ML20140D253

Affidavit of Wa Boyd Addressing Contention 10.Critical Analyses Performed for Spent Fuel Pool Expansion Amends Conform W/Applicable Stds & Criteria.Certificate of Svc Encl

ML20140D253
Person / Time
Site:	Turkey Point
Issue date:	01/20/1986
From:	Boyd W FLORIDA POWER & LIGHT CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20140C819	List: ML20140C815 ML20140C823 ML20140C831 ML20140C844 ML20140C858 ML20140C866 ML20140C878 ML20140D130 ML20140D137 ML20140D155 ML20140D181 ML20140D224 ML20140D233 ML20140D253
References
OLA-2, NUDOCS 8601290236
Download: ML20140D253 (34)

Text

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NUCLEAR REGULATORY COMMISSION [N -

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4 )

In the Matter of )

) Docket Nos. 50-250 OLA-2 FLORIDA POWER & LIGHT COMPANY ) 50-251 OLA-2

)

(Turkey Point Nuclear Generating ) (Spent Fuel Pool Expansion)

Units 3 & 4) )

)

AFFIDAVIT OF WILLIAM A. BOYD ON CONTENTION 10

1. My name is William A. Boyd. I am a Senior Engineer in Nuclear Design for the Nuclear Fuel Division of Westinghouse Electric Corporation. My business address is Westinghouse Electric Corporation, Monroeville Mall Office Building, P.O. Box 3912, Pittsburgh, PA, 15230. A summary of my professional qualifications and experience is attached hereto as Exhibit A, which is incorporated herein by reference.

2. The purpose of my affidavit is to address Contention

10. Contention 10 and the bases for Contention 10 are as follows:

Contention 10 That the increase of the spent fuel pool capacity, which includes fuel rods that are more highly enriched, will cause the requirements of ANSI NI6-1975 (sic} not to be met and will increase the 8601290236 860123 PDR ADOCK 05000250 G PDR

- . . . _ _ . ~ . _ _ _ _ _ _ . - _. _ ._ . . _ _ _ _ _ _ _ _ _ . _ _ - - . - - - -

probability that a criticality accident will occur

in the spent fuel pool and will exceed 10 C.F.R. Part 50, A 62 criterion.

Bases for Contention The increase in the number of fuel rods stored and the fact that many of them may be more highly enriched and have more reactivity will increase the chances that the fuel pool will go critical, and cause a major criticality accident, and perhaps explosion, that will release large amounts of radioactivity to the environment in excess of the 10 C.F.R. 100 criteria.

In particular, my affidavit demonstrates that the Turkey Point criticality analyses performed by Westinghouse for the Spent Fuel Pool Expansion Amendments conform with applicable industry standards, employ NRC approved methods, and provide results that meet NRC criteria.

3. This affidavit is divided into three primary parts.

l The first part discusses general principles of nuclear physics in order to provide a basis for understanding the criticality analyses performed for the Turkey Point spent fuel pool expan-sions. The second part provides a description of the provisions

for storing spent fuel assemblies as authorized by the Turkey Point spent fuel pool expansion amendments. The final part discusses the criticality analyses performed for the Turkey Point spent fuel pool expansions and shows that the results of those

! analyses are acceptable.

l i

1

i .

i I. General Princioles of Nuclear Physics

4. The primary fissile material in new fuel assemblies for a nuclear power reactor such as Turkey Point is an isotope of uranium called Uranium-235. Uranium-235 comprises less than 1%

of all naturally occuring Uranium. However, fuel assemblies for nuclear power reactors generally contain " enriched" Uranium, 12g., uranium which contains a greater percent of Uranium-235 than occurs naturally.

5. In general, when a neutron is absorbed by Uranium-235, there is an 80 percent probability, approximately, that the Uranium-235 will undergo fission, which results in the release of additional neutrons, and a 20 percent probability approximately, that the Uranium-235 will capture the absorbed neutron, creating l

Uranium-236 (a non-fissionable neutron absorber). In turn, these additional neutrons either can be absorbed by other Uranium-235 (producing additional fission), or can be absorbed by non-fissile material (resulting in no additional fission), or can escape without being absorbed (also resulting in no additional fission).

6. As is apparent from the preceding paragraph, not all neutrons released as a result of fission will cause additional fission. If fewer neutrons are being produced as a result of

(

fission than are escaping and being absorbed, the fission reaction will not sustain itself, and the condition is classified as being "subcritical". In contrast, if an equal or greater number of neutrons are being produced as a result of fission than are escaping and being absorbed, then the fission reaction will

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sustain itself and, in the latter case, increase in intensity.

These conditions are referred to as " critical" and " super-

, critical", respectively.

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7. The term " effective multiplication factor", designated

by the symbol k,gg and commonly called k-effective, has been devised as a measure of the ability of a fission reaction to sustain itself. K-effective is defined as the ratio of the number of neutrons per unit time resulting from fissions to the total number of neutrons lost per unit of time by absorption and leakage. If the value of k-effective is 1.0, a condition of criticality is attained and a self-sustaining chain reaction is possible, because at least as many neutrons are produced in fission as are lost by capture and leakage. If the condition is such that k-effective is greater than 1.0, the condition is supercritical; lig., more neutrons would be produced than are i lost, so that the neutron population, the fission rate, and hence the power generated in the fuel assemblies would all increase continuously with time. If k-effective is less than 1.0, the 1 system is said to be subcritical and the fission chain reaction would not sustain itself.

8. Changes in k-effective may be produced by several i

different methods. In general, k-effective can be increased by increasing the number of neutrons being absorbed by Uranium-235

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relative to the total number produced (by increasing the enrich-I ment of Uranium-235 or by increasing the storage density of the i

fuel assemblies) or by decreasing the number of neutrons escaping j

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o l or being absorbed by non-fissile material relative to the total number of neutrons produced (by decreasing the concentration of neutron absorbers called " poisons"). Conversely, k-effective can be decreased by performing the opposite actions identified above.

It is possible to maintain k-effective at a designated value by taking actions which decrease k-effective to counteract any actions which increase k-effective (g2g., an increase in k-effective caused by an increase in fuel enrichment can be negated by an equivalent decrease in k-effective caused by an increase in neutron absorbers).

9. In addition, it should be noted that, given a group of I

fuel assemblies with the same initial fuel enrichment of Uranium-235, the " reactivity" or k-effective of the new fuel j assemblies is greater than their reactivity after they are irradiated. During irradiation of a fuel assembly, the Uranium-235 in the fuel undergoes "burnup" which depletes the amount of l

Uranium-235 in the fuel and creates fission-product poisons.

Thus, an irradiated or spent fuel assembly will have a lower i effective enrichment and a greater concentration of poisons, and therefore a lower reactivity, than the same assembly which has not been irradiated.

II. Provisions for Storina Soent Fuel Assemblies

10. The purpose of the spent fuel pool expansion amendments for Turkey Point is to increase the amount of spent fuel that can 4

be stored in the existing spent fuel pools. The amendments

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authorize the replacement of the pre-existing spent fuel storage racks with new storage racks which can store spent fuel assem-blies in a higher density array, thereby increasing the number of spent fuel assemblies that can be stored in the spent fuel pools.

! Additionally, the new storage racks have been designed to accomodate the more highly enriched fuel which is now authorized for use at Turkey Point.

11. In order to permit storage of fuel assemblies with different fuel enrichments and fuel burnups, the spent fuel pools are divided into two regions which differ in design. Each region consists of new storage rec' c which provide a different high density fuel storage configuration and a different amount of neutron absorber than the racks in the other region. Figure 1 depicts the arrangement of the new storage racks within the spent fuel pools and identifies which of the storage racks are located in each of the two regions within the pools. Cross sectional views of the Region 1 and 2 fuel rack cells are shown schemati-cally in Figures 2 and 3, respectively.

12. The Region 1 fuel racks have sufficient capacity to permit the storage of 286 fuel assemblies, which is equivalent to about one and a half full cores. The Region 1 fuel racks are intended primarily for the storage of a full core of fuel assemblies removed from the reactor vessel, if necessary, shortly after a plant refueling when some of the fuel assemblies might have experienced very low Megawatt Day / Metric Ton (MWD /MTU) burnups. For this reason, the Region 1 fuel racks have been

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j designed to permit the storage of new, unirradiated fuel assem-blies with an enrichment of 4.5% of Uranium-235 and to accomodate j the reactivitly effects of this enrichment.

13. The Region 2 fuel racks have sufficient capacity to permit the storage of 1118 fuel assemblies. The Region 2 fuel racks are intended for the storage of irradiated fuel assemblies i

for the anticipated forty year life of the plant. Fuel assem-

! blies with an initial U-235 fuel enrichment of 4.5% can be stored

in Region 2 if their burnups are 39,000 MWD /MTU or higher (a fuel assembly with an initial enrichment of 4.5% and a burnup of 39,000 MWD /MTU has a reactivity equivalent to a fuel assembly j with an initial enrichment of 1.5% and zero burnup). Addition-ally, the Region 2 racks are designed to store fuel assemblies

! with lesser burnups and lower initial enrichment, as long as the

combination of burnup and initial enrichment in the assembly has a peak reactivity in the rack equivalent to or less than an 1.5% ,

I enrichment assembly at zero burnup. Finally, during the interim i period of installation of the new racks, the Region 2 racks are

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i designed to accomodate storage of fuel assemblies with a zero I 1

j burnup enrichment up to 4.5% as long as the assemblies are stored in a checkerboard pattern (itg., with every other adjoining cell in the storage rack remaining empty).

1 j .

14. The new spent fuel storage racks permit the storage of more highly enriched fuel assemblies in a denser array than the pre-existing racks. To counterbalance the reactivity effects of these changes, the new storage racks include Boraflex which i

4

4 contains Boron-10, a neutron absorber. Similarly, the Region 1 racks can accomodate fuel assemblies with lesser burnups (and thus higher reactivity) than the Region 2 racks. To counter-balance the reactivity effects of these differences, the Region 1 4 racks contain a less dense array than the Region 2 racks (10.6 inch versus a 9.0 inch center-to-center spacing between adjoining fuel assemblies, respectively) and contain more neutron poison than the Region 2 racks (Boron-10 area density of 0.020 gm/cm 2 2

versus 0.012 gm/cm , respectively).

i i

III. Criticality Analyses for the Turkey Point Soent Fuel Pool Exnansions A. Anolicable Criteria l

15. Criticality analyses for spent fuel pools are governed by General Design Criterion (GDC) 62 of Appendix A to 10 CFR Part 50, which states that " Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferrably by use of geometrically safe confi-gurations."

j 16. Guidance for preventing criticality in spent fuel pools is provided by NRC Standard Review Plan (SRP) Section 9.1.2, t paragraph III.2.a (Ref. 1); the NRC's "OT Position for Review and i

Acceptance of Spent Fuel Storage and Handling Applications"

- (April 14, 1978),Section III.1.5 (Ref. 2); and American National l

Standards Institute (ANSI) N210-76, Section 5.1.12.1 (Ref. 3).

This ANSI and NRC guidance states that the effective neutron multiplication factor in spent fuel pools should be maintained at i

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a value less than or equal to 0.95, including all uncertainties, ,

under both normal and accident conditions. This recommendation is referred to as the 0.95 criterion in the remainder of this affidavit.

17. The design basis for preventing criticality in the Turkey Point spent fuel storage pools is that, including uncer-tainties, there is a 95% probability at a 95% confidence level that k-effective of the fuel assemblies in the pools will be less than 0.95. Thus, the design basis of the Turkey Point spent fuel pool racks satisfies the 0.95 criterion provided in the NRC and ANSI guidance. In this regard, it should be noted that Turkey Point utilized the 0.95 criterion prior to issuance of the licensing amendments authorizing the spent fuel pool expansions.

Therefore, the amendments did not modify or increase the design basis k-effective limit for the Turkey Point spent fuel pools, and thus the amendments will not increase the probability of a l criticality accident.

18. In general, exceeding the design basis k-effective limit for the Turkey Point spent fuel pools is prevented by the design of the spent fuel storage racks. The design of the racks is such that the minimum separation distances between the stored fuel assemblies is fixed (thereby limiting their reactivity), and '

i includes nJutran absorbing Boraflex inserted between the assem-blies (which reduces their reactivity). As is demonstrated l

. 1 1

below, these design features ensure that the k-effective of the Region 1 and 2 storage racks will be less than 0.95 at a 95/95 probability / confidence level, including all uncertainties.

t

! B. Analytical Methods and Assumptions Employed

! for Calculatina k-effective y 19. Three computer programs were primarily employed to l calculate k-effective for the different initial enrichments and

burnups anticipated for fuel stored in the Turkey Point spent fuel pool. The actual criticality calculations were performed i

with KENO-IV (Ref. 4), which is widely used in the nucle.r

)i industry for the purpose of calculating the criticality of fuel racks, critical assemblies, and reactor cores. PHOENIX (Ref. 5) was used to calculate the isotopic compositions of the fuel as a function of irradiation history. CINDER (Ref. 6) was used to calculate the decay of fission products in the fuel and their l neutron capture effects during fuel storage. l l

20. The criticality calculation methods and cross section i a ,

j values for use in KENO-IV have been verified by comparison with I criticality experiment data for fuel assemblies similar to those l

l for which the fuel racks were designed. A set of 27 criticality i

experiments has been analyzed using KENO-IV to demonstrate its applicability to criticality analysis and to establish its method bias and uncertainties. The experiments range from water-i moderated oxide fuel arrays separated by various materials (Boral, steel, water) that simulate light water reactor (LWR)

fuel shipping and storage conditions (Refs. 7, 8), to dry, harder l

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t (higher energy) neutron spectra in uranium metal cylinder arrays with interspersed materials (Ref. 9). This benchmarking data is sufficiently diverse to establish that the bias and uncertainties of the method in KENO-IV apply to the conditions in the Turkey Point spent fuel pools. The average value of k-effective calculated using KENO-IV for these benchmark criticality arrays is 0.9998 which, when compared to the experimentally established k-effective value of 1.0000 for the arrays, demonstrates that there is no significant bias (systematic error) associated with the KENO-IV method when compared to measured values of k-effective. Furthermore, calculations using the benchmark data i

demonstrate that there is a 95% probability at the 95% confidence l level that the uncertainty in values of k-effective calculated using the method in KENO-IV is not greater than 1.3%. As is discussed below, this uncertainty has been accounted for in calculating the k-effective for the spent fuel storage racks for Turkey Point.

21. The accuracy of burnup dependent isotopic predictions calculated by PHOENIX has also been demonstrated by comparison with measurements obtained from fuel samples taken from the core l of the Yankee reactor (Ref. 10). These samples encompass the pellet size and enrichment of the fuel proposed for storage in the Turkey Point racks. The differences between the predicted and measured data are small, and the uncertainties associated j l

with the predictions are included in the final k-effective j factors calculated for the racks. The small difference observed l l

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4 between the measurements and predictions not only verifies the accuracy of the isotopic calculational methods in PHOENIX, they also verify the accuracy of the cross sections employed for the isotopes, and provide assurance that their reactivity worth will be accurately evaluated.

22. The CINDER computer program has been used widely in the nuclear industry for over 20 years. It has been well benchmarked by many sources and is accepted by the NRC. CINDER calculates I

j the production of fission products during irradiation and their decay with time after the fuel is discharged from the reactor.

I

! CINDER was used to evaluate fission-product decay and to examine I

the reactivity of spent fuel assemblies as a function of time

during pool storage. The time dependent concentrations of the f two fissian products with the highest neutron capture cross 4

sections, Xenon-135 and Samarium-149, are responsible for most of the variation in fuel assembly reactivity as a function of time during storage. Due to this variation in concentrations with 4

l time, the maximum fuel assembly reactivity occurs at approxi-i

! mately 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown, which is the design basis stated in the Spent Fuel Storage Facility Modification Safety Analysis Report for Turkey Point Units 3 and 4 (March 14, 1984), p.3-6. It was found that by setting the concentration of  ;

4 Xenon-135 to zero and by holding the concentration of Samarium- l 1

149 at its shutdown value, a conservatively high fuel assembly l

reactivity is obtained at the time of reactor shutdown which is  ;

greater than the fuel reactivity at any subsequent time during h

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storage (and which is conservatively higher than the maximum, design basis values at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown). These assump-tions are conservative because, in effect, no credit is taken for the neutron capture in Xenon-135, or for the increase in concen-tration with time after shutdown of neutron absorbing Samarium-149.

23. Specific analyses were performed for the fuel to be stored in both regions of the Turkey Point spent fuel pools.

These analyses employed the following common conservative assumptions:

a. Criticality was calculated assuming that the array of the particular type of fuel assemblies to be analyzed is infinite in lateral and axial extent. The value of k-effective is higher for an infinite array than for a finite array (such as the arrays in the spent fuel pool), because no credit is taken for loss of reactivity due to leakage of neutrons.

b. As described in paragraph 22, the reactivities of ,

the fuel assemblies were evaluated at the time of reactor shutdown with conservative assumptions concerning the l

concentrations of the neutron absorbing fission products Xenon-135 and Samarium-149, which provide assurance that the calculated values of k-effective for the fuel assemblies are conservatively higher than their k-effective values at any other time in their subsequent fuel pool storage.

c. No credit is taken for any neutron capture by Inconel spacer grids or Zircaloy spacer sleeves, which also leads to a higher value for k-effective.

d. No credit is taken for the presence of neutron absorbing boron in the spent fuel pool water, which assures that the calculated k-effectives are conservatively high.

e. The spent fuel pool water is assumed to have a density of 1.0 grams per cubic centimeter and a temperature of 68 degrees Fahrenheit. Because of the presence of the Boraflex neutron absorbing material in the Turkey Point spent fuel racks, the condition for " optimum moderation" (the water density corresponding to the maximum obtainable value of k-effective) is different than it is for fuel racks 5

that do not contain neutron absorbing material. In the Turkey Point spent fuel pool racks, any water density lower than the assumed value of 1.0 grams per cubic centimeter, which might be caused for example by a temperature increase, will result in a lower value of k-effective. Therefore the condition for optimum moderation in the Turkey Point spent fuel pool racks occurs at the water density assumed in the analysis, 1.0 grams per cubic centimeter.

24. The assumptions and unalytical methods discussed above conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section f

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 Reveiw Plan, Section 9.1.12; and the NRC guidance, "OT Postilon for Review and Acceptance of Spent Fuel Storage and Handling Applica-tions."

C. Realon 1 Criticality Analyses for Fresh Fuel

25. The Region 1 spent fuel racks were analyzed for the highest k-effective condition, which occurs when every available storage location is occupied by a fresh fuel assembly. In addition to the common assumptions I have described above, the criticality analyses for the Region 1 spent fuel racks employed the following specific assumptions.

a. All fuel rods contain uranium dioxide at the maximum authorized enrichment of 4.5 weight percent Uranium-235 over the assumed infinite length of each rod.

b. No credit is taken for any burnable poison in the fuel rods.

c. No credit is taken for any Uranium-234 or Uranium-236 (which are poisons) in the fuel, nor is any credit taken for the buildup of fission-product poison material (none are present in fresh fuel).

These assumptions will cause the calculated k-effective to be conservatively high.

The analysis was begun by calculating k-nominal, which is the reactivity obtained when nominal values are employed for the significant input parameters. The reactivity effects contributed by biases and by the uncertainties of the significant input parameters were then evaluated, combined employing standard statistical methods for the treatment of biases and uncertain-ties, and their net reactivity contribution was then added to k-nominal to obtain a final value of k-effective and its uncertain-ty. The net k-effective and its uncertainty obtained by this process were then compared with the 0.95 criterion and its 95/95 probability / confidence requirements.

Some of the biases were taken etre of directly in the k-nominal calculation by assuming worst case values. For exam-ple, since calculations show that the most reactive conditions occur when the fuel assemblies are centered in their storage cells, the k-nominal calculation assumed that the assemblies were in fact centered. Other examples are the width and thickness of the Boraflex poison material, which were assumed to have their minimum, worst case values in the nominal calculation. No additional bias was added to account for these effects, since they are explicitly accounted for in the k-nominal etic;1ation.

The other input parameters whose reactivity 0'4ttribu-tions were evaluated separately, statistically combitah 'in the uncertainty analysis and added to k-nominal, include the material and mechanical construction tolerances of the sneet metal cell walls, cell center-to-center spacing, cell bowing, and the

Boraflex neutron absorbing properties. The biases and uncertain-ties of the analytical methods were also taken into account in this manner.

26. The final k-effective for Region 1 provided by this analysis is 0.9403, which, including all uncertainties, is less than 0.95, with a 95% probability at a 95% confidence level.

Therefore, the 0.95 acceptance criterion is met for the storage of fresh fuel assemblies with an initial Uranium-235 enrichment of 4.5%.

D. Recion 2 Criticality Analyses for Presh Fuel

27. The Region 2 fuel racks were analyzed for the condition where, under administrative controls during installation activi-ties, fresh fuel assemblies are stored in a checkerboard pattern.

In addition to the common assumptions I have described in paragraph 23, the criticality analyses for the storage of fresh fuel in the Region 2 fuel racks employed the same specific assumptions listed in paragraph 25 for the analyses of fresh fuel storage in the Region 1 fuel racks.

28. The nominal case k-effective for Region 2 provided by this analysis is 0.8342, which is much less than 0.95. Calcula-tion of the remaining biases and uncertaintites was not deemed necessary in this case since assuming conservative values for these terms will result in a final k-effective for the checker-board configuration well below 0.95. Therefore the 0.95 accep-tance criterion is met for the checkerboard pattern storage of fresh fuel assemblies with an initial U-235 enrichment of 4.5%.

. 1 i

E. Reaion 2 Criticality Analyses for Soent Fuel

29. The Region 2 fuel racks were analyzed for storage of fuel assemblies with various amounts of burnup. In addition to the common assumptions described in paragraph 23, the criticality analyses for storage of spent fuel in the Region 2 racks also employed the specific assumptions listed in paragraph 25, except that a U-235 fuel enrichment of 4.5% was only one of the fuel l

enrichments evaluated, and credit was taken for the neutron absorbing fuel isotopes, such as U-234 and U-236 calculated by PHOENIX, and for the fission product poison isotopes calculated I

by CINDER. Additionally, the analyses of the Region 2 racks for storing spent fuel account for uncertainties and biases not present in the analyses of storage of fresh fuel, such as the reactivity uncertainties associated with the calculated concen-trations of plutonium and other isotopes that are a function of irradiation. These additional uncertainties and biases were evaluated and taken into account employing the methods described in paragraph 25 for the Region 1 criticality analysis.

30. The Region 2 fuel racks were analyzed for a set of conditions where fuel assemblies all having the same specified initial fuel enrichments and burnups occupy every available '

storage position. The purpose of these analyses was to determine r which set of initial fuel enrichments and burnups is acceptable 1

I for storage in the Region 2 racks.

l l

31. The analyses were conducted in an iterative process in which the fuel isotopic and fission product content was deter-mined by PHOENIX for a given initial enrichment and burnup. This information was used in the KENO-IV model to calculate k-effective for an infinite array of such assemblies. The calcu-lated k-effective was then compared with the 0.95 acceptance criterion to determine whether fuel assemblies with the given initial enrichment and burnup were acceptable for storage in the Region 2 racks. The process was repeated for several initial enrichments to determine the fuel burnup that produced a k-effective that satisfied the 0.95 criterion.

32. The results of the criticality analyses for the storage of spent fuel in the Region 2 fuel racks are presented in Figure

4. The curve, obtained by the iterative process I have described above, shows burnups in MWD /MTU for initial fuel enrichment that meet the 0.95 k-effective acceptance criterion for racks fully occupied by such fuel assemblies. Combinations of initial enrichment and burnup to the left of the curve are acceptable for fuel rack storage; combina'tions to the right of the curve are not acceptable for fuel rack storage. The calculated value of k-effective obtained in generating the curve is 0.9304 which, including all uncertainties, is less than 0.95 with 95% proba-bility at a 95% confidence level. Note that the intercept of the curve at zero burnup shows that fresh fuel with a 1.5% initial

S 40-ACCEPTABLE FOR FULL RACK STORAGE 32<-

C 2

3 3:

9.

S g 24<

S E

e

> 16- NOT ACCEPTABLE FOR id FULL STORAGE 3

8<

0 .

2.0 3.0 40 5.0 1.0 INITIAL ASSEMBLY AVERAGE ENRICHMENT (WIO U 235)

Fig,ure 4 MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION ll STORAGE

enrichment may be safely stored in Region 2, and that fuel with an initial enrichment of 4.54 and a burnup of 39,000 MWD /MTU may

also be safely stored in Region 2.

l F. Criticality Effects of Accidents i

33. The criticality effects of accidents involving the Turkey Point spent fuel pools were considered in accordance with the double contingency principle of ANSI N16.1-1975 (Ref. 11).

In effect, the double contingency principle states that it is not ll necessary to consider two unlikely, independent, and concurrent changes in conditions in performing criticality analyses. This principle has been adopted by the NRC's guidance "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applica-tions" (Ref. 2).

34. During normal conditions, the water in the spent fuel pools is borated with approximately 1950 ppm boron. This is equivalent to a negative reactivity change of about 0.30 in k-effective. As I mentioned in paragraph 23, all of the criti-cality analyses for the Turkey Point spent fuel pool were performed assuming the absence of this boron, which is one particular accident condition.

35. Under the double contingency principle of ANSI N16.1-l 1975, it is unnecessary to postulate an accident involving both '

the absence of this boron plus another independent change in i

conditions involving the spent fuel pool. Accordingly, the

~ _ . . . _ . . _ _ _ _.

analyses of the criticality effects of other types of accidents take credit, where appropriate, for the negative reactivity ,

present as a result of the borated water.

36. There are three basic types of postulated accidents that might have an impact upon k-effective. The first is absence of neutron poisons in the spent fuel racks. The second type of accident is one that would involve a change in the neutron moderating properties of the spent fuel pool water. The third type of accident is one that would involve mechanical damage, leading to a change in the geometric arrangement of the fuel rods, fuel assemblies, or storage cells. Each is discussed below.

37. The Boraflex poison plates in the new storage racks provide a negative reactivity change of about 0.25 in k-effec-tive. Thus, for postulated accidents involving the absence of the Boraflex, k-effective of the storage racks would still be less than or equal to 0.95 when, as noted in paragraph 34, the 0.30 k-effective value of the dissolved boron in the spent fuel pool water is accounted for. In other words, the 0.25 increase in reactivity associated with the absence of the Boraflex would be more than offset by the 0.30 decrease in reactivity associated with the presence of borated water, thereby assuring no net increase in the calculated k-effective.

38. If the temperature of the spent fuel pool wate': were to increase, its density would decrease. As noted in paragraph 24e, the condition for optimum moderation in the Turkey Point spent

l fuel pool occurs at a water density of 1.0 grams per cubic centimeter, which is the density assumed in the criticality analyses. Therefore any decrease in water density will result in a value of k-effective that is lower than the optimum value.

39. An inadvertent drop of an assembly between the outside periphery of the rack and the pool wall, a cask drop accident, an earthquake, damage to the fuel racks when empty modules are being installed, and other types of credible accidents which could change the mechanical or geometric configuration of the fuel assemblies or the storage racks might result in an increase in reactivity. However, any such changes in reactivity would be minor in comparison to the negative reactivity effects of considering the presence of borated water, because the undamaged geometic arrangement of the fuel rods is very close to the maximum achievable k-effective. Consequently, the decrease in reactivity associated with the borated water would more than offset any increase in reactivity associated with these acci-dents, thereby assuring no net increase in the calculated k-effective.

40. In summary, the criticality analyses for the Turkey Point spent fuel pool expansion assumed an accident condition involving the absence of boron in the pool water, and the results of these analyres were acceptable. Under the double contingency (i.e., single failure) principle, credit for the borated water may be taken when evaluating other types of accident conditions.

Since the negative reactivity of the borated water more than

--- - -g----- , - , - - -,-----,--c .n- -

re- -

offsets an increase in reactivity resulting from other accidents, these accidents would not cause the calculated k-effective to exceed acceptable limits.

IV. Summary and Conclusions

41. The criticality analyses performed for the Turkey Point spent fuel pool expansion amendments conform with applicable standards and criteria. These analyses demonstrate that fuel assemblies of authorized initial enrichments and burnups to be stored in authorized storage patterns in the Region 1 and Region 2 racks will have a k-effective of less than 0.95, including all uncertainties, under both normal and accident conditions.

1

REFERENCES

1. NUREG-0800, U.S. Nuclear Regulatory Commission Standard Review Plan Section 9.1.2, " Spent Fuel Storage," Rev. 3 (July 1981).

2. Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes, Assistant Director for Engineering Projects, Division of Operating Reactors, April 14, 1978, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications."

3. American Nuclear Society, American National Standard,

" Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Stations", ANS-57.2, ANSI N210-1976, April 12, 1976.

4. L. M. Petrie and N. F. Cross, " KENO IV--An Improved Monte Carlo Criticality Program," ORNL-4938 (November 1975).

5. A. J. Harris, et al., " A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors," WACAP 106, June, 1982.

6. England, T. R., " CINDER - A One-Point Depletion and Fission Product Program", WAPD-TM-334, August 1962.

7. S. R. Bierman, et al., " Critical Separation Between Subcrit-ical Clusters of 2.35 wt percent 235U Enriched U02 Rods in Water with Fixed Neutron Poisons," Battelle Pacific North-west Laboratories PNL-2438 (October 1977).

8. S. R. Bierman, et al., " Critical Separation Between Subcrit-ical Clusters of 4.29 wt percent 235U Enriched UO2 Rods in Water with Fixed Neutron Poisons," Battelle Pacific Northwest Laboratories PNL-2615 (March 1978). 1 1

9. J.T. Thomas, " Critical Three-Dimensional Arrays of U (93.2) ,

Metal Cylinders," Nuclear Science and Engineering, Volume )

52, pages 350-359 (1973). I l

10. Melehan, J.B., " Yankee Core Evaluation Program Final Report," WCAP-3017-6094, January, 1971.

11. American Nuclear Society, American National Standard, "American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,"

ANS-8.1/N16.1-1975, April 14, 1975.

l

FURTHER AFFIANT SAYETH NOT The foregoing is true and correct to the best of my knowledge, information and belief.

fI 1 WiTilim A. Boyd n hS -

//

STATE OF PENNSYLVANIA)

COUNTY OF ALLEGHENY )

$b Subscribed and sworn to before me this 7A day of M' , 1986. My commission expires: /d - /V-8 7. ,

d sw b. #

A

-NOTARY PUBLIC LORRAINE M. PIPLICA. NOTAEY PU3 tlc MONROEWlLE BOKO. ALLECHthY COUNTY MY COMMISSION EXPIRES CEC 14.1987 Member Pennsyhania Association of hotanes e

i i

l l

_.m-

EXHIBIT A STATEMENT OF PROFESSIONAL QUALIFICATIONS OF WILLIAM A. BOYD My name is William A. Boyd, and my business address is Westinghouse Electric Corporation, P. O. Box 3912, Pittsburgh, Pennsylvania, 15230. I am a Senior Engineer in the Core Engi-neering section of the Westinghouse Nuclear Fuel Division.

I graduated from Alliance College in 1973 with a Bachelors Degree in Mathematics. In 1975, I received a Masters Degree from Drexel University in Electrical Engineering. I received a Masters Degree in Nuclear Engineering from the Massachusetts Institute of Technology in 1977.

From 1977 to 1981, I was a Design Engineer at the General Electric Knolls Atomic Power Laboratory in Schenectedy, New York. My duties included the nuclear design and evaluation of a light water breeder reactor and certain navy propulsion reactors.

In June of 1981,. I joined Westinghouse in the Nuclear Design section of the Nuclear Fuel Division, as a Senior Engineer B. My duties included the reload nuclear core design of the Turkey Point Unit 4 reactor. I was later given the added responsibility of fuel rack and shipping container criticality coordinator of the Nuclear Fuel Division. As the criticality coordinator my duties included the direction, coordination, development and review of the methods used to perform all fuel l

l

l rack and shipping container criticality analysis for the Nuclear Fuel Division. In 1984, I was promoted to the position of Lead Engineer with the technical responsibility for the efforts of several engineers and technicians in the reload core nuclear design and analysis of the Point Beach Units 1&2, R. G. Ginna, and Prairie Island Units 1&2.

l l

=

. UNITED STATES OF AMERICA \ )g NUCLEAR REGULATORY COMMISSION G)/ ,

s2/ 2:

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD-

d4b'8 7 , "A Ec '4, Sim.

~4/

In the Matter of

)

)

'Qllfy.

N~~EL; pf

) Docket Nos. 50-250 OLA-2 FLORIDA POWER & LIGHT COMPANY ) 50-251 OLA-2

)

(Turkey Point Nuclear Generating ) (Spent Fuel Pool Expansion)

Units 3 & 4) )

)

CERTIFICATE OF SERVICE I hereby certify that copies of

1. Letter from Steven P. Frantz to Licensing Board Members (January 23, 1986).

2. Licensee's Motion For Summary Disposition Of Intervenors' Contentions (January 23, 1986).

3. Licensee's Statement Of Material Facts As To Which There Is No Genuine Issue To Be Heard With Respect To Intervenors' Contentions (January 23, 1986).

4. Affidavit of Rebecca K. Carr on Contention No. 3 (January 22, 1986).

5. Affidavit of Rebecca K. Carr on Contention No. 4 (January 22, 1986).

6. Affidavit of Harry E. Flanders, Jr.

on Contention Number 5 (January 23, 1986).

7. Affidavit of Leonard T. Gesinski on Contention No. 5 (January 21, 1986).

8. Affidavit of Rebecca K. Carr on Contention No. 6 (January 22, 1986).

9. Affidavit of Dr. Gerald R. Kilp on Contention No. 6 (January 20, 1986).

10. Affidavit of Eugene W. Thomas on Contention No. 6 (January 22, 1986).

) '

11. Affidavit of Rebecca K. Carr on Contention No. 7 (January 22, 1986).

12. Affidavit of Joseph L. Danek on Contention No. 7 (January 21, 1986).

13. Affidavit of Daniel C. Patton on Contention Nos. 6 and 8 (January 22, 1986).

14. Affidavit of William A. Boyd on Contention 10 (January 20, 1986).

in the above captioned proceeding were served on the following by deposit in the United States mail, first class, properly stamped and addressed, on the date shown below.

Dr. Robert M. Lazo, Chairman Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Richard F. Cole Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Office of Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Chief, Docketing and Service Section (Original plus two copies)

Joette Lorion 7269 SW 54 Avenue Miami, FL 33143

re Mitzi A. Young Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Norman A. Coll Steel, Hector & Davis 4000 Southeast Financial Center Miami, FL 33131-2398

/ *[

Stev'en P. Frantz V Newman & Holtzinger, P.C.

1615 L Street, N.W.

Washington, D.C. 20036 Dated: January 23, 1986 t

i I

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