Text

Response to NRC Request for Additional Information for Review of Proposed Inservice Inspection & Testing Programs
	ML19031A096
Person / Time
Site:	Salem
Issue date:	10/11/1977
From:	; Public Service Electric & Gas Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML19031A096 (109)
	v • d • e

RESPONSE TO NRC 6/Z8/77 REQUEST FOR ADDL INFO ON INSERVICE INSPECTION PROGRAM: W/A'ITACHED ENCIDSURES 1thru5 ....... IXJCKET NO. 50-272 RECEIVED WI'TII LETIER DATED 10/11/77 ACCESSION # 773000214 .

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RECORDS FACILITY BRANCH

ENCLOSURE 1

.. .- ADDITIONAL INFORMATION FOR REVIEW OF PROPOSED INSERVICE INSPECTION AND TESTING PROGRAMS SALEM NUCLEAR GENERATING STATION I. INSERVICE INSPECTION A. Examination categories missing from '!'able 1 (cfass 1 examinations} were omitted only because there are no components or examination areas in Salem Unit No. 1 which fall into these categories. Table 1 has been revised to include these categories with an appropriate notation and the revised. table is included in Attachment 1.

B. Omissions in Table 2 (Class 2 examinations) had the same basis as Table 1. ~able 2 has been revised in a like manner and included in Attachment 1.

C. Many components in Table 3 (Class 3 examinations} were incorrectly classified due to a misinterpretation of NRC Regulatory Guide *i. 26. These components have been designated as Class 2 for inspection purposes and moved from Table 3 to Table 2. Furthermore, some new com-ponents have been added to the revised Table 3 which is included in Attachment 1.

II. EXCEPTION TO SECTION XI REQUIREMENTS A. General

1. Examination of bolting under categories B-G-1 and B-G-2 will be performed in accordance with Code

requlrements. Since the 1974 Edition of Section IX does not require bolting in these categories to be removed for examination purposes only, examina-tion will be done with the bolts in place wherever possible. All bolts in Category B-G-1, except three bolts in each of the four reactor coolant pumps, will be examined ultrasonically in place during the inspection interval. The three 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 examina-tion reveals indications of distress, the bolts will be removed for surface and/or volumetric examination .

.e

ENCLOSURE 1 (CONTINUED)

2. Pumps and valves will be disassembled for visual examination of internal pressure boundary surf aces as required by Section XI if disassembly does not become necessary for other reasons.

3. 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 re-solved 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 Attach-ment 2, prepared by Southwest Research Institute as agents for PSE&G.

4. Ultrasonic examination of ferritic piping will be performed 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 3 above.

B. Nuclear Class 1 Components

1. Inaccessibility for examination was established during the Preservice Inspection and has been de-scribed in-Tables 1 and 2 of the Salem No. 1 Preoperational Baseline Examination Report. Those items which could not be examined due to inaccess-ibility have been extracted from these tables and compiled into an abbreviated table which has been included as Attachment 3 for convenient reference.

Comments in these tables are necessarily brief.

Information in greater detail has been provided in the way of additional comments, also included as part of Attachment 3. Consideration of alternate methods are discussed and in some cases sketches are attached for greater clarity.

2. Welds having limited 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 pre-pared at such time as the need arises in order to take advantage of the possibility of more advanced technology at that time.

ENCLOSURE 1 (CONTINUED)

C. Nuclear Class 2 Components

1. It is the position of PSE&G that tests designated in Section XI as hydrostatic pressure tests are not true hydrostatic tests in the usual connotation of that term in that they do not provide proof testing of systems already tested at 1.25 P0 and higher, but are merely leak tests. It had been the in-tention of PSE&G to limit all such tests to 1.10 Pn to avoid functional problems that might be associated with higher pressures, to provide con-sistency among all three classes of systems, and to provide consistency with anticipated changes in future Code editions. However, the anticipated changes, as presently voted by the ASME Code Committees, do not provide sufficient relief to warrant a departure from existing Code rules, and the Salem Plan will be revised to show a test pressure of 1.25 P0 for Class 2 systems.

2. Surface examination will be used to augment ultra-sonic examination whenever Section XI requirements cannot be fully complied with, such as on welds in Category C-F where UT examinations cannot be performed from either side of the weld.

III. INSERVICE TESTING OF PUMPS A. Statements made in Enclosure 2 of the 2/28/77 submittal regarding the applicable Code edition and addenda for pump testing were written to conform with the Salem Technical Specifications. These, in turn, were written to conform with the Standardized Technical Specifications submitted by the NRC, which in themselves were apparently in error. The Inservice Testing Program for Pumps is being revised to show conformance to the 1974 Edition and Addenda through the summer 1975.

B. Pump bearing temperature is included as a parameter to be measured in the Salem pump testing program as stated in the fourth paragraph on page 1 of pump testing program (Enclosure 2 of the 2/28/77 submittal).

C. The transducer type electronic flow measuring devices described in the Pump Testing Program for measuring pumped fluid flow rate have proven successful and al-ternate means of testing will not be necessary.

ENCLOSURE 1 (CONTINUED)

IV. INSERVICE TESTING VALVES A. Category A, B. (and C) Valves

1. Reasons for not full or part-stroke exercising Category C check valves identified by note (1) in enclosure 3 of the 2/28/77 submittal are described in detail in Attachment 4.

2. Reasons for exercise frequency of Category C valves identified by note (2) in enclosure 3 are also described in Attachment 4. In addition, some ex-ceptions indicated by note (1) have been deleted and enclosure 3 revised accordingly. The revised enclosure 3 is included with Attachment 4.

BB/LL:mlr 10/17/77 P77 79 16/19

Attachment l TABLE 1 CLASS l COMPONENTS, PARTS AND METHODS OF EXAMINATION Examination Category Item Table No. IWB-2500 Components and Parts to be Examined Method Reactor Vessel '

Bl.l B-A Longitudinal and circumferential shell welds in core region Volumet:ric Bl. 2 B-B r.Ongitudinal* and circumferential welds in shell (oth~r than Volumetric those of Category B-A and B-C) and meridional and circumferential seam welds in bottom head and closure Bl.3 B-C Vessel-to-flange and head-to-flange circumferential welds Volumetric Bl. 4 B-D Primary nozzle-to-vessel welds and nozzle inside radiused section Volumetric Bl. 5 B-E Vessel penetrations, including control rod drive and instrumentation Visual (IWA-5000) penetrations. *

Bl.6 B-F Nozzle-to-safe end welds Volumetric and Surface Bl. 7 B-G-1 Closure studs, in place Volumetric Bl.B B-G-1 Closure studs with nuts, when removed Volumetric and Surface Bl.9 B-G-1 Ligaments between threaded stud holes Volumetric Bl.10 B-G-1 Closure washers, bushings Visual Bl.11 .B-G-2 No components within this category B-H No components within this category

-Bl.l~

Bl.13 B-I-1 Closure Head cladding 1) Visual and surface, or

2) Volumetric Bl.14 B-I-1 Vessel Cladding Visual Bl.15 B-N-I Vessel Interior Visual Bl.16 B-N-2 Interior attachments and core support structures Visual Bl.17 B-N-3 Core-support structures Visual Bl.18 B-0 Control rod dirve housings Volumetric Bl.i9 B-P Exempted components Visual (IWA-5000)

Pressurizer '

B2.l B-B Longitudinal and circumferential welds Volumetric B2.2 B-D Nozzle-to-vessel radiused section Volumetric B2.3 B-E Heatar penetrations Visual (IWA-5000)

B2.4 B-F Nozzle-to-safe end welds Volumetric and surface B2.5 B-G-1 No components within this category B2. 6 B-G-1 No components within this category B2.7 B-G-1. No components within this category B2. 8 B-H* No components within this category B2.9 B-I-2 Vessel cladding Visual B2.10 B-P Exempted components Visual (IWl\-5000)

B2.ll B-G-2 Pressure-retaining bolting Visual

.eP77 71 40/43

Attachment 1 TABLE 1 CONTINUED Examination Category Item Table No. IWB-2500 Components and Parts *to be Examined Method Steam Generators B3.l B-B Longitudinal. and circumferential welds, including tube sheet-to-head Volumetric or shell welds on the primary side B3. 2 B-D Nozzle-to-head welds and nozzle inside radiused section on the Volumetric primary side B3. 3 B-F Nozzle-to-safe end welds Volumetric and surface B3.4 B-G-1 No components within this category 83.5 B-G-1 No components within this category 83.6 B-G-1 No components within this category

83. 7 B-G-1 No components within this catego(y B3.S B-I-2 Vessel Cladding Visual B3.9 B-P Exempted components Visual (IWA-5000) 83.10 B-G-2 Pressure-retaining bolting Visual Tubing Eddy current e Vessel Cladding (Steam Generator il4)

Piping Pressure Boundary Volumetri_c B4.l 8-F Safe-end to piping welds and safe~end in branch piping welds Volumetric and surface B4. 2 B-G-1 No components within this category B4.3 B-G-1 No components within this category B4.4 8-G-l No components within this category B4.5 B-J Circumferential and longitudinal pipe welds Volumetric B4.6 a.-J Branch pipe connection welds exceeding si'x in. ' diameter Volumetric B4.7 B-J Branch pipe connection welds six in. diameter and smaller Sur face B4. 8 B-J Socket_ welds Surface B4.9 B-K-1 Integrally welded supports Volumetric 84.10 B-K-2 Support components Visual 04:11 B-P Exempted components Visual (IWA-5000)

B4.12 B-G-2 Pressure-retaining bolting Visual

,\

Reactor Coolant Pumps B5.l B-G-1 Pressure-retaining bolts and studs, in place Volumetric*

BS. 2 8-G-l Pressure-retaining bolts and studs, when removed Volumetric and surface BS. 3 B-G-1 Pressure-retaining bolting Visual e

BS. 4 B-K-1 Integrally-welded supports Volumetric P77 71 44/47

Attachment 1 TABLE l CONTINUED Examination Category Item Table No. IWB-2SOO Components and Parts to be Examined Method BS.S B-K-'2 Support components Visual' BS. 6 B-L-1 Pump. casing welds Volumetric BS. 7 B-L-2 Pump casing Visual BS.8 B-P Exempted components Visual (IWA-SOOO)

BS.9 B-G-2 No components within this category Pump Flywheel Volumetric and surface Valves 06.l B-G-1 No components within this category B6.2 8-G-l No components within this eategory 86.3 8-G-l No components within this category 86.4 8-K-l No components within this category 36.S B-K-2 Support components Visual

-B6.7* B-M-2 Valv~ bodies Visual 06.8 B-P Exempted components Visual (IWA-SOOO) 06.9 B-G-2 Pressure-retaining bolting Visual P77 71 48/49

~

ATTACHMENT l TABLE 2 CLASS 2 COMPONENTS, PARTS,* ANO ME:'rHODS OF EXAMINATION Examination Category Item Table No. IWC-2520 Components and Parts to be Examined Method Pressure Vessels 1 Cl.l C-A Circumferential butt *welds Volumetric Cl.2 C-B Nozzle-to-vessel welds Volumetric Cl.3 c-c Integrally-welded. supports Surface Cl.4 c-o Pressure-retaining bolting Visual and.either surface or volumetric Piping C2.l .. C-F,C-G Circumferential butt welds Volumetric C2.2 c:..F, C-G Longitudinal weld joints in fittings . Vol umetr le C2.3 C-F, C-G Branch pipe-to-pipe weld joints Volumetric C2.4 c-o Pressure-retaining bolting *visual and either surface or volumetric C2.5 C-E-1 Integ~ally-welded supports Surface C2.6 C-E-2 Support components Visual e C3.l" C3.2 C-F, C-G c-o Pumps l Pump casing welds Pressure-retaining bolting Volumetric Visual and either surface or vol umetr le C3.3 C-E-1 Integrally-welded supports Surface C3.4 C-E-2 Support components Visual Valves C4*.l C-F, C-G No components within this.category C4.2 C.-D Pressure-retaining bolting Visual and either surface or volumetr le C4.3 C-E-1 Integrally-welded supports Surface C4.4 C-E-2 Support components Visual 1 CQmponents subject to examination:

Charg.ing Safety Injection Pumps 11, 12, and 13 No. 1 Reactor Coolant Filter No. 1 Excess Letdown Heat Exchanger (Tube Side)

No. 1 Regenerative Heat Exchanger No. 1 Letdown Heat Exchanger (Tube Side)

Accumulators 11, 12, 13, and 14 Boron Injection Tank Refueling Water Tank Safety Injection Pumps ll and 12 RHR Heat Exchangers ll and 12 RHR Pumps 11 and 12 Chemical Volume and Control Tank Head Tanks 11, 12, 13 and 14 Refueling water Storage Tank Heat Exchanger Refueling water Storage Tank Heating Water Recirc. Pump Containment Spray Pumps 11 and 12 Steam Generators 11, 12, 13 and 14 ~hell Side)

LL:jk P77 71 50/53

ATTACHMENT 1 TABLE 3 CLASS 3 COMPONENTS Examination In Accordance With IWD-2400 (Visual)

Reactor Coolant System Pr essu-r izer Relief Tank Chilled Water System

1. Chillers #11, 12 and 13

2. Chilled Water Strainers

3. No. 1 Expansion Tank

4. Chilled Water Pumps Chemical Volume & Control - Operations

1. Resin Fill Tank2

2. No. 1 Chemical Addition Tank2

3. No. l Boric Acid Batching Tank2

4. Boric Acid Tanks 11 and 122

5. Boric Acid Transfer Pumps3

6. Boric Acid Filterl

7. Seal Water Filterl

8. Seal Water Injection Filter 11 and 121

9. No. 1 Excess Letdown Heat Exchanger {Shell Side)2

10. No. 1 Letdown Heat Exch~nger {Shell Side)2

11. No. 1 Seal Water Heat Exchangerl

12. Mixed Bed Demineralizers 11 and 12

13. Deborating Demineralizers 11 and 121

14. Cation Bed Demineralizerl

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

l. Hold-up Tanks 11, 12 and 131

2. No. 1 Concentrates Holding Tank2

3. No. 1 Hold-Up Tank Recir. Pump3

4. Gas Stripper Feed Pumps 11 and 123

5. Concentrates Holding Tank Transfer Pump 11 and 123

6. No. 1 Concentrates Filterl *

7. No._l Ion Exchange Filterl

8. Evaporator Feed Ion Exchangers 11, 12, 13 and 141

9. Gas Stripper and Boric Acid Evaporator Packagel&2 Footnotes 1, 2 and 3 - See Page 4 P77 71 55

ATTACHMENT 1 TABLE 3 {CONTINUED)

Chemical Volume & Contrpl - Primary Water Recovery

1. Monitor Tanks 11 and 121

2. No. 1 Primary Water Storage Tankl

3. Monitor Tank Pumps 11 and 123

4. Primary Water Make-Up Pumps

5. Primary_ Water Storage Tank Heating Recirc. Pump

6. No. 1 Distillate Filterl

7. No. 1 Primary Water Storage Tank Heat Exchanger2

8. Evaporator Distillate Demineralizers 11 and 121 Containment Seray .

1. Spray Additive Tankl Auxiliary Feedwater

1. Auxiliary Feed Storage Tankl

2. Auxiliary Feed Pump 11 and 123

3. No. 1 Auxiliary Feedwater Storage Tank Heating Water Circulator Pump

4. No. 1 Feedwater Storage Tank Heat Exchanger2 Waste Disposal Liquid

1. Waste Monitor Tanks

2. No. 1 Reactor Coolant Drain Tank

3. No. 1 Spent Resin Storage Tank

4. Waste Monitor - Holdup Tank

5. Waste Holdup Tank 11 and 12

6. Auxiliary Building Sump Tank

7. No. 1 Reagent Tank

8. Laundry and Hot Shower Tank 11 and 12

9. No. 1 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. 1 Laundry Pump

15. No. 1 Chemical Drain Tank Pump

16. Waste Disposal Filter

17. No. 1 Waste Evaporator Footnotes 1, 2 and 3 - See Page 4 P77 71 56 Page 2

ATTACHMENT 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 11, 12, 13 and 13 and 141

7. Steam Generator Main Steam Sample Heat Exchanger! *

8. Pressurizer Steam Sample Heat Exchanger!

9. Pressurizer Liquid Sample Heat Exchange~!

10. Reactor Coolant Sample Heat Exchangerl Waste Disposal Solid

1. No. 1 Seal Water Tank

2. Evaporator Bottoms Hold-up Tank

3. Evaporator Bottoms Trans. Pump 1 and 2

4. Evaporator Bottoms Metering Pump 1 and 2

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

6. Waste Removal Pump 1 and 2 Component Cooling

1. Component Cooling Surge Tank2

2. Component Cooling Pumps 11, 12 and 13

3. Component Cooling Heat Exchangers2 Spent Fuel Cooling

1. Spent Fuel Pit Pumps 11 and 12

2. Spent Fuel Pit Skimmer Pump

3. Refueling Water Purification Pump

4. Spent Fuel Pit Skimmer Filter

5. Spent Fuel Pit Filter!

6. Refueling Water Purification Filter!

7. Spent Fuel Pit Heat Exchanger!

8. No. 1 Spent Fuel Pit Demineralizerl Service Water l*. Service Water *Pump*s* 11, 12, 13,- 14, 15 *and 16

2. Service Water Pump Strainers 11, 12, 13, 14, 15 and 161

3. Service Water Intake Sump Pumps Footnotes 1, 2 and 3 - See Page 4 P77 71 57 Page 3

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 -

Classified Nuclear Class 3 in accordance with NRC Regulatory Guide 1.26.

P77 71 58 Page 4

\

ATTACHMENT 2

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 I

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 following considera-tions on these two closely related questions, taking them in order:

A.~ The impracticality of recording/evaluation at the 20% reference level.

L The welded joints in nuclear p1p1ng frequently contain code-allowable wall thickness differences (12-1/2% of thickness) as well as allowable weld dropthrough and other conditions 9* 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 re-ference level (DAC) (see attached graph). Weld metal in stainless steel piping contains, in addition, reflectors due to metallurgical grain structure which can also produce indi-cations greater than 20% DAC. It appears that the incidence of geometric reflectors increases exponentially as the ampli~

tude 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%

a~d 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" adjacerif"r"oot *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 Attachment 2 Page 1 of 6

20% data would require that the personnel spend several times as much time in a radiation area to obtain additional ultra-sonic -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 0because of the inavailability of trained per-sonn0er: - The reauction or sampTe si0ie would ehave 0a 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 exami-nations on one or both sides 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 indications were recorded with the 60-degree examination in 5/8-inch of weld length while 10 indications were recorded with the 45-dgree examination in 1-7/8 inches of weld length. Only maximum amplitude positions were recorded and most indications were found in the 20-28%

DAC range. All indications of 20% DAC and greater were re-corded. 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 50% DAC recording and 100% evaluation criteria and several months of successful- operation without leaking confirmed that these examinations, like those per-formed at several other sites, were effective

B. Evaluation at 100% and greater provides equivalent sa:!;ety.

1. Equivafent safety-, - comparable to coCle* *:requirements*,- is assured by the recording/evaluation criteria developed, refined, and

It must be noted that the 100% evaluation criteria was (and is) the Code requirement for utilities committed 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 1976 Edition (IWA-2232) the ASME has reconfirmed its previous position by clearly requiring evalua-tion of only indications of 100% DAC or greater.

qualified by SwRI through many years of research and experi-ence. 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 investi-gated by a Level II or Level III operator to the extent necessary to determine the shape, i4entity, and location or the reflectors.

(c) Any indication 20% of DAC or greater and suspected by the operator to be other than geometric in nature, in-

. eluding 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 investigated and found to be other than geometric in nature shall be reported to the owner for evaluation artd 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 reproducibility of test data. We have shown data repro-ducibility to be a factor caused by many things including operator experience, training, procedure, equipment variations, environmental encumberances, test piece conditions, calibra-tion standards, etc. These facators are routine and will continue to occur during the application of examinations of this *nature on piping. This practice is believed 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:

(iP-) 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, th~ 100% reference level --eva:luation-criteria**of* Paragraph IS-213.S 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 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.

Page 3 of 6

(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 experi-enced leaking by way of a stress-corrosion crack between the periodic examinations.

(2) At least 48 piping cracks have been found and.re-paired in the early stages of propagation.

(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 indications 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 exam-ination 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 recommends to its clients that,

. in the interests of maintaining maximum nuclear power plant integrity and safety at minimum cost in time and personnel radiation exposure, any effort to institute a blanket 20% DAC recording/evaluation criteria Attachment 2 Page 4 of 6

be resisted. Instead, SwRI recommends a commitment to the SwRI recording/

evaluation practice which was set out in Paragraph B.l above and is re-interated below:

(a) All indications 50% of DAC or greater shall be recorded.

(b) All indications 100% of DAC or greater shall be investi-gated 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 deter-mine the shape, identity, and location of the reflector.

(d) Any indications investigated and found to be other than geometric in nature shall be reported to the owner for evaluation and disposition.

Pag~ 5 of 6

4-500 43C>o BASfD ON A TYPICAL ULTR.A.SONlC 4-100 'SURVEY Of:. EX.AM 11-JA..TlO~ R.ESULT5

~,i,,oo Le\/EL IL f0f2_ CLA '5 S PJPJNG, 3100 E"XAl'w'\11\Jf'R.<;

' )

35Do 5" To \0 TIMES 3300 3100 ks MANY 2'}00 INDICATIONS 27l>o WILL BE lL 2500 0 FovtJ[) IN THE 1~00 aJ.. 20'% To So'?.,.*

i.100 UJ (l) 1~00 'RAtJ9-E.THAN 2 1100 THE' .TOTAL

~

1500 Nvtl\BER. OF-2 1300 I"lDI GAT10r.J5l 1100 ABoVE 506lc, 900 100 SDO 300 100.l----.---w.~---1--l--l.--l--ll--l-J--l---l--l--ll--.l---l--l--l===t--ll==::t:=:i==1=:=1"--.....__---.---.-___;_

o* i.o* 40*

coo so 100 1z.o 150 200 2.50 310 400 .500 ~30 7'30 1000 PEl2.C£NT OF DAC

. .~ No RECOJ2.DED DATA AYAILA.BLE WI' Fog_ TABULATION,

~t-t-richment 2 e

Page 6 of 6

ATTACHMENT 3 WELDS EXAMINED DURING THE PRESERVICE INSPECTION OF SALEM 1 WHERE THE COMPLETE EXAMINATION AS PRACTICABLE COULD NOT SATISFY THE NORMAL CODE REQUIREMENTS (Extracted from. ~he Final Report)

Prepared for:

Public Service Electric and Gas Company 60 Park Place Newark, New Jersey 07101 August 1977

- SOUTHWEST SAN ANTONIO RESEARCH CORPUS CHRISTI INSTITUTE HOUSTON

SALEM NUCLEAR GENERATING STATION UNIT*l OATEI 08/03111

SUMMARY

Of NONDESTRUCTIVE EXAMINATIONS PAGI! I 001 0 CLASS l COMPONENTS

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SUMMARY

Of NONDESTRUCTIVE EXAMINATIONS PAGEi 001!

CLASS l COMPONENTS i

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~~SUMMARY~~
O' NONPESTR E EXAMINATIONS PAGEi Ol' ~.
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~~SUMMARY~~
Of NONDESTRU E EXAMINATIONS PAGEa DIS ~
0 CLASS 1 COMP ENT!
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~~SUMMARY~~
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~~SUMMARY~~
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~~SUMMARY~~
OF NONOESTRUCTIVf EXAMINATIONS PAG[ I oeo v CLASS 1 COMPONENTS 1 _: ~~~~rt>~.c~oL~~t.hatEM,,i.tNf~6~ '~ilc~uo CiitE f'iG~ C'!'~s>--~ -~~:~:--_-_-_ :I
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~~SUMMARY~~
Of NONDESTRUCTIVE !KAMINATIONS
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~~SUMMARY~~
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Of NONDESTRUCTIVE EXAMJNATION9 PAGE I Oi!'t CLASS ! COMPONENTS I PE~~t~~ ~~o~*NT.avstEM1 Li~E No,_e-~c-11~i (SEE~.~. c-~71 - I c* :I I. .. ~S'°'E. ~$ME .I '
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~~SALEM NU~LEAA GENE~NG~~
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~~SUMMARY~~
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SALEM NUCLEAR GENERATING STATION UNIT*l DUEi DBIDJ/71
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~~SUMMARY~~
Of NONDESTRUCTIVE fllAMJNATlONS STATION UNIT*l DAU.I QB/091, PAGEi OJlt
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~~SUMMARY~~
Of NONDESTRUCTIVE l!XAMINATIONI PAGEi 035 CLASS l COMPONENTS
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~~OALEM NUCLEAR GENERATING STATION UNIT*l~~
~~SUMMARY~~
~~Of NONDESTRUCTIVE EXAMINATIONS CLASS 1 COMPONENTS 91 DATEI 08/DJ/11 PAGEi DJb~~
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~~SUMMARY~~
OF NONDESTRUCTIVE EXAMINATIONS DATEI 08/0JIH PAGEi DH 0 CLASS 1 COHPONfNTl I . - SAfUY INJECJIOtii 1JYSfEl4~ LINE NO., 1i*aJ*UJ1 (8U:: flG~ c!ia}j - :.1 T.
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. .. *- ..--*---------- __ ,, ______________ _. ............... __ .............. THE.ELBOWeNO .. UTOW .. DU~ JO TH! .. ;I
. ---*-.. --------------------*---------- .. -------------.............. WHD CAOl'IN CONf-'IOURUJO"*
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Of NONDf.STRUCTJVf. [XAHl~ATIONS PAGEi O:H c CLASS l COMPONENTS I ~~ftH JNJ~ctJOtj ov*r~M. Ll~E-~Q~ ... *8J*lUi COEE flG~.C~bB)~* -- .I I
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~~SUMMARY~~
Of NONDESTRUCTIVE EXAMINATIONS CLASS l COMPONENTS OATEI OB/OJI.
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~~SUMMARY~~
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SUM~ARY Of NONDESTRUC~ El<AMINATIONS PAGEi D'tl -
CLASS 1 COMPONENTS L SAFETY .INJ!CitO~ ~Y$TEM, LINE N0 1 "belJ*lll1 (SEE* P10 1 .C*11J
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I f~.-- ___ , *__ : . : _ . *_ _. : . _. _.. -~~ ::*-~-~~:::~~* .-~~*:_ :_ :*::*-_ .*: ~-=*: :* .. :*: ~~-~. *:::-* . ------:_::_~::_:~--.:-=-:-::: ~::*:=:::::::-.-=.*_-:-_-_*-~--=-.===:=-~=-~==--~:-*. ___ :*---~*-.:_._=*:-:: ~ . : _ ~:. _: :-~-~*-: ~--* -~::*--. :* : -* -.: .. . 11 f C~ .~ _---** .~'!"~ _ . H!MS~! ~U !~P.L~'----* _______ ----*-** .. ___ ----*---*-*-_MI_ _____._'."_________.1!!,UL~ ..'!_-~---J!iJ!I_ ~uq __~E~P ~$- -~~c~U-JBLL ___ *-*-** .. .. .. ll]
___ .. _ . . . . . _ P.IPE LUG _ ... **---*-- _ _ **-**- .. __ ....... __ ---* ....... -***-- ___ .*-* _ **-**-*--- --*-- --* . _____ --****--*- -*----- __ IHIE. TO TH!; PENETR.lTION 1 _ .... ***-*-* -** ..... _ ,.....
r-:*--: ~ .* . -- -*:--* ::* -* . -*- .-.*. * *.: _:*~==*:-~ :* ~--_ ~-.--:* *--~-=* :::--=-:.:_*:.::*::.::::-:-:=---~:-=:.-::-=-::*:::*:-~:~-::*.==*:=~:-:*=*:_~:::-~-::-.:_-:_:=*=-~~-=--*.:- .: *:. -~-* :::::-:* .: -*. _* __ . _: _: .-. :-.:_:***::: :=*:.: :* .:.. Tl l 1.*.-.*
~~~~c*_* I!* * ~.. *.-.-.. **.**.~:!; H!~*~I! 1 ll~l!P.L-..IL __ . ----*-- **- - ---*-. - - *-. -*-* .... *-- ML. ______ JOO!".l/.l ___ , '~'~'l} __l{_~~~~~
_-_*... PIPE __'!' __! ______ ~E--!41 .f!rnM...~t!~_.U~tJT~~~,,_ HOE. _______ *-****-- *-----11 LUO .... -*- **---*-. _ ...... *- __ ... -*-**-**-*-*--- .... *-*-*--***-*-*- ***-***-----* ----*-****- .D !.TO THE .. !NETAATION, . _ ... -*-. .. 1.
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c*--~ ~ .*.* _* .. . - . . . . . ... - ... *-*** - . . . .. .. -
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DUE TO.THE.PENETRATION,_ . *-:-1 I __ *_ .. .~T .. REVEALfD . A.LINEAR.INDICA111 I
.... :.1 TION 1 CLEARED BY PS£10 ~ITH r *-**. _CNP' nu, ...
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~~~~SUMMARY~~~~
Of NONDESTRUCTIVE EXAMINATIONS OATEI DB/Dl/11~
PAGEi OS&
C) CLASS I COMPONENTS r-- - ~~I~ STUM OVSW?l1 LINE NO, ,,-~:"1s~1iii CBlf FIO, O*l U: _: :-- - -- -- -:-__ :~:-: - ;I f -- *- -!'~~~~~!'* .. *********-~-*-*****9:!'
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~~~~SUMMARY~~~~
- _R ... 8 l.H ____________ _ .Tl
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1 PIPE LUG ------*--- -------------- -* -*-- ---*---- oue:_TOTH!.PENETRUIQN, -- ~ I I -- :_ -- '.J I~
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~~~~~~SUMMARY~~~~~~
o~ NONDESTRUCTIVE !XAMJNATIONS CLASS I COMPONENTS DATE* 08/0J/11 PAGEi O!lb * ~ I 1
1* ~----~-*-**~~-*-**~***********-*******~*~-*~-~~~----~-~- .

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f. . ~si:it . ~SME.. .. . ... .. . .. . ..... _ . __ llWRI... IJU~f'IARY . R I E. IL........ . :I I
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I IT~M No CHOy_ .... E~-Ml~ATICIN.. Af![~_.o!NTlflCATJON Mt:Tt.too____ NOe/REV1 -~Uf!IM8_ C.G ~ R ____ 8E~ARK$ _____ ... 11
~~~~-!~ .. *!~-*~~.~~~~---~-~~~~~**~***~~**~**~***~~~~~~~~-~~~~---, ~~-~**~~~--~-~*~!~--- ~ ~-~-~-------~~*~~-~~*~~--~-~~~~,~~~~~~~~--~----~
I C*C Ji!*MS*l!li!l*IPL*l . :*-.au:. RMKIS -~ . . . IJOlOD .X... * * . . NO "IT our: ..TO. IN.ACCUSIBILITY iI P.lfE LUG -* -*-- -----*-*----------*** *----*----- --------* .. _ .. _______ .CAUSE BY PENETRATION, I r I I. Cl!. 't ie . . M6-21il*i!!PL;i_ . _ ..... ...... :......... -~i-lf.- :.~.:~-- :. iiiii~ii1::.._-:-_iiii1iiii.~_s_:~:~*' -_:-*No. Mt.i'aoi.i .. tHe:. uestR£At4 sii>E
. PIP.E LUO --*------- ... -*- __ **-***----. ______ *--------------------------*--------- ______ . __________ DUE . TO .THE PENETRATION, l-
'. 1_: . C:!., . C*C - U!~8*21i 1~'i!h~.----~_.:: _______ -* . *" .*::::: _:*:_-_-:_-:_:~::_ 1;1y_ ::*:-:::_-~-::J"iiii;ii?.~=~-:**!ii'ijJC *:::x !!-::..~::!.*.-~- :::No_ ijf_f.it6!rj~E. UP~T~tA~. SIDE .
-- .. - - - p IH LUG - ----- .. --- --- - .. -* - - ... -- ---*---- ... ----- --- --- *-- ..... - ------- - *------- ...... *-- ______ ... DVf TO TH!: .. PENnRATION,
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..... PIPE .LUO -----.*- ______________ -* _------*-----*--------- ---------------------------*-*----------*- .P!JE .. .JQ. lHE .PfNETRATION, ... _ ....... .
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c_ - ------- -* _ . Pl P.!: LUG. _ -------------* __ _____ _ __ ... ----------------*--------------------*-----------*DUE ..TO.. lHLPENETRA TION, ..... :l f - Ci!* 't. CliC *--- ]e~Mi.;;ei"zi,;2*ii[;-1:*:.*:-:*------ .. -~ -_.__ .. __ - ~.::*_~--=~~-_jlj-=~==~=ioii;;i-/f~_:_-iflo~eo::-::T.:~ .. ~-,;--_-*:-.ND.MT: i9oi4:'JHE UPST~f:AM iii>~ . - - . ** !I c-*. ~~~::_-_-::.*:
.. _____ PIP.E LUG --: --~=---~-~:--** ...... _. ____ -*-*----- --~--~:~~-- ________ ---* --------=~=-=:=-~-::*_-:=*--:_DUE TO THE .P.ENETRAllO!'I, .... _ ----* ---- --- .. i*I
!&*M8*!1!1*2PL*~-- ---* -~ -MJ. *:*: - - . ioo.;iij _._:* i1ottiL_-- ~ *:;.; ; -~---- NO ,.., . ;~oM. THE UPSJ~EAM. sioi! 1*1 l' - . PIP.E LUO -------* ... . _____ ---*---- ------ __ ___ . -------- ... *----- __ . ________ ... DUE TO THE .LUG .CONFIGURATION,.
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1** .. P.IPE LUO DUE.TD THE.PENETRATION, .. _ND .. HT iI IN THE COUNTERCLOCKWISE OlREC*
1_*- TION DUE TO THE PIPE OBSTRUC*
tlONe. I 1* Cl!,'t u .. Ml*lll ! l*i!PL~-l l PIPE LllG MT i!JDUi )( .*
~~~~~~NO MT FROM THE UPSTREAM SIDE DUE TO THE PENETRATION, I~~~~~~
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SALEM NUCLEAR GENERATING STATION UNIT*l DATEI 08/0J/1.
~~~~~~SUMMARY~~~~~~
OF NONDESTRUCTIVE EXAMINATIONS PAGEi 09?
CLASS Z COMPONENTS f- MAIN STEAM $~ST~~. LJN!.NO, Jl*MS~21Zl (SEE flG, D*ll) .. j*1 1-
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Ill!.,. ..... . . . . . . . . . . . . .,., . . . . . . . . . wCJ**'l~~******~ll!l'!'Wllllllip ...... ~P~!"* .. 11!1'~4!-P"**~ *"!!~~*""'~---*'!I 'II!. !l!I! .4!! ..... _ ******~~-"********~--_,.**.******~*****~.
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I PIH LUO --* DUE TD 1Ht PENETRATION, iI
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[' PIPE LUO STREAM .. OR.DO~NSTREA~_fNDS OF :I THE LUQ_DUE 10 THE PIPE .RE*
r-:-.- - : URAJNTS 1 : I c.c Ji?wM8*!lil*~PL*I - MT. _ .... Joo*1n. __ uoeoo __,. x .* !II * .NO.EXAHiNAtiDN.~ffO~ ~~E-~~~. :I PIPE LUG STREAH .. OR DOWNSTREAM.END~.OF.
THE LUO DUE TO THE* PIPE RE* 11 ST RUNT I -- I I
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PIPE LUG . STR~AM.OR OOWNSlREAM.~NO&_Df 11 I
.. THE LUO OUE TO.THE PIPE RE* :1
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... - ..... __ . WE.LUO DUE TO THE PIPE RE* !I I
.. -- _ .. --*--*** ... _. *-**-- ... *-*-*-*-*---* . STRAit-IT1.
l:I c.. c !12*Ml*2lll**PL*S. ... MT ___________ JOQ~l/1 ..... jjiiiJD:___ ~_- .K .* **....-. ~O . t:XAMi~~j.JON fRP~. i~~ VP! .. I PlPI! LUO STREAM.OR DOWNSTRl!A~_fNPS_Of :I l THE.LUO DUE TO THE PIPE RE*
BTRAINT 1 11
..*. i Ji!eMli*i! 11 l*'tPL*lt MT. JOO*l/l iJOB~O .. X * *
ND EXAMJNATJON.fRO~ THE UP*_ I I PIP! LUG . 8TREAH DR. DDWNSTREA~ ENDS_ Of I THE LUO .DUE TO THE PIPE RE* Il STRAINT 1
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NO EXAMJNATION.fRO~ .. THE .. UP~ ... I PIPE LUG STREAM OR DOWNSTREA~.LNDS.Of :I THE LUO DUE TO THE PIPE RE*
STRAJNT 1 I1 I
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SALEM NUCLEAR GENERATING STATION UNIT*l
~~~~~~SUMMARY~~~~~~
OF NONDESTRUCTIVE EKAHJN&TIONS
~~~~OT~*~ ~y~T~M, ~l~f NO *. Jl~~a~eiil (8El ~IG~ 0~11i -
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CLASS I COMPONENTS PAGEi use
~~~~~~OATEI DBIOJ/11~~~~~~
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~~~~~~SUMMARY~~~~~~
_fl I [ H I 3ECT ~* a~at ~* . . ~fL6 ~~~~[R *~DID~ - . . .. . EKAM*~. PROCEOIJllE .. IUiU:l. . E I !U: ....
.. ITEM NO CATOy. E¥AMl~ATION.AR!~ JDiNTlflCATION METHOD ... N0 1 111EV, NUMBER . . C* OM R.. REM All Kl :I
.I Ci!, 't .~i" ~***~~- -~~.--~-~ .. ioiJ~i/"1~* ... i JDi&ij . . ---~~-- * ** * . *-_-_ . _ijQ. t.~AMIN~.T 10~ _* f AQ~ ..*ii1i . uP~~ ..
...... _ . _.. ____ -* . _ . .. . UflEAM _O~ PDWNBTlliA!L U~Da. Qf \I
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~~~~~~SUMMARY~~~~~~
Of NONDESTRUCTIVE EXAMINATIONS PAGEi osq 0 CLASS I COMPONENTS l~~-~~iN*-~it~i ~~ST~~~ LINE NO *. J~ .. Ma*alll ia~~ ~iG, ~~i~> :I f L:*~- *****-**~*-***-*********************~*-**********~******
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~~~~~~SUMMARY~~~~~~
Of NONDE8TRUCTIVE"EMAMJNATION8 PAGEi OLD
~ CLASS I COMPONENTS 1~- M~J~ *t~AM *~s*f~* Ll~E No, ~0*~*~~1~j tii~E ~-~~ o~ai.~ I I I
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JOENTlflCAllON_ ...
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' r. .. .. P.JPE LIJG ... -*- . __ .... __ ---*- -*- . . ... .. .... DUI. TO THI: PIPE COLLAR*
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SALEM NUCLEAR GENERATING STATION UNIT*l O&T£1 DBIOJ/11
~~~~~~SUMMARY~~~~~~
OF NONDESTRUCTIVE EXAMINATIONS PAGEi 01.1 CLASS ! COMPONENTS
~
RESIDUAL HE~t RfMOVAL aySTEM, .LIN[ No. ~'*RH*!le' (S(( rlG~ D*JI) I
.N I _ 0 :*1
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( ITEM NO CAT~y E~AMINATION,AR!~ iDENTlflCATION. METHOD .... NOe/RU,. NU~Btft ___ C. G ILR. *.. ft.Ei:i~RK$ I I
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I 1..* C!,l *~~--~* C*f ***RH*!le,.s SE[" RHKS ** .. 1~1111 x. ~.. THIS.WELD.IS.WITHIN THE fLOOR .. I ELBOW TO PIPE __ ... . -**-* *-***---- **- -****-** --* .. -* ___ ...... -* _ AND U INACCUUBLE1 !I T
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~~~~~~SUMMARY~~~~~~
Of NONDESTRUCTIVE EXAMINATIONS HE~t R~MOVAL sySTEM,~LiNE No~ l'l~~H~lll,_C8£E ~.~;~;1~)
~*~-~--~~-~*~*****~--*******~**~*-**-~***************~---~---~--~~~
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ATTACHMENT 3 DETAILED EXPLANATION OF EXCEPTIONS TO THE ASME BOILER AND PRESSURE VESSEL CODE SECTION XI FOR THE SALEM UNIT NO. 1 INSERVICE INSPECTION PROGRAM Item Bl.2 Category B-B Closure Head ~eel Segmeht-To-Disc Ci~cumferential 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 1-RPV-6046 B in the exam-ination 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 require removing and re-in-stalling 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 1-RPV-4043 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 instrumenta-tion tube 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 bottom head. To gain access would require removing and re-install-ing numerous instrumentation tubes.
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 27.5-1110-1, 27.5-RPV-1120-1, 27.5-RpV-1130-1 and 27.5-RPV-1140-1 in the examination plan.
ATTACHMENT 3 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 an mechanized inspection device, these welds are inaccessible at this time. Examinations cannot be performed from the outside surface 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 towards the end of the inspection interval to facilitate their examinations, 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-RPV-1110-1, 29-RPV-1120-1, 29-RPV-1130-1 and 29-RPV-1140-1 in the examination plan.
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 performed from the inside surface using an automated inspection device, these welds are inaccessible at this time. Examina-tion cannot be performed from the outside surface 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 towards 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.
ATTACHMENT 3 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 examination will be per-formed when the Core Barrel is removed during the in-spection interval. The patches of cladding to .be ex-amined are identified as l-RPV-Patch-1, l-RPV-Patch-2, l-RPV-Patch-3, l-RPV-Patch-4, l-RPV-Patch-5 and l-RPV-Patch-6 in the examination plan.
Since this examination can only be performed from *the inside surface of the reactor vessel shell, the examination can only be performed when the Core Barrel is removed. The Core Barrel is scheduled to be removed towards 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 XI. These welds are identified as 31-RC-1110-2, 31-RC-1120-2, 31-RC-1130-2, 31-RC~ll40-2, 29-RC-1110-4, '29-R C-1120-5, 29-RC-1130-5 and 29-RC-114-5 in the examination plan.
The Reactor Coolant loop elbows are made of cast stainless steel which cannot be penetrated by ultrasonic examination techniques. Also, acceptable raaiographs cannot be ob-tained 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 welds is expected to be restricted during inservice in-spection due to inaccessibility. These welds are identified as 31-RC-110-5 and 6, 31-RC-1120-5 and 6, 31-RC-1130-5 and 6, and 31-RC-1140-5 and 6 in the examination plan.
ATTACHMENT 3 Due to anti-whip restraints that were installed after the preservice examination was completed, portions of these circumferential welds are expected to become inaccessible for inservice examination. The extent of inaccessibility will be determined during the first inservice examination of these welds.
Item BS.6 Category B-L-1 Reactor Coolant Pump Casing Welds Volumetric examination will not be performed on these welds.
A surface examination will be performed as well as a visual examination for leaks required by Section XI.
These welds are identified as 11-PMP-l, 12-PMP-l, 13-PMP-l and 14-PMP-l in the examination plan.
The Reactor Coolant Pump casings are made of cast stainless steel which can not be penetrated by ultrasonic examination techniques. Also, acceptable radiographs cannot be ob-tained due to fogging of the film 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.
BB/LL:mlr 10-18-77 P77 79 24/27
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ATTACHMENT 4
~ *Inservice Testing of Category A, B, (and C) Valves The following category C check valves cannot be full or part stroke exercised during normal plant operation for the following reasons:
Safety Injection System 11-14 SJ 17 Safety Injection Charging Line Cold Leg Check Valves
~~~~a. For operational modes 1 and 2, testing would require pumping 2,000 ppm borated water into the RCS. This would render the reactor subcritical and would also violate Tech. Spec. LCO 3.5.4.1.~~~~
~~~~~~b. For operational modes 3 and 4, testing would ultimately require significant RCS dilution and boric acid recovery operation.~~~~~~
~~~~~~It would also present a possible RCS over pressurization and would violate Tech. Spec. LCO 3.5.4.1 and certain operating procedures.~~~~~~
~~~~~~11-12 SJ 34 Safety Injection Pump Discharge Check Valves~~~~~~
~~~~~~a. For operational modes 1, 2 and 3, the RCS pressure is greater than the Safety Injection Pump shut-off head.~~~~~~
~~~~~~b. For operational mode 4, testing of these check valves could result in a possible RCS overpresurization and also contradict operating procedures.~~~~~~
~~~~~~11-14 SJ 43 Residual Heat Removal Discharge.Check Valves to Cold Leg~~~~~~
~~~~~~a. For operational modes 1, 2, 3 and 4, the RCS pressure is greater than the shut-off head of the RHR pumps.~~~~~~
~~~~~~11-14 SJ 55 Accumulator Discharge Check Valves~~~~~~
~~~~~~a. For operational modes 1, 2, and 3, the RCS pressure is greater than accumulator pressure.~~~~~~
b. For operational modes 4 and 5, testing would require significant RCS dilution and boric acid recovery. There is also the possibility of RCS overpressure and a con-tradiction of operating procedures.
~~~~~~11-14 SJ 56 Safety Injection and Residual Heat Removal Discharge to Cold Legs~~~~~~
~~~~~~a. For operational modes 1, 2, and 3, the RCS pressure is greater than the pumps shut-off head.~~~~~~
~~~~~~ATTACHMENT 4 (CONTINUED)~~~~~~
~~~~~~b. For operational mode 4, there is the possibility of RCS over-pressurization and a contradiction of operating procedures.~~~~~~
~~~~~~_________7 1 SJ 70 Refueling Water Storage Tank to Residual Heat Removal Pump Check Valve~~~~~~
~~a. For operational modes 1, 2 .and 3, this system is normally not in seryice and therefore requires testing immediately prior to returning to service (ASME Cod~ Section XI Par IWV 3410- Part (f) .~~
~~~~~~11-14 SJ 139 Safety Injection Pumps Discharge to Hot Legs~~~~~~
~~~~~~a. For operational modes 1, 2 and 3, testing is not possible since RCS pressure is greater than pump shut-off head.~~~~~~
~~~~~~b. For operational model 4 and 5 there is the possibility of RCS over-pressurization and testing would violate operating procedures.~~~~~~
~~~~~~11-14 SJ 144 Safety Injection to Cold Legs~~~~~~
~~~~~~a. For operational modes 1, 2, and 3 testing is not possible since RCS pressure is greater than pump shut-off head.~~~~~~
~~~~~~b. For operational model *4 there is the possibility of RCS over-pressurization and testing would vi_olate operating procedures.~~~~~~
~~~~~~1 SJ 150 Boron Injection Tank Discharge Check Valve to Cold Legs~~~~~~
~~~~~~a. For operational modes 1 and 2, testing of this valve would render the reactor subcritical and would also violate Tech. Spec. LCO 3.5.4.1.~~~~~~
~~~~~~b. For operational modes 3 and 4, testing would ultimately require significant RCS dilution and boric acid recovery and would present a possible RCS over-pressurization.~~~~~~
~~~~~~11-14 SJ 156 Safety Injection to Hot Leg~~~~~~
~~~~~~a. For operational modes 1, 2 and 3, the RCS pressure is greater than pump shut-off head.~~~~~~
~~~~~~b. For operational modes 4 and 5, there is the possibility of RCS over-pressurization and this procedure contradicts operating procedures.~~~~~~
~-- * ......
ATTACHMENT 4 (CONTINUED)
~ COMPONENT COOLING SYSTEM lCC 137 Component Cooling Check Valve from R.C.P. Lube Oil Coolers
a. For operational modes 1-4, component cooling cannot be isolated when any Reactor Coolant Pump is operating*
~~~~~~l CC 317 No. 11 Charging Pump - Mechanical Seal Heat Exchange Inlet 1 CC 320 No. 12 Charging Pump - Mechanical Seal Heat Exchange Inlet~~~~~~
~~~~~~a. For operational modes 1-3, Tech. Spec. requirement 3.5.2 requires both pumps be operable. Isolating these valves for test places these pumps in an inoperable condition.~~~~~~
~~~~~~RESIDUAL HEAT REMOVAL~~~~~~
~~~~~~~ ll-12 RH 8 Residual Heat Removal Pumps Discharge Check Valve~~~~~~
~~~~~~a. This system is normally out of service and requires testing immediately before returning to service (ASME Code Section XI Par IWV 3410 part (f).~~~~~~
~~~~~~13.14 RH.27 Residual Heat Removal Pump Discharge to 13 and 14 Hot Leg~~~~~~
~~~~~~a. See 11-12 RH 8.~~~~~~
~~~~~~AUXILIARY FEEDWATER SYSTEM 11-13 AF 8 Auxiliary Feed Pump Discharge Check Valve to Steam Generators~~~~~~
~~~~~~a. These check valves can be tested in any operational mode except model 1. The flow required in mode 1 is to great for Auxiliary Feed Pumps to maintain SG levels.~~~~~~
~~~~~~11-14 AF 23 Auxiliary Feed Check Valve at Steam Generators same conditions as 11-13 AF 8 CONTINMENT SPRAY SYSTEM 11-12 CS 4 Containment Spray Pump Discharge Check Valve~~~~~~
~~~~~~a. Check valves cannot be tested in modes 1-4 since Tech.~~~~~~
~~~~~~Spec. 3.6.2.1 requires a specific valve line up. A test would require a deviation from that line up.~~~~~~
~~~~~~b. The test would require a discharge to the reactor cavity.~~~~~~
The cavity is dry until mode 6.
ATTACHMENT 4 {CONTINUED) 11-12 CS 21 Eductor Suction Check Valves See 11-12 CS 4 Requirements.
11-12 CS 48 Containment Spray Header Check Valves See 11-12 CS 4 Requirements.
CHEMICAL AND VOLUME CONTROL SYSTEMS 11-13 BR 152 Discharge from Volume Control Relief Tank to 11-13 Holding Tank
~~~~~~a. Test requires remo~al of VCT relief valve for temporary source connection. This test will be done duing mode 6.~~~~~~
~~~~~~1 CV 23 Let Down to Mixed Bed Demineralizer~~~~~~
~~~~~~a. This valve is located in a concrete vault to protect~~~~~~
~~~~~~personnel from the radiological hazard. This valve will be tested in operational mode 6.~~~~~~
~~~~~~1 CV 36 Demineralizer Return to Volume Control Tank~~~~~~
~~~~~~a. See 1 CV 23 Requirements.~~~~~~
~~~~~~1 CV.42 Volume Control Tank Discharge Check Valve~~~~~~
~~~~~~a. To test requires closing valves 1 CV 40 and lCV 41.~~~~~~
Alternate source of water is from Refueling Water Storage Tank through valves 1 SJ 1 or 1 SJ 2. This is 2,000 ppm boated water which would render reactor sub-critical. Testing will be done in mode 6.
1 CV 74 Charging Safety Injection Pump Discharge to Regenerate Heat Exchanger
~~~~~~a. Tested in mode 6 because of the normally high radiation level.~~~~~~
~~~~~~b. Flow through valve required during normal operation.~~~~~~
~~~~~~11-14 CV 99 Seal Flow Check Valve to Reactor Coolant Pumps~~~~~~
~~~~~~a. Flow cannot be isolated during operational modes 1-4.~~~~~~
These valves will be tested during modes 5 or 6.
1 CV 176 Boric Acid Transfer Pumps Discharge Check Valve to Rapid Borate System *
~~~~~~a. Testing in operational modes other than mode 6 would cause a loss in reactor power due to high borate injection.~~~~~~
~~~~~~ATTACHMENT 4 -s-(CONTINUED) 1 CV 196 Chemical Tank Outlet Check Valve~~~~~~
~~~~~~a. Tested in operational modes 4-6 at which mode chemicals are added. Use of primary water would cause a reactivity change by dilution.~~~~~~
~~~~~~1 CV 198 Mixed Bed Demineralizer Check Valve to Hold up Tank~~~~~~
~~~~~~a. Tested is mode 6 because of normally high radiation levels.~~~~~~
~~~~~~SERVICE WATER SYSTEM 11-13 SW 5 Turbine Generator Service Water Heater Check Valves~~~~~~
~~~~~~a. Testing requires the isolation of one header to reduce pressure. This places the plant in a degrader mode of operation. Tested in mode 6.~~~~~~
(Note: These valves added as category tl) valves to original list).
12-14 SW 5 Nuclear Area Service Water Header Check Valves A. Testing requires the isolation of one hader to reduce pressure. This violates Tech. Spec. 3.7.4.1. Testing will be done in Mode 6.
11-12 SW 77 Containment Fan Coil Unit Discharge Check Valve
~~~~~~a. Testing can be done in any mode of operation. Category (1) will be deleted from these valves.~~~~~~
~~~~~~11-12 SW 79 Service Water Overboard Discharge Check Valves~~~~~~
~~~~~~a. To test these valves will require shutting down a major portion of the service water system. Testing will be done in operational mode 6.~~~~~~
ATTACHMENT 4 (CONTINUED)
Category C Valves The Category C check valves identified below will be full-stroke exercised not more often than once every three months during normal plant opeation in accordance with Article IWV-3520 of
Section XI of the ASME code, with the exception of those valves marked (1). These valves cannot be exercised during normal plant operation and will be full-stroke exercised during Mode 5 or '6 operatibn, not more often than once every nine months. Tho~e valves marked (2) will be exercised during Mode 5 or 6 *operation when relief valve 1CV241 has been removed for testing.*
The Category C safety/relief valved identified below will be bench tested with suitable hydraulic or pneumatic equipment in accordance with Article IWV-3510 of Section XI of the ASME Code.
Those valves marked (3) will be tested in place with hydraulic or pneumatic assist equipment. Test frequencies will be deter-mined in accordance with Table WV-3510-1.
Safety Injection System Check Valves Safety/Relief Valves 1SJ3 lSJlO llSJl 7 (1) 11SJ29 12SJ1 7 (1) 12SJ29 13SJ1 7 (1) 14SJ29 14SJ1 7 (1) 14SJ29 1SJ31 1SJ32 11SJ34 (1) 11SJ39 12SJ34 (1) 12SJ39 11SJ43 (1) llSJ 48 12SJ43 (l} 12SJ48 13SJ4'3 14SJ43 11SJ55 12SJ55 13SJ55 14SJ55 11SJ56 I 12SJ56 13SJ56' 14SJ56 (l}
1SJ70 (1) 1SJ107 11SJ139 (1)
ATTACHMENT 4 (CONTINUED}
~~~~~~e Safety Injection System Check Valves Safety/Relief Valves 12SJ139 (l}~~~~~~
13SJ139 (l}
14SJ139 (l}
11SJ144 (l}
12SJ144 (l}
13SJ144 (l}
14SJ144 (l}
1SJ150 (1) 11SJ156 (l}
12SJ156 (l}
13SJ156 (l}
14SJ156 (l}
Component Cooling System Check Valves Safety/Relief Valves llCCl 11CC14 12CC1 12CC14 13CC1 1CC34 1CC109 1CC40 1CC137 (l} 1CC51 1CC317 (l} 1CC58 1CC320 (l} 1CC63 1CC68 1CC75 1CC81 1CC112 11CC129 12CC129 13CC129 14CC129 1CC135 1CC138 1CC147 1CC156 1CC162 1CC165 1CC170 1CC193 1CC212 .
ATTACHMENT 4 -a-(CONTINUED)
Chilled Water System Check Valves Safety/Relief Valves 11CH13 12CH13 1CH61 Residual Heat Removal System Check Valves Safety/Relief Vaives 1RH8 (1) 1RH3 12RH8 (1) 13RH27 (1) 14RH27 (1)
Auxiliary Feedwater System Check Valves Safety/Relief Valves 11AF4 1AF99 12AF4 1AF128 13AF4 11AF8 (1) 12AF8 (1) 13AF8 (1) 11AF23 12AF23 13AF23 14AF23 Containment Spray System
.Check Valves Safety/Relief Valves 11CS4 (1) llCSS 12CS4 (1) 12CS5 11CS21 (1) 1CS26 12CS21 (1) 11CS48 (1) 12CS48 (1)
ATTACHMENT 4 (CONTINUED)
Main Steam System Check Valves Safety/Relief Valves llMSll (3) 12MS11 ( 3) 13MS11 (3) 14MS11 (3) 11MS12 ( 3) 12MS12 ( 3) 13MS12 (3) 14MS12 (3) 11MS13 (3) 12MS13 (3) 12MS13 (3) 13MS13 (3) 14MS13 ( 3) 11MS14 (3) 12MS14 (3) 13MS14 (3) 14MS14 (3) 11MS15 (3) 12MS15 (3) 13MS15 (3) 14MS15 (3)
Reactor Coolant System Check Valves Safety/Relief Valves 1PR25 1PR3 1PR4 1PR5 Chemical and Volume Control System Check Valves Safet:LLRelief Valves 1BR99 11BR81 11BR152 (2) 12BR81 12BR152 (2) 13BR81 13BR152 (2) 1CV6 1CV23 (1) 1CV43 1CV36 (1) 1CV115 1CV42 (1) 1CV124 1CV47 1CV141 1CV52 1CV241 1CV63 1CB253
ATTACHMENT 4 (CONTINUED)
Check Valves Safety/Relief Valves Check Valves 1CV74 (1) 11CV99 (1) 12CV99 (1) 13CV99 (1) 14CV99 (1) 1CV135 lCV137 1CV147 11CV154 12CV154 1CV173 1CV176 (1) 1CV180 1CV183 1CV189 1CV196 (1) 1CV198 (1) 1CV275 Service Water System Check Valves Safety Relief Valves llSW2 1SW109 12SW2 11SW242 13SW2 12SW242 14SW2 13SW242 15SW2 14SW242 16SW2 15SW242 11SW5 (1) 12SW5 (1) 13SW5 (1) 14SW5 (1) 11SW13 12SW13 13SW13 14SW13 15SW13 16SW13 11SW34 12SW34 11SW36 12SW36 11SW38 e 12SW38 11SW44 12SW44
~
ATTACHMENT 4 (CONTINUED) e Check Valves Safet~/Relief Valves Check Valves 13SW44 (l) 11SW47 12SW47 llSWSl 12SW51 11SW53 12SW53 llSW77 12SW77 11SW79 (1) 12SW79 (1) 11SW99 12SW99 13SW99 Nitrogen Supply System Check Valves Safety/Relief Valves 1NT26 lNTlS Primary Water System Check Valves Safety/Relief Valves 1WR81 1WR123 BB/LL:mlr 10/17/77 P77 79 09/15 & 20/23
r Attacbrnent 5 Revision 1 of the Following Diagrams System
~~~~~~1. 241825-A-1568-l Reactor Coolant~~~~~~
~~~~~~2. .241826-A-1568-l Stearn Generator Feed & Condensate~~~~~~
~~~~~~3. 241827-A-1568-l~~~~~~
~~~~~~'Main Reheat and Turbine By - Pass Stearn~~~~~~
~~~~~~4. 241829-A-1568-l Fire Protection~~~~~~
~~~~~~5. .241830-A-1568-l~~~~~~
~~~~~~Stearn 'Generator Drains & Blowdown~~~~~~
~~~~~~6. 241831-A-1568-l Chemical & Volume Control Operation 7.. 241832-A-1568-l Chemical & Volume Control Boric Acid Recovery~~~~~~
~~~~~~8. 241833-A-1568-l Chemical & Volume Control Primary Water Recovery~~~~~~
~~~~~~9. 241834-A-1568-l Component Cooling~~~~~~
~~~~~~10. 241835-A-1568-l Residual Heat Removal~~~~~~
~~~~~~11. 241836-A-1568-l Spent *Fuel Cooling~~~~~~
~~~~~~12. 241837-A-1568-l Safety Injection~~~~~~
~~~~~~13. 241838-A-1568-l Containment Spray~~~~~~
~~~~~~14. 241839-A-1568-l Auxilary Feedwater~~~~~~
~~~~~~15. 241840-A-1568-l Waste Disposal liquid~~~~~~
~~~~~~16. 241841-A-1568-l Service Water Nuclear Area~~~~~~
~~~~~~17. 241843-A-1568-l Demineralized Water Restricted Areas~~~~~~

ML19031A096

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