ML20044A086
| ML20044A086 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/15/1990 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044A085 | List: |
| References | |
| NUDOCS 9006280051 | |
| Download: ML20044A086 (13) | |
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ENCLOSURE 1 l
1 SAFETYEVALUATIONBYTHEOFFICEOFNUCLEARREACTORRW,LATION RELATING TO TOPICAL REPORT PECO FMS-006
' METHODS FOR PERFORMING BWR RELOAD SAFETY EVALVATIONS' PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM UNITS 2 AND 3 DOCKET NOS. 50-277 AND 278
1.0 INTRODUCTION
By letter dated May 30, 1989, the Philadelphia Electric Company (PEco) submitted the Topical Report PEco FMS 006 " Methods for Performing BWR Reload Safety Evaluations" for NRC review. Additional information was submitted on May 31, 1990, inresponsetoastaffrequest(Refs.Iand2).
The PEco-FMS-006 is the last of a six topical report series prepared by PEco l
forBWRreloaddesignandlicensing.-Previoustopicalreports(allofwhich have been approved by the NRC) include:
steady-state thermal hydraulic analyses.. transient critical power ratios, steady-state fuel performance, BWR systems transient analyses and BWR reactor physics. The subject topical report describes the methodology used by PEco for implementation of these individual topical reports in BWR reload design and licensing. Thisincludes(among others)reactorabnormaloperationaltransientanalysesfor:
reactor pressure increase, reactivity anomalies, coolant flow increase, coolant flow decrease, coolant temperature decrease and coolant inventory decrease.
In addition this L
report includes the PEco-FMS-004 interface requirements necessary to justify l
the acceptability of the RETRAN model for licensing applications. These y
requirements were imposed by the NRC in its PECo-FMS-004 safety evaluation.
Restrictions to be observed in the application of this topical report are listed in Section 3.5.
9006280051 900615 DR ADOCK 0500 7
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i 2.0
SUMMARY
OF THE TOPICAL REPORT This topical report describes the computer codes used in BWR reload design and safety evaluation followed by the qualification of these codes and methods. A series of abnormal operational transients are analyzed and evaluated. Each one of these areas is sumarized below.
l 2.1 PEco Computer Codes and Methods The primary analysis end many of the linkage codes utilized by PEco were supplied by the Electric Power Research Institute (EPRI) as part of their reactor analysis support package (RASP) (Ref. 3). The package provides extensive capabilities in BWR core physics, fuel performance and thermal hydraulics, allowing for steady-state analysis, transient analysis, fuel management, core design, plant technical specification limits, and core reload analyses.
PECo has benchmarked these codes against plant data and higher order calculations.
The exceptions are FROSSTEY and TCPPECO which were not part of
- RASP, FROSSTEY has been benchmarked and approved in the PECo Topical Report PEco-FMS-003(Ref.4). TCPPECO calculates the steady-state and the transient time variable CPR using the results of RETRAN-02. TCPPECO has been qualified by PEco and approved by NRC (Ref. 5).
The following codes are also used:
- MICBURN:
One-dimensional, cylindrical geometry depletion to develop burnup dependent cross sections including gadolinium (Ref.6).
- CASMO-1 PECO:
Multi-group, two-dimensional transport theory lattice physicsprogram(Ref.7).
- SIMULATE-E-PECO:
Three-dimensional nodal analysis program. Models the reactor core as a matrix of neutronically coupled nodes (Ref.8).
l 3
- NORGE B PECO:
Automated data link between CASMO and SIMULATE (P.ef.
9),
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- PDQ-7 E:
Fine mesh neutron diffusion theory program, to estimatethelocal(pin) fission (power) distribution withinagroupofadjacentBWRfuelassenblies(Refs.
10,11).
- COPHIN:
Automated data link between CASMO and PDQ 7/ HARMONY l
(Ref.12).
' PINUP:
Production code for the 2x2 assembly geometry fuel l
pin fission rate distribution. Qualification and applicationsaredescribedinPEcoFMS-005(Ref.13).
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- PWEASY:
PEco developed, SIMULATE E post-processor data reduction; its qualification is discussed in Ref. 13.
- SIGMA-PEC0; TOPS:
PEco developed SIMULATE-E post-processors to compare SIMULATE-E predictions to in-core measurements. Both codes are discussed in Ref. 13.
- SIMTRAN-E-PECO:
One of the RASP group of codes which collapses radially three dimensional cross section data for one dimensional kinetics, (Ref. 3).
t
- RETRAN-02-PECO:
One-dimensional, thermal hydraulic transient analysis program, qualified in Ref 9.
- FIBWR:
Steady-state BWR thermal hydraulic analysis.
Qualification and verification is described in Ref.
14.
4 The linkage between the multidimensional steady state results and the one dimensional or point kinetics transient analysis is accomplished by calculational procedures qualified and approved in the physics method, (Ref.
13).
2.2 Qualification of the Reload Safety Evaluation Methodology The components of the pEco steady state and transient analyses methods have been individually qualified and approved by the NRC (Refs. 9, 13).
In this report it is shown that the linkage between the steady-state and transient analysis models preserves the reactivity effects and the power distribution predicted by the physics model.
I The three-dimensional simulation is the most accurate representation of all parameters. A model is described which preserves to a high degree of accuracy the feedback parameters and retains consistency with the one-dimensional kinetics model.
In a similar manner the void and doppler reactivity feedback are derived from the three-dimensional model. The control rod worth treatment is straightforward, by inputting the three dimensional values to the one-dimensional or point kinetics models.
The verification of the linkage of the steady-state to the transient model is accomplished by comparison to the three-dimensional results such as power distribution. Radial leakage effect in the one-dimensional model are accounted by using an evaluation of the axial nodal buckling. Other power distribution adjustments are based on the fission cross section.
Verification for each of the important parameters is discussed.
2.3 Reload Safety Evaluation for Abnormal Operational Transients Utilizing previously qualified and approved methods PEco integrated the components in the analyses of abnormal operational transients to determine their consequences and to evaluate the plant's c6pability to accommodate such l
events.
The following categories of such postulated events were analyzed:
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increase in reactor pressure core coolant temperature decrease reactivity and power distribution anomalies reactor coolant flow increase
' reactor coolant flow rate decrease reactor coolant inventory decrease ASME reactor pressure vessel overpressurization The generator load rejection without bypass has been identified as the most severe event with respect to the increase in reactor pressure.
Extensive analyses and consideration of uncertainties has been performed.
The loss of feedwater heating was identified as the most severe event in the category of core coolant temperature decrease. A description of the event and a quantified system response evaluation are given.
For the reactivity and power distribution anon.aly, a continuous control rod withdrawal error Ond a rotated and mislocated bundle cases were analyzed. The analytical method and the r
results of the analysis are presented. The recirculation flow controller failure was identified as the limiting event in the increase in reactor coolant flow category. An evaluation of the analytical results is given. The trip of the two motor generator set was identified as the limiting event in the reactor coolant flow rate decrease category and specifically analyzed.
Finally, in the ASME vessel overpressure event category, the simultaneous closure of all main steam isolation valves with simultaneous failure of the i
direct position switch scram signal was analyzed. A number of conservative i
assumptions are made in the analysis. The consideration of this event is with respect to potential overpressurization rather than violation cf the MCPR limits.
3.0 TECHNICAL EVALUATION
It has been pointed out in the preceeding that the subject topical report integrates the methods on: steady-state thermal-hydraulics, physics, fuel performance transients, critical power correlation and BWR systems perfor-mance. Topical reports in these individual topics have been approved by the
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6 NRC. Thus, the subject topical report describes the qualification of the implementation of the individual topicals in BWR licensing. To this end the report examines the linkage between the physics and transient analysis models to assure that the reactivity effect and the power distributions (predicted by the physics model) are preserved and examines the reload safety evaluation procedures to assure conservative operating limits.
3.1 power Distribution and Reactivity Feedback Effects The point kinetics version of the RETRAN code requires input tables for core reactivity parameters, i.e., feedback as well as control rod dependent.
Verification and validation of the integrated reload safety evaluation methodology is accomplished by comparison to the results of the three-dimensional simulator (Ref. 8) which has been approved by the NRC and has been extensively qub11fied against plant operating data.
Forthemoderatordensityandfueltemperature(i.e.,voidanddoppler reactivity feedbacks respectively) predictions, two identical RETRAN-02 and SIMULATE-E cases are run with the feedback variables substituted in RETRAN from SlitVLATE. This substitution results in the preservation of the three.
dimensional reactivity characteristics.
In this manner SIMULATE E void and doppler reactivity predictions were tabulated as a function of RETRAN moderator l
density and fuel temperature.
By this process it was established that the predictions of transient changes of power differ less than 3.4 percent which is acceptable. The control rod worth is straightforward, i.e., the SIMULATE rod worth is input to the RETRAN point kinetics model.
No further verification was performed and this is acceptable.
For the one-dimensional version of the RETRAN code, the one-dimensional cross l
sections are collapsed from the three-dimensional solution using the SIMTRAN code. The radial leakage effects are accounted by the fast radial nodal buckling. User supplied albedos are used in the axial reflector regions.
(Note:
the one-dimensional RETRAN solves two group diffusion equations, the three-dimensional SIMULATE solves the' nodal multigroup transport.) Finally, l
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PEco added power distribution adjustments through the fission cross sections.
These last adjustments are small in magnitude as a function of burnup.
Examples of axial power distribution from the one-dimensional and the corresponding collapsed three-dimensional models show that the differences are smell, thus this method is acceptable.
However, the feedback variables predicted by the one dimensional RETRAN model represent a single channel core and may not necessarily agree with the collapsed three-dimensional variables. To assure that the transformation preserves the three-dimensional reactivity a method similar to the RETRAN point kinetics has been erployed. The one-dimensional control rod worth predictions are in excel'ent agreement with the three dimensional calculation.
To verify the above methods, reactivity comparisons were compiled at about the middle and end of cycle for peach Bottom (Cycle 7, which was computed by GE) with variable pressures and control rod insertions and the results were in excellent agreement and, thus, are acceptable.
.3.2 Abnormal Operational Transients As we have seen previously pEco has classified the abnormal operational occurrence in seven categories and analyzed the limiting event in each one..
The results of these analyses establish the bases for future licensing.
3.2.1 Generator Load Rejection Without Bypass Generator load rejection without bypass has been identified as the limiting event for the reactor pressurization category.
In this event turbine control valve closure takes place which initiates reactor protection, however, it is assumed that the turbine bypass valves fail to operate.
Reactor pressure will increase very rapidly to the safety and pressure relief setpoint which terminates the pressure increase. The objectives of the analyses are to establish that the P.CPR remains above the safety limit and that the safety valve relief capacity is adequate. This is typically the limiting event in
8 the pressurization category. Conservative assumptions were made in the reactor protection system setpoints and the relief valve capacity. Approved methods have been used in the calculation of delta-CPR and a conservative method have been used in calculating the thermal hydraulic model uncertainties and the estimation of the net reactivity change. The statistical adjustment factors calculated in the MCPR determination are conservative. Application of this methodology in the reference cycle, i.e., Peach Bottom Unit 3 Cycle 7, yielded a delta-CPR of 0.35, which is within the available margin, and a peak reactor pressure of 1,217 psig, which is below the safety limit of 1,375 psig.
3.2.2 Decrease of Core Coolant Temperature The two most severe events in this category are the feedwater controller failure and the loss of feedwater heating.
Feedwater controller failure which results in the maximum flow will cause:
reactivity insertion due to moderator feedback, increased water level, turbine trip, reactor protection system actuation, and opening of the safety and relief valves to terminate the pressure increese. The objectives of the safety analyses for this event is to assure that the CPR and the maximum pressure are within the allowable limits. Using conservative assumptions and approved methods, application to the reference cycle yielded a delta-CPR of about 0.06 which is a reasonable value for this reactor type and 8x8 reload fuel. The peak reactor pressure is 1,165 psig, i.e., well within the safety limit of 1,375 psig.
However, as shown in the loss of feedwater heater analysis, it envelops the loss of feedwater controller case, thus, it will not be reevaluated on a cycle-by-cycle basis.
l In the loss of feedwater heating, the decrease of feedwater temperature I
results from the loss of one or more of the steam extraction feedwater l
heaters. The feedwater heating system design is such that, under single failure conditions the temperature decrease is limited to less than 100'F.
Analysis using the SIMULATE code and conservative assumptions resulted in a 1
9 i
reference cycle delta-CPR as a function of burnup which is in the expected range for this type of fuel.
3.2.3 Reactivity and Power Distribution Anorelies This category includes the continuous rod withdrawal during power operation, fuel loading error due to a rotated fuel bundle and fuel loading error due to a mislocated bundle.
During rod withdrawal operating CPR margin will be lost due to overall power increase and power redistribution. The event is terminated by the rod block monitor or when the rod is fully withdrawn. The SIMULATE code is used for this analysis and its application for the rod withdrawal error in the reference cycle resulted in a maximum delta-CPR of 0.33 which is in the expected range for this event.
The mislocated and the rotated bundle can cause a highly localized power distribution anomaly. For the rotated bundle PEco approved physics, thermal hydraulics and transient CPR methods have been used with conservative initial condition and parameter assumptions. The results were identical with the vendor provided results for this event and cycle.
Finally the bundle mislocation event has been generically analyzed in Reference 13 which has been approved previously.
3.2.4 Increase in Reactor Coolant flow Rate Increase of the coolant flow rate can only be caused by increased recircula-tion pump speed which could result from the recirculation flow controller failure (PeachBottomisequippedwithafluidcoupler).
The most severe event' corresponds to an initial operating power level of 57 percent and 43 percent core flow. The analyses performed with approved codes and conservative assumptions showed that this event is bounded by the generator load rejection without bypass which has been discussed previously.
r o.
10 3.2.5 Decrease in Reactor Coolant Flow Decrease in reactor coolant flow could result in inadequate heat removal and, thus, violate MCPR. The most severe event results from the simultaneous recirculation pump trip, which will be followed by turbine and feedwater pump trip (high water level), turbine stop valve closure and reactor scram.
Analyses showed that the peak reactor pressure is well below the safety limit and the delta-CPR is conservatively bounded by the generator load rejection without bypass.
3.2.6 Decrease in Reactor Coolant Inventory 1.oss of reactor coolant inventory could result in loss of capability to remove the heat generated in the core, and thus, threaten fuel integrity.
Such an event can be caused by failure of all three feedwater turbines. Reactor scram will be initiated on low water level followed by recirculation trip and HPCIS initiation. Analyses with approved codes and conservative assumptions showed that the lowest water level during this event is 90 inches above the top of the active fuel. The delta-CPR is conservatively bounded by the generator load rejection without bypass.
3.2.7 Vessel Overpressurization Event A simultaneous closure of all main steam isolation valves could result in a challenge to the reactor vessel pressure integrity.
In such a transient the objective of the analysis is to assure the adequacy of the pressure relief system to limit pressure increases below the requirements of Section III of the ASME code.
For an event of this frequency a conservative limit is 1,375 psig(i.e.,10%abovethemaximumoperatinglimit). Making conservative assumptions, such as initial power at 102% power, relief valve capacities at the minimum ASME specified values and scram setpoint at technical specification limits, and using the RETRAN code it was estimated that the peak reactor i
pressure is 1,262 psig, i.e., below the safety limit.
l 11 3.3 Evaluation Sumary We have reviewed the topical report methodology for power distribution, reactivity feedback effects and transient and accident analyses. We concluded that the analyses are performed with conservative assumptions and approved methods, and the results of the application to the reference cycle, i.e., Peach Bottom Cycle 7, are within the expected range. Thus, we find the topical report, as supplemented by responses tn staff questions in a May 31, 1990 t
submittal, to be acceptable.
3.4 PEco-FMS-004 to PEco-FMS-006 Interface Requirements In the NRC safety evaluation of the PEco FMS-004 six items were identified as lacking sufficient information to justify licensing applications.
These items are:
(a)theuseofthejetpumpmodelwithflowreversal,(b)theuseofthe onedimensionalkineticsoption,(c)simulationofrapidfeedwaterflow excursiontosimulateturbinetests,(d)qualificationofthealgebraicslip model, (e) justification of licensing conditions and safety margins, and (f) statistical analysis to justify the uncertainty in delta CPR calculations.
These items were addressed in Sections 4.1.1.3, 4.1.1.4, 4.2.1.4, 2.2, and 3.2.
In our review of these sections we found the corresponding topics acceptable, thus, we consider the PECo-FMS-004 requirements as addressed adequately and the issues closed.
3.5 Limitations The application of this topical report will be subject to the following limitations:
(1) It is only applicable to the Peach Bottom Units 2 and 3, (2) It is applicable only in conjunction with the approved topicals in the same series, i.e., PEco-FMS-001 to PEco-FMS-005, and r-.-.
12 (3)
It is applicable for average fuel assembly enrichments equal or lower than 3.5 w/o U-235.
4.0 REFERENCES
i 1.
Letter from D. R. Helwig, Philedelphia Electric Company to USNRC, "In j
House Reload Licensing for Peach Bottom Atomic Power Station," dated May 30, 1989.
2.
Letter from G. A. Hunger, Jr., Philadelphia Electric Company to USNRC, "In House Reload Licensing for Peach Bottom Atomic Power Station, Uni'.s 2 and 3," dated May 31, 1990.
3.
EPRI NP-1761, "The Reactor Analysis Support Package," R. E. Engel et al.,
EPRI, Final Report, dated May 1986.
4 PECo-FMS-003, " Steady-State Fuel Performance Methods Report," J. F.
Buckley, PEco, July 1987.
5.
PECO-FMS-002, " Method for Calculating Transient critical Power Ratios for BoilingWaterReactors(RETRAN-TCPPECO),"K.R. Young,etal.,PEco, l
November 1985.
6.
Ahlin, A. and M. Edenius *MICBURN-Microscopic Burnup in Gadolinia Fuel Pins," Advanced Recycle Methodology, Chapter 7, Part II, AB Atomerergy for EPRI, November 1975.
l 7.
PECo-FMS-CCM-003, "CASMO-1: A Fuel Assembly Burnup Program, User's Manual," A. Ahlin, et al., revised by S. R. Hesse, PEco, August 1987.
8.
PECo-FMS-CCM-001, " SIMULATE-E: A Nodal Core Analysis Program for Light i
Water Reactors, Computer Code User's Manual," revised edition by S. R.
Hesse, PEco, August 1987.
V 13 9.
'PEco-FMS-CCM-004, "NORGE-B, Code Description," revised by W. G. Lee, PECo, September 1987,
- 10. Poetschat, G., "Abreviated PDQ-7/HARMONf User's Manual with EPRI-ARMP Modifications," Advanced Recycle Methodology, Chapter 9A, Part !!, EPRI, March 1983.
- 11. Rothleader, B., *PDQ 7/ HARMONY User's Manual," Advanced Recycle Methodology, Chapter 9B, Part II, EPRI, March 1983.
12.
EPRI-NP-1385 "COPHIN Code Description," R. D. Moste11er and R. S.
Borland, EPRI, April 1980.
j 13.
PEco-FMS-005, " Methods for Performin9 BWR Steady-State Reactor Physics Analyses," S. R. Hesse, PECo, January 1988.
14.
EPRI-NP-1923, "FIDWR: A Steady-State Core Flow Distribution Code for Boiling Water Reactors, Code Verification and Qualification Report," A.
Ansari, et al., YAEC for EPRI, July 1981.
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