ML20059A399

Forwards Response to NRC 900521 Request for Addl Info Re Plant Inservice Insp Program

ML20059A399
Person / Time
Site:	Braidwood
Issue date:	08/15/1990
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
References
NUDOCS 9008230039
Download: ML20059A399 (45)

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Dr-. T.E. Murley, Director m ',

Office Nuclear Reactor.egulation-i r

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.U.S. Nuclear Regulatory Commission Washington, D.C. ~20555

,i Aitn: Document Control Desk o,<s 1

Subject:

Braidwood Statin Units 1 and 2.

Inservice Inspecs ion Program (ISI) ffRC_Doj;ket Nos. 50-456 and 50-457 y~~

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Reference:

(a) May 21, 1990 S.P. Sands letter to T.J. Kovach Reference (a) provided the NRC requett for additional information in support of'the ongoing NRC review of the Braidwood Unit 1 and 2 ISI program, n

'In order to clarify what was required in this request, a-teleconference was i

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conducted on July 11, 1990~between Commonwealth Edison,- the NRC staff.and

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.3 Idaho National Engineering Laboratory (I.N.E.L.).

As-a result of this teleconference certain relief requests and notes contained in the Braidwood ISI Program have been revised or deleted..The enclosure describes these changes.and provides.the Edison response to the items requested in reference (a). Accordingly, enclosed are new revisions to be included in the NRC's copy'.

of the Braidwood:ISI Program.

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Please address any questions concerning this submittal to this office.

Very truly-yours, L. '

..</C. h e k i

L S.C. Hunsader Nuclear Licensing Administrator L

/Imw/scl/1D139 cc: Resident Inspector-BW S. Sands-NRR i

H. Shafer-RIII l'

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SO ENCLOSURE

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• M RESPONSE TO RE0 VEST FOR~ ADDITIONAL inh. A TION' BRAIDH000. UNITS 1 AND2-ISI PROGRAM

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The following is:the Commonwealth Edison (Edison) response to the NRC

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request for additional information:

A) NRC Request:

l Please provide the staff with Boundary Diagrams which define the ASME Code Class 1, Class 2 and Class 3 boundaries for the systems at the Braidwood Nuclear Power-Station, Units-1 and 2.

Edison Response:

The NRC Office of Nuclear Reactor Regulations (N.R.R.)

will supply I.N.E.L. with copies of the Braidwood Unit I and 2 Piping and Instrumentation Drawings (P&ID's).

Y B) NRC Request:

Address the degree of compliance with NRC. Regulatory Guide.l.150, " Ultrasonic Testing of Reactor _ Vessel Helds During Preservice and Inservice Examinations."

-t Edison Response:

Attachment A provides the Hestinghouse/ Commonwealth Edison position with respect to NRC Regulatory Guide 1.150.-

Westinghouse has been contracted to perform first period automated ultrasonic examination =of'the reactor pressure vessel for Braidwood Units 1 and 2.

C) NRC Re_ quest:

Note 10 of Section 2.3 of the ISI Program Plan states-s "The NRC has expressed a concern dealing with intergranular stress corrosion cracking in lines that contain stagnant borated water.

Braidwood Unit I will perform augmented volumetric examinations on class.two containment. spray (C.S.) welds.

The inspection shall include seven and one-half percent (7.57.) sampling of the welds in a single-train between the (C.S.) pump and the first weld beyond the isolation valve inside containment." A sampling of welds in the Braidwood, Unit 2, containment spray system should also be examined.

Verify that this note will 's revised to include examinations for Braidwood, Jnit 2, also.

Braidwood Response:

Braidwood will comply with the I.N.E.L. request to perform augmented examination on both the Unit I and 2 containment spray (C.S.) systems.

See revision 3 of note 10, page 2-27 contained in Attachment B.

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- e D) NRC Request:

In~ Commonwealth'Edt' son's-September 118L1986 submittal',-the Lir.ensee committed to examine a random sampling of 7.5% of the-large-bore-(greater than 4 inches) piping circumferential' welds in-the Safety Injection, L.9 mical and Volume Control, and.

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. Containment Spray systems.

int: submittal states that "These welds will be-examined over the ten year; inspection interval as described in the ISI Program and will te. tracked for the life of-the plant." However,'this. commitment =1s not reflected in'the Braidwood Nuclear Power Station, Units 1 and 2, First 10-Year Interval ISI Program Plan.

The staff notes that'the Chemical and Volume Control system and 6. 8',.12, 14 and:24-inch lines in the.

Safety Injection system have been exempted from inservice volumetric examinations based on the. pressure / temperature-s exemption criteria in Section XI'.

Verify that volumetric examinations of a 7.5% sampling of the-j Class 2 piping welds.in these systems will be performed,' as identified in the September 18, 1986 submittal, and that.the ISI-

-i Program Plan,will be revised accordingly.for.both Units 1-and-.2.

The Licensee should note that later Code editions and addenda _ do not permit pressure / temperature exemptions for RP'. and ECC systems.

Edison Response:

i During:the July 11, 1990 conference call between Braidwood Station, N.L.A...I.N.E.L. and N.R.R., 1t vas agreed that this

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item will remain open until,I.N.E.L. clearly defines its' specific requirements for augmented examinations, q

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.E) NRC Request:.

j Relief Request NR-4:1 Relief is requested from performing the Code-required volumetric examination of the Pressurizer and Steami Generator nozzle inside radius sections.' Byron Relief-Request, NR-3, which requests the same relief for Byron, Unit 1, was previously evaluated by the staff and relief.was denied due.to-

-insufficient justification.

By letter dated July 11, 1989, Byron j

Relief Request NR-3.was removed from the Byron.ISI Program.- The letter states.that Commonwealth Edison will continue'to evaluate.

the feasibility of performing the examination.

Relief Request NR-3 for Byron, Unit 1, discusses attempts to develop an 1

W applicable ultrasonic examination technique using a mockup.

How 0

-are the attempts to perform the examination for Byron applicable for'the'Braidwood plants? Also, provide a sketch which shows the location'of the closure ring with respect to the. steam generatcr primary nozzles listed in Relief Request NR-4.

p Edlson Response:

for the interim, Braidwood Station has removed Relief Request i

NR-4 from the ISI Program. Commonwealth Edison will continue to evaluate the feasibility of performing these examinations.

If at.

some later date we feel relief is justified, we shall submit a

. revised relief request to the NRC.

F) NRC Request:

Relief. Request NR-9: Confirm-that the Code-regulred surface

-examination will be performed on the Reactor. Pressure Vessel head,to-flange welds.

L Edison Response:

The Code-required surface examination will be performed on the t

Reactor Pressure Vessel head-to-flange welds.. Relief Request L

NR-9 has been revised to more clearly state'this fact, see pages

2-51 through'2-58 of Attachment B.

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i G) NRC_ Request:

Relief Request NR-12, Steam Generator Nozzles: !Reliefis-L requested from' performing the Code-required volumetric n

examination of the nozzle inside radiu!. sections of the. Stear Generator main feedwater. nozzles (Units 1 and 2).

The drawing attached to the. relief request shows that the main feedwater L

nozzle was designed with an internal multiple venturi type flow restrictor.

The flow restrictor area does not utilize'a radiused L

-nozzle as described in Figure IHC-2500-4, but'instead has.several L

Individual' inner radii, corresponding to each venturi.

L Therefore, the Code-required volumetric examination of these-t nozzle inner radius sections is impractical to perform.

However, l

the base of the nozzle is consistent with=ine radiused nozzle described in Figure IHC-2500-4 and should receive the Code-required volumetric examination.

What attempts have been made to perform the examination of the inside radius _of these nozzles? -Describe a "best effort" volumetric examination of this nozzle inside radius section that could be performed.

Residual Heat Removal Heat Exchanger Nozzles:

Relief-is y

requested from performing the Code-required _ volumetric l

o examinations of the' nozzle-to-shell welds and nozzle inside-radius sections of the subject nozzles.

Is the design of these i

no zles consistent with the sketch in Attachment B Page 5 of 5,.

of the July 11=, 1989 submittal for Byron Station, Unit-l? If not,'please provide a sketch showing the design of the subject nozzles.

Edison Response:

The portion of Relief Request NR-12 concerning the Steam Generator main feedwater nozzle has been removed.

Commonwealth r

p Edison will continue.to evaluate the feasibility of performing this examination.

If at some later date we feel relief is u

i justified, we shall submit a revised relief request to the NRC.

The portion of the relief-concerning the Residual Heat Removal Heat Exchangers Nozzles, has been revised to include a drawing of the subject weld configuration.

This configuration matches that which was submitted with the Byron Unit 1 ISI Program.

See pages 1

2-66 through 2-68 of Attachment B.

H) NRC Request:

Relief Request NR-13:

Please provide additional information which will justify the determination that the Code-required volumetric examination.is impractical to perform on the subject Reactor Coolant System nozzle-to-reducer welds. This information

.could consist of a dimensional sketch showing the precluding geometry.

Have smaller transducers (search units) been considered for use? If a "best effort" volumetric examination L

were performed, what percent of the Code-required volume could and would be examined?

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I Edison Response:

Draidwood Station has removed Relief Request NR-13 from the ISI Program. A "best effort" volumetric examination will be performed in accordance with the ISI schedule.

If the code required volume is not able to be examined, a relief request will

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be submitted to the NRC at that time.

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,7 I) IBIC hoguest:

u-Rel Request NR-17: ' Relief is requested from performing the Code-required volumetric examination of Class 2 pressure vessel shell and head circumferential welds with all thicknesses ranging from 0.252 inch to 0.327 inch.

The Braidwood, Units 1 and 2, first 10-Year Interval ISI Program Plan references the use of ASME Code Case N-435-1, " Alternative Examination Requirements for Vessels with Wall Thickness 2 inches or Less." Paragraph (C) of this Code Case states that, for welds in vessels with nominal wall thickness greater than 1/5 inch and les: :han or equal to 2 inches, ultrasonic examination may be A

performed using the rules of the Winter 1985 Addenda of Section XI, Division 1. Appendly III.

Other Licensees are able to perform ultrasonic examinations of vessel welds of this wall thickness range. The Licensee should either consider withdrawing this relief request or provide detailed technical information justifying the determination of impracticality.

Edison Response:

Braldwood Station has removed Relief Request NR-17 from the ISI Program.

In its place we are submitting Note 13 page 2-29.a. see Attachment B.

This note allows for exempting ASME class two vessels from the ISI Program M sed on cumulative inlet and cumulative outlet pipe cross-sectional areas not exceeding four inches in diameter.

J) IIRC Request:

Relief Request NR-18:

Relief is iequested from performing the Code-required volumetric examination of two Reactor Coolant System piping welds because the welds are encased in permanent whip restraints.

Please provide a dimensional sketch showing the subject welds and permanent whip restraints.

To what extent can the subject welds receive a phrtial volumetric examination?

Edison Response:

Relief Request NR-18 has been revised to include drawings of the whip restraint versus circumferential weld locations.

The test of the ralief has been revised to indicate the code required surface examination will be performed on, eld 2SI-09-17. see pages 2-76 through 2-76.b of Attachmen+ '.,.

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4 E) NRC Request 2 Relief Request NR-19:

The staff has recently noted significant improvements in the techniques being used for volumetric examination of branch pipe connection welds.

What attempts have been made to perform the Code-required volumetric examination of weld 251-02-457 r e cribe a "best effort" volumetric examination that could be performed on the subject weld.

Edison Response:

Braidwood Station has removed Relief Request NR-19 from the ISI Program. A "best effort" volumetric examination will be performed in accordance with the ISI schedule.

If the code required volume is not able to be examined, a relief request will be submitted to the NRC at that time.

L) NRC Request:

Relief Request CR-1:

Relief is requested from removing the insulation from all nonexempt coronent supports on Cooe Class I and 2 insulated lines for the sole purpose of performing a visual examination on the portion of he nonintegral or integral attachment within the insulation.

Based on the information submitted, the reviewer has no idea of the magnitude or number of component supports for which relief is being requested as compared to the component supports which may be receiving the Code-required examination.

The submittal requests relief for all nonexempt component supports on Code Class I and Class 2 insulated lines in the AF, CV, FH, MS, RC, RH, RY, and SI systems.

The regulations do not provide for granting generic relief requests.

Paragraph IW"

!00(e) states that "Hhere the mechanical connection of. nonintegral support is buried within the component insulation, the support boundary may extend from ie surface of the component insulation provided the support either carries the weight of the component or serves as a structural restraint in compression". ASME Code Section XI, Interpretation:

XI-1-86-11 (Interpretations No. 18), Question (5), provides clarification on the Code requirement for those componentt not excluded based on INF-1300(e).

Thereftre, the i

Licensee should provide additional information describing the analyses performed to determine which of the cceponetit suppods may be excluded based on IHF-1300(e) and evaluate the remaining supports to determine which supports may require reHef from the Code-required examination along with the technical 4

justifications.

Relief.is not required for component supports I

which are exempted based on IHF-1300(e).

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For those supports remaining after the above exemption, relief F

could be considered for.the-following: -(a) if the insulation is i

required by other regulations or the Technical Specifications to be in place (e.g. fire stops), or (b) if the Licensee can demonstrate that the failure of the component support would be obvious should the support fail with the insulation installed.

The method for determining item (b) above sho91d be described in-the request for relief.

If the above cannot be technically justified, it is expected that the insulation be removed for the examination.

Based on the above, it is concluded that:

(a) the Licensee should not ask for relief for supports wnich are exempted based on INF-1300(e), and (b) the 1.icensee she'.11d provide the additional information as described above as part of the technical justification for the granting of relief.

By letter dated July 11, 1989, Byron Relief Request CR-2, which requests the same relief for Byron 1, was withdrawn.

Relief Request CR-2 was previously evaluated by the staff and relief was denied due to insufficient justification.

Therefore, the Licensee should either provide the additional justifications or consider withdrawing Relief Request CR-1 from the Braidwood, Units I and 2, First 10-Year Interval ISI Program Plan.

Edison Response:

Braidwood Station has removed Relief Requesi CR-1 from the ISI

Program, M) NRC Request:

Relief Request SR-1:

Relief is requested from removing the insulation from all nonexempt safety-related sn'bbers on Code Class 1 and 2 insulated 11nos for the sole purptte of performing a visual examination on the portion of the inte9/a1 or nonintegral attachment within the insulation.

Paragraph INF-1300(e) states:

"Where the mechanical connection of a honiategral support is buried within the component insulation, the support boundary may extend from the surface of the component-insulation provided the support either carries the weight of the component or serves as a structural restraint in compression." ASME Code Section XI, Interpretation:

XI-1-86-11 (Interpretations No. 18), Question (5), provides clarification of the Code requirement for those components not excluded based on IWF-1300(e).

Therefore, the Licensee should provide additional information describing the analyses performed to determine which l

I of the component mpports may be excluded based on INF-1300(e),

and evaluate the remaining supports to determine which supports may require relief from the Code-required examination along with i

the technical justifications.

Relief is not required for component supports which are exempted based on IWF-1300(e).

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For those supports remaining after the above enemption, relief could be considered for the following:

(a) If the insulation is required by other regulations or the Technical Specifications'to be in place (e.g. fire stops); or (b) if the Licensee can demonstrate that the failure of the component support would be obvious should the support fail with the insulation installed.

l The inethod for determining item (b) above should be described ir the request for relief.

If the above cannot be technically

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justified, the insulation should be removed for the examination, j

Based on the above, it is concluded that:.fa) the Licensee should not ask for relief for supports which may be exempted j

based on INF-1300(e); and (b) the Licensee should provide the additional information as described above as part of the 1

technical justification for granting relief.

It is noted that Byron Relief Request SR-1, which requests the

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same relief for Byron, Unit 1, was previously evaluated by the staff and relief was denied due to insufficient justification.

The Licensee should either revise Relief Request SR-1 to include additional justification or consider withdrawing Relief Request SR-1 from the Braidwooo Units 1 and 2, first 10-Year Interval ISI Program Plan.

l Edison Response:

Relief Request SR-1 has been revised to provide nore informat;cn on and justification for, the proposed alternate exam methods in lieu of removing insulation on all safety-related snubbers.

See pages 4-40 through 4-41.a of Attachment B.

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WESTINGHOUSE PROPRIETARY I-O ATTACHWENT A i

NRC RG 1.150 IMPLEMENTATION

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BACKGROUND Westinghouse will perform the 40-month reactor vespl examinations in accordance with the following Westinghouse /Dynacon technical description for implementation of RG 1.150, Rev.1, " Ultrasonic Testing of Reactor Vessels During Preservice and Inservice Inspection".

1 DESCRIPTION The following technical description has been adopted by Westinghouse /Dynacon with respect to implemettation of RG 1.150 for reactor pressure vest.el 40 month examinations.

To satisfy Section XI and RG 1.150 requiren nts, Westinghouse will perform. examinations of the reactor vessel out10t nozzle-to-shell weld, and outlet nozzle bore regions. Nozzle-tysafe-end weids are not considered to fall under RG 1.150, Rev.-1 requ4ements.

Westinghouse /Dynacon implements Appendix A, where the Electric Power Research Institute Ad Hoc Committee recommendations are adopted as an acceptcHe approach to the positions recommended in the base document.

In all cases, what is summarized below S the minimum scope provided by Westinghouse /Dynacon to implement RG 1.130.

In some circumstances, the implementation exceeds the RG recommendations.

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This descr'ption has been divided into five sections, each one corresponding to the five main topics addressed in Appendix A.

Each s'ection contains a summary of those actions Westinghouse /Dynacon takes to implement or incorporate RG 1.150 into the reactor pressure vessel examinations.

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WESTINGHOUSE PROPRIETARY d'

Inspsetion System Performance Checks

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Performance checks of thf inspection system are divided into two categories:

pre exam and field por formance. For pre exam performance, the ultrasonic instruments and transducers are checked for proper operating characteristics.

The ultrasonic instruments are checked for' horizontal, screen height, and amplitude control linearity using the guidelines of ASME Section V, Article 1

4.

These checks are valid for 90-days and are usually performed immediately before deployment.

Spectrum analysis of the transducers to determine actual

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frequency and band width information is performed with a spectrum analyzer program or system. The spectrum analysis is performed eithin six-months of the reactor vessel examination.

These performance checks a.e kept current during the examination and are repeated as required.

Field performance checks include RF wa/eforms and a daily linearity check of the ultrasonic syshm.

Intermediate linearity checks are performed as an integral part of the calibration check.

RF <aveforms for each transducer are taken before and after each examination setup and are performed using the complete examination system.

Transducer characteristics relevant to VDRPS may be establishsd prior to or during calibration.

Calibration blocks supplied by Commonwealth Edison will be used in establishing these characteristics.

Angle beam profile characteristics are performed when necessary to provide additional information when determining the size of a recorded flaw.

This information is available from the recorded UDRPS data of the dynamic calibration.

.. w Calibration The initial and final calibrations are verified on site before and after ea'ch reacter vessel examinatien.

The final calibration check is performed with a

ombination of,a simulator and real reflector which verifies the entire er. amination system including search units.

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WESTINGHOUSETROPRIETARY

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Intermediate calibration checks using the complete examination system are performed nominally every 12-hours or when the examination setup is changed.-

l The calibration is performed using plant specific calibration blocks.

The calibration data is recorded dynamically with UDRPS after being established-statically.

Calibration over the examination volume may or may not be j

equalized using a time corrected gain depending on such variables as calibration block design, volume to be examined, or type of search unit used.

Westinghouse will review the plant's calibration blocks and,'if necessary, i

recommend new calibration blocks or modifications to the axisting calibration blocks, as required.

If an alternative or new conventional block is used, data from this block may be correlated with the original block at the owner's request for reference when comparing previous or current examinations.

Examinatior.

l The reactor pressure vessel examinations are performed using conventional ultrasonic equipment which has been modified to interface with UDRPS.

The

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three basic modifications required are external pulse trigger, video output and RF output.

Examinetions are performed with a minimum 25 percent scan overlap, based on the si:e of the transducer element. The type of examination and/or type of equipment determines whether the scan overlap is increased.

UDRPS and Westinghouse /Dynacon have qualified successfully in several EPRI programs for underclad and thick-wall specimen flaw detection and si-ing using a

,7 a variety of technicues.

These orograms meet the intent of RG 1.150 te be able to detect maximum code-allowaole si:e reflectors with the worst case orientation at the depth of concern.

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WESTINGHOUSE PROPRIETARY l

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.ouse perforts 10 examinations of the nozzle-to shell weld from the ne:Ile bore using at least one angle of inspection that is perpendicular, within pic: nr minut 15-degrees, to the weld / base metal interf ace.

An additional shallower angle is used to ensure coverage in the knuckle area of the outlet nozzle.

Recording and String i

UDRPS. is configured to make a record of all indications in the full examinstien volume. Due to its~ unique features the recording of indications is not amplitude dependent.

Therefore, VDRPS does not " gate out" any ultrasonic data.

The UDRPS record includes information on indication depth, metal path, location from a reference point, and amplitude.

This information is available for any point on a recorded indication. Using this information as an example, the beam profile of a search ur.t may be found on the through wall-dimension of an indication.

The varied di:tley formats of Ut)RFS facilitate easier analysis of indication features, such as tip diffraction, satellite pulses, and spectral reflections.

Indications may be sized based on the DAC (Distance Amplitude Correctien) threshold requirements of ASME Section V and XI, anc RG 1.150.

However, cecause UDRPS is not amplitude dependent, indications of any amplitude that a: pear to be relevant may be analyzed, t

3ecmetric indications from component geometry or due to component geometry are a'so included in the record of UDRPS data.

a Reporting 1

Eeocris are provided in accordance with the site recuirements, and will inc'iude any indications found to be reportable.

Included in the report is the cast estimate e,f the telerances concerning reported flaw si:e.

The tolerances

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I WESTINGHOUSE FROPRIETARY are based on inherent transducer characteristics and characterization of the transducer vertical beam width, which may include using, additional reflectors in the calibration block.

A final report of the reactor vessel examination, including copies of all procedures used and data generated during the performance of the examination, are provided. Estimates of the amount of and reason for missed coverage, and description of alternative techniques if used, are also included.

Copies of all personnel and equipment :ertifications, RF waveforms and any other pertinent data are also provided as part'of this final report.

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ATTACHMENT B 1

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Attached are revisions to the Braidwood Unit I and 2 Inservice Inspection Program.

Please insert them into your copy of the program as indicated below.

i Remove Insert j

pages 2-2 through 2-3, rev. 2 pages 2-2 through 2.-3, rev. 3-J page 2-7, rev. 2 page 2-7. rev. 3 page 2-29.a. rev. 0 I

page 2-29.b, rev. 0 pages 2-44 through 2-45, rev. 2 pages 2-44 through 2-45, rev. 3 pages 2-51 through 2-58, rev. 2 pages 2-51 through 2-58, rev. 3 pages 2-66 through 2-68, rev. 2 pages 2-66 through 2-68, rev. 3 page 2-69, rev. 2 page 2-69, rev. 3 page 2-75, rev. 2 page 2-75, rev. 3 page 2-76, rev. 2 page 2-76 through 2-76.b, rev. 3 page 2-77, rev. O page 2-77, rev. I page 3-12 and 3-13, rev. I page 3-12 and 3-13, rev. 2 page 4-64 and 4-65, rev. I page 4-64 through 4-65.a. rev. 2

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R;visita 3 i

TABLE OF CCEf7ENTS Inservice inspection Program Plan for Nondestructive Esamination I

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2.1 Program Description (Revision 2) 2-4

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2.2 -Program Tables (Unit 1) (Revision 1) 2-6 Alternate Exams 228-230 of 247 2.3 Notes (Revision 2 and 3) 2-10 j

(Rev. 2) Note 1 Main Steam Nossle Inner Radil

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(Rev. 2) Note 2 Category B-G-1 (Code Case N-419) 2-13 (Rev. 2) Note 3 Category B-G-2 (Code Case N-426) i 2-14 (Rev. 2) Note 4 Eddy Current Inspection of Steam Generator U-Tubes 2-15 (Code Case N-401)

(Rev. 2) Note 5 Augmented-High Energy Piping 2-16 (Rev. 2) Note 6 Augmented-Reactor Coolant Pump Flywheels 2-17 (Rev. 2) Note 7 Augmented-Pressuriser and Steam Generator 2-18 Vessel Welds (Unit 1)

(Rev. 2) Note 8 Augmented-Turbine Rotor 2-19 (Rev. 2) Note 9 Reactor Vessel Shell Welds, Limited 2-20 Examinations (Rev. 3) Note 10 Augmented-Stagnant Borated Water in Large Bore 2-27 Piping (Rev. 2) Note 11 Augmented Main Loop Weld (Unit 2) 2-28 (Rev. 2) Note 12 Category C-A (Code Case N-435.1)

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2-29 (Rev. 0) Note 13 Class Two Heat Exchanger Esemptions 2-29.a (Rev. 0) Note 14 Augmented 88-06 Examinations 2-29.b 2.4 Exempt Components (Units 1 & 2) (Revision 2) 2-30 2.5 Inservice Inspection Drawings (Unit 1) (Revision 1) 2-33 I

2.6 Rollef Requests (Revision 2 and 3) 2-38 f

(Rev. 2) NR-1 Main Steam Saddle Plates 2-39 (Rev. 2) NR-2 Cast Stainless Elbows to Cast Pumps or Valves 2-41 (Rev. 2) NR-3 Class One Valve Internal Inspection 2-42 (Rev. 3)- NR-4 Rollef Deleted 2-44 (Rev. 2) HR-5 Cast Stainless Elbows to Carbon Nossles 2-46 (Rev. 2) HR-6 Reactor Coolant Piping-to-rittings and Valves 2-48 Limited Examinations (Cast Stainless)

(Rev. 2) NR-7 Class One Piping Limited Ultrasonic Examinations 2-50 (Rev. 3) NR-9 Reactar *!,ssel Shell Welds, Limited Examinations 2-51 (Rev. 2) HR-10 Letdown Heat Exchanger, Limited Examinations 2-59 (Rev. 2) HR-11 Escess Letdown Heat Exchanger, Limited Examinations 2-62 2-2 1680m(072390)/11

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Revisito 3 i

TABLE OF CC3f7ENTS (Continued)

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taen (Rev. 3) NR-12 Steam Generator (Secondary Side) and Residual Heat 2-66 Removal Hest Enchanger Mossle Inner Radii.

1 (Rev. 3) NR-13 Rollef Deleted 2-69 (Rev. 2) NR-14 Reactor Coolant Pump Internal Inspectica 2-70 (Rev.'2) NR-15 Centrifugal Charging Pump and Residual Heat 2-71 Removal Pump Inaccessible Attachment Wolds.

(Rev. 2) NR-16 Class Two Piping Limited Examinations 2-74 (Rev. 3) NR-17 Rollef Deleted 2-75 j

(Rev. 3) NR-18 Class One Inaccessible Welds 2-76 (Rev. 1) NR-19 Rollef Deleted 2-77 r

2.7 Program Tables (Unit 2) (Revision 0) 2-78 Alternate taams 216-218 of 235 2.8 Inservice Inspection Drawings (Unit 2) (Revision 0) 2 79 e

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2-3 1680m(072390)/12

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i The NRC has espressed a concern dealing with latergranular stress

(

corrosion cracking in lines that contain stagnant borated water.

Braidwood Station will perform augmented volumetric examinations on class two containment spray-(C.8) welds on Unita 1==d 2.-

The ~

j inspection shall include seven and one-half percent (7.5%) sampling of the welds in a single train between t.be (C.S.) pump and the first weld beyond the isolation valve inside containment.

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In addition to the esempted components listed in Section(s) 2.4 of the Braidwood Inservice Inspection Pisa, traidwood will. also be esempting the following " lass 2 vessels from the volumetric and surface esamination requarements of IWC-2500:

i 1

""I'***t' Amplicable Exemption Regenerative Heat Enchangers (4 total)

IWC-1220(c) l 1/2CV03AA and 1/2CV03AB velume Control Tanks (2)

IWC-1220(b) 1/2CV01T Reactor Coolant Filters (2)

IWC-1220(c) 1/2CV03F l

Seal Water Return Filters (2)

IWC-1220(c) 1/2CV02F Excess Letdown Heat Exchangers (4)

IWC-1220(c) i 1/2C01AA and 1/2CV01AB Letdown Roheat Heat Eschangers (2)

IWC-1220(c) 1/2CV05A Horizontal Letdown Heat Exchangers (4)

IWC-1220(c) i 1/2CV04AA and 1/2CV04AB The Volume Control Tanks' operate-at a pressure less than 275 psig and a temperature less than 200 degrees F.

The remaining above listed Class 2 vessels are esempt since the cumulative inlet and cumulative outlet pipe cross-sectional areas for these vessels do not exceed a 4" NPS cross-sectional area.

This position has been clarified in later versions of ASME Section XI as well as in Code Case N-408-2.

All these vessels receive a periodic pressure test (VT-2) which assures their structual and operational integrity.

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1 2-29.a 1732m(072390)/21

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j In response to WRC Bulletin 88-08 " Thermal Stresses in Piping.

Connected to Reactor Coolant Systems," Ultrasonic esaminations utilising IGSCC sensitivity will be performed at specific locations l

on the Pressuriser Aus111ary line and the RHR discharge lines from i

the RCS.

Esaminations shall be performed at alternating refueling outages.

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EYETEMI - Reactor Coolant i

?>

NUMBER OF ITEMS 4

Wald Mt=her.

Floure Nt=har Meld Mumher 1RV-02-002 1,2,3, 2RV-02-002 (RV-002)

(RV-002) i 1RV-03-001 4,5,6 2RV-03-001 (RVCH 001)

(RVCH-001) 3.

A.E.M.E. CODE c'A**

1 4.

A.S.M.E. CODE SECTION XI REQUIREMENTS:

Subsection IWB, Table IWB-2500-1, Esamination Category B-A, Items 81.11 and B1.40 equire volumetric esamination of the regions described in rigures IWB-2500-1 and 5 respectively for welds in the reactor pressure vessel each inspection interval.

I 5.

BAEIE FOR RELIEft Lower Shell Course-to-Dutchman weld RV-002 (1RV-02-002),

a.

(2RV-02-002) has sia core support guide lugs welded to the interior surface of the reactor vessel approsimately 3.50" above the weld.

These lugs restricted the automated inspection tool from inspecting the required volume in the areas of tre lugs, shown in Figure 1.

All of the weld and heat affected sono received 100% coverage from at least one direction, however the required base metal was not fully inspected in the area of the core support guide lugs.

Figures 2 and 3 show esactly what was inspected. Note that the dimensions used for actual coverage are for transducer position, not volume inspected.

b.

Closure Head Flange-to-Dutchman Forging Weld RVCH-001 (2RV-03-001), (IRV-03-001) has the fasnge which physically obstructs the ultrasonic transducer from performing the required scan area.

Part of the tnree larger lifting luge also fall in the required scan area.

Figures 4 and 5 show the position of the weld and flange. A detailed diagram of the transducer position for actual and required coverage is shown in Figure 6.

The code required surface exams will be

(

performed on the accessible tress.

6.

ALTERNATE TEST METHODt None i

2-51 1734m(072390)19

B;vis!O3 3 I

i RELIEF REC 1!EST BR-9 7.

JUST1FICATICRi t. The Reactor Vessel is esamined remotely using l

the lauwersion technique.

Completion of the remalaing portions of the above listed welds is impractical and would result in

{

undue hardship without a compensating lacrease in safety.

By j

performing the limited ultrasonic esaminations and the leakage

'l test each refueling outage, an adequate level of structural integrity can be assured for plant operation.

8.

APPLICABLE TIME PERIODI Relief will be required for the first j

120 month inspection interval.

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,_ _, I

g4 Revislos 3 RELIEF REC 1!EST MR-12 1.

SYSTEMI Residual Heat Removal (Residual Heat Removal Beat Enchanger).

2.

NUMBER OF ITEMS 6

Component Attachment Restricted Ma=her Wald Na=her Ma=kers Emma 1kH02AA RHKN-01, RHIN-02 1

Inner Radil and possle to Vessel Wold 2RH02AA RHKN-01, RHIN-02 1

Inner Radil and Nossle to Vessel Wold 3.

A.E.M.E. CODE cfACE:

2 4.

A.S.M.E. CODE SECTION XI REOUIREMENTS:

Subsection IWC, Table IWC-2500-1, Esamination Category C-8, Item C2.22 requires volumetric esamination of the nossle inner radius and-Item C2.21 requires volumetric and surf ace esamination of the Nossle to Shell weld of the regions described in Figure IWC 2500-4(a) or j

(b), for nossles without reinforcing plate in vessels >1/2 in.

I nominal thickness. Esaminations shall be conducted on nossles at terminal ends of piping runs selected for examination under Examination Category C-F, each inspection laterval.

5.

EkEIS.FOR RELIEFt The nossles listed above contain inherent geometric constraints which limit.the ability to perfora meaningful ultrasonic camminations.

The Residual Heat Removal Heat Enchanger is approximately 7/8 in, nominal wall thickness with nossles of 14 inch diameter and approximately 3/8 in. In nominal wall thickness.

The configuration is best characterised as a fillet welded nossle

' I using an internal reinforcement pad and, thereby is not analogous to a full pone'cration butt welded nossle as shown la Figure IWC-2520-4.

In addition, the inner radius of the reinforcement pad would be representative of the nossle laner

"~

radius required for inspection.

The inherent geometric constraints of the nossle design prevent the performance of the required ultrasonic esaminations of the nossle-to-shell weld and the nossle inner radius, see attachment 1.

1 2-66 1734m(072390)34 1

-a

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R;visies 3 f

RELIEr REautsT ma-12

.[

F 6.

ALTERNATE *LST ME M The welds listed above will receive the i

required Section XI surface examinations. Visual examination (VT-1) of the nossle inner rodil shall be performed olther directly or remotely to the eatent practical when disassembly is required for maintenance purposes not to escoed once per inspection. interval.

In addition, visual examination (vt-2) i shall be performed each inspection period on all pressure retaining components.

7.

JUSTIFICATION: The VT-1 eammination will assure early detection of detrimental flaws.

Therefore, in performing the proposed alternative examinations during disassembly for maintenance, an 1

adequate level of structural integrity can be assured for continued plant operation.

8.

APPLICABLE TIME PERIOD:

This relief will be required for the first 120 month inspection interval.

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SYSTEM ? Reactor Coolant 2.

NUMBER OF ITEME:

2 Line N=har Wald Number Interferrina condition 2?C29AC-10" 281-09-17 Permanent Restraint 2RC2 C 10 2SI-13-28 Permanent Restraint E

3..

A.S.M.E.' CODE CLASS: 1 i

4.

A.S.M.E.

CODE SECTI(M EI REQUIREMENTS:

Table IWB-2500-1,-

Examination Category B-J,'

Item 9.11 requires volumetric and surface esamination of'the areas described in~ Figure IWB-2500-8 I

for essentially 100% of the weld length.

5.

BASIS FOR RELIEF: Weld number 2SI-13-28 is encased in a permanent whip restraint making it inaccessible to both surface and volumetric examination, sne attachment 1.

Weld number o

. 2SI-09-17 is ' adjacent to a permanent whip restraint making it.

accessible for surface examinatin but inaccessible for-volumetric examination since it is a valve to pipe weld on the upstream side, see attachment 2.

6.

ALTERNATE TEST hETHOD:. None.

7.

JUSTIFICATICE:

The structural integrity of this weld shall be-insured by:

a.

Performing a surf ace eremination of the regions accessible.

~

b.

Performing a system leakage test each refueling outage and a

.i system hydrostatic-test each interval.

8.

APPLICABLE TIME PERIOD:

Relief will be required for the first

- 120 month inspection interval.

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1.. SYSTEM ' All non exempt' portions of.saftty related piping systems.

2.,

NUMBER _OF_ ITEMS:

Snubbers in tho'following systems -

System Snubbers Attached =to Insulated Pipe-Unit l' Unit 2-AB 2

1 AF' 1

0 CS 0

9 CV 82 91

'ys' m

DO.

0 10

(

FW :

4' 42 li g

MS-12 12

.I RC 93 80 RH 24-28 RY 31 26 SD L15 21 1

Total: snubbers i

on insulated lines 264 320.

A o

i Total snubbar

population -

369-

-384 h

Table showing the number.of snubbers on insulated ~1ines vs.-the total number' of snubbers in -the population.

The above numbers'will vary.with-time.as a result of snubber reduction and other plant modifications..

j 3.,

ASME CODE CLASSt

~1,'2 and 3.

p 1

4.

AEME SECTION XI CODE REOUIREMENTt -The component support examination?

4

' boundaries are defined by IWF-1300 and Figure 1300-1 Per INF-1300e, the 11WF support exam boundary.for anubbers which have non-integral attachments c

ertends if rom.the contact surf ace between the component and the support to -

g the1 surface of the building structure.

]

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5. ! BASIS'FOR RELIEF:

The-visual examination of an Integral or non-lutogral

[i

,c-pipe attachment is limited by the insulation' installed on the piping.

lt t '

?

,:0

- would impose a great deal of hardship in terms of manpower, time and

- radiation exposure io remove insulation to visually inspect all snubber pipe-clamps, particularly if there are alternative methods - that provide an M

- equivalent means of determining pipe clamp integricy.

u l

4-64 1500(072390)

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The majority of-snubbers are ' located inside' Containment in high radiation-O

' Removing -insulation on all snubber pipe clamps would: require one :

. areas.-

Health. Physics Technician 1 o survey the insulation prior to its' removal t

and then a two man lasulator crew to remove the insulation. ' This would -

add, three people to the customary two man' inspection crew, which would t

i more than double the man rem espo:Jures for performing. the surveillance.-.

The CECO SPPM.VT-3/4 procedure allows remote inspections to be performed.

+

su on.snubbersfthat are out of reach for direct inspection.

Scaffolds or man s

baskets would'have to be used to remove insulation ~on remote-snubbers.

'L Estra scaffolds built for snubber pipe clamp lasulation removal would.

increase congestion in containment and increase the amount of material being handled.and surveyed during the outage..

It would also-produce-(

additional DAW and result in more scaffold material acquiring fined contamination duringsthe outage.

It would pose an additional burden on A

the examination in terms of manpower,. time, safety and radiation exposure.

It also defeats'the purpose of the remote examination methods i

allowed in the SPPM.

(

6.

ALTERNATE TEST METHQQ1 ASME Section XI Code, IMA-2240, allows for.

L alternate. examination' methods if they provide results that are equivalent to the specified method.

In lieu of removing insulation on snubber. pipe clamps, the,following alternate exam methods will be employed'on all snubbers that are accessible for direct examinations A hands-on inspection of the pipe clamp. will'be performed to verify a..

the clamp is tight.

i,

' b.

. Clamp. alignment with.the load pin axis will be observed to verify i

-alignment is within design toleran ~s.

l The load pin / stud will be inspor

'l to verify its' integrity.

This 9

.i c.

will insure that parts are in pl.;e and that the pin is tight.

If-the load pin is' obscured by insulation, the insulation will be removed' or modified to allow f or this inspection.

d.

Insulation will'be checked for evidence of damage due to slipped or loose clamps, If boric acid contamination or corrosion is observed, insulation e.

'will be removed to inspect the pipe clamp.

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7.; JUSTIFICATIONI iThis relief-request Is intended for non-integral;

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attachments on insulated lines., The ' visual inspection of snubbers are

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' performed-using the CECO SPPM VT-3/4 procedure.

The-inspections arel m.

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performed en a11' safety related snubbers every 18 months a 25%. Under

.this' procedure,: support-indications,to be observed and documented include-the:following L

- cracks, pitting-

--erosion, corrosion, wear

- loose,-missing, damaged parts contamination, debris-Ji

- weld' degradation Jh

- slipped clamps

- arc strikes, weld spatter, paint

- clearances, settings

- condition of spherical bearings i

The proposed alternate exam methods listed in part 6 of the Rollef Request

. enhance tne SPPM VT-3/4 inspection procedure.

The' hands on check combined-with the VT-3/4 precedure will insure that snubber pipe clamps are installed and secure on snubbers that can be reached for direct examination.

If there are any indications of degradation, the insulation 2

will 'be removed to allow for a total. clamp inspection.

I For the snubbers.that have pipe clamps completely tar. 3 in insulation, i

the ip;olation will be removed for a complete inspoetion._ Based on a review of the piping line list and snubber data bases this condition occurs primarily on-PSA-1/4 anubbers attached.to 1 in, or let.J piping

-covered with 2 in. thick or greater insulation.

Insulation removal to

= inspect these pipe clamps will be' accomplished.several ways.. When these l

snubbers are removed for functional testing, the inFulation will have to

.f be removed to unpin the snubber...The visual inspection-offthe pipe clamp j

will be performed at this. time.

Visuals on pipe clamps.will be coordinated with NDE inspections when insulation removel'to access welds also exposes the. pipe clamp for inspection.. Snubber exams are documented-so it can:be verified that all snubber pipe clamps completely-buried'in insulation receivs a visua.1 inspection within the ten year interval.

On snubbers that are inaccessible for direct examination, a remote exam will be performed per the SPPM VT-3/4, procedure.

This would include f

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. snubbers in heat exchanger pits or on overhead runs of piping that are-out of reach.

The number of snubbers in this category represents only about 5% of, the snubbers in Units I and 2..

None of these snubbers have pipe

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clamps completely buried in insulation.

They are not in high traffic areas where typically the most clamp slippage and ot.'er damage is

-experienced.

Boric acid spray from valves or flanges is also less likely on-the majority 'of these snubbers.

Because they.are in relatively safe areas, an indirect inspection verifying rx> outward indications are present

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will demonstrate that the pipe clamps are secure. When these snubbers are functionally tested, scaffold will be built to access them.

A hands on t

esam of the pipe clump will be done at that time.

If any of the,above conditions are present, scaffold will he built to more thoroughly investigate the indication.

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Previous inspe'etion esperience has not shown an:3ncreasing trend in regard?

I to loose) pipe' clamps. A review of earlier data indicates:that'the

- incidence: of -loose pipe clamps.is rare.

The alternate methods proposed ini u

this~ relief request combined with the commitment to remove insulation on-l

' those pipe clamps; completely obscured by insulation ~will provide a complete-esamination of the entire snubber population in Units 1 and 2..

L Thisfapproach meets the requirements of an alternate inspection and willl i

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provide a high degree of confidence:that the sntbber pipe clamps are in place and~ secure.

By limiting'the number.of people required toLperform the surveillance, the proposed methods--will minimise man rem exposures'.>

They will eliminate.the need=for additional scaffolding, which will lower the amount of contaminated material produced during. the outage, and' reduce theitraffic and congestion associated with moving scaffolding.in and out; of Contaimment., This will promote the efficient and cost effective-execution of the visual'aurveillances.

8.

APPLICATIONt This request for relief applies to visual examinations (VT-3/4) of non-integral attachments to snubbers.:

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IMPLEMENTATIQH* - Examinations may be performed d'aring plant operations or normal plant' shutdowns, l

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.10.-TIME-PERIOD This request for relief applies for the first ten year inte rval~.

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