ML19323B886

Proposed Tech Spec Changes to Sections 5.5-5.8 for Design Features.Criticality Analysis Encl

ML19323B886
Person / Time
Site:	Millstone
Issue date:	05/09/1980
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML19323B879	List: ML19323B878 ML19323B886
References
TAC-8947, NUDOCS 8005140380
Download: ML19323B886 (11)

Text

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8005140 % O O DOCKET NO. 50-336 ATTACHMENT 1 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 PROPOSED TECHNICAL SPECIFICATIONS i

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8 MAY, 1980

June 30, 1977 DESIGN FEATURES

'l',VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,060 + 700/-0 cubic feet.

5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for nomal degradation pursuant to the applicable Surveillance Requiremerts.

5.6 FUEL STORAGE CRITICALITY 5.6.1 The new and spent fuel storage racks are designed and shall be maintained with sufficient center-to-center distance between assen611es to ensure a keff < 0.95 with the storage pool filled with unborated water. The maximum fuel assembly. enrichment to,be stored .in_.the_ fuel pool will be 3.35 w/o.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to ,

prevent inadvertent draining of the pool below elevation 22'6". l s

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5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I Items in Section 5.1.1 of the FSAR shall be designed and maintained to the original design provisions contained in Section 5.8 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.8 METEOROLOGICAL TOWER LOCATION j 5.8.1 The meteorological tower location shall be as shown on Figure 5.1-1.

X MILLSTONE-UNIT 2 5-5

DOCKET NO. 50-336 t-f l

ATTACHMENT 2 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 CRITICALITY ANALYSIS FOR MILLSIONE UNIT NO. 2 SPENT FUEL STORAGE POOL MAY, 1980 l

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NUCLEAR CONSIDERATIONS 1.0 NEUTRON HULTIpLICATION FACT 0!1 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 to take advantage ,

of neutron absorption in water and stainless steel.

The design basis for preventing cr.iticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (K,ff) of the

, fuel assembly array will be less than 0.95 as recomended in ANSI N210-1976 and in "HRC position for Review and Acceptance of Spent Fuel Storage and

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Handling Spplications".

The following are the conditions that are assumed in meeting this design basis. .

1 1.1 NORMAL STORAGE

a. Th fuel assembly contains the highgt enrichment authorized without any control rods or any noncontained burnable poison and is at its most reactive point in life. The enrichment of the 14 X 14 Westing- 'l house fuel assembly is 3.35 w/o U-235 with no depletion or fission product buildup. The assembly is conservatively modeled with water replacing the assembly grid volume and no U-234 and U-236 in the fuel pellet.

b. The moderator is pure water at the temperature within the design limits of the pool which yields the largest reactivity. A

' 3 i conservative value of 1.0 gm/cm is used for the density of water.

No dissolved boron is included in the water. The nominal center-to-center spacing is 12.19 inches.

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c. The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design. The nominal case calculation is infinite in lateral and axial extent. Celculations show that the finite rack surrounded by a water reflector is less reactive than the nominal case infinite rack. Therefore, the nominal case of im infinite array of cells is a conservative assumption.

d. Mechanical uncertainties and biases due to mechanical tolerances during construction are treated by either using " worst case" conditions or by performir.g sens'itivity studies to obtain the appropriate values. The items included in the analysis are:

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- stainless steel thickness

- can ID

- center-to-center spacing ,

- asymmetric assembly position ,

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The calculational method uncertainty and bias is discussed in Section 1.3.

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e. Credit is taken for the neutron absorption'6nly in the full length stainless steel box wall- structure.

1.2 POSTULATED ACCIDENTS 1

Most accident conditions will not result in an increase in K,ff 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 i structure pertinent for criticality is not deformed and the assembly has more than eight inches of water separating it from the rest of the rack which precludesinteraction). - -

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However, accidents can be postulated which could increase the reactivity in the spent fuel storage pool such as an inadvertent drop of a fuel assembly between the outside periphery of the spent fuel rack and the spent fuel pool wall. The design of the Millstone Unit No. 2 spent fuel pool and associated equipment assures that the closest physically achievable approach of a fuel assembly to the side of the spent fuel rack will not increase the multiplication factor of the storage rack.

For fuel storage applications, water is usually present. However, accidental criticality when fuel assemblies are stored in the dry condition is also accounted for. For this case, possible sources of moderation, such as those ,

that could arise during fire fighting operations, are included in the analysis.

This " optimum moderation" accident is not a problem in fuel storage racks' 3

because possible water densities are too loa (10.01 gm/cm ) to yield K,ff values higher than for full density water and the rack design prevents the preferential reduction of water density between the cells of a rack (e.g.,

boilingbetweencells).

1.3 CRITICALITY ANALYSIS i

The calculational method and cross-section values are verified by comparison l with critical experiment. data for assemblies simila. to those for which the racks are designed. This benchmarking data is suff'iciently 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. .

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The design method which ensures the criticality safety of fuel assemblies in the spent fuel storage rack uses the AMPX system of codesD ,2] g;. cross- '

section generation and Kl..T I? E33 for reactivity detennina' tion.

The 218 energy group cross-section libraryO3 tha~t is the common starting point for all cross-sections used for the benchmarks 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.

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Energy and spatial weighting of cross-sections is performed by the XSDRNPM program [2] which is a one-dimensional NS transport theory code. These multi-group cross-section sets,are then used as input to KENO IV[3] which' is a three-diraensional Monte Carlo theory program designed for reactivity calculations.

A set of 27 critical experiments has been analyzed using the above method to demonstrate its applicability.to .aticality analysis and to establish the

, method bias anri variability. Theexperimentsrangefromwatermodeinted, oxide j fuel arrays separated by various materials (Boral, steel and water) (t,at l simulate LWR fu61 shipping and storage conditions [4,5] to dry, harder spectrum uranium metal cylinder arrays with various igterspersed materials [6] (Plexiglas.,

steel and air) tht demonstrate the wide rang'e of applicability of the method.

The average Kpff of the benchmurks is 0.9998 which demonstrates that there is no bias associated with the method.

The standard deviation of the K,ff 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 that the uncertainty in reactivity, due to the method, is not greater than 0.013 Ak.

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I The total uncertainty to be added to a criticality calculation is:- .

TU=[(ks)hthod +(ks)fsym

[(ks)2nominal where(ks) method is 0.013 as discussed above. (ks) nominal is the statistical uncertainty associated with the particular KENO calculation teing used, and (ks) is the statistical uncertainty due to asymmetric assembly position The reactivity effect of mechanical tolerances is ignored in this analysis  !

due to the assumption of " worst-case" geometry for the base case K,ff.

For Millstone Unit 2, the worst combination of mechanical tolerances results in a 0.24 inch can thickness and a spacing between cans of 2.625 inches. No bias or uncertainty is therefore applied due to mechanical tolerances.

Another center-to-center spacing reduction can be caused by' asymmetric assembly position within the storage can. The analysis assumes that groups of four assemblies are located within'the storage cans so that the spacing between them is a minimum. This results in a center-to-center spacing reduction of 0.54 inches and a reactivity increase of 0.0105 Ak. This'is conservatively taken as a bias even though the asynitetric position of an assembly within the storage can will be a random event.

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 (TV) and any biases, is le'ss than 0.95.

These methods conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurizad 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, i

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" Validation of Calculational Methods for Nuclear Criticality Safety"; NRC i 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".

1.4 RACK MODIFICATION -

The spent fuel storage rack is described in Reference 7. The worst case geometry assumptions (used to eliminate the sensitivity to mechanical tolerances) are identical to those used in previous spent fuel storage pool ~

criticality aralyses.

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For normal operation and usino the method in the above sections, the Keff for the rack is determin3d in the followina manner. ~

1 eff"bominal*Omethod + B,3y , + U ks)2 nominal +(ks)mthod + (ks),2 sym l

where: , ,

' knominal = n minal case KENO K,ff Beethod = method bias determined from benchmark critical comparisons B,3p = bias to account for asymetric assembly position ,

1 ks nominal = 95/95 uncertainty in the nominal case KENO K,ff I ksmethod = 95/95 uncertainty in the method bias ks asp = 95/95 uncertainty to account for asymetric assembly position l

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, Substituting calculated values, the result is: ,

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K,ff *= .9456 ,

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Since K,ff is less than 0.95 including uncertainties at a 95/95 probability /

confidence level, the acceptance criteria for criticality is met.

The tabular reactivity balance for the criticality analysis of the spent fuel storage rack is shown below. Calculated values of the multiplication factor iabeled nominal, minimum and most adverse correspond to calculations performed assuming, respectively. nominal rack dimensions, the most adverse concurrent combination of dimensional tolerances, ano the most adverse combination with the simultaneous displacement of the fuel assemblies into their most reactive positions within the cans.

l Nominal Minimum Most Adverse

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Multiplication Factor for Spent Fuel Storage Rack .875 .918 .946 Excess Margin .075 .032 .004 1.5 '

ACCEPTANCE CRITERIA FOR CRITICALITY -

For the purposes of this analysis, the acceptance criterion has been chosen to be Keff < 0.95, including all uncertainties and under all conditions.

The results of the analysis demonstrate that Keff for the Millstone Unit No. 2 spent fuel storage pool will meet the acceptance criterion with fuel enriched to 3.35 w/o U-235 for all normal and accident conditions.

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REFERENCES -

1. W. E. Ford III, et at, "A 218-Group Neutron Cross-Section Library in the AMPX Master Interfaca, Format for Criticality Safety Studies," '

ORNL/CSD/Tf t-4 (July 1976).

2. N. M. Greene, et al . " AMP : A Modular Code System for Generatir.g Coupled Pultigroup Neutron-Gamma Libraries from ENDF/B," ORNL/TH-3706 (March 1976).

3. L. St. Petrie and N. F. Cross, " KENO IV-An Improved Monte Carlo Critic 6If t,- Program," ORNL-4938 (November 1975).

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S. R. Bierman, et al, " Critical Separation Between Suberitical Clusters-235 of 2.35 wt % 0 Enriched UO Rods in Water With Fixed Neutron Poisons,"

2 Battelle Pacific Northwest Laboratories PNL-2438 (October 1977).

S. S. R. Bierman, et al. " Critical Separation Between Subcritical Clusters of 4.29 wt % 235U Enriched,UO Rods 'in Water with Fixed Neutron Poisons,"

2 Battelle Pacific Northwest Laboratories PNL-2615 (March 1978).

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6. J. T. Thomas, " Critical Three-Dimensional Arrays of U (93.2) - Natal Cylinders," Nuclear Science and Engineering,# Volume 52, pages 350 - 359 (1973).

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" Millstone Unit No. 2 Spent Fuel Pool Modifications" Docket 50-336, November 22, 1976.

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