ML20155B463

Safety Evaluation Supporting Amend 107 to License DPR-36

ML20155B463
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Site:	Maine Yankee
Issue date:	09/27/1988
From:	Office of Nuclear Reactor Regulation
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Download: ML20155B463 (12)

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. [ou ecg#'o, UNITED STATES

[ ' y "v. *h NUCLEAR REGULATORY COMMISSION

. WASHINGTON, D. C. 20566 cl

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 67 TO FACE lTYl# D ATT W [ICENSE NO. DPR-36 MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATMRDllf37ATIM

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00CKET NO. 50-309

1.0 INTRODUCTION

By letter dated June 30, 1988 (Ref. 1), the Maine Yankee Atomic Power Company (MYAPC) submitted an application to modi'y the Technical Specifications for Maine Yankee to permit operation for an eleventh cycle. A Cycle 11 core reload report (Ref. 2) was also submitted with the above letter. The fuels, physics, and thermal-hydraulic evaluations of this reload report are presented herein. In addition, those transients and accidents for which a new or revised analysis has been performed are evaluated in the Safety Analyses Section. An evaluation of the proposed Technical Specification changes is also presented.

J 2.0 EVALUATION OF FUEL O_E,S GN 2.1 Fuel Loading Pattern The Cycle 11 reload application involves fuel designs similar.to those previously considered for the Maine Yankee reactor. The Maine Yankee Cycle 11 core will consist of 217 fuel assemblies with fuel rods arranged in la by 14 arrays. Of the five fuel types proposed for use in the C fuel types were fabricated by Combustion Engineeringand (CE)ycle two were11 core, three fabricated by Advanced Nuclear Fuels Corp. (ANF). Two of the CE fuel types, Type N and P, consist of 136 previously irradiated fuel assemblies. The Type Q CE fuel consists of 72 unirradiated assemblies with an increased enrichment of 3.7 weight percent U-235. This higher enrichment was previously reviewed and approved by the staff (Ref 3). The ANF fuel, denoted Types L and l M, consists of 9 fuel assemblies in the Cycle 11 core which were previously l irradiated during Cycles 7, 8, 9, and 10.

As in Cycle 10, the Cycle 11 core will contain 81 control element assemblies (CEAs) of which four are nonscramable. In addition to the CEAs, the Maire Yankee Cycle 11 core will also contain burnable poison rods in selected assemblies. There will be 872 standard 8 4C-Al 3 burnable poison rods in Cycle 11 compared to 1072 in the previous cyc1 l

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2.2 Fuel Pechanical Design The mechanical design features of both CE and ANF fuel assemblies to be used in the Cycle 11 core are listed in Table 3.3 of Reference 2. The fresh reload fuel. Batch Q. being inserted in Cycle 11 is similar to the previously supplied reload fuel except for the increased U-235 enrichment. Since Batches L and M. the ANF supplied fuel, will achieve exposures hioher than previously encounteredatMaineYankee,anextendedburnupanalyses(Ref.4)was performed in order to demonstrate compliance with the appropriate design criteria of these higher exposures. The results of the extended burnup analyses for the ANF fuel batches are reported in Reference 2. These include evaluations of fuel cladding collapse, irradiation induced dimensional  ;

changes, cladding strain and fatigue analysis, maximum fuel rod internal pressure and fuel rod corrosion. The results indicate the primary stress in the cladding will not exceed the design stress limit, the collapse resistance '

of the fuel rods is sufficient to preclude collapse during the projected lifetime of the fuel, and the predicted fuel rod internal gas pressure remains below reactor coolant system (RCS) pressure throughout the projected lifetime <

of the fuel. These results were calculated with methods which hve been approved by the staff (Ref. 5) and are in conformance with the requirements of .

Standard Review Plan (SRP) 4.2. The CE fuel mechanical design was approved for earlier cycles and is applicable to Batch Q fuel. The staff, therefore, concludes that the Paine Yankee Cycle 11 fuel mechanical design is acceptable.

2.3 Fuel Thermal Design The licensee's analysis of the fuel thermal performance is the same as that used in previous reload analyses including the use of Dower history effects and burnup-dependent fission gas release. The Batch 1. fuel was not analyzed explicitly because its exposure and power histories were bounded by Batch M.

The fuel theraal design analyses have been performed using methodology previcusly approved by the staff and t'.e results are acceptable. As a result, the fuel thermal design analyses for Cycle 11 is acceptable. This finding includes both '

power-to-centerline melt and core average gap conductance calculations, i 3.0 EALUATIONOFpHYSICSDESIGN 3.1 Core Characteristics Maine Yankee Cycle 11 incorporates a low-leakage design, achieved by placement of fresh (unirradiated) fuel assemblies in selected enre interior locetions and burned (irradiated) fuel assemblies on the core periphery. In addition to reducing the irradiation exposure to the reactor pressure vessel, this low-leakage core design also produces a less severe moderator defect with cooldown at end-of-cycle (E00), improves the stability of the core to axial xenon oscillations near E00, and extends the achievable full power lifetire of the cycle. Cycle 11 is expected to attain a cycle average full pcwor lifetime of 12.750 MWD /MUT.

l 3.2 Power Distributions Hot full power (HFP) fuel assembly relative power densities are given in Reference 2 for beginning-of rycle (200) (50 WD/MTU) middle-of cycle (MOC)

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(6000 MWD /MTU), and EOC (14.000 MWD /MTU) co'nditions fo. both unrodded and rodded (CEA Bank 5 in) configurations. These results show that the unrodded maximum 1-pin radial peak power occurs at C0C when its value is 1.47. The proposed Technical Specification change, shown in Figure 3.10-4 and giving the allowable unrotided radial peaking, including 10 percent calculational uncertainty, as a function of average exposure for the Cycle 11 core, indicates radial peaks in the range from 1.770 to 1.731. Comparison of the radial peaks given in the above power distributions with the allowable values shown in the Technical Specifications demonstrates the adequacy of the results given in the core perfonnance analysis (Ref. 2). The staff, therefore. *inds this analysis to be acceptable.

3.3 Reactivity Coefficients and Kinetics parameters The moderator temperature coefficient (MTC), the fuel temperature (Doppler) coefficient, the soluble boron and burnable poison shim reactivity effects, and other kinetics parameters for the Cycle 11 core are compared with the corresponding values of Cycle 3 (reference cycle) and Cycle 10 (previous cycle) in the Cycle 11 Core Performance Analysis Report (Ref. 2).

The MTCs at nominal operating HFP and HZP. BOC conditions are slightly more positive than the encresponding previous cycle (Cycle 10) 'talues primarily because of the higher 300 critical boron concentration resulting from more excess reactivity in the core. The EOC values are more negative than in the orevious cycle due primarily to the increased core average exposure and the higher average enrichnent.

Technical Specification MTC limits are provided for Cycle 11 based on the LOCA analysis moderatur density defect curve. This curve infers specific MTC values in the operating range which must not be exceeded for the LOCA analysis to remain valid. The Cycle 11 Doppler coefficients are quite similar to the Cycle 3 and Cycle 10 values. The critical boron concentrations for Cycle 11 at BOC are higher than those of Cycle 10 because of the larger amount of excess reactivity in the core. Values of the delayed neutron #raction and prompt neutron generation time for Cycles 3.10. and 11 are comparable and the differences reflect the effects of core average exposure and power weighting.

Since the above data have been obtained using approved methods, are used in the safety analyses with appropriate calculational uncertainties applied in a conservative manner, and are included in the Technical Specifications, the staff finds the data to be acceptable.

3.4 Control Requirements The value of the required shutdown margin is dotennined either from the steam line break analysis (EOC) or from other safety analyses (BOC). Based on these values of required shutdown margin and on calculated available scram reactivity including a maximum worth stuck rod and appropriate calculational uncertainties, sufficient excess exists between available and required scram reactivity for all Cycle 11 operating conditions. These results are derived by approved methods and incorporate appropriate assumptions and are, therefore, acceptable. The Power Oependent insertion Limits (PDILs) for CEAs are given in the Technical Specifications and are required to provide for sufficient available scram reactivity at all power levels during the cycle.

An allowance for the 90ll CEA worth is made when determining the available scram CEA worth.

. o 4.0 EVALUATION OF THERMAL-HvDRAULIC DESIGN 4.1 Thermal-Hydraulic Analyrip, The steady-state and transient departure from nucleate boiling (DNB) analyses were performed using the CORRA-III C computer program. COBRA-III C was developed by Battelle Northwest Laboratory for use in the thermal-hydraulic analysis of nuclear fuel elements in rod bundles. The application of <

COBRA-III C to the Maine Yankee thermal-hydraulic design is described in  ;

References 6 and 7. The computer program was also used on a one-eighth core 1 assembly-by-assembly model to determine hot assembly enthalpy rise flow factors. This model accounted for the difference in hydraulic characteristics  !

between the CE and the ANF fuel assemblies. The inlet flow maldistribution .

1 imposed on the model was based on the results of flow measurements taken in  ;

scale model flow tests of the Maine Yankee reactor vessel as described in '

References 8 and 9. The resulting hot assembly flow factors for the CE l assemblies was 0.972 for those assemblies with bottom peaked power i distributions and 0.990 for those with top peaked distributions. A 0.95  ;

enthalpy rise flow factor was applied to the ANF fuel assemblies because of the higher spacer loss coefficients relative to the CE fuel. These factors l are applied to the inlet mass velocity in the hot channel model in predicting ONB performance.

4.2 Fuel Rod Bwinj I A parameter which is considered in the thennal-hydraulic design is rod-to-rod bowing within fuel assetnblies. The licensee has evaluated the maximum channel gap closure due to fuel rod bowing for both the CE and the ANF fuel assemblies with the highest burnup during Cycle 11. The maximum closure for CE fuel was calculated to be 24.8%. For the ANF fuel assemblies, the maximum gap closure due to fuel rod bowing was predicted to be less than 33%. Allowan:cs for rod  !

pitch and clad diameter variations due to manufacturing tolerances result in '

! an additional maximum channel closure of approximately 10% for the most adverse conditions. In accordance with the approved methodology Maine Yankee used, no penalty is to be applied to fuel if the predicted gap closure is less than 50%. lherefore, no rod bow penalty is required for any of the fuel in i 1 Cycle 11.

5.0 SAFETY ANALYSES Maine Yankee has reviewed the parameters which influence the results of the transient and accident analyses for Cycle 11 to determine which events, if any, require a reanalysis. The parameters of importance are initial operating j

, conditions, core power distributions, reactivity coefficients, shutdown CEA '

characteristics, and reactor protection system (RPS) trip setpoints and time j delays. In addition, the effect of 250 plugged tubes in each steam generator  ;

was evaluated for all events. For those events where the parameters for l 4

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P 5-Cycle 11 are outside the bounds considered in previous safety analyses, a new-or revised analysis was performed. These are:

(1) CEA Withdrawal (2) Boron Dilution (3) Excess Load (4) Loss of Feedwater (5) Loss of Coolant Flow s (6) Full Length CEA Drop (7) Steam Line Rupture (S) Seized Rotor (9) Loss o' Coolant Accident (LOCA) 5.1 CEA Withdrawal Event The CEA withdrawal is an anticipated operational occurrence (A00) for which the RPS is relied upon to assure no violation of the specified acceptable fuel design limits (SAFDLs). The most severe CEA withdrawal transient occurs for a J

ccabination of reactivity addition rate and time in core life that results in the slowest reactor power rise to the level just below the Variable Overpower ,

, Trip. The reference safety analysis parametric gtudy covered the range of MTCs from +0.5 x 10'4 delta k/k/*F4 to -3.0 x 10' delta k/k/*F and reactivity

] addition rates from 0 to 0.7 x 10 delta k/k/sec.

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The minimum DNBR for a CEA withdrawal event for Cycle 11 occurs for a bank withdrawal from an initial power level of 100% of rated power. Protection against violation of the SAFDLs is assured by the Variable Overpower Trip. ,

The minimum DNBR for this event is 1.37 as calculated with the YAEC-1 DNB correlation and the peak pressure is less than the American Society of Mechanical Engineers (ASME) design overpressure limit of 2750 psia.

This analysis, using approved methods and assumptions, assures that the SAFDLs

are not violated and is, therefore, acceptable.

5.? _ncontrolled Boron Dilution U

An inadvertent boron dilution will reduce the boron concentration in the i primary coolant which in turn will increase the reactor core positive

, reactivity. During power operation, the resulting reactivity insertion will increase the reactor power and cutomatic safety systems will act to shutdown 3

the reactor and maintain the plant within safety limits. However, a boron '

dilution event during shutdown will not be mitigated by any automatic safety

systems, if it is allowed to continue unmitigated it would result in reactor j recriticality unless the operator takes appropriate corrective action to stop j the dilution within the necessary time period.

, The licensee indicated that the boron dilution event was analyzed for the following operating modes:

1) refuelino

2) cold shutdown - filled RCS

3) cold shutdown - drained RCS (4) hot shutdown - filled RCS 1

(5) hot shutdown - drained RCS (6) startup ,

(7) hot standby (8) pcwer operation (9) failure to borate prior to cooldown.

The assumptions made in the Cycle 11 evaluation are consistent with those made in Reference 10 and 11. These events were evaluated using a mathematical model that has been previously reviewed and found to be suitably conservative.

For the refueling mode of operation, the limiting dilution was based on the maximum flow of the primary water makeup of 250 gpm. Based on the Cycle 11 core loading, the critical boron concentration under cold conditions (68'F) during refueling is 1346 ppm with the two most reactive CEAs withdrawn or 905 ppm with all CEAs inserted. The minimum initial reactor vessel boron concentration which will prevent an inadvertent criticality 'elthin 30 minutes is 1979 ppm with the two most reactive CEAs withdrawn or 1331 ppm for all CEAs inserted. There is, therefore, ample time for the operator to acknowledge the audible count rate signal and take corrective actio1.

Dilution during shutdown conditions with the RCS partially drained was addressed in References 11 and 12. The licensee hat shown the boron i concentrations reouired to meet the 5% delta k/k Technical Specification subcriticality requirement for shutdown conditions as well as the required initial RCS boron concentrations to allow 30 minutes margin to criticality  :

during drained RCS conditions. The 30 minute margin is assumed for drained conditions where the head is removed since these are classified as refueling condi tior.t in the Technical Specifications. The licensee has stated that administrative procedures ensure that the higher of these two values are used i and, therefore, a minimum margin to criticality of 30 minutes would be '

available for the operator to take appropriate action in the event of a l limiting boron dilution from drained conditions.  !

l Dilution during shutdown conditions with the RCS filled was addressed in '

Reference 12. The licensee has shown the boron concentrations required to meet the 5% delta k/k Technical Specification subcriticality requirement for shutdown conditions as well as the required initial RCS boron concentrations to allow 15 minutes margin to crit'cality during filled RCS conditions. The concentrations required by the Tectnical Specifications conservatively bound those required to meet the 15 minuti criterion for margin to criticality during boron dilution events from tiese conditions.

To evaluate the boron dilution event during hot standby, startup, and power operation for Cycle 11, the licensee 'ndicated that the same assumptions were used as in the analysis in Reference 11 except for the inverse bcron worth and highercriticalboronconcentration(l'/58 ppm)athotstandby. Based on the maximum reactivity insertion rate, it would take approximately 53 minutes of continuous dilution at the maximum charging rate to absorb the minimum Technical Specification shutdown margin of 3.2% delta k/k.

Failure to add boron during cooldown was evaluated based on conservative values of MTC, initial teeperature, and maximum cooldown rate. In order to I achieve criticality from these initial conditions, the temperat'jre reduction l requires approximately 61 minutes.

Based on the acceptability of the operator response times and comparison with Cycle 10 analysis, the staff concludes that the results for Cycle 11 are acceptable.

5.3 Excess load Event The excess load event occurs whenever there is rapid increase in the heat removal from the reactor coolant without a corresponding increase of reactor power. This power-energy removal mismatch results in a decrease of the reactor coolant average temperature and pressure. When the moderator temperature coefficient of reactivity is negative, unintentional increases in reactor power may occur. Therefore, the excess load event as reported in Reference 10 was analyzed over a wide range of power levels and negative MTCs todeterminetheminimummargintothelinearheatgegrationrate(LHGR)and DNBR SAFDLs. The most negative MTC value of -3.17x10 delta k/k/*F used is more negative than the value predicted for Cycle 11, including uncertainty.

The minimum DNBR for this transient is 1.33 and corresponds to an event initiated from the positive edge of the symetric offset band at full power and results in a power increase to the variable overpower trio setpoint. The closest approach to fuel centerline melt corresponds to an event initiated from the negative edge of the symetric offset band near full power and results in a power increase to the variable overpower trip setpoint.

The results of the analysis meet the SRP 15.1.1 criteria and, therefore, are acceptable.

5.4 Loss of Feedwater Event A loss of feedwater event could be caused by main feed pump failure or feed control valve malfunction. Loss of feedwater flow would result in a decrease in steam generator water level, increase in primary pressure and temperature and reduction in the secondary system capability to remove the heat generated in the reactor core. The event is a heatup transient. The minimum ONBR calculated for this event for Cycle 11 is 1.54 and peak RCS pressure is bounded by the loss of load transient of less than 2750 psia. For the loss of feed transient occurring from full power with the single failure of one auxiliary feedwater pump, the steam generator level reaches a minimum of 36.7%

of the tube bundle height 19.3 minutes after the low level trip occurs. This level provides adequate heat sink throughout the transient.

The results of the analysis meet the SRP 15.2.7 criteria and are, therefore, acceptable.

5.5 Loss of Coolant Flow The loss of coolant flow transient results are sensitive to initial overpower DNB margin, rate of flow degradation, low reactor coolant flow reactor trip setpoint, available scram reactivity, and MTC. For Cycle 11, the thermal power margin for the 100% power PDIL case is lower than the thermal margin for the FSAR design power distribution at full power conditions and, therefore, this event was calculated using the 100% power PDIL power distribution. The assumptions pertaining to rate of flow degradation, low flow trip setpnint, and MTC remain the same as in the reference safety analysis while the available shutdown margin assumed for Cycle 11 bounds the value assumed for the reference safety analysis. The minimum DNBR for the transient is 1.32.

l This value meets the criterion as stated in SRP 15.3.1 and 15.3.2 and, therefore, the staff concludes that the results of a loss of coolant flow event occurring during Cycle 11 are acceptable.

5.6 Full length CEA Drop Event The drop of a full length CEA is an A00 which relies on the provision of adequate initial overpower margin to assure no violation of the SAFDLs. The LCO syrinetric offset band is designed to restrict permissible initial operating conditions such that the SAFDL for DNB and fuel centerline melt are not exceeded for this event.

In order to cover all potentially limiting conditions, the CEA drop for Cycle 11 was analyzed from power levels ranging from 0 to 100% of full power.

Previous analyses (Ref. 10) have shown that the worst full length CEA drop with respect to DNB is the minimum worth CEA that results in the maximum increase in power peaking. Therefore, the Cycle 11 CEA drop evaluation was based on a CEA worth of 0.10% delta k/k. The results of the Cycle 11 DNB evaluation indicate that the limiting full length CEA drop is one initiated ,

from the positive edge of the 100% power symmetric offset LCO band. The minimum DNBR for this event is 1.37, well above the limiting minimum value of 1.20.

With respect to fuel centerline melt, the worst case full length CEA drop is one initiated from power distributions at the edge of the symmetric offset LCO band at each power level. The maximum allowable steady-state linear heat rate reouired to assure that the maximum linear heat generation rate after the drop does not violate the SAFDL of 23.2 kw/ft (for the fresh fuel) is used in deriving the LCO band on symetric offset for the RPS.

The safety analyses of the CEA drop event assumes tnat control of the turbine admission valves is performed menually. However, it is possible for the core power to return to a level higher than the pre-drop power level during a CEA drop transient if the turbine admission valves are in the automatic pressure control mode (IMPIN) of operation. Therefore, a separate Synnetric Offset operating band has been derived by assuming that the core power returns to the maximum level allowed by the Variable Overpower Trip Setpoint. This reduced cperating band applies to the Symetric Offset trip function whenever the IMP!N mode of turbine control is used.

The results of a CEA drop event meet the criteria stated in SRP 15.4.3 and are, therefore, acceptable.  :

1 5.7 Main Steam Line Break The main steam line break accident was analyzed in detail for Cycle 9 (Ref. i 13). The analysis was performed with RETRAN-02 MOD 2, which has been approved l for use by MYAPC. The analysis assumed a double-ended guillotine break in the l main steam line coincident with the worst single failure, a feedwater regulating  :

valve failure. The goal of the analysis was to determine if the core returns to I criticality after the initial reuctor trip. If the available trip reactivity and !

boron worth is larger than the reactivity due to moderator and Doppler defects )

at all times, adeouate margin exists to prevent recriticality.

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1 For Cycle 11, the nominal trip reactivity needed to avoid recriticality for HFP and HZP cases at BOC and EOC were determined. In all cases, the required trip reactivities are within the required shutdown margin Technical Specification for Cycle 11.

Since no return to criticality is predicted, the consequences of a main steam line break during Cycle 11 are acceptable.

5.8 Seized Rotor Accident The most significant safety parameters which affect the seized rotor accident 1

are the initial overpower DNB margin, core power distribution, radial pin power census, assumed rate of flow degradation, low reactor coolant flow trip setpoint, MTC and primary-to-secondary leakage flow rate. Most of these 3 factors remain unchanged for Cycle 11. The important differences are a j reduction in the initial overpower DNB margin, differerces in the radial pin i

power census, and a more limiting 100% power PDIL power distribution. The percentage of fuel experiencing DN8 using the Cycle 11 power distribution and the Cycle 11 pin census was less than 10.8% as compared to 10.3% for Cycle

10. The licensee also states that the radiological release analyses based on these figures would have consequences within the bounds of 10 CFR 100. The staff, therefore, finds this event to have acceptable consequences if occurring during Cycle 11, 5.9 Loss of Coolant

The Cycle 10 large break LOCA analysis is the reference LOCA analysis for 1

Cycle 11. The YAEC LOCA methodology used for Cycle 10 required break spectrum analyses for four different power shapes. Since CE supplied fuel is predominant in the Cycle 11 core as in Cycle 10, the hydraulic characteristics of the corc should be quite similar. Except for the moderator density defect, the physics parameters used in the Cycle 10 reference analysis are representative of Cycle 11. The variation of moderator density defect between Cycles 10 and 11 is well within the data uncertainty range. Therefore, the staff concurs that the Cycle 10 break spectrum analysis results are applicable to Cycle 11.

For each of the limiting breaks, a LOCA calculation was performed with input data specifically for Cycle 11. The results of the analysis for each axial powar shape indicate that the cladding temperature, cladding oxidation, and hydrogen generation values are in compliance with 10 CFR 50.46 Appendix K criteria.

Previous analyses have shown that small break LOCAs for Maine Yankee are nonlimiting. The results of these previous analyses are detemined primarily by the decay heat values which are insensitive to fuel type. In addition, sin':e the peak clad temperature was calculated to be well below the 10 CFR 50.46 criteria, it would not be significantly affected by slight differences in core configurations between cycles. The staff, therefore, concludes that the results of previous small break LOCA analyses for Maine Yankee are applicable to Cycle 11.

v 6.0 TECHNICAL SPECIFICATION CHANGES The licensee has proposed (Ref. 1) several changes to the Technical Specifications for the Cycle 11 reload core. The staff's review and evaluation of these changes follows with the numbering corresponding to that presented in Reference 1.

1. Technical Specification 2.2 (a) The steady state peak linear heat rates have been modified. This change is acceptable because the modification reflects the Cycle 11 SAFDLs for the prevention of centerline melting.

(b) The text has been slightly modified. This change is acceptable as it clarifies that each fuel type has its own LHGR limit.

2. Technical Specification 3.10 (a) A reduction in the allowable power dependent insertion limit (PDIL) for CEAs or a reduction in the LHR limits have been added as possible actions if the measured value of total radial peaking factor exceeds the value given in Figure 3.10-4. This is consistent with the Cycle 11 calculations of the sensitivity of LOCA limits to radial peaking factor given in the reload report and is acceptable.

(b) The equation for the calculated power level reduction if the incore monitoring system becomes inoperable has been modified. This is acceptable as it is consistent with the Cycle 11 power distributions and RPS setpoints.

(c) The PDIL given by Figure 3.10-1 has been modified. This change is acceptable because it reflects the Cycle 11 CEA insertion limits produced by the reload analysis.

(d) The allowable unrodded radial peak versus cycle average burnup, Figure 3.10-4, har been modified. This change is acceptable as it correctly reflects Cycle 11 radial peaking.

(e) The allowable power level versus the increase in total radial peak.

Figure 3.10-5, has been modified. This change is acceptable since it correctly reflects Cycle 11 power distributions and RPS setpoints.

(f) The required shutdown margin versus RCS boron concentration, Figure 3.10-7, has been modified. This change reflects the increased shutdown margin requirements at lower boron concentrations for Cycle 11 due to the increased moderator defect and is, therefore, acceptable.

(g) The reference power level versus nominal cold leg temperature, Figure 3.10-8, has been deleted since the inlet temperature programing has been eliminated from Cycle 11. This is acceptable since it provides a flexible core inlet temperature range from 500*

F to 552' F which is accounted for in the Cycle 11 safety analyses.

(h) Pages 3.10-2, 3.10-5, and Figures 3.10-9 through 3.10-12 have been revised. These revisions are acceptable since they are editorial in nature and provide consistency with the approved deletion of Figure 3.10-8.

7.0 EVALUAT_ ION FINDINGS The staff has reviewd the information presented in the Maine Yankee Cycle 11 reload report and in the MYAPC responses to the staff request for additional information. The staff finds the proposed reload and the associated modified Technical Specifications acceptable.

8.0 ENV!RONMENTAL CONSIDERATION This amendment involves a change in 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 involves no significant 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 Connission has previously issued a proposed finding that this amendment involves no significant hnards consideration and there has been no public concent on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), tio environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendinent.

9.0 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 conduc.ted in compliance with the Comission's regulations, and the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public. l

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10.0 REFERENCES

1. J.B. Randazza (MYAPC) letter to the U.S. Nuclear Regulatory Comission, MN-88-67, June 30, 1988.

2. "Maine Yankee Cycle 11 Core Perfonnance Analysis," YAEC-1648, Yankee Atomic Electric Company, July 1988.

3. USNRC letter to J.B. Randazza (NYAPC), Amendment No. 105 to DPR-36, June 23, 1988.

4 "Mechanical Design Report Supplement for Exxon Nuclear Maine Yankee XN-3 and XN-4 Extended Burnup," XN-NF-86-94(P) September 1986.

5. Safety Evaluation of the Exxon Nuclear Company Topical Report XN-NF-82-06(P), "Qualification of Exxon Nuclear Fuel for Extended Burnup," July 1986. t

6. "Maine Yankee Reactor Protection System Setpoint Methodology," YAEC-1110 September 1976. .

7. "Maine Yankee Core Thennal-Hydraulic Model Using COBRA III C." YAEC-1102, June 1976.

8. "The Hydraulic Perfonnance of the Maine Yankee Reactor Model,d TR-0T-34 Combustion Engineering June 1971.

9. Maine Yankee Atomic Power Station Final Safety Analysis Report (FSAR).

10. "Justification for 2360 MWt Operation of the Maine Yankee Atomic Power Station," YAEC-113?, July 1977.  :

11. Maine Yankee letter to USNRC, WMY 78-2, January 5, 1978.

12. Maine Yankee letter to USNRC, MN-82-53, "Boron Dilution During Hot and l Cold Shutdown (Mode 5 Operation)," March 18, 1982, i

13. "Maine Yankee Cycle 9 Core Performance Analysis," YAEC-1479, April 1985, i Date: September 27, 1988 l

Principal Contributor: 1.. Kopp, SRXB

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