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Home My WebLink About851173.tiff `EpR REGL,4 UNITED STATES O L9� NUCLEAR REGULATORY COMMISSION F REGION IV °r-� 4 y" 611 RYAN PLAZA DRIVE, SUITE 1000 'I'7 aG ARLINGTON,TEXAS 76011 AUG 2 0 1985 Docket: 50-267 NtI1! C!;fl q Public Service Company of Colorado AUo 2 a 1985 ATTN: 0. R. Lee, Vice President Electric Production P. 0. Box 840 GR€ELEY COLO. Denver, Colorado 80201-0840 Gentlemen: Our consultants at the Los Alamos National Laboratory have completed their review of three tasks for which we had requested their assistance. The results of their review are contained in the enclosed reports: 1. Review of Fuel Rod Metallography on Fort St. Vrain Fuel Element SN 1-2415 2. Post Irradiation Examination of Fort St. Vrain Cracked Fuel Element SN-1-2415. 3. Evaluation of Concrete Creep and Stress Redistribution in the Fort St. Vrain PCRV Following Rupture of Prestressing Tendons. We are providing you these reports for your information and/or comment. Since no specific reporting is required, OMB clearance is not required under P.L. 96-511. If you have any questions on this matter, please contact me. Sincerely, Philip C. Wagner Senior Project Manager Enclosures: As stated cc: See next page 851173 8 2 cc: Mr. D. W. Warembourg, Manager Nuclear Engineering Division Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 Mr. David Alberstein, 14/159A GA Technologies, Inc. P. 0. Box 85608 San Diego, California 92138 Kelley, Stansfield & O'Donnell Public Service Company Building 550 15th Street, Room 900 Denver, Colorado 80202 ! Chairman, Board of County Comm. of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1860 Lincoln Street Denver, Colorado 80203 Mr. H. L. Brey, Manager Nuclear Licensing/Fuels Div. Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 J. W. Gahm, Manager, Nuclear Production Division Fort St. Vrain Nuclear Station 16805 WCR 19f Platteville, Colorado 80651 L. Singleton, Manager, Quality Assurance Division (same address) Colorado Radiation Control Program Director Enclosure 1 Review of Fuel Rod Metallography on Fort St. Vrain Fuel Element SN 1-2415 NRC Fin No. A-7258 July 31 , 1984 Los Alamos National Laboratory William A. Ranken, Q-13 Robert D. Reiswig, MST-6 Responsible NRC Individual and Division J. Miller/ORB3 Prepared for the U.S. Nuclear Regulatory Commission Washington, D. C. 20555 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights. - 2 - ABSTRACT This report reviews the documentation submitted to the NRC by the Public Service Co. of Colorado, on the metallographic examination of a section of fuel stack 308, as found in cracked fuel element SN 1-2415 removed during the third refueling of the Fort St. Yrain HTGR. Los Alamos essentially concurs with the PSCo conclusions, in that the performance of the fuel during irradia- tion was acceptable, that no evidence of fuel and graphite element interaction has been found, and that no kernel migration was observed. The report provides comments to the NRC as to whether the licensee's technical information is cor- rect and consistent with the information gained by Los Alamos in previous fuel studies, and in their review of graphite slabs from Segment 2 cracked fuel element SN 1-2415. FORWARD This technical evaluation report is part of the technical assistance pro- gram, "Review of Selected Fort St. Yrain Issues", FIN No. A-7258, and is sup- plied to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, by Los Alamos National Laboratory. - 3 - Review of Fuel Rod Metallography on Fort St. Vrain Fuel Element SN 1-2415 Technical Evaluation Report Introduction During the Segment 2 reload of the Fort St. Vrain High Temperature Gas- Cooled Reactor, two colinear graphite fuel elements were found cracked axially down the B-face, and radially inward two and three webs deep. This report reviews the metallographic examination performed by GA Technologies on a fuel rod from a stack adjacent to the crack region of the more severely cracked fuel element. Background The GA Technologies report entitled "Metallographic Examination of a Fuel Rod from Segment 2 FSV Fuel Element 1-2415" (Ref. 1 ) describes the metallo- graphic examination of a fuel rod from the stack in fuel hole 308 of Fort St. Vrain fuel element SN 1-2415, which developed a crack during irradiation. Stack 308 was centered in the second row of fuel and coolant channels on the B-face side of the hexagonal graphite element, and was adjacent to the crack (which actually passed through fuel hole 307). The metallographic examination was conducted to see if any evidence of unusual thermal effects could be seen that would help in ascertaining the cause of the crack, and was performed in conjunction with a more general postirradiation examination on the fuel ele- ment. Each graphite fuel element is a hexagonal right prism, with 210 fuel holes and 108 coolant channels drilled parallel to one another in a triangular array with 0.74 inch pitch spacing, resulting in a basic ratio of two fuel holes for each coolant channel . Each fuel hole is drilled from the top surface of the element, to within about 0.3 inches of the bottom. Bonded rods containing coated fuel particles, each about 1/2 inch long, are stacked within the hole, and the hole is capped with a graphite plug. The fuel is in the form of carbide particles, coated with a highly reten- tive coating, and bonded into fuel rods with a coal tar pitch binder2. - 4 - The fuel rods contain a homogeneous mixture of fissile particles containing both uranium235 (93.15% enriched) and thorium, and fertile particles con- taining only thorium. As schematically shown in Figure 1 , the fuel particles are coated with a four-layer TRISO coating, sequentially consisting of an inner layer of porous pyrolytic carbon called the buffer layer; a high density iso- tropic pyrocarbon ( IPyC layer) ; a thin layer of SiC, which is highly impervious to metallic fission products; and an outermost layer of strong high density isotropic pyrocarbon (OPyC layer). Important particle parameters are repro- duced from the FSAR in Table 1 . Discussion Our general evaluation of the metallographic examination of this fuel rod, and the coated fission and fertile particles it contained, is that it appears to have been done competently. The main criticism relates to the selection of the examined fuel rod from a stack (308) adjacent to the stack in fuel hole 307, through which the crack in the fuel element passed. Presumably this se- lection was made because the thermal history of stack 307 might have been dis- torted by the crack and hence would be less representative than stack 308 of events that precipitated the cracking. Nevertheless, it would have been more informative if the fuel rods from 307 had been measured for dimensional changes and one of them chosen for metallographic examination in addition to the rod actually selected. Additional examination would have helped to rule out fuel nonuniformity effects as a cause for crack formation. The performance of the fuel was found acceptable in Ref 1 . Specific ob- servations and conclusions from the document are summarized as follows: 1 . Fuel rod 13 from stack 308 was in good condition although minor cracking in the matrix end caps and some debonding of particles from the rod sur- face were observed. ' 2. There was no fuel rod-block interaction as evidenced by the visual exami- nation of the rods, by the small pushout force of the stack, and by the small amount of debris collected from the emptied block. 3. The measured macroporosity for rod 13 was 17.5%, which was within the (14-29%) range of macroporosities observed for fuel rods from capsule F-30. - 5 - 4. A total of 231 fissile and 184 fertile particles from rod 13 were examined. For the (Th,U)C2 and ThC2 particles, respectively, the OPyC coating failure was 0.4% and 7.6%, and the SiC coating failure was 0.9% and 3.8%. However, these coating failure rates should be regarded as the upper limits since coating failures can be caused during manufacture and during the grinding and polishing procedure, as well as during irradiation. 5. The chemical behavior of the particles was acceptable. There was no kernel migration observed. However, there was evidence of fission product inter- action with the SiC coating. 3.9% of the fissile particles and 3.2% of the fertile particles showed fission product--SiC interaction with a pene- tration depth of % 5 pm, which was higher than the expected value of < 1 um. The increased penetration may have been caused by fuel dis- persion resulting from chlorine trapped in the buffer layer during the SiC coating and/or by higher operating temperatures than the average calculated for the element. 6. Fuel dispersion, which is attributed to chlorine diffusing through a low density IPyC into the buffer layer during the SiC coating process, and IPyC debonding were observed in some of the TRISO (Th,U)C2 and ThC2 particles. Fuel dispersion and IPyC debonding did not detrimentally affect the performance of the particles. In general the conclusions reached in the report are reasonable, but there is a need for caution. Visual examination, the ease with which the rods could be pushed out of the block, and the porosity measurements all confirm normal behavior for fuel stack 308. However, the fuel dispersion results, the amount of fission product interaction with the SiC layers, the inner pyrolytic carbon coating debonding (shrinkage of the pyrocarbon), the observed coating failure rates, and the lack of data on total coating failure (because of hydrolysis of the metallographic specimen mount') all raise questions. As stated in the report, one interpretation of the depth of fission product interaction (- 5 um versus < 1 pm expected) is that the operating tem- perature was higher than-the average calculated for this fuel element. Higher than expected operating temperature could presumably also explain the greater than usual amount of fuel dispersion observed in the sectioned fuel rod. Tem- perature may also have been a factor in the increased irradiation - induced shrinkage of the inner pyrolytic coatings that led to debonding. - 6 - As to questions of whether broken coatings were caused by metallographic sample preparation or by irradiation,.it would have been helpful if the GA investigators had cited past experience on sample preparation and used this to make a determination of the most probable cause of the observed cracking. The lack of total coating failure data is indeed unfortufate. It is not known whether the number of failures was a very small fraction of total par- ticles or a significant added population to that of particles with just one of the three containment layers cracked. Conclusions Despite the uncertainties in interpretation of some of the fuel rod metal - lographic data, there is little evidence that the thermal history was abnormal . The observed departures from the norm, particularly in fuel dispersion and fission product interaction, may indeed have been caused by variations in coated particle fabrication as suggested by GA. If the particles were uniform in their properties, one would expect a temperature effect to show up in all of them instead of just a fraction. On the other hand, the variation of par- ticle properties could mask a temperature effect, because one would not be sure, for instance, how much of the increased fuel dispersion was caused by chlorine and how much by a higher than expected temperature. It appears that fuel rod metallography is not, under the current conditions, a very sensitive determinant of fuel and thermal history. In any case the GA promise to monitor the fuel dispersion and fission product - SiC interaction in future FSV fuel surveillance should be seized upon. If any future work is done on fuel element 1 -2415, it should include dimensional checks and metallography on the fuel rods from stack 307. References _ 1 . "Metallographic Examination of a Fuel. Rod from Segment 2 FSV Fuel Element 1-2415", GA Technologies Doc. No. 906968, 1983. 2. Fort St. Vrain Nuclear Generating Station, "Updated Final Safety Analysis Report", Public Service Co. of Colorado. - 7 - MODEL TRISO COATED PARTICLE OUTER ISOTROPIC PYROLYTIC CARBON t'., ' ••441 A ,�` FUEL PARTICLE \ SILICON CARBIDE I 1/ �.+y: -T'` ; 1I • y t nr� : . n1F'�'� BARRIER COATING �I '�j�•�(.;,,$t'7 Y•<' t)�� i . 3 ,•'' . tEr4 i INNER ISOTROPIC FUEL PARTICLE I I PYROLYTIC CARBON `.� eniserar � BUFFER :, '• ._ ,�' / PYROLYTIC CARBON i Figure 1 . Model of TRISO Coated Fuel Particle Table 1 . Coated Fuel Particle Parameters Parameter Fissile Fertile Th:U 3.6:1,4.25:1 All Th Kernel composition (Th:U)C: ThC: Small Large Small Large Average fuel kernel diameter, micron 140 225 375 525 Average coating thickness: Buffer carbon layer, micron 50 50 50 50 Isotropic carbon layer, micron 20 20 20 20 SiC layer, micron 20 20 20 20 Isotropic carbon ' layer, micron 30 40 40 50 Average coated fuel _ diameter, micron 3B0 485 635 805 inches (0115) (.019) (.025) (.032) LA-UR -84-3503 Enclosure 2 Los Alamos Natrona' Laboratory is operateC by the University o1 CaMornla tar the United States DapenrnMt of Energy under Contract W-7405-ENG-36 TITLE POST IRRADIATION EXAMINATION OF FORT ST. VRAIN CRACKED FUEL ELEMENT SN1-2415 AUTHORS) Deborah R. Bennett Keith E. Dowler Jay H. Cook Robert D. Reiswig SUBMITTED TO US Nuclear Regulatory Commission _ By acceptance 01 this arncte the publisher recognizes that the U S Government retains a nonexclusive:royalty-tree license to publish Or reproduce the publilhed 102m 01 this contribution or to allow others to 00 so, for U.S Government purposes Inc Los Alamos National Laboratory requests that me publisher identify this article as work performed under tee auspices of the U S Department Of Energy 0 S Q Los Alamos National Laboratory _ 4 Los Alamos,New Mexico 87545 tanN NO 136 Ax St NO 3625 stet Post Irradiation Examination of Fort St. Vrain Cracked Fuel Element SN 1 -2415 NRC Fin No. A-7258 October, 1984 Los Alamos National Laboratory Deborah R. Bennett, Q-13 Keith E. Dowler, MST-14 Jay H. Cook, MST-14 Robert D. Reiswig, MST-6 t .. Responsible NRC Individual and Division J. Miller/ORB3 Prepared for the U.S. Nuclear Regulatory Commission Washington, D. C. 20555 DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of _ theiremployees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights. ABSTRACT This report reviews the post-irradiation examination performed by Los Alamos National Laboratory on three slabs from the Fort St. Vrain cracked fuel element SN 1-2415. The slab cracks were photographically documented, and radiological , fractographic and metallurgical examinations were performed. The results indicate that the element graphite microstructure shows no un- expected features, and that the cracks found in element SN 1-2415 are typical of cracks found in comparable artificial graphites. FORWARD This technical evaluation report is part of the technical assistance program, "Review of Selected Fort St. Vrain Issues", FIN No. A-7258, and is supplied to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, by Los Alamos National Laboratory. Post Irradiation Examination of Fort St. Vrain Cracked Fuel Element SN 1-2415 Technical Evaluation Report Los Alamos National Laboratory NRC Fin No. A-7258 October, 1984 1 .0 Summary Two adjacent graphite fuel elements removed from a common column of the Fort St. Vrain HTGR (High Temperature Gas-cooled Reactor) during the Segment 2 reload were found cracked axially along the 8-face of each element, and cracked radially two and three webs toward the center of each element. A post- irradiation examination (PIE) was performed at Los Alamos National Laboratory on three slabs from the element with the wider crack, fuel element SN 1-2415. The purpose of the examination was to investigate the crack microstructure, and to apply the results to a parallel analytical investigation of the crack- ing. The slab cracks were photographically documented, and radiological , fractographic and metallographic examinations were performed. The results indicate that the fuel element graphite microstructure has no unexpected fea- tures, and that the cracks found in element SN 1-2415 are typical of cracks found in comparable artificial graphites. This report describes the examina- tions performed, and the resultant conclusions based on the PIE findings. The parallel analytical evaluation of the crack mechanism will incorporate the results and conclusions of the PIE evaluation, and will be available at a later date. 2.0 Background 2.1 Core Position and Fuel Element Geometry The Fort St. Vrain core is composed of 37 regions, each region containing 7 columns of hexagonal reflector and fuel elements, arranged so that the center column, which houses the control rod pair for the region, is surrounded by the remaining six columns, Figure 1 . Each column is a stack of six active fuel elements, bounded by upper and lower reflector elements. Each fuel element, Figure 2, is a hexagonal right prism, standing about 30 inches tall , and measuring approximately 14 inches from flat to flat. Each face of the element has a letter designation, from A to F; face E on each of the six outer columns in a region is oriented to the region's center column. On the upper axial end of each element, a set of three dowels interlocks with a set of three dowel sockets on the bottom end of the next element. The dowel designation corresponds to the element face notation. Each graphite element has 210 fuel holes and 108 coolant channels drilled parallel to one another in a triangular array with 0.74-inch-pitch spacing, resulting in a basic ratio of two fuel holes for each coolant channel . All coolant channels and fuel holes are numerically designated as in the Fort St. Vrain Final Safety Analysis Report (FSAR)5 and shown in Figure 3. The cool - ant channels are 0.625 inches in diameter, and are axially aligned with the channels in each element. Each fuel hole, 0.500 inches in diameter, is drilled from the top surface of the element, to within about 0.3 inches of the bottom. Bonded rods containing coated-fuel particles, each about 1 /2 inch long, are stacked within the hole, and the hole is capped with a graphite plug. The fuel rods contain a homogeneous mixture of fissile particles containing both uranium235 (93.15% enriched) and thorium, and fertile particles containing only thorium. The carbide fuel particles are coated with a four-layer TRISO coating, and bonded into fuel rods with a coal tar pitch binder2. -5- 2:2 The Graphite Material The element material is H-327 graphite, a petroleum-based, artificial graphite that is extruded into logs. The anisotropic graphite has coal tar pitch acting as the binder for the coke filler particles. The material is a so-called "needle coke", because the coke filler particles are relatively long and thin, and highly ordered. During the extrusion process, the acicular coke filler particles are preferentially aligned with their long axes parallel to the longitudinal (extrusion) axis, producing the observed anisotropy in the graphite logs. The general porosity observed in the graphite microstructure is an artifact of the production process, formed during baking and graphitization, as the pitch binder is pyrolized--i .e. the binder shrinks as volatiles (hydrogen, oxygen, carbon, and sulfur in various forms) are driven off, and the pores grow in size. 2.3 Background of Element SN 1-2415 Following the irradiation and removal of Core Segment 2 from the Fort St. Vrain (FSV) core in September, 1981 , 54 fuel and reflector elements were sub- jected to a nondestructive examination1 at the FSV hot service facility in April , 1982. During that examination, two of the fuel elements from Region 8, Column 5, at adjacent Levels 6 and 7, were identified as being cracked. In both cases, the cracks were located in the center of the B-face and appeared to run the full length of the element (in Region 8, Column 5, the B face is oriented towards the core center). The crack extended radially to coolant channel 319, and projected on to some undetermined position under the B-face dowel . A review of the pre-irradiation inspection reports indicated that neither element had been cracked prior to insertion into the core, and there was no record of damage during handling. It was therefore assumed that the cracks had developed during irradiation. In the spring of 1983, the two cracked elements and three other elements were sent to GA Technologies (GAT) for more detailed visual examination. The visual examinations2,3 showed that the cracks were the only apparent damage to the elements, and a stacking demonstration revealed no abnormal interaction between the dowel and sockets of the two elements when stacked in their -6- in-core configuration. Dimensional measurements, calculated temperature and fluence levels, and gross gamma activity for the two elements are listed in Table 1 as reference, and as reproduced from Ref. 4. TABLE 1 Cracked Fuel Element Characteristics4 Item SN 1 -2415 SN 1-0172 Element Type Fuel Fuel Core Location 08.05.F.06 08.05.F.07 Top Crack Width, mm 0.20-0.25 0.13-0.15 Bottom crack width, mm 0.28-0.30 0.05-0.08 Temperature, deg C 650 700 (Time/volume avg) Fluence, x1025 n/me 1 .55 1 .28 (E < 29 f3 )HTGR Because the presence of the B-dowel obscured the extent of the radial cracking, it was decided that the element with the more pronounced crack, SN 1-2415, would be subjected to destructive PIE. The PIE plan included Los Alamos National Laboratory, who would perform radiographic, metallographic and fractographic analyses on three one-inch slabs in verifying crack and graphite microstructure; GAT would use the remainder of the element in acquiring mate- rial property data4. The second element, SN 1-0172, would be stored, and analyzed only if required. Initially, the extent of cracking in SN 1-2415 was confirmed by removing approximately a 1 /4" slice from the top of the element, thereby effectively removing the cemented B-dowel from the element socket, without interfering with the fuel stacks. Photo 1 shows the top of SN 1 -2415 after the removal of the first slice with the remainder of the B-dowel still in the socket. In this photo, the crack radially extends from the B face, across coolant channel 319, and disappears under the B-dowel . Photo 2, taken after the B-dowel section was removed, shows how the crack extended from coolant channel 319 to fuel channel 307, and on to coolant channel 295 under the B-dowel . The element was further sectioned to remove the fuel stacks, and then three one-inch slabs were removed from the remainder of the element, according to Fig. 4, for shipment to Los Alamos. -7- 3.0 Los Alamos Examinations 3.1 Visual Examinations The three slabs cut from element SN 1 -2415 were received at Los Alamos National Laboratory in August, 1983. The three slabs were labeled as Slab 1 , 2 and 4, corresponding to the original location in the element. The slabs were in good condition, except that the striation patterns from the bandsaw cutting at GAT were apparent on both faces of each slab, and made verification of the crack location and extent more difficult. A broad array of general and detailed photographs was taken of the three cracked regions, and provided a permanent record of the crack characterisitics on each slab prior to sectioning for other examinations. Photo 3. shows Slab 1 , as seen in the Los Alamos hot cell facility, mounted on the rotatable jig used for positioning the slabs. Photographs 4 a, b and c show the crack defi- nitely running from the B-face surface to coolant channel 319, over to fuel hole 307, and appatently traversing the graphite web over to coolant channel 295, for Slabs 1 , 2 and 4, respectively. It was noted that the crack width seemed to narrow for each successive slab approaching the core midplane. A tabulation of the crack widths is available in Table 2. TABLE 2 Crack Width Measurements Approximate Crack Widths, inches x 10-3 Slab B-face to 319 319 to 307 307 to 295 1 10-15 10-15 10-15 2 5-7 4-6 4-6 4 3-5 2-4 *Crack width not measurable by visual examination Measurements were taken of selected fuel and coolant channels on both the B-face side and the opposite side of the element, to determine channel post- irradiation roundness. Generally very little ellipticity was found in the measured holes around the B-dowel region, or in channels across the slab, except for the holes with cracks. In all three slabs, the cracked holes were -8- ellipsoidal with the major axis parallel to the B-face, with a maximum 0.005 inches variation between the major and minor diameters. Measurement values are available in Appendix A. 3.2 Radiological Measurements Radiological measurements were taken at the time of receipt, during which all three slabs indicated nominal radioactivity levels of 1 .5 mr/hr, one meter from the slab faces. Edge surface measurements averaged 72 mr/hr, 50 mr/hr and 45 mr/hr, for slabs 1 , 2, and 4, respectively, with a 40-90 mr/hr varia- tion. Average face surface measurements were 445, 232, and 212 mr/hr, again for slabs 1 , 2 and 4, respectively, with measurements ranging from 150-700 mrem/hr. Figures 5a, 5b and 5c indicate the relative radioactivity levels over each of the slabs, respectively. In general , the radioactivity levels of the three slabs were considered less than originally expected, allowing easier handling in the hot-cell environment. Radiological information about the coolant and fuel channel surface areas was acquired with filter paper swipe samples, which were gamma analyzed using a Ge(Li ) detector and a 4096 channel pulse height analyzer. A summary of the radioactive products found on the various surfaces is listed in Table 3. Fis- sion products of U233 result in beta and gamma emitters such as Co144, Csl34, Csl37, Mn54 and Co60. The alpha emissions are from the thorium daughter prod- ucts, but are at a negligible level . All radiological levels documented are well within anticipated ranges, and indicate that no unusual fission product migration to the coolant or fuel channel walls had occurred. TABLE 3 Fission Product Activity Levels* Location Cel 44 Csl 34 Csl 37 Mn54 Co60 Fuel Hole 307 0.1050 0.0155 0.0525 *** 0.0372 Channel 319 0.0218 0.0026 - 0.0123 0.0061 0.1980 B-Face 0.1290 0.0248 0.0805 0.0267 0.8250 Top Surface 0.1460 0.0306 0.1 01 0 0.0042 0.2550 Bottom Surface 0.0438 0.0112 0.0311 0.0084 0.3960 * Activity levels measured in microcuries -9- Autoradiographic techniques, where an x-ray type film is laid on the graphite surface and exposed by the resident radiation of the material , gener- ally showed a tendency for relatively higher radiological levels around fuel holes 307 and 308. However, this qualitative technique also shows other areas relatively hotter, and the higher radiological levels are most likely artifacts of the slab cutting process, rather than an indicator of a radioactive anomaly. 3.3 Fractographic Analysis Because Slab 4 was closest to the core midplane, it experienced higher power and fluence levels, relative to the other two slabs. For this reason, it was chosen for initial fractographic examination. The cutting plan in Fig- ure 6 was used to expose one side of the crack fracture surface for each of the affected webs. Photos 5(a) and 5(b) (4x original magnification, but photographically reduced 22%) show the fracture surface in the webs between the B-face and coolant channel 319 and between coolant channel 319 and fuel hole 307, respectively. In both cases, the fracture surface is considered typical of tensile cracking. It was verified at this time that indeed no crack existed in the web between fuel hole 307 and coolant channel 295. The SEM (Scanning Electron Microscope) photomicrograph in Photo 7 (original magnification as noted, but photographically reduced 4%) was taken on a sample of the fracture surface between the B-face and coolant channel 319, Slab 4, and shows how the graphite microstructure is a composite of dense, ordered coke filler particles, and unordered binder with porosity. With increasing magnification (50x, 100x, 200x) , the greater long-range order of the coke filler particle relative to the surrounding material becomes more distinct--and it is that long-range order that allows basal-plane cleavage to take place within the particle. Assuming the needle-like coke filler particles are basically aligned in a given direction, it is not difficult to imagine a stress field operating to fracture the material preferentially in that direction. -10- ' 3.4 Metallographic Examinations The metallographic analyses included examining the graphite microstructure for possible anomalies, looking for differences between longitudinal and transverse microstructures, and determining crack paths. The metallographic samples from locations indicated in Figure 6 were prepared for examination by pressure-potting techniques--i.e. an epoxy resin was forced, under pressure, into the surface-accessible pores in the sample. Once the epoxy had set, the samples were ground and polished for examination. Photo 6 (32x original magnification, but photographically reduced 50%) shows typical microstructures obtained as described. The light features that have "grain" resembling that of wood are the coke filler particles. The binder areas are usually slightly darker, but are best identified using polarized light. The dark gray spots are pores filled with the epoxy mounting resin (note appearance of mounting resin at edges of samples). The few black spots that are visible are pores that did not fill with resin, and are considered to be sample preparation artifacts. The H-327 graphite, as an anisotropic material , has distinctly different material property characteristics in the longitudinal (extrusion) and trans- verse directions, although this is not readily apparent in Photos 6(a) and 6(b) (32x original magnification, but photographically reduced 50%). This set of samples was taken from the web between the B-face and coolant channel 319, Slab 4, but is also typical of the longitudinal and transverse microstructure observed at several other locations in the slab. The final examination on Slab 4 verified that no crack actually existed between fuel hole 307 and coolant channel 295. In Photo 8(a) (original magnification as noted, but photographically reduced 51%), the presence of a crack can not be directly identified because of the saw striation marks. Again, in Photo 8(b) , no cracking could be found on the axial projection of fuel hole 307 or coolant channel 295 (not shown). But the photomicrograph of the web, shown in Photo 8(c), displays a string of pores between 307 and 295. Such a string is a good candidate path for crack propagation. Slabs 2 and 1 were also used in verifying the localized crack morphology. Photo 9(a) (32x original magnification, but photographically reduced 49%) is a photomicrograph of the crack between the B-face and coolant channel 319, and 9(b) shows the central part at higher magnification (128x) . At the center of -11 - 9.(b) the crack can be seen to follow a path of easy cleavage along the grain of a coke filler particle. Photo 10(a) illustrates another energetically- favorable crack path, in which the crack tends to propagate so as to link up pores. Interestingly, no crack was found in the web between fuel hole 307 and coolant channel 295 in Slab 2. Slab 1 was the only one of the three slabs to have cracks in the first three webs from the B-face, as shown in the sequence of Photos 11 and 12. Again, the general crack microstructure in Slab 1 echoed the patterns already observed in Slabs 2 and 4. Cleavage along the grain of filler particles and crack propagation between aligned, or even semi-aligned pores, seems to be the preferred mode of crack propagation in all slabs examined. This is consistent with what would be expected from the standpoint of lowest energy of formation, and is the crack mode usually found in artificial graphites. It should be mentioned that during the sectioning process, the web between fuel hole 306 and fuel hole 294, Slab 1 , broke. This area is within three webs of the B-face, but it is unclear whether the break was the result of propagating an incomplete, pre-existent crack, or if the crack was completely the result of handling. In examining Slabs 2 and 4, no evidence was found of cracking between fuel hole 306 and fuel hole 294. 4.0 Conclusions In general , the results of the post irradiation examination on fuel element SN 1-2415 indicate that the microstructure of the graphite element shows no unexpected features. The crack extended two webs deep in the Slabs 4 and 2 where the power density was higher, and extended three webs deep in Slab 1 . The results from the radiological examinations show that residual activity in the graphite is very low and well within anticipated ranges. More impor- tantly, the measurements indicate that no unusual fission product migration had occurred through any crack from fuel hole 307. The results of the fractographic analyses indicated that the cracks found are typical of cracks found in comparable artificial graphites6. The fact that the crack remained open by a finite amount at the surface of the slab, tapering to zero width at coolant channel 295 and fuel hole 307 indicates that a system of balanced residual stresses existed in the element after the time of fracture. However, a method to determine the time of crack initiation and propagation is not available. -12- The metallographic examinations documented the tendency of the crack to propagate between pores and along the easy cleavage direction (i .e. the basal cleavage plane of filler particles). In all cases, the crack tended to follow the path of least resistance, in other words, the path that was most energetically favorable between the aligned pores and along coke filler particle cleavage planes. Samples taken from all three slabs confirm this as the mode for cracking. In conclusion, the destructive post-irradiation examination of three graphite slabs from fuel element SN 1 -2415 showed that no excessive fission product migration had occurred through the cracks, that the overall graphite microstructure had no unexpected characteristics, and that the fracture sur- face is typical of thermally-induced cracking. References 1 . Saurwein, J.J. , "Nondestructive Examination of 54 Fuel and Reflector Elements From Fort St. Vrain Core Segment 2", GA-A16829, GA Technologies, October, 1982. 2. "Nondestructive Examination of FSV Fuel Element 1 -2415", GA-A906505, GA Technologies, May, 1982. 3. "Visual Examination Results of Segment 2 FSV Fuel Elements 1-2415, 1-0172, 2-2693, 1-0108 and 5-0801 ", GA-906577, GA Technologies, October, 1982. 4. "Test Procedure for the Destructive Examination of Fort St. Vrain Fuel Element 1 -2415", GA-A906770, GA Technologies, Febraury, 1983. 5. Fort St. Vrain Nuclear Generating Station, "Updated Final Safety Analysis Report," Public Service Co. of Colorado. 6. Private communication with Robert D. Reiswig, MST-6, Los Alamos National Laboratory, 1984. -13- R Y a= s p t Si 8u�it t i I I : I I _ W t E?1“; J S £ I I ` i . ..!;_i , I opoJ !!;P- �yyW W g:=W et Ate 4_.L,1. . I . . ! I . J Vii stp OCJt • i ' i i i i ' uW �' � ! _ ,7u{;TT'T� EIll _ t t t t t iii. : 1 W Y I r Ii 2: . 1 n u O 1i ,.. W O io p pt P! C`t J yDt N W W J +WD �ii h 1 W J� J_ WS t. Q �W 13 _�',� '. rt L C u tt L g "en.0000000000 ` • 00 00 a /% S3 ri L 00 �� '"0000000000000rAt 0000000000000 cC V\ 700000®00000 �"00 gi �. ®00 0000000 �, @40®®av®_+/fee 000 eeeee®0E7®e z®00000000000000iroseeneeoes000sese 0®®®e®®aae®®®e 00 i� .- 000000®0®00®000000 V000000000000000000e®e®e®ea � 0000®00 i ®000000000 I� W 000000 ee®00 e I II Vilaneatatt ®e®®®ae �� =We W Nd Y < , W, , f ii t C W ` Yo V[a6Dr s to ;it ;, t a IA, 1.--- i A COOLANT µDIE i'''S•'CIS'OSS 0.625 DIN.1102I C.' S^oSL D e€ c •••"s77•••;•4 T ob }S•• t cooky! µcLE ^S= c 1-': 0.500 D1A 16, S ;IgTcb'o;S: o 'in I! II1r&Iii c6 os6LErolsa+ • r7L^o9. M AOISe .;SS,S_oICUS ;SS�cbsossze S. OBfMOLESSS _ 0.500DIAI210) P01;C4,0 S' 'ST °3 '=S=S6°�S nrc� ;S �4. S_ efr SS'S'.'4:"1`6 baS cS .'-"- )S'ebe A CEMEMED 1 Iµ. GRAPHITE CLEARANCE T M rL.UG I TYP) FUEL µANDL IMG I �� / �r1cRUA n•:ij D04.1. PIN _ -a. fl "•11 a ! \ IIt 1 .. .I IN. I\I\ 15 Iµ. fr'� .1 y f� \ \ tiYy tyt 1 N \ � 1:M.4A \ \ \ µE..n i�',-'yl a�rY� g� - f II I j?\1\. � PLO. ITYr, _ -- - ]1.22 "} , y YYU4 , li } \ \ N '. D `\ \ 1 \ \ ►OISDN�E \ \ \ \ \ N 1 \ I 41 ��, \ \ N\ \ COOLANT i{ N \ N \ gl h1/4 tf 1 \ \ AWL ROD i� 4 '1 t I I\ \I I \I \ „s�N i 1� N.'. I >LERGTM ir'� \ I N I ti -- '',. ; Uc A-A \ DOwtk i, SOCKET Figure 2. Fuel Element Geometry D O.OO® oeoeoe • O C> O- .O t,1/// � a) oO SO0 e �of oGs OO O ® O tot. O t] i to O t. „ O o o O o0oo fa O Y se ., f) Oo fa O ft o e O o e O ® O O iC) ie O ® ® O 11) Ib O® O8 ® O ® ® O r.< ® - } o spa a e0® 80 F i • • it- �Sl f+-}]{' �a? • 1Ct -e- LL] O t<5 116 =•.zK1ip/F�MS +- 11\ / �./ 800887 fi Kf�s/N`/{ �Y li O ® H] ) J �S< 9) O IT et -�� pGD 0 0880880 SA O :II ):) O SM ® V ® Sa O ):J ® O 1) a< O O ossOO s {, c2000 Sf cOx= o O ® © oecO €eeoe ® O %) 00 - 008806090800 #000 a pti O 0 !>) O !)i ;; \O e © Dee O LS )Y I JC O IO 'a - L ® 08 e0 eoee Icy eeocy A ® ® O ® CD!m0 OO )<)�a B 295 Crack 307 30B Figure 3. Coolant Channel and Fuel Hole Designations 318 319 320 -Crack B-face Figure 4. Element SN 1-2415 Sectioning Plan 114—x �� - � Slab 1 'se se _ .- .00 - - - - - 6 -- - - - - - � 31 .22" 14" 1� �\ i e. i i___ 6" 1._ , " . 1— ,, .. - - - - _ _ _ - Slab 4 .... ... _ _ _ _ _ _i se.- 7" 1I O O to / ^^ .o `:J�V sor-)Q)a6ociPe zu o• "Ao384'o00000 it � ooOOO , Oo v J n en r: , u11:,•:� �' O OO�OG . " I of\ „ O6°G)Cm°® LI®oOo� O°O°owoou®ne(.,V® �'. �'O°O°rTh ! Ooo ; O I t, n ^ , r oLl/*I O Off } n-- 30C- Que'/8 n u ? � n O � O � �� ^ E pOpOrJOG("(1k) n 1 E 000 �(_,Op . pn 0 t, p " GO ! c m ^ (^ Or�OO O n F � O, O♦ O n O � , ` l`/) r O '^ O :; © o OC O�OOO�j OO\ `J " O r O ' co GOo`) � :: o � O ^ t l � a 10 r O a C S' N co Or,0O0 90OOO °° 0��0 4 O O d m m cc rt Is Z ++ E ElE a+ CV 4 c 4 n®O';')OQ)O : on) \ cr"uOuC ! cfl )O© c Uo 1o''O®O8; OOO'O°oo 1 N v 0006000/ Nvc1o®o®L}oowOQbo O i G rO. OO ^� n OOOOO , O r, O r. ��+ c�; o o° r O r, CJ. .pr Ii ODUC 8 0OO�O0w0 " OClOO ®®®()®®®®©©&©O O 0 G O � � r. Id O O°O°O°Cj°O° ;°Oo0o0c0 ` O L C O 0 O r 0 ® p O P. C) @ _, c n I 0 t0 o� d 0o0o/� 0:1! O`D =: OOO , 00 O Q�ricvr� O �;lr, J �J _, c� , 0�: � ® � ....,(5. F.0......,0....0 , 0 _ 0 _ 6 _ 0� ^ ^ ^ I Z a OrE �,O� �7 .;JO`oJOoO..J 0 � y v t_) .0 i r M is a, u 6 . , C oho-ex - 1- s. .,09:( T u 090(.2(90.ii V' U( ... , s CyuJO C�w0 ,• Od0 uG noo�0o�•JnOo� JO®O\ 05013 � pvbopc, pop ^ ?& ? 92GOnOO®�� „ 0p pOpOow0 I000p . 00000 f0CQ(,\CQpp COO O On •r QODOO �O�O�OO 02(7:2 CO\Ail 7/ Cl O 1 O O tJ 6„, O 6 O o 6) O Oo opopopo_.opononce% 0.& O 0 0 0 kv G ' p (o ;: cm0 qo 1 o84*(8 SCI(PCS8E!po0o c,)IcAO, g O�,Lio0v0O0O0 a fo O C' y� C ` / U t1 r • z e d C ` O C� 4 C v 7 Y .1 E > w �i E y W • Figure 6. Fractographic and metallographic Sectioning Metallograph 295 Slabs 1 , 2, ;`` 307 308 Metallograph 0 Slabs 1 , 2 Fractograph 318 319 320 S1ab4 Nipt +s� metallograph Slabs 1 , 2 B-face Metallograph Fractograph_ Slab 4 Slab 4 B-Face '. j - 319 -n`! r. G: r ' •f h....4 � 295 xxA,74: I- rr fr • SN 1-2415 Element- -Slice removed showing B dowel Photo 1 caK!. - , 307?"°'' . w , .y ' rr t '4, -4;.-4:3:8:::--i. 1,1",..c .,,,-,.....,_. ,,,,_,-- .. e.,..„,....„.°i .� . ....., ,.rAr. Top of GAT Slice, showing • crack in dowel socket Photo 2 • • • e '• - • • • 62.,hiDeoCeefack ♦ • • • 00e0Ot • • O . ( • • t • QSc e 0 • '•° • e • ♦ • • • • • eGCGCO • lies • • • • • • e o w c c e•1 Ar- ct e e F • 41 • • • • • e onene: • • • • • • • C11 •4-t) o • o • “ vo'w: . • ( . • • • • • • S MG . ) C : • s • • • • © e ® • ® s► OOC * c r • ! • • • • • °re • tt099sC9 ♦ • • • • QweS • • • sste i • • •4r • • 00 • O • s Vole / ‘ ^1 • • O • aO • • • ssi • is • . . • eeeeet s • , ^ • n 0 O • . • C' ,- • • wen B-Face 013-5 FT.ST. URAIN FUEL BLOCK S/N 1-2415 SECTION 1-TOP Photo 3 0 en c' M z x F O F ul o U W tip , ,w .. . :‘ .0' t tti • .. A , .,. 1r f, 1. �s R y ''�y'�R r Y: I� l Li. Li h \. cm 1'�S" W a • \ Gar N S Y # - . N•0 • T z C, O -- in U r80✓ N lli N S _ r .. fir , B-Face 319 i • 4�. 319 • 307 l i�ty 1 Photo 5(a) Photo 5(b) Fracture surface between Fracture surface between B-Face and 319, Slab 4 319 and 307. Slab 4 4x 4x MS New s1 , , J ReLn= li�' t t ♦ ' r ig-t 4� E. vl a14 ,),,u-4.•,,,,: e:7,M.'+ fry.. -. •� -•i I rw.�(y\S�� IN'', t y Gy j+,K •1 ,ro•w " \ -;/..4"'", ti( T piv‘ ` r� r. lrl� • X4- .t ,- ! ,v .• f"'� 't� r t rye, ' ♦ : v. `J� .c1 l c *"/ : = c. �r !.y \'P"Y_t.' yy .? �;' a •tw7S. f e Y {.... .it' lair yr "�i: ,r�r� • ..''E f�� . C �t,3�., •c - Poree S no epoxy t # ', f 4„,;1,4"` �!1_•A ° 'r tze � As • rr44 pr e P ii .6' +S; V.4•'‘ply Pees with 4: •_?�♦ ei i,�sf;g.'u eta} 1 .0'/ : Ay' apouy `� rt.?... r 'ice •"t .+A l ,•.•' ,s/ 7i (•. •ate' i +\+A : t t I . _ :rd 0y'� ,ji,t?-"ins• "C- Photo SOS l.1/2; 4 ° lti ir- ..Photomicropapt� of Transverse • cross section between 8-Face`.',I% " �� Photo ) end 319, Slab 4 Photanicroccaph of LonyhdE,el cross section between B-Face and 319, Slab 4 rite • Y SEM (Scanning Electron Microscope) Photomicrograph • .▪ e '- showing cleavage, Slab 4 .�M Fa1 /' s mot"'- oft Olt r''/t. -. , Photo 7(b) 100x Cleavage in dense area Jo S•�� T r + z`J''c t : tie., #0,,iitssee: t L.-----7".-- ttv �' T• yam... • ego iermirm - r' l f rC �a '+ t _ pf' r 7A+ �[ - ,,. . : Ill t" f- a * iiii,mi `N �. mac- .1, ." ---...k.. .t,st. ji; Photo 7(a) 50x Photo 7(c) 200x CA N1C yN it 3 1Jj•ti $.. e • \_ aas. i`••• , may. . CO N Cpm f w1 I�'' r , � �se j..-o,6,.,-..:-.r .y .ii B E a ' _ ill 3 t 4 co V to 'r N • ti ti 4- ^.y r+ ` a A s fk ; r,.1 1 •4 , " .-" Illiamilemisi, . . ' , 55 . f Mi � ..ak�4<tJ.5`g4t.:yr l L g;-- <Is. b �`'!�, ' ; FNS ' L ♦ n •• t•111 ' • • •w Ai•J t. b % X1'11'' c • art e• •- ii i� • a'. . .'- d s 1 . pg 6•• 41 r �'. .r .1* -1. in tip` {pc5.4 11.:44..,1%• ° y roK:l• M e' ' Ssa, . ` `+ ti�1s `•a.p``!,• .. p]{' r � » ti1S :.,-11....-7 Y'j •h •C • ,r !: .-;..'44','. ' 1• • �c+T\ �f•rJ . r ;• " 'ira 33 • s4.•:.•-•:1444' •'` f .Rt,; • ' _el�\ . .t„•-.y f 1. tw ti < tr 1• r es aa• • Ctl .} ,ii3 s el t .I� aA F {• ala ♦ • �.,Y . ASE" ----• ` ..� • I 1 ^« •'r. r. c. "��www =_ r i " •i4 ._ .. j" ♦4 a �; 1/47 „5511..1 s • ,- ..J (• tt•• f -.' a ,/ .. rr,.. • a-a'. 44 c V•�t�'4 .''t\ 4`f `- •... •• .... ••••••• •t4 2't 11y� ' ti RI J ( 1a'• 7 41'4 - S, l , a- M r 1.. Photo 10(a) Photomicrograph of Web 319 and 307, • 307 rnt • .S•.* a• ..`r . s• ` • ♦ - -C �T • f t. n • • ,:t• � "I.*:• "4 1.1 I . ' Photo 10(b) Photomicrograph of Uncracked Web between 307 and 295, Slab 2 1 No Ds x _i_f a s.'. .. N \ ? C‘i s i'4 4✓ a L J.5Y* .i..;fi x s r x. r •• �i apt I /'r f / A w '1 .1 '� {r /11,4( R a • - ' -t., ., v'. i 'r • at -4 17 e . � • . .4 .�, s a a �` { •.•• . �•��'*':Ire... •-•• ^••r:R i9 J i art r••••t...44.40910.-af; j ;; r '. p knLL x to CV • '! '• m • I 1.,isau . ' liggilitility... — , 1� • I to a• r .. ,„:: a .-_a sue''x 4:Z', ' F bI. ' ' + . st 6.... Ju•l ',.' `•',4• 1 7)y . 307 . . ..4 AN.? Photo 12(a) AN.? 0' Photomicrograph of Web between 319 and 307 Slab t-, `I`4efdi Y it4-9,,- —' Photo 12(b) '9,;-� +a. '" a s+4 r Photomicrograph of Web between a �y-.„-e'' Slab 1 e='t.;.' a• +• 1 307 end 295, M !. fAs. Appendix A . . . , h L-DA 17\) v 8 C ' I A I B 1 . ._ C I D Sicj Ju " 7; 0;O 1 11OLF_ # 316 • 6o O --1 ._(QZ/—.— — — :_62/ — . 1.r%.) _ _ --- 1301E # 317 I_._._y97--1 . y97€ - =` --I—•S92:---- BOLE # 318 •4/9 --= y91`--- _y97s .y9'?---- -- —_ BOLE # 319 t-( .2 -- --•to 2Y-- — =�Z/5 11 HOLE it 320 I- • '7'97 --- - -.. 07'97r_ -- ' 'I 9.7e I • A/9?1__---- HOLE # 321 y97- ---=- t7 ----L.Iy7c I • y97E - HOLE r 322 . / 21 - —=6 2L ____ t(p21-._.-__ _.2z i---- - - - HOLE # 295 - -/ '2-`- _ 223 _�zl: _ _�2-- - HOLE # 307 • .(:..9e --=1? _ • 'IZs •--- ___111_1?_____-- _ HOLE # 308 • 4/7 - =iMS------!_•..Oa__---- .-- . O7 _----_ f.6 12 __I-___•-(?-.2_ . G2z --- --- HOLE ._17---_._S.P ---- N9? - -..-._ I---=y9P_--- .11_1 ______ -- --- - ___ HOLF_-y- 1R----�`17 . . :In ay9P-- - -- . _._Y_9P - - - - HOLE # 30 1 . 6,22. __6?•'_2-- --- , 1:,2.2 - _62Z. -- \ . i'/ A Ci�P / C .5rd/3a 2 /OP 1 A B C I -" HOLE R 3161 .6).2 / _L_2d I _r- 21-- ------- -'2 / -- -- HOLE # 317 ( )1975- --_1125.—_-____I-:'5 — HOLF_ k 318 _�?7c _— . 99p( _.-__ii-21 r • yl g ------ HOLE P 319 . �&2s . GZ'I -- • 622 — — I &pi — — — HOLE P 32 0 y?, _ _998 • y 91 e-- - ----1 . '1'7_ _ HOLE P 321 — _, q97. . sift __ i °7, , i97 - HOLE ,'r 322 . 2/ --1....i_Le..ai.-- —•-60 2? ___ .--_4A'1 -- 2 .�2/s ---- - HOLE ° 295 ,�21 -- ---- -- .�22 _�.? — •-- I-3OLE !' 307 , a4�-- -' .ig — .- -91--------- if - HOLE k 309 . ,fiP _ _y97e I—_132.--____--- • y42Q----- _ -- HO1,E ¥ 6 - -I U . f Us - I ..l? 22S_, .._ — b.n. -- -• S- - - - ...a24S._-. ._ . HOLE J17 —__, y9 R--- - I . 9?r- _ _Y9P — - _ sift s--- e. -- - -- - -...�._,'�--- -- -- - HOLE-#-1g-- --'!�i- L'`-------=`.,,p er • • — • .97 •- ---- — l c L 2 2 s __,Az2 c--- -thfl s----- HOLE _ 30 _ G22= -_ _ X, <>9P ' ii‘ °CA icrt;at') iii y Zia__ A r s 11OLE it 316 slo 21 _ 9 HOLE # 317 , y7t , y98-- _ ys�s_ �_79P HOLE # 318 ,_197, —Yn7_s_____._ _ ._y97L___- Di --- HOLE # 319 =-(0 54 —•_,-f_-2•4 . (.D.22i . `._ �9. HOLE # 32 0 ' 8---- _ ill - --- 5 2 __ •±1? - HOLE # 321 I, y97, y983 _1_192s • y37 HO1.E # 322 (221 s .It215---_-_sid- f ______ -•-I?- HOLE # 295 (1.21 :6e21' - '12.22,- -- __-_4_zIt___---- - HOLE # 307_ . Yg?, _ _Fs----___11_ig---------=- 31._—_--- V9- - HOLE # 30c _yy7s =`�'7.`----- -_ _* i �.=__—_— • y9P- - HOLE #' 6. .- .-�. _ L%-2-2-s----•--- •-u 3---_. .. . _. _ _ _is) =--- ___...---=Gg21 _____ HQ1,_E g_12__ _._t y1.?l_ _._V Ps - - -- ' y9Ys- - - 2 till? r- --- • Y9 9 -------- -- °—y9_9--- -.-- -- =in. ----- i � 2 HOLE # 30 .1 . Lot • G_13— -- .(0225 . _ S _— A-34 � J \' ii:G�.. DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, of any information, apparatus, product or process disclosed in this report or represents that its use by such third party would not infringe privately owned rights. - i - ABSTRACT • During a routine lift-off force test of the prestressing tendons or Fort St.in the Vrain,pColoradoed concrete reactor, brokentstrandsel in(PCRV) n the several tendons were observed. These pretensioned tendons apply suffi- cient compression in the concrete to balance or exceed the cir- cumferential and vertical tension in the concrete that results from the internal pressure. A combined analytical and numerical study was undertaken to evaluate the evolution of these stresses, both to the initial prestressing and to subsequent partial and total rupture of these tendons. At the stress levels anticipated in the concrete, and for the anticipated operating life span of the PCRV, the concrete behavior can be modeled as a linear visco- elastic solid with the creep strain varying proportionally with the logarithm of time at constant stress throughout the projected reactor lifetime. A one-dimensional model of a long concrete column of rec- tangular cross-section with an embedded prestressing tendon along the length was used to evaluate the concrete and steel stresses as well as the hold-down and lift-off forces. These were evalu- ated for the intact tendons and the degraded tendons. The degree of tendon degradation is described through the ratio of the number of unbroken strands to the original number of strands. Initial time of rupture was varied from the time of initial pre- stressing to 400 days after emplacement. The formulation led to an integral equation, which was solved numerically. The hold- down forces decayed approximately with the logarithm of time and for both the extreme observed degradation (21 broken strands) and for a more extreme case (40 broken strands) the hold-down force still exceeded the minimum safety design requirements. In addition, several finite element calculations, using the finite element code NONSAP-C, were made to evaluate complete tendon failure in a 600 sector of the Fort St. Vrain PCRV. This code has an extensive material library of constitutive rela- tions to model the various properties of concrete, together with a 'specialized element model to simulate prestressing tendons. Two rows of vertical and arc row of circumferential tendons were incorporated in the model as a baseline calculation, the tendons were prestressed to 700 of the ultimate and an internal pressure of 775 psi was applied (this pressure is the internal pressure of the helium coolant in the 'HTGR) and the creep of the concrete and slow decay of the tendon stresses were evaluated out to 30,000 days. Then, three cases wherein one tendon was removed at one day were evaluated. First the middle vertical tendon in the outer row and in line with the outer buttress was removed. Second, an inner vertical tendon opposite the thinnest portion of the PCRV wall was removed. Finally, an inner layer circum- ferential tendon at midheight was removed. Stress redistribu- tions at 300 days after ruptures were calculated and shifts of the remaining tendon loads accommodate calculated. Regions oflocal tensilbroken tendon e and shear stress in were the concrete portion of the PCRV were identified and related to overall structural integrity. - ii - With all tei ns present, the mean vertical .ress was about -760 psi , the radial stress decreased from the applied internal pressure of -705 to about -1200 psi at the ring of circumferential tendons and the tangential stress ranged from -2400 psi at the inner wall to about -2200 psi at the same place. Removal of a vertical tendon reduced the mean axial stress by about +40 psi ,. the local tangential stress by -10 psi and did not materially affect the radial stress. Removal of a circumferential tendon reduced the mean tangential stress by +30 psi and the local axial stress by -B0 psi . The vertical hold-down force from zero days through 30,000 days decreased linearly and remained above the prescribed safety limit, as did the circumferential hold-down force. Comparison of the analytical solution and a small finite element problem simulating the analytical problem was made to verify the viscoelastic creep models and the tendon element in the NONSAP-C code. Excellent agreement for stresses, strains and hold-down forces was obtained. - iii - • FOREWORD This technical evaluation report is part of the technical assistance program, "Review of Selected Fort St. Vrain Issues", FIN No. A-7258, and is supplied to the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, by Los Alamos National Laboratory. - iv - Chapter I I. Introduction 1.1 Statement of the Problem During a routine inspection of the lift-off forces of the prestressing tendons in the Fort. St. Vrain reactor's prestressed concrete reactor vessel (PCRV), several of the tendons were observed to be partially degraded in that up to twenty one of the individual strands (of 169 total strands per tendon) had failed. Since these prestressing tendons carry a large tension, thereby placing the surrounding concrete in compression, the evaluation of the concrete stresses and the subsequent adjustment of these stresses to the degraded tendons, needs to be accomplished to assess the continued structural integrity and functional capability of the PCRV. Group Q-13 of the Los Alamos National Laboratory therefore undertook a systematic investigation of the evaluation of the concrete stresses in response to changing tendon configuration. The initial level of the stress in the concrete is such that the major feature of the concrete deformation is slow creep in compression. The approach to the desired engineering evaluation of the structural integrity is twofold. First, an analytical model of the tendon and concrete stresses with prestressing of the tendon, subsequent elastic behavior of the tendon and visco- elastic creep of the concrete is developed. The concrete creep behavior is generalized from the long time creep tests at constant stress on samples of the concrete in place. Tendon degradation is modeled by reduction of the effective cross-section area of the tendon proportional to the fraction of broken strands. Time evaluation of the hold-down and lift-off forces are then evaluated through solution of the resulting integral equation. The second step is the evaluation of the stresses in a sector of the PCRV with a number of initially prestressed tendons. The concrete stresses with all tendons intact is then compared with the stresses with selected tendons degraded, missing or removed at some time after initial pre- stressing. Solution of this problem is effected through the finite element code NONSAP-C, utilizing, in particular, the viscoelastic constitutive model for concrete creep and the tendon element model in the code. Connection between the analytical formulation and the finite element calculations is provided by finite element solu- tion of the analytical model . - 1 - With all tendons present, the mean vertical stress was about -760 psi , the radial stress decreased from the applied internal pressure of -705 to about -1200 psi at the ring of circumferential tendons and the tangential stress ranged from • -2400 psi at the inner wall to about -2200 psi at the same place. Removal of a vertical tendon reduced the mean axial stress by about +40 psi , the local tangential stress by -10 psi and did not materially affect the radial stress: Removal of a circumferential tendon reduced the mean tangential stress by +30 psi and the local axial stress by -10 psi. The vertical hold-down force from zero days through 30,000 days decreased linearly and remained above the prescribed safety limit, as did the circumferential hold-down force. - 2 - Chapter II I. Introduction The time evolution of the stresses in the prestressing tendons and surrounding concrete was accomplished for a simplified, one-dimensional model of a typical Fort St. Vrain PCRV tendon system. This study was done to investigate the decay of an intact tendon system and a degraded system wherein a small portion of the tendon strands had been broken at some tine during the structure's history. The concrete is modeled as a viscoelastic medium defined by creep tests at constant stress. Decay of the tendon stress, concrete stress, hold-down force and lift-off force were generated for a temperature range representative of the FSV operating environ- ment and for a number of degraded tendon conditions encompassing those observed during routine inspections. Figure 1 shows the schematic diagram of this problem. A steel tendon, of total cross-section area, As and initial length, Ls, is essentially stretched to an initial tension, as. This tendon is attached to base plates, which compress the concrete surrounding the tendon. The initial cross-section area of the concrete column is Ac and its initial length before compression is Lc. The initial com- pressive stress in the concrete is oc. We consider only the uniaxial stress component and uniaxial deformation in the tendon and concrete column. Stress equilibrium requires that Asos + Acct . 0 c through the history of the deformation. Equality of the lengths of the stretched steel tendon and the compressed concrete column requires Ls (1 + cs) = Lc (1 + cc) where es and cc are the strains in the steel tendon and concrete column, respectively. The steel is assumed to be linearly elastic throughout os Escs - 3 - while the concrete is assume. to be viscoelastic. Its prope' cies are defined through a creep test at constant stress. Appendix E of the updated FSAR (Ref. 2) yields the following form for the creep data for a time interval from t • 0 days to t = 300 days. 0 cc = e1l + aln (t + 1)] Where a is a constant dependent only on temperature. We assume this form of the creep equation will be valid for the range 0 ≤ t ≤ 11000 days (- 30 years) . The values for a were determined for the data presented in the FSAR, (ignoring the initial 2-8 day rapid variation) , and, using Ec = 5 x 106 psi , we found a = 0.12 for T = 75°F and a = 0.40 for I = 150.F. Generalizing the strain data at constant stress to that of a variable stress-time history through the Duhamel integral , we have: /itcc(t) = rl - E) oc )dE 0 where J(t) = 6(t) + tl+ 1) 1) with 6(t) denoting the Dirac delta functional . Eliminating as, ac and cs from the four equations, the following Volterra integral equation of the second kind is found for the concrete strain c (E)dE Acc (t) + B tE�E*-T + [C + D In (t + 1)] = o , with As EsL A = hy +1 , c c s - 4 - aA sE sLc S = — -r- ccs C = Ate_ C -1 , and cc s D = aC . The steel stress can be written in terms of the concrete strain as i es = Es [s (1 + £c)-1] the hold-down force, F, is given by F = Asps The area of the steel tendon, As is the product of the number of strands in the tendon, N, times the area , as, of each individual strand. As = Nas Each strand is 1/4" in diameter and, initially, there are 169 strands per tendon. The steel , with all strands in place, is stretched to 0.7 of its uniaxial yield strength, which is taken as 240,000 psi . The concrete is compressed to an initial compression, at t = 0, of -750 psi . Thus Ac/As = 224 , and Ls/Lc 0.99428202 . - 5 - The integral equation was solved numerically for N = 169 and the temperature of 75'F and 150.F. The hold-down forces for these two cases are shown in Figure 2. We also solved these equations for the case where the initial pre-tensioning is applied, and then M individual strands break at t = 04., Just after the initial load is applied and equilibrated. These are also shown in Figure 2, for M = 10 and M = 20. Also one case was evaluated when M = 0 for 0 s t ≤ 365 days, at which time 20 strands (AM = 20) were allowed to break. This was done for T = 150'F and this curve is also shown on Figure 2. From these solutions, it is seen that the stress decays almost linearly with log time for a > 0 where the load is applied at t = 0 and decreases faster with increasing a. For a = 0, the elastic solution is generated and there is no stress relaxation. When the condition M > 0 is applied at t = 0, the stress drops to a lower value and then the subsequent time behavior is as above. When M is changed at a later time, the stress relaxation curve changes rapidly, within a few days, from the initial relaxation curve to the other. In these calculations, F falls in 30 years from an initial value of 1394 kips to a low of 1130 kips for the case M = 20 and T : 1506F, which is well above the minimum requirement of 980 kips. The hold-down force for several tendons has been monitored for about 10 years, and there is reasonable agreement between the long term trend of the data and those calculations. The measured hold-down force (Ref. 1) (measured for six vertical tendons through load cells from t = 100 days to t = 4000 days) missed slightly the theoretical curve using the design data listed above. A second set of calculations was done using the measured data instead of the initial design data for the steel and concrete. The initial yield stress of the steel tendons was measured and found to be between 252000 and 256000 psi . We then used 250000 psi instead of 240000 psi and took the initial stress in the steel tendons to be 0.7 of that value. Re-examination of the concrete creep data in Appendix E of the FSAR (Ref. 2) yielded values of a = 0.25 and a = 0.50 for T = 75'F and 1506F, respectively, where t > 20 days. The rapid initial strain during the first 2-8 days was incorporated into the initial elastic strain and thus the effective initial concrete modulus was reduced to Ec = 4 x 106 psi . Figure 3 shows the time evolution of the hold-down force, using these values, and the comparison with the load cell data (Ref. 1) . The fit between the theory and measured values is much better. The current lift-off force can be calculated in terms of the current hold-down force, by assuming that only elastic forces are applicable during a short time - 6 - • lift-off test and applyin'^'ust enough axial tension to t steel (and thus axial strain) to reduce the current concrete compression to zero. Thus f • F (1 + s)/[(1 - F/AcEc)] • where f is the lift-off force and //JJ F is the hold-down force and B = As Es /Ac Ec For this problem F/AEc - 10-4 , c B = 0.0268 for the design parameters, and = 0.0321 for the measured parameters. Thus, f = 1.027 F (design parameters) or 1.032 F (measured parameters) . The relationship between the lift-off force and the number of broken strands, M, was determined by linear regression analysis to the data (Ref. 1) which yielded f = 1425 - 6.95M + 30 kips. - 7 - Chapter III III. The Simple Finite Element Model . An elementary test problem using the finite element code, N0NSAP-C, (Ref. 3) was formulated to evaluate the combined viscoelastic creep model for- concrete and the prestressing tendons. N0NSAP-C is a three-dimensional finite element code, derived from the N0NSAP code. It is specifically designed to evaluate reinforced concrete structures, with a number of concrete constitutive models specifically formulated for evaluation of various pertinent concrete properties. The main property of the concrete needed here is creep at low stress levels. In N0NSAP-C the creep compliance at constant stress is represented as a sum of exponentials, N a c = a o ' t r ( - exp(t/Ti)) i = 1 The maximum value of N allowed in the program is far. By choosing al = a, Ei Ec and T6 = 104T1, T4 = 10371, T3 10071, T2 = 1071, a good approximation to the logarithmic time fit, described in the previous fit is available. If T1 = 3 days, the fit is valid for 1 day c T < 60 000 days. The values of Eo and Ec were obtained from the fit of the Ft. St. Vrain concrete creep data (Ref. 2) , described in Section 2. The analytical problem of the previous chapter was simulated by a finite element structure with ten 20-node bricks of concrete in a five-high by two-wide column. A steel plate cap of two elements simulated the confinement. A single tendon element from the fixed base to the steel cap and passing through the mid-side nodes of the common concrete element was pretensioned. The finite element deformations and stresses were evaluated over a time period from 1 day to 30 000 days after pretensioning in logarithmic steps of one-half decade. Dynamic inertial terms were neglected. Figure 4 shows the axial hold-down force in the tendon versus time. The time decay of the tendon hold-down force is very similar in shape to that calculated from the analytical model. Slight differences •in the peak axial stress - 8 - above and below the mean due to the three-dimensional iture of this problem and the welded approximation to the steel plate-concrete interface. The numerical solution to the finite-element simulation of the analytical model is sufficiently close to the previous analytical solution in magnitude of stress and axial force, and in its time evolution, that we may confidently use the fit parameters in the NONSAP-C Dirichlet expansion for the creep for_the more complex models of the reactor containment wall described in the next chapter. Chapter IV FINITE ELEMENT ANALYSIS OF A SECTOR OF THE FORT ST. YRAIN PCRV CONTAINMENT WITH PRESTRESSED TENDONS 4.1 Finite Element Representation In this section, we will look at the stresses and tendon forces in a 60° sector of the outside wall of the containment vessel of the Fort St. Vrain reactor. Figure 5 shows the mesh used to mock up the PCRV wall . Figure 6 shows the horizontal cross-section of the mesh for the containment wall for this sector. From memory limitations of the program, the two inner rows of vertical prestressing tendons and the third row tendon are included in this model , and the 210 circumferential tendons are compressed into 5 mock-tendons at a radius of 255 in. , which corresponds to the inner row of the three actual layers of these tendons. Each of these circumferen- tial tendons mocks up 42 of the actual tendons, which is accomplished by using a cross-section and 42 times the actual value. The tendon locations are marked on the figure. The computational model has 1116 nodes, 190 20-node bricks, 20 tendon elements, 3049 degrees-of-freedom and the stiffness matrix has a band-width of 733. The NDNSAP-C program was run on the CRAY-XMP and averaged 9 minutes and 30 seconds per time step. A baseline calculation, described below, was made for an intact tendon struc- ture. This calculation will serve as the point of departure. In this calculation all of the tendons were pretensioned to 70% of the ultimate stress of the steel , which is 168 000 psi. An internal pressure of 705 psi, which is the operating pressure of the helium coolant in this HTGR was applied. The stresses were evalu- ated from one day to 30 000 days in logarithmic steps of one-half decade. Figures 7, 8, and 9 show the stresses on the mid-height plane at 300 days. The induced axial stress is approximate by uniform, about -750 psi between the inner wall and the ring of circumferential prestress, and is smaller outside. The radial stress varies from the internal pressure of -705 psi applied at the inner wall to a value of about -1200 psi at the inner row of the circumferential prestress. Outside this ring, the radial stress in this calculation goes rapidly to a tensile value and then returns to zero at the outside wall , with a concentration near the buttress - 10 - and cylindrical shell junction. This region of tensile stress is an artifact of the finite element tendon mockup and the very limited number of circumferential tendons in the model . In actuality, the 210 tendons are distributed over three rings of 70 tendons each with the outer ring coincident with the exterior boundary on the innter buttress boundary. The radial stresses between the inner tendon ring and the exterior tendon ring will vary approximately linearly from its peak compres- sive value to zero. The decay of the average vertical hold-down force are shown in Figure 10 for the center vertical tendon in the second row. Three calculations show the effects of removing three different tendons at the time of prestressing. The first calculation removes the center vertical tendon in the second row. The second removes the center vertical tendon nearest the thinnest part of the wall . The third calculation removes one inner circumferential tendon at mid-height. The stress contour plots for the three cases with a tendon removed are not plotted as the significant differences do not show up well on the scale as plotted. The differences are best illustrated by the several plots along the stated arcs or lines. The main results for each of these cases are best shown through comparisons of the three stress components, viz., the axial , radial and tangential stresses along certain arcs. Figures 11 through 13 show these components along an arc from 09 to 609 at mid-height through the outer ring of Gaussian integration points in the second radial row of elements. This arc is shown in Figure 6. On each graph, the three cases with one tendon missing are labeled I , II and III in the order described above. Figures 14 through 16 and 17 through 19 show the radial dependence at mid-height at o = 309 and o = 609, respectively. The former is the line from the inner wall through the middle of the buttress and the latter passes through the thinnest section of the wall . When one vertical tendon is removed, there are two basic effects. The first is that the mean axial compressive stress throughout the cross-section is reduced by approximately (0) (0 where bs is the stress in the tendon and As and Ac are the cross-section areas of a steel tendon and the concrete, respective- - ly. When one vertical tendon is removed the mean axial stress changes by *40 psi , and the tangential stress by -10 psi . When a circumferential tendon is removed, the mean tangential stress within the circumferential tendon radius - changes by •30 psi and the axial stress by -10 psi . The radial stress distribution is only slightly changed. - 11 - Ch— after V CONCLUSIONS AND RECOMMENDATIONS • This report represents a finite element calculation of the Concrete creep in a model of the Fort St. Vrain PCRV after prestressing and the subsequent stress redistribution when selected tendons are degraded or removed. The complexity of the finite element mesh used in these calculations was a compromise between exact physical detail and short computational time. Modelling of the complete PCRV with all of its prestressing tendons would have resulted in inordinately long compu- tational times, so fifteen vertical and five circumferential prestressing tendons were included in a 60' sector of the PCRV wall . The calculations of the creep and stress redistribution were accomplished for four cases. These were creep of the model with all tendons present, and creep of the model with one selected tendon missing. These latter three cases eliminated, in turn, the vertical tendon at 30' in the second row, the leftmost vertical tendon in the inner row, and one mid-height circumferential tendon in the inner ring. The viscoelastic creep model in NONSAP-C is a limited constitutive model in which only the uniaxial compressive Young' s modulus is allowed to vary with time. Poisson's ratio remains constant throughout the deformational process. Further, the creep rate varies linearly with stress. In view of the limited creep test data on the Fort St. Vrain concrete specimens, which is limited to uniaxial stress versus uniaxial strain data over a very limited compressive stress range, no further sophistication in the triaxial viscoelastic modelling is warranted. The upshot of the limited modelling accuracy imposed by computer memory and time limitations and the restricted viscoelastic creep model in the program (and further justified by the limited experimental data available) is that the numerical values of the calculated stresses are not to be interpreted as an exact representa- tion of the actual stresses in the structure. Rather, the orders of magnitude, their trends, and, in particular, the differences between the successive calcula- tions, are significant. The most significant departure in the structural stresses due to the limited numer of tendons included in the model is the tangential stress component. The solution without any tendons present indicates that the maximum value of o0 is - 12 - about +1850 psi at the ' er wall . With the mockup of t circumferential tendons in the model , an approximate uniform compression of -3700 psi is added. Comparison of the unprestressed creep and creep with all prestressing tendons in place show that the axial and the circumferential stress are altered by the impo- sition of an approximately uniform compression given by - As as g os while the lateral stress increment will vary from zero to approximately (-1/4N)(au), dependent on the overall boundary conditions. The radial stress is not affected significantly. See the discussion in the previous section for the adequacy of the radial stress component computation beyond the circumferential tendon ring. If one vertical tendon is removed, the axial stress redistribution has two components. The mean axial compression is reduced by approximately (1/N) (ao) and there is an additional local reduction of the same order confined to one vertical tendon spacing. The circumferential compression is reduced by the same amount locally, and the radial stress is only slightly affected if at all . Calculations for the one-dimensional deformation of a concrete column pre- stressed by a steel tendon yielded estimates of the time evaluation of the hold- down and lift-off forces in the column. The steel was assumed elastic and the concrete was assumed viscoelastic with a constant log time creep law under constant stress. The slowly time varying deformation and stress were evaluated numerically from the resulting Volterra integral equation. Two sets of material parameters were used in this evaluation, first, the design parameters of the steel and concrete and, then, the measured parameters as reported in the updated FSAR. The hold-down force decreases with time after the initial application of the prestressing load; this decay, for the first 20,000 days is approximately linear with log (t + 1) and is dependent on the number of broken,strands. The hold-down force decreases approx- imately linearly with the number of broken strands in the range up to 20. One case was evaluated to investigate the later breakage of tendon strands; initially, no broken strands were present at load application, then 20 strands were broken at t • 365 days. The hold-down force time history shifted rapidly from the M • 0 line to the M . 20 line and by t . 400 days, the subsequent equal deformation was indistinguishable from the M s 20 at t . 0. - 13 - In all cases, the p. Acted hold-down forces stayed well above 980 kips for the range of time 0 to 20,000 days, for the number of broken tendon strands from M : 0 to 20 and for the range of material parameters and temperatures (75e to 150'F) considered. The main conclusions of this evaluation are in two parts. First, we state the conclusions for the partial degradation of an isolated prestressing tendon. Then we examine the complete failure of one tendon in a region of multiple tendon. The hold-down force and the related concrete stress decreases proportionally with the fraction of broken tendon strands up to approximately one quarter of the strands broken. This decrease is independent of the time the strands break. The hold-down force decreases linearly with the logarithm of time due to creep. For the extreme combination of concrete creep at 150'F and tendon degradation over the lifetime of the structure, the prediction is that the hold-down force at the end of the lifetime will be higher than the minimum safety requirements. When we consider complete degradation or total failure of one tendon, we must consider the adjacent tendons. At the time of tendon failure there will be a uniform readjustment of the compressive stress. This mean change in the fraction of the compressive stress is the hold-down force for the one tendon divided by the total cross-section area. In addition, there is a slight local tension added in a region near the removed tendon, whose typical size is a circle with radius equal to the original tendon spacing. The magnitude of this localized stress change is 20 to 50% of the mean stress change. Over the lifetime of the structure, the change in the remaining hold-down forces is the same as if all tendons were present, and thus this change in the stress state will still be in the safe range. Combining these two sets of conclusions, either partial or complete degradation of one prestressing tendon will slightly reduce the mean compressive stress in the same direction in the concrete. In addition, a very slight local efffect is noticed. This stress and the hold-down force will remain within structurally safe bounds. - 14 - ACKNOWLEDGEMENT Mary Marshall , 0-13, painstakingly created the NONSAP-C mesh and performed the code runs on the PCRV wall structure. REFERENCES 1. Letter: Warembourg, D. W. , (Manager, Nuclear Production), Fort St. Vrain Nuclear Generating Station, to Collins, John (Regional Administrator), U.S. Nuclear Regulatory Commission, Region IV, April 12, 1984. 2. Fort St. Vrain Nuclear Generating Station, Updated Final Safety Analysis Report, Vol . 6, Appendix E-20. 3. Anderson, C. A. , Smith, P. D., and Carruthers, L. M. , "NONSAP-C: A Non- linear Stress Analysis Program for Concrete Containments Under Static, Dynamic, and Long-Term Loadings," Los Alamos LA-7496-MS, Rev. 1, NUREG/CR-0416, (1982) . - 15 - STEEL CAP STEEL TENDON • Es* As c CONCRETE COLUMN E , A , E• C C C z Fig . 1 . Schematic Diagram of the concrete Column With a Steel Prestressing Tendon . I I a Ls- 0 r In O n 41 VI 10 C C d ICI E 1- 0 'r ++ '- O Q N e O N C an N Y LL IL. L O 1- W L 1!7 pO § : O II I N V) F f f ).- i CV Q O'O G LI. C t a .- so + 0,- O W CC O N 2_ S O F O L r O �v -8 x o ff LL "Ow LL o dLL y 0 so N II .s n I- 7.— 10 C CA 10 .. 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C9 CT hr _E c a•+ F- C - e+ . o+ Y o 1 QI I I I I I 1 m El N 1 1 - - 58 I I I I I 7 7 Usd) SS3WS 9VIIN3DNtll SSINS No. : 6835 IN 85-68 UNITED STATES NUCLEAR REGULATORY COMMISSION WELD COUNTY COM!�fi, !Pr;{(t OFFICE OF INSPECTION AND ENFORCEME WASHINGTON, D.C. 20555 `i n August 14, 1985 �; AUG 2 3 1985 �"I ' L__ IE INFORMATION NOTICE NO. 85-68: DIESEL GENERATOR FAILURE AT CALVERT CLIFFS NUCLEAR STATION UNIT 1 Addressees: All nuclear power reactor facilities holding an operating license (OL) or a construction permit (CP). Purpose: This information notice is provided to alert licensees of a potentially signif- icant safety problem involving cracked interpolar connecting bars that connect the damper circuit of each rotor pole to the damper circuit of the adjacent rotor pole of the emergency diesel generator. It is suggested that recipients review this information for applicability to their facilities and consider actions, if appropriate, to preclude similar problems at their facilities. However, suggestions contained in this informa- tion notice do not constitute NRC requirements; therefore, no specific action or written response is required. Description of Circumstances: On May 14, 1985, Calvert Cliffs Unit 1 was shut down for refueling. An overspeed test on the No. 11 emergency diesel generator (EDG) was initiated with the generator not connected to its essential bus. During the test, plant personnel who were stationed in the proximity of the EDG* heard loud, metallic rapping sounds coming from the generator section of the EDG and immediately terminated the overspeed test. Examination of the generator revealed that insulation on the stator windings had been rubbed and abraded to the point where one stator winding had been exposed. Further examination determined the cause of the failure to be a broken interpolar connecting bar on the rotor. The bar initially had broken free on one end and damaged the stator windings. Because the generator was not connected to its essential bus, electrical field *Calvert Cliffs is a 2-unit PWR station with three EDGs that are designated as follows: Unit 1 (EDG 11), Unit 2 (EDG 21) , and shared (EDG 12). 8508120603 IN 85-68 August 14, 1985 Page 2 of 3 excitation had not been applied and no electric arcing occurred when the stator windings were damaged. The damaged generator was replaced with a spare genera- tor and a detailed investigation was undertaken involving the vendor, Louis Allis, and the licensee. A metallurgical analysis on the failed interpolar connecting bar determined the predominant cause of failure to be high stresses resulting from periodic centrifugal loading. To a lesser degree, bending of the connecting bar, during initial installation, and thermal expansion also were considered to be contrib- uting factors in the failure. The licensee reports that the analysis indicated the problem was of a design nature and not the result of a material defect. Radiographic testing on the damaged generator showed several cracks in the remaining interconnecting bars. The investigation into the generator design determined that the interpolar connections between rotor poles are not necessary (a) if the EDGs are not operated in parallel , which could cause power pulsations between units, and (b) if the plant is operating with a balanced three-phase electrical load. Calvert Cliffs does not operate its EDGs in parallel with each other. Analysis shows that the potential three-phase electrical load unbalance factor (i.e. , single-phase load) does not exceed 10% of the emergency three-phase load; therefore, it is not considered a concern. Consequently, the licensee initiated a program to remove the interpolar connecting bars from the three EDGs in service at Calvert Cliffs Units 1 and 2. On the basis of these findings, the licensee removed EDG No. 21 from Unit 2 service on May 26, 1985 to determine if the interpolar connections on that EDG were degraded in a similar manner. Test results indicated cracks were evident. On this basis, the EDG No. 12 (which applies emergency power to either Units 1 or 2) was declared inoperable. Under this set of conditions, Calvert Cliffs Unit 2, which was operating at 100% full power, started shutting down. In order to continue power operation of Unit 2, the replacement EDG from Unit 1 was connected and aligned to Unit 2 to provide the necessary emergency power source and the shutdown was terminated. The licensee' s corrective action, to terminate the limiting condition for operation, discussed above, involved the immediate removal of the interpolar connections from EDGs 12 and 21. Postmodification qualification testing, conducted on both EDGs, proved to be satisfactory. The licensee' s longer term corrective action plans are to remove the interpolar connections from the replacement (spare) EDG. Discussion: Typically, a diesel generator that employs the continuous damper circuit design (i . e. , using interpolar connecting bars) uses 16 connecting bars. These bars are installed so that each of the eight rotor poles on the generator has two connecting bars, one installed on the front and one on the back of each rotor pole. Noncontinuous damper circuit design does not employ interpolar connec- tions between the damper circuits on the rotor poles. IN 85-68 August 14, 1985 Page 3 of 3 The safety concern of this event is the common cause failure mechanism if similar cracked connecting bars exist on all diesel generators at a nuclear power plant (as was the case at Calvert Cliffs). Cracked connecting bars can lead to a condition that adversely affects the operating voltages which are necessary to operate essential equipment during accident conditions. Because of the generic implications of the Calvert Cliffs event, the vendor, Louis Allis, issued a 10 CFR 21 report on the potential problem to the NRC on May 21, 1985. A followup letter, dated May 29, 1985, from Louis Allis to the NRC, briefly discussed the problem and identified other facilities that use similar Louis Allis generator units. Colt Industries also forwarded a Louis Allis report dated June 3, 1985, that identified the major cause for interpolar connecting bars cracking and provided a basis for requesting removal of the interconnecting bars from similarly designed generators in service at other facilities. The June 3, 1985 report from Louis Allis to Colt Industries also was sent to end-users of. the Louis Allis generator units. As a result of the early notifications discussed above, the licensees of TMI/1, Vermont Yankee, and Peach Bottom Units 2 and 3 all report that they have either removed the interpolar connections or have established plans to have them removed from the affected generator units in service at their facilities. Although no similar interpolar connecting bar failures have been reported, related to other generator units, generators supplied by manufacturers other than Louis Allis may have similar design features. Additionally, licensees are reminded that if corrective actions are taken to resolve observed degradation of interpolar connecting bars this action should be reported to the NRC using the existing reporting requirements specified in 10 CFR 50. 72 and 10 CFR 50. 73. No specific action or written response is required by this notice. If you have any questions about this matter, please contact the Regional Administrator of the appropriate Regional Office or this office. , : 1 Edward L. Jordan, Director Division of Emergency Preparedness and lEngineering Response Office' of Inspection and Enforcement Technical Contact: Vincent D. Thomas, IE (301) 492-4755 Attachment: List of Recently Issued IE Information Notices Attachment 1 IN 85-42, Rev. 1 August 14, 1985 LIST OF RECENTLY ISSUED IE INFORMATION NOTICES Information Date of Notice No. Subject Issue Issued to 85-42 Loose Phosphor In Panasonic 8/12/85 Materials and fuel Rev. 1 800 Series Badge Thermo- cycle licensees luminescent Dosimeter (TLD) Elements 85-67 Valve-Shaft-To-Actuator Key 8/8/85 All power reactor May Fall Out Of Place When facilities holding Mounted Below Horizontal Axis an OL or CP 85-66 Discrepancies Between 8/7/85 All power reactor As-Built Construction facilities holding Drawings And Equipment an OL or CP Installations 85-65 Crack Growth In Steam 7/31/85 All PWR facilities Generator Girth Welds holding an OL or CP 85-64 BBC Brown Boveri Low-Voltage 7/26/85 All power reactor K-Line Circuit Breakers, With facilities holding Deficient Overcurrent Trip an OL or CP Devices Models OD-4 and 5 85-63 Potential for Common-Mode 7/25/85 All power reactor Failure of Standby Gas Treat- facilities holding ment System on Loss of Off- an OL or CP Site Power 85-62 Backup Telephone Numbers to 7/23/85 All power reactor the NRC Operations Center facilities holding an OL and certain fuel facilities 85-61 Misadministrations to Patients 7/22/85 All power reactor Undergoing Thyroid Scans facilities holding an OL and certain fuel facilities 85-60 Defective Negative Pressure 7/17/85 All power reactor Air-Purifying, Fuel Facepiece facilities holding Respirators an OL or CP OL = Operating License CP = Construction Permit tot moo, We o UNITED STATES ? S NUCLEAR REGULATORY COMMISSION .: WASHINGTON,D.C.20555 MD goyim C "iNcr ***„x44 August 2, 1985 A 9"f,4S L� cik2) Docket No. 50-267 AUG o 3198 GRZI-LEY, CALo. Mr. 0. R. Lee, Vice President Electric Production Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201 Dear Mr. Lee:The Commission has issued the enclosed Order confirming your commitments to implement those post-TMI related items set forth in Supplement 1 to NUREG-0737, "Requirements for Emergency Response Capability," which was sent to you by Generic Letter 82-33 dated December 17, 1982. This Order is based on commitments contained in your letters dated April 14, 1983 (P-83147), July 5, 1983 (P-83234), March 9, 1984 (P-84079) , and April 30, 1985 (P-85143) . The Order references your letters and, in its attachments, contains lists of the applicable NUREG-0737 items with your schedular commitments. Some of the items set forth in the attachment to the Order are subject to post-implementation review and inspection. Our post-implementation review and/or the development of Technical Specifications may identify alterations to your method of implementing and maintaining the requirements. Any identi- fied alterations will be the subject of future correspondence. A copy of this Order is being filed with the Office of the Federal Register for publication. Sincerely, Als Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing Enclosure: Confirmatory Order cc w/enclosure See next page - Mr. 0. R. Lee, Vice President Fort St. Vrain Public Service Company of Colorado C. K. Millen Albert J. Hazle, Director ` Senior Vice President Radiation Control Division_ Public Service Company Department of Health of Colorado 4210 East 11th Avenue P. 0. Box 840 Denver, Colorado 802220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A GA Technologies, Inc. P. 0. Box 840 Denver, Colorado 80201 J. K. Fuller, Vice President Public Service Company of Colorado P. 0. Box 840 Denver, Colorado 80201 G. L. Plumee NRC Senior Resident Inspector P. 0. Box 640 Platteville, Colorado 80651 Kelley, Stansfield & O'Donnell Public Service Company Building Room 900 550 15th Street Denver, Colorado 80202 R. D. Martin, Regional Administrator Region IV Parkway Central Plaza Building • 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1800 Lincoln Street Denver, Colorado 80651 J. W. Gahm Nuclear Production Manager Public Service Company of Colorado P. 0. Box 368 Platteville, Colorado 80651 UNITED STATES OF AMERICA NUCLEAR REGULAIURY COMMISSION t In the Matter of PUBLIC SERVICE COMPANY ) Docket No. 50-267 OF COLORADO (Fort St. Vrain Nuclear Generating Plant)• ORDER MODIFYING LICENSE CONFIRMING LICENSEE COMMITMENTS ON POST-TMI RELATED ISSUES Public Service Company of Colorado (PSC or the licensee) is the holder of Facility Operating License No. DPR-34 which authorizes the operation of the Fort St. Vrain Nuclear Generating Station (the facility) at a steady-state power level not in excess of 842 megawatts thermal . The facility is a high temperature gas-cooled reactor (HTGR) located at the licensee's site in Weld County, Colorado. II. Following the accident at Three Mile Island Unit No. 2 (TMI-2) on March 28, 1979, the Nuclear Regulatory Commission (NRC) staff developed a number of • proposed requirements to be implemented on operating reactors and on plants under construction. These requirements include Operational Safety, Siting and Design, and Emergency Preparedness and are intended to provide substantial additional protection in the operation of nuclear facilities and significant upgrading of emergency response capability based on the experience from the accident at TMI-2 and the official studies and investigations of the accident. - The requirements are set forth in NUREG-0737, "Clarification of TMI Action Plan Requirements," and in Supplement 1 to NUREG-0737, "Requirements for Emergency Response Capability." Among these requirements are a number of items consisting of emergency response facility operability, emergency procedure implementation, addition of instrumentation, possible control room design modifications, and specific information to be submitted. _ On December 17, 1982, a letter (Generic Letter 82-33) was sent to all licensees of operating reactors, applicants for operating licenses, and holders of construction permits enclosing Supplement 1 to NUREG-0737. In this letter, operating reactor licensees and holders of construction permits were requested to furnish the following information, pursuant to 10 CFR 50.54(f), no later than April 15, 1983: (1) A proposed schedule for completing each of the basic requirements for the items identified in Supplement 1 to NUREG-0737, and (2) A description of plans for phased implementation and integration of emergency response activities including training. III. PSC responded to Generic Letter 82-33 by letter dated April 14, 1983. By letters dated July 5, 1983, March 9, 1984, and April 30, 1985, PSC clarified some dates as a result of negotiations with the NRC staff. In these submittals, PSC made commitments to: 1) complete all of the basic requirements prior to returning to operation following the fourth refueling outage (currently scheduled to begin in January 1987) ; 2) perform a verification and validation effort to evaluate the completed activities; and 3) resolve any problems detected during the verification and validation effort by the fifth refueling outage. . 3 - The attached Table summarizing PSC's scheduler commitments or status was developed by the NRC staff from the Generic Letter and the information provided by PSC. PSC's commitments include (1) dates for providing required submittals to the NRC, and (2) dates for implementing certain requirements. The NRC staff reviewed PSC's April 14, 1983 letter and entered into = negotiations with the licensee regarding schedules for meeting the require- ments of Supplement 1 to NUREG-0737. As a result of these negotiations, the licensee clarified certain dates by letters dated July 5, 1983, March 9, 1984, and April 30, 1985. The NRC staff finds that the modified dates are reasonable, achievable dates for meeting the Commission requirements. The NRC staff concludes that the schedule proposed by the licensee will provide the required upgrading of the licensee's emergency response capability. In view of the foregoing, I have determined that the implementation of PSC's commitments are required in the interest of the public health and safety and should, therefore, be confirmed by an immediately effective Order. • IV. Accordingly, pursuant to Sections 103, 161i, 161o, and 182 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR 2.204 and 10 CFR Part 50, IT IS HEREBY ORDERED,_ EFFECTIVE IMMEDIATELY, THAT FACILITY OPERATING LICENSE NO. DPR-34 IS MODIFIED TO PROVIDE THAT THE LICENSEE SHALL: - c. 4 - Implement the specific items described in the Attachment to this ORDER in the manner described in PSC's submittals noted in ` Section III herein no later than the dates in the Attachment. Extensions of time for completing these items may be granted by the Director, Division of Licensing, NRR, for good cause shown. V. The licensee or any other person with an adversely affected interest may request a hearing on this Order within 20 days of the date of publication of this Order in the Federal Register. Any request for a hearing should be addressed to the Regional Administrator, Region IV, U.S. Nuclear Regulatory Commission, 611 Ryan Plaza Drive, Suite 1000, Arlington, Texas 76011. A copy should also be sent to the Executive Legal Director, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555. A REQUEST FOR HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER. If a hearing is to be held, the Commission will issue an Order designating the time and place of any such hearing. If a hearing is held concerning this Order, the issue to be considered at the hearing shall be whether the licensee should comply with the requirements set forth in Section IV of this Order. This Order is effective upon issuance. FOR THE NUCLEAR REGULATORY COMMISSION AlCd - ug . 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