Methods of Utilization of Information From

Home , Nuclear reactor core

NEACRP-A- q23 September 18, 1980

Methods of Utilisation of Information

from

Operating Reactors

prepared by

Burt A. Zolotar Electric Power Research Institute 3412 Hillview Avenue Palo Alto, CA 94303 U.S.A.

000600()1 Reactor Core Physics Desi n and Operating Data for Cycles 1 and 2 o 7 the Zion Unit 2 FWR Power Plant

t-JP-1232 Research Project 519-6

fine.1 Report, December 1979 a Prepared by

CARNEGIEMElLON UNIVERSITY Nuclear Science and Engineering Divisian Schenley Park Pittsburgh, Pennsylvania 15213

Principal investigators Albert J, Impink, Jr. B. Alan Guthrie Ill

Prepared for

Electric Power Research Institute 3412 Hilfview Avenue Palo Alto, California 94304

EPRI Projecr Manager Walter J. Eich Nuclear Power Division

00060002 ABSTRACT

This report contains a set of design and operating data relevant to Cycles 1 and 2 of the Zion Station Unit 2 pressurised water reactor (PWR). In general, these data constitute a substantial enhancement of fundamental information potentially usable in the process of validating PWR core analysis methodology. In particular, these data constitute the basis required for analysis that would allow specific compari- 0’ sons with the results of gaaana scans of selected assemblies at various axial posi- tions measured after Cycle I. These gasaaa scans are reflective of the local power history prior to measurement.

The design data collected herein is limited to the nuclear aspects of the core; thermal, hydraulic, and structural properties are sunanarized only insofar as they would be required for a nuclear analysis. The operational data includes core physics parameters measured at,the startups of Cycles I and 2 as well as the operating history over almost a four-year period. Measurements related both to the core power distribution as well as to core responses during certain maneuvers are included.

0006OQQ3 16 1600

All other banks out

Hot, zero power 1400

1200

...... \ l d .

. 400 0 . .

2 . zoo

. Grid Locations , I +I 1 I I 4 I 4 20 40 60 80 100 120 140 160 180 200 220 Steps withdrawn Oiagram 3.1.1-l Differential and Integral Norths of Control Bank II at Startup of Zion Unit 2, Cycle 1

3-5 00060004 -. 3-97 0006OOOfj GAMMA SCAN MEASUREMENTS AT ZION STATION UNIT2 FOLLOWING CYCLE 1

EPRI NP-509 (Research Project 130)

Final Report

October 1977

Prepared by

Nuclear Energy Systems Divbion GENERAL ELECTRIC COMPANY 175 Cwtner Avenue San Jose, California 95125

PRINCIPAL INVESTIGATORS G. R. Parkas G. F. Valbv

Prepared for

Electric Power Research lntiitute 3412 Hillview Avenue Palo Alto, California 94304

EPRI Project Manager Robert N. Whitesel

00060006 .

1. INTRODUCTION AND SUMMARY

An extensive gamma scan measurement of irradiaied fuel was performed following Cycle I at Zion Station Unit 2 as part of EPRI Project FiP-130, “Nuclear Reactor Core SenchmarkDala.“lheobjectiveof thesemeawrementz+was toprovide power distribution benchmark data for verification of those methods wed for predictive caiculationsandin-core monitoring of core power distributions. Combination of the benchmark data reported here and the design details and operating history of the core allows evaluation of the accuracy of power distribution calculations. These measurements constitute a detailed benchmark for power distribution in a PWR addressing the qu.estions of gross core shape, detailed axial shapa, core symmetry, and pobver sharing among assemblies of differing enrichments. Successful completion of a program of this magnitude is a resutt of excalient cooperation between Commonwealth Edison, EPRI, and General Electric. EPRI- $ponsored programs such as this will provide the users and manufacturers of nuclear supply systems with valuable benchmark data to imfxove performance and availability, and increase margin, by reducing uncertainties in the design and operation of plants.

Relative La-14Oactivity. reflective of recent power history, was measured for37 tuel assemblies at 12 to 24 elevations. A sodium iodide gamma ray detector, mounted inside a 3000-lb moveable collimator, was used to sequentially measure the relative intensity of the IS96 keV La-140 gamma ray at each elevation on the fuel. A minicomputer-based data acquisition system controlled the sequencing and reduced the data. From consideration of the accuracy of the measured intensities, small inherent biases in the technique, and the process by which calculated power distributions can be converted to La-140 0 distributions, it is concluded that benchmarking of power distribution methods within 4% is reasonably achievable from the data resulting from this program.

00060007 ZIQN-2 EOC1 BENCHMARK GAMMA SCAN FEBRUARY 1977 H ci F E 0 C 4. I *. + C50 7553ao.ct2 752335. o.a73 o.a70

. Al6 * . -I-A22 C&l aa7117. ma4a. 849601. co2 C63 alam. 640333. 0.992 0.946 0.740 I .025 1.120

. . . / A42 a59 A33 a21 / a55377 926263. a9362a. 913171. / 0.9a9 I.071 1.033 1.066 /’

/ AOf BT5 / as6276 946607. /

. OCTANT BOUNOARV A23 aa

/ / / . 1 655 A03 a34 911991. 863926. ,9la460. LEGEND: + EDGE ASSEMaLY. MAS”RE0 LA.140 f.054 0.999 1x62 MULTW’LIEO 6,’ 1.03 6EFORE NORMALZATION

* FL”X MAPPING lNSTR”MENTATlON !N THIS ASSEMBLY LOCATION A29 r-4 a59973. 2 12 ASSEMBLY SERIAL NO. XxXxXx UNNORMALIZED LA.140 0.994 x.xXx NORMAL!ZEO LA.740 A”ERAGE O”ER 28 ASSEM6LIES I634 EXCLUOEOP = 1.0

Figure A-6. Planar La-140 Distribution at 69 inches Above &ttom of Active Fuel

A-7 00~060008 ZION.2 EOC1 BENCHMARK GAMMA SCAN FEBRUARY 1977 . A51 922596. pig-fq~~~ ir

. 644 A59. *. + c46 910733. 9Lm*,, 613475. 0.709 I.053 I.249 1 t-

Figure B-6. Pknar la-740 Distribdon at 69 Inches Above &ttom of Active Fuel

B-7 CORE DESIGN AND OPERATING DATA FOR CYCLES 1 AND 2 OF QUAD CITIES 1

EPRI NP-246 (Research Project 497-l)

November 1976

Rincipal Investigators N. Ii. LWWI G. R. Parkas

Prepared for

Electric Power Researck Institute 2412 Hillview Avenue Palo Alto, California 94304

Project Manager Burt A. 2obtar 00060010 ABSTRACT

This report contains due design and operating data needed to define due fuel chae: teristics and reactor operation characteristics for Cyclas 1 and 2 of the fluad Cities 1 reactor. The purposa is tq pmvide reference quality data for use in the qualification of reactor cqre analysis metiqds and tq provide tie basis for the assessment of the irradiation envinwment of the plutonium recycle assemblies prasent.

The design data includes fuel assembly description, ewe component arranwments, control rod descriptions and cqra loading pattarns. Uydrauiic characteristics of tie assemblies and the inlet orifices are also provided. Operating data is compilad for 16 steady-state points during Cycle 1 and 13 during Cycle 2. Each state wint includes core average expqwre, tierma power, pressure, flux. inlet wbeoqling, cqn- tml configumtion and axial in-corn det&qr readings. In addition, benchmark cold critica/ data is wecified.

00060011 DATASET 03, JUNE &lS73

Reactor Cwdttiom

Ccfe Average Exp~we, 3636.0 MWd/t Core Thermal Power, 2320.16 h4W Core Pressure, P, 1032.24 psia Core Flow, 34.72 Mlb/hr inlet Subcwliw at P, 22.65 BWlb

Control ConfIguration

Legend: 48, Full Chit 0, Full In.

4B 40 40 40 40 40 48 48 40 48 48 40 40 48 48 48 40 4B 0 40 48 4B40 4836 40IQ 4020 406 4032 486 4820 4810 4036 40 40 48 40 48 840 440 0 40 8 48 4 48 0 40 48 48 48 48 20 48 36 48 20 40 36 48 7.0 48 48 40 40 40 8 40 0 40 4 48 4 48 a 48 8 40 48 48 20 40 36 48 24 40 36 40 24 48 36 4a 20 48 48 4a 0 40 0 48 6 48 6 48 a 48 0 48 40 48 28 40 36 48 24 48 36 48 24 40 36 40 28 48 48 48 8 40 0 40 4 48 440 84% 0 40 48 40 40 48 20 48 36 48 20 48 36 40 20 48 4B 40 40 48 0 40 4 40 8 4a 0 40 4 4a 0 48 40 40 48 40 36 48 20 48 32 40 20 40 36 48 40 48 40 48 40 48 10 40 6 48 6 40 10 40 40 48 48 48 40 48 48 48 40 48 48 4B 40 40 40 40 40 40

Axlai TIP Diqtribution d, 0 yc’ ! 1609 20.0 32.7 4if.5 58.4 72.0 07.4 105.4 110.9 105.0 102.1 97.5 92.3 I 04.6 03.6 82.2 74.6 72.6 75.7 75.2 69.8 66.0 53.7 39.3 25.5 ,; 2409 30.4 50.7 71.2 92.1 109.4 li8.2 125.1 130.1 123.0 119.1 117.4 110,O 102.9 106.7 110.2 110.0 106.3 100.0 100.9 09.1 06.9 76.0 56.6 39.0 3209 31.0 50.7 67.3 82.7 94.6 90.5 110.0 115.6 116.3 113.5 113.9 104.7 95.5 92.7 07.6 04.0 77.7 77.9 76.2 67.4 65.9 57.7 42.7 27.2 4009 20.7 47.3 62.9 77.9 92.5 100.9 ~100.4 109.2 97.0 91.5 93.2 09.9 05.9 09.9 94.0 95.5 91.1 91.5 80.5 05.6 78.8 04.0 45.0 27.4 4809~ 12.1 22.1 31.8 42.6 54.3 67.1 86.3 96.1 96.6 93.3 90.2 02.7 78.6 73.4 69.3 65.7 62.3 61.5 57.4 52.8 47.3 41.2 30.6 21.2 0817 30.6 53,3 71.4 00.6 100.0 104.0 90.7 102.0 94.5 09.0 02.0 01.5 75.6 70.9 69.9 60.1 04.0 63.0 62.7 50.6 50.B 5,3.1 41.6 31.7 1617 30.2 51.1 70.0 90.8 107.1 118.5 125.6 127.9 121.8 114.7 l10.3 109.3 103.4 103.3 111.6 112.1 109.9 108.1 104.5 92.3 85.2 76.0 58.0 44.2

B-22 Z.SD,.,% U22S BUNDLE AVERAGE

CONTROL ROD CORNER I-I --I-

ROO TYFE u235 w% -2’3 w,% N”MSSR OF RODS I

, 2.58 0 2s 2 134 0 lcl 3 1x4 0 S 4 1.33 0 , 6 2.56 2.50 3

t - TtE ROOS , - SPACER CAPTUREROO

Figure 4. Bundle Design for 7x7 L/O* &bad a.

940

I l-. EFFICIENCY 120

110

100

NOTCMES

L SUFICUOLING 2Kl - IO

0 llll~illll~lllll~lllflllllllllllll 7 , ,,,,,,,‘,I,~~~**~**~233333333 12346S7Se01234SS7S901234667S001234687

DAYS I GAMMA SCAN MEASUREMENTS

QUAD CITIES NUCLEAR POWER STATION UNIT 1 FOLLOWING CYCLE 2

NP-214 Uk.earch Project 1301

Final Report

July 1976

Prepared by

Nuclear Ener9y Systems Division General Electric Company 175 Cumer Avenue San he, California Sl25

Principal lnvesti9aton Martin B. Cutrone Geor9e F. Valby

Prepared for Electric Powr &sea& institute 3412 Hillview Avenue Paka Alto, Calfiornia 94304

Project Manqer Robert N. Whited 1. INTROOUCTION

An extensive gamma scan measurement of irradiated fuel was performed following Cycle 2 at Quad Cities Nuclear Power Station Unit 1 as part of two EPRI Projects : “Nuclear Reactor Core Benchmark Data, RP.130,” and “Quad Cities 1 Plutonium Recycle Nuclear and Fuel Performance Measurement, RP497.1.” The objective of these measure merits was to provide power distribution benchmark data for verification of those methods used for predictive calculations and in-axe monitoring of core power distributions. In addition, these data will be used to help assess the nuclear performance of plutonium recycle (MOzI bundles in a Boiling Water Reactor (BwRI environment. Utiiization of the benchmark data reported here requires knowledge of the design and operating history of the core. This information is contained in NP.240, Xore Design and Operating Data for Cycles 1 and 2 of Quad Cities 1,” prepared as part of RP497.1.

Relative La.140 activity, reflective of recent power history, was measured for 89 fuel bundles. The rod~to-rod distribution was measured in 5 assemblies 12 MO* and 3 UOz j by disassemblv and scanning of single rods. The rneas- urements const,itute a detailed benchmark for pov~er distribution in a BWR, addressing the questions of gross core shape, detailed axial shape, core symmetry, high exposure/reload bundle power sharing, local (rod-to.rodI distribution, and MO* performance. Successful completion of 8 progratn of this magnitude is 8 result of excellent cooperation between Commonwealth Edison, EPRI, and General Electric. EPRl.sponsored programs such as this will provide the users and manufacturers of nuclear steam supply systems with valuable benchmark data to improve performance and availability, arid increase margin, by reducing uncertainties in the design and operation of plants.

l-l 00060016 , 0 0

w. l 0 .-

!a Et r. D E F G ,,,,nF-.~. ~, ------~------WlI*E I I I x x. I I x x 1~ I I GW , ,7OY”Gl.I li’lEtY88.1 X X I 1617337. I xx I 172Il3IS.l 1734i33.1 11 I I x I I x I I I E1,NPL.E SERIRL NILIHEER GER 1SY , 1.nm I l.Eml I X).! I ..947 I xx I I.007 I l.cll5 I FOU E; FlW FLWW LRl40 DlSTFl8iJTlOti I I I x ,x I I x x I I I 2, I,,CHES QEO’VE BnT,TOll OF hf:~TIVE FIML EXPT lJtKEt?T”,tli-,’ = 1 7:: I I I I I I I 1 I lE.345X.l l63273O.l ,73fiSZ,.l lYO3WG.l ,87OSY7.I ,7.24Y3~., 164?213.I 21 I 1 I I I ‘I I I .332 I .356 I I.016 I I.114 I I w5 I I.@10 I .Y65 I T 1 I I I I I I

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Core Design and Operating Data for Cycle 1 of Hatch 1

NP-562 Research Project 130-3

Final Report, January 1979

Prepared by

GENERAL ELECTRIC COMPANY Nuclear Energy Engineering’ Division 175 Curtner Avenue San Jose, California 95125

Principal Investigators N. H. Larsen J. G. Goudey

Prepared for

Electric Power Research Institute 3412 Hillview Avenue i’aio Aito, California 94304

EPRI Project Manager f? N. Whitesel Nuclear ?ower Division

00060019 ABSTRACT

This report contains the design and operating data needed to define the fuel characteristics and reactor operation characteristics for Cycle 1 of the Hatch Nuclear Power Plant Unit 1. The purpose is to provide reference qua/ity data for use in the qualification of reactor core analysis methods and to provide the basis for the assessrnant of the irradiation environment up to and including the end of Cycle 1 gatnrna scan.

The design data incfudes fuel assembly description, core component arrangements, control rod descriptions and oore loading patterns. ffydraufic characteristics of the assemblies and the in/et orificee are a/so provided. Operating data is compiled for selected steady-state operating pbints during Cycle 1. Data listed for each state point include core average exposure, thermal power, pressure, flux, inlet subcooling, control configuration and axial in-core detector readings.

00060020 Gamma Scan Measurements at Edwin I. Hatch Nuclear Plant Unit 1 ~Following Cycle 1

NP-51 I Research Project 130-3

Final Report. August 1978

Prepared by

GENERAL ELECTRIC COMPANY~ Nuclear Energy Engineering Division 175 Curtner Avenue San Jose, California 95125

Princip~al Investigators L, M. Shiraishi G. I? Parkas

Prepared for

Electric Power Research Institute 3412 Hillview Avenue Palo Alto, California 94304

EPRI Project Manager Roberl N, Whitesel Nuclear~Power Diwsion

0006(lO21 ABSTRACT

A merxsurwnent program to obtain relative La-140 garnrna ray intensities from irradiated fuel bundles !nd rods was conducted at the Hatch 1 Reactor during the refueiing and maintenance outage at the end of the first fuel cycle. Of 106 bundles measured, 75 comprised a complete octant of the core and 35 were adjacent to partially inserted control e blades.

This report describes the measurement program and contains the gwnrna scan nwasurernent data, The data are reported as relative La.140 intensities for discrete locations along the fuel axis. The data are further formatted into various displays to accentuate particular features of the core, such as controlled bundle shapes, core radial shapes, and bundle power asymmetries. Evaluation of the measurement uncertainties yields values of 21.7% 110) for bundle nodal intensities and i2.0% (lo) for nwst of the single fuel rod intensities. , ‘;

Figure 5-3. Detailed Axial Plot of HX373 * * *

Figure5-3. Lletakl Axial Plot of HX373 CONTROL BLADE holcill4~

ROO HOLDER 4

lo-

4 0 , I I I I I I 0 2-l 40 Eo a0 100 li?J 1, ELEVATION AEOVE KTITOM OF ACTWE FUEL hChd

Figure 54. Oetaiied Axial Plots of Two Rods in i-IX.973

00060025 I? B c D E F G 7 6 5 4 3 2 I

------,, ,! ,.I83 l.E53 *.n3* I.062 .564 ! .BBl .556 .4a4 .258 .266 !I0 , I I! 21 1 .m4 l.BY6 I.054 I.185 ,.I33 I.,,8 .352 I .a31 .B27 .723 .639 .454 .465 .3l3 l!B I II 3! I.033 1.176 1.036 I.045 1.076 I .Yll .B23 .778 .72l .533 l!C !I 4, I.,,, ,.,Yl l.Em4 .Y86 I.834 I. 105 I 1 .E48 .YSl .B72 .770 ,705 .464 l!D 0 !! 5! I.155 I.867 ;%h l.E36 1.035 ; 1.025 .9BB . B67 .a73 ,811 ! !E I !! 6, I.158 I.203 I.188 I.862 ,.84B l.EY2 I.014 I .547 1.074 1.028 1.835 l.E38 .982 .748 IIF I I I 7! I .860 I .008 I. 134 .Y45 I.848 1 I.032 I.015 I.,,6 .972 .542 IG I I I ,------,------l I I G! .Vl5 .33Y l.ElY5 .VB4 I.047 I I.@35 1.015 1.215 l.BB4 l.18B I7 I I I- F,! .76S ,335 l.mm I.~335 ,.a42 1.893 .338 I l.844 I.,30 1.178 1.20R I.263 I.171 I.288 !6 ! 1 1 I E!! .I333 .B73 .36-5 .YY3 ,.a7a 1 I.147 I.150 I.162 I.220 1.360 I5 I ! I I D!! ,526 .m4 . B64 .352 I.845 I.221 I I.237 I.204 1.145 l.3lEi I.435 I.316 I4 !I I 1 C!! .623 . B44 .a82 .343 I.103 ! ,.2B6 l.225 1.279 I.385 I.252 I3 I ! 1 1 Bll .4l6 .564 .627 .844 .936 I.866 I.842 1 I.117 I.390 1.4Bl I.433 1.243 I.378 1.254 12 11 I I RI, ,377 .4l3 .5*4 .857 I.871 I l.222 1.236 1.344 1,296 1.4EiB !l ,,------I ------I 2 3 4 5 6 7 G F E D C B fi

BND ID BND PIA,? 163 ! 373 l.@7Y! .733 -- - - , ------,----- , 333 I 141 .081!,.247 21 IKHEV cAB0b-C BClTTO” OF RCTI’dE FUEL

FWR-B,W,“LE ROD-ROD PLWQR LR-,48 D,5TR,BLlT,LM ,,OR,WL,iXD m N,XWRED RODY IN THE PLONE Core Performance Benchmarking Edwin I. Hatch Nuclear Plant Unit 1, Cycle 1

NP-1235 Research Project 1178-l

Final Report, November 1979

Prepared by

SOUTHERN COMPANY SERVICES, INC Post Office Box 2625 Birmingham, Alabama 35202

Principal Investigators N. S. Folk W. !=I. Cobb

Prepared for

Electric Power Research Institute 3412 Hillview Avenue Pa10 Alto, California 94304

EPRI Project Manager W, J. Eich Nuclear Power Division

0006002? ABSTRACT

In order to qualify the CASMO-SIMULATE system for BWR core analysis, the codes were used to model cycle 1 of Hatch 1. Comparisons of the SIMULATE results were made to gamma scan measurements at the end of cycle land to the process computer at selected state points throughout the cycle to evaluate SIMULATE's ability to predict power distributions and reactitiity in an operating BWR.

00060028 1.9. 13 ‘-GG--j

1.3

0.3 - 4

I I I I 1 I I I I I I I I I I I I I I I I I I I , 2 3 4 5 6 7 8 9 IO 11 ,2 12 ,4 1s ,6 17 16 19 2ll 21 22 *a 24

NOOE

Figure 6-l 106-Bundle Average Axial h-140 Distribution I! ,.” -.- I I I I I I I I I I

Figure 6-Z Relative Bundle Integrated 6a-143 Distribution

6-4 00060030 0 0 , :

0.3 -

0.2 - :

0.1 -

1 2 3 4 5 6 7 6 9 10 18 ,2 ,3 14 15 ,6 17 16 ,6 20 2, 22 *3 24 Nom Figure 6-9 Octant Normalised Ba-140 Distribution for tlX 373 .,- -- 0 0 : 3.2 ,

Figure 7-8 Core Average Axial Power Distribution on March 7, 1977 at 9354 MWd/st 1.010

1.005

,.ooo

. l . . . . . l . . 7 $ 0.990 . ru . , . . .

0.980

0.970

I I I I I I I I I 0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0

ExPos”ttE wJ*,s,

Figure 8-l SIMULATE keff versus Exposure four Hatch 1 Cycle 1

. ,

EPRI NP-xxxx

Power Shape Monitoring System

Volume 2

Technical Description

0 to be issued Fall 1980

prepared by

Technical Staff at NSC, NAI, SCP ENC, B&W and EPRI Project Managers

F. Gelhaus A. Long 0 T. Oldberg B. Zolotar

NUCLEAR SERVICES CORPORATION A DIVISION OF QUADREX CORPORATION 1700 DELL AVENUE CAMPBELL, CA 95008

00060034. 0 SIMULATION OF JUNE STARTUP

. . Section 1

INTRODUCTION

SYSTEM OVERVIEW

The purpose of the Power Shape Monitoring System (PSMS)~is to reduce fuel rod failure incidence consistent with maximising power production. This purpose has been achieved by providing three distinct abilities; 0 (1) the ability to monitor the current and recent reactor core power shape, (2) the ability to predict reactor core power shape based on proposed control changes from the current (or previous) operating state, and (3) the ability to evaluate fuel rod reliability for each established power shape or reactor state.

The ability to monitor reactor power is based on the PRIME400-SIGMA3 Interface (PSI) providing data on the status of the reactor once every six seconds. This information is also maintained as a history of reactor states for the previous six hours. Additionally it is kept as a running average state which is filed along with previous average states as soon as a current reactor state varies from the current average by some fixed (preset) percentage or whenever a preset amount of time has elapsed 0 without such variations. The average state that is filed becomes input to a program that evaluates power distribution and fuel reliability.

The prediction capability is based on being able to start from any of 50 individual reactor states (on file). The user builds a file (inter- actively through a terminal) of reactor control maneuvers and sets flags to save any reactor state resulting from one of the planned maneuvers. In this way the user can evaluate many control maneuver combinations and pick the sequence that is most energy output effective with the least pertubation to~the reactor power shape.

00060036 THE IMPACT OF OPEPATIONAL EXPERIENCE ON R&D

Walter B. Loewenstein and Achilles G. Adamantiades

Electric Power Research Institute Palo Alto, California, U.S.A.

ABSTPACT

'Lhe gradual accumulation of operating experience data from nuclear plants is having a perceptible inpact on the direction of research and development. Four areas where this influence is best manifested are identified: increased awareness of systems interaction; the importance of operational data for code g-ualification; a sharper focus of separate effects; and the importance of well-defined scaled experiments. Illustra- tions from EPRI-sponsored and EPRI-conducted projects are presented.

INTRODUCTION

With the advent of the nuclear era , research and development expanded rapidly in an effort to bring nuclear power to practical application. Experimental and analytical work proceeded in tandem using as benchmarks the measurements from separate effects test assemblies and relatively small-scale reactors with simple heat removal arrangements. As a result, research work emphasised collection of nuclear data, analysis of individual phenomena in a static or dynamic mode, and design and evaluation studies on a variety of Potentially useful systems , most of which existed (and some of which still exist) only on a conceptual level. l'he need to harness nuclear energy for the production of electric power stimulated these activities much as the needs of society have historically stimulated scientific dis- coveries and technological innovations.

As a result of these studies, nuclear power production became a reality and an important component of energy systems in the U.S. and abroad. lhe light water reactor in its two versions (PWR and BWR) became the dominant reactor type. As of June 30, 19S0, there were 14 reactors operating in the U.S. and about 160 operating abroad, representing a total installed capacity of 125 GW. Together they had accumulated 1800 reactor- years of operating experience (550 in the U.S.) mostly of the LWR type. This large body of experience has provided new insights into the behavior of nuclear plants, influencing considerably the direction of R&D.

-i- 00060037 ---.. -~..-. -..-. ..-.. .- . ..- ;I r 4 I Ill I ,?I r . .- 1’4I . - -,“I

COMPARISON OF MEASURED AND PREDICTED T IP READINGS FOR 3/l/79 STATEPOINT

00060038 The main impact of this operational experience cc R&D can be sum- meri&d as follows:

1. llx importance of interaction between neutrcnic and thermal- :hydraulic effects and among plant subsystems has been highlighted. ccnse- quently, both experimental and analytical efforts have acquired a much stronger nsystensn orientation.

2. Cperaticnal data have assumed new importance, and the required quality of these data has been more clearly defined. Enhanced efforts are continuing tc acquire the operational data necessary for testing and quali- fying analytical tools.

3. Integral operational experience and data have sharpened the focus of separate effects R&D, resulting in better resource utilizaticn.

4. Scaled system data have emerged as particularly important where 0 plant limitations prevent the cdllecticn of adequate operational data. Such plant limitations mey arise from physical , regulatory, economic or other considerations.

Even though the above perceptions are relatively new, having gradually taken shape as experience has accumulated, they have already had a sigifi- cant influence cn research activities.

PLANT SYSTEX.5

The system.s effects can be aptly highlighted by investigating the balance-of-plant transient respnee and its effects on reactor perfcr- mance . A number of such dependencies were quanr+f+Td in the P.LCT=N code analysis of the Peach Sottan turbine trip tests - . During these tests extensive transient respcnse measurements were made using both regular and special instrumentation, along with a high-speed data acquisition system for later comparison with system response calculations performed with the RETRAN code. The results are particularly nctewcehy since the turbine 0 trip transients in a BWR provide a stringent test for any reactor system dynamic code because of: strong coupling between pressure, core void, and reactivity: dcminance of acoustic phenomena during the early part of the transient; and significance of the latter stages of transients when pressure begins to drop.

The dependence of the pawer peak on several balance-of-plant parameters was investigated with a series of sensitivity studies performed in the course of the analysis. Figure 1 shows the difference in stem dome pressure following a turbine trip with and without heat transfer to the bypass piping walls (which mey initially be near the temperature of the condenser). The slower pressurizaticn of the bypass line increases the period of critical flow through the bypass valve, resulting in pressure reduction throughout the system. The effect of delay time in bypass valve response xney be quite large cc transient peak power, as shown in Figure 2. A number of other parameters of steam flow In the bypass line (e.g., loss coefficients, bypass capacity , and steem flow model) were found to have considerable effect cn peak ~~wer and steam dome pressure.

-2- 00060039 No heat transfer !(0.2& sec. delay

950 l.A-LL+ 0 2 4 6 0 Time bet) Time (se@ figure 1. Steam Dome Pressure With and Without figure 2. Normalized Power Transients for Tr2, Heat Loss to Sypass PiPe Walk. Using Various Delay Times For Bypass Valve Response.

An ilbastration of the interaction between neutmnic calculations and the&l-hydraulic analysis is given in Figure 3, which shows the difference in peak power with and $thout direct moderator heating taken into account. Direct interaction heating by gammas and neutrons of the coolant occurs much faster than does heating by conduction, causing additional early void formation, 'Ibis tends to counteract the effects of pressuriza- tim and, hence, reduces the height of the power peak. The effect is highly nonlinear with much more pronounced differential changes in the area of low direct moderator beating fractions.

The system approach to nuclear plant anal&is has-&o highlighted the importance of proper interfacing 3 Wiihoui t direct between various codes describing parts 0 6-- L.. of the system. When preparing the necessary code packages, care must be exercised so that boundary conditions between them are matched, consistency of models and assumptions is maintained, and the necessary feedback among the various neutronic and thermal-hydraulic analyses flows in the proper sequerlce. An example of this~ interfacing is provided by the analysis of the same Feach Bzktom turbine trip test by usin various versions of the RAMONA code.T4) This code simulates in detail the in-vessel thernial-hydraulic and neutronic behavior in respxL5.e to a Time (set) plant disturbance in a BWR. The inter- figure 3. Nom?alized Power Transients with and action of the in-core behavior with the without Direct Moderator Heating. external plant is described by imposing

-3- 00060040 . . ‘>

the appropriate boundary conditions of 2.0 reactor vessel pressure and feedwater l-r3 flow on the in-core analytical model. When the recorded test data for these boundary conditions were used, the RAMONApredictions of the power peaks were very close to measured values. Steady-state avsrage axial power distributions are shown in Figure 4: transient power is shown in Figure 5. The three-dimensional RAMOMAIII prediction of average detector response agrees reasonably well with actual 0 measurements, but the shift of axial Sottom TOP power toward the top of the core figure 4. Average Axial Power Distribution, Tr3. (Level D) was underpredicted.

- -- Ramona I - - - Ramona II ------Ramcmalll Measurement

I 0 0.2 0.4 0.6 0.6 0 Time (set)

figure 5. Nuclear Power Transient, Tf3.

We have dwelt at some length on the Reach Sottom turbine trip tests and their analyses because they provide an excellent illustration of the importance of accurately modeling the ,various core phenomena (core neu- tronics, fuel iin thermodynamics, coolant flow loop hydraulics, etc.1 and their interactions. 'Ihey also point out that accurate interfacing of the core system with balance-of-plant by use of proper boundary conditions is essential to successful simulation of the transient. Another exampl~5~f the synergistic nature of the system is provided by recent calculations performed to simulate an anticipated transient without scram (ATWS) following loss-of-feedwater in a 2560 MWt Fi?R plant. Selected important plant variables during the transient are shown in Figure 6. Ihis sensitivity study was done to investigate the effect of varying the value of the moderator temperature coefficient CMTC). In the reference case a value of Ap/AT = -1.08 x 10m4 OC-’ (-0.6 x lO-4 OF-') was used, corresponding to the beginning of life of a FWR cycle when the boron concentration is relatively low. Runs with half and double this MX value

-4- 00060041 were also performed. It must be noted that as the cycle progresses and the boron concentration decreases, the ME falls toward an end-of-life value Of about -3.60 x lo-* V-’ (-2.0 x lO-4 OF-'). That condition would corre- spnd to a pressure transient with a peak lower than that of the solid cur"e on .FigUXe 6. lhe conclusion of this sensitivity study is that whereati the MT h+s little effect on the steam generator mixture level and

- - - Half MTC - 0.0 - - - Base case WJO 5

0 564x I-- SO-

I-

0 ?mo- -

250 O- -

2cc IO -- I

15cxl. - 0 0 0 20 40 60 so 100 120 l+.I Time (se@ Figure 6. PWR Transient Calculations fore case of LOSSOf Feedwater, Anticipated Transient without Scam; for three values of the Moderator Temperature Coefficient (MTC).

-5- 00060042 ,.

pressuriser liquid level, and a moderate effect on power, it can have a pronounced influence on the pressure transient in the primary coolant system. Furthermore, there is very strong coupling between the MTC and the steam and liquid relief capacity of the system in terms of the pressure transient. OPERATIONAL DATA

The second major impact of operating experience on RLD is the increased importance placed on having operational data that are detailed and accurate enocgh to use for code qualification. Before the Peach Bottom tests, no data existed at conditions approaching the limiting transients that are analysed with elaborate computer codes for safety and design analysis. &xt of the plant test data available were obtained under beginning-of-life conditions duzing startup testing for plent performance evaluation. By contrast, the EPRI-sponsored tests were specifically designed and instrumented to provide data for model and computer code 0 qualification.

An example of research activity oriented toward the collection of such plant data is the series of core stability tests at Peach Bottom 2(*), . which will be used for to test and qualify BWR stability codes. 'Ibe develoment of such a BWR stability code for utility use is now,underway with EPRI sponsorship. Fxperinlenta1 results of pressure perturbation tests from Peach Bottom 2 as well J.S from other European BWRs (Barse.bSzk 2, Sweden, and TV0 I, Finland) will be used in the code qualification. An example of these data is shown in Figure 7, which illustrates the transfer function of the magnitude of average neutron flux over core pressure. The PB-2 stability tests demonstrated BWR stability margins with minimal disturbance of plant operation,~provided better quality test data than those obtained from rod oscillator tests in BWR cores, and demonstrated safe operation at low flow conditions above the rated power/flow rod block line. The BWR stability code, when qualified, will provide the utility with an improved, independent, reliable I I lllllll I I I IllIll I 10-a analysis capability and is intended for 10-Z 10-l 100 use in the reload safety analysis meth- Frequency (Hz) odology ,package being assembled by EPRI. A general view of the various Figure 7. Peach Wtom-2 EOC2 Low+low Stability codes being considered for inclusion in Test (PT4) Measured Transfer Function Magnitude Average Neutron Flux (APRM A)/ the package and their interfaces is Core Ressure. shown in Figure 8. 000~60043 Model: Core Physics Fuel Sehavior Codes: ARMP* COMETHE, SPEAR ------oldputs: . Power distributions * Fuel temperatures * Control rod wotihs * Fission gas release * Decay heat parameters, -v-me---v * Strains etc. ’ Gap conductance

Steady state am’/ysis I I I

I I I 1 FlJei core system structures

FREY, FRAP COSRA, MEKIN RETRAN STEALTH, SOLA, WHAMS, ASAQUS 0

, .------, ~------~------, *,“cl”des Erw-CELL, wt.4 Poc!~7, sw.wLATE NooE+, -3, NORGE, e,c. Figure 8.~ Code Interfaces.

CORE ANALYSIS BENCENARXING AND G+MMA SCAN WSmENTS

The extensive work sponsored by EPRI in the erea of using gemmes scene for measuring power distribution in both wP.s and BWRs is anothe? illustration of the importance given to operational data in recent yeas. This work originated in the need of plant operators to ,heve at their disposal an 0 accurate, reliable, efficient, and user-oriented enalytical tool independent of the reactor vendor in order to optimise the plant economic performance by maximising its energy output: This e&ails the optimisation of both core design and operating strategy, including batch size, enrichment, and loading pettern. The optimisation of dey-to-day operating meneuvers can also increase the plant capacity factor.

In order to fulfil1 this need, e series of reactor analysis progrems wes spnsored known collectively es the Advanced Recycle Methodology Program (AFJlF’) package. This package continues to be expanded and refined. In order to test and qua~lify the developd codes, an extensive IP detector measurements wes undertaken ~~~~~h"~W~~1~~~n~~~~~~4~. llxe gemme sc!ans performed et Retch 1 provided the meet comprehensive set of power distribution deta currently available for a BWR. Figure 9 shows the 106 bundles on which gemua scans were performed - including e conplete core o&ant and six additional four- bundle cells chosen to reveal eny power asymmetries. The measured data

-7- 00060044 @amdIesdisaesembled for rod scanning

l Tip lkations WAFT Eundke identification

Figure 9. Octant Normalised Axial Power Distribution. (Edwin I. Hatch Unit 1 W/R)

-8- 0006QO45 were then compared with results from the code SIMULATE (lSl* (This code performs detailed simu~.#ons of the nuclear reactor core using 3s input results from the CASMO code, which calculates the nuclear characteristics of each fuel type.) Figure 9 also depicts normalised axial Power distributions for a characteristic bundle with a control blade inserted up to node 7, as shox,in: IMUIXl'S code results are in good agreement with In cases of deeply inserted control blades, however, the code tends to overpredict the Power Peak. Comparisons were also made between code predictions and axial power distribution derived from the process computer (either P-l or BUCm) at selected intermediate pints of the cycle as show in Fi9ure 10. Since lEJprocess computer distributions have been showm to be quite accurate these comparisons provide a good test for the predictive capabilities of'the code SIMULATE. Most comparisons are quite satisfactory. However, at the early point in life, agreement is lacking for a number of reasons. These include nonsteady state conditions early in life , and inaccwacies in process computer data or in SIMDLATE models providing some of those data. The latter include the impact of relatively flat axial shapes and input parameters at low-exposure power and flow. lb data from these tests, particularly from the gamma scan measurements, have provided the most complete and accurate set of benchmarks for qwalification of core calculation& methods. ll-ii.9 benchmarking is necessary in order to determine with confidence the uncer- tainty in the calculations and to allow for a possible reduction in the operating margins, thereby increasing both the safety and the productivity of the plant.

FOX.9 OF SEPARATE EFFECTS R&D

Operational data have narrowed the range of certain parameters and have provided better definition of factors that pertain to health hazards and environmental impacts fran operating nuclear plants. Tvm examples are discussed below. Neutron streaming in reactor vessel cavities has been studied both analytically and experimentally in both BWRs and PwRs"". The experimental work has been performed with a variety of detectors including activation foils, proton recoil chambers, helium accumulation, fission monitors, and track etch methods in order to provide a variety of redundant data. Calculations of neutron flux and spectrum at the same locations where the measurements have been made will ptiovide ~a validated, benchmarked analytical tool. Realistic calculations of neutron spectrum and flux can be very important in assessing the radiation embrittlement of the pressure v&se1 and supprt structures, in calculating the activation of components, and in estimating neutron streaming through penetration to the equipment chambers and the operating deck inside the containment build- ing. Shielding can thus be designed to more accurate conditions.

The second example can be cited from operational data regarding iodine releases and transport under both normal and accident conditions. An extensive measurements program, sponsored by EPRI, has contributed to a much improved understanding and a better quantification of iodine-131 behavior in PWRs(*') and BWRs(*') , which is essential for effective control of emanations and for accurate determina 'o of the potential expsure. 7327 Results frcm these studies have revealed that releases during normal

-9- . . A

1.6 1.4 8 1.2 $ 1.0 .g 0.8 f. 2 0.6 a 0.4 0.2 - 1.6 I- --I I

1.6 I I

1.8 I I 1.6 1.4 g 1.2 s2 1.0 ; 0.8 L+! 0.6 0.4 0.2 1 I I I I I 0 ‘4 6 12 16 2’3 24 0 4 8 12 16 20 24 Node Node - - - Calculated with the code SIMULATE

Obtained from process ComPUter

Figure IO. Core Average Axial Power Distributions.

-io- 00060047 operations ere relatively small (a total of 0.34 Ci/yr fron a BWR and 0.203 Ci/yr from e PWR, compared to the usual assumption of 0.6 to 1.0 Ci/yr); the majority of the in-plant sources ere locally treatable; spikes during transients ere roughly 10 times greater then steady-state concentrations; charcoal absorbers remain efficient over relatively long periods of tjme for elemental and inorgenic iodine forms, but not neces- sarily for organic forms; and surfaces ten have a significant effect on iodine releases.

Accidental r+eases from operating plants have also increased our insights into the transport and health impacts of iodine releases. The "Windscale Incident," caused by e graphite fire in the No. 1 Pile at the Windscale Establishment in October 1957, released large amounts of radioactivity, including 20 kCi of iodineyl31, some off which vas trenspa'ted long distances and was detected on the European continent. Despite the large emounts released, the maximum external exlxaure to the members of the public wes estimated et 50-75 mrem ti the area of maximum eqosure, and 0 many biostatistical studies failed to reveal any statistically significant health effects(23'. The nucleer accident at The Stationary Low-Power P.eactor No. 1 (SL-11 at Idaho Falls, Idaho in January 1961 also resulted in radioiodine releases to the environment. me total amount was estimated to be on the order to 70 Ci, "a significant amount but not alarming or dangerous"'24'25'. It must be noted that the total amount released wes only a small fraction of total inventory , even though the reactor had no engineered, airtight containment. Estimetes indicate that only 3-S% of total fission product inventory escaped from the pressure vessel, whereas dux released to the environment "es about 0.15% of total ;&~;12%:2+?.' EPRI is currently sponsoring research on simulating the SL-1 event using the in- ant and ex-plent consequence codes developed for the Reactor Safety Study@? 'Ibis will provide some benchmarking for these codes and their associat$g~odels for the predict+ of health impacts from nuclear accidents .

The recent experience at 'IMI-2 has also provided reassuring data regarding the quantities of radioactivity released to the environ- lllent~~~~. !&rther measurements and analyses of events which took ,place 0 inside the containment will provide realistic inform&ion on radioiodine transport under accident conditions. Very little radioiodine was released, although releases of kxyptcn end xenon through the same path emounted to between 2.5 to 13 megacuries. [It wes estimated that 16.7 Ci out of en available inventory of 70 MCi , or 2.4 x 10e5% wes releesed'3').]

SCALED EXPERIMBNTS

Operating plant data have also shed new light on the imPortace of scaled experiments. Conversely, insights gained from scaled experiments have provided useful guidelines for tests et full-scale operating plants.

The scaled facility tests serve two important purposes: first, to provide data where geps exist in operating plant data because of, for exemple, inadequate data collection systems; and, second, to extend the range of paremeters to accidental limits that an operating plant cannot be allowed to reach. A number of natural circulations tests have been and are scheduled to bs performed at operating LWR plants. Since a number of limitations exist in the performance of these tests , scale-model experiments are essential in supplementing the operational data. The 4-100~ natural circulation model, constructed at the Stanford Research Institute under EPRI contract, simulates the Trojan PWR 4-100~ plant. The 2-1~~ reflux boiling facility simulates an actual 2-100~ Combustion tigineering System 80 plant. Work with a model of the TMI-2 core showed that natural circulation was effective for a range of core resistances, primary water inventories, and secondary flow rates’32’. Experiments with injection of nowondensible gases (helium and nitrogen) showed that reflux boiling was stable even with the noncondensible gases present.

Operational guidelines for handling a number of off-normal conditions, e.g., small breaks, are currenty based on analytical tools, generally unverified under the conditions of interest. A phenomenological thermal- hydraulic model, in conjunction with scaled and full-&z test data, can Supplement the existing guidelines and provide an improved basis for opera- Such an analysis of core uncovering and phase separation ;;yg~~g;?331 and compared with single tube, rod bundle, and TMI-2 accident available data. Agreement between the theory, on one.hand, and te!t and actual accident data, on the'other, is encouraging.

The l/7-scale model simulation of a &'R U-tube steam generatw was also used to study the thermal-hydraulic responses due to loss of secondary feedwater'34'. Finally, expertients have been conducted with a scaled facility to ascertain the coolability of a bed of debris following an accident that causes severe degradation of the core.,

llx recent emphasis on this sort of scaled system research has been prompted by the TNI-2 accident. Because the observation of operating reactor behavior under accident conditions is rather uncommon, since accidents are to hs prevented rather than observed, the importance of properly designed and operated scaled models becomes evident.

CONCLUSION

The examples in this paper illustrate the impact that operational experience has had on nuclear R&D. 'I%e system character of a nuclear plant has been highlighted: the collection of data from operating plants has p?oved essential to the testing and validating of analytical tools; the breadth of experimental investigation of separate phenomena and the need for scaled tests have been better defined. All these aspects are having profound an impact on the direction and scope of the R&D required to support a mature technology.

-12- 00060049 ,.,. ..- . ,.,-..-- .-...

9 h-4.

1. K. V. MOO=, et al., RETPAN: A Program for One-Dtiensional Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, EPRI CCM-5 VOlS. 1-4, DW~XX 197a. 2. L. A. CARMICHAELand R. 0. NIBMI, Transient and Stability Tests at Peach Bottcm Atomic Power Station Unit 2 at End of Fuel Cycle 2, EPRI NP-564, June 1978. 3. K. HORNYIK and J. A. NASER, RETPAN Analysis of the nxbine Trip Tests at Peach Bottom Atomic Power Station Linit at the End of Fuel Cycle 2, EPRI NP-1076-SR, April 1979. 4. L. MOBERGet al., PAMONAAnalysis of the Peach Bottom-2 Turbine Trip Transients, EPRI &search Project 119-2, Scandpower Peport, ScP 6.35.35, June 1980. 5. J. A. NASER, B. R. SEHGALand L. J. AGEE, “Analysis of PWRATWS Transients with the P.ETw Code," submitted to the ANS/ENS International Conference, Washington, D.C., November 16-21, 1980. 6. M. B. CUTRONEand G. F. VALBY, Gsmma Scan Measurements at Quad Cities Nuclear Power Station I.init 1 Following Cycle 2, EPRI NP-214, July 1976. 7. N. M. LARSEN et al., Core !&sign and @crating &ta for Cycles 1 and 2 of Quad Cities 1, EPRI NP-240, November 1976. 8. K. W. BURKE, Special Tip D&e&or Measurements at !Zdwin I. EIatch Nuclear Plant wit 1 Prior to &d of Cycle 1, EPRI NP-540, Eecember 1977. 9. L. M. SHIRAISHI and G. R. PAPXOS, GammaScan Measurements at Edwin I. &xtch Nuclear Plant Unit 1 Following Cycle 1, EPRI NP-511, Aumst 1978. 10. N. H. LARSEN and J. G. GOUDEY,Core Design and Coeratinu Data for Cycle 1 of Hatch 1, EPP.I NP-562, January 19 11. G. R. PAFKES and A. F. VALBY, GanunaScan Measurements at Zion Station Unit 2 Following Cycle 1, EPRI NP-509, Cctobex 1977. 12. A. SCHAFER, &actor Core Physics Designs and Cperating Cata for Cycles 1, 2 and 3 of the Turkey Point #3 PWR Power Plant, EPRI NP-877, January 1978. 13. B. W. &&SON, Reactor Core Physics &sign and werating &&a for Cycles 1, 2, and 3 of Surry Vnit 1 PWR Power Plant, EPRI NP-79-2-LD, March 1979. 14. A. T. IMPINK, Jr. and B. A. GUTHRIE, III, Reactor Core Physics Cesign and Gperating Data for Cycles 1 and 2 of the Zion Unit 2 '&WRPower m, EPZU NP-1232, December 1979. 15. D. M. WR PLANCIC,Manual for the Reactor Analysis Rogram SIMULATE, Yankee Atomic Electric Company, YAEC 7158, 197s. 16. A. AiiLIN, M. EDENIUS, and H. HAGGBLOM,CASMO, A Fuel Assembly Burnup Program, Studsvik Bnergiteknik AB (formerly AB Atomenergi), AE-RF-76-415s (rev. ed.). 17. N. S. FOLK and W. R. COBB, Core Performance Benchmarking, Edwin I. Batch Nuclear Power Plant &it 1, Cvcle 1, EPRI NP-1235, Novenber 1979. 18. J. F. CAREW, Process Computer Performance Bvaluation Accuracy, Licensing Topical Report, General Electric Co., NEW-20340, 1974.

-13- NNl60050 F. RAHN and Ii. TILL, "Neutron Flux Determination in the Reactor Cavities of LWRs," Proceeding of e Special Session on "Radiation streaming in mwer Reactors,~ of the ANS Meeting, Washington, D.C., November 1978. Published by OFXL/RSIC-43, ANS/SD-79/16, February 1979. 20. C. A. PELLETIER et al., Sources of Padioiodine at Presswired Water Reactors, EPP.I NP-939, November 1978. 21. C. A. PELL.ETIER et al., Sources of Fadioiodine at Boiling Water Feactors, EPRI NP-495, February ~1978. 22. EI. TILL, "Fadioiodine in Gaseous Effluents from tAxlear Power Plants," International Symposia on Menagemat of Geseous Waters from Nuclear Facilities, Paper IAEA-SM-245/30, February 1980. 23. G. B. SCHOFIELD, "Environmental 8ealth and Windscale,," Proceedings of e Conference on "The Medical Basis for Radiation Accidents Preparedness," Oak Ridge, TN, October 18-20, 1979. 24. "Remarks by Dr. Frank K. Pittmen, Director, Division of Reactor Development, USABC," for presentation to representatives of industry, AFC Press Release, Germantown, Maryland,, January 24, 1961. 25. "ID0 Repart on the Nuclear Incident at the SL-1 Reactor, January 3, 1961 at the National Reactor Testing Station,m USAEC Report IDO-19302, Idaho Operations Office, 1962. 26. Final Report of SL-1 Recovery Operation, May 1961 through July 1962, Report IDO-19311, General Electric Co., Aircraft Nuclear Propulsion Devt.. 1962. 27. Nuclear Regulatory Commission, Reactor safety Study: An Assessment of Accident Risk in the U.S. CoiNnercial Nuclear Power Plants, Repxt WASH-1400 (NUR% 75/014) NTIS, October 1975. 28. "Consequence Analysis of the SL-1 Reactor Accident," EPRI Report in preparation. 29. ihe-PresidentIs Comtission, Report on the Accident et 'lM1, Washington, D.C., October 1979. 30. Assessment of Off-Site Radiation Doses frcm the Three Mile Island Unit 2 Accident, Pickard, Lowe & Garrick, Inc., Rqort TDR-TIM-116. 31. Y. ZVIRIN, et al., Experimental and Analytical Investigation of e PWR Natural Circulation Loog, ANS Topical Meeting, Thermal Reactor Safety, IXOxville, TN; 1980; also EPRI NP-1364-SR, March 1980. a 32. K. Ii. SUN, R. B. DUFFEY and C. M. PENG, The Prediction of Two-Phase Mixture Level and Kydrod~~cally~ontrolled Dryout under Lower Flow Conditions, EPRI NP-1359-SR, March 1980. 33. S. P. KALRA, R. 8. DUFFEY and G. ADAMS, Loss-of-Feedwater Transients in PWR U-Tube Stem Generators: Simulation Experiments and Analysis, EPRI NP-1367-SR, March 1980.

-14- 0006UO51