ML20005A957
| ML20005A957 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/12/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20005A955 | List: |
| References | |
| TAC-43380, NUDOCS 8107060108 | |
| Download: ML20005A957 (7) | |
Text
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Er iu ur e P ysics section Safety Evoluoliun
- ,,;,1c Sc e y Recort a
- illstor,e, Unit !;o. 2 Oc-kat i
- c. 50-335 1.0 SU1!!ARY OF REPOP.T (Physics)
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This report describes the reactor physics methods used by 'destinghouse for the performance analysis of the,lillstone Unit 2 reactor bui'. by Combustion Engineering (CE). These methous will be used for the nuclear design of Westingnouse reload fuel for Millstone 2 beginning with Cycie 4.
The report addresses the reactor model description, the calculationai oscel verification, and the application of the physics methods to both operating reactor conditions ind to reload safety evaluations.
E The calculational methods used to analyze the nuclear design of ;;illstone i.
consist of standard Westinghouse nuclear design procedures, codified to r
accommodate the differences between Millstor.e 2 and Westinghouse PWR cores. A description of this revised design procedure is given in the rep' ort with the B
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differences from standard Westinghouse design procedures emphasized. The following physics related computer codes are used:
1.
LEOPARD-CINDER, linked spectral codes which are used to obtain burnup dependent neutron cross sections for fuel cells.
HAIC!ER-AIM, linked spectral codes which are used to obtain neutron cross 2.
sections for the CEA rods.
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TURTLE, a three-dimensional neutron diffusion-depletion code used to obtain power distributions, fuel depletion, critical boron concentrations, h
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xenon distributions, reactivity coefficients, and control rod worths.
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. a ut;u-dia.e:mim:ll diffu-icn th00ry code ita 16eal feed 33CK 4.
FA wA, is used to oDrain axial pc.ier distributient, differential control wnicn rod w:rths, and ?xial.venon distributions.
a cwo-dimensional nodal code which is used te obtain power 5
P AL ADON,
distributions, fuel depletion, ' critical boron concentrations, reactivity coefficients, and control rod worths.
The Westingnouse standara nuclear design methods that have been adepted to i'ied oy benchmarking against Mills One 2 the Millstone 2 reactor are vec
=easurements cver the first three cycles of operation and by comparisons with higher order analysis. An addendum to the BSR (Ref.1) describes the power peaking factor uncertainty analysis utilized in the nuclear design of Millstone 2 and is also based on measured data from the first three cycles The following physics parrteters are addressed:
of operation.
- 1. : Control Rod worth comparison to_ measurement, isothermal temperature coefficient comparison to measurement, 2.
power distribution comparisons to maasureme it, and 3
critical boron concentration comparisons to measure >aent.
4.
For each parameter addressed the data base is presented, including comparisons between calculations and measurements, and conclusions are drawn regarding the suitability of the model to perform the calculations.
In addition to the calculational methods used, a description of the power distribution control phi'osophy adopted in Cycle 4, called felaxed Axial Offset Control (RAOC), is described.
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.Tne licensee > anaiysis of accican.s is provid:d in th:
demanstrate that Millstone 2 safety criteria are 2.Lisi'ed..hin ths :: :
is relced:d -ith Westinghouse fuel, and to establish a reference safety analysis for f"ture reloads.
2.0
SUMMARY
OF REVIEW (Physics) 2.1 Nuclear Design We have reviewed the information presented with regard to calculational methods and comparisons of calculations and experiment. Most of the pro-cedures are standard Westinghouse methods which have been used previously The and verified against critical experiments and Westinghouse cores.
slight modifications in these procedures due to differences between Mill-stone 2 and Westinghouse cores have been adequately described in the B5R.
Many of the computer codes used are acceptable industry-wide codes and, therefore, require no additional review. These 1'nclude the"LEOP'ARD, CINDER, HAMMER, and AIM codes which form the neutron cross section generator.
The TURTLE, PA"DA, AND PALADON codes have previously been reviewed and approved by the staff.
We have reviewed the comparison of predicted reactivity coefficients (moderator temperature, Doppler, and boron) with measured values from the first tnree cycles uf Millstone 2 and conclude that the Westinghouse model adequately predicts reactivity coefficients for the expected range, of operating a
conditions and burnups.
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"a nave reviewed the comparison of predicted enntrol rod wortns with measured values for Cycles 1, 2, and 3 and conclude that the cradel adequately predicts total and differential worths and trip reactivity and confirms the shutdown margin calculations.
We have reviewed.the comparison of predicteo power distributions with ceasured Cycle 1, 2, and 3 values and conclude that the method 1dequately predicts ra' dial power distributions and peak-to-average distributions for beginning, middle, and end-of-cycle conditions.
Comparisons of power peaking in fuel pins adjacent to CEA water holes using TURTLE (diffusion theory) and KENO (Monte Carlo) have shown an underprediction oy diffusion theory, as expected.- Due to the unavailability of experimental results on water hole peaking factors, the maximum bias was confirmed by comparisci;s of TURTLE and INCA results for Cycle 1, 2, and 3 (Ref. 2).
We find this water hole peaking correction to be acceptable.
The power distribution control philosophy to be used in iiillstone 2 in Cycle 4 and beyond is Relaxed Axial Offset Control (RAOC) which is similar to the Based on the information presented procedure used for Cycle 3 in most respects.
in the BSR and additional discussions with UNECO and Westinghouse, we find the RAOC procedure acceptable for providing power distribution control limits for Millstone 2 operation.
in addition, th'e Reload Safety Evaluation Report submitted by UNECO for Cycle 4 operation added to the data base for comparison of calculated and measured physics parameters and further verifies the nuclear design methods.
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m;;;0 h. s sub-itteM cn ?>denium to the BSR (pof. 1) which docerihas the pc cr wealing Io;;0r uncertainty cnGlysis us?d 10 t"? nucle 3r d*si7n of Millstone 2 beginning with cycle 4 operation.
The analysis uses measured data fe o.a the first 3 cycles and accounts for the errer in the Feurier fit for tne axial power shape used by INCA and includes a ccrrectica for three-dicensicnal effects on the pawer distribution.
Based on this analysis, we concur that the measurement uncertaint es of 6% for F and 7% for F are r
Q adequate.
I 2.2 Transient Analysis 2.2.1 CEA Withdrawal The CEA Withdrawal Event was reanalyzed from both the hot zero power condition and the full power initial condition. For the zero power case, two conputer programs were used. WIT-6 was used to calculate the nuclear power (reactivity) transient and FAC,TRAN was then used to obtain the thermal heat flux transient and the, fuel and clad temperatures.
The reactor trips on the Variable liigh
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Power Trip at 25% pcwer and the nuclear power does not overshoot the full power l
l nominal value. The core and the RCS are not adversely affected since the combination of thermal power and the coolant temperature result in a DNBR greater than the limiting value at 1.30.
For the 'ull power case, the LOFTRAN computer program is used. The thermal margin / low pressure trip provides protection for this case and terminates the transient before the DNBR falls below 1.30.
We have reviewed the initial conditions, the reactivity coefficients, and the CEA trip insertion characteristics and find the CEA withdrawal analyses and consequences acceptable.
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,2.2.2 Tre cts orop event was reanalyzed usino stancard W nuclear desigr. methods to compute s:cacy state power distribution 5.
The pi5 king f : tor: tere then. red in the THINC code to calculate tne OhbR. LOFTRAN aas used for the tr:::1ent analysis. The resalts indicate that following the drop of the.::r:t CEA, tne reactor-nay return to full power without exceeding the i
core tnermal limits. We have reviewed the arsumptions used for initial system conditions as well as the reactivity feedback coefficients and dropped CEA worths used and find them to be acceptable.
2.2.3 CEA Efecti:n The CEA ejection accident was reanalyzed for both full pov:er and zero power initial cor.ditions at BOC and E0C using the TWINKLE code in one-dimension (axial) for the average core channel calculation and the FACTRAN code for the hot fuel rod transient heat transfer calculation. The analysis performed for the more limiting HFP case predicted a maximum fuel stored energy of 172 cal /gm which is well within the Regulatory Guide 1.77 limiting criterion We have reviewed the analysis assumptions including the of 280 cal /gm.
Doppler and moderator temperature coefficients, delayed neutron fractions, initial fuel temperatures, ejected rod worths, hot channel factors and trip reactivity insertion and find the analysis to be conservative and the predicted consequences acceptable.
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3.0 EVALUATION PROCED'JRE (Physics)
We have reviewec the report within the guidelines provided by Section 4.3, Included 15.4.1,15.4.2,15.4.3, and 15.4.8 of the Standard Review Plan.
-7 in our review"2II the description of tne exasri:.ct:Lal Cat: 5:se, the cale::12: ions performed, and the' comparisons.T. ace to support tne conclusion that the NNECO reactor physics cathobs ar? d :'nte to calculate physics param:t:rs and reactivity transients for :u s istone 2 reload cores.
4.0 REGULATORY POSITION (Physics)_
We'have reviewed tne revised Westingnouse reactor.pnysics methods used l
by NNECO and benchmarked against Millstone 2 measurements over the first.
three cycles and find them acceptable to be used in Millstone 2 safety related calculations of those quantities described above.
We also find-tne reanalysis of the reactivity initiuted transients described in the BSR adequately defines the reference safety analysis and is valid for all future cycles of Millstone 2 provided that the reload safety analysis
. input parameters for any given cycle 1.e bounded by these reference analysis When a reload parameter is not bounded, further evaluation or a values.
reanalysis will be necessary.
The report may be referenced in licensing sutmittals by NNECO for the Hillstone 2 reactor.
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5.0 REFERENCES
(Physics)
NNECO submittal of a proprietary addendum to the BSR on nuclear 1.
uncertainties, W. Counsil to R. Clark, May 28, 1980.
NNECO answer to question 10 on power peaking in fuel pins, W. Cpunsil 2.
to R. Clark, July 22, 1980.
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