ML20116H318
| ML20116H318 | |
| Person / Time | |
|---|---|
| Issue date: | 10/31/1992 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-BR-0125, NUREG-BR-0125-V04-N2, NUREG-BR-125, NUREG-BR-125-V4-N2, NUDOCS 9211130059 | |
| Download: ML20116H318 (12) | |
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TECHNICAL nun-nos NEWSLETTER M ?d Development of In 11ouse Analytical pre-processor reads initiai geometry and boundary conditions, and matenal descriptions and loading condo Capability and Its Application Ior tion,,,ng,1
,,n,,,,,,, comput,tsnsi gria (m,3) for Advanced Reactor Reviews user venfication. The main proceaor then apphen the conservation laws for mass, momentum and energy, and Yong S. Kim and formulates an equation of state based on an appropriate Davbl C. Teng, DE constitutive relationship for determination of ktrains, dis-placementa and forces through apphcation of finite ele-Abstract ment technique. Die post-processor then interactively da-plays or prints time history plots, stress and strain re-As part of our effort to grengdien the Cml Engineenng sponses, and rigid body motions of the system calculated and Geosciences Dranch's (ECOD's) independent analyti-from the main processor, cat capabihty, the staff has developed a microcomputer-based, in-house analytical capability for use in performirig 1he program of ECOB has a variety of analytical capabJ-independent hcensing reviews and safety analysis of Cate-ties. Some of them are (1) linear elastic analysis for chil gory I structures. In the course of its development over the structures and mechanical components, using small deflec-past year, ECOD/DE accomplished the following: (1) pro-tion theory; (2) static and dynamic analysis,.nclud ng cured a commercially av.tilable computer program for dead weight, pressure, thermal, displacement, wind and structural analysis and design with provisions for training earthquake loads; (3) natural frequencies and mode of staff to learn about the theoretical background and shapes calculation; (4) response spectrum analysis; and knowledge of the detailed capabihties of the program, (2)
(5) time history and frequency-domain response analyses.
installed microcomputers on whkh to use the program, The latest version of the program has addiuonal capabili-and (3) demonstrated the accuracy and effectiveness of Lies such as: (1) nonhnear analysis with large deflection the developed capabihty by comparing with mainframe-theory; (2) nonlinear restraints; (3) fluid-structure inter-based structural analyses. At present, the staff is utihring action; (4) corrosion analysis; and ($) post-failure model-this analytical capability to review compbcated structural ing.
issues for operatmg and advanced reactors. This article desenbes the main features of the analytical capabihty and The DOS version of the program is installed on an 80386 its apphcation for advanced reactor reviews.
model microcomputer in Room 71119, ECGD/DE, and the UNIX version of the program is mstalled on a SUN Introduction SpARC STATION 2 computer in Room 7 C20 for staff use.
ECOB chose the computer program, ALGOR, as a struc.
tural analysis and design tool from among those commer-Application cially available microcomputer-ba:ed structural programs after a comprehensive review of the analyucal apabihty.
Linear clastic analyses have been performed using the the computer aided-drawing and engineering (CAD /CAE) program for reviewing two advanced reactor containments abihty, the cost, and sendor's technical support of each (System 60+ and ABWR) under the hmited loadmg conde computer program. ALGOR is a three-dimensional (3 D) tions. Phyucal conditions and assumptions (e g., boundary imite element program for structural analysis and design.
conditions, material properties, geometrical dimensions, and employs a Galerkin finite-element formulation with no penetration, etc.) used in the analyses were same as different element types (e.g., truss, beam, membrane, t',se conditions and assumpoons used by the vendors in plate /shell, brick, pipe, gap / cable and boundary ele-their analyses.
ments). The computational process of the program con-sists of three processors: (1) pre processor (CAD), (2)
The 1.75 inch thick, 200 foot-diameter steel containment mam-processor (CAE), and (3) post processor (CAD).
of System 60+ was analped using the program with 720, 1
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Development of In-liouse Analytical Capabihty Sumniary and its Apphcation for Advanced Reactor As part of 1:CGil/DE's ef fort to strengthen its independent Renews analytical capabthly, the staf f has developed a microcom-by Yong S. Kim and puter based, in-house analytical capabihty for structural Dand C. Jeng. DET 1
analpis and design review of operating and advanced re-actors. The developed capability hah bcen utilired for Sp-Station Blackout / Electrical Safeguards tem 80+ and AllWR steel containment reviews, and the Upgrade Program for Prairie Island linear clastic analyses of ECOU have independently con-by Armando hiasciantonio. PP 3-1 5
hrmed the resuhs of the vendor's analyses for hmited loading conditions. Although its application for other li-Technical Specifications improvement Pragram censing reviews and safety analpis of Category I structures by h1ary Lynn Reardon at d is still in progress, the bmited conbrmatory studies that Christopher 1. Grimes. OTSB 7
have been completed to date have shown encouraging re-sults, and have demonstrated that the accuracy ar.d effec-NRC Staff Inspection Actmtics involvinE tiveness of the microcomputer-based structural analysis hiotor-Operated Valves are as accurate as the mainframe based structural analyses by Thornas G. Scarbrough, DET 8
widely used by industries and universities. ECGD/DE envi-sions that this newly developed in house analytical capa-Two-Year Trial Program for Conducting bility together with an ongoing developrnent of the ad-Open Enforcement Conferences vanced analytical capabilities by the NRR staff will play an by Renee pedersen, OE
.. 10 important role in future independent safety review of reac-tors.
F
-~1-Table I 3 D shell elements. The containment was fixed at the hinximum Stress intensity for System 80+ under bottom, and the cork support near the bottom of the 1)esign Loading Condition (Dead Load + internal containment was modeled with equivalent spring boundary Pressure of 53 psig) conditions. The finite eternent mesh for the analpis was generated by the pre-processor (CAD) of the program and Computed General Computed Local is shown In l'igure 1. Table 1 shows the results of ECGill hjembrarm Stress hiembrane Stress DE's and Combustion Engineering's (CE's) analyses and Organization Intensity Intensity the stress in ansities obtained from the design rules of ash 1E Dailer and Pressure Vessel Code, Section 111 Divi ECGB/DE 18,490 ps:
21,570 psi sion 1, Subsection NE for Class htC components.1he CE analpis was done by CE's engineering contractor, Duke CE/ Duke 18.150 p t 20,850 psi Engineering, usmg the mainframe program, ANSYS. As can be seen from the table, the results of the two analyses ash 1E 22,000 psi (*)
33,000 psi (')
mostly agree, and the results also met the AShiE Section 111 Codeallowable stress intensay requirements for the
(,):
Maximum allowable stress intensity provided by ASML Doller design loading conditmn.
and Pressure vessel code, section 111, Division 1, subsection The 1.25-inch-thick, thin walled torispherical head for the AHWR drywell was also analyred using this program with 1,440, 3 D shell elements. Based on a parametric study, a large number of hnite elements needed to be used to take TM 2 into account the anticipated high compressive stress in the knuckle region. Fixed boundary condition was applied at htaximum Stress Intensity for AllWR under Severe the junction where the steel dome intersects the cylindrical Accident I.oading Condition (Dead Load + internal p
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reinforced-concrete structure lu,gure. shows the mesh used for analysis, and Table 2 shows the results of ECOB/
DE's and General Electric's (GE's) analyses for severe ac-Computed h1aximum hiembrane cident loading condition. The GE analyris was done by Organization Stress Intensity using the mainframe program NASTRAN. The agree-ments between the results of the two analyses are excel-ECGB/DE 23.070 psi lent, and they meet the allowable stress intensity provided by the ASN1E Boiler and Pressure Yessel Code, Section GE 23,365 psi 111, Division 1, Subsection NE for Class h1C components.
As an illustration of the CAD capability of the program, ash 1E 33,300 psi (*)
the hoop membrane stress distribution and deformed shape of the ABWR drywell are plotted as shown in Fig-(, D Maximum allowable stress intensity provided by ASME Boiler ures 3 and 4, respectively. Note that a color hard copy for and Pressure vessel code. Section 111. Division 1, subsection better graphicM illustration can be obtained by using the NE for Class MC components.
2
SYSTEM 80+
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Figure 1 Finite Element Mesh Used for SYSTEM 80+ Steel Containment Analysis ABWR CONTAINMENT (5-0 Finite Clement Nesh. 1440 Plate /Shell Elements)
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Figure 3 Iloop Membrane Stress Distribution of AllWR Torispherical IIcad ABWR CONiRINMENT Def'ormed (Shepe of Drywe'll.
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Station Blackout / Electrical Safeguards Unit 2 and as an ahernate ac source of power for unit i.
Upgrade Program For Prairie Island Th' '*i$ti"8 D * " d D 2 di* l E'"
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5 ' 8 5 dedicated emergency ac pcwer for Urut I and as an alter-Armando Mascianlonio, PD 3-1 nate ac s urce f p wu f r Unit 2.
The existing emergency diesel generator (EDG) configura-New Diesel Generator 11ullding tion for Prairie Island Nuclear Generating Plant is that of a A new Class I building will house the ocw diesel genera-two-unit site with two shared EDOs Because of the uncer.
tors, the support equipment, and a new switchgear for tamties associated with the reliability of onsite ac power.
Unit 2.
Northern States Power Company initiated a project to add two additional safegua:Js emergency diesels, upgrade the Upgrade of the 121 Cooling Water Pump e
safeguards electrical distribution system, and upgrade a cooling water pump to become a swing safeguards pump.
As part of the SBO project, a new 4-kV switchgear power feed, designed to be fed from either generator D5 or D6.
Although the diesels are being added in response to the will be provided for the 121 cooling water pump. This 1
station blackout (SBO) sale, the project will offer a num-pump is common to both units. The pump ruotor will be ber of benefits, namely, increased operational flexibility, ungaded to become safety related. This upgrade increases the availability of the cooling water system and elitninates 3
elimination of the two-unit shutdown exposure caused by the need for a two-unit shutdown in the event cf extended
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existing limiting conditions for operation, and increased emergency onsite power margin as well as margin for fu*
inoperability of one of the safeguards diesel-driven coohng water pumps.
ture loads, liowever, the most notable ber.efit is increased safety, owing to a drarnatic reduction in both total core-e Main Control Board Modification melt frequency and in the conutoution of loss of onsite ac power to core melt nsk. As shown In Figure 1, the total Instrumentation and control equipment for generators D$
core damage frequency for the unmodified plant is estj, and D6 and for controls for the additional switchgear ' vill mated to be 4.3E-5/ year, with station blackout being the be placed on a new "G" panelin the main control room.
primary contributrr accounting for $3 percent of the risks.
The core damage frequency is reduced to 9.9E-6/ year af-
- Simulator Modification ter the modificaticas, with station blackout contributing The modifications include changing the simulator software only 6 percent of tFe risk-and the simulator control board to inciude the new diesci
Background
generators and electrical switchgear control functions.
e Plant Interface in 1975, in the Reactor Safety Study (WASil-1400), the NRC recognized that station blackout was a significant Electrical, instrument, and control equipment will be in-contributor to the potential total risk from nrciaar power stalled to accommodate the new diesel generators, and plant accidents. The issue of station blackout was resolved modifications will be required to realign the existing die-with a final rule on June 21, 1988. The intent of the rule 5el5-was to reduce the contribution of station blackcut to the overall risk of a core meltdown. The Station Blackout in conjunction with the SBO Program modifications. the
- Coping Study, completed in September 1987, showed that electrical safeguards 4-kV and 480-V distribution systems Prairie Island can safely respond to a station blackout, will also be improved as follows:
coping for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with the condition, based on the con-figuration that will exist after'the installation of the new 4-kV safeguards buses in Unit 2 will be replaced
- diesels, and Unit I safeguards buses will be expanded.
The primary objective of the Station Blackout / Electrical
- 480-V safeguard buses in both Units I and 2 will be Safeguards Upgrade (SBO/ESU) Program for Prairie 1s-replaced.
land is to implement those plant modifications and soft-ware elements necessary to comply with the final NRC rule
- 480 V safeguard buses and 4-kV switchgear capac-on station blackout and to remove the regulatory uncer-ity wili be added for future loads.
tainty associated with the sharing of two safeguards diesel generators by the two units.
- Voltage regulation and circuit coordination will be improved.
SHO/ESU Program W
w diesel generators are tandem drne units manu-Prairie Island has proposed using an alternate ac power factured by SACM in Mulhouse, France. Each diesel source as the means of meeting the requirements of the generator consists of one generator driven by two diesel SBO rule. The alternate ac source of power is being added engines, mounted one at each end of the generator in a by means of the following plant modifications:
back-to-back arrangement. The two new generators are each rated at 5400 kW and will supplement the two exist-0 Diesel Generater Addition ing Fairbanks Morse units, which are each rated at 2750 kW, Although the Prairic Island installation is the first for Two new diesel generators. DS and D6, are being added SACM in the Umted States. SACM has supplied the to serve as a dedicated source of_ emergency ac power for -
world nuclear industry with some 250 engines m 80 3
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nuclear plants. The architect / engineer for the project is A license amendment modifying the Prairie Island 'lechni-Fluor Daniel.
cal Speenfications will be tequired to accommodate the changes made to the plant as a result of the SBO/ESU Staff Resiew project.
The staff has (3e:n reviewing the Praine Island SBO/ESU project since the design report was submitted to NRC for approval in November 1990. Staff from the following The current estimated cost of the pro;ect is $142.6 nul-branches are involved in the design review: Electrical Sys-tion. A dual-unit outage, scheduled for Nosember 1992, tems, Plant Systems, Structural and Geotechnical, Instru-will be required for final tie-in of the new diesels. All SBO mentation and Control Systems, Mechanical Engineering activities will be completed at that time. Work on the elec.
and Vendor Inspection. Region 111 staf f is providing addi-trical safeguards upgrade will continue into 1994 until all tional inspection support.
improvements have been completed.
Unmodified Plant SD0 63%
Plant After SB0 modifications in Place lACA $6%
4 ATYS 1%
% ATWS 0%
SD0 0%
1DCA 13%
Transient 31%
Transient 33%
Total CDP = 4.3E-05/ year Tottil CDP = 9.9E-06/ year ATWS - Anticipated Transient Without SCRAM CDP - Core Damage Frequency LOCA - Loss of Coolant Accident S00 - Station Blackout Figure 1 Prairie Island Risk Profile 6
Tecilitical Specifications Improvement About 60% of the comments were of an editorial nature.
Editonal comments were resolved by a review panel work-g.
.g ing m concert with the owners groups as they were devel-oping " writers guides" to consolidate style and format Mary Lynn Heardon and preferenen. Of the remaining
- technical" issues, about a Christopher I. Grimes, O'ISil hundred were of great enough significance to warrant de tailed discussion and disposition in public meetings with On September 29, 1992, NRR completed a significant task the owners groups. Of those, about a dozen issues were related to the Technical Specihcations improvement Pro-referred for appealin public meetings between NRR samior gram (TSIP). On that date, the Technical Specifications managers and the owners groups executive panel.
Branch issued Revision 0 of the improved Standard Tech-nical Specifications.
Ultimately, the staff intends to record the disposition of all of the comments on the draft STS, as well as future Since early 1987, NRR has been implementing a three.
changes to the STS, on a computerited data base in re-part program to improve technical specifications for com.
trievable form. The Techr'ical Specif cations Branch es-rnercial power reactors, consistent v'ith the Commir.sion's tablished a "bulle'in board system" (BBS) to provide intenm policy statement which created the TSIP. The ready access to the completed STS sections durmg the three elements of the TSIP are (1) development of irn.
crmment resolution process. Currently, the completed sets proved Standard Technical Specifications (STS), (2) im.
of the improved.STS are available on the BBS. In view of plernentation of generic improvements for existing 11 the volume of the improved STS (about 6,000 pages for censes, and (3) exploration of risk-based alternatives to all five sets), the BBS has substantially reduced the admin-technical specifications. Over the last year, the staff has istrative burdens on the staff, and substantially decreased accomplished major initiatives in all three of these areas.
communication time between the staff and industry.
These activ ties continue to progress toward a final Com-mission policy on technical specifications, due in February h approved Revision 0 of the improved STS will be used 1993. Recent accomplishments will have an immediate im.
by volunteer lead plants to convert their existing technical pact and a long term benefit for most of the staff in NRR spec f cations. The lead plants that have volunteered to and the regions.
convert to the improved STS are:
Crystal River Unit 3 (Babcock & Wilcox)
Standard Technica: Specifications San Onofre Uruts 2 & 3 (Combustion Engineering)
Regulations require that each commercial power reactor Zion Units 1 & 2 (Westinghouse) have technical specifications which are included as Ap-
}{atch Units 1 & 2 (BWR-4, General Electric) pendix A to the license (10 CFR 50.36), The Commis' A consortium of all four BWR-6 plants (General sion's 1987 policy statement on technical specifications Electric) improvements specifically defined the scope of technical specifications; the criteria of that definition have been ap.
The staff is discussing the logistics and workload aspects of plied to develop improved STS. The improved STS consist the conversion review process with the owners groups and of five NRC reports, one for each of the nuclear power representatives from lead plants. After it completes its re-plant owners groups representing: Babcock and Wilcox view of the lead plant conversion, the staff will develop (NUREG-1430), Westinghouse (NUREG-1431), Com-Revision 1 to the improved STS for use by the next group bustion Engineering (NUREG-1432), General Electric of plaints that volunteer to convert to the improved STS.
BWR/4 (NUREG-1433), and General Electric BWR/6 (NUREG-1434). These repons were issued in draft form The Technical Specifications Branch will also establish for public comment in January 1991. At that time, the means to update the STS to incorporate new generic re-staff also solicited comments from the NRR technical staff, quirements and "line item" improvements.
even though the NRC staff had already contributed sub-stantially to the development of the draft reports.
Generic Activities for Existing Technical Specifications The Commission's policy statement was explicit about The industry and NRC staff submitted 27.302 comments generic "line-item" imp ovements that could be voluntar-on the draft STS ranging in significance from major over-ily adopted by licensees in their existing technical specifi-hauls of particular limiting conditions for operation to in-cations during the development of the improved STS.
no< uous punctuation correctinns. hiany of the substantive "Itese improvements are issued as topical reports or as ge-comments involved issues that had been debated exten-neric letters accompanied by model specifications and a -
sively between 1988 (when the owners groups submitted safety evaluation. Licensees may adopt these changes in a their topical repon versions of the STS) and 1991 (when license amendment on a voluntary basis. Line-item the NRC issued the draft STS). The revival of these issues improvements are generated in a variety.of ways, hiany retlected the complexity of the underlying issues and a proposed amendments become lead plant' line item lack of conclusive positions on the details of the funda-actions because the licensee wants to adopt a particular mental requirements. Consequently, the Technical Speci-change that is being incorporated into the improved STS.
fications Branch developed a rigorous process for resoking Conversely, other generic actions result from corrective comments. By means of this process the staff was able to actions that a particular licensee includes in a license promptly decide and carefully momtor the disposition of amendment application, which could apply to a whole each comment and, ivnen necessary, to mvolve NRR class of reactors; these other generic actions may also hase management.
been adopted by the owners groups as part of their c
7 L
comments on the improsed STS or may base been re-to component agmg and maintenance, and risk impact of flected in sendor topical reports.
executmg action statements (operational mode changes),
as well as plant configuration control studies. Of particular The origmal practice for hne item improsements prohib.
interest is a pilot program to explore "real time" plant risk ited " piece-meal" adoption of the improsed STS Ilow-analysn and plant configuration control. The feasibihty of eser, some licensees incorporated changes in proposed 11-such a program was ou'
>d in a special study by Sciente cense amendments with safety analyses to justify the Applications Internatio wporation (SAIC). The real.
changes on a plant specihc basis when the changes could tirne apphcation of risk t..
's would ideally use an on-have been treated genencally. In order to encourage stan-line computerivd nsk aralyw tool based upon a plant-dardization and consistency, and to make more efficient specific pRA. SAIC has used si,~...r methods to genen-use of staff resources, the Technical Specihcations Branch cally examine changes in (quipment allowed outage times worked with the projects staff to develop a new procedure and surveillance intervals for the improved STS, based on for screenmg hcense amendments which will identify li.
generic PRA models.
cense amendments with genent apphcability and will use the existing and ia. proved STS as modek for the review At several domestic plants, programs hne been vuuntar-and approval of plant specific actions. Thn procedure was ily implemented to monitor plant operational risk; in part, usued to all NRR division directors in a memorandum these programs utilire computenred data files. These files dated June 1,1992. When implemented, this procedure usually include data derived from the configuration of the can substantially reduce the overall staff effort needed to plant, operating events and fa. lure information, and the process the steady flow of incoming license amendments, use of PRA fault trees. Use of this data results in varying as well as to complete the older actions in the licensing degrees of plant nsk based configuration control. These action inventory (an aserage of 128 licenung actions were approaches are based on a wide variety of risk assessment completed each month of the first two quarters of methods and level of detail in the data, and are pnmanly FY-92).
used for mamtenance planning and outage scheduhng.
The Technical Specif cations Branch staff is routmely con-Southern Cahfornia Ednon (SCE) has voluntected to ap-tacted wheneser a question anses about the interpretation ply PRA-based nsk analysis softwear to assess a "real-of technical specihcations requirements, usually for the time" alternative to technical specifications. SCE will in-older plants with custom techmcal specifications. In addi-stall the software at San Onotre Umts 2 and 3 to develop tion, the staff is developing a plan to revise pan 9900 of experience wnh real-time confi uration management and E
the inspection hianual 50 that the inspection guidance, in-ruk insight comparisons to technical specification require-cluding that which has evolved from technical specifica-ments.
tions interpretations, is better orgamzed and more clearly presented.
~
NRC Staff Inspection Activities Risk-Based Alternatives to Technical Specifications Techmcal specifications are generally specific determin Thomas G. Scarbrough istic requirements based on quahtauve nsk insights. Since Mechanical Engineering Branch, DIX, NIUt the cally consideration of safety goals, however, the sys-tematic application of nsk insights has played an eser-in-In June 1990, the NRC staff issued NUREG-1352, " Ac-creasing role in regulatory activities. The concepts of risk tion plans for Ntator-Operated Valves and Check Valves,"
management and configuration control based on risk sig-describing specific activities intended to help resolve the nihcance are broadly appealing. One of the progressive concerns about the perfumance of motor operated valves elements of the Commission's policy en technical specih-(MOVs) and check valves. A significant activity of the ac-cations improvement is a soluntary puot program to ex-tion plan is the implementation of Generic Letter (GL) plore alternate approaches usmg the direct apphcation of 89-10 (June 28,1989), " Safety 4 elated Motor-Operated probabihsuc nsk analysis (pRA) rnodels and raethods, Valve Testing and Surveillance " In GL 89-10, the staff asked holders of nuclear power plant operating licenses and construction permits to confirm the capabihty
.f in the Umted Kingdom, success has bcen reported at MOVs m safety related systems by:
liepham B with a nsk based approach to techn cal speen heations known :s the Enential Systems Status Monitor reviewing MOV design bases e
(ESSM). The British anticipate instalhng the ESSM at all verif ring MOV switch settings initially and periodically of their advanced gas-cooled reactors. In Finland, com-e puter software has been developed for comparing the risk testing MOVs under design-basis conditions where o
of continued operations to the risk of shutting down fol-practicable lowing a partial or total loss of residual heat removal (RilR). The Fmns are anticipating implementing this sys-mprovmg evaluations of MOV failures and necessary e
tem at one of their plants, in seseral other countnes, pro-correctise action, and trendmg MOV problems.
grams to apply a risk-based approach to regulatory con-The staff asked hcensees to complete the GL 89-10 pro-trols are being pursued. The NRC Office of Research has gram within three refueling outages or 5 years from the conducted seseral studies that directly support, or are in-issuance of the GL, whicheser was later ducctly related to, nsk-based alternatnes. Among these j
are the work on test strategies at Battelle National Labora-The staff issued Supplement I to GL 59-10 on June 13, I
tones to define surveillance test intervals, risk impacts due 1990, to transmit detailed information on the results of 8
I pubhc workshops held to discuu the generic 9ter. On e
lack of justification for assumpoons of the stem fue Aurust 3,1990, the staff issued Supplemen.
to GL tmn coefficient over the lubrication interval 89-10 to allow bcensees time to review the informatmn transmitted in Supplement 1 and to incorporate that infor-failure to consider rate-of loading effects (load sensi-e matmn into their programs. On the basn of the resuhs of tive behasior) tbst can reduce thrust deinered by the NRCoponsored MOV tests, the staff issued Supplement 3 motor operator under high differential pressure arid to GL 89-10 on October 25, 1990, asking licensees et flow conditions from the thrust delisered under no-load conditions boihng water reactor (BWR) nuclear plants to act in ad-vance o' the GL 89-10 schedule to resche concerns about failure to consijer inertit of the actuator m estabbsh-e the capabihty of MOVs used for contamment isolation in ing maximum torque switch settings the steam supply line of the high pressure coolant mjec-uon and reacter core isolauon cochng systems and in the Many bcensees were updating their degraded voltage stud-supply hne of the reactor water cleanup system. In Supple.
ies to ensure that the worst-case minimum voltage asad-ment 4 to GL $9-10, the staff indicated that BWR heen.
able at each MOV motor has been determined. The staff sees need not address inadvertent MOV operation in their found weaknesses among licen3ces in (1) procedures for GL 89-10 programs ahhough the staff belieses that con.
conducting the differential pressure and flow tests (2) the sideration of valve mispositioning would benefit safety.
acceptance criteria for the tests in evaluating the capabiht)
The staff is currently reconsidering the need to address of the MOV to perform its safety function under design-madsertent MOV operation as pirt of GL 89-10 at pres.
basis conditions. and (3) feedback of the test results into surited water reacto,- (pWR) nut: car plants.
the methodology used by the licensee in pred cling ihe thrust requirements for other MOVs. Many licensees stated that they will attempt to test MOW with ciapostic As an mtegral part of their GL 89-10 programs, most h.
equipment under zero differential pressure an1 Pow con-censees are relying on MOV diagnostic equipment to pro-ditions (static conditions) to demonstrate the continued vide mformation on the thrust required to open or close capabihty of MOVs to perform their safety functions un-the valve, as well as the thrust dehvered by the motor ac-der design-basis conditions. None of those licensees, how-tuntor. In information Notice (IN) 92-23. "Results of ever, justified applying the results of tests conducted under Validation Testing of Motor-Operated Valve Diagnostic static conditions to demonstrate design basis capabihty.
Equipment," the staf f alerted hcensees to concerns identi-One weakness concerned the use of motor current in post-ficd by industry tests to validate the accuracy of MOV di.
maintenance testing that may not reveal problems with agnostic equipment as claimed by the equipment vendors.
MOV capability following certain maintenance activities One such concern is the increased uncertainty of the diag 4 (such as valve packing adjustments). Another weakness nostic equipment when it is calibrated in one valve stroke noted: some licensees responded poorly to MOV failures direction, but is used to predict thrust in the other direc-and deficiencies. For example, some licensees had not tion. In particular, ITI MOVATS recently sent licensees a analyzed the root causes of MOV problems thoroughly.
methodology to evaluate the effect of increased inaccuracy Most licensees '
s attempting to improve the trendmg of of its " Thrust Measuring Device" caused by cabbration of MOV problems. Some licensees had not made adequate the device in the valve opening direction while using the progress toward resolving the MOV issue.
equipment to predict thrust m the closing direction. The Nuclear Management and Resources Council (NUMAPC)
In response to Supplement 3 to GL 89-10, DWR licensees has prepared guidelines on the ITI MOVATS document.
have completed their evaluations of the capability of The staff is considering the issuance of a supple.,
at to MOVs within the scope of the supplement to perform their OL 59-10 to ask licensees to address any increased uncer.
safety functions under design basis conditions. BWR ticen-tainty of MOV diagnostic equipment caused by these valve sees have reported that approximately half of the 200
)
stem directional effects. The staff intends to enderse the MOVs within the scope of Supplement 3 hase been, or NUM ARC gtaidelines with comment.
Will be, adjusted or modified to improve their capabiht).
Because of the analytical nature of the evaluations, BWR bcensees will need to continue to assess the capabihty of The staff has inspected programs being developed by li-MOW within the scope of Supplement 3 during the imple-censees at most nuclear power plants in response to GL mentation of their OL 89-10 programs.
89-10. The staff has followed Temporary Instructic i (TI) 2515/109, " Inspection Requirements for Generr Letter in addition to the GL 89-10 inspections, the staff's diag-89-10, ' Safety-Related Motor-Opecated Valve Testing nostic evaluation teams (DETs) continue to fmd weak-and Surveillance'," in performing the inspections. Part I nesses in licensee activities invohing MOVs. The staff per-of Tl 2515/109 invohes a review of the program beinE formed the most recent DET inspection at FitzPatrick in estabhshed by the hcensee in response to GL 69-10. Part the Fall of 1991. Among the MOV deficiencies found by 2 of the Tl involves a review of program implementation.
the DETs are a lack of engineering bases for torque switch settings, inadequate procedures for setting torque switches, a lack of control of torque switch settings, IN 92-17. "NRC Inspections of Programs Being Devel-inadequate procedures for disassembling MOVs, and m-oped at Nuclear Power Plants in Response m Generic adequate lubncation programs.
Letter 89-10," provdes sigruf cant results of these NRC inspections. The inspections to date have focused on re-Many hcensees implementing their programs in response viewmg the GL 89-10 programs (Part 1 of Tl 2515/109).
to GL 89-10 are fmding MOV problems. For example, in Among the weaknesses found by the staff at various facih-November 1991, the licensee of Wolf Crtek began to dis-ties in the area of MOV sizing and switch settings were:
cover numerous problems with its MOVs u;,on initiating its 9
GL $9-10 program. The problems included inadequate Two Yelir Tritil Progrfirii For thrust capabihty of some htOVs mcorrect spring pack and gggggggggb, O ien Enforcement I
nameplate data, and inabihty to manually declutch some N10Vs. The Wolf Creek bcensee delayed returning the Conferences plant to power for several weeks to resolve the immediate N10V problems. In Niarch 1992, the hcensee of San Henc,e Pedersen, OE Onofre reported that two of four high pressure coolant in-jection/ low-pressure coolant mjectmo combined miniflow The NRC's current policy on enforcement conferences is N10% in Umt 3 failed to close dunng design basis differ-addressed in Secuon V of the latest revision to the
- Gen-ential pressure and flow testing per formed m response to eral Statement of Policy and Procedure for Enforcement 01 S9-10. The San Onofre licensee shut down Unit 2 as a Actions" (Enforcement Policy),10 CFR Part 2 Appendix safety precaution and determined that the Umt 2 miniflow C, which was published on February 18,1992 (57 FR N10% would not han been able to operate in their old 5791). The Enforcement Pohey states that " enforcement conf gurat on under design basis conditions. In October conferences will not normally be open to the public."
1991, the hcensee of Crystal River tested an emergency However, on July 10, 1992, the Commission implemented feedwater (EFW) h10V under differential pressur-and a 2-year trial program to allow certain enforcement con-flow conditions as part of its program in response to GL ferences to be open to public observation (57 FR 30762).
59-10. During the test, the valve did not close electrically During the trial program, approximately 25 percent of all under the calculated design-basis differential pressure con.
eligible enforcement conferences will be open to publ,c ditions. At that time, the Crystal Rner licensee believed observation, at least one conference will be open in each that the test differential pressure was greater than the ac.
regional office, and open conferences will be conducted tual desigr"baus..!ferential pressure. On April 28,1992, with a variety of types of licerwees. Conferences that will the Crystal River hcensee notified the NRC that the Octo.
be open to public observation will be selected at random, ber 1991 test failure had properly revealed that the EFW by usually selecting every fourth eligible conference involv.
N10V was not capable of performing its safety function to ing a commercial operaung reactor, hospital, or other type close under cbsign basis conditions. The Crystal River ti, of licensee. Conferences may also be open in cases involv-censee then found the other three EFW h10Vs were aho ing an ongoing adjudicatory proceeding with one or more 8"I'fV'"Of5' incapable of closmg during design baus differential pres-sure and now tests.
Enforcement conferences will not be open to public obser-vation if the enforcement action being contemplated (1) may be taken against an individual or if the action, though Esen where h10Vs hase not failed to stroke dunng tests, not taken against the indwidual, depends on whether or licensees base found that many hiOVs require more thrust not an individual has committed wrongdoing; (2) involves to operate durmg differential ps isure and flow tests than significant personnel fadures and the NRC has requested predicted by the standard industry equation with typical that the individual (s) involved be present at the confer-vaht factors (such as 0.3 for Dexible wedge-gate valves) ence; (3) is based on the findings of an NRC O! report; or assumed m the past. For example, the licensee for Farley (4) involves safeguards, Privacy Act, or proprietary infor-found that les than half of the 55 Dexible wedge-gate mation. Enforcement conferences invoking medical mis-vahe tested under differential pressure and now condi-administrations or merexposures will be open if the con-tions nave thrust requirements bounded by the standard ference can be conducted without disclosing the individu-industry equation with a 0.3 valve factor. The industry test al's name. Enforcement conferences will be closed if the results confirm the conclusions of NRC-sponsored 510V conference will be conducted by telephone, or at a small research that previous industry methods of sirmg htOVs licensee's facility, or if good cause is shown, and setting their torque switches were inadequate for some NIOVs.
The NRC will verbally notify a licensee as soon as it deter-mines that an enforcement conference will be open to pubhc observation. The NRC will announce each open en-forcement conference to the pubhc normally 10 working Although heensees are beginning to implement their pro-days before the scheduled date of the conference through grams in response to GL 89-10, N10V problems and fail-the following means: (1) notices posted in the Public ures continue to occur. For example, at FitzPatnck in Document Room; (2) toll-free telephone inessages; and Stay 1991, h10Vs in both trains of the low-pressure cool-(3) toll-free electronic bulletin board messages. Although ant injection subsptem of the emergency core coobng sys-the toll-free electronic bulletin board message system is tem failed, resulting in the need to shut down the facibiy not yet available, the public may call (800) 952-9674 to for repairs and root-cause analysis. Further, during a GL hear a recording of upcoming open enforcement confer-89-10 inspection at Three Niile Island Unit 1 m June ences.
1992, the staff discosered a 5 il % crack m the actuator housing of an 5 V in the stearr
.ie to the EFW pump in accordance with normal practice, open enforcement turbine. The (
'uing problems and failures of 510Vs conferences will normally be conducted in the NRC emphasize the a for beensees to implement their pro-regional offices. Conference observers will be reminded gams in response to GL 89-10 in a timely matmer. The that the conference is open for public observation, not staff will begin its inspections of the GL 89-10 program participanon; the apparent violations discussed are subject j
1mplementation under Part 2 of Tl 2515/109 in early to further review and may be subject to change; the deci-
- 1993, sion to conduct the conference does not mean that the 10
NRC has determined that a violation has occurred or that The NRC wall monitor the two year program and make a enforcement action w11' be taken; and the opinions ex-recommendation to the Commission on whether to perma-pressed by NRC employees are not intended to represent nently adopt and policy based on the following assessment final determinations or beliefs. Conference observers wall criteria:
hase an opportunity to anonymously comment on 'he Whether the fact that the conference was open im-NRC's program by submitting a form available in the re-e gional offices.
pacted the NRC's ability to conduct a meaningful con-ference and/or implement the NRC's enforcemcnt program; Whether the open conference impacted the licensee's e
On August 20, 1992, the NRC conducted its first open participation in the conference; enforcement conference invohing a commercial operating Whether the NRC expended a significant amount of reactor, New York Power Authonty's Indian Point Unit 3.
in the NRC's Regin 1 office located in King of Prussia, res urces in making the conference p61ic; and Pennsylvania. The purpose of the enforcement conference The extent of public interest in opening the enforce-e was to discuss service water issues at the facihty.
ment conference.
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