ML20085E694
| ML20085E694 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 05/07/1965 |
| From: | Neidig R SAXTON NUCLEAR EXPERIMENTAL CORP. |
| To: | |
| Shared Package | |
| ML20083L048 | List:
|
| References | |
| FOIA-91-17 NUDOCS 9110210115 | |
| Download: ML20085E694 (14) | |
Text
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SAXTON IR1 CLEAR EXPERIMElirAL CORPORATION DOCEET NO 53 146 LICENSE DPL 4 Change Raquest /20
- 1. Applicant hereby submits Change Request No. 20 in comfisnee with paragraph 3B of License DPR.4 for change of the Technical Specifications to be authori::ed by the Cocaissiou as provided in 10 CFR 50 59 SAL 70N NUCLEAR EXPERIMEITTAL CORPORATION By _[c/ R. E. Neidig President ATTEST:
/s/ R. E. Sveher Secretary (S E A L)
Sworn and subscribed to before me this 7th day of May , 1965 l (S E A L)
/s/ Martin A. Kohr j Notary Public j Muhlenberg Township, Berks County l }ty Commission Expires Feb. h,1966 I
i 911021 ';15 910424 PDR 20IA l DEK0K91-17 PDR 1 e- -
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Dacket No. 50-146 DPR-4 Technical Specification Change Request No. 20 page 1 of 12 Pages
- 1) Description of Chance In Supplement No. 1 to Technical Specifications, page 2, change Item F.2.
to read:
Uranium oxide (Uo ) enriched to 5 7% of U-235 shall be used in the fuel assemblies, excepk that the test fuel assemblies listed below having enrichaents and other characteristics as described may be inserted in the reactor. In test fuel assemblies the fuel rods as described ma i replaced with regular fuel mds, that is, enriched to nominal 5 7% y be U-235 and constructed as described in Technical Specification F.3 Test Assembly No. i one 61 rod assembly containing rods of the numbers and types licted in the following table:
No. of Clad Pellet Rods Claddinj Thickness (1) Diameter Enrichment Peak Power
- 4) 304 SS 80 5 mils (11) 0.294 in. o.71 v/o 31kv/ft 4(5) 304 SS 80 5 0.294 0.29v/o 2.2 3 30L SS 15 0 357 (2) 16 3 30k SS 15 0 357 (2) 16 3 304 SS 15 o.357 (2) 16 3 30L SS 15 0 357 (2) 16 3 16-20 SS 15 0 357 =(2) 16 3 3k8 SS 15 0 357 (2) 16 3 304SS(9) 15 0 357 5.69w/o 13 5 3 304 SS 15 0 357 5.69v/o(3) 13 5 3 Zr 4(6) 23 7 o.337 5.69v/o 12.0 Zr-g(6) 5 23 7 o.337 6.1v/o 13 5 3 Zr-2(6) 23 7 o.337 (2) 14' 3 Zr-2(6) 23 7 o.337 (2) 10 (Ni free).
3 Zr-4(6) 23 7 o.337 (2) 14 3 Zr-k(lo) 23 7 o.337 (2) 14 3 Zr 4(8) 23 7 o.337 (2) ]4 3 Zr-L(7) 23 7 0 337 (2) 14 3 Zr-4(6) 23 7 0 337 7 3 v/o 16 l
4 DFR L Tecimical Specification Change Request No. 20 Pa ge 2 of 32 Pages Notes for Table l (1) All rods are free standing 0 391 in o.D. nominal.
(2) First 14 pellets 569v/o l next 5 penets 6.81v/o l next 12 pellets 6.45v/o l next 5 pellets 6.81v/o next 13 pellets 5.69 v/o
- 3) Contains approximately 100 ppm boron as zirconium diboride k) RCC element with perforated guide tube
- 5) RCC element with solid guide tube l (6) Autoclave pre-oxide on 0.D.
(7) Autoclave pre-oxide on 0.D. and I.D.
(6) Furnace pre-oxide on 0.D.
(9) Compartmented rod, 3 sections l (10) As pickled, no pre-oxide treatment (11) RCC rod O.D. is 0.461 in, nominal, l
Test Aseemblits 11 &nd iii
. Test Fuel Test Fuel Assembly No. 11 Assembly .'io. iii l 9-Rod 9-Rod i Subassembly Subassembly First 14 pellets 5.69% 5.69%
Next 2 pellets 9 19% 7 30%
Next 3 penets 8 57% 6.81%
Next 12 pellets 8.13% 6.46%
l Next 3 penets 8.57 % 6.81%
, Next 2 pellets 9 19% 7 30%
! Next 14 pellets 5.69% 5 69%
NOTE: The 9-rod subassembly in the f.iret column shall not be used at reactor power levels greater than 20 IGt.
l Test Fuel Assembly No. iv One 9-rod subassembly shall have four corner rods clad with zirealoy-4 having a nominal thickness of 23 7 mils and shall contain uranium oxide (UO2) l enriched to 6.1% U-235 The other five rods shall be clad with Type 304 l stainless steel having a nominal thickrass of 9 5 mils and shall contain uranium oxide (UO2 ) enriched to 5 7% U-235
) DPR-4 Technical Specification Change Request No. 20 Page __3_ of 12 Pages Test Fuel Assembly No. v One 9-rod subassembly shall have four corner rods clad with Zircaloy-4 having a nominal thickness of 23 7 mile and shall contain uranium oxide (UO2) enriched to 6.1% U-235. The other five rods -hall be clad with Type 304 t
stainless steel having a nominal thickness - uils and shall contain uranium oxide (U32 )having the same enrie' Test Fuel Assembly No. iii.
Test Fuel Assembly No. Vi One four-rod subassembly shall have rods clad with Type 304 stainless steel having a nominal thickness of 23 5 mils and shall contain uranium dioxide (UO 2
) fuel pellets uniformly enriched to 8 3 v/o U-235 One of these mds may contain up to 100 ppm boron initially as zirconium diboride. Two of these rods may be replaced with 23 5 mil thick Type 304 stainless steel clad rods containing vibration compacted uranium dioxide powder enriched to 6.0v/oU-235 These latter two rods shall initially contain approximately 500 ppm of boron as boron carbide (B4 0).
Test Fuel Assembly No vii One 9-rod subassembly chall have four corner rods and the center rod clad with Zircaloy-4 having a nominal thickness of 23 7 mils and shan contain uranium oxide (UO 2
) unifomly enriched to 7 3%. Two of the other rods shall be clad with Type 304 stainless steel having a nominal thickness of 15 mils and shall contain uranium oxide (UO2 ) unifomly enriched to 5 7%
U-235 One other rod stall be clad with Type 304 stainless steel having a no=inal thickness of 16.1 mils, shall contain uranium oxide (UO2 ) having a content of 0.29% U-235, and shall be concentrically located within a solid stainless steel guide tube. The remaining rod shall be clad with Type 304 stainless steel having a nominal thiczness of 16.1 mils, shan contain uranium oxide (UO 2 ) havin6 a content of 0 71% U-235 and shall be concen-trically located within a perforated stainless steel guide tube.
Tect Fuel Assembly No. viii One 9-rod subassembly shall have three corner roas clad vith Zircaloy-4 having a nominal thickness of 23 7 mils and shall contain vibrationally compacted uranium dioxide _(UO 2 ) enriebed to 7 2% U-235 and compacted to 86 3 2% theoretical density. The fourth corner rod and the central rod shall be clad with Type 304 stainless having a nominal thickness of 15 mile and shall contain vibrationally compacted uranium dioxide (UO2 ) enriched to 7 2%
U-235 and compacted to 8523 % theoretical dansity. Three of the remaining rods shall be clad with Zircaloy-l and shall ccatain uranium dioxide (UO2) pellets 0 337 inches in diameter which are enriched to 6.1% U-235 One of these mds shall have a previous irradiation exposure of ~7500 megawatt days per metric ton (MWD /M) and shall contain a 15 mil diameter hole machined through the clad. The second of these rods shall have a previous irradiation exposure of ~7500 MWD /E. The third of these rods shall have a 15 mil diameter hole machined through the clad. The final rod shall be clad with sensitized Type 304 stainless steel and shall contain uraaium dioxide (UO )
pellets enriched to 5.69% U-235 and the ten central pellets shall have 202 mil chamfers on both ends.
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DPR-4 -Technical Specification Change Request No. 20
, Page 4 of 12 Pages Following a period of irradiation, the two defected, Zircaloy-4 clad rods may be replaced by similar, defected, unirradiated Zirealoy-4 clad rods.
Test Fuel Assembly No. ix one 9 x 9 fuel assembly shan contain 51 rods clad vith Type 304 stainless steel of 15 mils thickness and shall contain umnium dioxide (UO 2
) fuel pellets of 5 69% U-235 enrichment. This assembly is made by removing the central 21 rods from a nomal 9 x 9 fuel assembly. The space left by removal of the central 21 rods shall be filled by a plug consisting of a stainless steel tube 0.125 inches thich and 2 75 inches in diameter velded to perforated stainless steel end plugs. The end plugs shall be designed so that flov through the plug vill experience the same enthalpy rise that is experienced by flow through a nomal fuel assembly. The plug shall contain three concentrically mounted stainless steel pipes 0.125 in, thick and of 2.125,1 50 and 0 75 in.
dinneters, respectively. Horizontal restraint for the plug shall be provided by the grids of the fuel assembly. Vertical support for the plug shan be provided by a 15 in, diameter stainless steel pipe extension of the reactor head port flange. When Change Request No.16 has been approved by the Atomic Energy Commission, this plug may be removed and replaced by a fueled super-critical loop pressure tube assembly.
, Test Fuel Assembly No. x One 9-rod subassembly shall have eight rods clad with Zircaloy-4 having a nominal thickness of 23 3 mils. The fuel shall be mixed natural uranium-plutonium dioxide enriched to 6.6 v/o pug 2 . Four of the rods shall contain vibration compacted fuel and the remaining four shall contain sintered ceramic penet fuel. The ninth rod position vill be a flux thimble for use with in-core instrumentation.
Test Fuel Assembly No. xi One 9-rod subassembly shall have the center rod and the four corner rods clad with Type 304 stainless steel having a nominal thickness of 15 mile and shall contain uranium dioxide (UO ) unifomly enriched to 5 7% U-235 Two of theremaining-odsshallbecladhithZircaloy4havinganominalthicknessof 23 mils and shall contain ursnium dioxide (UO 2 ) uniformly enriched to 6.45 v/o U-235 The third rod shall be clad with Zircaloy 4 having a nominal thickness of 23 mils and shall contain uranium dioxide (UO 2
) unifomly enri hed to 6.1
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v/oU-235 This rod shall have a previous irradiation exposure in excess of 13,000 KJD/IEU. The last rod chall be clad with Zircaloy-4 having a nominal thickness of 23 mils and containing uranium dioxide (UO 2 ) unifomly enriched to 7 3% v/o U-235 and shall have a previous irradiation exposure in excess of 3000 M4D/MTU. The test subassembly may be inserted into the core vhen Change Request No.18 is approved by the Atomic Energy Commission.
DPR-4 Technical Specification Change Ecquest No. 20 Page i of 12 Pages Test Capsules Test capsules containing non-fuel material may be inserted in any of the eleven du=my fuel locations adjacent to the reactor corc gion or in any of the eight irradiation sample tubes on the periphery of ! core provided that:
- 1) Prior to irradiation, the design of the test capsule has been evaluated by the SNEC Safety Committee and found acceptable with regard to physical, themal and hydraulic performance and effect on core reactivity, neutron flux and reactivity coefficients.
- 2) No foreseeable failure of a test capsule could result in damage to any core co=ponent or in any manner alter the ability of the control system to function.
In Supplement No. 1 to the Technical Specifications, page 3, change N.4.e.(4).(b):
(b) 4 rod spiked subasseubly, Btu /hr ft 577,600 In Supplement No.1 to Technical Specifications, page 4, change N.4.e.(8).(b):
(b) In 4 rod spiked subassembly, kv/ft 26 5
- 2) Purpose of Change The purpose of this change is to pemit movement of the 4-rod subassembly into the central location and to pemit replacement of two rods of the 4-md subassembly with new rods containing vibration compacted loose oxide fuel and a burnable poison (boron carbide).
The scope of the second experiment involves the insertion of the unmodified test fuel assembly vi into the N-1 centml location for irradiation at high specific power levels. During the irradiation, measurements vill be made to correlate reactor power level and subassembly peak-specific power. - Following the first period of irradiation, test fuel assembly vi vill be removed and two of the fuel rods replaced by two new fuel rods which are of the same I geneml mechanical design but are of lower enrichment and vill contain a burnable poison (boron carbide) intimately mixed with vibration compacted, loose powder oxide fuel.
The objectives of this experiment are to provide information on the behavior of loose oxide fuels containing B 40 when operated at high specific power ratings. In addition, the previously irradiated pelletized rods will be subjected to higher bumup.
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DPR-4 Technical Specification Change Request No. 20 Page 1 of lh Pages ,
The present experimental fuel subassembly vi was inserted into a peri-pheral location in the Saxton reactor and war described in Addendum 1 dated '
December 20, 1962 to the Safeguards Report for the Esxton Phase 1 Research and Development Program. A detailed description ol .he mechanical design and themal and hydraulic parameter for this subassembly in the peripheral location are presented in the Addendum.
- 3) Safety Considerations Details of the mechanical design features of the four rod subassembly are presented in A* $endum 1 to the Safeguards Report for Phase 1 of the Saxton Five-Year Research and Development Program, dated December 20, 1962.
The principal design features of the old and new rods are presented in Table 2.
TABLE 2 FUEL ROD CHARACTERISTICS k-ROD SURASSEMBLY vi Pelletized Fuel Loose Oxide Fuel
- I Number of Rods 2 2 Active Fuel Length, in. 33 2 33 2 Rod Outside Diameter, in. 0 5995 0 5995 Clad Thickness, in. 0.0235 0.0235 Pellet Diameter, in. 0 5435 -
l Cold Pellet-to-Clad-Gap, in. 0.009 (diametral) -
Pitch, in. 0 746 0 746 l Enrichment, w/o U-235 8.3_ 6.0 (Unifom throughout md)
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- The loose oxide fuel rod claddin6 is designed and fabricated according to the same specifications that vere used for the pelletized rods. The.
cladding is Type 304 stainless steel and is designed to be free standing, and at the maximum reactor pressure and tempezuture defomation vill not exceed the elastic limit.
+ Contains approximately 500 ppm of boron as B4C dispersed in the loose oxide fuel. Density of loose oxide fuel is 87 + 2fb theoretical density.
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DFB-4 Technical Specification Change Lequest No. 2 0 Page 7 of g Pages The principal thermal and hydraulic parameters calculated for the operation of the 4-rod subascembly in the peripheral location and in the central location are listed in Table 3 They are compared with the data for a spiked 3x3 in the central location.
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$$ TABLE 3 is THEINAL ANL HYDRAULIC DATA
- h-ROD SUBASSEMBLY v1
$; Peripheral 4 ~pe.l Location Centml Locotion 9'. 2x2 2x2 Spiked 9-Rod Subassembly Subassembly Subacsembly T i nlet, F 520 520 520 Core Pover, K4t ~23 5 20.0 3 23 5 Mode of Control Chemical Shim Chemical, Shim Chemical Shim ,
Ruinum Ther=al Output, kv/ft 20.0 26.5* 16.0 MaximumSurj*aceHeatFlux, 444,000 577,600 533,000 Btu /hr-ft Hot Channel Factors F 2 76 1 72 3 31 F
3 1.82 1.22 2 72 Min. mum D!B Batios by W-2 Correlation 100% Power - 2000 psia q"-DNBR = 2.81 q"-DIGR = 2. 59 q"-DNPn = 2.27 120% Power - 1800 psia q"-DNBR = 2.03 q"=DNBR = 1.86 q"-DIGR = 1.65 4
The potential for burnout of the 4-rod subassembly in the center position is less than that for the spiked 9-rod subsasemb'.y previously approved for oper1 tion in the central location. Table 3 lists the minimum DNB ratios for the normal operation of the 4-rod subassenhly as 2 59 calculated by the W-2 carrelation as co= pared with 2.27 for the spiked 9-rod subassembly. The overpower-underpressure DNB mtios for the 4-md and 9-rod subassemblies are 1.86 and 1.65, respectively.
The highest nominal specific power of a pelletized rod in the 2x2 subassembly with the Saxton reactor operating at 20 Mdt is calculated to be 2L.2Kv/ft. A calculational uncertainty of 10% is applied to make the maximum specific power 26 5 Kw/ft. These power levels are based on the following hot channel factors: F = 1 72, F3 1 The lover enrichment, vipacfuelrodsvilloperateatahminalmaxfmu=mo.22.f 19 2 Kv/rt and at a peak of 21.1 with the 10% calculational uncertainty factor added and based on the same hot channel factors.
- Maximum is calculated for 8.3 v/o pelletized rods o: 1y. Maximum for new 6.0 v/o loose oxide rods is 21.1 kv/ft.
DPR-4 Technical Specification Change Request No. 20
. Page 6 of 12 Pages Limited center melting of the pelletized fuel is expected if the rods are operatedat26.5Kv/ft. About 2% of the hot spot cross section vill be molten ,
under these conditions. No center melting is expected in the vipac rods at 21.1 Kv/ft because of their lower power level and because of the flux depres-sion caused by the BgC in the fuel. The maximum calculated center temperature <
for the vipac fuel at 21.1 Kv/ft is 2400 *C which is well below the melting point of 2800*C. Center melting in vipac fuel vould not be expected to occur unless a linear power density of 24 Kv/ft is reached.
Experimental data from G. E. , Hanford and Westinghouse vork indicate that there are no serious problems involved in operating fuels up to 30 Kv/ft with center melting. Hanford programs hav9 irradiated lov enrichment Pu0 2 -UO2 fuelpelletsupto38Kv/ft. G. E. programs in the GEIR using solid fuel pellets have experienced peak surface heat fluxes of 1 x 10D Btu /hr-ft (thic corresponded to an average rod power of 30 Kv/ft) without failure. Westinghouse programs in the Plum Brook reactor have run successfully at more than 30 kv/ft.
The maximum overpower tmnsient in Saxton might cause the power to go to 120% of the nominal before scram would occur. This considers errors in setpoints and instrument deadband. The peak power in the 2x2 vould then be about 31 Kv/ft and about 15% of the hot spot cross sectional area vould be in a molten condition. The solid portion of the fuel vill keep the molten fuel away from the clad and prevent any interaction between the two. The TREAT tests have shown that even in fairly rapid transients which involve only limited center melting that the molten fuel does not pass through the solid fuel and contact the cladding. These tests have also shovn that in all but the most rapid transients, the temperature profile acmss the fuel remains parabolic. Only the very shortest period (~ m see) transients could provide the required rapid energy input into the fuel to cause the outer layer of-fuel to melt and react with the clad.
Since tM mechanisms required to create rapid reactivity trcnsients are nrt present in the Saxton reactor, operation of the 2x2 vith 13mited center melting is considered u safe experiment and that gross failure of the 2x2 due to the presence of center melting is not credible.
(1) Deidentaum, B., "High Performance UO 2 Program, Quarterly Progress Reports,"
GEAP-3TT1-1 to GEAP-3771-15 (2) Roake, W. E., " Irradiation Alteration of Uranium Dioxide," IN-73072.
(3) Westinghouse data unpublished. To be presented at the June,1965 ANS Meeting.
DPR-4 Technical Specification Change Request No. 20 Page 9 of 12 Pages 4
In addition to the cafety margin calculated for the k-rod subassembly, the flux vire and pitot tube thermocouple probe vill provide measuremente of the actual operating conditions to assure that the specific power limit of 25:5 Kv/ft vill not be exceeded. The data vill allow correlction of the reactor power level and the maximum specific power of the subasse=bly. The reactor may be operated at power levels above the limit of 20 FWt as shown in Table 3 if the measurements indicate a lower specific power than calculated in Table 3 Modification of the 4-rod assembly to permit insertion of the new rods requires the removal of the flux vire thimble from the assembly. The presence of the new, lower enrichment rods vill not significantly alter the peak specific power in the old rods so that the reactor may operate with the modified 4-rod assembly up to the power level determined as limiting in the first irradiation perioa without exceeding the 26.5 kv/ft limit.
A study was performed using the Westinghouse W-2 DNB correlation to demonstrste that the 2x2 subassembly, when operated in the central location at specific power densities up to 26.5 kv/ft, vill not be damaged under accident conditions. The accident analysis for the 2x2 assembly when operated in a peripheral location at calculated power densities of 20 kv/ft covered the two cases of loss of flov accident and the boron removal accident. Because the loss of flow accident produced a lower DNB ratio in the 2x2 than does the boron removal accident, the boron removal accident vill not be repeated for this analysis.
The loss of flow accident was assumed to occur due to a loss of pump power and the flow coastdown curve of the main coolant pump is given in Figure I. The reactor and 2x2 parameters that were assumed for the accident are presented in Table 4. Reactor scram is assumed to start 1 5 reconds after the beginning of the accident and is completed at 2.4 seconds. The delay in scram initiation and completion is due to:
- 1) Flov decay to the low flow scram set point and error in the set point - 0 7 sec.
- 2) Inctrumentation delay - 0.2 sec.
- 3) Rod motion in a region of small effectiveuess - 0.6 see.
k) Time for rods to complete motion in region of effectiveness - o.9 sec.
The results of the loss of flow analysis are presented in Figure II and compared with the results for the previous analysis presented in Addendum 1.
For the 20 5 ha/ft operation, the minimum DNB ratio is a Q-DNBR of 159 that occurs at 2 7 seconds after the start of the accident. The Q-DNBR was evaluated assuming a rescetor coolant pressure of 1950 psia which is conserva-tP-c as the Q-DNBR decreases with decreasin6 pret sure.
In the event of a loss of flav accident with th 2x2 in the central position, an analy' sic vould be made at the time of the sceident based on the miasured core and floe coastdown characteristics to asceraain if damage could have occurred to the 2x2. If this analysis indicates the possibility of damage, the 2x2 may te examined before resuming operation with the 2x2.
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DPR k Technical Specification Change Request No. 20 Page lo of 12 Pagen TABLE 4 4-ROD SUBASSDGLY v1 ACCIDEIC ANALYSIS PARUCTERS Steady state inlet mass velocity, 1b/ft2 hr 1.157 x 106 2 0 Ma.ximum heat flux, Btu /ft hr 5 776 x 10 Ave n ge heat flux, Btu /ft2 hr 3 36 x 10 5 Equivalent diameter, based on vetted perimeter, ft o.0352 Equivalent dianeter, based on heat tansfer perimeter, ft o.0685 Head loss coefficients, based on flow area in fuel region Bottom end plate 1.13 Spacers, total of 4 3 20 Top end plate o.kB Steady state core pressure drop, psi 2.48 h) Health and Safety It is our opinion that the health and safety of the public vill not b; endangered by this change.
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100- DPR !+ Technical Specification
! Change Request Woo 20 Page 11' of M Puges FIGURE I PRIFARY CODI/dfr FLOW g_ 00ASTDOWN FOLLOWING IDSS OF PUMP POWER 60-I e
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DPR-b Technical Specification Change Request No. 20 -
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