ML20059H414
| ML20059H414 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 08/31/1990 |
| From: | Silberg J CLEVELAND ELECTRIC ILLUMINATING CO., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#390-10794 90-605-02-OLA, 90-605-2-OLA, OLA-2, NUDOCS 9009170155 | |
| Download: ML20059H414 (20) | |
Text
gp _, l D D CORRESPONDENCE-DOCKETED August 31#5b530 UNITED STATES OF AMERICA '90 SEP -4 A11 :21 NUCLEAR REGULATORY COMMISSION Of rlCf 0F SECRETARY Before the Atomic Safety and Licensina BoardDUCKrit% A MicVICf. i BRANCH In the Matter of ) 1
)
THE CLEVELAND ELECTRIC )
ILLUMINATING COMPANY, et al. ) Docket No. 50-440-OLA-'2 '
) ASLBP No.- 90-605-02-OLA-(Perry Nuclear-Power Plant, )
Unit No. 1) .)
)
1 LICENSEES' RESPONSE TO OHIO CITIZENS FOR .
RESPONSIBLE ENERGY, INC.'S INTERROGATORIES- 1 AND REQUEST FOR PRODUCTION OF DOCUMENTS Pursuant to-the provisions of 10 C.F.R. Sections 2.740, !
2.740b and 2.741, The Cleveland Electric Illuminating-Company, et al. (" Licensees") hereby answer Ohio Citizens for Responsible Energy, Inc.'s ("OCRE") Interrogatories and Request for Produc- i tion of Documents. - Licensees-hereby expressly reserve the right ,
to add to or amend their response to each-and every interrogatory contained herein, j INTERROGATORY NO. 1: Identify each person _ Licensees intend to call as a witness in this proceeding, and identify the subject i matter on which he/she is expected to testify, and the substance- 1 of the testimony.
ANSWER TO INTERROGATORY NO. 1: At this. time, Licensees have not yet determined the identity of each.and every: person whom i they intend to' call as a witness in this proceeding, nor do they know the subject matter or substance of any testimony which may be given by each and every'such person.
3 INTERROGATORY NO. 2: Identify each-person having knowledge <
of the subject matter of this proceeding.
ANSWER TO INTERROGATORY No. 2: The following groups of peo- ,
ple employed N' GE have knowledge of the subject matter of this proceeding: (i) Core Nuclear Eagineering, (ii) Fuel Design and i Development, (iii) BWR Technology, (iv) Fuel Rod Thermal and Mechanical Analysis, (v) Core Nuclear Design, (vi) Regulatory and Analysis Services and (vii) Doraestic Fuel Projects.
9009170155 PDR 900B31 9 ADOCK 05000?40 1806 c.
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c People employed by Licensees in the following sections have knowledge of the subject matter of this proceeding:
(i) Licensing and Compliance Section, which is part of the Perry Nuclear Support Department and (ii) Mechanical Design, Electrical Design, and Performance Engineering Sections, which are part of the Perry Nuclear Engineering Department.
INTERROGATORY NO. 3: Identify all documents relevant to the ,
subject matter of this proceeding, j 1
ANSWER TO INTERROGATORY NO. 3: Licensees object to OCRE's third interrogatory as being unduly burdensome; however, Licens-ees have identified the following groups of documents that are relevant to the subject matter of this proceeding: (i) Transient-Protection Parameters Verification for Reload Licensing Analyses
("OPL-3") form, (ii) the Fuel Release and Engineering Data
(" FRED") form and (iii) guidelines prepared by GE for generating i OPL-3 inputs.
INTERROGATORY NO. 4: For each of the following parameters, Linear Heat Generation Rate ("LHGR"), Minimum Critical Power ;
Ratio ("MCPR"), Power and Flow Dependent MCPR, Maximum Average Planar Linear Heat Generation Rate ("MAPLHGR"), Power and Flow '
Dependent MAPLHGR Curves, identify and describe the analytical methodology or methodologies used to calculate the values of the parameter. State whether the methodology uses computer models or hand calculations.
f ANSWER TO INTERROGATORY NO. 4: Reload license analyses are ;
t performed for each reload to satisfy the requirements for a !
safety evaluation of plant modifications. The results of the reload license analyses performed by GE Nuclear Energy for GE j supplied reload fuel assemblies are documented in a plant and i cycle-specific reload license analysis submittal. As a part of the reload licensing analysis process, core operating limits are established to demonstrate that the reload license analysis ;
results do not exceed the specific event analysis limits. These ,
core operating limits are developed consistent with the require-ments of the Nuclear Regulatory Commission's ("NRC"), " Removal of Cycle-Specific Parameter Limits from Technical Specifications (Generic Letter 88-16)," using NRC approved methodology.
The following is a description of the process used to estab-lish the minimum critical power ratio ("MCPR") and maximum aver-age planar linear heat generation rate ("MAPLHGR") operating lim-its in GE Nuclear Energy's reload license analysis process. ;
There are basically four in+errelated analysis methodologies used in the reload license analysis orocess to establish the specific values for the MCPR and MAPLHGR operating limits: (1) the safety L
l limit methodology; (2) the fuel thermal mechanical methodology; (3) the emergency core cooling system ("ECCS") evaluation method-ology; and (4) the transient analysis methodology. Each of these '
methodologies is comprised of one or more sets of methods and corresponding analysis inputs. The methodologies may be used to establish the value for an event analysis limit or to provide the results of a specific event analysis. All of the analysis
! results are generated using computer programs which have encoded the methodologies reviewed and approved by the NRC. The method-ologies reviewed and approved by the NRC are described in the following GE topical reports: (i) NEDE-240ll-P-A-9 and NEDE-24011-A-9, " General Electric Standard Application for Reac-tor Fuel," September-1988, (ii) NEDE-24011-P-A-9-US and l
NEDE-240ll-A-9-US, " General Electric Standard Application for Reactor Fuel (Supplement for the United States)," September 1988, (iii) N2DE-30130-P-A and NEDE-30130-A, " Steady-State Nuclear Methods," April 1985 and (iv) NEDE-20566-P-A and NEDE-20566-A,
" Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K," September 1986.
The safety limit methodology is used to establish the fuel cladding integrity ("MCPR") safety limit, which is used as one of the figures of merit in the transient analysis methodology for j
the evaluation of anticipated operational occurrences. In the analysis of anticipated operational occurrences, the calculated event analysis results are not allowed to violate the MCPR safety limit. The MCPR safety limit is established using the GE thermal analysis basis methodology. The thermal analysis basis methodol-ogy includes the use of a bounding core operating state and con-sideration of the core operating state uncertainties, fuel manu-facturing uncertainties, and critical power correlation uncer-tainties. The thermal analysis basis methodology, including the process for developing inputs to the methodology and treatment of the uncertainties in the methodology, has been reviewed and approved by the NRC. No changes to the NRC approved documented methodology can be made without prior NRC approval.
l The fuel thermal mechanical methodology is used to establish j the fuel rod thermal mechanical performance limits, which are the other figures of merit in the transient analysis methodology for i the evaluation of anticipated operational occurrences. These l
limits are developed consistent with the linear heat generation I rate ("LHGR") limit. In the analysis of anticipated operational l
occurrences, the calculated event analysis results are not allowed to exceed the thermal-mechanical performance limits, which are based on safety analysis fuel design limits as speci-fied by the NRC. These fuel rod thermal-mechanical performance
! limits are developed using the fuel rod thermal-mechanical meth-odology and are based on the fuel physical parameters and the plant operational requirements, including the uncertainties in these parameters. The fuel thermal-mechanical methodology, l
1
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e including the process for developing inputs to and the process for treating uncertainties in the methodology, has been reviewed and approved by the NRC. No changes can be.made to the NRC i approved documented methodology without prior NRC approval.
l The ECCS evaluation models are used in the analysis of the l loss of coolant accident as part of the ECCS evaluation methodol-ogy to demonstrate that the limits contained in 10 C.F.R. Section 50.46 are not exceeded. The ECCS evaluation is based on the plant operating state, the plant configuration, the ECCS perfor-mance parameters, single failure analysis considerations, and an initial or assumed MAPLHGR developed based on the LHGR limit.
Also assumed is an initial critical power ratic, which is used to determine the minimum allowable value for the operating limit MCPR, as described in the transient annlysis process below. '.
Using the ECCS evaluation methodology, toe peak. cladding tempera-ture and clad oxidation are calculated and compared to the limits of 10 C.F.R. Section 50.46. If the calculated results satisfy the event analysis limits, the initial MAPLHGR becomes the allow-able value for the MAPLHGR operating limit based on ECCS perfor-mance considerations. If the calculated results do not satisfy the limits of 10 C.F.R. Section 50.46, the assumed initial value-of MAPLHGR is reduced until the analysis results satisfy the event limits, and this MAPLHGR becomes the allowable value for the MAPLHGR operating limit based on ECCS performance consider-ations. The most limiting of the ECCS based MAPLHGR operating limit and the MAPLHGR developed using the thermal-mechanical methodology and demonstrated acceptable in the transient analysis process then becomes the operating limit MAPLHGR. The uncertain-ties in the ECCS evaluation models and their inputs are treated consistently with the. requirements of 10 C.F.R. Section 50.46.
The ECCS evaluation models, the process for developing inputs, and the treatment of uncertainties has been reviewed and approved i by the NRC. No changes can be made to the NRC approved-docu-mented methodology without prior NRC approval.
The transient analysis methodology is used in the analysis of anticipated operational occurrences and is an inherent part of the process used to confirm the MAPLHGR operating limit and establish the operating limit MCPR and the MCPR and MAPLHGR flow and power dependent limits. The potentially limiting anticipated operational occurrences that are evaluated to establish these limits include the most limiting rapid pressurization event (tur-l bine trip without bypass or the generator load rejection without bypass), the loss of feedwater heating, feedwater controller failure, recirculation flow increase _and control rod withdrawal error. The transient analysis methodology includes: (1) the lattice physics methods; (2) the three-dimensional BWR simulator; (3) the transient analysis model; (4) the steady state thermal hydraulics; and (5) the transient critical power calculational methodology. The transient analysis methodology, including the l
e process for developing inputs to the various models and the model and model input uncertainties, has been reviewed and approved by the NRC. No changes can be made to the NRC approvc4 documented methodology without prior NRC approval.
- The transient analysis process begins with the use of the lattice physics methods to develop the two-dimensional nuclear libraries which are required by the three-dimensional BWR simula-tor. To perform the required analyses, the lattice physics meth-ods require a fuel assembly description and data from the crose -
sect'on library. The lattice physics methods also provide the local peaking patterns used in the critical power correlation.
The three-dimensional BWR simulator is used to define the ,
core state and nuclear parameters used as input to the transient analysis model. In addition to the data from the lettice physics methods, the t.hree-dimensional simulator requires the reference core loading pattern, core operating state, and the steady-state thermal-hydraulic loss coefficients. These loss coef.ficients are -
developed using the steady-state thermal-hydraulics methodology and are derived from full scale fuel assembly specific pressure drop data as a function of power and flow. The three-dimensional simulator can also be used in the analysis of slow transients to-determine the change in critical power ratio (" Delta CPR") for these events.
The transient analysis is used to determine the peak tran-sient pressure, the transient change in power, and the transient heat flux and thermal-hydraulic parameter changes utilized in the transient crit ical power :aethodology. The transient analysis model utilizes the plant configuration and performance parame-ters. The trancient change in power is used to demonstrate con-formance with the applicable thermal-mechanical performance lim-its and used as part of the process for confirming the MAPLHGR ,
operating limit as described in the ECCS evaluation methodology -
above. Transient analyses are also performed at the limit;ng points on the power flow map to define the power / flow dependent MAPLHGR and MCPR operating limits.
The transient critical power calculational methodslogy is used to calculate the Delta CPR from the initial critical pover ratio ("!CPR") assumed as an initial condition for the transient being evaluated. This defines the Delta CPR during the tran-sient. The transient critical power calculational methodology requires the use of the critical power correlation, the fuel per-formance characteristics, power shape, number ot fuel assemblies, fuel assembly design characteristics, and an assumed ICPR. The MCPR calculated during the transient is compared to the safety limit. If the minimum value of CPR during the transient does not correspond to the safety limit, the transient critical power methodology results are used to define a new power level and flow distribution for an adjusted ICPR. This process is continued until the minimum value of CPR during the transiert corresponds to the MCPR safety limit. The uncorrected Delta CPR can Fe
- obtained from this last iteration. The Delta CPR is adjrsted for model and model input uncertainties and combined with the safety
'imit MCPR to establish the operating limit MCPR that is accept-uble from the transient analysis standpoint. The operating limit
' MCPR then becomes the highest value of the MCPR used in the ECCS evaluation methodology and the MCPR demonstrated acceptable in transient analysis methodology.
In summary, MCPR and MAPLHGR operating limits are estab-lished using the safety limit metnodology, the fuel thermal '
mechanical methodology, the ECCS evaluation methodology and the transient analysis methodology. Each of these methodologies has been reviewed and approved by the NRC. The approval includes the process for developing inputs, the various models and correla-tions used in the methodologies, the treatment of the model and model input uncertainties, and the application of the methodolo-gles. No changes can be made to the NRC approved documented methodology basis without prior NRC approval.
INTERROGATORY NO. St For each analytical methodology iden-titled in your answer to Interrogatory No. 4, identify all areas of uncertainty in the computational model and input values, the reasons for the uncertaint) and any limits on tne uncertainty.
ANSWER TO INTERROGATORY NO. 5: Uncertainties are considered in the methodologies used in the reload license analysis process described in the response to Interrogatory No. 4. The areas of uncertainty are the correlation and correlation input uncertain-ties, model and model input uncertainties, and measurement uncer-tainties. As discussed in the response to Interrogatory No. 4, each of these methodologies, incivding the treatment of uncer-tainties, has been approved by the NRC and no change can be made to the approved documented methodology without prior NRC approval.
Correlation uncertainties are the uncertainties associated with empirical correlations. These uncertainties are generally att *buted to differences between test configurations and the actual plant configuration and include the effects of data scat-ter, data reduction techniques, etc. Correlation input uncer-tainties are the uncertainties associated with the parameters that are used in empirical correlations. These uncertainties include uncertainties in the plant systems performance, uncer-teinties in the plant operating state, and msnufacturing variability.
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i Model uncertainties are due to the approximation in the !
analysis models. These uncertainties may be attributed to numer- )
ical solution techniques, model simulations, etc. Model input )
uncertainties include all uncertainties associated with the anal- 1 ysis model or methods inputs and may be separated into configura- )
tion and process uncertainties. Configuration uncertainties are the differences between the actual systems and components config-uration and the model representing their configuration, including flow areas, volumes, fuel and core characteristics, piping con-figurations, etc. Process uncertainties are introduced by model simplifications representing the physical plant processes includ-ing their dynamic effects. Process uncertainties include the effects of the process parameters or instrument system geometry on instrument response and the phenomenological differences between the process paremeter and the parameter simulation in the analysis process.
Measurement uncertainties are the uncertainties in the 1 instrument system or sensor performance. Measurement uncertain-ties include instrument system uncertainties, instrument drift, and uncertainties in the calibration proceFS. Instrument system >
uncertainty epresents the capability of the process measurement system considering instrument biases, environmental effects, electronic processing errors, and signal to noise uncertainties.
Instrument drift is the change in the output-input relationship of the instrument in the period of time between instrument cali-brations. Calibration uncertainty is the uncertainty in the pro-cess fot physically establishing the specified trip setpoint con-siderlag the instrument system.
Uncertainties are generally characterized as a statistical distribution which does not have any specific limits. This is acceptable as long as the uncertainties are appropriately treated
! in the analytical process. In the methodology for developing the MCPR and MAPLHGR operating limits, some uncertainties are trented deterministically; the remaining parameters are treated statisti-cally. The uncertainties in the LHGR limit are statistically considered in the development of the thermal-mechanical perfor-mance limits. In the deterministic process the inputs are selected in a manner to assure a conservative analysis result.
In the statistical analysis process uncertainties are statisti-cally treated in a manner to satisfy a statistical goal.
INTERROGATORY NO. 6: For each analytical methodology iden-tified in your answer to Interrogatory No. 4, identify each instance in which the user of the methodology is afforded discre-tion or judgment in its application, and any limits on this dis-cretion or judgment.
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6NSWER_TO INTERROGATORY NO. 6: The methodologies used to establish the MCPR and MAPLHGR are identified in the response to Interrogatory No. 4. As discussed in that response, each of these methodologies has been reviewed and approved by the NRC.
These approvals limit the application of the methodologies. No -
changes can be made to these NRC approved methodologies or appli-cations without prior NRC approval. Therefore, the user is lim-ited by this NRC approval in applying the methodologies.
INTERROGATORY ;L. a For each analytical methodology iden-tified in your anteer to .nterrogatory No. 4, identify all input variables and user cor+.011ed or -determined options, and any limits on the range a. values of these variables and options.
ANSWER TO INTERROGATORY NO. 7: As stated in tne response to Interrogatory No. 4, the process that GE Nuclear Energy uses to develop inputs to the methodologies that establish MCPR and -
MAPLHGR operating limits are approved by the NRC. No change to NRC approved documented methodologies and processes can be made without prior NRC approval. The input parameters used in the reload license analysis process are developed consistent with the NRC approvals of the specific methodologies and are based on the plant design and configuration described in the safety analysis report. The input parameters to the methodologies are developed from controlled design documents.
Two documents control the flow of information between GE and the utility during the reload design process, the Fuel Release and Engineering Data (" FRED") and Transient Protection Parameters Verification for Reload Licensing Analyses ("OPL-3") forms.
Guidelines for generating OPL-3 inputs are contained in GE Pro-l prietary Document NEDE-22061.
The FRED form specifies the principal operating data and analysis options that are to be used for the analysis of the operating cycle. (GE has developed generic reload license analy-sis approaches for different reactor designs, modes of operation, and operating margin improvement strategies. The FRED form basi-cally is used as a menu to specify which analyses are applicable based on the licensed plant conditions).
The OPL-3 form provides the parameters used in the reload design process. The information and review process documentation is presented in several columns. The first column is a tabula-tion of the parameter values used in the reload analysis for the l preceding cycle. The second column lists the values which are proposed by GE to be used for the cycle to be analyzed. Differ-ences in the two columns should exist only if design changes have been made or updated information is incor; orated. Other columns provide space for the utility to respond *n the GE proposed
-B-
values, and finally document the mutually recolved values that are the basis of the reload analysis for the analyzed cycle.
This review process ensures that the plant configuration provided cc input to the reload license analysis is currert (for example, updates to the initial cycle values occurred as a esult of the Startup Test Program and were included in the subssquent analyses to more accurately model the plant design and response).
This process, therefore, ensures that the analytical results rep-esent the actual plant configuration and is consistent with NRC regulatory requirements.
Further, the key reload license analysis input parameters are analytical limits which determine'the values found in the plant Technical Specifications. Therefore, these analysis inputs cannot be changed in a way which would reduce the margin of safety in the technical specification bases or would require a technical specification change without obtaining a license amend-ment consistent with the requirements of 10 C.F.R. Section 50.90.
The plant Technical Specifications are derived from the analyses and evaluations in the safety analysis report and its amendments (they also currently include other requirements that the NRC finds appropriate but which are not directly related to the analyses and evaluations). The Technical Specifications con-tain four categories of requirements that are used to develop inputs to the reload license analysis ptocesst (1) safety lit its; (2) limiting safety system settings; (3) limiting conditions for operation; and (4) design features. Safety limits are limits placed upon important process variables necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. Limiting safety system settings are settings for automatic protective devices relating to variables having significant safety func-tions. Design features are those features of the facility which, if altered, would have a significant effect on safety and are not covered by the other three categories. The key input parameters controlled by the Technical Specifications are listed in Table 1.
Any change in these inputs requires NRC review and approval of a license amendment.
The remaining input parameters fall into two general catego-ries: (1) plant configuration and (2) plant operating state, based primarily on a plant heat balance and consistent with the s power level and core flow conditions. These input parameters are listed in Table 2. As described above, the plant configuration is contained in controlled design documents. The plant heat bal-ance is restricted by the technical specification limits placed on reactor power level, feedwater temperature and vessel pres-sure. Therefore, changes to the parameters used as reload
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i license inputs identified in Table-2 without NRC review and
-approval are significantly restricted.
The input parameters are, therefore, based on the plant j design and configuration, as determined from design documents, j the safety analysis report, testing data and Technical Specifica- l tion limits. {
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Table i Key Input Parameters Controlled by the Technical Specifications Licensed Core Thermal Power Feedwater Temperature High APRM Neutron Flux Scram Setpoint High APRM Thermal Power Scram Setpoint Flow Referenced Thermal Power Monitor Setpoints Flow Referenced Thermal Power Monitor Time Constant Plow Referenced Signal Time Constant High Vessel Dome Pressure Scram Setpoint Dome Pressure Sensor Time Constant MSIV Position Switch Scram Setting l Turbine Stop Valve Position Switch Scram Setpoint Timing of Turbine Control Valve Fast Closure Scram Signal Turbine Stop Valve and Control Valve Fast Closure Scram Bypass Setpoint Reactor Protection System Delay Time control Rod Motion During Scram Recirculation Pump Trip Characecristics High Pressure Recirculation Pump Trip Setpoint High Water Level Scram Setpoint
. Low Water Level Scram Setpoint Low Water Level Sensor Time Constant t
Relief Valve Opening Setpoint Safety Valve Opening Setpoint High Pressure Core Spray System Capacity Low Pressure Core Spray System Capacity Low Pressure Coolant Injection System Capacity Number of Automatic Depressurization System Valves RCIC System Capacity Low Water Level ECCS Initiation Setpoints Low Pressure ECCS Permissive ECCS Time Delays Diesel Generator Starting Times High Containment Pressure ECCS Initiation Setpoint High Water Level RCIC and HPCS Trip Setpoints Low Water Level Isolation Setpoints Lov Water Level Recirculation Pump Trip Setpoints Low Main Steamline Pressure Trip of MSIVs Low Condenser Vacuum Setpoint for Trip of MSIVs MSIV Closure Time Reload Fuel Types
Table 2 ,
Plant Design Parameters Power Level Core Flowrate Steam Flowrate '
Feedwater Flowrote Recirculation System Characteristics ,
Operating Vessel Dome Pressure ,
Normal Reactor Water Level Dome to Turbine Pressure Drop -
Limiting Loss of Feedwater Temperature '
High Water Level Feedwater Pump Trip '
Steam System Configuration Recirculation System Configuration Vessel Configuration Feedwater System Flow Runout Limits Maximum and Minimum Condensate Storage Temperature Turbine Control Valve Characteristics Stop Valve Closure Characteristics Load Rejection Characteristics Turbine Bypass Valve Characteristics i Core and Fuel Design Characteristics e
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INTERROGATORY NO. 8: For each analytical methodology iden-tified in your answer to Interrogatory No. 4, identify the person or business entity who performs the calculations.
ANSWER TO INTERROGATORY NO. 8: All reload license analyses to establish the cycle dependent MCPR and MAPLHGR operating lim- ;
its are performed by GE Nuclear Energy.
INTERROGATORY NO. 9: For each analytical methodology iden- )
tified in your answer to Interrogatory No. 4, indicate whether there is any independent verification of the calculatien per-formed, and identify the person or business entity who performs this independent verification.
ANSWER TO INTERROGATORY NO. 9: All of the above analyses performed by GE Nuclear Energy are fully verified in accordance ,
with the requirements of 10 C.F.R. Section 50 Appendix B. This j verification is part of the overall GE Nuclear Energy Quality Assurance Program approved by the NRC. The verification process I for the above analyses include: (1) utility concurrence on the analysis inputs; (2) complete GE engineering verification of the analysis inputs (including verification of computer data entries) and results; and (3) a multi-disciplinary review of the results by GE engineering, licensing, and project personnel. These reviews must be complete and all items addressed prior to release of the information to the utility.
CEI reviews the results of the analyses. The review process includes a review of the GE reload licensing submittal prepared for that reload by the Reactor Engineering and Fuel Management Unit, and the licensing and engineering organizations. Addition-ally, any changes to the Core Operating Limits Report ("COLR")
are required by plant procedures to be reviewed by the Plant Operations Review Committee ("PORC") which provides a '
multi-disciplinary review of both the COLR changes and GE reload licensing submittal.
Also, as part of the technical review process for the reload license analyses, the CEI fuel management organization performs evaluations of GE Nuclear Energy as an independent verification of their activities. Additionally, the Perry Nuclear Power Plant Nuclear Assurance Department performs quality assurance audits of the GE reload program. The GE Quality Assurance Program is sub-ject to periodic audits by all reload fuel customers and the NRC.
1 INTERROGATORY NO. 10: For enth analytical methodology iden-tified in your answer to Interrogatory No. 4 that uses computer models, indicate whether the computer program will reject input values that are unreasonable or physically impossible. If so, state whether default values are used instead, and the value of and the basis for the def: ult value.
ANSWER TO INTERROGATORY NO. 10: The GE Nuclear Energy engi-neering computer programs used in the reload license analysis process do-not reject input data and substitute default values.
The analysis verification process described in the response to Interrogatory No. 9 provides the independent verification that assures the accuracy and consistency of the analysis inputs, the correct application of the methodologies and the accuracy of the analysis results.
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O REQUEST FOR PRODUCTION OF DOCUMENTS FRED and OPL-3 forms are available for inspection and copy-ing at the Cleveland Electric Illuminating Company by contacting Mr. Robert Newkirk at. (216) 259-3737, extension 5606. NEDE-22061 is also available for inspection and copying, subject to an appropriate proprietary agreement, by contacting Mr. Newkirk.
Respcrtfully submitted, SHAW, PITTMAN, POTTS & TROWBRIDGE 2300 N Str6et, N.W.
Washington, D.C. 20037 (202) 663-8063
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Ja . silberg Da M Becker J Cou e for Licensees Dated: August 31, 1990 4
B 330DMB5430.90 l
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AFFIDAVIT Robert.A. Nevkirk, being duly svorn, states that he is employed as Mcnager of the Licensing and Compliance Section, Perry Nuclear Support Department of the Cleveland Electric Illuminating Company; that he has read Responses 2, 7 and 9 of the foregoing " Licensees' Response to Ohio Citizens for Responsible Energy Inc.'s Interrogatories and Request for Production of Documents"; that he believes the ansvers contained therein to be true to the best of his knowledge and informations and that he is authorized to sign the foregoing ansvers on behalf of the Cle 41and Electric Illuminating Company.
Subscribed and svern to to before me this 1s' day of August,1990, Not an V Ablic W
My Commission Expires: __ lentary m
-r= - - - . =o.. no. ,
MM bi Lake County) i 3
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i RESUME !
Name Robert A. Nevkirk, Manager, Licensing and Compliance Section Formal Education and Training:
Bachelor of Science Degree, U.S. Naval Academy, 1964 Nuclear Power Training,'U.S. Navy, 1965-1966 Reactor Operator Course, General Electric Company BVR Training Center, Morris, Illinois SRO Licensed on Dresden 2 and 3 and Ovad Cities 1 and 2 Experience:
1985 - Present: The Clevelaad Electric Illuminating Company In July 1988 named Manager of the Licensing and Compliance Sectio.._ As such, is responsible for the overall Licensing and, compliance of the Perry Plant. He is responsible for Technical Specification and License-Ar endments, USAR Maintenance and Regulatory. compliance including LER preparation. In addition he is the primary project liaison'vith the Nuclear Regulatory Regulations (NRR), Vashington D.C. and Region III Chicago, Illinois. 'He reports _to the Director, Nuclear Support-Department.-
In June 1986 named Manager of the Technical Section. As such was responsible for all activities associated with providing technical support and services relating to monitoring plant performance, surveillance engineering, systems engineering, reactor engineering and core analysis.
In March 1986, named General Supervising Engineer. Electrical Design.
Responsible for directing all electrical and 16C design activities-associated with modifications and additions to the operating plant.
I In January 1986, was named General Suoervising Engineer. Nuclear Design and Analysis Section. As such, was responsible tor technical support on various Licensing, startup, and preoperational requirements. The Section has engineering responsibility for all systems turned over to operations and provided all design for modifications and additions to the. operating plant.
In September 1985 joined CEI as Senior Staf f Entineer in the Nuclear Engineering Department. Initial assignment was Chairman of the l Independent Safety Engineering Group (ISEG).
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t 4-1974 - 1985: Puget Sound Pover &~ Light' Assigned as assistant to Vice President, Engineering and operations.
'Vhile on laave of absence from February 1985 to May 1985, served as
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consultant to the Public Service Electric and Gas Company's Plant-Technical Manager at the Hope-Creek Nuclear Gener. ting Station. Also, was on loan to Portland General Electric for 6 months to develop and implement a performance mor.itoring program for Trojan Nuclear- Plant.
Served as a loaned employee to INPO f rom ' July 1980 to August 1982.-
Participated in 11 INPO evaluations of_ operating nuclear plant, primarily in the maintenance and technical support areas of PVRs and-BVRs.
Designated as Skagit Nuclear Plant Manager between' April 1979 and June ~
1980. -As Senior Project Engineer for the Skagit Plant from April 1974 to-March 1979, provided_ direction to the architect / engineer and performed design review for operations.
1970 - 1974 Comaonwealth Edison Company q Vas department head for technical staff .At the Ouad Cities Nuclear Power-Station from August 1971 to April 1974 during initial fuel loadingLon both units. Responsible for nuclest engineering, plant; modification control, performance monitoring, instrumentation and control maintenance, radiation protection and chemistry, and quality control ~. Licensed senior Reactor Operator for Quad Cities Units 1 and 2. . Also held SRO license at Dresden a startup engineer on shift during fuel load and power ascension testing at Dresden Unit 3.
1964 - 1970: Unites States Navy Served as Reactor Control and Communications Officer on USS Plunger (SSN-595) and Damage Control Officer on USS Puffer (SSN-652).
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AFFIDAVIT l 1
Janice S. Charnley, being duly sworn, states that she is i' employed as Marager, Fuel Licensing of GE Nuclear Energy; that she has read Responses 4, 5, 6, 7, 8, 9, and 10 of the foregoing " Licensees' Response to Ohio Cititetts ' for Responsible
- i Energy Inc.'s Interrogatories and Request for Producticn of Documents"; that she believes the answers contained therein to be true to the best of her knowledge and information; and that >
she is authorized to sign the foregoing answers on behalf of GE ,
Nuclear Energy. -J L .
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Subscribed and sworn to before me ) is j29 day f Augupf 1990, ed ?!1M/'.
' Notary Public
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ab My commission Expires: // / /98 I
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< . i STATEMENT'OF QUALIFICATIONS Janice S. Charn? ey graduated in 1969 from the University of Texas with a Bachelor of Science Degree in Engineering Science.
She began her career as a analytical test design-engineer at i the GE operated Knolls Atomic Power Laboratory where she was )
responsible for nuclear and thermal-hydraulic test analyses 1 including special heat transfer and' depletion studies. In l 1973, she transferred to the San Jose nuclear power division as a licensing engineer. In this position and subsequently as a senior licensing engineer, she successfully licensed the GE Thermal Analysis Basis and developed and licensed. the generic GESTAR reload fuel design and analysis documentation. In 1982, she was appointed the Manager, Fuel Licensing. In this ;
position,- she supervises the GE fuel licensing activities for all domestic and international plants. These activities.
include fuel and cora design and analyses safety assessments, generic fuel and core design and analytical methods licensing, and plant specific fuel and core analyses documentation. ,
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August 3SMYfo UNITED STATES OF AMERICA '90 SEP -4 A11 :21 .
NUCLEAR REGULATORY COMMISSION l
~rHCE N SEcallAn l Before the Atomic Safety and Licensino Board 10CKI1 NG A 'iltwlCI !
BR A NC61 l 4
In the Matter of )
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THE CLEVELAND ELECTRIC )
ILLUMINATING COMPANY, et al. ) Docket No. 50-440-OLA-2 i'
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ASLBP No. 90-605-02-OLA (Perry Nuclear Power Plant, ) i Unit No. 1) ) i
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CERTIFICATE OF SERVICE l l
I hereby certify that copies of the foregoing Licensees' Response to Ohio Citizens for Responsible Energy, Inc.'s Inter-rogatories and Request for Production of Documents were mailed, postage prepaid, this 31st day of August, 1990 to those listed on the attached Service List. >
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(. S11 berg niel for.-Licensees r
Bi3300MB5430.90 i
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' August 31, 1990 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensino Beard In the Matter of )
)
THE CLEVELAND ELECTRIC )
ILLUMINATING COMPANY, et al. ) Docket.No. 50-440-OLA-2
)* ASLBP No. 90-605-02-OLA (Perry Nuclear Power Plant, )
Unit No. 1) )
)
SERVICE LIST Docketing and Service Branch Ms. Susan Hiatt Secretary of the Commission 8275 Munson Rott U.S. Nuclear Regulatory Mentor, Ohio 44060 Commission Washington, D.C. 20555 Atomic Safety and Licensing Appeal Board Colleen P. Woodhead, Esq. U.S. Nuclear Regulatory Office of the General Counsel Comaission U.S. Nuclear Regulatory Washington, D.C. 20555 Com.niss ion Washington, D.C. 20555 Dr. Jerry R. Kline Atomic Safety and Licensing Chairman Board Atomic Safety and Licensing U.S. Nuclear Regulatory Board Panel Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 Dr. Frederick J.. Shen Atomic Safety and Licensing John H. Frye, III, Chairman Board Atomic Safety and Licensing U.S. Nuclear Regulatory Board Commission U.S. Nuclear Regulatory Washington, D.C. 20555 Commission Washington, D.C. 20555 B:330DMB5430.90 i
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