ML20065Q461

Forwards Description of Changes to Inservice Insp Program & Util Response to Remaining Item in NRC Re IGSCC in Lines That Contain Stagnant Borated Water

ML20065Q461
Person / Time
Site:	Braidwood
Issue date:	12/13/1990
From:	Checca A COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
References
ID139, NUDOCS 9012170213
Download: ML20065Q461 (7)

Text

,b Ctmmtnwealth Edisen a

-

• Z< 1400 Opus Place '

\U Downers Grove, Illinois 60515 December 13,1990 1

I Dr..T.E. Murley, Director Office Nuclear Reactor Regulation

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

-Attn: -Document Control-Desk

Subject:

Braidwood Station Units 1 and 2 Inservice Inspection Program (ISI)- ,

NRC_ Docket NosmS_0-156 and 50-452

.r

Reference:

(a) May,21, 1990 S.P. Sands letter  ;

to T.J. Kovach 4

-(b) August- 15, 1990 S.C. Hunsader letter to T.E. Murley Reference (a) provided the=NRC request for additional information in

-support of the-ongoing NRC review of the Braidwood Unit I and 2 ISI program.

In order to clarify what was required in this request, a teleconference was conducted on July 11,-1990 between Commonwealth Edison, the NRC staff and Idaho _ National Engineering Laboratory (I.N.E.L.). The data requested during thisiteleconference was pr_ovided in reference (b). One issue, left open on

'Julyfil, 1990, was~the topic of a teleconference between the same-parties, on November'14,=1990. _As a result of-this teleconference-certain' relief requests-andfnotes contained in;the Braldwood ISI Program have been revised or

deleted.

The enclosure describes.these-changes and.provides the Edison response to the remaining 1 tem requested in reference (a). Accordingly, ,1

enclosed

are new-revisions to be included in the NRC's-copy of the-Braidwood

'I SI--' Prog ram.

'Please address any questions concerning-this submittal to this-office.

Very truly yours, 6A A.% Checca Nuclear Licensing Administrator

/Imw/scl/ID139

-cc:' Resident; Inspector-BH I f

.S. Sands-NRR

.H,.Shafer-RIII 9012170213 901213 PDR O

ADOCK 05000456 PDR

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l NRC Request:-

InLCommonwealth Edison's September 18, 1986 submittal, the Licensee committed _to examine a random sampling of17.5% of the large bore-(greater than 4 inches) piping circumferential welds

,in the Safety Injection, Chemical and Volume Control, and

_ Containment Spray systems. _This submittal states that "These welds will be examined-over the ten year inspection interval as described in the ISI Program and will be tracked for the life of the plant." However, this commitment is 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-exemption criteria in Section XI.

Verify that volumetric examinations of a 7.5% sampling of the Class 2 piping welds in these systems will be performed, as identified in=the September 18, 1986 submittal, and that the ISI Program Plan will'be revised accordingly for both Units 1 and 2.

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

-Edison Response:

The_NRC has expressed a concern dealing with intergranular stress corrosion cracking (IGSCC) in lines that contain stagnant borated water. ;Braidwood Station will. perform augmented volumetric examinations on class two ECCS systems; Containment Spray-(CS),

Chemical and Volume Control (C.V.-), Residual Heat Removal (R.H.)

-and Safety _ Injection (SI) that are_not currently subject to volumetric-examination as required by code. The. inspections 2shall. include seven and one-half percent (7.5%) of the total population of-circumferential welds in piping greater than four

inches (4") nominal pipe size which:contains-stagnant borated

-water. Nominal pipe wall thickness and-pressure / temperature exemptions do not apply.

/scl:ID139:2

. R3 vision.4 TABLE OF CONTENTS 2.0 Inservice Inspection Program Plan for Hondestructive Examination Pa9t 2.1 Program Description (Revision 2) 2-4 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

-(Rev. 2) Note 1 Main Steam Nozzle Inner Radil 2-11 (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) 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-Pressuriter 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. 4) 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) 2-29 (Rev 0) ~ Note 13 Class Two Heat Exchanger Exemptions 2-29.a

-(Rev. 0) Note 14 Augmented 88-08 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 2.6~ Relief'Roguests (Revision 2 and 3) 2-38

'(Rev. 2) HR-1 Main Steam Saddle Plates 2-39 (Rev. 2) NR-2 Cast Stainless Elbows to Cast rumps or Vn1ves 2-41 (Rev. 2) HR-3 Class One Valve Internal Inspection 2-42 (Rev. 3) NR-4 Relief Deleted 2-44 (Rev. 2) HR-5 Cast Stainless Elbows to Carbon Hozzles 2-16 (Rev. 2) NR-6 Reactor Coolant Piping-to-Fittings and Valves 2-48 Limited Examinations (Cast Stainless)

(Rev. 2) HR-7 Class One Piping Limited Ultrasonic Examinations 2-50 (Rev. 3) NR-9 Reactor. Vessel Shell Welds, Limited Examinations 2-51 (Rev. 2) NR-10 Letdown Heat Exchanger, Limited Examinations 2-59 (Rev. 2) NR-11 Excess Letdown Heat Exchanger, Limited Examinations 2-62 2-2 1680m(112990)/11

'. Revision 4

- tLQTT J The NRC has expressed a concern dealing with intergranular stress corrosion cracking (IGSCC) in lines that contain stagnant borated water. _Draidwood Station will perform augmented volumetric exnminations on class two ECCS systems; Containment Spray (CS),

Chemical and Volume Control (C.V.), Residual Heat Removal (R.H.) and Safety Injection (SI) that are not currently subject to volumetric examination as required by code. The inspections shall include seven and one-half percent (7.5%) sampling > 4" Nominal pipe size of the total population of circumferential welds which contain stagnant borated water. Nominal pipe wall thickness and pressure / temperature exemptions do not apply.

2-27 1732m(112990)/18 l

R3 vision 4 RghlEr nrouEST wn-12

1. SYSIEM: Residual Heat Removal (Residual Heat Removal Heat Exchanger).

2. H1IMBER OF ITEMS 4 6 Component Attachment Restricted Numbe r _ Weld Number N umbe rs Er am 1RH02AA RHKN-01, RHXN-02 1 Inner Radil and Nozzle to Vessel Weld 2RH02AA RHKN-01, RHKN-02 1 Inner Radil and Hozzle to Vessel Weld

3. A.S.M.E. CODE CLASS: 2 g

4. A.S.M.E. CODE SECTION XI REQUIREMEHIS Subsection IWC, Table IWC-2500-1, Examination Category C-B, Item C2.22 requires volumetric examination of the nozzle inner radius and Item C2.21 requires volumetric and surface examination of the Nozzle to Shell weld of the regions described in Figure IWC 2500-4(a) or (b), for nozzles without reinforcing-plate in vessels >1/2 in.

nominal thickness. Examinations shall be conducted on nozzles at terminal ends of piping runs selected for examination under ExaA nation Category C-F, each inspection interval.

4

5. RASIS FOR RELLEft The nozzles listed above .contain inherent geometric constraints which limit the ability to perform meaningful ultrasonic examinations.

The Residual: Heat Removal Heat Exchanger is approximately 7/8 in nominal wall thickness with nossles of R14 inch' diameter and approximately 3/8 in. In nominal wall' thickness. The configuration is best characterized as a fillet welded nozzle

.using an internal reinforcement pad and, thereby is not analogous to a full penetration butt welded nozzle as shown in Figure IWC-2520-4. -In addition, the inner radius of the reinforcement pad would be representative of the nozzle inner radius required for-inspection. The inherent. geometric constraints of the nossle design prevent the performance of the requirad ultrasonic examinations of the nozzle inner radius, see attachmsut 1. A nozzle-to-shell UT may not achieve full ASME coverage cf the weld area.

2-66 1734m(082790)34

l

' Ravision 4 EELIEf_ EEQUISIJiR _12

6. ALIEENATE TESIjiElllQD: The welds listed above will receive the required Section XI surface examinations. Visual examination (VT-1) of the nozzle inner radil shall be performed either directly or remotely to the extent practical when disassembly is required for maintenance purposes not to exceed once per inspection interval. In :.0 21 t

• c , e ' r 't ' u.T.ination (VT-2) shall be performed each inspection period on all pressure retaining components. A best effort UT shall be performed on the RHR IDC notale-to-shell welds on a frequency consistent with ASME Section XI.

7. RSIlELChI10ti The VT-1 examination will assure early detection of detrimental flaws. Therefore, in performing the proposed alternative examinations during disassembly for niaintenance, an adequate level of structural integrity can be assured for continued plant operation.

8. AEP.LICABLE TIME PERIOD: This relief will be required for the first 120 month inspection interval.

2-67 1734m(082790)35

4 Revision 4 o'

NR-12 Attachtnent 1 unir 1 wo uwtr 2 -

W = : T --* (NOT TO SCALE)

VESSEL WALL T = 0.875 H0ZZLE -

/

/' (TYR)

N.0.375 LA\ \\\\ N a

-REINFORCEMENT PAD Wal.25

., (SEE DETAll A) 1/4 a 4F CHAMFER T YR(4) CORNERS

~ 2 3* ---*

I i , DIMENSIONS SHOWN ARE NOMINAL 20' -

4 - -

q ssett Axis OETAll A REINF PAD

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