ML20196J259
| ML20196J259 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/30/1997 |
| From: | Legrand R DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-01, GL-97-1, NUDOCS 9708040131 | |
| Download: ML20196J259 (8) | |
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.c Besvsr Vslisy Powsr Station Shippingport. PA 15077-0004 RONALD L LeGRAND (412) 393-7622 Division Vice President.
Fax (412) 393-4905 Nuclear Operations and Plant Manager i
July 30,1997 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 i
Subject:
Beaver Valley Power Station, Unit No. I and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73
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110 Day Response to Generic Letter 97-01 provides the Beaver Valley Power Station Units No. I and 2 response to Items 1 and 2 of the Requested Information section of Generic Letter 97-01,
" Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations."
i If there are any questions concerning this response, please contact Mr. J. Arias, Director, Safety & Licensing at (412) 393-5203.
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Sincerely, 1
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a Ronald L. LeGrand c:
Mr. D. M. Kern, Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr. D. S. Brinkman, Sr. Project Manager r
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9700040131 970730
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- t AFFIDAVIT COMMONWEALTH OF PENNSYLVANM.)
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i COUNTY OF BEAVER-
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Subject:
Beaver Valley Power Station, Unit No.1 and No. 2 l
BV-1 Docket No. 50-334, License No. DPR-66 l
BV-2 Docket No. 50-412, License No. NPF-73 i
120 Day Response to Generic Letter 97-01 i
Before me, the undersigned notary public, in and for the County and Commonwealth aforesaid, this day personally appeared Ronald L. LeGrand, to me j
known, who being duly sworn according to law, deposes and says that he is Division Vice President, Nuclear Operations and Plant Manager of the Nuclear Power Division, Duquesne Light Company, he is duly authorized to execute and file the foregoing submittal on behalf of said Company, and the statements set forth in the submittal are l
true and correct to the best of his knowledge, information and belief.
l Na Ronald L'. LeGrand Subscribed and sworn to before me on this day of
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K/tary Public
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DUQUESNE LIGHT COMPANY Nuclear Power Division Beaver Valley Power Station Units No. I and 2 120 Day Resnonse to Generic Letter 97-01 Introdudient Generic Letter 97-01, Degradation of ControlRod Drive Mechanism No:zle and Other Vessel l
Closure HeadPenetrations, was issued to request licensees to describe their program for l
insuring the timely inspection ofPWR control rod drive mechanism (CRDM) and other closure L
head penetrations. This response provides the Beaver Valley Power Station Units 1 and 2 information relative to the information requested by Generic Letter 97-01 (GL).
Prior to issuance of the GL, Beaver Valley Units 1 and 2 have woiked with the Westinghouse Owners Group, the Electric Power Research Institute (EPRI) and the Nuclear Energy Institute (NEI) to understand the operational experience, identify technical issues, cause factors, relative importance, and solutions to this issue. One of these tasks was the development of safety l
evaluations that characterized the initiation, propagation and consequences of this damage L
mechanism. The safety evaluations are contained in WCAP 13565 Revision 1 and WCAP 14219 i
Revision 1 and are applicable to Beaver Valley Units 1 and 2. The NRC reviewed these safety evaluations and issued a safety evaluation report (SER) to NEI (formerly NUMARC) on November 19,1993. Attachment A is a copy of the SER. The safety evaluations and the SER establish the basis for Beaver Valley Units 1 and 2 continued operation.
f Reauested Information Item 1: '
- 1. Regardinginspection activities:
1.1 A description of all inspections of CRDM nozzle and other VHPs performed to the date of this generic letter, including the results of these inspections.
1.2 If a plan has been developed to periodically inspect the CRDM nozzle and other VHPs:
a.
Provide the schedule for first, and subsequent, inspections of the CRDM nozzle and other VHPs, including the technical basis for this schedule.
b.
Provide the scope for the CRDM nozzle and other VHP inspections, including the l
total number of penetrations (and how many will be inspected), which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
l 1.3 If a plan has not been developed to periodically inspect the CRDM nozzle and other l
VHPs, provide the analysis that supports why no augmented inspection is necessary.
1.4 In light of the degradation of CRDM nozzle and other VHPs described above, provide l
the analysis that supports the selected course of action as listed in either 1.2 or 1.3,
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! (continued) 120 Day Response to Ocneric Letter 97-01 1
Page 2 4
above. In particular provide a description of all relevant data and/or tests used to develop crack initiation and crack growth models, the methods and data used to I'
validate these models, the plant-specific inputs to these models, and how these models substantiate the susceptibility evaluation. Also, if an integrated industry inspection program is being relied on, provide a detailed description of this program.
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Response to Item 1.1:
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Beaver Valley Units 1 and 2 perform several types ofinspections and examinations on Reactor Vessel head components. Liquid penetrant (PT) examinations of the CRDM housing weld is performed in accordance with ASME Section XI, B14.010. The following visual examinations (VT-3) are performed: Closure Head interior surfaces (related to Information j
Notice 90-92 issues), Marion clamp bolting and CRDM nozzles (ASME XI - B4.012).
Additionally, leakage examinations (VT-2) are performed routinely during power ascension from a refueling outage and during Mode 5 following power operation. These examinations identify not only active leakage, but also report evidence ofleakage, such as boric acid crystal deposits.
Other informal examinations are performed as part of regular work tasks.
An example of one of these routine work task inspections was demonstrated during the i
Sixth Refueling Outage at Beaver Valley Unit 2 (2R06), when the refueling crew identified i
evidence ofleakage on the reactor vessel head during the insulation removal process. An investigation of the leakage deposit found the leakage to be from a RVLIS connection that leaked during power ascension from 2R06 and was corrected at that time. The residue observed resulted from water that was trapped under the insulation prior to repair of the leakingjoint. A visual examination was performed of the reactor vessel head surface beneath the insulation and no evidence of degradation to the reactor vessel head was observed.
1 The examinations, resulting from formal eumination programs and those resulting from good work practices, demonstrate that the reactor head area is routinely examined for evidence ofleakage. As noted in Section 2.1.12 of the SER issued on November 19,1993 (Attachment A); 'there would be significant time between initiating a leak and experiencing wastage that would reduce the stmetural integrity margins of the reactor vessel head to below acceptable levels'; Beaver Valley believes that the above noted examinations would preclude the possibility that leakage onto the reactor vessel head would remain undetected for a period of time sufficient to allow significant wastage of the reactor vessel head.
The results from the above examinations and inspections performed to date have resulted in satisfactory resolution of the conditions identified or have met inspection acceptance criteria.
Response to item 1.2:
Beaver Valley Units 1 and 2 are participants in the Westinghouse Owners Group Reactor
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Pressure Vessel (RPV) Head Penetration Integrated Inspection program. The integrated program includes the volumetric inspection of the head penetrations that have been performed at
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- (continued) 120 Day Response to Generic Letter 97-01 Page 3 Point Beach Unit 1, D.C. Cook Unit 2 and North Anna Unit 1. The inspection results from these three examinations found no defects that exceeded the inspection criteria limit for structural integrity. Currently, Beaver Valley believes that the number of plants that have inspected is j
sufficient to demonstrate the adequacy of the Westinghouse Owners Group (WOG) integrated inspection program at this time. The need for additional volumetric inspections and/or re-irupections is being addressed by the orrgoing integrated inspection program. This need will be based upon inspection results from future inspections of WOG member plants and the inspections results from other Owners Group programs. See the response to Item 1.4 for more information
.concerning the WOG program.
t Response to Item 1,3:
See the response to Item 1.4.
Response to Item 1.4:
The data, tests and methods used in developing the crack initiation and crack growth models on which Duquesne Light's management strategy for addressing the Reactor Pressure e
Vessel Head Penetration (RPVHP) cracking issue is provided in Sections 2 and 3 of WCAP 14901, Revision 0 (attached).
Beaver Valley Units 1 and 2 are participants in the Westinghouse Owners Group analysis program in which plant specific probability analysis using the methodology described in Section 4 of WCAP 14901 has been performed. The plant specific input parameters to the analysis are provided in Table 1 for Beaver Valley Unit I and in Table 2 for Beaver Valley Unit 2. The analysis results will be incorporated into the WOG/NEI integrated inspection program for use in determining the need for additional plant specific inspections. This integrated inspection program includes all three PWR owners groups, the Electric Power Research Institute, and the Nuclear Energy Institute who are cooperatively working together. The program is compiling information on the estimated operating time from January 1,1997, needed to initiate and propagate a crack 75% through wall in a vessel penetration for all the PWR reactor vessel heads in the United States. This information will be evaluated by the WOG to determine if an adequate number of plants have or are planning to inspect in the near future. This evaluation will be completed and detailed inspection plans for the industry will be provided to the NRC by WOG/NEI by the end of1997.
Reauested Information Item 2:
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- 2. Provide a description of any resin bead intrusions, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
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2.1 Were the intmsions cation, anion, or mixed bed?
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- s. (continued) 120 Day Response to Generic Letter 97-01 Page 4:
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2.2. What were the durations of these intrusions?
l 2.3. Does the plant's RCS water chemistry Technical Specifications follow the EPRI i
guidelines?
J z 4 Identify any RCS chemistry excursions that exceed the plant administrative limits for the l
followmg species: sulfates, chlorides or fluorides, oxygen, boron and lithium.
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l-2.5 Identify any conductivity excursions which may be indicative of resin intrusions.
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Provide a technical assessment of each excursion and any follow-up actions.
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l-2.6 Provide an assessment of the potenti.>l for any of these intrusions to remit in a significant increase in the probability far IGA of VHPs and any associawd plan for l
l inspections.
Fesponse tt gjal:
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Duquesne Light Company has reviewed the plant historical records to determine if any incident of resin ingress similar to those which occurred in 1980 and 1981 at the Jose Cabrera
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(Zoritia) plant has occurred at either Beant Valley Power Station Unit 1 or Unit 2. A review of l
the records has indicated that no indicatioJ. ~a resin ingress that exceeded the screening eriteria I
desciibed below occurred at either Beaver Yahey Unit 1 or Unit 2.
l The data search was structured to identify all resin intrusion events into the primary coolant system with a magnitude greater than 1 ft.' (30 liters). The threshold of I ft.' was chosen l
as a conservative lower bound since it upresents less than 15% of the estimated volume of resin released into the reactor coolant system during the two events at Jose Cabrera.
1 For the period of plact operation prior to routine analysis for sulfate in the reactor l.
coolant, the data search was based on a review of the plant's reactor coolant cherristry records l
relative to specific conductance of the reactor coolant. An elevation of 28 mirco sicm increment I
in specific conductance was the value used as an indicator of cation resin ingress equivalent to a l
volume of I ft'.
1 Routine analysis for sulfate in the reactor coolant has been performed for Beaver Valley Unit I sc: October 9,1991 and since January 10,1992 at Beaver Valley Unit 2. A sulfate J
concentration in the range of 15 to 17 ppm peak concentration was used as the indicator of cation resin ingress. This concentration is approximately equivalent to a volume of I ft'.
l Had either specific conductance or sulfate ir reases indicated resin ingress to the i
magnitude of the threshold quantity identified above, additional data evaluation would have been conducted to ascertain a corresponding depression in pH or elevation in lithium as cwoborating
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information of the incident. In the case of the use of sulfate data as the indicator, spec ific conductance would also have been included as confirmatory data had a significant ic-leakage -
event been identified.
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... (continued) 120 Day Response to Generic Letter 97-01 Page 5 It is unnecessary to review plant records for boron, chlorides, fluorides and oxygen because these species are not valid indicatom of cation resin ingress and degradation within the
. primary coolant system of a PWR. Borate, chloride and fluoride anions could be associated with the anion portion of mixed bed resin (cation plus anion); however, if mixed bed resin leakage to the RCS occurred, the cation portion of the resin would contain the sulfate indicator described above. Detectable dissolved oxygen in the reactor coolant, during power operations with copropriate hydrogen overpressure on the volume control tank and specified residual dissolved 1:fdrogen in the reactor coolant, could not occur and, therefore, could not be associated'with resin in-leakage.
Beaver Valley Units 1 and 2 have utilized the EPRI PWR Primary Water Chemistry Guidelines since their inception and have implemented the changes recommended in subsequent revisions to the guidelines.
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120 Day Response to Generic Letter 97-01 Page 6 Table 1 Beaver Valley Unit 1 Input Values for Probabilistic Analysis Case Penetration #
Temperature Set up Angle Y.S. (ksi)
GBC (%)
1 58 thm 65 607'F 42.6 35.9 52.9 2
54 thru 57
'607'F 40.0 35.9 52.9 3
50 thru $3 607'F 38.6 48.5 23.0 l
4 46 thru 49 607 F 38.6 35.9 52.9 5
42 thru 45 607 F 37.3 35.9 52.9
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6 34 thru 41 607'F 33 1 35.9 52.9 7
31 thru 33 607'F 28.6 43.2 65.3 8
26 thru 30 607 F ie.6 35.9 52.9 9
24,25 607*F 27.0 43.2 65.3 10 22,23 607 F 27.0 35.9 52.9 11 18 thm 21 607 F 25.4 35.9 52.9 12 14 thm 17 607 F 19.8 35.9 52.9 13 10 thru 13 607'F 17.6 35.9 52.9 14 6 thru 9 607'F 12.4 35.9 52.9 1
Table 2 i
Beaver Unit 2 Input Values for Probabilistic Analysis Case Penetration #
Temperature Set up Angle Y.S. (ksi)
GBC (%)
1 58 thm 65 606'F 42.7 40.0 37.1 2
54,57 606 F 40.0 40.0 37.1 3
55,56 606'F 40.0 41.0 40.8 4
46 thm 53 606*F 38.7 40.0 37.1 5
42 thru 45 606'F 37.3 40.0 37.1 6
40,41 606'F 33.1 39.0 36.5 7
39 606 F 33.1 40.0 37.1 8
34 thru 38 606'F 33.1 36.0 45.5 l
9 26 thru 33 606'F 28.6 39.0 36.5 l-10 22 thru 25 606 F 27.0 40.0 37.1 l
11 21 606 F 25.4 40.0 37.I 12 18 thm 20 606 F 25.4 36.0 45.5 13 15 thru 17 606 F 19.8 36.0 45.5 14 14 606 F 19.8 40.0 37.1 15 10,11,13 606'F 17.6 36.0 45.5 16 12 606'F 17.6 39.0 36.5 17 5 thru 9 606'F 12.4 39.0 36.5
g33-UNITED STATED i
3 NUCLEAR RESULATCRY COMMISSION i
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wasmunon, o.c. sees un i
November 19, 1993 l
1 William Rasin, Vice President
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Director of he Technical Division i
Nuclear Management and Resources Council
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1776 Eye Street, N.W.
j Suite 300 j
Washington, D.C. 20006-3706 4
Dear Mr. Rasin:
The attached safety evaluation was prepared by the Materials and Chemical Engineering Branch, Division of Engineering, Office cf Nuclear Reactor d
Rowlttion, on the NUMARC submittal of June 16, Igg 3, addressing the Alloy 600 Control Rod Drive Mechanism (CRDM)/ Control Element Drive Mechanism (CEDM) press's ired water reactor vessei head penetration cracting issue.
This
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i setMttel addressed stress analyses, crack growth analyses, leakage i
nsuments, and wastage assessments for potential cr.ncking of the inside diameter of CRDM/CEDM nozzles. Bassd on the overseas inspection findin the review of your analyses, the staff has concluded that there is no gs and
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immediate safety concern fer cracking of the CROM/CEDM penetrations. This finding is predicated on the performance cf the visual inspection activities requested in Generic Letter 88-05. Also, special nondestructive examinations a
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are scheduled to cosamence in the Spring of 1994 to confim your safety analyses for each PWR owners group.
l Your submittals for each PWR type did not add:sss the kgey-3 flaw that was i
oriented approximately 30* cff the vertical exis nor a circumferential, J-
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groove flaw discovered at Ringhals.
Preliminary information supplied to the j
staff by Swedish authorities indicates that the J gro.we flaw may be asrociated with a fabrication defect. We are ce!i.ineing to work with the i
Swedish authoritiss to contim this.
From the iMomation available to us j
tcday, neither of these flaws would pose a threat to the integrity of the CRDM j
It is our understanding that you are, also reviewing these flaws and you will provide your assessment as to their significance and origin.
NRC will issue a supplemental safety evaluation after reviewing your supplemental l
assessment.
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The staff agrees that there are no unreviewed rafety questions associated with CRDM/CCDM penetration cracking. The staff agnaes that the flaw predictions based upon penetration stress analyses are in qualitative agreement with inspection findings. However, the stress analyscs so not address stresses These stresses, if large, g of CROM penetration t!bes during fabrication.
from possible straightenin could result in circumferential flaw orientations.
i The staff requests that you also address this issue !n your supplemental assessment.
Based upon information received from on:rseas regulatory I
authorities, your analyses, and staff reviews, the staff believes that catastrophic failure of a penetration is extremely unlikely. Rather, a flaw would leak before it reached the critical flaw size and would be detected I
during periodic surveillance walkdowns for boric acid leakage pursuant to j
Generic Letter 88-05. However, the staff recosamends that you consider 4
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!)WW 69(
William Rasin enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area.
The staff requests that you also address _the issue of enhanced leakage detection in your supplemental assessment.
The NRC staff has reviewed your July 30, 1993 submittal, which proposed flaw acceptance criteria to be used in dispositioning any flaws found during CRDM/CEDM inspections. The staff finds the proposed flaw acceptance criteria acceptable for axial cracks because the criteria conform to the American Society of Mechanical Engineers (ASME)Section XI criteria. The staff determined that flaws that are primarily axial (less tiian 45' from the axial l
direction) should be treated as axial cracks as indicated in Figure 1(b), (d),
and (f) of your July 30, 1993 letter.
Flaws more than 45' from the axial 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 flaws. Circumferential flaws which a licensee proposes to j
leave in service without repair, should be reviewed by the staff on a case-by-I case basis.
Sincerely, MMD William T. Russell, Associate Director for Inspection & Technical Assessment l
Office of Nuclear Reactor Regulation As Stated Distribution:
Central File JStrosnider WRussell EMCB RF BDLiaw RHermann JDavis JWiggins WKoo i
- SEE PREVIOUS CONCURRENCE
- EMCB:DE
- EMCB:DE
- EMCB:DE
- EMCB:DE
- DE:D NRR.;ADJ v. - u-JDavis WKoo RHermann JStrosnider JWiggins WRussell 09/23/93 10/25/93 10/27/93 11/18/93 11/19/93 0 / li/93 0FFICIAL RECORD COPY G:\\ DAVIS \\WOGSER.JAD (s:\\ DAVIS) 4 1
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SAFETY EVALUATION f.QE POTENTIAL REACTOR VESSEL HEAD ADAPTOR TUBE CRACKING s
1.0 INTRODUETION Primary water stress corresion cracking (PWSCC) of Alloy 600 was i
identified as an emerging issue by the NRC staff to the NRC Commission following a 1989 leakage from an Alloy 600 pressurizer heater sleeve j
penetration at Calvert Cliffs Unit 2, a Combustion Engineering designed pressurized water reactor (PWR). Severel instances of PWSCC of Alloy 3
j 600 pressurizer instrument nozzles had been reported to the NRC between
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the time period of 1986 to the present on domestic and foreign pressurized water reactors (PWR). The licensee at Arkansas Nuclear Operations, Unit 1, a Babcock & Wilcox (B&W) designed PWR, reported a leaking pressurizer instrument nozzle in 1990, after 16 years of i
operation.
Westinghouse PWR's do not use Alloy 600 for penetrations or i
nozzles in the pressurizers.
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According to the information provided to the staff by NUMARC at a public l
meeting held on July 5,1993, a leak was discovered in an Alloy 600 control rod drive mechanism (CRDM) adaptor tube penetration during a j
hydrostatic test at the Bugey 3 plant in France in 1991 after 12 years of operation. A visual examination of the CRDM adaptor tube penetration 3
indicated the presence of axial flaws in the inside diameter (ID) of the CRDM adaptor tube penetration. The remaining 65 CRDM adaptor tube l
penetrations were meined at Bugey 3 and 2 additional CRDM adaptor tube penetrations contained axial cracks on the ID of the CRDM adaptor tube j
penetrations. An examination of 24 CRDM adaptor tube penetrations at Bugey 4 revealed axial ID cracks in 8 CRDM adaptor tune penetrations.
l CRDM adaptor tube penetrations have been examined at 37 nuc, lear power l
plants in France, Sweden, Switzerland, Japan, and Belgium and 59 of the l
1,850 penetraticas have revealed short, axial crack indications.
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The primary safety cor.cern associated with stress corrosion cracking in i
Alloy 600 in CRDM penetrations is the potential for circumferential cracks.
Extensive circumferential cracking could lead to the ejection of a CRDM resulting in an unisolable rupture in the primary ecolant system. As indicated above, the inspections to date have identified j
short axial cracks. However, two other inspection findings are of particular interest.
First, the CRDM penetration that leaked during
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hydrostatic testing'at Bugey-3 was removed and examined metallurgically during December 1992. A secondary crack that was 0.120 inches long and 0.090 inches deep at about 30 degrees to the axial direction was observed on this CRDM.
Second, in early in 1993, a J-groove weld at the
- 1 Ringhals plant in Sweden was discovered to contain a circumferential i
crack.
Preliminary indications are that this flaw is a fabrication defect. Additional ' work is in progress by the staff at the Swedish j
Nuclear Power Inspectorate to confirm this.
The Westinghouse CRDM adaptor tube penetrations are similar in design to the European PWR's and use Alloy 600 for the penetrations. The NRC j
staff met with the WOG on January 7, 1992 to discuss the experience at 4
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2 the Bugey 3 plant and the relationship of the French design of the CRDM adaptor tube penetrations to the design'of domestic Westinghouse plants.
The WOG infomed the NRC staff tMt a program had been initiated in Decembar_1991 to: (1) determine tne root cause of the CRDM penetration cracking; (2) analyze the stress distributions in the CRDM penetrations of a typical domestic plant; (3) compare the design and operational characteristics of domestic and French plants to determine the likelihood for cracking; and (4) identify the need for additional efforts. The NRC staff also met with the Combustion Engineering Owners Group (CEOG) and the Babcock & Wilcox Owners Group (B&WOG) to discuss the PWSCC of CRDM adaptor tube penetrations. The Nuclear Management and Resources Council (NUMARC) coordinated the PWR Owners' Group efforts on this subject.
On June 16, 1993, NUMARC submitted safety assessments to the NRC from WOG, CEOG, and B&WOG for review by the NRC staff.
These safety assessments present stress analyses, crack growth analyses, leakage analyses, and wastage assessmerita for flaws initiating on the ID of CRDM adaptor tube penetrations. NRC requested additional information on the safety assessments by letter dated September 2, 1993.
NUMARC submitted the response to NRC on September 22, 1993. The safety usassments submitted to the NRC did not address the secondary flav observed at the Bugey-3 plant that was oriented approximately 30' from the longitudinal axis of the penetration nor the apparent fabrication flaw at the Ringhals plant.
Neither of these flaws posed a threat to the integrity of the CRDM penetrations. However, NUMARC has committed to submit a safety assessment relevant to this type of cracking. After this safety assessment has been reviewed by NRC, a supplement to this SER will be issued.
2.0 STAFF EVALUATION 2.1 WOG WCAP-13565. ALLOY 600 REACTOR VESSEL HEAD ADAPTOR TUBE CRACKING SAFETY EVALUATigg The WOG subuitted the, " Alloy 600 Reactor Vessel Head Adaptor Tube Safety EvalJation," through NUMARC on June 16, 1993.
The safety evaluation addresses the following elements:
1.
A susseary of. the stress analysis focusing on the type (orientation) of cracking that may be expected in the Alloy 600 material, and the stresses necessary for flaw propagation; 2.
A summary of the flaw propagation analysis along with the background of the flaw prediction method; 3.
An assass:rient of the WOG plants with respect to penetration flaw indication data from plant inspections at Ringhals, Beznau, and various Electricite de France plants, in which the key parameters for cracking are compared to WOG plants; h
3 4.
A leakage assessment summarizing leak rate.vs. flaw size, and postulating leaks for WOG plants for which leakage considerations l
l may apply; and, 5.
A N ssel head wastage assessment including the process that leads to wastage and an estimate of the allowable wastage.
l 2.1.1 REGULATORY BASIS AND DETERMINATION OF UNREVIEWED SAFETY OUESTIONS The WOG prepared safety evaluation addresses the potential for cracking and the ramifications of such cracking of the reactor vessel head adaptor tubes at Westinghouse designed NSSS plants. The WOG compared the results of this safety evaluation to the criteria in the Title 10, Code of Federal Regulations, Section 50.59 (10 CFR 50.59).
The WOG l
concluded that an unreviewed safety question did not exist.
Its l
evaluation considered the following:
i 1.
Continued plant operation will not increase the probability of an accident previously evaluated in the FSAR.
2.
The consequences of an accident previously evaluated in the FSAR are not increased due to continued plant operation.
3.
Continued plant operation will not create the possibility of an i
accident which is different than any already evaluated in the FSAR.
4.
Continued plant operation will not increase the probability of a malfunction of equipment important to safety.
1 5.
Continued plant operation will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR.
6.
Continued plant operation wil'1 not create the possibility of a malfunction of equionent important to safety different-than any already evaluated in the FSAR.
7.
The evaluation for the effects of continued plant operation with potentially cracked reactor vessel head adapters has taken into account the applicable technical specifications.
2.1.2 STAFF'S EVALUATION OF THE REG @ ATORY BASIS AND DETERMINATION OF UNREVIEWED SAFETY OUESTIONS The staff agrees that no unreviewed safety question exists, provided only axial flaws are found.
Those axial flaws would be expected to be short,- and they would most probably leak noticeably prior to the flaw size reaching unstable. dimensions.
The existence of any unexpected leaks would not-adversely affect plant operation, or accident / transient response.
No significant equipment degradation would be expected.
Details of the staff's evaluation that led to the above conclusions is discussed in the following sections.
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2.1.3 PENETRATION STRESS ANALYSIS 1
j The W0G conducted an elastic-plas"c. finite element analysis of a 4-loop IEG plant vessel head penetrations.
The WOG concluded that the 4-
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loop WOG plant is bounding since prior analyses showed that the j
operating and residual stresses are higher on a 4-loop plant than on 2 or 3-loop plants on the outermost penetrations. Three penetration f
i locations were modeled, the center location, the outermost location, and l
the location next to the outermost location..The stress history was simulated by using a load sequence of the themal load from the first welding pass, the thermal load from the second weld pass, the 1
fabrication shop cold hydrotest, the field cold hydrotest, and the j
steady state operational loading.
j-The highest stresses are found in the zone around the weld and are the highest in the penetration. farthest from the center of the vessel j
(peripheral penetrations). The highest stresses on that penetration are l
on the side of the penetration nearest to the center of the vessel i
(centerside) and on the side of the penetration farthest from the center of the vessel (hillside). Also, the stresses are the highest below the j
weld and decrease significantly above the weld. The ratio of peak hoop stress to axial stress at the same location at the outermost penetrations was about 1.4 compared to a value of about 1.6 estimated based on tha degree of ovaling measured on actual penetrations. The ratio of hoop stress to axial stress was about the same for center penetrations as for peripheral penetrations (1.6 for center penetrations compared to 1.4 for peripheral penetrations); however, the magnitude of j
the stresses at the peripheral penetrations was higher. The analysis i
indicates that axial flaws would be more likely than circumferential l
flaws, flaws are more likely below the weld than above the weld, and
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that axial flaws would appear at locations in the penetrations where they have been found in service.
2.1.4 STAFF EVALUATION OF THE PENETRATION STRESS ANALYSIS
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l The staff is in agreement with the results of the WOG stress analysis that predicts that the cracking will be predominately axial. These results are in qualitative agreement with field inspection findings.
However, the WOG did not address the effects of possible straightening of the CRDM penetration tubes during fabrication.
Such straightening operations could significantly alter the residual stress fields within I
i l
the penetration tubes. Results of inspections to date have not i
identified any problems directly related to this process; however, the staff requests that NUMARC address this issue for all three owners j
groups' plants.
2.1.5 CRACK GROWTH ANALYSIS: FLAW TOLERg[1 The WOG crack growth analysis was based on the assumptions that the flaw
}
would be caused by primary water stress corrosion cracking, and that the i
crack growth is controlled by the hoop stress. The maximum principal stress will be oriented at a slight angle to the hoop stress and flaws k
b i
would be expected to be perpendicular to the maximum principal stress.
However, all of the flaws found in service with two exceptions have been axially located. Hence, the WOG used the hoop stress as an approxiglation of the maximum principal stress. The outer-most penetration for a 4-loop Westinghouse plant wn selected for analysis i
since this location experiences the highest stresses. The highest stress was located along the inner surface just below the center side of the weld.
The calculated hoop stress through the wall of the penetration was used for flaw growth calculations.
The flaw growth data were obtained from steam generator field experience and laboratory data.
1 Based on the stress fields that exist in the CR31 penetrations, any flaw growth that occurs is expected to be predominately axial in nature.
Furthermore, the growth of any flaws inclined from the vertical would be limited in length due to the nature of the existing stresses. These i
conclusions are consistent with the inspection results described above.
Accordingly, there is no significant potential for failure of a peretration by ejection of the CRDM sleeve. With regard to axial crackir.g, WOG has concluded that the critical flaw length for an axial i
flaw for Allov 600 is sufficiently long that leakage would occur and be detected during surveillance walkdowns as required by GL 88-05.
Therefore, the consequences of cracking in the penetration sleeve are limited to the affects of leakage at discussed below.
The flaw growth analysis showed that under the most severe conditions of metallurgical microstructure, peak hoop stress, and operating temperature, it would take about five years for a flaw to grow through wall. Under the same conditions, it would take an additional 10 years for a through-wall flaw to grow 1 % inches above the weld on the lower hillside of the outomost head penetrations (Figure 3.2-2) and about the same time to grow two inches above the J-groove weld on the center side of the outermost penetrations (Figure 3.2-3).
The flaw growth analysis indicates that through wall flaws would essentially arrest.before growing a maximum of two inches above the weld. These flaws would be constrained within the hand and could not significantly open thus limiting the amount of Teakage that could occur.
2.1.6 STAFF EVALUATION OF THE CRACK GROWTH ANALYSIS.
The W0G stated that the crack growth analysis is in general agreement with the inspection findings. The crack growth rate data used in this analysis was limited, but the results predicted using these flaw growth data bound the results of the inspections.
Crack growth rates are difficult to determine precisely; however, the assumed growth rates compare well with inspection data available to date and the large margins that exisi, in the analyses will account for any possibly higher growth rates.
There are large margins of safety in the analyses and the CRDM penetrations are constructed of inherently tough material with a critical flaw size of approximately 13 inches in the free span above the reactor vessel shell.
Therefore, the staff concludes that catastrophic failure of a penetration is extremely unlikely because a flaw would be i
4
~
l i.
I 6
j-r detected during boric acid leakage surveillance walkdowns before it i
reached the critical flaw size.
l 2.
1.7 ASSESSMENT
OF WOG PLANTS
)
The WOG. compared the Ringhals and Beznau plants to the domestic 1
Westinghouse plants and developed a model for the relative l
susceptibility to PWSCC.
The WOG considered residual and operating i
stresses in the penetrations, the environment, material condition, j
operating temperature, and time-of-operation at temperature, and pressure.
Based on this evaluation, the WOG has evaluated domestic WOG j
PWR's with regard to their degree of susceptibility.
Based on what WOG considers to be conservative assumptions, the Ringhals plants envelope f
45 domestic plants. None of these plants are expected to have any flaws i
other than some short, shallow, axial flaws. Nine additional WOG plants 1
are not enveloped by the Ringhals plants.
Based on the stresses, operating temperattres, hours of operatirn, and the flaw growth curves i
provided in the WOG safety assessment, the WOG does not expect any CRDM i
penetration axial flaws to reach a length in excess of 1 inch before j
abont the middle of 1995.
)
2.1.8 STAFF EVALUATION OF THE WOG ASSESSMENT i
The susceptibility model developed by the W0G considers the appropriate l
parameters affecting IGSCC and should provide a reasonable ranki.'g of i
plant susceptibilities.
In addition, this evaluation indicates that it is unlikely that U.S. plants should exhibit any cracking significantly worse than that found in European plants.
l 2.1.9 LEAK RATE CALCULATIONS i
j The leak rates were calculated based on the assumption that the leak rate will be controlled by the flow rate through the flaw in the head i
penetration or by the flow through the penetration annulus, whichever is 4
smaller. WOG estimates the maximum leak rate would be 0.7 gpm for a 2 l
inch long flaw and an annular clearance of 0.003 inches.
Leakage above i
1.0 gpa is detectable in domestic WOG plants according to WOG. Growth of an axial flaw outside of the part contained within the reactor head will result in leakage greater than 1.0 gpm prior to reaching the 3
i critical flaw size. The WOG stated that an axial flaw would remain j
stable for growth up to 13 inches above the reactor vessel head.
J 2.1.10 STAFFS EVALUATION OF THE WOG LEAK RATE CALCULATIONS l
The staff agrees with the WOG assumptions about leakage and concludes, l
i that based on existing leakage monitoring requirements, there is reasonable assurance that leakage in excess of the 1.0 gpm technical specification limit would be detected prior to any unstable extension of the flaw.
i 4
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7 2.1.11 REACTOR VESSEL HEAD WASTAGE ASSESSMENT 1 This section assesses the potenti:
wastage of the reactor vessel head due to deakage of primary coolant through the CRDM penetrations. This assessment is based on wastage data from previous Westinghouse experiments and from the results of a penetration mockup test conducted by the Combustion Engineering Owners Group (CEOG).
This analysis assumed that coolant escaping from the penetration would flash to steam leaving boric acid crystals behind. WOG assumed that crystals would accumulate on the vessel head but would cause minimal corrosion while the reactor was operating. The head temperature would be about 500*F during operation and significant wastage of the reactor head by the boric acid crystals would not be expected. Dry boric acid crystals do not cause corrosion. Wastage would only occur during outages when the head temperature is below 212*F.
The CEOG provided all of the PkR owners groups with the results of pressurizer penetration mockup test results. The WOG examination of the CE0G mockup test results showed that the maximum penetration rate at the deepest pit was 2.15 inches / year while the average penetration rate was 0.0835inchgs/ year. The maximum total metal loss rate or wastage volume was 1.07 in / year, and the greatest damage occurred where the leakage lefttheannulus. The WOG considered the maximum wastage would be 6.4 in of vessel head material. The assumptions made were that any leakage over 1.0 gpm can be detected so only leak rates between 0.0 and 1.0 gpm were considered.
The WOG analyzed the situation using finite element analyses for a 2 loop 3 loop, and 4 loop reactor vessel head where a 1.0 gpa leak went undetected for 6 years and concluded that the ASME code minimum wall thickness requirement would be satisfied and that the stresses remain within the ASME code allowable stresses.
2.1.12 THE STAFF'S EVALUATION OF THE REACTOR VESSEL HEAD WASTAGE ASSESSMENTS The assumption used in the W0G corrosion assessment are based on experimental data and should provide a reasonable estimate of potential wastage of the reactor vessel head.
Based on these evaluations, there would be significant time between initiating a leak and experiencing wastage that would reduce the structural integrity margins of the reactor vessel head to below acceptable levels. Considering the length of time involved, there is reasonable assurance that leakage, manifested by the accumulation of moderate amounts of boric acid crystals would be detected during a surveillance walkdown in accordance with GL 88-05.
3.0 CEOG SAFETY EVALUATION The CEOG safety evaluation is essentially the same as the WOG safety evaluation.
The CEOG plants run at a slightly higher temperature than the European plants that have experienced cracking, have greater hillside angles, and have been in operation longer than many of the European plants.
The CE0G indicated that all of these factors would
i 8
increase the probability of cracking for the CEOG plants. However, the CEOG plants have significantly less weld metal in the J-groove welds and the CEOG stated that this would significantly re6tce the residual welding-induced stresses and would reduce the probability of PWSCC.
CEOG concluded that any PWSCC that formed would be short, axial flaws.
The CEOG states that they can detect a 0.12 gpm leak in the primary 1
coolant system.
CEOG also states that the boric acid accumulation as a result of a 0.12'gpm leak would not result in wall thinning below the code allowables in less than 8.8 years compared to 6 years for WOG plants and that surveillance walkdowns would detect boric acid crystals long before the 8.8 years.
3.1 STAFF EVALUATION OF THE CEOG SAFETY EVALUATION The staff has concluded that the potential for PWSCC of CRDM/CEDM for i
CEOG plants does not create an immediate safety issue as long as the surveillance walkdowns required by GL 88-05 continue and corrective action is instituted when leaks are discovered.
- The CEOG analyses indicating that the stresses would favor development of axial rather than circumferential cracks and that significant time would be required to reduce the wall thickness of the vessel head to below the ASME code i
allowables demonstrates that an immediate safety concern does not exist.
l 4.0 B&WOG SAFETY EVALUATION The B&WOG safety evaluation was essentially the same as the WOG and CEOG safety evaluations.
The B&WOG analysis indicates that B&WOG plants have essentially the same susceptibility to PWSCC as the European plants based on operating temperature, residual stresses, and operational life.
The B&WOG predicts short, axial flaws on the peripheral locations based on the results of finite element analyses. The B&EOG estimates that it would take 10 years from the time a flaw initiates on the inside diameter of a CRDM penetration until a leak appears. Once a leak l
starts, B&WOG concluded that it would take 6 years before enough corrosion would occur to reduce the wall thickness of the reactor vessel head to below ASME code minimums, and that this amount of leakage wculd l
be detected during surveillance walkdowns.
l-4.1 STAFF EVALUATION OF THE B&WOG SAFETY EVALUATION The staff has concluded that the potential for PWSCC of CRDM for B&WOG plants does not create an inmediate safety issue as long as the l
surveillance walkdowns required continue and as long as any leakage is corrected. The B&WOG analyses, indicating that the stresses would favor development of axial rather than circumferential cracks and that significant time would be required to reduce the wall thickness of the
~
vessel head to below the ASME code allowables, demonstrates that an imediate safety concern does not exist.
l i
i
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9 5.0 PROPOSED FLAW ACCEPTANCE CRITERIA on July 30, 1993, NUMARC submitted the proposed flaw acceptance criteria j
for fims identified during inservice inspection of reactor vessel upper head penetrations to the NRC for review. These criteria were developed by utility technical staffs and the domestic PWR vendors. NUMARC proposes that axial flaws are permitted through-wall below the J-groove weld and 75 percent through-wall above the weld.
There is no limit on the length of the flaws. NUMARC proposes that circumferential flaws i
through-wall and 75 percent around the penetration be allowed below the i
J-groove weld and that circumferential flaws above the weld could be 75 i
percent through-wall and 50 percent around the penetration.
Proximity j
rules found in ASME Section XI, Figure IWA 3400-1 are proposed for determining the effective length of multiple flaws in one location.
NUMARC proposes that the flaws be characterized by length and preferably depth. NUMARC proposes that if only the length is characterized, that-the depth be assumed to be one half of the length based on inspection findings to date.
5.1 STAFF EVALUATION OF THE PROPOSED FLAW ACCEPTANCE CRITERIA l
The staff finds the proposed flaw acceptance criteria acceptable for axial flaws because the criteria conform to the American Society of Mechanical Engineers (ASME)Section XI criteria.
The assumption that i
flaw depth is one half the flaw length for flaws whose depth cannot be detennined will limit the flaw length to 1.5 times the thickness of the penetration sleeve. However, it is expected that reasonable attempts will be made to determine flaw depths.
Flaws found through inservice inspection (ISI) that are primarily axial (less than 45' from the axial j
direction) will be treated;as axial flaws as indicated in Figure Ifb),
1 (d), and (f) of NUMARC'$ July 30, 1993 letter.
Flaws more than 45 from j
the axial direction are considered to be circumferential flaws.
Based upon information submitted to date and the more serious safety l
consequences of circumferential flaws, the staff has concluded that
)
criteria for circumferential flaws should not be pre-approved.
Detection of such flaws would be contrary to inspection results to date i
and to the conclusion of the Owners Groups evaluations. The curcumstances associated with such a flaw would have to be well understood. Therefore, any circumferential flaws found through ISI, which a licensee proposes to leave in service without repair, will be i
reviewed on a case-by-case basis by the staff.
)
6.0 LEAKAGE MONITORING l
NUMARC, through the owners groups' reports, determined that any leakage in excess of 1 gpa would be detected prior to any unstable extension of 1
axial flaws. Also, leakage at less than 1 gpa would be detectable over j'
time based on boric acid buildup as noted during periodic surveillance 1
wal kdowns.
Although NUMARC has proposed, and the staff agrees, that low i
level leakage will not cause a significant safety issue to ren:lt, the i
staff determined that NUMARC should consider methods for detecting i
smaller leaks to provide defense-in-depth to account for any pnential l
E i
l
1 10 uncertainty in its analyses. The reported leak rate at Bugey 3 was about 0.003 gpm and was detected using acoustic monitoring techniques during the performtnce of a hydrostatic test. The staff does not think that 11_is necessary to detect a 0.003 gpm leak, but does think that permitting leakage just below 1.0 gpm as currently proposed may be undesirable.
Leakage of this magnitude would produce significant deposits (thousands of pounds / year) of boric acid on the reactor vessel head. Further, most facilities' technical specifications state that no pressure boundary leakage is permitted.
The staff notes that small leaks resulting from flaws which progressed through-wall just prior to a refueling outage would be difficult to detect while the thermal insulation is installed. Although running for an additional cycle with that undetected leak would not result in a significant safety issue, the NUMARC should consider proposing a method for detecting leaks that are significantly less than 1.0 gpm, such as the installation of on-line monitoring equipment.
70 CONCLUSIONS Based on review of the NUMARC submittal and the overseas inspection results, the staff concludes that the CRDM/CEDM cracking at the reactor vessel heads is not a significant safety issue at this time as long as the surveillance walkdowns in accordance with GL 88-05 continue. The staff agrees with the NUMARC's determination that there are no unreviewed safety questions associated with stress corrosion cracking of CRDM penetrations. However, new information and events may require a reassessment of the safety significance.
Furthermore, there is a need to verify the conclusions of the NUMARC's safety evaluations.
Therefore, nondestructive examinations should be performed to ensure there is no unexpected cracking in domestic PWRs.
These examinations do 4
not have to be conducted immediately since only short, shallow, axial 2
flaws are likely to be present in the CRDM penetrations.
The industry has committed to conduct inspections at three units in 1994. They are:
(a) Point Beach I! nit 1 in the Spring of 1994, 1
(b) D.C. Cook Unit 2 in the third quarter of 1994, (c) Oconec Unit 2 in September 1994.
1 As the surveillance walkdowns proposed by NUMARC are not intended for detecting small leaks, it is conceivable that some affected PWRs could notentially operate with small undetected leakage at CRDM/CEDM penetrations.
In this regard, the staff believes it is prudent for.
l NUMARC to consider the implementation of an enhanced leakage detection method for detecting small leaks during plant operation.
The staff found NUMARC's flaw acceptance criteria acceptable for axial flaws but NRC review and approval of the disposition of any circumferential flaws will be required.
4 Technical Contacts:
Robert A. Hermann (301) 504-2768 William H. Koo (301) 504-2706 James Davis (301) 504-2713
l WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14901 Background and Methodology for Evaluation of Reactor Vessel Closure l
Head Penetration Integrity for the Westinghouse Owners Group i
1 W. H. Bamford i
B. A. Bishop l
l J. F. Duran l
D. E. Boyle July 1997 Reviewed by:
~1 A.
h G.V.Ra6 lineering and Materials Technology Approved:
D. A. Howell, Manager Mechanical Systems integration Westinghouse Electric Corporation Nuclear Services Division i
P.O. Box 355 Pittsburgh, PA 15230 C1997 Westinghouse Electric Corporation All Rights Reserved s
I July 1997 Rev.O c:\\1710. doc:1b:07/14/97 l
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TABLE OF CONTENTS l
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E x e e u tiv e S u m m a ry........................................................................................................ ii l
1.0 i n t r od u c tio n...................................................................................................... 1 1 l
2.0 Development of a Crack Crowth Rate Model for Alloy 600 Head Penetrations........ 21 l
3.0 Westinghouse Crack initiation Model Development and Crack initiation Testing.... 31 l
4.0 Technical Description of Probabilistic Model.......................................................... 4 1 1
5.0 References....................................................................................................51 j
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4 EXECUTIVE
SUMMARY
j This report is intended for use in response to NRC Generic Letter 97 01, Cracking in Alloy 600 reactor vessel head penetrations is a relatively new issue to the nuclear industry. The issue
)
was first brought to the world's attention in 1991 when, after 10 years of operation, a leak was j-detected during a hydrotest of the reactor coolant system at the Bugey Unit 3 power plant in J
Franco. Since then a significant number of studies and research programs have been funded by the industry to determine the causes of the problem and develop strategies for repair and management.
4 Through these programs and subsequent studies it was concluded that reactor pressure vessel j
head CRDM penetration cracking at Bugey Unit 3 is induced by is a thermally activated stress corrosion mechanism operative in primary water environments, more commonly known as primary water stress corrosion cracking (PWSCC). Based on conservative evaluation results, the NRC and industry concluded that PWSCC cracks were most likely to initiate from the inside surf ace of the penetrations, in the axial orientation, and would take at least six years to propagate through the wall under the typical plant operating conditions. Fracture mechanics evaluations have determined that the crack is non critical until its axiallength reaches 8.5 inches to 20 inches, depending on plant design. Therefore this issue is an economic one, and does not constitute a serious challenge to plant safety.
Extemal circumferential cracking is less probable. It may occur only in the presence of an above the weld through wall crack, with active leakage. Assuming coolant is present on the outer diameter of the penetration, one conservative analysis estimated that it would take more than 90 years before penetration failure would occur. In the presence of reactor coolant, corrosion wastage of the alloy steel RV head is possible. Conservative evaluations estimate that it would take longer than six years after a through wall crack occurs before the code structuralintegrity margin for the RV head would be impacted by corrosion. It was concluded that periodic visual inspection of the RV head in accordance with Generic Letter 88-05 is adequate to maintain plant safety, and sufficient to detect leakage prior to significant penetration cracking and vessel head corrosion.
Worldwide, approximately 5,200 Alloy 600 RV head penetrations have been inspected since the first cracking was observed in 1991. Approximately 2 percent of these penetrations are reported to be cracked. Most of the cracks were observed in French RV head penetrations, if the French inspection records are removed from the inspected populat;on, the percentage of head penetrations with indications is only about 0.5 percent. Only one plant worldwide has experienced PWSCC head penetration through walileakage, and this was from a single penetration.
Specialized NDE methods have been developed and verified using mock ups to ensure accurate inspectxms. Flaws were introduced into the mock up penetrations by artificial means.
The ability of these NDE methods to detect and size the potential PWSCC indications in the vessel head penetrations was demonstrated. Flaw acceptance criteria were establishad by the industry, and approved by the NRC staff.
The Westinghouse Owners Group has developed methods to ovaluate the PWSCC susceptibility and the probability of a penetration initiating a crack, or a leak, as a function of ii July 1997 Rev.O oM710. doc:10:07/14/97
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plint operation timo. This informition has bein us d to evtlutta fhe nesd for inspsetion of the reactor vess2I head penetrations or other appropriite actions.
Through participadon in WOG and U.S. industry programs the Westinghouse plant owners have taken a proactive approach to address the cracking issue in RV head penetrations. This approach is based on the conclusion that the issue is not an immediate safety concem, I
because (1) the PWSCC process is slow; (2) the allowable or critical flaw size is large; (3) leak-before break (LBB) will occur to allow safe shutdown of a plant and (4) at least six additional years of operation with a penetration leak is required before ASME Code structural ma rgins are challenged.
In addition to the material contained in this report, detailed integrity assessments have been i
completed for all Westinghouse plants, and these results are being incorporated into an integrated response to the Generic Letter 97 01, which is being prepared in cooperation with the Nuclear Energy Institute. This response will be transmitted to the NRC by the end of 1997.
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1.0 INTRODUCTION
1.1
SUMMARY
OF THE SAFETY EVALUATIONS The purpose of this section is to review the significance of cracking in pressurized water reactor (PWR) vessel head penetrations and to desenbe the management of the issue in response to the recently released NRC Generic Letter 97-01. This report covers the following areas:
l worldwide PWSCC history in head penetrations; safety evaluation conclusions reached by WOG and industry and approved by the NRC relative to PWSCC: and a number of supporting tasks performed by Westinghouse for the WOG concerning this issue. The latest findings on this subject are summarized, along with response to specific questions in Generic Letter 97 01.
In February of 1993, Westinghouse and the Westinghouse Owners Group performed an assessment of the continued safe operation of Westinghouse designed NSSS plants in light of the cracking that had been reported in French supplied ano operated plant reactor vessel head penetrations.
Westinghouse reviewed the available metallographic and fractographic data from the French plant and concurred with the EdF conclusion that the mechanism of degradation of the Bugey 3 reactor vessel penetration was due to pnmary water stress corrosion cracking.
The Westinghouse safety evaluation [1] provided the following elements:
- 1. A summary of the vesselhead penetration stress analyses that focuses on the nature and orientation of cracking that may occurin the Alloy 600 penetration material. The l
Westinghouse evaluation concluded that the penetration residual stress induced by welding into the reactor vessel head was the initiating source promoting crack initiation and growth in a susceptible microstructure.
l
- 2. A summary of the crack propagation analysis along with the basis of the prediction methodology. As indicated in Section 2 of this report, continued crack growth testing has confirmed the initial expectations. The analysis also predicted that cracking would be axial l
and any cracks formed would be limited in extent by the penetration stress field distribution.
l The crack lengths predicted were found to be much smaller than the length of cracking required for any instability. The existence of circumerential cracking is unlikely due to the nature of stress distribution in the penetrations (i.e., hoop stress dominates the stress field).
- 3. A description of an assessment of the Westinghouse Owners Group vessels with respect to crack indications reported at Ringhals, Beznau, and various Edf plants. Important parameters applicable for crack initiation (i.e., time, temperature, stress, and material) were l
compared to those of Ringhals, Beznau and EdF plants. A comparison of susceptibility predictions suggested that the WOG vessels were generally less susceptible than Ringhals.
However, several vessels were found to be more susceptible. Since this initial evaluation, three of these vessels were inspected for penetration cracking. One vessel head was found with cracking in a single penetration and no cracking was found in the penetrations of the other two plants. The level and depth of cracking was found to be covered by the Westinghouse Safety Evaluation.
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4.
A pinettstion Inkage asssssment summtrizing luk rSta vs. crsck size. Expectations from this evaluation were that (a) leakage would be detected well before cracks extended to their critical flaw size (through wall, and 8.5-20 inches long) and (b) Boron deposits would be significant enough from small flaws to be readily visible during a Generic Letter 88 05 walkdown.
I
- 5. A vessel head wastage and structural evaluation. The evaluation showed that the loss of i
approximately 1.0 in3 of vessel head material per year could be expected if cracks initiated l
and propagated through wall, however, vessel structural margins would be maintained for at least six additional years following the through wall leak.
1.2 HISTORICAL BACKGROUND l
l In 1991, during a hydrotest of the reactor coolant system at the Bugey Unit 3 power plant in t
i France, a leak from the reactor vessel head was detected by acoustic monitoring (2].
Subsequent investigation, by visual examination and destructive testing, revealed that the leak came from a through wall flaw in one of the head penetrations. Further inspections on this and many other plants in France led to the discovery of flaws in the head penetrations of several plants. Examinations confirmed that the problem was directly related to Primary Water Stress l
Corrosion Cracking (PWSCC).
EdF conducted additional CRDM (Control Rod Drive Mechanism) penetration inspections at its nuclear plants, using eddy current techniques for indication detection and ultrasonic methods for defect size determination. Inspection results and metallurgical examinations confirmed PWSCC in CRDM penetrations at several other EdF plants. This was a concern to the French regulatory authorities as well as to the other PWR owners and regulatory authorities around the world.
These incidents are similar in nature to what occurred to other Alloy 600 tubular parts used in the Reactor Coolant System (RCS). Over the past few years, cracks in Alloy 600 pressurizer heater sleeve penetrations and instrumentation nozzles (3,4] have been reported at non-Westinghouse supplied domestic and French PWR plants. In February 1990 the USNRC issued information Notice 90-10 on this issue (5). The Notice informed PWR utilities of a number of incidences of PWSCC of Alloy 600 in applications other than steam generator tubing and suggested that utilities review their Alloy 600 applications and implement an augmented inspection program as necessary. In 1990, EPRI issued a report (4) which suggested that utilities should identify locations where Alloy 600 is used on the primary side, review the material and f abrication records to assess material susceptibility to PWSCC in terms of microstructure, stress, and environment, and implement an inspection program to detect leakage or cracking with the view of repiacing susceptible components, as appropriate.
' The Westinghouse Owners Group (WOG) and Westinghouse initiated and helped to lead a joint industry owners group under NUMARC, now the Nuclear Energy institute (NEI), beginning in 1992. The group consists of all owners of Pressurizer Water Reactors in the USA along with EPRl. This group shared technicalinformation and developed consistent safety evaluations and evaluation procedures for flaws that may be found during iWoens. The group also worked with EPRI to develop inspection performance demonstrations for the head penetration inspections. The group demonstrated to the US Nuclear Regulatory Commission that cracking 12 July 1997 l
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on the head penetrations was not an immediate safety issue. The NRC concurred with the Westinghouse conclusion, stating that vessel head penetration cracking is not an immediate safety issue (5).
1.3 INSPECTIONS PERFORMED TO DATE In 1994, two WOG/ Westinghouse PWR plants in the US (Point Beach Unit 1 and D. C. Cook Unit 2) voluntarily performed inspections of the CRDM penetrations. The results showed that there were no indications found in Point Beach Unit 1. Three indications were found in a single l
penetration at D.C. Cook Unit 2. These were significant cracks but considerably smaller than the NRC approved acceptance.imit.
l In Spring of 1996, D. C. Cook Unit 2 re-inspected some of their penetrations that had been previously inspected and confirmed the same indications reported earlier. No new indications f
were found and the existing indication was successfully repaired. Meanwhile, North Anna Unit 1 inspected 20 out of the total complement of 65 penetrations. No indications were found.
j I
A large number of inspections have ben performed on Westinghouse supplied reactor vessel head penetrations throughout the world, and this section will document those inspections, and the findings to date.
ASME Code Section XI inspections (VT 3) have been performed for a number years on the.
l head penetration to reactor vessel partial penetration weld, and the weld between the head penetration tube and the control rod drive mechanism (CRDM) While these inspections do not f
cover the Alloy 600 inside diameter surface region of the head penetration directly, they do provide surveillance information on the head penetration region, and must be performed on every penetration once every ten years. To date no indications have been reported.
l i
A sec%d series of inspections which have been carried out regularly since 1988 involves visual l
sunteillance of the head for boron deposits which would be evidence of leaks, following NRC l
Generic Letter 88 05. Some boron deposits have been found by this surveillance, but the l
sources of the leakage were.ng,1 from cracked head penetrations. Generally these leaks have been associated with mechanical seals or canopy seals on the vessel head.
Westinghouse supplied NSSS plants in Spain, Sweden, Switzerland Belgium, Brazil, and Korea have conducted NDE inspections on Reactor Vessel Head Penetrations. By the beginning of 1996, some 5200 penetrations had been inspected woridwide. The results are i
. summarized in Table 1-1. On average, indications were found in approximately 2% of the penetrations that were inspected. Based on Table 1 1,it appears that the rate of indicati U.S. plants is significantly less than that of the French plants. The operating time for the pl of US manufacture where the inspections have been performed has in most cases been muc longer than for the French plants. Of all these inspections, only one penetration was fo have through wall cracking: the Bugey plant where cracking was first identified.
It will be of interest to examine the history of inspections of the plants of Westinghouse d worldwide, as well as the plants of Westinghouse design with US fabrication. A relatively lar l
number of these plants have been inspected, and very few indications have been found.
Outside of France, a total of 39 plants of Westinghouse design have been inspected. O i
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l approximately 1900 ptnetrations insp:cted, only 10 were reported to be cracked, emounting to a less than 0.6 percentage. Of the 39 plants,9 were manufactured in the USA, and for these
[
plants approximately 310 penetrations were inspected with only one reported to be cracked.
l Thus, for Westinghouse plants manufactured in the USA, only 0.3 percent of the penetrations have been found to be cracked.
Root cause evaluations concluded that the cracks were caused by PWSCC of the Alloy 600 l
matenal. Electricite de France (EdF) and Westinghouse concluded that the following factors contnbuted to the Bugey Unit 3 PWSCC, l
Susceptible microstructure produced during manufacturing Surf ace finish on the inside diameter surf ace of the penetration Stresses induced during welding, which caused ovalization of the penetration l
l l
l July 1997 14 Rev.0 o:0710. doc:1D:13.07.97
\\
TABLE 1 1 WORLOWIDE VESSEL HEAD PENETRATION PWSCC INSPECTION RESULTS*
Number of Total No. of Number of Penetrations Rate of Plants Penetrations Penetrations With Indication Country inspected in the plants Inspected Indications Detected" France 47 3225 3213 105 3.3%
Sweden 3
195 190 7
3.7%
Switzerland 2
72 72 2
2.8%
Japan 17 960 834 0
0 Belgium 7
435 435 0
0 Spain 5
325 102 0
0 Brazil 1
40 40 0
0 South Afnca 1
63 63 0
0 South Korea 1
65 65 0
0 United States 5
314 217 1*"
0.5%
Total:
89 5694 5231 115 2.0%
Based on cata available as of January 1996 (Europe) and July 1996 (U.S.).
Ratio of number of penetrations v.ith indications detected to number of penetrations inspected.
Oconee indications were not counted as cracks, because they had no measurable depth. Eddy current reinspection after one cycle did not indicate any growth i
l l
2 Rev. O
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1-5 July 1997 c:0710. doc:1D:13.07.97
F. -
1.4 WOG AND NUCLEAR INDUSTRY PROGRAMS
SUMMARY
A number of WOG programs were initiated to investigate the reactor vessel head penetration PWSCC issue. The key programs are summarized in Table 1-2. Additionally, selected utility programs have been responsible for the resolution of IGA due to sulfur species, and penetration attachment weld cracking. Domestically, the Babcock and Wilcox Owners Group (BWOG),
Combustion Engineering Owners Group (CEOG), Westinghouse Owners Group (WOG) and the Electne Power Research Institute (EPRI) agreed to combine their efforts as part of the Nuclear Energy Institute's (NEI) Alloy 600 CRDM Head Penetration Cracking Task Force. The purpose of the task force was to evaluate the issue and to recommend appropnate generic actions.
Through this effort, the Owners Groups (OGs) and EPRI have conducted the following tasks:
Performed safety analyses of vessel head penetration cracking Standardized flaw evaluation methods Developed flaw acceptance criteria Developed inspection methodologies to size indications in head penetrations Evaluated remedial measures and created probabilistic and economic decision making tools Evaluated leakage effects on the vessel head low alloy steel shell in addition, WOG has developed penetration repair techniques, plant inspection guidelines, and evaluated available leakage detection devices.
The NRC has evaluated the safety analyses and concluded that PWSCC of Alloy 600 head penetration is not an immediate safety concem (6].
i Under the programs, research on PWSCC was conducted domestically and overseas, for example, as shown in Refs. 3,7,8,9 and 10. The studies focused on material aspects and mechanics. Material aspects, thermomechanical processing effects, material properties, residual stresses, and microstructure were studied. A model of PWSCC susceptibility and cracking probability was developed (10).
Finite element analyses were performed to determine stresses in the penetracons. The finite element analyses performed included simulation of the whole spectrum of the mechanical fabrication sequences experienced by the RV head penetrations, such as the welding process, hydrotest, straightening and service loads. The finite element simulations allowed the determination of the applied as well as the residual stresses in the penetrations under any given specific geometrical, material, welding, temperature, and loading conditions. Based on the stress data, PWSCC initiation, crack propagation, and final failure were then evaluated. The analysis also fumished results for the time period required for the PWSCC to penetrate thro the wall thickness of the penetration and the critical crack size above which instability would occur. Initie crack growth behavior was assumed to be represented by the model developed by P. Scott (11).
i July 1997 1-6 Rev.0 o.0710.coc:10:13.07.97
._= - -
Confirmatory crack growth 1:boratory testing was immediately begun to verify that this initial assessment was correct. The integrity model was structured to be applicable to all penetrations regardless of product form or vessel fabricator. Subsequent testing to obtain comparison data in this area was initiated in 1996. The crack growth test results and preliminary crack initiation test results are discussed in Sections 2 and 3.
l l
1 I
l l
l t
Rm'. 0 1*7 d"IY1807 oM710. doc:1b:07/1497 l
t l
TABLE 12
SUMMARY
OF KEY TASKS PERFORMED BY WOG ltem Task Description Status 1
Root Cause of Cracking C
2 Key Matenal & Operation Parameters C
3 Elastic Finite Element Analysis:
C Residual / Operational 4
Elastic / Plastic Finite Element Analysts:
C Residual / Operational: 3 Locations 5
Crack Propagation / Acceptable Flaw Size C
Analysts 6
Penetration Leakage & Vessel Head C
Wastage Assessment C
7
, Safety Evaluation 8
Plant Screening / Susceptibility Cntena C
9 Matenal Microsttucture Charactenstx:s C
10
. Leakage Detection Methods Survey C
11 Evaluation of PWSCC Mrtigation Methods O
12 Gnnding Effect on Residual Stresses C
13 Development / Evaluation of Repaired C
Configurations 14 OD Crack Assessment C
15 Crack Growth Data and Testing O
16 Inspection Timing and Econome Decision C
Tools 17 Penetration Attachment Weld Safety C
Evaluation Report 18 Crack initiation Charactenzation Studies O
19 Residual Stress Measurements C
20 Development of PWSCC Susceptibility C
Ranking Models Key: C = Complete O = Ongoing.
I July 1997 18 Rev.0 c:0710.ec:10:13.07.97
f, 2.0 DEVELOPMENT OF A CRACK GROWTH RATE MODEL FOR ALLOY 600 HEAD PENETRATIONS l
Crack growth rate testing has been underway since 1992 to characterize the behavior of head penetration materials. The " modified Scott model," as described below was initially used for safety evaluation calculations in the NRC submittals made in 1992 and 1993. The goal of this section of the report is to review the applicability of that modelin light of the past five years of testing, during which over forty specimens have been tested representing 15 heats Alloy 600 of material. The original basis of the model will be reviewed, followed by all the available laboratory results, and finally a treatment of the available field results.
j The effort to develop a reliable crack growth rate prediction model for Alloy 600 began in the Spring of 1992, when the Westinghouse, Combustion Engineenng, and Babcock and Wilcox Owners Groups were developing a safety case to support continued operation of plants. At the tirpe there was no available crack growth rate data for head penetration materials, and only a j
few publications exsted on growth rates of Alloy 600 in any product form.
The best available publication was found to be that of Peter Scott of Framatome, who had developed a growth rate model for PWR steam generator materials (11). His model was based on a study of results obtained by McIlree and Smielowska (12] who had tested short steam generator tubes which had been flattened into thin compact specimens. His modelis shown in Figure 21. Upon study of his paper there were several ambiguities, and several phone conversations were held to clarity his conclusions. These discussions indicated that Reference 11 contains an error, in that no correction for cold work was applied to the McIlree/Smialowska data. The revision of the Peter Scott modelis presented below.
An equation was fitted to the data of Reference 12 for the results obtained in water chemistries that fell within the standard specification for PWR primary coolant. Results for chemistries outside the specification were not used. The following equation was fit'ed to the data for a temperature of 330'C:
= 2.8 x 10-" (K 9)' A' m/ sec dt where K is in MPa(m]". This equation impfies a threshold for cracking susceptibility, K,,c, = 9 MPa[m]". Correction factors for other temperatures are shown in Table 21.
The next step described by Sco'tt (11] in his paper was to correct these results for the effects of cold work. Based on work by Cassagne and Golpi(13), he concluded that dividing the above equation by a factor of 10 would be appropriate to account for the ettects of cold work. This step was inadvertently omitted'from Scott's paper, even though it was discussed. The revised crack growth model for 330'C then becomes:
= 2.8 x 10-12 (K - 9)' A' m / sec dt This equation was verified by Scott in a phone callin July 1992.
Rev.0 21 July 1997 o:\\37104cc:1b:07/14/77
IF D
9 Scott further corrected ibis mod 2l for the effects of temperature, but his correction was not used in the model employed. Instead, an independe.it temperature correction was developed based on service experience. This correction uses an activation energy of 32.4 kCal/ mole, which gives a smaller temperature correction than that used by Scott (44 kcaVmole), and will be discussed in more detail below.
Scott's crack growth model for 330'C was independently obtained by B. Woodman of ABB-CE (14), who went back to the onginal data base, and had a smaller correction for cold work. His equation was of a slightly different form:
$ = 0.2 exp (A + B in (in (K C)})
dt Where A = 25.942 B = 3.595 C = the threshold for cracking This equation is nearly identical with Peter Scott's original model uncorrected for cold work.
This work provided an independent verification of Scott's work. A further verification of the modified Scott model used here was providea by some operational crack growth rates collected by Hunt, et al [15).
The final verification of Peter Scott's model will come from actual data from head penetration materials in service, as will be discussed in detail below. To date 15 heats have been tested in carefully controlled PWR environment. One heat did not crack, and of the fourteen heats where cracking was observed, the growth rates observed in twelve were bounded by the Scott model.
Two heats cracked at a faster growth rate, and the explanation for this behavior is being investigated.
A compilation was made of the laboratory data obtained to date in the Westinghouse laboratory tests at 325'C, and the results are in Figure 2 3. Notice that much of the data is far below the Scott model, and a few data points are above the model. These results represent 14 heats of head penetration materials.
The effect of temperature on crack growth rate was first studied by compiling all the available crack growth rate data, for both laboratory and field cracking of Alloy 600. This information is summarized in Figure 2 2, where the open symbols are for steam generator tube materials, and the solid symbols are for head penetration materials. The results are presented in a simple format, with crack growth plotted as a function of temperature. The effect of stress intensity factor variation has been ignored in this presentation, and this doubtless adds to the scatter in the data. The remarkable result is a consistent temperature effect over a temperature range from 288'C to 370'C, more than covering the temperature range of PWR plant operation although there is a wide scatter band in the figure. The work done originally in 1992 results in a calculated activation energy of 32.4 Kcal/ mole, which has been used to adjust the base crack growth law to account for different operating temperatures.
2-2 July 1997 Rev. 0 c4710.coc:10:13.07.97 l
I
-- - - - ~ ~ -
f O
i A seriIs of crcck growth tests is in progrsss under car 2 fully controlled conditions to study ths temperature effset for head penetration materitis, and the results obtrinId to dite are shown in Figure 2 2. Sufficient results are available to report preliminary findings. The tests were performed with an applied stress intensity factor of 23 Ksi 5 (25.3 MPa[m}"), periodic unload / reload parameters of a hold time of one hour and a water chemistry of 1200 ppm B + 2 ppm Li + 25 cc/kg H,. The results are consistent with the previous steam generator and head penetration material work. In the case of heat 69, the three results in the middle of the temperature range,309'C,327'C rad 341*C have the same trend as the scatter band, almost exactly, while the high temperatue and low temperature results are both lower than would be predicted by the activation cn'):gy, as shown in Figure 2 2. The results for heat 20 show a similar behavior, with the ro.r!ts at 325'C and 340*C also within the scatter band and nearly parallel to the heat M weimens, but at a lower crack growth rate, as shown in Figure 2 2.
The effects of several different water chemistries have been investigated in a closely controlled l
series of tests, on two different heats of archive material. Results showed that there is no measurable effect of Boron and Lithium on crack growth.
The key test of the laboratory crack growth data is its comparison to field data. Crack growth from actual head penetrations has been plotted on Figure 2 2 as solid points. The solid circles i
are from Swedish and French plants and the solid stars are from a US plant.
i Figure 2 4 shows a summary of the inservice cracking experience in the head penetrations of French plants, prepared by Amzallag (16), compared with the Westinghouse laboratory data, corrected for temperature. This figure shows excellent agreement between' lab and Mid data, further supporting the applicability of the lab data.
Therefore it can be seen that the laboratory data is well represented by the Scott model j
corrected for temperature using an activation energy of 32.4 kcal/ mole. Also the laboratory results are consistent with the crack growth rates measured on actualinstalled penetrations.
Therefore the use of the modified Scott modelin the safety evaluations and other evaluations of head penetration integrity is still justifiabie, in light of both laboratory and 'ield data obtained to date.
i I
f 3
Rev.0 23 July 1997 o:\\3710.coc;1b:13.07.97
TABLE 2-1 TEMPERATURE CORRECTION FACTORS FOR CRACK GROWTH: ALLOY 600 Temperature Correction Factor (CF)
Coefficient (Co) l 330C 1.0 2.8 x 1C '
325 0.798 2.23 x 10"'
320 0.634 1.78 x 10"'
310 0.396 1.11 x 10"'
300 0.243 7.14 x 10"'
290 0.147 4.12 x 10"*
i l
i f
b = Co (K 9)"' m/ s i
di l
I where K is in MPa[m)"
j l
l l
l l
I Rev.0 2-4 July 1997 i
oM710. doc:10:13.07.97 l
l
v a
crack 46.drw 1 E-09 330*C en N
l E
LLJ 1E 10... -
4 W
I
}--
N O
l C
hd 1 E.11 g
g 9
e -
I I
'i i
O 20 40 60 80 100 1E 12 K - MPa SQRT(m)
Figure 21 Scott Model for PWSCC of Alloy 600 at 330*C, as modified from Reforc. 11
~
Rev. O
~
2-5 July 1997 c:0710 doc:1D 13.07.97
.=
TEMPERATURE, DEG. C i
372 352 333 316 214 282 I
f l
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900 2
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0.0040 0 00166 0.00170 0 00115 00040 RECIPROCAL TEMPERATURE,1/DEG. K m
t o :w rw o.e.
C80W Las Tosta l
& m 20 9 we 69 u.
...wem tw a tw 0.e.
l Figure 2 2 Comparison of Temperature Effects Results with Other Laboratory and Field Data 2-6 July 1997 Rev.0 oM710.coc:10:13.07.97
7.
'l-I l
crack 47 drw 1 E-09 330*C 325'C m
E I
. $.". j,;... g.c.
l.U 1E 10...
.e.,......
4
- a..
e g.
.,i
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x
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. p g..
a g
f......._..
u 4,.........
(...
..s p..
l 1E 12 O
20 40 60 80 100 l
K - MPa SQRT(m)
Figure 2-3 Summary of Available Westinghoun Laboratory Data for Alloy 600 Head Penetrations at 325'C 3
27 July 1997 Rev.0 c:\\3710.h.1bM/1497 I
l Comparison of Field & Laboratory Data 10 1
r.
~
e-E
,e_
a y
e 7
e e!
o*8 4*a*e F
O
-9 5 0.1 --
s m -
=
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__ __._;g _-., _=.. _ =.__=c___:_,._ _.= e F l(LD_.__... a wesongh 0 001 0
10 '
20
'O 40 50 60 70 80 Ki(Mpa m")
Figure 2-4 Comparisen of French Field Data and Westinghouse Laboratory Data
(__ results reduced to 290*C using Q = 130 KJ/moie)[16]
W Rev. 0 28 July 1997 c:GT10.coc:le:13.07.97
3.0 WESTINGHOUSE CRACK INITIATION MODEL DEVELOPMENT At / CRACK IN!TIATION TESTING 3.1 CRACK INITIATION MODEL Westinghouse advanced an Alloy 600 PWSCC initiation model for primary components in Pressurized Water Reactors (10). Briefly, the model incorporates three contributing factors for the prediction of crack initiation time; namely, material condition, stress, and temperature.
These are discussed below.
Material Condition and Microstructure As reported by several authors [17,18,19,20, and 21}, the Alloy 600 microstructure is a function of the thermomechanical history of the material heat as well as its carbon content.
Alloy 600 material heats subjected to mill annealing at low temperatures, i.e.,926*C or less, exhibit a fine grained microstructure with heavy transgranular carbide precipitation and little or no carbides precipitate on the grain boundaries. Such a microstructure is reported to be more susceptible to PWSCC. On the other hand, a high temperature mill anneal (>1000*C) tends to put more carbon into solution, increases grain size, produces grain boundary chromium carbide p.ecipitation and renders the material more resistant to resist PWSCC. Norring, et. al. [22), did not find a correlation between the total content of carbon and the crack initiation time, but they observed good correlation between the amount of grain boundary carbides and crack initiation time. The fact that grain boundary precipitation is beneficial to PWSCC has been reported by many researchers [23). Norring, et. al., [22], showed that the crack initiation time varied directly (linearly) with grain boundary carbides. Their data suggested that when the grain boun'dary carbide ; overage is increased by a factor of 3, the crack initiation time also increased by a similar factor (from 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />). Bandy and Van Rooyen [24), pointed out that in addition to grain boundary carbide coverage, other features relating to processing history variables such as carbon concentration gradients, substructural features, grain size distribution, cold work, intragranular carbide distribution and the grain boundary segregates all play an important role in the cracking behavior of the Alloy 600 material.
When considering the influence of rnicrostructure on the PWSCC susceptibility for the purpose of the current evaluation, to enable comparison of heats fabricated at different vendor shops, the thermomechanical processing history effect is separated from the grain boundary carbide coverage effects. In general, the influence of the grain boundary carbides is known and the coverage (G) can be easily measured directly from the microstructure. The infteence of other structural features due to processing history cr.nnot be assessed directly. These processing effects are represented in the current treatment by a single parameter (A) characteristic of the fabrication shop (vendor). This approach provides a means of comparing the PWSCC susceptibilities of Alloy 600 material heats from different vendor shops although they may contain similar grain boundary carbide contents.
Rev. 0 3-1 July 1997 cA3710.coc1b:13.07.97 l
l l
Influenc..ef Streaa Steady state tensile stress in the component, either due to residual and/or applied loads, has a strong influence on the PWSCC.
Bandy and Van Rooyen (24), reported that the time to failure varied inversely as the fourth power of applied stress in both annealed and coldworked specimens. They also reported data to support that coldwork reduces the resistance to PWSCC. The effective stress at a given Alloy 600 location is a function of the fabrication steps and their sequence, the yield stress of the material, and the service stress. In general, the local residual stresses resulting from fabrication can play a more significant role than the service stresses themselves.
l l
Temperature Effects Severalinvestigaters [17,24), examined the role of temperature on PWSCC. It is well established fror'n these results that the crack initiation time decreases exponentially with temperature and that they are related through an Arrhenius equation expressed as a function of the activation energy of the process. The experimental results confirm that Alloy 600 PWSCC l
is a thermally activated process and the activation energy for the process varies approximately between 50 to 55 kcal per mole. An activation energy value of 55 kcal/ mole is consistently applied throughout the current assessments, for crack initiation. A different value,32.4 applies for crack growth as was discussed ;n Section 2.
I i
3.2 THE WESTINGHOUSE CRACK INITIATION MODEL l
Consistent with the contributing factors discussed above, the crack initiation time (t,) or the rate l
of crack initiation (1/t,) is proportional:
14 a (Stress)"
l ae" a inverse of the grain boundary carbide coverage factor, (1/G) l o" e
- so that 14 a G
l Since the nature of the vendor thermomechanical processing is also a significant contributing factor, one can say that for a given fabri:ation process te*
1/t,= A (3-1)
G d
1 Rev. o 32 Juty1997
(
c:07104cc:1tr13.07.97 i
4 Tho proportionality constant "A" c:n b3 chtsen to repres:nt tha processing conditions representative of a given inanufacturing process or maraufacturer, and could include parameters such as yield strength as part of the expression.
"A" can be assessed for a given heat by substituting the parameters of a service component j
with a known cracking history for the heat of material. "A" will then represent the processing cond!fion (or the vendor) by the definition we have just established.
The parameters in the above rate equation (3-1) are described below:
A is a constant, relating to the processing, and fabrication conditions of the material l
G is the grain boundary carbide coverage factor is the effective tensile stress (resulting from applied and residual stresses) o is the stress exponent having a value ranging from 3.5 to 4.5 for Alloy 600 in primary water n
O is the activation energy for the crack initiation process and has an approximate value of 55 kcal/ mole R
is the gas constant (1.987 cal / mole degrees K)
T is the absolute temperature in degrees K, and t,
is the time to initiate cracking.
3.3 CRACK INITIATION TESTING Westinghouse currently 5% an ongoing autoclave test program to establish the PWSCC crack l
initiation behavior of archive Alloy 600 RV head penetration material heats from a variety of l
f abricators representative of microstructures of RV head penetrations that are currently in l
service. Tne objectives of the Program are:
To determine the effect of penetration microstructure and material type (vendor) on I
e the relative susceptibility to cracking.
To define a materialindex (A) to assist in plant maintenance planning.
i The program is sponsored by EPRI and the CE, W, and B&W owners groups. The accelerated testing is conducted under dense steam with hydrogen at 400*C and utilizes full size ring l
samples fabricated from RV head penetration tubing from different vendor shops. A listing of vendor shops representing the ring samples employed in the testing is provided in Table 3-1.
i July 1997 1
3-3 i
Rev.0 i
c:0710.coc:1D:13.07.97
V To provide referenc] benchmarking, sampi:s from ste:m gen:rator rolled transition tubing and Alloy 690 penetration material are also included in the test matrix. Penetration material specimens with known crack growth behavior measurements from previous test programs are included for comparison with other data.
This environment has been shown to provide adequate acceleration (up to 500x) to provide results within the test period. This will be verified using the specirrens from heats that have been tested previously. Test samples under the doped steam test will be inspected at 25,50, 100,200. 400, 800,1400 and 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />. inspection will include visual, metallographic and destructive examinations.
The ID surfaces of the ring samples are strained by controlled cyclic ovalization to simulate the residual hoop stresses in the plant. The stresses are quantified based on the ovalization. The final cycle of ovalization is calibrated to induce a 2mm difference in measured inside diameter.
This corresponds to the upper 95% of the measured ovality in the outermost penetrations in i
service. The cyclic straining procedure of the full nng samples is illustrated by the loading curve shown in Figure 31.
The testing is conducted under two phases. The first phase involves a cumulative exposure of up a 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in six exposure intervals. Periodic inspecttns are performed at 25,50,100, 200,400 and 800 cumulative hours of exposure. The second phase testing involves the exposure of specimens for a cumulative exposure of up to 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> with an interim inspection at 1400 hours0.0162 days <br />0.389 hours <br />0.00231 weeks <br />5.327e-4 months <br />. Currently, with the Phase I testing completed, the preliminary test results indicate clear trends in the initiation behavior. Out of the six heats of material tested, two of the heats consistently showed Nyer susceptibility to cracking; the worst heat being the heat that also showed the highest crack growth rate under the crack growth test program discussed in Section 2. Further useful trends in cracking behavior are expected at the end of the 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> exposure. The overall results of the program are expected to provide useful information for plant maintenance planning.
l Rev.0 3-4 July 1997 c:0710.coc:10 13.07.97
TABLE 3-1 MATERIAL HEATS EMPLOYED IN THE ALLOY 600 RVHP CRACK INITIATION TESTS S No.
Heat No.
Supplier Fabricator As Prod. Size 1
93510 B&iN B&W
%* (6 pes)
~
93510-R B&W B&W h* (6 pes) 2 3
91069 B&W B&W
%* (6 pes) 4 93511 B&W B&W
%* (6 pes) 5 WF675 B&W Creusot Loire 3-5/8* (1 pc) 6 WF151 Sizewell Creusot Loire 3 h*(1 pe) 7 M 78171 (EO.
CE Standard Steel 41/8"(1 pc) 6943#2) 8 R13 4 (NX64209)
CE Huntington 41/8* (1 pc) 9 NX810175 Huntington 6* (1 pc) 10 NX34C3 68 Huntington 6* (1 pc) 11 R177 Vattenfall Sanvik 6"(1 pc) l l
l l
l
=
i
- y Rev.0 3-5 July 1997 c%7to.210:13 07.77 l
l l
.3 Active and Residual Strains during Residual Stress introduction 40000 p.
_'1 1
g 30000
.----a,,--,----
. - - - p.- - -
/
/.
'g
~
4-
.E 20000
- - - /-
,x----
/
/
.g 2
?
10000 s
-.g-o x
f
--~\\.-
2 0l
-1 t
hw c
s 5
i
..----g---
10000 g 4"
-20000 0
0.25 0.25 0.75 0.75 1.25 1.25 1.75 1.75 relax relax relax relax Loading Cycle
- Load (pounds) - -a- - Strain (ue) 12-00/6:00 t.
Strain (ue) 3:00/9:00 Figure 3-1 Residual Stresses and Strains Induced During Controlled Cyclic Ovslization of RV Penetration Ring Samples 3S July 1997 Rev.0 o.0710. doc:1b.07/14/97 t
t 7
1 1
~
4.0 TECHNICAL DESCRIPTION OF PROBABILISTIC MODELS
)
l To calculate the probability of failure of the Alloy 600 vessel head penetration as a function of operating time t, Pr(t s t,), structural reliability models were used with Monte Carlo simulation j
methods. This section describes these structural reliability models and their basis for the -
primary failure mode of crack initiation and growth due to primary water stress corrosion cracking (PWSCC). The models used for the evaluation of head penetration nozzles are based I
upon the probabilistic and economic decision tools developed previously for the Westinghouse Owners Group (WOG). The capabilities of this software have already been verified in the following ways; i
j 1.
Calculated stresses compare well with measured stresses (see Figure 41),
e 2.
Crack growth rates agree with measured field data (see Figures 2-3 and 2 4).
l l
Recent improvements have also been made to the model in order to maximize its use for individual plant predictions. Among the changes were:
]
{
1.
The model accepts measured microstructure (replication) and also has the capability to ignore its effects,if desired.
j 2.
The relationship of initiation time to material microstructural e'fects and yield strength i
has been improved to more closely match the observations from the recent inspection at North Anna Unit 1, t
3.
Statistically based Bayesean updating of probabilities due to initial inspection results has I
been added (e.g. the lack of any indications at any given plant),
4.
The uncertainty on crack growth rate after initiation has been updated to reflect the findings observed in the recent Westinghouse test data and the recent in reactor measurement data to be published by EdF (16)(see Figure 4 2),
l 5.
All models have been independently reviewed by APTECH Engineering (Begley and Woodman)(25), and an improved model was developed for the effect of monotonic yield strength on time to initiation, and 6.
A wide range (both high and low values) of calculated probabilities are consistent with actual plant observations, as discussed below.
The most important parameter for estimating the failure probability is the time to failure, t,in hours. It is defined as follows:
t, = t, + (a,- a) / da/dt (41) where:
time to initiation in hours, t,
=
failure crack depth in inches.
a,
=
Rev.O 4-1 July 1997
~
eM710.coc:1tr13.07.97 l
crack d:pth at initiation in inch::s and a,
a crack growth rate in inch / hour.
da/dt =
In equation (4-1), both the crack depths at f ailure and initiation may be specified as a fraction of the penetration wall thickness, (w). The failure depth a, depends upon the failure mode being calculated. Since the failure mode of concern is axial cracks in the penetration that are deeper than th'e structurallimit of 75% of the penetration wall thickness (w),it would be specified as:
a, = 0.75 w (42)
Tho time to PWSCC crack initiation, t,in hours, consistent with the previous equation 3.1 by RAO [3] and is defined by:
1 t, =
exp (43) o%Sf
<RT, C,=
a log normal dist;ibution on the initiation coefficient, which was based upon the data of Hall and others (25) for forged Alloy 600 pressurizer nozzles, with only the uncertainty based upon the data of Gold and others (27],
C, =
coefficient for the effect of grain boundary carbide coverage, which is based upon the data of Norring and others (22),
o=
the maximum residual and operating stress level derived from the detailed elastic-plastic I
finite-element analysis from the WOG study of Ball and others (28] as shown in Figure 41, with its normally distributed uncertainty being derived from the variation in l
ovality from Duran and others (29)(see Figure 4-3), which is a trigonometric function of the penetration diameter and setup angle (local angle t'etween the head and longitudinal t
axis of penetration).
l S, =
yield strength of the penetration material, l
n,n, = exponents on stress and yield strength, respectivrq (n, = 4, n, = 2.5)
Q, =
the activation energy for crack initiation, which ia normally distributed, i
R=
universal gas constant, and T=
the penetration absolute temperature, which is uniformly distributed based upon the calculated variation of the nominal head operating temperature.
l l
Equation 4-3 is equivalent to the initiation equation by Rao [3] as listed in Section 3.2, where G/A = C, + (1 + C, P.yS,"*,
Either data from field replication [30] or the correlation model by RAO [31] can be used to determine the percent grain boundary carbide coverage, P,,,in squation (4-3). The model[31]
~
Rev. O 4-2 July 1997 oN1710. doc:1b:07M W97
WF H
is a statistical corr lation cf m csured valu;s with the following mrterirls c rtification parameters:
- Carbon content,
- Nickel content,
- Manganese content.
Ultimate tensile strength and
- Yield strength.
The uncertainty on this model, which is as shown in Figure 4 4, applies equally well to both the predicted and measured values.
The hours at temperature per operating cycle (year), which is normally distributed, is used to check if crack initiation has occurred. Once the crack has initiated,it is assumed to have a depth ct a, and its growth rate, da/dt, is calculated by the Peter Scott model, which matches the latest Westinghouse and EdF data and the previous data given in the WOG report on the industry Alloy 600 PWSCC growth rate testing results [32), as discussed in Section 2. The crack growth model is:
da fO '
7 = C (K, -Km).1s 3
ex M
l C, =
a log-normally distributed crack growth rate coefficient (see Figure 4-2),
K, =
the stress intensity factor conservatively calculated assuming a constant stress through the penetration wall for an axial flaw at the inside surface with a length 6 times its depth using the following form of the Raju and Newman equations (33):
K, = 0.932 + 1.006 (a / w)'s(x a) as (45) 0, =
activation energy for PWSCC crack growth, which is also normally distributed, and Km=
threshold stress intensity factor for crack growth The probability of failure of the Alloy 600 vessel head penetration as a function of operating time t, Pr(t s t,), is calculated directly for each set of input values using Monte-Carto simulation.
Monte Carlo simulation is an analytical method that provides a histogram of failures with time in a given number of trials (simulated life tests). The area under the simulated histogram increases with time due to PWSCC. Th# ratio of this area to the total number of trials is approximately equal to the probability of failure at any given time. In each trial, the values of the specified set of random variables is selected according to the specified distribution. A mechanistic analysis is performed using these vatuos to calculate if the penetration will fail at any time during its lifetime (e.g. 60 years). This process is repeated many times (e.g. 6000) until a sufficient number of failures is achieved (e.g.10 per year) to define a meaningful histogram, which is an approximation of the lower tail of the true statistical distribution in time to
}
Flev. o 4-3 July 1997 o:\\3710.coe:1b:13.07.97
....-~~_-.-
friture (sea Figure 4 5). The sh po of the distribu' ion d:pinds upon the input mtdian valuts and specified distnbutions of the random variables. It is not forced to be an assumed type of distnbution (e.g. Weibull) as is done for other non mechanistic probabilistic methods. For the worst penetration in one plant, the mean time to failure was greater than 160 years but its uncertainty was so large that the normalized area under the histogram (estimated probability) at 60 years was 8 percent.
To apply the Monte Carlo simulation method for vessel head penetration nozzle (VHPN) failure, the existing PROF (probability of failure) object library in the Westinghouse Structural Reliability ano Risk Assessment (SRRA) software system was combined with the PWSCC structural reliability models described previously. This system provides standard input and output, including plotting, and probabilistic analysis capabilities (e.g. random number generation, importance sampling). The result was program VHPNPROF for calculation of head penetration failure probability with time.
As reported previously [34), the Westinghouse SRRA Software System has been venfied by hand calculation for simple models and attemative methods for more complex models.
Recently the application of this same Westinghouse SRRA methodology to the WOG sponsored pilot program for piping risk based inspection has been extensively reviewed and verified by the ASME Research Task Force on RBI Guidelines [35] and other independent NRC contractors. Table 41 provides a summary of the wide range of parameters that were considered in this comprehensive benchmarking study that compared the Westinghouse calculated probabilities from the analysis (labeled SRRA) with those from the pc PRAISE program (36). As shown in Figure 4 6, the comparison of calculated probabilities after 40 years of operation is excellent for both small and large leaks and full breaks, including those reduced due to taking credit for leak detection.
In addition, the VHPNPROF Program calculated probabilities of getting a given crack deptn due to PWSCC were compared for four plants where sufficient head penetration information and inspection results were available. The four plants are identified it' Table 4 2 along with the l
values of the key input parameters and calculated failure probabilities. Table 4 2 also shows the agreement between the latest available inspection results and VHPNPROF predicted failure trends due to PWSCC.
l b
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Rev.0 4-4 July 1997 l
l oM710.coc:1D:13.07.97 l
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TABLE 41 PARAMETERS USED FOR THE pc PRAISE BENCHMARKING STUDY Type of Parameter Low Value High Value Pipe Matenal Ferntic Stainless Steel Pipe Geometry 6.625' O.D.
2 3.0" O.D.
0.562* Wall 2.5" W'all Failure Modes Small Leak, Full Break,
]
Through Wali Crack Unstable Fracture Last Pass Weld Inspection No X Ray Radiographic Pressure Loading 1000 psi 2235 psi Low-Cycle 25 ksi Range 50 ksi Range Loading 10 cycles / year 20 cycles / year High-Cycle
- 1 ksi Range 20 ksi Range Loading 0.1 cycles / min.
1.0 cycles /sec.
Design Limiting Stress 15 ksi 30 ksi Disabling Leak Rate 50 gpm 500 gpm Detectable Lsan Rate None 3 gpm
- Note: Mechancal Vibra: ion (Iow value of stress range and high value of frequency) for small pipe. Thermal Fatigue (high value of stress range ad low value of frequency) for large pipe.
i l
l i
Rev.0 45 July 1997 c:\\3710.coc:1D:07/1417
TABLE 4-2 COMPARISON OF VHPNPROF CALCULATED PROBADtLITIES W!TH PLANT OBSERVATIONS Parameters Almaraz 1 D. C. Cook 2 Ringhala 2 North Anna 1 Hours of Operation 85,400 07.000
?08,400 91.000 Setup Angle (*)
42.6 50.5 3E6 Temperature (*F) 604.3 598.5 605.6 600.0 Yield Strength (ksi) 37.5 58 51.2 51.2 Percent GBC 57.0 44.3 30 2.0 e
Flaw Depth, Wall 0.10 0.43 0.25
- 0. t 0 Initiation Probabihty 1.1 %
41.4 %
37.6%
15.3 %
Failure Probability **
1.1 %
38.1 %
34.6 %
15.3 %
1 3
0 (2 with scratches)
With Reported Indcations from ISI
- Calculations performed at an equivalent setup angle for the 2nd highest stress location that could be inspected.
- Defined here as the probability of reaching the specified flaw depth for the individual penetration 4-6 July 1997 Rev.0 o 0710 doc:1t> 13 07 97
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.0 4 11 July 1997 l
oM710. doc:16.07/1447 i
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Figure 4-6 Comparison of Calculated Piping Probabilities 4 12 July 1997 Rev.G c11710. doc.1b.07/14/97
.o'.
5.0 REFERENCES
(1)
WCAP 13565, Rev.1 " Alloy 600 Reactor Vessel Adapter Tube Cracking Safety Evaluation," February 1993 (Proprietary).
(2)
F. Hedin and P. Gasquet, " Alloy 600 Reactor Vessel Head Penetration Cracking: An Industrial Challenge," 12* SMIRT Post Conference, August 23 25,1993, Paris, France.
[3]
Rao, G. V., and Wright, D. A., " Evaluation and Resolution of the Primary Water Stress Corrosion Cracking (PWSCC) Incidents of Alloy 600 Primary System Pressure Boundary Penetrations in Pressurized Water Reactors," Proceedings of Fontevraud 11 Symposium on " Contribution of Materials investigation to the Resolution of Problems Encountered in PWR Plants," Royal Abbey of Fontevraud, France, September 1014,1990.
(4)
A. S. O'Neill and J. F. Hall, Combustion Engineering," Literature Survey of Cracking of Alloy 600 Components in PWR Plants," Report prepared for EPRI, January 1990.
(5)
U.S.,NRC information Notice No. 90-10 " Primary Water Stress Corrosion Cracking (PWSCC) of inconel 600, February 23,1990.
(6)
NRC letter frem William T. Russell to William Rasin of NUMARC (now NEI),
November 19,1993.
[7]
Pichon, C, Boudot, R., Benham' our, C., and Gelpi, A., " Residual Life Assessment of French PWR Vessel Head Penetrations through Metallurgical Analysis." Service Exoerience and Reliability Imorovement Nuclear. Fossil. and Petrochemical Plants, PVP Vol. 288, ASME,1994.
(8)
Lagerstrom, J., Wilson, B., Persson, B., Bamford, W.H., and Bevilacqua, B, " Experiences with Detection and Disposition of Indications in Head Penetrations of Swedish Plants,
" Services Exoerience and Reliability Imorovement: Nuclear. Fossil. and Petrochemical E.lgnJg, PVP Vol.288, ASME,1994, pages 29 to 40.
(9)
Bamford, W. H., Fyfitch, S., Cyboron, R. D., Ammirato, F., Schreim, M., and Pathania, R., "An integrated Industry Approach to the issue of Head Penetration Cracking for the USA,: Services Exoerience and Reliability Imorovement: Nuclear. Fossil. and Petrochemical Plants. PVP Vol 288, AGME,1994, pages 11 to 19.
(10)
Rao, G. V.,"Methodo'ogies to Assess PWSCC Susceptibility of Primary Alloy 600 Components in PWRs," Proceedings, Sixth Intemational Conference on Environmental Degradation of Materials in Nuclear Power Systems, NACE, August 1993.
(11]
Scott, P. M., "An Analysis of Primary Water Stresa Corrosion Cracking in PWR Steam l
Generators,' in Proceedings, Specialists Meeting on Operating Experience With Steam Generators, Brussels Belgium, September 1991, pages 5,6.
i 5
Rev.0 5-1 July 1997 eM710.coc:1b:13.07.97
. y '.
[12)
Mc Ilru. A. R., Asbik, R. B., Smialowsks S.,' Relationship of Stress Intensity to Crcek Growth Rate of Alloy 600 in Primary Water,' Proceedings Intemational Symposium j
Fontevraud 11, Volume 1, p. 258 267, September 10-14,1990.
[13)
Cassagne, T., Gelpi, A.," Measurements of Crack Propagation Rates on Alloy 600 Tubes in PWR Primary Water," in Proceeding of the 5th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors,'
l l-August 25-29,1991, Monterey, Califomia.
(14)
Personal Communication, Brian Woodman, Combustion Engineering, October 1993.
[15]
Hunt, S. L. and Gorman, J.," Crack Predictions and Acceptance Criteria for Alloy 600 Head Penetrations'in Proceedings of the 1992 EPRI Workshop on PWSCC of Alloy 600 in PWRs, December 1 3,1992, Orfando Fl (published in 1993).
(16]
Personal communication C. Amzallag to W. Bamford, Feb. 26,1997, t
(17]
G. Economy, F. W. Pement, Corrosion /89 Paper 493.
f l
(18)
H.. Tass et. Al.," Relation Between Microstructural Features and Tube Cracking Observed l
on Tube Samples of Doel 2 Steam Generator." EPRI Steam Generator Owners Group, SCC Contractors Workshop San Diego, CA, March 1985.
(19]
A. A. Stein, " Development of Microstructural Correlation and a Tubing Specification for i
Alloy 600." Paper presented at EPRI Steam Generator Owner Group SCC Contractors l
l Workshop, San Diego, CA, March 1985.
l
[20]
A. R. McIlree, "Results of Reannealing Studies of Trojan Doel 2, Ringhals 2 and 3, i
Ginna and Indian Point 3 Steam Generator Tubing" Paper as in Ref.18.
(21]
G. P. Airey, " Optimization of Metallurgical Variables to improve Corrosion Resistance of Inconel Alloy 600," Palo Alto, CA, Electric Power Research Institute, EPRI NP 3051, July 1983.
i
[22)
'Intergranular Stress Corrosion Cracking in Steam Generator Tubing, Testing of Alloy t
690 and Alloy 600 Tubes," Norring, Engstrom and Norberg, in Third Intemational Syrr'oosium on Environmental Degradation of Materials in Nuclect Power Systema -
j W. Nr Reactors Proceedings, The Metallurgical Society,1988
[23)
Z. Szklarska-Smialowska," Factors influencing IGSCC of Alloy 600 in Primary and Secondary Waters of PWR Steam Generators" Proceedings of the 5*ntematonal Symposium on "Erivironmental Degradation of Materials in Nuclear Power Systems '
Water Reactors." Nace Meeting. Edited by D. Cubicciotti, August 1989, p. 61.
[24)
R. Bandy and D. Van Rooyen," Stress Corrosion Cracking of inconel Alloy 600 in High Temperature Water - An Update" Corrosion, Vol. 40, No. 8, page 425 (1984).
/
Rev.0 5-2 July 1997 l'
ox371o.coe:te:tio7.sr l
'F i
r
<0
[25)
Litter, J. A. BIgl3y (APTF.CH) to B. A. Bishop,"Rtvisw of the Westinghous2 Structur:tl Reliability, Mod:1 for PWSCC of RV Head Psnetrations " June 23,1997.
- [26)
- Evaluation of Leaking Alloy 600 Nozzles and Remaining Life Prediction for Similar Nozzles in PWR Primary System Application,' Hall, Magee, Woodman and Melton, in Service &perience and Reliability Improvement, ASME PVP V01. 288,1994 l
[27]
'The Status of Laboratory Evaluations in 400'C Steam of the Stress Corrosion of Alloy 600 Steam Generator Tubing,' Gold, Fletcher and Jacko in Proceedings of 2nd Intemational Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operations,1986
[28)
WCAP 13525, Rev.1, RV Closure Head Penetration Alloy 600 PWSCC (Phase 2), Ball et al., December 1992 (Class 2)
(29)
WCAP-13493, Reactor Vessel Closure Head Penetration Key Parameters Comparison, Duran, Kim and Pezze, September 1992 (Class 2) l l
l
[30)
G. V. Rao,-Development of Surface Replication Technology for the field assessment of Alloy 600 micro structures in Primary Loop Penetrations," WCAP-13746, Westinghouse Class 2 report June,1993.
[31)
G. V. Rao and T. R. Leax, Microstructural Correlations with Material Certification Data in Several Commercial Heats of Alloy 600 Reactor Vessel Head Penetration Materials -
WCAP 13876,'Rev.1,1997.
[32}
WCAP 13929, Rev. 2, Crack Growth and Microstructural Characterization of Alloy 600 Head Penetration Materiais, Bamford, Foster and Rao, November 1996 (Clasa 2C)
\\
[33)
Newman, J.C. Jr. And Raju,1.S. " Stress Intensity Factors for Internal Surface Cracks in j
Cylindrical Pressure Vessels" Transactions ASME, Joumal of Pressure Vessel Tec.nology, Volume 102,1980, pp. 342 346.
WCAP 14572, Westinghouse Owners Group Application of Risk Based Methods to
[34)
Piping inservice inspection Topical Report, pp E-1 to E-6, March 1996 (Class 3).
Risk Based Inspection - Development of Guidelines, Volume 1, General Document, (35)
ASME Research Task Force on Risk Based inspection Guidelines Report CRTD Vol. 20-j 1 (or NUREG/GR-005, Vol.1), American Society of Mechanical Engineers,1991 j
i NUREGICR 5864, Theoreticaland User's Manual forpc-PRAISE, A Probabilistic
~
(36)
Fracture Mechanics Computer Code for Piping Reliability Analysis Harris and Dedhia, I
l July 1992 I
i i
I Juty 1997 53 Rev.0
~
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i
.