ML20214N008
| ML20214N008 | |
| Person / Time | |
|---|---|
| Issue date: | 07/09/1982 |
| From: | Lafleur J NRC OFFICE OF INTERNATIONAL PROGRAMS (OIP) |
| To: | Bock W Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML20214M958 | List: |
| References | |
| FOIA-86-336 ACRS-GENERAL, NUDOCS 8609150356 | |
| Download: ML20214N008 (3) | |
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i Jul. 91982 bec: A re le MEMORANDUM FOR: William M. Bock, ACRS FROM: Joseph D. Lafleur, Jr., Deputy Director Office of International Programs
SUBJECT:
PREHEATED SAFETY INJECTION SYSTEMS I am enclosing some descriptive infonnation on the Loviisa plant in Finland and on some of the USSR PWR's. I'm afraid it will not give the detailed ECCS infonnation you want.
Please let me know if you want us to do any international surveying to get more infonnation.
original signed by.
' Joseph D. Lafleur, Jr. -
Joseph D. Lafleur. Jr.
Deputy Director Office of International Programs
Enclosure:
As stated 8609150356 860905 PDR FOIA TAYLOR 86-336 PDR .
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% t>L w N86-est.1 00NTENTS Page ANNOTATION................................................ 2 INTRAREACTOR MEASUREMENTS AND MONITORING OF NEUTRON FIELD PARAMETERS IN ATOMIC POWER PLANTS by S. S.
Iomakin and V. I. Petrov................................ 3 ORGANIZATION OF RADIATION MONITORING SYSTEM POR ATOMIC POWER PLANT by K. B. Gochalleyeva, V. F. Fedulov and A. P. Dubkov............................................ 28 RADIATION METHODS OF MEASURING PRIMARY-CIRCUIT COOLANT FIDW RATE AND HEAT POWER OF WATER-MODERATED WATER-COOLED REACIORS by A. I. %marenko, A. A. Bolberov and V. V. Ips enko . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . W SYSTDI FOR MONITORING THE STATE OF METAL IN ATOMIC POWER PLANT EQUIPMENT by N. N. Shabanov, M. F.
Sheshenev and Yu. A. %bakova........................... 58 HYDRODYNAMIC ASPECTS OF THE PROBLEM OF VIBRATION OF PIPING AND THE MAIN PRIMARY CIRCUIT APPARATUS IN WATER-COOLED WATER-MODERATED NUCLEAR POWER PLANTS, Acoustic Emission Diagnostics by G. N. Nozdrin, A. A. Samarin, and G. P. Simanovskiy........ 71 METHOD OF CALCULATING THE EFFECT OF SHOCK WAVES AND THE MOVEMENT OF FLYING OBJECTS IN NUCLEAR POWER PLANT ACCIDENTS, by Yu. V.
Rzheznikov, M. S. Indurskiy, and E. V. Lifshits............... 89 l
l CRITICAL DISCHARGE OF HOT WATER by V. D. Keller, B. K. Mal'tsev, and D. A. Khlestkin........................................... 108 SOME PROBLEMS IN THE THERM 0 HYDRAULIC STUDY OF NUCLEAR POWER PLANT STEAM GENERATORS by N. S. Galetskiy, V. I. Grishakev, and O. B. 0tt................................................. 122 1
l l
l k
l Page CONTENTS (Continued)
RESULTS.0F'A STUDY AND WAYS OF IMPROVING STEAM SEPARATION IN STEAM GENERATORS AND DRUM SEPARATORS OF ATOMIC ELECTRIC POWER PLANTS 132 by Yu. V. Kozlov..............................................
STEAM DRYERS AND THEIR USE IN VARIOUS ATOMIC POWER PLANT EQUIPMENT 150 >
by Yu. V. Kozlov..............................................
THEORETICAL AND EXPERIMENTAL STUDIES OF THE POSSIBILITY OF OPERATING BLOCK II 0F THE BELOYARSK ATOMIC POWER PLANT IMENI I. V. KURCHATOV IN THE NATURAL COOLANT CIRCULATION MODE 159 ;
by A._M. Bukrinskiy and A. L. Khazanov........................
ALGORITHMS FOR CALCULATING TECHNICAL AND ECONOMIC INDICES OF ATOMIC POWER PLANTS WITH VVER-440-TYPE REACTORS, by Ye. M. Shvartsahteyn, V. P. Slutsker, M. P. Ryzhankov, B. L.
Bondarov, N. S. Galetskiy, V. F. Vlasik, A. N. Mitropol'skiy, G. D. Gil' gut, I. Kh. Tsukerman............................... 177 AUTOMATIC CONTROL OF NUCLEAR POWER PLANTS WITH WATER-COOLED WATER-N0DERATED REACTORS, by A. V. Naumov........................... 203 SELECTION OF THE NUCLEAR POWER PLANT CONTROL PROGRAM by A. V.
231 Na umov and V .' M. Sha mb erev . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
CIRCULATION SYSTEM FOR COOLING AND PURIFYING THE AIR IN THE CONTAINMENT SHELL OF A NUCLEAR REACTOR by G. V. Matskevich, B. M. Stolyarov, B. K. Mal'tsey, and Ye. M. Klement 'yeva. . . . . . 242 SOME WAYS TO IMPROVE THE RELIABILITY AND SAFETY OF NUCLEAR POWER PLANTS WITH GAS-COOLED REACTORS by Ye. M. Shvartsshteyn....... 249 THE MECHANISM OF CONDENSATION OF A JET OF STEAM IN A FREE VOLUME OF LIQUID by V. K. Bronnikov, b. K, Mal'esev, and I. P. '
261 Yermolenko....................................................
267 ABSTRACTS.......................................................
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llRC TRANSLATICfJ ,,1023
/, LA V LANGUAGE 0: FOREIGN DOCUEIR: cerman USSR:
SUMMARY
OF RELEVANT DATA TRANSLATED TITLE OF DOCUENT: ON SOVIET REACTOR MODELS k
UNTRANSLATED TITLE: UdSSR: ZUSAMMENSTELLUNG RELEVANTER DATEN S0WJETISCHER REAKTORTYPEN Aumm(S): ---
Association for Reactor Safety, Inc.
IRANSLATED NAME AND ADDRESS OF CCRPORATE AUTHOR:
UttrRANSLATED NAME AND ADDRESS cese11schaft fur Reaktorsicherheit (GRS) mbH OF CORFORATE AUTHOR: Glockengasse 2 - 5000 KBln
- OF ORIGINAL FOREIGN DOCLNENT: 1981 FOREIGN DOCtFF_NT ID NUMBER (S): 1981, Series J, Number 32 34BER OF PAGES IN TRANSLATION: 8 DATE 7RANSLATED Fm IRC: Mar 23, 1981 ,
f// th 0 l r / ? !I r'
' uv y < w / y j i sX Michael Barjansky ,
TRANSLATEDBY: 4921 Seminary Road - 11207 NAME AND Alexandria, VA 22311 (703)379-0576 ADDRESS
i GRS KUR2-INFORMATION [ NEWS IN BRIEF) 1981 Series: J No. 32 4
' USSR:
SUMMARY
OF RELEVANT DATA ON SOVIET REACTOR MODELS Tde USSR aims at constructing an additional Nuclear Power Station capacity of 24,000 to 25,000 MW between 1981 and 1985. This objective is to be achieved by expanding existing installations and by building new nuclear power station units in Zaporozje, Rostov, Ignalinsk, Balakovo. South Ukraine, Smolensk, Khmelnitskiy and Kalinin. The majority of the plants are to be equipped with pressurized water reactors of the WWER-1000 type (cf. GRS-Kurzinformation, 81/J/18, Issue 24+25/81) .
Very little is known in the West about Soviet' reactors. Accordingly, most tabular overviews under the heading USSR present only incomplete data, or none at all. For this reason, the following is a compilation of some available data on the Soviet reactors of the WWER-1000, RBMK-1000 and BN-600 types.
- WWER-1000 t
The WWER Pressurized Water Reactor, which has been used for years in the USSR and in other socialist countries, was developed in three stages, from the WWER-210 or WWER-265 to the WWER-440, and from there to the high performance reactor WWER-1000. The design changes leading to this latest stage involved a more ef ficient use of the reactor space. By levelling out the neutron flux, a more uniform b'urning in the core was achieved and, thus, a better utilization of the fuel.
The reactor is made up of pressure vessel, upper section with the drive mechanism for the control and shutdown system, and reactor core. The fuel used for the reactor is sintered uranium dioxide enriched with U 235. The fuel rods filled with uranium dioxide pellets are housed in the hexagonal guides of the core. Water containing boric acid serves as coolant and as moderator.
Biological radiation shielding of, the core is provided by a " dry shield" consisting of an annular vessel filled with serpentine concrete. Ducts in the " dry shield" contain instruments with which the intensity of the neutron flux is measured. In a clearance between the heat insulation and the reactor pressure vessel there are devices that make it possible to monitor by remote control the condition of the pressure vessel material.
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Publisher: Gesellschaft fur Reaktorsicherheit (GRS) mbH
[ Association for Reactor Safety, Inc.)
Glockengasse 2 - 5000 K51n - Tel.: (0221) 2068-1
6 Characteristics of the WRER-1000 Reactor power, MW
--electrical 1000
- thermal 3000
. qoolant temperature, *C
- reactor inlet. 288
- reactor outlet 322 Steam flow rate, ton /hr 4 + 1,470 Pressure in primary circuit, bar 160 Capacity of main coolant pump, ton /hr. 4 20,000 Saturated steam pressure before turbine, bar 60 Number of coolant loops 4 Main coolant tube, NW mm 850 Outside dimensions of reactor pressure vessel, m
- height ca. 10.90
- diameter ca. 4.50 9
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- 1. Upper section
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- 3. Control rod guide tubes
- 5. Reactor shaft'
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Sectional View of a WER-1000 Nuclear Power Station
- 1. Refueling machine 8. Reinforced concrete enclosure
- 2. Damping bassin 9. Replacement rods.for control
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Diagram of a WER-1000 Nuclear Power Station
- 1. Turbogenerator 5. Manostat
- 2. Steam generator 6. Emergency 8h"*d "" #*"
- 3. Main coolant pump
- 7. Pusher
- 4. Compressed air tank e RBMK-1000 The REMK-1000--a light water cooled, graphite moderated pressure pipe reactor--generates saturated steam with which turbogenerators are driven.
To prevent graphite oxidation and to enhance heat transfer, the reactor operates in a helium-nitrogen atmosphere.
The possibility is provide to exchange pressure pipes under cold, shutdown condition. By means of control and trim rods, reactor power can automatic-ally be maintained at a preset level, the neutron flux can be levelled out over the entire reactor core cross section and, consequently, the burn distribution can be made uniform throughout the reactor core.
- BN-600 The Fast Breeder Reactor BN-600 in Block 3 of the Belojarsk Nuclear Power Station drives three turbogenerators of 200 MW power, each. Inside the reactor tank there are: the reactor, pumps, the intermediate heat exchanger and an inside biological shield (consisting of steel blocks and pipes filled with graphite). The reactor tank--a cylindical vessel with an elliptical base and a conical top-rests on the foundations by means of a thrust ring on roller bearings. The upper part of the tank supports the revolving cover and the pivoting column.
5
3 Characteristics of the Fast Breeder Reactor BN-600 Reactor power, MW 1,500 thermal 600
- electrical' Temperature in the reactor coolant circuit, *C 380
-Iat inlet 550
- at outlet 24,000 Sodium throughput in the reactor coolant circuit, con /hr 6,800 Sodium throughput in a loop of the second circuit, con /hr Sodium temperature in second circuit, O' 320
- at heat exchanger inlet 520
- at heat exchanger outlet 1,840 Steam generator capacity, ton /hr Steam parameters 505
- temperature, O' 140
- pressure, bar Dimensions of core, m 2.05
- diameter 0.75
- height 370 Number of fuel elements in core 96 Dimension of fuel element enclosures (wrench width), mm 3,500 Length of fuel elements, mm ,
Fissionable fuel UO or (Pu0 + UO2 )
127 Number of fuel rods per fuel element Diameter of fuel rods, mm 6.9 Material for fuel rod sheaths stainless steel 700 Maximum temperature cf fuel rod sheaths, O' up to 10 Maximum burn Operating time per fuel element load, given a 450 utilization coefficient of 0.85, days ,
Work time between loads, days 150 1.4 Breed factor 6
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Sectional View of BN-600 Reactor l 1. Support ring
- 2. Reactor tank
- 3. Pump
- 4. Electrical drive for revolving cover l
- 5. Revolving cover ~
l 6. Upper fixed shield
- 7. Heat exchanger
! 8. Central column with mechanisms for control and shutdown system
- 9. Loading and unloading apparatus .
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Characteristics of the RLMK+1000 Reactor power
- electrical 1,000
.-- thermal 3,200 umber of pressure tubes 1,693
,Udhniumload, ton 192 Fnrichment, % 2 Averese burn, mwd / ton 21,500 Fuel . rod sheaths, diameter x thicknebs, mm 13.5
- 0.9 Material of fuel rod sheaths zircaloy Coolant throughput, ton /hr 37,500 1 Pressure in steam drier, bar 70 ,
Steam generation, ton /hr ~5,800 i Turbine steam consuaption, ton /hr ' 5,400 Steam parameters at turbine inlet
- pressure, bar 65
- temperature, O' 280 Heat is removed by means of a three circuit system: in the first two circuita, the circulating fluid is liquid sodium, in the third circuit it is water and steam. The sodium in the first circuit (reactor goolant circuit) 1s. cooled in the intermediate heat exchangers by the.sedium of the second circuit. The heat in the heat exchanger is transferred to the steam generators by means of three coolant loops of the third circuit.
The reactor coolant circuit comprises two parallel loops, each of which has two heat exchangers and one submerged circulation pump with two suction lines. Each pump is equipped with a back-pressure valve. The ,
sodium in each loop of the second circuit is circulated by means of a submerged centrifugal pump with a suction line.
The core zone consists of 370 fuel elements and the centrol and shutdown rods. This involves two adjustment rods,19 absorber rods. and six shut- --
down rods. At its upper surface and all-around the core zone is surrounded by breeding zones which consist of uranium dioxide grorpings.
In the USSR, too, efforts are under way to staudardir.e- the design of nuclear power stations. Plans along these lines are almost ecmpleted at the "Gidronproject" and "Teploelektroproj ect" institutes. It is significant that this involves advanced technical processes in the construction and utilization of nuclear power stations.
Source:-CRS Cologne 11.11.1981 8
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