR 89-14 (March 20,1997, Revision), Examination Category B-B, item 82.31, Steam Generator Circumferential Head Welds . . . . . . . . . . . . . . . . . . . . . . . . . . 21 ! 3.1.4 Piping Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 a'v L_._' . _ _ _ - _ - _ _ . _ _ ' - _ _ . _ . _ _ _ _.__-._______.__.__.____________.m.__ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ __.m.__ ._ .- . _ _ _ _ __ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ .- _____m_.____ _..___-.__b
u 3.1.4.1 Relief Request RR-89-06 (March 20,1997, Revision), Examination Category B-J, items 89.11 and B9.12, Circumferential and Longitudinal Butt Welds in Piping . . . . . . . . . . . . . . . . . . . . . . . 23 3.1.5 Pump Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.1.6 Velve Pressure Boundary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 ! 3 .1. 7 G e ne ra l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 2 5 ) 3.1.7.1 Request for Relief RR-89-16 (March 20,1997, Revision), Examination Category B-E, Jtems 84.10 and 84.20, Pressure Rstaining Partial Penetration Welds in Vessels ............. 25 3.1.7.2 Request for Relief RR 89-18 (March 20,1997, Revision), Examination Category B-G 2, item B7.80, Pressure Retaining Bolting 2-inch and Less in Diameter . . . . . . . . . . . . . . . . . . . . . 26 3.2 Cla ss 2 C om ponents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 3.2.1.3 Proposed Alternative N-522, Use of Code Case N-522, Pressure Testing of Containment Penetration Piping . . . . . . . . . . . . . . . . 27 3.3 Class 3 Components . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 3.4 Pre ssure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 8 3.4.1 Class 1 Systern Preisure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.4.2 Class 2 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.4.3 Class 3 System Pressure Tests . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29 3.4.4 General .............................................29 3.4.4.1 Request for Relief RR-13, IWA-4700(a) and (b), Alternative Pressure Test Requirements For Code Class 1, Class 2, and Class 3 Systems Following Repair, Replacements, and M odi fica tions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 3.4.4.2 Proposed Alternative to use Code Case N-4161, Altemative - Pressure Test Requirements for Welded Repairs or Installation of Replacement items By Welding, Class 1, 2, and.?, Section XI, { ) Division 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 3.4.4.3 Proposed Alternative to use Code Case N-498-1, Altemative Rules for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems, Section XI, Division 1 ......................31 3.4.4.4 Request for Relief RR 89-08 (March 20,1997, Revision), IWA-5250(a)(2), Corrective Action Resulting from Leakage at Bolted I C onnections . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 3.4.4.5 Request for Relief RR-89-17 (August 28,1997, Revision), IWA-5242(a), insulation Removal For VT-2 Visual Examination Of Bolting in Borated Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 3.4.4.6 Proposed A.w. stive to use Code Case N 546, Attemative Requirements for Qualification of VT 2 Examination Personnel, Section XI, Division 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 3.5 General ..................................................38 3.5.1 Ultrasonic Examination Techniques . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 3.5.2 Exempted Cumponents ...................................38 3 . 5 . 3 O the r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 3.5.3.1 Request for Relief RR-14, IWA, lWB, IWC, and IWF-4000 (IWX-4000), Repair Procedures, IWA, lWB, IWC, and IWF-7000 (IWX-7000), Replacements . . . . . . . . . . . . . . . . . . . . . . . . . . . 38 j 3.5.3.2 Request for Relief RR-89-19, Examination of Code Class ; Snubbers ....................................... 38 t v
3.5.3.3 Request for Relief RR-89 20, Removal of Pressure Relief Valves for the Purpose of Testing ...........................39 3.5.3.4 Request for Relief RR-89 21, IWA-4000 and IWA-7000, Requirements for Repairs and Replacements . . . . . . . . . . . . . . . 39 3.5.3.5 Proposed Alternative to use Code Case N-524, Altemative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping, Section XI, Division 1 ....................40 3.5.3.6 Proposed Alternative to use Code Case N-535, Altemative Requirements for Inservice Inspection Intervals, Section XI, Division 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41
- 4. C O N C L U S I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 '
- 5. R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 l
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- l. i l TECHNICAL EVALUATION REPORT ON THE L
THIRD 10-YEAR INTERVAL 1 INSERVICE INSPECTION PROGRAM PLAN: I 1 MILLSTONE NUCLEAR POWER STATION, UNIT 2 l NORTHEAST NUCLEAR ENERGY COMPANY I t DOCKET NUMBER 50-336
- 1. INTRODUCTION i Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.b5a(g)(4) (Reference 1) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class 3 meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components (Reference 2), to the extent practical within the limitations of design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system pressure tests conducted during successive 120-month inspection intervals shall comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the ~
date 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed therein. The components (including supports) may meet requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein, and subject to Nuclear Regulatory Commission (NRC) approval. The licensee, Northeast Nuclear Energy Company (NNECO), prepared the Millstone Nuclear Power Station, Unit 2. Third 10-Year IntervalInservice Inspection Program, Revision 2, to meet the requirements of the 1989 Edition, except that the extent of examination of Class 1 piping welds has been determined by the 1974 Edition through Summer 1975 Addenda as permitted by 10 CFR 50.55a(b). The third 10-year interval began December 26,1996, and ends December 25,2006. As required by 10 CFR 50.55a(g)(5),if the licensee determines that certain Code examination requirements are impractical and requests relief from them, the licensee shall submit information and justification to the NRC to support that determination. Pursuant to 10 CFR 50.55a(g)(6), the NRC will evaluate the licensee's d' determination that Code requirements are impractical to implement. The NRC may grant relief and may impose alternati , requirements that are determined to be authorized by law, will not l endanger life, proprty, or the common defense and security, and are 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. Alternatively, pursuant to 10 CFR 50.55a(a)(3), the NRC will evaluate the licensee's' determination that either (i) the proposed alternatives provide an acceptable level of quality and safety, or (ii) Code compliance would result in hardship or unusual difficulty without a 1
c compensating increase in safety. Proposed alternatives may be used when authorized by the NRC. The information in the Millstone Nuclear Power Station, Unit 2, Third 10-Year Interval inservice Inspection Program, Revision 2 (Reference 3), submitted by letter dated July 2, 1996, was reviewed, including the requests for relief from the ASME Code Section XI requirements that the licensee has determined to be impractical. The review of the inservice inspection (ISI) Program Plan was performed using the Standard Review Plans of l NUREG-0800_(Reference 4) Section 5.2.4, " Reactor Coolant Boundary inservice l Inspections and Testing," and Section 6.6, " Inservice inspection of Class 2 and 3 Components." In a letter dated January 21,1997, (Reference 5) the NRC requested additional
- information that was required to complete the review of the ISI Program Plan. The requested information was provided by the licensee in the " Response to Request for Information Related to the inservice inspection Program Plan", dated March 20,1997 l (Referunce 6). In this response, NNECO provided additional documentation and l clarification regarding questions on the program.
l !- By letter dated June 11,1997 (Reference 7), the NRC requested additional information j regarding the licensee's March 20,1997, submittal. The requested information was
. provided by the licensee in a letter dated August 28,1997 (Reference 8). The licensee submitted an additional request for relief by letter dated October 31,1997 (Reference 9).
, The Millstone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice Inspection l Program, Revision 2 is evaluated in Section 2 of this report. The ISI Program Plan is [ evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sample, (c) correctness of the application of system or ! component examination exclusion criteria, and (d) compliance with ISI-relateo ! commitments identified during the NRC's previous reviews. The requests for relief are evaluated in Section 3 of this report. Unless otherwise stated, references to the Code refer to the ASME Code, Section XI,1989 Edition. Specific l inservice test (IST) programs for pumps and valves are being evaluated in other reports. I i 2 . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ . - - -i
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- 2. EVALUATION OF INSERVICE INSPECTION PROGRAM PLAN This evaluation consists of a review of the applicable program documents to determine whether or not they are in compliance with the Code requirements and any previous license conditions pertinent to ISI activities. This section describes the submittals i reviewed and the results of the review.
2.1 Documents Evaluated Review has been completed on the following information from the licensee: Millstone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice Inspection Program, Revision 2, dated July 2,1996 (Reference 3). Response to Request for Additional information dated March 20,1997 (Reference 6). Response to second Request for AdditionalInformation dated August 28,1997 (Reference 8). Letter, dated October 31,1997, containing request for authorization to use Code Case N-522 (Reference 9). 2.2 Compliance with Code Requirements 2.2.1 Compliance with Applicable Code Editions The inservice inspection program plan shall be based on the Code editions defined in 10 CFR 50.55a(g)(4) and 10 CFR 50.55a(b). Based on the starting date of December 26, 1996, the Code applicable to the third interval ISI program is the 1989 Edition. As stated in Section 1 of this report, the licensee has prepared the Millstone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice Inspection Program, Revision 2 to meet the requirements of 1989 Edition of the Code, except that the extent of examination of Class 1 piping welds has been determined by the 1974 Edition through Summer 1975 Addenda, as permitted by 10 CFR 50.55a(b). In accordance with 10 CFR 50.55a(c)(3),10 CFR 50.55a(d)(2), and
. 10 CFR 50.55a(e)(2), ASME Code Cases may be used as alternatives to Code requirements. Code Cases that the NRC has approved for use are listed in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability (Reference 10), with any additional conditions the NRC may have imposed. When used, these Code cases must be implemented in their entirety. Published Code Cases awaiting approval and subsequent listing in Regulatory Guide 1.147 may b adopted only if the licensee requests, and the NRC authorizes, their use on a case-by-case basis.
The licensee's third 10-year ISI program includes the Code Cases listed below. With the exception of Code Case N 547, these Code Cases either have been approved for use in Regulatory Guide 1.147 or are included as requests for relief. 3
- Code Case N-401-1 Eddy Current Examination l
Code Case N-402-1 Eddy Current Calibration Standards Code Case N-416-1 Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding, Class 1, 2, and 3 (Evaluated in Section 3.4.4.2 of this report) Code Case N-432 Repair Welding Using Automatic or Machine Gas Tungsten-Arc Welding (GTAW) Temperbead Technique Code Case N-435-1 Alternative Examination Requirements for Vessels with Wall Thickness 2 in. or less Code Case N-457 Qualification Specimen Notch Location for Ultrasonic Examination of bolts and studs Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds Code Case N-461 Alternative Rules for Piping Calibration Block Thickness Code Case N-463-1 Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping That Exceed the Acceptance Standards of IWB-3514.2 Code Case N-471 Acoustic Emission for Successive Inspections Code Case N-481 Alternative Examination Requirements for Cast Austenitic Pump Casing Code Case N-498-1 Alternative Rules for 10-Year Systems Hydrostatic Testing for Class 1, 2, and 3 Systems (Evaluated in Section 3.4.4.3 of this report) Code Case N 508-1 Rotation of Snubbers and Pressure Relief Valves for the Purpose of Testing (Use of this Code Case is considered part of the Inservice Test (IST) Program and is, therefore, not included in this evaluation. This request for relief will be evaluated elsewhere by the Mechanical Engineering Branch of the NRC.) Code Case N-522 Pressure Testing of Containment Penetration Piping (Evaluated in Sec' tion 3.2.1.1 of this report) Code Case N-524 Alternative Examination Requirements for Longitudinal Welds in l Class 1 and 2 Piping (Evaluated in Sectico 3.5.3.5 of this report) l l Code Case N-535 Alternative Requirements for Inservice inspection Intervais (Evaluated in Section 3.5.3.6 of this report) Code Case N-546 Alternative Requirements for Qualification of VT-2, Examination Personnel (Evaluated in Section 3.4.4.6 of this report) 4
l Code Case N-547 Alternative Examination Requirements for Pressure Retaining Bolting in Control Rod Drive (CRD) Housings - Code Case N-547 is not acceptable as written. A similar alternative is evaluated in Section t 3.1.7.2 of this report. l 1 1 2.2.2 Acceptability of the Examination Sample Inservice volumetric, surface, and visual examinations shall be performed on ASME l ' Code Class 1,2, and 3 components and their supports using sampling schedules described in Section XI of the ASME Code and 10 CFR 50.55a(b). The sample size and weld ' selection have been implemented in accordance with the Code and 10 CFR 50.55a(b) and appear to be correct except for the volumetric examination of reactor pressure vessel (RPV) closure head studs, required by Examination Category B-G-1, item B6.20. Under Examination Category B-G-1, Note 1,' the examinations may be performed in-place or with ! the studs removed. Considering the latitude provided by the Code, these volumetric examinations should not be deferred until the end of the interval; a sample of the studs should be examined each period in accordance with Table IWB-2412-1. 2.2.3 Exemption Criteria The criteria used to exempt components from examination shall be consistent with Paragraphs IWB 1220, IWC-1220, IWC-1230, IWD-1220, and 10 CFR 50.55a(b). The exemption criteria have been applied by the licensee in accordance with the Code, as discussed in the ISI Program Plan, and appear to be correct. 1 l 2.2.4 Augmented Examination Commitments In addition to the requirements specified in Section XI of the ASME Code, the licensee has committed to perform the following augmented examinations: Examination of the reactor coolant pump flywheels in accordance with the requirements of Regulatory Guide 1.14, Revision 1 (Reference 11). Examination of the reactor pressure vessel in accordance with Regulatory Guide i 1.150 (Reference 12). l Visual examination at the end of the interval of two repaired reactor vessel core barrel support lugs. 4 Removal and examination of reactor vessel materialirradiation surveillance ! specimens for changes in material properties per Technical Specifications. Volumetric examination of an additional sample of 7.5% of thin-walled Class 2 l- piping welds in Examination Category C-F-1 distributed among the High Pressure Safety injection, Shutdown Cooling and Charging systems. An Erosion / Corrosion Program implemented in response to Generic Letter 89-08 (Reference 13) that monitors erosion / corrosion for both single-phase and two-phase flow systems, 5 - t
o e Repair and/or replacement activities for ASME Subsections IWE and IWL in ( accordance with the 1992 Edition and 1992 Addenda of Section XI. The one-time, augmented reactor pressure vessel examination required by 10 CFR 50.55a(g)(6)(ii)(A), was performed during the previous 10-year interval and, l therefore, is not required during the third 10-year ISI interval. I j f 2.3 Conclusions Based on the review of the documents listed above, no deviations from regulatory requirements or commitments were identified in the M/// stone Nuc/ ear Power Station, Unit 2, Third 10-YearintervalInservice inspection Program, Revision 2, with the exception of Code Case N 547, which is not acceptable as written, and the volumetric examination of the RPV closure head studs in accordance with Examination Category B-G-1, item 86.20, which have not been scheduled in accordance with IWB-2412-1. Under Examination i Category B-G-1, Note 1, the examinations may be performed in-place or with the studs removed. Considering the latitude provided by the Code, the volumetric examinations should not be deferred until the end of the interval, a sample of the studs should be examined each period in accordance with Table IWB-2412-1, l l I 4 l 6 l
- 3. EVALUATION OF RELIEF REQUESTS i
The requests for relief from the ASME Code requirements that the licensee has determined to be impractical for the third 10-year inspection interval are evaluated in the following sections. i
),
i 3.1 Class 1 Components \' 3.1.1 Reactor Pressure Vessel l 3.1.1.1 Request for Relief RR-89-01, Examination Category B-A, item B1.21, Reactor Pressure Vessel Circumferential Head Welds Code Requ/remont-Section XI, Table !WB-2500-1, Examination Category B-A, item B1.2% requires that essentially 100% of the accessible length of all reactor pressure vessel circumferential head welds be volumetrically examined as defined in Figure IWB-2500-3. L/censee's Code Relief Request-in accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric coverage of the reactor pressure vessel head Weld CHC-1. Licensee's Basis for Requesting Relief-
" Permanent obstructions due to the control rod drive mechanism housings make this weld inaccessible from the outside and inside surfaces of the Reactor Vessel Head. Alternative surface examination methods cannot be utilized for the same reasons."
Licensee's Proposed Attemative Examination-
"The integrity of this weld (CHC-1 as shown on the attached drawing #25203-29527 sheet 2') will be monitored each period utilizing a VT-2, visual examination during the system pressure test."
Evaluation-The Code requires 100% volumetric examination for the subject RPV head weld. However, this weld is surrounded by control rod drive (CRD) housings, and access for examination is restricted. As a result, the examination is impractical to perform to the extent required by the Code. Design modifications would be required to provide access for the Code-required examination. Imposition of this requirement would cause a considerable
, burden on the licensee.
The licensee proposes to monitor the subject weld by VT-2 visual examination during I pressure tests. In addition, there are other RPV welds that are receiving complete volumetric examination. Therefore, any significant patterns of degradation will be detected and reasonable assurance of structuralintegrity will be provided. l l Conclus/on-Examination of the subject RPV head weld is impractical because the CRD housings make it inaccessible. However, the system leakage test each period, in conjunction with the volumetric examination of other RPV welds, will detect any
- a. Drawings provided by the licensee are not included with this evaluation.
l
J significant patterns of degradation and provide reasonable assurance of structuraliritegrity. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(!). 3.1.1.2 Request for Relief RR 89-02, Examination Category B-A, item B1.22, Reactor j Pressure Vessel Meridional Head Welds i Code Requ/rement-Section XI, Table IWB-2500-1, Examination Category B-A, item B1.22 requires that essentially 100% of the accessible length of all reactor pressure vessel meridional head welds be volumetrically examined as defined in Figure IWB-2500-3.
- Lkensee's Code Refef Request-4n accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric coverage of reactor pressure
, vessel meridional head Welds CHM-1, -2, -3, -4, -5, and -6. l Licensee's Basis for Requesting ReNet (as stated)- '
" Permanent obstructions due to the control rod drive mechanism housing and R. V.
l head shroud make portions of these welds inaccessible from the outside and inside surfaces of the Reactor Vessel Head. Alternative surface examination methods l cannot be utilized for the same reasons." Licensee's Proposed Attemative Examination (as stated)-
"The accessible length of each of these welds (as shown on the attached drawing
#25203 29527 sheet 26) from the circumferential flange to head weld up to the Reactor Vessel Head shroud will receive the required volumetric exam. The remaining portions of these welds, within the confines of the shroud, will be monitored each period utilizing a VT-2, visual examination during the system l pressure test."
EveAuet/on-The Code requires 100% volumetric examination for the subject RPV head l . welds. However, access to these welds is restricted by the control rod drive housings and RPV head shroud. As a result, the volumetric examinations are impractical to perform to ' the extent required by the Code. Design modifications would be required to provide access for the Code-required examinations, imposition of this requirement would cause a considerable burden on the licensee. l The licensee can examine those portions of the welds outside the RPV head shroud. In addition, other RPV welds are receiving complete volumetric examination. These examinations,in conjunction with the periodic system pressure tests, should detect any significant pattems of degradation and provide reasonable assurance of continued structural integrity. Conclus/on--Examination of 100% of the length of RPV meridional head welds is impractical due to obstruction by CRD housings and the head shroud. It is further concluded that the examinations that can be performed will detect any significant patterns of degradation and that these examinations,in conjunction with the system leakage test, will provide reasonable assurance of structural integrity. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
- b. Drawings provided by the licensee are not included with this report.
8
r ' l~ i e 3.1.1.3 Request for Relief RR 89-03, Examination Category B A, items B1.11,81.12, B1.21, and 81.22, Reactor Pressure Vessel Shell Circumferential, Longitudinal, Bottom Head Circumferential and Meridional Wolds Code Requirement-Section XI, Table IWB-25001, Examination Category B-A, items B1.11, B1.12, B1.21, and B1.22 require that essentially 100% of the accessible length of all reactor pressure vessel shell longitudinal, circumferential, and head circumferential and i meridional welds be volumetrically examined as defined in Figure IWB-2500-1, -2, and -3, l as applicable. I i Licensee's Code Re#ef Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from the Code-required 100% volumetric examination of the reactor i pressure vessel shell longitudinal, circumferential, and head circumferential and meridional welds listed below. Componeht %$steg% M 2W J ,lQ3, lw e g t. " af;pp@g%7' s s f gp \ Q.f gg^ ,, ,gg g g g g p , HS-1 Lower shell to bottom 89 % Flow skirt and anti-rotation lug head interference BHV-1 Peel segment at 30' 45% Flow skirt interference BHV-2 Peel segment at 90* 52% Flow skirt interference BHV-3 Peel segment at 150v* 36% Flow skirt interference BHV-4 Peel segment at 210' 52% Flow skirt interference BHV-5 Peel segment at 270* 45% Flow skirt interference BHV 6 Peel segment at 330' 55% Flow skirt interference SC-2 Lower to middle shell 89 % irradiation specimen tube holder interference LSL-1 Lower shell longitudinal 77% irradiation specimen tube holder at 90* interference MSL-1 Middle shell longitudinal 55 % irradiation specimen tube holder at 90* interference USL-3 Upper shell longitudinal 88% Nozzle NS-1 integral extension at 330* interference Note: Welds listed in Request for Relief RR 89-03 as receiving essentially 100% Code coverage have not been included in this table. As confirmed by the licensee in the August 28,1997, submittal, Examination Category B-D welds originally included in this request have been excluded from evaluation. I l Licensee's Besis for Requesting Relief (as stated)- I ' Permanent obstructions located within the RV preclude a 100% volumetric exam of the weld length or essentially 100% of weld volume on several of the RV shell 9 l
l l l, welds. These obstructions are attributed to the following internally mounted components of the RV: Hot Leg Nozzle Extensions ! irradiati'on Specimen Tube Holders ! Flow Skirt l Anti-rotation Lugs l "These limitations are consistent with those experienced by other utilities with l Combustion Engineering designed RV's. Attachment 1* provides a Southwest Research
- institute report of the affective welds, the coverage limitations, weld figures, and the I
calculated percentages of UT volume for each weld examined during the Second Ten 1 l Year interval RV examination completed in 1994."
"After twelve operating cycles, the OD surface condition of the RV would not be conducive to a direct coupled UT technique without extensive surface preparation. ,
This preparation would unnecessarily expose personnel to a high radiation environment. { The radiation levels between the biological shield insulation and the RV middle shell section are estimated to be in the range of 5 to 15 R/hr. Thus full compliance to the requirements from the OD surface of the RV would result in unnecessary personnel exposure without a commensurate increase in the level of reliability, quality, or safety over the partial examinations reported in Attachment 1." Licensee's haposed Altemative Examination (as stated)-
"The examination volume and weld length percentages may be increased during the Third Ten Year RV examination due to emerging technology. However, as a minimum Millstone Unit 2 expects to obtain at least the same volumetric examination percentage as listed in attachment 1, during the Third Ten Year RV
. examination.
EveAuetion-The Code requires that the subject circumferential, longitudinal, and meridional reactor pressure vessel welds be 100% volumetrically examined during each inspection interval. The licensee has requested relief from the Code-required 100% volumetric coverage due to scanning limitations associated with the hot leg nozzle extensions, irradiation specimen tube holders, flow skirt, and anti rotational lugs. Because these items restrict scanning, it is impractical to perform the examinations to the extent required by the Code. To obtain complete volumetric coverage, design modifications would be necessary. Imposition of this requirement would result in a considerable burden on the licensee. The licensee proposes to perform the volumetric examinations to the extent practical on the subject circumferential, longitudinal, and meridional welds. The licensee estimates coverages from 36% to 89%. Based on the percent of coverage obtainable,it is concluded that any significant pattems of degradation will be detected. As a result, reasonable assurance of continued structuralintegrity will be provided. l Conclusions-Obtaining the Code-required volumetric coverage is impractical for the reactor pressure vessel welds at Millstone, Unit 2. However, the examinations that can be
- c. Attachments provided by the licensee are not included with this report.
. 10
l completed will provide reasonable assurance of continued structuralintegrity. Therefore, it i is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). 3.1.1.4 Request for Relief RR 89 04 (March 20,1997, Revision), Examination Category B D, items 83.90 and 83.100, Reactor Pressure Vessel Nozzle to-Shell and Nozzle Inner Radius Examinations Code Requirement-Section XI, Table IWB-2500-1, Examination Category B-D, items B3.90 and B3.100 require 100% volumetric examination of all reactor vessel nozzle-to-shell welds and nozzle inner radius sections each inspection interval as defined by Figure IWB-2500-7. At least 25% but not more than 50% (credited) of the nozzles shall be examined by the end of the first inspection period and the remainder by the end of the inspection interval. Licensee's ProposedNtemative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the Code requirement to examine at least 25% of the vessel-to-nozzle welds and nozzle inner radius sections during the first examination period. The proposed alternative is stated below.
" Perform the required volumetric examination on all six of the reactor vessel nozzle to vessel welds and inner radius sections during the Third Ten Year reactor vessel examination in accordance with the provisions of Code Case N-521."
Licensee's Basis for the Proposed Alismative (as stated)-
"During the last refueling outage of the Second Intervalin 1994, we examined all six reactor vessel nozzle to shell welds and their associated inner radius sections with no rejectable indications noted. These examinations were parformed during our second Ten Year Reactor Vessel examination. Perferming the required examination on two of the Category B-D nozzle welds during the First Period of the Third Ten Year interval would result in unnecessary personnel exposure with no commensurate increase in the level of reliability, quality, or safety over the nozzle examinations performed last refueling outage.
"As an alternative to the existing requirements, Code Case N-521, permits the deferial of nozzle-to-vessel welds and inside radius sections to the end of the inspection interval.
"No inservice repairs or replacements by welding have ever been performed on any of the Nozzle to Vessel Welds, inside Radius Sections, or Nozzle to Safe End Welds.
"None of the Nozzle to Vessel Welds,inside Radius Sections, or Nozzle to Safe End Welds contains identified flaws or relevant conditions that currently require successive inspections in accordance with IWB 2420(b)."
"MP 2, is not in the first inspection interval.
" Full compliance to the requirements of Table IWB 25001, Code items B3.90 and '
B3.100, would result in unnecessary personnel exposure and increase the risk associated with placing the volumetric examination delivery tool onto the reactor vessel 11
O flange without a commensurate increase in the level of reliability or safety over the proposed alternative examination schedule." Evaduet/on-Tha Code requires the examination of at least 2.5%, but not more than 50% of RPV nozzles and associated inside radius (IR) sections and nozzle safe ends during the first inspection period. The licensee has requened to use Code Case N-521 and defer examination of these areas until the end of the third 10 year interval. Code Case N-521 states that the examination of RPV nozzles, IR sections, and nozzle-to safe end welds may be deferred provided (a) no inservice repairs or replacements by welding have ever been performed on any of the subject areas, (b) none of the subject areas contain identified flaws or relevant conditions that currently require successive inspections in accordance with IWB-2420(b), and (c) the unit is not in the first interval. The licensee has confirmed that these conditions have been met, in addition, the licensee examined all the subject areas during the final refueling outage (the third period) of the second 10 year interval. By examining the nozzle and associated IR sections and nozzle-to-safe end welds at the end of the previous 10 year interval, the licensee has established a new sequence of examinations and will not exceed 10 years between examinations. Because NNECO is meeting the conditions in the Code Case and has repeated the examinations at the end of the previous interval, the proposed alternative will provide an acceptable level of quality and safety. ConcAusion-The licenses has met all the conditions stated in the Code Case and has examined all of the affected areas at the end of the previous interval, thereby establishing a new sequence of examinations. Since the time between examinations will not exceed ~ 10-years, the licensee's proposed alternative will provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N 521 should be authorized for the third 10-year interval at Millstone, Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use the Code Case with the limitations, if any, listed in Regulatory Guide 1.147. 3.1.1.5 Relief Request RR 8945 (March 20,1997, Revision), Examination Category B-G 1, item B6.10, Surface Examination of the Reactor Pressure Vessel Closure Head Nuts Code #etpuirement-Section XI, Table IWB-2500-1, Examination Category B-G-1, item B6.10 requires 100% surface examination of all reactor vessel closure head nuts. Licensee's Proposed Attemative (as stated)-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposes to perform s VT-1 visual examination in lieu of the Code-required surfece examination of the subject reactor vessel closure head nuts as cpecified in Table IWB 25001 of the 1989 Edition of ASME Section Xi. Licensee's Basis for the Proposed Alterative (as stated)- L "Later Addenda and Editions of the ASME B&PV Code, (1989 Addenda through the 1992 Edition) has changed the examination technique from a surface examination to a visual (VT-1) examination. Extensive surface preparation and cleaning of these reactor vessel closure nuts is required prior to performing an acceptable surface examination. I 12 l c _
The cleaning material and wet magnetic particle test solution results in ' mixed waste', i.e., contaminated and hazardous material that can not be economically disposed of. This extensive cleaning and surface examination technique has resulted in additional cost, inefficient use of available manpower resources, and are not consistent with the
'ALARA' program at Millstone Unit 2. NNECO believes that continuing to perform a -
surface examination on these closure head nuts will not provide any potential increase in plant safety margins and the additional costs of these examinations are no longer warranted based on later Addenda and Editions of the Code that changed the examination requirements from a surface exam to a visual (VT-1) examination.
" Additionally, a similar relief to perform a visual (VT-1) examination in lieu of the surface exam has boon granted to both Millstone Unit 1 and Connecticut Yankee Atomic Power Company by letter dated April 24,1994 and February 19,1992, respectively."
Evahaetton-Review of the examination requirerreents for Examination Category B-G 1 indicates that, with the exception of the reactor pressure vessel closure head nuts and the closure studs (when removed), items in this Examination Category require VT-1 visual examinations or volumetric examinations (as applicable). Typical relevant conditions that would require corrective action prior to putting closure head nuts back into service would include corrosion, deformed or sheared threads, deformation, and degradation mechanisms 3 (i.e., boric acid attack). The applicable Code examination requirement for closure head l nuts is a surface examination. Surface examination procedures are typically qualified for the detection of linear type flaws (cracks) and only have acceptance criteria for rejectable linear flaw lengths. When performing surface examinations in accordance with the 1989 Edition of the Code, item B6.10, the surface examination acceptance criteria are not provided, as they were in the course of preparation. Without clea.ly defined acceptance criteria, relevant conditions that require corrective measures may not be adequately addressed. The 1989 Addenda of Section XI, Article IWB-3000, Acceptance Standards, lWB-3517.1, Visua/ Examination, VT-1, describes relevant conditions that require corrective action prior to continued service of bolting and associated nuts. , included for corrective action in lWB-3517.1 is the requirement to compare crack-like flaws to the flaw standards of IWB-3515 for acceptance. Surface examination acceptance criteria are typically limited to linear type flaws (i.e. cracking, aligned pitting and corrosion). Because the VT-1 visual examination acceptance criteria include the requirement for evaluation of crack-like indications and other relevant conditions requiring corrective action such as deformed or sheared threads, localized corrosion, deformation of part, and other degradation mechanisms, it is concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut. As a result, the INEEL staff believes that VT 1 visual examination provides an acceptable level of quality and safety. In addition, the 1989 Addenda of Section XI changes the requirement for the subject reactor pressure vessel closure head nuts from surface examination to VT-1 visual examination and provides appropriate acceptance criteria. ConcAssion--As an alternative to the Code-required surface examination of re. actor pressure - vessel closure head nuts, the licensee proposed to perform a VT-1 visual examination.
]
Based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the 1989 Addenda and later editions of the Code require only a VT-1 13
E e. visual examination on reactor pressure vessel closure head nuts, an acceptable level of quality and safety is provided. Therefore, it is recommended that the proposed altemative VT-1 visual examination be authorized pursuant to 10 CFR 50.55a(a)(3)(i). 3.1.1.6 Request for Relief RR-89-07, Examination Category B-0, item B14.10, Control Rod I Drive (CRD) Housing Welds l Code Requirement-Table IWB-25001, Examination Category B-0, item B14.10 requires a i volumetric or surface examination to be performed on the welds of 10% of the peripheral CRD housings. Licensee's Code AeWe/ Aequest-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing 100% volumetric examinations on 10% of the periphery ! CRD housing welds. Licensee's Basis for Requesting ReGef (as stated)-
"There are 28 peripheral housings containing 5 welds (each) on the RV Head at Millstone Unit #2.
- 1. The lower CRD housing weld (W) is partially recessed into the Reactor Vessel Head and less than 40% of this weld can be examined by the above Code requirements. See attached drawing 25203 29527, Sheet 2 for details'.
- 2. Surface examinations would be extremely difficult to perform on these welds as the adjacent CRD housings are too close to one another to permit accessibility.
- 3. Subsequent cleaning of any surface examination residues would be questionable due to the crevice in the recessed area."
Licensee's heposed Akemative las stated)- l "The accessible welds, S, T, U, & V in the (CRD) housings will be examined in accordance with the specified Code requirements. The integrity of the inaccessible (W) welds will be verified during a system pressure test by a VT-2 visual examination." EveAustion-The Code requires 100% volumetric or surface examination of the welds in 10% of the peripheral CRD housings. However, based on review of the documentation proiided by the licensee, it has been determined that access to one of the five welds (Weld W) on each of the peripheral CRD housings is restricted, making the Code-required I examination imprac*ical. To perform the Code-required examination, design modification' I to allow accass for examination would be required. Imposition of this requirement would l cause a considerable burden on the licensee. j The licensee can meet the Code requirements for the four accessible CRD housing welds (S, T, U, and V). In addition, the CRD welds receive a VT-2 visual examination during the Class 1 system leakage test each refueling outage.~ Therefore, any patterns of l d. Drawings provided by the licensee are not included with this report. 14
t a degradation will be detected and reasonable assurance of structuralintegrity will be provided. Conclusion-Based on the evaluation above, it is concluded that CRD housing Weld W is impractical to examine at Millstone 2. It is further concluded that reasonable assurance of the continued structuralintegrity is provided by the examination of the accessible welds and the Class 1 system leakage test. Therefore,it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). 3.1.2 Pressurizer 3.1.2.1 Request for Relief RR-89-11, Examination Category B-D, item B3.120, Pressurizer Nozzle Inner Radius Examinations Code Requirement-Section XI, Table IWB-2500-1, Examination Category B-D, item B3.120 requires a 100% volumetric examination of all pressurizer nozzle inner radius sections each inspection interval as defined by Figure IWB 2500-7. Licensee's Code Relief Request--Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the Code-required volumetric examination to the extent required by the Code for the following pressurizer nozzle inner radius sections: ( Nonle? Coverage Direction? / Coverage! ! Total Coverage Spray Nozzle Axial 80.7 % 79.5 % Circumferential 100 % Safety Axial 62.8 % 81.4 % Nozzle Circumferential 100 % Surge Nozzle Axial 69.8 % 84.9 % Circumferential 100 % Licensee's Basis for Requesting Relief (as stated)-
"a. Volumetric examination of these sections is limited to the examination scans as depicted on the various nozzle inside radius figures within the above table.
"b. Volumetric Ultrasonic examination from the ID is impractical due to stainless steel cladding.
"c. Surface examination and radiography of the bottom head nozzle inside radius
, section is inaccessible. Surface examination and radiography of the top head ! nozzle inside radius sections are also impractical due to personnel risk / safety and would require access to a highly contaminated, high radiation area and would result in unnecessary exposure without a commensurate increase in the level of reliability, quality, or safety over the propose alternative examination listed below, t 15 l l L _ _ _ _. ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
a e'
"d. Volumetric examinations will be performed from the Pressurizer shell and nozzle OD to the extent practical."
Licensee's Proposed Altemative Examination (as stated)~
"The Pressurizer inside Radius sections will be examined to the maximum extent achievable using the latest ultrasonic examination techniques available during the i Third Inspection Interval. See attachment 1 [ paraphrased abovel, for calculated coverage and applicable figure'.
"The integrity of these areas will also be monitored during the required VT-2 examination associated with the system pressure test performed during the Third Ten Year Interval."
EveAmt/on-The Code requires 100% volumetric examination of the subject pressurizer j nozzle inner radius sections. However, examination of these areas to the extent required by the Code is restricted by component configuration which lim,its scanning. As a result, compliance with Code coverage requirements is impractical for these nozzle inner radius areas. To meet the Code requirements, design modifications would be necessary to provide access for examination, imposition of this requirement would cause a considerable burden on the lice ~nsee. The licensee can examine a significant portion (79-85%) of each inner radius section. Therefore, any significant patterns of degradation will be detected by the exam 5ations j that can be performed and reasonable assurance of structuralintegrity will be provided. ConcAssion-The Code examination requirements are impractical for the subject pressurizer nozzle inner radius sections. The volumetric examinations that can be completed will provide reasonable assurance of the continued structural integrity of these areas. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). ! 3.1.3 Heat Exchangers and Steam Generators 3.1.3.1 Request for Relief RR-89-10 (March 20,1997, Revision), Examination l Category B D, item 83.130, Steam Generator Nozzle-to-Shell Wald Examinations Code #eguiremant-Section XI, Table IWB-2500-1, Examination Category B-D, item-B3.130 requires 100% volumetric examination of all steam generator nozzle-to-shell welds as defined by Figure IWB-2500-7. Licensee's Code Ne#ef Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the Code-required volumetric examination of the steam generator nozzle to shell welds. The estimated coverage for each weld was provided by the licensee (Reference 14) and is listed in the table below. t
- c. Figures included in licensee's submittal are not included in this report.
16 . l
i i 1 n
'ISI Weld ID h -
? Coveraca ?
SG-1-NH-2 A 64.51 % SG-1-NH-4 A 57.43 % SG-1-NH 5 A 54.49 % SG 2-NH-2-A 64.51 % SG-2 NH-4 A 63.16 % SG 2-NH-5 A 54.58 % Licensee's Basis for Requesting ReWei(as stated)-
"a. The replacement (1992) steam generators were designed and manufactured )
utilizing a forged bottom head with an integral nozzle boss with a nozzle extension welded to that boss, see " Modified" Figure IWB-2500(d)'.
"b. There were no unacceptable indications detected during the preservice examinations. Two nozzle welds (SG-1-NH-2-A and SG-1-NH-4-A) each had indication that were recorded, evaluated as " spot" indications and found to be j acceptable to IWB-3512.
l "c. Volumetric examination of these welds is limited to the actual UT scans performed l j and the resulting percentage of weld volume examined as detailed on the attached ' Nondestructive Test Engineering " Coverage Estimate Calculation #25303-ER 0018"'. l "d. Ultrasonic volumetric and surface examinations from the ID are impractical due to the stainless steel cladding of the steam generator bottom head surfaces. These examinations if performed would also result in unnecessary personnel exposure in the range of 600 to 1225 MR/Hr. without a commensurate increase in the level of reliability.
"e. Radiographic volumetric examination is also impractical due to the expected high background radiation levels that would fog (prematurely expose) the film resulting in poor' quality (image and sensitivity) radiographs."
In the June 11,1997, RAI, the licensee was requested to discuss the basis for classification of these welds, and the possibility of performing e supplemental visual examination. In the August 28,1997, response to the NRC RAI, the licensee stated:
"in 1993, NNECO submitted an ASME. Code Inquiry IN 93-40, regarding the L '
Examination Category that should be assigned to the new designed nozzle welds of the i
- f. Modified figure provided by the licensee is not included with this report.
- g. Licensee's coverage estimate calculations not included with this report.
17
I replacement steam generators. The proposed answer to this inquiry was that the Code did not address this design. NNECO withdrew the inquiry on the basis that the responsible Working Group at ASME Section XI would develop rules to address this new design. Presently, there is a Code Case under development within the Working Group on inspection of Systems and Components that has classified these new designed welds as Examination Category B-D welds. There is sufficient information to , support the Code's conclusion at this point, that the Code's stress analysis review for ! this design and the thickness of these welds supports the Code's categorization as I Category B-D welds and so that is how we have addressed these welds within Relief Request RR-8910.
" Classifying these welds as Category B-D in lieu of Category B-J has some major
!- significance on the examination volume requirements under Section XI. The volume seguired for examination under Category B-D includes the entire weld thickness. 4 j Category B-J requires only the bottom 1/3 of the weld be volumetrically examined. A ' one sided examination is the only way these nozzle welds can be ultrasonically examined since the outside radius section of the nozzles is located too close to the weld and the material is clad on the inside diameter of the weld. Industry experience has shown that sound cannot be bounced off of this clad with meaningful results and thus limitations will be inherent when examining these welds. Millstone Unit No. 2 has . I taken the conservative approach and classified these welds as Category B-D welds since the preservice examinations were performed in 1992. The limitations that have been identified in Relief Request RR-89-10 have been directly taken from the preservice inspection results and have been determined to represent an examination that is being i l' performed to the extent practical.
"The replacement design for the new steam generators is in accordance with the
- inspectability requirements of 10 CFR 50.55a. For the design considerations of these ,
welds the following requirements were met: I
- 1. 10 CFR 50.55a(c)(1) states, ' Components which are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section ill of the ASME Boiler and Pressure Vessel Code except as provided in Paragraphs (c)(2), (c)(3) and (c)(4) of this section.'
- 2. Paragraph (c)(4) of this section states, 'For a nuclear power plant whose construction permit was issued prior to May 14,1984 the applicable Code Edition i and Addenda for a component of the reactor coolant pressure boundary continue to be that Code Edition and Addenda that were required by Commission regulators for such component at the time of issuance of the construction permit.'
l
- 3. 10 CFR 50.55a(g)(1) states, 'For a boiling water or pressurized water-cooled y nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (5) of this section to the extent practical.' This requirement applies to ;
Millstone Unit No. 2 because the construction permit was issued in December 1970.
- 4. Paragraph (g)(4) requires plants to meet the Section XI 'except design and access provisions and preservice examination requirements,' of subsequent editions, 'to l l l- 18 l
l: I
4 the extent practical within the limitations of design, geometry and materials of constructions of the components.'
- 5. Paragraph (g)(5) requires the inservice inspection program for the plant to be,
' revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section.'
" Based on these requirements and the fact that the vendor design process does not make allowances for reanalyzing and reforging its components to specifically address !
later ASME Section XI Code requirements, Millstone Unit No. 2 is in compliance with the regulations and has taken into account the limitations of these weld examinations by submitting Reliet Request RR 89-10.
" Millstone Unit No. 2 did not propose a VT-1 visual examination of the subject area of j
the nozzle interior as an alternative to the limited volumetric examination in Relief Request RR-89-10 for the following reasons: I
- 1. It is NNECO's understanding that there have been no industry failures in these nozzle areas.
- 2. At Haddam Neck Plant, for a similar relief request implementing a VT-1 visual examination, unwarranted man-rem exposure had to be expended to resolve non- l relevant indications.
- 3. Research data that has been documented shows that clad cracking may occur for I various reasons, but it is highly unlikely that it will propagate into the base metal l
or be associated with base metal underclad cracks that could lead to a component l failure. The basis for this statement can be found in the ASME PVP - Vol.28, 1994 Conference Paper, titled: ' Clad Failure Models for Underclad Flaws in Reactor Pressure Vessels,' authored by F. A. Simonen." j l Licensee's Proposed Mtemative Examination (as stated)-
"The steam generator nozzle to nozzle extension welds will be examined to the i maximum extent achievable using the latest ultrasonic techniques available during the Third Inspection interval. Specific weld coverage is expected to be equal to or greater than the percentages detailed in the attached Nondestructive Test Engineering ' Coverage Estimate Calculation #25303-ER-97-0018'.
"The integrity of these welds will also be monitored during the required VT-2 examinations associated with the system pressure test performed during the Third Ten Year Interval."
Evaluation-The subject welds are a result of steam generator replacement and are not specifically addressed by the Code. The steam generators were removed by cutting the nozzles, then rewelding them, at the heavy-walled section of each nozzle. As a result, the l licensee classified them as Examination Category B-D, item B3.130. This conclusion is supported by the Section XI Working Group on Inspection of Systems and Components, which is currently developing a Code Case to address these new weld designs. As currently classified, the Code requires 100% volumetric examination of the subject steam generator nozzle-to-vessel welds. However, complete examination is restricted by l 19 . i 1
component configuration that precludes examination from one side of the weld. Therefore, ! the Code coverage requirements are impractical to meet for these welds. To meet the ' Code requirements, the nozzles would have to be redesigned and modified to allow access for complete examination. imposition of this requirement would create an undue burden on the licensee. The licensee has examined a significant portion of the weldr,in three of the four Code-required directions. This has resulted in a cumulative coverage of 55% to 65% for each of the welds. Based on the extent of coverage obtained, the INEEL staff believes that any significant patterns of inservice degradation should be detected. Therefore, reasonable assurance of the structuralintegrity will be provided. ConcAusion--Based on the impracticality of meeting the Code coverage requirements and the reasonable assurance provided by the examinations that were completed, it is
- recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i).
l 3.1.3.2 Request for Relief RR-8912, Examination Category B D, item 83.120, Steam Generator Nozzle inner Radius Sections Code Requirement -Section XI, Table IWB-2500-1, Examination Category B-D, item B3.120 requires a 100% volumetric examination of all steam generator nozzle inner radii as defined by Figure IWB-2500-7. i licensee's Code Ma#ef #eguest-4n accordance with 10 CFR 50.55a(g)(5)(iii), the licensee requested relief from performing the Code-required volumetric examination of the following steam generator inlet and outlet nozzle inner radii. I
; yNorW% Coverage:Directionf (Cuvsregsj; iTotal Coverage:
Inlet Axial 68 % 84 % l Circumferential 100 % l Outlet Axial 88 % 94 % , Circumferential 100 % Licensee's Basis for Requesting #elief (as stated)- :
"a .' Volumetric examination of these sections is limited to the examination scans I as depicted on the inlet and outlet inside radius nozzle figure within the above I table.'
"b. Ultrasonic volumetric, examination from the ID is impractical due to the stainless steel cladding of the steam generator inside radius sections and the integral nozzle dam retaining ring. !
"c. Radiographic examination is also impractical due to the expected high (925 to l 1225 MR/Hr.) background radiation levels that would fog (prematurely expose) the ,
film resulting in poor quality (image and sensitivity) radiographs. l 20 I
Se "d. Surface examination techniques are also impractical due to the stainless steel cladding of the steam generator inside radius sections and would result in unnecessary personnel exposure (925 to 1225 MR/Hr.) without a commensurate increase in the level of reliability, safety, or quality, over the proposed alternative examinations listed below." Licensee's Proposed Attemative Examination las stated)- i "The steam generator inside radius sections will be examined to the maximum extent achievable using the latest ultrasonic techniques available during the Third inspection Interval. See attachment 1, for calculated coverage and applicable figures"." Ereduetion-The Code requires 100% volumetric examination of the steam generator inlet and outlet nozzle inner radius sections. However, examination of these areas is restricted by nozzle configuration which precludes examination to the extent required by the Code. ; Therefore, it is concluded that compliance with Code coverage requirements is impractical. l To meet the Code requirement:,, design modifications would be required to allow access for examination. Imposition of this requirement would cause a considerable burden on the ; licensee. 1 The licensee is examining a significant portion (84% and 94%) of the subject inner radius sections and inservice degradation, if occurring, should be detected. As a result, the examinations that can be completed provide reasonable assurance of structural integrity of the subject areas. ! ConcAusion-Based on the impracticality of meeting the Code coverage requirements and l the reasonable assurance provided by the examinations that were completed, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). 3.1.3.3 Request for Relief RR 89-14 (March 20,1997, Revision), Examination Category 8-8, item 82.31, Steam Generator Circumferential Head Welds Code Requirement--Section XI, Table IWB-2500-1, Examination Category B-8, item B2.31 ! requires 100% volumetric examination of one steam generator circumferential weld per ' l head as defined in Figure IWB-2500-3. Examination may be limited to one vessel among a i group of vessels performing a similar function. Licensee's Code Re#ef Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee , requested relief from the Code-required 100% volumetric coverage of the following steam l generator head welds. ; i 1
, {lSi Weld ID ? f Estimated Coveragei l SG-1-BHC-1 A 53.79% of 51% (- 27%)
SG 2 BHC-1 A 53.62% of 51 % (-27%) i
- h. Attachment provided by the licensee is not included in this report. )
i' 21 L.- . _ _ - - _ _ _ _ _ - - - - ___ _ _ - - - - - . - _ _ __ - - - - - - _ _ _ _ - - - - - - - - - _ _ _ _ _ _ - - - _ - - _ _ o
Licensee's Basis for Requesting Relief (as stated)-
"During 1992, Both Steam Generators were replaced. The Replacement Generators were designed and manufactured eliminating the original Bottom Head Meridional (Code item #B2.32), and Three Bottom Head Circumferential (Code item #82.31) welds. However, weld #SG-1-BHC 1 A and SG-2-BHC-1 A, could not be altered to assure 100 percent UT coverage due to the original reactor coolant system piping configuration.
" Partial volumetric examination of these welds is limited to a one sided exam due to the proximity of the Stay Cylinder Support Skirt Boss resulting in a 50 percent examination loss. The examination is limited further by the Hot Leg and Cold leg nozziss. The actual UT scans performed and the resulting percentages of the weld volume examined are detailed on the attached Nondes'ructive t Test Engineering
" Coverage Estimate Calculations # 25203-ER-97 0020 and Sketch #1'.
" Alternative inservice inspection radiographic examination of these welds is impractical due to the expected high background radiation levels (925 to 1225 MR/Hr.) that could fog (prematurely expose) the film resulting in poor quality (image and sensitivity) radiographs. The radiographic technique would further be impaired by geometric condition and access limitations.
"Since the inside surface of the Bottom Heads are cladded with stainless steel a meaningful surface examination cannot be performed on the root of these welds.
I "The ASME Section lil, radiographs were reviewed during the Preservice inspections to determine if any discontinuities are present that would require monitoring during inservice examinations. None are present." Licensee's Proposed Attemative Examination (as stated)-
"NNECO MP-2, will examine Weld #SG 2-BHC-1 A in generator #2. This circumferential weld will be examined to the maximum extent achievable using the latest ultrasonic techniques available during the Third Ten Year Inspection Interval.
. Specific weld coverage is expected to be equal to or greater than the rsrcentage detailed in the attached Nondestructive Test Engineering ' Coverage httmate Calculations #25203-ER-97-0020'.
"A magnetic particle examination will be performed on the OD surface of the weld.
"The integrity of these areas will also be monitored during the required VT-2 examinations associated with the system pressure test performed during the Third Ten Year Interval."
EveAustion-The Code requires 100% volumetric examination of one circumferential weld per steam generator head. However, examination of these welds is restricted by component configuration which precludes examination to the extent required by the Code. Therefore, it is concluded that compliance with Code coverage requirements is impractical. To meet the Code requirements, design modifications would be required to allow access
. Attachment provided by the licensee is not included with this report.
22
for examination. Imposition of this requirement would cause a considerable burden on the licensee. The licensee proposed to perform volumetric examination on approximately 27% of Weld SG-2-BHC 1 A In addition, a magnetic particle examination will be performed on the outside surface of this weld. However, because the coverage for this weld is low, the proposed examinations should be performed on both steam generators. The performance of these examinations on both steam generators, in conjunction with the VT 2 visual examination during system pressure testing, should detect any significant patterns of degradation. As a result, reasonable assurance of structuralintegrity will be provided. Concdusion--Based on the evaluation above, it is concluded that meeting the Code coverage requirements is impractical for the subject welds. However, due to the low coverage that can be obtained, it is further concluded that reasonable assurance of the structuralintegrity will be provided if the proposed examinations are performed on both' steam generators. Therefore, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i) provided that the proposed examinations are performed on both steam generator circumferential head Welds SG-1-BHC-1 A and SG-2-BHC-1 A. 3.1.4 Mping Pressure Boundary 3.1.4.1 Relief Request RR-89-06 (March 20,1997, Revision), Examination Category 8-J, items 89.11 and 89.12, Circumferential and Longitudinal Butt Welds in Piping Code #e_t; ant-Table IW8-2500-1, Examination Category B-J, items 89.11 and 89.12 require 100% volumetric and surface examination of circumferential and longitudinal butt welds as defined by Figure IWR-2500-8. l i Licensee's Code Re#e/ Request-Pursuant to 10 CFR 50.55a(g)(5)(iii), the licensee i requested relief from the Code-required 100% surface examination on the following welds: P-1-C-1-A P-5-C A P-3-C A P-10-C A i P-14-C-1 -A P-18-C A P-1-C-1 P-10-C-1 ! P-5 C-1 P-5-C-2 P-9-C-1 P-9-C-2 P-14-C-1 P-14-C-2 P-18-C-1 P-18 C-2 l
. l Licensee's Basis for Requesting Relief (as stated)-
" Plant design did not provide sufficient access to these welds to allow performance i of surface examinations. Surface examination was not required at the time of design or construction. The circumferential welds and the intersecting longitudinal welds are located in the annulus between the reactor vessel and the primary shield wall or within the primary shield wall. Access to these welds is completely blocked for more than 50% of the weld area by non-removable insulation. Access to the i remainder of the weld is physically hazardous and requires access to a highly contaminated, high radiation area. The area configuration is such that temporary scaffolding cannot be readily installed to permit physically safe access to the area.
Der:gn access to these welds was based on performing automated ultrasonic 23 l t
l i i examinations of the full weld volume from the outside diameter of the pipe using externally mounted scanners.
"Preservice examination experience and subsequent evaluation of this equipment have l identified that this technically outdated examination equipment is inadequate to l
perform the required surface or volumetric examinations. The tracks originally installed I to support these volumetric examinations further block access for the external surface ! examinations. The attached drawing (#25203-20146 sheet 97-1 ) shows details of the 1 access and interference configuration. l
" Relief Request #RR 10 and #RR 10, Rev.1, were submitted and approved during the Second Ten Year Interval permitting a full volumetric examination to be conducted from '
the inside diameter of the reactor coolant piping in lieu of performing the outside diameter surface examination. This technique was also satisfactorily demonstrated to be capable of detecting OD connected cracks (not notches) as required by the Safety Evaluation for the approved Relief Request #RR 10. I "During the Second Ten Year Reactor Vessel examination, Southwest Research Institute successfully performed the full volumetric ultrasonic examination from the piping inside diameter on four of the above Category B9.11 weld and their associated l Category B9.12 long seams."
- Licensee's Proposed Altemative Examination (as stated)-
"In lieu of performing the surface examination NNECO will continue to perform the previously approved (RR-10) full-volume ultrasonic examination from the intemal surface in accordance with the 'l989 Edition of ASME Section XI, Regulatory Guide 1.150 and Code Case N-524, for longitudinal welds. The ultrasonic examination
! willinclude 100% of volume 'A B-E-F' as shown on the attached copy of Figure
- IWB-2500-8. These welds will also be subjected to a system pressure test in l accordance with ASME Code Case N-498-1."
EvaJlustion-The Code requires a volumetric examination of the inner 1/3 volume and a surface examination of the outside (OD) surface for each of the subject welds. However, these welds are located in the annulus between the reactor vessel and the primary shield wall, or are located within the primary shield wall. Access is further restricted by non-removable insulation. Therefore, performance of the Code required surface examination to the extent required by the Code is impractical for these welds. Design modifications would be required to gain access for complete examination from the ou'-ide surface, imposition 6 of this requirement would cause a considerable burden on the licemee. As an altemative, ; the licensee has proposed to perform an automated ultrasonic examination from the interior surface that will examine the entire weld volume, including the outside surface of the weld.
'The proposed alternative examination was previously approved for the previous 10-year intervalin an NRC SER dated April 2,1992. To obtain relief for the previous interval, the ;
licensee demonstrated the effectiveness of the approach on OD connected cracks (not i notches). Considering the proven effectiveness of the proposed examination technique, j.' Attachment provided by the licensee is not included with this report. 24
\ , 64 the INEEL staff concludes that the proposed testing will provide adequate assurance of the structuralintegrity for the subject welds. Conclus/on-Based on the impracticality of meeting the Code surface examination requirements, and the adequata assurance of the structuralintegrity provided by the ultrasonic examination of the full volume from the interior surface, it is recommended that relief be granted pursuant to 10 CFR 50.55a(g)(6)(i). 3.1.5 Pump Pressure Boundary 1 No relief requests. 3.1.6 Valve Pressure Boundary i No relief requests. 3.1.7 Gerwral 3.1.7.1 Request for Relief RR-8916 (March 20,1997, Revision), Exart,ination Category B-E, items B4.10 and B4.20. Pressure Retaining Partial Penetration Welds in Vessels Code Requirement-Tablo IWB-2500-1, Examination Category B-E, items 84.10 and 84.20 require a VT-2 visual examination of pressura retaining partial penetration welds in vessels in conjunction with system hydrostatic tests. Licensee's Proposed A/temative-in accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative for tracking the subject Code items and from performing the required VT-2 visual examinations on these items, as stated.
" Code Class 1, Systern pressure integrity, including these Components shall be verified during the Ten-Year System hydrostatic test in accordance with (lWB-5222)."
Licensee's Basis for the Proposed Alternative-
" Subsequent Editions and Addenda of the ASME B&PV Code have deleted the Visual (VT-2) examination requirements for B-E Category items from the IWB Tables.
"NNECO, MP-2 believes that tracking the VT 2 examinations separately will not provide any potential increase in plant safety margins and are no longer warranted based on later Editions and Addenda of the Code deleting these items from the IWB-2500-1 Tables."
Eve /uetion-The Code requires a Visual (VT-2) examination for Category B-E components. The licensee has proposed to track the Code-required hydrostatic and VT-2 visual examinations for Examination Category B-E in conjunction with the vessel pressure tests. it is reasonable to track the Code required hydrostatic test and associated VT-2 visual 1 25
I examinations together. Therefore, the INEEL staff believes that the licensee's proposed altemative provides an acceptable level of quality and safety. Com:Ausion--Based on the evaluation above, the licensee's proposal to track the Code-required pressure test and VT 2 visual examinations for Examination Category B-E in conjunction with the 10-year system hydronatic pressure test for the reactor vessel will l provide an acceptable level of quality and safety. Therefore, it is recommended that the
.. proposed alternative be authorized pursuent tn 10 CFR 50.55a(a)(3)(i).
3.1.7.2 Request for Relief RR 89-18 (March 20,1997, Revision), Examination Category 8 0-2, item 87.80, Pressure Retaining Bolting 2-inch and Less in Diameter Note: In the July 2,1996, submittal, the licensee listed Code Case N-547, Altemative Examination Requirements for Pressure Retaining Bolting in Control Rod Drive (CRD) Nousings, Section X/, Division 1, and referenced this request for relief. As a point of clarification, this is not an evaluation of Code Case N-547, but of the licensee's proposed l alternative as stated below. As discussed in Section 2.2.1 of this report, Code Case N-547 is not acceptable as written. Code Regudremont--Section XI, Table IWB-2500, Examination Category B-G-2, item B7.80, requires a VT-1 visual examination of the surfaces of the bolts, studs, and nuts of control rod ddve (CRD) housings when the CRD is disassembled. Licensee's proposed Attemotive-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee !
. proposed an altamative to the Code required VT-1 visual examination of CRD bolts, studs, I and nuts, as stated.
"NNECO will perform a visual, VT 1 examination of this bolting prior to reassembly during routine maintenance, unless the bolting is replaced with new ones.
"A System pressure test and VT 2 visual examination shall be performed each refueling outage on all Class 1, bolted connections including this (CEDM) Greylock bolting."
Licensee's Basis for Requesting ReWef-
"ASME Code Case N 547, has stipulated that the 'VT-1 Visual examination of CRD housing bolts, studs, and nuts is not required'. Applicability: 1980 Edition with the Winter 1980 Addenda through the 1995 Edition.
"Although the actual CEDM components do not have any Code item B7.80, CRD Housing Bolts, Studs, or Nuts, Two CEDM's have been modified to include Greylock Fittings complete with one (1) set of four (4) studs and nuts each in support of the reactor vessel level monitoring system.
"These fittings are located at the very top of CRD housing #11 and #13, which are
- physically located in the second ring of housings from the center of the head. Access t
to these two CRD housings is limited. The Reactor Vessel Head is a high radiation area and these required VT-1 examinations are not consistent with the "ALARA" program at Millstone Unit 2. NNECO believes that the performance of these VT-1 examinations when the component is disassembled for maintenance will provide adequate assurance of plant safety margins." 26 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - - )
I Eva/uat/on-The Code requires a VT-1 visual examination of Control Rod Drive bolting, ! studs, and nuts when the CRD housing is disassembled. The licensee has proposed to , perform the VT-1 visual examination unless the botting is replaced with new bolts. I in the case of bolting that the licensee intends to reuse, it is imperative that the condition of thesc bolts, studs, and nuts be verified prior to reinstallation. The VT-1 visual
~
exarnination of the CRD bolting, when removed, does not place a significant burden on the licensee. However, performing examinations on bolting that will be repla.ned exposes plant personnel to unnecessary radiation exposure. Therefore, imposition of this requirement would result in a hardship without a compensating increase in the level of quality and ! safety. The VT-1 visual examination of previously used bolting prior to reassembly will detect any patterns of ongoing inservice degradation and will provide reasonable assurance {i of the continued structuralintegrity of the CRD bolting. ConcAuston-Based on the evaluation above, it is concluded that imposing the Code requirements on the licensee for bolting that will be replaced would result in a burden without a compensating increase in the level of quality and safet f. Therefore, it is recommended that the licensee's proposed alternative be authorized pursuant to 10 CFR { 50.55ata)(3)(ii). 3.2 Class 2 Components 3.2.1.1 Proposed Alternative N-522, Use of Code Case N-522, Pressure Testing of Containment Penetration Piping Code Requirement-Section XI, Table IWC-2500-1, Examination Category C-H, items C7.30, C7.40, C7.70 and C7.80 require a system leakage test at operating pressure for
~
pressure-retaining piping and valves. Licensee's Proposed Attemative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to use Code Case N-522 for pressure-retaining piping that penetrates L containment and is classified as Class 2 when the balance of the system is outside the scope of Section XI (non-classed). The licensee will comply with the following condition
- specified in the regulatory position of draft Regulatory Guide DG-1050, inservice i Inspection Code Case Acceptability, ASME Section XI, Division 1
" Code Case N-522 is acceptable subject to the following condition in addition to those specified in the Code Case. The test should be conducted at the peak calculated containment pressure and the test procedure should permit the detection and location of through-wall leakage in containment isolation valves (CIVs) and pipe segments between CIVs."
t Licensee's Basis for Proposed Attemative (as stated)- l "The performance of the hydrostatic and functional tests, as currently required by Section XI, would result in a substantial hardship to the plant in the form of additional expenditures of funds to perform the tests and additional radiation exposure to the l individuals involved in the tests without realizing c commensurate gain in safety." ' Evaluation-The Code requires that a VT-2 visual examination be performed during system ; pressure testing for Class 2 pressure-retaining piping. As an alternative, the licensee ! 27 L----_-____--.--_--_
1 proposes to implement the requirements of ASME Code Case N 522, Pressure Testing of Containment Penetration Piping, which allows the use of 10 CFR 50, Appendix J for Class 2 piping penetrations where that balance of the piping is outside the scope of l Section XI. l l The subject piping is classified as Class 2 because it penetrates primary reactor 1 containment and is considered an extension of the containment vessel. Since the piping I on either side of these penetrations is non-classed, the requirements of Appendix J are more appropriate than those of Examination Category C-H. Appendix J pressure tests verify the leak tight integrity of the primary reactor containment and of systems and components that penetrate containment by local leak rate and integrated leak rate tests. ! In Appendix J pressure tests, containment isolation valves and connecting pipe segments j must withstand the peak calculated containment internal pressure related to the maximum design containment. The NRC staff has determined that the Appendix J test frequencies are acceptable for assuring containment integrity. In addition, the licensee has committed to perform the Appendix J testing at no less than the peak calculated containment pressure and will use procedures that permit the detection and location of throught-wall leakage. Therefore, use of Appendix J provides reasonable assurance of the operational i readiness for the subject penetration piping. i Considering the acceptability of the proposed alternative, the performance of a system leakage test would result in excess radiation exposure to plant personnel. Therefore, it is concluded that imposition of the Code requirements for the subject containment i penetrations would result in a hardship without a compensating increase in the level of quality and safety. Conclusion-eased on the evaluation above, it is concluded that imposing the Code l requirements for Class 2 containment penetrations when the balance of the piping is non-classed would result in a burden without a compensating increase in the level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative to use Code Case N-522 be authorized pursuant to 10 CFR 50.55a(a)(3)(ii). The use of this Code l Case should be authorized for the third 10-year interval at Millstone Unit 2. or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use Code Case N-522 with the limitations, if any, listed in Regulatory Guide 1.147. I 3.3 Class 3 Components No relief requests. i 3.4 Pressure Tests 3.4.1 Class 1 System Pressure Tests No relief requests, l 3.4.2 Class 2 System Pressure Tests No relief requests. f L 1 28 L
O 3.4.3 Class 3 System Pressure Tests No relief requests. 3.4.4 General 3.4.4.1 Request for Relief RR-13, IWA-4700(a) and (b), Alternative Pressure Test Requirements For Code Class 1, Class 2, and Class 3 Systems Following Repair, Replacements, and Modifications Note: Request for Relief RR-13 was submitted separately in a letter dated June 12,1996 and is evaluated elsewhere by the NRC staff. This request is listed for j information purposes only. ' 3.4.4.2 Proposed Altemative to use Code Case N 416-1, A/temative Pressure Test Requirements for Welded Repairs orinstaHation of Replacement items By Welding, Class 1, 2, and 3, Section XI, Division 1 Note: The licensee requested authorization to use this Code Case for the previous 10-year interval by letter dated September 22,1994 (Reference 15). Use of this Code Case was subsequently authorized in an NRC SER dated October 25,1994. In the August 28, 1997, letter, the licensee requested authorization to use this Code Case during the third 10 year'ISI interval. Code Requirement-4WA-4700(a) requires that a system hydrostatic test be performed in' accordance with IWA 5000 after repairs by welding on the pressure retaining boundary. Licensec's Proposed Attemative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the,1censee proposed to apply Code Case N-416-1 as an alternative to the ASME Section XI requirements for welded repairs or installation of replacement items by welding in Class 1, 2, and 3 systems during the third 10 year interval in the August 28,1997, letter, the licensee stated:
"During the third 10-year interval NNECO plans to use this Code Case subject to the staff's approval with a commitment that additional surface examinations will be performed on the root (pass) layer of butt and socket welds on the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section ill of the ASME Code.
Licensee's Basis for the Proposed Attemative-
"The performance of the hydrostatic tests, as currently required by the code, would result in a substantial hardship to the plant in the form of additional expenditures of funds to performed the tests and additional exposures to the individuals involved in the tests without realizing a commensurate gain in safety. The issuance of this (these)
Code Case revision (s) by the ASME, with NRC acceptance, provides the basis for the approval of the altemative(s) which has been determined by the ASME consensus process to provide an acceptable level of quality and safety." Evaluation-Section XI of the Code requires a system hydrostatic test to be performed in accordance with IWA-5000 after repairs made by welding on the pressure-retaining 29
9 boundary. The licensee has proposed the use of Code Case N-416-1 in lieu of the Code requirements. In addition, the licensee will perform a supplemental surface examination on the root pass of Class 3 butt and socket welds when a surface examination is required by Section Ill. Code Case N-4161 specifies that NDE of the welds be performed in I accordance with the applicable subsection of the 1992 Edition of Section Ill. The Code Case also allows a VT-2 visua! examination to be performed at nominal operating pressure and temperature in conjunction with a system leakage test, in accordance with Paragraph IWA-5000 of the 1992 Edition of Section XI. The 1989 Editions of Sections ill and XI are the latest Code editions referenced in 10 CFR 50.55a. The NRC staff previously compared the system pressure test requirements of the 1992 Edition of Section XI to those of the 1989 Edition, in summary:
- 1) The test frequencies and the pressure conditions associated with these tests have not changed;
- 2) The hold times have either remained unchanged or increased;
- 3) The terminology associated with the system pressure test requirements for all three Code classes has been clarified and streamlined; and
- 4) The NDE requirements for welded repairs remain the same.
Piping components are designed to withstand the loading mechanisms that are . postulated to occur under the various modes of plant operation. Hydrostatic testing I subjects the piping components to a small increase in pressure over the design pressure and, therefore, does not present a significant challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components under pressure rather than a measure of the structuralintegrity of the components. Considering the NDE performed on Code Class 1'and 2 systems and that the hydrostatic pressure tests rarely result in pressure boundary leaks that would not have occurred during system leakage tests, the INEEL staff believes that the added assurance of integrity provided by the hydrostatic test is not commensurate with the associated burden, which typically includes the installation of blanks, cutting and removing supports for access, and removing insulation to prepare and restore the systems, all of whicn increase radiation exposure for plant personnel. For Class 3 components, there are no ongoing NDE requirements except for the visual examination for leaks in conjunction with the 10-year hydrostatic test and periodic pressure tests. Therefore, eliminating the hydrostatic test and only performing the system pressure test for Class 3 components is only considered acceptable if an additional surface examination is performed on the root pass layer of butt and socket welds on the pressure-retaining boundary during repair and replacement activities. The licensee has included this provision in their alternative when surface examinations are required by Section Ill. Therefore, the licensee's proposed alternative will provide reasonable assurance of the operation readiness of all effected systems. I l 30 , I 4 i
e Conclusion-Based on the evaluation above, it is concluded that compliance with the Code-required hydrostatic testing for welded repairs or replacements of Code Class 1,2, and 3 components would result in a hardship without a compensating increase in the level of quality and safety. Therefore, it is recommended that the proposed alternative, use of Code Case N-4161, be authorized pursuant to 10 CFR 50.55a(a)(3)(ii). The use of this Code Case should be authorized for the third 10-year interval at Millstone Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee may continue to use Code Case N-416-1 with the limitations, if any, listed in Regulatory Guide 1.147. 3.4.4.3 Proposed Alternative to use Code Case N-498-1, Attemative Kules for 10 Veer System Hydrostatic Testing for Class 1, 2, ami3 Systems, Section XI, Division 1 Note: The licensee requested authorization to use this Code Case for the previous 10-year interval by letter dated December 16,1994, (Reference 16). Use of this Code Case was subsequently authorized in an NRC SER dated January 18,1995. In the August 28, 1997, letter, the licensee requested authorization to use this Code Case during the third 10-year ISI interval. Code Requ/remont-Table IWB-2500-1, Examination Category B-P, Table IWC-2500-1, Examination Category C-H, and Table IWD-2500-1, Examination Categories D-A, D-B and D C, require system hydrostatic testing of pressure-retaining components in accordance with IWA-5000 once each 10-year interval. Licensee's proposed A/temative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to apply Code Case N-4981 as an alternative to the ASME Section XI requirements for system hydrostatic testing of Class 1,2, and 3 systems during the third 10-year interval. Licensee's Basis for the Proposed Attemative-
"The performance of these hydrostatic tests, as currently required by the Code, would result in a substantial hardship to the plant in the form of additional expenditures of funds to perform the tests and additional radiation exposure to the individuals involved in the tests without a commensurate gain in safety."
Evaluation-The Code requires the performance of a system hydrostatic test once per intervalin accordance with the requirements of IWA 5000 for Class 1,2, and 3 pressure-retaining systems. In lieu of the Code required hydrostatic testing requirements, the licensee has requested authorization to use Code Car,e N-498-1, Alternative Ru/es for 10-Year System Hydrostatic Testing for Class 1, 2, and 3 Systems, dated May 11,1994. Tiie system hydrostatic test, as stipulated in Section XI,is not a test of the structural integrity of the system but rather an enhanced leakage test." Hydrostatic testing only subjects the piping components to a smallincrease in pressure over the design pressure; l therefore, piping dead weight, thermal expansion, and seismic loads present far greater ! challenges to the structural integrity of a system. Consequently, the Section XI
- k. S. H. Bush and R. R. Maccary, ' Development ofIn-Service Inspection Safety Philosophy for U.S.A. Nuclear Power Plants, ' ASME,1971 31
l0 , l i hydrostatic pressure test is primarily regarded as a means to enhance leak detection during i the examination of components under pressure, rather than as a method to deterrnine the structural integrity of the components. In addition, the industry experience indicates that
)
leaks are not being discovered as a result of hydrostatic test pressures causing a preexisting flaw to propagate through the wall. In most cases leaks ere being found when the system is at normal operating pressure. Code Case N-498, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems, was previously approved for general use on Class 1 and 2 systems in ; Regulatory Guide 1.147, Revision 9. For Class 3 systems, Revision N-4981 specifies i requirements identical to those for Class 2 components (for Class 1 and 2 systems, the alternative requirements in N-498-1 are unchanged from N-498). In lieu of 10-year i hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a VT-2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test performed in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI. I Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a review of Class 3 system failures requiring repair during the last 5 years,' the most common causes of failures are erosion-corrosion (EC), microbiologically-induced corrosion (MIC), and general corrosion. In general, licensees have implemented programs for the prevention, detection, and evaluation of EC and MIC: therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms. System hydrostatic testing entails considerable time, radiation dose, and dollar resources. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove Code safety and/o relief valves (placing the system, and thus the plant, in an off normal state), erect temporary supports in steam lines, and expend resources to set up testing with special equipment and gages. Therefore, performance of system hydrostatic testing represents a considerable burden for the licensee. Giving consideration to the minimal amount of increased assurance provided by the increased pressure associated with a hydrostatic test versus the pressure for the system leakage test and the hardship associated with performing the hydrostatic test, the INEEL staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Conclusion--Based on the evaluation above, the INEEL staff concludes that compliance with the requirements of Section XI for hydrostatic testing would result in a hardship without a compensating increase in quality and safety. Therefore, it is recommended that the use of Code Case N-4981 for Code Class 1,2, and 3 systems be authorized pwsuant
- 1. Documented in Licensee Event Reports and the Nuclear Plant Reliability Data System i databases.
32
6 i. ! to 10 CFR 50.55a(a)(3)(ii). The use of this Code Case should be authorized for the third interval at Millstone, Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide. 3.4.4.4 Request for Relief RR 89-08 (March 20,1997, Revision),IWA 5250(a)(2), Corrective Action Resulting from Leakage at Bolted Connections Code Requirement--4WA-5250(a)(2) requires that the source of leakages detected during a system pressure test shall be located and evaluated by the Owner for corrective action. When the leakage is at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. L/cansee's Proposed A/temative-Pursuant to 10 CFR 50.55s(a)(3)(i), the licensee proposed an alternative to the ASME Section XI requirements for removal of bolting at leaking connections for VT-3 visual examination, as stated.
"If leakage is detected at a bolted Class 1,2, or 3 connection while performing visual I VT 2 examinations in accordance with IWA-5000, this leakage will be evaluated to determine the susceptibility of the bolting to corrosion and the impact on system operability. As a minimum this evaluation will consider the following items:
l 1. Location of leakage
- 2. Bolted connection material
- 3. System operability and history
- 4. corrosiveness of process fluid
- 5. Impact on other systems or components in area
- 6. Generic impact "When evaluation of the variables above indicate a need for further evaluation, the bolt closest to the source of leakage will be removed, visually VT-3 examined and evaluated in.accordance with IWA 3100 (a). If the examination results aro determined to be l rejectable in accordance with the applicable IWA-3100 (a) reference then g!! the remaining bolts in the connection shall be remove, examined and evaluated to the above one bolt requirements."
Licensee's Basis for the Proposed Attemative-
"NNECO, MP 2, believes that removing the complete set of bolting for a visual VT-3 examination and evaluation if leakage occurs at a bolted connection is not commensurate with the expense of manpower resources and in certain cases, l additional radiation exposure. Leakage at a bolted connection is not always indicative i of degraded botting that would require further examination or evaluation. Other variables to consider would be:
A. Orientation of the leak - not all of the bolts may be wetted. B. The leaking media - may not be corrosive C. Duration of the leak -length of time the bolts were actually in contact with the leaking media i 33 . w _ _______________________
D. Bolting material- may not be susceptible to corrosion. ! " Removing bolts from a leaking connection to perform a VT-3, visual examination may I not be the most prudent course of action until these variables are considered. The proposed alternative to the above requirements will provide an equivalent level of l quality and safety." Eve /ust/on--4n accordance with the 1989 Edition of the Code, when leakage occurs at' bolted connections, cil bolting is to be removed for VT-3 visual examination. In lieu of removal of all bolting to perform a VT-3 visual examination, the licensee has proposed to perform an evaluation of the bolted connection. The evaluation will consider the potential for bolting degradation as well as the cause of the leakage. if the evaluation indicates the need for a more detailed analysis, the bolt closest to the source of leakage will be removed, VT 3 examined, and evaluated in accordance with IWA 3100(a). The licensee's altemative to bolting removal when leakage occurs is based on the use of sound engineering judgment to evaluate the integrity of the bolted connection. Thus, degradation of the bolting,if occurring, should be detected. As a result, it is believed that the licensee's proposed alternative to the Code-required removal of botting at a joint when leakage occurs will provide an acceptable level of quality and safety. Conclusion--Based on the evaluation above, the licensee's proposed alternative to perform l an evaluation of the bolting when leakage occurs at a bolted connection providos an acceptable level of quality and safety. Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). 3.4.4.5 Request for Relief RR 89-17 (August 28,1997, Revision), IWA 5242(a), insulation Removal For VT 2 Visual Examination Of Bolting in Borated Systems Code Requ/rement--4WA-5242(a) requires that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retaining bolted connections for VT-2 visual examination. Licensee's Proposed A/temative-Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed an alternative to the ASME Section XI requirements for removing insulation during VT 2 visual examinations, as stated.
"a. Each refueling outage Borated Class 1 system connections shall be VT-2, visually examined at zero or static pressure with insulation removed.
) "b. Each inspection period Borated Class 2 system connections shall be VT-2, visually examined at zero or static pressure with insulation removed. l "c. A system pressure test and VT 2, visual examination shall be performed each refueling outage on all Class I bolted connections without removal of insulation.
"d. A system pressure test and VT-2, visual examination shall be performed each inspection period on all Class 2 bolted connections without removal of insulation.
34
t D
"e. Any evidence of leakage shall be evaluated in accordance with IWA 5250 as amended by Relief Request #RR 89-08."
. Licensee's Basis for Requesting ReHef-
"NNECO concludes that removing insulation from bolted connections in borated
{ systems for visual VT-2 examination is impractical and would result in unnecessary l
- personnel exposure without a commensurate increase in the level of safety, reliability, !
l or quality over the proposed alternative examinations.
"lWA 5242(a) examinations of over 150 valve bolted connections in high radiation and or hard to access areas are impractical to examine.
j "These borated system are very large systems covering large areas at various elevations within containment and its auxiliary building. Scaffolding would be required to access many of these bolted connections that are located in high radiation and/or I contaminated areas, insulation removal and replacement combined with additional scaffolding requirements willincrease personnel exposure and generate unnecessary radweste in the performance of these VT-2 examinations. These examinations are performed prior to start up when the reactor coolant temperature is greater than 500 degrees Fahrenheit.
"These activities associated with removing and reinstalling insulation and removing i temporary staging are difficult to perform relative to personnel safety and considered to be impractical when compared to the proposed alternatives."
Ersamt/on-Paragraph IWA 5242(a) requires the removal of all insulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. The licensee has j proposed an alternative similar to the requirements of Code Case N-533, Altemative ! Requirements for VT-2 VisualExamination of Class 1 Insulated Pressure-Retaining Bolted { Connections, for Class 1 and 2 bolted connections in borated systems. This Code Case i allows the VT-2 visual examination to be performed in conjunction with startup following a i 4-hour hold time at operating pressure with the insulation in place. A VT-2 visual examination is then performed each refueling outage during cold shutdown with the insulation removed. Requiring the licensee to remove insulation during the Class 1 system pressure test would create a safety hazard due to the elevated temperatures, and would also result in excess radiation exposure to plant personnel. Therefore, the requirements of IWA 5242(a) would create an undue burden on the licensee. The licensee's proposed alternative is essentially equivalent to Code Case N-533, except the proposed alternative was expanded to address both Class 1 and Class 2 bolted connections. Code Case N-533 is currently under review by the NRC staff and has not yet been approved for use by incorporation into Regulatory Guide 1.147, inservice Inspection Code Case Acceptability. For Class 1 systems, the licensee's proposed alternative provides a reasonable approach of ensuring the leak-tight integrity of systems borated for the purpose of controlling reactivity. First, the 4-hour hold time allows any significant leakage to ; penetrate the insulation. Second, by removing the insulation each refueling outage, the l C,,ensee will be able to detect minor leakage indicated by the presence of boric acid l I 35
r,
,, l crystals or residue. This two-phase approach provides reasonable assurance of the continued operational readiness of Class 1 bolted connections in borated systems.
For Class 2 and 3 systems, the frequencies proposed for insulation removal have not been found acceptable by the NRC staff. Therefore, the licensee's proposed alternative ; should not be authorized for Class 2 and 3 systems. j i ConcAuston-Based on the evaluation above, it is concluded that compliance with the Code I requirements for Class 1 systems would result in a burden without a compensating increase in the level of quality and safety. Furthermore, the licensee's proposed alternative provides reasonable assurance of continued operational readiness for Class 1 bolted connections. Therefore, pursuant to 10 CFR 50.55a(a)(3)(ii),it is recommended that the l licensee's proposed alternative be authorized for Class 1 systems. For Class 2 and 3 I systems, the proposed frequencies have not been found acceptable by the NRC staff. j Therefore, the licensee's proposed alternative should not be authorized for Class 2 and 3 ' systems. 3.4.4.6 Proposed Alternative to use Code Case N-546, A/temative Regdrements for Quem 6 cation of VT-2 Examination Personnel, Section XI, Division 1 Note: The licensee requested authorization to use this Code Case for the previous 10-year interval by letter dated March 20,1996, (Reference 17). Use of this Code Case was subsequently authorized in an NRC SER dated September 13,1996. In the August 28, 1997, letter, the licensee requested authorization to use this Code Case during the third 10 year ISI interval. l Code #egdromant-Section XI, IWA-2300, requires that personnel performing VT-2 and VT-3 visual examinations be qualified in accordance with comparable levels of competency as defined in ANSI N45.2.6. Additionally, the examination personnel shall have natural or corrected near distance acuity, in at least one eye, equivalent to a Snellen fraction of 20/20. For far vision, personnel shall have natural or corrected far distance visual acuity of 20/30 or equivalent. Licensee's Prqposed A/temative-Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee ; proposed to use Code Case N-546 as an alternative to the ASME Section XI qualification requirements for VT-2 visual examiners. Additionally, the licensee has committed to 1) implement procedural requirements to assure that consistent VT-2 examinations are performed,2) include a provision that an independent review and evaluation of the findings be performed by persons other than those that performed the examinations, and 3) document the qualifications, training and visual acuity of those persons performing VT-2 visual examinations. Licensee's Basis for Roguesting ReMef-
"NNECO requests relief from the qualification requirements for VT-2 visual examiners.
Specifically, NNECO requests approval to implement the alternatives to the Code j , requirements contained in Code Case N-546, which is not yet approved in Regulatory j Guide 1.147.
" Code Case N 546 allows experienced plant personnel, such as licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and 36
Ce t, nondestructive examination personnel, to perform VT-2 visual examinations without having to be certified to comparable levels of competency defined in ANSI N45.2.6. NNECO personnel performing visual examinations will be subject to the conditions provided in Code Case N-546. Using the alternative requirements provided in this Code l Case alleviates the need to contract certified VT-2 personnel to perform these examinations and reduces the administrative burden of maintain;ng a Section XI qualification and certification program for the contract personnel.
"Specifically, if the Code Case is approved, the examination personnel will have at least 40 hours of piant walkdown experience, receive a minimum of four hours of training on L Section XI requirements and plant specific procedures for VT-2 visual examinations, l and will pass the vision test requirements of lWA-2321,1995 Edition. NNECO concludes that the use of this Code Case will provide an acceptable level of quality and I safety based on the requirements in the Code Case and the additional requirements l contained in the commitment below, i
l "In addition to the requirements of Code Case N-546, additional procedure l requirements have been implemented to assure that consistent VT-2 visual examinations are performed among the Millstone Units that have approval to use this Code Case and these requirements shall include a provision that an independent review and evaluation of the findings be performed by persons other than those that performed i the VT-2 examinations. l Commitment: NNECO willimplement a program that documents the qualifications, L training, and visual acuity of persons selected to perform the VT-2 examinations and l will maintain records that all of the requirements in the Code Case are met.
"During the Third 10-Year Interval, NNECO plans to use this Code Case subject to the Staff's approval with the commitment as described above to implement additional
- procedure and documentation requirements. NNECO will use this Code Case until such time that this Code Case is published in a future revision of Regulatory Guide 1.147.
! At that time NNECO will follow all provisions in Code Case N 546, with limitations { issued in Regulatory Guide 1.147, if any." l- Evatustion-The Code requires that VT-2 visual examination personnel be qualified to levels L of competency comparable to those identified in ANSI N45.2.6. The Code also requires that the examination personnel be qualified for near and far distance vision acuity. In lieu l of the Code requirements, the licensee proposed to implement Code Case N-546 for l personnel performing VT-2 visual examinations. This Code Case includes the following requirements:
- 1. At least 40 hours plant walkdown experience, such as that gained by licensed and ,
nonlicensed operators, local leak rate personnel, system engineers, and inspection and I l nondestructive examination personnel. 1
. 2. At least four hours of training on Section XI requirements and plant specific procedures 1 for VT-2 visual examination. '
- 3. Vision test requirements of IWA-2321,1995 Edition. I i
I 37 1
l s. l 1 j The qualification requirements in Code Case N 546 are not significantly different from those for VT-2 visual examiner certification. Licensed and nonlicensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel typically have a sound working knowledge of plant components and piping layouts. This knowledge makes them acceptable candidates for performing VT-2 visual examinations. l In addition to meeting the requirements contained in Code Case N-546, the licensee has committed to use procedural guidelines for consistent, quality VT-2 visual examinations, verify and maintain records of the qualification of persons selected to perform VT-2 visual examinations, and perform independent reviews and evaluations of leakage by a person (s) other than those that performed the VT-2 visual examination. Based on a review of Code Case N-546 and the additional commitments made by the-licensee, the iNEEL staff believes that the proposed alternative to the Code requirements i will provide an acceptable level of quality and safety. I Conclusion-Based on the evaluation above, the INEEL staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's request to implement Code Case N-546 with the additional commitments be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of this Code Case should be authorized for the third interval at Millstone, Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide. 3.5 General ; 3.5.1 Ultrasonic Examination Techniques No relief requests. 3.5.2 Exempted Components No relief requests. 3.5.3 Other 3.5.3.1 Request for Relief RR-14, IWA, IWB, IWC, and IWF-4000 (IWX-4000), Repair Procedures, IWA, lWB, IWC, and IWF 7000 (lWX-7000), Replacements Note: Request for Relief RR-14 was submitted separately in a letter dated June 12,1996 and is evaluated elsewhere by the NRC staff. This request is listed for information purposes only. 3.5.3.2 Request for Relief RR-89-19, Examination of Code Class Snubbers ! This request for relief is considered part of the Inservice Testing Program (IST) and is, therefore, not included in this evaluation. The Snubber Testing Program will be evaluated by the Mechanical Engineering Branch of the NRC. 38 i
._-- _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . . . _ - _ _ _ ~
lo s. 3.5.3.3 Request for Relief RR 89-20, Removal of Pressure Relief Valves for the Purpose of I Testing This request for relief is considered part of the Inservice Testing Program (IST) and is, ! therefore, not included in this evaluation. The inservice Testing Program will be evaluated by the Mechanical Engineering Branch of the NRC. 3.5.3.4 Request for Relief RR 89-21, IWA-4000 and IWA-7000, Requirements for Repairs and Replacements Code Requirement-The Code requires that repairs and replacements be performed in accordance with IWA-4000 and IWA-7000, respectively. L/consee's prepa ed A/temative-in accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed an alternative to the ASME Section XI repair and replacement requirements, as stated.
" Invoke the use of Code Case N-544 as an alternative to ASME Code Section XI requirements when repairing or replacing NPS 1" and less piping, valves and fittings."
l Licensee's Basis for the Proposed Attemativa-
"The issuance of a Repair / Replacement Plan, conducting pressure testing, supplying the services of an Authorized inspection Agency or the completion of an NIS-2 form will not substantially increase the reliability, safety or quality of these items over the proposed alternative requirements. Code Case N-544 provides an alternative to current Section XI, Paragraph IWA 4000 Repair and IWA-7000 Replacement requirements in regard to smallitems. Millstone Unit 2, believes that meeting the requirements of this Code Case will adequately satisfy component compatibility and the intent of the repair and/or replacement requirements of the applicable Code Edition."
Evaluation-4WA-4110(b) states that the rules and requirements of this Article apply to welding items used for replacement (as defined in IWA-7110) to the system. IWA-7110 states, in part, " Replacement includes the addition of components, such as valves, and system changes, such as rerouting of piping, within the scope of this Division." The Code requirements that the licensee is requesting relief from are administrative requirements only. As such, the licensee has not demonstrated that the Code requirement is impractical i or represents a hardship or unusual difficulty. Further, the INEEL staff believes that the current Code requirements impart a level of quality and safety that would not be present under the licensee's proposed attemative. Therefore, whenever a repair or replacement involves welding, the rules of IWA-4000, Repair Procedures apply. l I ConcAus/ons-Based on the above evaluation, it is recommended that relief be denied. I l l l l k 39 l l
e, i, 3.5.3.5 Proposed Altemative to use Code Case N-524, A/temative Examination Requirements for Longitudinal Wekts in Ciess 1 and 2 Mping, Section XI, Division 1 Note: The licensee requested authorization to use this Code Case for the previous 10-year interval by letter dated June 2,1995, (Reference 18). Use of this Code Case was subsequently authorized in an NRC SER dated August 4,1995. In the August 28,1997, letter, the licensee requested authorization to use this Code Case during the third 10-year ISI interval. Code Requirement--Examination Category B-J, requires 100% volumetric and/or surface examination of longitudinal piping welds as defined by Figure IWB-2500-8. The
+
examination shall include at least one pipe diameter, but not more than 12-inches of each longitudinal weld intersecting circumferential wolds required to be examined by Examination Categories B-J and B-F. Examination Categories C F-1 and C-F-2, require 100% volumetric and/or surface examination, as defined by Figure IWC-2500 7,12 and - 13, for 2.5t of each longitudinal weld intersecting circumferential welds examined. Licensee's Proposed A/temative--{n accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-524 as alternative requirements for the examination of Class 1 and 2 longitudinal piping welds. Licensee's Basis for the Proposed Altemative-
"NNECO ISI ;,rogram requirement iv the examination of longitudinal piping welds l require that one pipe diameter in length, but no more than 12 inches, be examined for Class 1 longitudinal piping welds. The ISI program also requires that for a longitudinal pipe weld on Class 2 iping that a length of 2.5t (where t is the thickness of the weld) be examined. The longitudinal weld length is measured from the intersection of the circumferential weld and longitudinal weld. Code Case N-524, however, limits the volumetric and surface examination requirements of the longitudinal weld to the volume or area contained within the examination requirements of the intersecting circumferential weld.
" Longitudinal welds are produced during the manufacturing process of the piping, not in the field, as is the case for circumferential welds. The Code contains requirements for characteristics and performance of materials and products, and specifies the examination requirements during the manufacture of the subject longitudinal piping welds.
"The preservice examination and initial inservice examinations performed by NNECO have provided assurance of the structuralintegrity of ASME Code longitudinal welds during the service life of the unit to date. The experience in the United States has been that ASME Code longitudinal welds have not experienced degradation that would warrant continued examination beyond the boundaries required to Meet the circumferential weld examination requirements. No significant loading conditions or known material degradation mechanisms have become evident to date which specifically relate to lonCi tudinal seam welds in nuclear plant piping, if any degradation associated with a longitudind weld were to occur, it is expected that is would be located at the intersection with the circumferential weld. This intersection is inspected
' in accordance with the provisions of CWe Case N-524.
40
c, b
" Code Case N-524 directs the examination effort at weld intersections. It eliminates the longitudinal weld from examination, significantly reducing examination time requirements and pctential radiological exposure to examination personnel. Compliance with the existing ASME Section XI requirements, in lieu of the Code Case, results in unnecessary personnel exposure to complete the required examinations without a commensurate increase in the level of quality and safety."
Evabet/an-ASME Section XI requires the examination of one pipe diameter, but not more than 12 inches, of Cissa 1 longitudinal piping welds. For Class 2 piping welds, the length of longitudinal weld required to be examined is 2.5 times the pipe thickness. These lengths are measured from the intersection with the circumferential weld. The licensee's proposed alternative is to examine only the portions of longitudinal weld within the examination area of the intersecting circumferential weld in accordance with Code Case N-i 524, Altemative Examination Requirements for Longitudinal Welds in Class 1 and Class 2
. Piping.
Longitudinal welds are produced during the manufacture of the piping, not in the field as is the case for circumferential welds. Consequently, longitudinal welds are fabricated ( l under strict manufacturing standards, which provides assurance of structuralintegrity. These welds have also been subjected to the preservice and initial inservice examinations, which provide additional assurance of structuralintegrity. No significant loading conditions or material degradation mechanisms have been identified to date that specifically relate to longitudinal seam welds in Class 1 and 2 nuclear plant piping. The most critical region of the longitudinal weld is the portion that intersects the circumferential weld. If degradation associated with a longitudinal weld were to occur, it is expected that it would be located at the intersection with a l circumferential weld. Since this region will be examined during the examination of the i 'circumferential weld, the licensee's alternative provides reasonable assurance of the continued structural integrity. Carmbs/ons-Based on the evaluation above, it is concluded that the use of Code Case N 524 provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed alternative, to use Code Case N-524, be authorized pursuant ! to 10 CFR 50.55a(a)(3)(i). The use of this Code Case should be authorized for the third interval at Millstone, Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that tirne, the licensee must follow the L conditions, if any, specified in the regulatory guide. 3.5.3.6 Proposed Alternative to use Code Case N 535, Altemative Requirements for kseovice kspection Mtervals, Section XI, Division 1 Note: The licensee requested authorization to use this Code Case for the previous 10- , year interval by letter dated August 24,1995, (Reference 19). Use of this Code Case was l l subsequently authorized in an NRC SER dated April 5,1996. In the August 28,1997, ! I letter, the licensee requested authorization to use this Code Case during the third 10-year f- ISI interval.- Code Requirement-Paragraph IWA-2432 requires that successive inspection intervals be comprised of 10 years following the previous interval except as modified by Paragraph j 41 i i
l O l l IWA-2430(d), which allows an interval to be extended or reduced by as much as one year to coincide with an outage, thus changing the length of an interval. i Uconsee's Proposed Altemative-4n accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-535 as altemative requirements for scheduling the 10-year
' inspection interval.
i Ucensee's Basis for the Proposed Altemative-l "NNECO requests the use of an altamative to the ASME Boiler and Pressure Vessel l Code, Section XI Edition and Addenda, pursuant to the provisions of 10 CFR ! 50.55a(a)(3)(i). Specifically, NNECO requests approval to use the provisions of Code Case N 535, 'Altemative Requirements for inservice Inspection Intervals, Section XI, i Division 1,' for the Third 10-Year interval ISI Program Plan. i "This Code Case allows altamatives to the Section XI requirements now being used to schedule inservice inspections under inspection Program B in the Millstone Unit No. 2 l lSI Programs, it specifically clarifies that an inspection period may be reduced or extended by c much as one year to enable an inspection to coincide with a plant outage for any class of component. Code Case N-535 provides guidance to allow, among other things, adjustments to the inspection interval by as much as one year and performance of examinations for a successive inspection interval.
"NNECO has concluded that the use of this Code Case would provide an acceptable level of quality and safety, beer.sse the clarification of the requirements in the Code ;
Casa does not effect any phys' .11 process in the testing or examination requirements of : the ISI program. Therefore, the citaff's approval of this request will help us to relieve the uncertainty in trying to meet the existing requirements because they are not clear and do not provide the needed guidance that is contained in the Code Case.
"During the Third 10-Year interval NNECO plans to use the Code Case subject to the StafI's approval until such time that this Code Case is published in a future revisions of Regulatory Guide 1.147. At that time NNECO will follow all provisions in Code Case N-535, with limitations issued in Regulatory Guide 1.147, if any."
Evatustion-4nspection Program B of the Code requires inspection intervals 10 years in length, except as modified by IWA 2430(d), which allows an interval to be extenced or l reduced by as much as one year to coincide with an outage. The licensee proposes to apply the requirements of Code Case N-535 for the scheduling of intervals and examinations of Code Class 1,2, and 3 piping and components.
- - Code Case N-535 consists of four parts which can be summarized as follows
a) Each inspection interval may be reduced or extended by one-year. For extended intervals, neither the start or end dates nor the inservice inspection program for the successive interval need be revised. Thus, a successive interval may start prior to the end of the previous interval that was extended. b) Examinations performed to satisfy the requirements of the extended interval may be performed in conjunction with examinations performed to satisfy the i' requirements for the successive interval. However, examinations cannot be credited to both intervals. l 42
- o l 1.
c) Inspection periods may be extended or reduced to coincide with an outage. This adjustment shall not alter the requirements for scheduling inspection intervals. d) Examination records must identify which interval the examination was performed in. Part (a) of Code Case N-535 is the only change from current Section XI philosophy. The one year extension is independent of the plant operating cycle and two intervals can be enn concurrently during that year. Although slightly different from the current Code requirements, implementation of this Code Case does not change the number of examinations, acceptance criteria, or any other Code requirement, with the possible exception of distribution of examinations. However, this change would be slight and, therefore, the INEEL staff concludes that Code Case N-535 provides an acceptable level of quality and safety. Conclusions-Based on the evaluation above, it is concluded that the licensees' proposed alternative provides an acceptable level of quality and safety. Therefore, it is recommended that the use of Code Case N-535 be authorized pursuant to 10 CFR I 50.55a(a)(3)(i). The use of this Code Case should be authorized for the third interval at Millstone, Unit 2, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After thet time, the licensee must follow the conditions, if any, specified in the regulatory guide. l l e l 43
m
- 4. CONCLUSIONS Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee has determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code.
In these cases, the licensee has demonstrated that specific Section XI requirements are j impractical. Therefore, it is recommended that for Requests for Relief RR-89-01, RR ' 02, RR-89-03, RR-89-06, RR-89-07, RR-89-010, RR-89-11, RR-89-12 and RR-89-14, relief be granted as requested. Granting relief will not endanger life, 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. For Requests for Relief RR-89-04, RR-89-05, RR-89-08, RR-89-16, RR-89-18, and Proposed Alternatives N-416-1, N-4981, N-522, N-524, N-535 and N-546, it is concluded that, (a) the licensee's proposed alternatives will provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety. In these cases, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3). Request for Relief RR-89-17, should be authorized for Class 1 only. For Request for Relief RR-89-21, relief should be denied. Requests for Relief RR-13, RR 14, RR-8919 and RR-89-20 are evaluated elsewhere by the NRC staff. The licensee should continue to monitor the development of new or improved exasnination techniques. As improvements in these areas are achieved, the licensee should incorporate these techniques in the ISI program plan. Based on the review of the Milletone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice inspection Program, Revision 2, the licensee's response to the Nuclear Regulatory Comn.ission's RAl, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, no deviations from regulatory requirements or commitments were identified in the Millstone Nuclear Power Station, Unit 2, Third 10-Year Intervalinservice Inspection Program, Revision 2, with the exception of Code Case N-547, which is not acceptable as written, the volumetric examinations of reactor pret.sure vessel closure head studs, which have not been scheduled in accordance with Table IWB-2412, and as specified above. 44
s
- 4. CONCLUSIONS Pursuant to 10 CFR 50.55a(g)(6)(i), the licensee has determined that certain inservice examinations cannot be performed to the extent required by Section XI of the ASME Code. In these cases, the licensee has demonstrated that specific Section XI requirements are impractical. Therefore, it is recommended that for Requests for Relief RR-89-01, RR-89-02, RR-89-03, RR-89-06, RR-89-07, RR-89-010, RR-89-11, RR-89-12 and RR-89-14, relief be granted as requested. Granting relief will not endanger life, 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.
For Requests for Relief RR-89-04, RR 89-05, RR-89-08, RR-89-16, RR-89-18, and Proposed Alternatives N-416-1, N-498-1, N-522, N 524, N-535 and N-546, it is concluded that (a) the licensee's proposed alternatives will provide an acceptable level of quality and safety, or (b) Code compliance will result in hardship or unusual difficulty without a compensating increase in safety, in these cases, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3). Request for Relief RR-89-17, should be authorized for Class 1 only. For Request for Relief RR-89-21, relief should be denied. Requests for Relief RR-13, RR-14, RR-89-19 and RR-89-20 are evaluated elsewhere by the NRC staff. The licensee should continue to monitor the development of new or improved examination techniques. As improvemen!s in these areas are achieved, the licensee should incorporate these techniques in the ISI program plan. Based on the review of the Millstone Nuclear Power Station, Unit 2, Third 10-Year Interval inservice Inspection Program, Revision 2, the licensee's response to the Nuclear Regulatory Commission's RAI, and the recommendations for granting relief from the ISI examinations that cannot be performed to the extent required by Section XI of the ASME Code, the following deviations from regulatory requirements or commitments were identified in the Millstone Nuclear Power Station, Unit 2, Third 10-Year Interval inservice inspection Program, Revision 2.
- 1) Request for Relief RR-89-21 is denied.
- 2) Request for Relief RR-89-17 is approved for Class 1 only and denied for Class 2.
- 3) In accordance with Table IWB-2412, volumetric examinations of reactor pressure vessel closure head studs should be performed each inspection period, as discussed in Section 2.2.2 of this report.
- 4) Code Case N-547 is not acceptable as written and should be removed from the ISI Program.
i l l
m a N in Request for Relief RR-89-10 (Memo # 25203-ER-97-0018)
- 15. Letter dated September 22,1994, E. A. DeBarba (NNECO) to Document Control desk, requesting authorization to use Code Case N-416-1.
- 16. Letter dated December 16,1994, J. F. Opeka (NNECO) to Document Control Desk requesting authorization to use Code Case N-498-1.
- 17. Letter dated March 20,1996, F. R. Dacimo (NNECO) to Document Control Desk, requesting authorization to use Code Case N-546.
- 18. Letter dated June 2,1995, S. E. Scace (NNECO) to Document Control Desk requesting authorization to use Code Case N-524.
- 19. Letter dated August 24,1995, J. F. Opeka (NNECO) to Document Control Desk requesting authorization to use Code Case N-535.
l 46
O o NkC term 333 i U S Nuclear Regulatory Comnummn 1. REPORTNUMBER NPCM i102 (Assigned by NRC, Add Vol., Supp,, Rev., and 320s.3202 Addenkm Numbm,ifany) BIBLIOGRAPHIC DATA SIIEET INEEL/ EXT-97-01149 l 2. TirLE AND SUBTITLE 3. DATE REPORT PUBLISIIED t ! Technical Evaluation Report on the Third 10-Year Interval inservice M nth Year inspection Program Plan: Northeast Nuclear Energy Company, March 1998 Millstone Nuclear Power Station, Unit No. 2 4. FIN OR GRANT NUMBER Docket Number 50-336 JCN No. J2229 (Task order TWA A221
- 5. AUTilOR(S) 6. TYPE OF REPORT M. T. Anderson Technical l E. J. Feige A. M. Porter 7. PERIOD COVERED (laclusive Dates)
- 8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; ifcontrador, provide name and mailing addes)
INEEL/LMITCO P. O. Box 1625 Idaho Falls, ID 83415-2209 l l.
- 9. SPONSORING ORGANIZATION .NAME AND ADDRESS (If NRC, type"Same as above'*;ifcontractor, provide NRC Division,03im or Region U.S.
Nuclear Regulatory Conurussion, and mailing ad&ess) l Civil and Geosciences Branch Office of Nuclear Regulatory Commission U. S Nuclear Regulatory Commission Washington D. C. 20555
- 10. SUPPLEMENTARY NOTES
- 11. ABSTRACT (200 words or less)
This report presents the results of the evaluation of the Mi// stone Nuclear Power Station, Unit 2, Third TO-Year /nterval /nservice /nspection Program, Revision 2, submitted by letter dated July 2,1996, including the requests for relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI requirements that the licensee has determined to be impractical. The M/// stone Nuclear Power Station, Unit 2 Third 10-Year /nterva/ /nservice /nspection Program, Revision 2, is evaluated in Section 2 of this report. The ISI Program Plan is evaluated for (a) compliance with the appropriate edition / addenda of Section XI, (b) acceptability of examination sarnple, (c) correctness of the application of system or component examination exclusion criteria, and (d) compliance with ISl-related commitments identified during previous Nuclear Regulatory Commission reviews. The requests for relief are evaluated in Section 3 of this report. j 12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchm in locating the repon) 13. AVAILABILrrY STATEMENT i Unlimited l l 14. SECURITY CLASSIFICATION (Dus page) Unclassified (nus repon) Unclassified
- 15. NUMBER OF PAGES
- 16. PRICE L_________-______________________________ _ _ _ _ _ _ _ . _ _ _ _}}