ML20004F339
| ML20004F339 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/15/1981 |
| From: | Withrow G CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8106180181 | |
| Download: ML20004F339 (5) | |
Text
{{#Wiki_filter:r-m a_o ,e y% / 'T COR8958f5 L h POW 8r a ggmpa g%p,.,s. c ma Generet Offlees: 212 West Michigen Avenue, Jackson, MI 49201 e (517) 7884550 June 15, 1981 Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors 3 ranch No 5 U S Nuclear Regulatory Commission '4ashington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - REACTOR PHYSICS METHODOLOGY NRC letter dated April 24, 1981 requested additional information as described in its enclosure, pertaining to Big Rock Point reactor physics methodology. The majority of the questions involve the uncertainties in calculated and measurea core parameters. The application of uncertainties in Big Rock Point reactor physics methodology is based on common practice, engineering judgment and available operating data. No rigorous analysis exists to support fully the currently used uncertainties. Therefore, a benchmarking effort has been initiated to provide analytical justification for these uncertainties. The major part of this new effort will be verification of calculated power distributions and peaking facters and the uncertainties associated with flux vire measurements. The existance of little or no benchmark-qaality data results in the need to perform higher order cal-culations for much of the verification work. The need to perform additional analysis to address the requested information was brcught to the attention of the NRC Project Manager - Big Rock Point. It vis agreed that Consumers Power Ccmpany should provide as much information as possible within the kS-day response period and provide an estimate of v1.en a full response to all of the requested information could be submittei. Therefcce, to this letter provides cur current responses to the informatica request and it is esticated that a full response can be provided by December 15, 1981. Gregory C 'dithrow (Signed) ,.IeL,p Gregory C 'dithrov ',O~ / py- [Q*f [ Senior Licensing Engineer i CC JGKeppler, USNRC f ; M,%l7 n, J 7 URC Resident Inspector - Eig Rock Point /08 A 7I . g ~ e /Ig 8106180lD f
Ihi 3 ATTACH!ENT 1 - Responses to the' Enclosure of. NRC Letter dated April 24, 1981 Reactor Physics Methodology A first reading of the document describing the Big Rock Point Physics Methodology ~ leads to the conclusion that the GROK code, with input-preparation 'from CASMO, PDQ-7.and other. standard cross section preparation codes,.is used for both reload - design and analysis and as a replacement for or adjunct -to the process computer. When used for reload applications the code calculates power distributions, cri-tical rod patterns at zero power and at operating conditions,' rod. sequences and rod worths within the sequence, reactivity coeff1cients (void and Doppler), power distributions and margins to thermal limits. For core follov (process computer) applications the code is used to calculate. margins to thermal limits, segment i burnups, and Lflux vire activation profiles. (for comparison with measurements).
- 1.. Please verify, clarify, or modify the above understanding, as appropriate.
Response
This: conclusion is accurate. 2. ,For those parameters which are to be used as input to safety analyses please provide the following: a. The accuracy with which the parameter is calculated-including the support-ing data. Is the comparison with experiment, with higher order calculation, or=other? b. Comparison of results of GROK with those of the code used to calculate the parameter for the FSAR analyses or the currently used reference analysis for each event. c. Indicate whether the nominal value is used in the safety analysis or is the uncertainty added before comparison with safety analysis value?
Response
The parameters used as input to the safety analysis include: Beta effective Prompt neutron lifetime Void coefficient Doppler coeffecient Scram reactivity curve Margin to fuel damage during slow rod withdrawal (essentially a complete analysis) Analysis is underway to estimate the accuracy of these parameters. a. Results vill be supplied later. i b. The following table shows a comparison of values computed by GRCK for Cycle 17 and the values used in the Final Hazards Summary Report. y y---,-,, -,-p3. ~ .---s -m r y-, y -.~,,. +. - -- ,ws--r-
$ ;g ~ 2- ~ Parameter -FHSR' GROK, Cycle 17 ~ ' Doppler Coef -5.k2 ic 10-5. -10.67 x 10-5 (Ak/k/% power) 'Vo'id Coef BOC .20 .1h53 (Ak/k/ unit-void)
- EOC -'10
.1050 -Beta effective .007 .00577 h0 x 10-6 137'.3 x 10-6 Promktneftronseconds) life ime-A' comparison of the scram curve that GROK computed for Cycle 17. and : the reference ~ analysis that was ' submitted June 20,19Th (Docket 50-155, ~ License DPR-6, Proposed-Technical Specifications Change) is given by -Figure 1 attached. Tae dropped rod worth that yields 280 cal /gm energy deposition vas determined to be 1.8% Ak/k. The maximum in sequence -dropped worth for-Cycle 17 was ca.lculated to be 0.hT5 Ak/k. c. It'is our current practice.to co= pare best estimate values to those used in the safety analysis with no uncertainty applied. However, for the transient. event which typically deternines fuel operating limits (ie, turbine trip without bypass valve operation) the void coefficient is the most critica3-neutron kinetics parameter. The results of this accident become more severe with a larger negative void coefficient. Considerable margin exists between the Cycle 17130C void coefficient as predicted by GROK'and the value used in the FHSR. 3 A brief' description of the role GROK plays as a process computer surrogate. What parameters does it calculate? How are inputs obtained (ie, is there a direct connection from reactor to computer or is GROK used only on an off-line basis)? What is the accuracy of the calculated quantities? Are nominal values 'used to compare with operating limits (ie, the limits them-selves take into account the uncertainties) or are uncertainties added before comparing?
Response
As a. process computer surrogate, GR0K is used off-line to either update fuel exposure or to evaluate thermal margin and compute flux vire activation profiles. There is no direct connection from the reactor to GROK; all inputs must be done by hand. To evaluate thermal margin, the inputs are taken from a hand heat balance and consist of power level, inlet subcooling, core flow and control rod pattern. The quantities computed by GROK are: neat flux MAPLHGR MCEFR MCPR' assembly power The GROK calculation includes a k% uncertainty factor that is applied to heat flux, MCHFR and MCPR that accounts for fuel densification and mechanical variations in the fuel. Also, a 10% uncertainty factor is applied to enthalpy increase as the coolant' flows up the channel and is input to the MCHFR and MCPR calculations.
g. ~ ~, _4. 3-Before-- the: computed" values of; heat' flux,' MAPMGR' and MCHFR are ~ compared.to ;the ~ .Joperating' limits,'a'fluxvire correction factor is applied.~ The fluxvire corree-tion factor is' derived by comparing the GROK computed'fluxvire shapes ta the . measured : shapes and finding the'. largest difference 'between the two sets of ' j curves. The fluxvire measurement is taken at:the same time the heat balance lis made. The. single largest difference is then applied to the computed thermal limits (heat flux, MAPMGR & MCHFR) for the entire core. In addition, 2". is applied to these limits in addition to the fluxvire correction factor. The. fluxvire correction is not applied to bundle power, however,1". uncertainty is - applied. An analysis -to_ determine the uncertainty in the esiculated parameters will be performed.
- k. ' With ' respect to the' fluxvire measurements please provide estimates of the t
accuracy of these measurements for. axial power distribution in a single L ~. foar assembly cluster and for making radial power distribution determinations. (,
Response
The accuracy of the fluxvire system for measuring axial power distributions vill be analyzed.. There are only eight fluxvire holes, ~ and four of these are symmetric to the other four, so that the system is generally regarded as inadequate-for measuring radial power distribution. ' 5. Provide -information with respect to initial critical configurations in terms - of ak as well-as the number of notches. ~
Response
For Cycle lh.(Figure 4-1 of the Methodology Report) ak/k' =.00160. For Cycle'15 (Figure h-2) ak/k = .00003. l 6. The fluxvire - GROK comparison data in the report should be su=marized to . provide calculational uncertainties for radial and axial peaking factors and any other pover' distribution factors used in safety analyses or core follow.
Response
N An analysis to provide these uncertainties will be performed.
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