PLA-5702, Response to NRC 12/01/03 RAI Proposed Amendment 256 to License NPF-14: MCPR Safety Limits and Reference Changes
| ML040020346 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 12/22/2003 |
| From: | Shriver B Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-5702, TAC MB9902 | |
| Download: ML040020346 (8) | |
Text
Chief Nuclear Officer Berwick, PA 18603 t
- Tel. 570.542.3120 Fax 570.542.1504 4+*s
+<
blshriverlpplweb.corn * "
DEC 2 2003p p U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop OPl-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED AMENDMENT NO. 256 TO UNIT 1 LICENSE NPF-14: MCPR SAFETY LIMITS AND REFERENCE CHANGES PLA-5702 Docket No. 50-387
Reference:
- 1) PA-5638, B. L Shriver (PPL) to USNRC, "ProposedAmendment No. 256 to Unit I License NPF-14: MCPR Safety Limits and Reference Changes,"
dated July 01, 2003.
- 2) PA-5690, B. L Shriver (PPL)to USNRC, "ProposedAmendment No. 256 to Unit I License NPF-14:MCPR Safety Limits and Reference Changes Revised Core Composition," dated November 17, 2003.
- 3) USNRC to B. L Shriver, "RequestforAdditionalInformation (RAI) Regarding SSES I Minimum CriticalPower Ratio (MCPR) Safety Limits and Reference Changes (TAC No. MB9902), " dated December 1, 2003.
The purpose of this letter is to provide the PPL Susquehanna LLC (PPL) response to the NRC's December 1, 2003 request for additional information (Reference 3).
On July 1, 2003 PPL proposed a revision to the Susquehanna Steam Electric Station Unit 1 Technical Specifications (Reference 1). On November 17, 2003 (Reference 2)
PPL identified a change to the core composition table, provided in Reference 1, which is necessary to address design changes related to control cell friction mitigation. These proposed revisions to the Technical Specifications, if approved, would update the MCPR Safety Limit in Technical Specification Section 2.1.1.2, revise Section 4.2.1 to reflect the use of depleted uranium in reload fuel bundles, and revise Section 5.6.5.b to include NRC approved methodology which forms the basis for particular uncertainties used in the MCPR Safety Limit Analysis.
The need for additional information was identified during teleconferences held November 24 and 26, 2003. Attachment 1 to this letter provides the additional information.
4VA
Document Control Desk PLA-5702 Any questions regarding this request should be directed to Mr. Duane L. Filchner at (610) 774-7819.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on: zz)o3 B. L. Shriver Attachments:
- 1. Response to NRC Request for Additional Information cc: NRC Region I Mr. S. L. Hansell, NRC Sr. Resident Inspector Mr. R. V. Guzman, NRC Project Manager Mr. R. Janati, DEP/BRP
'M Attachment 1 to PLA-5702 Response to NRC Request for Additional Information
Attachment 1 to PLA-5702 Page 1 of 5 Response to NRC Request for Additional Information NRC Ouestion #1 Use of Depleted Uranium:
a) Define "depleted uranium," and explain how the use of depleted uranium differs from slightly or low enriched fuel.
b) The amendment request states that the use of depleted uranium is modeled in the NRC-approved design and licensing methodology. State the NRC-approved design and licensing method being used.
PPL Response Depleted uranium is defined as uranium with an initial U235 weight percent that is less than naturally occurring uranium. Depleted uranium is commonly referred to as "tails" or "residual" and is a byproduct of the uranium enrichment process. In contrast, slightly enriched uranium refers to the enriched uranium product resulting from the enrichment process. Depleted, natural, and enriched uranium fuel pellets are in the form of uranium dioxide, U0 2 . The use of depleted uranium does not differ from the use of natural or slightly enriched uranium fuel.
Figure 1 provides a flowchart for the various NRC approved methodologies used in the MCPR Safety Limit analysis. PPL performs cycle specific reload design and licensing analyses using the NRC approved methodology specified in PL-NF-90-001-A, including Supplements 1-A, 2-A, and 3-A. This methodology includes the use of CASMO3G/
SIMULATE-E for design and licensing applications. Depleted uranium, natural uranium, and enriched uranium fuel pellets are explicitly modeled in CASMO3G to determine fuel cross sections and local power distributions. Fuel cross sections are used by SIMULATE-E to determine core power distributions. These power distributions are input to the MCPR Safety Limit analysis performed by Framatome using NRC approved methodology, ANF-524(P)(A). Any revision to ANF-524(P)(A) or its supplements are consistent with Figure 1.
NRC Ouestion #2 There are restrictions delineated in the safety evaluation approving EMF-1997 (P),
"ANFB-10 Critical Power Correlation," and its supplements. For SSES 1, Cycle 14, provide confirmation that the fuel would not be operated outside the operating range that
Attachment 1 to PLA-5702 Page 2 of 5 the correlation is validated. Also, provide confirmation that the fuel performance analyses would meet the ANFB-10 critical power correlation restrictions specified in the staff's July 17, 1998, safety evaluation, transmitted to H. D. Curet from T. H. Essig.
PPL Response The NRC safety evaluation for EMF-1997(P)(A) and EMF-1997(P)(A), Supplement 1, specifies conditions on the range of applicability for the ANFB-10 correlation. The Unit 1 Cycle 14 MCPR Safety Limit analyses (two loop and single loop operation) utilize the ANFB-10 critical power correlation. The MCPR Safety Limit analyses implement the NRC conditions on the ANFB-10 range of applicability, including the additional uncertainty for local peaking greater than 1.5. Likewise, reload analyses performed by Framatome-ANP and PPL Susquehanna implement the NRC conditions on the ANFB-10 correlation range of applicability. Unit 1 Cycle 14 reload analyses are performed within the range that the correlation is validated. In addition, the fuel will be operated within the ANFB-10 correlation range of applicability during Unit 1 Cycle 14 operation.
NRC Ouestion #3 Power Distribution Uncertainties:
The SSES 1, Cycle 14 amendment request states the power distribution uncertainties based on the CASMO-4/MICROBURN-B2 code system are smaller than the corresponding uncertainties based on CASMO-3/MICROBURN-B code system.
a) Please, explain the difference in the power uncertainty distribution between the two versions of the code. Refer to the sections and the applicable tables of the CASMO-4/
MICROBURN-B2 submittal that specifies the reduced power distribution uncertainties.
b) Table 9.9 of EMF-2158(P) provides the power distribution uncertainty for the MICROBURN-B2. Explain if the acceptance criteria cross-tabulated against the MICROBURN-B2 is referring to the power distribution uncertainty associated with the earlier MICROBURN-B code system.
PPL Response Figure 1 provides a flowchart for the various NRC approved methodologies used in the MCPR Safety Limit analysis. Radial bundle power and local pin power uncertainties used in the MCPR Safety Limit analysis are based on NRC approved core simulation methods developed by Framatome-ANP. For the Unit 1 Cycle 14 MCPR Safety Limits,
i l Attachment 1to PLA-5702 Page 3 of 5 CASMO4/MICROBURN-B2 based uncertainties are used to be consistent with the implementation of the POWERPLEX-3 core monitoring system. Table 2.3 of EMF-2158(P)(A) lists the CASMO4/MICROBURN-B2 radial bundle power and local pin power uncertainties and also lists the associated CASMO3/MICROBURN-B uncertainties for C-lattice reactors. Based on the comparison provided in the table, the CASMO4/MICROBURN-B2 uncertainties are smaller than the CASMO3/
MICROBURN-B uncertainties.
Table 9.9 of EMF-2158(P)(A) also lists the CASMO4/MICROBURN-B2 radial bundle power uncertainty and the local pin power uncertainty. The acceptance criteria appearing in Table 9.9 lists the CASMO3/MICROBURN-B uncertainties (which are also shown in Table 2.3 of the topical report).
NRC Ouestion #4 For SSES 1, Cycle 14, the two-loop MCPR safety limit changes from 1.12 in Cycle 13 to 1.08. For single-loop operation that SLMCPR changes from 1.13 in Cycle 13 to 1.10.
These are significant reductions in the MCPR safety limit. The current amendment request attributes the lower MCPR safety limit to, (1) the reduction in the power distribution uncertainty and (2) the elimination of the factor of 2 factor applied to the ANFB-10 correlation. Please, discuss qualitative and quantitative effect of each contributor to the lower SLMCPR values.
PPL Response The reduction in the Unit 1 Cycle 14 MCPR Safety Limit is attributed to the elimination of the factor of 2 on the number of pins in boiling transition for the ANFB-10 correlation, the reduction in the radial bundle power and local pin power uncertainties, and cycle to cycle variability. Based on previous safety limit evaluations performed by Framatome for Susquehanna, the estimated contribution of each factor has been tabulated below.
A.
Attachment 1 to PLA-5702 Page 4 of 5 X____^___;___-___________ 2-Loop Single Loop.
Existing MCPRSL (TS 2.1.1.2) 1.12 1.13 Estimated contribution for -0.01 to -0.02 -0.01 to -0.02 removal of factor of 2 on the number of pins in boiling transition for the ANFB-10 correlation Estimated contribution for -0.02 to -0.03 -0.02 to -0.03 reduction in the radial bundle power and local pin power uncertainties Estimated contribution for +0.01 to -0.01 +0.01 to -0.01 Cycle-to-Cycle variation Proposed MCPRSL (PLA-5638) 1.08 1.10 NRC Ouestion #5 The November 17, 2003, supplement to the amendment request stated that the design changes implemented in order to address control cell friction issues resulted in a reduction in the unit's full power capability. Therefore, the fresh reload batch fraction is going to increase from 276 to 280 assemblies and the twice-burned assemblies would decrease from 172 to 168 assemblies. Discuss the bases for your assessment that replacing the three fresh bundles with three twice-burned fuel did not have any impact on the SLMCPR.
PPL Response The November 17, 2003 supplement to PLA-5638 provided the revised core composition for SSES Unit 1 Cycle 14. The changes made to the core composition involved replacing four twice burned assemblies with four fresh assemblies to maintain full power energy requirements after addressing control cell friction issues. Framatome-ANP performed a MCPR Safety Limit analysis explicitly for the revised core composition. The resulting single loop and two loop MCPR Safety Limit values remain unchanged from the values reported in PLA-5638 and assure that at least 99.9% of the fuel rods are expected to avoid boiling transition during normal operation or anticipated operational occurrences.
i ;
Attachment 1 to PLA-5702 Page 5 of 5 Figure 1 MCPR Safety Limit Methodology 1 ~~CRAMATMENPNP SLMCPR l To Be Supported I ~Plant, Fuel &I CPR Uncertanties I ~CorrelationUcetntsI l ANF-524 (PXA), Rev. 2, Supplement 1, Rev. 2
&Supplement 2 EMF-1997 (P)A, Rev. O &Supplement 1, Rev. 0 EMF-2158 (PXA) Rev. 0 Statistical Analysis I of Rods In Steady State Boiling Transition Core T/
(XCOBRA) Bundle Flow 2
vs Bundle A ANF24 (PXA), Rev. Power
- l I
a C PPL ~
I I Conservative I Power Profiles (SIMULATE-E /
CASMO-3G)
Radials, Axials, I ' 1 ANF-524 (PXA), Rev. 2, Supplement 1, Rev. 2, &
Supplement 2 PL-NF-90001-A & Locals Supplements 1-A, 2-A, 3-A I
I ~ I~~~~~~~~~~~~~~
01%~? l No I ~~~~~~~ncrease I ~~~SL.MCPR II Input SLMCPR Supported -Is I_ _ _ _ _ _ _ _ _ _. _