ML20039D519
| ML20039D519 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/27/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20039D514 | List: |
| References | |
| NUDOCS 8201050104 | |
| Download: ML20039D519 (10) | |
Text
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NUCLEAR REGULATORY COMMISSION j
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. DPR-28 VERMONT YANKEE NUJLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 1.0 Introduction By letter dated September 2,1981 (Reference 1) Vermont Yankee Nuclear Power Corporation (VYNPC or licensee) has proposed changes to the Technical Specifications of the Vermont Yankee Nuclear Power Station (VY), as supported by Reference 2.
The proposed changes relate to the core for Cycle 9 operation. Although the Cycle.9 reload involves the replacement of a number of irradiated fuel assemblies, the application does not involve physical changes to the fuel system design.
The safety analysis of Cycle 9 reload does, however, involve the first-time application of a number of analytical methods. These methods are described and supported in references 3-5, 9,13-15,17,19, 22-24, 26 and 27.
2.0 Evaluation 2.1 Fuel 2.1.1 Fuel Mechanical Design The Vermont Yankee Cycle 9 reload involves the insertion of 120 new fuel ~
bundles of the pressurized retrofit 8x8 design. These assemblies will be irradiated with a number of other prcssurized retrofit 8x8, unpressurized retrofit 8x8, and unpressurized standard 8x8 fuel assemblies already resident in the core. All fuel assemblies were fabricated by the General Electric Company. These three interchangeable fuel designs.have been approved for the previous cycle of operation at Vermont Yankee as well as at other boiling water reactors (BWRs). Cycle 9 involves no physical changes to the fuel design and is, therefore, acceptable.
2.1.2 Fuel Thermal Design The Cycle 9 fuel thermal performance analysis was performed with a new Yankee Atomic. computer code called FROSSTEY (refs. 3-5).
This code was used-to calculdte (a) incipient fuel centerline melt limits, (b) 1 percent cladding strain limits, (c) core average gap conductance and core average fuel temperature for initializing non-LOCA transient analyses, and (d) average gap conductance and average fuel temperature of the p.eak bundle for initializing hot channel calculations. Previous analyses provided by General Electric were used for the remainder of the fuel thermal analysis, including LOCA initial conditions.
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2 Our review of the.FR0SSTEY code is not yet complete.
Because this code was used te only a limited extent in the Vermont Yankee Cycle 9 analysis, we have reviewed the specific Cycle 9 fuel conditions predicted by the FROSSTEY code, rather than the generic methods employed by the code to predict thase conditions.
This has allowed us to make a finding on the Cycle 9 submittal without completing our review of the generic methods to berempicyed in later cycles.
We have compared local fuel integrity limits-calculated with FROSSTEY with those reported (ref. 6) by General Electric for the same~ conditions and find them to be similar, with slightly higher centerline melt and slightly lower cladding strain limits' being reported by Yankee Atomic.
Based on the similarity of the Yankee Atomic and General Electric (previously approved) results, and on the fact that these fuel integrity conditions are not limiting for Cycle 9 operation, we find these results acceptable.
The FROSSTEY code was.also used to calculate core average and hot channel average gap conductance and fuel temperature. We have audited these calculations using the results of an NRC fuel performance code called GAPCON-THERMAL-2 (ref. 7) as reported in NUREG-0559 (ref. 8).
The results of our calculations are very similar to those predicted by the FROSSTEY code, our code predicting slightly lower fuel temperatures and a slightly greater variation. in gap conductance as a function of burnup.
As a result, we conclude that the average gap conductance and fuel temperature predictions used in the Cycle 9 analysis are acceptable.
Because our calculations were based on Cycle 9 specific conditions, our approval does not apply to subsequent cycles of operation at Vermont Yankee.
Other fuel performance analyses used in the Vermont Yankee Cycle 9 Reload l
Report rely.on General Electric gener'ic analyses. These include fuel rechanical design, maximum linear heat generation rate (MLHGR, which remains at 13.4 KW/ft) and LOCA initial conditions.
These analyses are unchanged from those previously accepted for Vermont Yankee with the exception of the maximum average planar linear heat generation rate (MAPLHGR) operating limits.
The Cycle 9 MAPLHGR limits, which are documented in Reference 9, have been extended to 40,000 mwd /t for_some fuel types.
The General Electric Company has requested (refs.10-11) that credit for approved, but unapplied, ECCS evaluation model changes be used to avoid MAPLHGR penalties on operating reactors due to high burnup fission. gas release. Such penalities would apply to Vermont Yankee for burnups beyond 30,000 mwd /t.
However, we have conditionally accepted (ref.12) the GE proposal and, by letter of November 9,1981 -(ref.13), the licensee has subscribed to the conditions of our approval. We, therefore, find the Vermont Yankee Cycle 9 MAPLHGR limits acceptable as submitted.
2.1.3 Conclusions The NRC staff has reviewed those sections of the reload report for Vermont Yankee Cycle 9 dealing with the fuel system design and we find those portions of the application acceptable.
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3 2.2 Physics Considerations The nuclear design analysis of the core was performed with the SIMULATE.
code (ref.14) with input parameters calculated with the CASMO code (ref.15). VYNPC has provided reports which describe these codes and the analyses carried out to verify them for use..These reports are currently in review by the staff and evaluation reports will be issued.
Sufficient review has been completed to permit the conclusion that they are acceptable for use in the analysis of the Vennont Yankee Plant. This conclusion is based on comparisons of the results of calculations using these codes with measured data including that obtained from the first seven cycles of the Vermont Yankee reactor. These comparisons showed that power distributions, hot and cold reactivities, shutdown margins and reactivity parameters were calculated to within accuracies and precisions comparable 3
to those of other codes and techniques employed in the industry.
The licensee has provided a description of the core loading for Cycle 9 as well as analyses of anticipated power distributions, end of cycle exposure distributions, shutdown margin -values and cycle kinetics parameters.
Burnup calculations have been performed for both rodded and unrodded depletions. The end of cycle power distribution shows acceptable peaking factors for both rodded and unrodded depletion. The minimum shutdown margin during the cycle was calculated to'be 0.78 percent ak, an acceptable value. The cycle kinetics parameters are similai.to, those for earlier cycles and are acceptable.
We have reviewed the analyses of the rod withdrawal error, fuel misloading event and the rod drop accident. The analysis procedures employed by VYNPC are the same as those in current use for other operating boiling water reactors and are acceptable.
A bounding analysis of the rod withdrawal error is performed. A fully inserted high worth rod is assumed to be withdrawn continuously.
An assembly near the withdrawn rod is assumed to be operating on Technical Specification limits at the time of the withdrawal.
For the analysis, the maximum number of LPRMs (which make up the inputs to the Rod Block-Monitor) permitted by the Technical Specification is assumed to be inoperable. The response of the Rod Block Monitor is then calculated as a function of the distance the rod is withdrawn. When the rod block setpoint is reached, the rod is assumed to travel an additional two inches and then'to stop at the next notch.
The resulting change in CPR is tnen added to the safety limit value and the required MCPR operating l
limit for this event is obtained. For Cycle 9 of Vermont Yankee this value (1.29) is the most limiting value of this quantity (for the " measured" scram time) and establishes the operating MCPR value for the cycle. We conclude that th'e analysis of the rod withdrawal error is acceptable.
Two types of fuel misloading events are analyzed - misorienting an assembly in its proper location and mislocating a properly oriented bunale.
In the first~ (misoriented) event the bundle may be rotated by 900 or 1800 from its normal orientation. The worst case is chosen and the increase in linear heat generation rate (due to the presence of higher enrichnient rods near the wide water gap) and the decrease in critical power ratio w-
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(due to the effect of the tilting of the assembly and the change in local power distribution on the R factor) is determined for a large number of core locations. The limiting case is chosen and the operating MCPR limit required to prevent violating the MCPR safety limit when the misoriented bundle is placed on operating limits is obtained.
For Cycle 9 this value is 1.24 which is smaller than that required for the rod withdrawal event. The resulting linear heat generation limit is 17.5 kw/ft which is less than the 1 percent strain limit. We conclude that the analysis of the assembly misoriented event is acceptable.
The analysis of the mislocated bundle follows procedures used for other operating boil.ing water reactors and is acceptable. The procedure starts by substituting the higher enrichment reload bundle for various high burnup bundles throughout the core to obtain the highest change in CPR produced by the substitution. The ACPR is then added to the CPR values of all the bundles in the core at several times in the cycle:
Some bundles in the core (the least reactive ones) will not violate the MCPR safety limit when the core is at full power. These are dropped from consideration. The procedure is repeated with the high enrichment bundle being substituted for the least reactive of the remaining bundles.
Because these bundles have higher reactivity than the first group, the resulting ACPR will be smaller. The addition procedure is then repeated, additional bundles are dropped from consideration and the whole process is repeated. This iteration continues until all locations are shown to be above the MCPR safety limit or until a limiting location is identified.
For Cycle 9 all locations are shown to be above the safety limit (1.07) if the operating MCPR limit is 1.24.
We conclude that the analysis of this event is acceptable.
The analysis procedure for the rod drop accident is the same as that used for other boiling water reactors.and is acceptable. The Vermont Yankee reactor employs the banked position withdrawal sequence which is enforced with the Rod Worth Minimizer.
This sequence was examined for Cycle 9 and the maximum worth for a potential dropped rod was 0.86 percent.
The generic analysis perfcrmed by General Electric for this event (which is applicable to the Vsrmont Yankee reactor) would yield a maximum fuel enthalpy of about 140 calories per gram for this rod worth -
an acceptable value. We conclude that the analysis of this event is acceptable.
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2.3 System Analysis The licensee has provided its own analysis for Vermont Yankee cycle 9, independent of some parts of the GE modeling package, but making use of other parts.
In particular, YAEC has developed a methodology for the analysis of Verfnont Yankee full core transients making use of the RETRAN code in place of the ODYN code.
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Ample comparisons of RETRAN with experimental results exist (ref.16).
The NRC staff regards RETRAN as a substantially correct coupling of thermal hydraulics and neutronics packages but will require more experience with its use before approving it generically (expected about September 1982).
In the meantime the use of RETRAN in reload transient analyses is being reviewed on a case-by-case basis.
4 We have reviewed the use of RETRAN by VYNPC in Vermont Yankee Cycle 9 analysis to determine if the results and theory are sufficiently conservative-when compared with experimental data and with previously accepted methods.
1.
We have compared experimental results from the Peach Bottom. turbine trip without bypass test and calculations using the Vermont Yankee methodology (ref.17). The comparisons are close enough to produce a reasonable degree of confidence in the ability of the. Vermont Yankee methods to reproduce this sort of transi.ents and are best at the highest of the test powers (2275 MWt).
2.
We have compared experimental results from a generator load rejection test performed at Vermont Yankee with calculations of the same conservative. g RETRAN. The everall effect on core power appears to be transient usin 3.
We have considered comparisons based on a transient which simulates a turbine trip from full power without bypass flow, modeled on the Peach Bo'ttom Unit 2 (ref.18). The licensee has calculated this transient using Vermont Yankee methods (ref.19) and compared the result with ODYN and BNL-TWIGL calculations.
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4.
We have compared Cycle 9 ACPR values with Cycle 8 aCPR values (ref. 20) which were calculated using the REDY code (no longer i
approved).
The Cycle 9 aCPR values are more conservative than the.
corresponding calculated Cycle 8 values.
5.
We have considered the results of comparisons between calculations-t using RETRAN and RELAP4/ MOD 6.
In gen.eral, good agreement between the RELAP4/ MOD 6 results and VYNPC' results using RETRAN 01/ Cycle 15G and ANL results using RETRAN (ref. 21) was demonstrated.
In the course of this comparison, some of the steam line and steam dome modeling effects were studied.
The passage of the pressure waves along the steam line is clearly apparent from the RETRAN calculation, indicating an acceptable nodalization and computation of compressibility effects. The nonequilibrium model of the steam dome chosen in the VYNPC calculation was also determined to be more conservative than an equilibrium model.-
The theoretical' basis for incorporating three-dimensional physics effects in the RETRAN point kinetics model (ref. 22) has been reviewed. The VYNPC calculation's of the steady state core physics employ " state-of-the-art" methods for the 3D physics and for the collapse to equivalent.lD representa-tion and then to a "0" D' point kinetics representation for use.in the y
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RETRAN transient analysis.
Their use of the collapsed representation'is quasi-lD, not truly 1D as a comparable ODYN calculation would be, nor is it simply point kinetics as the old REDY calculation was.
The comparison of the Peach Bottom tests results with the VYNPC calculations, mentioned above, indicates the. conservative nature of the VYNPC methods.
The VYNPC method includes an implicit assumption that the set of steady state analyses adequately covers all possible transient conditions.
Because of the strong feedback from-the moderator density (boiling void) distribution and the changes that local voids can undergo during a transient, this general area requires further examination in the long term.
VYNPC has not missed something that others have included, however, as this is potentially a problem that ' Jst also be faced by GE. VYNPC has provided assurances that for the tri.csients analyzed in the Cycle 9 submittal, the -
snapshots used in their 3D calculations are representative of each specific transient analyzed.
In comparison to the Peach Bottom tests, VYNPC methods produce results that are more conservative than those of ODYN and agree well with the data.
VYNPC has not at this time submitted 'ACPR values from their Peach Bottom transient calculations. There are difficulties in producing a ACPR calculation for Peach Bottom using the VYNPC models, but a calculation df this value is being considered by VYNPC. The NRC stat
- believes that this could be a useful comparison with the GE calculations and should be pursued.
The licensee has proposed to use the measured scram as a basis in the Technical Specifications for determining Operating Limit MCPR's.
Technical Specifications require that these times are to be measured periodica.lly.
Sensitivity studies of the effect of scram time on ACPR have been included in their analysis supporting these Technical Specification changes. We find the methodology, as described above, and the proposed Technical Specification changes acceptable.
2.4 Core Thermal Hydraulics In support of the Cycle 9 reload application, the licensee has submitted YAEC-1273, Vermont Yankee Cycle 9 Core Performance Analysis (ref. 2), which utilizes results from new core thermal hydraulic computer codes. These codes are FIBWR (YAEC-1234, ref. 23) for steady-state core flow distribution calculations and MAYUO4-YAEC (YAEC-1235, Ref. 24) for rod bundle transient thermal hydraulic calculations.
In addition, the EPRI void model is used for the two-phase void fraction calculation. The GEXL ' critical quility-boiling length correlation is used for critical power calculation. The staff's review evaluations are described as follows.
FIBWR Code FIBWR is a steady-state th'ermal hydraulic analysis code which determines the flow and void distributions for a given power distribution and inlet flow conditions in a BWR core.
The staff has reviewed the FIBWR code and concluded it is acceptable for Vermont Yankee reload analyses.
The rev.iew of FIBWR will be addressed in a separate SER.
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'MAYUO4-YAEC Code MAYUO4-YAEC -(ref. 24) is a mcdified version of MAYU04 (ref. 25) which 4
analyzes one-dimensional single channel hydraulic and heat' transfer transients in rod bundles. The modifications made to the original
'MAYU04' include the use of the EPRI void model-' and-the GEXL' critical quality-boiling length correlation.
MAYUO4-YAEC is used to calculate _
hot channel thermal margins under transient conditions with the transient input provided by the RETRAN system response' analysis.
MAYU04-YAEC coupled with the GEXL correlation performs critical power ratio.calcu- -
lation for the reactor transients.
The staff-has reviewed the MAYUO4-YAEC code and requested the licensee to provide comparisons-using existing -transient ATLAS 4x4 data and MAYU04-YAEC with the EPRI-void model and GEXL correlation. For,these comparisons the code predicted poorly with data in the high void fraction
' range. The licensee.has -determined -that more work has.to be done.to identify the problem and fix1the code. -In the interim', the staff has concluded that MAYU0-4YAEC is not acceptable _for thermal margin-analysis and has informed the licensee that core wide transients using an acceptable code should be submitted by March 31,_1982.
In order to support reactor, operation without relying on the results'of
' MAYU04-YAEC, the licensee by letter dated November 23,1981 -(ref. 26) demonstrated that the most limiting transient prior to E0C-2000 MWD /t is the local control rod withdrawal error transient. A~comparisgn'ismade-forthe-ACPR values between rod withdrawal error and loss of 100 F feedwater -
heating transient for fuel burnup from B0C to E0C-2000: MWD /t for the previous
. Cycles 6, 7,-and 8.
In all cases,'the ACPR's -for RWE with rod block monitoring (RBM) trip setpoint "N" values of 41 and 42 percent are always greater than the loss of feedwater heating transient.
Since the loss of t
'feedwater heating transient is the most limiting among the core wi'de
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transients prior to E0C-2000 MWD /t, this comparison shows that RWE is the most limiting transient for the previous cycles.
Since Cycle 9 fuel.
loading is similar to the Cycle 8, it.is reasonable to assume that the t
RWE and RBM setpoint N value no less-than 41 percent is the most limiting
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transient prior 'to ~ E0C-2000 MWD /t in Cycle 9.
The RWE transient analys'is is performed with the three-dimensional -SIMULATE-code (ref.14) using the GEXL correlation in a quasi-steady state approach.
The review of the SIMULATE code is not complete, but has progressed ~
sufficiently. to conclude that SIMULATE is acceptable for Vermont Yankee MCPR prediction, the licensee.has provided a comparison (y'of the SIMU Cycle 9 reload analysis.
In order to-justify the validit ref. 27) between SIMULATE And FIBWR calculations on the core flow distribution in various bundles. The results show excellent agreement between the two codes in
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bundle flow predictions. This' demonstrates the-adequacy of the SIMULATE code in predicting the. thermal h' draulic conditions which are used in y
calculating'MCPR's associated with the RWE transient. Therefore, the ACPR calculated by SIMULATE for the RWE transient is acceptable.
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We have issued Technical Specifications which require that until further NRC approval is obtained, an RBM value of "N" greater than or equal to 41 percent shall be used,- and operation shall only be allowed until E0C-2 GWD/t.
Based on the above observations, the staff concludes that the operating limit MCPR based on the RWE transient analysis with RBM trip setpoint "N" value no less than 41 percent is acceptable until E0D-2000 MWD /t.
3.0 Environmental Considerations We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determina-tion, we have further concluded that the amendment involves an action which-is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR Section 51.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of the amendment.
4.'0 Conclusion We have concluded, based on the considerations discussed-above, that: (1) because the amendment does not involve a significant increase in the probability or consequences.of accidents previously considered.and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards ' consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and sec~urity or to the health and safety of the public.
Dated: November 27, 1981 i
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REFERENCES
- 1. - L. H. Heider (VYNPC) letter to -the Office of Nuclear Reactor Regulation (NRC) dated September 2,1981.
2.
" Vermont Yankee Cycle 9 Core Performance Analysis," Yankee Atomic Electric Company Report YAEC-1275, August 1981.
3.
S.- P. Schultz and K. E. St. John, " Methods for the Analysis of 0xide Fuel Rod Steady-State Thermal Effects (FROSSTEY) Code /Model Description Manual," Yankee Atomic Electric Company Report YAEC-1249, April 1981.
4.
S. P. Schultz and K. E. St. John, " Methods for the Analysis of 0xide Fuel Rod Steady-State Thermal Effects (FROSSTEY) code Qualifi-cation and Application," Yankee Atomic Electric Company Report YAEC-1265, June 1981.
5.
D. C. Albright, "H20DA: An Improved Water Properties' Package,"
Yankee Atomic Electric Company & port YAEC-1237, March 1981.
6.
" Generic Reload Fuel Application," General Electric Company Report NEDE-24011-P-A, July 1981.
7.
C. E. Beyer et al, "GAPCON-THERMAL-2: A Computer' Program for Calculating the Thermal Behavior of an Oxide-Fuel Road,
Battelle Pacific Northwest Laboratories Report BNWL-1898, November 1975.
8.
D. L. Acey and J. C. Voglewede, "A Comparative Analysis of LWR Fuel Designs," U.S. Nuclear Regulatory Commission Report NUREG-0559, July 1980.
9.
L. H. Heider (VYNPC) letter to Office of Nuclear Reactor Regulation (NRC) dated November 13, 1981.
10' R. E. Engel (GE) letter to T. A. Ippolito (NRC) dated May 6,1981.
11.
R. E. Engel (GE) letter to T. A. Ippolitio (NRC) dated May 28, 1981.
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12.
L. S. Rubenstein (NRC) memorandum to T. M. Novak (NRC) on
" Extension of General Electric Emergency Core Cooling Systems Performance Limits" dated June 25, 1981.
13.
R. L. Smith -(VYNPC) to T. A. Ippolito, " Additional Information on the Extension of Emergency'Co're Cooling System Performance Limits for Vermont Yankee," dated November 19, 1981.
14.
D. M. Ver Piahck, " Methods for the Analysis of Boiling Water Reactor Steady State Core Physics," Yankee Atomic Electric Company Report YAEC-1238, March 26,1981.
15.
E. E. Pilat, '.' Methods for Analyses of Boiling Water Reactor's Lattice Physics," Yankee Atomic Electric Company Report YAEC-1232, December 31, 1980.
16.
J. H. McFadden, et al, RETRAN Computer Code Manu !, EPRI CCM - 5 Volume 4, Energy Incorporated, December 1978.
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2 17.
A. Ansari, et al, Methods for the Analysis of Boiling Water Reactors, A Systems Transient Analysis Model (RETRAN), YAEC-1233, April 1981.
18.
M. S. Lu, et al, Analysis of Licensing Basis 1 Transients for a BWR/4, BNL - NUREG-26684, September 1979.
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19.
J. T. Cronin, et al, Yankee Atomic Boiling Water Reactor Analysis Methods: Analysis of a Typical BWR/4 Turbine Trip Without Bypass Transient, YAEC-1280, October 1981.
- 20. Supplemental Reload Licensing Submittal for Vermont Yankee Nuclear Power Station - Reload #7,' General Electric Company Y1003J01A02, July 1980.
- 21. Letter, P. Abramson, Argonne National Laboratory, to T. P Speis, NRC,
Subject:
Review of Vermont Yankee Cycle 9 Reload Analysis, with enclosure.
22.
J. M. Holzer, et al, Methods for the Analysis of Boiling Water Reactors Transient Core Physics YAEC-1239P, August 1981 (proprietary).
YAEC-1234, " Methods for the Analysis of Boiling) Water Reactors: Steady-State Core
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23.
Smith to U.S. NRC dated December 31, 1980.
- 24. YAEC-1235, " Methods for the Analysis of Boiling Water Reactors:
Transient Thermal Margin Analysis Code (MAYUO4-YAEC)," dated December 1980.
25.
W. C. Punches, "MAYU04: A Method to Evaluate Transient Thermal Hydraulic Conditions in Rod Bundles," GEAP-23517 dated March 1977.
26.
Letter from D. E. Vandenburgh to U.S. NRC, " Justification for the Vermont Yankee MCPR Operating Limits, BOC to E0C-200 MWD /t," dated November 23, 1981.
- 27. Letter from D. E. Vandenburgh (VY) to T. A. Ippolito (NRC),
" Validation of SIMULATE Thermal Hydraulics and Thermal Margin.
Calculations by Comparison to the FIBWR Code," dated November 23, 1981.
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