ML20149H155

Proposed Tech Specs Re Design Features Fuel Storage,Which Allows for Use of Higher Enrichment Fuel & Specifies Spent Fuel Storage Requirements for Regions 1 & 2 of Spent Fuel Pool

ML20149H155
Person / Time
Site:	Byron, Braidwood
Issue date:	11/07/1994
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20149H153	List: ML20149H149 ML20149H155 ML20149H159
References
NUDOCS 9411180092
Download: ML20149H155 (16)

Text

. _. -.

ATTACHMENT B r

PROPOSED CHANGES TO APPENDIX A, STATION TECHicCAL SPECIFICATIONS, OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72 AND NPF-77, BYRON STATION UNITS 1 & 2 BRAIDWOOD STATION UNITS 1 & 2 1

Revision to:

5-5 5-Sa i

)

k:/rila /brdwd/ f u.3 2.wpf /16 9411180092 941107 PDR ADOCK 05000454 P

PDR

DESIGN FEATURES 5.6 FUEL STORAGE

{

CRITICALITY lha,+ A 4rGrirl T he-spen t-fuel-sto rage-racks-a r+-desi gned-a nd-s ham-be-ma i n tained---

-with;

-a.

A k equivalent-te les than-cr equal-to-Gr95-when-flooded with=

gg

-unbera ted-wateWich-inc4edes-a-con servative-al4owanc+-for-uncer~

l

-tainties-as-descr4 bed-in-Section-9.1 ef the FSAR.

This is based nn i

spentJuel-storage in Region 2-with-enrichmentg and burngp in accgggi

+nc+-with-Mguec 5. 5-1-or-in-a-cheekerboard-pattern; and l

b.

^ nominal-10r32-inch-north-south-and-1&r42-inch-east-westMenter-te -

f

-center-distance 4etweer fuel ::cemblic: pliced-in-Regien-1-spent-fue+

-storage-racks-and-a-nominal-9J3 4nch-center-to center-distance--

4stween-fuel-assemblies-placed in Region 2 spentJuskstorage recks.

i 5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the l

spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

t DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to i

prevent inadvertent draining of the pool below elevation 423 feet 2 inches.

CAPACITY t

5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2870 fuel assemblies.,

l l

5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

j

}

l BYRON - UNITS 1 & 2 5-5 Amendment No. 25

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1.5 2.0 2.5 3.0 3.5 4.0

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4.5 FUEL ASSEMBLY INITIAL ENRICMfENT (W/0 U-235)

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FIGURE 5.6-1 MIEDfDM BURNUP VERSUS INITIAL ENBICHMENT FOR REGION 2 STORAGE BYRON - UNITS 1 f 2 5-5s Amendment No. 25

l 50,000

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45,000 -

Enrichment Burnup g

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(MWD /MTU) r 1.60 0 fT~

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40,000 --42.00 8,565 4_

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2.20 11,845 Acceptable 2.40 14,729

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2.60 17,397 Region E

4 8

2.80 20,085

-T-h hj 35,000 -- 3.00 22,742 i

-^h 3.20 25,132

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[ [ 3.40 27,810 T~

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_.. 3.60 30,179 q___

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3.80 32.651 0

4.00 35,047 l -

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h 4.20 37,389 t-~-

ca 4.40 39,655

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4.60 42,024 c) 25,000.

-_J 4. 8 0 44,290 Cn 5.00 46,442

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.ii 1.60 2,00 2.40 2.80 3.20 3.60 4.00 4.40 4.80 5.20 Fuel Assembly Initial U-235 Enrichment (w/o)

I Notes:

The use of linear interpolation between the minimum burnups reported above is acceptable.

FIGURE 5.6 - 1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FC>R REGION 2 STORAGE BYRON - UNITS 1 & 2 5-Sa Amendment No.

Insert A 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with a k n less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties as described in Section 9.1 of the UFSAR. This is ensured by controlling fuel assembly placement in each region as follows:

a.

REGION 1 1.

A nominal 10.32 inch north-south and 10.42 inch east-west, center-to-center distance is maintained between fuel assemblies placed in the spent fuel storage racks.

2.

Fue! assemblies may be stored in this region with a) a maximum nominalinitial U-235 enrichment of less than or equal to 4.2 weight percent, or b) a maximum nominalinitial U-235 enrichment of 5.0 weight percent with sufficient Integral Fuel Burnable Absorbers present in each fuel assembly such that the maximum reference fuel assembly koo is less than or equal to 1.470 at 68 F.

b.

REGION 2 1.

A nominal 9.03 inch center-to-center distance is maintained between fuel assemblies placed in the spent fuel storage racks.

2.

a)

Fuel assemblies may be stored in this region with a maximum nominalinitial U-235 enrichment of 1.6 weight percent with no burnup and up to 5.0 weight percent U-235 with a minimum discharge burnup as specified in Figure 5.6-1, or b)

Fuel assemblies with a maximum nominalinitial U-235 enrichment of greater than 1.6 and less than or equal to 4.2 weight percent that do not meet the minimum burnup specified in Figure 5.6-1, shall be loaded in a checkerboard pattern for storage in this region.

DESIGN FEATURES l

5.6 FUEL STORAGE 1

l 1

CRITICALITY T nsett k 5.5.1.1 The-opent fuel :ter:g: r::k: :re designed 2nd :h:!' be : int:f =d-

^k 0;uiv:1:nt t: ?::: th:r er :;u:1 t: 0.95 when ficoded with-err

=ber:ted w ter, which includ:: : ::n;ervativ; 11cwance for-7 ence-t:intie; a; described in 5;; tion 0.1 of th: FSAR.

Th i:-

7 4: b:: d Or : pent fu;l :terag in Regi:n 2 with errich: nt: a nd-

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burnep fa ::: rdant: with Figure 5.5-1, or in a checkecbcard

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1 7

-p:ttern; and -

L b.

^ = in:! 10.32 in 5 north :: tb :nd 10.d2 4^9 e: t-vett center-L t center di:tance between fuci a;;;;blia; placed in Regien 1 5 pent fuel rock: and ; nominel 0.03 in:5 ::nter te ::nt:r di:t:nt betw :r i

fuci ;;;;mblic: pland in hgion 2 :p:nt fuel :terag rc k:.

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5. 6.1. 2 The k,ff for new fuel for the first core loading stored dry in the

]

spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

(

DRAINAGE t

5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 423 feet 0 inches.

i CAPACITY i

5.6.3 The spent fuel storage pool is designed and shall be maintained with a c

storage capacity limited to no more than 2870 fuel assemblies.

j

5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

i l

i BRAIDWOOD - UNITS 1 & 2 5-5 AmendmentNo./

e Insert A 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with a k,,, less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties as described in Section 9.1 of the UFSAR. This is ensured by controlling fuel assembly placement in each region as follows:

a.

REGION 1 1.

A nominal 10.32 inch north-south and 10.42 inch east-west, center-to-center distance is maintained between fuel assemblies placed in the spent fuel storage racks.

2.

Fuel assemblies may be stored in this region with a) a maximum nominal initial U-235 enrichment of less than or equal to 4.2 weight percent, or b) a maximum nominal initial U-235 enrichment of 5.0 weight percent with sufficient Integral Fuel Burnable Absorbers present in each fuel assembly such that the maximum reference fuel assembly km is less than or equal to 1.470 at 68oF.

b.

REGION 2 1.

A nominal 9.03 inch center-to-center distance is maintained between fuel assemblies placed in the spent fuel storage racks.

2.

a)

Fuel assemblies may be stored in this region with a maximum nominal initial U-235 enrichment of 1.6 weight percent with no burnup and up to 5.0 weight percent U-235 with a minimum discharge burnup as specified in Figure 5.6-1, or b)

Fuel assemblies with a maximum nominal initial U-235 enrichment of greater than 1.6 and less than or equal to 4.2 weight percent that do not meet the minimum burnup specified in Figure 5.6-1, shall be loaded in a checkerboard pattern for storage in this region.

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.5 INITIAL EflRICHMENT, WT% U-235 FIGURE 5.6-1 Minimum Burnup Versus Initial Enrichment For Region 2 Storage BPAIDW9nD UNITS 1 & 2 AMEND' DENT NO. 20 5-Sa

50,000 45,000 Enricivncnt Burnup g

(w/o)

(MdD/MTU)

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1.60 OII

--~ - '

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1.80 4,635

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40,000 --

2.00 8,565 I

2.20 11,845 ddill l_m

_ --d_l 2.40 14,729 Acceptable l!

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2.60 17,397 Region O

2.80 20,085

-7

- i k 35,000 --

3.00 22,742 ii g

3.20 25,132

__I g

3.40 27,810

_ I_

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3.60 30,179 I

n 30,000 --

3.80 32,651 D

4.00 35,047 l, --

4.20 37,389 M

, ] 4.40 39,655 j~1-I e 25,000 --

4.60 42,024 I. ii i

g 4.80 44,290 I

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.g 5.00 46,442 g

I 20,000 O

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Unacceptable 1 i 3,

Region om

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i i i i i i i 1.60 2.00 2.40 2.80 3.20 3.60 4.00 4.40 4.80 5.20 Fuel Assembly Initial U-235 f.nrichment (w/o)

Notes:

The use of linear interpolation between the minimum burnups reported above is acceptable.

FIGURE 5.6 - 1 MINIMUM BURNUP VERSUS INITIAL ENRICHMENT FCR REGION 2 STORAGE BRAIDWOOD - UNITS 1 & 2 5-Sa Amendment No.

i i

ATTACHMENT C i

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS Commonwealth Edison (Comed) proposes to revise Section 5, Design Features, of Technical Specifications for Byron and Braidwood stations. The proposed changes would allow an increase to the allowable nominal fuel enrichment from 4.2 to 5.0 weight percent Uranium 235 (w/o U-235). Using higher enriched fuel in the design of reactor cores will result in fuel cost savings from the reduction in the number of new fuel assemblies required per reload and subsequent reduction in spent fuel storage space. The proposed changes are supported by a criticality reanalysis of the spent fuel pool. The criticality reanalysis considered the two storage regions and produced separate criteria for each region. The changes include (1) increasing the allowable j

storage enrichment in Region 1 and allowing the use of Integral Fuel Burnable l

Absorbers (IFBAs) for reactivity equivalencing, (2) revising the Region 2 discharge j

burnup curve to include nominal fuel enrichments up to 5.0 w/o U-235, and (3) making i

editorial changes.

Comed has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10CFR50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

l 1.

Involve a significant increase in the probability or consequences of an accident previously evaluated; or l

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.

l A.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to Section 5 of Technical Specifications do not affect any accident initiators or precursors and do not change or alter the design assumptions for the systems or components used to mitigate the consequences of an accident. The fuel enrichment increase will not affect reactor operation or the core design methods. The physical characteristics of the fuel assemblies are not changed, and fuel assembly movement will continue to be controlled by approved fuel handling procedures. Reload core designs will continue to be i

l performed on a cycle by cycle bases as part of the reload safety evaluation process, using NRC approved codes and methods. Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and Technical Specifications to ensure that reactor operation is acceptable.

kt/nla/bt e d/ fuel 2.wpf/13

The proposed changes are consistent with the analysis performed in the

" Criticality Analysis of Byron and Braidwood Station Fuel Storage Racks." The analysis was revised in June 1994 to include boraflex gaps and shrinkage. The revised analysis is provided in the proposed Technical Specification amendment.

The analysis methodology has been previously accepted by the NRC and is consistent with the appropriate standards to establish the K, limit for storage racks and to calculate the maximum K,. The reanalysis addresses the most limiting postulated accident of a misloaded fuel assembly and has shown that having at least 300 ppm of soluble boron offsets any positive reactivity impacts for any of the postulated accidents. The concentration of boron in the spent fuel pool water, which is administratively controlled, is sufficient to maintain K.,less than or equal to 0.95. The analysis is bounding for a dropped fuel assernbly on top of a rack or between rack modules, loss of cooling systems, and reduction the fuel pool temperature to less than 50*F. The proposed changes do not impact any other accident previously evaluated in the UFSAR. There is no postulated accident that could cause reactivity to increase beyond the analyzed conditions in the spent fuel racks.

There is no impact on the ability of the Spent Fuel Pool cooling system to maintain the bulk pool temperature within limits. The UFSAR analysis performed to calculate the maximum fuel cladding temperature and spent fuel pool cooling include assumptions which bound the use of more highly enriched fuel assemblies. Although fuel enrichment is not a specific assumption in any of these analysis, the heat load of a typical core offload may change with higher enrichments. The average burnup of the offload will be increased since few assemblies will be used per cycle; however, the new heat load will continue to be bounded by the UFSAR analysis because the spent fuel pool racks have been analyzed for a total core offload with all fuel assemblies having 4.5 years of operating time.

The radiological consequences analysis continues to bound the licensed fuel burnup and enrichment at Byron and Braidwood stations. The radiological l

consequences analysis results are a function of the core inventory of radioactive isotopes. Since the maximum fuel burnup limits and fuel peaking factors will not be exceeded, the assumed fission product inventory will remain valid; therefore,

)

the limits of 10 CFR 100 continue to be met. Additionally, Byron and Braidwood addressed the issue of the impact on the radiation levels at the pool surface to the worker during non-accident conditions. These conditions are not changed as the result of this submittal, because the average fuel assembly burnup limit (isotopic inventory) and maximum power produced in each fuel assembly will not be changed by the increased fuel enrichment.

k: inl a /t1 dwd/ f uel2.wpf /14

B.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

]

The proposed changes do not affect the design or operation of any system, structure, or component in the plant. There are no changes to parameters j

governing plant operation; no new or different type of equipment will be installed.

Each reactor core design will continue to meet all design requirements; operation of the core wil! not be affected. No modifications to the spent fuel pool are being 1

pursued and the fuel parameters used in the analysis remain bounding. The method and manner in which the fuel will be stored in the spent fuel pool has not changed. The proposed changes ensure that 17X17 (Optimized Fuel Assembly, VANTAGE 5, VANTAGE +, and PERFORMANCE +) fuel assemblies can be safely stored, maintaining a K., < 0.95 under full water density conditions, in both Regions 1 and 2 of the spent fuel pool. All design criteria and criticality l

acceptance criteria continue to be met. The reanalysis addresses the most limiting postulated accident (mistoaded fuel assembly) and has shown that having at least 300 ppm of soluble boron offsets any positive reactivity impacts for any of the postulated accidents. The level of boron in the spent fuel pool water,

(

which is administratively controlled, is sufficient to maintain K,less than or equal to 0.95. The reanalysis to increase the storage enrichment of fuel in Regions 1 and 2 of the spent fuel pool does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Additionally, approval of this amendment will not create a new accident with regards to the new fuel storage vault which is designed to handle the increased enrichment. The Byron and Braidwood new fuel vaults were previously analyzed using NRC accepted criticality analysis methodology in June 1989. This analysis was performed to increase the storage enrichment of the New Fuel Vault to 5.0 w/o U-235. The New Fuel Vault analysis was submitted to the NRC and is the current licensing basis, i

ll k : /nla /t>rdwd/ f uo] 2.wpf /15

C.

The proposed changes do not involve a significant reduction in a margin of safety.

The proposed changes do not affect the margin of safety for any Technical Specification. All reactor design criteria will continue to be met. The methodologies used in the accident analyses have been accepted previously by the NRC and all criticality acceptance criteria have been met under all assumed conditions (normal and accident). The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, K.,, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.2-1983 and OT Position Paper for Review and Acceptance of Spent Fuel Storage and Handling Applications, dated April 14, 1978. The analyses for both Regions 1 and 2 fuel storage were verified to meet the above design basis.

The criticality analysis for Regions 1 and 2 has been revised to allow for storage of fuel assemblies with enrichments up to 5.0 w/o U-235. The proposed Technical Specification changes include those changes necessary to maintain K,

less than or equal to 0.95, including conservative allowances for uncertainties and biases, when the pool is flooded with unborated water. The proposed changes include a requirement for fuel assemblies with enrichments above 4.2 w/o U-235 to contain sufficient integral fuel burnable absorbers such that the maximum reference fuel Km is less than or equal to 1.470 in unborated water at 68'F due to restrictions on spent fuel storage. Should a postulated accident occur which causes a reactivity increase in the Byron and Braidwood Spent Fuel Pools, K, will be maintained less than or equal to 0.95 due to the presence of at least 300 ppm of soluble boron in the spent fuel pool. The proposed changes do not affect any plant safety parameters or setpoints.

The proposed changes ensure that the design basis for preventing criticality in the fuel storage areas is preserved, and fuel cycle designs will continue to be analyzed using NRC accepted codes and methods to ensure the design bases are satisfied.

Therefore, based on the above evaluation, Commonwealth Edison has concluded that the proposed changes do not involve significant hazards considerations.

k: /nl a/t'rdwd/ f ue12.wpf /16

ATTACHMENT D ENVIRONMENTAL ASSESSMENT Commonwealth Edison has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10CFR51.21. It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10CFR51.22(c)(9). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10CFR50 and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards considerations As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards considerations.

(ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite As documented in Attachment A, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure The proposed change will not result in changes in the operation or configuration of the facility. Core design will continue to meet all core design cr;teria, and i

reactor operation will not be impacted. There will be no change in the level of l

controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposal result in any change in the normal radiation levels within the plant. Therefore,there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

4 k:/nla/brdwd/ fuel 2.wpf/17

ATTACHMENT E CHANGE

SUMMARY

Technical Described Specificatio_q Chanae Descrintion in item #

5.6.1.1 The first sentence from 5.6.1.1.a, which describes Keff, 3

was moved to 5.6.1.1. Change "FSAR" to "UFSAR".

The required distances in Region 1 were relocated from 3

5.6.1.1.a.1 5.6.1.1.b New section for Region 1 allowing fuel assemblies 4

5.6.1.1.a.2.a having a maximum nominalinitial enrichment of up to 4.2 weight percent New section for Region 1 allowing fuel assemblies 1

5.6.1.1.a.2.b having a maximum nominalinitial enrichment of 5.0 weight percent with minimum number of sufficient Integral Fuel Burnable Absorbers present in each fuel assembly such that the maximum reference fuel assembly Ke is no greater than 1.470 at 68'F 5.6.1.1.b.1 The required distances in Region 2 were relocated from 3

5.6.1.1.b New section for Region 2 allowing fuel assernblies 2

5.6.1.1.b.2.a having a discharge burnup in the " acceptable range" of Figure 5.6-1 5.6.1.1.b.2.b New section for Region 2 allowing fuel assemolies having a nominal U-235 enrichment of less than 4.2 2

weight percent that do not meet the minimum burnup in the " acceptable range" of Figure 5.6-1 to be loaded in a checkerboard pattern ;"

Figure 5.6-1 Replace Figure 5.6-1 with the new analysis curve 2

k : / n1 a/ tintwd/ f ue12.wp t /19

e o

P ATTACHMENT F I

t t

BYRON AND BRAIDWOOD SPENT FUEL RACK CRITICALITY ANALYSIS CONSIDERING BORAFLEX GAPS AND SHRINKAGE b

kt/nla/brdwd/ fuel 2.wpf/20

-