ML20086S617
| ML20086S617 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/28/1995 |
| From: | Richard Anderson NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20086S606 | List: |
| References | |
| NUDOCS 9508020065 | |
| Download: ML20086S617 (18) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION-NORTHERN STATES POWER COMPANY PRAIRIE ISIAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 ,
REVISED REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED JULY 28, 1995 BORON CREDIT IN THE SPENT FUEL POOL Northern States Power Company, a Minnesota corporation, requests authorization for changes to Appendix A of the Prairie Island Operating License as shown in the attachments labeled Exhibits A, B, C, D, E AND F. Exhibit A contains a ;
description of the proposed changes, the reasons for requesting the changes, the supporting safety evaluations and significant hazards determinations.
Exhibit B contains current Prairie Island Technical Specification pages marked up to show the proposed changes. Exhibit C contains the revised Technical Specification pages. Exhibit D contains a report on the applicability of the WCAP-14181 PRA results to Prairie Island, Exhibit E contains an information copy of Unit 1 Core Operating Limits Report Pages which incorporate spent fuel pool operating limits and Exhibit F contains an information copy of the Prairie Island spent fuel rack criticality analysis with credit for soluble boron.
This letter contains no restricted or other defense information.
NORTHERN STATES POWER COMPANY By f dith 41/ A4 W ,
" Roger 0 Anderson j Director Licensing and Management Issues On this N day of juE- M4Ibeforemeanotarypublicinandforsaid ;
County, personally appe$ ed Roger 0 Anderson, Director, Licensing and 1 Management Issues, 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.
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l Exhibit A l Prairie Island Nuclear, Generating Plant License Amendment Request Dated July 28, 1995 !
Evaluation of Proposed Changes to the i Technical Specifications Appendix A of Operating License DPR-42 and DPR-60 l Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90,.the holders of Operating l Licenses DPR-42 and DPR-60 hereby propose the following changes to Appendix A,- !
Technical Specifications:
)
Backaround .;
The purpose of this license amendment request is to improve the efficiency of ';
the storage of spent fuel assemblies in the Prairie Island spent fuel pool l (SFP) while retaining the margin of safety. 1 i
This submittal proposes to take credit for the soluble boron in the spent fuel l pool water to control the subcritical condition of the spent fuel assembly 1 array. The utilization of soluble boron, which is contained'in the plant spent fuel pool, provides a simple, direct method of ensuring suberiticality. 3 This control feature retains the necessary criticality safety requirements and ;
has many benefits. Credit for soluble boron is currently used for Mode 6 i reactivity control in the reactor vessel. )
i In order to obtain approval of this proposed license amendment, an exception is.taken to existing standards (References 1,2,3). This' exception is required ,
I because the existing standards require the limiting suberitical condition of the spent fuel pool be. achieved with the pool flooded with unborated water, j i.e., no credit for soluble boron is allowed. l
.1 While the proposed license amendment proposes use of credit for soluble boron j in the spent fuel pool criticality analysis, a storage configuration has been- J defined using maximum feasible Korg calculations, as described in WCAP-14416-P i (Reference 5) and Exhibit F, to ensure that the spent fuel rack Kett will be - I less than 1.0 with no soluble boron under normal storage conditions and.
assuming nominal fuel assembly parameters and fuel rack dimensions. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal
'l conditions and to provide subcritical margin such that the spent fuel pool Kerr is maintained less than or equal to 0.95. The Prairie Island. spent fuel storage racks were analyzed utilizing the Westinghouse Spent Fuel Rack
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Criticality Analysis Methodology described in WCAP-14416-P (Reference 5).
The Prairie Island spent fuel racks have been reanalyzed to allow storage of 'l fuel assemblies with nominal enrichments up to 4.95 w/o U-235 in all storage cell locations using credit for checkerboarding and burnup. The analysis also ignores the presence of the spent fuel rack Boraflex poison panels. The following storage configuration and enrichment limits resulted from the -l criticality analysis: l i
All Cell Storage Fuel assemblies with no.pnal enrichments no greater than i 1.95 w/o U-235 for Westidghouse OFA fuel assemblies and 1.85 w/o U-235 for Westinghouse STD and Exxon fuel I
eJ c s Exhibit A Page 2 of 16 assemblies may be stored in any cell location. Fuel assemblies with initial nominal enrichments greater than these limits must satisfy a minimum burnup requirement.
3x3 Checkerboard Westinghouse OFA assemblies with nominal enrichments less than or equal to 4.95 w/o U-235 may be stored in the center of a 3x3 checkerboard. The surrounding fuel assemblies must have an initial nominal enrichment no greater than 1.50 w/o U-235 for Westinghouse OFA fuel assemblies, 1.46 w/o U-235 for Exxon TOPROD fuel assemblies, and 1.38 w/o U-235 for Westinghouse STD and other Exxon fuel assemblies. Fuel assemblies with initial nominal enrichments greater than these limits must satisfy a minimum burnup requirement.
A copy of the Prairie Island specific spent fuel pool criticality analysis is attached as Exhibit F. This analysis is provided for the information of the NRC Staff.
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WCAP-14181, " Evaluation of the Potential for Diluting PWR Spent Fuel P'ols" o
(Reference 4), was transmitted to the NRC by Westinghouse Owners Group Letter OG-95-063. This report identifies potential events which could dilute the soluble boron contained in PWR spent fuel peols and quantifies the frequency of those dilution events via a probabilistic risk assessment. This PRA was provided to the NRC as supporting information for the Westinghouse Owners Group boron credit program. No formal review of that report was requested.
In WCAP-14181 (Reference 4), a generic methodology was developed to identify <
potential events which could dilute the soluble boron contained in PWR spent fuel pools and to quantify the frequency of those events. This methodology utilized a probabilistic risk assessment (FRA) of a composite plant model to calculate the event frequency of a dilution event. The results of the PRA .
concluded that the event frequency remained less than the NRC Safety Goal Policy Statement target risk objective of IE-6/ reactor year.
1 In order to reference the results of the composite plant PRA, the Prairie Island specific features related to the spent fuel pool were compared (Exhibit D) against the features assumed for the composite plant utilized in WCAP- )
14181. As a result of the comparison to the evaluation in WCAP-14181, it was l concluded that the results of the PRA completed for the composite plant in i WCAP-14181 bound Prairie Island. Because the WCAP-14181 evaluation is bounding for Prairie Island, the spent fuel pool boron dilution event frequency for Prairie Island is less than 1.0E-6/ reactor year, j Deterministic calculations were also performed (Exhibit D) in order to define the dilution times and volumes for Prairie Island. That data was then !
compared to analogous data for the composite plant utilized in WCAP-14181.
The dilution sources available at Prairie Island were also compiled and evaluated against the dilution volume calculated, to determine the potential !
of a spent fuel pool dilution event. The deterministic evaluations show that l a large volume of water is necessary to dilute the spent fuel pool to a soluble boron i
e - t Exhibit A Page 3 of 16 concentration where a Kerr of 0.95 would be approached in the Prairie Island spent fuel pool. For Prairie Island there is no single source of water which can provide the quantity of water necessary to dilute the spent fuel pool from ,
the current normal boron concentration down to the boron concentration limit that will be incorporated in the Unit 1 Core Operating Limits Report.
A dilution event large enough to result in a significant reduction in the spent fuel pool boron concentration would involve the removal of a large quantity of water from a dilution source and a significant increase in spent fuel pool level which would ultimately overflow the pool. Such a large water volume turnover, and the likely overflow of the spent fuel pool, would be readily detected and terminated by plant personnel.
In addition, because of the large quantities of water required, and the low dilution flow rates available at Prairie Island, any significant dilution of the spent fuel pool would only occur over a long period of time (hours to days). Detection of a spent fuel pool dilution via level alarms and/or visual inspections would be expected lo.,g before a sufficient dilution would occur.
The evaluations in Reference 4 and Exhibit D, which show that the dilution of j the spent fuel pool is a low probability occurrence, combined with the maximum feasible Kerr calculation, which shows that the spent fuel rack Kerr Will remain less than 1.0 when flooded with unborated water and assuming nominal fuel assembly parameters and fuel rack dimensions, provide a level of safety 1 comparable to the conservative criticality analysis methodology required by ;
References 1, 2 and 3. l The precedence of using soluble boron in water to provide criticality control aside from normal reactor operations has already been established. Credit for soluble boron in the spent fuel pool is permitted when considering abnormal or accident conditions. Also, during refueling, soluble boron in the reactor vessel is the only direct control utilized to ensure that the reactor remains subcritical.
This License Amendment Request proposes revisions to the Technical Specifications associated with controlling the storage of assemblies with differing initial enrichments and burnup. The proposed Technical Specification changes also include the addition of Limiting Conditions for Operation, Surveillance Requirements and Design Feature changes to control the boron concentration in the spent fuel pool water under normal situations.
Proposed Channes and Reasons for Channes The proposed changes to the Prairie Island Technical Specifications are described below, and the specific wording changes to Technical Specifications are shown in Exhibit B.
A. Proposed Channes to Technical Specification Definitions The definition of the Core Operating Limits Report is being revised as shown in Exhibit B to specify that the Unit 1 Core Operating Limits Report will contain the spent fuel pool operating limits.
e o s Exhibit A Page 4 of 16 B. Proposed Channes to Technical Specification 3.8.E Based on the results and bounding conditions of the new criticality analysis (Exhibit F) for taking credit for soluble boron in the spent fuel pool, Specification 3.8.E is revised as shown in Exhibit B.
The proposed changes to Specification 3.8.E, described below, relocate all spent fuel pool storage restrictions based on a combination of fuel assembly initial enrichment and burnup to the Unit 1 Core Operating Limits Report. The current spent fuel pool boron concentration requirement that protects against a misloaded fuel assembly accident is replaced with a requirement to always maintain the spent fuel pool boron concentration in accordance with the limit specified in the Unit 1 Core Operating Limits Report.
Specification 3.0.E.1 The spent fuel storage requirements of current Specifications 3.8.E.1.a and b are relocated to the Unit 1 Core Operating Limits Report.
Specification 3.8.E.1.a is revised to require fuel assemblies to be stored in accordance with the configurations specified in the Unit 1 i Core Operating Limits Report. The anticipated Unit 1 Core Operating 4 Limits Report spent fuel storage limitations are attached as Exhibit E for the information of the NRC Staff.
The reference to Specification 3.8.E.1.b is eliminated from Specification 3.8.E.1.c and Specification 3.8.E.1.c is renumbered to 3.8.E.1.b Specification 3.8.E.1.d is renumbered to 3.8.E.1.c
. Specification 3.8.E.2 Specification 3,8.E.2.a is revised to require that the spent fuel pool boron concentration be maintained in accordance with the limits specified in the Unit 1 Core Operating Limits Report. Current Specification 3.8.E.2.a which defines the requirements for the spent fuel pool boron concentration to compensate for a misloaded fuel assembly accident, is deleted in total. That requirement is replaced by the spent fuel pool boron concentration limit specified in the Unit 1 Core Operating Limits Report.
The spent fuel pool boron concentration limit being incorporated into the Unit 1 Core operating Limits Report (Exhibit E) establishes the minimum boron concentration (1050 ppm) permitted in the spent fuel pool. ,
This boron concentration is based on the new criticality analysis (Exhibit F).
Specification 3.8.E.2.b is revised to eliminate the reference to Specification 3.8.E.2.a and to incorporate a reference to the spent fuel )
pool boron concentration limit specified in the Unit 1 Core Operating !
Limits Report. Because the requirements for a spent fuel pool l l
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verification have been eliminated from the Specification 3.8.E.2.a, the action statement in Specification 3.8.E.2~b.2 is revised to eliminate-the' requirement to perform a spent fuel pool' verification in the event t the boron concentration requirements are not met. Verifications to 1
ensure that the spent fuel pool loading configuration meets the _ ,
- requirements'specified in the Unit 1 Core Operating-Limits Report'will -
be controlled by plant procedures.
Firure TS.3.8-1 i
Current Figure TS.3.8-1, which describes the spent fuel pool unrestricted region minimum burnup requirements, is being revi,ed and-relocated to the Unit 1 Core Operating Limits Report. '
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Bases for Specificatinn 3.8.E r
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The bases for Specification 3.8.E are revised in accordance with the. >
changes made in the specification as stated above. The changes to the {
bases are shown in Exhibit B. j C. Proposed Chances to Technical Specification Table TS.4.1-2B $
I In order to. ensure that the boron concentration limits specified in !
Specification 3.8.E.2.a and the Unit 1 Core Operating Limits Report are met, the current surveillance requirements in Table TS.4.1-2B are ,
revised. In the Table,' Item 13., " Spent Fuel Pit Boron Concentration", l currently specifies the frequency of sampling tests at monthly or ,
weekly. Weekly measurements are required in conjunction with :
Specification 3.8.B.l.c when a spent fuel cask is in the spent fuel pool .i (Note 7), and in conjunction with current Specification 3.8.E.2 when spent fuel verification has not been performed (Note 8). t Table TS.4.1-2B, Item 13 is revised to specify that weekly sampling is !
now required in all cases. The weekly frequency matches that required 'l for the boron concentration in.the refueling water. storage tank. The ;j weekly frequency is sufficient based on operating experience, and l because significant changes in the boron concentration in the' spent fuel . .
_i pool are difficult to produce without detection since the pool has such a large volume (inventory) of water (Exhibit D). Soluble boron l concentration reduction requires the inflow and outflow of large volumes of water which are readily detected. Pool inventory changes' provide a j good indication of potential boron concentration changes. The pool ;
water inventory is monitored by level indication and alarms. .!
Because Table TS.4.1-2B Item 13 has been revised to routinely require weekly sampling of the spent fuel pool boron concentration, Notes 7 and ,
6 are unnecessary and are being deleted, l
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Exhibit A :
Page 6 of 16 ;
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. . i D. Proposed Chances to Technical Specification 5.6.A i The information'in-Section 5.6.A is being revised as shown in Exhibit B t l
-to include the design feature elements required to take reactivity ~
credit for the soluble boron in the spent fuel pool.
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Soecification 5.6.A'.1.b i The current Specification 5.6.A.1.b is revised to eliminate the -l requirement that the spent fuel racks maintain a Kerr of less than 0.95 f when fully. flooded with unborated water. The design feature is changed .?
to include the condition of being flooded with borated water. This proposed change takes exception'to current industry standards which' currently require the limiting suberitical condition of the spent fuel !
pool be achieved with the pool flooded with unborated water, i.e., no !
credit for soluble boron is allowed.
Specifications 5.6.A.1.c and d Current Specifications 5.6.A.1.c and 5.6.A.1.d are deleted in total' -
because the spent fuel storage configuration requirements are relocated l
to the Unit 1 Core Operating Limits Report. A new Specification j 5.6.A.1.c is incorporated to require that K.fr be less than 1.0 if the !
spent fuel pool is fully flooded with unborated water, under maximum' .
feasible' conditions as described in WCAP-14416-P. !
Firures'TS.S.6-1 and 2 i
Current Figures TS.S.6-1 and TS.S.6-2, which describe the spent fuel j pool burned / fresh checkerboard cell layout and region minimum burnup - :
requirements, are being revised and relocated to the Unit 1 Core-Operating Limits Report.
Section 5.6 References
-1 A reference to the Westinghouse Spent Fuel Rack Criticality Analysis Methodology report, WCAP-14416-P is added to Section 5.6.
E. Proposed Chances to Technical Specification 6.7.A 6 4'
New Specification 6.7.A.6.h A new Specification 6.7.A.6.b is incorporated to specify that the spent.
fuel pool operating limits shall be established'and documented in the Unit 1 Core Operating Limits Report. New Specifications.6.7.A.6.b.1 and 2 specify the spent fuel pool operating limits which are to be located in the Unit 1 Core Operating Limits Report.
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3 Exhibit A Page 7 of 16 New Specification 6.7.A.6.d New Specification 6.7.A.6.d incorporates a reference to the analytical methodology (WCAP-14416-P) to be used to determine the spent fuel pool operating limits.
Old Speci fication 6. 7. A. 6.c Old Specification 6.7.A.6.c (new Specification 6.7.A.6.e) is revised to incorporate the spent fuel pool operating limits and to incorporate a reference to the spent fuel pool criticality limits.
Old Specifications 6.7.A.6.b. c and d Old Specifications 6.7.A.6.b, e and d are renumbered to accommodate the new requirements incorporated into Specification 6.7.A.6.
F. Relocation of Spent Fuel Pool Operatinn Limits Generic Letter 88-16 was issued to encourage licensees to prepare changes to Technical Specifications related to cycle-specific parameters. The generic letter provided guidance for the relocation of certain cycle-dependent core operating limits from the Technical Specifications.
While spent fuel pool operating limits were not specifically addressed in Generic Letter 88-16, the Technical Specification changes proposed by this License Amendment Request, which relocate spent fuel pool operating limits to the Unit 1 Core Operating Limits Report, are being submitted in accordance with the guidance provided in Generic Letter 88-16. The proposed changes reference the Unit 1 Core Operating Limits Report for spent fuel pool operating limits and ensure that the spent fuel pool is maintained within the limits of the Unit 1 Core Operating Limits Report.
The proposed changes to the administrative controls section ensure that the calculation of the spent fuel pool operating limits proposed for inclusion in the Unit 1 Core Operating Limits Report will be performed in accordance with NRC-approved methodologies.
Safety Evaluation The design basis for preventing criticality in the spent fuel pool is that, including uncertainties, there is a 95% probability at a 95% confidence level that the K.tr of the fuel storage assembly array will be less than 0.95 with full density moderation. This proposed license amendment includes an exception to the additional standard condition which states that the pool water is unborated.
Deterministic Evaluation For the storage of fuel assemblies in the spent fuel storage racks, the acceptance criteria for criticality requires the effective neutron multiplication factor, Kerr, be less than or equal to 0.95, including
b .. +
Exhibit A Page 8 of 16 uncertainties. The criticality analysis performed for the Prairie Island spent fuel storage racks shows that the acceptance criteria for criticality is met for the storage of 14 x14 fuel assemblies under both normal and accident conditions with soluble boron credit, no credit for the spent fuel rack ,
Boraflex poison panels and the storage configurations and enrichment limits described above.
The Prairie Island spent fuel storage racks were analyzed utilizing the Westinghouse Spent Fuel Rack Criticality Analysis Methodology described in WCAP-14416-P (Reference 5). The analytical methods employed in analysis conform with ANSI N18.2-1973, " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", Section 5.7 Fuel Handling System; ANSI 57.2-1983, " Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations", Section 6.4.2; ANSI N16.9-1975,
" Validation of Calculation Methods for Nuclear Criticality Safety"., and the NRC Standard Review Plan, Section 9.1.2, " Spent Fuel Storage". Exception is taken to the requirement that no reactivity credit may be taken for the presence of soluble boron in the spent fuel pool as stated in ANSI 57.2-1983 (Reference 2) and the NRC position paper (Reference 3).
While this License Amendment Request proposes use of credit for soluble boron !
in the spent fuel pool criticality analysis, a storage configuration has been l defined using maximum feasible Kort calculations to ensure that the spent fuel l rack Kerr will be less than 1.0 with no soluble boron under normal storage conditions and assuming nominal fuel assembly parameters and fuel rack j dimensions. Soluble boron credit provides significant negative reactivity in )
the criticality analysis which is used to offset uncertainties and tolerances and to provide suberitical margin such that the spent fuel pool K.fr is maintained less than or equal to 0.95. Soluble boron credit and storage configuration were also used to offset the reactivity increase when ignoring the presence of the spent fuel rack Boraflex poison panels.
Revised Specification 3.8.E.1 establishes the requirements for the spent fuel rack storage configurations. The actual storage rack configuration limitations will be contained in the Unit 1 Core Operating Limits Report.
Since the anticipated pool storage configuration limitations will be similar to those being eliminated from the Technical Specifications by this license amendment request, the new limitations will not have any significant effect on normal pool operations and maintenance.
Revised Specification 3.8.E.2 establishes the new boron concentration requirements for the spent fuel pool water. As stated above, the actual boron concentration limits will be contained in the Unit 1 Core Operating Limits Report. Since soluble boron has always been contained in the spent fuel pool, the new requirement will have little effect on normal pool operations and maintenance.
Pool systems, instrumentation, and supporting systems are not modified as a result of the proposed license amendment. The operations involving spent fuel pool water cooling and cleanup do not change. The procedures related to the spent fuel pool will be upgraded as necessary to ensure that the pool boron concentration is formally controlled during both normal and accident situations. The procedures will ensure that the proper provisions,
'o a Exhibit A Page 9 of 16 precautions, and instructions exist to control the pool boron concentration and water inventory.
The Prairie Island spent fuel rack criticality analysis also addressed i postulated accidents in the spent fuel pool. The accidents that can occur in t the spent fuel pool and their consequences are not significantly effected by taking credit for the soluble boron present in the pool water as a major suberiticality control element.
Most spent fuel pool accident conditions will not result in an increase in Kerr of the spent fuel racks. Examples of such accidents are the drop of a fuel assembly on top of a rack and the drop of a fuel assembly between rack modules or between rack modules and the pool wall.
From a criticality standpoint, a dropped assembly accident occurs when a fuel assembly in its most reactive condition is dropped onto the storage racks. At Prairie Island, the spent fuel assembly rack configuration has no openings between racks or the walls, a dropped assembly can only land on the top of the racks. The rack structure pertinent for criticality is not excessively deformed. Previous accident analysis with unborated water showed the dropped assembly which comes to rest horizontally on top of the rack has sufficient ,
water separating it from the active fuel height of stored assemblies to 1 preclude neutronic interaction. For the borated water condition, the interaction is even less since the water contains boron, an additional thermal neutron absorber. j l
The radiological consequences of a dropped assembly accident in the spent fuel j pool do not change because of the presence of soluble boron in the pool water. l The current USAR accident analysis assumes that the pool water is borated. In this analysis, a high burnup assembly is dropped onto the top of the racks, all fuel rods in the dropped assembly rupture releasing the gap radioactive gases. A large fraction of the halogen gases are entrained in the pool water limiting the off-site exposures.
With respect to the insertion of a fuel assembly between rack modules or between a module and the pool wall, design of the Prairie Island spent fuel racks is such that it precludes the insertion of a fuel assembly in other than ;
a normal storage location. l l
However, two accidents can be postulated for each storage configuration which could increase reactivity beyond the analyzed condition. The first postulated accident would be a loss of the fuel pool cooling system. The second would be the misloading of a fuel assembly into a cell for which the restrictions on location, enrichment or burnup are not satisfied.
The loss of normal cooling to the spent fuel pool water causes an increase in ;
the temperature of the water passing through the stored fuel assemblies. This causes a decrease in water density which would normally result in an addition of negative reactivity. However, since Boraflex is not considered to be present and the spent fuel pool water has a high concentration of boron, a density decrease causes a positive reactivity addition.
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. .s Exhibit A Page 10 of 16
. i The misloading of a fuel assembly accident involves having restricted storage locations based on initial enrichment and burnup requirements. Special administrative controls are placed on the patterned and region loading of assemblies into these restricted locations. The misloading of an assembly constitutes not meeting the enrichment and burnup requirements of that restricted location. The result of the misloading is to add positive reactivity, increasing K.tr toward 0.95.
The amount of soluble boron required to offset each of these postulated accidents was evaluated for both the all cell and the 3x3 checkerboard storage configurations. That evaluation established the amount of soluble boron necessary to ensure that the spent fuel rack K.tr will be maintained less than or equal to 0.95 should a loss of spent fuel pool cooling or a fuel assembly misload occur. The amount of soluble boron necessary to mitigate these events has been included in the spent fuel pool boron concentration limit being incorporated into the Unit 1 Core Operating Limits Report. ;
An event or sequence of events that reduces the amount of soluble boron in the spent fuel pool adds positive reactivity, thus increasing Kerr. Significant dilution of the spent fuel pool requires two elements: (1) The inflow of a '
large volume of water from a source of water with a boron concentration less than that of the pool, and (2) A location or place to which the large volume of pool borated water goes when it is removed from the pool. ,
1 Deterministic calculations were performed (Exhibit D) in order to define the dilution time and volumes for Prairie Island. The dilution sources available e at Prairie Island were compiled and evaluated against the dilution volume calculated, to determine the potential of a spent fuel pool dilution event.
The deterministic evaluations show that a large volume of water is necessary to dilute the spent fuel pool to a soluble boron concentration where criticality would be approached in the Prairie Island spent fuel pool. ,
The deterministic evaluation (Exhibit D) concluded that large volumes of water l i
are necessary to dilute the spent fuel pool water to less than the boron
< concentration limit of 1,050 ppm. The availability of such large water ,
supplies on site is limited. In fact, there is no single source of water at Prairie Island which can provide the quantity of unborated water necessary to dilute the spent fuel pool from the current normal boron concentration down to the boron concentration limit of 1,050 ppm. Also, the transferability of the available water supplies to the pool is very low due to the small number of possible flow paths and in many cases impossible due to the physical arrangement of the pool relative to the supplies.
A dilution event large enough to result in a significant reduction in the spent fuel pool boron concentration will involve the removal of a large quantity of water from a dilution source and a significant increase in spent fuel pool level which would ultimately result in pool overflow. Such a large water volume turnover, and the likely overflow of the spent fuel pool, would be readily detected and terminated by plant personnel.
1 In addition, because of the large quantities of water required, and the low dilution flow rates available at Prairie Island, any significant dilution of the spent fuel pool would only occur over a long period of time (hours to
w 4 e-yv-
. Exhibit A Page 11 of 16 days); Detection of a spent fuel' pool dilution via level alarms'and/or visual inspections would be expected long before a significant dilution.would occur.
Therefore,.it is highly unlikely that any dilution event in the spent-fuel' pool =could result the reduction'of the spent fuel pool to:less than.the 1,050
. ppm limit.
Probabilistic Evaluation In WCAP-14181, " Evaluation of the Potential for Diluting PWR Spent Fuel Pools" (Reference 4), a generic methodology was applied to identify potential' events which could dilute the soluble boron contained in PWR spent fuel pools and to quantify the frequency of those events. This methodology' utilized a
,probabilistic risk assessment (FRA) of a composite plant model to calculate the event frequency of a dilution event. The results of the PRA concluded that the event frequency remained less than the NRC Safety Goal Policy Statement target risk objective of 1E-6/ reactor year.
In order to reference the results of the composite plant PRA, the Prairie Island specific features related to the spent fuel pool were compared (Exhibit D) against the features assumed for the- composite plant utilized in WCAP-14181. As a result of the comparison to the evaluation in WCAP-14181, it was concluded that the results of the PRA. completed for the composite plant in-WCAP-14181 bound Prairie Island. Because the WCAP-14181 evaluation is bounding, the spent fuel pool boron dilution event. frequency for Prairie Island is less than 1.0E-6/ reactor year.
The results of these probabilistic evaluations demonstrate that the risk'of f exceeding a Kerr of 0.95 in the spent fuel pool when taking credit for the i soluble boron remains a very small contributor to the'overall risk associated -l with the operation of the Prairie Island plant. !
Relocation of Spent Fuel Pool Operatina Limits j a
The relocation of the spent fuel pool operating limits from the Technical i Specifications to the Unit 1 Core. Operating Limits Report has no impact upon.
plant operations or safety. No safety-related equipment, safety function, or plant operations will be altered as a result of the proposed changes. Since .
the applicable Updated Safety Analysis Report limits will.be maintained and l the Technical Specifications will continue to require operation of the spent j fuel pool within operating limits calculated by NRC-approved methodologies, -)
the proposed relocation of the spent fuel pool operating limits is administrative in nature. Appropriate actions to be taken if limits are violated will also remain in the Technical Specifications.
Exhibit A Page 12 of 16 Conclusion The combination of the following provide a level of safety comparable to the conservative criticality analysis methodology required by References 1, 2 and 3:
- 1. The maximum feasible Kerr calculation, which assumes normal storage conditions and nominal fuel assembly parameters and fuel rack dimensions, shows that the spent fuel rack K.tr will remain less than 1.0 when flooded with unborated water,
- 2. The unavailability of the large volumes of water that are necessary to dilute the spent fuel pool, and
- 3. The probabilistic evaluation shows that the dilution of the spent fuel pool is a low probability occurrence.
In conclusion, Northern States Power believes there is reasonable assurance that the health and safety of the public will not be adversely affected by the proposed Technical Specification changes.
Determination of Sinnificant 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 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 sinnificant increase in the probability or consecuences of an accident previous 1v evaluated.
There is no increase in the probability of a fuel assembly drop accident in the spent fuel pool when considering the presence of soluble boron in the pool water for criticality control. The handling of the fuel assemblies in the spent fuel pool has always been performed in borated water.
The consequences of a fuel assembly drop accident in the spent fuel pool are not effected when considering the presence of soluble boron. At Prairie Island, a dropped assembly can only come to rest on the top of the storage racks. Previous accident analysis with unborated water showed the dropped assembly which comes to rest horizontally on top of the rack has sufficient water separating it from the active fuel height of stored assemblies to preclude neutronic interaction, j There is no increase in the probability of the accidental misloading of spent fuel assemblies into the spent fuel pool racks when considering the presence of soluble boron in the pool water for criticality control. j The probability of misplacing a fuel assembly in the spent fuel pool is i not increased because fuel assembly placement will continue to be controlled pursuant to approved fuel handling procedures and will be in .
accordance with the spent fuel rack storage configuration limitations in j the Unit 1 Core Operating Limits Report. j i
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Exhibit A l Page 13 of 16,- .
There.is no increase in the consequences of'the accidental misloading of )
o spent fuel assemblies into the spent fuel pool racks because criticality j
, analyses demonstrate that_ the pool.will remain suberitical following .an l
'. ~
accidental misloading if:the pool contains an' adequate boron l concentration. The proposed Technical Specifications and the Unit 1 -i Core Operating Limits Report limitations will ensure that an adequate-spent fuel pool boron concentration will be maintained. ].
There is no increase in the probability of the loss of normal cooling to {
the spent fuel pool water when considering the presence of soluble boron ,j
- in the pool water for suberiticality control since a high concentration l of soluble boron has always been maintained in the spent fuel pool j water. i i
The loss of normal cooling to the spent fuel pool will cause.an increase 'i in the temperature of the spent fuel pool water. This will cause a '!
decrease in water density which would normally result in an addition of j negative reactivity. However, since Boraflex is not considered to be q present, and the spent fuel pool water has a high concentration of- t boron, a density decrease causes a positive reactivity addition. The; .!
amount of soluble boron required to offset this postulated accident was- j evaluated for.the allowed storage configurations. The amount of soluble ';
boron necessary to mitigate these accidents and ensure that~the spent i fuel rack K.tr will be maintained less than or equal to 0.95.has been i included in the spent fuel. pool boron concentration limit being l incorporated into the Unit 1 Core Operating Limits Report. Because 1 adequate soluble boron will be maintained in the' pool water, the .
consequences of a loss of normal cooling to the spent fuel pool will not j be increased. l Therefore, based on the conclusions of the above analysis, the proposed j changes will not involve a significant increase in the probability or l 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 previous 1v analyzed.
3 Spent fuel handling accidents are not new or different types of j accidents, they have been analyzed in the Upgraded Safety Analysis 'I Report.
Criticality accidents in the spent fuel pool are not new or different i types of accidents, they have been analyzed in the Upgraded Safety +
Analysis Report and in Criticality Analysis reports associated with -!
specific licensing amendments for fuel enrichments up to 5.0 weight ,
percent U-235. ;
i Since soluble boron has always been maintained in the' spent fuel pool ;
water, the possibility of a spent fuel pool dilution has always existed, l the implementation of controls for the soluble boron will not create the j possibility of a new or different kind of accident.
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Exhibit A Page 14 of 16 The control of the soluble boron in the spent fuel pool was considered- .
for the current Technical Specification requirements. The boron 4 concentration required for storage was well below that normally
, - maintained in the. spent fuel pool. A loss of soluble' boron from the
{
spent fuel pool accident was considered only on the basis of the double !
contingency rule relative to a misloaded assembly accident. j With the. soluble boron now a major factor in controlling criticality, an evaluation of spent fuel pool dilution events was completed. A generic methodology was applied (Reference 4) to identify potential events which {
could dilute the soluble boron contained in PWR spent fuel pools:and to' l quantify the frequency of those events. This methodology utilized a l probabilistic risk assessment (FRA) of a composite plant'model to' i calculate the event frequency of.a dilution event. The results of the r PRA concluded that the event frequency remained less than the NRC Safety Coal Policy Statement target risk objective of lE-6/ reactor year. ,
i The Prairie Island specific features related to the spent fuel pool were l compared (Exhibit D) against the features assumed for the composite .
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plant utilized in the generic evaluation (Reference 4). -As a result of the comparison to the generic evaluation, it was concluded that the i results of the PRA completed for the composite plant bound Prairie Island. Because the generic evaluation is bounding for Prairie Island, the spent fuel pool boron dilution event frequency for Prairie Island is less than 1.0E-6/ reactor year.
The results of these probabilistic evaluations demonstrate that the risk of achieving criticality in the spent fuel pool when taking credit.for the soluble boron remains a very small contributor to'the overall risk associated with the operation of the Prairie Island plant.
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Proposed Specifications 3.8.E.1 and 2, which ensure the maintenance of the spent fuel pool' storage configuration and boron concentration, do- ,
not represent new concepts but upgraded ones.
Current Specification 3.8..E.2, which covers'the storage of restricted fuel assemblies in an unverified condition, and Specification 3.8.B.1.c. .;
for the loading of fuel assemblies into a cask in the spent fuel pool, i contain requirements for spent fuel pool boron concentration. The l actual boron concentration in the spent fuel pool has always'been kept l at a higher value for refueling purposes. The criticality analysis- l (Exhibit F) determined that a boron concentration of 1,050 ppm results '
in a K.tr $ 0.95 including all the calculational uncertainties and additional margin to compensate for the possibility of loss of cooling.
Verifications will continue to be performed to ensure that the spent fuel pool loading configuration meets specified requirements. These
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verifications will be controlled by plant procedures.
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- Exhibit A Page 15 of 16 The proposed changes will not create the possibility of a new or different kind of accident. There is no significant change in plant configuration, equipment design or equipment. The safety analysis for dilution accidents has been expanded. The accident analysis in the Updated Safety Analysis Report remains bounding.
- 3. The nroposed amendment will not involve a sienificant reduction in the marnin of safety.
Proposed Specifications 3.8.E.1 and 2 and the associated spent fuel storage operating limits in the Unit 1 Core Operating Limits Report will provide adequate safety margin to assure that the stored fuel assembly array will always remain suberitical. Those limits are based on a plant specific criticality analysis (Exhibit F) performed in accordance the Westinghouse spent fuel rack criticality analysis methodology described in Reference 5.
While the criticality analysis utilized credit for soluble boron, a storage configuration has been defined using maximum feasible K.rt calculations to ensure that the spent fuel rack K.tr will be less than 1.0 with no soluble boron under normal storage conditions and assuming nominal fuel assembly parameters and fuel rack dimensions. Soluble boron credit is used to offset uncertainties, tolerances and off-normal conditions and to provide suberitical margin such that the spent fuel pool K.tr is maintained less than or equal to 0.95.
The loss of substantial amounts of soluble boron from the spent fuel i pool which could lead to exceeding a Kerr of 0.95 during accidents and. )
under adverse conditions has been investigated and shown to be highly l improbable. l The combination of the probabilistic evaluation which shows that the dilution of the spent fuel pool is a low probability occurrence, the maximum feasible K rt calculation which shows that the spent fuel rack K cc will remain less than 1.0 when flooded with unborated water and assuming nominal fuel assembly parameters and fuel rack dimensions, and the unavailability of the large volumes of water which are necessary to dilute the spent fuel pool, provide a level of safety comparable to the conservative criticality analysis methodology required by References 1, 2 and 3.
Therefore, the proposed changes in this license amendment will not result in a significant reduction in the plant's margin of safety.
Conclusion Based on the evaluation above, and pursuant to 10 CFR 50, Section 50.91, Northern States Power Company has determined that operation of the Prairic 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 50, Section 50.92.
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..: s Exhibit A l Page 16 of 16 l
.x ;;
f Environmental Assessment ;
Northern States Power has evaluated the proposed changes and determined.that: l
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- 1. The changes do not involve:a significant hazards consideration, ;
'2. The changes do not involve a significant change in'the types or .
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 cxclusion 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 ^ l aosessment of the proposed changes is not required. ,
i References -
l 1
- 1. USNRC Standard Review Plan for the Review of Safety Analysis Reports for f Nuclear Power Plants, LWR Edition, NUREG-0800, June 1987. .!
- 2. USNRC Spent Fuel Storage Facility Design Bases (for Comment) Proposed .
Revision 2, 1981, Regulatory Guide 1.13. ;
i
- 3. ANS, Design Requirements for Light Water Reactor. Spent Fuel Storage Facilities at Nuclear Power Stations,-ANSI /ANS-57.2-1983. i 4 WCAP-14181, " Evaluation of the Potential for. Diluting PWR Spent Fuel Pools", July 1995.
- 5. WCAP-14416 P, " Westinghouse Spent Fuel Rack Criticality Analysis ,
Methodology", June 1995. l l
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r Exhibit B Prairie Island Nuclear Generating Plant j License Amendment Request Dated July 28, 1995 Proposed Changes Marked Up p On Existing Technical Specification Pages '
Exhibit B consists of existing Technical Specification pages with the proposed changes highlighted on those pages. The pages affected by this License Amendment Request are listed below:
TS.1-2 TS.3.8-4 TS.3.8-5 i Figure TS.3.8-1 ,
Table TS.4.1-2B (Page 1 of 2)
Table TS.4.1-2B (Page 2 of 2)
TS.S.6-1 ,
TS.5.6-3 Figure TS.S.6-1 Figure TS.S.6-2 TS.6.7-4 TS.6.7-5 B.3.8-2 B.3.8-3 B.3.8-4 t
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