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Forwards Inservice Insp Program for Welds & Supports, Revision 0.Tech Specs Are Being Reviewed to Identify Conflicts W/Section XI of ASME Boiler & Pressure Vessel Code
	ML19208A312
Person / Time
Site:	FitzPatrick
Issue date:	09/10/1979
From:	Early P; POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T; Office of Nuclear Reactor Regulation
Shared Package
ML19208A313	List: ML19208A312; ML20136B967;
References
	JPN-79-57, NUDOCS 7909130356
	Download: ML19208A312 (2)
	v • d • e

r POWER AUTHORITY OF THE STATE OF NEW YORK 10 COLUMBUS CIRCLE NEW YonK. M. Y. 10019 (212) 307-0200

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b$fa September 10, 1979 JPN-79-57 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Inservice Inspection Program - Inservice Examination of Welds and Supports

Dear Sir:

As required by 10 CFR 50. 55a (g) , the Power Authority of the State of New York has performed a review of the Inservice Inspection Program requirements for the James A. FitzPatrick Nuclear Power Plant. This letter forwards the results of that review.

Under the provisions of 10 CFR 50.55a, inservice inspection and testing will be performed in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda which will be referenced simply as the " Code".

As further specified in 10 CFR 50.55a(b), the effective edition of the Code with regard to :his submittal is the 1974 Edition through the Summer 1975 Addcada. Based on the date of commencement of commercial operation, 10 CFR 50.55a requires that the program for inservice inspection examinations of welds and supports be updated at 40-month intervals to ensure compliance with the effect-ive Code. The ctarting date for the second 40-month period for the James A. FitzPatrick Nuclear Power Plant is November 28, 1978.

Enclosed as Attachmant 1 is the Inservice Inspection Program Update report for the James A. FitzPatrick Nuclear Power Plant.

This program has been prepared to comply with Code requirements to the extent practicable. Specific requests for relief from Code requirements are included in the Program report as provided by 10 CFR 50.55a(g) (5) (iii) . Additional exceptions to Code requirements 9g g

\ g 7909"o Q Y

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U.S. Nuclear Regulatory Commission may be requested of the Nuclear Regulatory Commission as the program is fully implemented. Additions and changes to the test programs which comply with Code Testing requirements will be implemented as needed without prior Nuclear Regulatory Commission notification.

The Technical Specifications for the James A. FitzPatrick Nuclear Power Plant are currently being reviewed with the objectiv e of identifying conflicts between Technical Specifications and Section XI Code requirements for inservice examination of wel65 and supports. Relief from any inconsistencies between the Techni:al Specifications and the Inservice Inspection Program will be stbmitted at a future date, as a request for change to Technical Specifications.

Very truly yours,

/ s' Pau J.

.hEarly Ass ~stant Ch'ef Engineer-Projects Attachment 1 - Inservice Inspection Program Update (3 copies)

Inservice Inspection Diagrams (1 set) b4)$17h

. i

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^

JAMES A. FITZPATRICK NUCLEAR POWER PLANT INSERVICE INSPECTION PROGRAM FOR WELDS AND SUPPORTS

} Revision 0 May 1979 5

Prepared By: [ Date: May 25, 1979 M. J Partridge [ eneral Physics Corporation Approved By: r Date: May 29, 1979

_ J. 11 . Beakes, General Physics Corporation i /

PAS?JY Approval: f(b/ oc.f f N c'lq[f( Tx( q Date: of /, / Y/[

/ ./AI V. Sorrentino d '

Power Authority of the State of tJew York m

4

1 INSERVICE INSPECTION PROGRAM FOR WELDS AND SUPPORTS

_ FOR Tile JAMES A. FITZPATRICK NUCLEAR POWER PLANT TABLE OF CONTENTS Section Title Page TABLE OF CONTENTS . . . . . . . . . . . i

1.0 INTRODUCTION

. . .. . . . . . . . . . 1 l.1 Program Update Description . . . . . . 2 l.2 Technical Approach . . . . . . . . . . 3 l.3 Key Technical Points . . . . . . . . . 4 1.3.1 Code Exemptions . . . . . . . . . . 4 l.3.2 Alternate Examinations for Exempt Components . . .. . . . . . . . . 4 1.3.3 Relief Requests . . . . . . . . . .

4 1.3.4 "In Lieu Of" Examinations . . . . .

4 1.3.5 Containment Penetrations . . . . . .

5 l.3.6 Pressure Tests .. . . . . . . . . .

5 o

1.3.7 Class 3 Systems and Components . . . 8 1.3.8 Repair Provisions , , , , , , , , ry Appendices A-1 EXAMINATION

SUMMARY

FOR ASME CLASS 1 SYSTEMS AND COMPONENTS . . . . . . .

A-2 EXAMINATION

SUMMARY

FOR ASME CLASS ?

i SYSTEMS AND COMPONENTS . . . . . . .

B PIPING WELD AND SUPPORT EXAMINATION PROGRAM . . . .. . . . . . . . . .

C INSERVICE DIAGRAMS (One set to be supplied with NRC Submittal) . . . .

s

INSERVICE INSPECTION PROGRAM FOR WELDS AND SUPPORTS FOR THE JAMES A. FITZPATRICK NUCLEAR POWER PLANT 1.0 Introduction In the past, there have been divergent opinions on the re-quirements for Inservice Inspection of nuclear power plant com-ponents. As specified in 10 CPR 50, the current Code in effect is the 1974 Edition of the ASME Boiler and Pressure Vessel Code Section XT i-hrough the Summer 1975 Addenda (herea f ter designated as " Code"). There is a general consenrus that the edition of the Code accepted by the Nuclear Regulatory Commission (NRC) provides a reasonable level of inspet on that wi11 help to ensure the pressure boundary integrity of essential systems of a boiling water reactor plant such as James A. FitzPatrick. However, there are areas needing further refinement.

For e : ample , the Code permits systems that perform an emergency core cooling function to be tota?ly exempted from

_ examination provided the control of chemistry of the contai.ned fluid is verified by periodic sampling and test; whereas it re-quires volumetric examination of many fillet welds that are not inspectable by that examination method. Further, since the ad )ption of the Summer 1975 Addenda, the ASME Section XI has issued new requirements in its later editions and addenda, and the NRC has released several guideline letters relating to in-service inspection. The new Code requirements and the guidelines promulgated by the NRC will affect future updated inservice

- innpection programn for tho Jamo: A. FitzPairiek P1 ant, and have

.= been consiuered in developing the Program deucribed herein. The Progran conforms to the purpose and intent of performing the Code required inservice inspection and will orovide assurance of tne i.nteg r i ty of the pressure boundaries of essential systems.

1 Md'!h

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I This report contains the complete Inservice Inspection weld l examination and component inspection program for the second 40-month operating period. The technical bases for any deviations from the Code are also identified. In accordance with the current requirements of 10 CFR 50, this Inservice Inspection program will be applicable for a 40-month period ending March 28, 1982. 1 Subsequent review and update will be in accordance with the 10 CFR 50 requirements in effect at that time.

1.1 Program Update Description The specific areas to be examined and the extent of examina-tions to be performed during the first 10-year operating interval are summarized in Appendices to this report as follows:

Appendix A Examination Summary for ASME Class 1 Systems and Components Appendix A Examination Summary for ASME Class 2 Systems and Components

Appendix B - Piping Weld and Support Examination Program To facilitate review, Appendices A-1 and A-2 are pre 'ed in a format consistent with that of Tables IWB-2600 and IWC-260u respectively, of the Code. These summaries provide the Item Number, Examination Category, Areas to be Examined, Examination Me thod ( s ) and Extent of Examination (s) to be Performed, as identified in the Code. Information on accessibility and other related considerations have been included along with references to Appendix B in the cases where Examination Method (s) and Extent of Examination (s) to be Performed are presented in the Piping Weld and Support Inspection Program.

1 A proposed 10 CPR 50 rule change recently published for comment in the Federal Register would leave the Program Update described in this report in effect for the remainder of the first full 120-month inspection interval which began with the commercial operation date of July 28, 1975. In the event that this proposed rule change is incorporated into 10 CFR 50, this Program Update will be effective until July 28, 1985.

2 ' '

4

Appendix B. includes: descriptive information and definitions of (1) Code Exemptions, (2) Relief Requests, and (3) Weld Summary Sheet Terms; a Weld Summary Sheet Index (listing of systems in-cluded in the Program); and Weld Summary Sheets for each system.

Appendix C. represents the color-coded Inservice Diagrams (ISD's) which define the ASME Class 1, Class 2, and Class 3

, boundaries for those systems listed in Appendix B.

i

.m The inspection program for ASME Class 3 systems and com-ponents is covered in Section 1.3.7 of this report.

3 1.2 Technical Approach In general, the inservice inspection requirements in the 1974 Edition through the Summer 1975 Addenda which are applicable

_ to ASME Class 1 components and systems are considered to be

( sufficient and workable to provide a reasonable and prudent level of inspection that will help to ensure the pressure boundary integrity of essential systems of a boiling water reactor plant.

However, the Code requirements applicable to ASME Class 2 com-ponents and systems are considered in many cases to need enhance-The ASME has recognized various aspects that need improve-

~

ment.

] nent in the 1974 Edition through the Summer 1975 Addenda and has 2 subsequently made many improvements in its later Editions and Addenda. These later Editions and Addenda (e.g., 1977 Edition, 2 Summer 1978 Addenda) are not currently recognized by 10 CFR 50, and therefore, are not used directly in the Inservice Inspection

__ Program Update.

- The Inservice Inspection Program Update was prepared to the

~

degree practicable in accordance with the Code requirements.

Where the Code requirements were determined to be either incom-plete or impracticable, specific requests for relief from Code requirements are identified. Each request for relief has been evaluated and an alternative examination is specified. In ne, r

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3

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m almost every case the alternative examination results in a higher level of confidence in system integrity than the original Code requirement (the exceptions to this are welds that are physically inaccessible for any type of examination.

l.3 Key Technical Points

~

1.3.1 Code Exemptions

-c Exemptions identified in IWB-1220 and IWC-1220 of the Code 7 are used in this Program with the exception of lWC-1220(c),

which exempts emergency core cooling systeins if the chemistry of their contained fluid is controlled.

1.3.2 Alternate Examinations for Exempt Components In accordance with Category B-P of Table IWB-2500 for Class 1 and IWC-2510 for Class 2, components exempt from volu-netric and surface examination will be visually examined during pressure test in accordance with Code requirements. In

-[ Appendix B, such examinations are listed under "In Lieu of Examination" where appropriate.

l.3.3 Relief Requests

=_

Where Code requirements are determined to be incomplete or impractical, requests for relief are documented and techni-cally justified in the notes and remarks in Appendices A-1 and A-2 and in Table B-2 of Appendix B.

_ l.3.4 "In Lieu of" Examinations Where relief is requested from examination of component,

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an "in lieu of" examination is specified. In most cases, the "in lieu of" examination results in a higher level of confi-3 dence in system integrity than the original Code requirements.

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I 1.3.5 Primary Containment Penetration We_lds Appendix B specifies a surface examination, if the weld is accessible, for all primary containment boundary to piping welds in all high energy steam or water containing piping greater than three inches nominal pipe size. This exceeds the Code requirement in some cases, and is therefore an "aug-mented" examination, but is considered necessary because of the importance of the integrity of the primary containment boundary.

I The containment penetration is considered to extend to the first isolation valve outside containment and the first seismic restraint past that valve.

l In addition, system piping greater than 4 inches nominal

.l pipe size that is directly open to the primary containment atmosphere, i.e., Containment Air Dilution (CAD) and Drywell to Suppression Pool Vacuum Breakers, will have " augmented" surface examination out to the outer containment isolation I> valve and the first seismic restraint past the valve. (Note that since these systems are air systems, they are not within the scope of ASME B&PV Code,Section XI.) The " Areas Subject to Examination" and the " Extent of Examination" of Examination Category C-G shall apply tc all above augmented examinations.

The typical containment penetration is constructed such that one circumferential butt weld inside the penetration is completely inaccessible for either surface or volumetric in-spection. This weld is shown in Figure 1-1. Each of these inaccessible welds is identified in Appendix B and an "in lieu of" examination is specified.

1.3.6 Pressure Tests In the Weld Summary Sheets, three different cypes of pressure tests are used Descriptions of the three types are given in Table 1-1. The system " Pressure" test identified as the "In 5 34hN

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Lieu of Test" for Relief Request R2 (Class 2 systems that operate above moderate energy less than one percent of plant operating time) is intended to increase the overall confidence in the pressure boundary integrity of those systems or portions of s ys ten.s using Relief Request R2. This test method will verify the integrity of 100 percent of the welds every 3 1/3 years rather than a small portion (can typically be as small as 2 percent of the welds during the same 3 1/3 year period) of the welds as required by the Code.

7_

Table 1.1 Pressure Test Descriptions

] Weld Stcmary Test "ame Sheet symbol Test Description ilydrostatic HYD The Code required hydrostatic pressure test 1.erfomed each interval as specified in Section XI Paragraphs 1WA-5000, IWB-5000, IWC-5000, and IWD-5000, as applicable. The hydrostatic test pressure shall be based on 5 the Coda requiremonts; however, overpressuri-zation of the most limiting component in a system or portion of a system is not permitted.

For example, piping attached to the reactor vessel will be tested wi.th the reactor ves-sel rather than to their own test pressure.

=

Pre >sure i% A pressure test performed on a system or portion of a system as the "In Lieu of Exam" for Relief Request R2. The test shall be performed every 3 1/3 years. For components that are not required to function during reactor operation, the system test pressure

_, shall not be less than 100 percent of the pressure developed during the conduct of a

._ periodic system inservice test [IWC- 52 20 (c ) ] .

Flow FLN A test performed to portions of systems that l cannot be hydrostatic pressure tested, such as lines connected to open tanks. The test shall be performed during system operation every 3 1/3 years The test pressure shall he that prer sure attaincd during normal system operation.

3

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Class 3 Systems and Components 1.3.7 Class 3 systems at the James A. FitzPatrick Plant fall into two categories: (1) the Reactor Core Isolation Cooling (RCIC) system, and (2) cooling water systems. The RCIC system has been designed to be able to be pressure tested and therefore, will be tested in full accordance with the Code requirements.

The cooling water systems were designed and constructed prior to the issuance of the first Code Edition that addressed Class 3 systems and components (the 1974 Edition). Since the systems were designed and constructed without the current inspection require-ments in view, isolation valves that would permit hydrostatic testing to current requirements were not installed.

_ An alternative inspection program described below is there-fore implemented to insure the integrity of the pressure boundary of these cooling water systems. This program incorporates features of the 1977 Edition of the Code through the Summer 1978 Addenda for visual inspection of piping systems and supports. To provide

] additional assurance of system integrity, these inspections will be condacted during system functional tests once per cycle (which exceeds Code requirement;). The inspection program for Class 3 cooling water systens is as follcws:

Cooling Water System Components Examination

_ Subject to Examination Method Reference 1 Piping and Components (Not Buried) VT-2 IWA-2212 Piping and Components (Buried) Flow Verification IWA-524 4 (c)

Piping Supports VT-3 IWA-2213

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^-

All references are from the 1977 Edition of ASME: Section 5:I throurJh the S um:ne r 1978 Mdenda.

. \

3%;s2 8

I I 1.3.8 Repair Provisions Because of the vintage of the James A. FitzPatrick Nuclear Power Plant, not all systems and components were originally de-signed and fabricated to the ASME Boiler and Pressure Vessel Code.

Systems and components re-classified herein in accordance with the inservice inspection requirements of 10 CFR 50, 50. 2 (v) , 50.55a(g) e and Regulatory Guide 1.26 as ASME Class 1, 2 and 3 shall be examined in accordance with this program. Consequently, relief is requested from the requirements of Articles IWA-4000 of ASME Section XI l Repair Procedures and the subsequent Articles IWB-4000, IWC-4000, and IWD-4000.

1 In lieu of these requirements in order to assure quality and e integrity of the repaired and/or replacement component, the following shall be accomplished:

1.3.8.1 Codes and Standards Repairs on the components of the systems clasrified in the inservice inspection program as ASME Class 1,2, and 3, shall be made in accordance with the Code or Standard uscd for the original fabrication (later approved Editions and Addenda may be used).

1.1.8.2 Quality Assurance All work pertaining to repairs and replacements shall be performed in accordance with the requirements of the exist-ing James A. FitzPatrick Quality Assurance Program.

1.3.8.3 Pressure Test After repairs by welding on the pressure retaining boundary of components (except repairs on cladding), a pressure test shall be performed in accordance with the provisions of IWA-5000, 1974 Edition, Summer 1975 Addenda.

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_ l.3.8.4 Re-oxamination Re-examinations shall include the method that detected the m

flaw requiring repair and shall be used to establish a new a

preservice record.

1.3.8.5 Procedure Review All repair procedures shall be available for review by the

, enforcement authorities having jurisdiction at the plant h site.

2

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18 6 10

Altf NDIX A-1 E X Af/IN ATION 5UYV ARY AEYE CLASS 1 SYST EMS AND COMPONENTS l'I LM EX AMIN A'llhN C()Mh >N EN IS ANI) l'ARIS L\ AMIN A ll(>\ EX ITNT OF LN AMIN A TION NO.I P A I WO R Y 1 10 bE EX AMINFl> Mi.m m3 l'F RUE N ! IN 10 YEARS I ACCl %IHil.lT Y NO'l IS Fr act or Ws se 1 r

81.1 b- A Feactor Iressule vessel lona.- Vol u: net r ic kr.y . S-10 The use of doors in the sacrifi- 1 tuJinal an c ti r c an t e r en t i a l welds Circ. d-5 cial shield and thermal '.uula-I4 in cure region Af ter expos n e to IO NVT tion provide O.D access to 5% of of fast ne ta r on s , both t i.e cir cumfer er.t t al and lut of the long, and circ. sean:s J or: git udinal welds.

tequire 50s examination Fl.2 b-b Closure head circumferential and Volumetric Circ. 5-5 -

u rili.nal welds Merld. S-10 Ecactor pressure vessel lorq 1- Vol ume t r i c I/>r.g . ^.-10 2 y tedinal and circurrferent ail weld, care. 5-5 y above sacrificial shield a

G -) kng . 5-10 Keactor pressure vessel welds Volumetrie 0.D. access to the welds within 3 a

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g within sacrificial shield and above Jet r timi, suprort [ late Circ. S-L the sacrificial shield will be rrovided by the opening of doors in the shield wall and plugs in r 3 the insulation.

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b i Bot tora ht.id c i rcumr er ent i al and voltimetric C2rc. S- 5 Access to the Lottom head welds -

me r ld lor.al we l ds - nr L1 t o Merld. S-lO outside the suiport skirt are I vessel weld and meridional welds severely limited due to insulation.

Nt>- outside the vessel s upp,: t er.irt structulal interferences, and l inst r ument ation leads.

b Bot t om head cir c umf er er.t id l ar? Vol u: net ri c Cire. S-S 0.D. access inside of the support 4 merilicnal welds inside sup;crt .Me r ld . C-19 skirt is provided by access ports Q skirt in the bottom head insalation ar.d

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4 - 18 i.:. access ports through the support skirt.

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Tdble IAa-2500, A4xP isoller and Pressur~ c el " b-1-e c t i on X: I M4 1:dition t hrouqh the Cumner 1975 Addenda

MPE NDIX A-1 EX AVINA TION SUYY ARY ASME CLASS 1 SYSTEMS AND COMPONENTS l EX A\1!N AII(>N (13v;.( 8N EN IS ,\.N D P \ h lS

( A ll t;(m) LXA\MNATMsN FAJENT(a u m m m l') bE FA.\MbF'O y n ; g gy gym g 3 9 3.g E1.3 B-C Vessel *o-flange citetmierential Volumetric 200 The vessel-tc-flange weld id weld 5 accessible from the flange mating surface.

!!ea l- t o- f l a n ge cir cunf er en t i al Volumetric 10; well Hea.bt>-flan += weld is m +ssible -

f rota cor respendir:9 sur f aces when the heal as removed fcr refueling.

Bl.4 B-D %zzle-tc-vessel attactcent welds: Volumetric 100 The O.D. by opening doors from -

Steam nozzle-to "tssel sacrificial shield and thermal FeeJwater nozzle-to-vessel insulation.

Core siray nozzle-to-vessel Pecir. inlet nozzle-to-vessel hecir. outlet nozzle-to-vesal

(

H Con t rol ru l dr ive wite r r et u r n 1.czzle-tn-vessel E

] Jet luxnp i n s t r um,-n t nozzle-to-vessel NOf L.a c j t i hvz zle-to-closure t.ca l a t r actanant Volune t ric 100 welds -

T 1 e i I nr. o r ra3ius coctions of the Volumet r ic 10, vessel n->zzles liste l atove Access to inner ra li us '*ctions of 6 the various vessel nozzles is sub-k3 3ect to the same conditions as b stated above for nozzle-t o-ve ssel W welds.

Inner radius sections at the volumetric 100 closure hrad nozzla Closure head nozzle inner rrili -

become accessible during refueling.

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1 Table IWB-2500, As'dE Boiler and Pressure Vessel Code,Section XI, 1974 Edition through the Suruner 1975 Addenda

AI'n'f D a X A - 1 LF AT/.;f MTiON FU'. r.uty M'M CL %S 1 SYSl f YS AND CO /f Curtsis r - - - - - - - -

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] {FN[_ IN 1 3% qgg g g g-93g Bl.S B-E helds in vessel at control rod Visual 25 Access is provided by cbservation -

drive penetratiens and in-core ports in the bottcn head insula-monitor housing tion.

(St ub tube-to-housing and ve ssel)

Bl.6 B-r Frimary nuzzles to saie-end welds: Visual, 100 Reactor pressure vessel safe-end -

Eecirculation inlet surfe c, welds are made accessible by re-I<ecir culat ion outlet and moving thermal insulation and

.l e t puq- i n s t r umen ta t ion volunetric sacrificial shield plugs.

core spray CiG Ret urn y Bl.7/

I Bl.8 B-G-1 Closure st uds and nut s Volorotric

" in i 100 Closure head stud and nuts are 7 surface.

accessible during refueling Bl.9 B-G-1 Lle7 wner.t s between thr eadsd st ud Volumetric 200 Ligaments are exposed by removal -

holes of the vessel head during re f ueling Bl.10 B-G-1 Closure washers and bushings Visual 100 Washers and bushings are accessi- 8 ble upon removal of the studs.

Bl.11 B-G-2 Reactor vessel Visual 100 Accessibility is provided upon 9 removal of insulation.

Bl.12 B-!i Sk i rt -to-vessel we i Volumetric 10 Access is possible frce the vessel -

g g O.D. by openings in the sacrifi-cial shield wall and removable C.)

y gg insulation panels.

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_u. A.a AWr ci4 1 Table IWB-2500, ASME Boiler and Pressure Vessel Code,Section XI, 1974 Pdition through the Summer 1975 Addenda

APeL NDIX A-1 E XAVINAllON SUWARY ASME CL ASS 1 SYS'IEMS AND COMPONEN TS

~ ~ ~

l i t'M EN AMIN A llt )N NOI COMN)N hN'IS A'.D P A R Ib EN AMIN A I!ON E.\ TENT OF EN AMIN A llON

('.\ l E( H MI TO 11 EN AMINED M Mib)D31 I El(( ENT IN 10 YE Altsi MTESS!Hil.lTY NOIES Bl.13 fi - I - l a> di l'l ical> 1 e . Tht closure -

head is not c la. i . _

Bl.14 b-I-l Ve ' se l c l +i. li ng - o l 'a t he s , ea rh Visual 103

% v; 2 ri . in ar*>a 10 Bl.15 B-N-1 The i nt er 2 0r vu f at es of the Visual At the first re f.2eling o 2 tag.

resetor ve nel an1 th, yac i The lower J. lend!" is accessible by 11 and th i e 2-yaar inte:vals. re:nval i #

four fuel Imndles, telcw th re3ctor core .$ bus i th(

bottom h< m] centrol rod, fuel sLpNrt piece and guide tube.

, Bl.16 B-N-2 The internal sulport ant

'[ at tachment s welded to t he vessel

'acual IN* of s i sually accessil.le No additional provision is in- -

7 welds and sur f aces of com- c lu h:J in the design to providae w -i l l and the internal om! or * :.t s por.en t s or:ce at or near the 4

access to t hose nr: terr.als that end of the i ris; .ect ion are not normally accessible trom interval, the vessel flange during refuel-i r.g .

Bl.17 P-N-3 Not Applicable.

(IMF only) _

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B l .154 B-O The CbD hou;1ng weld metal and f Vo l t.me t r i c 10G% of the welds in los of 7 ba >e riet al for one wall thick- or the Iori;heral CRD Lousings, 12 n*-ss beycod the e<lge of t he 'm l d Surface }mrformed at or riear the end a of th+ inspection n r,t erva l .

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Bl.19 B-P

.M[ Welds a:.1 su;'!cr ts nat otherwise Visuil 100 a~ examined -

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Talile IWB-2500, ASME Boiler and Pressuro Vessel Code,Section XI, 19 74 Edition through the Sumer 1975 Addenda

V X 'eDlX A -1 Ni */iN ATION L U' T ARY

/ ? .if CLT 55 1 SiSi L '/S AND CO' :'ON t.NIS .

1 I i \1 f \ \Ml%\ !It A n 1 l'ON LN IS .\N D l'.\ O'IS E N .\M I N.\ l im t N I ENI OF LN \ MIN ATION NO 3 C.' ! i t ;()l: YI 10 Bt EN .\\1I N E D M L l liOUS I PF RCI N !' IN 10 Y F.\RS I DIU

.\CCL W IBll { __

g ssurizer B 2 .1 - h ).1 1 All Not applicable (PWR only) -

Heat Ex hange rs and St e.un Generators B l.1-b 3.10 All t4 > t applicable (PWR anly) -

Pipirig Pre bure Rouaday a

w I

tn E4.1 B-F Pipinq pr essure boundary saf e- Visual, 100 Safe ends in branch welds and 13 ends in 1. ranch piping welds surface dissimilar wtal weld.s are made

6. accessible by removing piping volumetric thermal in s t21at ion .

Piping 1.ressure boundary welds between dissimilar metals (See Appendix B, Fiping Weld and Surtort Exa'ilnat ion Sumrnary f or Details) a al I I C B4.2/4.3/ B-G-1 Piping pressure boundary pressure _ -

4.4 retaining bolting. Not a pI li-cable. There is no bolting 2 in.

or larger in the piping.

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1 Table IWB-2500, ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition through the Sumer 197 5 Addenda

APPilDlx A-1 1 Xf AIN A TION SUP."N ARY ASME CL ASS 1 SYSTIr/S AND COMPGNENTS I~1 E\1 F.N iMI N A'i ti N ( (iyix, ,ps f g aui pas jh E N AMIN A l l( W ENIENI (>F EN AMIN ATION NO I ('XI r id )l0 1 143 p,p i,x,3y;3 y t 3 y, i g gg yg3 1 g.1- H Cl.N T IN 10 Y E AltS I ACCESM bill f Y NO I ES 34.5/B4.6 h-J Iorqitudinal, cirecr f e:ent ial, Visual, 25 Accessibility is r roviaM by 13 B 4 . 7/ f:4.8 br <i a b , a r.d w;cket welas in t h+ Carfa r removable insulatico to expose follcwing syttems: and/or 3 t he cir curaf orencial welds and, Fesidaal " eat F om w 31 P,ain st e un Volun. t r i:

') wharever appl 1 cable, 15 in. of the longituilnal welds.

To dwa t e-r

- - ~

Core ipray o

RW Cl s e.up P W Iw. ir olar ion / r a si trol rud drn }_Q _J Hi Jh i' r e i c ur e Cucla:J In je 1i- , y c

t.

b se t <n vre I .o l a t io:. Cc 11:

. y st en C (C . Appentix h, Pipi:,g hel d .nl Susport Ex2ninati,

iry 'or

'e t is i l s ) ,

w i

.urface 25 Acce.s will be provided by re- 13 B 4 . '> B-K-1 Piping I res mre to r.dary 1.,-

te<p a ll y welded external sur '.rt or movah)e insulation to expose the attachm.nts (se e Apt m..!i x 0, 11unet.ic = t t achmer.t wcids.

Piping W,1.1 ar i Su; pr t Lxami-r.ation 3 uma ry for M t ei ! I r. )

P4.lO E-F-2 E 1; in ' pr+ aur e b( undat han rs, u .a l 100 Access of supports is available 13 sa ul .be r s , ab s.o rbe r s , a r.d ct b v f or vi sual examinati on without enpror ts wnow st o ctural tv- providing CIccial removatle qrity . relied upn to s rtio:

w i th s t .ind As ig r. }oad ir *

'elsmic in laced di rpla roner.t ';

(E App.-ndtx II , Pipir.1 and Suttort Exc ir.a t i cn Qrva r'j for datail5)

B 1.1 1 B-P h e l:12 an.1 seiperts r.ot otherwise Visual 10J -

e x cur i n t d M

e <

%. 1.4.12 B-G-2 01;4ing 1 r enure bnundtry i r . r ~. - Visual 100 Accessibility is provi N pon 14

_ urn-retaining bolting rer. oval of insulation, WA

% o ' Table IwB-2500, ArMF. Haller and Pre ss ur e vessel Code, aection XI, . 'l Edition through the Sumer 1975 A?denda

APPE NDlX A-1 EXM?INATION SUWARY ASME CLASS 1 SYSTEMS AND COMPONENTS I I E.M EX AMINXi!ON ct iM h )N LN I's .tND l' A RTS EX AMIN A ! ION EXTENT OF EX AMINATION NO 3 CNitrd)RT1 qu pg Ex 3ylu p ypggg3pst I LHULNT IN 10 T EAhSt AFC LS31b1LITY SOlh5 Pug Pressura boundan a5.1/B5.7/

d5.3 B ';- l Pap pressur e bour Jary pressure- olumetric 100 Reactor recirculation prp casing 15 re tair.ir. ; bol t i ng and bolting is accessible darirq an Surface outage.

b5.4 B-K-1 Pump ir tegrally-weldei s ar rer t surface or 25 Access will be provided by re- 13 attachmenrs D',. e Ap rend ix B , volumetric movable insulation to expose the Illing Weld and Support E w ina- attachment welds, tion Su: mary f ar details)

> B5.5 B-F-2 Purp hangers, sn uller s , rdsorbers, Visual 10 Acct.ss of supports is available 13 1 t>Ie s'21 } crt s whose struct aral for visual examination without j 1r.tegrit y is relied uren to vith- ~J

~

[ roviding special removable stand design loads and melsmic sec t ion s induced displacements (See b -;al A p; *:nd i:: B, Piping Weld and C J Surport Examination Su:razy for CQ D details) r-

[Q/ .t 3 B5.6 B-L-1 NGt app 12 cable. There are no L -

puups with pressure retaining welds in the pump casing

[-

L_

B 5. ' B - L- 2 Ir.ternal pressare boundary sur- Visuol 50 Internal surfaces are accessible 16 faces of recirculation pumis for the required visual examina-tion upon disassembly nf the pumps. Renovable insulatior. will be provided for pump disasse.mbly.

g B5.8 B-P We1Js and support s n it otherwise Visual 100 examined b

e-se p by Table Iwn-25co, Asxt notler and Pressure Vessel Code, Secticn XI, 1974 Edition through the Summer 1975 Addenda

IdW MHX A -1 E X A?.*i N A T 6GN SU:s * *,H y s

.9 E CLASS 1 SYbi f MS AND CGP. TON L NIS l'11; l.N W ! N A 'l H 4N ( i n tl ()N L?. j y ,sg } 3 ;,

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lunctric CD m' E.uj p .i Ex a.1 1 oat ic- ":gr, s f. tw I e t 1114) w c'c '

bK b-K . Valv L.m yro reib! s s; ..rb- 'Jis a1 1: co n s .cf sur cr* is a v,ii l an ] e. 13 er_, , :y mrts utm e fer visaal ex a nativr, wItno t t r i. t m il i t. t . -., t- i t / 2s r e li. d travi.iinq s p-c i a l removaole i 4.n + , w i t ' e r u.s '

w ti r s a r. 3 -

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f/ '1 * :: i l d bt.G rem-1 , i ll ble " ore c4 r o -

val n:h rr ssu ' v.. tat 1 g veld in + < .a1 1. 3 'i es te.~' h-M-2 m !a ri su~

In e r ru l tre s at e 5 V i c o.11 ore '.rilv_ of ca.M cf th+rr I:t(rnal surfaces are aores.ible I fi f a. . of salve.. cx. e 315ng . qrc u o f v s l ve < f t he Sr the required vis m examira-ru sni r.a l 1 ile . - . rw l ". steam, mes co struct13n, de i ar. , r ion uper di t ass +>r bl y vf the i.ater, ore s j r ry , hi , manu:a.: r i r.1 ma t hn i , mane-fi vilves. Fa ovablo i n s ul a t l or.

pr.-ss u r . :,olant Ia)? tion, fatturer, an* ; er f orn i r.g will be prrvidel re;iduil at rencva l ar.1 similar f unc:t tons in the r r -i r c 4' *.cn pipi" pi a t e i s'r a .8%

n M '

T -s b l e IWB-2 s ' A L M F. holler und Itessare Vi< so l Cr, , on XI, 1974 Edit on hrow;a the Sa m r 1975 Aa h .d5 b

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APPENDIX A-1 NOTES NOTE 1: Relief from Code Requirements is requested.

With the core in place, high radiation levels in the vicinity of the subject welds make it impractical to attempt examination. Therefore, these welds will be examined in accordance with the Code only if and when the entire cc e is removed for any reason. Additionally, due to limited external accessibilitytothecircumferentialandlongituggnalweldsbecauseofstructuralinterferences, the 50% examination requirement after exposure to 10 NVT of fast neutrons cannot be performed. Vessel internal examination is impractical because the necessary scanning equipment ic not yet available, positioning is extremely difficult, and accurate interpretation of test results is impossible due to the cladding.

Various inspections and tests are performed and operational constraints are strictly observed to assure the integrity of this vessel. These include:

Item Reference (a) Visuur examination durino hydrostatic test Code required hydrostatic test

> each inspection interval Y

r o

(b) Minimum pressurization temperature restric- Technical Specification Paragraph 3.6.B tions and associated surveillance program (c) Thermal transient restrictions Tec'7ical Specification Paragraph 3.6.A (d) Reactor coolant system leakage monitoring Technical Specification Paragraph 3.6.D NOTE 2: Examination from the vessel internal surface is theoretically possible, but is not presently practical as indicated in Note 1, above.

NOTE 3: Relief from Code requirements is requested.

With the core in place, high radiation levels in the vicinity of the subject welds make it impractical to attempt examination. Therefore, these welds will be examined in accordance with (g the Code only if and when the entire core is removed for any reason. Examination from the 43 vessel internal surface is theoretically possible, but is not presently practical as indicated 1 in Note 1, above.

~. r m

gnA-

& .? I NOTE 4: Relief from Code requirements is requested.

Only visual examinatior during the 10-year interval hydrostatic test will be performed because the large number of penetrations in this area makes volumetric examination impractical.

NOTE 5: During refueling ultrasonic examination can be performed frcm the flange mating surface.

NOTE 6: Relief frce Code requirements is requested.

o acceptable procedure exists for examination of the inner radii from the nozzle O.D. due to technical problems with interpretation of results. Auto.'ated equipment is not available to perform examinations from the nozzle I.D.

NOTE 7: Closure head boltirg will be exanined under tension, when bolting is removed or when the bolting connection is disas:x mbl ed. Knen bolting is examined under tension, no surface exan is required.

urface ex.inination will onli Le nerforned in conjunction with volumetric cxam when bolting is removed.

NOTE 8: Visual examination will be performed only when studs are disasscmbled.

T NOTE 9: Visual examination will be performed only when the equipment is disassembled for maintenance or 7

overhaul.

W NOTE 10: Relief from Code requirements is requested.

The requirement for this examination has becn dropped from later addenda of the Code, such as the Summer 1978 Addenda. Visual examination of the internal surfaces of the reactor vessel is covered in Examination Category B-N-1.

NOTE 11: Visual examination of the lower plenum was performed at thc first refueling outage. During subsequcnt refueling outages if this area in made accessible during normal refueling operations, the lower plenum arca shall l>e examined. These exaninations shall not be made more frequently than at approximately 3-year intervals.

NOTE 12: Relief from Code requirements is requested for the exirsination of the stub tube to CRD housing

(,0 welda.

4%

h'? The stub tube to C M housing peripheral welds are inaccessil-le due to structural interferences and

._. cannot be surface or volunetrically exanined. The CED housing to flange welds are accessible j and will be examined in accordance with Code requirements.

T

g '

.. - - ~ - '

.' -

LUTE 13: Speci fic relief requestt are identified, as applicable, in Appendix B, Piping Weld and Support Sun:u ry.

I;OTE 14: Relief from Code requirenents is requested.

Visual exanination of relief valve and RHR systen flange bolting will Le ra da dming disassembly for naintenance. Vtsaal exanination of the recirculation systen decontanination flange bolting shall be nade whenever the flanao is renoved. Exaninationa need not be conducted more frequently than every 10 years.

IiJTE ';- heactor recirculation i'unp casing bolting will be exar'ined under tension, or when roncved or disav onbled. When bolting ir examined under tension, no surface exam ia required. Furface examinatica wil! only be perforned in conjunction with volumetric exan when boltin7 is remved.

!: ~fi'E 16: Relief fron Codo req ui rene nt e 1: rejuested.

Visual inspection of the internal nurfaces will be nade only when recirculation pu:rps are sisasceable:1 :or other reasons.

>i

~ N m 17- Relief fren Code requirenents is requested.

i w

PJ Visual exanination of r ecircula tion pun.p nerhanical seal bolting shall te done when disasserbled for maintenance L:::aninat iona need not be conducted nure frequent 1/ than every 10 years.

NOTE 18: Felief from Code requirements ir requested.

Visual exanination of the internal ;urface of valves will be n.ade when valves are cususenbled for other reasons. Exaninations. need not be canducted more frequently than ever; 3 /r years.

IFJTF 19: Eclief tror Code requirements is requested.

Visual exanination nf the valve boltiny wil? be rade when the valves are disasserbled for other reasons. Exaninations need not be cordactei co e frequently th an every 10 years.

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AP<'E N DI X A -2 E X A* % % TION SU'. . ' r.R Y AS'.iE CL ASS 2 SySIEMS AND COMPONEN TS I l l\f r X A\flN ATION c(is;IONF3 TS ND P '

NO.1 E' CAItGugyI To bt M 't 's* b p'\ \ !NAllON EN IENf UF EN A\t!NA!JO N

!LlllODSi FF lt( ENT IN 10 n tR33 Acct 3s;BILITY NC,T ES C2.5 C- E- 1 Pipir.g pressure bounde.ry integrally Surface 100% of cr.e eq_ivalent Generally accessible with reroval I welded external suTI=:rt attach- stream mer.ts (See Apgendix B, raring Weld of fe rt s nent insulation and Surjort Exam inat ion S'rnn.it y fur de tail s. )

22.6 C-E-2 F1pir.g pr essur e boundar y harge rs, Visual 10C4 of one equivalent m.40ers, absorbers, and other Access is provided 1 stream tut tert s whose struct ural int e gr i t y is r elie d uIon tt withstand design la.s i s and rel m.ic induced dis-placements (See Apperctix B, Pining Weld an! Sup png t Examination S urraar y for details.)

b]-

C t _ ,

_)

I

>J

{Y e (py s_ t_ i f-A> a i C3.1 C-F, C-C Not Applicable. d There are no pressure retaining I 9 L--, -

welds in purrps [

tw C3.2 C-D Fressore retaining boltlag in pumps Visual and Visual, 100 % o f one equiva-either ienerally accessible with ree val 2 stream of permanent insulation Surface or Surface or volumetric, 25%

Volumetric of the bolted jotrts with 10% of each joint botting examined C3 J C-E-1 Nat A F plicable.

There are no integrally welded -

supports on pumps

. p. C3.4 C-E-2 Inzmp support components

-

Visual 100% of one equivalent Access is provided -

. s tream

~

=~.%. (

1) (.D 1

Tele I a'P - 2 , C D , ,a IIer and Tiessure Ves al Code,Section XI, 1974 Edition ti raugh the E n ,er 19 75 AW f di

APh NDlX A --2 FX A'."lN ATION SUMYARY AL*.'E CL ASS 2 SYST E MS AND C O'. YON E N T S l l E\1 EX A\t1N A llON N O. 3 CON!!1)NEN TS AND PA RTS fX CAT t Golt Y3 10 BE t.X A5UNEI) NATIO.N EN I ENT OF EX A\tlNATiON st E t uopsi 1 t RCENT IN 10 YEARSt 3ccrylys;tt ry NOT ES Va ves C4.1 C-F, C-C Not Applicable.

There are no pressure retaining -

welds in valves

, C4.2 C-D Pressure retaining bolting in valves Visual ar.d Visual, 100% of one equiva- Generally accebrible with removal 2 either lent stream of permanent insulation sur face or sur f ace or Volumetric, 25%

volumetric of the bolted joints with 10% of each joint boating examined 3 24.3 C-E-1 Integrally welded valve supports Sur face 100% of one equivalent

[ (See Appendix B, Piping Weld and Accessible with removal of y Support Examination Sumnsry for stream permanent insulation 1 details.)

C4.4 C-E-2 Valve support components (see Appendix B, Piping Weld and visual 100% of or.e equivalent Access is provided 1 stream Support Examination Summary for details.)

(co) ,

CJ l A.J L .J h

( v%

e LJP k

m s' .

[

y Table I W 2570, A **5 Soile end I r -ssure Vessel Code, section XI, I M4 F di t t or. tbrcugh the Syrcr 197*. A.iknda

APPE!; DIX A-2 4

fiOTES

1. Sp<'c i f i c relief requests are identified, as applicable, in Appendix B, piping Weld and Support Examination Sunmary.

2. Relief from Code requirements in requested.

The Code requirement is to either sur face or volumetrically examine bolting exceeding 1 inch in diameter. This requirement is inconsistent and more restrictive than the Clasa 1 requiremon. (exaline bulting 2 inches and <'reater) and later Section XI Addenda, such as the Summer 1978 Addenda (examine bolting exceeding 2 inches in diameter).

The in-lieu-o f examir.a tion for Class 2 bolting will be to surface or volumctrically y examine bolting exceeding 2 inches in diameter.

I tJ l

45 CO A. ,

b* a a _0.

APPENDIX B PIPING WELD AND SUPPORT EXAMINATION PROGRAM TABLE OF CONTENTS Table Number Subject Page TABLE OF CONTENTS . . . . . . . . . . . B-i B-1 Code Exemptions . . . . . . . . . . . . B-1 B-2 Relief Requests . . . . . . . . . . . . B-2 B-3 Weld Summary Sheet Terms . . . . . . . B-5 B-4 Weld Summary Sheet Index and Cross Reference to Inservice Diagrams . . B-8 B-5 Weld Summary Sheets and Notes . . . . . B-9 B-i 34;; g 7

_n Table B-1 Code Exemptions Code Exemption Identifi er El Class 2 component connections, piping and associated valves, vessels (and their supports) in systems where both the design pressure and temperature are equal to or less than 275 psig

= and 200 F, respectively, are exempted from volumetric and surface examinations.

Reference:

Para IWC-1220 (a).

E2 Class 1 component connections, piping and associated valves, vessels (and their supports) , that are three-inch nominal pipe size and smaller are exempted from volumetric and surface examinations. (Sufficient normal makeup capacity using onsite pcwer exists for greater than a three-inch break.)

Reference:

Para IWB-1220 (b) (1) .

E3 Class 2 component connections, piping and associated valves,

~

and vensels (and their supports) , that are four-inch nominal pipe size and smaller are exempted from volumetric and surface examination.

,=

Reference:

Para IWC-1220 (d).

~~

E4 Open-ended portions of Class 2 lines in a non-closed system (e.g., suction line from a storage tank, or discharge line of a containment spray header) extending to the first shutoff valve are exempted from hydrostatic and pressure test require-ments.

Para IWC-5220 (d).

Reference:

E5 Class 2 circumf erential welds that are not structural dis-continuities nor within 3 pipe diameters of the centerline of rigid pipe anchors, or anchors at the penetration of the primary reactor containment, or at rigidly anchored components are exempted frca volumetric and surface examination.

Reference:

Table IWC-2520, Examination Categories C-F and C-G.

NOTE: References are to the indicated portion of ASME Boiler and Pressure

] Vessel Code,Section XI, 1974 Edition through the Summer 1975 77 Addenda.

+

as yn)' ?s u bs B-1 2O 3

Table B-2 Requests for Relief

=

Relief Request Code Required In Lieu of Identifier Relief Request Exam Exam R1 Class 2 component connc ctions, Volumetric Surface piping and associated valves, (if nominal and vessels, that are 0.5 pipe size inches nominal wall thickness exceeds 4 in.)

Z or less.

Basis for Relief Volumetric examination of thin walled pipe does not produce reliable results. In recogni-tion of this, later Addenda of Section XI (e .g . , 1977 Edition, Summer 1978 Addenda) establish the volumetric examination cut-

, off point as 0.5 inch for Class 2 systems.

R2 Class 2 components of systems Volumetric System or portions of systems that are (if nominal pressure not required to operate above pipe size test every

- a pressure oi 275 psig or exceeds 4 in.) 3 1/3 yrs.

temperature of 200 F except for limited periods which are far

~

less than one peretnt of normal plant operating time.

Basis for Relief This relief request encompasses the intent of Exerrption El, as provided by IWC-12 2 0 (a ) . Opera-ting conditions are specified in lieu of design conditions since

~

the latter are often overly con-servative. As a result, no additional burden is accrued in the form of added examinations which would normally be excluded had the designer not given extra consideration to conservatism.

"8 .

2 c> <(

Relief Request Code Required In Lieu of Identifier Re1ief Request Exam Exam

- R2 Situations where the stated (Cont.) pressure and temperature condi-tions are exceeded by a modest amount will be addressed on a case-by-case basis if the period where the threshholds are breached is of very small duration.

Further amplification will be provided in such cases.

R3 Component welds whose physical Volumetric Surface or location within the plant Visual, as

~

restricts access to the weld appropriate due to such factors as: b3ing located within a wall sleeve

= or penetration, in a high radiation area, very high in a room, adjacent to a wall or Surface Visual during other restriction without hydr; static sufficient clearance to perform testing once examinations. every 10 yrs.

Basis for Relief Specific reasons for requesting exemption from an examinc. tion requirement will he provided in each instance.

R4 Branch pipe to pipe welded Volumetric Surface joints that are Class 1 and greater than 6 inches in

- diameter or are Class 2.

Basis "'c Relief The physical design of branch connectior. does not permit meaningful volumetric examina-tion. This fact hau been recognized by isSME Section XI and the requirement for volu-metric examination of branch

= connections has been dropped from later Tiddenda of the Code (e.g., 1977 Edition, Summer 1978 Addenda),

B-3 S$M M ,

Relief Request Code Required In Lieu of Identifier Relief Request Exam Exam R5 Integrally welded supports Volumetric Surface

.n Class 1 systems.

Basis for Relief The physical design of integrally l welded supports (fillet or partial penetration welds) does not permit meaningful

_- volumetric examination. This fact has been recognized by ASME Section XI and the require-ment for only volumetric examina-tion of integrally welded supports has been dropped from later

-- Addenda of the Code (e.g., 1977 Edition, Summer 1978 Addenda).

R6 Class 1 component connections, Volumetric Surface piping and associated valves, (if nominal and vessels that are 0.375 pipe size inches nominal wall thickness exceeds 3 in.)

]; or less.

Basis for kelief Voluntetric examination of thin walled pipe does not produce reliable results based on the

__ material gecretry and on the cperator's ability, experience and other "numan factors"

_ intangibles that make the examination results nonreproduci-ble. In order to provide a greater factor of safety for Class 1 systems compared to Class 2 systema, the cutoff point for volumetric examina-tion is established as 0.375 inch for Class 1 systenc.

?

Id v (+ w.pr ...$

~~

[(,

Table B-3 Weld Summary Sheet Terms The following information is presented in the Weld Summary Sheets:

Item Description

_ " System" Identifies the system name.

"ISD No." Identifies the Inservice Diagram (ISD) number which corresponds to the subject system.

"ISD Rev." Identifies the revision of the ISD used to prepare

} the program.

"Line Number" Identifies a discrete portion of the system with piping having the same function, pressure / tempera-ture ratings, size, and pipe schedule. This number is the identifying number on the ISD's.

"Line Description" Describes the beginning and end of that particular line number.

" System Class" Identifies the ASME Class for the subject line.

Possible entries are:

1 - ASME Class 1 2 - ASME Class 2 3 - ASME Class 3

-- 2(A) - Augmented Class 2 (These systems are not within the scope of ASME B&PV Code

]-[ Section XI, but do receive inspection because of their function as containment pe ne tra tions . )

"ISD Coordinate" Identifies the coordinates on the ISD where the sun]ect line is located. Where only one ISD is associated with the system only one coordinate is given; e.g., C-8. Where more than one ISD is

- associated with the system, the unique part of the ISD number is given together with the coordinate;

e.g., 20B/D 1.

=J

" Weld Type" Identifies each type of weld found in the subject line. Possible entries are CIRCUM - Circumferential butt welds BRANCH - Branch welds in ASME Class 2 systems

_7 B-5 q,

m , w, g+

r m.

=

Table B-3 (continued)

Item Description

" Weld Type" BRANCH - Branch welds in ASME Class 1

-~

(continued) DIA _ 6 IN systems with the branch less

___ than or equal to 6 ir.ches nominal pipe size.

BRANCH - Branch welds in ASME Class 1 DIA > 6 IN v/ stems with the branch greater than 6 inches nominal size

- - - SUPPORT - The weld between a presst.re retaining component and its support LONGITUDINAL - Longitudinal fabrication welds CIRCUM - Circumferential butt welds that 7 BIMCTALLIC are also birtetallic.

"No. of Weld" Identifies the number of welds of that particular weld type for the subject line number.

" Required Code Exam" Identifies the examination that would be required by the Code (i.e., 1974 Edition of Section XI through the Summer 1975 Addenda) taking into account the permissible code exemptions, but not the Requests for Relief from Code requirements. The possible entries are VOL - Volumetric examination SUR - Surface examination HYD - Hydrostatic pressure testing

'.CN C - Applies to augmented systems that have no Code required examinations

_._ " Code Exemption" Identifies the applicable Code Exemption Identifier from Table 3-2 o f the inservice Inspection Program dc,cription report. Possible entries are Ei, E2, E 3, E4, and E5.

"Pequest for Relief" Identifier those welds for which a request for relief from Code requirements is made i

..o. of Welds" Identifies the number of welds of that particular weld type for which relief from Code requirements is re:;ue s t ed .

GM*

Jf.G n' 497*)

u!

B-6 2c%

B Table B-3 (continued)

Item Description

" Reason" Identifies the generic Relief Request Identifier from Table 3-3 of the Inservice Inspection 1 rogram description report. Possible entries are:

R1, R2, R3, R4, R5, and F6 "In Lieu of Identifies the examination to be performed in lieu B Examination: of the Code required examination for those welds for which relief is requested. Possible entries are:

SUR - Surface examination HYD - " Hydrostatic" pressure testing B FRS -

" Pressure" testing FLW - " Flow" testing NONE - No examination or testing is to be performed.

" Notes" Identifies the note number which presents supple-I 4Ptal or explanatory information for the Request for Relief or examinations performed in excess of Ccde requirements.

B l

l B

1 E

1

.E.

I 1

E

.L : i l

t

Table B-4 Weld Summary Sheet Index

_ and

=

Cross Reference to Inservice Diagrams E

ra -

Inservice Diagram Weld Summary System Number Sheet Pages

_=-

Drywell Inerting and Purge ll825-FM-18A, Rev. 18 1-3 Residual Heat Removal ll825-FM-20A, Rev. 16 4-13 11825-FM-20B, Rev. 15

~

ll825-FM-20C, Rev. 14 Il825-FM-20D, Rev. 14 Reactor Core Isolation Cooling ll825-F"-22A, Rev. 13 14-15 M

f Core Spray ll825-FM-23A, Rev. 13 16-19 Reactor Water Cleanup ll825-FM-24A, Rev. 12 20-21 liigh Pressure Coolant Injection ll825-FM-25A, Rev. 12 22-24 11825-FM-25B, Rev. 11 Reactor Water Recirculation ll825-FM-26B, Rev. 9 25-28

_ Control Rod Drive Return ll825-FM-27A, Rev. 8 29-30 Main Steam 11825-FM-29A, Fev. 13 31-33

" ll825-FM-34A, Rev. 15A Feedwater 34-35 E

zlo

, =s

m- -ni -

Page 1 of 35 System DR f.'E L L INERTING t. PURCE IYepared By M. J. PARTRIDGE 11-21-78 WELD

SUMMARY

SHEET

!!E25-FM-16A 18 Approved by H. H. ASKWITH, CHIEF ENGINEER 12-4-78 ISD No. nev.- James A. Fit 2 Patrick Nuclear Power Plant o.w Requer,t for Rehef Systein ISD No.of C(wle Cale No. of In Lmu of Line Number Luw D.wnption Class Coordinate Welil Ty pe hids Exam Liemptuen Wel.ls Renwn Examulation S utes 20"-N-152A-20 F ROM V ALVE 21- A0V-I I / TO 2(A) C-8 CIRCUM 12 NONE 10 -

SUR 1 SUPPRES$10N CHA"EE R BRANCH 4 NCNE - - -

SurPORT NONE -

SUR I FENETRATlUN X-205 5 5 24"-N-l52A-8 FROM VALVE 2 7- A0V- 112 TO 2(Ai H-7 CIRCUM 4 NONE 4 -

SUR I LINE 14'-N-152A-9 AND ERANCH 2 NONE -

LINE 18"-N-152A-6 FROM LINE 24"-N-152A-8 TO 2(A) E-8 CIRCUM NONE 11 -

SUR 1 14'-N-152A-9 13 PENETRATION x-/l SUPPCRT 2 NCNE 2 -

SUR 1 18'- N- 152 A-6 FROM L INE 24"-N-152A-8 TO 2(A) E-8 CIRCUM 4 NCNE 3 SUR I 5UPPORT 2 NCNE 2 -

SUR I to PENETRATION X 25 E

24"-N-152A-ILA F ROM VALVE 27- A0V-Il 3 TO 2(A) B-6 CIRCUM 9 NONE 7 SUR I LINE 18"- N- 152 A- 2 6 A BRANCH 4 NCNE l -

%R I SUPPORT 2 NONE 2 -

Et" I 18"-N-152A-268 F ROM 24"-N- 152A- 18A TO 2(A) C-6 CIRCUM 3 NONE 3 SUR I PENETPATICN K-26B SUPPORT 1 NONE 1 -

SUR I 18'-N-152-26A FROM 24"-N-152A-18A TO 2 ( A) C-6 CIRCUM 3 NONE 3 SUR 1 PENETRATION X-26A SUPPORT I NONE I -

SUR I 20"-N-152A-7 FROM VALVE 27-ACV-Il6 TO 2(A} C-8 CIRCUM 5 NONE 4 -

SUR )

PENETRATION X-220 BDANCH 2 NONE - -

SUPPORT 2 NONE 2 -

5UR I 73 sy h4 A o ,

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_ P I R P I R P l R P I R P I R E -

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Yc eb

JAMES A. FITZPATRICK Page 3 of 35 NOTES System: D rywe l l inerting and Purge

1. Containment inerting and purge lines are not within the scope of ASME Boiler & Pressure Vessel Code,Section XI. These lines are designated as Class 2 - Augmented, and all structural discontir.uity c i rcumfe ren t i a l , s uppo rt and branch (greater than 4 inch) welds will be surface examined. The pipe wall thickness is 0.375 inches.

2. Vacuum breaker lines are not within the scope of ASME Boiler & Pressure Vessel Code,Section XI, These lines are designated as Class 2 - Augmented, and all structural discontinuity circumferential, support, and branch (greater than 4 inch) welds will be surface examined. The pipe wall thickness is 0.375 inches.

?

C.0 o

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i--.---- -i-i.

Page 4 of 35 System RESIDJAL HE AT REMOVAL hepare<! By W. J. NEWELL 9-7-7S_

WELD

SUMMARY

SHEET ""

ISD No. Il825-FM-20A, Approved By H. H. A$rWITH, CHICP ENGINEE9 12-4-78 nev. 14-- James A. FitiPatrick Nuclear Power Plant o...

B, C, D Request for Rehef System 15D No..f Comte Code No of In Lieu ef Eane Numtier Line Ihstriptmn Claw Coordu. ate held T3 pe weld 3 Exam F u em ption welds Rea>on Exammatmn %@s 16"-W2 0- 302- 78 FROM RHR PUMP P-38 To L INE 2 0-5/200 CIRCUM 10 VOL 10 R2 PRS 1 20"-W20-302-8B BRANC H 4 VOL 4 R2 FRS 1 SUPPORT I SUR 1 R2 PPS 1 6"-W20-302-7D FROM RHR PUMP P-3D TO LINE 2 D-3/200 CIRCUM 9 VOL 9 P2 PRS I 2 0 '-W2 0- 302- 8 B BRANC H 4 VOL 4 R2 PRS I SUPPORT 1 SUR 1 R2 FRS I 20"-W20-302-88 FROM LINE 16'-W20- 302- 76 T O 2 L-5/20D CIRCUM 6 VOL 6 R2 PRS I LINE 24'-W20-302-IlB t RA NCH 3 VOL 8 R2 PRS 1 SUFPORT 3 SUR 3 P2 PRS 1 y 16"-WS-302-568 FROM VALVE RHR-177B TO LINE 2 J-l/200 CIRCUM l VOL 1 P2 PRS 1 C 24"-W20-302-IIB 16"-W20- 302-9 8 FROM LINE 20"-W20-302-8B TO 2 F-5/20D CIRCUM 2 VOL 2 R2 PRS 1 VALVE 10-MOV-65B BRANC H 2 VOL 2 R2 PRS 1 FROM RHR HEAT EXCHANGER E-2B G-5/20D CIRCUM 14 VOL 14 R2 PRS I 16"-W20-302-lCB 2 TO L I NE 20-W20-302-88 Cl# CUM 2 HYD ES - - -

B RANC H 4 VOL 4 R2 PRS 1 SUPPORT 3 SUR 3 R2 PRS I 2 F-5/20C CIRCUM 9 VOL 9 R2 PRS 1 16"-W20-302-7A FROM RHR PUMP 3A TO LINE BRANCH 4 VOL 4 R2 PRS 1 20"-W20-302-8A SUPPORT 1 SUR I R2 PRS I 2 F-3/20C CIRCUM 10 VOL 10 R2 FRS 1 ew 16"-W20-302-7C FROM RHR PUMP 3C TO LINE 20"-W20-302-8A BRANCH k VOL 4 P2 PRS 1 SUPPORT I SUR I R2 PRS I

>:$- N P

8 8 *

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Face 6 of 35 Sptem RESIDUAL HEAT PEMOVtl itepared By W. J. NEWELL 9-7-79 WELD

SUMMARY

SHEET **

ISD No. I I P,2 5- F M -2 0 A Rev. 14 Approved I!y H. H. ASKWITH, CHIEF EN:;1NEER 12-4-78 B, C,D ane A. FitiPatnck Nuclear Power Plant u. .

Request for Rebef System ISD No of C4 , = CWe No. of In Lieu of Line Number Lu;e Dncrmtion Class Coord u we Weld Type h eli t. Ex n Exemption Welds Remon Eumtnation N otes 24"-W20-302-118 FROM LINE 20"-W20-302-8B TO 2 l-2/20D CIRCUM 13 VOL 13 R2 FRS 1 LINE 24"-W20-302-14A CIRCUM 2 HYD E5 - - - -

SUPPORT I SUR 1 R2 PRS 1 bPANCH 6 VOL 6 R2 PRS I 16"-W20-302-15A FROM LINE 24"-W20-302-IIA TO 2 G-3/20A SUPPORT I SUR I R2 PRS 1 F I RST S E I SMI C St.,PPORT 16"-w20-302-15B F ROM L INE 2:e"-W2 0- 302-l l B T O 2 E-3/20B SUPPCRT I SUR l R2 PRS 1 FIRST SEISMIC SUPPORT 16"-W2 0- 3 02- 15A FROM LINE 24"-W20-302-llA TO 2 0-3/20A CIRCUM i HYD E5 - - - -

g 12"-W20-302-13A CIRCUM S V0L 5 R2 PRS 1 b

u 8 PANC H 2 VOL 2 R2 PRS 1 SUPPORT 4 SUR 4 R2 PRS 1 12"-W20-302-13A FROM LINE 16"-W20-302-15A TO 2 0-3/20A CIRCUM 6 HYD E5 LINE 10'-W20-302-12A CIRCUM 16 VOL 16 R2 PRS I BRANCH 4 VOL 4 R2 PRS 1 SUPPORT 1 SUR I R2 PRS 1 10"-W20-302-12A FROM LINE 12"-W20-302-ITA TO 2 G-3/20A CIRCUM 28 VOL 22 R2 AIR FLOW 1,2 CONTAINMENT SFRAY HEADER 6 R2 SUR 1,3 BRANCH 2 VOL 2 R2 ?RS i SUPPORT 4 SUR I R3 SUR 4 ONE SIDE 16"-W20-302-158 FROM LINE 24"-W20-302-ilB TO 2 E-3/208 CIRCUM 1 HYD E5 - - - -

12"-W20-302-13B CIRCUM 8 VOL 8 R2 PRS 1 Y BPANCH 2 VOL 2 R2 PRS 1 4 $UPPORT 8 SUR 8 R2 PRS I

W W W W MM M M m W W W Page 7 of 35 INSERVICE INSPECTION PROGR AM q Sy stem RESIOUAL HE AT REM /AL Prepared By W. J. NEWHL 9-7-73 gg ISD No. Il825-FM-2CA flev, 14 WELD

SUMMARY

SHEET

.\; proved By H. H. ASKWITH, CMIEF ENGINEER Na 12-4-78 B, C,D Jarnes A. FitzPatrick Nuclear Power Plant o,. .

Regant for Rehef Sptem ISD No of Cole Code No. of In Lieu of Lme Number Lme Dmription Clau Cm>nlinate hid Tyn. LIJ s Exam Exemption bids I< cason Exammation Notes 12"-W20-302-139 FR3M LINE 16"-W20-302-159 TO 2 E-3/2CS CIRCUM 4 HYO ES - - - -

LINE 10- W2 0- 3 02- 12 A CIRCUM 15 VOL 15 R2 PRS 1 ERANCH 4 VOL 4 R2 PRS I SUPFORT 7 SUR 7 R2 PRS 1 10"-W20-302-129 FROM LINE 12"-W20-302-138 TO 2 D-3/2C8 CIRCLM 31 VCL 6 R2 SUR 1,3 CCNTAIN*ENT SPRAY HEADER 25 R2 AIR FLOW I,2 EPANCH 2 VOL 2 R2 PRS I SLFFORT 4 5UR I R3 SUR 4 CNE SICE S"-W20-152-39 FROM VALVE FPC-34 TO LINE 2 C-2/2CD CIRCUM 23 HY3 [l g 20"-W20-152-2B BRANCH 3 HfD El - - -

O SUFPORT 7 HYD El - - - -

L't 2 4"- W2 0- 15 2- 3 B FROM PENETRATICN X-2258 TO 2 D-8/2CB CIRCLM 2 HYD E5 - - - -

LINE 20"-W20-152-6B CIRCUM 8 VOL 6 R2 PRS I E4 2 R2 SUR 1.10 PRANCH 3 VOL 3 R2 PRS I SUPPORT 2 SUR I R2 PRS 1 20"-W20-152-6B FROM L!sF 24"-W20-152-3B TO 2 B-4/200 CIRCUM 9 VOL 9 P2 PRS 1 RHR PUMP 38 2 RANCH I VCL 1 R2 PRS I 20"-W20-152-6D FROM LINE 24"-W20-152-3B TO 2 A-3/200 CIRCLM 9 VOL 9 P2 PRS I RMR PUMP 3D BRANCH I VOL I R2 PRS I 20"-W20-152-2B FROM LINE 20"-W20-152-2C TO 2 H-2/2CC CIRCUM 3 HYD ES - - - -

C LINE 20"-W20-152-69 CIRCUM 9 Vel 9 R2 PRS 1 A4 BRANCH 5 VOL 5 R2 PRS 1

'.' SLPPCRT 4 SUR 4 R2 PRS 1 dh

-w I

Para B of 35 INSERVICE INSPECTION PROGRAM W. J. NEVELL 9-7-78 System o,E5100AL HEAT PEMCwAL

. PreparM Hy t.

WELD

SUMMARY

SHEET ISD No. Il825-FM-2GA }(cy, 14 .QprmM Uy H. H. A5kWITH, CHIEF EN;lNEER 12-4-78 B, C, D James A. FitiPatrick Nuclear Power f' v.: ,,,,

l g Reque st for R*!wf Systtra 15D No of Cole Code No. of In I wu of I.ine N um ber 1.ir - I h-scnption Cla.* ('o, rd ina t.- Lid Typ.- wekts Exam E x em ption, k ehts ! tea >on Examn.atein N < .te s 20"-W20-152-2D FROM L I NE 20"-W20- 152-2 3 TC 2 C-3/200 CIRCUM 10 0L 10 R2 PRS 1 L I NE 2 0"-W2 0- 152-60 BRANCH 3 VOL 3 D? PRS 1 S U P F'O R T I SUR 1 R2 PP5 1 21e"- W2 0- I L 2 - 3A FROM PthETRATION x-225A 10 '

H-7/20A CIRCL'M 9 VCL 7 R2 PRS 1 LINE 20"-W20-152-6A E4 2 R2 SUR I.10 BFANCH 3 bOL 3 R2 PRS 1 SLFFORT 2 SUR I R2 PR5 1 20"-W20-152-6A FROM LINE 24"-W20-152-3A TO 2 1-5-20C CIRCUM 9 VOL 9 R2 PRS I RHR PUMP 3A BRANCH 3 VOL 3 R2 PRS 1 20"-W20-152-60 FROM LINE 24"-W20-152-3C TO 2 l-3/20C ClkCUM 9 VOL 9 R2 PR5 I e RHR PUMP 3C BRANCH 2 VCL 2 R2 PRS I a

w 20"-W20-152-2A FPDM LINE 20'-W20-152-2C TC 2 G-3/200 CIRCUM 8 VOL 8 F2 PRS 1 LINE 20'-W20-152-6A BR ANC H 3 VCL 3 R2 PRS I SUPPORT 1 SUR 1 R2 PRS 1 F ROM LINE NE AR M0/-10-MOV-17 TO 2 A-8/2CB CIRCUM 3 HfD E5 - - - -

2 0"- W2 0- 1.5 2 - 2 C LINE 20"-W20-152-6C CIRCUM 8 VOL 8 P2 PRS I ERANCH 7 VOL 7 R2 PRS 1 SUPPCRT 7 SUR 7 P2 PRS 1 2 0"- W20- 152 - 2 8 FPCM LINE 20"-W20-152-2C 10 NEAR 2 H-2/2CC CIRCUM 3 HYD E5 - - - -

LINE 20'-W20-152-68 BRANCH I VOL 1 R2 PRS 1 SUPPORT 3 SUR 3 Ri PRS 1 C

4 a s

%e 3%

4

@%N

~

Page C of 5:

ES ; R -EAT M":.A INSERVICE INSPECTION PROGRAM ;'-

q Spiem Pn pan d by =. J. NE-ELL gg isD No n ; F"-:,A Iter. lu WELD

SUMMARY

SHEET

.gr.n W Hy *

- ". ASK.iT . C*iEF EN0:NE:n .:

e'

u. C. ;

James A. FitzPat<<k Nuclear Power Plant - . .

fleg mt for It + '.# f Ss stem 13D No <f Cuje Code No of In La u o' {

Line N umber Lir.e Ik s< rt; in n ria m E m .rd e. ate bi1 T> pe Mh Exam E m en;ption wh heamn 1.un .c. n jN>

20"-W20-3;2-1 FROM VALVE E"P-ES TO I A-5/2:S Ci&CU" 9 Vt ! 83 Hv3 ,

.% L : 10 * ~A - 17 g a r.N ;- 2 LE CIA % 6 IN f i SuPPCT : .0L ; 05 5.- -

D alliam 4'-420-902-36 FROM WAttE 10-F0V-32 70 2 r; C-1/ CB CIRCJ" -YO 6 5.R 7 E3 -

VALVE 10-M0J-33 BRANCH HvD E3 - - - -

Su sPORT 2 WD E3 2 - 5,4 -

2 4- W2 0- 90 2 - 14 A FPCM VALJE PeR-BIA TO VALVE I F-L/2LA CIDCUM 13 VCL 1 P3 MYD

~

10 M04-25A Cl*Cv" 3 '. L L - - - -

BirETALLIC SvP l I

SLPPORT 6 V:t 5 55 S L *. -

I 5FANCH 2 SLR ,

3 DIA ( 6 IN '

I.

24'-W20-902-143 FROM VALVE RHR-8bB TC VAL'vE I E-3/2:9 CIACU" IL v:L I 53 -YO 10 MOV-258 CIRCJM i VCL /, - - - -

bt"ETALLIC 5.R I SLDPORT 6 VOL I

5 85 5.a -

SRANCH 3 5US ! - - - -

Ola 4 6 IN I

16"-W20-302-9A FPCM L I NE 2 0' -42 0- 3 02-8 A 10 2 E-5/20C Cl*CL" **O E5 - -

LINE 2C'-5LP-302-33A CIRCLF 12 V:L 12 8' 5.E -

5.PPOST 3 SUR - - - -

ERANCH 2 VOL 2 R4 5 ;R -

1 20"-W20-1504-42 FROM RECIRC LINE 32-2E-W8-GE-1B I B-5/20S CIRCU" " LCL ,

TO VALVE PhD-53 S RA'.: '

5.R - - -

(f e, m s s"n l

.gA ;iq:g. i g( g . . . .

(ss 5'"I'LL10 d.4 . ,

.y

'* I 31 i !

i

! I I

Page 10 of 35 INSERVICE INSPECTION PROGRAM Prepar.xf By W. J. NEwELL 9-7-73 g System RESIDUAL HEAT PEMCJAL La WELD

SUMMARY

SHEET gg Isp No_ ll825-FM-20A }tey, 14 Approm! Hy H. H. AS WITH, CHIEF ENGINEER 12-4 73 James A. FitzPatrw.k Nuclear Power Plant to B, C,D Reuuest for Rehef Itequired Sy stem ISD No of Caie C'"I" N" "I In b'#" 'I Cmrd .nate Welil Type Welati Exam Exemption W el.ls Rea.wn Exammatwn Notes Line Numtwr Line thwription th 4'-5LP-302-86A FROM LINE 20'-5LP-302-33A '10 2 C-4/2CC CIRCUM 1 HYD E3 - -

VALVE 10-SV-74A FRCM 20" x 6" REDUCER TO RHR 2 C-3/20C CIRCUM 3 VOL -

2 0"- 5 L P- 302- 3 3 A PRANCH I VOL i R4 SUR -

HE AT E XCHANGE R C-2 A 8"-5HP-902-32A FROM LINE 23-10' 5HP-9C2-34 TO 2 1-3/200 CIFCUM l HYD E5 1 CIRCUM 22 VOL 20 R1 SUR -

VALVE ID-PCV-63A SUFFORT 3 SUR ERANCH 2 VOL 2 R4 5UR -

FRCM VAtVE 10-PCV-69A TO C-3/200 CIRCUM 4 VOL 4 RI SUR -

6'-5HP-302-37A 2 w 20" x 8" RED'JCER BRANCH I VOL ] R :e SUR -

E x

C-4/20C CIRCUM 17 VCL 17 R1 SUR -

6"-5LP-302-34A FROM VALVE 10-SV-74A TO 2 BRANCH I VOL 1 Pk SUR -

LINE 16"-W20-152-5A SUPPORT 2 SUR - -

CIRCUM 3 NONE E4. E5 - - - -

20"-W20-152-2C FROM NE AR L I NE 20"-W20- 152-6C 2 A-8/208 CIRCUM l HYD E5 - - - -

TO 10-M0'/- 17 C I RC UM 10 VOL 9 R2 PRS I BRANCH k VOL 4 R2 PR5 1 LUPPORT 2 SUR 2 R2 PRS 1 l l

4"-W20-902-43 FROM RX VESSEL TC VALVE i B-2/20B CIRCUM 23 VOL - - - -

f 10-MOV-32 B RANC H I SUR -

1 DIA < 6 IN SUPPCRT 5 VOL 5 R5 $UR -

)

C

-;;4

%e i t

NWJ i b ID i G

'h Page i1 of 35 INSERVICE INSPECTION PROGRAM g System RE SIDUAL HE AT REM 0k AL 'repared By M. J. PAR'RIDOC 11-15-78 r=

Gg Il825-FM-20A WE LD

SUMMARY

SHEET ISD No. Rev. 14 Approved By H. H. A5FJ lTH , CHIEF tNGINEER 12-4-75 B, C, O James A. FitzPatnck Nuclear Power Plant i.

Request for Relief Sy stem 15D No. of Code Caic No. of in 1.tcu of lane Number 1,ine Description Chss Coord ma te hid Type Ltds Exam E x em pt wn hids Reason E x am tr.atwn No tes 16"- W2 0- 302 - 9 S FROM LINE 20"-5LP-302-338 TO 2 G- 3/2 CD CIRCUM 12 VOL 12 RI SUR -

j VALVE 10-MOV-65B ClkCUM 2 NYD - - - -

BRANCH 2 VOL 2 R !+ SUR -

h SUPPORT 3 SUR - - -

f i

20"-SLP-302-338 ASSEMBLY ON T6D OF RHR HEAT 2 0-3/20D CIRCUM 3 VOL - - - -

EXCHANGER E-26 BRANCH 3 VOL 3 Rh SUR -

f I

8"-5HP-902-328 FROM LINE 2 3- 10"-5HP-902- 34 TO 2 1-3/20D CIRCUM 27 VOL 27 Rt SUR -

l VALVE PCV-638 CIRCUN 2 HYD ES - - - -

I f

BR ANCH I VOL R4 $UR -

SUPPORT 3 SUR - - - -

to L

Pl 8"-SLP-302-378 FROM VALVE PCV-69B TO 2 H-3/20D CIRCUM 4 VOL 4 SUR -

LINE 20"-SLP-302-338 BRANCH I VOL 1 R4 SUR -

4"-SLP-302-E6B FROM LINE 20"-SLP-302-338 TO 2 G-4/20D CIRCUM i HYD E3 - - - -

VALVE SV-74B 6"- 5 L P- 3 0,2 - 34 8 FROM VALVE SV-74B TO 2 G-4/20D CIRCUM 18 VOL 18 Rt SUR -

16"-W20-152-58 CIRCUM 2 NONE E4. ES - - - -

BRANCH 2 VOL 2 R4 $UR -

SUTPORT 4 SUR - - - -

16"- W2 0- 302 - 15 B FROM VALVE 10-MOV-34B TO 2 E-3/20B CIRCUM 3 HYD El - - -

8 VALVE 10-MOV-34B BRANCH I HYD El - - -

8

$UPPORT 2 HYD El - - -

8 C

p.

e 3tt I

Page 12 of 35

' M. J. FARTRIOCE 11-21-78 g system "lSIDUAL H E A T R E **T. A L P.epared By t~

gp WELD

SUMMARY

SHEET ISD No. II825-FM-20A Itev. Ik .spprowd Hy H. H. A$ rw l T H , CHIE F ENGINEER 12-4-73 B, C. O James A. FitzPatock Nuclear Power Plant w.,

R yucst for Rehef Sptem ISD No. of Code Co.te No.nf In Lieu of 12nc Num!we Lme lhwription th Coorennate Weld Tne hits Exam F u emption H eids Reason Exammatwn Net + s FROM LINE '6"-W20- 302- 15e T O 2 C-3/2CB CIRCUM 3 HYD E1 - - -

8 6"- W2 0- 3 02- 16 B VA'.VE l0-M04-3ES ERANCH I HVD El - - -

8 6'-W20-152-4kB FROM VAlkE 10-M0 /- 3 CB T O G-6/205 CIDCOM 10 FLW El, E4 - - -

8 SUPPRE SS I ON C M*BE R ERANCH I FLW El. E4 - - -

8 F E N E T RA T I ON X-2118 SUFPCRT I FLW Et, E4 1 -

SLR 8,9 16"- W2 0- 151 - 5 5 FROM LINi 16"- W2 0- 30 2 - 15 8 TO 2 v6/208 CIRCUM 10 FLW Et, Ek - - -

8 SUPPRESSION CHAMBER SRANCH 7 FLW El, E4 - - -

6 FENETRATICN X-2108 StrPORT 2 FLW El, E4 1 SU9 8.9 16"-W20-302-15A FROM VALVE 10- MC V- 3 ; A TO 2 G-3/20A CIRCUM 5 HY3 El - - -

8 7 VALVE 10- MOV- 3 4 A ERENCH 3 HYD El - - -

8 o SUTFORT 4 HYD El - - -

8 6"- W2 0- 302 - 16 A FROM LINE 16"- W20- 302- 15 A TO 2 E-4/2CA CIRCUM 3 HYD El - - -

8 VALVE 10- MOV- 3 SA E-5/2]A CIRCUM 11 FLW El, E4 - - -

8 6" i20- 152-44 A FROM VALVE 10-MOV-33A TO 2 SLPPRESSION CMANSER SRANCH I FLW El, E4 - - -

8 PENETRATICN X-211A SUFPORT I FLW El, E4 1 -

SUR S.9 FROM LINE 16"-W20-302-15A TO 2 E-6/2CA CIRCUM 7 FLW El, E4 - - -

3 16"- W 2 0- 151 - 5 A SUPPRESSICN CHAMSER BRANCH 8 FLW El, E4 - - -

8 C PENETRATION X-210A SUPPORf 3 FLW El, E4 1 -

SUR 8.9 4

h*

N W*

MM I

JAMES A. FIT 2 PATRICK Page 13 of 35 NOTES System: Residual Heat Removal

1. Design temperature is 280 F, design pressure is 325 psig; however, in this portion of the system, the system operating pressure exceeds 275 psig for only approximately five (5) minutes ir the low pressure coolant injection mode of operation with unassisted low pressure coolant injection fo r med i um s i ze (0.3 ft2) breaks.

Accordingly, this portion of the RHR system is considered to be a moderate energy sys tem since it does not normally operate at this elevated pressure during normal operation, cold shutdown, or refueling modes of operation. This portion of the system is normally maintained at a low nominal pressure and a mechanism for piping failure is unlikely. This portion of the system is constructed of carbon steel and, therefore, is not s ubj ec t to intergranular stress corrosion cracking and is not subjected to cyclic stresses or high stresses that would result in a piping failure. This portion of'the system during the residual heat re mova l mode of operation also exceeds 200 F for only approximately four (4) hours at the beginning of operation in t h i s code . This elevated temperature operation represents less than one percent of system operating time and, therefore, also quali fies as a noderate energy system. This portion of the system will be " Pressure" tested every 3 1/3 years to confirm piping system integrity.

2. " Pressure' test or " flow" test not feasibie for containment spray line from inside drywell penetration to spray ring. An " airflow" test is conducted periodically.

7 33 Surface exam will be performed as augmented inspection of 6 circumferential welds near Penetration X-39A.

4. Penetration to pipe fillet welds are inaccessible from the side next to containment . Perform surface examination on outside weld only.

5 One ci rcumferential weld is located inside containment penetration X-12 and is inaccessible for volumetric or surface examination.

6. Six circumferential welds and two support welds will have augmented surface examination. One circumferential weld is located inside containment Penetration X-17 and is inaccessible for volumetric or surface examination.

7 One circumferential weld is located inside each of containment Penetrations X-13A and X-13B and is LO inaccessible for volumetric or surface examination.

54

~.

~ 8. Design pressure is 150 psig and design temperature is 187 F.

t p3 jj 9 Augmented surface examination will be performed at Penet rat ions X-210A, X-2108, X-211 A and X-211B.

\fl 10. Augmented surface exam will be perforned on two welds near each of the suppression chamner Penetrations X-225A and X-225B.

Page 14 of 15 i

lNSFRVICE INSPECTION PROGRAM M. PARTRfDCE Il-3-73 q System _P E AC T O R C O R E 150 Cc,,

a TIN; IWpared By J.

i-.

Gg WELD

SUMMARY

SHEET Il825-FM-22A Approved By H. H. A;kwlTH, CHIEF ENDINEER 12-4-78 ISD No. Rev 13

~~~ James A. F#tzPatnck Nuclear Power Plant % -

Request for Rchrf Sy stem ISD No. of Cale C< =1r No. of in Eaeu of Lane Number Line Dm ript.on Chun Counhr.a te M ehl Typ- R eLis Exam E xeml>uon % elds R eason Examinauon Nutra 4- W2 2 - 9 0 2 - 4 FROM VALVE 13-MOV-21 TO THEkMAL 1 F-5 CIRCUM 21 VOL -

BRANCH 3 5UR -

SLEEVE IN RCIC LINE DIA s 6 IN SUFPORT 7 VOL 7 P5 SUR -

C-5 CIPCUM 2 VOL 2 P6 $UR S 4" '! 2 902 A- 4A F ROM RC IC THE RMAL SLEEVE I TO THERMit SLEEVE IN F E E CWATE R LINE 34-IE"-6FP-902A-4A THERMAL SLEEVE TPERMAL SLEEVE BETWEEN RCIC I D-5 ClkCUM 4 VOL - -

LCNDITUDINAL 2 VOL - - - I LINES 4"-W22-902-4 AND 4"-W22-902A-4A H-6 CIRCUM 0 VOL 8 R1 SUR 2 m 8- 5LP- 152 - 2 2 FROM VALVE RCIC-5 TO SUPPRESSICN 2 i

CHAMBER PENETRATIGN X-212 (jp(UM i HYD - - - -

[

BRANCH 3 VOL 3 R4 SUR -

SUPPORT 1 SUR D-4 CIRCUM 21 HYD E2 5 SUR 3 3"-SHP-902-17A FROM MAIN STEAM LINE TO VALVE I SUPPORT 3 HYD E2 3 SUR 3 13-MOV-16 THROUGH CON T A 6 Na'E NT PENETRATION X-10 F-8 CIRCUM 2 HY D El 2 -

SUR 4 6"-W22-152-16 FROM VALVE 13-MOV-41 TO 2 SUPPOR* 1 HYD El 1 SUR 4 SUPPRESSION CHAMBER PENETRATION X-224 L.%:-

$t e *p

. m 96

.z C4

JAMES A. FITZPATRICK Page 15 or 35 NOTES System: Reactor Core isolation Cooling

1. Fabrication welds on thermal sleeve

2. Pipe is 8-inch std. wall pipe wi th a wall thickness of 0.322 inches.

3 Augmented inspection is being performed at Penetration X-10.

4. Augmented inspection is being performed at Penetration X-224.

w 5 Pipe is 4-inch Schedule 80 pipe with a wall thickness of 0.337 inches.

h CC 4

v.

MN

-o L;h J3

Pag, 16 of 35 INSERVICE INSPECTION PRO 6 HAM M. J. PARTRIDGE q System CCRE SPRAY Prepared By 11-3-78_

gg- WE LD

SUMMARY

SHEET ISD No. li325-FM-23A Rcy, 13 Approved Hy H. H. ASKWITH, CHIEF ENGINEER 12-4-78 James A. FitiPatrck Nuclear Power Plant w, Required Sy stm ISD No of Chie Code No. of In Lied of Line Nmatwr Lme Descry taon Clan Coorduute Weld Type welds Exam Exemptuen welds Reason Ex am mation Notes 16"-W23-152-IA FROM SUPPPESSION CHAMBER 2 B-7 CIRCUM 9 HYD El 2 -

SUR 1,2 PENETRATION X-227A 10 LINE BRANCH 2 HYD Ei - - -

1 16"-W23-152-3A SUPPORT 3 HYD El 2 -

SUR 1,2 16"- W2 3- 15 2 - 2 A FROM LINE 16"-W23-152-3A TO 2 H-7 CIPCUM i HYD El - - -

1 VALVE CSP-SA 16"-W23-152-3A

ROM LINE 16"-W23-iS2-IA TO 2 C-7 CIRCUM S HYD El - - -

1 CORE 5 FRAY PUMP 14P-IA BRANCH 6 HYD El - - -

1 S U F'PC RT I HYO El - - -

1 16"-W23-152-IB FROM SUFPRESSICN CHAMBER 2 E-7 CIRCUM 9 H/D El 2 -

SUR 1,2 7 PENETRATION X-227B TO BRA.4CH 2 HYD El - - -

I

$ LINE 16"-W23-152-3B SUPPORT 4 HYO El 2 -

SUR 1,2 16"-W23-152-2B FROM LINE 16"- W2 3- 152- 3 B TO 2 J-8 CIRCUM i rtY D El - - -

1 VALVE CSP-8B 16"-W23-l$2-3B FROM LINE 16"-W23-152-1B TO 2 J-8 CIRCUM 5 HYD El - - -

1 CORE SPRAY PUMP 14P-IB BRANCH 5 HYD El - - -

1 SUPPORT I HYD El - - -

1 FROM CORE SPRAY PUMP 14-P-IA 2 H-6 CIPCUM 25 VOL 25 R2 PRS 3 12 "- W2 3- 302- 4 A CIRCUM HYD ES - -

PRS TO VALVE 14-MOV-IIA 3 BRANCH 17 VOL 17 R2 PRS 3 SUPPORT 5 SUR 5 P* PRS 3 0

4-N' Hli ew

Page 17 of 35 System CORE SPRAY Prepared By M. J. FARTR10GE 11-5-78 Na WELD

SUMMARY

SHEET ISD No. Il825-FM-23A Rev.13 Approved By H. H. ASKWITH, CHIEF ENGINEER 12-4-78 James A. Fit:Patrck Nuclear Power Plant u.

Request for Relief Requ nd .

Sy ste,i ISD No.of Code Code No. of In Lieu of lane Numt er Line D*wrtpuon class Coordmate % eld Type wells Exam Exemption Welds Reason Examination Notes HYD b 10"- W2 3- 902- 5 A FROM VALVE 14-MOV-IIA T0 I D-5 CIRCUM 15 VOL 1 R3 VALVE CSP-14A BRANCH 4 SUR - -

DIA % 6 (N SUPPORT 3 VOL 3 R5 $UR -

10"-W23-1504-5A FROM VALVE CSP-14A T3 PfACTOR I B-3 CIRCUM 2 VOL & - - - -

VESSEL P.0ZZLE N-S ClPCUM VOL - - -

BRANCH I SUR -

DIA < 6 IN SUPPORT 2 VOL 2 F3 SUR -

E-4 CIRCUM 2 VOL 2 P2 PRS 3 10"-W23-302-8A FROM LINE 12"-W23-302-4A TO 2 in VALVE 14-MOV-26A SUPPORT 1 SUR 1 R2 PRS 3 E-4 CIRCUM 6 FLW El, E4 - - -

5 10"- W2 3- 15 2- 9A FROM VALVE 14-MOV-26A TO LINE 2 BRANCH FLW El, E4 - - -

5 10-I6"-W20-152-5A 2 SUPPORT I FLW El, E4 - - -

5 12"- 2 3- 304 - 4 B FROM CCRC SPRAY PUMP 14P-IB 2 1-6 CIRCUM 25 V0L 25 R2 PRS 3 TO VALVE 14-MOV-IIB CIRCUM 2 HYD E5 - - -

BRANCH 18 kOL 18 R2 PRS 3 SUPPORT 6 SUR 6 R2 PRS 3 9 .A

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A A

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JAMES A. FITZPATRICK Page 19 of 35 NOTES 3 System: Core Spray

1. Design pressure is 50 psig and design t empe ra t u re is 150 F.

2. From suppression chamber penetrations to first isolation valve, augmented surface examination will be performed.

3 Design temperature is 150 F, design pressure is 400 psig; however, system operating pressure exceeds 275 psig for only approximately six (6) minutes in the unassisted Core Spray Operating Mode for the worst case small break. Accordingly, this portion of the Core Spray System is considered to be a moderate energy system since it does not normally operate during normal operation, cold shutdown, or refueling modes of operation. This system is normally maintained at a low nominal pressure and a

.:echanism for piping failure does not exist. The system is constructed of carbon steel, and therefore, is not subject to intergranular stress corrosion cracking and is not subjected to cyclic stresses or high stresses that would result in a piping failure. This portion of the system th 11 be " Pressure" tested every 3 1/3 years to confirm piping system integrity.

4. One circumferential weld is located inside Containment Penetration X-16A and is inaccessible for volumetric or surface examinations.

5 Design pressure is 150 psig and design temperature is 150 F.

6. One ci rcumferential weld is located inside containment Penetration X-16B and is inaccessible for volumetric or surface examinations.

7 Eight ci rcumferential welds, two 3/4 inch branch welds, and one support weld are located on a pipe 15 feet above floor level in a room housina the RWCU sludge tanks and non-regenerative heat exchangers.

There is limited access to this room and radiation levels exceeding 10,000 mR/ hour exist. Since the configurations and service conditions,of Lines 10"-W23-902-5A and 10"-W23-902-5B are essentially identical, generic problems would be identi fied on Line SA. The pressure test specified will yield little reduction in detection of flaws pnd will minimize the unnecessary radiation exposure to inspec-tion personnel .

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O Page 22 of 35 System H I G H PA F ,sa- , . . P E COOLANT INSERVICE INSPECTION PROGR AM q __

ITeparni by M. J. PARTRIDGE 11-20-78 Gg INJECTION WELD

SUMMARY

SHEET -~-~rm.

ISD No. I!875-FM-25A Itey. 12 Ap roved By H. H. ASk1lTH1CHIEF ENCINEER 12-4-73 II 2$-FM-25b Res. 11 James A. Fit 2 Patrick Nuclear Power Plant m,,

g g Request for Rehef Sy stem ISD No. of C<xie Code No. of In beu of Une Numbei Line Descripton Class Coord mate Weld Type Weld s Exam Exemption Welds Iteason Exam. nation Nu tes 10"- SH P- 90 2- 19, FROM MAIN SilAM LINE I B-2/25A CIRCUM 16 VCL I R3 HYD 1 29-24" WP- D2-IC TO VAL M E P/sNC H 3 SM - - - -

2 3- MOV- 16 DI A A 6 IN SUPPORT 4 VOL 4 R5 SUR -

10"-SHP-902-19 FROM VALVE 23-M04-16 70 h*Cl 2 0-3/25A CIRCUM 26 VOL - - - -

TURBINE 23-TU-2 C RCUM i HYD E5 - - - -

BRANCH 7 VOL 7 R4 SUR -

SUPPORT 6 SUR - - - -

10'-SHP-902-34 FRCM LINE l T '- SHP-902 - 19 TO 2 C-2/25B CIRCUM 2 VOL - - - -

LlhE I F 8-SHP-902- 32 A AND 32 8 BRANCH l VOL 1 R4 SUR -

ti e

14"-W25-902-3 FR^M HPCI PUMP 23P-IA TO 2 C-3/258 CIRCUM 20 VOL 20 P2 PDS 2 VALVE 23-MOV-19 ClkcuM 4 HYD E5 - - - -

BRANCH 7 VOL 7 R2 PRS 2 SUFPORT 6 SUR 5 R2 PRS 2 14"-W25-902A-3A FROM VALVE 23-MOV-19 TO I E-5/25A CIRCUM 9 v0L - - -

LINE 34-18'-WFP-902A-48 BPANCH 2 SUR - - - -

DIA % 6 IN 10"-W25-902-4 FROM LINE 14"-W25-902-3 TO 2 F-5/25A CIRCUM 8 VOL 8 R2 PRS 2 VALVE 23-M0v-21 CIRCUM i HYD E5 - - - -

BRANCH I VOL I R2 PRS 2 SUPPORT 1 SUR I R2 PRS 2 es

%. e

h. . ,

..

Page 23 of 35 System H l id PRESSURE CUOLINT l'repared By M. J. PARTRIDGE 11-20-73 lNJELil0N WELD

SUMMARY

SHEET *"

ISD No. Rev. 12 Approved by H. H. ASKWITH, CHIE F ENGINE ER 12-4-78 IIB 2)-FM-25A Il825-FM-258 Rev. 11 aus A N#aM Nudear %s h ma Request for Rehef Syste m ISD No. of Code Cafe No. of In Lieu of Lme Numtwr Luw thwription Clu Cmrd mate Weld Type Welds Exam Exemption Welds Reason Exammation Notes 18"-5LP-152-37 FROM HPCI TURBINE EXHAUST TO 2 D-4/25B CIRCUM 1 VOL I R2 FLW 3 LINE 20"-SLP-152-25 20"-5LP-152-25 fROM LINE 18'-5LP-152-37 TO 2 D-4/250 CIRCUM 13 VOL 11 R2 FLW 3 LINE 16"-5LP-152-25A 2 R2 SUR 3,4 BRANCH 13 VOL 13 R2 F '.W 3 SUPPORT 4 SUR 3 R2 FLW 3,4 16"-5LP-152-25A FROM LINE 20"-5LP-152-23 TO 2 B-5/25A CIRCUM 4 VOL 4 R2 SUR 3,4 LINE 24"-5LP-152-26 PRANCH I VOL 1 R2 FLW 3 24"-SIP-152-26 FROM LINc 16"-5LP-152-25A To 2 B-5/25A CIRCUM 2 VOL 2 R2 SUR 3,4 7 PENETRATION X-214 BRANCH I V0L 1 R2 FLW 3 U SUPPORT l SUR I R2 SUR 3,4 16"-W25-152-17 FROM SUPPRESSION LHAMBER 2 C-8/25A CIRCUM 12 HYD El 2 -

SUR 5,6 PENETRATION X-226 TO BRANCH 3 HYD El - - -

5 VALVE 23-hJV-57 SUPPORT 2 HYD El 1 -

SUR 5,6 16"-WCP-15?-l FROM VALJE 23-MOV-17 TO 2 1-2/25A CIRCUM 24 HYD El - - -

5 HPCI PUMP 23P-1A BRANCH 8 HYD El - - -

5 SJPPORT 2 HYD El - - -

5 ew bv

h. .a-V ay v :

JAMES A. FITZPATRICK Page 24 of 35 NOTES System: High Pressure Coolant injection

1. One ci rcumfe rent ial we ld is located inside containment Penetration X-Il and is inaccessible for volumetric or surface examination.

2. This portion of the HPCI system is considered to be a moderate energy system since it does not normally operate during normal operat ion, cold shutdown, or refueling modes of operation. This system is normally maintained at a low nominal pressure and a mechanism for piping failure does not exist. The system is constructed of carbon steel and, therefore, is not subject to inter-granular stress corrosion cracking and is not subjected to cyclic stresses or high stresses that would result in a piping failure. This portion of the system will be " Pressure" tested every 3 1/3 year to confirm piping system integrity.

3 Same as Note 2 except that a " Flow" test must be performed instead of a " Pressure" test since these w lines are open to the suppression chamber.

O

4. Augmented surface exam will be performed on circumferential and support welds from Penetration X-214 through valve HPI-65 including the first seismic support past valve HPI-65 5 The design pressure is 100 psig and the design temperature is 'E' F.

6. Augmented surface examination will be performed on two ci rcumferential welds and one support weld between Penet rat ion X-226 and Valve 23-M0V-58.

(O 4

C' N

Page 25 of 35

^

g System RE ACTOR WATER REC I RCUL ATION Prepared By M. J. PARTRIDGE 11-3 '/8 gg ISD No. Il825-FM-26B Itev. 9 WELD

SUMMARY

SHEET Approved 14y H. H. ASKWITH. CHIEF ENGINEER 12-4-78 James A. FitzPatrick Nuclear Power Plant n..

g Itequest for Relief System ISD No. of Cafe Cmic No. of In Lieu of Lme Number Lme Desenption Class Coordmate Weld Type Welds Exam Exemption Wehls Reason Examination Notes 28"-WH-CE-1A FROM REACTOR VESSEL N0ZZLE N-IB I E-5 CIRCUM i VOL & - - - -

TO RECIRC PUMP 02-2-P-1A CIRCUM 10 VOL - - -

BRANCi1 5 SUR - - -

DIA < 6 IN SUPPORT 3 VOL 3 R5 SUR -

RECIRC PUMP REClRCULATION PUMP 02-2-P-IA I C-5 SUPPORT 7 VOL 4 R5 SUR -

02-2-P-IA 28"-WH-CE-2A FROM RECIRC PUMP 02-2-P-1A I B-7 CIRCUM 7 VOL - - - -

TO LINE 28"-WH-GE-3A BRANCH 3 SUR - -

DIA < 6 IN 7 SUPPORT 4 VOL 4 R5 SUR -

2 22"-WH-GE-3A RECIRC RETURN HEADER (ROM 1 D-4 CIRCUM 4 VOL - - - -

LINE 28" WH-GE-2A LRANCH 4 VOL 4 R4 SUR -

DIA > 6 IN SUPPORT 4 VOL 4 R5 SUR -

12"-WH-CE-4A FROM LINE 22"-WH-CE-3A TO I D-4 CIRCLc4 i VOL & - - - -

REACTOR VESSEL N0ZZLE N-2A CIRCUM 4 VOL - - - -

BRANCH I SUR - -

DIA < 6 IN 12"-WH-GE-5A FROM LINE 22"-WH-GE-3A TO I D-4 CIRCUM I VOL & - - - -

REACTOR VESSEL N0ZZLE N-2B CIRCUM '- VOL - -

Kai - -

, . BRANCH I OUR - -

DIA < 6 IN

Page 26 of 35 g System RE ACTOR WAT ER RECI RCULAT ION Prepared by M. J. PARTRIDGE 11-3-78 g'g WELD

SUMMARY

SHEET ISD No, il825-FM-268 Rev. 9 Approved Hy H. H. ASKWITH, CHIEF ENGINEER 12-4-78 James A. FitzPatrick Nuclear Power Plant u...

Request for Rehef Sy stem ISD No. of Code Code No. of In Lieu of Line Numtwr Line Ihwnptain Class Coord:nate Weld Type Welds Exam Exemption Welds Reason Examination Notes 12"-WH-GE-6A FROM LINE 28"-WH-GE-2A 10 1 D-4 _UM I VOL & - - - -

REACTOR VESSEL N0ZZLE N-2C CIRCUM 4 V0L - - - -

B RANC H I SUR - -

DIA 6 6 IN 12"-WH-GE-/A FROM LINE 2 2-WH- GE - 3 A T O I D-4 CIP. CUM I VOL & - - - -

I REACTOR VESSEL N0ZZLE N-2D CIRCUM 4 VOL - - -

BRANCH 1 SUR - - -

I'lA < 6 IN 12"-WH-GE-8A FROM LINE 22"-WM-GE-3A TO I D-4 CIRCUM i VOL & - - - -

CD PEACTOR VESSEL N0ZZLE N-2E i CIRCUM 4 VOL - - - -

a BRANCH I SUR - - - -

DIA 4 6 IN 4"- WH- G E - 9 A BYPASS LINE ARCUND VALVE 02-MOV-53A I B-7 N RC'JM 10 VOL - - - -

28'-WH-GErlB FROM REACTOR VESSEL N022LE N-lA I F-5 CIRCUM i VOL & - - - -

TO RECIRC PUMP 02-2-P-1B CIRCUM 11 'JO L - - -

BRANCH 7 SUR - - - -

OIA < 6 IN SUPPORT 3 VOL 3 P5 SUR RECIRC PUMP RECIRCULATION PUMP 02-2-P-1B i H-6 $UPPORT 7 VOL 4 R5 SUR -

(,$ 02-2-P-18 A

' p '

'l

~ '

l l l l l l l

Page 27 of 35 INSERVICE INSPECTION PROGRAM Pm m lid-M System REACTOR WATER REClRCULATICN WELD

SUMMARY

SHEET te.

ISD No. Il825-FM-268 Ikv. 9 A;> proved By H . H. ASKWITH, CHIEF ENGINEER 12-4-l8 James A. Fit 1 Patrick Nuclear Power Plant im, Request for Relief System ISD No of Code Code No.of In laeu of Lme Number Iane Ih,.cription Cl.ca Coordinate Wold Type Welds Exam F:xemption Welds item >n F:xamination Notes 28"-WH-GE-28 FROM RECIRC PUMP 02-2-P-IB TO I H-8 CIRCUM 7 VOL - - - -

LINE 28"-WH-GE-3A BRANCH 5 SUR - - - -

DIA < 6 IN SUPPORT 4 VOL 4 P5 SUR -

22"-WH-GF-3B RETURN RECIRC HE ADE R FROM 1 G-4 CIRCUM 4 V0L - - - -

LINE 28"-WH-CE-28 DI IN 4 VOL 4 R4 SUR -

SUPPORT 4 VOL 4 R5 SUR -

12"-WH-CE-48 FROM L I NE 22"-WH- GE- 3B 10 I G-4 CIRCUM i VOL & - - - -

REACTOR VESSEL N0ZZLE N-2F CIRCUM 4 VOL - - - -

BRANCH I SUR - - -

7 DIA A 6 IN 12"-WH-CE-58 F ROM L INE 2 2"-WH-GE- 3B TO 1 G-4 CIRCUM I VOL 6 - - - -

BIM TALLIC SUR REACTOR VESSEL N0ZZLE N-2C CIRCUM 4 VOL - - - -

BRANCH I SUR - - - -

DIA 4 6 IN 12'-WH-GE-6B FROM LINE 22"-WM-GE-2B 10 1 C-4 CIRCUM i VOL & - - - -

REACTOR VESSEL N0ZZLE N-2H CIRCUM 4 VOL - - - -

BRANCH I SUR - - - -

DIA A 6 IN (O

.:s e

4 h d':

8 ** 8 s 7 . e - - - - - -

5 - [*7- i to 5 4 3

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N e R ML M R T UL U C6 O UL U 6 O O YR u lc C A C N P C A C P I

d RT R A< P RT R 4 P T

C A N l e I CM E I C B A R

S U I CM E

C I

A S U

E M k W l I l I B

P M ic D B D S

N U t r

S a I

P e z

_ E C LO i F t ta n

. I E Did 5 $

V W .A R I S r o

1 J

E o s C S e N m I

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ms e s ts la 1 I yC S

T N

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E e O R g T C

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C A t m o O 7

p T T D 2 i r

C M c o 4 R s l l F e l I L

5 D - -

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D D R e R 4 R 8

T N l in C l C

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JAMES A. FITZPATRICK Page 30 of 35 NOTES System: Control Rod Drive Return

1. Nine circumferential welds and two support welds in between containment isolation valves CRD-110 and CRD-Il4 will have augmented inspection by surface examination. One circumferential weld is located inside Penetration X-36 and is inaccessible for surface examination.

7 to

.L

(.'

O

Page 31 of 35 System MAIN STEAM PreparM By M. J. PARTRIDGE 10-29-78 WELD S.,MMARY SHEET "'*

ISD No. I l82 5-F M-29A Rev, 13 Approved By H. H. ASKWITH, CHIEF ENGINEER 12-4-78 James A. FitzPatrick Nuclear Power Plant zm.

g g Request for Rehef Sptem !3D No. of Code Code No. of In Lieu of Line Number Line Description Claw Coordinate Weld Type Welds Exam Exemption Welds Reason Exammation Notes 6"-SHP-902-5 FROM LINE 24'-SHP-902-IC I D-3 CIRCUM 2 VOL - - - -

TO VALVE 29-RV- 71G 6'- 5HP-902- 55 FROM LINE 24"-SHP-902-IC I D-3 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-71E 6'-SHP-902-5T FROM LINE 24"-SHP-902-lC 1 D-3 CIRLUM 2 VOL - - -

TO V' 29-RV-71F 24"-SHP-902-ID FROM REACTOR VESSEL TO MAIN I E-4 CIRCUM 19 VOL 1 R3 HYD 2 STLAM ISOLATION V ALVE 29- A0V-86D BRANCH 10 SUR - - - -

SUPPORT 10 VDL 10 R5 SUR -

?

6"-SHP-902-SQ FROM LINE 24"-SHP-902-ID 1 E-3 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-71H 6'-SHP-902-5R FROM LINE 24'-SHP-902-ID 1 E-4 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-71) 6"-SHP-902-5W FROM LINE 24"-SHP-902-ID 1 E-4 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-71L

(.O L

h?

3 T .

N O s.

Page 32 of 35 MAGN STEAM INSERVICE INSPECTION PROGRAM q System Prepared By M. J. PARTRIDGE 10-29-78 6g WELD

SUMMARY

SHEET wi.

ISD No. Il825-FM-29A Itey, 13 Approved By H. H. ASKWITH, CHIEF ENGINEER 12-4-78 James A. FitiPatnck Nuclear Power Plant ,,,

Required Sy stem ISD No.of Cale Code No. of In Lieu of line Number Line Description C ss Coordmate Weld Type Welds Exam Exemption Welds Reason Exammation Notes a

24"-5HP-902-1A FROM REACTOR VESSEL TO MAIN 1 F-3 CIRCUM 20 VOL 1 R3 HYD 1 STEAM ISOLATION VALVE 29-A0V-86A BRANCH ll SUR - - - -

SUPPO RT iO VOL 10 R5 SUR -

6"-SHP-902-5N FROM LINE 24"-5HP-902-1A 1 G-3 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-718 6"-SHP-902-SP FROM LINE 24"-SHP-902-IA I F-3 C l u. ? VOL - - - -

TO VALVE 29-RV-7tA 6"-sHP-902-5X FROM LINE 24"-SHP-902-IA I F-3 CIRCUM 2 VOL - - - -

TO VALVE 29-RV-7tK a

24"-rip-902-IB FROM REACTOR VESSEL TO MAIN STEAM i F-4 CIRCUM 25 VOL 1 R3 HYD I ISOLATl0N VALVE 29- A0V-86B BRANCH 9 SUR - - - -

SUPPORT 11 VOL 11 R5 SUR -

6"-SHP-902-5U FROM LINE 24"-SHP-902-10 TO I F-4 CIRCUM 2 VOL - - - -

VALVE 29-RV-71C 6"-5HP-902-SV FROM LINE 24"-SHP-902-1B TO I F-4 CIRCUM 2 VOL - - - -

A Y VALVE 29-RV-71D e- '

24"-SHP-902-IC FROM REACTOR VESSEL TO MAIN STEAM 1 E-3 CIRCUM 25 VOL 1 R3 HYD 2 t

ISOLATION VALVE 29-A0V-86C BRANCH 9 SUR - - - -

h b} SUPPORT 11 VOL 11 R5 SUR -

M

JAMES A. FITZPATRICK Page 33 of 35 NOTES System: Main Steam

1. One circumferential weld is located inside cach of containment Penetrations X-7A and X-78 and is inaccessible for volumetric or surface examination.

2. One ci rcumferential weld is located inside each of containment Penetrations X-7C and X-7D and is inaccessible for volumetric or surface examination.

?

t t0

.O C.'

eb

i Page 34 of 35 INSERVICE INSPECTION PROGRAM ,

g System FEEDWATER n=.

Gg WELD

SUMMARY

SHEET ISD No. Il825-FM-34A Rev. I SA Approved 13y H. H. AS dlTH, CHIEF ENGINEER 12-4-78 James A. FitzPatnck Nuclear Power Plant o,,

Request for Relief Systnn iSD No.of Code Code No. of in Lieu of Line Number Lme Dewrtption Class Coordmate Weld Type Welds Exam Exemption Welds Heasan Examination Notes FROM VALVE NRV-Ill A TO LINES C-5 CIRCUM VOL R3 HYD 1 IB"-WrP-902A-4A I 19 1 12"-WFP-902A-5A AND BRANCH 3 SUR DIA < 6 IN 12"-WFP-902A-5C SilFPORT 3 V0L 3 R5 Sun -

LONGITUDINAL 2 VOL - - -

2 12"-WFP-902A-5A FROM LINE 18"-WFP-902A-4A TO I A-4 CIRCUM 8 VOL - - -

REACTOR VESSEL N0ZZLE N-kB SUPPORT 1 VOL I R5 SUR -

12"-WFP-902A-5C FROM LINE 18'-WFP-902A-4A TO I A-4 CIRCUM 12 VOL - - - -

REACTOR VESSEL N0Z2LE N-4A SUPPORT 5 VOL 5 R5 SUR -

FROM VALVE NRV-Ill8 TO LINES C-4 CIRCUM 16 VOL 1 R3 HYD I y 18'-WFP-902A-48 1 12"-WFP-902A-58 AND BRANCH 3 SUR -

C DIA < 6 IN 12"-WFP-902A-5D SUPPORT 3 VOL 3 R5 SUR -

12"-WFP-902A-58 FROM LINE 18"-WFP-902A-48 TO 1 B-4 CIRCUM 8 VOL - - -

REACTOR VESSEL N0ZZLE N-4C SUPPORT i VOL I R5 SUR -

12"-WFP-902A-5D FROM LINE 18"-WFP-902A-4B TO I B-4 CIRCUM 12 VOL - - - -

REACTOR VESSEL N0ZZLE N-4D SUPPORT 5 VOL 5 R5 SUR -

'D W

s - D

.5

JAMES A. FITZPATRICK Page 35 of 35 NOTES System: Feedwater

1. One circumferential weld is located inside each of containment Penetration X-9A and X-9B and is inaccessible for volumetric or surface examinations.

2. Fabrication welds on thermal sleeve.

T 0

t; 4

NP

.b.~p f

~

'^ *

~E B

-= APPENDIX C INSERVICE INSPECTION DIAGRAMS r

E One set of Inservice inspection Diagrams .

color-coded to show the ASME Class 1, '

Class 2, and Class 3 boundaries is included in one copy of this report being submitted to the Nuclear Regulatory Commission. -

- v.-

~ .B -

h LE

.g E ,

~B wu

,up,h.Umm/ Tith

/

. ~- g

ML19208A312: Difference between revisions

Revision as of 07:10, 2 February 2020

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1.0 INTRODUCTION

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ML19208A312: Difference between revisions

Revision as of 07:10, 2 February 2020

Text

Subject:

Dear Sir:

1.0 INTRODUCTION

SUMMARY

SUMMARY

Reference:

Reference:

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Reference:

SUMMARY

SUMMARY

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SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

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