ML20081K729

NRR Technical Newsletter.Volume 2,Number 3

ML20081K729
Person / Time
Issue date:	02/28/1991
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-BR-0125, NUREG-BR-0125-V02-N3, NUREG-BR-125, NUREG-BR-125-V2-N3, NUDOCS 9107010073
Download: ML20081K729 (11)

Text

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TECHNICAL nunsafen-o,2s NEWSLETTER eeMarssi NRR Staffing Growth by Thomas E. Murley, Director Office of Nuclear Reactor Regulation The NRC's budget includes provisions for growth interviews, reference checks, and followup calls in NRR staffing levels for FY-1991, and the Presi-to bring new employees on board.

dent's budget request to Congress for F%l092 includes further funding f( c substantial growth for it is especially important that we conduct the the NRR's programs.

entire recruitment process with efficiency and professionalism.

Much of the staffing increase is for the high priority programs of license renewal reviews and standard 2.

Please give special attention to help train and plant certification reviews, We will need to recruit coach new staff on how NRR performs its re-a number of experienced people from outside the sponsibilities.

agency it we are to meet these growth require-ments and to compensate for normalstaff attrition.

We expect to have as many as 30 interns in addition to the new journeyman staff hired for

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Although the NRR has made a major effort in the technical, project, and administrative recruitment since last summer, we entered FY-branches.

i 1991 well below our authorized staffing levels. I have emphasized to the senior NRR mananers the I appreciate your support in this vital activity, importance i place on recruiting new staff to rap-idly reach our full strength, and we are giving even more personal attention to this task, with impor-tant support from the Office of Personnel.

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What's in this issue?

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suaEe/ss-o,2s NEWSLE I I ER eener$bi NRR Staffing Growth by Thomas E. Murley, Director Office of Nuclear Reactor Regulation The NRC's budget includes provisions for growth interviews, reference checks, and followup calls in NRR staffing levels for FT -1991, and the Presi-to bring new employees on board dent's budget request to Congress for FY-1992 includes further fundin the NRR's programs. g for substantial growth forIt is especially important that we conduct the entire recruitment process with efficiency and professionalism.

Much of the staffing increase is for the hi h priority programs of license renewal reviews an standard 2.

Please give special attention to help train and plant certification reviews. We will need to recru;t coach new staff on how NRR performs its re-a number of experienced people from outside the sponsibilities, agency if we are to meet these growth require-ments and to compensate for normal staff attrition.

We expect to have as many as 30 interns in addition to the new journeyman staff hired for Although the NRR has made a major effo;t in the technical, project, and administrative recruitment since last summer, we entered FY-branches.

1991 well below our authorized staffing levels. I have emphasized to the senior NRR managers the I appreciate your support in this vital activity.

importance I place on recruiting new staff to rap-idly reach our full strength, and we are giving even more personal attention to this task, with impor-tant support from the Office of Personnel.

To be successful, we need the support of everyone in NRR.

What's in this issue?

Therefore, I am asking that the staff give special See theindex atter tion to this recruiting activity in two main inside at the top of Page 2 1.

Please do your part in reviewing applica-tions, and in conducting recruitment trips, 9107010073 910228 PDR NUREG BR-0125 R PDR

the safety significance of MOVs in developing their generic letter W TIHS ISSUE P' 8'"*'"h* ^"did ""' "* "'"*"d i" *" " *"*'8 " '

Supplements 1 or 2) that licensecs and permit holders establish any particular priority for MOVs within the program, llowever, the information recently obtained from the NRC-sponsored tests, NRR Staffing Growth however, has influenced the staff's position regarding the prioritics b) Th tras E. Murley -

1 for implementing the generic letter programs. Those tests were conducted as part of the stall s effort to resolve Generic Issue (GI)

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otor-Operated Vdec Testing and 87, " Failure of IIPCI [Iligh Pressure Coolant injection] Line Without isolation," which involves the evaluation of the capability by a a.

mh-2 of MOVs to mitigate the loss of reactor coolant inventory in the event of a pipe break outside of the reactor containment at imiling Operathnal Lxperience -- NRC Generic Communica-water reactor (BWR) plants. The particular MOVs within the in 's scope of C-187 are those used for containment isolation in the by Vernoi. C'. Ilodge and Carl 51. Berlinger 3

steam lines to the llPCI and Reactor Core Isolation Cooling (RCIC) systems and in the supply line to the Reactor Water bspec: ion Teams Combiac NRR and Region I Cleanup (RWCU) system.

Talents at Indian Point Unit 3 by Peter W. Esclgroth, Region I.

6 The MOV testing program for Gl 87 was conducted for the NRC Office of Nuclear Regulatory Research (RES) in two phases by the Palisades Steam Generator Replacement Project Idaho National Engineering Laboratory (INEL). Phase I was by John M. Jacobson, Region 111 ;

.7 performed in 1988 at the Wyle Laboratory facility in lluntsville, Alabama. The most significant tests in that phase consisted of Region \\, E,ng.meermg Managers' I,orum opening and closing two 6-inch flexible-wedge gate valves (manu-by F. Randall Iluey, Region V.

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NEWSLETTER CONTACT-those tests were discussed at a public meeting on February 1,1989, Valeria Wilson, NRR 492-1208 and are documented in NUREG/CR-5406,"BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualifica-tion and High Energy Flow Interruption Test."

Safety-Related Phase H of the MOV test program was performed in 198C at the Kraftwerk Union facility in the Federal Republic of Germany.

Motor-Operated Valve This phase consisted of opening and closing three 6-inch flexible-Testina and Surveillance

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"i"8 Vc' " "ad *^'* h) ""d three 10-inch flexible-wedge gate valves (Anchor / Darling, Powell, a

and Velan) against normal and blowdown flow conditions. On by Tom Scarbrough, EMEB December 26,1989, the NRC staff issued Information Notice 89-88,"Rccent NRC-SponsoredTesting of Motor-Operated Valves,"

On June 28,1989, the NRC staffissued Generic Letter 89-10, which alerted addressees to the tests and provided preliminary 1

" Safety-Related Motor-Operated Valve Testing and Surveil-results. On April 18,1990, the NRC staff held a public meeting to lance," w hich requests licensees and construction permit hold-discuss the results of Phase 11 of the MOV testing program. On ers to establish a program to provide for the testing, inspec-J une 5,1990, the staff issued Inform ation Notice 90-40,"Results of tion, and maintenance of safety-related motor-operated valves NRC-Sponsored Testing of Motor-Operated Valves," to provide (MOVs) and certain other MOVs in safety-related systems.

specific information regarding the MOV test program. One of the The staff requested that the test program be completed within significant conclusions was that the tested valves required more 5 years or thrse refueling outages and that MOV capability be thrust for opening and closing under high differential pressure and periodicalb, verified every 5 years thereafter. The staff held flow conditions than would have been predicted from standard public wor kshops to discuss the ge neric letter and prmided the industry calculations.

results of those workshops in Supplement 1 to Generic Letter 89-10 (June 13,1990). To allow time for licensees and permit Following the April 18 public meeting, the NRC staff conducted an holders to incorporate the information in Supplement 1 into informal survey of six BWR reactor units to determine the capabil-their programs, the staff stated in Supplement 2 (August 3, ity of the MOVs used at those plants for containment isolation in 19%) that inspections of generic letter programs would not the steam Ime of the llPCI and RCIC systems, and in the supply commence until January 1,1991.

line to the RWCU system. The staff compared the capability of the currently-installed MOVs to the results of the NRC-sponsored While suggesting that licensecs and permit holders consider tests. On May 24, the staff met with the BWR Owners' Group to 2

discuss the results of that survey, in resp <mse to staff concerns, As provided by 10 CFR 50.54(f), the staff required in Supplement the UWR Owners' Group on July 6 provided information re-3 that i WR licensees (1) notify the NRC staff within 30 days of the garding the capability of the A10Vs used for containment isola-date of Supplement 3 that a plant-specific safety assessment report tion in the steam lines to Ihe llPCI and RCIC systems and in the was available on site for staff review; and (2) provide the NRC staff supply line to the RWCU system for the remaining BWR reactor within 120 days with (a) the criteria, reflecting operating experi-units. Among the h10V data supplied by the BWR Owners' ence and the latest test data, that were applied in determining Group were the valve manufacturer, type, and si7e; motor sire; w hether deficiencies exist in t he 11PCI, RCIC, and RWCU h10Vs, actuator rating; fluid design-basis differential pressure and in the h10Vs in isolation condenser lines, and in any h10Vs temperature conditions; and thrust capability at the current considered to be more safety significant, as applicable, (b) a list of torque switch setting.

any h10Vs found to have deficiencies, and (c) a schedule for any necessarycorrective action. Any changes to the planned actions or From the statfs analysis of the BWR Ovmcrs' Group data, the schedule are to be submitted for the staff to revie'v and approve.

staff luame concerned that many of the h10Vs installed to perform a containment isolation function in the llPCI, RCIC, While the reporting requirements of Supplement 3 are addressed and RWCU systems at BWR plants might not be capable of to BWR licensees, the staffintends that alllicensees consider the performing that function. As a result, the staff prepared a safety applicability of the information obtained from the NRC-spon.

assessment to evaluate the need for prompt attention to the sored tests to other h10Vs within the scope of Gencric Letter 89-potential deficiencies in the installed h10Vs. Further, the staff 10, in addition, alllicensees should consider this information in the activated the BWR Regulatory Response Group and held a development of priorities for implementing the generic letter pro-public meeting with that group on August 1 to discuss the staffs gram.

analysis and safety assessment. At that meeting, the BWR Owners' Group presented its own safety as:icssment and de-The staff prepared Generic Letter 8910 in accordance with NRC scribed its plans in tesponse to the test results. Based on its safety procedures for the issuance of staff guidance containing backfit assessment, the BWR Owners' Group did not intend to advance provisions, including review by the Committee to Review Generic thc 5 year schedulc of Generic Letter 89-10 for the llPCI, RCIC, Requirements (CRGR), Because deficiencies might exist in and RWCU MOVs.

h10Vs installed to perform containment isolation functions at BWR plants,the staff determined that the issuance of Supplement in contrast to the BWR Owners' Group position, the staff 3toGenericLetter89-10wasnecessarytoprovideconfidencethat concluded that correction of any deficiencies in the llPCI, BWR facilities are in compliance with their safety analyses and RCIC, and RWCU MOVs needed to be given high priority in the NRC regulations such as described in Criteria 54 and 55 of 10 CFR implementation of Gencric Letter 8910. Therefore, the staff Part 50, Appendix A. Befereissuance of Supplement 3 toGeneric issued Supplement 3 to Generic Letter 89-10 on October 25, Letter 89-10, the staff discussed the proposed supplement with the 1990, which requested that BWR licensees assess the applicabil-CRGR and resolved its comments, ity of the data from the NRC-sponsored MOV tests; determine the capability of their ilPCl, RCIC, and RWCU M OVs; identify any deficiencies in those MOVs; and establish a schedule for the correction of anyidentified deficiencies. Because lines to isola-tion condensers also communicate directly with the reactor vessel, the staff requested in Supplement 3 that BW R licensees OPERATIONAL EXPERIENCE --

evaluate rhe MOVs used for containment isolation in those lines, NRC GENERIC where applicabic.

COMMUNICATIONS Based on the generic safety assessments prepared by the NRC staff and the BWR Owners' Group, the staff believes that justification exists for individual plants to which those safety by C. Vemon Hodge and assessments are applicable to take 18 months or to the end of the Carl H. Berlinger, OGCB first re fueling outage, following issuance of Supplement 3, w hich-everislater,to resolve anydeficienciesinthe HPCI,RCIC,and As part of its program to feed back information on operating i

RWCU MOVs. To verify the applicability of these generic safety experience to the industry, the NRC issues several types of generic assessments, the staff requested BWR licensecs to perform a communications. Since 1971, the NRC or its predecessor agency, plant-specific safety assessment. If a BWR licensee determines the Atomic Energy Commission, has issued about 1800 of these that a more limited time is mandated by its plant-specific safety generic communications in the form of bulletins, generic letters, assessment, the staffindicated that the licensee should use the circulars (which are no longer issued), and information notices more restrictive time. Where additional time is needed to (see Figures 1-3). Either of two offices may issue generic commu.

complete the corrective actions, the staff stated that the licensee nications: the Office of Nuclear Reactor Regulation (NRR) issues should submit the proposed schedule for review.

those communications pertaining to power plants or research reactors, and the Office of Nuclear Material Safety and Safeguards 3

(NMSS) issues those communi.

cations pertaining to fuel cycle BULLETINS ISSUED activitics or the uses of radioac-mnu tiw materialsinindustryor medi-NOVEMBER 1990 cine.

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l Addresseesmay substitute equally effective alternative ac-IN FORM ATION NOTICES ISSUED tions for the actions requested.

mnu The NRC staff wuukt issue these NOVEMBER 199O twes of communication with the concurrence of the CRGR. Re-m m

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presiously issued generic letters focused on Three hiite Island (TMI) Action Plan items, and GENERIC LETTERS ISSUED frequently these generic letters THnu requested changes in plant Tech-NOVEMBER 199O nical Specifications. For a few years after the accident at TMI,

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* " 2 occo t u u licensee actions in response to bulletins and genericletters. In March 1988 the NRC changed its procedures for these actions. Generally, preference is now cussed. A dBASE III program, "GC!.PRG," helps users find given to having licensees take the requested actions to address the documents in this large database. Copics of the GCI are loaded safety concern and then inform the NRC when the actions are on the common use microcomputers at workstations 8 D21 and complete. NRR occasionally finds that more than a confirmatory 11 A6 in the White Flint building. Copics are available to staff response from addressees is needed; for exampic, ifinspection o members and members et the public on request. The updated r

licensee actions is considered necessary, a temporary instruction GCI will be made available in NUREG/CR-4uX) and will be for regional inspection is developed along with the generic com-available for sale in computer use format through the National munication.

Energy Software Center at the Argonne National Laboratory.

The staff has closed out almost all the bulletins issued since the To provide rapid availability of the textual content of generic TM1 accident, and has published a NUREG/CR report for each communications, the staff has set up a bulletin board facility on bulktin closed. The staffis expeditiously proceeding with closure the t'ata General MV8000 called EMAIL, through which users of generic letters with appropriate annotations in the Safety Issues can now electronically view and download these documents.

Management System (SIMS). Most recent bulletins and generic Figures and signatures are not included because of memory letters are tracked and closed by project managers in the SIMS limits. Except for a few of the oldest documents, EMAIL has database. This effort is using significant technical assistance from texts for all the generic communications. EMAIL is available to contractors such as Parameter, Inc.

all who want it, including NRC staff. The staff plans to load both the GCI and EMAIL on the NRR hxal area network.

The NRR Generic Communications Branch has also developed a

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system to casily track information that has been published on a If a staff member needs a copy of a generic communication, the given subject. This system is the Generic Communications Index preferred source is NUDOCS/AD; a secondary source is the (GCI), which includes data on bulletins issued since 1971 (about NRR Generic Communications Branch (OGCB) files. Ilard 100), circulars issued from 1976 to 1981 (about 50), generic letters copies of almost all these documents can be foun, 'n labeled file issued since 1977 (about 500), and inform ation notices issued since cabinets netr the two computer locations previously mentioned.

1979 (about 1100). Each record includes data licids for general system or topic, specific component or topic, cause or defect, We thank Dick Kiessel,0GCB, for Figures 1-3. For more infor-potential effect, and vendors associated with the problem dis-mation, contact Vern ilodge (GCI) NRR/OGCB,492-1861.

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auembiininad7tieniiy iiried out ofibe core werc each suspenaea lnspection Teams Combine by a s.mgle locatmg pm. The locating pins at the top of each fuel NRR and Region I Talents assembly extend downward from the upper core piaic ana no,.

mally insert into two upper nonle holes in the fuelassembly when At Indian Point Un,it 3 the upper core iniernai structure is properiy angnea over ihe inp of the core. At each of the inadvertently suspended fuel assemblics, by Peter W. Eselgroth, Region I one locating pin was found severely bent and not engaged into the fuel assembly's top nonic hole; the other locating pin was found bent and suspending the assembly at an angle of approximately 7 On October 4,19W, Indian Point Unit 3 was in a refueling outage degrees.

wben two improperly grappled irradiated fuel assemblies were lifted out of the reactor core simultaneously. This safety. The NRC team resiewed the licensce's plans for retrieving the fuel significant event, which has wide potential applicability to other assemMes, and performed independent analysis of the radiologi-plants, called for an NRC response capable of rapid, onsite cal consequences and independent resiew of the fuel retrieval review of the situation,a technically thorough assessment of the pr cedures. A significant aspect of the NRC's review of the ft.cl licensee's pbn of action, competent communication about the anemMy recovery plan was that although the licensee had made an event with the press, and extensive followup actions.

analysis that concluded the fuel assemblics would probably not drop off during transit to the specially fabricated catch baskets, the The NRC response to this event comprised an initial contingency NRC did not concur with moving the assemblies until the licensee of technical personnel to respond to the fuel assembly recovery had analyzed and made contingency plans for the potential radio-efforts and a followup team assigned to delve into what had logic 1 e nsequences of a fuel assembly dropping.

caused the event and how to correct the problem. This NRC response drew upon personnel from Region I and NRR and The fuel retrieval process was implemented by the licensee as demonstrated the age ncy's ability to respond to such events with planned with the exception that during the movement of the UIP, a diverse and highly competent cadre of technical talent.

one f the two fuelassemblies dislodged and fellinto the prepared container when the brakes to the overhead crane were applied.

On October 4th, the licensee was attempting to remove the The licensee lowered the remaining assembly, and freed it without upper core support structure (upper internals package, UIp) further incident. The dropped assembly resulted in no radiological from the reactor vessel in preparation for refueling the core. rdcases or breach of fuelintegrity.

After raising the UIP out ofIhe reactor vessel, lateral movement of the UIP was commenced and then stopped when the licensee %c Regional Administrator also formed an augmented inspection discovered that two peripheral fuel assemblies were suspended te m (AIT) which was sent to the site following recovery of the from the bottom of the upper core plate whichis part of the UlP. inadvertently lifted fuel assemblics to document the relevant facts, llecause of poor lighting and improper camera location, the two to determine the probable cause(s), to evaluate the licensee's suspended fuel assemblies had not been recognized earlier w hen analysis and review of the event (including corrective actions), and the underwater video inspection was performed. The licensee to evaluate the potential generic implications of this event. The immediately suspended the manipulation of the U1P and noti-AIT comprised a team leader and four iIcadquarters and Regional ficd the NRC of this event.

specialists who had expertise in refueling operations and proce-dures, the design of reactor sessel internals and fuel assemblics, Initially, the NRC dispatched a special inspection team from core physics, training, and management controls.

Region I and NRR comprising engineering, operations, and health physics personnel; a supervisor; and management. The Since the two fuel assemblies were apparently lifted out of the team alsoincluded the senior resident inspector. A confirmatory reactor core on October 4,1990, because the fuel assembly upper action letter (CAL) was issued to confirm the licensce's commit-How nonles were hung up on upper core plate guide pins that were ments to develop a safe and controlled retrieval of the two fuel bent, one of the principal items of interest to the AIT was to find assemblics and to obtain NRC agreement before: (1) moving ut how the guide pins had become bent in the first place. This the upper core internals with the two assemblics attached and (2) miewinv Ived the consideration of a number of possibilities,such degrading the containment integrity and degrading vital safety s improper guide pin originallocation (installation)in the upper systems. The CAL also confirmed the licensee's commitment to e re pl te, loss of guide hole run-in due to excessive chamfering on restrict containment access to only those personnel required to the upper nonle outer edge, guide pin excess out-of-perpendicu.

monitor and recover the fuel.

larity (most likely due to bumping / bending during previous refu-cling operations), fuel assembly upper end misalignment associ-The two assembliesinadvertentlylifted out of the reactor core ated with assembly-to-assembly tolerance stackup, overall core /

were attached to the UIP by bent fuel assembly locating pins anemblics no longer "true" and causing misalignments due to (guide pins). By design, each fuel assembly is normally held in accumulation of debris or core baffle problems, and overall upper position in the reactor vessel by two upper and two lower guide me plate bowing due to the weight of the upper core internals.

pins attached to the upper and lower core plates. The two fuel 6

In parallel with the inspection and review efforts of the AIT, the concluded to be within existing analyses of the FSAR because of licensee formed a root-cause analysis team and took steps to the power bistory of the fuel assemblics. In the second instance,the ascertain the present condition of all upper core plate guide pins, relatively lower power density associated with the peripheralloca-fuel assemblics, visible reactor vessel internak, and associated tion of the two assemblics with debrmed fuel rads resuhed in the refueling equipment. In addition, the licensee thoroughly re-assemblics meeting the I5AR accident analyses during cycle 7 viewed all records associated with the 1989 refueling operation. operation. Ilowever, the potential for a heightened safety signifi-The hardware inspections revealed that three fuel assembly cance of this type of event beinginduced by such bent pins under upper flow noules had been bent down out of position, two fuel other circumstances is readily apparent. The safety significance of assemblics had bent fuel rods related to the upper now nonle this type of event has resulted in certain generic recommendations movement, and two guide pins were bent less than 7 degrees and by the AIT for precluding such occurrences.

could be straightened while three other guide pins were bent more sescrely. Of these latter three guide pins, one had de-Both of the team efforts described in this article are good examplcs tached during cycle 7 operation and had become hxiged in one of how the talents of NRR and a regional office were brought of the steam generators, onc tell off during the current 1990 together to respond to a problem and effectively address the refueling outage, and the third was cut off during this latest technical issues involved, outage.

The licensee's and the AIT's review of records from the 1989 refueling outage, including the videotapes of tefueling activities, revealed that the UIP had not been adequately raised from its PALISADES reaam cavity storage stand befme it was moved iainauy f" STEAM GENERATOR relandmg the UIP in the reactor vessel. The videotape faowing that the UIP had been bumped against the storage stand in the REPLACEMENT PROJECT lateral d, rect, ion, and the inspection of the angles of bend on the i

upper core plate guide pins led the licensee and the AIT to conclude that this was how the guide pins had been bent.

by John M. Jacobson, Region lll llaving determined how the guide pins had been bent, the Steam generator (SG) tube degradation appeared early in the licensee began preparations for reloading the reactor core and operatinglife of the Palisades Nuclear Plant. The first incidence of reinstalling the UIP. The AIT then became intimately involved tube corrosion failure occurred after less than 1 year of operation in reviewing the various aspects of licensee readiness to take in 1972. Aften approximately 4 years of operation,3,551 tubes out these steps. This review included an assessment of such factors of a total of 8,519 (over 20 percent) were plugged because d as: the acceptability of reactor operation with some fuel assem. corrosion, blics having missing guide pins, the adequacy of training and procedures, and review of the core reload analysis.

A change in water chemistry in mid-1974 slowed the corrosion rate such that only 378 additional tubes were plugged during the period From its review of the 1989 refueling outage activities as well as 1976 to 1987,1lowever, during the 16-month period beginning in from its observation of activities on site during the 1990 outage, December 1987, seven tube leaks resulted in six forced outages.

the AIT identified the following as the principal contributing factors to this fuel assembly event: inadequate monitoring by the A decision was made to replace the SGs during the plant refueling licensec of safety-relat ed work performed by a contractor; short-outage beginning Septembc. 15, 1990, continuing through mid-comings in ihe definition of responsibilities between the licensee February 1991. The replacement SGs were fabricated by the and its refueling contractor; and deficiencies in the level of detail original equipment nanufacturer, Combustion Engineering.

embodied in refueling training and procedures for the purpose of ensuring that the UIP guide pins are protected from damage To date, six U.S. plants (all of Westinghouse design) have com.

during transit and storage of the UIP outside of the reactor pleted SG replacement projects. At these piants, the containment vessel.

equipment hatch openings were big enough to accommodate the SGs.

In reviewing the safety significance of this event, the NRC inspectwn teams considered two areas of safety concern: (1) the The replacement project at Palisades is unique in that the signifi-potential of dropping two irradiated fuel assemblies when the cantly larger size of the Palisades SGs prevented their removal applicable analysis in the Final Safety Analysis Report (FSAR) from the containment through the equipment hatch. A specialcon-had been done for one fuel assembly and (2) the possibility that struction opening 26 by 28 feet had to be cut in the 3.5-foot thick the mechanical deformation of portions of the fuel rods in two containment wall. This approach was recently employed at the assemblics might have resulted in plant operation during cycle 7 Ringhals plant in Sweden, outside the design accident analyses and thermal-hydraulic con-siderations of the FSAR. In the first instance, the event was Another unique feature of the Palisadcs effort is the first U.S.

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reactor cavity and ;he polar crane rail was erected in q

j@j crane, a semi-gantry lift supported by a stiff leg over the

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g ty'4 As of December 1990, the 170-ton concrete bhxk cut f

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been removed and a temporary roll up door installed.

The SG piping and supports have been cut and the vessels

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removed and transported to the temporary onsite storage T

facility. The replacement SGs have been lifted into place W

d, and 100 yards of concrete has been poured to restore the J

iMM containment structure. Ilorizontal and vertical post.

M; tension tendons were then reinstalled.

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As previously mentioned, one of the unique features of g

M this project is the use of the " narrow gap" welding tech-nique. The " narrow-gap" process requires very precise 4

weld groove preparation and highly specialiied automatic welding equipment. The application of this process to heavy-wall nuclear system piping was the first of its kind in the United States. The process has been applied in e

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Europe with a high degree of success and most recently at h

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the Ringhals facility in Sweden.

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The " narrow-gap" procedure is based on the automatic hh,,q ri p ized equipment enablingthe welding torch to reach dceply gas tungsten arc welding (GTAW) process with special-4 t

s into a narrow weld groove preparation. The width of the groeve was approximately 0375 inch at the outer diame-ter of the 2.5-inch-thick (cold leg) and 4-inch-thick (hot.

gy leg) piping. To ensure a high quality weld, emme care gg@

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is required during the weld groove preparation and the joint fit up phase, gg ygg 1+

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This " narrow-gap" process includes the following advan-tages:

Region lli Reactor inspector John Jacobson and NRR Project o

Because of less heat input and smaller bead size, Manager Brian Hollan Inspect the stearn generator rigging plat-the residual stress distribution is favorable.

form, Significant reductions in shrinkage occur as a o

application of the " narrow gap" welding technique developed in Ger-result of a decreased weld volume, many. This technique was used to reattach the primary coolant piping to the SG nozzles, o

Low heat m.put and small bead s.ize produce a more favorable heat-affected zone.

The licensee contracted with Bechtel Construction, with welding exper-The weld preparation geometry provides an es-tise from Kraftmerk Union Atiengesellschaft (KWU), to perform the o

replacement, sentially constant, very narrow groove that does not require torch manipulation other than straight.

To accommodate movement of the 480-ton SGs, speciallift and trans-line travel. This benefit resultsin a reduced poten.

port equipment was required. A rigging platform 27 feet above ground ti i f r fusi n defects in the weld joint.

level was erected next to the containment opening to allow the SGs to bc winched out horizontally. A gantrylift was erected over the platforrn The " narrow-gap" process is in concert with the as low as to lift the vessels and then lower them onto a special heavy load is reasonably achievable (ALARA) concept in that the transport. The transport featured 44 axles, each capable of hydraulic welding equipment is operated remotely and the narrow level control, and a total of 176 tires.

weld groove requires significantly less weld volume, re-I a

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sulting in reduced welding time The welding

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exposure as a factor in the schedule incen-Q.

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tives offered to !!cchtcl. Decontamina-X Mg e

tion and shielding efforts,in conjunction

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with extensive training on a fullaire steam

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%,H(,1 ~p' pencrator channel head, haw proved highly

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successful in minimizing the accumulated s

exposure for the project. The current j exposure totalis more than 125 man tem a

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unforeseen problems, the final project to-l,

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and onsite inspections throughout the SG l replacement effort, The first replacement steam generator sits on the rigging platform, following its j

ren' oval from the transport vehicle.

1 REGION V "c*' h""' c nw fr nuhese nwdings l

ENGINEERING MANAGERS' (1)

DESIGN-!! ASIS DOCUh1ENT PROGRAhi GUIDE-LINES (issued hiay 1990)

FORUM This document was developed by a forum subcommittee whose members came from the NUhtARC design-basis by F. Randall Huey, Region V document (dud)workinggroup. The Region vdocument served as the foundation for the current NUhtARC guide-In the last few years, the NRC has increased its attention on assessing the quality of licensecs' engineering and technical These guidelines define the intended scope of DBD pro-work. This increase largely resulted from Safety System Func-grams and the evaluations needed to develop design-basis tionalInspection (SSFI)lindings. The SSF1-type inspection was information. In particular, the document establishes crite-particularly effective at highlighting implementation weaknesses ria for DBD scheduling, quality assurance, verification, and inlicensecs' engineering programs. Amongtheir most notewor-use. The document also addresses the need for tracking

, ny findings, these inspections consistently identified problems open items, for relating licensing bases to design bases, for

! invokinglicensecs' loss ofcontrolof the originaldesign ba:es for using pre-Appendix B design information, and for training

[ their plants and inadequate involvement of the engineering plant personnel on the use of DBDs. The document gives

organization in plant maintenance and modification processes, examples of typical design basis and design output docu-

subsequent to initial plant construction.

ments and describe 3 mechanisms for configuration man-agement,in order to ensure continued updating of design-l As a result of findings from several of their own SSF1-type basis information.

! inspections and in response to a heightened awareness of the

. need for more proactive and intrusive engineering involvement (2)

PROACTIVE ENGINEERING GUIDELINES (issued

' in support of nuclear plant operations, Region V licensees,in Alarch 1990)

! June 1988, initiated the Region V Engineering hianagers' Fo-i rum. Since itsinception, the forum has met six times and has held This document provides a format for promoting active, j numerous subcommittee meetings. To date, two formal docu-systematic, and timely involvement of engineering prin-l 9

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i ciples in plant operations. The document was given to o Ensur e effective and timely involvement of qualiGed per-

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INPO as well as to engineering management counter parts sonnel from both the engineering and operating staffs in in other regions. INPO commented favorably on it.

the capital and O&M budgeting process.

Develop an effective periodic design self assessment The primary strength of these guidelinesis the recognition o that:

process. It is important that engineering management get frequent feedback on the effectiveness of engineering Fundamentally, nearly every decision that affects plant improvement initiatives, operation has a design or engineering element. Therefore,

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it is beneficial for the engineering or3anizations to be Estabilshing Teamwork and involvement activelyinvolved in plant operation. Plant operation within o

Structure the engineering organization to promote the the design bases is a shared resp <msibility between the engineering and operation organi7ations. Three factors are team approach. Ensure that site and corporate person-I available to promote active and systematic involvement:

nel are working as a team.

j (1) Processes and Procedures: selection of the processes at the plant where engineering [ principles] can be effectively o Provi le job descriptions for engineers that clearly reflect and efficiently utilized is the first step and implementation the proactive engineering approach. Ensure an appro-i in accordance with the utility procedures assures consistent priate Erst-level super ision role in promoting team-l application (2) Availability: availabilityof the engineering work.

resources facilitates prompt participation, and (3) Culture:

Establish rotational and cross-training assignments.

teamwork is essential and is perpetuated by the culture o

endorsed by management. Cor} crate policies and actions Provide for periodic joint system engineer and design can endorse cooperative relationships, thus uniting the o

team.

engineer plant system walkdowns.

Some of the most notable recommendations of the document Establish an Engineering Oryanization That Is Resp <mshe to include:

Plant Needs Estabibhing Clear Responsibilities and Accountability In order to establish a credible working relationship with their operating plant counterparts, the engineering organization must Provide a formalized charter to clearly establish responsi. be responsive to plant needs, o

bilityand accountability and define relationships and shared boundaries betwee n engineering /t ech nical support organi-Provide cognizant engineer on calllists.

o rations and their plant operating and maintenance counter-

parts, o

Ensure that appropriate engineering personn J are physi-cally located at the plant.

Provide focused engineering resources. Establish effective o

Provide for periodic, joint engineering and operating long range planning and ensure that engineering resources o are efGciently applied to the right efforts at the right time.

organization review of engineering backlogs.

Estabibhing Effective Plant Support Programs and Processes Establishing EITecthe Communication Provide for such appropriate engineering participation Provide a workable root-cause analysis program that is o

o recognized as being effective in precluding recurrence of on plant review committees as: daily plant status meet-plant problems.

ings, restart readiness reviews, signiGeant event reviews, and plant management review meetiogs.

Develop an integrated design control process that can i

o Provide a clearly defined, easy-to-use and rigorously accommodate from the simplest to the most complex o changes to the plant in a disciplined and controlled manner, tracked non-conformance report (NCR) evaluation pro-Ensure appropriate design safety margin in all design

gram, actisitics.

o Establish a Grst-level supenisory interdepartmental o

Ensure readilyaccessible design basis information to front-rnecting process.

line engineering and technical support personnel. Ensure that the design basis document is used continually to sup-o Promote the use of electronic communication systems.

j port engineering activities.

The forum is currently working on other initiatives such as spare w

4 parts, backlog of engineering work, and qualification standards for engineers. The forum plans to work closely with and involve INPo in all ofits initiatives. To the degree that initiatives involve generic issues, requiring a close working relationship with the NRC, the forum plans to involve NUMARC (as was the case with the design basis document guidelines).

Region V forum members are enthusiastic and confident that their efforts will go far toward improving nuclear plant opera-tions. They are actively working with their counterparts in other regions and have indicated that both Region 11 and Region ill 4 utilitics are setting up similar forums.

l The ultimate measure of success for these initiatives willinvolve

how well the individual licensees translate the guidelines into

! clearly defined responsibilities and management expectations.

The Region V forum efforts to date appear to be a gomi first step.

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ti l). 5. 4+ernment Printany Of f s.e IH) 4 b 2 - 14 6 / 4 0014 a

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l 120'255139551 1 I t+ 1 US NRC- %3M DIV FOIA ( RUillICATIMas SVC3 ips P.] F - N J a E ',

D-223 WA3HING10N DC 20LLt i

THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER.

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