ML20035B913
| ML20035B913 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 03/23/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20035B911 | List: |
| References | |
| NUDOCS 9304050293 | |
| Download: ML20035B913 (17) | |
Text
ym UNITED STATES j
j NUCLEAR REGULATORY COMMISSION l
'(4 usmoros o e msssamt
- .....?
n r
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.107 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO.101TO FACILITY OPERATING LICENSE NPF-52 DUKE POWER COMPANY. ET AL.
CATAWBA NUCLEAR STATION. UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 t
i
1.0 INTRODUCTION
By letter dated December 15,1992 (Reference 1), as supplemented February 5 4
and March 18, 1993, Duke Power Company, et al. (the licensee) submitted a request for changes to the Catawba Nuclear Station, Units 1 and 2, Technical s
Specifications (TS).
The requested changes would revise the TS to reflect the reloading of Unit 2 with fuel manufactured by the Babcock & Wilcox (B&W) Fuel 4
Company, and analyzed using Duke Power Company methodology. The amendments also change the steamline and feedwater parameter setpoints and isolation times; the reactor makeup water pump minimum flowrate; and the pressurizer safety valve lift setpoint tolerance for both units.
The March 18, 1993, submittal requested that the amendments be issued prior to the expiration of the 30-day comment period provided in the Federal Reaister (58 FR 11260 dated February 24, 1993) and did not affect the proposed finding of no significant hazards consideration.
The submittal cited changed circumstances in that, through improved work planning and execution, its originally scheduled 68 day outage would be completed in 60 days, thus, necessitating an earlier than previously planned issuance of the amendments.
2.0 EVALUATION The Catawba Unit 2 plant operated in Cycle 5 with all fuel assemblies being of Westinghouse design.
Catawba Unit 2 Cycle 6 will be the first Catawba Unit 2 reload to contain a full reload batch of Mark-BW fuel assemblies (FAs) supplied by B&W Fuel Company (BWFC).
The use of Mark-BW fuel design in Catawba and the McGuire plants has been previously approved by the NRC (References 2 and 3).
2.1 General Description The Catawba Unit 2 reactor core consists of 193 FAs each of which is a 17x17 array containing 264 fuel rods, 24 guide tubes, and 1 incore instrument tube.
There are 117 burned FAs in the core, all of the Westinghouse Optimized fuel Assembly (0FA) design, and 76 fresh FAs consisting of the Mark-BW design. The Mark-BW fuel consists of dished end, cylindrical pellets of uranium dioxide.
9304050293 930323 PDR ADOCK 05000413 p
r i The nominal fuel loadings are 423.5 kg of uranium per fuel assembly for E'estinghouse fuel in batches 5A, 6A and 7A; and 456.2 kg of uranium per fuel assembly for the Mark-BW fuel in batch 8A.
The 8 batch SA, 33 batch 6A, and 76 batch 7A assemblies will be shuffled to
[
new locations. One batch 6A Ft. will be inserted into the center assembly location. The 76 fresh batch 8A assemblies will be loaded into the core in a symmetric checkerboard pattern. Cycle 6 will be operated in a' feed-and-bleed mode. Core reactivity is controlled by 53 rod cluster control assemblies (RCCAs), 640 Mark-BW burnable absorbers, and soluble boron shim.
T 2.2 Fuel System Desian The Catawba 2 Cycle 6 core will include 76 fresh Mark-BW fuel assemblies'with an enrichment of 3.75 wt % U235.
The re-inserted fuel assemblies in Cycle 6 will be 117 Westinghouse OFAs.
The Mark-BW 17x17 Zircaloy spacer grid fuel-assembly is similar in design to-the Westinghouse standard fuel assembly. The fuel rod outer diameter and guide tube top section, dashpot diameters,. and instrument tube diameter are the same as the Westinghouse standard 17x17 design. The unique features of the Mark-BW design include the Zircaloy.
i intermediate spacer grids, the spacer grid restraint system, and the use of Zircaloy grids with the standard lattice design. The Mark-BW-fuel is not unique in concept, nor does it utilize different component materials. -Thus, the chemical compatibility of all possible fuel-cladding-coolant assembly interactions for the fresh fuel is identical to that of the present fuel.
The mechanical analyses and thermal performance for the Mark-BW 17x17 design were performed by DPC using NRC-approved methodology (Reference 4); and-the-results of the cladding collapse time, the maximum cladding stress intensities, the diametral cladding strain and the maximum internal pin p
pressure are within the required limits. Therefore, the fuel system design is acceptable.
i 2.3 Nuclear Desian The core physics parameters for Cycle 6 were generated by DPC using NRC-approved methodology (Reference 5)-and are valid for the design cycle length (380 EFPD 10 EFPD). A representative relative power distribution for the beginning of Cycle 6 at full power was calculated as part of the design depletion using NRC-approved methodology (Reference 5) and assuming equilibrium xenon and rods in the All Rods Out (AR0) position. Calculated ejected rod worthi and their_ adherence to acceptance criteria were considered
~
during verification of the control rod insertion limits specified in COLR.
The adequacy of the shutdown margin with Cycle 6 stuck rod worths is also demonstrated including a 10% uncertainty on available rod worth.
The Reactor i
Protection System limits and operational limits for the core were verified by analyses for this fuel cycle using NRC-approved methodology (Reference 6) aLd i
are provided in the Technical Specifications. The results were derived by L
, t approved methods with appropriate assumptions. Therefore, the nuclear design is acceptable.
2.4 Thermal-Hydraulic Desian The generic and cycle-specific analyses supporting Cycle 6 operation were performed by DPC using NRC-approved methodology (Reference 7). Cycle 6 is the first Mark-BW transition cycle for Unit 2 and is analyzed using DPC's Statistical Core Design (SCD) methodology. Uncertainties on parameters that affect DNB performance are statistically combined to determine a statistical DNBR (departure from nucleate boiling ratio) limit (SDL). A generic SDL of 1.40 was calculated using NRC-approved BWCMV correlation (Reference 8) and a set of generic uncertainties given in Reference 7.
Reactor core safety limits for Cycle 6 were generated using BWCMV CHF correlation and SCD methodology for a full Mark-BW core and an enthalpy rise hot channel factor of 1.5.
The hydraulic compatibility of the Mark-BW fuel and the Westinghouse OFAs had been addressed in the approved topical report BAW-10173P-A, Revision 2 (Reference 2). The results of the hydraulic compatibility test indicated that the total pressure drop across the Mark-BW fuel is 2.4% lower than the total pressure drop across the OFA fuel. The licensee determined a generic transition core penalty by_modelling a conservative core configuration with one OFA assembly as the hot assembly located in a Ma-k-BW core. A number of statepoints and peaking conditions were analyzed, yielding a maximum DNBR penalty of 3.8% for the OFA fuel. The licensee addressed the transition core penalty for 0FA fuel by applying the 3.8% DNBR penalty against the 10.7%
generic margin included in the design DNBR limit (DDL), which is 1.55 for the generic Mark-BW and Catawba 2 Cycle 6 analyses.
Based on results of the analyses stated above, we have found the thermal-hydraulic desian is acceptoble since the approved methods are used.
2.5 Transient and Accident Analysis In order to determine the effects of this reload and to ensure that the thermal performance during hypothetical accidents is not degraded, the licensee has evaluated, using NRC-approved methodologies (References 5, 7, 9, 10,11 and 12), each Final Safety Analysis Report (FSAR) accident analysis sensitive to reload core physics parameters. The licensee has revised the i
licensing basis to reflect reanalysis of the thermal-hydraulic system transients including steam line break, turbine trip, feedwater line break, partial loss of forced reactor coolant flow, complete loss of forced reactor coolant flow, reactor coolant pump locked rotor, uncontrolled bank withdrawal from subcritical or low power startup condition, uncontrolled bank withdrawal' at power, dropped rod / rod bank, statically misaligned rod, single rod withdrawal, rod ejection, and steam generator tube rupture.
For the remaining FSAR thermal-hydraulic accident analyses sensitive to reload core physics parameters, e.g., LOCA, the current approved licensing bases are being retained.
In addition, the post-LOCA subtriticality evaluation and l
1
t !
boron precipitation evaluation have been performed by DPC in the FSAR for Catawba (Reference 13). The Catawba 2 Cycle 6 parameters have been reviewed with respect to the assumptions used in the subcriticality analysis.
The radiological consequences for the locked rotor, single rod withdrawal and rod ejection events were reanalyzed due to differences between the Mark-BW fuel and the OFA fuel fission product core inventories.
The results were reviewed and found acceptable Dy the staff (Reference 12). Review by the staff of Catawba Unit 2 Cycle 6 reload core physics parameters were found to be bounded by the accident analysis assumptions for all accidents which are sensitive to core physics parameters, and are, therefore, acceptable.
The LOCA analyses for Catawba Unit 2 transition cores with mixed Mark-BW and Westinghouse OFA assemblies and future core with all Mark-BW fuel have been reviewed previously by the NRC (Reference 3), and are, therefore, acceptable.
2.6 Technical Specification Chanaes The specifications on Catawba Units 1 and 2 TS pages are the same for both units, with a few exceptions, since the two units are identical in many respects. One of these exceptions involves the transition from fuel manufactured by Westinghouse to fuel manufactured by the B&W Fuel Company (BWFC) combined with a transition in analysis methodology to B&W and DPC methodology.
As these changes were first introduced into the Catawba Unit I plant, separate TS pages were generated for Units 1 and 2.
The changes for Unit 1 in Cycles 6 and 7 reflected the methodology change and transition to a mixed core of BWFC and Westinghouse manufactured fuel while separate pages for Unit 2 continued to reflect the Unit's reliance on Westinghot ce methodology and fuel.
A similar transition for Unit 2, beginning in its forthcoming Cycle 6, necessitates similar changes to its TS pages. This is accomplished by deleting the previous pages dedicated only to Unit 2 and making the previous pages dedicated only to Unit 1; again, applicable to both units. Thus, all of the changes discussed below, except for changes numbered 10, 13, and 14, were previously implemented for Unit 1 but are now applicable to both units.
Changes 10, 13, and 14 are first approved for Catawba in this amendment and are applicable to both units.
(1) TS 2.1.1 and Figure 2.1-lb This revision deletes Figure 2.1-lb for Unit 2 and identifies Figure 2.1-1 as applicable to both units to reflect the use of the BWCMV CHF correlation and DPC's SCD methodology with a 1.55 thermal design DNBR limit, which results in a decreased enthalpy rise hot channel factor from 1.55 to 1.5, and a reduced RCS minimum flow to 385,000 gpm.
Since the BWCMV CHF correlation and SCD methodology are based on the approved topical report, we conclude that the revised core safety limits are acceptable.
(2) TS Table 2.2-1 This revision changes the K values for the overtemperature and overpower delta T trip functions to reflect the use of BWCMV CHF correlation and k
i DPC's SCD methodology with 1.55 thermal design DNBR limit..The change to delete the Power Range Neutron Flux Negative Rate reactor trip function from the TS is acceptable, since it is no longer needed for the control rod drop event, when analyzed with the approved DPC methodology.
(3) TS 3/4.2.1 - Axial Flux Difference (AFD)
Baseload operation is deleted, since the reactor is not constrained to operate at a specified target AFD. The change is acceptable.
(4) TS 3/4.2.2 and 3/4.2.3 The changes were made to reflect the power peaking ' surveillance described in DPC-NE-2011P-A and are acceptable.
(5) TS 3/4.2.4 - Quadrant Power Tilt Ratio The change was made to provide quadrant power tilt ratio limits consistent with DPC methodology and is acceptable.
(6) TS 3/4.2.5 - DNB Parameters The change was made to provide DNB parameter limits consistent with DPC methodology and is acceptable.
(7) TS Tables 3.3-1, 3.3-2, and 4.3-1 The reactor trip on power range neutron flux negative rate is deleted from the Reactor Protection System, since it is no longer needed for the control rod drop event, when analyzed with the approved DPC methodology.
This change is acceptable.
(8) TS Table 3.3-4 Based on a reanalysis, the low steam line pressure setpoint for safety injection and main steam line isolation is changed from 725 psig to '775 psig.
The allowable value for this trip function is changed from 694 psig to 744 psig, maintaining the same 31 psig allowance for rack uncertainties, and the lead-lag controller for steam line pressure-low is deleted which eliminates spurious, ESF actuation on minor pressure increases in the secondary system. The TA, Z, and S columns are deleted.
from Table 3.3-4 which is consistent with the removal of these items from Table 2.2-1.
The allowable values associated with the RTD bypass system for the feedwater isolation on Tavg-Low and ESFAS P-12 interlock on Low-Low Tavg are deleted as a result of RTD bypass system being removed. -We find the changes to be acceptable.
(9) TS Table 3.3-5 The feedwater isolation response time is changed from 7 seconds to 12 seconds and the steam line isolation time is changed from 7 seconds to 10
i seconds.
The extended response times are consistent with or conservative for all licensing basis safety _ analyses. These two response times have been employed in the steam line break analysis (DPC-NE-3001P-A).
i Increasing these response times, from the current Technical Specification values, causes the primary system overcooling to worsen due to the extended blowdown of the intact generators and the additional mass of main feedwater delivered to the faulted generators.
Reanalysis (Reference 11) shows this transient does not exceed the imposed acceptance criterion of no DN3.
a (10) TS 3.3.3.11 and 4.3.3.11.2 The reactor makeup water pump flowrate limit foi Mode 5 is changed from 75 gpm to 70 gpm.
Each cycle, a bounding ratio of initial to critical boron concentration, is established from the reload design. This ratio is used to calculate the maximum reactor makeup water pump flowrate which satisfies the operator action time acceptance criteria of the Standard Review Plan (SRP).
The new flowrate limit is required to satisfy the operator action time acceptance criteria in the SRP and is acceptable.
(11) TS 3.4.2.1 and 3.4.2.2 This modification changes the allowable operational tolerances on the pressurizer safety valve setpoint from plus or minus 1% to +3%, -2% in all modes of operation. The actual bench test specification for these valves will remain at the 1% tolerance required by the staff, thereby ensuring accurate setpoints at the beginning of the operational cycle.
After verifying that the valves remain within the revised tolerances over several cycles, this larger tolerance will enable reduction of work in high radiation work environment by requiring only one valve to be tested per outage instead of three. The larger allowable deviation from the nominal lift setting is consistent with the licensing basis analyses. An increased pressurizer safety valve lift setpoint impacts the peak RCS pressure calculated for pressure increase transients, which is the result i
of a heatup in the RCS due to mismatch between the heat generated in the reactor core and the heat removed by the secor.dary system.
Three l
accident categories involving heat transfer mismatches, (1) decrease in secondary heat removal, (2) decrease in RCS flow rate, and (3) reactivity and power distribution anomaly transients, were analyzed by the licensee assuming a lift setpoint of 3% above the normal value.
These analyses using approved methodologies (Reference 9 and 11) showed that the peak RCS pressure criteria (110% of design pressure for.the feedline break and locked rotor events, and 120% of design pressure for the rod ejection transient) were met. The amount by which-the safety valve lift setpoint is allowed to drift downward is restricted to 2% of nominal in order to ensure that safety valve lift cannot freclude reactor trip on r
high pressurizer pressure. The licensee stated that reanalysis of DNB transients and the uncontrolled bank withdrawal at power and single rod I
withdrawal events showed that all acceptance criteria are met with the 2%
i i
downward setting. We find these changes are acceptable since the approved methodologies were used and the results of the analyses met the acceptance criteria.
l (12) TS 3/4.7.1.4 As stated in the technical justification for the proposed revision to TS Table 3.3-4, the valve stroke time, when added to the applicable C
instrumentation delays, yields the overall ESF response time. This response time is input to the steam line break transient analysis.
l Analysis using the approved DPC SCD methodology shows this transient does not violate the imposed acceptance criteria of no DNB.
l This change is acceptable.
i (13) TS 6.9.1.9 t
This change is to add NRC-approved Topical DPC-NE-1004A, " Nuclear Design
[
Methodology Using CASM0-3/ Simulate-3P," to the list of analytical methods used to determine the core operating limits.
This change'is acceptable.
l (14) TS Table 3.6-2a The proposed changes would revise the lists of containment isolation valves by changing the maximum opening time limitation from "s5" seconds to "NA" for the following valves:
Valves whose safety function is initiated by a Feedwater Isolation
[
Sianal:
Steam Generator Main Feedwater to Auxiliary Feedwater i
2 Nozzle Isolation Valves (CA-149 through CA-152)
Auxiliary Nozzle Temper Valves (CA-185 through CA-188) i Steam Generator Feedwater Containment Isolation Valves i
(CF-33, CF-42, CF-51, and CF-60)
Steam Generator Fe:dwater Purge Valves (CF-87 through l
CF-90)
Valves whose safety function is initiated by a Steamline Isolation Sicnal:
Main Steam Isolation Valves '(SM-1, SM-3, SM-5, and SM-7) i Main Steam Isolation Bypass Control Valves (SM-9, SM-10, SM-ll, and SM-12).
ii The licensee's technical justification for the changes is based on the following: (a) the valves' protective action functions are initiated by' signals other than containment isolation signals, and (b) credit for the
. operation of these valves is not taken in the dose analyses.
i J
b t
l
i
, The safety function of the above valves. is to prevent or mitigate the effects of excessive cooldown and its reactivity insertion effects during postulated events. Also, the valves serve as the second barrier in their' respective primary containment piping penetrations. The FSAR-Chapter 15
?
analyses for certain analyzed transients and design basis accidents involving cooldown are based on the assumption that these valves close i
within 5 seconds of receipt of a closure initiation signal.
Closure of the-valves is initiated by either a Feedwater Isolation Signal or a Steamline Isolation Signal as indicated above. Containment isolation signals (i.e., Containment Phase A and Phase B signals) do not themselves directly actuate any of the above valves, however, the Steamline Isolation Signal and the Feedwater Isolation Signal share' common monitored parameters with the Containment Phase A and Containment Phase B isolation signals. The $5-sec. closure time assumed in the FSAR analyses for the purpose of mitigating cooldown effect is more limiting than the time limit allowed for the containment isolation function which is typically 15 seconds.
Since the steam and feedwater isolation functions are safety-related functions for which the protective actions must be completed within an assumed time period, it is appropriate that the Technical Specifications include operability criteria which ensures that the protective actions can be accomplished within time limits assumed in the FSAR analyses.
Such assurance is provided elsewhere in the Technical Specifications.
Table 3.3-5 " Engineered Safety Features Response Times" specifies, for both feedwater and steamline isolation features, maximum protective action time limits which identify the maximum acceptable total time allotted for an entire protection action, encompassing the time from which a monitored parameter exceeds its setpoint, to the time at which the actuated valve is closed (Reference Technical Specification Section 1.13 " Definition of ENGINEERED SAFETY FEATURES RESPONSE. TIME).
Since the requirements established by Table 3.3-5 ensure the timely actuation of-the valves listed above, and the subject valves serve no other safety function, deletion of the closure time specification from isolation valve tables 3.6-2a/b is acceptable.
Based on the results of our review, we conclude that the proposed TS i
changes are acceptable.
The staff has reviewed the licensee's submittal to support Cycle 6 operation and proposed TS changes for the Catawba Unit 2. The staff concludes that the Cycle 6 operation and proposed TS changes are acceptable.
Further, we conclude that the combined TS are acceptable for Catawba Unit 1.
3.0 CHANGED CIRCUMSTANCES As stated in the Notice of Consideration of Amendments To facility Operating Licenses, Proposed No Significant Hazards Consideration Determination and Opportunity For Hearing published in the Federal Reaister (58 FR 11260, dated February 24,1993), the Commission will not normally issue the amendment until the expiration of the 30-day notice period (March 26,1993). The Notice also states that the Commission may issue the amendment before the expiration of
l
! the 30-day notice under certain circumstances provided that its final
[
determination is that the amendment involves no significant hazards
~
consideration. The licensee has requested in a letter dated March 18, 1993, that the amendment be issued by March 24, 1993, which is 2 days before the expiration of the 30-day notice.
The licensee cites changed circumstances in that its initially planned outage duration was 68 days with an expected readiness for MODE 3 on April 2, 1993, whereas, due to improved Sork planning and executica, the outage is now expected to last 60 days and the plant to be ready for MODE 3 on March 25, 1993.
Thus, the licensee requests the amendment l
to be issued before March 25, 1993. As a result of changed circumstances, there is a need for the Commission to act before the expiration of the 30-day comment period in order to permit the unit to proceed with startup operations upon completion of its refueling outage. Therefore, the NRC staff is making a i
final no significant hazards determination and will make public a notice of issuance and provide an opportunity for a hearing after issuance.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). 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; or (2) create the 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.
The following evaluation in relation to the three standards demonstrates that the proposed amendment does not involve a significant hazards consideration.
POWER DISTRIBUTION AND SAFETY LIMITS l
Catawba Unit 1 Cycle 6 was the first [ Catawba] Nuclear Station [ reload]
for which B&W Fuel Company (BWFC) supplied the reload fuel.
The Catawba Unit 1, Cycle 6 Reload Report presented an evaluation that concluded the core reload using Mark-BW fuel would not adversely impact the safety of a
the plant.
The Catawba Unit 1, Cycle 7 report was similar, but reflected that Duke Power performed the analyses in support of.the operation of Cycle 7 rather than BWFC. This reload for Catawba Unit 2, Cycle 6 is a compilation of the changes made for Unit I during Cycles 6 and 7 in that it justifies the use of Mark-BW fuel using Duke Power analysis.
The Catawba Unit 2, Cycle 6 Reload Safety Evaluation Report presents an evaluation which demonstrates that the core reload using Mark-BW fuel will not adversely impact the safety of the plant. During Cycle 6, the core will contain 76 fresh fuel assemblies supplied by B&W and 117 Westinghouse supplied Optimized Fuel Assemblies (OFA).
1 The changes to the Safety Limit and Power Distribution Technical Specifications presented in Section 8 of the Reload Report represent the application of previously approved methodology to Catawba Unit 2.
The
- D I
i
! changes to remove the power range neutron flux negative rate reactor trip, increase the low steam line pressure setpoint, increase feedwater isolation response time, increase steam line isolation response time, increase pressurizer safety valve lift setpoint tolerance, remove steam line pressure dynamic compensation,... and increase main steam line isolation valve stroke time reflect the use of Duke analysis, and have already been approved for Catawba Unit 1.
The changes described above include the deletion of references to specific units on individual Technical Specification pages, and delete pages which were previously for Unit 2 only. The implementation of unit specific references became necessary due to the transition from Westinghouse to B&W supplied fuel during Unit 1 Cycle 6 and for the Unit 1 Cycle 7 Reload due to the transition to Duke analysis methodology. The analysis which made the i
changes necessary in the Unit I reload submittal is generic, and as described in the technical justification, is equally applicable to both McGuire and Catawba units.
A LOCA evaluation for operation of Catawba Nuclear Station with Mark-BW fuel has been completed (BAW 10174, Mark-BW Reload LOCA Analysis for the Catawba and McGuire Units). Operation of the station while in transition -
from Westinghouse supplied 0FA fuel to B&W supplied Mark-BW fuel is also justified in this topical.
BAW-10174 demonstrates that Catawba Nuclear Station continues to meet the i
criteria of 10 CFR 50.46 when operated with Mark-BW fuel.
Large Break LOCA calculations completed consistent with an approved evaluation model (BAW-10168P and revisions) demonstrate compliance with 10 CFR 50.46 for breaks up to and including the double ended severance of the largest primary coolant pipe. The small break LOCA calculations used to license the plant during previous fuel cycles are shown to be bounding with respect to the new fuel design. This demonstrates that the plant meets 10 CFR 50.46 criteria when the core is loaded with Mark-BW fuel.
During the transition from Westinghouse OFA fuel to Mark-BW fuel, both types of fuel assemblies will reside in the core for several fuel cycles.
Appendix A to BAW-10174 demonstrates that results presented above apply to the Mark-BW fuel in the transition core, and that insertion of the Mark-BW fuel will not have an adverse impact on the cooling of the i
Westinghouse fuel assemblies.
~
Duke Power Company's Topical Reports DPC-NE-3000, DPC-NE-3001-PA, and DPC-NE-2004-PA provide evaluations and analyses for non-LOCA transients which are applicable to Catawba.
The scope of these analyses includes all events specified by sections 15.1-15.6 of Regulatory Guide 1.70 (Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants) and presented in the Final Safety Analysis Report for Catawba.
The analysis and evaluations performed for these topicals confirm that operation of Catawba Nuclear Station for reload cycles with Mark-BW fuel will continue to be within the previously reviewed and licensed safety limits.
I
F
, One of the primary objectives of the Mark-BW replacement fuel is compatibility with the resident Westinghouse fuel assemblies. The description of the Mark-BW fuel design and the thermal-hydraulics and the core physics performance evaluation demonstrate the similarity between the reload fuel and the resident fuel.
The extensive testing and analysis summarized in BAW-10173P shows that the Mark-BW fuel design performs, from the standpoint of neutronics and thermal-hydraulics, within the bounds and limiting design criteria applied to the resident Westinghouse fuel for the Catawba plant safety analysis.
Each FSAR accident has been reviewed to determine the effects of Cycle 6 operation and to ensure that the radiological consequences of postulated accidents are within applicable regulatory guidelines, and do not adversely affect the health and safety of the public. The design basis LOCA evaluations assessed the radiological impact of differences between the Mark-BW fuel and Westinghouse OFA fuel fission product core inventories. Also, the dose calculation effects from non-LOCA transients reanalyzed by Duke Power were evaluated using Cycle 6 characteristics.
The calculated radiological consequences are all within specified regulatory guidelines and contain significant levels of margin.
The analyses contained in the referenced Topical Reports indicate that the existing design criteria will continue to be met. Therefore, the enclosed TS changes will not increase the probability or consequences of an accident previously evaluated.
As stated in the above discussion, normal operational conditions and all fuel-related transients have been evaluated for the use of Mark-BW fuel at Catawba Nuclear Station. Testing and analysis was also completed to ensure that, from the standpoint of neutronics and thermal-hydraulics, the Mark-BW fuel would perform within the limiting design criteria.
Because the Mark-BW fuel performs within the previously licensed safety limits, the possibility of a new or different accident from any previously evaluated is not created.
The reload-related changes to the TSs do not involve a significant reduction in the margin of safety. The calculations ~and evaluations documented in BAW-10174 show that Catawba will continue to meet the eriteria of 10 CFR 50.46 when operated with Mark-BW fuel. The evaluation of non-LOCA transients documented in DPC-NE-3001 also confirms that Catawba will continue to operate within previously reviewed and licensed safety limits.
Because of this, the TS changes to support the use of Mark-BW fuel will not involve a significant reduction in the margin of safety.
The technical changes made to Table 2.2-1 reflect the use of the BWCMV CHF correlation and Duke Power's Statistical Core Design methodology with a 1.55 thermal design limit. These changes to Table 2.2-1 will not significantly increase the probability or consequences of an accident previously evaluated..the changes to the K values conservatively bound the allowable operating region, as defined by the new DNBR methodology.
i 1
It'can be concluded that these changes will not create the possibility of' any new accident from those previously evaluated.
It can also be concluded that since all new TS values are bounded by safety analysis assumptions that this change will not significantly decrease the margin.
of safety.
DELETION OF NEUTRON FLUX HIGH NEGATIVE RATE TRIP The removal of the Power Range Neutron Flux High Negative Rate trip will not result in any previously-reviewed accident becoming more probable or more severe. The trip is'a response to a pre-existing transient condition and would not initiate any accident. The trip is designed to
?
provide protection from a dropped control rod. However, in the event of a dropped rod, the reactor is assumed to trip on low pressurizer pressure. Therefore, the protection function is retained.
The consequences of a dropped rod have been analyzed and found to be within acceptable limits.
Likewise, the removal of this trip will not create a new accident not previously reviewed. The removal of a response to a transient will not initiate a new transient. There are no credible unanalyzed transients which will occur as a result of a dropped rod. The removal of this trip
?
will reduce the potential for spurious or unnecessary trips which may occur as a result of maintenance or the drop of a low-worth rod. There are no other hardware modifications or procedure changes that will be i
made as a result of this deletion which could create the possibility of a new accident.
No margin of safety will be reduced by this change. As noted above, if a dropped rod necessitates a trip, the trip function will be accomplished as a result of low pressurizer pressure.
For those dropped rods for which no trip is necessary, the removal of this trip will provide protection against an unnecessary transient.
LOW STEAM LINE SETP0 INT PRESSURE CHANGE Changing the Low Steam Line Pressure setpoint and removal of dynamic compensation will not increase the probability or consequences of any previously-reviewed accident. The higher steam line pressure setpoint is consistent with all licensing basis safety analyses.
This change, in i
conjunction with the removal of the dynamic compensation of the steam pressure signal, is intended to reduce or eliminate spurious Engineered Safeguards Features (ESF) actuations which are caused by minor (but i
rapid) pressure decreases in the secondary system.
The proposed amendment will not result in a new accident not previously reviewed. A change in steam line pressure is a response to an existing transient condition, rather than a precursor or initiating event. A change in the steam line pressure setpoint is also not a precursor or initiating event.
t t
_ The proposed amendment will not result in a significant decrease in a margin of safety. The reanalysis of the steam line break accident which was performed shows that all imposed Condition 11 acceptance criteria are met.
Based on the above, it is concluded that no significant hazards exist.
FEEDWATER AND MAIN STEAM LINE ISOLATION VALVE STR0KE TIME The proposed changes to the valve stroke times in Table 3.3-5 and Table 3.6-2a will not significantly increase the probability or consequences of any previously evaluated accident. The effects of the delays in isolation times on the various transients affected have been analyzed and found to be acceptable. Since these valves do not receive a containment isolation signal, and no credit is taken for operation of these valves in the dose analysis for a containment isolation function, a maximum stroke time does not apply for containment isolation.
The proposed changes will not significantly increase the possibility of a new accident not previously evaluated. Feedwater and main steam isolation are responses to ongoing transients, rather than initiators or l
precursors of transients. No equipment or component reconfiguration will occur as a result of this change.
?
The proposed changes will not significantly decrease any margin of safety.
As noted above, the effects of the longer isolation times have been evaluated and found to be acceptable.
Based on the above, it is concluded that no significant hazards exist.
INCREASE IN PRESSURIZER CODE SAFETY VALVE SETPOINT TOLERANCES The proposed amendment will not result in a significant increase in the probability or consequences of any previously analyzed accident. The valve lift setting is challenged only after a transient has been initiated and is not a contributor to the probability of any transient or accident. The transients which involve pressure increases which would potentially challenge the safety valves have been analyzed to determine the consequences of delayed or premature valve actuation at the extremes of the new setpoint tolerances. These analyses show that all applicable acceptance criteria are met using 'the wider tolerances.
The proposed amendment will not result in the creation of any new accident not previously evaluated.
As noted above, the setpoint tolerance only affects the time at which the safety valve opens following or during a transient, and is not a contributor to the probability of an i
accident.
i l
The proposed amendment will not result in a significant decrease in a j
margin of safety.
The limiting transient in each accident category has j
been analyzed to determine the effect of the change in lift setpoint l
i
L i tolerance on the transient.
In ecch case, the results of the analyses met all acceptance criteria.
Based on the above, it is concluded that no significant hazards exist.
CONTAINMENT ISOLATION VALVES The proposed changes to the valve stroke times in Table [s] 3.6-2a and 3.6[-]2b will not significantly increase the probability or consequences of any previously evaluated accident. The effects of the delays in isolation times on the various transients affected have been analyzed and found to be acceptable.
Since these valves do not receive a containment isolation signal, and no credit is taken for operation of these valves in the dose analysis for a containment isolation function, a maximum stroke time does not apply'for containment isolation.
The proposed changes will not significantly increase the possibility of a new accident not previously evaluated.
Feedwater and main steam isolation are responses to ongoing transients, rather than initiators or precursors of transients. No equipment or component reconfiguration will occur as a result of this change.
The proposed changes will not significantly decrease any margin of safety. The isolation times which are applicable to these valves are specified in Table 3.3-5, Engineered Safety Features Response Times. The effects of the isolation of these valves was evaluated based on their ESF function, not a containment isolation function, and determined to be acceptable, therefore there is no significant decrease in the margin of safety.
l BORON DILUTION MITIGATION SYSTEM TS 3.3.3.ll.a.2 is changed to reduce the allowable Reactor Makeup Water Pump flow in Mode 5 from 75 gpm to 70 gpm.
In the event that the Baron Dilution Mitigation System (BDMS) is inoperable the Reactor Makeup Water Pump flowrates are limited to ensure that operator action times required to terminate a dilution event can be met. The limits on reactor makeup water pump flowrates when the BDMS is inoperable are verified each cycle to ensure that the safety analysis assumptions for these parameters remain valid.
When the calculated Reactor Makeup Water Pump flowrate is found to be less than the existing flowrate limits, the flowrate limit must be reduced so that the operator action time acceptance criteria of Standard Review Plan 15.4.6 can be met.
Reducing the allowable Reactor Makeup Water Pump flow in Mode 5 does not involve a significant increase in the probability or consequences of an accident previously evaluated. The current TS flowrate does not allow enough time for the operator to terminate an uncontrolled dilution event i
when required operator response times are assumed.
The lower flowrate allows needed operator response times and is therefore more conservative.
I
l !
Reducing the allowable Reactor Makeup Water Pump flow in Mode 5 does not change the way_that any plant equipment is operated or maintained, therefore it does not create the possibility of a new or different accident.
i Reducing the Allowable Reactor Makeup Water Pump Flow in Mode 5 will not involve a significant reduction in the margin of safety.
This flowrate is more conservative, and ensures that safety analysis assumptions regarding operator actions times in response to the termination of an uncontrolled dilution event can be met.
CORE OPERATING LIMITS REPORT
]
The proposed change to TS 6.9.1.9 adds approved topical-DPC-NE-1004A to the list of analytical methods used to determine core operating limits.
This change is administrative, adding a topical report which has been approved for use on Catawba to the list of analytical methods used to determine core operating limits.
Since this change is administrative it has been determined that no significant hazards are involved..
In addition, the NRC staff finds that editorial and format changes, including deleting pages no longer applicable to Unit 2, and the renumbering and redesignation of remaining pages in certain sections as applicable to both units, will not involve a significant increase in the probability or i
consequences of an accident previously evaluated. These changes will not create the possibility of a new or different kind of accident from any l
accident previously evaluated for the use of BWFC fuel and the revised analysis methodology in the Catawba units. These changes will not involve a i
significant reduction in a margin of safety since they reflect the usage of fuel and analysis methodology that have been previously approved for the Catawba units.
t Based on the foregoing, the NRC staff has concluded that the standards of i
10 CFR 50.92(c) are satisfied. Therefore, the Commission has made a final determination that the proposed amendments do not involve a significant hazards consideration.
5.0 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.
6.0 ENVIRONMENTAL CONSIDERATION
)
The amendments change requirements with respect to installation or use of a
)
facility. component 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 i
l l
P !
occupational radiation exposure. The Commission has previously issued a f
proposed finding that the amendments involve no significant hazards l
consideration, and there has been no public comment on such finding (58 FR l
11260 dated February 24, 1993). Accordingly, the-amendments meet the i
eligibility criteria for categorical exclusion set forth in 10 CFR i
51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement-or r
environmental assessment need be prepared in connection with the issuance of the amendments.
7.0 CONCLUSION
l 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 i
activities will be conducted in compliance with the Commission's regulations,_
l 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:
T. L. Huang l
W. O. Long
}
i Date: March 23, 1993 l
l l
l i
t l
I i
4 f
l
, REFERENCES 1.
Letter from M. S. Tuckman (DPC) to USNRC, dated December 15, 1992.
2.
BAW-10173P-A, Mark-BW Reload Safety Analysis for Catawba and McGuire, Babcock & Wilcox, Revision 2, February 20, 1991.
3.
BAW-10174P-A, Mark-BW Reload LOCA Analysis for the Catawba and Mcbuire Units, Babcock & Wilcox, Revision 1, November 1990.
4.
DPC-NE-200lP-A, Rev.1, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Duke Power Company, October 1990.
5.
DPC-NF-2010A, McGuire Nuclear Station / Catawba Nuclear Station Nuclear Physics Methodology for Reload Design, Duke Power Company, June 1985.
6.
DPC-NE-20llP-A, Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors, Duke Power Company, March 1990.
7.
DPC-NE-2004P-A, McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01, Duke Power Company, December 1991.
8.
BAW-10159P-A, BWCMV Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, Babcock & Wilcox, July 1990.
9.
DPC-NE-3002-A, Mcguire Nuclear Station / Catawba Nuclear Station FSAR Chapter 15 System Transient Analysis Methodology, November 1991.
10.
DPC-NE-3000P, Duke Power Company, Thermal-Hydraulic Transient Analysis Methodology, Revision 2, February 20, 1990.
- 11. DPC-NE-300lP-A, Duke Power Company, Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology, Revision 2, November 1991.
12.
Catawba Nuclear Station Unit 1, Docket Number 50-413 and 50-414, Cycle 7 Relcad Submittal, Duke Power Company, April 13, 1992.
13.
Catawba Nuclear Station, Final Safety Analysis Report, Docket Nos.
50-413/414.
-_