ML20107B939
| ML20107B939 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/12/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20107B912 | List: |
| References | |
| NUDOCS 9604170150 | |
| Download: ML20107B939 (8) | |
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1 UNITED STATES
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- NUCLEAR REGULATORY COMMISSION
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WASNINGTON, D.C. 30086 4001 SAFETY EVALUATION'BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENOMENT NO. 133 TO FACILITY OPERATING LICENSE NO. NPF-12 SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY' VIRGIL C. SUMER NUCLEAR STATION. UNIT NO. 1 DOCKET NO. 50-395
1.0 INTRODUCTION
i By letter dated August 18, 1995, as supplemented.on November 1, 1995, and j
February 14, March 14.(there are.two supplemental letters with this. date),
j and March 25, 1996, (hereafter,' collectively referred to as power uprate.
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submittal) South Carolina Electric & Gas Company (the licensee) requested J
changes:to'the Facility Operating License (FOL) and. Technical Specifications (TS) for the Virgil C. Summer Nuclear Station, Unit 1 (VCSNS). The proposed amendment would revise the FOL and TS to increase allowed core power leve1 ~
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from' 2775 Megawatts _ thermal (MWt) to 2900 MWt.
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The original Federal Reaister notice included information in the licensee's j
i November 1, 1995 supplemental letter. The~ February 14, March 14,fand.
J March 25, 1996 supplemental letters provided clarification and amplificat3cn of the analysis in the November 1,1995 letter and were not outside th'e scope of the original Federal Reaister notice.
2.0- BACKGROUND i
' License Amendment No. 119, issued November 18, 1994, implemented changes to
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support VCSNS operation with replacement steam generators.- The majority of the supporting analyses-for the steam generator replacement were performed at the proposed core uprate power level-of 2900 MWt. Also, several'TS changes necessary for power uprate were approved in Amendment No. 119.
This safety l
evaluation (SE) covers the power uprate issues that were not addressed in the staff's SE supporting Amendment No. 119. The FOL and TS changes requested by the licensee in their power uprate submittal are:
FOL Paragraph 2.C.1 - Revise maximum power level to 2900 MWt core power.
TS Definition 1.25 - Revise Rated Thermal Powcr definition.to incorporate the -
increased power level.
.o TS Figures 3.4-2.and 3.4 Revise applicability from 14 effective full power years (EFPY) to 13 EFPY due to increased neutron fluence effect.
ENCLOSURE.
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'9604170150 960412 PDR ADOCK 05000395 P
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. TS.3'11.2.6 - Revise maximum quantity of radioactivity in each gas storage tank from 160,000 curies to 131,000 curies of Noble gas in. order to reference the current large break loss-of-coolant accident analysis.
TS 6.9.1.ll.c Revise methodology referenced by the core operating limits report that is used to determine the heat flux hot channel factor.
3.0 EVALUATION 3.1 Ucrate Issues Evaluated for Amendment No.119 The following table lists items previously evaluated in Amendment No.119 and found acceptable at the uprated power level of 2900 MWt. These items will not-be reevaluated for this amendment.
Evaluation SE Section.
Primary Components and Piping Support Considerations 2.2 Leak-Before-Break 2.2
. Nuclear. Steam Supply System Design Transients 2.3 Protection System Setpoints 2.3
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Small Break Loss-of-Coolant Accident (LOCA) 2.3-Post LOCA Long Term Core Cooling Subcriticality
'2.3 Hot Leg Switchover 2.3 Containment Considerations 2.4 Equipment Qualification Inside Containment 2.4 Radiological Consequences 2.5 3.2 :Ucrate Issues Not Previous 1v Evaluated for Amendment No. 119 3.2.1 Larae Break Loss-of-coolant Accident (LBLOCA)
IIn' its power uprate submittal, 1!he licensee stated the licensing basis
. analyses have consistently shown the double-ended cold leg guillotine (DECLG) break with C,=0._4 is the' most limiting DECLG break.
Previous analyses also showed that reduced vessel average-temperature produces the most limiting-results. -Therefore,. the licensee analyzed a DECLG break with a C =0.4 and a-reduced vessel average. temperature of 572*F using the Westinghous,e 1981 Evaluation Model with BASH (WCAP-10266-P-A, Rev.2, 1987, Including Addendum 2-A,1988). This analysis has been approved by the NRC for licensing applications and is applicable to VCSNS. The calculated peak cladding
- temperature.is 2099'F, the calculated maximum local metal / water reaction is -
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17.9 percent,Eand'the calculated core-wide metal / water reaction is less than
.1 percent. These results are within the criteria,specified in 10 CFR 50.46(b)
-(1 through 3,_ respectively).of 2200*F,.17 percent, and l ~ percent. The results Jensure the' core will remain amenable to cooling, as required by 10 CFR 50.46(b)(4).
In'its submittal for Amendment No.119..the licensee stated the; time of emergency core coolant. system (ECCS) hot leg switchover was determined
'by' analysis to be within.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This, combined with the VCSNS ECCS design, assures continued conformance with the long-term cooling requirement of 10. CFR 50.46(b)(5).7 The. licensee analyzed LBLOCA using bounding assumptions with an NRC-approved methodology and concluded it met the acceptance criteria of.
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- 10 CFR 50.46. Therefore, the licensee's proposal is acceptable.. The licensee.
proposed.to revise TS 6.9.1.'ll.c to add " Including Addendum 2-A, ' BASH METHODOLOGY IMPROVEMENTS AND. RELIABILITY ENHANCEMENTS,' MAY 1988." The staff agrees that the TS Administrative Controls section should include reference to the' BASH-addendum..Therefore,~the licensee's proposed TS change is acceptable.
3.2.2TResidual Heat Removal (RHR)- System
~The licensee indicated the RHR. system still has the ability to bring the plant from hot standby to coldLshutdown (defined as less than or equal to 200*F) within the TS required time of.30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Specifically, the licensee calculated the RHR system will require 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> to cool the reactor coolant system to 140*F.with two RHR heat exchangers and two RHR pumps ~ available. The staff: agrees,the RHR system can continue to perform its intended function in the uprated condition.
3.2.3 Increased Neutron Fluence
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The' licensee indicated the increase in core power will have an associated increase In the neutron fluence which interacts with the reactor vessel. To account. for this increase in neutron fluence, the licensee has p' roposed changing the applicability of the TS heatup and cooldown curves (TS Figures 3.4-2 and;3.4-3, respectively) from 14 effective full power years (EFPY) to 13.EFPY. The staff agrees that a 1 EFPY reduction in applicability is appropriate for the proposed power uprate.
Therefore,-the licensee's requested TS changes are acceptable.
3.2.4' Waste-Gas Decay Tank Ruoture In the Amendment No. 119 SE, the staff-independently assessed the potential' consequences of.the release of the contents of a waste gas decay tank. ' The
- gaffconcludedthelicensee'sassumedreleaseof. 160,000 curies-(Ci) of:
- Xenon (Xe) wg nonconservative and should instead be approxirrately i
131,000 Ci of Xe.= Specifically, the staff stated '"While this particular issue is not associated with the replacement of the D3 steam generators, the gcensee-should reevaluate the determination of the allowable TS quantity of Xe in the waste gas. decay tank."
As requested by the staff, the licensee reevaluated this issue..The licensee determined that a change to the maximum quantity.of radioactivity that can be stored.in the Waste Gas Stora5e Decay Tank is required.
In its power uprate i
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submittal, the licensee requested to change TS 3.11.2.6 by replacing "160,000 curies" with."131,000 curies." The.. licensee's proposal is consistent with the i
staff's previous conclusion and is therefore acceptable.- It is also noted thatthelicenseegtatedVCSNShasneverexceededanadministrativelimitof j
- 90,000 curies'of Xe in a gas storage tank.
3.2.5 Soent Fuel Pool Coolina System 4
L The spent fuel pool cooling system (SFPCS) was designed to remove:the decay
. heat released from the ' stored spent fuel assemblies and maintain the spent fuel pool (SFP) water temperature at acceptable levels. - The licensee evaluated three offload scenarios as part of its uprate analysis. The j
scenarios are 1)- a partial offload with a single failure, 2) a routine -
j refueling outage full core offload, and 3) an abnormal full core offload that
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occurs 36 days after a refueling outage in which 72 fuel assemblies were j
placed in the pool.
The. following are the.SFPCS heat loads resulting from the partial and full-core offloads and their corresponding calculated peak SFP temperatures -
resulting' from plant operations at the uprate power level:
SFPCS Heat Loads Peak Pool Temperature
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i Partial Core 21.23 150.2*F (assuming a single Offloaded active failure) s.
Full Core 44.76 152.6*F-(no single failure)
Offloaded - (offload.
occurs 36 days after 72 fuel ~ assemblies were placed in SFP) 4 Full Core
- 38.21 145.7'F (no single failure)
Offloaded 186.I'F (assuming a single.
active failure)
For the full core offload 36 days after 72 fuel assemblies have been placed in 4
the SFP-(i.e., abnormal offload), the calculated peak SFP temperature is i
152.6*F'which is ** ow the guidance in Standard Review Plan (SRP) 9.1.3.
Therefore,'the si finds the licensee's proposal acceptable.
!With the partial core offload SFPCS heat load case (assuming a single active failure), the calculated peak-SFP temperature is 150.2*F. Also, du;ing routine refueling' outages, the peak temperature could reach 186.1*F if a
' single active failure occurs. These temperatures are higher than the current VCSNS SFP Updated Final-Safety Analysis Report design temperature and the SRP 9.1.3 guidance-of 140*F for the SFP temperature limit during normal operating conditions. To address these. higher temperatures, the' licensee performed n
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evaluations of the:-
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SFPCS pipe; stresses-1
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SFPCS componentsi i
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Margin to localized boiling c6. ' Adequacy of. net positive suction head available for the-SFFCS pumps.
.i Based on'its evaluation, the licensee concluded the above peak temperatures, l
were acceptable.
The licensee provided its bases.for,this conclusion in a' March 14,'1996? supplemental letter. The licensee also stated that-suffit.ient-time exists to restore the SFPCS or provide make-up water to prevent'the spent fuel from being uncovered if boiling were to occur.
Based on the information I
provided by the licensee,.the staff finds:the licensce's proposal regarding
. normal SFP operations complies with General Design Criterion 61-Fuel storage
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'and handling and radioactivity control, and 1s therefore accentabie.
Based on1ourfreview,.thelicensee's evaluations listed above', and the i
experience gained from our review of power uprate applications for similar
. pressurized water reactor'(PWR) plants, we conclude that VCSNS operations at the proposed uprated power level is acceptable.
In'a related matter, an. issue: associated with spent fue1~ pool cooling. adequacy.
was identified in.NRC.Information Notice 93-83 and its Supplement 1,-
'" Potential Loss of' Spent Fuel Pool: Cooling Following a ' Loss of Coolant Accident (LOCA),"' dated October 7,s1993 and August'24,n1995, respectively, and.
.. inia 10 CFR Part 21 notification, dated November 27,' 1992.- The staff is
- evaluating this-issue,.as well as broader issues associated with. spent fuel:
storage safety, 'as part of:the NRC generic issue evaluation process. 'If the generic' review concludes that additional requirements in the area of spen'
. fuel pool safety are warranted, the staff will address those requirements to the license'under separate cover..:
3.2.6' Comoonent Coolina Water System The compon'ent cooling water system (CCWS) provides cooling water to various
. safety and non-safety systems during all phases'of normal plant operation, including startup:through cold shutdown and refueling, as well as following a Jstetton blackout event, loss-of-coolant accident (LOCA) or main. steam line.
break' accident..The CCWS is a closed loop system which serves as an intermediate barrier between the service water system and systems which contain radioactive' or potentially radioactive fluids in order to eliminate
'the. possibility of an uncontrolled release of radioactivity. - The licensee stated that the CCWS heat loads'resulting from 21 ant operations at the proposed uprated-power level will increase slig1tly. = The increases-in heat loads are'from the SFPCS during both power and refueling operations, and residual heat removal-(RHR) system during plant shutdown.
The licensee performed evaluations of the effects 'of those increases in heat loads' on CCWS and_ concluded that the CCWS has adequate cap city to accapt the minimal li
- increases from'SFPCS and CCWS heat loads.
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Based on our review ~and the' experience gained from our review of power uprate
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- applications for.similar PMR plants, the staff finds that plant operations at the proposed uprated powe" level do not change the design aspects and.
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. operations of the CCWS. ~Therefore, the staff concludes that VCSNS-operations at the proposed uprated power level is acceptable.
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3.2.7 Service Water System The service water system (SWS)'is designed to supply cooling water to various non-safety related components and heat exchangers in the turbine, reactor, and l
radwaste' buildings during normal plant operation, and to supply cooling water to safety related systems and other essential equipment during a station blackout event and a LOCA-or main steam line break accident.
Based on its-performed evaluations, the licensee stated that.the 'SWS as designed will
. supply sufficient water to remove the additional' heat loads resulting from plant operations at the proposed uprated power level.
Based on our review and_the experience gained from our review of power uprate ~
applications.for similar PWR plants, the staff finds that plant operations at s
the proposed uprated power level do not change the design aspects and 1
.' operations of the SWS. Therefore, the staff concludes that VCSNS operations!
j at the. proposed uprated power level is acceptable.
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.3'.2.8 Main Steam System-
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a Thelicenseestated--that-themainsteamsystemanditsassociatedcomponen)s 1
(e.g. main steam isolation valves, turbine steam bypass. system, etc.) were evaluated for a' reactor power level'of 2912 MWt and that VCSNS operations'i.t the proposed uprated power level have an. insignificant ~or no impact on the I
-main steam system and its associated compoaents.
j Based on our review and the experience gained from our review of power uprate
~ applications for similar PWR plants, the staff finds that plant operations at the proposed uprated power level do not change the design aspects and operations'of the main steam system. Therefore, the staff concurs with the
' licensee that VCSNS operations at the proposed uprated power level'is 1
acceptable.
'3.2.9 Condansate and Feedwater Systems The. licensee evaluate'd the condensate and feedwater systems for the plant-operations at 2912 MWt reactor power level to support the above cited
. replacement steam. generator TS change request and concluded that these systems satisfy their design bases for plant operations at the proposed uprated power level.
Since these systems do not perform any safety related function, the staff. has not reviewed the impact of plant operations at the proposed uprated power level lon the design and performance of these systems.
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i 13.2.10.
>circulatina Water / Main and Auxiliary Condensers / Turbine Auxiliar',v
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Systems 6
The~ circulating water, main and auxiliary condensers', and turbine auxiliary j
systems-are designed to remove'the heat rejected to the condenser by turbine exhaust.and other exhausts over the full. range of operating loads, thereby 1
-maintaining adequately low condenser pressure. The licensee stated that.
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-performance of these systems were evaluated for power uprate and determined
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- these systems are adequate for uprated power. level operation. Also, in order to solve the' problem of corrosion and fouling, the osen cycle cooling
~ i system for cooling various turbine auxiliary systems will
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- closed system coo'ed with a' modular forecd draft cooling tower..This j
' modification will solve the fouling problem, enhance performance, increase
' reliability and take a heat load off. the circulating water system.
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' Since the circulating water,. main and auxiliary. condensers, and turbine b
auxiliary systems do not perform any safety function, the staff has not
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reviewed, the impact of.the.uprated power level operation on the designs and
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performance of these~ systems.--
5" 3.2.11 Turbine-Generator-1
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.The licensee stated that evaluations for' turbine operations with respect to~
design' acceptance criteria ~to verify the mechanical integrity under the 4
cconditions imposed by the powerj uprate were performed.
Results of:the
' evaluations showed that there would be no increase in'the probability of turbine overspeed nor associated turbine missile: production.due to plant.'
. operations at the. proposed uprated power level.
Therefore, the licensee ~
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- concluded that' the turbine could continue to be operated safely at the proposed uprated power levels.
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. Based on our review,-the staff agrees with the licensee that operation ~of the
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iturbine at,the proposed uprated. power level is acceptable..
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'3.2.12. Systems Not Affected By Power Unrate 4
The licensee stated that various systems were evaluated and determined that /
-these systems were not affected by the power uprate. Those systems were evaluated for respective capacities, heat removal. capabilities, and in many cases no direct connection to plant uprate was found.. The following are. major plant systems 1that were not affected by power uprate:
auxiliary steam, l
condenser air removal.. emergency diesel generators.and auxiliaries, solid.and l liquid waste, fire' service, station / instrument air, reactor building cooling,
-generator gas ~and vent,'non-nuclear drains, plant waste, reactor building j
spray, and heating, ventilation and air-conditioning systems.
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- Since' plant operations at the proposed uprated power level do not change the
- design aspects and operations of these' systems and from the experience gained from our review of power uprate applications for other plants, the staff
- concludes 1that plant; operations at the proposed uprated power level is J
acceptable.
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~3.2.13 ~Eaui-t' Qualification Outside Containment In a letteridated October 17, 1994, the licensee stated that impacts on-l-
. environmental. conditions (inside and outside containment).due to high energy, 11ne' breaks were reconciled to ensure applicable equipment qualification t
3 srequirements continue to be met. -The licensee-also outlined the process of.
i ensuring environmental ~ qualification of equipment afterl replacement' steam
. generators. Thisil994 letter was part of the licensee's submittal for steam
. generatorf replacement; -This issue was' evaluated in this SE because the staff did not review this aspect of the licensee's submittal for Amendment'No.119 Based on our review, the staff concludes i. hat safety-related equipment outside'
. the containment will be qualified to operate..in an accident environment resulting from plant operations at the proposed uprated power level.-
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4.0 STATE CONSULTATION
.u In accordance with the Commission's regulations, the State of. South Carolina i
official was notified of. the proposed issuance of the amendment. The State-official had no comments.
5 5.0. E![VIRONMENTAL CONSIDERATION LPursu'antLto 10'CFR 51.21, 51.32, and 51.35, an environmental assessment and~
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! finding of no significant impact was published in the Federal Register on April 12, 1996 (611FR.16272).. In this finding, the~ Commission determined that issuance.of. these amendments would not have a significant effect on the s
quality of. the human environment.'
E 6'01 CONCLUSION lThe Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and-safety of the-j public'will not be endangered by operation in the proposed manner, (2) such 3 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 i
defense and security or to the health and-safety of the public.
Principal Contributors:.D.'Shum and S. Dembek Date: Apri1412, 1996
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