A05935, Responds to NRC Request for Addl Info Re May 1986 Request to Amend License DPR-65 to Allow Storage of Consolidated Spent Fuel in Spent Fuel Storage Pool.Qualification of Analytical Methods Used in Storage Rack Analyses Encl

Responds to NRC Request for Addl Info Re May 1986 Request to Amend License DPR-65 to Allow Storage of Consolidated Spent Fuel in Spent Fuel Storage Pool.Qualification of Analytical Methods Used in Storage Rack Analyses Encl

ML20211B659
Person / Time
Site:	Millstone
Issue date:	10/03/1986
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Thadani A Office of Nuclear Reactor Regulation
References
A05935, A5935, B12275, NUDOCS 8610170318
Download: ML20211B659 (16)

Text

. _ _ _ . _ _ _ . _ _ _ _

c ,

N UTILITIES o.n.,.i Ome.. . s.io.n sir..i. Boriin. conn.ciicui l e EIa == P.O. BOX 270 HARTFORD. CONNECTICUT 061414270 L L J C' ",.fllC'cC~

S (203) 665-50 @

October 3,1986 Docket No. 50-336 B12275 A05935 Office of Nuclear Reactor Regulation Attn: Mr. Ashok C. Thadani, Director PWR Project Directorate #8 Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Storage of Consolidated Spent Fuel In May,1986,(I) Northeast Nuclear Energy Company (NNECO) submitted to the NRC Staff a request to amend its operating license, No. DPR-65, for Millstone -

Nuclear Power Station, Unit No. 2, to allow the storage of consolidated spent fuel in the Unit No. 2 spent fuel storage pool. As a result of the NRC Staff review of this proposal the NRC Staff forwarded to NNECO a Request for Additional Information.(2) The purpose of this letter is to provide the NRC Staff the requested information.

Question #1:

Figure 3.9-3 shows the minimum required fuel assembly exposure as a function of initial enrichment for storage in Region 2 as consolidated fuel. If fuel rods from different assemblics and of different enrichments can be consolidated in one cannister, what value of initial enrichment is assumed in complying with Figure 3.9-37

Response

For the cannister under consideration, the pin with the highest enrichment determines the enrichment (assumed for compliance with Figure 3.9-3 of the license amendment request.1)

(1) 3.F. Opeka letter to A.C. Thadani, dated May 21,1986, " Millstone Nuclear Power Station, Unit No. 2 Proposed Change to Technical Specifications Storage of Consolidated Fuel.

(2) D.H. Jaffe !ctter to J.F. Opeka, dated July 25,1986, " Request for Additional Information on Storage of Con:olidated Fuel for Millstone Unit No. 2".

8610170318 861003 6 00/

DR ADOCK 0500

e- _

1 Question //2 How is the reactivity effect of less than a full consolidated storage box (less than 352 rods) accounted for?

Response

The consolidation process permits the placement of solid metal rods in positions where fuel rods are missing. For those instances were solid rods are not used, a limited number of fuel rods can be omitted based on the attached Figure 1.

Using Figure 1, the reactivity effect of less than a full consolidated storage box can be established by determining the maximum number of fuel rods that can be omitted while maintaining K-eff at 0.95 or less.

Question //3:

What are the values of the biases and calculated uncertainties referred to for Regions 1 and 2, and how were they derived?

Response

The value of the bias is 0.00138 and the 95/95 confidence level calculation uncertainty is 0.00714. The validation report is enclosed as Attachment I.

Question //4:

Explain in more detail how the Region 2 allowable burnup for each initial enrichment accounts for the underestimation of K-effective due to the assump-tion for uniform axial burnup.

Response

The non-uniform burnup distribution which produced the highest difference in reactivity in the Region 2 spent fuel rack when compared with the uniform distribution results is shown on page 3-6 of the license araendment request.(l)

The K-eff is 0.0114. This K-eff was used in Figure 3-3(l) to determine. the burnup needed to accommodate the increase in reactivity due to non-uniform buy,qup. The burnup was found for each initial enrichment shown in Figure 3-3.m The actual maximum burnup was 1,400 MWD /T for a K-eff of 0.0114.

For conservatism, the burnup correction for any initial enrichment was assumed to be 1,800 MWD /T. This value was added to the maximum uniform burnup requirement for each initial enrichment.

Question //5 If Figure 3-4 is based on an infinite array of consolidated fuel, justify why Figures 3-4 and 3-5 need not be derived based on the higher reactivity configuration of one storage pattern of consolidated fuel boxes surrounded by an infinite array of regular fuel assemblies.

p- ,

~

Response

Spent Fuel Pool Technical Specification - 3/4.9.20, SPENT FUEL POOL, will ensure that the K-eff of the spent fuel pool will always be less than 0.95 for any mix of unconsolidated or consolidated fuel. The Technical Specification requires that the blocked cell remain until the Region 2 STORAGE PATTERN of the spent fuel pool racks has been filled. At this time, consolidated fuel can be placed in a previously cell-blocked location only if it is completely surrounded by consolidated fuel. In this way, the unconsolidated fuel will be '

next to consolidated fuel which is stored in a 3 out of 4 pattern. The reactivity of consolidated fuel adjacent to the unconsolidated fuel is less than K-eff 0.95 since it is 3 out of 4, and not 4 out of 4.

Question #6:

Technical Specification 5.6.1.d should include additional wording to clarify that consolidated fuel can be stored in the 4th location of the storage rack only if the surrounding locations are occupied by consolidated fuel storage boxes.

Response

We propose that Technical Specification 5.6.1.d be modified to read as follows:

" Region II of the spent fuel storage pool is designed to permit storage of consolidated fuel in the 4th location of the storage rack and ensure a K-eff less than or equal to 0.95. Placement of consolidated fuel in the 4th location is only permitted if all surrounding cells of the STORAGE PATTERN are occupied by consolidated fuel."

The attached revised page 5-5 reflects incorporation of this change.

Question #7: ,

The NRC Staff recommends that a Technical Specification Surveillance Re-quirement be incorporated for consolidated fuel to verify the integrity of the fuel and structural elements before movement or placement in the spent fuel pool.

NRC Staff Clarification to Question #7:

What method does NNECO propose to verify the integrity of the storage cannister af ter it has been loaded with fuel rods?

Response; Section 4.6.2 of the license amendment request (l) describes the Quality Assurance requirements with which the design, procurement and fabrication of the consolidated fuel storage boxes will comply to ensure that all manufactur-ing and installation activities conform to the acceptable quality requirements throughout all areas of performance.

Static and impact analyses were performed to verify the adequacy of the consolidated fuel storage box design for all the service loads associated with both the consolidation operation and storage in the spent fuel racks. The

r

_4 results of the structural analysis demonstrates that the consolidation box and cover can be safely lif ted and transported using the cover as the lift point. The consolidation box is designed such that it will not be overstressed when subjected to a tensile load of 6000 lbs. The insert assembly supporting the

. weight of the fuel rods can withstand an impact of 5 Gs.

The cover assembly is a spring-loaded self-locking device which has a visual indicator when the cover has been engaged and locked in place. Finally, the cover is dimensionally similar to the upper end fitting of the fuel assembly, thereby permitting the consolidated fuel storage box to be transported by the fuel handling tool / system.

Additionally, it should be noted that, prior to placement of a consolidated storage box in the spent fuel racks, the consolidation operation will have transported the fully loaded consolidated storage box to the temporary racks within the Cask Laydown Area to permit access by the fuel handling tool / system, demonstrating that the fully-loaded consolidated storage box can be transported and placed in the racks while maintaining its integrity.

These measures were introduced into the design of the consolidated fuel storage box so that there would be no increase in the probability of a. fuel handling

-accident as a result of storing consolidated spent fuel. . We consider these me'.sures to be adequate without any augmentation of the previously proposed surveillance requirement.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY -

A F (LA_ U

3. F. bpeKa Senior Vice President

DESIGN FEATURES 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 normal degradation pursuant to the applicabic Surveillance Requirements.

5.6 FUEL STORAGE CRITICALITY 5.6.1 a) The new fuel (dry) storage racks are designed and shall be maintained with sufficient center to center distance between assemblies to ensure a Keff less-than-or-equal-to 0.95. The maximum fuel enrichment to be stored in these racks is 3.70 weight percent of U-235.

b) Region I of the spent fuel storage pool is designed and shall be maintained with a nominal 9.3 inch center to center distance between storage locations to ensure a K eff ess-than-or-equal-to l 0.95 with the storage pool filled with unborated water. Fuel assemblies stored in this region may have a maximum fuel enrichment of 4.5 weight percent of U-235. Consolidated fuel storage boxes may also be stored in this region.

c) Region 11 of the spent fuel storage pool is designed and shall be maintained with a 9.0 inch center to center distance between storage locations to ensure a Keff less-than-or-equal-to 0.95 with the ' storage pool filled with unborated water. Fuel assemblies stored in this region must comply with Figure 3.9-1 to ensure that at least 35% of the design burn-up has been sustained. The contents of consolidated fuel storage boxes to be stored in this region must comply with Figure 3.9-3.

d) Region 11 of the spent fuel storage pool is designed to permit storage of consolidated fuel in the 4th location of the storage rack and ensure a Keff less-than-or-equal-to 0.95. Placement of consolidated fuel in the 4th location is only permitted if all surrounding cells of the STORAGE PATTERN are occupied by consolidated fuel.

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".

CAPACITY 5.6.3 The spent fuel storage poolis designed and shall be maintained with a storage capacity limited to no more than 334 storage locations in Region I and 962 storage locations in Region 11 for a total of 1346 storage locations.

MILLSTONE - UNIT 2 5-5

4-ICrure L

Keff versus Number of Pins per b (for pms n on enrichments at mintmum 6enup or 3reder) l l i !I i I u! <

i l l

. ,8 l

-.m..

..o --o.

_m g

l p

l 5

/* c+,-

/ \

b

/ l o

/ ,

h i ,

z

q. ..

..7- - .: - ...- -. .. ...- -. - --.- --

t ,

I -

8 i

S e

1 1

{- 1

, . i

. ._ - -- - .... 1-I

_. I i

' l 5

W M Q 9 O

d[ )

ATTACHMENTI Qualification of Analytical Methods Used In Spent Fuel Storage Rack Analyses 1

{

1 1

1

r-1 r

Attachment I *

. QUALIFICATION OF ANALYTICAL METHODS USED IN SPENT FUEL STORAGE RACK ~ ANALYSES I.. Purpose -

- The purpose of this att.achment is t'o provide qualiitcation of the calculational model and evaluation of calculational uncertainties and/or bias fe.ctors used in analyzing spent [ fuel storage racks, especially the HI-CAPTM racks

-employing steel boxes and super HI-CAPS containing boro'n carbide poison.

This is . based on the analysis of a variety of reactor and laboratory experiments. The methods of cross-section generation are essentially those of C-E's physics design procedures modified appropriately for use in 'four group-transport, discrete ordinate method criticality calculations, and Monte Carlo codes.

II. Calculational Uncertainty and Bias -

The results of the analysis of a series of UO2 critical experiments are summarized in Table I. These are calculated using the methods described by Gavin (Reference _1) for CEPAK 2.3, which is used in present storage rack I

i I

1 calculations. ' Table I includes the mean'and standard deviation for this CEPAK model.

Although the spatial solution for the flux distribution was obtained by use of a diffusion theory code such as PDQ-7, transport corrections for the reflector.and heterogeneous lattice effects were employed. Thus, for example, in Reference 8,-

the 4.3 w/o U-235 infinite lattice of close-packed assemblies in room temperature water had a K-eff of 1.4547 in PDQ and 1.4568 in DOT, the conservative bias in DOT of 0.0021 will be ignored. These calculations support use of the differential cross-section data base and broad group cross section generation codes.

Since fuel storage arrays do involve the spacing of the fuel assemblies at larger separation distances than in typical PWR reactor lattices, the predictive capability of the calculational model was tested on the following experiments. In these analyses done for this memo, the spatial flux solution was obtained directly with the transport code, ANISN. To assess the accuracy of the calculational model in predicting the multiplication factor of fuel assemblies having a separation distance sufficiently large so as to be. isolated, analyses were carried out for a group of subcritical exponential experiments on clusters of 3.0 w/o UO 2 fuel pins clad with type 304 S.S. and moderated by H2 O (page 165 of Reference 9). The cluster sizes analyzed vary from 181 to 301 fuel rods so as to encompass the range of sizes typical of current PWR fuel assemblies. The multiplication factors for the lattices analyzed using axial bucklings deduced from the reported relaxation lengths are tabulated below.

e

No. of Fuel Rods K-eff I81 ~ 0.9966 j- 211 1.0011 235. 0.9966 265 0.9988 301 0.9984 These results indicate that the calculational model predicts the multiplication factor for small clusters of fuel rods in a water environment to a high degree of accuracy, i.e., a bias of .0017.

To ascertain whether the calculational mode can predict the reactivity character--

istics of ~ thick stainless steel plates and boron poisoned plates an analysis

) (Reference 10) was made of PNW experimental (Reference 11) critical separations of 2.35.w/o U-235 UO2 subcritical clusters. The results using the Monte Carlo code KENO IV are shown in Table II.

Method of Calculation The calculation methods for these experimental comparisons which are also used to determine reactivity for fuel rack storage, fuel shipping containers plus other fuel configurations found in fuel manufacturing. areas are based on CEPAK 2.3 (Reference 1) cross sections. Using an appropriate buckling value and taking proper

. account of resonance absorption,' three fast groups are collapsed from 55 fine energy mesh groups in FORM and the one thermal group is collapsed from 29 thermal energy groups in THERMOS. In addition, each component such as water

-, -----e-- -- ---.- _ w,--rg ,w . w ye- -- s----h--- -a y + -

- - -wwr- -trv-

. gap, or poison plate has its thermal cross section determined by a slap THERMOS calculation employing the proper fuel environment. FORM and THERMOS are sub-programs of CEPAK.

For one-dimensional analyses such as the BNL exponential ' experiments, the discrete ordinates code ANISN (Reference 12) is used. For two dimensional analyses, DOT-2W (Reference 13) is used. For three dimensional analyses (such as the critical separation experiments), KENO IV (Reference 14) is used.

Results

.The above analyses indicate a bias between predicted mean and measured multipli-cation factors of +.00138 and a calculational uncerta.inty of .00714 at the 95/95 confidence level for the complete series of UO2 experiments.~

Thus, using CEPAK 2.3 cross sections, we conclude the following:

Total Number of Results 41 Mean Value @) 1.00138 Standard Deviation = 6 0.00337 g Multiplier for 95/95 Confidence 2.11800 95/95 Confidence Level Uncertainty 0.00714 Bias (A- 1.0) +.00138 Uncertainty minus Bias .00575

^

It should be-noted that the seven no boron steel cases have a bias of 0.00207 (i.e., .

the calculated value is .00207 greater than the critical K-eff value of unity) which '

is greater than the mean bias. The three boral cases.have a bias of -0.00435 with

- unity- particle . self-shielding factor for the B4C. Because of the - size and distributiori of the boron carbide particles, the boron allows more transmission than an equivalent homogeneous boron carbide mixture. Neutron transmission experi-

[

1 ments conducted by the University of . Michigan for Brooks & Perkins, Inc.

(Reference 15) are consistent with using a 0.9 self-shielding factor in the third of four CEPAK neutron group and a 0.75 self-shielding factor in the thermal group.

These self-shielding factors which are used in designing boron containing fuel racks-

- make the bias for these boral cases +0.00008.-

4 Re'ferences:

4 i 1. - P.H. Gavin, "CEPAK 2.3 Mod 0," C-E Internal Report, 12/14/76.

1

2. T.C. Engelder, et al,' " Spectral Shif t Control Reactor, Basic Physics- Pro-gram," B&W-1273, November 1963.

- 3. R.H. Clark, et al, " Physics Verification Program Final Report," B&W-3647-3, March 1967.

4. P.W. Davison, et al, " Yankee Critical Experiments," YAEC-94, April,1959.

j 5. W.J. Eich and W.P. Rocacik, " Reactivity and Neutron Flux Studies in Multi-Region Loaded Cores," WCAP-1443,1961.

, - - - + -

-,m..-. m.- . . . - .. - --,.y,, - , - ~.m., ---,--.,--,,--,-,x-- .---c-,~

., w,---~,,----. -,

- 6. - F.3. Fayers, et al, "An Evaluation of Some Uncertainties in the Comparison Between Theory and Experiments for Regular Light Water Lattices, Brit.

Nuc. Eng. Soc. 3.,6, April 1967.

7. 3.R. Brown, et al," Kinetic and Buckling Measurements on Lattices of Slightly Enriched Uranium and UO2 Rods in Light Water," WAPD-176,1958.

8. 3. Handschuh, L.C. Noderer, R.C. for " Compact Spent Fuel Storage Critical-ity Analysis for Arkansas Power and Light, Unit 2 at- 680F," C-E Internal Report, April 8,1975.

9. G.A. Price, " Uranium - Water Lattice Compilation Part I, BNL Exponential Assemblies," BNL-50035 (T-449), December,1966.

10. L.C. Noderer, " Analysis of Critical Separation of Low Enriched Subcritical Clusters," C-E internal Report, May 11,1979.

11. S.R. Bierman, E.D. Clayton and B.M. Durst, " Critical Separation Between Subcritical Clusters of 2.35 w/o U-235 Enriched UO-2 Rods in Water with Fixed. Neutron Poisons," PNL-2438, October,1977.

12. Ward W. Engle, Jr.,"A Users Manual for ANISN, a One Dimensional Discrete Ordinates Transport Code With Anisotropic Scattering K-1693, March 30, 1967.

li.

~

R.G. Soltose, R.K. Disney, A. Collier, " User's Manual for the DOT-IIW Discrete Ordinates Transport Computer Code," WANL-TME-1982, December, 1969.

14. L.M. Petrie and N.F. cross, " KENO IV, An Improved Monte Carlo Criticality Program" ORNL-4938, November,1975.

15. James W. Bryson, John C. Lee and R. Robert Burn, " Neutron Transmission Through Boral Shielding Material: Theoretical Model and Experimental Com-parison," University of Michigan, Dept. of Nuclear Engineering, Michigan Memorial-Phoenix Project, prepared for Brooks and Perkins, Inc., April,1978.

O s TABLE I Results of Analysis of Critical'UO 2 Systems No[ lattice 0 ot E *

~ _

eff 1

Bau (2) I .08-2 1.00121 2

II

.172-2 1.00534 3

X

.79-2 .99838 4

XIII .701-2 5 1.00419 XX

.202-2 1.00550 6 B&W (3) 1

.861 -2

.t 2 1.00269

.420-2 1.00443 8

Yankee (4) 1 .408-2 9 1.00088 2

.531-2 1.00115 10 3

.633-2 1.0013G 11 Yankee,(5) 4

.688-2 1.00244 Winfrith,(6) 12 R1-20 .660-2 13 1.00214 Rl-80 .626-2 14 .'99942 R3

. 510-2 1.00422 15 Bettis (7) 1 .326-2 16 1.00053 2

17 .355-2 1.00046 3

.342-2 1.00106 Average

'1.00208

+

._.00206

• Using calculated radial bucklings and measured axial bucklirg . .

% -- ,- - ,y-e -

TABLE 11 Calculated keff Values For Separation Experiments Monte Carlo Expt i Type Poison Plate Keff 6(STD Deviation) 15 None 1.00227 .00534 04 None 0.99912 49

.00540 None 1.00221 .00473 18 None 1.00813 21

.00489 None 0.99589 .00461 28- 304 5 Steel 0.0 w/o Baron 1.003Q3 .00308 05* 304 5 Steel 0.0 w/o Boron 1.00329 .00303 29 304 5 Steel 0.0 w/o Baron 1.00271 .00302 12 7 304 5 Steel 0.0 w/o Boron 1.00418 .00273 26 304 S Steel 0.0 w/o Boron 0.99811 .00279 34 304 5 Steel 0.0 w/o Boron 0.99793 .00297 35 304 5 Steel 0.0 w/o Baron 1.00436 .00290 32 304 S Steel 1.05 w/o Boron 0.99970 .00524 33 304 S Steel 1.05 w/o Boron . 1.01173 .00491 4

38 304 5 Steel 1.62 w/o Baron 1.00289 .00512

39. 304 5 Steel 1.62 w/o Boron 1.00208 .00506

~.

20 Boral 0.99585 .00301 16 Boral 1.00020 .00288 -

17 Boral 0.99519 .00286 Mean Keff Value 1.00157 Std. deviation .00419

- .- - - . - .-_ - - . _ . ..