ML20112F945
| ML20112F945 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/06/1983 |
| From: | GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20112F941 | List: |
| References | |
| 907079, NUDOCS 8501160111 | |
| Download: ML20112F945 (113) | |
Text
{{#Wiki_filter:. _. _ . . . . . _ _ _._ -- GA 1483 (REV.10/82) ROLL 2011 N/C ISSUE
SUMMARY
TIRE O R&D 3 POSTIRRADIATION EXAMINATION AND EVALUATION OF APPROVAL LEVEL FSV FUEL ELEMENT 1-2415 O S GN DISCIPLINE SYSTEM 00C. TYPE PROJECT 00CUMENT NO. ISSUE N0/LTR. N 18 RTE ionn QUALITY ASSURANCE LEVEL SAFETY CLASSIFICATION SEISMIC CATEGORY ELECTRICAL CLASSIFICATION I FSV-I FSV-I N/A APPROVAL ISSUE PREP ^ ISSUE DATE DESCRIPTION / BY FUNDING APPLICABLE N EERING QA CWBS NO. PROJECT PROJECT 1 i r - o f. uc w, A' J
/1 gggg F.Mc ord 4.J. R. . Turner WBS #6415 020 204 Sche fel yyn) ff,f 4.p5 stansfield
.a t,$Ihf Sy- J.A. Simon
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N/C NOV 0 71984 F Mc ord .W. .//. s. - Dn M W, D.J.Kowal WBS #2970 104
'S effel D.Pettycor 1 y gy O$&.n/ t hy
$ N 0.M , f/g Stansfi(1d I
R.F.T er CONTINUE ON GA FORM 14851 NEXTINDENTURED DOCUMENTS (Ref-906770) PSC P0 N-5052 a ' $V I C) 8501160111 841217 DR ADOCK 05000 REV SH REV SH 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 REV SH I 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 i . lPAGE1 0F 113
TABLE OF CONTEN'I3 f.Aat
- 1. INTRODU CTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
'4
- 2. ELEMENT DESCRIPTION................................................ 9 i ,
3- .
- 3. 9 g IRRADIATION HIST 0RY................................................
- 4. RESULTS OF THE DESTRUCTIVE EXAMINATION. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 4.1 Disassembly of the Element.................................... 10
{ 4.1.1 Dri111ag............................................. 10 , 4.1.2 Removal of Top Three S11oe s. . . . . . . . . . . . . . . . . . . . . . . . . . 11 ' 13 4.1.3 Remov al of Fue l Rod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4.1.3.1 Initial Push-out Force Measurements.................. 14 ! 4.1.3.2 Stack Length Nessarements............................ 14 l 4.2 Examination of Psel Rods...................................... 15 4.2.1 Vi sua l Exam i na t i on. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 4.2.2 Fuel Stack Diame ter Mes sarement s. . . . . . . . . . . . . . . . . . . . . 16 4, .2 .3 Nota 11ographic Examination Results................... 17 4 4.3 Power Distribution Measurements............................... 20 ! 4.3.1 Description of the Gamma Scanning System............. 21 4.3.2 Axial Power Distributions............................ 22 l 4.3.3 Radial Power Distributions........................... 22 4.3.4 Fue l Rod Homo ge ne i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 A 23 . f/ 4.4 Barasp Nessarements........................................... w 4.5 Fluence kkasarements.......................................... 24 l
- 5. SECTIONING OF THE BL0CE............................................ 24 5.1 Cutting of the B1ook.......................................... 25 5.2 Activity of the B1ook......................................... 28 5.3 Colle c ting Na te rial Prope r ty Sample s. . . . . . . . . . . . . . . . . . . . . . . . . . 31
- 6. SUNNARY AND CONCLUSIONS............................................ 32 6.1 Fuel Element Performance Conclusions.......................... 33 6.2 Verification of HTUR Core De si gn Me thods . . . . . . . . . . . . . . . . . . . . . . 35 l
Page 2 l
OA TECGNOL08IES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./L13. N/C I I
- 7. ACKN OW LEDG MENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37
- 8. REFERENCES......................................................... 38 LIST OF TABLES 1-1 Compari son of Blocks w ith Cracke d Webs. . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 g
1-2 Obj ectives of the De structive Examination on Element 1-2415. . . . . . . 42 9 2 -1 Nominal Preirradiation Fuel Rod Attributes for FSV Fuel Element 1-2415............................................................ 43 2-2 Nominal Fissile Fuel Particle Mean Attributes for FSV Fuel Element S/N: 1-2415...................................................... 44 2-3 Naminal Fertile Fuel Particle Mean Attributes for FSV Fuel Element 1-2415............................................................ 45 4-1 Initial Fuel Stack Pushout Force Me a surement s. . . . . . . . . . . . . . . . . . . . . 46 4-2 Fuel Stack Length Measurements.................................... 47 4-3 Fuel Rod Diame te r Me a surement s for S tack 307. . . . . . . . . . . . . . . . . . . . . . 48 4-4 Fuel Rod Diame ter Me asurement s for Stack 308. . . . . . . . . . . . . . . . . . . . . . 49 4-5 Fuel Hole Diame ter Measurements for Hole s 307 and 308. . . . . . . . . . . . . 50 4-6 Fissile Particle Me ta11ographic Examination Re sul t s. . . . . . . . . . . . . . . 51 4-7 Fertile Particle Meta 11ographic Examination Resn1ts............... 52 A 4-8 Comparison of Measured and Calculated Composite Buranp. . . . . . . . . . . . 53 j/ LIST OF FIGURES w 1-1 FSV Element side f ace ide nti fica tion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 1-2 Core location of fuel element 1-2415.............................. 55 2-1 FSV fuel element 1-2415........................................... 56 4-1 Results of the first removed slab, crack is barely visible because of the dust from sawing. Notice the cut dowel pin and the graphite dust from the sawing.............................................. 57 Page 3
8A TECENOL08IE5 I N C. TITI2: POSTTBBAnIATION EIANINATION AND EVALUATION OF FSV PUEL RtRMNT 1-2415 l DOM NO. 907079 lISSOE NO./LTR. N/C i 1 I 4-2 Cracking pa t te ra nado r B dowe 1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 4-3 Remains of the B dowel pia following first out.................... 59 l 4-4 Fuel plugs still intact at an approximate axial location of 1 inch 60 l ! A 4-5 Bottom of the 3rd top slice at axial location 1-1/8 inch from the top. Note the this remalas of the fuel plugs and the cracking
- pattern........................................................... 61 I
t 4-d De s t i na t i o n o f f ue l rod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 4-7 Fee l rod pu shing ope r a tion se tup. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 4-8 TWo scales monated on each side of the deal receiving trough were use d t o me a s ure the s t ac k 1 ems t h s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 ) l 4-9 Compo sit e pho tograph s of fuel s tack 307. . . . . . . . . . . . . . . . . . . . . . . . . . . 65 l 4-10 Compo si te pho tographs of f uel s t ack 308. . . . . . . . . . . . . . . . . . . . . . . . . . . 68 4-11 Rod 14 of stack $9 showing more than normal surface debonding..... 71 4-12 Noamatching half rods which sugge st breakage at assembly. . . . . . . . . . 72 l 4-13 Loca tion of fuel rod used la me ta11ography. . . . . . . . . . . . . . . . . . . . . . . . 73 l 4-14 Photomicrographs representative of matrix phase of irradiated rod 13 from stack 308. The time average fuel temperature was approximately - 7650C at sa average fast finance of 1.6 x 1025 m/m3 (E > 29 fJ)gygg 74 4-15 Representative photomicrographs of composite of radial cross section o f f uel rod f rom FSV element 1-2415 ( le f t si de of rod) . . . . . . . . . . . . 75 i A 4-16 Photomicrographs of fissile and fertile particles.................. 78 l" [ . 4-17 Photomicrographs of fissile particle (Th,U)C2 showing fuel dis- ! persion. Note the dense phase la the buffer coatings and the low-density IRrc coating.............................................. 79 l 4-18 Photomicrographs showlas fission product interaction on the sic in the fissile and fertile particles................................. 80 4-19 Photomicrograph of an as-namaf actured defective particle in an irra-l disted inel rod. Note total coating failure with the matriz pressed 4 in the particle. Also, note that the kernel has been teached out.. 81 I i Page 4 1
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l 0A TOCDNOL00IES I N C. TITI2: POSTIRRADIATION BIANINATION AND EVALUATION OF FSV FUEL ELEIENT 1-2415 lBOCBIENT NO. 907079 115805 NO./L1R. N/C 1 l l j 4-20 Photomicrographs showing la pile partial coating f ailure. .. ... .. .. 82 i 4-21 Schematio of the gamma scanning system setup...................... 83 I 4-22 Measured and calculated axial power di stribution. . . . . . . . . . . . . . . . . . 84 i j s 4-23 Messared time averaged radial power normalized to the average Co-j 137 conte nt of the six f uel s t acks. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 85 q 4-2 4 Typ l o a l C s-13 7 t r ac e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 4-25 Normalized measured thermal finance distribution. . . . . . . . . . . . . . . . . . 87 5-1 Sectioning of PSV element S/N: 1-2415............................ 88 5-2 Composite of photographs at axial location 2-1/8 inches from top. This is a bottom view of the 1 inch slab removed from section 1... 89 5-3 Composite at axial location 9-1/8 laches. This is a bottom view of section 1. The orack on the B face extending into coolant hole 319 is c1 ear...................................................... 90 i { 5-4 Composite at axial location 9-1/8 at a different angle. Bottom j view of section 1. Note orack into coolant hole 319.............. 91 J . 5-5 Compo s i t e a t 16-1/ 8 inc h e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 92 4 l 5-6 Photographs at axial location 17-1/8 inches....................... 93 5-7 Photographs of the bottom portion of the 6 inch test slab. . . . .. . . . 94 5-8 Composite photograph of the central coolant channel on the B face i edge of the 6 inch test slab. The side faces were removed. Notice A the orack extending into the graphite web beyond the coolant chan- . ! nel. The extent of this orack i s not clear. . . . . . . . . . . . . . . . . . . . . . . 95 t
- 5-9 Composite photographs of the top of sostion 4 at axial location 23-1/8 isokes..................................................... 96 5-10 Composite photographs of a portion of sostion 4 taken at axial loostion 24-1/8 inches............................................ 97 5-11 Composite of the bottom of the block near the B dowel shows the ersoking patters on the bottom. The chip on the thin web between hole 319 sad dowel socket was consed by handling. The other cracks between the this webs and the sockets are believed to have been pr e se nt pr ior to de l111m s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 98 Page 5
GA TECEN0LOGIE5 INC. TITI2: POSTIRRADIATION EIAMINATION AND BVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBIENT NO. 907079 lISSEE NO./LTR. N/C I l 5-12 Botton view of the 2.5 he autoradiograph of the third slab removed from the top at axial location 1.125 inches. Light areas indicate regions of higher radioactivity................................... 99 s-a3 Autoradiograph of top of test slab. Dark areas indicate regions of higher radioactivity........................................... 100 5-14 Autoradiograph of bottom of test s1ab............................. 101 d 5-15 Location of holes for cleaning.................................... 102 5-16 Autoradiograph of the top of the test slab (6 inches) after re-amisag and bru shing sele c ted ho1e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 5-17 Autoradiograph of the bottom of the test slab (6 inches) after reaming and brushing of selected ho1e s. . . . . . . . . . . . . . . . . . . . . . . . . . . 104 5-18 Autoradiograph of bottom of test slab af ter roaming from top while vacnaming the dust from the botton............................... 105 5-19 Te st sl ab f or materi al prope rty sample s. . . . . . . . . . . . . . . . . . . . . . . . . . 106 5-20 Location of cored samples from the A face........................ 107 5-21 Loca tion of cored sample s f rom the B f ace . . . . . . . . . . . . . . . . . . . . . . . . 108 5-22 Location of cored samples from the C face........................ 109 5-23 Location of cored samples from the D face........................ 110 j i 5-24 Location of cored samples from the E face........................ 111 l l 5-25 Location of cored samples from the F face........................ 112 i 6-1 Port St. Vrain reactor power history: SURVEY analysis of cycles 1 and 2............................................................. 113 l I i { i Page 6
l GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL RIJ M NT 1-2415 lDOCUM NT NO. 907079 lISSEE NO./L72.N/C i i
- 1. INTRODUCTION Fif ty-four fuel and reflector elements from the Fort St. Vrain (FSV)* l core segment 2 underwent nondestructive metrological and visual inspections in A
April of 1982 in the hot service facilities at the Fort St. Vrain reactor site. These inspections were part of the FSV Fuel Surveillance Program spon-
- sored by the Department of Energy. The results of the inspections were re-ported in Reference 1.
Two of the fuel elements, serial numbers 1-2415 and 1-0172, were dis-covered to have similarly aligned hairline cracks extending the entire axial length in the face adjacent to the large single dowel (face B). See Figure 1-1. The cracked web in PSV fuel element 1-2415 was discovered during the on-site surveillance, whereas the crack in element 1-0172 was discovered later via viewing video tape s and photographs from the surveillance and was com- ; firmed by the visual examination performed in GA Technologies hot cell (Ref. 2). Both elements were loosted la refueling region 8, coluna 5. Element 1-2415 was located la axial layer 6 (active layer 3). See Figure 1-2. Element 1-0172 was located in axial layer 7 (active layer 4), directly beneath element 1-2415. l The preirradiation inspection reports indicated that neither element
- was cracked prior to insertion into the core, and there was no record of any damage having been done during hand 11ag. In addition, the fact that the O
oracks were colinear suggests that they developed during irradiation. There-l fore, to characterise better the oracks and determine why they occurred, the l two elements with oracked webs and three other elements of potential interest were shipped to the GA Technologies Inc. hot cell for further visual inspoo-tion and possible destructive examinations. The visual examinations have been completed, and the results were reported in Referesco 2. l *The Fort St. Vrain Neelear Generating Station is owned and operated by Public service Cgspany of Colorado (PSC). I Page 7 i
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2 8A TECENOLOGIES I N C. TI112: POSTIRRADIATION EXAMINATION AND EVALUATION OF PSV FUEL ELEMENT 1-2415 j lN NO. 907079 IISSUE NO./LTR. N/C I l Element 1-2415 was selected for destructive postirradiation examination
- l. (PIE). This element was chosen because it exhibited the larger crack and ex-periemoed larger fast finance and metrological changes than element 1-0172.
Table 1-1 compares the two elements with oracked webs. This report covers the i results of the destructive examination performed upon fuel element 1-2415. l I The destructive examination included the followlag tasks: l o drilling through to the fuel holes on the bottom of the block f o removing the fuel hole plugs by sawing off the top of the fuel element o removias the fuel stacks o measuring the feel stack push-out force and stack lengths t j o relocating the fuel stacks f ato a receptacle blo,k l o measuring the feel rod diameters on stacks numbered 307 and 30s o examinias the microstruotare of a fuel rod by oor aographic section l o gamma sessaing six fuel stacks I o sectioning and photographing of the empty element o cleaning and autoradiographing of a six inch section of the empty element o collecting the graphite axial cores for material property tests o collecting the graphite radial (transystse) cores for material property tests A The purpose of the destructive esamination was to provide esperimental data on the irradiated element, to verify its performance, and to seguire in-i, plie data. Tensile strength, thermal espansivity and thermal diffusivity sam-pies were taken from the irradiated graphite block. The samples were measured and characterized by the graphita laboratory at GA Technologies. The results of sample characterization are published as separate reports (Refs. 3 and 4). The fuel rod-graphite lateraction was measured by measuring fuel stack push- , out force. Gamma seams of a central fuel stack la each face were taken so Page a
I SA TECENGL0eIES I N C. l TI112: POSTIREADIATION EIANINATION AND EVALUATION OF FSV PUEL RI.RMRNT 1-2415 l lna-arf NO. 907079 lISSEE NO./LR* N/C l i l i 1 i i that power and barnap could be o.iarac teriz ed. These data were analyzed and made available for other scoping analyses that were done to develop possible scenarios leading to the oracked web. A section of graphite from the emptied i { element was cleased to do a qualitative study on the feasibility of cleaning spent graphite blocks for burial as low level waste. Spoolfic objectives of the destructive examination are given in Table 1-2. 4
- 2. ELE 8ENT DESCRIPTION The element was manuf actured from graphite grade B-3 27 by the Great Lake s Carbon Corporation. The manuf acture's serial number was 1-2415 with a fuel assembly type number of 101 and a drawlas member: 90R1801-101. Element S/N: 1-2415 consisted of a standard fuel body havlag 210 fuel holes, 6 bara-able poison holes, and 108 coolant chamaels (See Figure 2-1) . There was no lumped barnable poison lasta11ed in the element. The element oostataed 3132 fuel rods ocasisting of (Th,U)C2 RIS0* fissile particles and ThC2 3150 fer-tile particles bonded together by a carbonaceous matrix. The fuel rods were f carbonized at 18000C. The moniaal preirradiation dimensions of the fuel rods were 12.5 mm (0.49 in.) in disseter and 49.3 mm (1.94 la.) in length. Fuel rod and fuel particle monimal attributes are given in Tables 2-1, 2-2, and 2-3.
i A
- 3. IRRADIATION BISTORY I
l I ~ l Fort St. Vrain feel element 1-2415 was irradiated in region 8, oolana 5, oore layer 6 (third active layer from the top of the core). The element
*In the MISO particle de sign, a layer of sic is sandwiched between two 3
layers of high-density pyrolytto earbon, which provides a composite pressure vessel to retala gaseous fission prodnots. The Sic costing also provides a l ! barrier asalast the diffusion of metallio fission prodsets and increases the mechanical and dimensional stability of the particle during irradiation. As l inner low-density, or buf fer, coating adjacent to the fuel korsel provides a void volume to assommodate fission gases and kernet swelling and, la addi-tion, attenuates fission prodnet recoils. l Page 9 L_____.__ _ _ . _ ,-. _ _ _ _ _ .?
1 GA TECENOLOGIES I N C. TIT 12: POSTIERADIATION EIANINATION AND EVALUATION OF FSV FUEL RI.RastNT 1-2415 i lBOCB M NO. 907079 l ISSUE NO./L1R. . N/C I l was irradiated la thin oore location for cycles 1 and 2. From July 1976 to February 1,1979 (cycle 1), the element acommulated 174 effective full power days (EPPD)*. During eyolo 2 which was from May 26,1979 to May 13,1981, the element gained another 189 EPPD. For both cycles the element ensulative
- burasp was 363 EFPD. The average fast fluence ** accanalated by the element was approximately 1.55 x 1025 m/m2 (E > 29 fJ)mgR with a peak of about 1.8 x 4 1025 m/m2 The tima and volume-averaged graphite temperature was 6500C. The '
peak graphite temperature was 700cc. The time averaged fuel temperature was 7650C"'. (See section 6.2 for a description of the HIGE design codes which sissisted the irradiation history of the element).
- 4. RESULTS OF THE DESTRUCTIVE E3ANINATION 4.1 Disassemb1v of the Element The disassembly of FSV element 1-2415 was carried out la the high level hot sell at GA Teoksologies, Inc. While perfornias the destructive examina-tion, the procedural requirements of the test procedure (Ref. 5) were adhered to as closely as possible. Bowever, there were a few minor deviations which will be noted in the appropriate sections. The disassembly of the element is j described below.
1 4.1.1 Drillina - I i
- A d:1111 3 tool was developed and used to drill through to the fuel I holes on the bottom surface of the element. The tool was f abricated with l
l l 'An RFPD is the equivalent of 1 day of operation at full power (842 Nt). l
** Fast mestros fluence was obtained from the GATT code fuel secountability I ealesistloss of systes 1 and 2. The fast mestron flue nce was volume averaged.
I
"*These temperatures were obtained from the SURVEY code estestations based on
- the GAUGE eode depletion.
i I Page 10
l i l 6A TECEN0L0SIES I N C. TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCB E NT NO. 907079 l ISSUE NO./LTR.N/C I l aligning prongs so that it could be positioned and aligned using the coolant holes as a guide. Once the aligning prongs were in place, the six fuel holes surrounding a given coolant hole could be drilled without relocating the tool. The fuel holes were drilled to a depth of 8.9 mm (0.35 inch) and diameter of
- 9.5 mm (0.375 inch). [ NOTE: Reference 5 specified a drilling depth of 6.35 mm (0.25 in.), but at this depth the tip of the drill just pierced the conical 4 tip of the fuel hole. This did not permit enough of a hole clearance to in-sert the push rod. Therefore, an extra 2.5 mm (0.10 inch) was drilled.] All of the fuel holes except for the fuel holes under the dowel sockets were dril-led to this depth. For the fuel holes under the dowels, the drill was re-adj usted to penetrate to a depth of 35.3 mm (1.4. laches) from the bottom surface. When all of the fuel holes had been drilled, the graphite dust was vacuumed.
4.1.2 Removal of Ton Three Slices A modified band saw was placed in the high level hot cell. The saw was modified to fit into the hot cell and to clamp and hold down a FEV type ele-ment in a cradle during the cutting. Also installed on the band saw was a cleanup and vacuuming system. This saw was used to do all of the cutting of the irradiated block. In order to remove the fuel stacks, a slab approximately 12.7 mm (0.5 inch) thick was removed from the top of the block (NUTE: The removed slab was
- not the same thickness on all faces because the saw blade was slightly mis-aligned. The leading blade edge cut a thickness approximately 12.7 mm (0.5 inch) on the corner of faces B and C, and at the final corner of faces E and F the thickness was approximately 15.9 mm (0.625 inch)] . It took 3 minutes and 18 seconds for the band saw to make the cut.
Removal of the top slab bared the f uel in the holes which were not located under the dowels. Figure (4-1) shows the results of the removed slab. Page 11
i SA TBCON@LOGIES I D C. TI112: POETIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL Rt.RMFNT 1-2415 l lBOC M NT NO. 907079 l ISSUE NO./LTR. N/C
- 1 I f
I The bottom section of the B dowel fell out and exposed the sootion of the , l block which was located under the B dowel. Because the B dowel had previously obscured the cracked web area on the top surface, the orack had been charac-terized based upon its appearance on the top and bottom surfaces of the ele-
- meat and was believed to extend from the central coolant hole on the B face 8 edge to the oestral coolant channel of the B dowel. Instead, as can be seen
; in Figure 4-2, the oracked web on the top surf ace was shown, upon removal of i
j the B dowel, to extend from the outer edge of the central coolant chamael of the B face to feel hole 307 (located under the dowel) and to the central l j ooolant chamael of the B dowel. i The section of the B dowel which was removed from the top surface ! appeared not to be cracked, which suggested no problem with the fit of the i dowel-socket of element 1-2415 and the element above it (08.05.F.05) . Figure 4-3 shows the remains of the B dowel which had covered the B s ocke t. The stacking demonstration which was discussed la Reference 2 demonstrated that the dowel-socket clearance for element 1-2415 and the element directly beneath it (1-0172) was adequate. T' The removal of the first top slios permitted the removal of the fuel stacks whlok were not loosted under the dowels. When all of this fuel was o pushed ont (see section 4.1.3), a second slice was out from the top to espose the feel located mader the dowel [ NOTE: To correct for the ansven catting of the first slice, the leading blade edge out a thickness approximately 12.7 mm (0.5 inoh) on the corner of the faces B and C, and at the flaal corner of face E sad F the thickness was approximately 9.5 mm (0.375 inch). This second slice dif fered from the outtias sehese outlined la Reference 5 which called for a 15.9 mm (0.625 inoh) thick slab]. When the second slice was removed Page 12
GA TECENOLOGIES I N C. TITI2: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBE NT NO. 907079 l ISSUE NO./LTE. N/C I I from the block, the bottom of the fuel plugs under the dowels were still in-tact as shown in Figure 4-4. To expose the fuel under the dowels, it was decided to cut another thin
~
slice rather than push in the thin remains of the graphite plugs which were left in after the second slice. Therefore, a third slice of approximately 3.2
- um (0.125 inch) thickness was removed. Figure 4-5 shows this slice. The fuel under the dowels was exposed and the removal of the fuel proceeded as des-cribed in Section 4.1.3.
I 4.1.3 Removal of Fuel Rods All of the fuel rods were removed from element 1-2415. Most of the fuel was relocated into a unused H-327 graphite receptacle blocks however, some stacks were set aside for other tests. Figure 4-6 shows the destination of all the fuel. Two fuel stacks were removed f rom element 1-2415 at each pushing step I by pushing them from the back of the block into a metered dual-tube receiving ! trough attached to the front of the element. The initial forces needed to move each of the two stacks were measured with a force gauge. A 863.6 mm ('34 inch) long metal push rod was then used to further push the stack into the o trough. When filled, the trough was removed from the element, held upright, and visually examined via the Ko11 morgan periscope in the hot cell. Each of
~
the two stack lengths was measured and the condition of the rods determined. The rods were then transferred via the trough to the receptacle block. The i l dual-tube receiving trough was then reattached to the front of the element and the pushing procedure was continued. Details of the rod condition are discus-Page 13
GA TECENOL0GIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCDIENT NO. 907079 l ISSUE NO./LTR. N/C I l sed in section 4.2.1. The measurements of initial push-out forces and fuel stack lengths are discussed below. 4.1.3.1 Initial Push-out Force Measurements A 22.7 Kg (50 pound) push pull type force gauge was used to measure
- the initial push-out force. A 6.35 mm (0.25 inch) diameter 127.0 mm (5 inch) long extension push rod was at ttched to the gauge. The setup of the pushing operation is shown in Figure 4-7.
The initial push-out forces measured for fuel element 1-2415 are given in Table 4-1. The average measured initial force was 4.9 Kg (10.7 lbs) with a range from 0.7 Kg (1.5 lbs) to 15.0 Kg (33.0 lbs). In a few cases, the pushout forces are ac t believed to be true values, but are thought to be in-correctly higher because of pushrod-hole misalignment and the breakthrough force needed to remove the remains of the conical tip of the fuel hole. When-ever the operator suspected interference before encountering the fuel, this fact was noted. The sustaining forces needed to further push the rods out of the block were not measured however, none of the stacks of fered any signifi-cant resistance to pushing. Therefore, it was concluded that there was no appreciable fuel rod - fuel block interaction in element 1-2415. This was further corroborated by the small amount of debris collected from the emptied block (see Section 5.2). 4.1.3.2 Stack Lenath Nessurements A very rigorous determination of stack lengths was not performed since this was not a procharacterized surveillance element
- and therefore only nomi-
' Thirty two initial core surveillance elements were procharacterized before insertion into the initial core (Ref. 6).
Page 14
GA TECENOL06IES I N C. TIITE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./L1R. N/C l I nal preirradiation dimensions data were available. However, the measurements which were performed gave sufficient evidence to establish a general trend. The rods were initially pushed into the receiving trough as discussed in section 4.1.3. Two scales were mounted on each side of the dual trough as shown in Figure 4-8. The scales were in graduations of 0.02 inch and 0.01 inch each. The heights of the stacks in the metered trough were read via the Ko11 morgan and then recorded. The trough was then unloaded and cleaned of any loose particles before reinsertion into the unloading element. The results of the measurements are shown in Table 4-2. An approximate 4.0 mm (0.14 inch) ! decrease in average stack height from the nominal preirradiated stack height of 739.1 mm (29.1 inches) was observed for the element. 4.2 Examination of Fuel Rod _s 4.2.1 Visual Examination Following fuel stack removal, the fuel rods were individually viewed using the hot cell Ko11 morgan periscope system. Photographs were taken of all the rods in stacks 307 and 308. Figures 4-9 and 4-10 show these stacks, re spe ct ively. These photographs were typical of all the fuel rods taken from the element. In general, the appearance of the fuel rods was good, although consid-erable chipping at the end caps and some debonding of particles from the fuel rod surface were obse rved. Rods which were chipped and slightly debonded on the ends appeared similar to the acceptable preitradiated rods, and are not believed to have been damaged by reactor operation or by the extraction of the Page 15 L
GA TECENOL06IES I N C. TIT 12: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l DOM NO. 907079 l ISSUE NO./LTR. N/C I I stacks (see Section 4.1.3.1). One rod in particular (rod 14 of stack 59) was observed to have more than normal surface debonding. Figure 4-11 shows this rod. Six rods (0.2%) of the 3132 rods removed from element 1-2415 were broken. Three rods (0.1%) of the rods are thought to have been broken during
- irradiation or fuel removal becan'se the broken pieces matched. However, the remaining 0.1% were probably broken prior to assembly of the element because three of the broken fuel rods consisted of pieces with nonnatching fracture surf ace which would not match if turned relative to each other. Figure 4-12 shows these examples. During subsequent fuel handling operations (loading and unloading rods into pleziglass tubes, loading rods for other tests, etc.),
eight additional rods were broken by the operators. 4.2.2 Fuel Stack Diameter Measurements The diameters of the fuel rods and fuel holes of fuel stacks 307 and 308 were measured. The se two fuel stacks and' holes were selected because they were located near and/or on the cracked area as discussed in section 4.1.2. i The fuel rod diameter measurements were performed using a ring gauge, a dial gauge indicator and a standard calibrated pin. Diameter measurements were made at the top and bottom of each fuel rod at approximately 00 and 900 Tables 4-3 and 4-4 show the results of the measurements of stacks 307 and 308,
~
respectively. The fuel hole diameter measurements were performed using a depth gauge, a dial gauge indicator and a standard calibrated bore hole. Each hole was measured at depths of 1, 8,16, 22 and 28 inches at approximately 00 and 900 Table 4-5 shows the results of these measurements. Page 16
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUMNT NO. 907079 l ISSUE NO./LTR. g/C l i The results of the diameter measurements of the fuel rods and fuel holes show that there was sufficient fuel rod-fuel hole clearance in each hole. The average fuel hole diameter was 12.7 mm (0.499 inch) for both 307 and 308. The average fuel rod diameter was 12.5 mm (0.493 inches) for both fuel stacks. There was a 0.16 mm (6 mil) fuel hole-fuel rod diametrical gap
- for each hole.
4.2.3 Meta 11oaranhic Examination Results Irradiated fuel rod 13 from fuel stack 308 was subjected to metallo-graphic examination because it was expected to have a high temperature history relative to the fuel elsewhere in the element and, thus, would provide an indication of thermal effects in the fuel earlier than fuel rods from lower temperature portions of the element. This fuel rod, which was one rod length above the bottom of the fuel hole came from the axial location in the block where the power peaked (Ref. 7) . Fuel stack 308 was chosen because it was ! located under the dowel pin of f ace B adj acent to the cracted web. Figure t
- 4-13 shows the location of fuel stack 308 in reference to the cracked web.
Rod 13 was mounted in resin, ground, and polished in the metallographic hot cell. Prior to examination, all polished sections were passivated with a 50/50 solution of HNO3 and H2O to decrease the rate of hydrolysis of the ThC2 kernels. The entire polished surf ace of the rod was then examined. The fuel rod matrix was in good condition. Normal minor cracking was observed in the matriz end cap. The microstructure of the matrix after irra-diation is shown in Figure 4-14. The irradiated microstructure was similar to the microstructure observed for FSV fuel rods irradiated in the FSV fuel proof capsule F-30 (Ref. 8). The measured macroporosity for rod 13 was 17.5%. This
*Diametral gap = fuel hole diameter - fuel rod diameter.
Page 17 l l
d 4 i GA TECEN0LOGIGS I N C.
; TITI2: POSTIRRADIATION BIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 i
lN NO. 907079 lISSEE NO./LM. N/C I I i ! value is within the range (14 to 29%) of macroporosities observed for fuel rods from capsule F-30. Aa exasaple of a radial cross section showing the l macroporosity la the matrix is shows la Figure 4-15. '
- The results of the metallographic examination of fuel rod 13 from fuel
! stack 308 are presented in Table s 4-6 and 4-7. The irradiation performance of ' I , the fissile and fertile 11150 coated particles was satisf actory as evidenced by the moderate coating iallures and good thermal performance. The micro- , l structures of typical particles af ter being expo sed to a peak f ast neutrom , finence of approximately 1.8 x 1025 m/m$ (E > 29 fJ)H1GR and a time-average fuel temperature of 7650C, are showa la Figure 4-16. A total of 231 fissile ! and 184 fertile particles were examined. l Fuel dispersion was observed in 75% of the fissile am/. $% of the fer-tile particles. An example of fuel dispersion is shown la Fig. 4-17. Fuel j dispersion can be caused by chlorine which diffuses through a permeable IPyC into the buffer during the sic coating operation. Production records indicate { that this fuel %1 relatively low density IPyC and tendesoy to exhibit fuel di spe rsion. Psel made for segnient 4 and beyond did not have such low IPyC l density so fuel dispersion is espected to be reduced in the segments with more severe exposure than segments 1, 2, and 3. ,. i
# The primary purpose of the metallography was to examine the fuel for thermal effects. In this regard the chemical behavior of the 11150 particle a was neceptable. Eersel migration was not observed. Interaction of the sic i
conting with fission products was obse rved in both particle type s. Figure 4-18 shows examples of the fission prodset interaction upon the Sic. Fission prodset interaction into the Sic was observed in 3.9% of the (Th,U)C2 parti-eles and 3.3% of the ThC2 particles. The reaction penetrated about 5 pm into the'51C layer for both particle types. Page 18
9 i GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIANIN!. TION AND EVALUATION OF FSV FUEL Rt.RMRNT 1-2415 l lBOM NO. 907079 lISSER NO./LTE. N/C
- I -l This kind of interaction is expected la carbide fuel where rare earth fission products are released from the kernel and interact with the sic (Ref.
) 9). However, the observed depth of interaction (~5 pm) was larger than the (1 i
- pa expected value based on the time-temperature history of the fuel. This 1 .
rapid penetration rate may have been associated with fuel dispersion caused by chloriae trapped in the buffer layer during the sic coating operation as dis-
- cussed above, and/or a higher operating temperature than the average calen-1sted for the fuel element. Penetration depths on the order of 10 to 15 pm l
might lead to some volatile fission product metal release, but total coating 1l failure and fission gas release would not lacrease (Ref. 9). The penetration depth measured for these particles is significantly less than 10-15 pm. This phenomenon will be monitored la fatare FSV fuel surveillance so that any la-fluence on core performance can be properly assessed. l In the course of metallography the mechanical condition of the coating can be observed, but concissions from these data are difficult to draw because coating f ailure can be caused during manuf acture and during the grinding and polishing procedure in connection with making a metallographic monat. There was evidence that some of the coating f ailures can be attributed to as-mana-factured failures which occurred during coating or rod fabrication. This con-clasion is supported by the appe arance of failed particles. An example of this type f ailed particle is shown in Figure 4-19, where the particle had the j appearance of having been crushed and fuel rod matrix was pressed into the l coating cracks at the time of manufacture. In this case as-maasf actured feil-are rather than in pile fallare was indicated. However, salike the metallo-graphic results of Segment 1 (Ref. 10), there was partial coating failure which may have occurred la-pile as seen la Figure 4-20. Irradiation induced shrinkage of the pyrocarbon was apparently enhanced by the additional esposare of Segment 2, and both IPyC and OPyC failures were observed as espected from accelerated capsule resnits (Ref. 8). For purposes of comparing with prior 1 Page 19
8A TECEN0LOGIES I N C. T1112: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lN NO. 907079 lISSOE NO./LM. N/C I l work, conats of observed f ailed coatings are reported in Tables 4-6 and 4-7. These values should be considered apper limits for reasons stated above. The total coating f ailure with res,alting gas release can be expected to be very low because circulating gas activity has remained consistently low and indi-cates virtually no particle failure. However, as noted in Tables 4-6 and 4-7, the actual number of total coating fallares was not determined due to hydroly-sis of the kernels in the inel rod monat. 4.3 Power Distribution Measurements Six fuel stacks from FSV fuel element 1-2415 were samma scammed to e de termine the relative distributions of measurable radioisotopes. Figure 4-6 shows the location of these inel stacks. The data from these gamma scans can provide informaton on the power distributions in the element during irradia-tion. Power data was analyzed and made available for other analyses that were planned to evaluate various scenarios leading to the cracked web. The data was piso used to verify anclear design calculations and to characterize better the material performance. The measured radioisotope distribution of Cs-137 is useful. Cs-137, which has short-lived precursors, is approximated as a direct-yield isotope from the fission of U-235 and U-233. It has a half-life of 30 years, which is f far greater than the irradiation period of the element. If significant l amounts of Cs-137 did not escape from the fuel, its distribution is represen-tative of the time-averaged power distribution of the element (Ref. 11). Sinoe the feel was HISO coated and experienced relatively low temperature ((10000C during cycles 1 and 2) and low flue nce (I2.0 x 1025 m/m2 (E > 29 1 fJ) git;g] , the assumption that very little Cs-137 escaped the fuel is ressosa-ble. Page 20
J GA TECENOLOGIES I N C. i a TITI2: POST 1RRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 InamamT NO. 907079 lISSEE NO./L1R. !UC i i I l The measured radioisotope distribution of Zr-95 is representative of ( the normalized power distribution for the last 50 to 200 days
- of reactor operation (Ref. 11). Since no Zr-95 could be found in the fuel because of its short half-life (65.5 days), no end-of-life power profiles conid be ob-taine d.
A discussloa of the samma scanning setup and resnits of the gamma spec-troscopy is presented below. 4.3.1 Descrintion of the Ga - Scanalna System The samma scanning system consisted of a gamma scanning rig, an alumi-man collimator, an EG h G ORTEC high purity germanian (Ge) samma de t e c t or, 1 Nuclear Data (ND) 6620 data acquisition system, and a single channel analyzer ! (SCA)-ratemeter recorder system. An overview of the samma scanning system is ) shown la Figure 4-21. I i , i The samma scanning rig, which was placed in the low-level hot cell, positioned, rotated and slowly moved a thia-walled plexiglass tube in a direc-tion perpendicular to the collimator. The pleziglass tube housed a fuel stack of individsel rods separated by stainless steel spacers. A collimator, coa-
. structed of aluminum and having a length of 1079.5 mm (42.5 inches) and a 21.6 x 2.5 mm (0.85 x 0.10 inch) slot, was aligned with an out-of-cell high parity Ge samma de tector. The signal from the detector was sent to the ND6620 data acquisition sy stem and to the SCA ratemeter recorder system. The ND6620 collected entire spectra and stored them on disks where they were accessed and analysed by various ND spectral analysis programs. The SCA rateme ter recorder i system monitored and traced the 662 kev Cs-137 photo peak. ,
I
*Three half-line s assumed.
~
l Page 21 l i -- _ _ ~ _
GA TECENOLOGIE5 I N C. TITLE: POSTIRRADIATION EXANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCDE NT NO. 907079 lISSEE NO./LTR. N/C l I 4.3.2 Axial Power Distributions Axial scanning was performed by moving the fuel stack slowly past the i collimator. The speed on the gamma scanning ris was set so that spectra could i be acquired at a time interval approximately equal to the length of a fuel
- rod. The acquisition time was approximately 3 minutes'. All scanning was done in a semiautomated mode mader the direction of the ND6620 computer.
I Time averaged measured and calculated axial power distributions are shown in Figure 4-22. The measured time-average profile is a normalized Cs-137 profile obtained by averaging the results of the six axial side-f ace scans. The calculsted element time averaged power profile was obtained from the SURVEY code (see Section 6.2). The trends between measured and calculated proflies is in fair agreement except at the top end, where the measured pro-file is higher by approximately 9%. 1 4.3.3 Radial Power Distributions 1 i The normalized radial distributions of Cs-137 in FSV element 1-2415 are shown in Figure 4-23. ne normalized Cs-137 distribution came from the samma scanning of the fuel rods in each of the six measured stacks. The maximum j . observed tilt (difference between the highest and lowest relative power) was t 21% which was across faces B and E. 4.3.4 Fuel Rod Homoneneity The distribution of Cs-137 and other measured radioisotopes along the length of individual fuel rods was observed to be markedly U-shaped, with the activity near the ends being almost twice the activity in the middle. A portion of a typical Cs-137 trace for an axial scan is shown in Figure 4-24.
'The dead time, which is the time it take s the ADC (part of the ND6620 system) to convert the analog pulses to a digital data, was about 1.7% of the acquisition time.
Page 22
i i i GA TECENOLOGIES I N C. , TIT 12: POSTIRRADIATION EIANINATION AND EVALUATION OF PSV FUEL ELEMENT 1-2415 j i l l ISSUE NO./L1R. N/C l i lBOC8MNT NO. 907079 l l All of the scanned rods were observed to have this U-shaped profile, which suggested a manufacturing process that tended to segregate the fissile parti-l clos toward the ends of the rods. The same phenomena was observed in Segment Gamma scanning of unirradiated fuel (Reference 10) showed the U-235 i 1 rods. distributions of the fuel rods to have the same shape as the observed Cs-137 traces. a 4.4 Burnun Measurements Cs-137 can be used to establish a composite burnap which is defined as the number of fissions occurring per initial heavy metal atom (FINA) provided there is no significant Cs-137 loss. Since it is believed that there was no significant Cs-137 loss (see Section 4.3), the fuel burnup was measured via gamma spe ctrome try. Absolute calibration of the gamma scanning system using the known Cs-137 inventory of a fuel rod from the R2-K13 experiment (Ref.12) permitted composite barnap to be determined. The barnap can be calculated by: A FINAc " (Do+ Th o) Y \ l . where FIMAc = composite barnap A = activity in atoms
= where K
CPS = Cs-137 counts (counts per sec) L = rod length (inches) CPS-in K = detector efficiency Aton Page 23
GA TECENOLOGIES I N C. l ! TITIA: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBIWrf NO. 907079 lISSUB NO./LTE. N/C I l l l Do = number of atoms of uranium at beginning of life (BOL) The = number of atoms of thorium at beginning of life (BOL) Y = fractional fission yield of Cs-137 from U-235' l l
. Br.rnap data obtained from samma spectrometry are presented in Table 4-8.
Examination of the burnups determined by gamma spectroscopy yielded an element average composite burnup of 2.39%, whereas the calculated element average composite barnap using data from the fuel accountability derived GATT code (see Section 6.2) was 2.02%. The GATT code fertile and fissile particle burnups were ~0.61% and ~8.35%, respectively. 4.5 Fluence Measurements Cs-137 is a direct-yield isotope from the fission of U-235 and U-233, and its formation follows the thermal fluence profile linearly. Cs-134 is formed indirectly frosi fission thrcui;h the decay of Ie-133 to Cs-133, followed - by neutron activation. Therefore, Cs-134 is formed in a quadratic correlation with the fluence, and the Cs-134/Cs-137 ratio should follow the fluence in a l linear correlation. Figure 4-25 shows the normalized Cs-134/Cs-137 distribu-tion, which provides an indication of the relative thermal flux along the
- length of the element.
i
. 5. SECTIONIIJ OF TEE BihCK One of thr< main goals of the de structive examination was to acquire irradiated graphite samples for various tests and studies. With all of the
- Assume the same for U-233 Page 24 l
GA TECENOLOGIES I N C. l TITI2: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 IBOCBIW ff NO. 907079 l ISSUE NO./L'11. N/C I I fuel removed, the element was cut into sections using the modified band saw de scribed in se ction 4.1.2. The de tails of the cutting of the block, the activity of the empty element, and the collection of the irradiated graphite samples are discussed below. 5.1 Cuttina of the B19g.h The empty irradiated graphite element was placed in the low level hot cell and marked at the appropriate places for cutting. Placement marks were also made on the cradle of the band saw which was located in the high level cell. The marked block was then transferred to the high level cell and positioned in the marked cradle. Figure 5-1 shows the cutting scheme. [ Note: the cutting plan was modified from the specified plan of Reference 5 so that 25.4 mm (1.0 inch) thick slabs could be given to LANL' for testing.] In an effort to characterize further the axial and radial cracking j patterns of the element, photographs were taken of each cut section. However, in many cases cracking was very difficult to characterize because of the in-terference of the graphite dust from the sawing and because of the markings from the teeth of the saw blade. Nevertheless, a discussion of the cracking ! patterns by axial sections follows: [ Note: 0.0 mm (0.0 inch top): 792.5 as (31.2 inches bottom)] l l o Axial Section at (0 - 1 1/8) Inches l See section 4.1.2. o Axial Section at 21/8 Inches o The cracked web from the outer B edge coolant hole (319) was still visible.
*Los Alamos National Laboratory (LANL) is a national research center located in Los Alamos, New Mexico.
Page 25
l GA TECENOLOGIE5 I N C. TITI2: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCDENT NO. 907079 l15855 NO./L1R. N/C I 1 o The continuation of the crack from coolant hole 319 to fuel hole 307 was still visible. o The crack from fuel hole 3 07 to the central coolant hole under the dowel (295) was still visible. o No other cracks were discernible.
, Figure 5-2 shows a composite of photographs at this axial location.
1 I o Azial Section at 9 1/8 Inches i o At this location only the crack extending from the outer edge of the B face to coolant hole 319 was clearly visible. o No other cracks were clearly distinguishable. Figures 5-3 and 5-4 show axial location 91/8 inches. i o Axial Section at 161/8 Ing]Lu i o At this location only the crack extending from the outer edge of the B face to coolant hole 319 was clearly visible. o No other cracks were clearly distinguishable. Figure 5-5 shows axial location 161/8 inches.
- o Axial Section at 171/8 Inches o At this location only the crack extending frce the outer edge of the B
( f ace to coolant hole 319 was clearly visible, o No other cracks were distinguishable. 1 l l [ Page 26 i
I GA TECENOLOGIES I N C. TITI2: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUM NT NO. 907079 l ISSUE NO./LTR. N/C I I Figure 5-6 shows axial location 171/8 inches. o Axial Section from (17 1/8 - 23 1/8) Inches (6 Inch Material Pronerty 3 Snecimen Slab) l o The crack was still visible extending from the B f ace edge to coolant hole 319. o With the B face removed from the slab, the crack extending toward fuel i hole 307 was visible, but its extent inward was not characterized. o No other cracks were clearly distinguishable. Figures 5-7 and 5-8 show the cracking pattern at axial locations 17 1/8 - 23 1/8. o Axial Section 231/8 (Too of 4th Section) o Only the crack extending from the B f ace edge to coolant hole 319 was clearly distinguishable. o No other cracks were clearly distinguishable. I Figure 5-9 shows another view of the axial location at 23 1/8 inches. o Axial Section 241/8 (Too of 4th Section af ter the 1 Inch Slab Removed) o Only the crack extending from B face edge to coolant hole 319 was clearly discernible. o No other crack were clearly visible. Page 27
( - GA TECENOLOGIES I N C. TI112: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCU M NO. 907079 l ISSUE NO./LTI.N/C I l Figure 5-10 shows axial location 241/8 inches. o Axial Location 31.2 Inches (Bottom of Block) o The crack extended from the B face edge through coolant hole 319 and down into the socket midway between fuel holes 307 and 308.
~
o The chip on inner edge of coolant hole 319 and the begining edge of the dowel socket was made by the operator during handling. o Cracks were also observed on the thin webs between the coolant hole and the beginning edge of the dowel socket for coolant holes 319, 283, 267 I (two cracks here), and 280. [A careful examination, under a macro-scope, of photographs take n dering the visual examination (Ref. 2) corroborated that these cracks were indeed present before drilling.] Figure 5-11 shows a composite of the cracking pattern at the bottom of the block around and in the B dowel socket. 5.2 Activity of the Block The samma activity of the loaded element at 3 feet was approximately 4 80 R/ hr. When the fuel was removed from the block, the activity was taken by Health Physics and the surface samma and beta activity averaged 0.56 R/hr for face E, 0.57 R/hr for face D, 0.82 R/hr for face B, 0.60 R/hr for top surf ace (open empty fuel holes) and 0.60 R/hr at the bottom surface. These readings were believed to be high and possibly caused by loose particle contamination from the fuel rod removal and, also, from the loose debris in the block. Therefore, the block surf aces were lightly brushed to remove any clinging de-bris and then the block was placed in a vertical position so that any loose debris could fall out. As expected, since the rods pushed out easily, only a Page 28
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCDENT NO. 907079 l ISSUE NO./LTR.N/C l I small amount of debris fell out of the block onto the collection napkin. The surface samma and beta activity of the element was rechecked and the activity was 0.43 R/hr for face D, 0.57 R/hr for face E, 0.72 R/hr for face B, 0.50 R/hr for the top surf ace and 0.42 for the bottom surface. The light brushing and collecting of debris only reduced the activity on the average by 17%, which was not statistically significant. Autoradiographs were taken to see if the activity was surf ace contami-nation or the results of fission product migration into the graphite. Figure 5-12 shows a 2.5 hour autoradiograph of the 0.125 inch slab removed from the top of the block at axial location 1.125 inches. The autoradiograph was of the bottcm of the slab, and the bright exposed areas showed surf ace contamina-tion of the fuel and coolant holes. Also, there was contamination in the cracked area, but the graphite webs appeared relatively clean. This slab was not repre se ntative of the entire block since it came from a section of the block which was unfueled. The contamination in the, fuel holes could have been caused from hydrolysis of preirradiated fuel during the assembly into the graphite block, since fuel manuf&cturing s'pecifications allow a limited amount of such hydrolysis. Since the material property samples were to come from the 6 inch slab (see Section 5.1), the activity of this slab was monitored. The gamma plus beta surf ace activity ranged from 0.90 R/hr to 1.2 R/hr on the top surf ace and from 0.8 to 1.0 R/hr on the bottom surf ace. The top and bottom surfaces had the open empty fuel holes. The range of gamma plus beta activity at contact on the side face was 0.27 R/hr to 0.48 R/hr. Top and botton 15 minute autoradiographs (Figures 5-13 and 5-14, res-pectively) were taken of the 6 f ach (152.4 mm) slab to determine the distrib-ution of the activity. The activity which is revealed by dark regions in the 1 j Page 29
GA TECENOLOGIES I N C. TITI2: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBE NT NO. 907079 l ISSUE NO./LTR. N/C 1 I negative print again appeared to be on the surfaces of the fuel holes and not into the graphite webs. Therefore, a small scale cleaning task was launched to clean the hole surfaces. , l The cleaning task consisted of the abrading and reaming of selected fuel and coolant holes. He abrading was done by making two passes in the f
^
same direction through a fuel or coolant hole using a gun barrel cleaning brush. Reamers which took of f 0.13 mm (0.005 inch) wall thickness were used for the reaming. The brushing technique was done for fuel hole 294 and coolant hole 295 both located on the B face and for fuel hole 4 and coolant hole 3 of the E f ace. Raaming was done for fuel hole 293 and coolest hole 292 on the B f ace and for fuel hole 28 and coolant hole 27 on the E f ace. Figure 5-15 shows these locations. Autoradiographs of the top and bottom of the 157.4 mm (6 inch) slab were made prior to cleaning and af ter cleaning. Figures 5-16 and 5-17, show the results of the autoradiographs. Both cleaning techniques appeared to have worked on the top surface, but not on the bottom surf ace. This fact suggested that the cleaning techniques needed refining. An additional step of vacuuming from the bottom of the slab while ream-
- ing from the top downward was added. This allowed the contaminated dust to be removed rather than smeared into the walls as the reaming operation continued downward. Numerous holes on one half of the slab was reamed using this new technique. Figure 5-18 shows an autoradiograph of the bottom of the slab after using the new technique. The holes appeared clean. This further corro-borated that the holes had mostly surf ace contamination. This was an impor-tant fact because it meant that the material property samples could be handled i
Page 30
l l l 6A TECINOLOGIES I N C. l TITIA: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l lBOCBIENT NO. 907079 l ISSUE NO./LTR. N/C 1 l l in the glove box in the GA graphite laboratory, and because it showed that qualitatively it may be feasible to clean spent graphite blocks for possible burial as low level waste. 5.3 Collectina Material Pronerty Samoles Axial and transverse cores and buttons were take n from the 6 inch (152.4 mm) slab from element 1-2415 for material property studies. The but-l tons were used for thermal diffusivity samples using the heat pulse method. The axial and transverse cores were used for axial and transverse coef ficient l of thermal expansivity studies using the silica dilatometer method, and for axial and transverse tensile strength studies using the extensometer method. The graphite laboratory at GA Technologies, Inc. performed testing and pub-
- lished the res,ults in separate reports (Refs. 3 and 4) .
The samples were procured from the side faces of the 152.4 mm (6 inch)
- slab shown in Fit.sre 5-19. A R02A Eberline meter was used to checked the I gamma activity level on a collection of all the samples. The samma activity at contact was less than 3 mr/hr. This further corroborated the conclusions that radioactivity was limited to surface contamination as discussed in the previous section.
The cores were made using a drill press and coring drills. The approx-imate drill speed was 1850 RPM. Figures 5-20 through 5-25 show the location of each sample. ,In procuring the cored samples, the procedural requirements of Reference 5 were adhered to as closely as possible. However, there were a few minor deviations which are listed belcw: l Page 31
=
1 GA TECEN0LOGIE5 I N C. l TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBIENT NO. 907079 l ISSUE NO./LM. N/C 1 I Procedure Section of , i Deviation Ref. 5 Justification )
- 1. Thickness of pieces 5.5.2 Hot Cell personnel requested the l . ~0.437 inch lastead extra thickness for ease in dril-of 0.375 inch ling.
- 2. ~0.270 inch core drill 5.5.3 The obtained cores were large l
used instead of 0.312' enough to be machined to the test
, core drill sample diameter of 0.25 inches.
- 3. Double bagged samples 5.5.3 Requested by Graphite Laboratory in labelled enveloped instead of putting i them in bottles l 4. Cylindrical trans- 5.5.3 Requested by Graphite Laboratory verse cores 4 inches (5 inches not really needed) long instead of 5 inches
, 5. Transverse cores and Figure Because of the crack in the B buttons taken from 6(D) face, transverse cores would not lef t side of piece have been 4 inches long. In or-and axial cores taken der to be consistent in taking
- from right side of the cores on all faces, the piece axial cores had to be removed from the right side.
NGHE
- 1. There was no breakage of the axial cores.
- 2. There y3,1 breakage of the transverse cores, so extra cores were taken.
(See figures 5-20 through 5-25). The drill was sharpened, but this did not stop breakage.
- 3. The irradiated buttons were marked with white marke r showing the transverse direction.
- 6. SUlGIARY AND CONCLUSIONS l
FSV fuel element 1-2415 was irradiated in core location 08.05.F.06 for i 363 EFPD. The element experienced an average fast neutron fluence of about 1.55 x 1025 n/m2 (E > 29 fJ)lrIER and a time-averaged fuel temperature of about 1 i Page 32 1
6A TECENOLOGIE5 I N C. TITIA: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBENT NO. 907079 lISSUB NO./L1R. N/C I I 7650C. The element was ' removed from the reactor during the second refueling in May 1981. During the nondestructive metrological examination in the hot l service facility at FSV la April 1982, the element was discovered to have a crack extending the entfre length on the B face. Consequently, the element was shipped to GA Technologies, Inc. hot cell for de structive examination. The destructive examination of fuel element 1-2415 at GA was performed as part of a DOE-sponsored program. The purpose of this examination was to acquire data to characterisc better the cracking, to verify the performance of the fuel element, and to acquire in-pile data for verification of core design methods. The results of the destructive exaination of fuel element 1-2415 are summarized below. 6.1 Fuel Element Performance ConcInsions Specific observations of the performance of the element are as follows:
- 1. A hairline . axial crack extended the entire length of the element on the B face into coolant hole 319 and fuel hole 307 and then into the central coolant hole of the B dowel (295) on the top surface. This cracking pat-tern was clearly visible for 2-1/8 inches axially into the element, how-ever, af ter this axial location only the crack on the B face edge exten-ding into the central coolant hole (319) was clearly visible. Dust fron l
l the sawing and teeth marks from the saw made further characterization speculative. However, there was evidence that the cracking probably did not stop radially at coolant hole 319. A section of element taken from axial location 17 to 23 inches which had the side faces removed revealed that the crack did indeed go into the graphite web between coolant hole 319 and fuel hole 307, but the radial distance inward conid not be charac-terized. Page 33
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBE NT NO. 907079 l ISSUE NO./L13. N/C I l All other observed blemishes were surface markings only and had not etched the graphite to a harmful extent.
- 2. No evidence of mechanical interaction between fuel rods and fuel hole was found. Except in a few cases, very little force was required to push the fuel rods out of the block. Nisalignment of the force gauge pushrod with the holes, and the breaking through the graphite web using the force gauge pushrod are believed to be cause s of the occasionally high push-out forces.
! 3. The measured gaps between fuel hole and fuel rod of holes 307 and 308 in-dicated that there was sufficient clearance (approximate 6 mil .iametral j gap). l
- 4. The rods were in good condition although minor cracking in the matriz end caps and some debonding of particles from the rod surf ace were observed.
Six fuel rods (0.2%) were observed to be broken when removed from the element. Evidence indicated that half of those were broken prior to irradiation.
- 5. Approximately 231 fissile and 184 fertile particles were examined during metallography. The results of the metallography of fuel rod 13 from stack 308 indicated a little in pile failure. For the (Th,U)C2 and ThC2 parti-clos, respectively, the OPyC coating failure was 0.4% and 7.6% the sic coating f ailure 0.9% and 3.8% and the IPyC coating f ailure 3.0% and 10.3%.
However, evidence indicated that some failed coatings were as-manuf actured f ailures which occurred during coating or fuel rod f abrication.
- 6. The chemical behavior of the particles was acceptable even though chemical interaction of fission products with the sic coatings was observed on both Page 34
f GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION FIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lBOCEIWtr NO. 907079 l ISSUE NO./LTE. N/C 1 I particle type s. The percentage of particles experiencing chemical inter-action of fission products with the sic coating was 3.9% and 3.3% for the fissile and fertile particles, respectively. The depth of the interaction in both particles was approximately 5 pm, which was larger than the (1 pm expected value. The pene tration rate may have been caused by chlorine trapped in the buffer layer during the Sic coating or from higher than calculated fuel temperature. Fuel dispersion and IPyC debonding were ob- ! served in 75% and 5% of the TRISO (Th,U)C2 and ThC2 Particles, re spe c-1 tively. However, neither fission product-sic interaction, fuel di spe r-l sion, nor IPyC debonding detrimentally affected the performance of the particles. 6.2 Verification of HTGR Core Desian Methods Verification of HitiR core design methods cannot be accomplished from j comparisons of experimental observations and design code calculations fer one element. Instead, many such comparisons for core components which have col-l 1ectively experienced a wide range of irradiation conditions are required. The resnits of comparisons between measurements and design code calculations for element 1-2415 should be viewed with this in mind. R1tiR design codes used to calculate irradiation and performance parame-ters for fuel element 1-2415 are sammarized below: l 4 GADEE: A two-dimensional four group neutron diffusion and core depletion (Ref. 13) code which calculates column average power, neutron flux, and nu-clide inventories. Radial power distributions, neutron finance s, and fuel barasp can be obtained from the GAUGE output using the appropriate axial distributions obtained from another source. Page 35
4 GA TECENOLOGIE5 I N C. TIH2: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lBOCUMNT NO. 907079 l ISSUE NO./LTR. N/C I I I The power history for cycles 1 and 2 was represented by 433 time intervals of approximately uniform power. Cycles 1 and 2 were simulated with the GAUGE code using this detailed power history. PEVER: A one-dimensional, multigroup neutron diffusion and depletion (Ref.14) program for calcuating nuclide densities as a function of axial core i location. I BUS-2: A two-dimensional, multigroup neutron diffusion and depletion pro-t (Ref. 15) gram for axial power, neutron flux, and nuclide inventory di stribu-tions for fuel elements influenced by control rods in neighboring i elements. f GATT: A three-dimensional, four group neutron diffusion and core depletion l (Ref. 16) code for calcuatoing nuclide densities as a function of time and axial and radial core location. GATT is used to calculate fuel i I accountability. SERVEY: A program used to calculate temperature and fuel performance. SURVEY also calculates neutron fluences and fuel burnap by bringing together GAUGE, FEVER, and BUG-2 results. A SURVEY analysis of the segment 2 elements was performed based on the GAUGE results. The number of time intervals was further reduced from the GAUGE 433 to 69 for this analysis. Figure 6-1 shows the reactor power history used for the SURVEY analysis. The results of the comparisons are as follows:
- 1. Radial Power Distribution: The observed time-averaged power tilt (rela-tive to element average power) was 11%. The maximum time averaged intra-Page 36
_ . - . . . ~ , , . _ _ _
GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./LTR. N/C 1 I block tilting (difference between the highest and lowest relative power) was 21%. This was between f aces B and E with the higher power on the B face, which was the face with the linear cracked web.
- 2. Axial Power Distribution: The SURVEY calculated element average axial power distribution only had five axial points. Therefore. small power variations are smeared. None thele s s, the trends of both the calculated and measured curves are in fair agreement except at the top end, where the measured power profile is higher by approximately 9%.
- 3. Fuel Burman: The mesured composite burnup (indicative of total power generation) was 15.5% higher than calculated burnup, which was derived from the GATT code.
- 7. ACKNOWLEDGMENTS The author wishes to acknowledge the following contributors to this report.
Element Irradiation: Courtesy of Public Service of Colorado (PSC) Nuclear Calculations: V. Malakhof and S. C. Bachelor Data Acquisition at Hot Cell: H. O. Johnson and Staff, T. L. Smith and C. M. Miller Page 37
i SA TECEN0LOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL PLFMFNT 1-2415 lBOCBINDIT NO. 907079 lISSUB NO./L1R.N/C l l Consnitants on G=-= Scanninn: i Dale Hill and John Saarwein IIRlRA: M. Veerman
- 8. REFERENCES i
- 1. SAURWEIN, J. J. , Nonde structive Examination of 54 Fuel and Reflector Elements from Fort St. Vrain Core Segment 2, GA-A16829, October, 1982.
~
i BREf, f. L. (PSC) letter to JOHN T. COLLINS (NRC), ' ' Transmit tal of Noadest$sctive Fuel and Reflector Elements Examination Reports GA-A16829', 906505, 906577, P-83196, June 2,1983. ) 2. EETIERER, J. W., Visual Examination Results of Segment 2 FSV Fuel Elements 1-2415, 1-0172, 2-2693, 1-0108 and 5-0801' ', GA Document No. 906577, Issue 2, April 4,1983, i BREY, R. L. (PSC) letter to JOHN T. (X)LLINS (NRC), ' ' Transmit tal of Nondestructive Fuel and Reflector Elements Examination Reports GA-A16829, 906505, 906577 , P-83196, Jane 2,1983.
- 3. PRICE, R. J., Test Report: Tensile Properties of Fort St. Vrain Element 1-2415, GA Document No. 907057. Issue A September,1983.
Page 38
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./LTR. N/C 1 1 BREY, H. L. (PSC) letter to JOHN T. COLLINS (NRC), Transmittal of GA Fuel Block Test Reports 907057 and 907155 , P-84109, April 11,1984. , . 4. PRICE, R. J., Test Report: Thermal Properties of Fort St. Vrain Fuel Element 1-2415, GA Document No. 907155, October, 1983. l - BREY, H. L. (PSC) letter to JOHN T. COLLINS (NRC), Transmittal of GA Fuel Block Test Reports 907057 and 907155, P-84109, April 11,1984.
- 5. McCORD, F., Test Procedure for the Destructive Examination of Fort St. Vrain Fuel Element 1-2415, GA Document No. 906770, Issue A, February 25, 1983.
- 6. HUDRITSCH, W., W. J. SCHEFFEL and O. M. STANSFIELD, Preirradiation Characterization of Material for the Fort St. Vrain Core Surveillance Program, GA-A12455, June 30, 1975.
, 7. Public Service Company of Colorado Fort St. Vrain Nuclear Generating Station Updated Final Safety Analysis Report, Docke t No. 50-267, July 22,1982 (Fig. 3.5-10) .
- 8. SCOTT, C. B., and D. P. HARMON. Postirradiation Examination of Cap-sale F-30, General Atomic Report GA-A13208, April 1,1975.
- 9. IrltiR Fuel Technology Program Semiannual Report for the Period Ending September 30, 1982, DOE Report GA-A16919, November 1982.
- 10. SAURWEIN, J. J. , C. M. MILLER and C. A. YOUNG, Postirradiation Exami-nation and Evaluation of Fort St. Vrain Fuel Element 1-0743, DOE Re-port GA-A16258, May 1981.
Page 39
I SA TECHNOLOGIES I N C. TIT 12: POSTIRRADIA7 TON EIANINATMN AND EVALUATION OF FSV FUEL ELEMENT 1-2415 1 _ lBOCBIENT NO. 907079 lISSOE NO./L11. N/C I I LEE, O. R. (PSC) letter to THOMAS M. NOVAK (NRC), FSV Segment 1 Fuel Surveillance, P-81254, November 16, 1981.
- 11. HOLZGRAF, J . F. , F. McCORD, and C. F. WALLROTH, Gamma Spectroscopic Examination of Peach Bottom H1UR Core Compo ne nt s ' ' , DDE Report GA-A13453, April 1978.
- 12. YOUNG, C. A. , Preirradiation Report of GA Fuel for Capsule R2-K13 ,
GA Document No. 904669, Issue A, April 1980.
- 13. WAGNER, M. R., GAUGE, A Two-Dimensional Few Group Neutron Diffusion-Depletion Program for a Uniform Triangular Nesh, USAEC Report GA-8307, General Atomic Company, March 15, 1968.
i
- 14. TUK, F. W., and L. J. 1DN, ' ' FEVER / M1, A One-Dimensional Depletion Program for Reactor Fuel Cycle Analysis, General Atomic Report GA-l 9780, October 22, 1969.
- 15. DORSEY, J. P., R. FROERLICH, and F. 1DDT, BUG-2/BUGTRI, Two-Dimen-sional Multigroup Burnap Code s for Rectangular and Hexagonal Geo-metry, USAEC Report GA-8272, General Atomic Company, August 22, 1969.
i
- 16. KRAETSCH, H., and M. R. WAGNER, GATT, A Three-Dimenisional Few Group Neutron Diffusion Theory Program for a Hexagonal-x Mesh, USAEC Report GA-8547, General Atomic Company, January 1,1969.
t Page 40 4
--.m e~~--n-.-, _ . . - - . . _ _ . , _ _ , . .
- -- - -J-- __ _ . - - - - s -- . . - - - ,, s - 4 -m -
i' GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lBOCB E NT NO. 907079 l ISSUE NO./LTR. N/C l i Table 1-1 C0dPARISON OF BIACKS WITH CRACIED WEBS i Item S/N: 1-2415 S/ N: 1-0172 ! Element Type Fuel Fuel Core Location 08.05.F.06 08.05.F.07 Top Crack Width (mm) (Ref. 2) 0.20-0.25 0.13-0.15
. Botton Crack Width (mm) (Ref. 2) 0.28-0.30 0.05-0.08 l Temperature (CC)(a) 650 700 Fluence (x 1025 n/m2)(a) 1.55 1.28 Meas. Axial Strain (%)(b) i 1 , _o,337 1 0.027 -0.163 ! 0.010 Meas. Radial Strain (%)(b) la -0.257 ! 0.027 -0.089 1 0.019 Neas. Bow (mm)(b) 0.43 0.28 Gross y Activity (R/h) at 91.5 ca(c) 458 385
- (a) Temperature were obtaine d f rom SURVEY code calculations based on the GAUGE code depletion analysis of FSV Cycles 1 and 2. Fast Neutron fluences were obtained from the GATT code depletion ana-lysis of Cycles 1 and 2. Temperatures are time and volume aver-aged. The temperature uncertainty (la) is estimated at 10% of the difference between the block temperature and the (~335oC, time averaged) ga s-inle t temperature. The fast neutron fluence s (E >
, 29 fJ) H1GR are volume averaged. The uncertainty in the fast l fluence is 1 10% (la).
i (b) Measured values came from the results of the metrology robot. (Ref. 1). Axial strains are element averages. Bow is at element midplane. Radial strains are element-averaged strains at the top of the blocks. (c) Measured during nonde structive examination (see Ref. 1) with a
. Reuter-Stokes, Model RS-C4-1606-203, gamma ionization chamber.
l Page 41 s
GA TECENOLOGIES I M C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 1DOCBIENT NO. 907079 l ISSUE NO./LTR. N/C l l Table 1-2 OBJECTIVES OF THE DESIRUCTIVE EIANINATION 04 E1.ENENT 1-2415 Data Method Obj ec t iv e s A. FUEL RODS
- 1. Structural in- Visual examination Verify structural performance, tegrity look for signs of fuel rod graph-its interaction
. 2. Stack lengths Fuel-stack-length Nessure fuel stack shrinkage, messaring trongh provide data to compare with assumptions in core design codes
- 3. Fuel stack pash- Force messaring Messere the extent of fuel rod-ont force device graphite block interaction, com-i pare with core design assamptions
- 4. Diameter messare- Depth gauge and Measure the extent of fuel hole ments on holes dial indicator changes asially 307 and 30s
- 5. Diameter measure- Riss gange and Measure the extent of fuel rod-ments on fuel dial indicator fuel hole interaction rods from stacks 307 and Jos
- 6. Fission product Gamma scanning Determine asial and radial power inventories distributions, determine f uel barnap and fluences, compare these results with neslear design code calculations
- 7. Microstructure
- Neta11ography Verify fuel performance B. GRAPSITE BLOCE**
I
- 1. Structural intes- Visual examination Verify performance of H-327 grity Also asing data graphite block from Ref. 2
. 2. Thermal expan- Stitca dilatometer Verify axial and transverse CTE
! sivity and identify any differences between analysis and measured i values
- 3. Thermal diffusiv- Seat poise Verify thermal diffusivity ity identify any discrepancies be-tween analysis and measured vaine s
- 4. Tensile strength Estensometer Verify axial and transverse ten-site strengths compare to analysis v aine s
*Pablic Service of Colorado f unded this ta sk
* *S am pl e gathering was part of this tasks however, the actual tests were performed by the GA Graphite Laboratory and are reported in references.
Page 42
i GA TECRNOLOGIES I N C. TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL RIMNT 1-2415 lDOCBIENT NO. 907079 l ISSUE NO./LTR. N/C 1 1 ( l Table 2-1 i NOMINAL PREIRRADIATION FUEL ROD ATTRIBUTES FOR FSV FUEL ELEMENT 1-2415 j l Fuel blend type: CR 16N-10167 Blend 1 l Pronerty Accentance Value Preirradiation fission gas release, -
<3 x 10-5 Kr-85m at 11000C:
1 Heavy metal loadings U: 0.15 g/ rod Th: 4.13 g/ rod These requirements are for the average of all fuel rods in the core. Thorium contamination: 18x10-4 Impurities (ppa) B: 15 Fe: 1500 S: 11200 ' Ti: 150 l V: 150
. Residual hydrogen: 1200 Residual ash: 13000 H2 0: 1400*
C1. Firing temperature (OC): 1800 l
-denotes no available data l
. l 1 Page 43
GA TECENOLOGIES I N C. TITM: POSTIRRADIATION EIANINATION AND EVALUATICN OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./LTR. IUC I I Table 2-2 NOMINAL FISSILE FUEL PARTICLE NEAN ATIRIBUTES FOR FSV FUEL ELEMENT S/N: 1-2415
' Pronerty Accentance Value Kernel type: (Th,U)C2 Th/U mean ratio: 3.60 _
Kernel A nominal diameter: 100 to 175 pm Kernel B nominal diameter: 175 to 275 pm Particle type: TRISO Nean Coating Parameters Nean thickness: Buffer: 25-75 pm IPyC: 13-35 pa* Sic: 15-35 pm OPyC (Fissile A): >20 pm OPyC (Fissile B): [25pm M'ean density: Buf fer: 0.75-1.40 Ng/m3 IPyC: 1.70-2.00 Ng/m3 sic: 13.16 Ng/m3 OPyC: 1.60-2.00 Ng/m3 Nean IPyC OPTAF: 11.30 Nean OPyC OFTAF: 11.25 o ' Combined thickness of seal pins inner-iso l coatings. Page 44
GA TECRNOLOGIES I N C. TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1 2415 lDOCUIEDE NO. 907079 l ISSUE NO./LTR. N/C I l Table 2-3 NOMINAL FERTILE FUEL PARTICLE MEAN ATIRIBITTES FOR FSV FUEL ELEMENT 1-2415
- Pronerty Accentance Value Kernel type: ThC2 Particle type: 71150 Kernel A nominal diameter: 300 to 410 pm Kernel B nominal diameter: 410 to 500 pm Mean Coating Parameters Mean thickness Buffer: 35-75 pm IPyC: 13-40*
sic: 15-35 pm OPyC (A): 120 pm OPyC (B): 130 pm Mean density Buffer: 0.75-1.40 Mg/m3 IPyC: 1.70-2.00 Mg/m3 sic: 13.16 Mg/m3 OPyC: 1.60-2.00 Mg/m3 Mean IPyC OPTAF: 11.30 l Mean OPyC OPTAF: 11.25 ~
' Combined thickne ss of seal plus inner-iso coatings.
Page 45 I
e s . e Table 4-1 1MITIA1. FUEL STAG FUSE 0ITT PORG NEASURENENTS y Bots Force Bole Force sole Forse Bole Forse Note Forse Mole Force Bote Forse Sole Forse Bote Forse Bole Forse N Mumber (abs) Number (Ibs) Number (1bs) Number (Ibs) Number (abs) Number (Ibs) Number (Ibs) Number (Ibs) Number (Ibs) Number (1bs) 2 5.0 37 12.0 70 20.0* 102 10.0 136 8.0 172 N.N. 206 15.0' 240 9.0 274 5.0 307 10.5 M 4 5 6.0 2.0 38 40 4.0 3.0 11 73 10.0 30.0* 104 105 5.0 4.0 137 139 2.5 12.0 173 175 9.0 6.0 207 209 20.0* 13.0 242 243 13.5* 7.0 275 276 8.0 11.0* 308 310 7.0 4.0 kM S 7 8 3.5 4.0 41 43 9.0 6.5 74 76 22.0* 28.0' 107 108 21.0* N.N. 140 142 11.0 17.0* 176 178 5.0 7.5 211 212 13.0 3.5 245 246 11.0 26.0* 278 279 15.0* 14.0 311 5.0 k > 313 2.5 13 18.0* 44 7.5 17 9.0 110 15.0 143 27.0* 179' 5.0 214 3.5 248 10.0* 281 6.0 315 20.0* 12 14.0* 46 24.0* 79 5.0 111 18.0 146 4.5 182 22.0* 115 5.0 249 8.0 282 6.0 317 23.0* N 14 8.0 47 28.0* 30 2.0 113 12.0 147 15.0* 183 10.5 217 10.0' 251 33.0* 2 84 16.5 318 14.0* .- M 5.0 DS 15 49 7.0 82 8.0 114 7.0 *149 14.0* 185 7.0 218 12.0* 252 25.0* 285 N.N. 320 4.5 2 j 17 13.0 50 7.0 9.0 116 16.5 150 8.0 7.5* O 18 10.0 $1 3.0 83 85 7.0 118 12.0 152 5.5 1 86 188 12.0* 220 221 10.0 3.5 254 255 9.0 4.5 287 288 11.0* 3.0* 321 323 8.5 14.0* h M M 23 7.5 5.0 21 7.0 53 54 7.0 86 88 9.0 24.0* 119 121 6.0 4.5 153 156 5.0 N.N. 189 191 4.5 24.0* 223 224 4.5 27.5* 257 258 14.0* 20.0* 290 291 1.0 3.5 k 2 23 3 .5 , e56 14.0 89 24.0* 13.0 3.0 O 25 3.0 57 9.5 91 8.0 122 124 18.0* 157 159 4.5 192 194 N.N. 7.0 216 227 5.0 10.0* 260 262 13.0 293 5.0 $ 5.0 294 20.0* C3 P 25 8.0 59 21.0* 92 8.0 125 18.0 160 1.5 195 3.0 228 13.0* 263 4.0 296 N.N. 28 17.0* 60 24.0* 94 7.5 127 20.0* 162 3.5 197 4.0 7.5
- O 230 265 17.0* 297 12.0 29 8.0 62 9.0 95 , 6.0 128 N.N. 12.0 6.0 8 31 15.0 63 6.0 97 9.0 130 16.0 163 165 9.0 198 200 14.0 231 233 14.0*
7.0 266 268 17.0* 10.0* 299 300 11.5 3.0 4 h c3 M 32 4.0 65 2.0 7.0 9.0 34 7.0 67 4.0 98 99 5.0 131 133 11.0 166 168 14.0* 17.5' 201 203 9.0 15.0 234 236 16.0* 14.0' 269 271 6.0 10.5' 302 304 9.0 20.0* w h M D8 35 15.0 68 7.0 101 5.0 134 23.0* 169 4.5 204 8.0 21.0* 8.0* # 237 23 9 13.0* 272 305 21.0* $
-- O Grand Neam
- S.D. _ 10.7 ~* 6.6 Rasse 1.5 - 33.0 4 2
? *S.e,.. e4 1. e, re,s.se 3
N.C. = Not Neasured 5 (e f N
'13 N is C)
+
= %
(Je I i
' * . e
>l 3,
u mi Table 4-2 513EL ST6CE Lt.NG118 SEAscaEMENTS Bete Bole Leasth Nele Lee 8th gele Lea 8th Sole Leasth Bete Leasth Bete Leasth Belo Leesth Male Lee 8th Nele Lea 8th Lea 8th Numbes (Iach) Number (Insk) Stumber (lash) Number (lash) Number (Inch) Number (Insh) Number (lash) Number (Insk) Number (Insk) Numbe r (lash) ,.g M S 2 N. 8L. 37 29.00 70 29.00 102 28.97 135 N.N. 172 N.N. 206 28.96 28.97
.240 242 28.90 28.92 274 275 28.90 28.94 307*
308' 27.03 27.08 h 4 28.98 38' 27.08 11 28,97 104 28.94 137 28.97 173 28.96 207 3 28.99 40* 27.12 73 29.02 105 28.94 139 28.96 175 28.94 209 28.97 243 28.90 276 28.97 310 28.90 tus 7 28.98 48 29.02 74 29.02 147 28.97 140 28.95 176 28,95 211 28.96 28.96 245 246 28.95 28.94 278 279 28.97 28,95 311 313 28.95 N.M. pg 4 8 28.99 43 28.98 76 28.98 108 28.98 142 28.96 178 28,98 212 50 N.M. 44 28.98 77 28.98 ISO 29.00 143 28.97 179 28.97 214 28.94 248 29.22 281* 27.06 315 N.N. == M M 282* 27.06 317 28.94 Z 12 N.N. 46 29.08 19 28.97 111 28.95 146 28.98 182 28,96 215 28,97 249 28.95 y 84 29.00 47' 21.50 80 29.00 813 28.98 347 28.97 183 N.N. 217 28.96 251 N.M. 2 84 28.96 338 28.90 p.g 15 29.02 49' 27.30 82 28.90 134 28.96 149 29.00 185 28.94 218 28.97 252 28,93 285 28,98 320 28.94 M M 37 29.04 50 28.96 83 28,96 116 28.98 ISO 28.92 186 28.93 29.00 220 28.96 28.95 254 255 29.00 28.94 287 288 28.96 N.M. 321 323 28.94 N.N. h 48 29.02 51 28.84 85 28.98 118 28,96 152 28.80 Ist 221 20 29.00 53* 27.12 86 28.97 389 28.97 153 N.B. 189 N.N. 223 28.92 257 28.96 28,94 290 291 28.95 28.95 g C 21 J9.04 31 ' 31.07 88 28.98 121 28.98 156 9:. m. 198 28.97 224 28.92 258 g p 23 29.00 36 28.98 89 28.99 122 28.95 IST 28.94 192 28.97 226 28.96 260 28.96 28.98 293 294* 28.96 27.04
- h O 28.98 28.96 262 25*
26* 21.14 27.30 57 59 28.96 29.00 91 92 28.98 28.99 124 125 29.06 N.N. 359 360 28.96 28.94 194 195 28.94 227 228 28.99 263 28.95 296* 27.06 , Q g 28 29.00 de 29.00 94 28.96 327 N.M. 162 28,96 197 28.96 230 28.94 165 28.94 297 28.93 28.92 o 4 h C3 M 29 29.03 62* N.M. 95 28,95 328 28,97 143 N. N. Its 28.96 231 28.96 266 28.90 299 o 31 29.00 63* 27.14 91 28.96 130 28.94 les N.M. 200 28.96 233 28.98 268 28.92 300 28.90 g 32 29.00 65 28.98 98 28.94 131 28.96 166 29.00 tot 28.94 234 28,94 269 28.96 302 29.97 e p 34* 27.09 67 28.98 99 28.94 833 28.92 168 29.00 203 28.94 236 28.92 271 28.92 304 28.96 O 98 28,93 272 28.94 305 28.95 2 35' 27.10 68 28,98 101 28.98 134 N.N. 169 28,95 204 28.97 237 239 28.94 -- O M *r3 M Gesad Nees
- 3.D. I Gssedmess[S.D.* 28.96[*8.04 27.09 0.03 en M
n
*Lessted moder a dowel 23.C.
- Not measured 5 a L
e-h
.N h E
g
'12 N 2 U C3 >4 U H I
- to A b W
LA
6A TECENOL06IES I N C. TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL FLEMENT 1-2415 lDOCUlmff NO. 907079 l ISSUE NO./L11. N/C l i Table 4-3 FUEL ROD DIAME7ER MEASUREMENTS FOR STACK 307 Rod Fuel Top Botton Average Rod Diame te r Number 00 900 00 90 0 (In.) 1 0.4915 0.4917 0.4927 0.4935 0.4923 2 0.4937 0.4926 0.4926 0.4932 0.4930 3 0.4922 0.4922 0.4927 0.4930 0.4925 4 0.4922 0.4932 0.4927 0.4927 0.4927 5 0.4924 0.4932 0.4922 0.4922 0.4925 6 0.4910 0.4915 0.4930 0.4924 0.4919 7 0.4924 0.4927 0.4927 0.4924 0.4926 8 0.4920 0.4924 0.4932 0.4930 0.4926 9 0.4910 0.4902 0.4924 0.4924 0.4915 10 0.4907 0.4905 0.4927 0.4927 0.4916 11 0.4937 0.4937 0.4924 0.4927 0.4931 12 0.4924 0.4941 0.4934 0.4934 0.4933 13 0.4922 0.4927 0.4934 0.4931 0.4929 14 0.4947 0.4942 0.4927 0.4927 0.4936 Grand Average 1 S.D. = 0.4926 1 0.0006 Page 48 I
GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-241.5 lDOCEIENT NO. 907079 l ISSUE NO./L11. N/C i l Table 4-4 FUEL ROD DIAMETER MEASUREMENTS FOR STACK 308 t Rod l Fuel Top Botton Average Ecd Diameter. Number 00 900 ; 00 900 (In. ) 1 0.4920 0.4920 0.4927 0.4927 0.4923 2 0.4930 0.4940 0.4922 0.4922 0.4928 3 0.4917 0.4905 0.4930 0.4931 0.4921 4 0.4920 0.4917 0.4930 0.4927 0.4923 5 0.4910 0.4922 0.4927 0.4931 0.4922 6 0.4942 0.4934 0.4924 0.4927 0.4932 7 0.4920 0.4922 0.4927 0.4930 0.4922 8 0.4931 0.4939 0.4927 0.4930 0.4932 9 0.4930 0.4930 0.4930 0.4927 0.4928 10 0.4922 0.4917 0.4927 0.4927 0.4923 11 0.4924 0.4920 ;0.4924 10.4930 0.4924 12 0.4944 0.4937 0.4927 0.4927 0.4934 13 0.4931 0.4939 0.4927 0.4927 0.4931 , 14 ,0.4920 0.4914 0.4942 0.4930 0.4926 : Grand Mean 1 S.D. = 0.4926 1 0.0004 i e Page 49 _ . . - , . _ _ _ _ - . _ _ . . _ . _ _ _ _ _ _ _ __, _ ___. - ~ _-.__ _.- _ _ _ -.. ..._._--____.__.
GA TECENOLOGIES I N C. TITut: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBIENT NO. 907079 l ISSUE NO./LTR. N/C l I Table 4-5 FUEL HOLE DIAME7ER MEASUREMENTS FOR HOLES 307 AND 308 Hole Diameter (Inches) Depth * (In.) 00 900 l Average 00 900 Average 1 0.4986 0.4986 0.4986 ! 0.4988 0.4991 0.4990 8 0.4991 0.4588 0.4990 - 0.4988 0.4986 0.4987 16 0.4983 0.4988 0.4986 10.4988 0.4986 0.4987 22 0.4988 0.4986 0.4987 .0.4988 0.4986 0.4987 28 0.4991 0.4986 0.4989 0.4988 0.4988 0.4988 Grand Average 0.4988 0.4988 Standard Deviation 1 0002 1 0001
*Approximately 1.125 inch already removed from the ele-ment. These depths measurements are of the remaining element.
k Page 50
GA TECENOLOGIES I N C. TITIA: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCBENT NO. 907079 l ISSUE NO./L1R. N/C 1 1 Table 4-6 FISSILE PARTICLE NET /J LOGRAPHIC EXAMINATION RESULTS
- Prope rty Value Time avsrage fuel temperature (oC) 765(*)
Fast fluence x 1C25 n/m2 (E > 29 fJ)lntiR 1,79(b) Fissile particle FINA (%) 8.35(c) Number of particles examined 231 Buffer failure (%) 3.0(*) IPyC failure (%) 3.0(*) sic failure (%) 0.9(*) OPyC failure (%) 0.4(*) i Total coating failure (%) N.D.(d) Dispersion in buffer (%) 74.9 l Dispersion in IPyC (%) 1.3 Debonding in IPyC (%) 42.9 sic attack (%) 3.9 Flawed sic coating (%) 0.4 ! (a)Taken from the SURVEY code, axial point 5, local points 4 and 5 (fuel temperature time-averaged) . (b) Data from Ref. 1. t (c) Data from Ref. 16. (d)Not de termine d because the initial fuel rod mount hydrolyzed before any meaningful data could be gathered '-t the total coating failure. (e)For each particle with coating failure, one or more intact coatings remained on the particle even though one coating was observed f ailed. Page 51
.my ,- y -
,_.,___.,..m.,,. _ , _ _
.._,_.,__,._-,~_,__,,._,_,,._,,-,...,,,_._._..,,7_._-m,.____,m ,, , .w.m,y.,,,,.y.,y__,--,3,se.-- , ,.,
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELLLENT 1-2415 lDOCDIENT NO. 907079 l ISSUE NO./L11. N/C 1 i Table 4-7 FERTILE PARTICLE METALL0 GRAPHIC EIAMINATION RESULTS Prope rty Value Time average fuel temperature (OC) 765(*)
. Fast fluence z 1025 n/m2 (E > 29 fJ)RIUR 1,79(b)
Fertile particle FIMA (%) 0.61(c) Number of particle s examined 184 Buffer failure (%) 18.5(*) IPyC failure (%) 10.3(*) sic failure (%) 3.8(*) OPyC failure (%) 7.6(e) Total coating failure (%) N.D.(d) Dispersion in buf fer (%) 5.4 Dispersion in IPyC (%) 1.6 Debonding in IPyC (%) 12.0 SIC attack (%) 3.3 Flawed sic coating (%) 6.5 (a)Taken from SURVEY code, axial point 5, local points 4 and 5 (fuel temperature time-averaged) . (h)n... ,_-_ n., , (c) Data from Ref. 16. (d) fto t de termine because the initial mount hydro-lyzed before any meaningful data could be gathered for the total coating failures. (*)For each particle with coating failure, one or more intact coatings remained on the particle even though one coating was observed f ailed. b Page 52
GA TECENOL06IES I N C. TI112: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./ Lit. N/C 1 1 1 Table 4-8 COMPARISON OF MEASURED AND CALCULATED COMPOSITE BURNUP l Composite Burnup (%) Relative l Difference (%) Meas- Calcu- Calc
. Rod Cs-137 sured** 1ated*** -1 No. CPS * (Fe) S.D. (Fc) Meas 1 26.66 2.54 .24 2 25.64 2.44 .22 3 25.13 2.39 .18 4 24.35 2.32 .19 5 24.75 2.36 .19 6 24.31 2.31 .20 7 24.82 2.36 .19 8 24.73 2.35 .22 9 24.13 2.30 .23 10 24.23 2.31 .19 11 24.88 2.37 .20 12 24.74 2.35 .17 13 25.65 2.44 .18 14 26.73 2.54 .24 15 26.58 2.53 .26 Avg. 25.16 2.39 2.02 -15.48 S.D.**** .90 .08
- Corrected for background where:
** Measured composite burnup, Fe CPS = Cs-137 counts per sec.
Fe = [ CPS (L)/K]/(Uo+Tho)Y L = Rod length = 1.94 (in.) K = Dector efficiency =
+ 2.97 x 10-18(CPS-in/aton)
Do ,Tho = BOL loading for uranium and thorium (atom) 6 *** Calculated composite barnap, Y = fission yield of Cs-137 from Fc U-235 F5 = Fissile burnup Fe=F5 I+F3 (1-1) F3 = fertile burnup (Based on data from the fuel Do accountability) I= Uo+Tho
**** Standard deviation, S.D. j n 1/2 S.D. = ' Il -I i=1 '
' Page 53'
r GA TECENOLOGIES I N C. TITI2: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l DON NO. 907079 lISSEE NO./LTR. N/C l l E l
// .:=
f {-?~ \
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HANDl.ING
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.-,'CR AC KED A R E A x '-
B k. . ) __r - , F IG U R E I-1.. FbV ELEMENI SIDEFACE IDEt1TIFICATION Pese 54
GA TECENOLOGIES I N C. TITI2: POSTIRRADIATION EZAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l N NO. 907079 l ISSUE NO./L'11. N/C I l l N t ,
, vb fpsk 35 36 37 23 0 0'.'.' E L 34 18 19 21 * .
O f C 4 A >]i-p. 4 33 17 7 2 9 22 i FACE B {s , x. 32 ts i 3 13 23 g' % - Qr I\ ' 4 11 24 M pg,$ 31 15 5
' w q.g
.inTh;r 30 f4 13 12 23
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% \ (gg. TIS
" 3
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i
- FUEL ELEf. TENT 12415 .
REGION 8, COLUMN 5, CORE LAYER 6 (ACTIVE CORE LAYER 3) Fi g . 1-2 Core location of fuel element 1-2415 Page 55
GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./L11. N/C I I
. ALL DIMENSIONS IN mm H
A eosooGooso'. COOLANT HOLE OSOOSOO9 U 15.9 DIAM (102) o s c o s ~' s o oso .S o sooSoo so C00LANT HOLE so oso oso os osooS 'e6 12.7 DIAM (6) oosoosooso'Soosc. oso oso oso o'io oc ; oso soosoosoos cf,;osoos oosoosoo oosoosoo BURNABLE POISON 359 esoosocso oosooso s
'oscosooss cosoosoo HOLE 12.7 DIAM (6) so oso oso ca ;oosoosoos SO osoosooso
*O SOOSO PITCP*Osocsoosc S
Noso osofose osooscos FUEL HOLE so oso c' 40osooso 12.7 DIAM
.osoo asoosoo (210) soos a>osoos asooso osoo o S oSo oe eoosoe CEMENTED GRAPHITE
. -A FUEL HANDLING PLUG (TYP)
FICKUP HOLE g DOWEL PIN 3 : h?[!f[f'%# $j
! f f 5 34' 3 d: e 3 pHELluM FLOW j
b 793 ll. '
, f [ f
- s :s : s g nd i e -
] :q .m . s : .-
. d 0 y COOLANT j,y @c p : { [ ;
- CHANNEL' aw.c (g c s
- s : : ~
( . .. f FUEL ROD' D I T'
- : k f""
STACK 3y b E E { LENGTH = : , 7 fe f : : : 3 ( 749 3' (,+b ty ; ; ; ; ;s : 0 th. . s s s
' q S. .' : : s :
\p " LliNii i!BIq F SECTION A A DOWEL Fig. 2-1 FSV fuel element 1-2415 Page 56
) , GA TECENOLOGIES I N C.
I
; IIT12: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 A
lDOCUIENT NO. 907079 l ISSUE NO./L'IX. N/C I l l i I I - 1 I l l i 1 1
\' F. 'I
.*3s 'C j ..t m..., , ,.
N ,$[ ~i Mpf ,,yj yy x g,Sjyy . #. ~ .. ,,., ' .g'
. ~;:. *[.I '
; I ; $
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- ry ss.
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~>
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s ...s I , 6
. ,. % - S,; , :,
A
's., .,
a Fig. 4-1 Results of the first removed stab, crack is barely ' visible because of the dust from sawing. Notice the cut dowel pin and the graphite dust from the ! sawing. .I l Page 57 i l ...
i\ t , r l- l \ ,'l
! li il, l(ii' llII l e> dnnNzor. o**== g,O
. aGg@gdO gyH5z ga y @ Eg O o Ny@@mggYy5
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- GA TECENOLOGIE5 I N C.
l TITM : POSTIRRA6IATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l lDOCDIENT NO. 907079 l ISSUE NO./LTR. N/C l l l 1
,, ; [sNN4'I ' T,
+~
l e
- v. '. . I (a) 9"ewn=:5y. . " Dr l
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e t f.1 1 I:n:resesuWWtkl(N ; 1 (b) Fig. 4-3 Remains of the B dowel pin fallowing first cut. (a) Shows the l bottom of the dowel and (b) shows the cut portion. Note, the j dowel pin appears void of cracking. I Page 59
GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l lDOCUMDir NO. 907079 l ISSUE NO./L11. N/C B face l l
, ^ M.
c [K ':.. 4
.a. a e (9
$ .** ~ % ? );.
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7 B face (a)
, .s ?! , ,' , p' df, . ..
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. 93, x 9 . t. : , y! ~ 2
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s c. As s .v . _ _ , , ,
;j' y ,S (b)
Fig. 4-4 Fuel plugs still intact at an approximate axial location of 1 inch, (a) shows the B face with all plugs intact at the B dowel area and (b) is an enlargement of the area around the cracking pattern. Page 60
GA TECENOLOGIES I N C. TIT 12: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIENT NO. 907079 l ISSUE NO./LTR. N/C 1 - I B face w.
. . .v p.
. : ty. . . +*
,,. *NTN k
f ~ i' ,w. SIMS. .. . t4:; g. ,-h d D 3;; g m,u;- . m %;; p - :
~;f
.?
.,. s{ .%j'? +'
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-y'EY2.m$b*$~
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. f*y'..:. ?
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%. y-: . .a.:7s.u.hl*.,a.j! .*. ;s Fig. 4-5 Bottom of the 3rd top slice at axial location 1-1/8 inch from the top. Note the thin remains of the fuel plugs and the cracking pattern.
1 Page 61
GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 DOCBIENT NO. 907079 ISSUE NO./LTR. N/C
/e'e e e!
e
- e. 8 e e ele
. @e @@ e ec e e e e,, e e s e
,s e- - e ele e ee'-@., e s@e 8 e e @-e x
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#, Oc F - @ e 8 e@e s @ e@eseee ete 8 e e e e e s le8esesee ee8 >e 8eeeeeeee"e 1988e 8ee- i ee. ees e-8 e e esee 8esee=
e esse-- e se8eeeee-Geeeeeeee-Ge8eeees eeeeeee c / 08esessee:eeee8esee
'eeeeeeeeeeeeeeees c\e8eeeee8eeGeeeeeeele8ee ese "
lee 8ee see8 e8 ese i 88ee e . Ieip S e e, , ee8 ' 5"l'"^*" Ac t l
. 5 l % .t to KFA, FRG for recycle program use 9
l Gamma scanned (later located to receptacle block)
- Only three rods from this stack were sent to KFA, FRG for recycle program use '
All other unmarked holes to receptacle Fig. 4-6 Destination of fuel rods Page 62
HIGH LEVEL GELL
/ 9 BaHon.
3 y _ _ _ _ __ g s u i ___ _ . 1 n a2 B and Sa w I E
-gj .3Od . .
a a j viewinq g , y l Cradle winjo# l 3 [ s e -Granite Table n e- l $
' [!, D oj '
, _/ i E
- 15 *
;. I
!E o e I g j
g o Dual Tubi ' i) I E F *
' < a w R = civing T(c u gl.h .
S 4 m
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= g ,
.s g,o p
=
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'= i <N Fuel RJ Push;ng * * *
-l - ~Hlock l- 2.41 5 y< n j E * *
- _ % . 4 3
_m____
,- a 4
Force Mea surn.y - N 3 7 L P , p Opiiici 2 at OpoatofIat z g Vie w in-J %.Lw d Vie w.ng Win.iow I h 5
%o
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0 Fig. 4-7 Fuel rod pushing operation setup
. - - _ - _ - _ _ - _ .- . _- . _ _ - - - . = _ _ _ _ . _ _ . _ . . - - - _ - . . .
t GA TECINOLOGIES I N C. { TITLE: POSTIRRADIATION EIANINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 I IDOCUIENT NO. 907079 l ISSUE NO./L11. N/C l i l l l ! 0 l
. v I e ,
8 f ., o _. l
- Wt." Qc g
i t c j i 4 u . :.
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e
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Al [ l Fig. 4-8 Two scales mounted on each side of the dual receiving i trough were used to measure the stack lengths. l 1 I l Page 64 j f _ _ .- - __ _-- - _ _ _ _ _ _
! 6A TECENOLOGIES I N C.
l TITLE: POSTIRRADIATION EXAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 lDOCUIEfft NO. 907079 l ISSUE NO./L'II. N/C i I
,g
! s , 1 Il s T 1 l( . j ( ,
's. . .
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I : . li 1 y 1 Composite photographs of fuel stack 307
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i l GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l lDOCUIENT NO. 907079 l ISSUE NO./LTR. N/C i l l l l w
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GA TECENOLOGIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 IDOCUIEDfr NO. 907079 l ISSUE NO./LTR. N/C I I w N
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Page 81
I GA TECENOL0GIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l lDOCUIEDfr NO. 907079 l ISSUE NO./L'II. N/C _
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GA TECHNOL0GIES I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL ELEMENT 1-2415 l DOCUMENT NO. 907079 l ISSUE NO./L'IR. N/C I I Fig. 5-10 Composite photographs of a portion of section 4 taken at axial location 24-1/8 inches. (a) Shows the s bottom of the 1.0 inch section removed for LANL and (b) shows the top of the (a) remaining section. On both (a) and __ ] (b) only the crack extending in:o coolant hole 319 is clearlv "isi-le. I
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GA TECENOLOGIE5 I N C. TITLE: POSTIRRADIATION EIAMINATION AND EVALUATION OF FSV FUEL FJ.RMRNT 1-2415 l=W"I NO. 907079 lISSUE NO./ LEI N/_C I I OXiO! Cores. Tra ns verse Cores -
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