ML20215E529
| ML20215E529 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/01/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215E523 | List: |
| References | |
| NUDOCS 8610150356 | |
| Download: ML20215E529 (7) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
- j WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 14 TO FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO.
6 TO FACILITY OPERATING LICENSE NPF-52 CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DUKE POWER COMPANY, ET AL.
I.
INTRODUCTION By letter dated July 15,1986 (Ref.1), Duke Power Company (the licensee) made 4
an application to amend facility operating licenses NPF-35 and NPF-52 for Catawba Nuclear Station Units 1 and 2, respectively, to reflect the Cycle 2 refueling and related Technical Specification (TS) changes for Unit I and a TS change for both Units related to application of a positive Moderator Temperature Coeffi-cient (MTC). A second letter (Ref. 2) provided some TS pages inadvertently omitted from Reference 1.
The proposed TS change applicable to both Units consists of increasing the allowable positive MTC and most negative E0L MTC. The previous TS allowed a +5 pcm/*F MTC at power levels up to 70% power, and 0 at power levels about 70%. The proposed revision would allow an MTC of +7 pcm/*F up to 70%, decreasing linearly above 70% power to O pcm/*F at 100% power.
II.
EVALUATION 1.
General Design The Catawba Unit 1, Cycle 2 reactor core contains 193 Optimized Fuel Assemblies.
During the Cycle 1/2 refueling, 64 Region 1 fuel assemblies will be replaced with 64 Region 4 fuel assemblies. The mechanical design of the Region 4 as-l semblies is the same as that of Regions 1, 2 and 3 except for the use of 304L stainless steel sleeves on the top grid, a small downward axial shift of the fuel rods and minor top grid modifications. The Region 4 fuel has been designed according to the fuel perfomance model in WCAP-8785 (Ref. 3). The fuel is de-signed and operateo so that clad flattening will not occur as predicted by the Westinghouse model in WCAP-8377 (proprietary) and in WCAP-8381 (non-proprietary)
(Ref. 4). For all fuel regions, the fuel-rod internal pressure design basis, which is discussed and shown acceptable in WCAP-8964 (Ref. 5), is satisfied.
The licensee provided a Reload Safety Evaluation (RSE) for Catawba 1 Cycle 2 as an attachment to Reference 1.
The RSE presents a cycle-specific evaluation for Cycle 2 which demonstrates that the core reload will not adversely affect the safety of the plant. This evaluation was perfomed utilizing the approved reload design methods of WCAP-9272-P-A (Ref. 6).
2.
Nuclear Design The Cycle 2 core loading is designed to meet an [Fn (Z) x P] ECCS limit of 1 2.32xK(Z). Adherence to the F limit is obtainelf by using the F TS g
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, surveillance described in WCAP-10217-A (Ref 7). F surveillance is part of the Relaxed Axial Offset Control (RA0C) and replace 9 the previous F surveill by comparing a measured F, increased to account for expected plant
- Maneuvers,ance This pr0vides a more convenient form of assuring plant to the F limit.
operatioO below the F limit while retaining the intent of using a measured n
parameter to verify operation below TS limits.
The above discussion is consistent with Reference 7 which was approved. Thus, the staff finds that the TS change to F surveillance is acceptable.
q RA0C will be employed in Cycle 2 to enhance operational flexibility during non steady state operation. RAOC makes use of available margin by expanding the allowable AI band, particularly at reduced power.
RAOC is described in Reference 7 and was approved by the staff.
Thus, it is acceptable for use in Catawba Unit 1.
During operation at or near steady state equilibrium conditions core peaking factors are significantly reduced due to the limited amount of xenon skewing possible under these operating conditions.
The licensee proposes to use Base Load TS to recognize this reduction in core peaking factors.
The proposed Base Load TS are identical to those that the staff has previously approved for McGuire Units 1 and 2, and are therefore acceptable.
The RSE provides a table of Cycle 2 kinetics characteristics which are compared with the current limits based on previously approved accident analyses. The RSE also provides a table showing the results of the calculated Cycle 2 control rod worths and requirements at the most limi' ting condition during the cycle (end-of-life). These results include a standard 10% allowance for calculational uncertainty. From these results the staff concludes that sufficient control rod worth will be available to provide the required shutdown margin for Cycle 2 operation. Control rod insertion limits were increased for less than 100%
power for Cycle 2.
Since the required shutdown margin is maintained, the TS change proposed to reflect the increased insertion is acceptable.
A more positive MTC than the current value is specified for Cycle 2.
This is evaluated elsewhere in this SER.
3.
Thermal and Hydraulic Design The thermal hydraulic methodology, DNBR correlation and core DNB limits used for Cycle 2 are consistent with the current licensing basis described in the FSAR and approved by the staff.
The power distributions produced by the cycle-specific RA0C analysis were analyzed for normal operation and Condition II events. Limits on the allow-able operating flux difference as a function of power level from these considerations were found to be less restrictive than those resulting from LOCA Fn considerations. The Condition II analyses generate DNB core limits and reVultant Over-Temperature Delta-T setpoints.
These generated a change to the F(al) function in the TS. The change is acceptable because W reults s
from cycle-specific calculations using approved methods (Refs. 6 and 7).
Therefore, the staff concludes that the Cycle 2 thermal-hydraulic analysis is acceptable.
so o 4.
Accident Analysis All the Cycle 2 kinetics parameters fall within the bounds upon which the pre-vious applicable safety analysis is based, except for the proposed change to the positive MTC, and the following reanalysis.
1 The Uncontrolled Boron Dilution events for full power operation and startup operation were reanalyzed to show that there is greater than 15 minutes, from time of alarm for operator action to tenninate the dilution before the minimum allowable shutdown margin is lost. The events were reanalyzed because the Cycle 1 analysis was cycle specific. The results show 156 minutes are avail-able for full power operation with the reactor in automatic control and 65 minutes with the reactor at full power and in manual control. The latter re-sult bounds the case for startup operation. Thus the results of the reanalysis are acceptable.
The licensee provided a report on the effect of the MTC change on accident analysis as an attachment to Reference 1.
The analysis applies to both Catawba Units 1 and 2, and is evaluated below.
The licensee has assessed the impact of a positive MTC of 7 pcm/*F on the ac-cident analyses presented in Chapter 15 of the FSAR. Those incidents which were found to be sensitive to positive or near-zero moderator coefficients were reanalyzed. These incidents are limited to transients which cause the reactor coolant temperature to increase. Accidents not reanalyzed included those resulting in excessive heat removal from the reactor coolant system, for which a large negative moderator coefficient is more limiting, and those for which heatup effects following reactor trip are not sensitive to the moderator coefficient. The staff agrees with 'the licensee's conclusions about which transients did and did not require, reanalysis. The transients not reanalyzed are:
RCCA misalignment / drop.
Startup of an inactive reactor coolant loop.
Excessive heat removal due to feedwater system malfunction.
Excessive load increase.
Spurious actuation of safety injection.
i (6) Rupture of a main steam pipe.
(7) Loss-of-coolant accident (LOCA)
The incidents reanalyzed, with two exceptions, used a +7 pcm/ F moderator tem-perature coefficient This is conservatike,, assumed to remain constant for variations in temperature.
since the proposed change will require the coefficient to ramp to zero at full power.
The two exceptions are the rod ejection and the rod withdrawal from subcritical accidents, for which the computer model cannot accept a constant coefficient. The coefficient decrease which occurred during the transients was less than the proposed change, which is acceptable.
The transients reanalyzed and their results are:
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. A.
Boron Dilution l
Boron dilution accidents during refueling or startup are terminated by operator action.
The proposed MTC does not reduce the time available for operator action in these modes below the acceptable value of 30 minutes from the time the oper-ator is alerted to reactor criticality. This is acceptable. The dilution analysis for power corditions with the reactor in automatic control assumes operator action based on the rod insertion alarm. Analysis of the transient shows the time for operator action remains above the acceptable value of 15 minutes. The dilution from power with the reactor in manual control is bounded by the rod withdrawal transient. Boron dilution accident results will therefore remain acceptable with the proposed MTC.
B.
Control Rod Bank Withdrawal from a Subcritical Condition This transient results in an uncontrolled addition of reactivity leading to a power excursion causing a heatup of the moderator and fuel. The time the core is critical before a reactor trip is very short so that the RCS temperature does not increase significantly; hence the effect of a positive MTC is small.
The analysis results show a transient average heat flux which does not exceed i~
the steady state full power value and an increased core water temperature that remains below the full power value. The results show that the DNBR remains above the limit value during the transient, which is acceptable.
C.
Uncontrolled Control Rod Bank Assembly Withdrawal at Power 4
This transient produces a mismatch in steam flow and core power, resulting in an increase in RCS temperature. However, the results show that the nuclear flux and overtemperature AT trips prevent the core minimum DNBR from falling below the limit value for this transient, which is acceptable.
D.
Loss of Coolant Flow The most severe loss of flow transient is caused by the simultaneous loss of power to all four reactor coolant pumps (RCPs). This case was reanalyzed to determine the effect of the positive MTC on the nuclear power transient and the resultant effect on the minimum DNBR reached during the transient. The minimum DNBR remains above the limit value during the transient, which is acceptable.
E.
Locked Rotor The locked event was reanalyzed because of the potential effect of the positive MTC on the nuclear power transient and thus on the RCS pressure and fuel temperature. A positive MTC will not affect the time to DNB because DNB is conservatively assumed to occur at the beginning of the transient. The results show peak RCS pressure and peak pellet average and peak cladding temperatures less than the limits used in the previously approved FSAR analyses, which is acceptable.
p 4 F.
Loss of External Electric Load The loss of external electric load transient was reanalyzed for both the beginning-of-life (BOL) and end-of-life cases. Since the MTC will be negative at end-of-life, the end-of-life results were essentially the same as in the FSAR. Two beginning-of-life cases were analyzed:
(1) reactor in the auto-matic rod control mode with operation of the pressurizer spray and pressurizer power operated relief valves (PORV); and (2) reactor in the manual control mode with no credit for pressurizer spray or PORVs. The result of a loss of load is a core power that momentarily exceeds the secondary system power removal, causing an increase in RCS coolant temperature. The reactivity addition due to a positive MTC causes an increase in both nuclear power and RCS pressure. The result for the control rods in automatic control and assuming pressurizer spray and relief at BOL is a RCS high pressurizer pressure. pressure of 2518 psia following a reactor trip on A minimum DNBR well above the applicable limits is reached shortly after reactor trip. The result for the case of rods in manual control with no credit for pressure control is a peak RCS pressure of 2563 psia following a reactor trip on high pressure. The minimum DNBR increases through-out the transient. Since the DNBR remains above the applicable limits and the peak RCS pressure is less than 110% of the design value of 2500 psia, the con-clusions presented in the previously approved FSAR analysis are still applicable.
G.
Loss of Normal Feedwater/ Loss of Offsite Power These accidents are analyzed to show the ability of the secondary system auxil-iary feedwater to remove decay heat from the reactor coolant system. The results show that the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS. For the case without offsite power, the results verify that the natural circulation capacity of the RCS provide sufficient i
heat removal capability to prevent fuel or clad damage following reactor coolant pump coastdown, i
i H.
Rupture of Mair; Feedwater Pipe This accident is analyzed to demonstrate the ability of the secondary system auxiliary feedwater to remove heat from the RCS. The results show that the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS to prevent overpressurization or core uncovery. For the case without offsite power, the results verify the natural circulation capacity of the RCS to prevent overpressurization and fuel or clad damage following reactor coolant pump coastdown.
I.
Control Rod Ejection The rod ejection transient was reanalyzed only for 80C since the MTC will be negative at EOC and the existing FSAR analysis remains applicable for E0C. The higher nuclear power levels and hotspot fuel temperatures resulting.from a rod ejection are increased by a positive MTC. The results from the B0C.reana,1ysis show that the fuel and clad temperatures are within the limiting va' lues speci-fied in the existing FSAR analysis. The peak hotspot fuel centerline temperature 4
exceeded the melting temperature for the full power case; however, melting was restricted to less than the innermost ten percent of the pellet. The fuel and f
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. clad temperatures do not exceed the limits specified in the previously approved FSAR analysis. Therefore, the results of the control rod ejection reanalysis are acceptable.
J.
Accidental Depressurization of the Reactor Coolant System The acceptance criteria for the accidental depressurization of the RCS were shown to be satisfied by predicting a minimum DNBR above the limit value for this transient.
Since the reanalysis of the affected plant transients does not result in ex-ceeding any of the fuel limits or safety limits specified in the previously approved reference or FSAR analyses, the staff concludes the analysis supporting operation with a positive moderator temperature coefficient of +7 pcm/ F up to 70% power, and decreasing linearly from this to 0 pcm/*F at full power will not pose an undue risk to the health and safety of the public and is therefore acceptable. The analysis is applicable to both Catawba Units 1 and 2, and therefore the proposed revision of the TS to incorporate the MTC for both Units is acceptable.
5.
Technical Specification Changes The TS Changes proposed in the licensee's submittals (Refs. I and 2) involves the following changes for Catawba Unit 1 only:
1 1.
RAOC and Axial Flux Difference Limits 2.
F Surveillance n
3.
Base Load Technical Specifications 4.
Rod Insertion Limits 5.
OT ATf(aI)
Acceptability of items 1-4 was discussed in Section 2, Nuclear Design. Accept-ability of item 5 was discussed in Section 3, thennal and hydraulic design. The proposed changes are for Unit 1 only but the actual change pages involve both Units, making the changes for Unit 1 and leaving the Unit 2 TS unchanged. The revisions to the bases are also acceptable.
In addition, the positive moderator coefficient change for both Units was found acceptable in Section 4, Accident Analysis. A second change to the moderator coefficient revised the most negative E0C coefficient to the value used in the accident analysis for both Units and is, therefore, acceptable.
III. ENVIRONMENTAL CONSIDERATION The amendments involve a change in use of facility components located within the restricted area as defined in 10 CFR Part 20 and a change in surveillance require-ments. The staff has detennined 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 incrdase in in-i dividual or cumulative occupational exposure. The Comission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there have been no public coments on such finding. Accord-ingly, the amendments meet the eligibility criteria for categorical exclusion I
~
- 6 set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or enviromental assessment need be prepared in connection with the issuance of the amendments.
IV. CONCLUSION The Comission made a proposed determination that the amendments involve no significant hazards consideration which was published in the Federal Register (51 FR 30567) on August 27, 1986, and consulted with the state of South Carolina.
No public comments were received, and the state of South Carolina did not have any coments.
We have 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, and (2) such activities will be conducted in compliance with the Comission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
References (1) Letter to H. R. Denton (NRC) from H. B. Tucker (Duke Power), " Catawba Nuclear Station, Docket Nos. 50-413 and 50-414 Catawbc 1/ Cycle 2 Reload,"
July 15, 1986.
(2) Letter to H. R. Denton (NRC) from H. B. Tucker (Duke Power), " Catawba Nuclear Station, Docket No. 50-413, Catawba Unit 1 Cycle 2 Reload," July 24, 1986.
(3) Miller, J.V., (Ed.), " Improved Analytical Model Used in Westinghouse Fuel Rod Design Computations," WCAP-8785, October 1976.
(4) George, R.A., (et. al.), " Revised Clad Flattening Model," WCAP-8377 (Proprietary) and WCAP-8381 (Non-Proprietary) July 1974.
(5) Risher, D.H., (et. al.), " Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis," WCAP-8964, June 1977.
(6) Davidson, S.L., et. al., " Westinghouse Reload Safety Evaluation Methodology," WCAP-9272-P-A, July 1985.
(7) Miller, R.
W., (et al.), " Relaxation of Constant Axial Offset Control-F Surveillance Technical Specification," WCAP-10217-A, June 1983.
O Principal Contributors:
Kahtan Jabbour, PWR#4/DPWR-A Marvin Dunenfeld, RSB/DPWR-A c.
Dated:
October 1, 1986
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