ML20137C879
| ML20137C879 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/21/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20137C873 | List: |
| References | |
| NUDOCS 9703250155 | |
| Download: ML20137C879 (5) | |
Text
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i UNITED STATES s
j NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 30se6 0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMEN 0 MENT NO. 204 TO FACILITY OPERATING LICENSE DPR-57 AND AMEN 0 MENT NO.145 TO FACILITY OPERATING LICENSE NPF-5 i
EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By "HL-5230, Application for Amends to Licenses DPR-57 & NPF-5,revising SRs 3.1.7.7 & 3.4.3.1 & [[TS" contains a listed "[" character as part of the property label and has therefore been classified as invalid.,3.5.1 & 3.6.1.6 to Increase Nominal Mechanical Pressure Relief Setpoints for All SRVs & Allow Operation W/One SRV|letter dated October 7,1996]], Georgia Power Company, et al. (the licensee or GPC), proposed license amendments to change the Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant, Units 1 and 2.
The proposed changes would allow the units to be operated with one Safety / Relief Valve (SRV) out of service (Reference 1), and increased mechanical lifting setpoints from between 1120 and 1140 psi to 1150 psi. GPC had the General Electric Company (GE) perform a safety evaluation of the proposed revision and the GE report is included in the GPC submittal and is referenced as the justification for the proposed revision.
The proposed revision affects the sressure rise of the vessel during transient reactor behavior. Both of.these c1anges will tend to increase the peak pressure. The staff has evaluated the proposed revision by performing a series of audit c iculations to assess the effects of the changes. The staff
-used its TRAC-BFI evaluation model (Reference 2) with a generic BWR4 input deck. The deck design and its validation are discussed in Reference 3.
The staff modified the base deck by adding a more realistic control system model and adding a model of the steam line. Also, the SRV model was modified to model 11 valves instead of 13 for application to Hatch Units 1 and 2.
2.0 EVALUATION The licensee enclosed in its submittal (Reference 1) an analysis performed by GE of the impact of the proposed changes to the SRV setpoints and allowing one SRV to be out of service. The analysis considered a wide spectrum of events and equipment performance characteristics (High Pressure Coolant Injection, Automatic Depressurization System, etc.). The limiting Anticipated Operational Occurrence (the Turbine Trip without Bypass) was evaluated with the approved ODYN code (Reference 4). The analysis demonstrates that a margin of more than 50 psi exists to the American Society of Mechanical Engineers (ASME) Code limit of 1375 psi.
Finally, GE considered the effect of the 9703250155 970321 DR ADOCK 0500 1
., modification on the Loss-of-Coolant Accident and the plant's Anticipated Transient without SCRAM (ATWS) mitigation capability. GE concluded that the effect of the changes on the LOCA results would be minimal and that the ATWS response remains within acceptable limits.
As part of. its evaluation, the staff performed the following audit calculations:
1.
The Main Steam Isolation Valve (MSIV) Closure Anticipated Transient Without SCRAM (ATWS) using original SRV setpoints.
l 2.. MSIV Closure ATWS assuming all SRV's open at the relief setpoint of 1150 psi plus a 3% tolerance to account for valve drift.
l 3.
Same as 2 above except that one SRV is assumed out of service.
)
4.
A turbine trip without bypass transient with all valves at the relief l
setpoint of 1150 psi plus a 3% tolerance to account for valve drift.
l The recirculation pump trip is disabled.
5.
Same as 4 above except that one SRV is assumed out of service.
l These calculations provide the limiting pressure rise. The acceptance l
criterion for cases 1 through 3 is that the peak vessel pressure should remain below 1500 psi.
For cases 4 and 5, the acceptance criterion is that the peak vessel pressure should remain below 1325 psi (the 1375 psi ASME code limit minus a 50 psi margin). The use of the relief setpoints plus a 3% tolerance and, for the case of the turbine trip analyses, assuming a malfunction of the recirculation pump trip, are conservative assumptions because they increase the calculated peak pressure.
The licensee stated in its submittal that the purpose of the amendment is to l
reduce the potential for pilot valve leakage and the potential for forced outages due to an inoperable SRV during a fuel cycle. The staff concludes that by allowing one SRV to be out of service, the likelihood of having to shut down due to SRV failures will be reduced.
Furthermore, the proposed modification increases the margin between the mechanical relief setpoints and the. dome pressure to a value consistent with other units and should reduce the likelihood of pilot valve leakage.
The results of the staff's evaluation show that all the applicable acceptance criteria are met with considerable margin. The staff's audit calculation results are less conservative than the licensee's results, but this is consistent with the best-estimate nature of the staff's model.
3.0 TS CHANGES i
The following discussion of changes to the Limiting Conditions of Operation j
(LCO) and Surveillance Requirements (SRs) applies to both Hatch Units 1 and 2.
~
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! LC0 3.4.3 is changed to read that "10 out of 11" SRVs shall be OPERABLE l
from "10."
This change is acceptable based upon both the licensee's calculations, and the staff's audit calculations, which demonstrate that all of the applicable limits described above continue to be met.
l SR 3.'4.3.1 is changed to read that the nominal mechanical trip setpoints for all of the 11 SRV's are 1150 psig i 34.5 psi. This change is acceptable based upon both the licensee's calculations and the staff's audit calculations, which demonstrate that all of the applicable limits continue to be met. This also has an effect on the performance of l-Emergency Core Cooling Systems (ECCS) and the Standby Liquid Control System (SLCS). The licensee stated that the SLCS pumps are positive displacement pumps and the increase in dome pressure is within their capacity. The licensee's submittal states that the ECCS pumps can deliver l
the required flow at the elevated pressures associated with this chango.
This conclusion is based upon examination of the pump performance curves.
LCO 3.5.1 is changed to read that "six of seven" automatic l
depressurization system (ADS) valves shall be OPERABLE instead of "seven."
l The licensee stated that one less ADS valve will not affect the Large Break Loss-of-Coolant Accident (LBLOCA) results for Hatch because the l
plant depressurizes very fast. The staff agrees with the licensee's l
statement. GE studied the effect of one less ADS valve on the Small Break Loss-of-Coolant accident (SBLOCA) results for Hatch and concluded that the effect is minimal and the Large Break results remain limiting (with calculated LBLOCA Peak Clad Temperature (PCT) 400 degrees centigrade l
higher than SBLOCA results). This is due to the fact that during SBLOCA events, heat-up rates are very small (on the order of several degrees per second), and increasing the depressurization time by allowing one SRV to be out of service will not raise the limiting SBLOCA PCT by 400 degrees centigrade. The staff agrees with this conclusion.
LCO 3.6.1.6 is changed to state that "three of four" Lower Level Setpoint (LLS) valves shall be operable instead of "four." The LLS system is intended to reduce the probability of valve failures by reducing the number of valves that cycle. The LLS system consists of 4 SRVs designated
'to function in the LLS mode. LLS logic is armed following one full cycle of SRVs in the normal mode of operation. The reduction of the number of l
LLS valves could impact the probability of an SRV sticking open because the number of SRVs forced to open to maintain the system pressure could increase. The probability of an SRV sticking open is directly 1roportional to the number of valves open. However, this is acceptable acause control of the reactor if an SRV sticks open is provided for by l
the plant's Emergency Operating Procedures.
SR 3.1.7.7 is changed to reflect the fact that the SLCS pumps have to deliver their flow at a higher pressure (1232 psi). This change is l
consistent with the discussion of the changes to SR 3.4.3.1 and is, L
therefore, acceptable.
e o r 1 Based on its review of the licensee's submittal, the staff concludes that the proposed changes to Hatch Units 1 and 2 TS are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIROMENTAL CONSIDERATION The amendments change a requirement 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 l
consideration, and there has been no public comment on such finding (62 FR 129 dated January 2, 1997). Accordingly, the amendments meet the i
eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)t9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or l
environmental assessment need be prepared in connection with the issuance of l
the amendments.
6.0 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 Contributor:
A. Ulses Dated: March 21, 1997 i
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i RREFERENCES 1.
Letter from J. D. Woodward (GPC) to USNRC, " Request to Revise Technical l
Specifications: Safety / Relief Valve Setpoint Change," October 7, 1996, and attachments.
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2.
Borkowski, J. A., TRAC-BF1: An Advanced Best Estimate Computer Program for BWR Accident Analysis, NUREG/CR-4356, August 1992.
3.
Letter from David B. Matthews (USNRC) to K. P. Donovan (BWROG),
4
" ACCEPTANCE OF PROPOSED MODIFICATIONS TO THE BOILING WATER REACTOR (BWR)
EMERGENCY PROCEDURE GUIDELINES (TAC NOS. M89489 AND M89629)," June 6, 4
i 1996.
i 4.
Letter R. L. Tedesco (NRC) to G. G. Sherwood (EENE), " ACCEPTANCE FOR l
REFERENCING GENERAL ELECTRIC LICENSING TOPICAL REPORT NED0-24154/NEDE-l 24154P,"
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