ML20248J874
| ML20248J874 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/06/1989 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20248J877 | List: |
| References | |
| NUDOCS 8904170126 | |
| Download: ML20248J874 (13) | |
Text
-
7------------
l k N'>rthern States Power Company 414 Nicollet Mall Minneapohs, Minnesota 55401 Telephons (612) 330-5500 1
April 6, 1989 10 CFR Part 50 Section 50.90 Director of Nuclear Reactor Regulation ;
U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR CENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 License Amendment Request Dated April 6, 1989 Increased Fuel Enrichment Limit Changes Attached is a request for a change to the Technical Specifications, Appendix A of the Operating Licenses, for the Prairie Island Nuclear Generating Plant. This request is submitted in accordance with the provisions of 10 CFR Part 50, Section 50.90. i The Prairie Island Technical Specification currently limits fuel in the storage pool to a U-235 loading of less than or equal to 39.0 grams of U-235 per axial centimeter of fuel assembly (average). In order to support longer fuel cycles at Prairie Island, it is necessary to increase fuel enrichments beyond this limit. Based on the results of a criticality analysis included as Exhibit C to this request, the Prairie Island Technical Specifications are being revised to allow the storage of 4.25 wt% U-235 fuel in the new fuel pit and the spent fuel pool.
Exhibit A contains a description of the proposed changes, the reasons for requesting the changes and the supporting safety evaluation /significant hazards determination. Exhibit B contains pages of the Prairie Island Technical Specifications revised to show the proposed changes. Exhibit C contains a i Westinghouse report describing the criticality analysis performed on the Prairie Island new and spent fuel storage facilities. .
go0/
lll .
4
-o.,
, 1 i
Northem States Power Company Director of NRR April 6, 1989 Page-2 NRC review and approval of this License Amendment Request is requested by August l l, 1989 in order to support our current nuclear fuel management plans. Please l f\- contact us if you have any questions related to this License Amendment Request. !
l W
David Musolf Manager - Nuclear Support Services i
i c: Regional Administrator-III, NRC l NRR Proj ect Manager, NRC 1 Senior Resident Inspector, NRC MPCA Attn: J W Ferman C Charnoff i
i l
l l;
I l
f f
}
4 i i UNITED STATES NUCLEAR REGUIATORY COMMISSION ,
1 NORTHERN STATES POWER COMPANY f I
PRAIRIE ISIAND NUCLEAR GENERATING PIANT DOCKET NO. 50-282 l 50-306 i
f REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED APRIL 6, 1989
)
Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown on the attachments labeled Exhibits A, B, and C. Exhibit A describes the proposed changes, reasons for the changes, and a significant hazards evaluation. Exhibits B is a copy of the Prairie Island Technical Specifications incorporating the proposed changes. Exhibit C is a report supporting the requested changes.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY Du *~ M D -
By David Musolf i Manager-Nuclear Support Services .
On this d N ay'of d riAi/~
/G/9 before me a notary public in and for said County, personally appeared David Mdsolf, Manager-Nuclear Support Services, and being first duly ' sworn acknowledged that he is - 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 knowledge, information, and belief the statements made in it are true and that it is not interposed for delay. j WYYb /
(/~ / r!
M :'.='.v::::.a '
l JUDY L KLAPPER!CK NOTARY PUGUC--MINNESOTA ANOKA COUNTY My Commission Expues Sept 23,1993
- V: = :::: w :::::::::::: ::n ,
l I
l
) -
3
--- _--- _ J
.s- '\
t
.Q' Exhibit A ] 1
'i I j
Prairie Island Nuclear Generating Plant- ;
License Amendment Request Dated April 6, 1989 )
.1 Evaluation of Proposed Changes to the '
Technical Specifications Appendix A of Operating License DPR-42 and DPR-60 I
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:
- 1. Fuel Storage Enrichment Limit Channes Prooosed Changes Revise specification 5.6.A as shown on page TS.5.6-1 in Exhibit B to raise the fuel storage enrichment limit to a batch' average.4.25 weight percent (wt%) U-235 and to incorporate a reference to the limitations on storage-of low burnup fuel in proposed Technical Specification 3.8~E .
Add new specification;3.8 E, as shown on page 3.8-3 in Exhibit B, which specifies the requirements for the storage of low burnup fuel in the spent fuel pool.
Incorporate bases for proposed specification 3.8.E into the fuel handling
. specification bases as shown on page TS.3.8-5 in Exhibit B. j Revise specification 5.3.A.2 as shown on page TS.5.3-1 in Exhibit B to raise.the maximum batch average enrichment to 4.25 wt% U-235.
Reason For Chances Specification 5.6.A currently limits fuel in the storage pool to a U 235 ,
loading of less than or equal to 39.0 grams of U-235 per axial centimeter of fuel assembly (average). In order to support longer fuel cycles at Prairie Island, it is necessary to increase fuel enrichments beyond this limit. Based on the results of a criticality analysis included as Exhibit '
C to this request, specification 5.6.A is being revised to allow the storage of 4.25 wt% U-235 fuel in the new fuel pit and the spent fuel pool. l l
A-1 l
i
4 3
Proposed specification 3.8.E incorporates restrictions on the storage of low burnup fuel assemblies which resulted from the criticality analysis for the storage of 4.25 wt% fuel in the Prairie Island spent fuel pool (Exhibit C). The criticality analysis found that in order to assure a K,ff 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 shown in Figure 7 of Exhibit C or the spent fuel pool boron ;
concentration would have to be maintained greater than 250 ppm. l l
Specification 5.3.A.2 currently limits U-235 loading of a reload core to a nominal 39.0 grams of U-235 per axial centimeter of fuel assembly (average). As discussed above, in order to support longer fuel cycles at Prairie Island it is necessary to increase fuel enrichments beyond this limit. The proposed change to specification 5.3.A.2 increases the maximum batch average enrichment in the core to 4.25 wt% U-235. Cycle specific reload analyses using NRC approved methodology will confirm the acceptability of reactor operation with the higher enrichment reload fuel, i
i Safety Evaluation and 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 required by 10 CFR Part 50, Section 50.91 using the standards provided in i Section 50.92. This analysis is provided below: i
- 1. The proposed amendment will not involve a significant increase in the probability or consequences of an accident previousiv evaluated.
Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is !
done by fixing the minimum separation between assemblies and inserting neutron poison between assemblies.
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 K,g 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 as the new fuel racks, ;
K,ff must be less than 0.98 for low density optimum moderation conditions.
Scent Fuel Storage The criticality analysis performed on the Prairie Island spent fuel pool (Exhibit C) found K,g to be less than 0.95 including uncertainties at a 95/95 probability / confidence level for fuel enriched to 4.25 wt% U-235 if one of the following conditions was met:
1
- 1. Fuel with average assembly burnup less than approximately 4,300 MWD /MTU (Exhibit C, Figure 6) is only stored in the three out of four configuration shown in Figure 7 of Exhibit C, or
- A-2
75 y ,
+-
't i 4-4
- 2. Spent fuel' pool' boron concentrate 1on is maintained above 250. ppm.
Administrative controls implemented'by proposed. specification-3.8.E
'will ensure that-fuel'with average assembly burnup'less.than 5,000
-MWD /MTU will only be stored.in~the three out of four(configuration-4
.describedfabove.
In addition, proposed specification.3.8.E.will require the' fuel! pool i
- boron concentration to be maintained above 500 ppa'whenever fuel'with, j
average assembly burnup less than_5,000 MWD /MTU is, stored in the spent 1
' fuel pool. This requirement.will provids an additional margin of: 1 safety to ensure criticality will not_ occur even if new fuel assemblies 1 were-not stored in the required three out ofifour array.
The 5,0'0_
0 MWD /MTU and.500 ppm limits.specified in proposed specification-3.8.E provide an additional margin of safety over the; burnup and boron limits resulting from the criticality analysis, d Spent fuel gap activities, which are a function of fuel'asse'bly m burnup,'are not directly affected by an increase in fuel assembly enrichment._ The use of hi;her enrichment fuel is .not' expected to
-increase-the maximum fuel burnup'beyond the'value assumed in the fuel.
' handling accident' analysis. ,
1 Fuel assembly; decay = heat' production.is a' function of! core power = level, ,
and since the core power level remains' unchanged,' the decay heat' . f
. generated by_a spent fuel assembly will not be significantly impacted ."
by the proposed enrichment limits and the fuel handling accident ,
analysis willinot be affected.'
New Fuel'Storane
<The criticality analysis performed on the' Prairie Island'new fuel racks-(Exhibit. C) . found Kg,f. to' be less than 0.95 including uncertainties at .,!
a 95/95~ probability / confidence level for fuel' enriched to a batch average 4.25'wtt 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 (Exhibit C, Figure 4) was assumed to be completely filled. .;
Further evaluation found that if three rows of eleven new fuel storage cells were removed from service, thereby reducing the new fuel racks to j an array of-5 by 11 cells, the 0.98 K,,, limit would be met under low density optimum moderation conditions. Therefore, prior to placing any fuel assemblies with enrichments greater than currently approved, into the new fuel racks, the new fuel racks will be modified to preclude the storage of fuel assemblies in the three rows of eleven cells assumed to l be empty by the criticality analysis.
i L
A-3 l
. y
4 <
n Based on the results of criticality analysis described in Exhibit C, the implementation of the administrative controls and modifications described above will ensure that the Prairie Island new and spent fuel j
o storage facilities will remain substantially suberitical at all times i and the probability of a criticality event will not.be increased by the l storage of 4.25 wt% U-235 fuel. {
3 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.
- 2. The proposed amendment will not create the possibility of a new or y different kind cf accident from any accident oreviousiv analyzed. d As described above, the storage of higher enrichment fuel in the new fuel racks will require three rows of the racks to be modified to prevent insertion of new fuel assemblies. The modifications and their ;
installation will be minor in nature and as such will not create the i possibility of a new or different kind of accident. l 1
The administrative controls which will be impicmented to control the storage of higher enrichment fuel will only affect where spent fuel assemblies can be stored and the tequired spent fuel pool boron concentration. Limiting where fuel assemblies can be stored in the spent fuel pool vill 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 l spent fuel pool. Therefore, the implementation of these administrative controls will not create the possibility of a new or different kind of accident.
The requirement for maintaining spent fuel pool boron concentration greater than 500 ppm will ensure that K,,, will remain less than 0.95 even if new fuel assemblies were inadvertently inserted in the empty cells of the three out of four storage configuration.
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 the boraflex sheets 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 borsflex panels in the spent fuel 3 pool. This four inch gap is considered conservative Enned on neutron radioassay measurements of the boraflex poison mate;'ot. The centerline positioning of the gap is also conside.'c ; conservative because it resulted in the highest calculated K,ff.
A-4
- e s e
\..
Fuel assembly decay heat production is.a function of core power level, and since the core poe:r level remains unchanged, the decay' heat load
.on the spent fuel pool cooling system will not be significantly impacted by the proposed' enrichment limits.
As discussed'above, 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. .Therefore, 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 margin of safety.
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 Keg will remain .
less than~0.95. The spent fuel pool boron required by Proposed specification 3.8 E will provide an additional safety margin.to ensure criticality will not oc_ur even if new fuel assemblies were not. stored ,
in the required three out of four array. !
The. criticality analysis showed that Keg for the existing new fuel rack configuration would remain less than 0.95 with' full density moderation, i The modification to prevent storage of new fuel assemblies in three rows of the new fuel storage rack will assure that Keg will remain less than 0.98 when the new fuel racks'are under optimum moderation conditions.
Therefore, since the calculated values of Ke# have been shown to be below the regulatory limits and because they rnflect a substantial sub-critical configuration for both the fuel storage areas under adverse i' conditions, the proposed changes will not result in a significant reduction in the plant's margin of safety. 3 The Com.*ission has provided guidance concerning the application of the standards in 10 CFR 50.92 for determining whether a significant hazards consideration exists by providing certain examples of amendments that will
- likely be found to involve no significant hazards considerations. These examples wore published in the Federal Register on March 6, 1986 The changes to the Prairie Island Technical Specifications proposed above ;
are equivalent to NRC example (vi), because they involve changes which either muy result in some change in the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, i
A-5 1
l L f
h
- h .'
' or i
i
.but where,the results of the. change are clearly within~all. acceptable criteria with respect to the system or component specifiedlin the Standard 3 -Review Plan. Based on this guidance and the reasons discussed above, we s.
have' concluded that the proposed changes-do not involve a significant hazards consideration.
'I 2 '. Fuel Storaze'Desian' Description Channes Procosed Channes Revise specification 5.6.A as shown on'page TS.5.6-1 in Exhibit B to better define the new and spent fuel storage design K,,, values and to _
include a reference to Section 10.2 of the USAR for design information on new and spent fuel storage.
Revise Reference 1 on page TS.S.6-2 as shown in Exhibit B to refer to USAR Section 10.2 rather than Section 10.1.
Reason For Channes a
Specification 5.6.A is being revised to ensure that the discussion of the fuel storage design K,,, values is. consist'ent with the results of the criticality analysis The reference to'Section 10.2 of the'USAR is beir- cerporated to clearly.
A identify ~ the location of detailed design informatiu en the'new and spent l- fuel storage facilities.
Reference 1 on page TS.5.6-2 currently refers to USAR Section 10.1 for fuel storage design information. The correct reference for new and spent fuel design information is USAR Section 10.2.
Safety Evaluation and Determination of Significant Hazards Considerations The proposed change to the Operating License 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:
- 1. The proposed amendment will not involve a'significant increase in the probability or consequences of an accident oreviousiv evaluated.
Thel proposed administrative changes are intended to clarify the Technical Specifications by ensuring that the discussion of fuel storage design K values is consistent with the results of the
.criticalityanalfs'isandthatthelocationofdetaileddesign information on the.new and spent fuel storage facilities is clearly identified. -Because the proposed changes are administrative and only provide clarification they will not involve a significant increase in the probability or consequences of an accident previously evaluated.
A-6 l
\T'
- t. . .-
i u
- 2. The proposed'amondment will not create the possibility of a new or different kind of accident from any accident oreviousiv analyzed.
The proposed hhanges are administrative in nature. No safety analyses are affected by these. changes. No new or different type of accident is created. The accident analyses' presented in the Updated Safety _
Analysis Report remain bounding.
- 3. The proposed amendmentLwill not involve a significant reduction in the margin of safety.
~
1 The_ proposed changes involve only administrative changes intended to i clarify existing Technical Specifications. The margin to safety is not affected by these changes.
The Commission has provided guidance concerning the application of the <. I standards in 10 CFR.50.92 for determining whetheria significant hazards consideration exists by providing certain examples of amendments that will
~
likely be found to involve no significant hazards considerations. These examples were published in the Federal Register on March 6, 1986. 1 The changes to the. Prairie Island Technical Specifications proposed in this amendment request are equivalent to NRC example (i), because they involve purely administrative changes intended only to clarify existing Technical Specifications. Based on this guidance and the-reasons discussed above, we have concluded that the proposed change does not
' involve a significant hazards consideration.
- 3. Reactor Core Design De scription Channes Proposed Channes Revise specification 5.3.A.1 As shown on page TS.5.3-1 in Exhibit B to remove the reference to the approximate weight of uranium in the reactor Core.
Delete the reference to' reload core average fuel enrichment in specification 5.3.A.2 as shown on page TS.5.3-1 in Exhibit B.
Reason For Chances The amount of uranium contained in the reactor core and the average enrichment of the reload core are cycle specific variables which can be found in cycle specific reload design information and as such it is not
< necessary.to reference them in Section 5 of the Technical Specification.s.
'The deletion of this information'is consistent with the guidance provided ]
the Standard Technical Specifications.
1 A-7 i
i
. . i Safety Evaluation and Determination of Significant Hazards Considerations The proposed change to the Operating License 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:
- 1. The proposed amendment will not involve a significant increase in ;
the probability or consequences of an accident oreviously evaluated.
The proposed administrative changes are intended to clarify the Section 5 of the Technical Specifications by removing references to inaccurate cycle specific information. The deletion of this information is consistent with the guidance provided in the Standard Technical Specifications. Because the proposed changes are administrative, only provide clarification and are consistent with the Standard Technical Specifications, they will not involve a significant increase in the j 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 oreviously analyzed.
The proposed changes are administrative in nature. No safety analyses are affected by these changes. No new or different type of accident is created. The accident analyses presented in the Updated Safety Analysis Report remain bounding.
- 3. The proposed amendment will not involve a significant reduction in the margin of safety. _
The proposed changes involve only administrative changes intended to ;
clarify existing Technical Specifications. The margin to safety is not affected by these changes.
The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 for determining whether a significant hazards consideration exists by providing certain examples of amendments that will likely be found to involve no significant hazards considerations. These examples were published in the Federal Register on March 6, 1986.
The ch ages to the Prairie Island Technical Specifications proposed in this amendment request are equivalent to NRC example (1), because they involve purely administrative changes intended only to clarify existing Technical Specifications. Based on this guidance and the reasons discussed above, we have concluded that the proposed change does not l
involve a significant hazards consideration.
1 A-8
_-_______-----___7-_.
t;t ' ,' . .,,i~ #l .1
...g- p y -- ,,
_' :f l; , -
Environmental Assessment' I. i f
l- f .This license amendment. request does.not change, effluent types or total
' effluent amounts nor does'.it involve an increase in power level. Therefore, this change will not result in any significant environmental impact.
i A-9 l
___i-________ .. - - . _ . . .
4 Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated April 6, 1989 Proposed Changes Marked Up On Existing Technical Specification Pages Exhibit B consists of existing Technical Specification pages with the proposed changes written on those pages. Existing pages affected by this License Amendment Request are listed below:
TS.3.8-3 TS.3.8-5 TS.S.3-1 TS.S.6-1 TS.5.6-2