ML20214M978

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Trip Rept of 780205-18 Visit to Ussr to Discuss LWR Safety
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T Appendix B i

Technical Notes on Visit of NRC Delegation to the USSR To Discuss Light Water Reactor Safety February '5- 18, 1978

1. Introduction From February 5 to 18,1978, an NRC delegation visited the USSR for dis-cussions of light water reactor safety, and to explore topics of interest for future exchange if an agreement for LWR safety cooperation is achieved.

This Appendix summarizes the technical discussions held at the Ministry of Power and Electrification, the Teploelectroproject Institute, the Kurchatov Institute, the Dzherzhinskiy Institute, the Gi droproject Institute, the Izhorsk Manufacturing plant, the Leningrad Power Station, the Novovoronezh Power Station, and the Armenian Power Station.

Section 2 provides a general discussion of the impressions of the NRC delegation regarding the roles of the several Soviet organizations involved in the design, construction, operation and regulation. of nuclear plants in the USSR. Sections 3 through 11 summarize technical information obtained at each of the organizations and facilities visited. Attachment 1 provides an annotated list of documents obtained on the trip.

2. General Discussion In the Soviet Union, the design, construction and operation of a nuclear power station involves complex interfaces between a number of institutes and ministries, and the review and approval of a number of independent government agencies. It was stated that as many as forty different agencies must approve a nuclear project before operation is permitted. The Ministry of Power and Electrification (MPE) is responsible for the design, construc-tion and operation of nuclear and fossil-fired power plants in the USSR. MPE reports to the Council of Ministers. MPE initiates the design of nuclear power stations, to a large extent plays the role of an architect-engineering firm in the U.S. in integrating the inputs of the various design organizations, guides the project through the needed reviews and approvals and finally operates the plant. Al though the responsibilities of MPE are broader than those of HRC, and although certain regulatory and inspection functions are not their responsibility, it is clear to the HRC delegation that MPE is the appropriate counterpart organization on the
  • Soviet side for the purposes of an exchange agreement on light water reactor safety.

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The need for a nuclear power plant is determined by the Ministry of Power and Electrification. Regional growth patterns and the availability of alternate supplies of energy, such as fossil and hydroelectric, are considered. When it is decided that a particular region needs a nuclear plant to supply its energy needs, the planning for the nuclear plant is begun. A State Project Commission is appointed with the responsibility for the entire project (from the time of initial planning through construction, until the plant'is put in commercial operation).

The State Project Commission is responsible for getting the required approvals of the various. organizations and developing the various necessary documents, as will be discussed later. When the project is in its early planning stages, various other governmental . agencies are consul ted. For example, in the case of the Armenian Power Station (see discussion in Section 11) in a high seismicity area, the Institute of Engineering Seismology of the Soviet Academy of Science was involved early in the project in selecting the seismic design values for the site.

As the design criteria, reactor type and site are selected, the actual facility design is performed by a number of institutes for the flinistry of Power and Electrification.0ne of the largest of the institutes is the All-Union State Design Institute "Teploelectroproject" (TEP).

TEP reports to MEP, and. appears to play a central role in the overall plant design. Other institutes perform parts of the design; e.g., Gidroproject Institute does structural design and Kurchatov Institute does core physics and thermal-hydraulics c'esign.

As the design develops, a set of design documents is prepared. These documents are made available to a variety of agencies for review and approval.

There are three independent agencies that are primarily responsible for the State centrol and inspection of nuclear power plants for safety:

Gossannadzor (State Health Inspectorate), Gosstekhnadzor (State Technical Inspectorate), and Gossatomnadzor (State Atomic Inspectorate). The State Health Inspectorate, under the Ministry of Health, is responsible for radiation safety matters for nuclear plants, as well as for sanitary, .

public health, and occupational safety matters as at any industrial facility.

The State Atomic Inspectorate has responsibility principally for criticality safety and related matters, such as reactivity control systems, core design, and control of fissionable materials. The State Technical Inspectorate has review, approval and inspection responsibility for pressure vessels, primary system piping, mechanical handling equipment, fabrication, installa-

  • tion, in-service inspections, quality assurance, etc. Apparently, both the Health Inspectorate and Technical Inspectorate assign resident inspectors at each nuclear station, but the-Atomic Inspectorate has an inspector on site only for the initial startup, testing of plant operators, reloads,

, etc.

The Atomic Inspectorate approves and authorizes the initial criticality of the core, not only from the point of view of the adequacy of the nuclear equipment at the plant, but also on the basis of the competency of the plant staff to perform the safety functions (analogous to the NRC operator licensing function).* Following initial operation of the facility, there appears to be only periodic inspection of the facilities by this inspectorate. -

The Health and Technical Inspectorates are heavily involved through-out the entire design, construction and operation of the facility.

These inspectors have the authority to require design changes of the facility.

They have responsibility for setting standards and various requirements for facility design which appear to be similar to NRC Regulatory Guides, and have responsibilities that are a mixture of both inspection and licensing functions.

Technical inspectors are assigned to manufacturing facilities which supply plant components. They issue approvals for all project elements used in construction of the plant before these elements are used.

-The manufacturing of heavy components is governed by requiren>ents (similar in their intent to ASME codes) which the inspectors use for certification. A copy of a particular requirements document was shown at one meeting. Based simply on its thickness, it does not appear to have the same coverage as the ASME code, although it was asserted that this was, in fact, intended.

Each of the three inspectorates is involved with the plant from the time of conceptual design throughout facility life. They fonnally approve the initial siting of the facility, as well as a variety of other aspects throughout the construction of the plant.

The involvement of various other state agencies appears to be similar in scope to the involvement in the U.S. of the Dept. of Fisheries, _

Dept. of Interior, EPA, etc. Through the Council of Ministers, the various other agencies can exert influence to have a project changed -

or modified to accomplish a particular objective. For example, there was considerable discussion of a requirement to install ponds at some nuclear facilities to provide fish hatcheries for

  • It was stated that the examination of operators prior to the initial criticality is done by the Inspector, but that subsequent operator

__ examinations are done by the Chief Engineer of the plant organization and certified by the Inspector.

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developing new' strains of fish. When the need for a change in design of a project is stated by some agency outside of the Ministry of Power and Electrification, there is direct negoti-ation between the Ministry of Power and Electrification and the other agency to resolve the issue. If a resolution cannot be achieved at that level, it may be necessary to elevate the matter for resolu-tion. In such a case, the Council of-Ministers would be called on to resolve the issue, although it appears that this step is seldom necessary. A list was not provided of all the agencies that have to approve various aspects of a project to reflect special interests, but the overall discussion seemed to indicate about the'same breadth and depth of federal and state agencies as would be involved in the U.S. to achieve a similar objective. The Ministry of Power and Electrification has the responsibility for assuring an adequate supply of electricity. . Their interest in keeping a project streamlined so that they can get it online quickly obviously may conflict with other agencies which do not have this direct responsibility.

The Ministry of Power and Electrification seems to be heading toward a form of standardization. The 440 MW PWR design of Units III and IV at Novovoronezh

, appears to be equivalent to the " product line" approach used by NSSS vendors in the U.S., and the 1000 MW PWR unit now under construction at Novovoronezh seems to be intended as the prototype for a new product line. It appears that the Soviets have an incentive for standardization because of the desire to market their designs in other countries.

The Soviet Union relies heavily on " district heating", that is, central station heating plants that supply hot water or steam to residential and business buildings in their cities. It was stated that 55-60%

of all residential heating in the USSR is provided by district heating.

Both Moscow and Leningrad are 100% district heating. The central station plants are primarily fired by gas or crude oil at present. Because of increasing concern for conservation of fossil fuel resources and for environmental problems with fossil fuels, there is considerable interest in the potential for using nuclear plants for district heating. It is apparent that some early design work is underway on this concept, called

" Atomic Boilers". Although no details were provided, PWR's are considered to be most suitable. The Soviets stated that 25 to 30 km is about the maximum economic distance between the power station and the user, with substantial heat losses at that distance. This means, in effect, metropolitan siting; thus there is concern over. the difficulty of convincing the health organizations of the safety and

.- reliability of nuclear units for district heating. The general approach now under study seems to be simple low thermal power

- level, low pressure and temperature units to produce only heat and no electricity. It is thought that such low performance units could

  • be made simple and reliable.

Throughout the visit, it was apparent that there is a growing concern in the USSR for the quality of the environment. The use of cooling towers appeared widespread. The Armenian Power Station and part of the Novovo-ronezh Power Station had natural draft cooling towers. Control of the release of radioactive effluents in normal operation and monitoring of the environmental effects seem to be receiving substantial emphasis by designers. There also appears to be considerable emphasis within other government agencies to assure that efforts are made to minimize environ-mental impacts from nuclear power stations.

3. Visit'to Ministry of Power and Electrification

'in Moscow - February 6,1978 At the initial meeting at the Ministry of Power and Electrification (MPE) in-Moscow, Mr. L. H. Voronin, Chief Engineer of the Directorate for Atomic Stations, MPE, welcomed the NRC delegation and introduced representatives of many of the installations that the delegation would be visiting. (See Appendix A). The principal subjects of discussion involved the roles of the various organizations (summarized in Section 2 above), and the itinerary for the visit. The NRC delegation listed some of the major topics of interest, and Mr. Voronin responded by indicating which instal-lations could best. address these topics. It was indicated that the NRC delegation t ould be accompanied throughout the trip by Mr. Y. P. Karelin, Chief Specialist, Glavatomenergo, MPE, who is in charge of a small staff group at MPE dealing with operational safety matters, and by Mr. Victor Myagkov from the Novovoronezh Power Station, who served as an interpreter for the visit.

4. Visit to Teploelectroproject Institute in Moscow - February 6,1978 The NRC delegation was welcomed to the Teploelectroproject Institute (TEP) l by Mr. Vladimir N. Okhotin, Chief Engineer (See Appendix A for other contacts).

The TEP Institute is the largest single institute in the Soviet Union.

It reports to MPE, and is responsible for the design and development of all types of power stations: nuclear plants, fossil-fired plants and j hydroelectric plants. TEP has designed approximately 80% of all the r

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power stations in operation within the Soviet Union. They also have done design work for plants exported to 29 other countries. TEP has about 12,000 employees.

The institute is centralized, with its largest concentration of employees located in Moscow. There are 12 other local divisions in various cities within the Soviet Union. The institute appears to perform the overall design work for power projects and apparently has complete responsi-bility for this design. This could include transportation, dams, reservoirs and even the design and layout of a town to house people in the general area where a facility would be built. TEP does not manufacture equipment; e.g., pressure vessels and turbines are ordered to specifications from equipment manufacturing plants.

The delegation had lengthy discussions at TEP regarding design philosophies associated with reactor safety. The newest 1000-MWe PWR design (Novovoronezh Unit V) was used as a basis for these discussions.

The design basis accident for the 1000 Mil facility is an instantaneous rupture of the largest pipe anywhere within the plant (800 mm diameter). Three independent active systems (each said to be 100% capacity) are used for accident mitigation. The rationale for three systems is: one of the systems may fail and one may be out of service for one reason or another.

The basic design of the ECCS is apparently done by the Kurchatov Institute, with TEP having responsibility for the design from the pressure boundary outward.

The general philosophy appears to be that " active" systems should have. '

a redundancy of three, whereas " passive" systems should have a redundancy of two (for example, four 50% capacity accumulator tanks are used since this is a passive system -- two discharge into the vessel above the core and two into the downcomer.) The Soviets asserted that the three trains of the accident mitigating system are completely independent includ-ing separate and independent electrical systems, component cooling, lube oil, heat sinks, etc. The delegation did not see detailed drawings of these systems; but for one area of design where pump and hardware locations have been layed out, questions were asked regarding the extent to which equipment was separate and independent in redundant trair.s.

l The response was that every attempt was made to keep them independer,t. .

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It was not clear to the delegation that in all cases a truly independent arrangement of systems is in fact achieved. The need for 3 independent systems was discussed in more detail with respect to the assumption that one of the I systems fails. Since separate nozzles are used.for providing the ECCS {

cooling water to the reactor vessel, the failure of one of the lines i leading to the vessel can cause an accident of sufficient magnitude l that the ECCS would be required. Thus, a failure of one of the systems I would initiate the accident itself!' The Soviet design philosophy of l assuming one of the systems to be out of service for one reason or another appears to be equivalent to the U.S. assumption of a single failure, l with the remainder of the system being able to perform the function. ,

Other philosophical approaches used in the U.S., such as a loss of offsite power coincident with the accident, are also assumed for~ the Soviet design.

The Soviet approach for the ECCS appears to be quite similar to the approach used by the Germans. Simultaneous top and bottom injection is used. As is the case in the German designs, the coolant is intro-duced into the reactor vessel by direct injection to the vessel instead of into the piping systems. Redundancy is provided back through the' high pressure systems, including heat exchangers, diesels, etc. Three high pressure injection systems are provided, primarily for small breaks. Thgse do not provide recirculation from the sump, but from i three 100 m capacity tanks. Three low pressure injection systems j are provided with separate, independent storage tanks and heat exchangers.

l These provide for recirculation from the sumps, and provide feedwater j to the containment sprays. Three separate containment spray systems are, provided to avoid interconnecting the ECCS trains. The particular arrange-

[ ment of ECCS was discussed in some detail, but there appear to be a number l- of design features that are not yet resolved. For example, at present there is no cross-connect between the high pressure and low pressure ECCS systems, but it was stated that such a connection is being considered.

The arrangement of the containment in the 1000 MW unit has some interest-ing features. One is that all of the ECCS equipment (both high and low pressure pumps and their heat exchangers), is located in rooms immediately beneath the containment boundary. The-bottom of the containment itself is about 12 meters above grade level. The supply of water is taken from sumps which generally are in the rooms immediately above the' pump location and about 30 or 40 feet above -the elevation of the pump itself.

The overall space provided inside containment also seems to be more generous than in the containment designs used in the U.S. (See additional discussion of containment in Section 7).

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The subject of fire protection was explored in considerable detail.

The TEP representatives said that as a design basis, the following scenario is assumed: (1) an assumed design basis LOCA, (2) a simultaneous loss of offsite power, and (3) a fire in the control room. All of these events are considered to occur simultaneously, so that a separate auxiliary control room must be used to control the plant. The approach is to ass,ume that a LOCA. signal used to initiate actions normally going through the control room and back to the ECCS system occurs simultaneously with a signal that goes to the ECCS equipment itself. The design is such that the LOCA signal to actuate the safety trains takes precedence and overrides any effects of spurious actions that would come from a postulated fire in the control room. No specific infor_

mation was provided on how that design objective is accomplished.

It may be that they have not yet worked out the cetails. The question was posed as to whether, if there were no LOCA, but simply a fire in the contro1~ room, all necessary functions could be accomplished from the auxiliary control station. The somewhat confusing response to that question suggests that this design is still in a conc (ptual stage. This subject warrants further discussion in future exchanges to see how the objective is accomplished.

One fire protection criterion is to require the equivalent of a 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> fire Darrier between redundant equipment, without credit for active suppression. The rationale is that one and a half hours would be sufficient time to bring the fire under control and prevent redundant.

equipment from being damaged. The emphasis on fire protection is 6pparently aimed at fires that can originate as a result of overheating within cable trays. It was stated that no combustibles are assumed present in the cable area. It is unclear whether exposure fires are considered. Automatic suppression systems using gas, foam and water are used.*

[ Other design basis accidents are considered in the design of the 1000 MW

! facility, e.g., steam line rupture, loss of off-site power, and various l' transients ( rod withdrawal, etc.) . With regard to specific requirements I

  • 0uring the tour of Novovoronezh Units III and IV the delegation was permited to see the cable spreading room. The cable trays apeared

[ lightly loaded, with separation of redundant cables about equivalent to j ,

U.S. practice. There was no evidence of fire barriers between trays.

An automatically actuated foam system was present (See discussion in Section 10) .

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for environmental qualification of safety equipment, it was indicated

! that all equipment located within the containment must remain functional for relatively small or minor incidents for which the containment environ-ment might reach 90*C. For major pipe break accidents, the non-safety-related equipment would be permitted to fail, but any equipment necessary

to mitigate the accident is required to have been qualified to function i 'in that environment. No details were.,available at the time of our visit.

l The accident radiation environment is assumed to be that corresponding i to a 10 percent core melt. The source term associated with that amount i of melt is assumed present for qualification of equipment and is also used as a source term. for containment leakage dose calculations.

It was stated that the main steam line break accident is the limiting accident for setting the design strength of containment, but that in the Soviet design the loss-of-coolant accident produces a larger containment tempera ture. The Soviets felt that' their use of horizontal steam gener-ators probably accounts for the difference from U.S. calculations in this j regard.

l There is no single document corresponding to the Safety Analysis Report in the U.S. Separate volumes (about twenty in number) are prepared to

. describe the different areas of design in detail, and there is one volume

! that addresses safety matters explicitly. Each of the many (about 40?)

' reviewing agencies receive all the volumes. Each may, if it desires, comment on any subject. The evaluation of each reviewing agency is sent i to all the others. Any changes in design resulting from these reviews j are sent back to all agencies. The arbitrator of differences between ~

j agencies is the State Committee for Construction, or if differences exist

j. between Ministries, the Council of Ministers (in principle, at least).

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!- 5. Visit to the Kurchatov Institute in' Moscow - February 7, 1978 The NRC delegation was welcomed to the Kurchatov Institute by Dr. V.

Siderenkov, Director of the Department of Nuclear Reactors. (See Appen-dix A for other contacts). Kurchatov Institute was the first' nuclear

center established in the USSR. The Institute reports to the State j Committee for Atomic Energy. It was here, under the direction of f Dr. Kurchato'v,that the first nuclear reactor in the USSR was constructed and operated in 1946. The Institute now employs about 6,000 people.

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Dr. Siderenkov described the role of the Institute in nuclear power as being that of " scientific leader." It is involved in the development of

_ initial design philosophy, follows up during the design of projects and participates in the startup of initial prototypes.

The institute includes the followi~ng departments: Nuclear Physics (which includes solid state and super conductivity), Molecular Physics, Plasma Physics (fusion work), Nuclear Reactors (primarily work on thermal reactor physics), and Test and Experimental Reactors (including work on structural materials).

The Kurchatov Institute participate's In the startup of all the reactors '

built in the USSR. It is apparently considered as a part of the startup test team. The institute has a primary responsibility for the core design of all of the reactors in the USSR, hence the relationship to their startup testing role is evident.

The institute has been responsible for design of PWR reactors, graphite reactors, high temperature gas cooled _ reactors, fast breeder reactors (both of the liquid metal type and gas cooled type), and for various experimental reactors.

The institute is responsible for the development of calculational methods used for core design. These calculational methods include the physics and thermal-hydraulic codes for steady-state and transient processes.

~It is also responsible for the development of calculational procedures to evaluate various kinds of. reactivity accidents.

The institute has several operational computer codes which are used for the purpose of evaluating ECCS performance. Although they apparently do not have the' responsibility for validating the accuracy and adequacy of these l codes with research work, they do have the codes operational at the insti-tute and do run them. With regard to the steady-state calculational pro-cedures, the institute conducts experimental work to verify the correctness of the calculations.

L The Kurchatov Institute must formally and specifically approve a nuclear reactor facility before it is permitted to start up. Its representatives are present at all meetings of approving organizations, and it plays the role of an inspection organization at startups. There is some apparent duality of role since the institute is also responsible for the core design for each of the reactors. In the development of the core design, each of the important elements of the design must conform to the various regulatory criteria evolved within the Soviet government. The institute, in addition, participates in the development of the standards and

, regulatory criteria that are used in the design.

The institute has the responsibility for evaluating the operating experience of the facilities and factoring that experience back into new designs. This responsibility appears to be broad in l

. scope and includes evaluating experience with normal operations, and repair and maintenance operations, as well as failure data.

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The institute perfoms research on the transport of radiation within plants and research on criticality safety to support the Atomic Inspectorate (Gossatomnadhzor). It compiles documents on all aspects of nuclear safety. ,

In a brief discussion of the single failure criteria, it was stated that single failures are assumed for active components, but that there are no -

fim requirements for passive components. Some difficulty with the inter-1 pretation of what constitutes active components was acknowledged. With l regard to the use of quantitative reliability or risk analyses in making

decisions on design requirements, the Soviets stated that they do not yet have sufficient statistical information to develop quantitative bases l for evaluating any of the equipment in the plant. The designers, however,

! do use quantitative analyses in selecting between various options in their designs, on a relative basis. No specific reliability goals have been set in their standards, but such goals were said to be borne in mind in the

development of standards and criteria. The institute is collecting i

statistical information, but couldn' t predict when their standards

might become more quantitative. In the discussion of the quantitative

! approach to reactor safety design, WASH-1400 was specifically mentioned and Dr. Siderenkov indicated that it was his personal opinion, although he did not have quantitative information to back it up, that the actual probability of core melt accidents is lower than the number stated in  :

WASH-1400.

The design of the graphite-moderated, pressure tube reactors in the USSR was discussed at various institutes. (See also Section 9 discussion at Leningrad power station). At the Kurchatov Institute it was noted that the capability for the online refueling, and the capability to detect failed fuel elements in individual pressure tubes during operation,. allows prompt removsl of failed fuel without shutdown.. The -

pressure tube design also limits the maximum diameter of a postulated pipe failure.. That design also provides a large source of cooling water in the feedwater system that can be used for an ECCS source, and j the header system for the pressure tubes makes for a less complicated j situation in assuring reliable distribution of emergency coooling water.

The Test and Experimental Reactors Department of the institute has a

! program studying the properties of fuel elements for both PWR's and pressure.-tube reactors and of pressure vessel materials. The program uses the MR reactor located at the institute. The MR reactor first

, went critical in 1964. This very versatile test reactor is a pool-type (9 meter depth) with a graphite reflector, beryllium moderator blocks and the fuel in channels. Channels are also provided for

_ experiment loops. The core is a one-meter cube. The core fuel elements are concentric plates of U-Al alloy clad with A1. Each fuel "

element generates about 2000 kW. The core array can be rearranged

easily. There are a large number of control rods for flux shaping around a test position (25 are movable during operation). Twelve fuel channels can be inserted or removed during operation. The reactor can operate at 40 MW, plus 10 MW in the te loops. The neutron flux level in the center of the core is about 3 x 10y' n/cm2-sec and as high as 8 x 10 14 n/cm2 -sec in flux trop positions. In the present core configuration about 30 experi-ment loops are in operation. Twelye are loops operating in PWR conditions (300 C), one is at 800-1500 C for gas 2 cooled reactor fuel tests and another is being used for low temperature tests with liquid nitrogen coolant.

The MR reactor is used principally for tests of steady-state fuel per-formance at nominal operating conditions, but some transient tests are run in small loops, e.g., tests of release of radioactivity to the coolant from intentionally failed fuel elements. Some tests have been run with fuel irradiated to 80,000 MWD / ton burnup (U02, zircalloy-clad fuel of the type used in the 440 MW PWR's). It was stated that they have seen no evidence of pellet-clad interactions or bambooing of the cladding. The cladding wall thickness is 0.65 mm.

The pressure vessel steel used in the 440 MW PWR's is a Chrome-Molybdenum-Vanadium alloy which has high stability at high temperatures. Reference was made to a paper at the last Geneva conference which gave the compo-sition (Paper #705). The copper content is about 0.23%. They believe that their steel is better than most used elsewhere in the world--better radiation resistance and temperature embrittlement behavior. For the new 1000 MW PWR's a modified steel alloy is being used with nickel added and vanadium reduced, for more strength. The nickel content worsens radiation resistance, and they are doing experimental work to optimize the' properties. The actual compositions are proprietary. For the 1000 MW vessels there are no welds in the belt-line region. We later

[ learned at the Izhorsk plaiit that the central ring forging is 3.5 meters high with no longitudinal welds. The expected NDT shift in the belt-line region at the end of life is about 60 C with essentially no change in the upper shelf hardness. (See also the discussions at the Izhorsk plant in Section 8).

The delegation toured the MR reactor and the first Soviet reactor. The first reactor, a graphite pile reminiscent of CP-1, first went critical in 1946, and is still operated as a neutron calibration standard. Several critical facilities are also operated in this complex.

6. Visit to Dzherzhinskiy Institute in Moscow - February 7, 1978

. The Dzherzhinskiy Institute, also known as the All-Union Heat Engineering Research and Development Institute, reports to the Ministry of Power and l

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Electrification. It has about 2500 employees. Several daughter insti-tutes are located throughout the Soviet Union. This institute is responsible for a considerable amount of R&D work on power generation equipment. The institute operates a number of large boilers for the development of conventional boiler designs. The turbine department tests and helps to build turbines. There is also a metals research department, a water-chemistry department (waste treatment and decontam-ination), an automation department (instrumentation and controls for both fossil and nuclear plants), a fuel' department-(fossil fuel only), a physical-technical department (gas dynamics and heat exchangers), a nuclear power plant department and a nuclear physics department (radiation control and shielding).* The work of the institute is directed largely at operational problems and improvements for power stations. For example, the institute is responsible for developing in-service inspection tech-niques for examining all kinds of components within nuclear plants.

The activities of the surveillance laboratory were very impressive. This lab designed a large shielded vessel with viewing windows and manipulators that can be lowered inside a reactor pressure vessel for inspection and repairs (0.5 meter concrete wall with steel shell). This device was apparently developed originally for use in the removal of the thermal shield in the Novovoronezh Unit I (See discussion in Section 10), but is now intended for use for in-service inspections, and all new reactors will have such a device available.

1 The surveillance lab is also developing acoustic emission devices for

, installation on reactor primary system piping; remotely controlled magnetic crawler devices (" caterpillars") that crawl inside steam sep-arator drums and piping for UT inspections; eddy current probe devices;

. and gamma ray devices for internal radiographic inspections of nozzles and steam generators (" woodpeckers"). The lab also develops rigs for remote measurement of thermal growth and movement of pipes, and noise analysis equipment.

There was a brief discussion of two reactor containment concepts that have been patented in the U.S. by the Soviets [U.S. Patent 4056436]. One is a concept in which the reactor compartment is vented through a wet condenser. compartment, and then to an " air-trap" through check valves.

It was said to be planned for use in several power stations. The second concept was a variation of the ice condenser design using wet condensers instead of ice beds. No details of these designs were discussed.

  • This would seem to infringe on a specific area of responsibility of the State Committee on Atomic Energy and its institutes.
    • This device has been patented in the U.S. by the Soviets.

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7. Visit to the Scientific Research Station (named after S. Ya. Zhuk) of the Gidroproject Institute Tushino District of Moscow - February 8,1978 The Gidroproject Institute (Hydro-Project) is responsible for the design of hydro-electric power stations in the USSR, and for research and development on structures. It also a(sists TEP on the design and devel-opment on. structures for nuclear power' plants, both the pressure tube designs and the PWR's. The Deputy Director of the Station, Dr. Alexander P. Kirillov, chaired the discussions with the delegation. He speaks English. The principal subjects of discussion at Gidroproject were seismic design, containment design and pre-stressed concrete pressure vessel development.

The first Soviet nuclear station built in a high seismic area is the Armenian Power Station near Yerevan (See discussion of visit in Section 11).

The seismic design criteria for this station were based on a Grade IX intensity earthquake (on the MKS scale - essentially the same as Modified Mercalli). Gidroproject developed the design frequency spectrum - said to be a " synthetic spectrum" based on a collection of data for the region (maximum peak acceleration value 0.8g, static value 0.2g). Three-dimen-sional models were built and tested. The models simulated the masses of equipment and the rigidity of pipes and reactor components. To account for scaling factors, the materials of the model were selected, varying properties such as ductility, volume weight, and modulus of elasticity.

The materials selection for proper scaling was checked by some materials testing at several scales up to full scale. The scale model of the Armenian plants was tested dynamically, measuring frequency, amplitude, accelerations and stresses.

i At the reactor station on the Kola Peninsula where there is an operating station similar in design to the Armenian reactors, explosives were set off in the excavation pit for the second unit to simulate earthquake ground motion. Amplitude displacements, stress levels, and accelerations were measured on components and structures of the first unit for which con-struction was essentially complete. The results were compared with calcula-tions and with the results of the scale model tests. The results agreed to within about 15%. The largest discrepancies concerned the response of the steam generator supports.

The institute has a shake table capable of testing components and models

. up to 100 tons and 6 x 6 meters in size. The shake table is used for tests of electrical equipment and control systems for the calculated spectrum at their location in the plant. The equipment in a typical nuclear plant is divided into four seismic categories. The situation seems similar to that in the U.S. for various classifications of both safety and non-safety equipment. Category 1, safety-related equipment, is any equipment needed to control the release of radioactive materials within the plant.

A dynamic analysis of all such equipment is required.

The development of seismic design bases for nuclear facility designs was outlined. The Gidroproject Institute has its own seismology department that conducts geologic investigations to select the intensity grade (MCS) for the facility site and the appropriate spectrum and g-values. The Soviet

~

Academy of Sciences is responsible for evaluating the seismicity of all regions of the country and the collection of seismic data. The Academy prepares charts and maps showing earthquake intensity probabilities that are used by the institute. '" x For design purposes, two earthquakes are considered, a 1 in 100 year.

earthquake (apparently analogous to the U.S. operating basis earthquake),

and a 1 in 10,000 year earthquake for which there should be no failure of the primary pressure boundary and no loss of capability for safe shutdown and residual heat removal (analogous to the U.S. safe shutdown earthquake).*

Once a site is selected the Gidroproject Institute conducts site investi-gations (micro-seismicity, location of faults, etc.) and develops a site-specific design spectrum in consultation with the Academy of Sciences.

The Academy of Sciences has responsibility for the final approval on the adequacy of the analysis, and the geologic investigations are conducted in conjunction

, with personnel from the Academy of Science. The Academy may rely on investigations by the institute or they may conduct completely independent investigations if they choose.

With regard to proximity to known faults, it was stated that they would not approve construction of a facility on an existing fault, but there did not appear to be criteria for specific permissible distances from faults.

It was stated that the Armenian Power Station is located between two faults at distances of 2 kilometers and 5 kilometers from the station.

(Note discussion of this topic at the Armenian Station in Section 11, i which characterizes the faults differently.) Since there is good solid l rock foundation between the faults, surface displacement is not con-sidered in the design.

With regard to. stress criteria, for the smaller (1 in 100 year) earthquake, stresses are limited to code allowables (their codes). For the 1 in l 10,000 year earthquake, pipes and equipment should not crack or lose functionability, but may undergo plastic deformation. Structures may crack if the necessary leak-tight integrity is maintained. It is assumed that the plant may be unusable af ter the large earthquake. Earthquake and accident (largest pipe break) loads are combined, but it was not

  • clear how they are combined. Gidroproject Institute provides only the

-- seismic loads; others (apparently TEP) take into account the combinations.

  • In response to a question, Dr. Kirillov stated that for hydro dams for which a large population may be at risk, the 1 in 10,000 year earthquake is also used as a design basis.

The first Soviet pre-stressed concrete containment building is being built for the 1000 MWe PWR, Novoyoronezh Unit V. The dry containment is 45 meters internal diameter and 80 meters high. The cylindrical walls are 1.2 meters thick and the dome is 80 cm thick. There is a steel liner.

As noted previously, the ground floor of _ the containment is not within the leak-tight barrier, but consists of 12-meter-high compartments for pumps and other equipment. The design basis accident for the containment is a break of the largest pipe. The design ' pressure is 4.6 atmospheres over-pressure and the design temperature is 150 C. The internal compartments are not pre-stressed. The Gidroproject Institute' designed the contain-ment structures itself and performed experiments on a 1/5-scale model.

The subcompartment design is by TEP, but it was stated that their design is based on dynamic calculations of pressure and temperature versus time for design basis events.

It was noted that, because buttresses in pre-stressed concrete containments introduce high stress concentrations, the Soviet designs use diagonal cables ( spiraled 180 ). Each cable has-1000 ton capacity. This has simplified construction of the containments. Greased cables are used -- no grouting. Cable loads are measured periodically. At Novovoronezh Unit V, since it is the first of a kind, frequent measurements will be taken for at least the first year. Strain gages on the dome and walls will be monitored continuously.

The Gidroproject is developing a pre-stressed concrete pressure vessel for use in the reactor under development for district heating (See also discussion in Section 2). They have studied a Swedish ( ASEA) design for a pre-stressed concrete pressure vessel for a 900 MWe BWR and believe they have improved upon it. . The Soviet design employs a specially developed ceramic insulating liner (blocks), which is under full radial compression loading in the pre-stressed design.

t In the tour, the NRC team saw the shake table and numerous instrumented concrete models.

8. Visit to the Izhorsk Manufacturing Plant Near Leningrad - February 10, 1978

^

The Soviet hosts at the Izhorsk Plant were headed by the Acting Director General, Mr. Yuri Sobelev. The Izhorsk Plant was founded in 1721 by Peter

, the Great, as a shipbuilding plant, and has always been a state-owned

'~

industry. It now reports to the Ministry of Power iMchinery. It has been in operation for 257 years, including throughout the Leningrad siege of World War II,

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at which time the German lines were about 1 km away. Over 5,000 plant workers were killed during the siege. The plant now manufactures principally heavy equipment for power plants; reactor vessels, steam generator shells, i

i i

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.--,..,-_--------..-g,y- - - . ------------c--y. v -w-mev .*-v- r--w--m-- - - - - - - - -

pressurizers, steam drums, pump casings. It also manufactures hydraulic rod drive mechanisms. The plant is unusual in that it mills its own steel in the facility and does. extensive metallurgical development, as well as the fabrication of equipment.

It was stated that the plant has the capacity to produce 5 to 81000-MWe pressure vessels per year (both pressure tube and PWR types) . The plant does not have at. cess to waterways suitable for shipment of large pressure vessels. Thus the size limitations -(dimensions, not weight) of rail sh.ipments do pose difficulties. If vessels larger than 1000 MWe are to be built, other shipping means will have to be found.

The head of the Quality Acceptance organization at the plant is appointed direct-ly by the Ministry of Power Machinery. The Deputy Head of the Technical Inspectorate (Gosstekhnadzor) for the northwest region of Russia is resident at this plant.

All of the R&D work for the vessel materials is performed at research facilities within the plant.

The plant is responsible not only for the manufacturing of pressure vessels, but also for doing detailed design work on these components. The overall design parameters are established by another institute (TEP) and the final

-detailed design is produced at Izhorsk. The one exception where considerable overall design, as well as fundamental research and testing, is done is that associated with control rod drive mechanisms. These mechanisms are designed, manufactured and tested at this facility.

The reactor vessels currently being designed in the USSR are manufactured so that there are no welds in the belt-line region of the reactor vessel.

This is accomplished by forging cylindrical sections for which the largest section is 3-1/2 meters long and made from a' 180-ton billet. The nozzles in the reactor vessels and pressurizer vessels are pressed out of the billet during forging and no welds are required in the nozzle itself.

The facility appears to be well run and of a relatively modern vintage.

The vessel steel is manufactured on site using either an electric arc vacuum furnace or the electro-slag technique. The plant has extensive facilities for radiographic inspections (including linear accelerator and betatron) , and has large metallurgical R&D f acilities.

Standards for chemical composition of pressure vessel steels are set Ey the Technical Inspectorate. The copper content for steels that do not

__ contain nickel in the alloy is 0.2%. For the newer steels ( for the 1000+1We PWR) that contain nickel in the alloy, the copper content is held in the

- range 0.1 to 0.2% -- except that the cylindrical sections in the belt-line region of the reactor vessel are from specially selected batches of steel in which the copper is held to 0.05%.

The plant perfonns a 100% baseline inspection of all vessels after fab-rication, including 100% UT of base metal, and UT, X-ray, PT and magnetic

. particle testing of all welds.

Control rod drive housings and mechanisms for PWR's are designed and fabricated at Izhorsk. They also perform short cold and hot tests of all drives and life-test all prototypes. The drives are of the rack and

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pinion type. A brochure was provided that describes the mechanism.

9. Visit to Leningrad Power Station February 11, 1978 The Leningrad Power Station is on the Bay of Finland about two hours' drive from_ Leningrad, near the town of Sosnovyi Bor (popularion about 40,000).

The station has two 1000-MWe graphite-moderated, pressure tube, direct cycle boiling water reactors. Unit 1 went on line in December 1973, and Unit 2 in July 1975. In 1977 the plant generated 12.4 billion kilowatt hours at a capacity factor of 0.71.

Each reactor has two turbine generators with common auxiliary equipment in a common hall. There are 1600 pressure tubes per reactor and 179 control channels. The fuel rods are conventional type, uranium-oxide enriched to an average of 1.8 percent and clad in zircalloy tubes. The reactor is cooled with 8 circulating water pumps, 6 of which are required to achieve full power. The water is heated to steam in the core and separated in external steam drums. The steam from the drums is directed to the turbine generators. The system is essentially a once-through sys-tem in which water is separated in the steam droms, so there is no inter-mediate system. The graphite moderator is cooled by nitrogen and helium.

The turbine generator for the facility (3000 rpm) has one high pressure, and four low pressure, stages. There are four separate condensers per turbine. Full flow treatment is used for the return condensate to the plant. The condensers are cooled by sea water. The reactor is operated at 70 atmospheres steam pressure and 284 C with about a 0.1% carry-over factor. The steam generating capacity is about 5800 tons of steam per hour.

There is a total of 12 MWe diesel generating capacity for the standby power system, derived from 3 diesels, each of which is rated at 4 MWe.

The reactors are refueled on line. Apparently about 2-3 fuel assemblies

, per day are replaced. Each individual pressure tube can be monitored for fuel failures by a gamma-ray scintillation monitor which can be connected to

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any of the pressure tubes at the exit pipe upstream of the steam drum. The monitoring for the leak detection is not continuous for any tube because there

are only a few counters. Operation is permitted with 0.1% of the fuel assemblies failed (leaking). It was stated that the plant could operate with 1% failed fuel without radiological problems for workers. To date, since startup of the station there have been 35 cases where fuel assemblies have been removed because one or more fuel rods had leakage. They recently .

built a hot-cell at the station and are beginning to examine these 35 assemblies to determine causes of-failures.

  • The water chemistry for the plant is controlled at a pH of 7.2, with a conduc-tivity 0.1 micromho per centimeter. The chloride content is less than 4 micrograms / liter, hardness less than 0.5, and the oxygen content is less than 5 micrograms / liter. The primary piping is partly stainless steel and partly clad with stainless steel.

These reactors do not have conventional containment in the U.S. sense.

Rather, there are a variety of sub-compartments which are designed to withstand pressures of 4-5 atmospheres. The compartments are connected to a long corridor through check valves, so that if a pipe break occurs in one of the compartments the steam would be exhausted through the check valves into the corridor and thence into a special condensing

. building next to the reactor building with a spray system. The condensate would be collected and pwnped back into the coolant loop. The plant-is designed so that the largest pipe break in any one of these compartments is considered the design basis. In the event of a LOCA, the initial cooling would be by natural circulation. Then a compressed gas system drives water from tanks into each of 44 collector intake manifolds (22 on each side of the reactor) each of which serves several pressure tubes. Finally regular cooling water is pumped from a 400 m3 storage tank.

It was stated that pipe whip is taken into account in design. Main cool' ant pumps are in individual compartments (8). Each ECC pump is in a separate compartment with independent power sources and separate cables.

During the plant tour the NRC team noted that none of the turbines was l shielded. It was stated that the contact dose at the high pressure stage l of the turbine casing is about 0.8 prem/sec (about 3 mR/hr) and that the average plant worker exposure is about 0.8R per year. No direct reading or indicating radiation monitors were in evidence. Apparently the shift size is fairly high, a minimum of 80 employees per shift for the two-unit station.

._ The condenser tubes are copper-nickel alloy with a 2 mm wall thickness.

There have been several tube failures, but overall performance was stated l . to be satisfactory. The tube sheets are carbon steel covered with an epoxy. In the event of sea water ingress, it would be handled by the l

full- flow treatment. If ingress exceeds the capacity of the treatment system, power would be reduced by about 50 to 100 MW, the leaking condenser isolated and the tubes plugged. This can be done on-line since there are four condensers per turbine.

\

l 1

I l

The team was impressed with the cleanliness and general evidence of good management of the station. The Soviets stated that Unit I had operated continuously for a two-year period without shutdown.  !

10. Visit to the Novovoronezh Power Station February 15, 1978 s.

The Novovoronezh Power Station is located about 40 kilometers from the city of Voronezh. ( population 920,000), near the newly constructed town (principally for station workers) of Novoyoronezh. The station site was selected in 1957 because of its location on the Don River with very sandy soil and essentially no agriculture. It was originally to have been a fossil-fired station, but later it was decided to build a nuclear station.

4 The station has four operating PWR units and a new 1000 MWe fifth unit under construction. The basic characteristics of the units are given in Tables 1 and 2. The containment building for Unit V is up almost to the dome, and the pressure vessel is in place (but little

, else of the primary system) . The Soviets expect plant startup in late 1978 (!). The NRC team did not tour Unit V.

Unit I (210 MWe) was begun in 1957 and went into operation -in 1964.

Unit II (365 MWe) started up in 1969. These were the first two demonstration PWR-type reactors in the USSR. Improvements were made in each successive unit and Units III and IV are apparently the proto-types for the Soviet 440 MWe product line which has been built else-where ( e.g., Armenian statio'n ) and exported to Finland (Louisa station) .

Capital costs have decreased with each unit. The cost was 326 rubles per installed kilowatt for Unit I and is down to an estimated 199.6 rubles per installed kilowatt for Unit V (estimated in 1976). It was noted that the cost of labor has not changed significantly over the last 15 years.

The capacity factors for each of the units are posted throughout the plant, apparently to instill employee interest to do whatever they

can to improve the operating record. The capacity factor for last L year,1977, for all four units was said to be 0.79. It was also stated that Unit II operated for a 3-year period with no shutdown except

. for refueling. A typical refueling (remove 1/3 of core and shuffle

,_ fuel) is accomplished within 25-30 days and is scheduled annually.

The capacity factor quoted includes the varying output of the plant

_ because of changes in load demand, even though the unit was available for most of the time to generate at full capacity. The operating

' experience with fuel failure at the plant has been very good, with l no outage resulting because of fuel failure; the coolant activity I averages about 10- ciries per liter for the gross activity. Each of the fuel assemblies is checked during a refueling outage (both those

Table 1 Basic' Characteristics of Novovoronezh Nuclear Power Units I-IV I tem Unit I Unit II Units III & IV Initial Operation ..1964 1969 1971 & 1972 Electric output, MW 210' 365 each 440 Therm.al output of reactor, MW 760 1320 each 1375 Number of turbogenerators 3 5 each 2 Turbine plant rating, MW 70 73 220 Saturated steam pressure 29 30- 44 upstream of turbine stop (412.5 psi) (426.7 psi) (625.8 psi) valves, kgf/cm Coolant flow through the core, 36,500 48,000 40,000 m3 /h (160,706 gpm) (311,399 gpm) (176,116 gpm)

Primary water pressure, bar 100 105 125 (1450 psi) (1523 psi) (1813 psi)

Average inlet water temperature, 252 252 267 ,

C (486 F) (486 F) (513 F) -

Average coolant heating in 19.1 25.8 30 nuclear core, C (34.4 F) (46.4 F) (54 F)

_M_aximum nuclear fuel enrichment, % 2 3 3.6

' Average fuel burnup, MW-day /t 13,000 28,000 4.40,000 Production cost, Kopecs/kw-br 1 0.75 0.65

( 1.4 4/kw- hr) (1.1&/kw-br) (0.946/kw-hr) l Ef ficiency (gross), % 27 28 32 l

Table 2 Basic Characteristics .of Novovoronezh Unit Y Reactor capacity, MW:

elcctric 1000 thermal 3000 Turbogenerators 2 x 500 MW Main circulating pump 19,000 (83,655 gpm) capacity, m 3/h Steam generator capacity, t/h 1500 Primary Loop Pressure, Atm. 160 (2320 psi)

Steam Pressure at Stop Valves, Atm. 60 (870 psi)

Number of circulation loops 4 Fuel burnup, MW-day /t 40,000 Production cost, kopecs/kW-hr 0.5 (0.7 d/kW/hr)

Efficiency, % 33 Station needs, % 5.3 em G

l

that are removed and those that remain) for any leakage (dry and wet sipping). It was stated that these inspections are primarily to obtain statistical data and that only a few instances of leakers have been observed, primarily in fuel. exposed for three years.

The fabrication of the reactor vessels for the units differ in detail. Unit I has a stainless steel-lined reactor vessel,

, whereas Units II, III and IV have-no liner material on the carbon steel which has been exposed for the past several years. In the unlined vessels some very minor pitting corrosion was observed early in the operation of the unit, but now all of the vessels are covered with an oxide film that provides protection of the carbon steel.

All of the new 1000 MWe units will be designed with a stainless steel-lined vessel, even though no serious corrosion problems have been observed.

The design bases for Unit V appear to be similar, in tenns of safety philosophy, redundancy requirements, etc., to large PWRs being built in the U.S. There are some specific requirements not yet in final form for which complete information was not available at the site.

For example, the NRC team was unable to learn what leak rate limit would be required for the containment.

Units I through IV do not have containments in the U.S. sense. In Units I and II, the primary system is located within several inner compartments of thick-wall reinforced concrete. The compartments are designed for 3 atm. overpressure and are' interconnected through check valve vents. A dome over the top of the pressure vessel is also designed for 3 atm. overpressure. The compartments are lined with stainless steel . Their tightness was checked by pulling a vacuum and measuring in-leakage. There seem to be no specific leak rate requirements. For Units III and IV, all the primary system is within one compartment designed for an overpressure of 2 atm.

The team did not obtain specific information on the emergency core 4 cooling systems for Units I to IV. The statement was made that l early designs did not consider a large pipe break.

i For the new Unit V, it was stated that for pipe breaks with break flows of 300 m 3 /hr, or less, primary pressure could be maintained. For larger breaks (800 mn is diameter of largest pipe) pressure falls

and ECCS is needed. There are three 100% capacity low pressure injec-

. tion systems - separated and independent. The precise meaning of i _

100% capacity was not clear. The design basis seems to be prevention of overheating of the fuel, but no specific acceptance criteria were stated (although the figure of 1200 C maximum temperature was referred to in passing). The low pressure ECCS is injected into the top of

  • the reactor vessel, the bottom (through the downcomer), and into coolant pipes, all through special nozzles. There is apparently no control of the flow split. If the US-USSR exchange agreement is obtained, further detailed discussion of ECCS design, particularly for the 1000 MWe units, would be desirable.

Some of the Soviet criteria documents seen by the team indicated that two independent reactivity shutdown systems should be provided (preferably of diverse design). After questioning the intent of this criterion, it became apparent that, in practice, the two independent systems consist of (1) a rod system, which is capable of shutting down the reactor ~ under transient and accident conditions, and (2) a boron injection system. In the older units, the boron injection system appears to be the slow-acting type similar to systems presently used in the United States. A comment was made, however, that in the modern designs a " fast boron injection system" will be used. A characterization of the new system was that the boron would be capable of shutting the reactor down in several tens of seconds.

The. degree of separation of control and safety circuits in cabling systems was discussed. It was indicated that a separation of control and safety circuits completely back to the individual sensors was required. While visiting the units the team was permitted to go into the cable spreading room for Unit IV. We noticed a very light load of cables everywhere, which suggested that, if complete independence of control and safety were required, considerably more control cables should be evident in the cable spreading area.

The cable spreading room was clean and free of-combustibles, and was kept locked. There appeared to be about six or seven horizontal trays stacked about 6 inches apart vertically. No fire retardant coatings were noticeable. The room was equipped with an automatically actuated foam suppression system. The criterion is that the room should.

fill with foam in ten minutes. There is one distribution ring for the foam, fed by three,100% capacity pumps, it was stated that all cable tray areas and transformer areas have fire suppression systems. There is a special fire fighting station on the site, and shif t crews are trained for fire brigades.

It was stated that the new Unit V would have a separated control station for shutdown and residual heat removal in the event of fire, and that such a feature might eventually be retrofit on the older units.

At the Novovoronezh station, signals from radiation monitors all over the station are fed to a central dosimetry room with recorders and al a rms. The room is continuously manned. There appear to be few, if any, locally indicating monitors in the plant - only alarm lights

-- tripped from the central room.

O

The water chemistry for each of the units is controlled somewhat di f ferently. The major difference is that the pH level varies from a value of 7 in the Unit I vessel to values of 8 to 9 in Units II, III and IV, which are unclad steel. Unit I has no boric acid system. The parameters on the water chemistry for the secondary side are as follows:

chlorides 0.01 milligrams per liter, salts 0.003 milligrams per liter, oxygen 0.015 milligrams per liter, and a pH of 8.

...s During the tour of the station, the team had an opportunity- to look at the spent fuel rack arrangement in the Unit III spent fuel pool.

The pool was originally designed to be filled with spent fuel. Rather than providing a more densely packed fuel rack arrangement to increase capacity, they have chosen to build a second layer of storage racks above the bundles that are already present in the pool. The second layer for fuel storage has not been used yet but is installed in the pool. A comment was made that they are thinking of the possibility of building some dry storage facilties for storage of fuel away from the reactor.

The availability of onsite power for each of the units was discussed briefly. For Unit I, three diesel generators are provided, with two out of three being required for accident situations. In Unit

. II, four diesel generators are provided, and three out of four are required for accidents. For Units III and IV, three diesels are located in each of the units, with either one of three or two of three being required for the diesel. ( Answer was not clear) .

Units III and IV were said to have three completely independent diesels. The diesels are completely independent from the ac power distribution capability, and from the dc battery supplies for starting the diesels and supplying dc battery services elsewhere in the facility. There are no interconnections between the three diesels at each of the units, and no interconnections between the three buses; although apparently, each bus feeds both units. The statement was made that only one of the emergency buses was required to provide sufficient protection for the facility.

In Units III a.nd IV, the 73 control rods ( fuel-follower type) are arranged in twelve groups. There are three levels of scram pro-vided in the protection system (automatic and manual). Level 3 scrams two rod groups. Level 2 scrams each rod group sequen-tially (apparently with opportunity for the operator to interrupt the sequence). Level 1 scrams all rods. Because these were lead units of the 440 MWe design, large shutdown margins were provided.

.They normally control on boron concentration with rods fully out except

__ for one 6-rud group partially in. Unit IV has 1% excess k, cold with 0

4 all rods in at beginning of core life. Cold shutdown requires 12%

boron with all rods out. Insertion of sixteen rods is required to reach hot shutdown. It was said that they can achieve cold shutdown without boron addition after two to three months of operation. It was noted that the Armenian station has fewer rods (37) and less shutdown margin.

The 1000 MWe Unit V will have twelve nests of control rods - not of the fuel- follower type. The fuel rods _will bes, in cans with no cross-flow.

The most significant operating problem that has occurred at Novovoronezh was the failure of a thermal shield in Unit I in 1969, after five years of operation. The shield came loose and was' " bouncing" due to flow. The failure was discovered by the noise. The removal of the shield took about one year. The difficulty was that the shield was attached to the vessel inside the downcomer region and could not be removed without cutting it into pieces. The shielded device described in Section 6 was built and used for this repair. The shield was removed and not replaced. Analyses showed it to be unnecessary, and thermal shields were not put in the later units (although the downcomers were thickened).

Once again the NRC team was impressed with the cleanliness and general good housekeeping and management of this station.

11. Visit to Armenian Power Station near Yerevan, Armenia - February 13, 1978 The Armenian Power Station has two 440 MWe PWR units similar in design to the Novovoronezh Units III and IV. Construction began in 1971, and the first unit achieved criticality in December 1976.

The second unit is still under construction and expected to be on-line in 1979. The principal interest of the NRC team in this station results from the fact that this station is the only Soviet nuclear station in a high seismicity area.

The Armenian power grid exports power to the European part of the USSR through the Georgian grid. Formerly most of the power in Armenia was obtainad from hydroelecric dams. When additional power was needed, the difficulty of transporting crude oil to this remote region (altitude about 1800 meters) made nuclear a. preferred choice.

, Unit I is not base-loaded. It achieved full power operation in November 1977, but some testing is apparently still going on.

9

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4 From the outset it was recognized that the high seismic activity would be a problem in design. The region is characterized by Grade VIII i

on the MKS intensity scale. Extensive site investigations were i

performed, and historical records back to the 5 th Century A.D. were l

' studied. The area is rich in history and old structures. A cathedral near the station (the seat of the Armenian Catholic Church) dates back to the 3rd Century A.D. The Academy of Engineering Seismology ,

, of the Armenian Academy of Sciences was heavily involved in site investigations and in setting design parameters. The plant was designed for a Grade IX MKS intensity earthquake. In contrast to statements made at Gidroproject (see section 7), the people at the site said that the two features near the site were not " seismic faults" but were surface " wrinkles". Another "small fault - not j seismic" was said to be 10 km away; but they had not detected any evidence of movement from examination of 100-200 meter deep bore holes. No age data seemed to be known-to the people we spoke to.

The basement foundation under the site is basalt which had been

, covered by volcanic lava and then another layer of basalt. The site was excavated down to the second basalt layer (about -22.5 i meters elevation), the lava removed, and replaced with concrete.

i The lowest point of the reactor building foundation (beneath the i reactor vessel) is at elevation of -10 meters resting on the concrete fill. The center of the reactor core is about -1.8 meters

( grade level is "0") .

There have been three earthquakes in the region since 1977. The most recent (January 3,1978) had its epicenter in Georgia (Grade VII) and produced Grade IV intensity at the reactor site. The other two, with epicenters in Turkey and Iran, also produced Grade IV intensity at the_ site. These earthquakes resulted in no problems with equipment at the power station. The facility is equipped with extensive seismic instrumentation and complete records were obtained of the earthquakes that have occurred. There are 92 accelerometers located on equipment throughout the plant and a seismograph is located on the site. Some seismic research i

facilities are also located at the Armenian Power Station.

, Three categories of systems are specified in the design: Category 1 -

, the primary coolant system and reactor building structures; Category 2 -

! the cranes, auxiliary building and systems; and Category 3 - the turbine plant. Category 1 'was designed to the most stringent criteria

~and the other categories to less stringent criteria.

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a 1

O Ef fort was made in plant layout to keep major safety-related components at the lowest possible elevations. All switchgear ( relay cabinets and motor control centers) are located on the ground . floor. The Control Room, however, is et the 9.5 meter elevation ( above grade) . .

All equipment was said to be tested for seismic resistance (by full-scale or model tests); this includes control instrumentation, electronic controllers, equipment mounts and. connections, pumps, valves, motors, etc.

Some of these tests were done at Gidroproject, others at various.

manufacturing plants and institutes.

The reactor vessel is rigidly held in a support system tied to the foundation. An annular shield tank around the vessel, water-filled in other designs, was replaced with a dry shield.(material not speci-fied). The reactor internals are said to have special restraint l

devices to prevent vertical and rotational movement. The reactor vessel head has special screwjacks to prevent tilting and rotation.

Fuel assemblies were tested for seismic loads at the man"facturing plant, but no special reinforcement was found to be needed. .There is widespread use of hydraulic (oil) snubbers.

. The horizontal steam generators are suspended on straps (with vertical spring snubbers) and have horizontal snubbers ( two degrees of freedom) .

The reactor has a seismic scram circuit, with some readouts in the control room. It is apparently set now at about the Grade VIII intensity level, but may be changed as experience is obtained.

The subject of seismic design, and particularly the data available from the Armenian site studies, should be an important subject for future, more detailed exchanges.

9

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a. g ATTACHFENT 1 List of Documents Obtained on Visit to USSR- s February 1978 Copies of the 'ollowing f documents and brochures are available from F. Schroeder, DSS.

1.

Brochure entitled Teoloelectrooro.iectInstitute. Describes the activities of TEP . Written about 1962 - in English.

2. Booklet entitled Works of Gidroproject. A collection of 23 short technical papers, published in 1975 - in Russian.

3.

Paper entitled " Development of the Nuclear Power Industry in the USSR,"

by L. M. Voronin, E. P. Karelin, et al . Reprint from Nuclear Power and Its Fuel Cycle, IAEA, Vienna 1977 - in Russian with English abstract.

4. Paper entitled "On the Soviet-American Seminar on Designing and Construction of Large Dams in Seismic Areas" by A. P. Kirillov.

Prepared for International Commission on large Dams, Salzurg, Austria September 14-16, 1977 - in English..

5. Brochure entitled Novovoronezh Nuclear Power Station - in English.
6. Brochure entitled Drive Mechanism of Control and Protection System of Water-Moderated Water-Cooled Power Reactor (440 MWe) - in English.

7.

" Rules for Nuclear Safety of Atomic Power Station," 1976 ( Approved by Gossatomnadzor December 13,1974). . Informal translation by J. Lewin, ORNL.

l 8. Book entitled Atomic Electric Power-Stations by L. M. ' Voronin,1977 -

in Russian ( Appears to be a collection of papers).

9. Book - History of the Izhorsk Plant - in Russian.

l 4

I List of Documents Available from Joseph Lewin, ORNL

1. Pamphlet entitled Electric Power System of Armenian SSR.

Illustrated brochure of power installations '

and power grid in Armenia. -

2. Book entitled Problems of Design and Operation of Atomic Power Stations. Proceedings of the All-Union Heat Engineering Institute, Issue No.11, headed by A. S. Kon'kova, et al. Moscow " Energy" Publishing House,1977. Under translation at ORNL.
3. Book entitled Problems of Design and Operation of Atomic Power Stations. Proceedings of the All-Union Heat Engineering Institute, Issue No. 2, Moscow " Energy" Publishing House,1974. Four articles from this book are being translated at ORNL.
4. Book entitled The Control and Safety of Nuclear Power Reactors.

I. Ya. Emel'yanov. Moscow Atomic Publishing House, 1975. This is a personal ' copy belonging to Mr. Lewin.

5. Book entitled The Electrical Sections of Power Stations. Edited by S. V. Usov. Energy Press, Leningrad, 1977. Personal copy purchased by J. Lewin;.available for anyone interested.
6. Brochure entitled The Beloyarsk Atomic Power Station. Published in Moscow 1973. Extensive description of two super heating i reactors and one LMFBR in the Ural Mountains. In possession of J. Lewin from previous trip.

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