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| | issue date = 06/12/1981 | | | issue date = 06/12/1981 |
| | title = Forwards Responses to NRC 810409 Questions Re VEP-FRD-33, VEPCO Reactor Core Thermal-Hydraulic Analysis Using Cobra Iiic/Mit Computer Code. | | | title = Forwards Responses to NRC 810409 Questions Re VEP-FRD-33, VEPCO Reactor Core Thermal-Hydraulic Analysis Using Cobra Iiic/Mit Computer Code. |
| | author name = THOMAS W N | | | author name = Thomas W |
| | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) | | | author affiliation = VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| | addressee name = DENTON H R, EISENHUT D G | | | addressee name = Denton H, Eisenhut D |
| | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) | | | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) |
| | docket = 05000280, 05000281, 05000338, 05000339 | | | docket = 05000280, 05000281, 05000338, 05000339 |
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| {{#Wiki_filter:r W.N.THOMAS VxcE PRESIDENT FUEL RESOlrBCES | | {{#Wiki_filter: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 |
| * VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND~
| |
| VIRGINIA. | |
| 23261 June 12, 1981 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation Attn: Mr. D. G, Eisenhut, Director Serial No: 359 FR/KLB: plc Docket.Nos.: | |
| 50_;280 50_;281 50-,338 50-,339 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 | |
|
| |
|
| ==Dear Mr. Denton:== | | ==Dear Mr. Denton:== |
| TOPICAL REPORT VEP-FRD-,33 License Nos.: "VEPCO REACTOR CORE THERMAL-HYDRAULIC ANALYSIS . *us ING THE. COBRA. IIIC/MIT. | | |
| COMPUTER. | | TOPICAL REPORT VEP-FRD-,33 "VEPCO REACTOR CORE THERMAL-HYDRAULIC ANALYSIS |
| CODE" .. DPR-32 DPR-37 NPF-4 NPF-7 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 | | . *us ING THE. COBRA. IIIC/MIT. COMPUTER. CODE" .. |
| *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 cc: Mr. R. L. Tedesco Assistant Director for Licensing Division of Licensing f)oo I ..s I /1 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. | | Attachment 1 provides our responses to Nuclear Regulatory Commission (NRC). Staff questions on the Vepco topical rep~rt VEP-FRD-33," |
| 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. | | "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. |
| Three data points from each test series were chosen such that the range of key test parameters would be maximized. | | Thomas (Vepco), dated April 9, 1981. |
| 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. | | Should you have any further questions concerning this topical report, please contact us. |
| These limited data indicate that in order to meet a 957o probability/95% | | Attachment f)oo I cc: Mr. R. L. Tedesco Assistant Director |
| confidence level reactor design criteria, a minimum DNBR of 1.19 would be required. | | ..s for Licensing Division of Licensing I /1 r |
| 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 Table I: Range of Key Test Parameters Pressure Cpsia) Inlet Average Mass Velocity (Mlbm/h:c-ftZ) | | e PAGE 1 ATTACHMENT 1 Response to NRC Questions on VEP-FRD-33 |
| Inlet Tempe:catu:ce (OF) Local Heat Flux CMBTU/h:c-ftZ) | | |
| Ranges 1491-2433 1.05-3.66 433.0-617.0 0.563-1.063 PAGE 4
| | 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." |
| ,', . :,' 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. Mass Vel. -------------------------(OF) (Mlb/h:r-ftZ) qmeas qp:red. qmeas/qp:red | | ===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. |
| 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 1 8 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 1 8 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 . 0 0 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 . 0 6 3 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 . 0 6 6 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 1.0 0.8 0.6 0.4 0.2 o.o ---~ .... e FIGURE 1 COMPARISON OF DNB DATA WITH COBRA PREDICTIONS Heated Axial Flux Grid with Grid without Length Distribution Vane Vane 8' u sin u
| | 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 |
| * it. cosine u 0 -14' u sin u * -cosine u 0 -.:.:;.:::: | | |
| : ......... | | 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. |
| ':: .. ,.. :::_:*1:*.*,:1*:::::,::/:~::1*::*.*,'::::'::::
| | |
| ::::.i::::_i,./: | | e e PAGE 4 Table I: Range of Key Test Parameters Ranges Pressure 1491-2433 Cpsia) |
| *:i_::: :.: *. :*** / ---~--** ***-..... **** . , / :=~t~+/- :~j:~:~::
| | Inlet Average Mass Velocity 1.05-3.66 (Mlbm/h:c-ftZ) |
| :::J~:: :~J~~:=: | | Inlet Tempe:catu:ce 433.0-617.0 (OF) |
| ~~t::; ~~t(::
| | Local Heat Flux 0.563-1.063 CMBTU/h:c-ftZ) |
| :::;:::: ::: ::~: <Y=:: *:J:::: ::):::: :::~F:~1i?11::fJ::: | | :,' 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. |
| ::>:,~ . ::=J!:: ::~J~:: :::T!: ~~:J::: | | (OF) |
| ;~:::::~::IX
| | Mass Vel. |
| :,J)_:: :L/o o o L>:: ::\:; :;;:1;:: :~]~~:: ~+:=~~:;:t~== | | (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-!::
| | |
| :j[::. :~:=i. W :(<::;_:~::::~~;,,'
| | 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 |
| I o.o 0.2 I 0.4 I 0.6 I 0.8 q~NB Predicted (MBTU/hr-ft | | it. |
| : 2) I 1.0 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.
| | 14' u sin u cosine u *0 -- |
| ' .
| | ---~--** ***- .....':: ..**** |
| * e PAGE 8 HRC Question 492.3 "Confirm the applicability
| | ,.. . :::_:*1:*.*,:1*:::::,::/:~::1*::*.*,'::::':::: ::::.i::::_i,./: |
| :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.* | | , *:i_::: :.: *. :***/ / |
| 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 | | 1.0 |
| :range. | | :=~t~+/- :~j:~:~:: :::J~:: :~J~~:=: :~~J~:~y~ ~~t::; ~~t(:: =!~!:(:,:::;,?==-j:=:- |
| --. ---~--------------------------------, 4_ e | | :::;:::: ::: ::~: <Y=:: *:J:::: ::):::: :::~F:~1i?11::fJ::: :,:_::::,~ ::>:,~ . |
| * 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 | | ::=J!:: ::~J~:: :::T!: ~~:J::: ~:J~~ ;~:::::~::IX :,J)_:: :L/o o o L>:: |
| *
| | ::\:; :;;:1;:: :~]~~:: ~+:=~~:;:t~== ~::-:f-!:: :j[::. :~:=i. W :(<::;_:~::::~~;,,' |
| * PAGE 10 Refe:cences
| | 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 |
| : 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}} | | 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}} |
|
|---|
Category:CORRESPONDENCE-LETTERS
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Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr
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Submits Response to Violations Noted in Insp Repts 50-280/98-201 & 50-281/98-201.Corrective Actions:Visual Insps Were Completed on Accessible Coatings Inside Containment for Both Units 1 & 2
ML18152B3541999-10-12012 October 1999
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ML20212J9101999-10-0101 October 1999
Forwards SE Accepting Licensee 990916 & 27 Relief Requests IWE-3 for Plant.Se Addresses Only IWE-3 Due to Util Urgent Need for Relief.Requests IWE-7 & IWE-8 Will Be Addressed at Later Date
ML20216K1681999-10-0101 October 1999
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ML18152B3391999-09-27027 September 1999
Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC
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Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6)
ML18152B3381999-09-21021 September 1999
Forwards in Triplicate,Applications for Renewal of License for Bf Jurewicz & JW Heide.Without Encls
ML18152B3331999-09-17017 September 1999
Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves.
ML18152B6671999-09-17017 September 1999
Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03
ML18152B3341999-09-16016 September 1999
Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations
ML18152B4521999-09-14014 September 1999
Forwards Comments on Review of Preliminary Accident Sequence Precursor Analysis of Operational Event That Occurred at Plant,On 980508,as Reported in LER 98-009
ML20216F1381999-09-0808 September 1999
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ML18152B4501999-09-0808 September 1999
Submits in Triplicate,Application for Renewal of License for Rd Scherer,Iaw 10CFR55.57.Requests That Certification of Medical Exam by Facility Licensee,Nrc Form 396,be Withheld from Public Disclosure,Per 10CFR2.790(a)(6)
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ML20211L9151999-09-0101 September 1999
Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity
ML18152B4471999-09-0101 September 1999
Requests That NRC Remove Listed Labels from Distribution
ML20211J2211999-08-31031 August 1999
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ML18152B3711999-08-27027 August 1999
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Forwards Revised TS Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in Ufsar.Ref to Applicable UFSAR Section Included in TS Basis
ML20138B3241999-08-23023 August 1999
Forwards Draft Response to Question 1 Re NAPS USI A-46
ML20211D9041999-08-20020 August 1999
Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal
ML18152B3651999-08-20020 August 1999
Requests Removal of License Condition from Sh Wightman Operator License SOP-21538.Updated NRC Form 396 Is Encl.Form NRC 396 Withheld,Per 10CFR2.790(a)(6)
ML18152B3661999-08-20020 August 1999
Provides Medical Status Rept for E Washington,As Required by License Conditions.Summary of E Washington Current Physical Exam & Pertinent Lab Data Attached.Encl Withheld,Per 10CFR2.790(a)(6)
ML18152B3681999-08-20020 August 1999
Submits 30-day Rept Re Two Instances in Which Conditions of Approval in Coc Were Not Observed in Making Shipment.Two Type B Shipments Using Model CNS 8-120B Package Were Made After Expiration of QA Program Approval Between 990531-0628
ML18152B3801999-08-18018 August 1999
Forwards Technical Rept NE-1206,Rev 0, Surry Unit 2,Cycle 16 Startup Physics Tests Rept, Summarizing Results of Physics Testing Program Performed After Initial Criticality on 990525
ML20211B3871999-08-17017 August 1999
Requests Permission to Routinely Discharge from SW Reservior to Waste Heat Treatment Facility Under Existing Vpdes Permit Through Outfalls 108 & 103.Discharges Are Scheduled to Commence on 990907,due to High Priority Placed on Project
ML20210T0671999-08-13013 August 1999
Informs of Completion of Review of Proposed Revs of Schedule for Withdrawal of Rv Surveillance Capsules Submitted by VEPCO on 981217.Approves Proposed Revs.Forwards Safety Evaluation
ML18152B3781999-08-13013 August 1999
Forwards ISI Summary Rept for Surry Power Station,Unit 2 for 1999 Refueling Outage.Rept Provides Summary of Examination Performed During Outage for Third ISI Interval.No New Commitments Were Made
ML18152B3751999-08-13013 August 1999
Forwards LER 99-004-00,IAW 10CFR50.73.Commitment Made by Util,Listed
ML20210Q9841999-08-12012 August 1999
Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13
ML20210Q7661999-08-0606 August 1999
Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas
ML18152B4081999-08-0606 August 1999
Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves.
ML18152B4091999-08-0505 August 1999
Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d)
ML20210N2921999-08-0505 August 1999
Discusses Which Submitted Proposed TSs Bases Change for Containment Leakage. Licensee Changes to Bases May Be Subj to Future Insps or Audits
ML20210J8861999-08-0202 August 1999
Provides Clarification to Commitment Made in Identifying Extent by Which Existing Plant Design Complied with RG 1.97,specifically Re Variable, Radiation Exposure Rate
ML18152B4001999-07-29029 July 1999
Requests Relief from Certain Impractical Requirements of ASME Section XI Code Associated with Partial Exams Conducted During 1998 Surry Unit 1 Refueling Outage.Relief Request SR-020 Encl
ML18152B3981999-07-28028 July 1999
Forwards 60-day Response to GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal. Commitments Contained in Ltr,Listed
ML20210F6121999-07-28028 July 1999
Forwards Supplemental Info on Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit,Documenting Info Provided During 990624 Meeting & Suppl Original Submittal
ML18152B3971999-07-26026 July 1999
Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02
ML18152B3961999-07-23023 July 1999
Forwards Preliminary,Uncertified License Application & Medical Certification for License to Operate Surry Power Station Units 1 & 2 for Ds Cobb.Encl Withheld,Per 10CFR2.790 (a)(6)
ML18152B3991999-07-23023 July 1999
Requests That License for Jz Laplante Be Canceled as License Is No Longer Required
ML18151A6281999-07-23023 July 1999
Forwards Revised Epips,Including Rev 19 to EPIP-4.02,rev 14 to EPIP-4.16,rev 8 to EPIP-4.21 & Rev 7 to EPIP-4.30.EP & EPIPs Continue to Meet Stds of 10CFR50.47(b)
ML18152B3931999-07-16016 July 1999
Forwards Updated NRC Form 396 & Ltr,Which Documents Medical Status of Mb Gross,License SOP-20476-2.Encl Withheld from Public Disclosure,Per 10CFR2.790(a)(6)
ML20209E7621999-07-0909 July 1999
Provides Addl Info to Justify Use of Less than One Gpm Detectable Leakage Rate to Establish Required Margin for Crack Stability in LBB Analysis,Per 980623 Application on Reactor Coolant Loop Bypass Lines
05000338/LER-1999-005, Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl1999-07-0808 July 1999
Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl
1999-09-08
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML18152B3571999-10-22022 October 1999
Requests Relief from Temporary Repair of Through Wall Leak Discovered on 30 Inch Component Cooling Heat Exchanger Discharge Pipe Associated with Service Water Sys Common to Surry Units 1 & 2
ML20217K4391999-10-18018 October 1999
Provides Response to RAI to Support USI A-46 Program Submittal for North Anna Power Station,Units 1 & 2.Rev 10 to BNL Rept 52361,encl
ML20217H3301999-10-14014 October 1999
Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr
ML18152B3581999-10-14014 October 1999
Submits Response to Violations Noted in Insp Repts 50-280/98-201 & 50-281/98-201.Corrective Actions:Visual Insps Were Completed on Accessible Coatings Inside Containment for Both Units 1 & 2
ML18152B3541999-10-12012 October 1999
Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info
ML20216K1681999-10-0101 October 1999
Forwards Vols I-VIII of Rev 35 to UFSAR for Naps.Rev Also Includes Update to Chapter 17 of Ufsar,Which Contains Operational QA Program.Changes to Program Description Do Not Reduce Commitments Contained Therein
ML18152B3401999-09-27027 September 1999
Requests That Ma Walker Be Removed from List of Individuals Scheduled to Take Exam IAW Guidance Provided in NUREG-1021, Operator Licensing Exam Stds for Power Reactors
ML18152B3391999-09-27027 September 1999
Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC
ML20212G5091999-09-22022 September 1999
Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6)
ML18152B3381999-09-21021 September 1999
Forwards in Triplicate,Applications for Renewal of License for Bf Jurewicz & JW Heide.Without Encls
ML18152B6671999-09-17017 September 1999
Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03
ML18152B3331999-09-17017 September 1999
Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves.
ML18152B3341999-09-16016 September 1999
Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations
ML18152B4521999-09-14014 September 1999
Forwards Comments on Review of Preliminary Accident Sequence Precursor Analysis of Operational Event That Occurred at Plant,On 980508,as Reported in LER 98-009
ML18152B4501999-09-0808 September 1999
Submits in Triplicate,Application for Renewal of License for Rd Scherer,Iaw 10CFR55.57.Requests That Certification of Medical Exam by Facility Licensee,Nrc Form 396,be Withheld from Public Disclosure,Per 10CFR2.790(a)(6)
ML18152B4471999-09-0101 September 1999
Requests That NRC Remove Listed Labels from Distribution
ML20211L9151999-09-0101 September 1999
Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity
ML20211H4131999-08-27027 August 1999
Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint
ML18152B3711999-08-27027 August 1999
Forwards LER 99-005-00,per Plant TS Table 3.7.6.Rept Has Been Reviewed by Station Nuclear Safety & Operating Committee.Commitment Made by Util,Listed
ML18152B3851999-08-23023 August 1999
Forwards Revised TS Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in Ufsar.Ref to Applicable UFSAR Section Included in TS Basis
ML20138B3241999-08-23023 August 1999
Forwards Draft Response to Question 1 Re NAPS USI A-46
ML20211D9041999-08-20020 August 1999
Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal
ML18152B3681999-08-20020 August 1999
Submits 30-day Rept Re Two Instances in Which Conditions of Approval in Coc Were Not Observed in Making Shipment.Two Type B Shipments Using Model CNS 8-120B Package Were Made After Expiration of QA Program Approval Between 990531-0628
ML18152B3661999-08-20020 August 1999
Provides Medical Status Rept for E Washington,As Required by License Conditions.Summary of E Washington Current Physical Exam & Pertinent Lab Data Attached.Encl Withheld,Per 10CFR2.790(a)(6)
ML18152B3651999-08-20020 August 1999
Requests Removal of License Condition from Sh Wightman Operator License SOP-21538.Updated NRC Form 396 Is Encl.Form NRC 396 Withheld,Per 10CFR2.790(a)(6)
ML18152B3801999-08-18018 August 1999
Forwards Technical Rept NE-1206,Rev 0, Surry Unit 2,Cycle 16 Startup Physics Tests Rept, Summarizing Results of Physics Testing Program Performed After Initial Criticality on 990525
ML18152B3781999-08-13013 August 1999
Forwards ISI Summary Rept for Surry Power Station,Unit 2 for 1999 Refueling Outage.Rept Provides Summary of Examination Performed During Outage for Third ISI Interval.No New Commitments Were Made
ML18152B3751999-08-13013 August 1999
Forwards LER 99-004-00,IAW 10CFR50.73.Commitment Made by Util,Listed
ML20210Q9841999-08-12012 August 1999
Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13
ML20210Q7661999-08-0606 August 1999
Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas
ML18152B4081999-08-0606 August 1999
Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves.
ML18152B4091999-08-0505 August 1999
Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d)
ML20210J8861999-08-0202 August 1999
Provides Clarification to Commitment Made in Identifying Extent by Which Existing Plant Design Complied with RG 1.97,specifically Re Variable, Radiation Exposure Rate
ML18152B4001999-07-29029 July 1999
Requests Relief from Certain Impractical Requirements of ASME Section XI Code Associated with Partial Exams Conducted During 1998 Surry Unit 1 Refueling Outage.Relief Request SR-020 Encl
ML18152B3981999-07-28028 July 1999
Forwards 60-day Response to GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal. Commitments Contained in Ltr,Listed
ML20210F6121999-07-28028 July 1999
Forwards Supplemental Info on Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit,Documenting Info Provided During 990624 Meeting & Suppl Original Submittal
ML18152B3971999-07-26026 July 1999
Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02
ML18152B3991999-07-23023 July 1999
Requests That License for Jz Laplante Be Canceled as License Is No Longer Required
ML18152B3961999-07-23023 July 1999
Forwards Preliminary,Uncertified License Application & Medical Certification for License to Operate Surry Power Station Units 1 & 2 for Ds Cobb.Encl Withheld,Per 10CFR2.790 (a)(6)
ML18151A6281999-07-23023 July 1999
Forwards Revised Epips,Including Rev 19 to EPIP-4.02,rev 14 to EPIP-4.16,rev 8 to EPIP-4.21 & Rev 7 to EPIP-4.30.EP & EPIPs Continue to Meet Stds of 10CFR50.47(b)
ML18152B3931999-07-16016 July 1999
Forwards Updated NRC Form 396 & Ltr,Which Documents Medical Status of Mb Gross,License SOP-20476-2.Encl Withheld from Public Disclosure,Per 10CFR2.790(a)(6)
ML20209E7621999-07-0909 July 1999
Provides Addl Info to Justify Use of Less than One Gpm Detectable Leakage Rate to Establish Required Margin for Crack Stability in LBB Analysis,Per 980623 Application on Reactor Coolant Loop Bypass Lines
05000338/LER-1999-005, Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl1999-07-0808 July 1999
Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl
ML18152B4401999-07-0101 July 1999
Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co
ML18152B4371999-06-24024 June 1999
Forwards Response to NRC Request for Clarification of Relief Requests Submitted on 990212,requesting Relief from Performing Hydrostatic Testing for Certain Small Diameter Class 1,RCS Pressure Boundary Connections
ML18152B4361999-06-22022 June 1999
Forwards Response to RAI Re Surry & North Anna Power Stations,Units 1 & 2,per GL 96-06
ML20196F1151999-06-22022 June 1999
Forwards Relief Requests NDE-047 & NDE-048 for North Anna Power Station,Unit 1 Re ASME Section XI ISI Program
ML20212J2951999-06-22022 June 1999
Forwards Corrected Markup & Typed Version of Affected Pages. Requests That Attached Pages for Those Previously Provided in 990506 Submittal Be Replaced & Incorporated Into NRC Review of Proposed TS
ML20195J7011999-06-15015 June 1999
Forwards Revised EPIP 2.01 Which Corrects Typo That Was Found in Step 10 of Procedure.Rev Does Not Implement Actions That Decrease Effectiveness of EP
ML18152B4301999-06-0303 June 1999
Informs of Util Intention to Revise Schedule for Submittal of License Renewal Applications for Surry & North Anna Power Stations to March 2002
1999-09-08
[Table view]
Category:UTILITY TO NRC
MONTHYEARML18153C3661990-09-20020 September 1990
Forwards Topical Rept VEP-NE-3-A, Qualification of WRB-1 CHF Correlation in VEPCO Cobra Code.
ML20065D5671990-09-20020 September 1990
Forwards Rev 15 to Security Personnel Training & Qualification Plan.Rev Withheld
ML18153C3701990-09-18018 September 1990
Forwards Addl Info Re Facility Containment Isolation Valve Type C Test,Per 900914 10CFR50,App J Exemption Request
ML18152A2341990-09-14014 September 1990
Requests Exemption from 10CFR50,App J Section III.D.3 Re Local Leak Rate Testing During Every Reactor Shutdown. Basis & Justification for Exemption Encl
ML20059J3341990-09-13013 September 1990
Forwards Rev 15 to Nuclear Security Personnel Training & Qualification Plan.Rev Withheld
ML20059H7391990-09-10010 September 1990
Forwards Rev 15 to ISFSI Nuclear Security Personnel Training & Qualification Plan.Rev Withheld
ML20059H2931990-09-0707 September 1990
Advises of Review & Approval of ASME Section XI Inservice Insp Program Interval 1 Repair & Replacement Program Change to Applicable Code Edition & Addenda
ML20064A5201990-09-0707 September 1990
Responds to NRC Re Violations Noted in Insp Repts 50-338/90-15 & 50-339/90-15.Corrective Actions:Procedures Revised to Include Tolerance Bands for Temp Control Switch Setpoints & Involved Personnel Counseled
ML18151A2901990-08-31031 August 1990
Forwards Rev 10 to Updated FSAR for Surry Power Station Units 1 & 2,representing Second Updated FSAR Submitted This Yr
ML20059E5481990-08-31031 August 1990
Forwards Supplemental Info Re Deletion of Pressurizer Safety Valve Requirement During Mode 5.Current Tech Spec Requiring One Operable Safety Valve in Mode 5 Unnecessary & May Be Deleted
ML18153C3451990-08-29029 August 1990
Forwards Proprietary Semiannual Fitness for Duty Program Performance Data Rept for 900103-0630.Rept Includes Summaries of Mgt Sanctions Imposed,Actions Taken to Correct Program Weaknesses & Events Reported to Nrc.Encl Withheld
ML18153C3381990-08-22022 August 1990
Responds to NRC 900723 Ltr Re Violations Noted in Insp Rept 50-280/90-21 & 50-281/90-21.Corrective actions:as-found-as- Left Conditions of Auxiliary Feedwater Evaluated & Found Operable
ML18153C3391990-08-22022 August 1990
Requests Approval for Use of Plugs Fabricated of nickel- chromium-iron Uns N-06690 Matl (Alloy 690) to Plug Tubes in Steam Generators for Mechanical & Welded Applications
ML20063Q1241990-08-15015 August 1990
Provides Supplemental Info for 900115 Tech Spec Change Request Re Max Allowable Control Rod Drop Time.Retained DNBR Margin Tracked & Documented by Calculational Note
ML20058N5941990-08-0808 August 1990
Responds to NRC Re Violations Noted in Insp Repts 50-338/90-11 & 50-339/90-11.Corrective Actions:Procedure Revised to Limit Applicablity to Small Refueling Cavity Leaks & Listed Evaluations & Revs Will Be Made to EOP
ML18153C3171990-08-0101 August 1990
Resubmits Synopsis of Changes to Updated Operational QA Program Topical Rept Vep 1-5A
ML18153C3161990-08-0101 August 1990
Provides Supplemental Response to NRC 900629 Ltr Re Electrical Crossties,Load Shedding on Nonblackout Unit & Emergency Diesel Generator Reliability.Emergency Diesel Generator Reliability Program in Place,Per Reg Guide 1.155
ML18153C3091990-07-30030 July 1990
Provides Outline of Plan to Meet 10CFR50 App G Requirements, for Low Upper Shelf Energy Matls,Per NRC 900521 Request
ML18153C3061990-07-30030 July 1990
Forwards Revised Tech Spec Pages,Addressing Constitution of Quorum & Timeliness of Mgt Safety Review Committe Meeting Minutes,Per NRC Request
ML18153C3101990-07-26026 July 1990
Forwards Decommissioning Financial Assurance Certification Rept..., Nuclear Decommissioning Trust Agreement & Nonqualified Nuclear Decommissioning Trust Amended & Restated Trust Agreement, Per 10CFR50.75
ML18153C3051990-07-26026 July 1990
Advises That Util Submitted Decommissioning Funding Plan & Financial Assurance Info W/Isfsi License Application
ML18153C3041990-07-26026 July 1990
Responds to NRC 900626 Ltr Re Violations Noted in Insp Rept 50-281/90-20.Corrective Actions:Leaking Drain Plug & Upper Drain Plug on Motor Replaced W/Oil Drain Assemblies Composed of Piping & Valves
ML18153C3031990-07-26026 July 1990
Advises of Withdrawal of Request for NRC Review & Approval of Engineering Evaluation 8.Revised Evaluation Will Be Maintained Onsite for NRC Audit During Future Insps,Per Generic Ltr 86-10
ML20055H5651990-07-20020 July 1990
Requests for Temporary Waiver of Compliance from Requirement of Tech Spec 3.3.1.1,Table 2.3-2 Re Response Time Testing. Request Does Not Pose Significant Hazards Consideration
ML18153C2901990-07-18018 July 1990
Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Listed Transmitters Compiled.Transmitters Found Installed within Reactor Protection or ESFAS Have Been Replaced
ML20055F9361990-07-16016 July 1990
Advises That No 30-day Rept Will Be Made Re Degradation of Seismic Support Bolts of post-accident Hydrogen Recombiners 1-HC-HC-1 & 2-HC-HC-1
ML20044B3501990-07-13013 July 1990
Forwards Monthly Operating Repts for June 1990 & Corrected May 1990 Rept Pages for North Anna Power Station
ML18153C2861990-07-12012 July 1990
Requests Cancellation of Operator Licenses for Listed Individuals.Licenses No Longer Required
ML18153C2871990-07-11011 July 1990
Responds to Violations Noted in Insp Repts 50-280/90-18 & 50-281/90-18.Corrective Actions:Permanent Drain Line Installed & Matrix Which Describes Proper Ventilation Alignment for Plant Conditions Provided for Personnel Use
ML20055E8121990-07-0909 July 1990
Requests Cancellation of Licenses Listed
ML18153C2851990-07-0606 July 1990
Forwards Response to Generic Ltr 90-04 Re Status of Generic Safety Issues
ML18153C2831990-07-0303 July 1990
Advises That MW Hotchkiss No Longer Needs Operator License SOP-20548-1.Cancellation of License Requested
ML18153C2821990-07-0303 July 1990
Requests Exemption from 10CFR50,App J,Paragraph III.A.6(b), Which Requires That When Two Consecutive Periodic Type a Tests Fail to Meet Applicable Acceptance Criteria Type a Test Shall Be Performed at Plant
ML20055D4841990-07-0303 July 1990
Requests That Operator Licenses for Listed Individuals Be Cancelled.Licenses No Longer Required
ML20055D4831990-07-0303 July 1990
Requests That Operator License for Bl Shriver Be Cancelled. License No Longer Required
ML20055D7701990-06-29029 June 1990
Submits Supplemental Response to Violations Noted in Insp Repts 50-338/90-04 & 50-339/90-04 Re Failure to Take Timely Corrective Action for Repair of B Low Head Safety Injection Pump (1-SI-P-1B).No Addl Purchasing Procedure Required
ML18153C3591990-06-28028 June 1990
Responds to SALP Repts 50-280/90-16 & 50-281/90-16 for Period 890701-900331.Corrective Actions Focus on Issues of Maint Backlog,Maint planning,post-maint Testing,Staffing & Procurement
ML18153C2611990-06-21021 June 1990
Responds to NRC 900522 Ltr Re Violations Noted in Insp Repts 50-280/90-07 & 50-281/90-07.Corrective actions:as-built Configurations of 120-volt Ac & Dc Vital & Semivital Panel Breaker Installations Verified to Be Acceptable
ML20043H9421990-06-21021 June 1990
Forwards Proprietary WCAP-12351, North Anna Unit 1 Evaluation for Tube Vibration Induced Fatigue, Per Bulletin 89-001, Failure of Westinghouse Steam Generator Tube Mechanical Plugs. Rept Withheld (Ref 10CFR2.790)
ML18153C2591990-06-18018 June 1990
Forwards Response to NRC 900524 Request for Addl Info Re NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Engineering Will Initiate Study to Evaluate Enhancements of Cooldown & Possible Heatup Operation
ML18153C2581990-06-18018 June 1990
Forwards Reissued Semiannual Radioactive Effluent Release Rept,Jul-Dec 1989. Rept Contains Info Re SR-89,Sr-90 & Fe-55 Analytical Results for Liquid Composite Samples
ML20055C8661990-06-15015 June 1990
Forwards Application for Reissuance of NPDES Permit
ML18153C2541990-06-15015 June 1990
Forwards Corrected Tech Specs Page 3.1-3,per Identification of Typo in 900522 Application for Amends to Licenses DPR-32 & DPR-37
ML20043G5711990-06-15015 June 1990
Forwards Response to Request for Addl Info Re Insp Rept 50-339/89-20 Concerning Plant Safety Sys Outage Mod Insp. Response Withheld (Ref 10CFR73.21)
ML20043H1131990-06-15015 June 1990
Concludes That Removal of Temporary Loose Parts Monitoring Sys Does Not Increase Probability of Acccident,Create Possibility of New Kind of Accident & Does Not Reduce Margin of Safety
ML20043G8731990-06-14014 June 1990
Discusses Evaluation of ASME Section XI Testing for Outside Recirculation Spray Pumps
ML18153C2521990-06-14014 June 1990
Responds to NRC 900515 Ltr Re Violations Noted in Insp Repts 50-280/90-09 & 50-281/90-09.Corrective Actions:Abnormal Procedures & Fire Contingency Action Procedures Being Upgraded Via Technical Procedure Upgrade Program
ML20043H2091990-06-13013 June 1990
Forwards List of Corrective Actions & Current Status in Response to NRC Bulletin 88-004, Potential Safety-Related Pump Loss.
ML20043F4531990-06-0808 June 1990
Forwards, Core Surveillance Rept for North Anna 2,Cycle 7 W/Extended License Burnup Limit. Rept Contains Specific Power Distribution Limits Applicable for Cycle 7 Operation W/Burnup Limit Extension
ML18153C2501990-06-0808 June 1990
Confirms That Primary Policy Re Onsite Property Damage Insurance,Provided by Nuclear Mutual Limited
1990-09-07
[Table view] |
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
- 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
