ML20116M967
| ML20116M967 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1990 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| NUREG-BR-0112, NUREG-BR-0112-V03-N1, NUREG-BR-112, NUREG-BR-112-V3-N1, NUDOCS 9608210194 | |
| Download: ML20116M967 (12) | |
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OFFICE OF NUCLEAR REGULATORY RESEARCH U.S. NUCLEAR REGULATORY COMMISSION Volume 3, Number 1 January 1990 in Situ Testing of Motor The vast majority of MOV operators performed, and reverification of O erated Valves in in the United States are made by operability at intervals. This article P
the umitorque Corporation of discusses several aspects of in situ Nuclear Power Plants Lynchburg, Virginia. Most of the testing of MOVs.
j Owen O. Rothberg, DSIR/ElB remainder are made by Rotork Controls incorporated of Roches-The MGV operators in nuclear ter, New York. In most cases, plants are adaptations of a com-Motor-operated valves (MOVs) are identical MOVs are installed in the mercial design originally developed found in most fluid systems of paralleltrains of safety systems.
in the late 1930s. The design has every commercial nuclear power Where two MOVs are installed in a evolved, but the basic concept of a j
plant. Their failures and the train of a particular safety system to torque-limited motorized gear box subsequent consequences have act as two valves in series they are remains. The equipment design been concerns of the Nuclear usually identical. MOVs in parallel has not changed radically over the Regulatory Commission and its trains are usually served by identi-years. Several problems arose in staff for some time. Reports, cal power and control systems and the application of the operator Bulletins, Circulars, and Notices have similar operating environ-design to nuclear power plants, dating at least as far back as 1972 ments. These MOVs are serviced some of which were not recognized document failures and resulting by the same personnel using the for some time thereafter. Problems NRC recommendations to the same procedures in each particular that are inconvenient in the com-nuclear industry. RES is conduct-plant. Therefore, MOVs are espe-mercial field are intolerable in ing full-scale flow-interruption tests cially subject to generic and nuclear power plants. For example, (discussed in an accompanying common-mode problems. Since spring pack relaxation has only article) to determine the ability of the reliable positioning of valves in recently been recognized as a MOVs to close against fluid flows a nuclear power plant is critical to significant problem in the industry.
typical of a pipe-break accident.
the safe and economical operation Improperly loaded spring packs These tests indicate that the design of the plant, generic MOV problems can disable an operator or cause or adjustment of some MOVs may are a serious concern.
damage to a vdve.
be such that they will not isolate the flow. The critical problem is that in recognition of this concern, NRC MOVs are subject to loads and MOVs have been experiencing high issued Generic Letter 89-10, "In stresses from the control systems failure rates when called upon to Situ Testing of Safety Related and power systems that serve them operate against pressure or flow.
Motor-Operated Valves," in June as well as from the fluid systems in The major root causes of the 1989, in the letter, NRC recom-which they serve. MOVs are subject problem include:
mends that licensees develop and to partial damage or degradation implement a program to ensure that willleave them operable for
- 1. Defective parts or subcompo-that switch settings of safety-normal or no-load situations but nents.
related MOVs are selected, set, and may cause failure at design basis
- 2. Misadjustment of switches or maintained so as to ensure opera-demands.
components.
bility of the MOV at design basis
- 3. Poor coordination, on a life cycle conditions. Several other causes of MOVs are somewhat unique as a basis, of design, installation, inoperability at design basis class of components because they maintenance, and testing.
conditions such as sizing, exist at a junction of several sys-
- 4. Vague procedures for mainte-deterioration, and misadjustments tems and other components that i
nance, test, and failure analysis.
other than switch settings are must all work in order for the MOVs
- 5. Ignorance of the design basis addressed. NRC also recommends to function properly and without parameters that control MOV size trending of failures, prototype damage. There are a number of and adjustment.
testing if in situ testing cannot be effects that can operate either 9
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individually or synergistically to tain assurance'of design basis techniques, there were few in situ prevent an MOV from functioning, operability. This assumption was testing options available. Disas-either partially or fully, bolstered by the emerging signa-sembly is not a particularly attrac-ture analysis techniques that made tive option because the degraded in situ testing is only one aspect of it possible to diagnose misadjust-condition might not be detected a comprehensive program to ments,internaldamage, or wear and the MOV could be reas-ensure the operability of MOVs without extensive disassembly.
sembled incorrectly.
under all required conditions. The The assumption was weakened, role of in situ testing of MOVs is however, by results of the full-scale The advent of a system to record burdened by severalfactors. First, flow-interruption tests mentioned and allow subsequent detailed the testing that can be done in a previously. These tests indicate analysis of pertinent MOV electrical nuclear plant is limited by safety that some of the extrapolation and load parameters was truly and plant operation considerations. techniques used to predict MOV innovative. The recognition that Second, there is an implicit as-operability under design basis such data could provide insights sumption, justified or not, that a conditions may not be conserva-into the misadjustments and dete-particular component that is in tive. This may be due to the riorated conditions that might occur place has been properly designed, limitations of the existing diagnostic in the working mechanisms of an fabricated, and installed.
equipment as well as the variables MOV v.as a significant milestone.
associated with converting torque in situ testing may indicate that an to thrust in the MOV operator. The Signature analysis diagnostic MOV has deteriorated or is inoper-design basis conditions modeled in techniques provide unique insights able but may not be able to provide the tests were extremely severe.
into the on-line performance all of the information needed to Signature analysis diagnostic characteristics of an MOV and verify operability at design basis techniques, as they evolve, may allow detection of wear, deteriora-conditions. If the assumption of provide reliable extrapolations for tion, or misadjustment without design basis operability is to be less severe conditions. Certainly, major disassembly of the equip-validated by in situ testing, appro-signature analysis techniques are ment. The technology is develop-l priate test methods must be valuable to maintain operability of ing rapidly. For example, on-line available. The methods may be MOVs for normal service.
monitoring of MOV operating somewhat different from those parameters is expected to become used for the detection of degrada-A properin situ test of an MOV available in the near future.
tions or other anomalies. Thus the should provide objective assurance in situ test methods should provide of future operability under the The first signature analysis diag.
plant oDerators with techniques required conditions. The ASME nostic system for MOVs became that will verify that each MOV has Code (Section XI) stroke-timing available just about the time of the the ability to meet design basis test, which is the test mandated by Davis-Besse incident that prompted requirements ano allow diagnosis 10 CFR 50.55a(g), does not provide the staff to develop a bulletin on of anomalies as well. Such in situ such assurance. That test indi-MOVs (Bulletin 85-03). That testing may not be possible for a cates that an MOV may be capable signature analysis system was number of reasons, some of which of moving for a particular stroke but originally intended to serve as a are discussed below.
provides little information about diagnostic tool for misadjustments future operability. The Section XI or degraded conditions. The in spite of these limitations, the stroke-timing test is almost always system was used to accommodate NRC staff has focused attention on conducted under no-load (no flow the need to verify operability at in situ testing primarily because or no differential pressure) condi-design basis conditions by means virtually all of the components are tions. Very little information about of an in situ test that would allow in place and operational. Initially, wear, misadjustment, excessive extrapolation from the conditions the focus on in situ testing devel-loadings, broken parts, dete-that were encountered at the time oped when it was recognized that riorated parts, etc., can be gath-of a particular test to design basis the then existing MOV operability cred from the stroke-timing test. In conditions. The basic assumption test was inadequate and would severalinstances, MOVs have been was that a linear relationship exists have to be modified.
left inoperable after a stroke-timing between thrust, torque, and motor test and found to be inoperable load. A direct, if not perfectly linear, The assumption was made (and only when called upon to function relationship has been verified to still remains to be completely later. Such a situation is obviously exist between motor load and i
proven) that in situ testing can be unsatisfactory but, until the recent operatortorque. However,the used as the primary toolto main-development of signature analysis relationship of torque and thrust, as 2
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i applied to motor-operated gate operation. Generic Letter 89-10 greater than expected. This is valves, is not well understood. The extends the verification of operabil-believed to be due to large clear-usual practice in the industry has ity at design basis conditions to all ances between the valve compo-been to provide sufficient torque to safety-related MOVs. This will nents that permit the disc to tilt I
envelope the losses that occur in place several additional MOV types during closure and come in contact l
the operator when it converts under closer examination, and it is with the valve body. For some MOV torque to thrust. It was discovered expected that problems other than assemblies, this interference could l
that the losses may not have been those that have previously been cause the disc to stop before the l
conservatively estimated in the identified will be brought to our valve is completely closed, and the l
usual engineering methodology.
attention. These new problems continuing flow would cause the may require additional or modified safety concern described above.
The various vendors of diagnostic testing and evaluation schemes.
systems are now developing The tests have also revealed that strategies and hardware to meas-Flow-interruption the methods used by the utilities to ure thrust directly. There appear to determine the valve closing capa-Tests of Motor-be other factors such as the bilities could contribute to the number of strokes and time be-Operated Valves safety concem. specificaily, the l
tween strokes that affect the thrust Gerald H. Weidenhamer, DE/EMEB thrust required to close the valve is developed by an MOV. Again, it usually measured with the valve l
still remains to be shown that The RES Division of Engineering already installed in the plant.
required thrust can be reliably sponsored a series of 14 tests at Therefore, high-velocity flows I
extrapolated from a test conducted the Wyle Laboratory in Huntsville, typical of those that can be experi-l at less than design basis condi-Alabama, to evaluate the capability enced during a pipe break cannot j
tions (either in situ or prototype) of specified motor-operated gate be simulated. Instead, a utility will l
because of the variability of the valves (MOVs) to close against the subject the valve to low flows and l
thrust developed by the operator.
high-velocity flows that can occur in measure the closing or opening l
Factors such as valve stem design.
the event of a guillotine pipe break thrust at this reduced condition.
valve trim material, number of outside the containment building.
The measured thrust is then strokes, time between strokes, The main safety concern is that, if extrapolated to estimate the thrust l
temperature, wiring size, voltage, the MOVs fail to interrupt the high-that would be required at the accident condition. The concern is stem lubrication, and spring pack velocity flow, emergency equip.
adjustment, among others, are all ment needed for controlling the that, if the closing thrust character-pertinent. Load duration and reactor would be exposed to harsh istics of the valve are unpredict-intensity may be a major considera. environments that could cause the able, extrapolation will grossly l
tion. The individual and combined equipment to fail. Sufficient test underestimate the closing thrust l
effects of such conditions are not information at this flow condition requirements.
l completely understood.
was not available either in the open literature or from valve manufactur.
In addition to the information The action statements in Bulletin ers and utilities, and it was nec-discussed above, the tests pro-85-03 were focused primarily on essary to perform these tests to get vided information important to an switch setting adjustment. The the needed information. The understanding of valve behavior.
l implication was that switch adjust-information obtained from these All of this information is contributing ment was all that was necessary in tests is intended to contribute to to the resolution of GI-87 and is order to ensure design basis the basis for resolving a high-Providing the NRC staff with the operability. A number of safety-priority generic safety issue desig, basis for evaluating industry related MOVs are ball, plug, or nated GI-87," Failure of the High responses to regulatory actions in butterfly types, however, which do Pressure Coolant injection (HPCI) this area. However, the staff and not require the development of Steam Line Without isolation."
the industry believe that more thrust in the valve stem in order to information and data are required operate. These types of MOVs do to understand MOV performance under accident conditions and to not depend on the adjustment of a The tests showed that, although torque switch for the valve to the MOVs were able to interrupt answer other questions important achieve the proper closure posi-and shut off the high-velocity flow in to this and other safety concerns.
tion. The majority of MOVs covered all 14 tests, the thrust required to Therefore, additionaltests will be by Bulletin 85-03 are gate or globe move the disc to the closed posi.
conducted to obtain the necessary valves, which usually develop tion for one of the valves was data.
l thrust in the valve stem during i
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RES Revises The plan focuses research on operators were attempting to its Severe several near-term safety problems restart the recirculation pumps e
while maintaining the continuity of when the reactor tripped on high l
l Accident Research ongoing research fundamentalto neutron flux (118%).
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I improving the basic understanding l
Fam of severe accident phenomena for NRC was concerned whether a l
Frank A. Costanzi, DSR/AEB regulatory needs. In the near term, severe reactivity transient causing l
the severe accident research fuel damage could occur on restart In May of 1988, the NRC staff program will address the issues of of a recirculation pump. The presented to the Commission its l
plan for closing severe accident direct containment heating, resulting analysis provided another
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issues for nuclear power reactors.
meltthrough of BWR Mark I contain-example of the capability of RES-
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ment liners by molten core debris, developed codes to respond to The plan set forth specific objec-l tives for both the NRC and the and the effects of adding water to a operational events and to improve degraded core in the course of an our understanding of their cause l
nuclear utilities to be acheved over accident. Also in the near term, the and significance. BNL's BWR plant the next three to five years for dealing with reactor accidents more program will examine theissues analyzer was used to reproduce associated with extrapolating the actual LaSalle incident on a f
severe than the design basis results from small-scale experi-model of the Browns Ferry core.
accidents specified in the Comm.es-sion's regulations. Included were ments to develop an understanding Calculations showed that restart of of henomena dominating an the recirculation pump could lead
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some specific objectives relating.
actual reactor accident and the role to a power spike that would trip the primarily to issues of early contain-of computer codes in understand-reactor. The power spike is caused ment failure and accident manage-ing severe accidents. The continu-by the insertion of subcooled liquid
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ment.
ing research from which the near-from the lower plenum and j
term efforts will be drawn ad-downcomer into the core. In order In order to be sure that its severe dresses such issues as the details to more fully understand the i
accident research program sup-of core melt progression, interac-implications of the LaSalle event, ported meeting those objectives.
tion between molten core debris further calculations were made with the NRC's Office of Nuclear Regula-and concrete, fission product the BNL plant analyzer using input
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tory Research (RES) undertook an chemistry, and materials proporties that would more closely represent l
examination of NRC's severe at very high (core melt) tempera-the LaSalle core, e.g.,8x8 fuel rods, accident research. Experts from tures.
radial power peaking, and bundle within the NRC, the national exit pressure drop coefficient.
laboratories, universities, and Analysis of LaSalle These calculations showed that:
private enterprise were asked to review the present program with 2 Event
- 1. The LaSalle conditions lead to regard to the tasks that needed to Charles R. Troutman, DSR/RPSB limit-cycle oscillations; be done to support closure of
- 2. The LaSalle conditions, within severe accident issues and to On March 9,1988, at the LaSalle 2 the uncertainty envelope, pro-
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identify and recommend the plant (a BWR/5 in Illinois) both duce oscillations leading to 1
l research needed to achieve the recirculation pumps tripped on a automatic trip;
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l severe accident closure objectives.
false ATWS signal, a high-water-
LaSalle is caused by the com-
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dations of those experts as a point heating was automatica!!y reduced bination of radial power peak of departure, the RES staff revised because of the rapid power reduc-ing, flow reduction due to trip the Severe Accident Research tion (84% to 44%). The combina-of the recirculation pumps, FW Program Plan. The revised plan tion of reduced feedwater tempera-temperature reduction due to was reviewed extensively both ture and recirculation flow coast-reduced FW heating; and within and outside the NRC and by down to the low-flow /high-power
- 4. The amplitude of power oscil-the Advisory Committee on Reactor natural circulation region of the lations remains bounded even
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Safeguards. The resulting plan power / flow map resulted ir the if failure to tripis assumed.
l 1
tras presented to the Commission onset of neutron flux oscillations i
l in May of this year and was pub-due to thermohydraulic instability.
Nonuniform oscillations were i
lished as NUREG-1365, " Revised After about 5 minutes, the neutron studied at BNL using the RAMONA-i i
Severe Accident Research Program flux was oscillating from 20 to 50%
3B code on a Browns Ferry model Plan: FY 1990-1992."
on the average power range in order to gain insight on the monitor (APRM) scale. Plant causes and effects of multi-l l
4
s dimensional neutronic oscillations.
against the same data that were treatment for the thermal hydrau-RAMONA showed that hydraulic used with TRAC. The results lics, together with other necessary oscillations that begin in a few high-calculated by the two codes will modeling capabilities to describe a powered fuel bundles (given 60%
then be compared to confirm reactor system.
reactor power and natura! circula-technicalconclusionsof the tion flow) can cause regionwise analyses.
A report, NUREG/CR-5069, " TRAC-oscillations in the powe !A part of PF1/ MOD 1 Correlations and the core is increasing in power To avoid instability in normal Models," has been published to while at the same time the other operation, the nuclear steam complement the TRAC-PF1/ MOD 1 part is decreasing. The oscillation supply vendoris proposing the correlations and models utilized by shape is important since it affects redefinition and automation of the code. These correlations and the measured (viewed or recorded)
SCRAM setpoints of the power /
models provide the necessary local power range monitor (LPRM) flow regime map. Results of the coefficients required by the field output. Output from LPRMs at RES analysis of the La Salle 2 event equations and also provide models particular axial and radial positions are providing sensitivity analyses (e.g., decay heat, pump pressure) in the core is used to develop for independent NRR audit of the necessary for the transient simula-average power range monitor vendor's calculations. The results tions. The report also provides the signals and to trip the reactor on will also be used to confirm the technical basis for the code high flux. For a set of high-power /
adequacy of current procedures in through references to original low-flow states in the region of the event of ATWS. Here, research literature and a description of the minimum stability on the power / flow is supporting NRR by independ-development process, which lists map, power oscillation amplitudes ently determining the magnitude of the assumptions made. Because up to 300% (of rated) were calcu.
the power excursions and whether of the added generality in the lated with periods ranging from 2.6 suppression pool temperatures are models and closure relations, the to 2.9 seconds. Core-wide (in-kept within limits.
enhanced user convenience phase) osciliations were calculated, features, and the added capabili-as well as regionwise power Report Published on ties to track the boron solute in the distribution oscillations due to out-liquid phaseand theincondon-of-phase parallei :hannel flow TRAC Computer Code sable gas injected from accumula-instabilities. Two types of region-Models and tors, the TRAC-PF1/ MOD 1 code is wise oscillations were calculated:
Correlations recommended over previous (1) azimuthal oscillations in which versions for general use in ad-the power distribution shifted from dressing licensing problems and side to side across a core diameter To ensure the safety of the many questions relating to PWRs.
and (2) radial oscillations in which types of reactor plants in the U.S.
the power distribution oscillated for a wide range of normal and In the process of reviewing the between the center and periphery abnormal operations, the NRC.is details of the code and writing the of the core. For the cases calcu.
required to independently assess report, Los Alamos National Labo-each licensee's asseitions and lated, the regionwise oscillations ratory did not uncover any signifi-were driven by a few unstable fuel performance of its responsibility t cant errors. However, they did find channels. The excitation threshold design, construct, and operate a some minor inconsistencies in the of each mode of oscillation is a reactor with respect to the safety of coding and some correlations /
complex function of the local the plant for the complete spectrum models that today are not repre-channel inlet flow, subcooling, of credible operating conditions sentative of the best available in the and events. This need has led t bundle power, and axial subcritical-Iterature. The next (final) version of ity of the mode being excited.
the development of computer the code will resolve the incon-Additionalwork using the large codes designed to evaluate a sistencies and install new correla-thermohydraulic system codes broad spectrum of plant designs tions/models to replace those with TRAC-BWR and RAMONA is and plant transients. The current deficiencies, continuing. The TRAC effort code used by NRR to perform involves validating the code against these independent assessments for PWRs is TRAC-PF1/ MOD 1, an relevant experimental data followed p
by analysis of the LaSalle event.
advanced, best-estimate computer Modifications to the RAMONA program for calculating real and Examination Workshop code, including provision for more postulated transients. The code features a one-or three-dimen-The Commission's Severe Accident accurate two-phase flow reversal at the core inlet, will be validated sional two-fluid (liquid and gas)
Policy Statement called for system-5
l r
1 atic examinations of all plants to opinion that the meeting had been length of 1 micrometer) beam of discover plant-specific vulnerabili-very usefulin explaining the IPE radiation through the optical fiber.
ties to severe accidents. The process. NUREG-1335 was revised This " interrogating" radiation Individual Plant Examination (IPE) iollowing a careful review of the causes the trapped electrons to fall program, an integrated systematic workshop transcript and the written back to their ground state, which, in approach to examining each comments and questions received turn, causes the phosphor to nuclear power plant now operating and was issued for use by licen-luminesce, emitting radiation in the l
or under construction for possible sees in preparing their IPE submit-visible region. This visible radiation significant risk contributors that tals.
then travels back through the l
might otherwise be overlooked,is a optical fiber to the readout position key part of the implementation of NRC-Sponsored Small where the visible light is converted this Policy Statement. The staff into electric current by allowing it to issued Generic Letter 88-20 to all Business Research impinge on the photocathodes of a licensees in November 1988 Contract Successful; photomuttiplier tube. Conversion of requesting these examinations.
ele tri I signals t r di ti n doses Optical-Fiber-Coupled i
Guidance on what the NRC staff or dose rates is accompiished by l
will expect in the IPE submittals Radiation Dosimeter calibration in known radiation fields.
cas issued as a draft for comment Meets Project Goals The prototype dosimeter utilizing in January 1989 (NUREG-1335, "in-an optical fiber 0.5 mm in diameter dividual Plant Examination: Submit-exhibited a linear response to Work was completed on a small-tal Guidance"). The Genenc Letter wsbnM aM mbam gamma business innovative research stated that a workshop would be rays and to x-rays of energies from (SBIR) Phase 11 contract with the held to receive questions and 50 to E h W dewa@n fmm Quantex Corporation of Rockville, comments oq the IPE process and Unea% was @ sed mee the submittal guidance.
MD, to develop a prototype rateme.
measured ranges, which, for ter/ dosimeter capable of remotely am, was in a ss om N measuring radiation 'ields. It uses The IPE Workshop, held February mu de sh a solid state radiatior sensor 28 through March 2,1989, was rays and gamma rays is about 2 coupled to an opticalt5er to attended by an estimated 350 W
a measwemms wem rovide remote readout. The final representatives from the utilities' Iso made, but additional measure-report, NUREG/CR-5100, "Integrat-vendors, industry groups, consutt-n s wm N neededo estam l
ing firms, and the NRC staff and ing Fiber Optic Radiation Dosime-range of ha%. Mmm tec" was published in March 1989.
contractors. Representatives from detection limits for beta particles,s i
four foreign countries (The United e
aM m M aW 20 wh Kingdom, the Federal Republic of The research clearly demonstrated that a radiation-sensitive solid-state Germany, Japan, and Spain) als Potential uses for the fiber-optic-attended. In addition to the IPE storage phosphor coupled to an coupled radiation dosimeter l
process, the workshop included opticalfiber was capable of provid-e mmm radadon sensing at
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ing measurements of radiation discussions of staff plans for difficult-to-reach sites and possible fields at considerable distances accident management, the Inte-use as a swa posWon sensoda from the fields. The technology grated Safety Assessment (ISA) radiographic exposure devices. In involves the use of " electron program, and the relationship of viv r di ti n dose rate measure-s m in e m e the IPE to other parts of severe men s may M possde. De smaH accident issue closure such as the energy of the radiation field excites size of the detector and optical fiber electrons in the phosphor to higher Containment Performance improve-may make it possible to use energy trapping levels where they l
ment Program. A wide range of bundles of detectors,n parallelfor i
1 remain trapped until released by comments and questions reflect.ing Possible continuous readout of some external stimulus. The I
various degrees of understanding n
n, gamma, a,a d w ay number of electrons captured in of the intent of the IPE process and s.
these trapping levels is directly a wide range of utility experience rocortionalto the total amount of with probabilistic safety analyses Control of Water infiltra-radiatiun incident on the phosphor.
were received. The workshop was viewed as a very successful and tion Through Low-Level Stimulation of the trapped electrons informative meeting for all involved.
Waste Disposal Covers can W caded wt fmm a mm The staff received significant (kilometers) readout position by constructive comments, and transmitting a near-infrared (wave-The heavy rains of April and May numerous attendees expressed the 1989 provided a severe test of 6
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covers for controlling the entrance stressed condition, that is, an assurance (QA) procedures in the of water into low-level radioactive overdraft condition. The vegetation medical use of byproduct radioac-waste disposal units being evalu-actively scavenges the small tive material. Items in the regula-ated at Beltsville, MD. Water amount of water available in the soil tory guide on the use of radiophar-entering the disposal unit can below the cover. Pfitzer juniper, a maceuticals in nuclear medicine provide a mechanism for transport-drought-resistant evergreen, is and the use of sealed radioactive ing radionuclides beyond the used because of its hardiness and sources in teletherapy and bra-disposal site. One cover design, tolerance of wet and dry conditions. chytherapy were covered. The called bioengineering water man-Designs 1 through 3 will failif there participants - physicians, physi-agement, proved to be particularly is appreciable ground subsidence cists, dosimetrists, and technolo-effective not only in controlling whereas bioengineering water gists - provided usefulinformation water movement through the management is designed to on current GA practices in their covers but also in dewatering accommodate subsidence.
institutions and helpful suggestions flooded disposal units.
on how to improve the working Preliminary results for bioengineer-drafts. They indicated that they are The research to support these ing water management are most already following about 90% of the evaluations is sponsored by the encouraging. The two Beltsville QA procedures proposed in the Waste Management Branch, test cells with bioengineering draft regulatory guide.
Division of Engineering, and the covers were initially partially filled work is being performed co-with water (artificial water table) in The staff considers that the work-operatively by the University of 1987 to simulate flooded disposal shop was an informative and useful Califomia at Berkeley and the units. One test cell had 2 meters of exchange of information and ideas University of Maryland. The tests waterwhen data collection began.
with working-level medical profes-are carried out in large test cells The water level had been lowered sionals. RES carefully considered (45' x 60' x 12') that are instru-to 30 cm, and the heavy rains of the information obtained from the mented to measure infiltration and April and May 1989 resulted in a workshop in preparing the pro-to provide a complete water bal-rise of only 1 cm. The other test posed rulemaking package, in ance. Four cover designs are cell began with 1 meter of water, brachytherapy, the participants being assessed: (1) a resistive and it was completely dewatered suggested using clearly marked layer barrier system (i.e., a com-by the combination of enhanced storage spaces for each type of pacted earthen material such as runoff and stressed vegetation.
sealed source to identify the clay) beneath a rock cover (such as The waterlevelin that test cell did radionuclide and the source is used in uranium milltailings not increase following the heavy strength of the sealed sources.
covers) in one test cell and beneath rains. The soil beneath that cover This was included in the draft a vegetated cover in another test has been dewatered to such a regulatory guide. In teletherapy, cell, (2) a conductive layer barrier degree that the soil beneath the the participants suggested that the system that combines a conductive cover remains in an unsaturated measurements for each of the layer such as a fine sandy loam condition. The decline of the water treatment plans (e.g., tangential over a capillary break to wick water table in the bioengineering test breast, lung, rotation on a pros-around waste, (3) a cc-Mation of cells shows that bioengineering trate, and mantel field with irregular (1) and (2) in which the conductive water management could be used sources) were unnecessarily layer barrier is placed below the as remedial action (drying out) for complex and should be simplified, resistive layer barrier to wick away existing water-logged disposal The draft regulatory guide was the small amounts of water that will sites.
modified to incorporate this sug-pass through it, and (4) bioengi-gestion.
neering water management, which Workshop with Medical employs a surface cover of imper-Use Ucensees Exchange with United meable matenalto enhance runoff in conjunction with vegetation 9
placed in gaps between panels to A workshop with licensees who use Authority on Structural remove the small amounts of water radioisotopes for medical purposes that might leak through them. In was held on January 30 and 31, Integn,ty Research effect, the vegetation acts as a 1989, to discuss the working drafts DE/MEB pump powered by solar radiation.
of a proposed rule and regulatory Since the amount of water passing guide intended to reduce the The United Kingdom has begun an through the cover is small, the potential for misadministrations and earnest program to assess the vegetation is in an environmentally overexposures by improving quality structuralintegrity of the major 1
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components of its aging reactors.
Reactor Safety (JCCCNRS) ex-plastic fracture mechanics with.
Because this activity is very com-change meetings held in Rockville application to pressurized thermal piementary to work supported by in early June. In the subject area of shock analysis.
the RES Materials Engineering radiation embrittlement and anneal-Branch, RES is exploring a pro-ing, details were presented of the International Piping gram involving an exchange of background research used by the information and cooperation in Soviets as a basis for their current Integrity Research each other's projects and facilities program of annealing operating Group withrepresentativesof thelabora-reactors that have reached unac-EB l
toriesinvolved. The exchange has ceptably high levels of embrittle-already begun with independent ment. At the time of the meeting, analyses of the results of the first they had already completed five in April 1989, the italian Commis-sion for Nuclear and Altemative two UK spinning cylinder tests that vessel annealings and had two simulate stress buildup in vessels more in process. The Soviet Energy (ENEA) began its participa-through high-speed rotation and of representatives provided an tion in the Intemational Piping l
the first two pressurized thermal excellent overview of the electncal Integrity Research Group (IPlRG), a I
shock tests at Oak Ridge National resistance heater system devel, consatium of govemnwnt and Laboratory, oped for annealing the vessels and industrialorganizations formed to described howits effectiveness jointly fund research on the integ-j The British are also completing was verified both by testing of rity of piping subjected to seismic their Structural Features Test surveillance capsule specimens and dynamic loading, as well as, Facility that can apply 20 MN of and by direct hardness measure-other piping integrity issues within force biaxially to large plates of ments made by a workerlowered the group's area of interest.
steeland can also accept loading into the vesselin a shielded work in the third direction as well. The
- chamber, Italy thus joins with other partici-capacity and versatility of this pants, including representatives machine is such that NRC intends While the information presented by from Canada, France, Japan, to review its own program carefully each country was very interesting Sweden, Switzerland, Taiwan, and to determine if work of benefit to to the other, the direct applicability the United Kingdom, in funding the the U.S. should be conducted is limited because of design IPIRG Program at Battelle's Colum-using this machine. Specific areas differences between the two bus Division. Through their fi.
where the capabilities of this systems. The Soviets use a Cr-Mo nancialcontributions to the pro-machine may be usefulinclude steelthat is sensitive to P content gram (fmeign pa tners contribute evaluating the effects of biaxial for radiation embrittlement, while more than 2/3 of the total cost), all loading on crack behavior. The the U.S. uses a Mn-Mo steel that is pa ticipants gain valuable informa-results of such tests are used in more sensitive to Cu for radiation tion on the fracture behavior of full-validating analyses for pressurized embrittlement. Furthermore, the scale primary system pipe under thermal shock and for low-upper-Soviet reactors operate at about full operating pressure and tem-shelf welds. Other cooperation is 500 F while the U.S. reactors Perature conditions.
under way in the area of non-operate at 550 F; this temperature destructive examination, and more difference has a significant effect Final checkout of the Battelle facility is expected in the future as the on the amount of embrittlement, the for testing full-size pipe under entire area of primary system higher temperature yielding less dynamic loads simulating earth-integrity becomes more thoroughly embrittlement. Despite these quakes is under way in Columbus.
studied and the cooperation differences,it is expected that the Tests in this facility will be the develops.
information exchange will broaden culminating activity of the IPIRG NRC's irradiation effects data base Program. The facility incorporates Radiation Embrittlement and thus permit validation og a loop with anchor points, re-predictive models over wider straints, and bends representative Exchange with Soviets ranges of key parameters. This, o, of a typical piping ioop. Thus the DE/MEB course, would lead to more defen.
tests will give realistic indications of iP Pe perfumance in service. The sible regulatory decisions.
A highly interesting and informative results of these tests will be used in exchange was concluded with Plans were made for a second validating the flaw evaluation representatives of the Soviet Union meeting to try to sum up the analyses embodied in section XI of as part of the Joint Coordinating annealing data and experience and the ASME code and in validating Committee for Civilian Nuclear to initiate an exchanne on elastic, the pipe fracture analyses being 8
o included in NRC's leak-before-time / temperature / cooling-rate ent vibration level typical of the break evaluation procedures. The matrix to establish reference seismic excitations that might be initial shakedown tests have shown microstructures for comparison experienced in the U.S., i.e., from the system to work exceptionally with microstructures of the real 0.27 g to 2.0 g ZPA. The electrical well, with pipe system response lower head material when it be-capacity of the batteries and the matching very closely the analytical comes available in 1990.
resistance and capacitance of the predictions. The scheduled tests internal plates, conductors, and i
will subject cracked pipe to loads The two groups established for electrolytes were measured before simulating earthquakes.
oversight of this program, the and after each test.
Program Review Group and the Integrity of the TMI-2 Management Board, met early in Preliminary test results indicate that Vessel Lower Head September in Harrisburg, PA, to the batteries did not experience review the latest detaiis of defu-any loss in electricalcapacity as a DE/MEB eling and to make a visit to the result of the vibration that simulated TMI-2 site for a first-hand look at seismic excitation. Some local l
On January 1989, the NRC signed activities and new features being damage, however, was observed a new eleven-country agreement revealed by the final stages of the on the racks and on the batteries i
on a cooperative program to cleanup. Such information pro-where interference occurred during i
l examine and evaluate metal vided the basis for a new priority the tests at higher excitations.
specimens to be cut from the lower listing of locations for sample Visual inspection also revealed that head of the TMI-2 reactor pressure acquisition.
a small amount of fiber glass insu-vessel, which was damaged by tation came loose from some of the
]
melted fuelin the 1979 accident.
Seismic Testing of Plate separators; however, there l
was n widence of damage to the The program will determine the Naturally Aged Batteries lead plates or cracks in the cases properties of the steeland the
)
integrity of the vessel at the time of of any of the three battery groups, the accident and thus its margin to Ag.ing is a key concern with More thorough evaluations and failure. Providing a better estimate currently operating plants and will inspections are being performed at of the margin to failure could affect clearly be crucialto any assess-the Idaho National Engineering accident management develop-ment of the safety implications of Laboratory.
ment and will provide an improved license renewal Aging affects all j
basis for evaluating the course of reactor structures, systems, and Theinformation gained in these severe accidents and the time to compomnts and has the potential tests will add to the data base on vesselfailure. In addition, the to increase nsks to public health aging that is needed to serve as a program will provide a data base of and safety. One of the concerns is basis for regulatory decisions on related damage mechanisms to be the effect of aging on the capability license renewal. It may also prove used in support of a broad range of e stmctwe, system, or compo-useful in identifying possible contri-nont to withstand se,smic events.
butions to risk in the IPE program.
i severe accident evaluations and Seismic tests of 12 LCU-19 C&D accident mitigation programs.
batteries that were aged in actual service to determine whether these RESEARCH NEWS is published by While the TMI-2 defueling is still under way and sample extraction batteries could have adequate the Office of Nuclear Regulatory methods are being developed, the electrical capacity but inadequate Research, USNRC.
seismic ruggedness were com-integrity assessment activities were P eted at Wyle Laboratory in This issue was prepared by:
i started with the acquisition of steel Alabama. The batteries were Edward L Hill, Editor, and from the lower head of the aban-btained from Arkansas Power &
Charles B. Bartlett, doned Midland reactor vessel, a materialthat matches the TMI-2 Light Co. after they had been Technical Coordinator vessel materia!very closely, to removed from ANO-2, where they were used as Class 1E station Future issues will be prepared by:
serve as a source of archive material for the TMI-2 studies. It is batteries for over 13 years. The Robert L Shepard, MS NL-007 batteries are over 14 years old.
U.S. Nuclear Regulatory being used to intercalibrate allthe testing laboratories to ensure Commission, Washington, DC 20555 attenes wwe asseN &
uniforrnity in measurement and the three groups of four batteries; Comments, suggestions, and articles reporting of values;it is also being used as a source of materialfor a each group was tested at a differ-for futureissues should be directed to Dr. Shepard.
9
i i
REGULATORY GUIDES (CE802-5), Nuclear Criticality Safety i
for Steel-Pipe Intersections Con-Regulatory Guide 8.12, Revision 2
{
ACTIVE REGULATORY taining Aqueous Solutions of Fissile (CE801-5), Criticality Accident GUIDES ISSUED FY 1989 Materials. Issued 04/89.
Contact:
Alarm Systems. Issued 10/88.
K. Steyer, RES/DE/MEB.
Contact:
K Steyer, RES/DE/MEB.
M 815 e gn a 5 abr a ion Regulatory Guide 3.48, Revision 1 Regulatory Guide 10.9, Revision 1
{
Code Case Acceptability, ASME (CE406-4), Standard Format and (FC402-4), Guide for the Prepara-l Section 111, Division 1. Issued 07/
Content for the Safety Analysis tion of Applications for Licenses for 89.
Contact:
E. Woolridge, RES/
Report for an Independent Spent the Use of Self-Contained Dry j
DE/MEB.
Fuel Storage Installation or Moni-Source-StorageIrradiators. Issued j
tored Retrievable Storage Installa-12/88.
Contact:
S. Baggett, NMSS.
Regulatory Guide 1.85, Revision 26 0 n (Dry Storage). Issued 08/89.
Contact:
C. Nilsen, RES/DE/MEB.
DRAFT REGULATORY GUIDES (ME802-5), Materials Code Case Acceptability, ASME Section 111, Regulatory Guide 3.50, Revision 1 i
- r5ge, ES E
(CE402-4), Guidance on Preparing Regulatory Guide DG-1001, Mainte-j a License Application to Store nance Programs for Nuclear Power Regulatory Guide 1.114, Revision 2 Spent Fuel and High-Level Radio-Plants. Issued 08/89.
Contact:
M.
active Waste. Issued 09/89.
Dey, RES/DRA/ARGIB.
(HF601-4), Guidance to Operators
Contact:
W. Pearson, RES/DRA/
at the Controls and to Senior RDB.
Operators in the Control Room of a Regulatory Guide DG-1003, Assur-Nuclear Power Unit. Issued 05/89.
ing the Availability of Funds for
Contact:
- 1. Schoenfeld, RES/DSR/
Regulatory Guide 3.61 (CE306-4),
Decommissioning Nuclear Reac.
Standard Format and Content for a tors, issued 05/89.
Contact:
HFB.
Topical Safety Analysis Report for a F. Cardile, NMSS.
Regulatory Guide 1.147, Revision 7 Dry Spent FuelStorage Cask.
issued 02/89.
Contact:
W. Pear-Regulatory Guide DG-1005, Format (ME803-5), inservice Inspection son, RES/DRA/RDB.
and Content of Nuclear Reactor Code Case Acceptability, ASME Section XI, Division 1. Issued 07/
Decommissioning Plans. Issued 89.
Contact:
E. Woolridge, RES/
Regulatory Guide 3.62 (CE301-4),
09/89. Contacts: F. Cardile, Standard Format and Content for NMSS; K Steyer, RES/DE/MEB.
DE/MEB.
the Safety Analysis Report for Regulatory Guide 1.157 (RS701-4),
Onsite Storage of Spent Fuel Regulatory Guide DG-1006, Rec-Best Estimate Calculations of Storage Casks. Issued 02/89.
ords important for Decommission-
Contact:
W. Pearson, RES/DRA/
ing of Nuclear Reactors. Issued 09/
Emergency Core Cooling System RDB.
- 89. Contacts: F. Cardile, NMSS; K.
Performance. Issued 05/89.
Contact:
H. Tovmassian, RES/
Steyer, RES/DE/MEB.
DRPS/RPSB.
Regulatory Guide 3.64 (WM5034),
Calculation of Radon Flux Attenu-Regulatory Guide DG-3001, Rec-Regulatory Guide 1.158 (EE006-5),
ation by Earthen Uranium Mill ords important for Decommission-Qualification of Safety-Related Lead Tailings Covers. Issued 06/89.
ing for 10 CFR Parts 30,40,70, and
Contact:
G. Birchard, RES/DE/
72 Licensees. Issued 07/89.
Storage Batteries for Nuclear WMB.
Contact:
F. Cardile, NMSS.
Power Plants. Issued 02/89.
ct: S. Aggarwal, RES/DE/
Regulatory Guide 3.65 (CE304-4),
Regulatory Guide 1.9, Proposed Format and Content of Decommis-Revision 3 (RS802-5), Selection, Regulatory Guide 3.44, Revision 2 sioning Plans for 10 CFR Parts 30, Design, Qualification, Testing, and 40, and 70 Licensees. Issued 08/
Reliability of Diesel Generator Units (CE403-4), Standard Format and 89.
Contact:
F. Cardile, NMSS.
Used as Onsite Electric Power Content for the Safety Analysis Report for an Independent Spent Systems at Nuclear Power Plants.
Fuel Storage Installation (Water-Regulatory Guide 7.8, Revis. ion 1 issued 11/88. (Note: Active guide Basin Type). Issued 01/89. Con-(MS804-4), Load Combinations for in editing.)
Contact:
A. Serkiz, tact: W. Pearson, RES/DRA/RDB.
the Structural Analysis of Shipping RES/DSlR/RPSI.
Casks for Radioactive Material.
issued 03/89.
Contact:
W.
Regulatory Guide DG.7001, Frac-Regulatory Guide 3.45, Revision 1 Campbell, RES/DE/EMEB.
ture Toughness Criteria for Ferritic 10
RES PETITIONS FOR PRM-100-2, Population Density 05/02/89 (54 FR 18649).
Contact:
Criteria Near Nuclear Power Plants.
J. O'Brien, DE/SSEB.
RULEMAKING Petitioner: Public Interest Research Group, et al. Denial of the petition PETITIONS COMPLETED FY 1989 was published in the Federal PS06, Program for Resolution of Register on 12/14/88 (53 FR Generic issues Related to Nuclear PRM-31-4, Use of Phosphorus-32 in 50232).
Contact:
J.Telford, DRA/
Power Plants (10 CFR 50). With-Salmonella and Listeria Assays.
RDB.
drawal published in Federal Regis-Petitioner: Gene-Trak Systems.
ter on 06/07/89 (54 FR 24432).
Withdrawal of petition published in RES POLICY
Contact:
W. Milstead, DRA/ARGIB.
]
Federal Register on 03/14/89 (54 FR 10550).
Contact:
H. Scott, STATEMENTS DRA/RDB.
PS02, Education and Experience COMPLETED POLICY Requirements for Senior Reactor PRM-5048, Redefine " Testing Operators and Supervisors at ST ATEMENTS FY 1989 Facility" Based on the Function of Nuclear Power Plants (10 CFR 50).
the Facility Instead of its Power Final policy statement published in Level. Petitioner: University of PS14, Additional Applications of Federal Register on 08/15/89 (54 Missouri. Denial of petition pub-Leak-Before-Break Technology FR 33639).
Contact:
M. Fleishman, lished in Federal Register on 04/25/ (10 CFR 50). Final policy statement DRA/RDB.
89 (54 FR 17962).
Contact:
M.Au, published in Federal Reg, ster on i
DRA/RDB.
UNITED STATES mst etass uma NUCLEAR REGULATORY COMMISSION postaog,e es enio WASHINGTON, D.C. 20555 OFFICIAL BUSINESS PENALTY FOR PRIVATE uSE. $300 ef 19AI9819cjggy Y
f,P chCfN T pooy NCH LLj T %
y, d 5HINGT3N Oc 20555 l
i l
4 i
f 12 i
i 1
1 l
Steel Shipping Container Contain-Register on 02/03/89 (54 FR 5409).
Register on 11/28/88 (53 FR ment Vessels with a Maximum Wail
Contact:
J. Lambert, DE/WMB.
47822).
Contact:
M. Dey, DRA/
Thickness of FourInches (0.1m).
ARGIB.
i l
Contact:
W. Campbell, RES/DE/
RM192, Centralization of Material l
EMEB.
Control and Accounting Ucensing RM104, Education and Experience and Inspection Activities for Non-Requirements for Senior Reactor Regulatory Guide DG-7002, Frac-Reactor Facilities (10 CFR 70,74).
Operators and Supervisors at l
ture Toughness Criteria for Ferritic Final rule published in Federal Nuclear Power Plants (10 CFR 50, Steel Shipping Containers with a Register on 02/15/89 (54 FR 6876).
55). Proposed rule published in WallThickness Greater Than Four
Contact:
S. Dolins, DRA/RDB.
FederalRegister on 12/29/88 l
Inches (0.1m).
Contact:
W.
(53 FR 52716). (Note: The final l
Campbell, RES/DE/EMEB.
RM109, Flow Control Conditions for rulemaking was terminated on 05/
the Standby Uquid Control System 17/89.)
Contact:
M. Fleishman, REGUI.ATORY GUIDES in Boiling Water Reactors (10 CFR DRA/RDB.
l WITHDRAWN FY 1989 50). Final rule published in Federal l
Register on 04/03/89 (54 FR RM186, Palladium-103 foiintersti-Regulatoly Guide 1.74 (DS061-3),
13361).
Contact:
W. Pearson, tialTreatment of Cancer (10 CFR l
Quality Assurance Terms and DRA/RDB.
35). Proposed rule published in Definitions. Guide issued 02/74; Federal Register on 04/06/89 l
withdrawn 09/89.
Contact:
O.
RM024, Emergency Preparedness (54 FR 13892).
Contact:
A.Tse, Gormley, RES/DRA/ARGIB.
for Fuel Cycle and Other Radioac-DRA/RDB.
tive Material Ucensees (10 CFR 30, l
Regulatory Guide SC708-4 (Divi-40,70). Final rule published in RM112, Storage of Spent Nuclear i
sion 1), Qualification and #ccep-Federal Register on 04/07/89 Fuelin NRC-Approveo Storage tance Tests for Snubbers Used in (54 FR 14051).
Casks at Nuclear Power Reactor l
Systems important to Safety.
Contact:
M. Jamgochian, DSIR/
Sites (10 CFR 50,72,170). Pro-l Guide issued 02/81: withdrawn 04/
SAIB.
posed rule published in Federal 89.
Contact:
W. Campbell, RES/
Register on 05/05/89 i
l DE/EMEB.
RM182, Access to Safeguards (54 FR 19379).
Contact:
Information (10 CFR 73). Final rule W. Pearson, DRA/RDB.
I RES RULEMAKINGS published in Federal Register on 04/25/89 (54 FR 17703).
Contact:
RM135, Minor Amendments to FINAL RULEMAKINGS M. Au, DRA/RDB.
Physical Protection Requirements PUBUSHED FY 1989 (10 CFR 70,72,73,75). Proposed RM034, Disposalof Radioactive rule published in Federal Register RM114, Safeguards Requirements Wastes (10 CFR 61). Final rule on 08/15/89 (54 FR 33570). Con-l for Fuel Facilities Possessing published in Federal Register on tact: S. Dolins, DRA/RDB.
Formula Quantities of Strategic 05/25/89 (54 FR 22578).
Contact:
C. Prichard, DE/WMB.
RULEMAKINGS TERMINATED FY Special Nuclear Material l
(10 CFR 73). Final rule published PROPOSED RULEMAKINGS in Federal Register on 11/10/88 (53 PUBUSHED FY 1989 RM094, Deletion of Part 11 Require-FR 45447).
Contact:
S. Frattah, ment for Renewal of "R" Clear-RM109, Flow Control Conditions for ances (10 CFR 11). Termination of RM118, Alternative Methods for the Standby Uquid ControlSystem rulemaking approved by EDO on in Boiling Water Reactors (10 CFR 10/18/88.
Contact:
S. Frattah, Leakage Rate Testing (10 CFR 50).
Final rule published in Federal 50). Proposed rule published in DRA/RDB.
Federal Register on 10/24/88 (53 53 R 58
).
act: G. Arndt' FR 41607) (Note: The final RM104, Education and Experience rulemaking was published on 04/
Requirements for Senior Reactor DE/SSEB.
03/89).
Contact:
W. Pearson, Operators and Supervisors at DRA/RDR Ma&wMants OM %,
RM090, Criteria and Procedures for 55). The final rulemaking was ter-i Emergency Access to Non-Federal RM133, Ensuring the Effectiveness minated per a Commission deci-and Regonal WWeNaste of Maintenance Programs for sion on 05/17/89.
Contact:
l a@wMants Om mb M Nshan, MN pu in F Proposed rule published in Federal 11
!