ML20148P783
| ML20148P783 | |
| Person / Time | |
|---|---|
| Site: | 07001308 |
| Issue date: | 07/23/1975 |
| From: | Stevenson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20148P758 | List: |
| References | |
| NUDOCS 7811290042 | |
| Download: ML20148P783 (14) | |
Text
- _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
i JUL 2 31975 i
Mote to Files CEMERAL ELECTRIC C0!!PAHY (GE), !! ORRIS, ILLIUOIS,'REVISEDTUtrr:nunnuu .m
!!IDNEST FUCL RECOVERY PLAliT (!fFRp), DOCRET 70-1308, FEDRUARY 28,!IARCH 13, APRIL 14, A!TD JUME 2 (RECEIPT DATE FOR CRITICALITY SAFETY AttALYSIS),1975 The subject submittolo concern a proposed amendment to authorize revision of the storanc arrangements for irradiated fuel at IIFRP and storage of approximately 750 metric tons urantun instead of the presently authorized 100 tons. The new arrangement would be based on storage of BWR assemblico in 8 inch diameter, Schedule 10 stainlesa steci pipco and storage of PWR assemblics in 12 inch diancter, Schedulo 5 stainicos steci pipes. BWR asocmblics wnuld he limited to those of 1.40 maximum k while PWR assemblics sould be lh11ted to 1.35 maximum. The limiting h. values are to be calculated for asacnblics under the conditions of spacing in the reactor. The maximum keff value for the stored fuel in any case is not to exceed 0.95.
Initial nucicar ctiticality esfety revicw of the et _ittals han revealed unique features in the proposed controls. Some discussion of these unique features is needed to provide important perspective on the questions (attached list) arising from the review and is given in the following paragraphs.
(In view of the apparent technical campicxity of the review and to provide a more detailed record than would be appropriato in the body of this note, the description of the nucicar criticality safety review effort to date is presented as an attachment.)
The GE proposals are unique in at 1 cast two respects and these unique features have significant implications:
- 1. Limits on the fuel to be stored are based, as aircady noted, on k-infinity values for the two najor light water reactor types. In the writer's experience with controls for irradiated fuci storage, it has been customary to define the specific limiting parameters of the fuel, such as enrichnent, bundic size, rod diameter, pitch, etc. The controls of the latter type could of cource be readily established for storage at a particular reactor, whereas the varied fuel to be stored at itFRP makes limiting parameter definition more difficult.
If one excludes alternatives such as defining all fucis that ocet the k.
limits or limiting storsac to specific conforming fuel asceablics, the k, approcch secen to require that the ifconsce assume responsibility for k.
calculations. There would be a corresponding need for the licensee to define
, 7 8113 9004L4 , ,
J j
9 a control system for auch calculations in the application. For DWR ,
neocnblics this responsibility would appear to be rendered negligibic by the claim that the reactivity of the otored fuel is low at the maximum ,
k=. It should also be noted that GE is the sole U.S. manufacturer of BUR's.
Sinc $ the basic limit is to be the reactor k. value, it is neccesary for the applicant to demonstrate that the stored fuel is cuberitical when the k. limit is met.
- 2. The safety margin calculated for the proposed PUR storage appears to be snaller than that for DUR storage and a significant fraction of PUR assemblics may not meet the proposed k. limit for PUR asocmblics without '
allowance for burnup (the apparently smaller anfety margin incrococe the i inportance of necting the k. limit) . In the absence of calculations to support a icos restrictive approach, the acceptability of a burnup allowance would have to depend upon the control of individual assemblics by the utilitics whoce irradiated fuci is to be stored. If a limited fraction of PWR assemblica do not meet the h.11mit, it. would be far simpler to exclude storsgo of non-conforming assemblics than to intro-duce the compicxitico attendant on burnup allowance. (This comment on '
burnup applies to normal fuel storage conditions; there would be no cause for concern in assuning some defined and established icvel of burnup, for which the liconecc would be responsibic, to analyze such rare effecto as those postulated for tornado missilca.) .
It should be recognized that the significance of the potentini probleno >
identified in the foregoing paragraphs is blurred by a shortage of informa-tion in the GE application concerning details of administrative controlo, ,
types and quantitics of fuel to be stored, and other variabica. It should also be recognized that noot of the variables of significance to nucicar criticality acfety have been identified and analyzed in the subnittal roccived June 2, 1975. This submittal is the " Criticality Safety Basis for the !!FRP Project-I, Fuc1 Eundic Storage Baskets," a report summarizing t the analysis conducted for GE by Battelle Pacific Northwest Laboratories.
Perunal of the attached record of our review to date will show that we have generally confirned the validity of the information in the Battelle analyclo.. The attached questions arising from our review nay thus be rouably characterized as falling into two broad types:
1 4 '
- 1. Rcquests to fill what appcor to be' gaps in the Battelle scalysis.
- 2. Roquests for additional information necessary to clarify CE's intentions and related to unique features of the proposed amendment.
Robert L. Stevenson Fuci Cycic Licensing Branch 1 Division of thtorials and Fuci Cycic Facility Licensing Attachments:
- 1. Record of Initial liucicar Criticality Safety Review of Proposed Reviolon to ifFRP Pool Storage, Docket 70-1308,
- 2. Questions Concerning 11ucicar Criticality Safety, IfFRP Storage Pool Revision, Docket 70-1308. !
l l
cc: F. M. Empoon Distribution:
HMSS R/F MF:FCL#1 R/F ,
WTCrow .
LCRouse JMiller RLStevenson
- orrier
- MF - #1 .. _
.....c, RLSte eTson:mjb , ,_
. n . , , . 3, 7/33/75 - - ..... .
b
JUL 2 3 1975 1
1
, l Questions Concerning Uuclear Criticality Safety, MFRP Storage Pool Revision, Docket 70-1308 i
- 1. (Concerning p.3 of 2-28-75 appl.) What restrictions will be imposed on fuel not meeting Design Criterion 27
- 2. (Concerning p.3 of 2-28-75 appl. and proposed item 6A revision of SNM-1265 as given in the cover letter) a.) The application should be supplemented with a description of the administrative and technical controls to ensure that the
- k. limits are met. The controls should be sufficient to ensure fulfillment of the double contingency policy. b.) There appears to bc.a significant difference in the safety margin for BWR assembly storage and the margin for PWR storage, although some of the apparent margin for BWR storage may be needed to compensate for the dif ferences between assembly km values and rod lattice k='s. What was the basis for proposing a maximum PWR km value of 1.35? Does this permit
" unrestricted" storage of any particular fraction of PWR assemblies? What f raction of PWR assemblics to be stored have an initial enrichment of no more than 2.8%?
- 3. (Concerning p.3 of 2-28-75 appl.) Full details should be provided of the controls proposed to be applied to ensure meeting Criterion 5. However, in the absence of calculations to support a less restrictive approach (note comment no. 7 below), the acceptability of a burnup allowance would have to depend upon the control of individual assemblics by the utilitics whose irradiated fuel is to be stored. If a limited fraction of PWR assemblics do not meet the k. limit, it would be far simpler to exclude storage of non-conforming assemblies than to introduce the compicxities attendant on burnup allowance for normal fuci storage.
1 1
l i
- 4. (Concerning p. 10 of 2-28-75 appl.) The Battelle calculations for BWR assemblics were made for storage in 8-inch diameter, Schedule 10 SS pipes, so that the reference to Schedule 5 in the second paragraph appears to be a typographical error.
- 5. (Concerning p. 14 of 2-28-75 appl.) A reference should be given that domon-strates that current' BWR's are limited to fuel of maximum 1.30 k. , in support of the text of the middle paragraph.
- 6. (Concerning p. 15 of 2-28-75 appl.) Please provide a demonstration to show
, that a singic cycle of irradiation in a reactor will reduce the discharge reactivity to below k.1.25 for PWR fuel with a cold clear km of 1.39.
- 7. (Concerning' p. 16 of 2-28-75 appl.) a.) It was not demonstrated that accidental introduction into storage of one or more essentially unexposed PWR assemblies corresponding to a high k value is not credible. The analysis should therefore include demonstration that such an event would leave the storage array suberitical. b.) What is the purpose of the statement in the second paragraph concerning decay of short-lived fission products? c.) Picase provide copics of the BPNL correlat' ions referred to in the third paragraph, d.) (first paragraph) Who determines that the utility analyses are " equivalent to those employed by the fuci designer" and what criteria (e.g., concerning internni rev iewn of cnicul ntionn, fuct nnnembly controin, etc.) will be applied? Picaso provide information on the practicos of the utilitics from whom fuel will be roccived to show that this determination will be sufficicht to ensure that fuel reactivity limits will be met. Arc there no utilitics
, whose analyses are not "cquivalent"? It should be confirmed that no fuel will be' accepted f rom utilitics who do not meet the specified criteria.
ap* . MINb4 % me. sees * * = am-eae.- e e
. - -. .= - - -- . .
4..
)
l 1
- 8. (Concerning NED0-20825, p.6-5, para 6.3) It should be confirmed that the changes to be accomplished in accordance with para 4.7.1.0 (" Facility Change Control") of the FSAR, NED0-10178, exclude alterations that would affect the nucicar criticality safety of the proposed storage arrangements.
9.- (Concerning NED0-20825, second para. under 3.4.1.2, p.3-12) It should be demonstrated that the contents of a basket of assemblics could not be I
dropped out of the basket in the fuel unloading pit.
- 10. (Concerning NEDO-20825, p. 3-13, para. 3.4.1.3 and Missile Impact) Picase provide information on the analysis of three PWR assemblies in closc-packed configuration (details or specific reference to location in docketed informa-tion). It is not cicar how the calculation leads to the conclusion that three compacted PWR assemblies of K,1.35 would also be suberitical as part of the storage array. The logic should be explained for this conclusion and the one concerning infinite arrays of PWR assemblies at k.< 1.25 and spacings as low as two inches.
- 11. (Concerning NED0-20825, p. 3.13, 3rd para.) It is stated that subcritical conditions are ensured so long as fuci is maintained intact within the basket. To support the conclusion, please provide additional information on the deformation of the fuci expected in a drop accident. Additional nucicar criticality analyscs will be needed if significant deformat' ion is predicted.
- 12. (Concerning the Battelle report of May, 1975, p.12) The analysis for the storage od BRR assemblics purports to hold for the worst caso envelope.
4
The statements concerning evaluation for each bundle design thus require explanation as to their practical implications. This explanation should be part of a broader explanation of the demonstrated relationship between fuel rod lattice k. and the k. values to be calculated by the utilities. To permit ,
1 our confirmation of the relationship, we should be provided with information on the reactor and fuel assembly components, as well as fuel assembly spacing, and any other factors that are included in the calculation of the k= values that are to serve as limits.
13.- (Concerning the Battelle report of May, 1975) If storage of fuel with rods of dif fering enrichments in the same assembly is planned, additional informa-tion should be provided concerning such assemblics (such as method of analysis and validation).
- 14. (concerning the Battelle report of May, 1975) How was it determined that the reference fuel assembly designs (Table 1, p.4) represent the limiting cases?
l
- 15. (Concerning the Battelle report of May, 1975) The rods of highest enrichment '
in the experimental criticality data (Tables B.2 and B.3) directly used in validating the calcula'tional method were 2.73% enriched and were clad in stainicss steel. Considering that the stainicss steci cladding caused a marked reduction in reactivity, what was the basis for concluding that additional validation calculations were not needed for the higher enrich-ment range?
. - - _ - - _ - . . ~ - - - -
. .. . . .. _ ,. . ..= ,, . ,_ -., - _ . . , ._.
- 16. (Concerning all submittals) N'o information is provided on procedures and equipment for handing fuel assemblies with cladding failures. Such informa-tion and relevant safety analyses should be provided. Information should also be provided on the basis for nuclear criticality safety following accidental dropping of a fuel assembly on top of the stored fuel.
r 4
8 I
', , JUL -2 3 E/5 l
RECORD OF INITIAL NUCLEAR CRITICALITY SAFETY REVIEW !
0F PROPOSED REVISION TO MFRP P0OL STORAGE.
DOCKET 70-1308
- 1. Documents Reviewed (This is the list of major submittals read by the l l
reviewer;. additional.. references were used in reviewing particular details and these are identified where such details are discussed). 1
-a.- GE letter dated February 28, 1975, with attached " Nuclear Safety
' Evaluation, Morris Fuci Storage Expansion," February,1975. (Pa8es 4, 8, 9,10,11,12 end 13 were subsequently revised or deleted May 1975.) J
- b. GE letter dated March 13, 1975, with attached NED0-20825, " Safety Evaluation Report for Morris Operation Fuel Storage Expansion Phase I,"
March 1975.
- c. GE letter dated April 14, 1975, with attached report by Programmed and Remote Systems Corporation of St. Paul Minnesota, " Fuel Storage
~
System Design Report, G.E. Morris Operation, Job No. 74766," April 1975.
- d. Report received from GE on June 2, 1975,* prepared by Battelle Pacific -
Northwest Laboratories (BPNL), " Criticality F fety Basis for the MFRP Project-1 Fuci Bundle Storage Baskets," May 1975.
- 2. Review Effort on February 28, 1975 Submittal
- a. Check of fuel rod lattice kcoversus enrichment - Data points for PWR type rods calculated by A.T. Clark and C. Marotta were graphed along with the CE points and reasonabic agreement found.
4 Calculations reported by R. H. Odegaarden and C. R. Marotta ware also used to check the accuracy of the GE calculations for the rod lattice km of the PWR type rods. (See CONF-740901 Part 2, pages 616-633, Figure 1 in " Criticality Calculation Method for Multi-Element LWR
o .
=
Fuel Shipping Packages," in the Proceedings of the Fourth International Symposium on Packaging and Transportation of Radioactive Materials,"
, September, 1974.)
The rod lattice km versus enrichment values'from GE were also generally confirmed by reference to Figure III-13 in WCAP-2999, " Criticality Control Parameters and Supporting Calculations for Uranium Bearing Nuc1 car Fuel at Low Enrichment" by P.G. Lacey, October 1966.
.The rod lattice km for a BWR lattice at 3.2% enrichment was calculated by the reviewer to be 1.417, which compares reasonably well (2.5%
difference) with the value of 1.381 obtained by linear interpolation of the GE data. However, in view of the apparent conservatism of the reviewers PWR lattice km calculations, the 1.417 value should be the subject of further review.
- b. Check of keff's for stored fuel - The reviewer made calculatins for PUR fuel at three different enrichments under the proposed storage conditions (assemblics centered in SS pipes of various wall thicknesses, squarc lattice pattern). These calculations indicate that the PWR assemblics should be subcritical at enrichments slightly above 3%.
Subcriticality was'also indicated for a lattice of PWR assemblics at 2.65% cnrichment af ter severe deformation of the lattice and with icss than one-half the stainicss steel in the pipes.
I 1
~
A single calculation for nominal storage conditions (in 8 inch, Sched. 10 SS pipes) for BWR assemblies at 3.2% carichment gave a koff of 0.777, in good agreement with a value of 0.792' reported by CE for 3.42% enriched fuel of the same design. .
- 3. Review Effort on NEDO-20825, March 13, 1975 j l
- a. Check of basket geometry - A discrepancy was noted between the PWR l assembly guide design shown schematically here (Figure 3-2)and that shown in the later submittal of April 14, 1975. Howev.er, see para. 4 1
below. _
j
- b. , check of maximum PWR fuci enrichment - It was found that a significant' i l
fraction of PWR fuel assemblies with zircaloy clad rods have initial l
enrichments in excess of 3%, e.g., about one-third of the assemb for the Prairic Island Plant of Northern States Power are to be 3.4%-
l enriched. (
Reference:
application dated February 22, 1972 under l Docket 70-1315). The corresponding rod lattice k. at 3.4% is well above the proposed reactor km limit of 1.35 for PWR assemblics,
- c. Check of missile impact offects - Reference'to one document (note to files by A. T. Clark dated April' 26, 1973, Docket 70-1308) for the existing storage arrangement did not provide sufficient explanation for the criticality calculations mentioned in para. 3.4.1.3. Instead of the reviewer's making a further scarch of the docket, licensee will be requested to provide the information.
s
s ,
9 d Possible dropping of fuel assemblics from new baskets - This accident should be precluded by th'e requirement in the 5th paragraph on page 5-2 and Crit'erion No. 15 of the April 14, 1975 submittal. Th,ese requirements are consonant with present fuel handling practices. However, some doubt is raised by the text of the penultimate paragraph on page 3-12, which should be resolved by the applicant.
- e. Check of possible ' storage of PWR assemblies in the BWR baskets - In addition to the limits imposed by the fuel assembly guide plate in the baskets, the pipes in the baskets are too small for insertion of the smallest PWR assemblics.
- f. Check of possible storage of BWR assemblies in the PWR baskets - The low reactivity of the BWR assembly makes this possibility of little criticality safety concern. Physically, the most reactive arrangement would consist of pairs of BWR assemblics, but these pairs would be effectively isolated from other pairs and pairs would be individually sub eritical .
- 4. Review Effort on April 14, 1975 Submittal The only potential problem in this document relevant to criticality safety has to do with the assembly guides, aircady noted in 3(a) above, as revealed
.by Drawings D-17973-D and D-17974-D and sectional views in A-17988D and A-17987D. The concern with the guides is partially obviated by the calculations for "close packed" arrays since the calculations assume that the guides only ensure orientation of assemblics but do not ensure any spacing. Calculations indicate that the guides should ensure proper orientation of the smaller assemblics of each type.
[. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
o ;
s
- 5. Review Effor't on BPNL Report of May, 1975 The items reported under para. 2, above, also apply since this report is a basic source for the February submittal'. j
- a. Check of rod bundic versus infinite rod inttice K='s - The relationship l l
between rod bundic km and rod lattice km was not adequately demonstrated l (will be' the subject of a question to the license applicant) and addition l information will be required before independent check calculations can be made.
- b. Check of effect of thinner wall SS pipes on keff's for stored fuci - i The_ reviewer's independent computer calculations indicate that the conclusion on p. 13 of the Battelle report is correct.
- c. Effect of temperature on keff of fuel basket arrays - No' check calcula-tions were necessary because results of the. Battelle calculations were expected on the basis of (1) the reviewer's calculations for more closely spaced fuel assemblics, (2) the undermoderation designed into the fuel assemblics so that decreased isolation is compensated by decreased fuel assembly reactivity, and (3) similar results were calculated by Nucicar Fuel Services for a comparabic situation (submittal of March 29, 1974 under Docket 50-201).
.d . Check of effect of instrument tubes (water holes) on keff of PUR fuel bundic array _ - Three sets of calculations were used to check out the Battelle conclusions reported on p. 15 of the Battelle report:
(1) Rod lattice km values were compared on the basis of uniform smearing; e
of the water from the holes (a) using k. data from WCAP-2999 and (b) f using k data from the paper from CONF-740901, referenced in 2.a.
I l (2) Computer calculations were made by C. Marotta of Transportation
' Branch'for assemblies with and without water hoics.
(3) Calculations for comparable situations (and conclusions similar to those of Battelle) were reported by Babcock and Wilcox in the applicatiof dated November 21, 1974, under Docket 70-824.
- c. Effect of SS pipe ovality on keff of fuci bundic arrays - The modest effects reported by Battelle were anticipated on the basis of the calculations for arbitratily chosen smaller pipes.
- f. Effect of basket spacing on koff of fuel bundle arrays - All independent check calculations were based on a very large square lattice of assemblics in SS pipes in contac,t so that additional I
calculations of the type reported by Battelle (p. 17) were J unnecessary,
- g. Effect of burnup and fission products on fuel bundle reactivity - No attempt was made to confirm.the burnup effects (which could be compared with information from other sources) because of the uncertainty on the licensco's intentions.
The' accuracy possibic in such calculations was checked by reference to the literature, particularly the document " Reactor Burnup Physics, Proceedings of a Panel, Vienna,12-16 July 1971" published by IAEA,1973
__ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ . _ _ .