ML20059D578
| ML20059D578 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/30/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059D573 | List: |
| References | |
| NUDOCS 9009070107 | |
| Download: ML20059D578 (7) | |
Text
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, NUCLEAR REGULATORY COMMISSION t
W ASHINGTON, D. C. 20555 S I> v [. 8
\\....+f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.
34 TO FACILITY OPERATING LICENSE NPF-68 u
AND AMENDMENT NO. 14 TO FACILITY OPERATING LICENSE NPF41
-GEORGIA POWER COMPANY. ET AL.
DOCKETS NOS. 50-424 AND 50-425 V0GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2
1.0 INTRODUCTION
By letter dated May 29, 1990, GeorgiaPowerCompany(GPC)',et'a1.(thelicensee),
requested a Technical Specification (TS) amendment to revise the steam generator (SG) level instrumentation setpcints described in TS Tables 2.2-1 and 3.3-3.
Aaditional information regarding this TS' amendment request was provided on July 6, 1990. The TS amendment request stems f rom a proposed modification that
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will relocate the level of the SG' lower tap from 438 inches elevatioti to 333; inches, as neasured from the top of the SG tubesheet, increasing-the water level span from 128 inches to 233 inches.r.The wider instrumentation span changes the low-low, high-high and nominal operating level settings and enables
~GPC to increase the operating-margin to the low-low and high-high setpoints.
The low-low to nominal level increases by_about 20 inches and the high-high to nominal increases by about 9 inches. The proposed change should reduce the-nurther of spuricus reactor and turbine trips from the low-low and high-high.
level trips.
2.0 EVALUATION Because the TS amendment request and the associated plant modification affects the high-high and the _ low-low SG 1evel trip setpoints and the nominal: water level, many loss of coolant accident (LOCA) and non-LOCA transient analyses describedin-Chapter 15oftheVogtleElectricGeneratingPlant:(VEGP) Updated Final Safety Analysis Report (FSAR) are also potentially affected.
In the o
L following,-each category is examined separately with a brief description of the result and justification.
2.1 Non-LOCA' Evaluations Mass / energy release from secondary system pipe rupture inside containment (FSAR6.2.1.4)
The objective.is to assure that the peak containment pressure limit is not exceeded. The peak pressure depends (among other parameters) on SG inventory.
However, the revised inventory is estimated to be slightly lower than the current value, thus, no reanalysis is required.
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Mass / energy release from secondary systam pipe rupture outside containment (FSAR 6.2.1.4) l The objective is to assure that no adverse environmental conditions are created for equipment qualification. The revised mass / energy release calculations were' reviewed and the limiting temperature _ profile was not exceeded.
Feedwater system malfunctions which result in decreased feedwater temperature (FSAR15.1.1)
This transient is bounded by the "feedwater system malfunctions that result in increasedfeedwaterflow"(15.1.2),thusnoreanalysisisrequired.
Feedwater system malfunction that results in an increase in feedwater flow (zero power) (FSAR 15.1.2)
The existing analysn. demonstrates that the reactivity input is bounded by the
" uncontrolled rod cluster control assembly bank withdrawal from a suberitical or low-power startup condition" (15.4.1), thus, no reanalysis is required.
The following transients do not need a reanalysis because the SG trip functions are not challenged or are not modelec, thus not affected by the proposed modifications.
Excessive increase in secondary steam flow (FSAR 15.1.3)
Steam systen: piping failure (FSAR 15.1.5)-
Steam pressure regulator malfunction or failure-that results in decreasing steam flow (FSAR 15.2.1)
Loss of external electrical. load (FSAR 15.2.2)
Turbinetrip(FSAR(15.2.3)
Inadvertentclosureofmainsteamisolationvalves(FSAR15.2.4)
Loss of condenser vacuum and other events resulting in turbine trip (FSAR15.2.5)
Partiallossofforcedreactorcoolantflow(FSAR15.3.1)
Con.plete loss of forced reector coolant flow (FSAR 15.3.2)
Reactor coolant pump shaft seizure and locked rotor (FSAR 15.3.3)
Reactor. coolant pump shaf t break (FSAR 15.3.4)
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.c 3-Uncontrolled rod cluster control assembly bank withdrawal from a subcritic61 or. low-power startup condition (FSAR 15.4.1)
Uncontrolled rod cluster control assembly b6nk withdrawal at power (FSAR15.4.2)
Rod cluster control assembly misalignment-(FSAR 15.4.3)
Startup of an inactive reactor coolsnt pump (FSAR 15.4.4)
Chemical and volume control systeni malfunction which results in a decrease in the boron concentration in the reactor coolant (FSAR 15.4.6)'
Inadvertent loading and operation of_ a fuel assembly in aniimproper position (FSAR15.4.7)
Spectrum of rod cluster control assenibly ejection accidents (FSAR 15.4.8)
Inadvertent operation of emergency core cooling system during power operation (FSAR15.5.1)
Chemical and volume control.ystem malfunction that increases reactor coolant-inventory (FSAR 15.5.2)
Inadvertent opening of a pressurizer safety relief valve (FSAR 15.6.1) 2.1.1 Non-LOCA Transients. Requiring Reanalysis The following transients required reanalysis since the requested TS change and-associated plant iredificatior. impact the_ initial assumptions for the low-low, high-high and noninal water level for these analyses.
l Feedwater systen malfunctions which result in an increase in feedwater L
flow (full. power)(FSAR15.1.2) lhe objective of this analysis is to demonstrate that the MDNBR does not f all below the limiting value. -Using the same FSAR assumption (i.e., high-hish level trip setpoint at 100 percent of the narrow range span) it was estimated that the NDilBR does not fall below its limiting value.
Loss of non-emergency AC-power to the station auxiliaries / loss of normal feedwater (FSAR 15.2.6 and 15.2.7)
The objective in this analysis is to demonstrate that neither the prinary nor, the secondary systems will overpressurize following reactor trip, i.e., there is adequate heat reinoval capability.
The assumptions include:
510 gpm 1.
auxiliary ftedwater at 103*F, pressurizer spray and relief valves are assumed operable, the SG low-low level setpoint is at 30 percent, and 10 percent of the 50 tubes are plugged. The results demonstrate that the capacity of the auxiiiary -
' feedwater is sufficient to prevent overpressurization.
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4 Feedwater system pipe break (FSAR 15.2.8)-
The objectives of this analysis are the same as above. The assumptions, in-addition to those listed above, include:
power level of 3636 MWT, SG le low water level at 16 percent of the narrow range. span and a conservai.ae Leay heat model. The results demonstrate that the ' auxiliary feedwster <s adequate to remove decay heat and prevent overpressurization and for the loss of offsite.
power, natural circulation is sufficient to prevent secondary overpressurization l
and fuel cud damage.
2.2 LOCA Evaluation t
2.2.1 Large Break LOCA (FSAR-15.6.5)
The secondary trip setpoints are not modeled in.the large break LOCA analysis, thus, a change in the SG level tap, associated trip setpoints,'and initial I
water level will not have any affect on the large break LOCA.
2.2.2 Small Break LOCA (FSAR 15.6.5)
As in the large break LOCA, the trip setpoints are not modeled,- thus, a change-in the trip setpoint does not affect the 'small break LOCA analysis. However, the initial SG water level could affect the pressurizer low pressure trip which is modeled in the Westinghouse WFLASH (Refs. 3' and 4) Vogtle FSAR analysis. A conservative decrease of 10 inches.in the nominal SG level was investigated, and existing sensitivity studies showed an 11*F penalty in t h peak cladding-temperature. The peak cladding temperature for both units remains well, below I
the 10 CFR 50.46 limit of 2200*F.
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2.2.3 LOCA Blowdown Forces, Boron Precipitation'Due to Hot Leg Switchover and Post-LOCA Long Term tooling The following parts of LOCA related analyses ao not depend onithe SG' level trips nor the initial SG level, thus are~ not effected by the' proposed modifications:
The hydraulic forcing functions on the reactor vessel, vessel internals and the RCS loop piping. The peak forces occur well before the generation of any SG 1evel trip signals therefore, the proposed changes do not affect the hydraulic forcing functions.
The calculation of hot leg switchover time (ensures no boron precipitation following core boiling) depends primarily on power level but is independent of the proposed modifications.
The post-LOCA long tern cooling calculation _(to demonstrate the reactor remains shutdown) is independent of the assurrptions of the ~ secondary system, thus, the proposed modifications do not affect this calculation.
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l l r 2.2.4 Roo Ejection Mass Release The rod ejection mass release transient is similar to the small break LOCA transient. The small break LOCA sensitivity analysis showeo that the rod ejection is bounded by the small LOCA for the proposed SG tap modifications.
2.3 Steam Generator Tube Rupture The proposed modification to narrow range level indication increases the instrument span, thus; a SG tube rupture will be more easily detected at an earlier time in -%e transient, ultimately resulting-in an earlier-termination j
and reduction o'. off-site radiation doses.
2.4 Radiologice.i Consequences l
From the preceding discussion, we have seen that the proposed SG water level ucdifications do not affect primary mass releases and consequently do not i
affect off-site radiological doses. The FSAR findings remain valid.
1 2.5 Containment Integrity The proposed redifications do not affect the limiting conditions'in the containment anaiyses, thus, the FSAR conclusions remain valid.
2.6 Conclusions n
Review of the impact of the proposed changes an non-LOCA and LOCA accident J
analyses indicates that either the current analysis nf record remains bounding and reunalysis is not required, or that the Standaro Review Plan requirements are satisfied for those analyses reanalyzed.- In addition, the SG tube rupture, the off-site radiological consequences and the effect on containment' integrity have been reviewed with the conclusion that the FSAR findings remain valid.
1hus, we find the proposed TS changes resulting.from the moeffication to lower the SG level tap to be acceptable.
3.0 ENVik0hMENTAL CONSIDEPl. TION The amendu.ents involve changes-in requirements with. respect to the installation or use of facility components located within the restricted area as defined in
'0 CFR Part 20. The staff has cetermined that the amer.dments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be.releasea 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. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuantto10CFR51.22(b),
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no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
4.0 CONCLUSION
The Commission's proposed determination that the amendments involve no significant hazards consideration was published in the Federal Register
.i on July 25, 1990 (55 FR 30299). No public comments were received, and the-State of Georgia did not have any comments.
The staff has concluded, based on the considerations discussed above, tha't: (1)'
there is reasonable assurance that the health and safet endangered by operation in the proposed manner, and (2)y of the public will not be such activities will be conducted in compliance with the Consnission's regulations, and.the issuance of these amendments will net be inimical to the common cefense and security or to the.
i health and safety of the public.
i Principal Contributor:
L. Lois, SRXB/ DST Lated:
August 30, 1990
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