ML20246J793

Safety Evaluation Supporting Amends 120 & 123 to Licenses DPR-24 & DPR-27,respectively

ML20246J793
Person / Time
Site:	Point Beach
Issue date:	05/08/1989
From:	Office of Nuclear Reactor Regulation
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ML20246J787	List: ML20246J781 ML20246J793
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NUDOCS 8905170174
Download: ML20246J793 (20)

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g UNITED STATES f

g NUCLEAR REGULATORY COMMISSION j

WASHINGTON, D. C. 20555 7.

N l

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

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RELATED TO AMEhDMENT N05.12U AND 144 TO FACILITY OPERATING LICEh5E NOS. DPR-24 AND DPR-27

, WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAF PLANT, UNIT N05. 1 AND 2 DOCKL1 h05. 50-266 AND 50-301

1.0 INTRODUCTION

October 28 (Ref.- 2, November 30 (Ref 426,1988 (Ref.1)},as supplemented by le By letter dated Au ust December es modified January 17 1988 (sic) (Ref. 28), the Wisconsin Electric Power Company (the licensee),made application to change the Technical Specifications of the Point Beach Nuclear Plant, Units 1 and 2.

The proposed changes would permit the design and operation of future Point Beach Nuclear Plant (PBNP) reactor cores with enhanced Optimized Fuel Assembly (0FA) fuel and at higher core power pe6 king factors than are allowed by current Technical Specifications.

The higher power peaking factors will allow the use of a low-low leakage loading pattern (L4P) fuel management strategy which will result in a decrease in the fluence accumulation rate to the reactor pressure vessel.

Additional core design features included in the licensee's submittals are (1)useofPeripheralPowerSuppressionAssemblies(PPSA),(2)removalof fuel assembly thimble plugging devices, and (3) elimination of the third line segment of the K(Z) curve in Technical Specification Figure 15.3.10-3.

The use of PPSA's by the licensee is part of the L4P fuel management strategy.

The enhanced 14x14 0FA fuel design incorporates the following features:

(1) i Renovable Top Nczzles (RTN), (2) Integral Fuel Burnable Absorbers (IFBA), (3) axialblankets,(4)DebrisFilterBottomNozzles(DFBN),and(5) extended burnup geometry. These fuel features, with the exception of the DFBN, are a subset of the Westinghouse Vantage 5 fuel design. The bottom nozzle of the PBNP fuel will differ from the Vantage 5 bottom oczzle in that it will be fabricated from stainless steel rather than Inconel and the size and pattern of the flow holes have been changed. The DFSN will, however, meet all other design requirements. The Vantage 5 fuel design was generically approved with conditions (Ref.3).

The licensee has submitted a revised large-break LOCA analysis (Ref. 4) as part of the resolution of the Upper Plenum Injection (UPI) issue using the l

best estimate W COBRA / TRAC model (Ref. 5). This model has been reviewed and l

approved by the NRC (Ref. 6). The licensee also submitted a reanalysis of the small-break LOCA using the approved NOTRUMP code (Ref. 7). Based on generic small-break LOCA studies and results of analyses for the Northern States Power Prairie Island plants, thc licensee analyzed only the 4-inch cold-leg break.

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. The steam generator tube rupture (SGTR) event has been reanalyzed using the sane methodology that was previously used in the PBNP Final Safety Analysis Report (FSAR).

However, some revised input assumptions have been used.

l The licensee's analysis includes the use of a revised methodology, the revised thermal design procedure (RTDP). TheRTDP(Ref.8)hasbeenreviewedand approved by the staff (Ref. 9). All other analysis methodologies for the non-LOCA transients, except for the input changes noted for the SGTR event, i

j-are the same as those currently used in the PBNP FSAR analyses.

2.0 EVALUATION 2.1 Design Features and Parameters Future PBNP cores will contain 0FA, enhanced 0FA, and previously depleted Low-Par 6sitic (LOPAR) fuel. assemblies that are also known as standard (STD)

. fuel assemblies. The STD fuel assemblies will be bounded by the OFA fuel assemblies because the STD fuel assemblies will be previously irradiated fuel that will operate at lower power than the OFA fuel. The licensee will justify the use of previously irradiated STD fuel by cycle-specific reload analyses.

All of the fuel designs have a 14x14 geometry with 179 fuel rods and 17 guide tubes and an instrumentation thimble. The upgraded 0FA will include a number of Vantage 5 features:

(1)RemovableTopNozzles(RTN),(2)IntegralFuel BurnsbleAbsorbers(IFBA),(3)axialblenkets,and(4)extendedburnup In addition, the upgraded 0FA will include a Debris Filter Bottom geometry (DFBN). The licensee states that these 0FA upgrade features may not Nozzle all be used together in upgraded 0FA fuel but that the upgrade features used will be bounded by the reference analyses that have been submitted.

For its plant life extension (PLEX) program the licensee proposes to introduce a low-low leakage loading puttern (L4P) fuel management strategy. The L4P PBNP reactor cores will use a loading pattern that includes low power peripheral fuel assemblies and Peripheral Power Suppression Assemblies (PPSA's)

(Ref.28). The PPSA's are specially designed fuel assemblies that will be inserted on the core periphery to further reduce the fluence accumulation rate at specific reactor vessel welds.

The PPSA's will use the neutron absorber hafnium in the thimble tubes of the fuel assemblies. The hafnium will be a part-length design similar in mechanical design to the present Westinghouse hafnium Rod Cluster Control Assembly (RCCA) that is used in some Westinghouse plants.

In addition to its reduced fluence accumulation rate to the reactor vessel, the L4P provides PSNP with improved fuel utilization.

PBNP currently uses thimble plugging devices in some fuel assemblies to minimize core bypass flow through fuel assembly thimble tubes. The licensee states that the analysis will support the removal of these thimble plugging devices.

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3-y The current and proposed PBNP design parameters are as follows:

Current Proposed j

L Fuel Type (Westinghouse)

.STD, OFA STD, 0FA, upgraded 0FA i

Core' Power (MWt) 1518.5 1518.5 i

Average linear heat:.

i generation rate (kW/ft) 5.7 5.7

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System Pressure (psia) 2000(or2250) 2000(or2250) j 1

.CoreInletTemperature(*F) 545.3 545.3 Enthalpy rise hot channel peaking factor 11mit' (F-Delta H) 1.58 1.70 Total Peaking Factor Limit (F )

2.21 2.50 g

Totalthermaldesignflow(gpm) 178,000 178,000 Steam generator uniform tube 13%(Unit 1) plugginglevels(%)

14% (unit 2)

NOTE:

The LOCA and SGTR analyses used a 25% uniform tube plugging level ano the associated reduction in thermal design flow.

The proposed design provides for the removal of the third line segment of the Technical Specification K(Z) curve. ThisK(Z)curveisusedtoprovidethe required axial variation of the total peaking factor with core height such that at any core height the peaking factor limit will always be equal to or less than 2.50 (the Fn limit). The removal of the third line segment of the K(Z) curve is supportVd by the small-break LOCA analysis.

The staff has reviewed the design features and parameters proposed for future PBNP cores and concludes that they are acceptable because they are typical of the types of changes previously reviewed and approved for other plants-and because they lead to improvements in fuel utilization, fuel performance (for example, DFBN for the reduction of the passage of flow-entrained debris into the fuel assembly), and a reduction in the fluence accumulation rate to the reactor vessel.

j 2.2 Fuel Rod Design The increased power peaking factors affect the fuel rod design through increases in the steady-state fuel rod power histories and through the fuel rod transient duty. The licensee states that the fuel rod design criteria for the most limiting fuel rod design will be considered for PBNP including all cerbinations of Westinghouse STD, 0FA, and upgraded 0FA fuel. The fuel l

rod design criterie affected by this more sevore fuel duty are the rod internal pressure, cladding stress and strain, and cladding surface temper-l ature.

e t For the fuel rod internal pressure, Westinghouse uses an NRC-approved design l

limit that the internal fuel rod pressure of the lead fuel rod will be limited

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to a value below that which could cause (1) the diametrical gap to increase due to outward cladding creep during steady-state operation, and (2) extensive DNB propagation to occur. This fuel rod internal pressure limit is a function of system pressure. The design limit used for cladding stress is that the volume average effective stress is less than the Zircaloy 0.2% offset yield strength for Condition I and Condition II modes of operation, including the effects of temperature and irradiation. The design limit for cladding strain during steady-state operation is that the total plastic tensile creep and l

uniform cylindrical fuel pellet expansion due' to fuel swelling and thermal l

expansion are less than 1 percent from the irradiated condition.

For J

Condition II events the design limit for cladding strain is that the total tensile strain due to uniform cylindrical pellet thermal expansion during a transient is less than 1 percent from the pre-transient value.

The design limit applied to Zircaloy.claooing corrosion during steady-state and Condition II transients is to preclude a condition of accelerated oxidation. The controlling factor for Westinghouse reactors is the oxide-to-cladding interface temperature.

The fuel performance results for the PBNP are obtained using the approved PAD 3.3 (Ref. 12) and PAD 3.4 (Ref. 13) codes. The fuel rod design analysis is based on a best estinate plus uncertainty basis. The total uncertainty is based on a statistical convolution of the applicable individual uncertainties.

Appropriate power histories which define limiting duty for each of the fuel rod design criteria are used. The most limiting values of core inlet temperature and flow rate are used in the evaluations. The most limiting value of the system pressure (2000 or 2250 psla) for each of the fuel rod design criteria is also'used.

In addition to the fuel rod design criterie discussed above, the PBNP fuel will incorporate design changes to allow for extended burnup operation.

These changes are primarily concerned with the axial growth of fuel rods.

The staff has reviewed the fuel rod design for f uture reactor cores for PBNP and concludes that it is acceptable because (1) approved codes are used, (2) all applicable criteria are evaluated, and (3) the results for the increased power peaking f actors and increased fuel duty are acceptable.

2.3 Nuclear Design The licensee evaluated a reference core design that included the upgraded PBNP core features. A low-low leakage loading pattern (L4P) fuel management strategy was used. A cycle length of 10,500 mwd /l4TU was obtained through l

the use of 28 fresh fuel assemblies. Sixteen of the fresh assemblies were enriched to 4.0 w/o uranium-235. Twelve of the assemblics were enriched to 3.8 w/o uranium-235 and included 24 IFBA fuel rods per assembly for a total of 288 IFBA rods. The burnable poison coating of the IFBA rods was 56 inches in length and centered abcut the mioplane. All fuel assemblies, except the center assembly, contain axial blankets at the top and bottom of each fuel rod. The twelve assemblies on the core flats each contain a PPSA, with hafnium in the lower 6 feet of the guide tubes.

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The analysis was performed using the approved Westinghouse reload safety evaluation methodology (Ref. 14) and approved codes.

Because of the I

heterogeneous nature of this reference core design, a three dimensional core l

nodal model (Ref. 15) was used. The Relaxed Axial Offset Control (RA00) was l

performed with an approved methodology (Ref. 16).

The results of an analysis of the reference core design showed that the key safety parameters were insensitive to fuel type and primarily affected by the loading pattern. The results also indicated that future PBNP cores would require changes to the Technical Specifications for (1) an increase in the hotchannelfdctorlimit(F-DeltaH3),(3)andachangetotheallowablaflux total power peaking factor limit (F (2) an increase in the enthalpy rise difference operating envelope (RA0C delta flux band). The analysis assumed that the third segment of the Technical Specification K(Z) curve was removed (this will be confirmed later in our review of the small-break LOCA and large-

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break LOCA analyses). The licensee also changed the power-dependent rod insertion Technical Specification limits to ensure that the RA0C delta-flux difference band is conservative for future PBNP cores.

The staff has reviewed the nuclear design of the reference core and concludes that it is acceptable because (1) approved codes and methodologies have been used, (2) acceptable reactor core parameters have been obtained, and (3) appropriate changes to the Technical Specifications have been determined.

2.4 Thermal Hydraulic Design The licensee performed a thermal hydraulic analysis for the upgraded core features of PBNP. The analysis was performed for a nuclear enthalpy rise hot channel factor F-Delta H of 1.70 and for removal of thimble plugs. The increase in F-Delta H is the result of the L4P fuel management strategy. The increase in F-Delta H and the removal of the thimble plugs are accommodated by using the Departure from Nucleate Boiling Ratio (DNBR) design margin available in the safety analysis DNBR.

The current thermal-hydraulic analysis of 0FA fuel is based on the Improved Thermal Design Procedure (ITDP) (Ref. 17) and the Westinghouse WRB-1 critical heat flux correlation (Ref. 18). The analysis of the upgraded core features for future PBNP reloads is based on the Revised Thermal Design Procedure (RTDP) (Ref. 8) and the WRB-1 critical heat flux correlation.

However, for some transient events the Standard Thermal Design Procedure (STDP) is used witn the W-3 critical heat flux correlation.

The RTDP methodology removes some of the conservatism in the ITDP methodology by combining directly both system uncertainties and Departure from Hucleate Boiling (DNB) correlation uncertainty. The RTDP methodology safety analysis DNBR is 1.33 for both a typical cell and a thimble cell.

In addition this safety analysis DNBR includes 8.6 percent DNBR margin. The THINC IV code was used to perform the thermal-hydraulic calculations (Refs.19 and 20).

~6-The upgraded 0FA fuel is hydraulically identical to the OFA fuel and no transition core penalty is required. The use of STD fuel requires a small DNBR penalty on all the fuel. A rod bow penalty of less than 3% on DNBR 1s used in accordance with References 21, 22, and 23. This rod bow penalty is the maximum rod bow penalty for 14x14 0FA fuel at an assembly average burnup of 24,000 mwd /MTU. No rod bow penalty is taken for burnups greater than 24,000 mwd /MTU because credit is taken for the decrease in F-Delta H with 1

burnup.

The axial blankets and the increased allowable F affect the axial power n

distribution and, therefore, the DNBR.

For events not protected by the Overtemperature Delta-T (0 TDT) trip function, a limiting axial power i

distribution was used in the DNBR analyses. For these events, cycle-specific limiting axial power shapes will be evaluated and compared with this limiting axial power distribution for future PBNP reloads.

The licensee plans to remove the thimble plugs in addition to implementing upgraded core features. This removal of the thimble plugs results in an increase in the bypass flow from 4.5% to 6.5%.

There is a slight increase in the core flow rate which does not impact any mechanical design criteria. The removal of the thimble plugs results in a small decrease in DNBR margin. The licensee also evaluated the effect of thimble plug removal on fuel assembly hydraulic lift forces, fuel rod fretting wear, and control rod wear.

For these three areas, the licensee concluded that there were no significant effects caused by the thimble plug removal. The licensee has accounted for the increased bypass flow in both non-LOCA and LOCA safety analyses.

The staff has reviewed the thermal hydraulic design of the reference core and concludes that it is acceptable because (1) approved codes and methodologies have been used, (2) DNBR penalties resulting from the increase in peaking factor and removal of thimble plugs are offset by the present DNBR margin and the additional margin provided by the RTDP methodology, (3) rod bow penalty and any transition core effects are offset by DNBR margin available in the safety limit DNBR, and (4) all of the current thermal-hydraulic design criteria are satisfied.

2.5 Reactor Pressure Vessel Internals System Evaluation The licensee evaluated the effect'of removing the thimble plugs on the reactor I

pressure vessel internals system design requirements. Thimble plug removal f

leads to a reduction in core hydraulic resistance and to an increase in the portion of the bypass flow passing through the fuel assembly. The licensee's evaluations used operating, geometric and hydraulic characteristics of the PBNP with 14x14 0FA fuel and thimble plugs removed.

System pressures of 2000 and 2250 psia were considered.

The increased bypass flow in the fuel assembly and the reduction in core hydraulic resistance affects fluid system pressure drops, core bypass flow, baffle gap jetting momentum flux, closure head fluid temperature, internals component lift forces, and control rod drop times. The i

licensee determined that the core pressure drop would decrease by less than 10%, the total core bypass flow would be bounded by a value of 6.5%, the baffle

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gap jetting momentum flux is not adversely affected, the closure head fluid temperature is unaffected, the internals component lift forces are not adversely impacted (in fact they are reduced somewhat), and the control rod drop-times are not adversely impacted.

The staff has reviewed the licensee's evaluation of the reactor pressure vessel internals with respect to the thimble plug removal and concurs with the licensee's assessments.

2.6 Non-LOCA Accidents The licensee has evaluated the impact of the upgraded core features on the i

non-LOCA events presented in Chapter 14 of the PBNP FSAR. The licensee has used the approved reload core design methodology of Reference 14 and approved design codes.

In the present PBNP cores, thimble plugs are installed in all fuel assemblies which are not under control rod locations or do not contain neutron sources or burnable poisons. The removal of the plugs has two primary effects.

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increases the total core bypass flow and it reduces the core pressure drop somewhat. The events that have been reanalyzed have incorporated these effects. For the steamline break event end the mass and energy release to containment event, the licensee concludes that either the impact on the analyses is not significant or the conclusions of the previous analyses remain valid. Therefore, these two events were not reanalyzed by the licensee.

The licensee used the Revised Thermal Design Procedure (RTDP) in the analysis l

of a number of events. This extension of the ITDP methodology uses a safety 6nalysis DNBR limit of 1.33 for both typical and thimble cells. The licensee usedtheStendardThermalDesignProcedure(STDP)forthoseeventswhichdid not use the RTDP methodology.

The removable top nozzles (RTN) and debris filter bottom nozzles (DFBN) were designed to preserve core flow areas and loss coefficients. Therefore, no parameters important to the non-LOCA safety analyses were affected. The effects of integral fuel burnable absorbers (IFBA), axial blankets of natural uranium on the ends of the upgraded 0FA fuel rods, and extended burnup are taken into account in the reload design process. The effects of the L4P fuel management strategy and PPSA's (part length hafnium absorbers) in the core locations on the flats of the core periphery are also taken into account in the reload design process. These result in an increase in F-Delta H to 1.70 and F to 2.50 which impact the safety analysis.

q The licensee has also made a number of other design changes for future PBNP reloads. These changes include control rod power dependent insertion limits theeliminationofthethirdlinesegmentoftheTechnicalSpecificationK(Z}

curve, and a revised flux difference operating envelo The change to the rod insertion limits impacts (1) shutdown margin, (2)pe.tripreactivity,(3) j powerdistributionlimits,(4)ejectedenddroppedrodworths,(5) post l

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. ejected rod peaking factors, and (6) differential rod worths. The licensee states that nuclear design calculations with the proposed rod insertion limits ensure that the values for shutdown margin, trip reactivity, dropped rod worths, and differential rod worths assumed in the non-LOCA safety analyses are valid. The effects of new rod insertion limits have been included in the power distribution limits and rod ejection parameters for the affected events that were reanalyzed.

In addition, the shutdown margin and power distribution assumptions used in the steamline break analysis remain valid with the proposed rod insertion limits. The proposed changes to the K(Z) curve and the flux j

difference operating envelope could impact the power distribution assumptions used for the non-LOCA analyses. The licensee states that nuclear design analyses show that the power distribution assumptions of the non-LOCA analyses are ensured by the proposed K(Z) curve and the proposed flux difference operating envelope.

j The effect of the proposed power-dependent F-Delta H. limit does.not directly' affect' the system' transient response of the PBNP. This is because the PENP system response is determined with a point kinetics system coce which does not directly use F-Delta H as an input quantity.

Instead, the F-Delta H power dependent limit is used to determine DNBR for those events for which DNB is the acceptance criterion, once the plant's systems response has been determined. The licensee splits the DNBR limited events into two categories.

The first category includes those events in which the power-dependent value of F-Delta H is indirectly taken into account by the core limits. The second category includes those events which directly assumes the power-dependent value of F-Delta H in the analysis.

For events in the first category, the licensee used new Overtemperature Delta-T(0 TDT)andOverpowerDelta-T(0PDT)setpointequationswhichinclude the reviseo F-Delta H limit of 1.70.

Events which require either the OPDT or the OTDT trip functions were reanalyzed. These FSAR events are:

FSAR Section Event 14.1.2 Uncontrolled RCCA Withdrawal at Power 14.1.6 Reduction in Feedweter Enthalpy Incident 14.1.7 Excessive Load Increase Incident 14.1.9 Loss of External Electrical Load The uncontrolled rod cluster control assembly (RCCA) withdrawal at power event was reanalyzed at various power levels and reactivity insertion rates, for both minimum and maximum reactivity feedback cases. This transient is terminated by a reactor trip on either High Neutron Flux or Overtemperature Delta-T trip functions. The results of the reanalysis indicate that DNBR never falls below the safety analysis DNBR value. The DNB design basis has, therefore, been met.

The reduction in feedwater enthalpy event is bounded by the excessive load increase event and was not reanalyzed. The excessive load increase event was reanalyzed for both beginning-of-cycle (B0C) and end-of-cycle (E0C) conditions,

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I with and without automatic rod control. The results show that, for all cases, the reactor does not trip'but reaches a new equilibrium state. The DNBR value remains above the DNBR safety analysis value. The DNB design basis has, therefore, been met for this event.

The loss of external electrical load event was reanalyzed at both BOC and EOC' conditions, with and without pressurizer control and with the reactor in manual control. The reanalysis of this event shows that the DNBR remains above its safety analysis limit value.

In addition, the peak reactor coolant system pressure and secondary side pressure remain below 2500 and 1100 psia, respectively, that is, below the design pressure values. The DNBR value remains above. the safety analysis DNBR design limit for all. of the cases -

analyzed. The system pressure and DNB design bases have, therefore, been met for this event.

For events in the second category, the increased value for F-Delta H was used in the analysis of the following FSAR events:

FSAR Section Event 14.1.1 Uncontro11ec RCCA Withdrawal from Subcritical Condition 14.1.3 RCCA Drop 14.1.5 Startup of an Inactive Reactor Coolant Loop 14.1.8 Loss of Reactor Coolant Flow In general, an increase in F-Delta H results in a decrease in DNBR for a given set of thermal-hydraulic conditions.

The uncontrolled RCCA withdrawal from subcritical conditions event was analyzed assuming the most limiting axial ard radial power shapes associated with having the two highest combined worth sequential control rod banks in their highest worth position. The maximum withdrawal speed of 45 inches / minute is assumed in the analysis for a reactivity insertion rate of 100 pcm/second. The results of the reanalysis indicote that DNBR remains above the safety analysis DNBR limit.

Therefore, the DNB design basis for this event has been met.

The RCCA drop event consists of two separate events. These events are (1) a roddropevent,and(2)amisalignedrodevent. The analysis of the rod drop event was performed with an unapproved Westinghouse rod drop methodology (Ref.

24). The staff requested that an analysis be performed with an approved methodology. The licensee submitted an analysis based on the methodology currently used for PBNP (Ref. 28). A number of dropped rod cases were evaluated with respect to the DNBR design basis and acceptable results were obtained. The misaligned rod event results in an increase in the radial heat flux hot channel factor. However, the safety limit DNBR design limit is met.

The DNB design basis has, therefore, been met for the misaligned and dropped rod events.

The startup of an inactive reactor coolant loop event was analyzed at a reactor core power of 10% of full rated power.

The event is a reactivity

excursion which causes an increase in core heat flux. The transient is a relatively mild event with DNBR remaining above the safety analysis DNBR design limit. The DNB design basis has, therefore, been met for this event.

The loss of reactor coolant flow event consists of two separate events:

(1) the two-pump coastdown event, and (2) the one-pump coast down event. The loss of reactor coolant flow event caused by a locked rotor is reviewed elsewhere in this Safety Evaluation. The results of the analysis indicate that the safety analysis DNBR design limit has been met for these two flow coastdown events. The DNB design basis has, therefore, been met for this event.

In addition, the licensee considered the effect of the increase in F-Delta H on the steamline break accident which is discussed in PBNP FSAR Section 14.2.5.

The licensee performed the analysis of this event at hot, zero power conditions with the most reactive control rod stuck in its fully withdrawn position.. The increase in the power-dependent F-Delta H limit results in an increase in the stuck rod power peaking factor at zero power, with a resulting decrease in DNBR. The licensee states that its analysis shows that the safety analysis DNBR design limit is met. The licensee also concludes that the mass and energy release to containment event is not impacted by the increase in F-Delta H because the primary to secondary heat transfer characteristics of the event are not affected. The licensee concludes that this event is not impacted by the increase in F-Delta H.

The licensee analyzed two events affected by the increase in F.

These events q

are:

FSAR Section Event 14.1.8 Locked Rotor 14.2.6 Rod Ejection The locked rotor event is classified as an accident. The results of the analysisshowthatthemaximumreactorcoolantsystempressure(RCS)is2744 psia, the maximum claading temperature is 2166*F, the amount of zirconium-water reaction is 1.30% by weight, and less than 86% of the fuel rods in the core undergo DNB.

Because RCS pressure remains below the faulted condition pressure and because the core remains in place with no consequential loss of core cooling capability, the locked rotor event, therefore, meets all applicable safety criteria.

l The rod ejection event is also classified as an accident. For all of the cases analyzed, the maximum fuel stored energy is less than 200 cal /gm, and the maximum fuel melt at the hot spot is less than 10%. The analysis of the rod ejection event, including the effect of an increased peaking factor F,

J shows that the applicable criteria for this event have been met.

0 The boron dilution event was reanalyzed by the licensee although it was not

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directly affected by the upgraded core features. Dilution events were analyzed for refueling, startup, and power operation. The results obtained show that it 3

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l 1 l would take at leest 30.1 minutes before the loss of shutdown margin for the l

refueling dilution event. About 18.8 minutes would be avaflable for operator action before the reactor would become critical for the startup dilution event.

At least 16.2 minutes would be available for operator action for a boron l

dilution event during power operation. The results show that, for all cases, sufficient time is available for the operator to determine the cause and take corrective action before the required shutdown margin is lost.

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The loss of normal feedwater event was reanalyzed by the licensee even though it was not directly affected by the upgraded core features. The reactor is protected by a reactor trip either on low-low water level in either steam generator or on steam flow-feedwater flow mismatch coincident with low water level in either steam generator and by the auxiliary feedwater system. The I

results of the analysis show that the auxiliary feedwater system provides sufficient flow to the two steam generators to maintain heat transfer 4

capability to prevent water relief from the reactor coolant system relief or

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safety valves. Therefore, the loss of normal feedwater event does not lead to I

any adverse core conditions.

The loss of all AC power to the station auxiliaries was reunalyzed by the i

licensee even though it was not directly affected by the upgraded core features. The assumptions used in the analysis are similar to those for the loss of normal feedwater event except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip. The results of the analysis show that the natural circulation flow that is available is sufficient to provide adequate decay heat removal following reactor trip and reactor coolant pump coastdown.

No water relief occurs for this event through the pressurizer relief or safety valves. Therefore, the loss of AC power to the station auxiliaries event does not result in adverse core conditions.

The licensee reevaluated fuel handling accidents. The following conservative

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assumptions were made:

(1) all fuel rods in on assembly are assumed to be dameged, (2) the assembly power is assumed to be 1.8 times the core average assembly power (3) fission products released from the assembly consist of 3.6% halogens [as1-131)and30%noblegases(asKr-85)(thesevaluesare based on a conservative axial power distribution of 1.87 peak to average, corresponding to a peak linear assembly power of 15.6 kW/ft), (4) of the halogens released only 0.01 escape from the spent fuel pool surface to the environment,andf5)ofthenoblegassreleased,100%escapethespentfuel pool surface to the environment. The 2-hour site boundary thyroid dose is estimated to be 17.5 rem, based on the above assumptions. This dose is much less than the 10 CFR Part 100 guideline value of 300 rem. The integrated whole body dose for distances beyond the site boundaries is less than 10 rem, which is less than the 10 CFR guideline value.

The staff has evaluated the licensee's evaluation and analysis of the non-LOCA events, using the revised safety analysis assumptions associated with the upgraded core features, and concludes that they are acceptable because (1) approved methodologies and computer codes have been used, and (2) all applicable safety criteria have been met.

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2.7 Large-Break and Small-Break LOCA Analyses 2.7.1 Lerge-Break LOCA Analysis 1

The Point Beach Nuclear Plant (PBNP) is a Westinghouse-designed two-loop plant equipped with a low-pressure upper plenum injection (UPI) system as part of the emergency core cooling system (ECCS). The previous PBhP ECCS evaluation model assumed that the UPI water fell directly into the lower plenum without interaction with the core, and could th'erefore be treated as if it were a cold-leg injection plant.

In support of the proposed TS change to increase the power peaking factors, the licensee in a ' letter dated November 30, 1988 (Ref. 4) provided a new large-break LOCA (LBLOCA) analysis described in Addendum 2 to WCAP-10924-P, Revision 1, Volume 2.

This LBLOCA analysis uses a new Westinghouse ECCS' evaluation model developed for application to the two-loop UPI plants. This new ECCS mod:1, described in Westinghouse topical repcrt WCAP-10924-P (Ref. 5), uses a best-estimate thermal-hydraulic code WCOBRA/ TRAC and the approach described in SECY 83-472 (Ref. 25).

In using the SECY 83-472 approach, an estimate of the 95th percentile peak cladding temperature (PCT) is calculated using a best-estimate code and accounting for the uncertainties associated with the code ano application. Another calculation is also required to determine the " Appendix K PCT" by applying all the required features set forth in Appendix K to 10 CFR Part 50. The

Appendix K PCT" must then be shown to be greater than the 95th percentile PCT and remains below the 2200*F acceptance criterion. WCAP-10924-P has been reviewed and approved by NRC for referencing in the licensing calculations (Ref. 6), and has been used by Northern States Power Company for application to the Prairie Island (PI) unit which was the leso plant in using the method-ology of WCAP-10924-P.

Addendum 2 to Volume 11 of WCAP-10924-P provides the Point Beach plant-specific analysis to demonstrate that the methoc of analysis complies with the SECY 83-472 guidelines and Appendix K requirements, and that the acceptance criteria of 10 CFR Part 50.46 are not violated with the proposed higher peaking factors, i.e., the enthalpy rise factor, F-Delta H, of 1.70 and the total peaking factor, F, of 2.50.

The analysis follows the same procedure described in Volume II of WOAP-10924-Pwhichwasconewiththedataoftheleadplant,PrairieIsland.

In the PBNP plant modeling, the primary and secondary loop models are the same as the PI lead plant molel. The reactor vessel model follows the same approach and details of the PI unit in using the four-channel core model, but accounts for the differences in the reactor internals between the two plants. The major differences in the reactor internals are in the upper plenum configuration.

For example, the PBNP unit has free-standing mixers above some of the open holes on the upper core plate whereas the PI units have no free-standing mixer; and the PBNP has a flat upper support plate compared to the inverted l

top hat upper support plate for the PI units. Also the core barrel-baffle arrangements are such that, during steady state, the barrel-baffle flow is an upflow in the PBNP unit compared to a downflow in the PI units. These differences are reflected in the plant modeling.

In addition, since the FBNP upper plcnum configuration is different from the lead plant, a sensi-tivity study is required, as specified in the staff safety evaluation report

l

for WCAP-10924-P, to determine the upper plenum structure under which the hot l

assembly and hot rod would be placed to obtain the highest PCT. This sensi-tivity study, described in Section 5-3 of Addendum 2, is performed using a three-channel core configuration as wos done for the lead plant. The result of the sensitivity study justifies the locations of the hot assembly and hot rod in the PBNP reactor vessel model.

For the four-channel core model, the outer low power channel uses a conservatively high power factor to represent the flatter radial power profile expected for the PBNP reload designs. The use of flatter radial power profile is conservative because a sensitivity study has shown that'it will result in poorer core cooling and therefore higher PCT.

In accordance with the methodology of WCAP-10924-P, the PCT's are calculated for both the blowdown and reflooo peaks. The calculations are made for the realistic nominal condition, superbounded condition, and with Appendix K requirements. The 95th percentile PCT's at the blowdown and reflood peaks are obtained from the superbounded PCT's plus the respective code and application uncertainties.

In order to perform the superbounded calculation, conservative bounding values and assumptions of some plant parameters and models are used in the calculation.

Sensitivity studies would be necessary to determine the directions of conserva-tisu for the parameter uncertainties or assumptions, i.e., the directions to place the uncertainties and conservative assumptions that would result in higher PCT. The licensee asserted that the sensitivity studies performed in Volume II of WCAP-10924-P for the PI lead plant are bounding for the PBNP unit because the PI units have higher core power to ECCS flow ratio and therefore a greater PCT sensitivity.

In addition, only the direction of conservatism, instead of the magnitudes of the PCT sensitivity, of the parameters and assumptions are used in placing the conservative bounding values and conditions for the superbounded calculations. The stuff agrees with this observation that the PI sensitivity study results are applicable to the PBNP superbounded PCT calculation.

With regard to the Appendix K PCT calculation, the staff, in the evaluation report for acceptance of WCAP-10924-P for licensing application to Westinghouse two-loop UPI plants, required that the UPIrlicensees apply for exemptions to Items I.D.3 and I.D.5 of Appendix K to 10 CFR Part 50. These exemptions are necessary because Item I.D.3, which requires the use of a carryover fraction to calculate the reflood core exit fluid flow, and Item I.D.5, which sets specific requirements for refill and reflood heat transfer calculation, were intended for the conventional cold-leg injection plants and are not applicable to the UPI plants. The licensee, in its November 30, 1988 letter, requested an exemption to these two requirements and the exemption request has been granted (Ref.26).

The analysis was performed with the proposed enthalpy rise factor and total peaking factor of 1.70 and 2.50, respectively, and assuming a full cure of 14x14 optimized fuel assemblies (OFA) and a steam generator tube plugging level of 25 percent.

In addition, since PBNP units are licensed to operate at the nominal reactor system pressures of 2250 and 2000 psia, the nominal pressure of 2250 psia was used in the analysis. This is because the sensitivity study indicoted that the operating pressure of 2250 psia produced the highest PCT,

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and therefore the analysis using 2250 psia would be a bounding analysis. The results show a reflood PCT of 2023 F for the Appendix K calculation.

This PCT is higher than the 95th percentile PCT of 1932*F and 1892 F, respectively, for the blowdown and reflooo peaks, and below the 2200 F acceptance criterion.

In addition, the Appendix K calculation results shown in Table 6-4 of Addendem 2 indicate that both maximum local cladding oxidation and total hydrogen generation are below the acceptance criteria of 17 percent and 1 percent, I

respectively.

Since the analysis assumed a full core of 14x14 0FA fuel, this is inconsistent with the actual fuel loading of a transitional mixed core of standard, OFA and upgraded 0FA fuel assemblies, and an adjustment for the calculated PCT may be needed to account for the neglect of the effect of the hydrodynamic mismatch among the different fuel designs.

However, since both (1) the OFA and the u

{

graded 0FA fuel designs have the same hydrodynamic characteristics, and (2) p-the standard fuel has higher flow resistance (but is not the limiting fuel assembly), and would therefore increase flow into the more limiting 0FA fuel types, the analysis assuniption of a full core of 14x14 0FA fuel bounds the mixed core effects.

2.7.2 Small-Break LOCA Analysis The licensee has perforned a reanalysis of small-break LOCA (SBLOCA) using the approved method of WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP code." The analysis assumed a full core of 14x14 0FA fuel and 25 percent steam generator tube plugging, and used the proposed peaking factors of 1.70 and 2.50, respectively, for the enthalpy rise factor and total peaking factor.

In addition, the third segment of the K(Z) curve was not used, consistent with the proposed TS change. The analysis was performed for a 4-inch cold-leg break. Use of this break size and location as the limiting case was bosco on a previous generic Westinghouse two-loop plant analysis and Prairie Island SBLOCA analysis using the NOTRUMP code. The RCS pressures of both 2250 and 2000 psia were analyzed, the analyses demonstrated the 2000 psia case was limiting. The analysis result of the 2000 psia case shows a PCT of 809'F, far below the 2200*F acceptance criterion. Therefore, there is no concern that a SBLOCA would result in violation of the ECCS acceptance criteria of 10 CFR 50.46.

2.7.3 Staff Position on PBNP LOCA Analyses The staff has reviewed both LBLOCA and SBLOCA analyses in support of the PBNP technical specification changes for increased peaking factors, and concludes that the ECCS acceptance criteria set forth in Section B of 10 CFR 50.46 have been complied with.

2.8 Steem Generator Tube Rupture Accident Analysis The licensee reanalyzed the steam generator tube rupture (SGTR) event for PBNP using the same methodology as in the existing SGTR analysis with two key changes in the assumptions. These are (1) the increased peaking factors proposed by the licensee for PBNP and (2) that both safety in,)ection pumps

will run for 30 minutes to reflect a change in the safety injection (SI) termination portion of the SGTR recovery procedures. The reanalysis generates maximum radiological doses of 2.13 rem to the thyroid and 0.059 rem to the whole body. Although slightly higher than doses cited in the existing FSAR for the SGTR Accident Analysis (0.700 rem to the thyroid and 0.200 to the whole body), the doses calculated for the reanalysis of the SGTR accident remain a small fraction of the 10 CFR Part 100 exposure guidelines and are therefore acceptable. Because the methodology used in this reanalysis is the same as the staff-approved methodology used in the existing analysis, the methodology used" remains acceptable.

Departure from nucleate boiling (DNB) is not approached in the design basis i

SGTR analysis. Thus, the increase in peaking factors maintains acceptable results in tne SGTR scenario.

2.9 Technical Specifications (1) Specification 15.2.1.1 and Basis - This Specification and Basis were rewritten to eliminate reference to the transition :, ore safety limits.

The changes to page' ;5.2.1-1 and 15.2.1-2 are acceptable because future PBNP cores will not require transition core penalties, as discussed in the new PBNP safety analysis.

(2) Figures 15.2.1-1 and 15.2.1-2 are replaced with a revised Figure 15.2.1-1.

This is acceptable because a figure related to transition cores is removed and a new figure which corresponds to the new PBNP safety analysis is included.

(3) Specification 15.2:3.1.B.3 - The removal of the asterisk and footnote in

• his specification is acceptable because the safety analysis was I

performed at a bounding value of the pressure.

(4) Specifications 15.2.3.1.B.4 and 15.2.3.1.B.5 - These specifications have been revised to reflect the new setpoints used in the PBNP safety analysis. The changes are, therefore, acceptable.

(5) Specification 15.2.3 Basis - The references to the transition core are removed from page 15.2.3-6.

This is acceptable because transition core penalties will no longer be required for the PBNP.

j (6) Specification 15.3.1.C B6 sis - The assumed steady-state primary-to-secondary steam generator leakage rate is changed to make it consistent with the more conservative value in Specification 15.3.1.D.4.

The new

]

steady-state leakage rate is used as an input to the steam generator tube rupture event. This change is, therefore, acceptable, i

(7) Specifications 15.3.1.G.1 and 15.3.1.G.2 - The change to Specification 15.3.1.G.1 is made to reflect the value used in the PBNP safety analysis. The change to Specification 15.3.1.G.2 is made to reflect the i

location where the pressure indication is taken. The footnote is removed to reflect the fact that the safety analysis was performed at a bounding value of the pressure. These changes are, therefore, acceptable.

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ - = -

7 (8) Specification 15.3.3 Basis - An addition is made to the Basis to describe the basis for the RWST minimum volume and minimum boron concentration.

This change, is therefore, acceptable.

(9) Specification 15.3.10.B.1.a - This specification was changed to reflect the value of F9 qual to 2.50 and the value of F-Delta H equal to 1.70 e

that were used in the PBNP safety analysis. These changes are, therefore, acceptable.

1 (10) Specification 15.3.10 Basis - The Basis to Specification 15.3.10 is changed to reflect the new values of Fn and F-Delta H.

A clarification was also made to the basis of the Hot thannel Factor Normalized Operating Envelope and its use in the safety analysis. These changes are, therefore, acceptable.

(11) Specification 15.3.10-Figure 15.3.10-1 on control bank insertion limits was revised to obtain a wider delta-flux band. This change is acceptable because it is used in the PBNP safety analysis.

e (12) Specification 15.3.10 - Figure 15.3.10-3 on Hot Channel Factor Normalized AxialOperatingEnvelope(K(Z) curve)isrevised. This revised figure is acceptable because it results in an acceptable small-break LOCA analysis as well as other safety analyses.

(13) Specification 15.3.10-Figure 15.3.10-4 on Flux Difference Operating Envelope (delta-I band) is revised. This change is acceptable because adherence to the delta-I band limits will ensure that the power distribution limits of the safety analysis will be enforced.

(14) Specifications 15.5.3.A.2, 15.5.3.A.3, 15.5.3.A.4, and 15.5.3.A.5 -

Specification 15.5.3,A.2 has been revised and the current Specifications 15.5.3.A.3, 15.5.3.A.4 and 15.5.3.A.5 have been deleted to revise the description of the reector core by eliminating reference to a transition core. This is acceptable because transition cores will no longer be usec in future PBNP reloads.

(15) IFBA Description Addition - 5) edification 15.5.3.A.6 is deleted and Specification 15.5.3.A.3 in tie Reactor Design Features section of the Technical Specifications is inserted to describe the integral fuel burnable absorbers (IFBA's). This change is acceptable because it is one of the upgraded core features in future PBNP reloads.

(16) Specification 15.5.3.A.7 has been renumbered to 15.5.3.A.4.

This change is acceptable because it is an administrative change.

(17) Water Displacer/ Neutron Source Description Addition - Specification 15.5.3.A is revised to inciade a description of water displacer rods.

Specification 15.5.3.A.5 is added to describe the neutron source assemblies. The addition of the water displacer rods is acceptable because these water displacer rods have been previously used at PBNP.

lhe addition of the description of the neutron source assemblies is i

l acceptable because it provides a necessary description of a core l

feature.

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(18) Peripheral Power Suppression Assemblies Description Aodition - Specif1-cation 15.5.3.A.6 is added to describe.the peripheral power suppression assemblies (PPSA's) to be used in future PBNP reloads. The addition of the description of the PPSA's is acceptable because it provides a necessary description of a core feature.

3.0 FINDINGS The staff has reviewed the request by the Wisconsin Electric Power Company to operate the Point Beach Nuclear Plants, Units I and 2, with upgraded core features, including an increased total power peaking factor (F ) of 2.50 and 0

an increased enthalpy rise hot channel factor (F-Delta H) of 1.70. Based on this review, the staff concluded that appropriate material was submitted and l

that nortaal operation and the transients and accidents that were evaluated and reanalyzed are acceptable. The Technical Specifications submitted for this license amendment suitably reflect the necessary modifications for the.

cperation of' future P8NP reloads.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact has been prepared ano published in the Federal Register on March 28, 1989 (54 FR 12696). Accordingly, based upon the environmental assessment, the Commission has determined that the issuance of these amendments will rot have a significant effect on the quality of the human environment.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of the amendments will r.ct be inimical to the common defense and security or to the health and safety of the public.

i Principal Contributor: Daniel Fieno, Yi-Hsiung Hsii, Maria Angelt.s Gilbert 1

Dated: May 8, 1989 l

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6.0 REFERENCES

-1.

Letter (VPNPD-88-441) from C. W. Fay (WEPCO) to USNRC, dated August 26, 1988.

2.

Letter (VPNPD-88-526) from C. W. Fay (WEPCO) to USNRC, dated October 28, 1988.

3.

" Reference Core Report: Vantage 5 Fuel Assembly," WCAP-10444-P-A, September 1985.*

4.

Letter from C. W. Fay (WEPCO) to USNRC, " Dockets 50-266 and 50-301, Large-Break Loss of I.,oolant Accident Analysis for Technical Specification Change Request 127, increased Allowable Core Power Peaking Factors, Point Beach Nuclear Plant, Units 1 and 2," VPNPD-88-581, NRC-88-119, November 30, 1988.

5.

WCAP-10924-P, " Westinghouse Large Break LOCA Best Estirate Methodology,"

Volume I, June 1986, Volume 11 Revision 1, April 1988, Volume I Addendum 1 April 1988, Volume II Addendum 1, April 1988, Addenda 2 and 3, July 1988.

6.

Letter from Ashok C. Thadani (NRC) to W. J. Johnson (Westinghouse),

subject;

" Acceptance for Referencing of Licensing Topical Report, WCAP-10924, ' Westinghouse Large Break LOCA Best-Estimate Methodology,'"

August 29, 1988.

7.

N. Lee, et al., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.

8.

Friedland, A.J. and Ray, S., " Revised Thermal Design Procedure,"

WCAP-11397(Proprietary) and WCAP-11398 (non-Proprietary), February 1987.

9.

Letter from Ashok C. Thadani (NRC) to W. J. Johnson (Westinghouse),

subject:

" Acceptance for Referencing of Licensing Topical Report WCAP-11397, ' Revised Thermal Design Procedure,'" January 17, 1989.

10.

" Safety Evaluation for Increased Peaking Factors and Fuel Upgrade at Point Beach Nuclear Plant, Units 1 and 2," August 1988 (Attachment I to Reference 1).

l

11. " Final Report for Increased Peaking Factors and Fuel Upgrade Analysis -

Point 8each Nuclear Plant, Units 1 and 2," October 1988 (Attachment to Reference 2).

I

12. Miller, J. V., (Ed.), " Improved Analytical Models Used in Westinghouse l

Fuel Roo Design Computations," WCAP-8720, October 1976 (Proprietary) and WCAP-8785, October 1976 (Non-Proprietary).

l

7 i 13. Weiner, R.A., et al., " Improved Fuel Performance Models for Westinghouse Fuel Rod Design end Safety Evaluations," WCAP-10851-PA, (Proprietary) and WCAP-11873-A (Non-Proprietary), August 1988.

14. Davidson,S.L.(Ed.),etal.

" Westinghouse Reload Safety Evaluation Methodology,"WCAP-9272-P-A{ Proprietary)andWCAP-9273-A (Non-Proprietary), July 1985.

15. Davidson,S.L.(Ed.),etal.,"ANC: Westinghouse Advanced Nodal Computer Code," SCAP-10965-P-A (Proprietary) and WCAP-10966-A (Non-Proprietary), September 1986.

16. Miller, R.W., et al., " Relaxation of Constant Axial Offset Control,"

l WCAP-10216-P-A (Proprietary) and WCAP-10217-A (Non-Proprietary), June 1983.

17. Chelemer, H., et al., " Improved Thernal Design Procedure," WCAP-8567-P (Proprietary) and WCAP-8568 (Non-Proprietary), July 1975.

18. Motley, F.E., et al., "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids " WCAP-8262-P-A (Proprietary) and WCAP-8763-A (Non-Proprietary), July 1984.

19. Hochreiter, L.E., Chelemer, H.,

Chu, P.T., "THINC IV, An Improved Program for Thermal Fydraulic Analysis of Rod Bundle Cores," WCAP-7956, June-1973.

-20.

Hochreiter, L.E., " Application of the THINC IV Program to PWR Design,"

WCAP-8054 (Proprietary) and WCAP-6195 (Non-Proprietary), October 1973.

21. Skaritka, J., (Ed.) " Fuel Rod Bow Evaluation," WCAP-8691, Revision 1 (Proprietary) and WCAP-8692 (Non-Proprietary), July 1979.

22. " Partial Response to Request Number 1 for Additional Information on WCAP-8691, Revision 1" letter, E.P. Rahe, Jr. (Westinghouse) to J.R.

Miller (NRC), NS-EPR-2515, dated October 9,1981; " Remaining Response to Request humber 1 for Additional Information on WCAP-8691, Revision 1" letter, E.P. Rahe, Jr. (Westinghouse) to J.R. Miller (NRC). NS-EPR-2572, dated March 16, 1982.

23. Letter C. Berlinger (NRC) to E.P. Rahe, Jr. (Westinghouse), " Request for Reduc. tion in Fuel Assembly Burnup Limit for Calculation of Maximum Rod Bow Penalty," June 18, 1986,

24. Haessler, R.L. et. al., " Methodology for the Analysis of the Dropped Rod Event," WCAP-11394 (Proprietary), WCAP-11395 (Non-Proprietary), April 1987.

25.

Information Report from Willi 6m J. Dircks to the Commission, " Emergency Core Cooling System Analysis Methods," SECY-83-472, November 17, 1983.

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26. EnclosuretoLetterfromW.Swenson(NRC) toc.W. Fay (WEPCO), dated March 8, 1989, (54 FR 11095).

27. Letter (VPHPD-88-619) from C.W. Fay (WEPCO) to USNRC, dated December 23, 1988.

28, Letter (VPND-89-006) from C,W. Fay (WEPCO) to USNRC, dated January 17, 1988.

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