ML20081K731

Research News.Volume 4,Number 1

ML20081K731
Person / Time
Issue date:	05/31/1991
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-BR-0112, NUREG-BR-0112-V04-N1, NUREG-BR-112, NUREG-BR-112-V4-N1, NUDOCS 9107010074
Download: ML20081K731 (8)

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NUREG/BR-0112 f"g M..... RESEARCH XEWS OFFICE OF NUCLEAR REGULATORY RESEARCH U.S. NUCLEAR REGULATORY COMMISSION VOLUME 4, NUMBER 1 MAY 1991 U.S. Working Group 3 Participates vessel could be predicted under pressurized thermal shock or low-temperature overpressure accident in Reanneallng of Sov.iet conditions. A number of U.S. plants have vessels Novovoronezh-3 Reactor Vessel that will closely approach or possibly even exceed NDT limits during their origina! 40-year license Charles Z. Serpan, DE/MEB peri d; ther plants contemplating license renewal l

will almost certainly have to face the prospect that JCCCNRS Working Group 3 Agreement their vessel will reach or exceed the NDT limit.

During the October 1990 meeting of the Joint Reactor Wssd AnmaHng Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) between the USNRC and the Embrittlement of reactor vessels can, however, be Kurchatov Institute in Moscow, USSR, it was agretM reversed or eliminated by,, annealin g.,,

This is that U.S. Working Group 3 (WG-3) would send a accomplished by raising the vessel temperature as team of experts to witness the reannealing of the much as 300 F above the operating temperature for Novovoronezh-3 (NV-3) VVER-440 reactor vessel a time period of about one week. The Soviets find The purpose of this trip would be to follow in detail that such conditions will result in virtually complete and to evaluate the in situ annealing of the NV-3 recovery of the initial properties. Although no U.S.

reactor vessel and, as a result of this evaluation, reactor vessel has yet been found to be so embrittled provide an assessment of the applicability to U.S.

as to require annealing, nine Soviet VVER-440 plants of the many different procedures and activities reador ass 6 han akea@ Wn hnd m W undertaken and necessary for such an annealing in sufficiently embrittled to require annealing as a light of current U.S. t! ssel and plant designs, prerequisite for continued safe operation. The NV-3 industry practices, and regulatory requirements.

wssd was de Mst m M anneaW W W %Ms although the temperature was lower than that Reactor Vessel Embrittlement presently used for reasons of conservatism; it has now been annealed again but at a higher temperature Reactor pressure vessels are fabricated from low, to provide better assurance that the desired alloy ferritic steels, which are subject to properties were recovered and also for the Soviets to embrittlement caused by exposure to high energy check the rate of reembrittlement following the first neutrons emanating from the reactor core. The rate ameal of embrittlement is significantly affected by the exact chemical composition of the steel, the highest wet Reactor AnneaHng hperbnce rates being caused by the presence of copper and nickel in U.S. re ctors and phosphorus in Soviet gg reactors. Radiation embrittlement becomes apparent annealings of reactor vessels of the VVER-440 Model as an increase in the nil ductility transition 230 type. The Model 230 vessels are the most temperature (NDT) of the steel. Because the NDT important ones to be annealed because the temperature marks the important boundary between phosphorous and copper contents are higher in these ductility and brittleness in a reactor vessel, it is vessels than in the later Model 213 reactors. itis essential that the reactor operate above the NDT n tod in addition, however, that the newer Modal temperature at all times to ensure ductile behavior.

213 plants have certain safety improvements, Normal operation of reactors, however, causes the including ECCS systems. Both models nevertheless NDT temperature to increase, possibly to the poir't an smaMameter Wssels w% a m nanow where brittle rather than ductile behavior of the water gap; this coupled with the high P and Cu 9107010074 910531 PDR NUREO BR-0112 R PDR

content is the primary cause for early high U.S. WG 3 Team Schedule and Activities embrittlement in Soviet reactors. The following VVER 440 reactor vesseln have been annealed:

The actual reannealing of NV-3 took place during February and March 1991. The O % sent two teams 1.

Novovoronezh 3 (1987 and 1991),

of personnel to observe the sequence of operations 2.

Armenia 1 (1988),

conducted by the Soviets for this reannealing. The 3.

Nord 1, 2, 3 (in what was East Germany) first team observed the "tumplate" sampling (1988, 1990),

operations, the checkout and insertion of the 4.

Kozloduy 1, 3 (Bulgaria) (1989),

annealing rig, and the start of the annealing bestup i

5.

Kola 1, 2 (1989),

and temperature stabilization.

The first team returned to the U.S. after about three weeks at the site. The reactor was annealed for about four days, U.S. Preparations for Reannealing Observation and the cooldown required almost that much additional time. The second team arrived at the site A team of four U.S. personnel from NRC and a just as the reactor reached the end of the cooldown contractor, MPR Associates of Washington, D.C.,

period; they observed post anneal " template" traveled to Moscow in November 1990 to gather sampling operations and the start of the packaging of information about the activities related to Soviet the annealing rig for storage. Grinding operations

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vessel annealing and their sequence. The MPR were postponed to a later time. Personnel on the personnel were especially important to this two teams included representatives from the Offices information gathering visit because they have been of Research (RES) and of Reactor Regulation (NRR),

intimately involved with many different kinds of Region I, MPR Associates (who will be responsible activities conducted inside containment of a number for preparing the description of Soviet annealing in of reactors during shutdown and maintenance terms of application in the U.S.), U.S. utilit4s having operations. They are also very familiar with U.S.

PWRs that could require annealing in the future, the regulations conceming permissible operations inside Oak Ridge National Laboratory (for information

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containment and were able to quickly focus on the relative to metallurgical aspects of annealing), and 1

kinds of activities that were conducted for ann 6aling the Electric Power Research Institute.

l and how they were carried out to ensure safety and minimum radiation exposure.

Summary and Objectives The objective of this activity was to observe in detail Soviet Annealing Activities and to evaluate the in situ annealing of the NV 3 reactor vessel and, as a result of this evaluation, tc For the NV-3 reannealing. the Soviets first took provide an assessment of the applicability to U.S.

" template" samples of vessel material just under 1/4 plants of the many different procedures and activities inch thick by about 1 x 3 inches, from which they undertaken and necessary for such an annealing in have cut sub-size Charpy specimens to establish as-light of current U.S. vessel and plant designs, irradiated material properties.

(After annealing, industry practices, and U.S. regulatory requirements. -

additional " template" specimens were taken to

. In addition to observing the operations specific to establish the as-annealed condition.) For all previous annealing, the WG-3 team gained much information annealings, a manned shielded " cabin

• equipped with on other c, aerations related to the annealing of a hardness tester was lowered into the dry vessel to reactor vessels in the Soviet Union, including establish the as-irradiated material condition in a movement of the annealing rig into the containment, different way. (Such hardness measurements are removal of internals, placement of the seal between

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also taken after annealing to determine the vessel and refueling canal, decontamination, water

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effectiveness of the' annealing.) Finally, the heater chemistry, arid control of sadiation levels in the plant, assembly was lowered into the vessel.

The

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assembly had three rows of 18 heaters each and As result of these visits to the Soviet Union to covered almost six vertical feet of vessel surface.

- witness the reannealing of the NV 3 reactor _ vessel, The heaters raised the vessel temperature to 460 C NRC has gained important information on the actual (about 850 F) where it remained for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

conduct of reactor vessel annealing for application to Cooldown of the vessel required only a few days, if U.S. plants that might need to perform nneahng in done on a round-the-clock basis, the entire annealing the future.

The teams identified a number of cycle, including before and after " template differences in the application of the Soviet sampling," heatup, hold time, and cooldown, can be technology to U.Siplants. However, because of the accomplished in about 20 days.

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close interaction with the Soviet personnel and the Analysis and Evaluation of Operational Data. It is on site observations afforded by these visits, the also used by such Government Agencies as the i

differences were clearly defined. We now believe National A'wntutics and Space Administration and

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that positive consideration of anne " in the U.S.

the Departments of Energy and Defense for their can proceed on a much more inft

, asis and reliability and risk assessment work, e.g., Galileo From a mission and the space shuttle.

j with greater assurance of a

somewhat different viewpoint,

. provided s

the Soviets with an opportuniti

,nstrate their IRRAS version 1.0 was compted in February 1987.

capability for annealing to a a

) U.S. utility Over 400 copies have been

.ributed to national personnel.

The Soviets desirous of laboratories, utilities, Government Agencies, and marketing this technology in the v.i as an excellent other organizations. Version 2.0, which includes the

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way to earn "hard" currency, so the visits were as capabilities for accident sequence analysis, was important to the Soviets for economic reasons as distributed in the spring of 1990 and is documented they were to the NRC for technical reasons, in NUREG/CR 5111.

IRRAS 2.5 has just been 4

l completed and is being placed in the National Energy l

Software Center. A graphical event tree editor and PRA Code Developed for Use on an improved accident sequence cut set algorithm 4

were added in this version, which is documented in Personal Computers NUREG/CR 5300.

Dale M. Rasmuson, DSR/PRAB 5

New 10 CFR Part 20 Published in 1985 the NRC started to develop an easy-to use computer code to perform accident frequency Alan K. Roecklein, DRA/RPHEB analysis, At about that time, the increased power i

and capabilities of the personal computer (PC) made On December 13,1990, the Commission approved it possible for the PC to perform Jome of the data the issuance of the revised 10 CFR Part 20, preparation tasks for the primarily main-frame

" Standards for Protection Against Radiation," and program. However, the rapid increase in PC power the rule was published in the federal Register in and the new capabilities introduced made it possible December. Th,is comprehensive revision of the basic to do most PRA tasks and calculations on the PC.

radiation protection standards applies to all activities As a result, the Integrated Reliability and Risk licensed by the USNRC. Agreement States will be Analysis System (IRRAS), an integrated PRA analysis tool, gives the user the ability to create and analyze adopting similar regulations.

fault trees and event trees using a personal The revised 10 CFR Part 20 is based upon the 1977 computer. This program provides functions for fault recommendations of the International Commission on tree and event tree construction and analysis.

Radiological Protection and is generally consistent with the 1987 recommendations of the National The fault tree functions range from graphical fault Council on Rad;ation Protection and Measurements.

tree construction to fault tree cut set generation and quantification. The event tree functions include The revised Part 20 contains significant changes graphical event tree construction, linking of fault from past practice and procedures for estimating, measuring, combining, recording, and reporting trees, defining accident sequences, generating doses.

These changes are associated with the accident sequence cut sets, and quantifying them, IRRAS also contains capabilities for performing introduction of new concepts and methods of assessing doses.

uncertainty analyses, importance analyses, and sensitivity analyses.

The revised Part 20 differs from the previous Part 20 IRRAS is being used in many NRC projects. Some of in many ways, the most important of which are:

these are the seismic margins research program, 1.

The use of the " effective dose equivalent" (risk resolution of generic safety issues, and the development of risk-based inspection strategies. It weighting of organ doses) concept instead of is currently being used in the low power / shutdown the " critical organ approach" represents a major PRA's being done for RES by Sandia National departure from past practice.

2.

The concept of centrolling the sum of internal Laboratories and Brookhaven Nationallaboratory and is being evaluated for use in event assessment by and external doses will require new procedures and records.

the Offices of Nuclear Reactor Regulation and 3

3.

Reliance on annual instead of quarterly dose unit (one per train) with one of the two pumps being limits and elimination of a cumulative lifetime shared between the units via a crosstie. However, limit will necessitate tighter dose control by the ESW system support from Byron Unit 2 via the licensee and modification of operating crosstie between the two units was not available procedures, while Unit 2 was under construction. The insights 4.

New limits on doses to members of the public derived from that study indicated that the core and the embryo / fetus will also require new damage frequency due to the unavailability of a two-procedures and records for their train (one pump per train) ESW system could present implementation.

a significant risk to the public health and safety, 5.

Allowance for modification of dose estimates particularly if one ESW pump from the adjacent unit and airborne concentrations for such parameters via an ESW system crosstie is not available.

as actual particle size, solubility, and retention will require guidance on acceptable technical Fourteen units at seven sites having tae basic Byron approaches.

LSW configuration were evaluated as part of the 6.

Revised intake and concentration limits will resolution of this issue. Other design configurations require modification of procedures and dose of ESW systems, including those of single unit sites, calculation manuals.

will be evaluated under GI-153,

• Loss of Essential 7,

New recordkeeping and reporting requirements Service Water in LWRs.*

(Form 4 and Form 5) may call for additional instruction.

It should be noted that the success criteria for the ESW systems in providing adequate cooling A major staff effort is under way to deve8 new and capability during normal, accident, and postaccident their conditions are design specific, depending on the revised guidance to assist licens-conversion to the new standard. t ilatory plant configuration, the capacities of the ESW guides are being prepared on estima o -dding pumps, and equipment dependencies on ESW internal and external doses, calcula sse to cooling. Although the success criteria may be as the embryo / fetus, high and very high.a An areas varied as the ESW systems, this evaluation assumed at nuclear power reactors, the content of radiation a generic set of success criteria as a representative protection programs, controlling external doses from model for purposes of quantifying the events leading airborne radionuclides, recordkeeping and reporting to possible core damage accidents. These generic radiation exposure data, and estimating doses from criteria are discussed below and apply only to multi-bioassay measurements.

unit sites having two ESW pumps per unit with a crosstie between them.

Essential Service Water System During normal operation, one ESW pump per unit provides adequate cooling to the required systems Fa. lures at Multi-Unit Sites and components. The second ESW pump per unit is i

assumed to be normally in a standby mode. Because Demetrios Basclekas, DSIR/RPSIB of load shedding (isolation

.of_ nonessential equipment), one ESW pump per unit is assumed The essential service water (ESW) system is required capable of handling accident and cooldown heat to provide cooling in nuclear power plans during loads. With one plant in power operation and the normal operation and accident conditions. Typical second plant in the shutdown or refueling mode of equipment supported by the ESW system are the operation, the criteria assume that one ESW pump component cooling water heat exchangers (and can provide adequate cooling to shut down the therefore the reactor coolant pump seals),

operating plant through the crosstie connection containment spray heat exchangers, high pressure should the need arise.

injection pump - oil coolers, emergency diesel generators, and air conditiuning and ventilation A review of operational experience showed that a systems. Failure of the ESW function could lead to number of different components in the ESW system severe consequences.

may fail to perform their intended function in a variety of ways. However, some important failure A generic issue, GI-130, " Essential Service Water modes for the ESW system are associated with System Failures At Multi Unit Sites,* was identified f ailures of certain components. For example, f ailures in 1986 as a result of a probabilistic risk assessment of the traveling screens or other common-cause of the ESW system performed for Byron Unit 1.

problems at the intake structure can lead to the Multi-unit sites like Byron have two ESW pumps per partial or complete loss of the water supply. The 4

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ESW pumps and their electric power supply are other approximation msthod involving the combination of important contributors to the partial or total loss of the experience data with an analytical technique was the ESW system, used. A multi-unit ESW system fault tree similar to the existing model for Byron Unit 1 was developed.

i The comprehensive evaluation of the operating This modified model represents the unavailability of f

experience showed that, excluding system fouling the second unit to supply ESW to the first unit in the (sediment, biofouling, corrosion, erosion), there were evant of complete loss of ESW in the first unit.

a total of 12 events involving a possible complete loss of the ESW function in 667 reactor years.

To calculate the initiating event frequency for loss of l

System fouling data were noted but were excluded ESW, the total of 667 reactor years was divided into 1

from the current analysis because of the earlier 487 reactor-years at power and 180 reactor-years at j

resolution of Generic issue 51, " Improving the shutdown. The initiating event frequencies were Reliability of Open Cycle Service Water Systems."

calculated to be 1.1E-03 per reactor-year at power, 4

3.2E-02 per reactor-year at shutdown (with one pump running and one at standby), and 2.9E 01 per l

A detailed examination of the loss-of ESW events reactor-year at shutdown (with one pump running indicates that a number of them occurred during and the other in maintenance).

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shutdown. Some of these events may not have been a complete loss of ESW in terms of total The evaluation showed that the small LOCA due to stoppage of ESW flow, even though the ESW system failure of the reactor coolant pump seal is the j

may have been declared inoperable. The differences dominant acciderst sequence.

in the ESW system between power operation and shutdown are primarily the actual heat load and the Operating experience data consisting mostly of LER equipment affected by the loss of ESW. In addition, submittals showed that the duration of the ESW the different administrative requirements imposed by system f ailure has ranged from 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to a few days the technical specifications make these two before recovery. Approximately 70% of the ESW i

operational modes more distinct, failures were recovered within one hour. About 20 to 26% of all events involved more problematica.

i The evaluation of ESW system failures at multi-unit hardware or other failures and were recovered within sites included a determination of the initiating 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The remaining events involved the most frequency of loss of ESW, the core damage serious hardware-problems, and recovery took a frequency (CDF) due to ESW failure, and the dose relatively long time, it is estimated that, by the end

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i consequence.

of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, only about 1% of the events were not recovered, in order to estimate the core damage vulnerability caused by the failure of the ESW system, a full-scale For each of the operating states, a conditional core l

PRA model that included initiating event frequency damage probability was calculated by renormalizing categories, event tree and fault tree analysis, and the original base case with the respective support system dependencies was developed. The configuration-dependent initiating frequency and i

PRA model was then appropriately modified to reflect weighting the state-dependent initiating event various plant operating configurations to analyze the frequency. The dominant sequence is the reactor consequences of the loss of ESW function in each coolant pump seal LOCA with a CDF of 8.8E-05 per operating state in addition, the probability of a loss-reactor-year, which is about 60% of the total CDF of-coolant accident (LOCA) caused by failure of a due to loss of ESW of 1.5E-04 per reactor-year.

l reactor coolant pump seal was established based on a recent pump seal failure model.

The total CDF (1.5E-04 per reactor-year) is judged to be substantial compared to the Commission's subsidiary safety goal of 1.0E-4 per reactor-year, i

The initiating event frequency representing the loss of ESW for multi-unit site operations was derived Potential alternatives for improvements thet could from operational experience for single-unit PWRs.

lower this core damage frequency were selected by This initiating event frequency, as modified, is considering the dominant failure modes of the ESW assumed to be valid for multi-unit PWR sites and is system and the dominant accident sequences not specifically limited to single PWR units, As the contributing to the relatively high CDF.

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system configuration for various operating states may be different, the initiating event frequency for The following potentialimprovements were analyzed:

1 each state was datermined separately.

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1.

An additional crosstie to reduce the possibility is intended to critically review two potential i

of a malfunction of the cross connection mechanisms for volcanism in the Basin and Range, between units, the tectonic region containing Yucca Mountain, and 2.

An electric power cross-connection to increase to investigate the extent to which it is possible to I

the redundancy of the electric power supplies to assign a probability for future volcanic activity in the ESW pumps, region surrounding Yucca Mountain. Three studies 3.

A separate intake structure or bay with an will be performed in an attempt to address the above I

additional swing ESW pump to increase the issues. The first stuoy will analyze the existing redundancy of the ultimate heat sink or source volcanic and tectonic data of the Great Basin to of cooling and increase the availability of the identify spatial and temporal relationships between ESW pumps.

volcanism and tectonism.

4.

Changes in Technical Specification requirements and emergency procedures.

The second study will encompass investigations i

S.

Installation of an independent reactor coolant performed at several different analog sites pump seal cooling system, throughout the Southwest. Because of the size of 6.

A combination of the reactor coolant pump seal volcanic systems, e.g., with roots in the upper cooling system and changes in Technical mantle and a subareal extent of cubic kilometers, it

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Specifications and Procedures, is virtual!y impossible to understand the dynamics of a volcanic system through the study of one volcanic Costs, reductions in CDF, and collective dose averted field unless extensive drilling is performed at the site.

I have been calculated for each alternative.

The The alternative approach being used in the NRC work results are being evaluated, and a proposed involves investigations of at least two sites that resolution of GI-130 is being coordinated with other exhibit varying degrees of erosion and hence related generic issues, exposure of the internal plumbing system. The first investigation will examine the dynamics of a volcanic vent and the potential for polycyclic volcanism I

Volcanology Research Program through investigation of an analogous but deeply Plan Under Development eroded Tertiary basaltic volcano in the Sierra Nevada, i

Caiifornia, tnereby providing vitai information regarding the internal plumbing system. The second I

Linda A. Kovach, DE/WMB investigation will attempt to identify the various eruptive scenarios through consideration of the Volcanic activity has occurred in the vicinity of following issues: structural control on volcamsm Yucca Mountain during the Tertiary and Quaternary, within a volcanic field, coupled magmatic-hydrologic Evaluations of a high-level waste repository at Yucca processes, temporal and spatial distribution of Mountain will require an estimation of volcanic volcanic events, and the character of the crust and hazards in the Yucca Mountain area over the next upper mantle.

10,000 years. Current scientific methods have not demonstrated, with any certainty, the ability to The third study will be designed to integrate the predict future volcanic activity. This is due in part to results of the other work into the NRC's performance a lack of understanding of the controlling assessment efforts by modeling the mantle systems mechanisms of magma transfer and storage and that control the tectonic behavior of the Basin and I

even the rates of magma production. The external Range. This will provide the driving force that manifestations of magmatic activity in terms of defines the boundary conditions for the regional dramatic eruptions are abundant, but it is much less models evaluated in the first study. These models clear how this phenomenon is quantitatively related will then be used to estimate the probability and to deeper crustal processes.

probable type of volcanic events that may occur over the projected life of a repository.

The Department of Energy (DOE) and the State of Nevada are investigating specific aspects of the RES plans to begin this work in FY1991 with the potential hazard of volcanic activity at the Yucca placement of the first project at the Center for Mountain site. The Waste Management Branch of Nuclear Waste Regulatory Analyses. A proposal has the NRC Office of Nuclear Regulatory Research is been made to the Office of Research by the Johns also developing a modest research program to Hopkins University (JHU) to invite the author to provide an independent technical basis for the review come to JHU as a part-time Associate Research 1

and assessment of the claims of DOE and Nevada Scientist in the area of volcanology. If this proposal based on theirinvestigations. The proposed program is approved, she will begin work on the volcanic vent 6

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investigations as part of her duties at JHU. The National Seismographic Network studies of an extended volcanic field are designed to Ded.icated fohaw the volcanic vent studies if funds are available, and the final integration study is designed as a follow-on to the first study of available DE/SSEB information on the Basin and Range.

Division of Eng'ineering to Support On April 3, 1991, the National Seismographic Network (NSN) was officially dedicated at the Review of Advanced Reactors National Earthquake Information Center of the U.S, Geological Survey (USGS)in Golden, Colorado. The John A. O'Brien, DE/SSEB NSN was established as a cooperative effort between RES and the USGS. The USGS has long With the aim of assistmg NRR in its technical review been active in studying the seismicity of the United of evolutionary, passive, and other advanced States, especially the Western U.S., whereas the reactors, the RES Division of Engineering has NRC has placed more emphasis on the seismicity of developed a 5 year research plan tha* covers the Eastern U.S. The NRC, for over a decade and a structural, seismic, mechanical, electrical, and half, has sponsored regional seismographic networks materials engineering.

The plan has been covering areas of the Central and Eastern U.S. An coordinated with NRR, other RES divisions, and the interagency agreement signed in 1986 established Nuclear Safety Research Review Committee. Some the NSN, which will take over the function of the of the important issues treated in the plan are; regional networks in the Eastern and Central regions after 1992.

1.

Assessment of unique construction techniques and unusual structural configurations, The NRC was instrumental in identifying the need for 2.

Reliabihty of components in passive systems, establishing the NSN program and using more 3.

Definition of the OBE for advanced reactors, advanced earthquake monitoring technology than 4.

Qualification of advanced instrumentation and had been used previously. The NRC and USGS control systems, jointly directed the implementation of the program.

5.

Design critena for interfacing systems LOCAs, The NRC is primarily responsible for the portion of 6.

Deficiencies in ASME fatigue requirements the network east of the Rocky Mountains, whereas applied to advanced reactors with a 60-year life, the USGS emphasizes implementation of the western 7.

Extent of application of leak-before-break portion. The NRC is therefore a major partner in this behavior in advaiced reactors, pioneering effort that combines the rnost advanced 8.

Tornado design requirements for advanced seismograph stations with data transmission via

reactors, satellite to a central recording facility.

9.

Pressure vessel materials for advanced reactors,

10. Containment performance goal for severe accidents.

The dedication ceremony was attended by representatives from the Department of the Interior, Besides these topics, special sections of the plan the USGS, the NRC, and other interested parties, deal with reactor types very different from those including the news media and seismologists from all presently in service in the United States. These parts of the U.S.

Participants were addressed by, include the liquid-metal-cooled PRISM design of among others, Frank Bracken, Deputy Secretary of General Electric, the PIUS pool-type design being the Interior; Dallas Peck, Director of the USGS; and developed by Combustion Engineering, the heavy Eric Beckjord, Director, Office of Nuclear Regulatory water CANDU-3 being proposed by Atomic Energy of Research.

Dallas Peck and others strongly Canada Umited, and the modular high temperature emphasized that this undertaking has been marked gas reactor of General Atomics. The plan recognizes by excellent cooperation between the USGS and the that future licensing practices focus on design NRC, an achievement that does not always certification of standardized units and addresses accompany agreements between Federal agencies, questions resulting from the staff review of the EPRI advanced light water reacter requirements document.

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Feasibility of Using Collular end in - which altered oncogene properties could be involved in the loss of growth control that culminates j

Molecular Research to Reduce in tumorigeneses; and the progress that had been

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Uncertainties in Health Eff ects from made in the. genetic characterizations -of several Low-Level Radiation Studied h"**". and animal ne plasms. On the basis f this analysis, the study group concluded that, at the J

present time, it is feasible to mount a program of DRA/RPHEB radiation research directed at the mechanisms of radiation-induced cancer with special reference to RES has sponsored a study to examine the feasibility risk of neoplasia due to protracted low doses of of reducing the uncertainties in the estimation of risk sparsely ionizing radiation, due to protracted low doses of ionizing radiation through studies at the cellular and molecular levels.

A report on this study, NUREG/CR 5635, has bJen The idea that such studies would be timely was distributed to individual scientists and Federal derived from the discovery of oncogenes, the agencies working in this field The staff intends to development in the past 15 years of the techniques explore with them the possibility of developing a of recombinant DNA molecular biology, and the coordinated program of research designed toimprove evident and significant progress that has been made the understanding of the effects of low-level in the characterization of certsin human cancers in radiation.

genetic terms.

The study was porformed by Science Applications international Corporation with participation of RESEARCH NEWS is published by the USNRC Gifice eminent scientists from the Department of Radiology of Nuclear Regulatory Research, Edward L. Hill, and Radiation Biology, Colorado State University, Editor.

- under the leadership of Professor Mortimer M. Elkind.

Comments, suggestions, and articles for future in addressing the question of feasibility, the study issues should be directed to the Editor, RESEARCH group reviewed the cellular, molecular, and

NEWS, U.S.

Nuclear Regulatory Commission, mammalian radiation data that are available; the way Washington, DC 20555.

1 1A01AR19L19R 120555139531 A[puBLICATIONSSVCS N

TPS-POR-NUREG P-223 DC 20555 yASHINGTON 8

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