ML20002C456
| ML20002C456 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/21/1972 |
| From: | Sewell R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101100324 | |
| Download: ML20002C456 (11) | |
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IWic Dr. Peter A. Morris, Director 4
Re: Docket 50-155 Division of Reactor Licensing M 2 '1972tr License No DPR-6 United States Atomic Energy
%urony Proposed Technical Commission
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[gj fchu h 8 Specification CuRK Change 28 Washington, DC 20545 h^3
Dear Dr. Morris:
Transmitted herewith are three (3) executed and nineteen (19) conformed copies of a request for change to the Technical Specifications 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 28) will enable Consumers Power Company to use both original design and redesigned fuel channel and support tube assemblies in the Big Rock Point Nuclear Plant reactor vessel.
In addi-tion, this proposed change clarifies the wording gcverning the stack re-lease rate limits of Iodine-131 and particulates with half-lives greater than eight days so that the literal interpretation is consistent with the intent of the Technical Specifications and past interpretation.
It is our intention to install up to forty-two redesigned fuel channel and support tube assemblies in the Big Rock Point Nuclear Plant reactor vessel during our next refueling outage which is currently sched-uled for March 1972. We would, therefore, be most appreciative of an expeditious handling of this Request for a Technical Specifications Change so that we might receive approval before March 15, 1972.
Yours very truly, RBS/ map Ralph B. Sewell Nuclear Licensing Administrator
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CONSUMERS POWER CCEPANY B~M.. s:, Q L.. LI' 2l 'Q Docket No 50-155 Request for Change to the Technical Specifications i
Change No 28 i
License No DPR-6 i
i For the reasons hereinafter set forth, the following changes to the Technical Specifications of License DPR-6 issued to Consumers i
Power Ccepany on May 1,1964, for the Big Rock Point Nuclear Plant, are requested:
I.
Changes A.
Change Section 9.1.5(e) to read:
"(e) Total Weight Supported by Core Support Plate 84 Fuel Bundles @ Approximately 440 Lb/ Bundle 36,%0 Lb A total of 88 support tube and channel j
assemblies consisting of:
(1) Up to 86 support tube and channel assemblies with orifice bucket @
110lb/ assembly,or 9,460 Lb (2) Up to 86 support tube and channel assemblies with modified transition and orifice insert @ 107 lb/ assembly, and 9,202 Lb (3) 2 support tube and channel assemblies with channel plugs @ 110 lb 220 Lb 1 flow distributor assembly 2,500 Lb Total Weight 48,882-49,140 Lb" B.
Change the second and third paragraphs of Section 6.5.4(a) to read:
"The annual average stack release rate for Iodine-131 and particulate matter with half-lives greater than eight days (expressed in units of microcuries per second) shall not exceed the pemissible air concentrations for unrestricted 10 3
-areas, as given in 10 CFR 20, multiplied by 1.2 x 10 cm /
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second. Iodine and particulate sample filters shall be removed and analyzed at least weekly.
" Stack release rates for Iodine-131 and particulate matter with half-lives greater than eight days shall be based on these analyses, which shall be performed not sooner than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> nor later than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after filter removal."
II. Discussion - Fuel Channel and Support Tube Assemblies A.
General During the February-March 1971 refueling outage and several f
preceding refueling outages, it was noted that forces greater than a fuel bundle's immersion weight in some instances were required to remove fuel bundles from their fuel channels. The cause of these excessive forces having to be applied was traced to an oil-canning effect (inward deflection of the sides) in the sides of the fuel channel. This oil canning allowed the sides of the fuel channel to interfere with the lower tie plate of the fuel bundle when the fuel bundle was being inserted or removed frvm a fuel channel. Extraction forces have been monitored for all fuel bundles during the previous two refueling outages. The amount i
of force necessary to remove a fuel bundle increases gradually from one I
refueling outage to the next. As a result of this monitoring program, Consumers Power Company has been able to predict the number of fuel channels and support tube assemblies that require replacement at each I
refueling outage.
The cause of the oil ~ canning has been attributed to a radiation-induced stress relaxation. The stresses are residual stresses in the cor-ners of the fuel channels from fabrication. Fabrication procedures have been changed to help eliminate these residual stresses.
f When it became obvious that all of the original fuel channels and support tube assemblier were going to have to be replaced over the next three or four years, a study was initiated to determine what improve-ments, if any, could be made in the design of the fuel channel and support tube assemblies. The hydrodynamic characteristics were investigated in j.
detail because of the preferential crudding observed in the lower tier on i
the outer row of fuel pins. The Jersey Nuclear Company was engaged to mock up the fuel channel and support tube assembly and a 9 x 9 fuel array.
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In addition, they perfomed experimental work on the hydrodynamic behavior of the inlet region of the simulated Big Rock Point fuel bundle. This study showed coolant velocity fluctuations in the lower tier of the fuel bundle resulting from flow separation at sharp transitions or from the interaction of multiple bounded jets. However, in no instances did these flow effects cause any of the thermal-hydraulic operational limits to be exceeded on r. local fuel rod-local flow condition basis. Even so, as it will be necessary to replace most of the Big Rock Point fuel channel and support tube assemblies over the next three to four years, it was decided to redesign the fuel channel and support tube assembly orifice and tranci-tion zone.
The objective of this redesign was to obtain fully developed flow in the fuel bundle entrance region. This was to be accomplished while maintaining hydraulic compatibility such that both the redesigned and origi-nal fuel channel and support tube assemblies could be utilized in the reactor i
at the same time. In addition, the external configuration of the redesigned q
fuel channel and support tube assembly had to exactly duplicate the external j
configuration of the original assemblies. These objectiver were all met and forty-two of these redesigned fuel channel and support tube assemblies have been placed on order.
(Refer to Figures 1 and 2 attached.)
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B.
Redesign The redesigned support tube and channels were selected for the l
Big Rock Point rentor core on the basis of the results of the hydrodynamic l
tests. The results of these tests on a full-scale mock-up of the recom-mended design indicate the flow structure from the lower tie plate to the first spacer has been improved over the flow structure measured for the ex-isting fuel channel and support tube assemblies. Fl a structure includes assembly flow uniformity and turbulence level. The improved flow performance is the result of the combined flow effects associated with each of the prin-ciple redesign aspects of the support tube and channels:
(1) A transition piece which provides a smooth transition from circular support tube to square channel and increases inlet flow area to the fuel assembly by Wp over the existing design, (2) removal of grapple bar from the immediate proximity of the fuel bundle, and (3) lowering of the oriffce plate to 18 inches below the lower tie plate.
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The results of the hydrodynamic studies indicated that, if the transition free the circular cross section of the support tube to the square cross section of the fuel assembly were made more gradual, the highly chaotic flow structure of the existing support tube and channel design could be significantly reduced. The hydraulic tests performed on the transition piece for the redesigned support tube and channel support this initial finding.
4 The tranc.ition piece M= been decigned to be compatible with the existing support tube, channel and fuel bundle dimensions. The transition piece design is such that the elevation at which the transition from support tube to channels is made is unchanged from the existing design. The transi-tionpiecedoesraisetheelevationofthelowertieplate3/8inchabove the lower tie plate level for the existing design. This raises the upper tie plate surface by the same dir,tance, but it does not cause the upper tie plate to extend beyond the top end of the Zirealoy channel.
The attachment of the Zircaloy channel to the support tube is the same as the existing design. The actual point of attachment is made to the transition piece which is welded to the support tube.
The transition piece is approximately 2 95 inches long and pro-vides a smooth tre.nsition from the circular support tube to the square channel. The maximum wall angle fcr the transition piece is 10 degrees and the minimum wall angle is 2 degrees. The maximum wall angle is in the vicinity of the corners of the channel (the point of attachment of the transition piece and channel) and exists over a very small arc in the transi-tion piece. The majority of the transition piece has wall angles of 2 to 7.
degrees. This should enable the boundary layer to remain attached to the walls of the transition piece over the full transition length and results 4
in an increase in flow of coolant to the peripheral subchannels of the fuel bundle in comparison to the existing design.
The minimum flow area for the transition piece is 28.7 square inches which represents a 40% increase in assembly entrance flow area over the existing support tube and channel design. Eight support pads are built into the contoured transition piece surface, conforming to the geometry of the support legs on the lower tie plate of the existing fued. assemblies.
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5 The transition section walls extend approximately 3/4 inch into the Zircaloy channel and provide a restraint against any inward deflection a.
of Zircaloy channel.
In order to insure the redesigned fual channels and support I
tube assemblies were hydraulically compatible with the existing assemblies, it was necessary to size the orifice plate such that the pressure loss for the redesigned fuel channels and support tubes wits the same as the existing design. The orifice and lower tie plate loss coefficients were experimentally determined for both sizes of orifice buckets used with the 3
existing fuel channel and support tube assembly. These experimental values were used to size the orifices (refer to Figure 2 attached) for the rede-signed assemblies. By using this sizing technique, the differences in pressure losses due to differences in flow conditions at the lower tie i
plate of the fuel bundle are accounted for. Thus, the flows through the l
redesigned assemblies will be the same as the flows through an existing j
assembly and the total core flow will not be altered.
Each orifice plate Ms five equally sized holes. Four holes are located on a radius', A, from ti.e center of the orifice plate. Figure 2 shows orifice plate hole and radius dimensions for M-1 and M-3 orifice designs. The M-1 orifice plate hole diameter is 1.488 inches and the M-3 orifice plate hole diameter is 1.050 inches.
The M-1 and M-3 designs replace existing orifice bucket designs, i
G-1 and C-3, respectively.
1 The final feature of the orifice plate design is the lowering of the orifice plate to 48 inches below the lower tie plate and lowering
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the grapple bar to 18 inches.below the lower tie plate. These design a
j changes were made to eliminate component flow perturbations on the assembly
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flow structure, ie, jets from the orifice plate and flow separation about the grapple bar. Further hydrodynamic testing proved these design goals were realized.
C.
Quality Assurance Program A Quality Assurance Program has been implemented to assure that all activities affecting product quality have been perfomed in accordance withthequalityassurancerequirementssetforthinAppendixBof10CFR50.
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6 The vendor's Quality Assurance Program has been reviewed and found to be in confomance with AEC requirements. Design, manufacturing, and quality assurance requirements were delineated in the purchase docu-ments.
A program of periodic audits has been instituted by Consumers Power Company to insure that the vendor and his subcontractors are in compliance with the applicable Quality Assurance Programs, that Quality j
Control inspection and testing procedures are being properly implemented, and that the requirements designated in the material and product speci-fications are complied with. Audits are to be performed during each phase of the project which includes but is not limited to design, fabrichtion, inspection, and installation.
Consumers Power shall maintain or have access to records which assure conformance of the product to the purchase specifications and pro-vide evidence of activities affecting product quality.
D.
Hazards Considerations No changes in individual fuel channel and support tube assembly flow or total core flow will occur because the pressure loss for the rede-
~ signed assembly is the same as the pressure loss for the existing assembly.
Therefore, there is no change in the hydraulic parameters used in the l
Safety Analyses presented in the Final Hazards Summary Report for the Big Rock Point Plant.
In addition, the effects on reactor shutdown margin of raising the fuel bundles 3/8 inch with regard to their present location have been considered. It was concluded that this change will result in changes in reactor shutdown margin that are so small they cannot be measured.
III. Discussion - Iodine-131 Consumers Power Company has always interpreted the words halo-gens and particulate matter in the second and third paragraphs of Section 6.5.h(a) of the Technical Specifications as Iodine-131 and particulate j
matter with half-lives greater than eight days. The third paragraph of i
Section 6.5.4(a) states that the halogen and particulate sample filters will be analyzed between 48 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after filter removal. This im-l plies, and was intended to mean, that the halogen and particulate material i
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governing the allowable release rate limits are Iodine-131 and particulrtos with half-lives greater than eight days. This concept is specifically stated in the Technical Specifications of other nuclear plants.
This change in wording, if approved, will allow Consumers Power Company to be in ccanpliance with the precise wording as well as the in-tent of the Technical Specifications.
IV.
Conclusions Based on the considerations presented in Section II, we have concluded that the use of the redesigned fuel and support tube nssemblies
~ does not present a change in the hazards considerations described or implicit in the Final Hazards Summary Report. The change presented in Section III merely modifies wording to be comple'ely consistent with interpretations and practices.
CONSUbERS POWER COMP 1Y By_
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H. R. Wall Senior Vice President 3
Date: January 21, 1972 Sworn and subscribed to before me this 21st day of January 1972.
A CL(NLc.d Notary Public, Jackson County, Michigan My commission expires December 8, 1975 i
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CrEERs OKIG.: CCs SQ signed & 19 conf'd TOs Dr. Peter A. Morris "^75 ^" 5* 5"5 ' acuan uccess^"' O c "c"""E"c5 0 no action wretssany O cowurnt O sr> PO;r CfrICE TILE CQCE: CLAfstr,s 50-155 (INPUT) 5 nc. no, stronato to oxrr sectiven nr 04: tzscnirtioni(n. g.t g g : gi ann a-z69Z Ltr trans g$ w/9 cys for ACT10E 4(STRIBUTION: 8 ygg, C7 2nctosuats R2 quest for Change to Tech Space-Change ARC FOR No.18 for Lic. 3FR-6 motarised 1-21-72 W1 e (2) would allow CPC to see both original desi gn E -Re-F-506-A G redesigned feel charmel & swport tube asuntaing & staff 's casomblies in the'51g Rock Flaat w/st'ew Ami. alt Fig. 1 & 2........ 5 5,' p, j,r (2 Orig & 19 conf'd cys of emel rec'd) h[ u n *.d'U!I [* N' r.ijnt
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