ML20083K543
| ML20083K543 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 03/13/1984 |
| From: | Bauser M FLORIDA POWER & LIGHT CO., NEWMAN & HOLTZINGER |
| To: | Cole R, Lazo R, Luebke E Atomic Safety and Licensing Board Panel |
| References | |
| OLA, NUDOCS 8404160209 | |
| Download: ML20083K543 (39) | |
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4 NEWMAN Sc HOLTZINGER, P.vC. ,,4 -
1025 CONNECTICUT AVENUE,N.W.
JACn m.NEwMAN WAS H I N GTON, D. C. 2O O3 6 wituAM E.a&En.Jn.+
JOHN E.MO6TriNGEn,Jm. _ E.ontoon? BARNES MaxOLD r.ntiS - -
uoLAS L.eEnttrono M&uniCE ARELRAD J. A. Boun NiGMT, J a. \ [} N ET E. S.ECn E N TEVEN p.rnANTI PiuLM.sECn SRI AN R.GISM O EORG E L. E DG Am STEVEN C.OOLDBERG MATHLEEN M.SMEA . .- JILL E. GRANT j DAVID G.PowCLL retotaiC S.OmAv oOv LASo.o EEN March 13, 1984 - MOLLvN.L'NoEMAN mA800L LYN NEWMAN R EVIN J. UPSON JOH N T. STOUCM.Ja. DAVID S.RASalN JAM ES 5.VASILE JANE 4.AYAN MeCMAEL A.eAUSEm DON ALD J. SILVERM AN ALVIN M.GuTTERM AN JACOLv4 A.SsMMONS TMOM AS A.SCMMUT2 JOSEPH E. STUB AS NOgEnf 1.wMaTE NOBERTLowtNSTEsN of couantti
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Dr. Robert M. Lazo, Chairman Dr. Richard F. Cole Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing i Board Panel Board Panel U.S. Nuclear Regulatory U.S. Nuclear Regulatory 1
Commission Commission Washington, D.C. 20555 Washington, D.C. 20555 Dr. Emmeth A. Luebke Administrative Judge Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission j Washington, D.C. 20555 i
1 Re: In the Matter of FLORIDA I
POWER AND LIGHT COMPANY (Turkey Point Plant, Unit Nos. 3 and 4) - Docket i Nos. 50-250, 50-251 OLA-
Dear Administrative Judges:
Page 11 of " Licensee's Response to Petitioners' Brief" (Licensee's Response), dated March 14, 1984, states that "No approval to utilize fuel of a higher enrichment has been sought, however, and licensee has not even made a decision as to whether i to seek such approval." The purpose of this letter is to inform the Board and the parties that, under letter dated April 4, 1984,
, Florida Power & Light Company (FPL) requested a set of operating license amendments to accommodate the possible use of higher fuel enrichments at Turkey Point in future cycles.
A cooy of the application is enclosed. As stated at page 12 of Licensee's Response, FPL views this separate amendment appli-cation to be outside of the scope of this proceeding. However, 8404160209 840313
e 1 l
- NEWMAN & Hor.Tztwozo,P. C.
l l
Dr. Robert M. Lazo Dr. Emmeth A, Luebke Dr. Richard F. Cole March 13, 1984 Page Two this notice is provided anyway to keep the Board and the parties fully and completely apprised.
Sincerely, pl*.-,(/.,-_____._
Michael A. Bauser
q I O ,
. . FLORIDA POWER & LIGHT COMP ANY N APR 16 m :18 April 4, 1984 L-84-92 Office of Nuclear Reactor Regulation Atiention: Dorrell G. Eisenhut Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Re: Turkey Point Units 3 & 4 Docket Nos. 50-250 and 50-251 Proposed Amendment to Focility Operating License DPR-31 and 41 .
Fuel Storage U-235 Linear Loading Increase ..
Dear Mr. Eisenhut:
in accordance with 10 CFR 50.90, Florida Power & Light Company (FPL) submits herewith three signed originals and forty copies of a request to amend Facility Operating Licenses DPR-31 and 41.
FPL proposes to modify the existing Turkey Point Units 3 and 4 U-235 linear loading and delete the reactor core U-235 enrichment specification to accommodate storage of higher enrichments for possible use in future fuel cycles.
The proposed modification will involve Technical Specification changes os described below and as shown on the occompanying Technical Specification pages.
(Attachment 1).
Technical Specification Pope 5.2.1 The reactor core description specification is modified to reflect the deletion of the enrichment restriction.
Technical Specification Poqe 5.4.1 The fuel storage specification is modified to reflect the proposed increase in fuel storage U-235 linear loading and increase the limiting reactivity in the new fuel storage creo.
The proposed facility modification has been reviewed by the Turkey Point Plant :
Nuclear Safety Committee and the FPL Company Nuclear Review Board. FPL has concluded that operation of the Turkey Point Plants under the proposed license omendments would not:
- 1. Involve a significant increase in the probability or consequences of on occident previously evoluoted; or i
PtoPLE , .stRVING PtoPLE l
r s
Page 2 Office of Nuclear Reactor Regulation
! Mr. Dorrell G. Eisenhut Division of Licensing 1
- 2. Create the possibility of a new or different kind of accident from any occident I
previously evoluoted; or
- 3. Involve a significant reduction in a margin of safety.
Attachment 2 provides on evaluation of the proposed oction in light of three stonderds i contained in 10 CFR 50.92 (listed above, regarding the issue of no significant hazards l consideration.
In summary, FPL submits that the activities ossociated with the proposed amendments do not constitute a significotnt hozord to the public health and safety or to the environment and, therefore, that these omendments do not involve a significont hozords consideration. FPL respectfully requests therefore, that this opplicotton be processed pursuant to 10 CFR 50.91 and 50.92, os involving no significant hozords consideration.
4 Approval of these technical specification changes is desired by August I,1984 in order to meet our enrichment notification schedules to allow use of enrichments greater than 3.5 w/o in Turkey Point Unit 4, Cycle i1.
4 In accordance with 10 CFR 170.22, this proposed license amendment has been determined to represent a Class I and lll omendment. Accordingly, o check for j $4,400.00 will be forwarded under separate cover.
Very truly yours,
\ /= .
' d&Ws J. W. Williams, Jr.
Vice President Nuclear Energy
! JWW/ERK/ cob J
cc: Mr. J. P. O. Reilly Mr. Harold Reis, Esquire Lyle E. Jerrett, Administrator Radiological Health & Rehabilitative Services 1323 Winewood Boulevard
! Tollohossee, Florido 32301 i
Attachments i
! l. Proposed Technical Specification (pages 5.2 1,5.4-l) i l 2. No Significant Hazards Consideration l
- 3. Safety Analysis Report ;
J STATE OF FLORIDA )
) ss.
COUNTY OF DADE )
J. W. Williams, Jr. _, being first duly sworn, deposes and says:
of Florida Power & Light Company, the That he is a Vice President Licensee herein; That he has executed the foregoing document; that the statements mode in this document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute tne document on behalf of said Licensee, h h Y d 2 /~ v J. W. Williams, Jr.
Subscribed and sworn to before me this 4' . day of Anrt _, l 91L.
J _a Oy)M N h up y- g NOTARY PUBLIC, in and for the County of Dode, State of Florido.
M/ commission expires J-/3r-/ff*P_
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l Attachment 1 1
PROPOSED TECHNICAL SPECIFICATION
.__- Turker Point 3 and 4 5.2 REACTOR Reactor Core
- 1. The reactor core contains approximately 71 metric tons of uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy - 4 tubing to form fuel rods. The reactor core is made up of 157 fuel assemblies.
Each fuel assembly contains 204 fuel rods.
- 2. The average enrichment of the initial core is a nominal 2.50 weight per cent of U-235. Three fuel enrichments are used in the initial core. The ..
highest enrichment is a nominal 3.10 weight per cent of U-235.
- 3. Reload fuel will be similar in design to the initial Core.
I 4. Burnable poison rods are in the form of rod clusters,
' which are located in vacant rod cluster control guide tubes, are used for reactivity and/or power distribution control.
- 5. There are 45 full length RCC assemblies and 8 partial length
- RCC assemblies in the reactor core. The full
- Any reference to part-length rods no longer applies after the part-length rods are removed from the reactor.
This amendment effective as of date of issuance for Unit 3 and date of startup, Cycle 10, Unit 4.
5.2.1 l
l l
l
- j
e o Attachment 1 PROPOSED TECENICAL SPECIFICATION Turkew Point Units 3 and 4 i
5.4 FUEL STORAGE
- 1. The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class 1 structures. Each spent fuel pit has a stainless steel liner to ensure against leakage.
- 2. The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations. The fuel in the spent fuel pit is stored vertically
in an array with sufficient center-to-center distance between assemblies to assure Kerg equal to or less than 0.95 with new fuel containing not more than 52.4 grams of U-235 per axial centimeter of fuel assembly even if boron was not added to the pit water.
The fuel in the new fuel storage racks is stored vertically in an array with sufficient center-to-center distance between assemblies to assure Kerg equal to or lesu than 0.98 with new fuel containing not more than 57.7 grams of U-235 per axial centimeter of fuel assembly.
- 3. The boron concentration in the spent fuel pit is that used in the' reactor cavity and refueling canal during refueling operations, whenever there is fuel in the pit, except for initial new fuel storage.
5.4.1 t
I
- . 1 i
i
, ATTACEMENT 2 i
I Eo Siemific==t 8===eds Consideration 1 '
)
i ~ Florida Power 6, Light Company (FPL) presents this evaluation of the hazardrconsiderations involved with the proposed amend- ,
i ment,-focusing on the three standards set-forth in 10CFR 50.92(c) i as quoted below:
l "The Commission may make a final _ determination, pursuant i j to the procedures in 50.91, that a proposed amendment to (
an operating license for a facility licensed under 50.21(b) {
or 50.22 or for a testing facility involves no significant {
i j hazards considerations, unless it finds that operation of the facility in accordance with the proposed amendment j l j
^
would:
j 1. Involve a significant increase in the probability !
or consequences of an accident previously evaluated; l or j 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; '
j or
! 3. Involve a significant. reduction in a margin of safety."
1 l FPL submits that the activities associated with this amendment
- request do not meet any of the significant hazards consideration ;
j standards of 10 CFR 50.92(c) and, accordingly, a no significant i hazards consideration finding is justified. In support of this !
i determination, necessary background information is first provided, i
followed by a discussion of each significant safety hazards '!
i consideration factors with respect to the proposed amendments.
i BackEround The Turkey Point Plants were designed and constructed with t'o w i new fuel storage racks and two spent fuel storage pools, one of each associated with Unit 3 and one with Unit 4. The new
! fuel storage racks have a capacity of 54 new fuel assemblies.
- The spent fuel storage pools had a capacity for 217 spent fuel -
- assomblies (equivalent to 1-1/3 cores).
The Turkey Point Units 3 and 4 Final Safety Analysis Report ,
addressed the safety implications of these facilities and included !
relevant parameters associated with criticality, structural !
l integrity, and cooling. The Turkey Point Units 3 and 4 Sr.foty i Evaluation Report (Docket No.'s 50-250 and 50-251) found the i environmental and safety impacts of storage in these facilities .
l to be acceptable. l i
j
}
In 1976, a request to amend the Turkey Point operating licenses
- for increased spent fuel storage was submitted by FPL. By letter I l dated March 17, 1977, the Commission approved Amendments 23 L and 22 to facility operating licenses DPR-31 and DPR-41, respec-b
tively, for modification to Turkey Point Units 3 and 4 spent fuel storage. facilities. These modifications consisted of rerack-ing the Unit 3 and 4 spent fuel pools with high density fuel ,
storage racks which increased the storage capacity from 217 fuel assemblies to 621 fuel assemblies. Approval of the amend- 'l l
monts included _a detailed review and analysis of all relevant storage parameters and potential accidents. The analyses resulted in a finding that environmental and safety impacts were negligi-ble. !
The safety evaluation performed in support of the request to ;
amend the Turkey Point operating licenses to allow reracking o
[
of the Unit 3 and 4 fuel pools addressed the following:
i
- 1. Structural and Seismic Analysis j
- 2. Nuclear Criticality Analysis
- 3. Thermal-Hydraulic
- 4. Accident Analyses
^
- 5. Radiation Exposures ;
- 6. Spent Fuel Cask Drop Accident It was determined that the proposed modifications to the Unit i 3 and 4 spent fuel pools would be acceptable because: (1) the ,
rack structural design would withstand conditions during normal !
operation combined with the maximum earthquake, (2) the rack [
design would preclude criticality for any moderating condition, I (3) the existing spent fuel cooling system was determined to [
adequately cool the increased heat load and a redundant 1005 l capacity spare pump would be installed, (4) the increased radia- 1 tion doses, both onsite and offsite, would be negligible, and (5) spent fuel cask handling operations would not change from the original design. i The current spent fuel storage capacity at Turkey Point consists !
of 621 storage locations in each spent fuel pool. ;
With this application, FPL is requesting approval to increase [
the U-235 linear loading in all fuel storage areas and delete l the reactor core reload fuel U-235 enrichment specification, as set forth in the attached Safety Analysis Report.
Ermluatina The following' evaluation demonstrates (by reference to the analy-sis contained in the. attached Safety Analysis Report) that the l proposed amendment to increase the~ fuel storage U-235 linear loading does not exceed any of the three significant hasards consideration standards. The analysis of this proposed increase in fuel storage enrichment has been accomplished'using current ,
accepted codes and standards as specified in section 2.1 of' c l the attached Safety Analysis Report. The_results of the analysis 3 meet the specified acceptance criteria in these standards as ;
i
-.---~_m,.---..-----,,,-,-..r-...-_-.-,_-- ..,--b---__~~-,.... , . , - - - -
. o presented in the Safety Analysis Report. The basis of the proposed deletion of the reactor core reload fuel enrichment specification is that this specification is unnecessary and superfluous in that there are other provisions in the Technical Specifications which determine safe operating and fuel storage limits related to fuel enrichaynt. These other safe operating limits include dynamic paramsters, rod worths and peaking factors. In other words, specification of reload fuel enrichment has no bearing on the safe operation of the reactor core provided that existing safety limits and limiting conditions for operation (LCOs) are satisfied.
(1) Involve a significant increase in the nrobability or conse-auences of an accident nreviousiv evaluated.
In the course of the analysis, FpL has identified the following potential accident scenarios:
- 1. A fuel assembly drop in the spent fuel pool.
- 2. Loss of spent fuel pool cooling system flow.
- 3. A spent fuel cask drop.
For 1, "A fuel assembly drop in the spent fuel pool", the critical-ity' acceptance criterion is not violated as identified in Section 3.0 of the Safety Analysis Report. The radiological consequences of this type of accident in the spent fuel pool are bounded by the cask drop accident. Thus the consequences of this type accident will not be significantly increased from previously evaluated fuel assembly drops.
The consequences of 2, " Loss of spent fuel cooling system flow" will not be effected since this application is not intended to qualify the fuel for extended burnup operation. The use of a higher U-235 linear loading by itself will not affect the decay heat characteristics of the fuel assembly or the previous evaluation of the loss of spent fuel cooling system flow. The proposed amendment to increase the fuel storage U-235 linear ~
loading specification will not result in an increase in the probability or consequences of an accident previously evaluated for loss of spent fuel cooling system flow.
The consequences of 3, "A spent fuel cask drop", as previously evaluated will not be affected by an increase in fuel assembly U-235 linear loading since this application is not-intended to qualify the fuel for extended burnup nor does this amendment alter the configuration of the storage rac!cs.
The proposed amendment to increase the fuel storage U-235 linear loading will not result in an increase in the probability or consequences of an accident previously evaluated for a spent l fuel cask drop.
4 i
i l
Thus, it is concluded that the proposed amendment to increase i the fuel storage U-235 linear loading and deletion of the reactor
- core enrichment specification will not involve a significant i
increase in the probability or consequences of an accident previous-ly evaluated..-
j I (2) Create the nossibility of a new or different kind of accident from any accident oreviousiv evaluated.
I FPL has evaluated the proposed technical specification changes in accordance with the guidance of the NRC position paper entitled, "OT position for Review and Acceptance of Spent Fuel Storage and Handling Applications", appropriate j
NRC Regulatory Guides, appropriate NRC Standard Review Plans, and appropriate Industry Codes and Standards as 7 listed in Section 2.1 of the attached Safety Analysis Report.
4 j
As a result of this evaluation, FPL finds that the proposed ,
technical specification changes do not, in any way, create the possibility of a new or different kind of accident j
j from any accident previously evaluated for the Turkey Point
]
Fuel Storage Facilities.
1 i
(3) Involve a significant reduction in a marrin of safety.
The NRC Staff Safety Evaluation review process has established
~
i
' that the issue of margin of safety, when applied to modifica-tion, will need to address the area of nuclear criticality
- considerations.
I The established acceptance criteria for criticality is j
that the neutron multiplication factor, including all uncer-
' tainties, under all conditions:
(a) shall be less than or equal to 0.98 for the new fuel j storage facility; and i
(b) shall be less than or equal to 0.95 for the spent fuel pool. ,
This margin or safety has been adhered to in the criticality analysis methods for the spent fuel and new fuel storage, as 1 discussed in Section 3.0 and 4.0 of the attached Safety Analysis Report.
The methods to be used in the criticality analysis conform with applicable codes, standards, or pertinent sections thereof, as referenced in Section 2.1 of the Safety Analysis Report. ,
l In meeting the acceptance criteria for criticality in the Turkey i Point Unit 3 and Unit 4 fuel storage facilities such that:
- (a) K,gg is always less than 0.98, including uncertainties
! at a 95/95 probability confidence level in th6 new I fuel storage facility.
i J
l (b) Keff is always less than 0.95, including all uncertain-ties at a 95/95 probability confidence level in the spent fuel pool.
Increasing the limiting Keff in the new fuel storage facility to 0.98 is so+ely administrative and consistent with the value established in USNRC Standard Review Plan, NUREG-0800, Section 9.1.1.
The proposed amendment to increase the fuel storage U-235 linear loading and increase the limiting Keff in the new fuel storage area will not involve a significant reduction in the margin of safety for nuclear criticality.
In summation, it has been shown that the proposed increase in the fuel storage facility U-235 linear loading, increasing the limiting Keff in the new fuel storage facility, and deletion of the reactor core enrichment specification does not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; **-
or
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
FpL has determined and submits that the proposed amendments described do not involve a significant safety hazard and that the standards in 10 CFR 50.92 have been met.
A
SS-153 CRITICALITY ANALYSIS OF THE TURKEY POINT PLANTS UNITS 3 & 4 STORAGE RACKS WITH INCREASED ENRICHMENT Prepared for the Florida Power & Light Co.
by S. E. Turner, Ph.D., P.E.
M. K. Gurley 4
e February 1984 c
i I
- e TABLE OF CONTENTS .
Page 1
1.0 INTRODUCTION
AND
SUMMARY
3 2.0 CRITERIA AND METHODOLOGY FOR CRITICALITY ANALYSES...........
3 2.1 D e s i gn 8 a s e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 Reference Fuel Assemb1y............................. 4 2.3 Reference Fuel Storage Ce11.........................
7 2.4 An alyti c al Met hod s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
7 2.4.1 C al cul ati onal Model s . . . . . . . . . . . . . . . . . . . . . .
2.4.2 Calculational Bias and Uncertainty........ 8 3.0 CRITICALITY ANALYSIS OF SPENT FUEL STORAGE RACKS............ 10 10 ,,
3.1 Summary of Criticality Analyses.................... 13 3.2 Uncertainties Due to Manufacturing Tolerances.......
3.2.1 Fuel Enrichment and Density Variations.... 13 14 3.2.2 Inside Cell Dimensional Tolerance.........
Storage Cell Lattice Spacing Variation.... 14 3.2.3 14 3.2.4 Stainless-Steel Thickness Variations......
3.2.5 Eccentric Positioning of Fuel Assembly within Storage Rack....................... 14 15 3.3 Abnormal and Accident Conditions.................... 15 3.3.1 Temperature and Water Density Effects.....
3.3.2 Fuel Assembly Abnormally Located Outside Stora ge R ack . . . . . . . . . . . . . . . . . . . . . . 15 18 4.0 CR ITICALITY ANALY S I S OF NEW FUEL STOR AGE RACKS . . . . . . . . . . . . . .
REFERENCES 1
11 i
f i
a i
4 LIST OF TABLES l
. l Page No. ..
6 1 OPTIMIZED FUEL ASSEMBLY DESIGN SPECIFICATIONS................... l
SUMMARY
OF UNCERTAINTIES IN k, DUE TO TOLERANCES................ 10 l 2
3 INFINITE MULTIPLICATION FACTORS OVER ANTICIPATED RANGE OF 13 FUEL ENRICHM6'iTS AND DENSITIES.................................. r i
se i
I
$li !
I s
L _ -
LIST OF FIGURES No. Page 1 Reference fuel assembly and configuration of spent fuel s tora ge rack for the Tu rk ey Poi nt P1 ant. . . . . . . . . . . . . . . . . . . . . . . . . 5 2 Infinite multiplication factor of spent fuel storage rack for Va ri ous U-235 loadi ngs i n fuel a s sembly . . . . . . . . . . . . . . . . . . . . . . . . . 11 3 Effect of coolant temperature on reactivity of spent fuel storage racks................................................... 12 4 Variation in k with fuel assembly spacing (infinite lattice without stainl,ess 5 teel)........................................ 17 5 Fresh fuel storage rack configuration and analytical model...... 19 6 Reactivity effect of low-density moderator in fresh fuel --
i storage rack wi th fuel of 4.5% enrichmenmt. . . . . . . . . . . . . . . . . . . . . . 20 I
i i
tv
_a_ _ _ _ _ _ _ _ _ _ _ _ ..
l l
1.0 INTRODUCTION
AND
SUMMARY
Both the new fuel and spent fuel storage racks in the Turkey Point plant, Units 3 & .4."fre currently licensed to store fuel of 43.9 grams U-235 per axial centimeter of fuel assembly corresponding to 3.5 wt.% U-235 initial enrichment. The previous criticality analysis, submitted in support of the current Technical Specification limit on fuel enrichment, documented a neutron multiplication factor substantially below the NRC limiting reactivity value of 0.95 including all uncertainties. The evaluation reported here was prepared to justify the criticality safety of an increase in the Technical Specifica-tion limit on *uel enrichment in the existing storage racks for both new and spent fuel.
Results of the present evaluation confirm that the maximum reactivity af the spent fuel storage racks will be less than 0.95, including all uncertain-ties, with the racks fully loaded with fuel containing 52.40 grams U-235 per axial centimeter of fuel assembly and flooded with unborated water at a tem-perature corresponding to the highest reactivity, provided the UO2 stack density is no less than 10.08 g/cm3 (93% of theoretical density). The lim.
iting axial U-235 loading includes tolerances on fuel density and enrichment and corresponds to a nominal enrichment of 4.085 wt.% U-235 at a 002 pellet density of 97% of theoretical.
Criticality safety of the new fuel storage rack (a separate facility located near, but independent of, the spent fuel pool) was evaluated for fuel of 57.72 grams U-235 per axial centimeter, corresponding to a nomi_nal enrich-ment of 4.5% at a U02 pellet density of 971 of theoretical. Transport theory calculations confirm that me neutron multiplication factor under optimum moderating conditions (e.g., fog, spray, or foam) is substantially less than the limiting value of 0.98 specified in SRP 9.1.1, "New Fuel Storage."
Other than criticality safety, increasing the enrichment capability of the spent fuel storage pool and new fuel storage racks does not introduce any significant new or unreviewed safety considerations. On the basis of the analyses and evaluations presented herein, it is concluded that the spent fuel 1
i
pool and the new. fuel storage racks can safely accommodate fuel of 52.40 and 57.72 grams U-235 per axial centimeter of fuel assembly respectively "with no new or unreviewed hazard considerations under the guidelines of significant 10CFR50.92(c).'~~
e e
t 2
2.0 _ CRITERIA AND METH000 LOGY FOR CRITICALITY ANALYSES .
l 2.1 _Design Bases The objective in the spent fuel storage racks for the Turkey Point Plant is to assure that a neutron multiplication f actor (k,gg) equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest antic-ipated reactivity and flooded with unborated water at a temperature corre-The maximum calculated reactivity sponding to the highest reactivity.
includes a margin for uncertainty in reactivity calculations and in mechanical tolerances, statistically combined. such that the true k,ff will be equal to or less than 0.95 with a 951 probability at a 95% confidence level.
Applicable codes, standards and regulations, or pertinent sections .,
thereof, include the following. i e General Design Criterion 62 - Prevention of Criticality in Fuel Storage and Handling.
e NRC letter of April 14, 1978, to all Power Reactor Licensees - Of Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, including modification letter dated January 18, 1979.
e USNRC Standard Review Plan, NUREG-0800, Section 9.1.1, New Fuel Storage, and Section 9.1.2, Spent Fuel Storage..
e Regulatory Guide 1.13 Spent Fuel Storage Facility Design Basis (proposed), December 1981.
e Regulatory Guide 3.41, Validation of Calculational Method for Nuclear Criticality Safety (and related ANS! N16.9-1975).
e ANSI N210-1976, Design Objectives for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants.
e ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.
To assure the true reactivity will always be less than the calculated reactivity, the following conservative assumptions were made.
3 n- - _ - - - - - - - - - - _ _ - . - . _ _ _ .
! e Moderator. is pure, unborated water at a temperature corresponding to t
' the highest reactivity. ,
i no e Lattice of storage racks is infinite in all directions; i.e.,
credit-4 taken for axial or radial neutron leakage (except in the consideration of certain abnormal / accident conditions).
i.e.,
e Neutron absorption in minor structural members is neglected; spacer grids are replaced by water.
o Pure zirconium is used for cladding, control rod guide tubes, and instrument thimbles; i.e., higher neutron absorptica of alloying materials in Zircaloy is neglected.
I 2.2 Reference Fuel Assembly The reference design fuel assembly, illustrated in Fig.1, is a 15 x 15 array of fuel rods (Westinghouse design), with 21 ~ rods replaced by 20 controJ rod guide tubes and one instrument thimble. Two alternative fuel assembly designs have been used in the Turkey Point reactors: an optimized fuel assembly design with Zircaloy grids and an earlier design using inconel grids. The optimized fuel assembly is more reactive and has been used as the reference in the fuel rack criticality analyses. Table 1 summarizes the optimized fuel assembly design specifications and expected range of signi-ficant fuel tolerances.
2.3 Reference Fue1 Storage Cell
~
The nominal spent. fuel storage cell model used for the criticality anal-yses is shown in Fig.1. The rack is composed of 0.25-in, stainless-steel boxes of 8.790-in. inside dimension. The fuel assemblies are centrally located in each storage cell on a nominal lattice spacing of 13.659 in. The outer water space constitutes a flux-trap between the two steel plates. For two-dimensional X-Y analysis, a zero current (white albedo) boundary condition was applied in the axial direction and at the centerline through the outer water space (flux-trap) on all four sides of the cell, effectively creating an i infinite array of storage cells.
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Table 1 OPTIMIZED FUEL ASSEMBLY DESIGN SPECIFICATIONS s
Fuel Rod Data --:-
Outside dimension, in. 0.422 Cladding thickness, in. 0.0243 Cladding material Zr-4 Pellet diameter, in. 0.3659 Axial dishing factor 0.988 002 density, g/cm 3 10.08 - 10.514
% T.D. 93 - 97 Fuel Assembly Data Number of fuel rods 204 (15 x 15 array)
Fuel rod pitch, in. 0.563 Active fuel height, in. 144 Control rod guide tube Number 20 0.0., in. 0.533 Thickness, in. 0.017 Material Zr-4 Instrument thimble Number 1 0.0., in. 0.533 Thickness, in. 0.017 Material Zr-4 6
I
- r- -
1 2.4 Analytical Methods 2.4.1 Calculational Models
.-.?
Four different methods of calculation were used to enhance the credi-I bility of the analysis and to provide assurance that the true reactivity will be less than the limiting value of 0.95 including uncertainties. These methods of calculation include the following.
e CASMO - a two-dimensional multigroup transport theory codel (based upon capture probabilities), which provides the capability for a detailed geometric description of the storage cell and each fuel rod.
e AMPX-KENO - a multigroup Monte Carlo code package,3,4 using the 123-group GAM-THERMOS cross-section set developed by ORNL, and the NITAWL subroutine for U-238 resonance shielding effects (Nordheim integral treatment). AMPX-KENO has been benchmarked against a number of critM cal experiments (Refs. 5, 6, and 7) with generally good agreement for most critical experiments analyzed, although both Refs. 6 and 7 indi-cate an underprediction in reactivity in arrangements with large water gaps between fuel assemblies.
e AMPX-KENO - the Monte Carlo technique describgd above, but using the more recent 27-group SCALE cross-section set devejo by ORNL for criticality safety analysis. Benchmark calculations 'gindicate that the 27-group SCALE cross-section set consistently underpredicts reac-tivity by ~0.012 ok.
e Diffusion / Blackness Theory - a calculational technique based upon the multigroup cell homogenization code NULIF,Il to calculate diffusion theory constants for input to P0097. 2 A small correction, based upon blackness theory, was applied to the macroscopic absorption cross-section calculated by NULIF for stainless steel.
For investigation of reactivity effects due to uncertainties (e.g.,
mechanical and fabrication tolerances), the CASMO code was used to calculate the small incremental reactivity changes. Diffusion / blackness theory calcula-tions were used to estimate the reactivity effects of abnormal / accident con-ditions.
7
2.4.2 Calculational Bias and Uncertainty The infinite multiplication factor (k ) for the Turkey Point spent fuel storage rack is_ based upon CASMO calculations with fuel of the highest antici-pated react'ivity. Supporting calculations with both the 123-group and 27-are group AMPX-KENO code package confi rm that the CASMO calculations conservative. A diffusion / blackness theory calculation (NULIF-PDQ97) provides further confirmation of the conservatism in the reference CASMO calculation.
To illustrate the inherent conservatism of CASMO calculations, one case (57.72 grams U-235 per axial centimeter with the minimum cell box inside dimension) was selected for intercomparison. Results of criticality calculations with the four independent methods of analysis for this case are as follows:
Method Calculated k _ Bias Corrected k _
CASMO 0.9404 - 0.9404 123-gp AMPX-KENO 0.9052 t 0.0037* 0.0315 0.9367 t 0.0037*
(150,000 histories) (95%/95%)
27-gp AMPX-KENO 0.9095 t 0.0063* 0.012 0.9215 t 0.0063*
(50,000 histories) (95%/95%) l Oiffusion/ Blackness Theory 0.9340 - 0.9340 l
7 The 123-group AMPX-KENO calculational model has been benchmarked against critical experiments as nearly representative as possible of the Turkey Point fuel racks. These benchmark calculations indicate a nominal bias of 0.000 ! l 0.003 (95%/95%) plus a correction for the water gap thickness between storage cells. Linear extrapolation of the trend identified in reference 7 to the 4.7-in water gap of the Turkey Point spent fuel rack results in a bias cor- l rection of 0.0315 Ak, although linear extrapolation probably overestimates the water gap correction. Similar benchmark calculations of the 27-group AMPX-
- With one-sided tolerance factor for 95% probability at 95% confidence level.
8
r l . ,
KENO calculational model, reported in references 9 and 10 (and confirmed by independent calculations), indicate a bias of 0.012 Ak, with no apparent trend for the largest water gap (2.58-with water gap thickness observed at least Thus, the storage rack k, de-in.) measured in the critical experiments.
scribed above, as calculated by the Monte Carlo technique, probably lies '
between 0.9215 and 0.9367. Both values confirm the conservatism of the CASMO calculation (k. = 0.9404).
Diffusion / blackness theory calculations would normally be expected to provide a reliable estimate of the rack k , since the stainless-steel wall of the storage cells is a relatively weak absorbing medium. The value calculated by diffusion / blackness theory (k. of 0.9340) ten'ds to further confirm the conservatism of the CASMO calculated value (0.9404). CASMO benchmark calcula-tions on critical experiments representative of the Turkey Point racks (refee-ence 2, Section 2.1.3) suggest an overprediction of 0.006 Ak which, if applied to 'the reference CASMO calculation, would indicate good agreement with the diffusion / blackness theory -value. Nevertheless, for conservatism and to assure the true reactivity is less than the calculated value, the higher k ,
as calculated by CASMO, was used as the reference value for the Turkey Point spent fuel storage rack. All three of the alternate methcds of analysis indicate a lower value and therefore confirm that the reference CASMO value is conservative.
9 I
l l
l 1 9 l
i 3.0 CRITICALITY. ANALYSIS OF SPENT FUEL STORGE RACKS .
i 3.1 Summary of Criticality Analyses l ._.
Criticality analyses for the Turkey Point spent fuel storage racks were performed for several fuel densities and U-235 loadings, (in grams per axial centimeter of fuel assembly) and plotted as shown in Fig. 2. These data show that, for a given U-235 axial loading, the higher reactivity results for the lower assumed U02 stack density of 10.08 grams per axial centimeter. To this latter curve was added the total uncertainty in k, associated with manufacturing tolerances, 0.0243 Ak, as summarized in Table 2, to generate the upper curve in Fig. 2 identified as the " maximum with uncertainties." This curve was then read to determine the limiting U-235 loading of 52.40 grams per axial centimeter corresponding to the maximum allowble k, of 0.95. ..
Table 2
SUMMARY
OF UNCERTAINTIES IN k ,DUE TO TOLERANCES k_ Reference Uncertainties Fuel enrichment 10.0008 Section 3.2.1 Fuel density 10.0006 Section 3.2.1 Cell inside dimension t0.0020 Section 3.2.2 Lattice spacing t0.0230 Section 3.2.3 SS tolerance t0.0015 Section 3.2.4 Statistical combination to.0232 Eccentric positioning +0.0011 Section 3.2.5 Maximum uncertainty +0.0243 The design basis temperature for these calculations was 65.6*C (150*F) ;
which, as shown in Fig. 3 is the coolant temperature of the highest neutron -f multiplication factor. Lower pool temperatures expected for normal opera-tions, as well as neutron leakage from the finite size racks, provide addi-tional margin in k,ff below the limiting value of 0.95 for k , used in the analysis.
1 10
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51 52 53 54 55 56 57 55 GR AM S U-2 3 5 PER AXI AL CM
' Fig. 2 Infinite multiplication factor of spent fuel storage rack -for various U-235 loadings in fuel assembly.
11
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a U-235 axial loading of 52.40 grams per axial centimeter is the maxim 0m which This load-may be acconinodated in the Turkey Point spent fuel storage racks.
at 10.08 grams per ing may be. rhaiized by fuel of 4.261% U-235 enrichment axial centimeter UO2 stack density (93% T.O.) or lower enrichments at higher UO densities (e.g 4.085% enrichment at a UO2 stack density of 10.514 grams 2
per axial centimeter, 97% T.D.).
Credible abnormal or accident conditions will not result in exceeding the limiting keff of 0.95, with credit for the presence of soluble poison (nominally 1950 ppm boron).
3.2 Uncertainties Due to Manufacturing Tolerances n
3.2.1 Fuel Enrichment and Density Variations The maximum loading of 52.40 grams U-235 per axial centimeter can be realized over a range of enrichments and UO2 stack densities. Table 3 identi-fies the range considered and gives the k,, values for three combinations of enrichment and density.
Table 3 INFINITE MULTIPLICATION FACTORS OVER ANTICIPATED RANGE OF FUEL ENRICHMENTS AND DENSITIES Enrichment, Axial Loading k 002 Density g/cm
% T.O. g/cm 3 % U-235 (IASMO) 10.514 4.085 52.40 0.9232 97 10.297 4.173 52.40 0.9244 95 10.080 4.261 52.40 0.9256 93 These da:a show that the highest k,, occurs at a U02 density of 93% of theo-retical and an enrichment of 4.261 wt.% U-235. Thus, the low density case has been assumed as the design basis for the criticality safety evaluation.
Higher 002 densities, at the same axial loading in grams U-235 per axial centimeter, will always yield a lower k, .
l 13
In addition, there is a certain level of confidence to which the fuel '
enrichment and density are known. To evaluate the reactivity uncertainty, it is assumed that the U02 density is known to t0.05 g/cm3 and enrichment to t0.02 wt.% 4J-25. Evaluating the uncertainty for these tolerance limits (by differential CASMO calculations) yields an uncertainty of !0.0006 ak for density and t0.0008 ok for enrichment.
3.2.2 Inside Cell Dimensional Tolerance The stainless-steel inner box dimension, 8.790 t 0.125 in., defines the inner water thickness between the fuel and the inside box. For the tolerance of 10.125 in. on the bor inside dimension, the calculated uncertainty in k, is t0.0020 ak, with k, increasing as the inner stainless-steel box dimen-sion increases.
3.2.3 Storage Cell Lattice Spacing Variation Th9 storage cell lattice spacing between fuel assemblies is nominally 13.659 in. , positioned by a lattice of support grids intended to provide a nominal water gap between cells of 4.369 in. Receipt inspection of the racks confirmed that the water gap between adjacent storage cells is greater than 3.781 in. for all locations, indicating a toleranace of 0.588 in, in lattice spacing. Calculations with this minimum spacing between cells resulted in ar.
uncertainty in k, of 0.0230 ak due to the tolerance in lattice spacing.
3.2.4 Stainless-Steel Thickness Variations The nominal stainless-steel box thickness is 0.25 in. The maximum posi-tive effect on k, of the expected stainless-steel thickness tolerance varia-tion (10.01 in.) was calculated to be 50.0015 ak.
3.2.5 Eccentric Positioning of Fuel Assembly within Storage Cell The fuel assembly is normally located in the center of the storage cell. Nevertheless, calculations were made with adjacent fuel assemblies i 14 i
e
m-moved into the corner of the storage cell (four-assembly cluster at closest approach), resulting in a small positive effect on k,, (0.0011 Ak)." Fuel assembly bowing will produce a smaller positive reactivity effect locally.
The calculated reactivity increment due to eccentric positioning is considered an additive allowance, although eccentric positioning (if any) would normally be expected to be randomly distributed throughout the storage rack.
3.3 Abnormal and Accident Conditions 3.3.1 Temperature and Water Density Effects Increasing or decreasing temperature from the nominal temperature of 150*F (65.56*C) is calculated to decrease k,, in unborated water as indicated in Fig. 3 (reactivity effects calculated by CASMO). At 120*C (248'F), intro-ducing voids in the water internal to the storage cell (to simulate boiling) further reduced k,,' indicating a negative void coefficient of reactivity at the boiling temperature. Voids due to boiling will not occur in the outer (flux-trap) water region.
I 3.3.2 Fuel Assembly Abnormally Located Outside Storage Rack To investigate the possible effect of a fuel assembly abnormally located outside the rack, diffusion calculations were made for unpoisoned assemblies l separated only by water. Figure 4 shows the results of these calculations.
i From these data, the infinite multiplication factor will be less than 0.95 for any fuel assembly spacing greater than ~15 in, in the absence of any soluble poison or neutron-absorbing material other than water between assemblies.
For a drop on top of the rack, the fuel assembly will come to rest hort-zontally on top of the rack with a minimum separation greater than 15 in.
( ~24 in.). Consequently, fuel assembly drop accidents will not result in an increase in reactivity above that calculated for the infinite nominal design storage rack.
15
e .
An extraneous fuel assembly cannot physically be positioned outside the rack between the rack and pool wall or between rack modules. However, it is possible, although not likely, to position an extra fuel assembly adjacent to the rack in th region of the cask area. Two-dimensional PDQ calculations show that $ fresh fuel assembly positioned adjacent to the storage rack can increase the reactivity to 0.957 in the absence of soluble poison. However, soluble boron of ~1950 ppm is normally present in the spent fuel pool (for which credit is permitted under accident conditions)* and would reduce the maximum k to substantially less than 0.95. Consequently, it is concluded that the postulated accident conditions will not adversely affect the criticality safety of the Turkey Point spent fuel storage racks.
An implementing Technical Specification for 1950 ppm soluble boron has been
- submitted via L-84-71, dated March 14, 1984.
16
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4.0 FRESH FUEL STORAGE RACKS ,
The fresh fuel storage racks for the Turkey Point Plant consist of an "L"-shaped .. array of storage cells containing 54 cells on a 21-in. lattice ,
spacing. Figure 5 illustrates the fresh fuel storage cell arrangement and shows the geometry used in the criticality analysis. Although fuel is nor-mally stored in the dry condition, the criticality analysis considered flooding with clean, unborated water ranging in density from 1.0 to very low hypothetical values (e.g., fog, mist, or foam). Preliminary survey calcu-lations with diffusion theory suggested a second maximum in reactivity peaking at a hypothetical water density of ~10%-15%.
The criticality safety of the new fuel storage racks was evaluated for fuel of 57.72 grams U-235 per axial centimeter of assembly ( ~ 4.55 enrichment). Since diffusion theory is known to be inadequate in very dry i lattices, three-dimensional AMPX-KENO calculations were used in the low-moderator-density region to define the maximum keff under optimum moderating conditions. For these calculations, the array of fuel storage cells, as indicated in Fig. 5, was assumed to be reflected by full-density water on the outer boundaries and on both top and bottom of the array. Low-density water f was used within the storage boxes and between the array of storage cells. The XSDRNPM routine in the AMPX code package was used to homogenize the fuel assemblies for each moderator density calculated and to generate the weighted cross-section set for use in KENO.
Figure 6 shows the calculated keff values as a function of moderator density within and between the storage cells. These calculations indicate a low-density maximum reactivity of ~0.925 occurring at a water density of 0.10 g/cm3 . This low-density maximum reactivity is approximately the same as that for the fully flooded condition. In either event, the reactivity is sub-stantially less than the limiting reactivity of 0.98 specified in SRP 9.1.1 under optimum moderating conditions. Hence, it is concluded that unirradiated 7
fuel with a loading of 57.72 grams U-235 per axial centimeter (4.5% enrich-ment) may be safely stored in the new fuel racks of the Turkey Point Plant. ,
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I REFERENCES ,
! 1. A. Ahlin and M. Edenius, CASMO - A Fast Transport Theory Depletion Code
! for LWR ' Analysis, ANS Transactions, Vol . 26, p. 604,1977.
o CASMO-2E Nuclear Fuel Assembly Analysis, Application Users Manual, Rev.
A, Control Data Corporation,1982.
j 2. E. E. Pilat, Methods For the Analysis-of Boiling Water Reactors (Lattice
- Physics). YAEC-1232, Yankee Atomic Electric Co., December 1980.
t
- 3. Green, Lucious, Petrie, Ford, White, Wri ght, PSR-63/AMPX-1 (code package), AMPX Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENOF/B, ORNL-TM-3706, Oak Ridge National
' Laboratory, March 1976.
4 j 4 L. M. Petrie and N. F. Cross, KENO-IV, An Improved Monte Carlo Critical-j ity Program, ORNL-4938, Oak Ridge National Laboratory, November 1975.
~
- 5. S. R. Biermag Enriched al., Critical Separation Between Suberitical Clusters' i
of 4.29.wt1 U UO Rods in Water with Fixed Neutron Poisons,
{
NUREG/CR-0073, Battelle Paci,fic Northwest Laboratories, May 1978, with l
errata sheet issued by the USNRC August 14, 1979.
t 1 6. M. N. Baldwin et al., Critical Experiments Supporting Close Proximity
! Water Storage of Power Reactor Fuel, BAW-1484-7, The Babcock & Wilcox Company, July 1979.
I
- 7. S. E. Turner and M. K. Gurley, Evaluation of AMPX-KEN 0 Benchmark Calcu-lations for High Density Spent Fuel Storage Racks, Nuclear Science and Engineering,80(2): 230-237, February 1982.
l 8. R. M. Westfall et al., SCALE: A Modular Code System for Performing j Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, 1979.
i
- 9. R. M. Westfall and J. R. Knight, Scale System Cross-Section Validation with Shipping-Cask Critical Experiments, ANS Transactions, Vol. 33,
! p. 368, November 1979.
- 10. 8. F. Cooney, T. R. Freeman, and M. H. Lipner, Comparisons of 1 Experiments and Calculations for LWR Storage Geometries, ANS i Transactions, Vol. 39, p. 531, 1981.
L j 11. W. A. Wittkopf, NULIF - Neutron Spectrum Generator, Few-Group Constant Generator and Fuel Depletion Code, BAW-426 The Babcock & Wilcox l
Company, August 1976.
- 12. W. R. Cadwell, PDQ-7 Reference Manual, WAPD-TM-678, Bettis Atomic Power l
Laboratory, January 1967.
1 1
i
~
9 m
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION N D USNRC BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 14 Ast16 21:18 In the Matter of ) . . .
) 5"r 4 0F SElatiAn FLORIDA POWER & LIGHT COMPANY )
)
Docket Nod hh 50-251 OLA (Turkey Point Nuclear )
Generating Units 3 and 4 )
)
CERTIFICATE OF SERVICE I hereby certify that copies of a letter to the Members of the Licensing Board in the above-captioned proceeding, dated April 13, 1984, and attachments thereto were served on the fol-lowing by deposit in the United States mail, first class, pro-perly stamped and addressed, on the date shown below.
Dr. Robert M. Lazo, Chairman Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Emmeth A. Leubke Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Richard F. Cole Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Chief, Docketing and Service Section (original plus two copies)
Colleen P. Woodhead, Esq.
U.S. Nuclear Regulatory Commission '
Office of the Executive Legal Director Washington, D.C. 20555 w_
~ -,
s Mitzi A. Young, Esq.
Office of Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l Norman A. Coll, Esq.
Steel, Hector & Davis 1400 Southeast Bank Building 100 S. Biscayne Boulevard Miami, FL 33131 Martin H. Hodder 1131 N.E. 86th Street Miami, FL 33138 Dated this 13th day of April 1984.
A*** %
Michael-A. Bauser .
Newman & Holtzinger, P.C.
1025 Connecticut Avenue, N.W.
Washington, D.C. 20036 Telephone: (202) 862-8476
_____-_-__-__x__-__-_____-_____---_-____-_---