ML17157A956
| ML17157A956 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/21/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17157A955 | List: |
| References | |
| NUDOCS 9112060276 | |
| Download: ML17157A956 (30) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO TOPICAL REPORT PL-NF-90-001 "APPLICATION OF REACTOR ANALYSIS METHODS FOR BWR DESIGN AND ANALYSIS" PENNSYLVANIA POWER 8l LIGHT COMPANY SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1
AND 2 DOCKET NOS.
50-387 AND 50-388
- 1. 0 INTRODUCTION By letter from H.
W. Keiser to W.
R. Butler (NRC), dated August 8, 1990, Pennsylvania Power and Light Company (PP8L) submitted topical report PL-NF-90-001, "Application of Reactor Analysis Methods for BWR Design and Analysis," for NRC review.
These methods wi 11 be used to determine the Susquehanna 1 and 2 operating limit minimum critical power ratio, demonstrate compliance with the ASME overpressurization criteria and provide physics input to fuel vendor reload safety analyses.
The NRC staff was supported in this review by our consultant, Brookhaven National Laboratory.
The staff has adopted the findings recommended in our consultant's technical evaluation report (TER) which is attached.
2.0 EVALUATION The attached TER provides the evaluation.
3.0 CONCLUSION
S
'I The staff has reviewed the PP8L application topical report PL-NF-90-001 and the supporting documentation provided in response to our request for additional 9112060276 911121 PDR ADOCK 05000381 PDR
information.
Based on this review, the staff concludes that the proposed statistical combination of uncertainties (SCU) method is not acceptable for the reasons stated in TER Sections
- 3. 1.2 and 3.2. 1.
- Instead, the alternate method proposed in the August 29, 1991, letter from H.
W. Keiser (PP&L) to W.
R. Butler (NRC), PLA-3641, "Susquehanna Steam Electric Station Licensing Methods:
Plan for U1C7," should be used to determine the operating limit minimum critical power ratio for the rod withdrawal error, the generator load rejection without bypass, and the feedwater controller failure events.
In addition, the presently approved POWERPLEX power distribution uncertainties should be retained and should not be reduced.
Attachment:
Technical Evaluation Report Principal Contributor:
J.
Carew Date:
~T7 H
E TE HNI AL EVAL ATI N REP RT Topical Report
Title:
Application of Reactor Analysis Methods for BWR Design and Analysis Topical Report Number:
Report Issue Date:
Originating Organization:
PL-NF-90-001 August 1990 Pennsylvania Power and Light Company l.0 ~OU In Reference-l, the Pennsylvania Power and Light (PP&L) Company has submitted the PP&L BWR reactor design and analysis methods.
These methods are based on the PP&L transient analysis methods described in the PL-NF-89-005 topical report (Reference-2),
and the PP&L steady-state core physics methods given in the PL-NF-87-001 topical report (Reference-3).
While these steady-state physics methods and reactor transient methods provide best-estimate predictions, the proposed reactor applications methodology generally includes conservative adjustments to insure the required margin to fuel thermal and mechanical limits and to system performance criteria.
The proposed reactor analysis methods are intended for application to reload licensing evaluations for the Susquehanna Units 1 and 2.
The PL-NF-90-001 methods willbe used to determine the Susquehanna 1 and 2 operating limit minimum critical power ratio (OLMCPR), demonstrate compliance with the ASME overpressurization criteria and provide physics input to fuel vendor reload safety analyses.
The primary codes used in the PP&L methodology are the CPM-2 lattice physics code (Reference-4),
the SIMULATE-E (Reference-5) three dimensional steady-state core analysis
- code, the
RETRAN-02 (Reference-6) systems transient
- code, and the ESCORE (Reference-7) fuel performance code.
The Siemens Nuclear Power Corporation
-SNPC (formerly Advanced Nuclear Fuels) POWERPLEX code (Reference-8) is used for on-line core monitoring.
For typical Susquehanna 1 and 2 reload cores the potentially limitingevents are identified as the fuel bundle misloading error (FBME), loss of feedwater heating (LFWH), rod withdrawal error (RWE), generator load rejection without bypass (GLRWOB), feedwater controller failure (FWCF), recirculation flowcontroller failure (RFCF) and the main steam isolation valve closure (MSIV) events.
The reload application of the PL-NF-90-001 methods willtypically be limited to the analysis of these transients.
The topical report includes a sensitivity analysis of each of these events to key input parameters, as well as a detailed sample licensing analysis.
The proposed methods include a statistical combination of uncertainty (SCU) approach in which the CPR monitoring uncertainties are statistically combined with the transient b,CPR calculational uncertainties to determine an OLMCPR. This method is applied to the GLRWOB and FWCF transients, and to the analysis of the rod withdrawal event.
This review focused on the degree of conservatism included in the PL-NF-90-001 licensing methodology and the adequacy of the steady-state and transient methods for the specific events being analyzed.
The review included two meetings and extensive discussions with PP&L,
, and a detailed review of the topical report and PP&L response to the request for additional information.
The PP&L licensing methodology is summarized in the following section, the technical evaluation is given in Section-3, and the technical position is given in Section-4.
0 J
i
2.0 MMARY F THE T PI AL REPORT The PP&L reactor analysis methodology includes both steady-state and transient analyses.
The steady-state methods are applied to licensing events in which the final steady-state is limiting or the event is sufficiently slow that quasi-static methods apply. The transient methods are used to calculate RCPR (defined as hCPR/ICPR where ICPR is the initial CPR) for anticipated operational occurrences (AOOs),
and for the MSIV overpressurization
- analysis, The Susquehanna FSAR AOOs are evaluated and the GLRWOB, FWCF and RFCF events are shown to be the limiting transients for determining the OLMCPR.
2.l
~d-S 2.1.1.
Rod Withdrawal Error The rod withdrawal event results from the erroneous selection and withdrawal ofa control rod with a neighboring bundle at the MCPR operating limit. This rod withdrawal introduces a substantial amount of positive reactivity and causes a large increase in the bundle power and a reduction in MCPR. The transient RCPR is determined by the (flow-biased) rod block monitor (RBM) setpoint.
The RWE event is considered a quasi-static event and is calculated with SIMULATE-E.
The 'reduction in MCPR, RCPR, and the associated LPRM responses are calculated with SIMULATE-E(for a particular selected rod) and are input to the PP&L program RBM which calculates the RCPR as a function of RBM setpoint.
The RBM setpoint uncertainty, RBM
response and RCPR calculational uncertainty and LPRM failure probability are combined statistically with RBMSTAT to determine the RCPR distribution resulting from an RWE. Using the statistical combination of uncertainties approach STATOL combines the RWE RCPR distribution with the standard safety limit uncertainties to determine an OLMCPR.
In order to insure that all the important features are correctly modeled in the RWE analysis, PP&L has performed a series of sensitivity calculations.
These analyses indicate that the RBM setpoint for the RWE is relatively sensitive to the control rod pattern, error rod location, and the assumed LPRM failure rate and location.
2.1.2 Fuel Loadin Error Both the misloading of a fuel bundle into an incorrect core location and the rotation of a fuel bundle (by 90 or 180') in its intended location are analyzed as part of the Susquehanna Units 1 and 2 reload evaluation.
The fuel mislocation analysis is performed with SIMULATE-E and the largest'RCPR is determined considering all potentially limiting mislocations and all cycle exposure points.
An uncertainty allowance for the SIMULATE-Ecalculated RCPR is included.
The Susquehanna Units are C-lattice plants (i.e., with equal water gaps) and, consequently, the increase in RCPR resulting from a rotated bundle is relatively small. The RCPR resulting from a rotated fuel bundle is determined with SIMULATE-E.
A worst-case analysis has been performed which is expected to bound future reloads, and the applicability of the bounding analysis to a specific reload bundle design will be determined by comparing CPM-2 calculated peaking factors and S-factors.
2.1.3 s
f Feedwater Heatin The PP&L methodology treats the loss of feedwater heating event as quasi-static and calculates the transient RCPR using SIMULATE-E. A bounding feedwater temperature decrease of 100'F together with a 5 psi pressure increase is assumed in the LFWH analysis.
Sensitivity calculations for changes in pressure, rod pattern, cycle exposure and assembly reactivity were performed and used to determine a generic 95/95 transient RCPR correlation. This correlation will be verified and applied in reload licensing evaluations of the LFWH event.
2.1.4 Core Ph sics Parameter Core physics parameters are required for the reload evaluation to demonstrate compliance with the technical specifications, provide nuclear cross-section input to RETRAN transient
- analyses, and to provide input to accident analyses which SNPC will perform (LOCA, safety limit MCPR, control rod drop, and fuel storage criticality analyses).
These steady-state parameters are calculated with CPM-2 and SIMULATE-E. CPM-2 is used to calculate the pin-wise local peaking factors for the SNPC LOCA and SLMCPR analyses, and the data required for the POWERPLEX core monitoring system.
SIMULATE-Eis used to calculate (1) the core reactivity for shutdown and standby liquid control system analyses, (2) the scram and dropped rod reactivi ties and (3) the feedback coefficients for LOCA analyses.
SIMTRAN-E (Reference-9) uses the SIMULATE-Eradial flux solution to collapse the cross-sections to one-dimension for input to RETRAN-02.
PP&L has performed sensitivity calculations for the SIMULATE-Eanalyses, and either determines conservative physics parameters or includes an explicit 95/95 uncertainty allowance.
\\
2.2 Transient Anal es 2.2.1 enera r
d Reecti n With ut B as The generator load rejection without bypass is calculated with the Reference-2 RETRAN-02 core and hot-bundle models.
The calculation is performed for the case in which the transient is initiated by a fast turbine control valve closure, and for the case in which both the turbine control valves (TCV) and turbine stop valves (TSV) close.
Conservative assumptions based on PP&L GLRWOB sensitivity studies are employed.
These include neglect of the end-of-cycle (EOC) recirculation pump trip (EOC-RPT), technical specification mode operation of the safety relief valves and EOC all-rods-out conditions.
When the SCU method is not used in the GLRWOB analysis, the initial core power is conservatively increased to 104.4% (of rated) and the Reference-2 values of the 95/95 upper tolerance limits on RCPR are used to account for calculational uncertainties.
When the SCU method is used, the effect of a 2% standard deviation in core power is combined with the effect of a 0.2 ft/sec (plant-specifi) standard deviation in scram time using a two-dimensional RCPR response surface.
The resulting RCPR variation is then combined with the Reference-2 code uncertainty and POWERPLEX MCPR monitoring uncertainties using a Monte Carlo approach.
This statistical combination of uncertainties method yields a OLMCPR of 1.30 for the GLRWOB for Susquehanna Unit-2, Cycle-2 (U2C2).
2.2.2 Feedwater ontroller F ilure The FWCF is analyzed at EOC with all-rods-out and with a maximum allowed flow rate
failure. The event is analyzed with either the turbine bypassor EOC-RPT inoperable, and may be analyzed at earlier cycle exposures if exposure-dependent OLMCPR limits are required.
PP&L has performed a series of sensitivity calculations to quantify the effect of the important transient parameters.
Based on these studies a conservative methodology has been determined and assumes:
(1) the technical specification minimum scram insertion rate, (2) 100% (of rated) core flow and (3) 85% of the best-estimate TCV closure time.
The steam line uncertainties are combined with the RCPR calculational uncertainty of Reference-2 and a 95/95 upper tolerance limit on the transient RCPR is determined.
The SCU method may also be applied to the FWCF event and, in this case, a RCPR response surface will, be constructed and used to determine the OLMCPR in a manner similar to that used for the GLRWOB transient.
2.2.3 Recirculation Flow ontroller Failure The recirculation flow controller failure event results in increased core flow and is a potentially limiting MCPR event.
PP&L has evaluated both the master controller and single loop controller failure events, and has determined the master controller failure to be limiting.
The licensing calculations are performed on the 100% rod line since these have a larger RCPR and bound the lower powered (higher MCPR) statepoints.
An event-specific RCPR uncertainty has been determined,'based on the doppler and void coefficient uncertainties, and is applied to the calculated RFCF RCPR.
A limiting flow run-up rate will be calculated for the event and used in the Susquehanna reload licensing analyses.
2.2.4 v
re urization Anal si PP&L has evaluated both the GLRWOB and MSIV closure transients, and has determined that the MSIV closure is the limiting overpressurization event.
The MSIV transient is analyzed with the RETRAN-02 model of Reference-2 with improved MSIV and safety/relief valve models.
The analysis assumes that (I) the relief mode actuation of the SRVsis inoperable, (2) the six inoperable SRVs have the lowest pressure setpoints and (3) the SRVs have maximum opening times.
Sensitivity calculations were performed and indicate that the most important parameters are the initial core power, control rod insertion rate and the MSIV closure time.
In the PP&L analysis the core power is taken to be 104.4% (of rated) and the MSIV closure and control rod insertion time are based on the plant technical specifications.
I 3 0 TE HNI AL EVAL ATI N The reactor analysis methods Topical Report PL-NF-90-001 describes the methods that willbe employed in the PP&L reload licensing evaluations for Susquehanna Units 1 and 2. The initial review of this report resulted in a request for additional information (RAI) which was transmitted to PP&L in References 10-11.
This review included an evaluation of the proposed licensing analysis methods described in the topical report, as well as the PP&L responses to the RAI included in References 12-14.
The major issues and concerns raised during the review are summarized in the following sections.
3 ~ 1.1 Rod Withdrawal Error In the analysis of the RWE event, the maximum RCPR results when the selected rod yields the worst combination of control rod worth and RBM response.
Consequently, the limiting control rod location depends on the cycle-specific core loading.
For a given core reload, PP&L calculates the RWE transient RCPR for all control rods within a Sx5 control-cell region in the center of the core.
PP&L has evaluated the control rod locations outside this central region, including control rods having only two or three LPRM strings available, and has'etermined that these locations are not limiting. In Response-6 (Reference-13) PP&L indicates that in this procedure the worst-case combination of rod worth and RBM response yielding the
.limiting transient RCPR is determined.
3.1.2 A
lication f the SCU Method to he RWE Even In the application of the SCU method to the RWE, the POWERPLEX safety limit monitoring uncertainties'nd the rod block monitor (RBM) response uncertainties are considered to be independent and are combined statistically to determine the MCPR operating limit. Since the POWERPLEX and RBM systems use the same LPRM input and make use of similar neutronics solutions, the uncertainties associated with these systems are not believed to be independent and, therefore, should not be combined using the proposed SCU method.
In
'he POWERPLEX safety limit uncertainties are the POWERPLEX monitoring uncertainties (e.g.,
on bundle power) that are used in the statistical determination of the CPR safety limit.
addition, in the SCU approach the hCPR resulting from an RWE is calculated as a statistical average over all allowable (within technical specifications) LPRM failure states.
Consequently, the calculated average 2CPR is conservative for reactor states with a small number of LPRM failures and non-conservative for reactor states with many LPRM failures. This approach, therefore, does not provide protection for reactor states (with many LPRM failures) that are expected during normal operation and is considered unacceptable.
In Reference-14, PP&L has provided an alternate method for determining the transient ACPR for the RWE event.
This method assumes the worst-case combination of LPRM detector failures together with the worst-case RBM channel failure, and is consistent with NRC approved methods.
The proposed PP&L alternate method of Reference-14 for determining the d,CPR resulting from an RWE event is therefore acceptable.
- 3. i.3 ~ICh The RCPR resulting from a fuel bundle mislocation depends on the specific core location and the insertion of the neighboring control rods.
In the calculation of the limiting RCPR, PP&L assumes that all control rods are withdrawn.
In order to evaluate this approximation PP&L has calculated ninety-three combinations of misloading location and control rod pattern and finds only a slight (0.0017) underprediction of the limiting RCPR (Response-11, Reference-13).
This underprediction is considered to be negligible. In order to account for the r
variation due to fuel bundle location and control rod pattern, a 95/95 upper tolerance factor is applied to the RCPR calculation for the fuel bundle mislocation.
10
3.1.4 Lo s f Feedwater He tin The transient RCPR resulting from a loss of feedwater heating event depends on the time-dependence of the core power, pressure and flow. PP&L has measured these variables during a loss of feedwater heating transient and has found that the core pressure and floware essentially unchanged during the transient while the core power increased monotonically.
This confirms the PP&L assumption that the final statepoint is RCPR limiting during the LFWH event.
The local power and linear heat generation rate (LHGR) increase during the LFWH transient due to the increased core power and axial (bottom) peaking.
This increase is 20%
and is bounded by the generic LHGR transients.
3.1.5 ore Ph sics Parameters In the evaluation of the shutdown capability of the standby liquid control system, one of two methods are used to determine the boron reactivity worth.
The first approximate method includes a substantial margin of conservatism and typically overpredicts the boron-worth by a factor of two.
In the second method the cross-sections are adjusted to include the dependence on the soluble boron.
In certain
- cases, the conservatism of this cross-section adjustment procedure is small (-5%).
In order to provide additional conservatism in this calculation, PP&L has indicated (Response-18, Reference-13) that an additional 0.01 hk uncertainty allowance will be included when this method is used to determine boron reactivity worth.
The core shutdown margin is determined assuming the highest worth rod does not insert.
The control rod worth depends on the specific core loading and the location of the stuck rod, and, to reduce the number of calculations required to identify the strongest rod, an approximate
J I
g
calculation is performed with RODDK-E (Reference-15),
The uncertainty introduced by this approximation was determined by comparing RODDK-E to SIMULATE-E.
Seventy comparisons were made, including variations in core loading, cycle exposure and control and void histories, and indicated a rod worth discrepancy of less than 0.002 b,k (Response-5, Reference-13).
The SNPC LOCA analyses require reload-specific limitingcore void and doppler feedback reactivities.
In Response-19 (Reference-13)
PP&L has indicated that the cycle-dependent variations in the feedback reactivities are small compared to the EOC reduction in scram reactivity, and the EOC statepoint is limiting. RETRAN-02 transient LOCA calculations at BOC,
~
MOC and EOC were performed and 'demonstrated that EOC is limiting, PP&L intends to provide the core physics input for the SNPC reload LOCA, SLMCPR, control rod drop, and fuel storage criticality analyses.
SNPC has indicated (Response-7, Reference-13) that the input data provided is determined in a manner consistent with the approved SNPC methods and uncertainty treatment.
PP&L intends to use CPM-2 rather than the SNPC XFYRE lattice physics code to determine the neutronics input data for the POWERPLEX core monitoring system.
In order to determine the effect of this neutronics data change on the POWERPLEX power distribution uncertainties, PP&L has compared POWERPLEX predicted and measured TIP responses using both the CPM-2 and XFYRE data.
These comparisons indicate that POWERPLEX/
CPM-2 provides comparable or improved agreement with the TIP measurements, relative to POWERPLEX/XFYRE. PP&L proposed in Response-22 of Reference-13 to use reduced POWERPLEX/CPM-2 power distribution uncertainties, inferred from these comparisons, in the 12
0 7
N
SLMCPR analyses.
However, in the thermal margin licensing basis, the adequacy of all the SLMCPR uncertainties (including bundle power, feedwater flow and temperature, core flow, etc.) has been demonstrated as a group, and any reduction in an individual uncertainty will invalidate the set of uncertainties.
It is, therefore, concluded that the SLMCPR POWERPLEX power distribution uncertainties should not be reduced (as proposed in Response-22),
but should remain at their presently approved values of References 16 and 17 as originally proposed in Section-2.9.2 of the topical report.
Based on the above and the information provided in References 12-14, it is concluded that the steady-state analyses are acceptable with the limitations indicated in Section-3.1.1 concerning the SCU method, and in Section-3.1.5 concerning the POWERPLEX SLMCPR uncertainties.
3.2 Transient Anal se 3.2.1. A lication of the Method to the LRW B and FW F Even s The PL-NF-90-001 method for the analysis of the generator load rejection without bypass and feedwater controller failure events employs the SCU statistical combination of uncertainties method for including the calculational uncertainties in the transient DCPR. In this approach the transient hCPR calculational uncertainties are statistically combined with the POWERPLEX safety limit measurement uncertainties.
This SCU approach differs from presently approved methods and results in a substantial nonconservative reduction in CPR margin relative to the approved methods where the uncertainties are added separately.
The application of the SCU 13
r J
method to the GLRWOB and FWCF transients requires the mean, standard deviation and the distribution of the calculation uncertainty in the hCPR that occurs during the transient.
The mean and standard deviation are determined with three calculation-to-measurement data points derived from the Peach Bottom-2 turbine trip tests.
It is noteworthy that in typical statistical analyses the mean and standard deviation are determined from a relatively large data set.
The minimum number of data points required to determine a mean and standard deviation is three (3) and, therefore, in the present application a data base with the minimum number of data points is used.
Since the available data is not sufficient to determine the b,CPR uncertainty distribution, the b,CPR are taken to be distributed normally about the mean value d,CPR.
This assumption is important since the MCPR operating limit is sensitive to the details of the statistical representation used for the uncertainty in hCPR.
It is concluded that the three Peach Bottom-2 data points and the normality assumption, when used to combine the safety limit monitoring and transient hCPR calculational uncertainties using the proposed SCU method, do not provide the high confidence required to protect the specified acceptable fuel design limits.
In the PP&L application of the SCU method, the value of the safety limit MCPR is not calculated as part of the determination of the operating limitMCPR. The value of the operating limit MCPR is determined by the condition that 99.9% of the fuel rods are not expected to experience boiling transition.
PPE;L intends to use this condition as the safety limit in the Susquehanna Units 1 and 2 technical specifications.
This definition of the safety limit, as a condition, does not conform to the technical specification requirements of 10 CFR 50.36 (Reference-18) which states that the safety limits be "limits upon important process variables."
14
0 i
I a
~ ~
al
In Reference-14, PP&L has provided an alternate method for including the uncertainties in the calculated transient d,CPR for the GLRWOB and the FWCF transients and determining the OLMCPR.
The proposed method is the'same as the Siemens Nuclear Power Corporation approach which has been approved by the NRC.
In this method the b,CPR uncertainties are accommodated by increasing the transient integral power calculated by RETRAN-02 by 10%.
The transient hCPR resulting from this calculation is added algebraically to the SLMCPR to determine the OLMCPR. The SLMCPR is determined using the approved SNPC uncertainties and statistical methods.
This method provides adequate allowance for the transient ACPR calculation uncertainties and is based on NRC approved methods.
The proposed alternate method is therefore acceptable for calculating the OLMCPR for the GLRWOB and FWCF events.
In Reference-14 PP&L has indicated that certain cycle-specific evaluations will be performed.
The initial power used in the GLRWOB will be determined for each reload cycle via a parametric evaluation. The GLRWOB is identified as the limitingd,CPR transient resulting from core pressurization, however, PP&L has indicated (Response-l, Reference-14) that it will also evaluate the turbine trip without bypass for each cycle to determine the maximum transient d CPR.
3.2.2 Recirculation Flow Controller Failiire The recirculation flow controller failure is identified as one of the three limiting transient hCPR events.
The event involves a two pump runup which results in an increase in core power and a reduction in MCPR.
When the transient is initiated from a high power rod-line, the 15
maximum combined (steam) flow limit (MCFL) may be reached (depending on the MCFL setting) resulting in a loss of pressure control. However, the resulting pressure increase is slow and the moderator density changes are relatively small.
While the cross-section moderator density correction is not included in the calculation of this transient, it is indicated in Response-5 (Reference-14) that the neglect. of this correction results in a conservative overprediction of d,CPR.
The RFCF event is relatively slow ((100 seconds) and the DCPR uncertainties of Reference-2 (for overpressurization transients) are not applicable.
PP&L has used Doppler and void reactivity uncertainties of 20% and 50%, respectively, and a flow runup rate which gives, the maximum increase in hCPR.
In addition, in Response-9 (Reference-14)
PP&L has indicated that a conservative gap conductance corresponding to a value of 125% of rated power will be used.
3.2.3 ve res urization Anal sis PP&L has performed extensive sensitivity calculations for the MSIV closure overpressurization transient which indicate that the most important parameters are the core
Conservative values of these input are assumed; (1) the core power is taken to be 104.4% (of rated), (2) the technical specification (maximum) scram time is assumed and (3) the technical specification minimum MSIV closure time is assumed.
PP&L uses a flat axial power shape to determine the core average gap conductance, since this results in a minimum gap conductance and maximum transient pressure.
The moderator density correction is not included in the neutronic cross-sections used in the 16
MSIV analysis, since this also results in a conservative overprediction of the transient pressure (Response-7, Reference-14).
PP&L intends to perform a MSIV ovepressurization analysis for each reload cycle.
Based on the above and the information provided in References 12-14, it is concluded that the transient methods are acceptable with the limitation indicated in Section-3.2.1 concerning the SCU methods.
17
4.0 TE HNI AL P ITI N The PP&L methods described in PL-NF-90-001 and the additional information provided in References 12-14 has been reviewed in detail.
The proposed methods are intended for the analysis of the limitinghCPR quasi-static and transient analyses, the overpressurization
Based on this review it is concluded that the PP&L methods are acceptable for performing reload licensing analyses for Susquehanna Units 1 and 2, subject to the conditions given in Section-3 of this evaluation and summarized in the following.
(a)
The proposed statistical combination of uncertainties SCU method is not approved (Sections-3.1.2 and 3.2.1). Instead, the alternate method proposed in Reference-14 should be used to determine the OLMCPR for the rod withdrawal error, the generator load rejection without bypass, and the feedwater controller failure events (Sections-3.1.1 and 3.2.1).
b)
The presently approved POWERPLEX power distribution uncertainties given in References 16 and 17 should be used in the SNPC SLMCPR analyses (Section-3.1.5).
18
C,i References 1.
"Submittal of Topical Report PL-NF-90-001", Letter, Keiser, H.W. (PP&L) to Butler, W.R. (NRC), dated August 8, 1990.
2.
"Qualification of Transient Analysis Methods for BWR Design and Analysis",
PL-NF-89-005, Pennsylvania Power & Light, December 1989.
I 3.
"Qualification of Steady-State Core Physics Methods for BWR Design and Analysis",
PL-NF-87-001-A, Pennsylvania Power & Light, July 1988.
4.
D.B. Jones, "CPM-2 Computer Code User's Manual",
Part II, Chapter 6 of EPRI-NP-4574-CCM, February 1987.
5.
D.M. VerPlanck, "SIMULATE-E: A Nodal Core Analysis Program for Light Water Reactors",
EPRI NP-2792-CCM, March 1983..
1 6.
"RETRAN 02 - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems", EPRI-NP-1850-CCM-A, Computer Code Manual, June 1987.
7.
"ESCORE
- The EPRI Steady-State Core Reload Evaluator Code:
General Description", EPRI-NP-5100, February 1987.
8.
S.W. Jones, et. al., "POWERPLEX Core Monitoring System Software Specification for the Susquehanna Steam Electric Station Units 1 and 2", XN-NF-83-35(P), Revision 1,
August 1986.
9.
"SIMTRAN-E: A SIMULATE-E TO RETRAN-02 Data link", EPRI NP-5509-CCM, December 1987.
10.
"Request for Additional Information - Susquehanna Steam Electric Station Units 1 and 2,"
Letter, Thadani, M.C. (NRC) to Keiser, H.W. (PP&L), dated February 15, 1991.
11.
"Request for Additional Information on PP&L Topical Report PL-NF-90-001," Letter, Raleigh, J.J.(NRC) to Keiser, H.W. (PP&L), dated August 1, 1991.
12, "Initial Response to RAI on PL-NF-90-001 (SCU Questions)",
Letter, Keiser, H.W.
(PP&L) to Butler, W.R. (NRC), PLA-3566, dated April 23, 1991.
13.
"Final Response to RAI on PL-NF-90-001", Letter, Keiser, H.W. (PP&L) to Butler, W.R. (NRC), PLA-3578, dated June 4, 1991.
19
References Cont'd 14.
"Susquehanna Steam Electric Station Licensing Methods: Plan for U1C7", Letter, Keiser, H.W. (PP&,L) to Butler, W.R. (NRC), PLA-3641, dated August 29, 1991.
15.
"ARMP-02 Documentation: Part II, Chapter 8 SIMULATE-E(Mod. 3) Computer Code Manual", EPRI NP-4574-CCM, Part II, Chapter 8, September 1987.
16.
"Exxon Nuclear Critical Power Methodology for Boiling Water Reactors",
XN-NF-524-P-A, Revision 1, November 1983.
17.
"Exxon Nuclear Methodology for Boiling Water Reactor, THERMEX: Thermal Limits Methodology Summary Description", XN-NF-80-19-P-A, Volume 3, Revision 2, January 1987.
18.
"Code of Federal Regulations", Office of the Federal Register, Part-10, January 1991.
20
Mr. Harold W. Keiser Pennsylvania Power 5 Light Company Susquehanna Steam Electric Station Units 1 5 2
CC:
Jay Silberg, Esq.
Shaw, Pittman, Potts 5 Trowbridge 2300 N Street N.W.
Washington, D.C.
20037 Bryan A. Snapp, Esq.
Assistant Corporate Counsel Pennsylvani'a Power
& Light Company 2 North Ninth Street Al1entown, Pennsylvania 18101 Mr. J.
M. Kenny Licensing Group Supervisor Pennsylvania Power 5 Light Company 2 North Ninth Street Al 1 entown, Pennsy1 vani a 18101 Mr. Scott Barber Senior Resident Inspector U. S. Nuclear Regulatory Commission P.O.
Box 35 Berwick, Pennsylvania 18603-0035 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Resources Commonwealth of Pennsylvania P. 0.
Box 2063 Harrisburg, Pennsylvania 17120 Mr. Jesse C. Tilton, III Allegheny Elec. Cooperative, Inc.
212 Locust Street P.O.
Box 1266 Harrisburg, Pennsylvania 17108-1266 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale 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 Allentown, Pennsylvania 18101 Mr. Herbert D. Woodeshick Special Office of the President Pennsylvania Power and Light Company 1009 Fowles Avenue Berwi ck, Pennsylvania 18603 Mr. Robert G.
Byram Vice President-Nuclear Operations Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101