ML20216D513

Safety Evaluation Supporting Amend 190 to License NPF-6

ML20216D513
Person / Time
Site:	Arkansas Nuclear
Issue date:	03/12/1998
From:	NRC (Affiliation Not Assigned)
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ML20216D511	List: ML20216D506 ML20216D513
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NUDOCS 9803170090
Download: ML20216D513 (12)

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UNITED STATES

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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066HOO1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOl90TO j

i FACILITY OPERATING LICENSE NO. NPF-6 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT NO. 2_

DOCKET NO. 50-361

1.0 INTRODUCTION

By letter dated September 23,1997, as supplemented by the letters dated February 27 and March 4,1998, Entergy Operations, Inc. (the licensee) submitted a request for changes to the Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical Specifications (TS). Ths requested changes would reduce the minimum required reactor coolant system flow rate in TS 3.2.5 until the ANO-2 steam generators are replaced. The reduced reactor coolant system flow requirement will account for plugging of up to approximately 30 percent of the tubes in the existing steam generators at ANO-2.

The letters dated February 27 and March 4,1998, provided clarifying infom1ation that did not change the initial proposed no significant hazards consideration determination.

2.0 BACKGROUND

ANO-2 has an active damage mechanism affecting the steam generator tubing which requires the repair or the removal of tubes from service when they meet the repair criteria. The unit entered a mid-cycle outage in February 1998 in order to perform inspections of the steam generator tubes and perform plugging of those tubes found to heve exceeded the established plugging criteria. A reduction in the heat transfer surface area occurs for each plugged steam generator tube and requires an increased diEerential temperature acmss those tubes remaining in service in order to support continued operation at the rated thermal power of the reactor core.

The increased differential temperature is achieved by reducing the coolant temperature and steam pressure in the plant's power conversion systems. The plugged tube also increases the primary-side flow resistance through the steam generators which leads to a reduction in the reactor coolant system (RCS) flow available for core cooling. The reduced RCS flow and heat transfer surface areas in the steam generators result in changes to several primary and secondary parameters that affect the plant's response to design basis transients and accidents.

In anticipation of the effects of increased steam generator tube plugging, the licensee submitted 2 requests (both dated September 23,1997) for changes to the ANO-2 TS. The first proposed TS amendment requested NRC review and approval of a reduction in the steam geneistor low pressure setpoints in the Plant Protection System (PPS). The affected PPS functions initiate a reactor trip, main steam ano feedwater isolations, and emergency feedwater actuation. The 9803170090 980312

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4 proposed change to the steam generator low pressure setpoints is addressed in a separate TS amendment and associated safety evaluation. The second submittal that requested a TS

- amendment to support increased steam generator tube plugging involves a proposed reduction in the minimum required reactor coolant system flow rate from 120.4 million Ibm /hr to 108.4 million Ibm /hr. The requested duration of this amendment request is from the present until after the steam generators at ANO-2 are replaced (currently scheduled for an outage in the fall of 2000).

Given that the 2 amendment requests dated September 23,1997, are closely related and that the two parameters (RCS flow rate and steam generator low pressure setpoints) affect many of the same design basis transients and accidents, the evaluation provided below discusses both parameters. The separate TS amendment and safety evaluation for the reduction in the steam generator low pressure setpoint contains applicable parts of the following evaluation.

3.0 EVALUATION The licensee presented the results of analyses of the applicable SAR Chapter 6 loss-of-coolant accidents (LOCA) and Chapter 15 non-LOCA transients. The licensee identified the limiting case for each event category discussed in Chapter 6 and 15 of the SAR and evaluated the effects of reductions in the RCS flow and steam generator low pressure aetpoints on plant transients and accidents. For those cases bounded by the SAR cases, the licensee provided supporting rationale. For those cases that are affected by the steam generator tube plugging and the steam generator low pressure setpoint change, the licensee performed reanalyses and provided the results for NRC staff review.

3.1 Loss-of-Coolant Accident (LOCA) Analysis The licensee analyzed the LOCA accidents using the NRC-approved evaluation models (CEFLASH-4A (large break) and CEFLASH-4AS (small break) for system response calculations during the blowdown phase, COMPERC-Il for system behavior calculations during the refill and reflood phases, STRIKIN-Il for cladding temperature calculations for hot rods, HCROSS and PARCH for heat transfer calculations for steam cooling and COMZlRC for determination of the core-wide cladding oxidation).

3.1.1 Laroe Break Loss-of-Coolant Accident (LBLOCA) Analyses The LBLOCA analyses were performed with the core power of 2900 MWt (103% of the rated power), a peak linear bett generation rate of 13.5 kw/ft, and a reduced RCS flow of 107.8 x 10' lbm/hr (from 120 x 10'lbm/hr.) The fuel rod characteristics (gap conductance, fuel average and centerline temperatures and hot rod temperature) used in the analyses represented a rod average bumup of 40,000 MWD /MTU. In addition, the licensee assumed the hot leg temperature increased from 616.8 (used in the SAR analysis) to 622.7 *F to account for the RCS flow reduction. The LOCA case with a break size of 0.6 of the double-ended-cold-leg-guillotine (DECLG), previously identified as the limiting break that resulted in the highest peak cladding temperature (PCT), was performed. The results show a calculated PCT of 2158 'F, maximum cladding oxidation of 0.072 of the total cladding thickness, and metal-water reaction of less than 0.009g of the total amount of metalin the core.

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i As reported on August 14,1997, by ABB-CE, in accordance with 10 CFR Part 21, the 1975 l

generated curve of energy redistribution factors (ERF) as a function of pin / box peaking factor neglected to include the effects of voiding that occurs during the LOCA events. The licensee indicated that the current limiting value of the minimum pin / box factor applicable to the LOCA analyses for ANO-2 is 1.03. The ERF used in the current analyses, for pin / box factor of 1.03, l

was 0.960. With consideration of the voiding effects, the new ERF for a pin / box factor of 1.03 is 0.969. However, the licensee indicated that the pin / box factor of 1.03 is a conservative minimum applicable to the range of cycle designs..The actual ANO-2 Cycle 13 pin / box factoris 1.0632.

The corresponding ERF with inclusion of the void effects is 0.952. Since the ERF of 0.960 used in the current LBLOCA analyses is greater than the ERF of 0.952 applicable to the current fuel, the current analyaes are conservative. In addition, the licensee reduced the linear heat rate for Cycles 13 and 14 by 0.2 kw/ft, from 13.5 to 13.3 kw/ft to reserve more margin to compensate for the potential error in the ERF The licensee indicated that the reduction in the linear heat rate more than compensates for the maximum 40 'F change in PCT, and reduces the calculated PCT to 2158 *F, without consideration of the cycle specific pin / box factor.

Since the licensee used NRC-approved codes in its analysis, the values used for the input parameters are conservative, and the results show that the 10 CFR 50.46 criteria are met, the staff concludes that the LBLOCA analyses are acceptable.

3.1.2 Sma!I Break Loss-of-Coolant (LOCA) Accident The SBLOCA analyses were done assuming the core power of 2900 MWt, a maximum linear heat generation rate of 13.5 kw/ft and a radial peaking factor of 1.61. The licensee also assumed a tube plugging of 30% in each steam generator. SBLOCA analyses for various break 2

sizes (0.02,0.04,0.05, and 0.06 ft break in reactor coolant pump discharge leg) were performed. The results show that the limiting case is the 0.05 ft break. The limiting case results in a PCT of 2011 'F, maximum cladding oxidation of 0.055 of the total cladding thickness, and metal-water reaction of less than 0.0084 of the total amount of metal in the core. The limiting case does not exceed the acceptance criteria of 10 CFR 50.46 (PCT less than 2200. *F, l

maximum cladding oxidation of 0.17 of the total cladding thickness, and metal-water reaction of less than 0.01 of the total amount of metalin the core). The SBLOCA analyses were submitted in support of Amendment No.179 and were previously reviewed by the staff. The staff finds that the SBLOCA analyses are acceptable.

4 3.2 Non-LOCA Transient Analyses The staff's evaluation of the licensee's analyses of the SAR Chapter 15 non-LOCA transients is provided below.

3.2.1 Uncontrolled CEA Withdrawal from a Suberitical Condition The licensee reanalyzed this event using the CENTS code for cciculations of the system response and the CETOP code for departure from nucleate boiling ratio (DNBR) calculations.

Both codes are NRC-approved codes for licensing calculations. A steam generator tube plugging limit of 30% was modeled. To bound the maximum reactivity addition rates for CEA withdrawals near critical conditions, the licensee performed two cases with reactivity addition

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rates of 0.0002 and 0.00025 delta-rho /sec. The reanalysis assumed a conservatively small(in absolute magnitude) negative Doppler coefficient and the most positive moderator coefficient.

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-4 The results show that (1) the minimum calculated DNBRs are greater than the DNBR safety limit of 1.25, (2) the fuel centerline temperatures do not exceed the melting temperature of uranium dioxide, and (3) the calculated RCS pressures are less than 110% of the design pressure. The staff reviewed the reactivity addition rates and reactivity coefficients used in the analysis and concludes that the licensee used conservative values. The staff reviewed the calculated consequences of this event and concludes that they conform with the acceptance criteria in the standard review plan and are, therefore, acceptable.

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3.2.2 Uncontrolled CEA Withdrawals from Critical Conditions For the CEA withdrawal event from full power conditions, a sensitivity study performed by the licensee shows that the reduction in RCS flow does not have a significant impact on this transient and the existing SAR analysis remains applicable for the anticipated plant conditions following the plugging of additional steam generator tubes.

For the event from hot zero power conditions, the licensee performed a reanalysis using the NRC-approved CENTS and CETOP codes. The analysis assumed the maximum reactivity addition rate with the CEA withdrawal speed of 30 in/ minute. A conservatively small Doppler coefficient and the most positive moderator coefficient were assumed in the analysis. For conservatism, a bottom peak axial power shape was assumed for the scram reactivity model. A

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steam generator tube plugging limit of 30% was modeled. The results show that the calculated minimum DNBR is above the DNBR safety limits in the hot channel and the fuel temperatures are f

below those associated with centariine melt of the fuel. Since the codes used for the analysis are NRC-approved codes and the results show sufficient margin exists to prevent fuel damage from a control rod withdrawal event, the staff concludes that the reanalysis is acceptable.

3.4.3 CEA Droo Event For each reload cycle, the analyses calculating the required ove/ power margin (ROPM) are considered in the determination of Core Operating Limits Supervisory System (COLSS) inputr, and operating limits that assure that the DNBR safety limits will not be exceeded. The ROPM for CEA withdrawals, loss of RCS flow events, asymmetric stean; generator transient, full length CEA drops and other transients are determined to identify the most limiting value. The analysis shows that the full lengt!i CEA drop events produce relatively slow changes to the core power distribution, and are not limiting cases for determining COLSS inputs and operating limits. A specific reanalysis of the event to account for RCS flow reduction was not performed since other bounding transients (such as CEA withdrawals) were reanalyzed.

3.2.4 Boron Dilution Event As a result of additional steam generator tube plugging, both the RCS flow and steam generator volume will be reduced. The licensee evaluated the effect of reductions in the RCS flow and steam generator volume on the boron dilution anglysis presented in the SAR. The licensee's evaluation indicates that RCS flow is not a quantitative input to the analysis and thus, the reduction in RCS flow will not change the results of the boron dilution analysis. For the deboration events in Modes 3 through 6, the volume of the steam generators was conservatively not included in the dilution volume used in the SAR analyses. Therefore, the steam generator volume reduction has no impact in these operational modes. For the events in Modes 1 and 2, the total RCS volume (less the pressurizer and surge line) was included in the dilution volume.

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5-I However, boron dilution events during these modes result in a rapid reactor shutdown by the reactor protection system and the analysis results would not be changed significantly as a result of additional steam generator tube plugging. The staff finds that the licensee's evaluation is j

reasonable and acceptable.

l 3.2.5 Loss of Forced Reactor Coolant Flow Event I

This event is initiated by a loss of the power supplied to or a mechanical failure in the reactor l

coelant pumps. As a result, the core flow rate will decrease and core temperature wil' 8" crease.

j Before the reactor trip, the combination of the decreased RCS flow and increased tog ' re may violate the DNBR safety limits. The licensee reanalyzed events involving a totn is o: RCS flow. The partial loss of forced rescior coolant flow, resulting in similar loss in the DNf3R margin, is bounded by the total loss of forced reactor coolant flow. The reanalysis used the HERMITE code to calculate the system response and the CENTS code to calculate the RCS pump coastdown flow. Both computer codes are NRC-approved codes for licensing calculations. The analysis was done assuming the core power of 2900 MWt, an axi11 power shape of 0.3 and a radial peaking factor of 1.71. The licensee assumed a tube pluggh g of 30% in each steam generator. The initial core inlet temperature of 556.7 *F and RCS pressure of 2200 psia were selected to maximize DNBR degradation during a totalloss of RCS flow event. The core protection calculator (CPC) low reactor coolant pump (RCP) speed trip was credited for tripping the reactor. For conservatism, a bottom peak axial power shape was assumed for the scram reactivity model.

The results of the reanalysis show that the minimum DNBR is 1.29 which is greater than the

- minimum allowable DNBR limit of 1.25. Since NRC-approved codes are used in the analysis, i

l the values of the input parameters are conservative, and the calculated minimum DNBR is greater than the DNBR safetylimit (assuring no fuel cladding failure), the staff concludes that the reanalysis is acceptable.

l 3.2.6 Reactor Coolant Pump Rotor Seizure Event j

l To evaluate the effect of a steam generator tube plugging level of 30% on the RCP rotor seizure I

event, the licensee calculated the decreases in RCS flow from the initial state to the stable state during the transient. The NRC-approved CENTS and COAST codes were used in the calculations. The results show that the decrease in the RCS flow fraction in terms of the initial RCS flow for the 30% steam generator tube plugging case is equal to that calculated for the SAR case. Based on the sensitivity studies on thermal margin to the DNBR safety limit, the results of the same magnitude of decreases in the RCS flow fraction for the two cases assure that the l

minimum acceptable thermal margin to the DNBR limit in the SAR case remains unchanged.

l Since the licensee uses the approved codes and the evaluation showr that the SAR case I

remains conservative, the staff concludes that the evaluation is acceptable.

3.2.7 Idle Loop Startuo The Technical Specifications do not allow operations of ANO-2 at critical conditions with inoperable RCPs. Therefore, this is not a credible event and is'not reanalyzed.

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• l 3.2.8 Loss of Extemal Load. Turtpine Trio. Loss of Condenser Vacuum Events l'

This category of events results in a reduction of steam flow from the steam generators to the turbine generator. The loss of steam flow results in a rapid rise in secondary system pressure and temperature and a reduction of the heat transfer rate in the steam generators, which in tum i

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causes the RCS pressure and temperature to rise. The everts are terminated by a reactor trip on high pressurizer pressure. The licensee reanalyzed a loss of load event initiated by a turbine trip without a simultaneous reactor trip (the limiting case identified in the SAR for this category of events) with the approved CENTS code. Sensitivity studies were performed to assess the l

effects of a steam generator tube plugging and reduction in RCS flow. The results indicate that l

RCS flow has a very minor impact on the results of the analysis: higher RCS flow rates resulting in slightly higher peak RCS pressures and lower RCS flow rates resulting in higher peak steam generator pressure. The licensee presented results for two limiting cases: one maximizing the l

. peak RCS pressure and one maximizing the peak steam generator pressure. Since the reanalyses, performed with NnC-approved codes, calculated a peak RCS pressure of 2683 psia and a peak steam generatoc p ossure of 1195 psia, which are less than the acceptance limits of i

110% of the design pressures, the staff concludes that the reanalyses are acceptable.

3.2.9, Loss of Feedwater Row Event A loss of normal feedwater may be caused by feedwater pump failures, valve ma'Ifunctions or loss of ac power sources. Following a loss of normal feedwater event, the steam generator water inventory decreases as a consequence of continuous steam supply to the turbine. The mismatch between the steam flow to the turbine and the feedwater leads to a reactor trip on a low steam generator level signal, which also actuates the emergency feedwater system. The licensee reanalyzed the loss of normal feedwater event using the NRC-approved CENTS code.

The licensee presented the results for the limiting case with respect to the minimum steam generator water inventory. In the analysis, a steam generator tube plugging limit of 30% was modeled. The minimum RCS pressure in the normal operating range was assumed to delay the high pressure trip and extend the transient. The most negative moderator and Doppler

- coefficients were used to increase the post-trip core power and thus produce lower steam generator inventories. The results show that the peak pressurizer pressure does not exceed 110% of the design pressure and thus demonstrate that the steam generator heat removal capability is maintained. Therefore, the staff conclud: i that the reanalysis of the event is acceptable.

3.2.10 Loss of AC Power Event A loss of ac power event will result in a simultaneous loss of load, feedwater, and forced reactor coolant flow. The licensee's sensitivity studies show that reducing RCS flow has minimal impact on the analyses for total loss of RCPs, loss of extemal load and loss of normal feedwater events.

Furthermore, the SAR analyses show that the minimum DNBR considerations for this event are bounded by the totalloss of RCS flow event. Thus, the licensee did not reanalyze this event.

The staff finds that the licensee's rationale for not reanalyzing this event is acceptable.

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2 3.2.11 increase in Steam Flow Event An excess steam demand (ESD) event is caused Dy a failure of the main steam system that results in an increase in steam flow from the steam generr. tor. In the presence of a negative moderator temperature coefficient, the event results in an increase in core power and a reduction

- in DNBR. The system response to the event is dapendent on the rate of heat transfer through the steam generators. The reduced heat transfer area resulting from steam generator tube plugging will slow down the cooling of the RCS primary system. The reduced initial RCS flow tends to increase the rate of primary cooldown for a given rate of heat transfer. During the

. transients, the CPC will trip the reactor to avoid violation of the DNBR safety limit. To assure that the CPC can accurately sense the cooldown associated with the event, the licensee performed a CPC transient filter analysis for Cycle 13. In the analysis, the limiting conditions (design minimum flow reduced by 10% and no reduction in steam generator heat transfer area) identified by the licensee's sensitivity studies, were assumed for the ESD event. The results show that the minimum acceptable thermal margin to the DNBR limit in the SAR case remains available Since the results of the existing CPC transient filter analysis verify that CPC trip functions are conservative and demonstrate that the SAR case remains the bounding case, the staff concludes that the effects of a reduction in RCS flow and a steam generator tube plugging are appropriately considered for tne ESD event.

The licensee also assessed the impact of a lower steam caerator low pressure setpoint of 620 psia (reduced from 678 psia in the current SAR analysis) on the ESD event. The event assessed l

by the licensee is an inadvertent opening of atmospheric dump valves (ADV) event, previously identified as the limiting ESD case. The licensee's assessment shows that a lower steam -

generator low pressure setpoint delays isolation of the affected steam generator with an opened ADV and results in a 10% increase in the amount of steam release compared to the SAR case.

- However, the resulting total mass is well within those considered for the main steam line break (MSLB) event. With a greater steam release, the overcooling effect of the MSLB results in limiting core conditions that bound the ESD event. Since the results of MSLB analysis (discussed in section 3.2.13) show that the minimum DNBRs are greater than the DNBR safety limit, the licensee stated, and the staff agrees, that the results of the ESD event with a lower steam generator low pressure setpoint can meet the DNBR safety limit. Therefore, the staff concludes that the effects of reduced RCS flow rate and a reduction in the steam generator lor pressure setpoint are appropriately considered for the ESD event.

3.2.12 Failure of the Reaulatina Instrumentation Since ANO-2 does not have coolant flow regulators, a malfunction in reactor coolant regulators is not a credible event and is not analyzed.

3.2.13 Steam Line Break The licensee reanalyzed the MSLB event with consideration of the effects of a reduction in RCS flow and a decrease in the steam generator low pressure setpoint to close the main steam isolation valves (MSIVs). The analysis was performed with the NRC approved codes: CENTS for calculations of the system response, ROCS /HERMITE for calculations of the reactivity

- feedback and peaking factors for hot rods, and HRISE for the DNBR calculations. The licensee used RELAP5/ MOD 3 to calculate the feedwater flow for the MSLB at hot full power conditions.

RELAPS/ MOD 3 is not an approved code for licensing calculations. At the staff's request, the

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s licensee provided the feedwater flow rates calculated with the RELAPS/ MOD 3 code and compared them with the flow rates calculated with the HSTA code, a code used in the approved MSLB analysis to support the licensing amendments for ANO-2. The comparison shows that RELAP5/ MOD 3 predicts a higher flow rate throughout the transient. The use of higher feedwate-flow rates increases the overcooling effects and is conservative. The staff has determined that the use, in the SLB analysis, of feedwater flow rates that are higher than those calculated by HSTA is conservative and is therefore acceptable. The SLB analysis is, therefore, adequate and acceptable for ANO-2. The staff notes, however, that this action does not approve the use of RELAP5/ MOD 3 computer code for this or any other licensing analysis for ANO-2. Future use of REPAP5/ MOD 3 for licensing applications should be preceded by staff review and approval of the code and the its specific application.

The licensee performed analyses for 4 double-ended guillotine MSLB ( with break sizes of 6.357 2

ft ) cases in order to determine the limiting cases for approaching the fuel design limits. The 4 cases analyzed are:

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A large MSLB during full power (HFP) conditions in combination with a single failure, loss of offsite power and a stuck CEA.

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Case 1 with offsite power available.

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A large MSLB during zero power (HZP) conditions in combination with a single failure, loss of offsite power and a stuck CEA.

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Case 3 with offsite power available.

To maximize the overcooling effect, the licensee made the following assumptions: (1) the highest actuation pressure for a safety injection actuation signal (SIAS) was assumed to delay the in.iection borated water to the RCS, (2) the cooldown of the RCS was terminated when the affected steam generator blew dry, (3) a CPC low DNBR trip was credited for the loss of offsite power cases and the setpoint was based on 96.5% of the RCP design speed, (4) a low steam generator pressure was assumed at 620 psia to trip the reactor and to actuate the steam generator low pressure signal that closed the main steam isolation valves (MSIVs), main feedwater isolation valves (MFIVs), and back-up MFIVs, (5) the most negative moderator temperature and Doppler coet(icients were used to maximize the reactivity addition resulting from the cooldown effect, (6) two emergency feedwater pumps were assumed to be available to maximize the cooling potential of the EFW system, and (8) the boron from the safety injection tanks was not credited.

For single failure considerations, the analyses assumed that for the loss of offsite ac power cases, one emergency diesel generetor (EDG) failed to start. The fa; lore of an EDG resulted in the failure of one high-pressure safety injection (HPSI) pump and the MFIVs to close. For the HFP case with ac power available, a bus fast transfer failure was identified as the worst single failure. The single failure resulted in the failure of the back-up MFIVs and a HPSI pump. For the HZP case with ac power available, a single failure of a HPSI train was assumed.

The analyses show that the HFP cases remain subcritical throughout the post trip event and that the HZP cases show a return-to-criticality that is bounded by the SAR results. The calculated DNBRs for all cases are greater than the DNBR safety limit and, thus ensure that no fuel failure

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will occur. Since the licensee used NRC-approved codes for analyses, the values used for input parameters are conservative, and the results show that the minimum calculated DNBRs are greater than the acceptable safety limit to assure fuel integrity, the staff concludes that the analyses are acceptable.

3.2.14 Feedwater Line Break The licensee performed sensitivity studies of a 10% reduction in the RCS flow and 30% steam generator tube plugging on the feedwater line break (FLB) analysis presented in the SAR. The results show that changes in initial RCS flow have minimal effects on the FLB analysis, and that the cases without assumed steam generator tube plugging result in a slightly higher peak RCS pressure. Since a minimum design RCS flow rate without steam generator tube plugging are assemed in the SAR case, the licensee's sensitivity studies demonstrate that the SAR case remains conservative for the FLB analysis.

To assess the effect of a lower steam generstar low pressure setpoint ( 620 psia) to close the MSIVs during the events, the licensee reanalyzed the feedwater line break (FLB) event with loss of ac power, which is the limiting case identified in the SAR.

The licensee performed FLB analyses for various break sizes with the approved CENTS code and identified that the break of 0.24 ft resulted in the highest peak RCS pressure. To maximize the calculated peak RCS pressure, the licensee made the following assumptions: (*) the least negative Doppler coefficient corresporiding to the BOC core was used to maximize the power increase, (2) the initial plant conditions were assumed to be during full power operation with a loss of offsite power at the time the reactor trip breakers open, (3) a conservative CEA insertion curve corresponding to the axial power shape of +0.6 ASI was assumed, (4) a steam generator low pressure signal was assumed at 620 psia to actuate the MSIVs with a closure time of 3 seconds, (5) the blowdown of saturated liquid from the affected steam generator was assumed, (6) the tolerance for the safety valves and secondary safety valves was assumed to be +3% of the setpoints, and (8) the decay heat was maximized by assuming an equilibrium core.

The initial pressure and initial steam generator inventories were selected such that the low steam generator water level trip in the intact steam generator and the high pressurizer pressure trip occurred simultaneously with the dryout of the affected steam generator. The sensitivity study showed that this assumption resulted in a maximum peak RCS pressure after the trip.

The results of the reanalyses show that the peak RCS pressure is 2730.1 psia which is less than 110% of the design pressure. Since the licensee used NRC-approved codes for the analysis, the values used for input parameters are conservative, and the results show that the peak calculated RCS pressure is within the acceptance criteria of 110% of the design pressure, the staff concludes that the reanalyses are acceptable.

3.2.15 Inadvertent Loadina of a Fuel Assembiv into the Improper Position The licensee evaluated the effect of a reducticn in RCS flow for two cases: 1) the erroneous loading of fuel pellets or fuel rods of different enrichment in a fuel assembly and 2) the erroneous placement or orientation of fuel assemblies. Since neither case uses RCS flow as an input

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parameter, the licensee stated, and the staff agrees, that a reduction in RCS flow has no effect on the event.

3.2.16 Steam Gen.p_rgtor Tube Ruoture (SGTR)

The licensee evaluated the effects of additional steam generator tube plugging on the potential radiological releases during SGTR events. The licensee indicated that the effects of reducing RCS flow and plugging of the steam generator tubes result in slightly higher hot leg temperatures for a given cold leg temperature and lower steam generator pressures. An increase in hot leg temperatures results in a greater flashing fraction for the primary system fluid entering the steam generator. Higher hot leg temperLtures also result in more energy being stored in the RCS.

Both factors slightly increase the radiological rel6ase to the environment during an SGTR event.

Lower steam generator pressures at the start of the event result in a greater break flow from the primary to the secondary systems. The licensee's evaluation shows that a 10% reduction in RCS l-flow and 30% steam generator tube plugging result in an increase in the offsite dose by up to 30% but that the resulting radiological releases during a SGTR event remain within the l

10 CFR Part 100 limits. Since the licensee has quantified the effects of reducing RCS flow and

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plugging steam generator tubes on the SGTR event and the results of analysis for radiological releases continue to meet 10 CFR Part 100 limits, the staff finds that the analysis of the SGTR i

event is acceptable.

l 3.2.17 Control Rod Election Event l

l The control rod ejection events were reanalyzed to account for a 10% reduction in the RCS design flow. The method used is consistent with the approved method described in l

CENPD 190-A, "CEA Ejection, C-E Method for Control Element Assembly Ejection Analysis."

The initial conditions examined in the reanalyses range from zero power to full power with reactivity coefficients representative of beginning of cycle and end of cycle for these power level extremes. The results of the analysis show that all cases meet the criteria for clad damage and molten centerline temperature and are bounded by the SAR cases.

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Recent experimental data showed failures of high bumup fuels at lower enthalpy than the Regulatory Guide 1.77 fuel failure enthalpy limits. However, generic analyses performed by the nuclear industry showed that the fuel failures (and radiological consequences) of rod ejection accidents remain bounded by existing licensing basis analysis. The analyses were based on conservative treatment of the experimental fuel failure data applied to existing and planned core operations with approved bumup limits (62 GWd/MTU lead fuel rod exposure) for U.S. reactors.

In addition, thele is broad agreement among the staff, the industry and intemational community that bumup-dependent degradation in the margin to low enthalpy fuel failures is likely to be i

regained by ** application of more detailed 3-dimensional analysis methods to the evaluation of the fuel response to rod ejection accidents. Therefore, the staff concludes that, although the Regulatory Guide 1.77 fuel failure enthalpy limits are not conservative, detailed generic analyses provide seasonable assurance that the radiological consequences of rod ejection accidents will not exceed design basis licensing limits for reactor cores operating within current NRC approved bumup limits (62 GWd/MTU average in the peak rod.) The staff will not approve further extension of bumup limits until additional experimental information on fuel behavior is available to demonstrate that the fuel cladding will satisfy the regulatory acceptance criteria used in the rod ejection analyses for ikensing applications.

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g 3.2.18 Closure of a Sinale Main Steam Isolation Valve (MSM The asymmetric conditions resulting from the closure of only one of the two MSIVs is similar to a MSLB event since the primary loop associated with the closed MSIV experiences a heat up due to loss of heat sink and the primary loop associated with the open MSlV experiences a cooldown due to the load increase. The licensee performed reanalyses with the approved CENTS code. A CPC asymmetric steam generator (ASG) trip setpoint of 11 #F was assumed. The initial power was assumed at 90% of the design power that was reduced from 103% of the power assumed for the GAR case. The reduced initial power extends the duration of the transient by slowing the rate of increase of the cold leg temperature differences between the affected and intact steam generator and extends the time to reach the CPC ASG trip setpoint. The assumption is conservative because it allows more time for the distortion of the power distribution, producing a larger relative difference in DNBRs from the start of the event to the time of minimum DNBR and resulting in a larger ROPM. Since the licensee used NRC-approved methods, the values used j

for input parcmeters are conservative, and the results show that the ASG trip setpoint incorporated in the CPCs ensures that the DNBR safety limits are not exceeded in a steam generator transient, the staff concludes that the reanalyses are acceptable.

3,3 Conformance to SER Conditions As a result of the findings from the NRC's Maine Yankee Lessons Leamed Task Force, the staff requested that the licensee verify conformance to conditions stated in NRC safety evaluation reports (SERs) associated with the topical reports (TRs) reviewed and approved by the NRC staff for referencing in licensing applications. In response to the staff's request, the licensee and the fuel vendor, ABB-CE, evaluated their compliance with the conditions specified in the staff's SERs for those topical reports referenced in the submittals pertaining to the propnsed TS changes

. (reduction in minimum RCS flow rate and reduction in the steam generator low pressure setpoint). The licensee's evaluation determined that the SER conditions associated with the referenced topical reports had been satisfied. Accordingly, the staff finds that the licenseo has adequately addressed the request to ensure th ' previously approved topical reports are used consistent with the limitations established as a condition of NRC approval.

4.0 STATE CONSULTATION

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in accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The 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 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increast 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 previouly issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 4312). Accordingly, the 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

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environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

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 cperation in the proposed manner, (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.

Principal Contributor: S. Sun, NRR/SRXB Date: March 12, 1998 i

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