ML20079L979
| ML20079L979 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 10/30/1991 |
| From: | Schuster T COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-88-16, NUDOCS 9111070216 | |
| Download: ML20079L979 (9) | |
Text
{{#Wiki_filter:[O N-Ccmmonw:alth Edison I f 1400 Opus Place Downers Grove. Ulaois 60515 October 30, 1991 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commisston Washington, D.L. 20555 Attn: USNRC Document Control Desk
SUBJECT:
Byron Station Unit 1 Cycle 5_ Reload NRC_DocLel_Ho.,_50-454
REFERENCES:
See Attachment 3
Dear Dr. Murley:
Byron Unit 1 is completing a refueling outage that began September 6, 1991 following its fourth cycle of operation. Byron Unit 1 Cycle 4 attained a final cycle burnup of approximately 19,312 MHD/HTU. Cycle 5 is expected to commence in early November, 1991. This letter is to summarize Commonwealth Edison Company's (CECO) plans and evaluations regarding the Byron _ Unit 1' Cycle 5 reload core, and to provide the Cycle 5 Core Operating Limits Report per Generic letter 88-16 (Attachment 2). Attachment I describes the Byron 1 Cycle 5 reload and CECO's reload safety evaluation review, which is being performed in'accordance with the provisions of 10CFR50,59 as there are no unreviewed safety issues or necessar., Technical Specification changes. The Byron Unit 1 Cycle 5 core has been designed and evaluated using-NRC-approved methods. Commonwealth Edison performed the neutronic portion of i e BYlC5 reload design _ utilizing-NRC approved code package as described in W arence 2. The remainder of-the reload safety evaluation was performed by na Linghouse,in accordance with the methodology described in Reference 1, In summary, the Byron Unit 1 Cycle 5 reload design, including the development of the Core Operating Limits Report (COLR) pursuant-to the requirements of Technical Specific 6 tion Section 6.9.1.9, was generated and vertfled-by Commonwealth Edison using NRC approved methodology. I /Of I 9111070216 911030 PDR ADOCK 05000454 [I I .P-PDR ( { ZNLD/1304/2
October 30, 1991 Please direct any questions regarding this notification to this office. Very truly yours, 9 Gbwo.b +f ' A cv T. K. Schuster Nuclear Licensing Administrator cc: A. H. Hsia - Project Manager, I,dR A. B. Davis - Regional Administrator, RIII H. J. Kropp - NRC Resident Inspector, Byron .ZHLD/1304/3
AIIAC!iMENI_1 Byron _LCycle_LRehadJe.strlption The Byron Unit 1 Cycle 5 core is a standard " Low Leakage" design and is similar to the Cycle 4 core loading pattern. During the Cycle 4/5 refueling, eighty-eight (88) VANTAGE 5 fuel assemblies will be loaded into the core. The Byron Unit 1 Cycle 5 core will be composed of-176 Westinghouse 17x17 VANTAGE 5 assemblies (88 new and 88 once-burned) and 17 twice-burned 17x17 Optimized fuel Assemblies (0FAs). The NRC has approved the use of VANTAGE 5 at Byron Unit I for Cycle 4 and thereafter, under the provisions of 10CR50.90 in Reference 7. The Byron UFSAR (Rev. 2, December 1990) presently reflects the transition to VANTAGE 5 fuel. I The Byron Unit 1 Cycle 5 reload core was designed to perform under current nominal design parameters, Technical Specifications and related bases, and current Technical Specification setpoints such that: 1. Core operating characteristics will be equivalent or less limiting than those previously reviewed and accepted; or 2. For those postulated incidents analyzed and reported in the Updated Byron Final Safety Analysis Report (UF3AR) which could potentially be affected by fuel reload, reanalyses or reevaluations have been performed to demonstrate that the results of tiie postulated events are within allowable limite The reload VANTAGE 5 fuel assemblies have incorporated Westinghouse standardized fuel pellets, Reconstitutable Top Nozzles (RTN), extended burnup design features, snag resistant Intermediate Flow Mixers (IFM) grids and modified Debris Filter Bottom Nozzle (DFBN). Similar features have been successfully utilized in Byron Unit 1 Cycle 4 and other Byron /Draidwood units and are described in the UFSAR. Structural evaluations of these fuel features provided in the UFSAR have verified that the VANTAGE 5 assembly design is-structurally acceptable. Reference 6 and the Byron UFSAR fully justify the compatibility of Westinghouse OFA and VANTAGE b assemb'tes in a reload core, and verify compatibility with control rods, and reactor internals interfaces. ZNLD/1304/4-
AlitalHMU (cont'd) The reload fuel's nuclesr design has been evaluated generically in the UfSAR and in Reference 6. As OfA and VANIAGE 5 fuel have the same pellet aad fuel rod diameters, nost reactivity parameters are insensitive to fuel type. Changes in nuclear characteristics due to the transition from OfA to VANTAGE 5 fuel are within the sange normally seen from cycle to cycle due to fuel management effects and have been previously evaluated in Cycle 4 in Byron i Unit l's first transition cycle to VANTAGE 5 fuel. The loading pattern dependent parameters have been evaluated in 4 t 11 in the CECO /Westingnouse reload safety evaluation process. Commonwealth Edison has determined that all neutronic reload parameters remain within the previously established and recently revised reload safety and transient Lafety Parameter Interaction List (SPil) limits. These inc N h, but are not limited to, Spil items for non-LOCA and LOCA considerai M s, and have considered the resolution of Westinghouse issue PI-91-OF hich add esses the Boron Dilution re-analysis for Byron /Braidwood for CVCS malfunctions in Modes 3, 4 & 5 The thermal-hydraulic design f v ti.:.ycle 3 reload core has not slyntficantly changed from that of the previously reviewed and accepted Cycle 4 design. Tests and analysis have confirmed that the VANTAGE 5 assemblies are hydraulically compatible with the OFA assemblics reloadeI as Regicn 5. Cycle 5 has a majority (176 out of 193) of VANTAdt 5 fuel assemblies. The present Technical Specification FNDH limits of less than 1.55 for OFA assemblies and 1.65 for VANTML 5 assemblies ensure that the limiting ONB ratio during Normai Operation and Operational Transients (Condition 1 and Condition 11 events) is greater than or equal to the DNBR limit of the DNBR correlation being applied. Commonwealth Edison's reload safety evaluation process (RSE/SPIL review) is a verification to ensure that the previously reviewed and approved FSAR transient analyses are not adversely impacted by the cycle specific raload core design. Ceco's Byron Unit 1 Cycle 5 Reload Safety Evaluation (RSE) relied on previously reviewed and accepted analyses reported in the UFSAR, fuel technology reports, the VANTAGE 5 Reload Transition Safety Report, and previous RSE reports. A detailed review of the core characteristics was performed to determine those paransters affecting the postulated accident anelyses reported in the Byran UFSAR. Commonwealth Edison veriftad that for the accident analysel presented in the UfSAR, the conclusions were not affe W d by the reload core characteristics. ZNLD/1304/5
t AUActitKtiL1 (cont'd) l Westinghouse has concluded that the results of the Dyron Unit 1 Cycle 5 reload safety eva19attunt s e valid since there are no reloao related changes to the-current ie:hnical ';r~cifications required +o ensure safe operation during Cycle 5. farthermore, Westinghouse has concluded that the core design parameters and astumptions remain appropriate and the conclusions in UrTAR remain applicable. Finally, the Byron Unit 1 Cycle 5 reload core design will be verified per the standard reload startup physics tests. These tests include, but are not limited to the following: 1. A physical inventory of the fuel in the reactor by serial number and location prior to the replacement of the reactor head: 2. Control rod-drive tests and drop times; 3. Critical boron concentration measurements: 4. Control bank worth m9asurements using the rod swap tcchnique; 5. Moderator Temperature Coefficient (HTC) measurements; and 6. Startup power di'.a ibution measurements using the incore flux mapping system. In summary, CECO b;e of VANTAGE 5 fuel and use of advanced neutronics methods (as described in References 2 and 6, respectively) have been previously apprcyed by the NRC (References 3 and 7, respectively). Therefore, pending completion of the On-Site and Off-Site Reviews no additional prior NRO review and approval of the reload core analyses or -application for amendment to the Byron Unit 1 operating license, is required as a result of the cycle specific reload design for Cycle 5. 2 ZNLD/1304/6- ., _ ~ - -
ATTACHMENT 2 c Byron Unit 1 Cycle 5 Operating Limit Report - fxy Portion This Radial Pecking factor Limit Report is provided in accordance with Paragraph 6.9.1.9 of the Byron Unit 1 Nuclear Plant Technical Specifications. The fxy limits for RATED THERHAL P0HER within specified core planes for Cycle 5 shall be: a. For the lower core region from greattr than or equal to 0% to less than or equal to 501: 1. For all core planes containing bant "D" control roos: ! 1 2.16, for Bu 1 8000 MHD/HTU f 1 1.944, for Du > 8000 MHD/HiU 2. For all unrodded core planes: F'x 1 1.7701, for Bu 1 8000 MHD/HTU 1 1.6902. for Bu > 8000 MHD/H1U b. For the upper core region from greater than 50% to lens than or equal to 100%: 1. For all planes containing bank "0" control rods: RTP f 1 2.16, for Bu 1 8000 MHD/HTU xy 1 1.944, for Du > 8000 MHD/HTU 2. For all unrodded core planes: RTP f,y 1 1.8036, for du ( 8000 MHD/HTU 1 1.8048, fo;- Bu > 8000 MHD/H10 These fxy(z) limits were used to confirm that the heat flux hot channel factor FQ(z) will be limited to the Technical Specification values of: fg(z)1[L.50] [K(z)] for P > 0.5 and, P-fg(z)1[5.00) [K(z)) for P 1 0.5 assuming.the most limiting avi.' power distributions expected to result from N 1Hsertion and removal r'f Control Banks C and 0 during operation, including tne accompanying variations-in the axial xenon and power distributions as described in the " Power Distribution Control and Load following Procedures", .Therefore, these f limits provide assurance HCAP-8403. September, 1974-thattheinitialconditionsassumedintheLOCA$nalysisaremet,alongwith x the ECCS arteptance criteria of 10 CFR 50.46. See figures l'and 2 for plots of [F.Ppgj] vs. Axial Core Height. ZNL9/1304/8 ....-.-- - --. --.-.a.
~ FIC 1 BYRON 1 CYCLE 5 ~ ~ ~ Fxy UMIT ANALYSIS - BU ( 8000 2.60. h (" "h 2.50 xx xm x xxx xxx N 3 0 43i} 2.10 f.
- x P
N 'Na xx x x x X X d r x xx x x x 2.30 :_ x x V 210 ~ ^ x xx 2.10 : x r 2.00 : b 1.90 ! 2 v c 1.80 : Y e 1.70 : c-1 ~ C.60 f 'f!f ~ 1.50 : 1.40? a Tep Sund=c thit 1.30 ~~ I o Bcticm 9Jad:celimii x i 1.20 : j x CADCkdris Pobis l 1 l 1.10 _:. ~ i
- LOCA thithg Eny: lope x
1.00 ! O.90
- c
^ ^! !~ ^ B ^ I O 1 2 3 4 5 6 7 8 9 10 11 12 BOITOW CCREHBGliT(fI) l = k-_..___
s ~ L.~ l1 M x x es r x .-a n. m __m 3 K u rg .!a 4 e5 .G o T.1 a. x =u 1 M- -{8] . L" x 8 x 3 g M Q n. = 3 53 a a a + x __s x t: x a> g 7 x 9 uio x m a ao O tm x goA M cd x ~ c, - ao x w z l c o x g en m @W x x E x t-- x
- E x
U x n R u..' x g .x a EE a5 x x >C l x ,,. ::: n* = m-o ae S8 8?8?89R88?8 l ogg a ya s
a 9.0d 3AllVB8 X(603 L
1 l t it t - ~.. _.. -, _.
r AUACllMENL3 References 1. Westinghouse HCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology", dated October 1985. 2. Ceco /NFS topical report NFSR-0081, " Benchmark of PHR Nuclear Design Methods Using the PHOENIX-P and ANC Codes," (dated July 1990) submitted for NRC review by letter from J. Silady to T. E. Murley dated July 13, 1990. 3. NRC SER on Ceco's PHOENIX /ANC Topical (Ref. 2) dated March 11, 1991. 4. Westinghouse. topical report HCAP-11596-P-A, " Qualification of the PHOENIX /ANC Nuclear Design System for Pressurized Water Reactor Cores," dated June 1988. 5. Westirghouse topical report HCAP 10965-P-A, 'ANC: Westinghouse Advanced Nodal Computer Code," dated September 1986. 6. CECO subtalttal, R. A. ChrzanowsL1 to 7.E. Murley, " Byron Station Units 1 and 2 App 1tcation to facility Operating License NPF-37 -and NPF-66," dated July 31, 1989. 7. NRC Letter from L. N. 01shan to T. E. Kovach, Amendment No. 36 Use of VANTAGE 5 fuel," dated January 31, 1990. l l-l ZNLD/1304/7}}