ML20043B235
| ML20043B235 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/21/1990 |
| From: | Sands S Office of Nuclear Reactor Regulation |
| To: | Kovach T COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 9005250123 | |
| Download: ML20043B235 (11) | |
Text
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May 21, 1990 Docket Nos. STN 50-456 DISTRIBUTION:
and STN 50-457 LDocket r11esSNRC & Local PDRs PDIII-2 r/f JZwolinski RDudley-SSands CMoore OGC-WF1 EJordan ACRS(10)
Mr. Thomas J. Kovach PDIII-2 Gray Nuclear Licensing Manager Commonwealth Edison Company-Suite 300 OPUS West III 1400 OPUS Place Downers Grove, Illinois 60515
Dear Mr. Kovach:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION CONCEPHING BRAIDWOOD NUCLEAR POWER STATION, UNIT NOS. 1 AND 2 FIRST TEN-YEAR INSERVICE INSPECTION PROGRAM The NRC staff, with assistance from its contractor Idaho National Engineering Laboratory'(INEL), is currently evaluating the first ten-year interval Inservice Inspection Program for Braidwood Station, Units 1 and 2.
Enclosed is a request for additional information which is needed for the NRC staff to complete its review. After discussion with your staff, it was agreed that the response to the enclosed list of questions could be provided within 45
~ days from receipt of this letter.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under Pub. L.96-511.
Sincerely,
/s/
Stephen P. Sands, Project Manager Project Directorate III-2 D' vision of Reactor Projects - III, iV, V and Special Projects Offr.e of Nuclear Reactor Regulation
Enclosure:
As stated cc:
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Docket Nos. STN 50-456 and STN 50-457 Mr. Thomas J. Kovach Nuclear Licensing Manager Comonwealth Edison Company-Suite 300 OPUS West III 1400 OPUS Place Downers Grove, Illinois 60515
Dear Mr. Kovach:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATI0li CONCERNING BRAIDWOOD NUCLEt,R
. POUER STATION, UNIT N05.1 AND 2 FIRST TE!!-YEAR INSERVICE IKSPECTION PROGRAM The NRC staff, with assistance from its contractor Idaho National Engineering Laboratory (INEL), is currently evaluating the first ten-year interval Inservice Inspection Program for Braidwood Station, Units 1 and 2.
Enclosed is a reccest for additional information which is needed for the NRC staff to
-complete its review. After discussion with your staff, it was agreed that the response to the enclosed list of questions could be provided within 45 days from receipt of this letter.
The reporting and/or recordkeeping requirements contained in this letter-affect fewer than ten respondents; therefore, OMB clearance is not required under Pub. L.96-511.
Sincerely,
~
[4 4L' Steph P. Sands, Project Manager Project Directorate III-2 Division of Reactor Projects-- III, IV, Y and Special Projects Office of Nuclear Reactor Regulation e
Enclosure:
As stated cc: See next page
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9 Mr. Thoras J. Kovech Braidwood Station Commonwealth Edison Company Unit Nos. 1 and 2 cc:
Mr. Gabe Toth Ms. Lorraine Creek Westinghouse Electric Corporation Rt 1, Box 182 Energy Systems Business Unit Manteno, Illinois 60950 Post Office Box 355 Bay 236 West Pittsburgh, Pennsylvania 15230 Douglass Cassel, Esq.
109 N. Dearborn Street Joseph Gallo, Esq.
Chicago, Illinois 60602 Hopkins and Sutter 888 16th Street, N. W.
Mr. Thomas W. Ortciger, Director Suite 700 1111nois Emergency Services Washington, D.C. 20006 and Disaster Agency 110 East Adams Street Ms. Bridget Little Rorem Springfield, Illinois 62706 Appleseed Coordinator 117 North Linden Street Michael Miller, Esq.
Essex, Illinois 60935 Sidley and Austin One First National Plaza Mr. Edward R. Crass Chicago, Illinois 60690 Nuclear Safeguards and Licensing Division George L. Edgar Sargent & Lundy Engineers Newman & Holtzinger, P.C.
55 East Monroe Street 1615 L Street, N.W.
Chicago, Illinois 60603 Washington, D.C. 20036 U. S. Nuclear Regulatory Commission Attorney General Resident Inspectors Office 500 South 2nd Street RR fl Box 79 Springfield, Illinois 62701 Braceville, Illinois 60407 Regional Administrator, Region III EIS Review Coordinator U. S. Nuclear Regulatory Commission EPA Region V 799 Roosevelt Road, Bldg. #4 230 S. Dearborn Street Glen Ellyn, Illinois 60137 Chicago, Illinois 60604 Chairman Illinois Department of Will County Board of Supervisors Nuclear Safety Will County Board Courthouse Office of Nuclear Facility Safety Joliet, Illinois 60434 1035 Outer Park Drive Springfield, Illinois, 62704 Robert Neumann Office of Public Counsel State of Illinois Center 100 W. Randolph Suite 11-300 Chicago, Illinois 60601 I
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X, COMMONWEALTH EDISON COMPANY BRAIDWOOD NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NUMBERS 50 456 AND 50 457 MATERIALS AND CHEMICAL ENGINEERING BRANCH l
DIVISION OF ENGINEERING AND SYSTEMS TECHNOLOGY j
Reauest for Additional Information - First 10 Year Interval-Inservice Insoection Prooram Plan 1.
Scoce/ Status of Review 1
Throughout the service life of a water-cooled nuclear power facility, 10 CFR 50.55a(g)(4) requires that components (including supports) that are classified as American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Class 1, Class 2, and Class-3 meet the requirements, except design and access provisions and preservice examination requirements, set forth in the ASME Code Section XI, " Rules for Inservice Inspection of Nuclear Fower Plant Components," to the extent practical within the limitations of-design, geometry, and materials of construction of the components. This section of the regulations also requires that inservice examinations of components and system' pressure tests conducted during the initial 120-month inspection interval-comply with the requirements in the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to-the date of issuance of the operating license, subject to the limitations and modifications listed therein.
The comporents (including supports) may meet requirements set fort,h in sub' sequent editions and addenda of this Code which are incorportted by reference in 10 CFR 50.55a(b) subject to the limitations and modifications listed therein. The Inservice Inspection (ISI) Program Plan has been prepared to meet the requirements of the 1983 Edition, Stimmer 1983 Addenda of the ASME Code Section XI except that the extent of examination for Code Class 2 piping welds in Residual Heat Removal i_
(RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) i systems has been determined by the 1974 Edition through Summer 1975 Addenda as required by 10 CFR 50.55a(b).
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l As required by 10 CFR 50.55a(g)(5), if the Licensee determines that
-certain Code examination requirements are impractical and requests relief, the Licensee shall submit information to the Nuclear Regulatory Commission (NRC) to support that determination.-
The staff has reviewed the available information in-the Braidwood Nuclear Power Station, Units 1 and 2, First 10-Year Interval ISI Program Plan, Revision 2, and the requests for relief from the ASME Code Section XI requirements which the Licensee has determined to be impractical.
2.
Additional Information/ Clarification Reauired Based on the above review, the staff has concluded that the following information and/or clarification is required in order to complete the review of the ISI Program and relief requests:
A.
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.
B.
Address the degree of compliance with NRC Regulatory Guide 1.150,
" Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations."
C.
Note 10 of Section 2.3 of the ISI Program Plan states "The NRC has expressed a concern dealing with intergranular stress corr'osion 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 (7.5%) 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 be revised to include examinations for Braidwood, Unit 2, also.
2 i
D.
In Commonwealth Edison's September 18, 1986 submittal, the Licensee-committed to examine a random sampling of 7.5% of the large bore (greater than 4 inches) piping circumferential welds in the, Safety '
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Injection,. Chemical and Volume Control, and Containment Spray
_ systems. This submit,tal states that "These weldt 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 Nucient 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 not permit pressure / temperature exemptions for RHR and ECC systems.
E.
Relief Request NR-4:' Relief is requested from performing the Code-required volumetric examination of the Pressurizer and Steam-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 relief ~ request NR-3 was removed from the Byron ISI Program. The letter states that Commenwealth Edison will continue to evaluate the feasibility of performing the examination.
Relief Request NR-3 for Byron, Unit 1, discusses attempts to develop an applicable ultrasonic examinaticii technique using a mockup. How are the. attempts to perform the examinations for Byron applicable for the Braidwood plants? Also, provide a sketch which shows the location of the closure ring with respect to the steam generator primary nozzles listed in Relief-Request NR-4.
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F.
Relief Request NR 9:
Confirm that the Code required surfaee examination will be performed on the Reactor Pressure Vessel head-to-flange welds.
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G.
Relief Request NR 12: Steam Generator Nozzles:
Relief is requested from performing the Code-required volumetric examination of the nozzle inside radius sections of the Steam Generator main feedwater-nozzles (Units I and 2). The drawing attached to the relief request shows that the main feedwater nozzle was designed with an internal multiple venturi ~ type flow restrictor. The flow restrictor area does not utilize a radiused nozzle as described in Figure IWC-2500 4, but instead has several individual inner radii, corresponding to each venturi. Therefore, the Code-required volumetric examination of these nozzle inner radius sections is impractical to perform.
However, the base of the nozzle is consistent with the radiused nozzle described in Figure IWC 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 requested from performing the Code-required volumetric examinations of the nozzle-to-shell welds and nozzle inside radius sections of the subject nozzles.
Is the design of these nozzles consistent with the sketch in Attachment B, Page 5 of 5, of the July 11, 1989 submittal for Byron Station, Unit I? If not, please provide a sketch showing the design of the subject nozzles.
H.,
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 4
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4 effort" volumetric examination were performed, what percent of the Code required volume could and would be examined?
1.
Relief Request NR-17:
Relief is requested from performing the Code-required volumetric examination of Class 2 pressure vessel shell and head circumferential welds with wall 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 less than or equal to 2 inches, ultrasonic examination may be performed using the rules of the Winter 1985 Addenda of Section XI, Division 1, Appendix Ill.
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.
J.
Relief Request NR-18:
Relief is requested 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 partial volumetric examination?
K4 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 251-02-45?
Describe a "best effort" volumetric examination that could be performed on the subject weld.
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J Relief Request CR 1::-Relief is? requested from removing the L.
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l insulation from all nonexempt component supports on Code Class 1 Y 1-and 2' insulated lines for the sole purpose of performing a visual examination.on the portion of'the 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 1 and Class 2 insulated lines in the AF, CV, FW, MS, RC, RH, RY, and SI systems.
The regulations do not provide for granting generic relief requests.
Paragraph IWF-1300(e) states that "Where the mechanical connection of a nonintegral 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 on the Code requirement for those components not excluded based on IWF-1300(e). Therefore, the Licensee should provide additional informt. tion describing 1%
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analyses performed to determine which of the component supports may
'oe excluded based on IWF-1300(e) and evaluate the remaining supports to determine which supports may require relief from the Code required examination along with the technical justifications.
Relief.is not required for component supports which are exenpted based on IWF-1300(e).
For those supports remaining after the above exemption, 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 6
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failure of the-component support would be obvious should the support fail with the insulation installed. The method for determining item (b)-above should 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 which-are exempted based on IWF-1300(e), and (b): the Licenset should 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 1 and 2, First 10-Year
. Interval-ISI Program Plan.
M.
Relief Request SR-1:
Relief is requested from removing the insulation from all nonexempt safety-related snubbers on Code Class 1 and 2 insulated lines for the sole purpose of performing a visual examination on the portion of the integral or nonintegral attachment within the insulation.
Paragraph 1WF-1300(e) states:
"Where the mechanical conne'etion of a nonintegral 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 of the 7
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4" component supports may be excluded based on IWF-1300(e), and
,j evaluate the: remaining supports to determine which supports may require relief from the Code-required examination along with the technical justifications.
Relief is-not required for component supports which are exempted based on-IWF-1300(e).
For those supports remaining after the above exemption, 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. The method _for determining item (b) above should be described in the request for relief.
If the above cannot be technically justified,-the insulation should be removed for the examination.
Based on the above, it is concluded that:
(a) the Licensee should not ask for relief for supports which may be exempted based on IWF-1300(e); and (b) the Licensee should provide the additional information-as described above as part of the technical justification for granting relief.
- It is noted that Byron relief request SR-1, which requests the 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 add'itional justification or consider withdrawing Relief Request SR-1 from the Braidwood, Units 1 and 2, First 10-Year Interval ISI Program Plan.
The schedule for timely completion of this review requires that the Licensee provide, by the requested date, the above requested information and/or clarifications with regard to the Braidwood Nuclear Power Station, Units 1 and 2, First 10-Year Interval ISI Program Plan, Revision 2.
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