ML20049H883
| ML20049H883 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/01/1982 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20049H882 | List: |
| References | |
| NUDOCS 8203040410 | |
| Download: ML20049H883 (34) | |
Text
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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly conta.ining 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximun. T/s enrichment of 4.0 weight percent U-235. gi CONTROL R00 ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.
All control rods shall be clad with stainless steel tubing.
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5.4 REACTOR COOLANT SYSTEM \
DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is 680'F.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,612 + 100 cubic feet at a nominal T avg f 525 F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
SEQUOYAH - UNIT 1 5-4 8203040410 B20301 PDR ADOCK 05000327 P PDR
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DESIGN FEATURES 5.6 FUEL STORAGF, CFITICALITY - SPINT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. A Ke rr equivalent to less than 0 95 when flooded with unborated water, wnich includes a. conservative allowance of 1.42% delta k/k for uncertainties. However, for some accident conditions, the presence of dissolved boron in the pool water is taken into US account as a realistic initial condition. This assumption can be made by applying the doubic contingency principle of ANSI N16.1-1975 which requires two unJikely, independent, concurrent events to produce a criticality accident,
- b. A nominal 10.375-inch center-to-center distance between fuel assemblies placed in the storage racks.
CRITICALITY - NEW FUEL
- 5. 6.1. 2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to-center distance between new fuel assemblies
( such that k eff will n t exceed 0.98 when fuel having a maximum enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is 5/S assumed. //2 DRAINAGE 5.6.2 lhe spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft. '
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.
I 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT
, 5.7.1 The components identified in Table 5.7-1 are designed and shall be
( maintained within the cyclic or transient limits of Table 5.7-1.
o SEQUOYAH - UNIT 1 5-5
e PROPOSED TECHNICAL SPECIFICATIONS SEQUOYAH NUCLEAR PLANT UNIT 2
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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies withrod Each fuel each fuel assembly conta.ining 264 fuel rods clad with Zircaloy -4.
shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1766 grams uranium.
The initial core loading shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel shall be similar T/s in physical design to the initial core loading and shall have a maximum g enrichment of 4.0 weight percent U-235.
CONTROL R00 ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length The full length control rod assemblies shall contain control rod assemblies.
a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.
All control rods shall be clad with stainless steel tubing.
5.4 REACTOR'C00LANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of 2485 psig, and
- c. For a temperature of 650 F, except for the pressurizer which is 680*F.
VOLUME 5.4.2 The total water and steam volume of fthe reactor coolant system is 525 F.
12,612 + 100 cubic feet at a nominal T avg 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5-4 SEQUOYAH - UNIT 2
s DESIGN FEATURES i
I 5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:
- a. A Ke rf equivalent to less than 0 95 when flooded with unborated I water, which includes a conservative allowance of 1.42% delta k/k for uncertainties. However, for some accident conditions, the presence of dissolved boron in the pool water is taken into T/S account as a realistic initial condition. This assumption can be #1 made by applying the double contingency principle of ANSI N16.1-1975 which requires two unlikely, independent, concurrent events to produce a criticality accident.
- b. A nominal 10 375-inch center-to-center distance between fuel assemblies placed in the storage racks.
CRITICALITY - NEW FUEL 5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21.0 inch center-to center distance between new fuel assemblies such that kgff will not exceed 0.98 when fue! having a maximum enrichment of 4.5 weight percent U-235 is in place and optimum achievable moderation is assumed. T/S
- 2 DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 722 ft. '
CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1386 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be l
i maintained within the cyclic or transient limits of Table 5.7-1.
SEQUOYAH - UNIT 2 5-5
O 4 ,
G 9 e ENCLOSURE 2 JUSTIFICATIONS FOR PROPOSED TECHNICAL SPECIFICATIONS FOR SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2
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e JUSTIFICATION FOR PROPOSED TECllNICAL SPECIFICATIONS FOR SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 CIIANGE #1 INCREASE OF Tile MAXIMUM FUEL ENRICilMENT FOR RELOAD FUEL FROM 3.5 WEIGilT PERCENT OF U-235 TO 4.0 WEIGilT PERCENT OF U-235 I
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1.1
SUMMARY
OF CRITICALITY ANALYSIS FOR SEQUOYAH SPENT FUEL RACK Criticality of fuel assemblics in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly *
- interaction. This is done by fixing the minimum separation between assemblies and inserting neutron poison between assemblie:. .
The design bas ~is for preventing criticality outside the redctor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (Keff) of the fuel assembly array will be less than 0.95 as recommended in ANSI N210-1976 and in "NRC Position for Review and
- Acceptance of Spent Fuel Storage and Handling Application."
In meeting this design basis, some of the conditions assumed are:
fresh 17x17 Westinghouse standard fuel assemblies of 4.0 w/o U-235 are stored, the pool of water has a density of 1.0 gm/cm3, the storage array is infinite in lateral and axial extent which is more reactive than the actual finite array, mechan,ical and method biases and uncertainties are included, the minimum poison loading is used, and for some accident Conditions. credit for the dissolved boron in the pool water is taken.
The design method which insures the criticality safety of fuel assemblies in the spent' fuel storage rack uses the AMPX system of codes for cross-section generation and KENO-IV for reactivity determination. A set of 27 critical experiments has been analyzed using th'e above method to demonstrate its applicability to criticality analysis and to establish the method bias and varia-bility which are then included in the reactivity analysis of the rack.
-The result of the above considerations is that the nuclear design of the rack will meet the requirements of NRC guidelines and criteria.
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1.2 CRITICALITY ANALYSIS FOR SEQUOYAH SPENT FUEL RACK 1.2.1 NEUTRON MULTIPLICATION FACTOR Criticality of fuel assemblies in the spent fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between-assemblies and inserting neutron poison between assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (Keff) of the fuel assembly array will be les; than 0.95 as
- recommended in ANSI N210-1976 and in "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications".
The following are the conditions that are assumeo in meeting this design basis.
, 1.2.2 NORMAL STORAGE.
- a. The fuel assembly contains the highest enrichment at.thnrized without any control rods or any noncontained burnable poison and is at its most reactive point in life. The enrichment of the
~ fuel assembly is 4.0 w/o U-235 with no depletion or fission product buildup. The a .setiiblies are conserva tisi ly edeled with. water replacing the assembly grid volume ard no U-234-or U-236 in the' fuel pellet.
- b. The storage cell nominal geometry'is shown on Figure 1.2-1.
- c. . The moderator is pure water at the temperature within the~ design limits of. the pool which yields .the largest reactivity. A con-servative value of 1.0 gm/cm 3 is used for the density of water.
No dissolved-boron is included in the water.
- c. .. . . .
- d. The array is infinite in lateral and axial extent by appropriate choice of boundary conditions in KEN 0 IV.
- e. Mechanical uncertainties and biases due to mechanical tolerances ..
- during construction are treated by either using " worst case" ..
' conditions or by performing sensitivity studies to obtain the appropriate values. The items included in the analysis are:
~..
-- poison pocket thickness stainless steel thickness .
can 10
-- center-to-center spacing
-- can bowing .
The calculation method uncertainty and bias'is discussed in Section 1.2.4. .
- f. Credit is taken for the noutron absorption in full length structural materials and in solid materials added specifically -
for neutron absorption. The minimum poison loading @.0232 gm-B10/cm2)is assumed in the poison plates and B 4 C particle self-shielding is included as a bias in the reactivity .
calculation.
1.2.3 POSTULATED ACCIDENTS Most accident conditions will not result in an iricrease in Keff of .
the rack. Examples are the loss of cooling systems (reactivity decreases with decreasing water density) and dropping a fuel
- assembly on top of the rack (the rack structure pertinent for .
criticality is not deformed and the assembly has more than eight inches of water separating it from th'e rest of the rack which pre-cludesinteraction).
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l However, accidents can be postulated which would increase reactivity such as inadvertent drop of an assembly between the outside periphery of the rack and the pool wall. Therefore, for accident conditions, .
the double contingency principle of ANS N16.1-1975 is applied. This states that it shall require two unlikely, independent, concurrent events to produce a criticality accident. Thus for accident conditions, i
the presence of soluble. boron in the storage pool water can be assu'med as a realistic initial condition.
The presence of the approximately 2000 ppm borcn in the pool water will decrease reactivity by more than 30% tk. In perspective, this is more negative reactivity than is present in the poison pl.ste. *
(15% ak) so Keff for the rack would be less than 0.95 even if the poison plates were not present. Thus Keff f_ 0.95 can be easily met for postulated accidents, since any reactivity increase will be much ,
less than the negative worth of the dissolved boron.
For fuel storage applications, water is usually present. However, accidental criticality when fuel assemblies are stored in the dry condition is also accuanted for. For this case, possible sources of
' moderation, such as those that could arise during fire fighting' ,
operations, are includid in the analysis.
This " optimum moderation" accident is not a problem n fuel storage racks. The presence of poison plates removes the co iditions necessary for " optimum moderation" so that Keff continually decreases as moderator density decreases from 1.0 gm/cm3 to 0.0 gm/cm3 in poison rack design.
Figure 1.2-2 shows the behavior of Keff as a function of moderator density for a typical PWR poisoned spent fuel storage rack. .
1.2.4 METHOD FOR CRITICALITY ANALYSIS The calculation method and cross section value,s are verified by
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comparison with' critical experiment data for assemblies similar to those for which the racks are designed. This benchtnarking data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include strong neutron absorbers, .*
large water gaps and low moderator densities.
The design method which ensures the criticality saf 2ty of fuel assemblies in the s[ent fuel storage rack uses the AMPX system of codes [1,2) for crosg-section generation and KENO-IV[3] for reactivity determination. The 218 energy group' cross-section librarybl) that is ,.
the conynon starting point for all cross-sections used for the bench-marks and the storage rack is generated from ENDF/B-IV data. The NITAWL program [2] includes, in this library, the self-shielded resonance cross-sections that are appropriate for each particular geometry. The
- Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM program [2] which is a ,
- one-dimensional S transport theory code. These multi-group cross-n section sets are then used as input to KENO-IV[3] which is a three-dimensional fionte Carlo theory program designed for reactivity calculations.
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A set of 27 critical experiments has been analyzed using the above .
method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (Boral, steel and water) that simulate LWR' fuel shipping and storage conditions [4,5] to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials E0 (Plexiglass, steel and air) that demonstrate the wide range of applicability of the method.
The results and some descriptive facts about each of the 27. benchmark critical experiments are given in Table 1.2-1. The average Keff of the benchmarks is 0.9998 which; demonstrates that there is no bias associated with the method. The standard deviation of the Keff
.' q ** *
., .. values is 0.0057 ok. The 95/95 one sided tolerance limit factor for 27 values is 2.26. Thus, there is a 95 percent probability with a 95 percent confidence level th:t the uncertainty in reactivity, due to the method, is not greater than 0.013 ak.
The total uncertainty (TV) is to be added to a criticality calculation
is:
TU=[(ks)fethod+(ks)fominal+(ks)fech) where (ks) method is 0.013 as discussed above, (ks) nominal is the statistical uncertaint-y associated with the particular KEN 0 calcu-lation being used, (ks) mech is the statistical uncertainty associated with mechanical tolerances, such as thicknesses and spacings within .'
the storage cell. The most important effect on reactivity of the mechanical tolerances is the possible reduction in the water gap between the poison plates of adjacent storage cells. For Sequoyah, the worst combination of mechanical tolerances (i.e., sheet metal thickness, cell I.D. maximum, rack' grid assembly, and cell bowing) will result in a reduction of the water gap between adjacent cells by .189".
The analysis, for the effect of mechanical tolerances assumes a " worst" case of a rack composed of an array of groups of four cans where the ,.
water gap between the four cans is reduced to 0.784 inch (see Figure 1.2-3).
No reactivity increase was observed due to mechanical tolerances, but the uncertainty in the KENO result was used in the determination of the uncertain-ty in the final K,77
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Some mechanical tolerances are not included in the analysis because An worse case assumptions are used in the nominal case analysis.
example of this is eccentric assembly position. Calculations were -
performed which show that the most reactive condition is the tssembly Another centered in the can which is assumed in the nominal case.
No bias it example is the reduced width of the poison plates.
included here since ths nominal KEN 0 case mocels the reduced width * .
explicitly.
The final result of the uncertainty analysis is that the criticality design criteria are met when the calculated eff.cctive multiplication factor, plus the total uncertainty (TV) and any biases. is less than 0.95. .
These methods conform with ANSI Ni8.2-1973, " Nuclear Safety Criteria
- for the Design of Stationary Pressurized Water Reactor Plants",
Section 5.7, Fuel Handling System; ANSI N210-1976, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations",
Section 5.1.12; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety"; NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage"; and the NRC Guidance, "NRC
- Position for Review and Acceptance of Spent Fuel Storage and .
Handling Applications".
1.'2.5 CRITICALITY RESULTS The minimum The spent fuel storage rack is described in Section 2.0.
The sensitivity B
10 loading in the poison plates is 0.0232 gm/cm2 of storage lattice Keff to U-235 enrichment of the fuel assemoly, the storage lattice pitch, and B 10 . loading in the poison plates as requested by the NRC for poison racks is given in Figure 1.2-4 .
For normal operation and using the method described in the above sections, the Keff for the rack is determined in the following r
manner.
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- K,ff =Knominal + Omech + Umethod + dpart *
[(ksnominal) + (ksmech) + (ksmethoc) 3 r
where.
i nominal = . nominal case KENO Keff Bmech
" K eff bias to account for the fact that mechanical tolerances can result in water gaps between poison , ,
plates less than nominal.
Bmethod = - method bias determined from benchmark critical comparisons S
part
= bias to account for poison particle s1 elf-shielding ks nominal
= 95/95 uncertainty in the nominal case KEtt0 Keff ,
ksmech
= 95/95 uncertainty in the calculation due.to Kell 0 analysis of mechanical tolerances ks me: hod
= 95/95 uncertainty in the method bias Substituting calculated values, the results' are the following:
= 0.92304 + 0.0 + 0.0 + .0025 + [(.003859)2 (.004046)2 ,
+ (.013)23 l1/2 := .9397. ,
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is less than 0.95 includinn uncertainties at a 95/95 probability / .
Since Keff confidence level,.the acceotance criteria for criticality is met.
1.2.6 ACCEPTANCE CRITERIA FOR CRITICALITY Tne neutron multiplication facto.r in spent , fuel pools shall be less*
than or equal to 0.95, including all uncertainties, under all conditions. Generally, the acceptance crit'eria for postulated -
accident conditions can be Keff *.0.'98 because of the accuracy of the methods used coupled with the low pr>bability of occurrence.
For instance, in ANSI N210-1976 the acceptance criteria for the Hcwever, f or storage
" optimum moderation" condition is Keff < 0.98. ~
pools, which c'ontain dissolved boron, the use of realistic initial conditions ensures that Kgff <<0.95 for postulated accidents as -
Thus, for simplicit,y, the acceptance discussed in Section 1.2.3.
criteria for all conditions will be Keff < 0.95.
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.j j REFERENCES
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.I.t.
- i. 1. W. E,.,' Ford III, etfal,; "A 218-Group Neutron Cross Sec' tion Library 4
>. in the AMPX Master'.Interf ace Format for Criticali ;y Safety Studies,'
ORNL/CSD/TM-4(July 1976). ~ ' ,
O j, '
25 N. M. Greene, et al, "AMPX: A Modular Code System for Generating Cobpled Multir eutron-Gamma Libraries f rom ENDF/B," ORNL/Ti4-3706 -
(March 1976). ;
.. 3. L. M. Petrie and N. F. Cross, " KENO-IV" - Am Impr)ved Monte Carlo
-f *
- Criticality Program," ORNL-4938 (November 1975).
1
- 4. S. R. Bierman, et al, " Critical Separation Between Subcritical *
, Clusters of 2.35 wt %.235U Enriched U0 2 Rods in hater with Fixed Neutron Poisons," Battelle Pacific Northwest Labcratories PNL-2438 (October'1977).
4
- 5. S. R. Biennan, et al, " Critical Separation Betwer n Subcritical Clusters of 4.29 wt % 235U Enriched UO2 Rods in kater with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2614 (March 1978).
- 6. J. T. Thomas, " Critical Th'ree-Dimensional Arrays of U (93.2) -
Metal Cylinders," Nuclear Science ar.c Engineering,' Volume 52, Pages 350-359 (1973).
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,7 FIGURE 1.2-1 STORAGE CELL inWluAL DINttiSIONS Init SEi}uoVAll RACK ,- . .
P,f1 AIL "A" . .
WRAPPER PLATE (.036")
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w FIGURE 1.2-1 STORAGE CELL '10t1I!;AL DI. M Efts!0fts (C0flTiflUED) t
- DETAIL "A"
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j .b36"
( f 1.050" VOID l
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FIGURE 1.2-2 KgFF VS. .!ATER DENSITY r
FOR A TYPICAL " POISONED" SPElli FUEL STORAGE RACK Type ot Rack C-C, 10.25 Incn
'.0--
Poison Leading, 0.02 gm B 10/cm2 Fuel, 3.5 w/c H 17 x 17 0.9 ~ ~
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e 0.7 - - '
a .
1 0.6 .!
0.5 --
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0.6 0.8 10 0.2 0.4 ,
Moderator Density (gm/cm3)
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t ** '
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d 4--- 9. 4 020" -->
+1.16209 +1.1620". ,
-+ .784" 4--- '
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.784" n
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J, Each square represents one fuel storage cell.
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.,'- AS A FUNCTION OF C-C SPACING KEFF e, . .
POISON LOADING AND ENRICHMENT FIGURE 1.2-4 1.02 --
C-C Spacing
~~ **
1.00 - -
.98 -
K
.96- -
EFF Poison Loading .
Enrichment
.94- - .. __
.92- -
i
.90' - __
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l l .86 ! !
3.5 4.0 4.5
- nrichment (w/o 10.875 9.875 10.375 0-C Spacing Loadings (Inch) 2) .0132 (gm B10/cm .0232 .0332 Notes: For Enrichment Curve, C-C = 10.375", Loading = 0.0232 For Spacing Curve, w/o = 4.0, Loading = 0.0232 For Loading Curve, w/o = 4.0, C-C = 10.375" l
-t i Xi'l kill! l115 [ .,
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Ot f.CillGL.I.k t ilLAt .
t-General Enrichnent Separa t in9 Characterizin9 Seggralion(cm[ Ky g " .'.
Description w/o_t!235__ Reflector ,
, Ma te ri a,1, , _
2.35 water water 11.92 1.004 1 .'0154
- 1. U0 rod lattice 2
= " " "
8.39 0.993 1 .WM-2.
3.
6.39 1.005 i .,004 4.
4.46 0.994 1 .004 * .
5.
stainless steel 10.44 1.005 1 .004 6.
11.47 0.992 i .004 -
7.
7.76 0.092 1 .004 8.
7.42 ').004 i .004 9.
boral 6.34 1.005 1 .004 10.
9.03 0.992 i .004 11.
". 5.05 1.001 ! .004 12.
4.29 "
water 10.64 0.999 ! .005 13.
stainless steel 9.76 0.999 1 .005 14.
8.08 0.998 i .006 15.
" "~ "
boral 6.72 0.998 1 .005 I ti. U metal cylinders 93.2 bare air 15.43 0.993 ! .003 17.
paraffin air 23.84 1.00b i .005 ,
19.97 1. 6,., * .003 18.
bare air 19.
para f fin air 36.4/ 1.001 i .004 .
20.
bare a i n- 13./4 1.005 1 .003 21.
para i fin air 23.40 1.005 i .004 2.' .
hare p l e x t ij l a s., 15.74 1.010 ! .003 <
13.
. paraffin pl ex t iil.'in 24.43 l .00t. 1 .004 24.
bare plexiglass 21. 74 0.999 i .003
- p. ira f f in plexiillass 27.94 0.991 1 .005 2' 5.
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14.74 ,
- 1. 000 .t .003
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ii.i re s isiex tilla . , i t ei 1 1(i. 6 / 0.996 1 .003 4 . .
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JUSTIFICATION FOR PROPOSED TECilNICAL SPECIFICATIONS FOR SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 CilANCE #2 INCREASE OF Tile MAXIMUM FUEL ENRICHMENT FROM 3.5 WEIGilT PERCENT OF U-235 TO 4.5 WEIGHT PERCENT OF U-235 (NEW FUEL) r w
. - = _ _ _ . _ _ _ - _-
. . . l .
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- 1. NEUTR N MULTIPLICATION FACTOR Criticality of f uel assemblies in the new f uel stcrage rack is prevented by the design of the rack which limits fuel assembly interaction. This is_done by fixing the minimum separation between assemblies to take advantage of neutron absorption in water and stainless steel.
The design basis for preventing criticality outside the reactor is that, including uncertaintic't, there is a 95 percent probability at a 95 per-cent confidence level that the effective multiplication f actor (Keff) of the fuel assembly array will be less than 0.98 as recommenced in ANSI N18.2-1973.
The following are the conditions that are assumed in meeting this design basis for the Sequoyah new fuel storage racks.
- 2. NORMAL STORAGE
- a. The fuel assembly contains the highest enrichment authorized without ,
any control rods or any noncontained burnable poison and is at its most reactive point in life. The enrichment of the 17x17 Westing-house standard fuel assembly is 4.5 w/o U-235 with no depletion or fission product buildup. The assembly is conservatively modeled with the assembly grid volume removed and no U-234 and U-236 in the F
fuel pellet.
- b. The array is either infinite in lateral extent or is surrounced by a conservatively chosen reflector, whichever is appropriate for the design. The nominal case calculation is infinite in lateral and axial extent. Calculations show that the finite rack is less reactive than the nominal case infinite rack. Therefore, the nominal case of an infinite array of cells is a conservative assumption.
1603F:6
+- . .-
. . . . .r
- c. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using " worst case" condi-tions or by performing sensitivity studies to obtain the appropriate values. The items included in the analysis are:
-- stainless steel thickness
-- cell 10
-- center-to-center spacing
-- asymmetric assembly position The calculation method uncertainty and bias is discussed in Section 4.
- d. Credit is taken for the neutron absorption in full length stainless steel structural material.
- 3. POSTULATED ACCIDENTS Most accident conditions will not result in an increase in t ggf of the rack. An example is the dropping of a fuel assembly on top of the rack (the rack structure pertinent f or criticality is not deformed and the assembly has more than eight inches separating it from the active fuel in the rest of the rack which precludes interaction).
However, accidents can be postulated (under flooded conditions) which would increase reactivity such as inadvertent crop of an as embly between the outside periphery of the rack and pool wall. Therefore, for accident conditions, the double contigency principle of ANS N16.1-1975 is applied. This states that it is unnecessary to assume two unlikely, independent, concurrent even'; to ensure protection against a critica-lity accident. Thus, for accident conditions, the absence of water in the storage pool can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.
1603F:6
- e. -
The absence of water in the storage pool guarantees suberiticality for enrichmentslessthan5w/obl3 Thus any postulated accidents other than the introduction of water into the storage area will not preclude the pool from meeting the eK rr 3E 0 98 timit.
Because the most limiting accident is the introduction of moderation into {
l the storage pool, this accident will be considered in determining the maximume K rr for the storage pool. For this accident, possibler sources of moderation, such as those that could arise during fire flighting operations, are included in the analysis. This " optimum moderation" l
accident is not a problem in new fuel storage racks because physically achievable water densities (caused, for instance, by sprinklers, foam generators or fog nozzles) are considerably too low ((<0.01 gm/cm3) to yield Kerr values higher than full density water. The optimum achievable moderation occurs with water at 1.0 gm/cm3 Preferential water density t
reduction between cella (i.e., boiling between co11s) is prevented by the rack design. Also, the Tennessee Valley Authority (TVA) h:ts instituted administrative controls which will prevent the use of water fog or spray to combat a fire in the new fuel pool. These controls include the prohibition of aqueous foam fire extinguisher equipment from the refueling floor and the use of metal covers over the new fuel pool when fuel handling operations are not being performed. Although the covers are primarily for the control of dust and foreign objects, they will also prevent the entry of aqueous foam or mist. These administrative controls were instituted before the receipt of the Special Nuclear Materials License for Sequoyah -
NuclearPlantunit1inMarch1977[8]. In the Safety Evaluation Report 1603F:6
-= -
-rtcowase.c w.m._ e m m _-f m _ x . _ , . % _,
- g. , , ,
r,. . . . . . .. .. ...
which accompanied this license, NRC stated that TVA "has established reasonable and satisfactory precautions to avoid accidental criticality."
l s
- 4. METHOD FOR CRITICALITY ANALYSIS
)
The calculation method and cross-section values are verified by comparison with critical experiment data for assemblies similar to those for which the 1
racks are designed. This benchmarking data is sufficiently diverse to i eatablish that the method bias and uncertanity will apply to rack conditions which include strong neutron absorbers, large water gr. 'd low l moderator densities.
The design method which ensures the criticality safety of fuel assemblies in the spent fuel storage rack uses the f.MPX system of codes [2,33 for cross-section generation and KENO IV for reactivity determination.
The218energygroupcross-sectionlibrary[23 thit is the common starting point for all cross-sections used for th2 benchmarks and the storage rack is generated from ENDF/B-IV data. T:le NITAWL program adds to this library the shelf-shielded resonance cross-s3ctions that are
=1603F:6
r-e -
appropriate for each particular geometry. The Nordheim Integral Treat-ment is used. The 218 groups are reduced to 19 groups by energy and spatial weighting of cross-sections using the XSDRNPM[3] program which is a one-dimensional S transport theory code. These multi-group n
cross-section sets are then used as input to KEN 0 IV b3 which is a three-dimensional Monte Carlo theory program designed for reactivity calculations.
I A set of 27 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability. The experiments range from water moderated, oxide fuel arrays separated by various materials (Boral, steel and water) that simulate LWR fuel shipping and storage conditions [5,6] to dry, harder spectrum uranium metal cylinder arrays with various interspersed materials E73 (Plexiglas, steel and air) that demonstrate the wide range of applicability of the method.
The results and some descriptive facts about each of the 27 benchmark crilical experiments are given in Table 1. The aver age K eff of the benchmarks is 0.9998 which demonstrates that there is no bias associated with the method. The standard deviation of the K eff values is 0.0057 t.k. The 95/95 one sided tolerance limit f actor for 27 values is *!.26. Thus, there is a 95 percent probability with a 95 percent con-fidence level that the uncertainty in reactivity, due to the method, is not greater than 0.013ak.
The total uncertainty to be added to a criticality calculation is:
TU = [(ks),2ethod+@s(ominal where'(ks) method is 0.013 as discussed above, (ks) nominal is the statistical uncertainty associated with the particular KENO calculation being used.
1603F:6 l- .. .
.e,' -
The most important effect on reactivity of the mechanical tolerances is the possible reduction in the center-to-center spacing between adjacent assemblies. The nominal gap between adjacent cellt. for Sequoyah is 11.0 inches. The de;ign also guarantees that the average center-to-center storage cell spacing for a module of cells will be 21.0 inches. (See Figure 1). Therefore, any reduction of cell-to-cell gap on one side of a can will produce a gap increase on the opposite side of the can. The KENO model for the gap reduction analysis consists of an infinite array of clusters of 4 cells with the gap between adjacent cells in each cluster reduced to 10.97 inches.
Another center-to-center spacing reduction can se caused by the asym-metric assembly position within the storage cell. The inside dimensions of a nominal storage cell are such that if a fuel assembly is loaded into the corner of the cell, the assembly centerline will be displaced only 0.284 inches f rom the cell centerline. This means that adjacent asymetric fuel assemblies would have their center-to-center distance reduced by 0.568 inches f rom the nominal.
Analysis shows that the combined effect of the worst mechanical toler-dnces and the asymmetric assembly positioning may increase reactivity by 0.00lak. This will be treated as a bias although the individual deviations will be random.
The final result of the uncertainty analysis is that the criticality design criteria are met when the calculated effective multiplication factor, plus the total uncertainty (TU) and any biases, is less than
- 0. 9 8. .
These methods conform with ANSI N18.2-19/3, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", Section 5.7, Fuel Handling System; ANSI N16.9-1975, " Validation of Calculational Methods for Nuclear Criticality Safety".
1603F:6
- *=g , a + , ,
- 4 , M% . Q ,.
- t { ,
b
- 5. CRITICALITY ANALYSIS FOR RACK DESIGN For normal operation and using the method in the above section, the
- K,7f for the rack is determined in the following mar ner.
eff " nominal + Omech
- 0 method *
[(ks) nominal + .(ks)n ethod Where:
=
K nominal n minal case KEN 0 Keff B
eff as to account for W f ad mat mechanical mech tolerances can result in spacings tetween assemblies less than nominal B = method bias determined f rom benchmark critical compari-method sons ks = 95/95. uncertainty in the nominal czse KENO Keff n W nal ks = 95/95 uncertainty in the method bi.is tnethod Substituting calculated values in the order listed above, the result is:
.g K,ff = 0.9189 + 0.0010 + 0.0 + [(.0062)2 + (.013)2 l/2 = .9343 Since K,ff is less than 0.98 including uncertainties at.a 95/95 proba-
-bility/ confidence level, the acceptance criteria for criticality is met.
1603F:6 .
m
- .**. t ,; .
ee a e REFERENQES
- 1. J. T. Thomas, " Nuclear Safety Guide," NUREG/CR-0095 Rev. 2 (June 1978).
- 2. W.E. Ford III, et al, "A 218-Group Neutron Cross-Section Library in the AMPX Master Interf ace Format for Criticality Safety Studies,"
ORNL/CSD/TM-4 (July 1976).
- 3. N.M. Green, et al, "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B,"
ORNL/TM-3706 (March 1976).
- 4. L.M. Petrie and N.F. Cross, " KENO IV-An Impro/ed Monte Carlo Criti-cality Program," ORNL-4938 (November 1978).
- 5. S.R. Bierman, et al, " Critical Separation Between Subtritical Clusters of 2.35 wt % 235 U02 Enriched UO Rods in Water with 2
Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2438 (October 1977).
- 6. S.R. Bierman, et al, " Critical Separation Between Subcritical
. Clusters of 4.29 wt % 235 00 Enriched UO Rods in Water with 2 2 fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2614 (March 1978).
- 7. J.T. Thomas, " Critical Three-Dimensional Arrays of U (93.2) - Metal Cylinders," Nuclear Science and Engineering, Volume 52, pages 350-359.(1973).
- 8. Letter from L. C. Rouse, Chief, Fuel Processi:;g and Fabrication Branch, Division of Fuel Cycle and Material Safety, NRC, to Tennessee Valley Authority, ATTN: Mr. J. E. Gilleland dated
, March 25, 1977. Docket No: 70-2443.
l 1603F:6-
~: . . - ~_ , _ . , , , _ . _ _ fp _- ..
g __
TA8LE I BENOfiARKCMisCALEXPERIMENTS[4.5.6]-
l-Separating Characterizing -
General Enrictment Reflector Material Separation (c=) K Descriptten w/o U235 water water 11.92 1.004 + .004
- 1. UO rod lattice 2.35 2
- 6.39
- *
- 0.993 1 004
- 2.
- * *
- 6.39 1.005 + .004 3,
l * " *
- 4.46 0.s H + .004 4.
(6 - * *
- stainless steel 10.44 1.005 + .004 5.
- * *
- 11.47 0.992 1 004 6.
l * * * *
- 7.76 0.992 -+ .004
- 1, 4 * * *
- 7.42 1.004 f .004 8.
- *
- boral 6.34 1.005 1 004 9.
- * *
- 9.03 0.992 -+ .004 10.
- * *
- 5.05 1.001 1 004 11.
- - water .10.64 0.999 1 005-1 12.
- 4.29 0.999 + .005 l
- *
- stainless steel 9.76
- 13. l
- * * " 8.08 0.999 1 006 14.
- *
- boral 6.12 0.998 + .005 '
15.
bare air 15.43 0.998 2 003
- 16. U enetal cylinders 93.2
- * - paraffin air 23.84 1.006 + .005
. 11. j
- bare air 19.97 1.005 2 003 18.
19.
- paraffin air 36.47 1.001 1 004
- bare air 13.74 1.005 1 003 20.
21.
"
- paraffin air 23.48 - 1.005 1 004
- bare pleatglass 15.14 1.010 2 003 22.
" " paraffin- plemiglass 24.43 1.006 + .004 23.
- *
- bare plexiglass 21.74 0.999 + .003 24.
- paraffin plestglass 27.94 0.994 1 005 25.
- - a bare steel 14.74 1.000 1 003 26.
- bare plealglass. steel 16.67 0.996 1 003 27.
9.
e= tt 3 e
=
g e 9 .
a -
- ,, *t- a , 4 oo se FIGURE 1 STRUCTURE BARS INTERMEDIATELY SPACED ANGLE IRONS (FULL LENGTH)
(NOT INCLUDED IN KENO .TDEL)
REFLECTIVE p- _ _ _ _ ._ _ ._ . _
y I ,) ,,," _
n 2.0" I ' '-
hp #
Q t -
p n g _.
g g
f FUEL ASSEMBLY l E- 17 x 17 W STD. l 8.432" 9.0" 21.0" 0.25" 3 I - I l -
- l 9 '
l l 0.25"A s f - , c,,
.. s
- u n a a i; 4- -
L - - _ __ _ _ _ _ al .s e
. . . ..