ML17157A577
| ML17157A577 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/15/1991 |
| From: | Thadani M Office of Nuclear Reactor Regulation |
| To: | Keiser H PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| TAC-75999, TAC-76000, NUDOCS 9102280061 | |
| Download: ML17157A577 (15) | |
Text
Docket Nos. 50-387 and 50-388 Mr. Harold W. Keiser February 15, 1991 Senior Vice President - Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101
Dear Mr. Keiser:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1 AND 2 (TAC NOS. 75999/76000)
Our contractor, Brookhaven National Laboratory (BNL), has completed the initial review of two topical reports (PL-NF-005 and PL-NF-90-001) submitted by Pennsylvania Power and Light Company (PP5L) on reload analysis methodology.
As a result of that review, BNL has identified a set of questions and concerns summarized in enclosures 1 and 2.
These enclosures were informally transmitted to your staff on February 7, 1991, with a request that PPSL respond to the questions in a timely manner to support the review of Unit 2 reload.
If an acceptable response to questions in enclosure 1 is received by February 25, 1991, we would be able to complete our safety evaluation and provide you a response prior to Unit 2 restart scheduled in May 1991.
S incere ly,
Enclosures:
1.
Questions for Report PL-NF-89-005 2.
Questions for Report PL-NF-90-001 cc w/enclosures:
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Docket Nos. 50-387 and 50-388 Mr. Harold M. Keiser Senior Yice President - Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101
Dear Mr. Keiser:
SUBJECT:
PEQUEST FOR ADDITIONAL INFORMATION - SUSQUEHANNA ST M ELECTRIC STATION,, UNITS 1
AND 2 (TAC NOS. 75999/76000)
Our contractor, Brookhaven National Laboratory (BNL), h completed the initial review of two typical reports (PL-NF-005 and P -NF-90-001) on reload analysis methodology.
As a result of that review, B
has identified a set of questions and concerns summarized in enclosures 1
a d 2.
We request your response to the questions in a timely manner to s port the review of Unit 2 reload.
If an acceptable response to questions in enc osure 1 is received by February 25, 1991, we would be able to complete our safety evaluation and provide you a response prior to Unit 2 res4krt scheduled in May 1991.
- incerely,
Enclosures:
1.
Questions for Report PL F-89-005 2.
Questions for Report P -NF-90-001 cc w/enclosures:
See next Mohan C. Thadani, Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation DISTRIBUTION NRC
& Local PDRs PDI-2 Reading SVarga EGreenman MO'Brien(2)
- KDesai, SRXB
- LKopp, SRXB
- RBlough, RGN-I
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[TAC NOS 75999/76000]
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Docket thos.
50-387 and 50-388 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 Mr. Harold M. Keiser Senior Vice President - Nuclear Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101
Dear Mr. Keiser:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - SUSQUEHANNA STEAM ELECTRIC STATION, UNITS 1
AND 2 (TAC NOS. 75999/76000)
Our contractor, Brookhaven National Laboratory (BNL), has completed the initial review of two topical reports (PL-NF-005 and PL-NF-90-001) submitted by Pennsylvania Power and Light Company (PP&L) on reload analysis methodology.
As a result of that review, BNL has identified a set of questions and concerns summarized in enclosures 1 and 2.
These enclosures were informally transmitted to your staff on February 7, 1991, with a request that PPSL respond to the questions in a timely manner to support the review of Unit 2 reload.
If an acceptable response to questions in enclosure 1 is received by February 25, 1991, we would be able to complete our safety evaluation and provide you a response prior to Unit 2 restart scheduled in May 1991.
Sincerely,
Enclosures:
1.
Questions for Report PL-NF-89-005 2.
Questions for Report PL-NF-90-001 cc w/enclosures:
See next page ohan C. Thadani, Project Manager Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Mr. Harold M. Keiser Pennsylvania Power 8 Light Company Susquehanna Steam Electric Station Units I 5 2 CC:
Jay, Silberg, Esq.
Shaw, Pittman, Potts 5 Trowbridge 2300 N Street N.M.
Washington, D.C.
20037 Bryan A. Snapp, Esq.
Assistant Corporate Counsel Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. J.
M. Kenny Licensing Group Supervisor Pennsy1vania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Mr. Scott Barber Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.
Box 35 Berwick, Pennsylvania 18603-0035 Mr. Thomas M. Gerus ky, Director Bureau of Radiation Protection Resources Commonwea1th of Pennsylvania P. 0.
Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III A11egheny Elec. Cooperative, Inc.
212 Locust Street P.O.
Box 1266 Harr isburg, Pennsy1 vania 17108-1266 Mr. S.
B. Ungerer Joint Generation Projects Department At1antic E1ectric P.O; Box 1500 1199 Black Horse Pike Pleasantville, New Jersey 08232 Regioqal Administrator, Region I U.S. Nuclear Regulatory Commission 475 A11endale Road King of Prussia, Pennsylvania 19406 Mr. Harold G. Stanley, Superintendent of Plant Susquehanna Steam Electric Station Pennsylvania Power and Light Company 2 North Ninth Street Al1entown, Pennsylvania 18101 Mr. Herbert D. Moodeshick Special Office of the President Pennsylvania Power and Light Company 1009 Fowles Avenue Berwick, Pennsylvania 18603 Mr. Robert G. Byram Vice President-Nuclear Operations Pennsylvania Power and Light Company 2 North Ninth Street Al 1 entown, Pennsy1 vani a 18101
ADDITIONALINFORMATIONREQUIRED FOR THE REVIEW OF THE PENNSYLVANIAPO%ER AND LIGHTCOMPANY BWR TRANSIENTS 0
P RTP Is a SIMTRAN-ESIMULATE-Eone-dimensional cross section set determined for each licensing transient analyzed?
If not, what error is introduced by neglecting the dependence of the RETRAN-02 cross sections on the initial statepoint exposure distribution, rod pattern, power and flow?
What specific changes are made to the control system parameters during initialization?
Are the modified parameters consistent with the plant values and, ifnot, what effect does this difference have on the licensing analyses?
What is the effect of the (Appendix-A) SIMIRAN-ErtSIMULATFE void correction (to the RETRAN-02 cross sections) on the calculated ACPR and vessel pressure increase?
Since the correction was determined for an overpressurization transient, provide justification for the application of this correction to all non-pressurization transients for which it is intended.
Since the correction is not applied to all transients (e.g., the startup tests described in Sections 5. l-5.3), provide the criteria and basis used to determine when the correction is applied.
Were any adjustments made in the REZEMN4Z model or data (excluding those described in Sections 6.2, 6,3 and 7.2) to improve agreement with the Peach Bottom-2 initial conditions (e.g., axial power shape) and test results?
Was the ESCORE gap conductance calculation modified based on the Peach Bottom-2 measurement data? Ifso, what effect did this normalization have on the comparisons and are these adjustment procedures part of the PPL standard calculational method?
How did other variables such as core flowand steam flowcompare for the generator load rejection transient?
How are the case-specific inertia and loss coefficients of Table 3.1-2 determined?
What nodalization of the separator and upper downcomer was used in the RKJ~N-02 model (p-12)?
How will the constant carryunder assumption be validated when the vessel level falls below the separator discharge path exit (p-36)?
Provide the basis for the selected time constants of Figure 3.4-5. What restriction does this selection impose on the application of the RBTRAN42 model?
Discuss the validity of these constants for both rapid and slow transients?
Is the carryover negHgible under transient conditions as well as under the startup conditions (p-46)? Ifnot, how wi11 this be accommodated?
How is the uncertainty analysis of Chapter-8 used to determine the uncertainty in the transient vessel pressure increase?
Was data available for the Peach Bottom-2 tests (e.g., scram rod position versus time) that is not generally available for Susquehanna licensing analyses which allowed a more accurate calculation?
What effect did this have on the calculation-to-measurement comparisons'?
Since the steam lines are not modeled in the Chapter-8 uncertainty analysis, how willthe steam line, initial steam flow and bypass opening characteristics be determined in a conservative manner for licensing transients?
Discuss the adequacy of using the minimum number of samples (three) to determine the population mean (E) and standard deviation (S). What distribution oferrors was assumed in determining the 95/95 upper tolerance factor (7.655) and how was this justified?
How is the statistical uncertainty in the calculational bias of E-9%, which is based on only 3 samples, accounted for? What is the 95/95 upper tolerance limiton E and how is it determined?
In view of the absence of sufficient data to determine the distribution of the hCPR calculational errors, what distribution will be used in the safety limit Monte Carlo calculation and how willthis be justified Are the calculations performed by PPL and provided to ANP for input to safety analyses (e.g., LOCA, SLMCPR, rod drop and fuel handling) consistent with the accuracy and conservatism assumed in the approved ANP methodology?
Does the ANF methodology assume that an allowance for uncertainties is included in the data provided by PPL?
How are the differences between CPM-2/SIMULATE-E and the ANP physics codes accounted for?
~
~
o 18.
What is causing the overprediction of the core pressure and peak transient power of the Peach Bottom-2 TZ1 test?
19.
20.
21.
22.
23.
24, 25.
26.
What is causing the large overprediction of ECPR/ICPR in the TI3 test, Discuss why the bias E determined (to a large extent) by this overprediction is applicable to Susquehanna licensing calculations?
Was the SilVZMN-E/SIMULATE-Evoid cross section adjustment made for the TI'1, TX2, TI3 and LBT calculations?
The GE fuel temperature versus time curve has been omitted from Figurc 7.3.-8.
How does the GE and PPL fuel temperature compare for the licensing basis transient?
How is the effect of the time-dependence of the radial power and low distributions accounted for in the hot-bundle 4CPR calculation?
Are the initialSIMULATE-Epower and flow distributions used throughout the transient?
How is the uncertainty introduced by this simplification accounted for in the RETRAN42 uncertainty analysis?
The hot-bundle analyses (of Section 8.2) used in determining the "measured" and calculated hCPR are identical and, consequently, the measured-maculated hCPR/ICPR comparisons do not include the effect of the uncertainty in the hot-bundle calculation.
How is this additional uncertainty accounted for?
Willthe ESCORB gap conductance used in the Susquehanna safety analyses be consistent with (or conservative relative to) the methods described in Appendix-C (e.g., the use of an average fuel pin gap conductance and a core dominant fuel type)'?
Describe in detail what is causing the large variation in gap conductance between the Peach Bottom-2 tests TI'1, TI2 and TI3 given in Table C-2.
Is a similar variation expected in the Susquehanna licensing analyses?
How is the uncertainty introduced by the use of a single core-average gap conductance, based on a core-average power history and axial power shape, accounted for in the hot-bundle calculation?
27.
Ifthe core-equivalent gap conductance represents an average over the core fuel types, why is the Peach Bottom-2 Cycle-2 equivalent gap conductance ofTable CA larger than the conductances of both the 7x7 and 8x8 fuel types of Table C-3?
28.
What were the results of the pressure regulator failure at natural circulation test referred to in Table C-2?
29.
Is the flowingquality form of the Baroczy correlation less accurate at lower (approaching counter-current) flows? Ifso, how willthis additional uncertainty be accounted for?
30.
ls the PPL application of ESCORE consistent with the limitations of the ESCORE approval?
Has the U-238 resonance escape probability P been validated for the intended Susquehanna fuel designs?
How is the effect of the BSCORS reduced average fuel temperature and stored energy accounted for?
Have the plant-specific uncertainties in the ESCORE calculated gap conductance been deterrriined and accounted for?
31.
Is the PPL application ofREAD,N42 consistent with the limitations of the REI'RAN@2 approval?
Will the PPL application be within the model (boron transport, separator model, etc.) and bubbly flow regime limitations?
Has the conservatism of the heat transfer model for metal walls in non-equilibrium volumes been demonstrated?
Has the PPL Courant time step control been justified?
32.
Describe the calculation of the maximum expected h,CPR channel bowing adjustment of BCPR=0.02.
How has the Susquehanna-2 Cycle-5 loading, fuel and channel design and burnup history been accounted for in this determination?
33.
What are the differences between the channel bowing methods (correlation, statistics, etc.) used for Susquehanna-2 Cycle-5 and,the approved ANF methods?
How are the differences between CPM-2 and the ANP lattice physics code accounted for?
ADDITIONALINFORMATIONREQUIRBD FOR THE REVIEW OP THE PENNSYLVANIAPOWER AND UGHT COMPANY3%R TRANSIENTS TH PI P
There are two significant changes included in the analysis of the rod withdrawal error (RWB) event.
These are:
(1) the statistical treatment of the LPRM failures and (2) the statistical combination of the safety limit uncertainties and the SIMULATE-BRWE calculational uncertainties, In determining the core response to the RWE the hCPR is calculated as a
statistical average over all allowable (withintechnical specifications) LPRM failure states.
Consequently, the calculated average ACPR is conservative for cases oflow failures and non-conservative in the case of high LPRM failures.
Since this analysis is non-conservative for reactor states which cannot be precluded it is considered unacceptable.
The worst case condition of LPRM failures must be assumed in determining the hCPR resulting from a rod withdrawal error.
The basic assumption ofthe statistical combination ofuncertainties (SCU) method is that the POWEIU'LBX'afety limit uncertainties're independent of the RWE SIMULATORS I,CPR calculational uncertainties.
This assumption allows the statistical combination ofthe safety limitand SIMULATE-Ecalculational uncertainties, and results in a nonwonservative reduction in the operating limitMCPR for the RWB.
Since the POWERPLEX monitoring and the SIMULATE-E rod block response calculational models employ much of the same nuclear and thermal-hydraulic modeling data and LPRM input, they cannot be considered to be independent.
It is therefore concluded that the SCU method is not applicable to the rod withdrawal event, and the SIMULATE-E and safety limituncertainties must be applied separately.
How will the removal of the SCU methodology and the statistical treatment of the LPRM failures be accommodated in the rod withdrawal methodology7 The POWERPLBX'afety limit uncertainties are the POWERPLEX'onitoring uncertainties (e.g., on bundle power) that are used in the statistical determination of the CPR safety limit.
J C
2, The analysis of the generator load rejection without bypass (GLRVOB) event employs the SCU methodology, and statistically combines the POWERPLEX'afety limit uncertainties with the REIRAN/SIMULATEE ECPR calcuhtional uncertainties.
The statistical combination results in a significant reduction in the calculated MCPR operating limit. This statistical combination is valid only ifthe POWERPLEX'PR morutoring and the RETRAN/S?MULATE-EhCPR calculational uncertainties are independent.
The POWERPLEX'afety limit and RSTRAN/SIMULATE-E uncertainties are considered to be dependent because of (1) the common nuclear and thermal-hydraulic modeling data used to represent the reactor, (2) the similarityofthe PO%ERPLBX'nd SIMULATE-E calculational methods and (3) the adjustments made to REFEMN/SIMULATE-Esuch as the water density (kinetics parameter) correction.
WMle it is recognized that this interdependence does not result in a perfect correlation of the uncertainties, it is concluded that they cannot be considered to be independent as assumed in the SCU methodology.
Itis noteworthy that in the review of NEDO-24154 (Reference 21 ofPL.NF 001) the staff determined that a 5% probability of exceeding the CPR safety limit is acceptable.
That is, the acceptable calculational uncertainty was determined to be a 95%
probability/95% confidence level value.
The SCV methodology does not provide this assurance.
Based on the above, it is concluded that the SCU method is not acceptable for application to the overpressurization transients.
How willthe removal of the SCV methodology be accommodated in the GLRWOB transient operating limit MCPR calculations 3.
4.
The RERAN conservative bias ofE- -9% for the overyressurization transient is based on only three measurements, Adjust this bias in order to insure that the MCPR safety limitis not exceeded with a 95% probability and 9S% confidence.
%ithout the SCU methodology, the safety limit and event-specific RCPR will be calculated separately in determining the operating limitMCPR.
Modify the technical specifications to include the safety limitMCPR.
Provide the description and validation for the RQDDK-E code used to select the strongest worth control rod positions for the shutdown margin analyses.
Is the cycle-specific highest worth rod used in the rod withdrawal error analysis?
Demonstrate that the rod withdrawal error hCPR calculated for the central region of Figure 2.1-1 is bounding for all coze locations including those with only 2 or 3 LPRM strings.
In the RWE analysis identify the location, relative to the error rod, of the fuel bundle assumed to be on limits.
How do the LPRM and bundle power uncertainty used in the RWE analysis compare to the values approved for the ANF Susquehanna-1 and 2 safety limitcalculations?
Justify the use of smaller values.
Why is the tr,c,R = 0.032 uncertainty for the mislocated fuel bundle combined with the RCPR distribution?
How is the K a~,~ = 0.037 determined2 In the mislocated fuel bundle event, justify the use of the all-rodswut condition for determining ACPR.
In the loss-of-feedwater-heating (LFWH) event, how is the SIMULATE-8uncertainty in the change in bundle power (in addition to the core thermal power) accounted for?
In the SMULATE-ELFWH calculation is the xenon maintained at the initial value) If not, how is the effect on power peaking accounted for2 What is the increase in the LHGR during the LPWH event?
Demonstrate that the SIMULATE-Esteady-state analysis is bounding for the worst case transient feedwater flow, pressure and core power increase.
How willthe shutdown margin prediction uncertainty and design criterion be determined for a specific reload?
What is the basis for the 165 ppm boron margin to account for imperfect mixing of boron in the evaluation of the Standby Liquid Control System (SLCS)?
Demonstrate the conservatism (relative to SIMULATE-E) of the two approximate methods for determining shutdown margin in the SLCS analysis.
Demonstrate that the end-of-cycle reactivity calculations for the ANP LOCAanalyses are bounding for the entire cycle.
20.
21.
22, Is the definition of reactivity (on Page 73) identical to that used by ANP in the LOCA analysis? Ifnot, how willthis difference be accounted for?
How willthe maximum worth rod be determined for the control rod drop analysis?
What is the root-mean-square difference between the POWERPLEX'alculations with PPL and ANF nuclear data?
What increase in the ANF approved safety limitpower distribution uncertainties willthis ANF-to-PPL POWERPLEX'ifference introduce?
23.
How is the difference between the CPM-2 and the ANF lattice calculation of the ANF fuel vault k-infinity( 1.388 criterion accounted for?
24.
How do the static analyses of Chapter 2 and the transient analyses of Chapter 3 differ from the ANF treatment of these events?