ML20073H244
| ML20073H244 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/28/1991 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19298E441 | List: |
| References | |
| NUDOCS 9105070030 | |
| Download: ML20073H244 (20) | |
Text
________ ___ _
which are functions of temperature and has the capability to perform a sophisticated fuel-to-clad gap heat transfer calculation.
Additionally, the code can perform the necessary calculations to deal with post-DNB transient phenomena such as: film boiling heat transfer correlations, the zirconium-water reaction, and partial melting of the materials.
For the CEA ejection analyses, the nuclear power input is a time-dependent parameter from TWINKLE: however, the input parameters for pressure, flow, temperature and density are maintained at a i
constant conservative value so as to produce a limiting calculation in FACTRAN.
FACTRAN i<, further discussed in Reference 3.
3.0 lafety Limits The real physical limits of this accident are that any consequential damage to either the core or the reactor coolant system must not prevent long-term core cooling and that any offsite dose consequences must be within the guidelines of 10 CFR Part 100.
However, the magnitude of fuel failure will be determined by the following limits:
1.
The average pellet deposited energy at the hot spot is no greater than 200 cal / gram (clad damage threshold).
2.
The centerline enthalpy threshold for incipient melting is no greater than 250 cal / gram.
3.
The centerline enthalpy threshold for the fully molten condition is no greater than 310 cal / gram.
The criterion for determining the fraction of fuel rods that will release th11r radioactive fission products during the CEA ejection is the one quoted above for determining clad damage.
4.0 Rglerences 1.
Risher, D. H., Jr., "An Evaluation of the Rod Ejection ^.ccident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision I-A, January 1975 (Proprietary).
2.
Risher, D. H., Jr. and Barry, R. F.,
" TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A, January 1975 (Proprietary) and WCAP-8028-A, January 1975 (Non-proprietary).
3.
Hunin, C.,
"FACTRAN Code Description," WCAP-7337, Rev l-P, December 1989 (Proprietary).
9105070030 910419 PDR ADOCK 050002G5 P
t 4
Westinghouse Class 3 DEPORT ON THE ANALYSIS METHODS AND EVALUATION MODELS TO BE EMPLCfE0 IN THE LARGE BREAK AND SMALL BREAK LOCA ANALYSES FOR FORT CAUCOM UNIT 1 0
WESTINGHOUSE ELECTRIC CORPORATION SAFETY ANALYSIS TECHNOLOGY FEBRUARY 1991
Westinghouse Class 3 TABLE OF CONT'NTS t
Section T111g 1.0 INTR 00VCT10N 2.0 SYSTEM DESCRIPTION 3.0 LARGE BREAK LOCA 3.1 LARGE BREAX LOCA ECCS EVALUATION H0 DEL 3.1.1 SATAN-VI 3.1.2 MREFLOOO 3.1.3 COCO 3.1.4 BART 3.1.5 BASH 3.1.6 LOCBART 4.0 SMALL BREAK LOCA 4.1 SMALL BREAX LOCA ECCS EVALVATION H00EL 4.1.1 NOTRUMP 4.1.2 LOCTA-IV
5.0 REFERENCES
e E.
Wcstinghouse Class 3
1.0 INTRODUCTION
This report provides an overview of the Westinghouse Evaluation Models and analysis methodology which will be used to perform large Greak and Small Break LOCA licensing analyses for Fort Calhoun Unit 1.
Previous applications of Westinghouse LOCA methodology to a Combustion Engineering NSSS have relied heavily *n the approved Evaluation Models and methods employed for the analysis of Westinghouse plants.
Since most of the modeling and methodology is directly applicable to the analysis of a CE HSSS, the licensing precedent has been to directly reference the NRC-approved models and methods used for the analysis of Westinghouse plants.
The additional documentation required for licensing the application to a CE NSSS focused primarily on the modifications required to appropriately model the differences between the Westinghouse NSSS and CE NSSS designs, and the differences between the fuel designs for the two systems.
This report will assume a similar approach, referenc4ng existing NRC-Approved documentation, where possible, and focusing primarily on addressing plant and fuel difference relative to the LOCA analysis methodology.
Section 2.0 of this report contains a description of the CE (specifically Fort Calhoun) plant and fuel design, and the difference from a typical,Westingnouse plant and fuel design.-
The large Break LOCA analysis will be performed using a version of the 1981 Evaluation Model with BART/ BASH-(BASH EM), modified for application to a CE NSSS.
The BASH EM has been used for the analysis of Large Break LOCA in numerous licensing applications for Westinghouse fuel in Westinghouse plants. Modifications to the codes which constitute the BASH EM and special modeling methods will be developed and implemented for the analysis of Fort Calhoun Unit 1.
Proprietary and Non-proprietary versions of a topical report will be prepared describing the modifications.to the BASH EM for CE NSSS. This topical will be provided to the Omaha Public Power District (OPPD) and the Nuclear Regulatory Commission (NRC) in-support of the proposed Evaluation Model, analysis methodology and Fort Calhoun specific Large Break LOCA analysis.
}
Westinghouse Class 3 The Small Break LOCA analysis will be performed using the NRC-approved Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP code for the Combustion Engineering NS$5 (NOTRUMP EM).
The NOTRUMP EM has been used for the analysis of Small Break LOCA in numerous licensing applications for Westinghouse fuel in Westinghouse plants.
The version of the NOTRUMP EM developed for the analysis of a CE NSSS has previously been used to perform the Small Break analysis for Millstone Unit 2, which was subsequently reviewed and approved for licensing by the NRC.
Some modifications to the previously approved methodology will be implemented for'the analysis of Fort Calhoun Unit 1.
Proprietary and Non-proprietary versions of a topical report will be prepared describing the modifications to the NOTRUMP EM for CE NSSS.
This topical will be provided to OPPD and the NRC in support of the modified analysis methodology and Fort Calhoun specific Small Break LOCA analysis. A discussion of the various codes and modeling features of this model are contained in section 4.0 of this report.
2.0 System Descriotion The CE NSSS design has been reviewed to ensure the applicability of the models _and. methodology to be used for the Large and Small Break LOCA analyses.
Following is a brief summary of the unique features of the CE NSSS and proposed Fort Calhoun fuel design, highlighting the differences from a typical Westinghouse plant and 'f'uel design.
Loon layout: The CE design incorporates a system of two hot legs feeding two Steam Generators.
The two steam generators feed four cross-over. legs (loop seals) attached to four pumps which, in tyrn, feed four cold legs.
The cold leg diameter is slightly smaller than that in a Westinghouse NSSS design,- while the hot leg diameter is larger.
In addition,-the elevation Y
Westinghouse Class 3 of the loop seals is near the top of the core elevation, unlike the Westinghouse design, in v:Mch the loop seal elevation is near the core midplane.
Control Element Assembly (CEA) Desian: The CE design includes a variety of shroud designs for the CEAs. More than one design may be present within a given plant.
Vooer Head Bvoass/Vocer Head Temoerature: The CE design provides for negligible bypass flow from the upper downcomer to the upper head, resulting in an upper head temperature very near THOT' Core Shroud Bvoass: The core shroud region of the CE design differs from the Westinghouse Upflow Barrel / Baffle design in the type of flow paths available and relative flow resistance through the region.
For Fort Calhoun, for example, there are no flow holes through the centering plates (horizontal structures) in this region.
Slots at the interior-edge of the centering plates provide a bypass flow path from lower to upper plenum.
Reactor Coolant pumo: The reactor coolant pump design (single speed vertical shaft centrifugal pumps) for the CE plants are similar to those employed at Westinghouse plants, although the size of the pumps are somewhat smaller.
Steam Generator th,et Nozzle Anale: The Westinghouse NSSS features a Steam 0
Generator inlet nozzle inclined 40 from vertical', while CE designs have a greater angle of inclination.
Safety In.iection/ Accumulator Nozzle Anale: As in many Westinghouse plants, the CE design provides for a safet'y injection line which is connected to the cold leg at angle.
For Westinghouse plants, this angle is either or 90, while this angle is either 60 or 750 from horizontal 0
0 450 for the CE designs. The pressure differential due to the interaction of safety injection with steam flow in the cold leg piping is a function of this injection angle, b'
Westinghouse Class 3 Safety iniection Tank pressure: Westinghouse plants feature passive injection tanks (accumulators) with a cover gas pressure on the order of 600 to 700 psig.
Some CE designs may include Safety injection Tanks with a minimum cover gas pressure near 200 psig.
Fuel Assembly Desion: The CE fuel assembly design features guide tubes which are several times largar than the Westinghouse 14 x 14 guide thimbles.
The one instrumentation tube oer assembly is also much larger for the CE design. Grid parameters also vary slightly from the Westinghouse 14 x 14 designs.
Fuel Rod / Pellet Stack Desian: The cladding radial dimensions for the Westinghouse designed fuel for a CE NSSS differ somewhat from the Westinghouse designs, including open volumes inside the rod.
Also, the active pellet stack height for the CE fuel is often less than the typical 12 foot core length identified with most Westinghouse plants.
3.0 LARGE BREAK LOCA The Large Break LOCA analysis for Fort Calhoun will be performed using a version of the 1981 Evaluation Model with BART/ BASH (BASH EM), modified for application to a CE NSSS.
The BASH EM was developed to satisfy the requirements for the analysis of large Break Loss-Of-Coolant accidents as defined in 10 CFR 50.46(13, and employs the required models and assumptions' identified in 10 CFR 50, Appendix K(ll.
The BASH EM incorporates components.of the earlier.1981 Large Break LOCA ECCS' Evaluation Model[2), plus'BART technologyl33, to which the reflood thermal-hydraulic analysis capability of BASHI43 has been added. The codes used in ths BASH EM are SATAN-VI, MREFLOOD, C0CO, BASH and LOCBART.
-While not explicitly pa-t of the BASH series, the BART code has been modified and incorporated for specific application into the BASH and LOCBART codes.
Westinghouse Class 3 Among these codes, those which were included in the 1981 Evaluation Model (SATAN-VI, WREFLOOD, COCO and LOCTA-IV) have previously been modified for application to a CE NSSS.
Changes were made to the 1978 Evaluation Model(5) to 1) introduce new technology based on analytical techniques that had been approved for the modeling of plants with Upper Head Injection (VHI), and 2) incorporate the necessary features appropriate for the analysis of Large Break LOCA for Westinghouse fuel in a CE NSSS.(6)
The 1978 EM with the "UHI technology" eventually became the 1981 EM, and the modifications for the application to CE NSSS were' approved along with the model improvements for the 1981 EM in the NRC Safety Evaluation Report (SER) which appears in the Proprietary Version of the 1981 EM topical report (5).
The NRC SER for the 1981 Evaluation Model with BART[3]
states that the model, "may be applied to PWR's using Westinghouse fuel and which use only cold leg injection", which encompasses the Fort Calhoun large Break LOCA analysis. However, modifications for the modeling of a CE NSSS have, to date, not been identified and approved for the BASH and LOCBART codes, or for the methodology which constitutes the BMH EH.
Proprietary and Non-proprietary versions of a topical report will be prep red describing the modifications to the codes and methodology of the o
This topical will be provided to the OPPO and the NRC in support of the proposed Evaluation Model, analysis methodology and Fort Calhoun specific Large Break LOCA analysis.
3.1.1 SATAN-VI The analysis of the blowdown portion of the Large Break LOCA transient
-will be performed with the SATAN-VII7) code.
SATAN-VI has previously been. modified for application to a CE NSSS as part of the 1981 EM.
The primary modifications to this code ware 1) the addition of five new elements-(47 - 51) to model the intact cold side piping from the Steam Generator outlet plenum to the vessel '(4 elements), plus the SIT. attached to to the broken loop intact cold leg, and 2) the addition of two elements (52 and 53) to model the CEA shroud flow paths from the Upper Head to the Upper Plenum. A more detailed discussion of these modifications is contained in Reference 6.
]
Westinghouse Class 3 3.1.2 WREFLOOO Prior to the approval of the BASH EM, the MREFLOOD(8) code had been an integral part of the Large Break LOCA Evaluation Model, used for calculating the system thermal-hydraulics for the refill and reflood portions of the large Break transient.
In addition, WREFLOOD was interactively linked to the C0CO code providing reflood mass and energy release data for the calculation of the containment pressure transient.
With the BASH EM, the BASH code replaces WREFLOOD for the calculation of the reflood system hydraulics. WREFLOOO is still used, however, to calculate the system conditions during the refill portion of the transient, and still interactively provides reflood mass and energy rates to the C0CO code.
Two modeling/ methodology modifications were incorporated into EREFLOOD for modeling a CE NSSS as part of the 1981 EM. As with SATAN-VI, elements were added to model the intact leg in the broken loopl (auxiliary loop).
Because of the difference in the injection angle for Safety Injection flow into the steam-filled cold legs' (see Section 2.0), the NRC-specified 0
0 pressure drops of 0.4 and 1.5 psi for injection angles of 60 and 75,
respectively, are used to model the unrecoverable pressure drop due to steam / water mixing in the Reactor Coolant System (RCS). This feature of the CE NSSS design is accomodated in the MREFLOOO code via input. A more detailed discussion of these modifications is contained in Reference 6.
1 The intact leg of the broken loop is often referred to as the
" auxiliary loop' in Westinghouse documentation for CE NSSS analysis.
3
Westinghouse Class 3 3.1.3 [.Qf{.Q The C0C0 Code.is used to calculate the containment pressure transient during the Large Break LOCA.
COC0 uses code input, mass and energy information read from the SATAN-VI tape and information on the reflood mass and energy releases provided interactively from WREFLOOD. No modifications-(other than plant specific input data) are required to apply COCO to a CE NSSS.
3.1.4 SABI The BART(3) code was developed to provide a more mechanistic calculation of core heat transfer coefficients which had previously been determined from the FLECHT correlation (10) employed in the LOCTA-IV[lll code as part of.the 1981 EM. None of the features of the BART code specifically require modification for application to a CE NSSS. With the development of the BART EM, modifications were also added to the WREFLOOD and LOCTA-IV codes for consistency in modeling with BART.
The modeling of metal heat release in the-vessel downcomer and lower plenum was enhanced from a simple conservative exponential decay to a more mechanistic calculation.
The improved version of the WREFLOOD code models the various metal components of the lower plenum as slabs, spheres and cylinders and heat flux to the fluid in these regions is calculated by solving the conduction:
equation based on the modeled geometries.- In addition,-the core mass.
entrainment calculation in WREFLOOO was modified to account for'the effect of boiling below the quench front for compatibility with BART methods.
This version of the WREFLOOD code _is also used in the BASH EM. The BART code-is discussed further in Sections 3.1.5 and 3.1.6.
i
~ _.._
_ -_.__ _ ~__ _...
___._.m Westinghouse Class 3-LOCTA-IV'was"used to perform the rod heat-up calculation in Evaluation Models prior to BASH.
The only LOCTA-IV modifications were those required to read data from BART.
The rod heat-up calculation will be discussed further in Section 3.1.6.
The BART Evaluatien iiodel'was approved for application to PWR's using Westinghouse' fuel.and which use only cold leg injection by NRC SER which is printed-in-the Proprietary version of Reference 3.
This SER restricts the use of the BART~ spacer grid model, pending further-studies (12),
Aspects of the. spacer grid model later rec'eived separate approval [I33, and are included.as.part of the BASH EM.
Further discussion of fuel grid modeling=is contained in Section 3.1.6.
'3.1.5 H M 3
}
The BASH code wasideveloped to provide a more mechanistic analysis of the system thermal-hydraulics.during-the reflood portion of the Lar.ge Break--
LOCA transient' previously calculated by the WREFLOOO code, the~ BASH code is basically a synthesis of two existing codes - NOTRUMPII43 and BART.
NOTRUMP is a nodal: transient general network code which offers a more
~ detailed modeling and greater modeling flexibility than the WREFLOOD code..This flexibility: allows for' an improved modeling of upper plenum :
mass storage, steam generator heat transfer and loop -resistance:during the reflood transient. The non-equilibrium version of the NOTRUMP code (15) is used in the NRC-approved Westinghouse Small Break ECCS Evaluation Model dsing NOTRUMP(16).
The BASH' code is bullt upon1the.
- equilibrium version of this code. - :The NOTRUMP portion constitutes the
~ loop calculation in-the. BASH code, and;is complemented by a modifled version of the BART code which performs the BASH core calculations.-
i
Westinghouse Class 3 i
The BART code, described in Section 3.1.4, was modified for modeling an average core during the reflood transient (versus the hot channel modeling used in rod' heat-up calculations).
The modeling is very similar to that described for the BART code as part of the BART EM, except that thermal-hydraulic and core power conditions are input representing an (core-wide) average channel to model the core reflood.
The BART spacer grid model is conservative neglected in the BASH application.
Because BASH provides a more mechanistic calculation of the reflooding phenomena, it also demonstrates the oscillatory reflooding behavior predicted by a number of semi-scale tests.
The previously described BART model did not have the capacity to handle both positive and negative flows.
Therefore, modifications were incorporated into BART to model and numerically solve equations for reverse core flow during reflood.
The calculation of void propagation velocities and droplet size (droplet number density) were also modified.
Reference 4 provides a detailed description of the BART/NOTRUMP interface.
An additional difference between the BASH methodology and older reflood methodology is directly tied to the core reflooding rate used in the rod heat-up calculations.
Due to the oscillatory behavior of reflooding in BASH (which, if modeled explicitly would provide excellent heat transfer beyond that normal conservatisms of the traditional Appendix K approach),
flooding rates are not input directly into the rod heat-up calculations.
Instead, the oscillatory flooding rate is first undergoes a " smoothing" process which results in a reflooding rate as a representation of step functions used by the rod heat-up calculation.
A more detailed description of the flooding rate smoothing croredure is contained in References 4 and 18.
ll
Westinghouse Class 3 The Proprietary version of Reference 4 contains a copy of the NRC SER approval for the use of the BASH EM for licensing applications.
The conclusions of the SER state, "The staff and its consultants have reviewed the topical report its revisions and appendices. We find that it describes an evaluation model acceptable for the licensing analyses of large break loss-of-coolant accidents in Westinghouse plants." On this basis, additional documentation and NRC review will be required for the application of the BART EM to a CE NSSS.
Such documentation would describe any modifications required to adequately model the NSSS design and fuel design differences between Westinghouse and CE plants, and justification for the acceptability of models not revised in BASH EM for application to a CE NSSS.
3.1.6 LOCBART The LOCBART(17, 6) code replaces the LOCTA-IV code for performing the rod heat-up calculations in the BASH EM for Large Break LOCA.
LOCBART is a synthesis of two previously approved large Break LOCA codes, LOCTA-IV and BART.
The combination of these codes permits a direct exchange of transient fuel rod temperatures (from LOCTA) and heat transfer coefficients (from BART) at each time step.
This is an improvement over the BART EH methodology which required iteration and " hand transfer" of input between the BART and LOCTA-IV codes.
The LOCTA-IV portion of the of LOCBART is the same as that used in the BART EM. The BART portion of the code was modified for the modeling of reverse flow and droplet size as described in Section 3.1.5.
As with BASH, there is currently no licensing documentation supporting the use of the LOCBART code (as part of the BASH EM) for analysis of a CE NSSS.
It is however noted that the components of the LOCBART code had previously been modified for CE NSSS analyses,(6) and had been approved for this application.(2, 3)
\\1
Westinghouse Class 3 4.0 SMALL BREAK LOCA l
The Small Break LOCA analysis for Fort Calhoun will be performed using a i
version of the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code (16) (NOTRUMP EM) as modified for application to a CE NSSSil9).
The NOTRUMP EM will be used to demonstrate compliance with the requirements for the analysis of Small Break Loss-Of-Coolant accidents as defined in 10 CFR 50.46(Il, and employs the required models-and assumptions identified in 10 CFR 50, Appendix K(ll.
Specifically, the NOTRUMP EM was developed to address the requirements of NUREG-0737, Item-
!!.K.3.30.
The NOTRUMP EM consists of the NOTRUMP and LOCTA-IV codes.
These codes have previously been modified for application to a CE NSSS. Changes.were made to the NOTRUMP EM to incorporate the necessary features appropriate for the analysis of Large Break LOCA for Westinghouse fuel in a CE NSSS, and to address additional Small Break licensing issues not addressed in the Westinghouse version of.the NOTRUMP EM.
NUREG-0611 addressed NRC-concerns regarding Westinghouse NSSS design and the associated Small Break LOCA analysis methods, for which responses were provided in References 15 and 16.
The NRC SER on the NOTRUMP code and EM (contained in the Proprietary versions of References 15 and 16, respectively) evaluated the those' responses as being adequate.- NUREG-0635 addresses concerns similar to those identified'in NUREG-0611, but is specific to the CE_NSSS design and the associated Small Break LOCA analysis methods.
The additional concerns of NUREG-0635, not identified in NUREG-0611 are addressed specifically for.the NOTRUMP~EM applied to a CE-NSSS.
The responses to those additional-concerns, plus the NRC SER identifying the adequacy of the response are contained in Reference 19.
The NOTRUMP EM, as modified in Reference 19, for the application to CE NSSS has previously been used to perform the licensing basis Small Break LOCA calculation for Millstone Unit 2, demonstrating compliance with 10 CFR 50.46 and NUREG-0737, item II.K.3.31.
O
Westinghouse Class 3 4.1.2 NOTRUMP The NOTRUMP code is a nodal transient general network code used to model the system thermal-hydraulic transient during the postulated Small Break LOCA.
The NOTRUMP code has previously been modified for application to a CE NSSS.
The primary modifications to this code and methodology are identified below.
- 1) The system modeling was modified to accurately represent the CE two hot leg, four cold leg design. The NOTRUMP code uses a generalized input description and thus code modifications were not required to enable modeling of the CE piping geometry, Verification was also provided that the correlations and modeling techniques of the NOTRUMP code would be adequate and applicable for the larger diameter CE hot leg piping. This value is introduced to the code via input and requires no code modification.
- 2) The difference in Steam Generator inlet nozzle inclination was addressed and appropriate sensitivity studies were performed.
Based on comparisons of the flooding constants in the flooding correlation for the two designs, plus the results of the sensitivity studies performed, an appropriate modeling of the CE Steam Generator inlet plenum was defined.
l
\\d
Westinghouse Class 3
- 3) The hydraulic behavior of the loop seals during a Small Break LOCA are discussed in Reference 19.
Based on this discussion, it is verified that the noding scheme chosen for the analysis, in combination with geometric input representative of a CE NSSS (loop seal design) would adequately model the CE loop seal behavior.
- 4) The modeling of the flow paths from the Upper Head to the Upper Plenum and geometry of the CEA shrouds are examined in Reference 19.
Sensitivity studies were performed to verify that the noding scheme chosen for the analysis, in combination with geometric input representative of a CE NSSS would conservatively model the Upper Head ventir,g during a Small Break LOCA transient.
- 5) The adequacy of the modeling of the Upper Head bypass flow in the chosen noding scheme is addressed via comparison of the hydraulic resistances of the bypass flow paths in a CE and Westinghouse design.
This comparison serves as the basis for the accepted modeling technique and choice of Upper Head fluid temperature.
- 6) Despite differences in the fuel rod diameter, the steam cooling correlations used in the NOTRUMP code are applicable over a range encompassing both the CE and Westinghouse fuel rod design.
Accurate implementation of these correlations for application to the Westinghouse-designed fuel for a CE NSSS is included in the NOTRUMP EH modeling and methodology described in Reference 19,
- 7) The angle between the cold leg and the safety injection nozzle differ between Westinghouse and CE NSSS designs.
Since the angle of the Safety injection jet influences the amount of steam condensation in the loops, the choice of a safety injection angle for the analysis of a CE NSSS is such that loop steam condensation is conservatively modeled.
/f
=-.
Westinghouse Class 3
- 8) An evaluation of the behavior of the subcooled safety injection (SI) for a CE NSSS design is expected to be similar to that in a Westinghouse NSSS design.
Therefore, no modification to the NOTRUMP code was required to accommodate the CE SI geometry.
- 9) Examination of of data from accumulator (SIT) blowdown tests were used to determine a polytropic expansion coefficient for CE SITS with cover gas pressures of approximately 200 or 600 psig.
Based on this examination, a polytropic expansion coefficient was chosen which would be applicable to either of the CE SIT conditions.
- 10) Based on a comparison of the Reactor Coolant Pump (RCP) design, and comparison with applicable test data, an appropriate pump model was identified for use in NOTRUMP for the analysis of a CE NSSS, as documented in Reference 19.
A more detailed discussion of these modifications is contained in Referenco 19.
4.1.2 LOCTA-IV' The LOCTA-IV code is used to perform the rod heat-up calculation for the Small Break LOCA' analysis. The LOCTA-IV code is the same code identified
-in Reference 11, but has been modified slightly to allow consistency with the NOTRUMP output format.
The version of LOCTA-IV used in the NOTRUMP EM has also been modified to enhance the modeling of steam cooling and
-radiation in the rod heat-up calculations.
Discussions of.these modifications are contained in Reference 16.-
The primary modifications to this code-and methodology'~are identified below.
it
l Westinghouse Class 3
- 1) Despite differences in the fuel rod diameter, the steam cooling correlations used in the LOCTA-IV code are applicable over a range encompassing both the CE and Westinghouse fuel rod design.
Accurate implementation of these correlations for application to the Westinghouse-designed fuel for a CE NSSS is included in the NOTRUMP EM modeling and methodology described in Reference 19.
- 2) Fuel rod internal crack and dish volumes, normally calculated internal to the code, were replaced with values representative of a CE fuel design, rather than the standard Westinghouse values.
17
Westinghouse Class 3
5.0 REFERENCES
l l
1.
" Acceptance' Criteria-for Emergency Core Cooling Systems for Light
]
Water Cooled Nuclear Power Reactors,"10 CFR 50.46 and Appendix K of f
10 CFR.50, Federal Register, Volume 39, Number 3, January 4, 1974, j
l
-2.
" WESTINGHOUSE ECCS EVALVATION MODEL 1981 Version," WCAP-9220-P-A, Rev.'I (Proprietary) and WCAP-9221-A, Rev.'I (Non-Proprietary),
February 1982..
i 1
\\
~3.
Young, M. Y., et al., "BART-A1: A Computer Code for the Best Estimate
- Analysis of Reflood Transients," WCAP-9561-P-A (Proprietary) and WCAP-9565-A (Non-Proprietary), March 1984.
4.- Besspiata,-J.
J., et al., "The-1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A, Res. 2
'(Proprietary), March _1987.
- 5..B'rdelor, F. M., et. al., " SATAN-V1 Program: Comprehensive. Space-Time o
s
-Dependent' Analysis of loss of Coolant," WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974.
6.
" Westinghouse ECCS' Evaluation Model February 1978. Version,"
WCAP-9220-P-A'(Proprietary) and WCAP-9221-A (Non-Proprietary),
February.1978, 17.
Ferguson, K. L., and R. M. Kemper, "ECCS Evaluation Model for Westinghouse Fuel Reloads of Combustion Engineering NSSS," WCAP-9528
-(Proprietary), June 1979.-
8.
Kelly, R. 'D., et. al;, " Calculation Model for Core. Reflooding After A Loss-Of-Coolant Accident (WREFLOOO Code),_" WCAP-8170 (Proprietary) and WCAP-8171.(Non-Proprietary), June 1974.
lb
__________________.____._______.__________m,
Westinghouse Class 3 9.
Bordelon, F. M. and E. T. Murphy, " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary),
-June 1974.
10.- 10CFR50 Appendix K Section 1.0.3 11.
Bordelon, F. M., et, al., "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.
12.
Chiou, J. S. et. al., " SPACER GRID HEAT TRANSFER EFFECTS OURING REFLOOD," WCAP-10484 (Proprietary) and WCAP-10485 (Non-Proprietary),
March 1986.
13.
Letter from Cecil 0. Thomas (NRC) to E. P. Rahe (Westinghouse),
" Acceptance for Referencing of Licensing Topical Report WCAP-10484(P)/10485(NP), " Spacer Grid Heat Transfer Effects During Re flood. "
14.
Meyer, P. E., et. al., "NOTRUMP: A Nodal Transient General Network Code," WCAP-10076 (Proprietary), March 1982,
- 15..Meyer, P. E...et. al..."NOTRUMP: A Nodal Transient Small 8reak and General: Network Code,"~ WCAP-10079-P-A (Proprietary) and WCAP-10080-NP-A (Non-Proprietary), August 1985.
-16.
Lee, N.,
et, al.
" Westinghouse Small Break.ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054kP-A (Proprietary) and WCAP-10081-NP-A (Non-Proprietary), August 1985.
19.. Bajorek, S._M.
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