ML19296D979
| ML19296D979 | |
| Person / Time | |
|---|---|
| Site: | 07001100 |
| Issue date: | 10/17/1979 |
| From: | Lichtenberger H ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 8003140004 | |
| Download: ML19296D979 (29) | |
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C-E Power Systems T el. 203/GS8-t 911 Cornbustion Engineering. inc.
Te ex 99297 1000 Prcspect Hill Roa d Windsor. Connecticut 06095 7'~@ SYSTEMS POWER ih =a 4
t RECENED ll' S
License SiiM-1067 October 17, 1979
- cket 70-H00 U. S. Nuclear Regulatory Commission to C
Washington, D. C.
20555 m
.a Attention: Mr. L. C. Rouse, Chief Fuel Processing & Fabrication Branch Division of Fuel Cycle & Material Safety
Reference:
Amendment Application dated April 23, 1979 and Supplements dated June 27, 1979 and August 7, 1979 Gentlemen:
In our original amendment application dated April 23, 1979 it was stated that an independent review of all criticality safety calculations would be performed by the fluclear Safety Contiittee. That review, including in-dependent analysis, has teen completed and is hereby submitted as supple-mentary information to our original application.
One section of our present license allows storage of touching clad rods in horizontal storage packed in a hexagonal lattice to a maximum slab thickness of 15 inches. This section was inadvertently omitted in our amendment appli-cation and is included in this transmittal.
It is requested that Pace C-17 Revision 1 dated 7/16/79 be replaced by the corresponding attached page (Revi-sion 2) dated 10/8/79.
If you have any questions regarding this application, please contact i1r. G. J.
Bakevich of my staff on extension 3150.
Very truly yours, j
c,n ll % d
, A v' n : i W.. M t < - [] vQ H. V. Lichtenberger
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Vice President-Nuclear Fuel g
Nuclear Power Systems-Manufacturing 9
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Enclosures T*
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diameter pellets is 6.7".
Applying a safety factor of 1.2 yields a slab limit of 5.5 inches.
6.3 Close Packed Rods Touching clad rods in horizongal storage packed in a hexagonal lattice have been analyzed as reported in Figure II-5 of WCAP 2999.
For k = 0.99, e
the slab thickness with full water reflection is in excess of 19 inches, and an allowable slab limit of 15 inches will be applied.
6.4 Transfer of Material Material may be transferred on carts which accommodate one mass or slab limited SIU, or may be transferred by hand, one SIU at a time. Carts used for mass limited SIU's shall provide for centering of the unit, and shall measure at least three feet on a side.
Because most spacing areas do not extend beyond the physical boundary of the equipment, spacing between transfer carts and the equipment is of no concern.
In cases where the spacing area extends beyond the equipment boundaries, such as the storage facilities, the spacing boundary will be indicated with colored tape.
The tape may be crossed by carts only when they contain no more than one mass or volume limited SIU, and then only to permit an operator to transfer that SIU to an available storage position.
License No. SNM-1067, Occket 70-1100 Revision: 2 Date:
10/8/79 Page: C-17
Interoffice Correspondence 7
POWER 31 SYSTEMS H. V. Lichtenberger J. R. Dietrich October 15, 1979 Sub;e:t:
Review of Criticality Assessment in Amendment Acolication for License SNN1 1007 The Nuclear Safety Committee, through a subcommittee consisting of NSC members R. L. Hellens, S. Visner, and J. R. Dietrich, has directed a review of the criticality safety calc
- lations contained in the recently proposed amend-u ment to the Combustion Engineering hiaterial License No. SNht-1067 Docket 70-1100 (submitted 4-23-79, with additions submitted 6-27-79 and 8-7-79).
This proposed amendment raises the fuel enrichment limit to 4. I wt % U-235. Check calculations for selected cases and other analyses were made for the subcom-mittee by R. S. Harding, R. J. Kic.z, and L. C. Noderer, of the Nuclear Engi-neering Department, and the results are contained in the attached memorandum.
The calculations described in the memorandum were entirely independent of those contained in the license amendment application except for the generation of few-group cross sections, a portion of the analysis that was originally per-formed in the Nuclear Engineering Department for the calculations contained in the license amendment application. The end results of the calculations employing these cross sections were, however, examined critically by comparison with results of similar calculations made at other times.
The subcommittee of the Nuclear Safety Committee has gone over the attached memorandum in detail with its authors, reviewing both the methods used and the implications of the results relative to those reported in the license amendment application. The review and the associated analyses have focused on three areas encompassing the likely potential scurces of error which, in turn, might affect the validity of the criticality safety calculations.
.1.
Independent evaluation of those situations in which Safe Individual Unit (SIU) liraits are used.
2.
Verification that geometry and materials are correctly reprecented in hionte Carlo calculations.
3.
Evaluation of the error inherent in few-group calculations for moderated configurations.
)
- 11. V. Lichtenberge r 2
October 15, 1979 The judgement of the subcommittee was that independent computer calculations of most of the specific situations cited in the amendment were not required because the codes employed had already been subjected to benchmark checks reported in Exhibit D of the license amendment. The parametric s..dy (item 3, above) did, however, require a series of new calculations performed by the Nuclear Engineering Department of the C-E Nuclear Power Systems Division using a variety of design codes.
The results of the review revealed no errors in the original calculations pro-vided in the amendment which would significantly influence conclusions about the safety case for the facility at the new enrichment limits.
In particular, no disagreements were found with the proposed posting of work stations based on S1U limits. This is not surprising since the eight cases treated in this way are relatively simpic configurations and there is not much room for disagreement.
Differences did appear between the original and the reviewers' construction of some of the input tc the twelve Monte Carlo calculations, and in those cases in which the difference appeared to be potentially important, the somnuter input was changed and the cases rerun. The changes are cited in the attacheci e mo randum. In sho rt, it appears that uncertainties due to interpretation of geometry and material content were found to be less than 1. 5% in reactivity, an uncertainty sufficiently smaller than the level of subcriticality that it does not reduce the margin required for safe operation to a significant extent.
The calculations for most of the fuel rod configurations dealt with in the amend-ment used the "few-group" methods normally employed in core design work to rep-resent the effects of moderation by water mist and flooding conditions. These methods were developed to represent large regions of relatively uniform neutron spectrum typical of power reactor cores, and they ar a known to lose accuracy when applied to situations in which the neutron spectrum either varies rapidly in space or is determined by. strong interaction between two or more dissimilar regions. Two examples of this sort are d;scussed in the attrched memorandum to show the syste-matic trend of the multiplication constant with numbers of groups used in the spatial calculations and the dependence of the magnitude of the o rror on the physical con-figuration of the case. The cases represent infinite squc re arrays of fuct assemblies, moderated by mist, but arranged with differing interassembly gaps. The example representative of the facility arrancement shows a non-conservative error in the four-group calculation of about 3% A k /k for a situation which is roughly 20% sub-c ritic al. This error is judged by the subcommittee to be well within the expected and tolerable range for analyses of this sort.
The second exampic provides a mo: c severe test of the four-group analysis: it exhibits an error in the range of 7-8% A h/k but it is not representative of fuel assembly arrays permitted in the manufacturing facility by the licence. It does serve to indicate, however, that the errors associated witn few-group method 3 can be markedly reduced by using more advanced cross-section generation methr,ds for
II. V. Lichtenberger 3
October 15, 1979 few-group constants which take better account of spectrum interaction effects.
TS.csc advanced methods are just coming into use at C-E and will be applied to it.e NSC review function as occasions arise. Although they help to understand error trends, it is not felt that extensive analyses of this sort are needed to support the calculations presented in the current license amendment.
The subcommittee concludes from this review that: (1) the calculations in the license amendment application have been carried out with adecuate under-standing of the methods involved: and (2) although the calculational results are subject to some uncertainty, the calculated margins to criticality are adequate to cover the uncertainties. The review has developed information useful in the assessment of uncertainties due to limitations of computer codes. Beyond that it has found only minor discrepanc.cs in the original calculations: a drawing error and an error in taking information from a drawing. These have proved to be inconsequential and the subcommittee finds no fault in the criticality safety argument presented in the amendment application.
.a. Dn J. R. Dietrich Cn-.irman, Nuclear Safety Committee JRD:jd Att.
cc ws att. - B F. J. Pianki G. J. Bakevich cc, w/o att.
-B J. M. V! c s t R. S. Ha rding R. J. 1 lot 7.
L. C. Noderer W. E. Abbott V. C. Hall R. L. Hellens A. Stathonlos S. Visner
s Report to fluclear Safety Comnittee on Review of Nuclear Fuel Manufacturing Application For An Increase Ir. Limiting U02 Enr'.chment From 3.5 to 4.1 w/o U-235 R. S. Harding L. C. Noderer % ) 7(
R. J. Klotz
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(
Combustion Engineering, Inc.
Power Systems Groun Windsor, Connecticut i
October 5, 1979
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.A. Summary and Conclusions The arendrent to S':M License 1067 justifying an increase in the maximum enrichcent from 3.5 to 4.1 w/o U-235, which was submitted to the Nuclear Regulatory Comission under a cover letter dated April 23, 1979, and two subsequent transmittals of additional inforration dated June 27, 1979 and August 7,1979 were reviewed from the standpoint of the technical accen-tability of the criticality evaluations.
The methods of review included detailed checkina of input to selected analyses, checking of data sources cad application of data to assessments of criticality safety by SIU methodology, and evaluation of acceptability of results by comoarison with other analyses. The conclusions of this review are as follows:
1.
On the basis of examining each of the analyses empicying SIUs, 507, of the multigroup KENO calculations, and rationalization of the few group KEN 0 calculations, it was concluded that the original criti-cality analyses shown in the amendment were carried out in a satis-factory way. No significant fault was found in any calculation.
2.
Adequate subcritical margin exists in the various operations when handling 4.1 w/o enriched UO2 providing all administrative controls and area limits are adhered to by Manufacturing personnel.
3.
Detailed checking of selected Monte Carlo input resulted in only a few ambiguities which, when resolved, resulted in lower multipli-cation factors.
4.
The use of 4 neutron energy groups in criticality evaluations tends to underestimate the multiplication factor for a specific low moderator density condition (0.05 gr/cc) by an amount which is deoendent upon the geometry being analyzed. For the case of interest, in-plant storage of fuel assemblics, the calculated multiplication factors may be
~3*, ok nonconservative but the margin to criticality is calculated to be ~20% ok.
s
-1 B.
Introduction The Nuclear Manufacturing Facility in Windsor has prepared a license amendment to justify increasina the U-235 enrichment limit from 3.5 to 4.1 w/o U-235 in uranium dioxide. All steps in the fuel handling process were reviewed by Muclear Manufacturing personnel, criticality evaluations for various operations were carried out under the direction of the super-visor of Muclear Licensing and Safety, and a license amendment was ore-pared and transmitted to the U.S. Nuclear Regulatory Commission (NRC) under a cover letter dated Aoril 23, 1979.
In the cover letter it was stated that an independent review of all criticality safety calculations was being carried cut under the direction of the C-E Muclear Safety Comittee and the results of this review would be forwarded under seoarate cover. The purpose of this document is to sumarize the method and results of the review. The basic objective of this review was to determine whether the results of the criticality safety analysis were reasonable and acceptable.
The apolicable standards, regulatory guides, or branch technical positions which are available for guidance in this review appear to be limited; con-sequently precedent and guidance from the NRC reviewer would aapear to be a dominant factor in the overall review process. The only standard which appears to be pertinent is ANSI N 16.1-1975, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
Section 4.2.5, Subcritical Limits, states that where applicable data are available, sub-critical limits shall be established on bases derived from experiments, with adequate allowance for uncertainties in the data.
Plots of limiting values of U-235 mass, cylinder diameter, slab thickness, volume, and areal density are given as a function of w/o U-235; these curves are based on the data compiled by H. K. Clark in Reference 1.
Safe masses and dimensions are defined therein as those values resulting in an effective rultinlication f actor which lies 0.02 helow the averane curve defined by analysis and normalized to exoerimental data.
Deviations between individual experimental data coints and average values of keff given by the curves fall within t0.015 and for the most part within 10.01, consecuently one may deduce that an effective multipli-caNn factor as hich as 0.995 is acceptable under this standard.
For this license application added conservatism is introduced in the definition of safe masses and dimensions beyond that defined by this standard.
Criteria on safe masses and dimensions are used quite extensively in this license amendment for demonstrating safe conditions for relatively simnle geometries of fissile materials or for specific configurations which are either closely related to a configuration exainined experimentally or con-vertible to a simple ceometry through areal density techniques.
Criticality evaluations of the more complex geometries are modelled on the computer using the Monte Carlo computer code, KEN 0 IV.
t.
Extensive compilations of experimental data are available for model verifi-cation (see for examnie, References 2, 3, and 4).
However, there are noticeable nars, the most orominent can being for lattices moderated by lcv density hydrocenenus material so as to simulate so-called low density mist conditions. Mist conditions are costulated to occur for open arrays of fissionable material which mav he accessible to, for examnie, fire fichtino equipment such as sprinklers. foam or fog type nozzles.
Tyoically, the most reactive moderator distribution is costulated so as to define the upper bounds to the multialication factor.
in reality, most fire fighting techniques are not able to give a sufficiently high density of mist so as t'o meet the postulated conditions.
For examnie, overhead sprinklers can lead to conditions approximating water densities of 0.1% whereas fire fighting foams yield a hydrogen distribution couivalent to a water density of aporoximately 3", (special foam materials can go as high as 6%).
Over-head sprinklers are emoloyed in the manufacturing facility and fire hoses may be used but no foam is allowed for fire fighting in the facility.
Consequently there is the likelihood that low deasity mist conditions could occur and, under certain conditions high local concentrations of water cc;1d oci.ur for short periods of time, but the design and location of the facility is such that completeimmersion of manufacturing operations in an environment of water having an effective density of greater than 0.1 gr/cc is highly improbable.
The KEN 0 analyses carried out in support of this license amendment are of two types:
(1) 16 group calculations employing the Hansen-Roach cross-section libraries, and (2) 4 group analyses where the broad group cross-sections are derived by the CEPAK lattice code.
Reference 5 has been cited as a basis for validating the use of the KENO with the Hansen-Roach cross-sections; however, the benchmark analyses were primarily reflected and unreflected clastic moderated criticals with uranium enrichments in the U-235 isotooe of 5', or less and H/U-235 ratios in the range of ~130 to -970. The absence of benchmark experiments on large geometry fissile arrays moderated by low demity mist accears to be a persistent problem. The 4 neutron croup aonroach employing CEMK as the cross-section generator has been emolo.yed for criticality evaluations of snent fuel storace racks formany years. Appendix A provides information on the benchmarki'g of this tech-nique. However, there are still valid questions as to the uncertainty of this 4 group technique when aoolied to fissile configurations containing a mist environment. This uncertainty is covered by allowing a large mar-gin to calculated criticality.
Administrative controls olay an important part in assuring that safe conditions exist in the fuel manufacturing facility.
Since Un2 fuel of less than 65 w/o U-235 cannot achieve criticality in the absence of hydrogeneous moderator and many of the operations in manufacturing are dry, a key objective nf the administrative controls is to assure that in the event of floodino of the facility by water (so as to produce an environment comparable to part or full density water), fuel configurations are such that criticality safety is assured as demonstrated by the criti-cality safety analyses. Administrative controls and/or procedures include the following. All operations and changes in operations must be analyzed
, to establish safety limits and controls and these analyses must undergo independent review. Safety limits and controls are documented by written procedures. Signs defining the criticality safety limits are posted at each work station: in cases where the safe scacing area extends beyond the equipment boundaries, the boundary of the spacing areas are indicated by colored lines in the floor. All mass limited containers are labeled to indicate the enriebrent and uranium content; procedures require labeling so that the identity of all the fuel enrichments is known through-out the facility. Line management, includino the t!uclear Licensing and Safety supervisor, are responsible for enforcing all administrative con-trols.
3,
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6-C.
Review Approach 1.
CategoritatQnofCriticalityEvaluations For purposes of reviewinq the various criti ality analyses, they are divided into fcur cateoories as listed in Table 1.
The first cateqory contains those analyses which ase as their basis for demonstrating suberiticality the definition of Safe Individual Units (SIU).
As noted earlier, safe masses and diEensions are defined by using exoeri-mental data to determine the racnitude of these parameters such that a subcritical multiolication factor is assured.
The second category consists of these criticality evaluations which employed the KENO-IV computer code and the 16 group Hansen-Roach cross-section librarv. The ma.iority of these analyses involve homo-geneous mixtures of fuel and moderator. The third cateqory consists of those KEN 0 evaluations which employed four neutron group libraries generated by the CEPAK code. These cases involve heterogeneous arrays of fuel, moderator, and in some cases, structural materials which were not amenable to volume homogenization. The fourth category involves those operations which are deduced to be subcritical by comparison with other evaluations.
2.
Review Procedures a.
Safe Individual Units The most exceditious way of reviewing those analyses employing SIU limits was determined to be an examination of the data sources and a check of each anplication under the criteria adopted for this license amendment.
b.
16 Group Monte Carlo Analyses Four out of seven of the KEN 0 analyses in this category were selected for detailed review of the analysis (items C3.3, C3.4, C6.1 and~ C6.2) on the basis that they postulated moisture in the fuel and involved the more complex geometries.
c.
4 Group Monte Carlo Analyses Of the four analyses in this category, three exhibited maximum effective multiplication factors under full density water environ-ments and the fourth, In-Plant Storage of Fuel Assemblies (C-8.7),
exhibited its most reactive condition under a mist environment.
The review of this cateoory focused on the cuestion of calcula-tional uncertainties associated with the four grcuo approach for mist conditions; in particular, what bias may result from the use of only 4 neutron groups rather than a larger number.
g,
', TABLE I Categori:ation of Criticality Evaluations A.
Safe Individual Units 19.1 Limits for Individual Units 19.2 Interaction Analysis C-3.6 Pressing C-3.7 Devaxing and Sintering C-3.8 Final Sizing C-6.2 Pellet Storage Shelves - Additional Storage Evaluation C-6.3 Transfer of Material C-7.0 Pretreatment of Low Level Liquid Hastes C-8.9 Fuel Salvage C-L.10 In-Process Storage of Fuel Pellets in Containers C-8.11 Rod Transfer B.
16 Group Monte Carlo Analyses C-3.2 Virgin Pouder Storace Area C-3.3 Batch Make-Up C-3.4 Powder Preparation and Blending C-6.1 Concrete Block Storage Area C-6.2 Pellet Storage Shelves C-8.4 Fuel Rod Storage Area C-8.5 Double Shelf Rod Storage 1cks C.
4 Group Monte Carlo Analyses C-8.1 Pellet Alignment and Drying C-8.2 Rod Loading and Fuel Rod Transport Carts C-8.7 In-Plant Storage of Fuel Assemblies C-8.8 Shipping Container Storage D.
Comoarative Analyses C-3.5 Final Mixing C-8.3 Autoclave Corrosion Test C-8.6 Fuel Assembly Fabrication
's.
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D.
Results of Review 1.
Safe Individual Limits A safe individual unit limit is that mass or dimension which charac-terizes either a heteroceneous or homogeneous array of fuel and moderator known to be subcritical (safe) by ccmparison with experimental data. The definitien of safe masses and dimensions employed in this license aoolication have added conse tism beyond that defined in ANSI-N-16.1.
Section 19.1 quotes safety factors of 2.3,1.3,
'.1, and 1.2 for mass, volume, cylinder diameter and slab thickness, respectively, as existing in the definitinn of safe indivital unit limits. Attempts to verify the magnitude of the safety factors indicated that they may be underestimated relative to the data of Reference 2 but overestimated relative to the curves in ANSI-N-16.1.
In any event it is concluded that there is substantial conservatism still remaining relative to either source of data.
Additional conservatism has also been introduced in certain cases to meet criteria on maximum fraction critical values.
A review of section 19.2 concluded that Table 19.2 did not include the safe volume and spacinq area for two story operation with 4.1 w/o U02 fuel to support, for example, the pressing operation in Exhibit C.
Although the correct criteria are included in the text of 19.2 and the safe volume limit is given in Table 19.1, an additional statement in Table 19.2 is required for completeness.
Based on a review of the references of section 19.1, it was concluded that the single level spacing criteria of Table 19.2 meet the fraction critical criteria outlined on oage XIX-4. Comments on individual sections of Exhibit C employing the SIU approach are summarized below, a.
Section C-3.6 Pressino The writeun of this section of the license application is very brief and does not address criteria and controls for all asoects of the pressing operation.
For example, interactions between hoppers and Vesses are not discussed. Actually the hoooers are designed to be safe cylinders (3.5 w/o) or safe volumes (4.1 w/o) and are' located cn a mezzanine above the presses.
Criteria out-lined in Sections 19.1 and 19.2 for application to two story ooeration indicate that the safe volume hopoer for 4.1 w/o powder in combination with the already designated spacing areas for 3.5 w/o powder lead to safe conditions.
The statements concerning maximum height of boats and a limit of only one boat at a oress work station at any given time do not adequately describe the jus-tification for safety of this oneration. Actually the boats have a wall height of only 2 15/16 inches, the cellets are loaded to a point below the lip of the boat and a card is placed on top of the pellets giving certinent data on the feel. This procedure oro-vides reasonable assurance that the safe slab limit of 3.7 inches is met.
In addition, a boat loaded with pellets stacked to the height of the sidewalls of the boat contains approximadly 12 kg UO2 which is well below the safe mass limit of 24 kg for 4.1 w/o fuel.
-9 b.
Section C-3.7 - Oewaxina and Sintering The criterion for safety of this operation is a safe slab limit of 3.7 inches which is achieved through controls discussed above on the height of pellets loaded in the boats.
c.
Section C-3.8 - Final Sizing Here again the safety criterion is a safe slab linit of 3.7 inches.
In most staos of this operation the fuel is distributed in a layer having a thickness of -1 pellet daimeter.
The only part of the operation requiring investiqation was the infeeder where the pel-lets are dumped into a flooded bowl, one boat at a time to meet the 3.7 inch safe slab limit.
The volume of the infeeder bowl is 31.2 litres and for a volume ratio of H 0/U02 less than 1.5, the criti-2 cal volume is greater than 35 litres. Even if this bowl was fully filled with pellets it would be necessary to attain the large volume fraction of H2, consequently the bowl is a safe volume 0
for this operation. At the output of the grinder, the pellets are placed into " grinder trays" having a height of 1.5 inches and a cover assembly. The paragraph on drying of grinder sludge does not state what control is employed; corrents to the effect that the sludge is collected in volume limited SIU containers and trays having a maximum depth of 3.7 inches so that the oven is limited to a safe slab configuraticn, as stated in section 7.0, should have been included in the text.
The centrifuge and grinder coolant sump meet the safe volume and spacing area criteria of sections 19.1 and 19.2 although the spacing area of the grinder coolant sump (9 ft ) is not stated. For the storage rack (W.S. P-20), the 2
safe slab limit is achieved by storing grinder trays no more than 2 high.
d.
Section C-6.2 - Pellet Storace Shelves (W.S. P-15,19, 21)
An increased slab thickness limit was adopted for this storage area so as to accommodate 3 levels of grinder storage trays. The consequence of using this increased slab thickness is reduced con-servatism relative to section 19 criteria. The rationale for deducing the 5.5 inch slab limit was reviewed and found to be con-servative relative to exneriments by a factor of 1.20 on the slab thickness when the H 0/UO2 volume ratio is < l.0.
In reality, 2
three levels of grinder storage trays would result in a fuel height of less than 4.5 inches which is significantly less than
'w 5.5 inches but more than the 3.7 inch slab limit used in other operations.
e.
Section C-6.3 - Transfer of Material The SIU limit in combination with the cart dimensions meets the criteria in sections 19.1 and 19.2.
Transfers by hand occur in-frequently and are limited to one STU.
Attention is focused on cart transfers since carts may t'e left unattended and the dimen-N, sion of the cart assures a safe spacing area. When material is transferred by hand it is generally either for a very short dis-tance or consists of an amount of fuel less than contained in 1 SIU.
~. - -
f.
Section C-7.0 - Pretreatment of Low Level liquid Wastes The rationale employed to deduce the safety of this operation was reviewed and found to be accentable.
Although the diameter of the tank exceeds the safe cylinder limit (9.8" in Table 19.1) by 0.2", it is smaller than the critical diareter of the infinitely long, fully reflected cylinder having optimum moderation of the fuel water mixture by 0.8 inches.
This maroin in combination with the finite height of the settling portion of the tank (18 inches) should offer sufficient conservatism.
g.
Section C-8.9 - Fuel Salvace In addition to being mass limited, safe containers are employed to rec 2ive the recovered fuel.
h.
Section C-8.ll - Rod Transfer The rationale for the increase in the safe slab limit to 5.5 inches is discussed in paragraph Id, above, on section C-6.2.
2.
16 Grouo ffonte Carlo Analyses The detailed review of the following analyses included checking the fuel composition for the most reactive case in each series of calcu-lations for nuclide number densities, enrichment, potential scattering, dimensions, geometrical representations, and composition of other materials.
C 3.3 Batch Make-Up C 3.4 Powder Preparation and Blending C 6.1 Concrete Block Storage C 6.2 Pellet Storage Shelves The powder prenaration and blending station involves a complex three dimensional geometrical representation in KENO which makes verifica-tion difficult. However, spotchecking of the geometry was done by reconstructing from the generalized geometry equations. The following comments are provided for the indicated sections.
a.
Section C-3.2 - Viroin Powder Storace Area The fact that interseersed water moderation and flooding were not addressed does not belong in the listing of conservative assumo-tions. The design of the facility itself provides the principal argument that external moderation is of no concern in this case.
Internal moderation is addressed by the checking nrocedures of Section C-3.1 and the assumption in the criticality analysis of a higher value of moisture content than the limit defined in C-3.1.
en
b.
Section C-3.3 - Batch Make-Up The introduction of internal and external moderation as two separate variables raises questions as to the rate of convergence of the interative process and the validity of the process in general. Two iterations were carried out and there is no indication that the orocess is converaine. However, it would anpear from the trends of the analyses that any realistic conditions of noderation would result in much lower multinlication factors than corouted here since:
(1) the presence of near full density water external to the fuel containers is not probable under any conditions inside the hood, and (2) the ootimum moderation conditions postulated for the fuel containers simultaneous with the postulated external moderation conditions are not realistic for this area of the facility.
c.
Section C-3.4 - Powder Preoaration and Blending In subsection 3.4.2 under criticality analysis, the definition of optimum moderation should be stated more clearly as to the external moderation conditions. Once again the internal and external moderation variables approach is pursued.
However, in this case the calculations appear to be convergent.
The multi-plication factors indicate a high degree of suberiticality even with extremely high degrees of moderation in both regions.
Drawing NFM-C-4065 shows 6 inches le separation of the two one-half inch thick fuel layers on conveyor belt whereas the calculation used 9 inches. The correct dimension is 9 inches; the drawing should be modified.
In the discussion of the front end of the station, clarification of the term " optimum moderation" as employed in the conservative assumptions vould be beneficial. The iteration on internal and external moderation conditions was truncated after the first itera-tion on external moderator conditions and the resulting maximum multiplication factor is in the range of 0.94 to 0.95 for both enrichment cases. While there is less maroin to criticality than in the cases discussed above, it is difficult to see how the flooding of these areas with full density water could occur.
In the review of the KENO geometry, two small wedges of fuel at the intersection of the powder spread funnel and the powder tubes appeared to be improperly described in the original definition of the material regions. This should have a negligible effect upon reactivity but the case has been rerun to assess the impact on the calculated multiplication factor.
The revised multiplication factor was 0.8484 t 0.0086 versus 0.8684 t 0.0101 computed earlier, d.
Section C-6.1 - Concrete Black Storace Area i
The review of the calculations uncovered no problems or questions other than those raised above pertaining to the treatment of 1-internal and external moderation as separate variables.
[
l e.
Section C-6.2 - Pellet Storage Shelves j
r The discussion of the pellet storace shelves states that the shelves I
are limited to a slab thickness of'3.7 inches. Actually this is
assured by limiting the number of the covered grinder trays stacked at any position to two wnich implies that the maximum fuel height is <3 inches.
In the analysis, the iteration on internal and external moderation was truncated at the first iteration on external moderation and shewed a relatively low maximum multiplication factor (-0.82) for an external modera-tion condition which is higher than one could attain from fire fighting equipment or a sprinkler system.
In the determination of cross-sections for the pellet-water mixture, a volume homogenization procedure was employed which is non-conservative for low enrichment fuel.
However, it can be shown using data from Reference 2 that the non-conservatism of this approximation is more than offset by the assumption that the pellet water mixture is such that the fuel concentration is 2.7 gr U/cc.
For a random loading of the trays, one would expect a higher fuel density, i.e. of the order of 5.9 gr U/cc.
For 5 w/o UO the following critical masses (kg) are deduced from Reference 2:2 gr U/cc 2.7 5.9 Homogeneous
-97
-570 0.4" dia. rods -86
-230 From these data one can see that the conservatism associated with the assumption of a fuel density of 2.7 gr U/cc in the homogeneous approximation is equivalent to a reduction in the critical mass of 463 kg whereas the non-conservatism of using the homogeneous approximation evaluated for a fuel density of 5.9 gr U/cc is equivalent to only 340 kg.
The review of the KENO analyses indicated that the eight inch thick hollow concrete block wall was represented as full density concrete. The effect on reactivity is expected to be small but the case was rerun to evaluate the effect of reducing the wall thickness to five inches; the peak reactivity decreased from 0.8188 t 0.0088 to 0.7791 t 0.0081.
f.
Section C-8.4 - Fuel Rod Storage Area For U02 of 4.1 w/o enrichment and no hydrogeneous moderation, this area is clearly subcritical.
g.
Section C-8.5 - Double Shelf Rod Storace Racks This analysis is similar to that of section C-8.4 with the excep-tion that the spacing between storage boxes is greater and no physical barriers have been used to exclude mist or personnel from between storage boxes.
Previous analyses for the 3.5 w/o enrichcd fuel employed 4 neutron groups and the DOT code for the spatial calculation.
For the case of flooded boxes the analyses yielded a maximum multiplication factor of -0.87.
In the case of 4.1 w/o fuel but with no water inside the boxes, a multiplica-tion factor of -0.89 was obtained. These two sets of results are
t i
in reasonable agreement if one assumes that the non-conservatism inherent to the use of only 4 neutron groups with nist between boxes is nearly offset by the increase in enrichment from 3.5 to 4.1 w/o and the elimination of the relatively small amount of room temperature water inside the boxes.
3.
4-Group Monte Carlo Analyses a.
Validity of 4 fleutron Group tiodel The analytical model emploved in the criticality evaluation of fuel manufacturing operations considered in this section used CEPAK to generate 4 neutron aroup cross-sections and KEf;0 to solve for the snatial solution of the multiplication factor.
Validation of this model is discussed in Section 3.0 of Exhibit D of the license apolication.
Results discussed in Exhib.t 0 indi-cated very cood agreement for lattices employing full density water. Analyses discussed in Appendix A of this report provide added assurance of the validity of basic cross-section libraries and general methods of analyzing the reactivity of a broad variety of lattices. Thus, there is a high degree of confidence in the methodology not only at the full density water conditions but also at the reduced density conditions existing in a hot, full power reactor environment, i.e. water densities down to ~0.70.
A question arises as to the accuracy of the 4 group approach for mist conditions, viz. how strong is the dependence of the nalti-plication on the number of broad neutron grouos employed when two closely interacting recions of differing neutron spectrum are involved. To examine this coint, analyses of a 14x14 fuel assembly in a mist environment (0.05 gr/cc of H 0) were carried out using 2
the D0T code for the spatial solution and alternate cross-section generators to prepare broad groun cross-sections in differing num-bers of neutron groups.
It shmid be noted that the CEPAK lattice code is not emoloyed to generate broad group cross-sections in more than 3 non-thermal groups since it emoloys the MUFT type solution to solve the multigroue equations.
The GA'1 code was used as an alternate method of deriving the non-thermal broad grouo constants and, in one lattice geometry, the DIT code was employed to generate group constants.
Both the Gall-THERMOS and DIT calcula-tional models employed the same basic multigroup neutron cross section libraries (ENDF/B-IV), however, di f ferences in resonance self-shielding did exist between the two models.
The DIT code is a C-E proprietary code which solves for the multi-group neutron spectrum in 85 neutron groups throughout an entire fuel assembly and not simply the fuel pin cell as with CEPAK or GA!1.
It is oresently employed for reactor lattice calculations where it solves for the soatially dependent multigroup soectrum in various subregions of the heteroceneous fuel assembly and provides few groups,
cross-sections for scecified subrecions.
The DIT code is not cre-sently programmed to deal explicitly with the larne geometrical arrays encountered in the present calculations, but it can be used to generate few group cross-sections more representative of spectral variations than either CEPAK or GAM-TilERM05. This capability was
.~
14 exploited in the generation of 4 and 9 group cross-sections for a 00T calculation of an infinite array of 14x14 fuel assemblies having an edge-to-edge spacing of 12 inches and a 0.05 gr/cc mist both internal and external to the fuel assembly.
Figure 1 shows a comparison of the group dependence of the multi-plication factor computed with DIT and GAM-THERMOS derived cross-sections in the 00T code.
To expedite the GAM-THEPB05 calculations, a modified 14x14 cell definition was used to represent the fuel assembly which preserved the volume of structure, fuel and moderator but redistributed the moderator and structure from the CEA guide tube regions to the moderator region of the unit cells.
CEPAK and GAM were written to use slightly different resonance self-shielding algorithms; the former employs a fit to a Hellstrand experimental correlation whereas the latter used the Nordheim Integral Treatment based on resolved and unresolved resonance para-meters with a CEPAK derived Dancoff factor. The resulting difference in multiolication factors can be seen by comparing the 4 group "CEPAK(HETEROG)" point with the curve labeled GAM-THERf'OS(HOMOG).
To show the few group trend, the latter curve was displaced downward to pass through the CEPAK point and is plotted as a broken line. The points of interest in Figure 1 are as follows.
First, the DIT calculation shows very little "few group" depen-dence because the multigroup solution for the spatially deoendent neutron soectra includes the influence of the mist environment in the generation of the few grouo constants.
Second, the clas-sical tJigner-Seitz cell approximation for the fuel pin and a separate calculation of' reflector constants employed in the GAM-THERM 0S approach does not properly account for the fuel-reflector interaction in the generation of few group constants under the assumed mist conditions.
Increasing the number of neutron groups gives a better approximation to the energy dependent spatial flux solution but may still underestimate the multiplication f?ctor because of a failure to adequately represent the effect of the interassembly mist environment into the calculation of primarily the. resonance escape calculation.
Figure 2 shows the results of the D0T calculations for few group cross-sections derived by both GAM-THERMOS and CEPAK for the larger interassembly soacings emoloyed in the storage of fuel assemblies in the manufacturinq plant: no DIT based calculation could be carried out for this configuration since the problem size exceeds present capabilities. The first noint of interest is that the qroup dependence is much less than in the closer spaced geometry of Figure 1.
The second point is that the bias in the calculation due to using only 4 neutron groups appears to be in the range of 35 Ak.
Analyses at full density water conditions exhibit significantly less dependence on the number of "few groups" employed.
3.
I
b.
Comments on Applicable Sections (1) Secticn C-8.1 - Pellet Alignment and Drying The first paragraph states that the pellet configuration is limited to a 3.7 inch slab thickness; the text could state that this is an administrative control and the layer. of pel-lets on the table is generally one pellet high.
Item 5 of the assumptions stated that the aluminum pellet troughs were not included in the analysis; no reference was made about the remaining aluminum structure inside the 's inch thick stainless steel cylinder. The mean spacing of the pellet columns is sufficiently large that the assumed representation of the fuel / mist lattice is over-moderated in the fully flooded case.
Therefore, whether or not substitu-tion of moderator for aluminum structure is a conservative assumption may depend upon the actual configuration of fuel, structure and moderator.
In view of the latter question and the absence of data points between water densities of 0.5 and 1.0 in Figure 0-1.7.1, a review was made of the analyses carried out to develop Figure 0-1.9.3 (Rev. 2, 9/16/74) which is labelled " Critical and Safe Cylinder Diameters as a Function of Rod Spacing
- and Moderator Density". Although these analyses are for 3.5 w/o fuel rods, thev are informa-tive. They indicate that indeed the oven, if fully loaded and flooded with full density water, is overmoderated and more than 500 rods at the average spacing employed in the oven would be required for a critical configuration. The dif-ference in reactivity due to the uranium enrichment from 3.5 to 4.1 w/o should be more than offset by the introduction of the poison rods. Consequently, if the oven could be fully flooded, it should be subcritical.
Item 6 of the conservative assumptions states that four group cross sections were cenerated by the CEPAK code for fuel and poison regions; actually a HA!01ER-0TF sequence of calculations was used for these lumoed ooisons. The discussion at the top of page 19b is unclear as to whether the internal and external moderator conditions were varied simultaneously or indepen-dently; actually they were treated as a single variable.
(2) Section C-8.2 - Rod Loading and Fuel Rod Transoort Carts The description of the fuel configuration is not too clear; actually there are 250 fuel rods arranged in 5 concentric rings formed by 4 spacer rings within an annulus of -14.6" ID and 25.7" 00. The analytical results are judged to be reasonable on the basis of the relative large mean soacing of fuel rods, the maximum number of rods (250) and decoupled (annular) array of fuel rods in the fixture.
7,
- caption on figure says rod size other than rod spacing.
=.
. (3) Secticn C-8.7 - In-plant Storace of Fuel Assemblies The postulated representation of the fuel array for this cate-gory of analyses resulted in a maximum multiplication factor for reduced water density conditions (P 4. 0.5 gr/cc), of
~0.80.
Consequently a margin of approximately 0.20 ak exists to the critical condition. This margin should be more than adeouate to cover the bias associated with the use of 4 neu-tron groups as well as other biases and calculational uncer-tainties.
(4) Section C-8.8 - Shipping Container Storage If the shipoing containers are olaced in contact there will be a minimum of approximately 14 inches separating fuel assemblies in adjacent containers. Under fully flooded con-ditions this is adequate to orevent interaction. Based on analyses of fuel transfer tubes and upenders, the calculated
' multiplication factor is reasonable.
4.
Comoarative Analyses a.
Section C-3.5 - Final Mixing The writeup neglecgs to state that the spacing area for this operation is 27 ft which meets the requirement of Section 19 of the 1.icense.
b.
Section C-8.3 - Autoclave Corrosion Test The rationale for criticality safety contained in the license application is in error on one point, viz. that the fuel rods in a fuel assembly are spaced at the most reactive oitch.
Actually enaineering judgement would say that the autoclave is highly subcritical with 32 fuel rods present.
The previous defense of this operation for 3.5 w/o fuel showed that the system was subcritical with 120 fuel rods.
For 4.1 w/o fuel the number of rods has been reduced by 73", and the amount of U-235 has been increased by only 17"; in addition, the rod spacing has been increased so as to further decrease the infinite multi-plication factor below that which would exist if the rods were spaced so as to accommodate 120 rods in the autoclave.
Clearly this system will be subcritical.
c.
Section C-3.6 - Fuel Assembly Fabrication The logic emoloyed for deducing safety of this operation is S.
acceptable.
l.
References 1.
H. K. Clark, " Critical and Safe itasses and Dimensions of Lattices Rods in Water," DP-1014, Savannah River Laboratory (1966).
of U and U02 2.
UPAEA Fandbook AH5B(1)- Handbook of Criticality Data, Compilad by J. H. Chairers, G. L'alker, and J. Pugh, United Kingdom Atomic Enerav Authority (1965).
3.
G. A. Price, " Uranium-Hater Lattice Compilation Part I, GNL Exponential Assemblies," P.NL-50035, Brookhaven National Laboratory (1966).
4.
E. Bernocchi and R. Martinelli, " Light Uater Lattice Data,"
NEACRP-U-190, C NEN-Quaderno KIT /FIS (77)l.
5.
G. R. Handley and C. ft. Hopper, " Validation of the "KEM0" Code for Nuclear Criticality Safety Calculations of Moderated, Low-Enriched Uranium Systems," Y-1948, Oak Ridqe Y-12 Plant (1974).
7, I
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r APPEfIDIX A Oualification of Analytical Method Employed in Criticality Evaluation of Fuel Handling Goerations I.
Purpose The purpose of this Appendix is to orovide qualification of the calcula-tional model and evaluation of calculational uncertainties and/or bias factors used in performing criticality evaluations of fuel handling operations with structural and/or fixed ooisons in the fann of steel boxes and boron carbide plate. This qualification is based on the analysis of a variety of reactor and laboratory exoeriments. The methods of cross section generation are ntially those of C-E's physics design proce-dures modified appropr
- 1y 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 'able I.
These are calculated using the CEPAK 2.3 lattice code as a few group neutron cross-section generator. Table I includes the mean and standard deviation for this CEPAK model. These calculations support use of the differential cross-section data base and broad group cross-section ocneration codes.
To assess the accuracy of the calculational mndel in predicting the mul-tiplication factor of fuel assemblies having a separation distance suffi~
ciently larae so as to be isolated, analyses were carried out for a group of subcritical exoonential experiments on clusters of 3.0 w/o UOp fuel pins clad with tyoe 304 S.S. and moderated by H2O (Dage 165 of Reference 7).
The cluster sizes analyzed vary from 181 to 301 fuel rods so as to enccm-pass the rance of sizes typical of current P'.!R fuel assemblies.
In these analyses, the spatial flux solution was obtained directly with the trans-port code, AtlIS!. The multiplication factors for the lattices analyzea using axial bucklings deduced from the reported relaxation lengths are tabulated below.
No. of Fuel Rods Keff 181 0.9966 211 1.0011 235 0.9966 265 0.9983 301 0.9984 These results indicate that the calculational model predicts the multipli-caticn factor for small clusters of fuel rods in a water environment to a high degree of accuracy, i.e. a bias of.0017.
.L
~
A-2 To ascertain whether the calculational model can predict the reactivity characteristics of subcritical clusters of fuel separated by water channels of various thicknesses and, in some cases, with thick stain-less steel plates and boron poisoned plates inserted in the water channels, an analysis was made of the exceriments on critical separations of 2.35 w/o U-235 UO2 subcritical clusters reoorted in Reference 8.
The results using the Monte Carlo code KEil0 IV are shown in Table II.
The calculation methods for these experimental comoarisons, which are also used in criticality evaluations for fuel storace racks, fuel shioping containers, plus other fuel confiqurations found in fuel manufacturing areas, are based on CEPAK 2.3 cross-sectior.s. Using an acorooriate buck-ling value and taking nrocer account of resonance absorption, three fast groups are collansed from 55 fine energy mesh grouos in FORM and the one thermal group is collapsed from 29 thermal energy groups in THERMOS.
In addition, each comoonent such as water gan, or poison olate has its thermal cross-section determined by a slab THERMOS calculation employing an appropriate fuel environment. F0P14 and THEPfiOS are sub-programs of CEPAK.
For one dimensional analyses such as the Bril exponential experiments the discrete ordinates code ANISt! (Reference 9') is used.
For two dimensional analyses 00T-2W (Reference 10) is used.
For three dimensional analyses (such as the critical separation experiments) KENO IV (Reference 11) is used.
The abov analyses indicate a mean error between predicted and measured multiplication factors of +.00135 and a calculational uncertainty of 0.00714 at the 95/95 confidence level for the complete series of UO2 experiments.
Referen es:
1.
T. C. Engelder, et al, " Spectral Shift Control Reactor, Basic Physics Program," B&W-1273, November 1963.
2.
R. H. Clark, et al, " Physics Verification Program Final Report,"
B&W-3647-3, March 1967.
3.
P. W. Davison, et al, " Yankee Critical Experiments," YAEC-94, April,1959.
4.
W. J. Eich and W. P. Rocacik, " Reactivity and Neutron Flux Studies in Multi-Region Loaded Cores," FCAP-1443,1961.
5.
F. J. Fayers, et al, "An Evaluation of Some Uncertainties in the Com-parison Between Theory and Experiments for Regular Light Water Lattices, Brit. Nuc. En. Soc.
J., 6, April 1967.
6.
J. R. Brown, et al, " Kinetic and Buckling Measurements on lattices of Slightly Enriched Uranium and U02 Rods in Light Nater," UAPD-176, 1958.
7.
G. A. Price, " Uranium - Mater Lattice Compilation Part I, BNL Exponential Assemblics," BNL-50035 (T-449), December 1966.
A-3 8.
S. R. Gierman, E. D. Clayton and R. M. Curst. " Critical Separation Between Subcritical Clusters of 2.35 w/o U-235 Enriched U02 Rods in Water With Fixed Neutron Poisons," PNL-2438, October 1977.
9.
Ward W. Engle, Jr., "A Users fianual for A?IISN, A One Dimensional Discrete Ordinates Transoort Code With Anisotronic Scatterin9 K-1693, March 30, 1967.
10.
R. G. Sottesy, R. K. Disney, A Collier, " User's Manual for the 00T-IIll Discrete Ordinates Transoort Computer Code," NANL-TME-1982, December 1969.
11.
L. M. Petrie and ti. F. Cross, "KE'!0 IV, An Imoroved Monte Carlo Criticality Program," 0R.*ll-4938, November 1975.
't.
~
~~ ^~
A-4 TABl.E I Results of Analysis of Critical U02 Systems flo.
1.a t t i ce of eff 1
B&W (11 I
.88-2 1.00121 2
II
.172-2 1.00534 3
X-
.79-2
.99838 4
XIII
.701-2 1.00419 5
XX
.202-2 1.00550 6
8&W (2) 1
.861-2 1.00269 7
2
.420-2 1.00443 8
Yankce (3) 1
.408-2 1.00088 9
2'
.531-2 1.00115 10 3
.633-2 1.00136 11 Yankee (4) 4 688-2 1.00244 Win fri th '(5) 12 RI-20
.660-2 1.00214 13 Rl-80
.626-2
.99942 14 E3
.510-2 1.00422 15 Bettis (6) 1
.326-2 1.00053 16 2
.355-2 1.00046 17 3
.342-2 1.00106 Average 1.00208
+.00206
- Using calculated radial bucklings and measured axial bucklings.
't.
e 9
A-5 TABLE 11 Calculated kerr Values For Separation Experiments Monte Carlo Expt d Type Poison Plate Keff 6(STD Deviation) 15 None 1.00227
.00534
[
04 None 0.99912
.00540 49 None 1.00221
.00473 18 Mone 1.00813
.00489 21 None 0.99589
.00461 28 304 5 Steci 0.0 w/o Borcn 1.00393
.00308 05*
304 5 Steel 0.0 w/o Baron 1.00329
.00303 29 304 5 Steel 0.0 w/o Boron 1.00271 00302 27 304 S Steel 0.0 w/o Boron 1.00418
.00273 26 304 5 Steel 0.0 w/o Baron 0.99811
.00279 34 304 5 Steel 0.0 w/o Boron 0.99793
.00297 35 304 5 Steel 0.0 w/o Boron 1.00436
.00290 32 304 S Steel 1.05 w/o Boron 0.99970
.00524 33 304 S Steel 1.05 w/o Baron 1.01173
.00491 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 6
I t
4 e
l 1G437
!