ML20056H635
| ML20056H635 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/30/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20056H633 | List: |
| References | |
| NUDOCS 9309100239 | |
| Download: ML20056H635 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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WASHINGTC1, D.C. 20555-0001 v
y SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 152 TO FACILITY OPERATING LICENSE NO, NPF-7 i
VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POWER STATION. UNIT NO. 2 DOCKET N0. 50-339 l
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1.0 INTRODUCTION
By letter dated December 4,1992, the Virginia Electric and Power Company (the licensee) proposed a change to the Technical Specifications (TS) for the North Anna Power Station, Unit No. 2 (NA-2). The proposed changes would revise Table 2.2-1, Reactor Trip System Instrumentation Trip Setpoints, of TS 2.2-1 l
and Table 3.2-1, DNB Parameters, of TS 3.2.5 to allow a reduction in the i
minimum measured reactor coolant system (RCS) flow rate. The licensee has made this request in anticipation of increased steam generator tube plugging (SGTP) at NA-2.
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The licensee has examined the Chapter 15 accident analyses and determined
'l which transients required reanalysis and which required only reevaluation.
The criteria used by the licensee to determine whether to reanalyze or reevaluate was that (1) if the event is potentially impacted by RCS flow rate and also by other effects of SGTP, the event was either evaluated or reanalyzed, (2) if a departure from nucleate boiling ratio (DNBR) limited i
event is impacted by RCS flow rate hut not by other effects of SGTP, a DNBR oenalty is assessed, and (3) if an event is unaffected by RCS flow rate, it is not addressed.
2.0 DISCUSSION The primary consequences of increasing the SGTP are (1) increased reactor coolant system loop resistance, resulting in a lower RCS flow rate, (2) decreased steam generator tube heat transfer area, resulting in lower steam generator outlet steam pressure, and (3) a decreased total RCS volume.
The licensee measured the NA-2 flow in April 1992 at an average SGTP level of 7.0%
and found it to bc-293,321 gpm, which is greater than the TS limit of 284,000 gpm. To conservatively bound the flow rates reulting from future increases in SGTP levels up to approximately 18%, the licensee proposed a minimum
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9309100239 930830 DR ADOCK 05000339 PDR w
measured flow of 275,300 gpm, 3% lower than the current TS flow rate. The licensee selected this value based on experience gained during the NA-1 extended SGTP effort.
The increase in tube plugging results in a reduction in RCS total flow rate through the core.
This core flow reduction can challenge the DNB design limits.
The statistical DNBR limit (SDL) of 1.26 was achieved by combining key DNBR analysis parameter uncertainties in a statistical manner with the WRB-1 Critical Heat Flux (CHF) correlation.
The transient analyses results were assessed against a 1.46 design DNBR limit. The percentage difference between the design DNBR limit and the SDL represents the generic retained margin. This means that an additional 13.7% DNBR remains available to i
accommodate changing plant conditions.
t 2.1 REEVALVATED EVENTS The events that were not reanalyzed, but were evaluated and found to be affected by the flow reduction, had a DNBR penalty applied. The DNBR penalty of 4.8% was calculated by approved methods (reference 2) and is based on a flow reduction of 3.0% and a bounding WRB-1 DNBR partial derivative of 1.6%.
The 4.8% penalty is subtracted directly from the generic retained margin i
13.7%, leaving an 8.9% margin for other plant modifications. Those accidents accommodated by this single penalty are listed below:
15.2.1 Rod Withdrawal from Subcritical 15.2.5 Partial loss of Flow 15.2.10 Excessive Heat Removal 15.2.11 Excessive Load Increase 15.2.12 Accidental Depressurization of the RCS 15.2.13 Accidental Depressurization of the Main Steam System 15.2.14 Spurious Operation of the Safety Injection System 15.3.7 Single Rod Withdrawal at Power The 4.8% penalty is also extracted from the availablo Co.e Thermal Limit retained DNBR margin. This is to ensure that the bounding Core Thermal Limit protection is provided by the existing Overtemperature aT and Overpower AT reactor protection system for the reduced flow rate.
The disposition of the remaining Chapter 15 transients which were reevaluated is as follows:
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Main Steamline Break (15.4.2.1)
The main steamline break (MSLB) accident analysis uses the W-3 CHF correlation l
for DNBR calculations. The W-3 correlation has a different DNBR sensitivity to marginal changes in flow. When the conditions associated with MSLB were applied to the reduced flow rate, the flow reduction translated into a 4.3%
i penalty. This penalty was assessed against the generic retained margin.
Steam Generator Tube Rupture (15.4.3)
The steam generator tube rupture (SGTR) leads to increased contamination of i
the secondary sjstem due to leakage of radioactive coolant from the RCS. The analysis assumes the operator terminates the primary to secondary leakage within 30 minutes. The reduction in RCS flow rate will not adversely affect the operator's ability to control an SGTR.
Small Break LOCA (15.3.1)
The Emergency Core Cooling System (ECCS) was reanalyzed by approved methods (Reference 4), and found to meet the acceptance criteria of 10 CFR 50.46, for performance during a postulated small-break loss-of-coolant accident (SBLOCA).
The reanalysis included SGTP up to 35% in any single steam generator. This bounds the RCS conditions associated with the proposed flow reduction.
large Break LOCA (15.4.1) l i
l The licensee submitted a letter dated July 16, 1993 in compliance with the reporting requirements of 10 CFR 50.46. The licensee indicated that they have reanalyzed the LBLOCA event by approved methods and found the new LBLOCA peak cladding temperature (PCT) to be 2019'F. This PCT is within the 10 CFR 50.46 acceptance criteria of 2200*F and the analysis includes the reduced RCS flow rate of 264,400 gpm.
LBLOCA remains the bounding event for NA-l&2.
2.2 Reanalyzed Events The reanalyzed transients were initially reanalyzed and approved for the NA-1 extended SGTP evaluation (Reference 2).
The licensee reanalyzed the system transient portion of the analyses using the licensee's RETRAN system transient analysis code single and double loop models. The moiels were modified to reflect the effects of reduced RCS flow rate assoct&ted with increased SGTP.
Specifically, RCS flow rates, steam generator tube heat transfer areas (outside and inside the tubes), SG tube metal volume (heat capacity), and SG tube flow area were reduced to reflect plugging effects. The reanalyzed transients are detailed below.
Loss of External Load (Section 15.2.7)
The cases reanalyzed for loss of external load are beginning of cycle (B0C) with pressure control and B0C without pressure control. These cases represent the most limiting DNB and overpower cases, respectively. The calculated DNBR increased throughout the transient from the initial value of 2.15.
Both the m,
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RCS peak pressure and the main steam peak pressure remain below the acceptance 1
criteria.
Loss of Normal Feedwater (Section 15.L8 and 15.2.9) l The loss of normal feedwater was reanalyzed with consideration given to the availability of offsite power.
In both cases (with and without ovfsite power) the analysis demonstrated that increased steam generator tube. plugging levels did not adversely impact the ability of the auxiliary feedwater system to i
acequately perform its function.
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Rod Bank Withdrawal at Power (Section 15.2.2)
A wide range of initial plant conditions were reanalyzed to identify the most limiting rod withdrawal at power event cases. The results indicated that for all cases the minimum DNBR remained above the design limit value. Also, the reanalysis confirmed that the current TS setpoints for overtemperature and overpower AT trip continue to provide core thermal limit protection under extended SGTP conditions.
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l Complete Loss of Flow (Section 15.3.4) l The complete loss of flow event was analyzed for two cases, the complete loss l
of voltage at the RCP breakers, undervoltage (UV), and 5.0 Hz/sec decay rate of the supply frequency, underfrequency (UF). The transient DNBRs remained above the statistical DNBR design limit throughout the transient for both the i
UF and UV events.
l Locked Reactor Coolant Pump Rotor Event (Section 15.4.4)
The locked rotor event was analyzed in two parts: (1) a peak pressure calculation was performed assuming that no fuel rods experience DNB; and (2) i the calculation was repeated assuming a rod' experiences DNB with cladding failure. For the former case, the results indicate that peak RCS pressure remains within the acceptance limit of 2750 psia.
For the latter case, the criterion of less than 13% of fuel rods experiencing DNB at the limiting time in core life (for the current operating cycle) continues to be. met.
The Chapter 15 events that were not impacted by the extended SGTP and were not reanalyzed or reevaluated by the licensee are Inactive Loop Start-up (Section l
15.2.6) and Misloaded Fuel Assembly (Section 15.3.3).
3.0 EVALUATION The staff reviewed the licensee's submittal proposii.g the reduction of RCS flow rate in anticipation of increased SGTP. The licensee's proposal evaluated the Chapter 15 events for potential impact of reduced RCS flow rate and extended SGTP on accident analyses. Of the Chapter 15 events evaluated, five were specifically reanalyzed; (1) Complete Loss of Flow, (2) Loss of Normal Feedwater, (3) Rod Bank Withdrawal at Power, -(4) Locked Reactor Coolant Pump Rotor Event, and (5) Loss of External Load.-
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3 Those events that were evaluated, but not reanalyzed, accommodated the decrease in RCS flow by imposing a single 4.8% penalty on the retained DNBR margin. Finally, those Chapter 15 events that were not impacted by the reduced flow or increased SGTP were not addressed.
The events were reanalyzed by approved methods and the DNBRs were found to remain within the statistical DNBR design limits throughout the transients.
The single penalty was derived by approved methods (Ref(rence 3) and the DNBR for each of these events remained within the generic retained margin.
The licensee has shown that with the decreased RCS flow hA-2 continues to meet the acceptance cr',teria for the retained DNBR margin and also under those circumstances where the DNBR penalty was applied, the licensee continues to satisfy the limits of the generic retained margin.
By doing so, the licensee ensures that a 95% confidence level exists against DNB occurring on at least 95% of the limiting fuel rods. The staff, therefore, finds the proposed NA-?
TS RCS flow reduction to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendment. The State official had no comment.
5.0 ENVIRONMENTAL CONSIDERATION
This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 this amendment involves no significant hazards consideration and there has been no public comment on such finding (58 FR 7008). Accordingly, this amendment meets 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 this amendment.
6.0 CONCLUSION
The Commissica has concluded, based on the considerations discussed above, that:
(1) thave is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed menner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance: of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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7.0 REFERENCES
l 1.
Letter from W. L. Stewart, Virginia Electric Power Company, to NRC,
" Proposed Technical Specifications Changes Reduction in Minimum Measured RCS Flow Rate," dated December 4, 1993.
2.
Letter from G. S. Lainas, NRC, to W.R. Cartwright, Virginia Electric Power Company, Surry Units 1 and 2 and North Anna Units 1 and 2,
" Qualification of WRB-1 CHF Correlation in the Virginia Power COBRA Code," dated July 25, 1989.
1 3.
Letter from W. L. Stewart to NRC, " North Anna Power Station Unit 1,"
Supplemental Information Regarding Proposed Technical Specification Change for Reduced Minimum RCS Flow Rate Limit," dated January 31, 1992.
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4.
" North Anna Power Station Units 1 and 2 - Implementation of Extended SGTP Small Break LOCA Analysis," dated January 21, 1992.
5.
Letter from W. L. Stewart to NRC, " Report of ECCS Evaluation Model Changes and 30 Day Report Per Requirements of 10 CFR 50.49 North Anna Power Station Units 1 and 2" dated July 16, 1993.
Principal Contributor:
S. Brewer Date: August 30, 1993 l
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