ML20094P991
| ML20094P991 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/22/1995 |
| From: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML19317C138 | List: |
| References | |
| 95-605, NUDOCS 9511300296 | |
| Download: ML20094P991 (7) | |
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VINGINIA ELecTHIc AND Pownn COMPANY RICIIMOND, VIRGIN!A 20261 November 22,1995 U.S. Nuclear Regulatory Commission Serial No.95-605 Attention: Document Control Desk NES/ISI/PJN/EJW Washington, D.C. 20555 Docket No. 50-338 Ucense No. NPF-4 Gentlemen:
VIRGINIA El ECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 REACTOR VESSEL HEAD PENETRATIONS USE OF AN ALTERNATIVE REPAIR TECHNIQUE Inspections at pressurized water reactors have shown the presence of cracking in some roactor vessel head penetration tubes. This phenomenon has been followed closely by the Nuclear Energy Institute (NEI) and the owners groups. Baseo on a reassessment of the phenomenon by.he Westinghouse Owners Group (WOG), it appears that Virginia Electric and Power Company's (Virginia Powers) North Anna and Surry Power Stations may be more susceptible to the cracking mechanism than previously believed.
Because of the slow rate of crack growth and relative ease of detection, the issue appears to have a low safety significance but potential economic risk.
As a precautionary measure, it is our intent to perform a limited inspection at this time. The results of those inspections will be used to refine the WOG guidelines for reactor 4
vessel head penetration tube cracking and to determine the necessity for similar inspection activities in the future at North Anna and Surry.
Virginia Power currently plans to inspect North Anna Unit 1 during the February 1996 refueling outage. In the unlikely event that repairs are required as a result of the inspection, we request, pursuant to 10 CFR 50.55(a)(3), the use of the attached Alternative to Code Requirements. We also request that the NRC's review and approval of this alternate repair technique occur prior to mid-January 1996 in order to facilitate the upcoming North Anna Unit 1 outage, currently scheduled to begin February 9,1996.
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I The following documents are referenced in the attached Alternative to Code Requirements and are enclosed:
l
- 1. WCAP-13998, Rev.1, "RV Closure Head Penetration Tube ID Weld Overlay Repair" l
(Proprietary) (5 copies)
- 2. WCAP-14519, "RV Closure Head Penetration Tube ID Weld Overlay Repair" (Non-Proprietary) (5 copies)
Reector Vessel Head Adapter Tube Cracking," November 19,1993.
- 4. USNRC Letter, A.G. Hansen to R.E. Unk, " Acceptance Criteria for Control Rod Drive j
Mechanism Penetrations at Point Beach Nuclear Plant, Unit 1," March 9,1994.
I Also enclosed are a Westinghouse authorization letter, CAW-95-906, accompanying affidavit, Proprietary Information Notice, and Copyright Notice.
As item 1 contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit establishes the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations. Accordingly, the information which is proprietary to Westinghouse should be withheld from public disclosure in accordance with 10 CFR Section 2.790 of the Commission's regulations.
Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-95-906 and should be addressed to N. J. Liparulo, Manager of Nuclear Safety Regulatory &
Liconsing Activities, Westinghouse Electric Corporation, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.
The attached relief request has been approved by the Station Nuclear Safety and Operating Committee. If you have any questions concerning this request, please contact us.
Very truly yours,
)
,,y James P. O'Hanlon Senior Vice President - Nuclear Attachment
)
oc: U.S. Nuclear Regulatory Commission Region 11 101 Marietta Street, N.W.
Suite 2900 Atlanta, Georgia 30323 r
Mr. R. D. McWhorter NRC Senior Resident inspector North Anna Power Station l
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i ALTERNATIVE TO CODE REQUIREMEnlIS j
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- 1. IDENTIFICATION OF COMPONENTS Drawina - 11715-WMKS-RC-R-1.2 Gagg_1
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i j-Sagi Panatration #
Description 3
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Initial Sample Group 13 62 - 69 4" control rod drive tube (sleeved) l 12 58 - 61 4" control rod drive tube (sleeved) l 11 51,53,54,57 4" thermocouple tube (not sleeved) l 50,52,55,56 4" control rod drive tube, spare (not sleeved) r i
i 1
10 46 -49 4" control rod drive tube (sleeved) 9 38 - 45 4" control rod drive tube (sleeved) i 8
30 - 37 4" control rod drive tube (sleeved) l' j
7 26 - 29 4" control rod drive tube (sleeved) 6 22,23,24,25 4" control rod drive tube, rods removed (not sleeved) 5 15,17,19,21 4" control rod drive tube, spare (not sleeved) 4 10-13 4" control rod drive tube (sleeved) 3 6-9 4" control rod drive tube (sleeved) j.
2 2-5 4" control rod drive tube (sleeved) j-1 1
4" control rod drive tube, rods removed (not sleeved) 1 Ring number identifies the distance from the center of the reactor vessel head. The higher the ring i
l number the greater the distance from center and a higher probability of finding a flaw.
j l
2 xpansion scope - If an unacceptable flaw is found in the initial sample group, then the next rih 0 will be l
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examined Expansion will continue until all the penetration tubes in a ring are found to be acceptable.
I i
l 11.
IMPRACTICAL CODE REQUIREMENTS i
i The North Anna Unit 1 reactor vessel closure head penetrations are scheduled to be
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examined during the 1996 refueling outage, as shown above. The initial inspection j
scope will include the twenty penetrations in the outer three rings. The closure head penetration tube base material in the region of the attachment weld will be oxamined volumetrically using eddy current. Any identified flaws will be characterized by
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ultrasonics. There are no inservice acceptance standards established for this area since this examination is not required by ASME Section XI,1983 Edition, Summer 1983 addenda. As allowed by subparagraph IWA-3100(b) "If acceptance standards for a particular component, Examination Category, or examination method are not specified in this Division, indications that exceed the acceptance standards for materials and welds specified in the Section lli edition applicable to the construction of the component shall be evaluated to determine disposition. Such disposition shall be subject to review by the enforcement authority having jurisdiction at the plant site."
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i Acceptance criteria have been established by Westinghouse and reported in WCAP 14024, " Inspection Plan Guidelines for Industry / Plant inspection of Reactor Vessel Closure Head Penetration Tubes." The acceptance criteria have been reviewed and accepted by the NRC1,2, with comments. The NRC comments have been incorporated in WCAP 14024. Virginia Power and Westinghouse are developing repair techniques in the event repairs are required. The Code requires flaws exceeding the acceptance criteria to be removed or reduced to an acceptable size, as stated in subparagraph IWB-3112(c) " Components whose examination (IWB-2200) reveals flaw indications, other than the indications of (b) above, that exceed the standards of Table IWB-3410-1 shall be unacceptable for service unless such flaws are removed or repaired to the extent necessary to meet the allowable flaw Indication standards prior to placement of the component in service."
Thermal sleeves are installed in 48 of the 65 reactor vessel head penetration tubes.
Due to the penetration configuration and the available tooling, complete removal of flaws greater than 0.25 inches deep requires the removal of the thermal sleeve.
i Removal and reinstallation of the thermal sleeve is a very difficult process. Any removal and reinstallation method involves special tooling, a significant amount of remote machining / welding, radiation exposure, and uncertainty.
Ill.
BASIS FOR ALTERNATIVE TO CODE REQUIREMENTS An alternative to removing the thermal sleeve and totally removing the flaw is to partially removo the flaw and weld overlay to the original wall thickness. This technique is referred to as an " embedded flaw repair." This repair technique is described in the Westinghouse Annotated Letter and WCAP 13998 (attached), entitled "RV Closure Head Penetration Tube ID Weld Overlay Repair."
The weld overlay eliminates the exposure of the flaw to the reactor coolant environment, which stops further flaw growth and results in a subsurface flaw as defined by ASME Section yl, IWA-3320. Acceptance standards for flaws will be based on the NEl/NUMARC guicelines.
The penetration tube is sufficiently stiff, and constrained by the vessel head, so the integrity of the tube will be maintained by the weld overlay regardless of the extent of the flaw.
The other advantages to this type of repair verses a Code repair is that this technique results in lower residual stress than a complete excavation with a full weld build up and a better curface fcr reinspection than a complete excavation and a partial weld build up. Therefore, it is also advantageous to use this technique for unsleeved penetrations. Additionally the development of analysis and tooling for a single versatile repair technique is preferred.
1USNRC Letter, W.T. Rur* ell to Raisin, NUMARC, " Safety Evaluation for Potential Reactor Vessel Head Adapter Tube Cracking." November 19,1993.
2USNRC Letter, A.G. Hansen to R.E. Link, " Acceptance Criterla for Control Rod Drive Mechanism Penetrations at Point Beach Nuclear Plant, Unit 1," March 9,1994.
2
IV.
ALTERNATIVE TO CODE REQUIREMENTS i
The embedded flaw repair method, proposed and supported by the stated Westinghouse documentation, will be used as an alternative to the Code requirements If repairs are required, for axial flaws up to 75% through-wall in reactor vessel head penetration tubes. The flaw will be partially removed using electric discharge machining (EDM). The excavation will be based on the depth of the measured flaw and will range from 0.090 to 0.125 inches. A weld overlay will be performed to restore the tube wall thickness. The final weld will be examined volumetrically using eddy current and ultrasonics and surface examined using liquid penetrant. The reactor vessel head will be VT-2 examined without removing the insulation during startup at nominal operating pressure.
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Q Westinghouse Energy Systems Nuclear Technology Division '
Electric Corporation 3,333 Pittsburgh Pennsylvania 15230 0355 November 20,1995 CAW-95-906 Document Control Desk
' US Nuclear Regulatory Commission Washington, DC 20555 t
Attention: Mr. William Russell i
- APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE i
Subject:
"RV Closure Head Penetration Tube ID Weld Overlay Repair," WCAP-13998, Rev.1 (Proprietary)
I
Dear Mr. Russell:
I The proprietary information for which withholding is being requested is further identified in Affidavit CA W-95-906 signed by the owner of the proprietary information, Westinghouse Electric Corporation.
Te affidavit, which accompanies this letter, sets forth the basis on which the information may be l'
withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Virginia Power Company.
3 Correspondence with respect to the proprietary aspects of the application for withholding or the
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Westinghouse affidavit should reference this leter, CAW-95-906, and should be addressed to the undersigned.
j Very truly yours, b
k N. J. Liparulo, Manager Nuclear Safety Regulatory & Licensing Activities
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Enclosures ec:
' Kevin Bohrer/NRC (12H5) j _._
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I CAW-95-906 AFFIDAVIT i
a COMMONWEALTH OF PENNSYLVANIA:
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COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared James M. Brennan, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on 1
behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth j
in this Affidavit are true and correct to the best of his knowledge, information, and belief:
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M. Brennan, Manager Operating Plant Licensing Sworn to and subscribed before me this 2'rf day of 7
,1995 i
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y Notary Public manws w Lenalne M.
Rtge.
Mcsvoevee Boro.
County MyCommission Expron 14.1995 Hwmor,PenruyWwaAusomeonofhoues It$2C-SRO-1:ll3NS
. CAW-95-906 l
i (1)
I am Manager, Operating Plant Licensing, in the Nuclear Technology Division, of the f
Westinghouse Electric Corporation and as such, I have been specincally delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.
1 (2)
I am making this AfCdavit in conformance with the provisions of 10CFR Section 2.790 of the Commission's regulations and in conjunction with the Westir ghouse application for withholding accompanying this Affidavit.
)
(3)
I have personal knowledge of the criteria and procedures utilized by the Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or Gnancial information.
(4)
Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission a determining whether the information sought to be withheld from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a ratiopsi basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
l 4
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
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iB52C-SRO-2112095
L CAW-95-906 1
t (a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a i
competitive economic advantage over other companies.
4 i
(b)
It consists of supporting data, including test data, relative to a process (or i
component, structure, tool, method, etc.), the application of which ' data
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secures a competitive economic advantage, e.g., by optimization or improvd l
marketability.
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l (c)
Its use by a competitor would reduce his expenditure of resources er improve i
,i his competitive position in the design, manufacture, shipment, installation, i
l' assurance of quality, or licensing a similar product.
i (d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or costomer funded development plans and programs of potential commercial value to l
(f)
It contains patentable ideas, for which patent protection may be desirable.
i (g)
It is not the property of Westinghouse, but must be treated as proprietary by
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Westinghouse according to agreements with the owner.
i There are sound policy reasons behind the Westinghouse system which include the i
following:
i (a)
The use of such information by Westinghouse gives Westinghouse a
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competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
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a
. CAW-95-906 (b)
It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse
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I ability to sell products and services involving the use of the information.
(c)
Use by our competitor would pu+ Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
4 (iv)
'Ihe information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is i
appropriately marked in "RV Closure Head Penetration Tube ID Weld Overlay Repair", WCAP-13998 Rev.1 (Proprietary), November,1995 for North Anna Power Station Units I and 2, being transmitted by the Virginia Power Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the senc snanas 4
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. CAW-95-906 Document Control Desk, Attention Mr. William T. Russell. The proprietary information as submitted for use by Virginia Power Company for North Anna Power Station Units 1 and 2 is expected to be applicable in other licensee submittals in response to certain NRC requirements for potential reactor vessel head penetration i
repairs.
This information is part of that which will enable Westinghouse to:
(a)
Provide data supporting the acceptability of repairing reactor vessel head penetrations utilizing the " embedded flaw" technique.
(b)
Define the concept and benefits of the reactor vessel head penetration
" embedded flaw" weld repair approach.
(c)
Assist the customer to obtain NRC approval.
Further this information has substantial commercial value as.'ollows:
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(a)
Westinghouse plans to sell the use of similar information to its customers for l
purposes of meeting NRC requirements for licensing documentation.
l (b)
Westinghouse can sell support and defense of the technology to its customers in the licensing process.
i r
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar methodologies and licensing defense services for I
commercial power reactors without commensurate expenses. Also, public disclosure 1
of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
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1ssac-sRGS:n2005
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, CAW-95-906 -
f De development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
f In order for competitors of Westinghouse to duplicate this information, similar i
technical programs would have to be performed and a significant utanpower effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests.
t Further the deponent sayeth not.
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Copyright Notice'-
l De reports transmitted herewith each bear a Westinghouse copyright notice. De NRC is permitted
= to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance,' denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding
. restrictions on public disclosure to the extent such information has been identified as proprietary by -
.i L Westinghouse, copyright protection not withstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is j
insufficient for this purpose. Copies made by the NRC must include the copyright noi!ce in all instances and the proprietary notice if the original was identified as prop *.ietary.
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CAW 906/NSRLASS4L A
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Proprietary Information Notice i
Transmitted herewith is a proprietary document furnished to the NRC in connection with requests for ~
j generic and/or plant-specific review and approval.
I In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning I
- the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the Information that was contained within the brackets in the proprietary versions having been deleted).
The justification for claiming the information so designated as proprietary is indicated in both versions I
by means of lower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary i
or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).
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4 CAW 90WN9tLA5541.
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C VRA-95-121 C
Ba 355 t
Westinghouse Energy Systems Pinsburgh Pennsylvania 15230-0355 Electric Corporation November 20, 1995 i
Mr. R. W. Calder Reft RM07-1599 l
Supervisor - Materials Engineering Ref RM06-1602 Virginia Power Raft RM30571 Innsbrook Technical Center Reft BKI*14411 5000 Dominion Boulevard Refs VRA-95-122 Glen Allen, VA 23060 VIRGINIA POWER NORTH ANNA POWER STATION UNITS 1 AND 2 ANNOTATED LETTER ON REACTOR VESSEL HEAD PENETRATION EMBEDDED FLAW REPAIR Dear Mr. Calder See the eight (8) page attachment which provides a discussion, summary and conclusions for the reactor vessel head penetration embedded flaw repair.
)
If you have any questions or require anything further, please r
call me at (412) 374-3370.
Very truly yours, l
WESTINGHOUSE ELECTRIC CORPORATION i
D. R.
Beynon, Jr., Project Manager Chesapeake /Pittsburgh Area l
Operating Plant Programs l
Attachment i
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i ATTACEMENT TO WESTINGHOUSE LETTER VRA-95-121 l l
aus A.
Background
Inspections have shown the presence of cracking in reactor vessel j
i head penetration tubes in a number of-pressurised water reactors.
The cause of this cracking has been attributed to primary water s
j stress corrosion cracking (PWSCC).
Several methods are available for performing repairs to the penetration tubes should cracking j
be significant enough to warrant repairs.
These methods include excavation of the penetration tube to remove shallow flaws and, for deeper flaws, excavation and weld repair.
With respect to excavation and weld repair, two methods are available.
These
{
methods would be to 1) completely remove the crack by excavation l
1 followed by a full or partial weld buildup, and 2) partial j
removal of the flaw by excavation followed by a weld overlay
(" embedded flaw" repair).
4 B.
Introduction 1.
Wald Build-up Repair Technique 1
j Several issues are associated with the case of complete removal l
of the flaw followed by a weld buildup that have an undesirable affect on site schedule, personnel exposure, and component
}
adequacy-for continued operation.
These are discussed in the following paragraphs.
- a. Thermal Sleeve Removal Due to the spacial constraints associated with the j
design of the vessel penetration and thermal sleeve, thermal sleeve removal is necessary to completely remove a flaw that is deeper than 0.25 inches for those j
i penetrations which contain thermal sleeves.
Removal of i
l the thermal sleeve can be achieved by two methods.
The first method removes a portion of the thermal sleeve through the bottom of the penetration.
To accomplish this, first the thermal sleeve is cut at an elevation above the crack in the penetration.
- However, i
the distortion and ovality of the penetration produced i
by the original attachment weld may not permit removal of the thermal sleeve. The thermal sleeve contains an i
j alignment collar that has a small clearance to the penetration ID and may not pass through the bottom and without cutting the thermal sleeve into segments.
Following this cutting and removal, the repair is made to the penetration and the thermal sleeve subsequently reinstalled.
This reinstallation requires remote welding of the thermal sleeve followed by inspection to verify an acceptable weld as well as correct alignment.
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ATTACKMENT TO WESTINGHOUSE LETTER VRA-95-121 Although the technique for cutting and rowelding of the thermal sleeves has been developed in Europe, additional development and qualification of this process by Westinghouse would be required prior to its use at North Anna.
For those penetrations with ovality and distortion that will not permit thermal sleeve removal through the bottom end, the second method is to remove the thermal sleeve through the top of the penetration. This method requires remeval of the CRDM rod travel housing by cutting the canopy seal weld and threading the rod.
travel housing out of the CRDM latch housing, cutting the thermal sleeve above the thermal sleeve guide, and removal of the remaining thermal sleeve out of the top of the penetration.
Following the repair, it is i
necessary to reinstall the thermal sleeve through the top of the penetration, threading the guide to the bottom of the thermal sleeve and welding it to the thermal sleeve, reinstalling the rod travel housing and reweld the canopy seal weld.
Both of these methods involve a significant amount of remote machining / welding and radiation exposure associated with the removal and installation of the 4
l thermal sleeve.
b.
Penetration Residual Stress / Inspection Following Repair One method for application of the weld buildup is to i
completely fill the excavation and restore the ID of the penetration.
While this method provides a surface that can be readily inspected following repair, it will require the application of a significant amount of weld material which results in a significant increase in l
penetration residual stress which could adversely affect the susceptibility of the penetration to PWSCC.
An alternate method for repair is to apply a smaller i
amount of weld material and thereby minimize the amount j
of additional penetration residual stress and deformation.
However, this method has the drawback of not restoring the penetration ID and would result in a much more difficult surface for post repair inspection l
by UT (manual method only currently developed) and Eddy i
Current (development of method is required).
l While both of these repair techniques are in accordance with the ASME Code, the embedded flaw repair technique avoids the above mentioned drawbacks.
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ATTACHMENT TO WESTINGHOUSE LETTER VRA-95-121 i
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I 2.
Embedded Flaw Repair The embedded flaw. repair technique involves an excavation at the i'
inside surface of the penetration.
This excavation would be
}
sufficient to remove the portion of the crack which is exposed to j
the reactor coolant at the inside surface of the penetration.
The depth of the excavation, 0.125 inch or smaller, would be set such that following application of a weld overlay, the remaining portion of the flaw will qualify as a subsurface flaw according j
to the rules of ASME Section XI paragraph IWA 3310 (b).
The depth'of the excavation is controlled by utilising "hard stops" which are incorporated into the tooling to limit travel of the EDM electrode.
Following excavation and prior to welding, a dye l
penetrant test will be performed to verify that the excavation has covered the full length of the flaw.
The weld is applied and j
examined with dye penetrant, addy current and ultrasonics to j
verify an acceptable weld. This approach eliminates exposure of the flaw to the reactor coolant environment, which stops further l
l flaw growth due to PWSCC.
See the attached figure entitled " Head 1
Penetration Embedded Flaw Repair" for a schematic of the proposed l
repair configuration.
3.
North Anna Proposed Embedded Flaw Repair The North Anna 1 and 2 reactor vessel head penetrations are typical of those in Westinghouse designed plants.
These penetrations are nominally 4.0 inch OD with a 2.75 inch ID.
l Installed into the majority of the North Anna Unit 1 head penetrations are thermal sleeves.
While these thermal sleeves are generally similar to the standard Westinghouse design, they i
have a continuous collar located approximately at the elevation j
of the high side of the penetration attachment weld (see attached j
figure entitled " Standard Thermal sleeve Guides").
This collar i
is machined such that there is a very small clearance between the collar and the head penetration inside diameter to align the i
thermal sleeve to the penetrations.
This close clearance makes removal of the thermal sleeve through the bottom of the penetration uncertain.
The potential for interference between j
the collar and the lower portion of the penetration due to the l
ovalisation of the penetration resulting from the original
[
welding of the penetration into the head is the concern.
To eliminate the necessity for thermal sleeve removal, an excavation i
and weld overlay repair of the penetration is performed through a
" window" which will be cut into the thermal sleeve.
A local weld i
overlay (as opposed to 360* coverage) over the cracked area will l
be used to minimize penetration deformation and residual stresses.
This repair process will be equally useful for j
unsleeved penetrations, but it has particular advantages for 3 of 8 i
l w-wwww e
v-y.
w y-.
)
ATTACHMENT TO WESTINGHOUSE LETTER VRA-95-121 sleeved geometries.
Although this repair technique is considered to be practical for axial flaws with a depth up to through wall, it is currently being considered only for flaws which.have a
~
depth of.up to 75% of the wall thickness.
If application of this technique is considered for axial flaws greater than 75% wall thickness or for circumferential flaws, a separate submittal to j
the NRc will be required.
The flaw extent will determine the extent of the repair, and the flaw depth will determine the thickness of the repair veld.
The penetration tube is sufficiently stiff, and constrained by the vessel head, so the integrity of the tube will be maintained by the weld overlay i
I regardless of the extent of the flaw.
When the repair process is I
complete LL& ID surface of the penetration has been restored and j-is readily re-inspected.
I i
j The " embedded flaw" repair methodology has been developed using 1
. technology which has been demonstrated in McAP 13998 (attached),
entitled "RV closure Head Penetration Tube ID Wald Overlay Repair".
Although this report contains a number of approaches to 1
penetration tube repair, only some of these'are used in the embedded flaw repair technique.
Section C, below, will highlight 4
4 l
the key portions of the report that are used as the technical l
basis for the proposed repair.
l C.
Summary of key relevant topics of WCAP 13998 The technical basis for the embedded flaw repair methodology is i
developed as shown in report WCAP 13998.
The following i
paragraphs provide a summary of the key relevant topics of the i
report.
i l
The report contains all the elements of a repair design package, and an outline of the package is contained in chapter 2.
The potential repairs were performed on a. full scale mockup of a head l-penetration along with several mock penetration tubes. The l
preparation of these mockups is described in chapter 4.
l l
The welding process uses Alloy 52 filler metal, to maximize the corrosion resistance of the weld.
The development of the welding
]'
process and its qualification are shown in chapter 5, which also i
contains pictorial examples of overlay welds performed over flaws i
machined into the penetration using electrical discharge l
machining (EDM). Test results showed no cracks in the weld or j
cracking of the surrounding area.
The welding specification is j
contained in Appendix'A.
l l
A range of weld overlay thicknesses were investigated.
It was found that the thickest overlays produced measurable deformation I
of the tubes, as shown in chapter 6.
smaller deformations occur l
with a smaller amount of weld metal thickness.
One of the i
L l
j 4 of 8 l
l 0
ATTACKMENT TO WESTINGHOUSE LETTER VRA-95-121 I
benefits of the embedded flaw overlay is that with a smaller i
amount of weld deposit the deformation is minimized.
To verify the adequacy of the weld repair process, a series of residual stress measurements were also performed on excavated and j
repaired tubes, and these results are discussed in Chapter 7.
As l
expected, the residual stresses are increased as more weld metal l
is deposited.
The residual stresses produced by local weld 1
overlays were comparable to the unrepaired configuration for-excavation and weld deposit up.to 0.25 inches in depth.
The measured residual stresses also compara favorably to those of a
]
three-dimensional finite element analysis for residual stress.
These comparisons are shown in Chapter 7, Figures 7.4-1 through 7.4-4.
4 l
To complete the weld repair design package, a generic safety evaluation according to 10CFR50.59 was performed, and was provided as a separate document from the WCAP.
D.
Comparison of the embedded flaw approach and WCAP 13988 i
To produce an embedded flaw configuration, a weld overlay j
thickness of 0.090 to 0.125 inch is needed.
The embedded flaw i
repair will apply the weld in an axial direction.
The welding process which was utilized in the WCAP applied the weld in a circumferential direction relative to the longitudinal axis of l
the penetration.
It is judged that welding axially in this range of thicknesses l
will maintain the penetration ID surface residual stresses j
comparable to the unrepaired tube.
This judgement is based on the results listed in the WCAP that showed this comparable condition for weld thickness up to 0.25 inch.
Further, the residual stress measurement results and their j
favorable comparison to previous analysts (refer to Chapter 7 Figures 7.4-1 through 7.4-4 of WCAP 13's98) is sufficient to provide confidence that the penetration stresses after weld i
repair have been fully described such that additional testing for i
corrosion behavior is not necessary.
i l
In the early days of the Westinghouse program to evaluate small mmounts sinc additives to the RCS coolant, measurements were taken of the electrode potentials of the various primary side materials.
No difference was found between them, including 600 i
and 690 materials.
This is in agreement with the investigations by others in the high temperature electrochemistry area.
At high j
temperatures the potentials of all of these alloys tend towards j
the potential of the hydrogen electrode; i.e.,
there are no differences to promote any galvanic coupling effects.
I 5 of 8
\\
t m -
i j
e ATTACHMENT-TO WESTINGHOUSE LETTER VRA-95-121
)
In addition, Westinghouse has many years experience in laboratory tests and field exposures with alloys 600 and 690 intimately connected either mechanically or by welding in steam generator applications.
Exposures of approximately 15 years on hybrid expansion joints have not. produced any evidence of galvanic coupling.
sleeving and plugging exposures have not revealed any evidence of galvanic interaction over years (5 at least) of operation.
E.
Flaw Acceptability Although the flaw characterisation rules of section XI paragraph IWA 1300 are being used to establish sufficient weld overlay l
thickness to classify the repaired flaw configuration as l
?
subsurface, determinations about flaw acceptability will be based on the NEI/NUMARC guidelines.
These guidelines were accepted in a Safety Evaluation Report issued to Wisconsin Electric Power Co.
l on March 9, 1994 (Docket No. 50-226), and in a crevious Safety Evaluation Report issued November 19, 1993 to W.
Raison of NEI/NUMARC.
F.
Summary and conclusions The embedded flaw approach has been developed as a variation on the repair techniques documented in WCAP 13998.
The technique is versatile, in that it can be applied to the penetration tubes l
with or without thermal sleeves, and does not require the removal of the thermal sleeve.
There are a number of advantages to the technique.
It results in a permanent repair that seals the flaw from the water l
environment, and thus stops PWSCC.
There is no other mechanism i
of growth for cracks in these tubes because fatigue fluctuations l
are very small.
The small thickness of the weld minimizes i
deformation of the tube, as well as residual stresses in the
' surrounding region.
l t
5 I
6 of 8 d
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ATTACHMENT TO WESTINGHOUSE LETTER VRA-95-121 Weld Overlay Rema..ining Flow
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ATTACHMENT TO WESTINGHOUSE LETTER VRA-95-121 Standard Thermal Sleeve Guides
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UNITED STAvet t
NUCLEAR REGULATORY COMMISSION 1
i 5
wassemoven,e.c.mssent k*..a March 9, 19 %
l Decket No. 50-Iss Mr. Robert E. Link. Vice president Nuclear power Department l
Wiecessin Electric pesar Compent'9 331 West Michigan Street Rees pat j
N11weekee, Wisconsis 83E01 4
j Omar Mr. Links
SUBJECT:
ACCEFTANCE CRITRIA pot ComDL MD DRIVE MCHANISM pMETRATION IN5ptCTIONS AT p0!NT BEACH ElCLEAR ptANT. W IT 1 i
l
-On Julh30,1933,d acceptance criteria for flaws detected during(IRMWic)l re the Neelear Management end Resources Council submit propose centro i
drive mechaniss (CROM) penetration inspections to the IRC staff for review and i
concurrence. These proposed acceptance criteria were based on extensive i
safety assessments senducted he 5thcock & Wileen Owners Group (Baues), the te(CESS), and the WestWause Geners sreep Combustian Cngineering Geners ria were separated inte eriteria for (WO8p. The proposed asseptance axis flaws and for circesferential flams by location Jakeve or below the
. The eresesel der asial flaus was te
)
J-treeveweldontheCRONgenetration)1 below the J-ersove weld and axial i-allow through-wall axial flaws of any
}
flaws-75 percent through-wall ef any 1 at er above the J-Greeve weld... _______
iety of Mechanical Engineers (AINE Therefore, the staff has founs then) i These criteria confers to the Ameriean 3ection II criteria for flaws in piping.
j acceptable.
l The IslMARC proposal for circumferential flees was through-wall and 75 percent areend the circusferense below the J-erseve veld, and 75 percent through-well and 50 percent around the circeeference at er above the J-treeve weld. Based en the informatica submitted by the seners groups that circumferential flaws should not initiate and arou, and the mere serious consequences of
'circumferential flaws, the staff has not accepted the troposed criteria for citeenforential fins. The staff has further stated that acceptanse criteria fe tent'ft"4 flews voeld not be pre-approved and that any c
evential flaws would be rettawed en a case-by-case basis.
3 4.4 On Janus 31 19M, submitted supplemental safety assessments devel by he5
' s. These supplemental assespeents provided a more tailed ev tien of stress states in the mezzles and diacessed the j
circumferential fleus sheerved at Rieghals and Supey 3.
The Alaghals ciremeteren1(13 f1Els were attributed to fabricatten fleus sad were not related to m eers water stress serresien stacking (pWSCC). The Sugey 3 l
circasferen3141 Eaw initiated at the external surface of the CROM penetration l
above the J-Greeve weld, and propagated at an angle 30' free horissstal. All j
three owners groeps submitted assessments that incledad finite element f
l WF i
2
.P l
mw w 4 o, _ -, -, e f
1 i
Robert E. Link b reh e, 1994 L
analyses that indicated that short, circumferential cracks are possible although these flaws would not be expected to propagate through-wall M to sempressive stresses below the flaus.
Based se its review of the emners.
-lesental evaluations the staff has concluded that short, partial'
'h circumferential fla,ws are en the stress analyses presented in i
Es'ble in the CNel penetrations. Seemners groups reports and the length of time has been in operation, a shallow cireumferential flaw 10 percent of the ration seuld exist. Therefers the staff has i
circumference of the semeladed that si erential flaws ubese length, inc1 ding postulated crack I
does not aussed 16 percset of the growth during the nest emerating syslei through well, diam ciresoference and are in a location consistantwilh%lessthan75pereenfinite element analysis fontside accostable. These flaws would have te he re'nspected in subsessent l
exals aations consistaat with the retaspection approach of !WB. tote of AsflE j
section II.
l You will not be required to obtain llRC approval to continue operation if short llowever you will he required to report circumferential flaws are identified, to the leic the location, length, and depth of lhese flaus and any other flaws identified during the inspection. If tw depths of the flaws are not j
determined, you sty assume that the depth is one half f.f the length of the flaw.
j Any-flaws found during the inspections that are not-resulting.from pW5cc.._
should he evaluated in a manner consistaat with the appredch for flas evaluation in AsplE section II usine the assumptions in the proposed acceptance criteria submitted by IRSIARC to IgIC en July 30, 1993. IJeamples of these flows i
would he short, shallow fabrisation defects or manufacturing defects in locations not predicted by the finita element stress analysu. Should cheese to disposition any flaws (which ascoed ASME 5ection.II criteria)you i
by l
analysis, the staff will require that your evaluations be reviewed and approved prior to unit startup, If yee have any questions regarding this isses, please contact as at j
(301) 504-1390.
stecerely,
^
i Allen 4. Hansen, project Manager project Directorate III-3 Divisten of Reacter projects !!I/IV/V Office of fluslaar Reacter Regulation ecs see next page 4
l 4
J l'
Unma STATES 8
NUCLEAR REGULATORY COM444SSION
)
(*****
=as,mmeren, e.c. muss.ms, November 19, 1993
,i W1111am Rasin, Vice president Director of the Technical Division nuclear Management and Resources Council l
1776 tye Street, N.W.
i Suite 30s Weshington, D.C. 20006-3706 i
Ocar Mr. Rasin:
The attached safety evaluation was prepared by the Materials and Chemical 1
l Engineering tranch, Division of Engineering Office of Nuclear Reacter Raoulstion, en the lasWic submittal of June 16, Igg 3, eddmssing the Alley 600 control Red Drive Mechanism (CROM)/ control tienent Drivr Machanism (CEOM i
This pressurized water reacter vessel head penetratten cracking issue.
suleittel addressed stress analyset, crack growth analyses, leakage u sossa nts, and wastage assessments for potential cracking et the inside Based on the everseas inspection findings and i
diameter of CRON/CEM nozzles.
the review of your analyses, the staff has coacInded that there is no This immediate safety concern for cracking of the CAM /CEst penetrations.
l finding is predicated on the performance of thn visual inspection activities Also, sper.ial needestructive examinations requested in Generic Letter 88-05.
are scheduled to commence in the Spring of 1994 to confim your safety J
l analyses for each PWk emners group.
j Your submittals for uck PWR type did not addmss the Ampy-3 flam that was eriented appromientely 30' eff the vertical axis nor a c' rcunferential, J-l pro 11einary information supplied to the groove flaw discovered at Ringhals.
staff by Swedish authorities 'adicates that the J~ groove flew may be We are continuing to work with the assectated with a fabrication defect. Frem the internation available to us Swedish authorities to confim this.
i taday, neither of these finus would peso a threat to the integrity of the CRDN j!
It is sur understanding that you ars also reviewing these flaws penetrations.
NRC and you will provide your assessesnt as to their significance and origin.
will issue a supplemental safety evaluation after reviewing your supplemental j
assessment.
l The staff agmes that there are no unfeviewed sal'ety questions associated with The staff agmes that the flaw predictions CNB3/CEDM penetration cracking.
based upon penetratten stress analyses are in qualitative agreement with -
j j
inspection findings. Neuever, the stress analyses de not address stresses from possible straightening of CNBl tration t. dies during fabrication.
1hese stresses, if lag, could ress t in circumferential flaw orientations.
4 The staff regnests that you aise address this Issee in your supplemental I
assessment. Based upon infomation received free overseas regulatory i
authorities, your analyses, and staff reviews, the staff believes thatRather, a flaw l
catastrophic failure of a penetration is extremely unlikely.
~
would leak before it mached the critical flam size and umuld be detected J'
during periodic servet11ance walkdowns for beric acid leakage pursuant to Generic Letter 80L-06. However, the staff recommends that you consider i
ll l
l 12 pp (7390TO M 3 XA u W30ir
=
t 1
i William Rasin !
enhanced leakage detection ty visually examining the reactor vessel head untti i
either inspections have been completed showing absence of cracking er en-1tne leakege detection is installed in the head area. The staff requests that you aise address.1hs issue of enhanced leakage detectlen in your supplemental 4
assessment.
The NRC staff has reviewed your July 30, 1993 suiarittal, which proposed flaw J
acceptance criteria to be used in disposittening any flaws found during The staff finds the proposed flaw acceptance criteria CRON/CEDN inspections.
acceptable for axial cracks because the criteria confers to the American i
The staff Society of Mechanical Engineers (AEME)Section XI criteria.
i determined that flaws that are primarily asial (less tisan 45* frez the axial direction) should be treated as axial cracks as indicated in Figure 1(b)l,(d),
i i
and (f) of your July 30, 1993 letter. Flaus more than og* free the axia j
direction should be treated as circumferential flaws. However, based upon information submitted to date and the more serious safety consequences of circumferential flaws, the staff does not agree with your proposed criteria for circumferential flaus. Circumferential flaws which a licensee proposes to t
i leave in service without repair, should be reviewed by the staff en a case-by-i case basis, i
Stacerely,
(
MMW i
Idilliam T. Russell, Associate Director i
for Inspection & Technical Assessment j
Office of Nuclear Reacter Regulation i
Enclosum:
l As Stated 3
i DistrI M an:
i Central A le J5tresalder WRussell EMCS RF WLiam RHermann J0 avis JWiggins IAlee j
POR i
i 4
f
- SEE PREVIOUS CONCURRENCE
- ENCB:0E
- ENCS:DE
- DES:DE
- ENC 8:0E
- 0E:D NRR:ADI
,~
J0 avis
)RLee RHemann J5trosnider JWienins WRussell i
I 09/13/93 10/t5/93 10.ft7/93 11/10/93 11/19/93 il/19/93 FFICIAL Ammu Evn E:\\DAVS5\\humatn.JAD (s:\\DAVII) 4 4
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EAFETY FeatuaTI M pattmTtat erarTon vrartSan AnaptoR Tm rearvin l
3.0 INTMnifrTION I
I of Alloy 600 was l
primary water stress cPresten cracking (MISCC)f to the MC Commission identified as an emerging issue by the MC staf following a 1900 leakage from an Alley 500 oressuriser heater sleeve penetration at Calvert Cliffs Unit 1, a Coahustion Engineering designed i
pressurized mater reacter (PWR). Several instances of PWsCC of Alley 800 pressurizer instraset nozzles had been reported to the MC between l
the time period of 1905 to the present en dunestic and foreign pressurized water reacters (pWR). The licensee at Arkansas Nuclear i
Operations, Unit 1, a tabcock & Wilcox (SW) designed PWR, reported a l
1eaking pressurizer instrument nozzle in 1990, after 16 years of operation. Westinghouse PM's de not use Alley 600 for penetrations or mezzles in the pressurizers.
f Accordt to the information provided to the staff by IRSIAAC at a pub 1 k meeting 1d on July 5,1993, a leak was discovered in an Alley 600 centrolreddrivemechanism(CRDNpadaptertubepenetrattenduringa hydrostatic test at the Supey 3 p lant la France in 1991 after 12 years A visual examinaties of the CMll adapter tube penetration of operation.
indicated the presence of asial flams in the inside diameter (!O) of the CNNI adapter tube penetration. The remaining 45 CMBI adapter tube penetrattens were nremined at Omgey 3 and 2 additional CidlM adapter tube penetrations contained axial cracks en the 10 of the CISI adapter tube trattens. An examinettee of 24 CMBI adapter tube penetrattens at y 4 revealed asial ID cracks in 4 CMR adapter tube penetrattens.
j C
adapter tube penetrations have been examined at 37 nuc, lear power and Belgian and 59 of the plants in France, Smeden, Suitzerland, Japan j
1,850penetrationshaverevealedshort,axialcrackindicaticas.
i The primary safety concern associated with stress corrosien cracting in Alley 600 la CROM penetrations is the poteettal for circumferential i
cracks. Extensive circumferential cracking could lead to the ejection ef a CRDM resulting in en unteolable rupture in the pristry coolant l
systas. As indicated above, the inssections to date have identified i
short asini cracks. Neuever, two other inspection findings are of l
particular interest. First, the CM M penetratten that leaked during hydreetatic testing _at Sugey-3 was removed and examined metallurgically i
during Dessuber Issa. A secondary crack that was 0.!!0 inches long and
]
0.000 inches deep at about 30 degrees to the axial direction was observed en this CMWl. Second, in early in 1993, a J-groove weld at the i
Ringhals plant in Smeden was discovered to contain a circumferential crack. preliminary indications are that this flaw is a fabrication i
defect. Additional work is in progress by the staff at the Swedish i
Nuclear power Inspectorate to confire this.
The Westinghouse CMBI adapter tube penetrations are steiler in design to the European pWR's and use Alley 600 for the penetrations. The NRC i
staff est with the WOS en January 1, 1992 to discuss the experience at
!]
I i
j 4
r-
l~
i i
j i
I t
i' the Supey 3 plant and the relattenship of the French design of the CRDM adapter tehe penetrations to the design of domestic Westinghouse plants.
The WDS informed the NRC staff thtt a program had been initiated in i
Decembes.1931 to: (1) determine the rest cause of the CRIBl penetration analyse the stress distributions in the CRIBl penetrations cracking; (2) domestic plant; (3) compare the design and operational i
of a typical characteristics of domestic and French plants to detensine the itkalthood for cracking; and (4) identify the need for additional l
efforts. The NRC staff also met with the Cambustien Eneineering Guners j
Group (CEOS) and the Babcock & Wilces owners Group (SAN 0s) Management a to discuss the PWSCC of CalBI adapter tube penetrattens. The Nuclear Resources Comcil (Inm4AC) coordinated the ptm Ouners' Group efforts on this subject.
IRSIARC submitted safety sssessments to the NRC fron i
On June 16, 1993 These safety WOS, CE06, and B&WDE for review by the NRC staff.
assessments present stress analyses, crack growth analyses, leakage analyses, and wastage assessmetit. For flaws initiating en the 10 of CRDM adapter tube penetrations. NRC requested additional infensation on the safety assessments by letter dated September 2, 1993. NUMARC submitted The safety assessments i
the response to NRC en september it, 1993.
J submitted to the NRC did not address the secondary flam observed at the l
Eugey-3 plant that was oriented approximately 30' from the longitudinal amis of the penetratten nor the apparent fabrication flaw at the Ringhals plant. Neither of these flams a threat to the integrity l
of the CRDM penetrations. Houever, has committed te submit a After this safety safety assessment relevant to this type of cracking.
I assessment has been reviewed by IRC, a supplement to this SER will be J
l issued.
1 2.0 STAFF EVALu&Tind une uras.n585. ALLOY 8aa e_ratTOR Vrteri MEAD ma*TGR Tuar rmArv1na 2.1 5AFETy Evaluarren i
l The WDG submitted the. ' Alley 600 Reacter Vessel Head Adapter Tube Safety Evaluation.' through Inst 4AC en June 16, 1993. The safety i
evaluation addresses the following elementst l
A summary of the stress analysis focusing on the type (erientation) l 1.
of cracking that may he espected in the Alloy 400 material, and the l
stresses necessary for flaw propagation; i
A summary of the flam propagation analysis along with the background i
1.
of the flaw prediction mothed; An assessment of the WOG plants with respect to penetration flaw 3.
indication data from plant inspections at Ringhals. Bernau, and i
verless Electric 1te de France plants, in which the key parameters l
for cracking are compared to WDG plants; i
l l
l i
j j
i 3
4 4.
A leakage assessment summarizing leak rate vs. flew stre, and may apply;g leaks for WOS plants for which leakage censidoratiens postulatin l
- and, 4
5.
A Nasal head wastage assessment including the process that leads to j
wastage and an estiente of the alleunble mastage.
2.1.1 esanstio;; mit - im%maffas or a WI M G EarrTY m sTf m i
The WO4 prepared safety evaluation addresses the potential for cracking i
and the ramificattens of such crackt of the reacter vessel head adapter tubes at Westinghouse dest IIS$$ plants. The WDG cousared j
the results of this safety evaluat en to the criteria in the Title 10, j
Code of Federal Regulattens, Sectien 50.55 (10 CFR 50.H ).
The WD6 concluded that an unrevleued safety guestion did not exist. Its i
evaluation considered the felleming:
i i
1.
Continued plant operatten will not increase the probabiltty of an l
accident prevleusly evaluated in the FIAR.
j t.
The consequences of an accident previously evaluated in the FSAR are j
not increased due to continued plant operatten.
3.
Continued plant aparatten util not create the possibility of an accident uhtch is different than any already evaluated in the FSAR.
1 i
4.
Continued plant operation will not increase the probability of a j
malfunction of egulpment tapertant to safety.
Continued plant operstken will not increase h conseguences of a 5.
l malfunction of equipment important to safety prevleusly evaluated in the F5AR.
I l
E.
Continued plant operetten util not create the possiblitty of a i
malfunction of egutement tapertant to safety different than any i
already evaluated in the FSAR.
i The evaluatten for the effects of continued plant operatten with 7.
l i
potentially cracted reacter vessel head adapters has taken inte account the applicable technical specifications.
i 2.1.1 STAFF'E fvaluaTtau 3F "HE arnaM Atony anels aan DETEMINATfon OF
-rvrrwn 1&FETY art"Inut The staff agrees that ne unreviewed safety geestion exists, provided only amtal flaws are found. These utal flaws would be espected to be short, and they would aest probably leak noticeably prior to the flaw i
size reaching unstable dimensions. The existence of any unexpected leaks would not adversely affect plant aperatten, or accident /transtent j
lie significant equipment degradation would be expected.
response.
l Dotatis of the staff's evaluatten ht led to the above conclusions is I
i discussed in the following sections, l
l l
o i
~
i i
l 4
2.1.3 pflETRAflau ITRE$1 ANALYSIS The tElt condested an elastic-plastic, finite element analysis of a 4-I loop IEE plant vessel head penetrations. The Idos concluded that the 4-leap tf08 plant is bounding since prior analyses shound that the operating and resideal stresses are higher en a 4-leep plant than en 2 er 3-leep plants en the entermost penetrations. Three penetratten t
locations were endeled, the center locatten, the outernest 1ecetten, and
{
the location nest to the estemest locatten. The stress htstory was 1
staplated by using a lead sequence of the thems1 lead from the first 1
welding pass, the thermal lead from the second weld pass, the fabrication shop cold hydretest, the field cold hydrotest, and the
)
steady state operational leading.
3 l
The highest stresses are found in the rene around the sold and are the i
i i
highest in the penetratten. farthest free the center of the vessel l
(peripheral penetrations). N highest stresses en that penetration are l
en the side of the penetration nearest to the center of the vessel i
(contarside) and on the side of the penetration farthest from the conter Also, the stresses are the highest below the j
ef the vessel (ht11 side).
weld and decrease significantly above the weld. h ratie of peak heap stress to antal stress at the same location at the estemest penetrattens was about 1.4 compared to a vales of shout I.4 estimated The hased en the degree of evaling esasured on actual penetrations.
rette of heep stress to axial stress was about the same for center penetrattens as for peripheral penetrattens (1.5 for center penetrations compared to 1.4 for peripheral penetrattens); howeve, the magnitude of l
l the stresses at the peripheral penetrattees was higher. The analysts i
t l
indicates that axial fless sould be more likely than circumferentisi flaws, flaws are more likely below the weld than above the weld, and i
f that antal flaws would appear at locattens in the penetratten:; where j
l they have been found in service.
1' 2.1.4 STMF EvalHAT1ou # THE r= mATIoM mi-sce "" YE15 l
N staff is in t with the results of the 1806 stress analysts racking will be predentaately axial, hse that predicts tha i
results are la qualttative: agreement with field inspectlen findings, i
Hesever, the tiet did not address the effects of possible straf tosing i
Such stral tening of the CNBI penetration tubes during fabrication.
operattens could significantly alter the residual stress fiel s within Results of inspectfens to date have not the penetratten tubes.
identified any prehless directly related to this process; however, the j
j staff requests that IRRERC address this issue for all three emners groups' plants.
i 2.1.5 man w "TH ^^^^' YSTta Fl_aN TfM Funner N 1d06 crack growth analysis was based on the assumptions that the flaw l
would be causes by primary water stress corrosies cracking, and that the The nazieum principal crack growth is controlled by the heep stress.
stress will be oriented at a slight angle to the hoop stress and flaws i
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4 would be exeected to be perper.dtcular to the maalaun princip4} stress.
1 i
i However, all of the flams found in service witn two exceptfens have been I
axially located. Nance, the WDE used the heep stress as an apprenjaation of the maalaus principal stress. The outer-most l
penetratien for a 4-leep Westiagheese plant was selected for analys1s since this locatten experiences the hf st stresses. The highest l
stress was located along the inner ace just below the center side of i
the weld. The calculated heep stress through the wall of the penetratten was used for flaw growth calculations. The flaw growth data l
j were obtained from steam generator field esperience and laboratory data.
l i
Based en the stress fields that exist in the CW penetrattens, any flaw growth that occurs is expected to be predominately axial in nature.
i Furthermore, the growth of any fleus incitned from the vertical wegld be These i
limited in length due to the nature of the existing stresses.
conclusions are consistent with the inspection results described above.
there is ne significant potential for failure of a Accordingly,by ejection of W CRDN sleeve. With regard to axial penetratten cracking, WOG has concluded that the critical flaw length for an estal i
l flaw for A11ov 600 is sufficiently long that leakage would occur and be detected during serveillance walkdowns as required by SL 88-05.
Therefore, the conseguonces of cracklog in the penetration sleeve are limited to the affects of leakage as duscussed helow.
i l
l The flaw growth analysts showed that under the most severe condittens of I
metallurgical microstructure, peak heep stress, and operating l
temperature, it would take about five years for a flaw to grew through Under the same canditions, it would take an addittenal 10 years j
wall.
for a thin;;t =11 flew to grew 14 inches above the weld en the lower
{
hillside of the outermost head penetrations (Figure 3.2-2) and about the same time to grew two inches above the J
..x; us1d on the center side l
of the outeresst penetrattens (Figure 3.t-3). The flew growth analysis indicates that through well flaws would essentia11y arrest before i
These flaws would be j
growing a manteus of tue inches above the weld.
l constrained within the head and could not significantly open thus l
limiting the amount of feakage that could occur.
l
{
2.3.61TAFF Evaluattau er fut ram==#fu amatysts The WDE stated that the crack growth analysis is in general agreement The crack growth rate data used in this with the inspection findings.
analysis was limited, but the results predicted using these flaw growth Crack growth rates are data bound the reselts of the inspections.
difficult to determine precisely; however, the assumed growth rates cespare well with inspection data available te date and the large margins that exist in the analyses will account for any possibly higher a
There are large margins of safety in the analyses and the growth rates.
CMM penetrattens are constructed of inherently tough material with a critical flaw size of approximately 13 inches in the free span above the Therefore, the staff concludes that catastrophic reactor vessel shell.
failure of a penetratten is extressly unlikely because a flew would be l
t i
i l
i i
j 6
l detected during beric acid leakage surveillance walkdowns before it reached the critical flew size.
4 f
I.
2.1.7 ASSESM OF MDB PLAlffS N WOG cespered the Ringhals and Bosnas plants to the domestic Westinghouse plants and developed a model for the relative
)
susceptibility to pWSCC. m WDE considered residual and operating j
J sperating temperature, and tlas e environment, material conditt l
stresses in the penetrations thof-operation at temperaturw. and i
j Based on this evaluation, the Wo$ has evaluated domestic WDE pressure.
pWR's with regard to their degree of susceptibility. Based en what WOG considers to to conservative assumptions, the Ringhals plants envelop
)
45 demostic plants. None of these plaats are especLM to have av flaws other than same short, shallow, axial flaus. Nies additional WDE plants are not enveloped by the Ringhals plants. Based en the stresses, speratlag temperatures, hours of operation, and the flaw growth curves provided in the 458 safety assessmeet, the WOS does not expect any CROM 1
i penetration axial f1 sus to reach a length in escess of 1 inch before about the siddle of 1995.
.i I
g 1.3 tTArr tvaluafum DF TMF m E-d N susceptibility model developed iy the WOS considers the appropriate parameters affecting 163CC and should provide a reasonable ranking of In additten, this evaluaties indicates that it plant susceptittlit'es.
l is unlikely that U.S. plants should exhibit any cracking significantly i
worse than that found in European plants.
1 g.1.9 LEAK RATE CALCDLATICIl5 N 1eak rates were calculated based on the assuption that the leak i
rate will be controlled by the flew rate through the flaw in the head penetration er by the flew through the penetration annulus, whichever is l
Wes estimates the manismo leak rate would be 0.7 gpm for a 2 j
ses11er.
Leakage above inch long flaw and an annular clearance of 0.003 inches.
+
I Growth 1.0 sps 's detectable in desestic WOS plants accordlag to WOG.
of an axial flaw estside of the art contained within the reacter head l
will result te leakepe greater itan 1.0 gps prior to reaching the l
h WOG stated that an axial flav would rensin crittsal flaw stas, stable for growth op to 13 inches above the reacter vessel head.
i ETAf71 rvaluaTim M Tut undt trar naTE cae rulaftes 2.1.10 N staff agrees with the WOG assumptions about leakage and concludes.
t l
that based on existing leakage monitoring requirements, there is ressenable assurance that leakage in excess of the 1.0 gpa technical seecification limit would be detected prior to say unstable extension of th O w.
l i
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- _ _. _ _ _ _ _ _ ~ _ _ _ _ _ - - _
1
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7 his vrem uran useraar neweemrurg 2.1.!!
This section assesses the potentir! westage of the reacter vessel head This due te4eakage of primary coolant through the CNN penetrations.
assessment is based as wastage data from prvviews Westinghouse esperiments and free the results of a penetratten mockup test conducted j
by the Combustion Engineering Oumers troup (Clos).
This analysis assmed that coolant escaping from the penetratten would l
flash to steam leaving heric acid crystals behind. MS assumed that crystals would accuelate en the vessel head but would cause sinimalThe head i
cerrosien 211e the reacter was operating.
j he about 500'F during operattee and significant unstage of the reacter head by the beric ac9d crystals would not be espeseted. Dry heric acid i
crystals de not cause corresten. gastage would only occur during eutages when the head temperature is below tit *F.
a i
The CE08 provided.all of the PW emners groups with the results of The WDG examination of the pressuriser penetratten esckup test results.
CEOS mockup test results showed that the mastaus penetratten rate at the deepest pit was 3.15 inches / year while the average pewtration rate was The maximum total metal less rate er wast voies
}
0.0835 inchts/ year.
was 1.07 in /peor, and the greatest damage occurred where the 1 4
left the annulus. The WOS considered the anxieum wastage would he i
in of vessel head material.
i ever 1.0 tom can he detected se caly leak rates between 0.0 and 1.0 spe f
The 2 8 analyzed the situatten using finite element were cons <dered.
analyses for a 2 leep, 3 leep, and 4 leap reacter vessel head where a j
1.0 gea leak went undetected for 6 years and concluded that the ASME code minimum wall thickness ragstrument would be satisfied and that the i
stresses remain within the ASME code allemable stresses.
\\
TM ETAFF's fuuaTMN OF Tw erartna wretrt uran Erraar l
2.1.32 AHHMEEIA I
The assumption used in the W S corresten assessment are based en emperimental data and shoold provide a reasonable estiaste of potential f
wastage of the reacter vessel head. Based en these evaluations, there would he significast time between initiating a leak and esperiencing wastage that would reduce the structure 1 integrity margins of the reacter vessel head te belaw acceptable levels. Considering 4
i of time involved bytheaccumulatIenofmoderateamountsofhericacidcrystalswouldbe detected during a surveillance walkdown in accordance with GL 88-05.
i 3.0 CEoK 1AFEfY EV$LETION f
The CEOS safety evaluatten is essentially the same as the WoG safety The CESS pleets rue at a slightly higher temperature than l
evaluatten.
have greater the European plants that have esperienced cracking,than many of the hillside angles, and have been in operetten longerThe CEDG indicate European plants.
i I
i l
4
]
->v
j 2
i 4
5 1
i 8
b Neuever, W increase the probability of cracking for the CEOS plants.
i CEDE plants have significantly less weld metal in the J-groove welds and the CEOS stated that this would significantly reduce the residual weldtse4aduced stresses and would reduce the prehability of pW!CC.
l CE0S concluded that any PWSCC that formed weeld be short, axial flaws.
i The CEDS states that they can detect a 0.12 sps leak in the primary l
CEOS aise states that the beric acid accumulation as a coelaat system.
result of a 0.15 gas leak would not result in well thiantag below the code allowables in less than 8.8 years campared te a years for WOG i
plants and that servet11ance walkdsuns would detect borte acid crystals 4
long before the 0.8 years.
HAFF tvaluaTHIM nF M rFaa MFrTY swMMTfou 3]
l The staff has concluded that the poteattal for PWSCC of CRM/CEDM for f
CEOG plants does not create an tamediate safety issue as long as the surveillance walkdsuns regelred by GL 88-05 continue and corrective l
i actice is instituted when leaks are discovered. 'The CEDE analyses
)
indicating that the stresses would favor development of axial rather i
than circumferential cracks and that significant time would be required i
to reduce the wall thickness of the vessel head to below the A5ME co allevables demsestrates that an insediate safety concern does not exist.
1 i
4 4.0 EnlElE 14FETT EMalMTICE The RWes safety evaluation was essentially the same as the WOB and CEOG The 8W00 analysis indicates that SWOS platts have l
safety evaluattens.
essentially the same susceptibility to PW3CC as the Euro The SWOG predicts short, axial flaws en the seripheral locations based en the results of finite element analyses. The B&E06 estimates that it j
would take it years from the time a flew initiates en the inside Once a leak diameter of a CNN penetraties until a laak appears.
l starts SWOG concluded that it would take 6 years before enough correslon weeld eccer to reduce the wall thickness of the reacte i
l head to below ASE code sittaess, and that this amount of leakage would i
be detected dertog serveillance walkdomms.
j nAFF rwamattom er m % nFrty runattau l
4.1 The staff has concluded that the poteettal for pW5CC of C80N for SWOG plaats does not create an immediate safety issue as long as the i
surveillance walkesuns required continue and as 1 ens a i
corrected.
development of axial rather than circunferential cracks i
vessel head to below the hSNE code allenables, demonstrates that an l
l tenedtate safety concern does not exist.
l 4
J
1
]
i i
4 1
3 5.0
=-_-
a= Ft M acerpfawr estTFe1A i
On July 30 1933 IR8WAC submitted the proposed flew acceptance criteria for f1me Ident11'ied dering inservice inspection of reacter vessel upper These criteria were developed
{
head penetrattens to the IK for review.
IRMARC by utility technical staffs and the desestic pWR venders.
i proposes that axial flaus are permitted through-wall below the J-groove i
There is no limit en l
weld and 75 percent througlHen11 aheve the meld.
the length of the flaus. 25mAC proposes that circumferential flaws j
through-us11 and 75 percent around the penetratten be allound below the I
J-groove unid and that circumferential fleus above the weld could he 15 1
proximity j
percent incouglHea11 and 50 percent around the penetratten.
rules found in ASE Section XI, Figure Itm 3400-1 are proposed for determining the effective length of multiple flaus in one locatten.
IRMARC proposes that the flaus be characterised by length and preferably 4
InflAAC proposes that if only the length is characterized, that i
depth.
the depth he assumed to he one half of the length based en inspection findings to iste, i
ggf um4 Tram or int asi; rtas meerpTaaer cerTrata l
5.1 The staff finds the proposed flew acceptance criteria acceptable for axial flams because the criteria confers to the American sectety of i
Section XI criteria. The assesetten that IIschanical Eegineers (ASE)noe length for flaus dose depth cannot be flau depth is one half the j
determined will Itait the flaw length to 1.5 times the thickness of the penetratten sleeve. lieuever, it is espected that reasonable attempts Fless found through inservice r
1 will be ande te determine flau depths.that are primarily axial (less than 45' free l
inspection (151 direction) ull))he treated.as asial flaws as indicated in F1 Ijb),
i l
Flaus more 45 free i
(d), and (f) of IR8thC'S Aly 30, Igg 3 letter.
Based the axial directies are considered to be circueferential flaws.
l upon inforestian submitted to date and he more serious safety consequences of circumfermettal flaws, thi staff has concluded that l
criteria for circumferential fleus should not be pre-approved.
I Detection of such flaus usuid be centrary to inspection results to date l
The and to the conclusten of the Guners Groups evaluattens.
curcumstances assectated with sech a flaw would have to be well Therefore, any circumferential flaws fasad threegh 151.
understead.
l which a Itcenses proposes to leave in service without repair, will he revisued en a case-ty-case basis by the staff.
I i
I 6.0 LEAEARE$EMTMIES IRMARC, through the euners groups' reports, determined that any leakage in excess of I gps would he detected prior to any unstable extensten of i
Also, leakage at less th.em I gus would be detectable ever axial flaws.
ties based en beric acid tutidup as noted during periodic i
l 1evel leakage will not cause a sipalficant safety issue to result, the walkdeuns.
staff deteristand that IRSIAPC shou d censider methods for detectingsmaller leaks to provide l
(
4 l
l ND> 22 '93 assarr l
1 4
i 10 i
h reported leak rate at Segey 3 was uncertainty la its analyses, about 8.003 gps and was detected using acoustic monitoring techsteves i
The staff does not think during the performance of a hydrostatic test.that 11.is necessary to de j
permitting leakepe just below 1.0 gpa as currently proposed may be Laakage of this angnitude usuld produce significant undesirable.
deposits (themsends of peands/ year) of her$c acid on the reacter vessel i
Further, mest factiftles' technical specifications state that no head.
The staff notes that small ressere boundary leakage fs permitted, feaks resulting from flaws which progressed through-wall just prior to a refueling outage would be diffievit te detect while the therest i
insulation is installed. Although running fer as additional cycle with i
that undetected leak sesid met result le a significant safety issue, the NtalARC should consider proposing a method for detecting leaks that are
?
significantly less than 1.0 gps, such as the installation of on-line l
monitoring equipment.
8 l
70 CGICLU11Gts Based on review of the last4AC submittal and the everseas inspection
]
results the staff concludes that the CM/CEIN cracking at the reacter 4
vessel heads is not a signtficant safety issue at this time as long as The the surveillance walkdeums in accordance with GL 85-05 continue.
staff agrees with the NLMARC's detershatten that there are ne unreviewed safety guestions assectated with stress corresten cracking of l
Housver. new information and events may require a j
CRWI penetrattens.
Furthermore, there is a need reassessment of the safety significance.
f te verify the conclusions of the inflAAC's safety evaluations.
j i
Therefore, nondestructive examinattees should he performed to ensureThese exam there is no unospected cracking in desestic PIRs.
j not have to be conducted tunediately sface only short, shallow, axial The industry flaws are likely to be present in the Cast penetrations.
i has comettted to conduct fespectfees at three units in 1994. They are:
!l a) Point teach linit 1 in the Spring of 1994, b) D.C. Cesk Unit 2 in the thirti guarter of 1994 i
i c) Oconee Unit 1 in Septeober 1994.
l As the surveillanes walkdowns proposed by IRMAR t
l potentially operate with ses11 undetected leakage at CRON i
to consider the toplementatten of an enhanced leakage detection trattens.
l method for detecting small leaks during plant operetten.
)
The staff found instARC's flew acceptance criteria acceptable for axial flaws but IK, review and approval of the dispositten of any i
i circumferential flaws will he required.
i Technical contacts:
Robert A. Hermann (301 50 M 7sa W111tas H. Kee (301 504-3706 James A. Davis (301 504-2713 l
i l
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~ WESTINGHOUSE NON-PROPRIETARY CLASS 3
' WCAP-14519 l
i i
i RV Closure Head Penetration Tube
.ID Weld Overlay Repair
' A Westinghouse Owners Group Program Report -
i
' WESTINGHOUSE ELECTRIC CORPORATION Nuclear Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230 01995 Westinghouse Electric Corporation All Rights Reserved m \\2506w.wpf:1b-111095 i
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f s
TABLE OF CONTENTS Y
Section PageL Table of Contents
-i List ofIllustrations
' 111
' List'of Tables v
l Executive Summary -
vi-1.0 '
Introduction 1-1 2.0..
- =
Program Description 2-1 c 2.1 Objectives 2-1
- 2.2
- Weld Repair Program Outline 2-2
]
' 3.0 '
Approach for Development of Penetration Tube Weld Repair and 3-1 i
Overlay Designs i
3.1 Local Weld Repair -
3-1 3.2 360* Weld Overlay 3-2 i
3.3
- General Program Goal 3-2 4.0 Penetration hbe Sample & Reactor Vessel Head / Penetration 4-1 Tube Mock-Up Fabrication 4.1 Preparation of Penetration Tube Samples 4-1 4.2 Fabrication of Reactor Vessel Closure Head /
4-1 Penetration Tube Mock-Up 5.0 Weld Process SpeclScation 5-1 i-5.1 Selection of Welding Equipment 5-1 5.2 Qualification of the Welding Parameters 5-1 5.3 Welding of Penetration hbe Samples 5-3 1
5.4 Welding Reactor Vessel Closure Head / Penetration 5-4 hbe Mock-Up 6.0 Evaluation of Welded Penetration Tube Samples 6-1 6.1 Discussion of Diametral Measurements 6-1
)
- 7.0 Residual Stress Measurements on Reactor Vessel Head /
7-1 Penetration Tube Mock-Up.
7.1 Approach to Residual Stress Measurement 7-1 7.2 Hole Drilling Method 7-1
~ 7.2.1 Installation of Strain Oage Rosettes 7-2 7.2.2 Drilling Holes 7-2 7.3 Test Results 7-2 7.4
- Comparison of Test Results to Analysis 7-3 m:u506w.wpf:1b-111095 i
i x
n,
f TABLE OF CONTENTS (Continued)
- Section Page 8.0 Discussion of Post Weld Surface Treatment 8-1 8.1 General Discussion of Shot Peening 8-1 8.2 -
Shot Peen Parameters 8-2 8.3 Conclusions Regarding Post Weld Surface Treatment 8-3 L
9.0 Discussion of Weld Overlay Repairs 9-1 9.1 Penetration Tbbe Repair Parameters 9-1 9.1.1 Excavation Depths and Weld Thickness 9-1 9.1.2 Repair Geometry 9-2 9.1.3 Weld Surface Finish 9-3 9.1.4 ASME Code Approach to Weld Repair -
9-4 9.15 Post-Weld Inspection Requirements 9-4 9.2 Conclusions 9-5 i
10.0 References 10-1 Appendix A Welding Process Specification A-1 i
j Appendix B Weld Repair Drawing B-1 i
I Appendix C Data Package for the Penetration Mock-Up &
C-1 Penetration Mock-Up Sketches Appendix D Penetration Tube Dimensional Data D-1 l
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4 mA2506w.wpf:1b.t11095 li
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'.)<
LIST OF ILLUSTRATIONS.
. Figure
- Title Page Figure 3.0-1 Reactor Vessel Closure Head to Penetration Tube Geometry 3-4 Figure 4.1-1
' Excavation Geometry for the 10 Inch Penetration Tube Sample 43 j
Figure 4.1-2
. Full Size Mock-up Sketch Depicting "J" Preparation Excavation 4-4 Geometry. -
1 Figure 5.1 LWeld Head Used for Weld Repair Program 5-5 Figure 5.1-2 ' Weld Power Supply / Controller Used for Weld Repair Program 5-6 Figure 5.2-1 Joint Geometry for Qualification Samples -
5-7 Figure 5.3-1 Penetration Tube Sample No. 8 Cross-Section Showing Weld Repair 5-9 Over a Circumferential EDM Notch Figure 5.3-2 Penetration hbe Sample No. 8 Cross-Section Showing Weld Repair 5-10 Over the Longitudinal EDM Notches Figure 5.3 Penetration hbe Sample No. 7 Cross-Section Showing Weld Repair Over the 5 11 Circumferential (Bottom) and longitudinal (Top) EDM Notches
. Figure 5.4-1 Photograph Depicting Weld Tooling Set-Up In Full Size Penetration 5-12 hbe Mock-Up -
Figure 6.1-1 Deformation in Penetration hbe Sample #1 [
] *',
6-3
- Angle 90' i
Figure 6.1-2 Deformation in Penetration hbe Sample #2 [
]* C,
6-4 Angle 360'.
Figure 6.1 Deformation in Penetration hbe Sample #3 [
]*'C,
6-5
]
Angle 90' l
Figure 6.1-4 Deformation in Penetration hbe Sample #4 [
]*'* *,
6-6 Angle 360' -
' Figure 6.1-5 Deformation in Penetration Tube Sample #5 [
]***,
6-7 Angle 45* -
Figure 6.1-6 Deformation in Penetration Abe Sample #6 [..
]***,
6-8
' Angle 90' Figure 6.1-7 Deformation in Penetration Tube Sampic #7 [
],
6-9 Angle 90' l
m:u5064wpt:1b 111095 '
111 i-
LIST OFILLUSTRATIONS (Continued)
- Figure Title Page.
4 s
Figure 6.18 Deformation in Penetration Tube Sample #8 [
]*,
6-10 Angle 360' Figure 7.1-1 Overall Dimensions of Head Penetration Model and Air Abrasive Drill 7-6 Positioning Fixture ~
Figure 7.2-1 Location Map of Residual Stress Measurements for Step 1 - The 7-8 As-Received Condition i
Location Map of Residual Stress Measurements for Step 2 - After 7-9 Figure 7.2-2 Machining Weld Repair Areas
- Figure 7.2-3 Location Map of Residual Stress Measurements for Step 3 - After 7-10 Welding Repair Areas i
Figure 7.2-4 Test Setup for Residual Stress Measurements 7-11 Figure 7.2-5 Adjusting Hole Drilling Fixture 7-12 i
Figure 7.3-1 Hole Drilling Rosette Strain Gage Data 7-14
)
i.
Figure 7.3-2 Relationship of Principal Stress Directions to Rosette Gages 7-15 Figure 7.3-3 Residual Stress Versus Distance From End of Tube for Step 1 at 7-16 180' Location
)
Figure 7.3-4 Residual Stress Versus Distance From End of 7bbe for Step 1 at 7-17 0* Location Figure 7.4-1 Residual Hoop Stress As-Measured Compared to Analytical Estimates 7-18 of Hoop Stress for Center Side of Penetration Figure 7.4-2 Residual Hoop Stress As-Measured Compared to Analytical Estimates 7-19 4
of Hoop Stress for Hill Side of Penetration Figure 7.4-3. Residual Axial Stress As-Measured Compared to Analytical Estinctes 7-20 of Hoop Stress for Center Side of Penetration Figure 7.4-4 Residual Axial Stress As-Measured Compared to Analytical Estimates 7-21 j
. of Hoop Stress for Hill Side of Penetration 2
4
~
1 1
1 m:\\2506w.wpf:Ib-111095 iv
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' LIST OF TABLES Table Title Page Table 5.31 TEST MATRIX FOR [
]***# PENETRATION TUBE SAMPLES 5-8 Table 7.2-1 ROSETIE LOCATIONS ON THE ID OF THE TUBE -
7-7 Table 7.3-1 RESIDUAL STRESS MEASUREMENTS IN REACTOR VESSEL 7-13 HEAD /PENE1 RATION TUBE MOCK-UP t
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REVISION RECORD 11-10-95 Proprietary information marked in preparation of Class 3 Report No.14519 D. Boyle mA2506w.wpf:Ib-ll1095 vi
t i
i EXECUTIVE
SUMMARY
A technical approach to address the issue of primary water stress corrosion cracking (PWSCC) on the ID surface of reactor vessel closure head penetration tubes has been outlined by the Westinghouse Owners Group (WOG). In addition, the WOG has supported NUMARC at the industry level in taking a proactive role in ice >lution of this issue. In structuring an approach the WOG has supported root cause evaluations, investigated how WOG plants are impacted, submitted a generic safety evaluation, developed plant inspection criteria, and solicited volunteers to perform plant inspections. Also, via the weld overlay program authorization (MUHP-5017), the subject of this report, the WOG is providing generic guidelines applicable for penetration tube repair and potentially a methodology to mitigate PWSCC in the penetration tube ID.
'Ihis part of the program provides a weld design package which can be applied to repair reactor vessel j
closure head penetration tube ID initiated PWSCC. 'Ihe weld design package provides the criteria for the repair of the penetration tube ID either through the application a local weld repair or via the application of a 360* weld overlay. 'Ihe local weld repair process is targeted at restoring the minimum required design thickness of the penetration tube wall. 'Ihe 360* weld overlay is intended to provide a remedial measure to mitigate PWSCC in the Alloy 600 penetration tube ID by eliminating exposure of the highly stressed regions of the tube wall to the primary water environment.
i If an individual utility decides to perform volumetric inspections of vessel head penetrations
{
indications could possibly be encountered which would require disposition in order to permit plant l
start-up. Indications detected via penetration tube volumetric inspections need not necessarily i
immediately be repaired. Each penetration tube indication needs to be evaluated against the established industry acceptance criteria. Dependent on indication position, depth, and orientation it is quite possible no immediate corrective action is required. In fact no corrective measure may be required for the remaining design life of the plant. If however corrective action is required the first course of action would require removal of the defect by excavation. Excavation by itselfis an acceptable corrective measure as long as the minimum required design thickness of the penetration l
1 tube wall is not violated, approximately 0.3 inch. If the minimum required design thickness is violated the integrity of the penetration tube wall needs to be re-established, i.e. via the local weld repair.
1 i
i mA2506w.wpf:Ib-111095 vij l
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In support of the weld repair processes the following has been investigated; 1) Excavation geometries and various depths as related to flaw geometry, 2) Limitation of the weld repair with respect to crack length, 3) The definition of welding process parameters,4) 'Ihe definition of allowable weld filler metals,5) Weld surface finish, 6) Requirements relative to the surface profile of the penetration inside diameter,7) Industry. suggested parameters for shot peening, and 8) Weld inspection requirements. Shot peening was examined as a post weld surface treatment to mitigate the residual
. stresses induced by welding. In addition to support the repair process, Westinghouse performed a generic 50.59 Safety Evaluation such that a utility could license such a repair on an as needed basis.,
'Ihe definition of the above items along with the safety evaluation provides a comprehensive package such that the utilities can independently implement and or contract such services, i.e. local weld repair or 360* weld overlay.
Conclusions of the program are:
An acceptable weld overlay process has been developed and qualified to Section IX of the ASME Code.
'Ihe weldit g process specification developed as a result of the qualification is applicable for j
both locd weld repairs and 360* weld overlays in the reactor vessel closure head penetration tuber.
l i
Multiple repair geometries exist, each repair required should be individually specified. An j
e individual utility needs to specify repair requirements based on the technical merits and j
j economic impacts of each repair situ lon.
a An excavation only repair is suggested up to a depth of (
]* C#,
i It is suggested that if excavation to a depth of [
]*C# does not remove the entire defect, excavation should continue until the defect is removed or until [
]* C# inch of the penetration tube wall remains.
A weld overlay repair needs to restore the minimum required penetration tube wall design thickness.
I 1
m:uso6w.wpf:1b.111095 vill
~- -. -
I Repair welding provides an overall increase in the' surface principle stresses in the penetration
- tube. Dependent on weld thickness and circumferential extent the principle stresses will vary, t
1hese residual principle stresses for any of the geometries considered are comparable in magnitude to the residual plus operating stresses estimated via the elastic / plastic analysis for the j
outermost penetration tubes.
Areas of the penetration tube adjacent to the weld may be more susceptible to PWSCC than the alloy 600 base material not impacted by the welding process. However the susceptibility of j
adjacent material quickly dissipates due to the drop off of residual stresses as you move away i
{
from the weld.
The extent to which a utility wishes to pursue post weld surface treatment (s), such as shot peening needs to be an individual utility decision based on the t.:hnical merits and economic
<~
Impacts. The Westinghouse owner's group may consider such a program in the future.
The final geometry and surface finish of the repaired area needs to be such to facilitate baseline j
and potential future volumetric inspections.
Weld design depths, geometry, location, and circumferential extent can be varied in an attempt l
4-to minimize the impacts of the associated welding residual stresses. These variations are i
outlined on the associated design drawings provided in Appendix B. The WCAP report which follows is intended to provide the in depth information required to understand these impacts.
l J
I i
a i
mA25%w.wpf:Ib-111095 ix i
m.~
1.0 INTRODUCTION
Previously, leakage has been reponed from an Alloy 600 reactor vessel closure head penetration tube.
9 in a French plant during h).1ro testing at elevated pressure. Subsequent inspections of the leaking penetration indicate the presence of axial cracks on the inside diameter of the penetration tube. Cracks j
extend above and below the penetration tube to reactor vessel head attachment weld. The leakage has
)
been determined to result from an axially oriented through-wall crack in the penetration tube wall.
The cause of the axially oriented cracks has been attributed to primary water stress corrosion cracking j
(PWSCC), driven by both steady state operating and residual stress. 'Ihe residual stresses have been attributed to the ovality in the penetration tube which is a direct result of bending introduced in the l
1 penetration tube due to the offset geometry of the attachment weld. Reported data from inspections of head penetrations at additional plants (Both French plants and plants of Westinghouse design) has established the presence of axially oriented cracking in additional penetrations, f
f The plants of Westinghouse design with reported reactor vessel head penetration tube inside diameter j
PWSCC are [
]"##. A review of available inspection data would l
indicate that flaws have been detected in approximately 2% to 3% of the penetrations inspected. A f
review of the reported inspection results also indicates that the majority of flaws were detected in penetration tubes located at the periphery of the reactor vessel closure head. 'Ihis finding is consistent j
with estimate that residual stresses are greatest in the peripheral penetrations because the offset in the (or angle of) attachment weld is greatest at these locations.
i i
Reactor vessel closure head penetrations on all Westinghouse supplied plants are of similcr constmetion as that of the French plants and Westinghouse designed plants that have experienced cracking. 'Ihus, based on the character of the cracking and the known potential of the Alloy 600 material for susceptibility to PWSCC this phenomenon may be possible on all Westinghouse plants.
t Currently the WOG has undertaken an extensive program to examine and manage the phenomena of PWSCC initiated from the inside diameter of the reactor vessel Alloy 600 penetration tubes. 'Ihe f
WOG's position has been that U.S. industry should take a proactive but logical approach to addressing l
the issue. Thus the WOG has initiated various project authorizations, outlined below, which are intended to address the various aspects of this issue such that the issue can be tedmically and i
economically managed to a successful resolution.
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1 Understand the cause and extent of cracking experienced by the French in their plants. From this phase of the work the WOG concluded that the issue could impact selected US plants, I
however the extent and/or time frame could not be immediately quantified.
i" Assess the safety impacts of the issue. Detailed engineering analyses were conducted to i
understand the extent and safety impacts of cracking. A generic safety evaluation was
. performed and presented to the NRC. The conclusions were that the issue does not represent an immediate safety issue. The significance of cracking is that it can result in leakage which could e-result in wastage of the carbon steel vessel head. De WOG estimated wastage corrosion rates based on analysis performed by Westinghouse and experimental data provided by the Combustion Engineering Owners Group. De conclusion was that wastage could alter the reactor vessel head however the ASME Code Allowable stresses would be maintained for a minimum of 6 years.
ne experimental data used to estimate crack propagation for the thick-walled Alloy 600 penetradon tubes, which was used in the flaw tolerance evaluation portion of the safety evaluation, was based on thin-walled Alloy 600 tabing. De WOG chose to investigate crack propagation rates in thick-walled Alloy 600 tubing to verify that the crack propagation model for thin-walled tubing was valid for use. Dus the WOG initiated a crack propagation testing program to investigate this phenomenon. Ris work is scheduled to be completed in the fourth quarter of 1994.
De WOG had the opportunity to confirm the mechanism of cracking in the penetration tubes.
=
ne[
]*A' plant, a Westinghouse supplied plant, has also experienced cracking and has undertaken a program to investigate the cracking. As part of the Ringhals program boat samples were removed from the ID of a penetration which has experienced cracking. He WOG was offered the opportunity to perform a failure evaluation on one of these boat samples.
Westinghouse performed this work under authorization [
]"A*
His work further confirmed the French findings that the cause of cracking was PWSCC.
j ne WOG has authorized a report outlining a Flaw Evaluation Procedure which is intended to 1
identify the techniques required to estimate the propagation of any flaws detected by an inspection.
l m:\\2506w.wpf:lts111095 12
The WOG has supported an industry inidative coordinated by NUMARC to develop acceptance criteria for flaws detected along the inside diameter of reactor vessel closure head penetration tubes. _1hese acceptance standards have been provided as the standard for acceptance of any flaws detected during an in-plant inspection. Additionally, EPRI has applied these acceptance standards in developing a qualification program and standards for utilities to use in the qualification of vendors offering inspection services.
1he WOG has also solicited utility volunteers to perform pilot volumetric inspections of their reactor vessel closure head penetrations. The WOG intends to evaluate inspection results and assess the impact on the pilot and other W plants.
j Through these programs the WOG has attempted to determine cause, address the safety significance of
' this issue, develop inspection and acceptance criteria, provide a mechanism to qualify vendors offering inspection services such that interpretation of results across the industry is consistent, and make available pilot inspection results such that the future actions / requirements with respect to this issue relative to the U.S. nuclear industry can be formulated. Lastly, the WOG authorized a program to develop a weld repair methodology for penetrations which have experienced cracking. The following document outlines the program and reports on the results of the weld repair program.
m:u506w.wp01b.111095 '
13
i i-
- 2.0 PROGRAM DESCRIPTION i
2.1
-Objectives i
The objective of the program was to provide a weld desip package which can be applied to repair reactor vessel closure head penetration tube ID initiated PWSCC. The weld design study has investigated repair of partial through-wall and full through-wall cracks. The objective was to I
investigate a local weld repair process and a 360* weld overlay process as part of the weld design package. In addition to the weld repair process,information regarding excavation geometries and post weld surface treatment was investigated. Excavation serves two purposes; l) It provides access for application of the weld, and 2) It serves to remove any existing defects. For the purposes of this
- project authorization the post weld surface treatment investigated was shot peening. The objective of a post weld surface treatment such as shot peening is to negate / mitigate residual stresses induced by welding.
In support of the weld repair processes Westinghouse investigated; 1) Excavation geometries and -
various depths as related to flaw geometry, 2) Limitation of the weld repair with respect to crack length, 3) The definition of welding process parameters,4) The definition of allowable weld filler metals,5) Identification of the weld surface finish, 6) Requirements relative to the surface profile of the penetration inside diameter,7) Industry suggested parameters for shot peening, and 8) Weld inspection requirements. The definition of these items provides a comprehensive definition of the process such that the utilities can independently implement such a repair.
In support of the repair process, Westinghouse performed a generic 50.59 Safety Evaluation such that
)
a utility could license such a repair on an as needed basis. Also, this program provided engineering justification of the process through the preparation of a full size penetration mock-up to provide engineering data to enable evaluation of effects on penetration residual stress and deformation due to the weld overlay. The change in stress was measured using a Hole Drilling Strain Gage Method in accordance with ASTM E837. Mock-up testing was also used to investigate the extent of weld shrinkage associated with the weld overlay process and the extent that the weld overlay process impacts the shrink fit between the penetration tube and reactor vessel head, m:\\2506wiwpf:1b-111095.
2-1
2.2
. Weld Repair Program Outline
'Ihe development of a weld repair design packcge was structured to investigate specific weld process parameters and provide engineering justification for the various associated technical issues. In order to investigate the weld process parameters and technical issues several major program tasks were defined.
Each of these tasks along with a brief description follows:
Task 1 Development of Weld Overlay Repair Process Specification:
The Westinghouse weld repair process specification defines: A weld thickness of
[
]*## to [
]*## inches, defines critical weldi g process parameters, defines a
allowable weld filler metals [
]*## and identifies weld surface finish requirements and inspection requirements.
Also, shot peening as a post weld surface treatment available for mitigation of post weld residual stresses will be discussed. The documentation also defines target shot peening process parameters. Target shot peen process parameters were provided as a result of recommendations solicited from a commercial shot peen vendor and work performed by Westinghouse, independent of this program authorization.
Task 2 Define Penetration Excavation Geometry:
A drawing is supplied to compliment the penetration repair process to define such items as: the excavation geometry and depths for both an excavation only repair and excavation followed by a weld repair, the required ID profile of the penetration ID after the application of the weld overlay, and any limitations with respect to positioning the weld overlay relative to projected stress profiles in the penetrations.
l In addressing excavation of the penetrat?on two aspects were addressed: 1) It was imperative that the structural adequacy of the penetration was not compromise (, this was investigated via a review of available ASME code stress reports on the reactor vessel closure head, and 2) 'Ihe excavation geometry was defined such that adequate m:u506w.wpf:1trti1095 2-2
penetration material was removed such that, application of the weld does not restrict the flow area in the penetration or thermal sleeve movement is not impacted.
Task 3 Provide Evaluation of Applying Weld Overlay Over Existing Cracks:
1he effect of applying weld material over existing partial through-wall and full through-wall cracks was investigated. The applicable ASME Code paragraphs were investigated which discuss leaving cracks in the pressure boundary were reviewed. Also EDM notches were placed in mock-ups to assess impacts on the welding process.
. Task 4 Penetration Mock-up Tests:
A full size penetration mock-up was fabricated. The mock-up was fabricated using an
. alloy 600 penetration tube welded in a plate of low alloy carbon steel using the partial "J"-groove geometry for the attachment weld. The mock penetration tube was skewed to the surface of the plate to simulate the weld offset of actual penetration tube assembled in the reactor vessel closure head. The mock-up was used to investigate the application of weld material in a similar geometry to the penetration tube, and to quantify the addition of any residual stresses on the ID adjacent to the weld repair.
Several mock penetration tubes were also fabricated to investigate the application of various weld thicknesses and geometries. The various weld thicknesses were evaluated for cladding integrity via a cross-section taken through the weld thickness.
Task 5 Generic Safety Evaluation:
A generic 50.59 safety evaluation was performed to aid WOG members in implementing a weld overlay repair at their specific plant sites. The Safety Evaluation is provided as a stand alone document.
l In completing the above tasks the stated goal was to identify engineering justification and appropriate specifications for implementation of a local weld repair and a 360* weld overlay. Both weld repairs involve an appropriate amount of excavation from the penetration inside diameter followed by e
9 e
L
application of filler metal in the excavated area. In the case of the local weld repair the repair is targeted at restoring the minimum required penetration tube wall to maintain the pressure boundary.
For the 360* weld overlay the. Intent is to provide a remedial measure for mitigation of PWSCC. The 360' weld overlay would cover the entire inside surface of the penetration tube most susceptible to PWSCC over some given length.
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3.0 APPROACil FOR DEVELOPMENT OF PENETRATION TUBE WELD REPAIR AND OVERLAY DESIGNS In developing the weld application options for the reactor vessel closure head penetration tubes, two basic designs were targeted; 1) A local weld repair process and 2) A 360* weld overlay process. 'Ihe local weld repair process is tarpe'ed to restore the minimum required design thickness of the penetration tube wall. The 3ou weld overlay repair is intended to provide a remedial measure to mitigate PWSCC in the Alloy 600 penetration tube ID. Refer to Figure 3.0-1 for an overview of the reactor vessel closure head to penetration tube geometry.
3.1 Local Weld Repair In designing a local weld repair several considerations were taken into account:
The weld repair has to restore the minimum required design thickness. The governing design requirement with respect to the penetration tube is design pressure. An examination of a typical l
4-loop vessel head indicates that the required penetration tube thickness to meet design pressure requirements is approximately 0.29 inch.
l r
Slots were examined in [
]*## Reference 6, as a potential repair for the reduction of residual surface stresses in the penetration tube ID. The maximum slot depth examined was
[
]*## inch.
The industry flaw acceptance criteria developed for penetration tubes identifies the depth of an allowable flaw to be 75% of the tube wall thickness or [
]*## = [
]*## inch.
Thus a weld overlay repah !3 a penetration excavated to a depth of [
]*## inch may be required.
i In specifying the circui ferential extent of the local weld repair designs, the stress analysis l
results reported in WCAP-13525, Reference 5, were taken into account as well as the slot widths examined in [
]*## Reference 6. For the purpose of the local weld repair the intent was to position the toe of the weld in an area of the penetration tube ID having relatively low hoop stresses. Thus circumferential extents of 45' and 90* were selected, such m$2506w.wpf:1b-111095 3-1
that the toe of the weld could be approximately located on the 45* axis of the penetration tubes where the hoop stresses were estimated to be low.
Additionally, lengths of 4 and 6 inches were selected to investigate the variations which might occur due to changing the overall weld length.
Based on the above considerations local weld repair design geometries with varying weld thicknesses of[
]# inch, overall lengths of 4 to 6 inches, and having circumferential extents of 45' through 90' were considered for investigation.
3.2 360* Weld Overlay In performing a 360' weld overlay repair the two items taken into consideration were; 1) The weld overlay depth should be thick enough to provide a boundary which prohibits exposure of the Alloy 600 base material to the primary water over the applied length of the repair, and 2) The depth should be minimized such that any associated weld shrinkage minimizes the residual stress in the base material and does not negatively impact the interference fit on the OD of the penetration tube between the reactor vessel closure head and penetration tube.
Based on the above a [
]* *# inch weld thickness was judged as appropriate to meet the above two criteria. A thickness of [
]*'*# inch is approximately [ ]*'*# weld passes. However, an overlay need not be limited to [
]*'*# inch. Weld overlay thickness of [
]^*# inch were investigated for lengths varying from 4 to 6 inches.
'Ihe perceived advantages of the weld overlay are; l) Application of the weld overlay can be a continuous process using a spiraling application, and 2) both ends of the weld overlay can be readily j
positioned in lower stress regions of the penetration tube ID.
3.3 General Program Goal In order to evaluate the above defined design geometries a series of tests and measurements were 1
-identified for investigation of a weld process which could be qualified to the ASME Section IX Code requirements, Reference 2. Additionally, these test and measurements were used to assess technical m:\\2506w.wpf:1b-1 t1095 3-2 l:.
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impacts such that the specification of weld repair would not negatively impact the penetration tube geometry. These tests and measurements involved the fabrication of penetration tube samples and a full size reactor vessel closure head / penetration tube mock-up as well as the investigation of methodologies for performing weld overlay repairs. 'Ihe following sections provide the details and results of these investigations.
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PENETRATION. TUBE SAMPLE & REACTOR VESSEL HEAD / PENETRATION T 4.0 MOCK-UP FABRICATION 4.1 Preparation of Penetration Tube Samples Figure 4.1-1 depicts the geometry of the grooves machined in the 10 inch penetration tube samples.
The grooves were machined using electric discharge machining (EDM). As shown in Figure 4.1-1 the tube wall was machined to the defined depth and made use of a [
]*** taper to blend the excavation
~
depth into the original tube inside diameter (2.75 inch). The [
]"C taper was applied both circumferentially and axially. For the groove depths of[
]*' inch and [
]*** inch the [
]***
taper resulted in an acceptable geometry. However, for those samples with a groove depth of
[
]***# inch the taper was reduced to a ratio of [
]*****. The taper was reduced because the
[
]***# taper was impractical from the standpoint that it extended two far around the penetration circumictence, requiring too much weld filler metal to fill in the taper transition area. After l
performing weld repaim on the [
]***# taper geometry process time was still too long and too much weld filler metal was s'ill required, thus an alternative transition design was identified foi blending from the excavation d:pth to the inside surface of the penetration tube. The alternative transition is a j
typical weld "J" preparation applied in the industry and is depicted in Figure 4.1-2.
l 4.2 Fabrication of Reactor Vessel Closure flead/ Penetration Tube Mock-Up A full scale mock-up of the reactor vessel closure head and penetration tube was fabricated to depict the most peripheral penetration in a 4-loop reactor vessel head, thus indicative of a penetration tube with the greatest offset in the attachment weld, i.e. therefore the maximum residual stress. Fabrication sketches of the mockup are provided in Appendix C, Fabrication Data Package for the Penetration Mock-Up & Penetration Mock-Up Sketches. ' Die fabrication data package includes as-built dimensional data.
To validate the applicability of the mock-up, measwements were taken of the penetration tube inside diameter to measure the ovality which occurred as a result of performing the mock-up attachment weld. As in the actual reactor vessel head geomeay, s "J" groove weld prep was used for the attachment weld between the penetration tube and low carbon steel plate. The maximum ovality m:\\2506w.wpf:1b-111095 4-1
(major diameter less minor diameter) which occurred in the mock-up was [
]*## mils ([
]*##
inch) as compared to the maximum approximated ovality of [ ]*## mils ([
]*## inch) estimated from the linear regression equation for ovality which was developed based on actual plant ovality measurements, Reference 1.
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i 5.0 WELD PROCESS SPECIFICATION A welding process specification, which can be used for either the local repair or application of the 360* weld overlay in the reactor vessel closure head penetration tubes was generated and is attached in Appendix A of this WCAP report. The welding process specification is written to provide guidance for the qualification of welding procedures to be used for the performance of welding in Westinghouse PWR reactor vessel closure head penetration tubes. The parameters recommended in the specification were based on the welding operations performed for this feasibility study. 'Iherefore, the parameters were qualified for the intended applications to the extent as discussed in the following paragraphs.
5.1 Selection of Welding Equipment An automated pulsed gas tungsten arc welding (GTAW) system designed and manufactured by The
[
]* * (power supply model 215, Figure 5.1-1, and model 94 ID cladding and welding head, Figure 5.1-2), was selected for this program. The model 94 weld head is designed for spiral cladding and groove welding inside diameters as small as 2 inches. The model 94 provides arc rotation, axial (linear) travel, filler wire feed and arc voltage control (AVC) for are gap control. The combination of axial travel and arc rotation provides a spiralling effect directly applicable for use in a 360 degree weld metal overlay process.
To demonstrate the capabilities of the selected automated welding system and identify target welding parameters a pipe ID weld overlay was performed on a 2 inch nickel base alloy pipe with inconel 82 filler metal. The current design of the model 94 weld head feeds a 0.030 inch diameter weld wire.
The filler metal of choice for this program, [
]* *** was not available in 0.030 inch diameter at the time of the demonstration. A 20 lb. spool of 0.035 in. diameter [
]**** filler metal was obtained and reduced to the required 0.030 inch diameter.
5.2 Qualification of the Welding Parameters l
'Ihe intent of qualifying the parameters at the beginning of the program was to ensure that the starting parameters were appropriate for use with the [
]*** filler metal. 'Ihe starting parameters were based on the parameters used with the [
]**** filler metal during the demonstration of the welding system. This approach was taken due to the limited supply of the [
]* C
- filler i
1 m:U506w.wpf:1b-111095 5-1
i wire at the beginning of the program, and the long lead time required to reduce the diameter of the available weld wire to 0.030 inch. The weld wire situation prevented any practice welding to establish welding rarameters in advance with the [
]#.
Two alloy 600 pipe assemblies were welded using [
]*## filler metal to qualify the parameters to the ASME Section IX mechanical test requirements. Four 5-inch long pipe samples were machined with 37.5' grooves as shown in Figure 5.2-1. The 37.5* groove was machined starting from the ID of the pipe and finishing the groove at the OD of the pipe so that the groove could be welded from the pipe ID.
Starting process parameters for welding the pipe assemblies with the [ ]*## filler metal were those process parameters used in the demonstration with [
]*#. 'Ihe parameters were adjusted as welding progressed. Some difficulties were expedenced in welding the first assembly, during the initial two layers burn-through and stucl: wire in the weld puddle occurred. Once the parameters were adjusted based on the difficulties, there was no problem with the subsequent layers of the first assembly or the second assembly. Upon completion of welding the two pipe assemblies, mechanical test coupons, i.e., tensile and bend (face and root) specimens, were machined from each assembly in accordance with ASME Section IX requirements. All bend specimens were free of cracks with the exception of the root bend specimen of the first assembly. The failure of the root bend was attributed to the difficulties experienced as explained above.
as During welding of the qualification pipe assemblies it was observed that inconel 52 filler metal has a very sluggish characteristic, even worse than [
]"##. This may be due to higher contents of l
Cr, Fe and deoxidizers such as Al and Ti in [
]*## compared to [
]"##. The
[
]*## filler metal mixed well with the alloy 600 penetration tube producing a relatively smooth surface, as was observed in the first layers of the pipe assemblies. 'Ihe subsequent layers, however, started showing the sluggish characteristics which produced a relatively rough surface in comparison.
In general the surface condition of a weld is controlled by grinding or machining operations after welding. However, considering the actual field applications of this process it was desirable to improve the surface condition through weld process controls such that no gdading operation would be required after repair welding. As an attempt to improve the surface finish a [
]*##
m:u506w.wpf:lts111095 5-2 L_
mixture of shielding gas was tried during the welding of the second pipe assembly. The
[
]*## mixture gas was tried because it was readily available for a similar application on a nickel base alloy. The change in the shielding gas did not improve the surface finish of the as-welded condition. Thus the shielding gas was changed back to [
]*## gas. Welding process parameters were adjusted during welding of the subsequent test tube samples to maximize the quality of the final surface finish.
5.3 Welding of Penetration Tube Samples Table 5.3-1 shows the matrix of the eight [
]*## penetration tube samples and their respective geometries. Repair welding of the tube samples started with sample number 4, which had a 360 degree groove of [
]*## inches deep. Although the welding system was capable of welding die groove in one spiral operation the operation was stopped every one (1) inch or so to maintain the interpass temperature below [
]*## maximum. The [
]*## interpass temperature was selected becarse this is typical industry practice for minimizing distortion in stainless and nickel base alloys. Those samples with a partial groove, [
]*##, required a similar interpass temperature control. It should be pointed out that the samples with a partial groove took a much longer time to weld due to the setup required for every pass. Each weld pass was performed circumferentially for this program. The necessity of a setup for every pass could impose some i
difficulties on actual field applications for repair welding and special attention should be given in l
development of field tooling to minimize this impact.
t As explained in the previous section during welding of the [
]*## tube samples the parameters were adjusted to improve the weld surface finish, such that the smface smoothness could be maximized. Although surface finish appeared to be adequate, more improvement would appear to be possible. Welding of additional samples for further adjustment of parameters would be beneficial as well as investigating the use of other shielding gases. Another possible shielding gas would be a helium / argon mixture. Other options, such as a combination of [
]*## with [
]*## and/or
[
]*## on the last layer should be considered.
As indicated in the Table 5.3-1 tube sample number 7 and 8 included EDM notches in the repair area.
This was to study repair welding over [
]*##. Figures 5.3-1 through 5.3-3 depict the cross sections of repair wekts over the EDM notches. The notches were mA2506w.wpf:1twIi1095 5-3
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- approximately [
]*## laches deep and [
]*## inch wide. The metallography samples of the notches showed no cracks or indications generated in the surrounding area due to the welding.
Considering [
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5.4 - Welding Reactor Vessel Closure Head / Penetration Tube Mock-Up 4
h Two EDM grooves, Figure 4.1-2, were machined in the penetration mockup to simulate weld repairs in the plant. It was learned from the [
. ]*## penetration tube samples that a 360' groove would be much easier to weld repair as opposed to the partial groove with the welding system available.
l Thus, partial grooves were selected for the mockup to investigate the potential difficulties which might '
be experienced in a field application. 'Ihe [
]*## inch groove depth was selected for the partial grooves as the most probable thickness of weld overlay to be used in a field application.
Repair welding the excavation areas in the mockup were performed very much the same as in the f
penetration tube samples. Since the mockup, Figure 5.4-1, had more mass to transfer the heat during welding it was not necessary to stop the welding operation as often as in the [ )*## inch penetration tube samples, to meet the [
]*## interpass temperature requirement. It is estimated that the i
interpass temperature control may not be a concern with the field application due to the mass of the l
penetration tube and surrounding reactor vessel closure head.
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Table 53-1 TEST MATRIX FOR [
]*## PENETRATION TUBE SAMPLES i
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]*"
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i 6.0 EVALUATION OF WELDED PENETRATION TUBE SAMPLES i
The penetration tube samples were used to evaluate the feasibility of welding within the 2.75 inch diameter of the penetration tube and to evaluate the impacts of the various selected geometries. As defined in Table 5.3-1, eight penetration tube samples were selected to explore the various weld repair j
geometries. The overall weld length, circumferential extent and depth were varied.
f l
6.1 Discussion of Diametral Measurements i
To evaluate the penetration tube samples each sample had pre and post weld dimensional data taken.
The measurements were taken across both the inside and outside diameter in 0.5 inch increments over the entire length. The outside diameter measurements were used as the primary mechanism for comparison as opposed to inside diameter measurements in order to avoid variations resulting from the weld surface finish and the weld applied thickness. Figures 6.1-1 through 6.1-8 provide plots of the l
dimensional data. The dimensional data as-measured pre and post welding is provided in Appendix D, Penetration Tbbe Dimensional Data.
f The penetration tubes were scribed to retain the orientation of the axis,i.e. 0*,45*,90*, and 135*.
l The outside diameter measurements taken across each of these axis were very consistent and on the average were 4.000 +/- 0.001 inch. The pre-weld diametral measurements were averaged and plotted I
as a single line on Figures 6.1-1 through 6.1-8. Post weld measurements were taken across the same axis and are plotted individually on each of their respective figures.
l Based upon a review of Figures 6.1-1 through 6.1-8 the following observations were made:
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s Regardless of the weld length (4 or 6 inches) the diametral dimensions are impacted over a
=
length approximately 1 inch greater than the weld repair length. Recall the weld repair lengths do not include the taper length which is also filled with weld material, This would indicate that an approximately 0.5 inch transition zone exists from the end of the repair depth where weld shrinkage impacts the diametral measurements. This transition zone appears to independent of j
weld depth or taper length.
m:\\2506w.wpf:Ib.111095 6-1
1 A 360' weld repair results in deformation across each axis. The deformation is approximately uniform for each axis, Refer to Figures 6.1-2, 6.1-4, and 6.1-8.
l The deformation associated with a 90' weld repair also impacts each axis, particularly those
)
axis 45' from the weld centerline (i.e. primary axis). The 45* axis experiences deformation approximately 30% to 40% of the primary axis. The axis 90' from the primary axis appears to be the icast impacted. See Figure 6.1-3 and 6.1-5.
In all penetration tube samples the deformation, resulting from weld shrinkage, appeared to result in a decrease of the outside diameter except over a very few number oflocal positions, See Figures 6.1-1, 6.1-5, and 6.1-7.
On the average the deformations resulting from the various weld depths are:
Weld Depth Average Deformation Maximum Deformation (inch)
(inch)
(inch)
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]*# penetration tube samples.
It is judged that deformation in the actual plant penetration tubes would be less because of the available mass to dissipate the welding heat input.
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7.0 RESIDUAL STRESS MEASUREMENTS ON REACTOR VESSEL HEAD /.
j PENETRATION TUBE MOCK-UP 7.1 Approach to Residual Stress Measurement To determine the residual stresses buildup from welding the reactor vessel closure head and penetration tube full scale mock-up fabricated for this program was used. The fabrication of the mockup was described in a previous section of this report. De hole drilling method of residual stress 1
measurement was used for these measurements.~ A sketch of the head penetration model and test '
fixture.is shown in Figure 7.1-1. All of the residual stress measurements were made on the ID of the tube.
- De residual stress measurement program was divided into three steps:
[
0 4 -
1 ja c.e 1
i 7.2 Hole Drilling Method p
His method involves mounting a three strain gage rosette at the location the measurement is required.
A sma't hole is drilled at the center of the rosette and the rellerad strain is measured by the three gages ol'the rosette. De relieved strain and elastic constants of the material and constants for the rosette are used to calculate the residual stress. He rosette constants are obtained by calibration, either by the rosette manufacturer, or using the ASTM standard practices. The rosettes used were procured from Micro Measurements, gage model [
]***. His is a -
special three element 45' rosette in a circular pattern. The hole drilling method measures a near surface residual stress and is described in ASDI standard E-837-92. Stress is assumed to be uniform, i.
' or at worst,' varying uniformly through the thickness of the object measured. For a uniform stress field the accuracy is estimated within [
]* *.
mA2506w.wpf:lts!11095
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7.2.1 Installation of Strain Gage Rosettes Rosette locations for each step are shown in Figures 7.2-1,7.2-2, and 7.2-3. These figures depict maps of the inside surface of the penetration tube and show the angular position and distance from the inside end of the penetration tube. De rosette locations are also tabulated in Table 7.2-1. Rosettes were oriented with the number one gage in the axial direction of the tube.
The ID surface of the tube was prepared for installation of the strain gages by first cleaning with a chlorothen degreaser. The surface over which the strain gage rosettes were installed was dusted with micro sand blasters to give a mat finish for better adhesion of the strain gages. For mounting strain gage rosettes in the EDM machined areas, in Step 2, the surface was first smoothed [
] ***#. De welds in the weld repair area, in step 3, were ground to a flat surface suitable for strain gage installation. [
]*# adhesive was used to bond the gages.
7.2.2 Drilling Holes The setup for the residual stress measurements on the model is shown in Figures 7.2-4 and 7.2-5. Air abrasive machining was used to machine the holes. A special fixture (shown in Figure 7.1-1) was made to position the drill to target the center of the rosette. He rosettes are masked before drilling to protect them from the abrasive. Strain readings are taken before and after drilling. Hole depth is determined by air pressure, abrasive size, noz2le diameter and time. [
ja c.e 7.3 Test Results The principal stresses and directions were calculated using the relieved strains and equations in ASTM E 837. [
]C# Re relieved strains were corrected for transverse sensitivity and gage factor variations. Dese factors are provided by the strain gage manufacturer (see Figure 7.3-1). The equations for the calculation of residual stresses are:
m$2506w.wpf:1b-111095 7-2
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- Ihe equation for calculating the angle C from gage 1 of the rosette to the nearer principal stress is:
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Ihe relationship of principal stress directions to the rosette is shown in Figure 7.3-2.
f i
The results of the residual stress measurements are given in Table 7.3-1. Residual stress versus J
distance from the end of the tube is shown in Figures 7.3-3 and 7.3-4.
7.4 Comparison of Test Readts to Analysis i
I Elastic / Plastic analysis of the reactor vessel closure head / penetration tube geometry has been
)
performed and documented in WCAP-13525, Reference 5, also several repair geometries have been 2'
analyzed and documented in [
]**#, Reference 6. 'Ihe elastic / plastic analysis are of particular interest for comparison with residual stress measurements taken in the reactor vessel closure i
g i head / penetration tube mock-up, because the measurements serve to validate both the analysis and measurements. Also, the repair geometries examined local grooves (i.e. slots) as measures to reduce I
penetration tube residual stresses.
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While the hole drilling technique is a fairly accurate means for the measurement of residual stresses it should be noted that the measured values represent an average stress over the depth of the hole,
[
]***** inch in this case 'Ihus the measured stress value is slightly below the actual surface stress on the order of magnitude of 10%. The finite element analysis provides a calculation of the surface I
a l
' stress. Figures 7.4-1 through 7.4-4 provide plots of the penetration tube residual stress as calculated j
- after welding (as-opposed to the residual + operating stress) as compared with the measured stress l
values. Figures 7.4-1 and 7.4-2 plot hoop stresses while Figures 7.4-3 and 7.4-4 provide plots of the 1
axial stresses. Also, the plots distinguish between the penetration tube center side (180* orientation on Figures 7.2-1, 7.2-2, and 7.2-3) and hill side (0*/360' orientation). 'Ihe plots depict in general the t
4 2
same trends (peaks and valleys) between the measurements and the finite element calculations, also.
j*
fairly good quantitative agreement exists, particularly for the hoop stress values.
l J
Several other observations / comparisons were drawn from the hole drilling residual stress measurements and finite element calculations (It should be noted that the residual stress measurement locations in 1
Table 7.3-1. identified with the same numerical value are approximately positioned with the same t
coordinates):
7he machining of the grooves generally appeared to lowe'r stresses at the location measured.
l.
Hoop stresses were decreased at locations 4a, 8a, 9a,12a,14a and increased only at
)
I locations 3a. Axial stresses were decreased at locations 3a, 8a,9a,12a,14 and increased only
\\
j at location 4a. This generally supports the conclusions made in the analytkal study of repair configurations, Reference 6.
4 l
Weld repair areas have fairly high residual stresses, the greatest measured value being a 1
principal stress of [
]*'*'* ksi, see location Ilb on Figure 7.2-3. Although fairly high this I
i-l value is comparable with the calculated surface stresses.
i j
' Tensile stresses adjacent to the weld as indicated are fairly high but dissipate rather quickly, see l
locations lib and 16b on Figure 7.2-3.
Adjacent to the weld the axial / hoop stresses are l
[
]*'*** ksi respectively, but drop to [
J ksi less that 1 inch away.
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i-The penetration tube stresses approaching the 45' axis are expected by analysis to be low approaching compression. A review of these stresses after welding, see location 12B and 14b, in fact have compressive axial stresses of [
]*##and[.
]*## ksi with low hoop stresses l
of{
-]*## and [
]* *# ksi.
Axial and hoop stresses in the alloy 690 weld repair are higher than their corresponding values j
before welding, hoop stresses increasing by approximately [ ]*## ksi with the largest increase being in the axial stress components [
]**# ksi to [
]*## ksi and [
]*## ksi to
[
]*## ksi,'see locations 3/3b and 9/9b.
A review of measured principal stress in the penetration tube weld region prior to and Jter welding indicate an overall increase in surface stresses.
Although the individual measured stress components (axial and hoop) prior to and after welding indicate an overall increase in surface stresses the after welding values are comparable to calculated values. Again, Figures 7.4.1 through 7.4.4 provide the calculated and measured stress component values prior to welding.
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ROSETTE LOCATIONS ON THE ID OF THE TUBE a,c.e i
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Table 7.3-1 RESIDUAL STRESS MEASUREMENTS IN REACTOR VESSEL HEAD /
PENETRATION TUBE MOCK UP a,c.e l
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8.0 - DISCUSSION OF POST WELD SURFACE TREATMENT Post weld surface treatment of welds is typically performed to serve one or all of the following f
functions-
)
i 4
1)
Impmve the surface finish such that an acceptable surface is provided for performing post-weld l
Inspections and/cr future penetration tube inspections.
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2)
Provide an emeptable geometry such that the function of the component is not negatively
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Impacted. h me case of the penetration tube inside diameter, the inside diameter can not be J
5 4
reduced such that it impacts the thermal sleeve or reduces the flow path in the penetration tube j
ID to th?rmal sleeve OD annulus [
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3)
Mitigate the residual rtresses in the weld metal and adjacent base material which occur as a
{
I result of the welding process.
l In developing process requirements for welding, regardless if it is to be used as a mitigative measure i
i
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for PWSCC (360* overlay) or a repair to restore the penetration tube pressure boundary (local repair),
j i
items (1) and (2) above are intended to be addressed via process controls. The post weld surface l
3 i
l finish, item 1, and the post weld geometry, item 2, are intended to be controlled via weld process and I
inspections requirements. In order to address item 3 the WOG requested via the program authorization that shot peening be examined as the remedial measure for the mitigation of residual stresses induced by welding.
{
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8.1 General Discussion of Shot Peening l
l 4
^
Shot peening is a cold working process in which the surface of the material is bombarded with small j.
spherical media called shot. Each piece of shot striking the material acts as a tiny peening hammer, f
imparting to the surface a small indentation or dimple. In order for the dimple to be created, the l
surface fibers of the material must be yielded. The cold working process results in the application of beneficial compressive stresses being applied at orjust below the material surface. Compressive stresses are beneficial in increasing resistance to fatigue failures and stress corrosion cracking.
Benefits obtained due to cold working include hardening, intergrannular corrosion resistance and i
1 mA2506w.wpf:Ib ll1095 8.]
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' urface texturing., Westinghouse has investigated shot peening as a mitigative technique for s
spplication to the Alloy 600 penetration tube to increase the materials margin against primary water-stress corrosion cracking. (PWSCC).
The maximum value of the residual compressive stress is often cahed :1,e magnitude of the residual stress induced. Variations in the shot peening process have little effect on the magnitude of the compressive stress induced as long as the shot used is at least as hard as or harder than the material
' being peened. ' The magnitude of the compressive stress is primarily a function of the base material
- mechanical properties. As a general rule the magnitude of compressive s:ress induced has a value of at least one half the yield strength to a maximum of approximately 60% of the ultimate tensile
~ strength. For the minimum allowable mechanical properties listed for alloy 600 SB-166 & 167 this relates to a compressive stress range of [
]*".
The energy of the shot is a function of the media size, material, hardness, velocity and impingement angle. In order to specify, measure and calibrate peening energy a method utilizing SAE1070 spring steel specimens, called Almen strips, was developed. There are three standard Almen scales currently in use, each based on a different Almen strip thickness. The three scales are the "N", "A", and "C" scale in increasing order ofintensity. The depth of the compressive layer is proportional to the Almen intensity. It should be noted that the magnitude of the compressive stress induced is independent of the compressive layer depth. Peening depth needs to be examined from two aspects; 1) The greater the depth the larger the impact on surface or subsurface material imperfections, and 2) The stress distribution through the component has to be balanced, thus for the case of the penetration tube wall
. an increase of compressive stress on the ID results in an increase in tensile stress on the OD.
The maximum benefit of shot peening is realized when the surface is uniformly peened to a saturation energy level. Saturation is defined as the earliest point where doubling the exposure time produces no more than a 10% increase in Almen intensity.
8.2 Shot Peen Parameters Westinghouse performed a feasibility study to investigated shot peening for the reactor vessel closure head penetration geometry. The two primary objectives of the feasibility were; m:\\2506w.wpf:1b-111095 82
Show that tensile stresses on the inside diameter of the' test chamber (penetration tube mock-up) before shot peening were [
]*##.
Confirm that shot peening reduces these inside diameter tensile stresses to [
]*##.
'Ihe intent of these two objectives were to produce stresses in the penetration tube above the estimated threshold to PWSCC such that the penetration tube test chambers werc susceptible to PWSCC. The -
shot peen process investigated did successfully reduce the susceptibility of the test chamber sample material to stress corrosion cracking in a series of laboratory tests.
Through the specification of process control parameters an Almen intensities of [
]*## on the "N" scale were developed in the test chambers resulting in a compression layer depth of approximately
['
]***#. It was estimated that the magnitude of compressive stress induced was [
]*##, approximately [
]*## of the ultimate tensile strength of the material used in the test.
Subsequent discussions with commercial shot peen vendors have indicated that it should be feasible to develop Almen intensities of approximately 8 on the "C" scale resulting in approximate compressive depth layers of [
]*##. Although the Almen intensity scales can not be directly related the approximate reladonship between the two scales is: N = 0.lC or 10N = C.
8.3 Conclusions Regarding Post Weld Surface Treatment A properly controlled shot peening process should a reliable remedial measure for the mitigation of residual tensile stresses associated with a weld overlay repair, it appears feasible that a shot peen process can be developed which would apply a compressive stress to the surface of the base material on the order of [ ]*## ksi or greater dependent on the base material properties to a depth [
]*##. Such a process should increase the margin against PWSCC in the alloy 600 base material both in the heat affected zone adjacent to the weld and generally throughout the penetration tube ID.
Much investigation has been given to the development of approaches to provide margin against cracking in the weld toe profile. One common methodology is to grind the weld toe profile such that mA2506w.wpf:1b.111095 83
i the geometric discontinuities are removed from this area. This practice could also prove beneficial to 4 -
the penetration tube ID, either performed by itself or in combination with shot peening. The extent to l
which a utility wishes to pursue post weld surface treatment needs to be an individual utility decision based on the technical merits and econonic impacts. Clearly all post weld surface treatments add margin to weld life, each having its individual implementation costs and radiological impacts.
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- 9.0 L DISCUSSION OF WELD OVERLAY REPAIRS '
9.1 Penetration Tube Repair Parameters Generally, prior to the implementation of a weld overlay repair, any detected flaws will be evaluated I
against the industry acceptance standard using flaw evaluation techniques to determine if the flaws can be accepted as-is or need to be repaired. If repair is required or the' utility chooses to implement a f
repair, the next appropriate repair would be the removal of the defect. Ifit is either determined by j
volumetric inspection or during the course of defect removal that the minimum required penetration tube wall thickness is violated a repair of that penetration location would be required. As investigated via this WCAP report a weld overlay repair is viable option for that repair.
l 9,1,1'-
Excavation Depths and Weld Thickness As defined, the minimum required penetration tube wall thickness is approximately 0.3 inch.
Excavation de% which leave a remaining wall ligament of less than the required design thickness,
--0.3 inch, would require a build up of the penetration tube wall. Additionally, another factor should be consider in specifying excavation depths. Excavation of the penetration tube wall and subsequent repair welding could result in a heat affected zone in the reactor vessel closure head base material. To avoid having to perform a post weld heat treatment of the weld repaired area and adjacent reactor vessel closure head base material it is suggested that some minimum ligament be maintained in the f
penetration tube wall. [
]"A* It is judged that this thickness could be directly applied for use in repair of the penetration tube wall. Thus during excavation it is suggested that a minimum penetration tube wSI thickness (ligament) of [
]"A*
Inch be maintained. Based on the above discussion the following criteria are suggested for repair of
+:
reactor vessel closure head penetration tubes:
r P
Any defects detected in the penetration tube wall surface should first be repaired by excavation.
j No additional repair is required if the excavation depth does not violate the minimum required design basis thickness, approximately 0.3 inch.
f m:\\2506w.wpf:ltrit1095 91
?
If excavation to a depth'of [ -
- ]*## inch does not remove th. entire defect, excavation should continue until the defect is removed or until [
]*## inch of the penetration tube wall remains.
i A weld repair to restore the minimum required design thickness needs to take into consideration j
the remaining acceptable penetration tube wall thickness such that the acceptable tube wall after repair welding is 0.3 inch or greater. For example; i
If the flaw were through wall, no remaining acceptable penetration tube wall thickness would i
exist and the minimum required weld overlay thickness would be 0.3 inch.
i Conversely, If the remaining acceptable tube wall thickness were [
]*## inch, the minimum required weld overlay thickness would be [
]*## inch, such that the total thickness was 0.3 inch.
9.1.2 Repair Geometry l
As reponed welding does provide an overall increase in the surface principle stresses of the penetration tube. 'Ihese residual stresses are comparable in magnitude to the maximum residual plus operating stresses estimated via the elastic / plastic analysis for the outennost penetration tubes. It is difficult to quantify the impacts this increase in stress would have on the susceptibility of the alloy 600 base material to PWSCC. However,it would seem appropriate to estimate that the areas of the penetration tube adjacent to the weld would be more susceptible to PWSCC than the alloy 600 base material not impacted by the welding process. Of course, the alloy 690 weld filler metal should not be susceptible to PWSCC as compared with the base material.
l As discussed previously, the toe of the weld could potentially be positioned, by design,in areas of the t
penetration tube estimated to initially have relatively low stresses by comparison. The intent being that the increase in stress due to welding will result in final stresses of lower magnitude than if the toe I
of the weld were positioned in a high stress region initially.
i These considerations directly impact the selection of weld repair circumferential extent, f
5 i
. m:\\2506w.wpf:1b-111095 92 r
il
. As investigated in this WCAP, if it is desirable to locate the toe of the weld outside the comparably
- high stress zones in the penetration tube ID, the circumferential extent of the weld should be selected such that it falls along the [-
)."##' Or as discussed in Section 8.0, post weld-j surface treatment (s) could be used as a means to possibly mitigate the residual stresses induced by
}
welding.
t De specific local weld repair geometry a utility wishes to pursue needs to be an individual utility decision based on the technical merits and economic impacts. Westinghouse dawing [
],"##
attached in Appendix B, depicts the various weld repair geometry requirements and suggested repair j
I profiles.
Drawing [.
]"## also depicts the geometries associated with a 360* weld overlay. As stated earlier a 360* weld overlay repair was investigated to offer a remedial repair which could generally be j
implemented to mitigate PWSCC in the highly susceptible region of the penetration tube ID.
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9.1.3 Weld Surface Finish
'i i
The surface finish achieved in the application of a local weld repair or 360* weld overlay in the reactor vessel closure head penetration tubes is important from two aspects; 1) An acceptable weld l
surface finish is desirable to permit inspection of the weld and penetration tube base material, and
The intent as discussed in development of the weld process parameters was to refine the parameters such that the best possible surface finish could be achieved. De goal was to achieve a surface finish that would permit the volumetric inspection (ECT and/or UT) of the weld filler metal and base metal without having to rework the weld surface finish by some post weld machining operation. While rework of the surface is permissible the intent was to avoid the time and cost associated with rework of the surface. A realistic target surface finish judged to be achievable via the weld process and yet permissible for volumetric inspection was [
)."##
In the development work performed a[
]"_## was achieved over limited lengths of applied weld, but over the full 6 inch length weki applied in the penetation tube samples the [
]"## surface finish was i
not maintained.
m:uso6w.wpr.1b.111095.
93 i
-It is suggested that the final check / qualification of the applied welding process should be verification that the final weld geometry / surface finish could be volumetrically inspected, using ECT as a minimum.
9.1.4
. ASME Code Approach to Weld Repair Repair welding is intended to be performed to the guidelines established in Section XI of the ASME -
Code. However,Section XI does not specifically define guidelines for what depth of flaws must be repaired in the reactor vessel closure head penetration tube ID. In applying weld repair to re-establish the minimum required design thickness of the penetration tube wall no code ambiguities seem to exist for the case where the defects have been totally removed. [
ja.c.e a,c.e 9,1.5 Post Weld Inspection Requirements ASME Code Section XI Subsection IWA-4500 outlines the guidelines for inspections of repair welds made to pressure boundary components. 'Ihe code requires that a baseline volumetric inspection be perfonned of the weld repair for future reference, this is also consistent with the general guidelines m:us06w.wpf:1b.111095 94
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outlined for repair welds made to base metal by the component fabricator, ASME Section III i
. Subsection NB-4130.
9.2 - Conclusions -
i la summary the following conclusions are made:
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10.0 REFERENCES
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ASME Boiler and Pressure Vessel Code,Section IX, " Welding and Brazing Qualifications,"
1989 Edition, ASME, New York, New York, July 1,1989 r
' ASME Boller and Pressure Vessel Code,Section XI, " Rules for Icaervice Inspection of Nuclear 3.
I Power Plant Components," 1989 Edition, ASME, New York,New rork, July 1,1989 t
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APPENDIX A WELDING PROCESS SPECIFICATION l
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REPAIR WELDINO OF REACTOR VESSEL CLOSURE HEAD PENETRATIONS
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-SAFETY REQUIREMENTS: Personnel responsible for welaing application shall have a safety andi ladustrial hygiene program for handling hazardous materials and arc welding equipment (ANSI B7.1, Z43 and 249.1) r I
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'Ihis process specification is applicable to the Safety related ASME Code items. The i
applicable code issue and/or other requirements will be specified in the equipment specification or procurement document.
1.3.
This process specification is intended as a guide for the qualification of welding procedures and for the performance of welding on Westinghouse Nuclear Steam Supply System Components. Any exceptions to or deviations from the requimments of this specification must be documented in writing and submitted to WNTD (Westinghouse Nuclear Technology Division), Materials and Engineering Mechanics, at the time the welding procedure is submitted for approval.
2.0 REFERENCE DOCUMENTS t
2.1
. ASME Boller and Pressure Vessel Code Secdon IX " Welding and Brazing Qualifications".
2.2 ASME Code Case 2142.
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a,c,e 2.4 Additional documents that may be referenced in design specification, drawings and/or procurement documents.
3.0 MATERIALS
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3.1 Base Materials 3.1.1 Nickel-Chromium-Iron Alloy base materialin the solution annealed condition, ASME Section IX classification P-43.
3.2 Filler Materials i
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3.4 Shickline Gas 3.4.1 The shielding gas shall be welding grade [
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4.0 PROCEDURE REOUIREMENTS 4.1 Oualification 3
All weld procedure specifications and welding personnel shall be qualified to the
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requirements of Section XI of the ASME Code. Excepdons to this requirement will only be permitted by written approval of.W NTD, prior to any welding being performed on components.
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l 4.3 Joint Geometry & Preparation 4
4.3.1 Weld joint geometry shall be in accordance with the drawing number I
]*' attached in Appendix C.
4.3.2 The joint geometry shall be prepared by [
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4.4 '
Electrical Characteristics I
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Weldine Position l
All welding shall be done in the horizontal (2G) position where possible.
4.6 Preheat and Internass Temocrature a,c,e 3
4.7 Postweld Heat Treatment
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Postweld Heat Treatment (PWHT) is not required, nor permitted, unless specified in design specification, design drawings or other contractual documents.
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4.8 Techniaue 4.8.1 Filler metal diameter shall be suitable for the base material thickness and weld I
joint configuration used in the component. [
]* * diameter is required for the parameters in table 1.
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4.8.2 Deposition Method l
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4.8.3 Interoass Cleanine -
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Tooline & Fixturine I
4.9.1 Discretion shall be used in the selection of material for tooling and fixturing for parts being welded such that there will be no detrimental effects to the weldment due to contamination as a result of heating, rubbing, smearing or excessive clamping pressure.
a,c.e 5.0 OUALITY ASSURANCE 5.1 Fabricators Quality System, Quality Release Requirements, Data Packages, and witness and notification points, when required, shall be as specified in the procurement documents.
5.1.1 Weld procedures shall be submitted to W NTD, or its designee, for review and approval. Any deviation from the requirements of this specification shall be
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resolved by E NTD, Materials and Engineering Mechanics, as specified in Paragraph 1.3 of this specification.
4 5.1.2 All nondestructive examination procedures shall be submitted to E NTD, or its designee, for review and approval. Any procedure requirements not in compliance with Code or procurement document requirements shall be resolved by E NTD, as specified in paragraph 1.3 of this specification.
5.2 All welding inspections shall be in accordance with applicable code requirements j
and/or design specifications and drawings.
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APPENDIX B
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WELD REPAIR DRAWING 4
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APPENDIX C DATA PACKAGE FOR THE PENETRATION MOCK-UP PENETRATION MOCK-UP SKETCHES i
All of this section is proprietary '# f
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This Appendix C, pages C-1 thru C-31 contains detail dimensional data on the penetration mock up test piece and material catifications that apply to the components within the mock-ups. Also contained are proprietary Weinghase sub vendor information.
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APPENDIX D i
PENETRATION TUBE DIMENSIONAL DATA All of this section is proprietary **#8 This Appendix D, pages D-1 thru D-33, containc detail diametrical measurement data on the penetration mock-up tube samples before and after weld repair. Also contained are proprietary j
Waminghanse sub vendor infonnation i
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