ML18085A268

Forwards Inservice Insp Program

ML18085A268
Person / Time
Site:	Salem
Issue date:	10/29/1980
From:	Librizzi F Public Service Enterprise Group
To:	Miraglia F Office of Nuclear Reactor Regulation
Shared Package
ML18085A270	List: ML18085A268 ML19345A806
References
NUDOCS 8011250237
Download: ML18085A268 (27)

Text

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• Public Service Electric and Gas Company

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80 Park Place Newark, N.J. 07101 Phone 201;43'd:766cf;p ""'rnh Mr. Frank Miragila, Chief Licensing Branch 3 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

INSERVICE INSPECTION AND TEST PROGRAM UNIT NO. 2 SALEM GENERATING STATION DOCKET NO. 50-311 In accordance with 10CFR50.55A, we hereby transmit the Inservice Inspection Program for No. 2 Unit. The Inservice Program for Pumps and Valves was previously transmitted on April 23, 1980. Should you have any questions regarding this transmittal, do no hesitate to contact us. Very truly yours, Frank P. Librizzi General Manager - Electric Production 1oo:i-19iH Fl U:\llSIW SUl\lCE f 95-2001 (200M) 2-78

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l,'! INSERVICE INSPECTION PROGRAM UNIT 2 - SALEM GENERATING STATION INTRODUCTION: Commencing with the date of Commercial Operation, and for the duration of first 120 month inspection interval, the Inservice Inspection Program for Class 1, 2 and 3 components and systems is based on Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975. Section XI requirements are augmented by cer-tain special requirements as defined in the Salem Unit No. 2 Technical Specifications, Sections 4.4.5.0, 4.4.10.1 and 4.4.10.2 for examination of steam generator tubing, reactor coolant pump flywheels and steam generator channel head cladding. Repairs in the pressure retaining boundary of systems and components falling within the scope of this program will be made in accordance with Division 1 of Section XI, or, if such repairs are not specified in Division 1, repairs will be made in accordance with the provisions of the Code appli-cable to the construction of the component. SCOPE: Class 1, 2, and 3 components and the methods of examination for each component are listed in Tables 1, 2 and 3 respec-tively. Examination boundaries are defined on modified system piping diagrams entitled "ISI Boundary Diagrams" in Attachment 2. The schedule for examination of specific components is described in the long term inservice examination plan. REQUEST FOR RELIEF: Relief is requested from certain requirements of Section XI that have been determined to be impractical because of accessiblity and design limitations established prior to issuance of the 1974 edition of the Code, and for other con-siderations. These requirements and the affected components are itemized in Attachment 1. In some cases, alternate means of testing are described. It is tqe position of PSE&G that the relief will not compromise the integrity of safety related components. RFB:az LP4 la

ATTACHMENT 1 ITEMS FOR WHICH RELIEF IS REQUESTED

1. Bolts in Category B-G-1, except as many as four bolts in each of four reactor coolant pumps, will be examined ultrasonically in place during the inspection interval. Those bolts which are inaccessible due to interference from structural members and piping will be examined at or near the end of the inspection interval when the pumps are disassembled for maintenance. All bolts in Category B-G-2 will be examined visually during the inspection interval. If visual examination reveals indications of distress, the bolts will be removed for surface and/or volumetric examination.

2. Ultrasonic indications will be recorded at 50% of reference level (DAC) unless suspected by the examiner to be other than geometric in origin, in which case they will be recorded and investigated by a Level II or Level III Examiner if in excess of 20% of DAC. All indications above 100% DAC will be resolved as to their shape and identity by a Level II or Level III Examiner.

Justification for departure from a 20% reference level evaluation criterion for all UT examinations is explained in detail in Attachment lA, prepared by Southwest Research Institute as agents for PSE&G.

3. Ultrasonic examination of ferritic piping will be per-formed in accordance with Article 5 of Section V of the ASME Boiler and Pressure Vessel Code with the exception that all indications will be recorded and evaluated as indicated in paragraph 2 above.

4. Inaccessibililty for examination was identified during the Preservice Inspection and has been described in the Remarks column of Tables l and 2 of the Salem No. 2 Preoperational Baseline Examination Report. Those items which could not be examined due to inaccessi-bility have been. extracted from these tables and compiled into an abbreviated table which has been included as Attachment lB for convenient reference.

Comments in these tables are necessarily brief. P8023/05 01 Page 1 of 2

ATTACHMENT 1 (CONT'D) Information in greater detail has been provided in the way of additional comments, also included as part of Attachment lC. Consideration of alternate methods are discussed and in some cases sketches are attached for greater clarity.

5. Welds having limi~ed ultrasonic examinations are treated in the same manner and are included in the tables and comments mentioned above. No distinction is made between items having no accessibility and limited

            .inspectability except in the nature of the comment.

Detailed radiographic procedures for augmented or alternative examinations will be prepared at such time as the need arises in order to take advantage of the possibility of more advanced technology at that time.

6. The holding time after pressurization and prior to commencement of visual examinations during pressure tests required by IWA 5000 of Section XI will be 10 minutes when insulation is removed. This procedure is presently specified in recently approved editions of Section XI and is considered acceptable practice for an exposed pressure* boundary. The longer holding time required by the 1974 Edition of Section XI presents unnecessary exposure to radiation in an operating plant.

RFB:pac 7-16-80 P8023/05 01/02 Page 2 of 2

ATTACHMENT lA

Title:

Comments .Concerning the 20 Percent Versus 100 Percent Evaluation Level for Ultrasonic Examination of Nuclear Power Plant Piping Introduction I. The Nuclear Regulatory Commission (NRC) has asked several plant owners for detailed information to justify two things. A. That a 20% reference level evaluation criterion is impractical and B. That a 100% reference level evaluation criterion will provide a level of safety comparable to the Section V code requirements (of evaluation at 20%). Discussion II. Southwest Research Institute (SwRI) presents the fol~owing considerations on these two closely related questions, taking them in order: A. The impracticalit~ recording/evaluation at the 20% reference level.

1. The welded joints in nuclear piping frequently contain code-allowable wall thickness differences (12-1/2% of thickness) as well as allowable weld dropthrough and other conditions such as counterbore taper, crown, etc. These conditions can provide an extremely large number of geometric reflectors (with or without mode conversion) which produce ultrasonic examination (UT) indications greater than 20% of the UT reference level (DAC) (see attached graph). Weld metal in stainless steel piping contains, in addition, reflectors due to metallurgical grain structure which can also produce indications greater than 20%

Attachment lA Page 1 of 9 P8035/05 1

• .t DAC. It appears that the incidence of geometric reflectors increases exponentially as the amplitude is reduced.

2. Two stress-corrosion cracks are known to have been missed by SwRI normal examination techniques. However, they were not missed because of lack of detectability; indications of 141% and 159% DAC were obtained from these stress-corrosion cracks in the HAZ. They were not identified because of the large number of equally high amplitude geometric indications from the adjacent root area which, in effect, masked the test data to preclude identifying these cracks. Reducing the recording level to 20% will cause this problem to exist on a much larger scale in that the tremendous increase in recorded indication data will obscure real flaw indications.

3. During the performance of inservice inspections, significant radiation exposure is being experienced by all the inservice examination personnel. In SwRI's experience the examination staff receives essentially all of the legally allowed radiation exposure when recording 50% DAC data. To evaluate and record 20%

data would require that the personnel spend several times as much time in a radiation area to obtain additional ultrasonic data which is not practically decipherable and would require a proportional increase in radiation exposure to the available examination personnel. Therefore, these personnel would not be available for the performance of ultrasonic examinations of as many lines or at as many sites. Necessarily, this would force the industry to reduce the sampling rate of examinations because of the inavailability of trained personnel. The Attachment lA Page 2 of 9 P8035/05 2

• J reduction of sample size would have a detrimental effect on the monitoring of plant integrity through inservice inspection and would eliminate the non-mandatory examinations presently performed by the utilities in the interest of promptly examining known or suspected problem areas.

4. A typical example of the impracticality of the 20% level recording/evaluation practice involved the examination of the 4" Recirculation Bypass lines in a nuclear power plant. The job required both 45-degree and 60-degree angle-beam examinations on one or both side of 20 pipe welds having a total of 450 inches of weld examination length. To demonstrate the impact of the 20%

recording criteria to the utility, a small sample of a randomly selected weld was examined. Because of radiation levels, the demonstration was limited to one hour. In that hour, 15 separate iridications were recorded with the 60-degree examination in 5/8-inch of weld length while 10 indications were recorded with the 45-degree examination in 1-7/8 inches of weld length. Only maximum amplitude positions were recorded and most indications of 20'% DAC and greater were recorded. It is recognized that this was a very small sample, but it is believed to be typical of 4" bypass line welds. The three-man crew successfully completed the examinations recording at the 50% level within the one-day of examination time available on the unit. Evaluation time would have been increased proportionately with dubious conclusions due to the sheer volume of data. The welds were judged to be free of cracks based on the SO% DAC recording and 100% evaluation criteria and several months of successful operation without leaking Attachment lA Page 3 of 9 P8035/05 3

confirmed that these examinations, like those performed at several other sites, were effective. B. Evaluation at 100% and greater provides equivalent safety.

1. Equivalent safety, comparable to code*

requirements, is assured by the recording/evaluation criteria developed, refined, and qualified by SwRI through many years of research and experience. This criteria is embodied in current SwRI practice, which requires that: (a) All indications 50% of DAC or greater shall be recorded. (b) All indications 100% of DAC or greater shall be investigated by a Level II or Level III operator to the extent necessary to determine the shape, identity, and location of the reflect0rs. (c) Any indication 20% of DAC or greater and suspected by the operator to be other than geometric in nature, including all 20% or greater indications originating in the base metal, shall be recorded and investigated by a Level II or Level III examiner to the extent necessary-to determine the shape, identity, and location of the reflector.

   *It must be noted that the 100% evaluation criteria was (and is) the Code requirement for utilities commited to the 1971 Edition of Section XI (S71 Addenda). The inconsistency arose when the 1974 Edition of Section XI incorporated Section V by reference. In the Summer of 1976 Edition (IWA-2232) the ASME has reconfirmed its previous position by clearly requiring evaluation of only indications of 100% DAC or greater.

Attachment lA Page 4 of 9 P8035/05 4

(d) Any indications investigated and found to be other than geometric in nature shall be reported to the owner for evaluation and disposition. SwRI's long-standing requirement to record 50% DAC information reflects the necessity to record a sublevel of data below that point at which we feel a concern. The prime reason for recording this information is to allow for the known variation in r~producibility of test data. We have shown data reproducibility to be a factor caused by many thing including operator experience, training, procedure, equipment variations, environmental encumberances, test piece conditions, calibration standards, etc. These facators are routine and will continue to occur during the application of examinations of this nature on piping. This practice is belived to be more conservative than the intent of any edition of the Code and to provide greater safety at less cost in time, dollars, and radiation exposure to personnel than simply requiring the recording/evaluation of all 20% DAC data.

2. The adequacy of this practice is supported by the following:

(a) Except for a very limited number of applications of the 20% evaluation level criteria of Paragraph IX-3470 of the 1971 issue of Section III, the 100% reference level evaluation criteria of Paragraph IS-213.5 of the Summer 1971 Addenda to Section XI was in effect until the adoption in late 1976 of the 1974 Edition of section XI. The 100% recording criteria was endorsed by the Section V Subcommittee for Nondestructive Examination in a Attachment lA Page 5 of 9 P8035/05 5

• J Code Inquiry of 1973, and appeared in Paragraph T-544 of the 1974 Edition of Section v. There is no question of the overall success of the inservice examination program during these many years, and the 100% evaluation criteria was reconfirmed by ASME in Paragraph IWA-2232 of the Summer 1976 Addenda to the Section XI code.

(b) As a result of the different failure mode of austenitic piping (noted in Paragraph (d) below) SwRI had developed modified approaches in procedure to maintain assurance of maximum crack detection sensitivity. Search unit size, frequency and beam angle, as well as procedure, are optimized to take advantage of the known parameters of the type of failure to be detected and investigated in different situations. (c) While it has been demonstrated that significantly deep through-wall stress-corrosion cracking may give only a low amplitude response, it has been demonstrated by SwRI on multiple plants that the 100% evaluation criteria, augmented by operator investigation at the 20% level, can be applied with satisfactory results: (1) No component or pipe examined by SwRI has experienced leaking by way of a stress-corrosion crack between the periodic examinations. (2) At least 48 piping cracks have been found and repaired in the early stages of propagation. Attachment lA Page 6 of 9 P8035/05 6

.. ( J (d) Much experience has shown that the typical mode of failure in stainless steel piping is not in the weld metal, per se, but is "stress-corrosion cracking" in the adjacent heat-affected zone (HAZ) and base metal. A trained UT operator can distinguish the difference between the usual weld-metal geometric indications and the somewhat similar indictions due to stress-corrosion cracking by noting their location in the base metal of HAZ. This is true even when their amplitude is in the 20% to 50% range and even though indications in this range originating in the weld metal cannot be identified. (e) A prime example of the adequacy of the total SwRI examination technique is that Recirculation Bypass lines have been ultrasonically examined in accordance with RO Bulletin 74-10 in six nuclear power plants. Thirteen cracks were found in four plants and the findings were confirmed by other methods, including excavation, in all cases. As noted above, no component or pipe examined by SwRI has experienced leaking by way of stress-corrosion cracking between periodic examinations. Summary and Conclusions III. For the reasons enumerated above, SwRI reco~~ends to its clients that, in the interests of maintaining maximum nuclear power plant integrity and safety at ~inimum cost in time and personnel radiation exposure, and effort to institute a blanket 20% DAC recording/evaluation criteria be Attachment lA Page 7 of 9 P8035/05 7

I - resisted. Instead, SwRI recommends a commitment to the SwRI recording/evaluation practice which was set out in Paragraph B.l above and is reinterated below: (a) All indications 50% of DAC or greater shall be recorded. (b) All indications 100% of DAC or greater shall be investigated by a Level II or Level III operator to the extent necessary to determine the shape, identity, and location of the reflectors. {c) Any indication 20% of DAC or greater suspected by the operator to be other than geometric in nature, including all 20% or greater indications originating in the base metal, shall be recorded and investigated by a Level II or Level III examiner to the extent necessary to determine the shape, identity, and location of the reflector. (d) Any indications invest~gated and found to be other than geometric in nature shall be reported to the owner for evaluation and disposition. Attachment lA Page 8 of 9 P8035/05 8

4500 4-300 BASfO ON A T'( Pl CAL ULTJ2..ASON l C 4-100 'SURVEY Of:. EX.AM ll-JA.TtOt-J R.ESULT5 3,00 LE\/EL Ir. f012.. CLASS PIPING, V1 ?nob i?XAN'\INf Jl.Ci ) 2 3500 0 5 To \0 TIN.ES 33ao ~ 3100 As MAtlY u Q 2.'}00 I~D1c/\not-1s

z 2100 WILL BE" lL 2500 0 FovtJD IN THE 1100 cJ. 20% To So'?..

UJ 2.100 ro 1'100 '\iA~CfE THAN "L noo TUE TOTAL

J 1500 2 N v ti\ BER. OF 1300 I NDI GATIONS 1100 ABo\JE 5G'?c,

     '300 700 500 300 100 o*       zo*      40 .. coo dO IOO   IZD   150  200 2.50 310 400  .500 ~30 7'JO 1000 PEl2.CENT OF DAC
       ~  No RECOR.DED DATA AYAlLA.BLE t=Ol2- TABULATION,

I*.* ATTACHMENT lC

                       . DETAILED EXPLANATION OF EXCEPTIONS TO THE ASME BOILER AND PRESSURE VESSEL CODE SECTION XI FOR THE SALEM UNIT NO. 2 INSERVICE INSPECTION PROGRAM Item Bl.2  Category B-B Closure Head Peel Segment-To-Disc Circumferential Weld Volumetric examination will not be performed on this weld due to inaccessibility. A visual examination for leaks will be performed during pressure tests required by Section XI. This weld is identified as 2-RPVCH-6446 B in the examination plan.

As indicated in PSE&G Sketch No. 8-9-77 attached, this weld is located within the area covered by the Control Rod Drive Penetrations. This location prevents access to the weld for any type of volumetric or surface examination from either the inside or outside surface of the closure head. To gain access would required re-moving and re-installing numerous Control Rod Drive penetration tubes. Bottom Head Peel Segment-To-Disc Circumferential Weld Volumetric examination will not be performed on this weld due to inaccessibility. A visual examination for leaks will be performed during pressure tests required by Section XI. This weld is identified as 2-RPV-3443 in the examination plan. As indicated in PSE&G Sketch No. 8-9-77 attached, this weld is located within the area covered by the instrumentation tube penetrations. This location prevents access to the weld for any type of volumetric or surface examination from either the inside or out-side surface of the bottom head. To gain access would require removing and re-installing numerous instrumen-tation tubes. P802304 01 Page 1 of 5

ATTACHMENT lC (CONT'D) Item Bl.4 Category B-D Reactor Vessel Inlet Nozzle to Shell Welds Volumetric examination will not be performed when the Core Barrel is in place due to inaccessibility. The examinations will be performed when the core barrel is removed during the inspection interval. These welds are identified as 27.5-RCN-1210, 27.5-RCN-1220, 27.5-RCN-1230, and 27.5-RCN-1240 in the examination plan. As indicated in PSE&G Sketch No. 8-15-77 attached, the Core Barrel covers the weld area when it is in place. Since all weld examinations are performed from the inside surface using a mechanized inspection device, these welds are inaccessible at this time. Examina-tions cannot be performed from the outside surf ace due to the area being covered by insulation on the shell of the weld and the configuration of the weld joint on the nozzle side of the weld. The insulation on the shell portion of the reactor vessel is not designed for removability and to remove it would require the cutting away of the insulation support rings from the vessel. Since the Core Barrel is scheduled to be removed at or near the end of the inspection interval, and ASME Code Case N-73 (1647) allows these examinations to be deferred to the end of each inspection interval, it is PSE&G's position that the intent of the Code is being complied with. Reactor Vessel Outlet Nozzle to Shell Welds Volumetric examination will be performed with the Core Barrel in place except for the examination of the Reactor Vessel base metal due to inaccessibility. The base metal will be examined when the Core Barrel is removed during the inspection interval. These welds are identified as 29-RCN-1210, 29-RCN-1220, 29-RCN-1230, and 29-RCN-1240 in the examination plan. P802304 02 Page 2 of 5

ATTACHMENT lC (CONT'D} As indicated in PSE&G Sketch No. 8-16-77 attached, the Core Barrel covers the reactor vessel base metal when it is in place. Since all weld examinations are per-formed from the inside surface using an automated inspection device, these welds are inaccessible at this time. Examination cannot be performed from the out-sidesurface due to the area being covered by insulation which would require the cutting away of the insulation support rings to remove. Since the Core Barrel is scheduled to be removed at or near the end of the inspection interval, and ASME Code Case N-73 (1647) allows these examinations to be deferred to the end of each inspection interval, it is PSE&G's position that the intent of the Code is being complied with. Item Bl.14 Category B-I-1 Reactor Vessel Cladding Visual examination will not be performed on the Reactor' Vessel Shell Cladding when the Core Barrel is in place due to inaccessibility. The patches of cladding to be examined are identified as 2-RPV-Patch-l, 2-RPV-Patch-2, 2-RPV-Patch-3, 2-RPV-Patch-4, 2-RPV-Patch-5, and2-RPV-Patch-6 in the examination plan. Since this examination can only be performed from the inside surface of the reactor vessel shell, the exami-nation can only be performed when the Core Barrel is removed. The Core Barrel is scheduled to be removed at or near the end of the inspection interval. Item B4.5 Category B-J Reactor Coolant Loop Piping Longitudinal Welds Volumetric examinations will not be performed on the longitudinal welds in the Reactor Coolant piping elbows. Surface examinations will be performed as well as a visual examination for leaks required by Section P802304 03 Page 3 of 5

ATTACHMENT lC (CONT'D) XI. These welds are identified in the examination plan as the LU-1, LU-0, LD-1, and LD-0 welds on lines 31-RC-1240,31-RC-1230, 31-RC-1220, 31-RC-1210, 29-RC-1240, 29-RC-1230, 29-RC-1220, and 29-RC-~210. The Reactor Coolant loop elbows are made of cast stain-less steel which cannot be penetrated by ultrasonic examination techniques. Also, acceptable radiographs cannot be obtained by present techniques due to fogging of the film caused by long exposure time coupled with background radiation from the reactor coolant piping. Radiography will be considered as new techniques are developed that would make it possible to examine these welds in the future. Reactor Coolant Loop Piping Circumferential Welds Volumetric examination on eight reactor coolant loop piping ~~lds will be limited during inservice inspection due to restricted access. These welds are identified as 31-RC-1210-5 and 6, 31-RC-1220-5 and 6, 31-RC-1230-5 and 6, and 31-RC-1240-5 and 6 in the examination plan. Due to anti-whip restraints that were installed after the preservice examination was completed, portions of these circumferential welds are inaccessible for inservice examination. Item B5.6 Category B-L-1 Reactor Coolant Pump Casing Welds A visual examination for leaks will be performed. The Reactor Coolant Pump casings are made of cast stainless steel which cannot be penetrated by ultra-sonic examination techniques. Also, acceptable radio-graphs cannot be obtained due to fogging of the film P802304 04 Page 4 of 5

I Y. ATTACWtENT lC (CONT'D) caused by long exposure time coupled with background radiation from the pump casing. Radiography will be considered as new techniques are developed to examine these welds in the future. RFB:pac 7-16-80 P802304 05 Page 5 of 5

... } Attachment 2 Syptems ISI Boundary Diagrams

1. Reactor Coolant 249825 - A - 1777

2. Steam Generator Feed & Condensate 249826 - A - 1777

3. Main Reheat and Turbine By-Pass Steam 249827 - A - 1777

4. Chilled Water 241828 - A - 1568

5. Fire Protection 241829 - A - 1568

6. Steam Generator Drains & Blowdown 249830 - A - 1777

7. Chemical & Volume Control - Operation 249831 - A - 1777

8. Chemical & Volume Control - 249832 - A - 1777 Boric Acid Recovery

9. Chemical & Volume Control - 249833 - A - 1777 Primary Water Recovery

10. Component Cooling 249834 - A - 1777

11. Residual Heat Removal 249835 - A - 1777

12. Spent Fuel Cooling 249836 - A - 1777

13. Safety Injection 249837 - A - 1777

14. Containment Spray 249838 - A - 1777

15. Auxiliary Fe~d Water 249839 - A - 1777

16. Waste Disposal Liquid 249840 - A - 1777

17. Service Water 249841 - A - 1777

18. Sampling 249842 - A - 1777

19. Demineralized Water - Restricted Areas 241843 - A - 1568

20. Waste Disposal Solid 241844 - A - 1568 NM:az LBl 1-B

CLASS 1 COMPCNENI'S, PARIS AND ME.'rnOOO OF EXAMINATION e O* ' Examination I Category Item Table No. IWB-2500 Corrponents and Parts to be Examined Method

                                       ~actor   Vessel Bl.1     B-A        Longitudinal and circumferential shell welds in core region            VolUITetric Bl.2     B-B        Lcntitudinal and circumferential welds in shell (other than            Volurretric those of Category B-A and B-C) and rreridional and circumferential seam welds in oottom head and closure Bl.3     B-C        Vessel-to-flange and head-to-flange circumferential welds              VolUITetric Bl.4     B-D        Prirrary nozzle-to-vessel welds and nozzle inside radiused section     Volurretric Bl.5     B-E        Vessel penetrations, including control rod drive and instrurrentation   Visual (IWA-5000) penetrations.

Bl.6 B-F Nozzle-to-safe eoo welds Volunetric and Surf ace Bl. 7 B-G-1 Closure studs, in place Volurretric Bl.8 B-G-1 Closure studs with nuts, when renoved Volurretric and Surface Bl.9 B-G-1 Ligarrents between threaded stud holes Volurretric Bl .1,0 B-G-1 Closure washers, bushings Visual Bl.11 B-G--2 No carponents within this category Bl.12 B-H No conponents within this category Bl.13 B-I-1 Closure Head* cladding l)Visual and Surface or 2 )Volurretric Bl.14 B-I-1 Vessel Cladding Visual Bl.15 B-N-I Vessel Interior Visual Bl.16 B-N-2 Interior attachrrents and core support structures Visual Bl.17 B-N-3 Core-support structures Visual Bl.18 B-0 Control roo drive housings Volurretric Bl.19 B-P Exerrpted corrponents Visual (IWA-5000) Pressurizer B2.l B-B Longitudinal and circumferential welds Volurretric B2.2 B-0 Nozzle-to-vessel radiused section Volurretric B2.3 B-E Heater penetrations Visual (IWA-5000) B2.4 B-F Nozzle-to-safe and welds Volum=tric and Surface B2.5 B-G-1 No corrponents within this category B2.6 B-G-1 No corrponents within this category B2.7 B-G-1 No corrponents within this category B2.8 B-H No corrponents within this category B2.9 B-I-2 Vessel cladding Visual B2.10 B-P Exerrpted corrponents Visual (IWA-5000) B2.ll B-G-2 Pressure-retaining oolting Visual P8035/04 1/2

I ~.- .. TABLE 1 CCNI'INUED Examination Category I l tt:*:. Table

_

i. IWB-2500 Conponents and Parts to be Examined Method Steam Gi::!nerators B3.l B-B Longitudinal and circumferential welds, including tube sheet-to-head .Volurretric or shell welds on the prinary side B3.2 B-D Nozzle-to-head welds and nozzle inside radiused section on the Volurretric priirery side B3.3 B-F Nozzle-to-safe end welds Volurretric and Surface B3.4 B-{;-1 No oonponentswithin this category B3.5 B-G-1 No CQf!POnents within this category B3.6 B-{;-1 No conponents within this category B3.7 B-{;-1 No coirponents within this category B3.8 B-I-2 Vessel Cladding Visual B3.9 B-P Exerrpted corrponents Visual (IWA-5000)

B3.10 B-G-2 Pressure-retaining bolting Visual 4.4.5.0* Tubing Eddy Current 4.4.10.2* Vessel Cladding (Steam Generator #24) Volurretric Piping Pressure Boundary B4.l B-F Safe-end to piping welds and safe-end in branch piping welds Volurretric and Surface B4.2 B-G-1 No cooponents within this category B4. 3 B-G-1 No corrponents within this category B4.4 B-G-1 No cooponents within this category B4 .5 B-J Circumferential and longitudinal pipe welds Volurretric B4.6 B-J Branch pipe connection welds exceeding six in. diarreter Volurretric B4.7 B-J Branch pipe connection welds six in. diarreter and sireller Surface B-Lii B-J Socket welds Surf ace B4.9 B-K-1 Integrally welded supports Volurretric 84 .10 B-K-2 Support cooponents Visual B4.ll B-P Exerrpted cooponents Visual (IWA-5000) B4.12 B-{;-2 Pressure-retaining bolting Visual

                                     !Eactor Coolant Punps B5.l      B41         Pressure-retaining bolts and studs, in place                          Volurretric B5.2      B-{;-1      Pressure-retaining bolts and studs, when re11Dved                     Volurretric and Surface B5.3      B-{;-1      Pressure-retaining bolting                                            Visual B5.4      B-K-1       Integrally-welded supports                                            Volurretric Unit 2
                           *Salem Technical Specifications P2:J35/04 3/4

Method Visual Volurretric Visual Visual (IWA-5000) Volurretric and Surface Visual Visual Visual (IWA-5000) Visual P8035/04 5/6

TABLE 2 t,V*

        '      *Ir
  - -' - - - - - - - - - - - C I A S - - S - 2~' PARTS, AND ME.'IEODS OF EXAMINATION Examination I              Category Ite::-. Table

-:o. IWB-2520 Conponents and Parts to be Examined Method Pressure Vessels Cl.l C-A Cira.mtferential butt welds Volurretric Cl.2 C-B Nozzle-to-vessel welds Volurretric Cl.3 c-c Integrally-welded supports Surface Cl.4 C-D Pressure-retaining bolting Visual and either surface or volUJTEtric Piping C2.l C-F,C-G Circurnferential butt welds Volurretric C2.2 C-F,C-G Longitudinal weld joints in fittings Volurretric C2.3 C-F,C-G Branch pipe-to-pipe weld joints Volunetric C2.4 C-D Pressure-retaining bolting Visual and either surface or vollJJIEtric C2.5 C-E-1 Integrally-welded supports Surface C2.6 C-E-2 Support corrponents Visual Purrps l C3.l C-F,C-G Purrp casing welds Volurretric C3.2 C-D Pressure-retaining bolting Visual and either surface or volurretric C3.3 C-E-1 Integrally-welded supports Surface C3.4 C-E-2 Support corrponents Visual Valves C4.l C-F,C-G No corrp:>nents within this category C4.2 C-D Pressure-retaining bolting Visual and either surf ace or volurretric C4.3 C-E-1 Integrally-welded supports Surf ace C4.4 C-E-2 Support c011pOnents Visual l Class 2 Corrponents (subject to the exerrption criteria of IWC1220) Visual

                               - 01arging Safety Injection Purrps 21, 22 and 23
                               - No. 2 Reactor Coolant Filter
                               - No. 2 Excess Letdo.m Heat Exchanger (Tube Side)
                               - No. 2 ~enerative Heat Exchanger
                               - l'b. 2 LetdCMn Heat Exchanger (Tube Side)

Accuin.Jlators 21, 22, 23 and 24 Boron Injection Tank Refueling Water Tank Safety Injection Purrps 21 and 22

                               - RHR Heat Exchangers 21 and 22
                               - RHR Purrps 21 and 22
                               - Cllemical Volurre and Control Tanlc
                               - Head Tanks 21, 22, 23 and 24
                               - Refueling Water Storage Tank Heat Exchanger
                               - Refueling Water Storage Tank Heating Recirc. Purrp
                               - Contail'lllEnt Spray Puprrps 21 and 22
                               - Steam Generators 21, 22, 23 and 24 (Shell Side)

Pi~035/J.; 7/2,

TABLE 3 CLASS 3 COMPONENTS Examination In Accordance With IWD-2400 (Visual) Reactor Coolant System Pressurizer Relief Tank Chilled Water System

1. Chillers #21, 22 and 23

2. Chilled Water Strainers

3. No. 2 Expansion Tank

4. Chilled Water Pumps Chemical Volume & Control - Operations

1. Resin Fill Tank2

2. No. 2 Chemical Addition Tank2

3. No. 2 Boric Acid Batching Tank2

4. Boric Acid Tanks 21 and 222

5. Boric Acid Transfer Pumps3

6. Boric Acid Filterl

7. Seal Water Filterl

8. Seal Water Injection Filter 21 and 221

9. No. 2 Excess Letdown Heat Exchanger (Shell Side)2

10. No. 2 Letdown Heat Exchanger (Shell Side)2

11. No. 2 Seal Water Heat Exchangerl

12. Mixed Bed Demineralizers 21 and 22

13. Deborating Demineralizers 21 and 221

14. Cation Bed Demineralizer 1

15. Boric Acid Blender2 Chemical Volume & Control-Boric Acid Recovery

1. Hold-up Tanks 21, 22 and 231

2. No. 2 Concentrates Holding Tank2

3. No. 2 Hold-Up Tank Recirc. Pump3

4. Gas Stripper Feed Pumps 21 and 223

5. Concentrates Holding Tank Transfer Pump 21 and 223 No. 2 Concentrates Filterl

7. No. 2 Ion Exchange Filterl

8. Evaporator Feed Ion Exchangers 21, 22, 23 and 241

9. Gas Stripper and Boric Acid Evaporator Packagel&2 Footnotes, 1, 2 and 3 - See Page 4 P8035/04 9

  '"c:

TABLE 3 (CONTINUED) Chemical Volume & Control - Primary Water Recovery

1. Monitor Tanks 21 and 221

2. No. 2 Primary Water Storage Tankl

3. Monitor Tank Pumps 21 and 223

4. Primary Water Make-Up Pumps

5. Primary Water Storage Tank Heating Recirc. Pump

6. No. 2Distillate Filterl

7. No. 2 Primary Water Storage Tank Heat Exchanger 2

8. Evaporator Distillate Demineralizers 21 and 221 Containment Spray

1. Spray Additive Tankl Auxiliary Feedwater

1. Auxiliary Feed Storage Tankl

2. Auxiliary Feed Pump 21 and 223

3. No. 2 Auxiliary Feedwater Storage Tank Heating Water Circulator Pump

4. No. 2 Feedwater Storage Tank Heat Exchanger2 Waste Disposal Liquid

1. Waste Monitor Tanks

2. No. 2 Reactor Coolant Drain Tank

3. No. 2 Spent Resin Storage Tank

4. Waste Monitor - Holdup Tank

5. Waste Holdup Tank 21 and 22

6. Auxiliary Building Sump Tank

7. No. 2 Reagent Tank

8. Laundry and Hot Shower Tank 21 and 22

9. No. 2 Chemical Drain Tank

10. Reactor Coolant Drain Pumps

11. Waste Monitor Tank Pumps

12. Waste Monitor Holdup Tank Pumps

13. Waste Evaporator Feed Pumps

14. No. 2 Laundry Pump

15. No. 2 Chemical Drain Tank Pump

16. Waste Disposal Filter

17. No. 2 Waste Evaporator Footnotes 1, 2 and 3 - See Page 4 P8035/04 10 Page 2

1

• TABLE 3 (CONTINUED)

Sampling

1. Volume Control Tank Sample Vessell

2. Boron Sample Tankl

3. Pressurizer Steam Sample Vessell

4. Pressurizer Liquid Sample Vessell

5. Reactor Coolant Sample Vessell

6. Steam Generator Sample Heat Exchanger 21, 22, 23 and 241

7. Steam Generator Main Steam Sample Heat Exchanger!

8. Pressurizer Steam Sample Heat Exchanger!

9. Pressurizer Liquid Sample Heat Exchan~erl

10. Reactor Coolant Sample Heat Exchanger Waste Disposal Solid

1. No. 1 Seal Water Tank

2. Evaporator Bottoms Hold-up Tank

3. Evaporator Bottoms Trans. Pumps 1 and 2

4. Evaporator Bottoms Metering Pumps 1 and 2

5. Resin Slurry Metering & Trans. Pumps 1 and 2

6. Waste Removal Pumps 1 and 2 Com:eonent Cooling:

1. Component Cooling Surge Tank2

2. Component Cooling PUmps 21, 22 and 23

3. Component Cooling Heat Exchangers2 Spent Fuel cooling

1. Spent Fuel Pit Pumps 21 and 22

2. Spent Fuel Pit Skimmer Pump

3. Refueling Water Purification Pump

4. Spent Fuel Pit Skimmer Filter

5. Spent Fuel Pit Filterl

6. Refueling Water Purification Filterl

              *7. Spent Fuel Pit Heat Exchanger!

8. No. 2 Spent Fuel Pit Demineralizerl Service Water

1. Service Water Pumps 21, 22, 23, 24, 25 and 26

2. Service Water Pump Strainers 21, 22, 23, 24, 25 and 261

3. Service Water Intake Sump Pumps Footnotes 1, 2 and 3 - See Page 4 P8035/04 11 Page 3 L

Footnote #1 - Designed to 1968 Edition ASME Section III Class C - Classified Nuclear Class 3 in accordance with NRC Regulatory Guide 1.26. Footnote #2 - Designed to 1968 Edition ASME Section VIII - Classified Nuclear Class 3 in accordance with the 1970 Winter Addenda of ASME Section III, and NRC Regulatory Guide 1.26. Footnote #3 - Designed to the 1968 ASME Pump and Valve Code-Classif ied Nuclear Class 3 in accordance with NRC Regulatory Guide 1.26. PB035/04 12 Page 4}}