ML20032A085
| ML20032A085 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 10/23/1981 |
| From: | Moon C NRC |
| To: | DOHERTY, J.F. |
| References | |
| NUDOCS 8110280273 | |
| Download: ML20032A085 (18) | |
Text
{{#Wiki_filter:. 10/23/81 UNITED STATES OF A!! ERICA NUCLEAR REGULATORY COMt11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of HOUSTON LIGHTING & POWER CO. ) Docket No. 50-466 (Allens Creek Nuclear Generating 1 /p(Tr se i j{ flL b.y $ g, Station, Unit 1) ) g 3 V fl OCT2 719815 $ INTERVENOR DOHERTY'S FIRST SET OF v.s.maua mouuson
- INTERR0GATORIES TO STAFF ON CONTENTION #21 %.
mamm /h g}' The NRC Staff responds as follows to Mr. Doherty's firsthe qf interrogatories regardtng Contention No. 21. Interrogatory No.1 How much reactivity is calculcted to be introduced from the pressure spike observed as a result of the following transients? In addition, please give the per-cent void collapse if it is determined. a. Main Steam Isolation Valve Closure. b. Turbine trip with 10% bypass. c. Electric Load Rejection without bypass. d. Recirculation Control Failure-Increasing Flow e. Loss of offsite power. f. Turbine trip without bypass. g. Pressure regulator fails Closed.
Response
The following figures from the GESSAR II Nuclear Island are enclosed: a. Figure 15.2-6 MSIV Closure, Position Scram b. Figure 15.2-4 Turbine Trip with Bypass On Trip Scram c. Figure 15.2-3 Generator Load Rejection, without Bypass DESIGNAT ORIGINAL 8110280273 811023 DR ADOCK 05000 Certified By f( Of S*In 1
d. Figure 15.4-2 Fast Opening of One Recirc Valve at 30%/sec Figure 15.4-3 Fast Opening of Both Recirc Valves at 11%/sec e. Figure 15.2-9 Less of All Grid Connections f. Figure 15.2-5 Turbine Trip without Bypass Trip Scram g. Figure 15.2-1 Pressure Regulator Downscale Failure Reactivity versus time is shown on each of these figures, except for g. (Figure 15.2-1), which shows void fraction versus time. Although these data were computed using the REDY code instead of the ODYN code, the results are typical of those expected for Allens Creek using the ODYN code. Interrogatory No. 2 What was the power core density of the Peach Bottom-II plant when the turbine trip tests (1977) were done as reported in EPRI-NP-5637 R; ponse The design core power density at Peach Bottom at the time of the tests was about 50 kw/ liter. The three tests were run from initial conditions which corresponded to 23.7, 30.8 and 34.6 kw/ liter, respectively. Interrogatory No.3 Does Staff believe this Intervenor through this issue has raised Task Action Plan (TAP) B-197 What differences are there if not?
Response
Task No.19 involves the development of the analytical methods necessary for the Staff to perform independent calculations to check vendor analyses of thermal-hydraulic stability. The Applicant's thermal-hydraulic stability analysis is described.in Section 4.4.4.6 of the PSAR. I m
The Staff does not understand the contention to raise questions pertaining to thernal-hydraulic stability analytical methods. Interrogatory No. 4 Why was the Applicant for the LaSalle County nuclear plant not required to analyze the MSIV closure with CDYN but allowed to use the feedwater con-troller and turbine events for thermal limit determination? This party believes the MSIV closure transient is the most severe. Note: See p. 4-23 of NUREG-0519, $ER for LaSalle Cty. Plant.
Response
The Staff required (pg.15-5, NUREG-0519) 0DYN reanalyses for: (1) For Thermal Limit Evaluation a) Feedwater Controller failure-maximum demand b) Generator load rejection without bypass c) Turbine trip without bypass d) Main stehm isolation valve closures (2) For Overpressure Protection a) Main steam isoltion valve closure with position switch scram failure For thermal margin considerations, all of the events in category (1) above are bounded by the rod withdrawal error at power. Interrogatory No. 5 Why was the REDY code rejected by the NRC?
Response
The REDY code has not been rejected by the NRC. It is still used for most transient analyses. The 0DYN Codefis better for overpressure type transients and we thus require ODYN for overpressure transients. 8 _2_
Interrogatory No. 6 Have any Commission licensed energy producing reactors ever been de-rated due to suspected inability to insert negative reactivity sufficient to assure no nore than.1% of the fuel rods go into boiling transition?
Response
Yes, on a downward ramp of power vs. time near the end of a fuel cycle. Interrogatory No. 7 What (if any) conservatisms are built into the ODYH code?
Response
The ODYN code is a best estimate code. Conservatisms are introduced through conservatisms in the MCPR calculation and in conservative input parameters. Interrogatory No. 8 Does Staff believe power core density is an unimportant variable in the calculation of reactivity insertion by collapse of voids? On what documents does Staff rely?
Response
If by core power density intervenor means the average value in kw/ liter, then core power density is not directly a factor in the amount of reactivity inserted by void collapse. That amount is the product of the void coefficient of reactivity and the change in void volume resulting from the collapse. How-ever, the changes in design of reactor which pa:, nit a higher average core power density may also change the void coefficient of reactivity. In any event the change will be small. No specific documents are relied upon for this answer. 9 4
Interrogatory No. 9 a. What is a Haling mode of operation? b. Why is Haling mode of operation an important assumption sensitive for ODYN modeling? (See Attachment to Staff Response relevant to Contention #21 of 7/13/81)
Response
a. The Haling mode of operation refers to the operation of the reactor 50 that the power distriDution is the so-called Haling distribution. This distribution is based on the Haling principle which states that for any oower peaking factor is maintained given set of end of cycle conditions ti A at the minimum value when the power shape does not change during the operating cycle. The design end of power condition is that the core should be critical at full power with all rods withdrawn. The Haling power shape can be deter-mined by iterating between power distribution and exposure distribution un-til a consistent set is achieved with the core critical at full power. During the cycle control rods are manipulated so as to maintain this shape. A brief discussion of the Haling strategy is presented in Appendix 4.A to Chapter 4 of the PSAR (Amendment No. 56, March,1981). b. The Haling mode of operation is not an important assumption with regard to 0DYN modeling, i.e., no specific power shapes were defined as an integral part of the ODYH code. However, the results of an overpressurization transient are dependent on initial core power shape. In particular, at end of cycle where such transients tend to be limiting, the axial power shape is important. Vendor analyses have shown that assumption of a Haling power distribution at end of cycle leads to larger consequences for overpressuri-zation events than does assumption of power distributions that actually occur in practice. This result is due to the fact that the actual burnup at the a e
bottom of the core is less than that required by the Haling distribution. This means that the power is higher in the bottom of the core than would be the case for the Haling distribution. Thus negative reactivity is inserted sooner when the rods go in 0; scram than Nould be the case with the Haling distribution and the consequences of cha transient are thus reduced. Thus, as indicated in Table IV of the Staff response on Contention 21, dated July 13, 1981, the assumption of a Haling distribution is recommended as an input parameter to the ODYN calculation. Interrogatory No. 10 What is the GETAB safety limit for all transients stated in the Minimua Critical Power Ratio (MCPR) " units"?
Response
On page 15.1.6 of the PSAR, the Applicant states that maintaining MCPR greater than 1.07 is a sufficient, but not necessary condition to assure that no fuel damage occurs. This value is similar to values found acceptable in current operating license reviews. Interrogatory No. 11 Does ODYH use the SCAT code to calculate the change in CPR for transient events?
Response
No, the SCAT code used transient results from 0DYN or REDY to calculate changes in CPR. r t m
Interrogatory No.12 Does Staff believe that the ODYN code will accurately predict the pressure and power spike above 69% power for.the turbine trip event? Were any Peach Bottom tests at higher power conditions?
Response
Yes. The Staff believes that the ODYN code will accurately predict the pressure and power spike above 69% of power for the turbine trip event. The highest power level for the Peach Bottom turbine trip tests was 69.1% of full power. 't Interrogatory No. 13 a Will the HEKIH code be cited in Staff testimony on this Contention?
Response
No. a t-o 8
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I UllITED STATES OF A!1 ERICA tiUCLEAR REGULATORY C0:lilISSION BEFORE THE AT0i11C SAFETY AND LICEllSING BOARD ~ In the I-latter of ) i ) HOUSTO'l LIGHTIliG &-POWER C0. Docket No. 50-466 (Allens Creek fluclear Generating ) Station, Unit 1) ) i 1 AFFADAVIT OF CALVIN 11. !!00N k i I, Calvin 11. !!oon, do hereby depose and say under oath:- 1. I ara the Licensing Project lianager for the captioned application.- 2. The. foregoing liRC Staff's responses to interrogatories propounded by John F. Doherty were prepared by me or under qy supervision. 3. I certify that the answers i;i.en are true and correct to the best of my knowledge, information and belief. N MM Calvin W. ftoon ~ Subscribed and sworn to before me this& day of@Rn 3 , 1981. ~. 0?t$f l]. 2l[ YJ ttio'tary Public d ity Coa..ission Expires: M /,/064 .O y
UNITED STATES OF A!1 ERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ~ HOUSION LIGHTING AND POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating ) Station, Unit 1) ) CERTIFICATE OF SERVICE I hereby certify that copies of "INTERVENOR DOHERTY'S FIRST SET OF INTER-R0GATORIES TO STAFF ON CONTENTION #21" and " AFFIDAVIT OF CALVIN W. MOON" in the above-captioned proceeding have been served on the for.owing by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commission's internal mail system, this 23rd day of October,1981: Sneldon J. Wolfe, Esq., Chairman
- Administrative Judge Susan Plettman, Esq.
Atomic Safety and Licensing David Preister, Esq. Board Panel Texas Attorney General's Office U.S. Nuclear Regulatory Commission P.O. Box 12548, Capitol Station Washington, DC 20555 Austin, TX 78711 Dr. E. Leonard Cheatum Administrative Judge Hon. Jerry Sliva, Mayor Route 3, Box 350A City of Wallis, TX 77185 Watkinsville, Georgia 30677 Hon. John R. Mikeska Mr. Gustave A. Linenberger* Austin County Judge Administrative Judge P.O. Cox 310 Atomic Safety and Licensing Bellville, TX 77418 Board Panel U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. John F. Dohertv 4127 Alconbury Street ine Honorable Ron Waters houston, TX 77021 State Representative, District 79 3620 Washington Avenue, No. 362 Mr. William J. Schuessler Houston, TX 77007 5810 Darnell l Houston, TX 77074 J. Gregory Copeland, Esq. Baker & Botts One Shell Plaza i Houston, TX 77002 i t
Jack Newman, Esq. D. Harrack Lowenstein, Reis, Newman & 420 Mulberry Lane Axelrad Bellaire. TX 77401 1025 Connecticut Avenue, N.W. Washington, DC 20037 Texas Public Interest Research Group, Inc. Brenda A. McCorkle c/o James Scott, Jr., Esq. 6140 Darnell 13935 Ivymount Houston, TX 77074 Sugarland, TX 77478 Mr. Wayne Rentfro Rosemary N. Lemmer P.O. Box 1335 11423 Oak Spring Rosenberg, TX 77471 Houston, TX 77043 Lootis Johnston Carro Hinderstein 1407 Scenic Ridge Houston Bar Center Houston, TX 77043 723 Main Suite 500 ^ Houston, TX 77002 Margaret Bishop U.S. Nuclear Regulatory Commission J. Morgan Bishop Region IV, I&E 11418 Oak Spring 611 Ryan Plaza Drive, Suite 1000 Houston, TX 77043 Arlington, TX 76011 Stephen A. Doggett, Esq. Bryan L. Baker Pollan, Nicholson & Doggett 1923 Hawthorne P.O. Box 592 Houston, TX 77098 Rosenberg, TX 77471 Robin Griffith Carolina Conn 1034 Sally Ann 1414 Scenic Ridge Rosenberg, TX 77471 Houston, TX 77043 Mr. William Perrenad Atomic Safety and Licensing 4070 Merrick Board Panel * . Houston, TX 77025 U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketing and Service Section* Office of the Secretary Atomic Safety and Licensing U.S. Nuclear Regulatory Commission Appeal Board Panel
- Washington, DC 20555 U.S. Nuclear Regulatory Commission Washington, DC 20555 l'
m i \\AA. N Stepherf M. Schinki ' Counsel for NRC Staff a. i}}