ML20127D027

Proposed Tech Spec 5.6.1 Re Spent Fuel Pool Enrichment Limit

ML20127D027
Person / Time
Site:	Calvert Cliffs
Issue date:	09/01/1992
From:	BALTIMORE GAS & ELECTRIC CO.
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ML20127D020	List: ML20127D016 ML20127D027
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NUDOCS 9209110107
Download: ML20127D027 (12)

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4 ATTACJJ11ENT (JJ UNIT I TECIINICAL SPECIFICATION PAGE 5 5 l

9209110:07 920901 PDR ADOCK-05000317 l' P- PDR

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5.0' DESIGN FEATURES I VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 10,614 1 460 cubic feet at a nominal T.,, of 532'F.

5.5 METEOROLOGICAL TOW $R LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 _FU'!L STORAGE CRITICALITY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum-10 3/32" x 10 3/32" center-to-center distance between fuel assemblies placed in the storage racks to ensure a k ,, of 5 0.95 with the storage pool filled with unborated water. The k of50.95includesthe conservative allowances for uncertainties describ,e,d in Section 9.7.2 of the

'FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 5:0 weight percent.

l 49 CRITICALITY - NEW FUEL

1. " 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 inch center-to-center distance between new fuel assemblies such that k,f' will not exceed 0.95 when fuel having a maximum enrichment of 5.0 weight percent U-235 is in place and various densities of unbarated water are assumed including aqueous foam moderation and full flood conditions. The k,,, of 5 0.95 includes the conservative allowance for uncertainties described in Section 9.7.2 of the FSAR.

l DRAINAGE 1

5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.

l CALVERT CLIFFS - UNIT 1 5-5 Amendment No. 169

1 NITACHMENT (2)

UNIT 2 TECIINICAL SPECIFICATION PAGE 5 5

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5.0 DESIGN FEATURES l

f' 5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figurn 5.1-1.

5.6 FUEL STORAGE 1 CRITICALITY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum 10 3/32' x 10 3/32" center-to-center distance between fuel assemblies placed in the storage racks to ensure a k ,, of 5 0.95 with the storage pool filled with unborated water. The k of $ 0.95 includes-the conservative allowances for uncertainties describ,e,d in Section 9.7.2 of the -

FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 5 # weight percent. l

+s CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 inch center-to-center distance between new fuel assemblies such that k,,, will not exceed 0.95 when fuel having a maximum enrichment of 5.0 weight percent U-235 is in place and various densities of unborated water are assumed including aqueous foam moderation and full flood

• conditions. The k , of 5 0.95 includes the conservative allowance for uncertainties desc,ribed in Section 9.7.2 of the FSAR.

ORAINAGE 5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.

CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblics.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 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.

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l CALVERT CLIFFS - UNIT 2 5-5 Amendment No. 149

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NITACJIMENT (y i

l DESCRIPTION OF Analysis l

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1 i NITACilA1ENT (3)

DESCRIPTION OF ANALYSIS I

h10DELLING In order to accurately predict the multiplication factor of the storage arrays, reliable calculation of the spatial flux distribution, especially in the neutron absorbing regions, is essential. For this reason, a twu dimensional transport calculational model of the spent fuel storage racks is employed in which each component of the fuel storage array is explicitly represented. Thus, in the normal spent fuel

. storage cell calculation, the fuel assembly, and the water channel between the poison box and the wall are represented as separate regions.

The fuel assembly is represented as a 14x14 array of fuel pin cells containing moderator and either fuel pins, guide tubes or instrument tubes. Four neutron energy group cross sections are generated for each fuel assembly cell and for cach component of the storage cell with special attention given to the effect of adjoining regions on the spatial thermal spectrum and hence broad group thermal cross-sections of each separate region of the storage cell.

Cl(OSS SEC'llON GENER ATION The CEPAK lattice program (Version 2.3 hiod 4) is used to calculate four neutron energy group cro.ss sections for the fuct, water and steel regions. This program is the synthesis of a number of codes, e.g., FORM, TilERMOS, and CINDER. %cse programs are interlinked in a consistent manner with an extensive library of differential neutron cross section data. The data base for both fast and thermal neutron cross sections for this version of the CEPAK program is derived from several sources, mainly ENDF/B-IV. This data base gives good agreement with measured data from critical experiments and operating reactors.

Since CEPAK is a lattice cell code, the input geometric buckling must be indicative of the neutronic environment of the fuel assembly in the spent fuel rack. The geometric buckling supplied to CEPAK is derived from the DOT X Y transport solution for a fuel assembly in the rack environment.

The group degadent poison cross sections are generated by a 123 group, P-3, S-8 XSDRNPhi (Version 1.0 hiod 1) calculation. %c resulting set of four group cross sections are a function of the poison density, poison thickness and surrounding environment.

i TWO. DIMENSIONAL GENERATION The two-dimensional, discrete ordinates transport code DOT IV (Version 1.0 Mod 2) was used to determine the spatial flux solution and multiplication factor. An S-6 order of angular quadrature is used with a 1.(XX)5 convergence factor (the ratio of successive eigenvalues for each outer iteration).

In the storage cell calculations, an assembly is represented with at least one mesh interval for each fuel pin cell; the surrounding water channel, steel and water gap regions are calculated with 2 or more mesh intervals.

TilREE-DIMENSIONAL MONTE CARLO CALCUIATIONS The three-dimensional Monte Carlo Code KENO IV was used to determine the reactivity penalty associated with the assumed Boraflex gap distributions.

s

. NITACilMENT 0j 1)ESCillPTION OF ANALYSIS OUALIFICATION OF ANAlyJJCAL MI?filOI)S .

Oualifica tion of the calculational method and evaluation of calculational uncertainties and bias factors are based on the analysis of a variety of critical experiments. Results of this qualification, as revised to account for the epithermal shiciding of poison cross sections, are included in Appendix A.

Included are uranium dioxide critical experiments typical of reactor cores BNL exponential experiments typical ofisolated assemblics, and PNL critical separation experiments with spacings and poison inserts typical of fuel storage racks. For the total of 41 experiments the reactivity is over-predicted by 0.197 percent and has a 95/95 conGdeue level uncertainty equal to 0.714 percent.

L(ESUUrS l The analysis for the Unit 2 racks modeled assemblies with an enrichment of 4.30 w/o U 235. The results are as follows:

1

1. Nominal k-eff. (4.3 w/o,68 degrees, no biases, no uncertainties, no gaps) = 0.923082

2. Delta k-eff Uncertainties and Penalties -

Delta k-eff for temperature change to 40 degrees = 0.000563 Delta k-eff for minimum cell pitch = 0.003552 Delta k-eff for min wall thickness = 0.000785 Delta k-eff for 95/95 cale uncertainty = 0.00714 Delta k eff for boraflex gaps = 0.01126 Square root sum of Delta k-effectives squared = 0.013832

3. Methodology Bias = + 0.00197

4. Final k-eff(4.30 w/o) = 0.923082 + 0.013832 - 0.00197 = 0.934944

5. Max allowable enrichment (k-eff = 0.95) is calculated as follows, using a derivative of r arichment with Delta k-eir of 0.1464 w/o enrichment per pct delta k-eff.

Max enrichment = 4.30 + (0.95. 0.934944) f 100

• 0.1464 = 4.520 w/o The Unit 2 results envelope the Unit I racks in that the Boraflex penalty would not apply to the Unit I racks.

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APPENI)lX_A QUALIFICATION OF ANAIXrICAL MisfilODS USED IN SPENT FUEL STORAGE RACK ANALWES

1. PURPOSE ne purpose of this Appendix is to provide qualification of the calculational model and evaluation of calculational uncertainties and/or bias factors used in analyzing spent fuel storage racks. 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. CALCULATION UNCERTAIN'IY AND lilAS The results of the analysis of a series of UO2 critical experiments are summarized in Table I.

%csc are calculated using the methods described by Gavin (Reference 1) for CEPAK 2.3, which is used in present storage rack 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 j theory code such as PDO-7, transport corrections for the reflector and heterogeneous lattice effects were ernployed. Thus, for example,in Reference 8 the 4.3 w/o infinite lattice of close packed assemblics in room temperature water had a Kerr 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 assemblics at larger separation distances than in typical PWR reactor lattices, the predictive capability of the calculational l model was tested on the following experiments. In these analyses 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 assemblics having a separation distance sufficiently large so as to be isolated, analyses were carried out for a group of suberitical exponential experiments on clusters of 3.0 w/o UO2 fuel pins clad with type 304 S. S. and moderated by II 20 (page 165 of Reference 9). The cluster sizes analyzed vary from 181 to L 310 fuel rods so as to encompass the range of sizes typical of current PWR fuel assemblics. )

i The multiplication factors for the lattices analyzed using axial bucklings deduced from the reported relaxation lengths are tabulated below.

No. of Fuel Rods Kge 181 0.9966 211  :.0011 235 0.9966 265 0.9988 l 301 0.9984 These results indicates that the calculational model predicts the multiplication factor for L small clusters of fuel rods in a water environment to a high degree of accuracy, i.e., a bias I of-0 0017.

l To ascertain whether the calculational mode can predict the reactivity characteristics of thick stainless steel plates and boron poisoned plates an analysis (Reference 10) was made of PNW 1

4 4 APPENI)lX A QUALIFICAT!ON OF ANAIXi'ICAL Mirl'Il01)S USED IN SPENT FUEL STORAGE RACK ANALYSES experimental (Reference 11) critical separations of 2.35 w/o U-235 UO 2suberitical clusters.

The results using the Afonte Carlo code KENO IV are shown in Table 11 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 f rom 55 fine energy mesh groups in FORht and the one thermal group is collapsed from 29 thermal energy groups in TilERhiOS. In addition, each component such as water gap, or stainless steel has its thermal cross section determined by a slab TilERMOS calculation employing the proper fuel environment.

FORM and TilERMOS are sub. programs of CEPAK. Poison cross sections are generated with a 123 group, P 3, S-8 XSDRNPM calculation.

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

Itesults The above analyses indicate a mean error between predicted and measured multiplication factors of +0.00197 and a calculational uncertainty of 0.00714 at the 95/95 confidence level for the complete series of UO2experiments.

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

Total Number of Results 41 hican Value 1.00197 Standard Deviation 0.00337 Multiplier for 95/95 confidence 2.118 i 95/95 Confidence Level Uncertainty 0.00714 l Bias (Mean Value - t.0) +0.00197 Uncertainty Minus Bias 0.00517 It will be noted that the seven no boron steel cases have a bias of 0.00207 (i.e., the calculated value is 0.0020) greater than the critical Ke rrvalue of unity) which is slightly greater than the mean bias. The seven poisoned critical: have a bias of +0.00452.

ItEFERENCEN

1. P. I1. Gavin, "CEPAK 2.3 Mode 0 " PIlP-76-4SS, Decernber 14,1976

2. T. G. Engelder, et.al., " Spectral Shift Control Reactor, Basic Physics Program,"

B&W-1273. November 1%3

3. R.11. Clark, et.al., " Physics Verification Program Final Report," B&W-3647-3, March 1%7 2

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- - - - = . - . _ - . _ . - . . . - . . . - _ - - - - . - _ . - - . - _ - - - . - - . - .

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. o 3 APPENI)lX A QUALIFICATION OF ANAIXflCAL Mirrilol>S USED IN SPENT FUEL STORAGE RACK ANALYSES

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

5. W. J. Kich and W. P. Rocacik," Reactivity and Neutron Flux Studies in Muhi. Region Imaded Cores," WCAP-1443,1%1

6. F. J. Fayers, et.al.,"An Evaluation of Some Uncertainties in the Comparison Between Theory and Experiments for Regular Light Water Lattices," Brit, Nuc. En. Soc J.,6, April 1%7

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

8. J. Ilandschuh, L C. Noderer, R. C. for " Compact Spent Fuel Storage Criticality Analysis for Arkansas Power and Light. Unit 2 at 68 F," 6370-Pil-RC010, April 8,1985

9. G. A. Price, " Uranium - Water Lattice Compilation Part 1. BNL Exponential Assemblies," BNL-50035 (T449), December 1966

10. L C. Nodcrer," Analysis of Critical Separation of low Enriched Subcritical Clusters,"

P11D-79-38, May 11,1979 .

11. S. R. Bierman, E. D. Clayton and B. M. Durst, " Critical Separation Between Suberitical 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 Discretc Ordinates Transport Code With Anisotropic Scattering K 1693," March 30,1967.

13. R. G. Sottesy, R. K. Disnc y, A, Collier, " User's Manual for the DOT-11W Discretc Ordinates Transport Computer Code," WANL-TME-1982. December 1969

14. L M. Petria and N. R. 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 Comparison,"

University of Michigan, Dept. of Nuclear Engineering, Michigan Memorial. Phoenix Project, prepared for Brooks & Perkins, Inc., April 1978 3

t Al'I'ENDIX A '

QUALII'ICATION Ol' ANALYTICAL Mlil'IlODS USED IN SI'ENT l'UEL STOltAGE RACK ANALYSES TAllLE I RESULTS Ol' ANALYSIS Ol' CRITICAL UO2 SYSTEMS No. Lattice B2 tot Keff*

1 B&W (2) I .88 2 1.00121 l 2 11 .172 2 1.00534  !

3 X .79 2 0.99838 4 Xill .701 2 1.00419 5 XX .202 2 1.00550 6 B&W(3) 1 .861 2 1.00269 7 2 .420 2 1.00443 8 Yankee (4) 1 .408 2 1.(XX)88 9 2 .531 2 1.00115 10 3 .633-2 1.00136 11 Yankee (5) 4 .688-2 1.00244 12 Winfrith (6) R120 .660-2 1.00214 13 R1-80 .626 2 0.99942 14 R3 .510 2 1.00422 15 Bettis (7) 1 .326-2 1.00053 16 2 .355 2 1.00046 17 3 .342 2 1.00106 Average 1.00208 1 00206

• Using calculated radial bucklings and measured axial bucklings.

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Al'PENDIX A QUALIFICATION OF ANAIXflCAL METilODS USED IN SPENT FUEL STORAGE RACK ANALYSES TAlli.E il CALCULATED Kerr VALUES FOR SEPARATION EXPERIMENTS l

Monte Carlo Expt # Type Poison Plate Kerr (STD Deviation) 15 None 1.00227 .00534 N Nonc 0.99912 .00540 1 49 -None 1.00221 .00473  !

I 18 None 1.00813 .00489 21 None 0.99589 .00461 28 304 S Steel 0.0 w/o Boron 1.00393 .00308 05 304 S Steel 0.0 w/o Boron 1.00329 .00303 29 304 S Steel 0.0 w/o Baron 1.00271 .00302 27 304 S Stect 0.0 w/o Baron 1.00418 .00273 26 3N S Steel 0.0 w/o Baron 0.99811 .00279 34 304 S Steel 0.0 w/o Boron 0.99793 .00297 35 304 S Steel 0.0 w/o Boron 1.00436 .00290 32- 304 S Steel 1.05 w/o Boron 1.00367 .00288 33 304 S Steel 1.05 w/o Boron 1.00547 .00288 38 304 S Steel 1.62 w/o Baron 1.00P71 .00304 39 304 S Steel 1.62 w/o Boron 1.01123 .00277 20 Boral 1.00378 .00297 16 Boral 1.00179 .00283 17 Boral 0.99701 .00281 Mean Keff Value 1.00283 Std. dcvbtion .00419 S

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ttnis L A1xiN M,merq Dnector R0f tRT M LAnCAUCA, EMMelhys td%r PATRics L WARLICN Cer#e Datatt AUGUST 9,1992 -

, Time to phase out Trojan Trojan nudearplant is aging and unreliable; _

l PGE and dit region can build a better encryfuture widuout it 4}.y

~

ortland General Corp.should Whysink more money into a plant e .

pullthe. plug on the troubled

'Irojah duelear power plant sulTering from a metallic version of ,

t j I by the end of194 osteoporosis? Indeed, the sugestion

_ by PCE that Trojan could be fixed l' s Once free of this albatross and the 5 q

' and could operate for 19 more years 1L Dnanelal burdenit represents,the at a capacity rate of 71 percentis a P

\f,

• utihty could get on with providing its pipedrearn Onlyoneof16 pros- bj '

customers with a cheaper and mom surized water reactors in the nation reliable source of energy. l Trojan's closure, however, should not beimmediate.That would be has achieved that performance level.

Bosides, repa. iring Trojaais a bad buy, The $200 million that PGE would 3 s.1Q: j 7

devastating to Portjand GeneraPs e pay to replace four steam generators ,

customen t. It would force PGE to would be better spent on a strong {. +y g

scramble to bu replacement power . commitment to. conservation and u imm other util ties at higher prices, renewable-ene; sources, such as Closing Trojan now, moreover, wind, geotherma and solar power. 3 '

would severely disrupt the region's The on1 option that makes sense p power supply and could bankrupt when Port and General Corp 's board y s Ponland Generalif the utilityis not . of directors decides Trojan's fate f>

dfo 1- t later this month is an orderly phase-q g

'h recomme at on oph seout i

Trojan's operation over the next four years is a departure from The Ore-h te o "t, with 1996 as the tar-1o 1)y 1996. PGE can have online two IQ b, t gonian 5long-standingpositionin m.nred combustion turbines that support of this plant. would producer00rierawatts ofcloc-The Ortgonian stin believes that tricity -enoug:t to offset the loss of nuclear power plants represent a safo Trojan. Consultants have connrmed and clean source of energy;however, that thereis a verylarge natural gas

, Trojan nolonger makes economic supply at reasonable pricesin West.

sense or is a relfable energy source. ern Canada, making this source com-Trojan's 16 year history drives this . petitive with other fuels and avail-point home. Not only is the plant rat- able to the middle of the 21st century, ed overitslifetime as tha worst per- In the meantime, PGE can take forming pressurized waterreactorin carc ofits growth by promoting con-the nation, but it also is aging fast. servation and small scale renewable-We see )toblems ahead;more and energy plants and by Jooking for co-longer own times. Trouble. - generatlon opportunities. It also 1.ast year, for example, PGE found should consider advocating a rate cracksin steam generator tubes structure that does not tie profits to causN1 by stress and water corrosion. kilowatt +our sales.

PGE plugged 20 peNent of the 13,500 If the utility does its job wellin tubes, The U.S. Nuclear Regulatory buildim; a diversihd portfolio of Tro-Commission sta! Tis con 0 dent that Jan repiscement power,it might even these repairs would allow safe opera- be able tc rull the pic cn Trojan a tion of the plant though 19% That little earlier than 19%

supports the phase-out plan. Thisis a significant crossroadsin t- The problems with Trojan's steam- - the history of PGE. Trojan has been a generatot s)Tiem also'should be large part of the utility'simage over viewed as a warning sign that 7rojan the- bad.

the paat 16 years - both the good and . '

is aging more rapidly than expected. It's time to say goodbye toit.

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