ML19340A705
| ML19340A705 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/05/1962 |
| From: | Wade I COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| NUDOCS 8009030711 | |
| Download: ML19340A705 (49) | |
Text
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Robert Lowenstein, Director Civision of Licensing and Regulation U. S. Atomic Energy Commission Washington 25, D. C.
C ea r '.; r. Lowenstein:
Pursuant to Paragraph 3a(4) of ~ icense CPR-2, as amended ("CPR-Z"),
Commonwealth Zdison Company requests that Appendix "A" of-CPR-2 he amended as 411ows:
Amendment No.1.
Amend section "A.
INTRCOUCTICN" of Appendix "A" to CPR-2 to read in its entirety:
" A.
INTIOCUCTION "The following a re the principal design and periorance spec-ifications and operating limits and procedures of the Cresden Nuclear Power Station pertaining to safety.
" Sections 3 and C set forth the design and performance spec-ifications and operating limits and principles.
" Sections 2 and E specify the limitations to be observed during start-up, power ope ration, and refueling and maintenance oper-ations. In these sections.
.s well as in Section 3, where mumum or minimum limits are not given specifically, the values 2 ven are i
" design" values which are subject to normal mannfacturing and other tolerances.
" Sections F and C provide certain additional operating and testing procedures applicable to the control rod drive mechanisms.
Section I provides minimum requirements for certain inspections of the control rod drives, poison blader, and core grid structure."
l Amendment No. 2.
Amend item "2.
Nuclear Core" of section "B.
CESIGN FEATURES" of Appendix "A" to CPR-2 to read in its entirety:
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M r. Robert Lowenstein January 5,1962 "2.
Nuclear Core
" Maximum core diameter (circumscribed circle) 129 in.
Maximum active fuel length - cold 112 in.
Maximum number of fuel assemblies by types:
Type I 488 Type II 108 Type PF-1 through PF-12 (one each) 12 Maximum total number of fuel assemblies 488 "The various fuel assemblies may be located in any position of the reactor, provided over-all core symmetry is preserved and provided that fuel assemblies, Type PF-1 through PF-12 are each separated from any other such assembly by at least four Type I or Type II fuel assemblies.
"The reactor may be operated at any power up to and including rated power with any number of the va rious types of fuel as-semblies installed provided the maximum number and location are within the limits specified above."
Amendment No. 3.
Amend item "3.
Fuel" of section "B.
DESIGN FEATURES" of Appendix "A" to DPR-2 to read
]
in its entirety:
"3.
Fuel I
"Each fuel assembly consists of vertically positioned, rod-type fuel elements. The physical properties of each assembly are given in Table II. The number of fuel rods are given for a l
regular assembly. In several assemblies, fuel rods have been replaced with instrumentation tubes.
"The minimum fuel pellet density averaged over a fuel segment is 94% of theoretical for all fuel assemblies, except PF-7, PF-8, and PF-9 which is 90% of theoretical. "
Amendment No. 4.
Amend Appendix "A" of DPR-2 by deleting section
" D.
INITIAL LOADING AND CRITICAL TESTING" in its entirety and by redesignating the captions of subsequent sections as follows:
Change "E.
POWER OPERATION" to "D.
POWER OPERATION" Change " F.
REFUELING AND MAINTENANCE" to
" E.
REFUELING AND MAINTENANCE" Change "G.
TESTING AND INSPECTION OF CONTROL ROD DRIVES" to "F.
TESTING AND INSPECTION OF CONTROL ROD DRIVES"
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m M r.. Robe rt Lowenstein January 5, 1962 k
. Change "H.
OPERATING PROCEDURES" to "G.
OPERATING PROCEDURES" 7
Change "I.
PROJECTED INSPECTIONS OF CONTROL ROD DRIVE MECHANISMS, POISON BLADES, AND CORE GRID STRUCTURE" to "H.
PROJECTED INSPECTIONS OF CONTROL ROD DRIVE i -
-MECHANISMS, POISON BLADES, AND CORE GRID STRUCTURE" Amendment No. 5.
Amend item "1.
Power Test Prog ram" of section
" E.
POWER OPERATION" (to be changed to."D.
PO'VER OPERATION") of Appendix "A" of DPR-2 to read in its entirety:
" 1.
Approach to Rated Power "After any shutdown the approach te rated power shall be accom-p11shed in a g radual stepwise fashion and reactivity, power distri-4 bution, and stability shall be ca refully observed at all times."
Amendment No. 6.
Amend item "3.
Dete rmination of Maximum Reactor
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Powe r" of section "E.
POWER OPERATION" (to be I
changed to "D.
POWER OPERATION") to read in its enti rety:
i "3.
Determination of Maximum Reactor Power "The maximum reactor power is defined as that thermal power l
at which the rnaximum heat flux for any fuel rod is reached.
l This maximum heat flux, based on calculations and experimental l
data, will never exceed the following values in Btu /(hr)(sq ft):
Fuel Type I 350,000 Fuel Type II 445,000 Fuel Type PF-1 through PF-4 425,000 Fuel Type PF-5 through PF-9 415,000 Fuel Type PF-10, 11, and 12 475,000 "The peak rated heat flux and resulting rated reactor power are then set to 80% of their maximum values and, as indicated in item B.9, the high neutron flux sc ram setting will be no highe r than an indicated 120% of the rated reactor puwer. However, in no case will the high neutron flux setting he allowed to exceed an indicated reactor thermal power of 782 Mw (125% of the planned operational. power of the fully loaded core).
"The reactor will be operated within the above limits such that l
a burnout margin of at least 2.0 will be maintained in each type of fuel closest to burnout in the hottest channel in the core based
.on a uniform steam quality over the cross section of the channel."
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m Mr.~ Robert Lowenstein January 5, 1962-3 f
Amendment No. 7.
Amend paragraph b. of item "5.
Reactivity Limits" of Sectior.
"E.
POWER OPERATION" (to be changed 4
to "D.
POWER OPERATION") of Appendix "A" of DPR-2 to read in its entirety:
"b.
With the reactor in any condition, the following shutdown cri-terion shall be met:
" Stuck Rod" c riterion:
At every stage during load-l ing and in the fully loaded configuration, the control i
rods must provide a shutdown control'ma rgin of at
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least 0.01 ok with any rod wholly out of the core and completely unavailable.
"During core alterations after the first fuel cell is loaded,' the following ' cocked rod' c rite rion shall be met:
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" Cocked Rod Criterion:
The reactor must be sub-critical by at least 0.01 ok with at least one control rod fully withdrawn in the region of the alteration and available for rapid scram insertion."
Amendr. tent No. 8.
Amend paragraph d. of item "5.
Reactivity Limits" of section "E.
POWEgOPERATION"'(to be changed to d
i
" D.
POWER OPERA
N") of Appendix "A" of DPR-2 to read in its entirety:
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" d.
The void coefficient averaged over the core will always be negative. "
Amendment No. 9.
Amend item "3.
Shutdown Margin" of section "F.
RE l
FUELING AND MAINTENANCE" (to be changed to "E.
j REFUELING AND L. AINTENANCE") of Appendix."A" i.
to DPR-2 to read in its entirety:
"3.
Shutdown Margin "At every stage of refueling or maintenance, the minimum shut-down margin will satisfy the ' stuck rod' criteria discussed in items D. 5. b. and.D. 5. c.
4 "During movement of fuel in the core, or con':rol rod maintenance, the minimum shutdown margin will satisfy the ' cocked rod' criterion given in item D.5.b."
Amendm6nt No.10.
Amend section "F.
REFUELING AND MAINTENANCE" (to be changed to "E.
REFUELING AND MAINTENANCE").
of Appendix "A" to DPR-2 by adding the following item:
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"5.
Minimum Criticil Testing i
A critical core may be constructed for testing purposes, using any combination of the fuel which is available at the site, subject to the i
following restrictions:
"(a)
The minimum critical shall be located within the control rod patte rn.
"(b)
A minimum of three neutron sensitive instruments which are connected to the safety circuit shall be located inside the vessel in the vicinity of the small core. The circuitry shall be arranged so that any one of the three can actuate a scram.
Two of-these sense rs will sc ram on high neutron level,. and one will sc ram on sho rt pe riod. At least one low level neutron monitoring instrument shall be operating and i scated in the vicinity of the small core.
"( c )
The shutdown crite rion specified in item D.5.b will be met.
"(d)
Except for the final fuel increment, the size of each fuel increment will never exceed one half the estimated critical increment or one assembly, whichever is larger. This estimate will be based on neutron multiplication measurements made be-tween fuel additions. The final fuel increment will not exceed-one fuel assembly."
i Amendment No.11.
Amend TABLE II attached to Appendix "A " substituting therefor TABLE II (revised 12/31/61) attached he reto.
In accordance with paragraph 3.a.(4) of DPR-2, a Description and Hazard Evaluation Report of the proposed amendments to Appendix "A" is attached -
hereto as " EXHIBIT 1."
Your attention is directed to the application dated January 27, 1961, for
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amendment of DPR-2, (hereinafte r referred to as the."Janaary 27',Appli-l cation"),-in which changes in Appendix "A" were requested to permit the loading of 100 Type II fuel assemblies and 12 experim ental as s emblies identified as "PF-1" to "PF-12," inclusive (hereinafter referred to as.the
" experimental assemblies"). Subsequently, by letter dated February 17, 1961, it was requested that consideration of the January 27 Application be 4
l limited initially to amendments necessary to authorize the use of only two l
Type II fuel assemblies and 12 experimental fuel assemblies for power generation and' minimum critical tests with only Type II fuel assemblies.
Accordingly, the balance of the January 27 Application relating to the load-ing of 100 Type II fuel assemblies. changing.the burnout ratio from 2:1 to i
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M r. Robert Lowenstein January 5, 1962 1.5:1 and modifications of the temperature coefficient limitations was not considered at the subsequent hearings, nor authorized by the Fourth Sup-plemental Intermediate Decision issued May 27, 1961, in Docket 50- 10.
However, that decision and the amended Appendix "A" issued June 9,1961, pursuant thereto did authorize the loading of two Type II and 12 experimental as s emblie s.
This application and the attached Description axiHazards Evaluation Report is submitted for consideration in lieu of the balance of the January 27 Application not previously acted upon or considered. The most significant differences between this application and the January 27 Application (other than a change in the number of Type II fuel assemblies from 100 to 108) are:
(i) no re-quest is being made at this time for a change in the burnout ratio and (ii) no authority is sought for modification of the temperature coefficient limitations.
Because of these changes and the difficulties encountered in attempting to isolate those portions of the January 27 Application relating to loading of two Type II and 12 experimental fuel assemblies and the minimum critical testing, and in order to avoid confusion which might result in references to the January 27 Application, the attached Description and Hazards Evaluation Report has incorpor - u the analyses relating to such matters h the January 27th application.
Ingly, the January 27th application is ira-
-d in its entirety by this _eplication for amendment and the attached bescrip naand Haza rds Evaluation Report.
Ve ry truly yours, COMMONWEALTH EDISON COMPANY
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I. L. Wade Administrative Engineer Attachments Swcrn to before me this 51'4 day of @ w a.,
, 1962.
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Notary Public in and for the/ ounty of C
Santa Clara, California
.EVERETT H. LAYNE My Ccmmission Expires 08.00,1964
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( Revised 12/31/61)
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I Z,2 0.567 0.030 66 36 100 UO2 1.5 0498
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11 30455 0 441 0.019 7a7 40 100 U02 25 0 399 9
2.1 UO; 93 5 0.399 97.9 Th02 99.65 00
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PFI 30455 0.43 0 0.011 6=6 29
'o'$5 k,fd 0 457 0.35 Er20 3 g
99.65002 99.65 UO2 1'9 0.J53 M
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PF2 30455 0 430 0 011 6a6 29 0.35 E'20 3 0.35 Er203
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- 2. 3 U02 PF3 30455 0 433 0.011 6m6 29 100 UO2 26 0.453
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I 917 TH02 24UO PF4 30455 0 430 0 011 6=6 I
100 U02 26 C 451 7
93.5 0.J 57 79 916 TH 2TUDy PF5.6 30455 0 439 0 017 6=6 29 100 002 26 0 455 7
93.5 0 455 916 TH0y PF1, 8. 9 30455 0439 0 011 6a6 29 100 002 26 o 455 7
100 UO2 I7 C 455 99.65 002 99.65 UO2 1.5 0.35b PF IO 2,4 0 412 0.025 8a8 55 20 0 358 9
0 35 Eny0 3 0.35 E'2C3 18 UO PFil,12 Ze4 0 412 0 025 8=8 55 100 902 20 0.353 9
93.5 0.353 98.2 TH02 4
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R Exhibit 1 DRESDEN NUC LEAR POWE R STATION DESCRIPTION AND HAZARDS EVALUATION REPORT 4
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CONTENTS z
Section I Description of Proposed Amendments to Appendix A to DPR-2 Section II Physical Characteristics and Mechanical l
Design of Fuel i
Section III Thermal and Hydraulic CharacteristicsSection IV Nuclear Characteristics of the Fuel and Core Section V Safety Evaluation s
This report provides technical information in support of the attached application to an amendment to License DPR-2, as amended. It is not intended that the material contained here-in constitute " Technical Specifications" in the sense of the the licensing regulations (10CFR, Part 50, Section 50.36).
Dated January 5,1962 i
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SECTION I DESCRIPTION OF PROPOSED AMENDMENTS TO APPENDIX "A" TO DPR-2 Proposed Amendments 2, 3, 6. and 11 These amendments are proposed primarily for the purpose of obtaining authority to load up to 108 Type II fuel as semblies during the next refueling of the Dresden reactor, currently scheduled for the summer of 1962. The physical, thermo-hydraulic, and nuclear properties of the fuel and the safety implications involved in the use of 108 Type II fuel assemblies with Type I and experimental assemblies are coveredin Sections II through V of this Report.
It is also proposed by these amendments to simplify Appendix "A" by the elimina-tion of core classificaticns, such as Core I and Core I-Modified. Such classifi-cations serve no useful purpose and, if continued with the addition of each new type of fuel, will result in an unnecessarily complicated and confusing list of technical specifications. No attempt has been made to revise any concepts or specifications now contained in Appendix "A"; on the contrary, changes in the wording have been kept to a minimum consistent with the objective of simplifi-cation. Thus, for example, Table II has been enlarged to describe the physical properties of all types of fuel and item "3.
Fuel" of section "B.
DESIGN FEATURES" has been amended accordingly.
Proposed Amendments 1. 4, 7. 8. 9. and 10 It will be recognised that License DPR-2, including Appendix "A",
was issued first on November 12, 1959, prior to completion of the testing program required to dem-onstrate the safety of the Dresden reactor. Accordingly, Appendix "A" appro-priatelf prescribed in section "D.
INITIAL LOADING AND CRITICAL TESTING "
and item "1.
Powe r Test Program" of section "E.
POWER OPERATION" specific procedures for the initia11oading of the Dresden reactor, the performance of c ritical and other atmospheric tests and for the initial stepwise approach to full powe r ope ration. While such procedures were appropriate during the stages leading to the first full power operation of the Dresden reactor, it was never intended that such procedures would apply in their entirety to subsequent operations after the data and experience from such tests has been obtained and analyzed.
Therefore, for the purpose of simplifying the technical specifications by elimina-tion of portions which are no longer applicable, and thus avoiding potential sources of confusion, it is proposed by Amendment 4 that section "D.
INITIAL LOADING AND CRITICAL TESTING" be deleted in its entirety and the captions of succeeding sections be redesignated accordingly.
Since the " stuck-rod" and " cocked-rod" criterion stated in section "D" are appli-cable to power operation and refueling and maintenance procedures and are referred to in other parts of Appendix "A", it is proposed by Amendment 7 that these speci-fications be inserted in item "5.
Reactivity Limits" of section "E.
POWER i
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3 OPERATION" (to be redesignated "D.
K)WER OPERATION"). Similarly, since the specification that the void coefficient be negative continues to be important, proposed Amendment 8 incorpo rates such provisions in paragraph d. of item "5.
Reactivity I imits" of section "E.
POWER OPERATION" (to be designated "D.
POWER OPERATION").
Amendments 1 and 9 are proposed merely for the purpose of correcting refer-ences affected by Amendments 4 and 7.
Thus Amendment 1 reflects the deletion of s ection "D.
INITIAL LOADING AND CRITICAL TESTING" and the redesig-
, nation of succeeding sections, and Amendment 9 provides the correct references to the " stuck-rod" c rite rion and the " cocked-rod" c rite rion.
i By Amendment 10 it is proposed that the technical specifications for refueling and maintenance operations be enlarged to include procedures for constructing a mini-mum critical core which are substantially similar to the procedures for minimum critical testing in section "D" which will be deleted.
Since proposed Amendments 1, 4, 7, 8, 9, and 10 are not intended to, and in fact do not, effect any change in design or operating procedures respecting future operations of the Dresden reactor, their adoption will not involve any hazards g reate r than, or different from, those analyzed in the Hazards Summary Report.
The refo re, no further evaluation of such changes is submitted herewith.
l Proposed Amendment 5 As p reviously noted, item "1.
Power Testi ng Prog ram" of section "E.
POWER
,i OPE RATION" (to be redesignated as "D.
POWER OPERATION") specifies pro-J cedures that were followed in the initial stepwise approach to first full power operation and such procedures are not applicable to normal power operation at Dresden. Accordingly, it is proposed by Amendment 5 that item "1.
Power Test-ing Program" be deleted in its entirety and in lieu thereof Amendment 5 specifies procedures which would be applicable to a resumption of rated power operations after any shutdown. Such procedures are substantially similar to those now con-tained in pa rag raph b. of item "1.
Powe r Testing Program".
Because the provisions proposed to be deleted are no longer applicable and the pro-cedures proposed to be followed with respect to approach to rated power are sub-stantially a restatement of the procedures now applicable to Dresden Core I-Modi-fied which will be deleted, it is not considered that proposed Amendment 5 will involve any hazards greater than, or different from, those analyzed in the Hazards Summary Report, and no further evaluation of such changes is submitted herewith.
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i f SECTION II PHYSICAL CHARACTERISTICS AND MECHANICAL l
DESIGN OF FUEL 4
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Ten slightly different fuel assemblies have been designed for the Dresden reactor.
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- The differences are tabulated in Table II as part of the proposed amendments to l,
License DPR-2. All variations are associated with fuel rod design, design of the supporting hardware, and changes in material orfabrication techniques. (As of i
December,1961, all of these fuel types have been operated in the Dresden reactor and the description ' f the PF and Type:II fuel' pre'sented in..this section is intended o
to summaria:e and replace pre. iously submitted information. )
j The basic fuel for most of the fuel elements is a sintered, solid cylindrical pellet of about 94% theoretical density. The diameter is slightly less than 1/2 inch, and the length is slightly less than 3/4 inch. The exact pellet diameter varies with the fuel design and is given in Table II.
The pellets a re enclosed by a stainless-i steel or zircaloy jacket which forms a fuel element about nine feet long.
Three of the experimental fuel elements, PF-7, P F-8, and PF-9, have been de-signed around a new fabrication technique in which a homogenous matrix of UO2 powder is worked externally within the cladding to produce a continuous fuel rod i
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of about 90% theoretical density.
4 Each fuel rod is to be placed in a square array of six rows of six fuel rods, seven rows of seven fuel rods, or eight rows of eight rods. The desired geometry is maintained by means of a square channel about 4-1/4 inches inside dimensions with spacers at several positions along the' length which also serve to minimize denection and vibration.
i The fuel types may be divided into three categories as follows:
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Type I The first Dresden core fuel consisting of cylindrical zircaloy-clad fuel rods in a "6x6" mat rix. Details of the fuel assem-bly are shown in Figure 1 and Reference 1*.
l Type II A fuel of a,new and improved design con-stituting the reload of about 108 elements.
i Details are shown in Figure 2.
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Type PF-1 through PF-12 Fuel using advanced fabrication techniques.
A total of 12 experimental fuel assemblies are planned. Details are shown in Figures 3, 4, and 5.
I A.
Type II Each Type II fuel assembly consists of seven rows of seven nonsegmented fuel rods except, as stated below, the center rod is segmented as required for positioning spacers. The rods, clad with 304 stainless steel, are held in position at the top and bottom by stainless-steel tie plates. Between the plates, rod position is maintained by four wire-type spacers, spaced equally along the length of the fuel assambly.
The four spacers are held in longitudinal position by the segmented center fuel rod. Nine rods of each fuel assembly are loaded with thoria pellets.
The water-to-fuel ratio is 2. 67:1.
.'htails of the design a re given in Table II and Figure 2.
B.
I;:perimental Fuel Assemblies PF-1 through PF-12 i
These fuel assemblies incorporate fuel designs resulting from recent engineering developments in rod-type fuel. Figures 3, 4, and 5 show the essential design featu res of the new fuel.
The same general design for fuel Type I applies to the 12 experimental assemblies with the following variations:
a.
A removable fuel rod design is used.
b.
All fuel rods a re suspended f rom the top tie plate.
c.
" Corner" fuel rods in the fuel assemblies will contain either a Th02
- UO2 - Er2O3 mixture, or variable enriched UO2 mixture to reduce power peaking in the corners of the fuel assembly adjacent to the con-trol rod blade corner.
d.
Erbium-oxide (Er20 ) is incorporated in fuel assemblies PF-1, PF-2, 3
and PF-10 as a burnable poison which allows additional burnup without increasing the reactor control requirements.
Other fuel rod manufacturing processes, such as swaged powder e.
(fused), swa ged-over ground pellets, and swaged-over unground pellets, are incorporated in nine assemblies, PF-1 through PF-9, using stainles s-steel cladding.-
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PF-7 PR-8, and PF-9 stainless steel clad swaged fused powder The design utilizes a low cost fabrication process. The swaged powder process consists of filling a tube with " fused" UO2 Powder, plugging the ends of the tube, swaging the tube down to the required diameter, remov -
ing the end plugs, welding pennanent end plugs in place, and then per-forming the necessary quality control to insure that a sound fuel element has been manufactured. A UO2 density of 90% is obtained in this process.
The water-to-fuel ratio is 2. 9:1.
PF-10 zircaloy-4 clad with lower enriched'UO2 corner rods The "8 x 8" assembly is a nonsegmented continuous rod design, using conventionally loaded UO2 Pellets, as shown in Figure 4, with Erbium Oxide and high enrichment UO2 co rne r rods. The water-to-fuel ratio is 2. 4:1.
PF-11 zircaloy-4 clad with thoria corner rods The design is similar to PF-10 except burnable poisons are not used, and thoria corner rods are used to replace the highly enriched UO2 corner rods. The wate r-to-fuel ratio is 2. 4:1.
i PF-12 sircaloy-4 clad with thorta corner rods The mechanical design is shown in Figure 5.
The assembly is similar to PF-10 except segmented rods have been substituted for "through" rods, transverse rod spacing is accomplished by a single-plane wire spacer rather than a double-plane wtre spacer the spacer is positioned by a single-segmented rod, and springs have been added between the lower tie plate and fuel rods. Further, burnable poisons are not used, and thoria corne.r rods replace the highly enriched UO2 cover rods.
The typical core configurations, shown in Figures 6 and 7, represent two possible fuel loading schemes.
The fuel assemblies of Type PF-1 through PF-12 will be located at several positions in the core, spaced so that the performance of any such assembly will not significantly affect nuclear or thermal performance of the entire core.
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SECTION III THERM AL AND HYDRAULIC CHARACTERISTICS t
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i l l The thermal and hydraulic characteristics of the scattered and central loaded l
core,.shdwn in Figures 6 and 7, have been investigated. These investigations included varying primary and secondary steam flows at ratedand 125%of ratedpower conditions. The power distributions investigated are shown in Figure 8 for the central loaded Type II fuel and in Figure 9 for a scattered loaded core.
The means of achieving these power distributions are discussed in Section IV of this report.
l The Type II has greater thermal capability t.han the Type I fuel. This is ac-complished by (1) inc reasing the number of fuel rods per assembly f rom 36 to 49, (2) increasing the active heat transfer area per assembly, (3) re-t 4
duction of corner rod peaking factor by use of U-235 enriched thoria rods, and (4) inc reasing flow through the assembly by streamlining.
The central and scattered loading arrangements are considered for purposes of this analysis because they present the extreme thermal characteristics resulting f rom the use of,both Type I and Type II fuel in the same core as well a s the extremes of pos sible loading a rrangements. Thus, the extreme thermal cha racteristics for Type II fuel may occur with the central loading shown in Figure 6 where the maxirnum channel peaking Osctor is 1. 58 vs.1. 49 with the scattered loading shown in Figure 7.
Conversely, th e extreme for Type I j
fuel may occur with the scattered loading where the maximum channel peaking l
facto r is 1. 41 vs.1.10 with the cente r.cading. It is anticipated
- hat the actual
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loading to be used will resemble Figure 7 and the assembly peaking factors of both Type I and II fuel will fall between the extremes shown in Figures 8 and 9.
a Results of the the rmal and hydraulic analysis are shown in Tables 1 through 7.
Tables 1 and 2 consider rated power and 125% of rated power with maximum secondary steam flow, re spe ctively. Tables 3 and 4 consider the same power conditions, respectively, with maximum primary steam flow.
The maximum primary steam flow represents the worst condition from the stancipoint of bu rnout. The irtereased primary of course results m a greater void fraction in the reactor, as well as higher exit void fraction. The effec
- on minimum burnout ratio is very small, howeve r, since for t9 uit qualities and mass flows involved, th e burnout heat flux is not quality set.sttive.
The ef fec6 of the PF assemblies on the core is shown in Tables 5 and 6.
In 1
j these two tables, the 12 PF assemblies have been substituted for 12 Type I l [
a s s emblie s, anti separate peaking fa. tors are used for each PF assembly.
Table 5 considers ratedpower - 626 wiw with maximum primary steam flow.
Table 6 considers 125% of rated power with maximum primary steam flow.
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8 The peaking Actors for the PF assemblies, relative to the Type I fuel, are given in Table 11. These peaking factors apply to both central and scattered loading. The analysis assumes that the assemblies are located where the Type I fuel has the maximum (1. 41) relative power. So the maximum assem-bly peaking factors for the PF assemblies in Tables 5 and 6 are the produc.t of 1.41 and the appropriate factor shown in Table 11. The maximum heat flux peaking factor includes the axial peaking factor, corner rod peaking f acto r, and an area correction factor. The latter is included to account for the fact that the different assenablies have diffe rent areas, and it is convenient to include-it as a peaking factor so that one can obtain maximum values from the average values only by use of the appropriate peaking factors.
The PF as semblies would not be worked as ha rd in the central loading since the maximum Type I peaking factor is.l.10 in the central loading compa red tc 1.41 in the scatte red loading. Hence, the PF analysis is made for the
=catte red loading.
The PF assemblies produce a very slight change in the peak heat flux in the Type I and Type II fuels. Othe rwise, the PF as semblies have no disce rnible effect on the operation of the reactor.
The results of the analysis of the central loading is shown in Table 7.
The
- analysis is made at 125% of rated power for maximum primary steam flow.
In this analysis, it is noted that the Type :I fuel has a lower burnout ratio inan the Type I fuel but that the burnout rano for *he Type II fuel is highe r than minimum burnout ratio in the scattered loading for any type ^ fuel.
Table 8 presents a comparison of peak analytical results with license limits of peak heat flux and minimum burnout ratio for each type of fuel. It is con-cluded that all fuel can be operated within the limits specified in the license.
,g D**D g ]1 k Qi o Ju.
lfL oo i
m 9
. SECTION IV NUCLEAR CHARACTERISTICS OF THE FUEL AND CORE
- The detailed nuclear characteristics of the various fuel assemblies have been determined. This data was used to study the two types of core fue1LloadihgT which are of interest for this document. These arrangements are shown 'in Figures 6 and 7.
The results of these studies are given in the following paragraphs.
A.
Fuel Assembly Characteristics The basic physics lattice data for the Type I, Type II and PF-1 through PF-12 fuel assemblies are summarized in Tables 9 and 10. The data was developed by computing the nuclear characteristics of each type of i
fuel rod separately. Thermal group nuclear constants are averaged over the thermal neutron spectra calculated by the Wilkens equation.
Details of the thermal neutron behavior in and around individual rods are computed using the P-3 spherical harmonics approximation to the i
j neutron :ransport equation. Indivhiual fuel rods and surrounding. nod-a erator are then homogenized for the cell calculations. Epithe rmal group transport proper;ies are obtained from spectrum-weighted, multi-group cross section files:using the MUFT IV calculational model. Res-onance absorption and' fission parameters are derived from measured resonance integrals with 1/v or other appropriate components added to accommodate nuclear events in the energy range immediately above ther-mal. Fission in the higher energy group is treated by a fast fission factor multiplier of the diffusion theory source. The resonance absor;ition of U-23 8 is calculated from Hellstrand's measured resonance integral data. To account for fuel rod resonance interac*:ons including the effects of water gaps, a Dancoff interaction correction is apphed to the individual fuel rods through the U-23 8 resonance integral.
l The analysis described in the preceding paragraph provides three neutron group nuclear constants for the individual fuel rods which comprise a fuel as s e mbly. Nuclear constants for the remaining materials, i.e., water in the water gaps and channels, are obtained using similar techniques. The se data together with a geometric representation of the fuel assembly, are used to obtain assembly average lattice data. Two-dimensional three group diffusion theory is used to compute the assembly average lattice data. All materials in the calculation are treated as diffusing regions with the excep-tion of control blades. For this case, a logarithmic derivative boundary condition, simulating a near black neutron boundary, is applied to the ther '
mal neutron group and an absorption cross section is used to accommodate epithe rmal absorption in the control mate rial.
B.
Core Characterisdes
- 1. Power Distribution The central loading of Type II fuel shown in Figure 6 permits the Type II fuel to be operated at a higher average power density relative to the surrounding Type I fuel. This is illustrated by the gross radial porter distribution, computed for a central loading arrangement and a typi' al c
l
d p
,m 10 i
control rod arrangement, which is presented in Figure 8.
This power distribution was calculated using an R-Z geometry, three neutron group diffusion theory method which includes the effects of bo(ling.
1 A scattered loading arrangement, typical of what will undoubtedly be used, is shown in Figure 7.
A fuel assembly powe r distribution, com-puted for the loading arrangement indicated on Figure 7 and a typical control rod configuration, is given in Figure 9.
A three neutron group, two dimensional diffusion theory method was used to obtain the data.
The total core loading, the fuel loading arrangement, and the control l
rod configuration were represented in the computation.
To simulate the effects of boiling and three dircensions, the distribution given in Figure 9 was synthesized from distributions computed indepen-dently for the top and bottom regions of the reactor. It is significant to note that, except for the power difference between 'he Type I and i
Type II fuel assemblies, this distributien is almost identical to the distribution computed for the o'riginal core with the same control rod configuration. This results from the fact that, for a given control rod configuration. the uniform distribution of new fuel does not affect the gros s powe r uis tribution.
9 The PF assemblies, in accordance with the license provisions, will be separat.ed by at least four Type I or Type II fuel assemblies. Since this separhtion avoids any possibility of interaction between PF fuel assemblies, they will not affect the gross powe r distribution. The re -
fore the PF assembly powers may be computed by applying relative assembly powe r data to gross power distribution data. Detailed com-putations indicate that the PF assemblies do not significantly affect the local power density of the immediately adjacent fuel assemblies.
The assembly powers of each fuel assembly type relative to unirradiated Type I fuel was computed for scattered loading arrangements using a two-dimensional (x, y), three-neutron group diffusion theory method. The power of the various fuel assembly types relative to an unirradiated, xenon and Sm loaded, Type I fuel assembly are summarized on Table 11.
The relative assembly power data are essentially independent of void fraction and water temperature over the range of interest. This is true because the variation of the nuclear characteristics with temper-ature and void fraction are approximately the same for all fuel assemblies.
The power distribution within the various fuel assemblies was computed using two-dimensional (x, y) three neutron group diffusion theory methods The results of the analyses are illustrated by the relative power maps, Figures 10 an~dT1. The fuel rod peaking factors for the various fuel assemblies are given on Table 12.
4
- 2. Temperature and Void Coefficients Temperature and void coefficient data for Type I and Type II fuel are summarized on Tables 13 and 14. These da+2 indicate that the tem-perature and voTd' coefficients of the Type II fuel are more negative than the Type I fuel coefficients. Thus, the temperature and void co-efficients of a core composed of a uniform mixture of Type I and Type II s
3 1
E a
. _ ~
l y..
o t
gi r
fuel will be slightly more negative than the coefficients for a core com-posed of Type I fuel only. This will also be true for a core composed of an isolated region containing Type II fuel surrounded by a Type L fuel region.
After refueling is completed and before the core is brought to operating power levels, the temperature coefficient will be me asured. It will be verified by experimental measureme nt that the temperature coefficient is negative in accordance with the license requirements.
No temperature and void coefficient data are presented for the PF fuel as s emblie s.
Since these assemblies are scatte red throughout the core, i
the temperature and void coefficients will not be influenced significantly by the PF assemblies. However, it should be noted that the lattice data presented in Tables 9 and 10 indicate that the coefficients of the PF assemblies should be about the same as those of the Type I and II d
l as se mblie s.
I
- 3. Reactivity Control The lattice data presented on Tables 9 and 10 indicate that the lattice i
material buckling decreases and the control strengt h increases as the temperature and void fraction increases. This being true, the cold shutdown condition taxes the cortrol system more severely than any other condition. Thus, if the control system satisfies the cold shutdown margin requirements, adequate reactivity control is avail-able for any other operating condition.
Before loading is commenced, sufficient calculations will have been performed to provide the basic information required to demonstrate compliance with the shutdown margin recuirements.throughout the loading. During the loading, sufficient measurements shall be made to demonstrate compliance with the shutdown margin requirements throughout the loading process.
a.
Central Loading Arrangement Since the controlled ke of Type II fuel is in excess of 1.0 with Zr channels, Type II fuel cannot bc loaded in a localized region with l
Zr channels on all assemblies. If a central loading arrangement i
is used, it will be necessary to put SS channels on a sufficient number of the Type II fuel assemblies to assure compliance with the shutdown margin requirements. The exact numbeT of SS chan-nels required will depend upon the loading configuration selected.
The lattice data presented on Table 9 indicate that a configuration consisting of 50% Zr channels and 50% SS channels, in a uniform array, will meet the shutdown margin requirements for any central loading arrangement.
- b. Scattered Loading Configuration Detailed calculations have been performed to evaluate the shut-down margin for a typical scattered loading arrangement. The results of these calculations indicate that, for a uniform mixture,
1
~
s (3
. ~
12 consisting of.1/4 new Type.II fuel.in.Z r channels,1/4 new Type I fuel, and 1/2 Type I fuel at exposures in excess of 2000 MWD / T, the shutdown margin requirements are met in the controlled region of the core. Further, for the planned loading configuration, the calculations indicate that the shutdown rnargin requirements can be met on the periphery.of the control system for a 464 assembly
- core configuration.
4 i
k 4
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9
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~'
g.v yr y,_-+v e-y
r 13 SECTION V SAFETY EVALUATION The safety evaluation respecting the use of fuel elements described in the previous section consists of the examination of mechanisms by which power is rapidly in-creased and by which removal of heat is adversely affected. The two areas of fun damental inte rest are: (i) the addition of reactivity by withdrawal of control rods o r addition of fuel, and (ii) the sudden change in nuclear and hydraulic cha racte ristic s, together with the associated stability implications resulting f rom the loss of pumps.
With respect to these two areas, analyses have been made of the worst con-ditions assumed to occur in any mode of operation which lead to the closest approach to the thermal limits of the fuel. The conditions hypothesized are believed to actually have an extremely small probability of occurrence. Never-th el e s s, even if the as sumed conditions we re to occu r, the analyses show that ie re would be no fuel cladding failures.
In addition to considering the effect of additions of reactivity and loss of pumps, this safety evaluation also incluas analyses of the effects of: (i) Type II fuel on system stability, (ii) the new design on fuel cladding failures, (iii) fuel location error, and (iv) Typr fuel on the consequences of the maximum credible accident analyzed for the original core loading of Type I fuel.
A.
Additions of Reactivity On a comparative basis, it can be shown that the transients arising f rom additions of reactivity are less severe with the use of Type II fuel, solely or interspersed with Type I fuel, than were analyzed for the original core loading.
This conclusion is based upon the lattice data discussed in Section IV, above, and set forth in Tables 9 and 10, which show that the control rod worths are considerably smaller for Type II fuel than for Type I fuel.
It follows, therefore, that the total reactivity change and rate of reactivity addition resulting from withdrawing a control rod from a region of the core consisting of only Type II fuel is less than that of Type I fuel. Fu rthe r-more, any core consisting of inixtures of Type I and Type II fuel also must result in a maximum control rod worth less than the maximum control rod worth for a core composed of only Type I fuel. Therefore, the approach to thermal limits of the fuel during control rod runout or the step insertion of reactivity, will be less severe than was considered in the safety analysis of the original core loading (see GEAP-1044, p.131).
Rega rding the step insertien of reactivity, the original analysis involved computer studies of step inputs of reactivity of 0.4% to 0.6% ak. These results showed that even if a scram should not occur, the transient would
t 14 l
settle out very quickly (about two seconds) at a slightly higher power with no serious system effects and that even much larger step changes in reactivity would not produce serious results.
The mechanisms which limit the severity of the transient caused by additions of reactivity are (i) the prompt increase in neutron absorption 238 in the U by Dopple r broadening and (ii) the effects of void formation.
These mechanisms are applicable equally to both Type I and Type II fuel. In this connection it will be noted that all of the changes in design incorporated in Type II fuel tend to make these mechanisms more effective in limiting transients. The ThO2 rods in the Type II fuel produce a slight inc rease in ove rall Doppler reactivity coefficient, because the resonance 232 is slightly less while the integ ral and total resonance capture of Th Doppler broadening is greater than that of U238 Considering the effects of void formation, the slightly shorter time constant of the Type II fuel, compared to the Type I fuel, and the effect of the increased Doppler coeffi-cient result in a slightly lower flux peak in the Type II fuel.
An accident in refueling involving the maximum reactivity change has been examined. Again solely for purposes of considering the worst possi-ble condition, it was assumed that the entire core, less on,e Type II fuel assembly, has bean loaded as shown in Figure 6 and that 'zircaloy channels have been used incorrectly on all., Type II fuel assemblies. That this assumption represents a core-fuel arrangement making the most seve re loading accident possible, may be ve rified by obse rving that the controlled and uncontrolled lattice material bucklings of the Type II fuel with zircaloy channels (see lattice data in Section IV) are larger than those of the f resh Type I fuel. Therefore, the smallest possible uncon-trolled region which is just-critical with a vacent position in the center occurs vith Type II fuel in zircaloy channels. Any core composed of mixtures of Type I and Type II fuel with stainless-steel and zirconium channels will have a large r just= critical size. The reactivity worth of a single fuel assembly increases as the size of the just-critical core decreases.
Thus, the largest bundle worth which can be produced for any refueling condition has been assumed. Further assumptions made in the analysis of this hypothetical accident are:
1.
Control rods adjacent to the vacant fuel position are withdrawn, causing the reactor to become just-critical.
2.
The source level instrumen* flux counter fails, or the operator fails to observe increase in count rate upon withdrawal of the rods.
3.
A new Type II fuel assembly is inserted in the vacent fuel position at the maximum design rate of the hoist (~12 inch /sec. ).
4.
The pe riod sc ram circuitry fails.
5.
The mode rator tempe rature is 68 *F.
l lM
'm lS The maximum reactivity addition rate reached duringithis hypothetical accident is 0. 003 ok/sec. The resulting maximum center fuel tempera-ture is computed to be about 4600*F (based on a UO2 thermal conductivity 2
of 1.15 Btu /hr/ft 7. F).
Some cente r melting in the hottest fuel rod may occur, but no cladding failures are expected. The reactor is automati-cally shut down by the high neutron flux scram circuitry, with some localized boiling occurring.
Mechanical and procedural measures (presented in the original accident analysis of Dresden, see GEAP-3076, p. 66) are utili' zed Ao_ prevent this type.of accident.. Moreover,Leven thoup,h such measdres were circum-vented, the severity of the accident would be less than'. discuss'ed above unless the core were incorrectly loaded with Zr channels on all Type II fuel assemblies.
Even in the.unlikely event that all the accident conditions we re simultan-eously satisfied, the incident would not result in release of fission products.
B.
Loss of Coolant Of the several hypothetical accidents which adversely af fect the removal of heat f rom the core, the most severe in terms of the the rmohydraulic transient is the simultaneous loss of power to all recirculating pumps.
This limiting loss of coolant accident is analyzed for both the central and scatte red loading.
The analysis for this accident is based on the investigation of the steady state thermal-hydraulic performance of the core at the worst possible end condition. This condition assumes that the reactor is at full power and that the total recirculation flow is dete rmined by the water density dis-tribution in the total primary coolant loop prior to the pump trip. This analysis is considered to be conservative for the following reasons:
1.
The power of the reactor is not assumed to decrease with the associated increase in moderator void fraction, caused by the dec reased flow.
2.
The transient reduction in flow due to the kinetic energy of the coolant is neglected. This has the effect of instantaneously drop-j ping the flow rate to the natural circulation flow component of the total flow existing at the time of the pump trip.
3.
The additional flow which results f rom the accumulation of voids in the risers is neglected. The steady state natural circulation flow rate is higher than the value used in this analysis.
Without considering these three facto rs, the results obtained from the analysis are shown below:
1
t y
i L
Total Core Flow Rate 6 lb /h r.
in 10 Befo re Afte r Min. Burnout Ratio Pump Pump After Pump Type Loading Fuel T rip T rip T rip Type I 25.7 10.4 2.46 Type II 25.7 10.4 1.89 Type I 25.7 10.4 1.52 catte red Type II 25.7 10.4 2.05 In conclusion, the worst possible hydraulic condition following the loss of flow accident, due to simultaneous trip-out of all four circulating pumps, results in minimum burnout ratios in excess of the minimum burnout ratto previously analyzed for similar conditions for an all Type Iloading (see GEAP-1044, p.142 and GEAP-3009, p. 54).
Nuclear data and experience in operating the Dresden reactor indicate that the power in all regions of the core will increase uniformly with changes in load. Finally, the pump trip analysis with the central loading does not indicate that any adverse local changes on flow or power are produced.
C.
System Stability The operation of the Dresden Nuclear Power Station has adequately demon-strated the ability of the power plant to respond stably to a large variety of operation conditions. It has been experimentally proved that the Dresden reactor can be operated stably with void content in the core in excess of 30%.
- During startup, rod oscillating test, pressure regulator and system response tests, all indidated that a large margin from any instability at all ope rating conditions exists.
The current improvement in fuel design will have only a very minor effect on the ove rall system stability. This follows from the fact that of the numerous mechanical, thermal, and nuclear parameters affecting stability of the first Dresden core, only two are changed. Those are:
1.
The Type II fuel rod is slightly smaller in diameter which leads to the conclusion that heat generated within the fuel will reach the coolant slightly faste r than would be expected for the previous de-sign at the same power density. The resulting change in thermal response is reflected in the effective fuel time constant ** the. largest Report on the High Void Test, dated October 31, 1961, submitted with. lette r from I. L. Wade to R. Lowenstein on November 16,19P
- The fuel time constant is the time required to achieve 6. 3% of the steady state increase in heat flux in response to a step increase in reactor power.
D Q
component of which is 10 seconds for the Type II fuel and 13.5 seconds for the Type I fuel of the first Dresden core.
By com-parison, the VBWR has operated stably with a mixed core of both plate and rod fuel assemblies which had time constants of about 1/10 second and 13 seconds, re spectively. The variation in time constant in the mixed Dresden core is insignificant compared to the change that could be tolerated without approaching instability.
2.
The new improved fuel will naturally operate at a slightly higher power density than the fuel of current design. The accumulative effects of higher power, less flove restriction and larger entrance nozzle lead to a slightly higher velocity of the coolant in the fuel channel. The time required for a unit volume of coolant to travel through the core, i. e., " sweep time, " will dec rease slightly, tend-ing to increase the core stability. This effect, how ev e r, witi be so slight that no detectable change in stability is expected.
A further examination of flow distribution has been made as part of the
" loss of flow accident. " As pa rt of the study, it was of inte rest to determine the possible redistribution of flow between the Type I and Type II fuel. The study assumed the center of the core loaded with Type II fuel surrounded with Type I fuel, as shown in Figure 6.
The results of the loss of flow accident, reported in pa rag raph B above, show that there is a very small change in flow distribution between the two regions of the Nuclear data and experience in ope rating the Dresden reacto.r indicate core.
that tne power in all regions will increase uniformly with changes in load.
This is indicative of the response to be expected from sudden change in power of flow of the entire core. This evidence. taken togethe r, indicate s that the re should be no change in the fundamentally stable, predictable operation of the Dresden reactor with Type II fuel, and further, that the addition of the Type II fuel assemblies does not appreciably affect the core saf.ty or performance.
D.
Fuel Cladding Failure Evaluation of the original Dresden core indicated that about 4000 fuel element segments could be leaking at a rate of 10-6% of the noble gas activity per second without exceeding the maximum permissible stack emis sion rate. (See GEAP-1044, p.141 and GEAP-3076, p. 27).
This analysis is equally applicable to the new fuel with the exception that the rods are not segmented so that 1000 fuel reds wou!d result in approximately the same release as 4000 fuel segments. Although a single failure can result in an increase in gas release, the factor of 4 decrease in the number of welds increases the reliability adequately to compensate for the change in effect.
Relevant to this discussion are the approximately 1000 fuel rods con-taining ThO.
Assuming the same distributien of fission products 2
~_
. e m.N m T 18 235 233 and U
, the criterion for evaluating resulting f rom fission of U the change in hazard resulting from including ThO2 comer rods in the new fuel design is the retention of fission products by ThO. The 2
best available experimental evidence indicates that the fission gas retention of ThO2 is superior to that of UO. In addition, the thermal 2
conductivity is 2 times greater than that of UO2 and, unlike UO -
2 ThO2 has only one oxide state, hence it is stable in an oxidizing atmo sphe re.
r These physical properties indicate that the central temperature would
{~
be s'ightly lower in a ThO2 rod than a UO2 rod for the same surface heat flux and that cladding failure will result in release of a smaller quantity of fission gas than from a UO2 rod.
E.
Worst Fuel Assembly Location Error If, by error, a Type I fuel assembly with a small diameter orifice were placed in the Type II region of the central core loading, the fuel rods could overheat and rupture. If the rupture occurred during the approach to full power, the change in power which occurs relatively slowl/ and the longitudinal and radial peak-to-ave rage tempe rature within a fuel bundle 4
would cause a progressive failure of the fuel tubes in the misplaced assembly.
with inc reasing powe r.
The off-gas monitoring system would indidate a relatively large increase in activity within about two minutes of the first failure and the activity would continue to increase with increasing power.
The consequences that could result from this hypothetical accident are, therefore, limited to the relatively gradual rupture of a few fuel rods.
F.
.staximum Credible Accident The " maximum credible accident" for the Dresden plant is conceived as following a hypothetical instantaneous complete severance of one of the bottom inlet lines to the reactor while the reactor is in a " hot" standby 4
candition. (See GEAP-3076, p. 28. )
Following the accident, the fuel starts on a rising temperature transient and eventually all the fuel reaches the melting temperature.
Comparing the former analysis ~of the first Dresden core with the current i
analysis,-it may be observed that:
1.
The total release of energy resulting from the accident has not been changed and the probability of a stainless-steel water reaction is even less than the previously evaluated zirconium-water reaction.
m S
w 19 i
i 2.
A relatively minor change, resulting from the higher power density in the Type II fuel and the changes in fuel geometry, is that some SS/ ' cladding reaches melting temperature slightly faster and some 4
Zr cladding reaches melting temperature slightly slower, than would 3
have occurred in the previous core.
In summary, the changes in the core will not affect the total stored energy l
within the reactor vessel or the applicability of the original argument regarding simultaneous nuclear excursion or metal-water reactions.
There is, however, a relatively minor change in fission product celease 4
rates. But since the primary factor, i. e., power, has not changed,
{
the small increase in release rates will not appreciably affect the radio-logical aspects of the maximum credible accident.
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5 Talin E I TlifRMOllYDRAULIC ANAL'ISI.i AT RATLD P02R WITil MAXIM 1hi SECONDARY STEAM FLOW (Scattered Loading)
A. CENERAL' PIANT DATA
- 1. Reactor Output HW 626.4
- 2. Primary Steam Flow Rate 106 lb/ht 1.195 6
- 3. Secondary Steara Flow Rate 10 lb/hr 1.400
- 4. Feedwater Enthalpy BfU/lb 374.6
- 5. Core Inlet Enthalpy bfU/lb 4b5.6 i
B. CORE DESCHIPTION
- 1. Number of Fuel Assemblies 464
- 2. Fraction of Power Generated in Fuel 0.97
- 3. Total.lleat Transfer Area sq. It 22,481 92,245
- 4. Average ficat Flux BTU /hr-ftl
- 5. Total Core Flow Rate 106 i
lb/hr 24.633 6
- 6. Leakage Flow Rate 10 'lb/hr 1.972
- 7. Average Void Content in Channels 0.139 C. FUEL' DESCRIPTION
- 1. Type of Fuel I
11 I
1 1
I i
- 2. Orifice Zone o
I O
1 11 III
- 3. Number of Asse.Lblies 99 99 d6 68 34 78
- 4. Rod Diameter inches 0.567 0.44?
0.567 0.5b7 0.567 0.567
- 5. Active Fucl Length inches 106.5 110.5 106.5 106.5 106.5 106.5 6 Numbcr of Rods per Assembly 36 49 36 36 36 36
- 7. Ileat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43
- 8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.2 0.83
. 9. Maximum Heat Flux Peaking Factor 2.12 1.13 2.12 2.12 2.12 2,12 (Including Area Correction)
D. RESULTS FOR l'EAK CHANhEL -
3
- 1. Channel Active Coolant Flow Rate 10 lb/hr 53.95 54.47 50.47 39.78 35.73 ~
29.38
- 2. Channel Average Void Fraction 0.170 0.221 0.241 0.286 0.300 0.244
- 3. Channel Exit Void Fraction 0.503 0.574 0.597 0.646 0.666 0.601
- 4. Channel Exit Quality 0.070 0.096 0.107 0.136 0.148 0.110
- 5. Peak licat Flux 10 BTU /hr-ft2 241 238 276 248 235 162
- 6. Minimum Burnout Natio 3.74 3.67 3.22 3.42 3.55 5.01
TABLE 2 1
TilERMollYDRAULIC ANALYSIS AT 1257. OF RATED POWER WITil MAXIMUM SECONDARY STEAM FLOW
~
(Scattered Loading)
A. GENERAL PIANT DATA
- 1. Reactor Output FM 782.6
- 2. Primary Steam Flow Rate 106 lb/hr 1.847
- 3. Secondary Steam Flow Rate 106 lb/hr 1.400
- 4. Feedwater Enthalpy BTU /lb 374.6
- 5. Core Inlet Enthalpy BTU /lb 481.7 3
B. CORE DESCRIPIION
- 1. Number of Fucl Assemblies 464
- 2. Fraction of Power Generated in Fuel 0.97
- 3. Total Heat Transfer Area sq. ft.
22,481
- 4. Average lleat Flux HTU/hr-ft2 115,247
- 5. Total Core Flow Rate 10 lb/hr 24.861
- 6. Leakage Flow Rate 10 lb/hr 1.992
- 7. Average Void Content in Channels 0.197 C. FUEL DESCRIPTION
- 1. Type of Fuel I
II I
I I
I
- 2. Orifice Zone 0
I O
1 11 III
- 3. Number of Assemblics 99 99 86 68 34 78
- 4. Rod Diameter inches 0.567 0.442 0.567 0.567 0.567 0.567
- 5. Active Fuel Length inches 106.5 110.5 106.5 106.5 106.5 106.5
)
- 6. Number of Rods per Assembly 36 49 36 36 36 36 l
- 7. lleat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43
- 8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.2 0.83
- 9. Maximum lleat Flux Peaking Factor 2.12 1.73 2.12 2.12 2.12 2.12 (Including Area Correction)
D. RESULTS FOR PEAK CHANNEL
- 1. Channel Active Coolant Flow Rate 103 lb/hr 52.30 55.57 47.84 39.35 35.80 30.30
- 2. Channel Average Void Fraction 0.112 0.132 0.165 0.190 0.201 0.14 6
- 3. Channel Exit Void Fraction 0.605 0.640 0.688 0.718 0.731 0.664
- 4. Chanael Exit Quality 0.112 0.132 0.165 0.190 0.201 0.14 6
- 5. Peak IIcat Flux 103 BTU /hr-f t2 301 298 344 310 293*
203
- 6. Minimum Burnout Ratio 2.98 2.94 2.56 2.73 2.85 4.02
TABLE 3 TilERM0liYDRAULIC ANALYSIS AT RATED POWER WITil MAXIMUM PRIMARY STEAM FLOW (Scattered Loading)
A. CENERAL PIAhT DATA
- 1. Reactor Output MU 626.4
- 2. Primary Steam Flow Rate 106 lb/hr 1.601
- 3. Secondary Steam Flow Rate 100 lb/hr 1.000
- 4. Feedwater Enthalpy BTU /lb 374.6
- 5. Core Inlet Enthalpy BTU /lb 496.6
,j)
B. CORE DESCRIPTION
- 1. Number of Fuel Assemblies 464
- 2. Fraction of Power Generated in Fuel 0.97
- 3. Total Heat Transfer Area
.q.
ft.
22,481 2
- 4. Average lleat Flux BTg/nr-ft 92,245
- 5. Total Core Flow Rate 10 lb/hr 24.764
- 6. Leakage Flow Hate 106 lb/hr 1.981
- 7. Average Void Content in Channels 0.190 C. FUEL DESCRIPTION l
- 1. Type of Fuel 1
11 I
I I
I
- 2. Orifice Zone 0
I O
I II III
- 3. Number of Assemblies 99 99 86 68 34 78
- 4. Rod Diameter incleei 0.567 0.442 0.567 0.567 0.567 0.567
- 5. Active. Fuel Length inches 106.5 110.5 106.5 106.5 106.5_
106.5 J
- 6. Number of Rods per Assembly 36 49 36 36 36 36 d
- 7. Ileat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43
- 8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.20 0.83
- 9. Maximum lleat Flux Pee !ng Factor 2.12 1.73 2.12 2.12 2.12 2.12 (Including Area Correction)
D. RESULTS FOR PEAK CllANNEL
- 1. Channel Active Coolant Flow Rate 103 lb/hr 53.02 55.17 49.10 39.90 36.04 29.91
- 2. Channel Average Void Fraction 0.225 0.262 0.289 0.324 0.336 0.280
- 3. Channel Exit Void Fraction 0.558 0.603 0.636 0.673 0.685 0.624
- 4. Channel Exit Quality 0.091 0.111 0.130 0.152 0.163 0.123 2
241 238 276 248 235 162
- 5. Peak lieat Flux 103 BTU /hr-ft
- 6. Minimum Burnout Ratio 3.74 3.67 3.22 3.43 3.58 3.01
TAhlE 4 TliERM0 HYDRAULIC ANALYSIS AT 1257. OF RATED POWER WITl! MAXIMUM PRIMARY STEAM FLOW (Scattered Loading)
- [
A. CENERAL PLANT DATA
- 1. Reactor Output MW 782.6
- 2. Primary Steam Flow Rate 10gIb/hr 2.253
- 3. Secondary Steam Flow Rate 10 lb/hr 1.000
- 4. Feedwater Enthalpy BTU /lb 374.6
- 5. Core Inlet Enthalpy BTU /lb 492.5
)
i
,)
B. CORE DESCRIFTION
- 1. Number of Fuel Assemblies 464
- 2. Fraction of P wer Cencrated in Fuel
- f. 97
- 3. Total Heat Tr&nsfer Area sq. ft.
22,481 2
115,241 I
- 4. Average Heat Flux BTU /hr-ft
- 5. Total Core Flow Rate 106 lb/hr 24.937
- 6. Leakage Flow Rate 106 lb/hr 1.995
- 7. Average Void Content in Channels 0.241 C. FUEL DESCRIFTION 1
11 1
1 1
I
- 1. Type of Fuel
- 2. Orifice Zone 0
1 0
I II III
- 3. Number of Assemblics 99 99 86 68 34 78
- 4. Rod Diameter inches 0.567 0.442 0.567 0.567 0.567 0.567
- 5. Active Fuel Length inches 106.5 110.5 106.5 106.5 106.5 106.5 j
- 6. Number of Rods per Assembly 36 49 36 36 36 36
- 7. Heat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43
l
- 8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.20 0.83
- 9. Maximum Heat Flux Peaking Factor 2.12 1.73 2.12 2.12 2.12 2.12 (Including Area Correction)
D. RESULTS FOR PEAK CHANNEL
- 1. Channel Active Coolant Flow Rate 103 lb/hr 51.32 56.26 47.42 39.58 36.16 30.81
- 2. Channel Average Void Fraction 0.288 0.308 0.353 0.381 0.390 0.327
- 3. Channel Exit Void Fraction 0.540 0.664 0.710 0.734 0.741 0.681
- 4. Channel Exit Quality 0.132 0.146 0.184 0.205 0.214 0.159 3
- 5. Peak Heat Flux 10 BTU /hr-ft2 301 258 344 310 293 203
- 6. Minimum Burnout Ratio 2.98 2.94 2.56 2.73 2.85 4.02
TABLE 5
,~
TilERM0llYDRAULIC ANALYSIS AT MATED POWER WITil MAXIMUM PRIMARY STEAM FLOW (Scattered Loadind Including PF Eleuents)
CENERAL PIAlff DATA i1. Reactor Output HW 626
- 2. Primary Steam Flow Rate 106 lb/hr 1.601 6
- 3. Secondary Steam Flow Rate 10 lb/hr 1.000
- 4. Fe:dwater Enthalpy BTU /lb 374.6
}
i 5. Core Inlet Enthalpy BTU /lb 496.6 b
CORE DESCRIPTION
- 1. Numbsr of Fuel Assemblies 464 l
- 2. Fraction of Power Generated in Fuel 0.97
- 3. Total lleat Transfer Area sq. ft.
22,444 2
- 4. Avsrage lleat Flux BTg/hr-ft 92,338
- 5. Total Core Flow Rate 1A lb/hr 24.7 6
- 6. Leakage Flow Rate 1a lb/hr 1.981
- 7. Avartge Void Content in Chancela 0.201 FUEL DESCRIPTION 1.' Type of Fuel I
II I
I I
I PF 1-4 PF 5-9 PF 10-11 PF 12
- 2. Orifice Zone 0
1 0
I II III O
O O
O s
- 3. Number of Assemblies 87 99 86 68 34 78 4
5 2
1
- 4. Rod Diameter inches
.567
.442
.567
.567
.567
.567
.480
.489
.412 0.I,3 l
S. Active Fuel Length inches 106.5 110.5 106.5 106.5 106.5 106.5 105.81 111.94' 105.69 99- )
- 6. Numbar of Rods per Assembly 36 49 36 36 36 36 36 36 64 6J
- 7. Ileat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43 39.89 42.99 60.80 57.0
- 8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.20 0.83 1.75 1.69 1.52 1.59
- 9. Maximum lleat Flux Peaking Factor 2.12 1.73 2.12 2.12 2.12 2.12 2.06 1.93 1.41 1.78 (Including Area Correction)
RESULTS FOR PEAK CllANNEL
- 1. Channel Active Coolant Flow Rate 103 lb/hr 50.85 $3.97 48.09 39.90 36.04 29.91 87.7 91.8 58.7 58.7
- 2. Channel Average Void Fraction 0.257 0.283 0.315 0.324 0.336
.280 0.197 0.176 0.293 0.301
- 3. Channel Exit Vcid Fraction 0.592 0.621 0.6 54 0.673 0.685.624 0.513 0.480 0.615 0.629
- 4. Channel Exit Quality 0.104 0.121 0.141 0.152 0.163.123 0.074 0.062 0.116 0.124 2
- 5. Peak lleat Flux II IU/hr-ft 24L 238 276 249 235 163 333 301 198 262
- 6. Minimum Burnout Ratio 3.67 3.64 3.16 3.38 3.51 4.93 2.87 3.24 4.57 3.46
TABLE 6 THERM 0llYDRAULIC ANALYSIS AT 1257. OF RATED POWER WITil MAXIMUM PRIMAP.Y STEAM FLOW (Scattered Loading Including PF Elements)
CENERAL PLANT DATA
~1.
Rr. actor Output
}6i 782.7
- 2. Primary S eam Flow Rate 106 lb/hr 2.253 t
- 3. S:condary Steam Flow Rate 10 lb/hr 1.000
- 4. Ferdwater Enthalpy BTU /lb 374.6
}
- 5. Cora Inlet Enthalpy BTU /lb 492.5 CORE DESCRIPTION
- 1. Numb 3r of Fuel Assemblies 464 l
- 2. Fraction of Power Generated in Fuel 0.97
- 3. Total Heat Transfer Area sq. ft.
22,444
- 4. Avsrage Heat Flux BTU /hr-ft2 115.452
- 5. Total Core Flow Rate 106 lb/hr 24.937
~6. Leckrge Flow Rate 106 lb/hr 1.955
- 7. Avarage Void Content in Channels 0.250 FUEL DESCRIPTION j
- 1. Typm of Fuel I
II I
1 1
I PF 1-4 PF 5-9 PF 10-11 PF 12
- 2. Orifice Zone 0
1 0
I II III O
O O
O
- 3. Number of Assemblies 87 99 86 68 34 78 4
5 2
1
- 4. Rod Diameter inches 0.567 0.442 0.567 0.567 0.567 0.567 0.480 0.489 0.412 0.t * }
- 5. Active Fuel Length inches 106.5 110.5 106.5 106.5 106.5 106.5 105.81 111.94 105.69 9"
s
! 6. Number of Rods per Assembly 36 49 36 36 36 36 36 36 64 6d
- 7. Heat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43 47.43 34.89 42.99 60.80 57.0 i
'8. Maximum Channel Peaking Factor 1.23 1.49 1.41 1.27 1.20 0.83 1.75 1.69 1.52 1.59
- 9. Maximum Heat Flux Peaking Factor 2.12 1.73 2.12 2.12 2.12 2.12 2.06 1.93 1.41 1.78 (Including Area Correction)
RESULTS_FOR PEAK CHANNEL i
- 1. Channel Active Coolant Flow Rate 103 lb/hr 49.34 55.0 45.81 39.58 36.16 30.81 85.0 89.0 55.2 55.2
- 2. Channel Average Void Fraction 0.318 0.331 0.381 0.381 0.390 0.327 0.250 0.237 0.336 0.373
- 3. Channel Exit Void Fraction 0.665 0.680 0.729 0.734 0.741 0.681 0.606 0.570 0.695 0.710
- 4. Channel Exit Quality 0.147 0.158 0.199 0.205 0.214 0.159 0.109 0.095 0.171 0.182
- 5. Peak Heat Flux 103 BTU /hr-ft2 301 292 T+5 311 294 203 416 376 247 327 l
'6. Minimum Burnout Ratio 2.93 2 93 2.51 2.70 2.81 3.97 2.28 2.56 3.63 2.7:i i
TABLE 7 TIERM0llYDRAULIC A!M SIS AT 1257. OF RATED POWER WITil MAXIMUM PRIMARY STEAM FLOW (Central Loading)
A. GENERAL PIMr DATA
- 1. Reactor Output HW 782.7 6
- 2. Primary Steam Flow Rate 10 lb/hr 2.253 6
- 3. Secondary Steam Flow Rate 10 lb/hr 1.000
- 4. Feedwater Enthalpy BTU /lb 374.6
- 5. Core Inlet Enthalpy BTU /lb 492.6
)
i B. CORE DESCRIPTION i
- 1. Number of Fuel Assemblies 464 j
- 2. Fraction of Power Generated in Fuel 0.97 i
- 3. Total Heat Transfer Area sq. ft.
22,486
- 4. Average Heat Flux BTU /hr-ft2 115,236 5, Total Core Flow Rate 106 lb/hr 24.76
- 6. Leakage Flow Rate 106 lb/hr 1.98
- 7. Average Void content In Channels 0.184 C. FUEL DESCRIPTION t
- 1. Type of twel I
II I
I I
- 2. Orifice Zoae o
I I
II III c
- 3. Number of Aa.semblies 100 100 56 120 88
- 4. Rod Diameter inches 0.567 0.442 0.567 0.567 0.567
}
- 5. Active Fuel Length inches 106.5 110.5 106.5 106.5 106.5
- 6. Number of Rods per Assembly 36 49 36 36 36
- 7. Ileat Transfer Area per Assembly sq. ft.
47.43 52.21 47.43 47.43 47.43
- 8. Maximum Channel Peaking Factor 1.10 1.58 1.08 0.990 0.780
- 9. Maximum Heat Flux Peaking Factor 2.12 1.73 2.12 2.12 2.12 (Including Area Correction)
D. RESULTS _FOR PEAK CHANNEL
- 1. Channel Active Coolant Flow Rate 103 lb/hr 60.52 60.21 45.38 40.92 33.28
- 2. Channel Average Void Fraction 0.200 0.303 0.285 0.289 0.281
- 3. Channel Exit Void Fraction 0.534 0.659 0.636 0.642 0.631
- 4. Channel Exit Quality 0.082 0.144 0.130 0.134 0.127
- 5. Peak Heat Flux 103 IHU/hr-ft2 269 315 264 242 191
- 6. Minimuw Burnout Ratio 3.44 2.83 a.30 3.52 4.31
p l
?
TABLE 8 i
COMPARISON OF PEAK ANALYTICAL RESULTS WITH LICENSE LIMITS LICENSE LIMIT PEAK ANALYTICAL RESULTS i
Heat Flux Heat Flux 2
Fuel Type Burnout Ratio BTU /hr-ft Burnout Ratio BTU /hr-ft2 l
I 2.0 350,000 2.51 345,000 i
II 2.0 445,000 2.83 315,000 2.28 416,000 PF 1-4 2.0 425,000 PF 5-9 2.0 415,000 2.56 376,000 PF 10-12 2.0 475,000 2.76 327,000 i
1 3
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o
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~
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.i TABLE 10 ASSEMB LY AVERAGE LATTICE DATA
~
}
GREEN PF-1 T11ROUGII PF-12 FUEL ASSEMBLIES PF#
1 2
3,4 5, 6 7,8,9 10
- 11. 1 2 1
2 3, 4 5, 6 7,8,9 10 11, 1 2 Channel SS
.SS SS SS SS SS SS SS SS SS SS SS SS SS Ternperature (*F) 68 68 68 6,8 68 68 68 546 546 546 546 546 546 546
_s Void Fraction 0
0 0
0 0
0 0
0 0
0 0
0 0
0 1
K. Uncontrolled 1.055 1.080 1.11 8 1.085 1.113 1.008 1.058 1.057 1.083 1.143 1.111 1.13 8 1.0025 1.076 2(L it) 36.4 36.4 36.4 36.7 36.8~ 40.9 :11_0 59.6 59.8 59.9 57.9 58.1 65.8 66.1 2
M i
Channel Zr Zr Zr Zr Zr Zr Zr Zr Zr Ternperature (*F) 68 68 546 546 546 546 546 546 546 Void Fraction 0
0 0
0 0
0 0
0 0
K. Uncontrolled 1.13 6 1.192 1.160 1.194'I.260 1.7.21 1.261 1.12 2 1.209 M2 (L + t) 44.0 44.1 64.4 64.7 64.8~ 62.8 63.1 71.6 72.1 b
i
TABI.E 11 RELATIVE FUEL ASSEMBLY POWER DATA Fuel Type I
I I
II II II Channel Type Zr Zr Zr Zr Zr SS Temperature (*F) 546 546 546 546 546 546 Void Fraction
.0. 25
- 0. 25 O.25
.0.25 0
0
~ ')
Nxposure (MWD /T)
O 5000 10,000
,0 0
0 Relative Power
- 1.00 1.03
.92 1.17 1.17
- 1. 11 i
I Fuel Type PF-1 PF-2 PF-3,4 PF-5,6 PF-7, 8, 9 PF-10 PF-ll PF-12 Channel Type Z r-SS Zr-SS Zr-SS Z r-SS Z r-SS Zi-SS Z r-SS Zr-SS Temperature (*F) 546 546 S46 546 546 546 546 546 Void Eraction 0
0 0
0 0
0 0
0 Exposure (MWD /T) 0 0
0 0
0 0
0 0
i Relative Power
- 1.23-1.17 1.25-1.19.I.30-1.24 1.26-1.20 1.28-1.22 1.08-1.02 1.13-1.07 1.17 1.13 i
)
- -Relative Power = Fuel Assembly Power + Power of Green Sm and Xenon loaded Type 1 fuel assembly.
.. - ~_--.
TABLE 12 LOCAL (FUEL ROD) POWER PEAKING' FACTORS
- AT ZERO EXP06URE AND OPERATING TEMPERATURE
~.
l Fuel Type I
II II PF1 PF2 PF3-4 PFS-6 PF7,8,9
~
Channel Type
-Zr Zr SS Zr-SS Zr-SS Zr-SS Zr-SS Zr-SS Void Fraction 0
0.20 0.20 0
0 0
0 0
]
Local. Peaking Factor
- 1 33 1.22 1.17 1.19-1.15 1.14-1.12.
1.17-1.14 1.18-1.15-1.13-1.11 t!'
Fuel Type PF10
.PFll PF12
}-
Zr-SS' l'
Channel Type Zr-SS Zr-SS
'j.
Void Fraction.
0 0
0 i.
Local Peaking Factor
- 1.18-1.15 1.20-1.16' l.44-1.40 l-
} l Local peaking factor = Power of highest povered fuel rod in assembly + Power of average fuel rod in assembly s
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