ML20216D263
| ML20216D263 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/03/1997 |
| From: | Keaten R GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 6750-97-2223, NUDOCS 9709090268 | |
| Download: ML20216D263 (3) | |
Text
_ __. -_ - -_
l GPU Nuclear. inc.
(.
One Upper Pond Road NUCLEAR Parsippany, NJ 070541095 Tel 201316-7000 (201)316 7ll2 September 3,1997 6750-97-2223 E000 97 036 I
U, S. Nuclear Regulatory Commission Att: Document Control Desk l.
Washington,DC 20555 Gentlemen:
Subject:
Oyster Creek Nuclear Generating Station (OCNGS)
Docket No. 50-219 Facility Operating License No. DPR-16 XM-19 MaterialThsting Results
References:
(1) GPU Nuclear Letter C321-95-2175," Core Shroud Enhancement-Inservice Inspection and XM 19 Material Testing Program", dated June 7,1995.
(2) NRC Letter dated September 25,1995, "GPU Nuclear Corporation Material Testing Program for llot Rolled XM-19 Materials",
Reference 1 provided the NRC witt 'PU Nuclear's proposed hot rolled XM-19 material test program. The NRC SER accepting the proposed test program, Reference 2, imposed five conditions for acceptance, Attachment I to this letter (Lucius Pitkin, he. Report No. M96244) provides the details and results of this testing program and incorporates Conditions 1,2,3 and 5 of Reference 2.
Condition 4 of Reference 2 states,"The control specimens should include at least one sensitized XM-19 specimen from each heat of hot-rolled XM-19 to ensure that the testing conditions will J
produce IGSCC". As discussed with the NRC Staffin the telephone conferenceof February 12, f
- 1996, Condition 4 could not be met due to uncertairitiesin the ability to produce a sensitized hot rolled XM-19 material. Instead, one of the XM-19 specimens from one of the three heats included in the testing program was tested at a slower strain rate of about 5x10~'sec". This alternative was Dh found acceptable to the NRC Staffduring discussions held with New York Power Authority (James A. FitzPatrick Nuclear Power Plant) personnel on May 9,1996. Therefor, this accepted alternative
- to Condition 4 of Reference 2 was implemented in the Attachment 1 mateiial testing which represents a joint effort between GPU Nuclear and NYPA.
?,. y v< n 9709090268 970903 h
A 1
6750-97-2223 Page 2 0f 2 It should be noted that the testing program does satisfy Condition 5 of Reference 4, sir:ce water impurities were introduced into the BWR testing environment by the addition of sulfates which are the major contributor to water impurities at the OCNGS.
The results of the constant extension rate testing revealed that hot rolled XM-19, as evaluated using specimens under crevice conditions in a simulated BWR environment at 550 F, is og susceptible to intergranularstress corrosion cracking.
Since this XM 19 testing program was initiated in a joint effort between GPU Nuclear and NYPA, the results of this program are applicable to both OCNGS and FitzPatrick. Based on this, GPU Nuclear requests that the same NRC Staff member perform the review for both facilities.
If you have any questions or comments on this submittal, please contact Ron Zak, Regulatory Affairs,at (201)316-7035.
Sincen:ly, R. W. Keaten Vice President and Director Engineering c: Administrator, Region 1 Senior Residentinspector Oyster Creek NRC Project Manager
~
4 ATTACilMENT I Lucius Pitkin Incorporated Report No. M96244 HOT ROLLED XM-19 STAINLESS STEEL CORE SHROUD TIE-ROD MATERIAL -
CREVICE CORROSION INVESTIGATION GPUN No. NB70401073 (NYPA No. S94 62083, Task No. 96-4) 4
o Lucius Pitkin Gusu//ius tuviuun s\\1ctallurgical Gitanical Structural;\\leltanical 1ueoneon~,co f
testing Laboratories-Mexdrstnutie tramixation Serias 50 HUDSON STREET, NEW YORK, N,Y.10013 (212) 233 2737 FA X: (212) 406-1417 (212) 233-2558 O
Report No, M96244 O
HOT ROLLED XM 19 STAINLESS STEEL CORE SHROUD TIE-ROD MATERIAL -
CREVICE CORROSION INVESTIGATION NYPA NO. S94 62083, TASK NO. 96-4
[
GPUN NO. NB70401073 J
Prepared for New York Power Authority GPU Nuclear Corporation 123 Main Street 1 Upper Pond Road O
White Plains, NY 10601 Parsippany, NJ 07054 Attention: Messrs. William H. Spataro & Anthony Collado Pre ared by
/ /I, hccfsf Robert S. Vecchio, Ph.D., P.E.
Vice President 3
QA Review by c,9k r.
cam-s4ph P. Crosson, P.E.
QA Manager, V.P.
O April 3,1997 O
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O Lucius Pitkin O
1.0- EXECUTIVE
SUMMARY
O As a result of the widely recognized susceptibility of Type 300 series i
austenitic stainless steels to intergranular stress corrosion cracking (IGSCC) in light water reactor (LWR) coolant environments, the New York Power Authority (NYPA) and GPU Nuclear Corporation (GPUN) instituted repairs to the core shrouds of th,eir James A. Fitzpatrick (NYPA) and Oyster Creek (GPUN) boiling water reactor (BWR) nuclear power plants (NPP). Specifically, hot-rolled XM-19 0
stainless steel tie-rods (herein referred to as XM-19) were used to restrain the austenitic stainless steel shrouds within the reactor vessels in the event that the shroud circumferential welds would degrade to the point where load carrying capacity was compromised. To this end, tie-rod systems were installed in the James A. Fitzpatrick NPP in 1995 during the 11* refueling outage and in the Oyster Creek NPP in 1994 during its 15* refueling outage.
D Within the reactor vessel, the threaded tie-rods are exposed directly to the BWR coolant.
Since the tie-rods are threaded, a stress concentration and l
associated crevice condition exist in and around the thread roots.
Service conditions are such, therefore, that stress-corrosion crack initiation and subsequent propagation could occur if, of course, XM-19 is susceptible in the
,J BWR coolant environment. In this regard, Lucius Pitkin, Inc. (LPI) was requested to provide engineering services for determining the resistance of XM-19 to intergranular stress-corrosion cracking (IGSCC) in a simulated BWR coolant environment.
Constant extension rate testing (CERT) of XM-19 was performed in n
simulated BWR reactor coolant at 550*F in order to determine the resistance of XM-19 to IGSCC. The simulated coolant (as specified in Appendix B, part 4.4.)
was maintained at a dissolved oxygen concentration of approximately 10 ppm -
which is in excess of the 3 ppm saturation point for rapid IGSCC to occur in Type 304 stainless steel. Contaminant levels were controlled so as to maintain a a
conductivity of 0.4-0.5 pS/cm using sulfate additions.
Coolant pH was maintained in the range of 6.0 to 7.0, corrected to 70*F. Chemical analyses, indeed, indicated that the simulated BWR coolant chemistry remained within specified tolerances throughout the test program.
Control tests were performed in air to provide a baseline for determining the load (strength) and ductility (elongation) ratios for the BWR coolant tested
_U specimens. Load versus time curves and elongation data were obtained for each CERT specimen tested in the BWR coolant. In addition to establishing the preload levels for BWR coolant CERT specimen tests, the air results also O
D Lucius Pitkin New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g
provided the basis for calculating the coolant to air load (strength) and ductility ratios - both measures of resistance to IGSCC and, conversely the susceptibility, D
if any, of the XM-19 test specimens to IGSCC. After testing, all CERT specimens were examined visually and documented with 35 mm color photographs. In addition, the test specimens were examined metallographically and in the scanning electron microscope (SEM) to determine spec'imen fracture mechanisms.
D Results for the CERT specimens tested in air revealed XM-19 to exhibit i
substantially higher yield and ultimate load (strength) levels and lower elongations compared to Type 304 stainless steel. Since the air tests were performed at room temperature and the BWR coolant tests at 550 F, the air test ultimate load levels were adjusted to reflect the decrease in strength associated I
with increasing temperature. Based on review of allowable stress levels provided 3
in the ASME B&P_V Code for the subject materials at the respective test temperatures, a factor of 0.85 was applied to the room temperature ultimate tensile load levels. Scanning electron microscopy of the air tested specimens revealed a rough fracture morphology characterized by microvoid coalescence, as is typical of ductile overload fracture.
3 Results of the constant extension rate testing revealed that hot-rolled XM-19, as evaluated using threaded specimens under crevice conditions in a simulated BWR environment at 550*F, is Dqt susceptible to intergranular stress corrosion cracking.
That is, the relative resistance of XM-19 to IGSCC, as assessed by calculating the coolant to air load (strength) and ductility (elongation) ratios, revealed that XM-19 exhibits the same load and ductility 3
capacities in simulated BWR coolant as it does in air. It was also evident from the CERT program that sensitized Type 304 stainless steel is extremely susceptible to fracture in simulated BWR coolant. Moreover, the results of this investigation indicated that a reduction in test strain rate from 5 x 10# sec" to 5 x 108 sec" does not change the resistance of XM-19 to IGSCC. Finally, the 3
reproducibility of the results for the three different conditions of XM-19 evaluated herein attests to the homogeneity of the XM-19 material and the reproducibility of the constant extension rate test conditions.
In order to verify the mode of the XM-19 and Type 304 CERT specimen fractures, all test specimens were examined visually, metallographically, and by scanning electron microscopy (SEM).
Examination of the XM-19 CERT specimen fracture surfaces revealed a rough fracture morphology characterized by microvoid coalescence, characteristic of ductile overload fracture.
More importantly, the XM-19 specimens did not exhibit any evidence of intergranular J
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o Lucius Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
fracture. However, the Type 304 CERT specimen fracture surfaces exhibited an intergranular morphology, as is characteristic of intergranular stress corrosion l
'O cracking.
Clearly, by comparison, the XM-19 specimens were resistant to
)
IGSCC in the same environment in which sensitized Type 304 stainless steel exhibited extreme sensitivity to IGSCC.
l D
3 J
3 e
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0 Lucius-Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244
2.0 INTRODUCTION
p As a result of the widely recognized susceptibility of Type 300 series austenitic stainless steels to intergranular stress corrosion cracking (IGSCC) in light water reactor (LWR) coolant environments, the New York Power Authority (NYPA)_ and t#U Nuclear Corporation (GPUN) instituted repairs to the core shrouds of their James A. Fitzpatrick (NYPA) and Oyster Creek (GPUN) boiling water reactor'(BWR) nuclear power plants (NPP). Specifically, hot-rolled XM-19 j
stainless steel tie-rods (herein referred to as XM-19) were used to restrain the austenitic stainless steel shrouds within the reactor vessels in the event that the shroud circumferential welds would degrade to the point where load carrying i
capacity was compromised. To this end, tie-rod systems were installed in the
[
James A. Fitzpatrick NPP in 1995 during the 11'h refueling outage and in the th l
Oyster Creek NPP in 1994 during its 15 refueling outage, as shown in Fig 1.
D With regard to the shroud repair, it was reported that the most highly
- stressed (loaded) components are in-fact the threaded tie-rods which are loaded to a tensile stress of approximately 32.5 ksi or 31% of the minimum specified yield strength of 105 ksi for XM-19 (ASTM: A479, Gr. XM-19). Moreover, within the reactor vessel, the threaded tie-rods are exposed directly to the BWR h
coolant. Since the tie-rods are threaded, a stress concentration and associated crevice condition exist in and around the thread roots. Service conditions are such, therefore, that stress-corrosion crack initiation and subsequent propagation could occur if, of course, XM-19 is susceptible in the BWR coolant environment.
It is well known that three critical factors are necessary for the initiation of 3
stress corrosion cracking (SCC). These factors are: 1) a susceptible material,2) tensile stress, and 3) corrosive environment. Most austenitic stainless steels are susceptible to stress-corrosion cracking in aqueous environments depending upon the applied stress, material microstructure, and water chemistry. Of these factors, the applied tensile stress generally has the greatest effect on the time required for the initiation and propagation of SCC. Increasing the stress level 3
decreases the time for stress corrosion crack initiation.
For example, intergranular stress-corrosion cracking of sensitized Type 304 stainless steel in 0.2 ppm dissolved oxygen (O ) BWR coolant environment at 550*F occurs in less 2
than 20 hr at a stress level of about 90% of the ultimate tensile strength.
5 Whereas,10'to 10 hr are required when the applied stress is close to the yield strength.
3 Typically, BWR coolant is high-purity, 6.0-7.0 pH water containing approximately 0.2 ppm dissolved oxygen. This oxygen concentration has been
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o Lucius Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
shown to be sufficient to initiate IGSCC in sensitized austenitic stainless steels, increasing the dissolved oxygen content to approximately 3 ppm can increase
'O the rate of cracking. Anions such as chlorides, sulfates, and carbonates can further exacerbate the propensity for IGSCC. Moreover, crevice conditions, such as those which could develop in the tie-rod threads, generally exacerbate the environmental effects which, in-turn, reduces the initiation time for IGSCC.
Prior to the installation of the NYPA and GPUN core shroud repairs, a
- O number of evaluations (Ref.1) were performed to determine the adequacy of XM-19 in the BWR coolant environment. For example, initial susceptibility of XM-l 19 to IGSCC was assessed according to the requirements of ASTM
- A262.
l Additionally, constant extension rate testing was performed on smooth specimens of XM-19 in simulated BWR coolant conditions at 550*F. Results of these tests (Ref,1) revealed XM-19 to be immune to IGSCC for the evaluated lO smooth specimen test conditions. However, these initial tests did not simulate t
the stress concentration / crevice conditions potentially present at the tie-rod f
thread roots-In this regard, Lucius Pitkin, Inc. (LPI) was requested to provide engineering services for determining the resistance of XM-19 to IGSCC in mV simulated BWR coolant environment. NYPA and GPUN, in conjunction with LPI, developed a test program for the that purpose. To this end, LPI implemented a comprehensive constant extension rate test program with its subcontractor Babcock & Wilcox Research and Development Division, Alliance Research Center (ARC).
O Lucius Pitkin, Inc, maintains a Quality Control Program in accordance with the requirements of 10 CFR50 Appendix B "Qua!!ty Assurance Criteria for Nuclear Power Plants," which was followed for the subject investigation, as applicable to the services performed.
O e
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O Lucius Pitkin
........... p New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
3.0 XM 19 CREVICE CORROSION CERT EVALUATION PROGRAM o
3.1 Overview Constant extension rate testing (CERT) of XM-19 was performed in simulated BWR reactor coolant at 550*F in order to determine the resistance of XM-19 to IGSCC. The simulated coolant was maintained at a dissolved oxygen concentration' of approximately 10 ppm - which is in excess of the 3 ppm O
saturation point for rapid IGSCC to occur in Type 304 stainless steel.
Contaminant levels were controlled so as te maintain a conductivity of 0.4-0.5 S/cm using sulfate additions. Water pH were maintained in the range of 6.0 to 7.0, corrected to 70*F.
Load versus time curves and elongation data were obtained for each
.D CERT specimen tested in the BWR coolant, in addition, air tests were performed to provide a baseline for determining the strength and ductility ratios for the BWR coolant tested specimens. These ratios provide a measure of the susceptibility, if any, of the XM-19 test specimens to IGSCC.
After testing, all CERT specimens were examined visually and
'O documented with 35 mm color photographs. In addition, the test specimens were examined metallographically and in the scanning electron microscope (SEM) to
(
determine the fracture mechanisms.
Complete details of the test procedures and the test results are given in the following sections.
O 3.2 Materials As previously described, the XM-19 evaluation program was developed to determine the resistance of ASTM: A479, Gr. XM-19, to IGSCC in a crevice environment. In this regard, specification SP-1302-52-118, " Crevice Corrosion Testing For Hot Rolled XM-19 Material," was developed by GPUN for constant extension rate testing (CERT) of threaded (grooved) specimens machined from ASTM: A479, Gr. XM-19 bar stock, Jacketed with Type 304 crevice assemblies.
In addition, since sensitized Type 304 stainless steel is known to exhibit IGSCC in BWR coolant, sensitized Type 304 CERT specimens were evaluated concurrently so as to provide control specimens to-ensure that the test environment indeed was capable of inducing IGSCC. A copy of specification SP-1302-52-118 is given in Appendix A.
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O Lucius Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g
Three lots comprised of two heats of XM 19 and one lot /one heat of Type 304 stainless steel bar stock were submitted to LPI for evaluation, as described 3
in Table 1.
The test material was taken from the same stock bars (heats) as those placed in service in the respective BWR's addressed herein.
TABLE 1 TEST MATERIAL IDENTIFICATION LPI Bar Bar D
Ident.
Material Heat No.
Diameter (in.)
Lenath (in.)
Plant A
XM-19 AJ5018 1.494 7-5/16 JAF NPP (PC 18)
B-XM 19 AJ5018 1.740 5-5/8 JAF NPP (CLI 427819) l C
XM-19 A9098H 1.491 8-3/16 OC NPP D
Type 304 L27527 1.245 4-1/2 Control 1
L Prior to machining, each bar was marked and stamped with the J
appropriate letter designation. In order to obtain sensitized Type 304 stainless steel, bar D was heat treated at 'i250*F 110*F for 1 hr followed by furnace cooling. The time-temperature record for the sensitization heat treatment is given in Appendix B.
3 CERT specimens from ea_ch bar were machined with grooves to simulate the thread geometry according to the requirements of SP 1302-52-118 (Appendix A). -Each specimen was stamped with its letter designation and a sequence number as follows A1, A2,...; B1, B2,...; C1, C2,...; and D1, D2,....., etc. The dimensions of each CERT specimen were verified for conformance with the specified dimensions given in drawing No. 3B-SKM-722 (Fig. 2).
J Inasmuch as the shroud tie-rods are in contact with Type 304 austenitic stainless steel components, it was necessary to simulate a crevice environment which incorporated both the XM-19 and Type 304 materials. As such, crevice assemblies comprised of Type 304 austenitic stainless steel packing material and jacket were fabricated for each specimen, it should be noted, that Fig. 2 specifies Type 304 stainless steel wool for the crevice packing material, however,
)
austenitic stainless steel wool is not commercially available. In lieu of stainless steel wool, Type 304 stainless steel 150 x 150 mesh (0.0026 in. diameter wire) screen material was used for the crevice packing material. That is, the screen 3-7
O Lucius Pitkin New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
wires were separated and bundled to form a wool-like assemblage for wrapping around the specimen grooves. Moreover, the fine diameter of the screen wires O
provided significantly greater surface area of Type 304 material relative to the surface area of the XM-19 specimen grooves so as to enhance the effect of the environment.
The as-machined CERT specimens and crevice assemblies are shown in Figs. 3 through 7.
O 3.3 Test Procedures 3.3.1 Air Tests CERT specimens A1, 81, C1, and D1 were tested in air at room p
temperature in order to obtain (1) baseline ultimate loads and elongations for determining the BWR coolant to air load and ductility ratios, and (2) the specimen load at the onset of yielding for preloading of the CERT specimens in the BWR L
coolant solution (see Section 3.3.2). Inasmuch as the CERT specimens are threaded, the actual smooth specimen tensile stresses could not be obtained from the air test results. Rather, since the air and BWR coolant specimens are D
identical in size and shape, strength and ductility ratios were necessarily determined by a comparison of the ultimate loads and elongations.
Testing in air was performed on an Instron servo-hydraulic test frame at a d
loading rate of 0.003 see. Load and strain were measured using a load cell with an accuracy of approximately 11.0% and a 1-in. gage length extensometer mounted outside the specimen threads (grooves).
Fig. 8 shows the CERT specimen air test set-up.
Results of the air tensile tests are summarized in Table 2. As expected, the XM-19 exhibited substantially higher yield and ultimate load levels and lower elongations compared to Type 304 stainless steel. Load-strain curves for the air test specimens are given in Appendix C.
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O Lucius Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 O
Attn.: Messrs. W. H. Spataro & A. Collado M96244 TABLE 2 AIR TENSILE TEST RESULTS O
Specimen Yield Load (Ib)
Ultimate Elongation (%)
Ident.
Material (0.2% Offset)
Load (Ib)
(5/8 in. Gage)
A1 XM-19 1500 1680 17 B1 XM-19 1390 1580 15 C1 XM-19 1590 1700 12 D1 Type 304 740 1120 28
- g l
Evaluation of the CERT specimen air test results is given in Section 3.4.
3.3.2 BWR Coolant Tests l
b LPI provided ARC with ten threaded CERT specimens and associated crevice packing assemblies, as described in Table 3 and shown Figs. 4 through 7.
TABLE 3 BWR COOLANT CERT SPECIMENS O
Specimen ARC Test Ident.
Material Heat Frame A2 XM-19 AJ5018 (PC18)
SATEC 1 A-3 XM-19 AJ5018 (PC18)
SATEC 2 A-4 XM-19 AJ5018 (PC18)
CERT A-5 XM-19 AJ5018 (PC18)
Not Used
_)
B-2 XM-19 AJ5018 (CLI 427819)
SATEC 1 B-3 XM-19 AJ5018 (CLI 427819)
SATEC 2 C-2 XM-19 A90984 SATEC 1 C-3 XM-19 A90984 SATEC 2 D-2 Type 304 L27527 SATEC 1 3
D-3 Type 304 L27527 SATEC 2 The CERT program utilized dual 200 gallon feed-tanks which provided coolant solution to concurrently operating autoclaves, identified by ARC as
" CERT" and "SATEC" test frames, in a once-through mode, as shown in Figs. 9 and 10.
The dual feed-tanks and associated piping were constructed of austenitic stainless steel. Constant extersion rate testing was conducted in the five-specimen "SATEC" autoclave which is constructed from Alloy C-276. The one specimen " CERT" autoclave is constructed from Alloy 600, it should be noted that these materials have no effect on the simulated BWR environment.
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0 Lucius Pitkin
............ g New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g
Schematic diagrams and photographs of the "SATEC" and " CERT" autoclave systems are presented in Figs.11 through 15.
O Simulated BWR coolant solution was prepared in accordance with specification SP-1302-52-118, as follows:
Conductivity -
0.4 - 0.5 pS/cm 6.0 - 7.0 pH 0
Dissolved O -
saturated at room temperature (~8 ppm) 2 Chloride
<5 ppb sufficient to maintain conductivity and pH Sulfate Temperature -
~550*F.
In order to maintain the conductivity and pH within the specified ranges, a O
sodium sulfate / sodium bisulfate solution was prepared such that the total sulfate concentration was ~100 ppb. Following initial feed-tank preparation, the test 2
solution from both feed tanks was analyzed for conductivity, Cl, and SO4 These analyses were repeated following each subsequent feed-tank preparation.
l In addition, both feed-tanks were analyzed for conductivity at least once every two weeks. The pH of the test solution in the feed tanks was calculated from the C
rneasured conductivity assuming that the only constituents in the water were Na*
2 l
and SO4, and that the sodium sulfate to sodium bisulfate molar ratio was 3.5:1.
To maintain oxygen saturated conditions with a 6.0 - 7.0 pH, the solution inside the feed-tanks was covered with bottled CO2-free air. A small positive pressure was maintained to produce a dissolved oxygen concentration in the test g
solution of approximately 9-11 ppm. Had the test solution been exposed to air with CO2, the pH of the coolant solution would have been less than 6.
Results of the chemical analyses performed on feed-tank and autoclave effluent coolant solutions are given in Tables 4 and 5. A plot of the calculated 3
values for dissolved oxygen in the coolant solution (feedwater) during the entire test program is shown in Fig.16.
Simulated coolant dissolved oxygen concentration and pH were calculated as a function of temperature and conductivity, respectively, as shown in Appendix D.
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O Lucius Pitkin New York Power Authority /GPU Nuclear Corp.
April 3,1997 O
Attn.: Messrs. W. H. Spataro & A. Collado M96244 l
TABLE 4 CHEMICAL ANALYSES OF FEED-TANK COOLANT C
Feed tank 9 Feed tank 12 Conductivity Cr so4'8 Calculated Conductivity Cf ! so4'8 Calculated Date (psicm)
(ppb)
(ppb) pH (psicm)
(ppb)
(ppb) pH 9/10/96 0.387
<5 110 G.55 0.367
<5 100 6.57 (a)
(a)
(a)
(a) 6 49 0.458 6.49 g
9/20/96 0.456 (a)
(a) 10/2/96 0.470 6.48 0.459
<5 79 6.49 (a)
(a) 10/15/96 0.418
<5 77 6.53 0.410 6.53 Note: (a) Measurements not required on the noted date based on tank
,,V usage TABLE 5 CHEMICAL ANALYSES OF AUTOCLAVE COOLANT EFFLUENT
,0
" CERT" "SATEC" Autoclave Autoclave 2
cr soi' cr soi Date (ppb)
(ppb)
(ppb)
(ppb) 9/20/96
<5 96
<5 99 3
9/27/96
<5 73
<5 73 10/11/96
<5 74
<5 71 10/18/96
<5 76
<5 70 3
10'26/96 (a)
(a)
<5 73 Note: (a) Test completed between 10/18/96 and 10/26/96.
Clearly, these results indicate that the simulated BWR coolant chemistry remained witnin specified tolerances throughout the test program. Furthermore, a comparison of the feed-tank sulfate analysis results with the sulfate analyses of the autoclave effluent indicates that sulfate did not accumulate in the autoclaves during the test program.
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O-Lucius Pitkin New York Power Authority /GPU Nuclear Corp.
April 3,1997 O
Attn.: Messrs. W. H. Spataro & A. Collado M96244 3
Prior to testing, the autoclaves were thoroughly cleaned and rinsed with high purity water. The cleanliness of the autoclaves was verified by conductivity measurements. Following cleaning, specimen A-4 was inserted into the loading fixture in the " CERT" autoclave head, and specimens A-2, B-2, C-2, and D-2 were inserted into the loading fixture in the "SATEC" autoclave head.
The autoclave heads were-then bolted to their respective autoclaves and the D --
autoclaves were filled with the coolant solution. Coolant solution was pumped through the autoclaves at a rate of approximately 0.3 gal /hr. Autoclave pressure was maintained above the saturation pressure to insure that the autoclaves were completely filled with test solut'on at all times. The autoclaves were then heated
(
to approximately 550*F.
l)
Once the specified temperature, pressure and flow rates were established, l
the specimens were preloaded to approximately 75% of the 0.2% offset yield load level, as shown in Table 6.
It should be noted, that the specification indicated that the CERT specimens be preloaded to 90% of the yield stress.
1 However, a true yield stress could not be determined due to strain hardening affects associated with the presence of the thread root stress concentrations.
}
-Additionally, as shown in Appendix C,90% of the 0.2% offset yield load is well into the nonlinear regime of the load-strain curve. Accordingly, preload levels were selected for each threaded specimen group as the point on the load-strain curves just at the onset of nonlinear behavior.
TABLE 6 3
CERT SPECIMEN PRELOAD LEVELS Specimen Yield Load, CERT Specimen Series 70*F in Air (Ib)
Preload At 550*F (Ib) _
A 1500 1100 1
B 1390 1050 C
1590 1200-D 740 450
)
The CERT specimens were held at the indicated preload levels for 300 hr so as to permit the development of a suitable crevice environment at the stressed thread roots, except for specimen series D which failed prior to the end of the 300 hr hold period. When, during the 300 hr hold, the load on the specimens dropped below 90% of the specified preload, the load on the specimen was
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D Lucius Pitkin New York Power _ Authority /GPU Nuclear Corp.
April 3,1997 3
Attn.: Messrs. W. H. Spataro & A. Collado M96244 increased to the specified value.
D Following the 300 hr hold period, the specimens in the "SATEC" autoclave were loaded at a displacement rate of ~3.12 x 10'7 indsec, whereas the specimen in the " CERT" autoclave was loaded at a displacement rate of ~3.12 x 10
- indsec. These displacement rates corresponded to approximate strain rates of 5 x 10'7 sec" for specimens in the "SATEC" autoclave and 5 x 10-8 seco for the specimen in the " CERT" autoclave.
Displacement controlled loading was
)
continued until all specimens fractured.
During the course of testing, the foIP ting parameters were continuously monitored by a computerized data acquisition system (DAS) which collected data every fifteen minutes (see Appendix E):
. - Specimen load; Ram head / pull rod displacement; e
Autoclave temperature; and e
Autoclave pressure.
7 During testing in the " CERT" autoclave, the displacement transducer malfunctioned. Consequently, a dial gauge displacement indicator was attached i
to the " CERT" autoclave pull rod so as to provide continued displacement readings.
Dial gage readings did not reveal any changes in the " CERT" autoclave displacement rate.
3 After the first four specimens in the "SATEC" autoclave fractured, the autoclave was cooled and opened, and the specimens were removed. The specimens were dipped in methanol to promote drying, given a cursory visual examination, and protectively packaged without further cleaning. The crevice forming Jacket was left in place on one half of each specimen.
)
Subsequently, test specimens A-3, B-3, C-3 and D-3 were inserted into the "SATEC" autoclave. Specimen placement in the autoclave was such that specimens from the same series were not connected to the same load train (i.e.,
Specimen A-3 was not connected to the same load train as Specimen A-2 in the
_ previous test, etc.) in order to eliminate any bias in data due to position in the loading fixtures. The second test in the "SATEC" autoclave followed the same
)
procedures described previously for the first "SATEC" autoclave test. The test in the " CERT" autoclave continued uninterrupted during, and between, the two CERT tests in the "SATEC" autoclave.
)
13
O Lucius Pitkin
............ g_
New York Power Authority /GPU Nuclear Corp.
April 3,1997 g
Attn.: Messrs. W. H. Spataro & A. Collado M96244 The post-test procedures for the four specimens from the second test in the "SATEC" autoclave and the single specimen tested in the " CERT" autoclave D
followed those described above for the specimens from the first "SATEC" autoclave test.
3.4 Air Test Results Tensile results for the CERT specimens tested in air were previously 3
presented in Table 2. To reiterate, XM-19 exhibited substantially higher yield and ultimate load levels and lower elongations compared to Type 304 stainless steel.
In addition to establishing the preload levels for BWR coolant CERT specimen tests, the air results provide the basis for calculating the coolant to air load (strength) and ductility ratios - both measures of resistance to IGSCC. Since the air tests were performed at room temperature and the BWR coolant tests at h
550'F, the air test ultimate load (strength) leals were adjusted to reflect the decrease in strength associated with increasing temperature. Based on review of allowable stress levels provided in the ASME B&PV Code (Section ll, Part D, Table 1A) for the subject materials at the test temperature, a factor of 0.85 was applied to the room temperature ultimate tensile load levels. These adjusted
)
ultimate loads (strength) were used to calculate the load ratios for XM-19 tested in simulated BWR coolant.
Examination of the air tested CERT specimens (A1, B1, C1, and D1),
shown in Figs.17 and 18, revealed the fractures to be similar and to have occurred through the 2"d or 3'd thread root. In addition, the specimen fracture
)
surfaces were rough and irregular in appearance with no evidence of crack branching - features characteristic of ductile overload fracture. Scanning electron microscopy of the air tested specimens, shown in Figs.19 through 22, revealed a rough fracture morphology characterized by microvoid coalescence, as is typical of ductile overload fracture.
)
3.5 Constant Extension Rate Test Results Load versus time plots for the two "SATEC" autoclave tests and the
" CERT" autoclave test are given in Figs. 23 through 25. Maximum load and time to failure results for each specimen group are summarized in Table 7. Ram head displacement rate versus time (pull rod displacement versus time for the " CERT"
)
autoclave), autoclave temperature versus time, and autoclave pressure versus time plots for each test are given in Appendix E. It is evident from Appendix E, j
that pressure and temperature remained constant during the course of testing and that pull rod displacements increased uniformly up to the specimen fracture
)
14
o Lucius Pitkin
............ g
~
New York Power Authority /GPU Nuclear Corp.
April 3,1997 9
Attn.: Messrs. W. H. Spataro & A. Collado M96244 1
loads.
O TABLE 7 CONSTANT EXTENSION RATE TEST RESULTS "SATEC" Test 1 "SATEC" Test 2
" CERT"
($ x 10'7 sec)
(5 x 10'7 sec)
(5 x 10 a sec)
Specimen Time to Maximum Time to Maximum Time to Maximum Group Fall (hr)
Load (Ib)
Fall (hr)
Load (Ib)
Fall (hr)
Load (Ib)
A 368.3 1378 370.1 1410 891.7 1374 B
362.3 1323 359.9 1303 C
345.1 1363 346.4 1368 D
198.2 414 159.7 449 It is clearly evident from these results, that the Type 304 stainless steel specimens fractured prior to the end of the 300 hr hold period. In that the Type H
304 specimens were intentionally sensitized and fractured at a load significantly H
below their ultimate load, demonstrates that the simulated BWR coolant was sufficiently aggressive to promote stress corrosion cracking and thus suitable for evaluating the resistance of XM-19.
For the XM-19 specimens tested at a strain rate of ~5 x 10'7 sec, the g~
group C specimens failed first, followed the group B specimens and then the group A specimens in both "SATEC" tests. The maximum load attained by group A specimens was somewhat greater than that attained by the group B and C specimens. In fact, the CERT coolant exposed specimen failure loads ranked in the same order as the ultimate load results obtained in air, indicating little or no environmental effect on XM-19.
3 Similarly, the group A, XM-19 specimen tested at a strain rate of ~5 x 10-8 sec reached nearly the same maximum load as the group A specimens tested at ~5 x 10'7 sec and a load greater than the XM-19 group B and C specimens.
4 This latter result clearly indicates that there is no effect of strain rate on the resistance of XM-19 to IGSCC. Moreover, the reproducibility of the test results for the three XM-19 specimen groups attests to the homogeneity of the XM 19 material and the reproducibility of the CERT conditions.
The relative resistance of XM-19 to IGSCC was assessed by calculating S
15
0 Lucius Pitkin
............ g_
New York Power Authority /GPU Nuclear Corp.
April 3,1997 3
Attn.: Messrs. W. H. Spataro & A. Collado M96244 the coolant to air load (strength) and ductility (elongation) ratios. Results of such calculations are given in Table 8 and indicate that hot-rolled XM 19 exhibits the D
same load and ductility capacity in simulated BWR coolant as it does in air, it is also evident that sensitized Type 304 stainless steel is extremely susceptible to fracture in BWR coolant.
Had XM-19 been susceptible to stress corrosion cracking in the BWR coolant, then the XM-19 specimens would have exhibited significant reductions in both the load and ductility ratios.
3 TABLE 8 l
CERT LOAD AND DUCTILITY SUSCEPTIBILITY RATIOS Specimon Max Load in Coolant Elonnation in Coolan_t ident.
Material Strain Rate Max Load in Air Elongation in Air A2 XM 19 5x10 sec" 0.97 0.88 B2 XM 19 5x10 sec" 0.98 0.93 3-C2 XM-19 5x10 sec" 0.94 1.00 D2 Type 304 5x10 sec" 0.43 0.14 A3 XM-19 5x10 sec" 0.98 0.88 B3 XM-19 5x10 sec" 0.97 1.00 C3 XM-19 5x10 sec" 0.95 1.00
)
D3 Type 304 5x10 sec" 0.47 0.18 A4 XM-19 5x10 sec" 0.96 1.12 in order to verify the mode of the XM-19 and Type 304 CERT specimen fractures, all test specimens were examined visually, metallographically, and by 3
scanning electron microscopy (SEM). Prior to examination, the specimens were cleaned using benign cleaning techniques, that is, fiber brush scrubbing and ultrasonic agitation.
Visual examination was performed using a binocular microscope up to a magnification of 25X. SEM examination was performed at an accelerating potential of 20 kV.
)
Visual examination of the CERT specimens, shown in Figs. 26 through 30, revealed the XM-19 fractures to be similar and to exhibit rough and irregular fracture profiles and secondary cracking of the adjacent thread roots, as is characteristic of ductile overload fracture. In contrast, the Type 304 specimens exhibited relatively flat and granular fracture profiles, as is characteristic of stress corrosion cracking.
The Type 304 specimens also exhibited intergranular
)
secondary cracking in the adjacent thread roots.
SEM examination of the XM-19 CERT specimen fracture surfaces revealed a rough fracture morphology characterized by microvoid coalescence,
)
16
o Lucius Pitkin New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 nv characteristic of ductile overload fracture, as shown in Figs. 31 through 37. More importantly, the XM-19 specimens did not exhibit any evidence of intergranular O
fracture. However, as shown in Figs. 38 and 39, the Type 304 CERT specimen fracture surfaces exhibited an intergranular morphology, as is characteristic of stress corrosion cracking.
Clearly, the XM-19 specimens were resistant to IGSCC in the same environment in which sensitized Type 304 stainless steel exhibited extreme susceptibility to IGSCC.
O Metallurgical evaluation of the CERT test specimens was performed to (1) further assess the mode of fracture and (2) determine the nature of the secondary cracking in the thread roots adjacent to the fracture surfaces. One-half of each specimen was thus sectioned longitudinally so as to intersect the fracture surfaces, mounted in plastic, polished, and electrolitically etched for metallographic examination.
J The fracture profiles of the XM-19 specimens, shown in Figs. 40 through l
43, were, except for specimen A4, irregular and typical of ductile overload fracture. Specimen A4 exhibited a slanted fracture profile also characteristic of ductile shear fracture. The slanted nature of the A4 fracture is attributed to the development of coincident fractures in adjacent threads which merged at the V,
point of final fracture.
I In contrast, the Type 304 specimens exhibited intergranular fracture profiles with intergranular secondary cracks, shown in Fig. 44, as is characteristic of stress corrosion cracking.
O Metallographic examination of the secondary cracking in the thread roots adjacent to the fractures, shown in Fig. 45, revealed blunted ductile tears in the XM-19 specimens and branched intergranular cracking in the Type 304 specimens, further demonstrating the resistance of XM-19 to IGSCC and Type 304's sensitivity to the coolant environment.
O e
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- 7
0 Lucius Pitkin
............ g New York Power Authority /GPU Nuclear Corp.
April 3,1997 g
Attn.: Messrs. W. H. Spataro & A. Collado M96244
4.0 CONCLUSION
S O
Results of the XM-19 constant extension rate test program revealed that hot-rolled XM-19 stainless steel material, as evaluated using threaded specimens under crevice conditions in a simulated BWR environment, is aqi susceptible to intergranular stress corrosion cracking (IGSCC). That is, the relative resistance of XM-19 to,1GSCC, as assessed by calculating the coolant to air load and ductility (elongation) ratios, revealed that XM-19 exhibits the same load (strength)
O and ductility capacities in simulated BWR coolant as it does in air. It was also evident from the CERT program that sensitized Type 304 stainless steel is extremely susceptible to fracture in simulated BWR coolant.
1 Furthermore, the results of this investigation indicated that a reduction in test strain rate from 5 x 10 sec to 5 x 10-8 sec" does not after the resistance of 4
9 XM-19 to IGSCC. Finally, the reproducibility of the results for the three different l
conditions of XM-19 evaluated herein attests to the homogeneity of the XM-19 material and the reproducibility of the constant extension rate test conditions.
O RSVlib52/M96244 Final Report O
4 9
18
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l
O Lucius Pitkin
............ g_
~~
New York Power Authority /GPU Nuclear Corp.
April 3,1997 O
Attn.: Messrs. W. H. Spataro & A. Collado M96244
)
5.0 REFERENCES
O 1
- CERT Testing of Type XM 19 in Simulated BWR Environment," Document No. 51-1235147-00, B&W Nuclear Technologies, Lynchburg, VA, December 20,1994.
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April 3,1997 Attn. Messrs. W. H. Spataro & A. Collado M96244
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 lW C'9 l
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Lucius Pitkin
........,.. g
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 D
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the as-machined and crevice jacketed condition. Unjacketed specimen A1 (bottom) was tested in air.
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Lucius Pitkin New York Power Autho;lty/GPU Nuclear Corp.
April 3,1997
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the as-machined and crevice jacketed condition. Unjacketed specimen B1 (bottom) was tested in air.
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the as machined and crevice jacketed condition. Unjacketed specimen C1 (bottom) was tested in air.
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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Lucius Pitkin 7
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April 3,1997 Attn.: Messrs. 'N H. Spataro & A. Collado M96244
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado MD6244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 0
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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O Fig.16 Plot of the dissolved oxygen concentration in the feedtanks s a function of test duration.
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April 3,1997 Attn.: Messrs. W. H. Spataro & A Collado M96244 3
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!O Attn.: Messrs. W. H. Spataro & A. Collado M90244 1
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M06244
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April 3,1997 Attn.. Messrs. W. H. Spataro & A. Collado M96244
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 4
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............ g_
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 IO Attn. Messrs. W. H. Spataro & A. Collado M96244
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April 3,1997 Aun. FAessrs. W. H. Spataro & A CoHado FA96244 O
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O Lucius Pitkin
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado MD6244 O
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April 3,1997 Attn.. Messrs. W. H. Spataro & A. Collado M96244 g
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April 3,1997 Attn.: Messrs, W. H. Spataro & A. Collado M96244 O
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microvoid coalescence, as is characteristic of ductile overload j
fracture.
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............ g New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 e
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............ g-New York Power Authority /GPU Nuclear Corp.
April 3,1997
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 3
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........... g_
New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 g :p.
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............ g New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 y
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............ g_
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 O
Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 l
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April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O
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............ gj-
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New York Power Authority /GPU Nuclear Corp.
April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 e
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.,.q O Lucius Pitkin _gj-n. New York Power Authority /GPU !Juclear Corp. April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O O ta; .i.* : ':e; y m W '4.c y v m, n' r-+ t ' 4g u p ( , g t[ ' .) q.,. ;;.o . g i e' v:s. m e:. B -4, " c a 4 +c l, , $ ' ' * (; .f .s. k '! 4.^. n($,.t,.'r n., ' s.
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) Lucius Pitkin _gj- ~~ New York Power Authority /GPU Nuclear Corp, April 3,1997 Attn.: Messrs. W. H. Spataro & A, Collado M96244 ) ) m
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J Lucius Pitkin ............ g_ ~~ New York Power Authority /GPU Nuclear Corp. April 3,1997 Attn.: Messrs. W. H Spataro & A. Collado M96244 ) P A3 3 3 3 3 D3 s s s%_ ) Fig. 45 Photomicrographs (200X) showing the roots of threads adjacent to -) the fractures of specimens A3 (XM-19) and D3 (Type 304) in the as-polished condition. Specimen A3 exhibits a blunted ductile tear, whereas specimen D3 exhibits branched and oxide filled secondary cracks, typical of stress corrosion cracking.
Lucius Pitkin .........e.. " ' ~ ' New York Power Authority /GPU Nuclear Corp. April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 , APPENDIX A Specification For Crevice Corrosion Testing For Hot-Rolled XM 19 Material Oyster Creek And James A. Fitzpatrick Nuclear Power Plants SP-1302-52-118
/8L/8ot Nuclear sreciric4 Tion SP-1302-52-118 OUALITY CLASSIFICATION " REGULATORY REQUIRED
- SPECIFICATION FOR CREVICE CORROSION TESTING
_FOR HOT-ROLLED EM-19 MATERIAL OYSTER CREER NUCLEAR GENERATING STATION & JAMES A. FITEPATRICK NUCLEAR POWER PLA*T, CONTAINS GPUN PROPRIETARY INFORMATION b. DATE //-[-i6 PREPARATION A c /'A. Coll o/S. D. shnoff 'V bb ENGINEERING APPROVAL W DATE f q suu aanw DATE '/ N QA CONCURRENCE 'Ac..&79 REV_ 2 I miosime 1
suu suaa g DOCUMENT NO. r a has M SP-1302-52-118 TTTLE CREVICE CORROSION TESTING FOR HOT Rot. LED XM-19 MATERIAL REV
SUMMARY
OF CHANGE APPROVAL DATE W[1 g-//- 7 L I 1 Deleted propriet >,ry note f rom cover sheet. j Reference 2.3.1 was Revision 0. Deleted 7' Patent Panding f rom 4.2 and added "see y g contract details for proprietary definition'. In paragraph 4.3, 1.250 inches diameter was 3 ' inches. In Paragraph 4.4 300 hours was 7 [ days, added static loading, ground and thermal isolation requirements and testing of one additional specimen at lower strain rate. M* N8# Deleted pricing section 7.0. Revised 5.2.7. NF NMT S.2 /. g Added 6.2.11. 2' Reference 2.3.1 was revision 1. Rmsed 4.2 to spectfy that the 8Pecimens tested in air shall be at room temperature and that f fq[3s[9A heat treatment shall be ramped vs. Step change. Revtse 4.4 to ' + specify water chenustry momtoring and Dow rate requiremcats g,%,y L l Added stram rate requirements for specunens tested in air. This revision captures the "as-built" conditions. ref IM
- f. 39 p m
N0036 (03-90) la i
I SP.!302e52118 Rev.2 May 1996 Crve. Cor. Hot Rol'd XM.19 Page 2 of 9 TABLE OF CONENTS P. ass ~1.0SCOPE......................................................................................................................... 2.0 CO D ES AND STANDARDS........................................................... ..............3 2.1 American Society for Testing and Materials (ASTM).............................................................. 3 2.2 American Natiosal Standards Institute (ANSI)...................................................................... 3 2.3 References............................................................................................................................... 3. 0 G ENERAL REQ UIREMENTS................................................. .. 4 3.1 Work to be Provided by the Contractor............................................................................... 4 3. 2 W o rk t o be Pro vided by Others.......................................................................................... 4.0 D ETAI L E D RE Q UIREMENTS............................................... 4.1 Desc ription of I ntended Use......................................................................
- 4. 2 S pecimen P reparation........................................................................
4.3 Materials............................................................................................................ .6 4.4 Testin 8 and Ev aluation of Tests.................................................................................. 4.5 Cleaning.................................................................................................................................7 4.6 M atking & Identincation & Traceability................................................................................ 5.0 Q U ALITY ASSURANCE.......................................................... .. 7 6.0 INFO RMATION TO BE SUB MITTED.,.................................................................................... 8 6.1WithProposal.............................................................................................................................8 6.2 A fte r Aw ard of Cont ract.......................................................................................... P 014/204 12/2/96
SP-1303052118 Rev.2 May 1996 Crve. Cor. Hot Rol'd XM 19 Page 3 of 9 1.0 SCOPE This speci6 cation covers the requirements for crevice corrosion testag of bot-rolled XM 19 and 304 stainless steel matenals. Hot Rolled XM 19 material was used in Boder Water Reactor (BWR) shroud repaus. The 304 stainless steel testing is for control purposes only. This testag is to comply with the requuements of the U. S. Nuclear Regulatory C-imaion (NRC) for purpose of denonstratmg that hot-rolled XM 19 in the threaded (crevice) condition will perform satisfactordy in a BWR anvuonment. The matenal testing outlined herem addresses Hot Rolled XM 19 in service at the Oyster Creek Nuclear Generating Stanon operated by GPU Nuclear (GPUN), and the James A. Fitzpatrick Nuclear Power Plant, operated by the New York Power Authority (NYPA). 2.0 CODES AND STANDARDS Unless specified herein, the latest revision of the followmg codes and standards shall apply. 2.1 American Society for Testing and Materials (ASTM) 2.1.1 A479-92 Spectfication for Stainless and Heat-Resistant Steel Bar and Shapes for Use in Boders and Other Pressure Vessels. ~ 2.1.2 G49 Standard Practice for Preparation and Use of Direct Tension Stress-Corrosion Test Specimens. 2.1.3 G78 - Standard Guide for Crevice Corrosion Testag ofIran Base and Nickel Base Stamless Alloys in Sea Water and Other Chlonde Aqueous Envuonments. 2.1.4. A380 - Cleanmg and Descaling stainless Steel Parts, Fminment and Systems. 2.2 American National Standards Institute (ANSI) 2.2.1 N45.2.1 Cleaning Requaements - 1973 edition 2.3 References' ,2.2$' GPUN Drawing 3B-SKM 722, Rev. 2, Tension Test Specimen for Crevice 2 3./ Corrosion Testmg 014/204 12/2/%
l l SP 1302052118 Rev.2 .. May 1996 Crve. Cor, Hot Rol'd XM-19 Page 4 of 9 3.0- . GENERAL REQUIREMENTS , 3.1 - Work to be Provided by the Contractor a) Development of test procedwes. b) Preparation of test specunens, c)- Fie,4 test thedmes, toohns, ibawes and testag apparatus, d) Performance of tests. e) Evaluation and Reportug of test results in report form. f) Technical support (as required) in discussions and subesttals to the NRC. 3.2 Work to be Provided by Others a) Sufficient bar stock of test matenals (XM 19 HR and 304 SS). B) Providag tout certified materials W of the supplied matenal as required. c) Field verificanon of testag and source inspectica as requued. 4.0 DETAILED REQUIREMENTS 4.1 - - Description ofIntended Use - De matenal testas program required by this.A is to demonstrate the resistance ofXM-19 matenals to latergranular Stress Conosion Cracking (IGSCC) under creviced BWR environmental conditions. The crevice has been cmated into the design of reactor sternals repair in the form of machine-threadedjoints. 4.2 Specimen Preparation The specimen design shall be such that the established crevice is not lost dunni the testas process. Reference 2.3.1 shows an acceptable type specimen wrapper design. He Contractor may provide alternate specunen type subject to GPUN and NYPA approval. De specimen shall be cut (thermal cutting is prohibited) and machiM from the supplied-bars. 014/204 12/2/96 o
SP.!302.$2118 Rev.2 May 1996 Crve. Cor. Hot Rol'd XM 19 Pass 5 of 9 All Saal n=NW surthoes, including threads, and grooms shall be accomplisind with sharp tools and cutters that have not been prmously used on carbon steel. Ught sanding with approwd abrasiw cloth may be used to obtain surthes Snish supdrements. Final l machining pass / passes shall be linuted to.010" metal removal, nreads made by rollies are not permmed. I All abrasiw cleaning shall be performed with abrasiw materials that are not contaminated with residues other than austenitic stainisse and nickel chresne alloys. Cutting tools shall be property cooled with sufBeient Suid to prevent c ' N of tools and specimes insterials. Cutting lubriennis, tools, testing apparatus,6xtures, workers and other equipensat used la contact with the spooness on the testas Guid shall be controlled as to their contaminant levels. De following limits apply t3 thoes consumable products which contact the test specimen. CONTAMINANT LIMITS Zinc, its allon and/or w 4 50 m man. Total henw metals: 200 ppm, max. (Pb, Hs, Cd, Cu, their alloys and/or comoounds) Totalleachabis halogens: 100 ppm, max. (Cl., Br., F., etc.) Total halogens: 1,000_ ppm, max. (Cl., Br, F., etc.) Total sulfkr and its we 1,000 m max. When surbes are found to be contaminated by contact with product not meeting the abow limits, the surfaces hall be cleaned in accordance with a NYPA and GPLN - approved Contractor procedure. no Contractor and sub-tier wndors must use methods to prevent contamination of test matenais. De practices of ASTM A 380 shall be followed. Grinding or centerless grinding, incidental welding, or repair welding of the test specimen is not permitted. A total of three heats of XM 19 matenal require testing. Two specimens of each heat shall be tested in BWR coolant environment and one specimen per each heat shall be tested in air at room temperature as a control specimen. In addman, oce specimen of sensitized Tne 304 stamless steel shall be tested in the test anytroamem as e control to assure adequacy of the test environment to produce IOSCC. 014/204 12/2/96 d
SP 1302 521is Rev.2 May 1996 Crve. Cor. Hot Rol'd XM 19 Page 6 of 9 Sensitization of the Type 304 stainless steel shall be accomplished by heat treatment at 1250 *10'F for one hour, followed by furnace cool. Heat treatment shall be ramped versus step change. 4.3 Materials The XM 19 materials were purchased to ASTM A479, bot rolled with special corrosion
- test A262 Practice A or E.
The XM 19 as supplied by 'Others" has been previously subjected to a load conditioning at 80 Kai stress. l The 304 stainless steel materials were purchased to ASTM A479 with special corrosion l test A262 Praaice E. The 304 semialana steel was not load cc 4tioned. This material is in l the solution annaaled condition per ASTM A479. l l The XM 19 samples for specimen making will be supplied in one piece,1.5 inch diameter s by appmximately 6 inches long for each heat of material, lhe 304 stainless samples for the control specimen makmg will be supplied in one piece, l approxunately 1.250 inches diameter by 3 inches long. Other matenals are shown in Reference 2.3.1. 4.4 Testing and Evaluation of Tests lhe test medium shall be simulated BWR reactor coolant at 550'F *10'F with a 8 to 10 ppm oxygen. Ca*=mia=* levels shall be contmtled to maintain conductisity of the test medium in the range of.4.5 pS/Cm using sulfate addition. Water chmi*y must be monitored and recorded, once a week for the first month of the testmg (at the feedwater and effluent) and once e month thereaAer for the longer dumtion test to capture the equilibrium of bulk water conditions, in addition to oxygen, chlorides and sulfates shall be measured. Water pH range shall be 6.0 to 7.0, corrected to 25'C. Flow rate to be kept as slow as reasonable achievable while maintaining the required oxygen levels in order to simulate the oxygen differential in the crnice. Flow rate to be measured in A/sec. across the specimen surface. Test acceleration shall be accomplished by subjecting the specimen to slow strain rate 4 testmg at a low strain rate of 5x10 sec until failure. One additional spar % from whichever heat has Lafficient material, shall be subjected to slow strain at an 4 even lower strain rate of 5x10 sec until failure as a confirmation that the 5 strain rate accurately detects the potential failure mechanism. Elongation vs Time shall be recorded for all test specimens. Prior to strauung, the specimens shall be preconditioned for 300 hours in the elevated temperature test ensironment and statically loaded to 90% of yield stress for material during the preconditioning period. 014/104 12/2/96
SP.1302 52118 Rev.2 _ May 1996 Crve. Cor. Hot Rol'd XM 19 Page 7 of 9 Strain rate for the specimens tested in air shnu be in accordance with ASME range for tensile tests.1Amor bound rate shall be used. The strain rate achieved shnu be sta the Snalisport. Specimens shan be elecencally grounded and thermauy isolated 6am the environment. l Folkming the test, specimens shall be osamined using conventional ligk microscopy and sonaning electron miemoopy, n specimens win be -m fbr 1-im f stra o l corrosion cracking on the hacews surfbes and along the snues section. A minimum of two metallographic mounts will be evaluased ibr each specimen. Contractor to supply wntion evaluanon, graphs and photographs of test results. 4.5 Cleaning De comple sd specimens shau be cleaned per ANSI N45.2.1, Class B Cla==li prior to a insertion into the autoclaves. 4.6 Marking & Identincation & Traceabluty All XM 19 and 304 specimens shall be M Aarl and traced to their unique heat number for the material. Au markings shall be done with low stress vibrating tools. Marking in the specimen sage U "8' PomM Au specimens shnu be tracable to their bar stock whm the specimen wem out imm. De Contractor shall be responsible for maintaining a Quahty Control System that would preclude the mixing of ddrrent materials. 5.0 QUALITY ASSURANCE All work is classiBed as " Regulatory Required." De Supplier Quality Classification List requirements apply,
- Work shall be done under a Quality Assurance program which meets 10CFR50 Appendix B or an approwd program by GPUN and NYPA.
All work shall be performed in accordance with standard laboratory techniques. Calibration of mstruments shall be as requued by the National Bureau of St-lards / National Institute Standards and Technology. All fabrication of specimen and crmice corrosion testing activities shall be controlled by the Contractor. 014/204 12/2/96 A
SP.1302 52.lls Rev.2 May 1996 Crve. Cor. Hot Rol'd XM.19 Page 8 of 9 GPUN shall be noti 6ed of any non conformance to this specification or to any contractor requinments approwd by GPUN and NYPA. 4 GPUN and NYPA, or their agents, shall how complete access to the Contractor's and Subcontractor's facilities to' witness work and inspection in progress. GPUN and NYPA ahall have complete access to all data and records related to the fabrication, testing, and inspection of the spec e ns. l 6.0 INFORMATION TO BE SUBMITTED 6.1 With Proposal 6.1.1 Description of testmg facility. 6.1.2 Reference history list of sinular work p.fvi.. d. 6.1.3 Exceptions and/or alternatives to the requirement of this specification. 6.1.4 Schedule to perform the testing fran Purchase Order placemet to issuance of final report. 6.1.5 Certifications of qualified personnel to perform tests and evaluate test results. 6.1.6 Quality Assurance Program for information 6.2 After Award of Contract 6.2.1 Detail testmg procedun and test loop diagram, for approval prior to start of testms. 6.2.2 Records of all testmg activities. 6.2.3 Inspection Records on all test specimen materials. 6.2.4 All calibration records for calibration instruments used during the performance of the contract. 6.2.5 Specimen detailed drawings for appront prior to start of fabrication. 6.2.6 Method of extracting / cutting material for specimen making. 6.2.7 Technical evaluation of testmg results. l 6.2.8 Photographs of testing apparatus. 014/204 12/2/96
SP.1302 52.llI Rev.2 May 1996 Crve. Cor. Hot Rol'd XM 19 Page 9 of 9 6.2.9 Certi8catice of b6 oratory Teha' i== for informatica prior to start of testing. w 6.2.10 All documentation ofmaterials tested (i.e., Material Test Reports). l 6.2.11 DraA and Final Technical Rpts (four copies for OPUN and four copies for NYPA). l E e 9. I O e 014/204 12/2/96 o
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D Lucius Pitkin ............ g_ New York Power Authority /GPU Nuclear Corp, April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 3 ? J APPENDIX B i J Time Temperature Sensitization Heat Treatment Record For Type 304 Stainless Steel i J J J J J
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- O Lucius Pitkin
~~ New York Power Authority /GPU Nuclear Corp. April 3,1997 Attn.: Messrs. W. H. Spataro & A Collado M96244 !O O O APPENDIX C. O-Air Test Load strain curves O O O O O
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O Lucius Pitkin ............ g_ "~ New York Power Authority /GPU Nuclear Corp. April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O 0 0 l APPENDIX D Dissolved Oxygen Concentration And 3 pH Curves Used For The Simulated Coolant O i 3 9 4
D flabanL & Wilcun
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...i. RDD:97:l7567 00.1000:01 l' age 69 D D Dissolved Oxygen in Air Saturated Water (20.95% oxygen) l D to l \\ \\ A. 14 N ~ ,.. { 1 n h ~_C ~ ~~ ~~ 6 6 E to N' 6 w s d
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-.g ~"~ m' ^ ~ ~ m _N, "~ 6 30 50 70 90 110 130 150 Teriperature (F) 3 0 oz/sq in gage + 8 or/sq in. gage e-9 oz/sq vt gage 10 oz/sq vt gage + 11 oz/sq vt gage + 16 ot/sq in. gage D D 1 J
O Babcock & Il'ilcox asue aue-, RDD:97:47367 003 000:0) Page 7) O O pH versus Specific Conductivity for Sulfate / Bisulfate Solution i h 69 l -I5 l 5 67 5 g N } g E 's, 66 7 _, 2 ( %%'N l g 63 i 01 02 03 04 06 06 07 Specfc ConductNay (us/crr$ @ 25 C l Asomes a Sodium Sulfate to Socium Biutate Molar Ratio of 3 6.1 l b O e e
O Lucius Pitkin ........... g_ New York Power Authority /GPU Nuclear Corp. April 3,1997 Attn.: Messrs. W. H. Spataro & A. Collado M96244 O O O A PPENDIX E 1 lg Test Parameter Versus Time Plots For l The Simulated BWR Coolant Constant Extension Rate Tests O O O O O
O Babcock & Il'ilcox a u t w,.ar a..e,~, RDD.97:47367 003 0(10.01 O ) o5 '4 - 04 1 o3 y I h C o2 e 01 i i 300 320 340 300 380 TETTFE(FfE) O 3 e Ram 11ead Displacement Rate n Time for the First SATEC Test [ Strain Rate = 5 x 10'sec ') e
l 'O Babcock & IYilcox anim en w - RDD:97:47367 003 000:01 'O i I i l0 tes Il - 500 10 ~ E Ie 55 lO l i ~ 545 lO 1-90 i i 0-100 200 300. 400 muE(HE) 1 0 l l i-O- C-p Autoclave Temperature n Time for the First SATEC Test [ Strain Rate = 5 x 104 sec] - O. I k ..-,-_.e.,,_,,,_n,- ,,,n. y-e,-w.mn e,.w,n,-, ,g-y 7, ,,,-,.n.,,- .,,,,e .v n -,:..,,--a,+
Babcock & ll'ilcox atuenut. w, RDD:97:47367 003 000:0) J 3 1C00 D ivo 1500 ) 1450 J i@ t 1 i 0 100 200 300 400 TE>TTfE(HE) O O O Autoclave Pressure ys Time for the First SATEC Test [ Strain Rate = 5 x 10'sec '] O
O Babcock & It'ilcox a u t w.ar un,.n RDD:97:47367-003-000:01 l O 1 0 05 f04 O 8 s 5 1 Y h f l l g O2 O O Q1 i i n 2 m a a TEST TIME (mS) O O O Ram llead Displacement Rate n Time for the Second SATEC Test [ Strain Rate = 5 x 104 sec'] O (i
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O nabcock & Inicox ,, s u e...<<...e RDD:97:47567 003 000:01 O O - ~. l 66 5 l O I i I sao i G i 'O $ss i I Il~%gWf%$f#$W4M o 645 f i O !j i 64 0 - 1 0 200 400 600 000 1000 j TEST TIME (HRS.) m__.-_- .. ___._ j 3 e Autoclave Temperature n Time for the CERT Test on Specimen A 4 [ Strain Rate = 5 x 10'sec '] 9
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