ML20085D676
| ML20085D676 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 03/31/1964 |
| From: | SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 1653, NUDOCS 9110170051 | |
| Download: ML20085D676 (51) | |
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I ADDENDUM NO. 3 To :
SAFEGUARDS REPORT FOR PH;.SE 1 t OF THE SAXTON NUCLEAR EXPERIMENTAL CORPORATION FIVE-YEAR RESEARCH AND DEVELOPMElff PROGRAM i
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.t MARCH 196h
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i 9110170051 910424 PDR FOIA DEKOK91-t7 PDR
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TABLE OF CofffENTS DC8 1.0 INTROUJCIION . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 ColmLUSIONS.-. . . . . . ....
..............2-1 30 SPECIAL 3x3 SUBASSD4BLY E110HP TERM WEL IRRADIATION. 3-1 . . . .
4.0 SPECIAL .?x9 I410 TEIL4 FUEL IRI%DIATION . . . . . . . . . . . 4-1
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50 MODIFIED HOLIN FUEL ASSD4BLY AND PLUG FOR SUPERIIEAT POSITIO 6.0 MATERIAIS TESTING CAIBULES AND GAM 4A TIIEMIOKEIER . . . . 6-1 ..
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l M 'NTRODUCTION i .
I 1.1 SCOPE AND BACKGROUND i
The limiting categuards considerations governing the design and operation of special fuel assemblies in the Saxtra Reactor with chemical phim conditions have been previously analyzed in the Safeguards Report for Phase 1 of the Saxton Research and Development Program and its supplements, It is the purpose of thh report to present further additions and changes to the experimental program.
The report describes and evaluates the effect on reactor safety of the following additional experiments and changes.
(1) Insertion of a special 3x3 subassembly containing Zircaloy and stainless steel rods with two of the Zircaloy rods having intentional defects in the clad.
(2) Insertion of a special hollov 9x9 assembly containing both Zircaloy and stainless rods as well as eight rods simulating the design esafiguration and heat output of a rod cluster control assembly.
(3) Insertion of a hollov 9x9 standard ascembly with 21 rather than nine central rods removed for future use of a superheat loop.
A special removable plug is also included with the assembly to '
permit easy installation of the superheat loop.
(4) Insertion of materials irradiation capsules and a gamma thermometer to obtain data on radiation effects on structural properties of reactor component materials.
1.2 DEVELOPMENT OBJECTIVES The overall objective of these experimental programs is the testing and 1 evaluation of materials and procedures which are proposed for inclusion in large, central power st.ation systems. Ncv materials and 1-1 E
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techniques developed in the laboratory are subjected to many simulated reactor envirotunent tests to prove their reliability and worth. In the final analysis, hwever, operation in and exposure to actual reactor conditions is necessary to insure complete and proper evaluation.
I The specific objectives of each program are as follws:
(1) Special 3x3 Subnssembly '
1 (a) Hensurement of finsion product releese from known defect.-
(b) Performance evaluation of vibrationally compacted and chamfered pellet fuel.
(c) Comparison of crud deposition and corrosion resistance of Zr-L and stainless steel.
(d)
Evaluation of effect of chemical shim coolant on fuel and clad interior.
(2) Special 9x9 Assembly (a)
Evaluation of spiked, atainless steel clad rods, operating at 16 kv/ft.
(b) Evaluation of spiked, diresloy clad rods operating at 16kv/ft.
(c) Continued evaluation of.RCC concept and performance.
(d) Continued evaluation of performance of rods containing a burnable poison.
(3) Modified Assembl/ for Superheat Loop Instellation of a hollov 9x9 assembly with a special plug inserted +
to allow future superheat loop installation without head removal.
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(4) Materials Test Capsulec and Gamma Thermon.eter i
(a) Irradiation of various materials to obtain data on the ;
effect of radiation on structural properties.
Irradiation of various materials in high gamma and neutron
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(b) flux and determination of gantna heating. !
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i 2.0 C0!CLUSI0!i3 i
As this report shows, the proposed experiments and procedures may be carried f out without increasing the severity of any accidents previously analyzed and witnout creating any tin ,notential accidents.
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10' SPECIAL 3x3 SUBASSEMBLY SHORT TEhM WEL 1RRADIATION 3.1 PRO ~AAM OBJECTIVES, SCOPE AND DESCRIPTION The overall objective of this program is to study the effect of a pressurized water reactor environment on materials which are cor+.em-plated for use in large scale systene. The specific objectives may be outlined ass (1) Measurement of fission product release from a defect of known size and orientation.
(2) Performance evaluation of vibrationally compacted UO2 fuel.
(3) Evaluation and comparison of crud deposition on Zr-h and stainless steel cladding.
(h) Studying the interaction of a chemical shim coolant and U0p fuel.
(5) Evaluation and comparison of the general corrosion behavior of Zr-h and stainless steel in a chemical shim coclant.
(6) Tatermining if the presence of defects will accelerate the hydriding of Zr-h.
(7) Evaluation of new pellet design.
The scope of this program will consist of the removal of the present 3x3 RCC subassembly and insertien and irradiation of a special 3x3 sub-assembly in the center core position. The nir,e fuel elements, fou removable and five non-removable, are specially selected and designed for this experiment. -The inclusion of the removable elements allows the gathering of data on the corrosion, crud buildup and fission product release phenomena over several periods of time.
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4 The fuel rods are shown in Table 3-1. Two of the four rtr.ovable rods are defected, Zr-b clad type with the fuel in solid pellet
- form. One of these oefected rods has a history of prior irradiation ,
of about 7500 M4D/MT. The other defected rod is unirradiated and is used as a reference control. The defect is a nacl.ined hole, 15 mils in diameter. Following an irradiation period, these rods may be removed and replaced by similarly defected new rods.
These defected rods will allow an estimate to be made of fission product escape rate coefficients for defects of a kacwn size and orientaticn by fellowing the fission produ:t concentration in the coolant as a function cf time. Calculations indicate that the fiJsion product concentration resultir.g from these defects miCht be as much as 10 ue/cc. Saxton Technical Specifications allow operation with coolant fission product concentrations of up to 20 uc/ec of long-lived isotopes.
In addition to providing data on fission product release from defected elements, this program will also provide data on the general corrosion resistance, crud buildup and the effects of defects on the hydriding of Zr-h cladding.
One of the remaining removable rods is Zr-h clad with pellet fuel that has & histcry of about 7500 M4D/MT of prior irradiation. It is enriched to 6.1 w/o and should operate with a maximum power density of-13.5 kw/ft.
This rod will be used for a crud sample for Zr-h as well as for high ,
exposure studies on the fuel and clad. The last removable rod is a normal stainless steel clad rod whose cladding has beer. heat treated to sensitize it. Sensitizing the clad causes carbon to precipitate in the grain boundaries making the clad more sensitive to corrosion.
This rod will be used to study the effect of sensitizing or. corrosion of stainleas steel, and may be replaced by an identical rod after an irradiation period.
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In addition, ten fuel pellets of a modified design are to be ind uded in these rods. Fuel fabrication techniques have shown that a reducticn in the number of reject pellets can be effected by including a small chamfer on the edges of the pellots. It is proposed to insert some '
of these modified pellets in the rods to observe the effect of the chamfer cn pellet operation. A sketch of the new pellet type is shown !
in drawing 500Boh9. Heat transfer characteristics are not significantly dif ferent from those of a normal pellet.
Two of the five non-removable rods are stainless steel clad with i vibrationally compacted fuel. They are included to give data on the corrosion resistance and crud buildup cf stainless s, teel and performance of vibrationally compacted fuel. The remaining three-rodo~are Zr-b clad with vibrationally compacted fuel. They will give data on the ccrrosion resistance and crud buildup of Zx-h and performance of 1
vibrationally compacted fuel.
The crud buildup will be evaluated by quantitative crud removal. The corrosion resistance will be determined-by visual and micrometalle.
graphic examination of botn Zr-h and stainless steel and by chemically stripping the corrosion film for stainless steel. Performance evaluation of the vibrationally compacted fuel ie being carried out under 4 WAPD program.
3.2 MECHANICAL DI3 ION Three design changes have been included in this new 3x3 subassembly.
Some structural problems have arisen because of the ern perforation design _of the normal-3x3 assembly. Instead of using the standard elongated slots, the new can vill have offset rows of circular cutouts.
The new design will increase the structural rigidity of the can Jy approximately 70%. The percentage of metal removal from the can is slightly less using the circular cutout but this should in no way have any effect on the thermal and hydraulic characteristics of the subassembly.
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A cerond change is incit.nion of a rev 3x3 spring clip grid design. A sketch of this new clip is shavn on Draving SS2ID96. The new design supports the fuel rod with a curved spring clip rather than flat clips. \
Dimples are provided under each sprire to limit the amount of motion allowed the rods before they contact a colid backup. The allowed deflection of each spring is 7 mils. ,
The third design change has been to pin the top and bottom of the non-removuble and the bottom of the removable rods to reduce lateral motion of the rods. . Pin extensions are ir.21uded on the end plugs of the new rods, and these fit into machined recesses in the bottom and top end plates of the fuel assembly. The bottom recesses do not penetrate the end pinte so no nev flow channels nor hydraulic lifting forces are created. As the two previously irradiated rods do not have any pin extensions, they vill tiot be included in this change. When the defected, previously irradiated rod is replaced, its replacement will have the pin extension.
These design changes described above vill not alter the thermal and hydraulic conditions previously analyzed and reported for a spiked 3x3 subaar,embly.
33 NUCLEAR DESIGN CONSIDERATIONS
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Table 3-1 gives the per cent enrichment and the estimated maximum peak l
power density for each rod in the 3x3 subassembly. The maximum peak nover density anticipated at 23 5 W is 16 kv/ft whica vill occur in l
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the vibrationally compacted, stainless steel clad fuel. The Zr 4 clad loose oxide fuel vill opemte at a peak power density- of 14 kv/ft. The remaining-elements, pellet type, are expected to operate at 13 5 kv/ft j maximum power density. The enrichments and power densities proposed for this 3x3 element are consistent with the conditions analyzed in .
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the " Safeguards Report for Phase I of the Saxton Nuclear Experimentai Corporation Five4 car Research and Development Program" and its supple-tre nt t. . This subassembly therefore does not change the previou' sly described worst conditions.
34 SAFIL"Y CONSIDERATIONS The tsjor safety consideration involved with this experiment is the behavior of the tvo defected fuel rods while in the reactor. As previously stated, the r.nximum releases from the defected elements are expected to cause a coolant activity concentration of not more than 10 pc/cc which is within the Saxton operating
- mit of 20 pe/cc.
If the concentration becomes too hi6h due to the enlarging or-the defects, this subassembly can easily be removed and a backup teet assembly inserted in its place.
The possibility of the complete release of all of the fuel in both defected elements is extremely remote. Even if such an event should occur, it vould not present any hazard to the Iublic because the release vould be entirely cortained by the reactor coolant system.
As this experiment is a normal. 3x3 subassembly, no special instal.lation procedures or precautions r e necessary when inserting it into the core.
35 CONCLUSIONS It may be concluded that insertion of this cpecial 3x3 subassem)1y with defected fuel elements into the core introduces no new safety problems not previously analyzed.
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TABLE 3-1 PROPQiED FUEL ROP 'gICN .'.*?vdETERS FOR SAITON DFSFCT TEST 4
Rod Peak Previous Identi-- Power (ib~d/hT)
' ' Per Cent Furocse fication:
Number Type Ciad Fuel Enrich.cnt (kw/ft)_ Defect Excesure 13.5 No None a) Crud sarple Removable . Sensitized Pellets 5.09 c) N-w pe let evaluation 15 10 w/ ground 30h SS c) Clad evaluatica eds;e a) Cred sample No Ue ,,
Sensitized Pellets 5.69 13.5 Removable b) New pellet evaluation 30h S3 10 w/ ground edge c) Clad evaluatica Zr-h- Pellets 6.1 13.5 Tesb 7500 a) Defect after signifi-101 Removable cant burnup 6.1 13 5 Tesb None a) Defect test Removable Zr-h Pellets 7.2 No None a) Crud sample on /x-h 301 Non-removable 7.r-h vibrational lh
.b) Ferfornance of locse Conoacted ende fuel 7.2 lk- Ho None a) Duplicate of Rod 301 302 Non-removable Zr-h Vibrational Compacted 7.2 lh IIo None a) Duplicate of Rod 301, 303 'Non-removable .Zr-h Vibrational Compacted 302 16 No None a) Crud sanple JCL SS 311 Non-removable 30h SS Vibrational .2 ^
b) Performance of loose Cce.pacted oxide fuel 7.2 16 No None a) Duplicate of had 311 312 Non-removable Joh SS Vibrational Compacted
.No 7500 a) Crud sa=ple Zr-b Pellets 6.1 13 5 313- Removable b) Periornance of fuel cader high burnup 6.1 13.5 Tesb- None a) Low burnup control for
.. 321 -Removable- Zr-h Pellets Rod 101 Pei ats - 6.1 13.5 Ye d Ncne a) Defect tes*
Removable - 7.r-h b) Crud & ct, r . 31on t.es t d'15-mildisneterhela machined through clad vsll.
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L.0 SPECIAL 9x9 IDN3 TERM F'JEL IhRADIATION h.1 FROGRAM DESCRIPTION, OBJECTlVES AND SC0FE The objective of this program is to obtain long term, statistically me6ningful .4-radiation data to verify technical feasibility of naterials, prototype designs and fabrication variables and techniques preposed for fuel elements to be used in large plants with a chemical shim environment. Thir program combines into one test many items which have been tested in the past or are being tested now.
1 The scope of this pr ogram involves ir.sertion of_ a special hollow 9x9 fuel assembly into the central position. The special 9x9 fuel assembly contains 61 fuel rods which have been selected to provide the experi-mental data desired. The following variables are represented in the special assembly.
(1) Free standing stainless steel clad fuel rods, both spiked and unspiked, operating at peak power levels of up to 16 kw/ft.
(2) Static simulated control elements of the rod cluster control (RCC) concept.
(3) Stair.less steel clad rods utilizing 16 Ni - 20 Cr stainless steel.
(b) Normal stainless steel clad rods with a concentration of 100 ppm
' of boren used as a burnable poison in the fuel.
(5) Free standing Zircaloy clad rods operatir.g up co peak power levels of 16 kv/ft.
A cross section of the special essembly is shown in drawing ED SK 302100B and gives the rod position for the various groups of test elements. Table b-1 lists the details of clad type, enrichment and power level associated with each group.
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There are eight RCC test rods proposed for this special assembly, four for each of the two types presently being tested in Saxton. The design, 4
construction and Lupport for these new RCC elements is exactly the same as that for the elements in the core.
The assembly contains 12 fuel rods with free standing 30h stainless steel clad and all are spiked to operate at 16 kw/ft. Six of the rods have heliun in the gap while, of the remaining, three are air filled and three are argon filled. There are six more spiked rods with argon in the gap, three are clad with 3h8 stainless steel and three are clad with 16 Ni - 20 Cr stainless steel.
There are six normally enriched, argon-filled 30h stainless steel clad rods operating at 13 5 kw/ft. Three of these have been segmented into equal thirdu during manufacturing so that fission gas release studies may be made on different sections of the fuel rods. A fourth "od containa 100 ppm boron as zirconium diboride to gather more data on the performance of fuel containing a burnable poison. Borated fuel of the same bcron concentration but higher fuel enrichment is presently under test in Saxton.
The remaining 29 fuel rods are heliun filled and Zircalcy clad. They use fuel pellets of a slightly smaller diameter than those in the stainless steel rods which causes them to run at slightly lcwer power.
Eight of the Zircaloy rode are clad with Er-h and given an autoclave pre-oxide coating on the 0.D. Three are normally enriched (5.69 w/o) and run at 12 kw/f t peak power while the other five are enriched to 6.1 w/o and run at 13.5 kN/ft.
Six of the Zircaloy rods are Zr-2 clad with an autoclave pre-oxide on the 0.D. and are spiked to run at a peak power of ik kw/ft. The spiking pattern is the same as that used in the spiked stainless steel rods. Three of these Zv-2 rods are Ni free.
A Group of 12 Zr-4 clad rods are all spiked to operate at ik kv/ft.
Tncoe twelve are divided into four sub-groups which have received different surface treatments. Three rods have an autoclave pre-oxide on the 0.D., three rods have an autoclave fre-oxide on the 0.D. and I.D., three rods have a furnace pre-oxide on the O.D., and three rods are mounted an pickled with no pre-oxide treatment.
A fiust grour er threu 7.r-4 clad fuel rods is enriened to 7 3 v/o and l vik opeu.te nt 16 kv/ft. These rods also have an nut,oclave pre-oxide treatment. Two fuel red positions, indicated as J on ED SK 3021ooD, are left tar.rt in order to accotanodate flux mmsuring thimbles presently in these core pasitionr.
4.2 MDCMNICAL DESIGN With the exception of the presence of the eight RCC test elements, th':
mechanical design and fabrication of this test assembly is the same as
' that of a norml assembly. Installation of the RCC elements and te guide tubes vill follow exactly the same procedures used for those RCC elements presently in Saxton.
4.3 fMCLEAR, THERMAL AND HYDPAUIJC DESIGN The nuclear, themal snd hydraulic design parameters for the various proposed. fuel rods, including the RCC rods have been analp ed and reported in the " Safeguards Report for Phase I of the Saxton Nuclear Experimental Corporation Five-lear Research and Development Program" and its supplements and amendments. A thermal-hydraulic analysis of this special 9x9 assembly with RCC e;ements has been perfomed, and the results are shown in Table 4-2.
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i TABLE 4-2 CAlfU1ATED DNB AND !!0T CIIANNEL FACTORS Spiked Spiked 3x3 Spiked 9x9 RCC 9x9 RCC
- 1. Minimum Local q"-DNB Ratios _t i 100% Power (23 5 WT, 2000 psi 2 36 2.27 2.61 and 5200F inlet temperature) 0 1.65 1.87 120% ' rower (1800 psi and 524 F 1.67 inlet temperature)
- 2. Hot Channel Factors:
Heat 11ux, F q 3 34 3 31 3 31 2.81 2 72 2 72 Entha1/Fise,bH 1
3 Analysis Parameter:
c-Steady State Inlet Mass Flow, lb/hr-ft2 7 99 x 10 MaximumHeatFlux, Btu /hr-ft 2 5 38 x 105 Equivalent Diameter (Unit Cell Perimeter), ft. 0.0381 Equivalent Diameter (Total Vetted Perimeter), ft. 0.0333 Heat Loss Coefficient Based on Flow Area in Subassembly Bottom End Plate 4.05 Grida(ForFourLevels) 3 92 Top End P. late -5 83 l
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As Table 4-2 showc, operation of the spiked 9x9 acaembly with RCC )
vill be no more severe than the previously analyzed and reported spiked 9x9 assembly and spiked 3x3 subascembly with RCC elements. .
Table b-1 gives the maximum peak power expected in each of the rods than to be used. No rod vill operate at higher peak power density has been previously reported and approved.
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i 4.4 SAITrl CONSIDD%TIONS )
l With the exception of the as-pickled Zr 4 clad rods, all of the f l
variables represented in this test have been reported and analyzed in Operation of this the Phase I Safeguards Report and its amendments.
fuel assembly vil.1 not exceed the limiting safeguards considerations for the Saxton core nor does it increase the severity of any of the Prior to insertion of this assembly previously ana\yzed accidents.
containing three Zr k rods in the as-pickled condition, the 3x3 RCC subassembly vill be removed and the two Zr 4 as-pickled rods examined to determine that there is no visibly detectable excessive corrosion.
45 CONCIUSIOIC From the foregoing analysis and description, it may be concluded that insertion and operation of thic special 9x9 assembly may be accomplished with no undue hasards.
4-6
50 MODIFIED HOLis FUEL ASSDGLY AND PI110 WRjiUITRRFAT TOGITION 51 PETRAM DESCRIPI' ION, OBJECI'IVES AND SCOPE The ob,)ective of this program in the ins'callation of a modified hollow fuel assembly which vill be used to house a superhest loop thimble which vill be tested in the future. Also included is the installation of a c.oecial removable plug assembly in the honow acaembly which vill prevent excesrive flow by-pass and flux peaking. The primary purpose cf the plug is to permit future installation and testing of the super-heat ic o without requiring reactor vessel head removal.
The modified fuel assembly consists of a normal Saxton 9x9 fuel assembly which has the 21 center fuel rods removed. The removal of all but the four corner rods of the central 5x5 cluster provides the 21 rod central hole. Specially designed spring clips are attached to the normal spring clip grids to provide support for the four corner rods and the plug. The plug consists of four concentrienlly mounted stainless steel tubes. The outer tube is velded to perforated stain-less steel end pieces which are designed to limit flow through the plug. The three irterior tubes are provided in order to limit flux peak!ng due to a water gap.
52 MECHANICAL DESIGN 5 2.1 Du=my Plug A detailed view of the diumy plug is shown in Drawing 5%F311. The th ce concentric stainless steel tubes are tied to the outer tube at-fourelevations(onapproximately12in, centers)by1/2in.x1/8in.
bars. These bars are net velded to the inner tubes but are velded to the outer tube. At each of the elevations there are two bars placed
- at 90 degrees to each other.
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The bottom end piece is a 1 in. thick, tapered stainless steel disc perforated with ten holes of 0.129 in, diameter. The taper on the bottom end piece is provided to permit easy insertion of the dumn plug into the hollow fuel assembly. The top end piece, also perforated with ten holes of 0.129 in. diameter, provides an er ;' path for coolant as well as servirt, as the connection point for the plug's support tube. The 15 in. diameter stainless steel support tube provides vertical support for the plug assembly as it is an extension of the Conoscal port flange. Six slots,1/2 in. vide by 7 in. long, equally spaced around the support tube near its bottom allow coolant from inside the dummy to return to the main coolant stream.
5 2.2 Modified Fuel Assembly To provide space for the superheat loop or its dummy plug, the central 5x5 fuel rode have been removed from a standard 9x9 fuel assembly. To insure as little fuel removal as possikle, the four corner fuel rods of the 5x5 assembly are retained by providing speciauy designed spring clips.
In addition to providing lateral restraint for the four corner fuel rods, these spring clips also provide guidance and lateral support for the superheat thimble or the dummy plug. Because of the large mass of the dimray plug, as compared to a fuel rod, these spring clips are provided with two spring fingers. If the one finger that is normally in conte.ct vita. the plug should deflect more than 0.0015 in.,
the plug vill come in contact with the second finger which will then prodde additional restraint against further deflection.
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With the exception of the removal of the central 5x5 porttions from the opring clip grids and the inntallation of the special opring clip, the mechanical design of the modifitd necembly does not differ from that of a normal fuel assembly.
53 Tla2Wa, ErDRAlTLIC AND NUCIEAR DESIGN The dumy plug has been designed so that the flow through it vill experience the name temperature rice that vould be produced in a normal fuel assembly. Because of the lover heat generation rate of the plug assembly, a smaller than normal flow vill exist through the plug assenbly. This condition vil.1 slightly increase the flow rate adjacent to the plus which vill insure more than normal cool.ing of the adjacent fuel rods.
Because of the increased water void due to the dummy plug, a flux peaking of about 20 per cent above normal vill be experienced by the adjacent fuel rodo. Ao the N-4 superheat position is on the outer edge of the core, this flux peaking vill not produce any abnormal.ly hot rodo. The fuel in this area of the core operates at about 20 per cent below average so the added peaking vill only cause these few adjacent fuel rodo to operate as average roda rather than below average.
54 SAFIHT CONSIDEPATIONS Implementation of this program introduces no new hazards nor dees it increase the severity of any of the prqviously postulated hazards which vere analyzed and reports in the Safeguardo Report for Phase I of the Saxton Nuclear Erperimental Corporation Five-Year Research and Develop-ment Program and its supplemento.
55 CONCIUSIOF3 It is concluded, therefore, that the above described experiment may be inserted and operated in the- Saxton reactor without any undue hazards to the public.
5-3
)
_ .- - - - ~ _ - - . . . - . - . _ ~ - - . . - _ - . - - . _ . _
6.0 MATERIAIE2 TESTING CAf*iUlff> AND GAW.A THERMOVfrER 6.1 PHCGidM OBJECTIVE AND SCOPE The ovemil objective of this program is to provide data on the effects of radiation on structural materiala used in nuclear power plants.
Present designs of nuclear reactors are primarily based on limited test 3
data obtained in test reactors rather than in an environment of an ;
actuni operating nucicar power plant as proposed in this program.
Primrily, the " brittle fracture characteristics" of reactor vessel materials, those presently in use as well as new materials, vill be evaluated.
The scope of this program vill consist of insertion and irradiation of a device for the measurement of gamma heating of solids (gamma thermometer) followed by insertion and irradiation of materials testing The dumg specimens in a dummy assembly at the periphery of the core.
assembly, located in position N-6, 'is underneath a head port. The gemma thermometer and materials capsules may therefore be inserted and 2emoved I
without a head removal. Following the capsule irmdiation and removal, the specimens vill be removed from the capsule and examined in a hot cell.
In order to remove t.he limitation of using only one irradiation position for the many aspects of this program, two additional capsule types have been designed for simultaneous testing of additional-metallurgical samples. One of these capsules vill be for use in two of the dummy ,
fuel positions .on the periphery of the core while the other capsules will be inserted-in two of the tub 9s in the baffle outside of the core.
These two capsule types are non-removable and require head removal for '
insertion into and removal from the core. ED SK 313095-D shows the N-6, dumn fuel and baffle tube positions which vill be used for the irradiation.
6-1
- . ._._ _ _ .._ _ .__,_ _ ._ _ _ _ _ _ . , . . _ . . _ _ _ _ _ _ _ . , . _ , . ~ _ . _ _
I 6.2 Dir2CRIFfl0N OF TIST DEVICES 6.2.1 Gama Themometer The samma themometer is a device to measure heating of solid materials due to absorption of gama radiation. Its function is to detemine the heating, and thus the temperature distribution, in non-fuel and structural materials. The device vill monitor a given position in the core to detemine the gamma heating at that point as a constant or slowly varying function of time. Its use in this capacity is especially i valuable in predicting the temperature distribution in irradiation sur.
- veillance specimens which are usually too small to monitor directly with themocouples.
The gaton themometer aseembly consisting of the gamma themometer,
, associated leads, port closure penetration and support tubes are shown in Drawing 647J195 A u material in the assembly in contact with the primary coolant is 18-8 stainless steel.
The ar sembly, when installed in the core, vill fit within the central hole in the dummy fuel assembly beneath the N-6 head nozzle penetration.
Insertion and removal of the assembly vill be made through the N-6 nozzle and does not require removal of the reactor head.
The sama thermometer is in the form of an esacuated 1.625 in. 0.D. by 5125 in. long cylindrical bulb with a solid disk main body section and cylindrical end caps which are velded to the body. - A specimen rod -
of the material under test (in this case ASTM A 302B steel) is press fit axially through the central mounting block such that the rod ends project out each side of the block approximately 1/3 and 2/3 of the rod length from the block center. The two unequal legs of the specimen rods are suspended in a vacuum and therefore lose heat primarily by conduction to the central mounting block vhich is in thermal contact-with the coolant. Thermocouples are uced to measure the temperature 6-2
d at each end of the specimen nnd at the mounting point. The temperature rise thus determined between the ends of the rod and the mountinc point vill provide a measure of the heat generated by gamma absorptio) in each leg of the specimen. The unequal leg lengths allow a cort *ction for heat loca by radiation from the rod to the cylinder vallo.
The thermocoupleo used for measuring the rod temperatureo are of no small a mass na practical (0.041 diameter shetith) in order to minimize gamma heating of the couples. The thermocouples are brought out of the Camma thermometer hulb through a brazed, vacuum tight joint in the bulb center body section. The vertical run of the thermocouples from the bulb to the pressure tight penetrations in the Conoseal adaptor is made' within a 1.0 inch diameter pipe to protect the couples from coolant croco flow forces. The couplen are oupported within the pipe and tied to the pipe by brazing to spacers at intervals along their length.
When installed in the reactor, the gamma thermometer assembly is hung from the male conoseal adaptor at the head penetration and receives its vertical support through the support tube velded to the adaptor. lateral support for the assembly io provided by the 2.231 inch diameter holes in the top and bottom du::gy nozzle platen and the 2.625 diameter bore in the du==y top nozzle. Diametral clearances at these pointo are 0.031 and 0.025 inches nominal respectively.
6.2.2 1rradiation capsule Assemblies The irradiation capsule assemblies are of three types depending on the location in the core at which they are to be used and they are identi-fied by their location. The types and number of each proposed at this time for Saxton and the e9sembly drawing number for each are licted below:
i 6-3 i
. . l
t
!heber M Proposed Draviv Number N-6 Nozzle Location 1 832D')1'?
! Dummy Fuel Assembly Location 2 ED SK 282264J .
(Non-Removable) '
Baffle Tube Location 2 $40F335 Typical.ly, the assemblies consist of tubular capsule subassemblies designed to receive and contain the irradiation samples and the addi-tional structure necessary to support the capsules within the core. -
The 16-inch length in cach capsule assembly which vill straddle the mid-height of the core is reserved for the irradiation samples. The ;
camples vill be inserted into each assembly as four discreet bundles, j of which typical examplec are shown in ED SK's 313094 and 313093 A deceription of the design details, construction, and method of support of a typical sample bundle and for each type capsule assembly follove.
6.2.2.1 Irradiation Sample Bundles The irradiation bundles are made as integral units to be slipped into the irradiation capsules at final assembly. Each bundle vill consist j
of various combinations of Charpy impact specimens, tensile specimens, and biaxial test specimens which vill be banded together as shown in ED SK's 313093 and 313094. Four such bundles vill be used in each irradiation capsule assembly.
The banding arrangement is such that-the specimens are tied to6 ether around the periphery at the top, center and bottom of the bundles with stainless steel straps. Additional straps are inserted between the specimens _and brazed to the peripheral straps to provide gaps for coolant circulation. The peripheral straps are tied to each other ,
by longitudinc.1 straps which are brazed to the peripheral straps on each of the four sides. Circular, perforated end platec are brazed l
to each end of the peripheral straps to complete the bundles. The 6-4
1 effect of the banding arrangement is to form a basket which supports '
the specimens securely, spaces the specimens to allov coolant circula.
tion between them, and provides a convenient means of handling the specimens.
Dacimeter wires, including Ni, Co, Cd chicided Co, or Cu, vill be inserted and seal velded into either selected materials test samples or into type 304 stainless steel tubes. The latter tubes vould be velded to the stainless steel straps or end plates. Lov melting point fusible eutectic alloys and metals (alloys or pure metals of Zn, Cd, Sb, Sn, Ag, Fb and Bi) vill be used as temperature monitors. They v4.11 be enclosed in ghss tubes which vill be inserted into holes machined into selected mterials test samples. The holes vil.3 then be plugged and seal velded shut. In the event that a ghss tube breaks, the molten alloy or metal vill be entirely contained within the steel sample.
i It is phnned to insert the temperature monitors in a maximum of two biaxial and three impact samples per assembly. Each biaxial sample vould contain four wires 1/2 in. long by 1/16 in diameter and each impact sample vould contain two such vires. One dosimeter tube vill be used per capsule assembly. The tubes are k-1/2 in. long by 1/8 in. 0.D.
I and vill contain up to three dosimeter wires 1/32 in. in diameter.
' Dosimeter wires inserted and seal velded into impact samples vill be 1 in. long by 1/32 in. diameter. Six vires vill be used in each of a maximum of three samples per capsule assembly. The dosimeters and temperature monitors are consistent with ASTM-E-184, " Recommended Practice for Effects of Hi6h Energy Radiation on the Tencile and Impact Properties of Metallic Materials."
i l
l 6-5 l
l E _ . _ . . . _ _ _ . _ . . . . __ _. . _ _ _ _ . _ _ _
_ m.__._______.. _ . _ _ . . _ _ _ . _ _ . . . . . , _ _
a i 6.2.2.2 Irradiation Capsule Assemb}y 6 flozzle Location ,
i
' The irradiation capsule assembly for the 14-6 nozrie location is made to be interchangeable with the gan:m thermometer accembly and vill be inserted in the reactor upon removal of the gamma thermometer. The assembly consists basically of the capsule subassembly, support tube, and conoseal adaptor, all of vhich are constructed of type 18-8 stainless steels. They are shown in drawings Sl+0F326 and 822D012.
t Vertical support for the capsule subassembly in the reactor is provided through she support tube by the conoscal adaptor at the head penetration. llorizontal support for the capsule is provided by the top and bottom dummy nozzle plates and the central bore of the top _
dumy nozzle. The nominal diametral clearances between the capsule and the dumy fuel assembly are 0.031 in the nozzle bore and 0.025 at i the nozzle plates.
The capsule subassembly is constructed of 2.0 inch diameter stainless steel pipe which is perforated radially with 0 5 inch vide by 15 inch long slots to approximately 30% metal removal. The clots pravide for coolant circulation te ;he irradiation sach bundlea which a m con- .
tained within the cap',ule.
I In assembly,- support plates for the sample bundles-are velded to the I.D. of the capsule through the coolant slots. _ The lover support plate is first velded in place nine inches below the cove centerline.
The cample bundles are then inserted in the capsule and the top r
support plate inserted and velded in place. After insertion of the j
l _
bundles, the capsule subassembly is velded to the unport tube.
I i
The capsule is provided with a bottom end plate which is tapered to aid in cuiling the capsule into place during insertion. The plate also i w as a backup to prevent the. sample bundles from dropping out of the capsule in the event that the bundle support plate breaks loose.
6-5 l
l u , _a. . -. _ _ - . _ . . _ _ _ _ _ _ . . . _ _ . _ _ . - . . . _-..._.. _ _..._. _ . _. _ . _ . _ .-
6.2.2 3 Irradiation capsule Accembly - Dumray Fuel Assemblypeations i
The irradiation capoule accemb.'ics for all dummy fuel locations excepting the N-6 nozzle location are non-removable; i.e., can be installed or removed only when the reactor head and upper instrumentation
- package are removed. All material of construction for the acccmblies is type 18-8 stainless steel. This capsule ic chovn on drawing ED SK 2B226hJ.
As with the capsule assembly at the N-6 nozzle location, these assemblies fit within t.nd are supported horizontally by the dummy fuel asocablies. Diametral gaps are identical in both ca.ses. In order to obtain a simplified design, however, the non-removable capsules in the du:: ray fuel ascembliec have been designed withcut a faxed hold-down.
The assemblies are supported in the vertical down direction by the top dummy nozzles. A tapered surface ou the capsule latch makes a metal to metal fit within the conical inside surface of the nozzle to support the weight of the espsule as well as to restrict its horizontal motion. In the up direction, the capsule weight acts as a restraint against the hydraulic lifting force.
To obtain maximum veignt, the top and bottom ends of the non-removable capsules have been made .of alanost solid steel. As a result, with rated flow, the assembly weight is cvt r four times the hydraulic lift-ing forces. Assumin6., as a vorst condition, that the flow through the upper portion of the dummy assembly were completely cut off by the capsule assembly, the vef ght would still be tvice the hydraulic up load.
. In order to minimize horizontal motion at the bottom of the capsule assembly, leaf springs which bear against the I.D. of the bottom nozzle plate are mounted at four locations on the periphery of the capsule bottom plate. The springs are deflected nominally to a 4 lb. load at insertion of the assembly and restrain the assembly against small 7
- = _ _ _ _ _ _ _ _ _ _ - _ . - \
m36nitude vibrations. Diametral clearance between the capsule and noztle plate at this location is 0.039 inch. The spring clips also add an additional 5 6 to 10 9 lb. restrcining force against up motion.
The provisior.s for the sample bundles and method of loading in this assembly are the same as with the capsule at the N-6 nozzle location.
In this case also, the capsule bottom plate acts as a backup in the event the sample support plate breaks loose and is tapered to provide lead-in during insertion of the capsule in the reactor.
6.2.2.4 Irradiation Capsule Assembly - Irradiation Tube Location The capsule asacmbly to be used in the irradiation tube (baffle) location are of the non-removable type; i.e., require removal of the reactor head and upper instrumentation package for installation or removal of the capsule. The assembly consists of the tubular capsule subassembly and a spring loadedh31d-down device at the apper end.
I All material except for the spring is 18-8 stainless steel. The sprin6 is constructed of age hardened Inconel-X. The capsule is shown on Drawing 51 6 335 The capsule subassembly is similar to those in the dummy fuel assembly locations but of smaller diameter (1.25 inch schedule 10 l pipe). The sample bundles in this case are smaller in diameter but i otherwise identical to those for the other locations. In addition, l
l the method of loading and supporting the bundles in the capsule is the same as that in the other capsules.
l The capsule assemblies ir the irradiaton tube are supported vertically b; the center plate in the irradiation tube. A tapered ooss on the O.D.
of the capsules makes a metal-to-metal fit within a chamfer on the top surface of the central bore in the irradiation _ tube. This arrangement i supports the tube hn'4 tontally as ve.11 as vertica31y and prevents horizontal movement of the lower portion of the capsule, i
6-8 .
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9 The hold-down assembly on the top of the capsule is compressed by the i upper core plate to a nominal vertical load of 80 lb. This is more than nine times the hydraulic lifting force on the capsule. The contact between the hold-down and the core plate also provide a frictional force of k0 lb. to restrict horizontal motion of the upper end of the capsule.
63 S/JTTY CONSIDERATIONS Since the capsule assemblies and Samma thermometer vill be located outside of the core proper, they vill have only a small effect on the core. Based on calculations of the effect of the stainless steel material in the du=:::y assembl;les, the addition of the test materiale vill result it only a 0.1 to 0.2 per cent increase in reactivity and a 4 per cent decrease in power in those fuel assemblies adjacent to the du:may assemblies. However, with only three dummy positions in use, the increase in the central assembly power should be tuch less thar 1 per cent which is within instrument error associated with 3
power measurement.
Possible failure of the gamma thermometer during the test must be considered. Should the device fail, it will be by a short or open f in the thermocouples inside the capsule body, or by leakage past a veld into the cpsule. Neither event vill affect reactor operation nor create a path for radioactive materials (other than coolant) to leave the reactor, The effect of a possible increase in crud levels due to the use of -
unclad carbon steel specimens must be considered. The crud from unclad carbon steel specimens in the test capsules, estimated at 3000 sq. in., vill increase the crud level approximately 50 ppb.- This estimate is based on conservative calculations and takes no bredit for crud removal by demineralizers or filters. A limit of 2000 sq. in.
6-9
of test specimens has been set as not requiring corrosion protection.
It is planned that any specimens which would result in this limit being exceeded will be plated for corrosion protection. Hence it may be concluded that the insertion of the t -t capsule vill not siGnificantly affect crud levels.
The possibility of loss of irradiation samples has been a consideration in the design of the copsule assemblies. The samples vill be securely l
strapped together into bundles within the capsule assemblies. In addition, the capsule assemblies have been provided with bottom plates as backups to contain the bundles should the plates supporting the bundles in the capsules break loose. As a further consideration, should j
the bundles themselves break up and the individual samples fall out of the capsule assemblies, the samples vould be contained within the dummy fuel assemblies or the baffle irradiation tubes and would pose no safety problem for the reactor. Adequate coo.' ant flow is provided through the baffle tubes and dtm:my fuel assemblies to prevent excess ga=ma heatin6 j
of the irradiation samples and the ga m a thermometer.
6.4 CONCLU6 IONS Frca the above analyses, it may be concluded that insertion of the gamma thermometer and irradiation capsu'.e assemblies vill *.ntroduce no new hazards nor vill it increase the severity of eny r,re.iously -analyzed hczards.
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4 TABLE OF DP.AWINJS. -
Referred to on Page No. Subject Drawir.g No.
3-3 pellets 500B0h9 3-h spring clip 882D096 3021003 h-1, h-h assembly 5-1 dummy plug Sh0F311 3130953 6-1 dumn & barfle tube positions 6h7J195 6-2 gamma thermometer 882D012 6-h, 6-6 nozzle location capsule assembly 6-h, 6-7 dummy assembly location ;'
28226hJ 5h0F335 6-L, 6-8 capsule - battle tube location 6-h irradiation samples 31309k 4 i
5-b irradiation samples 313093
$h0F326 66 capsule assembly I
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March 2h, 196h
- 1 Docket No. 50-lh6 DPR-h Technical Specifications Change Request #1h (Page 1 of k pages)
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- 1. _ Description of Change In Supplement No.1 to Technical Specifications, page 2, change sectior labeled " Change Section F.2" to read:
i Uranium oxide (UO 2 ) enriched to 5.7% of U-235 shall be used in the fuel l assemblies, except that the test fuel assemblies listed below having enrichments l and other characteristics as described may be inserted in the reactor. In test fuel assemblies the fuel rods as described may be replaced with regular fuel rods, that is, enriched to nominal 5 7% U-235 and constructed as described in Technical Specification F.3 Test Assembly No. 1 One 61-rod assembly containing rods of the numbers and types listed in the following Table Clad Pellet Peak No. Power of Rods Cladding Thickness (l) Diameter Enrichment h(h) 30h SS 80.5 mils (ll) 0.29h in. 0.71w/o 3.1kw/ft. f 2.2 h(b) 30h SS 80.5 0.29h 0.29w/o l
- 3. 0.357 (2) 16 l 3 30h SS 15 0.357 (2) 16 3 30h SS 15 0.357 (2) 16 3 30h SS 15 0.357 (2) 16 3 30h SS 16-20 SS 15 0 357 (2). 16 3
15 0.257 (2) 16 i 3 3h8 SS . 3 30h SS(9) 15 0.357 5.69 w/o 13.5 f l 3 30a SS 15 0 357 5.69w/o(3) 13.5 Zr-h(6) 23.7 0 337 5.69w/o 12.0 3 y Zr-h(6) 23 7 0.337 6.1w/o 13.5 3 Zr-2(6) 23 7 0.337 -(2) lh 3 72-2(6)(Mi free) 23 7 0 337 (2) lu Zr-h(6) 23.7 0 337 (2) 1h 3 Zr-h(10) 23 7 0.337 (2) 1h 3 Zr-h(0) 23.7 0.337 (2) 1h 3 Zr-h(I) 23.7 0.337 (2) 1h 3 3 Zr-h(6) 23.7 0 337 7 3 w/o 16 1653
.- . - - . - . . .- - - . - --..- - . _ _ .~. -.
March 2h, 196h Docket No. 50-lh6 DPR-b Technical Specifications Change Request #1h (Page 2 of h pages) Notes for Table (1) All rods are free standing 0.391 in. O.D. nominal (2) First ih pellets 5.69w/o next 5 pellets 6.81w/o next 12 peitets 6.h5 w/o next 5 pellets 6.81w/o next 13 pellets 5.69 w/o (3) Contains approximately 100 ppm boron as circonium diboride (h) RCC element with perforated guide tube (5) RCC element with solid guide tube (6) Autoclave pre-oxide on 0.D. (7) Autoclave pre-oxide on 0.D. and I.D. (8) Furnace pre-oxide on 0.D. (9) Compart .ented red, 3 sections (10) As pickled, no pre-oxidc treatment (11) RCC rod 0.D. is 0.h61 in, nominal j Test Assemblies 11 and iii Test Fuci Test Fuel Assembly No. 11 Assembly No. iii 9-Rod Subassembly 9-Rod Subassembly First lh pellets 5.69% 5.69% Next ; pellets 9.19% 7.30% Next 3 pellets 8.57% 6.81% Next li pellets 6.13% '6.h6% Next 3 pellets 8.57% 6.61% Next 2 pellets 9.19% 7 30% Next 1h pullets 5.69% C.69% Note: The 9-rod subassembly in the first column shall not be uscd E.t reactor pcwer levels g eater than 20 MJt. Test Fuel Assembly No. iv
- One 9-rod subassembly shall have four cerner _ rods clad with Zircaloy-b having a nominal thickness of 23 7 mils and shall contain uranium oxide (UOp) enriched to 6.1% U-235 The other five rods shall be clad with Type 30h stainless steel having a nominal thickness of .5 mils and shall contain uranium oxide (UO3) enriched to C 7 % U-235 r
March 2h, 196h Docket No. 50-lh6 DER-h Technical Specifications Change Request plh (Page 3 of h pages) Test Ftel Assembly No. v One 9-rod subassembly shall have four corner rods clad with Zircaloy-L having a nominal thickness of 23.7 mils and shal3 contain uranium oxide (UO2)1 -iched to 6.1% U-235. The other five rods shall be clad with Type 30u stainless steel having a nominal thickness of 9.5 mils and shall contain uranium oxide (UO2 ) having the same enrichment as Test Fuel Assembly No. iii.
. Test Fuel 'ssembly No. vi One b-rod subassembly shall have rods clad with Type 30h stainlens steel having a nominal thickness of 23.5 mils and shall contain uranium oxide (UO2) fuel pellets uniformly enriched to 8.3% U-235. One of these roos may contain up to 100 ppm boron as circonium diboride.
? Test Fuel Assembly No. v11 One 9-red subassembly shall have the center rod and four corner rods j clad with Zircaloy-b having a nominal thickness of 23.7 mils and shall con-tain uranium oxide (UO 2 ) uniformly enriched to 7.3%. Two of the other rods shall be clad with Type 30h stainless steel having a nominal thickness of 15 mils and shall contain uranium exide (UO2 ) uniformly enriched to 5 7% U-235. One other rod shall be clad with Type 30h stainless steel having a nominal thickr.ees of 16.1 mils, shall contain uranium oxide (UO2) having a content of 0.29% U-235, at 1 shall be conenntrically located within a solid stainless steel guide tube. The remaining rou shall be clad with Type 30h stainless steel having a nominal thickness of 16.1 mils: shall contain uranium ox3de (UO 2
) having a centent of 0.71% U-235 and shall be con-centrically located within a perforat ed stainless steel guide tube.
Test Fuel Assembly No. viii One 9-rod subassembly shall have th ce corner rods clat with Zircaloy-h having a nominal thickness of 23.7 mils and shall contain vibrationally. compacted v. .niun dioxide (UO2) enriched to 7.2% U-235 and compacted to 86 2 2% theoretical density. Thc fourth corner rod and the central rod shall be clad with Type 30h stainless steel having a nominal thickness of 15 mils and shall contain vibrationally compacted uranium dioxide (U32 ) enriched to 7.2% U-235 and compacted to 8612% theoretical density. Three of the remaining uranium dioxide 2 (rods shall be clad with Zircaloy-h and shall containUO ) pelle to 6.1% U-235 One of these rods shall have a previous irradiation exposure of approximately 7500 megawatt days per metric ten (M4D/MT) and-shall contain a 15-mil diameter hole machined through the clad. The second of these rods shall have a previour irradiation exposure of approxi-mately 7500 MJD/MT. The third of these rocs shall have a lf-mil diameter hole machined through the clad. The final rod shall te clad with sensitized Type 30h stainless steel and shall contain uranium dioxide (UO2 ) Pellete enrithed to 5.69% U-235 and the ten central pellens shall have 20-mil chamfers en both ende.
l March 2h, 196h ;
. Docket No. 50-lh6 DPR-h Technical Specifications
, Change Request #1h l (Page h of h pages) Following a period of irradiation, the two defected, Zircaloy-b clad rods may be replaced by similar, defected, unirradiated Zircaloy-b clad rods. l ' Fuel Assembly No. ix One 9 x 9 fuel assembly shall contain 51 rods clad with Type 30h-stain-2 less steel of 15 mils thickness and shall containis uranium This assembly dioxide (UO ) made by removina fuel pellets of 5.69% U-235 enrichment. The space left by the central 21 rods from a normal 9 x 9 fuel assembly. removal of the central el rods shall be filled by a plug consisting of a stainless steel tube o.125 inch thick and The end 2.75plugs inches shallinbediameter designed welded to perforated stainless steel end plugs. so that flow
- rough the plug will experience the same enthalpy rise that is experiene by flow through a normal fuel assembly. The plug shall contain three concentrically mounted stainless steel pipes 0.125 Horicontal inch thick and of 2.125,1.50 and 0.75 inch diameters, respectively.
restraint for the plug shall be provided by the Erids of the fuel assembly. Vertical support for the plug shall be provided by a 1.5 inch diameter stainless steel pipe extension of the reactor head port flange. Uranium oxide being used has a melting point of approximately 5000 F.
- 2. Safety Considerations Safety considerations are studied in Addendum No. 3 to the SafeguardsIn I Report for Phase 1 of the Saxton Five-Year Research and Development Program.
our opinion the proposed change does not present significant hazards considera-t tions not described or Lmplicit in the Final Safeguards Report. l 3 Health and Safety of the Public It is our conclusion that-the health and safety of the public will not be endangered by this change. l l 1 l
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