ML20064J375

Forwards Responses to Matls Engineering Branch Questions on Preservice & Inservice Insp Dtd 781208.Will Incorporate Matl Into FSAR by Amend 60

ML20064J375
Person / Time
Site:	Sequoyah
Issue date:	02/14/1979
From:	Gilleland J TENNESSEE VALLEY AUTHORITY
To:	Varga S Office of Nuclear Reactor Regulation
References
NUDOCS 7902210257
Download: ML20064J375 (24)

Text

-

TENNESSEE VALLEY AUTHORITY CH AMANCCGA. rENNEssEr 374ot 500C Che.stnut Street Tower II FES 141979 Director of Nuclear Reactor Regulation Attention:

Mr. S. A. Varga, Chief Light Water Reactors Branch No. 4 Division of Project Management U.S. Nuclear Regulatory Consission Washington, DC 20555

Dear Mr. Varga:

/

N In the Matter of the Application of

)

Docket Nos. 50-327 Tennessee Valley Authority

)

50-328 Enclosed are response to the Materials Engineering 3 ranch (MEB) ques,. ions on preservice and inservice inspection transmitted b;'

S.

A. Varga's letter to N. B. Hughes dated December 8, 1978.

The response to MEB item 1, Requests for Relief from Testing Require-ments, is contained in Enclosure 1 and will be incorporated into the Sequoyah Nuclear Plant (SNP) Final Safety Analysis Report (FSAR) as a revision to question 5.9 And as question 5.29. is the response to ME3 item 2, Compliance with 10 CFR Part 50, Appendixes G and H, which will be incorporated into the SNP FSAR as the response to question 5.30.

MEB item 3, Steam Generator Inservice Inspection, is addressed in and will be incorporated into the SNP FSAR as the response

_ o question 5.31.

This raterial will be incorporated into the SNP FSAR by Amend:ent 60.

Very truly yours, w

. E. Gilleland Assistant Manager of Power l

Enclosure

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ENCLOSURE 1 Request for Relief from Certain Testing Requirements (KdB Item 1 of December 8, 1978, Letter from S. A. Varga to N. B. Hughes) 5.29 Specific Areas Identified in the Review which may Reauire Relief Based on our review of TVA's preservice and inservice inspection program for Sequoyah Nuclear Plant Unit Nos. I and 2, we have identified the following items that may reqaire relief:

1.

The NRC's position is that the inspection of like type of nozzels for the reactor vessel, Item Bl.4, Category B-D be evenly distributed over the inspection interval to meet the requirements of ASME Section XI.

Since your program indi-cates this requirement will not be met, relief is required for this item.

2.

Your program indicates that the control rod drive housing will not be examined per the requirements of Item Bl.18, Category B-0; therefore, a request for relief is required.

To support your relief request,'iscuss the maximum crack d

opening (largest size line break) in which the reactor can be shutdown and cooled by.tae reactor coolant makeup system only.

Compare the normal cool-down capacity with the maxi-mum opening created if the CRDM housing weld failed.

3.

Clarify the extent of v'olumetric examination that can be achieved for the cast stainless steel RCS piping.

If the complete volume is not examined f.or the Category B-J welds, relief from the requirement is needed. Lescribe the volume that will be examined.

4.

State if there is any bolting two inches in diameter or less (Category B-G-2) associated with the reactor vessel.

If there is, include it in the program.

5.

In Table A of the ISI program for Class A components, 2-1/2 inch diameter welds listed under " Safety Injection system, circumferential and socket welds" are scheduled for PT examination. Clarify if these welds are socket welds.

If the welds are not socket welds, relief is required from the volumetric examination requirements of Iten B4.5, Category B-J.

6.

In Table A of the ISI program for Class A components, 11-inch diameter stainless s: eel and 6-inch diameter Inconel piping welds are scheduled for PT examination.

State why these welds are not being examined in conformance with the requirements of Item B4.5, Category B-J.

Request relief for these items is necessary if the above stated requirements cannot be met.

R5.29 - 1

W 7.

State if the reactor coolant pump casings are welded.

If the pump casings are welded, include the examination of the welds

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in the preservice/ISI program.

8.

Since the inspection of valve bodies per the requirements of-Category B-M-2 will not be fulfilled by your program, relief from the requirement is needed. For preservice inspection, indicate if the visual inspections of the valve bodies have been performed by the manufacturer and if the test records i

are available.

9.

Section XI requires that an iudependent third party be utilized as the inspection agency.

In your program you state the TVA will perform the independent third party inspection function itself since it is an agency of the Federal government. TVA should provide the technical and/or legal justification that allows relief or exemption from having the third party inspec-tion performed by an independent agency, other than TVA.

10.

Provide details regarding the system testing that will be performed. Will the hydroatatic and leak testing meet the requirements of ASME Section XI?

11.

In your ISI program, you state the program will essentially comply with ASME Section XI, 1974 Edition through Summer 1975 Addenda. However, for the examination of piping you proposed to use Appendix III of Section XI from a later code edition.

Since the approved code,1974 edition including Summer 1975 Addenda requires ultransonic testing to Appendix I or Article 5 of ASME Section V, relief from the requirements of the code is needed.

12.

Table B of the licensee's ISI submission dated April 19, 1978, includes ASME Class 2 piping, Categories C-F and C-G..The total sample of the piping'is listed but not the sample to be tested or the test ' schedule for this interval. This informa-tion must be supplied to demonstrate compliance with code requirements.

13.

Clarify the program statement that integrally welded supports-and support components, Classes A, B, and C will be examined as prescribed by ASME Section XI but the examinations are not included in Tables A and B.

Response

1.

Section XI does not designate that examination of like-type nozzels for the reactor vessel, Item Bl.4, Category B-D, be distributed evenly over the inspection interval.

Informal discussions with the ASME Section XI Code Committee revealed that the intent of Section XI is that examination of nozzles, rather than like-type nozzles, be distributed over the inspection interval in accordance with IWB-2400.

i This position is further clarified in the 1977 Edition, Summer 1978 Addenda, of Section XI, Table IWB-2500-1, Items B3.10 and 33.90.

45.29 - 2

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2.

The Sequoyah inservice inspection program will be revised to include examination of the pressure-retaining welds in the peripheral control rod drive housings, item Bl.18, Category B-0.

3.

The extent of volumetric examination for the cast stainless-steel RCS piping includes the weld metal and base metal for one wall thickness beyond the edge of the weld in accordance with Item B4.5, Category B-J.

Ultrasonic (UT) examinations of each weld are perfor=ed paral-lel to the weld from two directions and perpendicular to the weld from two directions.

4.

There is no bolting two inches in diameter or less (Category B-G-2) associated with the reactor vessel.

5.

The 2-1/2-inch diameter velds listed in Table A of the ISI program under " Safety Injection System, Circumferential and Socket Welds" are scheduled for UT examination. The schedule for liquid penetrant (PT) examination was made in error.

6.

The 11-inch diameter stainless-steel piping welds in Table A of the ISI program (upper-head injection system) were listed in error and will be removed from the program. The 6-inch diameter Inconel piping welds listed in Table A are scheduled fer UT examination.

The schedule for PT examination was made in error.

7.

The reactor coolant pu=p casings are not welded.

S.

At least one valve in each group of similar valves, as identified in Examination Category B-M-2, will be visually examined for the preservice inspection. The ISI program will be revised to reflect these examinations.

See Q5.9, Exception 18.

9.

See Q5.9, Exception 1.

10.

The hydrostatic pressure testing program is in the course of pre-paration. Any requests for relief from Section XI requirements will be submitted pending completica of the program.

11.

References to App.sndix III of Section XI will be withdrawn.

UT testing will be performed in accordance with Appendix 1 of Section XI or Article 5 of Section V.

12.

The sample of Class B piping welds to be excained, Categories C-F and C-G, will be included in the ISI program.

13.

Installation of all Class A, B, and C integrally welded supports and support components has not been finished pending completion of cen-struction of various components. The examinations of the integrally welded supports and support components will be included in Tables A and B of the ISI program af ter completion of support installation.

45.29 - 3

TABLE QS.9-3 EXCEPTION 1 I.

Components 1

All components.

II.

Exception Taken The inspector will be an employee of TVA and will not be qualified in accordance with IRA-2130.

III. Justification Present TVA policy is to provide its own inspection services. TVA is a Federal agency and, as other Federal agencies, acts as its own j

inspector and is not subject to State or other non-Federal inspectors.

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IV.

Testing in Lieu of Section XI Requirements TVA will provide its own independent review of the Section XI program through its central office staff in Chattanooga.

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EXCEPTION 16 I.

Components Various class 1 and 2 piping systems.

II.

Exception Taken Calibration standards will not include the calibration reflector required by paragraph T-533.2(a) of Article 5 of ASME Section V.

III. Justification Paragraph T-533.2(a) of Article 5 of ASME Section V requires that i

the basic calibration block include a basic calibration hole drilled j

parallel to the contact surface. However, paragraph T-533.2(b) per-mits the use of other calibration reflectors provided equivalent respcases to that from the basic calibration hole are demonstrated.

In 1973 TVA and Nuclear Services Corporation performed an evaluation of notches versus side-drilled holes in basic calibration blocks.

A 5 percent notch calibration reference reflector was used to deter-mine the primary ultrasonic reference level on pipe having a wall thickness less than 1 inch.

It is demonstrated in figure Q5.9-1 that calibration with the 5 percent notch is on the average 3.5 dB more sensitive than calibration using a 3/32-inch side-drilled hole.

We conclude that calibration standards fabricated with notches provide a conservative examination.

IV.

Testing in Lieu of Section XI Requirements TVA at its discreti6n may use 5 percent notches in place of side-drilled holes as permitted by paragraph T-533.2(b) of Article 5 of ASME Section V.

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EXCEPTION 17 I.

Coeponents Reactor coolant pump (RCP).

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II.

Exception Taken All the RCPs will not be disassembled solely for visual examinatica of pump internal pressure boundary surfaces.

III. Justification In absence of required maintenance, disassembly of a reactor coolant pump solely to perform a visual examination of internal surfaces is impractical. This would represent an unnecessary employee exposure to high radiation and contamination areas and an excessive expense to TVA.

Time required for this major task of disassembly, examination, and reassembly would consume at least three weeks of 24-hours per day work. Radiation dose rates of the pump exterior will average 100-300 mrem / hour, and pump internal dose rates will average 10-20 rem /

hour.

l The benefit received from this major effort is minimal considering employee exposure, potential damage to safety related equipment, and cost in dollars.

In addition, the two units at Sequoyah Nuclear Plant will operate under similar conditions. Therefore, we feel that if a pump from one of the units is disassembled for maintenance during a 10-year interval, the visual examination performed will be ' representative of the pump condition for each unit. This would avoid unnecessary employee exposure to the high radiation dose rates noted above.

We conclude that if one pu=p is disassembled for maintenance during the 10 year interval, the visual examination performed satisfies examination category B-L-2 requirements for both units. Disassembly of the pump solely for visual examination is impractical.

IV.

Testing in Lieu of Section XI Recuirements The internal surfaces of the reactor coolant pump casing will be visually ex: mined whenever the surfaces are made accessible when a i

pump is disassembled for maintenance purposes.

If during the 10-year interval a pump from either unit is not disassembled for maintenance, a pump from cae unit shall be examined from the exterior. This shall be accomplished by ultrasonic thickness measurements of the pump

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25.9 - 45

TABLE Q5.9-3 (Continued)

EXCEPTION 18 I.

Components Class 1 valves exceeding 4-inch nominal pipe size.

II.

Exception Taken All of these valves will not be disassembled solely for visual examination of valve internal pressure boundary curfaces.

III. Justification During routine maintenance, visual examinations of valve body internal pressure boundary surfaces are performed and documented under existing plant administrative procedures. Most Class 1 valves, particularly containment isolation valves, are disassem-bled frequently for maintennace.

In addition, the two units at Sequoyah Nuclear Plant will operate under similar conditions.

If a valve from one of the units is disassembled for maintenance within a 10-year interval, we feel that the visual examination performed would be representative of both units and would be sufficient to satisfy the examination requirements for both units for that particular valve classification as defined in examination category B-M-2.

We conclude that if one valve in each group of valves of the same constructional design and manufacturer that perform similar func-tion is disassembled from either unit during the 10-year interval, the visual examination performed satisfies examination category B-M-2 requirements for both units.

IV.

Testing in Lieu of Section XI Requirements If a valve from a particular classification has not been disassem-bled as the end of the inspection interval approaches, a case-by-case study will be made to determine the practicality of disassembling a valve from one of the units solely for visual examination (determine if draining the vessel would be required, etc.).

If necessary, a request for relief will be issued at that time.

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45.9 - 46

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TABLE Q5.9-3 (Continued)

EXCEPTION 19 I.

Components Class I and 2 pressure-retaining welds in piping.

II.

Execption Taken Volumetric examination of longitudinal, circumferential, and pipe branch connection welds (examination categories B-J, C-F, and C-G) will not be done for all welds from both sides.

.III.

Justification In some cases it will be impractical to inspect all welds from both sides, i.e., nonremovable hanger interference or valve and pump casings adjoining the welds. These welds will be noted on the ultrasonic examination data sheets.

IV.

Testing in Lieu of Section XI Requirements In addition to the visual examination performec during system leakage and hydrostatic pressure tests, a "best effort" ultrasonic examination will be performed.

Where there is interference or problems from one direction only, consideration will be given to performing an angle beam examination for 2T + (greater than two wall thicknesses from one edge of weld) from the accessible side of the weld in order to examine the entire weld cross section.

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Rs.9 - 47

j TABLE Q5.9-3 (Continued)

EXCEPTION 20 I.

Components Steam generators (four per unit).

II. Exception Taken Volumetric examination of circumferential shell welds (examination category C-A) will not be completed for one weld.

III. Justification One circumferential shell weld on each generator is inaccessible due to the upper steam generator support brackets (weld nos. SGW-DI, SGW-D2, SGW-D3, and SGW-D4).

See figure Q5.9-2 for weld location.

Also, figures Q5.9-3 and Q5.9-4 show arrangements of the support brackets. One weld on one generator will be examined on a "best effort" basis for the baseline inspection and during the four inservice inspection intervals in accordance with IWC-2411.

IV.

Testing in Lieu of Section XI Requirements

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EXCEPTION 21 I.

Components Reactor pressure vessel.

II.

Exception Taken One-hundred percent preservice baseline volumetric examination of lower head dollar veld (examination category B-B) will not be made under conditions and with equipment and techniques equivalent to those expected to be employed during inservice inspection.

III. Justification TVA will employ automated remote inspection devices to examine most of the reactor vessel welds. These examinations will be conducted from the vessel inside diameter. However, the lower head weld on each reactor pressure vessel is partially inaccessible for examina-tion from the vessel inside diameter due to instrumentation tubes which penetrate the lower head (weld no. W01-02--see figures Q5.9-5 and QS.9-6).

Portions of the weld can be examined from one side (as permitted by 1-5121 of Section XI) and will include 100 percent of the examination volume in accordance with IWB-3511.1 of Section XI.

These portions of the weld will be reexamined during the inservice intervals in accordance with examination category B-B of Table IWB-2500.

IV.

Testing in Lieu of Section XI Requirements A 100 percent baseline examination of the weld will be conducted from the vessel outside diameter. This will be acommplished by performance of a manual ultrasonic examination. A remote ultra-sonic examination will be conducted from the vessel inside diameter on all accessible areas of the weld.

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TABLE Q9.5-3 (Continued)

EXCEPTION 22 I.

Components Steam generator nozzle.

II.

Exception Taken Volumetric examination of nozzle inside radius section on the primary side (examination category B-L) will not be made.

III. Justification Each steam generator consists of two integrally cast noizles and two integrally cast manways. The present capability of ultrasonic test-ing is not sufficient to examine cast material of this thickness and achieve meaningful results.

(See figure QS.9-7.)

IV.

Testing in Lieu of Section XI Requirements None.

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TABLE Q9.5-3 (Continued)

EXCEPTION 23 I.

Components Reactor coolant loop pipe welds.

II. Exception Ta' ken Volumetric examination of circumferential welds (examination category B-J) will not be done for all welds.

III. Justification Two circumferential welds in the reactor coolant loop piping (RC-2351 and RC-3151, loops 3 and 4) are located inside the reactor vessel shield wall and are inaccessible for baseline and inservice examina-tion (see figures Q5.9-8 and Q5.9-9).

Both welds have undergone shop radiographic examinations. Since the baseline inspection serves as a reference to future inservice inspections and both welds will be inaccessible for inservice inspections, the shop radiographic examina-tion coupled with the Section III hydrostatic test will provide ade-quate proof of integrit'.' of the system welds.

Inservice system leakage and hydrostatic testing till prove weld integrity during the life of the plant.

IV.

Testing in Lieu of Section XI Requirements None.

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ENCLOSURE 2 (MEB Item 2 of December 8, 1978, Letter from S. A. Varga to N. B. Hughes) 5.30 Compliance with 10 CFR Part 50, Appendices G and H 1.

State if the ferritic materials in the pressurizer and steam generators that are part of the reactor coolant pressure boundary meet the exact requirements of 10 CFR Fart 50, Appendix G.

Pro-vide the fracture toughness data available for these materials.

State the ASME Code Edition used for these components.

2.

State if all material for bolting and other fasteners greater than 1-inch diameter in the reactor coolant pressure boundary meet the exact fracture toughness requirements of 10 CFR Part 50, paragraph IV.A.4.

Provide the fracture toughness data available for all bolting greater than 1-inch diameter used for the reactor coolant pressure boundary.

Indicate the ASME Code Edition used for these components. The mechanical properties which have been submitted for the closure head bolting for Units Nos. I and 2 need not be provided again.

3.

State if the requirements of 10 CFR Part 50, Appendix G, para-graphs IIIB.3., 4, and 5 were met for the fracture toughness tests which were performed. Clarify if the testing was performed by an organization with a quality assurance program in conformance with 10 CFR Part 50, Appendix B.

4.

Although Westinghouse Topical Report WCAP-7924 has been accepted by the NRC, the method for determining RT shift was not NOT accepted. Provide information demonstrating that the pressure-temperature limits for reactor vessel heat-up and cool down will be constructed using the prediction for temperature shift con-tained in Regulatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

i

Response

It has been determined that retrieval of the requested information required Westinghouse Electrie Corporation support; therefore, the information will be forwarded to NRC by March 1,1979.

45.30 - 1

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1 l.

ENCLOSURE 3 l

(MEB Item 3 of December 8, 1978, Letter from S. A. Varga to N. B. Hughes) 4 1

5.31 Steam Generator Intervice Inspection Confirm that the preservice and inservice inspection of steam generator tubing will be conducted in accordance with Regulatory Guide 1.83 Revision 1.

If any of these examination requirements cannot be met, a complete technical justification to support your conclusions must be provided.

I

Response

Preservice and inservice inspection of steam generatcr tubing will be conducted in accordance with Regulatory Guide 1.83, Revision 1.

These requirements will be included in the ISI program.

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45.31 - 1

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