ML20137Y854
| ML20137Y854 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 09/30/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20137Y823 | List: |
| References | |
| NUDOCS 8510080106 | |
| Download: ML20137Y854 (15) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO AMENDMENT NO. 95 TO FACILITY OPERATING LICENSE NO. DPR-36 MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309
1.0 INTRODUCTION
By letter dated June 14, 1985 (Ref. 1), the Maine Yankee Atomic Power Company (MYAPC) submitted an application to modify the Technical Specifications for Maine Yankee to permit operation for a ninth cycle. A Cycle 9 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.
2.0 FUEL SYSTEM DESIGN The Cycle 9 reload application involves fuel designs similar to those pre-
"viously considered for the Maine Yankee reactor. The Maine Yankee Cycle 9 core will consist of 217 fuel assemblies with fuel rods arranged in 14x14 fuel types were fabricated by Combustion Engineering (CE)ycle 9 core, two arrays. Of the four fuel types proposed for use in the C and two were fabricated by Exxon Nuclear Company (ENC). The fresh reload fuel, Type N, consisting of 72 assemblies, was fabricated by Combustion Engineering. The other.CE fuel type, Type E, consists of a single fuel assembly previously irradiated during Cycle 2 and now loaded at the core centerline position for Cycle 9.
This assembly is expected to accumulate a maximum assembly average burnup of less than 33 GWD/MTU at end of Cycle 9.
The fresh reload fuel assemblies for Cycle 9 are similar in design to the CE Batch H fuel provided for Cycle 4, except the cold beginning-of-life shoulder gap has been increased by reducing the overall length of the fuel rod and the height of the lower end fitting. These changes will allow for additional fuel rod growth clearance.
The fresh fuel will be discussed further in the sections which follow.
The ENC fuel, denoted Types L, and M, consists of 144 fuel assemblies in the Cycle 9 core. These assemblies were previously irradiated during Cycles 7 and 8.
The ENC fuel exhibits some small mechanical differences from the CE fuel. These differences include thicker fuel rod cladding and slightly modified fuel pellet density and geometry.
In addition, Type L and M fuel assemblies are identical in mechanical design and have modified pellet length and dish geometries. The fuel rod plenum length in the Types L and M fuel has also been increased slightly due to the removal of the lower fuel rod insulator disc. The effect of the lower insulator disc was evaluated by 8510080106 850930 DR ADOCK O 39
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j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOP, REGULATION RELATING TO AMENDMENT NO. 85 TO FACILITY OPERATING LICENSE NO. OPR-36 MAINE YANKEE ATOMIC POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309
1.0 INTRODUCTION
By letter dated June 14,1985 (Ref.1), the Maine Yankee Atomic Power Company (MYAPC) submitted an application to modify the Technical Specifications for Maine Yankee to permit operation for a ninth cycle. A Cycle 9 core reload report (Ref. 2) das 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.
2.0 FUEL SYSTEM DESIGN The Cycle 9 reload application involves fuel designs similar to those pre-
'viously considered for the Maine Yankee reactor.
The Maine Yankee Cycle 9 core will consist of 217 fuel assemblies with fuel rods arranged in 14x14 arrays. Of the four fuel types proposed for use in the Cycle 9 core, two fuel types were fabricated by Combustion Engineering (CE) and two were fabricated by Exxon Nuclear Company (ENC). The fresh reload fuel, Type N, consisting of 72 assemblies, was fabricated by Combustion Engineering. The other CE fuel type, Type E, consists of a single fuel assembly previously l
irradiated during Cycle 2 and now loaded at the core centerline position for Cycle 9.
This assembly is expected to accumulate a maximum assembly average burnup of less than 33 GWD/MTV at end of Cycle 9.
The fresh reload fuel assemblies for Cycle 9 are similar in design to the CE Batch H fuel provided for Cycle 4, except the cold beginning-of-life shoulder gap has been increased by reducing the overall length of the fuel rod and the height of the lower end fitting.
These changes will allow for additional fuel rod growth clearance.
The fresh fuel will be discussed further in the sections which follow.
The ENC fuel, denoted Types L, and M, consists of 144 fuel assemblies in the Cycle 9 core. These assemblies were previously irradiated during Cycles 7 and 8.
The ENC fuel exhibits some small mechanical differences from the CE fuel. These differences include thicker fuel rod cladding and slightly modified fuel pellet density and geometry.
In addition, Type L and M fuel assemblies are identical in mechanical design and have modified pellet length and dish geometries. The fuel rod plenum length in the Types L and M fuel has also been increased slightly due to the removal of the lower fuel rod insulator disc. The effect of the lower insulator disc was evaluated by
. the fuel vendor, ENC, and was found to be inconsequential relative to the fuel The ENC fuel to be used in Cycle 9 was found acceptable J
design performance.in our evaluations for Cycles 7 and 8 and remains acceptable for irra in Cycle 9.
As in Cycles 7 and 8, the Cycle 9 core will contain 81 control element assemblies (CEAs) of which four are non-scrammable. A change for Cycle 9 is that the four part-strength non-scramable CEAs have been replaced with full-strength CEAs.
In addition to the CEAs, the Maine Yankee Cycle 9 core will also contain burn-There will be 1264 standard 8 C-AL 023 able poison rods in selected assemblies. burnable poison rods in Cycle 9 2.1 Fuel Mechanical Design The mechanical design features of both CE and ENC fuel assemblies to be used The results of in the Cycle 9 core are listed in Table 3.3 of Reference 2.
the fuel mechanical design analyses performed for the CE Batch N fuel are These include evaluations of fuel cladding collapse, reported in Reference 2. irradiation induced dimensional changes, cladding strain a The results indicate the primary and maximum fuel rod internal pressure.
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 coolant system pressure throughout the projected life-time of the fuel. These results were calculated with models approved for earlier reloads, and are in conformance with the requirements of SRP 4.2.
The ENC fuel mechanical design was approved for Cycles 7 and 8.
We, therefore, conclude that the Maine Yankee Cycle 9 fuel mechanical design is acceptable.
2.2 Fuel Thermal Design As discussed in Section 2.0 of this report, the fresh, Batch N fuel in the Maine Yankee Cycle 9 core is nearly identical to that previously irradiated The licensee's analysis of the fuel thermal performance is in the reactor.
also,the same as that used in previous reload analyses including the use of The fuel power history effects and burnup-dependent fission gas release.
thermal design analyses have been performed using methodology previously approved by the staff. As a result, we find the fuel thermal design analyses for Maine Yankee Cycle 9 acceptable. This finding includes both power-to-centerline melt and core average gap conductance calculations.
3.0 NUCLEAR DESIGN i
3.1 Core Characteristics Maine Yankee Cycle 9 continues to use a low-leakage design, achieved by placement of fresh (unirradiated) fuel assemblies in selected core interior locations and In addition to burned (irradiated) fuel assemblies on the core periphery.
reducing the irradiation exposure to the reactor pressure vessel this low-leakage core design also produces a less severe moderator defect with near E0C, and extends the achievable full power lifetime of the cycle. Cycle 9 is expected to attain a cycle average full power lifetime of 13,400 MWD /MTV.
I
' 1 3.2 Power Distributions Hot full power (HFP) fuel assembly relative power densities are given in Reference 2 for beginning of cycle (BOC) (50 MWD /MTU), middle of cycle (MOC)
(6000 MWD /MTU), and E0C (12,000 MWD /MTU) conditions and for unrodded and rodded (CEA Bank 5 in) configurations. These results show that the unrodded maximum radial peaks for Cycle 9 are comparable with the Cycle 8 values and the rodded radial peaking is 2-4% larger. The increased rodded radial peaking is the result of replacement of the four part strength CEAs in the reactor regulating non-scrammable bank subgroup with full strength CEAs. The calculated radial peaking (with uncertainties) versus exposure is included in the Technical Specifications (Figure 3.10-4). This ensures peaking will not exceed the valtas used in the safety analysis.
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 9 core are compared with the corresponding values of Cycle 3 (reference cycle Ref. 9) and Cycle 8 (previous cycle) in the Cycle 9 Core Performance Analysis Report (Ref. 2).
The MTC at nominal operating HFP and HZP B0C conditions is more positive than the corresponding previous cycle (Cycle 8) value primarily because of the increased 80C critical baron concentration resulting from more excess reactivity in the core. The EOC values are similar.
Technical Specification MTC limits are provided for Cycle 9 based on a new LOCA analysis moderator 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 9 Doppler coefficients are quite similar to the Cycle 3 and Cycle 8 values. The critical boron concentrations for Cycle 9 at BOC are greater than those of Cycle 8 because of the greater amount of excess reactivity in the core. Values of the delayed neutron fraction and prompt neutron generation time for Cycles 3, 8, and 9 are i
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, we find these data to be acceptable.
3.4 Control Reouirements The value of the required shutdown margin is determined either from the steam line break analysis (EOC) or from other safety analyses (BOC). Based t
on these values of required shutdown margin and on calculated available scram reactivity including a maximum worth struck rod and appropriate calculational uncertainties, sufficient excess exists between available and required scram reactivity for all Cycle 9 operating conditions. These results are derived by approved methods and incorporate appropriate 6
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=
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g.w.--g r,-
, assumptions and are, therefore, acceptable. The Power-Dependent 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 PDIL CEA worth is made when determining the available scram CEA worth.
For Cycle 9, group insertion limits are more restrictive than for Cycle 8, resulting from increased radial peaking under rodded and post-CEA drop conditions.
3.5 Augmentation Factors The augmentation factors provided in Reference 2 are incorporated as a power spike penalty in the calculation of the core power-to-incipient fuel center-line melt. The set of augmentation factors calculated for the new fuel used the approved CE model used for earlier cycles. Those for the ENC fuel used methodology approved for Cycles 5 through 8.
Based on the favorable comparison of the Cycle 9 with the Cycle 8 factors and because the factors have been derived using approved methods, we conclude that the augmentation factors are acceptable.
4.0 THERMAL-HYDRAULIC DESIGN 4.1 Thermal-Hydraulic Analysis The steady-state and transient departure from nucleate boiling (DNB) analyses were performed using the COBRA-IIIC computer program.
COBRA-IIIC was developed by Battelle Northwest Laboratory for use in the thermal-hydraulic analysis of nuclear fuel elements in rod bundles. The application of COBRA-IIIC to the Maine Yankee thermal-hydraulic design is described in References 3 and 4.
The computer program was also used on a one-eighth core assembly-by-assembly model to determine hot assembly enthalpy rise flow factors. This model explicitly represented each fuel assembly in the location it will reside for Cycle 9 operation.
In the Cycle 8 analysis, only the low flow region of the core was explicitly represented due to a limitation in the dimensional array size in the COBRA-IIIC code. For the Cycle 9 calculation, the dimensional array size was expanded to allow explicit representation of the entire one-eighth core. Test cases were run with the revised COBRA-IIIC code. These were compared with results from the original code and were found to be consis-tent. We find this analysis acceptable because COBRA-III-C is approved and the dimensional modifications made for the Cycle 9 analysis gave consistent results.
The inlet flow maldistribution 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 5 and 6.
The resulting hot assembly flow factors for the fresh CE assemblies as well as the single E-16 CE assembly was 1.0 for all power distributions. A 0.964 enthalpy rise flow factor is applied to all ENC assemblies due to higher spacer loss coefficients relative to the CE fuel. These factors are applied to the inlet mass velocity in the hot channel model in predicting DNB performance.
4.2 Fuel Rod Bowing A parameter which is considered in the thermal-hydraulic design is rod-to-rod j
bowing within fuel assemblies. For CE fuel assemblies, criteria for evaluating
l '
the effect of rod bow on DNBR do not require a penalty for bundle average burn-ups below 24000 MWD /MTU. No penalty is therefore required for the fresh CE Batch N fuel for Cycle 9, which will receive exposures well below this level.
For the single Type E assembly, the maximum channel gap closure due to fuel rod bowing was calculated to be 21.9% for Cycle 9.
For the ENC fuel assemblies, using the methodology of Reference 7 which was approved by Reference 8, the maximum gap closure due to fuel rod bowing is less than 30 percent.
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 percent. Therefore, no rod bow penalty is required for any of the fuel types.
4.3 Increase in Allowable Core Inlet Temperature For Cycle 9 operation the licensee has proposed an increase in the maximum allowable indicated core inlet temperature from 550*F to 552 F.
To accomodate this, the inlet temperature of 552*F was used in the accident analysis, which is acceptable.
5.0 SAFETY ANALYSES Maine Yankee has reviewed the parameters which influence the results of the transient and accident analyses for Cycle 9 to detemine which transients or accidents, if any, require a reanalysis. The parameters of importance are initial operating conditions, core power distributions, reactivity coefficients, shutdown CEA characteristics, and Reactor Protection System (RPS) setpoints and time delays.
For those events where the parameters for Cycle 9 are outside the bounds considered in previous safety analyses, a new or revised analysis was perfomed. These are:
(1) Boron Dilution (2) CEA Ejection (3) CEA Withdrawal (4) CEA Drop (5) Steam Line Rupture (6) Large Break LOCA In addition, other transients that require a partial reanalysis or review were the Seized Rotor, Excess Load, Loss of Load, Loss of Flow, Steam Generator Tube Rupture, and Small Break LOCA events.
5.1 CEA Withdrawal Event The CEA withdrawal is an anticipated operational occurrence (A00) for which the reactor protection system (RPS) is relied upon to assure no violation of the specified acceptable fuel design limits (SAFDLs). The most severe CEA with-drawal transient occurs for a combination of reactivity addition rate and time in core life that results in the slowest reactor power rise to the level just studycoveredtherangeofMTCsfrom+0.5xIggesafetyanalysis~2grametr below the Variable Overpower Trip. The referen
/ F and g /*F to -3.0 x 10 3#
reactivity addition rates from 0 to 0.7 x 10 e /sec. The CEA Bank 5 division into two separately moveable subgroups for Cycle 7 created a new class of CEA
bank withdrawal events, withdrawal of a CEA bank subgroup.
Complete withdrawals l
of each of the CEA Bank 5 subgroups were analyzed from initial conditions corresponding to the limiting power distributions within the symetric offset alarm bank for each power level. The withdrawal of CEA Bank SB, which was changed from part strength to full strength CEAs for Cycle 9, was the worst case. This subgroup is symmetrically located around the center in the inner portions of the core. Withdrawal of this subgroup causes an increase in peaking towards the center of the core which does not produce a full contribution to the excore monitors.
The MDNBR of 1.42 for withdrawal of CEA Bank 5B from its insertion limit at 47% rated power was the worst case. This is well above the MDNBR limit of 1.20.
This analysis, using approved methods and assumptions, assures that the SAFDLs are not violated and is, therefore, acceptable.
5.2 Uncontrolled Boron Dilution An inadvertent baron dilution will reduce the boron concentration in the primary coolant which in turn will increase the reactor core positive i
reactivity. During power operation, the resulting reactivity insertion will increase the reactor power and automatic safety systems will act to shut down the reactor and maintain the plant within safety limits. However, a boron dilution event during shutdown will nqt be mitigated by any automatic safety systems.
If it is allowed to continue unmitigated, it will result in reactor recriticality unless the operator takes appropriate corrective action
.to stop the dilution within the necessary time period.
The licensee indicated that the boron dilution event was analyzed for the following operating modes:
(1) refueling (2) cold shutdown - filled RCS (3) cold shutdown - drained RCS i
(4) hot shutdown - filled RCS (5) hot shutdown - drained RCS
(.6 ) startup (7) hot standby (8) power operation (9) failure to borate prior to cooldown The assumptions made in the Cycle 9 evaluation are consistent with those made in References 9 and 10. 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 9 core loading, the critical boron concentration under cold conditions (68'F) during refueling is 1086 ppm. The minimum initial reactor vessel boron concentration which will prevent an inadvertent criticality within 30 minutes is 1644 ppm, which therefore is required for refueling. There is, therefore, ample time for the operator to acknowledge the audible count rate signal and take corrective action. Dilution during shutdown conditions with the RCS filled was addressed in Reference 11 and remains unchanged for Cycle 9.
Dilution during shutdown conditions with the RCS partially drained was addressed in References 10 and 13. The licensee has shown the boron concentrations required to meet the 5% ak/k Technical Specification sub-criticality 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 licensee has stated that administrative procedures ensure that the. higher of these two values is used 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 limiting boron dilution from drained conditions.
To evaluate the boron dilution event during hot standby, startup, and power operation for Cycle 9, the licensee indicated that the same assumptions were used as in the analysis in Reference 9 except for the inverse boron worth and higher critical boron concentration (1557 ppm) at hot standby. Based on the maximum reactivity insertion rate, it would take approximately 56 minutes of continuous dilution at the maximum charging rate to absorb a 3.2% a k/k shutdown margin (minimum Technical Specification value is 3.1% A k/k).
Failure to add boron during cooldown was evaluated based on conservative values of MTC, initial temperature, and maximum cooldown rate.
In order to achie~ve criticality from these initial conditions, the temperature l
reduction requires approximately 63 minutes.
Based on the acceptability of the operator response times and comparison with Cycle 8 analysis, we conclude that the results for Cycle 9 are acceptable.
5.3 Excess Load Event The excess load event occurs whenever there is a 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. The transient which causes the most severe power excursion has been identified by the licensee as the steam dump and bypass system malfunction at hot standby and at E0C where the MTC is most negative.
The excess load transient had been reagalyzed for Cycle 4 in Reference 12.
In that analysis, a MTC of -3.17 x 10 A k/k"F was assumed. This value is more negative than that predicted for Cycle 9, including uncertainty. The MDNBR for the most limiting transient is 1.63 and corresponds to an event initiated from the positive edge of the synnetric offset band at approximately 100%
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 we therefore conclude that they are acceptable.
5.4 Loss of Load Event The loss of load event is an undercooling transient that results from station separation from the grid, turbine trip or electrical generator malfunctions.
Subsequent closure of the main steam stop valves causes a large mismatch between reactor power output and heat removal capacity. System parameters
. which have a major impact on the severity of the pressure excursion are the initial power level, initial RCS pressure, steam generator pressure, primary and secondary safety relief valve capacities and setpoint, and MTC.
For Cycle 9 two changes are being implemented which affect this transient. These are an increase of 2*F in allowable cold leg temperature and the number of steam generator tubes which could be plugged while maintaining a peak system pressure of 2750 psia.
The peak RCS pressure, 2750 psia, and MDNBR 1.69 were calculated with 180 tubes per steam generator plugged. Because these values are within the NRC acceptance criteria of SRP Section 15.2.1, we find this analysis acceptable.
5.5 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 heat up transient. The MONSR calculated for this event for Cycle 9 is 1.69.
Peak RCS pressure for this transient is bounded by the loss of load transient pressure of 2750 psia.
5.6 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 shutdown reactivity, and MTC. The limiting overpower DNB margin within the LCO envelope for Cycle 9 is greater than assumed for the FSAR analysis, while the rest of the parameters remain the same as in the Reference Safety Analysis (Ref. 9). The available shutdown margin assumed for Cycle 9 bounds that value =c.e
- Tor the reference analysis.
The MDNBR for the transient is greater than 1.51.
This value meets the criterion as stated in SRP 15.3.1 and 15.3.2 and, therefore, we conclude that the results of a loss of coolant flow event are acceptable.
5.7 Full Lenoth CEA Drop Event The drop of a full length CEA is an anticipated operational occurrence (A00) which relies on the provision of adequate initial overpower margin to assure no violation of the specified acceptable fuel design limits (SAFDLs). The LC0 syntetric offset band is designed to restrict pennissible initial operating conditions such that the specified acceptable fuel design limits (SAFDLs) for DNB and fuel centerline melt are not exceeded for this incident.
Previous analyses (Ref. 9) 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 9 CEA drop evaluation was based on a worth of 0.10 percent ap. The results of the Cycle 9 DNB evaluation indicate that the limiting full length CEA drop is one initiated from the positive edge of the 100 percent power symetric offset LCO alann~ band.
The MDNBR for this event is 1.29, 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 symetric offset LC0 band at each power level. The maximum allowable steady-state linear heat rate required to assure that the maximum linear heat generation rate after the drop does not violate the SAFDL of 23 kW/ft (for the fresh fuel) is used in deriving the LC0 band on symetric offset for the RPS.
The safety analyses of the CEA drop event assumes that control of the turbine admission valves is performed manually. 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 symetric 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 operating band applies to the Symetric Offset trip function whenever the IMPIN mode of turbine control is used.
J 4
The results of a CEA drop event meet the criteria stated in SRP 15.4.3 and are, therefore, acceptable.
5.8 Main Steam Line Break The main steam line break accident was analyzed in detail for Cycle 6 (Ref.14).
The analysis assumed a full dotble-ended guillotine rupture at the steam l
generator nozzle. The break results in a sudden decrease in back pressure and i
a large increase in the steam generation rate in the steam generator, i.e.,
increase in the heat removal from the RCS through the steam generator. The l
RCS temperature drops rapidly causing the RCS fluid volume to shrink. RCS cooldown and shrink will continue until the steam flow out of the break is i
reduced or teminated.
l The Cycle 9 steam line break accident was performed with RETRAN-02 M002, which has recently been approved for use by MYAPC (Ref.15). Five single failure cases were chosen for the steam line break to represent the worst possible 4
combinations of RCS cooldown and low boron concentrations for Cycle 9.
The case which gave the least margin (0.7% delta rho) was a feedwater regulation valve failure coincident with a steam line break from HZP. Since no return to criticality is predicted, the results of the steam line break analysis i
are acceptable.
5.9 Steam Generator Tube Rupture The analysis of the steam generator tube rupture event presented in reference 9 and the FSAR was reviewed for its applicability to Cycle 9.
Initial system pressure and the time of reactor trip have the most significant effect on l
primary system response.
For Cycle 9, the nominal operating pressure is the same as tnat used in the reference analysis.
In the reference analysis, the reactor trip occurred at the thermal margin trip setpoint. The results of the analysis adequately bound the primary system response to a steam generator tube rupture during Cycle 9.
l 4
1 I
i 5.10 Seized Rotor Accident The most significant safety parameters which affect the seized rotor accident i
are the initial overpower DNB margin, core power distribution, radial pin j
power census, assumed rate of flow degradation, low reactor coolant flow trip set-point, MTC, and primary-to-secondary leakage flow rate. Most of these factors remain unchanged for Cycle 9.
The important differences are a reduction in the initial overpower DNB margin Cycle 9 but is bounded by the assumed value of +0.5 x 10"js more p and differences in the radial power pin census.
The MTC w /*F.
The per-centage of fuel experiencing DNB using the FSAR power distribution and the The l
Cycle 9 pin census was less than 7.5% as compared to 4% for Cycle 8.
licensee also states that the radiological release analyses based on these j
figures would have consequences within the bounds of 10 CFR 100. We, there-i j
fore, find this event to have acceptable consequences, t
5.11 CEA Ejection Event f
As a result of the highest ejected CEA worths and the increased post ejection i
power peaking, the CEA ejection physics parameters have become more limiting i
i for Cycle 9 as compared to those assumed in the reference safety analyses 4
(Ref.9). Therefore, a reanalysis of the CEA ejection event occurring from both HZP and HFP for BOC and E0C core conditions was performed by MYAPC.
l This reanlaysis made use of recently approved revisions to the methodology t
(Ref. 16). All cases resulted in a radially averaged fuel enthalpy below j
the acceptable criterion of 280 cal /gm prescribed in Regulatory Guide 1.77
)
(Ref. 17). A bounding radiological release calculation has shown that the resulting off-site doses are within 10 CFR 100.
i 5.12 Loss of Coolant Accident (LOCA)
The LOCA analyses performed for Cycle 5 through Cycle 8 (Refs. 13, 18, and 19) are used as the reference analyses for Cycle 9.
The Cycle 9 core is slightly i
different from the reference core for the LOCA (Cycle 5) in the area of core
}
hydraulics and core physics. These differences were discussed in detail in l
Sections 3 and 4 of Reference 2.
Additional analyses were perfomed by the licensee to account for these minor differences. We reviewed these analyses l
and find them acceptable. Based on the results of these analyses, it is concluded that Appendix K criteria are met for Cycle 9 fuel types operating up to the I
following limits:
i 13.5kW/ftforfgreaterthan0.50andCAB Fresh Fuel less than or equal to 792 MWD /MTU i
j 14.0kW/ftforfgreaterthan0.50andCAB f
greater than 792 MWD /MTU 16.0 kW/ft for f less than or equal to 0.50
- - - + -...
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+,wn_ men-,---
-e.-n,,,-.,----~,an.-,-,,,--n--..,-v.-.~--..-~n.
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' Exposed Fuel 14.0 kW/ft for f greater than 0.50 16.0 kW/ft forf less than or equal to 0.50 WherefisfractionofcoreheightandCABiscycleaverageburnup.
Fresh fuel is comprised only of CE Type N fuel. Exposed fuel is comprised of ENC Types L and M fuel and CE Type E fuel.
6.0 TECHNICAL SPECIFICATION CHANGES The licensee has proposed (Ref.1) a number of changes to the Technical Specifications for Cycle 9 reload core. Our review and evaluation of these r
changes follows with the numbering corresponding to that presented in Reference 1.
1.
(a) Reference to.part-strength control assemblies has been removed.
This change is acceptable because it reflects a design change.
The full strength CEA replacements were used in the Reload Analysis.
2.
(a) The thermal margin / low pressure trip coefficients have been modified.
j This change is acceptable because it reflects Cycle 9 power i
distributions, with additional margin for future cycles.
(b) Figures 2.1-la and 2.1-1b have been modified. This change is acceptable because it reflects Cycle 9 power distributions, with additional margin for future cycles.
(c) Figures 2.1-2 has been modified. This change is acceptable because it reflects Cycle 9 analyses to prevent the fuel from exceeding i
linear heat rate limits.
It also contains additional margin for future cycles.
3.
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 9 Specified Acceptable Fuel Design Limits for prevention of center-line melting.
4 Technical Specification 3.10 (a) The description of the maximum reactor inlet temperature assumed in the safety analysis has been modified. This change is accept-able because it reflects the 2*F increase in cold leg temperature assumed in the Cycle 9 analysis.
(b) Figure 3.10-1 has been modified. This change is acceptable because it reflects the Cycle 9 CEA insertion limits produced by the reload analysis.
' (c) Figures 3.10-2 and 3.10-3 have been modified. This change is acceptable because it reflects Cycle 9 power distributions with additional margin for future cycles.
(d)
Figure 3.10-4 has been modified. This change is acceptable because it reflects Cycle 9 radial peaking.
(e) Figure 3.10-5 has been modified. This change is acceptable because it reflects Cycle 9 power distributions and RPS setpoints.
(f) Figures 3.10-6 and 3.10-8 have been modified. This change is acceptable because it reflects the 2*F increase in cold leg temperature assumed in the Cycle 9 analysis.
(g) Figures 3.10-9 and 3.10-10 have been modified.
This change is acceptable because it reflects Cycle 9 power distributions with additional margin for future cycles.
(h) Figure 3.10-11 has been modified. This change is acceptable because it reflects moderator temperature coefficients used in the Cycle 9 LOCA analysis.
7.0 EVALUATION FINDINGS The staff has reviewed the information presented in the Maine Yankee Cycle 9 reload reports and in discussion with MYPAC and Yankee Atomic personnel. We find the proposed reload and the associated modified Technical Specifications acceptable.
8.0 FNVIRONMENTAL CONCUISION 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 Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public cocrent 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), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
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 conducted in compliance with the Comission'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.
Date:
September 30, 1985 Principal Contributor:
M. Dunnenfeld
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10.0 REFERENCES
1.
J. B. Randazza (MYAPC) letter to the Director, Nuclear Reactor. Regulation (NRC), June 14, 1985, i
2.
" Maine Yankee Cycle 9 Core Performance Analysis", Yankee Atomic Electric Company Report YAEC-1396, April 1985.
3.
P. A. Bergeron, D. J. Denver, " Maine Yankee Reactor Protection System f
Setpoint Methodology", YAEC-1110, September 1976.
4 R. N. Gupta, " Maine Yankee Core Thermal-Hydraulic Model Using COBRAIIIC,"
YAEC-1102, June 1976.
j 5.
Combustion Engineering Report, TR-DT-34, "The Hydraulic Performance of the Maine Yankee Reactor Model", June 1971.
i 6.
Maine Yankee Atomic Power Station Final Safety Analysis Report (FSAR).
7.
" Computation Procedures for Evaluating Fuel Rod Bowing", XN-75-32(NP) 4 Supplement 2, July 1979.
i I
8.
Memorandum from L. S. Rubenstein to T. M. Novak, SERs for Westinghouse,
)
Combustion Engineering, Babcock and Wilcox, and Exxon Fuel Rod Bowing i
Topical Reports, October 25, 1982, f
9.
P. A. Bergeron, P. J. Guimond, J. DiStefano, " Justification for 2630
{
MWt Operation of the Maine Yankee Atomic Power Station", YAEC-1132, l
July 1977.
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o 4
I
- 11. Maine Yankee letter to USNRC, MN-82-53, " Boron Dilution During Hot and Cold Shutdown (Mode 5 Operation)", March 18, 1982, i
i i
i
- 13. Maine Yankee letter to USNRC, WMY 79-143, December 5, 1979; Attachment A, YAEC-1202, " Maine Yankee Cycle 5 Core Perfomance Analysis".
October 29, 1981.
- 15. Memo for G. C. Lainas from R. W. Houston, SER on YAEC-1447, August 16, 1985.
I
- 16. Memo for G. C. Lainas from L. S. Rubenstein, Safety Evaluation of l
YAEC-1464, June 20, 1985.
~
- 17. " Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors", NRC Regulatory Guide 1.77, May 1974.
1
- 18. " Maine Yankee Cycle 7 Core Perfomance Analysis", Yankee Atomic Electric l
Company Report YAEC-1324, September 1982.
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l
YAEC-1259, " Maine Yankee Cycle 6 Core Performance Analysis".
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