ML20151S914
| ML20151S914 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/03/1998 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TAC-MA1113, ULNRC-3893, NUDOCS 9809090111 | |
| Download: ML20151S914 (7) | |
Text
_. _
~_
~
l An,mn corpontion One Ameren Piar,a 1901 Chouteau Avenue PO Box 66149 I
St. Louis, MO 63166-6149 314 671.3222 September 3,1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station PI-137 ULNRC-3893 Washington, D. C. 20555 TAC NO. mal 113
[L Gentlemen:
gg DOCKET NUMBER 50-483 CALLAWAY PLANT RESPONSE TO NRC RAI LETTER CALLAWAY SPENT FUEL POOL RERACK
Reference:
- 1. ULNRC-3742, dated February 24,1998
- 2. ULNRC-3837, dated May 27,1998
- 3. ULNRC-3850, dated June 25,1998
- 4. ULNRC-3887, dated August 25,1998
- 5. NRC Request for AdditionalInformation Letter, dated July 31,1998, from K. M. Thomas to G. L. Randolph Reference 1 provided the original submittal of an amendment request to
/
revise the Callaway technical specifications to support modification to increase the spent fuel storage capacity at the Callaway Plant. References 2 through 4 provided additional information. Reference 5 transmitted requests for additional information. The responses to these requests are being provided as an enclosure to this letter.
If there are questions on the Enclosure or if additional information is required, please contact us.
jMj Very truly yours, (tr&
Alan C. Passwater Manager, Corporate Nuclear Services DJW/jdg s ',?,
4 Enclosure 9809090111 980903 "
PDR ADOCK 05000483 P
1
~
?
STATE OF MISSOURI )
)
Alan C. Passwater, of lawful age, being first duly
~' sworn upon oath says that he is Manager, Corporate Nuclear l
Services for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts j
therein stated are true and correct to the best of his l
knowledge, information and belief.
j l
DN By Alan C.
Passwater Manager, Corporate Nuclear Services l-I' SUBSCR ED and sworn to before me this 04 day of
!$rn&M
, 1998.
/
ThrYdtSLY
'f?h" gf..-....."".
t T[Wp W aqgg
,e.io w ease 9,,.
yp PATRICIAL REYNOLDS NOTARY PUIUC-41AN OF MIBOUN ST.Loul8 00UfftY MY COCES010N EXPUSE2,RMS I
cc:
M. H. Fletcher Professional Nuclear Consulting, Inc.
' 19041'Raines Drive' Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-80i34 Senior Resident Inspector Callaway Resident Office L
U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman,-MO 65077 Kristine M. Thomas (2)
Office of Nuclear Reactor Regulation l
U.S. Nuclear Regulatory Commission 1 White-Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department l
Missouri Public Service Commission
~,
P.O. Box 360 Jefferson City, MO 65102 Ron Kucera.
Department of Natural Resources l
P.O. Box 176 Jefferson City, MO 65102 Denny Buschbaum i
TU Electric l
P.O. Box 1002 Glen Rose, TX 76043 Pat Nugent Pacific Gas & Electric Regulatory Services P.O. Box 56 Avila Beach,-CA 93424 l
l l
ULNRC-3893 September 3,1998 Page1of4 Callawny Plant Response to Request for AdditionalInformation (RAI) Dated 7/31/98 Question 1:
The Westinghouse Improved Technical Specifications (ITS), in accordance with NRC Generic Letter 90-02, Supplement 1, allow limited substitution of fuel rods by zirconium alloy or stainless steel filler rods, but do not allow vacancies (water). Therefore, the proposed ITS 4.2.1 should delete the words, "or by vacancies."
Response 1:
The words "or by vacancies" will be removed from ITS Section 4.2.1, Reactor Core-Fuel Assemblies. These words were included in the initial ITS submittal to maintain consistency with present CTS Section 5.3.1. After further consideration, no situations could be identified where fuel assemblies with vacancies would be loaded (or re-loaded) into the reactor core. Therefore the ITS submittal will be revised as requested by this NRC RAI to be consistent with the Westinghouse Standard Technical Specifications and NRC Generic Letter 90-02, Supplement 1.
The Improved Technical Specifications amendment request will be supplemented to reflect this change.
Question 2:
Proposed ITS 4.3.1.2 specifies that the new fuel racks containing fuel enriched to 5.0 wt% U-235 are designed and maintained with k-effless than or equal to 0.95 if fully j
flooded with unborated water, or less than or equal to 0.98 if moderated by aqueous foam. Please submit for NRC review the criticality analysis, which justifies these specifications.
Response 2:
The Callaway fresh fuel rack criticality analysis, which supports the storage of up to 5.0 w/o enriched fuel, was performed by Westinghouse in December of 1989. The analysis and 10 CFR 50.59 evaluation are referenced in ULNRC-02130, dated December 28,1989. This ULNRC transmitted an amendment request to the NRC concerning several revisions for Callaway Plant's spent fuel pool storage. Included in the submittal letter is the statement that supplemental criticality analyses were performed to verify the storage of V5 fuel with maximum initial enrichments of 5.0 w/o U235 in the new fuel storage racks. Also stated is the fact that the results of a review performed under the provisions of 10 CFR 50.59 confirmed the fuel assemblies could be safely stored in the l
new fuel storage racks without exceeding criticality safety limits. The amendment request was approved by Amendment No. 54 for Callaway Plant in May of 1990.
l
ULNRC-3893.
September 3,1998 Page 2 of 4 The fresh fuel rack analysis is based on maintaining the neutron effective multiplication.
factor below the acceptance criteria for subcriticality. The analysis considers the storage of Westinghouse 17x17 STD, OFA, and VANTAGE 5 fuel with nominal enrichments up to 5.0 w/o U235 under full water density and optimum moderation conditions and including all uncertainties. By fixing the minimum separation between assemblies in the storage racks, fuel assembly interactions are limited and inadvertent criticality is prevented. The prevention of criticality is based on maintaining a 95 percent probability at a 95 percent confidence level that the La of the fuel assembly array, including all uncertainties, satisfies acceptance criteria for subcriticality.
L Because fresh fuel racks are maintained in a dry condition, the introduction ofwater into the fresh fuel rack storage area represents worst case accident scenarios. To ensure i
subcriticality K.amust satisfy the criteria for subcriticality for the dry condition, for a condition of full water density (flooded storage area with pure water of 1.0 gm/cm'-
- density at 68 F), and for the condition oflow water density (specifically, the optimum moderation density). When fresh fuel rack reactivities are calculated as a function of water density, there is a maximum value for Laat the optimum moderation density.
j For fresh fuel storage racks, under normal dry conditions or flooded with full density unborated water, La must be less than or equal to 0.95 (as recommer.ded in ANSI 57.3-1983 and in NRC Letter to All Power Reactor Licensees, from B.K. Grimes, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications",
j dated April 14, 1978). For fresh fuel storage racks, under optimum moderation t
conditions and with fuel of the highest anticipated reactivity, La must be less than or equal to 0.98 (as referenced in NUREG-0800, Section 9.1.1).
The analysis includes conservative assumptions for fuel parameters. Fuel assemblies are j
assumed to contain the highest authorized enrichment (5.0 w/o U235 nominal and 5.05 l
w/o U235 worst case) for all fuel rods, with no credit for burnable absorbers in the fuel i
r9ds. Fuel pellets are modeled with 96 percent theoretical density, without dishing or chamfers to bound the maximum fuel assembly loading. In addition no credit is taken for U234 or U236 in the fuel, and no credit is taken for spacer grids or spacer sleeves.
For each condition analyzed, the most conservative fuel design was used to calculate reactivity. For example, under full water density cor.ditions, the Westinghouse 17x17 OFA fuel assembly yields greater La values than Westinghouse 17x17 STD fuel when both assemblies have the same U235 enrichment. The VANTAGE 5 fuel design parameters relevant to criticality are the same as OFA fuel parameters and yield equivalent results. For low water density conditions the Westinghouse 17x17 STD fuel j-assembly is more reactive than the other fuel designs. The STD fuel assembly contains a higher uranium loading than the other designs and under optimum moderation conditions, i
i higher loadings result in higher reactivity.
4 a
ULNRC-3893 September 3,1998 Page 3 of 4 In April 1998, Callaway began using VANTAGE + fuel, which is similar to VANTAGE 5 fuel with the exception of the material used for fuel rod clad, instrumentation tubes, and guide thimbles. Based on a comparison of the VANTAGE 5 and VANTAGE + designs and material properties, VANTAGE + fuel is also bounded by the fresh fuel rack criticality analyses done using OFA fuel (for the full water density condition) and STD fuel (for the optimum moderation condition).
Under the Westinghouse criticality calculation method, the AMPX system of codes was F
used to generate cross-section libraries for input into the KENO-IV code for reactivity determination. The AMPX system ofcodes generates cross-section libraries from 4
I ENDF/B-V data; includes self-shielded resonance cross-sections appropriate for each particular geometry type; and provides energy and spatial weighting of cross-sections to l
generate multigroup cross-section sets. The multigroup cross-section sets are then used
)
ss input to KENO-IV for reactivity calculations based on three dimensional Monte Carlo theory.
I The criticality calculation method and cross-section values were verified by comparison I
to benchmarking data. The benchmark data is based on assemblies similar to those for which the racks were designed, and is sufficiently diverse so that method bias and i
uncertsinty will apply to rack conditions (strong neutron absorbers, large water gaps, and low moderator densities). A set of 33 critical experiments was analyzed using the criticality calculation method. From these calculations the method bias and variability l
were established.
The following results of the analyses demonstrate that fresh fuel storage up to 5.0 w/o
)
U235 is acceptable
- a. The normal, dry storage condition fresh fuel rack reactivity satisfies the acceptance criterion of Korless than or equal to 0.95.
- b. For the case using nominal rack dimensions and fuel enrichment and a condition of
)
full water density, the criticality calculation for fresh fuel storage racks results in an effective multiplication factor below the acceptance criterion ofLaless than or equal to 0.95. For the worst case model(using minimum rack thickness and maximum fuel enrichment) and a condition of full water density, the calculation also results in an effective multiplication factor that is below the acceptance criterion. When the method bias and uncertainties are statistically combined, the computation results in an effective multiplication factorthat remains below the acceptance criterion ofLaless than or equal to 0.95.
- c. For the worst case model and the low water density (optimum moderation) condition,
]
the calculation results in an effective multiplication factor that is below the -
4 I
acceptance criterion of K.aless than or equal to 0.98. When the method bias and uncertainties are statistically combined, the resulting La remains below the acceptance criterion of Leless than or equal to 0.98.
l
.~
.1 s
ULNRC-3893 i
_ September 3,1998 Page 4 of 4 In conclusion, for normal dry storage or for worst case accident scenarios, involving the introduction ofwater into the fresh fuel storage area, the calculated Lasatisfies the acceptance criteria of less than or equal to 0.95 and less than or equal to 0.98 for the optimum moderation condition.
Question 3:
1 Credit is taken for soluble boron under accident conditions. What are the plant' requirements (i.e. technical specifications, administrative, etc.) for minimum soluble
}
boron concentration? What is the minimum value and surveillance interval?
Response 3:
. Technical Specification 3/4.9.12, Spent Fuel Assembly Storage, provides restrictions on spent fuel assemblies stored in Region 2 of the Callaway Plant spent fuel pool to ensure an inadvertent criticality will not occur. Even though these restrictions exist, Callaway Plant analyzed a mispositioned fuel assembly in the spent fuel pool. FSAR Section 9.l A.3.4, " Accident Analysis", indicates credit is taken for a 2000 ppm boron concentration in the spent fuel pool. Calculations associated with Callaway Plant's requested change (Reference 1) demonstrate that, for the most severe mispositioning event, a soluble boron concentration of 500 ppm, in addition to the Boral contained in the racks, is adequate to maintain La less than 0.95.
The minimum allowable concentration of boron in the spent fuel poolis 2000 ppm as documented in FSAR section 9.1.2.2 and 9.1.3.2.3.1. Callaway Plant Procedure CDP-ZZ-00200, " Chemistry Schedule and Water Specifications", requires a weekly verification of greater than 2400 ppm boron in the SFP with an operating limit of 2020 ppm boron.
Should a mispositioning event occur, Action a. of Technical Specification 3.9.12 requires the boron concentration of the spent fuel pool be verified to be greater than or equal to 2000 ppm at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until the non-complying fuel assemblies are moved to Region 1.