ML20247E615

Safety Evaluation Supporting Amend 139 to License DPR-65

ML20247E615
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Site:	Millstone
Issue date:	03/20/1989
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.139 TO FACILITY OPERATING LICENSE NO. DPR-65 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO.

50-336

1.0 INTRODUCTION

By letters (Refs. I and 2) dated August 26, 1988 and November 15, 1988, Northeast Nuclear Energy Company (NNECO), the lice.1see for the Millstone Unit 2, submitted its safety analyses to support Cycle 10 operation. The reload application involves three fuel design related issues:

(1) the replacement of 60 spent fuel assemblies with 60 fuel assemblies provided by Advanced Nuclear Fuels Corporation (ANF formerly Exxon Nuclear), (2) the analysis of safety considerations involved in the determination of Cycle 10 operating limits, and (3) incorporation ui' new limits on the linear heat generation rate (LHR).

In support of the reload application, the licensee also provided a reload analysis submittal and supporting analyses. These submittals are as follows:

1.

License amendment to Technical Specifications (Ref. 2).

2.

Cycle 10 safety analysis report (Ref. 3).

3.

Fuel Design Report (Ref. 4).

4.

Transient analysis reports (Refs. 5 and 6).

5.

Steam Line Break analysis report (Ref. 7).

6.

Small break loss-of-coolant-accident (LOCA) analysis report (Ref. 8).

7.

Large break LOCA analysis report (Ref. 9).

The proposed Technical Specification (TS) changes and the supportin were based on an assumption that a minimum reactor coolant system (g analyses RCS) flow rate of 340,000 gpm, the current Technical Specification limit, can be maintained through the Cycle 10 operation. Subsequently, the licensee indicated (Ref. 29) that the RCS flow has been observed to be decreasing during the past several cycles of operation. As a result, the licensee stated that the margin between operating RCS flow rate and the existing TS limit is very small.

In order to increase the margin in the RCS flow, the licensee submitted the request for the TS change to allow operation of the Cycle 10 core with a minimum RCS flow rate of 325,000 gpm. A supporting analysis (Ref. 30)

ANF-89-011 (Millstone Unit 2, Reduced Flow, Standard Review Plan - Chapter 15 Event Analysis) was also submitted for review. The supporting analysis is to address the effect of the reduced RCS flow on the transient behaviors and the l

operating safety limits.

In Reference 30, the licensee identified and analyzed the events affected by the reduced RCS.

P

i The licensee's object ye for these submittals is to demonstrate that the proposed Technical Sper ification (TS) changes are appropriately reflected in the assumptions used in the Cycle 10 design and the analyses of transients and LOCAs in order to support its position that the Millstone Unit 2 can be operated safely at a rated core thermal power of 2700 MWt throughout Cycle 10.

The NRC staff has reviewed these submittals with analyses based on the current TS RCS flow rate of 340,000 gpm and the reduced RCS flow rate of 325,000 gpm for the reload application and prepared the evaluation as follows.

2.0 EVALUATION 2.1 Reload Description Millstone Unit 2 core consists of 217 assemblies, each having a 14x14 fuel rod array. The assemblies are composed of up to 176 fuel rods, 4 control rod guide tubes and I center control rod guide tube / instrument tube. The Millstone 2 Cycle 10 (H2C10) core will retain 157 Westinghouse assemblies from the previous cycle and add 60 unirradiated ANF fuel assemblies referred to as Batch M, ANF-1. The average enrichment for the added rods containing no burnable absorbers is 3.30 W/o U-235 (3.00 W/o around the guide and instrument tubes and 3.45 W/o elsewhere).

For the added rods containing 1.0 W/o gadolinia the enrichment is 2.85 W/o U-235.

For rods containing 6.0 W/o gadolinia the enrichment is 2.10 W/o U-235. The total batch burnable absorber requirement for ANF fuel is 576 gadolinia bearing rods.

The' fresh fuel is scatter-loaded throughout the core.

The fresh assemblies loaded in the core interior contain gadolinia-bearing fuel in order to control power peaking and reduce the initial boron concentration to maintain the moderate temperature coefficient (MTC) within its Technical Specification limit. The exposed fuel is also scatter-loaded in the center in a manner to control the power peaking.

2.2 Fuel Design The fresh ANF fuel assemblies are the first such ANF assemblies to be used in Millstone Unit 2.

The design of ANF fuel assemblies are described in Reference 10, " Generic Mechanical Design Report - Exxon 14xla Fuel Assemblies for Combustion Engineering Reactors," which was previously approved by the NRC.

l The licensee submitted, in Reference 4, the analysis pertinent to the Millstone Unit 2 ANF-1 reload fuel assemblies which are not covered in the generic report (Refs.10 & 28).

The analysis in Reference 4 is done using the NRC approved j

methods in References 11 and 12. " Qualification of Exxon Nuclear Fuel for Extended Burnup."

We have evaluated the analysis for fuel performance in Reference 4 and found that the fuel design parameters were calculated by using the NRC approved I

methods and were within expected ranges of a typical ANF fuel assembly in a PWR core. Therefore, we conclude the fuel design acceptable.

I

. l The proposed linear heat generation rate (LHGR) limit for the M2C10 reloa.' is 15.1 Kw/ft. Since this approved limit is obtained by ANF using the NRC approved EXEM/PWR evaluation model (Ref.13), is well below the design limit of 21 Kw/ft, and thus is acceptable.

2.3 Nuclear Design The nuclear design for the Millstone Unit 2 Cycle 10 reload has been performed by ANF with the approved methodologies (Ref. 12), which include the MICBl'RN-2/CASMO-2E and yTGPWR codes.

CASMO-2E is a two-dimensional transmission probability Lode for burnup calculations on PWR assemblies or pin cells.

MICBURN-2 is a multi-group one dimensional transmission probability code which calculates the microscopic burnup in an absorber rod containing initially homogeneously distributed gadolinia and generates effective cross-sections as a function of the gadolinia number density to be used in CASM0-2E assembly depletion. The MICBURN-2/CASM0-2E code generates the cross-sections to the XTGPWR code (Ref.14) which determines power and exposure distribution, reactivity feedback characteristics and cold shutdown margin.

The results of the reload analyses are given in Reference 4.

Since the M2C10 nuclear design parameters have been obtained with previously approved methods and fall within expected ranges, the nuclear design is acceptable.

2.4 Thermal-Hydraulic Design The objective of this review is to confirm that the thermal-hydraulic design of the reload has been accomplished using acceptable methods, and provides an acceptable margin of safety from conditions which could lead to fuel damage during normal operation and transient conditions. The review includes two areas:

(1) the core thermal-hydraulic design methodology, (2) boiling (ydraulic compatibility, and (3) minimum departure from nuclear thermal-h DNB). The licensee has submitted a reload report in Reference a for the M2C10 operation.

Discussion of the review of Reference 4 concerning thermal-hydraulic design is as follows.

2.4.1 Mixed Core Thermal-Hydraulic Design tiethodology The analytical tools used by ANF for the thermal-hydraulic design are the XCOBRA-IIIC code (Ref.15 & 25) and the XNB critical heat flux (CHF) correlation (Ref.16). Both methods were previously approved by NRC with a restriction that an adjustment of 2 percent of the minimum DNBR must be included for mixed cores containing hydraulically different fuel assemblies.

In the ANF thermal-hydraulic design analysis, the two steps in the calculations are firstly, an octant-core calculation is done on an assembly-by-assembly basis.

In this analysis the limiting bundle is placed at its allcwabl.e maximum radial peak of 1.61 while the remaining assemblies are at 111.7 percent over power.

Inlet flow maldistributions are accounted for by a reduction of 5 percent in the limiting assembly.

Cross flow between adjacent assemblies in t

4_

the open lattice core is directly model. The assembly specific single-phase loss coefficients are used to hydraulically characterize the assemblies in the mixed core.

The results of this calculation are the axial flow distribution for hot assembly and cross flow boundary conditions which will be used in the detailed subchannel model.

Next, an octant of the hot assembly is modeled on roo-by-rod basis to determine DNBR for the core.

In this model cross flow between the limiting and adjacent fuel assembly is accounted for via cro:s flow between adjacent subchannels. As 1

part of their subchannel analysis, ANF increases the peak rod heat flux by typically 3 percent to account for extremes in fuel rod manufacturing tolerances and uses a flat peaking distribution within the rod array except for the limiting rod which is placed at its maximum peak.

i 1

2.4.2 Hydraulic Compatibility of ANF and Co-resident Fuel Hydraulic performance differences between ANF and Westinghouse (W) fuel were assessed with pressure drop test performed in ANF's hydraulic test facility.

Using the loss coefficients from these tests, ANF determined that the overall assembly loss coefficient fo'r the ANF assembly exceeds that of the Westinghouse fuel assembly by 21 percent at typical full power plant operation.

For a mixed core, a larger hydraulic resistance causes a net flow diversion from the ANF assembly. This flow diversion from the ANF assemblies to the Westinghouse assemblies results in lower DNBRs for the ANF fuel and increased DNBRs for the Westinghouse fuel.

The staff finds that in accordance with our approved procedure, with an adjustment of 2 percent of the minimum DNBR to compensate for uncertainty in the mixed core methodology, the effect of the hydraulic differences between the ANF assemblies and Westinghouse assemblies on the calculated minimum DNBP are appropriately cor..idered and are acceptable.

2.4.3 DNBR Safety Limit The safety analyses for the Millstone Unit 2 Cycle 10 reload were done to analytically demonstrate that DNB can be avoided for the limiting rod in the core with 95% probability at 95% confidence level throughout each analyzed transient.

The 95/95 DNBR safety limit is 1.17 for the XNB correlation after rod bow penalties and the 2 percent adjustment for uncertainties in mixed core methodology are applied.

This DNBR safety limit was previously approved by the NRC staff, and therefore it is adequate for use in the M2C10 reload application.

2.5 Transient Analysis of FSAR Chapter 15 Events The transient and accident analyses discussed in Sections 2.5 and 2.6 are based on assumed RCS flow of 340,000 gpm. The safety analysis based on the reduced RCS flow of 325,000 gpm is discussed in Section 2.7.

Plant transient analysis

i is included in References 5, 6, and 7 to support operation of M2C10 with a mixed core of 60 ANF fuel assemblies and 157 W fuel assemblies.

Several physics parameters are more limiting than Cycle 9 due to the extension of the cycle *1ength for ANF fuel to 18 months. These parameters include (1) increased shutdown margin 2.9 to 3.6% delta-k/k, (2) increased maximum radial peaking factor (1.537 to 1.61) at full power, and (3) increased both positive and negative bounds on the moderate temperature coefficient.

In addition, the analysis also includes an effect of an anticipated need for a reduced core inlet temperature (12'F). The licensee evaluated (Ref. 3) all events described in Chapter 15 of the Standard Review Plan (SRP) and reanalyzed the events which are either limiting events or events with initiator or controlling parameters changed from the analysis of record so that the events need to be reanalyzed transient analysis (g application.

for current licensin The staff evaluation of the results of Refs. 6 & 7) is discussed as follows.

2.5.1 Increase in Steam Flow Event This event is initiated by a failure of the main steam system that results in an increase in steam flow from the steam generator.

To evaluate this transient, the licensee analyzed two cases:

one at full power and one at the hot shutdown condition.

In the analysis, the end-of-cycle reactivity feedback coefficients were used to maximize the challenge to the specific fuel design limits for both cases. Since the analytical results show that DNB is not expected to occur and the fuel centerline melt limit of 21 Kw/ft is not violated, the staff concludes the results are acceptable.

2.5.2 Loss of External Load Event ANF analyzed this event using the approved PTS-PWR code (Ref. 17). Two cases were analyzed for this event:

one maximizing the overpressurization, and one minimizing the fuel design limits.

In both cases the input parameters were conservatively assumed to maximize the increase in reactor power during the j

transient. However, for the overpressurization event the parameters and safety system actuation set-points were assumed to maximize the system overpressure, while for the low DNBR event the parameters and safety system actuation set-points were assumed to minimize the pressurization in order to result in a l

minimum DNBR during transient.

Since the results, using the approved code, show the peak pressure of 2697 psia, less than a limit of 2750 psia, a minimum DNBR of 1.39 and peak LHR of 17.6 Kw/ft resulting in no violation of the specific fuel design limits, the staff concludes that the results are acceptable.

2.5.3 Closure of a Single Main Steam Isolation Valve (MSIV)

From Reference 5, the limiting case is the event initiated from full power conditions.

For simultaneous closure of both MSIVs, the event is similar to the event discussed in Section 2.5.2.

The turbine stop valve closure time used in the Section 2.5.2 analysis (0.1 second) is much smaller than the MSIV closure time (6 seconds). Thus the consequences of event discussed in Section 2.5.2 will bound those of the dual closure event.

I The asymmetric conditions resulting from the closure of only are of the two MSIVs is similar to a steam line break event since the primary coolant associated with the closed MSIV experiences a heatup due to loss of heat sink and the primary coolant loop associated with the open MSIV experiences a cooldown due to the load increase.

The approved steam line break (SLB) methodology (Refs.18 & 19) was used to perform this analysis.

The neutronic parameters required to predict radial power distribution between the cold and the hot side of the core were based on the event specific XTGPWR (Ref.14) calculations. These calculations differ from the SLB calculations due to the difference in the power range of interest.

The end-of-cycle reactivity feedback coefficients were used to maximize the worst cooldown effect. The results indicated that the acceptance criteria are met.

Since the minimum DNB limit is not exceeded by this event, the peak LHR is less than the 21 Kw/ft limit to centerline melt and the maximum pressure is less than 110% of design pressure, the staff concludes that the results are acceptable.

2.5.4 Loss of Feedwater Flow Event Two cases were analyzed for this event: one maximizing pressurizer liquid level and one minimizing steam generator liquid inventory.

The analysis was performed with the approved SLOTRAX-ML code (Ref. 20).

The results showed that this event does not. result in the violation of safety DNBR limit, peak pressurizer pressure does not exceed 110% of the design pressure and primary liquid is not discharged through the safety valves.

It was also shown that the auxiliary feedwater system supplies adequate cooling water to allow a safe plant shutdown and prevent steam generator dryout.

Based on the above analytical results, the staff concludes the analysis of the event is acceptable.

2.5.5 Loss of Forced Reactor Coolant Flow Event This event is initiated by a loss of the power supplied to, or a mechanical failure of an RCS pump.

As a result, the com flow rate will decrease and core temperature will increase.

Prior to reactor trip, the combination of decreased flow and increased temperature may violate the DNBR safety limit.

In the analysis, the assumptions were made to minimize pressure which minimizes DNBR.

The steam bypass and atmospheric dump valves were both assumed not to operate to minimize the calculated DNBR. ANF used the approved statistical set-point methodeiogy (Ref. 21) to evaluate the DNBR consequences of this event.

The results (Refs. 6 & 2?) showed that no DNBR safety limit is violated and that the maximum LHR of 17.2 Kw/ft is below the acceptable limit of 21 Kw/ft.

The staff finds the results acceptable.

2.5.6 Reactor Coolant Pumo Rotor Seizure Event This event assumed the locked pump loss coefficient given by the homologous pump curve at zero pump speed. A statistical application of uncertainties demonstrated that the minimum DNBR for the loss of flow event is greater than XNB DNB limit.

Due to the presence of margin to the DNB LC0 limits for a loss

9 l

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1 of RCS flow event and the inherent similarity between locked rotor and loss of RCS flow event, ANF concluded that fuel failures are precluded for the locked rotor event.

The calculated peak LHR of 17.4 Kw/ft is less than the 21 Kw/ft limit to centerline melt. Since results demonstrated that no fuel failures were expected for this event, the staff concludes that the results are acceptable.

2.5.7 Control Rod Withdrawal Event ThePTS-PWRcode(Ref.17)wasusedtoagalyzeanuncontrolledrodwithdrawal for a reactivity insertion rate of 4x10- AK/K/sec from full power initial conditions. The minimum DNBR is 1.21 which is above the 95/95 acceptance limit of 1.17 using XNB DNB correlation. This transient tripped the reactor on themal margin / low pressure signal. The maximum peak pellet LHR is calculated to be 19.1 Kw/ft. Since the code used for analysis was approved and the results showed sufficient margin existed to prevent fuel damage from a control rod withdrawal event, the staff concludes that the results are i

acceptable.

2.5.8 Control Rod Drop Event In this event, the core power initially decreases due to the insertion of negative reactivity resulting from the dropped control rod.

Moderator and Doppler temperature feedback cause power to return to its initial state. The event results in a localized increase in the radial peaking factor, which causes DNBR to decrease for the case with the initial core condition at full power.

The DNBR consequences for this event were evaluated using the approved statistical set-point methodology (Ref. 21). The results showed that the DNBR safety limit will not be violated. Since the power initially decreases following the CEA rod drop event, no reactor trip occurs and protection of thermal margin limits is provided by the LCOs. The staff finds that the calculations using the approved methods show that fuel damage will not occur from a control rod drop event, therefore we conclude that the results are acceptable.

l 2.5.9 Single Control Rod Withdrawal Event This event results in a reactivity insertio'n and a localized increase in the radial peaking factor. The degradation of core condition characteristics of reactivity insertion transient, combined with an increase in local radial peaking, poses a challenge to the DNBR limit.

Since the calculational results l

show that the amount of fuel failure for this event (Ref. 23) is bounded by that of a control rod ejection event and the peak LHR is less than the acceptable limits of 21 Kw/ft, the staff concludes that the analysis is acceptable.

8 2.5.10 Boron Dilution Event The boron dilution analysis was used to confirm that the required shutdown margins, which would enable the operator to have at least 15 minutes from the time of the first safety alarm until criticality for Modes 1 through 5, and 30 minutes for Mode 6 (the refueling mode), are as required in the SRP.

These new shutdown margins were then reflected in the Technical Specifications.

The calculated shutdown margin requirements to meet the required operator response time in the SRP are less than or equal to 3.6% delta-k/k for Modes 1 to 4, 2%

delta-k/k for Mode 5 and 5% delta-k/k for Mode 6.

Conservative assumptions were used in the analysis (i.e., the operator response time, to terminate the source of boron dilution flow, of 141 minutes instead of 15 minutes was used to determine the shutdown margin requirement for Mode 3 and the reactivity insertion rates at the lower bounded values were used to terminate power excursion for Modes 1 and 2).

Therefore, the staff concludes the calculated required shutdown margins are adequate and acceptable.

2.5.11 Control Rod Ejection Event The control rod ejection event was analyzed with the approved methods described in Reference 24.

Energy deposition in the hot fuel pellet was evaluated for beginning of cycle and end of cycle conditions from hot zero power (HZP) and hot full power (HFP) initial conditions.

In the analysis, no credit was taken for the flux flattening effects of reactivity feedback to maximize the total peaking factors. The results of analysis show that the HFP case results in a highest energy deposition of 240.6 cal /gm, which is below the acceptable limit of 280 cal /gm. An analysis of the core pressure surge associated with the control rod ejection indicates a maximum pressure of 2671 psia, below the acceptable limit of 2750 psia.

The DNBR calculation shows that less than 11.5% of the core will experience fuel failure and the radiological consequences are within 10% of the 10 CFR 100 limits (Ref. 23).

The staff finds that the approved methods were used to demonstrate that the primary system integrity will be maintained, the energy deposition is within the l

acceptable limit and the radiological release is within the 10 CFR 100 limits.

Therefore, the NRC staff concludes that the results are acceptable.

2.5.12 Inadvertent Opening of a PWR Pressurizer Pressure Relief Valve Event This event is primarily considered a depressurization event.

The licensee determined that the limiting case is the event with the core operated at full power conditions, which was analyzed using the approved PTS-PWR code (Ref.17) for system response, the XCOBRA-IIIC (Refs.15 & 25) for the hot channel thermal hydraulic analysis and XNB correlation for determination of DNBR. The results demonstrate that the minimum DNBR is greater than the XNB DNB limit and i

the fuel centerline melt limit of 21 Kw/ft is not exceeded.

Since the approved i

methods were used to show that the results are within the acceptable fuel design limits, the NRC staff concludes that the analysis is acceptable.

?

]

-g-l 2.5.13 Steam Line Break Analysis The licensee provided the results of the steam line break (SLB) aralysis in Reference 9 for NRC staff review.

The analysis was performed with the ANF SLB methods (Ref. 7), which were previously approved by the NRC staff (Ref. 19).

The SLB methods used ANF-RELAP developed from RELAP5/MI"12 with ANF modifications, (Ref. 7) for system response, XTG (Ref.14) for calculation of the core and hot assembly power distribution, XCOBRA-IIIC (Refs. 15 & 25) for determination of the core and hot assembly flow and enthalpy distributions, and the modtfied Barnett correlation (Ref. 26) for the DNBR calculations.

The licensee performed four double-ended guillotine SLB analyses in order to determine the limiting case for the consequences approaching the fuel design limits.

The four cases were:

(1) A large SLB during full power operation in combination with a single failure, loss of offsite power and stuck CEA.

(2) Case 1 with offsite power available.

(3) A large SLB during HZP operation in combination with a single failure, loss of offsite power and stuck CEA.

(4) Case 3 with offsite power available.

The analyses determined that the HZP case with loss of offsite power is the iimiting DNBR case, and the H7P with offsite power is the limiting case from standpoint of centerline melt. The worst calculated DNBR of 1.18 and peak LHR of 20.9 Kw/ft provide margin to fuel failure during SLB. The NRC staff concludes that the results are acceptable.

2.6 Loss-of-Coolant Accident Analysis 2.6.1 Small Break Loss-of-Coolant Accident (SBLOCA)

The licensee provided the results of the SBLOCA analysis in Reference 8.

The SBLOCA analysis was performed with the NRC approved method, EXEM PWR Small Break Model (Ref. 27).

The analysis was done assuming 102% of the core power of 2700 MWt, a maximum LHR of 15.1 Kw/ft and a radial peaking factor of 1.61.

The licensee also assumed an average steam generator tube plugging (of 23.5% and a maximum asymmetry of 5.9% in the analysis.

Various break sizes 1%, 1.0%, 3%

and 4% of double ended cold leg guillotine (DECLG) breaks) were performed and the results show that the limiting case is 1.9% of DECLG break with 12 F reduction in primary coolant temperature. The limiting case results in the highest peak cladding temperature of 1811 F, well below the acceptance criteria of 2200*F. The staff concludes that the SBLOCA analysis is acceptable since the approved method was used to show the analytical results to be within the acceptance criteria in 10 CFR 50.46.

9

.' 2.6.2 Large Break Loss-of-Coolant Accident (LBLOCA) Analysis The licensee provided the results of a large break LOCA analysis (Ref. 9). The LBLOCA analysis was performed with 102% of the core power of 2700 MWt, and an l

average steam generator tube plugging of 23.5% with a maximum asymmetry of 5.9%.

In addition, the licensee assumed a primary coolant average temperature reduction of 12 F.

Various break sizes (0.4, 0.6, 0.8 and 1.0 DECLG and double-ended-cold-leg split (DECLS) were performed and results show that the worst break case is 0.6 DECLG break, resulting in a peak cladding temperature (PCT) of 2163 F of 2176 F for cases without and with 12"F reduction in primary coolant temperature, respectively.

The analysis was performed with the NRC approved ANF EXEM/PWR evaluation model (Ref. 13). The evaluation model used l

RODEX2 for computation of initial fuel stored energy, fission gas release, and gap conductance; RELAP4-EM for the system and hot channel blowdown calculations; CONTEMPT /LT-22 for computation of containment back pressure; REFLEX for compu-tation of system reflood and T00DEE2 for the calculation of fuel rod heatup durira rhe refill and reflood portions of the LOCA transient.

The stm f has reviewed the analysis.

As a result, the staff found that the approved analytical methods and computer codes were used and the results show that the peak cladding temperature, metal-water reaction and clad oxidation are within the acceptance criteria in 10 CFR 50.46 for LOCA analysis. The NRC staff, therefore, concludes that the results of the LBLOCA analysis are acceptable.

2.7 Effects of RCS Flow Reduction on Safety Analyses In order to increase the margin in the RCS flow the licensee proposed TS changes based on an assumed RCS flow of 325,000 gpm.

In the supporting analysis (Ref. 30), the licensee identified and reanalyzed five events which are significantly affected by the reduced RCS flow. These events are:

(1) loss of flow, (2) locked rotor, (3) control rod, (4) small break LOCA, and (5) large break LOCA.

The loss of flow event is reanalyzed because it is the limiting DNBR transient and demonstration of margin to the DNBR safety limit for this event will demonstrate sufficient margin for the other events. The locked rotor event is reanalyzed because in the original analysis, fuel failures were precluded for this event based on the available margin to the DNBR safety limit calculated in the loss of flow event. The reduction in the RCS flow has removed that margin. Thus, analysis of the potential fuel failure is needed for the locked rotor event. The control rod ejection event is reanalyzed because it is the limiting event with respect to the amount of predicted fuel failures.

Reanalysis allows an assessment of the effect of the reduced RCS flow on the magnitude of the predicted fuel failures. The small and large break LOCAs are reanalyzed because an increase in the predicted PCTs are expected as a result of the decrease in RCS flow rate.

I I

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- 11 In the analysis, the following assumptions were made in order to compensate for the effect of the reduced RCS flow on the DNBR, PCTs and the predicted fuel failures:

T 1.

Reduce the radial peak factor (Fr ) from 1.61 to that specified in Table 15.0.5-1 of Reference 30.

2.

Raise the thermal margin / low pressure trip set-point from 1750 psia to 1850 psia.

3.

Alter the low power density LC0 to allow power operation only to an axial shape index (ASI) of -0.04 instead of -0.06.

I 4.

Reduce the LHR limit of 15.1 Kw/ft to 14.5 Kw/ft for flow rates greater than or equal to 325,000 but less than 340,000 gpm.

3.0 TECHNICAL SPECIFICATION CHANGES Various changes to the Technical Specifications have been prrosed in order to operate the M2C10 core. The changes include changes ir, (1) iinear heat rate, l

(2) total integrated radial peaking factor, (3) total planar radial peaking factor, (4) moderator temperature coefficient, (5) shutdown margin and (6) low steam generator trip set-point. These changes include only the TS changes relating to the application dated November 15, 1989.

Changes related to the reduced RCP flow rate as requested.by letter dated February 1, 1989 will be in a subsequent amen'Nent. Our assessment of the proposed Technical Specifications (T changes is summarized as follows.

1.

Pages II, Y and XI of the index - Deletes references to the unrodded planar radial peaking factor (Fxy) because Fxy is removed from the proposed Technical Specifications.

2.

Page 1-5 of the definitions - Deletes the definition for Fxy because Fxy is l

being removed from the Technical Specifications.

3.

Section 2.1.1 (Figure 2-2 on page 2-2) - A footnote is added to specify the ruinimum reactor vessel flow of 325,000 gpm for the case with the reduced Fr as specified in Technical Specification 3.?.3.

This change is er,nsistent with the assumptions used in the acceptable analytical results (Ref. 30) and is acceptable.

4.

Section 2.2.1 - Increases the steam generator low pressure trip set-point and allowable value (Table 2.2-1, page 2-4) from 500 to 680 psia and 492 to 672 psia, respectively. The changes are to assure adequate protection against the asymmetric secondary side transient resulting from the enclosure of a single MSIV.

Footnote 2 of Table 2.2-1 (page 2-5) is changed to allow the steam generator low pressure trip to be manually bypassed from 600 to 780 psia. This change is to assure that the reactor is in safe condition whenever the trip is bypassed. Supporting analysis

1 (Ref. 6) was performed to demonstrate the acceptability of the changes.

The changes are acceptable.

[A footnote to specify the design flow rate which is a base for the low RCS flow trip set-point. When the measured flow rates are greater than 340,000 gpm, 340,000 gpm is used as the design fl ow. When the measured flow rate is between 325,000 and 340,000 gpm, the measured flow rate is used as the design flow.] The thermal margin / low pressure trip set-point is raised from 1750 to 1850 psia. The changes are supported by the acceptable results (Ref. 30) and are acceptable.

5.

Five changes to the bases of the Safety Limits and Limited Safety System Settings are as follows:

(i)

Page B 2 Changes the references for the DNB correlation to be consistent with the XHB co" relation used by ANF instead of the W-3 correlation used by Westinghouse.

(ii) Pages B 2-1, 3, 5, 6 and 8 - changes the DNBR limit determined by using statistical methods from 1.30 to 1.17 to be consistent with the DNBR safety limit of the XNB correlation used by ANF.

(iii) B 2 Deletes the steam generator operating pressure of 815 psia.

A qualitative statement without specific number provides clear definition of the technical base and is acceptable.

(iv) Page B 2 Changes the uncertainty factor for the thermal margin / low pressure trip from 67 to 72 psi which consists of a 22 psi pressure measurement error and a 50 psi time delay allowance.

The ci age is consistent with the assumption used in the transient analysis.

(v)

Pa.ge B 2 Changes the steam generator low pressure set-point from 500 psia to 680 psia to be consistent with the assumptions used in the supporting analysis for the reload application.

6.

Section 3/4.1.1.1 (page 3/41-1) - Changes the shutdown margin for Modes I through 4 from 2.90% to 3.60% delta-k/k to reflect the Cycle 10 fuel characteristics.

The change is supported by the transient analytical results and is acceptable.

7.

Section 3.1.1.4 (page 3/41-5) - The most positive moderator temperature coefficient (MTC) for the power less than or equal to 70% of the full 0.7x10-gd the most negative MTC at full gower are chagged from 0.5x10-4 power a to delta-k/k/*F, and from -2.4x10-to -2.8x10- delta-k/k/ F, respectively. The changes are to reflect the Cycle 10 fuel characteristics and are acceptable.

8.

Section 3/4.2.2 (pages 3/4 2-5 through 3/4 2-8) - Deletes the entire section regarding total planar radial peaking factor (Fxy) from the Technical Specifications.

This deletion is acceptable because ANF's 3-0

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- power distribution methodology (Ref. 31) does not require this parameter (Fxy). The axial shape index (ASI) tents in this Section are required for monitoring of the LHR. The LHR tent is therefore moved to the LHR specification (Section 3/4.2.1).

9.

Section 3/4.2.1 - The maximum linear heat rate (MLHR) is changed from 15.6 Kw/ft to 15.1 Kw/ft for the RCS flow rates greater than 340,000 gpm. The MLHR is 14.5 Kw/ft for the RCS flow rates greater than er equal to 325,000 gpm and less than 340,000 gpm.

Also, the end of cycle coastdown restrictions are removed. The changes are supported by ANF in their LOCA and set-point analyses and are acceptable (Refs. 6 & 30).

Figure 3.2-1 on page 3/4 2-3 with the current LHR value is deleted and the constant value is written into the text on pages 3/4 2-1 and 3/4 2-2.

The definition of LHR on Figure 3.2-1 is also written into the text.

Inclusion of the LHR limit in the text and deletion of the figure is proposed for simplifica-tion and does not affect the interpretation of the Technical Specification and is acceptable.

The LHR and the ASI requirements for monitoring of the LHR on Figure 3.2-2a are retained.

Figure 3.2-2a is renumbered as 3.2-2 (Ref. 29).

Figure 3.2-2b is deleted from the Technical Specification. The changes are consistent with the assumptions used for LOCA (Refs. 9 & 30) and set-point analyses and are acceptable.

The specific changes to the Technicc' Specifications for LHR monitoring using the excore detectors are discussed as follows:

(i)

Item a of Section 3.2.1 is deleted.

The deletion of item a, specifying the maximum allowable power less than 100% of rated thermal power at certain Fxy values, is consistent with the proposed Technical Specifications with deletion of Fxy.

(ii) Item b of Section 3.?.1 is combined with the introductory sentence.

References to the maximum allowable power limit are deleted as discussed in item (1) above. The two separate conditions following the "either" are separated as Items a and b.

References to Technical Specification 3.2.2 are changed to refer to Figure 3.2-2 as discussed above.

(iii) Section 4.2.1.2.b (page 3/4 2-2) - Reference to Technical Specification 3.2 2 is changed to Figure 3.2-2 as discussed above.

(iv) Section 4.2.1.2 (page 3/4 2-2) - Deletes Item C which requires verification every 31 days of operation within the limits of Figure 3.2-2.

Since the limits of figure of 3.2-2 are monitored continuously by the power ratio recorder per Technical Specification 3.2.1, we find that Item C is redundant to a more restrictive requirement and deletion of Item C is acceptable.

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? 10. Section 4.2.1.3 (page 3/4 2-2) - Deletes two penalty factors.

Tht/ are (1) Item 4.2.1.3.b.1, the flux peaking augmentation factor as shown in Figure 4.2-1 and (2) Item 4.2.1.3.b.4, the LHR uncertainty factor of 1.01 due to axial fuel densification and thermal expansion.

Based on the following reasons, we find the deletions acceptable.

(i) The flux peaking augmentation factors are a result of the fuel pellets interacting with the cladding prior to the time of maximum duel densification.

Since the licensee stated that the pellet and the clad will not occur prior to the time of maximum pellet densification in the ANF fuel and the NRC has previously approved the removal of this penalty for tae ANF fuel used in the Combustion Engineering reactor, we coNiude that the deletion of the flux peaking augmentatk n factors is acceptable.

(ii) The licensee stated that the LHR uncertainty factor due to axial densification and thermal expansion is included in the engineering uncertainty factor of 1.03 (Item 4.2.1.3.6.3 on page 3/4 2-2) in the ANF methodology (Ref. 4). We, therefore, agree that this penalty does not need to be included again in the Technical Specification for the ANF fuel.

The removal of these penalties only applies to the ANF fuel.

They shouId still be applied to the Westinghouse and Combustion Engineering fuel when used in the core.

Therefore, footnote is applied to these two factors (Itein 4.2.1.3.b.1 and 4.2.1.3.b.4 on page 3/4 2-2) specifying that they only apply to the non-ANF fuel.

This footnote is also added to Figure 4.?-1 (page 3/4 2-4).

11.

Section 3.2.3 (page 3/4 2-9).

The Total Integrated Radial Peaking Factor (Fr) of 1.537 is changed to 1.61.

Figare 3.2-2.b is replaced by Figure 3.2.2.

These changes are acceptable.

12. Section 3.2.4 (page 3/4 2-10) - Deletes the references to Fxy and the current Technical Specification Section 3.2.2 to be consistent with Item 8 of the proposed Technical Specifications discussed above.

In this page, two references to " Total Integrated Radial Peaking Factor" are no longer all capitalized to be consistent with the term used in the Technical Specifications.

13.

Section 3.3.1 - Changed footnote B on page 3/4 3-4 to be consistent with i

the changes regarding the set-point to bypass the steam generator low l

pressure trip as discussed in Item 3 of the proposed Technical Specifications.

14. Section 3.2.6 (page 3/4 2-14) - A footnote to the RCS flow of 340,000 gpm is added to indicate that the flow can be reduced to 325,000 gpm if there are reductions in the Fr as given in Technical Specification 3.2.3.

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. '15.

Section 3/4.3.3.2 (pages 3/4 3-30 and 3/4 3-31) - Deletes three references to the unrodded planar radial peaking factor (Fxy) to be consister.t with the deletion of Fxy.

16. Section 3/4.10.2 (page 3/410-2) - Deletes test exceptions allowed from the current Technical Specification Section 3.2.2 (Fxy) to be consistent with the deletion of Fxy.

Delete test exceptions allowed from Technical Specification 3.1.3.2 which dealt with the post length control rods because both the post length rods and Technical Specification were removed for Cycle 2.

17.

Section 3/4.1 (pages B 3/4 1-1 and B 3/4 1-2) - The base section for shutdown margin is changed to 3.6% delta-k/k from its current value to be consistent with the proposed Technical Specification changes discussed in Item 6 above.

Page B 3/41-4 refers to radial praking factors.

Since one of these two factors is deleted, the word " factor" is no longer pluralized.

18. Section 3/4.2 (page B 3/4 2-1) - Two changes are made in this section as follows:

(1) When the LHR is monitored by the excore detectors, the allowable radial power distribution is currently given by the total planar radial peaking factor of Technical Specification 3.2.2.

To be consistent with the proposed Technical Specifications, this is changed to the Total Integrated Radial Peaking Factor of Section 3.2.3.

(ii) The end of cycle coastdown restrictions are removed.

Therefore, the basis for this restriction is deleted.

References to Fxy and Technical Specification 3/4.2.2 are deleted to be consistent with the deletion of Fxy and Technical Specification 3/4.2.2 discussed in Item 8.

19.

Section 3/4.4.1 (page B 3/4 4-1) - Change the DNBR limit of 1.30 to 1.17 to be consistent with the proposed Technical Specification changes as discussed in Item 5(i).

20. Section 5.2-2 (Page B 3/4 6-2 and 5-4) - Correct the maximum design temperature in the containment building from 288 F to 289'F to be consistent with the value specified in the FSAR Section 6.4.1.1.

We have reviewed the Technical Specification changes and found that all of the changes reflect the characteristics of fuel in M2C10 and are supported by the assumptions used in the acceptable analytical results (Fefs. 5 through 9 & 30) we therefore, conclude the Technical Specification changes acceptable.

4.0 SAFETY ANALYSES CONCLUSIONS We have reviewed the reports submitted for the Cycle 10 reload of Millstone Unit 2 with the mixed core of the Westinghouse and ANF fuel assemblies, and the licensee's analytical results for transients and LOCAs.

Based on this review,

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we conclude that appropriate material was submitted and that the fuel design, nuclear design, thermal hydraulic design and transient and accident analyses are acceptable. The Technical Specification changes submitted for this reload are supported by the acceptable analytical results.

The operating limits associated with those changes and reload parameters are acceptable.

Table I summarizes a comparison of the consequences for the events analyzed to support the licensee's amendment request based on the RCS flow of 340,000 gpm (Ref.1) versus consequence calculated for the reduced flow analysis. The NRC staff has reviewed the safety analysis (Ref. 30) with the reduced RCS flow of 325,000 gpm and found that the assumptions are consistent with the proposed TS changes and the approved methods were used to demonstrate that the applicable fuel performance acceptance criteria are met in all cases. Therefore, the staff concludes that the results in Reference 30 support the operation of the N2C10 core at a rated thermal power of 2700 MWt with average steam generator tube plugging up to 23.5% of the steam generator tubes.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendment involve' no significant s

increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 651.22(c)(9).

Pursuant to 10 CFR 951.22(b), no environmental impact statement of environmental assessment need be prepared in coraeetion with the issuance of the amendment.

6.0 CONCLUSION

We have 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 amendment will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

Letter from R. J. Mroczka (Northeast Nuclear Energy Company - NNECO) to NRC, dated August 26, 1988.

2.

Letter with an attachment from R. J. Mrcozka (NNECO) to NRC, Millstone Nuclear Power Station, Unit No. 2 - Proposed License Amendment, Cycle 10 Reload, dated November 15, 1988.

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1 l

l 3.

ANF-88-126, Millstone Unit 2 Cycle 10 Safety Analysis Report, October l

1988.

4.

ANF-88-088 (Revision 1), Design Report for Millstone Unit 2 Reload ANF-1, August 1988.

5.

ANF-87-161, Millstone Unit 2 Plant Transient Analysis Report - Analysis of Chapter 15 Events, September 1988.

6.

ANF-87-161, Supplement 1, Millstone Unit 2 Plant Transient Analysis Report - Analysis of Chapter 15 Events, October 1988.

7.

ANF-88-127, Millstone Unit 2 Steam Line Break Analysis, October 1988.

8.

ANF-88-129, Millstone Unit 2 Small Break LOCA Analysis, October 1988.

9.

ANF-88-118, Millstone Unit 2 Large Break LOCA/ECCS Analysis, August 1988.

10.

XN-NF-82-09(P)(A), Generic Mechanical Design Report Exxon Nuclear 14x14 l

Fuel Assemblies for Combustion Engineering Reactors, dated November 18, 1983.

'11.

XN-NF-81-58(A)', Rev. 2, RODEX2 - Fuel Rod Thermal-Mechanical Response Evaluation Model, dated March 1984.

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12.

XN-NF-8'2-06(A), Rev.1, Supplements 2, 4, and 5 Qualification of Exxon Nuclear Fuel for Extended Burnup (PWR), dated October 1986.

13.

Letter from D. Crutchfield (NRC) to G. Ward (ENC), Safety Evaluation of Exxon Nuclear Company's Large Break ECCS Evaluation Model EXEM/PWR and Acceptance for Referencing of Related Licensing Topical Report, dated July 8, 1986.

14 XN-CC-28(A), Rev. 3, XTG: A Two Group Three Dimensional Reactor Simulator Utilizing Loose Mesh Spacing (PWR Version), January 1975.

15.

XN-NF-82-21(P)(A), Rev.1 Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations, September 1983.

16.

XN-NF-621(P)(A), Rev. 1, Exxon Nuclear DNB Correlation for PWR Fuel Designs. September 1983.

17.

XN-NF-74-5(A), Rev. 2 and Supplements 3-6, Description of Exxon Nuclear Plant Transient Simulation Model for Pressurized Water Reactors (PTS-PWR), October 1986.

18. ANF-84-93 (Su.oplement 1), Steam Line Break Methodology for PWRs, June 1988.

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0 19. Lettar from A. Thadani (NRC) to R. Copeland 3NF), Acceptance for Referencing)of Licensing Topical Reports, ANF-84-93(NP), and ANF-84-93(P, Supplement 1, Steam Line Break Methodology for PWRs, dated December 28, 1988.

20. XN-NF-85-24(A),SLOTRAX-ML: A Computer Code for Analysis of Slow Transients in PWRs, September 1986.

21.

XN-NF-507(A), Supplements 1 and 2, ENC Setpoint Methodology for CE Reactors:

Statistical Setpoint Methodology, September 1986.

22. Letter from E. Mroczka (NNECO) to NRC, Response te NRC Questions, dated February 3, 1989.

23. Letter from E. Mroczka (NNECO) to NRC, Response to NRC Questions, dated January 13, 1989.

24. XN-NF-78-44(A), A Generic Analysis of the Coolant Rod Ejection Transient for Pressurized Water Reactors, October 1983.

25. XN-NF-75-21(A), Rev. 2, XCOBRA-IIIC:

A Computer Code to Detennine the Distribution of Coolant During Steady-State and Transient Core Operation, January 1986.

26.

IN-1412(TID), A Correlation of Rod Bundle Cri.tical Heat Flux for Water in the Pressure Range 150 to 725 psia, July 1970.

27.

Letter from A. Thadani (NRC) to G. Ward (ANF), Acceptance for Referencing Topical Report SN-NF-82-49(P), Revision 1, Exxon Nuclear Company Evaluation Model - EXEM/PWR Small Break Model, dated July 12, 1988.

28. Letters from E. Mroczka (NNECO) to NRC, Response to Additional Questions, dated January 13 and January 23, 1989 (proprietary and non-proprietary versions).

29.

Letter with attachment from E. Mroczka (NNECO) to NRC, Proposed Revision to Techincal Specifications - Reduced Reactor Ccolant System Flow Rate, dated February 1, 1989.

30. ANF-89-011, Millstone Reduced Flow, Standard Review Plan, Chapter 15 Event Analysis, January 1989.

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31. ANF-507 (Addendum 1), Advanced Nuclear Fuels Corporation Setpoint Methodology for C.E. Reactors:

Three-Dimensional Axial Power Distribution Generation, June 1988.

Principal Contributor:

S. Sun Dated: March 20, 1989 L--

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