ML20043D268
| ML20043D268 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/31/1990 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9006070316 | |
| Download: ML20043D268 (11) | |
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PHILADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.
WAYNE, PA 19087 5691 (at s) s40 6000 May 31, 1990 Docket Nos. 50-277 50-278 U.S. Nuclear Regua tory Commission Attn: Document Control Desk Washington, DC 20555 o
SUBJECT:
In-House Reload Licensing for Peach Botton Atomic Power Station, Unita 2 and 3
REFERENCE:
Report: PECo-PMS-0006, " Methods for Performing BWR Reload Safety Evaluations,"
submitted to NRC on May 30, 1989 Dear Sir On May 21, 1990 representatives of Philadelphia Electric Company (PECo) met in Rockville, MD with representatives of the NRC Nuclear Reactor Regulation staff to discuss the referenced report.
At the conclusion of the meeting, the NRC representatives indicated a desire for PECo to respond in writing to several requests.
Each NRC request is restated in the attachment to this letter followed by PECo's response.
If you have any questions or require additional information, please do not hesitate to contact us.
Very truly yours,
^1j, O, i,
G. A. Hunger /"Jr.
Manager-Licensing Nuclear Engineering & Services Attachment cc T. T. Martin, Administrator, Region I, USNRC J. J. Lyash, USNRC Senior Resident Inspector 7
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Page 1 of 10 bTTACHMENT Docket Hos. 50-277 50-278 NRC Request No.
1:
The analytichl results presented in the report were determined primarily for the PB3C7 reference cycle.
Justify the application of the methods described in the report to future cycles and other BWRs.
Response
Typically, on a cycle-by-cycle basis, the configuration of the reactor core changes.
This change consists of the insertion of frech fuel that may differ from previous cye.l es in type and/or quantity. This may result in different core nuclear characteristics (i.e. void, Doppler, scram reactivity parameters) and different safety limits from the previous cycle.
The methods described in the subject
- report, however, explicitly account for these cycle-specific differences (changes in plant physical configuration that may affect reload licensing analysis are addressed in response to NRC Request 9).
Furthermore, PECo core monitorir.g activities routinely assess the accuracy of the steady-state physics models used in the evaluation of the core reactivity parameters.
- Thus, the methods employed to perform reload licensing calculations are continuously monitored to ensure their applicability to future reloads. This approach is consistent with the current NRC approved licensing basis for PECo reactors.
Thus, the methods described are applicable to future reloads (i.e. cycles).
The analytical methods described in the report are generic in nature and, thus, are applicable to other similar BWRs (i.e. Limerick).
However, PEco acknowledges that application of these methods to other BWRt: Will require prior approval from the NRC.
NRC Request No.
2:
The void and Doppler veritication for the RETRAN point kinetics model was performed by comparing the equilibrium end result of a slow transient to the 3-D simulator result.
Since the limiting licensing events are rapid power increase transients, explain the applicability of this verification to reload licensing.
Response
The application of the RETRAN point kinetics model is limited to the analysis of relatively slow transient events such as loss of f eedwater-flow. Consequently, the void and Doppler point reactivity i
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Page 2 of 10 verification discussed in section 3.1.2 of the subject report is restricted to the methods used in the analysis of relatively slow reactor transient events.
In addition, it should be noted that within the scope of the report, all transient events analyzed are less than 15 minutes in duration.
Thus, the transient effects of xenon are insignificant and may be ignored. The RETRAN 1-D kinetics option, which has been qualified against rapid power increase test data, is utilized for the analysis of rapid power increase events.
The fast events (e.g., generator load rejection) are typically the limiting licensing events.
The void and Doppler reactivity.
verification of the PEco 1-D methods is discussed separately in section 3.2.4 of the report.
NRC Request No.
3:
The qualification and application of the 1-D kinetics methods are described at end-of-cycle (EOC) and EOC-2000 MWD /T exposure statopoints only.
Since some transient events are more severe at other cycle exposures, explain why other exposure statepoints were not analyzed.
Response
The RETRAN 1-D kinetics option is utilized in the analysis of rapid power increase transients (i.e. GLRWOB, FWCF, MSIVC).
For GE designed BWRs (e.g. Peach Bottom, Limerick), it has been demostrated (1) that these events are most severe at end-of-cycle or near end-of-cycle conditions when the reactor control rod density (i.e.
number of control rods in the core) is minimal.
Certain transient events, such as the power anomaly events described in section 4.3 of the subject report, may be more severe at other cycle exposures.
These events are not analyzed with RETRAN but with the quasi steady-state methods outlined in section 4.3 of the reperc.
The methods explicitly account for the exposure depender.ce of these transient events.
(1)
General Electric, " Qualification of the one-Dimensional Core Transient Model for Boiling Water Reactors",
Volume
- III, NEDO-24154-A, General Electric Company, October, 1978.
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i Page 3 of 10 NRC Request No.
4:
Provide further justification that the uniform fuel pellet radial power distribution is conservative.
Response
A sensitivity analysis was performed utilizing RETRAN to demonstrate that a uniform fuel pellet radial power distribution is conservative for licensing application.
The analysis consisted of evaluating the fuel surface heat flux response to a step change in power generation with both a uniform radial power distribution and a linear, outside edge peaked (1.1 peaking factor) radial power distribution.
The fuel rod thermal time constant (time to reach 0.63 of surface heat flux change) for each case was compared.
The thermal time constant for the uniform power distribution case was determined to be approximately 5% greater than that for the edge peaked power distribution case.
Thus, the use of a uniform power distribution will result in slower thermal feedback (i.e. void feedback) and higher peak power when evaluating licensing events.
Therefore, it is demonstrated that a uniform fuel pellet radial power distribution is conservative.
NRC Request No.
5:
Justify PEco's use of the BSAFE code.
Response
The PECo methodology utilized in the determination of statistical adjustment factors (SAFs) is described in section 4.1.1.4 of the subject report.
This methodology requires a large quantity of l
numerical calculations to be performed.
The BSAFE computer code was designed to perform these calculations utilizing relatively simple and well known mathematical relationships.
BSAFE requires a limited amount of user input and does not require a high degree of user skill (such as for codes like SIMULATE or RETRAN) to execute.
Thus, PEco is justified in its use of BSAFE for application to reload safety evaluation activities.
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Page 4 of 10 NRC Request No.
6:
]
Provide further justification that the reactivity oscillation I
described in the analysis of the feedwater controller failure does not impact licensing results.
1
Response
The reactivity oscillation observed in the analysis of the feedwater controller failure (FWCF) event is the consequence of a RETRAN code limitation that results in a small step change in core reactivity when the core boiling boundary crosses a nodal junction.
The step change in reactivity is on the order 8.0E-5 Ak/k.
Experience has shown that the step change will result in one of two effects.
The core boiling boundary may recross the junction and correct the reactivity bias. This may still result in a small power oscillation, but d 2e to the large thernal time constant of the fuel, the surface heat flux does not change significantly.
Since critical power is driven by heat flux and not neutron power, the prediction of critical power ratio is not significantly affected.
If the boiling boundary does not recross the junction, the original change in reactivity resulus in a change in power (and heat flux) until sufficient reactivity feedback occurs to offset the step change.
At typical reacto; conditions, the power coefficient is approximately -0.03 (Ak/k)/ (AP/P). Thus, to offset a step change of 8.0E-5 Ak/k requires an increase in core power (and heat flux) of about 0.25%.
This small change in reactor power does not significantly affect the prediction of critical power ratio. Furthermore, it should be noted that the small change in reactivity associated with this code limitation is substantially bounded by the nuclear uncertainties utilized in the development of the conservative statistical adjustment factor for the FWCF event.
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NRC Request No.
7:
Provide further justification that the reactor recirculation controller failure event becomes less severe at higher initial core flows.
Response
The reactor recirculation controller failure (RFCF) event is driven by the increasing core flow resulting from the increasing speed of the failed recirculation M-G set. The analysis presented in section 4.4 of the subject report was performed at minimum core flow (i.e.
minimum recirculation M-G set speed). The magnitude of the core flow increase (and thus the magnitude of the reactivity insertion) is limited by the M-G set high speed stops.
To analyze the RFCF l
event at a higher initial core flow requires that the recirculation l
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Page 5 of 10 M-G set speed be at a higher initial speed.
Because the M-G set high speed stop is fixed, this reduces the magnitude of the core flow increase and thus reduces the reactivity insertion and severity of the event.
i NRC Request No.
8:
Assure that this report is complete with respect to chapter 15 of the NRC Standard Review Plan.
Response
The methods and analyses presented in the subject report address each of the six event categories lirted in Section 14.5 of the PBAPS UFSAR.
The limiting events in each category were analyzed by PEco with respect to the MCPR safety limit and the reactor coolant pressure boundary safety limit.
The relative severity of events in each event category was detarmined based on information contained in the PBAPS UFSAR and current industry experience.
PECo has reviewed the events contained in Chapter 15 of the NRC Standard Review Plan (SRP) and has determined that the subject report adequately addresses those events.
Some differences exist in the categorization of transient events between Section 14.5 of the PBAPS UFSAR and the NRC Standard Review l
Plan.
The table below lists the transient events in order as l
presented in the Standard Review Plan and indicates which analysis i
in the report addresses each event.
If an event is not evaluated in the subject report, either 1) the table lists the analyzed event i '
most similar in nature which is bounding, or 2) indicates that tho j
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event is not within the scope of the report.
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Standard Review Plan PEco-FMS-006 l
15.1.1 Decrease in Feedwater 4.2.2 Loss of Feedwater Temperature Heating 15.1.2 Increase in Feedwater 4.2.1 Feedwater (ontroller Flow Failute 15.1.3 Increase in Steam Flow Bounded by 4.2.1 and 4.2.2 15.1.4 Inadvertent Opening of a Not Applicable to BWRs Steam Generator Relief or Safety Valve 15.2.1 Loss of External Load 4.1.1 Generator Load Rejection Without Bypass 15.2.2 Turbine Trip Bounded by 4.1.1
Page 6 of 10 15.2.3 Loss of Condenser Vacuum Bounded by 4.1.1 15.2.4 Closure of Main Steam Bounded by 4.1.1 Isolation Valve (BWR)
(with position switch scram) 15.2.5 Steam Pressure Regulator Bounded by 4.1.1 Failure (Closed) 15.2.6 Loss of Nonemergency AC Bounded by 4.5.1 Power to the Station Auxiliaries 15.2.7 Loss of Normal Feedwater 4.6.1 Loss of Feedwater Flow Flow 15.2.8 Feedwater System Pipe Not Applicable to BWRs Breaks Inside and Outside Containment (PWR) 15.3.1 Loss of Forced Reactor 4.5.1 Two Recirculation M-G coolant Flow (Trip of Set Trip Pump) 15.3.2 Loss of Fo'i ced Reactor Bounded by 4.5.1 Coolant Flow (Flow Controller Malfunctions) 15.3.3 Reactor Coolant Pump Not in the scope of the Rotor Seizure report since this event is now considered an accident by NRC 15.3.4 Reactor Coolant Pump Not in the scope of the Shaft Break report since this event is now considered an accident by NRC 15.4.1 Uncontrolled Control Rod Bounded by 4.3.1 Assembly Withdrawal from a Subcritical or Low Power Startup Condition 15.4.2 Uncontrolled Control Rod 4.3.1 Continuous Rod Assembly Withdrawal at Withdrawal Error During Power Power Operation 15.4.3 control Rod Misoperation Bounded by 4.3.1 (system Malfunction or Operator Error) 15.4.4 Startup of an Inactive Bounded by 4.4.1 Loop or Recirculation Loop at an Incorrect Temperature l
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Page 7 of 10 4
15.4.5 Flow Controller 4.4.1 Recirculation Flow Malfunction causing an controller Failure Increase in BWR Core Flow Rate 15.4.6 Chemical and Volume Not Applicable to BWRs Control System i
Malfunction That Results in a Decrease in the 1
Boron Concentration in i
the Reactor Coolant (PWR) 15.4.7 Inadvertent Loading and 4.3.2 Fuel Loading Error -
I operation of a Fuel Rotated
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Assembly in an Improper 4.3.3 Fuel Loading Error -
Position Mislocated 1
15.4.8 Spectrum of Rod Ejection Not Applicable to BWRs i
Accidents (PWR)
)
15.4.9 Spectrum of Rod Drop Not in scope of the Accidents (BWR) report.
This event is generically evaluated by the fuel vendor.
15.5.1& Inadvertent Operation of Bounded by 4.2.2 15.5.2 ECCS and Chemical and Volume control System Malfunction That Increases Reactor Coolant Inventory 15.6.1 Inadvertent Opening of a Bounded by 4.6.1 PWR Pressurizer Relief VElve or a BWR Relief Valve 15.6.2 Radiological Consequences Not in scope of the t
of the Failure of Small report, addressed in I
Lines Carrying Primary UFSAR Coolant Outside Containment
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15.6.3 Radiological Consequences Not Applicable to BWRs of Steam Generator Tube Failure (PWR) l 15.6.4 Radiological Consequences Not in scope of the I
of Main Steam Line report, addressed in Failure Outside UFSAR Containment (BWR) 15.6.5 Loss-of-Coolant Accidents Not in scope of the Resulting from Spectrum report, addressed in of Postulated Piping UFSAR Breaks Within the Reactor Coolant Pressure Boundary I
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l Page 8 of 10 15.7.1 Waste Gas System Failure Has been deleted from NRC SRP 15.7.2 Radioactive Liquid Waste Has been deleted from System Leak or Failure NRC SRP 15.7.3 Postulated Radioactive Not in scope of the Helease Due to report, addressed in Liquid-Containing Tank UFSAR Failures 15.7.4 Radiological Consequences Not in scope of the of Fuel Handling report, addressed in Accidents UFSAR 15.7.5 Spent Fuel Cask Drop Not in scope of the Accidents report, addressed in UFSAR 15.8 Anticipated Transients Not in scope of the Without Scram report, addressed by vendor Closure of Main Steam Isolation Valve (with high flux scram) is the limiting pressurization event and is addressed in Section 4.7.1 of the subject report (this event is not contained in the NRC SRP).
NRC Request No.
9 l
Clarify the process by which the impact of plant modifications on reload licensing will be evaluated.
Response
The PECo staff responsible for reload licensing reviews all proposed plant modifications for potential impact on reload licensing analyses.
Any identified impact on reload licensing analyses is evaluated by the use of PECo's established 10CFR50.59 review process.
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NRC Request No. 10:
. Address the application of NRC approved safety limits for mixed fuel cores.
Response
PEco will utilize conservative safety limits in accordance with NRC i
approved methods as established in the fuel vendor's safety analysis report for mixed fuel cores.
Page 9 of 10 NRC Request No. 11:
Update thc ntatus of the BWR stability issue.
Response
PEco currently follows the recommendations outlined in GE SIL-380 and has inco.porated those recommendations into plant Technical Specifications and operating procedures.
PEco is an active member of the Boiling Water Reactor Owners Group subcommittee working on the long term resolution to the stability issue.
NRC Request No. 12:
Address the application of cycle operating flexibility options in future reload snalyses.
Response
Cycle operating flexibility options (e.g.
Increased Core flow, Extended Load Line Limit Analysis, Feedwater Temperature Reduction) will be analyzed in future reloads. These options utilize the same analytical methods as described in the subject report and consist of the limiting events evaluated at various off-nominal conditions (e.g.
at higher than rated core flow or at reduced feedwater temperature).
NRC Request No. 13:
1 Section 6 of the PEco Supplemental Reload Licensing Report should be revised to include values for Doppler coef ficient and core average fuel temperature.
Also, section 9 should be revised to include feedwater heaters out-of-service (FWHOOS) and section 10 should be revised to indicate the analysis methods employed (i.e.
PECo methods).
Response
PEco will revise the Supplemental Reload Licensing Report as requested when the approved version of the report is issued.
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e Page 10 of 10 NRC Request No. 14:
Provide a discussion on how the open items identified in the review of PEco-FMS-004 are resolved in PEco-FMS-006.
Response
The NRC Safety Evaluation of PECo-FMS-004 identified six items to be addressed in PEco-FMS-006.
These items are associated with analyses performed utilizing the RETRAN computer code. The following table lists the open items identified in the Safety Evaluation and indicates where they were addressed in PECO-FMS-006.
Onen Item PECo-FMS-006 Resoonse
- 1. Justify the use of the jet pump Section 4.4.1.3 model with flow reversal.
- 2. Justify the use of the one-Section 2.2 and dimensional kinetics option.
Section 3.2
- 3. Justify the feedwater system model Section 4.1.1.3 with ability to simulate rapid feedwater flow excursion such as encountered in the turbine tests.
- 4. Justify the qualification of the Section 4.1.1.4 algebraic slip model for licensing transients.
- 5. Justify the licensing conditions Section 4 and safety margins.
- 6. Provide a statistical analysis to Section 4.1.1. 4 (GLRWOB) determine the uncertainty Section 4.2.1.4 (FWCF) allowance applied to the ACPR calculations.
NRC Request No. 15:
Please revise the listing of topical reports in Section 6,
" References", to indicate their approved status.
Response
PECo will revise Section 6 of the subject report as requested when the approved version of the report is issued.
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