ML20045F021
| ML20045F021 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/30/1993 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20045F017 | List: |
| References | |
| NUDOCS 9307060325 | |
| Download: ML20045F021 (21) | |
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t UNITED STATES NUCLEAR REGUIATORY C0 EMISSION NORTIIERN STATES POWER COMPANY I
6 PRAIRIE ISIAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 i
50-306 I
REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 REVISION TO LICENSE AMENDMENT REQUEST DATED JUNE 11, 1993 Northern Str,tes Power Company, a Minnesota corporation, requests authorization
-j for changer to Appendix A of the Prairie Island Operating License as shown on-the attachments labeled Exhibits A, B, C and D.
Exhibit A describes the t
proposed changes, reasons for the changes, safety evaluation and a significant hazards evaluation.
Exhibits B and C are copies of the Prairie Island
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Technical Specifications incorporating the proposed changes.
Exhibit D is a report.upporting the requested changes.
l This letter contains no restricced or other defense information.
f NORTHERN STATES POWER COMPANY n
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' Roser U Andersen Director Licensing and Management Issues i
6 Onthishd~dayof
/993beforemeanotarypublicinandforsaid County,personallya@dbeingfirstdulyswornacknowledgedthatheis eared Roger O Anderson, Director Licensing and i
Management Issues, an i
authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his l
knowledge, informati:n, ar.d belief the statements made in it are true and that it is not interpose. fo delay.
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, l Exhibit A l
1 Prairie Island Nuclear Generating Plant June 30,1993 Revision to License Amendment Request Dated June 11, 1993 i
Revised Evalaation of Proposed Changes to the
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Technical Specifications Appendix A of
' j Operating License DPR-42 and DPR-60 Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A, Technical Specifications:
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- 1. Fuel Enrichmtn.t Limir Chanres f
Backcrour?
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I Technical Specifica t ion 5.6. A currently limits fuel in the spent fuel pool and new fuel storap racks to a maximum enrichment of 4.25 weight percent'
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4 U-235.
Technical Specification 3,8.E which places restrictions on the i
storage of low burnup fuel assemblies, resulted fron the criticality l
analysis for the storage of 4.25 weight percent U-23S fuel in the Prairie 3
j.
Island spent fuel pool.
Technical Specification 5.3.A.2 also limits the maximum enrichment in the reactor core to 4.25 weight percent U-235.
In l
t order to support longer fuel cycles at Prairie Island, it is necessary to l
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increase fuel enrichments beyond the 4.25 weight percent limit and to i
revise the Technical Specifiertion enrichment limits.
i j
The Prairic Island spent storage racks have been analyzed (Exhibit D) i to allow for the storage assemblies with enrichments up to 5.0
. i weight percent U-235 while u.a.caining K,y s 0.95 including uncertainties.
This criticality analysis utilized the following storage configurations or regions to ensure that the spent fuel pool will remain suberitical during the storage of fuel assemblies with all possible combinations of burnup and i
initial enrichment:
l 4
i i
1.
The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of l
8 5.0 weight percent'U-235.
This configuration stores " burned" and
" fresh" fuel assemblies in a 2x2 checkerboard pattern.
Fuel
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assemblies stored in " burned" cell locations must have an initial enrichment less than 2.5 weight percent U-235 (nominal) or satisfy a minimun burnup requirement.
The use of empty cells is also an acceptable option for the " burned" cell locations.
Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 weight percent U-235 with no requirements for burnup-or burnable j
absorberr.
I t
2.
The second region does not utilize any spt 4 21 loading pattctn.
Fuel j
assemblies with burnup and initial enrichments which fall into an j
unrestricted range can be stored anywhere in the region with no special placement restrictions.
Fuel assemblies which fall into an restricted burnup and. initial enrichment range must be stored in the checkerboard region.
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Exhibit A Page 2 of 20 Proposed Chanres and Reasons for Chances l
The propcsed changes to the Prairie Island Technical Specifications are described below, and the specific wording changes to Technical i
Specifications are shown in Exhibits B and C.
P A.
Pronosed Channes to Technical Soecification 3.8.E Technical Specification 3.8.E, which places restrictions on the storage of low burnup fuel assemblies, resulted from the criticality analysis for the storage of 4.25 weight percent fuel in the Prairie Island spent fuel pool.
That criticality analysis found that in order to assure a K,,, less than or equal to 0.95, high enrichment fuel with low average assembly burnup would have to be stored in the three out of four configuration described in Specification 3.8.E.
The storage requirements of the current Specification 3.8.E are not valid for the storage of fuel enriched to beyond 4.25 weight percent U-235.
Therefore, based on the results of the criticality analysis (Exhioit D) for the storage of 5 weight percent fuel in the spent fuel pool, t
Specification 3.8.E is being revised as shown in Exhibit B.
i The proposed changes to Specification 3.8.E, described below, delete I
the current requirements for the storage of low burnup fuel and
-i replace them with spent fuel pool storage restrictions which are based on the combination of fuel assembly burnup and initial enrichment.
The proposed changes to Specification 3.8.E and Table TS.4.1-2B are based on the guidance provided in Sections 3.17.6, 3.17.7 and 4.3.1.1 of the Westinghouse Standard Technical Specifications, NUREG-1431.
Section 3.8.E 1 Proposed Specification 3.8.i.1 incorporates new restrictions on the storage of fuel in the spent fuel pool. To be stored without l
restriction in the spent fuel pool, the burnup and initial enrichment of a fuel assembly will have to fall within the unrestricted range of proposed Figure TS.3.8-1.
Any fuel assemblies whose burnup and initial enrichment fall c.utside the unrestricted range of proposed Figure TS.3.8-1 will have to be stored in the checkerboard region, described above, in accordance with the requirements of proposed Specification 5.C.A.1.d.
1 Figure TS.3.8-1, which specifies the minimum burnup requirements for unrestricted storage in the spent fuel pool, is based on enrichments from 3.87 to 5.0 weight percent U-235.
Enrichments lower than 3.87 weight percent are conservatively bounded by the minimum burnup requirement for 3.87 weight percent U-235 which is 2000 MWD /MTU.
Therefore, Figure TS.3.8-1 has been drswn to require that fuel with an initial enrichment of less than 3.87 weight percent U-235 have 2000 l
MWD /MTU burnup or greater before unrestricted storage in the spent j
fuel pool will be allowed.
Proposed Specification 3.8.E.1 includes an action statement which specifies the action to be taken if the storage requirements of Specification 3.8.E.1 are not met.
The proposed action statement requires the immediate initiation of action to move any noncomplying fuel assembly to an acceptable location.
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Exhibit A Page 3 cf 20
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Proposed Specification 3.8.E.1 also includes a statement that the provisions of Specification 3.0.C are not applicable.
If the requirements of Specification 3.8.E.1 cannot be met during cold shutdown or refueling, Specification 3.0.C would not be applicable because the reactor would already be shutdown.
If the requirements of Specifichtion 3.8.E.1 cannot be met with the reactor above cold shutdown, any problems with respect to storage in the spent fuel pool would be independent of reactor operation and there would not be.
i sufficient reason to require a reactor shutdown.
Section 3.8.E.2 I
Proposed Specification 3.8.E.2 incorporates new restrictions on the spent fuel pool boron concentration.
Proposed Specification 3.8.E.2 will ensure that the spent fuel pool will contain adequate dissolved boron anytime fuel assemblies with a combination of burnup and initial enrichment in the restricted range of proposed Figure TS,3.8-1 are stored in the fuel pool and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the j
spent fuel pool.
The negative reactivity effect of the soluble boron would compensate for the increased reactivity caused by a l
mispositioned fuel assembly.
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Proposed Specification 3.8.E.2 includes an action statement which specifies the action to be taken if the spent fuel pool boron concentration is not within the limit of Specification 3.8 E.2.
The proposed action statement requires the immediate suspension of fuel
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movements in the spent fuel pool and initiation of action to restore' spent fuel pool boron concentration to within its limit or to complete a spent fuel pool verification.
Proposed Specification 3.8.E.2 also includes a-statement that the j
provisions of Specification 3.0.C are not applicable.
If the j
requirements of Specification 3.8.E.2 cannot be met during cold j
shutdown or refueling, Specification 3.0.C would not be applicable l
because the reactor would already be shutdown.
If the requirements of Specification 3.8.E.2 cannot be met with the reactor above cold shutdown, any problems with respect to the spent fuel pool boron concentration would be independent of reactor operation and there would not be sufficient reason to require a reactor shutdown.
j B.
Pronoced Chances to Technical Specification Table TS.4.1-2B In order to ensure that the boron concentration limits of new l
Specification 3.8.E.2.a are met, the existing surveillance requirements for spent fuel pool boron in Table TS.4.1-2B have been revised. A new Note 8 has been incorporated into Table TS.4.1-2B (page 2 of 2).
Note 8 specifies that the spent fuel pool boron concentration shall be verified weekly, by chemical analysis, to be within the limits of Specification 3.8.E.2.a. when fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the spent fuel pool and a spent fuel pool verification has not been performed.ince the last movement.
of any fuel assembly in the spent fuel pool.
A reference to Note 8 and the word " Weekly" has been added to the spent foal pool boron 3
surveillance requirement in Table TS.4 1-2B (page 1 of 2).
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Exhibit A j
Page 4 of 20 C.
Proposed Changes to Technical Specification 5.6.A s
Because they contain information not directly related to criticality considerations, the'first paragraph and the first sentence of the-second paragraph of Specification 5.6.A are being relocated to Section 5.6.B, " Spent Fuel Storage Structure".
j The remaining information in Section 5.6.A is being revised, as shown-
- Exhibit B, to a format consistent with Section 4.3.1.1 of.the tinghouse Standard Technical Specifications, NUREG-1431.
, lification 5.6.A.1.a Technical Specification 5.6.A currently Itmits fuel in the spent fuel pool to a maximum enrichment of 4.25 weight percent U-235.
In order:
to support longer fuel cycles at Prairie Island, it is necessary to increase fuel enrichments beyond the 4.25 weight percent limit.
Based i
on the results of a criticality analysis included as Exhibit D to this request, the maximum enrichment for fuel stored in the spent fuel pool has been increased to 5 weight percent U-235 in proposed Specification j
5.6.A.1.a.
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Specification 5.6.A.1.h Proposed Specification 5.6.A.l.b contains the spent fuel pool K,f, requirements from the current Section 5.6.A.
A reference to the Criticality Analysis for the uncertainties in the calculation of K,f, has been incorporated.
Specification 5.6.A_1.c New Specification 5.6.A.l.c is being incorporated to clearly state i
that new or spent fuel assemblies with a combination of burnup and initial enrichment in the unrestricted range of proposed Figure TS.3.8-1 are allowed unrestricted storage in the spent fuel racks.
This statement is consistent with the requirements of proposed Specification 3.8.E.1.a.
s Specification 5.6.A.1.d Per the criticality analysis (Exhibit D), fuel assemblies which fall e
into the restricted range of proposed Figure TS.3.8-1 must be stored in the checkerboard loading pattern discussed above.
This requirement is invoked by proposed Specification 5.6.A.l.d which states that new or spent fuel assemblies with a combination of burnup and initial enrichment in the restricted range of proposed Figure TS.3.8-1 will be stored in compliance with proposed Figures TS.S.6-1 and TS.5.6-2.
These figures will be utilized to implement the checkerboard loading pattern required by the spent fuel pool criticality analysis (Exhibit D).
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6 Exhibit A Page 5 of 20 Specification 5.6.A.2.a 1
Technical Specification 5.6.A currently limits fuel in the new fuel.
storage racks to a maximum enrichment of 4.25 weight percent U-235.
In order to support longer fuel cycles at Prairie Island, it is
.l necessary to increase fuel enrichments beyond the 4.25 weight percent i
limit. Based on the results of he criticality analysis (Exhibit D),
the maximum enrichment for fuel stored in the new fuel storage racks has been increased to 5 weight percent U-235 in proposed Specification 5.6.A.2.a.
Specifications 5.6.A.2.b and 5.6.A.2.c Proposed Specifications 5.6.A.1.b'and 5.6.A.2.c contain the new fuel storage rack Kg, requirements from the current.Section 5.6.A.
A l
reference to the Criticality. Analysis for the uncertainties in the calculation of Kg, has been incorporated in each specification.
Bases for Specification 3.8.E Bases for proposed Specifications 3.8.E.1 and 3.8.E.2 are being i
incorporated into the fuel handling specification bases as shown in Exhibit B.
D.
Proposed Changes to Technical Specification 5.3.A.2 i
Technical Specification 5.3.A.2 currently limits fuel in the reactor core to a maximum enrichment of 4.25 weight percent U-235.
In order to support longer fuel cycles at Prairie _ Island, the maximum enrichment for fuel in the reactor core has been increased to 5' weight percent U-235 in proposed Specification 5.3.A.2.
Safety Evaluation i
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95% probability at a 95% confidence level that the Kg, of the fuel assembly array will be less than 0.95 with full density moderation. Additionally for storage racks which are maintained in a dry condition, such ac the new fuel racks, Kg, must be
.i less than 0.98 for low density optimum moderation conditions.
i A.
Spent Fuel Storage
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The Prairie Island spent fuel storage racks have been analyzed (Exhibit D) to allow for the storage of fuel assemblies with enrichments up to 5.0 weight percent U-235 while maintaining K s
g, 0.95, including uncertainties, at a 95/95 probability / confidence level. This criticality analysis utilized the following storage configurations or regions to ensure that the spent fuel pool will remain suberitical during the storage of fuel assemblies with all possible combinations of burnup and initial enrichment:
1.
The first region utilizes a checkerboard loading pattern to accommodate new or low burnup fuel with a maximum enrichment of 4
5.0 weight percent U-235.
This configuration stores " burned" and
" fresh" fuel assemblies in a 2x2 checkerboard pattern.
Fuel assemblies stored in " burned" cell locations must have an initial
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Exhibit A Page 6 of 20 l
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t enrichment less than 2.5 weight percent U-235 (nominal) or satisfy
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a minimum burnup requirement. The use of empty cells is also an I
acceptable option for the " burned" cell locations.
Fuel assemblies stored in the " fresh" cell locations can have enrichments up to 5.0 weight percent U-235 with no requirements for burnup or burnable absorbers.
2.
The second region does not utilize any special loading pattern.
Fuel assemblies with burnup and initial enrichments which fall l
into the unrestricted range of proposed Figure TS.3.8-1 can be stored anywhere in the region with no special placement restrictions.
Fuel assemblies which fall into the restricted range of proposed Figure TS.3.8-1 must be stored in the first region using the checkerboard loading pattern.
The burned / fresh fuel checkerboard region can be positioned anywhere within the spent fuel racks, but the boundary between the checkerboard region and the unrestricted region must be either:
1.
separated by a vacant row of cells, or 2.
the interface must be configured such that there is one row carryover of the pattern of burned assemblies from the checkerboard region into the first row of the unrestricted region j
(Figure 6, Exhibit D/ Figure TS.5.6-1).
3 Spent fuel gap activities, which are a function of fuel assembly burnup, are not directly affected by an increase in fuel assembly enrichment.
However, the spent fuel gap activities are a function of j
fuel burnup, which will be increased by the use of higher enriched i
fuel.
Fuel burnup is not expected to increase beyond the value i
currently assumed in the accident analysis until late in 1996.
Prior to exceeding the currently analyzed maximum fuel burnup, the possible I
offsite dose consequences of extending fuel burnup will be evaluated J
to ensure compliance with 10 CFR Part 100 requirements.
This j
evaluation will be completed prior to the startup of the first cycle where the maximum fuel burnup currently assumed in the accident analysis is expected to be exceeded.
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Fuel assembly decay heat production is a function of core power level, and since the core power level remains unchanged, the decay heat j
generated by a spent fuel assembly will not be significantly impacted by the proposed enrichment limits and the spent fuel pool cooling j
requirements will not be affected.
The water in the spent fuel pool normally contains soluble boron, which results in large suberiticality margins under actual operating conditions. However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost.,
specify that the limiting k of 0.95 he evaluated in the absence of g,
soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a suberitical condition during normal operation with the regions fully loaded.
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4 Enhibit A Page 7 of 20 Most accident conditions do not result in an increase in the activity j
of either of the two regions.
Examples of_these accident conditions are the loss of cooling (reactivity increase with decreasing water i
density), the dropping of a fuel assembly on the top of the rack, and 1
the dropping of a fuel assembly between rack modules and wall (rack design precludes this condition). However, accidents can be postulated that could increase the reactivity.
For these accident
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conditions, the double contingency principle of ANS1 N16.1-1975 can be applied.
This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
The double contingency principle allows credit for soluble boron under abnormal or accident conditions, since only a single accident need be l
considered at one time.
Fcr example, the most severe accident scenario is the accidental misloading of a fuel assembly into a rack location for which the restrictions on location, enrichment or burnup
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are not satisfied.
This could potentially increase the reactivity in spent fuel racks. To mitigate these postulated criticality related accidents, Specification 3.8.E.2 ensures the spent fuel pool contains adequate dissolved boron anytime fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure TS.3.8-1 are stored in the fuel pool and a spent fuel pool verification has not been performed since the last movement of any fuel assembly in the spent '!uel pool.
The negative reactivity effect of the soluble boron would :ompensate for the increased reactivity caused by a mispositioned fus1 assembly, i
i The boron concentration requirements of Specification 3.8.E.2 are no longer imposed when no fuel raovements are occurring and a spent fuel i
pool verification has been completed, because the storage requirements l
of Specifications 3.8.E.1 and 5.6.A.l.d are then adequate to prevent criticality.
i Specification 3.8.E.2.a is not imposed when only fuel assemblies with j
a combination of burnup and initial enrichment in the unrestricted range of Figure TS.3.8-1 are stored in the spent fuel pool.
The
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requirements of Specification 3.8.E.2.a are not required in that case j
because with only fuel assemblies that have burnup and initial 1
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enrichment in the unrestricted range of Figure TS.3.8-1 it is not 3
possible to cause an inadvertent criticality by mispositioning a fuel
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assembly in the spent fuel pool.
l k' hen the requirements of Specification 3.8.E.2.a are applicable, and the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of. fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.
An acceptable alternative is to complete a spent fuel pool verification. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored.
j Exhibit A Page 8 of 20 j
A spent fuel pool verification is required following the last movement I
of fuel assemblies in the spent fuel pool, if fuel assemblies with a combination of burnup and initial enrichment in the restricted range j
of Figure TS.3.8-1 are stored in the spent fuel pool.
This verification will confirm that any fuel assemblies with a combination of burnup and initial enrichment in the restricted range of Figure i
TS.3.8-1 are stored in accordance with the requirements of Specification 5.6.A.l.d.
Based on the results of the criticality analysis described in Exhibit D, the implementation of the Technical Specification changes and administrative controls described above will ensure that.the Prairie i
Island spent fuel storage facilities will remain substantially suberitical at all times and the probability of a criticality event will not be increased by the storage of 5.0 weight percent U-235 fuel.
I Hew Fuel Storace Racks t
The criticality analysis performed on the Prairie Island new fuel racks (Exhibit D) found K,g to be less than 0.95 including uncertainties at a 95/95 probability / confidence level for fuel enriched to a batch average 5.0 weight percent U-235 and assuming full density moderation. However, the analysis of the new fuel racks under the low density optimum moderation conditions found that the maximum rack reactivity exceeded the design limit of 0.98 if the new fuel rack 8 by 11 storage cell array was assumed to be completely filled.
Further evaluation found that if 14 fuel storage cells in the center of the new fuel storage rack array were removed from service (Figure 2, Exhibit D), the 0.98 K,g limit would be met under low density optimum moderation conditions.
Therefore, prior to placing any fuel assemblies with enrichments greater than 4.25 weight percent U-235 into the new fuel racks, the new fuel racks will be modified to preclude the storage of fuel assemblies in the 14 central storage cells assumed to be empty by the criticality analysis (Figure 2, Exhibit D).
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Under normal conditions, the new fuel storage racks are maintained in a dry environment. The introduction of water into the new fuel storage rack area is the worst case accident scenario.
The full density and low density optimum moderation cases analyzed in criticality analysis (Exhibit D) are the bounding accident situations which result in the most conservative fuel rack K,g.
Other accidents can be postulated which would cause some reactivity increase, such as, dropping a fuel assembly between the rack and wall or on top of the rack.
For these other accident conditions, the absence of a moderator in the new fuel storage racks can be assumed as i
a realistic initial condition since assuming its presence would be a second unlikely event.
Since the normal, dry new fuel rack reactivity is less than 0.62 (Figure 5, Exhibit D), there is sufficient reactivity margin to the 0.95 limit to cover the above postulated accidents.
1 EdibitA I
' Page 9 of 20 Based on the results of the criticality analysis described in Exhibit D, the implementation of the Technical Specification changes,
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administrative controls and modifications described above will ensure that the Prairie Island new fuel storage facilities will remain substantially suberitical at all times and the probability of a criticality event will not be increased by the storage of 5.0 weight percent U-235 fuel.
Reactor Core i
The use of 5.0 weight percent U-235 fuel in the reactor core will be.
i evaluated as part of cycle specific reload analyses using NRC approved methodology. These cycle specific analyses will confirm the acceptability of reactor operation with the higher enrichment reload fuel.
'f Conclusion In conclusion, the health and safety of the public will not be adversely affected by the proposed Technical Specification changes.
i Determination of Significant Hazards Considerations The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as
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required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92.
This analysis is provided below:
1.
The proposed amendment will not involve a significant increase in the probability or conseauences of an accident previous 1v evaluated, Fuel Storage There is no increase in the probability of a fuel assembly drop accident in the new fuel storage area or the spent fuel pool since the mass of a fuel assembly does not increase when the fuel enrichment is increased, i
i There is not a significant increase in the consequences of a fuel assembly drop accident in the spent fuel pool since the fission i
product inventories in the fuel assemblies do not change significantly due to an increase in the fuel enrichment.
Spent fuel gap activities, which are a function of fuel assembly burnup, are not directly affected by an increase in fuel assembly enrichment. The spent fuel l
gap activities are a function of fuel burnup, which will be increased by the use higher enriched fuel. However, the increase in fuel burnup 4
anticipated with the proposed increase in fuel enrichment is not expected to significantly effect the fuel gap activity.
Additionally, fuel burnup is not expected to increase beyond the value currently assumed in the accident analysis until late in 1996.
The possible offsite dose consequences of extending fr.cl burnup during subsequent cycles will be evaluated to ensure corepliance with 10 CFR i
Part 100 requirements prior to the startup of the first cycle where the maximum fuel burnup currently assumed in the accident analysis is expected to be exceeded.
4.
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Exhibit A Page10cf 20 I
There is no increase.in the probability or consequences of misplacing-l fuel assemblies in the spent fuel pool or new fuel storage racks as a.
I result of an increase in fuel enrichment.
The probability of misplacing a fuel assembly in the spent fuel pool or new fuel vault is not increased because fuel assembly placement will be controlled pursuant to the current approved fuel handling procedures and the requirements of the proposed Technical Specifications. Additionally, l
there is no increase in the probability of. misplacing fuel assemblies 4
in the new fuel storage racks because the racks will be modified to.
prevent the insertion of fuel assemblies in the central 14 cell locations assumed to be open in the criticality analysis.
There is no increase in the consequences of misplacing fuel assemblies
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in the spent fuel pool because criticality analyses demonstrate that l
the pool will remain subcritical assuming misplacement does occur if j
the pool contains an adequate boron concentration. The proposed i
Technical Specifications will ensure that the adequate boron concentration is maintained when required.
l There is no increase in the consequences of misplacing fuel assemblies in the new fuel starage racks because for any such event, the absence of a moderator in the new fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.
Since the normal, dry new fuel rack reactivity is less than 0.62 (Fig. 5, Exhibit D), there is sufficient reactivity 1
margin to the 0.95 limit to cover any possible misplacement.
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i There is no increase in the probability of introducing optimum moderation conditions in the new fuel storage vault as a result of an increase in fuel enrichment. The increase in fuel enrichment will have no effect on the possible introduction of a moderating material into the new fuel vault.
There is no increase in the consequences of introducing optimum moderation conditions in the new fuel storage vault as a result of an increase in fuel enrichment. The new fuel vault has been analyzed under a range of moderation conditions from fully flooded to optimum moderation at the increased fuel enrichment. These analyses demonstrate that the new fuel storage racks remain suberitical under these moderation conditions.
Reactor Core Operation of Prairie Island Units 1 and 2 with 5.0 weight percent U-235 fuel in the reactor core does not involve a significant increase in the probability or consequences of an accident previously evaluated for the following reasons:
I 1.
The use of 5.0 weight percent U-235 fuel in the reactor core will be evaluated as part of cycle specific reload analyses using NRC a
approved methodology. These cycle specific analyses will confirm that reactor operation with the higher enrichment reload fuel will meet all applicable requirements and acceptance critieria.
l 2.
Neither actuation of safety systems nor accident mitigating capabilities will be adversely affected by operation of the Prairie Island reactors with 5.0 weight percent U-235 fuel.
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Exhibit A Page11of 20 i
,1 3.
The proposed enrichment increase does not pose a challenge'to l
installed safety systems. Therefore, no new performance requirements are being imposed on any system or component such that any design critieria will be exceeded.
Conclusions Based on the conclusions of the above analysis, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident nreviousiv analyzed.
Fuel Storate Spent fuel handling accidents are not new or different types of accidents, in that they are already analyzed in the Updated Safety Analysis Report. Criticality accidents in the new fuel storage vault or the spent fuel pool are not new or different types of accidents in that they are already analyzed in the Updated Safety Analysis Report for fuel enrichments up to 4.25 weight percent U-235.
Additional criticality analyses (Exhibit D) have been performed for fuel enrichments up to 5.0 weight percent U-235.
As described above, the storage of higher enrichment fuel in the new fuel racks will require the modification of 14 central cells of the new fuel storage racks to prevent insertion of new fuel assemblies.
The modifications and their installation will be minor in nature and as such will not create the possibility of a new or different kind of accident.
l The administrative controls which will be implemented to control the storage of higher enrichment fuel will only affect where spent fuel assemblies can be stored and the required spent fuel pool boron concentration.
Limiting where fuel assemblies can be stored in the spent fuel pool will have little affect on fuel handling operations and the boron concentration required for the storage of higher enriched fuel is well below the boron concentration normally maintained in the spent fuel pool.
Therefore, the implementation of these administrative controls will not create the possibility of a new or different kind of accident.
The Prairie Island spent fuel racks utilize boraflex sheets between the storage cells to assure suberiticality of the racks.
Even though the boraflex sheets in the spent fuel racks were not adhesively constrained during construction, which reduces the likelihood of gaps forming, concerns related to the possibility of gaps having formed in l
the horaflex theets due to radiation induced shrinkage, were addressed in the criticality analysis by assuming four inch axial gaps at the axial center of the active fuel in all the boraflex panels in the spent fuel pool.
This four inch gap is considered conservative based on neutron radioassay measurements of the boraflex poison material.
The centerline positioning of the gap is also considered conservative because it resulted in the highest calculated K,g.
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i Exhibit A F
Page12cf 20 i
Fuel assembly decay heat production is a function of core power level, j
and since the core power level remains unchanged, the decay heat load on'the spent fuel pool cooling system will not be significantly impacted by the proposed enrichment limits.
Reactor Core Operation of the Prairie Island reactors with 5.0 weight percent U-235 fuel vill not create ary initiators for accidents, including any accidents that may be different from those already evaluated in the Updated Safety Analysis Report.
Conclusions l
As discussed above, the preposed changes do not result in any significant change in the configuration of the plant, equipment design or equipment use nor do they require any change in the accident analysis methodology.
Therefore, no different type of accident is i
created. No safety analyses are affected.
The accident analyses presented in the Updated Safety Analysis Report remain bounding.
3.
The proposed amendment will not involve a significant reduction in the marnin of safety.
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Fuel Storage 1
The spent fuel pool storage configuration required by proposed specification 3.8.E will provide the administrative controls necessary to assure that fuel assemblies with the potential to form a critical array in the spent fuel pool are segregated such that K,g vill remain i
less than 0.95.
The spent fuel pool boron requ! red by proposed Specification 3.8.E will provide an additional safety margin to ensure criticality will not occur even if fuel assemblies were not stored in the required configuration.
The criticality analysis showed that K,g for the existing new fuel rack configuration would remain less than 0.95 with full density moderation.
1 The modification to prevent storage of new fuel assemblies in central 14 cells of the new fuel storage rack will assure that K,g will remain less than 0.98 when the new fuel racks are under optimum moderation conditions.
Therefore, since the calculated values of K,g have been shown to be below the regulatory limits and because they reflect a substantial sub-critical configuration for both the fuel storage areas under adverse conditions, the proposed changes will not result in a significant reduction in the plant's margin of safety.
Reactor Ccre Operation of the Prairie Island reactors with 5.0 weight percent U-235 fuel will not involve a significant reduction in a margin of safety because increasing the fuel enrichment in the reactor core does not change the conclusions of the accident analysis or safety limits of the plant. Additionally, the use of higher enrichment fuel will not j
s.
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4 Exhibit A Page13cf 20 adversely affect the operation of the fuel in the reactor. core'and does not decrease the margin of safety as described in the bases to
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any Technical Specification Conclusions Based on the conclusions of the above analysis, the proposed changes l
will not involve a significant reduction in the margin of' safety.
i Based on the evaluation described above, and pursuant to 10 CFR Part 50, f
Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant
.l hazards considerations as defined by NRC regulations in 10 CFR Part 50, l
Section 50.92.
Environmental Assessment r
Northern States Power has evaluated the proposed changes and determined that-i 1.
The changes do not involve a significant hazards consideration, i
2.
The changes do not involve a significant change in the types or j
significant increase in the amounts of any effluents that may be released offsite, or i
3.
The changes do not involve a significant increase in individual or i
cumulative occupational radiation exposure.
l Therefore, the proposed Technical Specification changes would not result in a significant radiological environmental impact.
With regard to potential nonradiological impacts, the proposed changes to the Technical Specifications involve components in the plant which are located within the restricted area as defined in 10 CFR Part 20.
They do not affect nonradiological plant effluents and have no other environmental i
impacts. Therefore, the proposed Technical Specification changes would not result in any significant nonradiological impact.
The environmental impacts of transportation resulting from the use of more highly enriched fuel and extended burnup rates have been discussed in the generic NRC staff assessment entitled "NRC Assessment of the Environmental j
Effects of Transportation Resulting from Extended Fuel Enrichment and Irradiation", dated July 7,1988, and published in the Federal Register l
(53 FR 30355). That assessment concluded that the environmental cost contributions of the extension of fuel burnup to 60 GWD/MT and increase _of fuel enrichment up to 5. 0 weight percent are either unchanged or may in fact be reduced from those summarized in Table S-4 asset forth in 10 CFR l
51.52(c). This NRC assessment bounds the proposed fuel enrichment increase because Prairie 1sland fuel enrichment will not exceed 5.0 weight I
percent U-235 and the increased fuel burnup resulting from the increased I
enrichment will not exceed 60 GWD/MT.
In conclusion, there are no significant radiological or nonradiological environmental impacts associated with the proposed amendment.
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i Exhibit A Page14 of 20 -
i 2.
Refueline Water Storane Tank Boron Concentration Limit Changes l
1 Background
-i The Refueling Water Storage Tank (RUST) supplies borated water to the chemical and volume control system during abnormal operating conditions and to the refueling canal during refueling.
During accident conditions,-
the RWST provides a source of borated water to the emergency core cooling system and the containment spray system.
The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a design basis accident, to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a design basis accident and to ensure adequate level in the containment sump to support emergency core cooling and containment spray pump operation in the recirculation mode.
Prairie Island Technical Specification 3.3.A.1.a sets the minimum RWST boron concentration at 1950 ppm.
The RWST boron concentration-requirements for recent core reload designs using 4.2 weight percent U-235 fuel assemblies have approached the current 1950 ppm limit.
It is expected that future core reload designs utilizing 5.0 weight percent U-235 fuel will require minimum RWST boron concentrations in excess of the current 1950 ppm limit in order to maintain reactor suberiticality following a LOCA.
Proposed Changes and Reasons for Changes
-l The proposed changes to the Prairie Island Technical Specifications are described below, and the specific wording changes to Technical j
Specifications are shown in Exhibits B and C.
Technical Specification 3.3.A.1.a currently limits the RWST minimum boron concentration to 1950 ppm.
The proposed revision to Specification
The increased RWST boron concentration will ensure that the reactor will remain suberitical following a LOCA for reload cores utilizing fuel e
enriched to 5.0 weight percent U-235.
Safety Evaluation During post-LDCA long term core cooling conditions, boron is necessary to maintain shutdown margin.
Evaluations of post-lDCA long term shutdown margin of Prairie Island reload cores are performed as a part of each Reload Safety Evaluation.
The RWST boron concentration requirements for recent core reload designs using 4.2 weight percent U-235 fuel assemblies have approached the current 1950 ppm limit.
It is expected that future core reload designs utilizing 5.0 weight percent U-235 funi will require minimum RUST boron concentrations in excess of the current 1950 ppmLlimit in order to maintain reactor suberiticality following a LOCA.
Increasing the minimum RWST boron concentration required by Technical Specification 3.3.A.1.a to 2500 ppm will provide adequate negative i
reactivity to ensure that the reactor will remain suberitical following a LOCA for reload cores utilizing fuel enriched to 5.0 weight percent U-235.
The evaluation of post-LOCA long term shutdown margin performed as a part of each Reload Safety Evaluation will provide continued assurance that the l
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s Exhibit A
. j Page1Sof 20 2500 ppm RWST boron concentration limit is adequate to maintain post-LOCA ihutdown margin.
Increasing the minimum RWST boron concentration to 2500 ppm will have no significant impact cui plant operations since the actual RWST boron concentration is normally above that concentration and because no change is required in the way RWST boron concentration is controlled or s
maintained.
In conclusion, because the increase in the minimum RWST boron concentration will provide additional assurance the reactor will remain-suberitical following a LOCA and because the proposed changes will not j
have a significant impact on plant operations, the health and safety of I
the public will not be adversely affected by the proposed Technical Specification changes.
- i Qptermination of Sirnificant Hazards Considsrations
}
The proposed change to the Operating bicense has been evaluated to determine whether it constitutes a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92.
This analysis is provided below:
i 1.
The proposed amendment will not involve a significant increase in l
the urobability or consecuences of an accident ureviousiv evaluated.
i i
An intrease in the required minimum RWST boron concentration has no 1
effect on the probability of any accident previously evaluated, j
l l
The increase in the required minimum RWST boron concentration will' ensure that the reactor will remain suberitical following a LOCA for reload cores utilizing fuel enriched to 5.0 weight percent U-235.
i Therefore, the proposed change will ensure the there is no increase in the consequences of a LOCA when fuel enriched up 5.0 weight percent U-235 is utilized in the core.
Therefore, based on the conclusions of the above analysis, the proposed changes will not involve a significant incrcase in the probability or consequences of an accident previously evaluated.
2.
The proposed amendment will not create the possibility of a new or l
djfferent kind of accident from any accident ureviousiv analyzed _
j Increasing the minimum RWST boron concentration to 2500 ppm will have no significant impact on plant operations since the actual'RWST boron concentration is normally above that concentration and because no change is required in the way RUST boron concentration is controlled and maintained.
Because the proposed changes do not result in any significant change-in the configuration of the plant, equipment design or equipment use nor do they require any change in the accident analysis methodology, no different type of accident is created. No safety analyses are affected. The accident analyses presented in the Updated Safety.
Analysis Report remain bounding.
.m Exhibit A Page 16 of 20 3.
The proposed amendment will not involve a significant reduction in the marcin of safety.
j Increasing the minimum RWST boron concentration required by Technical' Specification 3.3.A.l.a to 2500 ppm will provide adequate negative reactivity to ensure that the reactor will remain suberitical J
following a LOCA for reload cores utilizing' fuel enriched to 5.0 weight percent U-235.
The evaluation of posc-LOCA long term shutdown margin performed as a part of each Reload Safety Evaluation will provide continued assurance that the 2500 ppm RWST boron concentration limit is adequate to maintain post-LOCA shutdown margin.
l l
Therefore, since the increased RVST minimum boron concentration and cycle specific Reload Safety Evaluations will ensure that the reactor will remain suberitical following a*LOCA, the proposed changes will
)
l not result in a significant reduction in the plant's margin of safety.
l Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant i
hazards considerations as defined by NRC regulations in 10 CFR Part 50, l
Section 50.92.
i Environmenta! Assesmen+-
Northern States Power has evaluated the proposed changes and determined i
that:
l
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1.
The changes do not involve a significant aazards consideration, I
2.
The changes do not involve a significant change in the types or l
significant increase in the amounts of any effluents that may be l
released offsite, or 3.
The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
l Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).
Therefore, pursuant to 10 CFR Part 51 Section Sl.22(b), an environmental assessment of the proposed changes is not required.
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Exhibit A P2se17of 20 l
3.
Reactor Core Desien Description Channes Backcround Natural uranium has been previously used in the Prairie Island reactor cores in the form of axial blarikets and replacement fuel rods.
The use of l
natural uranium in the Prairie Island' reactor cores has.always been
~
evaluated as part of the reload analysia using NRC approved methodologies.
A new zirconium based fuel rod clad material called ZIRLD has been-j developed.
The Z1RLO clad material is expected to provide an improvement in corrosion resistance and dimensional stability under irradiation.
The initial batch of 5.0 weight percent fuel scheduled for insertion into the Unit 2 core in the Fall of 1993 will utilize ZIRLO clad material.
The NRC revised the acceptance criteria in 10 CFR Part 50, Sections 50.44 and i
50.46 (Federal Register dated August 31, 1992), relating to evaluations of emergency core cooling systems and combustible gas control applicable to zirealoy clad fuel to include ZIRLO clad fuel, f
Proposed Chances and Reasons for Channes The proposed changes to the Prairie Island Technical Specifications are described below, and the specific wording changes to Technical j
Specifications are shown in Exhibits B and C.
Section 5.3.A.1 is being revised to incorporate references to natural uranium and ZIRLD clad material as shown in Exhibit B.
These changes are.
intended to clarify that natural uranium and ZIRLO clad material may be used in the Prairie Island reactor cores. The incorporation of natural uranium and Z1RLD into the reactor core design description is consistent with the guidance provided in Section 4.2.1 of the Westinghouse Standard Technical Specifications, NUREG-1431.
Safety Evaluation The incorporation of natural uranium into the reactor core design description in Section 5.3.A.1 of the Prairie Island Technical Specifications is strictly a clarification. Natural uranium has been previously used in the Prairie Island reactor cores in the form of axial i
blankets and replacement fuel rods. The use of natural uranium in the Prairie Island reactor cores has always been evaluated as part of the reload analysis using NRC approved methodologies. Any future use of natural uranium in the reactor cores will also be evaluated with NRC' approved methodologies prior to use.
The use of natural uranium has no effect on the safe operation of the reactor. The incorporation of natural uranium into the reactor core design description is consistent with the guidance provided in Section 4.2.1 of the Westinghouse Standard Technical Specifications, NUREG-1431.
ZIRLD clad has a lower corrosion rate and reduced radiation induced growth which will enhance the safe operation of the Prairie Island reactors, Any use of ZIRLO clad fuel in the reactor cores will be evaluated with NRC approved methodologies prior to use.
The neutronic properties of ZIRLD are nearly identical to those of Zircaloy and therefore the use of ZIRLD is not expected to have any significant effect on the results of the core reload analyses.
The NRC revised the acceptance criteria in 10 CFR Part 50, Sections 50.44 and 50.46 (Federal Register dated August 31, 1992),
.,~
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Exhibit A Page 184,f 20 relating to evaluations of emergency core cooling systems and combustible gas control applicable to zircaloy clad fuel to include ZIRID clad fuel.
This revision to the federal regulations made ZIRLO an acceptable l
zirconium based cladding material along with zircaloy.
i In conclusion, the incorporation of natural uranium and ZIRLO clad into the Technical Specification reactor core design description will not have l
a significant impact on core reload analyses or. reactor' operation and the health and safety of the public will not be adversely affected by the proposed Technical Specification changes.
petermination of Sienificant Hazards Considerations The proposed change to the operating License has~been evaluated to 1
determine whether it constitutes a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in-Section 50.92.
This analysis is provided below:
l.
The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previousiv evaluated.
-l The incorporation of natural uranium and ZIRLO clad into the Technical-Specification reactor core design description and use of those materials in the reactor core will not effect the probability of any accident previously evaluated.
The incorporation of natural uranium into the reactor core design i
description in Section 5.3.A.1 of the Prairie Island Technical-Specifications is strictly a clarification. Natural uranium has been previously used in the Prairie Island reactor cores in the form of q
axial blankets and replacement fuel rods.
Natural uranium will i
respond to accident conditions in a manner similar to slightly enriched uranium. Additionally, fuel rods containing natural uranium instead of slightly enriched uranium will have lower gap activities I
which would slightly reduce the consequences of an accident.
l Therefore, the use of natural uranium in the reactor core has no significant effect on the consequences of an accident.
The use of ZIRLO clad material will not increase the consequences of an accident.
ZIRLO clad has improved mechanical properties such as a lower corrosion rate and reduced radiation induced growth which may improve the fuel clad response to accident conditions.
The NRC
[
revised the acceptance criteria in 10 CFR Part 50, Sections 50.44 and 50.46 (Federal Register dated August 31, 1992), relating to i
evaluations of emergency core cooling systems and combustible gas control applicable to zircaloy clad fuel to include ZIRLO clad fuel.
This revision to the federal regulations made ZIRLO an acceptable zirconium based cladding material along with zircaloy.
Therefore, based on the conclusions of the above analysis, the proposed changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.
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.o Exhibit A' Page19of 20 l 2.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident ureviously analvzed'.
t Because the proposed changes do not result in any.significant change in the configuration of the plant, equipment design or equipment use 4
nor do they require any change in the accident analysis methodology, no different type of accident is created. No safety analyses are affected.
The accident analyses presented in the Updated Safety Analysis Report remain bounding.
3.
The proposed amendment will not involve a significant reduction in the garnin of safety.
The incorporation of natural uranium into the reactor core design description of the Prairie Island fechnical Specifications is strictly a clarification.
Natural uranium has been previously used in the Prairie Island reacter cores in the form of axial blankets and replacement fuel rods. Any use of natural uranium in the reactor i
cores will be evaluated with NRC approved methodologies prior to use.
The use of natural uranium has no effect on the safe operation of the reactor.
The incorporation of natural uranium into the. reactor core design description is consistent with the guidance provided in Section 4.2.1 of the Westinghouse Standard Technical Specifications, NUREG-4
- 1431, ZIRLD clad has a lower corrosion rate and reduced radiation induced growth which will enhance the safe operation of the Prairie Island l
reactors. Any use of ZIRLD clad fuel in the reactor cores will be evaluated with NRC approved methodologies prior to use.
The neutronic properties of ZIRLO are nearly identical to those of Zircaloy and therefore the use of ZIRLO is not expected to have any significant effect on the results of the core reload analyses. The NRC revised the acceptance criteria in 10 CFR Part 50, Sections 50.44 and 50.46 (Federal Register dated August 31, 1992), relating to evaluations of emergency core cooling systems and combustible gas control applicable to zircaloy clad fuel to include ZIRLD clad fuel.
This revision to the federal regulations make Z1RLD an acceptable zirconium based cladding material along with zircaloy.
Therefore, the proposed changes will not result in a significant reduction in the plant's margin of safety.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, Northern States Power Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.
i t
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Exhibit'A Page20of 20 Environmental Assessmeng Northern States Power has evaluated the proposed changes and determined that:
1.
The changes do not involve a.significant hazards consideration, 2.
The changes do not involve a significant change in the. types or-l significant increase in the amounts of any effluents that may be released offsite, or-3.
The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51 Section 51.22(c)(9).
Therefore, pursuant to 10 CFR Part 51 Section 51.22(b), an environmental.
assessment of the proposed changes is not required.
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