ML18139B402

Forwards Responses to NRC 810409 Questions Re VEP-FRD-33, VEPCO Reactor Core Thermal-Hydraulic Analysis Using Cobra Iiic/Mit Computer Code.

ML18139B402
Person / Time
Site:	Surry, North Anna
Issue date:	06/12/1981
From:	Thomas W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Eisenhut D Office of Nuclear Reactor Regulation
References
359, NUDOCS 8106180259
Download: ML18139B402 (11)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND~ VIRGINIA. 23261

• June 12, 1981 W.N.THOMAS VxcE PRESIDENT FUEL RESOlrBCES Mr. H. R. Denton, Director Serial No: 359 Office of Nuclear Reactor Regulation FR/KLB: plc Attn: Mr. D. G, Eisenhut, Director Docket.Nos.: 50_;280 Division of Licensing 50_;281 U.S. Nuclear Regulatory Commission 50-,338 Washington, DC 20555 50-,339 License Nos.: DPR-32 DPR-37 NPF-4 NPF-7

Dear Mr. Denton:

TOPICAL REPORT VEP-FRD-,33 "VEPCO REACTOR CORE THERMAL-HYDRAULIC ANALYSIS

. *us ING THE. COBRA. IIIC/MIT. COMPUTER. CODE" ..

Attachment 1 provides our responses to Nuclear Regulatory Commission (NRC). Staff questions on the Vepco topical rep~rt VEP-FRD-33,"

"Vepco Reactor Core Thermal-Hydraulic Analysis Using The COBRA IIIC/MIT Computer Code", transmitted by theW. N. Thomas (Vepco) to H. R. Denton (NRC) letter, Serial No. 795, dated September *28, 1979, These questions on VEP-FRD-,33 were sent in a letter from R. L. Tedesco (NRC) to W. N.

Thomas (Vepco), dated April 9, 1981.

Should you have any further questions concerning this topical report, please contact us.

Attachment f)oo I cc: Mr. R. L. Tedesco Assistant Director

..s for Licensing Division of Licensing I /1 r

e PAGE 1 ATTACHMENT 1 Response to NRC Questions on VEP-FRD-33

PAGE 2 HRC Question 492.1 "Provide CHF predictions (plots and tables) versus the measured values for the test data (Ref. 1) using VEP-FRD-33 COBRA/W-3. Include at least two points from each set of tests in the Ref. 1 test data such that your test conditions will be similar to the limiting thermal-hydraulic operating conditions for Surry Units 1 and 2 and North Anna Units."

Response

COBRA models of the Ref. 1 3x3 and 4x4 test bundle geometries were created using code correlations and options consistent with VEP-FRD-33. Three data points from each test series were chosen such that the range of key test parameters would be maximized.

The data ranges are given in Table I. The North Anna and Surry allowable operating conditions are well within these ranges.A comparison of the COBRA/W-3 DNB predictions and the experimental DNB data is given in Table II. These results are presented graphically in Figure 1. The sample mean and standard deviation of the measured-to-predicted heat flux ratio are 0.982 and 0.0638, respectively. These limited data indicate that in order to meet a 957o probability/95% confidence level reactor design criteria, a minimum DNBR of 1.19 would be required. VEPCO intends

e PAGE 3 Response to NRC Question 492.1 (continued) to continue using the 1.30 minimum DHBR design criteria established by the original W-3 correlation data. These original W-3 correlation bounds are also indicated on Figure 1. Should we desire to use a minimum DHBR design basis other than 1.30, additional justi£ication would be provided.

e e PAGE 4 Table I: Range of Key Test Parameters Ranges Pressure 1491-2433 Cpsia)

Inlet Average Mass Velocity 1.05-3.66 (Mlbm/h:c-ftZ)

Inlet Tempe:catu:ce 433.0-617.0 (OF)

Local Heat Flux 0.563-1.063 CMBTU/h:c-ftZ)

.

,',

,' e PAGE 5 Table II: Comparison with DNB Data

-*.

Test Run P:ressu:re Inlet Inlet Avg. Local DNB (MBTU/h:r-ftZ)

Section no. Cpsia) Temp.

(OF)

Mass Vel.

(Mlb/h:r-ftZ)

---

qmeas qp:red. qmeas/qp:red

---

I 4 1502 468.0 1. 05 0.631 0.734 0.860 I 5 1502 480.0 2.03 0.803 0.839 0.957 I 6 1503 518.5 3.05 0.870 0.892 0.975 II 11 2100 567.0 2.55 0.819 0.768 1. 0 65 II 15 1808 579.0 3.55 0.801 0.791 1.013 II 60 2115 567.5 3.06 0.893 0.865 1. 0 32 III 21 1514 483.0 2.56 0.692 0.714 0.969 III 25 2091 544.0 2.SS 0. 623. 0.614 1.015 III 46 1799 559.0 3. 0 1 O.S66 0.573 0.988 IV 71 1509 507.0 2.58 0.779 0.818 0.952 IV 75 18 1 1 567.0 3.58 0.763 0.781 0.977 IV 81 2109 546.0 2.56 0.752 0.811 0.927 V 93 1502 476.0 1. 57 0.751 0. 711 1.056 V 99 18 1 1 553.0 3.64 0.794 0.870 0. 913 V 103 2109 560.0 2.58 0.722 0.749 0.964 VI 129 1541 540.0 3.63 0.829 0.843 0.983 VI 135 1813 560.0 3.57 0.820 0.819 1. 00 1 VI 144 2433 615.0 3. 1 3 0.608 0.612 0.993 VII 208 2026 536.7 2. 61 0.795 0.874 0.910 VII 211 1497 478.3 2.55 0.906 0.932 0.972 VII 216 1790 501. 0 2.07 0.777 0.839 0.926 VIII 219 1491 481.0 2.55 0.868 0.919 0.945 VIII 225 2105 565.3 2.55 0.690 0.769 0.897 VIII 235 2415 583.7 3.59 0.866 0.955 0.907 IX 254 1491 499.0 2.57 0.988 0.880 1 . 12 3 IX 264 2069 583.0 3.06 0.793 0.778 1.019 IX 270 2400 586.7 3.66 1. 063 0.995 1 . 0 68 X 275 1497 537.7 3.59 0.949 0.918 1.034 X 277 1799 558.3 3.58 1.008 0.892 1 . 130 X 290 2419 617.0 3.05 0.704 0.706 0.997 XI 346 1815 561 . 7 3.57 0.783 0.873 0.897 XI 356 2395 604.3 3.03 0.648 0.726 0.893 XI 378 1496 433.0 1. 53 0.778 0.825 0.943 XII 380 1854 568.0 3.53 0.662 0.621 1 . 066 XII 386 2098 578.0 3.08 0.576 0.589 0.978 XII 392 2403 584.0 2.53 0.563 0.559 1.007

F e COMPARISON OF DNB DATA WITH COBRA PREDICTIONS FIGURE 1 Heated Axial Flux Grid with Grid without Length Distribution Vane Vane 8' u sin u cosine u

• 0 it.

-

14' u sin u cosine u *0 --

.:.:;.:::: : .........

---~--** ***- .....':: ..****

,.. . :::_:*1:*.*,:1*:::::,::/:~::1*::*.*,'::::':::: ::::.i::::_i,./:

, *:i_::: :.: *. :***/ /

1.0

=~t~+/- :~j:~:~:: :::J~:: :~J~~:=: :~~J~:~y~ ~~t::; ~~t(:: =!~!:(:,:::;,?==-j:=:-

:: ::~: <Y=:: *:J:::: ::):::: :::~F:~1i?11::fJ::: :,:_::::,~ ::>:,~ .

=J!:: ::~J~:: :::T!: ~~:J::: ~:J~~ ;~:::::~::IX :,J)_:: :L/o o o L>::

\:; :;;:1;:: :~]~~:: ~+:=~~:;:t~== ~::-:f-!:: :j[::. :~:=i. W :(<::;_:~::::~~;,,'

0.8 0.6 0.4 0.2 o.o I I I I I o.o 0.2 0.4 0.6 0.8 1.0 2

q~NB Predicted (MBTU/hr-ft )

---~....

e PAGE 7 NRC Question 492.2 "Confirm that VEP-FRD-33 computer code will be used only for non-LOCA thermal-hydraulic analysis."

Response

The VEP-FRD-33 computer code will be used only for non-LOCA thermal hydraulic analysis.

' .

• e PAGE 8 HRC Question 492.3 "Confirm the applicability :range fo:r the key pa:ramete:rs such as temperature, quality, pressure, flow, etc., fo:r the use of your DNB co:r:relation."

Response

The :range of the key pa:ramete:rs associated with the W-3 co:r:relation, the L-g:rid factor, the cold wall factor and the non-uniform heat flux multiplier a:re included in Table III.

These a:re supported by the :references also indicated in Table III. Since these :ranges bound the operating conditions present in the Condition II, III, and IV transients fo:r which DNB is a concern, we intend to use the W-3 co:r:relation fo:r those transients.* Any DNB calculation pe:rfo:rmed at p:ressu:res less than 1500 psia will not include the spacer factor correction because of its limited p:ressu:re :range.

- - .- -

e

• 4_

PAGE 9 Table III: W-3 Correlation Limits Ref. Pressure Mass Equiv. Local J\Hial Inlet Correlation Ho. Range Velocity Diameter Quality Height Temp Cpsia) (Mlb/h-£ 2 ) Cin) (in) (OF)

W-3 1, 2 1000- 1. 0- 0.2- :50.15 10- >400 2400 5.0 0.7 144 F-factor 1, 2 1000- 1. 0- 0.2- :50.15 10-2400 3.0 0.7 144 Coldwall 1, 2 1000- 1. 0- :50.15 >10 Factor 3,4 2400 5.0

• _spacer 3,4 1490- 1.5- :50.15 96- 404-Factor 2440 3.7 168 624

• PAGE 10 Refe:cences

1. E.R. Rosal, J.O. Ce:cmak, L.S. Tong, J.E.Caste:cline, S.Kokolis, and B. Matze:c, "High P:cessu:ce Rod Bundle DNB Data with Axially Non-Unifo:cm Heat Flux," Nuclear Enginee:cing and Design Vol 31 (1974), No.1, pp.1-20

2. L.S. Tong, "Boiling C:cisis and C:citical Heat Flux.~ U.S. AEC C:citical Review Se:cies, 1972

3. F.F. Cadek, F.E. Motley, "Application 0£ Modified Space:c Facto:c to L-G:cid Typical and Cold Wall Cell DNB," WCAP-8030-A, Janua:cy 1975

4. F.F. Cadek, F.E. Motley, "DNB Test Results fo:c R-G:cid Thimble Cold Wall Cells," WCAP-7958 Add. 1, Janua:cy 1975