ML20070K490

Safety Evaluation Supporting Amend 60 to License NPF-49

ML20070K490
Person / Time
Site:	Millstone
Issue date:	03/11/1991
From:	Office of Nuclear Reactor Regulation
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ML20070K463	List: ML20070K460 ML20070K490 ML20070K512
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NUDOCS 9103180447
Download: ML20070K490 (20)

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I SAFETV J. VALUATION BY THEj f MCE OF NUCLEAR REACTOR REGULATip,N RELATED TO AMENDMENT NO. 60 TO FACILITY OPEPATING LICENSE NO. NPF-49 NORTHEAST tlUCLEAR ENERGY COMPANY, ET AL.

U!LLSTONE NUCLEAR POWER STATION,llNIT NO. 3 DOCKET NO bty23 i

1.0 [NLR,000CT10N By letter dated November 1, 1990, Northeast Nucleat Energy Company (NNECO or "the licensee") proposed to pend Operating)LHnse NPF-49 by incorporating changes to the Technicii Soet!fications (TS Millstone Unit No. 3.

Additional changes wer9 suamitted by letters u ted December 4, 1990 and f6bruary 15, 1991. These proposed changes werte made to support refueling and operation of Unit 3 for a fourth cycle. The proposed changes primarily result from changes in fuel design, the use of new analytical methodologies, and associated fuel / core related r.hanges.

A plant safety evaluation report for Millstone Unit 3 reload transition from the Cycle 3 core with 17x17 standard (STD)(SH) fuel design was also submitt Westinghouse fuei_ assemblies to a core also containing the VANTAGE 5 Hybrid support Cycle 4 operation and TS changes.

This report provided the results of*

the N l, nuclear, thermal-hydraulic, and accident analyses performed by the

licensee, in addition, a report describing instrument uncertainty methodology was submitted by letter dated November 2, 1990 and reviewed and evaluated by the staff in the following Safety Evaluation.

NNECO also requests.d removal of a cycle specific restriction on spent fuel

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storage. Spent fuel storage at Millstone 3 has been limited to fuel assemblies discharged through the end of Cycle 3.

This restriction was imposed ending reanalysis of the decay heat removal capability of the spent fuel poul, and evaluation of stresses in the associated piping.

This reanalysis was pravided by the licensee by letter dated November 30, 1990, with additional inforrntion provided by letter dated February 22, 1991.

Finally, by letter dated October 25, 1990, tafza requested removal of the-autoclosure interlock function from the residual heat removal _ system suction /

isolation valve.

Removal of_this function was suggested by Generic Le ter 88-17 as.a means to improve residual heat removal system reliability in pressurized:

water reactors.

NNECO provided additional information on this topic'by letter j

dated February 11, 1991.

. _ _ _ - _ _ - - 2.0 DISCUSSION AND EVALUATION 2.1 Qel_ Design Millstone Unit 3 Cycle 3 operated with Westinghouse 17x17 STD fuel.

Beginning with Cycle 4 Unit ? will also include VANTAGE SH fuel assemblies.

These assemblies incor flow mixer (lFM)porate the low-pressure drop Zirealoy grid and intermediate grid, and have been approved by the NRC.

In addition to these VANTAGE SH design features, Cycle 4 will also contain several VANTAGE 5 design features and other upgraded fuel design features previously approved by the NRC and used in the current Cycle 3 core.

These include the VANTAGE 5 integral fuel burnable abscrbers (IFBAs), the VANTAGE 5 axial blanket design, the VANTAGE 5 extended burnup design, the VA!!TAGE 5 reconstitutable tcp nozzle (RTN) design feature, debris filter bottom nozzles (DFBNs), snag-resistant grids, and standardized fuel pellets.

Although the NRC reviewed WCAP-10444, " Westinghouse Reference Core Report, VANTAGE 5 Fuel Assembly," and concluded that it was an acceptable reference to support plant-specific application of VANTACE 5, certain conditiens were specified. These conditions have been addressed by the licensee for this Cycle 4 application requeeting the use of VANTAGE SH fuel assemblies and are evaluated in the appropriate Sections of this Safety Evaluation.

In particular, for each plant application using VANTAGE SH fuel assemblies, it nust be demonstrated that the loss-of-coolant accident (LOCA)/ seismic loads considered in WCAP-9401 bound the plant in question; otherwise additional analysis will be required to demonstrate the fuel assembly strt ctural integrity.

Millstone 3 is not completely bounded by the LOCA/ seismic loads considered in WCAP-9401. Thus, in accordance with Condition 2 of the VANTAGE 5 HRC Safety Evaluation, additional plant-specific analyses were performed to demonstrate fuel assembly structural integrity.

An evaluation of 17x17 VANTAGE SH (with IFMs) fuel assembly structural integrity considering the lateral effects of seismic and LOCA accidents has been performed.

In accordance with NRC requirements in Appendix A of the Standard Review Plan (SRP) 4.2, the results show that the 17x17 VANTAGE SH is structurally acceptable for a transition core such as Cycle 4 consisting of both VANTAGE SH and STD fuel. The grids will not buckle due tc combined impact forces of a seismic /LOCA event. The core coolable geometry is maintained. The stresses resulting from seismic and LOCA induced deflections are within acceptable limits.

The reactor can be shut down under the combined faulted condition loads.

The fuel rod design evalua" ns for Cycle 4 were performed using NRC approved models and extended burnup oesign nethods. However, since the statistical convolution method for the evaluation of initial fuel rod to nozzle growth gap has not been approved by the NRC, worst-case fabrication tolerances and VANTAGE SH fuel rod and assembly growth were used to determine the initial fuel rod to nozzle growth gaps. All of the FSAR fuel rod design bases were satisfied.

Therefore, the staff conclude < ' hat the fuel design for Millstone 3 Cycle 4 is acceptable.

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, 2.2 Nuclear Design The effects of the VANTAGE SH fuel on the Cycle 4 physics parameters as compared to the STD fuel are small and the tii11 stone spent fuel pool criticality analysis allows for tN s+crage of VANTAGE dri fuel assenblies.

The nuclear design parenters characterizing the Cycle 4 core have been computed by methods previously used and approved for Westinghouse reactors Differences between the parameters for Cycle 3 and Cycle 4 are typical of the. normal cycle to cycle variations for fuel reloads.

However, in order to permit more flexibility in developing fuel management schemes (i.e., longer fuel cycles, improvement of fuel economy and neutron utilization, vessel fluence reduction) and to bound-futura cycles, some of these parameters have been further modified. ' This is allowabic because of the thermal margin gained through the use of the revised ehermal design methods end correlations discussed in Section 2.3.

The required shutdown margin for operating Modes 1 and 2'has been reduced from 1.6% delta k/k to 1.3% delta k/k.

The most restrictive shutdown margin requirement for these Modes is based on that required for the uncontrolled reactor coolant system (RCS) cooldown resulting from a postulated steam line break event at end-of-cycle (EOC) with T at no-load operating. temperature.

Thestaff'sevaluationofthiseventin$Ntion2.4showsthatthedesignbases-continues to be met with the reduced shutdown margin assumption.

For the shutdown Modes (Modes 3, 4. and 5), boron dilution analyses were performed to define the minimum shutdown margin requirements.

1 Beginning with Cycle 4, future cycles of operation for Millstone 3 will use increased peaking factors. The full power nuclear enthalpy--rise hot channel factor (F'N9i$u)mfon-loopheatfluxhotchannelfactor(F)L11mitwill 1.70.

ThN design limit will increase from-the current value of 1.55 to-increase from the current value of 2.32 to 2.60 while the ma91 mum -three-loop -

limit will increase from 2.25 to 3.0 at the maximum:75 percent rated thermal-power (RTP).

The increased limits on peaking factors continue to ensure that the design -limits on peak local. power density and minimum departure from nucleate boiling ratio (DNBR) are not exceeded during normal operation and i

anticipated operational occurrences (A00s) and that the peak clad temperature will not exceed the-2200'F emergency core cooling system (ECCS) ' acceptance -

t-criteria in the event of a loss of coolant accident (LOCA) as discussed'in Section 2.4.

i The IFH grid feature of the VANTAGE SH fuel design increases the core pressure i

drop.

One result is that the rod cluster control-assembly (RCCA) scram time has increased from 2.2 to 2.7 seconds.

The drop time is measured.from the-i beginning of decay of stationary Oripper coil voltage to dashpot entry.- The effect of this increase on the Cycle 4 safety analyses has been considered and found to be acceptable.

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Therodclusterccatralassembly(RCCA)allrodsout(AR0)'parkedpositionwill i

be redefined such that the RCCAs will be allowed to park.in the range of 222 to-231 steps and still be considered fully withdrawn.

This will help avoid control rod cladding wear--at the same-location as a result of contact between l:

4 the RCCA and the RCCA guide cards which are located in the upper internals.

Althcush this does not stop the wear-process, periodic-redefinition of the fully withdrawn " parked" position will spread the wear over a greater surface' area of the cladding.and minimize the probability of complete wear through the cladding at any spot.

Changing the RCCA to en ARO parked position which allows the rod to sit in the range of 222 to 231 steps presents no adverse effect the normal operation of Millstone 3 nor on the non-LOCA or 1.0CA analyses evaluated in Section 2.4 In addition, the rod. drop time criterion of 2.7 seconds discussed above is met for this range of parked positions'.

Therefore, the staff finds this acceptable.

Beginning with Cycic 4, Unit 3 will have the option of using relaxed axial offset control (RAOC) to define the allowed operation space of axial flux difference (AFD) versus thermal power for control of axial power distribution, This method of control has been approved by the NRC and allows a wider axial offset band compared to constant axial offset control (CAOC) which is the current control methoo being used by Unit 3.

Based on its review, the staff concludes that approved methods'have besa used, that the nuclear design parameters meet applicable criteria and are supported by design bases safety analyses discussed in Section 2.4 of this Safety Evaluation. Therefore, the nuclear design of Millstone 3 Cycle 4 is acceptable.

2.3 Thermal-Hydg ulic Design The thermal-hydraulic analysis of the Cycle 4 core containing-both Westinghouse 17x17 STD and VANTAGE SH fuel assemblies, incorporates the WRB-1 and WRB-2 DNB correlations, the Revised Thermal-Design Procedure (RTDP), and an improved THINC IV Modeling.

Each of these has been reviewed and approved by the NRC.

The W-3 DH correlation and the standard thermel-hydraulic methods described in the Millstone 3 FSAR are still used when conditions are-outside of-the range of the WRB-1 or WRB-2 DNB correlation und of the RTDP.

For the WRB-1 DNB correlation, the NRC has approved a 95/95 DNBR limit of 1.17-for the 17x17 STD fuel essemblies. The use of the WRB-2 Dim correlation with a 95/95 DNBR limit-of 1.17 is also applicable to the VANTAGE 5H fuel.

The W-3 correlation with a 95/95 DtlBR limit of 1.30 is used below the STD and VANTAGE SH fuel assembly first mixing vane grid. The W-3 correlation with a 95/95 DNBR limit of 1.4C is used for steam line break analyses in the-pressure range of 500 to 1000 psia.

With RTDp methodology, variations in plant operating-parameters,-nuclear =and thermal parameters, fuel fabrication paremeters, and DNB correlation predic-tions are considered statistically to obtain the overall DNBR uncertainty factor which is used to define the design limit DNBR that satisfies the DNB design criterion. This criterion is that the prcbability that DNB will not occur on the most limiting fuel rod is at least 95% (at a 95% confidence level) during normal operction or any A00.

Since the uncertainties are-all included in the uncertainty factor, the safety analyses are done with input parameters at their nominal values.

RTDP analyses use a new flow parameter, minimum l

_ _ - - - measured flow (MMF), equal to thermal design flow (TDF).plus a flow uncertainty.

Analyses by standard methods continue to use TDF.

The design limit DNBR values for 17x17 STD fuel are 1.25 for typical cells and 1.7.4 for thimble cells for both 4 and 3 loop operation.

For VANTAGE 5H fuel, the design limit DNBR values are 1.26 for typical cells and 1.24 for thimble cells for both 4 and 3 loop operation.

A rod bow penalty of 1.3% was calculated for the 17x17 STD fuel and for the lower assembly spans in the VANTAGE SH assemblies.

The maximum penalty is based on an assembly average burnup of 24,000 MWD /MTlj as approved by the llRC.

In addition, a transition core effect to account for localized flow redistribution from the VANTAGE SH assembly into the STD-assembly in a mixed core was calculatec using NRC approved methodology.

The DNBR penalty, which is a function of the number of VANTAGE EH fuel assemblies in the core, is less than 12.5% fer Cycle 4 operation.

For the Cycle 4 DNBR safety analyses the DNBR limits were increased to accourt fortheeffectsofrodbowandtransItioncorepenalties.

For 4-loop opera-tion, the VANTAGE SH fuel DNBR safety limits were 1.69 for typical cells and 1.65 for thimble cells. For 3-loop operation, the VANTAGE SH fuel DNBR safety limit values used were 1.75 for typical cells and 1.7? for thimble cells.

Therefore, the staff concludes that the rod bow and transition core penalties are adecuately covered by the margin maintained between the design and safety limit DNBR values. Maintenance of adequate DNBR margin to cover DNBR penalties is confirmed by the licensee on a cycle-specific basis during the reload safety esaluation process.

The thermal-hydraulic evaluation of the Cycle 4 fuel upgrade and peaking factor-increase has shown that 17x17 STD and VANTAGE SH fuel assemblies are hydraulically ccmpatible and that the CNB margin gained through use of the RTDp methodology and the WRB-1 and WRB-2 DNB coreelations is sufficient to allow an increase in-the design F exists in the safety limbOBN to cover any rod bow and transition core-f rom 1.55 to 1 penalties.

Approved methodology was used and all thermal-hydraulic design criteria were satisfied. Therefore, the staff finds the thermal-hydraulic design of Millstone-3 Cycle a acceptable.

2.a Transient and Accident Analyses The impact of the transition of Millstone 3 from Westinghouse 17x17 STD fuel to Westinghouse VANTAGE SH fuel as well as other changes proposed for Cycle 4 operation en the FSAR Chapter 15 accident analyses has -been reviewed by Westinghouse to determine which events need to be reanalyzed for Cycle 4 The review was based on event-specific sensitivities and a decision was made for each transient uith regard to the need for formal analysis as opposed to simply evaluating ti e impact of the subject features and assumptions. - Also considered was the possible difference in relative behavior of the 3-loop cases as compared to those with 4 locps operating. The 3-loop power rating used in the analyses represents 75% of the 4-loop RTP, which is. conservative with

respect to the current licensed maximum 3 loop pcwer for Millstone 3 of 65%

RTP. The safety reevaluations also assumed 10% uciform steam generator tube plugging with RCS thermal design flow rates of 378,400 gpm and 294,400 gpm for 4-loop and 3-loop operation, respectively, a core bypass flow incren from 6%

tc C.fA to account for complete thimble plug removel and the use of IFMs, grids,theuseofRelaxedAxialOffsetControl(RAOC).powerdistribution,and the increased peaking factors.

No one steam generator was assumed to exceed 10% tube plugging.

For 4-loop operation, a core inlet coolant temperature of 557.0 F and a coolant average temperature of 587.1'F were used.

For 3-loop operation, 550.2"F and 579.6'F were used for the inlet and average coolant temperature, respectively.

The RTOP methodology discussed above was used to define the initial conditions for most of the reanalyzed accidents to demonstrate that the DHB design basis is met. The other reanalyzed accidents used the Standard Thermal Design procedure (STDP) to obtain initial conditions by adding the maximum steady-state errors to nominal values.

The NRC requires a review of the temperature, pressure, power, and flow uncertainties used in the safety evaluations when using the RTDP. For Hillstone 3, the uncertainties have been calculated based on the installed plant instrumentation or special test equipment and on Unit 3 calibration and calorimetric procedures.

A report was submitted to the NRC with the Cycle 4 reload package which describes the-uncertainty evaluation and has been found to satisfactorily meet the NRC requirement for both 3-loop and 4-loop cperation.

The staff has reviewed each of the accidents which were reanalyzed or reevaluated for Cycle 4.

These reanalyses applied methods which have been previously found acceptable by the NRC. The results, which include transition core effects, show chana s in the consequences of transients and accidents

-previously analyzed.- b iver, the results remain within the required acceptance criteria. S aifically, for non-LOCA events, during normal opera-tion and anticipated op rational occurrences, there is at least a 95%

probability at a 95% confidence level that DNB will not' occur on the limiting fuel roo. During these operational modes, there is also a 95% probability at a 95% confidence that the peak kw/f t fuel rods will not exceed the melting temperature of UO taken as 4900 F (unirradiated) and 4800'F at-end of life.

For these events,2,eak RCS pressure does not exceed 110% of the 2250 psia p

design pressure.

Peak RCS pressure does not exceed 120% of the design pressure during accidents. The maximum average fuel pellet enthalpy was -less than 200 cal /gm for all control rod ejection events, thus meeting the staff criterion of-less than 280 cal /gm.

Except for the fuel tandling accident, the radiological consequences of those accidents reanalyzed because of the Cycle 4 changes remain bounded by the current reference analyses.

For the VANTAGE SH fuel, the new peaking factor of 1.7 was assumed in the Cycle 4 safety analyses. This is different from the current peaking factor of 1.65.

The only accident for which a peaking factor is used to calculate doses is the fuel handling accident.

Therefore, NNEC0 reanalyzed this accident using the NRC-developed TACTIII code and found an increase in dose of 3% due to the peaking factor increase.

However, the doses

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-7 remain well within the staff's acceptance criteria of 75 rem to the thyroid and 6 rem to the whole body.

In addition, due to the differences in the codes used by NNECO and Stone and Webster, the new doses are also at or below the doses that the PRC reviewed from the previous Stone and Webster calculation included in the FSAR.

The large break LOCA analysis for Millstone 3, applicable to a full core of VANTAGE SH fuel assemblies, was performed to develop specific peaking factor limits. The approved Westinghouse 1981 ECCS evaluation model was used and a spectrum of cold leg breaks were analyzed. The LOCA analysis considered.4-loop and 3-loop operation as well as transition core effects. The worst case peak clad temperature (PCT) was 2133*F for 4-loop operation and 1874*F for 3-loop operation. A transition core penalty of 50*F was applied to the PCT of the VAtlTAGE SH assembly. Therefore, the results demonstrated that the PCT acceptance criterion of 2200*F as well as the criteria related to clad oxidation and maximum hydrogen generation contained in 10 CFR 50.46 continue to be met.

In addition, the core remains amenable to cooling during and after the LOCA.

The small break (less than one square foot) LOCA analysis for 4-loop operation was performed with the approved Westinghouse ECCS small break evaluation model using the fl0 TRUMP-code.

The analysis assumed a full core of VANTAGE 5H fuel to determine PCT for a spectrum of cold leg breaks.

The results demonstrate that one centrifugal pump and one high head safety injection aump, together with the accumulators, provide sufficient core flooding to meet tie acceptance criteria limits of 10 CFR 50.46.

A reanalysis of ECCS performance for a small break LOCA with 3 loops in operation was not performed.

Since operation with a loop out-of-service would improve the ratio of pumped safety injection to power and a small break-is not as limiting as a large break, the staff concludes that small break LOCA analyses for 3-loop operation are not necessary to demonstrate compliance with the criteria of 10 CFR 50.46, in 6 letter dated January 20,.1988, the flRC requested that for future Millstone 3 reload core cycles, NNECO provide an expected anticipated transients without scram (ATWS)mcderatortemperaturecoefficient-(MTC)atequilibriumxenon conditions, pending the. staff's evaluation of a forthcoming Westinghouse Owners Group (WOG) response to a llRC request for information-on ATWS MTC dated June 12, 1987.

For the Cycle 4 core, the hot full power (HFP) MTC at equilibrium xenon conditions will be -7.2 pcm/*F.

-In' addition, the MTC will become more negative as cycle burnup increases.

This Cycle 4 value is more conservative than the -5.5 pcm/'F value used in the generic ATWS~ studies performed for 4-loop Westinghouse plant designs and which yielded a limiting peak pressure of-3200 psig.

Since Millstone 3 is, from an ATHS point of view, similar to the 4-loop class of Westinghouse plants for which the studies were performed, the staff concludes that ATWS considerations are not significantly impacted, in addition, the Millstone 3 Cycle 4 physics startup tests will provide a hot, zero power MTC reference point which can be used to assess the core design HFP MTC with equilibrium xenon conditions.

_. -. _ _ _.._ _-_.. 2.5 Spent Fuel Storage Spent fuel storage at Hi11 stone 3 has been limited to fuel. assemblies discharged through the end of Cycle 3.

This restriction was imposed pending reanalysis of the decay heat removal capability of the spent fuel pool, and evaluation of stresses in the associated piping. The reanalysis results were provided by the licensee by letter dated November 30, 1990, The licensee reevaluated the heat removal capability of the spent. fuel cooling i

system using a projected worst-case end-of-life decay heat loed.

The heat load was based on the expected spent fuel pool capacity, and projected reelistic:

operating conditions.

Heat exchanger performance was predicted by-a computer model based upon the actual heat exchanger construction, as opposed to earlier analysis, which had used lers precise methods based on the heat exchanger design specification sheet.

The licensee states the models were developed and used with appropriate quality assurance controls.

The reanalysis predicted spent fuel cooling system temperature! 1111 remain below.

those assumed in the original piping stress calculations. The-ore, cycle-specific restrictions on spent fuel storage can be removed.

ne staff has-l reviewed the licensee's evaluation-of November 30, 1990, and the additional information provided on February 22, 1991, and-accepts the-proposed change to TS 5.6.1.1.

2.6 Removal of the Autoclosure Interlock for the Residual Heat Removal System Suction / Isolation Valves

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By letter dated April 22,1988,-theWestinghouseOwnersGroup(WOG) submitted Topical Report WCAP-11736 entitled " Residual Heat Removal System Autoclosure.

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Interlock Removal Report for the Westinghcuse Owners Group" for NRC review.

WCAP-11736 documents the analyses performed to justify deletion-of,the autoclosure interlock.(ACI) on the Residual' Heat Removal System (RHRS) suction /

isolation valves at four reference plants:

Salem Unit 1, Callaway Unit 1, North Anna Unit 1, and Shearon Harris Unit 1.-

The reference plants represent the lead plant in each of four groups into which WOG participating plants were categorized based on similarity of RHRS ; configuration 'and design characteristics.

The proposed ACI deletion addresses NRC concerns regarding potential failure'of ACI circuitry resulting in isolation of the RHRS-with attendant loss of decay heat removal capability durino cold shutdown and refueling.

A Safety Evaluation Report (SER) documenting the. NRC review of WCAp-11736 was-issued on August 8, 1989. The SER concluded that a net safety benefit would result from removal of the RHRS ACI provided that five-plant improvements delineated in the SER are implemented, in addition, the SER concluded that the information contained in WCAP-11736 may be-referenced to supplement licensees' plant-specific submittals requesting removal of the RHRS ACI. However, such reference only would be used to show compliance with those items that are

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9 generic to the WOG plants. A plant-specific submittal would be required of each licensee seeking approval to remove the RHRS ACl. The above referenced plant improvements are listed below:

(1) An alarm will be added to each RHR suction valve which will actuate if the valve is open and the pressure is greater than the open permissive setnoint and less than the RHR system design pressure minus the RHR pump head pressure.

(2)- Valve position indication to the alarm must be provided from the stem-mountedlimitswitches-(SMLSs)and affected by power lockout of the valve. power to the SMLSs must not be (3) The procedural improvements described in WCAP-11736 should be implemented.

Procedures themselves are plant specific.

(4) Where feasible, power should be removed from the RHR suction valves prior to their being leak-checked (plant specific).

(5) The RHR suction valve operators should be sized so that the valves cannot be opened against full system pressure (plant specific).

By letter dated October 25, 1990, the licensee submitted an application to revise Technical Specification 4.5.2.d.1. Supplemental information was provided by _ letter dated February _11,1991. The proposed revision would delete the surveillance requirements to verify operability of ACI for the RHRS suction /

isolation valves on high RCS pressure.

Elimination of ACI surveillance is.a result of the licensee's plans to remove the RHRS ACI during the February 1991 outage.

As noted above, the NPC-approved report UCAP-11736 provides the underlying basis for justifying the licensee's planned action.

The WCAP-11736 reference plant for Millstone, Unit 3 is Callaway Unit 1.- The licensee's October 25, 1990 subnittal incivdes a plant-specific analysis of_the-planned ACI deletion-as a supplement to WCAP-11736.

The submittal includes an identification-of the differences existing between-the Millstone, Unit 3 RHRS configuration and design characteristics and those of.

the WCAp-11736 reference plant. At Millstone. Unit 3, there are three in-series POVs in each of the two RHPS pump suction lines, two within containment and one-located outside. The inboard MOV closest to the containment wall (MV8701A and 8702B) as well as the outboard MOV (MV8701B and 8702A) are provided with an ACI feature which the licensee proposes to remove. These valves are, in addition,.

furnishedwithanopenpermissiveinterlock(OPI)whichpreventsopeningwhen the RCS pressure plus the RHR pump head is above the RHRS design pressure (600 psi).

RCS pressure in this case is approximately 375 psi.

The OPI will remain intact and its surveillance TS will be unaffected by the proposed ACI removal.

Additionally, these valves are equipped with key lock handswitches which are administrative 1y controlled; position ir.dication and associated computer points

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1 are available at all tirms. A relief valve is located inside containment in between the two valves.

The third in-series MOV (MV8701C and 87020), located inboard and closest to the RCS hot leg, is not provided with ACI or OP1 i

features.

During power operation, this valve is closed and, unlike the other two MOVs, de-energized at the motor control center; position indication and associated computer points are available at all times, includirg during power lockout.. No changes have been proposed for this valve.

Only the two inboard MOVs are classified as pressure isolation valves (P1Vs).

The outboard valve (MV8701B and 8702A) is not designed to withstand RCS pressure for extended i

periods and is not leak tested.

In comparison, for the WCAP-11736 reference plant, there are two in-series motor-operated valves (POVs) located in each RHRS suction line.

Each valve is equipped with ACI and OPI features and is classified as a P1V. During power operation, both of these valves are closed and de-energized.

A relief valve is located downstream of these valves.

4 Probabilistic risk assessments (PRAs) addressing ACI removal and the j

installation of new alarms were provided in WCAP-11736 for the four reference plants.

Because of the differences in RHRS configuration and design characteristics existing between Millstone, Unit 3 and its reference plant (Callaway), the Callaway PRA is not directly applicable to Millstone, Unit 3.

Accordingly, in order to demonstrate that these deviations do-not invalidate i

the conclusions reached in WCAP-11736, the licensee submitted a plant-specific-PRA for Millstone, Unit 3.

This PRA addresses the risks associated with an i

intersystem loss of coolant accident (ISLOCA), RHRS short-term and long-term cooling unhvailability, and overpressure transients due to both heat input and mass input.

The results of the Millstone, Unit 3 PRA are consistent with those of UCAP-11736 and indicate an overall reduction in risk as a result of ACI s

removal combined with the installation of new alarms.

With regard to the five plant improvements discussed above, the licensee's October 25, 1990 and February 11, 1991 submittals have provided the following responses:

I Concerning Improvement 1, the licensee plans to add an alarm-to each of the RHRS suction valves currently equipped with an ACI feature.

The alarm will actuate if the valve is not fully closed when'RCS pressure is above the alarm setpoint.

The setpoint will be determined in accordance with WCAP-11736.

Also, in accordance with WCAP-11736, j

the existing OPI circuitry for these valves will remain intact and unchanged. The licensee does not plan to install an alarm on-the i

inboard MOV (MV8701C and 8702C) closest to the RCS hot leg. This valve, as noted above, is not furnished with an ACI or OPI feature and is de-energized during power operation.

The current procedures direct the operator to verify valve closure indications for all three in-series MOVs prior to de-energization.

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i Concerning Improvement 2, the licensee proposes to use existing I

Limitorque limit switches on the actuator cam /rctor rather than l

installing stem-mounted limit switches for valve position indication to the new alarms.

Use of the existing limit switches provides direct position indication and has been previously approved by the NRC for i

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- other plants, furthermore, as noted earlier, position indication and associated computer points are available at all times for each of the three in-series MOVs.

• Concerning Improvement 3, the licensee has reviewed the Millstone, Unit 3 operating procedures to determine the impacts of ACI removal and the installation of new alarms,- and has identified the necessary procedural revisions. This includes modification of the alarm response portion of the plant startu power operation.p procedures as well as the procedures applicable to In addition, a surveillance procedure will be written to provide for periodic testing of the new alarms.

• Concerning Improvement it was noted above that onl 4,(87020) and MV 8701A (8702B)y the two inboard suction valves, MV 87010

, are classified as PlVs and are, accordingly, leak tested. The licensee does not plan to de-energize MV 8701A (8702B) prior to testing. As stated earlier, this valve remains energized at all times and is equipped with a key lock handswitch which is administrative 1y controlled.

Position indication is available at all times. The leak testing procedure requires that MV 8701C and MV 8701A first be closed.

MV 8701A is then tested by gradually opening MV 8701C.

Subsequently, MV 8701C is closed, de-energized (i.e. its normc1 state for power operation) and leak tested. During the testin suction valve MV 8701B (8702A) g of both inboard valves, the outboard is key-lock closed and remains in that state during power operation.

' Concerning Imarovement 5, the licensee's February 11, 1991 submittal l

states that tie existing motor operators on the RHRS suction valves are -

indeed capable of opening these valves against full RCS pressure..

However,asnotedearlier,valvesMV8701A(8702B)-andMV8701B(8702A) are provided with an OPI feature which prevents opening when RCS pressures exceed 375 psia. The OPl is tested according to Technical Specification 4.5.2,d.1, once every -18 months.

Additionally, and as noted above, valve MV 87010-(87020) is de-energized during power operation. Therefore, the likelihood of an ISLOCA scenario-owing-to an inadvertent open signal is extremely-low. -On this basis, the licensee does not plan to downsi::e the motor. t.atuators.

We have completed our evaluation of the licensee's October PS, 1990 and February 11, 1991 submittals and have concluded the following:

• The licensee has adequately. identified differences in RHRS configuration and design characteristics that exist between Millstone, Unit 3 and the reference plant-(Callaway) addressed in WCAP-11730. By performing a plant-specific PRA, the licensee'has satisfactorily demonstrated that ACI removal combined with the l

installation of new alarms results in an overall reduction in risk.

l This is consistent with the results provided in WCAP-11736 for the reference plant.

V

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• The licensee has adequately addressed the five plant improvements requested -by the NRC.for plant-specific application 'of WCAP-11736.

Where deviations between these improvements and the licensee's proposed actions were identified, the licensee has adequately demonstrated that the proposed actions provided at least an_ equivalent level of_ safety.:

• The proposed change to TS 4.5.2.d.1 is consistent with the licensee's-plans to remove the RHRS suction / isolation valveLAC1. The licensee-has stated that no other TS requires revision as a-result of-the-proposed action.

On the basis of the above' evaluation, we find the proposed TS changes and the proposed plan for RHRS ACI removal to be acceptable.

3.0 TECHNICAL SPECIFICATION CHANGES tiilistone 3 Technical Specifications have been modified to support Cycle 4:'

operation. The specific changes are evaluated below.

(1) Definitions 1.a3 and 1.44 have been added for APL and APLOL.:

These are required as part of the usexof the approved RAOC for the-control of: axial power distribution and are acceptable.

(?) The Reactor Core Safety Limits specified in Figures 2.1-1 and 2.1-2 are modified.

The changes reflect the DNB correlation used for the standard and VAllTAGE SH fuel and are acceptable.

(3) The Over Temperature Delta T-(OTdT) and Over Power Delta T (0PdT) trip-setpoints specified in Table-2.2-1 have been revised.-

i The implementation of. VANTAGE SH fuel and the inclusion of more-conserva-tive uncertainty values in temperature and pressure cause:the DNB core limits to change. These core limit changes;and implementation of,RAOC result in OTdT-and'0PdT reactor -trip-setpoint -changes.- -These setpoint changes are -ref.lected in.the -Cycle 4 - safety' analyses, which resulted in--

acceptable consequences, and are, therefore, acceptable.

(4) Bases Sections 2.1.1 and 2.P.1.have beei. revised to incorporate the WRB-1 and HRB-2 DNB correlations.

The DtlB analysis of the core _ containing both the. standard and -VAtlTAGE -5H--

fuel assemblies _ incorporates-the NRC-approved WRB-1 and WRB-2 correlations and has been shown to be acceptable in Section 2.3.

The change is, therefore, acceptable.

(5) The shutdown margir. for Modes 1 and 2 specified in TS 3.1.1.1.1 has been reduced from 1.6% delta k/k to 1.3t delta k/k.

A shutdown margin of 1.3% delta k/k was used in the non-LOCA safety _

analyces for Cycle 4 for all events initiated from Modes 1 or 2.

For all events, including the most limiting steam line break, the design bases continue to be met with the reduced shutdown margin assumption.

Therefore, the change is acceptable.

(6) The shutdown margin for Modes 3, 4, and 5 (loops filled) has been relocated to TS 3.1.1.1.2.

Figures 3.1-1(4-loops-Mode 3),3.1-2 (3-loops, Mode 3), 3.1-3 (Mode 4) and 3.1-4 (Mode 5 with loops filled) c have been added and show the shutdown margin as a function of critical boron concentration.

For these shutdown Modes, boron dilution analyses were performed to define the minimum shutdown margin requirements. The results show that the minimum time intervals available to the gerator before a loss of shutdown margin occurs meet those required by SRP Secticn 15,4.6. - As required by the SRP, these minimum time intervals are calculated from the time an alarm alerts the operator to_a dilution, not from the time the dilution begins. The new shutdown margin monitors which provid.e-the alarm are discussed in item (27) below. Therefore, the p1oposed thenges are acceptable.

(7) The shutdown margin for Mode 5 with the loops not filled 'specified in TS 3.1.1.2 has been modified to either require a larger shutdown margin per new Figure 3.1-5 or to preclude a boron dilutien event by securing dilation flow paths and maintaining the same shutdown _ margin as required l

for Mode 5 with loops filled per Figure 3.1-4.

The boron.di btion analysis perfoimed for Pode 5 drained shows that-the required M nimum 15 minutes for operator action from the time of the flux multiplication alarm to loss of shutdown margin is met with the proposed TS'Fisure 3.1-5.

In addition, when the water level is drained down from a filled and vented condition in Mode 5, an uncontrolled boron dilution event may be prevented by administrative controls which isolate the RCS from the potential source of'unborated water.

This _is' accomplished in.the revised TS by an option which allows dilution source valves to be secured.

The proposed change is,.therefore, acceptable.

(8) The wording of TS 3.1.2.1 and 3.1.2.2 has been changed to allow the use of either one boric acid tank or two.

The proposed change of wording is necessary to make these specifications consistent with the Unit 3 design and with the wording of TS 3.1.2.5 and 3.1.2.6. 'Therefore, the proposed change is acceptable.

(9) The Action statement of TS 3.1.2.4 has been revised to reflect the new Figure 3.1-4.

. _ _ _ _ _ _ _ _ This Figure has been previously approved.

Therefore, the proposed change is acceptable.

(10) The minimum boron concentraticn during shutdown Modes 5 and 6 in the refueling water storage tank (RWST) specified in TS 3.1.2.5 has been l

increased from 2300 ppm to 2700 ppm.

The boron capability; required below 200 F must be sufficient to provide a shutdown margin of 1.3% delta k/k after xenon decay and cooldown from 200*F to 140'F. This can be accomplished with either a usable volume of 4100 gallons of 6300 ppm borated water from the boric acid storage tanks or 250,000 gallons of 2700 ppm borated water from the RWST.

The usable volume in each BAST is 1300 gallons.

In addition, the 2700 ppm boron concentraticr limit in conjunction with the limit on contained water volume ensure a pH value of between 7.0 and 7.5-for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The change is, therefore, acceptable.

(11) The range in the RWST boron concentratior, during Modes 1, 2, 3 -and 4 specified in TS 3.1.2.6 has been changed from 2300-2600 ppm to 2700-2900 ppm.

In addition, " contained volume" has been changed to " usable volume."

In order to c.aintain the core subcritical via boron addition after a large break (greater than 3.0 sq. ft.) LOCA, an increase in the boron concentra-tion range in the RWST and accumulators is necessary. - Based on the safety analyses performed for Cycle 4, an increase in the minimum RWST boron has no adverse impact on the non-LOCA or LOCA accident analyses and provides adequate shutdown margin. However, the required post-LOCA pH range of 7;0 to 7.5 cannot be assured with the increased boron concentration range-unless there is an increase in the amount of sodium hydroxide (Na0H) transferred from the Chemical Additive Tank (CAT) to containment via the s3 rays. The CAT volume and NaOH concentration requirements were therefore cianged in TS 3.6.2.3 to ensure that the minimum sump pH will be in the required range. The maximum expected boration capability; requirement-occurs-at E0L from full power equilibrium xenon conditions and requires a usable volume of 21,020 gallons of 6300 ppm borated water from the BASTS or 1,166,000 gallons of 2700 ppm borated water from the'RWST.

A minimum RWST volume of 1,166,000 gallons is specified to be' consistent with the ECCS requirement. Therefore, the change is acceptable.

(12) The maximum allowable control rod drop time specified in TS 3.1.3.4 has been increased from 2.2 seconds to 2.7 seconds due to the fuel design changes.

The effect of this increase on the Cycle 4 safety analyses has been considered and found to be acceptable.

The drop times are demonstrated through measurement prior to reactor criticality. Therefore, the change is acceptable.

-15 (13) The axial flux difference (AFD) specified in TS 3.2.1.1 for 4-loops operation and in TS 3.2.1.2 for 3-loops operation will be modified to allow either relaxed axial offset control (RAOC) or base load operation for the control of axial power distribution..

Currently, Millstone 3 uses the constant axial offset control (CAOC) procedure for the centrol of axial power distribution.

Beginning with Cycle 4, Unit 3 will have an option of the RAOC or base load. The RAOC' procedure defines the allowed operating space of AFD versus thermal power as described in Section 2.2 above. This procedure has been approved by the NRC and the proposed TS Sections for 4-loop and 3-loop operation are consistent with the existing Westinghouse Standard TS.

Therefore, the change is acceptable.

(14) The Surveillance Requirement Sections for heat flux hot channel factor (Fn) specified in TS 3.2,2.1 for 4-loop operation and in TS 3.2.2.2 for 34 00p operation will be modified to incorporate RA00 operation.

In addition, the F measurement for 3-loop operation will be based on a full n

power value of 75%.

As previously stated, Unit 3 will be using RAOC.

The proposed surveillance sections are based on the Westinghouse Standard TS. The F measurement for 3-loop operation is now based on a safety analysis fu110 power of 75% of the 4-loop RTP. Therefore, the proposed changes are acceptabic.

(15)TheRCSflowratespecifiedinTS3.2.3.1for4-loopoperationandinTS 3.2.3.2 for 3-loop operation will be modified to denote the minimum measured flow.

In addition, the uncertainty for P,CS flow for 4-loop operation will be increased to 12.4% from the previous rl.8% and to 2.8%

from t2.0% for 3-loop operation.

With the use of RTDP for the VANTAGE SH transition analysis, the non-LOCA safety analyses performed to confirm the DNB design basis used the minimum measured flow (MMF) values proposed in the modified TS. As mentioned in Section 2.3 above, MMF is equal to the thermal design flow (TDF) plus a flow uncertainty. The flow uncertainties proposed in the revised TS are those reported in WCAP-12621 and are approved by the NRC and used in the Cycle 4 safety analyses. Those Cycle 4 events reanalyzed using the STDP continued to use the TDF for initial conditions. The proposed TS flow and uncertainty values were used in the Cycle 4 safety analyses and all design criteria were met.

It should be noted that the Millstone 3 TS will require a flow penalty of 0.1% for undetected fouling of the feedwater venturi if the venturis are inspected and. cleaned at least once every 18 Therefore, the proposed changes are accep. table.

montns.

(16)iheDNBparameters(indicatedRCST and indicated pressurizer pressure) for4-loopand3-loopoperationspeB59iedinTable3.2-1havebeen modified.

16-The DNB parameters are modified due-to the revised instrument uncertain-ties. The revised RCS average temperature uncertainty allowance is

+6.0/-6.6 *F as compared to the previous 6.1 'F.

The revised pressurizer pressure uncertcinty is +50/-E3 psi as compared to the 445 psi-noted currently in the FSAR. The first value given is for Veritrak transmitters and the second value is for Rosemount transmitters: both of which are used at Millstone 3.

The. instrument uncertainty analysis described in WCAP-12621 has been reviewed and approved by the NRC.

The reported instrument uncertainties have been used in the Cycle 4 safety analyses.

Therefore, the proposed chances are acceptable.

(17) The response tirne for the pressurizer water level-high reactor trip has been specified as less than or_ equal to 2 seconds in Table 3.3-2.-

The current licensing basis analysis for this event did not model the actuation of the high pressurizer water level reactor trip.

However, this value has been used in the revised Cycle 4 safety analysis for the RCCA bank withdrawal at power transient and ensures that the pressurizer does not go water solid.

The change is therefore, acceptable.

(18) Engineered safety feature instrumentation setpoints in Table 3.3-4 are revised to incorporate revised uncertainty analysis.

The licensee proposes splitting the existing ~four channels of pressurizer pressure measurement into separate channels for the Veritrak and Rosemont instrunientation. This channel designation will have different sets of uncertainty values.

The licensee proposes this revision because the existing TS does not address the differences between the channels that are due to the characteristics of the instrumentation.

The staff has determined that this approach will not degrade the safety margin already i

existing in the TS, and therefore finds the changes to_be acceptable.

The. licensee revised their uncertainty analyses for the pressurizer pret sure low and steam line pressure low setpoint total allowance (TA),

2, and sensor error (S) values in the technical specifications. ~The licensee states that the revisions are based on data that indicates there l

is an additional error due to decreased cable insulation resistance at elevated temperatures.

The new setpoints account for the resulting error-in the pressure measurement.

The staff evaluated the proposed revisions and finds them acceptable.

(19) TS 3.4.1.2 has been revised to require at least three reactor coolant loops to be in operation during Mode 3 when the reactor trip system breakers are closed.

A single reactor coolant loop provides sufficient heat removal capacity-if an RCCA withdrawal transient can be-prevented, i.e., by opening the reactor trip system breakers.

However, with-the trip breakers closed, an RCCA withdrawal transient must be considered in Mode 3.

Cycle 4 analyses have shown that the DNB-design basis continues to be met for an RCCA N

+

w-

-e r.,,w withdrawal in Mode 3 if three reactor coolant loops are in operation.

The l

proposed change is, therefore, acceptable.

(20) TS 3.4.1.4.2 has been revised to require certain valves in the chemical and volume control system (CYCS) to be secured.

A list of these valves is included in this revised TS and requires verification at least once per 31 days that these valves are closed or locked under administrative controls.

The licensee's boron dilution analysis, submitted on December 4,1990, originally included controls on valve V119 in surveillance requirement 4.4.1.4.2.3.

In a letter dated February 15, 1991, the licensee noted that this is an air-operated valve in a high radiation area.

The licensee proposed to use valve V120, which is a manual valve in a low radiation area. The staff accepts use of valve V120 instead of valve V119 in this surveillance requirement.

As mentioned in item (7) above, certain valves in the CVCS may be secured in order to prevent a boron dilution event.

Verification that these valves are secured is required every 31 days.

The staff considers this to be acceptable.

(21) The boron concentration of the isolated loop specified in TS 3.4.1.6 for Modes 5 and 6 is increased from 2300 ppm to 2600 ppm.

The 2600 ppm value is sufficient to bound shutdown margin requirements for Modes 5 and 6 and provide for boron concentration measurement uncertainty between the loop and the RWST.

The applicable Mode 5 and 6 shutdown l

margin TS are referenced.

The proposed change would allow the isolated loop to be brought into service as long as the water used to fill the isolated loop has a boron concentration at least equal to the rest of the RCS.

Verification of the boron concentration in an idle loop _ prior to opening the stop valves provides a reassurance of the adequacy of the boren concentration in the isolated loop. Therefore, the proposed change is acceptable.

(22) Thc boron concentration in the accumulators specified in TS 3.5.1 is increased from 2200 - 2600 ppm to 2600 - 2900 ppm.

As discussed in item (11) above, an increase in the boron concentration in the accumulators and the RWST is necessary to meet the post-LOCA shutdown requirements.

The increase in the accumulator boron concentration along with an increase in the minimum RWST boron concentration (see below) has been shown to have no adverse impact upon the Cycle 4 non-LOCA safety I

analyses.

For the LOCA related analyses, the higher concentrations decrease the allowable time for operator action to initiate hot leg recirculation. However, the resulting time requirement of 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> is still mere than adequate to assure those operator actions can be accomplished.

Therefore, there is no adverse effect on LOCA related accidents for the proposed boron concentration increases. These analyses are applicable to both 3.co; and 4-1 cop operation.

The proposed changes i

are, therefore, accer table.

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L (23) Surveillance Requirement (SR) 4.5.2.f.2 reflects a 10%- reduction in safety injection (SI) flow and SR 4.5.2.h.2 reflects a 10 gpm SI pump flow 1

l imbalance.

A 10% reduction in SI and charging flow was incorporated in the LOCA and non-LOCA analyses as well as a 10-gpm SI and char 1-The prcposed changes are, therefore, acceptable. ging pump flow imbalance.

l (24) The boron concentration in the RWST specified in TS 3.5.4 is increased from 2300-2600 ppm to 2700-2900 ppm.

As discussed in item (11) above, the effect of this increase on the Cycle 4 safety analyses has been considered.

Therefore, the proposed change is acceptable.

(25) The range of sodium hydroxide concentrations in the chemical addition tank ~

(CAT) specified in TS 3.6.2.3 will increase to 3.4-4.1 percent from the previous range of 2.4-3.1 percent.

In addition, the volume (level) in the CAT is reduced to a range of 17,760-18,760 gallons from the previous range of 18,000-19,000- gallons.

As discussed in item (11) above, the r_equired post-LOCA pH range of 7.0 to 7.5 cannot be assured with the-increased boron concentration range unless l-there is an increase in the amount of sodium hydrolide transferred from l

the CAT to containment via the sprays..The revised CAT volume and Na0H l

concentration requirements ensure that the minimum sump pH will.-be in the l

required range. Therefore, the proposed changes are acceptable.

(26) The boron concentration in the filled por_ tion of the RCS and refueling canal during Mode 6 specified in TS 3.9.1.1 is-' increased to 2600 ppm.

In addition, the CVCS valves of Specification 4.4.1~.4.2.3 are required-to be closed and secured ia position.

This minimum required boron concentration ensures that the core will remain subcritical by at least the required SL delta k/k during refueling and includes a conservative uncertainty allowance of 50 ppm boron.

In addition, the occurrence of a boron dilution event is precluded by the requirement that the boron dilution source valves be secured.

The proposed change is, therefore, acceptable.

-(27) Existing function unit 6b of Table 3.3-1 (high flux monitor at shutdown) will be deleted and a new functional unit 21 -reflecting the proposed-shutdown margin monitor will be added to the Table.

For Cycle 4, two shutdown margin-monitors, one per train, will be added_ to the design. These monitors will m asure the count rate obtained from the l

GAMMA-METRICS wide range neutron-flux monitoring system and provide an alarm when the count rate increases by an amount equal to the alarm ratio set into the monitors.

These monitors will provide an alarm only and will not perform a protective function such as a reactor trip. These monitors

%p i !

serve to meet the requirements of SRP 15.4.6 for redundant alarms to alert the operator of any unplanned boren dilution event and have been credited in the Cycle 4 boron dilution safety analyses during shutdown conditions.

The alarm setpoint of 2.0 corresponds to a doubling of the_ neutron flux and a 10 second delay for alarm response has been assumed.

Based on these considerations, and on the addition of a quarterly surveillance requirement in Table 4.3-1 to verify the correct alarm setting for the shutdown margin monitor, the proposed changes are acceptable.

(28) The Bases associated with the above TS changes are being revised to reflect the proposed changes.

The proposed TS changes have been approved and, therefore, the associated Bases are acceptable.

(29) TS Sections 6.9.1.6 a.2, 6.9.1.6 a.3, 6.9.1.6.a.4, 6.9.1.6.a.5, and 6.9.1.6.a.6 are being revised.

The proposed revisions correct the references to the approved modified TS l

c numbers and make them consistent with the revised TS. Therefore, the revisions are acceptable.

(30) TS Section 6.9.1.6 b is being revised +n include additional references.

The referer.ced methodology reports are those used in the Cycle 4 safety analyses for the upgraded VANTAGE SH fuel and have been reviewed and approved by the NRC.- Therefore, the proposed change is acceptable.

l (31) Several TS figures are being redrawn for clarity and TS aages are being renumbered to accommodate the approved TS changes described above.

These changes are editorial in nature and are acceptable.

(32) TS 4.5.2.d.1 is being revised to delete surveillance of the automatic closure function, i

The proposed change is expected to improve reliability of the residual heat removal system and reduce overall plant risk, and is acceptable.

(33) TS 5.6.1.1 is being revised to delete cycle-specific restrictions on spent fuel storage.

The licensee has demonstrated the spent fuel pool will be operated within previously accepted design -limits.

Therefore, the proposed-change is acceptable.

4.0

SUMMARY

The staff has reviewed the reports submitted for the Cycle 4 operation of Millstone 3 and concludes that appropriate material was submitted and that the

m I

~20-fuel design, nuclear design, thermal-hydrculic design, and transient and accident analyses are acceptable. The TS changes submitted for-this reload suitably reflect the necessary nodifications for operation in this cycle.

The staff has reviewed information provided-in the licensee's submittals of November 30, 1990 and February 22, 1991 supporting removal of cycle-specific limitations on spent fuel storage. The staff concludes elimination of_these restrictions is acceptable.

The staf f has reviewed information provided in the licensee's submittals of October 25, 1990 and February 11, 1991, supporting removal of the autoclosure interlock from the residual heat removal system suction / isolation valves. The staff concludes the licensee has adequately addressed improvements requested for plant-specific application of WCAP-11736. Therefore, the staff concludes removal of the interlock,-combined with installation of new alarms, results in an overall reduction in risk, and is acceptable.

5.0 ENVIRONMENTAL CONSIDERA, TION Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact has been prepared and~ published in the Federal-Register on Januar 10 1991 (56 FR 1033).

Accordingly, based upon the ~~

environmental assessmen,t, we have determined that the issuance of the amendment will not have a significant effect on the quality of the human environment.

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6.0 CONCLUSION

l Wehaveconcluded,basedontheconsiderationsdiscussedabove,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 i'

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 orfto the health and safety of the public.

Principal Contributors:

L. Kopp M. Waterman H. Abelson J. Williams Date: March 11,1991 L

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