ML17310A342
| ML17310A342 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 06/14/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17310A341 | List: |
| References | |
| NUDOCS 9306180300 | |
| Download: ML17310A342 (11) | |
Text
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)&**~4 UNITED STATES NUCLEAR REGULATORY COMM)SSION WASHINGTON, D.C. 20555.0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO ON-SITE AUDIT OF PALO VERDE RELOAD ANALYSIS METHODOLOGY ARIZONA PUBLIC SERVICE COMPANY DOCKET NOS. 50-528 50-529 AND 50-530
- 1. 0 INTRODUCTION In a letter of April 6, 1993, Arizona Public Service Company (APS) transmitted the Reload Analysis Methodology Report for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 to the U. S. Nuclear Regulatory Commission (NRC) for review (Ref. 1).
This report summarizes the program which was undertaken by APS to develop the capability to independently perform the analysis of a PVNGS reload cycle.
In order to develop the capability to independently perform the analyses required for the design, licensing, operation and surveillance of a reload fuel cycle, APS contracted Asea Brown Boveri/Combustion Engineering (CE) to provide a training program referred to as the Reload Technology Transfer Program.
This program consisted of classroom lectures, on-the-job training, and independent analysis.
The scope of the program included all reload engineering technology except loss of coolant accident (LOCA) analysis, fuel mechanical
- design, and fuel fabrication engineering.
These areas remain the responsibility of the fuel vendor, currently CE.
The models and methods used have been previously approved by the NRC for use by CE.
The independent analysis was the final phase of the training program, whereby APS performed parallel reload analyses independent of CE.
In order to evaluate the ability of APS to properly utilize the computer codes and methods for reload core design, the NRC conducted an on-site audit at the APS offices in Phoenix, Arizona, from May 10 through May 12, 1993.
The NRC staff performing the audit consisted of Charles Trammell, Edward Kendrick, and Laurence Kopp from Headquarters, and Dennis Kirsch from Region 5.
The following CE-developed computer codes are used in the physics analyses:
(1)
ROCS (Ref. 2),
a coarse-mesh, two-energy group higher order difference diffusion theory neutronics code which can model all aspects of reactor operations from startup to refueling; (2)
MC (Ref. 2),
a fine-mesh, two-energy group diffusion theory neutronics code which calculates fine-mesh (pin-wise) flux, power and burnup distributions through the application of the nodal imbedded method to 9306280300 9306i4'DR ADDCN 05000528 P PDR j
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(3)
(4)
(5) individual fuel assemblies using inter-assembly currents calculated by the coarse-mesh ROCS code; HERMITE (Ref. 3),
a few-group, space and time-dependent neutron diffusion code which includes feedback effects of fuel temperature, coolant temperature, coolant density and control rod motion; gUIX (Ref. 4),
a two-group, one-dimensional diffusion code used for axial shape analysis; and VISIONS (FLAIR) (Ref. 5),
a three-dimensional, fast-running PWR simulator used to evaluate the response of the excore detectors to core power shape variations.
The core thermal-hydraulic analyses use the following CE codes:
TORC (Ref. 6),
a three-dimensional, open-lattice core thermal-hydraulic code used to determine the local coolant conditions
- and, in turn, the minimum departure from nucleate boiling ratio (DNBR) for the core; and (2)
CETOP-D (Ref. 7),
a fast-running variant of the TORC code used as a
design code in thermal margin analysis.
The fuel performance analyses use the following CE codes; FATES3A (Ref. 8),
a fuel evaluation code which predicts the steady-state fuel rod temperature distribution, gap conductance, fuel and clad dimensions, plenum pressure, and stored energy for CE-designed
- fuel, and includes the NRC-required grain size restriction in the fission gas release calculation; and (2)
FATES3B (Ref. 9),
a revised version of FATES3A with an improved predictive capability at high burnup.
In addition to the HERMITE and CETOP-D codes mentioned
- above, the non-LOCA transient and accident analyses are performed with the following codes:
CESEC-III (Ref. 10),
a system code which incorporates point kinetics, reactivity feedback, and core thermal-hydraulics to calculate system parameters including core heat, flow, pressure, and temperature, during a transient; and (2)
STRIKIN-II (Ref. 11),
a code which provides a single, or dual, closed
,channel model of a core flow channel to calculate the clad and fuel temperatures during a transient.
I
- 2. 0 EVALUATION ELOAD DESIGN The specific disciplines required of APS to implement a reload design are the following:
I (1)
(2)
(3)
(5)
(6)
Physics design Core thermal-hydraulics design Fuel performance design Transient and accident analyses (except LOCA)
Generation of Core Operating Limit Supervisory System (COLSS) and Core Protection Calculator (CPC) setpoints, data base constants, and core operating margin assessment The Reload Analysis Nethodology Report presented comparisons of characteristic physics parameters calculated by APS for PVNGS 3 Cycle 3 with those calculated by CE.
These included critical boron concentration, beginning-of-cycle (BOC) and end-of-cycle (EOC) boron worths, moderator temperature coefficients (HTCs), Doppler coefficients, CEA reactivity worths, and fuel assembly relative power densities.
The comparisons show that the results of the CE and APS analyses agreed very closely.
The minimal differences can be attributed to the difference in previous cycle burnup assumed by APS and CE.
The comparison of core thermal-hydraulics parameter results between APS and CE for PVNGS 3 Cycle 3 were essentially identical.
The low power physics tests and the power ascension tests currently performed at BOC cover sufficient physics parameters to reasonably assure that the core is operating as designed, and adequate shutdown margin is available.
In addition, the Core Follow Program currently performed by APS will be maintained to monitor physics and thermal-hydraulic parameters throughout core lifetime.
During the staff audit, selected calculations were examined and discussions were held with the APS staff to clarify specific points.
These calculations included fuel performance
- design, margin setting events (anticipated operational occurrences) such as loss of flow, CEA withdrawal, CEA drop, and inadvertent boron dilution, and fuel failure events (accidents) such as steamline
- break, locked rotor, and CEA ejection.
For some of the analyses
- examined, CE relied on calculations performed for a previous cycle.
The corresponding APS calculations, on the other hand, were more detailed and demonstrated a comprehensive understanding by APS of reload technology.
The APS design calculations. for their independent analysis of Unit 3 Cycle 3
were also forwarded to CE for a technical review to verify the ability of APS to correctly implement CE models and methods.
In a letter from N. J.
Breckenridge (CE) to P.
F. Crawley (APS), dated November 30, 1992 (Ref. 12),
CE concluded that the final APS Reload Topical Report and the underlying recorded calculations demonstrate proper application by APS Nuclear Fuel Management staff of the transferred CE methods and models to the PVNGS.
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CONTROLS ON CPCS AND COLSS CONSTANTS The NRC also reviewed the procedures that APS uses for review, implementation, and control of the constants that are installed in the Core Protection Calculator System (CPCS) and the Core Operating Limits Supervisory System (COLSS).
These procedures were determined to be acceptable.
CO PUTER FACILITIES APS reload design calculations will be run on Hewlett Packard 9000 Series 433 workstations.
The three current APS workstations are configured identical to the workstations used by CE and will be directly connected to APS's existing Sun Hicrosystems LAN.
All CE codes will be installed on each workstation in a controlled directory.
Although CE maintains the source code for APS, both the CE/APS Fuel Contract and the APS Software gA Procedure contain provisions for APS to maintain the source code independently of CE should APS decide to do so.
Any source code changes made by APS will be governed by 05AC-ONFll, "NFH Software guality Assurance Program for Non-Process Computer Software,"
and by 05DP-ONF09, "NFH Analysis Controls."
Should APS make any source code changes, independently of CE, which change the calculated values of any safety related parameters, APS will perform a thorough engineering evaluation, validation, and verification prior to use of the modified source code for licensing related activities.
If necessary, a topical report covering the code modifications would be prepared for NRC review and approval.
CHANGE IN FUEL VENDOR As long as CE remains the PVNGS fuel vendor, qualification of models and methods will be performed using the CE qualification process.
A change in fuel vendor will require an evaluation of any changes required to the physics and safety analysis methodology to accommodate that vendor's particular fuel designs.
Changes of this type would undergo a thorough engineering evaluation, validation, and verification prior to use of the new fuel design.
DESIGN CONTROL The Nuclear Fuel Management (NFM) process is controlled by several procedures ranging from program description procedures to detailed implementing procedures.
These procedures were sampled and reviewed to assess the degree to which the NFH program implements the requirements of the quality assurance
- program, in general, and the commitments to ANSI N45.2. 11 addressing design control.
The staff concluded that the NFH program and implementing procedures adequately implement these requirements and commitments.
The auditors examined two quality assurance audits which addressed various aspects of the NFH program and the resolution of several findings and recommendations.
The staff concluded that the oversight of the NFH resulted in finding and correcting the minor problems.
TRAINING The NFM program for training and qualifying the staff assigned to perform core reload design work was examined and found to be adequate to assure a well qualified staff.
The training program procedure was reviewed and found to detail an appropriate mix of training lectures, job performance
- measures, and knowledge assessment techniques to assure an adequately qualified staff.
Written training materials used by the instructors were sampled and reviewed and found to contain the requisite level of detail to accomplish the training lecture purpose.
Training records of several engineers were reviewed and found to demonstrate that the NFM organization contains a staff which has been adequately trained and certified to perform the required functions.
Four engineers were interviewed to assess their perceptions regarding their preparation to perform their tasks; all professed an adequate knowledge level and supportive management.
ADDITIONAL EXAMPLES OF APS RELOAD DESIGN ABILITY APS engineers have participated with CE in all phases of the Unit 2 Cycle 5
reload design.
In addition, APS engineers have performed specific portions of reload calculations used by CE for Unit I Cycle 3, Unit 2 Cycle 3, and Unit 3 Cycle 3 reload analyses.
APS engineers perform the fuel management for current reload cycles and have performed this phase of design for a number of cycles for each of the PVNGS units.
In addition, APS engineers have been involved in a number of fuel management scoping studies including investigations of high or extended burnup and alternative burnable absorber designs.
APS has developed a set of "Fuel Management Guidelines" which were reviewed and accepted by CE.
These guidelines were based on APS reload analysis knowledge and provide up-front guidance to avoid safety analysis problems by placing limits on key core physics parameters.
APS has supplied in-house support for many plant events and issues including development of reload specific Core Operating Characteristics
- Reports, reload startup tests, development of core physics for Core Follow Program, Core Data Book and startup test predictions.
APS also has the ability to independently complete
- JCOs, 50.59s, Nuclear Safety Assessments following plant events, and Safety Assessments of Technical Specification changes.
These activities provide a satisfactory assessment of the NFM organization's readiness to independently perform core reload analyses.
- 3. 0 CONCLUSION Based on the results of this audit, and recognizing that APS has participated in the CE Reload Technology Transfer
- Program, we conclude that the APS staff has the capability to use the CE codes under discussion for non-LOCA reload analyses of the CE-fueled PVNGS cores.
The Topical Report "Reload Analysis Methodology for the Palo Verde Nuclear Generating Station," Revision 00-P (Ref. I) describes the reload design process and the scope of the analyses
which may be performed by APS, and is acceptable for referencing in PVNGS licensing applications.
4.0 REFERENCES
(1)
"Reload Analysis Methodology for, the Palo Verde Nuclear Generating Station," Revision 00-P, February 1993.
(2)
"The ROCS and DIT Computer Codes for Nuclear Design,"
CENPD-266-P-A, April 1983.
(3)
"HERMITE, A Multi-Dimensional Space-Time Kinetics Code for PWR Transients,"
CENPD-188, March 1976.
(4)
"QUIX User's Manual," CE-CES-79, REV. 0-9, Hay 1987.
(5)
Bollacasa, D.
and J.
C. Stork, "VISIONS - Versatile, Interactive Simulator of Nuclear Systems,"
American Nuclear Society Meeting, Nov.
29 Dec.
3, 1981.
(6)
"TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core,"
CENPD-161-P-A, April 1986.
(7)
"CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3," CEN-160-(S),
Rev.
1-P, September 1981.
(8)
"Improvements to Fuel Evaluation Model," CEN-161(B)-P-A, August 1989.
(9)
"Improvements to Fuel Evaluation Model," CEN-161(B)-P, Supplement 1-P-A, January 1992.
(10)
"CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System,"
CENPD-107, April 1974.
(11)
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program,"
CENPD-135, April 1974.
(12)
Letter from N. J. Breckenridge (CE) to P.
F. Crawley (APS), V-92-237, "ABB CE Review of the APS Reload Topical Report,"
November 30, 1992.
Principal Contributors:
L. Kopp E. Kendrick D. Kirsch C. Trammell Date:
June 14, 1993