ML20085D382
| ML20085D382 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 10/31/1963 |
| From: | SAXTON NUCLEAR EXPERIMENTAL CORP. |
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| ML20083L048 | List:
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| References | |
| FOIA-91-17 NUDOCS 9110160131 | |
| Download: ML20085D382 (24) | |
Text
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l ADDENDtid NO. 2 TO SAFEGUARDS REPORT FOR PHASE I OF THE SAXTON FIVE-YFAR RESEARCH AND DEVELOWINT PROGRAM
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9110160131 910424 PDR FOIA DEKOK91-17 PDR 7n o 9-
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TABLE OF CONTDTTS_
Page .
1 I. Introduction 3
II. Rod Cluster Control Assembly Program Objectives and Scope 3 A.
h B .- Program Description Subassembly Design 5 C.
Safety Considerations - 11 D.
Conclusions 14 E.
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i III. Irradiation of a Barnable Poison Fuel Rod 15 l
Program Objective and Scope 15-A.
Description of Burnable Poison Rod 15 B.
- 16 C. Safety Considerations D. Conclusions. '18
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I. Introduction 4
The limiting safeguards consideration governing the design and operation of special fuel assemblies in the Saxton core were first analyzed in the Safeguards Report for Phase I of the Saxt.on Research and Development Program, submitted with the application for Amendment No, 10 to the provisional operating license, dated January 6, 1962.
This analysis was augmented by Addendum No. 1 to the above Safeguards Report which was submitted with the application designated "Amendmen*
No.10 - Supplement No. 2" to the provisional operating license, dated December 20, 1962. It is the purpose of the present report to describe further additions to and changes in the test program. The pienent report describes and evaluates the effect on reactor safety of the following experiments:
- 1. The insertion of a special 3 x 3 subassembly containing zirealoy, and thin clad stainless fuel rods, and two rods simulating the design configuration and heat output of the elements of a rod cluster control assembly.
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- 2. The insertion of a rod containing a small quantity of burnable poison in the previously described 2 x 2 subassembly.
The foregoing program introduces the following changes and additions to the previously described program.
- 1. The special 3 x 5 subassembly presently in the central core position (Position I of Figure 1) vill be replaced by the new subassembly containing simulated rod cluster control elements,
- 2. The previously described insertion of a 2 x 2 subassembly had been delayed due to procurement difficulties. It is still planned to insert this assembly in Position IV (see Figure 1).
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4 W ROD CLUSTCH CONTROL SUBASSCGLY 2
.: ROD-OSOILLATOR SUBASS'T i 7
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2 x 2 SUBASS'Y i ~
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FIGURE L-FLAN VIEW OF SAXTON CORE SHOWING LOCATION OF SPECIAL '
SUB ASSEMBLIES.
This subassembly when installed vill contain a rod with small amounts of burnable poison for a portion of its reactor exposure. ,
None of the abor +ests significantly change any of the accident analytes pre ' resented. The detailed de516n considerations are preseni. - subsequent sections of this report, 2
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II, Rod Cluster Control Subassembly ,
A, Program Objectives and Scope At present, Westinghouse is developing a new control rod concept.
In this concept, which has been called rod cluster control, the absorber material for cach mechanism is contained in a number of small tubes which are distributed throu6hout a fuel nssembly. The cluster of tubes is connected by means of a spider to the actuatin6 mechanism.
4 This concept is of considerable interest since it provides a more uniform distribution of poison over the core and hence can lead to lover peaking factors. In addition, the amall water slots mean that followers are not required. This leads to a significant saving due to the reduced reactor vessel length.
The individual absorber rods are guided through the fuel assembly by hollow tubes. The clearance between the guide tube and absorber- rod is small. One could postulate a si6nificant accumulation of crud in this area and hence a marked increase in the withdrawal force requirements. It is thus desirable to conduct an in-pile test which demonstrates that this problem does not arise.
. Extensive mechanical tests are to be conducted out-of-pile. . Thus a simple static test in a reactor is sufficient. The present test will provide a simulation of the proposed absorber operation. Two sbnulated rod cluster control (RCC) rods will be contained within guide tubes and irradiated. The required withdrawal forces will be measured both before and after irradiation.
. The objective of these tests will be accomplished by incorporatin6 the RCC elements into a 9 rod test subassembly. In addition to the RCC elements, two stainless steel clad and five zirealoy clad fuel rods will also be present.
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The objectives of this test may be summarized as follows:
1- To obtain information on the effect of in-pile exposure on the force required to move simulated RCC elements.
- 2. To continue the observation of the effect of in-pile exposure under checical shim conditions on the surfaces of stainless steel clad fuel elements.
3 To extend the observation of the effect of in-pile exposure under chemical chim conditions on the surfaces of zirealoy clad fuel elements and to observe the effects of operating such elements at a peak power of 16 kw/ft.
B. program Description i i l A cross sectional view of the test nubassembly is shown in Figure 2.
The subassembly contains tuo sLmulated RCC elements. In order to simulate the heat 6eneration within the absorber rods without the use of absorber material, the rods are filled with depleted uranium.
When the absorber rods are fully inserted there is a hiSh resistance to coolant flow in the annulus between the absorber and guide tube.
In a large reactor core, a solid guide and fully inserted absorber can result in superheated steam issuing from the annulus. To avoid this condition, the RCC design calls for the guide tubes to be perforated. This allows for mixing between the coolant in the annulus and main stream and hence reduces the exit enthalpy. With the absorber fully withdrawn, the total flow through the perforated guide tube -
would increase and could lead to excess bypass flow. -To prevent this, some designs call for a perforated guide tube in the lower region of 1
the core and a solid tube in the upper section. The Saxton test assembly will provide a test of both sections. One RCC element vill utilize a solid guide tube and the second a perforated one.
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ITEM
- l. RCC TEST ELEMENT WITH PERFORATED GUIDE TUBE
- 2. RCC TEST ELEMENT WITH SOLID GUIDE TUBE
- 3. THIN CLAD STAINLESS STEEL FUEL ROD TUBE .361 IN.:1.D. X .0095 IN. WALL FUEL .3571N. DI A UO2 PELLETS ENRICHED TO 14 W/F1 POWER OUTPUT
- 4. ZlRCALOY - 4 CLAD FUEL RODS TUBE .3435 IN. l.0 X .023B IN. WALL ' -.
PELLETS ENRICHED T016 W/FT POWER OUTPUT FUEL - .337 IN DIA. U02 E. D. . S K. 302099-B--
FIGURE 2 - ROD CLUSTER C013 TROL SUBASSEMBLY -
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' Five zircaloy clad fuel rods are included. These are enriched to
- 7. 3 v/o U235 in rder to produce a maximum heat generation of 16 kv/f t. In addition, two stainless steel clad fuel rods are
! included. These contain fuel which was originally enriched to the normal fuel enrichment of 5.69 These rods vill be obtained fram the subassembly now in the central core position (position I of Figure 1).
It is expected that during November 1963 the special subassembly now in the central core position vill be removed. Two of the removable stainless clad fuel rode vill be removed and placed in the RCC assembly. The RCC subassembly will then be inserted in the vacant core position and the irradiated subassembly shipped to the West 1nghouce Walt: Mill Site for hot cell examination.
C. Subassembly Design
- 1. General Design The RCC (Rod Cluster Control) Test Subassembly is basically a 3 x 3 Removable Fuel Assembly modified to incorporate 2 RCC test elements in place of 2 removable fuel rods. The test assembly l is interchangeable with existins 3 x 3 Removable Fuel Assemblies and vill be supported in the Saxton reactor with a latch assembly of existing design. The 3 x 3 Test Assembly with RCC elements installed is shown on Figure 2.
The RCC test elements consist of a guide thimble which is made an integral part of the fuel assembly and a stmulated absorber rod which is suspended statically in the thimble and supported
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t from the upper end. The method of support of the absorber rods is s'uch that, with the test assembly in the fuel storage rack and the latch assembly removed, the rods are free for removal or withdrawal force measurement.
Fuel assembly modifications required to incorporate the test elements consist oft (a) removal of grid spring fingers at the RCC locations, (b) enlargement of 2 existing holes in the upper end plate for insertion of the absorber rods, and (c) addition of 2 holes in the bottom end plate for mounting of the test element guide tubes.
- 2. RCC Test Element Construction The absorber rods consist of depleted U02 pellets contained within stainless steel cladding. The U0 Pellets were used in 2
place of the poison material which would normally be used in the absorber rods. This is done in order to obtain the req.uired test element heat flux without affecting the Saxton reactor l control. A growth marker is provided between the top of the pellet stack and the upper end plug to pensit determination of the maximum LD temperature when the test elements are removed 2
from the reactor. The marker dimensions are set to provide for a minimum of .400 inch UO to clad differential growth (sufficient 2
for better than 2500 F U0 2temperature ).
The guide tubes are constructed of thin vall stainless steel.
The lower 4 inches of the tube is svaged down to simulate the t.
.015 inch diametral gap in the dashpot area at the bottc. of l
a prototype absorber. The tight gap on the test element will l permit evaluation of crud buildup to be expected in the prototype dashpot.
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The .031 diametral Sap between the absorber rod and guide tube
' over the remainder of the 6 21de tube length was set to obtain
= test elements.
the desired coolant outlet temperature frao The test is desi6ned to simulate the peak heating in the sections of the BCC element which have perforated and solid guide tubes.
Since two separate elements are used to shmulate these sections, pellets, used in place of the different enrichments of the U02 absorber material, are required. The lowest enrichment is The highest contained in the element with the solid guide tube.
enrichment is contained in the guide vhich is perforated to The cross sectional dimensions, 25 per cent metal removal.
materials of construction and other information pcrtinent to the RCC elements are listed in Table I.
3 RCC Test Element Support The test element guide tubes are made a permanent part of the test assembly and are supported vertically by velding of the In guide tube end plugs to the test assembly bottom end plate, order to permit use of as large a guide tube as possible and to provido lateral support for the guide tubes, the guide tube diameter was set to provide a nominal .0005 interference with the grid straps.
In order to permit measurement of forces required for axial l
' movement of the absorber before and af ter installation in
- Saxton, the absorber rods support is vertically independent of the guide tubes. The vertical support is provided by a caged spring assembly mounted on the upper end plug of each absorber.
The springs are compressed between the test assembly upper end l plate and the latch assembly during assembly of the latch to the test assembly. The compression of the springs is set to give a 7
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j TABLE I_
Thermal and Mechanical Des 16n Data for RCC Elements _
Solid Guide Perforated Guide Tube Unit __
_ Tube Unit __
DIMENSIONS (COLD 1
.294 .294 Pellet O. D., in. 4595 .4595 Sheath Tube O. D., in.
.2985 .2985 I. D., in. 5145 5145 Guide Tube O. D. (Upper Section), in. 4995 5075 (Lo.ter Section), in. .4905 _4815 I. D. (Upper Section), in. .h745 .4745 (Lower Section), in.
Diametral Gap, in. .0045 .0045 Pellet to Clad Sheath Tube to Guide Tube .031 .022 (Upper Section) .015
.015 (Lower Section)
- MATERIALS, UO Fuel Material 0d UO 0d Enrichment, w/o U 235 93 93 OxideDensity,%ofTheoretical 304 SST 304 SST Fuel Cladding (Sheath) 304 SST 304 SST j
Guide Tube Material THERMAL DESIGN _
55,500 75,500 Maximum Heat Flux, Btu hr-ft 2 35,800 k8,700 Average Heat Flux, Btu hr ft l
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nominal downward restraining load of 10 lb. on each rod. In the vorst case at elevated temperatures, the minimum calculated end load af ter creep vill still be 8.4 lb., which is better than 5 times the 1,67 lb Weight of a submerged absorber rod.
Upward motion of the absorber rods beyond the spring limit is further restrained by met *? to metal contact between the upper end of the absorber rod and the latch assembly.
Lateral movement of the rod is restrained by an approximate 1.7 D. frictional force at the upper end of the rod and is limitei by the 0.015 diametral gap at the bottom.
- 4. Nuclear and Themal Design The nuclear and thermal design of the RCC elements was based on achieving the same heat production rate as expected in a large reactor using such control elements.
Two heat outputs,195 and 2.65 kv/ft, vere chosen as representative of the maximum heating in the ful1 scale rod covered by the solid and perforated guide tubes.
The vorst dimensional tolerances anticipated during operation were used in the computation of the heating effect.
The effects of gamma heating of the pellet and cladding vere included. The fuel parameters vere established so that the desired kv/ft values were met on a nominal basis (i.e., no uncertainty factors were included).
The estimated uncertainty in the calculated values is +- 20% and was included in detemining maximum temper safety evaluation.
The fuel parameters selected were as follows:
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Fuel for Solid Thimble Elements (1 95 O.35 kv/ft)
Pellet Density 9%
U Enrichment 0.29w/o 235 Fuel for Perforated Thi2nble Elements (2.65 0.45 kv/ft_)
Pellet Density 9$
U g39 Enric kent 031v/o The analytical methods used in the study included two-dimensional calculations in an x-y plane at the axial power peak. These indic%ted that the insertion of the RCC subassembly (3 x 3) does not increase the maximum power in the other fuel rods in the core.
The calculations also indicated that the spiked fuel rods in the RCC 3 x 3 subassembly and the 9 x 9 assembly will not exceed
- 16 kv/ft including maximum tolerances, with the normal *uel in thecoreremainingbelowitslimitof14.1kv/ft.
The performance of the special test assemblies is dependent upon the control rod configuration. For desi6n the chemical shim conf 16uration was chosen since this results in maximum heating.
! If the Saxton reactor were to be returned to rodded operation, l
l the special assemblies vould produce less power. ,
l 3 Zirealoy Rods The five permanent zircaloy clad fuel :c .s have been designed to produce a maximtun of 16 kv/ft with the core operating at
- 23 5 MW. This maximum is consistent with t he conditions snalyzed in the " Safeguards Report for Phase I of the Saxton Nuclear Experimental Corporation Five-Year Research and Development Program" and supplements thereto. These elements tilerefore do not change the previously described worst conditions.
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The zirealoy clad rods will contain 002 pellets which are uniformly enriched to 7 3 v/o U235 he zireal y cladding vill be free standing under the conditions of test. The fuel elements vill conform to the following nominal cold dimensions:
Pellet O. D. o.337 inches Rod O. D. 0.391 inches Cladding Thickness 0.0237 inches
- 6. Stainless Steel Clad Rods The two removable stainless steel rods in the subassembly are obtained from the subassembly now occupying the central core position. These rods were loaded with pellets initially enrichedto5.69w/oU he power output of these rods will 235 notexceed14.1kv/ft. The dimensions and character 1ctics of these are described in Addendum No.1 to the Phase I Safeguards Report as corrected.
D. Safety Considerations Three questions must be considered in assessing the operation of tne proposed RCC test, viz.
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- 1. Is the design of the RCC subascembly such that the fuel elements operating at 16 kv/ft impose more stringent restrictions than previously7
- 2. Do the RCC test elements themselves impose more stringent limitations?.
3 Does the RCC assembly affect the remainder of the core adversely?
e TABLE II Comparison of Perfomance of RCC Subassembly and Full Spiked Assembly .
Spiked 3x3 Spiked with RCC 9x9
- 1. Minimum Local q"-DNB Ratios:
100% Power (23 5 MWt, 2000 psi and 52'BF inlet temperature) 2 36 2.27 120% Power (1800 psi and 52k F inlet temperature) 1.67 1.65
- 2. Hot Channel Factora_:
For Heat Flux, F q 3 34 3.31 For Enthalpy Rise, Fg 2.81 2.72 3 Accident Analycis Parameters:
5 Steady State Inlet Mass Flow lb/hr ft 7 99 x 10 5
Maximum Heat Flux Btu /hr ft 5 38 x 10 Equivalent Diameter (Unit Cell Perimeter) f t. .0381 Equivalent Diameter (Total Wetted Perimeter), ft. .0333 Heat Loss Coefficient Based on Flow Area in Subassembly Ibtton End Plate k.05 Grida (For Fcur Levels) 3 92 Top End Plate 5.83 l
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To answer the first of these questions, the themal and hydraulic conditions in the 3 x 3 subaseembly have been detemined and are shown in Table II. The local q"-DNB ratio for the 16 kv/ft spiked fuei rods was calculated to be slightly hi6 er h than the ratio calculated for the centrally located 16 kv/ft spiked 9 x 9 assembly reported in the previous Hazards Analysis. Thus, the 3 x 3 spiked fuel rods are not worse than the spiked 9 x 9 rods from a DNB consideration. The enthalpy rise hot channel factor Fg for the 3 x 3 subassembly was found to be slightly hi6her than the value reIorted for the 9 x 9 spiked assembly. For the 3 x 3, the Fg was computed by estimating the coolant flov in the subassembly assuming it to be a closed channel and the total power output of the seven rods. The enthalpy rise was then compared to the overall core average for 23 5 mit operation.
This procedure is conservative. Therefore, the overall themal and hydraulic performance of the spiked fuel rods in this proposed 3 x 3 RCC test subassembly is somewhat better than the perfomance of the spiked escembly previously analyzed and reported.
The second question required a study of the behavior of the RCC elements during accident conditions. It is found that the element with the perforated guide tube does not cause any problems since the coolant in the annulus mixes with the coolant in the remainder i of the subassembly.- This, coupled with the lov heat flux, assures that the RCC element does not reach burnout during the conditions of any credible accident.
The solid sheath element, however, is more critical. The annular diametral gap selected, 31,+ 1 mils, was based upon the thermal ,
and hydraulic steady-state condition that the peak coolant temperature is just below saturation at nominal power. Thus local boiling vill exist on the absorber rod. If the most adverse ]imit of the uncertainty range is assumed (i.e., 20% higher than nominal power per unit length of absorber), bulk boiling is obtained in
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the annulus. Under this condition, the maximum calculated steady 1 state surface temperature of the absorber rod is 900*F. The loss-of-flow accident was assumed to occur with the solid sheath RCC element at thir> initial steady state condition.
In the modul assu=ed for analysis of loss-of-flows in this case, reactor serem occurs 15 seconds after the start of coastdown.
Simultaneously, the boiling regine in the RCC annulus is replaced by a stagnant steam condition, with a corresponding film coefficient of50 Btu /ft-hr'F,andthepellet/cladSapconductancedropsto 2
250 Btu /ft-hr*F, These assumptions are conservative, since the thermal conditions corresponding to the stagnant steam condition vould actually develop gradually over a finite time interval. The calculated peak surface temperature for this case is 1000*F. The 0.080-inch vall of the absorber rod cladding is free-standing up
.. to a temperature of 1600'F; therefore, no rupture of the cladding would be anticipated in this accident.
The last of these questions posed can be answered by referring to Section II C-4 The nuclear analysis showed that the presence of I the RCC element did not. increase the Tuinum power in the remainder i
of the core.
l l E. Conclusions
'n the basis of the above it may be concluded that the presence of the RCC test subassembly _does not adversely affect the remainder of the core. Furthermore, no restrictions are *. aced on operation oftheadjacent16kv/ftchannelsduetothepresenceoftheRCC
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elements. Finally, under the conditions of a loss-of-coolant flov accident, there vould be no permanent deformation or rupture of l the elements of the RCC test subassembly. It is believed, there-fore, that the proposed experiment can be conducted with safety.
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t III, Irradiation of a Durnable Poison Fuel Rod A. Program Objective and Scope The objective of this test is to determine the in-pile behavior of a fuel rod containing a small quantity of the burnatie poison, zirconium diboride. This vill be accomplished by replacing one of the fuel rods in the 2 x 2 subassembly (to be located in Position IV as shown in Figure 1) by a fuel rod containing a small amount of ZrB 2' After a few months irradiation, the four rod subassembly will be recoved from the reactor. The burnable poison rod vill be removed from the subassembly and shipped to the Westinghouse Waltz Mill Site for hot cell examination. The vacant space in the subassembly will be filled by a fuel rod without poison and the subassembly returned to the reactor.
B. Description of Burnable Poison Rod The burnable poison fuel rod. vill contain fuel having the same enrichment (8.3%) as the other rods in-the subassembly. In addition, it will be constructed in the - same malner and conferm to the same dimensions as the other rods in the four rod sub-assembly (see Supplement No. 2 to Safeguards Report for Phase I of the Saxton Nuclear Experimental Corporation Five-Year Research and Development Program for detailed description).
I The burnable poison fuel rod vill contain 100 ppm natural boron in the form of circoniu:.1 diboride. Th( zirconium diboride is added to the i'uel in The fora of small particles, coated with tunEsten or niobtum, prior to pressing and sintering the pellets.
The homogeneity of the dispersion after sintering vill be assured by chemical and metallographic analysis of sample pellets.
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C. Safety Considerations The effect of a burnable poison element on neutron flux and power distribution is minimal. The low level of poison vill result initially in a very small power reduction in the rod to which it was added. The effect of the burnable poison rod on the power level of adjacent rods is negligible. Thus, the burnable poison rod will not perturb the neutron flux and power distribution in a way that will cause closer approach to limiting conditions than currently permitted in license.
T No dimensional changes excessive stress or fatigue in the rod cladding are anticipated as a result of irradiation. The quantity of helium which is generated by the fissioning of all-the boron atoms will be only approximately 1% of the total fission gases which are produced when the UO2 element is irradiatedtoaburnupof20,000 MWD /MTU. An ample reservoir has been provided for fission gases including helium. . In other respects, no problems are anticipated since the stresses to which the cladding is subjected vill be identical to those in the other three rods.
It is not anticipated that the distribution of burnable poison in the rod will change to a significant extent during irradia-tion. Since the pellets vill be sintered at 1700'C for over four hours, it is expected that no redistribution vill occur-below this temperature. Above 1700*C some redistribution is l
possible, however, any movement will be slight and in a radial direction to a cooler position. The effect of redistributing-the boron in fuel operating above IVOO'C'on neutron flux and power distribution appears to be insignificant.
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NASA has recently investigated the .
! compatibility of UO various diborides (NASA-TN-D-262)
ZrB2 .
2 and is compatible with UO Their results showed that Experiments at Westinghouse have sho2 up to 7200*C, in the bu of ZrB 2 are mixed with 2 UO , a noticeable reactiwn that when sm i
lover temperature,1600*C. on occurs at a a coating to prevent such a lov temper tTungsten s or niobiu known that tungsten and UO a ure reaction. It is point of 2UO , 2750'C, and niobium and 002 are compatible, e perature of 7200'C. 2 are _ compatible at a tem-niobtur or tungsten coated ZrBThus, no reaction en the is expected b reaction vere to occur, it would only b2 particles 2 If a and the UO .
fuel operating at temperatures e e in a small volume of xceeding 2000*C.
If it vere assumed that all of th ZrB 2 particles were to react and become ce niobium and tungsten coa of the fuel rod in contact with the cl d oncentrated at be anticipated.
i region of high boron concentrationThe n the neutron
! generation rates in that region .
, resulting in lover heat depleted of boron, the power level In the other areas, which are other three rods in the subassembly. s would approach those of the the allowable limits. Such operation is within 1 Since the maximum operating tempe cladding is only 700'F, no reaction brature of the stainless steel stainless steel cladding is expected etween 2.and the ZrB the obtained by General Electric. w according to the data Even if-it were assumed that a 1
J. Belbe, "UO oo 2 - Preparation and Properties" -91US AEc 16 -
A.' N. Holden, E. W. Hoyt, W. V .
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" Radiation Damage to Boron-Cont iCummings, and D. C. Zimmerman' Peactor Material anda ning Control Rods. P theadiation Effects of R Damage" roperties of
- Buttervorths, 1962 I
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October 18, 1963 Docket No. 50-lh6 DPR-h Technical Specifications Change Request #10 (Page 1 of 2 pageu)
- 1. Description of Change In Supplenent No.1 to Technical Specifications, page 3, change Item 6(b) to read:
Uranium oxide (UO 2 ) enriched to 5.7% of U-235 shall be used in the fuel assemblies, except that the test fuel assemblies listed belew having enrichments as described may be inserted in the reactor.
Test Fuel Test Fuel Test Fuel Assembly No. 1 Assembly No.ii Assembly No.111 63-Rod Hollow 9-Rod Sub- 9-Rod Sub-Assembly and assembly or assembly First lb pellets 5.69% 5.69% 5.69%
Next 2 pellets 7.30% 9.19% 7.30%
Next 3 pellets 6.81% 8.57% 6.81%
Next 12 pellets 6.h6% 8.13% 6.h6%
Next 3 pellets 6.81% 8.57% 6.81%
Next 2 pellets 7.30% 9.19% 7 30%
Next lh pellets 5.69% 5.69% 5.69%
NOTE: The 9-rod subassembly in the second column shall not be used at reactor power levels greater than 20 MWt.
Test Fuel Assembly No. iv One 9-rod subassembly shall have four corner rods clad with Zircalcy-h having a nominal thickness of 23.7 mils and shall contain uranium oxide (UO 2 ) enriched to 6.1% U-235. The other five rods shall be clad with=
Type 30h stainless steel having a nominal thickness of 9 5 mils and shall contain uranium oxide (UO2 ) enriched to 5.7% U-235 Test Fuel Assembly No. v One 9-rod subassembly shall have four corner rods clad with Zircaloy-b having a nominal thickness of 23.7 mils and shall contain uranium oxide (UO2 ) enriched to 6.1% U-235 The other five rods shall be clad with Type 30h 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. 1.
Test Fuel Assenbly No. vi One h-rod subassembly shall have rods clad with Type 30h stainless steel having a nominal thickness of 23.5 mils and shall contain uranium exide (UO 2 ) fuel pellets uniformly enriched to 8.3% U-235._ One of these rods may contain up to 100 ppm boron as zirconium diboride.
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Oatober 16, 1963
' D:cVet No, f0-lh6 DPR-b Tt:hniesl Specifir e'.a< 5 Change Request #1C (Page 2 of ? pages)
Teet Fuel Assembly No. tii One 9-rod subassenbly shall have fcur corner rods clad with Zirealcy-h having a nominal thickness of 23.7 mils and shall centain urani* a oxide (UO 2 ) uniformly enriched te 7.3%.
Two of the other rods shall be clad with Type 30h stainless steel having a nominal thickness of ?.5 mils and shall contain ur anium oxide (UO2) uniformly enriched to 5.7% U-235. one other rod shall be clad with Type 30h stainless steel tavdog a nominal thickness of 16.1 nils, shall contain uranium oxide (FC2) having a content of 0.29.% U-235, and shall te concentrically located within a solid e*airl. are steel guide tube. The remaining rod shall be clad with Type 30h stainless steel having a nominal thicknese of 16.1 mils, shall centain urand t exide =
(UO 2 ) having a content of 0.71% U-235 and shall be contentrically located within a perforated stainless steel guide tube.
Uranium oxide at the density being used has a meltir.g point of apprcxi-mately 50000F.
- 2. Safety Considerations i
Safety considerations are studied in Addendun No. 2 to Safeguarde l
Report for Phase 1 of the Saxton Five-Year Research and Development Program.
I In our opinion the proposed change does not present significant hazards i
considerations not described or implicit in the Final Safeguards Report.
I 3 Health ay_d S:r e ty of the Public It is our conclusion that the heslth and safety of tt.2 public vill not be endangered by this change.
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- h. Schedul+-
Approv.1 of this change is requested by November 15, 1963
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