ML20052F238
| ML20052F238 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/03/1982 |
| From: | Bordine T CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8205120264 | |
| Download: ML20052F238 (83) | |
Text
{{#Wiki_filter:% e 3 C0HSu m 8IS +y POW 8r Company General offices: 1945 West Parnale Road. Jackson, MI 49201 * (517) 788-0550 May 3, 1982 ,p d.b[iD. y sNy x' / N D c-s,, j,,% ({q l Dennis M Crutchfield, Chief '8, Operating Reactors Branch No 5 -'7 ~ f 7 Nuclear Reactor Regulation h
- 's US Nuclear Regulatory Commission
\\., f7 / Washington, DC 20555 ',x /\\ lig },& DOCKET 50-155 - LICENSE DPR BIG RCCK POINT PLANT - RESPONSE TO INSERVICE INSPECTION REQUEST FOR ADDITIONAL INFORMATION AND REQLTSTS FOR RELIEF FROM SPECIFIC ASME CODE REQUIREMENTS NRC letter dated February 17, 1982 included a listing of previous submittals from Consumers Power Company which requested inservice inspection relief. The letter requested our review of the list of submittals as well as identifica-tion of any additional relief requests for both exempted components under the terms of the ASME Code and relief from the requirements of the Code. In addition, information pertaining to our November 17, 1980 submittal was later requested by NRC letter dated March 3, 1982. Consumers Power Company has revised our November 17, 1980 submittal in response to the NRC February 17, 1982 request for identification of additional relief and the March 3, 1982 request for additional information. This docu-ment is enclosed as Attachment 1. Our detailed response to NRC letter dated March 3, 1982 is provided as Attachment 2 of this letter. This letter ful-fills our commitment made by our submittal of March 19, 1982 which requested a 30 day extension in order to allow us sufficient time to prepare an adequate response. Aool L.. s Thomas C Bordine f Staff Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector-Big Rock Point Attachments-2 nr0582-0001a142 8205120264 820503 gDRADOCK 05000155 m m
r-March 4, 1982 A'ITACHMENT I Revision 2 PROPOSED REQUEST FOR RELIEF FROM PROVISIONS OF ASME B&PV CODE - SECTION XI, PURSUANT TO 10 CFR 50 SECTION 50.55a(g)(6)(i) rp0382-0621a-42-48 1
e o March 4, 1982 Revision 2 BIG ROCK POINT RELIEF REQUEST IMPACT STATEMENT Request for Relief From Provisions of ASME B&PV Code Section XI Purusant to 10 CFR 50.55a(g)(6)(1) The relief requests present the detailed information required to appropriately implement the requirements of 10 CFR 50.55a(g)(4) as specified in the newly revised Technical Specifications Section 9.0, which will become effective, per 10 CFR 50.55a, at the end of ISI 6 (February 1979 refueling outage). An integral part of each relief request is the " Alternative Inspection or Test" which describes why the proposed alternate examination (s) establish pressure, boundary integrity in a manner equivalent to the actual examination method specified in Section XI. The NRC must review these relief requests and write an impact statement noting that the pressure boundary integrity will be pro-tected as noted above. With this chain of events, it can safely be concluded that the design integrity and operational characteristics will be maintained as adequately with the proposed alternate examination schedule as with the Section XI specified examination schedule. i rp0382-0621a-42-48
e e March 4, 1982 Revision 2 TABLE OF CONTENTS Section Title Page 1.0 Description of the Big Rock Point I I-14 40-Year Inservice Inspection Master Plan 2.0 Applicable Editions and Addenda of II II-51 ASME Boiler and Pressure Vessel Code - Section XI 3.0 Relief Requests III III-64 Subsection A - Class 1 Subsection B - Class 2 Subsection C - Class 3 Subsection D - Pressure Tests Subsection E - General s rp0382-0621a-42-48 I-1
e M:rch 4, 1982 Revision 2 SECTION
1.0 DESCRIPTION
OF THE BIG ROCK POINT 40-YEAR INSERVICE INSPECTION MASTER PIAN rp0382-0621b-42-48 I-2
i March 4, 1982 Revi.'7n 2 DESCRIPTION OF THE BIG ROCK POINT 40-YEAR INSERVICE INSPECTION MASTER PLAN I. INTRODUCTION This document is a plan for inservice examinations (ISI) to be performed over a 40-year period on Class 1, Class 2 and Class 3 components and systems (and their supports) of Consumers Power Company's (CP Co) Big Rock Point Nuclear Power Plant. A. HISTORICAL BACKGROUND The initiation of commercial service for Big Rock Point was on December 8, 1962. The start of the first 10-Year Inservice was on January 1, 1972. During the first two 40-month intervals, in order to comply with Section 9 of the Technical Specifications of the " Operating License (DPR-6) for Big Rock Point Nuclear Plant," which discusses ISI requirements of Class 1 components and systems, the nondestructive examinations were performed to satisfy the requirements of the ASME Section XI Code,.'971 Edition including the Winter 1972 Addenda. In February 1976, the NRC ammended Paragraph 55a(g) of 10 CFR 50 to require nuclear power plants to upgrade their Technical Specifications in the areas of the ISI requirements and the functional testing of pumps and valves. By amending Paragraph 55a(g) and by invoking Regulatory Guide 1.26, the NRC required nuclear plants to upgrade their systems to include not only Class 1 systems, but also Class 2 and Class 3 systems in their ISI programs. B. UPGRADING CRITERIA The construction of this plan was based on the following documents: 1. Big Rock Point Nuclear Plant's Piping and Instrument Diagrams and Plant Q-List. l l 2. Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for l Inservice Inspection of Nuclear Power Plant Components," as identi-fled in Section 2.0 of this document. 3. USNRC " Rules and Regulations, Title 10, Chapter 1, Code of Federal Regulations-Energy," Part 50.55a. 4. Applicable sections of Section 9.0 of the Technical Specifications of the " Operating ;icense (DPR-6) for the Big Rock Point Nuclear Power Plant." l l Components were scheduled for examination in accordance with the plant's l Technical Specifications and Section XI of the code. Examinations are l conducted in accordance with Section 2.0 of this document. l l l i rp0382-0621b-42-48 I-3 l L
e March 4, 1982 Revision 2 C. REFERENCES 1. 10 CFR 50.55a(g). 2. Operating License DPR-6, Technical Specifications for the Big Rock Point Plant, Docket 50.155, Section 9.0, as modified by Technical Specifications change request dated July 27, 1978. 3. ASME Boiler and Pressure Vessel Code, Section XI. 4. Consumers Power Quality Assurance Program Procedure for Operations, No 10-52. 5. Consumers Power Nuclear Plant Support Department Procedure NPS-05. D. GENERAL 1. This inservice inspection plan for the 4 year inservice in-tervals has been developed by Consumers Power Company for use at Consumers Power Company's Big Rock Point Nuclear Power Plant. This plan incorporates all periodic surveillance requirements of Refer-ences C-2 and C-3 for the 40-year service lifetime. The length of the second three and one-third year period has been extended by approximately six to seven months to permit the next inspection to be coincident with the scheduled refueling outage. Therefore, the start of the third interval, which is the date the update becomes effective, is approximately April 1, 1979. 2. Responsibility for the maintenance of this plan and the development of subsequent plans rests with the Nuclear Plant Support Department as defined in Reference C-4. Only copies marked " Controlled" will be issued revisions. 3. In view of the fact that the Big Rock Point Nuclear Plant went into commercial operation well before the issuance of Section XI of the ASME B&PV Code, the Inspection Access requirements of IS-142,1971 Edition, were not available to impact the plant design parameters. The Technical Specifications / Relief Request section of this plan details specific code requirements which cannot be met. 4. Examination methods delineated in the tables are intended to be representative of past ISI practice or of preservice methods uti-lized. In either case, it should be recognized that either UT or RT are acceptable volumetric exams and either PT or MT are acceptable surface exams. Unique weld joint parameters may, of course, dictate more restrictive selection criteria; eg, high background radiation I will preclude RT, stainless materials will preclude MT, etc. It is intended that the process which selects exam methods for inspections under this plan treat UT and RT as interchangeable and PT and NT as interchangeable with consideration given to past practice in light of the reproductibility of results. I The following table summarizes the inspections to be performed at the l Big Rock Point Nuclear Plant per the updated ISI 40-Year Plan: rp0382-0621b-42-48 I-4 l
~ Ita rch 4 1982 Rivision 2 TABLE l-1 BIG ROCK PolNT INS MYlCE INS _P_[CTION PROGRAM Class } Item Number Section XI Section XI Number To Be Reller Request Compon_enJs To Be Examined Eem_lio__ Cat Totals inspected Number Bean ot_yessel Circuarerential Welds 81.11 0-A 2 0 RR-A1, RR-A2 Longitudinal Welds 81.12 B-A 4 2 RR-A1, RP-A2 RV itead Meridional Welds B1.22 B-A 8 4 RR-A3 RV ifead Circumferential Welds 81.21 B-A 2 1 RR-A3 Vessel-To-Flange Weld 81.30 B-A 1 1 llead-To-Flange Weld 81.40 B-A 1 1 Nozzle-To-Shell Wolds 83.90 B-D 21 11 RR-A4, A10, A6, A7 Nozzle inside Radius B3.100 B-D 15 10 RR-A4, A10, A6, A7 Control Rod Drive Nozzles (J-Welds) B4.13 8-E 32 32 Nozzle-To-Safe End Welds 85.10 B-F 15 10 RR-A9, A10, All, A12, A13 Closure Ifead Studs 86.20 B-G-1 42 42 Closure Head Nuts 86.10 B-G-1 13 2 42 Closure liead Washers B6.50 B-G-1 42 42 Flange Ligaments 86.40 B-G-1 42 42 Nozzle Flanges 87.10 B-G-2 10 10 Vessel Supports 88.10 B-H 20 0 RR-A14, AIS, A16 Vessel Interior B13.10 B-N-1 All All Interior Attachment Welds B13.20 B-N-2 A11 See RR RR-A17, A18 Core Support Structure B13.21 B-N-3 All All yea _t_pchartge_rs/ Steam _ Gene ralogs A. [me rgenc1Conde_r!se_r Circumferential Welds B2.51 B-B 4 4 rp0382-0621c-42-48 l-5
Ma rch 4, 1982 Revision 2 1 TABLE l-1 (Contd) Item Number Section XI Section XI Number To Be Relief Request Components To pe_ Examined item No Cat Totals _ inapected Mumbe r 3 Nozzle-To-Shell Welds 83.150 B-D 4 4 Nozzle inside Radius 83.160 B-D 4 See RR RR-A19 Bolting B7.40 B-6-2 2 2 B. Regenerative Heat Exchangers (4) Circuarerential Welds B2.51 B-B 16 16 Nozzle Welds 83.150 B-D 16 16 Nozzle inside Radius 83.160 B-D 16 0 RR-25 Supports 88.40 B-la 16 16 C. Non recene ra t ive Hea t Exchange r Ci rcuarerential Welds 82.51 B-B 4 4 Nozzle Welds 83.150 B-D 2 0 RR-A23 Nozzio inside Radius 83.160 B-D 2 0 RR-A23 Vessel Supports B8.40 B-ll 2 2 D. Steam Drum Ci rcumferential Wolds 82.51 B-B 4 See RR RR-A20 Longitudinal Welds 82.52 B-B 6 6 RR-A20 Nozzle-To-Shell Welds and Nozzle inside Radius 83.130 B-D 34 17 RR-A21 Nozzle-To-Safe r7d Welds B5.30 B-F 10 10 RR-A26 Bolting B6.90 B-G-1 6 6 Vessel Supports B8.30 B-H 18 4 RR-A29 I E. Clean-Up A min Tank Ci rcumfe ren t ia l Welds 82.5i B-B 3 3 I longitudinal Wolds 82.52 B-B 2 2 i Nozzle-To-Vessel Weld B3.150 B-D 5 5 rp0382-0621c-42-48 l-6 I
--_ - _.____ _ _ _ _ = Ma rch 4, 1902 R visitn 2 TABLE l-1 (Contd) Item Number Section XI Section XI Number-To Be Relief Request Components To Be [wamined _ Item No_ Cat lotals Inspected Numbe r Nozzle inside Radius B3.130 B-D 5 0 RR-A24 Bolting B7.40 B-6-2 1 1 Vessel Supports B8.40 8-H 3 3 ffping 4 System Identirication s. Liquid Poison System (LPS) 2. [mergency Condenser System ([CS) 3. Shutdown Cooling System ( SCS) 4. Reactor Clean-Up System (RCS) 5. Main Steam System (MSS) 6. Main Reci rc System (MRS) 7. Core Spray System (CSS) 8. Redundant Core Spray (ROC) 9. Control Rod Drive (CRD) 10. f eedwater System (f WS) 11. Reactor Depressurization (RDS) Dissimila r Meta l Welds 85.50 8-F RR-A31, A32 LPS 2 2 SCS 2 1 RR-A30 RCS 4 4 MSS 9 9 l CSS 1 1 RDS 19 10 Total 28 27 Pressure Retainine Boltine (< 2"1 B7.50 B-G-2 RCS 4 4 MSS 14 14 5 CSS 1 1 RDC 2 2 RDS _h _h Total 25 25 i 1 rp0382-0621c-42-48 l-7
Ma rch 4, 1982 R2visicn 2 TABLE l-1 (Contd) Ites Number Section XI Section XI Numbe r To Be Roller Request Components To Be Examined item No Cat Iqtals_ Inspected Number Ci rcuarerent ia l Welds 12 4") 89.11 B-J ECS 56 16 SCS 43 11 RR-A32, A33 MSS 37 10 RR-A33 MRS 124 39 RR-A32 CSS 18 6 RDC 31 9 IWS 15 4 RDS E J Total 353 ?O3 [pongitudinaI Weids f 3 4"l 89.12 B-J MRS 4 2 Ci rcumferentia l Welds f < 4") 9.21 B-J LPS 76 22 RR-A32, A34 RCS 183 52 RR-A27, A28 MSS 63 18 RR-A32 CSS 4 0 RR-A32, A33 CRO J J Total 329 95 Rocket Welds 89.40 B-J RCS 50 18 MSS 181 45 RR-A35 CRD 81 21 RR-A35 RDS JD 1} Total 362 97 Branch Connections f > 2") B9.31 B-J MRS 9 5 Branch Connections iI 2"1 89.32 B-J [CS 1 1 MSS 1 1 CSS 1 0 RDS 3-TotaI 8 5 rp0382-0621c-42-48 l-8
1 Ma rch 4, 1922 R3 vision 2 TABLE l-1 (Contd) Item Number Section XI Section XI Numbe r To Be Roller Request Components To Be Examined _Ilem No . Cat Totals inspected Number Melded Suppot13 B10.10 B-K-1 ECS 5 2 LPS 3 1 SCS 2 O RCS 7 4 MSS 3 1 MRS 28 4 RR-A36 CSS 1 0 CRD 1 1 RDS _3 _J RR-A37 Total 54 14 pechanjcally Attached Support 811.10 B-K-2 ECS 10 10 LPS 9 9 RR-A38 SCS 8 5 RR-A38 MSS 18 18 MHS 60 58 RR-A36 CSS 5 RDC 4 4 CR0 13 13 FWS 3 3 RDS 15 15 RCS _19 _10 Total 175 170 Puggs Bolts, Studs and Nuts 87.60 B-C-2 3 2 Component Supports 811.20 B-K-1 11 11 Pump Casing Welds B12.10 B-L-1 3 2 RR-A39 Pump Casing B12.20 B-L-2 3 2 yalves Component Supports B11.30 B-K-2 1 1 Bolts, Studs and Nuts 87.70 B-G-2 56 56 Bolting B6.230 B-G-1 8 8 rp0382-0621c-42-48 l-9
O. Ma rch 4, 1982 P4 vision 2 TABLE I-1 (Contd) Item Number Section XI Section XI Numbe r To Be Roller Request Components To Se Examined item No Cat lqtals Inspected Number Valve Body Welds 812.30 B-M-1 27* 27 Valve Body 812.40 8-M-2 52 46 RR-A40 i e i i. eExistence or these welds requires verification, rp0382-0621c-42-48 1-10 I
0 Ma rch 4 1902 Rsvision 2 TABLE l-1 (Contd) BIG ROCK POINT IN}[RVICE INSPECTION PROGRAM Class 2 Item Number Section XI Section XI Number To Be Relier Request Components in De Fxamined item No Cat Totals Inspected Number Yem elg Class 2 vessels to be examined include: 1. liigh-Pressure Itea ter 2. Shutdown lleat Exchangers (2) 3. Core Spray lleat Exchanger 4 Liquid Poison Tank Pressure Vessel Inspections per Table IWC-2600-1 f.I.11L9 P 1 Control Rod Drive (CRD) RR-BI, d2 2. Cnre Spray System (CSS) 3. Emergency Condenser ( ECS) 4. Feedwater System ( FWS) 5. Liquid Poison System (LPS) 6. Main Steam System (MSS) 7. Post-Incident Cool ing ( PB S) 8. Reactor Clean-Up System (RCS) 9. Shutdown Cooling System (SCS) Cirguererential Welds f I_]/2") C5.11 C-F CSS 45 45 PIS 32 32 SCS 71 71 C i rcum fe ren t i a l Welds ft > I/2") C5.21 C-F CRD 18 18 FWS 65 65 MSS 20 20 Pipe Hranch Connections C5.31 C-F FWS 1 1 rp0382-0621c-42-48 I-11
Ma rch 4, 1982 Ravisicn 2 TABLE l-1 (Contd)- Item Number Section XI Section XI Number To Be Relief Request Components To se Examined item No Cat Totals inspecttg Numbe r Co*R2nent Suppo rt s C3.50 C-E CRD 1 1 CSS 8 8 fWS 15 15 MSS 3 3 PBS 4 4 SCS 10 10 Integrally Weided suggort C3.40 C-E IWS 5 5 MSS 3 3 SCS 2 2 Pump 3 Class 2 pumps required for inspection: 1 Shutdown Heat Exchan9er Pumps (2) Yalvgs . The following valves are subject to inspection: '1 10-fWS-201 - CV 4000 2. - FW 24 3. - FW 25 4, 6-fWS-204 - IW 300 5. - FW 6 6. 6-FWS-205 - FW 301 F. - FW 3 8. 12-MSS-201-CV-6599 9. 10-MSS-204-MO-7067 10. -CV-4200 II. 6-PIS-201 - IP-29 12. - PI-302 13. 6-PIS-203 - SW-3 14. 8-SWS-201 - MO-7059' 15. 8-SWS-202 - MO-7057 16. 8-SWS-203 - SC-301 17. - SC-4 18. 6-SWS-204 - SC-3 19. 6-SWS-205 - SC-300 20. - SC-2 21, 6-RSS-206 - SC-1 rp0382-0621c-42-48 l-12
a 4 Ma rch 4, 1982 RsvisiCn 2 TABLE l-1 (Contd) Item Number Section XI Section XI Number To Be Reller Request Compo.nents To De Examined item No Cat Totals inspected Number i Elains Class 3 piping systems subject to NDE requirements are as rollows: 1. Condensate Pump Suction (CPS) 2. Condensate Tank Piping (CTP) 3. Hesctor Cooling Water (RCW) 4. Se rv ice Wa te r Sys tem ( SWS) 5. Heactor Shutdown System (SCS) Gomagnent_ Supports & Restraints D1.2 D-A D2.2 D-B RR-C1 D3.2 D-C i CPS O O CTP O O HCW 69 69 SWS 5 5 SCS 1 1 f 4 rp0382-0621c-42-48 l-13
Ma rch 4, 1982 Rivisicn 2 TABLE l-1 (Contd) 31C ROCK POINT INSERVICE INSPECTION PROGRAM System Pressure Tests item Number Section XI Section XI Number To De Reller Request Components To Be Examined item No Cat Totals inspected Number Class 1 Vessels, Piping, Pumps and Valves B-P RR-D1 Applies to All Three Classes or Pressure tests to be completed: Systems 1. System teakage Test ( IWO-5221) 2. System flydrostatic Test (LWD-5222) C_Lans_2 Vessels, Piping, Pumps and Valves C-H Pressure tests to be completed: 1 System Pressure Test (IWC-5221) 2. System Hydrostatic Test (1WC-5222) Class 3 Vessels, Piping, Pumps and Valves Pressure tests to be completed: 1. System Funct ion Test ( IWD-5222) D-A, 2. System Inservice Test (IWD-5221) D-B, 3. System Hydrostatic Test D-C (IWD-5223) f I j rp0382-062tc-42-48 l-14 I l l
March 4, 2982 Revision 2 SECTION 2.0 APPLICABLE EDITIONS AND ADDENDA 0F ASME BOILER AND PRESSURE VESSEL CODE - SECTION XI s l l l 1 l l \\ l t rp0382-0621d-42-48 II-15
March 4, 1982 Revision 2 2.0 APPLICABLE EDITIONS AND ADDENDA 0F THE ASME BOILER AND PRESSURE VESSEL CODE - SECTION XI Pursuant to Paragraph 50.55a(g) of 10 CFR Part 50, amended by the US Nuclear Regulatory Commission effective November 1, 1979, the inservice examination requirements applicable to nondestructive examination and system pressure testing at the Consumers Power Company Big Rock Point Nuclear Plant are based upon the rules set forth in the 1977 Edition of Section XI of the ASME Boiler and Pressure Vessel Code, including Addenda through Summer, 1978. As permitted by the amendment, the Summer 1975 Addenda may be applied in lieu of later Addenda for determining the extent and frequency of in-service inspection of Class 1 and Class 2 pipe welds. The amendment also requires the use of the Summer 1975 Addenda for establishing the pipe welds to be examined in the residual heat removal system, the. emergency core cooling system and the containment heat removal system. Consumers Power Company elects to adopt Subsection IWF - requirements for Class 1, 2 and 3 component supports - issued in the Winter 1978 Addenda to serve as guidance for inservice inspection of snubbers. This Subsec-tion, IWF, provides improved inspection guidance not contained in earlier Code Editions and Addenda. l rp0382-0621d-42-48 II-16
Mirch 4, 1982 Revision 2 THIS PAGE DELETED t i I I i f i rp0382-0621d-42-48 II-17 i I l L
March 4, 1982 Revision 2 CODE USE FOR THIRD PERIOD FIRST INSERVICE INSPECTION INTERVAL Consumers Power Company Code Applicability Statement 1
Subject:
Piping Examination I. The basic requirements for ultrasonic examination of piping shall be in accordance with the following: Type of Material Nominal Thickness Applicable Code Austenitic and fer-Greater than.4 to Article 5 of ASME Section V ritic piping and 6 inches with Adienda through dissimilar metal Summer 1978 welds Austenitic and .1 to.4 inches Appendix III with Supple-ferritic piping and ment 7 of ASME Section XI dissimilar metal with Addenda through welds Summer 1978 II. The following modifications of the requirements of Article 5, Section V, and Appendix III, Section XI, 77S78 Addenda are applicable. A. The following Code Cases shall be utilized: 1. Code Case N-211, Recalibration of Ultrasonic Equipment Upon Change of Personnel 2. Code Case N-234, Time Between Ultrasonic Calibration Checks, Sec-tion XI, Division 1 3. Code Case N-235, Ultrasonic Calibration Checks per Section V, Section XI, Division 1 B. For examinations conducted in accordance with Article 5 of ASME Section V, the basic calibration block for production material thickness 1.0 inch or less shall be as follows: The basic calibration block shall have the same nominal thickness as the production material or no more than 25% less than the nominal production material thickness or closer in thickness to the production material than the 3/4-inch alternate thickness allowed by Article 5. This modification will assure a more accurate calibration than the basic calibration block design allowed by the Code. C. Subparagraph T-535.1(d) (1) and (2) of Article 5 requires that transfer (attenuation compensation) be accomplished between the production material and the basic calibration block and a correction made for the difference. In accordance with Code Interpreta-tion V-78-01 and Code Case 1698 (N-92), the transfer method will not rp0382-0621d-42-48 II-18
o M2rch 4, 1982 Revision 2 be used. Attenuation measurements will be recorded and considered during analysis and evaluation of indications; however, no attempt will be made by the examiner to compensate for any observed differ-ence before or during the ultrasonic examination. D, For examination conducted on welds, the following requirements from Appendix III, Section XI, 77S78, shall be used in lieu of Para-graph T-530, Section V, 77S78. III-2300 Written Procedure Requirements i III-2400 General Examination Requirements III-3100 Instrument Calibration (to include Supplements 5 and 6) III-3210 System Calibration - General Requirements III-3300 Calibration Confirmation III-3500 Calibration Data Record III-4000 Examination E. Data shall be recorded as per IWA-2232(b). i l l rp0382-0621d-42-48 I!-19
March 4, 1982 Revision 2 s CODE USE FOR THIRD PERIOD FIRST INSERVICE INSPECTION INTERVAL Consumers Power Company Code Applicability Statement 2
Subject:
UT Piping Calibration Blocks 1 I. All new piping ultrasonic calibration blocks will be procured, designed and fabricated as follows: 1 - A. Material Basic calibration block material shall conform to the requirements of Paragraph III-3411, Section XI. B. Radius of Curvature Basic calibration blocks shall conform to the radius of curvature required by Paragraph III-3410, Section XI. C. Thickness Basic calibration blocks shall conform to the requirements of Para-graph III-3410. D. Calibration Reflectors 1. Thickness: .10 inches to.40 inches Calibration reflectors shall conform to the requirements of Paragraph III-3430, Section XI. 2. Thickness: Greater than.4 inches to 6 inches. t Calibration reflectors shall be holes cnd shall conform to the requirements of Paragraph T-533 (a), Article 5, Section V, 77S78, ASME B&PV Code. i i k i 1 rp0382-0621d-42-48 II-20 . -,. ~ -, ,,,n ,--.n. n. .n-
March 4, 1982 Revision 2 SECTION 3.0 Requests for Relief Pane Subsection A - C? ss 1 III-22 Subsection B - Class 2 III-58 Subsection C - Class 3 III-61 Subsection D - Pressure Testing III-63 (Classes 1, 2 and 3) Subsection E - General III-64 rp0382-0621d-42-48 II-21
March 4, 1982 Rsvision 2 SUBSECTION A - CLASS 1 Reactor Vessel Category: B-A Item No: Bl.11, Bl.12, Bl.21, Bl.22 LONGITUDINAL AND CIRCUMFERENTIAL SHELL WELDS, MERIDIONAL AND CIRCUMFERENTIAL HEAD WELDS RR-Al i Vessel Core Region Basis Statement Longitudinal and circumferential welds in the core region are not accessible for volumetric examination Remarks The reactor vessel is closely surrounded by concrete so examina-tion from the outside is not possible. Inner wall of reactor vessel is inaccessible in the core region due to the presence of a thermal shield, which is immovable. Design clearance between the thermal shield and reactor vessel wall is 1.65 inches. The thermal shield extends 66.50 inches below the beltline circumfer-ential weld. Alternative Inspection or Test Each refueling outage, a hydrostatic test (prestart-up hydro) is performed at 1.1 times the operating pressure. This is more con-servative than the prestart-up leak test at operating pressure required by the Code (IWB-5221). Inspection of the six riser nozzles and vessel-to-flange weld will provide some indication of reactor vessel integrity, along with results of vessel coupon testing. During operation, a failure of any weld in the reactor vessel would be readily detectable by level indication, dew cell indication, makeup water flow and/or temperature indication in adequate time for safe shutdown. References CE Drawings E-201-801-5 and E-201-794-8 GE Drawing 141F797 Bechtel Drawing 0740G20128, Rev C (C-128) rp0382-0621d-42-48 II-22
March 4, 1982 Revision 2 Implementation Relief requested. Relief was previously approved May 4,1972 pursuant to April 21, 1972 request. RR-A2 11 Vessel Shell (Above Beltline Cire Weld) Basis Statement Longitudinal welds in vessel shell are generally inaccessible for volumetric examination. Remarks Essentially 60* of the length of the longitudinal welds are inac-cessible for volumetric inspection because of the thermal shield and core spray sparger and baffle supports. Access from the out-side is not possible due to concrete wall. Vessel shell-to-flange weld is accessible for inspection and has been ultrasonically examined. Alternative Inspection or Test Volumetric examination of accessible areas (UT - mechanized). Hydro (See RR-A1.) References CE Drawings E-201-794 and E-201-793 l l Bechtel Drawings 0740G20128, Rev C Implementation Relief requested on unaccessible areas of longitudinal welds as previously granted. RR-A3 iii Vessel Bottom Head Basis Statement l Meridional and circumferential welds in the bottom head are inaccessible for volumetric inspection. Remarks Meridional welds are inaccessible from inside reactor because of the core support plate. Poor geometry due to the 40 penetrations on the bottom head makes the meridional welds inaccessible from the outside. The circumferential shell-to-bottem head weld is not j accessible with existing equipment from the inside and the I rp0382-0621d-42-48 II-23
Mtrch 4, 1982 Revision 2 ccncrete wall prevents inspection from the outside. The lower head circumferential weld (793-1) is also inaccessible from inside because of the core support plate structure and from the outside because of the CRD penetrations. An access port, fabricated during the February 1982 outage through the high-density aggregate trays, may provide extremely limited access to a small portion of these welds. Further access engineering in 1983 will determine if the port lies in the vicinity of these welds and if examination of portions of the welds is feasible. Alternative Test or Inspection Hydro (See RR-A1.) References CE Drawings E-201-794 and E-201-793 Bechtel Drawing 0740G20128, Rev C (C-128) Implementation Relief requested as previously granted. Category: B-D Item No: B3.90, B3.100 PRIMARY N0ZZLE-TO-VESSEL WELDS AND N0ZZLE INSIDE RADIUS SECTION RR-A4 1 20-Inch Recirculation (Recire) Nozzles (796-1A. B) Basis Statement These nozzles are inaccessible for 100% volumetric examination. Remarx. i l l Concrete shield wall and high-density aggregate trays prevent full l access to these nozzles from the outside. Diffuser plates prevent volumetric inspection of the nozzles from the inside. During the inlet baffle repair in 1979, a visual inspection (remote) was com-pleted on these nozzles with no indications noted. Approximately one-third of the external surface of one nozzle can be accessed through an access port fabricated in the high-density aggregate trays during the February 1982 outage. l l l l rp0382-0621d-42-48 II-24
March 4, 1982 Revision 2 Alternative Inspection or Test Hydro (See RR-A1.) Volumetric (UT) limited to approximately one-third of external surface of Nozzle 796-1B. Remote visual examination on accessible internal surfaces. References CE Drawings E-230-791-2, E-201-794-8, E-201-795-5 Bechtel Drawing 0740G20128, Rev C (C-128) Implementation Relief requested on nonaccessible areas. RR-A5 11 (Deleted) RR-A6 11 3-Inch Notzles (1) Core Spray 795-6 (2) Poison Inlet 796-6 Basis Statement Volumetric techniques are not available to inspect these nozzles. Remarks The poison inlet nozzle is located within high-density aggregate trays and is not acces ible. Access to the tray area is further prevented by 32 - 3/4-2 ach CRD hydraulic lines and incore detector lines located just below the trt.ys. Reactor vessel exterior cooling lines also restrict access to the poison inlet nozzle. Core spray nozzle access from outside of reactor vessel is pre-vented by concrete wall. A mechanized volumetric inspection is not possible from the inside because of the thermal sleeves present. During the baffle repair in 1979, these nozzles were examined visually (remote) with no indications noted. Attempted removal of the core spray nozzle sleeve in 1979 failed to dislodge it. Alternative Inspection or Test Hydro (See RR-A1.) Remote visual examination of accessible internal surfaces. rp0382-0621d-42-48 II-25
March 4, 1982 Revision 2 References CE Drawings E-201-795-5 and E-230-791-2 Bechtel Drawing 0740G20128, Rev C (C-128) Implementation Relief requested as previously granted. RR-A7 111 8-Inch Shutdown Unloading Nozzle (795-15) Basis Statement This rozzle is not accessible for mechanized ultrasonic inspection. Remarks A direct manual volumetric examination is not possible due to in-accessibility of nozzle. Interference with the core spray sparger prevents use of the mechanized ultrasonic device. During the baffle repair in 1979, this nozzle was visually (remote) inspected with no indications noted. Alternative Inspection or Test Hydro (See RR-A1.) Remote visual examination of accessible internal surfaces. References CE Drawing F-230-791-2 GE Drawings 212E456, 104R175, Sheet 2 Bechtel Drawing 0740G20128, Rev C (C-128) Implementation Relief requested as previously granted. RR-A8 iv 3-Inch Instrument Nozzles (955-1A-E) (This request has been deleted due to equipment developments.) rp0382-0621d-42-48 II-26
March 4, 1982 Revision 2 Category: B-F Item No: B5.10 N0ZZLE-TO-SAFE END WELDS ~ RR-A9 1 20-Inch Recire Nozzles (796-1A, B) Basis Statement Refer to RR-A4 and RR-A32. Remarks i Refer to RR-A4. Alternative Inspection or Test Surface and volumetric examination limited to approximately one-third of external surface of Nozzle 796-1B. References Refer to RR-A4. Implementation Refer to RR-A4 RR-A10 11 14-Inch Steam Outlet Nozzles (795-11A through 11F) Basis Statement These nozzles are inaccessible for surface examination from outside due to concrete walls which enclose these nozzles. Remarks During refueling, these nozzles are submerged in the shielding water making die penetrant or magnetic particle testing impossible.* A mechanized ultrasonic examination is performed underwater along with the prestart-up hydrostatic test. These nozzles are also visually inspected, also underwater, per Code requirements.
- Total core unloading and 1 to 2 R/hr fields also make surface examination from inside the nozzles impossible.
rp0382-0621d-42-48 II-27
March 4, 1982 Revision 2 Alternative Inspection or Test (a) Mechanized Ultrasonic Inspection (b) Visual (VT-1) (c) Prestart-Up Hydro (See RR-A1.) References CE Drawing F-230-791-2 Bechtel Drawing 0740G20128 (C-128) GE Drawings E-201-794-8, E-201-195-5 Implementation Relief requested from performing surface examination. Ultrasonic testing and hydrostatic testing will be completed. RR-All lii 8-Inch Shutdown Unloading (795-15) Basis Statement See RR-A7. Remarks See RR-A7. Alternative Inspection or Test See RR-A7. l l References i See RR-A7. l Implementation See RR-A7. l RR-A12 iv 3-Inch Nozzles (1) Core Spray 795-6 (2) Poison Inlet 796-6 Basis Statement l Refer to RR-A6. r i rp0382-0621d-42 II-28
March 4, 1982 Revision 2 Remarks Refer to RR-A6. Alternative Inspection or Test Hydro (Refer to RR-A1.) References Refer to RR-A6. Implementation Refer to RR-A6. RR-A13 v 3-Inch Instrument Nozzles (795-1A-E) Basis Statement Refer to RR-A1. Remarks Refer to RR-A1. Alternative Inspection or Test Refer to RR-A10. References Refer to RR-A1. Implementation Refer to RR-A10. Category: B-H Item No: B8.10 RR-A14 i Suspension Rod Brackets (801-1A through IM) Basis Statement j Plant design does not allow access to the 12 suspension rod l brackets. l l rp0382-0621d-42 II-29
March 4, 1982 Revision 2 Remarks The integrally welded support is located nine feet below the flange and at this location there is one foot between the reactor vessel wall and shield cooling jacket. Alternative Inspection or Test For welds in which defects would affect the integrity of the reactor vessel boundary, the prestart-up hydro serves as an alternative inspection (see R-1). Remote visual equipment will be secured as a result of access engineering in February 1982. A limited visual exam is scheduled for 1983. Limitations will be specified upon attempted examination. References Bechtel Drawing 0740G620128, Rev C (C-128) CE Drawing E-201-801-5 Implementation Attempt remote visual examination limited to accessible areas. During the 1983 outage, limitations for remote visual examination shall be explored. RR-AIS 11 Vessel Hanger Lug (801-2A through 2D) Basis Statement Volumetric examination of the hanger lugs is not possible due to accessibility and configuration. Remarks The hanger lugs, whose purpose is to prevent torsional movement of the reactor vessel, are approximately three feet above the suspen-sion rod brackets or about six feet below the flange. Direct access to this lug is impossible due to concrete shield wall. Remote visual access is also limited due to lug support structure whose disassembly is impossible for safety considerations. Alternative Inspection or Test l l For welds in which defects would affect the integrity of the l reactor vessel boundary, the prestart-up hydro serves as an alternative inspection (see R-1). The lug's purpose is to re-strict movement in a torsional direction. During normal reactor operation, these lugs are essentially nonload bearing. If suffi-l cient torsional movement is encountwied during operation, a remote visual exam should identify any operations induced defect. Also, rp0382-0621d-42 II-30
March 4, 1982 Revision 2 remote visual equipment will be secured prior to the 1983 outage. A visual examination will be attempted at that time and limita-tions specified pending attempted visual examination. References Bechtel Drawing 0740G620128, Rev C (C-128) CE Drawing E-201-802-9 Implementation Attempt remote visual examination limited to accessible areas. During the 1983 outage, limitations for a remote visual inspection shall be specified pending attempted visual examination. RR-A16 111 Stabilizer Brackets (801-3A through 3D) Basis Statement Volumetric examination of the stabilizer brackets is not possible due to accessibility. Remarks These brackets are approximately 5.5 feet above the bench mark of the vessel and are inaccessible for direct inspections due to concrete shield wall. Forces caused by the stabilizers are com-pressive in nature, thereby eliminating shear forces. Since there are no shear forces, the weld is essentially nonload bearing. Defects in the weld would, therefore, have no effect on the intended use of the lug. l Alternative Inspection or Test } For welds in which defects would affect the integrity of the i reactor vessel boundary, the prestart-up hydro serves as an alternative inspection (see R-1). I References j Bechtel Drawing 0740G62128, Rev C (C-128) CE Drawing E-201-802-9 Implementation Relief requested as previously granted. l rp0382-0621d-42 II-31 l
O March 4, 1982 Revision 2 Category: B-N-2 Item No: B13.20 INTERIOR ATTACHMENTS AND CORE SUPPORT STRUCTURES RR-A17 i Sparger and Core Sprav Supports (801-6A through 6H) (This request has been deleted.) RR-A18 11 Thermal Shield Supports (801-4A-F) and Inlet Diffuser Brackets (801-7A-F) (This request has been deleted.) RR-A19 iii Core Support Brackets (801-5A through 5D) (This request has been deleted.) HEAT EXCHANGERS Category: B-B Item No: B2.51, B2.52 LONGITUDINAL AND CIRCU'!FERENTIAL STEAM DRU't WELDS RR-A20 Basis Statement The 170-240 manrem additional exposure required during the in-terval to conduct a full Category B-B examination of the steam drum will not provide sufficient additional information over that provided by the partial Category B-B exam proposed to warrant the excess exposure. Remarks \\ s N 4 / 2aswes / s.in M. I \\ s 8: 'm e.:.'* ~ m m.u rx aw Wti n*
- seemet 2.3A/e neu.w u si
>= na e rp0382-0621d-42 II-32
e March 4, 1982 Revision 2 Access to the upper areas of the steam drum, within approximately 30' of arc either side of the top longitudinal center line, is good without extensive scaffolding. The general field is rela-tively low, approximately 0.06 R/hr. The combination of good access and low general field will permit shell weld exams without excessive exposure. Access to the lower portions of the steam drum, within approximately 150' of are either side of the bottom longitudinal center line, is poor without extensive scaffolding. The general field in this area is approximately ten times (0.6 R/hr) that of the upper portion. No shell weld exams will be conducted in the lower portions due to the combination of high fields and poor access; ie, approximately 20 to 25 man-hours will be required (scaffolding, insulation, weld preparation, exam) to perform each exam in the lower areas; with six of the ten longicu-dinal and circumferential welds being located entirely in this region, an additional 170-240 manrem exposure will be required (28-40 manrem per weld) to perform a complete Category B-B exam over the interval. However, by restricting inspection to the upper portions of the steam drum, the Code exam requirements can be fulfilled on four of the ten welds without accumulating the excess exposure. Alternative Inspection or Test The exams in the upper portions will disclose the condition of those areas representative of the most severe service conditions to which the steam drum is exposed and, therefore, will disclose any incipient general degradation. The hydrostatic examinations performed each refueling outage (prestart-up hydro) and the nozzle-to-safe end exams of the risers and downcomers will provide additional indication of the steam drum structural integrity. Failure of any weld in the steam drum shell would be ductile in nature most likely induced by a stress-corrosion pr fatigue mechanism and, therefore, not catastrophic. Such a failure would i be readily detected during operations by level indication, dew cell indication, makeup water flow, and/or temperature indication in adequate time for safe shutdown. l i References See Master Plan Isometrics A-14 and A-15 and CE Drawing E-230-101-9. l l Implementation Relief is requested to perform a partial Category B-B inspection i with exacs limited to those welds in the upper areas of the steam L drum within 30* of are either side of the top longitudinal center l line. 1 I rp0382-0621d-42 II-33 i
D March 4, 1982 Revision 2 f N0ZZLE-TO-SHELL WELDS AND N0ZZLE INSIDE RADIUS SECTIONS RR-A21 Steam Drum Basis Statement Nozzle-to-shell weld exams and nozzle inside radius section exams will be performed as follows: Nozzle-To-Shell Nozzle Inside Nozzle Weld Radius Section i Manway A-1, A-2 Examine Examine Steam Outlet D-1, 2, 3, 4 Examine Examine Safety Relief F-1, 2, 3, 4, 5, 6 Examine Examine Decontaminating H Examine Examine Gauge Glass J-1 Examine Thermal Sleeve Precludes Exam Reactor Vent L Examine Examine Level M-1, 3 Examine Examine The downcomer (B-1, 2, 3, 4), riser (C-1, 2, 3, 4, 5, 6), feedwater (E-1, 2), condensate (G-1, 2), gauge glass (J-2) and l 1evel (M-2, 4) nozzle-to-shell welds and nozzle inside radius sections will not be inspected due to the 595-850 manrem addi-tional exposure required during the interval to perform these inspections. The additional information to be gained by per-forming a full Category B-D examination over that provided by the partial inspection described is not considered sufficient to war-rant the additional exposure. Remarks I l Examinations of nozzle-to-shell welds and nozzle inside radius sections will be performed on only those nozzles located within approximately 30' of arc either side of the top longitudinal center line of the steam drum for the reasons described in Relief Request 18 above. Performing these exams on nozzles in the lower portions of the steam drum will require 595-850 manrem exposure over the interval (35-50 manrem per nozzle for 17 nozzles). These figures are based on t - (easral field in the area of the nozzles. t Contact readings on *te czzle-to-shell weld area are on the order of three times tb c*t f t.e g2neral field. Alternative Inst. tio. ,, Test l The exams in the upper portions will disclose the conditions of l those areas representative of the most severe service conditions to which the steam drum is exposed and, therefore, will disclose any incipient general degradation. The hydrostatic examinations (prestart-up hydro) performed each refue}iag outage and the nozzle-to-safe end exams of the risers and 17wncomers will provide rp0382-0621d-42 II-34
March 4, 1982 Revision 2 additional indication of the steam drum structural integrity. In addition, in the case of the risers and downcomers, the nozzle-to-safe end exams scheduled to be performed represent investigation of an area of the nozzle generally subjected to higher stress levels than those encountered in the nozzle-to-shi l '4 elds and inside radius sections. Any failure encountered t t* nozzle-to-shell welds or inside radius sections would be ductile t, nature most likely induced by a stress-corrosion or fatigue mecLtcism and, therefore, not catastrophic. Such a failure would be readily detected during operation by level indication, dew cell indica-tion, makeup water flow and/or temperature indication in adequate time for safe shutdown. References See Master Plan Isometrics A-14 and A-15 and CE Drawing E-230-101-9. Implementation Relicf is requested to perform a partial Category B-D inspection with exams limited to those nozzle-to-shell welds and inside radius sections of nozzles in the upper areas of the steam drum within 30' of arc either side of the top longitudinal center line. RR-A22 Clean-Up Regenerative Heat Exchanger Units A, B, C and D Basis Statement The channel inlets (A, B, C, D-5) and outlets (A, B, C, D-6), and the shell inlets (A, B, C, D-7) and outlets (A, B, C, D-8) inside radius sections cannot be UT inspected due to geometry. Remarks The nozzle-to-shell welds on the channel inlets and outlets and shell inlets and outlets will be inspected in accordance with the Code requirements. These examinations will provide adequate information concerning the general internal condition of the nozzle and shell. The inside radius section exams have been unsuccessfully attempted and will, therefore, not be scheduled. 1 Alternative Inspection or Test See Remarks section. References See Master Plan Isometrics A-9, 11, 12 and Southwestern Engineering Drawings EM-65149, EM-86924 and DM-86366. l rp0382-0621d-42 II-35
March 4, 1982 Revision 2 Implementation Relief is requested to perform a partial Category B-D inspection with exams limited to the nozzle-to-shell welds. RR-A23 Clean-Up Nonregenerative Heat Exchanger Unit E Basis Statement The channel inlet (E-5) and outlet (E-6) nozzles, nozzle-to-shell welds and inside radius sections cannot be UT inspected due to the close proximity of shell circumferential welds and internal baffle arrangements. Remarks L The channel inlet and outlet nozzle-to-shell welds will be sub-jected to visual and surface exams in lieu of the Code required volumetric. In addition, the prestart-up hydro will provide a routine monitoring of the condition of these welds. Alternative Inspection or Test See Remarks section. References See Master Plan Isometrics A-10, 11, 13 and Southwestern Engineering Drawings EM-65161, BM-86365. Implementation Relief is requested to schedule a partial Category B-D inspection by performing visual and surface exams in lieu of volumetric exam i on the channel inlet and outlet nozzle-to-shell welds. l l RR-A24 Clean-Up Demin Tank Nozzle Inside Radius Sections l i Basis Statement l The clean-up demin tank nozzles are fabricated by welding squared-ended pipe nipples into the tank shell. The portion of the pipe nipple corresponding to a nozzle inside radir.s section has a radius of essentially zero and, therefore, cannot be UT inspected. l Remarks I The examination of the nozzle inside radius section was incorpo-rated into the Code primarily to detect thermal cycling stress degradation. The primary system water flowing into the clean-up demin tank first passes through the clean-up regenerative and non-l regenerative heat exchangers to reduce and stabilize the water i rp0382-0621d-42 II-36 t
March 4, 1982 Revision 2 temperature to below 110*F to prevent decomposition of the demin-eralizer resins. This tank is, therefore, not subject to signifi-cant thermal cycling, and inspection of the nozzle inside radius sections is, therefore, not required. Alternative Inspection or Test The prestart-up hydro and the remainder of the Category B-D testing will provide adequate indication of tank structural integrity. References See Master Plan Isometrics A-18, 19 and Infilco Drawings Y-30-4858-3, H28-4-2. Implementation Relief is requested to exempt tank nozzles from inside radius section examination. Further research by Consumers Power Company is being implemented to substantiate the basis for this request and should be available prior to the 1983 outage. RR-A25 Emergency Condenser Nozzle Inside Radius Basis Statement The nozzles are of an " insert type saddle" arrangement. This configuration may not be readily inspectable by ultrasonic methods. Implementation Sufficient reference drawings are not available for the inner radius section. Possibilities for inspecting the inner radius l will be studied. No relief requested. l [ Category: B-F l Item No: B5.40 l l N0ZZLE-TO-SAFE END WELDS RR-A26 Steam Drum l Basis Statement i All steam drum dissimilar metal nozzle-to-safe end welds have been classed as B5.40, B-F welds; these include: l rp0382-0621d-42 II-37 1 I \\
March 4, 1982 Revision 2 Downcomers 17-MRS-111,-112, 113, 114-1 Risers 14-MRS-101, 103, 104, 105, 106-6 and 14-MRS-102-7 These welds are scheduled for examination in accordance with Code requirements. All steam drum similar metal nozzle-to-safe end welds have been classed as B9.11, B9.21 or B9.40, B-J welds since there is no safe end or dissimilar metal weld; these include: Steam Outlet 8-MSS-101, 102, 103, 104-1* Feedwater 8-FWS-102, 103-3, 5* Safety Relief 3-MSS-107, 1, 2, 3, 4, 5, 6 Condensate 4-ESS-103, 104-15 Gauge Glass 1.5-MSS-110-1, 22 Rx Vent 1.5-MSS-117-41** Level 1.5-MSS-111-17 1.5-MSS-112-17 4-MSS-111, 112-1 The similar metal nozzle-to-safe end welds are scheduled for examination within the Code requirement to exam 25% of the weld joints so as to exclude those similar metal nozzle-to-safe end or pipe welds associated with nozzles within approximately 150' of erc either side of the bottom longitudinal center line of tha steam drum as defined in Relief Requests RR-A20 and RR-A21. Remarks See Remarks for Relief Requests RR-A20 and RR-A21. Alternative Inspection or Test Not applicable. l References BRP ISI Master Plan l CE Drawings E-230-101, E-230-103, E-230-104 I t Implementation i No relief is required. l l l l ?
- These are nozzle-to-pipe welds.
i
- The Rx vent lir.e has a dissimilar metal weld 1.5-MSS-117-40, elbow-to-saiu end.
l l l rp0382-0621d-42-48 II-38
March 4, 1982 Revision 2 RR-A27 Clean-Up Regenerative and Nonregenerative Heat Exchanger Basis Statement Each clean-up regenerative heat exchanger is fitted with four forged carbon steel nozzles. There are no safe ends on the clean-up regenerative heat exchangers. There are four carbon steel nozzles on the clean-up nonregenerative heat exchangers (two of the four nozzles are unclassed). There are no safe ends on the clean-up nonregenerative heat exchanger. Remarks All similar metal nozzle-to-pipe welds will be examined as B9.21, B-J welds. Tha carbon steel nozzles to carbon steel piping welds are as follows: a. Clean-Up Regenerative Hx 3-RCS-101-23, 24, 25, 26, 29, 30, 30 and 32 b. Clean-Up Nonregenerative Ex 3-RCS-101-37, 38 Alternative Inspection er Test Not applicable. References See Master Plan IS0s A-9, 10, 11, 12, 13, 32, 37 and Southwestern Engineering Co Drawings EM-65149, EM-86924, DM-86366, EM-65151 and BM-85365. Implementation i No relief required. I RR-A28 Clean-Up Demineralizer Tank Basis Statement l The clean-up demineralizer tank is fitted with four carbon steel j nozzles (A-234 WPB). There are no safe ends on the clean-up I demineralizer tank nozzles. l Remarks l All similar metal nozzle-to pipe welds will be examined as B9.21, l B-J welds. The carbon steel nozzle-to pipe welds are: l r rp0382-0621d-42-48 II-39 I t
March 4, 1982 Revision 2 3-RCS-101-89 3-RCS-102-1 2-RCS-110-5 2-RCS-111-1 There are no dissimilar metal nozzle-to pipe welds adjacent to the clean-up demineralizer tank. Alternative Inspection or Test Not applicable. References See Master Plan IS0s A-18, A-19, A-35, A-36 and A-43 and Infilco Drawing Y-30-4853-3. Implementation No relief required. Category: B-H Item No: B8.40 INTEGRALLY WELDED STEAM DRUM SUPPORTS RR-A29 Basis Statement I The 392-560 manrem additional exposure required, during the in-terval, to conduct a full Category B-H examination of the steam drum will not provide sufficient additional information over that provided by the partial Category B-H exam proposed to warrant the excess exposure. I Remarks i j Access to the upper areas of the steam drum, within approximately 30' of arc either side of the top longitudinal center line, is good and supports T2-1, 2, 3 and 4 are inspectable. Access to the lower portions of the steam drum coupled with high radiation fields, as stated previously for the steam drum, pre-cludes examination of the following steam drum supports: l T1-1, 2, 3, 4 V1-1, 2, 3, 4 V2-1, 2, 3, 4 H1, 2 rp0382-0621d-42-48 II-40
March 4, 1982 Revision 2. Alternative Inspection or Test See Remarks above. References See Master Plan ISO A-16 and CE Drawings E-230-101-9 and D-230-213. Implementation Relief requested to perform a partial Category B-H inspection with exams limited to those welds in the upper areas of the steam drum within 30' arc either side of the top longitudinal center line weld. PIPING PRESSURE BOUNDARY Category: B-F Item No: B5.50 SAFE END DISSIMILAR METAL WELDS RR-A30 1 6-SCS-101-7 Basis Statement Pipe weld is not physically accessible for Nondestructive Testing due to plant design. Remarks See Remarks section for 6-SCS-101-1 through 7 and 8-SCS-101-8 through 12 (RR-A33). Alternative Inspection or Test A hydrostatic test (prestart-up hydro) is performed before each start-up at 1.1 times the operating pressure. This is more con-servative than the prestart-up leak test at operating pressure as required by the Code (IWB-5221). Inspection of the remaining ac-cessible portion of this line will provide some indication of the piping integrity. References See Master Plan ISO A-28. rp0382-0621d-42-4f. II-41
March 4, 1982 Revision 2 Implementation Relief requested to examine 6-SCS-101-7 solely hydrostatically during prestart hydro. RR-A31 11 Dissimilar Metal Socket Welds Basis Statement Section XI does not address the subject of dissimilar metal socket welds. Plant piping contains 11 dissimilar metal socket welds. Meaningful volumetric examination of socket welds is not possible. Remarks The following dissimilar metal socket welds have been designated B5.50, B-F and will be inspected by surface examination: 2-RCS-101-64 2-RCS-111-8, 9 1.5-RDS-112-5, 6 1.5-RDS-113-5, 6 1.5-RDS-114-5, 6 1.5-RDS-115-5, 6 1.5-MSS-117-41 Alternative Inspection or Test See Implementation. References See ISI Master Plan. Implementation Relief requested to examine dissimilar metal socket welds by surface examination. RR-A32 111 Similar Metal Safe End-to-Pipe Welds Basis Statement The following safe end-to-pipe welds join two stainless steel components and are, therefore, similar metal welds. These safe end-to-pipe welds are, therefore, classed as B9.11 or B9.21, B-J welds rather than B5.50, B-F welds: rp0382-0621d-42-48 II-42
M rch 4, 1982 Revision 2 20-MRS-121-20 B9.11 B-J 20-MRS-122-20 B9.11 B-J 14-MRS-101-2 B9.11 B-J 14-MRS-102-2 B9.11 B-J 14-MRS-103-2 B9.11 B-J 14-MRS-104-2 B9.11 B-J 14-MRS-105-2 B9.11 B-J 14-MRS-106-2 B9.11 B-J 6-SCS-101-1 B9.11 B-J 3-LPS-102-47 B9.21 B-J 2-MSS-121-1 B9.21 B-J 2-MSS-131-1 B9.21 B-J. 2-MSS-134-1 B9.21 B-J 2-MSS-124-1 B9.21 B-J 3-CSS-104-20 B9.21 B-J Remarks All of these welds with the exception of the 4 inch instrument nozzles, the 14-inch main recire lines and approximately one-third of the external surface of 1 inch recire line are inacces-sible for examination. See the following relief requests for the basis of this statement: a. 20-Inch Recire Welds RR-A4 b. 3-Inch (1) Core Spray RR-A6 (2) Poison Inlet c. 8-Inch Shutdown Unioading RR-A7 Alternative Inspection or Test Prestart-Up Hydro (See RR-A1.) Volumetric (UT) and surface limited to approximately one-third of the external surface for 20-MRS-122-20. 1 References CE Drawings E-201-795-5 and E-201-796-5 Big Rock Point ISI Master Plan rp0382-0621d-42-48 II-43
March 4, 1982 Revision 2 Implementation Reclass the welds discussed above from B5.50, B-F to 39.11 or B9.21, B-J and relief requested on the safe end-to-pipe welds as discussed in the Remarks section. Category: B-J Item No: B9.11 CIRCUMFERENTIAL PIPE WELDS IN PIPING 2 4 INCHES RR-A33 Basis Statement Physical plant layout precludes Nondestructive Testing of the following welds: a. 6-SCS-101-1 through 7 b. 8-SCS-101-8 through 12 c. 3-CSS-101-17 through 19 d. 12-MSS-105-9 and 10 Remarks a. Welds 1 through 6 of the 6-inch shutdown cooling system are embedded in the concrete reactor shield wall and are, there-fore, not accessible for Nondestructive Testing. Welds 7 through 12 are inaccessible for Nondestructive Testing due to plant layout. The portion of Lines 6-SCS-101 and 8-SCS-101 containing Welds 7 through 12 is greater than 35 feet above the floor. This line is at least 15 feet from all walls except the concrete reactor shield. It is not possible to i erect scaffolding or to utilize a ladder to access this line due to the sloping floor at the base of the containment sphere. Access from above is precluded due to a straight, uninterrupted drop of 35 feet from the steam drum environs. b. Access to 3-CSS-101, Welds 17 through 19, is possible only through small crawl spaces made available by removing blocks l of shielding material. An additional 99-117 manrem exposure l is required to prepare for conduct and cleanup after the examinations. c. Welds 9 and 10 of Line 12-MSS-105.are located on a vertical l run of pipe adjacent to the south wall of the main recircula-tion pump room. Plant design and existing components prevent scoliolding from below with required safety standards. Access from the top (steam drum level) involves the risk of falling l from 25 feet to 60 feet. i The radiation exposure and personnel safety risks involved to per-l form these inspections are not justifiable. Information gained from Nondestructive Testing of accessible portions of these rp0382-0621d-42-48 II-44 I
March 4, 1982 Revision 2 systems, along with hydrostatic testing, assures the integrity of these welds. If conditions evolve which warrant additional in-spection of these welds, efforts shall be made to obtain suitable access. Alternative Inspection or Test A hydrostatic test (prestart-up hydro) at 1.1 times the operation pressure is performed in these systems each outage before start-up. Nondestructive inspection of the accessible portion of these lines will provide some indication of the piping integrity. References BRP ISI Master Plan Isometrics A-28, A-47 and A-72 Implementation Relief requested from nondestructive examination of these welds. Perform hydrostatic test only. Category: B-J Item No: B9.21 CIRCUMFERENTIAL WELDS - NOMINAL PIPE SIZE < 4 INCHES RR-A34 Basis Statement Physical plant layout precludes examination of the following welds: a. 3-LPS-102-18 through 23 b. 3-LPS-102-44 through 47 l Remarks a. Welds 18 through 23 of Line 3-LPS-102 are inaccessible for examination. The line is located 50 feet above the floor of the containment. Scaffolding from below is prevented by l existing plant components and piping and containment design l without violating safety standards. This line is also located l underneath a ledge which makes access from the top impossible. Radiation fields in this area are in the order of 1 R/hr. b. Welds 44 through 47 of Line 3-LPS-102 are just above 32 CRD hydraulic lines and located inside high-density aggregate trays. The hydraulic lines and aggregate trays would require removal to access these welds. Therefore, because of personnel safety and radiation experure, j examination of these welds is not justified. Results of rp0382-0621d-42-48 II-45 i
March 4, 1982 Revision 2 nondestructive inspection of accessible portion of the line and hydrostatic tests will give some indication of the integrity of these welds. References Big Rock Point ISI Master Plan - Isometric A-26 Bechtel Equipment Location Drawing 0740G40102 (M-102) Implementation Relief requested to examine these welds with hydrostatic testing only. Category: B-J Item No: B9.40 SOCKET WELDS RR-A35 Basis Statement Physical plant layout precludes examination of the following welds: a. 2-MSS-121-2, 3 b. 2-MSS-131-2, 3 c. 2-MSS-134-7, 8 d. 2-CRD-101-22 e. 1.5-MSS-117-17 through 22 and 29 Remarks a, b, c. These main steam piping welds are directly adjacent to the reactor vessel (2 R/hr field). An additional 99-117 manrem exposure is required to prepare and conduct the inspection. Access to the welds is possible only through a small crawl space made available by removing blocks of shielding material, d. Weld 22 of Line 2-CRD-101 is embedded in concrete and inac-cessible for inspection. l i e. Welds 17 through 22 of Line 1.5-MSS-117 are located approxi-mately 60 feet above the containment floor. Radiation fields in this area are approximately.5 to 1 R/hr. Scaffolding from the steam drum walkway would require an additional 1,600-2,000 mR exposure. Weld preparation and inspection would also total about 1,000 mR exposure. Weld 29 of Line 1.5-MSS-117 is directly adjacent to a pipe restraint which prevents access to the weld for surface rp0382-0621d-42-48 II-46
March 4, 1982 Revision 2 examination. The support cannot be moved without cutting and rewelding. Radiation fields are approximately 50-100 mR/hr. Alternative Inspections or Tests These systems are part of the prestart-up hydrostatic pressure test (at 1.1 times the operating pressure) performed every outage. A VT-1 examination vill be performad on Weld 1.5-MSS-117-29 in place of a surface examination. Nondestructive examination re-suits of accessible portions of these lines will provide indica-tions of the integrity of these welds. References See the corresponding ISI Master Plan isometrics: a. ISO A-56 b. ISO A-59 c. ISO A-61 d. ISO A-26 e. ISO A-74 f. ISO A-55 Implementation Relief requested to perform alternate inspections described above. Category: B-K-1 Item No: B10.10 INTEGRALLY WELDED ATTACHMENTS RR-A36 i Main Recire System Basis Statement The following pipe attachments are not accessible for volumetric or surface examination due to plant design and high radiation fields: I 14-MRS-103-3PL 1 through 8 14-MRS-105-3PL 1 through 8 Remarks These supports are located 30 feet above the lower deck area of containment. The floor in this area is curved due to containment configuration and the use of scaffolding or ladders is not pos-sible. Access from above is also not possible due to plant layout. rp0382-0621d-42-48 II-47 t
March 4, 1982 Revision 2 The general radiation field in this area is 0.6 R/hr, and 11-13 manrem exposure per support per interval would be absorbed solely in preparation for the exam, ie, weld prep and insulation removal, performing the exam and postexam cleanup, provided access to the supports were possible, which is not the case. Alternative Inspection or Test For the case in which a weld defect would affect the integrity of the piping pressure boundary, tha prestart-up hydro serves as an alternative inspection. A remote visual examination will be made of the support to try and determine its integrity as a support. References See Master Plan ISO A-64. Implementation Relief requested to perform alternate examinations. RR-A37 11 Reactor Depressurization System Basis Statement The following pipe attachments are inaccessible for volumetric or surface examination due to configuration of pipe support. These attachments are also nonload bearing: 12-RDS-101-3PL 12-RDS-101-13PL l Remarks Pipe lugs are covered by pipe support and are not accessible for volumetric or surface examination as shown below. Support is accessible for visual examination. l 1 I \\ I l ^ 1 / rp0382-0621d-42-48 II-48
March 4, 1982 Revision 2 The pipe attachments are welded to the pipe and during normal operation, do not contact the pipe support structure. The pipe attachments serve to resist torsional movement and are nonload bearing. Alternate Inspection or Test This system is hydrostatically tested during the prestart-up hydro. References Big Rock Point ISI Master Plan Isometric - A-82 Suntac Nuclear Corporation drawings: 12-RDS-101-13PL, A-18505, A-13507 12-RDS-101-3PL, A-18502, A-13512 Implementation This support is not required to be inspected per Section XI because of its nonload bearing function. 'l Category: B-K-2 Item No: B11.10 RR-A38 Basis Statement The following component supports are not accessible for direct visual examination due to plant design and high radiation areas: a. 3-LPS-102-21PR-1, 21PR-2, 23PR b. 6-SCS-101-6PS c. 8-SCS-101-8PR, 10PR Remarks l a. The supports are located 50 feet above the lower deck area of containment. The floor in this area is curved due to contain-ment configuration and the use of scaffolding is not possible. Access from above is also not possible. The general radiation field is 0.6 R/hr and 11-13 manrem l exposure per support per interval would be abosrbed before, during and after examining these welds, provided access to the support were possible, which is not the case. b, c. Access to these components is discussed previously in this report. Refer to RR-A33 (a) for remarks. rp0382-0621d-42-48 II-49
March 4, 1982 Revision 2 Alternative Inspection or Test Inspection shall be done remotely as allowed by Para-graph IWA-2213(c) of the Section XI Code (77 Edition). i References Big Rock Point ISI Master Plan Isometrics: a. A-26 b. A-28 c. A-28 Implementation Relief requested from direct visual examination. Category: B-L-1 Item No: B12.10 PUMP PRESSURE BOUNDARY i RR-A39 Pump Casing Welds Basis Statement Volumetric examination of these welds (clean-up pump, main re-circulation pump) is not possible due to configuration. Remarks During the Big Rock Paint 1982 refueling outage, access engi-neering was conducted on the clean-up pump and main recirculation pump casing welds by Southwest Research Institute. Due to in-ternal configuration, the use of ultrasonic examination on these welds is not practical. [ l Alternative Inspections l Surface examinations to be conducted on accessible arecs. i See RR-A1. l References Byron Jackson Drawing IF-4614-3 Implementation Relief requested from volumetric examination. l rp0382-0621d-42-48 II-30 l
March 4, 1982 Revision 2 Category: B-M-2 Item No: B12.40 VALVE PRESSURE BOUNDARY VALVE INTERNAL EXAMINATION RR-A40 Basis Statement . Main recirculation pump discharge and butterfly valve are not fully isolable from reactor and are, therefore, not inspectable. Remarks The following valves are not isolable from the reactor.: Line Valve Number 5-MRS-131 MO-N002A 5-MRS-132 MO-N002B 20-MRS-121 MO-N001A 20-MRS-122 MO-N001B 20-MRS-121 MO-N006A 20-MRS-122 M0-N006B Examination of the above valves reqaires complete draining of the reactor vessel which is not practical and requires 100*. core off-loading which creates tremendous maintenance and adds exposure problems. i Alternative Inspection or Test A hydrostatic test (prestart-up hydro) conducted at 1.1 times the operating pressure will serve as an alternate test. The inspec-tion of other valves and the results will also give an indication i of the condition of the internals of these valves. Should it be necessary to disassemble these valves for maintenance, we are prepared to commit to visual examinations. References See Master Plan IS0s A-66, A-67, A-68 and A-69. l j Imolementation Relief requested to examine these valves by hydrostatic testing and to exempt these valves from the VT requirements of B-M-2. I i ( rp0382-0621d-42-48 II-51 1 l I
March 4, 1982 Revision 2 SECTION 3.0 i GENERIC RELIEF REQUESTS rp0382-0621e-42-48
Ma rch 4, 1942 R2Visicn 2 TABLE lil-1 RELIEr R[ QUESTS BIC ROC _k POINT NUGl[AR PLANT CLASS 1 COMPONENTS Arsa of Section XI Alternative Test Relier R91ief _Ca tego ry Congonent function Roller Reauested and Examination Imniementation Request 31.11 B-A Vessel Pr ima ry Longitudinal & a. Presta rt-Up flydro(1) a. Tech Spec RR-Al Bl.12 Core Pressure C i rcumfe ren t ia l b. Rema inder of B1.11, b. Master Plan Region Bounda ry Seam Welds In-B1.12 Testing 1 accessible for Volumetric Exams B1.11 B-A Vessel Prima ry Longitudinal & a. Presta rt-Up Hydro (1) a. Tech Spec RR-A2 D1.12 Shell Pressure Ci rcumfe rent ia l b. Rema inder or 81.11, b. Master Plan Region Bounda ry Seam Welds In-B1.12 Testing accessible for Volumetric Exams 01.21 B-A Vessel Prima ry Lower licad Ci rcumfer-a. Presta rt-Up Hyd ro(1) a. Tech Spoc RR-A3 81.22 Botton Pressure ential & Longitudinal b. Rema inder of 81.21, b. Haster Plan Head Bounda ry Seam Wo ld s Gene ra l ly 81.22 Testing inaccessible for Volu-metric Exams 83.90 B-D 20" Recirc Prima ry Nozzle-to-Vessel Weld a. Presta rt-Up flydro(1) a. Tech Spec RR-A4 B3.100 Nozzles Pressure & Nozzle inside b. Rema inder of B3.90, b. , Master Plan (796-1A-B) Boundary Radius Are inaccessi-B3.100 Testing ble for Volumetric Examination 83.90 B-D 3" Nozzles Prima ry Nozzle-to-Vessel a. Presta rt-Up flydro(1) a. Tech Spec RR-A6 B3.100 1. Core Spray Pressure Weld & Nozzle inside b. Rema inder or 83.90, b. Master Plan (795-6) Bounda ry Radius Are inaccessi-B3.100 Testing 2. Poison in-ble for Volumetric let (796-6) Examination B3.90 B-D 8" Shutdown Prima ry Nozzle-to-Vessel Weld a. Presta rt-Up Hyd ro(1) a. Tech Spec RR-A7 83.100 Unloading Moz-Pressure & Nozzle inside b. Rema inder of B3.90, b. Haster Plan zie (795-15) Bounda ry Radius Aro inaccessi-B3.100 Testing ble for Volumetric Examination 83.90 B-D 3" Instrument Pri ma ry Deleted a. Pre sta rt-Up* Hyd ro( 1 ) a. Tech Spec RR-A8 B3.100 Nozzles Pressure b. Rema inder of 83.90, b. Master Plan (955-1A-D) Dounda ry B3.100 Testing NOIE: All hydrostatic tests referenced in the body or this report are the prestart-up hydros listed in the alternative test and examination column. rp0 382-062 3 r-42-48 111-52
Ma rch 4, 1982 R3 vision 2 TABLE lil-1 (Contd) Area or Section XI Alternative Test Roller Emiler ca tego ry connonent runction Reller Reauested and Examination implementation Request B5.10 B-r 20" Recire Prima ry Nozzle-to-Sa re Ends a. Pres ta rt-Up l'yd ro( 1 ) a. Tech Spec RR-A9 Nozzles Pressure inaccessible b. Rema inder of B5.10 b. Master Plan (796-1A,B) Bounda ry Testing B5.10 B-F 14" Steam Out-Prima ry Nozzle-to-Safe End a. Presta rt-Up ifydro(1) a. Tech Spec RR-A10 let Mozzles Pressure Welds b. Ultrasonic Exam b. Master Plan (,95-11A Bounda ry Through 11F) B5.10 B-F 8" Shutdown Un-Prima ry Nozzle-to-Safe Ends a. Presta rt-Up Hyd ro(1) a. Tech Spec RR-All loading Nozzle Pressure inaccessible b. Rema inder of 85.10 b. Master Plan (795-15) Bounda ry Testing $5.10 B-f 3" Nozzles Prima ry Nozzle-to-Safe Ends a. P re s ta rt-Up Hyd ro( 1 ) a. Tech Spec RR-A12 1 Core Spray Pressure inaccessible b. Rema inder of 85.10 b. Master Plan (795-6) Bounda ry Testing 2. Poison in-let (796-6) E5.10 B-r 3" Instrument Prima ry Nozzle-to-Sa re Ends a. Presta rt-Up Hyd ro( 1 ) a. Tech Spec RR-A13 Nozzles Pressure inaccessible b. Rema inder of 85.10 b. Master Plan (955-1A-D) Bounda ry Testing 88.10 B-ll I nteg ra l ly Prima ry Supports Are inacces-a. Prestart-Up ifydro(1) a. Tech Spec RR-A14 Welded Ves-Pressu re sible for Visual b. Attempt Remote sel Supports Bounda ry Examination Visual Exam (801-1A-M) 88.10 B-H Vessel Hanger Prima ry Lugs inaccessible a. Presta rt-Up uydro a. Tech Spec RR-A15 Lugs (810-2A-D). Pressure for UT b. Remote Visual b. Haster Plan Bounda ry Attempt B8.10 B-H Stabilizer P rima ry Brackets inaccessi-a. Presta rt-up ifydro a. Tech Spec RR-A16 B racke t s Pressu re bio for UT b. Remote Visual b. Master Plan (801-3A-D) Bounda ry Attempt B13.20 B-N-2 Sparger & Core Prima ry Deleted None - Inspected as a. r40-Yea r Plan RR-A17 Spray Supports Pressure Possible (801-6A-H) Bounda ry 313.20 B-N-2 Thermal Shield Prima ry Deleted None - Inspected as a. 40-Yea r Plan RR-A18 Supports & In-Pressu re Possible let Dirruser Bounda ry B racke ts (801-4A-F) (811-7A-F) rp0382-0621r-42-48 111-53
~__ Ma rch 4, 1982 R3 vision 2 TABLE Ill-1 (Contd) A ros or Section XI Alternative Test Roller ggtjar Category. Component Function Relier Reauested and Examination Implementation Request G13.20 B-N-2 Core Support P rima ry Deleted None - Inspect as a. 40-Yea r Plan RR-A19 B rac ke ts Pressure Possible a (801-5A-D) Bounda ry 02.51 B-B Steam Drum Lon-Prima ry Sys-Exempt from Volu-a. Presta rt-Up Hyd ro( 1 ) a. Tech Spec RR-A20 gitudinal Welds tem Pressure metric Exam Due b. Rema inder or 82.51 b. Master Plan 101-2,3,5,6,8,9 Bounda ry to Accessibil-Testing ity & Exposure C2.52 B-B Steam Drum Cir-Prima ry Sys-Pe rfo rm Vol ume t ric a. Presta rt-Up Hyd ro( 1 ) a. Tech Spec RR-A20A conferential tem Pressure Only in Upper b. Rema lprier ur B2.52 b. Master Plan Welds Bounda ry 60* Arc or Drum Testing 101-1,4,7,10 ci rcuarerence 83.150 B-D Steam Drum Prima ry Sys-Exempt from Volu-a. Presta rt-Up Ilyd ro( 1 ) a. Tech Spec RR-A21 B3.160 Nozzle-to-Vessel tem Pressure metric Exam Due b. Remainder or 83.150 b. Master Plan & Inside Radius Bounda ry to Accessibility Testing c. Haster Plan Sections Down- & Exposure c. Rema inder of B3.160 come rs B-1,2,3,4 lesting Risers C-1,2, 3,4,5,6 Feedwa te r E-1,2 Condensate C-1,2 Cauge Class J-2 Rx Vent L tevel M-1,3 03.150 B-D Clean-Up Regen-Prima ry Sys-Exempt f rom Volumetric a. Presta rt-Up Hydro (1) a. Tech Spec RR-A22 83.160 e ra t ive Hea t tem riow Exam Due to Geometry b. Remainder or B3.150 b. Master Plan Exchanger Regulation Testing Units A,B,C D Nozzle inside Radius Sections on Nozzles A,B,C,D-5; A,B,C,D-6: A,B,C.D-7; A,B,C,D-8 03.150 B-D Clean-Up Non-Auxi l ia ry Perform Visua l & a. Visual & Surface (1) a. Master Plan RR-A23 regene ra t ive System Tem-Surface in Lieu b. Presta rt-Up *Hyd ro b. Tech Spec Heat Exchanger pera ture Con-or Volumetric Unit E Nozzle-trol to-Shell Wolds E-5 & E-5 rp0382-0621r-42-48 111-54
Ma rch 4, 1982 R3 vision 2 TABLE 118-1 (Contd) Arca er Section XI Alternative Test Relier Religr Category. Component Function Reller Reauested and Examination implementation Eequest C3.160 B-D Clean-Up Non-Aux i l i a ry Exempt From Examina-a. Pre s tc.rt-Up Hyd ro( 1 ) a. Tech Spec RR-A23 regene ra t ive System Tem-tion b. Remainder of B-D b. Haster Plan llea t Exchanger pe ra ture Con-Testing Unit E Nozzle trol Inside Radius Section C3.160 B-D Clean-Up Demin Aux i l ia ry Exempt from Examina-a. Presta rt-Up Hydro (1) a. Tech Spec RR-A24 Tank Nozzle System Water tion b. Rema inder of B-D b. Master Plan inside Radius Purification Testing Sections 83.160 B-D [mergency Con-Prima ry Sys-Investigating Feasi-a. None - Examine as a. 40-Yea r Plan RR-A25 denser Nozzle tem Imergency bility of Exams Scheduled inside Radius Heat Removal Sections E5.40 B-F Steam Drum Simi-Prima ry B3.3, B-F Dissimilar a. None - Examine as a. 40-Yea r Plan RR-A26 l a r Me ta l Pressure Metal Wolds Will Be Required Nozzle-to-Safe Bounda ry Examined as Scheduled. End Welds 85.40, B-F Similar Metal Welds Shall Be Reclassirled as 89.11, B9.21 or B9.40, B-J Welds. These Simi-lar Metal Weld in-spections Shall De Chosen on the Basis of Accessibility & Low-Radiation Fields. D3.40 B-F Clean-Up Regen Prima ry None - No B5.40, B-F a. None - Examine to a. 40-Yea r Plan RR-A27 & Monregen Heat P re ssu re Welds Requi rements of Exchanger Bounda ry B9.21, B-F Welds Mozzle-to-Safe End Welds 05.40 B-F Clean-Up Demin Prima ry Nono - Ao B5.40, B-F a. Nono - Examine to a. 40-Yea r Plan RR-A28 Tank Nozzle-Pressure Welds Requirements of to-Safe End Bounda ry 89.21, B-F Holds Welds 88.40 B-H I nteg ra l ly Prima ry Pa rt ia l Ca tego ry B-H a. None - Examine as a. 40-Yea r Plan RR-29 Welded Steam Pressure inspection Required Drum Supports. Bounda ry rp0382-0623r-42-48 181-55 1
Ma rch 4, 1982 Gsvisien 2 TABLE 111-1 (Contd) A rca of Section XI Alternative Test Relief R311er Ca t ego ry Component function Relier Requested and Examination I mp lement a t ion Request C5.50 B-F Socket Welds P ri ma ry No Access for Inspec-a. Presta rt-Up Hydru(1) a. Tech Spec RR-A30 6-SCS-101-7 Pressure tion Dissimilar Bounda ry Metal C5.50 B-F Socket Weld Pri ma ry inspect With Surface a. resta rt-Up itydro(1) a. Tech Spec RR-A31 Dissimilar Pressure Exams Only Metal Bounda ry 85.50 B-F Reactor Vessel Pri ma ry Not Accessible for a. Presta rt-Up Ilyd ro( 1 ) a. Tech Spec RR-A32 Safe End to Pressure Volumetric inspection, b. Remainder of Testing b. 40-Yea r Plan Piping Welds Bounda ry Reclass as 89.11 or on Reactor Vessel Dissimilar 99.01, B-J Welds. E9.11 B-J Butt Welds 2 4" P ri ma ry inaccessible for a. Presta rt-Up Hydro (1) a. Tech Spec RR-A33 Pressu re inspection b. Remaining Welds on b. 40-Yea r Plan Bounda ry Same Lines C9.21 B-J Butt Welds < 4" Prima ry Welds inaccessible a. Presta rt-Up Hydro a. Tech Spec RR-A34 P ressu re for Direct inspection b. Rema inder of B9.21 b. 40-Yea r Plan Boundary Butt Welds on Same Lines 89.40 B-J Socket Welds Prima ry Welds inaccessible a. Presta rt-Up Hydro (1) a. Tech Spec RR-A35 Pressure for Volumetric b. Remainder of Welds b. 40-Yea r Plan Bounda ry inspection on Same Lines B10.10 B-K-1 I nteg ra l Pipe Prima ry Supports Are Inacces-a. Presta rt-up Hydro (1) a. Tech Spec RR-A36 Supports Pressure sible for inspection b. Other Supports on b. 40-Year Plan 14-MRS-103-3PL Bounda ry Same Lines 14-MRS-105-3PL B-K-1 I nteg ra l ly Prima ry Volumetric Exam Not a. Presta rt-Up Hyd ro(1) a. Tech Spec RR-A37 Welded Supports Pressure Required. Nonload b. 40-Yea r Plan 12-RDS-101 Bounda ry Bea ring Lugs. PL-1 Through 8 & 13-PL-1 Through 4 B11.10 B-K-2 Support Com-Pri ma ry Not Accessible for a. Remote Visual in-a. 40-Yea r Plan RR-A3B ponents - Pressure Direct Visual Exam spection St ruc tu ra l Bounda ry rp0382-0623r-42-48 111-56
Ma rch 4, 1932 R3 vision 2 TABLE 111-1 ( Contd ) A rca or Section XI Alternative Test Roller Rilier Ca tego ry _ Component Function Relier Reauq1ted and Examination implementation ggqttgal S12.40 B-M-2 Valve Internals Prima ry Valves Cannot Be a. Presta rt-Up Hydro (1) a. Tech Spec RR-A40 MO-N002A Pressure Isolated From Reactor MO-N0028 Doundary So inspection Not MO-N001A Possible MO-N001B MO-NOO6A MO-N0068 012.10 B-L-1 Pump Casing P rima ry Reller Requested From a. Prestart-up Hydro a. A 40-Year Plan RR-A39 Welds Pressure Volumetric Exams b. Surface Exam or Bounda ry Accessible Areas +50 (1)The conduct of the prestart-up hydrostatic test at a pressure or 1500 -0 psia (> 1.1 operating pressure) will be accompanied by a visual examination, VT-2, to detect the presence or any through-wall leakage, ir any, of those pressure-retaining components where access for volumetric examination is impractical. This test and examination provides assurance that any structural deg ra-dation that may have occurred in service will be detected and corrective actions taken prior to Plant start-up. Any s t ruc tu ra l degradation that may develop while the component is in service is further monitored by fluid level indica tions, dew cel l in-dications, makeup water flow and/or temperature indications as provided by instrumentation in the Big Rock Point Nuclear Plant for timely detection and safe shutdown of the Plant. rp0382-0621r-42-48 111-57
March 4, 1982 Revision 2 SUBSECTION B - CLASS 2 Paragraph IWC-1210: The examination requirements of IWC shall apply to Class 2 Pressure Retaining Components (and their components). A. CLASS 2 PIPING AT CONTAINMENT PENETRATIONS BETVEEN CLASS 3 OR NONCLASSED (NONNUCLEAR) COMPONENTS RR-B1 Basis Statement Implementation of this requirement is not appropriate at certain con-tainment penetrations. Remarks The following penetrations are affected by this statement: Penetration Piping on Either Side Number P&ID No of Penetration H-12 M-111 Class 3 H-13 M-111 Class 3 H-15 M-108 Nonclassed/ Class 3 H-17 M-108 Nonclassed H-18 M-110 Class 3 H-21 M-108 Nonclassed H-22 M-107 Nonclassed/ Class 3 H-23 M-107 Nonclassed H-31 M-121 Nonclassed The classification of the Class 2 portion of the piping is governed by the leak tightness requirement associated with the containment barrier, I while the Class 3 or nonnuclear classification is dictated by system function under pressure and temperature operating conditions. In all cases, the system operating conditions impose loading on the components I that are more severe than the external loadings imposed when the con-l tainment functions are tested. l The system pressure tests and visual examinations of Subsection IWD of the ASME Code are intended to detect any service induced degradation resulting from the more severe operational loads that the components are expected to sustain over its service lifetime. Accordingly, the system function governs the classification under which the applicable rules of ASME Section XI Code must be applied. This position is con-sistent with the revision of IWA-1300(f) approved by ASME Subcommittee l on Nuclear Inservice Inspection (Agenda No ISI-77-14) which states: "The portion of piping penetrating a containment vessel required by Section III to be constructed to Class 1 or 2 rules, and whie.h may differ from the classification of the balance of the system, need not I rp0382-0621g-42-48 III-58
March 4, 1982 Revision 2 ~ ~ affect the overall system classification which determines the applica-ble rules of this section." Accordingly, where the components beyond the containment penetration areas delineated on the referenced P and I diagrams are classified as ASME Class 3, the Class 2 process piping and associated pressurization pipe will be subjected to the requirements of Subsection IWD of ASME Section XI Code. Where the components beyond the containment penetration area are classified nonnuclear class, the Class 2 components and associated pressurization pipe will not be subject to any Section XI requirements. Alternative Inspection or Test Since the proposed examinations or the Class 2 portions of the above referenced systems are in accordance with code intent, no alternate in-s spections or tests are considered necessary. This request "for relief is submitted primarily to obtain confirmation of IVA-1300(f), which is expected to be issued in a forthcoming addenda to the ASME Section XI Code. References Penetration No P&ID Coor H-12 M-111 0-10 H-13 M-111 H-10 H-15 M-108 H-17 M-108 > x H-18 M-110 J-23 H-21 M-108 H-22 M-107 S-15 H-23 M-107 D-6 H-31 M-121 Implementation Inspection as required by code. B. CLASSIFICATION OF CLASS 2 VENT AND DRAIN LINES BEYOND THE NORMALLY CLOSED VALVE RR-B2 Basis Statement Implementation of this requirement is not appropriate for certain piping. l l Remarks i l The portion of the nonregen and regen heat exchanger channel vents and l drains and shell vents and drains beyond the second normally closed valve would generally be classified as Class 3. Since these vents and rp0382-0621g-42-48 III-59
March 4, 1982 Revision 2 e 4 h drains could have been classified as Class 3, and since the venting or draining does not relate to any of the functions listed in ~ Table IWD-2500 1, these portions of the vents and drains will be classified as nonnuclear for the purposes of inservice inspection and will not be subjected to examination under Section XI of the ASME Code. Alternative Inspection or Test .I None. Reference l Bechtel Drawing 0740G40107, Revision M. Implementation Not applicable. \\ t l i i l I' r i 1 i i l rp0382-0621g-42-48 III-60 i t
March 4, 1982 Revision 2 SUBSECTION C - CLASS 3 Table IWD, Items D1.2, 2.2, 3.2: INTEGRALLY WELDED SUPPORTS FREQUEho.' 0F EXAMINATION RR-C1 Basis 2tatement Table IWD required visual examinations (VT-3) to be performed durir.g each inspection period. Implementation of this requirement is imprac-tical in view of the examination requirements for Class 1 and Class 2 components. Remarks The Section XI Code requirements for Class 1 component supports (Table IWB-2500, Categories B-H and B-K-1) and Class 2 component supports (Table IWC-2520, Categories C-C and C-E-1) require less frequent examinations of component supports than the frequency required for supports for Class 3 components. The requirements can be summarized as follows: Class Table Category Examination Frequency 1 IWB-2500 B-H 100% for the First Two Intervals B-K-1 25% in Each Interval 2 IWC-2520 C-C 100% in Each Interval C-E 100% in Each Interval 3 IWD-2500-1 D-A, D-B, 100% in Each Period D-C l In view, then, of the less frequent examination requirements for l Classes 1 and 2 component supports and nonintegrally welded supports, examination of 100% of the integrally welded supports during each interval in lieu of examination of 100% during each period should be adequate. Alternate Inspections or Tests None. rp0382-0621g-42-48 III-61
O March 4, 1982 Revision 2 References Big Rock Point ISI 40-Year Master Plan, Class 3 examinations. Implementation Not mplicable. I I i l l l l l l l rp0382-0621g-42-48 III-62 l l
March 4, 1982 Revision 2 SUBSECTION D - PRESSURE TESTS CLASSES 1, 2 AND 3 Table IWB-2500 (Category B-P), Table IWC-2520 (Category C-H) and Table IWC-2500-1 (Items D1.2, D2.1 and D3.1) SUBSTITUTION OF SYSTEM HYDROSTATIC TESTING FOR SYSTEM PRESSURE TESTING RR-D1 Basis Statement Table IWB-2500 (Category B-P), Table IWC-2520 (Category C-H) and Table IWD-2500-1 (Items D1.2, D2.1 and D3.1) require the performance of both system pressure tests and system hydrostatic tests during the third inspection period. Implementation of the requiremant for system pressure tests during the third inspection period is inappropriate in view of the performance of a system hydrostatic test. Remarks In view of the higher pressures applied during the system hydrostatic test, it represents a more severe test than the system pressure tests which are required by the Code; therefore, performance of the system hydrostatic test would make performance of the system pressure test redundant and unnecessary. Alternative Inspection or Test None. References l Big Rock Point 40-Year ISI Master Plan. Implementation Not applicable. l rp0382-0621g-42-48 III-63 I t
March 4, 1982 Revision 2 SUBSECTION E - GENERAL 1. Items 5.50, B9.11, B9.12, B9.51, C5.21 and C5.22 (Pressure Retaining Piping - Volumetric Examination). Basis Statement ASME B&PV Code, Section XI, 1977 Edition with Addenda through Summer 1978, Paragraph IWA-2232(b) requires the utilization of Appendix III, Section XI for ultrasonic examinations of piping. Appendix III, Paragraph III-1100(d) permits the utilization of alternative examination techniques and calibration block design in accordance with IWA-2240 (Alternative Examinations). Consumers Power Company elects to utilize hole type reflectors in lieu of notch reflectors required by Paragraph III-3430 for ultrasonic examinations performed on piping thicker than 0.400". In order to effect the change in reflector type, Consumers Power Company elects, per IWA-2232, to conduct ultrasonic examinations of piping welds in accordance with Article 5, Section V, ASME Codes. The requirements of Article 5, Section V will be modified by the attached code applicability statements. Discussion Utilization of hole type reflectors for material thicker than 0.400" in lieu of notch type reflectors will increase the sensitivity of the examination by a factor to two (2) over the standard 45* angle beam examination. Consumers Power Company will implement the two code applicability statements attached in order to affect this change in reflectors. The code applicability statements will enhance procedure and calibration control and control of calibration block procurement /fabri-cation. Relief Requested It is requested that relief be granted from the provisions of IWA-2232 with respect to calibration block design and calibration procedures and that the provisions of the two (2) code applicability statements be sub-stituted in lieu thereof. rp0382-0621g-42-48 III-64
o ATTACICENT II REQUEST FOR ADDITIONAL INFORMATION INSERVICE INSPECTION PROGRAM Big Rock Point 4 Request 1. Request for Relief, RR-Al (B-A; Bl.11, Bl.12) This relief request concerns volumetric examination of longitudinal and circumfer-ential pressure retaining velds in the core region of the reactor vessel. Access to these velds is prevented on the outside by the concrete shield and on the inside of a non-removable thermal shield. The following additional information is requested: (a) Identify the portions of any velds which might be accessible for examination because of their proximity to any access ports or other penetrations in the concrete shield. (b) Is it possible to perform surface or remote visual examination on these velds to supplement volumetric examinations? Please provide an estimate of the percentage of each veld which may be accessible for these supplementary examinations.
Response
(a) The surrounding concrete shield and plant design makes external access to longitudinal and circumferential reactor vessel velds i physically impossible. The concrete sheild surrounds (actually encloses) the reactor piping and varies in thickness from 6 to 10 feet where piping penetrations are located. Removal of this concrete shield would be sn enormous task in a high radiation area (500 to 1500 mr/hr). j (b) The concrete shield and exterior insulation on the reactor vessel vould not permit surface or remote visual examination of longitud-inal and circumferential velds from outside the vessel. The non-removable ther=al shield and stainless cladding inside the vessel fully cover the lower longitudinal velds and circumferential velds and therefore cannot be seen via visual examinatien from in-side the vessel.
2 Fequest 2. Request for Relief, RR-A2 (B-A; Bl.12) This relief request concerns volumetric examination of the longitudinal pressure-retaining velds in the reactor vessel shell above the beltline circumferential veld. Apparently, these velds are totally inaccessible from the outside due to the external shield vall, and 70 to 75% of the length of these velds is inaccess-ible from the inside due to interferences by internal components. For the 25 to 30% of these velds which is accessible from the inside of the vessel, no technique is claimed to be available for examination of the longitudinal velds. The following additional information is requested: (a) Please identify the techniques for volumetric examination of the longitudinal velds that you have surveyed which are not appropriate for this application, and justify your position that no techniques are available for the examination of the longitudinal velds. (b) Is it possible to perform surface or remote visual examinations on these velds to supplement volumetric examinations? Please provide an estimate of the percentage of each veld which may be accessible for these supplementary examinations.
Response
(a) Since our last submittal, Consumers Power Company requested SvRI to ( modify the programmable and remote (par) mechanised inspection device. [ With some modification to the containment crane it is expected that about 30% to 50% of the longitudinal velds above the beltline circum-- ferential veld will be accessible for volumetric examination. The longitudinal velds (#749-1A & Th9-1B on ISO-A1) above the beltline veld have about 60% of their length obscured by internal components s (core spray sparger, baffle plates, thermal shield top ring and ther-mal shield). (b) The concrete shield and exterior insulation on the reactor vessel vould not permit surface or remote visual examination of the longi-tudinal and circumferential velds from outside the vessel. The
3 non-removable thermal shield inside the vessel fully covers the circumferential beltline veld and a portion of the upper longitud-inal velds. The upper longitudinal velds are also obscured in some areas by Lne core spray ring and primary core spray line. These velds also covered with a stainless steel cladding their full length and therefore cannot be seen via visual examination from inside the vessel. I Portions of these velds have now been identified for volumetric examinatien in RR-A2 (Rev 2). Request 3. Request for Relief, RR-A3 (B-A; Bl.21. Bl.22) Relief is requested from volumetric examination of the meridional and circumferential velds in the reactor vessel bottom head. The meridional velds are inaccessible from inside the reactor vessel because of the core support plate. Poor gecmetry due to LO penetrations on the bottom head makes the meridional velds inaccessible ufran the outside. The circumferential shell-to-bottom head veld is not accessible with existing equipment from the inside and the concrete vall prevents inspection from the outside. The following additional information is requested: (a) Identify the portions of any velds which might be accessible for 1 examination because of their proximity to any access ports or other penetrations in the concrete shield. (b) Is it possible to perform surface or remote visual examination on these velds to supplement volumetric examinations? Please provide an estimate of the percentage of each veld which may be accessible for these supplementary examinations.
Response
l I (a) & (b) No access ports or penetrations are known to afford access to the j meridional and circumferential velds in the reactor vessel bottom i head. A recently fabricated access port in aggregate trays belov l the reactor vessel may provide extremely limited access to a small l l ~
o k portion of some of these velds and will be investigated during our 1983 refueling outage. It is considered at this time that adequate access is not available through this newly fabricated port _ to allow for external preparation and examination activities on these velds. Inside visual examination of these velds is not feasible due to core support plate structures and stainless steel cladding. Request k. Requests for Relief, RR-Ah, 6-9, 11-13 (B-D, B3 90, B3.100: B-F, B5.10 ) These relief requests concern volumetric examination of primary nozz1 -to-vessel t velds and nozzle inside radius sections (Category B-D) and nozzle-to-safe-end welds (Category B-F). In each. case, access from outside the reactor vessel is prevented by the concrete shield. From the inside, mechanized volumetric inspection is not possible due to interferences and equipment limitations. During baffle repair on some of these nozzles, remote visual inspections were co=pleted with no indica-tions noted. The following additional information is requested: (a) Are you prepared to commit to a remote visual examination of the accessible areas of these nozzles? Please provide an estimate of - the percentage of each area which may be accessible for a remote visual examination. Respgnse (a) Since our last submittal, Consumers Power Company requested SvRI to i modify the par device to accommodate the reactor at Big Rock Point. We anticipate accessibility to examine most of the nozzle-to-safe-end and nozzle-to-vessel velds, including incide radius sections. The exceptions vill be the primary core spray nozzle, the two 20" recirculating water inlet nozzles, the 3" liquid poison inlet nozzle and the shutdown unloading nozzle; due to the internal configuration of the reactor. We are prepared to co=mit to a remote visual exami-nation of the accessible area of these nozzle which can't be ultra-sonic tested.
a 5 Dequest 5. Request for Relief. RR-A5 (B-D; B3.90, B3.100) This request for relief applies to surface examinations of these velds and inside radius sectionc on the ik-inch steam outlet nozzles. These nozzles are inaccessi-ble for surface examination, as they are submerged in shielding water during refuel-ing. Mechanized ultrasonic and visual examinations are proposed as alternatives to surface examination of these areas. Table IWB-2500-1 of the code (Ref. 2) requirec only volumetric examination for these areas. Since your request states that ultrasonic examination is possible, please explain why you believe a request for relief'io necessary for this item.
Response
This relief request hae been deleted. Request 6. Request for Relief, BR-Alb (B-H; E8.10) Relief is requested from volumetric examination of the 12 suspension rod brackets which are integrally velded to the reactor vessel. Their locations are not accessi-ble to ultrasonic examination equipment. Access for remote visual examination was to be explored during the next outage. The following information is requested. (a) What information has been learned concerning your ability to perform a remote visual inspection in these areas, since Ref. 1 was issued? (b) Will you commi,t to performing remote visual examinations as alter-natives to the required volumetric examinations?
Response
(a) Basically our position remains unchanged, the use of ultrasonic equipment current with present technology indicates thisoexam is inconceivable due to the non-removable thermal shield. Remote visual equipment may offer some access to a portion of the 12 suspension rod (bolts) and hanger plates (MK #802-1 on GE Dvg E-201-801-5). However, the insulation which covers the reactor
e 6 externals make the support brackets unaccessible for remote visual exam where integrally velded to the vessel exterior. (b) We vill commit to performing a limited remote visual exam per the restrictions described above. We have scheduled the use of this equipment (remote visual) for our 1983 refueling outage. Request 7. Requests for Relief, RR-A15, 16 (B-H; B8.10) Relief is requested from volumetric examination of the vessel hanger lugs and stabilizer brackets. The locations of these items make them inaccessible to volumetric examination. These components are said to be non-load-bearing in that under normal operation' they are subjected to no shear stresses. Defects in these velds are claimed to have minimal effect on the components' support functions. The following additional information is requested: (a) The velds which attach these support components to the vessel are under tension as is the base material of the vessel. These velds are also areas of stress concentration; please show why a flaw in one of these velds would not propagate to the vessel base material-and, therefore, affect the primary pressure boundary. (b) Have you considered a rigorous visual inspection program as an alternative to the volumetric examination of these velds?
Response
(a) The hanger lugs (RR-A15) and stabilizer brackets (RR-A16) are non-load bearing (CE Dvg E-201-801-5) and restrict movement.in a torsional direction where integrally veldea to the reactor vessel. The applica-tion of the par mechanized inspection device scheduled for the 1983 outage mechanized inspection may allow examination of these bracket and lug velds and base =aterial from inside the reactor. The problem still being considered is how credible the ultrasonic test results would be on a veld of this nature. i
e: 7 We expect that through rigorous testing of the par device using mock-ups of the same composition confidence vill be acquired. Until then, the reactor vessel base material integrity is proven by pre-startup hydro testing. This test is performed at 1.1 times the nor-mal operating pressure. (b) The remote visual exam scheduled for the 1983 outage for the support rods will allow evaluation to this area of the vessel. However, the insulation on the exterior surface may render visual methods ineffec-tive for the hanger lug and stabilizer brackets veld examinations. Request 8. Requests for Relief, RR-A17,18,19 (B-N-2; B13.20) The code requires visual examination of the sparger and core spray supports, the thermal shield supports and inlet diffuser brackets, and the core support brackets as they become accessible during a normal refueling outage. With the exception of items which were made accessible for other reasons, your requests for relief pro-pose the visual examination of only accessible portions of the components. Your requests appear to be consistent with the code requirements of Ref. 2, Table IWB-2500-1 for this item. Please state why relief from code requirements is nee-essary on these items. i
Response
l l Relief requests have been deleted. Request 9 Request for Relier, RR-A20 (B-B, B2.51, B2.52) RR-A21 (B-D; B3.150, B3.160) RR-A29 (B-H; B8.h0) l l These requests for relief apply to examination of various velds on the steam drum. Concern is expressed that very high radistion exposures would result from perform-ing these code examinations on all required welds of the steam drum. Alte rnatively, it is proposed that only velds in these groups which are located on top of the steam drum within 30 degrees either side of top-dead-center should be examined. The pro-posed examinations are said to give sufficient information as to the integrity of the steam dru=. Please provide drawings of the steam drum which show the locations L L
8 of the longitudinal and circumferential velds, and the locations of all the pene-trations and their details. Drawings showing the location and details of the steam drum supports should also be provided. If these drawings have previously been furnished to the NRC, please document by reference.
Response
(a) As indicated on CE Dvg E-230-101-9 and Isometric'Dvg A-lh & A-15, circumferential velds are located starting 3 inches from the head approximately 11 feet, 5 inches apart. The longitudinal velds is located down the center line of the drum. The location of steam drum penetrations and details, steam drum supports and details can be found on the drawings noted. Request 10. Request for Relief, RR-A2h (B-D; B3.150, B3.160) Relief is requested to exempt inside radius sections of the cleanup demineralizer tank nozzles from volu=etric inspection requirements. The demineralizer tank nozzles are fabricated by velding square-ended pipe nipples into the tank shell. The portien l of the pipe nipple corresponding to a nozzle inside radius section has a radius of l essentially zero, and apparently cannot be UT inspected. It is also noted that these l l nozzles are not subjected to the thermal cycling stresses co= mon to other primary system components because the inlet temperature is required to be less than 10d*F. For this reason, it is claimed that the examinations are not required. The following additional information is requested: (a) Provide justification for your position that these examinations are not required for components that do not experience significant thermal cycling. 1 (b) Provide an estimate of the additional radiation exposure which would be incurred if these inspections, along with the re=ainder of the Category B-D testing, were to be attempted.
Response
(a) The examination of the nozzle inside radius section was incorporated ?
9 into the code primarily to detect thermal cycling stress degradation. The primary system water flowing into the clean-up demin tank first passes through the clean-up regenerative and non-regenerative heat exchangers to reduce and stabilize the water temperature to below 0 100 F to prevent decomposition of the demineralizer resins. This tank is therefore not subject to significant thermal cycling and inspection of the nozzle inside radius sections is therefore not required. (b) The additional radiation exposure which would be incurred according to Big Rock Point Radiation Protection personnel would be:
- 1) 1 1/2 R/hr on contact
- 2) 600-800 MR/HR General Field Request 11.
Request for Relief RR-A30 (B-F; B5.50) RR-A33 (B-J; B9.11) RR-A3h (B-J: B9 21) RR-A36 (B-K-1; B10.10) These requests for relief apply to nondestructive testing of the following pipe velds: (a) 6-SCS-101-1 Thru 12 (c) 12-MSS-105-9 and 10 (b) 3-CSS-101-17 thru 19 (d) 3-LPS-102-18 thru 23 and the folleving pipe attach:nents velds: (e) 1h-MRS-103-3PL-1 thru 8 (f) 1h-MRS-105-3PL-1 thru 8. Relief is requested from examination of these welds due primarily to various combi-nations of long vertical drop with attendant inaccessibility and high radiation fields. The following additional information is requested: t (a). For. each of the syste=s listed above, provide the percentage of welds which can be inspected. (b) ' Provide more details on why scaffolding cannot be built on a eloping floor with proper restraints. '(c) Provide more information on the methods that were considered for reaching the velds with long, vertical drops. Also, based on current-industry i
10 safety standards, show why each of these methods was rejected.
Response
(a) The percentage of welds en line 6-SCS-101-1 thru 7 and line 8-SDS-101-8 thru 12 which can be examined is 0%. However, of the "SCS" Shutdevn Cooling System these velds represent only 26% making Th% of the exams accessible. Lines 3-CSS-101-17 thru 19 and 12-MSS-105-9 & lo, approximately 95% of the required exams can be completed. The velds on line 3-LPS-102-18 thru 23 represent a small portion of welds requiring inspection. Approximately 98% of this line vill be accessible. On the Main Recir-culation System (1h-MRS-103-3PL-1 thru 8 and 1h-MRS-105-2PL-1 thru 8), approximately 98% of this system vill be accessible during normal inservice inspection activities. The above percentages were obtained using Table I-1, of the Request for Relief (Rev 2) (proposed) and the Big Rock Point h0-Year Master. (b) The plantdesign is such that a minimum structure of 35' to 50' in elevation would need to be erected for most areas requiring inspection. The base of each structure erected may require velding to the concaved internal surface of the containment sphere to stabilize it on the slop-ing floor. This could jeopardize containment integrity. The structure would then ascend interlaced among numerous pipes and lines impairing the structure's stability, and could not be erected in a straight vertical line. (c) Alternate methods such as overhead rigging and boatsmen chairs were considered but no vertical drops could be obtained using structures available. This would result in suspending a person using block-and-tackle and svinging them to the area to be examined where they would lash themselves to the line with rope or equivalent safety harnesses to minimize the predominant safety hazards. The benefit is not con-sidered worthy of the additienal man rem and personnel safety hazard. -.,r-r
11 Plant maintenance staff and radiation protection personnel vere consulted for alternate means of access and confirmed the methods considered would only jeopardize human safety and propogate excessive radiatien exposure. Request 12. Request for Relief, RR-A39 (B-L-1; B12.10) This item concerns volumetric and surface examination of cleanup pu=p and main recirculation pump casing velds. Relief from these requirements is not requested because the locations of the velds in these pump casings cannot be deter:dned. The following additional information is required: (a) To date, have you determined the locations of these velds? Have you determined whether or not relief from volu=etric examination require-ments is required? (b) If relief is required, show why ultrasonic examination of these velds cannot be performed.
Response
(a) During the Big Rock Point 1982 refueling outage, access engineering was conducted on the clean-up pump and =ain recirculation pump casing velds by Southwest Research Institute (SvRI). SwRI indicated that a surface and/or visual examination could be perfor=ed. The veld in question (=ain recirculation pump) is identified on the enclosed Byron Jackson pump drawing 1F h61h-3 l (b) Due to the internal configuration (FINS) of the pump, the use of ultrasonic examination on these velds is not practical. Request 13. Request for Relief, RR-Aho (B-M-2; 311.h0) Relief is reauested from these requirements to visually inspect the internal surfaces of various main recirculation pump discharge and butterfly valves. Examination
12 5 of the valves requires complete draining of the reactor vessel. Are you prepared to commit to a visual inspection of any of these valves' internals if_ they are required to be disassembled for maintenance?
Response
The main recirculation pump discharge and butterfly valves are theoretically not isolable without draining the reactor. However, should it be necessary to dis-assmble these valves for maintenance, we are prepared to commit to visual inspection.
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