ML20073R452
| ML20073R452 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/31/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20073R451 | List: |
| References | |
| NUDOCS 9106070204 | |
| Download: ML20073R452 (7) | |
Text
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f 'c UNITED STATES
$ *D.. j, +! i NUCLEAR REGULATORY COMMISSION 5
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WASHINGTON D C. 20%5 j
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SAFETY EVALOATION EY THE OFFICE OF HUCLEAR REACTOR REGULATION RELATED TO AMENDMEt?T NO. 86 TO FACILITY OPERATING L1CFt!SE tiPF-35 Atl0 At<ENDMEtlT 110. PO TO FACILITY OPERATit!G LICENSE t!PF-52 DUKE POWER COMPANY, FT AL.
CATAWBA NUCLEAR STAT 10t', UNITS 1 AND ?
DOCKET !!05. 50-413 AND 50-414
1.0 INTRODUCTION
Ey letter dated January 9,1991 (Ref.1), Duke Power Company, et al. (Duke or the licensee) requested amendments to the Technical Specifications (TSs) appended to Facility Operating License Nos. NPF-35 and NPF-52 for the Catawba Nuclear Station, Units 1 and 2 (CNS1 and CNS2). The TS changes are primarily for operation of CNSI during Cycle 6 with reload fuel and safety and operational analysis provided by the Babcock and Wilcox Fuel Company (BWFC). There are also several administrative changes to the CNS1 TSs, and these are also reflected in administrative changes to the CNS2 TSs.
Included with Reference I were (1) the proposed TS changes along with related Bases changes and a sample Core Operating Limits Report (COLR), (2) a discussion of the technical justification for the changes, and (3) a CNSI Cycle 6 reload report prepared by BWFC (Ref. 2).
The reload for CtlS1 Cycle 6 is one of the first applications of BWFC providing the reload fuel and the operational and safety analysis for a Westinghcuse designed reactor. The changes to the CNSI TSs are largely a result of changing the BWFC operational methodology and analysis. The characteristics of the Cycle 6 core are not significantly different from those which would have resulted from a reload with Westinghouse standard fuel. BWFC operational methodology, while generally similar to Westinghouse methodology, differs in and E many details. This requires extensive changes to the TSs for the Fq delta H power distribution peaking factors as well as sone changes to other TSs related to power distribution and departure from nucleate boiling (DNB).
BWFC operational methodologies (and corresponding TSs) for Westinghouse reactors are similar to BWFC methcds for B&W reactors, which have been reviewed and approved by the NRC. They differ primarily in surveillance related areas because Westinghouse reactors do not have the fixed incore neutron detectors of B&W reactors to provide continuous detailed power distribution information.
BWFC reload design and safety analysis methods for Mestinghouse reactors are also similar to those used and approved for B&W reactors.
In addition to the new fuel for Cycle 6, BWFC has provided a reload analysis and a corresponding Reload Report (Ref. 2) describing the fuel, nuclear and thernal hydraulic design and characteristics, the transient and accident analysis justifying the operat' ions of CilS1 for Cycl 3 6, and TS changes 9106070204 910531 PDR ADDCK 05000413 P
, necessary to accompany the charges to the operational methodology and Cycle 6 specific analyses. The changes to the operating methodology and TSs include changes to the operating limit parameters assigned to the COLR, and a sample COLR indicating the nature of those changes is provided in the Reload Report.
The nethodologies used by BWFC for reload design, safety and operational analyses of Westingbouse designed reactors have been documented in topical reports and reviewed and approved by the NRC. This includes both generic reports and reports specific to the Duke Catawba and McGuire reactors..These reports are listed as references in the Reload Report (Ref. 2).
Several of the primary generic reports are listed at References 3 through 7 of this revier: and the prinary Duke specific reper ts in References 8, 9, and 10.
2.0 EVALUATION The Cycle 6 reload for CtiS1 will be the first.:si of the BWFC flark-BW fuel assembly in the Duke reactors. This fuel is similar in most respects to the Westinghouse standard fuel assembly design. Tnere wih be 72 of these assemblies with an enrichment of 3.55 percent, along with 121 remaining Westinghouse optimized fuel design (0FA) assemblies from the previous cycle.
T$ _ mixed (transition) core aspects of the loading are similar to previously reviewed mixtures of Westinghouse standard and 0FA fuel assemblies, and they have been considered in the review of the Duke specific topical report (e.g.,
-Reference 9).
Limits of the enthalpy rise peaking factor for the two fuel types (0FF limits are 96 percent of the Mark-BW limits) have been provided and no additional penalty is needed for CNSI Cycle 6 (Refs, 9 and IP).
The Mark-BW fuel design has been reviewed and approved by the NRC (e.g.,
References 3 and 8). The open items in the conclusions of Reference 8 were answered in Reference 11. The answers are acceptable and the NRC review is complete and the fuel design approved, as indicated in the staff review in Reference 11.
The nuclear design and parameters for Cycle 6 have been calculated with standard BWFC methodology, based on the NRC approved N0ODLE code (Ref. 4).
The core design includes a low-leakage fuel pattern, and the cycle length (350 effective full power days) is slightly longer than CNSI previous cycles. For this design the physics parameters calculated for the cycle, including control rod worths, reactivity coefficients and shutdown margins, fall within expected ranges and are acceptable.
The thermal-hydraulic design analysis used the generically applicable Statistical Core Desian and the BWCMV Critical Heat Flux correlation discussed (and approved by the ARC) in References 5 and 6.
The Cycle 6 safety limits are based on enthalpy rise peaking factors of 1.55 for the tiark-BW fuel and 1.t,9-for 0FA fuel. The statistical design limit was determined to be 1.345 for the CNS1 core and the thermal design limit was 1.50, providing a thermal margin of 10.3 percent. The analyses used approved models and reasonabic input values, and the thermal-hydraulic design evaluation for Cycle 6 is acceptable.
l
l 3
The Duke specific reload transient and accident safety analyses and evaluations were presented in References 9 and 10, with Reference 9 examining all events except loss of coolant accident (LOCA), which is considered in Reference 10.
The scope of the events considered is consistent with that addressed in the final Safety Analysis Report (FSAR) for Catawba.
The evaluations considered the effects of mixed (transition) cores using Westinghouse and Park-BW fuel.
4 The reports, evaluations and analyses have been reviewed and approved by the
!!EC staff.
The cenclusions of the review of Reference 9 listed several conditions (o be considerert in using the results of the analyses for cycle specific cases.
These conditions were responded to in Reference 12 and with recent information en a benchmark calculation of the sternline breat without offsite power. A review of these responses (along with the information in the Reload Peport) has concluded the indicated conditions have been satisfied, and the analyses of Reference 9 are acceptable for CNSI Cycle f.
The conclusions of Reference 10 accepting the LOCA analyses also contain 4
several conditions.
However, most of these are indicated as being satisfied for Catawba, and the others were satisfied by the information provided in the Reload Report.
Thus, the analysis provided in Reference 10 is acceptable for i
CNSI Cycle f.
The Reload Peport presents values of key parameters used in the analyses of i
References 9 and 10 and compares them to values determined for CNSI Cycle 6.
All ytlues are within limits, and the analyses are therefore acceptable for Cycle 6.
Cycle specific statepoint analyses and dose calculations were performed for the events classified as accidents to confirm that the analyses were applicable and to provide cycle specific dose levels. The dose values met all required criteria and are acceptable.
The new core power distribution related core operating limit methodology developed by BWFC for Westinghouse designed reactors using BWFC fuel, reload analysis and cycle operation centrol, and adopted by Duke for CNS1 for Cycle 6, is the primary cause for most of the significant changes to the CNS1 TSs for this cycle. The methodology was presented in the BWFC topical report BAW-10163 (Ref. 7) and has been previously reviewed and approved by the NRC.
The methodology was developed for reactors such as Catawba with BWFC fuel and analysis and is acceptable for use in CNSI Cycle 6.
Reference 7 discusses the methodology, the relevant terminology (changed in a number of respects from previous Catawba-Westinghouse terminology) and the resulting changes to the power distribution TSs. The report provides example TSs for the revised me'chodology. The primary TS changes resulting from the methodology are to TSs 3/4.2.2 and 3/4.2.3 on F and F limits and surveillance. There are q
delta H less extinsive changes to other related TSs. The CNSI use of the methodology.
and the changes to the TSs generally follows the description in Reference 7, but there are some changes in terminology for some parameters, some rearrangement of the TS text to what Duke considers tn be a more logical pattern, and a few variations in action and surveillance times, usually more conservative or to maintain previously approved times from the Westinghouse 1Ss. The surveillance methodology for F and F departs from the q
delta H Reference 7 description by bypassing the first tier surveillance (comparison
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. t..
3 of measured to predicted design peating) and proceeding directly with the second I
tier peating margin determination. The two tier method would involve performance of the first tier calculation and if its result was not within expected deviations i
then the second tier calculation of pealing margin would be performed.
Bypassing of the first tier calculation is a conservative change since it eliminates a step involving uncertainties in these (urveillances.
This is an acceptable deviation, as are the other c'eviations indicated above.
The methodology adopted for CHS1 Cycle C for power distribution operational limits a.id surveillance is based directly on an NRC epproved BWFC methodology, applicable to the CNSI Cycle 6 reactor, ard the proposed TSs flow directly from on approved example se*. of TSs. Deviations have been explained, and the current review has conciuded that the methodology change and the expression of Oat change in the revised TSs is acceptable.
P.1 TfCHNICAL SPECIFICAT10ff CHANGES In addition to the changes in the TSs resulting from the operational methodology, there are changes retuiting f rom necessary changes to the Core Operating Limits Report (COLR) as well as unrelated minor technical end administrative chances.
CNS1 has an approved COLR but the change in methodology requires modifications to the TS transferring parameters to the COLD as well as ertensive changes to the COLR itself to describe and list the relevant parameters and associated data.
Additions to the list of approved topical reports providing the bases of the methodology are also required.
Because CHS1 and CNS? currently have common TSs and only CNSI is receiving modified TSs at this time, there are administrative TS changes noting that the CNS2 TSs are onchanged, requiring in some cases separate pages for a given TS for the two reactors.
The following changes are proposed by Duke for the CNS1 and CUSP TSs. Duke has presented, in Reference 1, an extensive, detailed discussion of, and basis for the changes.
The review has concluded that this discussion is complete, correct and acceptable. The changes related to the revised operational methodology will not be discussed in detail here since they have been, for the most part, approved (as an example) in the review of Reference 7.
(1)
Index, pages 111, IV, V, Vlli, X111; administrative changes adding titles and page numbers for new material.
This is acceptahle.
(2) TS ?.1 Safety Limits; a new figure 2.1-1 is added for CHS1 for the reactor core safety limits because of the new BWFC saf ety analysis methodology. The current figure is retained for CNS2. This is acceptable.
(3) Table 2.2 1 Trip Setpoints; there are small changes to the Loop Minimum flow, Neutron Flux High and Coolant flow Low Setpoint (both conservative) and various TA, Z and Sensor Error values.
These are based on values used in the safety analyses and a reexamination of plant-specific uncertainty values.
They apply in general to both units.
Values of constants in the Unit 1 Op delta T and OT delta T trip functions were i
i l
..., -. _ _ _ _ - _ _ _ _. _ -. -., _, - _ -, _ _,, - -,, _ - ~.
. revised so that f2 (delta 1) could be set to Zero (as indicated and approved in Reference 71.
(Unit 2 is already set to zero.) The changes are reasonable and acceptable.
Reference to RTD bypass was removed to reflect the removal of that system.
Removal of the system was evaluated in amendment 40 to license Npf-35 and aa ndment 33 to license NPF-52, issued on February 17, 1988.
This is an acceptable administrative change.
(4) TS 3/4.2.1, Axial flux Difference; references to Westinghouse featurrs such as RA00 and baseload operation are removed.
This is acceptable.
These changes are applicable to Catawba Unit 1 only.
(5) TS 3/4.2.2, Heat Flux Hot Channel Factor: and (6) TS 3/4.2.3, Nuclear Enthalpy Rise Hot Channel factor; the Westinghouse methodology is removed from both of these TSs and the BWFC methodology inserted.
These changes have been discussed previously and are acceptable, the changes apply only to Unit 1.
(7) TS 3/4.2.4, Quadrant Power Tilt; the statement in the LCO about applicability is moved to the Applicability section.
The power penalty for excess tilt is changed to reflect a penalty on thermal power only above a tilt ratio of 1.02.
This is acceptable because a 1.02 tilt penalty has been included in the design safety analyses.
A footnote is added to the Applicability statement indicating it applies after detector calibration after refueling, and a statement that TS 3.0.4 is not applicable is also added. These changes, except for the tilt penalty, are administrative.
These changes to TS 3/4.2.4 are acceptabic. The chances apply to Unit 1 only.
(8) TS 3/4.2.5 DND Parameters; the reactor coolant flow rate limit is moved from TS 3.2.3 to this TS, and is incorporated in a new Figure 3.2-1, where it is combined with the power level to provide permitted, restricted or prohibited operating regions. This figure defines trade-offs in power and flow and has been verified by a number of thermal evaluations.
It provides comparable margins to those provided in the previous Westinghouse design TS 3.2.3, which is now revised.
It is acceptable. The changes apply to Unit i only.
(9) The Bases for the TSs which have been changed for Unit I have been revised to reflect the new methodology. These revisions present the changes and reasons for the changes in a satisfactory manner and are acceptable.
Current Bases are retained for Unit 2.
l.
(10)TS3/4.5.1.1, Accumulators,3/4.5.1.2,UpperHeadinjection;thesehave i
been changed or deleted to reflect the removal of the UHI system.
This is applicable to both units.
This change is acceptable. The administra-tive change of the word " pressurizer" to " Reactor Coolant System" is also acceptable.
I (11) TS 4.5.2.h, ECCS Subsystems; flows for the centrifugal charging pump lines and for the safety' injection pump lines, each with a single pump running are changed slightly (in appropriate directions) to match values 1
o 6-used in the LOCA analyses. The present configuration meets the revised limits.
The change is acceptable for both units, (12) TS 6.9.1.9, COLR; footnotes have been added to indicate that some of the references are only partially applicable, and new references have been ddded for the new CWFC methodology.
These additions are References 4, 7 and BAW-10168P.
These changes are acceptable.
2.2
SUMMARY
The NRC staff has reviewed the rep 0rts and other reference material submitted for the justification of the operation of Cycle 6 of CNSI. the use of the BWFC methodology for operation of that cycle and the associated TS revisions, and the safety analysis methodologies and results provided by BWFC.
Based on this review, the staff has concluded that appropriate material was submitted and that the operations, fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable.
The TS changes submitted for this reload for CNS1 and for the administrative changes to Units 1 and 2 reflect the necessary modifications for operation of Unit 1 Cycle 6.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
These amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment-on such finding (56 FR 20031). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of these amendments.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
H.Richings,SRXB F. Orr, SRXB Date:
May 31, 1991
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7
5.0 REFERENCES
1.
Letter, and attachments, from H. Tucker. Duke Power Company, to USNRC, dated January 9,1991, " Technical Specification Amendment, Unit 1 Cycle 6 Reload."
?.
BAW-2119 " Reload Peport, Catawba Unit 1, Cycle 6 " October 1990.
3.
PAW-10162P-A, " TACO 3, Fuel Pin Thermal Analysis Code," Babcock & Wl1407, Lunchburg, Virginia, November 1989.
4.
PAW-10152A, "N0ODLE -- A 11ulti-Dimensional Two-Group Reactor Simulator,"
Babcock & Wilcox, Lynchburg, Virginia, June 1985.
5.
BAW-10170P-A, " Statistical Core Design for Mixing Vane Cores," Babcock &
Wilcox, Lynchburg, Virginia, December 1988.
6.
BAW-10159P=.A. "BMCMV Correlatior of Critical Heat flux in Mixing Vane Grid Fuel Assemblies," Babcock & Wilcox, Lynchburg, Virginia, July 1990.
7.
PAW-10163P-A, " Core Operating Limit tiethodology for Westinghouse-Designed PWRs," Babcock & Wilcox, Lynchburg, Virginia, J. qe 1989.
8.
BAW-10172P, " Mark-BW Mechanical Design Report," Babcock & Wilcox, Lynchburg, Virginia, July 1988.
9.
BAW-10173P, Revision 2, " Mark-BW Reload Safety Analysis for Catawba and McGuire," Babcock & Wilcox, flovember 1990, 10.
BAW-10174, Revision 1, " Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units," Babcock & Wilcox, November 1990.
II. DPC-NE-2001P-A, " Fuel Mechanical Peload Analysis Methodology for Mark-BW Fuel," Revision 1, January 1990, Approved October 1990.
12.
Letter and attachment from M. Tuckman, Duke Power Company, to USNRC, dated fiarch 14,1991, " Response to Conditions Relative to the Use of Topical Report BAW-10173."
~
DATED:
M3v 31. 1991 AMENDMENT NO. 86 TO FACILITY OPERATitlG LICENSE NPF Catawba Nuclear Station, Unit 1 AMENDMENT N0. 80 TO FACILITY OPERATING LICENSE NPF Catawba Nuclear Station, Unit 2 DISTRICUTION:
Docket File NRC & Local PDRs PD 11-3 R/F Catawba R/F S. Varga 14-E-4 G. Lainas 14-H-3 R. Ingram 14-H-25 R. Martin 14-H-25 OGC-WF 15-B-18 D. Hagan MNBB 4702 G. Hill (8)
P1-37 W. Jones MNBB 7103 C. Grimes 11-F-22 ACRS(10)-
P-135 GPA/PA 17-F-2 OC/LFMB MNBB 4702 L. Reyes R11 H. Richings SRXB F. Orr SRXB l-i
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