ML20002D873
| ML20002D873 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 12/18/1970 |
| From: | Walke G, Wall H CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P, Skovholt D US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101230244 | |
| Download: ML20002D873 (10) | |
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C0mp2Dy General Off ces: 212 West Michigan Avenue..lackson. Michigan 49201. Area Code S17 788-0550 December 18, 1970 Dr. P. A. Morris, Director Re: Docket 50-155 Division of Reactor Licensing DPR-6 ZEK United States Atomic Energy Commission Proposed Tech Spec Washington, DC 20545 Change 23 s
Dear Dr. Morris:
Attention:
Mr. D. J. Skovc'"
Transmitted herewith are three (3) executed and thirty-seven (37) conformed copies of a request for a change to the Technical Speci-fications of License DPR-6, Docket No 50-155, issued to Consumers Power Company on May 1,1964, for the Big Rock Point Nuclear Plant.
This proposed change (No 23) vill enable Consumers Power s Company to insert into the rea'etor at Big Rock Point an instrumented' fuel rod which will allow cladding temperature in a representative fuel bundle to be monitored. This information vill be useful in the analysis of crud laydown rates and heat transfer phenomena associated therewith.
It is our intention to insert the instrumented rod into the Big Rock Point Reactor during our next refueling outage whicP is cur-rently scheduled for February 1971. We would, therefore, be most appre-ciative of an expeditious handling of this Request for a Technical Specifications Change so that we might receive approval befcre January 15, 1971.
Yours vny truly, D
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f Change No 23 License No DPR-6 For the reasons hereinafter set forth, the-following changes to the Technical Specifications of Licer.se DPR-6 issued to-Consumers-Power-Company on May 1,1964, for the Big Rock Point Nuclear Plant are requested:
I.
Changes:
Section 8 A.
In Section 8.1, change the first paragraph to read as follows:
"The general dimensions and confi uretton of the developmental 6
fuel designs shall be.as shown in Figures 8.1 through 8 5 Principal design features shall be essentially as on Table 8.1."
B.
In Section 8.1, add Figure 8.5, Big Rock Point F Fuel with Instrumented Fuel Rod.
C.
Add a new Section 8.1 3:
"8.1 3 Instrumented Fuel Bundle One reload-F fuel bundle may be modified to include. an instrumented fuel rod. The instrumented fuel rod shall incorporate reload-F design and fabrication features except that provision shall be mace to locate thermo-coup 1M in the fuel rod and to extend thermocouple leads from the rod to a penetration seal in the reactor pres-sure vessel head.
Nuclear and thermal hydraulic characteristics of the instrumented fuel bundle shall be the same as for the reload-F bundles."
D.
In Section 8.2, add the attached colurn and footnotes to Table 8.1.
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. Section 8.2, Table 8.2 (Column and footnotes to be added):
" Reload (F).With Instrumented Fuel Rod General
- Geometry 9x9
- Rod Pitch, Inch-.
0 707
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Standard Fuel Rods per Bundle 69 Special Fuel Rods
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per Bundle 12(b)
Spacers per Bundle.
Fuel Rod Cladding Material Zr-2 Standard Rou Tube Wall, Inch 0.040 Special Rod Tube Wall, Inch 0.040 and 0.060(9)
Fuel Rods Standard. Rod Diameter, Inch 0 5625 Special Rod Diameter, Inch 0 5625 Fuel Stacked Density,
-94 pellet [g) 1 Percent Theoretical Active Fuel Length, Inches - Standard Rod TO 64 9 central
- Special Rod 68.4 Instrumented
- Fill Gas Helium 4
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Same as E-G fuel except that the bundle contains an instrumented fuel rod.
~ (9) Instrumented fuel rod is clad with tubing having 0.060-inch wall. Tubing contains axial grooves on inner surface. The depth of the groovec is 0.020 inch."
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3 II.
Discussion - Instrumented F Fuel Rod-1 A. - Program Description' One reload-F. fuel bundle will-be modified to include an instrumented fuel. rod.
A: summary of program' elements-is shown below:
1.
Modify'a reload-F fuel bundle la. -Remove the upper tie-plate and one fuel rod.
-b.
Install an instrumented fuel rod and a new upper tie-plate which has been modified to permit extension.
of the thermocouple Icad out of the fuel rod.
~2.
Irradiate the instrumented fue'l bundle for one to three
. reactor cycles.
3.
Remove the instrumented fuel rod and conduct a post-irradiation examination.
- 4. -Depending on the bundle reactivity after irradiation of the instrv.ented rod, the bundle will either be stored as spent fuel or returned to.the core as a power producing element.
If the bundle is returned to the core, a conventional type-F fuel rod and upper tie-plate will be installed.
B.
Fuel!and Instrumentation Description i
The design, nuclear, and thermal-hydraulic characteristics of the instrumented F bundle will be essentially the same as standard Reload-F-bundles. However, slight changes will be required to permit placement of thermocouples in the instrumented fuel rod and extension of thermocouple leads from'the fuel rod to a penetration seal in the reactor head.
' Details are shown in Figures 1, 5.13 and the fuel characteristic table.
Differences between the instrumented fuel bundle and Reload-F bundles are summarized below:
1.
A standard _ type-F upper tie-plate will be modified. The modification will consist of the addition of a sleeve which permits extension of
_ thermocouple Icads through the' tie-plate. The modification will not
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interfere with the function of the tie-plate as a structural' element and will not affect nuclear or thermal-hydraulic characteristics of the bundle.
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2.
The upper end-plug of the fuel rod will be changed to permit extension of thermocouples out of the fuel rod. A tandem-extruded Zr-2 to stainless steel transition will be welded to
.the end-plug..Thermocouples will extend through the end-plug and
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transition and be sealed in a stainless steel over-sheath. -The over-sheath will be welded to the Zr-2 to stainless transition.
The over-sheath _ and thermocouples are part of the lead-out assembly which will extend from the instrumented fuel rod to the reactor
-head.
Both the lead-out assembly and' tandem-extended transition incor-
.porate design features-from previous in-core development-activity and current General Electric reactor control instrumentation.
Tandem-extruded-joints between Zircaloy and austenitic stainless steel have been tested and found suitable for use in fuel elements for thermal spectrum reactors.1) The lead-out assembly is.similar to the instrumented fuel assembly used in the Big Rock Point Reactor during the Research and Development phase and incorperates design features of the current General Electric in-core sensors.
3.
Thickness of the fuel cladding will be increased to 0.060 in.
Axial grooves will be cut on the inner surface of the cladding.
Stress analyses of the cladding for the instrumented fuel rod indicate that cladding integrity will not be compromised by the use of thick wall tubing and the presence of axial grooves.
4.
.The diameter of the UO2 Pellets in the instrunented fuel rod is less than the diameter of Reload-F fuel.
Fuel enrichment will be increased so that the power factors for the instrumented rod are nearly egn. to standard F fuel.
5.
Thermocouple leads will extend from the instrumented fuel rod'to a penetration seal in the reactor head.. The leads will run from the central region of the core upwards through an existing opening in the steam baffle. The routing will be to the same as was used during irradiation of the instrumented fuel assembly listed in Section 3.0 of the Technical Specifications.
C.
Nuclear Design Analyses of the instrumented F bundle indicate that the nuclear proper-ties of the instrumented rod and the instrumented bundle will be the same as Reload-F fuel.
1 Busboom, H. J. and M. E. Snyder, " Fabrication and Testing of Dissimilar Metal Joints for Thermal Spectram Superheat Reactors," GEAP-4756, December 1964.
3-D., Thermal Design Thelthermal properties of'the instrumented F bundle are_ essentially identical to the Reload-F fuel..As in the-F-fuel', a 122% overpower
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condition will produce 1fue1 centerline temperatures which exceed.
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the melting point-of UO. The wo'rst condition exists at the'end of 2
Cycle 1 when operation at 122%; overpower, 500,000 BTU /hr-ft2, would produce a ? fuel' centerline teraperature.of 5276*F 2).
However, fuel porosity;and 'the presence of 3% ' dishes in each fuc1 pellet provide sufficient' space to absorb volumetric increases due to transforma-tion of state by the UO2 fuel.
E, - Thermal-Hydraulic Design-
- Thermal-hydraulic characteristics of thclinstrumented'F bundle'are i
the same as Reload T' fuel. The instrumented rod has the same local and axial power factors as the F-rod which it replaces. Consequently.
no changes in'the critical heat flux margin will be encountered.
F.
Acci fent Analysis 1.
Reactivity Excursion Analysis:
The response of:the instrumented-F bundle to all postulated reactivity accidents is.the same as the response of Reload-F fuel. The instrumented fuel rod was designed
?to be loaded-into a Type-F fuel bundle without perturbing.nucicar and thermal-hydraulic characteristics.
The consequences of u re--
activity excursion are no more severe-for the instrumented-F' bundle than for the Reload-F fuel.
- 2.
Primary System Integrity: Analysis of primary system integrity is-needed because thermocouple leads extend from the instrumented fuel rod through the reactor pressure vessel. Requirements which must.be satisfied are that -fission gasses be contained in the instrumented fuel-rod and that coolant be contained in the pressure vessel.
Fission gas and primary coolant containment requirements will be-satisfied in a straightforward manner. As described previously, thermocouple leads will extend out of the fuel rod into an over-sheath. The over-sheath will be sealed against gas or fluid leakage.by braze joints at both ends and at an intermediate position.
The sheath will be attached to the instrumented fuel rod by a fusion weld and to the pressure vessel penetration by a braze joint.
2)
The maximum centerline temperature for the instrumented rod is 5276*F and is greater than the corresponding temperature in the F-fuel (VIZ 5203*F)~ because of differences in cladding thickness and in fuel diameter.
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Fission products will be contained in the fuel' rod by the braze seal at. the lower end of the over-sheath.
The seal will prevent fission gasses from traveling past the upper fuel rod end-plub-
'Redundaat containment is provided by two additional braze seals located midway between the rod and the reactor penetration and at the upper end of_the over-sheath.
1CoolantLwil1~_ue contained in the primary system by the over-sheath and pressure vessel penetration seal. The over-sheath._is protected from physical damage by a second sheath which consists of a flex-ible metallic bellows.. The braze joint at the upper end of the over-
~
sheath:provides a redundant' seal to contain reactor coolant if the
-over-sheath should be damaged.
The reactor penetration and instrument lead-out assembly is_ based on proven designs. The over-sheath assembly incorporates features used on the current General Electric in-core instrumentation. The configuration of the pressure vessel penetration seal and the-method of extending leads out of the reactor from the instrumented rod was taken from the Instrumented Fuel Assembly used in the Big Rock Point Reactor during the Research and Development phase. Primary system integrity will not be compromised by the instrumented fuel rod.
3.
Loss of Coolant Accident: The mechanical, nuclear and thermal-hydraulic characteristics of the instrumented-F fuel bundle w311 be equal'to Reload-F. fuel.
Consequently, the response of the instrumented-F bundle to emergency procedures-during a loss of coolant. accident will be the same as the response of Reload-F fuel.
The safety of F-fuel during a loss of coolant accident was shown previously in data submitted for license change numbers 14 and 16.
G.
Conclusions.
Based-on the preceding analysis and comparison with "F" fuel the following can be concluded:
1.
The mechanical design of the instrumented fuel rod is the same as-
"F" rods except for changes required to install thermocouples.
The changes will not adversely affect the performance of the instrumented fuel rod or the fuel bundle in which the rod is loaded.
2.
The nuclear and thermal-hydraulic characteristics of the instru-mented fuel rod are equal to the characteristics of "F" fuel.
Nuclear and thermal-hydraulic properties of the "F" fuel will not be changed by the addition of the instrumented fuel rod.
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Conclusions-(Cont.)
3.
The-consequences of a reactivity excursion or loss of coolant
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accident are no more severe with the instrumented fuel rod in the core than with standard "F" fuel.
_4.
The integrity'of the primary reactor system will not be com-
,romised. Redundant coolant and fission" gas seals protect the primary system in the event ~of equipment damage.
Based on the preceding considerations, we conclude that the addition of an instrumented fuel' rod in the' Big Rock Point Reactor does not present a signifi-cant change in the hazards consideration described or implied in the Final-Hazards Summary Report.
CONSUMERS POWE3 COMPANY' 8
-By d&<.
Senior Vice President Date:
December 17, 1970 Sworn and subscribed to before me this 17th day of December 1970.
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j Notary Public, Jackson-County, Michigan f
My Co= mission Expires January 15, 1972 1
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s DATE OF DOCUMENT.
DATE RECEIVED NO.:
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Cecsumers Erewer Co.
12-18-70 1-4-71
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Jackson. Mich.
Gerald J.
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U rec NO-REFERRED TO DATE R EC Eiv E D s 't DATE Ltr trams the following req. ehange la Ziemann 1-4-70 Tech Specs to auth insertion of am W/9 avs for action Amstrmeented fuel rod which will al low _._cla ddiar tee.
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