ML20215C709
| ML20215C709 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/12/1987 |
| From: | Cutter A CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20215C712 | List: |
| References | |
| 87TSB07, 87TSB7, NLS-87-103, NUDOCS 8706180220 | |
| Download: ML20215C709 (11) | |
Text
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f" Carolina Power & Light Company P, o. Box 1551 e Raleigh, N. C. 27602 j
(919) 8364231 JUN 121987 -
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SERIAL: NLS-87-103 4
Nuclear Engineering & Ucensing 10CFR50.90 '
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87TSB07-j United States Nuclear Regulatory Commission -
1 ATTENTION: Document Control Desk
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Washington, DC -20535-l BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 i
REQUEST FOR LICENSE AMENDMENT FUEL ASSEMBLY / FUEL STORAGE DESCRIPTIONS I
Gentlemen:
SUMMARY
In accordance with the Cor' f Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light C iany hereby requests a revision to the. Technical Specifications (TS) for the b.unswick Steam Electric Plant (BSEP), Unit Nos. I and 2.
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The proposed changes affect the description of fuel assemblies found in TS 5.3.1 and the enrichment limitations imposed in-TS 5.6.1.1 and TS 5.6.1.2. These changes'will j
facilitate the handling and storage of improved, higher enrichment fuel assemblies which the Company intends to use for upcoming operating cycles. The current TS, by'ingosing.
limits on U-235 axial gm/cm and enrichment, do not recognize the reactivity hoici down associated with integral burnable poisons. The _ conversion to a limit based upon the maximum allowable infinite lattice multiplication factor accounts for the additional poisoning effect of integral burnable poisons, allowing high enrichment fuel assemblies while retaining the assurance that adequate margin to criticality is maintained in the-new and spent fuel storage rack configurations. Through the use of integral burnable poisons, fuel assemblies with U-235 axial gm/cm and enrichment higher than that allowed by the current TS will actually be less reactive than the fuel assemblies analyzed to meet current TS limits and, therefore, will be bounded by the existing analyses. Carolina Power &. Light Company requests issuance of the proposed amendment by October 5,
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1987 in order to support receipt of new fuel for the upcoming BSEP-2 Reload 7 outage.
1 DISCUSSION Currently, Specification 5.3.1 delineates the fuel types existing in the reactor core and provides maximum average enrichments for the initial core loading and reload fuel assemblies. This specification has been revised to limit the fuel assemblies contained in the core to those whose mechanical design is consistent with designs approved by the -
NRC as described in the fuel vendor's generic fuel design licensing topical reports. In addition, the limitations on maximum fuel enrichment have been removed. The fuel's l
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i acceptability for operation is ensured by the performance of a cycle-specific safety I
analysis for each reload. The revised Specifications 5.6.1.1 and 5.6.1.2 impose infinite lattice multiplication factor (k-infinity) limitations which ensure the fuel can be safely j
handled and stored.
Specification 5.6.1.1 requires that the new fuel storage racks be maintained such that there is sufficient center-to-center distance between stored fuel assemblies to ensure a keff equivalent to less than 0.90 when dry and less than 0.95 when flooded with unborated water. The proposed change specifies that this requirement is met by limiting the -
k-infinity of new fuel assemblies to less than or equal to 1.31 in an infinite core geometry lattice. The new fuel storage racks at both BSEP-1 and BSEP-2 are General-Electric, low-density new fuel storage racks. New fuel assemblies with an initial k-infinity limit of 1.31 will be subcritical in the new fuel storage racks without concern for their initial U-235 enrichment. This limit has been approved by the NRC and is.
j documented in Section 3.3.2.1.4 of GESTAR II, NEDE-240ll-P-A-8.
4 The final change involves Specification 5.6.1.2. Currently, Specification 5.6.1.2 requires that the spent fuel storage racks be maintained such that there is sufficient.
center-to-center distance between stored fuel assemblies to ensure a keff equivalent to less than 0.95 with the storage pool filled with unborated water..The current specification also limits the maximum enrichment of fuel contained in the spent fuel pool to 3.2 weight percent U-235 for PWR fuel assemblies and 3.0 weight percent U-235 for BWR fuel assemblies and the U-235 per axial centimeter to 41 gm/cm for PWR assemblies and 15.6 gm/cm for BWR assemblies. (The proposed revision restricts the k-infinity of the fuel rather than the U-235 enrichment and axial gm/cm.
The BSEP-1 and BSEP-2 spent fuel pools contain three types of spent fuel storage racks:
high-density, poisoned General Electric designed BWR fuel storage racks; high-density,-
unpoisoned BWR fuel storage racks; and high-density, unpoisoned PWR fuel storage racks. The NRC has approved a generic limit for the k-infinity of fuels stored in the General Electric supplied fuel racks. This limit is documented in Section 3.3.2.1.4 of GESTAR II, NEDE-240ll-P-A-8. The limiting k-infinity is 1.33; associated with the General Electric high-density, poisoned rack design.
The criticality analysis for the remaining two rack designs employed at the Brunswick Plant was approved by the NRC upon issuance of Amendments 8 and_30 to the Facility Operating Licenses for Brunswick Plant, Units 1 and 2 on August 26,1977. This criticality analysis assumed spent fuel stored in the high-density, unpoisoned PWR and BWR racks had a maximum assembly average loading of 3.2 weight percent U-235 for PWR fuel assemblies and 3.0 weight percent U-235 for BWR fuel assemblies. General 1
Electric Company has performed an analysis (Enclosure 1) and determined that 3.2
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weight percent U-235 PWR fuel has a corresponding k-infinity of 1.416 i.005 and 3.0 weight percent U-235 BWR fuel has a corresponding k-infinity of 1.344
.005. The revised k-infinity limits of Specifications 5.6.1.2.a and 5.6.1.2.b are conservatively established at 1.41 and 1.33 respectively.
SIGNIFICANT llAZARDS ANALYSIS The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated,(2) create the
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possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety. Carolina Power & Light
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Company has reviewed this request and determined that:
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The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Specification 5.3.1 was revised to require fuel assemblies contained in the reactor core be designed consistent with designs already approved by the NRC as described in the fuel vendor's generic fuel design licensing topical reports. Performance of a cycle-specific safety analysis for each reload ensures the fuel's acceptability for operation.
The revisions to Specifications 5.6.1.1 and 5.6.1.2 impose limitations which ensure that fuel can be safely handled and stored. Basing these limitations on k-infinity rather than the maximum U-235 enrichment and axial gm/cm<
prevents inadvertent criticality while allowing the use of improved, higher enrichment fuel assemblies in upcoming operating cycles. The proposed changes to Specifications 3.6.1.1 and 5.6.1.2 are more restrictive than the existing TS. The revision to Specification 5.6.1.1 provides a k-infinity limit to ensure the k limits are met. The k-infinity limit established in Specification 5"d.l.2.a is 1.41; slightly less than the k-infinity determined by General Electric at the lower 2a tolerance for the PWR fuel assembly analyzed. The k-infinity limit established in Specifications 5.6.1.2.b for BWR fuel is 1.33, the more restrictive k-infinity limit for General Electric designed, high-density, poisoned BWR storage racks.
2.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated. As stated above, the proposed revisions allow the use of improved, higher j
enrichment fuel assemblies, however, the revision to Specification 5.3.1 l
requires that this fuel be of a design which is approved by the NRC. In addition, the revised Specifications 5.6.l.1 and 5.6.1.2 are more restrictive than those currently existing in the TS. This amendment does not affect the
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method in which any safety-related equipment achieves its safety function.
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The proposed amendment does not involve a significant reduction in the j
margin of safety. The proposed k-infinity limitations provide margin which is equivalent to that which currently exists for the PWR fuel and more restrictive than currently exists for the BWR fuel.
l Based on the above reasoning, Carolina Power & Light Company has determined that the proposed amendment does not involve a significant hazards consideration.
ADMINISTRATIVE INFORMATION 1
The revised BSEP TS pages are provided in Enclosures 2 and 3. The Company has j
evaluated this request in accordance with the provisions of 10 CFR 170.12 and j
determined that a license amendment application fee is required. A check for $150 is enclosed in payment of this fee. Carolina Power & Light Company requests issuance of the proposed amendment by October 5,1987 in order to support receipt of new fuel for
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the upcoming BSEP-2 Reload 7 outage.
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TDocument Control Desk :
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rNLS-87-103 / Page 4' Please refer any questions regarding this submittal to Mr. Sherwood R. Zimmerman at
.(919) 836-6242.
.You s very tr I,
"A. B. Cutter -
MAT /lah -(5197 MAT)
Enclosures cc:
Mr. Dayne H. Brown Dr. 3. Nelson Grace (NRC-Ril)
Mr. W. H. Ruland (NRC-BNP)
- Mr. E. Sylvester (NRC) l A. B. Cutter, having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his in-formation, knowledge and belief; and the sources of his information are
. officers, employees, contractors, and agents of Carolina Power & Light
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Company.
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ENCLOSURE 1 TO NLS-87-103 GENERAL ELECTRIC K-INFINITY ANALYSIS i
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GENERAL ELECTRIC CCMPANY Attachment to LMQ:87-090 April 27, 1987
Subject:
Brunswick-1/2 Fuel Storage k-infinity Conversions
SUMMARY
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There are four types of fuel storage rack designs at the
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Brunswick 1 and 2 site.
A low density new fuel storage rack design and one type of modular high density fuel storage rack l
1 design, both supplied by GE.
There are two additional spent fuel J
rack designs, one for BWR fuel using a round tube configuration and one for the H. B. Robinson PWR spent fuel.
The infinite l
neutron multiplication factor (k-infinity) has been calculated for the design basis fuel bundles used in the nuclear safety analysis of the non-GE supplied BWR and PWR spent fuel storage racks.
The calculations consisted of an infinite array of design basis fuel in the cold uncontrolled reactor. core geometry.
The resulting k-infinities, including the critical benchmark bias, are:
STORAGE RACK k-infinity (I 26)
BWR (Round Tube Type) 1.344 0.005 PWR 1.416 i 0.005 i
Assuming a Normal distribution, a lower bound value can be calculated using 95/95% statistics.
It is recommended that the 95/95% lower tolerance limit values of k = 1.34 and k = 1.41 be used as the Technical Specification limits in place of the
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existing Uranium-235 enrichment limits, for the BWR (round tube type) and PWR spent fuel storage, respectively.
ANALYTICAL METHOD The calculation was performed with the General Electric MERIT monte carlo neutron transport computer program.
The MERIT
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program is a monte carlo program for solving the linear neutron c
transport equation as a fixed source or an eigenvalue. problem in three space dimensions.
The cross sections in MERIT are processed from the ENDF/B library in the multigroup and resonance parameter formats.
Thermal scattering in water is represented by the Haywood kernel obtained from the ENDF/B library.
The MERIT program utilizes 190 full spectum cross section energy groups.
The types of reactions considered in MERIT are fission, elastic, inelastic and (n,2n) reactions.
Absorptions are implicitly treated by applying the non-absorption probability to neutron weights on each collision.
QUALIFICATION The MERIT program has been thoroughly verified for programming, sampling procedured, particle tracking, rardom number generation, fission source distribution, statistical evaluation, resonance cross section evaluation, edits and other functions of the program.
The overall performance of MERIT and the cross section data was evaluated by comparison against critical experiments which include:
1.
CSEWG thermal reactor benchmark problems:
TRX-1,TRX-2,ORNL-1,ORNL-2,PNL-1,PNL-2 2.
Babcock and Wilcox Small Lattice Facility 3.
Jersey Central Gamma Scan Experiments 4.
BWR Gadolinia Critical Experiments 5.
Battelle Critical Experiments with Fixed Neutron Poisons 6.
Nippon Atomic Industrial Group (NIAG) Critical Experiments with BWR Control Rods 2
The analyses of these benchmark calculations indicate that MERIT based on the ENDF/B-IV cross sections under-predicts k-effective by approximately 0.5% for thermal reactor criticals.
The CSEWG j
evaluation of the ENDF/B-IV data concluded that the experinent is under-predicted by 0.5% for the high moderator-to-fuel ratio in i
water moderated uranium latti'ces.
The MERIT results confirm the.
i biases supporting the CSEWG conclusions.
The MERIT results l
1 reported in this su= mary include the ok bias correction of 0.005 i 0.002 (ld).
I CALCULATIONS An infinite lattice neutron multiplication factor was calculated for the fuel bundles used in the Brunswick 1 and 2 BWR and PWR l
spent fuel storage rack nuclear safety analysis.
The fuel bundle parameters used in the BWR spent fuel storage (round tube type) analyses are contained in Table-l.
The PWR fuel bundle parameters are contained in Table-2.
The BWR fuel bundle k-infinity was calculated using the reactor core geometry consistent with Brunswick 1 and 2 without the presence of control i
rods.
The PWR fuel bundle k-infinity was calculated assuming no additional water gap between the bundles and no control or instrumentation in the water rods.
Since an unirradiated high enrichment lattice without burnable poisons is overmoderated in the reactor core geometry, the calculation was performed at 20 C, resulting in the minimum design basis k-infinity.
The calculations were perfor'ed with 55 batches of 1000 neutrons, m
starting with a uniform fission source distribution.
The first 5 batches were discarded to eliminate the bias resulting from the initial source distribution.
The final results were based on the remaining 50 batches or 50,000 neutron histories.
The resulting k-infinity values are: '
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I j-FUEL RACK.
'k-INFINITY (d.26).
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BWR;(Round' tube type) 1.344 i.005
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i P.. van.D emen LSE, Fuel'and Core Support M/C 735, (408) 925-6160 m
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f TABLELl..BWR Fuel Bundle-Paraneters
-PARAMETER
~VALUE' r
. Pellet-.O.D. (inch)
-0.416 I
Pellet Density (%T.D. )-
94.0-Fuel Rod 0.D.E(inch).
O.493-1 Fuel Rod I.D.:(inch) 0.425 Fuel' Rod. Array =
8-x 8 Fuel Rod. Pitch (inch) 0.640' Average U-235 Enrichne:rit ' (%)
3.00 Rod Enrichment Distribution uniform Burnable Poisons none' Number of Water Rods 1
Water Rod O.D. (inch) 0.493 Water Rod I.D.- (inch) 0.425 Channel Thickness (inch)
- 0.080 '
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. TABLE'2 'PWR Fuel Bundle Parameters-PARAMETER VALUE Pellet 0.D.-(inch) 0.3659 Pellet Density (%T.D.)
95.0-
' Fuel' Rod O.D. (inch) 0.422 Fuel Rod I.D. (inch)'
0.3734 Fuel Rod Array 15 x 15 Fuel Rod Pitch (inch) 0.563 Average U-235 Enrichment (%)
3.20 Rod Enrichment Distribution uniform.
Burnable Poisons none Number of Water Rods 21 Water Rod O.D..(inch) 0.546 Water Rod I.D. (inch) 0.512
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