ML17212B670
| ML17212B670 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 06/07/1982 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| References | |
| NUDOCS 8206140119 | |
| Download: ML17212B670 (15) | |
Text
Docket No.:
50-389 Dr. Robert E. Uhrig Vice President Advanced Systems 5 Technology Florida Power 5 Light Company Post Office Box 529100 Miami, Florida 33152
Dear Dr. Uhrig:
7 1982 RHartfield, MPA BSheron OELD IE JGuttmann Reg.
2 Distribution:'0-389 NRC PDR L PDR NSIC TIC ACRS (16)
LBb3 File DEisenhut FMiraglia VNerses JLee RTedesco RVollmer SHanauer RMattson HThompson
Subject:
St. Lucie Plant, Unit 2 FSAR - Request for Additional Information in Support of Confirmatory Issues Relating to CESSEC
- Enclosed, please find additional request for information we require before completing our confirmatory reviews, as documented in our SER for St. Lucie 2.
Responses to the enclosed request should be submitted by August 1, 1982.
If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses.
Please contact Mr. Nerses (301/492-7318),
St. Lucie 2 Project Manager, if you desire any discussion or clarification of the enclosed report.
Sincerely, Original signed by
Enclosure:
As stated cc:
See next page 8206140119 820b07 PDR ADOCK 05000389 E
PDR Robert L. Tedesco, Assistant DIrector for Licensing Division of Licensing l
p@CP.V~ED oi suN 09 l9~~~
4 g @NEST Rt~ospa >++OE ggggf QQ4MA N 6
woe OFFICE/
SURNAME/
OA~E f>
D:
3 V
2t
~ \\
DL.
FMir na 6/
'82
'L R
sco
/.,2
~ ~ ~ 0
~ ~
~ ~ ~ ~ 0 ~ ~ ~ 0
~ ~
~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~
~ ~ \\
~ ~ ~
~ ~ ~ ~ i
~
~ i
~ ~ ~ ~ ~ ~ ~ ~,,
NRC FORM 318 II0/SOI NRCM 0240 FFICIAL RECORD COPY C
- USGPO; 1980 329.821
7 rl l$
e S~iil y
C E
I sr g
I
LUG IE Dr. Robert E. Uhrig, -Vice President Advanced Systems and Technology Florida Power
& Light Company P. 0.
Box 529100
. Miami, Florida 33152 Harold F. Reis, Esq.
Lowenstein,
- Neman, Reis, Axelrad
& Toll 1025 Connecticut
- Avenue, N.W.
. Washington, D.C.
20036 Norman A. Coll, Esq.
Steel.Hector
& Davis 1400 Southeast First National Bank Building Miami, Florida 33131 Mr. Martin,H. Hodder 1131 N.
E. 86th Street Miami, Fl orida 33138 Resident Inspector St. Lucie Nuclear Power Station c/o U. S. Nuclear Regulatory Comnission 7900 South AlA
. Jensen Beach,. Fl.orida 33457 U. S. Nuclear Regulatory Comnission 101 Marietta Street Sui te 3100 Atlanta, Georgia 30303
CONFIRMATORY QUESTIONS REGARDING CESEC*
Section 1.0 440.83 Describe in detail the relationship between CESEC-SAR, CESEC-ASS, CESEC-SLB and CESEC-III with an emphasis on the diffsrences.
440.84 There is no discussion of DNBR calculations.
If the code does compute DNBR, provide details.
440.85 Describe the self-initialization procedure.
440.86 How is the closure head bubble modelled'ection 2.0 440.87 Does assumption
(.1) of Appendix B, assuming pressure to be spa-tially uniform throughout the entire primary coolant system, imply that no differentiation is made between pressurizer pressure and system pressure in the derivation of the T/H equations, eq.
(B-30) through eq (B-78)7 Where and how is each pressure used'ection 3
440.88 How are the crossflows, bypass flows, mixing flows, head flows and leak flows incorporated into eq. (3-1), the equation for the pump flow2 440.8g How does the inertia term in eq. (3-1) take account of the flow split..for parallel pumps'ection 4
440.90 Justify the use of the Semiscale degraded two phase'pump data to model CE pumps.
- Note:
guestions are organized in Sections, corresponding to CENPD-107, "CESEC."
Section 6.0 440.91 Describe how the level in the pressurizer is determined and Wow the external heat transfer/mass flow terms are divided between the steam region and the liquid region.
Section 7.0 440.92 Why does the gravity term in the surge line momentum equation, eq.
(7.1) contain expressions for the pressurizer when the inertia term is written only for the surge l.ine?
440.93 Show the table of f.values used 'in eq. (7-1) when Re > 15000.
Section 8.0 440.94 Why is the pressurizer pressure time derivative and not the RCS pressure time derivative used in eq.
(8-1)?
Section 9.0 440.95 Are the 13 nodes referred to in the wall heat transfer model radial nodes?
Provide a figure for the model to illustrate it.
440.96 How is heat conduction through the steam generator tubes modelled?
440.97 Is eq. (9-2) for the shroud heat capacity solved simultaneously with the thermal hydraulic equations of Appendix B?
Section 10.0 440. 98 For the Doppler and the moderator reactivity feedback calculation does the core have only one axial node?
How is the split core ac-counted for?
440.99 How is the moderator temperature/density calculated for the reac-tivity feedback2 440. 100 How is the fuel temperature for the Doppler feedback calculated2 440.101 Describe in detail the 3-D reactivity feedback model used for steam line breaks.
440.102 Explain the homogenization procedure for the third radial node of the fuel rod heat conduction model.
440.103 Is the heat transfer correlation, given by eq. (10-1), for the clad-coolant interface, actually used for all pressures and temp-eratures2 440.104 What is the difference between T and Tw1 in Fig. 10-62 440.105 Is gw in Fig. 10-6 the source term used in the thermal-hydraulic equations of Appendix B2 Section 11.0 440.106 Is the letdown fluid temperature at the heat exchanger user input as stated by %11.0 or calculated in accordance with eq.
(F.-3)2 440.107 In the iterative procedure described in Appendix F-is the critical flow set equal to the D'Arcy flow2 Why assume saturation at the RHX exit2 440.108 What is the database for the heat transfer correlation given by eq. (F-6)2
Section 13.0 440.109 (a)
Why does the suppression factor S, given by eq. -(13-9)- got correspond with the formula given in Table 2 of the Hoeld paper referenced'b)
Provide references/explanation for the difference in func-
.tional dependence between Chen's suppression factor
[f(Re "F '3 and Hoeld's expression [f(1 - x)
Re F '.
(c)
Justify the choice of Hoeld's formula for the Reynold's number factor, F, eq.
( 13-8) over the original Chen values.
440.110 Justify the use of the Dittus-Boelter cor relation, eq.
(13-3) for film heat transfer.
)
440.111 (a)
Justify the assumption of two phase flow with condensation in the steam generator primary for all cases of forward heat
'transfer.
(b)
Explain why the extrapolation of the Akets,
- Deans, and Crosser correlation to the water system is valid.
440.1j2 (a)
Present the database for CE's modification of the Rohsenow pool boiling correlation, eq.
(13-10),
and discuss the range
'f validity.
{b)
Should the term (P - 800) be (Psec - 800) in the second expression for KR on page 13-57 440.113 Justify the assumption of free convection in the steam generator seco ndai y during reverse heat trans fer.
A.. Hoeld, "A Theoretical Hodel for the Calculation of Large Transients in Nuclear Natural Circulation U-Tube Steam Generators (Digital Code UTSG),"
Nuclear Eng.
and Design, 47, pp. 1-23, 1978.
- J.
C. Chen, "Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," I&EC Pr'ocess Design'and Development, 5, pp. 322-329.
OO Section 15.0 440. 114 Is flow through the valves of the-steam system assumed to be;.
choked even when the sink pressure is higher than the throat I
pressure?
440.115 Justify the expression for ATUB in Fig. 15-1A.
440.116 Correct the equation for WgL in Fig. 15-1A.
Section 16.0 440.117 Justify the use of CRITCO for steam discharge flow when. the refer-ence+ quoted in the CESEC report is for two phase mixtures; 440. 118 Is an orifice coefficient used in eq.
(16-4) only when the steam generator tube rupture option is selected2 Section 17.0 x
440. 119 Show why the steam generator load dependency of the steam gener-ator water level, required in the steady state situation, is not needed to determine level during transients.
A endix B
440. 120 Equations (B-31), (B-33), (B-35) and (B-54) should be corrected.
440. 121 Is reverse flow allowed in the cores'40.122 Is W25 = 2W25 77 A. N.Nahavandi and M. Rashevsky, "Computer Program for Critical Flow Discharge of Two Phase Steam-Mater Mixtures," CVNA-128, February, 1962.
40
~g 440.123 Describe the algorithm CESEC uses to trace the state of the pres-surizer and to maintain continuity as the state changes.
Is there an automatic time step adjusterl 440.124 Justify the identification, in state 8, of WB-with the vapor por-tion of the critical flow through the pressurizer valve.
How is this consistent with the absence of WB in state 7?
'40.125 Provide references for the two phase pressure drop correlations, eqs.
(C-1) - (C-5) and a comparison with the Baroczy or Chisholm correlation.
4'40. 126 Are the CEFLASH-4A water properties applicable to the supercriti-cal regionT Provide a copy of'the'eport A
endix D
440. 127 Explain why eq.
(D-12A) can be neglected.
440.128 (a)
How is the reference exit temperature in the steam generator node calculated and how is it used'b)
How is the exiting enthalpy computedl A'ndix E
'440.129 Prove that eq. (E-4) converges.
+
C. J. Baroczy, "A System Correlation for Two Phase Pressure Drop," AIChE reprint f37.
Paper presented at the 8th Hat. Heat Transfer Conf., Los
- Angeles, Aug. 1965.
D. Chisholm, "The Influence of Mass Velocity on Frictional Pressure Gradients During Steam-Water Flow," Paper 35 presented at the 1968 Thermodynamics and Fluid Mechanics Convention, Bristol,.1968.
CEWPD-133, "CEFLASH-4A, A FORTRAN IY Digital Computer Program for Blowdown Analysis," August, 1974.'
W 440.130 How is the moderator feedback divided between the void feedback and the density feedback?
440.131 (a)
Provide additional information about the core coolant flew and. temperature calculation as the connection between CORE(
and LOOPEg is not clear.
Which moderator temperature and density is used for the reactivity feedback?
(b)
Mhat fr'action of instantaneous coie power is absorbed by the coolant?
r 440. 132 Justify the reactivity flux weighting method (i.e.; flux or flux**2).
440. 133 Mhy does the void reactivity calculation employ static quality when the Nartine1li-Nelson correlation referred to uses flowing quality?
440. 134 (a)
Is the quantity g used in Teff for the Doppler an input'onstant?
(b)
How is it determined?
440.135 Justify the CESEC/PDg-TH calibration scheme for weighting fac-
~
tors.
440. 136 Explain why in the formulation of the T/H nodal equations the pressure p is used but in the determination of water properties the pressure p + apsuge is utilized.
440. 137 Present the derivation/assumptions used to reduce the T/H nodal equations to a 19 equation set.
N.
C. Sher, "Review of Martinelli-Nelson Pressure Drop Correlation,"
Westinghouse Electric Corp. Report WAPD-TH-219 (July 1956).
440.138 In the T/H model is the instantaneous, core power entirely ab-sorbed by the coolant with no lieating of the fuel?
440.13g Are the sprays 100" efficient2 440.140 Discuss the DNBR calculation in more detail; in particular the open/closed channel aspect.
440. 141 Describe the modelling of the steam bubble.
What effect does 'the.
assumption of a uniform RCS pressure have2 440.14's there only boiling heat transfer on the secondary side of the steam generator.
and only film heat transfer on the primary side2 440.
143 (a)
Explain the steam generator dryout heat transfer criterion and the calculation of UA.
(b)
How is the steam flow calculated2 (c)
How is tube heat transfer area related to the mass inven-.
tory, the recirculation flow, and the quality calculation2 (d)
How is quality calculated2 440. 144 Is it correct that the heat transfer in the steam generator is UA(T. - T )
s UA = 1.0 UA forward transfer
= input= reverse transfer T
+ T UA(
- T) when T.> T and T T
1 s
0
-s with UA = 0.8 UA of previous timestep where UA = overall heat transfer coefficient Ti = primary side inlet temperature To = primary side outlet te'mperature Ts
= secondary side temperature
440.I4g (a)
Does the code use the cold edge temperature or. the-cold leg temperature for the moderator feedback?
(b)
How is the cold edge temperature calculated?
4403.46.
Is there an iteration between the pump flow calculation and the energy/mass bal ance calcul ations?
440.147 (a)
How is the input flow fraction for the outlet plenum to closure head flow determined?
(b)
Are the plena crossf1ows or the bypass flows user specified?
If so what is the basis for the input values?
440. 148 How is the UA parameter used in the steam generator heat transfer determined at the minimum mass inventory?
440.149 It fs tat fntent. that CESEC-I II be used for future analyses fnstead of CESEC-I or CESEC-I I.
Pl ease d~nstrate that CESEC-I II fs capabl e of per fo~-
fng the required analyses by suhafttfng an analysfs of the followfng tran-sfents for the CESS'R desfyi usfng CESEC-III:
( a) sterol fn
- break, (b) feemter lfne break, (c) loss of feedwater ATMS, and (d) steam 9 nerator tube ruptu're.
The results obtained xfth CESEC-III should be over'lafd Ath the results ob-tain ned vfth CESEC-I and CESEC-I I.
440.150 In g nerH
~ ft fs cer fxyressfon that the data
) presented by Fla. P.hL.
fs fnsufffcfent to support a ~n ral afxfng mde3 sfnce only one flee condf-tfon'xas measured.
Furthe~re, the par tfcular Am condftfon chosen has a
Reynolds.nurser nearly one full ord r of aagnf tude belm operatfng condftfons.
The experfmnt ~nsented each core asseebly hy a sfngle flm tube and used aft as fts sfeulant flvfd.
SQ2 ~as fn$ected fpto the-afr flat of one (of
- 4) reactor vessH.inlet nozzles on a roughly I/O scale model.
The SO~ concen-tratfon was measured at the exft of each of the "core tubes and fn each of the No reactor vessel outlet'nozxles.
"Although the r~thod appears reasonable for obtafnfng fnfomatfon on reactor vess l outlet Am, unfortunately, data
~ fs presented for only one cperatfng condftfon.
Furthe~re, ve are concerned about the f~act of constrafnfng the core f}cr4 fn tubes Den substantfal cross flpc fs to be expected.
4 Thus ft fs car general opfnion'hat thfs sfngle dam pofnt fs not an adequate basfs upon which to bufld a c~uter co& fidel fntended to sedel a
.vfde range of flm condftfons.
The fol3owfng s~mcf ffc questfons should be addressed by the applfcant.
4 (a )
Thfs data fs only for one Re nu"..ber, representfng only one operatfng condf tfon.
Upon Rat grads does CE utflfre thfs data for other flow condftfons such as p~ coastdmn or loss of one pmp.
(c )
Hcx ms thfs data frcplerwnted fn the CESEC cmyuter codes, 5ustf 4y and expl a fn fn dipth.
Dfscus: the <~act of having done these experi~~~nl 5 fn a g c
'entry Nn cn pr'ohfb3ts cross flc.. b ~en ass~lf~s.
Hm fs the cross f1o~ expected to f~act the fl&splft fn the exft no-zles.
(T)
Test Pep-~i on Flufd nfx3ng fn a Seal~" R~ac.cr Vessel Flee Rvel,"
CEH-16g(L)-P, July, 154"I, Cc-.'-us fcn Engfn ring.
r
~
~
~
~
~
Furthe~re, hm does thfs lack of cross-flan @pact the accu-racy 0f sfmlatfon of core moderator temperature during n remen to POSSE'?
(g)
Oiscuss eccuracy of the experiments - &at ere the fractional error.
fn the 5 of flmr throv~~ various "core tubes"7 I~
The total X of flic going canto the bro outlet nozz1es fs 38
+
$ s ft not 50~2 Nat mre the experimental errom7
~
~
~
v
~
~